Page 1 of 52

DOE Review Release 04/27/2018 

Internal Dose Reconstruction

ORAUT-OTIB-0060 Rev. 02

Effective Date: 04/20/2018

Supersedes: Revision 01

Subject Expert(s): Elizabeth M. Brackett

Document Owner 

Approval: Signature on File

James P. Griffin, Document Owner

Approval Date: 04/18/2018

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 04/18/2018

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 04/20/2018

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 2 of 52

 

PUBLICATION RECORD

EFFECTIVE REVISION 

DATE NUMBER DESCRIPTION

02/06/2007 00 This technical information bulletin provides guidance in the 

performance of internal dose reconstructions. It incorporates the 

requirements of ORAUT-PROC-0003, Internal Dose Reconstruction, 

which will be cancelled. Incorporates internal formal and NIOSH 

review comments. Incorporates additional NIOSH review comments. 

There is no change in the assigned dose and no PER required. 

Training is required: As determined by the Task Manager. Initiated 

by Elizabeth M. Brackett.

09/08/2014 01 Revision to bring the document up to date with current practice and 

to provide additional guidance to dose reconstructors. Incorporates 

formal internal and NIOSH review comments. Constitutes a total 

rewrite of the document. Training is required: As determined by the 

Objective Manager. Initiated by Elizabeth M. Brackett. 

04/20/2018 02 Revision to address comments from the Procedures Review 

Subcommittee and to provide additional guidance and information to 

the dose reconstructors on current methods. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training is required: As determined by the Objective 

Manager. Initiated by Elizabeth M. Brackett. 
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TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

1.0 Introduction .............................................................................................................................. 8

1.1 Purpose........................................................................................................................ 8

1.2 Scope........................................................................................................................... 8

2.0 Terminology ............................................................................................................................. 9

2.1 Systemic Versus Nonsystemic...................................................................................... 9

2.2 Metabolic Versus Nonmetabolic Organs....................................................................... 9

2.3 Bioassay....................................................................................................................... 9

2.3.1 In Vitro Bioassay ............................................................................................ 9

2.3.1.1 Urinalysis..................................................................................... 9

2.3.1.2 Fecal Samples .......................................................................... 10

2.3.2 In Vivo Bioassay .......................................................................................... 10

2.3.2.1 Whole-Body Counts .................................................................. 10

2.3.2.2 Chest Counts ............................................................................ 10

2.3.2.3 Organ Counts............................................................................ 11

2.3.2.4 Wound Counts .......................................................................... 11

2.4 Positive and Negative Results .................................................................................... 11

2.4.1 Definitions .................................................................................................... 11

2.4.1.1 Decision Level........................................................................... 11

2.4.1.2 Minimum Detectable Amount..................................................... 11

2.4.1.3 Reporting Level ......................................................................... 11

2.4.2 Determination of a Positive Result for Dose Reconstruction......................... 12

2.4.3 Occupational Uranium Intakes ..................................................................... 12

2.4.4 Nonoccupational Cesium Intakes ................................................................. 12

2.5 Types of Doses and Associated Distributions ............................................................. 12

2.5.1 Philosophy of Internal Dose Assignment ...................................................... 13

2.5.2 Missed Dose ................................................................................................ 13

2.5.3 Fitted Dose................................................................................................... 14

2.5.4 Environmental Intakes.................................................................................. 14

2.5.5 Coworker Dose ............................................................................................ 14

3.0 Intake and Dose Assessment Guidance................................................................................. 15

