High-Level Waste Environmental TECHNICAL REPORT
Standards Program EPA-520/3-80-007
Technical Support Document
RADIATION EXPOSURES FROM
SOLIDIFICATION PROCESSES FOR HIGH-LEVEL
RADIOACTIVE LIQUID WASTES
William F. Holcomb
William N. Crofford
Raymond L. Clark
and
Frederick C. Sturz
OFFICE OF RADIATION PROGRAMS
U.S. ENVIRONMENTAL PROTECTION AGENCY
WASHINGTON, D.C. 20460
MAY 1980
-------
EPA REVIEW NOTICE
The Office of Radiation Programs, U.S. Environmental Protection
Agency, has reviewed this report and approved it for publication.
Mention of trade names or commercial products does not constitute an
endorsement.
ii
-------
PREFACE
The Office of Radiation Programs, U.S. Environmental Protection
Agency, carries out a national program to evaluate individual and
population exposure to ionizing and non-ionizing radiation and to
promote development of controls for the protection of public health and
the environment.
This report is technical support for EPA's high-level radioactive
waste environmental standards; it estimates the potential environmental
effect of solidification of high-level radioactive liquid wastes.
The Office of Radiation Programs invites readers to report omissions
or errors, submit comments, or request further information.
OFFICE OF RADIATION PROGRAMS
U.S. ENVIRONMENTAL PROTECTION AGENCY
ill
-------
TABLE OF CONTENTS
EPA Review Notice i:L
_ ,. iii
Preface
List of Figures v
List of Tables v
viii
Summary
x
Glossary
1.0 Introduction
2.0 Solidification of Radioactive High-Level Liquid Wastes 2
3.0 Generic Solidification Plant 5
4.0 Off-Gas Treatment, Decontamination Factors and Discharge
Rates at the Generic Solidification Plant 8
4.1 Off-Gas Treatment °
4.2 Decontamination Factors 9
A. Tritium Removal 10
B. Iodine-129 Removal 10
C. Ruthenium-106 Removal 10
D. Particulate Removal 12
4.3 Discharge Rates of Generic Solidification Plant 13
5.0 Estimates of Annual Dose Equivalents 15
5.1 Computer Code Input Parameters 15
5.2 Results 16
6.0 Discussion and Conclusions 20
23
7.0 References tj
Appendix A: Waste Calcination and Classification Processes A-l
A. 1 Calcination A"2
A.2 Waste Classification A"6
IV
-------
Appendix B: DOE and NRC Generic Solidification Plant Studies B-1
B.1 NRC Contract Studies n 0
c—d
B.2 DOE Contract Studies D r-
D-D
Appendix C: Selected Tables From AIRDOS-EPA Computer Program C-1
LIST OF FIGURES
Figure 4.1 Off-Gas Treatment System Decontamination Factors
(DF) in Waste Calcination and Classification
Process
Figure A.1 Fluidized-Bed Calciner
Figure A.2 Spray Calciner
Figure A.3 In-Can Melter
Figure A.4 Continuous Melter
LIST OF TABLES
14
A-4
A-5
A-8
A-9
Table 3.1
Table 3.2
Table 4.1
Table 4.2
Table 4.3
Table 4.4
Table 5.1
Table 5.2
Radionuclide Inventory of Spent Fuel Prior to
Reprocessing and Solidification
Radionuclide Inventory of the HLLW Feed to the
Generic Solidification Plant
Approximate Decontamination Factors for
Iodine-129 Removal Technologies
Approximate Decontamination Factors for
Ruthenium-106 Removal Technologies
Approximate Decontamination Factors for
Particulate Removal Technologies
Discharge Rates and Decontamination Factors
for the Generic Solidification Plant
Annual Individual Dose Due to Releases from
Generic Solidification Plant
Annual Population Dose Due to Releases from
Generic Solidification Plant
11
11
12
13
18
19
-------
Table 6.1 Comparison of Annual Dose Equivalents from the
Generic Solidification Plant with the Annual Dose
Equivalent Limit under the UFC Standards 22
Table 6.2 Comparison of Releases from the Generic
Solidification Plant with Release Limits Under
the UFC Standards 22
Table B. 1 Decontamination Factors Expected During the
Calcination and Classification of HLLW B-3
Table B.2 Maximum Annual Dose Equivalents to an Individual
Due to Gaseous Releases from a Generic HLLW
Solidification Facility B-3
Table B.3 Maximum Annual Dose Equivalents to an Individual
at the Site Boundary Due to Gaseous Releases from
HLLW Solidification Plant B-M
Table B.4 Summary of Estimated Decontamination Factors for
Solidification Processes B-5
Table B.5 Maximum Annual Dose Equivalents to an Individual
Due to Gaseous Releases from Calcination and
Classification Facilities B-6
Table C.1 List of Input Values for Radionuclide-Independent
Variables C-3
Table C.2 Computed Values for the Area (Rural and Urban) C-5
Table C.3 List of Input Data for Nuclide H-3 C-6
Table C.4 List of Input Data for Nuclide Sr-90 C-7
Table C.5 List of Input Data for Nuclide Ru-106 C-8
Table C.6 List of Input Data for Nuclide 1-129 C-9
Table C.7 List of Input Data for Nuclide Cs-134 C-10
Table C.8 List of Input Data for Nuclide Cs-137 C-11
Table C.9 List of Input Data for Nuclide Pu-239 C-12
Table C.10 Maximum Annual Dose to a Rural Individual from
One-Year-Decayed Spent Fuel C-13
vi
-------
Table C.11 Annual Dose to the Rural Population from
One-Year-Decayed Spent Fuel c_m
Table C.12 Maximum Annual Dose to an Urban Individual from
One-Year-Decayed Spent Fuel c_15
Table C.13 Annual Dose to the Maximum Urban Individual from
Ten-Year-Decayed Spent Fuel C_-J5
Table C.14 Maximum Annual Dose to a Rural Individual From
Five-Year-Decayed Spent Fuel c_17
Table C.15 Annual Dose to the Rural Population from
Five-Year-Decayed Spent Fuel C-18
Table C.16 Maximum Annual Dose to an Urban Individual from
Five-Year-Decayed Spent Fuel C-19
Table C.17 Annual Dose to the Urban Population from
Five-Year-Decayed Spent Fuel c_2o
Table C.18 Maximum Annual Dose to a Rural Individual From
Ten-Year-Decayed Spent Fuel C-21
Table C.19 Annual Dose to the Rural Population from
Ten-Year-Decayed Spent Fuel C-22
Table C.20 Maximum Annual Dose to an Urban Individual from
Ten-Year-Decayed Spent Fuel C_23
Table C.21 Annual Dose to the Urban Population from
Ten-Year-Decayed Spent Fuel c_24
vii
-------
SUMMARY
The Office of Radiation Programs, U.S. Environmental Protection
Agency (ORP/EPA), has prepared this analysis as technical support for
EPA's proposed environmental radiation protection standards, 40 CFR 191,
concerning the management and disposal of high-level radioactive wastes.
For Subpart A of 40 CFR 191, waste management and storage operations,
EPA proposes to extend the limitations of 40 CFR 190 to these
operations.
EPA/ORP developed a generic high-level liquid waste solidification
plant and assessed the potential environmental impact of atmospheric
discharges during normal operations in four solidification processes:
fluidized-bed calcination, spray calcination, and glassification by
in-can melting and continuous melting. We used a newly developed
computer code, AIRDOS-EPA, to perform the assessment.
Our assessment involves seven radionuclides that account for 885& of
the doses due to the solidification process: H-3, 1-129, Ru-106, Cs-134,
Cs-137, Sr-90, and Pu-239. We estimated the decontamination factors for
typical off-gas equipment components to remove these radionuclides and
an overall off-gas cleanup system decontamination factor.
For purposes of comparison, we based our assessment on two
hypothetical plant sites with widely different population size, food
sources, and weather: an urban site, St. Louis, Missouri; and a rural
site in the southeastern United States typified by the South Carolina
sites of the Federal Government's Savannah River Plant and the
commercial Barnwell Nuclear Fuel Plant.
We estimated off-gas releases during normal operations of the
generic solidification plant and the resulting annual individual and
population dose equivalents. We compared the doses to individuals and
viii
-------
the quantities of radioactive materials released with the limits in
EPA's standards for the Uranium Fuel Cycle (UFC), 40 CFR 190.
Our assessment indicates that for fuel decayed for one year the
maximum annual doses to an individual due to releases from a
solidification facility at a rural site would be lower than the 40 CFR
190 standards; that maximum annual doses from a facility at an urban
site would exceed the UFC standards. In the case of the radionuclide
waste products that have decayed for five years or longer, the maximum
annual dose to an individual at either site is lower than the 40 CFR 190
standards. Additional off-gas treatment for the solidification facility
can also reduce the maximum annual doses.
IX
-------
GLOSSARY
ABBREVIATIONS
AEC - U.S. Atomic Energy Commission
40 CFR 190 - Title UO, Part 190, Code of Federal Regulations
Ci - Curie
DF - Decontamination Factor
DOE - U.S. Department of Energy
ERDA - U.S. Energy Research and Development Administration
EPA - U.S. Environmental Protection Agency
GWe - Gigawatts electrical; giga is a thousand million
HEPA - High-Efficiency Particulate Air Filter
HLLW - High-Level Liquid Wastes
INEL - Idaho National Engineering Laboratory
LLI - Lower Large Intestine
LWR - Light-Water Reactor
MWd - Megawatt days; mega is a million
MTHM - Metric tons of heavy metals (i.e. uranium and plutonium)
MTU - metric tons of uranium
NRC - U.S. Nuclear Regulatory Commission
ORNL - Oak Ridge National Laboratory
ORP - EPA's Office of Radiation Programs
PNL - Pacific Northwest Laboratories
UFC - Uranium Fuel Cycle
WCF - Waste Calcining Facility
-------
TERMS
Actinides - The series of elements beginning with element No. 89,
actinium and continuing through element No. 104.
Annual Dose - The dose received by an individual or a population from
one year's release. It is the sum of the external dose
received that year plus the 70-year dose commitment from
internal radioactive material.
Burnup - A measure of reactor fuel consumption. It is usually
expressed as either (a) the percentage of uranium atoms
that have undergone fission or (b) as thermal energy
produced per quantity of nuclear fuel (i.e. megawatt-days
per metric ton).
Calcination - method of converting the solids in solution to a solid by
atomizing and coating the liquid on small granules and
heating to drive off the water.
Calcine - The resulting solid granule product from calcination.
Curie - The basic unit to describe the intensity of radioactivity
in a material.
Decay (Radioactive) - The spontaneous transformation of one nuclide into
a different nuclide or into a different energy state of
the same nuclide usually resulting in the release of
ionizing radiation. The process results in a decrease,
with time, of the number of the original radioactive atoms
in a sample.
Decontamination Factor - The ratio of the amount of a given type of
radioactive material entering a process (or process step)
to that which leaves the process (or process step).
Deposition Velocity - The ratio of the deposition rate to the
ground-level concentrations.
Dose Commitment - Radionuclides which enter the body through ingestion
or inhalation remain in the body as a continuing source of
exposure for a length of time determined by biological and
physical factors. The dose is cumulative and is
evaluated in this report for 70 years and is included in
the annual doses.
Dose Equivalent - A term used to express the amount of effective
radiation when modifying factors have been considered such
as quality and distribution factors.
xi
-------
E+00 Format - Throughout this report, numeric values are frequently
~~ expressed in a modified scientific format. For example,
0.00456 = 4.56 X 10"^ may be expressed as 4.56 E-03 and
78900 = 7.89 X 10 as 7.89 E+04.
Fission Products - The radionuclides and their decay products formed by
the fission of heavy elements.
Fuel Enrichment - Material such as uranium in which the percentage of a
given isotope present has been artificially increased, so
that it is higher than the percentage of that isotope
naturally found in the material.
Fuel Reprocessing - The processing of spent reactor fuel to recover the
unused fissionable uranium and plutonium.
Fluidized Bed - A cushion of air or hot gas blown through the porous
bottom slab of a container which can be used to float a
powdered material as a means of drying, heating or
calcining the immersed object.
Generic - Characteristic of a whole class.
Glass Frit - The calcined or partly fused materials of which glass is
made.
Classification - Incorporation of wastes into a glass matrix.
High-Level Liquid Waste - The aqueous waste resulting from the operation
of the chemical extraction systems in a facility for
processing spent nuclear fuel.
High-Level Waste - High-level liquid waste, or the products from
solidification of high-level liquid waste, or spent fuel
elements, if discarded without processing.
Off-Gas - The normal gasborne discharge from any process vessel or
other process equipment.
Scavenging Coefficient - The fraction of material reaching the ground
per unit time due to the collection of particles and gases
by cloud or precipitation droplets.
Spent Fuel - Any fuel removed from a nuclear reactor after it has been
irradiated, usually to the extent that it can no longer
effectively sustain a chain reaction.
XII
-------
1.0 INTRODUCTION
The Office of Radiation Programs (ORP), U.S. Environmental
Protection Agency (EPA), is proposing generally applicable environmental
radiation protection standards for management and disposal of spent
nuclear fuel, high-level and transuranic radioactive wastes (1). These
proposed standards would become Part 191 of the Code of Federal
Regulations. Title 40-(40 CFR 191).
Subpart A of the proposed standards applies to normal waste
management operations, which include preparation for storage or disposal
(solidification of high-level liquid wastes, packaging of spent fuel),
storage, and emplacement in a disposal repository.
Since the UFC standards exclude waste management operations,
ORP/EPA prepared this radiation exposure analysis of airborne emissions
as technical support for EPA's proposed 40 CFR 191 standards. For
practical purposes the basic assumption for this analysis is that the
only radioactive materials entering the general environment from a
solidification facility are airborne discharges to the atmosphere;
liquid releases or accidental releases were not considered. For Subpart
A of the proposed standards, EPA proposes to extend the limitations of
40 CFR 190 to these operations.
-------
2
2.0 SOLIDIFICATION OF RADIOACTIVE HIGH-LEVEL LIQUID WASTES
High-level liquid wastes (HLLW) are generated during the chemical
reprocessing of spent nuclear fuel to recover uranium and plutonium. As
of 1977, the inventory of high-level liquid wastes from Federal
reprocessing of spent fuel amounted to about 0.3 million cubic meters,
containing about 400 to 600 million curies. Most of these wastes have
been reduced to solids or semi-liquids in the form of salt cake,
crystals, sludges, and calcine. However, the government will probably
continue to generate liquid wastes at a rate of a few thousand cubic
meters per year. The Department of Energy (DOE) stores the wastes from
Federal reprocessing plants at the Hanford Reservation in Washington,
the Savannah River Plant in South Carolina, and the Idaho National
Engineering Laboratory (INEL) in Idaho (2-6).
