Recommended Screening Limits for Contaminated Surface Soil and Review of Factors Relevant to Site-Specific Studies: Recommendations of the National ... Protection and Measurements

NCRP REPORT No. 129

RECOMMENDED SCREENING LIMITS FOR CONTAMINATED SURFACE SOIL AND REVIEW OF FACTORS RELEVANT TO SITE-SPECIFIC STUDIES Recommendations of the NATIONAL COUNCIL O N RADIATION PROTECTION AND MEASUREMENTS

Issued January 29,1999

National Council on Radiation Protection and Measurements 7910 Woodmont Avenue / Bethesda, M D 20814-3095

LEGAL NOTICE This Report was prepared by the National Council on Radiation Protection and Measurements (NCRP). The Council strives to provide accurate, complete and useful information in its documents. However, neither the NCRP, the members of NCRP, other persons contributing to or assisting in the preparation of this Report, nor any person acting on the behalf of any of these parties: (a) makes any warranty or representation, express or implied, with respect to the accuracy, completeness or usefulness of the information contained in this Report, or that the use of any information, method or process disclosed in this Report may not infringe on privately owned rights; or (b)assumes any liability with respect to the use of, or for damages resulting from the use of any information, method or process disclosed in this Report, under the Civil Rights Act of 1964, Section 701 et seq. as amended 42 U.S.C. Section 2000e et seq. (Title VZZJ or any other statutory or common law theory gouerning liability.

Library of Congress Cataloging-in-PublieationData Recommended screening limits for contaminated surface soil and review of fadors relevant to site-specific studies. p. cm. - (NCRP report ; no. 129) "January 1999." Includes bibliographical references and index. "SC a-20." ISBN 0-929600-61-4 1. Radioactive pollution of soils -Standards. 2. RadioisotopesEnvironmental aspeds. 3. Radiation dosimetry. I. National Council on Radiation Protection and Measuremenb. 11. Series. TD879.R34R43 1998 98-48643 628.5'5-dc21 CIP

Copyright O National Council on Radiation Protection and Measurements 1999 rights reserved. This publication is protected by copyright. NO part of this publication may be reproduced in any form or by any means, including photocopying, or utilized by any information storage and retrieval system without written permission from the copyrightowner, except for brief quotation in critical articles or reviews.

Preface The decision regarding the need to cleanup surface soil contaminated with radionuclides can be complicated. The use of this Report is intended to assist in this decision. This Report provides screening limits which can be applied to sites where the surface soil is known to be contaminated with radionuclides. The screening limits are calculated using methods which are chosen to be conservative under most conditions. Their use will allow reasonable judgments to be made regarding whether additional action is needed. Further action will generally not be required if the surface soil concentration is below the suggested limits. If the soil concentration is above the screening limit, a site-specific dose assessment is recommended. It is emphasized that doses calculated with the dose factors in this Report da not represent estimates of doses to particular or typical individuals or thresholds for possible adverse effects. Thus, the screening doses calculated using the methods of this Report are inappropriate for use in calculating population exposures or to estimate health effects. The calculation of doses to actual individuals requires the use of site-specific and individual-specific parameters in the formulas used for the calculations. This Report was prepared by Scientific Committee 64-20 on Contaminated Soil. Serving on the Committee were:

Harold L. Beck,Chairman Environmental Measurements Laboratory New York, New York Members

David G Baker Richland, Washington

William L. Robison Lawrence Livermore National Laboratory Livermore, California

Andre Bouville

Joseph H. Shinn

National Cancer Institute Bethesda, Maryland

Lawrence Livermore National Laboratory Livermore, California

F. Owen Hofhnan SENES Oak Ridge, Inc. Oak Ridge, Tennessee

Steven L. Simon National Academy of Sciences Washington, D.C.

NCRP Secretariat E. Ivan White, Senior Staff Scientist Cindy L. O'Brien, Managing Editor The Council wishes to express its appreciation to the Committee members for the time and effort devoted to the preparation of this Report and to the U.S.Department of Energy for the financial assistance in developing this Report.

Charles B. Meinhold President, NCRP

Contents ..................................................................................... 1. Introduction ........................................................................ 1.1 Scope of Report ............................................................. 1.2 Approach ......................................................................... 1.3 Land-Use Scenarios ........................................................ 1.3.1 Agricultural (AG).................................................. 1.3.2 Heavily Vegetated Pasture (PV).......................... 1.3.3 Sparsely Vegetated Pasture (PS) ......................... 13.4 Heavily Vegetated Rural (RV) ............................. 1.3.5 Sparsely Vegetated Rural (RS)............................ 1.3.6 Suburban (SU) ...................................................... 13.7 No Food Suburban (SN) .......................................

Preface

1.3.8 Construction. Commercial. Industrial (CC)

iii

........

. Soil .........................................................................................

2 Recommended Screening Limits for Contaminated

2.1 Recommended Screening Limits ................................... 2.1.1 Multiple Nuclide Contamination ......................... 2.1.2 Alternate Limiting Doses ..................................... 2.2 Application of Screening Limits .................................... 2.2.1 Applicability of Tabulated Land-Use Scenarios ............................................................... 2.2.2 Special Situations ............................................... 23.3 Nonapplicable Contamination Scenarios ............ 2.2.4 Site-Specific Dose Assessments ........................... 2.3 Scientific Basic for Screening Limits ............................

. 3.1 Dose Model .................................................................... .........................................

3 Dose from External Exposure

34 34 36 36 37

3.2 Discussion of Model Parameters ................................... 3.2.1 Concentration in Soil ............................................ 3.2.2 Effective Dose Factor ............................................ 3.2.2.1 Effective Dose versus Effective DoseEquivalent. Organ Dose .......................... 38 3.2.2.2 Exposure Rate. Free-in-Air Kerma versus Effective Dose .............................. 39

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CONTENTS 3.2.2.3 Dependence of Dose Factor on

Orientation ............................................... 3.2.2.4 Dependence of Dose Factor on Nuclide Depth Distribution .................................. 3.2.2.5 Dependence of Dose Factor on Soil Composition .............................................. 3.2.2.6 Dependence of Dose Factor on Areal Extent of Contamination ......................... 3.2.2.7 Accuracy of Dose Factor Calculations .... 3.2.2.8 Dose Factors Chosen for Screening Dose Calculations ............................................. 3.2.3 Dependence on Soil Moisture and Bulk Density 3.2.4 Shielding by Dwellings ......................................... 3.2.5 Indoor and Outdoor Exposure Times .................. 3.2.6 Age. Sex Correction .............................................. 3.3 Summary of Parameter Values for Screening Calculations .................................................................... 3.4 Calculated Screening Doses ...........................................

.

4 Dose from Inhaled Radionuclides .................................. 4.1 Introduction to the Resuspension-Migration Pathway 4.1.1 General Classes of Resuspension ........................ 4.1.1.1 Wind-Driven Open-Environment

Resuspension

...........................................

4.1.1.2 Localized and Direct Contamination

.....

4.1.2 Parameters of Contaminants that Affect Open-

Air Resuspension .................................................. 4.1.2.1 Particle Size ............................................. 4.1.2.2 Physical-Chemical Form Meeting Availability ............................................... 4.1.2.3 Availability with Time ............................ 4.2 Resuspension and Dose Models ..................................... 4.2.1 Dose Model ............................................................ 4.2.2 Resuspension Models ............................................ 4.2.2.1 Method 1, Estimation of C , by Modified Mass Loading ........................................... 4.2.2.2 Method 2, Estimation of C h by Resuspension Factor ................................ 4.2.2.3 Method 3, Estimation of C& by MassLoading-Derived Resuspension Fadors ... 4.2.3 Estimating the Uncertainty in Values of Air Concentration .......................................................

1

CONTENTS 46.4 Derivation of Radionuclide Concentrations in

Air for Screening

................................................... ...

4.2.5 Example of Calculation of Air Concentrations

4.3 Discussion of Dose Model Parameters .......................... 4.3.1 Average Nuclide Concentration in Soil ............... 4.3.1.1 Concentration in Soil for a Site-Specific

Assessment

...............................................

4.3.1.2 Concentration in Soil for Screening

.......

4.3.2 Indoor versus Outdoor Concentrations in Air .... 4.3.3 Inhalation Dose Factor ......................................... 4.3.3.1 Dependence of Dose Factor on Lung-

Absorption and Particle Size

..................

4.3.3.2 Dependence of Dose Factor on Age ........ 4.3.3.3 Organ Doses versus Effective Dose ........ 4.3.3.4 Uncertainty and Variability in Dose

Fadors ...................................................... 4.3.3.5 Recommended Dose Factors for

Screening

..................................................

Dose Calculations

.................................................

4.3.4 Usage Factors ....................................................... 4.3.4.1 General ..................................................... 4.3.4.2 Screening Values ..................................... 4.3.5 Age Dependence of Dose.Chi1d. infant Screening 4.3.6 Dose from Inhalation of Airborne Radon and

Progeny

..................................................................

4.4 Summary of Recommended Parameter Values for

Inhalation Dose Estimation

...........................................

4.5 Calculated Screening Inhalation Doses ........................

.

5 Dose from Ingested Radionuclides ................................ 5.1 Dose Model ...................................................................... 5.2 ~ k u s s i o nof Model Parameters ................................... 5.2.1 Eladionuclide Concentration in the Soil :,............ 5.2.2 Human Diets ......................................................... 5.2.2.1 Variability in Human Diet ...................... 5.2.2.2 Screening Values for Human Diet ......... 5.2.3 Soil to Vegetation Transfer Factors .................... 5.2.3.1 Effect of Soil Depth Profile on Soil to

Vegetation Transfer Factor

.....................

5.2.3.2 Soil to Vegetation Transfer Factor

Values for Screening

...............................

5.2.4 Animal Diets ......................................................... 5.2.5 Meat, Milk Transfer Factors ................................

~ i i

viii

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CONTENTS

5.2.6 Decay Correction for Delay from Harvest to Consumption ......................................................... 5.2.7 Committed Effective Dose Factors ...................... 5.2.7.1 Dependence of Dose Factor on Age ........ 5.2.7.2 Dependence of Dose Factor on Gastrointestinal Uptake ......................... 5.2.7.3 Uncertainty in Biokinetic Models .......... 5.2.7.4 Recommended Ingestion Dose Factor Values for Screening ............................... 5.2.7.5 Dose Factors for Site-Specific Dose Assessments ............................................. 5.2.7.6 Organ Dose versus Effective Dose ......... 5.2.8 Uncertainty in Ingestion Dose Factors Chosen for Screening Calculations ................................... 5.3 Direct Ingestion of Soil ................................................... 5.3.1 Factors of Inadvertent Ingestion and the Etiology of Purposeful Intake .............................. 5.3.1.1 Inadvertent Intake .................................. 5.3.1.2 Geophagia ................................................. 5.3.2 Review of Literature on Soil Intake .................... 5.3.3 Recommended Ingestion Rates ............................ 5.3.4 Use of Soil Ingestion-Rate Data in Screening Calculations ........................................................... 5.3.5 Calculation of Screening Doses ........................... 5.4 Dependence of Committed Effective Dose on Age ....... 5.5 Summary of Recommended Parameter Values for Screening (Ingestion Pathway) ...................................... 5.6 Calculated Screening Doses ...........................................

.

6 Determination of Radionuclide Concentration in Soil ......................................................................................... 6.1 Factors to Consider in the Design of Screening and Sampling Programs for Radionuclide Concentrations in Contaminated Soil ..................................................... 6.2 Instrumental Measurement Techniques ....................... 6.2.1 In Situ Gamma-Ray Spectrometry ...................... 6.2.2 Exposure Rate Measurements ............................. 6.3 Soil Sampling .................................................................. 6.3.1 Soil Sampling Methodology .................................. 6.3.2 Sample Preparation and Analysis ....................... 6.4 Air Sampling ................................................................... 6.5 Strategy of Determining Radiouclide Concentrations for Screening ...................................................................

6.5.1 Estimating Soil Concentrations by In Situ

Spectrometry or Exposure Rate Measurements

...

6.5.2 Sampling Soil for Screening .................................. 6.5.3 Site-Specific Soil Sampling ....................................

.

7 Calculation of Screening Doses........................................ 7.1 Distribution of Individual Doses ..................................... 7.2 "Maximum" Dose Estimates for Screening .................... 7.3 Site-Specific Dose Assessments .......................................

Appendix k Calculated Screening Doses ..........................

. Appendix C. Dose Factors. Shielding Factors .................. Appendix D. Transfer Factors ..............................................

Appendix B Radionuclide Decay Data...............................

Glossary ....................................................................................... References .................................................................................. The NCRP ................................................................................... NCRP Publications................................................................... Index .......................................................................................... 352

1. Introduction Surface soil can become contaminated with radionuclides through many different mechanisms such as airborne deposition, spills and leaching from contaminated material stored above ground. Current activities associated with the cleanup of contaminated weapons production and storage facilities and the future decommissioning of nuclear facilities might result in additional soil contamination as well as in the discovery of past contamination. Even after cleanup of known contaminated land, some residual contamination will remain. No matter how the contamination occurred, the issue becomes whether the level is ~ ~ c i e n thigh, l y either before or after cleanup, to warrant action or restriction on the use of the contaminated site. The primary purpose of this Report is to provide screening limits (in Bq kg-') which can be applied to sites where the surface soil is determined to be contaminated with one or more radionuclides. These screening limits, which can be defined a s a conservative method of relating an effective dose ( E ) limit for a critical group to a corresponding soil contamination level (EPA, 1990a), can be used to allow reasonable judgments to be made regarding the need for possible (further) action based on present soil radionuclide levels. Such judgments need to be consistent with the recommendations of the National Council on Radiation Protection and Measurements (NCRP) regarding radiation exposure to members of the general population, with applicable regulatory limits, and with the principle of ALARA (as low as reasonably achievable). If the surface soil concentration is below the suggested limits, then no further action will generally be required. If the concentration is above the suggested limit, a site-specific dose assessment should be conducted. I t is emphasized that the doses given in this Report are strictly for comparison with a limiting value to establish a screening level and do not represent estimates of doses to particular or typical individuals or threshold values for possible adverse effects. The calculated doses are deliberately designed to conservatively represent the maximum dose to any individual. Thus,these doses are inappropriate for use in calculatingpopulation exposures or to estimate health effects. The calculation of doses to actual individuals requires the use of site-specificand individual-specificparameters in the formulas used for the calculations.

2

1

1.

INTRODUCTION

To justify the guidance provided, this Report reviews in some detail the scientific basis for estimating both site-specific and generic doses to individuals from all pathways that could result from direct or indirect exposure to the contaminated soil. This review includes discussions of the uncertainty and variability in all the important parameters included in the calculational models. The Report also provides guidance on how to determine the site average radionuclide concentration in surface soil to be used for applying these generic screening criteria. Factors important in site-specific studies are cited in many of the cases discussed.

1.1 Scope of Report Only surface soil contamination is considered. Buried wastes are not considered in this Report. Surface soil refers only to depths comprising the plow layer, i.e., down to a depth of about 30 cm. The guidance herein is not intended to be used for evaluating the implications of an ongoing contamination episode such as a continuing airborne deposition, which is treated in NCRP Commentary No. 3 (NCRP, 1989) and NCRP Report No. 123 (NCRP, 1996). Only radionuclides whose half-lives are longer than 30 d (or are supported by a precursor with a half-life >30 d) are considered. Ground water contamination is not calculated explicitly, although it is recognized that the latter could be an important dose pathway for some sites (see Section 2.2.2).The area contaminated is considered to be relatively extensive (hundreds of square meters or more). All other important dose pathways are considered including external radiation exposure, beta-ray skin dose, ingestion of contaminated foodstuffs, direct and indirect ingestion of soil by humans and animals, and both indoor and outdoor inhalation of resuspended material. Examples of contamination scenarios for which these limits are applicable include widespread contamination from fallout from weapons tests and nuclear facility accidents (such as occurred a t Chernobyl) as well as more localized contamination resulting from nuclear facility operations andlor decontamination and decommissioning. An example of the latter would be the plutonium contamination of soils downwind from the Rocky Flats plant (Krey and Hardy, 1970).

1.2

Approach

For screening purposes, it is appropriate to establish as a goal the limitation of the maximum annual E to a member of the most

1.2 APPROACH

/

3

critically exposed group, i.e., to insure that no individual is likely to receive a dose that exceeds some recommended limit. The approach used to calculate the screening doses in this Report is to first review the current models for estimating the dose to individuals for each pathway. Then, based on an extensive literature review, recommended parameter values are presented for eight different land-use scenarios describing the present or intended use of the contaminated site. These chosen parameter values are used to calculate the highest annual E from external exposure or committed effective dose [E(.s)l from inhalation or ingestion that would be delivered by both the nuclide and its progeny for each radionuclide considered for each land-use scenario.' Prudently conservative values are suggested for the uncertainty in each parameter value used in the models and scenarios. This uncertainty reflects both true variability due to biological and environmental variations, human lifestyle differences, etc., and lack of knowledge as t o the correct mean or central tendency. If possible, this uncertainty is characterized by a distribution parameter such as a standard deviation or geometric standard deviation. Sometimes,only a range can be estimated for a particular parameter because of a lack of adequate information regarding the true distribution of potential values. In those cases, a triangular distribution is assumed, encompassing the estimated 5 to 95 percent limits. To estimate the likely median and maximum dose to an individual in the critical group, the distribution of potential individual doses from each pathway and the total dose from all relevant pathways is calculated stochastically using the estimated uncertainty for each chosen parameter value. The 95th percentile of the calculated total dose distribution was then used as an estimate of the maximum dose to any individual in the critical group. Section 7 of this Report describes the methodology used for these stochastic calculations. These "maximum" dose estimates and a postulated "acceptable" maximum annual dose to any individual from exposure to any single contaminated site is used to estimate the recommended soil concentration screening limits presented in this Report. Each of the calculated screening doses can be thought of as a conservative estimate of the maximum annual E(T)to a representative member of the most 'Usually the maximum annual dose occurs in the first year of exposure. However, for a few very long-lived nuclides, the ingrowth of progeny may result in a higher annual dose sometime in the future. For those nuclides, the highest annual dose in any year over the next 1,000 y is calculated, and this value is used for setting screening criteria. A 1,000 y limit is consistent with delays used in the past for screening guidance (NCRP, 1996). In any case, only the dose for a few very long-lived radionuclides, primarily members of the naturally occumng uranium and thorium series, will continue to increase beyond 1,000 y (see Appendix A).

4

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1. INTRODUCTION

exposed population group. The most exposed population will depend not only on the land-use scenario, but also on the particular radionuclide. Separate calculations are made for children and for adults and the highest dose is used for the guidance presented in this Report. The screening limits and total screening doses are tabulated in Section 2. The calculated E(z) for each pathway and land-use scenario are tabulated in Appendix A Using specific guidance provided in this Report, the reader can use the individual pathway doses for site-specific assessments and for the summation of doses from sites contaminated with more than one nuclide. By following this approach, the calculated annual E(z)and screening limits provided in this Report are expected to be much more realistic than previous screening models, yet still conservative. Previous NCRP screening recommendations (NCRP, 1989; 1996) assigned a single safety factor of 10 to the calculated total screening dose which was already a conservative value. The more extensive uncertainty analysis used for this Report provides separate land use, pathway and nuclide dependent screening limits. This approach eliminates the need to assume a single overly conservative overall safety factor to cover the range of all possible exposures and thus avoids undue conservatism in recommending screening soil concentration limits for particular sites and nuclides. NCRP Commentary No. 8 (NCRP, 1993a), for example, states that the screening limits given in Commentary No. 3 (NCRP, 1989) for airborne deposition scenarios were likely to be too conservative for external exposure while perhaps not conservative enough for ingestion of food grown on nutrient-depleted soil. Thus, by providing only a single screening limit for all land uses, the screening guidance in Commentary No. 3 is probably too conservative for certain land-use scenarios, for example where no food is grown. Finally, the approach taken here assumes that if the measured nuclide concentration in soil can result in an annual E(z) higher than the recommended screening limit, a site-specific dose assessment will be performed. Such an assessment might consider not only annual doses but also the lifetime risk to potentially exposed populations.

1.3 Land-Use Scenarios Separate calculations have been made for eight different land-use scenarios. The intent is to differentiate among uses where different dose pathways might dominate, the most exposed population group

1.3 LAND-USE SCENARIOS

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5

might differ, or the range of a particular critical parameter might be more limited, while keeping the number of different scenarios considered to a manageable level. This results in a narrower range of possible individual doses for each projected land use. Thus, the screening limits for particular land uses are lower than would be required to maintain a comparable level of conservatism if all possible land uses were lumped together. The land-use scenarios chosen are described below.

1.3.1 Agricultural (AG) This scenario is intended for sites used primarily for food production for human consumption, e.g., vegetables, fruit, grain, etc. It is assumed that there are no dwellings on the contaminated site itself in order to more clearly distinguish this type of land-use from a rural land-use scenario. Thus, only adults are assumed to be exposed via inhalation and/or external radiation, although children and infants, in addition to adults, may be exposed via ingestion of contaminated food produced on the site. The most exposed individuals would depend on the dose pathway and radionuclide. If external exposure and/or inhalation is the most important dose pathway, the most exposed population would be farmers working on the site. If the ingestion pathway is significant for the nuclide in question, infants and children might be the most exposed population, even if they do not live on the site. Separate calculations of the maximum dose are made for adults and for children and infants and the highest dose used for the screening limits presented for this scenario.

1.3.2

Heavily Vegetated Pasture (PV)

This scenario is for sites used primarily for milk or meat production. Again, it is assumed that there are no dwellings and thus no direct external or inhalation exposure to children or infants. The most exposed population for the external radiation and inhalation dose pathways would be adults working on the land. For the ingestion dose pathway, the most exposed population group would be individuals who obtain most of their meat and milk from animals ingesting fodder grown on the site. Thus, the most exposed population group for the ingestion pathway will depend on the particular radionuclide since adults ingest more meat than children or infants but the dose per unit intake for many radionuclides is higher for infants and children (see Section 5). As for the agricultural scenario, separate

6

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1. INTRODUCTION

calculations of the maximum dose from all relevant pathways are made for adults and for children and infants and the highest total dose used for the screening limits presented for this scenario. This scenario should also be used for agricultural sites used to grow feed for meat or milk production.

1.3.3 Sparsely Vegetated Pasture (PS) This scenario is identical to the previous except that the site is assumed to be located in an arid area. Resuspension of surface soil is assumed to be higher at these sites. A typical site might consist of open range land. Grazing animals are assumed to get less of their total diet from the site than for more heavily vegetated sites. This scenario should be used for sites where one might expect higher than average resuspension. It should also be used for sites with a higher than average potential for ingestion of soil by both animals and humans, either directly or via resuspension onto vegetation.

1.3.4 Heavily Vegetated Rural (RV) This scenario is intended to include open fields and forested sites. Some ingestion of contaminated food is assumed from gardens, wild game, or fruits and mushrooms fmm forests. It is assumed that there may be dwellings directly on the contaminated site. The most exposed population group will depend on the particular radionuclide but would most likely be children and infants living on the site that ingest milk from a backyard cow. This scenario can also be used for farms where persons live on site, as contrasted to the agricultural land-use scenario where it is assumed that no people live on site.

1.3.5 Sparsely Vegetated Rural (RS) The rural-sparsely vegetated scenario is similar to RV except, as for the sparsely vegetated pasture scenario, the site is likely to be in an arid area. Less food is assumed to be produced on these types of sites. However, resuspension of surface soil is assumed to be higher than for more heavily vegetated sites.

1.3.6 Suburban (SU) This scenario is for residential properties. Some minor food production such as that from vegetable gardens is assumed to occur. The

1.3 LAND-USE SCENARIOS

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7

most exposed population would likely be children who live on the site, play outdoors, and ingest vegetables grown on the site and perhaps some contaminated soil.

1.3.7

No Food Suburban ( S N )

This scenario is identical to SU except that no food is produced from the site. Parks, schools, developed recreational sites and residential lawns would be included in this land-use scenario. The most exposed individuals would likely be children playing outdoors, inhaling resuspended soil and possibly ingesting contaminated soil.

1.3.8

Construction, Commercial, Industrial (CC)

This scenario is for sites where the soil is likely to be disturbed due to present or future construction activities or activities involving earth moving or for sites used for industrial or commercial purposes. It is assumed that there are no dwellings on the site and no children are exposed. The critically exposed group consists of adult workers from external radiation exposure andlor inhalation and ingestion of suspended contaminated soil. The doses from construction and earth moving activities are likely to be short term and thus the screening limits will be somewhat more conservative than for long-term exposures.

2. Recommended Screening Limits for Contaminated Soil This Section lists screening guidance for over 200 radionuclides with half-lives >30 d. This guidance is based on the screening dose calculations described in Sections 3 through 7. If the exposure is from a single site, as for the doses calculated in this Report, the NCRP recommends that the dose to the maximally exposed individual from any single set of sources, e g . , at a particular site, should not exceed 0.26 mSv y-I. This is intended to insure that the total dose rate to that individual from all man-made sources other than medical exposures does not exceed 1mSv y-I. Thus, to remain consistent with these recommendations, the screening limits recommended in this Report are based on limiting the maximum E(z) rate to any individual to 0.25 mSv y-l. Dose limits that appear to be lower than that used here are currently under consideration by regulatory bodies such as the U.S. Nuclear Regulatory Commission (NRC) and U.S.Environmental Protection Agency (EPA). However, the limits proposed by NRC and EPA, which are intended for cleanup of contaminated sites, are based on the median dose to an individual in the most critically exposed population rather than the maximum dose to any individual as used in this Report. Thus, the screening limits proposed here will generally be considerably higher (i.e.,more conservative) than the values proposed for regulatory purposes. It is again emphasized that the guidance proposed in this Report is for use in screening and is not intended for use a s cleanup criteria, since the conservative nature of the guidance given here could result in greater amounts of soil being removed than would be necessary with realistic, site-specific calculations.

2.1 Recommended Screening Limits

Table 2.1 lists the recommended screening limits for each nuclide considered in this Report. The limits were calculated by dividing 0.25 mSv by the calculated "maximum"screening total dose per unit

Isotope

T,,,(d)

TABLE2.1 -Recommended screening limits (Bq

Isotope

AG PV

PS

RV

RS

SU

TABLE 2.1 -Recornmended screening limits (Bq kg-').a (continued) SN

Isotope

.-Recommended screening limits (Bq kg-')." (continued)

Gd-153 Eu-154 Eu-155 Tb-157 Tb-158 Dy-159 Tb-160 Ho-166m Yb-169 Tm-170 Tm-171 Hf-172 Lu-173 Lu-174m Lu-174' Hf-175 Lu-176 Lu-177m Hf-178m Ta-179 Ta-180 Hf-181 W-181 Hf-182 Ta-182' Re-184m Re- 184' 0s-185 W-185 Re-186m

Isotope

TABLE 2.1 -Recommended screening limits (Bq kg-').' (continued)

9.4E + 04 1.9E + 02 5.3E t- 02 2.OE + 01 2.1E + 02

3.2E + 02 1.3E + 05 (4.8E+ 03 2.5E + 06 3.33 + 05

Bk-249 Cf-249 C f-25W Cm-250 Cf-251

RV 9.8E + 04 2.OE + 02 7;2E+B2 3.2E+01 3BE + 02

PS 1.2E + 05 2.4E + 02 6.73 + 02 3.2E+ 01 2.8E + 02

W

2.OE + 06 3.9E + 03 8.7E + 03 3.OE + 02 4.OE + 03 5.3E + 04 l . l E + 02 2.8E + 02 1.4E + 01 1.3E + 02

RS l . l E + 05 2.3E + 02 1.OE + 03 4.7E + 01 3.7E + 02

SU

1.3E + 05 2.7E + 02 2.OE + 03 7.7E +01 5.5E + 02

SN

8.8E + 04 1.8E+ 02 4.6E + 02 2.4E + 01 2.2E + 02

CC

"Average Bq kg-' dry soil measured over top 5 cm of soil (see Section 6). bSee Section 2.2.3 for additional guidance regarding the screening limits for naturally occurring radionuclides. 'A decay product of another radionuclide with half-life >30 d. If this nuclide is present only as a result of parent decay, its dose is included in that of i t s parent and only .the screening limit for the parent need be applied.

AG

Tvz (d)

Isotope

TABLE2.1 -Recommended screening limits (Bq kg-').a (continued)

2.1 RECOMMENDED SCREENING LIMITS

1

17

soil concentration in Sv (Bq kg-')-' from Table 2 Z 2The doses given in Table 2.2 represent the median and maximum annual dose (calculated using the median dose and safety factor) over the next 1,000 y from all pathways from the listed parent nuclide and all of its progeny for the most critically exposed population group. The dose distributions were calculated using a Monte Carlo technique as described in Section 7. The calculated doses for each dose pathway and landuse scenario are given in Appendix A along with the delay indicating the year of maximum exposure and whether the most exposed population consists of children or of adults. Progeny with half-lives >30 d are also listed separately in Tables 2.1 and 2.2 since the progeny may be present at the time the contamination is discovered even though the parent has already decayed away or the contamination scenario may have been such that the particular decay product was separated from its parent prior to release to the soil. However, if a radionuclide is present in the soil only as a result of decay of a precursor also present in the soil, only the screening limit for the parent should be used because the dose from the daughter product is included in that of the precursor. As described earlier, the screening limits given in Table 2.1 are designed to restrict the total annual dose to any exposed person from a single contaminated site to <0.25 mSv y-I as recommended in NCRP Report No. 116 (NCRP, 1993b). The units a r e Bq kg-' dry soil. The radionuclide concentration in the soil is assumed to have been determined in a manner consistent with the guidance given in Section 6. The half-lives and radionuclide transformation data used for the dose calculations were taken from ICRP (1983) and are listed in Appendix B. The screening limits in Table 2.1 for the transuranics are, in general, less conservative than presently used. For example, EPA guidance (EPA, 1990b)recommends a screeninglimit for 239pUof 8 kBq m-2 in order to limit the dose to <0.1 mSv. If the plutonium is uniformly distributed to a depth of 5 cm, this would correspond to a concentration of about 100 Bq kg-' as opposed to 300 to 7,000 Bq kg-' recommended here based on a limiting dose of 0.25 mSv. For perspective, the limits given in Table 2.1 may be compared with actual levels measured in soils from some documented contamination scenarios. Uranium-238 concentrations measured outside the Fernauld site averaged 400 Bq kg-' in soils up to 1 k m from the site boundary with a maximum of 1,100 Bq kg-' at the fence line Table 2.2 actually lists the calculated median dose and the safety factor which is the ratio of the 95th percentile to the median. The maximum dose used is the product of these two entries.

Dose

Rb Dose

Rb Dose

Rb Dose

Rb Dose

Rb Dose Rb

Dose

CC

Rb

2.9E+OOA

I.4E-W

Si32

2.6E+WA I.IE+OlA

6.6E-b9 3.4E-12

2.8E+COA

26Ec00A 2.6E+OOA

I.1E-07 8.7E-08

2.9E 08 2.6E+OO A

-

4.6E -07 2.6E+OOA

6.OE-128.OE+WA

2.6E+cOA

- 13 9 . m + 0 0 A 2.3E-07

6.9E

3.OE - 13 9.3E+OO A

5.4E+OOA

1.m-11 2.B-07

8.9E -07 2.6E+OOA

8.2E+OOA 2.8E+OOA

6.92-12 2.m-lo

2.6E+OOA

1.1E-06

AI-28

3.8E+OOA 7.9E 07 2.6E + W A

Ns-22

-

2.G-10

Rb

RS

Be- 10

Dose

RV

4.OE-W2.5E+WA

d

PS

Be-7

Nuclide T ,

PV

TABLE 2.2-Total dose, Sv (Bq kg-')-', and ratios of 95th percentile to median for each land-use scenario."

2

2

V1

Bu

1

C)

crl 0

2 8 k'

C)

rn

N .

Nuclide T ,

d

2.4E+OOA 2.4EtM)A 2.4EcOOA

9.3E-07 3.5E-06 a.9E-08

2.5E+OOA

Rb

1.4E-07

Dose

AG

a.lE-08

2.7E-08

7.1E-07

1.1E-07

Dose

Rb

2.6EcOOA

2.5E+OOA

2.6E+OOA

2.6E+OOA

PV

3SE-08

2.8E-08

7.4E-07

1.lE-07

Dose

Rb

2.6E+OOA

2.5E+mA

2.6E+OOA

2.6E+OOA

PS

8.OE-08

6.8E-08

18E-06

2.8E-07

Dose

Rb

2.2E+OOC

23E+OOC

2.2E+OO C

2.4E+OOC

Rvb

8.9E-08

76E-08

2.OE-06

3.LE-07

Dose

Rb

2.3Et00 C

2.3E+OOC

2.2E+OOC

2.4E+OOC

RS

land-use ~ c e n a r w(continued) .~

7.8E-08

6.8E-08

1.8E-06

27E-07

Dose

Rb

2.2E+OOC

2.3E+MC

2.2E+OO C

2.3E+OOC

SU

7.8E-08

6.I)E-08

1.8E-06

2.7E-07

Dose

Rb

2.2E+OOC

2.3E+OOC

2.2E + 0 0 C

23E+OOC

SN

TABLE 2.2 -Total dose, Sv (Bq kg-') -',and ratios o f 95th percentile to median for each

2.6E+OOA

Rb

3.2E-08

2.5E+OOA

2.8E-082.5E+WA

7.3E-07 2.6E t 0 0 A

1.U-07

Dose

CC

I9

2 0

E g

C]

vl

PS

\

o

Nuclide T , , d Dose

Rb Dose

PV

Rb D08e

PS Rb Dose

RV Rb DOB~

RS

Rb

land-use scenario." (continued)

Dose

SU Rb Dose

SN Rb

TABLE 2.2-Total dose, Sv (Bq kg-')-', and ratios of 95th percentile to median for each

Dose

CC

Rb

F)

Rs-186m

W-185

Os-185

Kc-l&lC

Re-184m

Nuclide Tm.d Dose

Rb

DOBe

Rb h e

Rb h

e

Rb

land-usescenario." (continued)

TABLE 2.2-Total dose, Sv (Bq kg-')-', and mtios of 95th percentile to median for each

Dose

Rb

2

m

3.IE-01 2.5E-08

2.8E+02

1.OE+O2

5.6E+01

Es-264

Pm.267

Md.258

1.OE +01 C

2.7E+OOA

2.4E+OOA

8.9E+OOC

8.4E+OO C

29E-08 2.OE-09

2.5E-07

1.9E-08

3.1E-08

Z.BE+OO A

1.4E-08

2.6E+OO A

7.IE+WA

6.9E+mA

3.OE- 10 4 4 E + W A

2SE-01

2.7E-09

3.43-09

6.7EcOOA

3.1E+OOA

26E+OOA

6.1E+OOA

6.5E+OOA

2.6E-09

4.4E-08

6.9E-07

6.OE-08

4.23-08

7.9E+OOC

2.6E+OOC

2.3E+OOC

8.5E+OOC

7.7E+OO C

2.8E+OO C 5.3E+OO C

6.7E-08 8.3E -09

5.3E+OOC 6.1E+OOC 2AE+ OO C

1.4E-07 laE-07 6AE-07

2.3E+WC

5.9E-07 2.lE-OB

6.3E+W C

2.4E+WC

8.1E+00C

3.6E-08 4.2E-08

6.1E+OO C

3.43-08

2.4E+OOC 5.9EcOOC

3.7E-08 8.6E-10

7.5E+OOC 9JE+OOC 2.3E+OOC

1.3E-08 1.OE-08 6.8E-07

2.6E-07 2.7E-09

7.OE+OO A

3.4E+OOA

6.8E+OOA 2.6E+OOA

2.4E-08 2.6E-08

7.3E+OOA

4.OE-08

'Dosc?i are median &(T) per Bq kg-' dry soil to s member of critical population. R = mtia of 95th ponsntile to median dose. 'AD A following mtio idieatoa critical population condsts of adultq a C indicates critical population consists of children andlor infanta Soil nuclide mncenbation ia assumed unifonn lor 30 em 'A decay pmducl whose dose is also included in parent done. Doses indude contribution fmm p-y aod are marimurn annual dose over a 1,000 y interval See Appndi. A for delay times a d individual pathway dosea

2.9E-09

6.6E-08

6.1E+01

Cf-254

6.OE-08

9.6E+O2

Cf-262

26

1

2. SCREENING LIMITS FOR CONTAMINATED SOIL

(Stevenson and Hardy, 1993) while the screening limits in Table 2.1 range from 600 to 4,000 Bq kg-'. Plutonium measured in soil in the environs of Rocky Flats were (75 Bq kg-' offsite, but levels as high as 750 Bq kg-' were measured at the inner fence line (Krey and Hardy, 1970). Global fallout plutonium levels in soil are generally about 1 Bq kg-' (UNSCEAR, 1988) while plutonium measured in soils i n populated a r e a downwind of t h e Nevada Test S i t e were usually <5 Bq k g ' (Beck and Anspaugh, 1991). Similar values were measured downwind of Chernobyl to a distance of 50 km (UNSCEAR, 1988). These values can be compared to the screening limits in Table 2.1 for 2 3 9 Pwhich ~ range from 300 to 7,000 Bq kg-'. Cesium-137 concentrations measured in some soils in Sweden from the Chernobyl accident exceeded 1,000 Bq kg-' (UNSCEAR, 1988) compared to the screening limits in Table 2.1 which range from 100 to 450 Bq kg-'. The screening limit for '37Cs was also exceeded in many areas of eastern Europe as well as in the areas of the former Soviet Union directly downwind from Chernobyl (Likhtariov et al., 1996; UNSCEAR, 1988). For contrast, '37Csconcentrations from global fallout ranged from 25 to 75 Bq kg-' (UNSCEAR, 1988) and for sites downwind of the Nevada Test Site from 25 to 35 Bq kg-' (Beck and Anspaugh, 1991). Although the median screening doses calculated for 238U,226Raand 210Pbare comparable to values calculated in previous NCRP guidance (NCRP, 1984a), the screening limits for naturally occurring radionuclides 238U,232111and 40Kand their progeny represent a special situation which is discussed later in Section 2.2.3

2.1.1 Multiple Nuclide Contamination If a site is contaminated by more than one nuclide or nuclide series, e.g., 137Csand lS4Cs,then it will be necessary to ensure that the total of their maximum annual E(z) will not exceed 0.25 mSv. In general, this can be accomplished by ensuring that: where: Si SFi

the measured concentration of nuclide i the corresponding screening limit for nuclide i from Table 2.1 The result will be approximate since the maximum (95th percentile) doses are summed algebraically rather than stochastically combined (e.g., the median and 95th percentiles of a sum of approximately lognormally-distributed doses will generally not equal the corresponding sum of the medians or 95th percentiles, particularly if one = =

2.2 APPLICATION OF SCREENING LIMITS

1

27

or more of the distributions are very skewed). However, the result is sufficiently conservative for screening. For site-specific assessments, it would be appropriate to recalculate the doses and uncertainties and sum them stochastically allowing for nuclide-related correlations that might result in a broader distribution of doses. As mentioned previously, if both a parent and its decay products are present in the soil, the latter as a result of the decay of the parent in the soil, only the screening limit for the parent need be used since the dose from all decay products is included in the parent dose. 2.1.2 Alternate Limiting Doses

As discussed in Section 2.1, the screening recommendations in this Report are predicated on limiting the annual E(z) to any individual from a contaminated soil site to 0.25 mSv. If subsequent NCRP guidance or federal or state guidance modifies this recommended limit, the applicable screening limits in Table 2.1 can be derived simply by scaling the values to the chosen dose limit, L, i.e.,

SFnew = (SF x L,J0.25

mSv)

(2.2)

2.2 Application of Screening Limits

In general, if the measured surface soil concentration is below the applicable value specified in Table 2.1, it is unlikely any individual will exceed an annual E(z) &om that source of 0.25 mSv. Thus, further action would not be required. If the concentration in soil exceeds the screening limit, a site-specific assessment should be carried out. Such a site-specific assessment should follow a graded approach that takes into account by how much the measured concentration exceeds the screening limit. Even if the measured concentration in the soil is hlgher than the screening recommendation, a simple evaluation of the applicability of the calculated screening doses and assumed landuse scenario to the particular site may be sufficient to show that the dose to any individual will actually be less than recommended limits (see Section 2.2.4). If not, a more extensive site-specific dose assessment may be required, including a more elaborate assessment of the true soil concentration levels and spatial variation. 22.1 Applicability of Tabulated Land-Use Scenarios

This Report provides guidance for a limited number of land-use scenarios. Although they cover a broad range of situations, they are

28

/

2. SCREENING LIMITS FOR CONTAMINATED SOIL

not all inclusive. Actual sites may have mixed use, i.e., a residential site may be subject to some construction activity. Other sites may be modified in a way that reduces the exposure from contaminated soil to nil, e-g., paved over. Judicious application of the tabulated screening limits would mandate that for sites with land use not exactly conforming to those for which the guidance is formulated, one should apply the screening limits for the closest most conservative match. If none of the land uses examined are appropriate, alternate screening limits should be derived. This can be accomplished easily using the individual pathway screening doses and safety factors given in the appendices. One should note that for nuclides where the external dose pathway dominates, the dose is less sensitive to the land use than for nuclides where either the inhalation or ingestion dose predominates. In addition, the ingestion dose is sometimes very sensitive to whether milk or meat is produced on the site. Other factors to consider in applying the screening limits for the various land-use scenarios are the true spatial extent of the contamination and the precision and accuracy of the soil radionuclide concentration determination. The projected future, as well as the present, utilization of the site should also be considered, particularly if the contamination is due to very long-lived radionuclides. If a site could be used in the future in a manner that would result in more restrictive limits, for example to grow food, one may need to apply more restrictive screening limits than those based on present utilization. Whether people may eventually live on the site and whether infants andlor children may be exposed are important as well.

23.2

Special Situations

Some special situations might result in individuals receiving doses higher than allowed by the screening limits in Table 2.1. The reason for this is that incorporating special situations which generally affect only a very small number of people can result in higher doses and that the available models for estimating doses to individuals can have a very large uncertainty in special situations. If these doses, with their accompanying uncertainty, were folded into the doses from the more important pathways, the calculated screening limits that would result for many of the land-use scenarios considered would be so low as to not be of practical use as a screening tool. In particular, as discussed below, if a large public aquifer underlies an area of widespread contamination, a site-specific assessment will probably be necessary. Similarly, if individuals are known to drink

2.2 APPLICATION OF SCREENING LIMITS

/

29

significant quantities of contaminated goat's milk produced on a contaminated site, rather than cow's milk,the recommended guidance for nuclides and sites where ingestion of cow's milk is a critical pathway may not be sufficiently conservative and a site-specific assessment could be required. The following paragraphs discuss some of the contamination scenarios where the screening limits given in this Report may not be applicable and, therefore, that a site-specific assessment should be carried out.

2.2.3

Nonapplicable Contamination Scenarios

The guidance given in this Report is not to be applied to sites where potential contamination of an aquifer or drinking water source is likely. However, some previously published screening guidance assumes ground water contamination to always be a major contributor to potential dose for a number of important radionuclides resulting in very conservative screening levels (NCRP, 1996; NRC,1994). Because the uncertainty in predicting ground water contamination for generic sites is so large, these models generally use very conservative assumptions that often result in the ground water pathway dominating the calculated screening doses (Kennedy and Strenge, 1993; Yu et al., 1993). The potential for ground water contamination due to leaching from surface soil depends on a large number of factors, all of which are very site specific. These include the extent of contamination, type of soil, soil porosity, infiltration rate (generally the amount of precipitation), the depth, size and location of the aquifer, the amount of dilution by uncontaminated water, the number, depth and location of wells, and the amount of drinking water obtained from those wells (EPA, 1996a;Kennedy and Strenge, 1993; NCRP, 1996; Yu et al., 1993). The presence of drainage channels or fissures in the underlying soil and rock are also important factors in the rate at which leached activity will be transported to an underlying aquifer. In general, most soils retain most radionuclides very tightly. The retardation factor, i.e., the relative rate of movement of the radionuclide through the soil compared to the movement of infiltrating water, is characterized by the soil partition coefficient, Kd (Bq kg-I soil per Bq L-' of solution). For "Sr and 13?Cs,values of Kd range from 20 to 2,000 which implies a range of approximately 100 to 500 for the corresponding retardation factors ( M A , 1994). For sites with an average annual precipitation of 30 to 90 cm y-I, this implies that the migration velocity of gOSrand I3?Cs,would be <1 cm y-'.

30

1

2. SCREENING LIMITS FOR CONTAMINATED SOIL

Thus, except for a few elements with very low Kd and very long halflives, leaching of radionuclides from contaminated surface soil is not likely to contribute significantly to the dose even a t sites with an underlying aquifer. Contamination of drinking water may also occur via surface runoff to a surface water source which is consumed directly or feeds an underground aquifer. Again, this would be very site specific. At any rate, for most contaminated sites, the conservatism built into the screening factors given in thls Report should be sufficient to account for any contribution from this pathway. Furthermore, any significant dose from contamination of ground water will be offset by a reduction in dose from the other pathways due to removal of the leached radionuclides from the surface soil. However, for sites overlying a large public aquifer, a site-specific evaluation should be carried out, particularly if the size of the contaminated area is also large. Guidance on modeling contamination of ground water a t specific sites have been provided by Yu et al. (1993) and EPA (1996a; 1996b). The screening limits are also not intended to apply to sites with sources buried below the surface. NCRP Report No. 123 (NCRP, 1996) provides screening guidance for buried waste. and Screening limits are given for the natural emitters, 226Ra (and their decay products) in Table 2.1. These screening limits are consistent with previous NCRP guidance for natural uranium, 226Ra and 'lOPb when normalized to the same dose limit (0.25 mSv y-l) (NCRP, 1984a) and are intended to apply only to contamination in excess of the natural background for that site. Because of the conservatism provided in the calculated screening doses in order to limit the annual dose to any individual to <0.25 mSv, which is about one-fourth of the average dose from natural background (excluding radon exposure) (NCRP, 1988; 1993b), the recommended screening limits (4 to 20 Bq kg-' for 226Raand 2 to 70 Bq kg-' for 232Th)are for some land-use scenarios less than average background soil concentrations for the United States. Background concentrations are about 10 to 70 Bq kg-I for both 226Raand with an average of about 40 Bq kg-' (NCRP, 1988; UNSCEAR, 1988). Thus, these screening limits may be undistinguishable from background a t some sites without more intensive soil sampling and analysis. Gogolak et al. (1997) discusses sampling techniques for distinguishing contamination from background levels. Since, a s discussed in Section 7 of this Report, a major part of the conservatism in the screening factors (about a factor of two) is due to the uncertainty in the measurement of the nuclide concentration in the soil. Once the actual increase in

2.2 APPLICATION OF SCREENING LIMITS

1

31

soil concentration has been estimated more pre~isely,~ it would be acceptable to apply a value of about twice that given in Table 2.1 for naturally occurring radioisotopes such as 226Raand 232Thas a screening level to determine if a more extensive site-specific evaluation is required. Again, it is important to remember that the purpose of these screening limits is to provide guidance as to whether any further action is required. The fact that the calculated screening limits may be less than the ambient background level at some sites only suggests that for those radionuclides, some additional sitespecific analysis will usually be required. The contribution of radon and its progeny to the inhalation dose, as shown in Table 2.1 and discussed in Section 4.3.6, has a major impact on the screening limits for 226Ra,accounting for most of the estimated "maximum" dose for some of the land-use scenarios. The exhalation of radon from the contaminated soil and average groundlevel concentration in air can vary greatly from site to site depending on local meteorology, soil type, and other factors. Thus, for any soil known to be contaminated with 226Ra,some additional site-specific assessment will generally always be required. Screening limits are also given for 40Kin Table 2.1. These range from about 600 to 1,500 Bq kg-'. However, since the amount of potassium in the body, and thus the dose from ingested and inhaled 40K, is under tight homeostatic control, only the dose due to external irradiation was included in calculating the indicated screening limits. Again, the screening limits are intended to be applied only to levels exceeding the natural background concentration which averages about 400 Bq kg-' (NCRP, 1988; UNSCEAR, 1988). Finally, screening limits are not given in this Report for 3H and 14C.NCRP Report No. 76 (NCRP, 1984b)discusses the special cases of these radionuclides resulting from their ubiquitous redistxlbution once released to the environment due to rapid and uniform mixing with their stable counterparts in the vicinity of the contaminated area. Because of this mixing and rapid redistribution, these nuclides are not likely to remain fmed in the surface soil. The behavior of 3H and 14Cin the environment have both been covered in previous NCRP guidance (NCRP, 1979; 1985b).

Site-Specitic Dose Assessments If the measured concentration in soil of a nuclide exceeds the guidance given in Table 2.1 or from Equation 2.1, a site-specific dose

2.2.4

3Naturallyoccurring emitters are generally in equilibrium with their decay products and additions due to a soil contamination scenario will probably not be, so one should be able to distinguish fairly small incremental concentrations of these long-lived radionuclides.

32

1

2. SCREENING LIMITS FOR CONTAMINATED SOIL

assessment must be carried out. A graded approach is recommended since (as discussed in Section 7) the screening limits are based on conservative dose estimates. It may be relatively simple to show that the potentially "maximally-exposed" individual for the site in question would receive a much lower dose than that used to determine the screening limit. First, the individual pathway dose estimates given i n Appendix A should be examined to see if the controlling pathway doses are actually relevant for the particular site and population. This may result in an immediate determination that the actual likely maximum dose based on present or planned use will not exceed the recommended guidance. For instance, if no milk,meat or vegetables are or will be produced on a site, or if no children or infants will be exposed, the dose for many nuclides for that site may be much lower than reflected in the recommended screeninglimits. If a particular parameter such as the pwture season length or the fraction of diet derived from the site is clearly less than the conservative values used for calculating the screening guidance, one may also be able to quickly demonstrate that the corresponding site-specific dose will be lower. If animals do not graze directly on the contaminated land, the effect of soil adhesion and uptake will be lower, and if milk is not obtained from a backyard cow, the actual milk ingestion pathway doses may be much lower than reflected in the guidance. The size of the contaminated area will clearly be a factor in judging the applicability of some of these factors. Second, if more than one pathway is important for the nuclide in question, one should determine whether the same individual is likely to be exposed to both pathways. As discussed earlier, the screening limits assume that the same individual may be exposed to all pathways. Thus, they allow for the possibility for the rural land-use scenarios that the infant or child drinking milk, for example, could also be the individual most exposed via the inhalation and external exposure pathways. For convenience in canying out this initial site-specific assessment, the individual ingestion pathway doses (milk, meat, vegetables, soil) are tabulated separately in Appendix A for each nuclide. If the limiting annual dose is still exceeded after the actual site usage and possible individual exposures are considered, it will be necessary to calculate site-specific "maximum" doses using the guidance given in Sections 3 through 5 of this Report. One should also estimate the uncertainty in these new dose estimates. If the soil concentration is significantly above the guidance in Table 2.1, a detailed assessment of the actual extent and distribution of radionuclides in the soil should also be carried out (see Section 6).

2.3 SCIENTIFIC BASIS FOR SCREENING LIMITS

1

33

Finally, even if the concentration in soil is below the screening limit recommended in this guidance, the principle of AWlRA should still be applied. If reasonable cost-effective measures can be taken to assure lower exposure to individuals than provided for by this guidance, such measures should be considered. The Council's recommendation on the application of ALARA to exposure from environmental contamination is discussed in NCRP Report No. 116 (NCRP, 1993b). 2.3 Scientific Basis for Screening Limits

Sections 3 through 5 of this Report discuss the scientific basis for estimating both generic and site-specific doses from each of the three major pathways: direct exposure, inhalation and ingestion. The models used to estimate dose are presented along with a discussion of each of the model parameters. Recommended values, with associated uncertainty, are provided for use in screening and site-specific assessments. Examples of calculated doses using these parameters are presented for selected important radionuclides. Section 6 discusses how to determine the average nuclide concentration in soil for use in applying the screening recommendations given in this Report. The importance of the variations in concentration across the site is discussed. Also discussed is how to use direct measurements of gamma-ray flux or free-in-air kerma to estimate surface soil radionuclide concentrations. Section 7 discusses the calculation of the distribution of possible doses to a representative individual for each pathway using Monte Carlo techniques and the uncertainty estimates. Finally, the appendices list the calculated E(z) for each land-use scenario for the most exposed population (adults or children) and the nuclide-dependent transfer and dose factors used for the screening calculations. The individual pathway doses are tabulated in Appendix A (including the individual doses from ingestion of vegetables, milk, meat and soil and the total dose) along with the delay used to calculate the highest annual dose.