3.1 Selection of Appropriate Organs for Dose Estimate.................................................... 15

3.2 Selection of Parameter Values ................................................................................... 15

3.2.1 Intake Mode ................................................................................................. 16

3.2.2 Particle Size Distribution .............................................................................. 16

3.2.3 Material Type ............................................................................................... 16

3.2.3.1 Standard Types......................................................................... 16

3.2.3.2 Additional Plutonium Types ....................................................... 17

3.2.3.3 Intakes of Mixtures .................................................................... 17

3.2.3.4 Ingestion Intakes ....................................................................... 18

3.2.4 Radiation Type............................................................................................. 19

3.2.5 Normalization of Bioassay Data ................................................................... 19

3.2.6 IMBA-Specific Parameters and Information.................................................. 20

3.2.6.1 Data Type.................................................................................. 20

3.2.6.2 Measurement Error ................................................................... 20

3.2.6.3 Error Distribution ....................................................................... 21
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3.2.6.4 IMBA Limitations ....................................................................... 21

3.3 Determination of Missed and Unmonitored Intake and Dose ...................................... 22

3.3.1 Potential for Exposure.................................................................................. 22

3.3.2 Missed Dose Determination ......................................................................... 23

3.3.3 Unmonitored Dose ....................................................................................... 24

3.3.4 Missed Versus Unmonitored Dose ............................................................... 24

3.3.5 Coworker Data ............................................................................................. 25

3.3.6 Example Assignment of Coworker and Unmonitored Doses ........................ 26

3.3.7 Short-Duration Missed Dose ........................................................................ 26

3.4 Fitting Positive Bioassay Results ................................................................................ 27

3.4.1 General Philosophy...................................................................................... 27

3.4.2 General Guidelines ...................................................................................... 27

3.5 Overall Assessment Methods ..................................................................................... 29

3.5.1 Overestimate................................................................................................ 29

3.5.1.1 Generic Overestimates.............................................................. 29

3.5.1.2 Individual-Specific Overestimates.............................................. 29

3.5.2 Underestimate.............................................................................................. 30

3.5.3 Best Estimate............................................................................................... 30

3.5.3.1 Performing a Best Estimate Using Bioassay Data ..................... 30

3.5.3.2 Multiple Cancers........................................................................ 31

3.6 Working with Multiple Bioassay Types........................................................................ 31

3.6.1 Fitting Multiple Bioassay Types.................................................................... 31

3.6.2 Projections ................................................................................................... 32

3.7 Specific Issues ........................................................................................................... 32

3.7.1 Plutonium Mixtures....................................................................................... 32

3.7.1.1 Background Information ............................................................ 32

3.7.1.2 Assumptions for Use in Dose Reconstruction............................ 33

3.7.1.2.1 Mixture of Material ............................................... 33

3.7.1.2.2 Age of Material .................................................... 33

3.7.1.3 Americium Ingrowth to Plutonium Mixture.................................. 33

3.7.2 Methods for Assessing Isotopic Uranium Results......................................... 35

3.7.2.1 Alpha Spectroscopy .................................................................. 35

3.7.2.1.1 Underestimate ..................................................... 35

3.7.2.1.2 Overestimate ....................................................... 35

3.7.2.1.3 Best Estimate ...................................................... 36

3.7.2.2 Inductively Coupled Plasma-Mass Spectrometry....................... 36

3.7.3 Assignment of Thoron and Radon Dose....................................................... 37

3.7.4 Assessment of Mixtures of Radionuclides .................................................... 37

4.0 Additional Internal Dosimetry Guidance Documents............................................................... 38

4.1 OCAS-IG-002, Internal Dose Reconstruction Implementation Guideline..................... 38

4.2 OCAS-TIB-009, Estimation of Ingestion Intakes ......................................................... 38

4.3 DCAS-TIB-011, Lung Dose Conversion Factor for Thoron WLM ................................ 38

4.4 DCAS-RPT-005, Alternative Dissolution Models for Insoluble Pu-238 ........................ 38

4.5 ORAUT-OTIB-0001, Maximum Internal Dose Estimates for Savannah River 

Site (SRS) Claims....................................................................................................... 38

4.6 ORAUT-OTIB-0011, Tritium Calculated and Missed Dose Estimates ......................... 38

4.7 ORAUT-OTIB-0014, Assignment of Environmental Internal Doses for 

Employees Not Exposed to Airborne Radionuclides in the Workplace........................ 39

