WASH 1209

THE POTENTIAL RADIOLOGICAL

IMPLICATIONS OF NUCLEAR

FACILITIES IN THE UPPER

MISSISSIPPI RIVER BASIN IN THE

YEAR 2000

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A AIR ENVELOPE mEoNVE

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MINNEAPOLIS 

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STUDY MADISON DETROIT

SOxCT REONCHICAGO. CLEVELAND

NORTH PLATTE DESNE

OMAHA AIR ENVELOPE J COLUMBUSREG ION

ST.LOUIS

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LITTLE 

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J N A Y 1

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DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

U.S. ATOMIC ENERGY COMMISSION

metadc784182

Study

WASH 1209

UC-12

THE POTENTIAL RADIOLOGICAL

IMPLICATIONS OF NUCLEAR

FACILITIES IN THE UPPER

MISSISSIPPI RIVER BASIN IN

THE YEAR 2000

The Year 2000 Study

JANUARY 1973

DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

U.S. ATOMIC ENERGY COMMISSION

For sale by the Superintendent of Documents, U.S.Government Printing Office

Washington, D.C., 20402

PREFACE

The increasing commitment by the utility industry to the use of nuclear reactors for the

generation of electrical power in the United States clearly indicates that nuclear power has

become a principal option for helping to satisfy the national need for a clean, safe and reliable

energy supply. There are many factors that may be associated with the selection of any of the

clean energy options for widespread use. These factors can best be assessed through the

cooperative efforts of those involved in planning for the use of energy in this country. To this

end, the Atomic Energy Commission (AEC) has had underway a number of interrelated and

interdependent studies that consider the environmental effects and implications of the use of

civilian nuclear power plants and the associated manufacturing, shipping, reprocessing and

waste disposal activities. These studies also include examination of alternatives to nuclear

power, the economics of introducing nuclear power in various combinations with other proven

types of power plants, and provide an assessment of the environmental implications of these

mixed economies.

As with all power generation methods, nuclear power plants impinge to some extent on the

biosphere. Although the controlled and accidental releases of relatively small amounts of

radioactivity from operating nuclear power and reprocessing plants have been maintained well

below specified limits, it is important that the proper measures be taken in sufficient time to

ensure that the anticipated future widespread use of nuclear power does not present an

infringement upon these limits nor a potential hazard to the plant operators or to the public.

Thus, among other efforts in this area, considerable emphasis has been placed on developing a

capability, including improved analytical methods, for assessing the potential radiological

impact of nuclear facilities individually and on a regional basis. Accordingly, the AEC initiated

a study in January 1970 to develop a computer model (HERMES) and the necessary data to

permit estimates through the year 2000 of the radiation potentially received* by individuals

and population groups as a result of various estimated radioactive release rates from reactors

and fuel reprocessing plants. This study is being conducted by the Hanford Engineering

Development Laboratory (HEDL) at Richland, Washington with the assistance of Battelle

Memorial Institute under the direction of the AEC Division of Reactor Development and

Technology and the Director of Regulation. Among its many objectives, this study is intended

to provide information that may be used to:

" Inform involved organizations and the public of the potential radiological impact of

widespread application of nuclear energy for the generation of electricity;

" Guide the AEC and the nuclear power industry in planning research and development

programs to reduce further the already low levels of radioactive waste emissions;

" Assist in the further development of design criteria and regulatory standards related to

radioactive waste and emissions;

" Aid in evaluating the effects of potential improvements in waste treatment and

engineering technology;

" Guide the nuclear industry and other agencies and organizations, such as the

Environmental Protection Agency (EPA), the Office of Science and Technology (OST),

the Federal Power Commission (FPC) and state agencies concerned with the environmental aspects of the electric power industry in planning for new nuclear facilities; and

" Provide improved understanding of the contributions of specific radionuclides to

radiation potentially received by the population.

For the purpose of this study, "potentially received" indicates the maximum radiation that an individual

might receive in the situation under consideration. The radiation actually received would probably be less.

The overstatement of radiation received is occasioned by the conservative assumptions relating to the living

habits of the affected population as well as those utilized in determining the movement of the radionuclides

in their particular pathways to man.

1
This report analyzes the radiation potentially received by the population of the Upper

Mississippi River Basin (UMRB) resulting from the operation of nuclear facilities with their

assumed radwaste treatment systems which might be located in this region in the year 2000.

The year 2000 was chosen as the base year for the study because it is far enough in the future

that nuclear energy can be projected to be a major, and perhaps dominant, factor in electric

generation; yet it is near enough to reflect the effects of decisions and commitments presently

being made. It is anticipated that nuclear power plants now under construction will still be in

operation at that time.

This initial study was focused on determining the relative importance of each of the

radionuclides and pathways considered in the study and obtaining an indication of the radiation

potentially received by the population in the UMRB. Calculations performed by the model are

dependent upon the types of nuclear facilities and radwaste treatment systems assumed,

uncertainties in the role of various processes in transporting radionuclides through the

environment leading to human exposure, and approximations required in the mathematical

modeling of these processes. In a general study such as this, it is virtually impossible to take

into account all variations which might occur; however, sensitivity studies were performed to

evaluate the effects of uncertainties in the parameters believed to be most significant in their

effects on dose.

The complex of nuclear power plants operating in the year 2000 was assumed for the study

to consist of Light Water Reactors (LWRs), both the Pressurized Water Reactor (PWR) and the

Boiling Water Reactor (BWR), and the Liquid Metal Fast Breeder Reactor (LMFBR).

Gas-cooled reactors were not considered in the study. Depending on a number of factors, the

inclusion of gas-cooled reactors in the study could change the results to the extent that their

radiological release characteristics differ from those of the reactor types considered. However,

data obtained in the past two years suggests that the use of gas-cooled reactors would not result

in an appreciable change in the overall conclusions of the study. The radwaste treatment

systems chosen for LWR nuclear power plants are similar to those which are presently being

designed to conform to the recent guidelines for radionuclide release from nuclear reactors (10

CFR 50 Appendix I). In the study the LMFBR power plants and reprocessing plants, besides

utilizing the radwaste system applied to LWRs, utilize on a standard basis a noble gas

absorption system and solidification of all liquid wastes. Since operating data are not available

from the use of such systems, anticipated performance was estimated based on analysis of

ongoing development programs, system components and typical estimated releases as contained

in Safety Analysis Reports (SARs) for current generation nuclear plants.

The radwaste treatment systems in future generations of nuclear plants are expected to

exhibit substantial decreases in radionuclide releases as compared with those assumed in the

study. The trend appears to be toward the adoption of treatment systems and operating

techniques which would result in the release of essentially no radionuclides except through

leakage. Liquid wastes may be concentrated and solidified, and gases may be retained or

"bottled" for long-term storage. In order to examine the effects of this assumption, some

calculations were performed with the assumption that the reprocessing plants' radiation waste

treatment systems incorporated equipment which reduced the release of tritium by a factor of

100.

As with other studies pertaining to long-range applications, the reader is cautioned to bear in

mind that this study is broad in scope and includes many varying assumptions. Moreover, the

reader should note that future decisions on each plant type and each site location will be made

by individual utilities throughout the country, as has been the practice to date. The AEC

studies take cognizance of the fact that the utility industry, in conjunction with the appropriate

regulatory agencies, determines the extent and rate of commercial investment in electric

generating plants, taking into account the technical, economic and environmental options

available at the time each decision is made. The HERMES model, or segments of it, are being

utilized by the nuclear industry to assist in assessing the possible radiological impact of certain

plants.

Hypothetical extrapolations of individual utility decisions on plant siting, size and type have

been used in these studies mainly to identify trends, recognizing that the extrapolations are at

best uncertain and can lead to serious errors if used out of context. Further, it is equally

important to emphasize that such extrapolations are but one major element in any attempt to

assess the total range of environmental implications associated with energy growth projections.

For these reasons, the overall civilian power studies, including this one, will be frequently

revised and updated as the input data and implied results change with time. This will permit

ii
current examination of the environmental implications of the future nuclear industry including

the potential range of impact of the advanced reactor types. The range of possibilities may be

further explored by identification and incorporation of those detailed environmental effects

normally delineated in specific environmental statements. It is believed prudent to subject the

mathematical model and the results of the present study to continuing critical scrutiny. To this

end, the Oak Ridge National Laboratory (ORNL), in cooperation with the Tennessee Valley

Authority (TVA) and the National Oceanic and Atmospheric Administration (NOAA), has

critiqued the model. Recommendations for improving the model are being utilized in the

second application of the model, a study of the potential radiological implications of nuclear

facilities in the TVA region in the year 2000. In addition, study results will be compared to

actual field measurements when practical to validate all or part of the model. This may be difficult, if not impossible, in some instances because of the very low amounts of radionuclides expected to be present in any given area under normal conditions.

While this study concerns the potential exposure of the public to the radiation emitted

externally from nuclear facilities, the design and operating practices for them are also

controlled to ensure that potential radiation exposure to the operating and maintenance

personnel is as low as practicable. Similar consideration is also, of course, given to the

protection of industry personnel and the public for all other aspects of the nuclear fuel cycle;

uranium mining and milling, enrichment, fuel preparation and fabrication, and transportation

of fuels and wastes.

It is most important that those studies required to explore the potential environmental

impact of the options of providing for the Nation's energy demand be conducted on an

expedited basis. The legitimate concerns of the public that the quality of the environment be

preserved and that health not be endangered can only be alleviated by providing the public with

information that will allow a meaningful comparison of the total benefits and costs associated

with programs which are proposed as essential to the future of the Nation.

In the interest of improving the report, critical comment was solicited from selected

representatives of involved Government agencies, members of the nuclear community, and

interested public. Particular attention was given to obtaining comments from those who are

directly concerned with the environmental implications of the growth of nuclear power in the

UMRB.

The response to the request for comment was gratifying and we are deeply indebted to the

many reviewers for the time and effort spent in critically reviewing the draft report. In

preparing the report for final publication, all comments were carefully considered and

addressed to the maximum extent practicable. Many of the comments will also aid us in

improving future studies using the HERMES model, including the study of the potential

radiological implication of nuclear facilities in the TVA region in the year 2000.

We wish to thank all those that have submitted comments. Their efforts, which reflect their

interest in assuring that the Nation's energy needs are met without sacrificing the quality of the

environment, have resulted in an improved report.

Milton Shaw, Director

Division of Reactor Development and

Technology

iii

TABLE OF CONTENTS

Page

I. INTRODUCTION. ...................................................... I-1

II. SUMMARY ............................................................... II-1

Regional Estimates... ..................................................... II-1

Relative Importance of Pathways ............................................. 114

Radionuclide Contributions............................................... 114

Power Projections.........................................................11-8

System Design Operation Variants ............................................. 11-8

Effects of "Advanced" Treatment Systems.....................................I..11-8

Tritium Diffusion Effects......... .. .................................... II-11

Radionuclide Transport Considerations ........................................ II-11

Conclusions ............................................................. 11-12

III. DESCRIPTION OF STUDY AREA.............................................I..111-1

Physical Description ....................................................... III-1

Area Boundaries ...................................................... I..111-1

Topography ........... ................................... ....... .... 111-1

Hydrology ............................................ ........... 111-1 I.. 

M eteorology. ...................................................... 111-3

Power Supply and Demand ................................................ 111-3

Power Supply Characteristics ........ ................................... 111-3

Power Demand Projections ............ ................................. 111-3

Generation and Load Centers .............................................. 111-7

Seasonal Variation in Load............................................. 111-7

Population and Living Patterns............................................. 111-7

Population and Population Grouping.......................... .............. 111-7

The Centroid Concept .................................................... III-10

Food Production, Consumption and Transportation ............................ . III-10

Production.............................................................111-12

Consumption............................................................111-12

Food Sources..........................................................111-13

Recreation Patterns .. ................................................. 111-13

IV. POWER GENERATION PROJECTIONS.......................................... IV-1

Forecast of Capacity Additions................................................ IV-1

Projection of Generating Plant Mix.......................................... IV-1

Reprocessing Plant Requirements.. ........................................... N-3

Plant Siting Considerations................................................... IV-3

Monthly Average Plant Factor Synthesis ...................................... IV-3

V. RADIONUCLIDE RELEASE CONSIDERATIONS ................................. .V-1

Nuclear Power Plants ...................................................... V-I

Reprocessing Plants. ........................... ... ...................... V-3

"Source Map" Release Patterns................................................ V-9

v
TABLE OF CONTENTS (Continued) Page

VI. AIR TRANSPORT OF RADIONUCLIDES ...................................... VI-1

Diffusion................................................................VI-1

Plume Depletion...........................................................VI-2

Radionuclide Build-up on Surface.............................................. VI-2

Results of UMRB Study.....................................................VI-2

VII. WATERBORNE TRANSPORT OF RADIONUCLIDES.............................. .. VII-i

VIII. RADIATION EXPOSURE PATHWAYS..........................................VIII-1

Food Concentrations ...................................................... VIII-1

Calculations of Average Diets................................................. VIII-2

Consumption of Waterfowl and Game Birds..................................VIII-2

Consumption of Fresh Water Sports Fish...................................... VIII-3

Inhalation Rates...........................................................VIII-3

Recreation Habits..........................................................VIII-3

Calculation of Potential Radiation Dose.......................................... VIII-3

External Exposure.......................................................VIII-4

Air Submersion.........................................................VIII-4

Contaminated Ground.................................................... VIII4

Water Immersion........................................................VIII-5

Transpiration of Tritium.................................................. VIII-5

Inhalation.............................................................VIII-5

Ingestion..............................................................VIII-7

Dose Summation..........................................................VIII-7

IX. POTENTIAL RADIATION DOSE BASE CASE RESULTS............................. IX-1

Model Restraints..........................................................IX-1

General Trends and Patterns................................................ IX-2

Radionuclide Contributions to Potential Dose ................................... IX-8

Pathway Composition of the Potential Regional Dose ............................. . IX-8

X. SENSITIVITY STUDIES: EFFECTS OF PARAMETER VARIATIONS ..................-. X-1

Major Effects on Dose .................................................... X-1

Tritium Release Considerations.......................................... . X. -1

Effect of Inclusion of Chicago Area ....................................... ... X-3

Effects of "Air Envelope" Boundary Region ................................. ... X-3

Effects of Air Deposition on Water Concentrations ............................. .. X-3

Spent Fuel Cooling Time............................................... X-5

Plant Design and Operation Effects......................................... . X-5

Improved Treatment System Introduction Date ............................... ... X-6

BWR Gaseous Effluent Holdup Time.......................................... X-6

Noble Gas Treatment System Variants......................................... X-7

Factors Affecting Radwaste System Performance ................................. X-7

The Steam Generator Leakage ........................................... ... X-7

Recycling of PWR Floor Drain Flows ...................................... ... X-8

BW R Flow Variations................................................. X-8

Demineralizer and Evaporator Performance .................................. ... X-8

Variations in Plant Mix. .............................................. . X-8

Nuclear/Fossil Ratio................................................. . X-9

PW R/BW R Ratio. ................................................... X-9

Clustering of Plants ....................................................... X-9

Reprocessing Plant Size Variation......................................... X-10

Effects of Fuel Behavior In Fuel Diffusion Coefficients. ........................ X-13

vi
TABLE OF CONTENTS (Continued) Page

X. SENSITIVITY STUDIES: EFFECTS OF PARAMETER VARIATIONS (Continued)

Diffusivity Dependence on Burnup ........................................ X-13

Fuel Failure Probability............................................... X-13

Radionuclide Transport Considerations ....................................... X-14

Variations in Air Transport Parameters Mixing Depth and Stability .................... X-14

Wind Direction and Speed .............................................. X-14

Height of Release................................................... X-17

Rainfall Variation................................................... X-17

Water Transport Parameter Variations ...................................... X-17

Solution/Sediment Transport Relationships .................................. X-17

Runoff Variation................................................... X-17

Food Pathway Variants................................................. X-25

Depletion of Available Radionuclides ...................................... X-25

Effects of Dairy Management Techniques .................................... X-26

Relative Effects of Plant Uptake and Direct Deposition ........................... X-26

REFERENCES............................................................ R-1

APPENDICES

Appendix A: Glossary of Terms . . . . . . . . . . . . . . . . . . . . . . . .. ..................... A-1

Appendix B: Study Area Characteristics ........................................ B-1

Hydrology and Meteorology of Study Area ........................... B-3

Electrical Energy and Power Demand Projections ....................... B-18

Food Production and Food Consumption Patterns ....................... B-26

Appendix C: Potential Radionuclide Release Rates ................................. C-1

Appendix D: Potential Dose Rate Patterns in Study Area ............................. D-1

Relative Contribution of Various Radionuclides in Radiation Dose ............ D-3

Tritium ................................................ D-3

Krypton and Xenon ........................................ D-3

Strontium-90 ............................................ D-4

Zirconium and Niobium ..................................... D-4

Iodine-129 and - 131 ........................................ D-4

Cesium-134 and -137 ....................................... D-4

Other Radionuclides ........................................ D-5

Appendix E: Potential Radionuclide Concentrations and Dose Rate Comparisons for

Eight Typical Centroids .......................................... E-1

vii
LIST OF TABLES

Table No. Title Page

II-1 Ratio of (Radiation Potentially Received by Organ)/(Radiation Potentially Received

by Total Body)........................................................ 11-4

III-1 Annual Energy Demand Summary: Projection for Extended Study Area, 1970-2000.......-..III-7 111-2 Central Study Area Population.......................................... . III-10

III-3 Characteristics of Representative Centroids .................................. . .. II-12

IV-1 Study Area Annual Reprocessing Requirements ............................... ... IV-3

V-1 Fuel Cycle Characteristics for Plant Types Considered ........................... ... V-1

VIII-1 Exposure Pathways and Foods........................................... VIII-2

VIII-2 Maximum and Average Water Intake of Adult, Teen, and Child ..................... .. VIII-3

VIII-3 Values of Parameters Required for Calculation of Radiation

Dose to the Thyroid from Radioiodine......................................... VIII-6

VIII-4 Organ Mass and Effective Radius for Child and Teen ............................ ... VIII-6

VIII-5 Solubilities Selected for Inhaled Radionuclides.................................. VIII-6

IX-1 Average Annual Doses and 50-Year Dose Commitments Received

from Radionuclide Releases by Population of Study Area........................... IX-4

IX-2 Summary of Radionuclides Found to Contribute to Dose........................... IX-9

IX-3 Pathways Found to Contribute to Dose and Dose Commitment.. ..................... IX-12

X-1 Contributions of Air Deposition to River Water Activity ......................... ... X-4

X-2 LWR Fuel Reprocessing Plant Effluent Variation with Preprocessing Cooling Time..........X-4

X-3 LMFBR Fuel Reprocessing Plant Effluent Variation with Preprocessing Cooling Time...... .. X-5

X-4 Ratio of LMFBR Reprocessing Plant Effluent Activities with 30 Days Versus

150 Days Preprocessing Cooling......................................... . X-5

X-5 Changes in Radionuclide Contribution to Dose Caused by Delay in

Installation of Improved Treatment Systems..................................... X-6

X-6 Effect of Delay Time on BWR Noble Gas Releases ................................ X-7

X-7 Typical Release Rates of 8 5 Kr from Reactors.. ................................. X-7

X-8 PWR Secondary Waste Recycle Effects on Releases ............................. ... X-8

X-9 Effect of Reprocessing Plant Size on Contributions to Regional Integrated Dose ......... .. X-13

X-10 Parameter Variations for Vertical Diffusion Sensitivity Studies ............. ........ C-14

X-11 Missouri River Basin Annual Flow and Depletion Rates .......................... ... X-25

X-12 Effects of Radionuclide Removal from Soil on Total-Body Doses ...................... X-25

X-13 Ratio of Radiation Doses to the Average Child Received Under

Two Dairy Management Systems............................................. X-26

X-14 Fraction of Total-Body Dose from Ingestion of Foods Which Results from

Direct Deposition of Radionuclides....................................... . X-26

viii
LIST OF FIGURES

Figure No. Title Page

II-1 Comparison of Radiation Potentially Received From Nuclear Facilities

with that from Other Sources........................................... . 11-2

11-2 Distribution of Total-Body Radiation Received by the Population ...................... 11-3

11-3 Contributions by Pathways (Average Individual)...............................11-5

11-4 Detailed Contribution by Pathways (Average Individual) ......................... ... 11-6

11-5 Contributions by Radionuclides (Average Individual)........................... .... 11-7

11-6 Typical Boiling Water Reactor Radwaste Treatment System .......................... 11-9

11-7 Typical Pressurized Water Reactor Radwaste Treatment System .................... .. II-10

III-1 Study Area Boundaries............................................... . 111-2

111-2 Mean Mixing Depth, Meters, for January and July ............................... 111-4

111-3 Percentage Frequency (Percent of Total Hours) of Occurrence of Low Level

Inversions (Below 500 Feet)............................................ . III-5

111-4 Federal Power Commission Power Supply Areas Included in the Study Area ............ . III.6

111-5 Major Load Centers in Study Area ............................................ 111-8

111-6 Contributions of Plant Type to Monthly Demand.............................. .... 111-9

111-7 Designated Centroids in Study Area ....................................... ... 111-11

IV-1 Capacity Projections for UMBR Extended Study Area ........................... .. IV-2

IV-2 Reprocessing Plant Requirements......................................... IV-4

IV-3 Plant Sites Assumed in Study Area. ...................................... . IV-5

IV-4 Annual Nuclear Energy Generation Pattern for Study Area ....................... ... IV-6

V-1 Nuclear Power Plant Model. ............................................ V-2

V-2 Pressurized Water Reactor Model Schematic..................................V-4

V-3 Boiling Water Reactor Model Schematic .................................... ... V-5

V-4 Liquid Metal Fast Breeder Reactor Model Schematic ............................ V-6

V-5 Midwest Fuel Recovery Plant Flow Chart ................................... ... V-7

V-6 Dual Purpose Reprocessing Plant Flow Chart ................................. ... V-8

V-7 8 5 Kr Release Rates for January 2000 AD ...................................... V-10

V-8 3H Release Rates for January 2000 AD..................................... V-li

V-9 Total Radionuclide Release Rate to Air for January 2000 AD ...................... .. V-12

VI-1 Accumulated Concentration Pattern of 1 37 Cs in the Year 2000

Due to Long-Term Buildup.. ............................................. VI-3

VI-2 3H Air Concentration Pattern Calculated for the Month of July 2000 AD .............. .. VI-4

VI-3 Monthly Ground Deposition Rate of 13 7 Cs Calculated to Occur in July 2000 AD.......... VI-5

VII-1 Upper Mississippi-Lower Missouri Principal Stream Drainages ...................... ... VII-2

VII-2 Upper Mississippi-Lower Missouri Subbasins..................................'VII4

VII-3 Typical Water Concentrations Calculated at Centroid 34 ......................... . VII-5

IX-1 Annual Total-Body Dose Rate Distribution .................................. ... IX-3

IX-2 Regional Pattern of Total-Body Annual Dose Rate (Average Adult) .................. IX-5

IX-3 Relative Radionuclide Source Distribution, Total Airborne Release Rates.............. .. IX-6

IX-4 Regional Pattern of Bone Annual Dose Rate (Average Adult) ...................... IX-7

IX-5 Age Group Dependence of Thyroid Dose Rate for Average Individual in Region.......... .. IX-11

X-1 Effects of In-Reactor Tritium Diffusion on Reprocessing Plant Effluents.............. .. X-2

X-2 Effect on Total-Body Dose Rate Distribution of Various Clustered Plant Configurations .... X-11

X-3 Reprocessing Plant Sites: 5, 10 & 15 MT/Day ................................ . X-12

X-4 Range of Concentrations for Calculated Variations of Vertical Diffusion Parameter, oz ..... X-15

X-5 Concentration-Distance Relationship for Pasquill Stability Categories................ .. X-16

X-6 Effect of Wind Direction on Air Concentration of 85 Kr.......................... .. X-18

X-7 Effect of Wind Speed on Air Concentration of 8 5 Kr ............................ . X-19

X-8 Effect of Stack Height on Air Concentration of 85Kr (Multiple Sources).............. .. X-20

X-9 Effect of Stack Height on Air Concentration of 85Kr (Single Source)................. .. X-21

X-10 Effect of Precipitation Frequency on Air Concentration of 90Sr .................... X-22

X-11 Effect of Distribution Coefficients, Kd, on Radionuclide Concentrations in

Mississippi River Water (Centroid 87) ...................................... ... X-23

X-12 Effect of Distribution Coefficients, Kd, on Radionuclide Concentrations in

Mississippi River Suspended Sediments (Centroid 87)........................... ..X-24

ix

I. INTRODUCTION

This study considered the region comprising the

Upper Mississippi and Lower Missouri River basins. This

region is approximately 300,000 square miles in area,

and extends through eight midwestern states. To assess

the contributions to the radiation potentially received

by the populace within the area from airborne radionuclides originating in adjacent areas, an "air envelope"

zone, some 200 miles wide and surrounding the major

study area, was also considered.

This region is particularly well suited to a study of

this type. The Mississippi-Missouri Valley area includes

several major centers of population and industry, as well

as some of the most important agricultural production

areas of the United States. It also encompasses one of

the major power supply/demand regions of the country,

a region in which nuclear power is expected to play an

increasingly important part in electrical generation.

Furthermore, many of the characteristics of the region,

needed for the study, were already available. These

included demographic data, power and industrial projections, and regional environmental information. Finally,

this region includes within its bounds substantial parts of

two large river basins so that the influence of major river

systems is included.

The study area has a population today of about 18

million, roughly 70% of which may be classified as

urban, 10% as farming, and 20% as rural nonfarm. By

the year 2000 this population is projected to reach 29

million. Today and in the year 2000, the area accounts

for about 10% of the total electric energy generation and

consumption of the United States; with the "air envelope" boundary region added, about 25% of total U.S.

generation is represented. Recent increases in nuclear

energy generation are expected to continue; nuclear

energy will likely be the predominant source for power

generation in the year 2000, when it is estimated that

over 80% of the projected generation of 2,500 billion

kilowatt hours will be provided by nuclear plants.

The study area encompasses much of the eastern

portion of the Great Plains and consists primarily of flat

or undulating prairie land, interrupted by local hilly

areas and incised by the region's waterways. The

hydrology is dominated by the Mississippi-Missouri

drainage systems, and about 80% of the surface water

runoff is supplied by precipitation within the area. As

might be expected from the generally flat terrain, the

climate is rather uniform, meteorological differences

within the area reflecting primarily differences in latitude or movement patterns of continental air masses.

The Chicago metropolitan area was not included in

the base study because of the substantial computational

difficulties involved in adding Lake Michigan to the

region being studied. (Lake Michigan could play a major

role in determining the radiation potentially received by

the populace in Chicago.) Sensitivity studies which

estimated the radiation potentially received by the

populace in the Chicago area indicated that if they were

to be included in the study region, the radiation

potentially received by the average individual would not

increase. However, the impact of Chicago on the power

demand and food transportation patterns of the study

has been included and the results properly account for

its presence and its influence.

The boundaries of the study area were established

along county lines and the majority of the data were

collected at the county level. The 479 counties comprising the study area were combined into 300 environmental and demographic units. Within each such area, all

activities pertaining to the transport of radionuclides and

the exposure of the population to radiation were

assumed to occur at a single point, designated as the

centroid of the area. Calculations were made in accordance with the centroid structure and consisted of the

following major five steps:

1. Using projected power plant and reprocessing

plant siting patterns and rates of radionuclide

releases, based on assumed selection of radiation

waste treatment systems, a "source map" of

release locations was established as a function of

time.

2. The deposition, by centroid, of the released

radionuclides after atmospheric transport was

computed. This was then modified to account for

water transport. The effects of radioactive decay

were included.

3. The uptake of deposited radionuclides and their

subsequent concentration in the various food

types was estimated.

4. Transportation of food to meet subregional demand (by centroid) was simulated.

5. Accounting for work and recreational habits, as

well as living habits, the radiation potentially

received by each population group in each centroid was calculated, based on anticipated direct

exposure to airborne, deposited, and dissolved

radionuclides as well as those ingested or inhaled.

The preparation and accumulation of the requisite

information for the study entailed an effort of major

proportions. The quantity of individual bits of data

handled by the computer program can be gauged by the

approximately three and one-half hours of running time

on a UNIVAC 1108 computer required for a single set of

I-1
calculations. The following are some further details of

the study that will assist the reader in placing in

perspective the results, conclusions, and detailed descriptions of the major steps of the program given in the

succeeding sections.

Using linear programming techniques, a mix of

generating plant types was synthesized in a manner to

provide adequate capacity to meet projected electrical

power demands to the year 2000. The mix included

approximately 440,000 MW of capacity, consisting of

PWR, BWR, and LMFBR nuclear plants and also fossil

fueled plants. A number of reprocessing plants, sufficient to serve only the nuclear plants in the area, were

also included. All the plants, both generating and

reprocessing, were "sited" to supply the various load

centers in the area, with appropriate consideration being

given to particular siting requirements for each type of

plant (proximity to fuel, availability of cooling water,

transmission distance, nuclear plant siting criteria, etc.).

Month-by-month operation of each power plant to

account for seasonal variations in the many subregions

was simulated.

The model simulates generation of fission products in

reactor fuel, as well as activation products formed in the

coolant and reactor structural materials. It then estimates the diffusion of fission products from the fuel

into the coolant, the transfer of radionuclides to the

radiation waste treatment system, and the treatment of

radionuclides as they pass through each component of

the treatment system. Finally, it calculates the rates of

release of radionuclides through gaseous and liquid

discharge streams, and the inventories remaining as

concentrated wastes for subsequent disposal. The reprocessing plants are treated in an analogous manner, with

appropriate consideration of plant feed rate and fuel

type (to distinguish between LMFBR fuel and PWR or

BWR fuel). The releases to the regional environment are

the result of releases associated with the normal operation of nuclear power plants and nuclear fuel reprocessing plants. Other releases were not considered in this

study.

The air transport part of the model, which simulates

the spreading of the released radionuclides into the

environment via atmospheric mechanisms, uses a onedimensional calculation to estimate vertical diffusion of

the radionuclides based on wind speed and atmospheric

stability criteria. Lateral distribution, on a monthlyaverage basis, is assigned individually to each of 16

compass-point sectors. During the transit of the radionuclides, depletion by radioactive decay and the introduction of daughter nuclides are taken into account.

Depletion of airborne concentrations by wet and dry

deposition processes is calculated, as are the ground

concentrations resulting from deposition. Additionally,

calculations are made of ground concentrations in the

year 2000 which result from previous years' operation of

each nuclear facility considered in the study, based on

the age and load history of each plant.

The water transport segment of the model, which

simulates dispersion by hydrologic phenomena,

estimates the transport and resulting concentrations of

radionuclides in both streams and sediments. It also

calculates radionuclide additions to streams from surface

runoff, and concentrations in ground water - and in

standing surface water. For these calculations it uses the

deposition data calculated by the air transport portion.

For both stream transport and surface runoff calculations, the distribution of radionuclides between solution

in water and sorption on sediments or soil is estimated

based on soil types.

Together, the air and water transport models provided comprehensive data on radionuclide concentrations in air and water throughout the study area. In

addition, water concentrations were calculated at locations considered to be important potential sources of

radiation-municipal water intakes, recreation areas,

water used for irrigation water supply, and the like.

Major factors considered in projecting the radiation

potentially received by the populace include eating

habits, food production techniques, recreational habits

and population growth patterns. A massive data bank

containing information pertinent to the population and

living patterns of the study area was established and

used. Included were data on population distribution by

location, age group, and urban-rural classification; dietary habits and work and recreational patterns; drinking

water supply and treatment; and food production,

consumption, transport within the study area, and food

imports from outside the region. These data were

obtained from many involved cognizant organizations

including the U.S. Geological Survey, the Corps of

Engineers, NOAA, the Bureau of Census, the Department of Agriculture, state Fisheries Departments, and

regional planning and conservation committees.

Ten separate pathways leading to possible human

exposure were considered. These included contributions

from direct exposure to radionuclides in air and water

and on the ground, air inhalation, ingestion of water and

the ingestion of some 35 separate categories of foods.

Food-chain relationships of radionuclide concentrations

were calculated for each food type of each of the 45

separate fission and activation products included in the

study. The radiation potentially received was examined

for total body and for each of six vital organs; 50-year

commitments were calculated in addition to the annual

calculations. Definitions of 1. the manner of calculating

the radiation potentially received in the course of a year

and 2. the 50-year commitment are:

1. The rate of potential exposure to radiation (millirem/yr) is summed from 12 monthly values.

External exposure was calculated from concentrations in the environment during each month.

Internal exposure from each month's deposition of

radionuclides was calculated for the succeeding 12

months. A portion of the internal exposure was

thus calculated beyond the year 2000. For nuclides with effective (radiological or biological)

half-lives of less than 1/2 year, this extension of

the calculation very closely approximates the

I-2
exposure in the year 2000 from previously accumulated body burdens.

2. The radiation commitment was the long-term

contribution due to uptake during the year 2000.

This value, termed 50-year dose commitment in

this report, was the radiation an individual would

be committed to during the following 50 years due

to the body burden from radioactive material

ingested during the year 2000.

This procedure actually overstated the radiation

received in the year 2000 from materials ingested in that

year; however, for nuclides with effective half-lives less

than 1/2 year, it compensates for the body burden

accumulated prior to the year 2000.

Individual values of radiation potentially received

were calculated for average individuals and for individuals representing maximum-exposure population

groups in each of three age groups (child, teenager, and

adult). Also, average values were calculated for the study

area total population and for the population in each

centroid. Finally, contributions by isotope and by

pathway were calculated for each centroid.

The calculations were repeated for different values of

major parameters. This allowed identification of critical

pathways and important radionuclides as well as an

assessment of the effect of uncertainties in input data.

And this, in turn, permitted conclusions to be made

relative to the potential effects of technological advancements in disposal and treatment of radioactive waste

materials, and to the sensitivity of many factors that

might be large contributors to the population exposure.

Though this study does not involve the safety aspects

of nuclear power plants, it is felt that the following short

discussion on the safety principles for nuclear power

plants is appropriate.

The nuclear industry has been and is one of the safest

industries in the Nation. This has been, in large part, due

to the AEC emphasizing safety from the beginning.

Unlike most other industries, the nuclear industry has

not waited for serious accidents before adopting rigid

safety controls.

To assure that nuclear power plants are built and

operated safely and that reactor accidents are prevented,

a "defense-in-depth" technical concept is used. It consists of three basic levels of safety as follows:

The first level concerns the intrinsic features of the

design of the nuclear plant and the quality,

redundance, testability, inspectabiity, and fail safe

features of the components of the reactor and

plant. The design must be such that the plant is

unquestionably safe in normal operation and has a

maximum tolerance for errors, abnormal operation

and component malfunction. Analyses are made

and test programs conducted to find those types

of malfunctions or faults that could affect safety

so that they can be guarded against by design,

quality assurance, or fail-safe features as appropriate. A reactor plant built in this way and routinely

tested and monitored not only provides protection

for the heavy investment made by the utility

owner which would be seriously jeopardized by a

severe accident, but also provides a maximum of

protection for the operating staff and the public.

The second level concerns such incidents as partial

loss of flow, reactivity insertions, failure of parts

of the safety system, or fuel handling errors, which

are assumed to occur in spite of the care taken in

design, construction, and operation. This second

level provides fault detection equipment and design features which enable such occurrences to be

arrested or accommodated safely. Conservative

design practices, adequate safety margins, and

parallel, independent, redundant arrangements of

detecting and actuating equipment (so that if one

fails, others will be available to provide protective

action) are used in the design and operation of

such reactor protection systems. In addition, these

systems are designed to be readily inspected and

tested so that there is a high degree of assurance

that they will operate reliably in the infrequent

event they are required.

The third level concerns the postulated failure of

protective safety systems simultaneously with the

accident they are intended to control. The consequences of such hypothetical accidents must be

evaluated and understood. Furthermore, if a practical means can be found to provide an additional

measure of safety to mitigate the accident or

accommodate the consequences, that also may be

considered.

The AEC also has a vigorous licensing process to

review and assure the safety of nuclear facilities. The

regulatory organization includes three main independent

units; the regulatory staff, the independent Advisory

Committee on Reactor Safeguards, and the independent

Atomic Safety and Licensing Boards.

The expertise of the Commission's regulatory organization is first brought to bear as part of the process for

the review of applications for construction permits for

nuclear power reactors.

These applications undergo four separate reviews: (1)

by the AEC's regulatory staff, (2) by the independent

Advisory Committee on Reactor Safeguards, (3) by an

Atomic Safety and Licensing Board after a formal "trial

type" public hearing, and (4) by the Atomic Safety and

Licensing Appeal Board or the Commission itself, based

on the hearing record. A similar procedure is followed

for applications for operating licenses for nuclear power

reactors, except that there is no mandatory public

hearing in the absence of a request for one or a decision

by the Commission, on its own motion, to order such a

hearing.

The AEC's safety program and that of the commercial nuclear power industry has been frequently cited as

an example of the ability to provide an excellent level of

safety. The industry's safety record is due in large part

to the rigid standards applied to the design, construction, testing, operation and maintenance of nuclear

power plants. These involve high quality materials and

workmanship and safety systems designed both to

I-3
prevent serious accidents and to limit the consequences

in the remote event that they occur. With respect to

safety, President Nixon said in his Energy Message to

Congress on June 4, 1971:

".. . the safety record of civilian power reactors in

this country is extraordinary in the history of

technological advances. For more than a quarter of

a century-since the first nuclear chain reaction

took place-no member of the public has been injured by the failure of a reactor or by an accidental

release of radioactivity. I am confident that this

record can be maintained...."

I-4
II. SUMMARY

Regional Estimates

This study considers the radiation potentially received by the population of the UMRB study area which

might result from the operation of nuclear power and

fuel reprocessing facilities in the year 2000. Those

facilities which were assumed to be operating in that

year, in the UMRB study area and the surrounding "air

envelope" region, included an aggregate nuclear generating capacity of 356,000 MWe (assumed to consist of

46,000 MWe of BWRs, 138,000 MWe of PWRs, and

172,000 MWe of LMFBRs), plus nine nuclear fuel

reprocessing facilities (8 plants of 5 tonne per day

capacity plus the existing 1 tonne/day Morris facility).

Radionuclide releases from these facilities were estimated, transport through the region was modeled, and

the resulting concentrations were used to calculate

radiation exposure to the 29.1 million inhabitants of the

UMRB study area.

The results of the study of the UMRB based on the

selected radiation waste treatment systems showed that,

on the average throughout the region, the potential

radiation an average individual could receive in the year

2000 would be increased by roughly 0.2 millirem/year

because of the presence of nuclear facilities. This may be

brought into perspective by realizing that it is only

slightly more than one-tenth of one percent of the

roughly 140 millirems/year received from natural sources,

about two-tenths of one percent of that received from

medical and dental sources such as diagnostic X-rays,

and only one percent of that received from other

manmade sources such as television set components and

luminous watch dials. This is shown graphically in Figure

II-1. In a 1972 report published by the United Nations

entitled "Ionizing Radiation: Levels and Effects," it is

estimated that by the year 2000 the annual average

radiation potentially received by the world population

from nuclear facilities might be as high as two-tenths of

one percent of that received annually from natural

sources.

As might be expected over a region as large and

diverse as the study area, the spread in estimated

exposures is substantial. Nevertheless, as indicated in

Figure 11-2, 99 percent of the population was estimated

to receive a potential total body radiation of less than

0.5 milirem/year. A very small fraction of the population could receive up to 1.2 millirems/year-a value which

is still only one percent of the natural background.

Patterns for the three age groups considered (adult,

teenager, child) varied only slightly from that for the

entire population.

A few individuals, living near the boundaries of

nuclear sites, might potentially receive somewhat more

radiation than those indicated. Proposed AEC regulations are intended to limit the dose to any organ of

individuals offsite to 5 millirems per year from radioactive material in liquid effluents and 5 millirems per

year to any organ of individuals offsite from radioactive

material in gaseous effluents. For greater distances, it

was estimated in the study that a person living one mile

from a nuclear power plant might receive radiation to

the total body of 0.02 to 0.15 millirem/year from that

one installation, depending on the type of plant installed. A person living one mile from a nuclear fuel

reprocessing plant in the UMRB could potentially

receive a total body radiation of approximately 7

millirems/year (present regulations permit 170 millirems

per year); even this value is only about 5 percent of

natural background radiation. It is expected that only a

very few individuals would live as close as one mile to

these plants. Thus, when calculating the radiation

potentially received by the average individual in the

region, little error is introduced by omitting the dose

calculations for these few individuals. The calculations

using the centroid concept are felt to be representative

of radiation potentially received by typical persons

throughout the region.

The average radiation potentially received by the

total body of an individual in the study area in the year

2000, resulting from operation of the assumed facilities,

was calculated to be 0.17 millirem. Of this, roughly 82

percent was contributed by nuclear facilities within the

study area while the remaining 18 percent was from

airborne releases in the peripheral "air envelope" region.

The calculations also indicated the amount of radiation potentially received by various organs. Because of

the different dietary habits and also because of differences in organ size, the radiation potentially received

by the organs for the three population groups varies

somewhat. Table II-1 indicates the relationship between

radiation potentially received by individual organs and

the radiation potentially received by the total body for

an average individual in each population group within

the UMRB.

The natural background figures include roughly 25

millirems/year due to natural radiation sources within the

body (Carbon-14, Potassium-40, etc.). The incremental

body burden of radionuclides resulting from operation

of nuclear facilities is only a very small fraction of this.

For a person exposed for one year to the average

radionuclide concentrations estimated in this study for

the year 2000, the total body commitment resulting

II-1
RANGE

IN UMRB

REGION

NATURALLY OCCURRING

BACKGROUND RADIATION

POTENTIAL RADIATION FROM

NUCLEAR FACILITY EFFLUENTS

(NORMAL OPERATION)

_- . 1

MEDICAL-DENTAL SOURCES

OF RADIATION (AVERAGE)

MISCELLANEOUS MAN-MADE

SOURCES OF RADIATION

(TV, WATCH DIALS, ETC.)

Figure 11-1.-Comparison of Radiation Potentially Received from Nuclear Facilities with that from Other Sources.

1 50

LUJ 100 

LL

W/ V0-4 

I0

F- 50 r'

0
1.5 F-

- 1.2 - MAXIMUM* (AVERAGE ADULT)

1.0 - 1 .035 - - MAXIMUM* (AVERAGE PERSON) - -

*MAxIMUM: HIGHEST AVERAGE DOSE CALCULATED FOR

ANY OF THE 300 CENTROIDS CONSIDERED

I

Ow

J >-

I-LW

CC1- -Z

W -4

0

C7

- 0.17

- 0.13 ---

1

99 PERCENTILE - - --- - -

AVERAGE I-- MEDIAN -- 

1 1 1 1 1 l 

40 50 60 70 80 90

PERCENT OF REGIONAL POPULATION (CUMULATIVE)

Figure 11-2.-Distribution of Total-Body Radiation Received by the Population.

0.5 F 0.50 --

0

10 20 30 100

1m
TABLE Il-1

Ratio of

(Radiation Potentially Received by Organ)/

(Radiation Potentially Received by Total Body)

Organ

GI tract

Thyroid

Bone

Lungs

Skin

Liver

Child

.71

8.41

.36

.97

1.09

1.20

Teen

.87

4.03

.25

.97

1.12

1.11

from the decay of the added radionuclide burden over a

50-year period, would be less than 0.2 millirem-less

than one percent of the annual radiation received from

the natural body burden.

Relative Importance of Pathways

Not all pathways contribute equally to the radiation

potentially received by the population. One finding of

the study was that the contribution is governed primarily by air transport rather than by water transport or

shipment of foodstuffs. The study disclosed that four of

the pathways considered would provide the bulk of the

radiation potentially received: submersion in air, inhalation of air, transpiration of tritium and exposure to soil.

All of these pathways are principally related to air rather

than water transport.

The relative contributions of the various pathways are

shown in Figure 11-3 for the organs considered in the

study. Note that air inhalation, which involves the intake

of air and its associated radionuclide burden, is the

major contributor for most organs, especially the lungs.

Air submersion involves external radiation to the body

from these same airborne radionuclides. Not too surprisingly, it is a factor in the radiation potentially

received by the skin. Soil exposure is also external

radiation, but from radionuclides deposited on the

ground. As expected, it is also a contributor to the

radiation potentially received by the skin. The perhaps

unexpected relative magnitude of this contribution to

the radiation potentially received by the bone results

from the low radiation absorbed by the bone from other

sources (total radiation potentially received by the bone

is only about 20% of the total radiation potentially

received by the body). Tritium transpiration is a

relatively large factor for all organs except bone.

Generally, only about 10% of the potentially received

radiation is contributed by pathways other than the four

already mentioned, although for some organs it is larger,

reaching 25% for bone and 67% for the child's thyroid.

Of these other pathways, ingestion of drinking water,

milk, and beef are the most important. A more detailed

breakdown of pathways for bone and thyroid in the

average adult is shown in Figure 11-4.

Radionuclide Contributions

Of the 45 radionuclides considered in the study,

isotopes of only three elements contributed to the bulk

of the potentially received radiation. These three elements, tritium, iodine, and cesium, contribute about

95% of the radiation potentially received by the total

body and 80% or more of the radiation potentially

received by the organs considered. The breakdown of

radionuclide for the various organs considered is given in

Figure II-5.

Examination of the figure reveals tritium to be a

major contributor for many of the organs and for total

body. This is due partly to the fact that tritium is the

only gaseous radionuclide for which extensive gaseous

waste stream removal was not assumed. Efforts have

recently been initiated to study the development of

gaseous tritium removal systems. When the present study

was begun, however, such systems could not be considered "present technology," and thus the base study

does not assume the use of gaseous tritium removal

systems. The potential impact of tritium removal was

examined in sensitivity studies. There are some indications based upon examination of some test fast reactor

mixed oxide stainless steel clad fuels, that most of the

tritium generated in the fuel diffuses into the primary

sodium coolant system. As a result, tritium removal may

be an important consideration for LMFBRs as well as

reprocessing plants.

Iodine is of principal interest in relation to the

thyroid. The contribution is relatively large despite

radwaste treatment because of the affinity of the

thyroid for iodine. It also contributes to other organs

because it is a major fission product-that is, a relatively

large fraction of fissions yield iodine as a fission product,

and in spite of effective removal systems, some iodine

releases still occur. The major pathways for iodine were

determined to be air inhalation, ingestion of drinking

water, fresh milk, beef and lamb.

Cesium isotopes are also major fission products and

contribute to the dose rate in bone, liver, total body,

and GI tract. The principal pathways involved were

drinking water and exposure to soil.

Krypton-85, considered by many to be a "problem"

radionuclide, was found to be an important factor only

in the radiation potentially received by the skin. This

could be expected partly because of the short-range beta

emission which contributed only to the skin and also

because of the assumed long-term retention of noble

gases in the radiation waste treatment processes in

operation in the year 2000. Similar comments may be

made about xenon isotopes of which Xenon-133 is the

major contributor.

Strontium-90, also generally considered to be a

radionuclide of concern, was not found to be an

important factor. With the nuclear plant emissions

assumed in this study, its contribution was only about 2

percent of the radiation potentially received by the

bone. It was a contributor in the GI tract because of its 9 0 Y daughter.

11-4
0.169

OTHER

DRINKING

WATER

MILK

FOOD

INGESTION

EXPOSURE

TO SOIL

AIR

INHALATION

TRITIUM

TRANSPIRATION

AIR

SUBMERSION

TOTAL BODY

0.181

OTHER

DRINKING

WATER

MILK

FOOD

INGESTION

EXPOSURE

TO SOIL

AIR

INHALATION

TRITIUM

TRANSPIRATION

AIR

SUBMERSION

LIVER

0.163

OTHER

EXPOSURE

TO SOIL

FOOD

INGESTION

DRINKING

WATER

AIR

INHALATION

TRITIUM

TRANSPIRATION

AIR

SUBMERSION

LUNGS

0.133

OTHER

EXPOSURE

TO SOIL

DRINKING

WATER

FOOD

INGESTION

AIR

INHALATION

TRITIUM

TRANSPIRATION

AIR

SUBMERSION

G.I. TRACT

0.753

OTHER

DRINKING

WATER

FOOD

INGESTION

MILK

EXPOSURE

TO SOIL

AIR

INHALATION

TRITIUM

TRANSPIRATION

AIR

SUBMERSION

THYROID

0.034

OTHER

DRINKING

WATER

FOOD

INGESTION

MILK

EXPOSURE

TO SOIL

AIR

INHALATION

AIR

SUBMERSION

BONE

0.179

OTHER

EXPOSURE

TO SOIL

DRINKING

WATER

FOOD

INGESTION

AIR

INHALATION

TRITIUM

TRANSPIRATION

AIR

SUBMERSION

SKIN

Figure 11-3.-Contributions by Pathways (Average Individual).

DOSE RATE

MREM/YR

100r

801-

60F-

- FUC/

CA

hW~ 40

20

01
0.753

100 r_ OTHE

DRINKING

WATER

PORK

BEEF

EGGS

MILK

WHEAT

FRESH VEG.

POTATOES

FRUIT

*7.- SOIL

EXPOSURE

AIR

INHALATION

TRITIUM

TRANSPIRATION

AIR

SUBMERSION

THYROID

Figure 11-4.-Detailed Contribution by Pathways (Average Individual).

11-6

DOSE RATE

MREM/YR 0.034

80 1-

60

U

u--

o- 40

20 k

01-..

BONE

- - - - -

1
DOSE RATE

MREM/YR

1 00

0.169

OTHER

CESIUM

XENON

IODINE

ZIRCONIUMNIOBIUM

TRITIUM

TOTAL BODY

0.181

OTHER

CESIUM

XENON

IODINE

ZIRCONIUMNIOBIUM

TRITIUM

LIVER

Figure I1-5.-Contributions by Radionuclides (Average Individual).

0.163

OTHER

CESIUM

XENON

IODINE

ZIRCONIUMNIOBIUM

TRITIUM

LUNGS

0.133

OTHER

CERIUM

CESIUM

XENON

IODINE

STRONTIUM

ZIRCONIUMNIOBIUM

TRITIUM

G.I. TRACT

80 

60-

wi

ILw

L/)

O

IU

0.

40

20 h

0 -L

w 0.753

OTHER

IODINE

0.034

OTHER

CESIUM

XENON

IODINE

IODINE

RUTHENIUM

STRONTIUM

ZIRCONIUMNIOBIUM

BONE

0.179

OTHER

CESIUM

XENON

IODINE

KRYPTON

ZIRCONIUMNIOBIUM

TRITIUM

SKIN

TRITIUM

THYROID
Cerium-144 was a factor to be considered only in the

GI tract. Its half-life is sufficiently long to survive the

transit times involved, and its short-lived daughter,

144Pr, also contributes to the radiation potentially

received by the total body.

Transuranium elements were not considered in this

study. A subsequent study has been undertaken to

evaluate the effects of these elements; preliminary

results indicate that these effects will be minimal except

for a small contribution to the dose rate to the lungs.

Power Projections

For the purpose of the study, approximately 360,000

MW of the 440,000 MW of electric generating capacity

required for the region was assumed to be nuclear.

Reprocessing plants in the area were assumed to service

only these nuclear power plants. It was determined that

for the radiation waste treatment systems assumed for

the nuclear facilities, the primary source of the population exposure was the reprocessing facilities and not the

power plants.

Additional studies showed that the ratio between the

PWR and the BWR was relatively unimportant in dose

rate determination. Studies were made of the effects of

clustering nuclear sites into small "nuclear parks" with

sufficient reactors to support various sizes of reprocessing plants. Although indications were obtained that

some slight decrease in dose may be achievable, the

sensitivity of results to the specific sites studied made

generalization difficult.

System Design and Operation Variants

Radioisotopes present in the gaseous and liquid

process streams of a nuclear reactor are removed to a

very high degree prior to the discharge of any portion of

these streams from the plant. This removal is accomplished in radiation waste (radwaste) treatment systems

designed to recycle liquid or gaseous streams when

possible, and to highly purify any portions of those

streams which are discharged. Flow charts for typical

BWR and PWR radwaste systems are shown in Figures

11-6 and 11-7.

Several types of radwaste treatment equipment have

been developed, and improved types are expected to be

developed in the future. In the base calculations of this

study, it was possible to include only a limited number

of the radwaste treatment options available or potentially available in the year 2000. Performance of several

of these systems, in the absence of actual operating data,

was calculated based on experimental results to date as

reflected by estimated radionuclide releases as published

in SARs.

The use of different radwaste treatment systems,

differences in the performance characteristics of individual radwaste components, and differences in design

flow rates or in rates of leakage, could change the

estimates of radiation received by the populace.

Because of the high degree of recycling assumed for

reclaimed water in LWR radwaste systems, liquid releases were not found to be sensitive to variations in

flow rates of individual streams or to differences in

leakage rates (as taken for PWR steam generators).

However, if floor drain streams (which include steam

generator blowdown in this analysis) are not recycled,

the liquid releases of radionuclides other than tritium

may increase by about a factor of 8 or 10 in PWRs. BWR

floor drain flows normally are not recycled, but lower

concentrations in the treatment system due to higher

rate of primary coolant cleanup and the absence of

blowdown contributions reduce releases from this

source.

Increasing the delay time of BWR charcoal-bed gas

retention systems (e.g., by low-temperature operation)

can significantly decrease noble gas activity. A ten-fold

increase in delay would lower xenon activity by a factor

of 1000, and could effectively remove all krypton

isotopes except 85 Kr.

Effects of "Advanced" Treatment Systems

The effects on results of the study which might be

expected if the use of alternate types of waste treatment

systems were assumed, are indicated by the following

results:

" All nuclear plants in the study except LWRs were

assumed to be equipped with systems for bottling

noble gases, with an assumed decontamination

factor of 300. If the LWRs were similarly

equipped, small reductions in radiation potentially

received by the skin would occur; negligible

changes would be seen in the radiation potentially

received by other organs or to the total body.

" If no reprocessing plants or power plants in the

study had bottling systems for noble gases, the

average total body radiation potentially received

by the regional population could increase by

approximately 30%-primarily from the additional

8 5 Kr released from reprocessing plants.

" If the introduction of radwaste treatment systems

meeting the guidelines given in 10 CFR 50

Appendix I for radionuclide releases from nuclear

reactors and similar guidelines for reprocessing

plants were to be delayed until 1985, the average

total body radiation potentially received by the

regional population would be increased by about

35%. Again, a substantial fraction of the increase

would be due to additional 85Kr releases from

reprocessing plants without krypton bottling facilities. Specific radionuclide releases-especially

those in the releases to water from reprocessing

plants (permitted in this test) would increase by

factors of 100 to 3000, but even at the increased

release rates they would play a minor role in dose

contributions as compared to tritium and krypton.

" There is some economic incentive for shortening

the "cooling" time of fuel elements between

11-8
REACTOR NOTE: CIRCLED NUMBERS REFER TO COLUMN I NUMBERS IN TABLES C-2 AND C-3

FUEL

SOLIDLIQUID

WASTES

COOLANT PHALECTO ITRDMNRLZRTN

I TU I EAAA

SOLIDRELEASES

SOLID 'SOLID

WASTES WASTES

COAPAFiu re

SOLIDLI

L__ _ ___WATTE_. .