3. Dose from External Exposure One of the principal dose pathways resulting from contaminated surface soil is external exposure from radiation emitted from the radionuclides present in the surface soil. The dose to individuals from external exposure will depend on a number of factors including the type and energy of the emitted radiations, the distribution of the source with depth in the soil, the relative amount of time spent outdoors versus indoors, the amount of shielding provided by on-site structures and the body itself, and even the moisture content of the soil. The values of most of these parameters will vary with projected land use. This Section discusses models for estimating the external E and the key parameters used in these models and their uncertainty or variability. The values used to calculate screening doses are discussed, and examples of calculated screening doses are given for the various land-use scenarios for selected nuclides likely to be found in environmental contamination scenarios.

3.1 Dose Model The average annual E from external radiation exposure, E,,, to a person standing on contaminated land can be calculated from the following expression: Ed = Df,,, where:

X

[To,,

+ (Ti,

X

SF)] x (CIA) x

W,X S(z, t ) (3.1)

the average annual E to an exposed individual per unit radionuclide concentration [Sv (Bq kg-')-'] from external radiation = the dose factor for a particular radionuclide in [Sv y-' Df, (Bq kg-')-' of dry soil]. It provides the free-in-air E for a person standing on top of an extended contaminated soil surface. The value of Df,, depends on the concentration depth profile; S(z, t ) S(z, t) = the radionuclide concentration in the soil averaged over the 1y interval for which the dose is calculated (Bq kg-') E,,

=

3.1 DOSE MODEL

1

35

of dry soil. The concentration varies with depth, z, beneath the ground surface = the mean fraction of the time spent outdoors on the conTo,, taminated site = the mean fraction of the time spent indoors in a dwelling Ti, located on the contaminated site SF = the shielding factor or ratio of the dose indoors to the unshielded outdoor dose = a density correction due to soil moisture, i.e., the ratio W, of the dry soil density to the actual in situ bulk density (CIA) = the ratio ofthe external dose to children to that for adults when children are present The factor that is the most important determinant of external dose is S(z, t ) . The distribution of radionuclides with depth in the soil is particularly important since it also affects the value for Df,. Structure shieldingcan result in a si@cant reduction in exposure. Indeed many screening calculations have ignored this correction completely. The soil moisture density correction, W,, is less significant. The above formulation applies only to external exposure due to radionuclides in the soil. The low penetration of most beta rays is such that E from pure beta emitters in the soil layer is relatively insignificant compared to that from a gamma emitter in the soil. Although the dose from beta rays emitted from the soil surface is included in the calculated dose factors, pure beta-emitting radionuclides generally do not produce significant penetrating radiation exposure compared to gamma emitters due to their short ranges (ICRU, 1996). The range of a 1 MeV beta ray is about 0.4 g cm-2 or about 3 m in air and 0.3 cm in soil. Furthermore, the beta-ray component of total exposure from a mixed beta-gamma source is small compared to that from the gamma rays. Because of their low penetrating power, pure beta-emitting radionuclides may produce large free-in-air exposure rates and relatively high skin doses, but they generally produce relatively small E. External exposure can also occur as a result of soil becoming attached to an individual's skin surface during outdoor activities. The dose from radionuclides deposited directly on the skin surface is not calculated explicitly for the screening guidance provided in this Report. This pathway is generally negligible compared to the dose from the soil radioactivity itself, particularly for gamma emitters. An exception might be if the skin of crawling infants or other persons spending significant time outdoors becomes contaminated with a beta-ray emitter and the radionuclide is allowed to remain on the skin for some time. This would result in a higher dose to the skin than might occur from

36

1

3. DOSE FROM EXTERNAL EXPOSURE

just standing on a contaminated surface. However, the calculated external doses are believed to be ~ ~ c i e n tconservative ly to allow for any measurable contribution from this pathway. A site-specific dose assessment may need to examine this potential pathway in more detail, however (Haywood and Smith, 1992). Exposure from soil brought indoors on shoes and clothing may also be a potential source of population dose, both from external exposure and via inhalation of indoor dust. For the purposes of screening calculations, however, this dose component has been accounted for by use of a conservative value for the housing shielding factor (SF).

3% Discussion of Model Parameters In the following sections, each of the important parameters needed to estimate the external E, including their uncertainty and variability, is discussed. 3.2.1 Concentration in Soil

The concentration in the soil, S(z, t), as a function of depth and time must be determined for each contaminating radionuclide. This value will change with time due to radioactive decay and buildup from decay of precursors. The soil radionuclide concentration will also change with time from disturbance by animals and from erosional processes. Depending on the circumstances of the original contaminating event, the type of soil, chemical form of the radionuclide and the amount of precipitation, S(z, t) is also time dependent to allow for percolation into the soil or redistribution by wind or water. Other factors that may act to modify the soil radionuclide concentration in the future, such as human activity, may also need to be considered. For screening, it is assumed only that the average concentration in the top 5 cm averaged across the site is measured at some time to with a coefficient of variation of <50 percent (see Section 6). As discussed later, a dose factor is then chosen with an uncertainty estimate that allows for various depth distributions of radionuclide concentration extending to a depth of 30 cm. An average annual concentration is then calculated for use in Equation 3.1. The average concentration over the first year for the parent is given by S = So x 11- exp( - ht1)Y(At') where So is the measured concentration in soil a t time to, t' = 365 d, TI&the nuclide half-life in days and A is

3.2 DISCUSSION OF MODEL PARAMETERS

/

37

ln2/Tm. If the total dose from all pathways for a particular nuclide and its progeny is higher in some other year over the succeeding 1,000 y due to ingrowth of progeny, then the concentration in soil of the parent and its progeny decayed to that year is used for calculating the individual pathway screening doses.4Thus, as discussed earlier in Section 1, for nuclides that may grow in from precursors, this average concentration is calculated using a delay such that one obtains the maximum annual total dose from both parent and progeny which might occur in any year during the first 1,000 y after the initial soil contamination is measured. The dependence of the calculated dose on the actual distribution of the radionuclide in the soil is discussed in the following paragraphs. 3.2.2

Effective Dose Factor

E to individuals standing on a contaminated soil surface has been calculated by a number of investigators (DOE, 1988; Eckerman and Ryman, 1993;Jacob and Meckbach, 1990;Jacob et al., 1990;Petoussi et al., 1991) for a large number of nuclides. The models assume a particular depth profile and then trace the emitted radiations from various depths in the soil through the soil interface, intervening air, and body organs, to calculate the energy absorbed in various body organs and from that, E to a "standard" man. The more recent calculations (Eckerman and Ryman, 1993; Jacob et al., 1990) utilize sophisticated transport theory methodology such as Monte Carlo or Discrete Ordinate techniques to simulate transport of the emitted particles or photons. Earlier methods (eg., DOE, 1988; Kocher, 1983) relied on a less accurate buildup factor approach. The dose factors recommended for use in previous screening models (DOE, 1988; NCRP, 1989) usually assumed a surface plane source, and thus provided very conservative estimates of the dose for real environrnental contamination scenarios. The calculations of Jacob et al. (1990) and Petoussi et al. (1991)are also for a single depth profile. However, they assume a plane source buried at a depth of 0.3 cm to account for the effects of ground roughness and initial penetration. Although the buried plane source dose factors are more realistic than the ideal plane source assumption, they also will generally provide conservative estimates. The calculations presented in Federal Guidance Report (FGR) No. 12 (Eckerman and Ryman, 1993)give the external E equivalent 'NCRP (1996) and Kennedy and Strenge (1993) give general formuli for calculating both the parent activity and that of the progeny as a function of the delay time.

38

/

3. DOSE FROM EXTERNAL EXPOSURE

(see Section 3.2.2.1 below), HE, to a Reference Man for over 800 radionuclides. These calculations were performed for a number of assumed depth profiles ranging from a plane surface source to an infinite homogeneous medium, including slab sources of various thickness. The calculations account for physical processes not always included in other compilations such as the contributions to dose from beta rays, bremsstrahlung, and x rays. They use the most current set of data on particle and photon emissions, branching ratios, and decay constants. Consequently, they are the current preferred dose factor data. Although source distributions that decrease exponentially with depth are not explicitly calculated in this compilation, they can be estimated by superposition of the tabulated data for other distributi0ns.j The dose factor calculations require a number of assumptions such as the spatial distribution both laterally and with depth; the size, shape, composition and orientation of the exposed target human subject; and the composition and density of the soil. They also require accurate data on radionuclide decay chains. The effect of these assumptions on the calculated dose factors is discussed below. 3.2.2.1 Effective Dose versus Effective Dose-Equivalent, Organ Dose. All of the aforementioned calculations use ICRP Publication 26 (ICRP, 1977)tissue weighting factors to calculate E (designated effective dose equivalent, HE). The NCRP (1993b) and the ICRP have recommended that ICRP Publication 60 (ICRP, 1991) tissue weighting factors be used for calculating E. Zankl et al. (1992) demonstrated that for external exposure, the newer weights result in <10 percent error for most photon sources. The only significant differences (-30 percent) arise when the source energy is very low ( < l o keV), and for these energies external exposure is generally insignificant. For all source energies in excess of 10 keV, the use of the older weights is conservative, i.e., results in a higher dose estimate. For some low-energy gamma emitters or for pure beta emitters, a site-specific assessment may wish to consider the dose to a particular organ (e.g.,bone surface or skin) using the detailed organ dose factors

T h e E for a 3 cm relaxation length exponentially decreasing soil radionuclide concentration can be approximated by 1.5 x (1.0 cm slab E ) + 1.1 x (5 cm slab E ) + 0.2 X (15 cm slab E ) where the 1, 5 and 15 cm slab doses in units of Sv (Bq m-3)-L are from Eckerman and Ryman (1993). Using data shown later in Table 3.4, it can be shown that this approximation slightly underestimates the dose from a true exponentially decreasing source by about 10 percent but preserves the total inventory down to 1 , 5 and 15 cm.

3.2 DISCUSSION OF MODEL PARAMETERS

1

39

provided in the above referenced reports. Table 3.1 compares the highest organ dose to E for a few selected nuclides. 3.2.2.2 Exposure Rate, Free-in-Air Kerma versus Effective Dose. The E in humans can often be inferred quite accurately from the Gee-in-air exposure rate or air kerma rate by multiplying the air exposure or kerma rate by a n energy-dependent estimate of Sv Gy-'. This conversion factor, which depends slightly on the assumed angle of incidence, can be obtained by comparing calculations of kerma and E equivalent for the same depth profile. The conversion factor is relatively source independent and averages about 0.8 for adults for most photon energies for rotational incidence (Chen, 1991; ICRP, 1996a; Jacob et al., 1986; 1990). A similar value of 0.7 has been recommended by UNSCEAR (1988; 1993) for the general population. Table 3.2 lists some estimates reported by various investigators of Sv Gy-' for adults as a function of source energy and angle of incidence. All the ratios in Table 3.2 are based on calculations for a plane surface source except those from Jacob et al. (1990) which are for a buried (0.3 cm beneath surface) plane source. However, considering how little the values vary with source energy above 100 keV, one would not expect the values for more deeply buried sources to differ sigmficantly from the values in the table. The ratio of E to air kerma, Sv Gy-', is somewhat dependent on body size. Table 3.3 compares values for children with those for an adult for the case of a buried plane source a t 0.3 cm below the surface.

TABLE 3.1 -Ratio of highest organ h s e to E from external radiation exposure for selected radionucl&s." Nuclide

Ratio

Organ

Bone Bone Skin Bone Bone Bone Bone Bone Skin Bone Bone

surface surface surface surface surface surface surface surface surface

"Based on dose factor for adults from Eckerman and Ryrnan (1993) for 15 cm slab. Note that the organs for which doses are given in Eckerman and Ryman (1993) are limited to gonad, breast, lung, bone marrow, bone surface, thyroid, and skin.

40

/

3. DOSE FROM EXTERNAL EXPOSURE

TABLE 3.2-Ratio of E to air kerma (Sv Gy-I) for various gamma emitters as a function of angle of incidence.

"Antero-posterior (radiation incident on front orthogonal to long axis of body) (ICRP, 1996a). bPosterior-anterior(radiation incident on back orthogonal to long axis of body) (ICRP, 1996a). 'Rotational (broad beam incident orthogonal to long axis of body while body rotates at uniform rate about the long h s ) (ICRP, 1996a). dIsotropic(fluence per unit solid angle independent of direction) (ICRP, 1996a). eStanding man, actual incident spectrum from buried (0.3cm) plane source (Jacob et al., 1990).

Tmm 3.3-Ratio of E to k e n a (Sv Gy-I)for selected radwnlsclides." Nuclide

Adult

Child

"From Jacob et al. (1990) and Petoussi et al. (1991) for a buried plane source at depth 0.3 cm. Comprehensive compilations of free-in-air exposure rate or air kerma from various concentration depth profiles for a wide selection of radionuclides can be used to estimate E (Beck, 1980;ICRU, 1994; Saito and Jacob, 1995).These data sets offer the advantage of allowing E estimates to be made for depth profiles other than the single distribution used by Jacob et al. (1990)or the restricted set of depth profiles offered by Eckennan and Ryman (1993).

3.2.2.3 Dependence of Dose Factor on Orientation. As indicated in Table 3.2, the ratio of the dose to a particular organ (and thus E

3.2 DISCUSSION OF MODEL PARAMETERS

1

41

to air kerma) is somewhat dependent on the angle of incidence of the irradiation. The more realistic transport calculations (Eckerman and Ryman, 1993; Jacob et al., 1990) assume the subject is standing and utilize the actual calculated angular incidence. The actual angular incidence was shown by Beck and de Planque (1968) to be weighted toward rotational incidence as opposed to isotropic incidence. A more isotropic source incidence results in longer pathlengths and thus more attenuation of gamma radiation in the body. The net effect is a 10 to 15 percent lower value of Sv Gy-I than for a rotational incidence. Isotropic incidence might be more appropriate for exposure indoors, particularly for low-energy sources. However, the standing adult model generally results in a more conservative dose estimate. 3.2.2.4 Dependence of Dose Factor on Nuclide Depth Distribution. The dependence of the dose estimate on the true source distribution is far more important, however, than is the receptor geometry. Table 3.4 illustrates the change in the calculated E for selected nuclides for the same total inventory (Bq m-2)distributed differently with depth. The dose from nuclides which emit primarily x rays, beta rays, and/or bremsstrahlung, e.g., T c , lZ9I,2a%, are more sensitive to the depth distribution than primarily gamma-ray emitters such as 13'Cs. It has been observed from contamination incidents that even a few weeks aRer deposition, radionuclides percolate significantly into the soil. Furthermore, even for fresh deposition, for most sites, ground roughness results in a reduction in the dose at I m height equivalent to that from assuming a relaxation length of about 0.5 to 1cm (Jacobet al., 1986). UNSCEAR (1988)suggests that radionuclides deposited on the surface can be represented by an exponential distribution with relaxation length of about 1cm between a few weeks and 1 y after deposition, increasing to a relaxation length of about 3 cm after about 1y. This trend is based on both weapons fallout data and data from the Chernobyl accident (Likhtariov et al., 1996). Miller and Helfer (1985) found that 13'Cs from global fallout percolated rapidly at first but stabilized in most open fields after about 2.5 y. Of course, at more arid locations, or for soils with a very high clay content, radionuclides might remain closer to the surface for a longer period. Some radionuclides, due to their physicalchemical form, can be bound in the soil structure and thus be less mobile. For example, highly calcareous soils will tend to bind strontium and cesium can replace potassium in the lattice structures of most clays and thus become less mobile. (However, at very arid locations, even clay soils might crack when dry allowing significant

2.OE - 02 2.5E - 05 4.6E - 04 6.7E - 06 2.33 - 03 1.3E-01 5.OE - 02 1.2E - 01 2.1E-02 6.OE - 04 1.3E - 02 2.5E - 03 7.2E - 05 3.2E - 05 2.4E - 04

Co-60 Sr-90 Y-90 Tc-99 1-129 CS-134 CS-137 Bi-214 Pb-214 Ra-226 U-235 Np-237 F'u-238 Pu-239 Am-241

-

2.OE - 04 8.83 - 02 3.3E - 02 7.8E - 02 1.4E - 02 4.OE - 04 8.3E - 03 1.2E - 03 -

9.OE - 04

-

3.8E - 03 5.OE - 04 1.2E - 01 4.3E - 02 9.9E - 02 1.8E - 02 5.OE - 04 1.OE - 02 1.6E - 03 -

1.OE - 03

-

-

BP

RL=O.l cmb 1.3E - 02 1.1E - 05 2.7E - 04 2.5E - 06 5.1E - 04 8.5E -02 3.1E - 02 7.9E - 02 1.4E - 02 3.63 - 04 8.3E - 03 1.2E - 03 5.5E - 06 4.8E - 06 9.9E - 04

1cm slaba

3.8E - 04

5.5E - 02 2.1E -02 4.6E - 02 8.1E - 03 2.OE - 04 4.5E - 03 7.OE - 04

2.OE - 04 2.OE - 04

RL = 3 cmb

7.7E -02 5.1E - 06 1.4E -04 1.OE - 06 1.2E -04 4.9E - 02 1.8E -02 4.6E - 02 7.8E -03 2.OE - 04 4.6E -03 6.0E - 04 1.3E -06 2.OE - 06 3.7E -04

5 cm slab"

4.2E -02 2.1, -07 6.63-05 3.9E -07 4.OE - 05 2.7E-02 9.6E -03 2.5E- 02 3.9E - 03 9.4E - 05 2.2E -03 2.4E - 04 4.73 -07 8.7E - 07 1.3E - 04

15 cm slaba

"From Eckerman and Ryman (1993). bFrom Beck and de Planque (1968) using Sv Gy-' from Jacob et al. (1990). RL refers to the relaxation length, i.e., the depth a t which the concentration is reduced to l/e of the value a t the surface, assuming an exponential decrease with depth. 'From Jacob et al. (1990). BP refers to a buried plane source a t 0.3 cm.

Plane"

Nuclide

True Depth F'rofile

actual depth distribution.

TABLE 3.4-E per unit inventory (nSv d-I per Bq m-2)for selected radionuclides as a function of

b P

rn

P

rn

H

5

rn

p

P

u

w

%

h3

3.2 DISCUSSION OF MODEL PARAMETERS

1

43

percolation to greater depths.) Studies at many sites tend to indicate that regardless of the distribution with depth, the soil in the top 2.5 cm tends to become rapidly mixed due to a number of physical and biological processes (see Section 4.2.2.2). This argues against the assumption of an ideal plane source, even for a very fresh surface deposition. As time passes, the depth profile may be modified by human activity, such as plowing or excavation, by animal and insect burrowing, or by physical forces such as cracking, frost heaving, and colloidal transport. The true total inventory (Bq m-2), surface soil concentration at z = 0 cm, and average concentration over the top 2.5 cm for various depth profiles can be calculated by assuming that the average concentration is 1 Bq kg-' over the top 5 cm (see Table 3.5). If the actual depth profile is much shallower than a uniform 5 cm slab, the surface soil concentration will be much higher, leading to a higher dose rate even though the total inventories are the same. If the depth distribution decreases exponentially, e.g., with a 3 cm relaxation length, the total inventory will be only slightly higher while the surface concentration will be about a factor of two higher than for a uniformly mixed thick slab. If the concentration is uniform with depth but extends below 5 cm, the total inventory will also be higher but the surface concentration will be the same. Since the highest TABLE 3.5 -Total inventory and surface soil concentration as a function of actual depth distribution." Actual Profile

Inventory (Bq m-a)

Concentration, 0 cm (Bq kg-')

Concentration, 0-2.5 cm (Bq kg-')

Buried planeb 1 cm slab 0.1 cm RL" 1 cm RL 2 cm RL 3 cm RL 5 cm slab 10 cm RL 15 cm slab "Assuming an average concentration of 1 Bq kg-' measured in the top 5 cm (see Section 6) and a uniform soil density of 1.6 g ~ r n - ~ . bInfinite plane source a t 0.3 cm (Jacob et al., 1990). 'RL is the relaxation length, i.e., the depth where the concentration has decreased by lle of the surface concentration assuming an exponential decrease with depth.

44

1

3. DOSE FROM EXTERNAL EXPOSURE

proportion of the dose from radionuclides in the soil comes from radionuclides in the first few centimeters, as can be seen from Table 3.6, a layer of source below 5 cm contributes much less to the total dose rate than the same layer above 5 cm, particularly for lowenergy gamma-ray emitters. Conversely, because most of the dose is due to nuclides close to the surface, even a thin layer of inactive soil will reduce the above-ground dose rate significantly. 3.2.2.5 Dependence of Dose Factor on Soil Composition. To calculate dose factors, a soil composition and moisture content must be assumed. Generally, for high-energy gamma rays, the calculated doses will be fairly insensitive to the chosen composition (Beck and de Planque, 1968; Eckerman and Ryman, 1993; Saito and Jacob, 1995). For low-energy gamma rayshrehmsstrahlung, the choice of soil composition can result in large differences in calculated dose due to the large differences in interaction cross sections, which vary as a function of atomic number, Z, a t energies of <50 to 100 keV. For example, the addition of a small amount of a high-Z element such as iron, can increase the photoelectric absorption of low-energy gamma rays significantly, resulting in a greatly reduced exposure rate and E (Eckerman and Ryman, 1993). Organs sensitive to lowpenetrating radiations such as the skin, will be particularly affected. Similarly, a high moisture content can reduce the average Z of the bulk soil. The dose from beta rays is strongly dependent on the soil composition, soil density, and moisture content. Thus, the published dose factors for pure beta-ray emitters should be used with caution since they are based on a particular composition which may not be appropriate for the particular site being evaluated. The dose factors adopted for screening assume a fairly low Z soil composition (Eckerman and Ryman, 1993). Thus, the low-energy

TABLE 3.6 -Approximate percent reduction in free-in-air exposure rate with amount of clean soil cover, z cm.qb

"Adapted from Beck and de Planque (1968). bEquivalentto fraction of total exposure rate from radionuclides in first z cm.

3.2

DISCUSSION OF MODEL PARAMETERS

1

45

gamma and beta values are generally conservative. However, a sitespecific dose assessment may need to consider the true soil composition if the contamination is from a low-energy gamma or pure beta emitter. Furthermore, for low-energy gamma and beta radiation, even differences in the density of the atmosphere for high-altitude locations compared to sea-level sites can result in differences of several percent in the respective doses at 1m above ground level.

3.26.6 Dependence of Dose Factor on Areal Extent of Contamination. AU the calculational models assume a source of infinite areal extent. For the high-energy gamma rays (>500 keV), about 80 percent of the dose for a source uniformly distributed with depth results from sources within a radius of about 10 m (Artuso, 1981; Beck and de Planque, 1968). The exact area from which photons reach the receptor depends on the source energy and the depth distribution. For more shallow source distributions, a greater fraction of the dose will come from larger distances. For example, the exposurerate in the center of a 100 m diameter contaminated surface source of medium energy gamma rays (0.5 to 2 MeV) is at least 85 percent of that for an area of infinite extent. The diameter sufTicient to approach an effective infinite area is even less for a source distributed uniformly with depth in the soil. In this case, about 95 percent of the exposure rate is due to sources within a radius of 10 m. For low-energy photons and x-ray or beta-ray emitters, the radius of the area that can be considered effectively infinite is even smaller than the above. Thus, even for contaminated sites of limited area, the infinite area calculations provide a reasonable estimate of the dose to persons located within the contaminated area. As shown in Figure 3.1, a s one

0

50

100

150

200

Distance from Center, m Fig. 3.1. Variation in dose rate versus distance from the center of a contaminated area of radius 90 m, adapted from Hubbell (1963) for plane source of 1.25 MeV gamma rays.

46

1

3. DOSE FROM EXTERNAL EXPOSURE

approaches and crosses the boundary of a contaminated area, the exposure rate drops off to about 50 percent of the maximum and then decreases rapidly. This example is for a plane surface source of 1.25 MeV gamma rays; the decrease from the boundary value would be even more dramatic for a more deeply distributed or lowenergy source. For screening purposes, it is assumed that the contamination is extensive and the critical population are those exposed a t the maximum level when onsite. Thus, if only a very limited area is contaminated, the guidance will be conservative. The guidance will also be conservative in that it assumes the entire area consists of contaminated soil. Many "soil" sites actually contain a great number of rocks. The rocks are generally not contaminated and thus the actual extent of the contamination, and the resultant external dose rate, is overestimated.

3.2.2.7 Accuracy of Dose Factor Calculations. All of the recent calculations rely on essentially the same decay scheme data (energies, branching ratios, etc.) and for most radionuclides of interest, the uncertainty in these basic data are small compared to the uncertainty arising from other factors. Thus, for a given nuclide depth profile, reported calculations of kerma tend to agree to better than 10 percent for the important high-energy gamma emitters and better than 25 percent for the low-energy gamma and the beta and x-ray sources (Beck, 1980; Chen, 1991; Eckerman and Ryman, 1993; Saito and Jacob, 1995). Beck (1980) estimated that his transport theorybased calculations of exposure rate in air for gamma emissions only were accurate to better than 5 to 10 percent for most nuclides for a given soil composition and depth profile. Differences between published values are probably due primarily to different assumed soil compositions (Eckerman and Ryman, 1993). The results are fairly insensitive to the assumed composition for the high-energy gammaray emitters. However, this is not the case for low-energy photons and beta rays. These results are highly dependent on the true soil composition and the assumed angle of incidence, both of which contribute to variability. The variability due to differences in soil composition are even larger for sources distributed with depth in the soil. Calculations of dose tend to be less certain, and reported values differ more than for kerma due to differences in the assumed geometry of the target organs (phantom geometry) and angle of incidence (Eckerman and Ryman, 1993). This is particularly true for the lowenergy gamma ray emitters. The results of Eckerman and Ryman often differ from those of Jacob et al. because of differences in organ geometry and the fact that Eckerman and Ryman include the dose from electrons and brehmstrahlung while the results of Jacob et al.

3.2 DISCUSSION OF MODEL PARAMETERS

1

47

are for gamma rays only. Based on comparisons of published dose factors, one can conclude that the dose Eactors reported by Eckerman and Ryman are probably accurate to within 10 percent for predicting E to adults for a given depth profile for high-energy gamma emitters (>I00 keV) and better than 25 percent for both low-energy gamma emitters and beta emitters. However, the true dose a t a given site may differ because the depth profile, soil composition, and soil moisture are not the same as those assumed.

3.2.2.8 Dose Factors Chosen for Screening Dose Calculations. In this Report, screening doses are calculated and screening criteria recommended based on an assumed average of 1 Bq kg-' (dry soil). In a screening program, the true depth profile or total inventory may not have been measured (see Section 6). Thus Table 3.7 compares the calculated E rate as a function of various depth profiles, assuming the average concentration in the top 5 cm is 1 Bq kg-'. Note that when the dose rate is calculated in this manner, the sensitivity to the true depth profile is greatly reduced, particularly for the highenergy (>300 KeV) photon emitters. For screening purposes, therefore, the maximum of the values tabulated by Eckerman and Ryman (1993) for a 1cm slab source and a uniformly distributed to infinity source has been used as a prudently conservative estimate. Assuming all the radionuclides are in the top 1cm provides a relatively conservative estimate comparable to that provided by the calculations of Jacob et al. (1990) for low-energy gamma emitters. The conservatism allows for possible contributions of x-ray and bremsstrahlung emissions from beta- and alpha-emitting nuclides that are not explicitly accounted for. In contrast, assuming a uniform source provides a conservative estimate for the high-energy photon sources. Since, as discussed above, the soil is frequently well mixed down to 2.5 cm, one can discount the ideal plane source as a reasonable model of reality. Note, as shown in Table 3.7, that if the radionuclides are actually distributed with a relaxation length of 3 cm, one will have overestimated the dose by about 30 to 40 percent for the high-energy gamma ray emitters and as much as 70 to 80 percent for the lowenergy gamma and x-ray emitters, and the beta emitters. Thus, in most cases, the external dose estimates will still be quite conservative. In order to estimate the possible variability in the dose factor from site to site, the maximum difference between E for the 1cm slab, the infinite slab, and the 3 cm relaxation length (from Eckerman and Ryman, 1993) was calculated. The variations in dose factors due to differences in soil composition and angular incidence are also dependent on source energy. Thus, they should be closely correlated to the variations with depth profile. Since the chosen depth profile

-f

1.6E - 2 7.OE + 0 2.6E + 0 8.OE - 2 l.lE+O 6.2E + 0 2.8E - 2

4.OE - 2 9.3E + 0 3.4E + 0 9.8E - 2 1.4E+O 7.9E + 0 3.7E - 2

-f -I

-f 3.4E - 4

-f

1.1E + 1

1.4E + 1

BP

1.6E + 1 2.OE - 5 3.6E - 2 5.4E - 4 1.8E - 1 1.OE + 1 3.8E+O 1.4E - 1 1.7E+O 9.8E + 0 4.4E - 2

=

Co-60 Sr-90 Y-90 Tc-99 1-129 (28-134 (25-137 Ce-144 Pb-214 Bi-214 Ra-226

RL 0.1 cm'

Planeb

Nuclide

l.lE+ 1 9.OE - 4 2.1E - 2 2.OE - 4 4.1E-2 6.8E + 0 2.5E + 0 7.9E - 2 l.lE+O 6.3E + 0 2.9E - 2

I cm slabb =

7.6E + 0 5.1E-4 1.4E-2 1.OE -4 1.5E -2 4.8E + 0 1.8E+ 0 4.93 - 2 7.5E- 1 4.6E + 0 1.9E-2

3 cme

RL

Actual Depth Distribution

distribution for selected nuclides."

6.2E+ 0 4.1E-4 1.2E -2 7.9E - 5 9.6E - 3 3.9E + 0 1.4E+O 4.OE - 2 6.2E - 1 3.7E + 0 1.6E- 2

5 cm slabb

l.OE+l 5.1E -4 1.7E-2 9.33-5 9.6E - 3 6.2E+O 2.3E + 0 5.2E - 2 9.3E - 1 6.OE + 0 2.3E - 2

15 cm slab'

+

1.2E+1 5.2E-4 1.8E-2 9.3E-5 9.6E-3 7.OE+O 2.4E + 0 5.3E - 2 9.9E - 1 7.3E 0 2.4E - 2

;labb

inf.

TABLE3.7-Variation of Eper unit nuclide concentration in top 5 crn [nSv d-' (Bq kg-')-'] with actual depth

UI

2

m

$

?j

M

8

8

3

Y

2: m

a

aAssumes an average concentration of 1Bq kg-' is measured in top 5 cm (see Section 6). Thus, if the actual depth profile is an infinite plane on the surface, E for 60Co will be 16 nSv d-l as opposed to 6.2 nSv d-I if the source is uniformly distributed to 5 cm and 12 nSv d-I if the source is uniformly distributed to infinite depth. bFrom Eckerman and Ryman (1993). Since the values in Eckerman and Ryman are given in terms of Bq m-3 rather than Bq kg-', they were converted assuming a n in situ bulk density of 1.6. The infinite plane source conversion is not dependent on the choice of density. However, the 1cm slab conversion is, a s discussed in Section 3.2.3. cFrom Beck (1980) using Sv Gy-' from Table 3.2. RL is the relaxation length, i.e., the depth where the concentration has decreased by lle of the surface concentration assuming an exponential decrease with depth. dFrom Jacob et al. (1990) for a buried plane source a t 0.3 cm below the surface. "om superposition of slab sources from Eckerman and Ryman (1993) (see Footnote 5,Section 3.2.2) 'No data.

5 C

8

5

8Z

m

50

1

3. DOSE FROM EXTERNAL EXPOSURE

is already conservative and the uncertainty in the calculation itself is expected to be minor compared to variations due to soil composition, it was assumed that the uncertainty in the dose factor could be conservatively accounted for by assuming that one-third of the calculated difference between E for the 1cm slab, the infinite slab, and the 3 cm relaxation length is a reasonable estimate of the standard deviation (S.D.) of the total distribution of possible dose factors. Using this assumption, the estimated variability in the dose factor ranged from about 10 percent (1S.D.) for medium and high-energy gamma emitters to around 30 percent (1 S.D.) for low-energy beta and x-ray emitters (see Section 7). Although this uncertainty estimate is somewhat ad hoc, a sensitivity analysis (see Section 7) indicates that the uncertainty in the Df,, is not a major contributor to the overall uncertainty in the external dose. The Df,, values used for screening, along with their estimated uncertainty, are listed in Appendix C for all nuclides considered in this Report.

3.2.3 Dependence on Soil Moisture and Bulk Density

The in situ or bulk density of the surface soil will determine the actual source concentration in any particular depth increment (see Table 3.5). Calculational models usually assume either a uniform bulk soil density of 1.5 to 1.6 g cm-3 or a fixed inventory of one gamma ray emitted per cubic centimeter of soil surface, integrated to some depth. Beck and de Planque (1968) and Chen (1991)discuss how to correct for differences in density for certain types of depth profiles. The calculated doses for the distributed sources in Table 3.7 are based on a concentration of 1Bq kg-' over the top 5 cm and the inventories from Table 3.5 that assume a density of 1.6. A lower density would imply a lesser mass (kg) of soil in each centimeter depth increment and thus a lower source concentration per cubic centimeter, than assumed for the dose calculations. However, the reduced attenuation due to the lower density would result in the reduction of dose contributions from the surface layers being offset by additional contributions from deeper layers. Thus, the calculated dose for the uniform distribution would not be affected. As long as the soil concentration is measured per unit soil mass, the effect on the calculated dose factors due to typical variations in bulk soil density will generally not be significant. Often, however, the density of the very top few centimeters of soil is lower than 1.6, usually due to a greater concentration of organic matter. If the density over the top few centimeters is <1.6, but the average density over 5 cm is still 1.6, the dose calculation for the 1cm slab and buried plane

3.2 DISCUSSION OF MODEL PARAMETERS

1

51

source calculations may be slightly too low for low-energy emitters due to the reduced attenuation. However, the organic matter is also generally deficient in radionuclide concentration relative to the contaminated soil. Therefore, minor differences in density from site to site should not impact on the screening guidance in this Report. The measured concentration is usually given per kilogram of dry soil. Actual soils might have annual average water contents ranging from a few percent in very arid climates to as much as 40 percent in areas of high rainfall. Thus, it is necessary to correct the estimated dose rate for the true in situ specific activity based on average water content. For example, not correcting for a uniform 30 percent by weight moisture content in situ results in a 30 percent overestimate in surface soil concentration in each depth increment and therefore about a 30 percent overestimate in the dose rate. This correction is approximate since the additional water also changes the atomic composition and thus the radiation attenuation properties of the soil, particularly for low-energy gamma-ray emitters and beta rays (see Section 3.2.2.5). For the screening dose calculations reported here, reasonable yet conservative values were chosen for average soil moisture (Table 3.8). The variability in the chosen values was conservatively represented by a S.D. of + 0.05 for all land-use scenarios. This S.D. was chosen so that 3 S.D. encompasses the observed lower range of typical average annual soil moisture content. It allows, with high probability, for completely dry soils for the sparsely vegetated landuse scenarios. Shielding by Dwellings Assessments of the external dose pathway ofken fail to adequately address the importance of shieldingby dwellings [NCRP Commentary TABLE 3.8-Soil moisture corrections (W,) used for screening dose calculations. 3.2.4

Land Use

Drylln Situ"

S.D.

Agriculture Heavily vegetated pasture Sparsely vegetated pasture Heavily vegetated rural Sparsely vegetated rural Suburban Construction, etc. "Ratio of dry soil weight to in situ soil weight, i.e., a ratio of 0.85 is equivalent to a soil moisture content of 15 percent by weight.

52

1

3. DOSE FROM EXTERNAL EXPOSURE

No. 8 (NCRP, 1993a)l.The resultant dose estimates are then usually overly conservative for United States' populations since most people spend most of their time indoors. The shielding offered by dwellings varies widely depending on the type of construction, the height above ground, and other factors. Studies from Chernobyl and from areas affected by weapons fallout generally indicate that even for lightly constructed housing, the exposure rate from the high-energy gamma emitters is reduced to about 0.4 of the outside free-in-air value. For more massive buildings, such as apartment houses, the indoor level can be reduced to < 1 percent of the outdoor level. Table 3.9 gives some estimates of typical shielding factor, i.e., ratio of dose indoors to dose outdoors, reported in the literature. The values listed are based either on actual measurements or on transport calculations. Clearly the true shielding factors for any real exposure scenario is highly site-dependent. For the purposes of calculating screening doses, values were chosen as discussed below for each of the landuse scenarios where dwellings are assumed to exist. The critical population in each case is assumed to be persons living in the most lightly constructed housing. Four different values were chosen and used for different energylparticle sources. The choice of these four values was governed by the relative penetrating capability of a given nuclide's emissions. This penetrating capability was inferred from the ratio of the calculated Df, for a uniform distribution and a 1cm slab for that nuclide with the corresponding Df& for an exponentially distributed source of 3 cm relaxation length (see Section 3.2.2.8). As mentioned previously, the maximum difference between the uniform depth distribution, the 1cm slab and the 3 cm relaxation length for most high-energy penetrating gamma sources is <40 percent. For lesser penetrating radiations, the difference increased to over 80 percent. The higher the hfference, the greater the attenuation of that nuclide's radiation in the soil, air and body. Thus, the radiation from those nuclides would be expected to also be the most attenuated in passing through the walls of dwellings. For those nuclides (gamma emitters of energy >I00 keV) with dose factors exhibiting the smallest differenceswith assumed depth profile, i.e., <40 percent, a shielding factor of 0.4 was assumed. For nuclides where the maximum difference in depth profile dose factors was between 40 to 50 percent, correspondingto low-energy gamma (<100 keV) or high-energy beta emitters (average energy >I00 keV), SF = 0.3 was used, between 50 to 60 percent, corresponding to pure beta (average energy
NCRP (1993a)

Literature review Literature review Literature review

0.02-0.6 (0.33 avg.) 0.02-0.4 most likely 0.2 0.2; 0.001-0.5

Residential

Average

Mean; range

Ttatio of indoor dose to outdoor dose.

UNSCEAR (1988)

71 measurements 72 measurements

Brick houses Wood houses

Kennedy and Strenge (1993)

Likhtariov et al. (1996)

Burson and Profio (1977)

Measurement (fallout)

Profio (1977) Profio (1977) Profio (1977) Profio (1977)

Burson and Burson and Burson and Burson and

Steinhausler (1987)

Calculation (cloud) Measurement (fallout) Measurement (fallout) Measurement (fallout) Measurement (fallout)

Jacob and Meckbach (1987)

Measurements

Semi-detached Prefab housing Single family, two story house

Light construction Heavy construction Cars Wood frame, one story Large buildings

Catsaros and Vassiliou (1987)

Calculations

Cars, buses

Reference

Le Grand et al. (1987; 1990)

Data Source

Calculations

Shielding Factor"

Average Worst case

Type

TABLE 3.9 -Typical shielding factor reported in the literature.

0

01

1

fi

W

t'

z

5

g

g

2

54

1

3. DOSE FROM EXTERNAL EXPOSURE

was also assigned to each chosen value. This uncertainty estimate is conservative in that it allows the highest shieldng factor (i.e.,the minimum shielding) to range from about 0.2 to 0.6 and the lowest (i.e.,maximum shielding) to range from 0.004 to 0.16, assuming the range is equivalent to about 3 S.D. Table 3.10 gives the resulting shielding factor and associated S.D. for selected radionuclides. The same screening values were used for each land-use scenario where dwellings are assumed to be on the contaminated land. For screening, a maximum shielding correction corresponding to S F = 0.1 is assumed to be sufficiently conservative to also account for exposure indoors due to soil accidentally brought inside on shoes and clothing (Cannell et al., 1987), particularly since contamination tracked in is unlikely to remain for long periods. The shielding factor chosen for screening purposes are consistent with the attempt to choose prudently conservative parameter values appropriate to the most exposed population group. The values chosen may be too conservative for site-specific assessments, however, based on the data in Table 3.9. The shielding factor used for calculating screening doses for each of over 200 nuclides covered in this Report are listed in Appendix C.

3.2.5 Indoor and Outdoor Exposure Times

Clearly, one of the more important parameters affecting the potential external dose received from exposure to contaminated soil is the actual time exposed, either outdoors or indoors on the contaminated site. UNSCEAR (1993) estimates that on average people spend 80 percent of their time indoors either a t home or work. Brown (1985) estimates that indoor occupancy is 90 percent on average for the

3.10-Shielding factor used for screening dose calculations TABLE for selected radionuclides." Nuclide

SF

"SeeAppendix C for complete listing.

S.D.

3.2 DISCUSSION OF MODEL PARAMETERS

1

55

entire United Kingdom population. He estimates 75 percent of the time is spent indoors a t home and that females spend somewhat more time indoors a t home than males with little seasonal variation (about 80 percent versus 66 percent). Kennedy and Strenge (1993) estimate that for average residential scenarios, individuals spend about 55 percent of the time in the house and 19 percent outdoors. Outdoor workers average about 65 h week-' outdoors while indoor workers spend only about 15 h week-' outdoors. Women caring for children probably spend almost 150 h week-' indoors and only about 18 outside. Children may average 35 h week-] outdoors and about 100 h indoors (Kennedy and Strenge, 19931, particularly in temperate climates. Table 3.11 gives the indoor and outdoor exposure times used in this Report for each of the land-use scenarios considered. The estimates used here are conservatively based on the data from the literature. They are intended to reflect the situation for the most critically exposed populations for each land-use scenario, those who spend a higher than average proportion of their time outdoors. This would generally be the outdoor worker or, for the residential scenarios, children playing outdoors. Thus, it was conservatively assumed that for sites with dwellings, the critical population averages at least 90 percent of their time on the site and 30 to 40 percent outdoors, as shown in Table 3.11. The range assigned, however, allows for the possibility of some individuals spending up to 70 percent of their time outdoors on the contaminated land. Site-specific assessments may need to gather detailed demographic data on the time various individuals might now, or in the future, spend on the contaminated

TABLE3.11-Indoor and outdoor exposure times" used for screening dose calculations. (Percent of time) Land Use

Ti.

Agriculture Heavily vegetated pasture Sparsely vegetated pasture Heavily vegetated rural Sparsely vegetated rural Suburban Construction, etc.

0 0 0 50 50 50 0

the contaminated site.

Range (in)

20-90 20-90 20-90

To,,, Range (out) 40 30 30 40 40 40 30

0-70 0-60 0-60 0-70 0-70 0-70 0- 60

56

3. DOSE FROM EXlERNAL EXPOSURE

land, both inside and outdoors. Some screening calculations make the very conservative assumption that the critical population group consists of individuals who stand outdoors in the center of a contaminated area for the entire exposure period. In reality, it is unlikely that any individual will spend all of his or her time outdoors on the site. For sites without dwellings, e.g., agricultural or construction sites, the time exposed will be generally limited to working hours. For other sites, for example in tropical or subtropical areas, some individuals, e.g., foresters, children a t play, etc., may spend much of their waking hours outdoors. However, in more northern areas, the average annual exposure for outdoor workers or children would be much more limited.

3.2.6 Age, Sex Correction

Most of the published dose factor calculations are for adults (Reference Man). However, Jacob et al. (1990) and Petoussi et al. (1991) also provide calculations for children and babies and compare values for males versus females. From their results, it appears that E to children ranges from 10 to 30 percent more than that for typical adults, with a larger ratio for low-energy gamma-, beta- and x-ray emitters. Doses to babies lying on a contaminated soil surface range up to 50 percent higher than the corresponding dose to adults. However, the length of time exposed in this manner is likely to be small for most western lifestyles. Since the child to adult (C/A) ratio of external dose factors varies little from nuclide to nuclide, the adult (Reference Man) Df, values used in this Report were multiplied by a constant factor of 1.3 -+ 0.1 for screening calculations for children for the land-use scenarios (RV, RS, SU, SN) where children are potentially the most exposed population.

3.3 Summary of Parameter Values for Screening Calculations

Land-Use Independent Dfext Shielding factor Soil bulk density CIA correction Soil concentration, S

see Table 3.7, Appendix C see Table 3.10, Appendix C 1.6 +. 0.1 (1S.D.) g ~ r n - ~ 1.3 + 0.1 (1S.D.) 1.0 t 0.5 (1S.D.)Bq kg-'

3.4

CALCULATED SCREENING DOSES

Land-Use Dependent Soil moisture correction, W, Indoorloutdoor exposure times

3.4

1

57

see Table 3.8 see Table 3.11

Calculated Screening Doses

Using Equation 3.1 and the parameter values described in the preceding paragraphs, a screening dose was calculated for each nuclide for each of eight land-use scenarios. Table 3.12 shows the calculated median and range (5 and 95 percentile) of doses to a member of the most exposed population group for some selected nuclides often encountered in environmental contamination scenarios. These were calculated stochastically using the estimated uncertainties for each parameter (see Section 7). An additional possible pathway for direct radiation exposure not directly treated in this Report is from radon decay products in the atmosphere as a result of radon exhaled from soil contaminated with 226Ra.Generally, the dose from this pathway is small compared to the dose from the gamma rays emitted from the decay products remaining in the soil (Beck, 1972; 1974). The increase in the dose due to airborne radionuclides is also offset by a generally greater decrease because of the loss of gamma emi1;ters from the surface soil (Beck et al., 1972). This dose contribution was thus not included in calculating the external exposure screening limits for members of decay series. the 226Ra The differences in the annual E versus land-use scenario illustrated in Table 3.12 reflect mainly the differences in the estimates of the fraction of time spent outdoors versus indoors or away from the site. The largest contributor to the range is the uncertainty in soil radionuclide concentration (see Section 7). The complete set of calculated external E for all radionuclides considered in this Report is given in Appendix A.

PV

9.63- 7 (3.0-24) 1.8E - 9 (0.6-4.6) 1.4E - 7 (0.5-3.6) 5.9E - 8 (2.0-15) 5.OE - 7 (1.7-13) 2.1E - 7 (0.7-5.3) 1.5E - 8 (0.5-3.8) 6.9E - 7 (2.3-18) 6.53 - 9 (2.1- 17)

AG

1.2E-6 (0.5-3.0) 2.4E - 9 (0.9-5.7) 1.8E - 7 (0.7-4.3) 7.4E - 8 (2.9-18) 6.5E - 7 (2.5-16) 2.6E - 7 (1.0-6.4) 1.9E- 8 (0.7-4.6) 8.73 - 7 (3.4-21) 8.4E - 9 (3.1-21)

1.OE - 6 (0.3-2.5) 1.9E -9 (0.6-4.9) 1.5E - 7 (0.5-3.7) 6.2E - 8 (2.1-16) 5.2E - 7 (1.7-14) 2.2E - 7 (0.7-5.5) 1.5E - 8 (0.5-3.9) 7.2E - 7 (2.4-18) 6.73 - 9 (2.2-18)

PS

2.4E - 6 (1.0-5.6) 4.73 -9 (2.0-11) 3.53 - 7 (1.6-8.0) 1.5E - 7 (0.6-3.4) 1.3E - 6 (0.6-2.9) 5.2E - 7 (2.2-12) 3.7E - 8 (1.6-8.7) 1.7E - 6 (0.7-4.0) 1.4E - 8 (0.5-3.4)

RV

2.7E - 6 (1.2-6.2) 5.1E - 9 (2.2-12) 3.9E - 7 (1.7-8.8) 1.5E - 7 (0.6-3.4) 1.4E - 6 (0.6-3.2) 5.8E - 7 (2.5-13) 4.1E - 8 (1.8-9.6) 1.9E - 6 (0.8-4.4) 1.5E - 8 (0.6-3.6)

RS

2.4E - 6 (1.0-5.6) 4.73 -9 (2.0-11) 3.53 - 7 (1.6-8.0) 1.5E - 7 (0.6-3.4) 1.3E - 6 (0.6-2.9) 5.2E - 7 (2.2-12) 3.73 - 8 (1.6-8.7) 1.7E - 6 (0.7-4.0) 1.4E - 8 (0.5-3.4)

SU/SN

1.OE-6 (0.3-2.5) 1.9E-9 (0.6-4.9) 1.5E-7 (0.5-3.7) 6.2E - 8 (2.1-16) 5.2E - 7 (1.7-14) 2.2E-7 (0.7-5.5) 1.5E - 8 (0.5-3.9) 7.3E - 7 (2.4-19) 6.73 - 9 (2.2-18)

CC

"Median dose to representative member of critically exposed population. 5 to 95 percent quantiles given in parentheses. The dose includes the contributions from progeny as well as from the parent and is the maximum annual E over the next 1,000 y. See Section 3.2.2.8 for a discussion of the assumed radionuclide depth profile. See Appendix A for complete listing of external doses.

Nuclide

Land-Use Scenario

TABLE 3.12-Annual E [Sv (Bq kg-')-'] for selected radionuclides-external e ~ p o s u r e . ~

$

8

W

$

8m

2g

M

u $

w

\

03

vl

4. Dose from Inhaled

Radionuclides Another potential pathway for human doses from contaminated sites is inhalation of airborne radionuclides on suspended contaminated soil. Unlike external exposure, which contributes only to the annual dose in the year exposed, inhaled radionuclides can result in a dose in subsequent years as well, depending on how long the radionuclide remains in the body. Thus, for radiation protection purposes, one generally calculates E(z)over a 50 y interval (or until age 70) resulting from each year's annual intake. The potential exposure from this pathway depends on a number of factors including the average activity of the airborne resuspended soil, the length of time exposed either inside or outdoors, the particle size distribution of the suspended soil, the nuclide and its chemical form, and the age and breathing rate of the person exposed. The resuspension pathway is generally not a significant contributor to long-term exposure for most sites andlor nuclides. However, it is potentially important for dusty sites where the nuclide remains in the surface soil and for nuclides which are insoluble and thus are likely to remain in the lung for long periods of time after inhalation. For alpha-emitting nuclides such as plutonium, which do not emit any high-energy penetrating radiation and are not readily absorbed from the GI tract, the external radiation and ingestion pathway doses are usually minimal. In these cases, inhalation is often the most significant potential dose pathway. Because of the complexity of the resuspension process itself, the estimation of the airborne nuclide concentration, as well as the resulting dose, is discussed in some detail in this Section.

4.1 Introduction to the Resuspension-MigrationPathway

The general goal of this introduction will be to review the atmospheric resuspension process. Knowledge of the resuspension process has its roots in two major areas. The first is the physics of blowing sand and soil erosion. The second is related to the resuspension of plutonium following dispersal from an accident involving nuclear weapons. It is commonly accepted that after deposition on the

60

1

4. DOSE FROM INHALED RADIONUCLIDES

ground, the primary hazard of the dispersed plutonium is resuspension and subsequent inhalation of the plutonium. However, in this Report, ingestion of contaminated soil is also discussed (see Section 5.3). Resuspension is a general term that is used to indicate the process whereby material is transferred from the soil surface to the atmosphere. Typically, the radioactive material, or some portion of it, is returned back to the atmosphere, or resuspended, if it originally was deposited via an airborne pathway. However, the concept is equally applicable to situations where the material arrives at the storage soil by entirely different means, such as being contained in a liquid effluent. In this case, resuspension implies that the radioactive material is being transferred to the atmosphere for the first time. It is obvious that the physical state of the radionuclide is of great importance in determining whether it is likely to be a hazard via the resuspension pathway. A single solid source of large physical size will not generally resuspend. As will be discussed later in this Section, suspended radionuclides can generally be assumed to be described by a lognormal particle size distribution with a median diameter of from 2 to 6 pm and a geometric S.D. of about five. Furthermore, in general, radionuclides associated only with particles of the size of a few millimeters or greater will not constitute an inhalation hazard. However, the weathering process that causes particles to break or material to dissolve from them can be rapid in effecting a change in the particle-sizedistribution of the contaminant.

