EPA-520/3-75-02J
TRANSPORTATION ACCIDENT
RISKS IN THE NUCLEAR
POWER INDUSTRY 1975-2020
OFFICE OF RADIATION PROGRAMS
US. ENVIRONMENTAL PROTECTION AGENCY
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TRANSPORTATION ACCIDBTT
RISKS IN THE NUCLEAR
POWER INDUSTRY 1975-2020
OFICE OF RADIATION PROGRAMS -
U,S, BMRONIWTAL PROTECTION AGENCY
WASHINGTON, B.C. 20460
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EPA Review Notice
This report has been reviewed by the EPA and
approved for publication. Approval does not
signify that the contents necessarily reflect the
views and policies of the EPA, nor does mention
of trade names or commercial products consti-
tute endorsement or recommendation for use.
11
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FOREWORD
The Office of Radiation Programs carries out a national program
designed to evaluate the exposure of man to ionizing and nonionizing
radiation, and to promote the development of controls necessary to
protect the public health and safety and assure environmental quality.
The analyses presented in this report were made for the Office
of Radiation Programs, .Environmental Protection Agency, by Holmes and
Narver, Inc., under contract. This report represents one of the
first efforts to quantitatively assess the potential impact of
accidents occurring in the transportation of radioactive materials
associated with the nuclear power industry through the year 2020.
Technical data from numerous sources were collected and analyzed to
produce the results reported herein. While not all of the radiological
aspects of transportation analyzed in the report are covered in the
detail which may be ultimately necessary, each area has received
sufficient analysis to provide information useful in environmental
impact statement reviews and other activities of the Agency. The
results of this study will also provide an input into a planned EPA
review of the need for additional protection standards for the trans-
portation of radioactive materials.
This publication is made available as a resource to the
scientific community and the public generally. Because of the
intended uses, the study may be of considerable interest to a large
number of persons; therefore, it is likely that interested parties
may wish to comment on the report, or certain aspects of it. Comments
may be' submitted to the Environmental Protection Agency, Office of ^
Radiation Programs, Washington, D.C. 20460.
W. D. Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs
111
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ABSTRACT
A quantitative assessment was made of the accident risks associated with
the transportation of radioactive material in the nuclear pow'er industry
to the year 2020. A scenario was developed for the period from 1975 to
2020 for shipments between nuclear reactors, chemical reprocessing
plants, fuel fabrication facilities and waste repositories. Technical data
from numerous sources were collected for shipments of spent fuel,
recycled plutonium, high-level radioactive solid waste and fission-product
gases. Assumptions were developed regarding transport modes, shipping
containers, transport pathways for dispersion of released radioactivity, and
population distribution;.
Fault tree analysis was used to estimate the probability of release from
shipping containers taking into consideration the transport mode, severity
of the accident and damage to the package. Estimates were made of the
fraction of the radioactive contents which could be relased as a result of
each class .of accident. Calculations were made of the annual frequency of
accidents and the frequency of radiological releases. Risks were charac-
terized by the probability of release and the magnitude of the resulting
exposure. Both individual exposure and population doses were calculated.
The results of the study indicate that by 2020 the number of shipments
exceed 37, 000 per year with a total shipping distance of more than 17 million
miles. As a result, the number of accidents is in excess of 20 per year.
However, the frequency of release is small, the expected value being less
than one per 13 years in 2020. Releases from accidents involving spent
fuel'dominate the release frequency during the entire period studied.
The average annual population dose from transportation accidents of the
nuclear power industry is insignificant--less than one person-rem per
year from all sources in 2020. A release, if it should occur, is most
likely to result in a relatively small exposure fo individuals near the
accident, on the average.
IV
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This report was submitted by the Nuclear & Systems Sciences Group of
Holmes & Narver, Inc. , to the Office of Radiation Programs, Environ-
mental Protection Agency, in fulfillment of Contract No. 68-01-0555.
Work was completed as of November 1974.
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CONTENTS
Section Page
I SUMMARY AND CONCLUSIONS 1
II RECOMMENDATIONS 4
III STUDY PLAN 5
Introduction 5
EPA Objectives for the Study 5
Study Methodology 6
Elements Included in the Study - 9
Organization of Study 11
IV THE UNITED STATES NUCLEAR INDUSTRY 12
Introduction 12
Population Growth in the United States 14
Demand for Electricity in the United States 14
Nuclear Supply of Electricity 16
Nuclear Fuel Cycles 19
V NUCLEAR TRANSPORTATION FORECASTS 30
(1975 to 2020)
Annual Fuel Requirements 30
Annual Spent Fuel Transportation 30
Annual Plutonium Transportation 42
Annual Transportation of High Level .50
Radioactive Solid Waste
Annual Transportation of Gaseous 57
Fission Products
Summary 61
VI TRANSPORTATION ACCIDENT RISKS . 63
Introduction 63
Accident Probability and Severity 64
Containment Failure in Shipment Accidents 67
Calculations of Release Probabilities 84
VI
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CONTENTS (continued)
Section
VII CONSEQUENCES OF RADIOACTIVITY RELEASES 91
Introduction 91
Release Fractions 92
Dispersion of Radioactivity 94
VIII RISKS FROM TRANSPORTATION ACCIDENTS 100
Introduction 100
Accident Frequency 100
Individual Exposures 105
Population Doses 108
IX REFERENCES 116
VII
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FIGURES
Figure
1 Overall Organization of Evaluation of Risk from 7
Transportation Accidents in the Nuclear Power
Industry
2 Detailed Organization of Evaluation of Risk from 8
Transportation Accidents in the Nuclear Power
Industry
3 Projected Nuclear Electric Capacity 13
4 Projected Electricity Demand 17
5 Projection of Generating Capacity of Various 18
Nuclear Reactor Types
6 Schematic Fuel Cycle Diagram for Light Water 20
Reactor
7 Material Flow in Typical LWR Fuel Cycle Without 24
Plutonium Recycle
8 Material Flow in Typical LWR Fuel Cycle With 26
Plutonium Recycle
9 Material Flow in Typical HTGR Fuel Cycle 27
10 Material Flow in Typical LMFBR Fuel Cycle 28
11 . Projection of Annual Nuclear Fuel Fabrication 32
Requirements
12 Projection of Annual Nuclear Fuel Reprocessing 33
Requirements
13 Projection of High-Level Radioactive Solid Waste 52
Shipping Requirements
14 Summary of Annual Transportation Activity 62
15 Schematic Diagram of Spent Fuel Shipping 70
Container
16 Schematic Diagram of Recycled Plutonium 70
Shipping Container
17 Schematic Diagram of High Level Radioactive 72
Solidified Waste Shipping Container
18 Schematic Diagram of Fission Gas Shipping 72
Container
Vlll
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FIGURES (Continued)
Figures
19 Fault Tree Diagram for Spent Fuel Shipping 74
Container
20 Fault Tree.Diagram for Plutonium Shipping 75
Container
21 Fault Tree Diagram for Solid Waste and Noble 76
Gas Shipping Containers
22 Frequency of Transportation Accidents 101
23 Frequency of Transportation Accident Releases 103
24 Annual Whole Body Population Dose 110
25 Annual Lung Population Dose 111
26 Annual GI Tract Population Dose 112
27 Annual Thyroid Population Dose 113
IX
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TABLES
Table
1 Projected Population of the Conterminous United
States
2 Forecast of Nuclear Generating Capacity 16
3 Annual Production of Fabricated Fuel 31
4 Annual Discharge of Spent Fuel After Cooling 34
5 Assay of Nuclear Fuels at Time of Reprocessing 35
6 Typical Radioactivity in Fuel and Wastes at Fuel .36
Reprocessing Plants
7 Radioactivity of Annually Transported Spent Fuel 38
8 Spent Fuel Shipment Capacities 39
9 Spent Fuel Transportation Scenario 40
10 Distances Between Nuclear Power Reactors 41
Operating in 1970 and Known Chemical Process
Plant Sites
11 Distances Between Sea Ports for Spent Fuel 43
Shipments
12 Summary of Annual Spent Fuel Shipping Data 43
13 Annual Plutonium Generation in Chemical 44
Reprocessing Plants
14 Specific Activity of Plutonium 46
15 Radioactivity of Annually Transported Plutonium 46
16 Plutonium Shipment Capacities 47
17 Plutonium Transportation Scenario 48
18 Distances Between Chemical Reprocessing Plants 49
and Fuel Preparation Plants in 1973
19 Summary of Annual Plutonium Shipping Data 50
20 Annual Transportation Requirements for High 51
Level Radioactive Solid Waste
21 Radioactivity of Annually Transported High-Level 54
Radioactive Solid Waste
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TABLES (Continued)
Table Page
22 Shipment Capacities for High-Level Radioactive 55
Solid Waste
23 Distances Between Chemical Processing Plants 56
in 1973 and Possible Federal Waste Repositories
24 Summary of Annual Shipping Data for High-Level 56
Radioactive Solid Waste
25 Radioactivity of Annually Transported Noble Gas 58
26 Shipment Capacities for Noble Gas 59
27 Summary of Annual Shipping Data for Noble Gas 60
28 Probabilities of Accidents 64
29 Accident Frequency Statistics 65
30 Accident Severity Classes 66
31 Conditional Probabilities of Impact, Puncture, 68
and Fire
32 Fault Tree Input Data for Spent Fuel Shipping 77
Container on Truck
33 Fault Tree Input Data for Spent Fuel Shipping 78
Container on Rail
34 Fault Tree Input Data for Spent Fuel Shipping 79
Container on Barge
35 Fault Tree Input Data for Plutonium Shipping 80
Container on Truck
36 Fault Tree Input Data for Plutonium Shipping 81
Container on Rail
37 Fault Tree Input Data for Shipping Container of 82
High Level Radioactive Solid Waste or Noble
Gas on Truck
38 Fault Tree Input Data for Shipping Container of 83
High Level Radioactive Solid Waste or Noble
Gas on Rail
39 Selected Release Sequences for Spent Fuel Shipping 85
Container
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TABLES (Continued)
Table
40 Selected Release Sequences for Plutonium 86
Shipping Container
41 Selected Release Sequences for High Level 87
Radioactive Solid Waste or Noble Gas Shipping
Container
42 Release Probabilities for Shipping Containers 89
43 Release Probability for Adopted Transportation 90
Scenario
44 Release Fractions in Transportation Accidents 95
45 Dose Coefficients 97
46 Average Population Density 99
47 Average Annual Release of Radioactivity 104
48 Average Annual Individual Doses at 0. 1 Mile 106
49 Conditional Individual Doses at 0. 1 Mile 107
50 Worst Case Individual Doses in 2020 at 0. 1 Mile 109
51 Conditional Population Doses in 2020 115
XII
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ACKNOWLEDGEMENTS
Contributions to the study were made by Mr. J. H. Wilson, data collection;
Dr. O. C. Baldonado, methodology; Dr. S. Kaplan, review and comment;
and Dr. B. J. Garrick, project direction. Many other individuals and
organizations contributed to the study. Special thanks are extended to
Mr. W. A. Brobst of the U. S. Atomic Energy Commission and to
Messrs. J. Nichols and L. B. Shappert of the Oak Ridge National Laboratory
for their help.
The assistance and cooperation of Mr. J. L. Russell, Project Officer for
the Environmental Protection Agency, in supplying documents and project
direction is gratefully acknowledged.
Xlll
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SECTION I
SUMMARY AND CONCLUSIONS
The nuclear power industry is expected to grow rapidly during the next
50 years and thus the concomitant transportation activity also will increase.
Associated with the transportation of radioactive materials is a risk of the
possible release of radioactivity from accidents occurring during transport.
This radiological risk will increase with the growth in total transportation
of nuclear fuels and other radioactive materials of the nuclear power
industry. The purpose of this study was to assess risks to the public from
the accidental release of radioactivity during transportation of radioactive
materials in the nuclear power industry to the year 2020.
The usual method of calculating risks is to use information based on
statistical records. However, little experience exists on transportation
accidents for shipments of radioactive materials from nuclear power
facilities because the number of shipments to date has been small and the
probability of an accident during transportation is low. Nevertheless, it
is possible to estimate the risk by a process of synthesis, because shipments
of radioactive material in the nuclear fuel cycle are not fundamentally
different from shipments of other hazardous materials transported in routine
commerce. In this study, the amount of damage, as a result of an accident
to a container used for the shipment of radioactive materials, was synthe-
sized by consideration of the design of the container and the severity of the
accident. Fault tree analysis was used to obtain quantitative estimates of
the release probability resulting from an accident.
The general definition of risk used in this study takes into account both the
probability of occurrence of an undesirable consequence and the magnitude
or value of the consequence. The risk of a specific consequence is the
probability of weighted sum of all events leading to the consequence, i. e. ,
the expected value or average. The consequences evaluated in this study
were radioactivity released and population doses.
Assumptions were made about the timely introduction of plutonium recycling
programs, breeder reactors, fuel reprocessing and fabrication facilities,
and waste disposal management in order to quantitatively describe the
magnitude and rate of growth of the nuclear economy from 1975 to 2020.
Spent fuel, plutonium, high-level radioactive solid waste, and fission-
product gases were chosen as the radioactive materials requiring transpor-
tation. The transportation modes considered were trucks, rails, and
barges. Four parameters were evaluated: frequencies of accidents from
accident statistics; frequencies of releases from fault tree analysis;
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amounts of radioactivity released from postulated release fractions; and
individual and population doses from an environmental dispersion model.
The results are presented as annual averages, i.e., averaged over the variety
of transportation modes, accident severities and package damage severities,
and as worst-cases for accidents which occur very infrequently.
As a result of the assumptions made regarding the growth of the nuclear
power industry, the transportation activity will increase each year during
the entire 50-year period. Nuclear power transportation activity will
exceed one million miles in 1980 and ten million miles after 2000. By
2020, there will be 37, 000 shipments per year averaging 470 miles per
shipment - a total shipping distance of more than 17 million miles per
year.
The probability of an accident occurring in commercial transportation is
small. Based on current accident statistics, there will be about 1. 3
accidents per million vehicle miles in 2020. The total accident frequency
will be less than one per year in 1975, exceed one per month after 2000,
and reach almost two per month in 2020.
In the overwhelming majority of cases, these accidents will not result in
the release of radioactivity. On the average, only about 0. 3 percent of
the accidents in 2020 result in releases. The average release frequency
ranges from approximately one release per 250 years in 1975 to one release
per 13 years in 2020. The approximate average release rate ranges from
one release per 100, 000 shipments in 1975 to one release per 500, 000
shipments in 2020.
The maximum average releases occur in 2020 when the transportation
activity is the largest for the period under study. The average annual
release of radioactivity from spent fuel is 13 curies of Kr-85, 0. 031 curies
of 1-131 and 6. 3 curies of mixed fission products. The average annual
release from other materials transported in 2020 are: 0.017 curies from
a plutonium shipment accident; 0. 19 curies of fission products from high-
level radioactive solid waste; and 11 curies from noble gas shipments.. The
expected values of the doses in 2020 to a hypothetical individual at a
distance of 0. 1 mile on the centerline downwind from an average ground
level release from a transportation accident under average weather condi-
tions are moderate.
_4
The expected value of the annual population dose in 2020 is 6.4 x 10 person-
rem whole body, 2.2 x 10~* person-rem to the lungs, 1.8 x 10~^ person-
rem to the GI tract and 1. 9 x 10~3 person-rem to the thyroid.
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The averaging process distributes the releases over all shipments. The
actual release, when it occurs, will be larger than the average values. A
typical actual release from an accident involving spent fuel would be about
1,000 curies of Kr-85, 1 curie of 1-131, and 400 curies of mixed fission
products. The release frequency would range from approximately one
release per 250 years in 1975 to one release per 13 years in 2020. A
typical actual release from a plutonium shipment accident would occur
about once per 400 years in 2020 and release about 10 curies. A
typical actual release from high-level solid radioactive waste would be about
2,000 curies with a frequency of occurrence in 2020 about once in 2,000 years.
The smallest actual release from a noble gas shipment can be no less than
the contents of one gas cylinder - about 180,000 curies of Kr-85. The
frequency of occurrence is once in 14,000 years in 2020.
The highest lung exposure (|.£Mrem) arises from plutonium with a frequency
of occurrence in 2020 of about once in nine million years. An exposure of
9,900 rem to the lung could result from a high-level solid radioactive
waste accident about once per seven million years in 2020. The lung exposure
from spent fuel release could be 2,300 rem with a frequency of about once
p'er 90,000 years. ^
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SECTION II
RECOMMENDATIONS
This study has shown that the risks from the release of
radioactivity from transportation accidents in the nuclear power industry
is relatively small. This result is due primarily to the low probability
of accidents, the small fraction of the accidents resulting in release of
radioactivity, and because the majority of releases are relatively small
fractions of the radioactive contents. Nevertheless, when a release does
occur, the amount of radioactivity released is sufficient to raise issues
of public concern. Further effort is warranted to define an acceptable
level of risk for activities whose consequences are highly statistical in
nature.
The following recommendations are made to verify the low expected
values of the consequences and to assure that uncertainties in some of the
parameters would not alter the conclusions:
1. Accidents should be monitored to verify release probabilities and
the relationships between accident severity and container
damage.
2. Experiments should be performed to develop data for modeling
energy transfer during accidents, i. e. , the relationships
between accident severity and container damage over a wide
range of accident severities. The probability and severity of
container damage should be used to evaluate the adequacy of
regulatory standards for shipping containers.
3. Investigations should be conducted to verify the relationships
between container damage severity and release fraction.
4. A data bank and associated data handling methodology should be
implemented to accommodate distributional variations in the
parameters used in this study. Among these parameters are
nuclear power growth, facility locations, transportation scenarios,
accident severity, container damage severity, release probability,
release fractions, dispersion models, and population distribution.
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SECTION in
STUDY PLAN
INTRODUCTION
The energy demands of the United States are increasing. Traditional
energy sources are being depleted, and this fact means that there will be
an increase in the fraction of energy supplied by nuclear fuel. By the
year 2020, nuclear fuel is expected to provide as much as three-fourths
of the electrical power of the United States. By that time, electrical
power is expected to supply as much as one-third of the total energy
requirements of the country.
The increased use of nuclear fuel will result in more mining, fuel
enrichment, fuel fabrication, fuel reprocessing, and nuclear waste
disposal. The facilities required to carry out these activities will not
necessarily be located within the same area. It will be necessary,
therefore, to transport nuclear fuels and wastes in a variety of forms
and levels of radioactivity as it goes through the fuel cycle.
Associated with the transportation of these radioactive materials is a risk
of the possible release of radioactivity during normal conditions of trans-
port and from accidents. This radiological risk will increase with the
growth in total transportation of nuclear fuels and radioactive materials
of the nuclear power industry. The purpose of this study is to assess the
risks from transportation accidents in the nuclear power industry to the
year 2020.
EPA OBJECTIVES FOR THE STUDY
The responsibility for assessing and minimizing the detrimental environ-
mental impact from many of man's activities rests with the U. S.
Environmental Protection Agency (EPA). As a part of these responsi-
bilities, EPA has undertaken the assessment of the total environmental
impact resulting from the production of nuclear power. The transporta-
tion of nuclear materials may represent a significant fraction of the total
impact resulting from the nuclear power industry. As the nuclear trans-
portation industry grows, a larger burden of radioactivity may have to be
borne by the public and the environment.
In the present study, the possible radiological releases as. a result of
transportation accidents and the consequences of such releases are
analyzed. The purpose of the study is to assist the EPA in developing
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base information for use in establishing policies for the transportation
of radioactive materials generated by the nuclear power industry.
Additional objectives of the present study are to identify and evaluate
important variables and to apply fault tree analysis to determine the
frequency of releases of radioactive materials under varying accident
conditions.
STUDY METHODOLOGY
The method used in this study is essentially a mapping mechanism.
Figure 1 contains a diagram which illustrates the overall action of the
mapping. The idea is to map a function, the amount of radioactive
material being shipped, into a set of risks from accidents to the ship-
ments involved. The amount of material being shipped is a function of
several variables concerning the nuclear power and transportation
industries.
In this study, risk is defined as the probability of an undesirable conse-
quence times the magnitude or value of the consequence. Among these
consequences are: release of radioactivity, population doses, etc. The
total risk to the public is the expected value of the consequence under
consideratiori. The expected value is the probability weighted sum of all
values of the consequence.
The overall view described by Figure 1 is given in more detail in Fig-
ure 2. A series of calculations produced the final mapping. First, the
amount of radioactive material being shipped is calculated from such vari-
ables as the number and power of nuclear generators, the capacity of
chemical reprocessing plants, the number of metric tons of fuel burned
and the isotopic composition of the residues. Second, the probability of
accidents is calculated based on available accident statistics by trans-
portation mode. Third, the probability of release from a given accident
is calculated by means of fault tree analysis. Fourth, the environmental
distribution of radioactivity released from accidents is calculated by
means of a dispersion model. The results of this calculation are esti-
mates of the dose to a hypothetical individual in an assumed scenario and
the dose to the population.
The fault tree simulation model is based on the representation of the
shipping container as a series of barriers that are breached with some
computable probability. The use of a barrier model in a fault tree is a
way to calculate the conditional probability that radioactive material is
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Materials Modes :
Spent Fuel Trucks
Recycled Plutonium Trains |
High Level Radio- Barges
active Solid Waste <
Fission Product .
Gases
•Data
Nuclear Facilities
Transport Modes (Acci-
, dent Probabilities)
Transport Routes
Radioactive Materials
• Container Designs
; Population Density
Nuclear ;
Reactors <
Waste
Repositories ,
Chemical
Reprocessors
Fuel
Fabricators
V
Mapping Methodology
A
Nuclear Power Transportation Network
A
Fault
Tree
Simulation
Radio-
activity
Dispersion
Model
Radioactivity
Released
Acute Dose
Population Dose
, Risks
FIGURE 1: OVERALL ORGANIZATION OF EVALUATION OF RISK FROM TRANSPORTATION ACCIDENTS
IN THE NUCLEAR POWER INDUSTRY
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Nuclear Industry Transportation Network
Nuclear Facilities
Nuclear Reactors
Chemical Processing Plants
Fuel Fabrication Plants
Waste Repositories
Transportation Modes
Trucks
Trains
Barges
Materials
Spent Fuel
Recycled Plutonium.
High Level Radioactive Solid
Waste
Fission Product Gases
Data Inputs
Number of Facilities
Power of Generators
Capacity of Reprocessors
Isotope Composition of
Materials
Storage Policies
Radioactivity of Materials
to Be Shipped
Capacity of Shipments
Distance Between Facilities
Accident Frequencies by
Transport Modes
Container Designs
Population Density
Fault Tree
Simulation
Model
Radioactivity
Dispersion
Model
Mapping
Methodology
Radioactivity
Released
Indivddual
Dose
Population
Dose
Risks
FIGURE 2. DETAILED ORGANIZATION OF EVALUATION OF RISK FROM
TRANSPORTATION ACCIDENTS IN THE NUCLEAR POWER INDUSTRY
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released, given that an accident has occurred. These fault trees require
input data in the form of conditional probabilities of elementary events.