4.8 ORAUT-OTIB-0018, Internal Dose Overestimates for Facilities with Air 

Sampling Programs .................................................................................................... 39

4.9 ORAUT-OTIB-0022, Guidance on Wound Modeling for Internal Dose 

Reconstruction ........................................................................................................... 39
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4.10 ORAUT-OTIB-0033, Application of Internal Doses Based on ClaimantFavorable Assumptions for Processing as Best Estimates ......................................... 39

4.11 ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly Retained in the 

Lung ........................................................................................................................... 40

4.12 ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal 

Dose-Related Gross Beta and Gross Gamma Analyses............................................. 40

4.13 ORAUT-OTIB-0066, Calculation of Dose from Intakes of Special Tritium 

Compounds................................................................................................................ 40

5.0 Attributions and Annotations .................................................................................................. 40

References ........................................................................................................................................ 41

ATTACHMENT A RADIATION TYPES BY NUCLIDE FOR ENTRY INTO IREP.......................... 44

ATTACHMENT B ICRP PUBLICATION 68 DEFAULT MATERIAL TYPES ................................. 45

ATTACHMENT C RETENTION OF MATERIAL IN THE LUNGS................................................. 48

LIST OF TABLES

TABLE TITLE PAGE

2-1 Mean body burden of 137Cs in the United States .................................................................... 13

3-1 Selected absorption types for recycled uranium contaminants ............................................... 18

3-2 Selected f1 values for recycled uranium contaminants........................................................... 19

3-3 Nuclides for which IMBA calculates incorrect organ doses..................................................... 22

A-1 Radiation types by nuclide for entry into IREP........................................................................ 44

B-1 Publication 68 default material types...................................................................................... 45

LIST OF FIGURES

FIGURE TITLE PAGE

3-1 Using IMBA to determine Publication 68 default material types.............................................. 17

3-2 IMBA window showing uranium default material types and f1 values ..................................... 18

C-1 Lung retention following an acute inhalation intake of 1 Bq of type F material........................ 49

C-2 Lung retention following an acute inhalation intake of 1 Bq of type M material ....................... 50

C-3 Lung retention following an acute inhalation intake of 1 Bq of type S material........................ 50

C-4 Lung retention of a 365-d chronic inhalation intake of a type F material ................................. 50

C-5 Lung retention of a 365-d chronic inhalation intake of a type M material ................................ 51

C-6 Lung retention of a 365-d chronic inhalation intake of a type S material................................. 51

C-7 Lung retention of a 1,000-d chronic inhalation intake of a type M material ............................. 51

C-8 Lung retention of a 1,000-d chronic inhalation intake of a type S material.............................. 52
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ACRONYMS AND ABBREVIATIONS 

BB bronchial region

Bq becquerel

CAD Chronic Annual Dose (tool)

CFR Code of Federal Regulations

d day

DCAS Division of Compensation Analysis and Support

DL decision level

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ET1 extrathoracic region 1

ET2 extrathoracic region 2

g gram

GI gastrointestinal

GSD geometric standard deviation

hr hour

ICD-9 The International Classification of Diseases, 9th Revision

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Assessment

ICP-MS inductively coupled plasma–mass spectrometry

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

L liter

LLI lower large intestine

LN(ET) extrathoracic lymph nodes

LN(TH) thoracic lymph nodes

LOD limit of detection

MDA minimum detectable amount

MDD minimum detectable dose

mL milliliter

MPC maximum permissible concentration

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

PER Program Evaluation Report

PID Principal Internal Dosimetrist

POC probability of causation
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REF radiation effectiveness factor

SI small intestine

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

TIB technical information bulletin

U.S.C. United States Code

ULI upper large intestine

wk week

WLM working level month

μCi microcurie

μg microgram

μm micrometer

μrad microrad

§ section or sections
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1.0 INTRODUCTION

Technical information bulletins (TIBs) are not official determinations made by the National Institute for 

Occupational Safety and Health (NIOSH) but are rather general working documents that provide 

historical background information and guidance to assist in the preparation of dose reconstructions at 

particular sites or categories of sites. They will be revised in the event additional relevant information 

is obtained about the affected site(s). TIBs may be used to assist NIOSH staff in the completion of 

individual dose reconstructions.