TURHINEDUPINCHLTR TERAOU

O

SOLID 'OI

L _WASTES WSE

Figue WterReator I-6.Typial adwsteTreamen oilng Sytem
NOTE: CIRCLED NUMBERS REFER TO COLUMN

NUMBERS IN TABLES C-4 AND C-5

jGAS HOLDUP LVT L GASEOUS

_ - -KrRELEASES

REACTOR 0

SOLID WASTES

FUELCCOLLECTION DEGASIFIER HOLUP FILTER

1COOLANTTAKAN

|| o o |SOLID

DEMINERALIZER H EVAPORATOR TANK WSE

COOLANT

CLEANUP

SYSTEM

SOLID SOLID

SLDWASTES 

WASTES

WASTES

I _

I TURBINE

FLOOR

DRAIN FILTER TANK

TANK*

SOLID J WASTES

*INCLUDES STEAM

GENERATOR BLOWDOWN

Figure 11-7.-Typical Pressurized Water Reactor Radwaste Treatment System.

0-

0

LIQUID

RELEASES
irradiation and reprocessing. The shorter cooling

time results in sharp increases in concentration of

the fission products with shorter half-lives that

must be handled by the reprocessing plants. For

example, reduction of the cooling time of LWR

and LMFBR fuels from 150 days to 30 days

increases 1311 inventories by about 10,000, and

those of 1 3 3Xe by about 10 million. Increases in

reprocessing plant releases from LMFBR fuels,

alone, by these factors would increase the average

regional total body radiation potentially received

by about 40% (tritium would still be the major

contributor), unless compensating improvements

in treatment systems were included in facility

design.

0 Addition of a volatile oxidizer (voloxidizer) to the

initial processing stages of a reprocessing plant

would permit collection and treatment of gaseous

radionuclides in relatively concentrated states,

before they contact the process fluids. This could

result in reduced releases. A reduction in release of

about 100, for example, may result for tritium,

krypton and xenon, if this system is used. The

radiation potentially received due to tritium would

then be reduced to the extent that it would

account for about 20% of the total and the total

itself would be reduced by about a factor of five.

Tritium Diffusion Effects

In the base study, most of the tritium produced in

reactor fuel was assumed to remain in the fuel until

released during reprocessing. Thus, the release of tritium

at reprocessing facilities became the dominant factor in

contributing to radiation potentially received by the

population in the study area since no tritium removal

systems were assumed for these plants.

At the time when a large fraction of the nuclear

capacity is provided by the LMFBR (about 40% in year

2000) much of the gaseous tritium that is assumed to be

released from reprocessing plants in the base case may,

as noted in the section entitled "Radionuclide Contributions to Dose Rate," diffuse into the LMFBR sodium

coolant systems. Tritium-removal systems for the

LMFBR would then be a major factor in reducing

tritium releases.

The tendency of tritium to hydride sodium leads to

its possible removal in cold traps (as assumed in the

study). Other potential recovery means-sorption in

zirconium or other metals, entrapment in molecular

sieves, or oxidation followed by recovery and disposal of

the tritiated water-may apply equally to LMFBRs or to

reprocessing plants.

Sensitivity studies, in which 90% of the tritium was

assumed to have diffused from LMFBR fuels, indicated

that gaseous releases from reprocessing plants (handling

combinations of LWR and LMFBR fuels) decreased to

15%-25% of the activity that would result if tritium were

largely retained in the fuel (over 95% retention was

assumed in the reference case). If the diffused tritium

were assumed to escape from the LMFBRs, no large

change would occur in radiation potentially received

throughout the region, but the distribution would be

expected to change because of the more diffuse,

widespread releases. Tritium treatment at the LMFBRs

(for example by the cold traps assumed in the study)

could in fact substantially reduce regional distribution of

the potential radiation.

Radionuclide Transport Considerations

Because of the large contributions from airborne

radionuclides, uncertainties in meteorological parameters

can have relatively large effects on radiation potentially

received by the population. In the study, average

meteorological conditions for each month were derived,

for each location involved, by evaluation and interpretation of long-term recorded weather data. Although

large deviations from these mean conditions may occur,

they would not be expected to be long-lasting nor to

affect large portions of the study area. Thus, these

deviations would not be expected to affect the regional

radiation distribution calculations. For that reason, more

complicated models which allow consideration of these

deviations were not considered to be justified for this

study.

A series of calculations was made to estimate the

effects of variations in meteorological conditions on

airborne radionuclide transport. Changes were assumed

in mixing layer depth, atmospheric stability, wind speed

and direction, and frequency of precipitation. Extreme

variations of any one parameter could affect concentrations by a factor of ten or greater. However, variations

of this type would be expected to be short-lived and

would not affect the season-average values used in the

study.

Calculations of releases from stacks of varying height

indicate that the concentration difference between

releases at ground level and at 100 meters are relatively

small (using season-average meteorology) at distances

beyond about 5000 meters.

In the waterborne transport of radionuclides, few

data are available on the sorptivity of radionuclides on

various types of soils or waterborne sediments, or on

conditions (temperature, pH, etc.) which might affect

sorption. Distribution coefficients defining the relative

concentrations of radionuclides adsorbed on sediments

and those in solution were chosen based on best

available data. Because of the large uncertainties in

radionuclide behavior, however, the coefficients are

subject to considerable uncertainty.

To test the effects of these uncertainties, radionuclide

concentration calculations were run in which the distribution coefficients of radionuclides subject to sorption

were varied over a large range. Changes in dissolved

radionuclide concentrations were relatively smallusually, by factors of 10 or less. However, the calculations

indicated that ground water concentrations and conII-i1
centrations retained in soils and sediments could potentially vary over several orders of magnitude. In this

respect, the assumptions on ground water concentrations

as used in the base study appear to be conservative.

Tests of the contribution of air-deposited radionuclides to river water activity showed wide variations,

regionally and within the region. The greatest part of the

activity in rivers from strontium, zirconium-niobium,

and 129I (typically released mainly to air from reprocessing plants) resulted from air deposition. For other

nuclides, air deposition was less important than direct

release into streams.

Surprisingly, there appeared to be very little coupling

between airborne tritium and its activity in river waters.

Apparently, this is due to the dilution of tritium oxide

by natural water in the atmosphere (only a small

fraction of atmospheric water is precipitated).

Conclusions

The Year 2000 Study represents a detailed and

thorough analysis of radiation potentially received by

the public due to nuclear facility effluents, projected to

a time when the generation of electricity in nuclear

power plants will exceed the total amount of electricity

generated by all types of plants at the present time.

Present practices in the treatment of effluents from

nuclear facilities are such that discharges of radionuclides are held to very low levels. Contributions to the

radiation potentially received by the population from

these discharges represent only a small fraction of the

radiation received from natural background radiation.

The results of the study indicate that with relatively

modest changes in equipment and practices for effluent

control, radiation potentially received from emissions

from the nuclear industry can be maintained at an

extremely small fraction of the unavoidable radiation

received from the natural background. Moreover, the

study shows that with more "advanced" effluent control

technology-which is now being developed-radiation

from the nuclear industry could be held to even lower

levels.

While differing regional conditions may cause some

difference in the calculated radiation potentially received by the population from sector to sector, it does

not appear that large differences are likely to occur.

Thus, it appears that radiation potentially received by

the population from the nuclear industry will remain

low throughout the country, as the industry grows. The

significance of the projected average potential radiation

potentially received by the population of 0.2 millirem/

year in the year 2000 from nuclear facilities should be

assessed with the realization that this is added to a base

of about 200 millirems/year from all other sources.

Therefore, it can be concluded that the potential

radiation received by the population from the operation

of potential nuclear facilities in the year 2000 would

present no hazard to their health and safety.

II-12
III. DESCRIPTION OF STUDY AREA

PHYSICAL DESCRIPTION

Area Boundaries

The area selected for investigation, in this pilot study

of radiation dose in the year 2000, comprises the

watersheds of the Upper Mississippi River (above its

confluence with the Ohio River) and of the lower

portion of the Missouri River (below Pierre, South

Dakota). The study area encompasses 303,230 square

miles and includes the entire state of Iowa and portions

of Minnesota, Wisconsin, Illinois, Missouri, Kansas,

Nebraska, and South Dakota (Figure III-1). Headwaters

of some western tributary rivers, not important to the

study, were excluded from the area.

Boundaries of the study area were defined along

county lines, approximating closely the actual watershed

boundaries. Organization of the area in this manner

facilitated the use of large amounts of data available

primarily on a county basis.

It was recognized that an evaluation would be needed

of the effects on potential radiation dose to the

population resulting from the airborne contribution of

radionuclides from adjacent areas. To account for this

contribution, a peripheral zone some 200 miles wide and

surrounding the basic study area was defined and

designated as the "air envelope." This is also shown in

Figure III-1. The generation of airborne radionuclides in

this area, and their influx into the central study area,

were included in the study. Waterborne radionuclides

released in the "air envelope" area were not considered;

because of regional stream flow patterns, these radionuclides would bypass the central area-except in the

extreme northwest sector of the envelope area, where no

significant nuclear buildup was projected.

The "air envelope" region, like the central study area,

was defined along county lines-in this case to assure

that load centers and generation centers which might be

important to the study were included for consideration.

In western portions of the area, it was convenient to

extend the air envelope boundaries to match FPC Power

Supply Area (PSA) boundaries rather than attempt to

subdivide available FPC power forecast data.

Topography

The study area encompasses much of the eastern

portion of the Great Plains, and much of the area

consists of flat or undulating prairie land, interrupted by

local hilly areas and incised by the waterways of the

region. The area extends on the south to the northern

reaches of the Ozark Highlands of Missouri, and on the

north to the lake country of Minnesota and the low hills

of the Minnesota-Wisconsin area.

The lack of rugged terrain features in this area and

the presence of well-defined waterways were influential

in the early establishment of the area as one of the great

agricultural and industrial areas of the United Stateshence, one which is of interest in this study. Likewise,

the generally uniform topography simplifies the movements of weather fronts, storm systems, and the like,

thus avoiding undue complications in modeling the

meteorology.

Hydrology

The hydrology of the study area-including both surface and ground waters-is dominated by the MississippiMissouri drainage systems. Nearly all surface water

runoff in the Upper Mississippi (93,000 cfs) is supplied

by precipitation within the area. In the Missouri River

drainage, however, approximately 40% of the present

annual average of 75,000 cfs flows into the study area

from upstream sources. Runoff varies from one inch per

year in the northwest part of the study area in South

Dakota to 15 inches in the northeast and southeast. Less

than 5% of the precipitation leaves as runoff in the

northwest in contrast to 40% in northern Wisconsin.

Average runoff for the area is approximately seven

inches or ~23% of annual precipitation. The distribution

of runoff within the region is shown in Appendix B.

Sediment load varies considerably over the basin,

ranging from concentrations of 10 ppm in Mississippi

headwater streams of Minnesota to over 20,000 ppm in

the Nishnabotna River of Iowa. Appendix B includes a

map of average annual sediment concentration in the

study area. Average concentrations in excess of 2000

ppm are found in an east-west belt extending from

central Illinois across southern Iowa and northern

Missouri into Nebraska and Kansas.

Ground water pumped from wells in the region is

generally developed from shallow, unconsolidated sand

and gravel aquifers or more deeply seated rock aquifers,

which frequently are under artesian pressures. Unconsolidated aquifers generally capable of yielding 50

gallons per minute (gpm) have their broadest distribution in Nebraska, where they are utilized as major

sources of irrigation water, primarily along the Platte

River. Kansas City and St. Louis withdraw significant

quantities of water from shallow alluvial aquifers along

the Missouri and Mississippi Rivers.

Rock aquifers are major sources of supply in the

Minneapolis-St. Paul, De Moines, Peoria, Rockford and

III-1
. AIR ENVELOPE

REGION ~MINNEAPOLIS 4 E A

CENTRAL

STUDY KADISON DETROIT

SIUXCIY REGION CICGO

T CIAGCLEVELAND DEUS 

MOIES

NOT LTEOMAHA MI SAI ENVELOPE .,COLUMBUSREGION

ST.LOUIS.LOUISVILLE

DODGE CITY W CHITA

NASHVILLE

KLAHOMA CITY

LITTLE ROC K

Figure I11-1.-Study Area Boundaries.
Chicago surburan areas. The northeastern group of

aquifers consists of interstratified sandstones, shales and

dolomites which dip away from a major structural

feature, the Wisconsin Arch. This feature forms a

highland and relatively impervious pre-Cambrian basement rock in north-central Wisconsin. Two highly

productive artesian systems, the Ironton-Galesville formation and the St. Peter sandstone, are among this

group. The principal artesian aquifer in South Dakota is

the Dakota sandstone. Cavernous limestones and sandstones in the Ozark Highland are the major rock aquifers

in Missouri. Rock aquifers have been highly developed as

a source of community water supply in the study area.

The distribution of consolidated and unconsolidated

aquifers can be found in Appendix B.

Meteorology

As might be expected from the generally flat terrain

of the study area, there are no dramatic changes in

climate in the area. The meteorological differences

which are present occur as gradual changes, reflecting

differences in latitude or patterns of movement of

continental air masses.

The mean surface temperatures(2) vary from winter

lows of 0-35F to 70-85F in the summer. Mean

minimum winter temperatures of -5F occur in January

in the northern portions of North Dakota and Wisconsin.

Mean maximum summer temperatures of 90F occur

during July in the southern part of the study area.

Throughout the study area precipitation occurs in the

form of both rain and snow. Normal amounts increase

from approximately 20 inches in the northwest to 45

inches at the southern tip of Illinois with an average of

about 30 inches. Generally, rain is the larger contributor

to annual precipitation, with June the peak month.

Spatial and seasonal patterns of precipitation are presented graphically in Appendix B. Only the western

fringe of the area has less than a 50% probability of

receiving at least 0.1 inch per week, and that only during

the winter months. The weekly probabilities of receiving

1.0 inch or more of precipitation generally increase from

west to east but are lower than 50% throughout the area.

In June the probabilities are at their maximum and range

from 30% to over 40%. In January the prevailing wind

direction is from the northwest with a mean speed

ranging from 8 to 13 mph in the Upper MississippiLower Missouri region. North of a line from Chicago to

Kansas City, April winds are most frequently from the

northwest at speeds of 10 to 15 mph; south of this line

winds are predominantly southerly with speeds of 10 to

15 mph. July winds are characteristically from the south

throughout the basin, with the exception of the Mississippi headwaters in northern Minnesota where northwest

winds prevail. Speeds range from 7 to 11 mph. October

patterns tend to be southerly at 8 to 12 mph. Wind rose

patterns for a more detailed picture are included in

Appendix B.

The variation in atmospheric stability across the

study area is not well documented. However, Holzworth(3) has shown that the mean maximum mixing

depth variations follow a pattern as depicted in Figure

111-2. Hosler(4 ) has also shown that the percent frequency of occurrence of inversions in the area follows a

pattern as shown in Figure 111-3.

POWER SUPPLY AND DEMAND

Power Supply Characteristics

The Upper Mississippi-Lower Missouri basin region

encompasses one of the major power supply/demand

areas of the United States. The study area proper

accounts for approximately 10 percent of the total

power demand of the contiguous United States. With the

"air envelope" region included, about 25 percent of the

total U.S. demand is represented. The extended area

(including the "air envelope") for which power supply

projections were made, completely or partially covers 19

of the 48 PSAs designated by the FPC (Figure 111-4).

Electric power in the study area is supplied primarily

from steam plants. Hydroelectric generation is utilized

to some extent throughout the region, but has only local

importance, and that mainly in the upper reaches of the

Missouri River.

Coal has been the dominant fuel for power production in this region, although oil and gas are important in

the southwest portions. Commercial nuclear energy was

introduced in this area with the Dresden-1 plant, which

became operational in 1960. In the past decade, substantial increases in the use of nuclear-generated electric

energy have developed in this region. This trend is

expected to continue and nuclear energy may be

dominant in regional electric generation by 1980 or

shortly thereafter.( 5-

8 ) Nevertheless, portions of the

study area are underlain by major fossil fuel reserves; in

these regions fossil fuels will probably remain an

important factor in power generation, at least for the

period covered by the study.

Power Demand Projections

The FPC has published regional power surveys and

forecasts covering the contiguous United States. The

power forecasts are published for each power supply

area, with some of the information given for individual

load centers. Thus, for those power supply areas which

are traversed by the boundaries of the "air envelope," it

is possible to derive forecasts for those load centers

contained within the boundaries.

The FPC projection of energy demand covers the

period 1970-1990, with forecast demand reported in

10-year increments. For use in The Year 2000 Study,

the forecasts obtained for the individual power supply

areas were interpolated and extrapolated (beyond 1990)

to provide annual demand data for each year of the

period 1970-2000.

The resulting projection for the extended study area

is summarized in Table III-1.

111-3
4 6 t

240 ----- 80 . {,..___..- " 440 4 ~

660 20-."340 43,

410

- ---- --- - 40460

0 

------ JIO - - -- &--

-"0*340/40 

- -4--0 370__ , 

:060- 0 --

440 --

710 430 46 ,(3 _ 55 * 2 - 520

43

6460

0 520 390-- 4-

--- 0 500 6730 10

\ 1240

-64 6 JANUARY

402 126 980 

0 120 24.,.r24 8 28

8 28 __9 24 20 2 8 1 -'1 2290 -130 - 40 n4

2600-0--- --- 3 0- ) 900 0 3 1408 4' 0

1720 

41400.--,

416092-0 12 1410

273 - --

322 3 -0 1380 450 -

3 13io6401307}I8

4900 

1820 ;" 

46, 37 o1630--8

3400 1590 -' 0

-" 1490 '1640 1200

2 4 2760 -" -"- 

2 2 740 2290

281300

026 Q- _- 1830 130 4

024 0JULY 1

I8 .1280

161

Figure 111-2.-Mean Mixing Depth, Meters, for January and July.

111-4
20 o 5 z2502

25 .. ' 22 25 30 35.0

35-

40 t_- 1 1

45 --... X-.3-.0- 5

- - 0

35 3

50 45 25

\ 2015

40 .10

WINTER 5

25 /-30 3

- - -- - s . ---- 25 - 2

021

20 ----- 1.J 43' 

35 C. - ~ ~

3 22 

t 20

1 15

1015 20 0 10

SUMMERFigure II-3.-Percentage Frequency (% of Total Hrs) of Occurrence of Low Level Inversions

(Below 500 ft).

III-5
26

- 16 

27

13

31 28 17

- 12 

~5 40

29

r'-- 19 0

~-----~-33

) 1 20f

25--

'*r -- I\~S

Figure 111-4.-Federal Power Commission Power Supply Areas Included in the Study Area.
Annuw

fo

Area

Total

*An exp

Appendix

General

Since

for the generation of electricity, generation centers tend

to coincide with load centers; other things being equal,

the tendency is to minimize transmission costs by

building generating stations close to the load centers.

Only in the case of large hydroelectric installations on

the Upper Missouri River is there appreciable displacement of generation centers from load centers.

Nuclear plants also tend to be located near load

centers, although their degree of approach is, of course,

limited by siting regulations.

The tendency toward coincidence of load centers and

generation centers is likely to continue in future years. It

may be tempered, however, by the availability of cooling

water, at least in the later years of the study and-at

least for nuclear plants-by a tendency toward clustering

plants in sites more remote from load centers.

The FPC survey for the West Central Region,( 5

which covers the major part of the extended study area,

identifies some 50 generation centers-areas where

power plants are now located or are likely to be located

in future years. Again, these areas coincide closely with

load centers in the region. This and the other surveys

referenced,( 5 -8) identify over 100 load centers within

the extended study area, as shown in Figure I1I-5. Many

of the smaller centers identified are combinations of

several small, close-by centers.

Seasonal Variation in Load

Seasonal variations in load reflect the changing

electrical demand throughout the seasons of the year.

Generally, winter heating loads and summer air conditioning loads produce peaks in the electrical demand

or load curve.

Predictions of the probable monthly power generation in"the year 2000 are, of course, quite conjectural.

They were derived by extrapolation from FPC projection reports for the period 1970-1990. Of the 19 PSAs

included in this study area, details of projected monthly

POPULATION AND LIVING PATTERNS

Evaluations of population groupings, dietary habits,

patterns of food production and consumption, and

patterns of work and recreation activities are required

for an accurate overall evaluation of exposure to

radiation in the environment. Considerable effort was

required to provide adequate coverage of these population-related factors. At the same time, the size and

diversity of the area required categorization and grouping of much of these data to reduce the total quantity to

a manageable size, without detracting from the validity

of dose evaluations.

Population and Population Grouping

The central study area includes 479 counties extending over eight midwestern states. The 1970 population

of this area was 18 million,( 9 ) ranging from the highly

urbanized residents of the large cities included in the

area to the rural inhabitants of sparsely-settled regions in

the western portion of the study area. The regional

population density varies from 2.1 persons per square

mile in Wheeler County, Nebraska to nearly 10,000 per

square mile in St. Louis.

The 1970 population was projected to the year 2000,

county by county, based on changes in the area's

population over the last 20 years(9-1 1) and on likely

future trends. Table 111-2 summarizes the present population and that projected for 2000, for those portions of

the states included in the study area. The projected

growth rates appear high for several of the states because

the study area included only the fastest growing portions

of those states. Hence, the projected year 2000 population figures reflect both the total growth expected for

the midwestern states and an extension of the current

trend toward increased urbanization.

111-7

TABLE Ill-1 power generation in 1990 were provided only for

PSA-20; and only two FPC regional reports, for the East

al Energy Demand Summary: Projection Central and South Central regions, gave current seasonal

r Extended Study Area, 1970-2000* variations. However, the peak capacity requirements for

1990 were provided for all PSAs.

Annual Demand, Millions of Use of summer and winter peak capacities and their

Kilowatt-Hours ratio, together with the assumption that monthly power

generations are proportional to monthly peak capacities,

1970 1980 1990 2000 provided the basic information needed to generate four

curves of seasonal variations. These variations, expressed

in units of monthly percent of annual generation, can be

422,073 826,033 1,511,027 2,428,736 found in Appendix B.

The assignment of each PSA to one of the four curves

anded regional demand projection is included in depicting monthly variation in electrical demand was

ix B. done by reference to the summer and winter peak

capacities and their ratio. These assignments of PSA to

monthly generation curves are shown graphically in

Appendix B. The contributions to monthly generation in

tion and Load Centers the study area, by the different plant types considered,

were estimated in the course of this study. These

this area utilizes steam plants predominantly estimated contributions are shown in Figure 111-6.
0 0)

0 0 0

O0

0 0

10

0 T

AREAOF CRCLEINDIATE

RELAIVE IZEOF DMAN

CETR NYER201

Figure 111-5.-Major Load Centers in Study Area.
FOSSIL

EJ LWR

SLMFBR

'N

V)

I

O-

-J

-E

100

80

60

K

N

K

JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC

YEAR 2000 A.D.

Figure 11-6.-Contributions of Plant Type to Monthly Demand.

111-9

140

120 }-

K K

40

20

I

I

I
The Centroid Concept

Central Study Area Populatio

Stated 1

Illinois

Minnesota

Iowa

Missouri

South Dakota

Nebraska

Kansas

Wisconsin

Total

Actual

1970

3,580,742

3,291,667

2,789,893

3,938,474

399,947

1,193,159

983,662

1,693,398

17,870,942

(1) Includes only the portion of state included in s

Much of the data on food consumption

tional activities in the area is categori

population groupings as defined and used

Bureau of Census:

" Urban-Households in localities wi

25,000 inhabitants, and in closely-s(

areas surrounding cities of 50,000 or n

ants.

" Rural non-farm-Households outside

calities but without a farm operator.

" Rural-Households outside of urban lc

a farm operator.

Classification of the entire population

area would be a major undertaking for which

data are available. However, the Agricultu

Service of the U.S. Department of Agric

survey of dietary habits in the North Centr

the United States( 1 2) (including the study

wished estimates that 70% of the population

fell in the Urban classification, 20% into Rur

and 10% Rural farm.

Using these guidelines, the study area po

classified by counties, assigning each cou

population type. In this process the counti

ordered by population density. Counties

placed in the Urban classification, in orders

population density, until 70% of the total

tion was in that category. Counties repr

following 20% of population were classifi

non-farm, and the remaining counties wer

Rural-farm. In reality, of course, few if any

consist of a single category of households;

three categories will be represented. In the

additional data, however, the categorizatio

provided a reasonably accurate means of

dominant population group in each county.

dure thus simplified the selection and des

"typical individual" from each region, asv

for dose evaluations.

in The total of 479 counties in the study area, even after

categorization by dominant household type, still reProjected quired that an undersirably large number of calculations

2000 be performed in regional dose evaluation. To aid in

reducing the model to manageable size, the concept of

5,073,300 the "centroid" was introduced.

5,851,600 A centroid is essentially a unit in which a county, or a

3,350,100 group of counties, is represented by a single point. The

7,675,500 entire population of the centroid is assumed to be

433,000 concentrated at that single point, and the population1,831,100 related activities in the centroid are also assumed to

2,491,900 occur there. Thus, with judicious selection of the point

2,375,300 representing the centroid, calculations leading to dose

estimates for the centroid need only be carried out

29,091,800 relative to that point. The centroid concept also permitted the grouping of certain non-urban counties,

study area. having like categorization (Rural non-farm or Ruralfarm) and similar characteristics (crop types, agricultural

practices, etc.) into single centroids. Thus, the 479

and recrea- counties were combined into 300 centroids. These

zed by the centroids, and their categorization, are shown in Figure

by the U.S. 111-7.

The centroids were identified by number, in north-toth at least south, east-to-west progression, to facilitate scanning

settled fringe routines performed by elements of the HERMES comnore inhabit- puter code.

The representative point of each urban centroid was of urban lo- defined as the location of the largest city in the centroid.

For non-urban centroids, the location of a city or town

)calities with nearest the geographical center of a centroid was chosen

as the definitive point. For convenience, each centroid

)f the study was given the name of the city in which its definitive

insufficient point fell.

ral Research Eight centroids were chosen for detailed study of

ulture, in a radionuclide concentration effects and dose relational Region of ships. These "representative" centroids, and some of

area), pub- their characteristics, are shown in Table 111-3. Use of

of the region these eight centroids permitted the use of detail in

al non-farm, reporting the results which would not have been feasible

if attempted for all 300 centroids. The "representative"

pulation was centroids include four of Urban classification, three

nty a single categorized as Rural non-farm, and one Rural-farm

es were first centroid. They were chosen to represent different

were then geographical regions of the study area, as well as

of decreasing different population groupings. A preponderance of

area popula- urban centroids was selected because of the greater

esenting the overall importance of urban centers in regional dose

ed as Rural evaluations.

e designated

counties will Food Production, Consumption, and

in many, all Transportation

absence of

n by county The food consumption patterns of individuals form

selecting the an important aspect of the overall living patterns.

. This proce- Evaluation of potential radiation dose received through

cription of a food pathways, moreover, requires detailed analyses of

was required dietary habits and of the relationships between production and consumption of foods. Considerable attention,

III-10

TABLE E111-2
193

62 6 133

227

195 14 134

6 54

262 22811 135 20

994

284 26 229 17 '137

264 3 2 95 21

285 f2651 202 t42 58

295 59 25 2

iax 204 1 143 103 104 60 26 3

286 268 235 234 203 v62 61

269 236 237. 206 144'106 105 63 28 17 4

287 17 145 109 m a 64 65 29 30 6 5

96238 208 148 146 110 6 33 7

289 240 201 182 149 112 1.1 68 3

13 9 34 1

297 24.8 150 70 q

29 3n

74 247 246 213' 15 17 f 2

q9 X 7 4 - 85 15 838 37 1

00 276 "'248 57 39 1

291 2 725 214, 19 120 4 ,1

27 252 21 7 t'1 4

2386 19 154 79 42

81 8 43

280 255 21 7 - 12 44 1

188 84 3

283 86

Urban 260 190 is 126 5 11

259

94

15 9 5,4

Rural

Figure lII-7.-Designated Centroids in Study Area.

III-11
TABLE 111-3

Characteristics of Representative Centroids

Number Name County State

20 Gleason Langlade

Lincoln

34 Princeton Bureau

68 Davenport Scott

87 St. Louis

Popular

Classification 1970(a)

Wisconsin Rural Non-Farm 41,440

Illinois Rural Non-Farm 38,080

I owa Urban

St. Louis City(d) Missouri Urban

nation 1970

Square Population

2 00 0(b) Miles Density(c)

37,500 1,748 23.71

38,600 866 43.97

141,360 235,800 454 311.37

607,720 452,600 61 9,962.62

143 Charles City Chickasaw

Floyd

Howard

Mitchell

166 Minneapolis Hennepin

222 Kansas City Jackson

Iowa Rural Non-Farm 58,530

Minnesota Urban

Missouri Urban

54,300 1,446 40.48

955,620 1,605,400 567 1,685.40

644,950 870,500 603 1,443.62

238 Mapleton Ida

Monona

I owa Rural Farm 21,290 11,100 1,130

(a) Bureau of Census, 1970 Census of Population, rounded to nearest 10.

(b) Projected. Rounded to nearest 100.

(c) Individuals per square mile.

(d) Includes only the "central city" of St. Louis, and not the surrounding suburbs.

therefore, was directed to evaluating these relationships

in the study area.

Food production and consumption data for the

1964-1970 period, used to establish the basic patterns

for the area, were assumed to be escalated uniformly to

meet requirements of the year 2000 population. This

procedure skirts the question of the effects of evolution

of farming practices-and of dietary preferences-during

the intervening period. However, the assumption of

uniform escalation of today's patterns was made deliberately, to avoid introducing too many unknowns to the

problems, and to provide a positive frame of reference.

Production

The Upper Mississippi-Lower Missouri basins include

some of the most important food production areas of

the United States. Although some regions are devoted

primarily to staple grain production (mainly wheat and

corn) the area as a whole has a highly diversified

agricultural economy.

A basic list of 73 food types considered important to

the study was developed and can be found in Appendix B.

(This list contains several food types not produced in

the study area.) Production data for these foods( 1 3)

were gathered at the county level and combined, where

needed, to provide production figures by centroid.

The units of measurement between production and

consumption were not uniform. Since consumption data

were represented by pounds and that unit of measure

proved to be most workable for dose calculations, all

production data were converted to pounds except for

eggs which retained the common unit of dozens. For

example, production of beef was given as the number of

beef cattle sold for slaughter. The assumption was made

that each steer weighed approximately 1000 pounds

when sold and that a 43% loss in processing occurs

resulting in 570 pounds available for consumption. In

the area of produce, production may have been reported

as pounds, bushels or acres in production. Agricultural

statistics for 1967(14) published by the U.S. Department of Agriculture were utilized to provide basic

production figures for the conversion to common units

of pounds.

Consumption

The pattern of food consumption was found to vary

with the degree of urbanization. Families in urban areas

tend to select more commercially prepared foods; farm

111-12

18.84
families use more ingredients for home-prepared

foods.( 1 2 ) The 73 food types utilized for food production data were condensed into 35 categories of food

consumption. This categorization provided a convenient

model for handling of data. The addition of processed

foods in most fresh food categories allows for the

subsequent dose calculation for consumption of a food

type on a year-round basis. In Appendix B food

consumption per household is tabulated for Urban,

Rural non-farm and Rural-farm areas..

The effect of crop irrigation was included because

irrigation involves a waterborne route, other than rainfall, for the potential deposition of radionuclides.

Appendix B includes a table summarizing the importance of irrigation in food crop production in the study

area. For example, 56 of the 300 centroids grew fresh

berries and approximately 50% of the berries grown

were irrigated.

Food Sources

The food consumption habits of modem man, be he

urban or rural, are cosmopolitan. A perusal of the food

consumption categories in Appendix B would reveal

several food products such as fresh citrus fruit that

obviously cannot be supplied from within the study

area. It is also apparent that urban centroids will not be

able to produce sufficient foods to satisfy their consumptive demand.

There is an obvious correlation to be made between

the proximity of a food production area to an environmental pollution source and the degree of pollutant

loading that may be found in the marketable crop.

However, it is not obvious that the effect of the

pollutant loading will be applied only to food consumed

in the immediate vicinity of the polluting source.

Foodstuffs are transported, sometimes over great

distances, to various marketing regions for consumption.

To accommodate these factors (shipment between

centroids, imports from outside the study area and

production-consumption within centroids) a transportation algorithm( 1) was developed which adhered to the

following procedures:

" Production of a centroid is applied first to satisfy

consumption requirements of that centroid (this

step sets up a pattern of food surpluses and

deficits).

" Deficits in a centroid are satisfied preferentially by

food shipment from the nearest centroid having a

surplus.

" Any deficits remaining after intra-area shipments

are satisfied from outside the study area.

" Fresh food deficits are satisfied first; any remaining surplus is available to satisfy demands for

processed food.

Three production sources were allowed to fill each

deficit: the centroid under consideration and two

sources from outside the centroid. If the deficit could

not be satisfied from within the study area, it was

assumed that sufficient imports from outside the area

(assumed to be free of radionuclides) to satisfy the

demand would be obtained. Eventual study of other

parts of the country, where these imported foods are

grown, will permit evaluating their contribution.

Recreation Patterns

An individual's activities create exposure patterns

that contribute to the overall dose he might receive.

Standing or sitting on a river bank while fishing may

bring an individual in contact with material that has

been transported by the river and deposited upon the

river bank. Likewise, an individual submerged in water

while swimming may be exposed to different radioactive

material.

Recreational habits, though not significant in contribution to the total body dose, were included to account

for all likely pathways. The activities included were

swimming, boating and water skiing, fishing, water fowl

hunting, and upland game bird hunting. The category

"boating and water skiing" includes canoeing, sailing,

and barge or boat fishing, whereas "fishing" includes

only bank fishing. An estimate was made for the number

of hours per month a typical person in a certain segment

of the population would devote to such recreational

activity. The seasonal characteristics of each activity

were considered in estimating the hours involved.

111-13

IV. POWER GENERATION

PROJECTIONS

FORECAST OF CAPACITY ADDITIONS

The projected capacity requirements to satisfy electrical energy demands for the study area were computed

by first interpolating available FPC projections (provided

on a 10-year basis to 1990) and then extrapolating to

later years. This was done for each PSA: the individual

results were summed to give the total study area

capacity requirements. This total capacity was calculated

annually from 1970 through the year 2002 and for

two-year periods thereafter until 2020, 20 years past the

study period. The calculations were extended past the

year 2000 to avoid possible "end-effect" perturbations

to the calculated capacity requirements.

Annual energy requirements for each PSA were

calculated between the specific data points in the energy

projections by assuming a constant growth rate between

the specified points. The energy produced by plants

existing before the study period was then backed out

and the resultant energy was that to be provided by new

plants built and brought on-line in the study period

1970-2020.

Annual plant factor curves describing the operation

of all plant types over their 30-year lifetimes were those

used in cost-benefit studies of the U.S. LMFBR

program.(1 55)

Two plant factor curves were used: one to simulate a

base-loaded plant and the other a load-following plant.

Generally, base-loaded plants provide a more or less

constant base energy supply and therefore have higher

plant factors, whereas the load-following plants fill in

the peaks in demand and have lower plant factors. The

30-year average plant factors for base-loaded and loadfollowing plants were 62% and 46%, respectively. The

maximum annual plant factors of these plants were 80%

and 75%, respectively.

With the energy demand for all time periods and the

plant factor curves available, the projected capacity

required was determined using the following basic

equation:

1 =

where

Ci = capacity required in the ith period, Gw(e).

Ei = energy demand in the ith period, Gw hr.

ti = time in ith period, hrs.

fpi = average plant factor in ith period.

Additional input required was the ratio of new

base-loaded plants to total new plants. This plant ratio

was assumed to be 0.66 in 1970, slowly increasing to

0.80 in the year 2020. This allowed the total system

plant factor to stay between 60% and 65%. The resultant

total system capacity projection for the entire UMRB

extended study area is shown in Figure IV-1.

PROJECTION OF GENERATING PLANT MIX

Projections of the generating plant mix assumed for

the year 2000 were obtained using a modified version of

a linear programming optimization model of the national

power economy. Descriptions of this model may be

found in References 15, 16, and 17.

Basically, the model minimizes the total system cost

of the power economy over a specified time span subject

to constraints on the annual capacity of new plants

built, uranium price, the fissile plutonium stockpile, and

the rate of introduction of LMFBRs. Nuclear plants

actually scheduled to become operational through 1976

(reflecting current commitments) are specifically accounted for in the projections.

Plant types considered in the model were LWRs,

LMFBRs, and fossil-fired plants. Gas-cooled reactors

were not included in the study. Depending on a number

of factors, the inclusion of gas-cooled reactors in the

study chould change the results to the extent that their

radiological release characteristics differ from those

reactor types considered. However, data obtained in the

past two years suggest that including gas-cooled reactors

would not result in an appreciable change in the overall

results of the study.

The area bounded by the line around the outer air

envelope was considered a closed system. The energy

demand was projected for that region and the cost data

pertinent to the region were generated.

Other cost data were taken from Reference (15). The

linear programming model yielded optimized solution

consisting of the generating plant mix through the year

2000. In meeting the year 2000 capacity requirements a

total of 184,000 MWe of LWR capacity and 172,000

MWe of LMFBR capacity was needed in the entire study

area including the air envelope. In the central study area

alone, 59,000 MWe of LWR capacity and 74,000 MWe

IV-1
1000

800

600

400--

V)

200 

TOTAL

ADDED

20

0 1970 1980 1990 2000 2010 2020

YEAR

Figure IV-1.-Capacity Projections for UMRB Extended Study Area.
of LMFBR capacity were calculated to meet the power

requirements.

REPROCESSING PLANT REQUIREMENTS

The number of reprocessing plants within the study

region is a function of the individual plant designs

assumed and the amount of fuel to be processed. Several

important constraints were imposed on the fuel reprocessing system in this study. All the spent fuel generated

in the region (including the air envelope) was assumed to

be reprocessed within that region. Fast reactor (LMFBR)

fuel was assumed to be reprocessed in aqueous separation plants at one-half the design feed rate for LWR fuel.

The decreased feed rate allows LMFBR fuel reprocessing

in plants designed for LWR fuels with additional product

shielding and standard methods of criticality prevention.

The reprocessing plant characteristics were based on

1500 tonne/year aqueous separations facilities, except

for the Midwest Fuel Recovery (MFR) plant, which is an

existing 300 tonne/year facility employing the Aquafluor process.

Within the constraints outlined above, the number of

reprocessing plants was determined from the annual

spent fuel load as follows:

1. The MFR is loaded with the first 300 tonnes of

LWR fuel. (This plant was assumed not to accept

LMFBR fuel.)

2. One aqueous separations plant is built for each

1500 tonnes of equivalent LWR fuel (each typical

1000 MWe LWR discharge 30 tonnes of fuel each

year).

3. A tonne of equivalent LWR fuel is one tonne of

LWR fuel or one-half tonne of LMFBR fuel.

All aqueous separations facilities built in the region

have the same ratio of LWR fuel to LMFBR fuel. Thus,

the aqueous reprocessing plants are in effect dualpurpose facilities, and are represented as such for the

source map.

The start-up dates for each plant were used in the

calculation of the likely background concentration of

long half-life radionuclides on the ground. The allocation

of fuel to reprocessing plants within the constraints

outlined above resulted in the start-up shcedule shown in

Figure IV-2. In the year 2000 the region requires nine

reprocessing plants: the MFR plant in Morris, Illinois,

and eight 1500 tonne/year aqueous-process plants. The

quantity of spent fuel generated in the study area, and

thus the reprocessing requirements of this area, were

determined by calculating the annual spent fuel discharges by reactor type from all nuclear power plants

included in the study. The spent fuel discharges (reprocessing requirements) derived from these calculations are

shown in Table IV-1.

PLANT SITING CONSIDERATIONS

Operation of the HERMES model to provide estimates of population dose requires that the nuclear

TABLE IV-1

Study Area Annual Reprocessing Requirements

LWR Fuel,

Year Metric Tons

1970

1975

1980

1985

1990

1995

2000

2005

LMFBR Fuel,

Metric Tons

0

433

1,078

2,636

4,929

6,545

5,721

4,276

0

0

0

0

113

786

2,455

4,483

facilities selected by linear programming techniques be

assigned to specific sites in the study area. The dependence of dose on the locations of nuclear sites

introduces the potential for considerable uncertainty in

the final estimates, and makes it essential that judgment

be used in the selection of nuclear sites.

Criteria used for selection of nuclear facility sites for

the study area were:

" Conformance with AEC siting criteria as expressed

in Part 100, Title 10 of the U.S. Code of Federal

Regulations (10 CFR 100).(18)

" Sufficient water supply availability for plant

operation (as a minimum, sufficient supply for

makeup to recirculating, evaporative cooling systems).

" Minimizing of transmission costs to load centers.

" Meeting of power demand projections for each

PSA, and distribution of demand within each PSA.

Plant sites assumed for the year 2000 in the UMRB

area are shown in Figure IV-3. In accordance with usual

utility practice, these sites were preponderantly designated as multi-plant sites, accommodating up to four

plants (or equivalently, 4000 MWe of capacity) per site.

It should be understood that these site identifications

are purely hypothetical for plants not yet built or under

construction. The study assumes location only for the

purpose of permitting analysis.

MONTHLY AVERAGE PLANT FACTOR

SYNTHESIS

The plants selected by the linear programming model

were assigned to sites in a one-to-one correspondence by

ordering the power plants on the basis of year built and

the sites in the order defined by site priority numbers.

Fossil plant sites were preferentially located in

low-cost fossil fuel areas. These sites were restricted to

fossil fueled plants only. Other sites had no fossil fueled

plants, but were allowed to contain any mixture of

nuclear plants. The separation of fossil and nuclear sites

was made to simplify the computerized selection

process.

IV-3
T I

}-50

40

I.

00

I20

CD

00

1970 1980 1990 2000

YEAR

Figure IV-2.-Reprocessing Plant Requirements.

IV-4

m 0 - - - - - - - - -0009r

I
A FOSSIL POWER PLANT

" NUCLEAR POWER PLANT

a REPROCESSING PLANT "" a n

Figure IV-3.-Plant Sites Assumed in Study Area.
140 I*I I

® BWR

P W R

S,9 P W R

120 LMFBR

100

= 80

C60

0

J

40

20

0

JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC

YEAR 2000 A.D.

Figure IV-4.-Annual Nuclear Energy Generation Pattern for Study Area.

IV-6
In the plant siting process, 25 percent of the LWR

plants were randomly designated as BWRs (assuming

each current vendor to have an equal share of the year

2000 market); the rest were designated as PWRs.

The study area was divided into four subregions based

on characteristics of the variation in power demand

through the year. By computer techniques, each plant in

a subregion was assigned an average plant factor for each

month of the year so as to satisfy (1) monthly and

annual energy demands of the subregion, and (2) the

prescribed annual average plant factor for each plant. All

power plants were required to shut down once each year

for a six-week maintenance and (for nuclear plants)

refueling period. Base-loaded plants were not allowed to

shut down during peak demand periods. Resultant

annual patterns of generation by nuclear plants in the

study area are shown in Figure IV-4.

IV-7

V. RADIONUCLIDE RELEASE CALCULATIONS

The patterns of anticipated radionuclide releases in

the study area, the magnitudes of these releases, and

their variation throughout the year, are essential quantities in the estimation of the potential radiation dose to

man in the year 2000. These data were developed

through the use of the HERMES model codes,

RADREL, DOSIS, and REPARE.( 1

RADREL utilizes the data on site locations, siting of

plant types, and plant monthly operating status, as

discussed in Section IV, to provide the framework for

the "source map"-the compilation of data defining the

patterns of radionuclide release. Using the DOSIS code

as a subroutine, it calculates the month-by-month release

rates of radionuclides at all nuclear power plant sites,

factoring in the monthly plant factor for each plant.

REPARE is run separately, using the fuel reprocessing

requirements derived from the calculated fuel throughput of power plants (see Section IV), to determine the

number of reprocessing plants required in the study area,

the fuel mix processed by each plant, and the estimated

rates of radionuclide release from each plant. These data

are transmitted to RADREL for inclusion in the source

map.

NUCLEAR POWER PLANTS

The DOSIS code is used to calculate release rates

typical of those expected in the year 2000 for the

reactor types expected to be operating. This code is

designed to simulate the major nuclear plant characteristics affecting radionuclide release to the environs. The

majority of the factors that reflect the anticipated

performance of plant systems were obtained from SARs.

Advanced treatment components were then assumed to

be added to the plant radwaste systems, such as are

anticipated to be required to meet Federal guidelines.

The program was used to evaluate release rates with

these advanced treatment systems for both the LWRs

and LMFBRs, as well as to include plant factor

variations in the consideration of the regional problem.

The subdivision of potential reactor types for the

study region and year (BWR, BWRA*, PWR, PWRA*,

and LMFBR) was made by the codes preceding DOSIS

in recognition of the potential for simultaneous employment of the several types of nuclear plants for electrical

energy production in the study region. The DOSIS code

*The "A" designates advanced: designating reactor models assumed to be built after 1985.

considers each reactor type separately. Code segments

for the advanced LWRs include the effects of anticipated

increased fuel rating.

The DOSIS computer program follows the radionuclides included in the study from their point of origin

in the reactor system through the coolant and radwaste

systems to their eventual point of capture for solid waste

handling or their point of release from the reactor

(Figure V-1).

At each step in this path, various equations, generally

with different constants for each radionuclide, are used

to calculate the effect of that part of the system on the

eventual release rate for the particular radionuclide to

the environs. As used in the UMRB study, the constants

defining radionuclide transport between the point of

generation and the coolant were adjusted to yield

coolant concentrations similar to those published in

typical SARs.

Fission product generation rates in the fuel are

calculated, release fractions from the fuel matrix and

fuel failure probability distributions are factored in, and

the resulting release rates to the coolant determined

using the assumed fuel cycle characteristics shown in

Table V-1. Activation products are generated in the

vessel, coolant, fuel clad and spacers. These activated

materials are also assumed to be released at specified

rates to the coolant. The coolant is treated by a cleanup

system; in addition, a portion is continually removed to

the radwaste system. Gaseous releases from various tanks

as well as from the primary coolant system yield the gas

treatment system input stream. Liquid streams go by

various paths through the radwaste systems. Specified

fractions of the cleaned-up streams are recycled to the

coolant; the remainder is released to the environment.

TABLE V-1

Fuel Cycle Characteristics for

Plant Types Considered

BWR PWR BWRA PWRA LMFBR

No. of Reload

fractions

Goal Burnup

(MWD/MT)

Power Density

(MW/MT)

Max. Linear

Heat Rate

(kw/ft)

4 3 4 3 5

27,500 30,000 27,500 30,000 80,000

22 35 24 39 150

18 18 20 20 20

V-1
NUCLEAR POWER PLANT MODEL

FUEL mi COOLANT S ACTIVATED 14m MATERIALS

C)

SOLID

WASTES

'-'-

COOLANT

CLEANUP

SYSTEM

4

I

I

--

S

I

1

LIQUID

RELEASES

GASEOUS

WASTE

SYSTEM

i

GASEOUS

RELEASES

Figure V-1.-Nuclear Power Plant Model.

LIQUID

RAD. WASTE

SYSTEM

qw
The radwaste systems for each reactor type are also

simulated by the DOSIS code. Flow charts for the three

types considered are given in Figures V-2 to V-4. The

numbers indicate points at which radionuclide flow rates

were calculated. Detailed listings of these flows are

included in Appendix C. The PWR liquid treatment

system flow chart (Figure V-2) illustrates the two paths

available for waste streams to follow. One, shown as

starting in the waste collection tank, is the path used for

streams that could be characterized as chemically clean.

These streams are assumed to have radionuclide concentrations similar to that of the primary coolant. The other

path, starting in the floor drain tank, is used for wastes

that are low in radionuclide concentrations and are

contaminated to a high enough level with nonradioactive

materials so that discharge, rather than cleanup and

return to the liquid system, is desirable. Additionally,

steam generator blowdown flows are treated through the

floor drain-flow path in this simulation.

Simulation of the BWR liquid treatment system in

the program (Figure V-3) is slightly different from that

of the PWR system, with an additional flow path

considered. These paths, shown in the figure, include:

" Treatment of liquid effluents from the coolant

cleanup system. These effluents, after phase separation and configuration, are routed to waste

collection tanks.

" Treatment of miscellaneous waste flow streams

which are routed directly to the waste collection

tanks. Combined with the effluents from the phase

separator, centrifuge, and evaporator, these wastes

are filtered and demineralized. Part of the treated

stream is returned to condensate storage; the

remainder is released.

" Treatment of floor drain streams, similar to those

in PWRs, but without the steam generator blowdown flows.

The BWR gas treatment system includes a routine for

calculating an appropriate delay time for the major gas

stream isotopes, as would be provided by a charcoal bed

system.

The LMFBR waste treatment system simulation

(Figure V-4) is much simpler than the LWR treatment

systems because only small quantities of aqueous liquid

wastes are expected to be generated; these were assumed

to be concentrated and solidified for disposal. In

addition, several radionuclides of interest are assumed to

react with the sodium coolant and to be removed from

the coolant in the cold trapping system.

The LMFBR gas treatment system is modeled in the

routine called FREON. As with the other gas treatment

systems, provisions are made for a delay time and for gas

filtration. The gas stream is then directed to the FREON

subroutine where a removal decontamination factor

(DF) of 300 is assumed with the majority of the noble

gases being bottled for storage.

The assumptions used for the treatment systems as

applied to the different reactor types were:

" BWRs use a charcoal delay filter system on the gas

stream as the principal means of reducing releases.

The filter is assumed to be operated at room

temperature.

" All PWRs have a reduced release of tritiated

wastes, such as would result from the use of a

system such as the Westinghouse Boron Thermal

Regeneration System. Also, a fraction of the

treated floor drain flow is recycled to the reactor

coolant.

" LMFBRs contain all liquid wastes and treat all

gaseous releases in an advanced treatment system

with decontamination factors of 300 for Krypton

and Xenon isotopes, such as might be expected

with a refrigerant absorption system featuring

bottling of noble gases.

" The major difference between advanced and present day LWR designs is the maximum linear heat

rate of the fuel. This affects the diffusion coefficient of the fission products in leaking fuel. The

higher linear heat rate results in higher potential

release rates for the advanced plants.

REPROCESSING PLANTS

The nine reprocessing plants in the study region

include the MFR plant at Morris, Illinois, and eight

dual-purpose aqueous-process plants. The dual-purpose

plants process a mixture of LWR fuel and LMFBR fuel

at a rate equivalent to 1500 tonnes/yr of LWR fuel or

750 tonnes/yr of LMFBR fuel. Reduced feed rates for

LMFBR fuel allow LWR fuel separations plants to

process LMFBR fuel with additional plutonium cycle

shielding and standard criticality prevention provisions.

Potential releases from the reprocessing plants were

calculated by the REPARE computer program included

in HERMES.( 1) The REPARE code was designed to

model the major components in the several process

streams which ultimately lead to products, waste storage

or effluents. No attempt has been made to model the

separations process; instead, this code addresses itself to

modeling the effluent treatment systems. Each of the

major components other than the separations process is

represented by a program subroutine. This component

modeling system allows flexibility to simulate various

decontamination schemes and also facilitates use of

decontamination data on an overall system basis or on

an individual component basis.

The characteristics for the base case reprocessing

plants were drawn from the MFR plant design data(19)

and the projected 5-tonne/day aqueous plants described

in the Preliminary SARs by Allied Chemical Corporation( 2 0 ) and Atlantic Richfield Company.(2 1) Schematics for the flow sheets for the MFR plant and the

aqueous plant simulated in REPARE are shown in

Figures V-5 and V-6.

The MFR and aqueous plants have both high- and

low-level waste solidifications systems. All the effluents

from the plant enter the atmosphere through the stack.

The plant's waste treatment systems provide several

levels of decontamination from the separations process

to the waste solidification systems.

V-3
NOTE: CIRCLED NUMBERS REFER TO COLUMN

NUMBERS IN TABLES C-4 AND C-5

GAS HOLDUP O GASEOUS

FITANKRFILTER RELEASES

REACTOR

SOLID WASTES

FUEL COLLECTION DEGASIFIER HOLDUP FILTER

COOLANT

|~ 

SOLID

WASTES'

DEMINERALIZER EVAPORATOR TANK

COOLANT

CLEANUP

SYSTEM SOLID WASTES SOLID WASTES

SOLID

WASTES

TURBINE

L

FLOOR

DRAIN FILTER TANK

SOLID WASTES

Figure V-2.-Pressurized Water Reactor Model Schematic.

0o LIQUID

RELEASES

*INCLUDES STEAM

GENERATOR BLOWDOWN

C

i
- - REACTOR NOTE: CIRCLED NUMBERS REFER TO C

R TI NUMBERS IN TABLES C-2 AND

I FUELWAT

COOLANT COLLECTION FILTER DEMINERALIZER TANK

SOLID SOLID

COOLANT PHASE WASTES WASTES

I CLEANUP SEPARATOR 

SYSTEM CENTRIFUGE

SOLID

WASTES

FLOOR

TURBINE DRAIN FILTER TANK

I TANK

| SOLID

(__ ._________WASTES

HOLDUP CHARCOAL FILTER

SOLID SOLID

WASTES WASTES

Figure V-3.-Boiling Water Reactor Model Schematic.

COLUMN

C-3

0

LIQUID

RELEASES

0

GASEOUS

RELEASES
REACTOR

I FUEL

COOLANT I1 O L

COOLANT

~CLEANUPSYSTEM

SOLID

WASTES

DELAY FREON OR GASEOUS

CRYOGENIC FILTER RELEASES GAS SYSTEMHD45

SOLID

WASTES

SOLID

WASTES

NOTE: CIRCLED NUMBERS REFER TO COLUMN

NUMBERS IN TABLE C-6

Figure V-4.-Liquid Metal Fast Breeder Reactor Model Schematic.
AIR

RELEASES --- GAS

LIQUID

SOLID

SAND FILTER -STACK

LOW-LEVEL

WASTE TREATMENT

HIGH-LEVEL

SOL ID WASTE r -- - - -1

LOW-LEVEL

SOLID WASTE

DISSOLVER

ywi

AQUAFLUOR

PROCESS

WASTE

CALCINER

Figure V-5.-Midwest Fuel Recovery Plant Flow Chart.

OFF-GAS

TRAIN

I

I
- - AIR

SAND FILTER RELEASES - GAS

DISSOLVER STACK- LIQUID

OFF-GAS SYSTEM

Fr CELL OFF-GAS

LOW-LEVEL SYSTEM

LIQUID WASTES

I

I I

LOW-LEVEL HIGH-LEVEL

SOLID WASTES SOLID WASTES

HIGH-LEVEL I

L-- LIQUID WASTES I

DISSOLVER

'PUREX' PROCESS

Figure V-6.-Dual Purpose Reprocessing Plant Flow Chart.
The decontamination factors used in the models were

chosen to provide effluents representative of activity

levels suggested by Federal regulations.( 1 8) The effluents computed for various plant and fuel combinations, at full plant capacities, are listed in Appendix C.

The radionuclide input to each reprocessing plant

model is obtained by multiplying the input fuel load by

the appropriate inventory value. Separate inventories are

available for LWR and LMFBR fuels.

The component parameters used in REPARE were

calibrated to provide effluents representative of activity

levels for existing and proposed commercial reprocessing

plants which would be in accordance with the population exposure limits established by Federal regulations.