Examples of elementary events might be occurrence of puncture force
greater than that which the barrier can withstand, or failure of a seal
due to heat from a nearby fire. Such data are obtained from laboratory
or field tests, distribution functions, .statistical tabulations for similar
events, and theoretical estimates. Once the fault tree has'been com-
pleted and elementary probability data has been assigned, the release
probability can be computed by Boolean algorithm or Monte Carlo
simulation.
ELEMENTS INCLUDED IN THE STUDY
Nuclear facilities which produce or handle radioactive materials requiring
transportation are confined in this study to the following:
1. Nuclear power reactors.
2. Chemical reprocessing plants.
3. Fuel fabrication plants.
4. Radioactive waste repositories.
The transportation modes which are considered in this study are:
1. Truck.
2. Rail.
3. Barge.
Operation of nuclear fuel cycles involves transportation of many materials.
The following selected list of transported materials is chosen for study,
since they are assumed to represent the highest potential risks under
accident conditions:
1. Spent fuel.
2. Recycled plutonium.
3. High-level radioactive solidified waste.
4. Fission product gases .
Results of the analysis are presented as a set of risks: ,
1. The frequency of transportation accidents.
2. The radioactivity released from an accident.
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3. The dose to a hypothetical individual exposed at an assumed
distance from the point of release.
4. The population dose.
In this study, annual risks are obtained for nuclear facility related trans-
portation within the conterminous United States between 1975 and 2020.
The set of risks is generated at five-year intervals, beginning in 1975.
The fundamental parameters of the problem are listed as follows:
1. Capacities of shipments .
2. Number of shipments.
3. Average distance between facilities.
4. Transport mode mix.
5. Radionuclide composition of material cargo.
6. Physical form of material cargo.
7. Physical nature of accidents.
8. Accident severity.
9. Probability of loss of containment (release probability).
10. Fraction of radionuclide content of cargo released after loss
of containment (release fraction).
11. Dispersion conditions in environment.
12. Biological response to specific radionuclides.
13. Population density distribution.
In the present risk analysis of accidents, the principal parameters varied
are population density, transport mode mix, material cargo, accident
severity, and release fraction.
The severity of accidents is divided into minor, moderate, and severe
categories. These classifications are functions of the relative velocity
of colliding vehicles and of the time duration of fires associated with
10
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accidents. The release fraction is divided into small, medium and large
categories.
ORGANIZATION OF STUDY
The potential risks associated with radioactivity released from nuclear
transportation accidents in the nuclear power industry are identified and
quantified in the following way: In Section IV, a reasonable picture of the
United States nuclear industry from 1970 to 2020 is presented. This
information is used in Section V to predict the amount and type of nuclear
material transportation. The probabilities for the various types of
accidents which can lead to release of radioactivity is calculated in
Section VI. The release fraction, environmental dispersion, and population
dose models are described in Section VII. The evaluation of public health
risks from transportation accidents to the year 2020 is summarized in Section
VIII.
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SECTION IV
THE UNITED STATES NUCLEAR INDUSTRY
INTRODUCTION
To evaluate the environmental impact of accidents occurring during
transportation of nuclear material," ifTs necessary to know the numbers,
origins, and destinations of the shipments; the types of vehicles and
shipping containers used; and the expected population densities along the
routes. These parameters are dependent on projections of the develop-
ment of the nuclear industry and the population during the 50 years
covered by this study.
The aspects of the nuclear industry which are considered in a projection
are:
1. The magnitude of the installed nuclear power capacity.
2. The relative contributions of each type of reactor to this
capacity.
3. The long term disposal (or storage) policies for radioactive
waste.
• 4. The economics of the fuel cycle.
Projections should be reevaluated periodically to make use of the most
recent data. An example of an updated projection is given in Figure 3
(Reference 1). This figure shows the AEC's forecasts of the nuclear
generating capacity in the decade 1970 to 1980. The first estimate was
made in 1962. The forecast was revised in 1964, 1966, 1967, and 1969;
each time it was revised upward. A more recent projection by the staff
at Oak Ridge National Laboratory (Reference 2) indicates that the nuclear
generating capacity in 1980 will more closely agree with the upper esti-
mate made in 1966, 114 thousand megawatts.
In the 50-year period of this study, society may change its pattern of
energy consumption. Technological breakthroughs may occur which will
invalidate the underlying assumption of the projections. Nevertheless,
since decisions must be made today, it is necessary to make predictions
based on the best information and opinions available today.
The development of the United States nuclear industry over the next 50
years will be governed primarily by the demand for electrical energy.
The energy crisis may result in a decrease in total electricity demand,
but such a decrease is regarded as a fluctuation which does not influence
the long term upward trend of demand. Military and scientific influences
12
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THOUSAND MEGAWATTS
200
160 -
120 -
1970
1975
1980
1964
1962
ESTIMATED
INSTALLED
CAPACITY
Reference: "The Nuclear Industry - 1970," U. S. Atomic
Energy Commission
FIGURE 3. PROJECTED NUCLEAR ELECTRIC CAPACITY
13
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on the development of the nuclear industry are expected to be small
compared to the influence of the domestic needs for more energy. In
this section, the following aspects of the nuclear electric industry for
the next 50 years are discussed:
1. Demand for electricity in the United States between 1970
and 2020.
2. Ways in which the electrical energy demand will be met.
3. Numbers, sizes, and types of facilities needed to support the
nuclear energy requirements.
POPULATION GROWTH IN THE UNITED STATES
One of the major forces governing energy demand is the growth of
the United States population. Table 1 shows the population"of the
conterminous United States in 1970 and the projected populations through
the year 2020. This projection is based on an average of 2, 775 children
per 1,000 women at end of child bearing and a net annual immigration
of 400, 000. This birth rate is higher than that recently released by the
Census Bureau. They report a rate of 2,040 children per 1,000 women
(Reference 3).
The population of the United States is expected to increase significantly
during the next 50 years. The population'in 2020 is expected to be about
79 percent higher than in 1970. If the per capita energy demand were to
remain constant, an increase of 79 percent in the overall energy demand
would be expected by the year 2020.
The conterminous United States population density projections are also
displayed in Table 1. They are based on a constant land area of 3. 04
million mi .
DEMAND FOR ELECTRICITY IN THE UNITED STATES
The standard of living and life style in the United States uses large
amounts of energy for each person. It has been estimated that while
having 6 percent of the world's 1970 population, the United States
consumed one-third of the world's 1970 annual production of energy.
The demand for energy increases when the population increases or when
the per capita demand for energy increases. Demand for electricity is
only part of the total demand for energy, but it is the part for which
nuclear reactors can provide an important supply.
14
-------
TABLE 1. PROJECTED POPULATION OF THE
CONTERMINOUS UNITED STATES
Year
1970
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Totala
Population
(10° Persons)
204
218
233
251
269
281-
292
310
327
346
365
Average0
Population Density
(Persons /mi^)
67. 1
71.7
76.6
82.6
88.6
92.4
96.0
102.0
107.6
113.8
120.0
Statistical Abstract of the United States, 1972, U. S.
Government Printing Office, July, 1972. Census Series C
projections are used, based on a constant annual immigra-
tion rate of 400, 000 and a constant fertility rate of 2, 775 ^
children per 1, 000 women at end of child bearing.
°Based on constant land area of 3. 04 x 10° mi .
The per capita demand for electricity was recently calculated from a
projection of national demand (Reference 2). While the population is
expected to increase by 1.8 over the next 50 years, the per capita
demand for electricity is expected to increase 7.-5 times from 21.5
kilowatt-hours per day (kwh/day) in 1970 to 160 kwh/day in 2020. Stated
differently, each person used on the average about 0.327 megawatt-days
(Mwd) of electricity during 1970, and would use about 2.4 Mwd in 2020.
The total demand for electric energy in 1970 was 67 million Mwd com-
pared with an estimated 888 million Mwd in 2020.
15
-------
To illustrate the relations between electricity demand and population,
the graphs of population, per capita demand, and total demand have been
superimposed on the same figure in Figure 4.
NUCLEAR SUPPLY OF ELECTRICITY
i
The nuclear role in the electric economy will be played by both nuclear
converter and breeder reactors. At the present time, the most important
converter designs in the United States appear to be the light water reactors
(LWRs), which include both the pressurized water reactors (PWRs) and
the boiling water reactors (BWRs), and the high temperature gas cooled
reactors (HTGRs). The most active development in the breeder field is
currently the liquid metal cooled fast breeder reactor (LMFBR). Other
breeder designs may account for appreciable contributions to the nuclear
power picture within the next 50 years, but these will require further
development.
A recent projection of installed nuclear generating capacity (presented in
Table 2 and Figure 5) based on a study of economic competition between
LWRs,'HTGRs, LMFBRs, and fossil plants (Reference 4), was utilized
for the analyses in this report.
TABLE 2. FORECAST OF NUCLEAR
GENERATING CAPACITY
Year
1970a
1975
1980
1985
1990
1995
2000
2005
2010b
2015
2020
Power Capacity (10^ Mwe)
LWR
5
45
112
210 '
345
476
546
533
520
500
- 490
i
HTGR
0.3
2
35
93
151
201
273
350
455
590
LMFBR
16
113
370
756
1, 140
1,590
1,960
Total
5
45
114
245
454
740
1,117
1,562
2,010
2,545
3,040
aORNL-TM-4224.
Graphical extrapolation for years after 2005.
16
-------
350
a
o
-------
1970
1980
1990
2000
2010
2020
FIGURE 5. PROJECTION OF GENERATING CAPACITY OF
VARIOUS NUCLEAR REACTOR TYPES
18
-------
Earlier studies of optimum mixtures of generating plant designs have been
published (References 5, 6). These forecasts differ only slightly from
the one given above.
NUCLEAR FUEL CYCLES
To understand the transportation requirements of the nuclear power
industry, the fuel cycles associated with the reactors producing the power
must be understood. Transportation occurs between all plants in the
nuclear fuel cycle since almost all the plants are geographically
separated. In the future, construction of nuclear parks may obviate the
necessity for transportation services for some materials, such as spent
fuel and recycled plutonium, but the need for transporting wastes,
uranium ore, thorium ore, and enriched fuel is not likely to be reduced.
Permanent waste repositories are in the development stage. One location
in southeastern New Mexico is being examined as a possible site. A first
coring sample of the salt bed information is expected to be taken in 1974
to allow a choice to be made between sites in New Mexico and Kansas. It
is not anticipated that pilot plant operations will be started prior to 1982,
while regular operation of the facility on a nonexperimental basis is not
expected to take place before 1993. It is anticipated that if this project
is successful and the salt formation disposal technique becomes publicly
accepted, at least one other national disposal site will be in operation by
the year 2020 (Reference 7). -
Light Water Reactor
The fuel cycle of an LWR is schematically illustrated in Figure 6. The
cycle begins with the mining of uranium ore. The uranium is separated
from the ore by means of pulverizing, leaching, and calcining operations
in a uranium mill. The mill's product is called yellowcake after its color
and texture. The yellowcake consists of different oxides. Between 70
and 90 percent of the oxides contain uranium of varying stoichiometric
composition, but they are collectively written as 11303 (Reference 8).
The uranium in the yellowcake consists of a mixture of isotopes. The
composition of natural uranium is 99. 28 weight percent U-238, 0. 72
weight percent U-235 and a trace of U-234.
19
-------
MINE
URANIUM ORE
MILL
YELLOWCAKE
99. 18 % U-238
0.72 % U-235
UF6 PRODUCTION
PLANT
NATURAL UF6
0.72 % U-235
UNIRRADIATED FUEL CONTAINING
oLtlOHTLi
UNIRRADIATED FUEL CONTAINING
RECOVERED PLUTONIUM
FUEL FABRICATION
PLANT
LIGHT WATER
REACTOR
UO2 PRODUCTION
PLANT
RECOVERED
PLUTONIUM (Pu)
LOW LEVEL,
RADIOACTIVE
SOLID WASTE
SPENT FUEL
FISSION PRODUCT (NOBLE) GASES
CHEMICAL REPROCESSING
PLANT
SLIGHTLY
ENRICHED UFfc
2 TO 4% U-235
RECOVERED
URANIUM „
CONVERSION TO
UF6
HIGH LEVEL
RADIOACTIVE SOLID
WASTE
LOW LEVEL
RADIOACTIVE SOLID
WASTE: SOME INTER-
MEDIATE LEVEL
RADIOACTIVE SOLID
WASTE
RETRIEVABLE
SURFACE STORAGE
FACILITY
AND/OR
FEDERAL WASTE
REPOSITORY
COMMERCIAL
BURIAL GROUND
UF6 TAILS
0.20 % U-235
FIGURE 6. SCHEMATIC FUEL CYCLE DIAGRAM FOR LIGHT WATER REACTOR
-------
Yellowcake is transported to a refining complex of plants for purification
and conversion to uranium hexafluoride gas. The gas is then used as
feedstock in a gaseous diffusion plant. The depleted tails product (-0.2
percent U-235) is stored for future uses, such as future breeder cycle
fuel and as shielding material (Reference 9).
The enriched product (2 to 4 percent U-235) is transported to a fuel
preparation plant where it is transformed to uranium dioxide. The
dioxide powder is sintered and compressed into pellets. Rods of these
pellets are then assembled to form fuel elements. Some fabrication
plants combine the dioxide production and fuel element fabrication
operations.
At operating equilibrium, LWRs are charged annually with about a fourth
of a new core, which may utilize either U-235 or Pu-239 as the fissile
isotope. In those LWRs utilizing recovered plutonium, only about a third
of the new elements contain plutonium; the rest contain regular uranium
fuel that is slightly enriched in U-235. The plutonium elements are
mixtures of oxides of natural uranium and plutonium. Not all the
plutonium consists of fissile isotopes, but the fissile isotopes Pu-239
and Pu-241 usually account for 70 to 90 percent of the plutonium mass.
When the reactor is shut down for refueling, some of the spent fuel is
discharged. After a brief cooldown period to allow radioactive decay,
say 150 days for LWRs without Pu recycle and 90 days for LWRs with
Pu recycle, the spent'fuel is shipped to a chemical reprocessing plant.
At the reprocessing plant, metal clad and metal end boxes are separated
mechanically from the fuel. The fuel is dissolved in nitric acid followed
by several chemical processes designed to separate radioactive fission
products, uranium, plutonium, and other actinide elements.
The recovered uranium and plutonium are either recycled into new
fuel or are stored. The uranium product is usually converted to uranium
hexafluoride and shipped to the enrichment plant. The plutonium product
is shipped as PuO2 powder to a fuel preparation plant, where it is mixed
with powder of natural. UO2- The mixture is then sintered into fuel pellets,
The aqueous waste solutions found at the end of the separation process
contain some fission products and actinides and are classified by specific
radioactivity levels into three categories (Reference 10):
High: >lCi/£
Intermediate: 100 jiCi/jP - 1 Ci/S.
Low:
21
-------
In two of the three known reprocessing plants, the waste liquid is stored
in large tanks for up to five years to allow radioactive decay and then
solidified.. One of the plants produces a solid waste. The waste is
stored in solid form until it is shipped to a permanent storage site. By
regulation, all the waste must be shipped within ten years after it is
generated in the separation processes.
The permanent storage site has not been selected at this time. Plans to
construct a Retrievable Surface Storage Facility are being formulated
until such time as a suitable site is found for the Federal Waste Repository.
The cladding and other fuel element hulls are contaminated with a small
amount of fission products and actinides that have diffused from the fuel
pin matrix into the cladding matrix during irradiation. While these
materials present a radioactive hazard, they do not provide as large a
heat removal problem as does high-level radioactive solid waste.
Calculations (Reference 6) of the composition of mass, radiation, and
thermal power of irradiated Diablo Canyon PWR fuel assemblies yield
insight into the distinction between high and intermediate levels of radio-
active waste. After the fuel has decayed 150 days after discharge from
the reactor, the fission products and actinide elements in high-level
waste and activation products in the intermediate waste (assumed to be
the Zircaloy-4 cladding and Inconel spacers taken from the assemblies)
may be compared as follows:
Thermal
Radiation Power
Waste (MT) (MCi) (kw)
Fission Products in High-Level Waste 0.035 4.390 19.300
Actinide Elements in High-Level Waste 0.006 0.018 0.646
Activation Products in Intermediate- 0.271 0.028 0.224
Level Waste (Cladding)
From this comparison, the solid wastes may be classified qualitatively
on the basis of radioactivity and heat removal as shown on the following
page (Reference 11).
22
-------
Radioactivity . Heat
Lievel Containment Problem Removal Problem
High Substantial Substantial
Intermediate Significant Not serious
Low Not serious Not serious
The methods for disposal of intermediate level waste have not been fully
resolved. Currently, it is lumped together with low level waste and
transported from the chemical processing plants to commercial burial
grounds for permanent disposal in the earth. However, future quantities
of intermediate level waste may be disposed at the Federal Waste
Repository, along with high-level waste.
Low level waste includes not only the solid made from the liquid waste of
little radioactivity, but also sludges, resins, contaminated equipment,
clothing, packaging, and sundry items. All this material is transported
to commercial burial grounds for disposal.
In the course of processing, some of the fission products, isotopes of
iodine, xenon, and krypton, evolve as off-gases. Current practice is to
vent these gases to the atmosphere, where they are diluted to very small
concentrations. Future production levels of these gases may become
great enough to warrant the collection of these gases. Presumably, the
collected gas will be bottled and shipped to the Federal Waste Repository
for storage along with other fission products. As yet, no rules for these
actions have been formulated.
LWR Without Pu Recycle - A quantitative example of the material flows
in an LWR fuel cycle with no plutonium recycle is presented in Figure 7.
The annual flow rates are based on the Diablo Canyon PWR design.
Wastes from a BWR are expected to be comparable.
The amount of waste that must be transported from the chemical reprocess-
ing plant to the Federal Waste Repository is a function of the specific power
and burnup characteristics of the reactor. The fuel cycle pictured in
Figure 7 is based on a specific power of 37.5 Mwt/MTU with an 80 percent
load factor.
LWR With Pu Recycle - Recovery of plutonium generated in the irradiation
part of one LWR fuel cycle for use as fissile fuel in another cycle is
economically advantageous. With Pu recycle, about a third of the fuel
23
-------
MINE
27.350 U
3.2 % U-235
27. 350 HEAVY METAL
121.208 U
MILL
LIGHT WATER REACTOR
l.OOOMWe
3. 077 MWt
32, 873 MWtD/MTU
37. 5 MWt (FULL POWERt/MTU
(CHARGED)
0. 5 % LOSS
0. 609 U
26.137 U
0. 92 % U-235
0. 255 Pu
70. 66 % FISSILE
0. 817 FISSION PRODUCTS (FP)
0. 141 ACTINIDES (ACT)
121.208 U
UF6 PRODUCTION
PLANT
:120.599 U
0.72 % U
FLOW RATES IN MT/YR
AT 0. 80 REACTOR LOAD
FACTOR
RETRIEVABLE
SURFACE STORAGE
FACILITY
AND/OR
FEDERAL WASTE
REPOSITORY
0.254 Pu
70. 66 % FISSILE
124.468 U '
0.2 % U-235
FIGURE 7. MATERIAL FLOW IN TYPICAL LWR FUEL CYCLE WITHOUT
PLUTONIUM RECYCLE. (SOURCE: TABLE 2, ORNL-TM-4244. )
-------
elements in the annual charge to the reactor contain plutonium. Using
the steady state replacement of 1.21 atoms fissile plutonium for each
atom U-235 in 32.4 percent of the fuel elements (Reference 2), the
quantitative fuel cycle shown in Figure 8 results. The quantities of
fission products and actinides are based on an 80 percent load factor
and a specific power of 37.5 MWt/MT(U+Pu).
High Temperature Gas Cooled Reactor
The fuel cycle of the HTGR is similar to the LWR fuel cycle, except that
the fissile material is nearly fully enriched U-235 and the fertile material
is primarily thorium. Thus, there is the additional requirement for the
mining and shipment of thorium. A quantitative graphical representation
of the HTGR fuel cycle is given in Figure 9 (Reference 2).
More efficient fuel utilization in operation of the HTGR is possible if the
fissile material is U-233 instead of U-235 (Reference 12). However, the
U-233 will not be available until produced in HTGRs or breeder reactors
using thorium. One breeder well suited for this purpose is the Gas Cooled
Fast Reactor (GCFR) using modifications of HTGR technology. Since it
is not as well developed as the LMFBR, use of the GCFR in the nuclear
economy has not been anticipated in this forecast.
The fuel in the HTGR consists of'particles of uranium or thorium coated
with layers of pyrolytic carbon. The coatings contain the fission products
as they form during irradiation. Such a ceramic fuel leads to a different
set of techniques for separation of recoverable fuel and fission products
than those used for LWR oxide fuel. The HTGR fuel is crushed and
burned to free the fuel from the graphite. The U-235 recycle particles
are separated by mechanical screening. Chemical processes are then
applied to the other fuel particles to separate the fission products and
actinides in the fuel.
Liquid Metal Cooled Fast Breeder Reactor
A quantitative representation of the fuel cycle for LMFBR is illustrated
in Figure 10. The annual flow rates are based on the Atomics International
Follow-on Design (Reference 2). The reactor is composed of two regions:
a core containing fissile nuclei and a blanket containing fertile nuclei. The
core fuel elements contain about 15 weight percent fissile isotopes of
plutonium. The blanket utilizes depleted uranium tails left from the
enrichment process.
25
-------
SLIGHTLY ENRICHED U ELEMENTS (67.6 % OF CHARGE)
NAT U + Pu ELEMENTS (32. 4 % OF CHARGE)
0. 2 % LOSS
to
8.451 U
0.442 Pu
0.3% LOSS
8.409 U
0.72 % U-235
0.441 Pu
61.30 % FISSILE
18. 500 U
3. 2 % U-235
LIGHT WATER REACTOR
1,000 MWc
3.077 MWt
32. 873 MWtD/MT (U + Pu)
37. 5 MWt (FULL POWER)/MT
(U + Pu) (CHARGED)
89.619 U
0.2 % U-235
8. 190 U
0.32 % U-235
0. 273 Pu
55. 36% FISSILE
0.273 FP
0. 114 ACT
17.679 U
0.93 % U-235
0.172 Pu
70. 65 % FISSILE
0. 553 FP
0. 096 ACT
CHEMICAL
REPROCESSING
PLANT
0.826 FP
0.210 ACT
25. 740 U
0.74 % U-235
0, 3 % LOSS
10.077 U
26.909 U
2. 42% U-235
0.441 Pu
61.30 % FISSILE
27. 350 HEAVY METAL
25.869 U
0.74 % U-235
0. 446 Pu
61. 30 % FISSILE
0.826 FP
0.210 ACT
27. 350 HEAVY METAL
RETRIEVABLE
SURFACE STORAGE
FACILITY AND/OR
FEDERAL WASTE
REPOSITORY
CONVERSION
TO UF6
FLOW RATES IN MT/YR
AT 0.80 REACTOR
LOAD FACTOR.