4.1.1

General Classes of Resuspension

Resuspension can be divided into two general classes, wind-driven open environment and local direct.

4.1.1.1 Wind-Driven Open-Environment Resuspension. A major source of large-scale resuspension of soil particles results from interaction of wind with the soil surface. This natural form of resuspension is generated from any very large area source. It is assumed that the suspended material a t any point represents that created from large surface areas located upwind. Thus, when radionuclides are deposited onto the soil surface and combine with soil particles, the winddriven resuspension process can provide an airborne source of radionuclides which can possibly be inhaled and deposited in the lungs. The wind can suspend smaller size particles directly into the air and cause larger particles to jump into the air where they are borne by the wind until they land again on the soil surface (saltation).

4.1 RESUSPENSION-MIGRATIONPATHWAY

1

61

Saltation causes collision and fragmentation of the surface particles which, in turn, are propelled into the air by the energy of the saltating particles. Thus, resuspension in the open environment depends on meteorological conditions, soil type, soil surface characteristics, weathering processes, and contaminant characteristics. The natural meteorological processes can create an open environment background of airborne particles that range from a few micrograms per cubic meter of air to situations where visibility is limited. An important factor to be considered in this range of possible resuspension situations is the fraction of the total resuspended material that is in the respirable size range. The natural large-scale, wind-driven, open-environment resuspension can also be enhanced by events that alter the soil surface to an extent that soil particles are suspended and carried by the wind over relatively large areas or regions. This immediate suspension generally lasts only as long as the event causing it; it is usually produced in a relatively small area, although particles may be transported over a much larger area. Enhanced resuspension can be created, for example, by the plowing of fields, vehicular traffic on fields or dirt roads, movement of large herds of animals, scraping of soil surfaces, and use of earth moving equipment. Smaller sources of suspended material can be generated by individuals or animals. These disturbances not only provide a source of suspended material for large-scale dispersal but create a localized suspension that may be inhaled by the persons creating the source. This localized resuspension will be discussed below. Processes that involve soil moving, plowing or scraping of surface soil not only create a n immediate suspension for the duration of the disturbance but can leave the soil surface more vulnerable to the natural wind-driven resuspension processes. This results in creating a higher background of resuspended material over the long term than would have been present otherwise. The enhanced suspension due to mechanical disturbance of the soil surface often results in a different class of resuspended material being present due to the subsurface soils and broken aggregates suddenly being available for resuspension. This soil upheaval can, for example, result in a smaller average particle size being suspended from the particle size inventory available in the soil. This can also lead to a change in the radionuclide concentration per unit mass of suspended material. The depth from which subsurface soil may be brought to the surface, mixed with surface soil, etc., is of course dependent on the type of disturbance. 4.1.1.2 Localized and Direct Contamination. Human activity can create localized resuspension. Actions by individuals that create local

62

1

4.

DOSE FROM INHALED RADIONUCLIDES

sources of respirable resuspended material in the immediate vicinity of the individual range from relatively large suspension sources such a s those created by a person plowing a field or operating earthmoving equipment to those created by adults and children in recreational activities, persons digging in t h e backyard, clearing weeds and brush, or sweeping the back porch or a room of a house. All such actions create a source of respirable material that is unique to the localized source and its creator (or people in the immediate vicinity of the activity) and that is supplemental to the large-scale, winddriven, natural form of resuspension. Localized sources are very important in the overall evaluations of the potential exposure to radionuclides due to inhalation because they often create a mass loading (microgram of dust per cubic meter of air) that is significantly higher than that from natural resuspension and can generate a smaller particle size distribution that is more readily deposited in the lung. Because of the diverse nature of localized sources of suspension, they are much harder to define and evaluate than the natural, large-scale, wind-driven resuspension.

4.1.2

Parameters of Contaminants that AfjCect Open-Air Resuspension

Particle Size. One of the most significant parameters affecting potential exposure to airborne, soil-associated radionuclides is the size of particles to whch the radionuclides are attached. In general, the smaller the particle size the greater the deposition in the thoracic region of the respiratory tract (ICRP, 1994a). For example, the fractional deposition of the activity in a given volume of inspired air in the thoracic region (bronchii, bronchioles and alveolar interstitiurn) for a n activity median aerodynamic diameter (AMAD) of 0.1 ym is about 38 percent. It is about seven percent for the extrathoracic region. For particles of 50 Fm AMAD, the fractional deposition in the thoracic region drops to about 0.3 percent while that in the extrathoracic region reaches about 56 percent. At 100 Fm AMAD the extrathoracic region remains a t about 56 percent while the thoracic region drops to about 0.05 percent. Consequently, the exposure and dose will vary for the different regions of the respiratory tract as a function of the particle size inhaled and the mechanisms and rates of clearance from each region. These factors, in conjunction with the relative sensitivity of the various tissues, will determine the detriment. In recent years, there has been more interest in evaluating the role of larger particles in the overall exposure and dose from resuspended particles because

4.1.2.1

4.1 RESUSPENSION-MIGRATIONPATHWAY

1

63

they can provide localized sources of radiation exposure not only in the extrathoracic region, but also in the bronchii region. 4.1.2.2 Physical-Chemical Form Affecting Availability.

The physical and chemical form of the radionuclide generally determines the ultimate size of the particle or aggregate with which the radionuclides become associated. For example, oxides of transuranic radionuclides have been dispersed into the environment that include sizes too large to be a problem for resuspension or inhalation. Thus, the most likely pathways of bodily contamination for these particles might be ingestion of soil or direct entry into the body via a n open wound. These larger size particles or pieces of PuOz can, however, degrade with time to smaller particle sizes. I n other cases, transuranics have been released in ionized or reactive form, or as very small particles, such that they can be incorporated into or onto small soil particles or organic complexes immediately. Some chemical forms of 239h are very insoluble and relatively unavailable to degradation for years. Other chemical forms of 2 3 9 P ~ , and other radionuclides, might be readily available in, or degrade rapidly into, forms which can be incorporated onto particles in the respirable size range. Knowledge of the chemical form may help in decisions about judicious treatment and management of contaminates. As discussed later, it may also allow one to make a more accurate estimate of the inhalation dose. Although the original physical-chemical form of the radionuclide is of interest in trying to estimate its long-term stability, it is not of much use from a theoretical point of view for predicting the availability to resuspension. To accurately determine the amount of the radionuclide that may be resuspended, and thus inhaled, requires direct measurement of the airborne material under the associated ecological and meteorological conditions.

Availability with Time. The availability of radionuclides to the resuspension process can change as a function of time after deposition. This change in availability can occur due to both chemical and physical processes that make the radionuclide less available for resuspension. One of the major processes which reduces the availability of surface deposited radionuclides for resuspension is the "weathering" or subsequent movement of the radionuclides deeper into the soil. The solubility of the radionuclide complex is important to this process because as the solubility increases, the movement down the soil column will be further enhanced by solubilization in rainwater as the precipitation percolates through the soil. This process may 4.1.2.3

64

1

4 . DOSE FROM INHALED RADIONUCLIDES

eventually transfer a portion of the initial deposition all the way to the groundwater except for very deep aquifers. Other physical processes such as soil cracking, frost heaving, root growth and decay, and soil fauna movement can also cause weathering. Some portion of the radionuclide may aIso be bound to the soil matrix and be relatively unavailable for vegetation recycling back to the soil surface as is the case for cesium in clay soils. Thus, these processes, with time, tend to stabilize some fraction of the initial deposition in such a way that it is essentially unavailable for resuspension.

4.2 Resuspension and Dose Models

Dose Model

4.2.1

E(z)resulting from the inhalation of a radioactive material during a given year can be calculated from the following expression: Elnh = Dfinh

X Cair X [Rout X Tout

+ (110)X

Rin X

Tin1 (4.2)

where: Einh

DLh C, R,,,

= committed effective dose for inhalation (Sv y-') = inhalation dose factor (Sv Bq-') = average annual outdoor air concentration (Bq m-9

= average breathing rate outdoors (ms d-')

&,I

= average

I10

= ratio

breathing rate indoors (m3d-') of nuclide concentration in air indoors versus outdoors To,, = days per year spent outdoors on contaminated land = days per year spent indoors on contaminated land Ti,, The calculation of the median radionuclide concentrations in air per unit soil concentration resulting from resuspension is described in some detail in the following section. The other parameters are described in detail in later paragraphs. 4.2.2 Resuspension Models

The purpose of this Section is to present a rationale for the estimation of mean annual concentrations of airborne radionuclides from resuspended soil particles, C ~ ,and , to show what experimental evidence exists for the validation of the methods and their limitations and uncertainties. A discussion of airborne particle size is also

4.2 RESUSPENSION AND DOSE MODELS

1

65

included, so that estimations of inhalation exposure can be made based on the expected lung deposition. 4.2.2.1 Method 1, Estimation of C, by Modifid Mass Loading. The primary objective is to estimate the concentration of airborne radionuclide, Cai,(Bq m-Y. Three methods are generally available. Each will be discussed below. The modified mass loading method is based on the relationship:

where: EI

S M

the enhancement factor, defined as the ratio of airborne particle concentration (Bq kg-') to total surface soil concentration (Bq kg-') = the total surface soil concentration (Bq kg-' dry) = the total suspended particulate concentration (kg m - 9 also known as the "mass loading" =

It would be preferable that the three factors in Equation 4.3 be measured. At many sites with contaminated soil, parameters such a s M and S a r e monitored routinely. However, i t is implicitly assumed that the suspended particles have their origin in the soil where S is measured. There are many cases where this does not occur and good judgment must be used to assure that the airborne contaminants are truly representative of the contaminated soil. The "footprint" of a highvolume air sample is the area defined by the distance to the upwind fetch locations (x) enclosing the region of vertical soil flux which contributes to particles having trajectories reaching the air sampler a t height z. It is a requirement for the estimate of soil contamination, S, to be averaged over this region, in order for the air sampler to be representative of a single value of S. The upwind fetch, x, can be calculated for any desired percentile collection, and any combination of air sampler height and roughness condition, using the method of Shuepp et al. (1990). In this method, a numerical solution is found for the relative contribution to the vertical flux of particles originating from an infinite crosswind source of unit width as a function of upwind distance. The effect of wind speed is expressed as a drag coefficient that depends on "roughness length." Table 4.1 indicates that about 53 to 64 m would be a common range of distances that a typical air sampler would represent for 75 percent of the particles collected. Note that the effect of wind "drag" on the surface is dependent on the type of vegetative cover.

66

1

4. DOSE FROM INHALED RADIONUCLIDES

TABLE4.1 -Fetch distance, x, calculated for 75 and 90 percent of upwind contribution with the air sampler intake at the 1.13 m height. ff(mP

Surface Type

xdm)

xdm)

Bare soil Short grass Tall grass "H i s height of vegetation.

Estimates of the factors Ef and M in Equation 4.3 can be made from observations made at other contaminated sites, but S must be measured locally. The following general rule may be applied to estimate Ef where there are no measurements of the enhancement factor: Er Er Er

= = =

0.7 in the case of undisturbed surface soil 4 in the case where soil is recently disturbed6 0.01 in the immediate vicinity of a surface nuclear blast site

These values of the enhancement factor were taken from high volume and cascade impactor air sample observations of radionuclide resuspension under different environmental conditions by Shinn (1992). The undisturbed soil value is a median value from observations a t sites a t Bikini Island, Nevada, South Carolina, and California. The range of values was 0.21 to 1.04. The value for recently disturbed soil is the median of observations during activities such as soil thawing, bulldozer blading, vacuum cleanup, wildfire, and raked-off surfaces, that had a range from 2.2 to 6.5. Values of the enhancement factor calculated from radionuclides in suspended dust during the agricultural tractor operations near Chernobyl gave a median of 4.4 for medium to highly contaminated soil, and had a range of values from 2.8 to 8.4 (Loshchilov et al., 1992). In the case of a surface nuclear blast site, where some of the radioactive material is held in a substance known as "shot glass," values reported by Shinn (1992) had a range of 0.002 to 0.024. Such sites are rare but could be managed differently than others. When the total suspended particulate concentration, M, cannot be measured, it may be estimated from the observations by Shah et al. 'Fhxently disturbed refers to a period beginning from redeposition to a few days later.

4.2 RESUSPENSION AND DOSE MODELS

/

67

(1986) who found annual averages of 28 and 75 ~g m-3 for rural and urban locations, respectively, for undisturbed conditions. During cleanup or disturbed soil conditions, however, one would expect that M values could increase dramatically. Loshchilov et al. (1992) found that M values behind a tractor had a median of 15 mg m-3 and a range of 0.3 to 200 mg m-3. (Nuisance dust is considered anything over 0.15 mg m-3). There is also a seasonal effect on values of M. For example, Shinn (1992) shows a change from 7 to 30 kg m-3 from spring to summer in rural Nevada. As discussed earlier, the particle-size distribution is needed for estimating inhalation exposure from the estimated value of C&. For site-specific assessments, measurements of particle size should be made if possible. Cascade impactors are useful for direct measurement of the activity size distribution. While the effects of different particle-size distributions on the measured or estimated exposure concentration, C*, are of concern, there is evidence that the effects are predictable. There are usually three modes of size-component distributions, each approximately log-normal mass distributions with particle diameter. Milford and Davidson (1985) studied size distributions of 11crustal elements from suspended soil and found a median diameter of about 5 pm. This is supported by observations from Shinn (1992) a t many sites that had median suspended-soil aerodynamic diameters of 2 to 6 pm. A recent review by Dorrian (1997) concludes that the median diameter for resuspended material is about 6 pm. A coarse component, median diameter of about 15 pm, is sometimes also found when the soil is disturbed or when very strong winds are present. This coarse component should be considered transient, however, because the gravitational settling velocities of the coarse particles are greater than the suspension velocities and their residence times in the atmosphere are short. PMlois the fine component of the particle-size distribution (particles below 10 pm AMAD) represented by the total suspended particulate obtained from M. A study by Rodes et al. (1985) found that the ratio of M/PMIois typically about 1.7 with a range of 1.3 to 2.2. Under disturbed conditions, this ratio will increase. In urban areas, Milford and Davidson (1985), as well as many others, found a very fine size component with a median diameter of about 0.5 pm. This component is almost universally not of crustal origin, but is derived from combustion sources. Typically, the very fine component is in a state of transformation to larger particles by the process of coagulation. The very fine-size component contributes very little to M, and in most cases would not carry locally derived radionuclides. In the absence of measurements, it is recommended that one should expect all the suspended radionuclide is on soil particles having one mode of

68

1

4. DOSE FROM INHALED RADIONUCLIDES

particle size with a median diameter in the range of 2 to 6 km and lognormally distributed with a GSD of about five. In the special case of disturbed soil, when complete measurements of size-distributions cannot be made, PM,, measurements are preferred. Another option is to estimate the respirable size range: PMIo = M I 1.7 (disturbed case only)

(4.4)

The coarse component contributes to the total suspended particulate estimate of M (but not PM,o);however, it is not as important as an estimator for deposition in the respiratory tract. As discussed above, particles of soil origin are lognormally dispersed (GSD > 4) but as discussed earlier, the larger median-diameter distribution affects mostly the upper respiratory deposition. ICRP Publications 66 and 71 (ICRP, 1994a; 1995a) should be consulted for calculating sitespecific dose endpoints from estimated particle-size distributions. Method 1has been shown to hold for many sites of aged deposits and provides more site-specific information than the alternative Method 2 described below. Before considering the alternative methods of estimation of C,,, it is important to emphasize the requirement for representative measurement of S . The soil radionuclide concentration will have natural variability that cannot be characterized by a few samples. This natural variability is due to the complex nature of the original deposition, surface effects, and variability of the weathering process after initial deposition. Experience has shown that within a larger region of relatively homogeneous deposition there will be spatial variability depending on the scale of the soil-surface sampled. That is, the area of the soil-surface that is sampled is the critical dimension. Since the air sampled for estimates of C& contains particles whose trajectories came from many upwind locations, it is best to use a large area (tens of meters in radius) to determine an average of S based on many soil samples or from a n in situ gamma spectroscopic measurement that represents a large view of the soil surface. More detail on soil variability and methods for determinings are presented in Section 6. 4.2.2.2 Method 2, Estimation of C,,, by Resuspension Factor. The second method of estimation of C& is known as the resuspension factor method:

where:

Sr D

= the = the

resuspension fador (m-') total decay-corrected soil deposition (inventory) in Bq m-2

4.2 RESUSPENSION AND DOSE MODELS

1

69

By comparison with Equation 4.3, one sees that less information is and, in fact, by rearranging available about factors that influence Cd,, Equation 4.5 one obtains the definition of the "resuspension factor," which is really an air concentration normalized to D and has units of inverse length (m-I). Method 2 is more difficult to relate to site properties and thus less specific than Method 1.The deposition, D, can be related to soil surface concentration, S. Experimental evidence abounds that indicates the vertical distribution of radionuclides in the soil within a few months after a surface contaminating event will have a n approximately exponential distribution with depth: A = A, exp ( - 0 2 )

(4.6)

where: a

= known as the inverse relaxation depth (1RL-')

z

=

A,

= the extrapolated value of soil surface concentration

the depth below ground surface level (Bq kg- '1

Actually, the value of A at the surface is probably uniform with depth over some shallow depth, zl.The surface is likely to be wellmixed in the first few centimeters, because of geophysical and biological processes such as rain drop impact, shrinking-swelling cycles, wind saltation, root growth, or freezing-thawingthat are very active in the span of a only a few years (Shinn, 1992). Furthermore, in practice, soil is sampled from the top down in thin layers, usually at least 2.5 cm in depth, and, in fact, it becomes very difficult and sometimes arbitrary to decide what is the appropriate "zero" depth a t very shallow depths, simply because of natural microtopography. Let us define A,, in terms of S, the average concentration in the shallow layer of depth 21, assuming for practical reasons that the radionuclide is distributed uniformly through the layer. Then from Equation 4.6:

& = Sexp(azl/2)

(4.7)

By definition, the total deposition (inventory)D (Bq m-2)is obtained by summing over all soil layers:

D

= pX

(Ai &i)

(4.8)

where: P

Ai

= the soil bulk density (kg m-7 = the average activity in soil layer i in Bq kg-'

Thus, for site-specific assessments, soil sampling should provide the vertical profile distribution A, in order to determine D, and to solve

Equation 4.6 for the inverse relaxation depth. Analyzing a large number of depth increments from a single sampled site may be prohibitive, so a curve-fitting least squares analysis of three to four depth increments is usually employed to solve Equation 4.6. D can be expressed in terms of the inverse relaxation depth by integrating Equation 4.6 to great depth, and substituting Equation 4.7: D

=

pShe

(4.9)

where h,, which is defined as the "characteristic" or effective depth of deposition contributing to resuspension, has the properties: h,

=

(1 1 a)exp (az,1 2 )

(4.10)

he can be estimated easily for the reasonable values of z , (-2.5 cm), and for relaxation depths commonly observed a few months or more after deposition, 0.2 cm-' < a < 0.7 cm-' (1.4 cm < RL < 5 crn); he -5 k 1.5 cm (4.11) Note that heis not equal to the RL, i.e., l / a ,because of the presence of the well-mixed upper layer with average concentration S. Thus, estimates of deposition, D, can be made by applying Equation 4.9, with knowledge of the average surface-soil concentration, S, a reasonable judgment of penetration of the radionuclide so that h from Equation 4.11 applies, a n d using t h e soil bulk density p 1.6 g ~ m - One ~ . is now in a position to utilize the resuspension factor method of Equation 4.5 to estimate Cak,by applying available estimates of Sf.Resuspension factors for the first two months can be derived from the experiments by Garland (1982), those from 2 to 42 months from the observations following the Chernobyl event summarized by Garland et al. (1992), and for long times, from the results of Anspaugh et al. (1975). The latter authors also show a method for estimating resuspension factors for all times, but the more recent evidence from the Chernobyl accident shows a more rapid decline in resuspension factors with time in the short to intermediate times than their formula predicts. Based on these data, where no measurements of this resuspension factor are available, one should use the following estimates:

-

Sf = 1 0 - V t for the first 1,000 d

(4.12a)

Sf =

(4.12b)

for longer times, out to many years,

where:

Sr

=

t

=

has units of m-I the time in days since the contaminating event

4.2 RESUSPENSION AND DOSE MODELS

1

71

4.2.2.3 Method 3, Estimation of C,,, by Mass-Loading-Derived Resuspension Factors. A combination of Methods 1and 2 makes it possible to estimate C,.,, the concentration of radionuclides in air by means of a site-specific resuspension factor derived from the modified mass loading approach. This method results in much less uncertainty for purposes of comparing scenarios in human health risk assessments. The uncertainty due to variations in surface soil concentration S is eliminated, and more conservative, less uncertain variables are used in the estimation of the resuspension factor Sf. Recalling Equation 4.5 and Equation 4.9,

C,i,

=

Sr D

(4.5)

where:

Sr = the site-specific resuspension factor now defined by inserting Equations 4.5 and 4.9 into Equation 4.3 as follows:

sr

= (Er M ) I (p x h,)

(4.13)

To apply this method, Cai, i s estimated from Equation 4.5, with D from Equation 4.9, but with Sr estimated by Equation 4.13. This method is recommended for undisturbed soil surface conditions only, and the uncertainty of estimating CG,in that case is usually lower than for Methods 1or 2. 4.2.3

Estimating the Uncertainty in Values of Air concentration

The Methods 1, 2 and 3 are products, so that an estimate of the expected relative error in C~ can be obtained by the square root of the sum of the squares of the relative errors of each factor. One can provide an estimate for each of the factors under certain conditions. However, these factors are usually the median value of some distribution. Thus, estimates of the relative errors of the above factors are provided by estimating the coefficients of variation (CV), defined as the ratio of the sample S.D. to the mean. I n the case where there is empirical evidence of a logarithmic distribution, the parameters of the log-distribution were converted to CV (Gilbert, 1987):

cv = [exp (ln2s,) where: Sg

In

= =

the GSD the natural logarithm

-

Illn

(4.14)

72

/

4. DOSE FROM INHALED RADIONUCLIDES

Values of CV determined empirically for the modified mass loading and resuspension methods, respectively, are provided in Tables 4.2 and 4.3. The CV values of the enhancement factor, Ef, in Table 4.2, parts a and b, are calculated from sample GSDs of Elvalues reported by the authors cited. The CV values of S for soil sampling were taken from a study where Shinn et al. (1993a) compared 63 soil profiles after the data had been normalized to the deposition D, so that the samples were naturally statistically stratified, and there was no difference in analytic precision between sample locations. The CV values of S for in situ gamma spectroscopy were estimated by Shinn et al. (1993b1, where the CV was shown to be dependent on the combined uncertainty in the instrument calibration and the variability in bulk density, p, inverse relaxation depth, a,and characteristic depth, he. Even the uncertainty in the isotope ratios commonly needed by the method, such as the plutoniumlamericium ratio, contributes to the total CV. The CV values of mass loading, M, i n Table 4.2 came from the sample GSDs of the cited authors' data. Thus, it can be seen from Table 4.2 that the expected minimum error in Ca using the mass loading method can be expressed as a coefficient of variation approximately equal to 0.77.This CV is a combination of the smallest relative errors in Er,S and M shown in the available literature, in either the undisturbed or disturbed case. Furthermore, the expected maximum error in Cai,would be represented by a coefficient of variation of 3.4 or greater if soil sampling, rather than in situ gamma spectroscopy is used, or if mass loading is more uncertain, such as in the case of highly disturbed soil conditions. If an alternative method is used, the expected minimum error (CV)in Cabusing the resuspension factor method is 0.86 soon after a contaminating event, and increases in time (see Table 4.3). At later times, the CV value of the resuspension factor dominates over the CV of deposition (which has nearly the same expected error as the soil measurement). The reason for the differences between the CV values of long-term resuspension factors that were estimated for Chemobyl fallout and for studies in Nevada is likely due to the differences in horizontal representativeness of the Cai,measurements. For Nevada, the site studied was limited in size and the air samplers sometimes were not perfectly located with respect to the wind. It should be expected that this was not a problem where the fallout was more nearly horizontally homogeneous as in the Chernoby1 case. From Chemobyl data, the expected maximum error in Cai, using the resuspension factor method corresponds to a coefficient of variation of about 3.3. This value, obtained when soil sampling is used, is based on the long-term resuspension factor estimates under optimum conditions. However, the error increases drastically if the

CV

0.40 0.40 2.6 0.53 0.36 10.5 0.77 11.1

0.43 2.0 2.6 0.53 0.36 0.52 1.02 0.77 3.4 Disturbed soil Agriculture tractor Soil sampling In situ gamma Urban samplers Agriculture tractor Expected minimum Expected maximum

Undisturbed soil Nuclear event site By soil sampling By in situ gamma Urban samplers Rural samplers Replicate sampler Expected minimum Expected maximum

Description

"n is the number of measurements. bValues obtained by combining several error terms.

Cair

M M cot

S

Er Er S

Disturbed Soil Case

CBir

c~,

M

Er Ef S S M M

Undisturbed Soil Case

Variable s,

(GSD)

Method 1-Modified Mass Loading n'

Reference

Shinn (1992) Loshchilov et al. (1992) Shinn (1993a) Shinn (1993b) Shah et al. (1986) Loshchilov et al. (1992) Calculated Calculated

Shinn (1992) Shinn (1992) Shinn et al. (1993a) Shinn et al. (1993b) Shah et al. (1986) Shah et al.(1986) Luna et al.(1971) Calculated Calculated

TABLE 4.2 -Empirical CV for variables used in estimating resuspension.

cv 0.43 2.0 0.36 0.52 1.02 0.64 2.3

Er

12 14 14 14 20 115

1.83 1.98 2.24 2.77 3.7 10.3 4.2 1.76 2.1 4.9 10.4

-

-

-

nu

s, (GSD)

Undisturbed soil Nuclear event site Urban samplers Rural samplers Replicate sampler Expected minimum Expected maximum

Description S,

6 2 46 28 137

1.51 3.5 1.42 1.63 2.32 1.8 3.8

-

nn

(GSD)

8

P

1

Shinn (1992) Shinn (1992) Shah et al. (1986) Shah et al. (1986) Luna et al. (1971) Calculated Calculated

Reference

ul

'3

2c

i

i

Garland (1982) a Garland et al. (1992) m Garland et al. (1992) Garland et al. (1992) Garland et al. (1992) Kercher and Anspaugh (1993)b From S and h errors From S and h errors m Expected minimum Long term expected Expected maximum

Reference

"n is the number of independent sample periods or sites. bKercher, J.A. and Anspaugh, L.R. (1993). Personal communication (Health and Ecological Division, Lawrence National Laboratory, Livermore, California).

Cair

cair

M

3M

First 2 months At 12 months At 24 months At 42 months Long t e d c h e r n o b y l Long termNevada Soil sampling I n situ gamma First 2 months Long terdchernobyl Long term/Nevada

Description

Method 3 -Mass-Loading-Derived Resuspension Factor Method. Undisturbed Case Only.

0.61 0.77 0.96 1.35 2.1 15.1 2.6 0.61 0.86 3.3 15.3

CV

Variable

c,

cam c&

D

D

Sr Sr

s f

Sr Sr Sr

Variable

used in estimating resuspension.

Method 2-Resuspension Factor Method. Undisturbed Case Only.

TABLE4.3 -Empirical CV for variables

4.2 RESUSPENSION

AND DOSE MODELS

1

75

measurement is confounded by wind direction problems (e.g., for Nevada data). In summary, the uncertainty in the estimation of C& depends highly on several key variables. The surface soil concentration, S, contributes uncertainty to all methods of estimation. S has a large variance if it is derived from a limited number of soil samples and a smaller variance if it is obtained from measurement by in situ gamma spectroscopy. If measurements are made of the mass loading or the resuspension factor, then one must be concerned about the representativeness of the air sampler location. If no measurements are available, then estimates may become highly uncertain in the cases where either long-term resuspension or disturbed soil conditions is the subject in question. The latter is particularly true if C ~ , is being estimated for a soil cleanup exercise. The CV values provided here are based on the empirical evidence obtained from available contaminated sites. More experience in the future should assist in providing a better estimate. 4.2.4

Derivation of Radionuclide Concentrations in Air for Screening

The concentrations C,L for screening for the land-use scenarios discussed in Section 1were estimated from the resuspension factor Sfand Equations 4.5, 4.9 and 4.13. The combination of variables found in Table 4.4 were chosen as parameters for these estimates. The values for M, Efand Sf are annual averages. At a minimum, the mean concentration S, over the top 5 cm is assumed to be measured (Bq kg-'). The other parameters are estimated. The importance of obtaining a representative and areaaveraged value of S cannot be overemphasized (see Section 6.5, Strategy of Determining Radionuclide Concentrations for Screening). The values for Slare derived from Equation 4.6, values for Ef are from observations of resuspension from normal and disturbed soils and values of M from monitoring data in the literature. The bulk density, p, is assumed, conservatively, to have a common value of TABLE 4.4-Selected parameter values for screening calculations. Land-Use Scenario

AG

PV

PS

RV

RS

SUtSN

CC

76

/

4.

DOSE FROM LNHALED RADIONUCLIDES

1.6 g c m 3for topsoil, although the surface soil density is often less than this. The effective depth of deposition, h, can thus be estimated from Equation 4.9 using a sampling depth of the surface soil, 0.025 to 0.05 m. The value of a for most soils will fall in the range 10 to 100 m-l, so that h calculated from Equation 4.10 will have a value on the order of 0.05 to 0.1 m. For sandy soils, soils that have been tilled, or soils subject to construction after the contaminating event, h has a value closer to 0.1 m. The values for h in Table 4.4 should thus be prudently conservative. The diameters of airborne radionuclides have been determined primarily by the diameters of the host soil particles to which the radionuclides have become firmly attached. As discussed earlier, observations have shown that the diameters of airborne radionuclide particles originating from the soil are distributed approximately lognormally for the range of aerodynamic diameters between 0.3 and 10 pm. These distributions are broadly dispersed (GSD > 4), and have a median aerodynamic diameter usually between 2 and 6 pm. For screening purposes, a value of 1 pm is used. This favors the conservative estimate of pulmonary (deep lung) retention and corresponds to the value generally recommended for calculating inhalation dose factors for members of the general public (see Section 4.3.3;ICRP, 1994a; 1996a). The values given in Table 4.4 have been observed a t enough sites to permit determination of the uncertainty in each parameter for these selected scenarios. The major source of uncertainty is usually the combined natural variability of the observed parameters and not measurement imprecision. Thus, most of the parameters have an approximately lognormal frequency distribution, since concentrations of contaminants in soil and air also have lognormal frequency distributions. Observed estimates of the CV derived from these distributions are given in Table 4.5. The uncertainty estimates chosen TABLE4.5-Uncertainty in screening parameter values (CV. Parameter

CV

"Observed in soil contamination zones stratified by in situ gamma spectroscopy measurements.

4.2 RESUSPENSION AND DOSE MODELS

1

77

for use in screening refer to the variability in average annual values of the parameters. Thus the values are closer to the minimum values observed and smaller than would be appropriate for calculating exposures over shorter time intervals. Example of Calculation of Air Concentrations The first step in the estimation of concentrations of radionuclides in air is to calculate the effective deposition (inventory),D, given the radionuclide concentration, S, in the topsoil (representative, areaaveraged soil samples). Equation 4.9 is used, and Table 4.4 provides selected values for h. For example, assume the topsoil concentration, S, is 1Bq kg-' a t a site that is slated for development of commercial buildings. Using the construction land-use scenario, because the soil surface will be stirred by earth-moving and traffic activity, the estimate of D from Equation 4.9, using values from Table 4.4, is D = (1 Bq kg-') X (160 kg m-2) = 160 Bq m-2. The next step is to estimate radionuclide concentrations in air, C&, from Equation 4.5. The value of D is as calculated above and values of the site-specific Sfare calculated, from the method shown in Equation 4.13, or from values given in Table 4.4. Given the value of S = 1Bq kg-', the estimate of D = 160 Bq m-2, and site-specific SF= 4 x m-l, the estimate of C& is C&, = 6 x Bq m-3. = Sf x D = (1.6 x lo2) x 4 x The uncertainty of C& can be estimated by determining the coefficient of variation of C& as a figure of merit. The CV is calculated from the square root of the sum of squares of each of the CV in Sf, ph and S given in Table 4.5: CV = [(0.78)2 + (0.41)2 + (0.5)21* = 1.0. (4.15) The average annual outdoor concentrations in air were estimated in this manner for each of the land-use scenarios considered. These values, in turn, were used to estimate the annual effective committed doses using Equation 4.2. These estimated average annual air concentrations and associated uncertaintylvariability are given in Table 4.6. Note that the values are per unit radionuclide concentration 4.2.5

TABLE 4.6-Radionuclide concentrations in air used in screening dose ~alculations.~ Land-Use Scenario

AG

PV

'Per 1 Bq kg-' average in top 5 cm.

PSRS

RV

SUISN

CC

78

1

4. DOSE FROM INHALED RADIONUCLIDES

in soil. The CV value does not include the uncertainty in the soil concentration. The uncertainty estimates in Table 4.6 are used in Section 7 to estimate the overall uncertainty (range) of inhalation dose to the most exposed populations.

4.3

Discussion of Dose Model Parameters

Once the average annual radionuclide concentration in air has been estimated, one can use Equation 4.2 to estimate the annual E(T).Each of the remaining required parameters is discussed below.

4.3.1 Auerage Nuclide Concentration in Soil In order to calculate the annual dose commitment, one needs to determine the concentration of the particular nuclide at the time of inhalation, and the integral activity inhaled. The activity inhaled is in t u n directly proportional to the surface soil nuclide concentration, S . 4.3.1.1 Concentration in Soil for a Site-SpecificAssessment. For a site-specific dose assessment one should determine the local average depth distribution (see Section 6) and, if possible, the average concentration in the top 2.5 cm. Preferably, one should consider all the factors discussed in Section 4.1 and, if at all possible, measure the average daily airborne radionuclide concentrations as well as concentration in soil. 4.3.1.2 Concentration in Soil for Screening. For screening purposes, the median and maximum annual committed doses to a member of the most exposed population group must be estimated. It is assumed that the average concentration of each contaminant in the top 5 cm has been measured a t some time ,t in a manner sufficient (see Section 6.5) to achieve an acceptably small CV. It is assumed for screening that the radionuclide is, and remains, uniformly distributed over the top 5 cm. As discussed in Section 3, however, the depth distribution soon after the original deposition may be such that the concentration near the surface is somewhat higher and the concentration in the surface soil will decrease with time due to percolation to deeper layers or due to erosion. The additional uncertainty resulting from this assumption is likely to be minor, since, as discussed in Section 4.2.4, a conservative estimate for the effective depth of penetration is assumed. For most situations, as

4.3 DISCUSSION OF DOSE MODEL PARAMETERS

1

79

discussed in Section 3.2.2.4, the radionuclide will likely have penetrated significantly into the ground over the course of the first year. As discussed earlier for the external dose pathway (see Section 3.2.1.1), the concentrations will also change with time due to decay and buildup of progeny. Thus the soil concentrations used for calculating screening doses were corrected to provide the maximum decay series dose from all pathways in any year over a 1,000 y interval. 4.3.2

Indoor versus Outdoor Concentrations in Air

Even when indoors, one may still be subjected to contaminated air. Studies have shown that indoor/outdoor concentration ratios can vary widely depending on the tightness of the dwelling (air exchange rate), the type of structure, etc. (DOE, 1990; Roed and Camell, 1987). Radionuclides can also be brought into the building on shoes and clothing and then resuspended and inhaled. For site-specific studies, it may be useful to actually compare indoor versus outdoor airborne radionuclide concentration measurements under various meteorological and living conditions. For screening studies, however, an average indoor to outdoor ratio of 0.3 (GSD = 1.45) was chosen as a reasonable yet prudently conservative value for screening. The estimated uncertainty corresponds to an estimated range of from about 0.1 to 0.9. Both the ratio and range are consistent with results from a number of studies (NCRP, 1993a). Note that although rac2lonuclide concentration levels are generally lower indoors than out, total resuspendable dust levels indoors may be comparable to or even higher than outdoors due to indoor activities (Thatcher and Layton, 1995). 4.3.3

Inhalation Dose Factor

The E(z) to a reference human due to inhalation of a given amount of radionuclide has been reported for a large number of radionuclides (EPA, 1988; ICRP, 1979-82; 1996b; Phipps et al., 1991) using biokinetic models developed by the ICRP (1977; 1994a). These calculations account for the buildup and decay of daughter products in the body. The biokinetic models consider the particle size distribution, transfer from lung to other body organs, and uptake from the GI tract. The calculated individual organ and effective doses are for one or more particular lung absorption types based on the likely chemical form of the inhaled radionuclide. The ICRP inhalation dose model has recently been revised (ICRP, 1994a) and the ICRP has published calculated E(z) for both adults

80

1

4. DOSE FROM INHALED RADIONUCLIDES

and children as members of the general public of various ages for over 800 radionuclides (ICRP, 1995a; 1996b).E(T)for workers have also been published based on the new inhalation dose model (ICRP, 199410). Adult ICRP organ dose and dose-equivalent estimates for most radionuclides, as a function of clearance class based on the ICRP Publication 30 lung model and using ICRP Publication 26 organ weighting factors, are reported in Federal Guidance Report No. 11 (EPA, 1988). The National Radiological Protection Board (NRPB) of the United Kingdom has also published calculations of effective dose factors based on the ICRP Publication 30 lung model (Phipps et al., 1991). The NRPB,however, used ICRP Publication 60 organ weighting factors and also provided estimates as a function of age for a large number of nuclides. 4.3.3.1 Dependence of Dose Factor on Lung Absorption and Particle Size. The physical-chemicalform of the radionuclide andlor matrix material and size of the particle upon which it is imbedded determines the absorption and clearance properties of the material in the various compartments of the respiratory system after it is either inhaled or ingested. For example, if plutonium is in the form of h 0 2 , it is insoluble and less available for transfer across the gut wall and the lung-blood barrier than if it is in an ionized or other more chemically soluble form or bound to small particles or organic complexes of some type (ICRP, 1979-82; 1994a; 199413). Often, however, it is not possibleto determine the chemical form. Thus, a conservative approach, used in previous NCRP screening models (NCRP, 1996), is to assume that the inhaled material estimated from resuspension measurements is of the absorption type that results in the highest E(.r). The estimated dose for some nuclides varies significantly depending on the assumed absorption type as shown in Table 4.7. The new ICRP Publication 66 inhalation dose model no longer uses only three discrete clearance-rate classes (ICRP, 1994a); however, the three absorption types generally listed, slow, moderate and fast, roughly correspond to the ICRP Publication 30 day, week and year clearance rates, respectively. The ICRP (1996133lists recommended absorption types to be used in calculating E(z) to the general public for various chemical compounds likely to be encountered in the environment as well as for unspecified compounds. When the dose factor has been calculated for more than one absorption type or particle size, the dose factor corresponding to the recommended absorption type clearance rate should be used for site-specificstudies if the chemical form of the nuclide in the soil is known. Similarly, if the size distribution is known to be significantly different from that assumed for screening

4.3 DISCUSSION OF DOSE MODEL PARAMETERS

81

1

TABLE4.7-Comparison of adult inhalation E(T)factors (Sv Bq') versus lung absorption type for selected radionuclides." Absorption Type Nuclide

F (fast)

M (moderate)

S (slow)

"Data from ICRP Publication 72 (ICRP, 199613)

purposes, one should use a dose factor calculated for particles nearer that size rather than the ICRP default values. Table 4.8 compares the calculated dose factors for an adult for 1Fm versus 5 km median particle sizes. Note that for absorption types S and M, the use of dose factors for 1 krn AMAD particles is generally conservative but may underestimate the dose for fast absorption. However, since the recommended absorption type used in this Report for the nuclides where inhalation is likely to be an important dose pathway is usually type M (see Appendix C), the dose factors used should be conservative. The ICRP recommends and uses a median particle size of 1 km TABLE4.8-Variation of inhalation dose factor (Sv Bq-') with median particle sizea Absorption Type-F

Absorption Type-M

Nuclide

1 pm

5 pm

1pm

5 pm

Co-60 Sr-90 Zr-95 Ru-103 Ce-144 U-238 Pu-239

5.23-9 2.43-8 2.53-9 4.93-10

-b

9.63-9

7.1E-9

3.03-8 3.OE-9 6.83-10

-b

-b

4.53-9 2.33-9 3.43-8 2.63-6 4.73-5

3.63-9 1.93-9 2.33-8 1.63-6 3.23-5

-b

-b

4.93-7

5.83-7

-b

-b

"Data from ICRP Publication 68 (ICRP, 1994b). bNodata.

Absorption Type-S 1pm

5pm

82

1

4.

DOSE FROM INHALED RADIONUCLIDES

for members of the general public and 5 p m for adult workers (ICRP, 1996b),although, as discussed earlier, resuspended material is likely to have an AMAD higher than 1 pm. Dependence of Dose Factor on Age. The dose factor also varies significantly with age, as shown in Table 4.9 which gives the ratio of the infant (1y) and child (10 y) dose factors to the corresponding adult dose factor. Thus, a site-specific dose assessment should calculate separate inhalation doses for infants, children and adults. This requires age-specific breathing rates, which are discussed in Section 4.3.4.A site-specific assessment should also be based on realistic exposure times for each age group.

4.3.3.2

Organ Doses versus Effective Dose. For site-specific dose assessments it may also be appropriate to estimate the doses to specific organs rather than E. For some nuclides, the highest organ doses differ significantly from E. Table 4.10compares the dose factor for the most critical organ with E(T).

4.3.3.3

Uncertainty and Variability in Dose Factors. The NCRP (1998)has estimated the uncertainty in the biokinetic models used to determine inhalation dose factors. Table 4.11 summarizes this uncertainty for some selected nuclides of interest. Because of the number and complexity of the assumptions needed to calculate the dose factors and the lack of adequate biokinetic data for many nuclides, the uncertainty estimates given in Table 4.11 are based TABLE 4.9 -Variation of inhalation dose factor with age for selected radi~nuclides.~

4.3.3.4

Nuclide

Sr-90 Se-75 Tc-99 1-129 CS-134 CS-137 Ra-226 U-238 Pu-238 Pu-239 Am-24 1

Absorption Type

M F M F F F

M M M M M

Infant I Adult

Child / Adult

3.1 6.0 3.3 2.4 1.1 1.2 3.1 3.2 1.6 1.5 1.6

1.4 2.5 1.4 1.9

0.8 0.8 1.4 1.4 1.O 1.0 1.0

LDatafrom ICRP Publication 72 (ICRP, 1996b). The infant values are for age 1 y while the child values are for age 10 y. See Appendix C for a complete listing for all covered radionuclides.

4.3 DISCUSSION OF DOSE MODEL PARAMETERS

1

83

TABLE 4.10-Ratio of dose to critical (most exposed) organ to effective committed inhalation dose for selected radionuclides." Absorption

Nuclide

Type

Critical Organ

Sr-90 Zr-95 Tc-99 1-129 Ce-144 Cs-137 U-238 Pu-239

M M M F M

Lung Lung Lung Thyroid Lung Respiratory tract Lug Bone surface

F M M

Ratio

'For adults, data from ICRP Publication 71 (ICRP, 1995a).

TABLE 4.11-Estimated uncertainty in inhalation dose factors for selected radwnuclides." Estimated Range Nuclide

Male Adult

Othersb

10 (female teen) 10 10

5 20 10 20 20 5 10 10 5 10 10 10 10 10 10 20 "dapted from NCRP (1998). The estimated range R can be interpreted as indicating that the dose factor for some individuals may be a s much as a factor of R higher or lower than the dose factor recommended by the ICRP. bSpecial population groups generally consisting of diseased people, or of infants or children.

84

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4.

DOSE FROM INHALED RADIONUCLIDES

on expert judgment rather than a mathematical calculation. The uncertainty ranges given use ICRP Publication 30 values as a starting point and include uncertainty in the appropriate gut transfer factor as well as the uncertainty in the biokinetic factors used in the ICRP Lung Model. They reflect the degree of reliability that a particular group of experts has in the ICRP Publication 30 values and thus may include an element ofbias. The column labeled "others" reflects the larger uncertainty in model parameters for infants or children or for groups suffering from certain diseases. Although the uncertainty estimates are based on the ICRP Publication 30 model rather than the newer ICRP Publication 66 model, we have utilized them for estimating the uncertainty in the inhalation dose factors used for calculating the screening doses in this Report since no similar estimates have been made for the newer model. 4.3.3.5

Recommended Dose Factors for Screening. In this Report,

E(T)factors published by the ICRP (1996b) for members of the general public are used for calculating screening doses. These dose factors are listed in Appendix C. If a value for a given nuclide was calculated for more than one clearance rate or median particle size, the value recommended for use by the ICRP for unspecified compounds was used. It was also assumed that all the airborne material was in the respirable size range covered by the dose model which assumed a 1p m median diameter particle size. As discussed earlier, this is generally conservative since the median diameter of suspended material is likely to be higher than 1 bm. GSDs were chosen to reflect the estimated range for the male adult for use with the adult dose factors discussed above, assuming the range about the median corresponds to about 3 GSD. A conservative estimate of uncertainty (GSD = 1.4 to 2.2) inferred from the uncertainty ranges presented in Table 4.11 was assigned to E(T) factor. A GSD of 1.7 was used for all nuclides listed in Table 4.11 except that for the well-studied nuclides, 90Sr,=9pU and 137Cs.For '"Sr, ='Pu and L37Cs a GSD of 1.4 was used. A GSD of 2.2 was assigned for all other radionuclides. As indicated in Table 4.9, the mean dose factors for some nuclides for infants or children may be much higher than those for adults. However, as will be shown later in Section 4.3.5, this higher dose factor is compensated for by a much lower inhalation rate. 4.3.4

Usage Factors

4.3.4.1 General. The true inhalation dose depends on the time a person is exposed to resuspended soil and the amount inhaled over

85

/

4.3 DISCUSSION OF DOSE MODEL PARAMETERS

the course of the exposure. The time exposed, both indoors and out, is assumed to be the same as the time exposed to external radiation (see Table 3.12). The amount inhaled, however, will depend on the breathing rate, which will vary with the type of activity and with age. Table 4.12 gives reported average breathing rates for various types of activity versus age. Screening Values. Table 4.13 gives the chosen values for breathing rates and associated variability used in the screening calculation for both indoors and outdoors. These values represent averages for outdoor and indoor exposure over the course of a year. The outdoor rate is based on the assumption that the individuals making up the critical population are active outdoor workers. The inhalation rates are for adults to correspond to the use of adult dose factors.

4.3.4.2

TABLE 4.12-Average breathing rates (m3d-I) for various types of activity. Reference

Adult

Child (10y)

Infant(1 y)

UNSCEAR (1993)-all activities

22

15

4

15 7 27 43-53

5 4 8

ICRP Publication 66 (1994~) -all activities 16-20 Resting 8-11 Light activity 30-36 Heavy activity 65-72

-

ICRP Publication 66 ( 1 9 9 4 ~-average ) annual indoor and out Housewife 18 Sedentary male worker 22 Sedentary female worker 18 Outdoor worker 25

TABLE 4.13 -Inhalation rates used for calculating adult screening doses for land-use scenarios considered (m3d-3. Land-Use Scenario

Outdoor-median GSD Indoor-median GSD

35 1.2

35 1.2

35 1.2

30 1.2

30 1.2

25 1.2

35 1.2

-

-

-

20 1.2

20 1.2

20 1.2

-

86

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4. DOSE FROM INHALED RADIONUCLIDES

Table 3.12 gave the chosen indoor and outdoor exposure times and associated uncertainty assumed for the screening calculations for each land-use scenario. These times apply to both external exposure and inhalation exposure. Based on information reviewed in NCRP Report No. 76 (NCRP, 1984b), minute breathing rates tend to be distributed lognormally with a GSD of about 1.3 to 1.5. Since the screening criteria in this Report are based on average annual doses, and the variability in inhalation rate over the course of a year will tend to average out, the GSD of 1.2 assigned to the chosen estimates is believed to be reasonable and conservative. It allows for the possibility t h a t a member of the critical group will spend a significant fraction of hisher time in heavy activity, both while indoors or outdoors. 4.3.5 Age Dependence of Dose-Child, Infant Screening Dose Calculations

Using the age variation in the inhalation dose factor given in ICRP Publication 72 (ICRP, 1996b) (see Table 4.9 and Appendix C) and the differences in average (all activities) inhalation rate given in Table 4.12 for infants (1y old) and children (10 y old) versus a d ~ l t s , ~ the committed effective inhalation dose was also calculated for children (age 10 y) and infants (age l y) for each nuclide for the landuse scenario sites where children a r e allowed to be exposed. Table 4.14 shows the ratio of the inhalation dose for children and infants to adults corrected for breathing rate for some selected radionuclides. As can be seen, the ratio of infant or child dose to adult dose ranges from about 0.2 to 1.7. The child (age 10 y) to adult dose ratio, in particular, is sometimes >1 for some nuclides despite the much lower breathing rates. Because the adult-child-infant differences are small compared to the uncertaintylvariability in the biokinetic factors, applicable clearance class, and usage factors, the same uncertainty estimates for breathing rate and dose factor are used for the calculations of the childlinfant doses as for the adult calculations. Since the calculated screening doses use an adult outdoor breathing rate based on an active worker, the maximum dose calculated based on these uncertainty estimates should still include the dose to even the most exposed infant or child. However, for site-specific assessments where children or infants might be exposed, it is recommended that whenever possible, appropriate age-specific parameters and 'A ratio of 0.25 was used for the infant to adult (VA) breathing rate and 0.8 for the child to adult (CIA) rate.

4.3

DISCUSSION OF DOSE MODEL PARAMETERS

1

87

TABLE4.14-Ratio of child to adult a n d infant to adult committed effective inhalation dose for selected radionuclides." Nuclide

Se-75 Sr-90

Tc-99 1-129 (3-134 CS-137 Ra-226 Pu-238 Pu-239 Am-241 U-238

Child/Adult

InfanVAdult

1.7 1.0 1.0 1.3 0.6 0.6 1.0 0.7 0.7 0.7 1.0

1.1 0.6 0.6 0.4 0.2 0.2 0.6 0.3 0.3 0.3 0.6

"Estimated using child (age = 10 y), infant (age = 1 y), adult dose factors from ICRP Publication 72 (ICRP, 1996b), and relative average breathing rates from Table 4.12. related uncertainty be used. Note that ICRP (199613) provides dose factors for a wider range of ages than the three age groupings used for this Report. For the land-use scenarios where children can be present, the separate inhalation, ingestion and external exposure doses calculated for them were summed stochastically, as described in Section 7, and the highest dose (for either children or adults) used to calculate the screening guidance given in Section 2. Note that in Appendix A, separate calculations for infants and children are not tabulated since only the highest inhalation dose (generally that for children) was conservatively combined with the childlinfant doses calculated for the other dose pathways. 4.3.6

Dose from Inhalation of Airborne Radon a n d Progeny

The potential doses due to exposure to radon and its progeny were reviewed in NCRP Report No. 78 (NCRP, 1 9 8 4 ~and ) by UNSCEAR (1988; 1993). Most of the inhalation dose ascribed to 226Rain the surface soil will be from inhalation of radon progeny. The screening dose calculations for 226Raused the values given in UNSCEAR (1993) relating E(T)from radon and its progeny to a given outdoor air radon concentration, i-e., the average 226Ra concentration measured in soil of 40 Bq k g 1 to the average outdoor radon gas air concentration of

88

/

4. DOSE FROM INHALED RADIONUCLIDES

10 Bq me3,resulting in an annual E ( d of 650 pSv y-I from radon and its progeny if exposed 100 percent of the time (UNSCEAR, 1993). Thus, an average 226Ra soil concentration of 1Bq kg-' would correspond to an inhalation dose of about 15 pSv y-l. For sites where children might be the most exposed population, UNSCEAR recommends using a dose 1.5 to 2 times this value. The inhalation screening doses for a26Ragiven in this Report thus include a contribution of 15 pSv y-I (Bq kg-')-' for adults and 25 pSv y-' (Bq for children (for sites with dwellings) for the fraction of time spent outdoors on the contaminated site, and 8 pSv y-l (Bq kg-')-' for adults and 13 pSv y-I (Bqkg-')-' for children for time spent indoors.'