The decontamination factors for each component of the

waste treatment systems as modeled by REPARE, as

well as other details of the model, are provided in

Reference (1).

"SOURCE MAP" RELEASE PATTERNS

The source map is the product of all phases of the

model which precede it. The economic model which

selects plant types, the site evaluation which selects

typical sites, and the monthly plant operating schedule

evaluation all feed information to RADREL which

assigns probable monthly release rates for each reactor

and reprocessing plant in the region.

The DOSIS and REPARE models, discussed in the

previous sections, calculate the reactor and reprocessing

plant releases. The RADREL code uses these codes or

the results of calculations with them to assign releases to

each plant in the region.

All LWRs, but neither LMFBRs nor reprocessing

plants, were assumed to have liquid releases. All plants

had gaseous releases, emitted from a 100 meter stack at

each plant site. Each power plant site was allowed up to

a maximum of four plants (or equivalently 4000 MWe of

capacity) at that site; all releases at a site were combined

as if being released from one stack or from one liquid

outfall.

Regional release patterns for two radionuclides, 85 }(r

and 3 H, are shown in Figures V-7 and V-8, respectively.

Total estimated air releases are shown in Figure V-9.

V-9
C

0

0

. o1.0 x 1012

Figure V-7.-85 Kr Release Rates for January 2000 AD.

RADIA'

(PICOC
U

RADIATION RELEASE

TO AIR

(PICOCURIES/SECOND)

1.0 x 1012

O 1.0 x 1010

o 1.0Ox 108

o 1.0 x107

L

0 f

I

" " 

" 

Figure V-8.-3H Release Rates for January 2000 AD.
RADIATION RELJ

TO AIR

(PICOCURIES/SE

O 1.0 x

0 1.0 x

EASECOND)

1010

100

do

FiueV9-otlRdouldeRlaeRtet i o anay200D
VI. AIR TRANSPORT OF

RADIONUCLIDES

Radionuclides released to the atmosphere are transported from source to receptor sites by air currents.

Patterns of transport, and thus of local concentrations,

are highly dependent on local wind velocities and

direction. In the transport process the initial concentrations of radionuclides are modified by turbulent diffusion, deposition processes, and radioactive decay.

The air transport code ARTRAN, a part of the

HERMES model, simulates these transport processes and

estimates the potential resulting concentrations of airborne and surface-deposited radionuclides at specific

receptor sites. ARTRAN was developed as a comparatively simplified model so that simulation of air transport of radionuclides within a geographic region

containing many source and receptor sites could be

accomplished in reasonable computation times. A onedimensional version of the bivariate normal equation was

used, with meteorological variables assessed on a

monthly-average basis, the potential variation in stability

parameters limited, and the behavior of a discrete plume

near the site fenceline ignored. Further simplifications

were introduced elsewhere in the HERMES model in the

representation of county-size areas by a single-point

centroid.

Concentration relationships calculated by ARTRAN

are sufficiently accurate, on a regional basis, for the

requirements of the dose calculations which are also

made on a monthly average basis. Such calculations also

ignore the "close-in" (fenceline) dose. These close-in

doses, which are characteristically somewhat higher than

the average dose to the population, are themselves quite

low; and since it can be expected that very few people

will live near enough to facility fencelines to receive

"close-in" doses, little error is introduced into a regional

study such as this one by omitting their calculation. The

capability of calculating "close-in" doses is being added

to the HERMES model for use where warranted in

further studies. Estimates of "close-in" dose for typical

nuclear facilities in the UMRB are given in Chapter IX.

The ARTRAN model is capable of handling more

complicated calculations, including consideration of a

wider range of meteorological variables. These computations could be carried out only at some sacrifice in

problem size and with materially increased computer

running time. As an example of the problem size

encountered, a typical calculation involved dispersion of

35 airborne nuclides from 185 source sites into 300

receptor centroids for conditions representative of each

of the twelve months. The structure and use of the

ARTRAN code are defined in Reference (1). Those

detailed features pertinent to the UMRB study are

summarized below.

DIFFUSION

The basic equation for diffusion modeling in

ARTRAN is a one-dimensional version of the bivariate

normal equation:

().5 0.01 f

az ii (27rx/n)

exp-1(h2) (1)

Where:

X - mean air concentration (pCi/m3

f - frequency of occurrence of wind direction

(monthly average) for compass sector being

considered (percent)

Q - monthly average rate of radionuclide release

(pCi/sec)

Qz - vertical diffusion variable characterizing plume

distribution

u - mean wind speed (m/sec)

x - distance downwind of source (m)

n - number of compass rose sectors considered

h - stack height at source (m)

In the UMRB study 16 compass sectors were used.

Data on source release rates and stack heights at each

site were provided from other portions of the HERMES

code (in this study a constant stack height of 100 meters

was assumed). Meteorological variables for the study

area were provided on a pre-processed data tape as

interpolated values from actual data at weather stations

within the area. Data derived for atmospheric stability

categories C and D, as defined by Pasquill( 2 2 '

2 3

) were

used. For other uses (where computer time does not

become prohibitive), any or all of the 7 Pasquill stability

categories may be used in the code.

Values of the diffusion parameter az were obtained

by interpolation of the so-called "Pasquill" curves given

in Reference (24), as modified in a review by Waterfield,( 2 5 ) for downwind distances of up to ~30 km. For

distances over 30 km, uz values are estimated by the

VI-1
method suggested by Dickson et al.,( 2 6 ) employing the

equations:

(S-XL)

Qz = 0.465 L + (XL)(.335L) for XL

XL < S 2XL

and

(2)

Qz = 0.8L for S > 2 XL (3)

where

S - is the downwind distance (meters)

L - is a seasonal mixing depth (meters)

XL - is the downwind distance beyond which

vertical mixing is inhibited. (This is also the

downwind point where the vertical distribution of radionuclides is considered to

commence the change from Gaussian to uniform.)

2XL - is the downwind distance at which the

vertical distribution of radionuclide concentration is assumed to be uniform.

In the diffusion/deposition calculations no credit is

taken for plume rise above the stack at any site. This

provides a somewhat conservative estimate of resultant

concentrations; however, Briggs( 2 7 ) shows that the

prime driving force in plume rise is the density

difference or temperature difference between the

effluent air and that of the surroundings. Since the

effluent from a nuclear facility is typically a relatively

cool plume, the density difference is small and the plume

rise similarly small.

PLUME DEPLETION

As the plume moves downwind from source to

receptor, the radionuclide inventory is modified by

radioactive decay and by deposition of radionuclides on

the ground. These depletion processes are described by

the equation:

X = Xoe-t (4)

where the term A is separately defined, for each process,

as (1) the radioactive decay constant, (2) a

dry-deposition coefficient defined as a normalized

velocity of descent of particulate radionuclides, and (3)

a coefficient for washout of materials by rain or snow.

Values for deposition coefficients are experimentallyderived values reported in the literature.(2 8 -

3 1) The

time variable, t, involved in the radioactive decay process

is a function of both the source-to-receptor distance and

the mean wind speed.

RADIONUCLIDE BUILD-UP ON SURFACE

Retention of the longer-lived radionuclides on the

ground surface after deposition is a function of their

individual water soluble characteristics, subsequent

precipitation events, resuspension by wind forces, and

redistribution within the surface layer by farming

activities. In addition, the individual radioactive decay

characteristics of each radionuclide also acts as a removal

process. ARTRAN estimates the build-up of activity on

the surface in monthly increments, for each centroid

location, by "back calculating" the contribution from

each nuclear plant in the years since its startup. A

depletion estimate is made and surface retention

resulting from the particular deposition event is

evaluated as a function of solubility. The loss due to

radioactive decay is then calculated. No estimates are

made in ARTRAN for removal from the surface by

subsequent heavy precipitation events, wind induced

resuspension or redistribution by farm activities.

The loss from each increment of deposition by

radioactive decay is then calculated, yielding the residual

surface burden of radionuclides in the year of interest.

Decay chains involving successive decay steps are

handled using standard techniques.

RESULTS OF UMRB STUDY

Representative data on air and ground concentrations

as calculated for the Upper Mississippi study area are

shown for eight centroids in the area in Figures VI-1

through VI-3.

Figure VI-1 shows, for Cesium 137, the accumulated

ground concentrations resulting from deposition in past

years. Accumulated concentrations of cesium and iodine

are fairly uniform among the centroids except for those

situated near the northeastern and western boundaries

of the study area (and relatively remote from nuclear

installation). Strontium deposits, on the other hand,

appear to reflect more closely the location of a centroid

with respect to reprocessing plant sites.

Monthly-average air and ground concentrations in the

year 2000 for tritium and Cesium-137 are shown in

Figures VI-2 and VI-3. The concentrations indicated are

the calculated total of contributions from all nuclear

facilities in the study area and the surrounding "air

envelope" region.

VI-2
- \

j7 40 20-

0 0

00

CONTOUR INTERVALS

10 PICOCURIES/M 2 0,

Figure VI-1.-Accumulated Concentration Pattern of 137Cs in the Year 2000 Due to Long-Term Buildup.

VI-3
00

C0

oCONTOUR INTERVALS

50 PICOCURIES/M3

Figure VI-2.- 3 H Air Concentration Pattern Falculated for the Month of July 2000 AD.

VI-4
.0

40.

40--

oe 

COTU INERAL

20~~~~~ ~? x 02PCCRISM/OT

Fig ure~ ~~ 0.4-o t l r u d e o i i n R t f 3 C a c l t d t O c r in J l 2 0 D

VI-5

VII. WATERBORNE TRANSPORT OF

RADIONUCLIDES

The water transport subelement of the HERMES

code, described in Reference (1) and called WTRAN, is

designed to calculate concentrations in streams of both

dissolved radionuclides and radionuclides absorbed on

sediments. In addition, concentrations of dissolved

radionuclides are calculated for lakes fed by local runoff,

for shallow ground water recharged locally, and for

artesian ground water recharged from distant sources.

WTRAN logic provides the calculational routines by

which the river water transport of radionuclides is

simulated for each stream in the Upper Mississippi

system and concentrations associated with suspended

sediment, sediment deposition, and dissolved ions are

calculated. For each successive river reach downstream

where a reactor site is located, nuclides are, when

appropriate, added to the stream. If an overland runoff

point or a tributary is encountered, nuclides are added.

If the sedimentation rate is changing, a change in nuclide

load in suspension and deposition or scour is calculated;

and if a point of interest (for example, a municipal water

intake) is encountered, requested information is stored

for later use or printed.

For computational purposes, the rivers were

subdivided into reaches (a "reach" is a portion of a

watercourse of arbitrary length, usually a few miles in

extent, in which the river characteristics are uniform).

Detailed dimensions of a river were modified for

computer use by transformation into equivalent trapezoidal sections where the surface and cross-sectional

areas were conserved. Historical streamflow patterns,

modified by projected year 2000 depletions and regulations, were the basic hydrologic data used in the

study.(3 2 -

4 7

Ground water and local surface water concentrations

were calculated by a different method than river water

concentrations. Monthly runoff and atmospheric

deposition in each centroid provided the input data

required for the calculations. Artesian conditions were

not considered in the computations. Their inclusion

would have required data which are only partially

available (such as regional definition of porosity,

transmissive characteristics and piezometric surface of a

given aquifer). Since ground water velocities are

generally less than 5 miles per year in the this area,

neglecting artesian conditions produce only a very small

error.

For the Upper Mississippi Basin, which extends from

northern Minnesota (see Figure VII-1) to the confluence

of the Mississippi and Ohio Rivers, concentrations of

dissolved radionuclides in a given stream were calculated

by the algorithm:

Cx,t = [Qx-ox,t-At Cx-Ax,t-At e71t

+ 21QiCi] (1

where

Cxt

Qxtx

QiCi

concentration in reach x, at time t

stream flow in reach x, at time t

decay constants (1n 2)/half-life

radionuclide contribution added in reach x

(tributary flow, cooling water from power

plants, etc.)

Similar calculations were performed for major

tributaries. In cases where a tributary stream was large

enough to be considered as a subbasin (Figure VII-2), a

concentration was calculated at its mouth for each

month and stored with its associated flow for input to

the trunk stream. The calculations were made for the

entire Upper Mississippi system, from upstream to

downstream, and were repeated for each nuclide.

In order to account for adsorption of radionuclides

onto stream sediments, the distribution coefficient

concept was used to calculate initial concentrations:

Cs = Qw

QS + --D

(2)

D = I (3) -KD Qs + Q

where

Cs

Cw

I

Qs

Qw

KD

Ki, K2

- concentration on sediments

- concentration in water

- input rate of a particular radionuclide

- Ki(Qw)K2 = sediment flow rate

- stream flow rate

- Distribution coefficient

- constants

VII-1

(1)
OTTER TAIL Co

-WNG

0MlD sTV~s oM EsnZ o

sAFT ;avl Ilos oown TALO ILWLL(KAA

IS.FTI KO bl

ST" .AaO MReUSS-e 'x 0 

cuR E T An Mi na p0 Pi e Au

ssaSTCR00 DUNs 3A e 00 H DEUL to t is ark, LARE''Iato

NAiIM rtt nM luo c a t sT au aua

tata1 comu RFaas ~mn su

A I ~ m*( A s T D TY D O M A oa Eo . C o l s r t r « A I ( T k I N o h s e " W i o n M a"o u

o Ct sou I 

-iAl 'b rpwTN IYCII An s 

AvTlr s teeE on /AG satc 11ton -- D c u a wntasoB

M~tr sonCity p cu OAE IMaT sNE I NA rL- Otom wus AAmEI 1004m Cron VIT NVSCas ~t 

w il"e(" ow C iy D a nprt Lo g 'd i 0 D YiSS30~ h

f01 STN , a C~ r [h( y( r* Ro tswand 0 C . ID. s.SI ste a~as i us cu u ucs oncE NEHE- ME T ts e agE ans lan as n nNsi . an

Ioa ino n Iow Citym D'e n .or IAi SA'" M AN i Lu . r t T t e i c ~ e m n E - a ,n t a e b r t p o i 

Ill s unc a + as"C olive"o r. Isp rCE Rock Isad Q s RA

SpringfKeTd K' c~u S oAM N 

Otturn l1 Li nL " ma Atuso en rTmon -

'mnum .'.. ... .Lus

,IT.s er 

Wrt" . ,it

IIOFI ~aE A a" rEi

Fgure, VC-1-U pe Missip-' e Misour Prnipprtem r ingfes. A

VII-2
The distribution coefficient, KD, is the ratio of the

fraction of ions sorbed per unit weight of sediments to

the fraction of ions remaining in solution per unit

volume of solution. In the present study no provision

was made to vary KD as a function of pH or

temperature, primarily because of insufficient data to

develop the functional constants. Problems of acid water

drainage in association with coal mining are found

largely in Illinois and to a lesser extent in Iowa, Kansas,

and Missouri. It is not an extensive problem in the Basin

except on Pond Creek in the Big Muddy Subbasin where

a low pH of 2.4 has been recorded.

Sediment transport for the area was formulated from

a set of equations that describe the relationship between

suspended sediment load and streamflow. Where

long-term field surveys have been carried out, the

equations provide a good correlation between

streamflow and sediment load rate on a monthly basis.

The main-stem Mississippi, Missouri, Rock, Cedar-Iowa,

Nemaha, Kansas, Osage and Glasconade Rivers were

included in this group. Sediment load equations for the

remaining subbasins shown in Figure VII-2 were based

on the assumption that sediment load is directly

proportional to streamflow (i.e., K2 = 1.00). Estimates

of annual sediment yield for each subbasin by a U.S.

Government Interagency Task Force on Sedimentation(4 0 ,44 ) were utilized in the formulation of the latter

group of equations.

Typical calculated concentrations of selected radionuclides in surface and ground waters, rivers, and river

sediments are summarized for a representative centroid

in Figure VII-3 for the year 2000.

VII-3
OTTER TAIL WING

Xna nNE

,ow t. cun L

$MI. .au I NY uLR AK

" 

NULI - LI D a" , P ulyfir CAR uoo CIM ..

tai~~~~ o 5cutse 

Nae YNNORN , Z U M 8 R 0 ...

AIRY. kocnester Wi coosu

61 ' em -5 ilW csmo ..... V' f .FO TNIUBAL IrOt MY asc -

qs T E & N i t t A o e M

M City Ti 1A10Mahi

naSioux Pa oge terlo sow" Aenso city Rock ord"

NON Bn QR Cedar pids eol 064CLNO

"'Lt PL ATTE (WEST ,n - YT- p/- .o- ", qI1NO ln ~se 

E K 11610 F IneS Iowa'City* Da1 pat Ll[ il

c ruC~r JN NYLE YaT" cass "Ma " w.NONI Moline em P u - 1'' Coun """ SKUNK ,,oh DYIA MecRc kIln NeAME

Li " r? A Thy umw rlingt 'a Galesbur Peoria

""" Nrr W OY FBIUS I O,

KA. N A " .11400 M NGRAN D 5".. x0.1-s OIL NOIS

L E le . 441 N N u. - Quincy " , 1pL Z A N Decatur

n,,o can e Springfield 'TNI

Zooc.so .ues a1In sw

"A N" a.a *o ise w: MY~4 Ye""" MISU Lou .~ Kansa ittL T

IIYIs N .. I c oo EI C , r I AM""y IN.NeIN '"LIMNE l ;KA UDDYI uu- ,OL wn "cullil

FMgurS V0-2-Upe Missip-+e Missour Subba is.r~l

VII-4
Wn O 0 a

N 6X+ 

W W I p 1 11 1

IN I ' IIJ

cc 1 1 11 1

o

I Ii I : * II I I I , I

'L 'i

Figure VII-3.-Typical Water Concentrations Calculated at Centroid 34.

VII-5

Il i i h u h i i i i I t.

II-

. :I I 1 0 II: I II. I 

N

U

n

N

L)

v

rf

ri

a

N

7

C

O

O

a

z

4M

rn

L

Ln

is

L

Lm

to

a

s

u

I m

N

U

a

M

rn

N

7

OC

10

O

a

z

rn

L

N

NY

01

L

N

O

S

M1

N

L)

n

t f

N

U

a

r

o

N

7

K

10

O

a

Z

O

L

N

Cr'L

N

O

01

S

M

N

W

J

6-r

z

W

6-

0 '0 '0

'a+ '0 'r 'C) "m

VIII. RADIATION EXPOSURE

PATHWAYS

Dose evaluations for the UMRB study were performed by the DOSE code of the HERMES model. This

code, described in Reference (1), uses the radionuclide

concentrations calculated by the ARTRAN and WTRAN

codes, and the demographic data for the area provided in

the Living Patterns Model, as a base for its calculations.

The DOSE code evaluates the various pathways leading

to population dose, and calculates radiation dose and

dose commitment to population groups as listed below,

as well as the integrated population dose (in manrems/yr) and 50-year dose commitment for the entire

area. The dose rate and dose commitment values

calculated were:

1. The dose rate (millirem/yr) was the summation of

12 monthly exposure values. The external exposure was due to the concentrations found in the

environment during each month. The internal dose

rate included contributions from each month's

internal deposition for the succeeding 12 months,

even though the deposition may occur late in the

year 2000. This procedure actually overstated the

dose in the year 2000 from materials ingested in

that year; however, for nuclides with effective

half-lives less than 1/2 year, it compensates for the

body burden accumulated prior to the year 2000.

2. The dose commitment was the long-term contribution due to uptake during the year 2000. This

value, termed 50-year dose commitment in this

report, was the dose an individual would be

committed to during the following 50 years due to

the body burden from radioactive material

ingested during the year 2000.

The dose calculations consider the accumulation over

previous years of long-lived nuclides in the environs.

Potential doses are calculated for both external and

internal exposure to the whole body and to certain

selected organs (skin, lungs, GI tract, bone, thyroid,

liver). Dose to the gonads was not calculated. However,

this dose very closely approximates the total body dose,

and the approximation of gonad dose through evaluation

of total body dose is recommended by the Federal

Radiation Council.( 1 5 2)

The following groups were considered in the study:

" Average child (age 4)

" Maximum child

" Average teenager (age 14)

" Maximum teenager

" Average adult (age > 18)

" Maximum adult

The maximum individuals are members of populations living at each population center of interest whose

dietary and living habits tend to maximize the radiation

doses they could receive.

Integrated dose (defined in this report in terms of the

"man-rem") is the sum, over all segments of a given

population, of the products of the number of people in

each segment of a population and the dose received by a

person in that segment. The man-rem concept facilitates

the resolution of average doses and dose distribution

over large segments of the population, and is a useful

tool in the evaluation of sensitivity studies. For purposes

of calculating integrated man-rems for either a population center or for the entire region, the population is

divided into three age categories: 0 to 11, 12 to 17, and

18 years or older. (The teenager age grouping was chosen

based on both noticeably different dietary habits and

activities of this age group.)

The integrated (man-rem) dose for the first category

is calculated as the dose to the average child, multiplied

by the number of persons in the range of 0 to 11 years.

Similarly, the calculation for the second and third

categories employs the dose to the average teenager and

the average adult, respectively.

Doses are calculated for each of several pathways of

exposure (Table VIII-1). Air and water concentrations

are used directly to calculate doses to various body

organs from inhalation of air, submersion in air, consumption of water, and immersion in water. Concentrations in river bottom sediment are used to calculate

expected external exposure from fishing, hunting, picnicking, etc., along the rivershore. Deposition on farmland from air and from irrigation practices is used to

estimate external exposure received while farming or

hunting.

This deposition is also used as input to the FOOD

PATHWAYS MODEL (FPM) subroutine which calculates possible concentrations of radionuclides in animal

feed crops, animal products, and food crops consumed

by humans. Once the concentrations in the several foods

have been determined by the FPM, then the potential

dose to man from consumption of the various foods can

be calculated from factors for annual dose rate and for

fifty-year dose commitment.

FOOD CONCENTRATIONS

Calculations of the radionuclide concentrations in

food take into account the corresponding concentrations

VIII-1
TABLE VIII-1

Exposure Pathways and Foods

AIR SUMMERSION AQUATIC FOODS

TRITIUM TRANSPIRATION 'Sports Fish

AIR INHALATION 'Waterfowl

DRINKING WATER *Fresh Ocean Fish

SWIMMING 'Processed Ocean Fish

BOATING *Shell Fish

EXPOSURE TO SOIL *Secondary Water

EXPOSURE TO SHORELINE SILT

TERRESTRIAL FOODS

Fresh Berries

Processed Berries

Fresh Tree Fruit

Processed Tree Fruit

Fresh Melons

Fresh Potatoes

Fresh Root Vegetables

Processed Root Vegetables

Fresh Green Leafy Vegetables

Processed Green Leafy

Vegetables

Fresh Other Vegetables

Processed Other Vegetables

Grain and Grain Products

Rice and Rice Products

Wheat and Wheat Products

Fresh Milk

Milk Products

Butter

Beef and Lamb

Fresh Pork

Processed Prok

Fresh Whole Eggs

Poultry

Upland Game Birds

*Fresh Citrus Fruit

'Processed Citrus Fruit

*Fresh Tropical Fruit

'Processed Tropical Fruit

'Although available in the model, the pathways designated (')

were not utilized for this region. The foods involved are grown

and processed outside the study area. For purposes of this

study they were assumed to be free of radioactivity (only

radionuclides originating within the study area or its "air

envelope" were considered). "Secondary water" intake was

combined with tap water intake because of the nature of the

data available, as described in the text.

in air, water, and soil (including long-term accumulation). These calculations estimate the concentrations in

food at the time of production or harvest. Some losses

of radionuclides occur during home or commercial

processing, or (by radioactive decay) during the time

between harvest and consumption. These losses are

separately considered in the overall dose calculation.

Several limitations on the accuracy of the estimates

of food concentrations generated by the FPM are

apparent. Current diet and agricultural practices were

modeled. However, these will certainly change in some

difficult-to-define fashion in future years. Another limitation is the paucity of information about radionuclide

behavior in soil, plants, and animals. Most research has

concentrated on the behavior of a few selected radionuclides. Frequently, this lack of data has encouraged

the utilization of concentration factors obtained for

stable elements, regardless of any potential effects from

differences in chemical and physical forms between the

stable and radioactive isotopes. For this study, it was

also necessary to apply one set of median or average

concentration factors across the UMRB region in spite of

the variations in soil and climatic conditions and farming

practices.

A complete development of the general equations

derived for calculation of radionuclide concentrations in

foods and the application of the equations to the

particular foods used for this region is contained in

Reference (1).

CALCULATIONS OF AVERAGE DIETS

Household food requirements for the three population categories considered for the study area-Urban,

Rural non-farm, and Rural-farm-were defined based

on data from Reference (12), as was described in Section

III of this report. These requirements, expressed in units

of pounds of food per week per household, were

converted into quantities consumed per person through

reference to data in U.S. Department of Agriculture

reports.( 1 10,111) When UMRB regional data were not

available, the results of local dietary surveys conducted

at Hanford were taken into consideration.( 1 12,113)

The ICRP( 1 18) has defined the water intake of the

Standard Man as 1.2 liters/day of tap water plus 1.0

liter/day of other fluids such as milk, fruit, etc., plus 0.3

liter/day of water created as an oxidation product of

foods. Dietary studies in the U.S. indicate, however, that

actual water intake values are slightly different from the

ICRP results. Total tap water intake including that

mixed with frozen juices, canned soup, powdered milk,

tea, and coffee has been studied for adults and children

at Hanford.( 1 12,115) Several studies of the intake of

children principally in relation to proposals for fluoridation of water were reviewed by Walker, et al.( 1 7) From

these data, total tap water intake by the average and

maximum persons was defined as listed in Table VIII-2.

Consumption of Waterfowl and Game Birds

Detailed data on populations and harvest of waterfowl are maintained by Federal agencies since such birds

may migrate across several states. Data for duck and

goose kill, number of hunters, and days of hunting for

1965 were selected( 1 19) because this was the year

VIII-2
TABLE VIII-2

Maximum and Average Tap Water Intake

of Adult, Teen and Child(112,115,117,118)

Average Maximum

households, considering the fraction of population residing in each type.

INHALATION RATES

The ICRP( 1 18) has defined the inhalation rate of

Age liters/day kg/mo liters/day kg/mo Standard Man as 20 m3/day. This rate was adopted for

both the average and maximum adult. Data on inhala4 yrs 0.7 21.4 1.4 42.7 tion rate versus age are limited.( 8 1,126,127)

14 yrs 0.7 21.4 1.4 42.7 Values selected for this study were those derived by

Adult 1.0 30.5 2.0 61.0 Rohwer and Kaye( 1 2 8 ) from oxygen consumption,( 12 9

oxygen concentration in ambient air (20.9%), and

oxygen content of exhaled air (130) The data of Rohwer

covered in the survey of diets.( 1 2) The waterfowl data

were reported by state and hence covered a larger area

than the North Central Region defined for the dose

program, therefore, the bird harvest was weighted by the

total population (as of 1960) in the eight states to yield

an average consumption of 0.008 lb/wk/household. (The

fact that some birds may have been given away by the

hunter does not change the average consumption across

a large area.)

UMRB data on upland game bird consumption were

not available; however, data from the Washington State

Department of Game for the period 1960-1964 and

from local diet surveys at Hanford both indicated that

the upland game bird consumption was about twice that

for waterfowl.

This ratio yields an upland game bird consumption of

0.016 lb/wk/household for a total of 0.024 lb/wk. (This

value is proably a little high because the 1965 population, which was higher than that for 1960, should have

been used if available to derive the average consumption

from the total kill.) Data from local surveys at Hanford

indicate that the ratio of maximum to average consumption of gamebirds is about 20 to 1. These data were

prorated on the basis of the limited data in Reference

(12) to each of the three types of households in the

study region.

Consumption of Fresh Water Sports Fish

Several studies of fishing activity and harvest have

been made in the past decade in states that constitute

the North Central Region.( 12 0 -124) Data are available

on number of fishing trips, hours per trip, catch per hour

of various species, and whether the fishing was from the

bank or from a boat or barge. Some of the surveys also

determined the seasonal distribution of the fishing

activity and harvest. Data from these reports yield an

overall average of 4.5 fish/yr/person in the total population. Based on the species most frequently harvestedcatfish and panfish-a reasonable estimate of edible

weight would be 0.5 lb/fish.( 125) These data lead to an

estimate of 0.15 lb/wk/household overall average. Again,

the ratio of maximum to average consumption was

assumed to be 20 to 1.

Data from the North Central Region( 1 2) were used to

prorate this consumption into the three types of

and Kaye for male and female children were averaged to

yield inhalation rates of 14 m3 /day for the teen (14 yr

old) and 7.3 m3 /day for the child (4 yr old). Maximum

and average rates were assumed to be similar.

RECREATION HABITS

Many of the same references used to estimate fish and

game bird consumption also contained data on the

location of various activities, where people who used

certain sectors came from, and the hours spent in various

activities during each season.( 1 19 ,

1 2 0, 1 2 1, 1 2 3

, 13 1) The

data from these reports generally related only to persons

12 years and older. Data from local surveys( 1 16,132)

were used to estimate the age distribution of recreation

hours, and the ratio of maximum to average hours, and

to sharpen estimates of hours for each separate category.

The above procedure estimated the hours of recreation only for the four aquatic categories. The fifth

category, Upland Game Bird Hunting, was not covered

in sufficient detail in available reports but was estimated

from the waterfowl hunting activity on the basis of local

estimates that the total hours per year spent hunting

upland birds were four times those for waterfowl.

Data on where sportsmen traveled to hunt and fish

were taken from recreation reports(119,120,121,123,131)

for the North Central Region. In addition, maps were

used to estimate the location of water that probably

could be used for swimming (considering size of the

body of water and existing pollution).

Upland game bird hunting was assumed to occur in

the centroid of residence except for certain small urban

areas where adjacent centroids were selected. The

location of waterfowl hunting was estimated from the

data on location versus number killed,( 1 2 2, 1 2 3, 13 3) and

hunter travel information.( 1 3 1)

CALCULATION OF POTENTIAL

RADIATION DOSE

Potential radiation doses were calculated by multiplying a series of Dose Factors by the previously generated

information on concentration of radionuclides in air,

water, and foods, and the hours of exposure or

VIII-3
consumption rate. Dose factors were derived for each

pathway-person type-organ-radionuclide combination.

External Exposure

of body size. This follows from an assumption that the

presence of the person does not significantly perturb the

radiation field.

Contaminated Ground

The dose factors for external exposure were derived

on the assumption that the contaminated media were

large enough in dimension to be considered an "infinite

volume" relative to the range of the emitted radiations.

Under this assumption the energy emitted per cc of

media is equivalent to the energy absorbed per cc of

media. All that is required then is to convert MeV per

disintegration to rem and to correct for the differences

in energy absorption between air or water and tissue,

and for physical geometry of each specific exposure

situation.

Air Submersion

The dose from submersion in air is an external dose

to the total body and/or the skin, depending upon the

penetrating power of the radiation emitted from the

airborne radionuclides. Beta radiation was not considered to irradiate internal organs, and only that

energetic enough to penetrate 7 mg/cm2 of tissue

(generally accepted absorption value for the epidermis)

was considered to irradiate the skin. Ratios of surface to

depth dose were estimated for each individual beta

energy (or mono-energetic electron) by the methods

given by Hine and Brownell( 1 3 4) and summed for each

nuclide. Gamma radiation dose at 5 cm depth in tissue

was used for total-body dose and dose to internal organs.

The 2r geometry was used for a person standing on

the ground surrounded by a very large (half-infinite)

hemisphere of contaminated gas for both beta and

gamma radiation. This geometry follows from the

half-infinite volume for gamma. For the beta with

shorter ranges in air, the physical arrangement approaches the infinite volume (4r geometry), but since

the beta is of limited penetrating power, it will irradiate

the skin from only one side, not two sides as with

penetrating gamma radiation. The equation for calculating the dose factor for submersion is given below.

(D.F.)submersion = 8.87 X 10-7 (E + E7 ) (1)

where

D.F. - the dose factor (mrem/hr per pCi/m 3

Ej - the effective beta energy per disintegration

(Mev) at the appropriate depth in tissue.

Ey - the effective gamma energy per disintegration

(Mev) at the appropriate depth in tissue.

The constant takes into consideration the density of

air as well as the conversion from MeV to rem.

In application of this equation, and of the ones for

exposure from water immersion and standing on

contaminated ground, the dose factors are independent

Material deposited from the air or from irrigation

water onto the ground represents a fairly large, nearly

uniform, thin sheet of contamination. Dose factors for

such a situation have been calculated previously for most

important fission products.( 1 35) A standard height of

one meter was selected as the appropriate point at which

to calculate the dose factors. Only gamma radiation is

considered important at this height. Not all of the

contamination will be on the surface of the ground.

There will be some subsurface long-lived contamination

mixed into the plow layer from previous years'

deposition.

If the subsurface contamination were significant, the

calculation of dose would have to consider the source to

be a slab of finite thickness and infinite lateral

dimensions. Such a calculation is given in the literature,

but would require computer solution for each energy

increment.

The conservative assumption was made that all

deposited material was on the surface; however, a factor

of 0.5 was introduced into the equation to account for

ground roughness and any small loss through heavy

clothing.( 1 36,137)

Similar considerations apply to the calculation for a

person standing at the edge of a river on top of river

sediments. To insure that sufficient data are available for

calculation of the dose if the thickness of the deposit

were later found to be significant, the water transport

program specifies not only the nuclide content of the

river sediment but its thickness as well.

The equation for calculation of the Dose Factors for

external exposure from contaminated ground or river

banks is given below.

(D.F.) ound = (0.5)(0.869)

x 2(A)(Ri)(Pi)

i=1

(2)

where

D.F. - the dose factor in units of mrem/hr per

pCi/m2

Ai - Fractional abundance of photon (i) in the

radionuclide under consideration.(138)

R; - Exposure rate at 1 meter above infinite

smooth plane (mR/hr) contaminated to one

pCi/m2 .(139)

Pi - Fraction of surface dose which penetrates

to skin depth (7 X 10- 3 cm) or total-body

depth (5 cm) as the case may be.(140)

VIII-4
0.5 - Ground roughness factor.

(rads in air) rem

0.869 -(R in air)X 1 - (140)

The utilization of a standard height of one meter for

calculation of exposure from contaminated ground may

tend to underestimate slightly the exposure to the child

from this source. A more typical height for calculation

of this exposure to children might be 50 cm. The error

introduced by using a one-meter height, however, is not

significant to the study. For the radionuclides contributing to ground exposure, the calculated external dose

to the child if the 50 cm height were used would be

increased only by about 5 to 10 percent.( 1 39) External

exposure represents only a portion of the total dose to

the individual; however, the net change in dose would be

considerably less. (For the present study, the net

difference is estimated at about 1.0 to 1.5% for total

body dose and most organ doses, and 1.5 to 3% for bone

dose.)

Water Immersion

Exposure from water immersion (swimming) will

depend upon the amount of time spent in the water and

the concentrations of the nuclides present at the

location of the immersion. As for air submersion

exposures, the dose factors do not vary with person size.

The assumption was made that the swimmer was

completely submerged and surrounded on all sides by a

large volume of water, even though he might spend most

of his time near the surface. For gamma radiation, the

physical arrangement leads to a 41r geometry. For beta

radiation, the geometry is approximately 21r as explained

for air submersion, regardless of whether the person is

near the surface or deeply submerged. The following

equation can be used to calculate the dose factor for

each nuclide.

(D.F.)water immersion = 2.13 X 10-6

X (El + 1/2EQ) (3)

where

loss of the radiation in the distance between the water

surface and the center of the body of the person

exposed or the shielding provided by a boat. This is

compensated for by the fact that a water skier spends a

fraction of his time in the water.

Transpiration of Tritium

Although the transpiration of tritiated water through

the skin is known to be rapid, actual rates of transpiration are not available in the literature. Therefore, for this

study the dose from transpiration of tritium was

assumed equal to that from tritium inhalation in

accordance with ICRP recommendations.( )18)

Tritium was assumed to be transferred to most but

not all parts of the body following transpiration. The

total-body dose calculated for tritium transpiration was

therefore applied to all other organs except bone.

Mineral bone is relatively low in water content; thus,

tritium concentrations would be correspondingly low.

The skin, lungs, and GI tract all represent interfaces

where tritium enters the body. The equilibrium concentration of tritiated water in these organs was taken to be

similar to that in the total body.

Inhalation

The inhalation dose was calculated by multiplying the

appropriate air concentration (pCi/m 3 ) times the person's breathing rate (m3 /hr), an occupancy factor

(hrs/month), and the dose factor (mrem/pCi intake).

The occupancy factor was assumed to be 24 hr/day X

30.5 days/month, assuming that.the person spends the

major portion of his time in the centroid where he

resides.

The equations for calculation of internal dose factors

were derived from the equations given by the ICRP(118)

for calculation of the allowable Body Burdens and the

Maximum Permissible Concentrations of radionuclides in

air.

The dose factor varies with nuclide, person type and

organ. A separate factor is required for the annual dose

and the dose commitment.

The dose factor for inhalation can be calculated from

the following set of equations:

D.F. - dose factor (mrem/hr per pCi/liter)

E7 - effective gamma energy per disintegration

(MeV).

E - effective beta energy per disintegration

(MeV).

As before, dose factors were derived for skin and

total-body exposure using effective values of (E7 ) and

(E3), which account for the fractional penetration of the

radiation to the skin depth and the total-body depth.

Water surface exposure is received as a result of

activities such as water skiing and boating. Dose factors

for exposure of this type were taken as one-half those

for water immersion. The factor of one-half ignores the

(D.F)ntal orns interal organs

0.074ETfa

m

X 1 

(D.F .)inhalation,

GI Tract-LLI

-0.693 t

e

0.0256 r' Ef*fa

m

X e-0.693 t'

TR

VIII-5

(4)

(5)
TABLE VIII-3

D.F. - Dose Factor (mrem/pCi inhaled).

e - effective energy of the specific nuclide in the

specific organ under consideration (MeV/dis).

f* - fraction of the materials which escapes absorption in the GI tract ahead of the LLI

(lower large intestine). For insoluble material

f* = (1.0 - f1) where (fl) is defined by the

ICRP(118) as the fraction passing from the

GI tract to the blood. If (fl) was given as 1.0

completely soluble nuclide), then (f*) was

taken as 0.05 rather than zero.(1 41

fa - the fraction of the inhaled material that

reaches the organ under consideration. For

the GI tract, (fa) was taken as 0.5 for any

soluble material and as 0.625 for any insolu-

- ble material.

T - the effective half-life of the nuclide in the

organ under consideration (days).

t - length of time over which the dose is

calculated (days). For the present application, t was either 1 year or 50 years.

TR - radioactive half-life of the nuclide under

consideration (days).

t' - time of travel from mouth to entrance of LLI

(days).

T - travel time through LLI (days).

m - mass of the organ (grams), or mass of

contents of LLI.

Values of the parameters required for these equations

were taken from the ICRP(118) for the adult. Adult

values (with few exceptions) were also used for the teen

and child. Literature data on biological half-life of

cesium in the total body versus age(14 2-145) indicated

that an appropriate value for the 4 year old would be 20

days; for the 14 year old, it would be 30 days. Applying

these to Cs-134 and Cs-137 yields values of T for the

child of 19 and 20 days, respectively; and for the teen,

29 and 30 days, respectively.

Values of (fa), (fw), biological half-life, and thyroid

mass (m) selected from the literature are listed in Table

VIII-3.

Effective decay energies were derived for children's

organs by utilizing the model of the ICRP, whereby the

nuclide is considered to be concentrated in the center of

a spherical organ of effective radius (x).(1'18) The values

of (x) were determined from organ weights, using the

assumptions that (x) was proportional to the cube root

of the organ weight and the proportionality factor was

the same as that for the corresponding adult organ.

Decay schemes chosen were those of Lederer, et al.( 13 8)

Average beta energies were estimated from the graphs of

E/E0 given by Slack and Way.( 149 ) Organ weights

derived from the literature are listed in Table VIII-4.

Appropriate values for absorption coefficients for

muscle and bone were taken from the National Bureau

of Standards Handbook 85.(140)

Values of Parameters Required for

Calculation of Radiation Dose to"

the Thyroid from Radioiodine

Parameter 4 years

0.2(128)

0.5(146)

Biological

half-life,

days

14 years

0.2(128)

0.37(146)

13.(128,146) 30.(128)

m (grams) 5.(81,146-148) 15.(81,146-148)

0.23

0.3

138.

20.

TABLE VIII-4

Organ Mass and Effective Radius

for Child and Teen

Child (4 yr)

Radius

Organ Mass (g) (cm)

Total Body

Thyroid

Bone

GI Tract

Liver

Lungs

16,400

5

1,640

35(a)

530

300

20

2

3

3

7

7

Teen (14 yr)

Radius

Mass (g) (cm)

49,000

15

4,900

100(a)

1,200

580

27

2.7

4

4

9

8

(a)Assumed to be proportional to total body weight.

TABLE VIII-5

Solubilities Selected for

Inhaled Radionuclides(150)

Nuclide

H-3, C-14, Na-22, Na-24,

1-129, 1-131, 1-132, 1-133, 1-135,

CS-134, CS-137

Mn-54, Fe-55, Fe-59,

Co-58, Co-60, Ni-63,

Sr-89, Sr-90, Zr-95, Nb-95

Ru-103, Ru-106, Te-132

Cr-51, Cu-64, Zn-65,

Mo-99, Ba-140, La-140

Ce-141, Ce-144

Solubility

Soluble

Insoluble

Very Insoluble

VIII-6

where
Values of parameters for the GI Tract were assumed

to be proportional to the total body weight. Values of

(t') and (r') calculated on this basis were 0.12 and 0.18

days, respectively, for the child; and 0.36 and 0.50 days,

respectively, for the teen.

It was also necessary to determine the relative

solubility of the various airborne radionuclides in order

to choose proper values of (f1 ) for calculating inhalation

dose factors. Probable chemical and physical forms were

determined for reactor and fuels separation plant releases, and then solubility was determined from data of

the ICRP Task Group on Lung Dynamics.( 1 5 0) The

solubilities selected are tabulated in Table VIII-5.

Ingestion

Equations for calculation of dose factors for ingestion

were derived in a manner similar to those for inhalation

and are given below.

(D.F.)ingestion,

internal organs

0.074 ETfw

m

X 1 - e 0.693t 6)

D.F. -0.0256 Er'f*

(D.F.)ingestion, m

GI-LLI

( TR-0.693t'

Xe T (7) R

where

fw - the fraction of the ingested nuclide reaching

the organ of interest (see Table VIII-3).

The other symbols are the same as defined previously

for Equations (4) and (5). The dose factors for ingestion

are independent of the media ingested, and Equations

(6) and (7) apply for both food and water. All nuclides

in these media were assumed to be in soluble form.

The internal and external dose factors for the various

combinations of nuclide-person-organ are tabulated in

Appendix D.

DOSE SUMMATION

Annual internal and external doses, calculated as

discussed above with the value of (t) set to 365 days,

were summed to estimate the potential total dose in the

year 2000 for each person type at each of the

approximately 300 centroids. Utilization of a value of

(t) of 50 years yields 50-year internal-dose commitment

from material ingested during the year 2000. No

external exposure was included in the dose commitments. Dose commitments for the child were calculated

in three parts, representing child, teen, and adult

parameters, and time span; and then were summed. The

three time periods were 4 through 11, 12 through 17,

and 18 through 54 years. The two time periods used to

calculate the dose commitment for the teen were 14

through 17, and 18 through 64 years. Because the

external pathways are so significant for certain centroids, the anomalous situation is possible where the

dose commitment appears to be less than the annual

dose.

Radiation doses calculated for each of the approximately 300 centroids by the methods described previously were used to calculate potential population doses

in units of man-rems in the year 2000, and man-rems

over the succeeding 50 years (from internal dose only).

The dose (and dose commitment) for the average

child in a given centroid, multiplied by the population of

children (0-1 1 years) estimated for the year 2000 in that

centroid, yields the dose (and dose commitment) to

children. Similarly, calculations for average teens (12-17

years) and average adults (> 18 years) were also made.

The sum of the three age-related doses (and dose

commitments) yields the potential dose (and dose

commitment) in man-rems for that centroid. The total

potential doses and dose commitments for the study

area were obtained by summing the appropriate values

for each of the 300 centroids.

VIII-7

IX. POTENTIAL RADIATION DOSE

BASE CASE RESULTS

The quantitative evaluation of the potential radiation

dose to people in the study region in the year 2000

requires the summation of contributions from all nuclides through all exposure pathways from all nuclear

facilities. Annual dose rate and fifty-year dose commitment to the total body and to six specific organs were

calculated for average and maximum individuals in three

age groups at each of the 300 centroids. Detailed results

show the contributions to dose from each of the 45

isotopes and 10 pathways considered. Summary information provides individual and integrated (dose times

population) total-body dose data for each centroid. The

expected integrated total-body dose rate to the region in

the year 2000 is estimated to be 4940 man-rems/year.

Since the projected total population of the region in that

year is 29.1 million, the average total-body dose rate

would be about 0.17 millirem per year per person, while

the median (50 percentile) dose is 0.13 millirem per year

per person. These results are very similar to those

presented in a 1972 United Nations report entitled

"Ionizing Radiation: Levels and Effects" which estimates the average worldwide total-body dose from

nuclear facilities utilized for electric power production

in the year 2000 at about 0.2 millirem per year.

Similarly, the regional 50-year dose commitment to the

total body from the intake of radionuclides in the year

2000 is estimated to be 4430 man-rems, which yields an

average of about 0.15 millirem per person. The average

total-body dose in the region from natural background

radiation, medical and dental sources, and all other

sources not associated with the nuclear fuel cycle is

about 236 millirems/yr; in the year 2000 the integrated

total-body dose rate due to these sources would be

about 6,800,000 man-rems/year. Thus the supplementary dose rate from the assumed nuclear facility

operation in the region is approximately one-tenth of

one percent of the overall dose rate.

MODEL RESTRAINTS

The HERMES model estimates the potential radiation

dose to the population of the study region from

conditions assumed for the year 2000. Limitations to

the absolute accuracy of the estimates have been

discussed in detail in the HERMES descriptive document,O1) but a review of some limitations will clarify the

detailed discussion of the dose evaluation to follow.

" The model provides essentially a steady-state

calculation of radioactive effluents. Both the

release rates of radionuclides and their eventual

concentrations are calculated as monthly averages

(in the case of reprocessing plants, the monthly

average effluent rates are based on annual average

release rates and are thus constant over a one-year

span).

" Doses to large groups of individuals were calculated according to the conditions at a specific

geographic point, each point representing a population centroid.

" Release rates for presently designed light water

reactors were computed with parameters based

primarily on Safety Analysis Reports. However,

advanced effluent treatment systems were simulated to provide releases consistent with the

proposed Federal guidelines.( 1 5 1 ) Potential uncertainties in release rates arise from the potential

differences between SAR release estimates and

releases in practice, advanced treatment systems

performance uncertainties, and the behavior of

fission products in the LMFBR sodium coolant

and cleanup systems.

" Radiation dose contribution from 45 radionuclides

were considered. The list of radionuclides

(Appendix C) includes fission and activation

products, but not transuranium nuclides. The capability to evaluate the radiation dose from plutonium and other transuranium nuclides has recently

been added to the HERMES model, and preliminary results have been obtained. These results

indicate that these nuclides would add slightly

(about 3%) to dose to the lungs, through the air

inhalation pathway; dose contributions to other

organs would be insignificant. However, additional

supporting data are needed on behavior of transuranium nuclides within the body.

" Living habits for the study area were characterized

by the Living Patterns Model. This model defined

the sources of foodstuffs and the sites for recreational activities for the population of each centroid. If a particular food source or recreation site

lay outside the study area, the associated radionuclide concentrations were assumed to be zero.

Thus, only radioactivity from within the study

area was considered to contribute to dose through

food pathways.

IX-1
" Biological behavior and radiological characteristics

as well as other radionuclide characteristics

throughout the model were based on existing but

often limited data. In many cases extension and

extrapolation of data from chemically similar

radioactive and/or stable isotopes were necessary to

determine the parameters needed to complete the

model. Details of the assumptions and values used

are given in the HERMES document.( 1

" Tritium transpiration has emerged as a major

contributing pathway to total body dose. The dose

attributed to this pathway was estimated as being

equal to the tritium dose from inhalation. Rohwer

and Kaye(1 28) indicate that the doses from

tritium transpiration and tritium inhalation are

approximately equal, and the assumption of

equality is recommended by ICRP.(1 18) However,

some data(15 8) indicate that the ICRP recommendations may overestimate the tritium transpiration dose by as much as a factor of two.

" Initial formulation of the model assumed that the

air submersion dose to people close to the radioactive emission source (close-in dose) was not a

major contributor to the regional integrated dose.

The assumption is manifested in both the centroid

population representation and in the atmospheric

transport model.

Additional calculations were made to evaluate the

added contribution of close-in dose using release rates as

discussed in Chapter V. These contributions were calculated for each site in the study area, using the

meteorology for each site and assuming that the distribution of population near the site is proportional to

distance from the boundary (actual nearsite population

distributions were not available). Similar calculations

were performed for each of the reprocessing plants in

the study area.

From these calculations, it was estimated that an

individual living one mile from a 1000 MWe BWR would

receive an additional dose contribution of 0.15 mrem/yr

from the close-in dose. Somewhat lower supplemental

doses were calculated for other reactor types because of

the reduced release of 133Xe from these plants. This

reduction results from a longer delay time for gases

released from the PWR and from the noble gas treatment

system assumed for the LMFBR. The supplemental dose

rate at one mile from either of these facilities was

estimated at 0.02 mrem/yr. For reprocessing plant sites,

the dose rate at a distance of one mile was estimated at

about 7 mrems/yr. Tritium was indicated to be the major

contribution to dose in areas adjacent to these plants.

Doses of this magnitude, of course, would be received

only by those few individuals who might live near the

plant boundaries.

The contribution of the close-in dose to the regional

integrated population dose was estimated in the same

series of calculations. The total contribution to the

regional total body dose was estimated to be about 240

man-rems/yr, or 5% of the integrated dose as calculated

with the regional HERMES model. Only a small fraction

of the population lives near plants, although those living

at the fenceline of reprocessing plants were estimated to

receive up to 7-10 mrems/yr. This estimate of close-in

integrated dose is only an approximation, but, it is small

enough to indicate that omission of the close-in dose in

the HERMES model does not significantly affect the

estimate of the total population dose.

GENERAL TRENDS AND PATTERNS

The dose rates calculated throughout the region had a

range of 3 orders of magnitude. Average adult total body

dose rates for the various centroids, for example, ranged

from 1.1 X 10- to 1.2 mrems/yr. The average totalbody dose distribution in the total population of the

region (Figure IX-1) shows that 99% of the people could

be expected to receive a total-body dose less than 0.50

mrem/yr. The total-body dose for adults is slightly

higher than the average individual indicated; that for

children and teens is somewhat lower.

Radiation doses were also calculated for six specific

organs-GI tract, thyroid, bone, lungs, skin and liver.

The average radiation doses for the region for all organs

and for total body are tabulated for both annual dose

rate and 50-year dose commitment for children, teens

and adults in Table IX-1. Note that the relative

magnitude of the doses calculated for the three age

groups considered is not constant for different organs.

Distribution of the annual dose rate by organ for average

age groups is shown in Appendix D. Since external

pathways contribute the same exposure to individuals in

each age group, the variations in dose among the age

groups are affected only by internal pathways. The

internal pathways are in turn affected by age-dependent

diet patterns and organ masses. Radiation dose distribution without age dependence would be represented by

that of the skin dose, which results from tritium

transpiration and external exposure only, i.e., exposure

to soil and air submersion.

As is indicated in Table IX-1, doses to individual

organs may be either greater or smaller than to the total

body. These variations are dependent on the specific

radionuclides contributing to dose, and on such factors

as radionuclide concentration in specific organs, differences in size of the various organs, the extent to

which internal organs are shielded from external radiation, and on the relative effectiveness of absorption of

radiation within the tissues of an organ.

These differences can be illustrated by consideration

of the thyroid, which has a high affinity for iodine. The

decay of, say, one picocurie of radioiodine within the

thyroid contributes a much greater dose to that organ

than would be contributed to the total body if the same

decay energy were to be absorbed uniformly by the

body. On the other hand, the thyroid is much less

affected by external radiation from 85 Kr than is the

skin.

IX-2
1.5 

- 1.2 - - MAXIMUM* (AVERAGE ADULT) - -

1 .035 - - MAXIMUM* (AVERAGE PERSON) - -

*MAXIMUM: HIGHEST AVERAGE DOSE CALCULATED FOR

ANY OF THE 300 CENTROIDS CONSIDERED

TI.

C

C

0

> o0

o.

0

S

0

D

0.5 -0.50 --

- 0.17

- 0.13

40 50 60 70

PERCENT OF REGIONAL POPULATION (CUMULATIVE)

1.0

C)

0

0 

WU

'-LU

Wz ------ - 99 PERCENTILE -

AVERAGE --- - --

MEDIAN 

I I I 1 1 1 I

0

10 20 30 80 90 100
TABLE IX-1

Average Annual Doses and 50-Year Dose

Commitments Received from Radionuclide

Releases by Population of Study Area

Average Individual

Average of

[Maximum Population Group] 

Organ Child Teen Adult Child Teen Adult

ANNUAL DOSE RATE, mrem/yr

0.122

0.087

1.038

0.044

0.119

0.133

0.146

0.119

0.104

0.601

0.031

0.116

0.133

0.133

0.199 0.132 0.125

0.159 -

0.889 1.680 0.797

0.030 0.065 0.040

0.191 -

0.208 -

0.206 0.175 0.150

0.211

1.090

0.037

0.223

50-YEAR DOSE COMMITMENT, mrem

0.106

0.104

0.940

0.039

0.229

0.097

0.128

0.101

0.109

0.442

0.021

0.227

0.095

0.115

0.181

0.190

0.637

0.017

0.285

0.170

0.187

0.117 0

1.580

0.073

0.

0.

0.158 0.

.109 0.195

638 0.893

039 0.030

132 0.205

*The maximum population group consists of those individuals in

each centroid whose dietary habits and patterns of work and

recreation tend to maximize their exposure to radionuciides in

the environment. This group should not be confused with the

few individuals living near the boundaries of nuclear plant sites.

Radiation dose limits have been promulgated by

various public councils to provide guidelines for the

protection of the nuclear industry work force and the

public at large. Proposed AEC design guidelines( 15 1) are

selected to provide reasonable assurance that radioactive

effluents from light water power reactors would contribute less than 10 mrems/yr to people living at the site

boundary and less than 1 mrem/yr to the public at large.

The Federal Radiation Council regulations( 1 5 2) limit

the whole body exposure to 500 and 170 mrems/yr for

maximum and average individuals, respectively. The FRC

recommends limits of 1500 and 500 mrems/yr to the

bone and thyroid for maximum and average individuals,

respectively. The difference in the recommended limits

for total body and other organs acknowledges the

varying radiosensitivity of specific organs.

The doses listed in Table IX-1 are those estimated to

result from the release of radionuclides from nuclear

facilities (power plants and reprocessing plants) which

are assumed to be operating in the year 2000. The

computed doses from these assumed multiple sources are

about 0.1% of background. The ratios of the doses for

specific organs to total-body dose rate calculated for this

region are similar to the ratios of the recommended

limits for specific organs to total-body dose rates in the

FRC regulations.