In this document, the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy (DOE) facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 [42 U.S.C. § 7384l(5) and (12)].

1.1 PURPOSE

The purpose of this TIB is to provide information and guidance for reconstructing internal dose and to 

document the rationale for selection of certain default parameters.

1.2 SCOPE

There are many approaches that can be taken when reconstructing the internal dose component for a 

claim. These options depend on several factors including the employment site, type and number of 

cancers, and availability of monitoring data for the energy employee. There may be approaches in 

this document that are not applicable or appropriate to a specific claim. 

The terminology and methods in this TIB are applicable for reconstructing doses on the Project and 

are not necessarily reflective of standard internal dosimetry practices in an operational program, 

where the program is being conducted to meet regulatory standards. Quantities of interest are 

different, and methods in this document are intended to be favorable to the claimant when a 

parameter is unknown rather than using the typical operational approach of a “most likely” value when 

there is uncertainty.

This document provides default values for use only when there is no better information to be found in 

the claim file or the site profile. Worker information also takes precedence over default values in the 

site profile. For example, most site profiles contain tables of minimum detectable amounts (MDAs) for 

urine sample results. If the claim file includes an explicit or implied MDA or reporting level in the 

bioassay result listing (e.g., “<” appears in front of a value), this specific value should be applied to the 

claim rather than the default value from the site profile. Note that this direction does not apply if that 

information is determined to be invalid, such as in the case of laboratory results that have been 

demonstrated to be inaccurate.

Section 2.0 defines terminology. Section 3.0 provides guidance for internal dose reconstruction. 

Section 4.0 provides a listing of additional documents containing internal dosimetry guidance and 

documentation, along with a description of the content. Attributions and annotations, indicated by 

bracketed callouts and used to identify the source, justification, or clarification of the associated 

information, are presented in Section 5.0. Attachment A lists radiation types by nuclide for entry into 

the Interactive RadioEpidemiological Program (IREP), Attachment B lists the International 

Commission on Radiological Protection (ICRP) Publication 68 (ICRP 1995a) material types for all 

elements of interest, and Attachment C addresses retention of material in the lungs. 
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2.0 TERMINOLOGY

2.1 SYSTEMIC VERSUS NONSYSTEMIC

“Systemic” organs and tissues are those to which radioactive material is transferred through blood 

circulation.

“Nonsystemic” organs are those in which material is directly deposited in the organ, specifically the 

respiratory tract through inhalation and the gastrointestinal (GI) tract through ingestion. The lymph 

nodes are included in the nonsystemic organs. The following regions are defined by the ICRP: 

• The respiratory tract includes extrathoracic regions 1 and 2 (ET1 and ET2), lungs, thoracic 

lymph nodes [LN(TH)], extrathoracic lymph nodes [LN(ET)], and bronchial region (BB) 

(ICRP 1994).

• Components of the GI tract include the stomach, small intestine (SI), lower large intestine 

(LLI), upper large intestine (ULI), and colon (note that the colon is simply a weighted average 

of the LLI and ULI) (ICRP 1979).

2.2 METABOLIC VERSUS NONMETABOLIC ORGANS

Several organs are included in the general models in ICRP Publication 78 (ICRP 1998). However, for 

a given element, only a specific subset of these organs is included in the metabolic or biokinetic 

modeling. Organs that are specifically modeled are referred to as “metabolic” for this Project. Others 

for which a dose is calculated but are not specified by the ICRP element-specific model are referred to 

as “nonmetabolic.” The biokinetic models are based on the behavior of the particular element in the 

body, so the metabolic organs vary with the element of interest. For all elements, the metabolic 

organs include the GI tract (ICRP 1979) and, in the case of inhalation, the respiratory tract (ICRP 

1994), because material is always deposited in these regions. In addition, if any part of the bone is 

specified in the ICRP model for an element, all bone parts, including red bone marrow and bone 

surface, are considered to be metabolic.