4.4 Summary of Recommended Parameter Values for

Inhalation Dose Estimation Land-Use Dependent Outdoor air concentrations Indoor/outdoor ratio Breathing rates Indoor/outdoor occupancy Land-Use Independent Dose factors Age correction factor

see Table 4.6 0.3 (GSD = 1.45) see Table 4.12 see Table 3.12 see text, Appendix C (GSD = 1.4 to 2.2) see text, Table 4.14, Appendix C

4.5 Calculated Screening Inhalation Doses

Table 4.15 lists the calculated median annual E(z) from inhalation for an adult member of the most exposed population group for each land-use scenario for a number of important radionuclides likely to be found in contaminated soils. Estimates of the likely range (5 to 95 percentiles), calculated stochastically based on the uncertainty analysis discussed in Section 7, are also given. The complete listing of inhalation screening doses for all nuclides considered is given in Appendix A. 'For screening, the indoor gas concentration is arreumed to be equal to the outdoor level but since the equilibrium equivalent radon concentration indoors is about 40 percent compared ta 80 percent outdoors (UNSCEAR,1993) the dose when indoors L halved.

AG PV

PS

RV

RS SU/SN

CC

"Median dose to representative adult member of critically exposed population. Ranges in parentheses represent 5 to 95 percent percentiles. Doses include contributions of decay products and are maximum annual doses over a 1,000 y interval. Nuclide concentration is per kilogram of dry soil assumed uniform over top 5 cm. See Appendix A for complete listing of inhalation doses. Doses are also given for 226Rawithout the contribution from 2a2Rnto illustrate contribution from radon.

Nuclide

Land-Use Scenario

TABLE 4.15-Annual E(T.from inhalation of selected radionuclides [Sv (Bq &-')-'I."

m w

.

2

90

1

4.

DOSE FROM INHALED RADIONUCLIDES

The inhalation doses vary more with land-use scenario than do the external pathway doses, and exhibit a much larger likely range. However, the inhalation doses generally control the screening limits only for nuclides such as the transuranics that emit nonpenetrating radiations and generally only for relatively dusty sites such a s sparsely vegetated and construction sites.

5. Dose from Ingested Radionuclides If contaminated land is used for the production of food for either human or animal ingestion, the ingestion of this food, or of the by-products (meat, milk) from animals eating forage grown on the contaminated land may result in a radiation exposure and E(z). Ingestion of contaminated soil may also result in E(z). This Section discusses models which can be used for estimating such ingestion doses both for screening purposes and for subsequent site-specific assessments, the relevant parameters used in these models, and the calculated screening doses.

5.1 Dose Model The annual committed effective dose, incurred from the ingestion of a given quantity of food containing a particular radionuclide can be estimated from the following expression: E,,, = Z,(Ci x Ri) x fi

x exp[- A(ti - to)] x Dfi,,

(5.1)

where: of a particular radionuclide in foodstuff i at harvest =the average annual intake of foodstuffi (kg) Ri = t h e fraction of R i derived from the contami6 nated site exp[ - A(ti- to)]= a correction factor to account for radioactive decay between harvest (to)and ingestion (t) = the committed effective dose for ingestion (Sv per Dfi, Bq) that would result from an intake of 1 Bq of this nuclide

Ci

= the concentration in Bq kg-'

Foodstuff i can be vegetation grown in the contaminated soil or milk or meat from animals which have ingested vegetation grown on the contaminated soil. The foodstuff can also take the form of contaminated soil ingested by humans either directly or as a result of eating

92

1

5. DOSE FROM INGESTED RADIONUCLIDES

unwashed contaminated vegetables or fruit. The total dose from ingestion is obtained by summing the individual contributions from each nuclide for each foodstuff and from soil itself. The concentration in a given foodstuff, Ci,depends on the foodstuff and the particular pathway by which the radionuclides in the soil were transferred to the foodstuff. For edible vegetation, eg., vegetables, grains and fruits directly ingested by humans, the contamination can result from direct root uptake by t h e plant or from contamination due to resuspension of contaminated soil with subsequent deposition and adhesion onto the plant surfaces and also possible translocation into its edible portions. The concentration (Bq kg-') in a given type of vegetation due to root uptake and resuspension processes can be estimated for screening calculations from the following expression:

B,

= an empirically determined soil to vegetation transfer factor

for root uptake usually expressed as Bq kg-' of wet vegetation or dry animal fodder per Bq kg-' of dry soil B,, = a similar transfer factor representing the net effect of all resuspension processes S = the concentration Bq kg-' (dry) at harvest of the soil to which the vegetation is exposed (for screening S is assumed to be the surface soil concentration) In the case of contamination by resuspended soil, the mechanisms can be quite complex. Not only can airborne resuspension and subsequent redeposit contaminate the vegetation, but also phenomena such as rain splash, saltation and mechanical disturbances during harvest. Thus, B,, must also be determined empirically for the particular site or type of site and the type of vegetation. Previous NCRP screening models [e.g., NCRP Commentary No. 3 (NCRP, 1989; 1996)lfocused on contamination of vegetation by direct airborne deposition. Thus, the contamination was calculated using an estimated interception fraction and deposition velocity. Such an approach is not valid for contamination dominated by resuspension. In this case, the major source of contamination is due primarily to contaminated soil being deposited onto, and adhering to, the vegetation or to root uptake. The major dose pathways are thus direct ingestion of the contaminated soil adhering to the vegetation by humans or animals or ingestion of edible portions contaminated by root uptake or translocation of radionuclides from plant surfaces to edible portions. Incorporation of the radionuclide into the plant itself via resuspension is generally of much smaller importance than for

5.1 DOSE MODEL

1

93

the case of direct deposition. Thus the use of the airborne deposition screening model formulation for contamination by resuspension will generally significantly underestimate the concentration in or on unwashed vegetation. Vegetation for human consumption is usually washed, although not always thoroughly enough to remove all attached soil. However, soil adhesion can be particularly important for forage or pasture vegetation eaten by animals because, unlike vegetation consumed by humans, it is unwashed. The concentration CmilLWt of a particular nuclide in contaminated animal products depends on several additional factors. C& ,, can be calculated from the following expression: C*meat

=

Cfodder

X Qmilk, meat X

T Q ~ Imeat I I ,X F d ,meat

(5.3)

where: = the nuclide concentration (Bq kg-') in the dry fodder

Cfodder

Qrmlk, meat

2Qw

,,

F d ,meat

eaten by the animal, nuclide concentrations in fodder can be calculated from Equation 5.2 = the average daily intake (kg d-') of the animal of dry contaminated fodder = the fraction of the animal's total feed derived from the site = an empirically determined transfer factor representing the equilibrium concentration in meat or milk (d kg-' or d L-') resulting from a given daily intake of radionuclide by the animal (C* ,, is thus given in Bq kg-' or Bq L-' of meat or milk, respectively.)

If the animal consumes different types of fodder, i-e., grain, grass, etc., one first uses Equation 5.2 to calculate the nuclide concentration in each type of vegetation. The average fodder concentration is calculated by summing the different radionuclide concentrations, weighting by the fraction of the animals diet from each fodder type. The dose incurred by an individual ingesting foodstuffs or soil from a contaminated site will, of course, depend on the type and amount of food grown on the site (the land-use scenario). The dose will also depend on the fraction of hidher total diet that originates from the contaminated site. The type of site, the type of soil at the site, the particular vegetation, and the distribution of the radionuclides in the soil will influence the amount of radionuclide available for root uptake or that may be resuspended and deposited onto the vegetation. The amount of soil ingested will also depend on the type of site and the particular lifestyles and habits of the inhabitants (see Section 5.3). The soil to vegetation and feed to milklmeat transfer factors in the expressions given above also differ considerably from

94

1

5. DOSE FROM INGESTED RADIONUCLIDES

nuclide to nuclide and vary depending on the nuclide's chemical form in the soil. For the screening models used in this Report, the effect of many physical and chemical processes are often grouped together into a single transfer factor for which a steady-state annual average is assumed. One may wish to consider using more sophisticated foodchain pathway models such as PATHWAY (Whicker et al., 1990) for site-specific assessments. Dynamic models such as PATHWAY simulate the time-dependence of root uptake and soil adhesion during the course of plant growth and harvest and more readily allow for inclusion of site-specific conditions. Models such as PATHWAY also allow one to treat temporal variations in animal feed practices, pasture season, etc., in much more detail than do the screening models described in this Report.

5.2 Discussion of Model Parameters Each of the important parameters needed to estimate E(T)from ingestion of radionuclides, including the recommended values for both screening and site-specific assessments, uncertainties and variability, is discussed in the following paragraphs.

5.2.1 tiadionuclide Concentration in the Soil To calculate the dose from Equations 5.1 through 5.3, one needs to know the relevant nuclide concentrations in the soil. For sitespecific assessments one should use a suitably spatially averaged nuclide concentration a t the time of interest, i.e., the concentration in the surface soil during above-ground vegetation exposure for contamination via resuspension or the concentration in the root zone at harvest for contamination via root uptake. The decay and ingrowth of progeny during the growing period should also be considered. The NRC (Kennedy and Strenge, 1993) has provided guidance on how to calculate the concentration of each nuclide and its progeny in growing vegetation from measurements in soil. For the purposes of the screening calculations described in this Report, it is assumed that the average concentration of each radionuclide of concern in the top 5 cm of soil has been measured a t some time to (see Section 6). The screening calculations assume a mean concentration of 1 Bq kg-' over the entire root zone. As discussed in Sections 3 and 4 for the external and inhalation dose pathways,

5.2 DISCUSSION OF MODEL PARAMETERS

1

95

the average annual concentration in soil of that nuclide and each of its progeny for each year over the succeeding 1,000 y period were then calculated from the decay data given in Appendix B. Screening doses were computed for each radionuclide using Equations 5.1 through 5.3, and the highest series sum used for the screening criteria reported in Section 2. Thus, for calculating ingestion screening doses from Equations 5.1 and 5.2, the average concentration over the entire year of highest dose was used for the concentration in soil a t harvest, under the assumption that crops are grown and ingested throughout the year. The average concentration in the top 5 cm is assumed to have been measured according to the guidance given in Section 6, resulting in an uncertainty in the mean concentration within certain bounds (i.e., with CV 5 0.5).

5.2.2 Human Diets

In order to calculate the average annual dose from ingestion, it is necessary to determine the average quantity of vegetables, fruits and grains grown in the contaminated soil that were consumed during the year. One must also determine the quantities of meat and milk consumed annually from animals that ingested fodder or other vegetation grown on the site. Human diets vary considerably, both with locale as well as with age. For site-specific studies, it may be possible to determine the specific diet of individuals who may be impacted. For these screening calculations, however, generic diets were used. 5.2.2.1 Variability in Human Diet. Reported per caput annual consumptions of major food groups as a function of age for the United States are listed in Table 5.1. Also listed in the table are some estimates of the annual consumption of each food group as a function of age. Site-specific dose assessments may require a survey of the exposed population to determine their dietary habits and the fraction of their diet that is derived from the contaminated site. It is also important to consider the potential exposure to populations who do not live on or near the contaminated site but who may be exposed as a result of ingesting food produced on the site. On average, beef is generally consumed in slightly greater quantities than other types of red meat (about 28 kg y-I versus 22 for pork and <1 for lamb and veal (USDA, 1994). However, in recent years the United States population has been consuming increasingly greater quantities of poultry (about 30 kg y-I in 1994) compared to previous years.

Included in other vegetables 58 + 62 (potatoes)

USDA, 1994b per capita Child

Infant

Included in meat 105 110 120

Included in other vegetables 140 90 45 50 35 15

Adult

UNSCEAR,1993

Child

Infant

Included in meat 110 200 170

Included in other vegetables Included in other vegetables 95 59 37

Adult

RG1.109 (NRC, 1977Y

(wet weight).

"Population weighted data from 1977-78 USDA survey, Kennedy and Strenge (1993)based on data from Pao et al. (1985). bU.S. Department of Agriculture, Economic Research Service, Food Consumption, Prices, and Expenditures, Annual (USDA, 1994). 'All vegetables, grain; as reported by NCRP (1984b). dRoots and fruits. "Fruit, vegetables and grain.

Poultry Fluid milk (Ly-l)

Fruit Grain, cereal Red meat

Leafy vegetables Other vegetables

Food Group

Kennedy and Strenge, 1993'

TABLE 5.1 -Average annual consumption rates of various food groups (kg y-')

P

6.2 DISCUSSION OF MODEL PARAMETERS

1

97

5.2.2.2 Screening Values for Human Diet. For screening purposes, it is convenient to group dietary components into just three categories: vegetables, meat and milk. The rationale for this is that the variability in the transfer of radionuclides from soil to vegetation between various species of vegetation is generally much smaller than the variability in the same species from site to site, as will be shown later. Furthermore, for most nuclides, separate data are not available for each animal and plant. Table 5.2 gives the annual ingestion of vegetables, meat and milk used for the screening calculations along with the chosen GSD. These estimates are somewhat higher than the per capita values given in Table 5.1 since they are intended to represent the median ingestion rates for the most exposed adult populations. They also were chosen conservatively to account for the fact that the values in Table 5.1 do not account for certain processed foods. For example, USDA (1994) indicates that the consumption of processed milk products is about equal to fluid milk consumption. NRC regulatory guide 1.109 (NRC, 1977) suggested that for screening, one should assume a maximum ingestion r a t e for milk of 300 L y-' for adults and 400 L y-I for children, for meat 110 kg y-', and for fruit + vegetables + grain 520 kg y -I. The GSDs in Table 5.2 were chosen to include these upper limits a t the 3 S.D. or (GSD)'level. The fraction of an individual's total diet obtained from the contaminated site will vary depending on the land-use scenario, size of the area of contamination, climate, types, number and variety of crops grown, and other factors. Most screening guidance in the past has generally assumed that a n individual derives hisher entire diet from the contaminated site. However, this assumption is usually overly conservative, since no distinctions are made regarding land use. In this Report, the usage parameter values and associated ranges used to estimate the mean screening dose for each land-use scenario (Table 5.3) were chosen somewhat arbitrarily to reflect the fact that it is unlikely that any individual will derive hisher entire diet of vegetables, fruit and grain from a single agricultural site (and particularly TABLE5.2-Human consumption rates used for calculating screening doses. Food Group

Median Consumption Rate, Adults

Vegetables, fruits, roots, grains (kg y-I wet weight) Milk (L y-I) Meat (kg y-l)

200 100

300

GSD

1.20 1.25 1.15

98

1

5. DOSE F'ROM INGESTED RADIONUCLIDES

TABLE5.3 -Usage factors (percentage of diet from site) for calculation of screening doses. Land-Use Scenario

Food Type

Agricultural

Vegetables, fruits, roots, grains

Heavily vegetated pasture

Milk Meat

Sparsely vegetated pasture

Milk Meat

Heavily vegetated rural

Vegetables, fruits, roots, grains Milk Meat

Sparsely vegetated rural

Suburban

Average from Site, %

Range, 46

50

20-80

30

10-50

100 30

10-50

-

Vegetables, fruits, roots, grains Milk Meat

15

5-25

100 10

0-20

Garden vegetables

15

5-25

-

not from a backyard garden). It is even more unlikely that any individual would get all of hisher vegetables, grain and fruit from a rural, nonagricultural site. This is particularly true since, for most agricultural sites, only a limited variety of crops will be produced and crops will not be grown year round, even though some families may freeze produce for later consumption. Thus, the values chosen for vegetables and meat from rural sites, although
5.2 DISCUSSION OF MODEL PARAMETERS

1

99

land-use scenarios, it is assumed that not even the most exposed populations would receive all of their meat and vegetables from the contaminated site. The parameter values chosen for the quantity of meat consumed from rural sites are intended to account for the possible ingestion of wild game that grazes on the contaminated site as opposed to domestic livestock. The pasture scenario should be used for estimating ingestion dose rates from meat for the latter. Similarly, the vegetable values for rural and suburban sites are intended to include contributions from backyard gardens or, in the case of forested sites, from wild plants such a s berries or mushrooms. These parameter estimates are based on expert judgment rather than hard data, but are believed to be reasonable estimates of the central tendency for a member of the critical group. However a conservative estimate of uncertainty was assigned for use in calculating doses to the most exposed individual.

5.2.3

Soil to Vegetation Transfer Factors

The soil to vegetation transfer factor B,for a given type of vegetation and for a given radionuclide can vary considerably from site to site with season and time aRer c~ntamination.~ These variations depend on such factors as the physical and chemical properties of the soil, environmental conditions, and chemical form of the radionuclide in the soil. Furthermore, soil management practices such as plowing, liming, fertilizing and irrigation can also affect the uptake of radionuclides by vegetation (IAEA, 1994;NCRP, 1984b, Simmonds et al., 1995). Often, values for B,and B,have been derived empirically from controlled experiments in greenhouses. The reported values are conservative in that they generally assume the activity in the plant has reached equilibrium with the activity in the soil, which might not always occur in actual field situations. For most nuclides, only sparse data are available and only for selected vegetation types. Even for the most studied radionuclides such as lS7Csand 'OSr, the reported data vary over more than a n order of magnitude. The available data have been compiled in an IAEA handbook (IAEA, 1994).

gThe subscript p, i.e., B,, is used to distinguish the transfer factors for animal fodder from those for vegetation grown for human consumption.

100

/

5. DOSE FROM INGESTED RADIONUCLIDES

Table 5.4 lists some examples of the variability in B, and B, reported in that handbook for selected nuclides for various plant species and soil types. Also shown are the ranges of reported values. A range of uncertainty was estimated in LAEA (1994) (eg., x 10) where insuiEcient data were available to indicate a measured range. The values listed have units of Bq kg-' of dry vegetation per Bq kg-' dry soil. The values can be converted to fresh vegetation units using dry to wet ratios reported in the literature (see Table 5.5). As can be seen, the values for a given nuclide vary by a factor of about 10 or more of the expected value for the same vegetation type. The variation among vegetation types is usually much less (about a factor of two to three generally). Thus, for screening, it is reasonable to use a single conservatively weighted average for all vegetation types. [Note that data for different soil types are available only for the more widely studied nuclides such as lS7Csand 90Srbut indicate that the uptake tends to be substantially higher for sandy and peat soils than for clay soils. Similar data are also available for other crops than shown (MA,1994; See1 et al.,1995)l. The large values shown in Table 5.4 are likely due in part to soil adhesion from resuspension processes as opposed to actual root uptake. However, plants grown in some nutrient depleted soils are known to take up much larger amounts of certain radionuclides from the soil. In particular, plants grown in sandy soils deficient in potassium will take up increased amounts of cesium while plants grown in soils deficient in calcium can take up larger amounts of strontium. The bioavailability has also been found to decrease with time as a nuclide either becomes fixed in the soil matrix or migrates to depths beyond the root zone. However, the bioavailability can also increase with time in some cases as contaminated particles weather ( M A , 1994). Because of the large variability in B , and B, from site to site and with age of contamination and plant type, the uncertainty in this parameter tends to dominate the uncertainty in the ingestion dose from plants contaminated by root uptake. Fortunately, for most radionuclides, the root uptake by plants is quite small. Even though the root uptake of many nuclides is quite low, it has been found empirically that the concentration on unwashed plants averages about 5 to 10 percent of the surface soil concentration. Thus, for many nuclides, the contribution to the vegetation activity from soil adhesion can be much greater than the direct root uptake, particularly for nuclides such as plutonium where the root uptake is very small. The concentration can be as high as 25 percent of the surface soil concentration for very dusty sites, presumably as a result of soil adhesion following resuspension, saltation, rain splash, and

Clay Sandy Clay Sandy All Clay Sandy Clay Sandy Clay Sandy

Other

Roots

Fruit

Grain

Fodder

Grass

0.11 0.24

0.017 0.29

0.010 0.026

0.22

0.040 0.010

0.017 0.094

0.18 0.46

Expected Range

0.011-1.1 0.024-2.4

0.0017-0.17 0.029-2.9

All

0.40-2.9 0.35-7.8

1.1 1.7

76

8.1

All

0.079-7.9 0.1-10

0.022-0.66 0.032-1.4

0.73

0.24

All

0.11-11 0.14-14

All

4.3

All

0.34-4.9 0.53-9.4

0.05-0.8

200

Expected

All

0.74-1.0 0.30-30

Range

Soil Type

0.79 1.0

0.12 0.21

0.2

0.001-0.1 0.0026-0.26

1.1 1.4

1.3 2.2

2.7 3.0

Expected

0.004-0.4 0.001-0.11

0.002-0.14 0.001-0.75

0.019-1.7 0.047-4.5

13'Cs Range

10-760

0.81-81

0.073-3.7

0.1-2.4

10-43

10-2,000

OTc

3.43-04

5.OE-04

8.63-06

9.1E-04

7.3E-05

7.33-05

Expected

kg-' dry vegetation per Bq kg-' dry soil. Range is 95 percent confidence interval (IAEA, 1994).

Clay Sandy

Leafy

"q

Soil Type

Vegetable Type

TABLE 5.4-Examples of variability in soil to vegetation transfer factors."

5E-05-6.5E-01

2E-06-5E-02

3.5E-07-4.2E-01

9.1E-05-9.1E-03

7.3E-06-7.3E-04

7.33-06-7.33-04

Range

\

CI)

3

22

u

0

z

E

2

E

C(

E

102

/

5. DOSE FROM INGESTED RADIONUCLIDES

TABLE5.5 -Average dry / wet (fresh) weight ratios for various food types. Food Type

Till and Meyer"

Leafy vegetables Other vegetables Fruits Grains Forage, hay

0.20 0.25 0.18 0.91 0.22

0.10 0.17 0.17 0.70 0.25

5%

5% 5%

20% 10%

"Till and Meyer (1983) bIAEA (1994)

mechanical disturbance (IAEA, 1994; Pinder and McLeod, 1989; Pinder et al., 1990). Vegetation consumed by humans is generally washed. Thus the radionuclide concentration in the edible portions is unlikely to remain significantly increased due to soil adhesion (IAEA, 1994, Simmonds e t al., 1995). However, the quantity of a radionuclide ingested by an animal as a result of soil adhering to plants, as well as by direct ingestion of soil a t some sites, can be as great a s 25 percent for very dry, sparsely vegetated sites. The IAEA (1994) reviewed estimates of soil adhesion to plants and also available data on direct soil ingestion by animals. Although animals will often ingest soil directly, it is difficult to separate the contribution of direct soil ingestion from the soil adhering to vegetation. However, estimates in the literature indicate that on average an animal on pasture all year long will ingest an amount of soil equivalent to about six percent of its total dry matter intake (Fries, 1987; IAEA, 1994).

5.2.3.1 Effect of Soil Depth Profile on Soil to Vegetation Transfer Factor. For calculating consewative screening doses, the estimate of the root uptake factor assumes that the concentration throughout the root zone is the same as that measured in the top 5 cm soil sample (see Section 6.3.1). If all the radionuclide is close to the surface, the mot uptake for some types of vegetation may be reduced and the actual concentration in the vegetation overestimated. However, the effect of resuspension and subsequent soil adhesion to vegetation, as well a s direct ingestion of soil by animals (or humans), may be enhanced due to the higher than calculated surface soil concentration. For screening purposes, it is thus calculated that the concentration throughout the root zone is the same as the average concentration measured in the soil sample. For many surface soil contamination scenarios, the concentration in the lower portion of the root zone is likely to be less than the average in the top 5 cm

5.2 DISCUSSION OF MODEL PARAMETERS

1

103

and the root uptake estimate is likely to be conservative. Furthermore, the screening calculations also assume that the concentration in the plow layer or root zone will not be reduced by factors other than radioactive decay. However, for nuclides where the screening guidance is driven by the ingrowth of progeny such that the maximum dose occurs sometime in the future, processes such as leaching may well remove much of the original deposition to depths beneath the root zone before the theoretical maximum dose rate is reached. The estimates of B,, and B,, assume that the specific activity of the resuspended soil is the same as the average over the top 5 cm. Often, the resuspended soil may have a higher (enhanced) specific activity as discussed in Section 4.2.2, even if the radionuclide in the top 5 cm is fairly well mixed, particularly if the contamination is fairly fresh. This fact was taken into consideration, as discussed below, in choosing fairly conservative values of B,, and B,
Soil to Vegetation Transfer Factor Values for Screening. The values of B , recommended in NCRP Report No. 123 (NCRP, 1996) have also been used for the screening calculations in this Report. These recommendations, which reflect the conclusions and data presented in IAEA (1994) as well as other recent data, are conservative estimates of the median uptake factor for all crops consumed by humans and for all forage consumed by animals. Table 5.6 gives the values used for selected radionuclides in farm

5.2.3.2

TABLE5.6-Recommended soil to vegetation transfer factors for selected radionuclides (NCRP, 1996). Nuclide

Fresh vegetable^".^ B.

Pastureb.' BP

GSD

"Median (Bq kg-' fresh weight per Bq kg-' dry soil) for entire "vegetable" diet. b T ~times o value shown used for sparse vegetation land-use scenarios (PS, RS) (see Section 5.2.3.2). 'Median (Bq kg-' dry weight per Bq kg-' dry soil) for all types of fodder.

104

/

5. DOSE FROM INGESTED RADIONUCLIDES

vegetables and pasture grass along with an estimate of the uncertainty (GSD). The tabulated values were multiplied by a factor of two for the sparsely vegetated sites (PS, RS)to reflect the possibility that the sparse vegetation could be due to a lack of nutrients in the soil or to a more sandy soil. The complete set of parameter values for all elements is given in Appendix D. Even though the chosen values are conservative, an uncertainty represented by a GSD ranging from 2.5 to 3 was assigned to each estimate. The lowest GSD was generally assigned to elements for which fair amounts of good experimental data are available. A value of 2.7 was assigned for elements for which only a small amount of data are available or the estimates are based on mathematical models. The highest uncertainty (GSD = 3) was assigned to elements for which few or no data or estimates are available (see Appendix Dl. Because of their higher variability with type of soil and soil nutrients, a value of 2.7 was also used for 137Csand 90Sr.These estimated GSDs are somewhat lower than the values of 3.5 t o 4 for I3?Cs and 90Sr quoted in NCRP Report No. 76 (NCRP, 198413) as representative of the entire range of variation about the median for all crops. However, the central tendency parameter values adopted here for calculating screening doses are already conservative (close to the top of the measured range for most nuclides) and an even higher median value is used for sparsely vegetated soils. Thus, the GSDs used here should reasonably reflect the upper range of possible values. The estimates for B,, and B,, used for calculating the effect of soil adhering to vegetation or directly ingested by animals along with an estimated uncertainty (possible range) are given in Table 5.7. The estimates for vegetation for human consumption a r e much TAE~LE 5.7 -Transfer factors from soil to vegetation due to soil adhesion used for calculating screening doses." Fresh Vegetable, B,, Land Use

Agricultural Heavily vegetated pasture Sparsely vegetated pasture Heavily vegetated rural Sparsely vegetated rural Suburban

Dry Fodder, B,,

Estimate

Range

0.001

x 10

-

-

-

0.05 0.1 0.05 0.1

~3 x3 x3 x3

-

-

-

0.001 0.004 0.001

x 10 x 10 x 10

Estimate

Range

"Bq kg-' fresh weight per Bq kg-' dry soil for fresh vegetables, Bq kg-' dry weight per Bq kg-' dry soil for fodder. A range of three is assumed to correspond to a GSD of 1.4 and a range of x 10 to a GSD of 2.2.

5.2 DISCUSSION OF MODEL PARAMETERS

1

105

smaller than those for animal fodder since most vegetables will be washed or the nonedible portions containing much of the adhering soil will be discarded. Furthermore, ingestion of soil by individuals is also calculated separately in a conservative manner (see Section 5.3). The values assigned for animal fodder have been conservatively chosen to reflect the fact that animals will ingest soil directly, particularly on sparsely-vegetated sites, and that the resuspended soil may be enhanced in specific activity compared to the measured soil specificactivity (see Section 4.2.2). Even so, a fairly high uncertainty is assigned to each parameter value estimate to reflect the large potential variability from site to site and from one type of vegetation to another, as well as the variability in resuspended soil specific activity (enhancement factor). values were used for sparsely vegetated sites Higher B,! and compared to highly vegetated sites to account for possible greater resuspension and redistribution, and thus greater soil adhesion, at such sites. The higher values also reflect a greater likelihood of direct soil ingestion by animals for sparsely vegetated sites as well as a higher enhancement fador for resuspended soil (see Section 4.2.4). Although the amount of soil ingested by animals can be quite high and can vary widely, E(z) calculated for screening based on the transfer factors in Table 5.7 should be conservative since the screening calculations assume that the radionuclides in the ingested soil are absorbed from the animal's GI tract as efficiently as nuclides incorporated in plants (see Section 5.2.7.4). In actuality, the bioavailability is likely to be lower (ICRP,1993). All the soil to vegetation transfer factor estimates used in this Report for screening were chosen to be sufficiently conservative to allow for variations with time and soil typercondition. The actual transfers fiom soil to vegetation for a particular site may well be much lower, particularly if the radionuclide has had time to become stabilized in the soil. Site-specific assessments may be able to use less conservative values based on information regarding soil type and composition. B,l

5.2.4 Animal Diets

As with human diets, animal diets vary widely, depending on the type of pasture, type of animal, etc. IAEA (1994)estimates that milk cows consume an average of 16 kg d-I of dry feed, while beef cattle consume only about 7 to 8 kg d-'. Other domestic animals consume

106

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5. DOSE FROM INGESTED RADIONUCLIDES

lesser amounts of feed per day (see Table 5.8). For milk and dairy cows, the fraction of the diet from fresh pasture as opposed to stored feed can vary widely depending on the season of the year, location and pasture practices. On average, milk cows obtain about 75 percent of their feed from fresh pasture during the pasture season and are on pasture about 150 d y-I (IAEA, 1994). The pasture season length varies considerably, however, with location (climate) and local practice. Beef cattle are usually fed about 25 percent stored hay and 75 percent stored grain and consume only about half the total feed of the average milk cow (Kennedy and Strenge, 1993).Generally, fresh pasture consists of grass, grain or hay, etc. Site-specific assessments should include the actual diet, fraction of feed from fresh pasture (or grown on the contaminated site) and pasture season length appropriate to the site. For screening, it is prudent t o choose conservative values.

TABLE 5.8 -Animal diet for

screening dose calculations. Land-Use Scenario

Percent of total diet from site Milk cows 75 range 60-90 Beef cattle, game 75 range 60-90

75 60-90 66 50-80

Total dry matter intake (kg d-I)'

Milk cows

16

range Beef cattle range

8-25 8

4-12

Total dry matter intah fir other animals (kg d-I)" Pigs 2.4 range 2-3 Calves 1.9 range 1.5-3.5 Lamb 1.1 range 0.5-2.0 Chicken 0.07 0.05-0.15 range "See IAEA (1994)for references.

75 60-90 25 15-35

50

30-70 15 10-20

5.2 DISCUSSION OF MODEL PARAMETERS

/

107

The average diets assumed for milk and beef cows for the screening dose calculations along with a n estimate of uncertainty are listed in Table 5.8. For screening, i t is conservatively assumed that all animals are on pasture continuously. Only milk cows and beef cattle diets are used in the screening calculations although parameter estimates for other animals are given in the table. The fraction of the animal's diet from fresh pasture or stored feed grown on the contaminated site is assumed to vary with land use. However, the fraction of the animal's total diet derived from the site is assumed to average 4 , since even animals on pasture year round would likely receive some supplemental feed. This is particularly so for animals raised for meat production on sparsely vegetated sites. For rural sites, the parameter estimates for meat are intended to reflect the fraction of feed that wild game might obtain from feeding on the contaminated site. The large uncertainty assigned, particularly for pasture sites and for milk cows, is intended to reflect the possibility that the more exposed animals could receive stored feed that was grown previously on the site. Therefore, most of their diet could derive from the contaminated land. For the purposes of the screening guidance in this Report, the pasture site scenarios should also be used for contaminated land used to grow fodder for animal consumption even if the animals do not graze directly on the site.

5.2.5

Meat, Milk Transfer Factors

As was the case for the transfer from soil to vegetation, the transfer factors relating a given intake by an animal and the concentration in meat (or milk) exhibit considerable variability. The available data on feed to meat (or milk) transfer factors have been summarized in IAEA (1994). The published data assume conservatively that the activity in the meat (milk) has attained equilibrium with the activity in the feed, i.e., that milk and meat are produced continuously, and that the animals are on contaminated feed year round. The variations within an animal type are generally smaller than the differences from one animal type to another, as shown in Table 5.9 for a few selected elements. IAEA (1994) and Simmonds et aZ. (1995) discuss the difficulty in applying these transfer factors under the assumption of equilibrium having been attained. Metabolic homeostasis, variations in gut uptake with the chemical form of the radionuclide, and lack of information for young animals all influence variability noted in the literature for the transfers from feed to meat or milk.

108

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5 . DOSE FROM INGESTED RADIONUCLIDES

Tmm 5.9-Examples of variability in relating a given animal intake to concentration in milk (d L-I) and transfer factors for meat (d kg-') for selected radionuclides." Nuclide

Sr-90

Beef

best 0.008 min 0.0003 max 0.008 Cs-137 best 0.051 min 0.015 max 0.056 Ru-106 best 0.051 mm 1E-4 max 0.051 Pu-239 best 2E-5 min 2E-7 max 2E-4

Lamb

Goat Milk

Eggs

Milk

0.33

0.18

-

0.18

-

0.64

0.003 0.028 0.001 0.006 0.003 0.039

Pork

Fowl

Veal

0.49 0.10 1.6 1.5 0.01 1.5 0.003 -

-

"ata from IAEA (1994). As can be seen, the transfer per unit intake for some nuclides is generally higher for younger animals, i.e.,lamb, veal. However, these higher transfer factors are balanced by the lower feed intakes for young animals (see Table 5.8). The transfer to goat's milk is particularly high, especially for nuclides such as the iodines. Site-specific dose assessments should calculate separate doses for each type of animal that provides meat or milk for human consumption and that receive significant portions of the diet from vegetation grown on the site. Transfer factors appropriate to the site should be used, and corrections should be made if the animals are not on pasture continuously (see IAEA, 1994; Kennedy and Strenge, 1993). For screening, we assume all the meat consumed by humans is beef since beef is generally consumed in greater quantities in the United States than is lamb, veal, pork or poultry (see Section 5.2.2.1). Furthermore, cows are more likely to ingest forage directly grown on contaminated soil than are other animals. The possible error due to the different feed-to-meat transfer factors for poultry, sheep or other animals compared to beef is discussed below. The FmihVmeat values used for calculating screening doses are listed in Table 5.10 for some of the most common radionuclides. The variability in these transfer factors is assumed to be lognormal, and an estimate ofthe GSD is also given. For most elements, these prudently

6.2 DISCUSSION OF MODEL PARAMETERS

1

109

TABLE5.10- Transfer factors used for screening calculations relating a given animal intake to concentration in milk and meat for selected radionmlides. Nuclide

F ~ ( L-I) d

GSD

F-dd

kg-1)

GSD

conservative estimates for beef and cow milk are identical to those recommended by the NCRP (1996) for use in screening for airborne releases. In order to account for the possible large variations, including the f a d that pork, veal, lamb or poultry may constitute a substantial fraction of the diet of some individuals, a fairly conservative nuclide-dependent uncertainty estimate (GSD)was assigned to each of the chosen values based on the data reviewed in IAEA (1994). The GSDs assigned for transfer to meat range from 1.2 to 2.8. The lowest value, GSD = 1.2, was assigned to elements such as calcium, potassium, phosphorus, sodium and magnesium whose uptake is controlled by homeostatic processes. A GSD of 1.5 to 2 was assigned for well-studied nuclides such as lS7Csand 90Sr.Higher GSDs were assigned to elements for which fewer data are available or for which the transfer factor varies more widely with age or type of animal. The highest values were assigned for elements for which few actual data exist. For transfer to milk, the GSDs range from 1.6 to 2.5. Again, the lowest values are for the most widely-studied nuclides and the highest for elements for which few data are available. The GSDs used here are somewhat lower than the range quoted in the literature. For example, in NCRP Report No. 76 (NCRP, 1984b)the GSD for F,,, for 13'Cs was estimated to be about 2.3. However, as was the case for the soil to vegetation transfer factors, the values chosen for calculating screening doses are already considered to be conservative, i.e., biased toward the upper end of the distribution, in keeping with the screening philosophy of calculating doses to the most exposed population. Even though the lack of good data for many nuclides required that these choices be based on the best judgment of the experts preparing this Report and NCRP Report No. 123 (NCRP, 1996), the GSDs assigned should adequately reflect the

110

1

5. DOSE FROM INGESTED RADIONUCLIDES

actual upper limit of the true distribution for even the most exposed population groups. Appendix D lists the transfer factors to beef and milk chosen for each element along with the assigned uncertainty estimates. 5.2.6

Decay Correction for Delay from Harvest to Consumption

For site-specific assessments, decay corrections should be made for the delay between harvest and ingestion of contaminated food. Corrections should also be made for decay for stored fodder used to feed animals. This is only important for the shorter-lived radionuclides (TI&< 60 d). The actual hold-up time will vary depending on whether the food is consumed onsite or shipped offsite. Table 5.11 gives some recommended delay estimates (ti - to) and associated estimated uncertainty (range) for use in site-specific assessments. Site-specific corrections should account not only for decay of the parent radionuclide, but also any buildup of progeny. Since these delay corrections are significant only for vegetables and meat, and only for the shorter-lived nuclides considered in this Report, the screening calculations in this Report neglect this decay and buildup during the hold-up period. The lack of a correction for hold up is conservative, particularly when considered in the light of the conservative assumptions regarding the concentration in the soil a t harvest and the assumption that crops are produced year round. 5.2.7

Committed Effective Dose Factors

The committed doses to particular body organs and the resultant E(z) for an intake of 1Bq of a particular nuclide have been calculated for a Reference Man for over 800 nuclides (ICRP 1993; 1994b; 1995a;

TABLE5.11 -Recommended hold-up times, (ti+,) for calculating ingestion doses. Food Type

Leafy vegetables Other vegetables Fruits Grains Meat Milk

Average Delay, Days Between Harvest and Consumption

1

14 14 14 20

1

Range, Days

0- 14 0-30 0-30 0-60 0-60 0- 7

5.2 DISCUSSION O F MODEL PARAMETERS

1

111

199513; 199613). These calculations are based on biokinetic models adopted by the ICRP (1993; 1995b) for both adults (Reference Man) and children. The effective dose factors include the dose resulting from ingrowth of decay products after ingestion of the parent nuclide and use ICRP Publication 60 tissue weighting factors. Thus, one need only to calculate the concentration of the parent nuclide a t the time of ingestion in order to use the dose factor to solve Equation 5.1. Individual organ dose estimates are reported in ICRP Publications 67 and 69 (ICRP, 1993; 1995b) while a summary of total committed effective ingestion doses as a function of age are reported in ICRP Publication 72 (ICRP, 1996b). Similar calculations were published previously for a more limited number of nuclides by the NRPB (Phipps et al., 1991). These compilations sometimes provide a range of estimates for some nuclides, using differing GI uptake factors, since the g u t uptake usually depends on the chemical form ingested. 5.2.7.1 Dependence of Dose Factor on Age. Table 5.12 shows the

ratio of infant (1y) and child (10 y) E(T)to that for an adult (20 y) reported by the ICRP for some selected nuclides. In all cases, the committed dose is calculated to age 70. Note that for many nuclides, E factor for children or infants per unit intake is sigdicantly higher than for adults. Ingestion doses were first calculated for adults using Equation 5.1 with parameters appropriate for adults. The dose to a child or infant was then calculated separately as discussed in Section 5.4. 5.2.7.2 Dependence of Dose Factor on Gastrointestinal Uptake. The biokinetic models used to estimate ingestion dose contain a parameter (f,) which represents the fraction of the nuclide transferred to the blood from the gut (ICRP, 1979-82; 199613). The appropriate value for fi will vary depending on the chemical form of the radionuclide ingested. Table 5.13 shows E(T)reported by the ICRP for a few important selected nuclides for different assumed gut uptake factors. As can be seen, the dependence of E on GI uptake can be very significant. E does not always scale directly as the GI uptake since, for some nuclides with low uptake and short half-lives, such as 95Zr, the dose to the relatively radiosensitive GI tract itself may control E calculation. Values for fi as a function of age have been recommended for members of the public for use in site-specific assessments by the ICRP (1996b) for environmental compounds of unspecified chemical composition. Some examples are listed in Table 5.14.

Uncertainty in Biokinetic Models. NCRP (1998) has reviewed the uncertainty in the biokinetic models used to determine

5.2.7.3

112

/

5. DOSE FROM INGESTED RADIONUCLIDES

TABLE 5.12-Ratio of infant to adult, child to adult ingestion E(T)

for selected radionuclds." Nuclide Fe-55 Co-60 Zn-65 Se-75 Sr-90 Zr-95 Nb-95 Tc-99 Ru-106 1-129 CS-134 CS-137 Ce-144 Ra-226 U-238 Pu-239 Am-241

InfanVAdult

ChildIAdult

7.3 7.9 4.1 5.0 2.6 5.9 5.4 7.5 7O . 2.0 0.8 0.9 7.5 3.4 2.7 1.7 1.9

3.3 3.2 1.6 2.3 2.1 2.0 1.9 2.0 2.1 1.7 0.7 0.8 2.1 2.9 1.5 1.1 1.1

"Data from ICRP Publication 72 (ICRP, 199613). See Appendix C for a complete listing for all covered radionuclides. The infant values are for age 1 y while the child values are for age 10 y. The values are ratios of dose factors (Sv Bq-I). The ratios of dose will generally be lower due to the smaller quantities ofvegetablesand meat consumed by children and infants compared to adults (see Section 5.4). ingestion dose factors. This uncertainty includes the uncertainty in the gut uptake factor, f i . The estimated uncertainty in the dose factor for application to the general public for some selected nuclides is summarized in Table 5.15. Because of the number and complexity of the assumptions needed to calculate the dose factors and the lack of adequate biokinetic data for many nuclides, the uncertainty estimates given in Table 5.15 are based on expert judgment. The uncertainty given includes uncertainty in the GI transfer factor and other factors used in t h e biokinetic models. The column labeled "other populations" reflects the larger uncertainty in model parameters for infants or children or for groups suffering from certain diseases. The dose factor calculations require accurate data regarding the half-lives and decay products of each nuclide and its progeny. The ICRP calculations utilize the best available data, and the uncertainty in the dose factors due to uncertainty in the physical data is, for most

5.2 DISCUSSION OF MODEL PARAMETERS

1

113

TABLE5.13-Adult E(z) per unit intake, Sv Bq-I, versus GI uptake ( f l )for selected radwnuclides."

f1

E(T)

0.8 (unspecified compounds) 0.05 (elemental and selenides)

2.63-9 4.13-10

0.01 (SrTi03) 0.3 (unspecified compounds)

2.73-9 2.83-8

1.0 (methyl mercury) 0.4 (unspecified compounds)

1.93-9 1.13-9

5E-4 (unspecified compounds) 2E-4 (oxides, hydroxides)

2.23-7 9.23-8

0.02 (unepecified compounds) 0.002 (tetravalent compounds)

4.53-8 7.63-9

1E-6 (insoluble oxides) 1E-4 (nitrates) 5E-4 (unspecified compounds)

9.03-9 5.33-8 2.53-7

Nuclide

Hg-203 (organic)

Pu-239

"Data from ICRP Publication 72 (ICRP, 1996b). Indicated f l values (fraction of nuclide transferred to the blood) are recommended for use with chemical compounds shown in parentheses.

TABLE 5.14-GI uptake factors (fl) recommended for use in calculating ingestion closes to members of the public (age > 1 y) for selected elements." Element

fi

fI

Element

Cobalt

0.1 (adults) 0.3 (age 1-15 y)

Antimony

0.1

Strontium

0.3 (adults) 0.4 (age 1-15 y)

Iodine

1.0

Zirconium

0.01

Cesium

1.0

Niobium

0.01

Lead

0.2 (adults) 0.4 (age 1-15 y)

Technetium

0.5

Neptunium

5E-4

Ruthenium

0.05

Americium

53-4

"Data from ICRP Publication 72 (ICRP, 1996b) for unspecified chemical composition.

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5. DOSE FROM INGESTED RADIONUCLIDES

TABLE5.15- Estimated uncertainty in ICRP ingestion dose factors." Estimated Range Radionuclide

Fe-55 Co-60 Se-75 Sr-90 Zr-95 Nb-95

Ru-106 Sb-125 CS-137 Ce-144 Pb-210 Po-210 Ra-226 U-234 Np-237 Pu-239 Am-241

Adult Male

Other Populationsb

5

10

10

20

5 3

10

10

5 20

5

10

10 10 2

20 20 5 30 20

10 10 10

5

10 10

10 10 10 10

20 20 20 20

"Adapted from NCRP (1998).The estimated range, R, can be interpreted as indicating that the dose factor for some individuals may be as much as a factor of R higher or lower than the dose factor recommended by the ICRP. bSpecialpopulation groups generally consisting of diseased people, or of infants or children. radionuclides, minor compared to the uncertainty and variability in the biokinetic factors. 5.2.7.4 Recommended Ingestion Dose Factor Values for Screening. In this Report, ingestion dose factor values based on the current ICRP biokinetic models (ICRP, 1993; 1995b; 1996b) have been used for calculating screening doses (see Appendix C). If values were given for more than one GI uptake factor, the value used for the screening dose calculations was the one recommended by the ICRP (1996b) for members of the general public of a particular age. 5.2.7.5 Dose Factors for Site-Specific Dose Assessments. Sitespecific dose assessments should consider the age distribution of the exposed population. This is particularly true when ingestion of milk is important. For some nuclides, the dose factor for infants < l y old is even higher t h a n those for t h e 1y old. ICRP (199613) gives dose factors for a wider range of ages than used for the screening

115

1

5.2 DISCUSSION OF MODEL PARAMETERS

calculations in this Report. For site-specific dose assessments, it would also be appropriate, if possible, to choose dose factors where the assumed GI uptake is based on the true form of the radionuclide likely tc be ingested. If infants andlor children do not ingest food produced on the site (particularly milk), a site-specific assessment may result in significantly lower estimated doses than the screening doses calculated here. Correction factors for adult dose factors are given in Table 5.16 to correct for age dependence. 5.2.7.6 Organ Dose versus Effective Dose. For site-specific assessments, it may be useful for some nuclides to also use particular organ dose estimates as well as E in making judgments, particularly when a given organ is particularly at risk. The compilations cited provide individual organ committed doses as well as E(z).Doses to the most exposed organ are compared with E(2) in Table 5.17 for some selected nuclides. Note that the critical organ doses for some nuclides are almost 50 x E(T). TABLE 5.16-Ratio of average annual intake used in screening dose calculations to correct adult dose factors for age dependence. Food Group

InfanVAdult

Vegetable, fruit, grain Meat Milk Soil

0.34 0.3 1.1 --see

ChildIAdult 0.65 0.7 1.0 Table 5.19-

-

TABLE 5.17 -Ratio of most exposed organ dose to E(T)for selected radionuclides." Nuclide

Organ Dose 1 E(T)

Critical Organ

Co-60 Sr-90 Zr-95 Tc-99 1-129 Cs-137 Ra-226 Pu-239

4 15 8 6 19 1.2 47 33

Lower large intestine Bone surface Lower large intestine Lower large intestine Thyroid Lower large intestine Bone surface Bone surface

"Data from ICRP Publication 67 (ICRP, 1993).

5.2.8

Uncertainty in Ingestion Dose Factors Chosen for Screening Calculations

In order to estimate the uncertainty in the final screening doses, a GSD ranging from 1.25 to 2.5 was assumed to represent the uncertainty in the ingestion dose fador Df,. These GSDs (see Appendix C) were chosen by assuming that the uncertainty estimates (ranges) for adult males based on NCRP (1998), shown in Table 5.15, correspond to about the (GSD)S.The lowest GSDs were used for 13'Cs and 'Sr whose biokinetic behavior in the body has been studied much more extensively. A GSD of 1.7 was assigned to isotopes of iron, selenium, niobium and radium, while a value of 2 was assigned to the other nuclides listed in Table 5.15. A value of 2.5 was assigned to nuclides not specifically evaluated by the NCRP. Although no additional uncertainty was used for calculating the doses to "children," the assigned GSDs are believed to be sufficiently conservative since the new ICRP age-specific dose factors were used for the screening dose calculations. The increased uncertainty due to age dependence implied by Table 5.15 is explicitly corrected for by the use of the IIA and C/A dose factor ratios discussed above.

5.3 Direct Ingestion of Soil There are numerous pathways by which humans can be exposed to contaminants in the environment and by which their bodies can become contaminated with chemicals, toxins or radioactivity that are resident in the soil. It is also well known that young children are particularly susceptible to direct intake of soil-borne substances whether it be pesticides, bacteria or radioactivity. During infant years, children normally exhibit mouthing behavior whereby their hands are repeatedly moved in and out of their mouth during the day; in some cases, young children and even adults intentionally ingest soil. Some previous reports which have recommended limits for actinide contamination of soil (e.g., Healy, 1974; EPA, 1990b; 1990~) have superficially considered soil ingestion by humans, but this pathway has not been comprehensively evaluated in the open literature for purposes of radiation risk assessment. In addition, little empirical data has been published from which ingestion rate and other related model parameter values can be derived for specific occupations, lifestyles, environments or different ethnic/cultural groups. This Section reviews some of the available literature on soil ingestion which has reported a variety of types of data including qualitative observations, quantitative studies of soil ingestion,

6.3 DIRECT INGESTION OF SOIL

1

117

estimates of intake based on a variety of assumptions and indirect measurements, and literature reviews. Considerations for modeling the direct intake of soil, whether it be intentional or inadvertent, are also discussed for a variety of lifestyle scenarios. An examination of the critical group is also provided. Intake rates are recommended for use in calculating screening doses. The circumstances of soil ingestion can be divided into those in which soil is ingested (1)inadvertently by swallowing dirt or dust, (2) via food items contaminated with soil (discussed earlier), (3) from mouthing of dirty hands or other contaminated nonfood items, or (4) intentionally, i.e., deliberate consumption of clay, soil or any earthen materials. Though we briefly discuss intentional ingestion, for the purpose of calculating screening exposures, the very rare cases of intentional ingestion of large amounts of soil are not considered. The phenomenon of soil ingestion has been discussed in the literature with various clinical and public health endpoints in mind. Most literature which has reported on the public health concerns of soil ingestion are focused on the health implications which arise due to the intake (via soil) of a variety of nonradioactive environmental toxins. Considerable attention, for example, has been drawn to positive correlation between soil intake, anemia and pregnancy. The literature has generally emphasized intentional rather than inadvertent soil ingestion because those cases generally have higher morbidity rate and more severe clinical endpoints. It is useful in evaluating the available literature to understand the terminology used to differentiate intentional from inadvertent soil intake. In particular, the phenomenon of purposeful soil ingestion has received its own terminology. Taber's Medical Dictionary (Thomas, 1993) defined pica to be: "An eating disorder manifested by a craving to ingest any material not fit for food, including . . . clay. . . ." Lanzkowsky (1959) also defined pica to be a perversion of appetite with persistent and purposeful ingestion of unsuitable substances (also see Danford, 1982). Geophagia is more specific and was defined to be: "A condition in which the patient eats inedible substances, as chalk, clay or earth" (Thomas, 1993). Most authors agree with Reid (1992) that geophagia or earth eating is a special case of pica. Halsted (1968) defined geophagia as the habit of eating earth including clay and other types of soil, whereas pica means the eating of any foreign substance and thus includes eating of soil. Barltrop et al. (1974) defined soil pica subjects to be those children who habitually put fingers or toys in their mouths while playing in their gardens as well as putting soil directly in the mouths and included children who had pica for other substances in addition to

118

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5. DOSE FROM INGESTED RADIONUCLIDES

soil. Calabrese et al. (1991) defined soil pica as ingestion of soil in amounts far exceeding those observed in the average child. In this Report, the terms "inadvertent ingestion" and "geophagia" will be used to distinguish between accidental and intentional ingestion of soil. The definition of the critical group for soil ingestion is important because, in general, intake rates are assumed as age and sometimes gender dependent. Depending on the lifestyle scenario to which the screening calculation applies, the critical group is likely to be children -because of likely higher intake rates, pregnant women- because of historical associations with geophagia among culturally disadvantaged groups, or adult men working in occupations which routinely contact the soil.