At Hanford, DOE is converting its own high-level liquid wastes to a
salt cake, which is temporarily stored in underground tanks along with
residual sludge and liquor; cesium-137 and strontium-90 are separated
and stored in aboveground facilities. At the Savannah River Plant, DOE
converts its HLLW to salt cake, without separating cesium and strontium.
DOE's facility at INEL converts its HLLW to a granular calcine and
stores it in specially designed underground vaults (2-6).
Federal policy at the present time is to defer commercial
reprocessing of spent fuels from the nuclear power industry (7).
Therefore, most spent fuel from commercially operated reactors is
unprocessed and in temporary storage. As of 1976, commercial spent fuel
in storage amounted to about 2343 metric tons of uranium (8). If the
Federal Government permits reprocessing in the future, each metric ton
processed will produce about five cubic meters of high-level liquid
wastes (9).
-------
3
A small amount of high-level liquid wastes from commercial
reprocessing of spent fuel - about 17 thousand cubic meters, containing
approximately 40 million curies - is stored at the Nuclear Fuel
Services fuel reprocessing plant near West Valley, New York (10).
Nuclear Regulatory Commission (NRC) regulations require that
commercially produced high-level liquid wastes be converted to a stable
solid form within five years after they are generated and then
transferred to the Federal government for permanent disposal (11).
Solidification immobilizes the wastes in order to isolate them from the
environment. It also reduces the volume of wastes requiring storage by
80 to 90 percent.
Numerous solidification processes have been developed throughout
the world; many have been demonstrated by pilot-plant or plant-scale
operation (see Table B.4). Of the many technologies two seem to have
emerged as the most prominent - calcination and glassification.
As part of the Government's waste management program, DOE has
developed solidification alternatives for commercial and Federal HLLW.
Among the solidification processes DOE has proposed are calcination, "
which converts HLLW to a granular powder; and glassification, which'
incorporates the powder into a solid matrix that serves as an engineered
barrier in preventing or delaying migration of the radionuclides to the
environment (12-18).
During normal operations of a solidification plant, some of the
radionuclides in the wastes are released to off-gas streams as volatile
gases and particulates. Before release to the atmosphere these
off-gases are routed to treatment systems to remove the radionuclides.
The amount and concentration of radionuclides in the plant's exhaust
stack discharge depends on the amount and concentration of radionuclides
in the high-level liquid waste feed to the plant and on the
-------
effectiveness of the treatment systems in removing the radionuclides
from the off-gas streams before their release to the atmosphere.
Several major factors can affect the potential radiation dose to
individuals and populations as a result of the discharge: proximity to
the plant, the pathways by which the radionuclides can reach them, the
length of time during which the radionuclides continue to pose a health
hazard, decay time, meteorological factors, plant capacity, and off-gas
treatment. The radioactive decay of the fission products and actinides
in the fuel during storage before reprocessing and in the liquid waste
before solidification causes a significant reduction in the amount of
radioactive materials.
-------
5
3.0 GENERIC SOLIDIFICATION PLANT
For the purposes of this analysis, the EPA developed a generic
solidification plant for the calcination and glassification of
high-level liquid wastes. Input data on actual solidification plant
experience came from the Government's Waste Calcination Facility (WCF)
in Idaho; input data on hypothetical HLLW were developed from proposed
commercial spent fuel reprocessing plants. The analysis applies to both
Government and commercial wastes. They contain the same major
radionuclides. However, Government wastes are less radioactive and less
thermally active because of different decay time for the fission
products, different enrichments and burnup percentages, and different
reactor operational characteristics (3, 13-15).
We chose the four most promising and advanced solidification
processes: fluidized-bed and spray calcination; and glassification by
in-can melting and continuous melting. (See Appendix A.)
From the hundreds of fission-product and actinide radionuclides, we
selected seven for our analysis: tritum (H-3), iodine-129 (1-129),
ruthenium-106 (Ru-106), cesium-134 (Cs-134), cesium-137 (Cs-137),
strontium-90 (Sr-90) and plutonium-239 (Pu-239). We selected them
because of their adverse health effects, high dose-equivalent conversion
factors, half-lives, high release rates, and the fraction of the nuclide
released to the environment. These seven radionuclides account for more
than an estimated 88% of the maximum doses to the major organs of adults
due to releases from the solidification of HLLW (19).
The feed rate to the generic solidification plant is the HLLW
generated from the reprocessing of 1500 MTHM (metric tons of heavy
metal) per year of spent fuel from light-water reactors.
-------
The radionuclide inventory of this HLLW feed is determined by the
radioactive inventory of the spent fuel, the length of time during which
the spent fuel decayed before reprocessing, and the length of time the
HLLW and fission products were in storage before solidification, and the
radionuclide percentage carryover from spent fuel reprocessing.
The initial radionuclide inventory of the spent fuel prior to
reprocessing is based on an average burnup in a commercial pressurized
light-water reactor of 33,000 megawatt days thermal per MTHM at a
continuous power of 38.4 megawatts per MTHM. The original fuel
enrichment averaged 3.3%. Table 3.1 gives the calculated inventory of
the seven selected radionuclides, in the spent fuel after decaying for
one, five, and ten years (3, 20).
The radionuclide carryover in the HLLW from spent fuel reprocessing
is 5% of the tritium, 5% of the iodine, over 99% of the nonvolatile
fission products, and 1% of the plutonium (21). Table 3.2 gives the
calculated radionuclide inventory of the HLLW feed to the generic
solidification plant.
-------
TABLE 3.1
RADIONUCLIDE INVENTORY OF SPENT FUEL PRIOR TO
REPROCESSING AND SOLIDIFICATION
RADIONUCLIDE
HALF-LIFE
1 YEAR
DECAY PERIOD
5 YEARS
10 YEARS
H-3
1-129
Ru-106
Cs-137
Cs-134
Sr-90
Pu-239
(years)
12.3
1.7 E+07
1.01
30.0
2.05
28.1
2.40 E+04
(curies per MTHM)
6.91 E+02
3.77 E-02
3.23 E+05
1.06 E+05
1.92 E+05
7.49 E+04
3.31 E+02
5.5 E+02
3.77 E-02
2.12 E+04
9.70 E+04
4.98 E+04
6.78 E+04
3.31 E+02
4.16 E+02
3.77 E-02
6.50 E+02
8.64 E+04
9.18 E+03
6.00 E+04
3.31 E+02
TABLE 3.2
RADIONUCLIDE INVENTORY OF THE HLLW FEED
TO THE GENERIC SOLIDIFICATION PLANT
RADIONUCLIDE
H-3
1-129
Ru-106
Cs-137
Cs-134
Sr-90
Pu-239
DECAY PERIOD
1 YEAR
5.19 E+04
2.84 E+00
4.80 E+08
1.59 E+08
2.88 E+08
1.12 E+08
5.00 E+03
5 YEARS
(curies per
4.13 E+04
2.84 E+00
3.18 E+07
1.46 E+08
7.39 E+07
1.02 E+08
5.00 E+03
10 YEARS
year*)
3.12 E+04
2.84 E+00
9.75 E+05
1.30 E+08
1.36 E+07
9.00 E+07
5.00 E+03
*Jased on a reprocessing plant capacity of 1500 MTHM per year of spent
-------
8
4.0 OFF-GAS TREATMENT, DECONTAMINATION FACTORS AND DISCHARGE RATES
AT THE GENERIC SOLIDIFICATION PLANT
The reduction of discharge rates from any solidification process
occurs based on three factors: the off-gas treatment; the off-gas
treatment system's decontamination factors; and the radionuclide decay
time measured from the time the spent fuel was discharged from the
reactor.
4.1 Off-Gas Treatment
Off-gas treatment reduces the discharge of airborne radioactive
materials to the environment. The equipment and systems discussed in
sections 4.1 and 4.2 present a brief review of the existing
technologies. Additional and more detailed information on equipment is
presented in references 9, 23, 53.
During calcination and glassification of HLLW, the tritium, iodine,
and part of the ruthenium will volatilize; the cesium, strontium,
Plutonium, and a small fraction of the ruthenium will become entrained
as particulates in the process1 off-gas streams going to the plant's
off-gas treatment system. Off-gas treatment technologies are readily
available, and operational information is available on many components
and systems.
Gaseous radionuclides are usually removed by chemical treatment
systems, such as sorption techniques, catalyst reactions, or
distillation. Particulates are usually removed by inertial separation
(cyclone or gravity settling), filtration (fabric, glassfil, sandbeds,
HEPA), precipitation (electric, thermal), sonic agglomeration, or liquid
-------
scrubbing. Final filtration is either through deep beds of sand,
fiberglass filters, or compact high-efficiency particulate air (HEPA)
filters.
Off-gas treatment systems are used at reactors, spent fuel
reprocessing facilities, fuel fabrication facilities, and the INEL's
Waste Calcination Facility. The Waste Calcination Facility off-gas
treatment system removes both particulates and gaseous products (except
tritium) and includes scrubbing, filtering and absorption (9, 22-28).
^•2 Decontamination Factors
The effectiveness of an off-gas treatment component or system in
removing a particular radionuclide from a plant's off-gas streams is
measured by the decontamination factor (DF), which is the ratio of the
concentration of a radionuclide before treatment to that after
treatment. The estimated DF of a total treatment system includes the
DFs of individual components or integrated systems.
The best available technology for off-gas treatment was chosen.
The DFs are taken from the available literature. The overall DF for the
EPA generic solidification plant assumes that the calcination and
glassification are a combined process and that the off-gasses pass
through several systems which selectively remove the various
radionuclides. Existing technology permits radionuclide removal systems
of almost any design. In some cases DF ranges are shown for the
technologies because of differences in the data sources, varying
operating conditions, and EPA conservatism.
-------
10
A. Tritium Removal
The technical and economic feasibility of tritium control is still
under investigation. Therefore we will assume that the DF for tritium
in the calcination and glassification processes is one (9, 23, 29).
B. Iodine-129 Removal
Removal processes for radioactive iodine include aqueous scrubbing
(reactive sprays, towers, wet filters) and adsorption (charcoal,
activated charcoal, silver and other metallic zeolite adsorbents). Table
4.1 lists the known DFs for the various iodine-129 removal technologies
(9, 23, 29-33). For iodine removal, the generic plant off-gas system
consists of a mercuric nitrate-nitric acid scrubber and a
silver-impregnated adsorber. The overall DF is estimated to be 1 E+03.
C. Ruthenium-106 Removal
In a high-temperature solidification process, ruthenium may be in
the off-gas stream as both a gas and a particulate. Personnel at the
Waste Calcining Facility at the Idaho Chemical Processing Plant
estimated that DF for the total off-gas treatment system for volatilized
ruthenium is about 1.0 E+07 (9, 22, 23, 25-32, 34-36). Table 4.2 lists
the known DFs for the various ruthenium-106 removal technologies, for
ruthenium removal, the generic plant off-gas system consists of the
process cyclone, quench tank venturi scrubber, silica gel adsorber and
HEPA filters. For particulate ruthenium the overall DF is estimated at
1 E+10.
-------
11
TABLE 4.1
APPROXIMATE DECONTAMINATION FACTORS
FOR IODINE-129 REMOVAL TECHNOLOGIES
TECHNOLOGYDECONTAMINATION FACTORS"
Caustic Scrubbing 2 E+00 to ^ E+Q1
Silver-Impregnated Adsorbents 1 £+02 to 1 E+05
Metallic Zeolite Adsorbents (non-silver) 1 E+01
Mercuric Nitrate-Nitric Acid Scrubbing 1 E+01 to 1 E+02
lodox Process ! E+04 to ^ E+Q6
Charcoal Filters ! E+01 to ^ E+Q2
TABLE 4.2
APPROXIMATE DECONTAMINATION FACTORS FOR
RUTHENIUM-106 REMOVAL TECHNOLOGIES
DECONTAMINATION FACTOR^'
TECHNOLOGY AND PARTICULATE VOLATILIZED
COMPONENTS RU VU^IILI^LU
Calciner and Cyclone 1 E+01 to 4 E+01 1 E+03 to 1 E+04
Scrubbing System 4 E+01 to 6 E+02 1 E+01 to 2 E+01
Silica Gel Adsorbers 3 E+00 to 8 E+00 8 E+02 to 1 E+03
HEPA Filters 1 E+03 1 E+00
-------
12
D. Particulate Removal
The major radioactive particulates associated with the
solidification processes are cesium-134, cesium-137t strontium-90,
ruthenium-106 and actinides such as plutonium-239. Table 4.3 lists
approximate DFs for several types of filtration components (9, 22-25,
27, 32, 34, 36-39). For particulate removal, the generic plant off-gas
system uses the existing process cyclone, the wet scrubbing system and
the adsorbers; for final filtration and particulate removal the off-gas
system relies on HEPA filters and either a deep bed glass filter system
or sintered metal filters. For particulate removal the overall DF is
estimated at 1 E+10.
TABLE 4.3
APPROXIMATE DECONTAMINATION FACTORS
FOR PARTICULATE REMOVAL TECHNOLOGIES
COMPONENT
Prefilters
Sand Bed Filters
Deep Bed Glass Filters
HEPA Filters
Sintered Metal Filters
Scrubbing Systems
DECONTAMINATION FACTOR
6 E+00 to 1 E+01
1 E+01 to 1 E+02
1 E+02 to 1 E+04
1 E+03
1 E+03 to 1 E+05
1 E+01 to 1 E+02
-------
13
*'3 Discharge Rates of Generic Solidification Plant
After determining the treatment system DFs for the seven major
radionuclides we estimated the discharge rates to the atmosphere during
normal operations of the generic solidification plant. Table 4.4 shows
the DFs and the discharge rates based on decay periods of one, five and
ten years. Figure 4.1 is a schematic of the off-gas treatment systems
DFs for the HLLW calcination and glassification processes.