By considering the rate of nuclear energy buildup

before the year 2000, one can estimate the dose received

during that year from body burdens accumulated during

previous years. Such an analysis can provide a ratio (the

accumulation ratio) between the dose rate which

includes the contribution of body burden from previous

years to the dose rate calculated for year 2000 intake

alone.

Values of this ratio for total body varied from 1.0 to

10; 3

H with its short effective half life exhibited the

value of 1.0 and 9 0Sr the value of 10. Cesium had a

value of 1.07 and all radionuclides found to make

substantial dose rate contributions had values closer to

1.

Due to the nature of the calculation made for internal

dose (the summation of the dose over the succeeding

year due to each month's internal deposition) the dose

rate calculated for year 2000 environmental concentrations is essentially the same as if one considered previous

body burdens for all organs except bone. Since about 2%

(0.0007 mrem/yr) of the bone dose rate was calculated

to be from 9 0 Sr, and the accumulation ratio for 9 0

Sr is

11 in bone, the total dose from 9 0Sr including the

previously accumulated burden would be 0.0075

mrem/yr, or about 16% of the total corrected dose.

Similarly, the bone dose ratio from cesium (accumulation ratio of 1.16) would increase from 0.015 to

0.018 mrem/yr. The corrected bone dose rate, considering contributions from previously accumulated

body burden, would thus be about .044 mrem/yr or

about 30% higher than the rate calculated for year 2000

intake alone.

Annaul total-body dose rates and 50-year dose

commitments, as calculated for the study area, display

similar regional distributions. The distribution of potential annual total-body dose rate for average adults

(Figure IX-2) shows the general pattern of the dose

variation. Superposition of radionuclide source strengths

for airborne releases (Figure IX-3) illustrates the correlation of source distribution and dose distribution.

Regional dose distributions calculated for average adult

GI tract, thyroid, lungs, liver and skin, as shown in

Appendix D, display patterns similar to those for the

total-body dose. However, the distribution patterns for

bone dose for the average adult (Figure IX-4) appear to

be more related to the location of waterways than to

facility location.

Dose to the total body and the liver result primarily

from tritium intake through inhalation and transpiration. Air inhalation also plays a major role in the dose to

GI tract, thyroid and lungs. Tritium contributes heavily

to the air inhalation and transpiration dose for lungs, but

relatively less so for the GI tract and thyroid. The

thyroid gland receives a significant portion of its total

IX-4

Total Body

GI Tract

Thyroid

Bone

Lungs

Skin

Liver

Total Body

GI Tract

Thyroid

Bone

Lungs

Skin

Liver
193

1 2 161 133

2212

19 94 134

128 96 163 54

262 228

165 13613 20

284 , 229 197 16 39

2939 6 138 95 96 55

:2 4 65 23"! 1 "171 113 139 98 23 22

a65 23 i01 14 40 100 10 7 56

24 285 266 265 202 76 4 4 2

,295 1 6 1 1 1 2 1 2 0158 2

261 r7 59 25 2

204 13 103 104 60

286 268 235 234 20718 61 26 3

269 236 152 144 106 105 62 7

287 _ 63 28 7

1 270 X 45 109 108 6'2 0

296

288 238 208 8017 148 10 31

289 273 239 24011m ?09 182 49 11 1 61 33 3

9 2991 24 a. 243 241 211 21 18 50 4?0 3

296 299 2724 247 246 213 24515 16173

3011 276 29248 15 15.3 152 11 73 1 3 3 1

291 78 xrn251x 75 74 39 1

2 27 ,2 2i -19 120 16 40 3

... 7 X92 253 15 122 7942

254 123 $ 43

2 80 255 256 21gZ0 1 9 ,-, 2- 4 1 88 1s 04 -31

759 18q 26 121 85 86

" 25 24 ,19 89 - 53 16

1. 2xl00 to 10 19134

mrem/yr "0 191 43

1 5 2in Iq 5 tn 132 92

10~ to 10 2 r _9 8 1

mrem/yr

010-2* to 10-3

Qmrem/yr

Figure IX-2.-Regional Pattern of Total-Body Annual Dose Rate (Average Adult).

IX-5
0

0

0

00

" O

0 0

" O

oo

00 0

00

00

(10 0

o o 

F 0

o 0

0 00

0

Figure IX-3.-Relative Radionuclide Source Distribution, Total Airborne Release Rates.

IX-6
193

227162 161 133

195 194 134

262 228 196 16135 54

165 136 20

284 263 229 197 164 166 137 9 94 2

-293 231 230 198 Ilo e 6 3 95 9 55 2

24 265 232 207171 173 139 98 23 22

265 23320 17 14 405 201175174 140 100 101 57 5

295 25 26 25 22 176 141 142 102 58 267 24 177 59 25 24 2

286 268 235 234 203 178 143 10 0 0 61 26 3

269 236 237 205 206 79 144 106 105 62 6 82

27 270 207 145109 lob 64 6529 30 6 5 296 6

288 238 208 180 147 148 146 110 107 31 36

7

289 273 272 239 240 212 209 182 183 181 149 112 68 33 32 8

297 290 244 242 243 241 211 210 184 150 11 3 0 4 1

1 5 72 JO 35 1

298 299 274 247 246 213 245 151 116 117

300 275 276 29248 185 153 152 118 73 7138 37 11

291 27278 250 251 214 19 2075 74 39 12

2279 252 215 116 773 420 

292 253 186 154 22 121 78 412 14

5 5 18 8 123 gl8 79 4

2802525621921 220 21 8 2 44

282 283 257 223 222 188 155 124 125 84 83 15

-260 259 224 190 189 1 2562 8 89 r8s 53 16

2x10- to 1X101 "5'21129 128 99088 45 17

mrem/yr 26 91 4 9 46 47 18

x10-1to 1210- 226 192 19 131 132 92

o11- 9x0 48 19

F mrem/yr 5 /y ~160 sls50

l x10-2 to l x10-4

mrem/yr

Figure IX-4.-Regional Pattern of Bone Annual Dose Rate (Average Adult).

IX-7
dose from 1311. Exposure to the soil is a contributing

pathway to some extent for all the organs, but most

significantly for skin and bone. The principal contributors to the dose from exposure to soil were 1311, 134Cs,

9 5Nb and 9 5Zr. Similarly, air submersion was a contributing pathway for all organ doses, but primarily for skin

and bone.

RADIONUCLIDE CONTRIBUTIONS

TO POTENTIAL DOSE

Forty-five fission product and activation product

radionuclides were considered in this study. These were

selected for study based on the probability of their

occurrence in nuclear facility releases and on their

potential contribution to human radiation dose via at

least one of the exposure pathways considered. Results

of the study, however, indicate that only a few of the

nuclides considered are likely to contribute significantly

to dose. A summary of data on those nuclides is given in

Table IX-2.

Tritium was indicated to be the largest contributor to

dose. It accounts for about 85% of the total-body dose

to the average adult-82% if all age groups are considered-and for a preponderance of dose to all organs

except bone and thyroid as well. It also contributes

heavily to the estimated 50-year dose commitment

(about 93 percent of total-body dose commitment).

Dose contributions from noble gases are primarily

those from 8 5Kr and 133 3 Xe. These gases contribute

mainly to skin dose- 8 5 Kr accounts for about 6.3% of

dose to skin, and 133Xe about 4%. Xenon also contributes to a small extent to the dose to other organs and

to total-body dose (about 1% of total-body dose, 9% of

bone dose).

The long term buildup of 8 5 Kr in the atmosphere

would also contribute to dose rate in the year 2000.

Two simplified calculations were made to estimate the

effect of this atmospheric buildup on dose. The first

calculation considered only releases of 8 5 Kr from

nuclear facilities within the study area, but assumed that

all the released gas remained in the air space above the

area-a very conservative assumption. This calculation

indicated a contribution to the average individual's

total-body dose rate of about 0.007 mrem/yr and a skin

dose rate contribution of about 0.4 mrem/yr. For

comparison, the 8 5 Kr contributions for the year 2000 as

calculated by the HERMES model averaged 2 X 10-4

mrem/yr to the total body, and 0.01 mrem/yr to the

skin.

In the second calculation, worldwide releases of 8 5 Kr

were considered, and worldwide dispersion in the atmosphere was assumed. This calculation assuming the

same ground rules for bottling of noble gases as were

used in the study, indicated average dose rate contributions of 5 X 10-4 mrem/yr to the total body and 0.04

mrem/yr to the skin. If no bottling of noble gases were

assumed, these dose rate contributions would be 0.04

mrem/yr to the total body and about 3 mrems/yr to the

skin.

The nuclides 9 0Sr, 9 5 Zr, 9 5Nb, and 1 0 6 Ru-often

considered as "problem" radionuclides-were indicated

in this study to contribute only to a small extent to

human dose. The aggregate contribution of the four

nuclides included 15% of thyroid dose, 2.8% of dose to

the GI tract, and about 1% of total-body dose. However,

9 0

Sr, which accounted for 2% of the calculated annual

dose rate to bone for the year 2000 (16% if prior internal

deposition were included), did contribute about 32% of

the 50-year dose commitment to bone.

Iodine isotopes are potentially important contributors to dose because of the strong affinity of iodine

for the thyroid. Five isotopes of iodine (129, 131, 132,

133, and 135) were considered in this study. Of these,

only 1311 was indicated to contribute significantly to

dose, although the other isotopes did make "trace"

contributions to thyroid dose. Inaddition to its heavy

contribution to thyroid dose, 1311 was calculated to

contribute substantially to dose to the GI tract, bone,

lungs, and skin.

The two cesium isotopes studied- 1 34 Cs and

1 3 7 Cs-contributed to some extent to all organ doses,

but most heavily to the bone dose. Together, these

isotopes accounted for about 42% of annual dose rate to

bone, as well as 60% of the 50-year dose commitment to

bone.

The isotopic contributions to dose were observed to

differ among the three age groups considered. In most

cases these differences were minor. For the thyroid,

however, significant differences in isotopic contributions

to dose were calculated for the three age groups. These

differences are indicated in Figure IX-5.

Radionuclide contributions to dose are discussed in

greater detail in Appendix D.

PATHWAY COMPOSITION OF THE

POTENTIAL REGIONAL DOSE

Ten pathways of potential radiation exposure are

included in the dose evaluation model. The results of the

study indicate that under the conditions assumed for the

year 2000 and the limitations of the model as previously

discussed, only a few of the pathways contribute

significantly to the radiation dose. The pathways or

foods which were found to contribute to specific organ

doses are listed in Table IX-3. Dose commitment is

assumed to result only from internal pathways; thus, air

submersion and exposure to soil contribute only to the

annual radiation dose rate.

The greatest part of the potential annual dose rate is

contributed by pathways associated with airborne or

air-deposited radionuclides. Four pathways-air submersion, air inhalation, tritium transpiration, and exposure

to soil-contribute, as an average for the region, some

91% of the total-body dose. They contribute heavily to

specific organ doses as well, average contributions

IX-8
TABLE IX-2

Summary of Radionuclides Found to Contribute to Dose

3 85 90 95 9 5 106 129 131 133 134 137 144 H Kr 5r Zr Mb Ru I I Xe Cs Cs C

RAUFLIFE 12 YEARS 1076 YEARS 65 AYS 35 DAYS RS HALFLIFE HUO I 6X107YEARS .05 DAYS S.OTDAYS 2.IYEARS YEARS 28.5 DA

TYPE O MATERIAL FOUND AS GAS, TRITIATED WATER, NOBLE GAS ALKALINE EARTH METAL SIMILAR RELATIVELY INSOLUBLE MTAL METAL. SIMILAR TO ZIRCONIUM OF TAE PLATINUM METALS REAIVEY TNE TYPE OP MATERIAL SOLUBLE HALOGEN POWERFUL SOLUBLE HALEGD. POWERFUL NOBLE GAS VERY REACTIVE ALKAL METAL. MOST VERY REACTIVE ALKALI METAL ST OAL LOARTH

OR SODIUM HYDRIDE TO CALCIUM INSOLUBLE, HAS VOLATILE OXIDE FORM OXIDIZING AGENT OXIDIZING AGENT SALTS WATER SOLUBLE SALTS WATER SOLUBLE INSOLUBLE OXDE

LWR AFFECTED BY LWR TREATMENT SYSTEMS DUE TO LONG HALF-LIFE RELATIVELY REMOVED BY LWR TREATMENT SYSTEMS REMOVED BY LWR TREATMENT SYSTEMS REMOVEDBY LWR TREATMENT SYSTEMS REMOVED RY LWR TREATMENT SYSTEMS LR ASSUMED TO BE REMOVED BY LWR REMOVEDBY LWR DEMINERALIZERS LWR DELAY SYSTEMS REDUCE EFFECTIVELY REMOVEDBY LWR EFFECTIVELY REM VED BY LWR EFFECTIVELY REMOVE OB.Y BY THEIR ABILITY TO LIMIT ANAFFECTEDBY LWR TREATMENT IN ACTIVE COMPONENTS FILLER IN ACTIVE COMPONENTS FILTER IN ACTIVE COMPONENTS FILTER INACTIVE COMPONENTS FILTER DEMINERALIZERS AND FILTERS. AND FILTERS. RELEASE ALSO RELEASES SOMEWHAT. BWR CHARCOAL DEMINERALIZERS DEMINERALIZERS DMINERALIZRS LIQUID DISCHARGES AND RECYCLE SYSTEMS DEPENDENT ON DELAY FOR REMOVES INSOLUBLE PORTION, REMOVES INSOLUBLE PORTION. REMOVES INSOLUBLE PORTION, REMOVES INSOLUBLE PORTION DAY HAS NO EFFECT REDUCED RY DELAY DELAY HOLS UP FORT DAYS. PWb TRITIUM TO PRIMARY COOLANT DECAY DEMINERALZEF FOR SOLUBLE DEMINERAAIZER FOR SOLUBLE DEMINERALIZER FOR SOLUBLE DEMINERALIZER FOR SOLUBLE TANKS PROVIDE TO DAY HOLDUP PORTION PORTION PORTION PORTION

LMR REMOVAL BY COLD TRAP IN LLMBR ASSUMED TO BE REMOVED WITH A REMOVAL BY COLD TRAP COOLANT REMOVAL BY COLD TRAP COOLANT REMOVAL BY COLD TRAP COOLANT REMOVAL BY COLD TRAP COOLANT LMFBR ASSUMED TO REMOVED FROM ASSUMED TO BE REMOVED FROM ASSUMED TO BE REMOVED WITH A DF ASSUMED TO BE REMOVED FROM ASSUMED TO BE REMOVED FROM ASSUMED TO BE REM COOLANT SYSTEM IS ASSUMED WITH F OF D FROM LMEBR AND TREATMENT SYSTEM IS ASSUMED TREATMENT SYSTEM IS ASSUMED TREATMENT SYSTEM IS ASSUMED TREATMENT SYSTEM IS ASSUMED Na COOLANT BY COLD TRAP WITH Na COOLANT BY COLD TRAP WITH OF 3 FROM LUMFBA AND REPROCESSING LMR COOLANT BY COLD TRAP LABR COOLANT BY COLD TRAP LMBR COOLANT BY OF 10REPROCESSING PLANT EFFLUENT WITH A OFO 0 WIH A OFOF A A ID WIH A OF OF 10WITH A OFOFA O DF OFI10 A OF 10 PLANT EFFLUENT STREAMS BY TREATMENT SYSTEM WITH A F TREATMENT SYSTEM WITH A 0 TRETMENT SYSTEM'

STREAMS RY REFRIGERANT REFRIGERANT ABSORPTION SYSTEM OF F0DIDOF 10 ABSORPTION SYSTEM OR EVEIVALENT OR EUIVALENT PRODUCTION PRODUCED WITHIN FUEL AS A PRODUCED WITHIN FUEL AS A FISSION PRODUCT, RELEASE AT MUCH ACTIVATION OF FUEL CLAD IN LWR'S DAUGHTER OF Zr95. REASED FISSION PRODUCT DIFFUSES FROM PRODUCTION FISSION PRODUCT. DIFFUSES FISSION PRODUCT. DIFFUSES FISSION PRODUCT AND AS DAUGHTER FISSION PRODUCT. DIFFUSES FROM FISSION PRODUCTS DIFFUSES FROM FISSION PRODUCT. D FISSION PRODUCT. ASSUMED TO FISSION PRODUCT. DIFFUSES FROM SLOWER RATE THAN A GASEOUS AND CORROSION HOWEVER SIMILAR TO Z FUEL A REDUCED RATE RELATIVE TO AT A REDUCED RATE RELATIVE TO AT A REDUCED RATE REATIVE TO OF 1331, DIFFUSES FROM FUEL AND FUEL AT A REDUCED RATE RELATIVE FUEL AT A REDUCED RATE RELATIVE FUEL AT A REDUCED DIFFUSE LIKE A NOBLE GAS FROM LEAFING FUEL AND FROM COOLANT FISSION PRODUCT WOULD BE CORROSION RATE IS LOW. ALSO NORLE GASES NOLEGASES NOBLE GASES IS AFFECTED ONLY BY DELAY IN TO GASEOUS F MISSION PRODUCTS TO GASEOUS FISSION PRODUCTS TVGCASEOUS FISSION LEAKING FUEL PRODUCED AS AN TO GAS TREATMENT SYSTEM PRODUCED AS AFI SSION PRODUCT LWR TREATMENT ACTIVATION PRODUCT IN PWR'S UNAFFECTED BY ANY CLEANUP SY STEMS WITHIN FUEL DIFFUSING AT MCH

FROM SOLUBLE BORON CONTROL . RELEASED AT SAME RATE AS SLOWER RATE THAN GASEOUS FISSION

RELEASED FROM FUEL IN REACTORS PRODUCTS

REPROCESSING PLANT 3H FOUND IN FUE ENTERING K, NOT RELEASED FROM REACTOR EFFECTIVELY REMOVED BY LIQUID EFFECTIVEY REMOVED BY LIEUID EFFFCTIVELY REMOVED BY LIQUID VOLATILE OXIDE REMOVED BY FILTERS. REPROCESSING PLAN T MODERATE F IN DISSOLVER MODERATE 0 ATTAINED IN SILVER DECAYED CONSIDERABLY IN EFFECTIVE REMNVEDRY ION EFFECTIVE RENRVEDRY ION EFFECTISELY REMOVE REPROCESSING PLAN ISRELLASED LEAKING FUR IS RELEASED TO TREATMENT SYSTEM ION EXCHANGER TREATMENT SYSTEM ION EXCHANGER TREATMENT SYSTEM ION EXCHANGER SOLUBLE FORM REMOVED BY ION OFF GAS SILVER REACTOR REACTOR CONSIDERABLY REDUCED REPROCESSING CVELING PERIOD EUCUANGE IN TREATMENT SYSTEM EXCHANGE IN TREATMENT SYSTEM EXCHANQ INTALUTI AS A GAS FROM THE REPROCESSING REPROCESSING PLANT TREATMENT EXCHANGER BY DECAY AND ADDITIONAL DF OF 3 FROM PLANT SY STEM, WHICH PROVIDES A HOBO GAS TREATMENT SYSTEM

DE O30 PLANT SYSTEM DECONTAMINATION PLANT SYSTEM DECONTAMINATION FACTORS FACTORS RELEASE TO COOLANT RELEASE TO COOLANT RELEASE TO SURROUNDINGS RELEASE TO SO UNOINGSA

LWR 1.0-1.01 1.00-1.01 .EIE-I. 5X0D 5 

4. -I.XI1 o.AO- I. X10I. X 108

-U. 4x106 L R.41 9

3

-3.E31D 1.xT1O-.9I2.52-U. 10103 1.901 -2.TX1O 1.IR -2. A0El SVI - 2.

LMER I.DIO6 3D.DI02 4.010 5.5X10 4.IDI6. 2X 10

T LMFBR 5.OI N.A10 6.,10 5 

I.DOISI.0x10o

REPROCESSING PLANT 1.01 I. 1X10 .5X10 9.0X106 A T9X16E.6XIPOCESSING PLANT .0 . 1 08

INU O LN INPUT TV PLANT U.AIlRYI~H II~EODROI .05 INPUT (RE I 1IA U~N DIN GS ) R ELEA SE TO SURR OU NDING S

AIR TRANSPORT CHARACTERISTICS FOUND AS TRITIATED WATER VAPOR A NOBLE GAS FOUND AS A SMALL PARTICULATE FOUND AS A SMALL PARTICULATE FOUND AS A SMALL PARTICULATE FOUND AS A SMALL PARTICULATE AR TRANSPORT CHARACTERISTICS FOUND AS A SMALL PARTICULATE, AND NOBLE GAS FOUND ASA SMALL PARTICULATE FOUND AS A SMALL PARTICULATE FOUND A A SMALL

WASHOUT IS CALCULATED AS A NO WASHOUT DIFFUSES AS METHYL IODIDE. A VPOR NO WASHOUT DIFF USES AS A GAS DUE TO DIFFUSES AS A GAS DUE TO DIPPUSES AS A T ASDA GAS DUE TO SMALL DIFFUSES AS A GAS DUE TO SMALL DIFFUSES AS A GAS DUE TO SMALL DIFFUSES AS A GAS DUE TO SMALL SMALL PARTICLE SIZE SMALL PARTIES SIZE SMALL PURTICLU SI FUNCTION OF DIFFUSIVITY PARTICLE SIZE PARTICLE SIZE PARTICLE SIZE PARTICLE SIZE DIFFUSES USA GAS DUE TO SMALL COEFFICIENT WASHOUT PARTILE SIZE WASHOUT CALCULATED US A FUNCTION WASHOUT CALCULATED USSFNCIO WASHOUT CALCULAE SAHSIV UU CALCULATED AS A FUNCTION WASHOUT CALCULATED AA FUNCTION WASHO UT CALCULATED AS A FUNCTION WASHCACUAHAUECIOAUNO FUCTION O EAON T AUP

NO SURFACE RETENTION. WASHED WASHOUT CALCULATED AS A FUNCTION. O OF O 0WASOUTALCUTEDASANCIONC OF FUNCIONf FUNTIm1F OFF BY PRECIPITATION DROP SIZE DROP SIZE DROP SIZE DROP SIZE TROP SIZE DROP SIZE DROP SIZE DROP SIZE DROPYSIZE

PRECIPITATING INTENSITY PRECIPITATION INTENSITY PRECIPITATION INTENSITY PRECIPITATION INTENSITY PRECIPITATION INTENSITY PRECIPITATION INTENSITY PRECIPITATION INTENSITY PRECIPITATION INTENSITY PRECIPITATION

PARTICLE CHARACTERISTICS PARTICLE CHARACTERISTICS PARTICLE CHARACTERISTICS PARTICLE CHARACTERISTICS PARTICLE CHARACTERISTICS PARTICLE CHARACTERISTICS PARTICLE CHARACTERISTICS PARTICLE CHARA

9MH SURFACE RETENTION Rn SURFACERETNION RONSURFACE9RETENTION 7O SURFACE RETENION 63N SURFACE RETENTION 65S SURFACE RETENTION 9 SURFACE RETENTION 99% SURFACE RETENTION 99 SURFACE RELENT

WATER TRANSPORT CHARACTERISTIC TRITIATED WATER ENTIRE LOAD HIGH DEGREE O ADSORPTION ON MODERATE ADSORPTION O MODERATE ADSORPTION ON HIGH DEGREE OF ADSORPTION ON WATER TRANSPORT CHARACTERISTICS NO SEDIMENT TRANSPORT, NO SEDIMENT TRANSPORT, MODERATE DEGREE O ADSORPTION MADERATEDEGBEE O ADSORPTION MOATE DEGREE

TRANSPORTED BY STREAMFLOW AND SEDIMENTS. LOAD EQUALLY DISTRIBUTED SEDIMENTS. MAJOR PORTIONOF SEDIMENTS MAJOR PORTION O SEDIMENTS. LOAD EQUALLY ENTIRE LOAD TRANSPORTED ENTIRE LOAD TRANSPORTED ON SEDIMENTS. MAJOR PORTION ON SEDIMNTS. MAJOR PORTION ON SEDINTS MA

GROUNDWATER NO ADSORBED SH BETWEEN WATER AND SEDIMENT LOAD CARRIED DISSOLVED IN LOAD CARRIED DISSOLVED IN DISTRIBUTED BETWEEN WATER AND DISSOLVED IN STREAM FLOW DISSOLVED IN STREAM FLOW OfULOAD CARRIED BY STREAM FLOW OF LOAD CARRIED BY STREAM FLOW LOAD CARRIED B

CARRIED BY SUSPENDED SEDIMENT TRANSPORT STREAM FLOW STREAM FLOW SEDIMENT TRANSPORT

DISTRIBUTION COEFFICIENT" 00 SA 52 IS LI DISTRIBUTIN COEFICIENT D0 DO D 0 241 41 05

PERCENT OF AVERAGE ADULT PERCENT OF AVERAGE ADULT ORGAN DOSE BY ALL PATHWAYS ORGAN DOSE BY ALL PATHWAYS TOTAL BODY 4.U DE 003 0.25 D U.UI TOTAL BODY - 5.93 1.3 3.76 308 0.01 GI TRACT 3.E6 0O9 0.49 0.33 113 0.54 El TRACT - 5..5951.34 1 41 063 3.4 THYROID U.H 80.02 0.0 0E1.14 0.0 THYROID ODE 672. 0. 0. 40.160 .D

RNE ---- 0.61 0.82 1.65 3. 0.T BONE --- 404 H.12 18.8 22. 0.0 LENG 88.8 0.01 0.26 0.52 0.01 LUNG --- 5.18 1.4 1.0 1.03 0.01 SKIN 81.6 5.18 001 0 2036 002 SKIN -.- 6.1 3.34 1.58 0.63 0.01 LIVER 81.7 0.09 0.04 0.24 0.48 0.5 LIVER 5.98 1.31 4H9 4.0 0.01

PATHWAYS FOLLOWED IN APPLYING TRANSPIRATION PATHWAYS FOLLOWED IN APPLYING EXPOSURE TO SOIL EXPOSUREATRSOIL

TOAL ODY DRINKING WATER AIR SUBMERSION EXPOSURE TO SOIL EXPOSURE TO SOIL OTAL ODY REEXPOSUREEF0LMAIRDSUBMERSIONULDRINING WATER DRINKING WATER

BEEF 8 LAMB

AIR NHAATIO BEFA AMB OE IIC AER AIR SUBERSION EXPO SURE TO SOLt EPoSU TOOILl GI TRACT ARINKINGWATER PAIR SUBMERSION AIR INHALATION AIR INHALATION AIR INHALATION AIR INHALATION GC TRACT EINGURE TO EIL DRINKING WATER DRINKING WATER SAIR INHALE TRANSPIRATION EXPOSURE TO 501L EXPOSURE TO IL E

DRINKING WATER

THYROID AIR INHALATION AIR INHALATION 3

R TRANSPIRATION A:R INHALATION AIR INHALATION AIR INHALATION FRESH MILTK FRESH MILK AIR INHALATION AIR INHALATION DINING WATERREXPOSURE TO SOIL EXPOSURE TO SOIL BEEF AND LAMB BEEF AND LAMB EXPOSURE TO SOIL EXPOSURE TO SOIL BEEF K LAMB EXPOSURE To SOIL

EXPOSURE TO SOIL AIR INHALATION AIR INHALATION

REEF K LAMB REEF A LAMB EXPOSURE TO SOL EXPOSURE TO SOL EXPOSURE TO SOIL AIR INHALATION FRESH MILO FRESH MIL BONE AIR SUBMERSION AIEXSURETSIL AIR INHALATION EONE FRESH MILK WHEAT WKAT PROD. WHEAT WHAT PROD. WHEATUWI-AT PROD. AIR SUBMERSION AIR SUBMERSION DRINKING WATER DRINKING WATER WHEAT A WIAT PROD. EPOSURE TO SOIL EXPOSURE TO SOIL

AIR INHALATION BEEF & LAMBWATER LUNG DRINKING WATER AIR SUBMERSION EXPOSURE TO SOIL EXPOSURE TO SOIL LUNG EXPOSURE TO SOIL AIR SUBMERSION EXPOSURE TO SOIL EXPOSURE TO SOIL 3H TRANSPIRATION

SKIN 3 TRANSPIRATION AIR SUBMERSIOSKIN EUPOSUR TA SOIL AIR SUBMERSION EXPOSURE TO SOIL EXPOSURE TO SOIL

AIR INHALAT ION AIR INHALATION DRINKING WATER EXPOSURE TO SOL EXPOSURE TO O

HTRANSPIRATION EXPOSURE TO SOIL EXPOSURE TO SOIL LIVER DRINKING WATER DRIVING WATER AIR INHALATION7BED ANDAREEF IIAMB

" DISTRIBUTION COEFFICIENT K IS THE RATIO OF THE FRACTION OF IONS ABSORBED PEA UNIT MIGHT OF SEDIMENT TO THE FRArTION OF

IONS REMAINING IN SOLUTION PER UNIT VOLUME UP SOLUTION.

IX-9

e

SS

D BY LWR

ROVED FROM

COLD TRAP

WITH A DF

IFFUSES FROM

ATE RELATIVE

N PRODUCTS

D BY ION

MENT SYSTEM

5X 105

PARTICULATE

S DUE TO

IZE

TED ASA

INTENSITY

CTERISTICS

TION

F ADSORPTION

JOR PORTION

F STREAM FLOW

ATION

LTIG

CHILD

'71 OTHER 

'. C

CD

.I1

G)0c0

0.

CD0

a,

CD

CD

o'

\\

//

ADULT

OTHER

+-- 

1

3

1

I

- 3H

LU

I-0

LUW

I--J

LW

131

I

1 l 1 1

TEEN
TABLE IX-3

Pathways Found to Contribute to

Dose and Dose Commitment

ANNUAL DOSE FIFTY YEAR DOSE COMMITMENT

GI

Tract Thyroid

Total

Bone Lungs Skin Liver Body

GI

Tract Thyroid Bone Lungs Skin Liver

Air Inhalation

3H Transpiration

Exposure to Soil

Air Submersion

Drinking Water

Foods

Beef & Lamb

Fresh Milk

Wheat and Wheat

Products

X X X X X X X X X X X X X X

X X X X X X X X X X X X X

X x x x x x x

X x x x x x x

X x x x x x x x x x x x x x

X x x x x x x x x x x x x x

X x x x x x x x x x x x x x

X x x x x x x x x x x x x x

ranging from 74% of bone dose to 99.9% of skin dose

(although they account for only 32% of dose to the

child's thyroid). Two of these pathways-air inhalation

and tritium transpiration-were calculated to contribute,

for the average adult, virtually all the 50-year dose

commitment to the total body (99.2%) and lungs (81%),

as well as most of the dose commitment to the liver

(87%) and thyroid (58%).

The pathways discussed to this point depend solely

on air transport. Centroids in close proximity to

radioactive sources tend to receive larger doses because

receptors further downwind are subjected to lower

concentrations of radionuclides, diminished by decay,

diffusion, and removal by dry deposition and washout.

The nuclides 1311, 1 34 Cs, and 1 3 7Cs, which contribute

the major portion of the dose from exposure to soil, are

accumulated on the ground principally by deposition

from the air, rather than by deposition with irrigation

water.

The drinking water pathway primarily includes radionuclides released with liquid effluents from power

reactors; however, a mechanism exists, both physically

and in the model, for transfer of air-deposited radionuclides from ground surfaces to waterways in the

runoff of rain water. Drinking water contributes to the

annual dose

50-year dose

GI tract.

to all the organs considered and to the

commitment of all the organs except the

The radionuclide contributions to the dose from

drinking water vary for the different organs considered.

Tritium and cesium account for most of the drinking

water dose to total body, GI tract, lungs and liver. The

bone receives most of its drinking water dose contribution from cesium, while the thyroid dose is primarily

from 1311.

The remaining pathway of significance is the food

pathway, i.e., fresh milk, beef and lamb, wheat and

wheat products. Foods contribute radiation doses which

reflect a combination of the concentration of radionuclides deposited on vegetation from the air and the

concentration of radionuclides in water. The isotopic

composition of the doses from the food pathway is

comparable to that of the drinking water pathway. The

ingestion of foods and water plays a particularly important role in the dose to the child's thyroid, contributing

about 67% of the total dose.

Additional data detailing the pathway contributions

to the various organ dose rates and dose commitment are

included in Appendix E.

IX-12

Total

Body
X. SENSITIVITY STUDIES:

EFFECTS OF PARAMETER VARIATIONS

The evaluation of the potential radiation dose to

people in the Upper Mississippi and Lower Missouri

River basins from possible nuclear facilities operating in

the year 2000 was accomplished with the use of the

HERMES digital computer code. The model on which

HERMES is based provides computation of radiation

dose from forty-five radionuclides and ten pathways.

The particular radionuclides and pathways were selected

on the basis of their potential contributions to regional

radiation doses. The results of the present study indicate

that twelve radionucides and six pathways contributed

nearly all of the computed radiation dose, as was

detailed in Chapter IX. Parameter variations were made

and analyzed to understand better the mechanisms in

the model and the input data which affect the significance of various radionuclides and pathways. The

sensitivity studies were designed primarily to indicate

critical parts of the calculation as well as the effects of

the various assumptions made.

MAJOR EFFECTS ON DOSE

Most of the potential radiation dose in the region can

be attributed to radioactivity transported from source to

receptor via the air. Air inhalation, air submersion,

tritium transpiration and exposure from radionuclides

deposited on the ground from airborne effluents account

for about 90% of the calculated dose. The remainder of

the dose comes from ingestion of foods and drinking

water. The drinking-water pathway involves primarily

radionuclides from liquid effluents; all other major

pathways primarily involve airborne radioactivity. Reprocessing plants contribute 80% of the airborne tritium

dose to the region, but less than 2% of all other

contributions to dose from airborne effluents.

Tritium Release Considerations

The greatest portion of the calculated total-body dose

for the study region is contributed by airborne tritium

released from reprocessing plants. Approximately 95%

of the activity released from these plants was estimated

to consist of tritium. About 40% of this tritium

originates in LMFBR fuels.

In the study, the rate of diffusion of tritium from the

fuel element into the coolant was assumed to be small in

both LWR and LMFBR fuels; hence, most of the tritium

generated within the fuel was released at the reprocessing plant. However, recent experiments involving the

dissolution of stainless steel-clad LMFBR spent

fuel( 15 7) indicate that substantial diffusion of tritium

from these fuels may occur.

Figure X-1 indicates the effects on reprocessing plant

releases, with different LMFBR fuel cooling times,

which would result if as much as 90% of the tritium

generated diffuses from the LMFBR fuel during its

residence in the reactor.

It appears to be feasible to remove tritium from

LMFBR coolant systems, as sodium hydride, by cold

trapping. Cold trap efficiencies of from 75 to 90 percent

(i.e., decontamination factors of 4 to 10) may be

attainable on a single-pass basis. In a continuously

recirculating cleanup loop, with continuous removal of

wastes from the trap, much greater effective decontamination factors may be achievable. However, diffusion of

tritium into the environment (through pipe walls, etc.)

may decrease the effective decontamination factor somewhat.

Similarly, the introduction of volatile oxidizers

(voloxidizers) for head-end treatment of LMFBR fuels at

reprocessing plants may afford decontamination factors

up to 100 in reducing tritium and noble gas releases.

Since tritium is indicated in the study to contribute

so strongly to dose, the development of effective

treatment systems at both LMFBRs and reprocessing

plants could result in major reductions in human dose.

The potential effects of various combinations of

tritium behavior and reprocessing plant treatment on

dose were evaluated with a number of calculations over a

wide range of tritium diffusion behavior and treatment

system effectiveness. The results indicate that:

" If 70 percent of the LMFBR-generated tritium

were lost from the fuel into the sodium coolant

through diffusion, and subsequently treated by

cold trapping, the potential total-body dose in the

study area would be reduced to an estimated 3700

man-rems, as compared to 4940 man-rems for the

reference case (diffusion of 90% of the tritium

would lower the dose further to about 3350

man-rems).

" The addition of voloxidizers with associated tritium treatment system having decontamination

factors of 100, even in the absence of in-reactor

diffusion of tritium, would lower releases from

reprocessing plants to the extent that potential

X-1
REPROCESSING FUEL LOAD

150-DAY-COOLED LWR FUEL AND

150-DAY-COOLED LMFBR FUEL

REPROCESSING FUEL LOAD

150-DAY-COOLED LWR FUEL AND

30-DAY-COOLED LMFBR FUEL

90% TRITIUM

DIFFUSION

85Kr

-J

3H

O

E--

I-

-J

I

& -

REFERENCE

85Kr

3H

90% TRITIUM

DIFFUSION

OTHERS

l Xe 1311

85Kr

133Xe

-J

Ih-

-J

LW

REFERENCE

1 33Xe

f, OTHERS

13lmXe

131

85K

- -- 

Figure X-1.-Effects of In-Reactor Tritium Diffusion on Reprocessing Plant Effluents.

OTHERS

100

80

1-1

1-1

LL.

-J

LL.

LU

CC

60 t40 F20 F0 - -

X

N
regional total-body dose would be reduced to 937

man-rems (tritium would account for only about

17% of this dose).

" The combined effects of 70% in-reactor diffusion

(in LMFBRs) and voloxidizer-associated treatment

of tritium at reprocessing plants would lower the

potential regional total-body dose to 923

man-rems-only slightly lower than for the case

where no diffusion was assumed.

Effect of Inclusion of Chicago Area

The base case calculations omitted the metropolitan

Chicago area, which is outside the region defined as the

UMRB area, although it is immediately adjacent to it. To

evaluate the effects of inclusion of this major metropolitan area, an estimate was made of the dose received

by people of the Chicago area. In making this estimate,

it was not possible to evaluate the radionuclide transport

in Lake Michigan waters in the same detail as was done

for rivers in the study area. Thus, a simplified estimate

of radioactivity in lake waters was made, introducing

uncertainties into dose contributions related to the lake.

However, since air releases were estimated to make the

major contribution to dose rate, the effect of these

uncertainties tended to be minimized.

The Chicago area, included as two centroids, was

estimated to have a year 2000 population of 9.5 million.

These people received an average dose rate of 0.12

mrem/yr, compared to the value of 0.17 for the UMRB

base case study. If the Chicago area were included in the

UMRB area, the overall average dose rate would be

0.16 mrem/yr.

Inclusion of the Chicago area was estimated to

increase the integrated dose value to 6050 man-rems,

from the 4940 man-rems estimated in the base case.

Effects of "Air Envelope" Boundary Region

The dose calculation results show that air transport

mechanisms have a major effect on the regional radiation

dose. The dose to any region of the country will be

affected by sources in surrounding regions, primarily by

interchange of radionuclides through the air.

To account for these added contributions, an "air

envelope"-a band roughly 200 miles in width, surrounding the central study area-was established. Nuclear

industry growth projections in this peripheral zone were

made along with those of the central study area, and the

estimated airborne releases from plants in this zone were

considered along with airborne and waterborne radionuclide releases from plants within the central study area

in the estimation of population dose in the central area.

To determine the sensitivity of dose in the central

area to those peripheral plants, a calculation was made in

which all releases in the "air envelope" region were

omitted. The integrated total-body dose calculated for

this case was 4027 man-rems, a decrease of about 18%

from the 4940 man-rems of the base case.

Comparisons of the calculated doses with and without the air envelope for several individual centroids

shows, as expected, the dependency of air envelope

effects on prevailing wind patterns. Doses in centroids

on the eastern side of the study area are virtually

unchanged by elimination of contributions from the air

envelope; those in centroids on the southern and

southwestern perimeter are sharply reduced.

It would appear that a smaller air envelope might be

used without greatly detracting from accuracy of the

calculated dose; however, some accounting of the air

envelope contribution appears to remain necessary.

Effects of Air Deposition on

Water Concentrations

To evaluate the contribution of air deposition of

radionuclides to their concentrations in water, a water

transport model calculation was made, identical in other

respects to the base case, in which contributions from air

deposition were deleted. With the deletion of air

deposition, concentrations of radionuclides in groundwater and standing surface water dropped to zero (air

deposition is the only contamination source for these

waters). Contamination of river waters decreased to a

greater or lesser degree, depending on the magnitude of

releases to water at upstream points and on the ratio of

flow derived from surface runoff within the area to flow

originating outside the area. For example, concentrations at Centroid 238, on the Missouri River in northwestern Iowa, dropped to zero when air-deposited

contributions were removed; no releases to water were

assumed to occur above that point.

For five centroids fronting on major rivers in the area,

the percentages of total activity contributed by air

deposition, for several radionuclides, are shown in Table

X-1. The activity due to tritium and the cesium isotopes,

which contributes the greatest portion of dose from

drinking water in most centroids, appears to be derived

only to a small extent from air deposition. Other

radionuclides show greater dependence on air deposition, although in most cases they contribute relatively

little to dose. For ruthenium the air-derived contribution

is small for three of the centroids, but rises to

appreciable percentages at St. Louis and Kansas City

(probably caused in each case by upstream air deposits

from reprocessing plants). Activities of the other

isotopes investigated are appreciably affected by air

deposition.

The behavior of 1311 is unique; contributions of

air-deposited activity for this nuclide are relatively small

for most of the centroids investigated, but are responsible for over 90% of total 1311 activity in the water at

St. Louis. This apparently results from deposition of

airborne material relatively close by, so that this

relatively short-lived nuclide has little time to decay in

traveling downstream.

Air-deposited contributions to the activity of 1291

were uniformly high; the major source of this nuclide as

X-3
TABLE X-1

Contributions of Air Deposition to River Water Activity

Centroid 34 68 87 166 222

River Illinois Mississippi Mississippi Mississippi Missouri

River Mile 240 500 195 844 367

Isotope Percent Contribution from Air Deposition

3H 0.10 0.51 1.30 1.08 1.82

9 0Sr 41.6 69.1 93.2 82.3 98.4

9 5Zr 56.0 82.8 95.5 89.6 97.4

9 5N b 67.9 87.8 93.1 93.1 98.5

106R u 3.8 4.5 20.2 8.8 32.1

1201 95.3 83.0 96.5 91.8 98.9

1311 21.7 24.0 93.1 12.2 23.4

134Cs 5.4 5.0 10.1 2.7 16.6

137Cs 1.8 1.5 3.0 7.8 4.2

TABLE X-2

LWR Fuel Reprocessing Plant Effluent Variation with Preprocessing Cooling Time

(Release Values in pCi/sec)

COOLING TIME

180 Days 150 Days 120 Days 90 Days

3H 3.27E + 10* 3.29E + 10 3.30E + 10 3.32E + 10

8 5Kr 1.75E + 09 1.76E + 09 1.76E + 09 1.77E + 09

9 0Sr 8.18E + 05 8.19E + 05 8.21E + 05 8.23E + 05

9 5Zr 2.01E + 06 2.76E + 06 3.80E + 06 5.23E + 06

95mNb 3.92E + 06 5.22E + 06 6.85E + 06 8.83E + 06

106R 3.66E + 05 3.87E + 05 4.10E+05 4.34E + 05

1291 4.47E + 02 4.47E + 02 4.47E + 02 4.47E + 02

1311 1.44E + 03 1.90E + 04 2.52E + 05 3.33E + 06

133Xe 3.97E + 01 1.96E + 03 1.01E + 05 5.22E + 06

135Xe 0 0 0 0

134Cs 1.23E + 05 1.26E + 05 1.29E + 05 1.33E + 05

137Cs 6.32E + 04 6.34E + 04 6.35E + 04 6.36E + 04

*3.27E + 10 means 3.27 X 1010.

X-4
TABLE X-3

LMFBR Fuel Reprocessing Plant Effluent Variation with

(Release Values in pCi/sec)

COOLING TIME

3H

85Kr

90Sr

95

Zr

95mNb

106Ru

1291

1311

13 3Xe

135Xe

134Cs

137Cs

150 Days

2.17E + 10

7.87E + 08

2.30E + 05

2.92E + 06

5.49E + 06

2.24E + 06

3.12E + 02

1.90E + 04

7.07E + 02

0

7.69E + 03

3.2E + 04

120 Days

2.19E + 10

7.91E+08

2.31E + 05

4.02E + 06

7.21E + 06

2.37E + 06

3.12E + 02

2.52E + 05

3.66E + 04

0

7.90E + 03

3.21E+04

90 Days

2.20E + 10

7.96 + 08

2.31E + 05

5.53E + 06

9.28E + 06

2.51E + 06

3.12E + 02

3.33E + 06

1.89E + 06

0

8.12E + 03

3.22E + 04

Preprocessing Cooling Time

60 Days

2.21E + 10

8.00E + 08

2.32E + 05

7.61E + 06

1.16E + 07

2.66E + 06

3.12E + 02

4.41 E + 07

9.76E + 07

0

8.34E + 03

3.23E + 04

30 Days

2.22E + 10

8.04E + 08

2.32E + 05

1.05E + 07

1.38E +07

2.82E + 06

3.12E + 02

5.84E + 08

7.49E + 09

0

8.57E + 03

3.23E + 04

calculated by the HERMES model was in the gaseous

releases from reprocessing plants.

Spent Fuel Cooling Time

The cooling time for both LWR and LMFBR fuel was

taken to be 150 days in the study. Five months is

considered to be a reasonable time between removal

from the reactor and chemical reprocessing for LWR

fuels. Application of this cooling time to LMFBR fuels

allows them to be reprocessed at one-half the design feed

rate in LWR fuel reprocessing plants.( 1 5 5) However, the

economic penalties associated with fissile plutonium

inventories in spent fuel may make it desirable to reduce

the reactor-to-reprocessing time to as little as 30 days.

Such short-cooled fuel will require special chemical plant

designs to provide sufficient decontamination for shortlived radionuclides. The calculational model was augmented with simulation of a predissolution oxidizer to

provide additional treatment for the noble gases, iodine,

and ruthenium.( 1 55)

Release rate changes resulting from variation in

cooling time, for those radionuclides which contribute

significantly to dose, are shown for LWR fuel in Table

X-2 and for LMFBR fuel in Table X-3. The releases show

significant variation in effluent activity for 1311; this

activity increases by slightly more than an order of

magnitude each time the cooling is reduced by a month.

However, there is less economic incentive to reduce the

reactor-to-reprocessing time for LWR fuels.

As an indication of the sensitivity of regional dose

values to LMFBR fuel cooling time, a calculation was

made whereby LMFBR fuel was assumed to be cooled

30 days prior to reprocessing. LWR fuel was cooled 150

days as in the base case calculation. This change resulted

in the potential regional integrated population dose

TABLE X-4

Ratio of LMFBR Reprocessing Plant Effluent

Activities with 30 Days Versus 150 Days

Preprocessing Cooling

3H

85Kr

90Sr

9 5Zr

95mN b

106Ru

1291

1311

133Xe

135Xe

134Cs

137Cs

1.02

1.02

1.01

3.59

2.52

1.26

1.00

3.07 X 104

1.067 X 107

1.11

1.01

being increased by a factor of 1.4 to 6940 man-rems.

This calculation assumed no change in reprocessing plant

radwaste system performance to permit comparison with

the base case. The majority of the increase was contributed by Iodine 131 and Xe 133 as one would expect

by examining the ratio of LMFBR reprocessing plant

effluent activities with 30 days versus 150 days

reprocessing cooling given in Table X-4.

PLANT DESIGN AND OPERATION EFFECTS

The HERMES model, of necessity, includes assumptions as to types of plants included in the region and the

operation of these plants. To understand better the

X-5
sensitivity of the model to changes in these assumptions,

several calculations were made with changed parameters

relating to plant design and operation, to note the

variation in results.

Improved Treatment System

Introduction Date

The treatment systems included on the nuclear power

plants and fuel reprocessing plants (discussed in Chapter

V) included treatment systems such as are being designed in many cases to meet Federal guidelines. Of

interest is the effect of a delay in introduction of such

systems. The effect on dose of such a delay until 1985

on reactors and of not having an improved system on

reprocessing plants was evaluated. In this case the light

water reactors built before 1985 were assumed to be

operating without the improved systems discussed in

Chapter V. All LMFBRs and advanced LWRs were

assumed to have advanced systems in use since these

reactor types did not come into use until the 1985 date.

Purex reprocessing plants were assumed neither to

include the augmented gas treatment system nor to

solidify all liquid wastes; thus, they released some liquid

effluents as well as a larger amount of noble gases.

The total releases from the nuclear operations in the

region were changed considerably; the potential total

regional releases, including those in the air envelope, are

compared to those for the base case in Appendix C.

As one would expect, doses calculated in this study

were changed from those in the base case; the patterns

observed in accumulating those doses also changed

markedly. Noble gas releases from reprocessing plants

became more significant, as did the short-half-life noble

gases from BWRs with shortened holdup time, for

centroids near these reactors. The air submersion pathway was a negligible contributor to total-body dose in

the base case. In this sensitivity run, air submersion was

found to contribute up to 30% of the total-body dose at

centroids near BWR facilities.

The calculated effects of the delayed introduction of

the improved treatment systems on average adult totalbody dose rate at one centroid are shown in Table X-5.

At this centroid 8 5 Kr appears to contribute about

one third of the 30% increase in total-body dose

between the base case and the sensitivity study; most of

the remainder can be attributed to the nuclides of

iodine. The increased contribution of 8 5 Kr, and part of

the increased dose from iodine, result from the larger gas

releases from reprocessing plants. The remainder of the

iodine contribution can be attributed to differences in

the releases from the older (pre-1985) LWRs.

The integrated population dose was calculated to be

6670 man-rems as compared to the base case value of

4940 man-rems. The effect of the added close-in dose

from increased emissions of short-half-life noble gases

was not included.

TABLE X-5

Changes in Radionuclide Contribution to Dose

Caused by Delay in Installation of

Improved Treatment Systems

Contribution to Average Adult

Total Body Dose (mrem)

(Centroid 143)

Base Delayed Treatment

Contributing

Radionuclides

H-3

Kr-85

I (all isotopes)

Cs (all isotopes)

Other Radionuclides

Total

0.0592

0.001

0.0084

0.0047

0.0028

0.0761

0.0592

0.0086

0.0241

0.0047

0.0051

0.1017

BWR Gaseous Effluent Holdup Time

Time delay is one of the fundamental ways of

reducing the radioactive material to be handled or

released. All plant types utilize some means of delaying

both gaseous and liquid wastes to eliminate short-halflife radionuclides and to reduce inventories of radionuclides with intermediate half lives. Radionuclides with

long half-lives are removed from effluent streams, when

possible, and shipped to long-term storage centers. As far

as reactor releases are concerned, however, the delay of

interest is that used for the elimination of short-half-life

radionuclides and reduction of those with intermediate

half-lives.

Time delay provisions are of principal interest for

gaseous releases because the large volume of the reactor

coolant and the collection of liquid wastes for treatment

as batches provides sufficient delay to eliminate the

shorter-life radionuclides from the liquid wastes. Shorthalf-life gases are also of common interest because the

close-in whole-body dose from early-vintage boiling

water reactors results mainly from the release of steam

jet air ejector gases, with their short half-life gas

concentrations, after only a 30 minute delay.

All reactors operating in the year 2000 were assumed,

in the base case, to be equipped with advanced treatment systems. For the most part, components for these

systems are currently under development or in use.

BWRs were assumed to be equipped with charcoal

beds operated at room temperature to provide a 15-hour

delay for krypton isotopes and a 7-day delay for xenons.

Releases are considerably reduced with use of this

system. If the bed were operated at lower temperatures,

greater delays and hence lower releases could be obtained. To show the effect of these changes, a case was

run with delays increased by a factor of 10-to about

X-6
TABLE X-6

Tenfold Increase

(in Delay)

-0-

3700.

-0-

-0-

3.1

.006

4.7

-0-

Typical Release Rates of 85 Kr from Reactors

(Release Rates at 0.8 Plant Factor)

Plant Type

BWR

BWRA*

PWR

PWRA*

LMFBR (with FREON System)

LMFBR (without F R EON System)

Effect of Delay Time on BWR Noble

Gas Releases

Noble Gas Release (Curies/year)

at 0.8 Plant Factor

*"Advanced" LWR designs differ from "standard" designs

primarily in higher heat rate of fuel which results in increased

diffusion from the fuel. The radwaste treatments systems for

the two designs are similar.

15 hr delay for Krypton, 7 day delay for Xenon.

3 days for kryptons and 70 days for xenons. The

releases for such a system are compared to base case

releases in Table X-6. The indicated reduction in 133Xe

release is particularly interesting; such a reduction would

considerably lower the air submersion contribution to

close-in dose.

Changes in gas system delays for other reactor types

had less effect; the base-case delay of about 30 days

assumed for these types sufficiently lowered the releases

of the short-half-life radionuclides that the added delays

had comparatively little effect.

Noble Gas Treatment System Variants

If all LWRs were assumed to be equipped with

refrigerant-absorption or cryogenic gas treatment

systems, their calculated releases of krypton and xenon

isotopes would be reduced by a factor of 300 (the DF

provided in the model to characterize these systems).

Such a reduction, as previously noted, would decrease

the close-in dose from nuclear facilities. However, on a

regional or centroid-average basis, the incremental reduction would be small (about 10%) for the skin dose, and

negligible for dose to all other organs. (These gases

contribute primarily to skin dose.)

If the LMFBR radwaste treatment systems are

assumed not to include the FREON refrigerantabsorption system, gas releases from these reactors

become comparable to those from LWRs. Comparisons

of releases of 85Kr are shown in Table X-7. For the

reactor types not equipped with FREON gas treatment

systems, the 85Kr release rates noted are the same as the

rates of leakage from the fuel. These release rates in turn

reflect differences in generation rates with different fuel

types and flux spectra, and the effects of fuel power

density and burnup on failure probabilities and on

coefficients of diffusion.

Factors Affecting Radwaste

System Performance

Calculations were made to demonstrate the effect of

changes in various parts of the treatment systems on

power plant releases and to determine the effect on

release rates of various assumptions made in the release

model. A more detailed description of the reactor

models and the base case releases is found in Chapter V

and is discussed in even greater detail in the HERMES

descriptive document.( 1

The Steam Generator Leakage

The computer model of the PWR treatment system

has two flow paths for treatment of liquid effluent. The

first, starting at the waste collection tank (Figure V-2),

principally receives effluent from the primary coolant

system. The second path, shown as originating at the

floor drain tank, accepts both floor drain wastes and

steam generator blowdown (a simplification of actual

PWR systems). Radionuclides in the entering stream for

this path are taken to be representative of those in the

secondary coolant system. With flow rates for the two

paths as used in base-case calculations, the eventual

releases of radionuclides other than tritium were about

equal for the two paths (56% was contributed by the

floor drain path). Most of the tritium release (about 90%)

was contributed by the waste collection tank path. When

the flow in the floor drain path was assumed to double,

simulating increased steam generator leakage and the

attendant increase in blowdown, a 50% increase was

observed, as expected, for radionuclides other than

tritium. Since the model assumed steady state conditions

and tritium is not treated in the radwaste system, rates

of tritium release remain constant.

X-7

Release Rate

(Curies/year)

3700.

5700.

2900.

4400.

17.

5100.

Isotope

Kr-85m

Kr-85

Kr-87

Kr-88

Xe-131m

Xe-133m

Xe-133

Xe-135

Base Case*

140.

3700.

.35

76.

120.

88.

18000.