6.3.1 Factors of Znadvertent Ingestion a n d the Etiology of Purposeful Intake Radiation risk to individuals is normally assumed to be a function of the exposed person's age, their intake rate of contaminated substances, and their personal biokinetics, if they are known. However, for population-based risk evaluations or for screening calculations for the typical population, the factors of casual soil intake are relevant t o assessing the likelihood of soil intake and, hence, to defining acritical group for scenarios in which soil ingestion might be a significant contributor to radiation dose. Current knowledge does not yet allow precisely predicting the likelihood of soil ingestion or quantitative intake rates among various occupations or socioeconomic settings that are at risk, yet, an understanding of those factors which appear to be causal may be helpful for that purpose.

6.3.1.1 Znadvertent Intake. Because inadvertent soil ingestion does not quallfy a s a disease, it has no associated etiology. However, there are environmental conditions which may enhance the likelihood and degree of accidental intake. In particular, inadvertent soil ingestion is enhanced by the same conditions which encourage inhalation of dust: dry, dusty climates; dusty dwellings; open air dwellings; primitive or dry living conditions in which little access to washing of foods and the body is possible; living conditions in which children may play on the floor; and occupations which involve direct contact with soil or moving soil by mechanized means (ATSDR, 1992). 5.3.19 Geophagia. Because geophagia is considered an aberrant behavior, i.e., a mental disease or compulsion (Hattis and Burmaster, 1994) or can result in damage to the health of individuals, it has an

5.3 DIRECT INGESTION OF SOIL

1

119

associated etiology. The etiology of this habit continues to draw the attention of some medical practitioners, public health specialists, psychologists, and anthropologists. Extensive literature before 1970 discussed the observed relationship between soil pica and iron deficiency. The theory that iron deficiency was a casual factor for soil ingestion was mainly derived from uncontrolled clinical observations. This theory was eventually refuted (see for example, Ansell and Wheby, 1972) and it was concluded that iron deficiency was a statistical rather than casual relationship. Some of the factors suspected of being involved in the etiology of soil pica were divided by Mitchell et al. (1977) into physiological and psychological factors. Although typically associated with urban andlor poor living conditions in the southern United States, a noticeable incidence of geophagia was documented as late as 1970 in New York City (Roselle, 1970 cited by Gelfand et al., 1975).Crosby (1976)and Wedeen et al. (1978) also note that the cultural tradition of eating clay has recently been transplanted to northern metropolitan areas and is not limited by socioeconomic or racial background. More recently, general pica and geophagia were reviewed by PanyJones and Parry-Jones (1992) and Reid (1992), respectively. ParryJones and Parry-Jones (1992) discussed a wide variety of historical interpretations of pica and they conclude that in the developed world, pica is now more rare and occurs mostly among infants and young children, particularly in low socioeconomic groups. Those groups at higher risk can be either children or adults of either sex but mainly are those in poor, rural areas as well as the adult groups of pregnant women and mental defectives. However, they also note that the habit or phenomenon, which is culturebound in Third World countries, still persists. Reid, who discussed cultural and medical aspects of geophagia, provides documentation that anemia is associated with, but not the cause of, geophagia. In some cultures, those groups who are a t higher risk for anemia, e.g., pregnant women, are culturally sanctioned and encouraged to eat clay. The phases of life when geophagia is most likely practiced, i.e., childhood and periods of growth, pregnancy and lactation (Lanzkowsky, 1959;Vessal et d.,1975),are the same periods when radiation exposure may have the most serious consequences. This reason is probably sufficient for including geophagia in radiation risk calculations in groups of low socioeconomic status or in undeveloped countries. However, as earlier stated, this practice is considered highly unusual for typical United States' populations and thus is not explicitly considered in the screening guidance provided here.

120

1

5.3.2

Review of Literature on Soil Intake

5. DOSE FROM INGESTED RADIONUCLIDES

Soil ingestion, whether inadvertent or intentional, is difficult to verify and quantify. In particular, those instances of inadvertent intake are, by definition, accidental and likely not to be observed. However, by the same arguments that it is known that dusty environments result in inhalation of dirt, it can be reasonably assumed that these environments can also lead to swallowing, i.e., ingestion of soil. Inadvertent soil ingestion can only be observed and, hence verified under unusual circumstances. Both types of intakes have been discussed in the literature though little empirical data is available for either. There are many dozens, if not hundreds, of citations in the literature which document qualitative and semi-quantitative observations of soil intake (Simon, 1998).Several previous reviews of historical literature have been published (e.g., Cooper, 1957;Halsted, 1968; Laufer, 1930). Most of the studies cited refer to intentional intake, i.e., geophagia and apply to lifestyle scenarios encountered infrequently in the United States. There is a paucity of studies which have been specifically designed to determine soil ingestion rates; yet, a number of publications have reported quantitative values of intake rates. A summary of published values is provided in Table 5.18 from which recommended intake rates are derived (Table 5.19). Only four rigorously conducted empirical studies to quantlfy soil ingestion are noted in the English literature: Binder et al. (1986), Calabrese et al. (1989), van Wijnen et al. (1990), and Davis et al. (1990). These studies focus on inadvertent intake and probably provide ingestion rate estimates that are likely suitable for typical United States or western populations. These four studies do not address, however, the special situations of high-risk occupations,primitive living conditions, or indigenous peoples living in a subsistence economy; furthermore, these studies address only to a limited degree the child or adult who intentionally eats soil. The empiricallybased studies are not without their own problems. In particular, widely varying intake rate estimates for the same group have been observed depending on the trace element measured. For example, Binder et al. (1986), in one of the first carefully conducted quantitative investigations of soil ingestion, studied 59 children, l to 3 y of age, in East Helena, Montana. Their estimates were based on the results of fecal sampling for soil trace elements, although they varied widely depending on the trace element studied. The soil ingestion estimates (based on fecal analysis) were 181 and 184 mg d-I using aluminium and silicon tracers, respectively, and 1,834 mg d-I using titanium. The other three empirical studies also exhibited the same problem to some degree.

A

U

Fisher et al. (1981)

G

40 y old American Indian woman, ingestion rate estimated, confumed from fluoride analysis of bone biopsy

Estimate of ingestion of street dust

C

U

NAS/NRC (1980)

I

CIA

R

Vermeer and Frate (1979)

Nutritional questionnaire to determine consumptio?l of clay

G

A

R/U

Wedeen et al. (1978)

Theoretical assessment calculations for mid-west U.S.environment (Mound Laboratory in Ohio) Theoretical assessment calculations for exposure to transuranic radioactivity 46 y old black woman with lead poisoning and anemia

Study Summary1 Analysis Methods

G

G

C

R

Healy (1977)

I

A

R

(m)

Rogers (1975)

Reference

Rural or Urban

Geophagia or Child Inadvertent or Adult Ingestion (CIA) (GA)

typically 50 g d-'

40 mg d-'

E, IM

E

E

10 to 150 g d-', average of 50 g d-I

2-4 g d-"or

E

A

12 yb

20 g d-'

8-11 mg d-'

Basis for Ingestion-Rate Estimates Ingestion-Rate Estimates (A, E, IM, M, LRIa

geophagia), ordered by publication date (Simon, 1998). (continued)

TABLE 5.18-Selected literature reporting quantitative estimates of soil ingestion-rate (inadvertent and

2

2

g

z

m

%

?'

N h3

w

Fecal analysis for thorium in stool samples collected &om 2 farms in Pocos de Calda plateau of Brazil Fecal analysis for soil trace elements for 18 nursery school children compared to 6 hospitalized children who did not play outside Compilation of literature

I

I

I

CIA

C

CIA

U

U

Not specified

Linsalata et al. (1986)

Clausing et al. (1987)

LaGoy (1987)

Fecal analysis for soil trace elements of 59 Montana children aged 1-3 y

I

C

U

Binder et al. (1986)

Theoretical assessment calculations for exposure to TCDD contamination

I

CIA

U

Kimbrough et al. (1984)

181 mg d-I (All 184 mg d-I (Si) 1,834mg d-I (Ti)

9-18 mo: 1 g d-I 18-42 mo: 10 g d-I 42 mo to 5 y: 1g d-' > 5 y: 100 mg d-I

M

E

?

U

U

Calabrese et al. (1989)

Calabrese et al. (1990)

(m)

Wong (19881, Wong et al. (1988)

Reference

Urban

Rural or

A

C

C

I

I/G

IIG

Study Summary/ Analysis Methods

6 adults studied as part of a validation program in a child soil ingestion study (Calabrease et al., 1989)

Fecal analysis for soil trace elements during 8 d period for 65 children, 1-4 y of age in greater Amherst, MA area; one soil pica child identified

- 50 mg d-'

non-pica, 5-8 g d-I for pica child

9 to 40 mg d-I for

M

M

M

Basis for Ingestion-Rate Estimates Ingestion-Rate Estimates (A, E, IM, M. LRY

25% > 100 mg d-I Observations made for 4 14% > 200 mg d-I months in 2 government 4% > 300 mg d - L C supervised homes in Jamaica for children awaiting foster home placement, analysis by fecal sampling for silica

Geophagia or Child Inadvertent or Adult Ingestion (CIA) (GA)

geophagia), ordered by publication date (Simon, 1998). (continued)

TABLE 5.18-Selected literature reporting quantitative estimates of soil ingestion-rate (inadvertent and

u

m

el

2 m

2

8

z

5: m

VI

ZI

P

E3

R/U

U

van Wijnen et al. (1990)

Calabrese et at. (1991) C

C

G

0 to 90 mg d-I (geometric mean) for day care groups, 30 to 200 mg d-' for camping groups

10 g d-I for all age groups (noted as unrealistic for 3 mo old infants)

200 mg d-I for children 100 mg d-' for adult

39 mg d-I (Al) 82 mg d-I (Si) 245 mg d-I (Ti)

Fecal analysis for soil trace 10 to 13 g d-I during elements for one child (3.5 2nd of 2-week y, female) of 64 (see observation period Calabrese, 1989) who displayed pica during 2nd of two week observation period

Fecal analysis for titanium of two different groups in the Netherlands during summer compared to hospitalized children

I

C/A

R

Haywood and Smith (1990)

Based on observation of aboriginal population and estimates of consumption of soil contaminated water and other sources

Regulatory guidance values

I

CIA

R/U

EPA (1989a; 1989b; 1989c; 1990aj

I/G

Fecal analysis for soil trace elements of 104 normal school children randomly selected, 2-7 y of age in S.E. Washington state

I

C

R/U

Davis et aE. (1990)

M

M

E, A

LR, A, E

M

U

Finley and Paustenbach (1994)

C/A

C

U

de Silva (1994)

I

I

I/G

C/A

I

C

R/U

U

(R/U)

ASTDR (1992)

Thompson and Burmaster (1991)

Reference

Rural or Urban Study Summary1 Analysis Methods

Fitted child data of Calabrese and Stanek (1992) to parametric (lognormal) distribution

Blood levels of lead from Barltrop (1979) and Barltrop et al. (1975) used to determine ingestion-rate of lead contaminated soil

Regulatory guidance values

Re-analysis of data of Binder et al. (1986) using actual stool weights, fitted parametric (lognormal) distribution

Geophagia or Child Inadvertent or Adult Ingestion (CIA) (GO)

LR

LR, A, E 50-100 mg d-I for non-geophagic child, 50 mg d-' for adult, 5-10 g d-I for geophagic child

Arithmetic mean value of 16 mg d-', 90 mg d-' S.D.

LR

59 mg d-' (geometric mean), 91 mg d-I (arithmetic mean), 126 mg d-I (S.DJd

Basis for Ingestion-Rate Estimates Ingestion-Rate Estimates (A, E, IM,M, LRY

geophagia), ordered by publication date (Simon, 1998). (continued)

TABLE5.18-Selected literature reporting quantitative estimates of soil ingestion-rate (inadvertent and

2

i4

i2

m

9 $

g

2

$ m

?'

\

Q, to

+'

C/A

U

U

Sheppard (1995)

Stanek and Calabrese (1995)

Assumed for a case study of 0-1.5 y: 0.1-10 mg d-I A TCDD risk assessment at 1.5-5 y: 9-50 mg d-I 1 x loM6risk 6-12 y: 5-50 m g d-' 13-30 y 0.1-50 rng d-I 0.1-30 mg d-I outdoors A 0.1-10 mg d-I indoors I Back-calculated from median = 84.5 mg d-I IM measurements of urinary mean = 261 mg d-I arsenic using EPA ingestion S.D. = 625 mg d-' and inhalation models 95th% = 986 mg d-I I/G Recommended values, all pica child: 500 mg d-I LR geometric mean values (GSD = 12); 2.5 y: 50 mg d-I 6 y: 20 mg d-' adult gardener: 20 mg d-I adult (indoors): 0.4 mg d-I I/G Revision of estimates 75 m g d-I (median M presented in Calabrese et al. daily)" (1989)

"Basisfor ingestion-rateestimates: Assumption (A), estimate (E),indirect measurements (IM),direct measurements 0, literature review (LR). bDerivedfiom lead ingestion-rates reported by Wedeen d al. (1978). 'Young children, 470 ? 370 m g d-I ; 10.5 percent of young children > 1 g d-' , 21 percent > 1 g d-I on at least one occasion, intakerates for single time geophagia incidences of 3.7 to 61 g d-'. dEquivalentto a geometric mean of 59 mg d-' and GSD of 2.81. "nge of median soil ingestion of 64 subjects over 365 d: 1 to 103 mg d-'. Range of average daily soil ingestion for 63 subjects over 365 d: 1 to 2,268mg d-l. Median of the daily average soil ingestion for 64 subjects: 75 mg d-'. Range of upper 95 percent soil ingestion estimates: 1 to 5,263rng d-'. Median upper 95 percent soil ingestion estimate of 64 subjects over 365 d: 252 mg d-I.

C

C

U

Lee and Kissel (1995)

Construction site

Residential

2

+

1

sr

m

Z

0

=!

rn

2

4

gC1

g

E

128

/

5. DOSE FROM INGESTED RADIONUCLIDES

TABLE 5.19 -Soil ingestion rates and number of days of exposure (TIused for screening dose calculations. Land-Use Scenario

Adult

Child

T ( d Y-9

Range

Agricultural Heavily vegetated pasture Sparsely vegetated pasture Heavily vegetated rural Sparsely vegetated rural Suburban Construction, etc.

0.1 0.05 0.1 0.05 0.1 0.05

0.1 0.2 0.1

0.1

-

270 270 270 270 270 270 180

180-360 180-360 180-360 180-360 180-360 180-360 90-270

-

-

% GSD of 4.2 was used for all child dose calculations. A GSD of 3.2 was used to calculate adult doses.

Calabrese and Stanek (1994; 1995) discussed the source of these variations and attributed it to high soillfood ratios, i.e., conditions in which the signal-to-noise ratio for the tracer was excessively high. Errors in specifying the fecal sampling period relative to the ingestion period relative to the ingestion period were also implicated in causing calculational errors. Most of the quantitative estimates reported, other than the previous four studies discussed, have little empirical basis. The few that do [e.g., Linsalata et al. (1986) or Wong et al. (198811, may not be easily generalized except within the similar conditions in the country of origin. The reported intake rate estimates in any of the papers cited here apply either to single individuals or to small groups which were observed for short periods of time. These and other factors contribute to the difficulty in generalizing to population. Some studies indicate unusual exposure and intake conditions. For example, the intake rate estimates by Haywood and Smith (1990; 1992) for the Aboriginal population of the Maralinga district in Australia (up to 10 g d-l) point out the risk to which indigenous people may be subjected by virtue of lifestyle attributes and environment. However, not all rural living conditions will result in extraordinarily high intakes. Linsalata et al. (19861, for example, found relatively modest intakes of 200 mg d-l for a farming community in Brazil, an environment considerably different than the Marlinga Desert. Presumably the degree of vegetative ground cover, a function of the regional climate, as well as life-style, are important factors that influence intake rates. Cultural norms for geophagia, even in the United States, may vary significantly.Vermeer and Frate (1979)investigated geophagia

5.3 DIRECT INGESTION OF SOIL

1

129

in rural Mississippi in 50 households containing 229 individuals. In this study, the individual had to admit to consumption of clay to be considered a practitioner of geophagia. Therefore, the data were believed to represent a minimum of the true incidence. The nutrition survey administered did not uncover any geophagia among adult males or adolescents of either sex but identified 57 percent of the women and 16 percent of the children (ages 1 to 4 y) as geophagic. In particular, 50 percent of the pregnant women practiced geophagia. Average daily consumption of clay was reported to be 50 g; the values were determined as a portion or multiple of the amount held in a cupped hand. Despite beliefs of some authors that geophagia in rural areas or among children with normal mental faculties is diminishing, geophagia has recently been noted in some countries not to be a rare event. In reports by Wong (1988) and Wong et al. (1988), children institutionalized in two homes in Jamaica were studied for geophagous behavior over a period of four months to assess the risk of exposure to geohelminth infection. Wong determined from analysis of fecal levels of dietary silica that in the two homes, 33 and 66 percent of children, respectively, practiced geophagia and that some children ingested up to 10 g soil d-'. However, 20 percent of the children accounted for >60 percent of the soil ingested. Several mentally disturbed children consumed, on occasion, up to 60 g soil d-'. In the study by Wong, 21 percent of the children with normal mental capabilities, with an average age of 3.1 y, displayed geophagia occasionally with ingestion rates >1 g d-' and in 10.5 percent of all observations of 24 children, ingestion was > 1 g d-'. Calabrese and Stanek (1993) reviewed the dissertation by Wong (1988) and concluded that the high prevalence of geophagia among these normal children challenges previous assumptions that geophagia is always a rare condition. The literature quoted in Table 5.18 has derived intake rates for specific groups of study by a variety of methods. The weakest estimates are those produced by assumption, e.g., guessing or assuming the number of times children might put candy which has collected surface dirt or dirty hands in their mouth. These studies would include Gelfandetal. (1975),Wedeenet al. (1978),Vermeer and Frate (1979), NASINRC (1980),Kimbrough et al. (1984), and Haywood and Smith (1990; 1992). Some literature has quoted parametric distributions based on an assumption in which the spread of the distribution represents lack of knowledge uncertainly, eg., Martin and Bloom (1976), Rogers (1975), and Healy (1977). Other literature has quoted intake rates based on measurements of fecal and urinary excretion of an environmental contaminant followed by a back calculation to

130

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5. DOSE FROM INGESTED RADIONUCLIDES

determine the likely rate of soil ingestion. Because these estimates were not produced by planned intake studies, these estimates are referred to as "indirect estimates." These studies would include Lepow et al. (1974), Day et a.!, (1975), Fisher et al. (19811, and de Silva (1994). The most reliable estimates of soil intake are from planned intake studies which have inferred their findings from the quantity of naturally occurring trace elements measured in feces (Binder et at., 1986; Calabrese et al., 1989; Clausing et al., 1987; Davis et al., 1990; Linsalata et al., 1986; van Wijnen et al., 1990) under conditions designed to determine reliable estimates. These studies are termed "direct estimates"in Table 5.18. The main distinction of these studies from "indirect measurements" is in their planning and design. The more careful of these studies subtract the amount of trace elements assumed to have been ingested from non-soil sources, e.g. food, toothpaste, etc. Finally, some reports have simply reviewed previous literature, e.g., Finley et al. (1994) and Sheppard (1995). The uncertainty of even well planned studies is emphasized by the recent report of Stanek and Calabrese (1995) in which they revised their previous estimates of ingestion rate from their 1989 study. Their new estimates were acknowledged to be a striking departure from their previous recommendations as well as EPA recommended default values. In particular, their new estimate of the upper 95 percent of the distribution of intake rates over a year's time was 1,750 mg d-' compared to 200 mg d-' recommended by EPA, a value that is generally interpreted to approximate the upper 95 percent of the distribution for children (Stanek and Calabrese, 1995). The recent analysis of Stanek and Calabrese also determined that 10 percent of the subjects might ingest 1.2 g d-I on average and that 33 percent of the children would ingest >10 g of soil on 1 to 2 d y-' and 16 percent would ingest >1 g of soil on 35 to 40 d y-I. The estimated median of the 64 subjects averaged over 365 d was 75 mg d-I and the range of average daily soil intake rates varied from 1 to 2,268 mg d-I. Most of the publications cited in Table 5.18 refer to populations in the United States; the exceptions reported findings in Brazil (Linsalata et al., 1986), Jamaica (Wong et al., 19881, Australia (Haywood and Smith, 1990; 1992),and the Netherlands (van Wijnen et al., 1990). A few of the papers include nonwhite populations, e.g., Black (see Gelfand et al., 1975; Wedeen et al., 1978; Vermeer and Frate, 1979), Native American (see Fisher et al., 1981), Jamaican (see Wong et al., 1988), and Aboriginal (see Haywood and Smith, 1990; 1992).Those studies noted in Table 5.18 which have empirical

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measurements from children are probably the most reliable. Intakes of soil by adults is much more controversial among reporting authors. Soil ingestion among adults in western societies is likely to be mainly a function of occupation. These studies cited in Table 5.18 do not address the special situation of the pica child or adult (e.g., pregnant women) who might intentionally eat soil. Ingestion rates for specially defined critical interest groups andlor non-United States' populations must be examined carefully and estimated with sufficient knowledge of the living conditions and cultural attitudes of the population of interest.

5.3.3 Recommended Ingestion Rates Some recommendations have been offered in the literature and by United States' regulatory agencies for default soil ingestion rates [for example, ATSDR (1992) Exhibit D.3; Kimbrough et al. (1984), EPA (1989a; 1989b; 1989c) Exhibit 6-14; EPA (1990a), Calabrese et al. (1992), Calabrese and Stanek (19941, Sheppard (1995)l. For example, ATSDR (1992) provides soil ingestion values for CERCLA (Federal Comprehensive Environmental Response, Compensation and Liability Act) and RCRA (Federal Resource Conservation and Recovery Act) mandated assessments. The Agency for Toxic Substance and Disease Registry (ATSDR) Public Health Assessment Guidance Manual (ATSDR, 1992) gives values of 50 to 100 mg d-I for non-pica children, 50 mg d-' for adults (Calabrese et al. 1990) and states that 5 to 10 g d-I are possible intakes of children with pica behavior. Their primary sources of data were Calabrese et al. (19891, Davis et al. (1990) and Calabrese et al. (1990). In a series of risk assessment guidance publications, EPA (1989a; 1989b; 1 9 8 9 ~recommends ) the use of similar values for soil ingestion: 200 mg d-' for children and 100 mg d-' for adults. Calabrese et al. (1992) and Calabrese and Stanek (1994), in later publications, made the following recommendations which were intended to be conservative, upper-bound estimates: (1)assume that two percent of children aged 1 to 6 y exhibit soil pica of 1g d-', (2) assume that 0.2 percent of children ingest 10 g d-' from ages 1 through 6 y. They also provided estimates intended to be more realistic: (1)assume one percent of children exhibit soil pica of 1 g d-I for 4 d week-' and 500 mg d - ' for 3 d week-' for 4 y, (2) assume that 0.2 percent of children ingest 10 g d-' for 3 d week-' and 200 mg d-' for 4 d week-' for 4 y. The recent revisions (Stanek and Calabrese, 1995)of the Calabrese ingestion r a t e s emphasize t h e need to have wide uncertainty

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distributions that allow for high values to occur with finite probability. The recommendations here take into account the near universal values of 50 to 100 mg d-I as a n estimate of central tendency. Furthermore, a lognormal distribution has been assigned to represent the range of possible alternative values with a GSD of 4.2 for children and 3.2 for adults. This degree of uncertainty results in an upper 95th percentile equal to 100 mg d-I x (4.2)2 or 1,764 mg d-' for children, a value in agreement with the Stanek and Calabrese (1995) revised estimates. NCRP recommendations for intake rates for the various scenarios are presented in Table 5.19. In light of the sparse empirical data for many scenarios, Table 5.19 is intended to recommend a wide range of possible alternative values. The ranges may be assumed to cover approximately 95 percent ( + 2 S.D.) of the values likely to occur in a typical United States' population. The purpose of these suggested ranges is for use in assessments which fit the scenario description. It should be understood that the values and ranges presented in Table 5.19 represent a subjective determination of the central tendency and spread of numerous literature citations rather than a statistical analysis of available data.

5.3.4

Use of Soil Ingestion-Rate Data in Screening Calculations

The screening criteria in this Report are defined for critical population subgroups that are composed of individuals whose occupations, recreational activities or lifestyles would likely bring them into contact with soil contaminated with a particular radionuclide. These assumptions take into account exposure situations for people with life habits that are likely to be substantially different from many typical residents of the United States. The definition of critical groups, in relation to soil ingestion, requires careful consideration. Children, because of smaller body mass may be the obvious choice; however, in some circumstances, only adults may be exposed (eg., at construction and agricultural sites). In general, the (unidentified) child who practices geophagia is a t the highest risk. However, whether such individuals are numerous enough to consider in setting guidelines for acceptable soil contamination levels is debatable. The degree of realism or conservatism desired, regardless of the subgroup of interest, is another area requiring consideration. To some extent, this problem is handled in the guidance presented in this Report by a detailed uncertainty analysis that attempts to differentiate between true stochastic variability (i.e., variation

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among individuals) and uncertainty (e.g., the relevance of a parameter value to a particular critical interest group). Finally, for population exposure considerations, it is difficult to know what is the appropriate proportion of the population that is a t risk from soil ingestion. However, just recognizing that there is a segment of the population that practices pica.geophagia is the first step to defining a heretofore neglected critical interest group. Similarly, recognizing that soil ingestion occurs inadvertently is a sufficient argument for including soil ingestion in screening calculations of radiation risk to populations living on, or near, contaminated lands.

5.3.5 Calculation of Screening Doses To calculate the annual E(r)(Sv y-I) in terms of a unit concentration of a radionuclide in soil: where:

CWil

I,d T Dfw OF

average concentration in top 5 cm of soil (Bq kg-') average ingestion rate of soil during the exposure period (kg d-'1 = exposure duration (d y-I) = ingestion dose factor (Sv Bq-I) defined earlier = occupational exposure modification factor = =

In the analysis presented in this Report, a number of land-use scenarios have been defined which specify whether the land is rural or urban and whether the vegetation is heavy or sparse. For each of these land-use scenarios, specific values for the parameters IBOil and T in Equation 5.4 have been chosen (Table 5.19). The same ingestion dose factors used earlier for ingestion of foodstuffs are also used for ingestion of soil. (For sites where children are likely to be the most exposed population, the CIA correction factors described earlier were applied.) The concentration of the nuclide in the ingested soil is assumed to be equal to the mean measured over the top 5 cm. Even if the concentration in the soil actually ingested is higher than the average, the bioavailability of many radionuclides (transfer from gut to blood) is likely to be much lower than for the same nuclide incorporated in vegetables, meat or milk, and thus the calculated E from soil ingestion are still likely to be conservative. A site-specific dose assessment where the soil ingestion pathway is significant should consider the actual distribution of the nuclide with depth in

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the soil and whether dose factors based on lower gut uptake factors than used in the screening model might be more appropriate. For the screening calculations in this Report, the use of median parameter values is based on soil ingestion by normal adults or children since pica children make up < 1 percent of all children and <2 percent of children ingest more than 1g d-' of soil. However, we assign a GSD for purposes of calculating the range in total ingestion dose which allows for the possible maximum ingestion of up to 1g d-' of soil for adults and over 3 g d-' for children for dusty sites. Children are considered to be the most exposed population where dwellings are present on the site. Only normal (non-geophagic) adults are assumed to be exposed a t the sites without dwellings. The OF is not used directly for the screening calculations but is included in the land-use dependent choice of ingestion rate range.

5.4 Dependence of Committed Effective Dose on Age

As pointed out in the discussion in Section 5.2.7, the ingestion dose factor for some radionuclides varies considerably with age. For the purposes of calculating doses for screening, it is important to consider the most exposed population. Thus, when the total dose to children or infants is higher than the total dose to adults, the screening guidance has been based on those childlinfant doses. For most food categories except milk, the intake for infants and children is usually much lower than that for adults and, therefore, the actual E(T)will not vary as much with age as is shown in Table 5.12. Even though the intake of vegetables and meat is smaller for children and infants than for adults, the large difference in the dose factors for some nuclides still often results in a higher E(T),particularly when a large fraction of the dose is from milk consumption. Thus, for the land-use scenarios where children or infants might be the most exposed population group v i a the ingestion pathway, correction factors were applied to the calculated adult dose to estimate the child and infant doses from the ingestion of vegetables, meat, milk and soil. These correction factors were obtained by weighting the ratio of the infant to adult and child to adult E(2) factors reported by the ICRP (1993; 1995b; 1996b) and given in Appendix C, by the ratio of the average annual intakes of each food category for infants, children and adults reported by the UNSCEAR (1993) (see Table 5.16). The highest resultant "corrected" dose (i.e., either for infants or for children) was then used as the ingestion screening dose for "children." For nuclides where ingestion of milk is the most important pathway,

5.6 CALCULATED SCREENING DOSES

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135

the highest dose will be to infants. For vegetables and meat, the highest dose will generally be to children. As discussed in Section 7, the total ingestion dose calculated for "children" was combined with the corresponding inhalation and external exposure doses for children (for sites where applicable) to determine a conservative value for the total dose to a "child" from all pathways for each land-use scenario where children can receive a dose. The infanvadult or child, adult ratios used for each radionuclide considered in this Report are listed in Appendix C, along with the correspondingadult dose factors. Table 5.20 compares the calculated total ingestion doses for adults with the corresponding doses for infants and children versus landuse scenario to illustrate the magnitude of the average age correction on the final ingestion screening doses. The differences in the overall screening guidance will be smaller for nuclides where the external and inhalation pathways contribute a significant fraction of the total committed dose. 5.5 Summary of Recommended Parameter Values for Screening (Ingestion Pathway)

Land-UseIndependent Concentration in soil

Bv Uptake to milk, meat transfer Adult dose factors Human diet Animal diet Hold-up factors

Land-Use Dependent Human soil ingestion Bv*,B,. Fraction of diet from site (humans) Fraction of diet from site (animals) CIA correction factors

1 Bq kg-', CV = 0.5 see Table 5.6, Appendix D see Table 5.10,Appendix D see Appendix C see Table 5.2 see Table 5.8 see Table 5.11 see Table 5.19 see Table 5.7 see Table 5.3 see Table 5.8 see Table 5.16, Appendix C

5.6 Calculated Screening Doses Using Equation 5.1 and the parameter values described in the preceding paragraphs, an ingestion screening dose for each nuclide

IfA

CIA

VA

PV CIA

YA

PS CIA

IIA

RV CIA

VA

RS CIA

IfA

SU CIA

"Based on age-dependent dose factors from ICRP Publication 72 (ICRP, 1996b) times UNSCEAR (1993) age-dependent ingestion rates (see Table 5.1).

Nuclide

AG

selected radionuclides.a

TABLE 5.20-Ratio of infantladult total ingestion dose and childladult total ingestion dose for

6.6 CALCULATED SCREENING DOSES

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137

was calculated for eachland-use scenario. Table 5.21 lists the median and estimated range (5 to 95 percentiles) for some selected nuclides often encountered in environmental contamination scenarios. See Appendix A for comparisons of these ingestion doses with the doses from other pathways and for a breakdown by food type. These doses represent conservative estimates of the dose to a representative member of the most exposed population group. For example, for sites where ingestion of milk is calculated, the most exposed group is generally children and infants. Also shown in the table is the estimated range of possible ingestion pathway doses to an individual in this population, based on the 5 to 95 percent confidence level of the calculated doses. The confidence interval was estimated from the ranges or uncertainties assigned to each parameter in the dose model as discussed in Section 7. It can be seen that the ingestion dose varies considerably depending on assumed land use. This variation reflects the variability in root uptake, soil adhesion, and type and amount of food produced. The range of doses is larger than the range for the external and inhalation pathways, reflecting primarily the additional large uncertainty in the transfer fadors from soil to vegetation and vegetation to milk or meat and in the ingestion dose factor. As shown in Appendix A, the dose from ingestion of soil is usually small compared to that from ingestion of food except for radionuclides such as plutonium. This is reflected in the differences in estimated doses for the two suburban land-use scenarios, one with and one without a vegetable garden. The actinides generally have low uptake into food, but effective dose factor is relatively high, particularly for children, due to a relatively high dose to the GI tract itself. Thus, the dose to a pica child from plutonium-contaminated soil could be quite significant. However, as discussed earlier, the calculated doses from soil ingestion are believed to be conservative. The complete set of ingestion screening doses is given in Appendix A.

"Median dose to representative member of critically exposed population. Ranges in parentheses represent 5 to 95 percentiles. Doses include contributions of decay products and are maximum annual doses over a 1,000 y interval. Nuclide concentration assumed uniform over top 30 cm. See Appendix A for complete listing of ingestion doses and identification of critical population.

Nuclide

Land-Use Scenario

TABLE5.21-E(z)[Sv (Bq kg-')-'] fiom ingestion for selected nuclides."

w

L

OD w

Determination of Radionuclide Concentration in Soil To apply the guidance in this Report, it is first necessary to determine the level and extent of the soil contamination. This Section discusses the factors to consider in designing screening and sampling programs to accomplish this task.Acceptable methods of soil analysis and guides to sampling soil, including statistical considerations with regard to the number and spacing of samples, are also discussed. A strategy is recommended for determining the initial screening concentration levels in the top 5 cm soil layer. This strategy is designed to ensure that the uncertainty in the mean value determined for the site will be low enough to reasonably apply the screening guidance given in Section 2.

6.1 Factors to Consider in the Design of Screening and

Sampling Programs for Radionuclide Concentrations in Contaminated Soil

An adequate assessment of present and future potential radiation exposures and doses to persons living on or near contaminated soil (or ingesting foods grown in such soil) requires a determination of what the contaminating nuclides are and their mean concentration and spatial distribution in the surface soil. Additional information may also be required to adequately assess how individual nuclides might translocate with time. Translocation might occur vertically by penetration into the soil by both natural processes and as a result of human activities (e.g., plowing) and laterally via resuspension, erosion or human disturbance. Because of the wide diversity of possible contamination scenarios, one cannot define a single procedure or set of procedures for characterizing the concentration of a radionuclide in contaminated soil. Nor can one define the degree of measurement precision required for site-specific dose assessments. Usually each site must be considered on a case by case basis. However, the major factors that must

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be considered in designing a program to characterize the soil radionuclide concentrations resulting from any contamination scenario can be identified. Procedures, techniques and instruments currently available for carrying out typical programs can also be specified. The scope and magnitude of the monitoring program will, of course, be governed by the particular scenario, the potential hazard, the extent of the contamination, and the present and projected use and location of the contaminated land. Generally, one can infer the probable nuclides of concern (for example plutonium around the Rocky Flats Plant, uranium and radon daughters around mill tailings piles, etc.) and the probable extent of contamination from the contamination scenario. The character of the contaminated land will likely influence the final geographical distribution of deposited radionuclides. The more diverse and complicated the terrain, the more likely an original uniform deposition will be redistributed by wind, water, etc. This redistribution will complicate the characterization of the total quantity of a radionuclide present and its spatial distribution. Conversely, deposition onto a relatively flat, grasscovered meadow or field might be easily characterized with only a minor survey effort. The type of sampling conducted and number of samples required will depend on what specific nuclide analyses are planned, the precision required and what other physical information is required to adequately assess the potential hazard ( e g . ,particle sizes, chemical composition, mass isotopic ratios, etc.). Such a priori decisions will dictate the number and sizes of soil samples required, or indeed, if soil needs to be sampled at all. The following sections describe equipment and techniques available for characterizing soil contamination once a basic scope and level of effort have been established.

6.2 Instrumental Measurement Techniques ORen, it is possible to determine the concentration in soil quite adequately without extensive soil sampling, utilizing in situ instrumental techniques. Two methods, in particular, have been used successfully, particularly for gamma emitters, in situ gamma-ray spectrometry and free-in-air exposure or kerma rate measurements. 6.2.1 In Situ Gamma-Ray Spectrometry In situ gamma- and x-ray spectrometric techniques have been used for a number of years for characterizing contaminated soil

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(Anspaugh, 1976; Beck et al., 1972; Gogolak, 1982; ICRU, 1994; Miller and Shebell, 1993; NCRP, 1976). Theae techniques allow one to rapidly survey large areas, locate hot spots, identify contaminants in the soil, and, with proper calibration, estimate area contamination (e.g., Bq kg-' or Bq m-2). The major advantage of in situ spectrometry is that a large number of analyses can be carried out in a relatively short time compared to the effort required for soil sampling, preparation and analysis. Furthermore, large areas of soil are, in effect, sampled in each measurement compared to the relatively small area represented by a single soil sample. For example, a detector placed at a height of 1m above the ground effectively "sees" an area of soil of diameter of about 10 m (Beck et al., 1972; Miller and Shebell, 1993). Unfortunately, however, only photon-emitting nuclides can be detected, and the detector usually cannot distinguish the location of the radionuclides in the soil, i.e., the point of origin of the detected flux. Thus, unless some a prwri knowledge of the distribution of the radionuclides in the soil is available, one cannot usually obtain precise quantitative estimates of the nuclide concentration. Often, however, the nuclide concentration can be assumed to be fairly uniformly distributed horizontally within the field of view of the detector, and data from a limited number of soil samples can provide s f i c i e n t information on the depth profile to allow one to utilize the in situ measurements as a primary means of characterizing the concentration in soil at a contaminated site (Beck and Krey, 1980; 1983; Miller and Helfer, 1985; Dickson et al., 1976). Miller et al. (1994) discuss how the depth distribution can be estimated from the in situ spectrum alone for some nuclides emitting gamma rays of more than one energy, using the example of the measurement of 23Win the environs of the DOE Fernald site. Even if the depth profile for a site was not measured, one can obtain a reasonable estimate of the surface soil concentration in the top 5 cm for most reasonably expected depth profiles by making the assumption that the radionuclide is uniformly distributed with depth. As illustrated in Table 6.1, this provides a conservative estimate of the true nuclide concentration over a wide range of source energies, since, even for high-energy gamma emitters and deeply distributed sources, only a small fraction of the total uncollided flux is from depths >5 cm. The worst situation is if the radionuclide is actually distributed uniformly, but completely within a 5 cm slab. In this case, assuming an infinitely deep uniform distribution will result in underestimating the concentration of high-energy gammaray emitters by about 30 percent. However, this depth profile is unlikely to occur in real environmental contamination scenarios.

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TABLE 6.1-Ratio ofpredictedltrue average concentration in soil in top 5 cm using in situ gamrna-ray spectrometry and assuming a uniform distribution with depth. l3ICs (662 keV)

'We (144 keV)

-'Am (60 kev)

a-' = 0.1 cm a-I = 1 ern a-I = 3 cm a-I = 10 cm 1 cm slab

2.4

1.4 1.0 1.0

3.3 1.8 1.2 1.0

1.6

2.2

5 cm slab

0.7

5.6 2.3 1.3 1O . 3.0 1.0 1.0

True Depth Distributiona

0.9 Infinite slab 1.0 1.0 a a- I is the relaxation length (RL),i.e., the depth at which an exponentially decreasingconcentration is reduced to l/e of the surface concentration. Note that for slabs less than infinite thickness but greater than 5 cm, the predicted concentration will always be slightly less than the actual concentration. All calculations assume a bulk density of 1.6 g c ~ r - ~ . The calculations shown in the table assume that the radionuclides are not covered by even a thin layer of uncontaminated soil. A few centimeters overburden of uncontaminated soil will likely also result in the true concentration in the top 5 cm being underpredicted, especially for low-energy emitters, due to the additional attenuation of the emitted photons. The uncollided flux will be reduced by a greater percentage than the reduction in average concentration over the top 6 cm. The estimates in Table 6.1 are for a uniform soil bulk density ~ . exponential depth profiles scale as d p (Beck et al., of 1.6 g ~ r n - The 1972)so, for example, a distribution with a-' = 3 cm a t 1g ~ m - ~ would be equivalent to a distribution with a-' = 1.9 cm at 1.6 g ~ r n - ~ instead. Similarly, a 5 cm slab at 1g cm- would be equivalent to a 3.1 cm slab a t 1.6 g ~ m - ~ . The significant problems associated with representatively sampling a large inhomogeneous (both in ground cover andfor nuclide distribution) area through a limited number of soil samples, each of relatively small area, further illustrate the advantages of utilizing in situ gamma spectrometric techniques to supplement the soil sampling. Besides being faster and more economical, t h e detector "samples" a much larger area of ground, providing a much more representative measurement. An example of the relative precision of sampling by in situ gamma-ray spectrometry versus soil sampling is available from measurements made on Rongelap Island in 1991 to 1993 (data courtesy of the Marshall Islands Nationwide Radiological Survey). Rongelap received fallout in 1954 from weapons tests on

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Bikini, about 150 krn away. Variability was studied by comparing gamma-spectrometric measurements of 13'Cs made a t 64 equally spaced grid points, 200 m apart, with composites of three 15 cm x 15 cm x 5 cm deep soil samples taken within 10 m of the spectrometer. The CV observed for the gamma-spectral measurements was 0.42 compared to a value of 0.60 for the soil sample data. A comparison on a smaller scale made by dividing four 200 m x 200 m grid cells into twenty-five 40 m x 40 m subcells resulted in the CV for the gamma-spectral data ranging from 0.2 to 0.4, compared to CVs for the soil sample data of 0.4 to 0.6. The four CVs for the gammaspectral data for the smaller grids ranged from 50 percent to 80 percent of that for the corresponding soil data. Thus, in situ gamma spectrometric sampling can often provide mean radionuclide concentrations in soil with confidence intervals much narrower than those obtainable from a tractable number of soil samples. In addition, the data provided offer immediate guidance on the relative uniformity of contamination over the site and thus on the number and spacing of (soil) samples required either to estimate total inventory to a particular degree of certainty, or to locate all hot spots to a given probability. Unfortunately, as discussed earlier, some nuclides can only effectively be sampled by soil collection and subsequent laboratory analysis. Often, however, one can infer concentrations of a particular radionuclide from measurements of a second "surrogaten radionuclide, based on expected relative abundances of known sources (eg., weapons grade plutonium from measurements of 241Amor fallout 90Srfrom measurements of 137Cs) (Beck and Krey, 1983; Beck et al., 1976; Ibrahim et al., 1995; 1996).

6.2.2

Exposure Rate Measurements

When one or more of the soil contaminants is a beta or gamma emitter, external radiation exposure may constitute a significant dose pathway as discussed in Section 3. In this case, it may be beneficial to directly measure current exposure rate levels on and around the contaminated area. These measurements can be used to estimate potential doses or for verifying theoretical estimates based on measured soil activities. If the exposure rates are significant, continuous monitoring may also be needed to document temporal variations due to decay, translocation or changes in soil moisture. If the activity in the soil is from a single radionuclide, a gammaray exposure rate measurement can also be used in lieu of or in conjunction with soil sample analyses. This is particularly useful if the depth profile is known, or can be roughly estimated, since

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tabulated values for conversion between concentration in soil and gamma-ray free-in-air exposure rate are available (Beck, 1980; ICRU, 1994).In addition, if the relative fractions of various multiple contaminants are known, as was the case for fallout from weapons testing at the Nevada Test Site (Beck and Anspaugh, 1991),one can derive conversion factors relating the measured total air dose to the concentration in soil of each radionuclide. If the actual depth profile is not known, assuming a uniform distribution with depth to 5 cm will, generally, provide a conservative estimate of the average concentration in the top 5 cm regardless of the true distribution. Thii is shown in Table 6.2 for a selection of gamma-emitting nuclides spanning a wide range of source energies. Use of the 5 cm slab as the default, rather than the infinite slab assumption used for the flux in Section 6.2.1, will assure a conservative estimate for high-energy photon emitters. (See Table 3.2 for examples of the variation in E,, with depth profde. The air kerma rate will vary similarly.) Again, however, this assumption may not hold for sites with a few centimeter overburden of uncontaminated soil, since, particularly for low-energy emitters, the reduction in exposure rate due to the additional attenuation will be greater than the reduction in average concentration over the top 5 cm. Conversion factors from free-in-air dose rate to concentration in soil for nuclides most commonly expected in the environment are listed in Table 6.3. These factors can be used for making initial conservative estimates of the average concentration in soil from freein-air dose rate measurements. TABLE 6.2 -Ratio of predicted ltrue average concentration in soil in top 5 ern versus true depth profile when predicted value is inferred from Fee-in-air dose rate measurement assuming a uniform 5 cm slab depth profile. True Depth Dietribution8 a-' = 0.1 cm @--I

- 1 cm

a-' = 3 cm a-' = 10 cm

1 cm slab 5 cm slab 15 cm slab Infinite slab

137Cs (1.3MeV)

(662keV)

2.0 1.4 1.2 1.4 1.7 1.0 1.6 1.7

2.3 1.5

1.3 1.5 1.7 1.0

'Ye (144keV)

2.9 1.8 1.3 1.3 1.3

1.0

1.6

1.3

1.8

1.3

241Am

(60keV)

8.0 2.5

1.5 1.3 2.7 1.0 1.1 1.1

'a-'is the relaxation length (RL), i.e., the depth at which an exponentially decreasing concentration is reduced to l/e of the surface concentration.

6.3 SOIL SAMPLING

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TABLE 6.3 -Free-in-air kerma rate (pGy h-') per Bq kg-' for selected radionuclides." Nuclide

Kerma Rate

Nuclide

Kerma Rate

"The free-in-air kerma rate assumes a uniform 5 cm slab depth profile and was estimated by dividing the 5 cm slab E equivalent from Federal Guidance Report 12 (Eckerman and Ryman, 1993) by the appropriate value of Sv Gy-I from Table 3.3. A soil density of 1.6 g C I X - ~is assumed. Values include contributions of decay products.

For most contamination scenarios, dose rates from beta emitters in the soil are small compared to the gamma component. However, the flux of beta rays in air near the ground can be quite high for nuclides deposited near the surface of the soil, as can the free-in-air ionization rate. For example, the free-air ionization rate a t 1m above the ground from beta rays from naturally occurring radionuclides in the soil is about 50 percent of that from the gamma emissions from the same sources (O'Brien et al., 1958; Minato et al., 1978). If the total exposure rate is measured with an improperly shielded (thin-walled) detector, a large beta flux may result in the gammaray component being overestimated. Because the beta flux close to the air-ground interface will vary tremendously with the amount and atomic number of material between the source and detector, and thus with the amount of soil moisture, it is generally imperative to adequately shield out the beta component when using air dose rate measurements to screen for soil contamination.

6.3 Soil Sampling Once the location and extent of a contaminated site have been established and initial screening completed, sufficient numbers and types of soil and vegetation samples may need to be collected to satisfy remaining characterization goals. For detailed site-specific assessments, it is usually beneficial to plan to obtain an adequate number of samples to allow for various types of analyses, replicates

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6. DETERMINATION O F RADIONUCLIDE CONCENTRATION

for quality assurance and for samples to be archived for future re-analysis. The major objective is to collect soil samples in a manner that will provide a valid estimate of the mean radionuclide concentration or inventory for the contaminated axea and provide reasonable estimates of the probability that significant anomolies (hot spots) have not been overlooked. The number of samples required for a given degree of certainty will depend on the local terrain, uniformity of deposition, and extent of contamination. Parkinson and Horrill (1984) have assessed the variability in "global" plutonium fallout in grazed and ungrazed pastures, illustrating that a large number of samples are required to adequately characterize even a fairly simple contamination situation. Nyhan et al. (1983) reported on considerations involved in choosing the sizes and numbers of samples to determine inventories of 13'Cs in natural terrain near the Trinity site. Fowler et al. (1977) discuss the general philosophy of sampling soils for radionuclides. What constitutes a representative sample or number of samples for the area in question may be difficult to determine without some preliminary data, such as in situ gamma spectrometric or free-inair dose rate measurements. Often the number of samples that can be taken and subsequently analyzed will be limited by economic factors, and the problem is how best to locate the allotted number of samples to maximize the results. The degree of statistical rigor needed for choosing the number and location of sample sites depends greatly on the program's specific objectives. To determine if guidelines for acceptable levels of environmental contamination are met, an adequate number of representative, properly selected measurements is required. However, the process of determining the proper number of measurements, how the sampling sites should be selected and how the statistical comparison with the screening or other guidance should be made, opens many complex questions. Gogolaket al. (1997)discuss the problems related to spatial inhomogeneity and suggest non-parametric sampling strategies to be used for sampling to verify cleanup of contaminated sites.

6.3.1

Soil Sampling Methodology

Various methods of sampling have been reported in the literature. The procedures manual of the Environmental Measurements Laboratory, U.S. Department of Energy (EML, 1997) provides numerous additional literature references on soil sampling technology as well

6.4 AIR

SAMPLING

/

147

as a discussion of site selection criteria for estimating inventories and depth profiles. ASTM standard C988-90, Standard Practice for Sampling Surface Soil fir Radionuclides (ASTM, 1990) provides guidance on sampling soil for radionuclides. For site-specific assessments, the required sampling depth will depend on the particular contamination scenario and how long the radionuclides have been in the soil. Generally, sampling to depths >30 cm is not required for contamination due to airborne deposition. For example, "globaln fallout that was deposited primarily in the early 1960s is now generally distributed exponentially with depth with a relaxation length of only 5 or 6 cm (ICRU, 1994; Miller et al., 1990). Conversely, fresh fallout i s generally confined to t h e top few centimeters of soil (ICRU, 1994).

6.3.2 Sample Preparation and Analysis Expensive and careful efforts expended in collecting large numbers of samples in the field may be completely wasted by improper preparation of the samples in the laboratory, resulting in the analyses being unrepresentative of the actual average concentration in the sample or of the area and depth of in situ soil sampled. Techniques for preparing soil samples for analysis have been described in HASL-300 (EML, 1997). NCRP Report No. 58 (NCRP,1985a) reviews procedures for analysis of radionuclides in environmental samples and also discusses radiochemical separation procedures. A critical element in any sampling program is a well-conceived quality assurance program governing all phases of the sampling, soil preparation, and analysis including the preservation, integrity and identification of samples. This program should ensure that all equipment and instruments functioned properly, that samples were not cross-contaminated or otherwise compromised, and that adequate records were kept of all measurements and results. The program should also allow a reasonable determination of both the precision and accuracy of all radioanalytical measurements as well as all important subsidiary data. The required accuracy and precision will vary depending on the contamination scenario, characterization goals, and types of measurements. The key is to decide what is acceptable for the particular situation and then to demonstrate compliance. 6.4

Air Sampling

For site-specific assessments, it may be necessary to make direct measurements of average air concentrations to evaluate the dose

148

/

6. DETERMINATION OF RADIONUCLIDE CONCENTRATION

from resuspended material. Generally, large quantities of air must be sampled (>10,000 m3) to collect sufficient sample for a precise analysis. Methods for air sampling are discussed in ACGIH (1983) and in EML (1997). (See also Section 4.2.2.1 for a discussion of the "footprint" of a high-volume air sampler.) A common omission in air sampling programs is the failure to determine the airborne particle concentration (Bq kg-'). This quantity is important for diagnostic purposes and risk assessment and is easily determined by weighing the impaction surface (usually a filter) before and after exposure. Taking precautions with hygroscopic media, one can then determine the mass loading, M (kg m-3),before analyzing the filter for radionuclide concentration, C& (Bq m-3).Then the particle concentration is just CIM (Bq kg-').