TABLE 4.4
DISCHARGE RATES AND DECONTAMINATION FACTORS
FOR THE GENERIC SOLIDIFICATION PLANT
RADIONUCLIDE
DF
1 YEAR
DECAY PERIOD
5 YEARS
10 YEARS
H-3
1-129
Ru-106
Cs-137
Cs-134
Sr-90
Pu-239
1 E+00
1 E+03
1 E+07
1 E+10
IT"1 A f\
E+10
1 E+10
1 E+10
5.19 E+04
2.84 E-03
4.80 E+01
1.59 E-02
2.88 E-02
1.12 E-02
5.00 E-07
4.13 E+04
2.84 E-03
3.19 E+00
1.46 E-02
7.39 E-03
1.02 E-02
5.00 E-07
3. 12 E+04
2.84 E-03
9.75 E-02
1.30 E-02
1.36 E-03
9.00 E-03
5.00 E-07
-------
OFF-GAS TREATMENT SYSTEM DECONTAMINATION FACTORS (DF)
IN WASTE CALCINATION AND CLASSIFICATION PROCESS
STACK
EFFLUENT
RAW
FEED
(HLLW)
CALCINATION
AND
CLASSIFICATION
PROCESSES
^-
SCRUBBING
SYSTEM
O
-^~
FF-GAS CLEANUP
ADSORBENTS
-
SYSTEM
HEPA
FILTERS
SPECIAL
FILTERS
i
SPECIES
TRITIUM
IODINE-129
RUTHENIUM 106
PARTICULATES:
e.g. CS-134/137
Sr-90
Ru-106
Pu-239
COMPONENT DFs
1
1
3.2 x 103
10
1
10
10
102
1
102
3.2 x 102
10
1
1
1
103
1
1
103
TOTAL DF
1
103
107
1010
Figure
-------
15
5.0 ESTIMATES OF ANNUAL DOSE EQUIVALENTS
We estimated the annual dose equivalents (hereafter referred to
simply as annual doses) to individuals and populations due to discharges
from the generic solidification plant. For purposes of comparison, we
based our assessments on two hypothetical sites with widely different
demographic, meteorologic, and agricultural characteristics: an urban
midwestern site at St. Louis, Missouri; and a rural site in the
southeastern United States located adjacent to the commercial Barnwell
Nuclear Fuel Plant and the Government's Savannah River Plant (5, 12,
40). We assumed the discharges listed in Table 4.4 and estimated annual
doses to individuals and to the population within 80 kilometers of the
plant at each site. The estimates include doses to the total body,
thyroid, red bone marrow, lungs, endosteal cells, stomach wall, lower
large intestine wall, liver, kidneys, testes, and ovaries. The computer
program we used evaluates seven pathways: immersion in air containing
radionuclides, exposure to contaminated land surfaces, immersion in
contaminated water, inhalation of radionuclides in air, and ingestion of
meat, milk, and leafy vegetables and fresh produce grown in the area
(41).
The use of the reference site, rural or urban, should not be
construed as an endorsement of any particular region for siting of
radioactive waste management facilities, but rather as a means of
dealing with site-specific aspects for comparative radiation exposures.
5.1 Computer Code Input Parameters
A newly developed computer code called AIRDOS-EPA performed the
calculations (41). Appendix C contains the AIRDOS-EPA computer code
printouts relevant to the input data and the annual doses to individuals
and the population.
-------
16
Meteorological input data and other characteristics for the rural
site came from the final environmental impact statements on the Barnwell
Nuclear Fuel Plant and the Savannah River Plant (5, 42). The
meteorological input data for the urban site came from the National
Climatic Center in Asheville, North Carolina.
All of the releases from the generic solidification plant are
through a 62-meter high stack. The gravitational fall velocity in all
cases is zero; the deposition velocity, 0.01 meter per second (except
zero for tritium); and the scavenging coefficient, 1.19 E-05 per second.
NRC developed the information used as agricultural input data (43).
The characteristics of the generic urban site and the population data
for both sites were taken from information developed for EPA (44). The
individual and population doses calculated by the AIRDOS-EPA computer
program include both an annual external dose and a 70-year internal dose
commitment from one year's release. ORNL developed the dose conversion
factors for the seven pathways as input data for each radionuclide and
reference organ. These dose conversion factors were used by the
computer code to calculate dose commitments from one year's release
(41).
5.2 Results
Table 5.1 shows the annual individual dose to the most significant
organ of interest and the radionuclides delivering the highest
percentage of the dose. Table 5.2 shows the annual dose to the
population within 80 kilometers of the generic plant at the rural and
urban sites. The doses are due to exposure to the radionuclide waste
products from spent fuel that have decayed one year, five years, and ten
years before reprocessing and solidification.
-------
17
Many factors affect the dose received by an individual:
meteorological patterns, radionuclide activity at time of exposure, the
significant pathways for exposure, and the proximity to the source of
release. The maximum annual individual doses at the urban and rural
sites occur at approximately 1000 and 3000 meters, respectively, from
the release point. The main reason individuals receive higher doses at
the urban site than at the rural site is because they are closer to the
plant. Population doses are also higher at the urban site because there
are more people closer to the plant.
Tables C.10 through C.21 in Appendix C show the annual individual
dose and population dose for each of the eleven body organs and seven
radionuclides. If the spent fuel decays for one or five years before
reprocessing and solidification, most of the dose is due to exposure to
Ru-106, with tritium the second largest contributor. The largest organ
dose is to the lower large intestine (LLI) wall.
If the spent fuel decays for ten years before reprocessing and
solidification, most of the dose is from tritium. The largest organ
dose is to the thyroid because of exposure to tritium and 1-129.
The pathways through which the highest percentages of dose are
deposited follow the same general trends no matter whether the target is
urban or rural, or an individual or a population. One year decayed fuel
delivers its dose mainly through the surface and ingestion pathways. As
the fuel is decayed longer the importance of the surface exposure
decreases to a very small percentage (less than 10*) of the total dose
while the ingestion pathway grows in importance (70 - 80% with ten year
decayed fuel). Also with longer decayed fuel the inhalation pathway
gains importance to a maximum of 15 - 25%.
-------
18
TABLE 5.1
ANNUAL INDIVIDUAL DOSE DUE TO RELEASES
FROM GENERIC SOLIDIFICATION PLANT*
NUCLIDE
TOTAL BODY
LLI WALL
THYROID
1 YR
DECAY
5 YR
DECAY
10 YR
DECAY
(millirem per year)
1 YR
DECAY
5 YR
DECAY
10 YR
DECAY
(millirem per year)
10 YR
DECAY
H-3
Ru-106
1-129
All
0.5
1.5
2.1
H-3** 1.6
Ru-106 12.5
1-129
All 14.5
0.4
0.1
0.6
2.5
0.6
3.3
0.3
0.0
0.3
1.9
0.02
2.1
RURAL SITE
0.5
23.1
23.7
URBAN SITE
1.5
190
192
0.4
1.5
1.9
1.3
12.2
13.7
0.3
0.05
0.4
1.9
0.3
2.3
0.3
0.2
0.5
1.9
1.0
3
*Includes only the most significant organ doses from the radionuclides
delivering the highest percentage of dose.
**The H-3 doses do not decrease as would be expected because of the
method by which the AIRDOS-EPA computer program evaluates the summary
results. The program selects the highest individual dose to the organ
of interest and reports the dose contribution of each radionuclide to
that individual. Therefore, the doses listed are not necessarily the
maximum dose from that radionuclide to any individual in the population
but rather are the dose contributions to the individual receiving the
highest organ dose from all nuclides. In the case of H-3. a different
individual was involved for each decay period.
— Negligible
-------
19
TABLE 5.2
ANNUAL POPULATION DOSE DUE TO RELEASES
FROM GENERIC SOLIDIFICATION PLANT*
NUCLIDE
TOTAL BODY
1 YR
DECAY
5 YR
DECAY
10 YR
DECAY
(man-rem per year)
LLI WALL
1 YR 5 YR
DECAY DECAY
10 YR
DECAY
(man-rem per year)
THYROID
10 YR
DECAY
H-3 8.6 6.8 5.1
Ru-106 19.6 1.2 0.04
1-129
All 28.9 8.6 5.6
H-3 66.9 53.3 40.2
Ru-106 195 12.6 0.4
1-129
All 269 71 44.8
RURAL SITE
8.6
135
144
URBAN SITE
67.6
1072
1146
6.9
8.7
16
54
69
127
5.2
0.3
5.8
40.9
2.2
46
5.2
1.5
7.0
40
10.5
54
•Includes only the most significant organ doses from the radionuclides
delivering the highest percentage of dose.
— Negligible
-------
20
6.0 DISCUSSION AND CONCLUSIONS
Under the EPA environmental standards for the uranium fuel cycle
(UFC), 40 CFR 190, normal operations are to be conducted in such a
manner as to provide reasonable assurance that: (a) the annual dose
equivalent does not exceed 25 millirems to the whole body, 75 millirems
to the thyroid, and 25 millirems to any other organ of any member of the
public as the result of exposures to planned discharges of radioactive
materials to the general environment from uranium fuel cycle operations
and to radiation from these operations; (b) the total quantity of
radioactive materials entering the general environment from the entire
uranium fuel cycle, per gigawatt-year of electrical energy produced by
the fuel cycle, contains less than 50,000 curies of krypton-85, 5
millicuries of iodine-129, and 0.5 millicuries combined of plutonium-239
and other alpha-emitting transuranic radionuclides with half-lives
greater than one year (45).
Since the UFC standards exclude waste management operations,
ORP/EPA prepared this analysis as technical support for EPA's proposed
environmental radiation protection standards, 40 CFR 191, concerning
management and disposal of high-level radioactive wastes. For Subpart A
of 40 CFR 191, waste management and storage operations, EPA proposes to
extend the limitations of 40 CFR 190 to these operations.
We compared the estimated maximum annual doses to an individual at
the two plant sites with the annual dose limits under the UFC standards.
(See Table 6.1). We also compared estimated releases from the generic
solidification plant with the release limits under the UFC standards.
(See Table 6.2).
-------
21
In the case of the radionuclide waste products that have decayed
one year, our assessment indicates that maximum annual doses to an
individual due to releases from a solidification facility at a rural
site would be less than the 40 CFR 190 standards; that maximum annual
doses from a facility at an urban site would exceed the UFC standards.
However, in the case of the radionuclide waste products that have
decayed for five years or longer, the maximum annual dose to an
individual at either site would be less than 15 millirem.
Our assessment of the releases of the radionuclide waste products
that have decayed for one year indicates that releases of krypton-85,
iodine-129, and plutonium-239 are less than the allowable UFC release
limits by at least a factor of 100.
The quantity of radionuclides in releases from a solidification
plant is primarily determined by the DFs of the radionuclide removal
systems. Additional off-gas components will change a plant's DF and
reduce the quantity of radionuclides released to the environment. Plant
siting is an important factor, as shown by the comparisons of doses due
to releases from urban and rural plant sites. Urban characteristics
(e.g. population, food crops produced or imported for local consumation,
meat and diary animals) contribute to larger doses. The length of time
the spent fuel's radionuclide waste products decay before reprocessing
and solidification is also an important factor. Increasing the decay
time from one to five years, for example, will reduce the dose to the
lower large intestine wall by an order of magnitude. (Tables 5.1 and
6.1)
Since many solidification processes, as well as final waste forms
(i.e. crystalline, cement, and metal matrices), are under development
throughout the world, improvements are possible. All limitations have
not necessarily been identified.
-------
22
TABLE 6.1
COMPARISON OF THE ANNUAL DOSE EQUIVALENTS FROM THE GENERIC
SOLIDIFICATION PLANT WITH THE
ANNUAL DOSE EQUIVALENT LIMIT UNDER THE UFC STANDARDS
UFC
ORGAN DOSE LIMIT
ESTIMATED ANNUAL DOSES
TO MAXIMALLY EXPOSED INDIVIDUAL
(one-year-decayed fuel)
RURAL SITE URBAN SITE
(millirem/yr)
Total body
Thyroid
Other organs
lungs
liver
bone
endosteal cells
stomach wall
kidneys
lower large
intestine wall
testes
ovaries
25
75
25
25
25
25
25
25
25
25
25
(millirem/yr)
2.1
2.2
3.5
2.1
2.4
2.7
2.1
2.1
23.7
2.3
1.8
(millirem/yr)
14.5
15.7
21.3
14.8
17.3
19.6
14.2
14.9
191.9
16.3
11.8
TABLE 6.2
COMPARISON OF RELEASES FROM THE GENERIC SOLIDIFICATION
PLANT WITH RELEASE LIMITS UNDER THE UFC STANDARDS
RADIONUCLIDE
UFC STANDARDS
RELEASE LIMIT
UFC STANDARDS
RELEASE
LIMIT EQUIVALENT(a)
ESTIMATED
GENERIC
SOLIDIFICATION
PLANT RELEASE
(One-Year decay)
Krypton-85
Iodine-129
Alpha (Pu-239)
H-3
Ru-106
Cs-137
Cs-134
Sr-90
(Ci/GWe-yr)
5 E+04
5 E-03
5 E-04
(b)
(b)
(b)
(b)
(b)
(Ci/yr)
2.27 E+06
2.27 E-01
2.27 E-02
— —
(Ci/yr)
0
2.94 E-03
5.02 E-07
5.21 E+04
4.80 E+01
1.59 E-02
2.88 E-02
1.12 E-02
(a) The conversion from Ci/GWe-yr to Ci/yr is based on an LWR operating
at 33% thermal efficiency and producing approximately 33 MTHM of
spent fuel at a burnup of 33,000 MWD/MTHM; all of the releases are
assumed to be from a 1500 MTHM per year fuel reprocessing plant.
(b) Not included in UFC standard
-------
23
7.0 REFERENCES
1. "Environmental Radiation Protection Standards for High-Level
Radioactive Waste," Federal Register. Vol. 41, No. 235, Monday,
December 6, 1976, page 53363.
2- Report to the President by the Interagency Review Group on Nuclear
Waste Management, U.S. DOE Report TID-29442, U.S. Department of
Energy, Washington, D.C., March 1979.
3. Technical Support of Standards for High-Level Radioactive Waste
Management, Volume A; Source Term Characterization, Report No.
EPA-520/4-79-007A, Office of Radiation Programs, U.S. Environmental
Protection Agency, Washington, D.C., 1979.
**• Final Environmental Impact Statement, Waste Management Operations,
Idaho National Engineering Laboratory, Idaho, Report ERDA-1536,
U.S. Energy Research and Development Administration, Washington,
D.C., September 1977.
5« Environmental Statement, Waste Management Operations, Savannah
River Plant, Aiken, South Carolina, Report ERDA-1537, U.S. Energy
Research and Development Administration, Washington, D.C.,
September 1977.