2.3 BIOASSAY

2.3.1 In Vitro Bioassay

In vitro bioassay refers to samples that are removed from the body for analysis. The most common 

sample type is urine, followed by feces. Other examples include blood, breath, and tissue samples; 

these are typically used as indicators of an intake but are generally not used in a dose assessment.

2.3.1.1 Urinalysis

Urine sampling is the most common bioassay method across the complex, and was the first bioassay 

method implemented at most of the covered facilities. Samples can be collected in the workplace or 

at home. Translocation of material to the bladder and subsequent elimination via urine occurs over 

many hours, so urine collected within 24 hours of an acute intake must be used with extreme caution, 

and those collected within a few hours of intake are often not representative of the intake.

Samples can be in a variety of forms:

• Spot. These consist of a single voiding of the bladder.
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• 24-hour. This includes all voids collected over a 24-hour period. By convention, the sample 

date (and sometimes time) of the last void is used to indicate the sample date and is entered 

in IMBA as the measurement date.

• Simulated 24-hour. This varies among sites, but most often consists of samples collected over 

two consecutive nights. This includes the last void before going to bed, the first void of the 

morning, and all voids in between. This is intended to replace a true 24-hour sample, which is 

less convenient for the worker to collect.

• Minimum volume. A worker can be given a bottle and asked to fill it rather than being given a 

specified time. Common requested volumes are 500 mL and 1 L.

While urinalysis is not always the most sensitive bioassay technique, it can detect many nuclides for 

long periods after an intake and excretion is far more consistent than in feces, so this is often the 

preferred measurement type for an intake assessment.

2.3.1.2 Fecal Samples

Fecal analysis is a very sensitive procedure because more than half of the material deposited in the 

upper respiratory tract clears rapidly to the stomach and GI tract. Total fecal elimination for the first 

few days after an exposure can provide the earliest and most accurate assessment of an intake, 

assuming all excreta for several days is collected. For this reason, it is an important part of a special 

bioassay program that is implemented in response to an incident. However, interpretation of a result 

without a known time of intake, which is the case with a routine or confirmatory sample, can yield 

large uncertainties because fecal voiding patterns are highly variable, both among individuals and for 

a given individual.

2.3.2 In Vivo Bioassay

In vivo bioassay is the direct measurement of activity from radioactive material that is deposited in the 

body. Techniques include a whole-body count, chest count, or specific organ measurement, the most 

common being thyroid. In vivo counting is a late entry into bioassay methods, having been developed 

in the late 1950s, and was not common until the 1960s.

2.3.2.1 Whole-Body Counts

A whole-body count measures high-energy gamma rays being emitted from anywhere in the body, 

although some counters are collimated such that the view is the torso only. The set point for the lower 

end of the energy range depends on the nuclides of interest, the particular counter, and the sitespecific instrumentation settings, but is often around 100 keV.

2.3.2.2 Chest Counts

Chest counts are intended as a measure of material in the lungs but include any activity in the ribs. 

Low-energy gammas are of interest in these counts, and they are primarily used for the detection of 

long-term deposition of plutonium or uranium in the lungs. Note that type F material is retained in the 

lungs for a very short time, so a chest count is not an appropriate technique for monitoring or 

assessing intakes of type F material, either acute or chronic. Attachment C contains additional 

discussion and graphs on the retention of all three material types in the lungs.
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2.3.2.3 Organ Counts

Thyroid counts are frequently used to assess intakes of iodine, which concentrates in the thyroid. 

IMBA has an option for performing an assessment from thyroid measurements.