6.5 Strategy of Determining Radionuclide Concentrations for Screening

As discussed earlier, the natural variation in soil contamination is due to factors such as the manner of the radionuclide deposition, the microtopography and weathering of soil, and the dilution of individual particles by host soil particles. Among stratified soil contamination zones sampled by areaaveraging in situ gamma spectroscopy methods, the coefficient of variation of specific soil activity, S , is generally about 0.5 (see Table 4.5). If, on the other hand, the stratified soil contamination zone is sampled by removing cores of small surface area, it would not be unusual to obtain a coefficient of variation for the mean concentration greater than twice that, as illustrated in the previous section. The effect this would have on the estimate of mean radionuclide concentration in air, for example, can be expressed by examining the CV of C& in Section 4.2.5 with a value of two for the CV of S rather than the value of 0.5 which was assumed. For C, the resultant CV = 2.2 as opposed to the value of one calculated in Section 4.2.5. Thus, in this case, the uncertainty in C& would be almost entirely due to the variability in S. A similar result would be obtained for the dose from external radiation exposure. For this reason, one must approach the problem of a determination of the screening estimate of S with a specific strategy in mind designed to assure that the CV of the measured nuclide concentration is 5 0.5. In light of the significant problems discussed in the previous section with regards to determining the mean concentration and variability for a contaminated area, it is appropriate to define a strategy

6.5 D E E F M N M N G RADIONUCLIDE CONCEWl'RATIONS

1

14

of representative, area-average sampling for purposes of applying the screening criteria given in this Report.

6.5.1 Estimating Soil Concentrations by In Situ Spectrometry or

Exposure Rate Measurements

If applicable, in situ spectrometry or free-in-air dose rate measurements should be used in lieu of soil sampling for the initial screening. For screening, it was shown in Table 6.1 that assuming a uniform depth distribution will generally provide a conservative estimate of the surface soil concentration. Similarly, if only a single nuclide is known to be present, or if the relative activities for multiple nuclide contamination are known, one can use the conversion factors given in Table 6.3 to estimate the surface soil concentration from free-inair exposure rates.

6.5.2

Sampling Soil for Screening

If in situ gamma-ray spectrometry is not applicable or not available, a preliminary, relatively simple soil sampling protocol is recommended for obtaining a reasonably precise estimate of the mean surface soil concentration for the contaminated site. The area over which the average soil nuclide concentration should be determined depends on the dose pathway of most importance. For example, an individual inhalation exposure at the site will be the result of radionuclide particles each with a trajectory originating from a different upwind distance. As a rule of thumb, when the wind over a long time averages over many wind directions, about 75 percent of the inhalation exposure particles will come from within a radius of about 55 m (see Table 4.1), an area of about one hectare (10,000 m2). This same rule applies to a typical high volume air sampler placed on the site (see Section 4.2.2.1). The dose from external exposure will be due to radiation from a comparable,if somewhat smaller, radius while the dose from the ingestion pathway will depend on the average concentration in soil over at least as large an area if not larger. Thus, each site should be divided into land-use zones, based on the current or likely utilization of the site. If a site is suspected of having contamination on an area <1 hectare, then that should be considered one land-use zone. A single 2 to 3 kg soil sample should be collected as the screening sample for each land-use zone. The method described in ASTM Standard C988-90 (ASTM, 1990) can be

150

/

6. DETERMINATION OF RADIONUCLIDE CONCENTRATION

used as an alternative. The sample should be obtained by pooling at least twenty to twenty-five 5 cm deep surface-soil subsamples of fixed volume, not
6.5 G D -

RADIONUCLIDE C O N C ~ T I O N S 1

151

although perhaps not as accurate, could probably be obtained for the same site from 5 to 10 in situ measurements. 6.5.3 Site-Specific Soil Sampling

If the screening analysis shows that the site falls into the category requiring a site-specific assessment (see Section 2), then other more expensive but more effective soil sampling strategies should be applied a s discussed in Section 6.3. The sampling protocol may require detailed mapping of the contamination pattern to determine the variability and thus the optimum number of samples. Detailed concentration depth profiles may be required as well as separate instead of a single pooled sample analysis. Gogolak et al. (1997) provides guidance on sampling sites to determine the mean activity as well as the likelihood of hot spots exceeding some predetermined level. Soil screening guidance has also been provided by the EPA (1996a; 1996b). The Multiagency Radiation Survey and Site Investigation Manual (MARSSIM) (NRC, 1997) developed jointly by the NRC, U.S.Department of Defense, DOE and EPA provides information on planning, conducting and evaluating surface soil radiological surveys for demonstrating compliance with dose or risk-based regulations or standards.

7. Calculation of Screening Doses This Section describes the methodology used to estimate the median and "maximum" dose to a member of a n assumed critically exposed population group and discusses why these "maximum" doses are believed to result in conservative dose estimates to most individuals in the critically exposed population. I n recommending screening limits, one wishes to estimate the likely maximum dose to which an individual might be exposed as well as the median dose for a member of the most exposed group. To accomplish this task, an estimate was also made not only of the central tendency of each parameter used in the dose models described in Sections 3 through 5, but also of the uncertainty or variability in the central tendency value. Both the central tendency and variability estimates are the best judgments of this Report, based on the available information from the literature. When few actual data were available, both the central tendency estimates and ranges were chosen conservatively. As discussed in Sections 3 through 5, the uncertainty was generally expressed as a likely range or as a S.D.if the distribution is known or assumed to be normal or lognormal. To estimate the likely distributions of doses to a member of a critical population group for various nuclides and exposure pathways, a Monte Carlo analysis was carried out for each radionuclide considered. The Monte Carlo simulation simultaneously calculated the doses from decay products a s well. Separate calculations were carried out for adults and for children for land-use scenarios where children might constitute the critical group. The Monte Carlo simulations provided a distribution of possible doses to a member of the critical group for each land-use scenario and nuclide. Separate dietributions were obtained for each dose pathway as well as for the total dose from all pathways. The distributions of doses were generally quite broad due to the large uncertainty in the average or central tendency of the various parameters entering into the dose determination. However, the parameter uncertainties were chosen to be SUEciently conservative to ensure that the calculated doses adequately reflect the possible distribution of doses to a representative member

7.1 DISTRIBUTION OF INDMDUAL DOSES

1

153

of the most exposed population except, perhaps, for individuals with highly unusual behavior.

7.1 Distribution of Individual Doses The equations and associated parameters used to calculate the doses to exposed critical populations from each important pathway were presented and discussed in Sections 3 through 5. Recommended central tendencies for screening were also presented along with the chosen estimates of the uncertainty or variability. One possible method of estimating the dose to the maximally exposed individual would be to use the maximum value of each parameter in the dose equation rather than an estimate of the mean or median. However, this approach would result in unrealistic, exceedingly conse~ative maximum doses. Such a procedure implies that the same individual who spends most of hisher time outdoors on the site ingests the most soil directly, drinks the most milk from a backyard cow or goat, has the highest radiosensitivity, the highest biokinetic transfer factors, et.. The approach taken here is to estimate the approximate distribution of doses from each pathway, and from all pathways, for each land-use scenario. This information is then used to produce a distribution that represents a more reasonable, yet still conse~ative, estimate of the dose to a representative member of the critically exposed population. Because the uncertainty and variability of many parameters entering the dose calculations are not well known, the stochastically determined dose distributions are only approximations themselves. However, by using prudently conservative estimates for the ranges or S.D., the calculated distributions are also consemative, i.e., likely to overestimate the true range of doses. The stochastic calculations were carried out using a Monte Carlo simulation. Assumptions were made regarding the shapes of the various parameter distributions. The parameter values and uncertainty estimates used are summarized in Table 7.1. (Wherethe exact parameter value depends on the particular radionuclide, refer to the appropriate tables in the text or appendices for specific values.) Also given are the probability distributions used for carrying out the Monte Carlo analysis. Where the true uncertainty distribution was not known and could not be reasonably inferred, a triangular distribution was assumed with mode equal to the estimate and minimum and maximum given by the range. If a S.D.or CV was estimated, a

0 0.10 0.05

1

0 0.15 0.05 1

R,, (mSd-l) GSD = 1.2 R, (m3 d-') GSD = 1.2 35

-

-

-

-

20

-

35

-

-

(YO)",

range, GSD

40

-

-

30 10 70

40 10 60 -

0

PV

of

0

AG

Ce (nBq m-9 C.V.= 0.88

INHALATION:

(CIA), S.D.

Soil water, (1-W.) S.D.

Time offsite (%) range Percent time indoors, T,

EXTERNAL: SF S.D.:Nuclide dependent (see Appendix C) CV = 0.2 Percent of time out, T , S.D. (%)

(a) LAND-USE DEPENDENT PararnekrNse:

TABLE 7.1-Summary

-

35

20

-

20

0.3 0.3 ----(0.1-0.9 GSD 30 30

-

400

1.3 0.1

10 0-20 50 0.05 0.05

40 10

=

0.1 to 0.4

RS

10

1.3 0.1

40 10 10 0-20 50 0.15 0.05

0.1 to 0.4

RV

400

1

-

0 0.05 0.05

-

70

30 10

0

PS

20

0.3 1.45)---25

50

1.3 0.1

40 10 10 0-20 50 0.15 0.05

0.1 to 0.4

SUISN

parameter values and associated uncertainty"

-

35

-

600

-1

0 0.05 0.05

-

30 10 70

0

CC

LN

LN

LN

LN

N

N

-

A

N

N

Diatributionb

%

u

Q

2 Z

E m

n

C13

S

z

a

1 Y

1

g

'+

7.1 DISTRIBUTION OF INDMDUAL DOSES

o

o

a

0

LO

5

0 1

0

2 2

%

r-&2&gl w

w

5

o i 3

5

L O C U

0

w

F)

~4

0

1

X r l

f

2

3

LO

0

0 % r - I

'El

" g *

8

8 'El

0

g

I

LO

8

LO 0 0

155

dependent (see Appendix D) dependent (see Appendix D) dependent (see Appendix D) dependent (see Appendix Cf dependent (see Appendix C) Max L0.65 (CIA)@, 0.34 (I/A)i.,P Max [1.0 (CIAL,, 1.1(YA),,,I' Max t0.7 (CIA)iw,0.3 (I/A)ingY

Nuclide Nuclide Nuclide Nuclide Nuclide

300 200 100 16 8

Nuclide dependent (see Appendix C) Nuclide dependent (see Appendix C) Nuclide dependent (see Appendix C, Section 4.3.5)

1.0 (mean) 1.6

Value

A =

triangular distribution.

% m

U LN

LN

3

5

Em

C) m

8

z

S

C)

-1

1

LN

LN

A

A

LN

LN

GSD = 1.2 GSD = 1.2 GSD = 1.15 Range 8-25 Range 4- 12 GSD = 2.5-3.0 (see Appendix D) GSD = 1.2-2.8 (see Appendix D) GSD = 1.6-2.5 (see Appendix D) GSD = 1.3-2.5'

Distributionb LN N N LN

Uncertainty S.D. = 0.5 S.D. = 0.1 CV = 0.1-0.3 (see Appendix C) GSD = 1.4-2.2

"See Equations 7.1 through 7.8 for explanation of symbols. bAssumed probability distribution: N = normally distributed, LN = lognomally distributed, 'See Section 5.2.7.5.

p (g

cm-9 ofmt[Sv (Bq kg")-'] Dfk(Sv Bq-') (clA),.h R , (kg Y-') R,,, (L y-') R-1 (kg Y-') Qmiu (kg d-I) Qm,, (kg d-I) B., B, (Bq kg-' per Bq kg-') F,,, (d kg-') Fmw(d L-l) Df,, (Sv Bq-') (clA)iw, (uAhog (CIA), (C/A),n (CIA)-,

S (Bq kg-' dry)

(b) LAND-USE INDEPENDENT Parameter

TABLE7.1-Summary of parameter values and associated uncertaintya (continued)

P

m

U1

7.1 DISTRIBUTION OF INDIVIDUAL DOSES

1

157

normal distribution was assumed, while if a GSD was specified, a lognormal distribution was assumed and the central tendency estimate was assumed to represent the median. Even if only a range rather than a particular distribution is specified in the text, whenever reasonable, a normal or lognormal distribution was assumed rather than a triangular distribution. The S.D. or GSD was estimated from the range by assuming that the maximum corresponds to about 3 S.D. or (GSDI3.The Monte Carlo analysis was carried out using a computer program that randomly selected a value for each parameter from its assumed probability distribution. These selectionswere then used to arrive a t an initial solution for each of the individual pathway doses as well as the total dose from all pathways for each landuse scenario. This process was repeated 5,000 times, resulting in distributions of individual pathway doses and total doses, from which could be extracted a median dose and the 5 and 95 percent quantities used for establishing the screening limits given in this Report. The results of each of these calculations for each nuclide is given in Appendix k 1 0 The equations used for calculating the screening doses are summarized below where E(z) = committed effective dose (Sv y-')." External:

E,,

=

S

X

DC

X

DL,, x W,X (CIA),,,

l00nly doses for the most critical group for each land-use scenario are tabulated in Appendix A, i.e., if the 96th percentile dose for a particular nuclide for a n agricultural site was higher for a child or infant ingesting food grown on the site than for a n adult who was exposed via all pathways, then only the ingestion dose for the child is tabulated and the doses from inhalation and external radiation are zem. For sitespecific assessments where children will not be the most exposed population group, the appropriate adult doses from all pathways may be estimated from the tabulated value for other land-use scenarios using the relative parameter values given in Table 7.1 "Note that a deterministic calculation of the dose using these equations will sometimes result in somewhat different values than the median of the distribution calculated stochastically. Usually the difference is <20 percent. However, for the ingestion pathway where the total dose is sometimes a sum of several doses, the deterministic estimate is sometimes significantly lower (50 to 80 percent) than the median dose. The reason for this is that combining (and in particular summing) central tendencies of distributions of parameters, particularly skewed distributions, does not always provide a good estimate of the median of the combination. For example, the median of a distribution that represents a sum of lognormal distributions is not equal to the sum of the medians.

158

/

S DC

Df,, SF

Tout Tin

w8

7. CALCULATION OF SCREENING DOSES = mean concentration in soil (Bq kg-' dry) = buildup and decay correction factor to convert S to aver-

age annual concentration in year of maximum parent plus progeny dose12 = adult external radiation dose factor (Sv d-' per Bq kg-') = childfadult correction factor -used only when calculating doses to infants and children = housing shielding factor = fraction of time (days per year) outdoors on contaminated land =fraction of time (days per year) indoors on contaminated land = soil moisture correction (dryhulk)

Inhalation:

Dfioh

= adult inhalation dose factor (Sv Bq-') = childadult correction factor-used only when calculat-

(Ar

= annual outdoor air concentration (Bq m-3 per Bq kg-')

ing doses to children (see text) breathing rate outdoors (m3d-') breathing rate indoors (m3d-') (intout) = ratio of concentration indoors versus outdoors

Rout Rin

= adult = adult

Ingestion:

(7.3) S X DC X Df,, x (CIA),, x I,, x Td Dfbg = adult ingestion dose factor (Sv Bq-') (C/A)bg = childadult correction factor -used only when calculating dose to infant or child (see text) = daily soil intake (kg d-') Lil

E,i1

=

',If the parent concentration a t time 0 is A,, the average annual concentration of the parent after a delay oft' days will be A, = A, [exp(- A1 t') - exp( - A, (t' + 365))Y (365 A,). The average annual concentration A, of the first daughter is given by A, = A, f [Az/(A2- A,)] [{exp( - A, t ' ) - exp( - A,(t' + 365))/(365 A,) - {exp(- A, t') exp( - Adt' + 366))/(366 Arrl where f is the fraction of parent decays to this daughter and A is ln(2)lhalf-life. See NCRP (1996) for a general formula to calculate the average concentrations of subsequent generations.

7.1 DISTRIBUTION OF INDMDUAL DOSES

1

159

per year exposed to soil from site (d)

Tg ,

= days

C,

= concentration in vegetables, fruit a t harvest (Bq kg-'

R, Tv, (CIA),,

= vegetable, fruit intake (kg wet y- ')

wet) = fraction of total vegetables, fruit from site = chilainfant correction factor used only when calculat-

-

ing dose to infant or child (see text)

cr

= concentration in fodder at time of harvest (Bq kg-'

F d k &milk

TQmilk

Rmdk Tmn

Fmat Qmeat

T&,; Rm,t Tm,t (CIA),,,

dry) =feed to milk transfer factor (d L-I) = milk cow feed intake (kg dry d-I) = fraction of total feed from site = annual adult milk intake (L y-I) = fraction of total milk from site = childhnfant correction factor -used only when calculating dose to infant or child (see text)

= feed

to meat transfer factor (d kg-')

= animal feed intake (kg dry d-') = fraction of total animal feed from = annual meat intake (kg y-I) = fraction of total meat from site

site

= chilainfant correction factor -used only when calcu-

lating dose to infant or child (see text)

C,

=

S x DC x (B,

+ B,)

(7.7)

B,

= root uptake factor (Bq kg-'

B,,

= resuspension/soil adhesion (Bq kg-' wet vegetable per Bq

wet vegetable per Bq kg-' dry soil)

kg-' dry soil)

Bf = root uptake factor (Bq kg-' dry fodder per Bq kg-' dry soil) Bfl = resuspension/soil adhesion (Bq kg-' dry fodder per Bq kg-' dry soil)

A sensitivity analysis to determine the parameters most affecting the total uncertainty showed that for all pathways, the uncertainty

160

/

7. CALCULATION O F SCREENLNG DOSES

in the measurement of average concentration in soil was a significant contributor to the total variance. For the external pathway, the variability in the time exposed outdoors is also important. The variations in radionuclide concentration in the soil and air and the uncertainty in the dose factor dominate for the inhalation pathway. For the ingestion pathway, the variations in soil concentration and in transfer from soil to vegetation and from feed to milk and meat and the uncertainty in dose factor dominate. The variability in the other parameters contributed very little to the overall variance in total doses. Known correlations were accounted for in the analysis. For example, the same realizations (set of Monte Carlo random numbers) for concentrations in soil were used for all pathways and for all decay products. The same realization was used for the fraction of time indoors and outdoors for the external and inhalation pathways, and the same dose factor was used for calculating the dose from each ingestion component. The individual pathway and total doses were all calculated using the same realizations. If a decay product of the same element was formed, the same realizations were used for parameters (i.e., soil to plant transfer factors, etc.) that depend only on chemical behavior. The time spent outdoors and offsite were sampled and the sum subtracted from one to calculate the time spent indoors. Detailed examples of the results (ratios of 2.5, 5, 95 and 97.5 percentiles of distribution to median) of the Monte Carlo calculation for 137Cs, 'Sr and239Puare shown in Table 7.2. Results are presented for each land-use scenario for each dose pathway and ingestion subpathway as well as for the total dose. The ratio of the mean to median for each distribution is also shown. The values in Table 7.2 are normalized to the median to allow easier comparison of distribution shapes from one pathway and landuse scenario to another. The distribution of total dose from 13'Cs for the RV scenario is shown in Figure 7.1 to illustrate the skewed nature of the total dose distributions. The results in Table 7.2 illustrate the differences in the distributions of dose for each pathway and subpathway and also the difference a s a function of land use. Although the individual pathway ratios do not depend strongly on land use, the ratios for the total dose do, reflecting the different mix of dominant dose pathways for different land-use scenarios. The broadest distributions are for the dose from ingestion of contaminated food, followed by those for the dose from inhalation, while the external dose pathway has the least uncertainty and variability. These results show the advantage of

7.1 DISTRIBUTION OF INDMDUAL DOSES

/

161

Table 7.2-Ratio of various percentiles and mean to median dose." Percentile

AG

W

PS

NA NA NA NA NA

0.15 0.19 1.0 6.3 8.6 1.9

0.15 0.20 1.0 6.3 9.0 1.9

RV

RS

SU

SN

CC

Ce-137 External 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Inhalation 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Ingestion-vegetable 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Ingestion-milk 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0%

0.13 0.10 NA 0.18 0.15 NA 1.0 1.0 NA 6.6 7.2 NA 9.6 10.4 NA 1.9 2.1

NA NA NA NA NA

NA NA NA NA NA

Ingestion-mil 2.5% 0.08 0.08 0.08 0.05* 0.05* 0.05* 0.05* 0.08 5.0% 0.12 0.12 0.12 0.07* 0.07* 0.07* 0.07* 0.11 50.0% 1.0 1.0 1.0 1.0* 1.0* 1.0' 1.0* 1.0 95.0% 8.4 8.4 8.4 12* 14* 12* 12* 8.4 97.5%13 13 13 19* 21* 19* 19* 13 rnean/50.0% 2.3 2.3 2.3 3.1* 3.3* 3.1* 3.1* 2.2 Total ingestion 2.5%

5.0% 50.0% 95.0% 97.5% mean/50.0%

0.11 0.16 1.0 6.6 9.0 2.0

0.16 0.21 1.0 5.9 8.1 1.8

0.16 0.22 1.0 5.8 8.4 1.8

0.18 0.24 1.0 4.8 6.4 1.6

0.18 0.12 0.05* 0.08 0.24 0.16 0.07* 0.11 1.0 1.0* 1.0 1.0 6.6 12* 8.4 4.7 9.5 19* 13 6.7 1.4 1.9 3.1* 2.2

162

1

7. CALCULATION OF SCREENING DOSES

Table 7.2-Ratio of various percentiles and mean to median dosea (continued) Percentile

AG

W

PS

RV

RS

SU

SN

CC

2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0%

0.31 0.38 1.0 2.7 3.3 1.2

0.28 0.35 1.0 3.4 4.5 1.3

0.27 0.34 1.0 3.9 5.4 1.4

0.36 0.42 1.0 2.5 3.0 1.2

0.36 0.41 1.0 2.5 3.2 1.2

0.37 0.43 1.0 2.4 2.7 1.1

0.37 0.43 1.0 2.3 2.7 1.1

0.26 0.34 1.0 2.5 3.1 1.2

2.5% 5.0% 50.0% 95.0% 97.5% mead50.W

0.31 0.37 1.0 2.5 2.9 1.1

0.24 0.33 1.0 2.6 3.1 1.2

0.24 0.32 1.0 2.6 3.1 1.2

0.36 0.44 1.0 2.3 2.6 1.1

0.36 0.43 1.0 2.3 2.6 1.1

0.36 0.43 1.0 2.3 2.6 1.1

0.36 0.43 1.0 2.3 2.6 1.1

0.24 0.32 1.0 2.6 3.1 1.2

2.5% 0.14 0.12 0.11 5.0% 0.19 0.17 0.17 50.0% 1.0 1.0 1.0 95.0% 5.0 5.3 5.4 97.5% 6.7 7.3 7.4 rnean/50.0% 1.6 1.6 1.7

0.13 0.19 1.0 5.5 7.0 1.7

0.13 0.18 1.0 5.1 6.7 1.6

0.13 0.19 1.0 5.0 6.7 1.6

0.13 0.19 1.0 6.1 6.7 1.6

0.12 0.18 1.0 5.3 7.0 1.7

0.10 0.10 0.13 0.14 1.0 1.0 6.7 6.4 9.1 8.9 1.9 1.9

0.11 0.15 1.0 6.5 9.3 2.0

NA NA NA NA NA

NA NA NA NA NA

NA NA NA NA NA

NA NA NA NA NA

NA NA NA NA NA

NA NA NA NA NA

NA NA NA NA NA

NA NA NA NA NA

Total dose

Sr-W)+ Y-90

External

Inhalation

Ingestion-vegetable

NA NA NA NA NA

NA NA NA NA NA

2.5% 5.0% 50.0% 95.0% 97.5% mead50.0%

0.09 0.14 1.0 6.5 9.8 2.0

2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0%

NA NA NA NA NA

2.5% 5.0% 50.0% 95.0% 97.5% mead50.0%

NA 0.10 0.10 0.09 NA 0.14 0.13 0.13 NA 1.0 1.0 1.0 NA 7.0 7.6 7.8 NA11 11 12 2.0 2.1 2.1

Ingestion-milk 0.08 0.09 0.08 0.08 0.13 0.14 0.13 0.12 1.0 1.0 1.0 1.0 7.7 8.2 7.7 8.0 11 12 12 12 2.2 2.2 2.2 2.2

Ingestion-meat 0.07 0.10 1.0 8.4 13 2.3

Ingeetion-soil 2.5% 0.08 0.08 0.08 0.05* 0.05* 0.05* 0.05* 0.08 5.0% 0.12 0.12 0.12 OM* O.W* 0.07* 0.07* 0.11 50.0% 1.0 1.0 1.0 1.0* 1.0* 1.0* 1.0* 1.0 95.09'0 8.4 8.4 8.4 12* 14* 12* 12* 8.4 97.5%13 13 13 19* 21* 19* 19* 13 mead50.0% 2.3 2.3 2.3 3.1* 3.3* 3.1* 3.1* 2.2

7.1 DISTRIBUTION OF INDIVIDUAL DOSES

1

163

Table 7.2-Ratw of various percentiles and mean to median dose~continued) Percentile

Total ingestion 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Total dose 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0%

Pu-239 External 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Inhalation 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Ingestion-vegetable 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Ingestion-milk 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0% Ingestion-meat 2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0%

AG

PV

PS

RV

RS

SU

SN

CC

164

/

7. CALCULATION OF SCREENING DOSES

Table 7.2-Ratio of various percentiles and mean to median dosea(continued) Percentile

AG

PV

PS

RV

RS

SU

SN

CC

Ingestion-soil 2.5% 0.06 0.06 0.06 0.04* 0.04* 0.04* 0.04* 0.06 5.0% 0.09 0.09 0.09 O.M* 0.07* O.M* 0.06* 0.09 50.0% 1.0 1.0 1.0 1.0* 1.0* 1.0* 1.0* 1.0 95.0%11 11 11 15* 16* 15* 15* 12 97.5% 18 16 18 27* 25* 27* 27* 18 mean/50.0% 2.9 2.7 2.9 4.4' 4.1* 4.4* 4.4* 3.0 Total ingestion 2.5% 5.W? 50.0% 95.0% 97.5% mean/50.0%

0.12 0.10 0.10 0.16 0.15 0.14 1.0 1.0 1.0 5.5 8.8 9.4 7.4 13 14 1.7 2.4 2.5

0.12 0.16 1.0 5.9 8.6 1.7

0.12 0.11 0.04* 0.06 0.17 0.16 0.06* 0.23 1.0 1.0* 1.0 1.0 15* 12 6.4 6.2 9.8 27* 18 8.8 1.8 1.8 4.4* 2.9

2.5% 5.0% 50.0% 95.0% 97.5% mean/50.0%

0.17 0.22 1.0 5.0 6.6 1.7

0.14 0.19 1.0 5.3 7.2 1.7

0.19 0.24 1.0 4.1 5.2 1.4

Total dose 0.17 0.22 1.0 8.0 7.0 1.7

0.14 0.19 1.0 5.1 7.0 1.6

"Ratios are for adult doses except when followed by bNA = not applicable.

0.19 0.25 1.0 4.4 5.6 1.5

0.18 0.24

1.0 4.6 6.0 1.6

0.13 0.19 1.0 5.0 6.9 1.6

*.

Fig. 7.1. Frequency distribution of total annual dose to a child from 13'Cs for RV land-use scenario.

7.3 SITE-SPECIFIC DOSE ASSESSMENTS

1

165

calculating separate limits for each nuclide in setting screening criteria based on land use.

7.2 "Maximum"Dose Estimates for Screening Based on the fact that the uncertainty estimates for the parameters used for calculating the screening doses in this Report are conservative, the total ranges of the dose distributions themselves are likely quite conservative. Thus, except under certain extraordinary circumstances, discussed in Section 2, it is safe to assume that no individual is likely to receive an annual dose in excess of the dose represented by the 95th percentile of the Monte Carlo calculations [see NCRP Report No. 76, page 225 (NCRP, 198413) for a discussion of the effect of bias on the shape of the distribution]. This is particularly true since it is unlikely that the same individual will be a member of the most exposed population group for all pathways a t any given site as is assumed for the screening dose calculations which sum the doses from all pathways. For example, few persons will derive both all their milk and all their meat from the same site, etc. Therefore, the 95th percentile dose was chosen as a reasonable estimate of the "maximum" dose to an individual of the most critically exposed population from all pathways. This dose was used to determine the screening guidance given in Section 2.

7.3 Site-Specific Dose Assessments It is important to carry out an adequate uncertainty analysis to account for the variability in the values of the various pertinent parameters in estimating the dose to a particular individual or group of individuals for a specific site. However, the range of possible doses will usually be much narrower and better defined since some of the uncertainty due to variability in life style, diet, etc. will be removed. The total dose distribution may still have a large range, however. This will be particularly true if the ingestion pathway is critical, due to the large uncertaintylvariations in human and animal biokinetic factors that even a site-specific study cannot reduce.

Q,

6,

4.93 - 09 1.2E - 08 O.OE + 00 O.OE + 00 9.53 - 07 2.33 - 06 1.4E - 06 3.33 - 06 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Be-7

C1-36

S-35

Si-32

A1-26

Na-22

Be-10

External

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median:

Agricultural (AG)

2.OE - 15 1.5E - 14 O.OE + 00 O.OE + 00 2.OE - 13 1.5E - 12 3.53 - 12 2.7E - 11 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Inhalation

4.53 - 12 4.4E - 11 2.OE - 09 2.1E - 08 1.9E - 08 2.OE - 07 2.53 - 09 2.63 - 08 9.OE - 07 1.1E - 05 4.53 - 08 5.63 - 07 5.63 - 06

Veg.

Soil

1.4E - 13 1.8E - 12 O.OE + 00 O.OE + 00 6.7E - 11 8.7E - 10 8.4E - 11 l.lE - 09 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Meat

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Milk

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Ingestion

Calculated Screening Doses [Sv (Bq kg-')-']"

APPENDIX A

Total Ingestion

Totalb

Delay, d'

median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %:

External

Agricultural (AG) (continued) Inhalation

Veg.

Meat

Ingestion

Milk Soil

Total Ingestion

52

$

2u

\

0

0

0

0

0

0

0

0

0

o *

36500

O

Delay, d'

median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

median: 95th %: Tc-99 median: 95th %: median: Rh-101 95th %: Rh-102111 median: 95th 8: Rh-102d median: 95th %: median: Ru-103 95th %: median: Ag-105 95th %: median: Ru-106 95th %: median: Pd-107 95th %: Ag-108m median: 95th %: Cd-109 median: 95th %: Ag-1lorn median: 95th %:

Tc-98

Nuclide

External

Agricultural (AG) (continued) Inhalation

Veg.

Milk Meat

Ingestion

Soil

Total Ingestion

Delay, dc

CL

\

0

45

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %:

median: 95th %:

median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Agricultural (AG) (continued) Inhalation Veg.

Milk

Meat Soil

Total Ingestion

Delay, dc

-.

h3

k

Pm-148m

Gd-148

Eu-148

Sm-147

Pm-147

Sm-146

Pm-146

Gd-146

Sm-145

Pm-145*

Pm-144

Ce-144

Pm-143

Ce-141

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 96: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

4.43 - 09 l.lE-08 9.1E - 08 2.23 - 07 1.9E - 08 4.63 - 08 5.33 - 07 1.3E-06 7.53 - 09 2.OE - 08 1.3E - 08 3.OE - 08 2.53 - 07 6.23 - 07 3.33 - 07 7.93 - 07 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 2.23 - 07 5.63 - 07 O.OE + 00 O.OE + 00 2.OE - 07 4.93 - 07

Nuclide

median: 95th %:

median: 95th %:

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Agricultural (AG) (continued) Inhalation

Veg.

Meat

Ingestion Milk

Soil

Total Ingestion

Delay, dc

1

6 4

F

median: 95th %: Ho-166111 median: 95th %: Yb-169 median: 95th %: median: Tm-170 95th %: median: Tm-171 95th %: Hf-172 median: 95th %: median: Lu-173 95th %: Lu-174m median: 95th %: L ~ - 1 7 4 ~ median: 95th %: Hf-175 median: 95th %: median: Lu- 176 95th %: Lu-177m median: 95th %: Hf-178m median: 95th %: median: Ta-179 95th %:

Tb-160

1.5E - 07 3.63 - 07 8.23 - 07 2.OE - 06 1.7E - 08 3.93-08 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 7.73 - 07 1.9E-06 4.1E - 08 1.OE -07 1.3E - 08 3.23-08 4.63 - 08 l.lE-07 4.63 - 08 l . l E -07 2.33 - 07 5.63 - 07 2.33 - 07 5.73 - 07 l.lE - 06 2.83-06 9.43 - 09 2.53 - 08

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Agricultural (AG) (continued) Inhalation

Veg.

+

+

+

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE 00 O.OE+OO O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Milk

+

+

O.OE 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE+OO O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

+

Meat

Ingestion

2.OE - 11 2.7E - 10 4.3E - 12 5.4E - 11 O.OE + 00 O.OE + 00 1.3E - 10 1.4E - 09 1.5E - 11 1.8E - 10 2.7E - 11 3.4E - 10 3.6E - 12 4.9E- 11 4.3E - 12 5.8E - 11 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Soil

4.3E - 10 3.9E - 09 1.2E - 10 l.lE-09 7.1E - 09 7.5E - 08 3.4E - 09 2.3E - 08 3.1E - 10 2.6E - 09 3.1E -08 3.8E - 07 4.OE - 09 5.2E - 08 7.5E - 10 8.9E - 09 3.4E - 08 3.5E - 07 1.4E - 07 1.7E - 06 3.3E - 10 4.4E - 09 2.2E - 07 2.OE - 06

Total Ingestion

Delay, dc

1

median: 95th %: 1 ~ 1 9 2 ~ median: 95th %: median: Pt-193 95th %: Hg- 194 median: 95th %: Ir-194m median: 95th %: 0s-194 median: 95th 8: median: Au-195 95th %: median: Pb-202 95th %: median: Hg-203 95th %: median: T1-204 95th %: median: Pb-205 95th %: median: Bi-207 95th %: median: Bi810m 95th %: Pb-210d median: 95th %:

Ir-192m

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Agricultural (AG) (continued) Inhalation

Veg.

Milk

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Meat

Ingestion Soil

4.23 - 07 2.53 - 06 3.OE - 06 1.4E - 05 2.73 - 06 1.5E - 05 1.3E -05 9.43 - 05 l.lE-06 8.63 - 06 2.63 -06 1.9E - 05 2.23 - 06 1.OE - 05 4.33 - 06 2.43 - 05 1.6E-05 l.lE-04 1.5E - 06 9.23 - 06 2.63 - 07 1.6E - 06 2.63 - 08 1.5E -07

Total Ingestion

4.2E - 07 C 2.53 - 06 9.6E - 06 A 2.73 - 05 2.7E - 06 C 1.5E - 05 1.3E-05 C 9.43 - 05 l.lE-06 C 8.63- 06 2.6E-06 C 1.9E-05 2.2E - 06 C 1.OE - 05 4.3E - 06 C 2.43 - 05 1.6E -05 C 1.1E-04 1.5E- 06 C 9.23 - 06 2.6E - 07 C 1.6E - 06 2.7E - 08 A 1.5E-07

Totalb

365000

365000

3650

18700

73000

365000

90

0

0

90

36500

0

Delay, dc

8*

2!

%Cd

1

00

4

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

O.OE + 00 O.OE + 00 8.2E - 08 2.OE - 07 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 4.8E - 11 1.3E - 10 1.OE - 07 2.63 - 07 3.53 - 09 8.93 - 09 4.6E - 11 1.2E - 10 1.4E - 08 3.3E - 08 4.2E - 11 1.OE - 10 4.7E - 11 1.l E - 10 8.43 - 09 2.1E- 08 2.9E - 08 7.33 - 08 2.3E - 10 6.1E - 10

External

5.1E - 09 1.2E - 08 O.OE + 00 O.OE + 00 3.8E - 11 1.OE - 10 9-73- 08 2-43- 07 5.63 - 08 1.4E - 07 4.6E - 11 1.2E - 10 1.6E - 07 3.93 - 07 4.33 - 08 1.1E - 07 4.3E - 11 1.2E - 10 4.6E - 08 l . l E - 07 1.7E - 07 4.1E - 07 O.OE + 00 O.OE + 00

Nuclide

Am-242m median: 95th %: Cm-242d median: 95th %: Pu-242 median: 95th %: Am-243 median: 95th 8: Cm-243 median: 95th %: Cm-244d median: 95th %: Pu-244 median: 95th 8: Cm-245 median: 95th %: Cm-246 median: 95th %: Bk-247 median: 95th %: Cm-247 median: 95th %: Cf-248 median: 95th %.

Agricultural (AG) (continued) Veg.

7.43 - 08 3.73 - 07 4.23 - 09 2.63 - 08 7.43 - 08 4.73 - 07 6.43 - 08 4.03 - 07 4.43 - 08 3.53 - 07 3.63 - 08 2.73 - 07 8.43 - 08 4.63 - 07 1.33 - 07 6.93 - 07 6.53 - 08 5.03 - 07 1.13- 07 9.23 - 07 6.93 - 08 4.63 - 07 1.2E - 08 8.73 - 08

Inhalation

8.1E - 09 4.4E -08 O.OE + 00 O.OE + 00 8.83 - 09 4.4E - 08 7.23 - 09 4.33 - 08 5.43 - 09 3.1E - 08 4.73 - 09 2.83 - 08 9.OE - 09 5.33 - 08 1.4E - 08 7.33 - 08 7.53 - 09 4.43 - 08 1.2E - 08 8.9E -08 8.OE - 09 4.5E -08 O.OE + 00 0.03 + 00

Meat

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 0.03 + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 0.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Ingestion

Milk Soil

Total Ingestion Totalb

Delay, d'

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

95th 8:

median:

95th %:

median:

95th %:

median: 95th %: median: 95th %: median:

95th 8:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th 8:

median:

95th %:

median: 95th %: median:

External

Inhalation

Heavily vegetated pasture (PV) Veg.

Meat

Ineestion Milk

Soil

Total Ingestion

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %. median: 95th %: median: 95th '?k median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

O.OE + 00 O.OE+OO O.OE + 00 O.OE + 00 2.23 - 07 5.63 - 07 O.OE + 00 O.OE+OO 4.33 - 07 l.lE-06 O.OE + 00 O.OE+OO 1.OE - 07 2-63- 07 8-43- 08 2.23 - 07 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Heavily vegetated pasture (PV) (continued) Veg.

Meat

Inzestion

Milk Soil

Total Ingestion

Delay, dc

1

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 96th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Ru-103

Nuclide

median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

2.3E - 14 1.4E - 13 8.1E - 15 6.2E - 14 1.3E - 12 7.8E - 12 O.OE + 00 O.OE + 00 3.7E - 13 2.7E - 12 O.OE + 00 O.OE + 00 3.1E - 13 2.2E - 12 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 1.2E - 13 8.7E - 13 O.OE + 00 O.OE + 00

Inhalation

Heavily vegetated pasture (PV) (continued) Veg.

Meat

Ingestion Milk

Soil

Total Ingestion Total

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 96:

External

Inhalation

Heavily vegetated pasture (PV)(continued) Veg.

Milk Meat

Ingestion Soi1

Total Ingestion

0

0

0

Delay, dc

rD

8!2

%'-a

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Tm-171

Tm-170

Yb-169

Ho-166m

Tb-160

Dy-159

Tb-158

Tb-157

Eu-155

Eu-154

Gd-153

Gd-152

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %.

Inhalation

O.OE + 00 O.OE + 00 8.63 - 14 6.43 - 13 3.23 - 12 2.6E - 11 4.23 - 13 3.23 - 12 7.93 - 14 6.OE - 13 3.OE - 12 2.3E - 11 l . l E - 14 8-63- 14 1.3E - 13 9.63 - 13 8.OE - 12 5.9E - 11 2.53 - 14 2.OE - 13 O.OE + 00 O.OE + 00 7.73 - 14 5.93 - 13

External

O.OE + 00 O.OE + 00 1.9E - 08 4.83 - 08 4.63 - 07 1.2E - 06 1.8E -08 4.83 - 08 5.8E - 10 1.5E -09 3.OE - 07 7.43 - 07 4.93 - 09 1.4E - 08 1.2E - 07 3.OE - 07 6-53- 07 1.7E - 06 1.3E - 08 3.4E - 08 O.OE + 00 O.OE + 00 1.5E - 10 3.9E - 10

Veg.

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE+ 00 O.OE + 00 O.OE + 00

Heavily vegetated pasture (PV)(continued)

1.9E - 09 2.53 - 08 2.43 - 12 3.OE - 11 2.8E - 11 3.5E - 10 4.33 - 12 5.8E - 11 4.83 - 13 6.53 - 12 1.6E - 11 2.OE - 10 6.63 - 13 9.33 - 12 6.63 - 12 7.9E - 11 3.OE - 11 3.9E - 10 1.3E - 12 1.6E - 11 6.9E - 11 9.OE - 10 1.4E-12 1.8E - 11

4.53 - 09 6.73 - 08 2.OE- 11 3.OE - 10 2.4E- 10 3.1E - 09 3.6E - 11 5.1E- 10 4.1E - 12 5.8E - 11 1.3E- 10 1.8E - 09 5.73- 12 8.4E - 11 5.4E - 11 7.5E - 10 2.4E - 10 3-53- 09 1.1E- 11 1.5E- 10 1.5E - 10 2.33 - 09 1.1E-11 1.4E - 10

Meat

Ingestion

Milk

O.OE + 00 O.OE + 00 2.OE - 12 2.5E- 11 2.3E - 11 3.1E - 10 3.6E - 12 4.8E - 11 3.9E - 13 5.OE - 12 1.3E - 11 1.7E - 10 5.8E - 1 3 7.3E - 12 5.4E - 12 7.2E - 11 2.4E - 11 3.2E - 10 1.1E - 12 1.4E - 11 O.OE + 00 O.OE + 00 l . l E - 12 1.6E - 11

Soil

7.83 - 09 8.9E - 08 3.OE - 11 3.5E - 10 3.6E - 10 3.73 - 09 5.2E - 11 6.2E - 10 5.9E - 12 6.7E - 11 1.9E - 10 2.OE - 09 8.4E - 12 9.9E - 11 7.9E - 11 8.7E - 10 3.5E - 10 4.1E -09 1.6E - 11 1.7E - 10 2.6E - 10 3.1E-09 1.6E - 11 1.7E - 10

Total Ingestion

Delay, dc

\

Ta-182d

Hf-182

W-181

Hf-181

Ta-180

Ta-179

Hf-178m

Lu-177m

Lu-176

Hf-175

Lu- 174*

Lu-174m

Lu-173

Hf-172

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

median:

External Inhalation

Heavily vegetated pasture (PV)(continued) Veg.

Meat

Ingestion

Milk Soil

Total Ingestion

9

0

0

0

0

0

720

0

0

0

0

3 0

54

%"u

m

1

0

Delay, dC

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00

O.OE + 00 O.OE + 00 1.3E - 06 4.OE - 06 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 2.OE - 14 1.3E - 13 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

O.OE + 00 O.OE + 00 2.4E - 07 6.23 - 07 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 2.9E - 10 7.5E - 10 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

median: 95th %: median: Th-230 95th %: median: Pa-231 95th %: Th-232 median: 95th %: median: U-232 95th %: median: U-233 95th %: median: U-234 95th %: median: Np-235 95th %: median: U-235 95th %: Np-236A median: 95th %: P ~ - 2 3 6 ~ median: 95th %: median: U-236 95th %:

Th-229

Veg.

Inhalation

External

Nuclide

Heavily vegetated pasture (PV)(continued)

7.73 - 07 6.33 - 06 1.4E - 07 6.73 - 07 7.OE - 07 4.73 - 06 3.93 - 06 2.73 - 05 6.1E - 07 3.63 - 06 l . l E - 07 6.43 - 07 2.OE - 08 1.6E - 07 1.3E - 13 1.2E - 12 4.4E - 08 2.43 - 07 6.43 - 08 3 - 3 3- 07 2.43 - 08 1.3E - 07 2.OE - 08 1.6E - 07

5.73 - 08 4.63 - 07 1.3E - 07 6.1E - 07 4.93 - 08 3.83 - 07 4.OE - 07 3.23 - 06 6.83 - 08 4.OE - 07 l . l E - 08 6.43 - 08 3.63 - 09 2.93 - 08 3.23 - 12 2.7E - 11 5.53 - 09 3.43 - 08 7.73 - 09 4.1E - 08 2.63 - 09 1.5E - 08 3.43 - 09 3.OE - 08

Meat

Ingestion Milk

O.OE + 00 O.OE + 00 1.3E - 08 1.2E - 07 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 4.73 - 13 5.OE - 12 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Soil

Total Ingestion Total

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th%: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

8.23 - 08 2.1E-07 2.83 - 09 7.33 - 09 3.6E - 11 1.OE - 10 O.OE + 00 O.OE + 00 3.3E - 11 8.5E - 11 3.7E - 11 9.1E-11 6.53 - 09 1.7E - 08 2.33 - 08 6.OE - 08 1.8E - 10 5.1E - 10 4.93 - 09 1.3E - 08 2.2E - 11 6.OE - 11 3.OE - 11 8.4E - 11 7.53 - 08 1.9E - 07 4.43 - 08 1.2E - 07

Cm-250

Cf-250d

Cf-24gd

Bk-24gd

Cm-248

Cf-248

Cm-247

Bk-247

Cm-246

Cm-245

Pu-244

3.6E - 11 1.OE- 10 1.3E - 07 3.33 - 07 3.33 - 08 8.53 - 08 3.3E - 11 9.6E - 11 3.63 - 08 9.43 - 08 1.3E - 07 3.53 - 07 2.7E - 11 7.4E - 11 2.5E - 11 7.OE - 11 2.5E - 10 6.5E- 10 1.2E - 07 3.23 - 07 3.4E - 11 9.5E - 11 1.6E-07 4.23 - 07

Cm-244d

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Nuclide

1.7E - 09 l . l E - 08 3.43 - 09 2.OE - 08 4.83 - 09 2.83 - 08 2.73 - 09 1.6E - 08 4.43 - 09 3.43 - 08 2.83 - 09 1.6E - 08 4.5E - 10 3.23 - 09 9.73 - 09 5.83 - 08 9.53 - 12 7.2E- 11 4.83 - 09 3.43 - 08 2.23 - 09 1.6E- 08 5.1E-08 3.23 - 07

Inhalation

2.9E - 11 3.1E - 10 2.1E - 10 1.3E - 09 1.1E - 10 6.5E - 10 5.2E - 11 5.4E - 10 8.8E - 11 1.OE- 09 5.4E - 11 5.OE - 10 5.83 - 12 5.2E - 11 1.8E - 10 2.OE - 09 1.8E - 13 1.8E- 12 9.OE - 11 1.OE - 09 3.9E - 11 3.2E - 10 l.lE-08 1.2E - 07

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE+OO O.OE +00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 6.7E - 11 8.3E - 10 8.8E - 10 4.23 - 09 4.3E - 10 2.23 - 09 1.3E - 10 1.6E - 09 2.2E - 11 2.6E - 10 1.7E - 10 1.4E - 09 1.4E - 11 1.6E - 10 4.5E - 10 5.53 - 09 1.3E - 12 1.5E- 11 6.4E - 10 7.53 - 09 2.8E - 10 2.43 - 09 2.63 -09 3.33 - 08

Meat

Ingestion Milk

Veg.

Heavily vegetated pasture (PV)(continued) Soil

Total Ingestion Total

Delay, dc

\

Md-258

Fm-257

Es-254

Cf-254

Cf-252

Cf-251

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

SC-46

Ca-45

Ti-44

Ca-41

K-40

C1-36

S-35

Si-32

Al-26

Na-22

Be-10

Be-7

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: 4.OE - 09 9.9E - 09 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 l.lE - 06 2.9E - 06 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 6.5E - 08 1.7E -07 O.OE + 00 O.OE + 00 8.9E - 07 2.3E - 06 O.OE + 00 O.OE + 00 2.6E - 07 6.8E - 07

External

1.5E - 14 1.1E - 13 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 2.6E - 11 2.1E - 10 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 1.6E - 10 1.2E -09 O.OE + 00 O.OE + 00 2.8E - 12 2.1E - 11

Inhalation

Sparsely vegetated pasture (PS) Milk Meat

Ingestion

1.4E - 13 1.8E - 12 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 8.4E - 11 l.lE-09 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 1.4E - 10 1.8E - 09 O.OE + 00 O.OE + 00 1.lE- 11 1.5E - 10

Soil

Total Ingestion

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Lnhalation

Veg.

Sparsely vegetated pasture (PS) (continued) Meat

Ineestion Milk

Soil

Total Ingestion

Delay, dc

1

+

median: O.OE + 00 95th %: O.OE + 00 median: 3.3E - 11 95th %: 9.1E - 11 median: 3.3E - 11 95th %: 9.2E - 11 median: 6.33 - 07 95th %: 1.6E - 06 median: 4.33 - 08 95th %: l.lE-07 median: O.OE + 00 95th %: O.OE +00 median: 1.5E- 07 95th 6: 3.73 - 07 median: O.OE + 00 95th %: O.OE 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: 8.83 - 08 95th %: 2.23 - 07 median: l . l E - 07 95th %: 2.83 - 07 median: 7.33 - 07 95th %: 1.9E - 06

median: O.OE + 00 95th %: O.OE+OO median: O.OE + 00 95th 8: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th 8: O.OE + 00 median: O.OE + 00 95th %: O.OE +00 median: 1.8E - 07 95th %: 4.73-07 median: O.OE + 00 95th %: O.OE+OO median: 1.5E- 07 95th %: 3.73 - 07 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE+ 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation Veg.

Sparsely vegetated pasture (PSI (continued) Milk

Meat

Ingestion Soil

Total Ingestion

Delay, d'

1

median: 1.OE - 08 95th %: 2.73 - 08 median: 2.OE - 07 95th %: 5.43 - 07 median: 2.63 - 07 95th %: 6.83-07 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: 1.8E - 07 95th 8: 4.73 - 07 median: O.OE + 00 95th %: O.OE + 00 median: 1.6E - 07 95th %: 4.23 - 07 median: 6.43 - 09 95th%: 1.7E-08 median: 5.63 - 07 95th %: 1.5E - 06 median: 7.83 - 09 95th %: 2.OE - 08 median: O.OE + 00 95th %: O.OE + 00 median: 4.43 - 07 95th %: 1.2E-06

Nuclide

median: 95th 46: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External Inhalation

Veg.

Sparsely vegetated pasture (PS)(continued) Ingestion Milk Meat Soil

Total Ingestion Total

0

0

0

0

0

0

0

0

0

* 0

!?

2 u

% Cd

\

0

0

Delay, dc

Ta-182d

Hf-182

W-181

Hf-181

Ta-180

Ta-179

Hf-178m

Lu-177m

Lu-176

Hf- 175

Lu-174d

Lu-174m

Lu-173

Hf-172

median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

median: 95th %: median: Re-184d 95th %: 08-185 median: 95th %: W-185 median: 95th %: Re-186m median: 95th %: median: Re-187 95th %: W-188 median: 95th %: median: Ir-192m 95th %: 1 ~ 1 9 2 ~ median: 95th %: Pt-193 median: 95th %: Hg-194 median: 95th %: median: Ir-194m 95th %:

Re-184m

Nuclide Inhalation

O.OE + 00 O.OE+ 00 O.OE + 00 O.OE + 00 7.OE - 13 5.73 - 12 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 6.1E - 11 4.2E - 10 2-43- 12 2.OE - 11 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 8-73- 12 6.8E- 11

External

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 9-23- 08 2-43- 07 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 3.73 -07 9.73 - 07 8.9E - 08 2-33- 07 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 4-73- 07 1.2E-06

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE+OO

Veg.