6- Final Environmental Statement, Waste Management Operations, Hanford
Reservation, Richland, Washington, Report ERDA-1538, U.S. Energy
Research and Development Administration, Washington, D.C., December
1975.
7. Statement by President J.E. Carter on "Nuclear Power Policy," April
7, 1977, White House, Washington, D.C., (see also Nucleonics Week
Vol. 18, No. 15, April 14, 1977).
8- LWR Spent Fuel Disposition Capabilities 1977-1986, Report
ERDA-77-25, U.S. Energy Research and Development Administration,
Washington, D.C., May 1977.
9' Alternatives For Managing Wastes From Reactors and Post-Fission
Operations In The LWR Fuel Cycle, Volume 2: Alternatives For Waste
Treatment, Report No. ERDA-76-43, Vol. 2 of 5, U.S. Energy Research
and Development Administration, Washington, D.C. May 1976.
10. Alternative Processes for Managing Existing Commercial High-level
Radioactive Wastes, Report NUREG-0043, U.S. Nuclear Regulatory
Commission, Washington, D.C., April 1976.
-------
24
11. "Licensing of Production and Utilization Facilities — Policy
Relating to the Siting of Fuel Reprocessing Plants and Related
Waste Management Facilities, "Code of Federal Regulations, Title
10, Chap. I, Part 50, Appendix F, U.S. Government Printing Office,
Washington, D.C., 1977.
12. Alternatives for Long-Term Management of Defense High-Level
Radioactive Waste, Savannah River Plant, Volume 2, Report
ERDA-77-42, U.S. Energy Research and Development Administration,
Washington, D. C., May 1977.
13. Alternatives for Long-Term Management of Defense High-Level
"Radioactive Waste, Idaho Chemical Processing Plant, Report
ERDA-77-43, U.S. Energy Research and Development Administration,
Washington, D.C., September 1977.
14. Alternatives for Long-Term Management of Defense High-Level
"Radioactive Waste, Hanford Reservation, Report ERDA-77-44. U.S.
Energy Research and Development Administration, Washington, D.C.,
September 1977.
15. J.M. Lukacs, et al., Compatibility of Two Idaho Chemical Processing
Plant Glasses with Electric Melting Processes, U.S. DOE Report
PNL-2751, Battelle Pacific Northwest Laboratory, Richland,
Washington, December 1978.
16. C.C. Chapman et al., Vitrification of Hanford Wastes in a
Joule-Heated Ceramic Melter and Evaluation of Resultant
Canisterized Product, U.S. DOE Report PNL-2904. Battelle Pacific
Northwest Laboratory, Richland, Washington, August 1979.
17. E.J. Wheelwright et al., Technical Summary Nuclear Waste
Vitrification Project, U.S. DOE Report PNL-3038, Battelle Pacific
Northwest Laboratory, Richland, Washington, May 1979.
18. Immobilization of Defense High-Level Waste: An Assessment of
Technological Strategies and Potential Regulatory Goals, U.S. DOE
Report SAND-79-0531 (2 volumes), Sandia Laboratories, Albuquerque,
New Mexico, June 1979.
19. A.H. Kibbey, H.W. Godbee and G.S. Hill, "Estimated Radiological
Doses from the Gaseous Effluents of a Model High-Level Waste
Solidifcation Facility", Back End of the LWR Fuel Cycle Proceedings
of the American Nuclear Society Topical Meeting, Savahhah, Georgia,
March 19-23, 1978, Report No. CONF-780304, American Nuclear
Society, Inc., La Grange Park, Illinois.
20. M.J. Bell, ORIGEN - The ORNL Isotope Generation and Depletion Code,
Report No. ORNL-4628, Oak Ridge National Laboratory, Oak Ridge,
Tennessee, May 1973.
-------
25
21. H W Godbee and A. H. Kibbey, Source Terms for Radioactive
Effluents from a Model High-Level Waste Sol irii
Keport NO. ORNL/NUREG/TM-67. Oak Ridge National Laboratory, Oak
Ridge, Tennessee, November 1976.
22 ' Tec""ical Support of Standards for High-Level Radioactive Waste
Management, Volume B; Engineering Controls. Reportlfo
Engineering Controls. epor
EPA-520/4-79-007B, Office of Radiation Programs, U.S. Environmental
Protection Agency, Washington, D.C., 1979.
23 ' Technology for Commercial Radioactive Waste Management. 5 Volumes
°°E/ET"002 " °f Energy» Washington, D.C., '
24. W.F. Holcomb, A Survey of the Available Methods of Solidification
for Radioactive Wastes, Report Technical Note nRP/TAn_7g-i — 0~S —
Environmental Protection Agency, Washington, D.C., November 1978.
25. L.T. Lakey and B.R. Wheeler, "Solidification of High-Level
Radioactive Wastes at eh Idaho Chemical Processing Plant »
Management of Radioactive Wastes from Reprocessing. Proceedings of
Symposium by ENEA/IAEA, Paris, France, November 27-December 1?
I y i c. .
26. J.A Wielang and W.A. Freeby, The Fifth Processing Campaign In The
Waste Calcining Facility FY-1972. USAEC itennrt. Mn Trp.moi T^nho
National Engineering Laboratory, Idaho Falls, Idaho, June 1973.
27. J.A. Wielang et al . , The Fourth Processing Campaign In The Waste
Calcining Facility FY-1071. USAEC Report MO -rrp^nn,, Trhho -
National Engineering Laboratory, Idaho Falls, Idaho, March 1972.
28. W.F. Holcomb, "Uses of tne Fluidization Bed Process," Combustion
Vol. 48, No. 10, page 31, April 1977. ~ - *
29 • Environmental Analysis of the Uranium Fuel Cycle; Part III -
Nuclear Fuel Reprocessing. Report Nn. F PA _c; ? 0/0,73^03 P — U~S
Environmental Protection Agency, Office of Radiation Programs'
Washington, D.C., October 1973.
30- Environmental Analysis of the Uranium Fuel Cycle: Part IV-
Supplementary Analysis - 1976. Report No.EPA-Rpn/a-7ft_ni7 n s
Environmental Protection Agency, Office of Radiation Programs "
Washington, D.C., July 1976.
31. R.B Hower et al . , Radioactive Airborne Effluent Measurement and
Monitoring Survey of Reporcessing and Waste Treatment Facilities
Report COO-3049-9, Science Applications, Inc., Prepared for hte '
Harvard Air Cleaning Laboratory (USERDA), September 1977
-------
26
32. J.D. Christian and D.W. Rhodes, Ruthenium Containment During
Fluid-Bed Calcination of High-Level Waste From Commerical Nuclear
Fuel Reprocessing Plants, USERDA Report ICP-1091, Idaho National
Engineering Laboratory, Idaho Falls, Idaho, January 1977.
33. R.A. Brown et al., (Ed), Reference Facility Description for the
Recovery of Iodine, Carbon and Krypton From Gaseous Wastes, U.S.
DOE Report ICP-1126, Idaho National Engineering Laboratory, Idaho
Falls, Idaho, April 1978.
31*. B.J. Newby and D. W. Rhodes, Ruthenium Behavior During Calcination,
U.S. DOE Report ICP-1164, Idaho National Engineering Laboratory,
Idaho Falls, Idaho, September 1978.
35. B.J. Newby and V.H. Barnes, Volatile Ruthenium Removal From
Calciner Off-Gas Using Solid Sorbents, U.S. ERDA Report ICP-1078,
Idaho National Engineering Laboratory, Idaho Falls, Idaho, July
1975.
36. W.A. Freeby, Off-Gas Cleanup System Considerations For
Fluidized-Bed Radioactive Waste Calcination at the ICPP, U.S. DOE
Report ICP-1162, Idaho National Engineering Laboratory, Idaho
Falls, Idaho, August 1978.
37. W.F. Bonner et al., Spray Solidification of Nuclear Waste, USERDA
Report No. BNWL-2059, Battelle Pacific Northwest Laboratories,
Richland, Washington, August 1976.
38. W.J. Bjorklund, Development and Use of Sintered Metal Filters with
Fluidized Bed and Spray Calcination of Simulated High-Level Waste,
US ERDA Report BNWL-2074, Battelle Pacific Northwest Laboraatories,
Richland, Washington, July 1976.
39. R. E. Schindler, Removal of Particulate Solids From the Off-Gas of
the WCF and NWCF, U.S. DOE Report ICP-1157, Idaho National
Engineering Laboratory, Idaho Falls, Idaho, June 1978.
40. Environmental Monitoring at Major U.S. Energy Research and
Development Administration Contractor Sites, Report ERDA-77-104/2,
U.S. Energy Research and Development Administration, Washington,
D.C., 1976.
41. R.E. Moore, et al., AIRDOS-EPA: A Computerized Methodology for
Estimating Environmental Concentrations and Dose to Man from
Airborne Releases of Radionuclides, Report EPA-520/1-79-009, U.S.
Environmental Protection Agency, Office of Radiation Programs,
Washington, D.C., December 1979.
42. Final Environmental Impact Statement on the Barnwell Nuclear Fuel
Plant of Allied-Gulf Nuclear Services, Docket No. 50-332, U.S.
Atomic Energy Commission, Washington, D. C., January 1974.
-------
43
47.
48.
49.
50.
51.
52.
53.
27
Memo from Eckerman, K.E., Dayem, N., Emch, R., Radiological
Assessment Branch, Division of Technical Review, Nuclear Regulatorv
Commission, "Code Input Data for Man-Rem Estimates," October 15 "
44.
"f "adionucllde. Into
l-mary Report, Report No. EPA
— - , - - —~...*...wi j n^friy^i u, ncpur u WO. LrA
-79-006, U.S. Environmental Protection Agency, Office of
rinn D»»/-MTV»^»»,« i.r—1_.- j ~ _ . rf * -*-~^ \j±
,..
Programs, Washington, D.C., August 1979.
45.
46.
Environmental Radiation Protection Standards for Nuclear Power,"
mn n Ver Re*ulat1™*' Title 40, Chap. I, SubChapter F, Part
190, U.S. Government Printing Office, Washington, D.C., 1978.
R.B. Keely and W.F. Bonner, "Technology Status of Spray/
Vitrification of High-Level Liquid Waste for Full-Scale
/rHPreSe^e^ at 7°th AnnUal Meeting of the American
10.7 Chemical Engineers, New York, New York, November
1 977 •
Laboratories, Richland, Washington, June 1977.
D W Readey and C.R. Cooley (Eds), Ce£amic_andGlass Radioactive
Waste Forms, Report No. CONF-770102, U.S. Energy Research and
Development Administration, Washington, D.C., January 4-5, 1977.
Bat with Waste Vitrification Systems
Battelle-Northwest," Radioactive Wastes From the Nuclear Fuel
C££le, Symposium Series No. 154, Vol. 72, American Institute of
Chemical Engineers, New York, New York, 1976.
J.L. Buelt and C.C. Chapman, Liquid-Fed Ceramic Melter: A General
terthfft10? h ^T' U'S- ^ Hep°rt PNL"2^5, Battelle Pacific
Northwest Laboratory, Richland, Washington, October 1978.
H.T Blair, Vitrification of Nuclear Waste Calcines by In-Can
T^^9 USERDA Report No. BNWL-2061, Battelle Pacific Northwest
Laboratories, Richland, Washington, May 1976.
W P. Bishop and F.J. Miraglia, Jr., (Ed), Environmental Survey of
r.hP R o nr>»* n a o e< •! v\ r, ^^j t.r i. ^ >. . _ *- J VJL
,, c «-» » -• • i \«-^/, j^tivj.1 i>iJiiieiiua.L ourvev
the Repressing and Waste Management Portions of the LWR Fuel
to WASH-1^8). U.S. Nucle
_ — ~-—f-r,,~^, .11 »^/i j— ic_-ruy, U.O. 1
Regulatory Commission, Washington, D.C., October 1976.
J.D. Christian and D.T. Pence (Scientific Advances, Inc.) Critical
Assessment of Methods for Treating Airborne Effluents from
High-Level Waste Solidification Processes. Report Mo PMT,?^
i, Richland,
-------
28
54. Environmental Aspects of Commerical Radioactive Waste Management, 3
Volumes, Report DOE/ET-0029, U.S. Department of Energy, Washington,
D.C., May 1979.
55. Draft Environmental Impact Statement, Management of Commerically
Generated Radioactive Waste, Report DOE/EIS-0046-D, 2 volumes, U.S.
Department of Energy, Washington, D.C., April 1979.
-------
APPENDIX A
WASTE CALCINATION AND CLASSIFICATION PROCESSES
A-1
-------
APPENDIX A
WASTE CALCINATION AND CLASSIFICATION PROCESSES
A.1 CALCINATION
Calcination, the conversion of high-level liquid wastes to a
calcine powder, is the most likely first step in the solidification
process. This section covers the two most promising and advanced
calcination processes (9, 22-24).
A.1.1 Fluidized-Bed Calcination
Fluidized-bed calcination was the first technique developed for the
conversion of radioactive waste solutions to solids. The Atomic Energy
Commission (AEC) sponsored its development in 1955 and built the Waste
Calcining Facility (WCF) at the Idaho National Engineering Laboratory as
part of the Federal Government's Idaho Chemical Processing Plant (4).
Fluidized-bed calcination solidifies radioactive high-level liquid
wastes by pnuematically atomizing the waste solution into a bed of
fluidized solid granules. In-bed combustion of kerosene with oxygen
generates temperatures of 500 C. The waste solution is sprayed into the
fluidized heated bed; water vapor and volatile gases flash from the
spray droplets, depositing the oxides of metallic salts in the waste on
bed particles. At equilibrium conditions, the effect of particle growth
is balanced by the formation of new seed particles and by removal of the
calcine product. The powdery solids and granules are continuously
removed from the calciner and pneumatically transported to an integrated
on-site storage facility. The off-gas from this process is composed
primarily of the fluidizing air, the transport gas, and the gaseous
A-2
-------
reaction products (9, 22-28). Figure A.I shows the type of
fluidized-bed calciner used at the Idaho National Engineering
Laboratory.
A.1.2 Spray Calcination
This process has been under development at DOE's Hanford
Reservation for over 15 years. The Battelle Pacific Northwest
Laboratories is now testing it with simulated wastes.