Some sites performed skull counts in order to determine how much activity was deposited in the 

bones. This information was then used to correct chest counts for activity deposited in the ribs rather 

than in the lungs. This count type is not common and is not used in the dose assessment. 

2.3.2.4 Wound Counts

Some sites might have performed wound counts to help determine how much activity was deposited 

at a wound site and to aid in the decision for actions to be taken (e.g., cleansing of the area, excision, 

and chelation). These results are not used directly in an assessment but can be used to help 

determine if an intake has occurred.

2.4 POSITIVE AND NEGATIVE RESULTS

2.4.1 Definitions

2.4.1.1 Decision Level

Decision level (DL) is defined in ANSI/HPS N13.30-2011, An American National Standard – 

Performance Criteria for Radiobioassay, as the number of counts measured or final instrument 

measurement of a quantity of analyte at or above which a decision is made that the analyte is 

definitely present (HPS 2011). The term “definitely present” used in the standard is somewhat 

ambiguous, but what is meant is that the analyte is deemed to be present with a given probability α of 

being wrong. The probability α is often called the false-positive rate, and it describes the long-run 

frequency of deciding that analyte is present in a sample when in reality none is present. In an 

operational setting, the result of a specific analysis is compared to the DL in order to decide if analyte 

is present in that sample.

2.4.1.2 Minimum Detectable Amount

Minimum detectable amount (MDA) is defined in ANSI/HPS N13.30-2011 as being the smallest 

amount (activity or mass) of an analyte in a sample that will be detected with a probability β of 

nondetection (Type II error) while accepting a probability α of erroneously deciding that a positive 

(nonzero) quantity of analyte is present in an appropriate blank sample (Type I error) (HPS 2011). In 

other words, the MDA is the amount of analyte in samples that would produce results below the DL a 

fraction β of the time over the long run. The MDA is used to characterize the detection capabilities of 

the process, and HPS (2011) states that it should not be compared to a specific analytical result to 

make the detection decision. However, it is appropriate for the determination of missed dose, the 

purpose of which is to account for the detection capabilities and limitations of the bioassay monitoring 

that was performed. 

2.4.1.3 Reporting Level

For this Project, a reporting level is defined as the level below which a value is not recorded for a

bioassay result in the worker’s record, i.e., the result is censored at the reporting level. In some 

instances, a site might have applied a reporting level that is greater than the MDA. In such cases, 

only measurements with values that exceed the reporting level are recorded in the employee files. 

That is, results less than the reporting level are considered less-than values and are recorded as 0 or 

“<[the reporting level].” In these cases, the reporting level, not the MDA, is used to determine the
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missed dose. This practice of censoring the data is most common when the nuclide is easily 

detected, such as 3

H, and a result at the MDA produces a very small dose.

2.4.2 Determination of a Positive Result for Dose Reconstruction

When discussing bioassay analyses for dose reconstruction, a “positive result” typically refers to one 

that is greater than the MDA or reporting level while a “negative result” is less than that value. In the 

early years of the weapons complex, most sites used the MDA to determine if activity had been 

detected. In more recent years, sites have used DL to make the decision about the presence of 

activity. Results between the DL and MDA are accounted for in the missed dose calculation so are 

not considered to be positive when performing a dose reconstruction.

It is also possible for a result to be a true negative value (i.e., less than zero). Both uses of these 

words can be found in Project documents so it is important to be aware of which meaning is being 

discussed.

Many sites recorded the value of the MDA in the employee records for results less than the MDA. In 

such cases, a result recorded with a value of the MDA is considered to be negative. If it is clear that 

this is not the case, a result equal to the MDA should be treated as positive. If it is unclear, a result 

equal to the MDA is considered negative. 

A sample-specific MDA takes precedence over the generic MDA in the site profile.