.054

TABLE X-7
TABLE X-8

PWR Secondary Waste Recycle Effects

on Releases*

PWR Liquid Release Rates (curies/year)

H3

Mn 54

Fe 55

Sr 90

1 131

Cs 134

Base Case

3.91 X 102

3.47 X 10~4

2.68X 10- 3

5.36 X 10- 5

1.17

1.64X 10-1

*Release values include contributions from

steam generator blowdown.

Sensitivity Study

(without recycle)

4.01 X

3.03 X

2.24 X

4.42 X

10.2

1.36

102

10-3

10-2

10 

floor drains and

Recyling of PWR Floor Drain Flows

In the model used in this regional evaluation, the

PWR floor drain flows are treated in the same way as the

waste collection tank stream, in that 95% of the treated

streams are recycled to the coolant. Although BWR floor

drains are not recycled, these varying assumptions

between the two reactor types do not appear unreasonable considering the lower coolant concentrations

obtained in the BWR with its higher flow coolant

cleanup system.

To show the effect of floor drain recycling on PWR

releases, a case was run in which the treated floor drain

stream was released and not recycled. The plant liquid

releases, as would be expected, were increased by about

a factor of 10 for radionuclides other than tritium. As

discussed in the previous section, the tritium releases

remained essentially unchanged with the varied flows.

Typical changes in release rates are shown in Table X-8.

The increased release would triple the drinking water

contribution to dose, thus increasing the potential

integrated regional dose about 6%.

BWR Flow Variations

The BWR plants in the UMRB study had treatment

systems which used two flow paths: one fed by the

waste collection tank with coolant radionuclide concentrations, the other path passing through the floor

drain tank and having radionuclide concentrations

reduced to represent the concentration reduction due to

the boiling process. In the base-case calculation, essentially all of the releases of radionuclides, other than

tritium, came from the waste collection tank path

through the radwaste system. The small contribution of

the floor drain path results from the effectiveness of the

coolant cleanup system and the high flow rate through

that system, providing low coolant concentrations, and

the effectiveness of the boiling process in containing

coolant impurities in the reactor and in isolating them

from the turbine portion of the plant. Tritium is neither

affected by the coolant cleanup system nor contained by

the boiling process; thus, 90% of this nuclide is released

by the floor drain path.

To show the reaction of the model to a modification

in waste stream flow, a calculation was performed in

which the base case flows of 0.79 liter/sec to the waste

collection tank and 0.40 liter/sec to the floor drain were

changed to 1.1 and 0.1 liters/sec, respectively. The liquid

releases of non-tritium radionuclides increased slightly,

as would be expected from the increased flow using the

waste collection tank path, which was observed to

provide most of the contribution to release of these

radionuclides. The tritium releases were essentially unchanged due to the steady state nature of the model.

The changes in flow rate assumed for this case resulted

in a lower total volume of liquid release. To accommodate the steady-state conditions for production and

release of tritium, its concentration in the coolant

increased to some 2.6 times that in the base case.

Demineralizer and Evaporator

Performance

In the BWR the majority of the non-tritium release

contribution is from the path through the waste collection tank; this stream is treated by a demineralizer. In

the PWR essentially equal fractions of the release were

assumed to be provided by the two radwaste system

flow paths, and both had equal fractions recycled to the

coolant; thus, a change in demineralizer efficiency would

affect only about half of the release. Thus, a tenfold

improvement in demineralizer decontamination factor

would decrease BWR releases by a factor of ten, but

would reduce PWR releases only by about 40%.

The evaporator in the PWR radwaste system (Chapter

V) treats the outlet flow from the demineralizer; changes

in evaporation decontamination factors thus affect the

releases in a manner similar to that discussed for the

PWR demineralizer.

The total decrease in effluent activity which would

result from a tenfold increase in decontamination factors

of these components would lower the integrated potential regional dose by only about 1%.

Variations in Plant Mix

The sequence of introduction of various plant types

in the study area was synthesized by use of the linear

program calculation technique developed in modeling

the national power economy.( 15

,16, 1 7 ) This method

minimizes the present worth of total system power cost

over the period considered in the study. The mixture of

plants selected in the LP calculation depends on the

projection of economic and technical data into the

future; changes in these parameters would affect the

plant mix. Changes in other factors-the degree of

X-8
clustering of nuclear plants, unit size of reprocessing

plants, or the assumed date of introduction of improved

treatment systems, for example-can affect dose. Several

studies were performed to estimate the effects of such

changes in the plant mix on the resulting dose.

Nuclear/Fossil Ratio

Obviously, an increase in the number of fossil plants

operating in the region whould reduce the number of

nuclear plants and thereby reduce the population dose.

(No radioactive releases from fossil plants were considered in the study.) This sensitivity case was intended

to quantify the degree to which changes in the nuclear/

fossil ratio change the resulting doses and to serve as a

check on the model. To accomplish this, a case was run

with the number of fossil plants approximately doubled

from 88-in the base case to 175 fossil plants operating in

the region (including the air envelope). This reduced the

nuclear capacity by 25 percent (from 349 GW to 262

GW). The integrated dose was calculated to be within 2%

of the anticipated 25% reduction, verifying the direct

dependence of dose on nuclear capacity.

PWR/BWR Ratio

In the base case there were 3 times as many PWRs as

BWRs, corresponding to the present-day ratio of the

number of manufacturers of each plant type. Since the

releases which characterize each plant type are different,

it was felt necessary to examine the sensitivity of dose to

the split between LWR types. This was accomplished by

comparing releases for two extreme cases-one in which

all LWRs were represented by PWRs, and one with all

BWRs. In general, the overall organ doses are relatively

insensitive to variations in LWR type, principally

because of the overriding influence of the reprocessing

plant gaseous 3 H release.

One would expect the all-BWR case to exhibit a

higher contribution to "close-in" dose (dose to persons

near plant site boundaries), as discussed in Chapter IX.

A calculation of the sensitivity of close-in dose to reactor

type was made. In the base case, the close-in dose

contribution in the region was estimated to be about

240 man-rems per year, of which about 16 man-rems/yr

were contributed by the nuclear power reactors in the

region (the balance of 224 man-rems/yr accrued near

reprocessing plant sites).

When all LWRs are assumed to be BWRs, the

contributions to close-in dose from reactors increase

from 16 to 35 man-rems/yr. If they are all assumed to be

PWRs, this contribution drops to 5. man-rems/yr. It

should be noted that an increase in the delay time in the

charcoal bed on the BWRs, as evaluated in a sensitivity

study discussed in a previous section of this chapter,

results in reduced BWR xenon and krypton releases; the

resulting close-in dose would be similar to that from

PWRs.

Since such a large proportion of the total-body dose

is contributed by tritium through transpiration and air

inhalation, and since essentially all the tritium in the air

comes from reprocessing plants, the potential total-body

dose is relatively unaffected by variations in light water

reactor type. Of more significance are doses to some of

the organs and their contributing pathways and

radionuclides.

One of the major differences between the reactor

types is in the liquid tritium release rates, which shows

up in the dose to some organs from the drinking water

pathway. Tritium in drinking water contributes to dose

mainly at centroids where drinking water is obtained

from rivers and at points downstream from a number of

plants. Since only LWR plants were assumed to have

liquid releases and there is little coupling between the air

releases and waterway concentrations of tritium (this

was discussed previously in this section), any change in

reactor sources which affects the tritium liquid releases

will directly affect the drinking water dose contribution.

Replacing all the PWRs with BWRs would therefore

reduce the tritium liquid releases by a factor of about

2.3 from 28,000 curies/yr to 12,000 curies/yr. Consequently, the dose contribution from tritium through

the drinking water pathway would also be reduced in

roughly this proportion (the effect on potential total

dose, however, would be minor). Replacing all BWRs

with PWRs would increase the total liquid tritium

release from 28,000 to 34,000 curies/yr, effecting a

slight increase in dose from tritium.

The fraction of the skin dose contributed by 133Xe is

increased by a factor of 3 in the all-BWR case because of

the shorter delay time in the BWR charcoal bed when

compared to the PWR tanks. The 8 5

Kr contribution to

dose is unaffected by the variations in LWR type

because of the long half-life of this nuclide.

Releases of 131I to the atmosphere for the all-BWR

case are nearly double the base case release rate. Iodine

contributes to the dose for all organs by the exposureto-soil pathway and is a major contributor to thyroid

dose but primarily from ingestion. The increased 1311

release rates in this case result in potential dose rates to

the total body about 8% above those in the base case;

dose to the liver is increased by about the same extent.

Thyroid dose is increased by 70% (90% for the child's

thyroid), bone dose by 35%, and that to the GI tract,

skin and lungs by 7%. Detailed tabulations of the

releases used in these cases are included in Appendix C.

Clustering of Plants

The "nuclear park" concept, featuring relatively large

numbers of nuclear power plants located at a common

site with fuel fabrication and reprocessing facilities, has

attracted considerable interest in regard to the potential

economic advantages of such arrangements. The concept

of clustering can have important effects on population

dose, depending on the specific sites considered for

clustering. Some of the potential effects include:

X-9
" Doses to individuals living near clusters may be

expected to be higher than would be found with a

more conventional plant siting pattern.

" Clustered plant sites will tend to be more remote

than conventionally sited power plants; therefore,

clusters at well-chosen sites may potentially reduce

integrated population dose.

To observe the effects of clustering nuclear plants on

dose, two cases were run for comparison with the base

case. In each of these cases, all nuclear power plants,

other than those known to be actually committed for

construction, were assumed to be built at "nuclear park"

clusters. A typical clustered site was assumed to include

a reprocessing plant and enough nuclear power plants

(when the site is "filled") to utilize its fuel reprocessing

capacity. In the first "cluster" case, the reprocessing

plants were assumed to have a capacity (based on LWR

fuel) of one metric ton per day; in the second case, plant

capacities of five metric tons per day were assumed.

In the base case, some 132 nuclear generating sites

were included, with an average site capacity of 3000

MW. In addition, nine reprocessing plants were

included-eight plants of five tonnes per day capacity

each, and the one-tonne/day facility at Morris, Illinois.

Some 55 generating sites and six reprocessing plants (five

5-tonne plants and the Morris facility) lay within the

central study area. For the clustered case assuming

one-tonne plants, some 66 sites were required, having an

average aggregate capacity up to 7500 MW. Thirty-four

of these sites contained reprocessing plants.* Within the

central study area some 24 sites were assumed; 16 of

these had reprocessing plants. In the case of clusters

associated with five-tonne/day plants, 12 clustered sites

were assumed for the year 2000; 10 of these sites had

reprocessing plants (the Morris facility was also assumed

to operate). Of these, four sites, all with 5-tonne/day

reprocessing plants, were within the central study area.

Maps showing the sites selected for all of these options

are included in Appendix B.

The average adult total-body dose rates calculated in

the case with the clusters with 5-tonne/day reprocessing

plants, as shown in Figure X-2, were higher than found

in the base case as one would expect with individual

source strengths increased. The highest values were

calculated at centroids near these large clusters. In the

case with smaller clusters with 1-tonne/day reprocessing

plants, the largest single source strength is reduced; thus,

the maximum dose rate calculated is also reduced.

The considerations of the patterns of potential dose

contribution become even more important in the selection of sites for nuclear clusters than when more diffuse

siting patterns are used. Locations selected for individual

sites having concentrations of facilities typical of clusters, with attendant higher release rates, can significantly

affect the contributions to population dose.

*Reprocessing plants were assumed to be added at a cluster

when the installed power plants could make use of 50% of the

reprocessing facility capacity.

Reprocessing Plant Size Variation

Because reprocessing plants are potentially large

sources of radionuclides, special attention was paid to

their dose contributions. The sites selected for these

plants had particularly low population densities, as is

fitting with their higher release potential.

In the base case, the 350-plus nuclear power plants

considered (including those in the "air envelope" region

around the central study area) were served by eight

Purex-type plants of nominal 5-tonne/day capacity

(based on LWR fuel) plus the 1-tonne/day Morris,

Illinois facility.

To evaluate the effects of variation of reprocessing

plant size on dose, three cases (without reactor releases)

were run in which the nominal Purex-type plant size was

set, respectively, at 5, 10 and 15 tonnes/day. Since the

total reprocessing load for the region remained unchanged, the number of required plants was inversely

dependent upon plant size: 8-5-tonne/day plants of the

Purex type (plus the 1-tonne/day facility at Morris,

Illinois) in one case, 4-10-tonne/day plants in another,

and 3-15-tonne/day plants in the third. Sites selected

for the plants of various sizes are shown in Figure X-3.

Efficiencies of the radwaste systems were assumed to

be the same for all sizes of plants; thus, with the same

fuel reprocessing load, the total radionuclide releases

would be the same in the three cases. The primary

differences in integrated population dose in the study

area would therefore be those associated with the

operation of fewer, more concentrated radionuclide

sources. With fewer reprocessing plants assumed in the

test cases, the siting of individual plants assumed greater

importance in their contributions to total dose.

In the base case (5 MT/day plants), 26 tonnes per day

of reprocessing capacity are located within the study

area. Releases from one additional plant, located in the

southwest portion of the air envelope, probably also

contribute to dose. The two plants located in the eastern

portion of the air envelope probably do not contribute

appreciably, since prevailing winds near those plants

tend to carry releases away from the study area. Thus, a

total capacity of about 31 tonnes/day might be expected

to contribute to dose.

In the case assuming 10-tonne/day plants, 21

tonnes/day of capacity are located within the area;

again, 31 tonnes/day might contribute to dose. The 15-

tonne/day case assumes the location of 31 tonnes/day

within the study area; potentially, the entire capacity of

46 tonnes/day could contribute to dose.

The contributions of reprocessing plants to the

integrated population dose in the study area are shown

for each case in Table X-9. Tritium releases from the

plants account for approximately 97% of the doses

shown.

If the integrated dose for the 15-tonne/day case were

normalized to the probable "effective" capacity indicated in the other cases, this dose would be 3155

man-rems, rather than the 4681 man-rems contribution

X-10
2.1

2.011 MAXIMUM* (2)

1.8 

1

(1) BASE CASE

1.5

(2) .- -- 5 METRIC TON/DAY REPROCESSING PLANT CLUSTERS

(3) ------ 1 METRIC TON/DAY REPROCESSING PLANT CLUSTERS

1.2- *MAXIMUM: HIGHEST AVERAGE DOSE CALCULATED FOR

ANY OF THE 300 CENTROIDS CONSIDERED

1.035 MAXIMUM* (1)

0.9 0.899 MAXIMUM* (3)

0

0

0.6

0.3

1 - - - - ---

0.0

0 10 20 30 40 50 60 70 80 90 100

POPULATION (PERCENT)

Figure X-2.-Effect on Total Body Dose Rate Distribution of Various Clustered Plant Configurations.
o\'

Z IC

0 5 MT PLANT+ 1

10 MT PLANTr

0 15 MT PLANT

MORRIS,ILL. PLANT "% N 

X (1 MT)

Figure X-3.-Reprocessing Plant Sites: 5, 10 & 15 MT/Day.
TABLE X-9

Effect of Reprocessing Plant Size on Contributions to Regional Integrated Dose

No. of Contributions

Plants(a) to Integrated

in Central Total Body

Area Dose Man-Rem

6

3

3

3391

3197

4681

Plant Capacity

Probably Effective(b)

in Contributing

to Dose

31

31

46

Normalized(c)

Dose

Man-Rem

3391

3197

3155

(a) Including the 1 MT/day facility at Morris, Illinois.

(b) Reprocessing plant capacity within the study area or so located in the air envelope that discharges are likely to enter the area.

(c) Dose normalized to 31 MT/day "effective" reprocessing capacity.

calculated in the case (the "effective" plant sites are the

same for the 10-tonne and 15-tonne cases). Thus, for the

sites assumed in the study, the use of fewer plants of

larger capacity appears to result in slight lowering of the

total potential integrated dose in the study area. It

should be emphasized, however, that the potential dose

will be highly sensitive to locations of the individual

plants. Since the largest individual source strength is

increased for the larger plant sizes, the individuals near

these sources could be expected to experience higher

dose rates.

Effects of Fuel Behavior

In-Fuel Diffusion Coefficients

The relationships used for the diffusion of fission

products within reactor fuel were based on the most

applicable data found in the literature.( 1 5 3 ,

1 54) The

effect of changes in values of diffusion coefficients on

the release of fission products was investigated by first

uniformly reducing the diffusion coefficients by a factor

of 10, and then increasing them (over the base case) by a

factor of 10. The diffusion rates of fission products

through the fuel are assumed to be proportional to the

square root of the diffusion coefficients. Eventual

releases of fission product gases to the environment were

observed to also follow this dependency. Tripling reactor

releases would increase potential regional total-body

dose by about 60%.

Diffusivity Dependence on Burnup

The diffusivity of fission products in fuel may be

expected to vary with burnup of the fuel. In the

HERMES code this variation of diffusivity was accounted for by including in the diffusion calculations a

"burnup factor" defined as:

Fbu = e0. 0 0 0 2 1 8 (Bu - 18000)

where Bu is fuel burnup in megawatt-days per metric

ton. The burnup factor Fbu was constrained to a

maximum value of 100, corresponding to the assumption that once burnup exceeds 40,000 MWD/MT no

further effect on diffusivity occurs. The limiting of Fbu

is based on the hypothesis that once the fuel has

undergone a certain amount of structural change, the

release rate of fission products would be diffusioncontrolled; further structural change would then have

little influence.

To test the sensitivity of releases to this assumption,

the limit on the burnup factor value was increased to

200, corresponding to a cutoff value of 62,000

MWD/MT for effects of burnup. This change has the

effect of assigning considerably higher diffusivities to

fuel at high exposure. Release rates relative to the base

case were increased by a factor of 1.4 for all radionuclides assumed to diffuse from the fuel. Such an

increase would increase potential regional integrated

total-body dose by 10 percent.

Fuel Failure Probability

The release rate of all fission products would be

expected to be directly proportional to fuel failure

probability. To make this intuitive check of the program,

a case was run with the fuel failure rate doubled. As

expected, the release rate was doubled.

The fission product release calculations include an

equation which calculates the distribution of fuel

failures as a function of burnup and linear heat rate. The

equation, of the form:

P = 1. - exp(-c (Burnup)Q (Linear heat rate)')

used a value of 3.4 for 0 in the base case. To test the

sensitivity of releases in the shape of this function a test

case was run with 03= 1.7, while the total fuel failure

probability was held the same.

X-13

Reprocessing

Plant Unit

Capacity

MT/Day

5

10

15

Total

Number(a)

of Plants

9

5

4
The effect of this change is to reduce the dependence

on burnup or spread the failures earlier into the fuel

life. As would be expected, this increased the release rate

of fission products; the increase was about 50% for this

assumed change in the burnup dependence. This increase

in reactor releases would increase potential regional

integrated total-body dose by about 10 percent.

RADIONUCLIDE TRANSPORT

CONSIDERATIONS

Transport of radionuclides from sources to receptors

involves both coupled and uncoupled air and water

mechanisms. Data and assumptions for the natural air

and water parameters which affect radionuclide transport were varied to determine their effect on the

concentration of radionuclides in the region with base

case source terms.

Variations in Air Transport Parameters

Mixing Depth and Stability

Establishment of the vertical diffusion variable, OZ, is

accomplished by two methods. For relatively short

travel distances the "Pasquill" graphical methods attributed to Hilsmeier and Gifford( 2 4 ) are used, while for

longer distances the methods developed at the National

Reactor Testing Station (NRTS)(26) are used. In the

present study the latter case dominates by far because of

the magnitude of the travel distance involved. A test was

made of the effect of varying the parameters which are

used to define the values of oz. The winter case was

chosen as a test case to preserve the other variables and

provide a uniform set of test conditions. It is assumed

that any variability shown would also be found in any

other month. Also, since krypton-85 is unaffected by

any depletion mechanism-other than diffusion-this

radionuclide was chosen for the test.

The variables representing mixing layer depths and

critical distances for inhibition of vertical diffusion were

varied, in self-consistent groupings, as shown in Table

X-10. The resulting concentration-distance relationships

formed a family of curves in which concentrations

decreased as oz values increased, as expected. The

envelope of these curves is shown in Figure X-4, and is

compared with concentration-distance relationships for

the base case.

Figure X-5 shows the calculated concentrationdistance relationship for each of the seven stability

categories as defined by Pasquill. These curves, calculated for radionuclide release from a 100-meter stack,

exhibit the characteristic peaking of concentrations at

relatively short distances from the release point, followed by gradual declines as diffusion proceeds.

Wind Direction and Speed

An investigation was made of the effect of varying

the percent frequency of occurrence of wind direction,

with other parameters remaining at their "base case"

TABLE X-10

Parameter Variations for Vertical Diffusion Sensitivity Studies

Vertical Mixing Depth, L"

Winter Day Winter Night

(meters) (meters)

400

500

600

700

800

900

1000

1100

1200

1300

1400

100

150

200

250

300

350

400

450

500

550

600

Critical Distance, XL 

Winter Day

(meters)

1,000

2,000

3,000

4,000

5,000

6,000

7,000

8,000

9,000

10,000

11,000

Winter Night

(meters)

4,000

5,000

6,000

7,000

8,000

9,000

10,000

11,000

12,000

13,000

14,000

*As used in the ARTRAN Model(I) formulas:

az = 0.465L + (S-X) )(0.335L) For XL s S < 2 XL XL

and Q = O.8L for S> 2XL

X-14

Case

I I

III

IV

V (Base Case)

VI

VII

VIII

IX

X

XI
100 111 1

BASE CASE

10-1 CASE I, TABLE X-10

0

a 0 2

CASE XI, TABLE X-10

1 0 - 3- -- -- --

1 10 100

DISTANCE, KM

Figure X-4.-Range of Concentrations for Calculated Variations of Vertical Diffusion Parameter, OZ.

X-15
100

B

A

C

10'1

Ei \

-t 

Z

(G--

0C

W- E

1 0

G

w

1 1 E

10 0 10 20 30 40 50 60 70

DISTANCE, KM

Figure X-5.-Concentration-Distance Relationship for Pasquill Stability Categories.

X-16
values. The results are shown in Figure X-6. The

systematic increase in air concentration is the expected

result.

A similar test for the effect of wind speed likewise

shows the expected results. The mean wind speed was

varied from 0.5 m/sec to 33 m/sec (-66) mph. The

upper limit is rarely experienced and certainly is far

beyond any possible climatological mean. Figure X-7

shows the results. The steady systematic decrease in air

concentration demonstrates the increased diffusion as a

result of having wind speed as a factor in the dilution.

Height of Release

In the study the stack height was preset at 100 meters

although the capability exists to vary this value. In the

model this factor appears in an exponential multiplier.

To investigate the code response to variation of this

parameter, a test was conducted whereby the stack

height was incremented from 0 (surface) to 110 meters.

The results, a systematic but slow decrease in air

concentration, are shown in Figure X-8. Effects near a

site are more pronounced but vary with local

meteorology, as shown in Figure X-9.

Rainfall Variation

The percent frequency of rainfall has a direct impact

upon the ground accumulation as a result of wet

deposition processes. The corollary result is a decrease in

air concentration as the depletion mechanism increases

in effectiveness. Figure X-10 illustrates the expected

result as the percent frequency of precipitation increases

from 0 to 100%.

Water Transport Parameter Variations

In modeling the waterborne transport of radionuclides, lack of factual data required estimation of

several factors affecting the transport. Since the values

of some of those factors can potentially affect the

calculated dose, studies were performed to test the

sensitivity of estimated radionuclide concentrations to

the values assumed. In addition the effects of variation

in runoff of the river systems were investigated.

Solution/Sediment Transport Relationships

The relationship between concentrations of dissolved

radionuclides and the extent of their sorption on

sediments is expressed through the distribution coefficient, KD. This coefficient defines the ratio of the

quantity of radionuclides sorbed per unit weight of

sediments to the quantity remaining in solution per unit

volume of solution. Data on the solution/sorption

relationships are available for only a few radionuclides,

and these vary considerably depending on the nature of

the sediments in a stream. Consequently the values of

the distribution coefficient used in the study, as listed in

Appendix B, may be subject to considerable uncertainty.

In order to evaluate the effects of changes in these

values on radionuclide transport, test calculations were

made in which all non-zero coefficients were set first at

zero, then at 5000. Results of these calculations for an

example centroid are shown in Figures X-l1 and X-12.

Results are shown for the month of June; concentrations

in other months for the most part showed similar

relationships, although base concentrations varied with

flow, as would be expected.

The suspended-sediment concentrations of 9 0Sr and

10 6 Ru show small but variable changes for increases in

the distribution coefficient above the base-case value of

1050. This behavior was also exhibited by other nuclides

of high absorptivity. In three of the centroids

examined-Centroid 68 (Davenport, Iowa), 87 (St.

Louis, Missouri) and 238 (south of Sioux City,

Iowa)-the concentrations of these nuclides on the

suspended sediment, as well as in solution, decreases

slightly with increase in the distribution coefficient

value. The river reaches in these centroids have high

sediment loads, and in each case, tributaries entering

upstream of the centroid investigated also have high

sediment loads. Consequently, except during high-flow

periods, there is considerable deposition in these river

reaches. Two radionuclide removal mechanisms are

active: sorption on sediments newly introduced by

tributaries and deposition of sediment load. Thus, the

concentration of dissolved radionuclides is decreased as

well as the concentration of sediment-borne nuclides.

Scouring during periods of flood brings the deposited

nuclides back into the flow stream, restoring the mass

balance. The anomalous behavior thus is highly seasonal.

Data for November, when the flow/sediment relationships are different, indicate that for two of the three

anomalous centroids the suspended-sediment concentrations of 9 0Sr increase as would normally be expected

when the distribution coefficient is increased.

Runoff Variation

At St. Louis, Missouri, the average stream flow of the

Mississippi River for the period 1861-1960 was 175,000

cubic feet per second (cfs). During the decade of

1951-1960 the average was 159,000 cfs. The difference

is largely the result of flow variations, long-term regulation and depletions in the Missouri Basin, although

Mississippi River flows were 3,000 cfs below normal for

the period. Table X-11, summarizes U.S. Army Corps of

Engineers' projections of average annual flow in the

Missouri Basin.

Reservoir storage in the Missouri Basin upstream of

the study area is approximately three times greater than

the average annual runoff entering the area. During the

1951-1960 period, part of the runoff was retained for

initial filling of reservoirs in the system. Thus, if the

1951-1960 period is used as the base for flow comparisons, the net effect of additional depletions, relative to

X-17
103

102

F1-

-I

CONTRIBUTIONS TO CENTROID NO. 1

FROM ALL SOURCES

1006

0 20 40 60 80 100

FREQUENCY OF DIRECTION

Figure X-6.-Effect of Wind Direction on Air Concentration of 8 5Kr.

X-18
1 0"

CONTRIBUTIONS TO CENTROID NO. 1

FROM ALL SOURCES

102- ---

0

}-4_

O 

00

011

0 2 4 6 8 10 12

MEAN WIND SPEED, M/SECOND

Figure X-7.-Effect of Wind Speed on Air Concentration of 85 Kr.

X-19
46 I I

45

- CONTRIBUTIONS TO CLNTROID NO.1

FROM ALL SOURCES

cL- 44

41-

4-

w 43

LU

LU

w 42-

41

40

0 20 40 60 80 100

STACK HEIGHT, M.

Figure X-8.-Effect of Stack Height on Air Concentration of 8 5Kr (Multiple Sources).

X-20

46

I I I I
102

STACK = 10M

1 01

100

x STACKK==150M

U

0

100 1000 10,00

DISTANCE, METERS

Figure X-9.-Effect of Stack Height on Air Concentration of 8 5Kr (Single Source).

X-21
I9

00

C.D

LUC

1W 

I

CONTRIBUTIONS TO CENTROID NO. 1

FROM ALL SOURCES

0 20 40 60 80

PERCENT FREQUENCY

Figure X-10.-Effect of Precipitation Frequency on Air Concentration of 90Sr.

X-22

100

1

10-8 T T T

I I i I
SEQUENCE OF BARS FOR EACH ISOTOPE

LEFT - KD = 0

CENTER - BASE CASE

RIGHT - KD =5000

At

" r

{

x 

I' "r 1

r"

:r, s;

t r {\

FPO

7'. 4K

r;

16 L

.?,rte I

r"

tih<.

, 

x.,

Hy,

9 0 Sr 95Zr

yY

J.

95 Nb

fit

lt

r tyI

~tir

,L1 vY

r'a

I

A.

v ..

V

"1

100

10-1

i IV

is

r

wr~

:ray

Y. 

s

.. 

sa b

. w

a

,j

'.t 1

r

10 6 Ru 1 34 Cs 13 7 Cs

Fiugre X-11.-Effect of Distribution Coefficients, Kd, on Radionuclide Concentrations in

Mississippi River Water (Centroid 87).

X-23

7

K

10-2 I-

-J

LU

0a

0

UOi

L)

10

104

06

I

r

F

C

.r

r

r

r

r

c

r

}

S

r

r

L ! I

LMMMARRIMMINabaw 1 km _n rr6"

IL - --- a A hNNAN l w X& I LNi km6ma 6WAM l L

r

i

f

i

f

t

1

1

S

C

N

6

1

3

1
SEQUENCE OF BARS FOR

LEFT - K = D

CENTER - BASE

RIGHT

C)

2.

vAo

I!

O

1M 

f I

O 

Y C.

V,

1'-T

t"

A. 

46

ti

Y w

'-C

a .+1

r I'f' 

- I..

90Sr

EACH

0

CASE

ISOTOPE

- K = 5000 D

9 5 Zr

r 

I)

-a..

1"

C)

1

r

i

9 5 Nb 1 0 6 Ru 1 3 4 Cs 1 3 7 Cs

Figure X-12.-Effect of Distribution Coefficients, Kd, on Radionuclide Concentrations in

Mississippi River Suspended Sediments (Centroid 87).

X-24

103

10

101

100

D

C)

FFLU

C)

U)

1 10

10 2

10

I

O

1

r.

J

1

C

C

Y

l'

t,

L

E

6

7

r 

r:% w. _ j L i, W

'. 

I

I til 6b" .4

L.r'.til

I

I O)

-I

t\

\

\

\

if\

!\

9

\

\

$0

r

i

F

i

r

Y,

i\

i \

5 \

C\

R
TABLE X-11

Missouri River Basin Annual Flow and

Depletion Rates( 4 3

Year

1860

1890

1910

1940

1949

1970

1980

1990

2000

2020

Depletion

Rate [cfs]

0

1,700

7,500

8,700

9,300

16,000

22,000

25,000

29,000

38,000

Annual Flow

[cfs] [%]

91,000

89,300

83,500

82,300

81,700

75,000

69,000

65,000

62,000

53,000

100

98

92

91

90

82

76

72

68

58

the 1951-1960 mean, will be less than might be

indicated by Table X-11. For example, the mean annual

flow for the year 2000 at St. Louis is expected to be

155,000 cfs-only 4000 cfs less than the 1951-1960

mean. 62,000 cfs of the year 2000 flow will be

contributed by the Missouri River system and 93,000 by

the Upper Mississippi system.

In order to approximate monthly Missouri River

flows in the year 2000 for simulation purposes, the

Corps of Engineers' projections of navigation season

(April-November) and non-navigation season (DecemberMarch) streamflow at Sioux City, Iowa for the year

2000 were used for control with downstream tributary

inflow based on average streamflow adjusted to 1970

depletions. For the Upper Mississippi upstream of the

confluence with the Missouri, 1951-1960 mean flows

were utilized for base case simulations.

For sensitivity studies, two low-flow cases were

examined. One utilized the monthly flow of the year

1956, when the average at St. Louis was 94,000 cfs, and

the other the low extreme of the 1951-1960 monthly

groupings for an annual average of 86,000 cfs. As

expected, the changes in concentration for radionuclides

in solution were inversely proportional to the changes in

flow. A 60% decrease in flow relative to the (average)

base case flows resulted in an overall increase of 70% in

concentrations, with a 100% increase for radionuclides

other than tritium.

FOOD PATHWAY VARIANTS

The potential regional radiation dose from steadystate releases by nuclear facilities is contributed primarily by external exposure pathways and drinking

water, However, in the case of an acute activity release,

the food pathways may become important. As was

discussed previously, the major mechanism for concentrating radionuclides in plants consists of direct

deposition, rather than uptake from soil. However, for

those cases where uptake from soil does contribute

significantly, assumptions as to the potential removal of

radionuclides from availability in the soil become

important. Consideration of acute release incidents or

attainment of decreased exposure through external

pathways may also warrant closer examination of the

effects of varying techniques of food production.

Among these, the techniques of managing dairy herds

are of interest because of their potential effects on dose

to the thyroid.

Depletion of Available Radionuclides

A net reduction in human dose would occur if

mechanisms for removal of radionuclides from availability after deposition in the soil, other than decay,

were involved. Such a mechanism might be downward

migration of radionuclides to a level where they are no

longer available to the plant's roots. Another could be a

chemical or physical change of state which would make

the radionuclide less available to the plant. Either

mechanism postulated has the effect of shortening the

disappearance half-time for the radionuclides. The effect

of the additional mechanism was modeled by a "removal

half-life" in addition to the radioactive decay. The decay

constant used for long-term accumulation is then the

sum of the constants for radioactive decay and for

removal.

Two separate runs were made with the removal

half-life set first to 10 years and then to 15 years. The

observed decrease in the integrated dose and dose

commitment was small, which is consistent with the

observation that the total food pathway contribution to

integrated dose is small. The effect of simulated removal

of radionuclides from the soil can be seen more clearly

by comparing doses from individual food pathways.

TABLE X-12

Effects of Radionuclide Removal from Soil

on Total Body Doses

(Based on Calculations for Centroid 143)

R 

(TL = 15yr)

Fresh Tree Fruit 0.985

Fresh Potatoes 0.970

Processed Potatoes 0.993

Fresh Other Vegetables 0.945

Processed Other Vegetables 0.939

Fresh Milk 0.949

Milk Products 0.951

Butter 0.950

All Others 1.

R* = Total-body dose in sensitivity study

Total-body dose in reference case

R

(TL = 10yr)

0.981

0.970

0.993

0.934

0.926

0.937

0.944

0.943

1.

X-25
TABLE X-13

Ratio of Radiation Doses to the

Average Child Received Under Two Dairy

Management Systems

(Storage Feeding/Forage Feeding)

Organ Total Body

Pathway Fresh Milk

Centroid Milk Products Butter

20

34

68

87

143

166

222

.97

.97

.97

.97

.95

.98

.97

.98

.98

.99

.98

.97

.99

Effects of Dairy Management Techniques

Dairy management techniques can affect the radionuclide content of dairy products. "Storage-feeding" is a

dairy management technique which employs no fresh

forage. The effects of this particular technique on

concentrations of radionuclides in milk products is of

interest since the trend of most large dairies is toward

use of stored feeds. The relative changes in radioactive

dose to the total body and thyroid of an average child

are shown in Table X-13. Generally the total body and

thyroid doses decreased only about 2% and 15%,

respectively, when the storage feeding technique was

simulated in place of forage feeding.

.98

.98

.98

.96

.99

Thyroid

Fresh Milk

Milk Products Butter

.85

.87

.87

.88

.83

.92

.86

.85

.87

.87

.88

.83

.93

.86

.85

.87

.87

.83

.91

Relative Effects of Plant Uptake and

Direct Deposition

To study the relative importance of plant uptake

from soil and direct deposition on plant surfaces as

mechanisms for radionuclide buildup in plants, two cases

were defined. The first case eliminated all uptake from

soil and the second eliminated direct deposition. The

ratio of the dose for each case to the reference dose

reflects the relative importance of each mechanism.

Uptake from soil accounts for only a small fraction of

the radionuclide concentration in food crops. The

majority of dose for these pathways is contributed by

radionuclides directly deposited on the crops, as shown

in Table X-14.

TABLE X-14

Fraction of Total-Body Dose

from Ingestion of Foods Which Results

from Direct Deposition of Radionuclides

Centroid

Food 87 143 238

Fresh Berries

Fresh Treefruit

Fresh Melons

Fresh Potatoes

Processed Potatoes

Fresh Root Vegetables

Processed Root Vegetables

Fresh Green Leafy Vegetables

Processed Green Leafy Vegetables

Fresh Other Vegetables

Processed Other Vegetables

Grain

Wheat

0.842

0.981

0.657

0.967

0.966

0.528 0.719

0.835 0.882

0.923 0.931

0.922 0.950

0.981 0.984

0.717 0.794

0.707 

0.755 0.823

0.748 

0.621 0.633

0.615 0.634

0.972 ' 0.977

0.970 0.975

X-26
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R-1
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R-2
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1965 Illinois Licensed Fishermen, Division of

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in Wisconsin, 1969, Wisconsin Dept. of National

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Radionuclides in Milk, Food and Air," Health

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E. L. Saenger, "Radionuclide Doses to In

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128. P. S. Rohwer and S. V. Kaye, Age-Dependent

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130. Handbook of Respiration, W. B. Saunders Co.,

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132. J. F. Honstead, Recreational Use of the Columbia

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134. G. J. Hine and G. L. Brownell, Radiation Dosimetry, Academic Press, Inc., New York, 1956 (2nd

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136. H. A. Knapp, Gamma Ray Exposure Dose to

Non-Urban Population From the Surface Deposition of Nuclear Test Fallout, USAEC Report

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137. J. M. Ferguson, Ground Roughness Effects for

Fallout-Contaminated Terrain: Comparison of

Measurements and Calculations, U.S. Navy Report

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R-6
APPENDIX A

Glossary of Terms

A-1
APPENDIX A

GLOSSARY OF TERMS

Absorbed Dose: The energy imparted to matter by

ionizing radiation per unit mass of irradiated material at

the place of interest. The unit of absorbed dose is the

rad. One rad equals 100 ergs per gram.

Absorption: (1) The process by which radiation

imparts some or all of its energy to any material through

which it passes. (2) The entrapment of gases or liquids in

the internal structure of solid bodies with which they are

in contact. For example, a sponge absorbs water.

Absorption Coefficient: Fractional decrease in the

intensity of a beam of X or gamma radiation per unit

thickness (linear absorption coefficient), per unit mass

(mass absorption coefficient), or per atom (atomic

absorption coefficient) of absorber, due to deposition of

energy in the absorber.

Activation Product: An unstable isotope of an element formed in a nuclear reactor by processes other

than fission.

Activity: The number of nuclear transformations

occurring in a given quantity of material per unit time.

(See Curie).

Adsorption: The adhesion in an extremely thin layer

of gases, dissolved substances, or liquids to the surfaces

of solid bodies which they contact

Air Envelope: A geographic band approximately 200

miles in width around the perimeter of the Upper

Mississippi-Lower Missouri River Basin region. Radioactivity released in the zone is allowed to travel into the

study area via air transport to minimize isolation effects

on the regional dose calculation.

Atmosphere Stability: A characterization scheme that

describes the turbulent nature of the atmosphere (in this

report specifically the vertical component of turbulence)

within certain preset criteria.

Aquifer: A geological formation containing groundwater.

Average Adult (teen, child): A person whose dietary

habits represent the average of his age group at his

centroid of residence and whose recreation habits

similarly represent the average of the habits for his age

group at his centroid.

Background: Radioactivity found in the environment

which emanates from sources other than nuclear reactors

or fuel reprocessing plants.

Base-Loaded Plant: A power plant operated at a high

plant factor for much of its life to meet the relatively

constant portion of the energy demand. In the economic

model two operating curves are available for the linear

program to select, one called the base-loaded curve and

one called the load-follower curve. The plants required

to optimally meet the region's requirements consist of a

mixture of these types.

Beta Particle: Charged particle emitted from the

nucleus of an atom, with a mass and charge equal in

magnitude to that of the electron.

Boating: Recreational activity that includes water

skiing, sailing, canoeing, boating, and fishing from a boat

or barge.

Body Burden: A particular amount of radioactivity

retained within the body of an individual.

BWR (Boiling Water Reactor): A water-cooled nuclear reactor in which steam (for turbine operation) is

produced by boiling of the coolant within the reactor

vessel.

BWRA: An advanced BWR (as used in this study)

featuring increased linear heat rate.

Centroid: A designation used in this evaluation

whereby a county or group of counties is represented by

a single point. The entire population of the area and its

characteristics are represented by typical individuals and

their activities referenced to that point.

Close-In Dose: The radiation dose received by people

living within a few miles of a nuclear facility.

Cold Trapping System: The coolant cleanup system

assumed used in reducing radionuclide concentrations in

LMFBR systems. Such a system removes impurities from

the sodium coolant by reducing the temperature of a

small stream of coolant. Impurities less soluble at the

lower temperature are thus removed.

Curie: A unit of radioactivity. One curie equals

3.700 X 1010 nuclear transformations per second.

(Abbreviated Ci.) Several fractions of the curie are in

common usage.

Microcurie: One-millionth of a curie (3.7 X 104

disintegrations per sec.) (Abbreviated

iCi.)

Millicurie: One-thousandth of a curie (3.7 X 107

disintegrations per sec.) (Abbreviated

mCi).

Picocurie: One-millionth of a microcurie (3.7 X

10-2 disintegrations per second or

2.22 disintegrations per minute).

(Abbreviated pCi; replaces the term

Daughter: Synonym for decay product.

Decay Radioactive: Disintegration of the nucleus of

an unstable nuclide by spontaneous emission of charged

particles and/or photons.

A-2
Decay Constant: (See Disintegration Constant.)

Decay Product: A nuclide resulting from the radioactive disintegration of a radionuclide, formed either

directly or as the result of successive transformations in

a radioactive series. A decay product may be either

radioactive or stable.

Decontamination Factor: The ratio of the amount of

undesired radioactive material initially present to the

amount remaining after a suitable processing step has

been completed. Decontamination factors may refer to

the reduction of some particular species of radionuclide,

or to the gross measurable radioactivity.

Disintegration Constant: The number which characterizes the rate of decay of a radionuclide X in the

equation N = Neest, where No is the initial number of

atoms present, and N is the number of atoms present

after some time, t.

Disintegration Nuclear: A spontaneous nuclear transformation (radioactivity) characterized by the emission

of energy and/or mass from the nucleus. When numbers

of nuclei are involved, the process is characterized by a

definite half-life.

Distribution Coefficient: For any dissolved substance,

the ratio of concentrations adsorbed on soils or suspended sediments in contact with water to the concentration of the substance in solution.

Dose: A general term denoting the quantity of

radiation or energy absorbed. For special purposes it

may be appropriately qualified. In this report, it is

generally used synonomously with Dose Equivalent.

Dose Commitment: The dose received during a

specified time (e.g., 50 years) following intake of a

quantity of a radionuclide, caused by retention of that

radionuclide in the body. As used in this report, dose

commitment is calculated for radionuclide intake in the

year 2000 only, and does not include previous or

subsequent intake of radionuclides.

Dose Commitment Factor: A number which can be

used to calculate the radiation dose received over a long

period of time from an intake of a radionuclide. In this

report, the dose commitment factor has units of

mrem/50 yrs per pCi/yr intake and the dose commitment calculated is the dose received in the ensuing 50

years from the first year's intake. No external dose

commitment factors have been used in this report.

Dose Equivalent: A quantity used in radiation protection. It expresses all radiations on a common scale for

calculating the effective absorbed dose. It is defined as

the product of the absorbed dose in rads and certain

modifying factors denoting the biological effectiveness

for the particular type and energy of radiation involved.

(The unit of dose equivalent is the rem.)

Dose Factor: A number which can be used to

calculate radiation dose to a person (or organ) from the

intake of radionuclide (internal dose factor) or from

exposure to given concentration of radionuclides in

environmental media (external dose factor). In this

report, internal dose factors have units of mrem/yr per

unit intake of radionuclide and the dose calculated

represents the dose in the first year from the first year

intake. External dose factors have units of mrem/unit

concentration of a radionuclide in air (pCi/m 3 ), water

(pCi/liter), or on the ground (pCi/m 2 ). The value

calculated represents dose rates received during time of

exposure.

Dose Rate: Absorbed dose delivered per unit time.

Dose rates in this study refer to the summation of 12

monthly values made up of (1) an external exposure

contribution from radionuclide concentrations in the

environment during each month of the year and (2) an

internal dose calculated as the dose contribution, over

the following 12-month period, from radionuclides

taken into the body during each month of the year.

Dry Deposition: The depletion of an airborne contaminant plume by interaction process at the ground-air

interface (i.e., gravitational settling, impaction, chemical

interaction, etc., with surface-based material).

Effective Energy: In this report, a term used to

denote the actual energy deposited at the point of

interest per disintegration of a nuclide. It is found by

summing the products obtained by multiplying the

average energy of the beta particle (or gamma photon)

emitted, the fraction of disintegrations which give rise to

that particular beta particle (or photon), and the

fraction of the initially released energy which is absorbed at the place of interest.

Effective Radius: The radius of a hypothetical spherical organ, used for calculational simplicity, in which the

absorbed fraction of the gamma rays emitted from a

point source in the center is the same as the absorbed

fraction of the gamma rays emitted by the same

quantity of a radionuclide uniformly distributed

throughout the volume of the real organ. For such a

sphere, the absorbed fraction is equal to (1-e-Mx) where

x is the effective radius in cm, and 1 is the Absorption

Coefficient in cm-1

Equilibrium, Radioactive: In a radioactive series, the

state which prevails when the ratio between the amounts

of successive members of the series remains constant.

Exposure: A measure of the ionization produced in

air by X or gamma radiation. It is the sum of the

electrical charges on all ions of one sign produced in air

when all electrons liberated by photons in a volume

element of air are completely stopped in air, divided by

the mass of the air in the volume element. The special

unit of exposure is the roentgen. In this report, exposure

has been used synonomously with dose and dose

equivalent, especially when received from external

sources.

Exposure Pathway: The means by which a person

becomes exposed to, and receives a radiation dose from,

radionuclides released to the environment. Generally,

the pathway followed by the radionuclide between the

time of release and the time of exposure.

Exposure Rate: Exposure delivered per unit time.

External Dose: A radiation dose received by a person

(or organ) which is delivered by radiation originating

outside of the person.

Fishing: Recreational activity that includes, in this

study, fishing from the bank only.

A-3
Fission, Nuclear: A nuclear transformation characterized by the splitting of a nucleus into at least two

other nuclei and the release of a relatively large amount

of energy.

Fission Product: An unstable isotope formed during

the fissioning process.

Fission Yield: The percentage of fissions leading to a

particular nuclide.

Food Deficit: Excess of food consumption over

production for a given food type (within a centroid).

Food Surplus: Excess of food production over

consumption for a given food type (within a centroid).

Fuel Burnup: A measure of the degree to which a

nuclear fuel is consumed before replacement with new

fuel, commonly expressed in megawatt days per metric

ton of fuel.

Gamma Ray: Short wave length electromagnetic

radiation of nuclear origin (range of energy from 10 KeV

to 9 MeV) emitted from the nucleus.

Generating Plant Mix: The mix of nuclear and fossil

plants which produce electric power for the study area.

Gigawatt-Electrical (GWe): One billion watts of electrical energy (often used as a unit of electric generating

capacity).

Groundwater: Water residing below the surface of the

land in formations saturated with water. The "water

table" is the upper surface of groundwater.

Half-Life, Biological: The time required for the body

to eliminate one-half of an administered dosage of any

substance by regular processes of elimination. Approximately the same for both stable and radioactive isotopes

of a particular element.

Half-Life, Effective: Time required for the radioactive

element in the body to be diminished 50 percent as a

result of the combined action of radioactive decay and

biological elimination.

(Biological half-life)

EaX (Radioactive half-life) Effective Half-Life (Biological half-life)

+ (Radioactive half-life)

Half-Life, Radioactive: Time required for a radioactive substance to lose 50 percent of its activity by

decay. Each radionuclide has a unique half-life.

Induced Radioactivity: Radioactivity produced in a

substance after bombardment with neutrons or other

particles.

Integrated-Dose: The summation of all the products

of each particular dose level times the number of people

exposed at each particular dose level.

Internal Dose: A dose received by a person (or organ)

from radiation originating from material contained

within the person.

Ion: Atomic particle, atom, or chemical radical

bearing an electrical charge, either negative or positive.

Ion Exchange: A chemical process involving the

reversible interchange of ions between a solution and a

particular solid material such as an ion exchange resin

consisting of a matrix of insoluble material interspersed

with fixed ions of opposite charge.

Ionization: The process by which a neutral atom or

molecule acquires a positive or negative charge. 

(LWR) Light Water Reactor: A general designation

applied to thermal reactors using normal water as a

coolant and moderator, and including both pressurized

(PWR) and boiling (BWR) water reactors.

Linear Heat Rate: The quantity of heat produced in a

unit length of reactor fuel rod.

Linear Programming: A mathematical technique that

finds the solution to a set of simultaneous linear

equations which results in the maximum (or minimum)

value of a particular linear function which is designated

as the objective function.

LMFBR (Liquid Metal Fast Breeder Reactor): A

nuclear reactor type, using liquid metal coolant (usually

sodium) designed to operate with an unmoderated

("fast") neutron spectrum and to "breed" more nuclear

fuel materials than it consumes.

Living Pattern: Observable features that characterize

an individual, i.e., where he lives, water he drinks, food

he eats, air he breathes and his recreational habits.

Load-Following Plant: A power plant whose output is

varied to enable a power system to meet changes in

demand.

Maximum Adult (teen, child): A person whose

dietary and recreation habits tend to maximize the

radiation dose which he (or she) receives. In this report,

the maximum individual is assumed to reside at the same

location within the centroid as does the average person.

For this reason, doses higher than those calculated for

the maximum person can be postulated, e.g., the air

submersion dose from full-time residence close to the

site boundary of a nuclear facility.

Man-Rem Dose: A number expressing the total

integrated total-body dose received by a large population. It is obtained by multiplying the average dose

received by a given subgroup of the population by the

number of persons in that subgroup and then summing

over all subgroups (see Integrated Dose).

Megawatt-Electrical (MWe): One million watts of

electrical energy (often used as an electric generating

station capacity unit).

MFR: Midwest Fuel Recovery plant, a fuel reprocessing plant near Morris, Illinois.

Micron.: Unit of length equal to 10-6 meters.

(Symbol: p or more recently pm).

Millirem (mrem): A submultiple of the Rem, equal to

one one-thousandth of a Rem. (See Rem).

Mixing Depth: The depth of that portion of the

atmosphere immediately contiguous with the earth in

which vertical motion is not significantly inhibited.

Monitoring: Periodic or continuous determination of

the amount of ionizing radiation or radioactive contamination present in an occupied region.

Noble gas: A member of the family of inert gases

(helium and its chemical homologues), whose orbital

electron configurations cause them to have little tendency to react chemically with other materials. For most

A4
purposes they can be assumed to be totally non-reactive

or inert chemically, although they can be affected by

essentially physical processes (e.g., adsorption).

Nuclide: A species of atom characterized by the

constitution of its nucleus. The nuclear constitution is

specified by the number of protons (Z), number of

neutrons (N), and energy content; or, alternatively, by

the atomic number (Z), mass number A = (N + Z), and

atomic mass. To be regarded as a distinct nuclide, the

atom must be capable of existing for a measurable time.

Thus, nuclear isomers are separate nuclides, whereas

promptly decaying excited nuclear states and unstable

intermediates in nuclear reactions are not so considered.

Organ: Group of tissues which together perform one

or more definite functions in a living body.

Parent: A radionuclide which, upon disintegration,

yields a specified nuclide-either directly or as a later

member of a radioactive series. (See also Daughter).

Pathway: An environmental mechanism or series of

mechanisms by which radionuclides are transported

from a source to a receptor.

Plant Factor: The ratio of the actual energy output in

the period of time considered to the energy output

which would have occurred if the plant had been

operating at its full rating throughout the total time in

the period.

Plant Factor Curve: A curve defining the plant factor

of a plant throughout its expected lifetime.

Plant Ratio: The ratio of base-loaded plant capacity

to the total plant capacity built in a specified period of

time.

Poison: Material of high absorption cross section

which absorbs neutrons unproductively and reduces the

reactivity of a reactor.

PWR (Pressurized Water Reactor): A water-cooled

reactor in which the coolant is pressurized to the extent

that boiling in the reactor core is inhibited. Steam is

generated in a secondary system, using a separate heat

exchanger.

PWRA: An advanced PWR (as used in the study)

featuring increased linear heat rate.

Rad: The unit of absorbed dose equal to 0.01

Joule/Kg (100 ergs per gram) in any medium (see

Absorbed Dose). (Written: rad.)

Radionuclide: Any isotope (of one of the chemical

elements) which undergoes radioactive decay.

Radwaste: Radioactive waste.

Radwaste System: A combination of process equipment components used in a nuclear facility to clean up

the reactor coolant and other liquid and gas streams

before reuse, storage, or discharge.

Receptor: An individual for whom a radiation dose is

calculated.

Rem (roentgen equivalent man): A biological dose

unit defined by the amount of energy absorbed from a

radiation source times the relative biological effectiveness of the particular type of radiation.

Release Fraction: The fraction of a fission product

generated in the fuel which is released from the fuel

matrix. This fraction is released to the coolant in failed

fuel rods.

Reprocessing Plant: Chemical processing plant which

separates useful isotopes from waste products in spent

nuclear fuel.

Roentgen (R): The special unit of exposure. One

roentgen equals 2.58 X 104 coulomb per kilogram of

air. (See Exposure.)

Rural Non-Farm Area: A centroid designation used

for areas outside of urban localities but without a farm

operator.

Rural Area: A centroid designation used for areas

outside of urban localities with a farm operator.

Sensitivity Study: The part of the overall examination of a technical problem where evaluations are made

of the effect of changes in particular parameters of

interest on the behavior of the system being examined.

Source: A point of discharge for radioactive materials

to the environment.

Source Map: A compilation of all regional sources of

radioactive releases from nuclear facilities.

Spent Fuel: The depleted nuclear fuel which must be

discharged from nuclear reactors periodically to be

replaced by fresh fuel.

Surface Water: Water on the surface of the land.

Total Body: The whole body taken as an organ for

purposes of computing a radiation dose.

Transpiration: The absorption of water vapor through

the skin; thus, tritium transpiration is the absorption of

tritiated water vapor (HTO).

Transuranium Isotope: An isotope with an atomic

number greater than that of uranium.

Turbulent Diffusion: The diffusion of an airborne

contaminant by eddies in a turbulent flow.

Urban Area: Localities with at least 2500 inhabitants

and in closely settled fringe areas surrounding cities of

50,000 or more inhabitants.

Voloxidizer: A system to oxidize the volatile components of spent fuel prior to dissolution in a reprocessing plant to permit more effective trapping of the

volatile effluents.

Wet Deposition: The depletion of an airborne contaminant plume by precipitation (rainfall, etc.).

A-5

APPENDIX B

Study Area Characteristics

B-1
APPENDIX B

Study Area Characteristics

A large amount of data has been accumulated to

characterize the region and the people who live in it.

These characteristics cover the anticipated power development in the area, the hydrology and meteorology of

the area, the food production characteristics of the area,

Figure No.

B-1

B-2

B-3

B-4

B-5

B-6

B-7 - B-9

B-10 - B-13

B-14

B-15

B-16

B-17-B-18

and the food consumption, living patterns, and distribution of the people in the area. Chapter III discusses such

regional characteristics from the regional modeling viewpoint.