The liquid wastes are pneumatically atomized and sprayed into the
top of a cylindrical calciner chamber, the walls of which have been
heated to 700 C. The atomized liquid wastes are sequentially
evaporated, dried, and calcined as they fall and are then discharged
from the lower cone of the chamber (9, 22-24, 37, 46). Figure A.2 shows
the type of spray calciner system used at Hanford.
A-3
-------
TO CYCLONE
FOR FINES
REMOVAL TO
PRODUCT STORAGE
AND
TO OFF-GAS CLEANUP
SYSTEM
CALCINER
VESSEL
i
-Cr
WASTE FEED
NOZZLE
ATOMIZING AIR
OXYGEN
KEROSENE TO
FUEL NOZZLE
FLUIDIZING
AIR
PRODUCT OVERFLOW
TO STORAGE
Figure A.1
FLUIDIZED-BED CALCINER
-------
FURNACE
ATOMIZING AIR
AND
WASTE FEED NOZZLE
CALCINER
CHAMBER
OFF-GAS
OUT
VIBRATOR
SINTERED STAINLESS
STEEL FILTERS
CALCINE POWDER OUT
Figure A. 2
SPRAY CALCINER
A-5
-------
A. 2 WASTE CLASSIFICATION
Classification is a solidification process that incorporates
high-level wastes in a solid matrix. The wastes and glass frit are
combined, melted, and canned; the melt cools and solidifies. Over the
past 20 years, many countries have developed various glassification
processes. The two most promising candidates for commercial use in the
United States are in-can melting and continuous melting (9, 22-24,
47-51).
The calcination and glassification processes can be coupled.
Glassification is a batch process, and calcination is a continuous
process; however, with diverter valve and multiple melting furnace
canisters, the coupled systems become semicontinuous. The Battelle
Pacific Northwest Laboratories is developing a tandem unit that combines
spray calcination and in-can melting. France and West Germany have
coupled the continuous melting system with the spray calciner and with
another calcination process called rotary-kiln.
A-6
-------
A.2.1 In-Can Melting
The Battelle Pacific Northwest Laboratories is developing the
in-can melting batch process for the Department of Energy. In this
process, the calcine powder and specially formulated glass frit fall
directly into a close-coupled melter canister. The frit and the calcine
are melted together in a metal canister in a multizone furnace at
processing temperatures of 1000-1100 C. In-can melting offers several
advantages: (a) simplicity in process steps and equipment; (b)
non-transfer of melt; (c) complete fixation in glass of everything
entering the melter except some volatile species; (d) disposability of
the melter canister; and (e) sufficient flexibility to accommodate
calcine products from a wide range of processes, such as spray or
fluidized-bed calcination (9, 22-24, 48, 51). Figure A.3 illustrates
the in-can melting process.
A.2.2 Continuous Melting
The Batelle Pacific Northwest Laboratories is also developing a
continuous (or joule-heated) melter process that is similar to
commercial electric-glass melter processes. It can be coupled with
different kinds of waste calciners and can even receive liquid wastes
directly (4, 9, 13). The process is carried out at temperatures ranging
from 1000 to 1200 C in a refractory-lined melter with internal
electrodes; the molten glass acts as its own electric-resistance heating
element. Flexibility in glass composition and controlled draining of
the glass-waste mixture from the melter permit changes in the final
waste form package (9, 22, 24, 47, 48, 50). Figure A.4 shows the
continuous melter.
A-7
-------
CALCINE
GLASS FRIT ADDITION
DIVERTER VALVE
r r
-i "
•J J
• -*
II II
Lr'
fl
rl
C^
" *
STORAGE
CANISTER
MOLTEN WASTE/GLASS MIXTURE
IN-CAN
1*1 k 1m 1 k rl
J FURNACE
Figure A.3
IN-CAN MELTER
A-8
-------
i
vo
CALCINE OR LIQUID WASTE
AND
GLASS FRIT
OFF-GAS
MOLTEN GLASS
AND
WASTE MIXTURE
M«-J»S»»
MOLTEN GLASS
TO
STORAGE
CANISTER
ELECTRODES
Figure A.4
CONTINUOUS MELTER
-------
-------
APPENDIX B
DOE AND NRC GENERIC SOLIDIFICATION
PLANT STUDIES
B-1
-------
APPENDIX B
DOE AND NRC GENERIC SOLIDIFICATION PLANT STUDIES
Both the Nuclear Regulatory Commission (NRC) and the Department of
Energy (DOE) have contracted studies involving the gaseous discharge
from high-level waste solidification processes (19, 21, 23, 52, 53, 54).
B.1 NRC Contract Studies
Oak Ridge National Laboratory (ORNL) personnel conducted studies
for NRC. They evaluated the gaseous effluents released from a generic
high-level waste solidification facility similar to the New Waste
Calcining Facility at the Idaho National Engineering Laboratory; this
generic facility also glassified the calcine. Estimated decontamination
factors are one for tritium, 100 for iodine, and 5 E+08 for ruthenium.
The DFs for other radionuclides range from 5 E+09 to 1 E+10.
Table B.1 shows the ORNL list of likely radionuclide source terms
and expected decontamination factors during the calcination and
glassification processes. The generic facility is based on HLLW from a
5 MTU/day spent fuel reprocessing plant that processes 213- day-decayed
material irradiated to 29,000 MWD/MTU.
B-2
-------
TABLE B.J
DECONTAMINATION FACTORS EXPECTED DURING THE CALCINATION
AND CLASSIFICATION OF HLLW (19) 1WA11UN
Decontamination Factor
Radionuclides
Tritium
Iodine
Particulates
Ruthenium
Cestum
1.0 E+00
1.0 E+02
1.0 E+10
5.0 E+08
1.0 E+10
(b)
(b)
1.0 E+12
3.8 E+11
7.7 E+09
1.0 E+00
1.0 E+02
9.9 E+09
5.0 E+08
4.4 E+09
/-
C.
• — "o -«--I-VIMJLVJ waouc d
the calcined solids with a glass frit, and
the mixture by heating to about 1000°C.
waste at about
(b) All tritium and iodine are volatized in the calcination step.
TABLE B.2
MAXIMUM ANNUAL DOSE EQUIVALENTS (a) TO AN INDIVIDUAL (b)
DUE TO GASEOUS RELEASES FROM A GENERIC HLLW
SOLIDIFICATION FACILITY (c) (19)
Adult
Organ
Dose
(millirem)
Major Nuclides Causing Dose"
(% of Total Dose)
H-3 Sr-90 Ru-106 1-129 Cs-134 & -137
Total Body
G.I. Tract
Bone
Thyroid
Lungs
Liver
Kidney
Testes
Ovaries
3-9 E-01
1.3 E+00
5.0 E-01
8.0 E-01
3.8 E-01
3.8 E-01
3.9 E-01
4.0 E-01
3.0 E-01
58
18
45
29
60
60
58
56
75
10
75
10
13
52
30
28
27
30
26
32
18
Fifty-year dose commitment for 1-year exposure.
Ingestion is the principal exposure pathway.
and
B-3
-------
TABLE B.3
MAXIMUM ANNUAL DOSE EQUIVALENTS TO AN INDIVIDUAL AT THE SITE
BOUNDARY DUE TO GASEOUS RELEASES FROM HLLW SOLIDIFICATION
PLANT (52)
Adult Organ
G.I. Tract
Bone
Thyroid
Lung
Total Body
Dose (millirem)
1.7
0.7
1.1
0.5
0.52
After evaluating the source terms and the control technology
available and necessary for dose reduction, ORNL personnel prepared a
table of the maximum annual total-body dose and organ doses to an
individual due to gaseous effluent releases from their generic facility
(see Table B.2).
Battelle Pacific Northwest Laboratories personnel prepared Table
B.3 for the NRC in a review of environmental impacts from the release of
radionuclides during the operation of a generic HLLW solidification
plant; the plant processes high-level liquid waste from a 2000 MT/year
spent fuel reprocessing plant that processes 160-day-decayed material
irradiated to 33,000 MWd/MT (52).
B-4
-------
B.2 DOE Contract Studies
The DOE contract studies were based primarily on the assessment of
control technology for treating airborne effluents from the
solidification processes. The assessment included processes developed
throughout the world. (See Table B.4) Additional assessments included
decontamination factors for off-gas cleanup and some estimated annual
doses from gaseous effluents. Table B.5 shows the dose assessment.
TABLE B.4
SUMMARY OF ESTIMATED DECONTAMINATION FACTORS FOR
SOLIDIFICATION PROCESSES (53)
Process
Particulates
Feed-to-Atmospheric Release DF
" Volatilized Ru
USA ICPP Fluid-Bed^
USA PNL Fluid-Bed*" '
Eurochemic LOTES
USA PNL Spray
German VERA
PNL Pot
British FINGAL
British HARVEST
French PIVER^eJ
Italian Pot
Phosphate Glass )
Borosilicate
USA PNL Phosphate Glass
French Rotary. Kiln
German FIPS }
German PAMELA (e)
USA PNL Proposed
2 E+10
1 E+10
6 E+08
1 E+12,
1 E+12(d)
1 E+12
1 E+15
1 E+13
1 E+13 to E+14
1 E+14 to E+15
1 E+14 to E+15
1 E+12
1 E+10 to E+11
1 E+11 to E+12
1 E+13 to E+14
2 E+11
1 E+06
1 E+07
1 E+10
1 E+13
1 E+10
1 A- • T 1 W
.1 E+13
P * " ~ 1 J
1 E+08
1 E+10
1 E+10
1 A-i T | W
1 E+09
1 E+10
« *-l T 1 \J
1 E+06
1 E+05
1 E+09
(c)
to E+09
to E+11
to E+07
1 E+12 to E+14
1 E+10
(a)
(b)
(c)
(d)
(e)
(f)
With a second HEPA filter.
Including final HEPA filter.
Data are for total Ru, but since total Ru DF is 0.01 times total
Cerium DF, one may assume the majority of released Ru is in
volatized form.
** finfKHEPA filter is included in ventilation system, particulate
DF will be increase by a factor of approximately E+02.
Waste evaporator (concentrator) included in integrated system.
USA PNL spray calciner with in-pot melter assumed. DF for iodine
equals 1 E+03.
B-5
-------
TABLE B.5
MAXIMUM ANNUAL DOSE EQUIVALENTS TO AN INDIVIDUAL DUE TO GASEOUS
RELEASES FROM CALCINATION AND CLASSIFICATION FACILITIES (a)(23, 54, 55)
Adult Organ
Total Body
Thyroid
Lung
Bone
Calcinaton
2.8 E-01
2.4 E-01
2.4 E-01
8.7 E-06
DOSE (millirem)
Classification
2.4 E-01
2.5 E-01
2.4 E-01
1.4 E-05
(a) The doses are based on a uranium-piutonium recycle scenario which
used the additive pathways of air submersion, inhalation, and
ingestion.
B-6
-------
APPENDIX C
SELECTED TABLES
FROM AIRDOS-EPA
COMPUTER PROGRAM
C-1
••*
¥
-------
APPENDIX C
SELECTED TABLES FROM AIRDOS-EPA COMPUTER PROGRAM
The following tables contain selected data developed in the
AIRDOS-EPA computer calculations of annual dose equivalents to
individuals and populations (41, 44).
Tables C.1 and C.2 presents the input data for the rural and urban
sites. Tables C.3 through C.10 lists the input data for each
radionuclide studied.
Tables C.11 through C.22 list the total dose equivalents to
individuals and populations near rural and urban sites for one year,
five years, and ten years. The tables also list the radionuclide
contributions to the organ doses by percentage.