2.4.3 Occupational Uranium Intakes

Uranium is naturally occurring in the earth’s crust and can be found in the urine and feces of 

individuals with no occupational exposure. This uranium background can vary greatly dependent on 

local geology, eating habits, and water source. While some sites have set levels below which they 

consider a result to be attributable to background, any result other than a baseline sample that is 

positive (as defined in Section 2.4.2) is assumed to be due to an occupational exposure for the 

purposes of dose reconstruction.

2.4.4 Nonoccupational Cesium Intakes

Fallout affected everyone in North America, and body burdens of 137Cs measurable in whole-body 

counters were common in the 1960s and 1970s. Table 2-1 provides mean body burdens of 137Cs for 

the United States from NCRP Report 94 (NCRP 1987) for the years when this fallout was most likely 

to produce interference with occupational whole-body count results. If whole-body count results show 

detection of only 40K and 137Cs, and the 137Cs result is less than the value in Table 2-1, the 137Cs 

results can be assumed to be due to fallout. Values exceeding these levels are assumed to be due to 

occupational intakes, and the full value is used in the assessment (i.e., the values in the table are not 

subtracted from the measured result). If other fission or activation product radionuclides are present 

in the whole-body count, assume any positive 137Cs result is from occupational sources and assess 

based on the full value. 

2.5 TYPES OF DOSES AND ASSOCIATED DISTRIBUTIONS

This section describes the different types of doses that can be assigned to an individual, which 

depend on several factors including the site of employment, available data, and potential for 

exposure. Appropriate distributions to be entered into IREP are discussed. Section 3.3 discusses the 

applicable conditions for assigning each type of dose and provides guidance on the calculations.
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2.5.1 Philosophy of Internal Dose Assignment

The general assumption for internal dose reconstruction is that an individual was constantly exposed

to at least some level of airborne radioactivity throughout employment. At a minimum, ambient 

Table 2-1. Mean body burden of 137Cs in the United States.a

Year

Body burden 

(Bq)

Body burden 

(nCi)

Annual 

absorbed dose 

(μrad)

1953 10 0.27 40

1954 40 1.08 130

1955 80 2.16 260

1956 160 4.32 570

1957 190 5.13 660

1958 240 6.48 820

1959 300 8.1 1,000

1960 250 6.75 860

1961 170 4.59 590

1962 220 5.94 770

1963 400 10.8 1,400

1964 700 18.9 2,400

1965 590 15.9 2,000

1966 360 9.72 1,200

1967 210 5.67 730

1968 130 3.51 450

1969 100 2.7 360

1970 100 2.7 360

1971 100 2.7 360

1972 100 2.7 360

1973 100 2.7 340

1974 60 1.62 210

1975 40 1.08 140

1976 60 1.62 200

1977 40 1.08 140

a. Source: NCRP Report 94 (NCRP 1987, Table B5).

background levels are assigned. This might not be an accurate representation of the exposure 

pattern, but it best approximates the intakes given the available information and provides for 

consistency across all individuals. Note that if a fitted dose has been calculated for all periods, there 

is no missed dose and it is not calculated.

2.5.2 Missed Dose

Missed dose is the potential dose that could have been received by a bioassay program participant 

but, because of limitations in the monitoring system or a policy decision to use a reporting level, was 

undetected (i.e., activity went undetected because it was less than the detection limit or unreported 

because it was less than the reporting level). Missed dose is assigned using actual worker in vivo or 

in vitro bioassay measurements that are less than the MDA and worker-specific employment 

information. In an operational setting, the application of dosimetric and biokinetic models to the MDA 

of a bioassay technique is referred to as the minimum detectable dose (MDD). The MDD is used for 

selecting bioassay programs to meet given monitoring requirements and is not assigned to workers. 

However, for the EEOICPA program, this dose is included in the dose reconstruction assessment and 

is assigned as the missed dose to the worker. The “true” dose is assumed to fall between 0 and that 

which would result in the prediction of bioassay results equal to the MDA. In many instances, data 

below the MDA were censored and only the MDA value was reported. Therefore, in order to 
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standardize the missed dose calculation, the EEOICPA program imputes this data by using one-half 

the censoring value (e.g., half of the MDA or half of the reporting level). A value of one-half the MDA 

is substituted for the result and used to calculate the missed dose. A triangular distribution is then 

assigned to the dose to account for uncertainty in the bioassay result. The triangular distribution is 

assigned in IREP with a minimum of 0, a maximum based on results at the MDA, and a mode based 

on results at one-half the MDA.