FIGURE 8. MATERIAL FLOW IN TYPICAL LWR FUEL CYCLE WITH
PLUTONIUM RECYCLE. (SOURCE: TABLE 2, ORNL-TM-4244. )
-------
FERTILE Th, U-233 MAKEUP
FISSILE U-235 FRESH MAKEUP
FISSILE U-235 RECYCLED MAKEUP
0. 104 U
29. 09% U-235
0.403 U
92.56 % U-23S
0. 358 U
8.27 % U-235
0.217 U-233
8. 434 Th
HTGR
1, 160 MWe
3. 000 MWt
94. 264 MWtO/MT (U + Th)
80. 65 MWt (FULL POWER)/
MT(U + Th)
(CHARGED)
0.070 U
3.71 % U-235
0. 10S U
29.25% U-235
0. 366 U
8.38 % U-235
0.219 U-233
7.819 Th
1 % LOSS
CHEMICAL REPROCESSING PLANT
0.046 U
FISSILE
0. 103 U
FERTILE
0. 362 U
FLOW RATES IN MT/YR AT
0. 80 REACTOR LOAD FACTOR.
0.865 U
50.03 % U-235
0.217 U-233
8.434 Th
9. 516 HEAVY METAL
0.541 U
11.84 % U-235
0.219 U-233
7.819 Th
0.010 Pu
21.00 % FISSILE
0. 907 FP
0. 020 ACT
9. 516 HEAVY METAL
RETRIEVABLE
SURFACE STORAGE
' FACILITY
AND/OR
FEDERAL WASTE
REPOSITORY
FIGURE 9. MATERIAL FLOW IN TYPICAL HTGR FUEL CYCLE.
(SOURCE: TABLE 2, ORNL-TM-4244.)
-------
co
0. 5 % LOSS
0.037 U
0.009-Pu
CORE FABRICATION
5%
RECYCLE
8. 349 U
1.760 Pu
0.422 U
0. 088 Pu
0. 5 % LOSS
0.043 U
0.009 Pu
CORE FUEL PREPARATION
CORE
7.890 U
0.20% U-235
1.663 Pu
'71..92 % FISSILE
AXIAL
6.571 U
0.20 % U-235
RADIAL
2.702 U
0. 18 % U-235
17. 163 U
0.20 %
1.663 Pu
71. 92 % FISSILE
18. 826 HEAVY METAL
LMFBR
CORE
AXIAL
RADIAL
MWt
2,219
107
74
1,002 MWe
2,400 MWt
MWtD/T
67,594
4,739
7,970
MWt/T T
116.1 19.1
8.1 13.2
4.7 15.7
TOTAL 2,400 37,098
50.18 48.0
7.255 U
0.11 % U-235
1.655 Pu
69. 06 % FISSILE
6.415 U
0.17 % U-235
0.137 Pu
97. 08 % FISSILE
2. 543 U
0. 12 % U-235
0.126 Pu
94.44 % FISSILE
18.030 DEPLETED UFfc
0.20 % U-235 (ASSUMED AVAILABLE)
FLOW RATES IN MT/YR AT 0. 80 REACTOR LOAD FACTOR.
16.213 U
0.14 % U-235
1.918 Pu
72. 73 % FISSILE
0.679 FP
0.016 ACT
18.826 HEAVY METAL
RETRIEVABLE
SURFACE STORAGE
FACILITY
AND/OR.
FEDERAL WASTE
REPOSITORY
FIGURE 10.
MATERIAL FLOW IN TYPICAL LMFBR FUEL CYCLE.
(SOURCE: TABLE 3, ORNL-TM-4244. )
-------
The transportation services required in the LMFBR fuel cycle include
shipping depleted UF^ from enrichment plant to the fuel preparation
plant as well as recycling uranium and plutonium recovered from fuel
preparation and fuel fabrication plants. The uranium from chemical
processing plants is shipped to waste storage.
Separation techniques for LMFBR fuel are still being developed. The
process is similar to that for LWR fuel in that the fuel elements are
mechanically disassembled. Volatile fission products are then removed
by heating. The oxide fuel is then separated from the stainless steel
cladding by dissolution in nitric acid. Chemical separation processes
are then employed to separate the plutonium fuel from the fission products,
other actinides,' and alpha-contaminated wastes.
29
-------
SECTION V
NUCLEAR TRANSPORTATION FORECASTS (1975 to 2020)
ANNUAL FUEL REQUIREMENTS
The nuclear power economy forecast in Table 2 involves large require-
ments in fuel fabrication from plutonium produced in nuclear reactors
and from uranium and thorium mined from the earth. In 1979, some
LWRs are supposed to begin using recycled plutonium in their fuel ele-
ments. Consequently, fabrication of these elements is scheduled for 1978.
Introduction of LMFBRs is forecast for 1987. Recycle of plutonium in
LWR fuel is presumed to stop then. Use of LMFBRs is expected to
increase rapidly after 1990; half or more of all fuel fabricated after 2005
will be LMFBR fuel. Use of LWRs is anticipated to increase until year
2000 and then to decrease at a slow rate. Some of the plutonium bred
in the LWRs will be used in LWRs, but most of it will sustain the LMFBRs.
The commercial operation of HTGRs is expected to flourish by 1980 and
continually increase through the balance of the 50-year period.
The forecast for the amounts of these fuels that will be used annually in
the fabrication of fuel is given in Table 3 and Figure 11. The total amount
of fuel fabricated is expected to double every 5 years until about 1985.
The LWRs using plutonium recycle are charged with two types of fuel:
one in which the fuel pins contain natural uranium and plutonium and one
in which the fuel pins contain slightly enriched uranium only. In reactors
irradiating recycled plutonium, only about one-third of the fuel elements
actually contain plutonium. The plutonium constitutes about 3 weight
percent of these fuel elements.
ANNUAL SPENT FUEL TRANSPORTATION
The estimated annual discharge of spent fuel from power reactors for the
next 50 years is presented in Table 4 and Figure 12. These numbers were
calculated under the assumption of an 80 percent load factor. They provide
the annual quantities of spent fuel that must be transported from the
reactors to the processing plants with cooling times taken into account.
Values for the cooling times and burnup values have been assumed as
follows:
Reactor Type Cooling Time (days) Burnup (Gwtd/MT)
LWR 150 33
HTGR 365 94
LMFBR '90 37
30
-------
TABLE 3. ANNUAL PRODUCTION OF FABRICATED FUEL
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Amounts of Component Fuel Materials (10 MT Heavy Metal)
LWR
No Plutonium
Recycle -Slightly
Enriched Uranium
1.57
2.99
6.01
10.14
11.76
11.42
9.75
7.80
5.80
3.80
Plutonium Recycle
Natural
Uranium
0.76
0.51
Plutonium
0.04
0.03
Subtotal
1.57
3.79
6.55
10.14
11.76
11.42
9.75
7.80
5.80
3.80
HTGR
Uranium
0.01
0.07
0.13
0.16
0.21
0.26
0.30
0.32
0.35
Thorium
0.11
0. 54
0.96
1.27
1.64
2.03
2.30
2.48
2.75
Subtotal
0.12
0.61
1.09
1.43
1.85
2.29
2.60
2.80
3.10
LMFBR
Uranium
0.54
2.91
7.14
13.20
19.44
25.54
31.65
Plutonium
0.04
0.27
0.68
1.27
1.87
2.47
3.06
Subtotal
0.58
3.18
7.82
14.47
21.31
28.01
34.71
Total
1.57
3.91
7.16
11.81
16.37
21.09
26.51
31.71
36.61
41.61
aData derived from ORNL-TM-4224. Data for 2010-2020 are extrapolations.
-------
36
34
32
30
28
26
24
22
CO
o
~ 20
Q
W
5 18
u
KH
« . 16
W 14
D
12
10
.2
TOTAL,
\
/
FORECAST
EXTRAPOLATION
1970
1980
1990
2000
2010
2020
FIGURE 11. PROJECTION OF ANNUAL NUCLEAR FUEL
FABRICATION REQUIREMENTS
32
-------
H
2
Q
W
OT
«
W
u
o
u
ID
fn
H
2
W
FORECAST
EXTRAPOLATION
-W
1970
1980
1990
2000
2010
2020
FIGURE 12. PROJECTION OF ANNUAL NUCLEAR FUEL
REPROCESSING REQUIREMENTS
33
-------
TABLE 4. ANNUAL DISCHARGE OF SPENT FUEL AFTER COOLING
Amount of Spent Fuel (10 MT Heavy Metal)3"
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
LWR
LWR
Without
Plutonium
Recycle
0.85
2. 12
3.57
7.28
10.48
11.04
10.64
9.30
7. 50
6.00
LWR
With
Plutonium
Recycle
0. 27
1.18
0. 37
Subtotal
0. 85
2.39
4.75
7.65
10.48
11.04
10. 64
9. 30
7.50
6. 00
HTGR
0.01
0.17
0. 59
1.04
1.42
1.81
2.20
2.60
3.00
LMFBR
0. 18
1.47
5. 13
10.68
15. 60
20.80
26. 20
Total
0.85
2.40
4.92
8.42
12.99
17.58
23. 13
27. 10
30. 90
35. 20
aORNL-TM-4224. Data for 2010-2020 are extrapolations.
The spent fuel contains varying amounts of uranium, plutonium, other
actinides, fission products, and thorium, depending on the type of reactor.
The assays of the different types of fuel at time of processing are derived
from data in Figures 7 to 10 and are collected in Table 5.
The radioactivities of major fission product nuclides, actinide nuclides,
and activation product nuclides present in the spent fuel of each reactor
type at the time of processing are listed in Table 6. Also shown in
Table 6, is the radioactivity in waste after 10 years of storage. The
totals were applied to the data in Table 4 to obtain the amounts of radio-
activity transported each year in spent fuel shipments. The results of
this calculation are posted in Table 7.
34
-------
TABLE 5. ASSAY OF NUCLEAR FUELS AT TIME OF REPROCESSING
OJ
en
Fuel Component
Uranium
U-235
U-233
U-238
Plutonium
Fissile
Nonfissile
Thorium
Fission products
Actinides other
than U or Pu
TOTAL
Weight Percent in Different Fuels
LWR
Not
Containing
Recycled Pu
95.56
0.92
99.08
100..00
0. 93
70. 66
29. 34
100. 00
-
2.99
0. 52
100. 00
Containing
Recycled Pu
92. 54
0. 32
99.68
100. 00
3.08
55. 36
44. 64
100. 00
-
3. 08
1. 30
100.00
HTGR
7.99
11.84
28.82
59. 34
100. 00
0. 10
-
82. 17
9.53
0.21
100. 00
LMFBR
86. 12
0. 14
99. 86
100. 00
10. 19
72.73
27. 27
100. 00
-
3. 60
0.09
100. 00
-------
TABLE 6. TYPICAL RADIOACTIVITY IN FUEL AND WASTES AT FUEL
REPROCESSING PLANTSa
Nuclide
H-3
C-14
Kr-85
Rb-86
Sr-89
Sr-90
Y-90
Y-91
Zr-93
Nb-93m
Zr-95
Nb-95
Tc-99
Ru-103
Rh-103m
Ru-106
Rh-106
Ag-llOm
Ag-110
Cd-113m
Cd-115m
Sn-123
Sb-124
Sb-125
Te-125m
Te-127m
Te-127
Te-129m
Te-129
1-129
1-131
Xe-131m
Xe-133
Ca-134
Cs-135
Cs-136
Cs-137
Ba-137m
Ba-140
La- 140
Concentration (Curiea/MT Heavy Metal Charged to Reactor)
LWR-U
Fuel
692
11.000
1.90
97,200
76,900
76,900
161.000
1.89
0.181
277,000
520,000
14.3
88; 200
88,200
410,000
410,000
2,440
317
10.3
49.1
3,860
71.8
7,950
3,200
6,150
6,080
2,710
1.740
0.0374
2.18
3.19
214,000
0.286
20.5
107,000
99,900
431
496
Waste
5.22
67,900
67,900
1.89
0. 564
14.3
13,000
13,000
16.4
2.13
8.03
0.154
2,200
913
0.055
0.055
0.00005
39, 500
0.286
95,100
89,000
LWR-Pu
Fuel
908
6,850
0.654
67,700
45,400
45,400
119,000
1.46
0.135
255,000
478,000
14.5
99,200
99,300
682,000
682,000
5,080
660
21.6
61.3
4.940
111
13,100
5,320
7.700
7.610
3,000
1,920
0.0480
2.27
3.30
187,000
0.524
29.3
110.000
103,000
409
470
Waate
6.85
40,100
40, 100
1.46
0.432
14.5
21,700
21,700
34.1
4.43
16.8
0.197-
3, 640
1.510
0.070
0.070
0.00006
34, 400
0.524
97,800
91,500
HTGR
Fuel
4.040
38.6
58,800
0. 0075
22,600
284, 000
284, 000
40, 600
6.56
0.924
66,100
140, 000
33.7
2.030
2.030
95, 000
95, 000
447
58.1
12.9
2.13
2,680
20.9
17,000
7,030
5,680
5,610
89.0
57.0
0.125
564, 000
0.678
0.00
298, 000
279, 000
0.0078
0.0090
Waate
30.5
251,000
251,000
6.56
2.19
33.7
3,020
3,020
3.00
0.39
10.0
0.107
4,710
1,950
0.051
0.051
0.0001
104, 000
0.678
266,000
248,000
LMFBR
Fuel
58
9,400
229
252,000
53,200
53,200
416,000
1.76
0.122
843, 000
1,390.000
18.2
511,000
511,000
1,190,000
1,190,000
1,250
162
172
584
11,100
723
26,200
10,200
19,800
19,600
11,200
7, 190
0.0429
604
160
20.7
43, 300
1.40
1,370
141,000
132,000
16,300
18,800
Waste
47,000
47, 000
1.76
0.49
18.2
37,900
37,900
8.36
1.09
134
0.443
7,260
3,010
0. 179
0.177
0.0001
7,980
1.40
126,000
118,000
36
-------
TABLE 6. (Continued)
Nuclide
Ce-141
Pr-143
Ce-144
Pr-144
Nd-147
Pm-147
Pm-148m
Sm-151
Eu-152
Eu-154
Eu-155
Tb-160
Th-228
Pa-231
Pa-233
U-232
U-233
U-234
Np-237
Np-239
Pu-238
Pu-239
Pvi-240
Pu-241
Pu-242
Am-241
Am-243
Cm-242
Cm-244
TOTAL,
Concentration (Curies /MT Heavy Metal Charged to Reactor)
LWR-U
Fuel
56,400
679
771,000
771,000
50.3
98,000
3,270
1,250
12.2
6,870
6,400
303
0.34
0.754
0.34
18.0
2,820
323
475
102,000
1.37
153
18.0
17,700
2,390
4. 5xl06
Waste
8,930
8,930
26,100
1,200
9.16
5,530
943
0.34
0.34
18.0
99.5
1.62
3.48
404
0.007
156
18.0
14.8
1,970
0.3xl06
LWR-Pu
Fuel
52, 000
625
658,000
658,000
46.7
105,000
4.440
1,640
35.0
9,270
9,380
538
0.085
0.003
0.324
0. 0848
514
18,900
735
1,940
607,000
16.2
1,580
514
240, 000
136,000
5. 5xl06
Waste
7,630
7,630
27,900
1,570
26.2
7,460
1,380
0.087
0.017
0.087
514
1,260
3.75
72.9
2,400
0.081
1.590
514
284
112,000
0. 5xl06
HTGR
Fuel
1,500
0.033
1,050,000
1,050,000
146, 000
118
698
2.98
13, 100
9,240
21.5
196
0.818
4,500
290
221
61.9
1.57
7.26
18, 700
15.0
31.9
10,400
0.413
27.8
7.26
875
1,600
4.6xl06
Waste
12,200
12,200
39,000
671
2.23
10,600
1,360
1.38
0.818
1.57
1.42
1.11
1.04
1.57
7.26
18,000
15.0
32.6
8,200
0.413
101
7.26
0.739
1,320
1.2xl06
LMFBR
Fuel
265,000
21,900
1,040,000
1,040,000
3,330
350,000
17,200
6,100
26.8
2,050
56,000
1,740
0.016
0.172
0.043
0.085
0.173
56.9
K.900
3,270
3,860
346, 000
10.0
1,450
56.9
52,800
2.390
10. IxlO6
Waste
12,100
12,100
93,400
5,860
20.1
1,650
8,260
0.0029
0.175
0.004
0.175
56.8
312
16.4
20.4
1,370
0.050
1,450
56.8
111
1,970
0.6xl06
; Based on calculations in ORNL-TM-4224 and ORNL-TM-3965. Fuel at time of reprocessing.
i Waste after 10 years storage.
37
-------
TABLE 7 . RADIOACTIVITY OF
ANNUALLY TRANSPORTED SPENT FUEL
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Radioactivity of Spent Fuel (103 MCi)
LWR
'Without
Pu
Recycle
3.82
9.54
16.06
32.76
47. 16
49.68
47. 88
41.85
33.75
27.00
LWR
With
Pu
Recycle
1.48
6.49
2.04
Subtotal
3.82
11.02
22. 55
34.80
47.16
49.68
47.88
41.85
33.75
27.00
HTGR
0.05
0.78
2.71
4.78
6.53
8. 33
10. 12
11. 96
13.80
LMFBR
1.82
ji
14.85
51.81
107.87
157.56
210.08
264.62
Total
3.82
11.07
23.33
37.51
51.94
56.21
164.08
209.53
255.79
305.42
Spent fuel may be transported in various sizes of shipment. The several
states impose restrictions on weights of shipments made by motor freight.
The legal weight limit in most states is about 33 MT (73, 000 Ib) gross
vehicle weight. Such a shipment can accommodate a spent fuel shipping
cask that weighs about 22 MT (48, 000 Ib) when unloaded and contains
0. 5 MTU equivalent of spent fuel. Special permits are issued in some
states to admit truck shipments of heavier gross vehicle weights up to
71 MT (156, 500 Ib), although most states issuing such permits post a
50-MT (110,000 Ib) limit (Reference 13). A compendium of information
on possible spent fuel shipment sizes is included in Table 8.
The most popular mode will probably be rail shipments utilizing the
GE IF 300 cask since it is the largest of the currently licensed containers.
This cask may hold 7 PWR elements or 18 BWR elements, or equivalently
a spent fuel mass of 3. 15 MTU. A larger cask is being designed that will
38
-------
TABLE 8. SPENT FUEL SHIPMENT CAPACITIES
Designation
Mode
Size Data
GE IF100B129a
DOT SP 5369
Truck, rail, or barge
GE IF300b
Inter modal: overweight
truck + rail or barge
GE IF400b
Weatinghousea
Yankee cask
Rail or barge
Rail
Cask: 1.1 m dia x 3. 7 m long
Cavity: 0. 3 m dia x 3. 3 m long
Shield: 0.2 m Pb 0. 04 m ateel
Contents: 0.45 MTU
Number of elements: 4 BWR or 1 PWR
Tare weight: 20 MT (44, 000 Ib)
Number of casks per shipment: 1
Cask: 1.6 m dia x 5. 3 m long
Cavity: 1.0m dia x 4. 6 m long
Shield: depleted uranium, stainless steel, water
Contents: 3.15 MTU
Number of elements: 18 BWR or 7 PWR
Number of casks per shipment: 1
Cask: under development
Contents: more than 6 MTU
Number of elements: 32 BWR or 15 PWR
Number of casks per shipment: 1
Cask: 1. 6 m dia x 3. 9 m long
Cavity: 1.0m dia x 3. 0 m long
Shield: 0.2mPb
Weight: 68. 0 MT loaded, 62. 6 MT empty
Number of casks per shipment: 1
aDirectory of Shipping Containers for Radioactive Materials. United States Atomic Energy Commission,
Washington, D. C. October 1969.
Transportation of Nuclear Fuel. Report by Southern Interstate Nuclear Board. Atlanta, Georgia. December 1972.
-------
hold 15 PWR elements or 32 BWR elements, or equivalently a spent
fuel mass of 6. 75 MTU.
Smaller shipments by truck will still be used to serve those reactors
without access to rail transportation and those reactors close to the
chemical processing plants. Two truck shipment sizes are considered
here: a legal weight shipment of 0.45 MTU capacity and an overweight
shipment of 0. 90 MTU capacity.
Some shipments by barge are anticipated. These shipments will involve
a modal interchange since not all reactors or chemical processing plants
will have access to water transportation. If barge shipments are made,
they will probably be used only in conventional shipping lanes such as the
Atlantic and Pacific seacoasts, the large river systems (e.g., Mississippi,
Missouri, Ohio, and Hudson), and the Great Lakes.
The transportation scenario for spent fuel adopted for this study is shown
in Table 9. Using this scenario, the annual number of spent fuel ship-
ments was calculated from the data in Table 4. The results are shown
in Table 12. , .
TABLE 9. SPENT FUEL TRANSPORTATION SCENARIO
Mode
Legal Weight Truck
Overweight Truck
Small Cask on Rail
Large Cask on Rail
Barge and Overweight Truck
AVERAGE
Use
(Percent)
5
5
70
15
5
100
Capacity
(MTU Equivalent)
0.45
0.90
3.15
6.75
3.15
2.34
Data on the distances between nuclear power reactors operating in 1970
and the three known sites of chemical processing plants are presented
for truck and rail transport modes in Table 10. The distance over which
spent fuel shipments must travel varies strongly with the location of the
40
-------
TABLE 10. DISTANCES BETWEEN NUCLEAR POWER REACTORS
OPERATING IN 1970 AND KNOWN CHEMICAL PROCESS PLANT SITES
Reactor Site
Shippingport, Pennsylvania
Buchanan, New York
Rowe, Massachusetts
Haddam Neck, Connecticut
Toms River, New Jersey
Scriba, New York
Waterford, Connecticut
Peach Bottom, Pennsylvania
Big Rock Point, Michigan
Lagoona Beach, Michigan
Harts ville, South Carolina
Two Creeks, Wisconsin
Morris, Illinois
Genoa, Wisconsin
San Clemente, California
Humboldt Bay, California
Richland, Washington
AVERAGE
Chemical Processing Plant Sites
West Valley,
New York
Distance
by Truck
(Mi)a
217
450
322
450
400
120
450
365
640
354
853
748
570
838
2,646
2,862
2,477
868
Distance
by Rail
(Mi)b
217
433
426
480
424
154
575
346
647
345
948
864
608
870
2,750
3, 100
2,569
927
Barnwell,
South Carolina
Distance
by Truck
(Mi)a
593
932
1,090
962
858
1,064
962
601
1,080
750
113
1,129
900
1,193
2,452
3, 018
2,852
1,209
Distance
by Rail
(Mi)b
894
810
1,002
862
731
1, 127
997
697
1, 121
844
117
1,129
978
1,286
2,559
3,202
2,911
1,251
Morris,
Illinois
Distance
by Truck
(Mi)a
501
923
885
953
842
719
953
762
459
315
1,033
259
0
323
2,095
2,429
1,874
901
Distance
by Rail
(Mi)b
519
979
1,018
1, 108
900
707
1,031
770
450
292
989
294
0
308
2,077
2,476
2,049
874
aRand McNally Road Atlas : United States, Canada, and Mexico, 43rd Annual Edition,
Rand McNally & Company (1967).
bRand McNally Handy Railroad Atlas of the United States, Rand h
IcNally & Company (1971).
-------
reactors. If a shipment of spent fuel from any of these reactors were to
be randomly sent to either of the processing plants, it must traverse an
average distance of 993 miles by truck or 1017 miles by rail.