C-2
-------
Table C.I
LIST OF INPUT VALUES FOR R AOIONU CLIDE-iNDE;3 ENDENT VARIABLES
SOMBER OF N'JCLIDES CONSIDERED
TIME DELAY--IKGESTION OF PASTURE GRASS BY ANIMALS (HE)
TIME DELAY--INGESTION OF STORED FEEE BY Atlif.ALS (HIi)
TIME DELAY — INGESTION OF LEAFY VEGETABLES BY BAN (HP)
TIME DELAY--INGESTION OF PRODUCE BY MAN (I1R)
REMOVAL RATE CONSTANP FOR PHYSICAL LOSS BY WEATHERING (PER HOUR)
PERIOD OF EXPOSURE DJRING GROWING SEASON — PASTURE GRASS (HR)
PERIOD OF EXPOSURE DURING GROWING SEASON—CROPS OB LEAFY VEGETABLES (HR)
AGRICULTURAL PRODUCTIVITY BY UNIT AREA (GBASS-COW-MILK-HAN PATHWAY (KG/SQ. METER)}
AGRICULTURAL PRODUCTIVITY BY UNIT AREA (PRODUCE OR LEAFY VEG INGESTED BY MAN (KG/SQ. METER))
FRACTION OF YEAR ANIMALS GRAZE ON PASTURE
FRACTION OF DAILY FEED THAT IS PASTURE GRASS WHEN AKIMAL GRAZES ON PASTURE
CONSUMPTION RATE OF CONTAMINATED FEED OE FORAGE BY AN ANIMAL IN KG/DAY (HET WEIGHT)
TRANSPORT TIME FROM ANIMAL FEED-KILK-MAN (DAY)
SATE OF INGESTION DF PRODUCE BY MAN (KG/YR)
RATE OF INGESTION OF MILK BI MAN (LITERS/YB)
RATE OF INGESTION DF MEAT BY MAN (KG/YH)
RATE OF INGESTION OF LEAFY VEGETABLES BY EAN (KG/YR)
AVERAGE TIME FROM SLAUGHTER OF MEAT ANIRAL TO CONSUMPTION (DAY)
FRACTION OP PRODUCE INGESTED GBCWN IN GARDEN OF INTEREST
FRACTION OP LEAFY VEGETABLES GROWN IN GARDEN OF INTEREST
PERIOD OF LONG-TERM BUILDUP FOR ACTIVITY IN SOIL (YEARS)
EFFECTIVE SURFACE DENSITY OF SOIL (KG/SQ. M, DRY WEIGHT) (ASSUMES 15 CM PLOW LAYER)
VEGETABLE INGSSTIOS RATIO-IMMEDIATE SURROUNDING AREA/TOTAL WITHIN AREA
MEAT INGESTION RAPID-IMMEDIATE SURROUNDING AREA/TOTAL BITUIN AREA
MILK INGESTION EATIO-IMMEDIATE SUBRCUNDIKG AREA/TOTAL WITHIN ASEA
MINIMUS FRACTIONS OF FOOD TYPES FROM OUTSIDE AREA LISTED BELOW ARE ACTUAL FIXED VALUES
MINIMUM FRACTION VEGETABLES INGESTED FROM OUTSIDE AREA
MINIMUM FRACTION MEAT INGESTED FROM OUTSIDE AREA
0.0
0.2160E+OU
0.3360E+03
0.3360E+03
0.2100E-02
0.7200E+03
0. 1U40E + 01*
0.2800E+00
0.7160E+00
0.4000E+00
O.U300E+00
0.15602+02
O.UOOOE+01
0. 1760E+03
0. 1120E+03
0.9400E+C2
0.1800E+02
0.2000E+02
0.1000E+01
0.1000E+01
0.1000E+03
0.2150E+03
0.1000E+01
0.5000E+00 *
0.1000E+01
0.5000E+00
0.1000E+01
0.5000E+00
0.0
0.2000E+00
0.0
0.20002+00
-------
o
Table C.I continued
MINIMUM FRACTION MILS INGESTED EROM OUTSIDE AREA
0.0
INHALATION RATS OF MAN (CUBIC COTIMETEBS/HR) 0.2000E*00
' 0.9167E+06
BUILDUP TIME FOE EADIONUCLIDES DEPOSITED ON GROUND AND HATEE (DAYS) 0 3650£+05
DILUTION FACTOR FOR WATER FCR SWIMMING (CM)
0.1524E+03
FRACTION OF TIME SPENT SWIMMING
0. 1000E-01
MUSCLE MASS OF AMIHAL AT SLAUGHTER (KG)
0.2COOE+03
FRACTION OF ANIMAL HEED SLAUGliTEEED PER DAY
0.3810E-02
MILK PRODUCTION OF C3W (LITERS/CAY)
0. 1100E+02
FALLOUT INT2RCEPT1DN FRACTION-VEGETABLES
0.2000E+00
FALLODT INTERCEPTION FfiACTICN-EASTURE
0.5700E+00
PSiCIIOH OF BADI01CIIVIIY EETAINED OH LEWI VEGETABLES ASD PKODUCE AFTEB WASHING 0..1000E*01
dose
-------
Table C.2
COMPUTED VALUES FOR THE AREA (Rural)
TOTAL POPULATION
TOTAL NUMBER OF MEAT ANIMALS
TOTAL NUMBER OF MILK CATTLE
TOTAL AREA OF VEGETABLE FOOD CROPS (SQUARE METERS)
TOTAL MEAT CONSUMPTION (KG PER YEAR)
TOTAL MEAT PRODUCTION (KG PER YEAR)
TOTAL MILK CONSUMPTION (LITERS/YEAR)
TOTAL MILK PRODUCTION (LITERS/YEAR)
TOTAL VEGETABLE FOOD CONSUMPTION (KG PER YEAR)
TOTAL VEGETABLE FOOD PRODUCED (KG PER YEAR)
477127.0
202801
0.3740E*08
0.4485E«08
0.5641E«08
0.5344E*08
0.5?83E*08
0.9256E*08
0.2678E*08
O
I
COMPUTED VALUES FOB THE AREA (Urban)
TOTAL POPULATION
TOTAL NUMBER OF HEAT ANIMALS
TOTAL NUMBER OF MILK CATTLE
TOTAL AREA OP VEGETABLE FOOD CRCPS (SQUARE METERS)
TOTAL MEAT CONSUMPTION (KG PER YEAR)
TOTAL MEAT PRODUCTION (KG PER YEAR)
TOTAL HILK CONSUMPTION (LITEHS/YEAE)
TOTAL MILK PRODUCTION (LITERS/IEAfi)
TOTAL VEGETABLE FOOD CONSUMPTION (KG PER YEAR)
TOTAL VEGETABLE FOOD PRODUCED (KG PER YEAR)
2486049.0
689632
38000
0.1638E+09
0.2337E+09
0.19182*09
0.2784E+09
0.1526E+09
0.4823E+09
0.1173E+09
-------
Table C.3
LIST Of INPUT DATA FOR NUCL1DE H-3
RADIOACTIVE DECAY CONSTANT (PER DAY)
ENVIRONMENTAL DECAY CONSTANT—SURFACE (PER DAY)
ENVIRONMENTAL DECAY CONSTANT—WATER (PER DAY)
DOSE CONVERSION FACTOR FOR FOOD INGEST10N (PEM-CC/PC1-YEAR )
DOSE CONVERSION FACTOR FOR WATER INGESTION CREM-CC/PCI-YEAR)
ORGAN
TOT.PODY
S WALL
LLI HALL
LUNGS
KIDNEYS
LIVER
OVARIES
R MAR
ENDOST
TESTES
THYROID
INHALATION INGESTION
(REMS/MICROCURIEMREMS/MICROCURIE)
0.125E-03
0.125E-03
0.133E-03
0.125E-03
0.129E-03
0.124E-03
0.124E-03
0.12
-------
n
i
ORGAN
Table C.4
LIST OF INPUT DATA FOR NUCLIOE SR-90
RADIOACTIVE DFCAY CONSTANT (PER CAY)
ENVIRONMENTAL DECAY CONSTANT—SURFACE (PER PAY)
ENVIRONMENTAL DECAY CONSTANT—WATER (PER DAY)
AVERAGE FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH L OF MILK (DAYS/L)
FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH KG CF FIESH (DAYS/KG)
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FROM SOIL FOR PASTURE ANC FORAGE
UN PCI/KG DRY HEIGHT PER PCI/KG DRY SOIL)
CONCENTRATION FACTOR FOR UPTAKE CF NUCLIDE FROM SOIL BY EDIBLE PARIS OF CROPS
(IN PCI/KG WET WEIGHT PER PCI/KG DRY SOIL)
GI UPTAKE FRACTION (INHALATION)
GI UPTAKE FRACTION (INGESTION)
PARTICLE SIZE (MICRONS)
SOLUBILITY CLASS
INHALATION INGESTION
(REMS/MICROCURIEMREMS/MICRDCURIE)
DOSE CONVERSION FACTORS
SUBMERSION IN AIR
(REMS-CUBIC CM/
HICROCURIE-HR)
SURFACE EXPOSURE
(REMS-SCUARE CM/
MICROCURIE-HR)
C.O
0.0
0.2400E-C2
0.3000E-C3
0.12COE+C1
0.2900E+GC
0.2000E+CO
O.^OOOE+CO
0.1000E+01
SUBMERSION IN HATER
(REKS-CUEIC CM/
MICROCURIE-HR)
TOT.BODY
S WALL
LLI WALL
LUNGS
KIDNEYS
LIVER
OVARIES
R MAR
ENDOST
TESTES
THYROID
0.241E+00
0.197E-03
0.141E-01
0.969E-02
0.146E-01
0.146E-01
0.146E-01
0.110E+01
0.220E+01
0.146E-01
0.146E-01
0.945E-01
0.876E-03
0.778E-0!
0.594E-08
0.599E-02
0.571E-02
0.599E-02
0.43C6+00
0.859E+00
0.599E-02
0.599E-02
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
C.O
0.0
o.c
0.0
0.0
0.0
0.0
o.c
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
-------
o
I
00
Table C.5
LIST OF INPUT DATA FOR NUCLIDF RU-106
RADIOACTIVE DECAY CONSTANT (PER DAY)
ENVIRONMENTAL DECAY CONSTANT — SURFACE (PER DAY)
ENVIRONMENTAL DECAY CONSTANT— WATER (PER DAY)
AVERAGE FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH L OF MILK (DAYS/L)
FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH KG OF FLESH (DAYS/KG)
FQR
py
ORGAN
TOT.BODY
S WALL
LLI KALL
LUNGS
KIDNEYS
LIVER
OVAR IES
R MAR
ENDOST
TESTES
THYROID
GI UPTAKE FRACTION (INHALATION)
GI UPTAKE FRACTION (INGESTION)
PARTICLE SIZE (MICRONS)
SOLUBILITY CLASS
INHALATION INGESTION
(REMS/MICROCUR1EMREMS/MICROCURIE)
0.618E-01
C.696E-02
0.137E+00
0.385E+01
0.895E-02
O.I15E-01
0.767E-02
0.937E-02
0.100E-01
0.697E-02
0.919E-02
0.59<»E-02
0.6
-------
o
I
ORGAN
TOT.BODY
S WALL
LLI WALL
LUNGS
KIDNEYS
LIVER
OVARIES
R KAR
ENDOST
TESTES
THYROID
Table C.6
LIST OF INPUT DATA FOR NUCLIDE 1-129
RADIOACTIVE DECAY CONSTANT (PER DAY)
ENVIRONMENTAL DECAY CONSTANT — SURFACE (PEP DAY)
ENVIRONMENTAL DECAY CONSTANT—WATER (PER DAY)
AVERAGE FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH L DF MILK (DAYS/L)
FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH KG OF FLESH (DAYS/KG)
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FRCP SOIL FOR PASTURE AND FORAGE
(IN PCI/KG DRY WEIGHT PER PCI/KG DRY SDH)
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FROM SOIL BY EDIBLE PARTS OF CROP1:
(IN PCI/KG WET WEIGHT PER PCI/KG DRY SOU)
GI UPTAKE FRACTION (INHALATION)
GI UPTAKE FRACTION (INGESTION)
PARTICLE SIZE (MICRONS)
SOLUBILITY CLASS
INHALATION INGESTION
(REMS/MICROCURIEHREMS/MICRCCURIE)
0.2C5E-02
0.461E-04
0.428E-04
0.788E-03
0.449E-03
0.466E-03
0.378E-03
0.605E-03
0.564E-03
0.357E-03
0.497E+01
0.318E-02
0.784E-04
0.670E-04
0.179E-03
0.702E-C3
0.724E-03
0.592E-C3
0.942E-03
0.879E-03
0.558E-03
0.778E+01
DOSE CONVERSION FACTORS
SUBMERSION IN AIR
(REMS-CUPIC CM/
MICROCUR1E-HR)
0.554E+01
0.234E-*01
0.827E+00
0.288E+01
0.3I5E+01
0.2295*01
C.217E+01
0.788E+01
0.1C9E-»02
0.724E+01
SURFACE EXPOSURE
(REMS-SCUAPE CM/
P1CROCUR1E-HR)
0.3C1E-C2
0.127E-C2
0.449E-C3
0. 156E-02
0. 171E-02
0.124E-02
0.11FE-C2
O.A28E-C2
0.590E-02
0.393E-G2
0. 312E-02
0.12G9E-GS
C.G
0.0
0.99CCE-02
0.7GCOE-02
0.2000E+OC
0.55COE-01
G.95COE-»C'C
0.1000E+03
SUBMERSION IN WATER
(REMS-CUBIC CM/
MICROCURIE-HRJ
0.13 JE-01
0.554E-02
0.196E-02
0.682F-02
0.7<»7E-02
C.514E-02
0.187E-01
0.25PE-01
0.172E-01
0. 1366-01
-------
o
M
O
Table C.7
LIST OF INPUT DATA FOR NUCLIDE CS-134
RADIOACTIVE DECAY CONSTANT (PER DAY)
ENVIRONMENTAL DECAY CONSTANT—SURFACE (PER DAY)
ENVIRONMENTAL DECAY CONSTANT—WATER (PER DAY)
AVERAGE FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH L OF MILK (DAYS/L)
FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH KG OF FLESH (DAYS/KG)
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FROM SOIL FOR PASTURE AND FORAGE
(IN PCI/KG DRY WEIGHT PER PCI/KG DRY SOIL)
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FROM SOIL BY EDIBLE PARTS OF CROPS
(IN PCI/KG WET WEIGHT PER PCI/KG DRY SOIL)
GI UPTAKE FRACTION (INHALATION)
GI UPTAKE FRACTION (INGESTION)
PARTICLE SIZE (MICRONS)
SOLUBILITY CLASS
ORGAN
TOT.BODY
S WALL
LLI WALL
LUNGS
KIDNEYS
LIVER
OVARIES
R MAR
ENDOST
TESTES
THYROID
INHALATION INGESTION
(REMS/MICROCURIEHREMS/HICROCURIE)
0.455E-01
0.326E-01
0.371E-01
0.338E-01
0.677E-01
0.699E-01
0.645E-01
0.616E-01
0.589E-01
0.513E-01
0.519E-01
0.684E-01
C.499E-01
0.575E-01
0.ft68E-01
0.102E+00
0.105E+00
0.974E-01
0.926E-01
G.e86E-01
0.773E-01
0.781E-01
DOSE CONVERSION FACTORS
SUFMERSION IN AIR
(REMS-CUBIC CM/
MICROCURIE-HR)
0.96EE-»03
0.893E+03
C.670E+03
0.909E+03
0.870E+03
0.827E+03
0.466E+03
0.105E + CH
0.119E+04
0.980E*03
0.765E+03
SLRFACE EXPOSURE
(REMS-SCUARE CK/
MCROCUR1E-HR)
0. 19?F-»00
0.177E+CO
0.133E+CO
O.IPOE+OO
0.17?F+00
0. 16AE+CO
0.923E-01
0.206E+CO
0.235-E+CO
0.)9^E+00
0.151E+00
0,920fE-C3
0.0
0.0
0.5600E-C2
O.l^OOE-01
0.1400E+CO
0.91COE-C2
0.95COE*CO
0^.95 OOE*CO
0.1000E+01
SUBMERSION IN WATER
(REMS-CUBIC CM/
KICRDCURIE-HR)
0
c
0
0.196E-»01
0.187E+01
0.178E+01
0.100E+01
0.227E+01
0.255E+01
0.211F+01
192E+01
-------
Table C.8
LIST OF INPUT DATA FDR NUCLIDE CS-137
RADIOACTIVE DECAY CONSTANT (PER DAY) 0.6293E-C4
ENVIRONMENTAL DECAY CONSTANT—SURFACE (PEP CAY) 0.0
ENVIRONMENTAL DECAY CONSTANT—WATER (PER DAY) G.O
AVERAGE FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH L OF MILK (DAYS/L) 0.56COE-02
FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH KG OF FLESH (DAYS/KG) 0.14COE-01
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FROM SOIL FOR PASTURE AND FORAGE G.MCOE + CC
(IN PCI/KG DRY WEIGHT PER PCI/KG DRY SOIL)
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FROM SOIL BY EDIBLE PARTS OF CROPS 0.91COE-02
(IN PCI/KG WET WEIGHT PER PCI/KG DRY SOIL)
GI UPTAKE FRACTION (INHALATION) 0.95COE«CO
GI UPTAKE FRACTION (INGESTION) 0%95CCE-»CO
PARTICLE SIZE (MICRONS) 0.1000E*01
SOLUBILITY CLASS D
DOSE CONVERSION FACTORS
o UKbAN
1
l_i
H
TOT. BODY
S WALL
LLI WALL
LUNGS
KIDNEYS
LIVER
OVAR IES
R MAR
ENDOST
TESTES
THYROID
INHALATION INGESTION
(REMS/MICROCURIEHREMS/MICRCCURIE)
0.326E-01
0.139E-01
0.160E-01
0.162E-01
0.513E-01
0.523E-01
0.500E-01
0.491E-01
0.531E-01
0.444E-01
0.447E-01
0.491E-01
0.218E-01
0.259E-01
0.199E-01
0.773E-01
C.767E-01
0.75.4E-C1
0.738E-01
0.799E-01
C.668E-01
0.672E-01
SUBMERSION IN AIR
(REMS-CUBIC CM/
MICROCURIE-HR)
0.37CE-»03
0.345E+C3
0.257E+03
0.3^»7E + 03
0.332E-»03
0.316E+03
0.166E-*03
0.408E+03
0 .