Sparsely vegetated pasture (PSI (continued)

7.1E - 08 1.3E - 06 8.63 - 09 1.6E - 07 2.OE - 11 3.23 - 10 3.83 - 09 6.63 - 08 1.6E - 07 3.23 - 06 3.9E - 10 6.73 - 09 7.53 - 08 8.53 - 07 4.43 - 12 6.23 - 11 9.23 - 13 1.4E - 11 6.1E - 11 l.lE - 09 3.1E - 07 4.7E - 06 2.63 - 12 4.3E- 11 2-23- 08 4.23 - 07 2.53 - 09 4.73 - 08 8.9E - 11 1.6E - 09 3.2E - 08 4.83- 07 4.73 - 08 9.53 - 07 l . l E - 10 2.33 - 09 1.8E- 07 2.63 - 06 9.6E - 10 1.5E - 08 1.9E - 10 3.OE - 09 7.53 - 12 1.4E - 10 4.63 - 07 9.23 - 06 5.4E - 10 9.93-09

Meat

Ingeetion Milk

O.OE + 00 O.OE + 00 O.OE + 00 0.03 + 00 4.33 - 12 5.8E - 11 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 4.8E - 11 5.6E - 10 9.93 - 12 1.3E - 10 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 2.7E - 11 3.5E- 10

Soil

Total Ingestion Totalb

Delay, dc

\

+

median: 3-43- 08 95th %: 8.73 - 08 median: 1.4E - 08 95th %: 3.83 - 08 median: 1.8E - 07 95th %: 4.53-07 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th 8: O.OE + 00 median: 6.1E - 07 95th %: 1.6E - 06 median: O.OE + 00 95th %. O.OE + 00 median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th 8: O.OE+OO median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th 8: O.OE + 00 median: O.OE + 00 95th 96: O.OE 00 median: O.OE +00 95th %: O.OE + 00

U-236

Pu-236d

Np236A

U-235

Np-235

U-234

U-233

U-232d

Th-232

Pa-231

Th-230

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 3.93 - 13 2.5E - 12 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

+

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median: 95th %: median:

95th %:

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 3.OE - 10 7.9E - 10 O.OE +00 O.OE +00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Th-229

median: O.OE 00

Inhalation

External

Nuclide

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Veg.

Sparsely vegetated pasture (PS) (continued) Ingestion Milk Meat Soil

Total Ingestion Total

Delay, de

\

median: 8.53 - 08 95th %: 2.23 - 07 median: 2.93 - 09 95th %: 7.63 - 09 median: 3.8E - 11 95th 8: 1.OE - 10 median: O.OE + 00 95th %: O.OE + 00 median: 3.4E - 11 95th %: 8.9E - 11 median: 3.8E-11 95th 8: 9.5E - 11 median: 6.73 - 09 95th %: 1.8E - 08 median: 2.43 - 08 95th %: 6.33 - 08 median: 1.9E-10 95th 8: 5.2E - 10 median: 4.1E - 09 95th %: l . l E - 08 median: 2.3E - 11 95th %: 6.2E - 11 median: 3.1E - 11 95th %: 8.8E - 11 median: 7.93 - 08 95th %: 2.OE - 07 median: 4.63 - 08 95th %: 1.2E-07

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %:

External

Inhalation

Veg.

Sparsely vegetated pasture (PS) (continued) Meat

Ingestion Milk

Soi1

Total Ingestion

4.4E - 08 A 2.4E - 07 2.4E - 07 A 7.4E - 07 1.6E - 07 A 6.9E - 07 6.9E - 08 A 3.8E - 07 1.6E - 07 A 8.5E - 07 2.3E - 07 A 6.9E - 07 1.1E-08 A 6.6E - 08 2.4E - 07 A 1.4E - 06 5.4E - 10 A 2.1E-09 2.73 - 07 A 1.OE - 06 5.53 - 08 A 3.7E - 07 1.6E - 06 A 7.8E - 06

Total

0

0

0

1825

0

0

365000

0

0

365000

365000

0

Delay, dc

E

% 8

\

Md-258

Fm-257

Es-254

Cf-254

Cf-252

Cf-251

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Heavily vegetated rural (RV) Veg.

Milk Meat

Ingestion Soil

Total Ingestion

0

0

0

0

0

0

0

90

0

b

O

5;!

3

% 8

\

0

0

Delay, dc

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %.

+

O.OE + 00 O.OE+OO O.OE + 00 O.OE + 00 5.63 - 07 1.3E - 06 O.OE + 00 O.OE + 00 1.l E - 06 2.53 - 06 7.OE - 08 1.6E - 07 2.53 - 07 5.93 - 07 2.1E - 07 4.93 - 07 O.OE 00 O.OE + 00 2.43 - 06 5.63 - 06 2.53 - 06 5.83 - 06 O.OE + 00 O.OE + 00 3-63- 07 8.43 - 07 6.OE - 07 1.4E - 06

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Heavily vegetated rural (RV)(continued) Veg.

Meat

Ingestion Milk

Soil

Total Ingestion Total

Delay, dc

1

median: 3.6E - 10 95th %: 9.2E - 10 median: 5.9E - 11 95th %: 1.5E - 10 median: 6.1E - 11 95th %: 1.5E - 10 median: 1.5E - 06 95th %: 3.53 - 06 median: 1.OE - 07 95th %: 2.43 - 07 median: 1.6E - 07 95th %: 3.63 - 07 median: 3.53 - 07 95th %: 8.OE - 07 median: 2.1E - 10 95th %: 5.5E - 10 median: 4.6E - 10 95th %: 1.2E - 09 median: 1.4E - 06 95th %: 3.23 - 06 median: 3.5E - 11 95th %: 8.6E - 11 median: 2.1E - 07 95th 8: 4.83-07 median: 2.73 - 07 95th 8: 6.23 - 07 median: 1.8E - 06 95th %: 4.OE - 06

Nuclide median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 9%: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Heavily vegetated rural (RV)(continued) Veg.

Milk

Meat

Soil

Total Ingestion Totalb

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

1.OE - 10 2.5E - 10 1.1E - 09 2.93 - 09 1.OE - 09 2.63 - 09 3.43 - 09 8.1E - 09 5.33 - 08 1.2E - 07 3.53 - 09 9.23 - 09 4.33 - 07 1.OE - 06 2.1E - 09 5.OE-09 3.53 - 07 8.1E-07 1.7E - 09 4.43-09 1.9E - 06 4.23-06 2.93 - 09 6.93 - 09 6.53 - 09 1.5E -08 8.43 - 09 2.OE - 08

Nuclide

median: 95th %:

95th %:

median:

95th %:

median: 95th %: median: 95th %: median:

median: 95th %:

median: 95th %:

95th %:

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median:

External

Inhalation

Heavily vegetated rural (RV)(continued) Veg.

Ingestion Milk Meat Soil

Total Ingestion

3.5E - 07 C 8.1E- 07 1.5E - 06 C 3.4E - 06 2.3E - 08 C 1.2E - 07 6.73 - 07 C 1.6E - 06 6.9E - 09 C 1.7E - 08 1.2E - 06 C 2.8E - 06 6.1E - 08 C 1.4E - 07 8.8E - 09 C 2.OE - 08 1.8E - 07 C 4.2E - 07 4.3E - 08 C 1.OE - 07 1.OE - 06 C 2.4E - 06 l.lE - 08 C 2.9E - 08

Totalb

0

0

0

0

0

0

0

0

0

0

0

0

Delay, dc

*

851

%

\

0

h3

h3

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %:

median: 95th %:

median: 95th %:

95th %:

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median:

External

Inhalation

Heavily vegetated rural (RV)(continued) Veg.

Meat

Ingestion Milk

Soil

Total Ingestion

Delay, dc

\

median: 95th 96: median: 95th 8 : median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %median: 95th %: median: 95th %.

median: 95th %: Re-184m median: 95th %: Re-184d median: 95th %: median: 0s-185 95th %: W-185 median: 95th %: Re-186m median: 95th %: Re-187 median: 95th %: W-188 median: 95th %: Ir-192m median: 95th %: 1 ~ 1 9 2 ~ median: 95th %: median: Pt-193 95th %: Hg-194 median: 95th %:

Ta-182d

Nuclide

External

Inhalation

Heavily vegetated r u r a l (RV)(continued) Veg.

Ineestion Milk Meat Soil

Total Ingestion

Delay, dc

\

/

CALCULATED SCREENING DOSES

225

ma, me- ma, me- me- e-w o m cob e-w wua e-w wua (Dm uaua 00 00 00 00 0 0 00 00 00 0 0 0 0 00 00 7 4 0 00

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WW W W WW WW WW W W WW W W WW WW WW WW W W WW 1t"?=? N'? r!'? = ? 0E:c? 1 9 Lo'? *'? '9L- Cqo! L-9 '9* r!? + m cod m b e m m m m m wrl rlm mua +coma, rlw ma, rlm

$g.ig$sig$ggg$gig$g$gig$g$g*g . - A 3 5 3 5 3 5 3 5 3 4 3 A .-5 3 A 3 A 2 A 3 s 3 A 4 A % g a m a,', a m a m a g a i g Z u a aig a $ ig aig a & a $

a m a m a m am am am am a m a m a m am am am am

Nuclide

median: 95th %: median: 95th 8: median: 95th %: median: 95th 56: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Inhalation

(RV)(continued)

External

Heavily vegetated rural Veg.

Meat

Ingestion Milk

Soil

Total Ingestion

Delay, d c

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

7.1E-11 1.8E - 10 1.6E - 07 3.63 - 07 7.OE - 09 1.7E - 08 5.2E - 11 1.3E - 10 2.73 - 08 6.23 - 08 5.9E - 11 1.4E - 10 5.3E - 11 1.2E- 10 1.l E - 08 2.53 - 08 5.73 - 08 1.4E - 07 3.OE - 10 7.4E - 10 l . l E - 08 2.73 - 08 4.1E - 11 l . l E - 10 4.3E - 11 l . l E - 10 1.9E - 07 4.43 - 07

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Heavily vegetated rural (RV)(continued) Veg.

Meat

Ineestion Milk Soil

Total Ingestion

Delay, dc

\

Md-258

Fm-257

Es-254

Cf-254

Cf-252

Cf-251

Cm-250

median: 95th 8 .

95th %:

median: 95th %: median: 95th 8. median: 95th %: median: 95th %: median: 95th %: median:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Sparsely vegetated rural (RS) Veg.

Meat

Ineestion Milk

Soil

Total Ingestion

0

0

0

0

0

0

0

90

0

0

0

0

Delay, dc

8 9

5?

%w

\

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %. median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Sparsely vegetated rural (RS)(continued) Veg. Milk

Meat

Ingestion Soil

Total Ingestion

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Cd-115m

In-1 14m

Sn-113

Cd-113

Cd-113m

Ag-llOm

Cd-109

Ag-1O8m

Pd-107

Ru-106

Ag-105

7.43 - 08 1.7E - 07 8.63 - 08 1.9E - 07 1.6E - 07 3.73 - 07 O.OE + 00 O.OE+ 00 1.3E - 06 3.OE - 06 3.93 - 09 9.53 - 09 1.9E- 06 4.33 - 06 1.7E - 10 4.OE - 10 2.7E - 11 6.5E - 11 l.lE-07 2-43- 07 1.8E - 08 4.23 - 08 4.63 - 09 l . l E - 08

Ru-103

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Nuclide

8.93 - 13 5.2E - 12 3.53 - 13 2.63 - 12 4.7E - 11 2.7E - 10 1.2E - 12 9.33 - 12 1.4E - 11 9.7E - 11 1.7E - 11 1.2E - 10 1.2E - 11 8.5E - 11 2.OE - 10 1.5E - 09 2.3E - 10 1.7E - 09 2.6E-12 2.OE - 11 7.6E - 12 5.7E - 11 3.OE - 12 2.3E - 11

Inhalation

Veg.

6.3E - 10 4.83 - 09 7.5E - 11 6.OE - 10 3.OE - 08 2.43 - 07 7.7E - 10 9.43 - 09 1.6E - 09 1.3E - 08 9.73 -08 1.2E -06 1.6E -09 1.3E - 08 7.53-07 9.1E -06 9.63 - 07 1.3E - 05 1.7E-08 2.OE - 07 9.8E - 10 8.73-09 4.53 -08 5.83-07

Sparsely vegetated rural (RS)(continued)

1.2E - 11 1.OE - 10 1.2E - 09 1.4E - 08 5.8E - 10 4.63 - 09 5.2E - 11 9.5E - 10 2.6E - 08 2.83 - 07 5.OE - 08 8.8E - 07 2.7E - 08 3.OE - 07 3.53 - 07 6.73 - 06 1.5E - 07 2.73 - 06 6.33-09 1.2E - 07 2.6E - 10 2.83 - 09 2.33 - 08 4.33- 07

2.33 - 12 2.7E - 11 1.2E - 12 1.4E - 11 l . l E - 10 1.4E - 09 1.9E - 13 3.63 - 12 2.3E - 11 3.OE - 10 4.4E - 11 8.3E - 10 2.5E - 11 3.OE - 10 4.1E - 10 8.63 - 09 4.9E - 10 9.73 - 09 1.2E- 10 2.83 - 09 9.63 - 13 1.4E - 11 2.1E - 11 4.4E - 10

Meat

Ingestion Milk

1.1E - 11 1.8E - 10 7.OE - 12 1.3E - 10 5.1E - 10 7.83 - 09 3.8E - 12 7.5E - 11 1.7E- 10 3.1E - 09 1.3E - 10 2.331 - 09 1.6E- 10 2.93 - 09 1.4E - 09 2.53 - 08 6.1E - 10 7.63 - 09 3.1E- 11 5.8E - 10 8.4E- 11 1.4E- 09 5.9E - 11 1.2E- 09

Soil

7.OE - 10 5.OE - 09 1.4E - 09 1.4E - 08 3.43 - 08 2.53 - 07 9.3E - 10 1.OE - 08 3.OE - 08 3.OE - 07 1.9E - 07 2.OE - 06 3.OE - 08 3.1E - 07 1.5E - 06 1.5E - 05 1.4E - 06 1.6E - 05 3.OE-08 3.23 - 07 1.7E - 09 1.3E - 08 9.1E -08 1.OE - 06

Total Ingestion

7.6E - 08 C 1.7E - 07 8.9E - 08 C 2.OE - 07 2.1E - 07 C 5.53 - 07 9.3E - 10 C 1.OE - 08 1.4E-06 C 3.2E - 06 2.OE - 07 C 2.OE - 06 2.OE - 06 C 4.53 - 06 1.5E - 06 C 1.5E - 05 1.4E - 06 A 1.6E - 05 1.5E-07 C 4.93 - 07 2.1E - 08 C 5.1E-08 9.6E - 08 C 1.OE -06

Totalb

E

5 z!

y

0

0

0

0

0

0

0

0

0

o >

0

O

Delay, dc

tu o +

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

l . l E - 10 2.8E - 10 1.2E - 09 3.23 - 09 l . l E - 09 2.83 - 09 3.83 - 09 8.93 - 09 5.93 - 08 1.4E-07 3.83 - 09 1.OE - 08 4.83 - 07 l.lE-06 2.33 - 09 5.53 - 09 3.83 - 07 8.83 - 07 1.9E - 09 4.83 - 09 2.OE - 06 4.63 - 06 3.1E - 09 7.63 - 09 7.1E - 09 1.7E - 08 9.33 - 09 2.1E - 08

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %

External

Inhalation

Sparsely vegetated rural (RS)(continued) Veg.

Meat

Ingestion Milk

Soil

Total Ingestion Total

Delay, dc

1

median: 95th %: Gd-146 median: 95th %: median: Pm-146 95th %: median: Sm-146 95th %: median: Pm-147 95th %: Sm-147 median: 95th %: Eu-148 median: 95th %: Gd-148 median: 95th %: Pm- 148m median: 95th %: Eu-149 median: 95th %: Eu-150B median: 95th %: median: Gd-151 95th 8: median: Sm-151 95th %: Eu-152 median: 95th %:

Sm-145

+

2.1E - 08 4.83 - 08 5.43 - 07 1.3E-06 7.1E - 07 1.6E-06 O.OE 00 O.OE + 00 1.4E - 11 3.4E - 11 O.OE + 00 O.OE + 00 4.93 - 07 l.lE-06 O.OE + 00 O.OE + 00 4.43 - 07 1.OE-06 1.6E - 08 3.73-08 1.5E - 06 3.53 - 06 1.9E - 08 4.43 - 08 6.1E - 1 3 1.5E - 12 1.2E - 06 2.83 - 06

Tm-171

Tm-170

Yb-169

Ho-166m

Tb-160

Dy-159

Tb-158

Tb-157

Eu-155

Eu-154

Gd-153

Gd-152

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

O.OE +00 O.OE +00 4.73 -08 l.lE -07 1.3E - 06 3.OE - 06 4.63 -08 l.lE -07 1.2E - 09 3.OE - 09 8.1E-07 1.9E-06 1.OE -08 2.73 - 08 3.33 - 07 7.43 -07 1.8E - 06 4.1E-06 3.63 - 08 8.2E - 08 2.OE - 09 4.93 - 09 3.OE - 10 7.4E- 10

Exkmal

3.83- 17 2.93 - 16 3.83 - 12 2.8E - 11 9.7E - 11 6.9E - 10 1.3E - 11 9.4E- 11 2.23 - 12 1.8E - 11 7.8E- 11 6.1E- 10 4.2E - 1 3 3.2E - 12 4.33 - 12 3.OE - 11 2.1E - 10 1.6E-09 8-43- 13 6.43 - 12 7.83 - 12 5.8E - 11 2.63 - 12 1.9E - 11

Inhalation

Veg.

1.5E -08 1.3E - 07 1.4E - 10 1.2E - 09 1.4E - 09 1.3E - 08 2.6E - 10 2.43 -09 2.8E - 11 2.4E - 10 7.1E - 10 6.43 -09 3.7E - 11 3.4E - 10 3.6E - 10 3.OE - 09 1.2E - 09 l.lE-08 7.2E - 11 6.OE - 10 5.4E - 10 4.83 - 09 8.4E - 11 7.4E- 10

Sparsely vegetated rural (RS) (continued)

2.53 - 09 3.23 - 08 2.4E - 11 3.OE - 10 2.4E - 10 3.33 - 09 4.2E - 11 5.9E - 10 4.53 - 12 6.OE - 11 1.2E - 10 1.6E - 09 6.33 - 12 8.8E - 11 5.8E - 11 8.2E - 10 1.9E - 10 2.83 - 09 1.3E - 11 1.5E - 10 9.2E - 11 1.2E - 09 1.4E-11 1.8E - 10

Soil

2.6E - 09 5.OE - 08 1.7E - 11 2.7E - 10 1.9E - 10 3.6E - 09 3.1E - 11 5.5E - 10 3.2E - 12 6.OE - 11 1.OE - 10 1.9E - 09 4.7E - 12 8.6E- 11 4.3E - 11 8.8E - 10 1.8E - 10 3.8E - 09 8.8E - 12 1.5E - 10 6.OE - 11 1.2E - 09 9.3E - 12 1.6E- 10

Meat

1.6E - 10 2.73 - 09 1.4E - 12 2.5E - 11 1.5E- 11 2.3E - 10 2.63 - 12 4.2E - 11 2.93 - 13 4.43 - 12 7.43 - 12 1.2E - 10 3.83 - 13 6.53 - 12 3.73 - 12 5.4E - 11 1.2E - 11 2.OE - 10 7.43 - 13 1.OE - 11 5.33- 12 8.8E - 11 8.1E- 13 1.2E - 11

Ingestion

Milk

2.73 - 08 2.1E - 07 2.3E - 10 1.7E - 09 2.4E - 09 1.9E - 08 4.2E - 10 3.4E - 09 4.5E - 11 3.5E - 10 1.2E - 09 9.4E - 09 6.4E - 11 5.OE - 10 5.9E - 10 4.5E - 09 2.OE - 09 1.6E - 08 1.2E - 10 8.7E - 10 9.1E - 10 7.1E -09 1.4E - 10 1.1E-09

Total Ingestion

2.7E - 08 C 2.1E - 07 4.83-08 C 1.1E -07 1.3E - 06 C 3.OE - 06 4.7E-08 C l.lE -07 1.3E-09 C 3.2E - 09 8.1E-07 C 1.9E - 06 1.OE - 08 C 2.73 - 08 3.3E - 07 C 7.4E - 07 1.8E - 06 C 4.1E - 06 3.63 - 08 C 8.23 - 08 3.23 - 09 C l.lE-08 4.9E - 10 C 1.7E - 09

Total

Delay, dc

1

Re-184m

Ta-182d

Hf-182

W-181

Hf-181

Ta-180

Ta-179

Hf-178m

Lu-177m

Lu-176

Hf-175

Lu-174d

Lu-174m

Lu-173

Hf-172

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Sparsely vegetated rural (FBI (continued) Veg.

Meat

Ingestion Milk

Soil

Total Ingestion Total

Delay, dc

\

9.1E - 08 2.1E-07 3-53- 08 8.33 - 08 4.73 - 07 l.lE - 06 4.3E - 08 9.9E - 08 l.lE - 09 2.53 - 09 4.43 - 12 1.2E - 11 1.6E - 06 3.83 - 06 2.63 - 07 6.23 - 07 1.7E - 09 4.1E - 09 4.1E - 12 1.OE - 11 1.9E - 06 4.43 - 06 3.93 - 07 9.1E - 07 1.2E - 06 2.83 - 06 median: 1.5E - 06 95th %: 3.53 - 06

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

/

242

APPENDMA

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I

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I

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zz z z 2 2 2 2 22 2:

2 2 2 2 2: 2 2 "?s 2 2

b(D b(D t - W (Dm b(D t-t- Cob M r l Cob Klb c o t - C C l b 0 0 0 0 0 0 0 0 0 0 0 0 0 0 rlrl 0 0 0 0 0 0 0 0

I

I

I

I

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I 1

I

I

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I

I

I

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WW W W WW WW W W W W WW WW W W WW WW WW

9 1 =?=!1 l n Y = ?9 % 1 % PC?lnc? cq- 1".- c q '4" m a m v Q,(D m m a * r l c o N N rlrl m m c o * m r l N N (Dm

2

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(D(D (Dm (Dm t-CD t - C D cot00 00 00 00 00 00 00

I

I

I

1

I

I

I

I

I

1

I

I

I

I

o m cotd o 00 I

I

I

I

t-t- c o t - cot00 00 00

I

I

I

I

I

I

W W W W WW WW W W WW WW WW W W W W WW WW

5 ; 2 %Z? 2 2 Z? =?z 2 2 2 2 2 %2 2 2: 2 2 (Dm (Dm

c o t - ma, m a c o b

coo3

mco mco

.

t-t-

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W W WW WW WW WW WW WW WW WW W W WW W W t-(D d m m a m t - o-e m m ot- m m o m cow r l w 41 A 6 & A A 6 A G , - i d d , - i054 G 4 4 d A 6 4 d G *

2

-

a

m

E

0

t-t- t-t-

00 00 0 0 0 0 00 00 00

I 1

I

I

I

I

I

I

I

I

I

I

I

I

d.4

I 1

00 00 00 00

I

I

I

I

I

I

I 1

(D* coco 0 0 o m t-t- t-t- c o t - - 0 o o o o o o o o o o o o rlr( d o o o o o o o d r l I I I I I I I I I I 1 1 I I I I I I I I I I I I

t-t- t-(0 t-t- ( D C D

W W WW WW WW W W WW WW WW W W WW WW WW " Y cqY Y Q ! r:" u?". qrq ".". "cq cq* cq" I T cqq m b (Dd * m NCD d m m(D d m b d d * N(D (Dd t - N

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th%: median: 95th %: median: 95th %: median: 95th %: median: 95th%: median: 95th %: median: 95th %: median: 95th %:

1.7E - 07 4.OE - 07 7.83 - 09 1.8E - 08 5.8E - 11 1.5E - 10 2.93 - 08 6.83-08 6.5E - 11 1.5E - 10 5.8E-11 1.3E - 10 1.2E - 08 2.83-08 6.33 - 08 1.5E-07 3.3E - 10 8.1E - 10 8.53 - 09 1.9E - 08 4.6E - 11 l.lE-10 4.7E - 11 1.2E - 10 2.1E - 07 4-83- 07 1.2E - 07 2.93 - 07

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Sparsely vegetated rural (RS)(continued) Veg.

Milk

Meat

Ingestion Soil

Total Ingestion

Delay, dc

1

Md-258

Fm-257

Es-254

Cf-254

Cf-252

Cf-251

median: 95th %: median: 95th %: median: 95th %. median: 95th %: median: 95th %: median: 95th 5:

Nuclide

External

median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th I: median: 96th %. median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Suburban (SU) Inhalation Veg.

Milk Meat

Soil

2.8E - 12 2.6E - 11 7.4E - 10 7.3E - 09 9.2E - 09 9.9E - 08 2.OE - 09 1.9E - 08 2.7E - 07 3.5E - 06 1.3E - 08 1.7E - 07 1.7E - 06 2.1E - 05 O.OE + 00 O.OE + 00 6.OE - 09 7.1E - 08 1.2E - 09 l.lE-08 1.7E-08 2.1E-07 1.6E - 10 1.4E - 09

Total Ingestion Total

Delay, dc

\

median: O.OE + 00 95th %: O.OE + 00 median: O.OE + 00 95th 8: O.OE + 00 median: 5.63 - 07 95th %: 1.3E - 06 median: O.OE + 00 95th %: O.OE + 00 median: l . l E - 06 95th %: 2.53 - 06 median: 7.OE - 08 95th %: 1.6E - 07 median: 2.53 - 07 95th %: 5.93 - 07 median: 2.1E - 07 95th %: 4.93 - 07 median: O.OE + 00 95th %: O.OE + 00 median: 2-43- 06 95th %: 5.63 - 06 median: 2.53 - 06 95th %: 5.83 - 06 median: O.OE + 00 95th %: O.OE + 00 median: 3.63 - 07 95th %: 8.43 - 07 median: 6.OE - 07 95th %: 1.4E - 06

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Suburban (SU) (continued) Inhalation Veg. Meat

Ingestion Milk

Soil

Total Ingestion

Delay, d c

1

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

2.9E - 10 7.3E - 10 5.9E - 11 1.5E - 10 6.1E - 11 1.5E - 10 1.5E - 06 3.53-06 1.OE -07 2.43 - 07 1.6E - 07 3.63 - 07 3.53 - 07 8.OE - 07 2.1E - 10 5.5E - 10 4.6E - 10 1.2E - 09 1.4E - 06 3.23 - 06 3.5E - 11 8.6E - 11 2.1E - 07 4.83 - 07 2.73 - 07 6.23 - 07 1.8E - 06 4.OE - 06

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

S u b u r b a n (SU) (continued) Inhalation

Veg.

Meat

Ingestion Milk

Soil

Total Ingestion Total

Delay, dc

\

median: 95th Ok: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8:

External

Suburban (SU) (continued) Inhalation

Veg.

Milk Meat

Ingestion Soil

Total Ingestion

B

1

ti

\

N

0

0

0

0

0

0

0

0

0

O D

0

0

Delay, dc

E3

en

Eu-152

Sm-151

Gd-151

Eu-150B

Eu-149

Pm-148m

Gd-148

Eu-148

Sm-147

Pm-147

Sm-146

Pm-146

Gd-146

Sm-145

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

95th 8:

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median:

External

Suburban (SU) (continued) Inhalation Veg.

Milk

Meat

Ingestion Soil

Total Ingestion

Delay, dc

\

Ta-182d

Hf-182

W-181

Hf-181

Ta-180

Ta-179

Hf-178m

Lu-177m

Lu-176

Hf-175

Lu-174*

Lu-174m

Lu-173

Hf-172 median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

median: 95th %: Re-184* median: 95th 8: median: 0s-185 95th %: W-185 median: 95th %: Re-186m median: 95th %: Re-187 median: 95th %: W-188 median: 95th %: Ir-192m median: 95th %: 1 ~ 1 9 2 ~ median: 95th %. Pt-193 median: 95th %: Hg-194 median: 95th %: median: Ir-194m 95th %:

Re-184m

Nuclide

External

Suburban (SU)(continued) Inhalation

Veg.

Ingestion Meat

Milk Soil

2.OE - 08 2.4E - 07 2.3E - 09 3.OE - 08 4.1E - 10 4.5E - 09 1.OE - 08 l.lE - 07 4.3E - 08 5.3E - 07 1.OE - 10 1.2E - 09 6.4E - 08 6.OE - 07 5.1E - 09 5.OE - 08 l.lE-09 1.3E - 08 3.2E - 10 4.1E - 09 5.9E - 07 7.6E - 06 2.5E - 09 2.9E - 08

Total Ingestion

I 8 0

0

0

0

0

365

0

0

0

0

O *

%

\

0

Delay, dc

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

U-236

Pu-236d

Np-236A

U-235

Np-235

U-234

U-233

U-2326

Th-232

Pa-231

Th-230

2.9E - 07 6.7E - 07 6.2E-07 1.4E-06 4.OE - 07 8.9E - 07 2.4E - 06 5.73-06 1.4E-06 3.1E-06 2.7E -08 5.9E - 08 1.2E - 10 3.1E-10 6.5E - 10 1.6E-09 1.6E-07 3.6E -07 2.5E-07 5.73-07 5.6E - 08 1.3E - 07 7.1E-11 1.8E- 10

Th-229 median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 96th %: median: 95th %: median: 95th %:

External

Nuclide

Suburban (SU) (continued) Veg.

8.OE - 07 5.73 - 06 6.63 - 07 2.93 - 06 1.3E - 06 7.1E-06 4.631 - 06 3.23 - 05 4.53 -07 2-73- 06 7.93 - 08 5.1E - 07 6.83 - 09 4.1E - 08 8.3E- 11 6.6E - 10 3.93 - 08 1.9E - 07 6.73 - 08 3.33 - 07 1.7E - 08 1.OE - 07 6.1E-09 4-33- 08

Inhalation

1.9E - 08 1.OE - 07 5.1E - 06 1.3E - 05 1.6E - 07 l.lE-06 1.6E - 08 8.53 - 08 l.lE-08 5.93 - 08 2.83 - 09 1.3E - 08 8.7E - 10 4.73 - 09 8.1E- 14 4.93 - 13 4.33 - 09 2.5E - 08 2.OE - 09 9.631- 09 4.5E - 10 2.43 - 09 7.8E- 10 4.43 - 09

Meat

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 0.03 + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Ingestion Milk

3.3E - 08 4.73 - 07 5.8E - 08 7.6E - 07 8.9E - 08 1.2E - 06 1.2E - 07 1.6E - 06 2.5E - 08 3.6E - 07 5.5E - 09 7.OE - 08 1.8E-09 2.8E - 08 2.1E - 12 3.1E- 11 3.8E - 09 4.9E - 08 3.4E - 09 4.7E - 08 9.6E - 10 1.3E- 08 1.8E - 09 2.5E - 08

Soil

8.9E - 07 6.1E - 06 8.0E - 07 3.4E - 06 1.5E - 06 8.1E - 06 5.OE - 06 3.3E - 05 5.3E - 07 3.OE - 06 9.4E - 08 5.5E - 07 l.lE - 08 6.8E - 08 9.OE- 11 6.8E - 10 4.8E - 08 2.2E - 07 7.7E - 08 3.6E - 07 2.OE - 08 1.1E - 07 9.7E - 09 6.2E - 08

Total Ingestion

0

7300

182500

365000

0

365000

365000

3650

18700

73000

365000

90

Delay, dc

!2

a 5

%

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Suburban (SU) (continued) Inhalation

8.6E - 09 6.8E - 08 1.8E - 08 1.OE - 07 2.8E - 08 1.4E- 07 2.OE - 08 1.5E - 07 2.8E - 08 2.4E - 07 1.5E - 08 1.OE - 07 3.6E - 09 2.5E - 08 5.1E - 08 3.9E - 07 5.7E - 11 4.9E - 10 2.8E - 08 2.3E - 07 1.7E - 08 1.4E - 07 2.8E - 07 2.4E - 06

Veg.

Milk Meat

Ingestion Soil

Total Ingestion

0

0

0

3650

0

0

365000

0

0

365000

3 0

k

%'u m

\

0

Delay, dc

Md-258

Fm-257

Es-254

Cf-254

Cf-252

Cf-251

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

No food suburban (SN) Inhalation

Milk

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Veg.

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Meat

Ingestion Soil

2.7E - 13 4.7E - 12 5.8E - 11 l.lE-09 1.2E - 10 1.9E - 09 1.7E - 10 3.2E - 09 1.8E - 10 2.83 - 09 1.3E - 11 2.2E - 10 4.5E - 11 8.4E - 10 O.OE + 00 O.OE + 00 1.1E- 11 2.OE - 10 2.6E - 10 5.OE - 09 2.OE - 11 4.2E - 10 2.2E - 11 4.OE - 10

Total Ingestion

9.7E - 09 C 2.1E - 08 3.9E - 10 C 1.4E - 09 1.9E-06 C 4.33 - 06 2.8E - 06 C 6.1E - 06 2.8E - 09 C 7.3E - 09 1.5E - 11 C 2.2E - 10 6.3E - 10 C 1.8E - 09 1.6E - 07 C 3.8E - 07 1.1E - 11 C 2.OE - 10 2.2E - 06 C 4.9E - 06 3.3E-11 C 4.4E - 10 6.3E - 07 C 1.5E - 06

Total

8E 0

o b

%

\

0

Delay, dc

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 96: median: 95th 8: median: 95th %. median: 95th %. median: 95th %: median: 95th %:

External

Inhalation

No food suburban (SN)(continued) Veg.

Meat

Ineestion Milk Soil

Total Ingestion

Delay, dc

\

median: 95th %; median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th 8: median: 95th %. median: 95th %: median: 95th%: median: 95th %. median: 95th %: median: 95th %: median: 95th %: 3.9E - 10 9.9E- 10 5.9E - 11 1.5E - 10 6.1E - 11 1.5E - 10 1.5E - 06 3.53 - 06 1.OE - 07 2.43 -07 1.6E- 07 3.63 - 07 3-53- 07 8.OE - 07 2.1E- 10 5.5E- 10 4.6E - 10 1.2E- 09 1.4E- 06 3.23-06 3.5E - 11 8.6E - 11 2.1E - 07 4-83-07 2.73 - 07 6.23 - 07 1.8E - 06 4.OE - 06

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

N o food suburban (SN) (continued) Veg.

Meat

Ingestion Milk

Soil

Total Ingestion

Delay, dc

1

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Pm-145d

Pm-144

Ce-144

Pm-143

Ce-141

Ce-139

La-138

La-137

Cs-137

Cs-135

Cs-134

8.33 - 13 5.9E - 12 7.73 - 13 4.OE - 12 l.lE - 13 5.1E - 13 1.2E - 12 5.83 - 12 1.9E - 12 1.4E - 11 2.9E - 11 2.2E - 10 2.OE - 13 1.2E - 12 9.93 - 14 5.63 - 13 2.63 - 13 1.9E - 12 6-33- 12 3.6E - 11 1.5E- 12 1.2E - 11 7.33 - 13 5.1E - 12

3.43 - 07 7.93 - 07 1.3E - 06 2.93 - 06 1.2E - 11 2.7E - 11 5.23 - 07 1.2E - 06 6.83 - 09 1.7E - 08 1.2E - 06 2.8E - 06 6.1E - 08 1.4E - 07 8.63 - 09 2.OE - 08 1.8E - 07 4.23 - 07 3-73- 08 8.73- 08 1.OE - 06 2.43 - 06 1.1E - 08 2.83 - 08

Ba-133

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Inhalation

External

Nuclide

No food suburban (SN)(continued)

+

+

O.OE + 00 O.OE + 00 0.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 0,OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Meat

Ingestion Milk

Veg.

l.lE - 10 1.9E - 09 2.7E - 10 3.33 - 09 4.3E - 11 5.9E - 10 2.5E - 10 3.OE - 09 3-93- 12 7.2E - 11 4.2E - 11 7.5E - 10 6.OE - 12 8.7E - 11 4.73 - 12 7.2E - 11 6.43 - 12 1.2E - 10 1.8E - 10 2.83 - 09 3.3E - 11 5.6E - 10 5.53 - 12 1.OE - 10

Soil

1.1E - 10 1.9E - 09 2.7E - 10 3.3E - 09 4.3E - 11 5.9E - 10 2.5E - 10 3.OE - 09 3.9E - 12 7.2E - 11 4.2E - 11 7.5E - 10 6.OE - 12 8.7E - 11 4.7E - 12 7.2E - 11 6.4E - 12 1.2E - 10 1.8E - 10 2.8E - 09 3.3E - 11 5.6E - 10 5.5E - 12 1.OE - 10

Total Ingestion

3.4E - 07 C 7.9E - 07 1.3E- 06 C 2.9E - 06 5.6E - 11C 6.1E - 10 5.2E - 07 C 1.2E - 06 6.8E-09 C 1.7E - 08 1.2E - 06 C 2.8E - 06 6.1E-08 C 1.4E - 07 8.6E - 09 C 2.OE - 08 1.8E - 07 C 4.2E - 07 3.8E - 08 C 8.9E - 08 1.OE - 06 C 2.4E - 06 l.lE-08 C 2.9E - 08

Totalb

*

O

0

0

0

0

0

0

0

0

0

z

E

% m"

1

O

0

Delay, dc

h3

median: 95th %: median: 95th %: median: 95th %: medan: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %:

External

Inhalation

No food suburban (SN) (continued) Veg.

Meat

Ineestion Milk

Soil

Total Ingestion Total

Delay, dc

1

median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th Z: median: 95th %: median: 95th 8: median: 95th %: median: 95th %:

4.7E - 07 l.lE - 06 1.2E - 07 2.9E - 07 2.3E - 07 5.2E - 07 3.3E - 11 7.7E - 11 2.6E - 08 6.2E - 08 O.OE + 00 0.OE + 00 1.5E - 08 3.4E - 08 9.OE - 07 2.OE - 06 2.2E - 07 5.OE - 07 3.2E - 12 8.2E - 12 1.OE - 06 2.3E - 06 1.1E - 06 2.7E - 06

Re-184m

median: 95th %: Re-184* median: 95th %: 0s-185 median: 95th %: W-185 median: 95th %: Re-186m median: 95th %: median: Re-187 95th %: W-188 median: 95th %: median: Ir-192m 95th %: 1 ~ 1 9 2 ~ median: 95th %: median: Pt-193 95th %: median: Hg-194 95th %: Ir-194m median: 95th %:

External

Inhalation

(SN)(continued)

Nuclide

No food suburban Veg.

Soil

5.2E - 11 9.6E - 10 6.3E - 12 1.2E - 10 8.3E - 12 1.7E - 10 7.OE - 12 1.1E - 10 1.1E- 10 2.OE - 09 2.4E - 13 4.5E - 12 5.7E - 11 9.8E - 10 8.8E - 11 1.5E - 09 1.9E - 11 3.3E - 10 1.7E - 12 2.8E - 11 1.6E - 09 2.9E - 08 5.3E - 11 9.5E - 10

Meat

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Milk O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 0,OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

Ingestion Total Ingestion

R

%'d k? u

\

h3

0

0

0

0

365

0

0

0

0

o *

0

0

Delay, d'

h3

-a

median: 95th %: median: 95th 46: median: 95th 96: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %; median: 95th 96: median: 95th' %:

8.3E - 08 1.9E - 07 3-23- 08 7.53 - 08 4.33 - 07 9.83 - 07 3.93 - 08 9.OE - 08 9.7E - 10 2.23 - 09 4.1E - 12 1.OE - 11 1.5E -06 3.5E - 06 2.43 - 07 5.53-07 1.6E - 09 3.73 - 09 3.83 - 12 8.93- 12 1.7E - 06 4.OE - 06 3.63 - 07 8.23 - 07 1.5E - 06 3.43 - 06 1.4E - 06 3.23 - 06

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8:

External

Inhalation

No food suburban (SN) (continued) Veg.

Milk

Meat

Ingestion Soil

Total Ingestion Total

0

7300

182500

365000

0

365000

365000

3650

18700

9

3 365000 73000

% 71 m

\

90

Delay, dc

median: 95th %: median: 95th %: median: 95th %: median: 95th %. median: 95th 8: median: 96th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 96: median: 95th %:

2.OE - 07 4.73 - 07 7.OE - 09 1.7E - 08 6.8E - 11 1.7E - 10 2.73 - 08 6.23 - 08 7.6E - 11 1.8E - 10 6.9E - 11 1.6E - 10 1.43 - 08 3.33 - 08 5.7E - 08 1.4E - 07 3.9E - 10 9.6E - 10 1.2E- 08 2.9E - 08 4.1E- 11 l.lE- 10 5.6E - 11 1.5E - 10 1.9E - 07 4-43- 07 l.lE -07 2.63 - 07

APPENDIX A

U

U

coca

P-P- P-P-

U

U U U U U U cot- b P - P-E- m m mP- 0

0 0 00 0 0 00 0 0 0 0 00 0 0

I I

I I

I I

I I

I I

I

I

I I

I

I

U

-8 I

I

V

U

P-P- C O + b W 0 0 0 0 00

I 1

I l

I I

WW WW WW WW WW W W WW WW WW WW W W WW

23 2 2 2: 2: 2 5 2: $2

2 2 ~2 2; 2:

mP- c o b at- ma3 a3P- d o COP- Q)Q) b(O 8 0 COP0 0 0 0 0 0 0 0 0 0 0 0 drl 0 0 0 0 0 0 I I I I I I I I I I I I I I I I I I I I I I I I P-

g%

W !3 , W2 2W,gg52W,w,g% Re w g 553g;w c6w G A A 4 uj4 ,-iA u j A 4 A 4~ ci4 A e i u j & . - i e i

ma3 m b COP- me- COP-m ~ -ma3 c o t - d o a,r- ma3 P - w 00 00 00 00 00 00 00 00 4rl 0 0 00 0 0 I 1 I 1 I I I 1 I I I I I 1 I I I I I I I I I 1

WW WW WW WW WW W W WW WW WW WW WW WW c?9Y-! -'? 'f?9 --! = ? 9 9 %F?'? Yc? t r ! qC9 9F? mCD P - 4 rlrl lnrl

FIN

mrl

r(*

NCI)

eav

FIN

U3m r l w

8 8 8 8 8 8 88 8 8 8 8 800 8 8 8 8 80" 80" 80"

++ ++ ++ ++ ++ ++ ++ ++ ++ ++ ++ ++

w0w0 wow0 33 gi2 EWo Wow0 !2E3Wo Wow0 Wo!2 Wow0 WOE d o do do d od o od do do do do do do 0 0 0 0 00 0 0 0 0 00 0 0 00 0 0 0 0

0 00 00 80 00 00

00 0 0 00 0 0 00 0 0 00 0 0

++ ++ ++ ++ ++ ++ ++ ++ ++ ++ ++ ++

WW WW WW WW WW WW WW WW WW WW WW WW 99 99 99 99 9 9 99 99 99 99 99 99 99

0 0 00 00 0 0 0 0 0 0 00 0 0 00 0 0 0 0 0 0 000 0000 8 8 800 8 8 0 0 0 8 00 0 0 ++ ++ ++ ++++ ++ ++ ++ ++ ++ ++ ++ 3Wo 8i2Wow0 Wag w0Wo gf2gw0 38 Wo3 Wow0 EWo Wow0

8 8 8 8 8 8 8-8"o

--

do do do do dodo do do do d ood do ma3 mCO m a , mco corn mco m a , C O b d o b m a bb 00 0 0 00 00 0 0 0 0 0 0 0 0 80 00 00 I I

I

1

I I

I I

I

I

I

I

I I

I

I

I I

I

I

I

I

I I

WW WW WW WW WW WW WW WW WW WW WW WW

$2 2 2 22 22 22 2 2 2 2 22 2~ 2~ 2~ 2:

Md-258

Fro-257

Es-254

Cf-254

Cf-252

Cf-251 median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Veg.

Construction, commercial, industrial (CC) Milk

Meat

Ingestion Soil

Total Ingestion Total

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 2.33 - 07 5.93 - 07 O.OE + 00 O.OE + 00 4.53 - 07 1.2E - 06 2.93 - 08 7.63 - 08 1.l E - 07 2.73 - 07 8.73 - 08 2.33 - 07 O.OE + 00 O.OE + 00 1.OE - 06 2.53 - 06 1.l E - 06 2.73 - 06 O.OE + 00 O.OE + 00 1.5E - 07 3.83 - 07 2.53 - 07 6.33 - 07

Nuclide

median: 95th 96: median: 95th 8: median: 95th 96: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %:

External

Inhalation

Veg.

Meat

Ineestion

Milk

Construction, commercial, industrial (CC) (continued) Soil

Total Ingestion

0

0

0

0

0

0

0

0

0

0

0

0

Delay, do

*

R

z

% Cb

\

median: 2.2E - 10 95th %: 5.8E - 10 median: 3.3E - 11 95th %: 9.1E - 11 median: 3.3E - 11 95th 8: 9.2E - 11 median: 6.33 - 07 95th %: 1.6E - 06 median: 4.33 - 08 95th %: l . l E - 07 median: 6.43 - 08 95th %: 1.6E - 07 median: 1.5E - 07 95th 8: 3.73 - 07 median: l . l E - 10 95th %: 3.2E - 10 median: 2.6E - 10 95th %: 7.1E - 10 median: 5.73 - 07 95th %: 1.5E - 06 median: 1.8E - 11 95th 8: 4.6E - 11 median: 8.83 - 08 95th %: 2.23-07 median: l.lE - 07 95th %: 2.83 - 07 median: 7.33 - 07 95th %: 1.9E - 06

-

2.8E - 08 6.9E - 08 3.2E - 08 8.1E - 08 6.2E - 08 1.6E - 07 O.OE + 00 O.OE + 00 4.8E - 07 1.3E - 06 1.9E - 09 5.23 - 09 6.9E - 07 1.8E - 06 7.1E 11 1.8E - 10 1.6E - 11 4.1E- 11 3.9E - 08 1.OE -07 6.7E - 09 1.7E - 08 1.7E - 09 4.5E - 09

Ru-103 median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Nuclide

Inhalation

0.OE + 00 O.OE + 00 0.OE +00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 0.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 0.OE + 00 O.OE + 00 O.OE + 00

Veg.

Meat

Ineestion

Milk

Construction, commercial, industrial (CC) (continued) Soil

Total Ingestion

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

External

Inhalation

Veg.

Meat

Ingestion Milk

Construction, commercial, industrial (CC) (continued) Soil

Total Ingestion Total

Delay, dc

\

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 8: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

Nuclide

95th %:

median:

95th 8:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

External

Inhalation

Veg.

Construction, commercial, industrial (CC) (continued) Meat

Ingestion Milk

Soil

Total Ingestion Total

Delay, dc

\

CALCULATED SCREENING DOSES

4

4

4

4

4

4

4

4

4

4

4

4

/

287

4

4

r - w m m m m moo m m r-r- r-r- r - w cnm r - c - m m m m r - w r-r0 0 00 0 0 00 0 0 00 0 0 00 0 0 0 0 0 0 0 0 0 0 0 0

I

I

I I

I 1

I

I

**

I I

I

I

I 1

I

I

I

I I

I

I

I

I

I

I

I

I

I

WW W W W W W W WW W W WW W W WW WW WW WW WW WW *C9 * V ! 1* C:C: a!* a ! 9 9 9 C 9 1 11 U?* 9 * 1 C 9 1 N . w + m m d m mcn mcn r l w r l m m m r - m m m m m m r l w r l m m

r l o e a r l e a r l e a r l earlrlo

40

do

r l o m . 4 mea r l o eao

03.4

rlrl rlrl r l d d d rlrl rlrl d r l r l d d r l rlrl 4 4 d d

I

I

WW

I

I

ww

I

I

WW

Ic!* 1 ' 9 cq?

mm m* *m d o earl

~

rlrl rlrl r l d

I

I

I

I

1

1

I I

I

I

I

I

, 4 7 4

I

I

m rll m r l r l o r l r c rlrl rlrl

r

I

I

I

I

rlrl r l r l 4 r l

I

I

I

I

I

I

I

I

I

ww W W ww ww ww ww c:? * * a!=? qo, ?a! 9 m * rcea m m r n r l r-cn corn

r l o mrc e a r l mrc m~ 4 4

I

I

I

I I

I

I

I

I

I I

I

I

rld

I

I

I 1

w*'? w wcqcqw wO Nw. WU?9W w* Nw

r l r n m m m w mcn cnrl

m d

mrl mm

4 0 rlo do rlo N O rlrl rlrl r l 4 d d rlrl rlrlrlrl rlrl

I

I

I

I

I

I

I I

I

I

I

I

I

I

I

I

o c n d o o m mea r l o m r l w m o m earl 0 4 d o rlrl r l d rlrl d r l 4 0 r l d 7

4

I

0

I

I I

I

I

I

I

I

I

I

I

I

I

I

I

I

I

WW W W W W W W WW W W W W WW WW WW W W WW W W W W C : l C:=? C:* 1 ' 9 11 c ? 1 V!* =?9 ?a! 9a! C9* N.? c?9 a!* m w m m w m r-m w m

rlrl rlrl

*w mw mm

rlrl

m r l w m r-m

r - w m m m m m m m m r-r- r-r- r - w cnm r-r- corn cnm r - w r-r00 00 00 00 00 00 00 00 00 00 00 00 00 00 I I I 1 I I I I I I I I I I I I I I I I I I I I I I I I

WW W W WW W W WW W W WW WW W W WW WW WW WW WW *C9 =?U? 1 * * * C:C: a!* e 9 o,c? C 9 1 1 d : U?* 9 * Y V ! 1 N . w d m m d m mcn m m r l w r l m m m r - e a N N m m m r l w d ~m

5

a-

w

0

aE

; z q " 4; 5

re a -r n r

S

4

4

4

4

S

g

E

E

2

mo

r

7 l " d "m

p wB g $

k

2

&

Nuclide

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %. median: 95th %: median: 95th %:

External

Inhalation

Veg.

Meat

Ineestion Milk

Construction, commercial, industrial (CC) (continued) Soil

Total Ingestion Total

Delay, dc

N

1

OD OD

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %:

3.43 - 08 8.73 - 08 1.4E - 08 3.8E - 08 1.8E- 07 4.53 - 07 1.6E - 08 4.1E - 08 4.4E - 10 1.1E-09 2.2E - 12 6.2E - 12 6.1E - 07 1.6E - 06 9.73 - 08 2.6E-07 7.7E - 10 2.OE - 09 1.6E - 12 4.OE - 12 7.33 - 07 1.9E - 06 1.5E - 07 3.83 - 07 6.2E - 07 1.6E-06 5.63 - 07 1.5E - 06

Nuclide

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

95th %:

median:

External

Inhalation

Veg.

Meat

Ingestion Milk

Construction, commercial, industrial (CC)(continued) Soil

Total Ingestion

Delay, dc

1

median: 8.53 - 08 95th %: 2.23 - 07 median: 2.9E - 09 95th %: 7.63 - 09 median: 3.8E - 11 95th %: 1.OE-10 median: l . l E - 08 95th 8: 2.83 - 08 median: 3.4E - 11 95th %: 8.9E - 11 median: 3.8E - 11 95th %: 9.5E - 11 median: 6.73 - 09 95th %: 1.8E - 08 median: 2.43 - 08 95th %: 6.33 - 08 median: 1.9E - 10 95th %: 5.2E - 10 median: 5.1E - 09 95th %: 1.3E - 08 median: 2.3E - 11 95th %: 6.2E - 11 median: 3.1E - 11 95th %: 8.8E - 11 median: 7.93 - 08 95th %: 2.OE-07 median: 4.63 - 08 95th %: 1.2E - 07

Nuclide

OE+L W-.

median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th %: median: 95th 5: median: 95th %: median:

External

Inhalation Veg.

Meat

Ingestion Milk

Construction, commercial, industrial (CC) (continued) Soil

Total Ingestion Totalb

Delay, dc

\

median: 95th %: median: 95th %: median: 96th %: median: 95th %: median: 95th %:

3.6E - 11 1.OE - 10 2.83 - 14 8.OE- 14 2-33-07 6.1E - 07 1.4E - 08 3.63 - 08 9.3E - 11 2.5E - 10

3.53 - 08 2.83 - 07 1.9E - 08 1.5E-07 1.3E - 08 9.1E - 08 9.43 - 09 5.93 - 08 2.53 - 09 1.9E - 08

O.OE + 00 O.OE + 00 O.OE + 00 O.OE+OO O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 O.OE + 00 O.OE +00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00

O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 O.OE + 00 1.3E - 09 1.7E - 08 1.5E - 09 2.1E - 08 3.8E - 10 4.53 - 09 1.6E - 10 1.7E - 09 4.5E - 11 5.6E - 10 1.3E -09 1.7E - 08 1.5E - 09 2.1E-08 3.8E - 10 4-53- 09 1.6E - 10 1.7E - 09 4.5E - 11 5.6E - 10

4.OE - 08 A 2-93- 07 2.4E -08 A 1.6E -07 2.6E -07 A 6.83 - 07 2.6E - 08 A 8.83 -08 2.7E - 09 A 1.9E -08

0

0

0

0

0

.Median and 95th percentile annual committed effective dose per Bq kg-' dry soil for a member of the critical exposed p u p . bAnA following the total dose indicates the critical group consists of adults while a C indicates the critical exposed group consists of children andlor infants. The doses are the maximum annual over a 1,000 y interval. 'The delay indicate the number of days elapsed prior to the beginning of the year of maximum dose. *Daughter nuclide whose dose is also included in listed dose for parent.