If additional chemical processing plants were built in the vicinities of
Salt Lake City, Utah; Dallas, Texas; Cincinnati, Ohio; and St. Paul,
Minnesota, and nuclear power plants were constructed fairly uniformly
throughout the country, the average spent fuel shipment distance would
decrease from the 1970 value of about 1000 miles to about 400 miles.
The future behavior of this parameter is uncertain. It may stay rather
constant if future reactor construction is centered around a few process-
ing plants, or it may decrease nearly to zero if the construction of
nuclear parks becomes the vogue. For purposes of this study, the
average shipment distance is assumed to decrease linearly from 1000
miles at the start of 1970 to 400 miles at the end of 2020. The distinction
between truck and rail distances is negligible for 1970 and no distinction
between these modes for the future is anticipated.
To ascertain a value for the shipment distance by barge transport, the
average is taken of such voyages as those from northeast ports or Florida
ports to Steel Landing near the Barnwell, South Carolina, chemical
processing plant down the Atlantic seaboard, applicable Great Lakes
routes, applicable Mississippi and Ohio River routes, and possible
shipments from the Southern California or Puget Sound areas to
San Francisco port. These data are listed in Table 11. This distance
parameter is assumed to remain constant at 565 miles over the fifty
year period under investigation, since the location of the ports that will
be used for modal interchanges between truck and barge or rail and barge
is expected to be insensitive to the location of reactors.
The average shipment distance with respect to all modes is obtained by
means of the modal use spectrum postulated in Table 9 is shown in
Table 12.
By way of summary, the annual shipping data for spent fuel are collected
in Table 12. The annual shipping units, which are measured in units of
a million shipment-miles, grow until 2005, increasing about one million
shipment-miles every five years. After 2005, the annual shipping units
remains essentially constant at about 6 million shipment-miles per year.
ANNUAL PLUTONIUM TRANSPORTATION
The annual amounts of plutonium appearing in the chemical reprocessing
plants was calculated by combining the data in Table 4 with the Pu data
from Table 5. The results are tabulated in Table 13.
42
-------
TABLE 11. DISTANCES BETWEEN SEA PORTS
FOR SPENT FUEL SHIPMENTS
-
Origin Port
New York Harbor, New York
Miami, Florida
Portland, Maine
Cleveland, Ohio
Green Bay, Wisconsin
New Orleans, Louisiana
St. Louis, Missouri
Los Angeles, California
Seattle, Washington
Destination Port
Steel Landing, South Carolina
Steel Landing, South Carolina
Philadelphia, Pennsylvania
Buffalo, New York
Chicago, Illinois
Cincinnati, Ohio
St. Paul, Minnesota
San Francisco, California
San Francisco, California
AVERAGE
Distance
(Mi)a
890
460
550
180
255
1,000
575
370
810
565
aBarge distance data computed from map of United States of America,
Federal-Aid Highways, U. S. Department of Transportation,
Washington, D. C. (1970), and The World Almanac Book of Facts,
Newspaper Enterprise Association, Inc. , New York (1974).
TABLE 12. SUMMARY OF ANNUAL SPENT FUEL SHIPPING DATA
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Mass
Transported
(MT)
850
2,400
4,920
8,420
12,990
17, 580
23, 130
27, 100
30, 900
35,200
Radioactivity
Transported
(MCi)
3,820
11, 070
23, 330
37, 510
51,940
56,210
164,080
209, 530
255,790
305,420
Number of
Shipments
363
1,026
2,103
3,598
5,551
7,513
9,885
11,581
13,205
15,043
Average
Shipment
Distance
(Mi)
920
860
810
750
690
640
580
520
460
410
Shipping
Units
(106 Shipment-
Mi)
0. 33
0.88
1.70
2.70
3.83
4.80
5. 73
6. 02
6.07
6.16
43
-------
TABLE 13.
ANNUAL PLUTONIUM GENERATION IN CHEMICAL
REPROCESSING PLANTS
-
Year
1975
1980
1985
1990
1995
2000
2005
.2010
2015
2020
Amount of Plutonium (MT Pu)
LWR
LWR
Without
Pu
Re cycle
7.90
19.72
33.20
67.70
97.46
102. 67
98.95
86.49
69.75
55.80
LWR
With
Pu
Recycle
4.40
19.23
6.03
Subtotal
7.90
24. 12
52.43
73.73
97.46
102.67
98. 95
86.49
69.75
55.80
HTGR
0.01
0. 17
0. 59
1. 04
1.42
1.81
2. 20
2. 60
3.00
LMFBR
18.34
149.79
522.75
1088.29
1589:64
2119.52
2669.78
Total
8
24
53
93
248
627.
1189
1678
2192
2729
-------
The plutonium separated in chemical reprocessing plants will eventually
be shipped to fuel fabrication plants for the production of fuel for LWR
with plutonium recycle or LMFBR. However, according to the scenario
adopted for this study as shown in Table 3, the production of plutonium
fuel is not expected to begin until 1978 for LWR and until 1988 for LMFBR.
Excess plutonium will be stored~until required for use. For the purposes
of this study, it is assumed that storage will be at the fuel fabrication
plant. All transportation between the chemical reprocessing plant and
the fuel fabrication plant is assumed to take place during the same year
that the plutonium is generated in the chemical reprocessing plant.
Therefore, the quantity of plutonium transported each year is given by
the data in Table 13.
The specific activity of the plutonium transported can be calculated from
the plutonium data in Table 6 and Table 5. The results have been tabulated
in Table 14. ""^ "
The specific activity of plutonium was applied to the generation data
in Table 13 to obtain the amounts of radioactivity transported each year
in plutonium shipments. The results of this calculation are tabulated in
Table 15.
The AEC (10 CFR71C) and DOT (49 CFR173. 393) regulate shipments of
fissile fuel to prevent the possible attainment of critical mass or loss of
containment by controlling the mass and geometry of fissile material
allowed in individual packages and the number of packages allowed in a
shipment. In the nuclear fuel cycle, these regulations pertain to move-
ments of enriched uranium, unirradiated fabricated fuel, spent fuel, and
recycled plutonium.
Plutonium may be shipped in several container sizes, depending on its
isotopic assay, and in several forms: single isotopes or mixtures of
isotopes in powder or bulk solid form, as plutonium metal or in dry
compounds, usually oxides, of plutonium. Powdered plutonium oxide
will probably be the most frequently shipped form since this form is
readily incorporated into fuel. Some technical data on possible shipment
sizes are included in Table 16. For purposes of this study, two of these
sizes are picked for each of truck and rail modes of transportation: a
small shipment of 17 packages of the DOT SP 5332 design, with each
package containing 4 kgm of plutonium metal, and a large shipment of 133
packages of the DOT SP 5795 design, with each package containing "
15.2 kgm of powder composed of atleast 60 weight percent Pu-239, less
than 1. 5 weight percent Pu-241 (a beta emitter), more Pu-240 than
Pu-241, and less than 1 weight percent Pu-238.
45
-------
TABLE 14. SPECIFIC ACTIVITY OF PLUTONIUM
Reactor
LWR-U
LWR-Pu
HTGR
LMFBR
Concentration
(MCi Pu/MT
Heavy Metal)
0.1056
0.6286
0.0292
0.3640
Assay
(Weight
Percent Pu)
0.93
3.08a
0.10
10. 19
Specific Activity
(MCi Pu/MT Pu)
11.4
14.3b
29.2
3.6
a32.4% of core; balance is LWR-U.
^Average.
TABLE 15.. RADIOACTIVITY OF
ANNUALLY "TRANSPORTED PLUTONIUM'
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Radioactivity of Plutonium (103 MCi)
LWR-U
0.09
0.22
0.38
0.77
1. 11
1.17
1. 13
0.99
0.80
0.64
LWR-Pu
0.06
0.27
0.09
LWR
0.09
0.28
0.65
0.86
1.11
1.17
1.13
0.99
0.80
0.64
HTGR
.0.02
0.03
0.04
0.05
0.06
0.08
0.09
LMFBR
0.07
0.54
1.88
3.92
5.72
7.63
9.61
Total
0.09
0.28
0.65
0.95
1.68
3.09
5. 10
6.77
8.51
10.34
46
-------
TABLE 16. PLUTONIUM SHIPMENT CAPACITIES
Material
Designation
Mode
Size Data
Plutonium shipped
from chemical
processing plant to
fuel preparation
plant
Metal
Single or mixed
plutonium isotopes
as metal or oxide:
weight composition:
260% Pu-239
SI. 5% Pu-241
Pu-240 > Pu-241
Pu-238
Model 2030-la-b
DOT SP 5332
Truck or rail
Foamglasc»"
shipping container
DOT SP 5795
Truck or rail
•Drum: 0.5m dia x 0. 8 m high
Cavity: 25 cm dia x 18 cm high
Number of cavities: 2
Shield: Celotex (thermal)
Contents of single drum: 4 kgm total in
both cavities
Number of drums per shipment: 17 (fissile
class II), 42 (fissile class III), (other
sizes of shipment are also possible. )
Drum: 0.6m dia x 0. 9 m high
Cavity: 12 cm dia x 46 cm high
Shield: 0. 2 m Foamglas (thermal)
Contents of single drum:
metal: 0.0069 MT Pu-239; 0.0115 MT Pu
oxide: 0.0091 MT Pu-239; 0.0152 MT Pu
Number of drums per shipment: 133 (other
sizes of shipment are also possible. )
aDirectory of Shipping Containers for Radioactive Materials. United States Atomic Energy Commission.
Washington, D. C. October 1969. p. IA19. Iff.
bTechnical Documentation for Model 2030-1 Shipping Container. RFP-1867. Dow Chemical U.S.A.,
Rocky Flats Division, Golden, Colorado. November 1972.
cReference a, p. IA33. Iff.
dSpecial Tests for Plutonium Shipping Containers 6M, SP5795, and L-10. SC-DR-720597. Sandia
Laboratories, Albuquerque, New Mexico. September 1972.
-------
Most of the smaller shipments of plutonium will probably go by truck,
although a string of rail cars containing small shipments may be both
economical and convenient on occasion. The larger shipments are
expected to be delivered by rail in the majority of instances. The
plutonium transportation scenario for purposes of this study is shown in
Table 17. Using this scenario, the annual number of plutonium shipments
was calculated from the data in Table 14. The results are shown in
Table 19.
TABLE 17. PLUTONIUM TRANSPORTATION SCENARIO
Mode
Small Shipment by Truck
Large Shipment by Truck
Small Shipment by Rail
Large Shipment by Rail
AVERAGE
Use
(Percent)
40
10
10
40
100
Capacity
(MT Pu).
0.068
2.022
0.068
2.022
0. 132
The shipping distance for recycled plutonium refers to the trip between
chemical processing plants, where plutonium is separated from other
components of spent fuel, and fuel preparation plants, where it is mixed
with uranium to fabricate new fuel. The distances between these types
of facilities existing in 1973 are given for truck and rail modes in
Table 18.
If additional chemical processing plants are built in the vicinities of
Salt Lake City, Utah; Dallas, Texas; Cincinnati, Ohio; and St. Paul,
Minnesota, and fuel fabrication plants were constructed fairly uniformly
throughout the country, the average plutonium shipment distance would
decrease from the 1970 value of about 1, 000 miles to about 400 miles. ~
The future behavior of this parameter is uncertain. It may stay rather
constant if future fuel fabrication plant construction is centered around
a few processing plants, or it may decrease nearly to zero if the con-
struction of nuclear parks becomes the vogue. For purposes of this
study, the average shipment distance is assumed to decrease linearly
from 1, 000 miles in 1970 to 400_miles_in 2020. The difference~between""
truck and rail is small for 1970 and no significant difference is antici-
pated between these modes in the future.
48
-------
TABLE 18. DISTANCES BETWEEN CHEMICAL REPROCESSING PLANTS AND
FUEL PREPARATION PLANTS IN 1973
Fuel Preparation Plant Site
Hematite, Missouri
Crescent, Oklahoma
Erwin, Tennessee
Apollo, Pennsylvania
Columbia, South Carolina
Wilmington, North Carolina
Richland, Washington
AVERAGE
Distance (Mi) to Chemical Reprocessing Plant Sites
West Valley,
New York
Distance
by Truck
(Mi)a
800
1,310
850
220
720
760
2,550
1,030
Distance
by Rail
(Mi)b
760
1, 340
820
200
870
1,030
2, 560
1,083
Barnwell,
South Carolina
Distance
by Truck
(Mi)a
750
1, 130
260
650
50
260
2,810
844
aRand McNally Road Atlas: United States, Canada, and Mexico,
Distance
by Rail
(Mi)b
870
1,020
480
850
65
280
2,910
925
Morris,
Illinois
Distance
by Truck
(Mi)a
290
850
650
540
830
970
1,930
866
Distance
by Rail
(Mi)b
320
840
700
480
865
1,280
1,960
921
43rd Annual Edition,
Rand McNally & Company (1967).
DRand McNally Handy Railroad Atlas of the United States, Rand McNally & Company (1971).
-------
A summary of annual transportation data for plutonium is given in
Table 19. The number of shipment-miles grows with a doubling time
of five years until 2000. After 2005, the growth rate for plutonium
production is about 5 percent per year.
TABLE 19. SUMMARY OF ANNUAL PLUTONIUM SHIPPING DATA
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Mass
Transported
(MT)
8
24
53
93
248
627
1, 189
1,678
2,192
2,729
Radioactivity
Transported
(MCi)
90
280
650
950
1,680
3,090
5, 100
6,770
8, 510
10, 340
Number of
Shipments
60
183
400
704
1,887
4,764
9,037
12,755
16,658
20,737
Average
Shipment
Distance
(Mi)
940
880
820
760
700
640
580
520
460
400
Shipping Units
(10° Shipment-
Mi)
0.06
0. 16
0.33
0.54
- 1.32
3.05
5.24
6.63
7.66
8.29
ANNUAL TRANSPORTATION OF HIGH LEVEL
RADIOACTIVE SOLID WASTE
The quantities of high level radioactive solid waste are functions of the
total heavy metal weight in the fuel charged to the chemical processing
plants (References 2, 6).
All the solid waste is assumed to be stored at the chemical processing
plants for ten-years before it is shipped to the Federal Waste Repository.
The annual volumes of high level waste to be transported are calculated
using the data in Table 4. The results are displayed in Table 20 and
Figure 13.
50
-------
TABLE 20. ANNUAL TRANSPORTATION
REQUIREMENTS FOR HIGH LEVEL RADIOACTIVE SOLID WASTE
Year
1985
1990
1995
2000
2005
2010
2015
2020
Amount of High Level Radioactive Solid Waste
(103 ft3)a
LWR
LWR
Without
Pu
Recycle
1.7
4.2
7. 1
14. 6
21. 0
22. 1
21. 3
17.5
LWR
With
Pu
Recycle
0.5
2.4
0.7
Subtotal
1.7
4.7
9.5
15.3
21.0
22.1
21.3
17.5
HTGR
0. 1
1.0
3. 5
5.7
8.5
10. 9
13. 3
LMFBR
0.5.
4.4
15.3
32 . 0
. 48.7
Total
1.7
4.8
10.5
19.3
31. 1
45.9
64. 2
79. 5
a ~
Specific volumes of wastes from various spent fuels
(Reference 6) are:
Fuel
Specific Volume
of Waste (ft3/MT) '
LWR - No Pu Recycle
LWR - Pu Recycle
HTGR
2
2
6
LMFBR 3
51
-------
FORECAST
EXTRAPOLATION
U
1970
1980
1990
2000
2010
2020
FIGURE 13. PROJECTION OF HIGH-LEVEL RADIOACTIVE
SOLID WASTE SHIPPING REQUIREMENTS
52
-------
The total amount of high level waste accumulate(d over 50 years is about
1.3 million ft . This amount of waste would occupy about 30 acres,
assuming the allotment rate of 1 acre/50, 000 ft given by a commercial
burial operation (Reference 6). This allotment rate actually applies to
solid waste other than high level waste, and is used here only to indicate
that the total quantity is small.
The radioactivity contents of high level radioactive solid waste produced
from processing a metric ton of fuel from the different reactors under
consideration are given in Table 6. Combining this information with the
data in Table 20, the annual amounts of radioactivity transported in the
form of high-level radioactive solid waste are calculated and tabulated in
Table 21.
Technical data describing shipment capacities for high level radioactive
solid waste are given in Table 22. The container which will most likely
be used for transportation and storage purposes is the one with largest
capacity, 1. 6 m.3 (56. 5 ft^) shipment.
Since the Federal Waste Repository will most likely be located in the
western United States, where population density is lowest, the shipping
distances for high-level waste transportation will be long. For this
reason, most of the waste will probably be shipped by rail, since long-
haul traffic is more convenient by rail than by truck. For purposes of
this study, 75 volume percent of the waste is assumed to be shipped by
rail and 25 volume percent by truck, each with a capacity of 56. 5 ft^
per shipment. Using this scenario, the annual number of high-level
radioactive solid waste shipments was calculated from the data in
Table 20. The results are shown in Table 24.
Probably only one Federal Waste Repository, but possibly two, will be
built in the next 50 years. The three sites used in Table 23 appear to
be good candidates for such a facility. Distance data for high level waste
shipments made in the near future to these sites are given in Table 23.
The average distance is 2, 130 miles by truck or 2, 190 miles by rail;
for the purposes of this study, these values were adopted for the period
1970-2000. By year 2000, construction of chemical processing plants
is expected for the southwestern and western United States. If they are
built within the vicinities of Dallas, Texas, and Salt Lake City, Utah, the
average distance will be 1,730 miles by truck and 1, 840 miles by rail;
these values were adopted for the purposes of this study for the period
2000-2020.
53
-------
TABLE 21. RADIOACTIVITY OF ANNUALLY
TRANSPORTED HIGH-LEVEL RADIOACTIVE SOLID WASTE
Year
1985
1990
1995
2000
2005
2010
2015
2020
Radioactivity of High Level Waste (103 MCi)a
LWR
LWR
Without
Pu
Recycle
0.26
0.63
1.06
2.19
3.15
3.32
3.20
2.62
LWR
With
Pu
Recycle
0.12
0.60
0.18
Subtotal
0.26
0.75
1.66
2. 37
3. 15
3.32
3. 20
2.62
HTGR
0.02
0.20
0.70
1.14
1.70
2.18
2.66
LMFBR
0.10
0.88
3.06
6.40
9.74
Total
0.26
0.77
1. 86
3. 17
5.17
8.08
11.78
15.02
aSpecific radioactivity levels of wastes from various spent fuels are:
Fuel
Specific Activity
of Waste (MCi/ft3)
LWR - No Pu Recycle
0. 15
LWR - Pu Recycle 0. 25
HTGR 0.20
LMFBR 0.20
54
-------
TABLE 22. SHIPMENT CAPACITIES FOR HIGH-LEVEL RADIOACTIVE SOLID WASTE
Material
Designation
Mode
Size Data
Ul
Ul
Shipment of high
level radioactive
solid waste from.
chemical
process plant to
Federal Waste
Repository
GE cask 701 or 702a
DOT SP pending
All
Salt vault cask1-
Truck
Not available0
Truck, rail,
or barge
Cask: 1.5m dia x 1. 6 m high
Cavity: 0.4m dia x 1. 0 m deep
Volume of contents: 0. 13 m^
Shield: 0.3 m Pb
Radioactivity of contents: 10^ C:
Number of casks per shipment: 1
Cask: l.lm dia x 3. 4 m long
Cavity: 0.5m dia x 2. 3 m long
Shield: stainles.s steel interior, steel
exterior 0.2 m (equivalent) Pb
Number of canisters: 8
Radioactivity of contents: 1.855 MCi
Number of casks per shipment: 1
Cask: 1.5m dia x 3. 7 m long
Cavity: 1.2m dia x 3. 1 m long
Shield: carbon steel interior with water
channels 0.2m Pb. 0. 04 m carbon steel
Canister: 0. 3 m dia x 3. 0 m long
Volume of canister contents: 0. 18 m^
Number of canisters: 9
Volume of cask contents: 1.6 m^
Number of casks per shipment: 1
aDirectory of Shipping Containers for Radioactive Materials. United States Atomic Energy
Commission. Washington, D. C. October 1969.
"Reference a.
cBlomeke, J. O. Magnitude of the Waste Management Problem. Oak Ridge National Laboratory.
Lecture given at UCLA. ORNL-J3WG-71-3841. July 1972.
-------
TABLE 23. DISTANCES BETWEEN CHEMICAL PROCESSING PLANTS
IN 1973 AND POSSIBLE FEDERAL WASTE REPOSITORIES
Chemical
Processing
Plant Site
West Valley,
New York
Barnwell,
South Carolina
Morris,
Illinois
AVERAGE
Transport
Mode
Trucka
Railb
Truck
Rail
Truck
Rail
Truck
Rail
Distance to Selected Federal
Waste Repository Sites (Mi)
Nevada
2,560
2,570
2,640
2,720
2,000
2,000
2,400
2,430
Hanford
2,550
2,570
2,810
2,860
1,930
1,960
2,430
2,463
New Mexico
1,860
1,980
1,460
1,560
1,380
1,500
1,567
1,680
aRand McNally Road Atlas: United States, Canada, and Mexico,
43rd Annual Edition, Rand McNally & Company (1967).
bRand McNally Handy Railroad Atlas of the United States,
Rand McNally & Company (1971).
TABLE 24. SUMMARY OF ANNUAL SHIPPING DATA FOR
HIGH-LEVEL RADIOACTIVE SOLID WASTE
Year
1985
1990
1995
2000
2005
2010
2015
2020
Volume
Transported
(ft3)
1,700
4,800
10, 500
19,300
31, 100
45,900
64,200
79,500
Radioactivity
Transported
(MCi)
260
700
1,860
3,170
5, 170
8,080
11,780
15,020
Number of
Shipments
31
87
189
348
560
827
1, 156
1,431
Average
Shipment
Distance
(Mi)
2, 175
2, 175
2, 175
2, 175
1,810
1,810
1,810
1,810
Shipping Units
(10° Shipment-
. M )
0.07
0. 19
0.41
0.76
1.01
1.50
2.09
2.59
56
-------
The summary of annual transportation data for high level waste appears
in Table 24.
ANNUAL TRANSPORTATION OF GASEOUS FISSION PRODUCTS
Some fission products are in the form of gas. Reference to Table 6
indicates that krypton and xenon are the predominant gaseous radio-
nuclides at the time of processing. The other radioactive gases have
essentially decayed away by the time of processing. Since krypton
and xenon belong to a family of chemically inert species called noble
gases, the gaseous fission products are called noble gases.
The longest lived component of fission gases is Kr-85, which has a half
life of 10. 7 years, which makes it the most significant gaseous radionuclide\
in nuclear fuels. One possible management strategy for the gaseous wastes
is to store them temporarily and then release them to the atmosphere. In
this study, the premise is adopted that such release provides an unaccept-
able solution and shipments of fission gases are assumed to be part of the
transportation scenario (Reference 22). Research is being conducted to
find a feasible method of entraining the gases in a solid matrix for trans-
port purposes. When such solidification processes are available, the
shipments of gases would be counted as shipments of high level radio-
active solid wastes. Such shipments would probably be safer than
shipments of pressurized cylinders of gas since the amount of gas
released in an accident would be much less. For purposes of estimating
the risk, the management of gaseous waste is assumed to include trans-
port of gases in pressurized cylinders and not release to the atmosphere
of the gases produced in fission or chemical processes. All of the
gaseous waste is assumed to be stored at the chemical processing
plants for ten years before it is shipped to the Federal Waste Repository.