-------
ORGAN
Table C.9
LIST OF INPUT TATA FOR NUCLIDE PU-239
RADIOACTIVE DECAY CONSTANT (PER DAY)
ENVIRONMENTAL DECAY CONSTANT — SURFACE (PER HAY)
ENVIRONMENTAL DECAY CONSTANT—WATER (PER DAY)
AVERAGE FRACTION OF ANIMAL»S DAILY INTAKE OF NUCLIDE WHICH APPEARS IN EACH L OF MILK (DAYS/L)
FRACTION OF ANIMAL'S DAILY INTAKE OF NUCLIDE WHICH APPEARS IK' EACH KG CF FLESH (DAYS/KG)
CONCENTRATION FACTOR FOR UPTAKE CF NUCLIDE FFOM SOIL FOR PASTURE AND FORAGE
(IN PCI/KG DRY WEIGHT PER PCI/KG DRY SOIL)
CONCENTRATION FACTOR FOR UPTAKE OF NUCLIDE FROM SOIL BY EDIBLE PARTS OF CROP*
(IN PCI/KG WET WEIGHT PER PCI/KG DRY SOIL)
GI UPTAKE FRACTION (INHALATION)
GI UPTAKE FRACTION UNGESTION)
PARTICLE SIZE (MICRONS)
SOLUBILITY CLASS
0.0
0.0
INHALATION INGESTION
(REMS/MICROCURIEMREMS/MICROCURIE)
DOSE CONVERSION FACTORS
SUBMERSION IN AIR
(REMS-CUEIC CM/
n
i
M
TOT. BODY
S WALL
LLI WALL
LUNGS
KIDNEYS
LIVER
OVARIES
R MAR
ENDOST
TESTES
THYROID
0.169E*03
0.272E-02
0.115E+00
0.58GE+03
0.103E+03
0 . 79 7E +0 3
0.363E+01
0.599E+03
0.416E+04
0.114E+02
0.585E+01
0.951E-01
0.4^i2E-02
0.196E+00
0.935E-C7
0.633E-G1
0.<(91E + 00
0.225E-02
0.372E+00
0.258E*01
0.707E-02
0.363E-02
MICROCURIE-
C.486E-01
C.238E-01
0.2C9E-01
0.317E-01
0.255E-01
0.267E-01
C.161E-01
0.611E-01
0.730E-01
0.^^2E-01
0.422E-01
SURFACE FXPOSUPE
(REKS-SCIARF CM/
FICROCURIE-HR)
0.9«S9E-04
0.527E-04
0.31PE-C<»
0.1P1E-C3
0.41COE-C6
0.22COE-C2
0.2CCOE-G3
0.3000E-C«4
0.3000E-OA
>
O.IOOOE+C1
SUBMERSION IN WATER
(REMS-CLiPIC CM/
KICROCURIE-HR)
O.I1PE-03
0.482E-04
0.729E-0<>
0.371E-OA
0.141E-03
0.16PE-C3
0.1C2E-03
0.972E-OA
-------
Table C.10 Maximum Individual Dose to a Rural Individual
from Qne-Year-Decayed Spent Fuel
TOTAL DCJSE TO EACH ORGAN THROUGH ALL PATHWAYS
ORGAN
TOT.BODY
R MAR
LUNGS
ENDOST
S MALL
LLI WALL
THYROID
LIVER
KIDNEYS
TESTES
OVARIES
DUSE(REMS)
O.P106E-02
G.2436E-02
0.3511E-02
0.2698E-02
0.2057E-02
0.2368E-01
0.2237E-02
0.2127E-02
0.2145E-02
0.2313E-02
0.1760E-02
n
CONTRIBUTORS TO ORGAN CCSES
NUCLIOE
H-3
PU-239
1-129
RU-106
CS-137
CS-134
SR-90
TOT. BODY
24.6135
O.QOOO
0.0257
72.8232
1.3556
0.7348
0.4*71
R MAR
21.2492
0.0001
0.0283
74.8595
1.3433
0.7611
1.7584
LUNGS
14.762C
0.0001
0.0071
84.1493
0.7143
0.3672
O.COOO
ENDOST
18.6414
0.0009
0.0349
76.0782
1.3525
0.7199
3.1722
S WALL
25.1907
c.oooo
0.0097
72.9445
1.2172
0.6337
C.0042
LLI WALL
2.2083
O.OOOC
0.0003
97.628C
O.OS15
C.0493
0.0326
PERCENT
THYROID
23.1389
O.OOCO
7.91C4
67.1853
1.GSC6
0.6477
0.0267
LIVER
24.3355
O.CC02
C.C099
73.541 1
1.2590
0.8275
C.C267
KIDNE YS
24.2676
o.coco
0.0132
73.5695
1.2956
0.8262
0.0278
TESTES
22.4C56
O.OOCO
0.0271
75.5113
1.32C4
0.7099
O.C258
OVARIES
29.4C69
0.0000
0.0112
68.8571
0.9231
0.7678
0.0339
-------
Table C.ll Annual Dose to the Rural Population
from One-Year-Decayed Spent Fuel
TOTAL DOSE TO EACH ORGAN THROUGH ALL PATHWAYS
ORGAN
TOT.BODY
R hAR
LUNGS
ENOOST
S WALL
LLI HALL
THYROID
LIVER
KIDNEYS
TESTES
OVARIES
DCSt (MAN-REPS)
0.2687E+02
0.3164E+02
0.5940E*02
0.3413E+02
0.2774E+02
0.1442E+02
0.2765E*02
C.2707E+02
0.2741E+02
0.3025E*02
0.2043E+02
NUCLIDE TOT.BODY R
MAR
LUNGS
CONTRIBUTORS TO ORGAN DCS.5
PERCENT
ENDOST
S KALL
H-3
PU-239
1-129
RU-106
CS-137
CS-134
SR-90
29.8487
0.0001
0.0236
67.6867
1.5377
0.7095
0.1887
27.183*
0.0002
0.0351
69.6624
1.5S32
0.7522
0.7835
14.5081
O.CCOl
O.COda
84.5110
0.6745
0.2994
0.0000
24.0343
0.0012
C.0446
72.0616
1.63g7
0.7476
1.4511
31 .0680
o.oooc
0.011 7
66.6374
1.4393
0.6418
0.0018
"• *- * " ** L. L.
6.063C
0.0000
0.00 OF
93.-39C6
0-2105
0.1043
0.0308
i n i KU i L
31 .1068
o.occo
5.2765
61 .6153
1.33C7
0.6582
0.0125
LIVER
31 .7777
C-OC03
C.0122
65.S234
1.4838
C.7905
C.0122
KIDNEYS
31 .6650
o.ooco
0.0164
65.9626
1.5278
0.7955
0.0126
TESTES
28.4828
o.ccco
C.C335
69.2162
1.5485
0.7076
0.0114
OVARIES
42.0957
o.ocoo
0.0153
55.9950
1.1306
0.7*65
0.0169
-------
Table C.12. Maximum Annual Dose to an Urban Individual
from One-Year-Decayed Spent Fuel
TOTAL DOSE TO EACH ORGAN THROUGH ALL PATHWAYS
ORGAN
TOT.BODY
R MAR
LUNGS
ENOOST
S WALL
LLI WALL
THYROID
LIVER
KIDNEYS
TESTES
OVARIES
DOSE(REMS)
0.1447E-01
0.1734E-01
0.2134E-01
0.1956E-01
G.1422E-C1
0. 1919E«OC
0.1570E-01
0.1479E-01
C.1492E-01
0.1633E-01
0.1177E-01
O
H
Ul
CONTRIBUTORS TO ORGAN DCSES
NtiCLIDE
H-3
pU-239
1-129
RU-1C6
CS-137
CS-134
SR-90
fOT.BODY
10.8021
0,0000
0.0308
86.1276
1.62*5
0.6801
0.5348
R MAR
9.CC60
0.0001
0.0328
86.4952
1.5545
0.8803
2.0322
LUNGS
14.8768
o.ccoi
O.C067
84.0918
0.6766
0.3479
O.COOO
ENDGST
7.7523
0.0005
0.0397
86.2594
1.5358
0.3170
3.595*
S WALL
10.9919
0.0000
C.0116
66.7865
1.4503
0.7547
0.0050
LLI WALL
0.6217
o.oocc
0.0003
99.012C
0.082F
0.050C
0.0332
PERCENT
THYROID
9.9 * 58
C.CCGO
9.2bCl
78.7C33
i *?7S9
C.7597
C.0312
LIVER
1C. 5589
C.COO 1
C.0117
86.S265
1.4915
C . 9 79 7
C.0316
KIDNEYS
10.5190
0.0000
0.0i56
86.9219
1 .5333
0.9773
0.0329
TESTES
9.5739
0 .OQCO
0.0316
87.9962
1 .5«C5
0.6278
0.0300
OVARIES
13.2588
o.ooco
0.0138
84.6C52
1.1 361
0.9445
0.0417
-------
Table C.13. Annual Dose to the Urban Population
from One-Year-Decayed Spent Fuel
TOTAL DOSE TO EACH ORGAN THROUGH ALL PATHWAYS
ORGAN
TOT.BODY
R PAR
LUNGS
ENDOST
S WALL
LLI WALL
THYROID
LIVER
KIDNEYS
TESTES
OVARIES
DOSE(MAN-REPS)
0.2693E+03
0.2944E«03
C.5933E«03
G.3182E*02
0.2572E*03
0. 1146E*04
0.2496E*02
0.2477E*03
0.2513E+03
0.2810E+03
0.1780E+03
CONTRIBUTORS TO ORGAN DCS^S
NUCL1DE
H-3
FU-239
1-129
RU-106
CS-137
CS-134
SR-90
TOT.BODY R MAR
24.8642
0.0001
C.0312
72.5373
1.6844
0.7370
0.1459
22.6922
0.0002
0.0389
74.1578
1.7294
C.7742
0.6072
LUNGS
11 .2372
O.CC01
0.007C
87.7147
0.6S61
0.29*9
O.COOO
ENOOST
19.7993
0.0013
C.0495
76.4639
1 .7887
0.7749
1-1224
S WALL
26.0336
C.OOOO
0.0131
71.6748
1.5967
0.6784
O.Q014
.1 WALL
5.9486
o.oooc
0.00 1C
3.6251
o.2?ie
0.125£
0.027S
PERCENT
THYROID
26.7657
G.CCCO
4.2620
66.7795
1 .49C6
0.6922
0.0100
LIVER
26.9699
C.C003
C.C137
7C.561 3
1 . 6 36 2
C.809I
O.C096
KIDNEYS
26.8883
0.0000
0.0184
70.5815
1 .6e49
0.8170
0.0099
TESTES
23.8272
o.ooco
0.0373
73.6972
1.6955
0.7340
0.0089
OVARIES
37.5340
o.ocoo
0.0180
60.3727
1.2848
0.7765
0.0140
-------
Table C.14. Maximum Annual Dose to a Rural Individual
from Five-Year-Decayed Spent Fuel
TOTAL DOSE TO EACH OfiGAM TUB0UGH ALL PATHWAYS
ORGAN
TOT.EODI
B (1AB
LUNGS
EHDOST
S RAIL
LLI HALL
THYROID
LIVES
KIDHETS
TESTES
OTABIES
DOSE CRESS)
0.5513E-03
0.6044E-03
0.6300E-OJ
0.6498B-03
0.5366B-03
0. 1931E-02
0.7133E-03
0.5434E-03
0. 54778-03
0.5592B-03
0.5101B-03
O
CONTRIBUTORS TO OBHAN DOSES
NUCLIDE
N
H-3
EU-239
1-129
BU-106
CS-137
CS-134
SR-90
TOT. BODY
7 5. 02 8 5
0.0002
0.0983
17.8G51
4.7219
0.7260
1.5401
E NAP
68.3589
0. C006
0. 1141
19.4030
4.r ^80
0.7936
6.3919
LUNGS
65.6495
0.0005
0.0394
30.1514
3.6298
0.5293
0.0002
ZNUOST
61.7693
0.0038
0.1450
20.3106
5. 1211
0.7732
11.8770
S WALL
77.0006
O.OOCO
0.0373
17.9832
4.255S
0.6285
0.0146
LLI WALL
21.6065
0.0000
0.0037
76.9615
0.9112
0. 1563
0.3608
PERCENT
THYROID
57.9194
0.0000
24.8104
13.5497
3.1196
0.5255
0.0754
LIVEE
76.0237
0.0009
0.0388
18.5101
4.4944
0.8379
0.0944
KIDNEYS
75.8568
0.0001
0.0517
18.5283
4.6278
0.8371
0.0982
TESTES
73.9667
0.0000
0.1121
20.0845
4.9808
0.7596
0.0962
OVABIES
60.9870
0.0000
0.0387
15.2787
2.9048
0.6854
0.1055
-------
Table C.15. Annual Dose to the Rural Population
from Five-Year-Decayed Spent Fuel
TTTIT nncw r» B««t. *>»«... .