LIST OF FIGURES

Title Page

Average Annual Precipitation Runoff in Study Area........................B-4

Average Suspended Sediment Concentration in Rivers..................... .. B-5

Consolidated Rock Aquifers in Study Area............................. B-6

Unconsolidated Aquifers in Study Area............................... B-7

Normal Annual Precipitation in Study Area............................ .. B-8

Normal Monthly Distribution of Precipitation .......................... B-9

Probability of Receiving Precipitation, 0.1, 1.0, 2.0 Inches or More . . . . . . . . . . . . B-10 - B-12

Wind Rose Patterns, January, April, July, October ....................... . B-13 - B-16

Predicted Seasonal Variations in Power Generation (2000 A.D.).............. .. B-19

Assignment of Typical Demand Curves of Power Supply Areas (2000 A.D.)...... .. B-20

Location of Nuclear Facilities in the Base Case Configuration................ .. B-22

Location of Nuclear Facilities in the 1 and 5 MT/Day Clustered Configuration.......B-23 - B-24

LIST OF TABLES

Table

Number

B-i

B-2

B-3

B-4

B-5

Title

Water Transport Distribution Coefficients Used in Study ........................

Annual Energy Demand Summary: Projection for Extended Study Area, 1970-2000.... 

Food Types Considered in Study..........................................

Weekly Food Consumption Per Household...................................

Irrigated Food Production.............................................

Page

B-17

B-25

B-27

B-28

B-29

B-2
Hydrology and Meteorology

of Study Area

B-3
4 5 6 7

8

9

10

2 3 t 1

QS 

T1215 15 14

nv.s ,,, V 13

Sw'T Na Lt

NT ' At ii ,C

Ci.. Minnea S 12 a[ F xt. L.ols P , lcl lsr[0 fuc ,,h. lire *" . u ing a crves, # . t 4 ]

lp C11111g Io Cr'1

r rlrt- r oiro

... A" -R ter n 9 L ~" "oc ROE Iac '" '"'Met 'alsoa " t ha

-- R10 7

R 11Rht r t OSCIora T NOSSYIO NME- WORT IC I A

sloesoi

. n o a .~I

cggs .. Beloi

nta ~ ~ ~ y r1[R wa onl waeo u ias aOlis ^ne a"' I am

sNler CI *" ~. K Rock oslan $' . ms

- co rcs aa tne "I onox saa a a aso

N.-. - - w Cit Daeno

,.T~ . ., ... O "u *..N o Dic

R ~ ~ ~ ~ ~ ~ ~ ~ e Tu 3 C ITwt okIl n YeOd i

v~R ...,B ni A,1 1 20111

Q aa oumbi n., s

a ~ ~~~~ us CitGae eri lt.

curo nco

7 f. . " 15o

10 1.1

r ..T YI iY. "SO l 0[121 .. q 1S SCNT~t11

Fiur B-.AeaeAnaYrcptto uofi td ra

B-4
SEDIMENT CONCENTRATION

PARTS PER MILLION

6500 TO 15,000

2000 TO 6500

®280 TO 2000

LESS THAN 280

.~ r1

C yCf~

-_ j.... o :6i

m a ws Hftiyri fj, _" n

L4 'tck M#1and 1a

- ": -!( ':- l- M a -c n*

/~~an L d okIln

-- uo ::pn

Ln n .01tu., 'u c1o " aa

oovms . -lG lsbr

.11 SSli

oseph ecs : r1Rt Noeee

F"gur B-.-veag Supne SeietCnetrto nR r.

B-5
CONSOLIDATED ROCK AQUIFERS

C

U

S

L

C

Figure B-3.-Consolidated Rock Aquifers in Study Area.

B-6

COLOED PTTENS IDICTE AEASSANDSTONE: INCLUDES SOME

OLORD PTTERS IDICAE AEASUNCONSOLIDATED SAND UNDERLAIN BY ONE OR MORE AQUIFERS

GENERALLY CAPABLE OF YIELDING] CARBONATE ROCK: LIMESTONE

TO A WELL AT LEAST 50 GALLONS QAND DOLOMITE

PER MINUTE (OR 3-1/2 LITERS PER o ADTN N ABNT OK SECOND) OF FRESH WATER (GENERALLY ISADTN ADCRB AEROK

LESS THAN 1000 PARTS PER MILLION OTE alClow OTTER TAIL lg OF DISSOLVED SOLIDS.

too w

a losa

(O tac anCE

SWIFT w \ A LR LANGLA

Ill ruoo ' ic : rt ausau

fai l - -,

\ r . \

\ \ 000 WAYw .l~lyi. V, a ~ft .?"> \ \

004Triwie { CIl3tt . W ..... I$W LOI\ t AIY,'A(*tOrl: 

c La Fa.. ls

Co.ro on pri fe

AATIL ~~ I- Lit \ MOA,...: .. r uu an fhdepne Co'lu. 1t[S, R r al"

Kt~ ~ ~ ~ ~ ~ ~ ~ ~~t .. c.,.i L. t toTOd:. . :.

'"" S --s: -- O earnr \\ . Loui YU

sow~~e - , Be l crvirr [ ao[Myts h..ry N tvijb ?"1

IPK 0 1 , VI~ l . " .
CO

UN

GE

TO

PE

SE

LE

DI

B-7

LORED PATTERNS INDICATE AREAS UNCONSOLIDATED AQUIFER

DERLAIN BY ONE OR MORE AQUIFERS NERALLY CAPABLE OF YIELDING A WELL AT LEAST 50 GALLONS

R MINUTE (OR 3-1/2 LITERS PER

COND) OF FRESH WATER (GENjERALLY

SS THAN 1000 PARTS PER MILLION OF

SSOLVED SOLIDS.

Sro AI- *IA ICL I AUS

I 4 G ( :" M i o w StT . D a v e n p o lf . :A."LL

OEUE -t - o iAT-.r - V o a N A ,, '. -t --.- :i' u l o - -u c -w-

" ':>A',: NsRfi "IAIot s AAI

;:.~~ ".- ----- l ~ s w - -o a -P --

DEUEL o- Rlaul - CLA -f -~ na st-W u ca s ar t r

c Rochester W Wnon sA

AoR NR ~ AI A AAA LAIY LUCAS ro ( IIAS SI."".LU SI

Ottu1 RIC PALA

A&/ IAALU T rIRI NOT IC S1in E Al " AEA

/ ."::: .: :.

K::L:" C ::SAE R : * ALn Cit .A L S A ECA ml

IO C .'ck- .ILR n N POIIA U CA

IS/ OWSllr~i afN R~

LAS AU.LL' J.4(ansas 

- la - - * * c. c wNA -A[o.B li

WAECUSE LUIL *AAf OI ~ ortASU Dodg Ep ba sCALLAAS :

VL- -A V ERS BY STREA 'UEdS

- -', -CD -ln 

SAND-ANDGRAVEL:DNdINTER

::.8 ". :. ; : : OQ : :."' T pw . :.. s - s N rs -: Dav npor - - -- --

:-:+. VAL ... ABANDONED OR

-' it CLAI"I LUCAS .o" - ~ lR N :.No ST- r -- lm~afto --- ALU A VALY ALV L - - .I-AS- >:;- -- --- -~O - - t w um N .nu Q an Gakesbu -N-

-- :'::.:::- - . :: . O ""

- , .. ,,p . ... - -- : r ' :."o ril 

.h - .w -..t ..M - -. -.-.u - - co,oY -... - - Springfield':: lt -- : - Arcel . N . .. -s . : R" fYsnl s

- K -:("" :":Kansas 

WATERCOURSE: ALLUVIAL UTA " AND ICEVALLEY TRAVERSED BY STREAM T E E I FROM WHICH RECHARGE CAN BE 'uBn n11 " -is.iS Ae t

INDUCED -t~ . ALI. 

SAND AND GRAVEL: IN INTER- "" . - .a Cit t

MONTANE VALLEYS; ABANDONED OR 04 c of r"" w". K

BURIED ALLUVIAL VALLEYS; ALLUVIAL soa.oR r(RROR Nlcsa,

TERRACES; SAND DUNES; COASTAL PLAINS AND u« --fP Q 

GREAT PLAINES; GLACIAL OUTWASH AND ICE- 'ot" Act

CONTACT DEPOSITS OF GLACIATED REGIONS .m -.-

Figure B-4.-Unconsolidated Aquifers in Study Area.
& 24

O AI 28

cnws OTTER TAIL wI MG

DD Z; IM32 36 3

ILA ANT DOUGLAS 1 at cli - RA L 

0 Q -E *""

SION( loud- URI RARO Rano us ainc

RANT SWIFT ANDIRON Ro ANK TAYLOR LANGLA

NO HDut ,F St. Loui Pr.( A MRAN

IN lll ""'(0 .,Vlar W ausau,

ral~ vitL DICIN I rCE[ AllA

RPEP f O T

A..N.. A E -G OONN .YO SI- RPhs Wioays cs~ reco ' 

NlCOSII Suit FaIsACO au

AUDA/i \ SnCOy RE

2SO "'OK rl w o' ' o ch s e W inna Ros 

SioLu W~c D A CATIN FAIAUU rtO ' lim it FOt to Lauroe OWs s

s'''t Citt Fal vtrtO Dodg I"t o o s XMnr N-T ,--O Roct -r IrORf "" a OSrt C fI N.m RICNL s A ND0

MO~Cit Cit 0 1r

c[DcR F La Y. 1p Tcm _ "U

ITru rouol u~ to ssc Cass Ita - ?io5=a s on 2 0ca ' IT ao 1A ~Ok slA nd 4 ' - Ta

s N Jas. g s U'0 CaE tcsa nesill 32 AM ~~ ~ ~ ~Otu Bu n 

D A A OM" NN (11 ... ~ [)I ANt Gtr I orts Dog IN Ca wa -I oo' aS t S bu o O l, I

wrt sr Ro.c ford

I"11Ct Ci,. -6 tC RUR(AU Ru I Aua TA i CS DIR MAIONr. " Ipr o n gfe

tlw l .rsG! c k Isa d I Q ' TARA r[1tICRasso .a C 

Q~ Li -- TA LR a "r *oD r 4 0B 9 ( a e b r

es tons '-so s ;

I6 Os -3- .. ,., n . fu,,,..s

cE ,+I tUrl- \ Sw IG

[t L Ilru r " F1N I g r B- . N r al An u Pr cp t to . .Qinc y rADc

B-8
J F M A M J J A S O N D

KANSAS CITY

KANSAS, MISSOURI

J F M JTHADSOTN D

HURON, SOUTH DAKOTA

JF M A M J A SON D

LINCOLN

NEBRASKA

J F M A MJ J A SO0N D

SPRINGFIELD, ILLINOIS

J F MA Mi J A S O N D

ST. LOUIS, MISSOURI

J F MA MI A SON D

MINNEAPOLIS, MINNESOTA

J F MA MI EJA S 0 N D

DEMO INE S, IOWA

5

4

3

2

0

JiF M A M JI A S ON D

DUBUQUE, IOWA

Figure B-6.-Normal Monthly Distribution of Precipitation.

B-9

5

4

3

7

0
r- -40 _ _ 50 60

- - JAN 10-16

70

--

CI - - 80

20> 8

j --- - - 30 40 '50' 60 ----- --- 7

-- U 

'N1-

r-- 60 70 APR 26-MAY 2

- -- --- 7 ------- 0

60 70

60

70 -- - --- 80

I J I

\ l

----- 7--------

- -- 60- -

6.. 707

40

-. 60 MAR 15-21

70

I (.

808

40 

90

--- _ Io I c i

S90090

404 50 60 70 - -- -

1 I o- -- 

r -" - -- - - JUNE 14-20

9 80

-- 70 90 ---

890

j \" i F . i 9

40

400

-

-- 6 ""..-- ------ " 0 

Figure B-7.-Probability of Receiving Precipitation, 0.1 Inch or More.

B-10
- -- - - JAN 10-16 k 1 ^. 05 v

10 10 2I 

i \

01

~ 20 -303~ 

r - .- .. 10 - 20 APR 26-MAY 2

1 - ---. Is

20

~ 

30

40 20 -- - 30 -------4--- _ - ------- - J

r- -- _ 10 20 I "-- - JULY 26-AUG 1

10'

- - - -- 20

-- 20 3

-- -- - - -- - 20 \

I 20 

30

i -05

S- -- . 10 MAR 15-21

20

- . - 

\ 40

10 20

r---- 30

4 0 4 JUNE 14-20

I I 40 40

-- -- 30

zo -- 20-

S----- 30

40

-20 - 30 40 ~ 40 ~

r-- _..-10 20

- - SEPT 13-19

20

- .A

10\

j l 3

Figure B-8.-Probability of Receiving Precipitation, 1.0 Inch or More.

B-11
- JAN 10-16

L 

01 ------- 1

.05 ~01

01

- .. -_--- ---------

L - ----

I--- 05

I --

I'-- -- 

10 -_ _- 10 I \.

___ 10~ 1

----- MAR 15-21

- 

LESS THAN 01

20 r

j------- - 0

9-5------ -

0 - - JUNE 14-20

~ -0

05-

- -10'

020

j O5 SEPT 13-19

i 05

05

------ t- - - ._ __-

Figure B-9.-Probability of Receiving Precipitation, 2.0 Inches or More.

B-12
SEA LE

POKANE 'LUMARCK BIGSEATFAL S A 1T

QEDFORD y RAPID MINEAPOLI GREA ME' R I1J cq '~ CITY -ST. PAUL AL AN

CASSPER E CHCAGO 0 5 A

RENO SALT LAKE CITY N RTH PLATTE 4WSIGTON

ST LOUIS 1020

SAN o , OUISVIL.E NOFL

FRANCISCO , FRESNO 002 KANHSAATT

10A0 ER '%V WICHITA CI. L LL E A EG H

LSANGELES AMARILLON

ALBUQUERQUE , OKLAHOMA C

131 ,, PH1 N 2 CITY MEMPH I A IN C AR LSTON

DALLAS 2 JACKSON 2KSO

EW ORLEANS * TLAIS

AUSTIN ' WOEN 'TALLAHASSE

I GALVESTON2

1020

BROWN SV ILLE

Figure B-10.-Wind Rose Patterns, January.
SEA Tf 

HIGH 8 GRE AT FALLS BISARC SPUTS ARI L1GO 0

CADUUTER DES OP'l

MED 80,

v AS\GTON

CASPER DES

RENO SALT LAKE CITY MOINES CIAO CLMU

SAN DENVER SLLAENORTH o PLATTE ST. ST/OI LOUIS

FRNIC RSO2LOU ISV ILLE NORFOLK

WICHITA CSAS

LOS ANGELES AMA~il0KNOXVILLE GH

ALBUQUERQUE 

j TARK AH M

E AAMEMPHISNTA CHARLESTON

DALLAS ' JACKSON ! 9 MONTGOMERY

EL PASO 2

ow AUSTIN NEW ORLEANS 9

AULESTN- TALLAHASSEE

'GALVESTONBROWNSVILLE

Figure B-11.-Wind Rose Patterns, April.
SEA T7LE

SPOKANE 1014 BISMARCK SpL T A 0, URITOP jv

GREAT FALLS DULUTH 

MEDFORDSE RAPID MINEAPOLIS B O\/ CITY -ST. PAUL At~

GPER o MOINES CHICAGO C OL (0U

RENO SALT LAKE CIY 3 NORTH PLA s \S

s 01 DENVER ,5. LSA\ LOUlSV tEFOL

FRANCIS , RE 0 2 C A/NIHIA SAS

101oWICHITA CIY2

KNOXV LER G

LOS ANGELE AMARILLO KLAHOMA <S

ALUUQE 22' CITY MEMPHIS T CHRLESTON

%PHOEN\ IY ~ p~

Low S DALLA JACKSO ' ONTGOMERY JAC NL

EL PASO I 0

NEW ORLEAN , E

AUSTIN T 

- 5

s

BROWN SV ILLE

Figure B-12.-Wind Rose Patterns, July.
SEA L

SPOKANE : S ASMARCK 1 SO 0

aa~aALL S ST BISMDFAK DULUTH 50

DFg 0 EPD MN PL S GC G0O

R BUNES pL BAOSTON10

SA 16DENVER 2 syLE2 0ft

FRANCISCO FREGR A/

WICHITA ' N C XIC tE O \ G N

AMAILL SATLK IK OT LA WAS

LOS ANGES ALBUUERUE A L 

H ENEE' OROx

1 0 1 4D A L L A S jA C K S W IGHI T

BROWNSVILKE

Figure B-13.-Wind Rose Patterns, October.
TABLE B-1

Water Transport Distribution Coefficients*

Used in Study

Values in (pCi/kg sediment)/

(pCi/liter solution)

Isotope Coefficient

H-3 0.

C-14 10.0

Na-22, 24 0.

Cr-51 35.0

Mn-54 59.0

Fe-55, 59 1050.

Co-58, 60 3057.

Ni-63 30.

Cu-64 15.

Zn-65 108.

Sr-89, 90 1050.

Zr-95 52.

Nb-95 108.

Mo-99 30.

R u-103, 106 1050.

Te-132 43.

1-129, 131, 132, 135 0.

Cs-134, 137 241.

Ba-140 28.

La-140 151.

Ce-141, 144 305.

The relationship between concentrations of dissolved radionuclides and the extent of their adsorption on sediments is

expressed through the distribution coefficient, K 0 . This coefficient defines the ratio of radionuclides adsorbed per unit

weight of sediment to the quantity remaining in solution per

unit volume of solution.

B-17
Electrical Energy

and

Power Demand Projections

B-18
12

11 

0

8 -o

= SUMMER PEAK

6--

---.-- SUMMER INTERMEDIATE

WINTER INTERMEDIATE

5--WINTER PEAK

0

JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC

Figure B-14.-Predicted Seasonal Variations in Power Generation (2000 A.D.).
26

-- 16

11

-15

r 

'~~l SUME INEMDAT

WITE IN ER EDAT . - - - --

... l

, --- /t tt I ~y. . "! at 7 .;' ; 

Figure B-15.-Assignment of Typical Demand Curves of Power Supply Areas (2000 A.D.).
A sensitivity study, described in Chapter X, considered potential patterns of clustered plant configurations.

These cases had various reprocessing plant sizes with

power plants clustered at the same site consistent with

these reprocessing plant capacities. The following three

maps show the plant sites used in considering the effect

of clustering power plants.

B-21
NORTH DAKOTA

MINNESOT A

0

1

SOUTH DAKOTA

S1

IOWA

NEBRASKA

Q0

KANSAS

WER PLANT

MT REPROCESSING PLANT

MT REPROCESSING PLANT (MFR)"0

0

Figure B-16.-Location of Nuclear Facilities in the Base Case Configuration.

B-22

" P0

5

1

A11
NORTH DAKOTA

MINNESOTA

WISCONSIN

SOUTH DAKOTA

0

IOWA

NEBRASKA

- .ILLINOIS - .. _.. 

MISSOURI

* NON-CLUSTERED ITE

o CLUSTERED SITE KANSAS

CLUSTERED SITE WITH

1 MT REPROCESSING PLANT 0 0

A MFR PLANT (1 MT/DAY)

0

Figure B-17.-Location of Nuclear Facilities in the 1 MT/Day Clustered Configuration.

B-23
NORTH DAKOTA

MINNESOTA

WISCONSI

SOUTH DAKOTA

IOWA

NEBRASKA

ILLINOIS

* NON-CLUSTERED SITE MISSOURI

o CLUSTERED SITE KANSAS

CLUSTERED SITE WITH

5 MT REPROCESSING PLANT

MFR PLANT (1 MT/DAY)

Figure B-18.-Location of Nuclear Facilities in the 5 MT/Day Clustered Configuration.

B-24
TABLE B-2

Annual Energy Demand Summary: Projection for

Extended Study Area, 1970-2000

Annual Demand, Millions of Kilowatt-Hours

1970

23,874

56,210

63,120

27,400

52,300

20,400

23,800

28,980

4,047

42,392

5,640

4,300

4,800

9,100

6,640

720

18,150

11,200

19,000

422,073

1980

45,380

102,000

123,000

50,500

105,000

43,800

52,800

55,680

6,867

76,409

12,080

8,400

9,500

18,100

12,020

1,585

41,912

23,000

38,000

826,033

1990

86,921

181,003

233,204

91,302

195,003

85,101

105,002

108,682

12,627

101,494

22,454

16,100

18,300

33,301

20,730

3,189

80,411

42,901

73,301

1,511,027

*For those power supply areas marked

lying within the extended study area.

(*), the demand shown is that of the portion of the PSA

B-25

PSA*

9*

11

12

13

14

15

16

17

19*

20*

25*

26

27

28

29*

31*

33*

34

40

Area Total

2000

142,651

285,158

380,634

145,280

314,780

140,737

175,827

180,176

20,275

137,123

36,264

26,432

30,125

53,485

32,166

5,374

132,083

69,413

120,754

2,428,736
Food Production

and

Food Consumption Patterns

B-26
TABLE B-3

Food Types Considered in Study

Strawberries

Other Berries

Grapes

Plums

Pears

Apples

Peaches

Apricots

Cherries

Cantaloupe

Other Melons

Bananas

Pineapple

Lemons

Limes

Grapefruit

Avocado

Oranges

Irish Potatoes

Sweet Potatoes

Carrots

Beets

Turnips

Dry Onions

Green Onions

26. Other Root Vegetables

27. Cabbage

28. Lettuce

29. Other Leafy Vegetables

30. Pumpkin

31. Squash

32. Tomatoes

33. Asparagus

34. Lima Beans

35. Wax Beans

36. Peas

37. Celery

38. Cucumbers

39. Cauliflower

40. Sweet Corn

41. Broccoli

42. Other Vegetables

43. Rhubarb

44. Grain Corn

45. Oats

46. Rye

47. Barley

48. Rice

49. Wheat

50.

51.

52.

53.

54.

55.

56.

57.

58.

59.

60.

61.

62.

63.

64.

65.

66.

67.

68.

69.

70.

71.

72.

73.

Fresh Milk

Frozen Milk

Cream

Cheese

Processed Milk Desserts

Butter

Eggs

Beef

Veal

Lamb

Liver

Fresh Pork

Cured Pork

Canned Pork

Pork Lunch Meat

Chicken

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.

B-27

Turkey

Water Fowl

Fresh Ocean Fish

Processed Ocean Fish

Shell Fish

Sports Fish

Upland Game Birds

Secondary Water
TABLE B-4

Weekly Food Consumption Per Household(8)

Population Category

Food Type Urban Rural Non-Farm Rural Farm

Fresh Berries .420 1.020 1.290

Processed Berries .070 .130 .120

Fresh Tree Fruit 1.880 2.260 3.260

Processed Tree Fruit 3.940 3.610 3.830

Fresh Melons 1.610 .940 1.280

Fresh Tropical Fruit 1.510 1.610 2.070

Processed Tropical Fruit .240 .290 .280

Fresh Citrus Fruit 3.020 2.810 3.140

Processed Citrus Fruit 12.250 12.370 10.770

Fresh Potatoes 4.920 6.410 9.230

Processed Potatoes .720 .730 .630

Fresh Root Vegetables 1.560 1.630 1.960

Processed Root Vegetables .230 .190 .170

Fresh Green Leafy Vegetables 2.150 2.220 2.390

Processed Green Leafy Vegetables .160 .110 .100

Fresh Other Above Ground Vegetables 2.880 3.640 4.910

Processed Other Above Ground Vegetables 3.990 4.060 4.030

Grain and Grain Products .830 1.020 1.390

Rice and Rice Products .320 .260 .240

Wheat and Wheat Products 9.920 11.400 14.180

Fresh Milk 19.180 21.100 25.300

Processed Milk Products 3.810 4.150 5.230

Butter .500 .500 1.010

Eggs 2.530 2.840 3.720

Beef and Lamb 6.290 6.220 8.650

Fresh Pork 3.910 3.780 2.000

Processed Pork 4.000 4.150 4.840

Poultry 2.740 2.130 2.500

Water Fowl .003 .017 .023

Fresh Ocean Fish .450 .450 .400

Processed Ocean Fish .060 .090 .050

Fresh Fish .050 .050 .100

Sports Fish .120 .210 .220

Upland Game Birds .007 .033 .047

Secondary Water 32.260 33.430 31.770

(a)Pounds per week per household, except eggs in dozens per week per household.

B-28
TABLE B-5

Irrigated Food Production(13)

Number of Percent of

Centroids Crop

Food Type Producing Irrigated

Fresh Berries 56 50%

Fresh Melons 8 39%

Fresh Potatoes 43 53%

Fresh Green Leafy Vegetables 8 64%

Fresh Other Above Ground Vegetables 75 31%

Grain and Grain Products 19 36%

Wheat and Wheat Products 8 2%

B-29

APPENDIX C

Potential Radionuclide Release Rates

C-1
APPENDIX C

Potential Radionuclide Release Rates

LIST OF TABLES

Table

Number Title Page

C-1 Radionuclides Considered in Study....................................... C-3

Radionuclide Release RatesC-2 Boiling Water Reactor............................................ C-5

C-3 Advanced Boiling Water Reactor .................................... C-6

C-4 Pressurized Water Reactor........................................... C-7

C-5 Advanced Pressurized Water Reactor.................................. C-8

C-6 Liquid Metal Fast Breeder Reactor..................................... C-9

C-7 Calculated Stack Effluents from Reprocessing Plants with 150-day Precooled Fuel.........C-10

C-8 Radionuclide Release Rates-Dual-purpose Reprocessing Plant with 150-day

Precooled Fuel.........................................................C-11

C-9 Effects on Regional Release Rates of Delayed Introduction of Improved

Treatment Systems..................................................... C-13

C-10 Effect on Regional Air Release Rates of PWR/BWR Ratio ....................... .C-14

C-11 Effect on Regional Water Release Rates of PWR/BWR Ratio ..................... .C-15

C-2
TABLE C-1

Radionuclides Considered in Study

Serial

Number*

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

30

31

32

33

34

35

36

37

38

39

40

41

42

43

44

45

Nuclide

H-3

N-13

C-14

Na-22

Na-24

Ar-39

Ar-41

Cr-51

Mn-54

Fe-55

Fe-59

Co-58

Co-60

Ni-63

Cu-64

Zn-65

Kr-85m

Kr-85

Kr-87

Kr-88

Sr-89

Sr-90

Zr-95

Nb-95

Mo-99

R u-103

Ru-106

Te-132

1-129

I-131

1-132

1-133

I-135

Xe-131m

Xe-133m

Xe-133

Xe-135m

Xe-135

Xe-138

Cs-134

Cs-137

Ba-140

La-140

Ce-141

Ce-144

Considered in Release

Calculations

Half Life Air Water

1 2.26 y

9.96 m

5730. y

2.6 y

15.0 h

270. y

1.83 h

27.8 d

312. d

2.4 y

45. d

71. d

5.24 y

92. y

12.9 h

243. d

4.4 h

10. 7 6 y

76. m

2.8 h

50.6 d

28.8 y

65. d

35. d

67. h

40. d

1. y

78. h

1.6X 107 y

8.05 d

2.3 h

21. h

6.7 h

12. d

2.3 d

5.27 d

16. m

9.2 m

17. m

2.1 y

30. y

12.8 d

40.2 h

32.5 d

285. d

*Order of consideration in computer programs.

C-3

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X
Release rates for each type of nuclear power plant

and reprocessing plant, whose calculations were described in Chapter V, are shown in the following tables.

The sample release rates are for 10 year old, 1000 MWe

reactors operating with a 100% plant factor. The dual

purpose reprocessing plant is a nominal 5 Tonne/day

plant and the MFR releases are for its 1 Tonne/day

capacity. The tabulation of reactor releases is keyed to

the flow chart schematics found in Chapter V.

C-4
TABLE C-2

Radionuclide Release Rates-Boiling Water Reactor

Solids Retained'

(1) (2) (3) (4) (5) (6) (7) (8) (9) (10)

Recombiner,

Fuel Release Phase Collection Floor Condenser, Total Total

Generation to Separator, Tank Drain Charcoal Gas Liquid Gaseot

Isotope Rate Coolant Centrifuget Filter Demineralizer Filter Filter Filter Releases Releas

H-3 1.4533+01 1.3102-01 -0- -0- -0- -0- 1.0587-04 -0- 1.2938-01 2.6466-

N-13 -0- 1.5512+01 -0- -0- -0- -0- -0- -0- -0- 3.6597-

C-14 -0- 1.1125-12 -0- 1.0585-12 -0- 5.3595-16 -0- -0- 5.3461-14 -0-

Na-22 -0- 5.2077-09 -0- -0- -0- -0- -0- -0- 2.8316-09 -0-

Na-24 -0- 7.8840-01 -0- -0- -0- -0- -0- -0- 1.7973-05 -0-

Ar-39 -0- 3.1110-02 -0- -0- -0- -0- -0- -0- -0- 3.1110-

Ar-41 -0- 3.3752+00 -0- -0- -0- -0- -0- -0- -0- 8.9667-

Cr-51 -0- 4.3974-02 4.3479-02 1.1534-04 1.0380-04 1.2007-08 -0- -0- 5.8257-07 -0-

Mn-54 -0- 3.7105-03 3.6710-03 1.9248-05 1.7323-05 1.0286-09 -0- -0- 9.7180-08 -0-

Fe-55 -0- 1.0650-01 1.0537-01 5.7671-04 5.1904-04 2.9550-08 -0- -0- 2.9121-06 -0-

Fe-59 -0- 6.4312-03 6.3603-03 2.2456-05 2.0210-05 1.7671-09 -0- -0- 1.1331-07 -0-

Co-58 -0- 5.1030-01 5.0476-01 2.1108-03 1.8997-03 1.4075-07 -0- -0- 1.0651-05 -0-

Co-60 -0- 7.8235-04 7.7407-04 4.2915-06 3.8623-06 2.1714-10 -0- -0- 2.1671-08 -0-

Ni-63 -0- 6.9819-03 6.9081-03 3.8694-05 3.4824-05 1.9383-09 -0- -0- 1.9541-07 -0-

Cu-64 -0- 3.4093-07 3.2997-07 2.8913-26 2.6022-26 4.0633-14 -0- -0- 2.1319-14 -0-

Zn-65 -0- 2.1308-03 2.1081-03 1.0846-05 9.7615-06 5.9042-10 -0- -0- 5.4756-08 -0-

Kr-85M 4.6111+07 2.5097+00 -0- -0- -0- -0- -0- -0- -0- 2.1835

Kr-85 4.9514+02 4.4134+00 -0- -0- -0- -0- -0- -0- -0- 4.4129+

Kr-87 3.1171+08 2.6207+00 -0- -0- -0- -0- -0- -0- -0- 5.4313

Kr-88 1.9841+08 5.4827+00 -0- -0- -0- -0- -0- -0- -0- 1.1822-

Sr-89 6.1154+05 5.0410-01 4.9857-01 1.8529-03 1.6676-03 1.3867-07 -0- 1.2599-06 9.3497-06 1.2599-

Sr-90 3.6066+03 1.9277-02 1.9073-02 1.0669-04 9.6019-05 5.3514-09 -0- 4.8192-08 5.3877-07 4.8192

Zr-95 6.2014+05 8.8452-02 8.7489-02 3.5612-04 3.2051-04 2.4381-08 -0- 2.2108-07 1.7970-06 2.2108

Nb-95 1.1909+06 8.4563-02 8.3668-02 4.4392-04 3.9953-04 2.3488-08 -0- 2.1143-07 2.2394-06 2.1143-

Mo-99 1.4350+07 2.7453+02 2.6478+02 1.6255-03 1.4630-03 6.2767-04 -0- -0- 5.6171-04 -0-

Ru-103 4.8780+05 8.5540+01 8.4568+01 5.0725-01 5.5797-02 4.2241-05 -0- -0- 3.2631-05 -0-

Ru-106 7.1086+03 2.9555+01 2.9227+01 2.7852-01 3.0637-02 1.4743-05 -0- -0- 1.7098-05 -0-

Te-132 8.5220+06 3.9787+01 3.9206+01 3.6627-04 3.2964-04 9.6004-06 -0- -0- 1.0305-05 -0-

1-129 1.1137-03 3.3148-05 3.2782-05 -0- 3.3042-07 -0- 3.0303-08 -0- 1.8522-09 3.3670-

1-131 2.3507+06 3.4542+02 3.4078+02 -0- 2.5947-01 -0- 1.6958-01 -0- 1.5625-03 1.8842-

1-132 2.8905+08 9.6247+01 8.8758+01 -0- 6.7863-04 -0- 5.3120-24 -0- 1.0972-05 5.9022-

1-133 4.9720+07 8.6137+02 8.3978+02 -0- 4.0413-10 -0- 2.8507-03 -0- 1.9105-04 3.1674-

1-135. 1.4373+08 8.8065+02 8.3471+02 -0- 3.3543-30 -0- 1.8447-08 -0- 2.7702-05 2.0496-

Xe-131M 9.2102+03 2.6555-01 -0- -0- -0- -0- -0- -0- -0- 1.8492

Xe-133M 4.5236+05 1.0941+00 -0- -0- -0- -0- -0- -0- -0- 1.3645-

Xe-133 8.2533+06 6.9252+01 -0- -0- -0- -0- -0- -0- -0- 2.7979+

Xe-135M 1.0534+09 8.5550-01 -0- -0- -0- -0- -0- -0- -0- 6.4039

Xe-135 1.0688+08 3.6993+00 -0- -0- -0- -0- -0- -0- -0- 8.4676

Xe-138 3.3051+09 2.9393+00 -0- -0- -0- -0- -0- -0- -0- 3.4384-

Cs-134 5.6775+04 7.6140+00 7.4635+00 7.6535-02 6.8882-02 2.0720-05 -0- -0- 4.0324-04 -0-

Cs-137 3.6822+03 3.3841+01 3.3175+01 3.4889-01 3.1400-01 9.2148-05 -0- -0- 1.8365-03 -0-

Ba-140 3.2003+06 5.7849-01 5.7162-01 6.3127-04 5.6814-04 1.5494-07 -0- -0- 3.2367-06 -0-

La-140 2.4616+07 1.1111-01 1.1312-01 7.2628-04 6.5365-04 6.0947-08 -0- -0- 3.6285-06 -0-

Ce-141 1.2213+06 1.5213-02 1.5043-02 4.4449-05 4.0004-05 4.1638-09 -0- -0- 2.2440-07 -0-

Ce-144 1.2321+05 4.4960-02 4.4482-02 2.3176-04 2.0859-04 1.2461-08 -0- -0- 1.1701-06 -0-

*All units are curies/year/MWe.

t Includes coolant cleanup demineralizer residue.

is

,es

-05

-03

-02

-01

-01

-00

-04

-01

-06

-08

-07

-07

-09

-02

-25

-04

-09

-01

-01

-01

-09

-05

-31
TABLE C-3

Radionuclide Release Rates-Advanced Boiling Water Reactor

Solids Retaned

(1) (2) (3) (4) (5) (6) (7) (8) (9) (10)

R ecormbiner,

Fuel Release Phase Collection Floor Condenser, Total Total

Generation to Separator, Tank Drain Charcoal Gas Liquid Gaseou

Isotope Rate Coolant Centrifuget Filter Demineralizer Filter Filter Filter Releases Release

H-3 1.4533+01 2.0223-01 -0- -0- -0- -0- 1.6340-04 -0- 1.9973-01 4.0851-C

N-13 -0- 1.5512+01 -0- -0- -0- -0- -0- -0- -0- 3.6597-C

C-14 -0- 1.0198-12 -0- 9.7034-13 -0- 4.9132-16 -0- -0- 4.9008-14 -0- Na-22 -0- 5.2077-09 -0- -0- -0- -0- -0- -0- 2.8392-09 -0-

Na-24 -0- 7.8840-01 -0- -0- -0- -0- -0- -0- 1.9618-05 -0- Ar-39 -0- 3.1110-02 -0- -0- -0- -0- -0- -0- -0- 3.1110-C Ar-41 -0- 3.3752+00 -0- -0- -0- -0- -0- -0- -0- 8.9667-C Cr-51 -0- 4.3973-02 4.3479-02 1.1534-04 1.0380-04 1.2007-08 -0- -0- 5.8258-07 -0- Mn-54 -0- 3.5592-03 3.5213-03 1.8464-05 1.6617-05 9.8663-10 -0- -0- 9.3218-08 -0-

Fe-55 -0- 9.9499-02 9.8446-02 5.3881-04 4.8493-04 2.7608-08 -0- -0- 2.7208-06 -0-

Fe-59 -0- 6.4307-03 6.3599-03 2.2454-05 2.0209-05 1.7670-09 -0- -0- 1.1331-07 -0-

Co-58 -0- 5.0955-01 5.0401-01 2.1076-03 1.8969-03 1.4054-07 -0- -0- 1.0635-05 -0- Co-60 -0- 7.8218-04 7.7391-04 4.2906-06 3.8615-06 2.1710-10 -0- -0- 2.1667-08 -0- Ni-63 -0- 6.4117-03 6.3440-03 3.5534-05 3.1980-05 1.7800-09 -0- -0- 1.7945-07 -0-

Cu-64 -0- 3.4093-07 3.3018-07 2.8932-26 2.6039-26 4.0659-14 -0- -0- 2.1333-14 -OZn-65 -0- 2.1308-03 2.1081-03 1.0846-05 9.7615-06 5.9042-10 -0- -0- 5.4756-08 -0- Kr-85M 4.6111+07 4.9591+00 -0- -0- -0- -0- -0- -0- -0- 4.3146-C Kr-85 4.9514+02 6.8185+00 -0- -0- -0- -0- -0- -0- -0- 6.8178+( Kr-87 3.1171+08 5.1785+00 -0- -0- -0- -0- -0- -0- -0- 1.0732- Kr-88 1.9841+08 1.0834+01 -0- -0- -0- -0- -0- -0- -0- 2.3360-C Sr-89 6.1154+05 9.9611-01 9.8519-01 3.6614-03 3.2953-03 2.7402-07 -0- 2.4896-06 1.8475-05 2.4896-C

Sr-90 3.6066+03 3.3530-02 3.3175-02 1.8557-04 1.6701-04 9.3080-09 -0- 8.3824-08 9.3712-07 8.3824-C Zr-95 6.2014+05 1.7478-01 1.7288-01 7.0370-04 6.3333-04 4.8178-08 -0- 4.3686-07 3.5508-06 4.3686-C Nb-95 1.1909+06 1.6710-01 1.6533-01 8.7718-04 7.8947-04 4.6413-08 -0- 4.1779-07 4.4252-06 4.1779-C Mo-99 1.4350+07 5.4247+02 5.2385+02 3.2160-03 2.8944-03 1.2418-03 -0- -0- 1.1113-03 -0- Ru-103 4.8780+05 1.6903+02 1.6711+02 1.0023+00 1.1026-01 8.3469-05 -0- -0- 6.4479-05 -0-

R u-106 7.1086+03 4.8109+01 4.7575+01 4.5337-01 4.9871-02 2.3999-05 -0- -0- 2.7831-05 -0- Te-132 8.5220+06 7.8619+01 7.7479+01 7.2383-04 6.5144-04 1.8972-05 -0- -0- 2.0365-05 -0-

1-129 1.1137-03 4.1276-05 4.0821-05 -0- 4.1144-07 -0- 3.7734-08 -0- 2.3064-09 4.1926-C 1-131 2.3507+06 6.8255+02 6.7342+02 -0- 5.1274-01 -0- 3.3511-01 -0- 3.0877-03 3.7234-C

1-132 2.8905+08 1.9018+02 1.7574+02 -0- 1.3411-03 -0- 1.0518-23 -0- 2.1684-05 1.1686- 1-133 4.9720+07 1.7021+03 1.6601+03 -0- 7.9890-10 -0- 5.6352-03 -0- 2.7767-04 6.2613-C 1-135 1.4373+08 1.7402+03 1.6515+03 -0- 6.6365-30 -0- 3.6497-08 -0- 5.4807-05 4.0552-C Xe-131M 9.2102+03 5.2473-01 -0- -0- -0- -0- -0- -0- -0- 3.6527-4 Xe-133M 4.5236+05 2.1619+00 -0- -0- -0- -0- -0- -0- -0- 2.6888- Xe-133 8.2533+06 1.3684+02 -0- -0- -0- -0- -0- -0- -0- 5.5241+C Xe-135M 1.0534+09 1.6905+00 -0- -0- -0- -0- -0- -0- -0- 1.2670-C Xe-135 1.0688+08 7.3098+00 -0- -0- -0- -0- -0- -0- -0- 1.6292-C Xe-138 3.3051+09 5.8080+00 -0- -0- -0- -0- -0- -0- -0- 6.7944- Cs-134 5.6775+04 1.2408+01 1.2163+01 1.2473-01 1.1225-01 3.3767-05 -0- -0- 6.5716-04 -0- Cs-137 3.6822+03 5.2152+01 5.1125+01 5.3767-01 4.8390-01 1.4201-04 -0- -0- 2.8303-03 -0- Ba-140 3.2003+06 1.1431+00 1.1296+00 1.2474-03 1.1227-03 3.0617-07 -0- -0- 6.3959-06 -0- La-140 2.4616+07 2.1955-01 2.2334-01 1.4352-03 1.2916-03 1.2039-07 -0- -0- 7.1701-06 -0- Ce-141 1.2213+06 3.0061-02 2.9726-02 8.7832-05 7.9048-05 8.2278-09 -0- -0- 4.4341-07 -0- Ce-144 1.2321+05 8.3432-02 8.2545-02 4.3008-04 3.8707-04 2.3125-08 -0- -0- 2.1713-06 -0-

*All units are curies/year/MWe. tIncludes coolant cleanup demineralizer residue.

B

s

35

D3

02

D1

01

00

03

31

06

08

07

07

39

02

24

04

09

01

01

01

08

04

31
TABLE C-4

Radionuclide Release Rates-Pressurized Water Reactor

Solids Retained'

(1)

F uel

Generation

(2)

Release

to

Isotope Rate Co

H-3 1.4533+01 6.64

N-13 -0- 

C-14 -0- 9.32

Na-22 -0- 5.20

Na-24 -0- 7.88

Ar-39 -0- 3.11

Ar-41 -0- 2.10

Cr-51 -0- 1.35

Mn-54 -0- 9.69

Fe-55 -0- 6.72

Fe-59 -0- 1.48

Co-58 -0- 5.03

Co-60 -0- 8.54

Ni-63 -0- 8.76

Cu-64 -0- 3.40

Zn-65 -0- 2.13

C Kr-85M 4.6111+07 3.44

-. Kr-85 4.9514+02 3.43

Kr-87 3.1171+08 3.59

Kr-88 1.9841+08 7.52

Sr-89 6.1154+05 6.91

Sr-90 3.6066+03 1.42

Zr-95 6.2014+05 1.21

Nb-95 1.1909+06 1.16

Mo-99 1.4350+07 3.76

Ru-103 4.8780+05 1.17

Ru-106 7.1086+03 2.87

Te-132 8.5220+06 5.46

1-129 1.1137-03 2.57

1-131 2.3507+06 4.74

1-132 2.8905+08 1.32

1-133 4.9720+07 1.18

1-135 1.4373+08 1.20

Xe-131M 9.2102+03 3.64

Xe-133M 4.5236+05 1.50

Xe-133 8.2533+06 9.50

Xe-135M 1.0534+09 1.17

Xe-135 1.0688+08 5.07 Xe-138 3.3051+09 4.03

Cs-134 5.6775+04 4.20

Cs-137 3.6822+03 2.56

Ba-140 3.2003+06 7.94

La-140 2.4616+07 1.52

Ce-141 1.2213+06 2.08

Ce-144 1.2321+05 4.43

*A11 units are curies/year/MWe. t Includes steam generator blowdown.

olant

186-01

-0-

54-13

)77-09

40-01

10-02

52+00

81-01

)29-02

38-01

26-01

61+00

.36-02

68-04

93-07

08-03

150+00

10+00

373+00

60+00

97-01

86-02

42-01

08-01

84+02

42+02

759+01

14+01

703-05

15+02

12+02

24+03

88+03

451-01

18+00

)61+01

43+00

79+00

347+00

94+00

21+01

08-01

?51-01

383-02

388-02

(3)

Collection

Tank

Filter

-0-

-0-

4.8785-13

-0-

-0-

-0-

-0-

3.2346-04

4.7023-04

3.4114-03

4.7591-04

1.9290-02

4.3933-04

4.5567-06

7.4112-27

1.0135-05

-0-

-0-

-0-

-0-

2.3434-03

7.4145-05

4.5216-04

5.6853-04

3.0425-03

6.3866-01

2.5366-01

3.6230-04

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

1.9157-01

1.2046+00

7.5794-04

8.7202-04

5.5660-05

2.1392-04

(4) (5)

Demineralizer Evaporator

-0- -0-

-0- -0-

-0- 4.3906-13

-0- 4.7171-09

-0- 1.5919-17

-0- -0-

-0- -0-

2.9111-04 2.9111-05

4.2321-04 4.2321-05

3.0703-03 3.0703-04

4.2832-04 4.2832-05

1.7361-02 1.7361-03

3.9540-04 3.9540-05

4.1010-06 4.1010-07

6.6700-27 6.6700-28

9.1211-06 9.1211-07

-0- -0-

-0- -0-

-0- -0-

-0- -0-

2.1091-03 2.1091-04

6.6731-05 6.6731-06

4.0694-04 4.0694-05

5.1168-04 5.1168-05

2.7383-03 2.7383-04

7.0252-02 6.3866-04

2.7903-02 2.5366-04

3.2607-04 3.2607-05

2.4017-07 2.4017-08

2.9921-01 2.9921-02

6.7124-04 6.7124-05

2.1948-10 2.1948-11

4.8713-31 4.8713-32

-0- -0-

-0- -0-

-0- -0-

-0- -0-

-0- -0-

-0- -0-

1.7242-01 1.7242-02

1.0842+00 1.0842-01

6.8214-04 6.8214-05

7.8482-04 7.8482-05

5.0094-05 5.0094-06

1.9253-04 1.9253-05

(6)

Floor

Drain

F iltert

-0-

-0-

3.0064-15

-0-

-0-

-0-

-0-

4.2116-06

3.0975-06

2.1528-05

4.6558-06

1.5933-04

2.7370-06

2.8098-08

2.9256-12

6.8036-08

-0-

-0-

-0-

-0-

2.1779-05

4.5783-07

3.8367-06

3.7238-06

3.2228-02

6.6198-03

1.6549-03

1.3398-03

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-a1.2131-03

7.4378-03

2.3706-05

1.2390-05

6.5066-07

1.4181-06

(7)

Gas

Filter

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

1.1471-06

3.5643-08

2.2045-07

2.7776-07

-0- 2.1277-06

1.3341-06

6.6075-12

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

(8)

Coolant

Cleanup

System

-0-

-0-

-0-

-0-

-0-

-0-

-0-

1.3140-01

9.5727-02

6.6492-01

1.4468-01

4.9395+00

8.4520-02

8.6756-04

1.5462-07

2.1032-03

-0-

-0-

-0-

-0-

6.7628-01

1.4136-02

1.1899-01

1.1494-01

1.1147+02

1.1437+02

2.8409+01

4.5211+01

2.5438-05

4.3490+02

5.6424+01

6.7794+02

3.6524+02

-0-

-0-

-0-

-0-

-0-

-0-

3.7847+00

2.3198+01

7.4866-01

3.1527-01

2.0269-02

4.3830-02

(9)

Total

Liquid

Releases

4.8601-01

-0-

2.5895-15

2.7848-11

9.7620-07

-0-

-0-

3.6846-07

3.8963-07

2.7812-06

4.6774-07

1.7540-05

3.5646-07

3.6832-09

8.5460-14

8.4590-09

-0-

-0-

-0-

-0-

2.2478-06

5.9962-08

4.1606-07

4.7001-07

1.4544-03

4.0035-05

1.0595-05

6.1460-05

2.1552-10

1.4686-03

6.6730-05

1.1324-03

1.4908-04

-0-

-0-

-0-

-0-

-0-

-0-

1.5638-04

9.7419-04

1.5294-06

1.1332-06

5.9829-08

1.7769-07

(10)

Total

Gaseous

Releases

-0-

-0-

-0-

-0-

-0-

3.1103-02

8.3693-31

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

1.3695-30

3.4129+00

1.4310-30

2.9919-30

1.1471-06

3.5643-08

2.2045-07

2.7776-07

-0-

2.1277-06

1.3341-06

6.6075-12

6.3300-10

1.8442-04

1.3617-11

2.0563-13

1.7346-33

1.0728-01

7.1732-04

3.4911+00

2.5164-27

1.7107-21

1.6040-30

-0-

-0-

-0-

-0-

-0-

-0-
TABLE C-5

Radionuclide Release Rates-Advanced Pressurized Water Reactor

Solids Retained

(1) (2) (3) (4) (5) (6) (7) (8) (9) (10)

Fuel Release Collection Floor Coolant Total Total

Generation to Tank Drain Gas Cleanup Liquid Gaseous

Isotope Rate Coolant Filter Demineralizer Evaporator Filter t Filter System Releases Releases

H-3 1.4533+01 7.1971-01 -0- -0- -0- -0- -0- -0- 5.3682-01 -0-

N-13 -0- -0- -0- -0- -0- -0- -0- -0- -0- -0-

C-14 -0- 8.3699-13 4.3786-13 -0- 3.9408-13 2.6984-15 -0- -0- 2.3242-15 -0-

Na-22 -0- 5.2077-09 -0- -0- 4.7486-09 -0- -0- -0- 2.8034-11 -0-

Na-24 -0- 7.8840-01 -0- -0- 1.7668-17 -0- -0- -0- 1.0834-06 -0-

Ar-39 -0- 3.1110-02 -0- -0- -0- -0- -0- -0- -0- 3.1103-0

Ar-41 -0- 2.1052+00 -0- -0- -0- -0- -0- -0- -0- 8.3693-3

Cr-51 -0- 1.3581-01 3.2417-04 2.9175-04 2.9175-05 4.2209-06 -0- 1.3168-01 3.6927-07 -0-

Mn-54 -0- 9.0680-02 4.4001-04 3.9600-04 3.9600-05 2.8984-06 -0- 8.9574-02 3.6459-07 -0-

Fe-55 -0- 6.1350-01 3.1129-03 2.8016-03 2.8016-04 1.9644-05 -0- 6.0673-01 2.5378-06 -0-

Fe-59 -0- 1.4807-01 4.7595-04 4.2835-04 4.2835-05 4.6562-06 -0- 1.4469-01 4.6777-07 -0-

Co-58 -0- 4.9943+00 1.9147-02 1.7232-02 1.7232-03 1.5815-04 -0- 4.9029+00 1.7410-05 -0-

Co-60 -0- 7.7278-02 3.9739-04 3.5765-04 3.5765-05 2.4757-06 -0- 7.6452-02 3.2244-07 -0-

Ni-63 -0- 7.8804-04 4.0959-06 3.6863-06 3.6863-07 2.5257-08 -0- 7.7984-04 3.3108-09 -0-

Cu-64 -0- 3.4093-07 7.8356-27 7.0520-27 7.0520-28 3.0932-12 -0- 1.6347-07 9.0355-14 -0-

Zn-65 -0- 2.1308-03 1.0137-05 9.1235-06 9.1235-07 6.8054-08 -0- 2.1037-03 8.4613-09 -0- n Kr-85M 4.6111+07 6.9526+00 -0- -0- -0- -0- -0- -0- -0- 2.7640-3

00 Kr-85 4.9514+02 5.2899+00 -0- -0- -0- -0- -0- -0- -0- 5.2621+0

Kr-87 3.1171+08 7.2600+00 -0- -0- -0- -0- -0- -0- -0- 2.8862-3

Kr-88 1.9841+08 1.5189+01 -0- -0- -0- -0- -0- -0- -0- 6.0382-3

Sr-89 6.1154+05 1.3965+00 4.7352-03 4.2617-03 4.2617-04 4.4007-05 2.3151-06 1.3665+00 4.5420-06 2.3151-0

Sr-90 3.6066+03 2.4570-02 1.2752-04 1.1477-04 1.1477-05 7.8742-07 6.1303-08 2.4313-02 1.0313-07 6.1303-0

Zr-95 6.2014+05 2.4504-01 9.1342-04 8.2207-04 8.2207-05 7.7507-06 4.4490-07 2.4038-01 8.4050-07 4.4490-0

Nb-95 1.1909+06 2.3426-01 1.1478-03 1.0330-03 1.0030-04 7.5149-06 5.5996-07 2.3197-01 9.4877-07 5.5996-0

Mo-99 1.4350+07 7.6052+02 6.5836-03 5.9253-03 5.9253-04 6.9737-02 -0- 2.4120+02 3.1471-03 -0-

Ru-103 4.8780+05 2.3697+02 1.2909+00 1.4200-01 1.2909-03 1.3381-02 4.3008-06 2.3119+02 8.0923-05 4.3008-0

Ru-106 7.1086+03 4.6556+01 4.1071-01 4.5178-02 4.1071-04 2.6794-03 2.1600-06 4.5997+01 1.7154-05 2.1600-C

Te-132 8.5220+06 1.1022+02 7.4333-04 6.6899-04 6.6899-05 2.7489-03 1.3557-11 9.2759+01 1.2610-04 1.3557-1

1-129 1.1137-03 3.2203-05 -0- 3.0090-07 3.0090-08 -0- -0- 3.1871-05 2.7001-10 7.9306-1

1-131 2.3507+06 9.5692+02 -0- 6.0830-01 6.0830-02 -0- -0- 8.8417+02 2.9857-03 3.7302-0

1-132 2.8905+08 2.6663+02 -0- 1.3772-03 1.3772-04 -0- -0- 1.1733+02 1.3692-04 2.7937-1

1-133 4.9720+07 2.3862+03 -0- 4.6240-10 4.6240-11 -0- -0- 1.4283+03 2.3857-03 4.1523-1

1-135 1.4373+08 2.4397+03 -0- 1.0565-30 1.0565-31 -0- -0- 7.9213+02 3.2332-04 3.5014-3

Xe-131M 9.2102+03 7.3565-01 -0- -0- -0- -0- -0- -0- -0- 2.0878-0

Xe-133M 4.5236+05 3.0309+00 -0- -0- -0- -0- -0- -0- -0- 1.3820-0

Xe-133 8.2533+06 1.9185+02 -0- -0- -0- -0- -0- -0- -0- 6.8472+C

Xe-135M 1.0534+09 2.3700+00 -0- -0- -0- -0- -0- -0- -0- 4.9120-2

Xe-135 1.0688+08 1.0248+01 -0- -0- -0- -0- -0- -0- -0- 3.3402-2

Xe-138 3.3051+09 8.1426+00 -0- -0- -0- -0- -0- -0- -0- 3.2371-3

Cs-134 5.6775+04 6.6559+00 3.0314-01 2.7283-01 2.7283-02 1.9195-03 -0- 5.9888+00 2.4745-04 -0-

Cs-137 3.6822+03 3.9709+01 1.8671+00 1.6804+00 1.6804-01 1.1528-02 -0- 3.5955+01 1.5099-03 -0-

Ba-140 3.2003+06 1.6026+00 1.5369-03 1.3832-03 1.3832-04 4.8071-05 -0- 1.5181+00 3.1012-06 -0-

La-140 2.4616+07 3.0780-01 1.7682-03 1.5914-03 1.5914-04 2.4417-05 -0- 6.1337-01 2.2682-06 -0-

Ce-141 1.2213+06 4.2145-02 1.1255-04 1.0129-04 1.0129-05 1.3157-06 -0- 4.0985-02 1.2098-07 -0-

Ce-144 1.2321+05 8.1309-02 3.9194-04 3.5275-04 3.5275-05 2.5982-06 -0- 8.0304-02 3.2555-07 -0-

*All units are curies/year/MWe. includes steam generator blowdown.