The radioactivity of the fission-product gases at the time of shipping was
calculated from the radioactivity of the various spent fuels at time of
processing in Table 6. The radioactivity that maybe transported
annually in the form of pressurized cylinders of radioactive gas was
calculated. The results are given in Table 25.
Technical data describing the containers and shipment capacities for noble
gas are presented in Table 26. The container adopted for the purposes of
this study consists of 6 cylinders holding 0. 18 MCi each. The annual
number of cylinders of gas to be transported was calculated from the
annual radioactivity data in Table 24. The results are presented in
Table 27.
57
-------
TABLE 25. RADIOACTIVITY OF ANNUALLY
TRANSPORTED NOBLE GAS
Year
1985
1990
1995
2000
2005
2010
.*
2015
2020
aAfter
Radioactivity of Noble Gas (MCi)a
LWR
LWR
Without
Pu
Recycle
5
13
22
45
64
68
65
57
LWR
With
Pu
Recycle
1
5
2
Subtotal
5
14
27
47
64
68
65
57
HTGR
1
6
19
34
47
60
73
LMFBR
-
1
7
25
53
77
Total
5
15
33
67
105
140
178
207
10 years storage. Fission Product Gases
Fuel (Percent Radioactivity)
LWR - No Pu
Recycle 0. 136
LWR - Pu Recycle 0.074
HTGR 0.718
LMFBR 0.049
58
-------
TABLE 26. SHIPMENT CAPACITIES FOR NOBLE GAS
Material
Designation
Mode
Size Data
Gaseous Fission Products
Shipped from Chemical
Process Plant to Federal
Waste Repository
Gas Shipping3-
Container
BE 374
Not Available13
Ul
vO
Truck
Rail
Barge
Truck
Rail
Barge
Drum: 0. 6 m dia x 1. 1 m long
Cavity: 0.3m dia x 0. 8 m long
Shield: 0.15m Phenolic Foam
0 to 8 cm Pb
Cylinder: 10 cc to 52
Radioactivity of Contents: 1000 Ci
Number of Drums per Shipment: 1
Cask: Structure with 2 cm steel wall
large enough to contain 6 standard 502
(2200 psi) cylinders in hexagonal array
Shield: Steel raschig rings, water
Radioactivity of Cylinder Contents: 0. 18 MCi
Number of Cylinders per Shipment: 6
aDirectory of Shipping Containers for Radioactive Materials.
Commission. Washington, D. C. October 1969.
United States Atomic Energy
'-'Blomeke, J. O. Magnitude of the Waste Management Problem. Oak Ridge National Laboratory.
Lecture given at UCLA. ORNL-DWG-71-3839. July 1972.
-------
TABLE 27. SUMMARY OF ANNUAL SHIPPING DATA FOR NOBLE GAS
Year
1985
1990
1995
2000
2005
2010
2015
2020
Radioactivity
Transported
(MCi)
5
15
33
67
105
140
178
207
Amount
Transported
(Cyl)
28
84
184
373
584
778
989
1150
Number of
Shipments
5
14
31
63
98
130
165
192
Average
Shipment
Distance
(Mi)
2, 175
2, 175
2, 175
2, 175
1,810
1,810
1,810
1,810
Shipping Units
(106 Shipment-Mi)
0.01
0.03
0.06
0. 11
0. 18
0.24
0.30
0. 35
-------
Transportation of noble gas is expected to have the same scenario as that
of high-level solid waste, since noble gas is a waste product that is
destined for storage in the Federal Waste Repository. The use frequencies
of truck and rail modes of transport and the average shipment distances of
high-level waste transportation are adopted for noble gas transportation.
The summary of annual transportation data for noble gases is given in
Table 27.
SUMMARY
The transportation activity data in Tables 12, 19, 24, and 27 are repre-
sented graphically in Figure 14. Nuclear power transportation activity
exceeds one million miles in 1980 and ten million miles after 2000. Up
until about year 2005, spent fuel transportation will dominate. Plutonium
transportation increases dramatically after 1995 and exceeds spent fuel
transportation after 2005.
The transportation of radioactive waste does not exceed ten percent of
the total nuclear power transportation activity until after the year 2000.
Shipments of radioactive gases comprise less than two percent of the
transportation activity.
61
-------
w
CO
H
I
H
*
W
a
CO
NO
4
H
H-i
H-i
U
O
I—1
EH
§
0
High-Level
Solid Waste
1970
1980
1990
2000
2010
2020
FIGURE 14. SUMMARY,OF ANNUAL TRANSPORTATION ACTIVITY
•62
-------
SECTION VI
TRANSPORTATION ACCIDENT RISKS
INTRODUCTION
One risk associated with the shipment of radioactive materials from the
nuclear power industry is the release of radioactivity from the shipment
package as a result of an accident during transportation. However, there
is very little statistical data on which to assess this risk. Within the
United States over the past 25 years, there have been about 300 reported
accidents in transportation involving packages of all kinds of radioactive
material. About 30% of those accidents involved release of radioactive
material from medical and industrial radiochemicals. None resulted in
perceptible injury or death attributed to the radiation aspects. There
have been no releases from nuclear power shipments.
At present, shipments of radioactive materials in the nuclear power
industry move in routine commerce on conventional transportation
equipment. Therefore, shipments are subject to the same transportation
environment, including accidents, as nonradioactive cargo. Consequently,
the frequency and severity of accidents involving shipments of radioactive
materials can be estimated from accident statistics involving the trans-
portation of nonradioactive materials.
The public may receive a radiation exposure from a transportation
accident if radioactivity is released from the shipment package. For
the purposes of this study, it is assumed that if the package containing
radioactive material is sufficiently damaged, there will be a release
of radioactivity. Therefore, the risk of this release can be directly
related to the probability and severity of damage to the package. This
damage is related to the severity of the accident, the form and amount of
energy sustained by the package as a result of the accident, and its ability
to withstand those forces. This ability is a function of packaging design.
This section describes transportation accidents risks. The parameters
to be estimated are the probability and severity of accidents by trans-
portation mode, the probability and severity of damage to the package,
and the probability of release of radioactive materials. The probability
and severity of accidents will be based on reported statistics for com-
mercial vehicles. Each accident is described by a sequence of events
which can damage the package and thus lead to the release of radioactivity.
The combinations of events necessary for release is described by a fault
tree. Each event in the sequence is a statistical variable which could be
described by a probability distribution if enough information were available.
In this report, best estimate values are used as input data. The probability
of release in a shipment accident is determined by fault tree analysis.
63
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ACCIDENT PROBABILITY AND SEVERITY
The probability and severity of accidents to shipments of radioactive
materials depends on the characteristics of the transport vehicles. In
this study, the shipments are expected to go by truck, rail, and barge
on paths that are used also for shipments of other products of commerce,
industry, and defense. Accident statistics for these transport vehicles
maintained by Federal agencies, such as the Department of Transportation
and the Coast Guard, provide readily available estimates of the accident
probabilities for nuclear shipments. Specific values of these probabilities
are shown in Table 28.
TABLE 28. PROBABILITIES OF ACCIDENTS
Mode
Truck
Rail
Barge
Probability3- .
[(106 Shipment-Miles)" J
1.69
0.8
1.8
aReference 14.
The accident statistician recognizes three broad categories of traffic
accidents: collisions, noncollisions, and other events. Collisions involve
interactions of the transport vehicle with other objects, whether moving
vehicles or fixed objects. Noncollisions are accidents in which the trans-
port vehicle leaves the transport path or deviates from normal operation
in some way, such as by rolling over on its top and side. Accidents
classed as other events include personal injuries suffered on the vehicle,
records of persons falling from or being thrown against a standing
vehicle, cases of stolen vehicles, and fires occurring on a standing
vehicle. The probabilities with which these categories of events have
been observed in truck and rail freight accidents are indicated in
Table 29.
All these accidents may involve conditions which lead to damage of the
cargo. In particular, physical forces arising from impact, puncture,
vibration, and fire events tend to reduce either the ability of the shipment
64
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TABLE 29. ACCIDENT FREQUENCY STATISTICS
Parameter
Speed
Fire
Duration
Mode
Truck
TOTAL
Rail
TOTAL
Truck
Rail
Magnitude
0-30 mph
30-50 mph
50-70 mph
>70 mph
0-30 mph
30-50 mph
50-70 mph
>70 mph
0 hr
<0.5 hr
0. 5-1 hr
>1 hr
0 hr
<0.5 hr
0.5-1 hr
>1 hr
Fraction of Accidents2-
Collision
0.221
0.422
0.153
0.005
0.801
0. 123
0.067
0.020
0.0002
0.210
Noncollision°
0.008
0.'065
0.022
0.095
0.410
0.224
0.065
0.001
0.700
Other0
0. 104
0.090
0.090
0.98430
0.01556
0.00012
0.00001
0. 98500
0.01275
0.00210
0.00015
aBased on data from Reference 14.
^Noncollision accidents for trucks are classified as run-offs and roll-
overs, while for rail transport, they are classified as derailments.
cOther accidents include events assumed to be of no significant to
release of radioactivity.
package to contain radioactivity and shield radiation or the ability of the
package to dissipate heat or both. The ability of the materials and struc-
tures in the shipment package to resist these forces depends on the mag-
nitudes of the forces. These magnitudes vary with the severity of the
accident, as does the frequency with which they occur. The variation of
the frequency of accident severity can be formulated on the basis of
accident experience in the cases of truck and rail transport, but less
data is readily available in the case of barge transport.
65
-------
The statistical data indicate that collision and noncollision truck and rail
accidents may be classified into four ranges of vehicle speeds. In addi-
tion, all accidents may be classified into four ranges of time duration of
fires that may or may not break out. For convenience, one of these time
ranges is defined so as to describe all those accidents that do not involve
fire. The compositions with respect to these parameters of the frequency
of collision and noncollision accidents to truck and rail freight shipments
are given in Table 29 .
The partition of the entire set of freight traffic accidents into severity
classes is somewhat arbitrary. In this report, all accidents have been
divided into one of three severity classes. Minor accidents include those
collisions and noncollisions at low speed and short duration of fire and
those other accidents with short fires. Moderate accidents include
combinations of speed and fire of intermediate intensity. Severe
accidents include combinations of speed and fire of large intensity. The
data in Table 29 have been classified in Table 30. Barge accident proba-
bilities have been classified on the basis of the duration of fires and
actual data on cargo damage in Reference 7.
TABLE 30. ACCIDENT SEVERITY CLASSES
Severity
Minor
TOTAL
Moderate
TOTAL
Severe
TOTAL
Speed
0-30 mph
30-50 mph
Other
0-30 mph
30-50 mph
50-70 mph
Other
0-30 mph
30-50 mph
50-70 mph
>70 mph
Other
Fire
0-1 hr
l/2->l hr
l/2->l hr
0->1 hr
>1 hr
Fraction of Accidentsa
Truck
0.229
0.479
0. 104
0.812
0.008
0. 175
0.183
0.005
0.005
Rail
0.531
0.287
0.090
0.908
0.002
0.004
0.085
0.091
0.001
0.001
Barge
0.900
0.080
0.020
aFractions less than 0. 001 are not entered in table.
66
-------
CONTAINMENT FAILURE IN SHIPMENT ACCIDENTS
In order for the public to be exposed to radioactive material contained in
the shipping containers, at least one of several possible sequences of
events must occur. The shipping containers may be represented as
concentric series of envelopes. Each envelope must be broken in some
way to provide a pathway for the radioactive contents to spill outside the
container. Even if a pathway to the environment is provided, a physical
force must1 be available to drive the contents outside. Evaluation of the
probability of occurrence of these accidental conditions is facilitated by
fault tree analysis-. The shipping container is represented by a fault tree
diagram, which shows the logical relationships between barriers that
must be faulted and pathways that must be provided for releases to occur.
Probability values are attached to each of the primary events represented
in the fault tree diagram.
Mechanisms Leading to Containment Failure
The principal mechanisms in an accident which may initiate sequences
of events leading to containment failure used in this report are:
1. Impact.
2. Puncture.
3. Thermal mode.
4. Vibration.
5. Equipment failure.
6. Human error.
(
Although the list is not exhaustive, these mechanisms are the primary
events used in the fault trees for each shipping container.
The probabilities of impact and puncture for truck and rail accidents
are estimated directly from accident statistics. The probability of
impact is assumed to be the fraction of accidents denoted by collisions.
In the noncollision category, which for truck accidents includes run-off s
and roll-overs and for rail accidents includes derailments, the shipment
package is more likely to encounter sharp objects than in collisions or
other accidents. Thus, the probability of puncture is assumed to be the
67
-------
fraction of accidents denoted by noncollisions. . It is assumed that both
impact and puncture can occur as a result of a barge accident. The
probabilities of impact and puncture for barge accidents were estimated
by engineering judgement.
Failure in the thermal mode requires a high temperature environment
for the shipping container. Two causes of high temperature are recog-
nized: a heat source outside the container and a heat source inside the
container. The external heat source is provided by the cargo when the
cooling capability of the shipment package has been impaired. The
probability of an external heat source is estimated from the incidence of
fires in the accident statistics.
The information gain from accident statistics can be formulated to
reflect the conditional probability of events as a function of accident
severity. In particular, the conditional probabilities of impact, puncture
and external heat source can be estimated. For example, the conditional
probability of impact, given an accident of a specific severity, is the
fraction- of accidents of that severity which are collisions. Other condi-
tional probabilities are evaluated similarly. The results are presented
in Table 31.
TABLE 31.
CONDITIONAL PROBABILITIES OF IMPACT,
PUNCTURE, AND FIRE
Mode
Truck
Rail
Barge0
Accident
Severity
Minor
Moderate
Severe
Minor
Moderate
Severe
Minor
Moderate
Severe
Probability
0.812
0. 183
0.005
0.908
0.091
0.001
0.90
0.08
0.02
Conditional Probability3-
Impact
0.783
0.874
0.987
0.208
0.234
0. 194
0.20
0.50
1.00
Puncture
0.089
0. 126
0.002
0.693
0.763
0.800
0
0.30
0.80
Fire
0.006
0.057
0.033
0.009
0.068
0.449
0
0.0065
0.065
aDerived from Tables 29 and 30, this report.
^Impact and puncture 'numbers estimated by engineering judgment.
-68
-------
Accident statistics are not available for vibration, equipment failures
and human error. The conditional probabilities of vibration, equipment
failures and human error for severe accidents are taken from another
study (Reference 15). As conservative estimates, the probabilities are
assumed to be the same for minor and moderate accidents. This study
does not include the release pathway described by diffusion of radioactivity
through a barrier because little information on the phenomena is available
and the release rates are small.
Description of Shipping Containers
Shipping containers that are expected to be used for transporting spent
fuel, recycled plutonium, high level radioactive solidified waste, and
fission gases were selected for fault tree analysis. Schematic diagrams
for these containers are presented in Figures 15 through 18.
The diagram for the spent fuel shipping container in Figure 15 is a simpli-
fied schematic description of the General Electric IF 300 cask (Refer-
ence 16). Briefly, the cask consists of a fuel cavity lined with a steel
shell, surrounded by a gamma shield of depleted uranium and an outer
steel shell. The fuel is cooled with water or solution inside the cavity.
Outside the outer shell, the cask is fitted with a jacket of water, which
serves to shield the public from fast neutron fluxes and to further cool
the cask. Additional cooling by circulation of air is provided, but since
such a blower system presents no significant barrier to radioactive
material release, and since the cask is designed to withstand the thermal
rigors of the heat load from both the fuel and the sun on a 130 F day with-
out the benefit of these auxiliary blowers (Reference 16), this blower
system is disregarded in the fault tree diagram.
Essentially, the cask may be regarded as a double containment of the
fuel. The cladding provides one barrier and the combination of inner
shell, gamma shield, and outer shell provides the other. The second
barrier may be penetrated through one of three paths: blowdown of the
pressure relief valve, loss of sealing capability of the closure head, and
breachment of the cask. Breachment of the cask is considered to require
breachment of each of two subbarriers: the inner shell and the layer of
combined shield and outer shell. Neither the primary coolant fluid in the
cavity nor the neutron shield water and jacket are regarded as significant
barriers to material release.
The diagram for the recycled plutonium shipping container in Figure 16
is a simplified schematic description of the Dow Chemical Model 2030-1
container (Reference 17). Briefly, the container consists of a 30-gallon
drum lined with boards of Celotex insulation. The Celotex boards are cut
69.
-------
r
Pressure Relief Valve
Closure(.
Seal
Outer Shell
Inner Shell
Cladding (Typical)
FIGURE 15. SCHEMATIC DIAGRAM OF
SPENT FUEL SHIPPING CONTAINER
12 LN. SQUARE POROUS .
REFRACTORY FIBER INSULATION f
TO LNSVRE TIGHT TIT ICERATELT) N
INSULATING CELOTEX RINGS (Kl
I-IN. THICK (MAY DC ARRANGED
TO ENCLOSE 2 7.IN. IIIC.H
CONTAIN?.ITNT VESSELS OR
1 ll-IN. IIIOI CONTAINMENT
VESSEL)
CONTAINMENT VESSELS (2)
10 IX. DIAMETER X T IN.
MUSH 12 CALCr ,0.109 LX)
STAINLESS STEEL
LID IS 0.2 > IN. THICK
'S/////////77// ••:''
\v^\^\\\\\\\\\\
I IN. DIAMETER VCNT HOLE SCALED WEATHERTXCHT
BY MOLDED PLASTIC PLUO
OOT-I'X DRUM 1)0 GALLON) 20 W. DIAMETER X »« DC
H:CH it GAUGE 10.041 IN. i STEEL
ICUICX CONNECT COUPUXG TOR EVACVATLVG AND
BACxriLLLNG VESSEL OITII INERT CA> OR DR1T AIR
COMPOUND GAUGE INDICATING LVTERNAL PRESSURE
' TO It PSt AND VACUUU TO 10 IN. I(G
FLANGE SOLTS
• t/ltlN. CAP SCREWS
U ON EACH CONTAINMENT VESSEL
VITON O-UNC SEAL UP TO SERVICE TEMPERATURE
' or
-------
into rings and can be arranged to form a single large cavity to accom-
modate a large containment vessel, or to form two smaller cavities to
accommodate two small containment vessels.
In this container, the plutonium must penetrate three barriers to reach
the atmosphere: the containment vessel, the Celotex insulation, and the
drum. Pathways may be provided by a break in the vessel walls or the
loss of seal at the interface of the vessel lid and the vessel wall, by
breakage, slippage, or warping of the Celotex boards, and breakage, loss
of seal at the edge of the lid, or removal of the molded plastic plug in the
lid vent opening of the drum.
In Figures 17 and 18, schematic diagrams are shown for possible shipping
containers for high level radioactive solidified waste and gaseous fission
products based on Oak Ridge National Laboratory designs (Reference 10).
These containers are similar to spent fuel shipping containers in that the
materials are packed in individual elements which are mounted in a
shielded, cooled cask. In the case of spent fuel, the elements are con-
tained by cladding and end boxes, whereas the waste products elements
are contained by metal cans or cylinders.
Container Damage and Release Probability
The probability and severity of damage to a container in an accident is
related to the severity of the accident, the form and amount of force
applied to the container as a result of the accident and the ability of the
container to withstand those forces. The form and amount of force
transmitted to the package in an accident depend on the design features
of the transport vehicle and the mounting of the container on the vehicle.
The ability of the container to withstand these forces resulting from the
accident depends on the design of the container.
The severity of container damage in an accident determines the size of
the release pathway formed and thus, the quantity of radioactivity
released to the environment. In this study, three discrete values of the
container damage severity were postulated: small, medium and large.
Container damage consisting of microscopic openings such as hairline
cracks and pinholes is considered to be small damage severity. Medium
damage severity is taken to be openings on the order of the size of a
fill port. An opening several times larger than a fill port is assumed
to be large damage severity.
Estimates have been made elsewhere of the fraction of containers
expected to be damaged and the severity of that damage for various
accident severity classes (Reference 7). Some of the accidents result in
71
-------
• LEAD SHIELD
CARBON STKEL CASKET
WITH COOLANT CHANNELS
CASKETED CLOSURE
CAVITY KILLED WITH
WATER AND ENERGY
AnsonniNG KINS
(I)OTH ENDS)
FIGURE 17. SCHEMATIC DIAGRAM OF HIGH LEVEL
RADIOACTIVE SOLIDIFIED WASTE SHIPPING CONTAINER
1/4-INCH STEEL CASKET
WALL WITH EXTERNAL
INSULATION
KILLED WITH STEEL
RASCIIIG RJNGS AND
WATER
FIGURE 18. SCHEMATIC DIAGRAM OF
FISSION GAS SHIPPING CONTAINER
72
-------
no breach of containment and thus, no release of radioactivity to the
environment. This integrity has been attributed to Federal regulations
which include testing requirements for containers used for the shipment
of radioactive materials.
In this study, the probability of release of radioactivity is assumed to
be directly related to the probability and severity of damage to the
container. The release severity is thus classified as small, medium
and large corresponding to the discrete values of container damage
severity. The conditional probability of release as a function of accident
severity is estimated by fault tree analysis. Because accident statistics
are not available, the conditional probabilities of release are assumed
to be constant with accident severity.
Fault Tree Diagrams and Data
Fault tree diagrams of the spent fuel and plutonium shipping containers
are shown in Figures 19 and 20, respectively. The fault tree in Figure 21
applies to both high-level radioactive solid waste and noble gas shipping
containers. Each fault tree identifies the combinations and sequences of
events which must occur for the release of radioactivity.
Fault tree input data-for the spent fuel shipping container are presented
for each of the three transportation modes in Tables 32, 33, and 34. The
data are numbered sequentially corresponding to the numbers on the fault
tree diagram of Figure 19. Six mechanisms are identified corresponding
to the principal accident mechanisms leading to failure of a containment
barrier.
The first entry for each mechanism is the conditional probability of
occurrence of the mechanism given an accident of a specific severity.
Subsequent entries are the conditional probabilities of release of a
specific severity given that the mechanism has actually occurred in an
accident of specified severity. These entries are the probabilities of
barriers yielding to the force of a mechanism, the so-called inhibit gate
probabilities. The probability data were adopted from Reference 15. The
barriers are arranged in the tables from inside the container outward as
much as possible, although certain pathways, such as pressure relief
valve lines and closure seals, may span two or more barriers. Basic
fault tree input data for each of the fault trees for the remaining shipping
containers and their associated transportation modes are presented in
Tables 35, 36, 37, and 38. The input data for noble gas shipping containers
are assumed to be the same as for solid waste containers, since their
fault trees are similar and the inhibit gate probability information is not
established well enough to distinguish between the two containers.