Md-258

Fm-257

Es-254

Cf-254

Cf-252

0

wl

u

m

2 Gc3

APPENDIX B

Radionuclide Decay Dataa Nuclide

Tmb

Daughter

Fraction

Daughter

Fraction

RADIONUCLIDE DECAY DATA

1

295

Radionuclide decay data (continued) Nuclide

Tmb

Daughter

Fraction

Daughter

Fraction

Radionuclide decay data (continued) Nuclide

Tmb

Daughter

Fraction

Daughter

Fraction

RADIONUCLIDE DECAY DATA

1

297

Radionuclide decay data (continued) Nuclide

Pm-148m Pm-148 Eu-149 Eu-150B Gd-151 Sm-151 Eu-152 Gd-152 Gd-153 Eu-154 Eu-155 Tb-157 Tb-158 Dy-159 Tb-160 Ho-1661-11 Yb-169 Tm-170 Tm-171 Hf-172 Lu-172 Lu-173 Lu-174m Lu-174 Hf-175 Lu-176 Lu-177m Lu-177 Hf-178m Ta-179 Ta-180 Hf-181 w-181 Hf-182 Ta-182 Re-184m Re-184 0s-185 W-185 Re-186m Re- 186 Re-187 Re-188 W-188

Tmb

Daughter

Fkaction

Daughter

F'raction

Radionuclide decay data (continued) Nuclide

Tlnb

Daughter

Fraction

Ir-194

1.0

Tl-202

1.0

Daughter

Fraction

RADIONUCLIDE DECAY DATA

1

299

Radionuclide decay data (continued) Nuclide

Tmb

Daughter

Fraction

Daughter

Fraction

300

/

APPENDIXB

Radionuclide decay data (continued) Nuclide

Am-242m Am-242 Cm-242 F'U-242 Am-243 Cm-243 Pu-243 Cm-244 F'U-244 Cm-246 Am-245 Cm-246 Am-246 Pu-246 Bk-247 Cm-247 Cf-248 Cm-248 Bk-249 Cf-249 Cm-249 Bk-250 Cf-250 Cm-250 Cf-251 Cf-252 Cf-253 Es-253 Cf-254 3s-254 Fm-257 Md-258

Tmb

1.52E + 02 y 1.60E+01 h 1.63E+02 d 3.763 +05 y 7.383 +03 y 2.853+01 y 4.963+00 h 1.81E+01 y 8.263 + 07 y 8.503 + 03 y 2.053+00 h 4.733+03 y 3.903 + 01 m 1.09E+01d 1.383+03 y 1.56E + 07 y 3.343 + 02 d 3.393 + 05 y 3.203 + 02 d 3.51E + 02 y 6.423 + 01 m 3.223 + 00 h 1.313+ 01 y 6.90E + 03 y 8.983 + 02 y 2.643 + 00 y 1.78E + 01 d 2.05E + 01 d 6.05E + 01 d 2.763 + 02 d 1.013 +02 d 5.503 + 01 d

Daughter

Fraction

Daughter

Fraction

Np-238 Cm-242 Pu-238 U-238 Np-239 Pu-239 Am-243 Pu-240 U-240 Pu-241 Cm-245 Pu-242 Cm-246 Am-246 Am-243 Pu-243 Cm-244 Pu-244 Cf-249 Cm-245 Bk-249 Cf-250 Cm-246 Pu-246 Cm-247 Cm-248 Cm-249 Bk-249 Cm-250 Bk-250 Cf-253

0.0048 0.827 1.0 1.0 1.0 0.998 1.0 1.0 1.0 1.0 1.0 0.9997 1.0 1.0 1.0 1.0 1.0 0.9174 1.0 1.0 1.0 1.0 0.9992 0.25 1.0 0.9691 0.0031 1.0 0.0031 1.0 0.9979

Am-242 Pu-242

0.9952 0.173

Am-243

0.002

Bk-250

0.14

Es-253

0.9969

'Data from ICRP Publication 38 (ICRP,1983). bs = second,m = minute, d = day, y = year.

W

Isotope

Tin (dl Df&

CV (%) SF

Dose Factors, Shielding Factorsa Dfa CIA

l.fA

Dfw

CIA

VA

Isotope

'I'm

(4

of-

Dose factors, shielding factors (continued)

CV (%) SF Dfm C/A

IfA

Dfi,

Isotope

TI,(d)

of-t

Dose factors, shielding factors (continued)

CV (%)

SF Dfinh

CIA

YA

Dfins

CIA

YA

2

0

Te-1291x1 Te-129 Ba-133 CS-134 CS-135 Ba-137m CS-137 La-137 La-138 Ce-139 Ce-141 Pm-143 Ce-144 Pm-144 Pr-144 Pr-144m hn-145 Sm-145 Eu-146 Gd-146 Pm-146 Sm-146 Pm-147 Sm-147 Eu-148 Gd-148 Pm-148m Pm-148 Eu-149 Eu-150B

Isotope

T , (dl Dfext

Dose factors, shielding factors (continued)

I/A 4.1 2.5 2.4 2.8 5.7 2.8 3.3 2.5 2.2 4.6 3.6 2.1 3.3 4.0 4.1 4.1 4.4 3.6 3.6 3.3 3.8 2.4 3.3 3.4 2.2

CIA 1.5 1.1 1.2 1.3 1.9 1.2 1.3 1.2 1.1 1.6 1.4 1.1 1.4 1.6 1.4 1.5 1.8 1.5 1.5 1.4 1.5 1.1 1.4 1.4 1.2

6.5 6.5

2.1 2.0 1.9 1.3 2.1 2.1 2.1 2.0 1.9 2.1 2.1 1.8 2.1 2.2 2.1 2.0 1.9 2.0 2.1 2.1 2.0 1.9 2.1 2.3 1.7

6.7 6.0 6.9 6.5 5.4 6.4 6.3 4.7 6.5 7.5 7.1 6.1 5.4 6.2 7.2 6.3 5.9 6.1 6.5 7.4 4.0

:

I/A

CIA

0

38

%

\

Isotope

Tin (dl Dfd

Dose factors, shielding factors (continued)

CV(%)

SF Dfinh

C/A

I/A

Dhng

Isotope Tllz

(dl W s x t

Dose factors, shielding fadors (continued)

CV (%) SF

CIA

llA Dfm

CIA

llA

\

' D f , [nSv d-' (Bq kg-')-'] is external exposure dose factor (see Section 3.2.2). Dfw (Sv Bq-l) is inhalation dose factor, and Df,, (Sv Bq-') is ingestion dose factor. CIA is ratio of dose factor for 10 y old to adult, I/A is ratio of dose factor for 1y old to adult. Inhalation and ingestion dose factors from ICRP Publication 72 (ICRP, 1996).

CL

CL w

--

APPENDIX D

Transfer Factorsa Fmilk

Element

BJwetIb B,(dry)'

GSD

(d L-'1

Fmt

GSD

(d kg-')

GSD

TRANSFER FACTORS

/

Transfer factors (continued)

Element Ha Hf Hg Ho I In Ir K La Li Lr Lu Md Mg Mn Mo Na Nb Nd Ni No NP 0s

P Pa Pb Pd

Pm Po

Pr t'F Pu Ra Rb Re Rf Rh Ru S Sb Sc

Se Si Sm

GSD

Fdk (d L-'1

FWt

GSD

(d kg-')

GSD

314

/

APPENDMD

Transfer factors (continued)

Element

BJwetIb

B,(dry)e

GSD

Fdk (d L-')

GSD

Fmt (d kg-?

GSD

Sn Sr Ta

0.3 0.3 2E-03 2E-03 5.0 0.1 1E-03 5E-04 0.2 2E-03 2E-03 2E-03 0.8 2E-03 2E-03 0.4 1E-03

1.0 4.0 0.02 0.05 40.0 1.3 1E-03 5E-03 0.6 0.05 0.1 0.01 3.0 0.05 0.05 1.0 5E-03

3.0 2.7 3.0 3.0 2.5 2.7 2.5 3.0 3.0 3.0 2.5 3.0 2.7 3.0 2.7 2.5 2.7

1E-03 2E-03 5E-06 6E-05 1E-03 5E-04 5E-06 0.01 3E-03 6E-05 4E-04 5E-04 3E-04 6E-05 6E-05 0.01 6E-07

2.5 1.6 2.5 2.5 2.0 1.8 2.5 2.5 2.5 2.5 1.8 2.5 2.5 2.5 2.5 2.5 2.0

0.01 0.01 5E-06 2E-03 1E-04 7E-03 1E-04 0.02 0.02 2E-03 8E-04 0.01 0.04 2E-03 2E-03 0.1 1E-06

2.8 1.5 2.8 2.8 2.0 2.5 2.8 2.8 2.8 2.8 2.0 2.8 2.5 2.5 2.8 1.4 2.5

Tb TC Te Th Ti Tl

lh U V W Y Yb Zn Zr

"Adapted from NCRP (1996). bBqkg-' wet vegetation per Bq kg-' dry soil for vegetables, fruit and grain. 'Bq kg-' dry vegetation per Bq kg-' dry soil for fodder and grass.

Glossary absorbed dose: The mean energy imparted by ionizing radiation to an irradiated medium per unit mass. Units: gray (Gy); formerly the unit was the rad (1 Gy = 100 rad). AZVIAD: Activity weighted median aerodynamic diameter. aquifer: A formation or group of formations, or part of a formation that contains sufficient saturated permeable material4.a yield significant quantities of water to wells and springs. becquerel (Bq): The special name for the unit of activity. 1 Bq = 1s-'. committed effective dose [E(z)]:The committed equivalent doses to individual tissues or organs resulting from an intake multiplied by the appropriate tissue weighting factor (wT) and then summed. E(d = ZwTHT(z)where HT(z)is the committed equivalent dose in tissue T, W Tis the weighting factor for tissue T and z is the integration period in years. confidence interval:A measure of the reliability of a risk estimate or other parameter. A 90 percent confidence interval means that 9 times out of 10 the unknown true value of the risk would be within the specified interval. conservative bias: A tendency to overestimate rather than underestimate. coefficient of variation (CV):The ratio of the standard deviation to the mean. default value: A value prescribed for a model parameter in the absence of data directly relevant to the assessment situation. dispersion coefficient: A measure of the spreading of a flowing substance due to the nature of the porous medium. distribution coefficient: The quantity of the radionuclide sorbed by the solid per unit weight of solid divided by the quantity of radionuclide dissolved in the water per unit volume of water. dose: See absorbed dose, equivalent dose; effective dose; committed effective dose. dose rate: Dose delivered per unit time. effective dose (E):The sum over specsed tissues of the products of the equivalent dose in a tissue (T)and the weighting factor for that tissue (wT),i.e., E = 2wTHT.

316

/

GLOSSARY

a):

equivalent dose A quantity used for radiation-protection purposes that takes into account the different probability of effects which occur with the same absorbed dose delivered by radiations with different wRvalues. It is defined as the product of the averaged and the radiation absorbed dose in a specified organ or tissue (DT) weighting factor (wd. The unit of equivalent dose is joules per kilogram (J kg-') and its special name is the sievert (Sv). extrapolation: The projection of model calculations to situations outside the realm of past experience or known data. Model calculations performed within the realm of experience and pertinent data are considered to be interpolations unless verified by measurements. geometrical standard deviation (GSD):The geometrical standard deviation reflects the deviations of the logarithms of the individual values (the xi)from the geometric mean (p). It is equal to the base of the natural logarithms (e) raised to the power u, where u is the standard deviation of the logarithms of the individual values. Mathematically, we have u = E(lnx, - p)' 1 (n - 1)Im and GSD = em. gray (Gy):The special name for the unit of absorbed dose, k e m a and specific energy imparted, 1 Gy = 1J kg-'. homogeneity:The properties, or conditions ofisotropy or anisotropy are constant from point to point in the groundwater medium. kerma (kinetic energy released per unit mass of material):A unit of energy released, expressed in gray, that represents the kinetic energy transferred to charged particles per unit mass of irradiated medium when indirectly ionizing (uncharged) particles, such as photons or neutrons, traverse the medium. If all of the kinetic energy is absorbed 'locally," the kerma is equal to the absorbed dose. log-normal distribution: A set of values whose logarithms are normally distributed is said to be lognormally distributed. The lognormal distribution is asymmetric. It is bounded at the lower end by zero because the logarithm is defined only for positive numbers. mean: The mean value of a set of measurements of a quantity is the sum of the measured values divided by the number of measurements. The mean value is also often called the (arithmetic) average value. The mean of a distribution is the weighted average of the possible values of the random variable. median: Of a set of n values, the median is the value that is as frequently exceeded (by other values in the set) as not. To determine the median value, arrange the set of values in order to generate the sequence: x, 5 x, I x, 5 . . . I x,. When n is odd, the median is the value in the center of the sequence; there will be

GLOSSARY

1

317

(n - 012 values that are smaller and (n - 1)12 values larger than the median. The median value of a distribution is the 50th percentile. mode: The value that is measured (or estimated) most frequently, i.e., the mode is the most probable value. model: A mathematical abstraction of an ecological or biological system, sometimes including specific numerical values for the parameters of the system. model prediction: The result or dependent variable produced by a model calculation. model structure: The conceptual design, mathematical equation, and set of algorithms that control the results or predictions produced from a given set of input data or assumptions. Monte Carlo calculation:The evaluation of a probability distribution by means of random sampling. normal distribution: The normal distribution is an unbounded symmetric distribution, characterized by its mean (m) and standard deviation (S.D.). In this distribution, the median and the mode are both equal to the mean. parameters: Any one of a set of independent variables in a model whose values determine the characteristics or behavior of the model. plow layer: The depth to which soil is generally tilled; in this Report taken to be 30 cm. porosity: The property of containing interstices. Total porosity is expressed as the ratio of the volume of interstices to total volume. Effective porosity refers to the porosity through which flow occurs. radiation weighting factor (wB): A factor used for radiationprotection purposes that accounts for differences in biological effectiveness between different radiations. The radiation weighting factor (w,) is independent of the tissue weighting factor (w,). relaxation length:The distance by which an exponentially decreasing quantity is reduced to Ve of its initial value. resuspension:A general term used to indicate the process whereby material is transferred from the soil surface to the atmosphere. screening: The process of rapidly identifying potentially important radionuclides and exposure pathways by eliminating those of known lesser significance. screening models:Simple models employing conservative assumptions for the expressed purpose of screening out radionuclides and exposure pathways of negligible importance. shielding factor (SF): A ratio of the exposure rate or dose indoors to that outdoors due to attenuation of radiation by the structure components.

318

/

GLOSSARY

sievert (Sv): The special name for the unit of effective dose and equivalent dose, 1 Sv = 1 J kg-'. site-specific data: Data used in radiological assessment models which are obtained to describe the particular location for which the assessment is being performed. When site-specific data are not available, default values must be used. soil-to-plantconcentration ratio (B,):The ratio of the concentration of a radionuclide i in fresh vegetation to that in dry soil. Cri, the ratio of the concentration of a radionuclide i in dry vegetation to that in dry soil. stochastic model:Any model whose input and output are expressed as random variable. Contrast with deterministic model. tissue weighting factor A factor for a particular tissue representing the fraction of the detriment (cancer plus hereditary effects) attributed to that tissue when the whole body is irradiated uniformly. transfer coefficient to milk (Fmilk): The fraction of element i ingested daily by a cow that is secreted in milk a t steady-state of equilibrium. transfer coefficient to meat (F,,&: The fraction of element i ingested daily by an herbivore that can be measured in 1 kg of animal product at steady-state or equilibrium. triangular distribution: A bounded probability distribution that may be either asymmetric or symmetric. It is usually characterized by three parameters. In a symmetrical distribution the mean, median and mode are all equal. uncertainty: The lack of sureness or confidence in the predictions of models. uncertainty analysis: Analysis of the uncertainty in model predictions. variability: Reflects the fact that the value of a given parameter such as the transfer of a particular ingested radionuclide from diet to bone or the average breathing rate might differ from individual to individual even though the mean or central tendency for a particular group of individuals might be well known.

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The NCRP

The National Council on Radiation Protection and Measurements is a nonprofit corporation chartered by Congress in 1964 to:

1. Collect, analyze, develop and disseminate in the public interest information and recommendations about (a) protection against radiation and (b) radiation measurements, quantities and units, particularly those concerned with radiation protection. 2. Provide a means by which organizations concerned with the scientific and related aspects of radiation protection and of radiation quantities, units and measurements may cooperate for effective utilization of their combined resources, and to stimulate the work of such organizations. 3. Develop basic concepts about radiation quantities, units and measurements, about the application of these concepts, and about radiation protection. 4. Cooperate with the International Commission on Radiological Protection, the International Commission on Radiation Units and Measurements, and other national and international organizations, governmental and private, concerned with radiation quantities, units and measurements and with radiation protection.

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THE NCRP / 335

Presant V i President Secretary and Assistant lkeasurer Assistant Secretary Beusurer

Members S. JAMES ADELSTEIN LARRYE. ANDERSON LYNNR. ANSPAUGH JOHN W. BAUM HAROLDL.BECK MICHAELA. BENDER B. GORDON BLAYLWK BRUCEB. BOECKER JOHN D. BOICE,JR ANDM BOWILLE LESLIEA. BRABY JOHN W. BRAND DAVIDBRENNER ANTONEL. BROOKS PATRICIA A. BUFFLER CHUNGKWANG CHOU JAMES E. CLEAVER J. DONALDCOSSAIRT ALLEN G. CROFF PAULM. DELUCA GAILDE PLANQUE SARAH S. DoNALDSON WILLIAM P. WRNSIFE KEITH F. ECKERMAN MARC EDWARDS H. KEITH m R I G THOMAS F.GESELL ETHELS. GILBEE~T JOHN D. GRAHAM JOEL E.GRAY RAYMOND A. GUILMET~E ERIC J. HALL NAOMIH. HARLEY WILLIAMR. HENDEE DAVIDG. HoEL F. OWENHOFFMAN GEOFFREY R. HOWE DONALDG. JACOBS KENNETH R. KASE AMY K R O m E R G

CHARLES E. RICHARD L E G G ~ HOWARD LlBER JOHN B. LI'ITLE RICHARDA. LUBEN ROGER0.MCCLEU BARBARAJ. MCNEIL CHARLESB. MEINHOLD FRED A. METTLE& JR CHARLESW. MILLER KENNETHL. MILLER DAVIDS. MYERS RONALDC. PETERSEN JOHN W. F~S'TON,SR ANDREWK POZNANSKI R JUL~AN PRESTON GENEMEVES. ROESSLER MARVINROSENSTEIN LAWRENCE N. ROTHENBERG HENRYD. ROYAL MICHAELT. RYAN JONATHAN M. SAMET STEPHENM. SELTZER ROY E. SHORE KENNETHW. SKHABLE DAVID H.SLINEY PAULSLOVIC LOUISE C. STRONG RICHAFtD A. TELL THOMAS S. T E ~ R D E JOHN E. nu LAWRENCE W. TOWNSEND ROBERTL. ULLRICH RICHARD J. VET~ER DAVIDA. WEBER F.WARD WHICKER CHRISG. WHIPPLE J. FRANKWILSON SUSAN D. WILTSHIRE MARVIN C. ZlSKIN

336 / THENCRP Honorary Members LAURISTON S. TAYLO~ Honorary President WARRENK. SLNCW President Emeritus W. ROGERNEY,Executiue Director Emeritus SEYMOUR ABR~UIAMSON

EDWARD L. ALPEN JOHN A. AUXIER WILLIAM J. BAIR VICTOR P. BOND ROBERTL. BRENT REYNOLDE BROWN MELVIN C. CARTER RANDALL. 5. CASWELL P.COWAN FREDERICK JAMESF. CROW GERALDD. DODD PATRICJA W. DURBIN

THOMASS ELY R~CHARDF. FOSPER HYMERL. F'REDELL R.J. MICHAELFRY ROBERT0. GORSON ARTHUR

w. Guy

JOHN W.HEALY BERNDKAHN W=B.MANN

DADEW. MOELLER A. ALAN MOGHISSI

KARL2.MORGAN ROBERTJ. NELSEN

WESLEY L. NYBORG CHESTERR. RICHMOND HARALD H. ROSS] WILLIAMh RUSSELL JOHN H. RUST

EUGENEL. SAENGER WILLW J. SCHULL

J. NEWELL STANNARD JOHN B. STORER A R m C. UPTQN GEORGE L. VOEU EDWARDW. WEBSTER HAROLD0. WCKOFT

h u r i s t o n S.l'hylor Lecturers

The Squares of the Natural Numbers in Radiation Protection Why be Quantitatiue about Radiation Risk Estimates? Radiation Protection--Concepts and Tkak m s From -Quantity of Radiation" and "Dose" to #Exposure" and "Absorbed DoseB-An Historical Review JAMES F.CROW How Well Can WeAssess Genetic Rbk?Not Very EUGENEL. SAENGER Ethics, M - o f f sand Medical Radiation MERRILEISENBUD The Human Environment-Past, h s e n t and Future Limitation and Assessment in Radiation Protection n u t h (and Beauty) in Radiation Measurement Biological Effects of Non-ionizing Radiations: Cell u h r Properties and Interactions How to be Quantitative about Radiation Risk Estimates How Safe is Safe Enough? Radiobiology and Radiation Protection: The Past Century and Pmpects for the Future Radiation Protection and the Internal Emitter saga When is a Dose Not a Dose? Dose and Risk in Diagnostic Radiology: How Big? How Little? WARRENR SLNCLAIR Science. Radiation Protection and the NCRP RJ. MICHAELFRY Mice, Myths and Men AIBRECHT KELLERER Certainty and Uncertainty in Radiation Protection SEYMOURABRAHAMSON 70 Years ofRadiation Genetics: h i t Flies, Mice and Humans W I L L I J. ~ BAIR Radionuclides in the Body: Meeting the Challenge( Eruc J. HALL EZom Chimney Sweeps to Astronauts: Cancer Risks in the Workplace

THENCRP /

337

Currently, the following committees are actively engaged in formulating recommendations: Basic Criteria, Epidemiology, Radiobiology and Risk SC 1-4Extrapolation of Risks from Non-Human Experimental Systems to Man SC 1-6Linearity of Dose Response SC 1-7Information Needed to Make Radiation Protection Recommendations for Travel Beyond Low-Earth Orbit SC 1-8Risk to Thyroid from Ionizing Radiation Structural Shieldmg Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies Up to 10 MeV Operational Radiation Safety SC 46-8Radiation Protection Design Guidelines for Particle Accelerator Facilities SC 46-10Assessment of Occupational Doses from Internal Emitters SC 46-11Radiation Protection During Special Medical Procedures SC 46-13Design of Facilities for Medical Radiation Therapy SC 46-14Radiation Protection Issues Related to Terrorist Activities that Result in the Dispersal of Radioactive Material SC 57-10Liver Cancer Risk SC 57-15Uranium Risk SC 57-17Radionuclide Dosimetry Models for Wounds Environmental Issues SC 6417Uncertainty in Environmental Transport in the Absence of Site-Speafic Data SC 6418 Ecologic and Human Risks from Space Applications of Plutonium SC 64-19Historical Dose SC 6421 Decontamination and Decommissioning of Facilities SC 64-22 Design of Effective Effluent and Environmental Monitoring Programs SC 6423 Cesium in the Environment Biological Effects and Exposure Criteria for Ultrasound Radiation Protection in Mammography Guidance on Radiation Received in Space Activities Fbk of Lung Cancer from Radon Hot Particles in the Eye, Ear or Lung Radioactive and Mixed Waste SC 87-1Waste Avoidance and Volume Reduction SC 87-2Waste Classification Based on Risk SC 87-3Performance Assessment SC 87-4Management of Waste Metals Containing Radioactivity Fluence as the Basis for a Radiation Protection System for Astronauts

338 / THENCRP SC 89

Nonionizing Electromagnetic Fields SC 89-1Biological Effects of Magnetic Fields SC 89-3Biological Effects of Extremely Low-Frequency Electric and Magnetic Fields SC 89-4Biological Effects and Exposure Recommendations for Modulated Radiofrequency Fields SC 89-5Biological Effects and Exposure Criteria for Radiofrequency Fields SC 91 Radiation Protection in Medicine SC 91-1Precautions in the Management of Patients Who Have Received Therapeutic Amounts of Radionuclides SC 91-2Radiation Protection in Dentistry SC 92 Public Policy and Risk Communication SC 93 Radiation Measurement and Dosimetry In recognition of its responsibility to facilitate and stimulate cooperation among organizations concerned with the scientific and related aspects of radiation protection and measurement, the Council has created a category of NCRP Collaborating Organizations. Organizations or groups of organizations that are national or international in scope and are concerned with scientific problems involving radiation quantities, units, measurements and effects, or radiation protection may be admitted to collaborating status by the Council. Collaborating Organizations provide a means by which the NCRP can gain input into its activities from a wider segment of society. At the same time, the relationships with the Collaborating Organizations facilitate wider dissemination of information about the Council's activities, interests and concerns. Collaborating Organizations have the opportunity to comment on draft reports (at the time that these are submitted to the members of the Council). This is intended to capitalize on the f a d that Collaborating Organizations are in an excellent position to both contribute to the identification of what needs to be treated in NCRP reports and to identify problems that might result from proposed recommendations. The present Collaborating Organizations with which the NCRP maintains liaison are as follows: Agency for Toxic Substances and Disease Registry American Academy of Dermatology American Academy of Environmental Engineers American Academy of Health Physics American Association of Physicists in Medicine American College of Medical Physics American College of Nuclear Physicians American College of Occupational and Environmental Medicine American College of Radiology American Dental Association American Industrial Hygiene Association

THE NCRP / 339 American Institute of Ultrasound in Medicine American Insurance Services Group American Medical Association American Nuclear Society American Pharmaceutical Association American Pediatric Medical Association . American Public Health Association American Radium Society American Roentgen Ray Society American Society of Health-System Pharmacists American Society of Radiologic Technologists American Society for Therapeutic Radiology and Oncology Association of University Radiologists Bioelectromagnetics Society Campus Radiation Safety Officers College of American Pathologists Conference of Radiation Control Program Directors, Inc. Council on Radionuclides and Radiopharmaceuticals Defense Special Weapons Agency Electric Power Research Institute Electromagnetic Energy Association Federal Communications Commiseion Federal Emergency Management Agency Genetics Society of America Health Physics Society Institute of Electrical and Electronics Engineera, Inc. Institute of Nuclear Power Operations International Brotherhood of Electrical Workers National Aeronautics and Space Administration National Association of Environmental Professionals National Electrical Manufacturers Association National Institute of Standards and Technology Nuclear Energy Institute Office of Science and Technology Policy Oil, Chemical and Atomic Workers Union Radiation Research Society Radiological Society of North America Society of Nuclear Medicine Society for Risk Analysis United States A . Force United States Army United States Coast Guard United States Department of Energy United States Department of Housing and Urban Development United States Department of Labor United States Department of Transportation United States Environmental Protection Agency United States Navy

340 / THENCRP United States Nuclear Regulatory Commiseion United States Public Health Service Utility Workers Union of America The NCRP has found its relationehips with these organizations to be extremely valuable to continued progress in its program. Another aspect of the cooperative efforts of the NCRP relates to the Special Liaison relationships established with various governmental organizations that have an interest in radiation protection and measurements. This liaison relationship provides: (1)an opportunity for participating organizations to designate an individual to provide liaison between the organization and the NCRP; (2) that the individual designated will receive copies of draft NCRP reports (at the time that these are submitted to the members of the Council) with an invitation to comment, but not vote; and (3) that new NCRP efforts might be discussed with liaison individuals as appropriate, so that they might have an opportunity to make suggestions on new studies and related matters. The following organizations participate in the Special Liaison Program: Australian Radiation Laboratory Central Laboratory for Radiological Protection (Poland) European Commission Federal Office for Radiation Protection (Germany) Health Council of the Netherlands Institute de Protection et de Surete Nucleaire (France) International Commission on Non-Ionizing Radiation Protection Japan Radiation Council Korea Institute of Nuclear Safety National Radiological Protection Board (United Kingdom) National Research Council (Canada) Russian Scientific Commission on Radiation Protection South African Forum for Radiation F'rotedion Ultrasonics Institute (Australia) The NCRP values highly the participation of these organizations in the Special Liaison Program. The Council also benefits significantly from the relationships established pursuant to the Corporate Sponsor's Program. The program facilitates the interchange of information and ideas and corporate sponsors provide valuable fiscal support for the Council's program. This developing program currently includes the following Corporate Sponsors: 3M Health Physics Services Amersham Corporation Commonwealth Edison Consolidated Edison Duke Energy Corporation Landauer, Inc.

THENCRP /

341

New York Power Authority Nuclear Energy Institute Southern California Edison Westinghouse Electric Corporation The Council's activities are made possible by the voluntary contribution of time and effort by its members and participants and the generous support of the following organizations: Agfa Corporation Alfred P. Sloan Foundation Alliance of American Insurere American Academy of Dermatology American Academy &Health Physics American Academy of Oral and Madlofacial Radiology American Association of Physicists in Medicine American Cancer Society American College of Medical Physics American College of Nuclear Physicians American College of Occupational and Environmental Medicine American College of Radiology American College of Radiology Foundation American Dental Association American Healthcare Radiology Administrators American Industrial Hygiene Association American Insurance Services Group American Medical Association American Nuclear Society American Osteopathic College of Radiology American Podiatric Medical Association American Public Health Association American Radium Society American Roentgen Ray Society American Society of Radiologic Technologists American Society for Therapeutic Radiology and Oncology American Veterinary Medical Association American Veterinary Radiology Society Association of University Radiologists Battelle Memorial Institute Canberra Industries, Inc. Chem Nuclear Systems Center for Device8 and Radiological Health College of American Pathologists Committee on Interagency Radiation Research and Policy Coordination Commonwealth of Pennsylvania Consumers Power Company

342 / THE NCRP Council on Radionuclides and Radiopharmaceuticals Defense Nuclear Agency Eastman Kodak Company Edison Electric Institute Edward Mallinckrodt, Jr. Foundation EGCG Idaho, Inc. Electric Power Research Institute Federal Emergency Management Agency Florida Institute of Phosphate Research Florida Power Corporation FuJi Medical Systems, U.S.A, Inc. Genetics Society of America Health Effects Research Foundation (Japan) Health Physics Society Institute of Nuclear Power Operations James Picker Foundation Martin Marietta Corporation Motorola Foundation National Aeronautics and Space Administration National Association of Photographic Manufacturers National Cancer Institute National Electrical Manufacturers Association National Institute of Standards and Technology Picker International Public Service Electric and Gas Company Radiation Research Society Radiological Society of North America Richard Lounsbery Foundation Sandia National Laboratory Siemens Medical Systems, Inc. Society of Nuclear Medicine Society of Pediatric Radiology United States Department of Energy United States Department of Labor United States E n v i r o ~ l e n t dProtection Agency United States Navy United States Nuclear Regulatory Commission Victoreen, Inc. Initial funds for publication of NCRP reports were provided by a grant &om the James Picker Foundation. The NCRP seeks to promulgate information and recommen-dations based on leading scientific judgment on matters of radiation protection and measurement and to foster cooperation among organizations concerned with these matters. These efforts are intended to serve the public interest and the Council welcomes comments and suggestions on its reports or activities from those interested in its work.

NCRP Publications Information on NCRP publications may be obtained from the NCRP and fax website (httpd/www.ncrp.com) or by telephone (800-229-2652) (301-907-8768). The address is: NCRP Publications 7910 Woodmont Avenue Suite 800 Bethesda, MD 20814-3095 Abstracts of NCRP reports published since 1980,abstracts of all NCRP commentaries, and the text of all NCRP statements are available a t the NCRP website. Currently available publications are listed below.

NCRP Reports No. 8 22

23 25 27 30 32 35 36 37 38

41

Title Control and Removal of Radioactive Contamination in Laboratories (1951) Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure (1959)[Includes Addendum 1 issued in August 19631 Measurement of Neutron F l u and Spectra fir Physical and Biological Applications (1960) Measurement of Absorbed Dose of Neutrons, and of Mixtures of Neutrons and Gamma Rays (1961) Stopping Powers for Use with Cavity Chambers (1961) Safe Handling of Radioactive Materials (1964) Radiation Protection in Educational Institutions (1966) Dental X-Ray Protection (1970) Radiation Protection in Veterinary Medicine (1970) Precautions in the Management of Patients Who Have Received Therapeutic Amounts of Radionuclides (1970) Protection Against Neutron Radiation (1971) Specification of Gamma-Ray Brachythrapy Sources (1974)

344 / NCRP PUBLICATIONS Radiological Factors Affecting Decision-Making in a Nuclear Attack (1974) Krypton-85 in the Atmosphere-Accumulation, Biological Significance, and Control lkchnology (1975) Alpha-Emitting Particles in Lungs (1975) h'tium Measurement lkchniques (1976) Structural Shielding Design and Evaluation for Medical Use o f X Rays and Gamma Rays of Energies Up to 10 MeV (1976) Environmental Radiation Measurements (1976) Cesium-137 from the Environment to Man: Metabolism and Dose (1977) Medical Radiation Exposure of Pregnant and Potentially Pregnant Women (1977) Protection of the Thyroid Gland in the Event of Releases of Radioiodine (1977) Instrumentation and Monitoring Meththods for Radiation Protection (1978) A Handbook of Radioactivity Measurements Procedures, 2nd ed. (1985) Operational Radiation Safety Program (1978) Physical, Chemical, and Biological Properties of Radiocerium Relevant to Radiation Protection Guidelines (1978) Radiation Safety Training Criteria for Industrial Radiography (1978) Tritium in the Environment (1979) Tritium and Other Radionuclide Labeled Organic Compounds Incorporated i n Genetic Material (1979) Influence of Dose and Its Distribution in Time on Dose-Response Relationships for Low-LET Radiations (1980) Management of Persons Accidentally Contaminated with Radionuclides (1980) Radiofrequency Electromagnetic Fields-Properties, Quantities and Units, Biophysical Interaction, and Measurements (1981) Radiation Protection i n Pediatric Radiology (1981) Dosimetry of X-Ray and Gumma-Ray Beams for Radiation Therapy in the Energy Range 10 keV to 50 MeV (1981) Nuclear Medicine-Factors Influencing the Choice and Use of Radionuclides in Diagnosis and Therapy (1982) Operational Radiation Safety-Training (1983) Radiation Protection and Measurement fir Low-Voltage Neutron Generators (1983) Protection in Nuclear Medicine and Ultrasound Diagnustic Procedures in Children (1983) Biological Effects of Ultrasound: Mechanisms and Clinical Implications (1983) Iodine-129: Evaluation of Releases from Nuclear Power Generation (1983)

NCRP PUBLICATIONS / 345

77 Exposures from the Uranium Series with Emphusis on Radon and Its Daughters (1984) 79 Neutron Contamination from Medical Electron Accelerators (1984) 80 Inductwn of y r o Cancer by Ionizing Radiation (1985) 81 Carbon-14 in the Environment (1985) 82 SZ Units in Radiation Protection and Measurements (1985) 83 The Experimental Basis for Absorbed-Dose Calculations in Medical Uses of Radionuclides (1985) 84 General Concepts for the Dosimetry of Internally Deposited Radionuclides (1985) 86 Biological E f f i t sand Exposure Criteria for Radiofrequency Electromagnetic Fields (1986) 87 Use of Bioassay Procedures for Assessment of Internal Radionuclide Deposition (1987) 88 Radiation Alarms and Access Control Systems (1986) 89 Genetic Effects from Internally Deposited Radionuclides (1987) 90 Neptunium: Radiation Protection Guideliws (1988) 92 Public Radiation Exposure from Nuclear Power Genemtion in the United States (1987) 93 Ionizing Radiation Exposure of the Population of the United States (1987) 94 Exposure of the Population in the United States and Canada from Natural Background Radiation (1987) 95 Radiation Exposure of the US. Population from Consumer Products and Miscellaneous Sources (1987) 96 Comparative Carcinogenicityof Ionizing Radiation and Chemicals (1989) 97 Measurement of Radon and Radon Daughters in Air (1988) 98 Guidance on Radiation Receiued in Space Activities (1989) 99 Quality Assurance for Diagnostic Imaging (1988) 100 Exposure of the US. Population from Diagnostic Medical Radiation (1989) 102 Medical X-Ray, Electron Beam and Gamma-Ray Protection for Energies Up to 50 MeV (Equipment Design, Performance and Use) (1989) 103 Control of Radon in Houses (1989) 104 The Relative Biological Effectiveness of Radiations of Difirent Quality (1990) 105 Radiation Protection for Medical and Allied Health Personnel (1989) 106 Limit for Exposure to Wet Particles" on the Skin (1989) 107 Implementation of the Principle of As Low As Reasonably Achievable (ALARA)for Medical and Dental Personnel (1990) 108 Conceptual Basis for Calculations ofAbsorbed-DoseDistributions (1991) 109 Effects of Ionizing Radiation on Aquatic Organisms ( 1991)

346 / NCRP PUBLICATIONS

110 Some Aspects of Strontium Radiobiology (1991) 111 Developing Radiation Emergency Plans fir Academic, Medical or Industrial Facilities (1991) 112 Calibration of Survey Instruments Used in Radiation Protection for the Assessment of Ionizing Radiation Fields and Radioactive Surface Contamination (1991) 113 Exposure Criteria for Medical Diagnostic Ultrasound: I. Criteria Based on Thermal Mechanisms (1992) 114 Maintaining Radiation Protection Records (1992) 115 Risk Estimates for Radiation Protection (1993) 116 Limitation of Ezposure to Ionizing Radiation (1993) 117 Research Needs for Radiation Protection (1993) 118 Radiation Protection in the Mineral Extraction Industry (1993) 119 A Practical Guide to the Determination of Human Ezposure to Radiofrequency Fields (1993) 120 Dose Control at Nuclear Power Plants (1994) 121 Principles and Application of Collective Dose in Radiation Protection (1995) 122 Use of Personal Monitors to Estimate Effective Dose Equivalent and EffectiveDose to Workersfbr External Exposure to Low-LET Radiation (1995) 123 Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground (1996) 124 Sources and Magnitude of Occupational and Public Exposures from Nuclear Medicine Procedures (1996) 125 Deposition, Retention and Dosimetry of Inhaled Radioactive Substances (1997) 126 Uncertainties in Fatal Cancer Risk Estimates Used in Radiation Protection (1997) 127 Operational Radiation Safety Program (1998) 128 Radionuclide Exposure of the EmbryolFetus (1998) 129 Recommended Screening Limits for Contaminated Surface Soil and Review of Factors Relevant to Site-Specifi Studies (1999) Binders for NCRP reports are available. Two sizes make it possible to collect into small binders the "old series" of reports (NCRP Reports Nos. 8-30)and into large binders the more recent publicati~ns(NCRP Reports Nos. 32-129).Each binder will accommodate from five to seven reports. The binders cany the identification "NCRPReports" and come with label holders which permit the user to attach labels showing the reports contained in each binder. The following bound sets of NCRP reports are also available: Volume I. NCRP Reports Nos. 8,22 Volume 11. NCRP Reports Nos. 23,25,27,30 Volume 111. NCRP Reports Nos. 32,35,36,37 Volume n!NCRP Reports Nos. 38,40,41

NCRP PUBLICATIONS / 347

Volume V. NCRP Reports Nos. 42,44,46 Volume VI. NCRP Reports Nos. 47,49,50,61 Volume VII. NCRP Reports Nos. 52,53,54,55,57 Volume VIII. NCRP Report No. 58 Volume M.NCRP Reports Nos. 59,60,61,62,63 Volume X. NCRP Reports Nos. 64,65,66,67 Volume M.NCRP Reports Nos. 68,69,70,71,72 Volume XI. NCRP Reports Nos. 73,74,75,76 Volume XIII. NCRP Reports Nos. 77,78, 79,80 Volume XTV. NCRP Reports Nos. 81,82,83,84,85 Volume XV.NCRP Reports Nos. 86.87,88,89 Volume XVI. NCRP Reports Nos. 90,91,92,93 Volume XVII.NCRP Reports Nos. 94,95,96,97 Volume XVIII. NCRP Reports Nos. 98,99,100 Volume XTX. NCRP Reports Nos. 101,102,103,104 Volume XX. NCRP Reports Nos. 105,106,107,108 Volume XXI.NCRP Reporta Nos. 109,110,111 Volume XXII.NCRP Reports Nos. 112,113,114 Volume XXIII.NCRP Reports Nos. 115,116,117,118 Volume XXTV. NCRP Reports Nos. 119,120,121,122 Volume XXV.NCRP Report No. 1231 and 12311 Volume XXVI. NCRP Reports Nos. 124,125,126,127 (Titlea of the individual reports contained in each volume are given above.)

NCRP Commentaries No. 1 4

5 6 7

8 9

Title Krypton-85 in the Atmosphere-With Specific Reference to the Public Health Signijkance of the Pmposed Controlled Release at Three Mile Island (1980) Guidelines for the Release of Waste Water from Nuclear Facilities with Special Reference to the Public Health Significance of the Proposed Release of k t e d Waste Waters at Three Mile Island (1987) Review of the Publication, Living Without Landfills (1989) Radon Exposure of the US. Population--Status of the Problem (1991) Misadministration of Radioactive Material in Medicine-Scientific Background (1991) Uncertainty in NCRP Screening Models Relating to Atmospheric Transport, Deposition and Uptake by Humans (1993) ConsiderationsRegarding the Unintended Radiation Exposure of the Embryo, Fetus or Nursing Child (1994)

348 / NCRP PUBLICATIONS Advising the Public about Radiation Emergencies:A Document for Public Comment (1994) Dose Limits for Individuals Who Receive Exposure from Radionuclide Therapy Patients (1995) Radiation Exposure and High-Altitude Flight (1995) An Introduction to E m in Diagnostic Radiology and Nuclear Medicine (Justification of Medical Radiation Exposure) (1995) A Guide for Uncertainty Analysis in Dose and Risk Assessments Related to Environmental Contamination (1996) Evaluating the Reliability of Biokinetic and Dosimetric Moclels and Parameters Used to Assess Z n d i v i d d Doses for Risk Assessment Purposes (1998)

Proceedings of the Annual Meeting No. 1

Title Perceptions ofRisk, Proceedings of the Fifteenth Annual Meeting held on March 14-15, 1979 (including Taylor Lecture No. 3) (1980) Critical Issues in Setting Radiation Dose Limits, Proceedings of the Seventeenth Annual Meeting held on April 8-9, 1981 (including Taylor Lecture No. 5) (1982) Radiation Protection and New Medical Diagnostic Approaches, Proceedings of the Eighteenth Annual Meeting held on April 6-7,1982 (including Taylor Lecture No. 6) (1983) Environmental Radioactivity, Proceedings of the Nineteenth Annual Meeting held on April 6-7,1983 (including Taylor Lecture No. 7) (1983) Some Issues Important in Developing Basic Radiation Pmtection Recommendations, Proceedings of the Twentieth Annual Meeting held on April 4-5,1984 (includingTaylor Lecture No. 8) (1985) Radioactive Waste, Proceedings of the Twenty-first Annual Meeting held on April 3-4, 1985 (includingTaylor Lecture No. 9x1986) Nonionizing Electromagnetic Radiations and Ultrasound Proceedings of the Twenty-secondhnua.1Meeting held on April 2-3,1986 (including Taylor Lecture No. 10)(1988) New Dosimetry at Hiroshima and Nagasaki and Its Implications for Risk Estimates, Proceedings of the 'henty-third Annual Meeting held on April 8-9,1987 (including Taylor Lecture No. 11)(1988) Radon, Proceedings of the Twenty-fourthAnnual Meeting held on March 30-31, 1988 (including Taylor Lecture No. 12) (1989)

NCRP PUBLICATIONS

/ 349

11 M i a t i o n Protection Ibday-The NCRP at Sixty Years, Proceedings of the Twenty-fifth Annual Meeting held on April 5-6, 1989 (including Taylor Lecture No. 13) (1990) 12 Health and Ecological Implications of Radioactively Contaminated Environments, Proceedings of the Twenty-sixth Annual Meeting held on April 4-5, 1990 (including Taylor Lecture No. 14) (1991) 13 Genes, Cancer and Radiation Protection, Proceedings of the Twenty-seventh Annual Meeting held on April 3-4, 1991 (including Taylor Lecture No. 15) (1992) 14 Radiation Protection in Medicine, Proceedings of the Twenty-eighth Annual Meeting held on April 1-2, 1992 (including Taylor Lecture No. 16) (1993) 15 Radiation Science a n d Societal Decision Making, Boceedings of the Twenty-ninth Annual Meeting held on April 7-8, 1993 (including Taylor Lecture No. 17) (1994) 17 Environmental Dose Reconstruction and Risk Implications, Proceedings of the Thirty-first Annual Meeting held on April 12-13, 1995 (including Taylor Lecture No. 19) (1996) 18 Implications ofNew Data on Radiation Cancer Risk, Proceedings ofthe Thirty-second Annual Meeting held on April 3-4, 1996 (including Taylor Lecture No. 20) (1997)

Lauriston S. Taylor Lectures No. 1 2

3 4

5 6 7

Title The Squares of the Natural Numbers in Radiation Protection by Herbert M. Parker (1977) Why be Quantitative about Radiation Risk Estimates? by Sir Edward Pochin (1978) Radiation Protection-Concepts a n d Dude Offs by Hymer L. F'riedell(1979) [Available also in Perceptions of Risk, see abovel From "Quantity of Radiation" and "Dose" to "Exposure" and "Absorbed Dosem-An Historical Review by Harold 0.Wyckoff (1980) How Well Can We Assess Genetic Risk?Not Very by James F. Crow (1981) [Available also in Critical Issues in Setting Radiation Dose Limits, see above] Ethics, Pa&-offs and Medical Radiation by Eugene L. Saenger (1982) [Available also in Radiation Protection and New Medical Diagnostic Approaches, see abovel !l7w Human Environment-Past, Present and Future by Merril Eisenbud (1983) [Available also in Environmental Radioactivity, see above]

350 / NCRP PUBLICATIONS Limitation and Assessment in Radiation Protection by Harald H . Rossi (1984) [Available also in Some Issues Important i n Developing Basic Radiation Protection Recommendations, see above] Truth (and Beauty) i n Radiation Measurement by John H. Harley (1985) [Available also in Radioactive Waste, see abovel Biological Effects of Non-ionizing Radiations: Cellular Properties and Interactions by Herman F! Schwan (1987)[Available also in Nonionizing Electronagnetic Radiations and Ultrasound, see abovel How to be Quuntitative about Radiation Risk Estimates by Seymour Jablon (1988) [Available also in New Dosimetry at Hiroshima and Nagasaki and its Implications for Risk Estimates, see above] How Safe is Safe Enough? by Bo Lindell(1988) [Available also in Radon, see above] Radiobiology and Radiation Protection: The Past Century and Prospects for the Future by Arthur C. Upton (1989) [Available also in Radiation Protection M y , see above] Radiation Protection and the Internal Emitter Saga by J. Newel1 Stannard (1990) [Available also in Health and Ecological Implications of Radioactiuely Contaminated Environments, see above] When is a Dose Not a Dose? by Victor P. Bond (1992) [Available also in Genes, Cancer and Radiation Protection, see abovel Dose and Risk i n Diagnostic Radiology: How Big? How Little? by Edward W. Webster (1992)[Available also in Radiation Protection in Medicine, see above] Science, Radiation Protection and the NCRP by Warren K Sinclair (1993)[Available also in Radiation Science and Societal Decision Making, see above] Mice, Myths and Men by R.J. Michael Fry (1995)

Symposium Proceedings

No. 1

Title The Control of Exposure of the Public to Ionizing Radiation in the Event of Accident or Attack, Proceedings of a Symposium held April 27-29,1981(1982) Radioactive and Mixed Waste-Risk as a Basis for Waste Classification, Proceedings of a Symposium held November 9, 1994 (1995) Acceptability of Risk from Radiation-Application to Human Space Flight, Proceedings of a Symposium held May 29,1996 (1997)

NCRP PUBLICATIONS / 35 1

NCRP Statements

No.

Title

1

"Blood Counts, Statement of the National Committee on Radiation Protection," Radiology 63,428 (1954)

2

"Statements on Maximum Permissible Dose from Television Receivers and Maximum Permieeible Dose to the G2n of the Whole Body," Am. J. Roentgenol., Radium Ther. and Nucl. Med. 84, 152 (1960) and Radiology 75,122 (1960)

3

X-Ray Protection Standards for Home 7klevision Receivers, Interim Statement of the National Council on Radiation Protection and Measurements (1968)

4

Specifhation of Units of Natural Uranium and Natural Thorium, Statement of the National Council on Radiation Protection and Measurements (1973)

5

NCRP Statement on Dose Limit fir Neutrons (1980)

6

Control of Air Emissions of Radionuclides (1984)

7

The Probability That a Particular Malignancy May Have Been Caused by a Specified Irradiation (1992)

Other Documents The following documents of the NCRP were published outside of the NCRP report, commentary and statement series:

Somatic Radiation Dose for the General Population, Report of the Ad Hoc Committee of the National Council on Radiation Protection and Measurements, 6 May 1959, Science, February 19, 1960, Vol. 131, No. 3399, pages 482-486 Dose Effect Modifying Factors In Radiation Protection, Report of Subcommittee M-4(Relative Biological Effectiveness)of the National Council on Radiation Protection and Measurements, (1967) Brookhaven National Report BNL 50073 (T-471) Laboratory (National Technical Information Service Splulgfield, Virginia)

Index AMAD 52,315 Aquifers 28, 30 Buried sources 30

Cesium-137 26 Cleanup of contaminated sites 8 Critically exposed group 3 Distribution of doses 152-165 Distribution parameter 3, 316, 318 standard deviation 3 geometric standard deviation 3, 316 triangular distribution 3, 318 Dose factor 301-314

construction, commercial, industrial 7 heavily vegetated pasture 5 heavily vegetated rural 6 no food 7 parameter values and uncertainty 154-156 screening dose 57,58 sparsely vegetated pasture 6 sparsely vegetated rural 6 total dose 18 Maximum annual dose 3, 8 Maximum dose rate 2,8, 17 conservatism 2, 17 Monte Carlo analysis 17, 152-157,317

External dose model 34-36, 36-57 model parameters 36-57

NCRP screening

Geometric standard deviation 3, 316 Geophagia 117-134

Pica 117-119 Plutonium 17, 26

Health effect calculation 1

Resuspension-migration model 59-78 Retardation factor 29

Ingested radio nuclides 91-138 direct ingestion of dirt 116-134 dose model 91-135 model parameters 99-135 Inhaled radionuclide model 59-90 dose model parameters 78-85 screening doses 88-90 Land-use scenarios 3-7, 18, 27, 28, 57,58,154-156, 166-293 agricultural 5 calculated screening doses 166-293

Natural emitters 30,31 recommendations 4, 17

Safety fador 17-25 Screening doses 57, 58 Screening guidance 2-33 scientific basis 33 Screening limits 3, 4, 27 Shielding factors 301-314, 317 Site-specific dose assessment 1, 2, 31, 318 Soil partition coefficients 29 Soil radionuclide concentration 139-151 sampling programs 139

INDEX strategy 148-151 Soil screening limits 1,4,8 recommended 8 Special situations 28-31 aquifers 28,30 buried sources 30 milk 29 natural emitters 30,31

/

353

Triangular distribution 3,318 Uncertainty 3, 30, 152-157, 318 Uranium 17,26 Variability 3, 318