)2

1

30

)0

30

30

)6

)8

)7

17

)6

)6

11

10

14

11

13

3

1

13

10

27

21

30
TABLE C-6

Radionuclide Release Rates' -Liquid Metal Fast Breeder Reactor

(1) (2) (3) (4) (5) (6)

Fuel Release Coolant Freon or Solids Gaseous

Generation to Cleanup Cryogenic Retained Release

Isotope Rate Coolant System System by Filters Rate

H-3 1.1080+01 3.4352-01 3.4348-01 -0- -0- 3.4192-07

N-13 -0- -0- -0- -0- -0- -0-

C-14 -0- 6.2527-13 6.2526-13 -0- -0- 6.2526-18

Na-22 -0- 1.3616-05 -0- -0- -0- 1.3321-09

Na-24 -0- 6.2810+02 -0- -0- -0- 2.2088-16

Ar-39 -0- 1.0401-04 -0- -0- -0- 1.0398-04

Ar-41 -0- 5.7532+02 -0- -0- -0- 2.2872-28

Cr-51 -0- 3.7328-02 3.6624-02 -0- -0- -0-

Mn-54 -0- 5.2574-02 5.2484-02 -0- -0- -0-

Fe-55 -0- 5.4836-03 5.4802-03 -0- -0- -0-

Fe-59 -0- 6.9407-04 6.8592-04 -0- -0- -0-

Co-58 -0- 1.7679-01 1.7547-01 -0- -0- -0-

Co-60 -0- 7.3100-04 7.3080-04 -0- -0- -0-

Ni-63 -0- 8.8302-05 8.8300-05 -0- -0- -0-

Cu-64 -0- -0- -0- -0- -0- -0- Zn-65 -0- -0- -0- -0- -0- -0-

Kr-85M 1.8658+07 2.6877+01 -0- 1.0649-29 -0- 3.5616-32

Kr-85 2.0073+02 6.1926+00 -0- 6.1396+00 -0- 2.0534-02

Kr-87 1.0708+08 2.3826+01 -0- 9.4403-30 -0- 3.1573-32

Kr-88 5.9485+07 4.3504+01 -0- 1.7237-29 -0- 5.7649-32

Sr-89 1.7630+05 3.8462+00 3.8060+00 -0- 6.3761-06 6.3761-06

Sr-90 9.9509+02 4.0393-02 4.0391-02 -0- 1.0078-07 1.0078-07

Zr-95 4.0414+05 1.5256+00 1.5132+00 -0- 2.7700-06 2.7700-06 Nb-95 7.5064+05 1.4341+00 1.4353+00 -0- 3.4505-06 3.4505-06

Mo-99 1.0443+07 5.2873+03 4.4371+03 -0- 7.6902-06 7.6902-06

R u-103 7.2515+05 3.3655+03 3.3211+03 -0- 5.0024-05 5.0024-05

Ru-106 6.3836+04 1.3613+03 1.3593+03 -0- 3.2148-05 3.2148-05

Te-132 5.3377+06 6.5955+02 5.6635+02 -0- 1.0991-05 1.0991-05

1-129 1.0189-03 5.7657-05 5.7655-05 -0- -0- 1.1531-09

1-131 2.0938+06 8.1430+03 7.6355+03 -0- -0- 1.2301-02

1-132 1.9449+08 1.7140+03 7.4074+02 -0- -0- 2.2650-05 1-133 3.2295+07 1.4808+04 9.1899+03 -0- -0- 1.4127-11

1-135 1.1188+08 1.8143+04 6.2218+03 -0- -0- 1.4425-31

Xe-131M 8.2606+03 6.3034+00 -0- 1.7110+00 -0- 5.7225-03

Xe-133M 3.0175+05 1.9315+01 -0- 8.4251-03 -0- 2.8178-05

Xe-133 5.3608+06 1.1905+03 -0- 4.1566+01 -0- 1.3902-01

Xe-135M 9.8153+08 2.1097+01 -0- 3.5606-26 -0- 1.1908-28 Xe-135 8.7947+07 3.8299+02 -0- 2.5058-20 -0- 8.3805-23

Xe-138 2.0997+09 4.9421+01 -0- 1.9581-29 -0- 6.5490-32

Cs-134 3.6878+04 8.6808+01 8.6745+01 -0- 8.4484-04 8.4484-04

Cs-137 2.9429+03 9.1780+01 9.1773+01 -0- 9.1606-04 9.1606-04

Ba-140 1.9750+06 9.4488+00 9.0699+00 -0- -0- -0- La-140 1.5096+07 1.8033+00 3.5621+00 -0- -0- -0-

Ce-141 6.4105+05 2.1134-01 2.0792-01 -0- -0- -0-

Ce-144 6.1404+04 2.8252-01 2.8199-01 -0- -0- -0-

*A1l units are In curies/year/MWE.
TABLE C-7

Calculated Stack Effluents from Reprocessing Plants

with 150-Day Precooled Fuel

(Curies/Year)

MFR Plant

(LWR Fuel Only)

2.07 X 105

2.67 X 10-12

1.22 X 10-6

3.56X 10~7

3.62X 10-6

4.19X 10-8

4.16 X 10-6

1.25 X 10~5

1.15 X 10-6

1.11 X 104

4.85X 10-1

4.32 X 10-1

9.92 X 10-2

1.87X 10-1

-0-

2.05 X 10-1-

1.05

-0-

1.45 X 10-1

5.07

-0-

3.47

-0-

-0-

7.75 X 10-2

3.91 X 10-2

1.15 X 10~4

1.33 X 10- 4

1.84X 10-2

2.78X 10-1

Aqueous 'PUR E X' Plants

LWR Fuel

1.04X 106

2.43 X 10-5

1.11 X 10-5

3.24 X 10-6

3.31 X 10- 5

3.78 X 10-7

3.78X 10-5

1.14 X 10~ 4

1.05X 10~5

5.54 X 104

2.90 X 101

2.58 X 101

8.69 X 101

1.64X 102

-0-

2.50

1.22 X 101

-0-

1.41 X 10-2

5.99X 10-1

-0-

1.736 X 10+1

-0-

-0-

3.96

1.99

5.67X 10- 3

6.52X 10-3

9.29 X 10~1

1.42 X 10~1

FBR Fuel

6.851 X 105

5.468 X 10-4

1.167 X 10-5

9.923 X 10~4

3.780 X 10-4

8.656 X 10-6

8.332 X 10-4

1.194 X 10-5

6.328 X 10~7

2.480 X 104

1.878 X 101

7.258

9.185X 101

1.731 X 102

6.965 X 10-15

1.428 X 101

7.062 X 101

2.696 X 10 9

9.837 X 10-'

5.991 X 10~

2.778X 10-

2.323 X 101

4.649X 10-15

2.229 X 10-2

2.422 X 10~1

1.011

5.188 X 10~4

5.969 X 10~4

9.296 X 10~1

8.773

C-10

H3

C14

Cr51

Mn54

Fe55

Fe59

Co58

Co60

Ni63

Kr85

Sr89

Sr90

Zr95

Nb95m

Mo99

Ru 103

Ru 106

Te132

1129

1131

1132m

Xe 131

Xe 133

Xe 133

Cs134

Csl 37

Ba140

La140

Ce 141

Ce 144
TABLE C-8

Radionuclide Release Rates -Dual-Purpose Reprocessing Plant with 150-Day Precooled Fuel

(1) (2) (3) (4) (5) (6)

Filters, Bottled

Stack Resins, Solidified Noble Fuel

Isotope Inlet Release etc. Wastes Gases Hulls

H-3 8.616+05 8.616+05 -0- -0- -0- -0-

N-13 -0- -0- -0- -0- -0- -0-

C-14 2.287+01 2.859-04 -0- 2.259-02 -0- 2.285+01

Na-22 -0- -0- -0- -0- -0- -0-

Na-24 -0- -0- -0- -0- -0- -0-

Ar-39 -0- -0- -0- -0- -0- -0-

Ar-41 -0- -0- -0- -0- -0- -0-

Cr-51 1.070+06 1.138-05 1.175+01 9.625+02 -0- 1.045+06

Mn-54 4.043+08 4.983-04 5.039+02 3.961+04 -0- 4.030+07

Fe-55 2.664+07 3.314-04 3.346+02 2.623+04 -0- 2.659+07

Fe-59 3.999+05 4.522-06 4.629+00 3.728+02 -0- 3.939+05

Co-58 3.715+07 4.631-04 4.440+02 3.538+04 -0- 3.680+07

Co-60 5.043+06 6.291-05 6.348+01 4.973+03 -0- 5.036+06

Ni-63 4.462+05 5.579-06 5.629+00 4.406+02 -0- 4.459+05

Cu-64 -0- -0- -0- -0- -0- -0-

Zn-65 -0- -0- -0- -0- -0- -0-

Kr-85M -0- -0- -0- -0- -0- -0-

Kr-85 1.215+07 4.008+04 -0- -0- 1.211+07 -0-

Kr-87 -0- -0- -0- -0- -0- -0-

Kr-88 -0- -0- -0- -0- -0--

Sr-89 1.162+08 2.393+01 2.437+08 1.079+08 -0- -0-

Sr-90 7.361+07 1.655+01 1.664+06 7.190+07 -0- -0-

Zr-95 4.264+08 8.950+01 9.089+06 4.002+08 -0- -0-

Nb-95 7.969+08 1.689+02 1.713+07 7.524+08 -0- -0-

Mo-99 7.748-08 3.488-15 4.409-10 2.989-08 -0- -0-

Ru-103 1.503+08 8.269+00 1.234+06 1.386-08 -0- -0-

R u-106 6.938+08 4.131+01 6.118+06 6.812+08 -0- -0-

Te-132 1.205-05 1.350-09 3.564-07 5.046-06 -0- -0-

1-129 4.478+01 1.198-02 2.599+00 4.226+01 -0- -0-

1-131 3.278+03 5.992-01 1.381+02 2.148+03 -0- -0-

1-132 1.243-05 1.391-09 3.658-07 5.203-06 -0- -0-

1-133 -0- -0- -0- -0- -0- -0-

1-135 -0- -0- -0- -0- -0- -0- Xe-131M 6.150+03 1.391+01 -0- -0- 6.150+03 -0-

Xe-133M 7.064-03 2.327+00 -0- -0- 7.033-13 -0-

Xe-133 1.274+01 4.194-02 -0- -0- 1.268+01 -0-

Xe-135M -0- -0- -0- -0- -0- -0-

Xe-135 -0- -0- -0- -0- -0- -0-

Xe-138 -0- -0- -0- -0- -0- -0-

Cs-134 1.697+08 2.107+00 2.129+06 1.668+08 -0- -0-

Cs-137 1.205+08 1.507+00 1.520+06 1.189+08 -0- -0-

Ba-140 3.532+04 3.103-03 3.280+03 2.841+05 -0- -0-

La-140 4.068+05 3.564-03 3.784+03 3.280+05 -0- -0-

Ce-141 8.546+07 9.303-01 9.587+05 7.789+07 -0- -0-

Ce-144 9.335+08 1.151+01 1.167+07 7.789+07 -0- -0-

* All units are curie/year.
The following tables show the potential total regional

release rates for two of the sensitivity studies. In the

delayed treatment case reactor treatment systems built

before 1985 were less elaborate. All reprocessing plants

also had less elaborate treatment systems. A description

of the differences is found in the discussion in Chapter

X. The light water reactor type sensitivity study releases

are also included in the following tables. The release

rates shown in the tables are the sum of all releases from

the nuclear power plants and fuel reprocessing plants in

the region.

C-12
TABLE C-9

Effects on Regional Release Rates of Delayed Introduction of Improved Treatment Systems

Total Regional Release Rates (Ci/year)

Air Water

H3

N13

C14

Na22

Na24

Ar39

Ar41

Cr51

Mn54

Fe55

Fe59

Co58

Co60

Ni63

Cu64

Zn65

Kr85m

Kr85

Kr87

Kr88

Sr89

Sr90

Zr95

Nb95

Mo99

Ru 103

Ru 106

Te132

1129

1131

1132

1133

1135

Xe131

Xe133m

Xe133

Xe135m

Xel35

Xe138

Csl34

Cs137

Ba 140

La140

Ce141

Ce144

Base

7.09 X 106

1.10X 102

2.28 X 10- 3

-0-

-0-

7.88 X 103

2.70 X 104

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

8.22 X 103

9.58 X 105

2.05 X 101

4.44 X 103

1.93 X 102

1.33 X 102

7.15 X 102

1.35 X 103

-0-

2.84 X 101

1.16X 102

1.26

2.02X 10-1

2.15 X 103

2.60

1.19X 101

7.72X 10-5

1.99 X 104

5.23 X 103

1.47X 106

2.42 X 10-5

3.12

-0-

1.25 X 102

1.41 X 102

2.49X 10-2

2.87 X 10-2

7.47

9.23 X 101

Delayed Advance

Treatment Use Date

7.09 X 106

2.12 X 104

2.28X 10-3

-0-

-0-

7.88 X 10 3

4.79 X 104

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

-0-

2.22 X 104

1.00 X 10

1.33X 104

3.59 X 104

1.93X 102

1.33X 102

7.15 X 102

1.35 X 10 3

-0-

2.87 X 101

1.16X 102

1.26

2.03X 10-1

3.75 X 103

2.60

3.65 X 102

9.64X 10-1

7.02 X 104

1.18X 104

1.75 X 106

3.94X 10

1.69X 105

5.73 X 103

1.25X 102

1.41 X 102

2.49 X 10-2

2.87X 10-2

7.47

9.23 X 101

C-13

Base

5.13X 104

-0-

-0-

6.84 X 10~5

6.65 X 10~1

-0-

-0-

5.04X 10-2

4.13X 10-2

5.23 X 10-1

4.57 X 10-2

1.85

4.76 X 10~1

1.40 X 10-1

-0-

2.38 X 10-3

-0-

-0-

-0-

-0-

6.14 X 10-1

3.69X 10-2

1.16 X 10-1

1.39X 10-1

2.01 X 102

5.92

1.84

7.65

9.10 X 10~5

2.31 X 102

8.32

1.44 X 102

1.96

-0-

-0-

-0-

-0-

-0-

-0-

5.13 X 101

2.08 X 102

3.01 X 10-1

2.68 X 10-1

1.54X 10-2

7.21 X 10-2

Delayed Advanced

Treatment Use Date

5.42 X 104

-0-

-0-

6.84X 1 O1.12

-0-

-0-

1.84 X 10-1

2.07X 10-1

2.31

2.12X 10-1

7.75

7.50 X 10~

1.67 X 10-1

-0-

5.07 X 10-3

-0-

-0-

-0-

-0-

1.07X 102

7.25 X 101

3.94 X 102

7.43 X 102

5.45 X 102

9.80 X 102

4.88 X 10

2.51 X 101

3.53 X 10-3

6.21 X 102

2.72 X 101

4.73X 102

6.39 X 101

-0-

-0-

- 0 -+

-0-

-0-

1.25X 102

4.28X 102

6.99X 10-1

5.45 X 102

6.87

8.38X 101
TABLE C-10

Effect on Regional Air Release Rates of PWR/BWR Ratio

Total Air Releases (Ci/year) for January 2000

H3

N13

C14

Na22

Na24

Ar39

Ar41

Kr85m

Kr85

Kr87

Kr88

Sr89

Sr90

Zr95

Nb95

Ru 103

Ru 106

Tel 32

1129

1131

1132

1133

1135

Xe 131

Xe133m

Xe 133

Xe135m

Xe135

Xe138

Csl 34

Cs137

Ba 140

La140

Ce141

Ce144

Base (3 PWR: 1 BWR)

7.09 X 106

1.10X 102

2.28X 10-3

-0-

-0-

7.88 X 103

2.70X 104

8.22 X 103

9.58X 105

2.05 X 101

4.44 X 103

1.93 X 102

1.33X 102

7.15 X 102

1.35X 103

2.84 X 101

1.16X 102

1.26

2.02 X 10-1

2.15 X 103

2.60

1.19X 101

7.72X 10-5

1.99 X 104

5.23 X 103

1.47X 106

2.42 X 10- 5

3.12

-0-

1.25X 102

1.41 X 102

2.49 X 10-2

2.87 X 10-2

7.47

9.23 X 101

All BWR

7.09 X 106

4.13 X 102

2.28 X 10-3

-0-

-0-

7.88 X 103

1.01 X 105

3.31 X 104

1.07 X 106

8.22 X 102

1.79X 104

1.93 X 102

1.33 X 102

7.15X 102

1.35X 103

2.82 X 101

1.16X 102

1.26

2.02 X 10-1

4.28 X 103

2.60

4.79 X 101

3.11 X 10~4

2.89 X 104

2.06 X 104

4.25 X 106

9.70 X 10- 4

1.25 X 101

-0-

1.25 X 102

1.41 X 102

2.49X 10-2

2.87 X 10-2

7.47

9.23 X 101

C-14

All PWR

7.09 X 106

1.02 X 101

2.28X 10- 3

-0-

-0-

7.88 X 103

2.51 X 103

1.66 X 102

9.20 X 105

4.10 X 10~1

8.98 X 101

1.93 X 102

1.33 X 102

7.15 X 102

1.35 X 103

2.84 X 101

1.16X 102

1.26

2.02

1.46 X 103

2.60

2.40 X 10-1

-0-

1.70 X 104

2.13X 102

5.64 X 105

-0-

6.43 X 10-1

-0-

1.25 X 102

1.41 X 102

2.49 X 10-2

2.87 X 10-2

7.47

9.23 X 101
TABLE C-11

Effect on Regional Water Release Rates of PWR/BWR Ratio

Total Liquid Releases (Ci/year) for January 2000

Base All BWR All PWR

H3 5.13 X 104 2.34 X 104 6.05 X 104

C14 -0- -0- -0-

Na22 6.84 X 10~5 2.54 X 10~4 - 0 

Na24 6.65 X 10-1 2.16 1.70X 10-1

Cr51 5.04 X 10-2 6.96 X 10-2 4.41 X 10-2

Mn54 4.13 X 10- 2 1.42 X 10- 2 5.01 X 10-2

Fe55 5.23 X 10-1 5.17 X 10-1 5.26 X 10-1

Fe59 4.57 X 10-2 1.59 X 10-2 5.58 X 10-2

Co58 1.85 1.27 2.04

Co60 4.76 X 10-1 1.55 1.18 X 10-1

Ni63 1.40X 10-1 4.79X 10-1 2.82X 10-2

Cu64 -0- -0- -0-

Zn65 2.38 X 10-3 6.21 X 10-3 1.12 X 10-3

Sr89 6.14 X 10~1 1.42 3.53 X 10-1

Sr90 3.69X 10-2 1.05X 10-1 1.43X 10-2

Zr95 1.16 X 10-1 2.72 X 10-1 6.55 X 10-2

Nb95 1.39 X 10-1 3.40 X 10-1 7.43 X 10-2

Mo99 2.01 X 102 8.63 X 101 2.39 X 101

RulO3 5.92 4.98 6.21

R u106 1.84 2.50 1.63

Te132 7.65 1.61 9.61

1129 9.10X 10~5 2.56X 10~4 3.65X 10-s

1131 2.31 X 102 2.38 X 102 2.29 X 102

1132 8.32 1.72 1.05

1133 1.44 X 102 2.99 X 101 1.81 X 102

1135 1.96 X 101 4.32 2.45 X 101

Cs134 5.13X10 1 8.91X10 1 3.91 X 101

Cs137 2.08 X 102 3.13 X 102 1.74 X 102

Ba140 3.01 X 10-1 4.91 X 10~1 2.39 X 10-1

La140 2.68 X 10-1 5.51 X 10-1 1.76 X 10-1

Ce141 1.54X 10-2 3.40 X 10-2 9.42 X 10-3

Ce144 7.21 X 10-2 1.94 X 10-1 3.21 X 10-2

C-15

APPENDIX D

Potential Dose Rate Patterns in Study Area

D-1
APPENDIX D

Potential Dose Rate Patterns in Study Area

LIST OF TABLES

Table

Number Title Page

D-1 Inhalation Dose Factors, Average Child...................................... D-6

D-2 Inhalation Dose Commitment Factors, Average Child .......................... D-7

D-3 Inhalation Dose Factors, Average Teen .................................... D-8

D-4 Inhalation Dose Commitment Factors, Average Teen .......................... D-9

D-5 Inhalation Dose Factors, Average Adult ................................... . D-10

D-6 Inhalation Dose Commitment Factors, Average Adult .......................... D-11

D-7 Ingestion Dose Factors, Average Child .................................... D-12

D-8 Ingestion Dose Commitment Factors, Average Child ............................. D. -13

D-9 Ingestion Dose Factors, Average Teen....................................... D-14

D-10 Ingestion Dose Commitment Factors, Average Teen ........................... . D-15

D-1I Ingestion Dose Factors, Average Adult .................................... D-16

D-12 Ingestion Dose Commitment Factors, Average Adult........................... .. D-17

D-13 Dose Factors for Submersion in Air ...................................... D-18

D-14 External Dose Factors for Immersion in Water................................ .. D-19

D-1 5 External Dose Factors for Exposure While Standing on Contaminated Ground

or River Bank ....................................................... D-20

LIST OF FIGURES

Figure

Number Title Page

D-1 Regional Pattern of Total-Body Annual Dose (Average Adult) .................... D-21

D-2 Regional Pattern of GI Tract Annual Dose (Average Adult) ...................... D-22

D-3 Regional Pattern of Thyroid Annual Dose (Average Adult)....................... .. D-23

D-4 Regional Pattern of Bone Annual Dose (Average Adult)......................... .. D-24

D-5 Regional Pattern of Lung Annual Dose (Average Adult)......................... .. D-25

D-6 Regional Pattern of Skin Annual Dose (Average Adult) ......................... .. D-26

D-7 Regional Pattern of Liver Annual Dose (Average Adult)......................... .. D-27

D-8 Average Individual Total-Body Dose Distribution ............................. . D-28

D-9 Average Individual GI Tract Dose Distribution ............................... . D-29

D-10 Average Individual Thyroid Dose Distribution ............................... . D-30

D-11 Average Individual Bone Dose Distribution ................................. . D-31

D-12 Average Individual Lung Dose Distribution ................................. . D-32

D-13 Average Individual Skin Dose Distribution.................................... D-33

D-14 Average Individual Liver Dose Distribution ................................. . D-34

D-2
APPENDIX D

POTENTIAL DOSE RATE PATTERNS IN STUDY AREA

RELATIVE CONTRIBUTION OF VARIOUS

RADIONUCLIDES IN RADIATION DOSE

The original list of 45 radionuclides was chosen on the

basis of their potential contribution to human radiation

doses from at least one of the several pathways of

exposure considered. It was conceded that some of the

nuclides might not make a significant contribution from

any pathways. The relative importance of the individual

nuclides could only be determined by actually

performing the calculations for the entire list. As it

turned out, with the releases and waste treatment

systems assumed, only a few of the radionuclides

originally considered contributed significantly to the

calculated doses.

A discussion of the contributions of selected nuclides

to dose is presented below to provide insight into the

relative importance of various portions of the regional

model and to identify radionuclides of potential interest

for other purposes. Information concerning the

radionuclides which contribute the major portion of the

doses is summarized both in this Appendix and

Appendix E.

Tritium

Even a cursory examination of the results of the

regional study indicates that tritium would make a

substantial contribution to many of the organ doses,

including the total-body dose. Tritium is produced as a

fission product in both LWRs and LMFBRs. It is also

produced as a coolant activation product in both BWRs

and PWRs, but at a much greater rate in PWRs due to

their use of soluble boron shim control. In LWRs the

tritium is unaffected by the liquid waste systems and,

under steady state conditions, tritium release rates are

equal to the sum of the release rates from the fuel plus

the activation rates in the coolant. The tritium not

released from fuel within the reactor is released later at

fuel reprocessing plants. LMFBR fuel is assumed to

behave the same as LWR fuel, although experience( 1 5 7)

suggests that tritium can be expected to diffuse through

the stainless clad to a greater degree than through LWR

zirconium clad. In the LMFBR, a portion of the tritium

which leaks into the coolant is assumed to be removed

by the cold trap system; a small fraction is released with

the gaseous effluent. At the reprocessing plants all of the

tritium is assumed to be released with the gaseous

effluent. In all instances, tritium is assumed to be in the

form of its oxide by the time it reaches the point of

human exposure.

Tritium entering the region's waterways with the

liquid effluent from either type of reactor is assumed to

be unaffected by any concentration mechanisms as it is

transported through the environment.

Concentrations of tritium in the atmosphere at the

points of possible human exposure were calculated from

the quantities released with gaseous effluents using the

Air Transport model. Three pathways-air inhalation,

transpiration, and drinking water-were the major means

by which tritium was calculated to contribute to human

radiation dose (it should be noted that transpiration

dose was taken as equal to the inhalation dose as

discussed previously).

The total tritium release estimated in the region, 7.1

million curies per year in the year 2000, is primarily

released to the air; most of this is from reprocessing

plants. The release rate to the water (32,500 curies per

year) accounts for less than 0.6% of the total tritium

release rate to the region.

The degree to which the air releases effect the

concentrations of 3H in drinking water was investigated.

This study, reported in Chapter X of this report, showed

the air-water coupling for tritium to be small.

Krypton and Xenon

Krypton 85, a noble gas with a long half-life, has been

considered by many investigators to be one of the

"problem" radionuclides. Noting that calculations in this

study do not include the long-term buildup in the earth's

atmosphere, Krypton 85 was not found to make a

significant contribution to potential total-body dose. It

did make a somewhat larger contribution to the skin

dose, as one would expect, since most of the decay

energy of 85Kr is from beta radiation.

The Krypton 85 released from LWR fuel was assumed

to be essentially unaffected by the radwaste system.

Both LMFBRs and reprocessing plants were assumed to

employ an advanced gas treatment system with an

effective decontamination factor of 300 for kryptons

and xenons. This assumed decontamination factor (DF),

attainable with a refrigerant adsorption or cryogenic

system with provisions for bottling, reduces 85 Kr

releases from reprocessing plants to the same order of

magnitude as those from LWRs. The LMFBRs releases

are correspondingly less. If the LMFBR did not employ

the advanced gas waste system, the 8 5

Kr releases from it

D-3
would be slightly larger than from a LWR because of the

higher fuel burnup in the LMFBR and the resulting

effects on gas diffusion from the fuel.

Because of the assumed removal of 8 5 Kr from the

reprocessing plant gases, it was not found to contribute a

significant part of the dose for any organ except the

skin.

Xenon-133, with a five-day half-life, emits both beta

and gamma radiation in its decay. It contributes

significantly via air submersion to the dose to skin

(although it contributes less than 8 5Kr). It has smaller

but discernible contributions to the dose to total body

and to the internal organs.

BWRs in the region were assumed to be equipped

with charcoal beds which provided a seven-day delay for

xenon isotopes and a 15-hour delay for krypton

isotopes. This delay reduced the 1 33 Xe releases by a

factor of more than two. Releases of 8 5 Kr were not

reduced by- this delay, but those of the shorter-lived

isotopes of both krypton and xenon were significantly

reduced.

Strontium-90

Strontium-90 is also generally considered to be one of

the radionuclides of concern. Demineralizers in the

treatment systems of the light-water reactors and

reprocessing plants were assumed to be effective in

removing strontium from the process streams; thus, the

release rates were low. Strontium was also removed from

the LMFBR coolant by the cold traps, reducing the

gaseous releases from the LMFBR.

Strontium is a long-lived, bone-seeking element,

however, and much of the concern is over this aspect of

its behavior. Its contribution, however, appears to be

relatively minor; overall, it contributes about 2% of the

dose rate to bone. In the calculations it appears as a dose

contributor to the GI tract from the air inhalation

pathway because of its 90 Y daughter.

The relative contributions to dose commitment for

bone, however, are somewhat different. About 50% of

the commitment to bone at typical centroids results

from 90Sr. This high percentage results from the

combination of the long radiological half-life of 90 Sr

and its long biological half-life in bone. It must be

remembered, however, that the dose and dose

commitment calculated by the current model represent

only those from one year's intake and do not include the

additional dose (or dose commitment) from previous or

future intake and retention. A measurably higher bone

dose from 9 0 Sr (and also from certain transuranium

nuclides) would in fact be calculated if several years'

intake were compounded into a total bone burden.

Inclusion of previous year contributions to year 2000

bone burden would increase the year 2000 bone dose

rate due to 9 0Sr by a factor of 11 and would increase

the bone dose rate by 15 percent. Other radionuclides in

the original list of 45 have sufficiently short effective

(radiological or biological) half-lives so that the organ

burden after several years' intake is not significantly

higher than after the first year's intake.

Zirconium and Niobium

Zirconium and Niobium were found to make

relatively small contributions to certain organ doses.

9 5 Zr is formed as a fission product and as a clad

activation product in the LWR. 95 Nb is formed directly

as a fission product and also is the daughter of 95 Zr

decay. The clad corrosion rate in LWRs is low and

reactor releases of these radionuclides are low. They are,

however, released from the reprocessing plant stack in

sufficient amounts to make them contributors to the

doses to the total body, GI tract, bone, skin and liver

from external exposure to soil. They also appear as

contributors to the GI tract dose from both inhalation

and air submersion.

Iodine-129 and -131

1291 and 1311, as would be expected, are of principal

interest in relation to thyroid dose. Treatment systems

of all reactor types remove iodine from the coolants and

radwaste process streams. Reprocessing plants have

treatment system components designed particularly for

iodine removal. Nevertheless, because of the affinity of

the thyroid for iodine, 131I still contributes measurably

to the thyroid dose. Iodine-129 is of interest because of

its long radioactive half-life. It does not make a

significant contribution to thyroid dose rate due to the

short biological half-life (in the child's thyroid 13 days).

The base calculations included an environmental

concentration contribution due to operation of the

nuclear industry in past years. The year 2000 dose rate

value is for an individual in equilibrium with this

concentration buildup due to the short biological

half-life of iodine.

Four pathways contributed notably to iodine doses in

the thyroid: drinking water, air inhalation, fresh milk,

and beef and lamb ingestion. The air inhalation pathway

contributed the majority of the average adult dose and

about 50% of this pathway's contribution was 131I. For

the child, with a proportionately larger milk consumption and a smaller thyroid gland, both the dose and the

fraction of the dose coming from the milk pathway were

higher.

Cesium-134 and -137

13 4 Cs and 13 7

Cs are major contributors to a number

of organ doses by a variety of pathways. They are

assumed to be released in LWR liquid effluents and in

LMFBR and reprocessing plant gaseous releases. Even

though effective removal by treatment system

components is assumed at all plants, the amounts

released are sufficient to make significant contributions

to dose. The principal pathways followed by cesium are

drinking water (liquid release from LWRs) and exposure

to soil (gaseous release from LMFBRs and reprocessing

D-4
plants). The cesium isotopes make their most substantial

contributions to dose to bone, liver, total body, and GI

tract.

Other Radionuclides

A number of the original 45 radionuclides have short

half-lives and are eliminated by decay between the

generation point and the receptor, usually prior to their

release. These included 13 N, 41 Ar, 8 5mKr, 8Kr, 1321*,

1331, 13 5 1,131Xe,1 3 3 mXe,135 mXe, andl13 8 Xe, none

of which contributed to dose. In a sensitivity study

which showed the effect of not having advanced

treatment systems on all plants, the short-half-life noble

gases were found to contribute to doses in centroids

close to the plant site (Chapter X).

Another group of the original nuclides are activation

products: 5 1

Cr, 5 4 Mn, 5 5Fe, 5 8 Co, 6 0Co, 6 3 Ni, 6 4 Cu,

6 5 Zn. These nuclides are present only in liquid effluents

and are efficiently removed by demineralizers in the

coolant cleanup system and radwaste systems. The

nuclides 64 Cu and 6 5Zn, which have been troublesome

in some early nuclear power plants, are assumed to have

been eliminated by retubing condensers and feedwater

heaters with materials which reduce the generation and

release of these nuclides (e.g., stainless steels). This trend

is apparent in the design of latest-generation LWRs.

The remaining nuclides have a combination of

characteristics which result in low contributions to

potential dose. Carbon-14 was included on the original

list primarily for consideration if nitride fuels were

included. Since only oxide fuels were considered in the

study, the releases of 14C are very low.

The sodium isotopes 2 2 Na and 2 4 Na were included

primarily for consideration as activation products of

LMFBR coolant. The release rate of sodium through the

LMFBR gas treatment system was felt to be very small.

In addition, 24 Na also has a short half-life.

39 Ar and 4 1 Ar were included since argon may be

used as a cover gas in the LMFBR. Release estimates of

these two radionuclides were small enough that a

notable dose contribution was not calculated.

The strontium isotopes, 8 9 Sr and 9 0 Sr, are also

assumed to be effectively treated by the coolant cleanup

and radwaste systems. As mentioned earlier, 9 0 Sr does

contribute somewhat to dose. The effect of 8 9Sr is

reduced because its half-life is significantly shorter than

that of 9 0Sr. This shorter half-life reduces long-term

*The 1321 is, however, maintained in the environment as a

decay product of 1 3 2Te (78 hour half-life).

**Since 14C is formed mainly by (n,p) reaction on 1 4N, the use

of fuels containing more than trace amounts of nitrogen could

significantly increase releases of this nuclide.

buildup in soil, lowers concentrations in stored foods,

and lowers the dose per picocurie intake relative to 90 Sr.

The nuclide 99 Mo is found in relatively large amounts

in reactor coolants but is not present in gaseous

effluents. It has a half-life of only 67 hours and is

effectively removed by the treatment systems. As a

result, it does not contribute measurably to the calculated radiation dose.

Ruthenium is removed efficiently by the treatment

system components. The shorter half-life of the 10 3 Ru

(40 days) compared to that of 106 Ru (1 year) results in

a reduction of the ratio of these two nuclides in the

environment. In addition the dose factors for 1 0 6 Ru for

the various organs are as much as ten times higher than

those for 103Ru because of the significantly higher

decay energy of 1 06 Ru. The difference in radiological

half-lives of the two isotopes does not significantly

affect the relative magnitude of the dose factors since

the biological half-life of ruthenium in most body organs

is short (for example, 7.3 days in the total body and 16

days in the bone).

Because of the short half-life (78 hours) of 1 3 2 Te and

because of the assumed effectiveness of the treatment

systems for removal of tellurium, this nuclide does not

contribute significantly to dose.

The nuclides 14 0 Ba, 14 0 La, 1 4 1 Ce, and 1 4 4Ce were

all assumed to be effectively removed by treatment

system components. The half-life of 14 4Ce (290 days) is

considerably longer than those of the other three, and it

was the only one to make even a small contribution to

dose. The major part of this dose is from the 144Pr

daughter of 14 4 Ce. Since both of these nuclides are rare

earths and 14 4Pr has only a 17 minute half-life, the

parent and daughter are in radioactive equilibrium both

in the environment and in the body. Cerium-144 does

make a small contribution to dose via the air-inhalation

pathway, mainly to the GI tract, since it is classed as

insoluble.

Dose factors and dose commitment factors provide a

means of calculating radiation dose to a person (or

organ) from the intake of radionuclides (internal dose

factor) or from exposure to a given concentration of

radionuclides in environmental media (external dose

factor). In this report the value calculated using the dose

factors in the following tables represents a dose received

over a twelve-month period from the amount of material

taken into the body. In this study intake is calculated

for each month of the year and the resulting dose values

are summed for the annual dose rate. For external

exposure, the dose factors are used to calculate the dose

rate received during the time of exposure. The value

calculated using a dose commitment factor represents

the dose received in the ensuing 50 years from the intake

calculated for the study year.

D-5
TABLE D-1

INHALATION DOSE FACTORS, AVERAGE CHILD

(mrem/microcurie) 

H3

N13

C14

NA22

NA24

AR39

AR41

C951

MN54

FE55

FE59

CO58

C060

163

CU64

ZN65

KR8SM

KR8S

KR87

KR88

SR89

SR90

LR9b

N895

M099

RU103

KU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

CS134

C5137

1A140

LA140

CE141

CE144

T BODY

.2030

2.2561

45.0430

4.6498

2.1117

6.2607

.5755

1.8914

1.1604

58.5118

33.8414

THYRD

.2030

2.2561

45.0430

4.6498

GI-LLI

.2039

2.2561

1.2835

2.5644

.2930

6.2020

.6203

17.9249

9.7839

25.3034

1.4693

9.7770

4.5696

45.6383

93.0545

15.4894

8.9577

34.0690

11.4039

118.3329

68.4261

.1580

.7167

1.0086

1.4666

1.3681

1.0202

.921b

25.2426

62.1482

14.6560

107.7167

BONE

2.2561

45.0430

4.6498

8.0353

12.3184

.6407

4.4319

1.5045

LUNGS

.2030

2.2561

45.0430

4.6498

4.6158

413.8736

22.0939

339.6559

305.2507

1673.3727

56.1680

2.6972

272.1261

596.2059

3557.3090

520.8785

151.8350

36.7632

168.6832

3680.4301

108.2768

148.0564 28.3667

196.1583 23.0194

473.8368

50.1187

133.1946

3175.6015

*1 curie= 106 microcuries

1 microcurie = 106 picocuries

These tables were computer generated, only the first two figures are significant.

D-6

SKIN

.2030

2.2561

LIVER

.2030

2.2561

45,0430

'4.6498

1847.0400

3107.9999

182.7800

1126.8719

407.5920

4.2222

12.6553

1.2296

5.2250

2.6125

258.1007

204.7833
TABLE D-2

INHALATION DOSE COMMITMENT FACTORS, AVERAGE CHILD

(mrem/microcurie)

THYRD

.2030

2.2561

45.0430

4.6498

BONE

2.2561

45.0430

4.6498

LUNGS

.2030

2.2561

45.0430

4.6498

SKIN

.2030

2.2561

LIVER

.2030

2.2561

45.0430

4.6498

4.6158

437.7840

24.3874

339.8080

306.3600

1873.0879

63.9360

2.6972

284.1600

H3

N13

C14

NA22

NA24

AR39

AR41

CR51

MN54

FL55

FE59

C058

C060

N163

CU64

1N65

KR85M

KR85

KR87

KR88

SR89

SR90

4R95

NB95

M099

RU103

kU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

C5134

C5137

8A140

LA140

CE141

CE144

8.0353

12.3184

.6407

4.4319

1.5045

596.7360

4049.2798

522.1440

151.8480

36.7632

168.7200

3916.0798

108.2768

4.2222

12.6553

1.2296

5.2250

2.6125

269.7509

217.4623

167.7183 32.1338

233.5061 27.4022

473.8368

50.1187

133.2000

3346.2799

D-7

T BODY GI-LLI

.2030

2.2561

45.0430

4.649$

1847.0400

3107.9999

182.7800

1126.8719

407.5920

2.1117

6.2607

.5755

1.8914

1.1604

58.5119

33.8415
TABLE D-3

INHALATION DOSE FACTORS, AVERAGE TEEN

(mrem/microcurie)

THYRD

.1057

.7551

17.8166

1.8349

M3

113

C14

NA22

NA24

AR39

AR41

CR51

MN54

FE5b

FE59

C058

C060

NI63

CU64

LN65

KR85M

KR85

KR87

KR88

SR89

5R90

LR95

NB95

M099

RU103

RU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

C5134

C5137

BA140

LA140

CE141

CE144

BONE

.7551

17.8166

1.8349

2.6894

4.2045

.2243

1.4964

.5230

LUNGS

.1057

.7551

17.8166

1.8349

2.6297

239.2576

11.5763

190.7427

171.6177

958.2738

29.0524

1.3951

152.4845

SKIN

.1057

.7551

308.3824

1839.9874

288.6642

90.9357

19.0154

96.4342

1916.6209

59.8022

1.8648

5.7964

.6169

2.4359

1.2887

132.9936

95.6143

53.3291 15.8951

67.2287 12.2468

255.7440

27.4484

68.8938

1642.5525

D-8

LIVER

.1057

.7551

17.8166

1.8349

T BODY

.1057

.7551

17.8166

1.8349

.8325

2.4148

.2238

.6966

.4495

35.4689

19.0244

710.3999

1133.8773

66.5507

388.3520

146.5200

GI-LLI

.1057

.7551

1.4396

2.1253

.3539

7.8502

.6030

20.5195

11.8624

29.3594

1.4285

7.4544

5.7081

44.2289

90.4682

16.6018

11.0315

31.9315

11.8292

115.7869

69.8212

.1536

.7135

.1917

1.2031

.9422

1.1196

.9600

22.2041

58.8312

14.1772

104.6632
TABLE D-4

INHALATION DOSE COMMITMENT FACTORS, AVERAGE TEEN

(mrem/microcurie)

THYRD

.1057

.7551

17.8166

1.8349

BONE

.7551

17.8166

1.8349

LUNGS

.1057

.7551

17.8166

1.8349

SKIN

.1057

.7551

LIVER

.1057

.7551

17.8166

1.8349

2.6297

253.0816

12.7786

190.8281

172.2414

1072.7387

33.0745

1.3951

159.2279

H3

N13

C14

NA22

NA24

AR39

AR41

CR51

MN54

FE55

FE59

Cubs

C060

NI63

CU64

ZN65

KR8SM

KR85

KR87

KR88

SR89

SR90

ZR95

N895

M099

RU103

RU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

CS134

CS137

BA140

LA140

CE141

CE144

2.6894

4.2045

.2243

1.4964

.5230

308.6565

2094.7197

289.3655

90.9434

19.0154

96.4552

2039.3535

59.8022

1.8648

5.7964

.6169

2.4359

1.2887

60.4197 18.0083

90.0660 14.5845

255.7440

27.4484

68.8965

1730.8412

138.9970

101.5350

D-9

T BODY GI-LLI

.1057

.7551

17.8166

1.8349

710.4000

1133.8773

66.5507

388.3520

146.5200

.8325

2.4148

.2238

.6966

.4495

35.4747

19.0286
TABLE D-5

INHALATION DOSE FACTORS, AVERAGE ADULT

(mrem/microcurie)

THYRD

.1269

.5286

13.2566

1.3605

GI-LLI

.1269

.5286

1.6314

1.9166

.4159

9.5114

.6029

22.8238

13.4206

34.9116

1.4285

5.8989

6.8938

44.1229

90.4671

18.1480

12.5626

30.5353

13.3643

115.7475

72.8504

.1536

.7637

.0566

1.08Q0

.7113

BONE

.5286

13.2566

1.3605

M3

N13

C14

NA22

NA24

AR39

AR41

CR51

MN54

FE55

FE59

C058

C060

i163

CU64

N65

KR8bM

KR85

KR87

KR88

5R89

SR90

4R95

NB95

M099

RU103

RU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

CS134

C5137

1A140

LA140

CE141

CE144

LUNGS

.1269

.5286

13.2566

1.3605

1.8062

.167.9840

6.6282

125.1875

119.4459

663.3728

16.8504

.8567

108.8504

178.8618

1067.1927

193.4692

59.9349

11.2687

61.2586

1119.1512

38.5392

SKIN

.1269

.5286

LIVER

.1269

.5286

13.2566

1.3605

1.3163

4.2377

.4578

1.7496

.9520

100.5834

71.1407

39.6428 10.6375

47.7954 7.6951

159.6624

17.3941

42.1783

952.6804

D-10

4735.4186

1448.7423

59.6381

355.3776

135.8196

T BODY

.1269

.5286

13.2566

1.3605

4.5038

2.5308

.1667

.5311

.3419

66.2300

42.3205

1.8826

2.9718

.1639

1.0613

.3782

1.2794

1.0240

20.1424

57.7968

14.1237

104.6172
TABLE D-6

INHALATION DOSE COMMITMENT FACTORS, AVERAGE ADULT

(mrem/microcurie)

THYRD

.1269

.5286

13.2566

1.3605

H3

N13

C14

NA22

NA24

AR39

AR41

CR51

MN54

FE55

FE59

C058

C060

N163

CU64

ZN65

KR8SM

KR85

KK87

KR8d

SR89

SR90

iR95

NB95

M099

RU103

RU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

C5134

CS137

BA140

LA140

CE141

CE144

BONE

.5286

13.2566

1.3605

1.8826

2.9718

.1639

1.0613

.3782

LUNGS

.1269

.5286

13.2566

1.3605

1.8062

177.6888

7.3162

125.2435

119.8800

742.5456

19.1808

.8567

113.6640

SKIN

.1269

.5286

179.0208

1214.7839

193.9392

59.9400

11.2687

61.2720

1190.8080

38.5392

LIVER

.1269

.5286

13.2566

1.3605

1.3163

4.2377

.4578

1.7496

.9520

105.1235

75.5453

44.9074 12.0502

56.8954 9.1602

159.6624

17.3941

42.1800

1003.8839

D-11

T BODY GI-LLI

.1269

.5286

13.2566

1.3605

5637.0237

1448.7423

59.6381

355.3776

135.8196

5.3613

2.5308

.1667

.5311

.3419

68.6931

45.9857
TABLE D-7

INGESTION DOSE FACTORS, AVERAGE CHILD

(mrem/microcurie)

T BODY GI-LLI THYRD BONE LUNGS SKIN LIVER

H3 .2030 .2030 .2030 .2030 .2030 .2030

N13

C14 2.2561 2.2561 2.2561 2.2561 2.2561 2.2561 2.2561

rA22 60.0573 2.5671 60.0573 60.0573 60.0573 60.0573

NA24 6.1998 5.1288 6.1998 6.1998 6.1998 6.1998

AR39

AR41

CR51 .0090 .4726 .0050 .0077

MN54 2.8508 9.0029 10.9183

1E55 .6747 .9005 2.8935 .3682 2.5171

FE59 12.6823 26.0200 15.5995 5.6609 26.4298

C058 5.5717 11.0463 1.8883

C060 15.6889 28.5684 5.1995

N163 5.1084 1.6589 128.6428 9.9407

CU64 .1495 11.3539 .2554

ZN65 17.2183 6.6333 9.8587 37.8653

KR85M

KR85

KR87

KR88

SR89 37.8802 51.5271 142.3238

SR90 381.8032 137.3173 1430.1810

ZR95 .0217 24.9829 .1024 .0021

N895 .0060 14.4479 .0197 .0080

M099 3.3787 10.2233 13.6674

RU103 .2770 17.8415 1.1956

RU106 1.4912 185.1338 11.6631

TE132 5.8659 82.7735 6.8376 10.3230 4.7779

1129 2.8156 .3160 4617.5998 10.6127 5.6296

1131 8.3476 1.4333 7769.9997 16.2696 16.8737

1132 .7673 2.0172 456.9500 .8462 1.6394

1133 2.5219 2.9332 2817.1799 5.8534 6.9666

1135 1.5472 2.7362 1018.9800 1.9870 3.4833

XE131

XE133M

XE133

XE135M

XE135

XE138

C5134 78.0157 2.0405 197.4085 37.0000 361.3410

C5137 45.1218 1.8432 261.5444 30.0253 286.6966

BA140 4.8763 38.6782 86.0117 .0395 .0728

LA140 .0012 100.2390 .0107 .0036

CE141 .0027 23.6386 .0361 .0182

CE144 .0848 173.7365 1.4022 .5579

D-12
TABLE D-8

INGESTION DOSE COMMITMENT FACTORS, AVERAGE CHILD

(mrem/microcurie)

T BODY

.2030

2.2561

60.0573

6.1998

.0090

2.8508

1.5245

12.7163

5.5717

15.6889

15.6612

.1495

23.6348

GI-LLI THYRD

.2030

2.2561

60.0573

6.1998

BONE LUNGS

.2030

2.2561

60.0573

6.1998

.0050

M3

N13

C14

NA22

NA24

AR39

AR41

CR51

MN54

FE55

FE59

C058

C060

N163

CU64

ZN65

KK85M

KR85

KR87

KR88

SR89

SR90

ZR95

NB95

M099

RU103

RU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

C5134

C5137

bA140

LA140

CE141

CE144

78.0159

45.1219

4.8763

.0012

.0027

.1155

6.8376

4617.5998

7769.9997

456.9500

2817.1799

1018.9800

8.4006

15.6487

451.4811

13.9426

143.2712

17039.0670

.1040

.0197

1.1956

11.6631

10.3230

10.6127

16.2696

.8462

5.8534

1.9870

223.6244

311.3414

86.0117

.0107

.0361

2.1677

2.2561

60.0573

6.1998

.0077

.4065

5.6635

SKIN

.2030

2.2561

41.9136

35.7420

.0395

D-13

LIVER

.2030

2.2561

60.0573

6.1998

10.9185

4.8897

26.4913

1.8883

5.1995

24.6468

.2554

38.7034

38.1299

4342.6808

.0220

.0060

3.3787

.2770

1.4912

5.8659

2.8156

8.3476

.7673

2.5219

1.5472

.0021

.0080

13.6674

4.7779

5.6296

16.8737

1.6394

6.9666

3.4833

377.6513

304.4472

.0728

.0036

.0182

.6778
TABLE D-9

INGESTION DOSE FACTORS, AVERAGE TEEN

(mrem/microcurie)

w O IAIA

.4 NNE-U

N

z

-4

Z

J

0

.4

t

U,

O

O *P- 000.4.4

Sf% .4 41 t " N O'

W1

InU,

u, In IA In

MItN

N

z

O

O

(O

NO%

M7 .4.4

I',

.4 IA

Wa IAUto z I*-I"

N

0

MJ

-I

(9

0

0

c.-

PtI

O

0

.4

vIA

O

O

0

.4

W.4LU

LA LA a NA 4t

0144

N

NU

to IA in LA

fP- 4

N

N

U'U, O D 0.4 I 0% InU A 0 O% IA O N IA ~ 0N U, It) U U 

0

N

0

0

C onovhr'.. to u .O r-f N N#0@0 O 0

.4 0'It)It)400N

. (VJ.4 

iD t vi f*0O0 0IA in U OI)00' V) 0 .' N -+ 0n COU't

0.4N0DN t'+00

.4 IAN 94 FO 

0D oN .4-.40NV)UN'f hnN .OONV)0 N

O M')C OOO (ftn0 ! O PAN 0; 0( Ln W4~D e@g. S~egeegeg e It)It)It)I .4 ee e e 

U' nnha

OD00% 0' .4 to 00N f - fo .4N%o

LAIt) U'EN0.-U r%' -4 I

S 0% N.r-. 3-)0O4"'rN

.-4N

0 0'Ovi'40UNN4 N 

ItNP.'C'(OLADONOOO

O; DF"': ;0 0.4U' .4N "r :tNN.-4 .0 f.4N"4 .4o0

01'.-ONOUPfl--00'a000

I"00o000f".-4"'4 0NO

0Ovi'00D.O4.4LA)N4cN'LAf

N f- .4 N.4t

-4N

"-I

Z )'D N -4I')It)IALAO'f"0O 4- NU'0'.4-4 U 0 ' & 00 ALALAI"-t00'01f A0'O00 I ) OoO ' CAN-- .4N V It)n n pr)W) ) 0' S N N n IA n n to IAO NOO %Oc0 co 0 co 0 O C% O ON f " "- M n n a a M -1 4 -4 4 '4 -4 .e4 -4 e- M

MIt).4.4 4 < ZaJJOd0 - 3z EEE 3 .- 4-4 -4-"4IJ WJWLJAJLIIJ 4 4LaJIwi

ZZ JZ Z44uZLLUUZ U MMYeJI NZZE b -4*-M C'-(UMXUU.JUU

D-14

00'o

004

Oo'0 0 0 

Id OlA O

t@0NN.4

o)*I G0Nl*

NNE.K).4

C00OOfln* 0 %0)"0 U -40 000 N

'O4-cin W4 W4 .4.4O

MtiN

Ot!