73
-------
RELtASEOF flADIQACT
MATE filALfROM WENT
SHIPPING CONTAINER
VE
FUEL
1
FAILURE OF
FUEL
CLADDING
• - - •
•- 1 v : • • •-
• -.-I
FAILURE OF *
CONTAINMENT
srsTtw
r
H1QM
PRESSURE
1
MELTING
1
HEAT
SOURCE
1
EXTERNAL
HEAT
SKOUBCE
I
MECHANICAL
MODE
THERM
MODE
AL
n
FIGURE 19. FAULT TREE DIAGRAM FOR SPENT FUEL
CMIPPIII6 CONTAmiR
74
-------
RELEASE OF RADIOACTIVE
MATERIAL FROM PLUTONIUM
SHIPPING CONTAINER
FIGURE 20. FAULT TREE DIAGRAM FOR PLUTONIUM
SHIPPING CONTAINER
75
-------
HE 16 AM OF RADIOACTIVE MAT I RIAL
FAOM SMimNO CONTAINED FOR
HIGHLEVKL RADIOACTIVE
SOUOtHiO WASTE Ofl FISSION OAS
FIGURE 21. FAULT TREE DIAGRAM FOR SOLID WASTE AND
NOBLE GAS SHIPPING .CONTAINERS
76
-------
TABLE 32. FAULT TREE INPUT DATA FOR SPENT FUEL SHIPPING CONTAINER ON TRUCKa
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
• 26
27
28
29
Mechanism
Impact
Thermal
Puncture
Vibration
Equipment
Defects
Human
Error
Input Event Name
Impact Occurs0
Impact > Cladding0
Impact > Inner Shellc
Impact > Shield. Outer Shell0
Impact > Closure Seal
Fire Occurs0
Pressure > Claddingc
Preasure > Inner She llc
Pressure > Shield, Outer Shellc
Preasure > Closure Sealc
Temperature > Cladding
Temperature > Inner Shell
Temperature > Shield. Outer Shell
Temperature > Closure Seal
Pressure > Slowdown Setting of
Pressure Relief Valvec
Puncture Occurs0
Puncture > Inner Shellc
Puncture > Shield, Outer Shellc
Puncture > Closure Sealc
Puncture > Neutron Shield Jacket0
Coolant Leaks Out°
Vibration Occurs
Vibration > Cladding
Vibration > Inner Shell
Vibration > Shield. Outer Shell
Vibration > Closure Seal
Defective Seal
Inadequate Coolant
Improper Closure
Conditional Release Probability
Small Release
Minor
Accident
0.783
O.ZxlO"3
0. IxlO"3
0. lxlO'2
0. IxlO'2
0.006
0. IxlO'2
0.3xlO-3
0.3xlO-3
0.3xlO-3
0. 1
0. IxlO-2
0.1
0.1
0. IxlO'2
0.089
0.1x10-3
0. 1x10-3
0.1x10-3
1.0
0.089
O.S
0.3xlO-5
0.15x10-4
0.15x10-4
0.75x10-5
0.5x10*3
0.3x10-4
0.5x10-3
Moderate
Accident
0.874
0.057
0.126
0.126
Severe
Accident
0.987
0.033
0.002
0.002
Medium Release
Minor
Accident
0.783
O.lxlO-3
0.5x10-4
0.5x10-3
O.SxIO-3
0.006
0.5x10-3
0.15x10-3
0. 15x10-3
0.15x10-3
0.05
0.5x10-3
0.05
0.05
O.SxIO-3
0.089
0.089
0.3x10-6
0.15x10-5
0. 15x10-5
0.75x10-6
0.5x10-3
0.3x10-4
O.SxIO-3
Moderate
Accident
0.874
0.057
0.126
0. 126
Severe
Accident
0.987
0.033
0.002
V
0.002
Large Release
Minor
Accident
0.783
0.2x10-4
O.lxlO-4
O.lxlO-3
0. IxlO-3
0.006
0. IxlO-3
0.3xlO-4
0.3x10-4
0.3x10-4
0.01
O.lxlO-3
0.01
0.01
0. IxlO-3
0.089
0.089
0.3x10-7
0. ISxlO-6
0. 15x10-6
0.75x10-7
O.SxIO-3
0.3x10-4
0. 5x10-3
Moderate
Accident
0.874
0.057
0.126
0.126
Severe
Accident
0.987
0.033
0.002
0.002
aBas.ed on data in Reference 15, except where noted. Misaing values for moderate and severe accidents, same as for minor.
bData derived in Table 31.
cEstimated by engineering judgment.
-------
TABLE 33. FAULT TREE INPUT DATA FOR SPENT FUEL SHIPPING CONTAINER ON RAILa
-j
00
No.
1
Z
3
4
5
6
7
8
9
10
11
12
13
14
IS
16
17
18
19
20
21
22
23
24
25
26
27
28
29
Mechanism
Impact
Thermal
Puncture
Vibration
Equipment
Defects
Human
Error
1
Input Event Name
Impact Occurs t>
Impact > Claddingc
Impact > Inner Shellc
Impact > Shield. Outer Shellc
Impact > Closure Seal
Fire Occursh
Pressure > Claddingc
Pressure > Inner Shellc
Pressure > Shield, Outer Shellc
Pressure > Closure Sealc
Temperature > Cladding
Temperature > Inner Shell
Temperature > Shield, Outer Shell
Temperature > Closure Seal
Pressure > Slowdown Setting of
Pressure Relief Valve'
Puncture Occursb
Puncture > Inner Shellc
Puncture > Shield, Outer Shellc
Puncture > Closure Sealc
Puncture > Neutron Shield Jacketc
Coolant Leaks Outc
Vibration Occurs
Vibration > Cladding
Vibration > Inner Shell
Vibration > Shield, Outer Shell
Vibration > Closure Seal
Defective Seal
Inadequate Coolant
Improper Closure
Conditional Release Probability
Small Release
Minor
Accident
0.208
0.4x10-4
0.2x10-4
0.2xlO-3
0.2x10-3
0.009
0.8x10-3
0.2x10-3
0.2x10-3
0.2x10-3
0.8x10-1
0.8x10-3
0.8x10-1
0.8x10-1
0.8x10-3
0.693
0.2x10-3
0.2x10-3
0.2x10-3
1.0
0.693
0.5
0.4x10-5
0.2x10-4
0.2x10-4
0. 1x10-4
0. 5x10-3
0.3x10-4
0.5x10-3
Moderate
Accident
0.234
0.068
0.763
0.763
Severe
Accident
0.194
0.449
0.800
0.800
Medium Release
Minor
Accident
0.208
0.2x10-4
0.1x10-4
0.1x10-3
0.1x10-3
0.009
0.4x10-3
0.'lxlO-3
0.1x10-3
0.1x10-3
0.4x10-1
0.4x10-3
0.4x10-1
0.4x10-1
0.4x10-3
0.693
0.1x10-3
0.1x10-3
0. 1x10-3
0.693
0.4x10-6
0.2x10-5
0.2x10-5
0.1x10-5
0.5x10-3
0.3x10-4
0.5x10-3
Moderate
Accident
0.234
0.068
0.763
0.763
Severe
Accident
0.194
0.449
0.800
0.800
Large Release
Minor
Accident
0.208
0.4x10-5
0.2x10-5
0.2x10-4
0.2x10-4
0.009
0.8x10-4
0.2x10-4
0.2x10-4
0.2x10-4
0.8x10-2
0.8x10-4
0.8x10-2
0.8x10-2
0.8x10-4
0.693
0.2x10-4
0.2x10-4
0.2x10-4
0.693
0.4x10-7
0.2x10-6
0.2x10-6
0. 1x10-6
0.5x10-3
0.3x10-4
0.5x10-3
Moderate
Accident
0.234
0.068
0.763
0.763
Se ve re
Accident
0.194
0.449
0.800
0.800
aBased on data and footnotes In Table 32 with the following adjustment*:
Numbers Multiplier Xl
"2-5 0.2
7-15 0.8
17-19 2.0 .
23-26 1.33
-------
TABLE 34. FAULT TREE INPUT DATA FOR SPENT FUEL SHIPPING CONTAINER ON BARGE*
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27 '
28
29
Mechanism
Impact
Thermal
Puncture
Vibration
Equipment
Defects
Human
Error
Input Event Name
Impact Occurs h
Impact > Claddingc
Impact > Inner Shellc
Impact > Shield, Outer Shellc
Impact > Closure Seal
Fire Occursb
Pressure > Claddingc
Pressure > Inner Shellc
Pressure > Shield, Outer Shellc
Pressure > Closure Sealc
Temperature > Cladding
Temperature > Inner Shell
Temperature > Shield, Outer Shell
Temperature > Closure Seal
Pressure > Slowdown Setting of
Pressure Relief Valvec
Puncture Occursb
Puncture > Inner Shellc
Puncture > Shield, Outer Shellc
Puncture > Closure Sealc
Puncture > Neutron Shield Jacketc
Coolant Leaks Outc
Vibration Occurs
Vibration > Cladding
Vibration > Inner Shell
Vibration > Shield, Outer Shell .
Vibration > Closure Seal
Defective Seal
Inadequate Coolant
Improper Closure
Conditional Release Probability
Small Release
Minor
Accident
0.2
0.2x10-4
0.1x10-4
0. 1x10-3
0. 1x10-3
0
0.8x10-3
0.2x10-3
0.2x10-3
0.2x10-3
0.8x10-1
0.8x10-3
0.8x10-1
0.8x10-1
0.8x10-3
0
0.5x10-4
0.5x10-4
0.5x10-4
1.0
0
0.05
0.1x10-5
0.5x10-5
0.5x10-5
0.2x10-5
0.5x10-3
0.3x10-4
0.5x10-3
Moderate
Accident
0.5
0.006
0.3
0.3
Severe
Accident
1.0
0.065
0.8
0.8
Medium Release
Minor
Accident
0.2
0.1x10-4
0.5x10-5
0.5x10-4
0.5x10-4
0
0.4x10-3
0. 1x10-3
0. 1x10-3
0.1x10-3
0.4x10-1
0.4x10-3
0.4x10-1
0,4x10-1
0.4x10-3
0
0.25x10-4
0.25x10-4
0.25x10-4
0
0.1x10-6
0.5x10-6
0.5x10-6
0.2x10-6
O.SxIO-3
0.3x10-4
O.SxlO"3
Moderate
Accident
0.5
0.006
0.3
0.3
Seve re
Accident
. 1.0
0.065
0.8
0.8
Large Release
Minor
Accident
0.2
0.2x10-5
0.1x10-5
0. 1x10-4
0.1x10-4
0
0.8x10-4
0.2x10-4
0.2x10-4
0.2x10-4
0.8x10-2
0.8x10-4
0.8x10-2
0.8x10-2
0.8x10-4
0
0.5x10-5
0. 5x10-5
0.5x10-5
0
0. 1x10-7
0.5x10-7
0.5x10-7
-0.2x10-7
O.SxlO"3
0.3X10"1
0. SxlO'3
Moderate
Accident
0.5
0.006
0.3
0.3
Severe
Accident
1.0
0.065
0.8
0.8
a Based on data and footnotes in Table 32 with the following adjustments:
Numbers Multiplier
2-5 0. 1
7-15 0.8
17 - 19 0.5
23-26 o.33
-------
TABLE 35. FAULT TREE INPUT DATA FOR PLUTONIUM SHIPPING CONTAINER ON TRUCKa
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
Mechanism
Impact
Thermal
Puncture
Vibration
Equipment
Defects
Human Error
Input Event Name
Impact Occurs"
Impact > Vesselc
Impact > Celotex0
Impact > Drum0
Fire Occursb
Pressure > Vessel0
Pressure > Celotex0
Pressure > Drum0
Temperature > Vessel0
Temperature > Celotexc
Temperature > Drumc
Puncture Occurs^
Puncture > Vessel0
Puncture > Celotexc
Puncture > Drum0
Vibration Occurs
Vibration > Vessel0
Vibration > Celotex0
Vibration > Drum0
Defective Vessel Seal
Defective Penetrations
Improper Closure
~- Conditional Release Probability
Small Release
Minor
Accident
0.783
0. 5xlO-2
0.1
0. 1
0.006
0.01
0. 1
0. 1
0. IxlO-4
0.05
0.05
0.089
0. 5xlO-2
0. 1
0. 1
0. IxlO-4
0.3xlO-6
0.5
0.01
0. 5xlO'3
0.5xlO-3
0.5xlO-3
Moderate
Accident
0.874
0.057
0.126
Severe
Accident
0.987
0.033
0.002
Medium Release
Minor
Accident
0.783
0.25xlO-2
0.05
0.05
0.006
O.SxIO-2
0.05
0.05
0.5xlO"5
0. 25X10"1
0.25xlO-J
0.089
0.25x10-2
0.05
0.05
0. lxlO'5
0.3xlO-7
0.05
0. IxlO-2
0. 5xlO'5
0. 5xlO"5
0. SxlO"5
Moderate
Accident
0.874
0.057
i
0.126
Severe
Accident
0.987
0.033
0.002
Large Release
Minor
Accident
0.783
0. SxlO-3
0.01
0.01
0.006
0. IxlO-2
0.01
0.01
0. IxlO-5
0.5x10-2
0. SxlO-2
0.089
O.SxIO-3
0.01
0.01
0. IxlO"6
0.3xlO-8
O.SxIO-2
O.lxlO-3
0. SxlO"5
O.SxlO'5
O.SxIO-5
Moderate
Accident
0.874
0.057
0.126
Severe
Accident
0.987
0.033
0.002
aBased on data in Reference 15, except where noted. Missing value for moderate and severe accidents, same aa for minor.
bData derived in Table 31.
cEstimated by engineering judgment.
oo
o
-------
TABLE 36. FAULT TREE INPUT DATA FOR PLUTONIUM SHIPPING CONTAINER ON RAIL*
oo
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
Mechanism
Impact
Thermal
Puncture
Vibration
Equipment
Defects
Human Error
Input Event Name
Impact Occurs'1
Impact > Vessel0
Impact > Celotex0
Impact > Drumc
Fire Occursb
Pressure > Vessel0
Pressure > Celotex0
Pressure > Drumc
Temperature > Vessel0
Temperature > Celotexc
Temperature > Drumc
Puncture Occurs*3
Puncture > Vessel0
Puncture > Celotex c
Puncture > Drum0
Vibration Occurs
Vibration > Vessel0
Vibration > Celotex0
Vibration > Drum0
Defective Vessel Seal
Defective Penetrations
Improper Closure
Conditional Release Probability
Small Release
Minor
Accident
0.208
O.lxlO-2
0.02
0.02
0.009
O.SxIO-2
0.08
0.08
0.8x10-5
0.04
0.04
0.693
0.01
0.2
0.2
0. IxlO"4
0.4xlO-6
0.7
0. IxlO'1
O.SxIO-5
0. SxlO-5
O.SxIO-5
Moderate
Accident
0.234
0.068
0.763
Severe
Accident
0.194
0.449
0.800
Medium Release
Minor
Accident
0.208
0.5xlO"3
0.01
0.01
0.009
0. 4xlO.-2
0.04
0.04
0.4xlO-5
0.02
0.02
0.693
O.SxIO-2
0. 1
0. 1
0. IxlO-5
0.4x10-7
0.07
0. IxlO-2
0. 5xlO-7
0. 5xlO~7
O.SxIO-7
Moderate
Accident
0.234
0.068
0.763
Severe
Accident
0. 194
0.449
0.800
Large Release
Minor
Accident
0.208
0. IxlO-3
0.2x10-2
0.2x10-2
0.009
O.SxIO-3
0.8x10-2
O.SxIO-2
O.SxIO-6
0.4x10-2
0.4x10-2
0.693
0. IxlO-2
0.02
0.02
0. IxlO-6
0.4xlO-8
0.7x10-2
0. IxlO"3
0. 5x10-7
O.SxIO-7
0.5x10-7
Moderate
Accident
0.234
0.068
0.763
Severe
Accident
0. 194
0.449
0.800
aBased on data and footnotes in Table 35 with the following adjustments:
Numbers Multiplier
2-4 0.2
6-11 0.8
13 - 15 2.0
17 - 19 1.33
-------
TABLE 37.
FAULT TREE INPUT DATA FOR SHIPPING CONTAINER OF HIGH LEVEL
RADIOACTIVE SOLID WASTE OR NOBLE GAS ON TRUCKa
00
to
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
Mechanism
Impact
Thermal
Puncture
Vibration
Equipment
Defects
Human
Error
Input Event Name
Impact Occursb
Impact > Can or Cylinders
Impact > Caskc
Impact > Closure Sealc
Fire Occursb
Pressure > Can or Cylinders
Pressure' > Caskc
Pressure > Closure Sealc
Temperature > Can or Cylinders
Temperature > Caskc
Temperature > Closure Sealc
Puncture Occursb
Puncture > Can or Cylinders
Puncture > Caskc
Puncture > Closure Sealc
Puncture > Cooling System^
Coolant Leaks Outs
Vibration Occurs
Vibration > Can or Cylinder
Vibration > Caskc
Vibration > Closure Seal
Defective Seal
Inadequate Coolant -•
Improper Closure
Conditional Release Probability
Sn
Minor
Accident
0.783
0. 1x10-4
0.1x10-4
0.01
0.006
0.5x10-2
0.5x10-2
0.05
0. 1x10-4
0.1x10-3
0. 1x10-2
0.089
0. 1x10-3
0. 1x10-3
0. 1x10-2
1.0
0.089
0. 1x10-4
0.2x10-5
0. 1x10-4
0.2x10-5
0.5x10-3
0.3x10-4
0.5x10-3
nail Releas
Moderate
Accident
0.874
0.057
0.126
0.126
e
Severe
Accident
0.987
0.033
0.002
0.002
Medium Release
Minor
Accident
0.783
0.5x10-5
0.5x10-5
0. 5x10-2
0.006
0.25x10-2
0.25x10-2
0.025
0.5x10-5-
0.5x10-4
0. 5x10-3
0.089
0.5x10-4
0.5x10-4
0. 5x10-3
0.089
0. 1x10-6
0.2x10-7
0. 1x10-6
0.2x10-7
0. 5x10-5
0.3x10-6
0.5x10-5
Moderate
Accident
0.874
0.057
0.126
0. 126
Severe
Accident
0.987
0.033
0.002
0.002
Large Release
Minor
Accident
0. 783
0. 1x10-5
0. 1x10-5
0. 1x10-2
0.006
0. 5x10-3
0.5x10-3
0. 5x10-2
0. 1x10-5
0^ 1x10-4
0. 1x10-3
0.089
0. 1x10-4
0. 1x10-4
0. 1x10-3
0.089
0. IxlO'6
0. 2xlO"7
0.1x10-6
0.2x10-7
0.5x10-5
0.3x10-6
0. 5x10-5
Moderate
Accident
0.874
0.057
0. 126
0. 126
Severe
Accident
0.987
0.033
0.002
0.002
Based on data in Reference 15, except where noted. Missing value for moderate and severe accidents, same as for minor.
Data derived in Table 31.
CEstimated by engineering judgment. . '
-------
TABLE 38.
FAULT TREE INPUT DATA FOR SHIPPING CONTAINER OF HIGH LEVEL
RADIOACTIVE SOLID WASTE OR NOBLE GAS ON RAILa
No.
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
Mechanism
Impact
Thermal
Puncture
Vibration
Equipment
Defects
Human
Error
Input Event Name
Impact Occursb
Impact > Can or CylinderC
Impact > Caskc
Impact > Closure Scale
Fire Occursb
Pressure > Can or Cylinders
Pressure > Caskc
Pressure > Closure SeaK
Temperature > Can or Cylinder0
Temperature > Caskc
Temperature > Closure SeaK
Puncture Occurs0
Puncture > Can or Cylinders
Puncture > Caskc
Puncture > Closure Sealc
Puncture > Cooling SystemC
Coolant Leaks Outc
Vibration Occurs
Vibration > Can or Cylinder
Vibration > Caskc
Vibration > Closure Seal
Defective Seal
Inadequate Coolant
Improper Closure
Conditional Release Probability
Small Release
Minor
Accident
0.208
0.2x10-5
0.2x10-5
0.2x10-2
0.009
0.4x10-2
0.4x10-2
0.04
0.8x10-5
0.8x10-4
0.8x10-3
0.693
0.2x10-3
0.2x10-3
0.2x10-2
1.0
0.693
0. 1x10-6
0.2x10-7
0. 1x10-6
0.2x10-7
0.5x10-5
0.3x10-6
0.5x10-5
Moderate
Accident
0.234
0.068
0.763
0.763
Severe
Accident
0. 194
0.449
0.800
0.800 .
Medium Release
Minor
Accident
0.208
0. 1x10-5
0.1x10-5
0.1x10-2
0.009
0.2x10-2
0.2x10-2
0.02
0.4x10-5
0..4xlO-4
0.4x10-3
0.693
0.1x10-3
0.1x10-3
0.1x10-2
0.693
0.1x10-8
0.2x10-9
0. 1x10-8
0.2x10-9
0.5x10-7
0.3x10-8
0.5x10-7
Moderate
Accident
0.234
0.068
0.763
0.763
Severe
Accident
0. 194
0.449
0.800
0.800
Large Release
Minor
Accident
0.208
0.2x10-6
0.2x10-6
0.2x10-3
0.009
0.4x10-3
0.4x10-3
0.4x10-2
0.8x10-6
0.8x10-5
0.8x10-4
0.693
0.2x10-4
0.2x10-4
0.2x10-3
0.693
0. 1x10-8
0.2x10-9
0. 1x10-8
0.2x10-9
0.5x10-7
0.3x10-8
0.5x10-7
Moderate
Accident
0.234
0.068
0.763
0.763
Severe
Accident
0. 194
0.449
0.800
0.800
Based on data and footnotes in Table 37 with the following adjustments:
Numbers Multiplier
2-4 0.2
. 6-11 0.8.
13-15 2.0 __
19-Z1 0.01 ' -.Jr-7
-------
CALCULATIONS OF RELEASE PROBABILITIES
The logical relations and probability values in the fault tree descriptions
of the shipping containers and transport vehicles under study have been
programmed onto a computer. The probability that a particular sequence
of events in an accident of given severity leads to a loss of containment,
called the release probability in this report, is given by the appropriate
combination of input probability values according to the rules of Boolean
algebra. In particular, the probabilities for those sequences consisting
of only a few events are calculated exactly in the program that was used.
For sequences with a larger number of components, Monte Carlo simula-
tion of statistical failure rates was used in the program to give the
release probabilities. The simulations are carried out by examining the
fault tree with randomly failed components in a large number of trials.
An approximation to the release probability for each sequence is obtained
from a count of the number of containment failures in the computer
trials, from biasing factors, and from input data.
Failure Sequences
Selections from the complete set of calculations of failure sequences for
each of the shipping containers are presented in Tables 39, 40, and 41.
Only those sequences with probabilities greater than one percent of the
largest value in each set are listed in the tables. The program computes
many more sequences of lesser probability, but in all accident situations
studied, no more than twenty sequences were required to include all
sequences greater than one percent of the largest value.
In the case of a spent fuel cask, for example, the most probable sequence
of events leading to a release is described in Table 39. The identification
code of the component events of a fault tree in Table 39 is keyed to that of
Table 32. An accident occurs in which the corrugated jacket containing
the neutron shield water is punctured, causing the loss of the water. This
is represented by Event 21, which in Table 39 is given the name "coolant
leaks out" with a probability equal to that of Event 16, "puncture occurs. "
Loss of the neutron shield implies that heat from the spent fuel is not
dissipated adequately. The resultant internal heat source raises the
temperature of the cladding and the closure seal. Release of radioactive
material from the shipping cask occurs only when both the cladding and
the closure seal fail, according to Figure 19. These failure events are
represented by the code numbers 11 and 14, respectively.