„„ AW tnv.,, unbaH XK PATHWAYS
OBGAN
TOT.BODI
R MAR
LUNGS
ENDCST
S HALL
LLI WALL
THTBOID
LIVEB
KIDNEYS
TESTES
0?AEIES
DOSE (MAN- HEMS)
0.8649E + 01
0. 9035E*01
0.1052E+02
0.9168E+01
0.8il83E*01
0. 1601E*02
0. 9805E + 01
0.8536E*01
0. 8720B*01
0.7857B*01
CONTEIBUTORS TO ORGAN DOSES
NOCLIDE
N
H-3
PO-239
1-129
RO-106
CS-137
CS-134
SB-90
TOT. BODY
79.5U9
0.0002
0.0956
1U.5277
4.6809
0.6126
0.5681
R HAB
*» una
75.9769
0. 0006
0,1228
15.6872
5.0565
0.6815
2.«7
-------
Table C.16. Maximum Annual Dose to an Urban Individual
from Five-Year-Decayed Spent Fuel
TOTAL OOSB TO EACH OBGAH THROUGH ALL PATHWAYS
ORGAN
TOT.EODI
£ MAfi
LUNGS
£NUOST
S MALL
LLI HALL
THYBOIO
LIVEE
KIOHEYS
TESTES
OTABIES
DOSE(BEMS)
0.3327E-02
0.3631E-02
0.3839E-02
0.3889E-02
0.3241E-02
0.1370E-01
0.4259E-02
0.3281E-02
0.3306E-02
0.3371E-02
0.3089E-02
fV
COMTRIBUTOBS TO OBGAN DOSES
•OCLIDE
H-3
PU-239
1-129
10-106
CS-137
CS-134
SI-90
TOT. BODY
76.1476
0.0002
0.0938
17.0915
4.50U8
0.6926
1.4696
BM A n
nAH
69.6780
0.0006
0.1093
18.5945
4.7313
0.7605
6.1258
LUNGS
65.9827
0.0005
0.0372
30.0500
3.4293
0.5001
0.0002
vunrtcf*
EI* Wa 1
63.2152
0.0040
0.1395
19.5426
4.9265
0.7438
11.4284
S HALL
78.1529
0.0000
0.0355
17.1425
4.0561
0.5990
0.0139
LLI HALL
9.1833
0.0000
0.0043
89.1549
1.0575
0.1813
0.4188
PEBCENT
THYBOIO
59.4170
0.0000
23.9261
13.0692
3.0082
0.5068
0.0728
LIVEB
77.1291
0.0009
0.0370
17,6577
4.2862
0.7991
0.0900
KIDNEYS
76.9686
0.0001
0.0494
17.6758
4.4139
0.7985
0.0937
TESTES
75.1381
0.0000
0. 1071
19.1814
4.7562
0.7254
0.0919
OVAEIES
61.9184
0.0000
0.0368
14.5308
2.7620
0.6517
0.1003
-------
Table C.17. Annual Dose to the Urban Population
from Five-Year-Decayed Spent Fuel
TOTAL DOSE TO EACH ORGAN THROUGH ALL PATHWAYS
ORGAN
TOT.BOD1
B NAB
LUNGS
EHDCST
S MALL
LLI WALL
THYFOID
LIVES
KIDNEYS
TBSIfS
OVABIES
DOSE(HAN-HEHS)
0.7109E + 02
0.7433E+02
0.91 17E+02
0.7515E + 02
0.6954E+02
0. 1269E*03
0.7855E + 02
0.6884E+02
0.6979E*02
0.7177E+02
0.6273E+02
O
to
o
CONTBIBUTOBS TO OBGAN DOSES
NOCLIDE
N
H-3
PO-239
1-129
80-106
CS-137
CS-134
SB-90
"OT. BODY
75.1727
0.0002
0.1182
17,6692
5.8190
0.7223
0.1498*4
R MAR
71.7442
0. 0008
0.1512
18.8902
6.2479
0.7934
2.1694
I UN n C
*• un iio
58.6218
0.0006
0.0457
36.7039
4.1313
0.4964
0.0002
END03T
66.9189
0.0056
0.2098
20.8222
6.9082
0.8489
4.2865
S WALL
76.8563
0.0000
0.0485
17.0483
5.3931
0.6492
0.0047
LLI HALL
42. 8699
0.0000
0.0095
54.3627
2.2370
0.2939
0.2270
PERCENT
THYROID
67.8895
0.0000
13.5456
13.6470
4.3203
0.5691
0.0286
LIVER
77.4658
0.0012
0.0493
16.3293
5.3702
0.7532
0.0311
KIDNEYS
77.2664
0.0001
0.0664
16.3414
5.5325
0.7610
Ort 1 1 •»
• vjJLZ
TESTES
74.4668
0.0000
0.1463
18.5571
6.0550
0.7435
0.0313
OVARIES
85.0025
0.0000
0.0511
11.0158
3.3249
0.5700
0.0358
-------
Table C.18. Maximum Annual Dose to a Rural Individual
from Ten-Year-Decayed Spent Fuel
TCTAL DOSE TO EACH CRGAN THROUGH ALL PATHWAYS
O
I
to
I-1
ORGAN
TOT.BODY
E MAR
LUNGS
ENDCST
S WALL
LLI WALL
THYSOID
LIVER
KIDNEYS
TSSJES
OVABIES
DOSE (REMS)
0.3463E-03
0.37712-03
0.3384E-03
0.4064E-03
0.3354E-03
0.3835E-03
0.51 18E-03
0.3372E-03
0.3399E-03
0.3414E-03
0.32772-03
NUCLIDE
TOT.BODY
R KAB
LUNGS
COHTETBUTOKS 10 OR3A.H DOSES
ENDOST
S WALL
?ESCENT
LLI WALL THYhOID
LIVER
KIDNLY:
OVARIES
H-3
PU-239
I- 129
RU-106
CS-137
OS-1 34
jjR-9 0
39.8030
0.0003
0.1565
0.90-41
6.7398
0.2 14 1
2. 1824
32.3773
0.0010
0. 1829
0.9876
7.0963
0.2356
9. 1196
91.9009
0.0009
0. 0733
1.7327
b.0595
0. 182o
0.0003
74.2560
0.0061
0.2319
1.0313
7.3416
0.2290
16.9043
92.7152
0.0000
0.0597
0.9136
6.1 047
0.1 d63
0.0208
81.7985
0. 0000
0. 0187
12.3060
4. 1139
0. 1458
1 .6174
60.6920
0.0000
34.5808
0.5997
3.3983
0 . 1 3 57
0.0936
92.1098
0. 0014
0.0625
0.9472
6. 4938
0.2501
0.1354
91.8925
0.0002
0.0334
0.94oO
6.6854
0.2499
0.1409
91.0365
0.0000
0. 1836
1.0446
7.3146
0.2305
0.1403
94.7865
0.0000
0.0603
0.7553
4.0543
0.1 976
0.1462
-------
Table C.19. Annual Dose to the Rural Population
from Ten-Year-Decayed Spent Fuel
TOT A I nn«-r- •»»•. —._.
•— -uoc ,u tM.H ORGAN THROUGH ALL PATHWAYS
ORGAN
TOT.BODY
R MAR
LUNGS
ENDOST
S HALL
LLI WALL
THYROID
LIVER
KIDNEYS
TESTES
OVARIES
DOSE (MAN-REHS)
0.5635E+01
0.5837E+01
0.5613E+01
0.5855E*01
0.5547E+01
0.58l3£«oi
0.6967E*01
0.5542E*01
0.56C5E*01
0.5619E+01
0.5386E*01
O
NJ
NJ
CONTRIBUTORS TO ORGAN DCSES
MJCLIDE
H-3
PU-239
1-129
RU-106
CS-137
CS-134
SR-90
TOT. BODY
91.753%
O.OOC3
O.M6S
C.7080
6.4^13
0.17*2
C.7761
R MAR
68.4160
0.0010
0. 1901
0.7710
7.0172
0.1954
3.4093
LUNGS
92.1134
0.0010
0.0717
1 .8259
5.8360
0.1518
0.0003
ENDOST
84.0616
0.0069
0.26.02
0.8579
7.6148
0.2088
6.7897
S WALL
93.2134
0.0000
0.0587
0.6824
5.8845
0.1538
0.0072
LLI WALL
90.2337
o.oooc
0.0199
4.7399
4.2687
0.123S
0.614C
PERCENT
THYROID
74.0758
o.coco
20.9418
0.4993
4.3182
0.1252
0.0398
LIVER
93.1231
C.C014
C.0597
C.6574
5.S251
0.1850
C.0477
KIDNEYS
92.9159
0-0002
0.0803
0.6589
6.1089
0.1864
0.0494
TESTES
92.0103
0.0000
0.1805
0.7609
6.8165
Q.1826
0.0493
OVARIES
95.8147
0.0000
0.0579
0.4337
3.5066
0.1357
0.0514
-------
Table C.20. Maximum Annual Dose to an Urban Individual
from Ten-Year-Decayed Spent Fuel
TOTAL DOSE TO EACH ORGAN THROUGH ALL PATHWAYS
O
to
u>
ORGAN
TOT.BODY
R MAR
LUNGS
ENOOST
S WALL
LLI WALL
THYROID
LIVER
KIDNEYS
TESTES
OVARIES
DGSE(REMS)
0.2108E-02
0.2285E-02
0.2064E-02
0.2450E-02
0.2045E-02
0.2323E-02
0.3061E-62
0.2056E-02
0.2072E-02
0.2080E-02
C.2C01E-02
CONTRIBUTORS TO ORGAN OCSES
NUCLIDE
N
H-3
PU-239
1-129
RU-106
CS-137
CS-134
SR-90
TOT. BODY
90.3539
0.0003
0.1480
0.8566
6.3743
0.2025
2.0645
R MAR
83.2555
0-0010
0.1737
0.9384
6.7417
0.2239
8.6659
LUNGS
92.2642
0.0010
0.06S2
1.7747
5.7184
0.1723
0.0003
ENOOST
75.4174
0.0063
0.2214
0.9847
7.0090
0.2186
16.1424
S WALL
93.1220
0.0000
0*0563
0.8627
5.7635
0.1759
0.0197
LLI WALL
82.6993
0.0000
0.0178
11.6977
3.909t
0.1385
1.5371
PERCENT
THYROID
62.1566
0.0000
33.2922
0.5774
3.7528
0.13C6
0.09C1
LIVER
92.5474
C-C014
C.0590
C.854
-------
Table C.21. Annual Dose to the Urban Population
from. Ten-Year-Decayed Spent Fuel
TOTAL DOSE TO BACH OBGAN TUHOUGH ALL PATHWAYS
ORGAN
TOT.BODY
ft NAR
LUNGS
ENDOST
S HALL
LLI WALL
THYROID
LIVEB
KIDNEYS
TESTES
OVAEIES
DOSE (MAN-KEMS)
0.4478E+02
0.4636E+02
0.4474E+02
0.4610E+02
0.4404E + 02
0.4597E+02
0.5421E*02
0.4391E+02
0.4453E+02
0.4472E+02
0.4230E*02
O
l
to
CONTRIBUTORS TO ORGAN DOSES
NUCLIDE
N
H-3
PO-239
1-129
BU-106
CS-137
CS-134
SR-90
TOT. BODY
89.7222
0.0004
0.1876
0.8907
8.2825
0.2124
0.7043
B MAB
86.4774
0.0013
0.2473
0.9617
8.9810
0.2356
3.0958
LUNGS
89.7966
0.0013
0.0931
2.3747
7.5467
0.1873
0.0004
ENDOST
82. 001^
0.0091
0.3420
1.0777
10.0950
0.2563
6.2188
S WALL
91.2373
0.0000
0.0766
0.8548
7.6349
0.1899
0.0066
LLI WALL
88.9650
0.0000
0.0261
4.7648
5.5361
0.1503
0.5578
PERCENT
THYBOID
73.9462
0.0000
19.6248
0.6278
5.6117
0.1527
0.0369
LIVEB
91.2983
0.0018
0.0773
0.6128
7.5477
0.2187
0.0434
KIDNEYS
91.0427
0.0002
0.1040
0.8133
7.7741
0.2209
0.0449
TESTES
89.8423
0.0000
0.2347
0.9456
8.7117
0.2210
0.0447
OVARIES
94.7805
0.0000
0.0757
0.5188
4.4212
0.1566
0.0472
-------
BIBLIOGRAPHIC DATA
SHEET
1. Report No.
EPA-520/3-80-007
3. Recipient's Accession No.
4. Tide and Subtitle
RADIATION EXPOSURES FROM SOLIDIFICATION PROCESSES FOR
HIGH-LEVEL RADIOACTIVE LIQUID WASTES
5. Report Date
MAY 1980
6.
7. Author(s) William F. Holcoinb, William N. Crofford,
Raymond L. Clark and Frederick C. Sturz
8. Performing Organization Rept.
No.
9. Performing Organization Name and Address
OFFICE OF RADIATION PROGRAMS (ANR-460)
U. S. ENVIRONMENTAL PROTECTION AGENCY
WASHINGTON, D.C. 20460
10. Projcct/Task/Work Unit No.
11. Contract/Grant No.
12. Sponsoring Organization Name and Address
Office of Radition Programs (ANR-460).
U. S. Environmental Protection Agency
Washington, D.C. 20460
13. Type of Report & Period
Covered
14.
15. Supplementary Notes
16. Abstracts The office of Radiation Programs, U.S. Environmental Protection Agency (ORP/EP.
has prepared this analysis as technical support for EPA's proposed environmental
radiation protection standards, 40 CFR 191, concerning the management and disposal of
high-level radioactive wastes. For Subpart A of 40 CFR 191, waste management and storage
operations, EPA proposes to extend the limitations of 40 CFR 190 to these operations.
EPA/ORP developed a generic high-level liquid waste solidification plant and
assessed the potential environmental impact of atmospheric discharges during normal
operations in four solidification processes: fluidized-bed calcination, spray calcination
and glassification by in-can melting and continuous melting. We used a newly developed
computer code, AIRDOS-EPA, to perform the assessment.
17. Key Words and Document Analysis. I7o. Descriptors
7b. Idcntifiers/Opcn-Ended Terms
High-level radioactive liquid wastes
Solidification processes
AIRDOS-EPA
Atmospheric discharges
Environmental radiation protection .standards
Off-gas releases
7c. COSATI Fie Id/Group
B. Availability Statement
19. Security Class (This
Report)
UNCLASSIFIED
20
Security Class (This
P"*e
AsstFirn
21. No. of Pages
22. Pri<
KNDOKSED HY ANSI AND UNhSCO.
THIS FORM MAY UK Kl-.PROUUCKD
USCOMM-OC 6269-P74
-------
------- |