N "-

aP- 4 

0) O' W4D 4(9 NOO' S ee @5 g

M(701

r GCN

K)( N 0O%0-4

e g e egg

0' W-4LU)NO.40 0' .4LAnO'

0% D4 OON NI')co 000 S e S eg

U N
TABLE D-10

INGESTION DOSE COMMITMENT FACTORS, AVERAGE TEEN

(mrem/microcurie)

T BODY

.1057

.7551

23.7555

2.4465

.0036

1.1743

.5515

5.0299

2.2834

6.3700

5.9958

.0546

9.7073

H3

N13

C14

NA22

NA24

AR39

AR41

CR51

MN54

FE55

FE59

C058

C060

N163

CU64

ZN65

KR85M

KR85

KR87

KR88

SR89

5R90

ZR95

NB95

M099

RU103

RU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XEI35

XE138

CS134

CS137

BA140

LA140

CE141

CE144

47.2996

25.3714

1.8421

.0005

.0010

.0388

GI-LLI THYRD

.1057

.7551

23.7555

2.4465

.0020

2.4568

1314.2399

2097.6730

123.1187

718.4512

271.0620

BONE

.7551

23.7555

2.4465

3.1734

5.5690

178.9124

6.1025

LUNGS

.1057

.7551

23.7555

2.4465

.0044

.2130

3.1805

SKIN

.1057

.7551

48.1233

10216.4347

.0371

.0072

.4001

3.9090

3.5961

3.5520

5.5532

.2963

1.9764

.6907

80.5595

106.7546

29.2966

.0037

.0121

.7334

23.4890

19.0233

.0213

D-15

12.7618

2556.6088

.0086

.0025

1.1961

.1067

.5056

2.2774

1.1100

3.2198

.2984

.9288

.5994

LIVER

.1057

.7551

23.7555

2.4465

5.9570

2.2436

13.3718

.9790

2.7886

11.8446

.1194

19.9430

.0011

.0043

6.1677

2.3785

2.4864

7.7286

.8225

3.2479

1.7183

194.5958

142.1490

.0350

.0018

.0085

.2996
TABLE D-11

INGESTION DOSE FACTORS, AVERAGE ADULT

(mrem/microcurie)

T BODY

.1269

.5286

17.6754

1.8141

.0027

.8734

.1601

3.7366

1.6783

4.7302

1.1968

.0398

5.2293

GI-LLI

.1269

.5286

3.2627

3.8333

.6693

13.8069

.8752

33.1313

15.1523

39.4163

1.6128

6.8504

10.0071

H3

N13

C14

NA22

NA24

AR39

AR41

CR51

MN54

FESS

FE59

C058

C060

N163

CU64

ZN65

KR85M

KR85

KR87

KR88

SR89

5R90

ZR95

NB95

M099

RU103

RU106

TE132

1129

1131

1132

1133

1135

XE131

XE133M

XE133

XE135M

XE135

XE138

C5134

CS137

bA140

LA140

CE141

CE144

88.3067

56.4274

1.3461

.0003

.0007

.0199

2.5587

2.0479

30.8633

93.2207

22.7802

168.7375

THYRD

.1269

.5286

17.6754

1.8141

BONE

.5286

17.6754

1.8141

.0016

.6779

4.1174

30.1392

3.4646

1.9092

6176.6329

1889.6639

77.7888

463.5360

177.1560

LUNGS

.1269

.5286

17.6754

1.8141

.0030

.1105

2.0865

33.3444

335.0710

.0270

.0056

.2801

2.7401

2.6089

2.4864

3.9250

.2165

1.4017

.4995

52.8571

63.7272

20.8639

.0026

.0085

.3290

13.8750

10.0370

.0133

D-16

SKIN

.1269

.5286

8.8748 49.8162

89.4510 121.8516

.0063 29.2710

.0018 20.2622

.8525 8.9546

.0787 20.9087

.3562 181.0888

1.6766 88.1255

6.0051 .3072

3.3744 1.5273

.2223 .1133

.7082 2.1699

.4558 1.4226

LIVER

.1269

.5286

17.6754

1.8141

4.6053

.7847

10.1233

.7679

2.1421

3.0992

.0843

15.7401

.0008

.0033

4.3536

1.7673

1.7551

5.6503

.6104

2.3328

1.2693

140.8167

99.5969

.0255

.0013

.0060

.1739
TABLE D-12

INGESTION DOSE COMMITMENT FACTORS, AVERAGE ADULT

(mrem/microcurie)

T BODY GI-LLI THYRD BONE LUNGS SKIN LIVER

113 .1269 .1269 .1269 .1269 .1269

N13

C14 .5286 .5286 .5286 .5286 .5286 .5286

NA22 17.6754 17.6754 17.6754 17.6754 17.6754

1A24 1.8141 1.8141 1.8141 1.8141 1.8141

AR39

AR41

CR51 .0027 .0016 .0030

MN54 .8734 4.6054

FE55 .3625 1.9892 .1219 1.5251

FE59 3.7466 4.1304 2.0874 10.1468

C058 1.6783 .7679

C060 4.7302 2.1421

N163 3.7220 108.7039 7.7099

CU64 .0398 .0843

LN65 7.1780 4.8999 16.0885

KR85M

KR85

KR87

KR88

5R89 8.9333 33.5664

SR90 1836.7080 7448.1297

LR95 .0064 .0274 .0008

NB95 .0018 .0056 .0033

M099 .8525 4.3536

RU103 .0787 .2801

RU106 .3562 2.7401

TE132 1.6766 1.9092 2.6089 1.7673

1129 7.1484 7352.6397 2.4864 1.7551

1131 3.3744 1889.6639 3.9250 5.6503

1132 .2223 77.7888 .2165 .6104

1133 .7082 463.5360 1.4017 2.3328

1135 .4558 177.1560 .4995 1.2693

XE131

XE133M

XE133

XE13bM

XE135

XE138

CS134 91.5909 59.8766 15.7176 147.1729

CS137 61.3143 75.8606 11.9480 105.7634

BA140 1.3461 20.8639 .0133 .0255

LA140 .0003 .0026 .0013

CE141 .0007 .0085 .0060

CE144 .0271 .5086 .2113

D-17
TABLE D-13

DOSE FACTORS FOR SUBMERSION IN AIR

(mrem/hr per pCi/m 3)(a)

Nuclide Skin Total-Body(b) Nuclide Skin Total-Body(b)

H-3(c) 0 0 Mo-99 5.8 E-7 2.2 E-7

N-13 1.4 E-6 8.5 E-7 Ru-103 4.9 E-7 4.1 E-7

C-14(c) 3.5 E-9 0 Ru-106 1.6 E-6 1.7 E-7

Na-22 2.3 E-6 1.8 E-6 Te-132 2.3 E-7 1.8 E-7

Na-24 4.5 E-6 3.5 E-6 1-129 4.0 E-9 1.1 E-9

Ar-39(d) 1.2 E-7 3.3 E-10 1-131 4.9 E-7 3.1 E-7

Ar-41 1.6 E-7 1.1 E-6 1-132 2.7 E-6 2.0 E-6

Cr-51 2.9 E-8 2.4 E-8 1-133 8.8 E-7 4.4 E-7

Mn-54 8.2 E-7 7.0 E-7 1-135 2.0 E-6 1.5 E-6

Fe-55(d) 1.5 E-10 2.7 E-11 Xe-131m 4.8 E-8 2.8 E-9

Fe-59 1.2 E-6 1.0 E-6 Xe-133m 6.0 E-8 2.7 E-8

Co-58 1.1 E-6 8.2 E-7 Xe-133 6.9 E-8 2.5 E-8

Co-60 2.5 E-6 2.0 E-6 Xe-135m 5.0 E-7 3.5 E-7

Ni-63 0 0 Xe-135 4.9 E-7 2.1 E-7

Cu-64 2.8 E-7 1.7 E-7 Xe-138 1.7 E-6 1.2 E-6

Zn-65 5.6 E-7 4.9 E-7 (Cs-138) (3.1 E-6) (1.8 E-6)

Kr-85m 3.2 E-7 1.3 E-7 (Xe+Cs-138) (4.8 E-6) (3.0 E-6)

Kr-85(d) 1.5 E-7 2.2 E-9 Cs-134 1.7 E-6 1.3 E-6

Kr-87 2.7 E-6 1.3 E-6 Cs-137 7.0 E-7 4.7 E-7

Kr-88 2.0 E-6 1.5 E-6 Ba-140 4.4 E-7 2.2 E-7

(Rb-88) (2.7 E-6)(e) (5.6 E-7) La-140 2.7 E-6 1.9 E-6

(Kr+R b-88) (4.7 E-6) (2.1 E-6) Ce-141 1.5 E-7 5.0 E-8

Sr-89(d) 4.9 E-7 2.1 E-7 Ce-144 4.1 E-8 1.4 E-8

Sr-90(d) 1.3 E-7 2.4 E-10 (Pr-144) (1.2 E-6) (2.6 E-8)

(Y-90) (8.7 E-7) (5.8 E-7) (Ce+Pr-144) (1.2 E-6) (4.0 E-8)

(Sr+Y-90) (1.1 E-6) (6.0 E-9)

Zr-95 8.4 E-7 6.8 E-7

Nb-95 7.5 E-7 6.4 E-7

(a) Same factors apply for adult, teen, and child.

(b) Total-body factors also apply to internal organs.

(c) Not including penetration of oxide into skin.

(d) Including bremstrahlung.

(e) Values in parentheses are for radionuclides not on the reference list (Appendix C) but which are daughters of radionuclides on the

list.

D-18
TABLE D-14

EXTERNAL DOSE FACTORS FOR IMMERSION IN WATER

(mrem/hr per pCi/liter) (a)

Nuclide

H.3(c)

N-13

C-14(c)

Na-22

Na-24

Cr-51

Mn-54

Fe-55(d)

Fe-59

Co-58

Co-60

Ni-63

Cu-64

Zn-65

Sr-89(d)

Sr-90(d)

y-90(d)

(Sr+Y-90)(d) (e)

Skin

0

2.6 E-6

3.8 E-6

4.8 E-6

9.3 E-6

6.4 E-8

1.8 E-6

3.6 E-10

2.6 E-6

2.3 E-6

5.4 E-6

0

5.2 E-7

1.2 E-6

5.4 E-7

1.5 E-7

9.6 E-7

1.1 E-6

Total-Body(b)

0

1.9 E-6

0

4.0 E-6

7.9 E-6

5.2 E-8

1.5 E-6

6.4 E-11

2.2 E-6

1.8 E-6

4.6 E-6

0

3.7 E-7

1.1 E-6

4.6 E-9

5.4 E-10

1.3 E-8

1.3 E-8

Nuclide

Zr-95

Nb-95

Mo-99

Ru-103

Ru-106

Te-132

1-129

1-131

1-132

1-133

1-135

Cs-134

Cs- 137

Ba-140

La-140

Ce-141

Ce-144

Pr-144

(Ce+Pr-144) f)

Skin

1.8 E-6

1.6 E-6

9.1 E-7

1.1 E-6

1.9 E-6

4.8 E-7

6.1 E-9

9.5 E-7

5.5 E-6

1.5 E-6

4.0 E-6

3.5 E-6

1.4 E-6

7.6 E-7

5.3 E-6

2.4 E-7

6.2 E-8

1.3 E-6

1.4 E-6

Total-Body(b)

1.5 E-6

1.4 E-6

4.7 E-7

8.9 E-7

3.8 E-7

4.0 E-7

2.1 E-9

6.8 E-7

4.4 E-6

9.6 E-7

3.3 E-6

2.9 E-6

1.0 E-6

4.6 E-7

4.1 E-6

1.3 E-7

3.0 E-8

5.6 E-8

8.6 E-8

(a) The some factors apply to Adult, Teen and Child.

(b) Total-Body factors also apply to other internal organs.

(c) Not including penetration of oxide into skin.

(d) Includes bremstrahlung.

(e) Use these factors for 90

Sr unless 80Y concentration is given separately.

(f) Use these factors for 14 4Ce unless 144Pr concentration is given separately.

D-19
TABLE D-15

EXTERNAL DOSE FACTORS FOR EXPOSURE WHILE STANDING ON

CONTAMINATED GROUND OR RIVER BANK(a)

(mrem/hr per pCi/m 2

Nuclide Skin Total-Body(b) Nuclide Skin Total-Body(b)

H-3 0 0

C-14 0 0 R u-103(c) 4.2 E-9 3.6 E-9

Na-22 1.8 E-8 1.6 E-8 Ru-106(c) 1.8 E-9 1.5 E-9

Na-24 2.9 E-8 2.5 E-8 Te-132 2.0 E-9 1.7 E-9

Cr-51 2.6 E-10 2.2 E-10 1-129 9.6 E-12 7.0 E-12

Mn-54 6.8 E-9 5.8 E-9 I-131(c) 3.4 E-9 2.8 E-9

Fe-55 0 0 1-132 2.0 E-8 1.7 E-8

Fe-59 9.4 E-9 8.0 E-9 1-133 4.5 E-9 3.7 E-9

Co-58 8.2 E-9 7.0 E-9 1-135 1.4 E-8 1.2 E-8

Co-60 2.0 E-8 1.7 E-8 Cs-138(d) 2.4 E-8 2.1 E-8

Ni-63 0 0 Cs-134 1.4 E-8 1.2 E-8

Cu-64 1.7 E-9 1.5 E-9 Cs-137 4.9 E-9 4.1 E-9

Zn-65 4.6 E-9 4.0 E-9 Ba-140 2.4 E-9 2.1 E-9

Rb-88(d) 4.0 E-9 3.5 E-9 La-140 1.7 E-8 1.5 E-8

Sr-89 6.5 E-13 5.6 E-13 Ce-141 6.2 E-10 5.5 E-10

Sr-90 0 0 Ce-144 1.4 E-10 1.2 E-10

Zr-95 6.2 E-9 5.9 E-9 Pr-144 2.3 E-10 2.0 E-10

Nb-95 6.8 E-9 5.8 E-9 (Ce+Pr-144)(e) 3.7 E-10 3.2 E-10

(a) The same factors apply to Adult, Teen and Child.

(b) Total-Body factors also apply to internal organs-Bone, Thyroid, GI Tract, Liver, Lungs.

(c) Includes daughter radiations.

(d) Not on nuclide list but parent is.

(e) Use these factors for Ce-144 unless Pr-144 daughter concentration is given separately.

D-20
93

162 161 133

227

195 19 34 .- -

16 54

221 196

282165 1e3520

284 229 191 164 9 ?84 22 166 137 95 2

293 131 '0 t418 138 96 55

264 265 Il 1 39 9823 2

_33.a ., t 140 ioo iai 57 5

p 2g5 266 265 202 176 141 142 102

295 58 2

267 i 7 59 25 2

204 143 103 104 60

286 268 235 234 203 178 61 26 3

.. 1 _. -62

269 236 _205 2tl 144 106 105 63 28 27 4

7 1 145S 0 iaa 6d 6

296 0o

26 288 238 208 ' 1 47 148, 103 67 -.

289 273 ' 39 2410 21m 182 183 11 49 12l 68 33 2

29 X4. z 2 43 241 211 210 184 1130

9 29 7 274 15 1 2 0 3 0 

9824 7 246 23 245 1 1 16 7 36

300 29 75276 2428 18 53 1521 18 73 71 3 37I

291 278 Zi 251. 75 74 39 1

277 7 5 .215 11 20 76 40 3

292 253 -- - 14 22 .86 79 42 1

254 118 123 81 43

280 2524

12 8 1584 283 257 5823222 .., .. 187z 8 8

Z60 29224 19 8 n 2 2 8 s 53 16

1 .2x100 to 10 19284 90 88 45 17

mrem/yr 46 47 18

226 192 158 '13' 132 92 4

10- to 10- 49 1

m2rem/yr 5

l10-2 to 10-3

mrem/y r

Figure D-1.-Regional Pattern of Total-Body Annual Dose (Average Adult).

D-21
227

262 22

284 263 229

293

cO* L24L Ji1 1

295

296 238

2 8 

297

298 299 

x00 25

193

62161 133

195 194 134

16163 54

165 13613 20

64137 95 94 21

198 ( 1 138 96 55

199 196823 22

; pr 171 173 139 9f' 99

201 175174 140 100 101 57 5

a. 76 141 142 102 58 2 925 2 177 59

204 a143 103104 60

61 2

7 144 106 105 62 63 28 27 4

145 109 108 64 65 29 30 6 5

208 146 1077

17 : 1 r 36

1507

ill8

S214

214

l x10-2 to 1x10 3

mrem/yr

Figure D-2.-Regional Pattern of GI Tract Annual Dose (Average Adult).

D-22

C
193

62 161 

227

19 94 134

_...19 16 -- .. 54

228 13592

24 263 22 66 17 95 942

2 3231 2311 a; 1

26 25 232 121l7 3 39 9 I23 ?

b 233 01 17 4 100 101 57 56

24 285 26 2 76 141 142 10, 24

295 267 59 2 2

- 204 1 7 43 103 104 0 2

28 268 235 ? 18 61 2

269 236 237 205 2 06 144 lob 105 63 78 ?

287 ; 2 0 07 7 45 log a t 30 5

296 288 38 208 X1 1 7 148 146 110 107 . 3

289 273 ? ?40 st v182 18 149 li 68- 8

297 9 243 24 21. 210 184 l50 ; ;1 570

298 9274 347 2,61 132 5 5 1 7 7

275 27 49 3 18 5 I t21 8 3l

291 27 7 278 ? 1 4 f

292 253 186 154 l22 l 9 42 4

t 23 8 80 4

280 25 4274

188__ 84 83

283 25 5 2 2 2 5 8 6

,3 x100 to 1 x100 2 12 28 8 1

mrem/yr

225 226 192 158 13 9 3 4 48 19

1 x100 to 1x101 19 5

mrem/yr 160

U1 x10- 1 to I x10- 2

mrem/yr

Figure D-3.-Regional Pattern of Thyroid Annual Dose (Average Adult).

D-23
193

162 161 133

227

95 19 34

262 228 96165 .1520

284 263 2?9 197 164 166 137 95 942

93231 230 1968o i 138 9b 55

264 265 232 19394 - 2

?00121113 19 8 9923 ?

265 233 201 1 - 140 100 101 57 56

295 19' 285 266 2665 20,' 176 ldl 142 102 5

29 6 75 25 24 2

204 143103 104 60

286 268 235 234 203 118 1361 26 3

269 236 237 205 206 144 106 105 62 63 28 27 4

281270 207 1 145 109 ]oa 64 65 29 30 6 5

296 288 238 208 180 147 148 146 110 107 66 31 7

289 273 272 239 240 m ?O9 182 183 181 149 112 1 68 3 32 8

13 9 34 10

297 290 ?44 4 243 ?41 211 210 184 150 14 1 5 72 70 is

299 257 4 7249 2185 153 152 118 73 11 38 37 11

2979 25 2 4 215 9 120 757 4 76 40 13

292 253 186 154 122 7 2 i

254 16 123 81 8 43

8280 255 26 1 18io 217 18 s 44

21128 124 15 84 83 15

783 257 258 223 222 26..1-5 8

260 259 224 190 ] sa 89 -87 3

2xl0 1 to 1 x10-1 1 ? 128 9 9088 45 17

mrem/yr 261 1 30 46 47 /8 i

225 2?6 192 58 1 1 i32 92 48 1

F 1x101 to 1 x10- 159

]mrem/yr 1 .5

1x10-2 to 1x10-

Qmrem/yr

Figure D-4.-Regional Pattern of Bone Annual Dose (Average Adult).

D-24
227

262 228

284 263 229

293

M 23

L1x10- 2 to 1x10-3

mrem/yr

Figure D-5.-Regional Pattern of Lung Annual Dose (Average Adult).

D-25

193

161 133

16626

195 194 134

7 163 

54 28 165 tae13

164

132 95 94 21

198 16 138 96 55

19 9 e 168 2 999 23 2

171 173 139 99

201 ti5 te 140 100 101 57 56

l76 141 142 102 5

y.; 59 25 24 2

177 5

143 103 104 60 26 3 61

7 ~44 106 105 6 8 2

145 109 toa 64 65 29 30 6 5

208 146 177
[2j|y 285 ,. r+

295 

3.

2S23

28

2 t

296 21

227

262 228

284 263 229

293

lx10- 2 to 1x10-3

mrem/yr

Figure D-6.-Regional Pattern of Skin Annual Dose (Average Adult).

D-26

L

193

162 161 133

195 194 134

16163 54

165 136135 20

164 137 95 94 21

18 16 138 96 55

1.99 196823 22

171 17 3 139 98 g

996

201 lts lta 140 100 101 57 5

176 141 142 102 5

.,: 8 5925 24 2

13 103 104 60

8 4361 26 3

_. 1..62

7 , 44 106 105 63 28 27 4

145109 log 64 65 29 30 6 5

208 146 10

289 :.. .. 0

297 24? 1241,?11 2It1184 15 114 15

298 2915 1f z X7 7.1

214
193

J-- 161 131

162

227

195 19 34

19 63 54

165 6 i 20

284 7 166 3 95 942

293 99 43 168 23 22

24 265 232 1 71 173 139 9

233 'O I 17 140 100 101 57 56

24 285 26 252 - 76 141 142 102 58 5 5 24 2 2952659 22

267 204 13 103 104 60

786 268 235 234 203 178 4362 61 26 3

269 236 0 206 4 106 105 63 28. 27 4

2$7 21 270 2 7 19 14 5109 ios cs 55 29 30 6 5

28 881811 14 7 148 10 631

289 273 :2 39 240 tr 182 183 181 149 112 il 68 33 2 8

291 zaz 24 3 241 211 210 J84 150 114 7 0 34 0

247 246 213 245 151 11 71 71 8 3 2 9 2 274 249 135 15 5 8 7

300 276 248 

181 2 8 7

30039 12

291 278 5 5 '5 74

292 253 1 154 22 79 42

280 255 254 s4

256 124. ez $ 44

z 188 ts ls 84 83 1

283 257 5$23 222 126 I27 85 86

264 5 224 10 84 i89 -9 53 1

-x ^1 t 1 -1 129 128 9 9088 45 11

mrmyr11 3 4 7 8225 226 192 158 Ii 12 9 49 48 14

1 x10~ to 1 x10 > s 1 t

m rem/yr

1 x10-2 to 1 x10-3

mrem/yr

Figure D-7.-Regional Pattern of Liver Annual Dose (Average Adult).

D-27
1.5

1.2 TOTAL BODY DOSE

" AVERAGE ADULT

o AVERAGE CHILD

0.9

J

r-J

0.6

0.3

0.0

0 10 20 30 40 50 60 70 80 90 100

POPULATION (PERCENT)

Figure D-8.-Average Individual Total-Body Dose Distribution.
1.5

GI TRACT DOSE

1.2

o AVERAGE CHILD

* AVERAGE ADULT

0.9

J

o 0.6

0.3

0.0

0 10 20 30 40 50 60 70 80 90 100

POPULATION (PERCENT)

Figure D-9.-Average Individual GI Tract Dose Distribution.
THYROID DOSE

5.0

4.5

4.0

3.5

3.0

2.5

2.0

1.5

1.0

0.5

0.0

10 20 30 40 50 60

* AVERAGE ADULT

AVERAGE CHILD

- ii Ill~ll

70 80 90 100

POPULATION (PERCENT)

Figure D-10.-Average Individual Thyroid Dose Distribution.

d

J

J

r-4

V.)

0

m

0

1 1 V f 1 1 1 l l 1 l 1 1 l 1 l 1 1 l L_ 1 l

I I I I l l l l l l l l _1 1 1 L-
0.30

0.25 BONE DOSE

~ AVERAGE ADULT

0.20 - 0 AVERAGE CHILD

J

J

0.15

0.10

- O

0.05

0.00

0 10 20 30 40 50 60 70 80 90 100

POPULATION (PERCENT)

Figure D-11.-Average Individual Bone Dose Distribution.
1.5 

1.2 LUNG DOSE

o AVERAGE CHILD

* AVERAGE ADULT

0.9

J

J

-- I

O 0.6

0.3

0.0

0 10 20 30 40 50 60 70 80 90 100

POPULATION (PERCENT)

Figure D-12.-Average Individual Lung Dose Distribution.
1.5

I I I I I rI 1 1I I l 1 I I1I ] I1 I I1I 1 1 III1I11 1 1 If 1 1 1 I I I1 1 1 1 1 1 I 1 1 1 l 

SKIN DOSE

1.2 0 AVERAGE CHILD

AVERAGE ADULT

0.9 ix

J

0.6

0.3

0.0

0 10 20 30 40 50 60 70 80 90 100

POPULATION (PERCENT)

Figure D-13.-Average Individual Skin Dose Distribution.

vI
1.5

1.2 LIVER DOSE

_ AVERAGE ADULT

o AVERAGE CHILD

0.9

J

0.6

V',

0

0.3

0.0

0 10 20 .30 40 50 60 70 80 90 100

POPULATION (PERCENT)

Figure D-14.-Average Individual Liver Dose Distribution.
APPENDIX E

Potential Radionuclide Concentrations

and Dose Rate Comparisons

for Eight Typical Centroids

E-1
APPENDIX E

Potential Radionuclide Concentrations and

Dose Rate Comparisons for Eight

Typical Centroids

LIST OF FIGURES

Figure

Number Title Page

E-1 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 20) ....... E-3

E-2 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 34) ....... E-4

E-3 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 68) ....... E-5

E-4 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 87) ....... E-6

E-5 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 143).......E-7

E-6 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 166).......E-8

E-7 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 222).......E-9

E-8 Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 238).......E-10

E-9 Monthly Average Radionuclide Concentrations in Water (Centroid 20)............... .. E-11

E-10 Monthly Average Radionuclide Concentrations in Water (Centroid 34)............... .. E-12

E-11 Monthly Average Radionuclide Concentrations in Water (Centroid 68)............... .. E-13

E-12 Monthly Average Radionuclide Concentrations in Water (Centroid 87)............... .. E-14

E-13 Monthly Average Radionuclide Concentrations in Water (Centroid 143).............. .. E-15

E-14 Monthly Average Radionuclide Concentrations in Water (Centroid 166).............. .. E-16

E-15 Monthly Average Radionuclide Concentrations in Water (Centroid 222).............. .. E-17

E-16 Monthly Average Radionuclide Concentrations in Water (Centroid 238).............. .. E-18

E-17 Radionuclide Contributions to Dose (Centroid 20) ............................ E-19

E-18 Radionuclide Contributions to Dose (Centroid 34) ............................ E-20

E-19 Radionuclide Contributions to Dose (Centroid 68) ............................ E-21

E-20 Radionuclide Contributions to Dose (Centroid 87) ............................ E-22

E-21 Radionuclide Contributions to Dose (Centroid 143) ........................... .E-23

E-22 Radionuclide Contributions to Dose (Centroid 166) ........................... .E-24

E-23 Radionuclide Contributions to Dose (Centroid 222) ............................. E-25

E-24 Radionuclide Contributions to Dose (Centroid 238) ........................... .E-26

E-25 Pathway Contribution to Dose (Centroid 20) ................................ E-27

E-26 Pathway Contribution to Dose (Centroid 34) ................................ E-28

E-27 Pathway Contribution to Dose (Centroid 68) ................................ E-29

E-28 Pathway Contribution to Dose (Centroid 87)................................. .. E-30

E-29 Pathway Contribution to Dose (Centroid 143) ............................... .E-31

E-30 Pathway Contribution to Dose (Centroid 166) ............................... .E-32

E-31 Pathway Contribution to Dose (Centroid 222) ............................... .E-33

E-32 Pathway Contribution to Dose (Centroid 238) ............................... .E-34

E-33 Pathway Contribution to Dose Commitment (Centroid 20) ...................... E-35

E-34 Pathway Contribution to Dose Commitment (Centroid 34) ...................... E-36

E-35 Pathway Contribution to Dose Commitment (Centroid 68) ...................... E-37

E-36 Pathway Contribution to Dose Commitment (Centroid 87) ...................... E-38

E-37 Pathway Contribution to Dose Commitment (Centroid 143) ................... E-39

E=38 Pathway Contribution to Dose Commitment (Centroid 166) ..................... .E-40

E-39 Pathway Contribution to Dose Commitment (Centroid 222)..................... .E-41

E-40 Pathway Contribution to Dose Commitment (Centroid 238)..................... .E-42

E-2
BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

104

102

100

10-2

10

10-6

1 0-8

1311

riLrfl

90Sr 1311

GROUND CONCENTRATIONS

Figure E-1.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 20).

106

E'\ 

* -o "r

C7

85Kr 90Sr

AIR CONCENTRATIONS

I

1. 5. 1. 1. 1. 1. 1. 1. 1 0. FlT] 1. 1- 0- 0- 1- 1- 2- 0-I-I-l- 0- t- t- I -t -I -I -I -0 -0 -l -0 
106

104

102

0

Ur 10-2

D 10 

LC

10 o 4.- 

10-.

10- 8

BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

1' 1 L1 1 I IL"I I I I 1 1 1L1________1 1 1 1'1 .1 1.1 Il

85Kr 90Sr

AIR CONCENTRATIONS

1 3 1 1 90Sr

GROUND CONCENTRATIONS

Figure E-2.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 34).

1311

_

_
BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

106

104

102

100

10-2

10

10-6

10-8

131I

_ ___ 0 1 I 1I , .......... ..___... ._I__._.__. _ _.

9 0Sr 131

GROUND CONCENTRATIONS

Figure E-3.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 68).

-4T

85Kr 90Sr

AIR CONCENTRATIONS

E~ 2

i. * C)

0~

C

7

I
BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

104

102

E100

'- >u -2

(10

V d

"r 10- 2 0-

10-6

10-8

131I 90 Sr 1311

GROUND CONCENTRATIONS

Figure E-4.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 87).

106

7

85

Kr 90Sr

AIR CONCENTRATIONS

T

ON

. . . I. I I. I I . I I. I. I. I. I. I. I I I I I.I.I.I.I. I I.I.I.I I I.
N~ M~V E

"S

U O_.r

Q -O

C

rI

BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

7

100

10-2 10-4

10

10-

131I

106

104

102

85Kr 90Sr 131I 3H 90Sr

AIR CONCENTRATIONS GROUND CONCENTRATIONS

Figure E-5.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 143).

l l i l 1 L_
BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

7

_IIII A L L1 1 11 l A A 4

85Kr 90Sr

AIR CONCENTRATIONS

1311 90Sr 131I

GROUND CONCENTRATIONS

Figure E-6.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 166).

106

104

102

M~ E

E

"- V.

L- C

.~0

S1-

CD

100

10

10-6

10-8

K
104

102

S E100

0-2

1

1

ioC

10-6

10-8

BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

Tf

7

. . . 0. l l. . .L.L. 1. .1.1.1.I.11. . . . . . . . . . . . . .

85Kr 90Sr

AIR CONCENTRATIONS

1311 90Sr

GROUND CONCENTRATIONS

Figure E-7.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 222).

TrI

1311

106
106

10

1024

N

10

010-

10

10

10o8

BARS LEFT TO RIGHT

JAN, APR, JULY, OCT

85Kr

AIR CONCENTRATIONS

90Sr 1311 90Sr

GROUND CONCENTRATIONS

Figure E-8.-Monthly Average Radionuclide Concentrations in Air and on Ground (Centroid 238).

t1

C

1311

_
RADIONUCLIDE CONCENTRATION IN GROUND WATER, pC./LITER

NO RIVER WATER

H

100

10-3

16

10

90Sr 95Zr 95Nb 106Ru 1291 1311

Figure E-9.-Monthly Average Radionuclide Concentrations in Water (Centroid 20).

103

r

I

134Cs

.:. ::.

". ".

".": .,.Oo ...

0000.

:.: :.

:. 

.00 .00 .". ":

.: :": ":

. ":: ... :

.: :": 

.: ":: ".::

... ...

:" :.: :.:

.. .. ...

". :.

": ":: :.

". ,0.

'000.0

boo ... .. 

0

; 

.. 

:.:

":

::

::.

". 

:":

::

".

:":

::

:":

..

:":

:":"

.:"

..

:":

:.

:.

::,

::, ..

".

":

":

.. ..

":

":

":

::

:.

:.

::

.:

:":

::..

". 

:":

::

".

:":

:":

::

::

::

..

:":"

:":..

::"

::"

::

::

..

..0

A

04

.

14

14

.;

.;

.;

a4
103 RADIONUCLIDE CONCENTRATION IN GROUND WATER, pC./LITER

RADIONUCLIDE CONCENTRATION IN GROUND WATER, pC./LITER

RADIONUCLIDE CONCENTRATION IN ILLINOIS RIVER, pC./LITER

CONCENTRATION OF ADSORBED RADIONUCLIDES _

SUSPENDED SEDIMENT, pC./Kg

00

10

l+,. r.-

11O16

16

109

3H 90Sr 95Zr 95Nb 0R 1291 1311 134Cs 3s

Figure E-10.-Monthly Average Radionuclide Concentrations in Water (Centroid 34).
103

RADIONUCLIDE CONCENTRATION IN GROUND WATER, pC./LITER 1

RADIONUCLIDE CONCENTRATION IN MISSISSIPPI RIVER, pC./LITER

*: CONCENTRATION OF At)SORBED RADIONUCLIDES, pC./Kg

SUSPENDED SEDIMENT 1

100-.-

10

106

10o

IN-.tr - - 111

0 --- .... :--. .- 

3H-... r ...... .-..- - . 1 -... 1...3..... ..... ...-

Figure E-11.-Monthly Average Radionuclide Concentrations in Water (Centroid 68).
103

RADIONUCLIDE CONCENTRATION IN GROUND WATER, pCi/LITER

RADIONUCLIDE CONCENTRATION IN MISSISSIPPI RIVER, pC./LITER

CONCENTRATION OF ADSORBED RADIONUCLIDES, pC./Kg SUSPENDED SEDIMENT 1

100

10 1 :2: " 

10 9 --- 95N b -" 6 R ' -29 - 1 -- 4

3H 90Sr 95Zr 95Nb 106Ru 1291 1311 134Cs 137Cs

Figure E-12.-Monthly Average Radionuclide Concentrations in Water (Centroid 87).
RADIONUCLIDE CONCENTRATION IN GROUND WATER, pC./LITER

NO RIVER WATER

95Zr 95 1291 Nb 1311

Figure E-13.-Monthly Average Radionuclide Concentrations in Water (Centroid 143).

103

100

10 3

4-1

1 6

10

1 2

0-2

90Sr 1 37Cs

...

..

-...

.....

.. ... ... ...

... ... ... .. ...

.. ... ... V.

.. ... ... H ...

...

0.00

.. ... ... ... ...

.. ... ... ... ...

.. ... ... ...

0.0

... ... ...

00.0 oo...o

... ... ...

.00,0 0 00000

... ... ... . .. . . . .. 

... ... ... ... ...

...... ... ...

... ... ... ... ... ...

... ... ... ... ... ...

... ... ...

... ... ... ...

%EMMMM.M--a i

...

....

196Ru
1 3 -RADIONUCLIDE CONCENTRATION IN GROUND WATER, pC./LITER

RADIONUCLIDE CONCENTRATION IN MISSISSIPPI RIVER, pC./LITER 

J CONCENTRATION OF ADSORBED RADIONUCLIDES, pC./Kg

SUSPENDED SEDIMENT

100

10

06

-9* --

1V

10

3H 90Sr 95Zr 95Nb 106Ru 1291 1311 134Cs 3C

Figure E-14.-Monthly Average Radionuclide Concentrations in Water (Centroid 166).
RADIONUCLIDE CONCENTRATION

RADIONUCLIDE CONCENTRATION

IN GROUND WATER, pC1/LITER

IN MISSOURI RIVER, pCi/LITER

CONCENTRATION OF ADSORBED RADIONUCLIDES, pC./Kg

SUSPENDED SEDIMENTS

":

...

::... ...

.::

:".

:.:

::.

:.:

:.:

::...

":

:":

":

":

::"

":

":

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:":

":

":

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...

::.

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::.

:".

:.:

.

::.

::.

::.

::.

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".

":

...

..

...

..

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...

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..

...

..

...

.": ..

:":

.

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...

."."... ..... ..... .....

..

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::

:"

::

::

..

": 

..

::

::

::

:"

..

": 

": 

": 

.. .....

":

...

::.

.

129i

100

1C

10-

10-

Figure E-15.-Monthly Average Radionuclide Concentrations in Water (Centroid 222).

103

90Sr 131i 95Zr

:::

:" ... ...

:::...

". 

...

:::

:::

:::

:::... ...

:":

...

". 

": :.

..

:::

::"

":

...

:.:

.. ...

..

...

::.

::.

::.

..

":

...

..

":

...

..

...

..

...

..

...

..

...

..

...

:.".

..

...

..

...

::.

::.

":

...

::.

..

":

...

::.

..

".

...

..

".

...

...

...

..

...

:.:

..

...

::.

...

...

...

::"

":

::"

...

::.

::.

134Cs

.:.

:t :

:.

:

...

P, Y.04

0:::

:": 

I .,.1 S:: J ". _"r' ' . ; i i;":: .'cr. '"i r. '".... ; " .r i _"YXY 1 J

95Nb

.: es.

::.

LX7L. L

..

..

..

..

.

..

..

..

:".

..

..

..

..

..

..

..

..

.":

..

..

..

..

..

..

..

..

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..

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": 

..

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..

:."

:."

::

..

..

..

..

..

..

': 

::"

::

:.L

1

"1
103

RADIONUCLIDE CONCENTRATION IN GROUND WATER, pC./LITER 1

RADIONUCLIDE CONCENTRATION IN MISSOURI RIVER, pC./LITER

! CONCENTRATION OF ADSORBED RADIONUCLIDES, pC./Kg

00:SUSPENDED SEDIMENT 1

3 ..-.. :

100

10 ........ . 9 ... .... 9 131. ...

9"-...... .. .. -.--. -.--... . . ....

-12 " i:: :"-1

3H 90Sr 95Zr 95Nb 106Ru 1291 1311 134Cs 137Cs

Figure E-16.-Monthly Average Radionuclide Concentrations in Water (Centroid 238).
0.006

OTHER

137Cs

133Xe

131i

?85Kr

TOTAL BODY

DOSE RATE

MREM/YR

100

80

60

0.003

OTHER

f141-144Ce

137Cs

131I

G.I. TRACT THYROID BONE LUNGS

Figure E-17.-Radionuclide Contributions to Dose (Centroid 20).

SKIN

0.075

OTHER

-4--

LIVER

40

TI

W

I0

U

L

20

0

0.009

OTHER

141-144Ce

134-1 37

Cs

133Xe

1311

0.063

OTHER

1311

0.004

OTHER

141

-144

Ce

134-1 37 Cs

133Xe

131I

85Kr

3H

0.006

OTHER

141-144Ce

134-1 37

Cs

133Xe

1311

85Kr

134

Cs

133X

131 I
0.178

100

80

0.193

OTHER

134Cs

133Xe

Jt95Nb- Zr

1311

3H

TOTAL BODY

0.919

OTHER OTHER

144Ce

Xe

131i

134Cs

137C

95Nb-Zr

3H

GI TRACT THYROID

0.036

OTHER

1311

103-1 06

Ru

95Nb-Zr

89-90Sr

85Kr

BONE

0.168

OTHER

133Xe

131i

137Cs

_134Cs

3H

LUNGS

0.196

OTHER

133Xe

1311

85Kr

134Cs

137

1 Cs

3H

SKIN

0.185

OTHER

OTHER

-34- 137Cs

134

Cs

Xe

1311

195Nb-Zr

3H

LIVER

Figure E-18.-Radionuclide Contributions to Dose (Centroid 34).

DOSE RATE

MREM/YR

1311

O

60

40

Lii

L)

LFW~

20

0
0.049

OTHER

DOSE RATE

MREM/YR

100

80

-.

Figure E-19.-Radionuclide Contributions to Dose (Centroid 68).

r1

60

UO

IC-)

a 40

0.223

OTHER

137Cs

134Cs

1311

133Xe

95 Nb-Zr

3H

TOTAL BODY

20

0

0.220

OTHER

144Ce

1311

134Cs

L3Cs 1

33Xe

95 Nb-Zr

3H

GI TRACT

1.01

OTHER

131i

0.202

OTHER

131I

L137Cs 134 s

133Xe

3H

LUNGS

0.223

OTHER

133Xe

1311

85

Kr

134Cs

137C

3S

SKIN

133Xe

1311

103-106Ru

95Nb-Zr

89- 90S r

85Kr

39-41Ar

THYROID BONE

0.240

OTHER

137Cs

134Cs

J

133Xe

1311

95Nb-Zr

3H

LIVER
0.314

100

80

60

1.01

OTHER

131 k

OTHER

137Cs

134C5

1311

T95Nb-Zr

3H

TOTAL BODY

0.049

OTHER OTHER

1311

144Ce

137Cs

134Cs

95Nb-Zr

S89-90Sr

3H

GI TRACT

0.299

OTHER

1311

_37Cs

134Cs

3H

LUNGS

0.314

OTHER

L131

85Kr

134Cs

137Cs

--133Xe

3H

SKIN

Figure E-20.-Radionuclide Contributions to Dose (Centroid 87).

DOSE RATE

MREM/YR

THYROID

T1

0.320

133Xe

1311

103-106Ru

95Nb-Zr

89-90 Sr

85Kr

BONE

0

IC

40

0.336

OTHER

137Cs

1 34

Cs

1311

1 33Xe

95

Nb-Zr

3H

LIVER

20

0
0.075

100

80

0.079

OTHER

144Ce

131I

137Cs

133Xe

-- 95Nb-Zr

OTHER

134Cs

133 X 137Cs

1311

f9NbZr

3H

TOTAL BODY

0.318

OTHER

1311

THYROID

0.016

OTHER

137Cs

134Cs

1311

103-1061

95Nb-Zr

89-90Sr

85Kr

BONE

0.072

OTHER

OTHER

133Xe

131I

LUNGS

0.085

OTHER

133Xe

1311

85Kr

134Cs

137Cs

SKIN

Figure E-21.-Radionuclide Contributions to Dose (Centroid 143).

DOSE RATE

MREM/YR

GI TRACT

t~ii LLJ

Ucr.

60

40

0.077

OTHER

134Cs

3 137Cs

3

t95Nb.Zr

3H

LIVER

20

0
0.099

OTHER

100

80

60

0

I')

F-)

a 40

20

0

0.077

OTHER

-4--

134Cs

133Xe

137C

1311

99 M

3H

GI TRACT

1.15

OTHER

1331

1311

T3H

THYROID

0.052

OTHER

1 34

Cs

134Cs

133Xe

$133

131

99M

85Kr

3H

TOTAL BODY

0.075

OTHER

137Cs

134Cs

133Xe

1311

LUNGS

0.088 OTHER 0.119

OTHER

134Cs

137Cs

1311

85

Kr

SKIN

Figure E-22.-Radionuclide Contributions to Dose (Centroid 166).

DOSE RATE

MREM/YR

133 1

1311

E 85Kr

BONE

OTHER

1 34

Cs

1l33Xe

*133i

1311

99mo

85 Kr

3H

LIVER

0.088 0.119
0.134

OTHER

13Cs

- 137Cs

133Xe

131 I

0.149

OTHER

13Cs

13 3Xe

137Cs

131 I

OTHER

137Cs

134Cs

S133 e

131

0.154

100

80

60

LL

N

a 40

Figure E-23.-Radionuclide Contributions to Dose (Centroid 222).

DOSE RATE

MREM/YR 0.145

OTHER

3 Cs

144Ce

137Cs

133Xe

1311

Nb-Zr

3H

GI TRACT

OTHER

1331

1311 TIl

0.141

85

Kr

3H 3H

LUNGS SKIN

0.765

OTHER

THYROID

0.039

OTHER

137Cs

1 34

Cs

133Xe

1311

t799

85Kr

39-41 Ar

3

H

LIVER

1337

1311

103-1 06

Ru

95Nb-Zr

88Kr

85Kr

39-41Ar

BONE

20

0

TOTAL BODY

134Cs
0.246

100

80

60

0.252

OTHER

Xe

- 131I

3H

TOTAL BODY

0.288

OTHER

131I

OTHER

144CeT

3H

GI TRACT

0.002

OTHER

137-

1311

103-1 06

Ru

95Nb-Zr

B9-90Sr

85Kr

BONE

DOSE RATE

MREM/YR

Figure E-24.-Radionuclide Contributions to Dose (Centroid 238).

THYROID

aN LL.

C

F_

Wi

L)

Wi

Q- 40

20

0 L.

0.246

OTHER

3H

LUNGS

0.251

OTHER

85Kr

3H

SKIN

0.247

OTHER

3H

LIVER
DOSE RATE

MREM/YR

100

80

3H

TRANSPIRATION

AIR

SUBMERSION

TOTAL BODY

0.006

EXPOSURE

TO SOIL

AIR

INHALATION

AIR

SUBMERSION

SKIN

0.006

FOODS

DRINKING

WATER

BEEF 

LAMB

EXPOSURE

TO SOIL

AIR

INHALATION

Figure E-25.-Pathway Contribution to Dose (Centroid 20).

3H

TRANSPIRATION

0

Li

LU

60

40

20

0

0.003

FOODS

BEEF 

LAMB

EXPOSURE

TO SOIL

AIR

INHALATION

AIR

SUBMERSION

BONE

0.063

FOODS

MILK

EXPOSURE

TO SOIL

DRINKING

WATER

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

THYROID

0.009

FOODS

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

GI TRACT

0.004

FOODS

BEEF 

LAMB

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

LUNGS

0.075

FOODS

BEEF 

LAMB

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

LIVER
DOSE RATE

MREM/YR 0.178

FOODS

DRINKING

WATER

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

TOTAL BODY

0.196

FOODS

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

SKIN

0.036

r

0.919

100

80

60

0.193

EXPOSURE

TO SOIL

DRINKING

WATER

FOODS

DRINKING

WATER

FOODS

BEEF 

LAMB

EXPOSURE

TO SOIL

AIR

INHALATION

AIR

SUBMERSION

BONE

DRINKING WATER

FOODS

MILK

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

THYROID

0.168

FOODS

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

LUNGS

0.185

FOODS

DRINKING

WATER

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

LIVER

Figure E-26.-Pathway Contribution to Dose (Centroid 34).

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

GI TRACT

TI

8

ti

0

GC

W

cd

40

20

0
DOSE RATE

MREM/YR

100

80

0.223

DRINKING

WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

TOTAL BODY

0.223

FOODS

EXPOSURE TO

SOIL

DRINKING

WATER

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

SKIN

0.049

DRINKING

WATER

FOODS

WHEAT 

PRODUCTS

EXPOSURE TO

SOIL

AIR

INHALATION

AIR

SUBMERSION

BONE

1.01

DRINKING

WATER

FOODS

MILK

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

THYROID

0.220

DRINKING WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

GI TRACT

0.202

FOODS

EXPOSURE TO

SOIL

DRINKING

WATER

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

LUNGS

0.240

DRINKING

WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

I

LIVER

Figure E-27.-Pathway Contribution to Dose (Centroid 68).

TN

60

40

LL

0

CC)

IL

20

0
0.320

DRINKING

WATER

FOODS

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

TOTAL BODY

0.314

DRINKING WATER

FOODS

BEEF & LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

SKIN

0.049

100

80

60

1.01

DRINKING

WATER

FOODS

DRINKING

WATER

FOODS

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

AIR SUBMERSION

BONE

3H

TRANSPIRATION

THYROID

0.314

DRINKING. WATER

FOODS

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

MILK

EXPOSURE

TO SOIL

AIR

INHALATION

AIR SUBMERSION

0.299

DRINKING WATER

FOODS

BEEF & LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

-UG_

LUNGS

0.336

DRINKING

WATER

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

LIVER

Figure E-28.-Pathway Contribution to Dose (Centroid 87).

DOSE RATE

MREM/YR

3H

TRANSPIRATION

AIR SUBMERSION

GI TRACT

WI

O

W

C'

LL

0

FLU

0lC

40

20

0
0.075

FOODS

EXPOSURE TO

SOIL

100

80

60

3H

TRANSPIRATION

AIR

SUBMERSION

TOTAL BODY

0.085

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

SKIN

0.016

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

AIR

SUBMERSION

BONE

0.318

DRINKING WATERFOODS

MILK

EXPOSURE

TO SOIL

AIR

INHALATION

AIR

SUBMERSION

3H

TRANSPIRATION

THYROID

0.079

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

GI TRACT

0.072

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

-SUBMERSION

LUNGS

0.077

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

LIVER

Figure E-29.-Pathway Contribution to Dose (Centroid 143).

DOSE RATE

MREM/YR

AIR

INHALATION

tI1 0

LC)

LU

0~ 40

20

0
0.099

100

80

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

TOTAL BODY

0.088

DRINKING

WATER

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

SUBMERSION

SKIN

0.052

DRINKING

WATER

FOODS

BEEF 

LAMB

1.15

DRINKING

DRINKING

WATER

FOODS

DRINKING

WATER

FOODS

BEEF & LAMB

EXPOSURE TO

SOIL

AIR INHALATION

AIR

SUBMERSION

BONE

EXPOSURE

TO SOIL

AIR

INHALATION

3H

TRANSPIRATION

IAIR

.SUBMERSION

THYROID

0.077

DRINKING

WATER

BEEF & LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

TRANSPIRATION

AIR SUBMERSION

GI TRACT

0.075

DRINKING

WATER

FOODS

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

LUNGS

0.119

DRINKING

WATER

FOODS

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

LIVER

Figure E-30.-Pathway Contribution to Dose (Centroid 166).

DOSE RATE

MREM/YR

MILK

MI

W

60

40

w

IU

GC

20

0
0.145

DRINKING

WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

TOTAL BODY

0.149

100

80

AIR

SUBMERSION

SKIN

0.039

DRINKING WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

0.765

DRINKING

WATER

DRINKING

WATER

FOODS

BEEF 

LAMB

EXPOSURE TO

SOIL

AIR INHALATION

AIR

SUBMERSION

BONE

MILK

EXPOSURE

TO SOIL

AIR

INHALATION

3 H

TRANSPIRATION

AIR

SUBMERSION

THYROID

0.141

DRINKING

WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

GI TRACT

0.134

DRINKING

WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

TRANSPIRATION

AIR

SUBMERSION

LUNGS

0.154

DRINKING

WATER

FOODS

BEEF & LAMB

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBM RSION

LIVER

Figure E-31.-Pathway Contribution to Dose (Centroid 222).

DOSE RATE

MREM/YR

FOODS

rTIi

60

40

Wi

0

IL1i

20

0
DOSE RATE

MREM/YR

100

80

60

0.246

FOODS

1 DRINKING

WATER

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

TOTAL BODY

0.251

FOODS

AIR

INHALATION

3H

TRANSPIRATION

AIR SUBMERSION

SKIN

0.002

DRINKING

WATER

I-(uuS

EXPOSURE TO

SOIL

AIR

INHALATION

AIR

SUBMERSION

BONE

0.288

FOODS

DRINKING

WATER

MILK

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

THYROID

0.252

FOODS

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

0.246

FOODS

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

0.247

DRINKING

WATER

FOODS

EXPOSURE TO

SOIL

AIR

INHALATION

3H

TRANSPIRATION

AIR

SUBMERSION

LIVER

Figure E-32.-Pathway Contribution to Dose (Centroid 238).

h'1 Li

0

ILU

CC)

LU

a~ 40

20

0
50 YEAR DOSE

COMMITMENT, MREM 0.005

100 001 F - FOODS

BEEF 

LAMB

80

60-

AIR

INHALATION

3H

TRANSPIRATION

TOTAL BODY

0.002

FOODS

BEEF 

LAMB

AIR

INHALATION

BONE

0.062

FOODS

FRESH MILK

DRINKING WATER

AIR

INHALATION

3H

TRANSPIRATION

THYROID

0.057

BEEF 

LAMB

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

0.007

FOODS

BEEF & LAMB

AIR

INHALATION

3

TRANSPIRATION

LIVER

0.002

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

0.002

AIR

INHALATION

3H

TRANSPIRATION

SKIN

Figure E-33.-Pathway Contribution to Dose Commitment (Centroid 20).

tT1

.1w

1-

W

CL 40

20

0

7
50 YEAR DOSE

COMMITMENT,N

100 r- I

80

MREM 0.154

FOODS

AIR

INHALATION

3H

TRANSPIRATION

TOTAL BODY

0.015

FOODS

BEEF 

LAMB

FRESH MILK

WHEAT 

PRODUCTS

AIR

INHALATION

BONE

0.895

DRINKING WATER

FOODS

FRESH MILK

AIR

INHALATION

3H

TRANSPIRATION

THYROID

0.386

FOODS

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

0.161

FOODS

AIR

INHALATION

3H

TRANSPIRATION

LIVER

0.141

FOODS

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

0.141

t

FOODS

AIR

INHALATION

3H

TRANSPIRATION

SKIN

Figure E-34.-Pathway Contribution to Dose Commitment (Centroid 34).

60

40

Wi

OI

03

U

cr

Wi

0l20

0
MREM 0.201

DRINKING

WATER

FOODS

AIR

INHALATION

3H

TRANSPIRATION

TOTAL BODY

0.033

DRINKING

WATER

FOODS

BEEF 

LAMB

WHEAT 

PRODUCTS

AIR

INHALATION

BONE

0.990

50 YEAR DOSE

COMMITMENT,I

100 E

80 

0.369

DRINKING

I WATER

FOODS

AIR

INHALATION

DRINKING

WATER

FOODS

FRESH MILK

AIR

INHALATION

TRANSPIRATION

THYROID

0.217

DRINKING

WATER

FOODS

AIR

INHALATION

3H

TRANSPIRATION

LIVER

0.172

DRINKING

WATER

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

0.172

DRINKING

WATER

AIR

INHALATION

H3

TRANSPIRATION

SKIN

Figure E-35.-Pathway Contribution to Dose Commitment (Centroid 68).

3H

TRANSPIRATION

LUNGS

rT

60

LU

0

LU

CC)

Q-40

20

0
50 YEAR DOSE

COMMITMENT, MREM 0.301

DRINKING

WATER

FOODS

BEEF 

LAMB

AIR

INHALATION

3H

TRANSPIRATION

TOTAL BODY

DRINKING

WATER

100

80

60

BEEF

LAMB

FRESH MILK

WHEAT AND

PRODUCTS

AIR

INHALATION

BONE

DRINKING

WATER

FOODS

FRESH MILK

AIR

INHALATION

3H

TRANSPIRATION

THYROID

FOODS

BEEF 

LAMB

DRINKING

WATER

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

DRINKING

WATER

FOODS

BEEF 

LAMB

AIR

INHALATION

3H

TRANSPIRATION

LIVER

FOODS

DRINKING WATER

AIR

INHALATION

3

H

TRANSPIRATION

GI TRACT

FOODS

)RINKING WATER

AIR

INHALATION

3H

TRANSPIRATION

SKIN

Figure E-36.-Pathway Contribution to Dose Commitment (Centroid 87).

0.049 0.982 0.420 0.312 0.270 0.270

FOODS

0

1-,

40

20

0
50 YEAR DOSE

COMMITMENT, MREM 0.063

F OODS

AIR

INHALATION

3H

TRANSPIRATION

TOTAL BODY

0.005

DRINKING

WATER

FOODS

BEEF

LAMB

FRESH MILK

FRESH

POTATOES

AIR

INHALATION

BONE

0.065

FOODS

AIR

INHALATION

0.305

DRINKING

WATER

FOODS

FRESH MILK

AIR

INHALATION

3H

TRANSPIRATION

THYROID

0.059

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

0.059

AIR

INHALATION

3H

TRANSPIRATION

SKIN

Figure E-37.-Pathway Contribution to Dose Commitment (Centroid 143).

80

60

LL.

0

FLii

d~ 40

20

0

0.125

FOODS

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

3H

TRANSPIRATION

LIVER
50 YEAR DOSE

COMMITMENT,

100F

80

60

LL

MREM 0.076

DRINKING

WATER

FOODS

BEEF 

LAMB

AIR

INHALATION

TRANSPIRATION

TOTAL BODY

0.054

DRINKING

WATER

0.032

DRINKING

WATER

FOODS

BEEF

LAMB

AIR

INHALATION

BONE

0.054

DRINKING

WATER

1.13

DRINKING

WATER

FOODS

FRESH MILK

AIR

INHALATION

3H

TRANSPIRATION

THYROID

0.080

DRINKING

WATER

FOODS

BEEF & LAMB

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

Figure E-38.-Pathway Contribution to Dose Commitment (Centroid 166).

0.097

DRINKING

WATER

FOODS

BEEF

LAMB

AIR

INHALATION

3H

TRANSPIRATION

LIVER

T1

Wi

UCC

W~ 40

20

0

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

FOODS

AIR

INHALATION

3H

TRANSPIRATION

SKIN

FOODS
50 YEAR DOSE

CTfl4TMENT MREM' A 118

DRINKING

WATER

FOODS

AIR

INHALATION

3H TRANSPIRATION

TOTAL BODY

0 014 ,DRINKING

DRINKING

WATER

FOODS

BEEF & LAMB

FRESH MILK

WHEAT AND

PRODUCTS

AIR

INHALATION

BONE

0.737

DRINKING

WATER

FOODS

FRESH MILK

AIR

INHALATION

3H

TRANSPIRATION

THYROID

0.174

FOODS

DRINKING

WATER

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

0.127

DRINKING

WATER

FOODS

BEEF & LAMB

AIR

INHALATION

3H

TRANSPIRATION

LIVER

0.104

DRINKING

WATER

FOODS

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

0.104

DRINKING

WATER

FOODS

AIR

INHALATION

TRANSPIRATION

SKIN

Figure E-39.-Pathway Contribution to Dose Commitment (Centroid 222).

100

80

60

40

0

FW&

U~

20

0

I
50 YEAR DOSE

C TMMITMFNT MRFM A 246 A 87R n Ong2 n OAK , .1iI .I . U. V. 40U. 244U '.244

FOODS

AIR

INHALATION

3H

TRANSPIRATION

TOTAL BODY

DRINKING

WATER

FOODS

BEEF & LAMB

WHEAT

PRODUCTS

PROCESSED

OTHER

VEGETABLES

AIR

INHALATION

BONE

FOODS

DRINKING

WATER

FRESH MILK

AIR

INHALATION

3H

TRANSPIRATION

THYROID

FOODS

AIR

INHALATION

3H

TRANSPIRATION

LUNGS

FOODS

AIR

INHALATION

3H

TRANSPIRATION

LIVER

FOODS

AIR

INHALATION

3H

TRANSPIRATION

GI TRACT

FOODS

AIR

INHALATION

3H

TRANSPIRATION

SKI N

Figure E-40.-Pathway Contribution to Dose Commitment (Centroid 238).

100 r80

60

40

T1.

W

LA- O~

20

0

n n3 A f1 7AA 03AA


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