Summary of Release Probabilities
The conditional releas.e probability is the fraction of accidents yielding
release. This probability of release incorporates both the conditional
84
-------
TABLE 39. SELECTED RELEASE SEQUENCES FOR SPENT FUEL SHIPPING CONTAINER3
Conditional Release Probability
Sequence of Events
11 14 21
11 14 16 20
6 11 14
11 15 16 20
11 15 21
7 14 16 20
7 14 21
1 5 11 16 20
11 16 20 29
11 16 20 26
11 21 29
11 21 26 -
11 14 29
6 11 15
611 26
6 11 29
6 7 14
11 14 26
Truck
Minor
Accident
'Small
Release
0.9x10':*
0.9x10';?
0.6xlO"Z
0.9x10"?.
0.9x10";?
0.9x10"?.
0.9xlO~b
Moderate
Accident
Small
Release
0.1x10'^
0.1x10'^
0.6x10';?
0.1x10";
O.lxlO"4
O.lxlO'J
0.1x10"^
Medium
Release
0.3x10':*
0.3x10':?
o.ixio';?
0.3xlO~b
0.3xlO'b
0.3xlO'b
0.3xlO'j?
0.3x10
OHIO'S
0.3xlO"b
Severe
Accident
Medium
Release
0.6xlO~j?
0.6xlO"b
O.SxlO"4
r*
0.1xlO"b
0.9x10"°
0.9x10'°
0.9x10'°
0.9xlO'b
Large
Release
0.2x10'*?
0.2x10')?
0.3xlO~b
-j
o.5xio~;
0.3x10"'
0.2x10'°
0.2x10'°
0.3xlO~;
0.5xlO"7
Rail
Severe
Accident
Large
Release
0.5xlO"J
0.5xlO'J
o!5xlO"^
0.5x10"°
/•
0.5xlO"b
Probability values for certain release sequences that are less than one percent of the greatest
probability value appearing in each column are not entered in the table.
-------
TABLE 40. SELECTED RELEASE SEQUENCES FOR PLUTONIUM SHIPPING CONTAINER
00
Sequence of Events
1234
12 13 14 15
1 3 4 22
1 3 4 20
1 3 4 21
1 3 12 13 15
1 2 3 12 15
1 4 12 13 14
1 .2 4 12 14
1 3 4 12 13
5678
13568
13456
12 14 15 22
12 14 15 21
12358
1 3 5 6 11
1 4 5 6 10
5 6 10 11
1 2 5 8 10
1 2 12 14 15 -
5 6 12 14 15
5 8 12 13 14
5 7 12 13 15
5 6 8 12 14
5 6 7 12 15
5 11 12 13 14
5 10 12 13 15
5 6 11 12 14
5 7 8 12 13
5 6 10 12 15
5 6 7 11
5 6 8 10
5 10 11 12 13
5 7 11 12 13
5 8 10 12 13
Conditional Release Probability
Truck
Minor
Accident
Small
Release
0.4xlO-4
0.4xlO-5
0.4x10-5
0.4x10-5
0.4x10-5
0.3x10-5
0.3x10-5
0.3x10-5
0.3x10-5
0.3x10-5
0.6x10-6
0.5x10-6
0.5x10-6
0.4x10-6
0.4x10-6
Severe
Accident
Small
Release
O.SxlO"4
0.5x10-5
0.5x10-5
0.5x10-5
0.3x10-5
0.3x10-5
0.2x10-5
0.2x10-5
0.2x10-5
0.8x10-6
0.8x10-6
Minor
Accident
Large
Release
0.4xlO-7
0.4x10-9
0.4x10-9
0.4x10-9
0.3x10-8
0. 3x10-8
0.3x10-8
0.3x10-8
0.6x10-9
0.5x10-9
0.3x10-8
Rail
Minor
Accident
Large
Release
0.3x10-6
0.6x10-8
0.6x10-8
0.6x10-8
0.2x10-8
Severe
Accident
Large
Release
0.3x10-6
0.6x10-8
0.2x10-7
0.6x10-8
0.6x10-8
0.6x10-7
0.6x10-7
0.5x10-7
0.5x10-7
0.3x10-7
0.3x10-7
0.2x10-7
0.2x10-7
0.2x10-7
0. 1x10-7
0. 1x10-7
0.6x10-8
0. 1x10-7
0. IxlO-7
Severe
Accident
Small
Release
0.3x10-3
0.6x10-5
0.6x10-5
0.2x10-4
0.6x10-5
0.6x10-5
0. 1x10-3
0.6x10-4
0.6x10-4
0.5x10-4
0. 5x10-4
0.3x10-4
0.3x10-4
0.2x10-4
0.2x10-4
0.2x10-4
0. 1x10-4
0. 1x10-4
0.6x10-5
•Probability values for certain release sequences that are less than one percent of the greatest probability value appearing in each column are not entered in the
table.
-------
SELECTED RELEASE SEQUENCES FOR HIGH LEVEL, RADIOACTIVE SOLID WASTE OR
NOBLE GAS SHIPPING CONTAINER21
Sequence of Events
6 8 16
6 8 12
6 7 16
6 12 15
8 12 13
6 11 16
6 11 12
568
1 4 6 12
1 4 6 16
6 7 12
567
5611
6 21
6 23
1 4 6 16
124
Conditional Release Probability
Rail
Minor
Accident
Small
\ Release
0. 1x10-3
0. 1x10-3
0. 1x10-4
0. 6x10-5
0.6x10-5
0.2x10-5
0.2x10-5
0. 1x10-5
0. 1x10-5
0. 1x10-5
0. 1x10-4
Moderate
Accident
Small
Release
0. 1x10-3
0. 1x10-3
0. 1x10-4
0. 6x10-5
0. 6x10-5
0.2x10-5
0.2x10-5
0. 1x10-4
0. 1x10-5
Medium
Release
0.3x10-4
0.3x10-5
0.2x10-5
0.2x10-5
0.6x10-6
0.6x10-6
0. 3x10-5
0.4x10-6
-
Severe
Accident
Medium
Release
0.3x10-4
0.3x10-4
0.2x10-5
0.2x10-5
0.6x10-6
0.6x10-6
0.2x10-4
0.2x10-4
0.4x10-6
•
Large
Release
0. 1x10-5
0. 1x10-6
0.6x10-7
0. 6x10-7
0.3x10-7
0.3x10-7
0. 7x10-6
0. 1x10-6
0. 7x10-7
0. 1x10-7
Truck
Severe
Accident
Large
Release
0.6x10-8
0.6x10-8
0.8x10-7
0. 1x10-8
0.2x10-8
0. 2x10-8
0.2x10-8
0. 1x10-8
0. 1x10-8
aProbability values for certain release sequences that are less than one percent of the greatest
probability value appearing in each column are not entered in the table.
00
-------
probability of an accident severity given an accident and the conditional
probability of a container damage severity given an accident severity.
The conditional release probabilities for spent fuel, recycled plutonium,
noble gas and high level radioactive solid waste shipping containers have
been calculated. Results are presented for a variety of accident
conditions in Table 42.
The values in Table 42 amount to calculations of sensitivity of a basic
failure probability data set to changes in accident conditions and to
changes in the logical connections of the data. The release probabilities
for plutonium and solid-waste containers are not significantly different.
Compared to the spent fuel shipping container that was modeled, the
plutonium and waste containers are calculated to be less likely to lose
their contents. This comparison is largely related to the internal heat
source in the spent fuel container that is lacking in the plutonium
container and the higher probability assigned to fuel cladding failure than
that assigned to waste can failure.
For the transportation scenario adopted in Section V, the fraction of
accidents of given severity associated with damages of given severity is
summarized in Table 43. On the average, these data indicate that about
one percent of spent fuel accidents yield releases, and about 0.02
percent of accidents to other shipments yield releases.
The most severe cases that would involve the highest release, large
damages in severe accidents, occur in about 0.01 percent of spent fuel
accidents, about 4 x 10 of recycled plutonium accidents, and about
2 x 10 of accidents to high level radioactive solid waste shipments and
noble gas shipments. The releases shown to be of greatest frequency for
most material shipment accidents are those characterized by small
damages in severe accidents.
88
-------
TABLE 42. RELEASE PROBABILITIES FOR SHIPPING CONTAINERS
Shipping
Container
Spent Fuel
Recycled
Plutonium
High Level
Radioactive
Solid Waste
or Noble
Gas
•
Transport
Mode
Truck
Rail
Barge
Truck
Rail
Truck
Rail
Release
Severity
Small
Medium
Large ,
Small
Medium
Large
Small
Medium
Large
Small
Medium
Large
Small
Medium
Large
Small
Medium
Large
Small
Medium
Large
Conditional Release Probability
Minor
Accident
1.9E-3
4.7E-4
2.1E-5
9.2E-3
2.3E-3
9.7E-5
6.7E-6
1.7E-6
4. 3E-8
8.2E-5
8.1E-6
6.0E-8
3.0E-4
3.8E-5
3.0E-7
5.9E-5
1.4E-5
5.3E-7
2.5E-4
3.9E-5
2.6E-6
Moderate
Accident
3.2E-3
8.1E-4
3.4E-5
l.OE-2
2.7E-3
1.1 E-4
4.0E-3
l.OE-3
4.3E-5
1.5E-4
1.6E-5
1.2E-7
4. 1E-4
5.0E-5
3.9E-7
9.4E-5
2.4E-5
2.2E-7
3.0E-4
4. 2E-5
4.4E-7
Severe
Accident
4. OE-4
l.OE-4
4. 4E-6
1.3E-2
3. 4E-3
1.5E-4
1. 1E-2
2.7E-3
1.2E-4
7. 9E-5
9.7E-6
7.2E-8
8. 5E-4
8.8E-5
7.6E-7
1.8E-5
3. 1E-6
l.OE-7
3.7E-4
9.0E-5
2.6E-6
Average3"
Accident
2. IE -3
5. 3E-4
2.3E-5
9. 3E-3
2. 3E-3
9.8E-5
5.5E-4
1.4E-4
5.9E-6
9.4E-5
6. 6E-6
7. 1E-8
3.1E74
3.9E-5
3.1E-7
6. 5E-5
1.6E-5
V4. 7E-7
2. 5E-4
3.9E-5
2.4E-6
Accident severities are weighted for each transport mode as shown in
table 31. "
89
-------
.TABLE 43. RELEASE PROBABILITY FOR
ADOPTED TRANSPORTATION SCENARIO'
Shipping
Container
Spent
Fuel
i -ab
Recycled
Plutonium
High Level
Radioactive
Solid Waste
or Noble -
Gasc
Release
Severity
Small
>v
Medium
Large
Small
Medium
Large
Small
Medium
Large
Conditional Release Probability
Minor
Accident
8E-3
2E-3
8E-5
2E-4
2E-5
4E-7
2E-4
3E-5
2E-6
Moderate
Accident
9E-3
2E-3
1E-4
3E-4
3E-5
3E-7
2E-4
4E-5
4E-7
Severe
Accident
1E-2
3E-3
1E-4
5E-4
5E-5
4E-7
3E-4
7E-5
2E-6
Average
Accident
8E-3
2E-3
9E-5
2E-4
2E-5
2E-7
2E-4
3E-5
2E-6
a
Distribution of transport modes assumed to be 10 percent
trucks, 85 percent rails, and 5 percent barge-truck combina-
tions (split 75 percent barges and 25 percent trucks).
Distribution of transport modes assumed to be 50 percent
trucks and 50 percent rails.
c
Distribution of transport modes assumed to be 25 percent
trucks and 75 percent rails.
90
*-»•
-------
SECTION VII
CONSEQUENCES OF RADIOACTIVITY RELEASES
INTRODUCTION
Among the consequences of a release of radioactivity from a nuclear
transportation accident are: quantity of radioactivity released, and population
dose. The risk of a release from a shipment of radioactive material is the
expected value of the consequences of the release.
In an accident leading to a release, only a fraction of the total cargo of
radioactivity is assumed to be released. The severity of container dam-
age directly determines the release fraction of the radioactive contents
and thus, the quantity of radioactivity released to the environment.
The released radioactive material may expose the population through a
variety of mechanisms:
1. Solid or liquid material spilled on the ground near the
accident.
2. Gaseous or particulate material dispersed into the
atmosphere.
3. Fallout of gaseous or particulate material dispersed into
the atmo s phe re.
4. Contamination of potable water supplies or the food chain.
The population dose may be acquired by at least four important pathways:
1. External exposure.
2. Immersion in a radioactive cloud.
3. Inhalation from a radioactive cloud.
4. Ingestion through the food chain.
The population dose is thus dependent on the radioactivity, chemical, and
physical properties of the shipment; the population distribution; and
quantity of released material transported from the accident to the
population.
91
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Public health effects can be determined from the population doses and
published information on the biological effects of ionizing radiation. The
health response can include illness and fatalities from acute and long-
term effects, somatic and genetic.
.RELEASE FRACTIONS
In an accidental release of radioactivity from a shipment, not all the
radioactive contents can be expected to leave the shipping container.
Estimates of the amount of radioactivity that actually escapes from an
accidently opened container depends on the nature of radioactivity under
consideration. For the purposes of estimating the risk in a conservative
manner, the release fractions for all accidents are estimated for large
releases assuming easily dispersible materials. The values of the release
fractions that are applicable to accidents leading to small and medium
releases are estimated as fractions of the large values. The scenarios
discussed below provide an upper limit estimate of the public risk from
accidental releases from shipments of the materials under study.
Spent Fuel
In accidents involving spent fuel shipments, the most likely releases
involve gaseous fission products or contaminants present in the cask
coolant or leached through the surface of the fuel cladding (Reference 6).
Fission gases tend to accumulate in the void space within the fuel rods
during reactor operation and during the cooldown period-after discharge
from the reactor. Certain solid fission products diffuse from the fuel
matrix and concentrate on the inner cladding surface. Upon cladding
perforation, these products are available for release from the fuel element.
A possible source of contaminants in a release is the cask coolant to which
fuel elements in the cask are exposed. In LWR spent fuel shipments, the
coolant is taken from the water in the storage pool containing the fuel
elements during the cooldown period. This situation occurs because the
shipping cask is opened, loaded and closed while submerged in the storage
pool water. Since some fuel elements incur cladding failures during reactor
operation, some radioactive fission product contaminants naturally enter
the water. These contaminants are available for release in the event of
shipping cask failure.
One possible accident to a spent fuel shipment involves the heating and
subsequent perforation or melting of the fuel cladding. Under these con-
ditions, all the fission product gas accumulated in the void spaces and some
of the other fission products are available for release. Negligible amounts
of actinides, activation products, and corrosion products are assumed to be
92
-------
released in these circumstances (Reference 18). The quantity of fission
products available for release is assumed to be 0. 1 percent of the inventory
in the fuel elements being carried. An independent estimate based on
consideration of leaching processes in the outer thirty microns of the
uranium oxide fuel used in LWRs and LMFBRs holds that no more than
0. 01 percent of the fission product inventory is available for release.
However, since no experimental confirmation exists for this number, the
higher value is used. The solid fission products of greatest concern in the
release are cesium and ruthenium, due to their relatively long half lives
and because of their volatility.
The principal volatile fission products released from spent fuel are Kr-85
and 1-131. Approximately thirty percent of the Kr-85 in the fuel element
inventory is estimated (Reference 18) to migrate to the void space in the
fuel element, and approximately two percent of the 1-131 is estimated
(Reference 18) to collect there.
Recycled Plutonium
Plutonium recycled from spent fuel to freshly fabricated fuel will likely
be shipped from chemical processing plants to fabrication plants in oxide
powder form. In this form, Pu can be economically mixed with natural
uranium before being sintered into pellets that will be assembled into fuel
elements. In an accident, the containment vessel lid can be broken or
warped away from the vessel walls and some Pu powder released from the
vessel, Celotex insulation, and drum. Since a release fraction is not"
available from the literature, a fraction of 0. 1 percent is used typical of
nonvolatile materials.
High Level Radioactive Solid Waste
A possible accident to a shipment of high level radioactive solidified waste
would involve immediate loss of the cask cooling capability and the breakage
or puncture of several of the waste canisters. The fraction of waste
released in such an accident is complicated by the cooling requirement. If
the waste melts, significant fractions of the cesium and ruthenium radio-
nuclides may be released (Reference 19). Since a release fraction for such
an accident is not known, a value of 0. 5 percent is arbitrarily chosen as
the release fraction for mixed fission products.
Noble Gas
Of the gaseous fission products, only Kr-85 is generated in sufficiently
large quantities and has sufficiently long half life to merit consideration
for separation from spent fuel elements, accumulation, and transportation
93
-------
from the chemical processing plants to the Federal Waste Repository.
This gas would probably be shipped in standard cylinders pressurized
to 2200 psig. In a credible accident, any loss of containment will mean
a total release of the gas in a shipment.
When one of the gas cylinders is breached and the cask is breached, all
the gas from that cylinder will eventually be lost from the shipment.
However, not all the cylinders of a shipment necessarily lose their con-
tainment property in an accident. The concept of release fraction is
restricted to refer only to the number of damaged cylinders in the case
of noble gas accidents, whereas its meaning in the case of other ship-
ments encompasses not only the number of damaged containers within a
single shipment, but also the released fraction of the contents of a single
container. Since a noble gas shipment carries an integral number of
cylinders by hypothesis, the release fraction in accidents leading to
releases less severe than a large release must be smaller by an integral
number of cylinders. In particular, a small release fraction can be no
smaller than one-sixth of the shipment.
f
Summary of Release'Fractions
Release fractions in transportation accidents by material and severity are
summarized in Table 44. The table also lists the release fraction averaged
by transportation mode.
DISPERSION OF RADIOACTIVITY
Accidental releases of radioactivity from shipments may endanger the
population through a variety of mechanisms. Releases from spent fuel,
plutonium, and solidified waste shipments might contain solid material that
spills onto the ground and provides a ground deposit source of external radia-
tion. The most serious example of such a release is for part or all of a
spent fuel element to exit its shipping cask. Such releases have been discussed
elsewhere (Reference 18), and will not be considered further in this report.
In all the. shipment accidents under consideration, the primary source of
radiation exposure is assumed to result from atmospheric dispersion.
Radionuclides may leave the shipping containers as gases or particles
small enough to be transported through the atmosphere. These constitute
clouds of airborne sources of external radiation. Only part of the solid
material spilled from the container will be dispersed away from the
scene of the accident by the atmosphere, since some particles of the solid
material may be too large or too dense for airborne transport. The release
fractions used in this report (Table 44) apply to releases of easily dispersible
material.
94
-------
TABLE 44. RELEASE FRACTIONS IN TRANSPORTATION ACCIDENTS
in
Material
Transported
Spent Fuel
Recycled
Plutonium
High-Level
Solid Waste
Noble Gas
Released
R adionuclide
Kr-85
1-131
Fission Products
Pu
Fission Products
Kr-85
Release Fraction
Large
Release
0.30a
0. 02a
0. 001
0. 001
0. 005
1. 0
Medium
Release
0. 15
0. 01
0. 0005
0.0005
0. 0025
0. 5
Small0
Release
0. 003
0.0002
0.00001
0. 00001
0.00005
0. 17
Average**
Release
0.034
0. 0023
0. 00011
0.000054
0. 00049
0.23
Reference 18.
Assumed to be one -half the value for large releases. .
Assumed to be one percent the value for large releases.
Weighted for each release severity for the transportation scenario adopted
(Table 43).
-------
The public may also absorb radiation by means other than external radiation
sources. Some of the radioactive particles transported from the scene of a
transportation accident to positions occupied by the public may become
sources of radiation inside human bodies through the pathways of inhalation,
ingestion, or open wounds. Of the possible internal pathways, only the
inhalation pathway will be considered in this report. Plutonium is assumed
to be insoluble.
Atmospheric Dispersion Model
In this study, the standard Gaussian plume diffusion model (Reference 20)
was used to compute atmospheric transport of radioactivity. In this
model, possible kinds of weather are characterized by seven Pasquill
stability categories, A through G, with associated weather probabilities
and average wind speeds. Published calculations (Reference 18) were used
to simplify the computations. In every case, it was assumed that the
radioactive material is released in a very short period of time at ground
level.
For worst case assumption in nuclear safety analysis, it is common to
use very conservative dispersion coefficients corresponding to Pasquill
Type F, moderately stable meteorological conditions. For this study,
the more realistic assumption of average weather conditions was used.
The average value lies between Pasquill Types E and F. For some
calculations, the weather condition occurring with the greatest probability, '
Pasquill Type D, was used. Pasquill Type G stability category was not /
used because of its low frequency, short duration, and calm wind /
conditions.
Dose Coefficients
The meteorological diffusion model calculates the concentration at a
receptor on the ground as a function of distance, direction and time from
radioactivity released in a very short period of time at ground level. The
exposure to the receptor is expressed in terms of the integrated concen-
tration resulting from the passage of the airborne cloud of radioactivity.
The dose coefficient is the factor which transforms the integrated con-
centration of radioactivity at the receptor position into a dose absorbed
by the receptor.
The dose coefficient for external radiation sources depends on the energy
of the radiation under consideration, the energy absorption coefficient of
the absorbing medium, the scattering properties of the absorbing medium
and the geometry of the emitter-receptor situation. The coefficient for
internal radiation sources must also include the mass of the critical organ
96
-------
under consideration, information on intake, retention, distribution, trans-
location, and the radioactive decay rate of the radionuclide under considera-
tion. Dose coefficients are given in Table 45 for several critical organs
for radionuclides in the four materials undergoing transportation.
TABLE 45. DOSE COEFFICIENTS
Material
Transported
Spent Fuel
Recycled
Plutonium
High-Level
Radioactive
Solid Waste
Noble Gas
Nuclide
Kr-85
1-131
Fission
Products
Pu 238-242
Fission
Products
Kr-85
Dose Coe
Whole
Body
1.2E-la
1.2E-la
1.2E-la
1.2E-la
Lung
1.6E2b
1.2E4t>
1.6E2C
3
rr • i /rem-m \
\ Ci-sec /
Gastrointestinal
Tract
1.6Elb
7.0EOb
Thyroid
4. 76E2C
3. 18E2d
aReference 20
^Reference 6
cReference 18 - Child
dReference 18 - Adult
Individual Doses
In evaluating the consequences of radioactivity released from transporta-
tion accidents, it is necessary to calculate the dose that an individual
might receive. The dose at a receptor on the ground varies with the
distance from the release point in exponential fashion. The radioactive
material is distributed downwind from a release such that the isopleths
(contour lines of equal dose) are cigar shaped. The dose from an accident
97
-------
can be estimated by assuming a ground level release under average weather
conditions and calculating the exposure dose delivered along the centerline
of that pattern, i. e. , the direction in which the highest exposures would
occur.
The average dose in all directions from the accident, will be significantly
less than the centerline dose. For average weather conditions, this is
about five percent of the centerline dose at 50 meters (Reference 18).
Population Dose
The population dose resulting from radioactivity released from a trans-
portation accident depends on both the spatial distribution of the dose and
the number of persons exposed. The spatial distribution of the dose was
estimated from published data (Reference 18) for Pasquill Type D, the most
probable weather condition. The population dose is calculated by an
approximate integration of the spatial dose and the population density in
the vicinity of the accident.
The population density varies widely along the route of a shipment. Most
facilities in the nuclear fuel cycle are located remote from population
centers. Nuclear power plants, however, will be located nearer to
urban centers. The probability distribution of the population density
within 50 miles of presently operating reactors has been described
(Reference 18) for the 1980 time period. The average population density
near power plants is about three times the national average and i'si stated
to be probably typical of the Eastern United States. About 75% of the area
in a 50-mile radius is greater than the average United States population )
density; about 20% is greater then ten times; and about one percent is
greater than 100 times.
The average distances of shipments in the nuclear power industry postu-
lated in Section V of this report range from 410 miles for spent fuel in
2020 to 2, 175 miles for radioactive wastes in 1975. Spent fuel and plu-
tonium shipments, with an average shipping distance of 500 miles, pre-
dominate the transportation picture during the entire period. If accidents
were presumed to occur randomly along the route, the population density
distribution near nuclear power plants would be applicable to about ten per-
cent of the route. The balance of the population density is assumed to have
a distribution typical of the whole United States. About 35% of the land
area has greater than the average United'States population density; about
ten percent of the area greater than three times; about one percent greater
than 20 times; and about 0. 1 percent greater than 100 times. The population
98
-------
model is summarized in Table 46. The probability distribution of the
average population density in the vicinity of an accident as multiples of
the national average density is assumed to apply to all shipments and all
times studied in this report.
TABLE 46. AVERAGE POPULATION DENSITY
Multiple of the National
Average Population Density
<1
1- 3
3- 10
10- 20
20-100
>100
Probability
0.650
0.250
0.080
0.010
0.009
0.001
An estimate of the population dose within a reasonably large area of an
accident is facilitated by zoning the area according to population distribution.
For an average accident site, the wind direction and population distribution
are assumed to be independent and thus, the zones are assumed to be
circular areas. No one will ordinarily be located within the zone containing
the wrecked cargo. Some individuals may happen to enter this zone, which
is assumed to extend radially from the accident to form a circle with an
outer radius of about 500 feet. A uniform density of population (Table 1) is
assumed to exist in the rest of the area, which is assumed to extend radially
from the accident to form a circle of area 10 mi . The outer radius of
this assumed population environment is thus about 18 mi.
99
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SECTION VIII
RISKS FROM TRANSPORTATION ACCIDENTS
INTRODUCTION
The risk to the public from transportation accidents to shipments of
radioactive material from the nuclear power industry takes into account both
the magnitude or value of the consequences of the accidents and their fre-
quency of occurrence. Two types of consequences have been analyzed in this
study; the quantity of radioactivity released and, the radiation doses resulting
from the releases. Since many events can lead to the same consequence, risk
is defined by the average or expected value of the consequences of release,
i.e., the probability weighted sum of all values of a consequence. Risks are
I-
projected for the next fifty years.
The estimates of risks to follow begin with transportation accidents: the
frequency of accidents and the frequency of radiological releases. A
scenario is postulated in which a hypothetical individual is exposed to a
release from a transportation accident at a distance of 0. 1 mile. The
exposures are presented as annual averages, i.e. , averaged over the
variety of transportation modes, accident severities and package damage
severities, and as worst cases for accidents which occur very infre-
quently. This is followed by the calculation of population doses. Finally,
the risks from transportation accidents in the nuclear power industry are
.compared to other sources of radiation exposure.
ACCIDENT FREQUENCY
The frequency of accidents was expressed in Section VI (Table 28) as the
number of accidents occurring in a million shipment-miles of transporta-
tion activity. The transportation scenarios given in Section V and the
projections given for the annual shipment-miles for transportation of
spent fuel (Table 12), recycled plutonium (Table 19), high level radio-
active solid waste (Table 24) and noble gas (Table 27), were used to
calculate the frequency of transportation accidents from 1975 to 2020.
The average number of accidents per year and the average time between
accidents are presented in Figure 22.
100
-------
10
Recycled
Plutonium
High-
Level
Solid
Waste
10
-1
10
0
10-
PI
-------
The total accident frequency is less than one per year in 1975, exceeds
one per month after 2000 and reaches two per month in 2020. Shipments
of spent fuel dominate the accident picture until about the year 2000. The
approximate accident rate for spent fuel transportation ranges from one
accident per 900 shipments in 1975 to one accident per 2000 shipments
in 2020. This reduction is due entirely to the reduction in shipping
distances during the period. The annual shipments of plutonium exceed
the annual shipments of spent fuel after 2005 because of the rapid growth
of the LMFBR. By 2020, the average number of accidents to shipments
of recycled plutonium exceeds one per month. The approximate accident
rate for plutonium transportation ranges from one accident per 700 ship-
ments in 1975 to one accident per 1700 shipments in 2020. The accident
rates for high-level solid waste and noble gas transportation are approx-
imately one accident per 500 shipments for the entire period.
It is important to note that in the overwhelming majority of cases, these
accidents will not result in the release of radioactivity. As indicated in
Table 43, only about one percent of the spent fuel accidents result in
releases and about 0.02 percent of the accidents involving other materials
result in releases. The effect of the release probability (for the trans-
portation scenario adopted) on the frequency of transportation accident
releases is shown in Figure 23. Also shown is the average time between
releases.
Releases from accidents involving spent fuel totally dominate the release
frequency. The release frequency ranges from approximately one
release per 250 years in 1975 to one release per 13 years in 2020. The
approximate release rate for spent fuel ranges from one release per
90,000 shipments in 1975 to one release per 200,000 shipments in 2020.
The average release rate for plutonium is approximately one release per
5,000,000 shipments. The release rates for high-level solid waste and
noble gas are approximately one release per 2,000,000 shipments for
the entire period.
The release fractions tabulated in Section VII (Table 44) were combined
with the transportation data in Section V and the release probabilities in
Section VI (Table 43) to calculate the amount of radioactivity released
from transportation activity in the nuclear power industry from 1975 to
2020. The results were averaged by mode, accident severity, release
probability and package damage severity. The average annual release of
radioactivity is summarized in Table 49T The largest average releases of
radioactivity occur from spent fuel.
The averaging process distributes the releases over all shipments; however,
only a small fraction of all shipments is associated with releases. The
102
-------
10
-1
lO-2
04
W
W
en 10-3
W
W
o •
04
W
PQ
10-4
W
O
s
w
10-5
10-6
1970
Recycled
Plutonium
High-Level
Solid
Waste
101
10-
w
04
<;
w
to
w
w
W
W
W
W
104 ~
W
O
w
10-
10C
1980
1990
2000
2010
2020
VV
FIGURE 23. FREQUENCY OF TRANSPORTATION ACCIDENT RELEASES
103
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TABLE 47. , AVERAGE ANNUAL RELEASE OF RADIOACTIVITY
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
.Average Annual Release (Ci
Spent Fuel
Kr-85
5.5E-1
1.4EO
2.9EO
5.7EO
8.8EO
1.0E1
1.2E1
1.3E1
1.3E1
1.3E1
1-131
7.4E-6
2.0E-5
3.7E-5
4. 4E-4
2.7E-3
8.9E-3
1. 7E-2
2.2E-2
2. 7E-2
3. 1E-2
Fission
Product
1.6E-1
4.6E-1
9.3E-1
1.4EO
2.2EO
3.4EO
4.4EO
5.5EO
5.9EO
6.3EO
Plutonium
3. 5E-4
9. 5E-4
2. 1E-3
2.9E-3
4.4E-3
7.6E-3
1. 1E-2
1 . 4E -2
1.6E-2
1. 7E-2
High-Level
Radioactive
Solid Waste
6. 5E-3
1.7E-2
3.7E-2
6.7E-2
8.6E-2
1.2E-1
1.6E-1
1.9E-1
Noble
Gas
2. 7E-1
7.6E-1
1.8EO
3.8EO
5.9EO
7. 6EO
1.0E1
1. 1E1
-------
actual release occurs very infrequently as indicated in Figure 23. When it
occurs, it will be larger than the average values shown in Table 49*
The average release fractions in Table 44 were applied to typical actual
shipments. Because the transportation scenario is assumed constant over
the entire period under study, the variation in radioactivity transported
per shipment is small and due only to the nuclear reactor mix postulated in
Section V. Consequently, the typical actual releases are relatively constant
from 1975 to 2020.
A typical actual release from an accident involving spent fuel (SF) would be
about 1000 curies of Kr-85, one curie of 1-131 and 400 curies of mixed
fission products (FP). A typical actual release from a plutonium shipment
accident would be about 10 curies and from high-level radioactive solid
waste (HSW) about 2000 curies. As discussed in Section VII, the smallest
actual release from a noble gas (NG) shipment can be no less than the
contents of one gas cylinder. This corresponds to about 200,000 curies of
Kr-85. The average frequency of these actual releases is presented in
Figure 23.
. INDIVIDUAL EXPOSURES
The meteorological information discussed in Section VII was used to
calculate the average dose to a hypothetical individual exposed on the
centerline downwind from an average ground level release from a transpor-
tation accident at a distance of 0.1 mile under average weather conditions
(Reference 18). The calculations were based on average annual releases
from 1975 to 2020 shown in Table 49 and the dose coefficients given in
Table 45. The results are presented in Table 50.
A more realistic calculation, however, would calculate conditional individual
doses from infrequently occurring actual releases corresponding to the
frequencies shown in Figure 23. The conditional dose was estimated for a
ground level release under average weather conditions at a distance of 0.1
mile on the centerline of the exposure pattern. The results are shown in
Table 51. The average conditional dose to all persons in all directions
from the accident would be about one percent of the centerline dose at 0.1
mile for average weather conditions.
Comparison of the results between Tables 50 and 51 show conditional doses
from actual releases significantly larger than average doses. Conditional
doses from spent fuel releases are less than conditional doses from high-
level solid waste releases. Since spent fuel releases occur much more
frequently (Figure 23) than high-level solid waste releases, the average
annual doses from spent fuel accidents are greater than the average annual
105
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TABLE 48. AVERAGE ANNUAL INDIVIDUAL DOSES AT 0.1 MILE.
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
•>
Material
SF
Pu
SF
Pu
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
Critical Organ Dose (rem)
Kr-85
Whole
Body
7.5E-5
1.9E-4
4.0E-4
3.8E-5
7.8E-4
l.OE-4
1.2E-3
2.5E-4
1.4E-3
5.2E-4
1.6E-3
7.6E-3
1.8E-3
l.OE-3
1.8E-3
1.4E-3
1.8E-3
1.6E-3
Fission Products and Pu
Whole
Body Lung
2.2E-5
6.~3E-5
1..3E-4
8.9Er7
1.9E-4
2.3E-6
3.0E-4
5.1E-6
4.6E-4
9.2E-6
6.0E-4
1.2E-5
7.5E-4
1.6E-5
8.1E-4
2.2E-5
8.6E-4
2.6E-5
2.9E-2
4.8E-3
8.4E-2
1.3E-2
1.7E-1
2.9E-2
1.2E-3
2.15E-1
4.0E-2
3.1E-3
4.0E-1
6.0E-2
6.7E-3
6.2E-1
l.OE-1
1.2E-2
8.0E-1
1.5E-1
1.6E-2
l.OEO
1.9E-1
2.2E-2
1.1EO
2.2E-1
2.9E-2
1.1 EO
2.3E-1
3.5E-2
GI Tract
2.9E-3
8.4E-3
1 ..7E-2
5.2E-5
2.6E-2
1.4E-4
4.0E-2
3.0E-4
6.2E-2
5.3E-4.
8.0E-2
6.9E-4
l.OE-1
9.6E-4
1.1E-1
1.3E-3
1.1E-1
1.5E-3
1-131
Thyroid
Adult
4.0E-6
1.1 E-5
2.0E-5
2.4E-4
1.5E-3
4.8E-3
9.2E-3
1.2E-2
1.5E-2
1.7E-2
Child
2.7E-6
7.2E-6
1.3E-5
1.6E-4
9.8E-4
3.2E-3
6.2E-3
8.0E-3
9.8E-3
1.1E-2
106
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TABLE 49. CONDITIONAL INDIVIDUAL DOSES AT 0. 1 MILE
Year
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Material
SF
Pu
SF
Pu
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
SF
Pu
HSW
NG
Critical Organ Dose (rem)
Kr-85
Whole
Body
1.3E-1
1.2E-1
1.3E-1
3.0E1
1.6E-1
3.0E1
1.8E-1
3.0E1
1.6E-1
3.0E1
1.6E-1
3.0E1
1.6E-1
3.0E1
1.6E-1
3.0E1
1.6E-1'
3.0E1
Fission Products1 and Pu
Whole
Body
3.4E-2
3.6E-2
2.7E-2
3.6E-1
3.6E-2
3.5E-1
4.0E-2
3.4E-1
4.8E-2
3.3E-1
5.3E-2
3.3E-1
6.3E-2
3.1E-1
6.6E-2
2.9E-1
7.0E-2
2.7E-1
Lung
4. 6E1
2.5E2
4. 7E1
2.5E2
4. 9E1
2.7E2
4. 7E2
4. 7E1
2.2E2
4. 6E2
5.3E1
1.5E2
4. 6E2
6.4E1
1. 1E2
4.4E2
7. 1E1
9. 1E1
4.4E2
8.4E1
8.7E1
4. 2E2
8. 8E1
8.3E1
3.8E2
9.3E1
8.4E1
3.6E2
GI Tract
4. 6EO
4. 7EO
4. 9EO
2. 1E1
4. 7EO
2.0E1
5. 3EO
2.0E1
6.4EO
1.9E1
'
7. 1EO
1.9E1
8.4EO
1.8E1
8.8EO
1.7E1
9. 3EO
1.3E1
1-131
Thyroid
Adult
3. 5E-3
3.8E-3
3. 5E-3
2.6E-2
1.2E-1
2.9E-1
4/6E-1
5.8E-1
7.2E-1
8. 1E-1
Child
1.3E-3
1.4E-3
1.3E-3
i
9.8E-3
4. 3E-2
1. 1E-1
1.7E-1
2.2E-1
2.7E-1
3.0E-1
"^tfga.
107
-------
doses from high-level solid waste accidents. The average annual doses
from Kr-85 released from.noble gas accidents are much smaller than the
average annual doses from Kr-85 released in spent fuel accidents. Although
the release frequency from noble gas shipments is very low, the conditional
releases are many times larger than the conditional releases of Kr-85 from
spent fuel and thus the actual dose, when it occurs, is significantly larger.
For comparison purposes, the worst-case releases in 2020 and their
expected frequencies of occurrence were calculated for large container
sizes in severe accidents with large release fractions. The results are
shown in Table 52. As expected, the releases and doses are significantly
greater than those in Table 51. These worst-case releases are approxi-
mately three to four orders of magnitude less probable than the conditional
values shown in Table 51.
POPULATION DOSES
The population dose was calculated by the method described in Section VII.
The results are presented in Figures 24 through 27 for whole body, lung, .
GI tract and thyroid doses, respectively, from 1975 to 2020. Spent fuel
exposures dominate the population doses in every category of exposure for
the entire period.
The annual population dose from transportation accidents in the nuclear
power industry can be compared with the annual doses received by the
population from natural and man-made radiation sources. The expected
value of the total annual whole body population dose from accidents in the
nuclear power industry in 2020 is less than 10 person-rem per year. The
routine exposure of the population from environmental sources such as
cosmic rays and natural radioisotopes in the earth's crust averages about
0. 125 rem per year in the coterminous United States. The principal source
of man-made radiation exposure has been from medical use of diagnostic
x-rays averaging 0. 1 rem per person per year. The total,estimated annual
population dose in the United States would be about 80 x 10 person-rem per
year in 2020. The average annual transportation accident population dose
is thus a very insignificant fraction of the total annual population dose
received from natural background and from other man-made radiation
sources.
The population doses shown in Figures 24 through 27 assume the uniform
population density of the United States presented in Table 1. The population
density distribution of Table 46, however, indicated that 35 percent of the
conterminous United States has a population density greater than average.
If an accident were to occur in an area where the population density is
108
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TABLE 50. WORST CASE INDIVIDUAL DOSES IN 2020 AT 0. 1 MILE
Material
Spent Fuel
Recycled
Plutonium
High-Level
Solid Waste
Noble Gas
Nuclide
Kr-85
1-131
FP
Pu
FP
Kr-85
Release
(Ci)
2.9E4
7.0E1
1.4E4
7.8E4
5.4E4
1.0E6
Frequency
. (10~6 yr-1)
11.00
11.00
11.00
0. 11
0. 16
1.4
Critical Organ Dose (rem)
Whole
Body
4. OEO
1.9EO
7.3EO
1.4E2
Lung
2.3E3
1.6E6
9.9E3
GI
Tract
2.6E2
4. 3E2
Thyroid
Adult
2.0E1
Child
7. 6EO
o
10
-------
10
-3
10
-4
c
10'5
5.
w
C5
W
W
OT
o
Q.
2-
o! io-6;
W
u
W
10
-7
10'8j
High-Level
Solid Waste
pent Fuel
Fission
Products
1970
1980
1990
ZOOO
2010
2020
FIGURE 24. ANNUAL WHOLE BODY POPULATION DOSE
110
-------
10°
10-1
w
w
rf
w
w
w _
o io-2
Q
I
H
P
On
O
w
o
w
10
-3
10-4
1970
High-Level
Solid
Waste
Spen
pent Fuel
Fission
Products
Recycled
Plutonium
1980
1990
2000
2010
2020
FIGURE 25. ANNUAL LUNG POPULATION DOSE
111
-------
W
>•<
5
H
o
w
ti
W
w
w
o
Q
O
>-l
H
D
A
O
A
W
W
10
-2
10
-3
10-4
10
-5
Spent Fuel
Fission
Products
High-Level
Solid
Waste
1970
1980
1990
2000
2010
2020
FIGURE 26. ANNUAL GI TRACT POPULATION DOSE
112
-------
10
1970 1980 1990 2000 2010 2020
FIGURE 27. ANNUAL, THYROID POPULATION DOSE
113
-------
greater than average, the population dose would be proportionately higher.
The probability of the higher exposure would be correspondingly less, based
on the population model of Table 46. For example, there is one chance in
50 that the population dose would be" ten times greater than the values
presented in Figures 24 through 27; there is one chance in a thousand that
the population dose would be 100 times greater.
The population doses shown in Figures 24 through 27 are annual averages
corresponding to the average annual releases of radioactivity presented
in Table 49. For comparison purposes, the conditional population doses
in 2020 from infrequently occurring actual^releases and their expected
frequencies of occurrence are shown iri Table 53.
114
-------
TABLE 53. CONDITIONAL POPULATION DOSES IN 2020
Material
Spent Fuel
'
Recycled
Plutonium
High-Level
Solid Waste
Noble Gas
Nuclide
Kr-85
1-131
FP
Pu
FP
Kr-85
Frequency
(yr"1)
0.075
0.075
0.075
0.0025
.00053
.000071
Critical Organ Dose (person-rem)
Whole
Body
3.9E-3
1.7E-3
7.8E-3
3.1EO
Lung
2.4EO
1.4E1
1.0E1
GI
Tract
*
2.4E-1
4. 5E-1
Thyroid
Adult
2.5E-2
Child
6.9E-3
-------
SECTION IX
REFERENCES
1. "The Nuclear Industry, 1969, 1970, 1971, " U. S. Government
Printing Office.
2. Nichols, J. P. et al. Projections of Radioactive Wastes to be
Generated by the U. S. Nuclear Power Industry. ORNL-TM-3965,
February 1974.
3. "Statistical Abstracts of the United States, 1972, " U. S.
Government Printing Office, July 1972.
4. Blomeke, J. O. and J. P. Nichols. Commercial High-Level
Waste Projects. ORNL-TM-4224, May 1973.
5. Potential Nuclear Power Growth Patterns. WASH-1098,
December 1970.
6. Siting of Fuel Reprocessing Plants and Waste Management
Facilities. ORNL-4451, July 1971.
7. Brobst, W. A., The Probability of Transportation Accidents, U. S.
Atomic Energy Commission, November 1972.
8. Atomic Fuel. U. S. Atomic Energy Commission, Division of
Technical Information, December 1964.
9. Environmental Survey of the Nuclear Fuel Cycle. U. S. Atomic
Energy Commission, Fuels and Materials, Directorate of
Licensing, November 1972.
10. Nichols, J. P. and J. O. Blomeke. Nuclear Fuel Reprocessing,
Fuel Transportation, and High-Level Radioactive Waste Storage.
Oak Ridge National Laboratory. Lectures given at UCLA,
July 1972.
11. Environmental Analysis of the Uranium Fuel Cycle: Part III -
Nuclear Fuel Reprocessing. EPA-52019-73-003-D, October 1973.
12. Fortescue, Peter. A Reactor Strategy: FBRs and HTGRs.
Nuclear News, April 1972.
116
-------
13. Transportation of Nuclear Fuel Material in the United States.
Report by the Nuclear Assurance Corporation, 24 Executive
Park West, Atlanta, Georgia, 1970.
14. Shappert, L. B., et al., 'Nuclear Safety _14 (6), 595(1973).
15. Garrick, B. J. , et al., "A Risk Model for the Transport of
Hazardous Materials, " Holmes & Narver, Inc., HN-204,
Anaheim, California, August 1969.
16. USAEC Safety Evaluation by the Transportation Branch,
Directorate of Licensing, General Electric Company,
Model IF-300, Shipping Cask, September 24, 1973.
17. Adcock, F. E. , J. D. McCarthy, and W. F. Wackier, Technical
Documentation for Model 2030-1 Shipping Container (DOT SPECIAL
PERMIT 5332), Dow Chemical U.S.A., Rocky Flats Division,
RFP-1867, Golden, Colorado, November 16, 1972.
18. "Environmental Survey of Transportation of Radioactive Materials
to and from Nuclear Power Plants, " WASH-1238, Directorate of
Regulatory Standards, U. S. Atomic Energy Commission,
December 1972.
19. Gera, F. , and D. G. Jacobs, "Considerations in the Long-Term
Management of High-Level Radioactive Wastes, " ORNL-4762,
February 1972.
20. Meteorology and Atomic Energy 1968, D. H. Slade, editor,
TID-24190, U. S. Atomic Energy Commission, Division of
Technical Information, July 1968.
21. The Effects on Populations of Exposure to Low Levels of
Ionizing Radiation, Report of the Advisory Committee on the
Biological Effects of Ionizing Radiation's, Division of Medical
Sciences, National Academy of Sciences, National Research
Council, November 1972. (Commonly referred to as the BEIR
report.)
22. Lushbaugh, C. C. , F. Comas, C. L. Edwards, G. A. Andrews,
"Clinical Evidence of Dose-Rate Effects in Total-Body Irradia-
tion in Man, " in the Proceedings of a Symposium on Dose Rate
in Mammalian Radiation Biology, Oak Ridge, Tennessee,
April 29 to May 1, 1968.
117
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