When there are multiple bioassay types available for a given nuclide, the missed dose is calculated 

using one bioassay type followed by a projection to the other type to ensure that the results are not in 

disagreement. This is done by taking the intake calculated from the first bioassay type and using the 

“Intakes to Bioassay” tab in IMBA to calculate the expected excretion or retention on the dates of the 

second bioassay type. The projection is compared to half of the MDA of the second bioassay type. 

That is, the intake calculated from the first bioassay type does not disagree with the second type if the 

projected values are less than or equal to half of the MDA.

2.5.3 Fitted Dose

Fitted dose is the term for doses that are calculated from positive bioassay results (although some 

<MDA results may also be used in the fitting of the intake), which typically confirms that an intake did 

occur. This is entered as a lognormal distribution in IREP. A geometric standard deviation (GSD) of 3 

is assigned to account for the uncertainty in the bioassay measurements and the biokinetic models.

2.5.4 Environmental Intakes

Internal radiation doses to some employees were limited to doses from inhalation of airborne 

radionuclides in the ambient environment from site operations or contamination rather than from 

localized airborne radionuclides from uncontained radioactive materials in the workplace. For these 

employees, assignment of environmental dose only is appropriate. Environmental dose is also 

assigned to monitored individuals if there are nuclides in the environmental suite of radionuclides that 

have not been assigned through some other method in the dose reconstruction. 

Environmental intakes are documented in the site profile. These are assigned as a lognormal 

distribution with a minimum GSD of 3 to account for the uncertainty noted in Section 2.5.3.

2.5.5 Coworker Dose

Coworker exposure distributions are developed from available dosimetric data from DOE or atomic 

weapons employer sites. ORAUT-PLAN-0014, Coworker Data Exposure Profile Development,

(ORAUT 2004b), provides a generic discussion on the development of these datasets. ORAUTOTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment (ORAUT 2005b), 

contains more specific information on the derivation of internal dosimetry data parameters. Additional 

documents specify the statistical techniques to be used in the development of these models. Sitespecific TIBs and site profiles are available for some sites and provide assumptions for intakes based 

on coworker dosimetry analyses.

Typically, workers with a significant potential for intake are assigned doses at the 95th percentile with 

a constant distribution, while those with less potential are assigned the 50th percentile with a 

lognormal distribution. The GSD is determined in part by the fitting of the data, but a minimum of 3 is 

assigned even if the modeling yields a smaller value.
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3.0 INTAKE AND DOSE ASSESSMENT GUIDANCE

3.1 SELECTION OF APPROPRIATE ORGANS FOR DOSE ESTIMATE

Dose must be calculated to the organ in which the cancer originated. ORAUT-OTIB-0005, Internal 

Dosimetry Organ, External Dosimetry Organ, and IREP Model Selection by ICD-9 Code (ORAUT 

2012), correlates the codes in The International Classification of Diseases, 9th Revision, Clinical 

Modification, Sixth Edition (ICD-9; PMIC 2007) with the appropriate organs and tissues to be modeled 

in the Integrated Modules for Bioassay Assessment (IMBA) computer program. For claims where the 

organ of cancer origination is not included in IMBA, the use of “highest nonmetabolic organ” is 

specified. In such situations, the dose to be assigned to the organ is the largest dose among the 

reported organs in IMBA that are not part of the ICRP (1998) metabolic model for the particular 

radionuclide.

In practice, dose reconstructors typically do not need to make this determination. The Chronic Annual 

Dose (CAD) tool, also referred to as Web CAD, selects the highest nonmetabolic organ for the 

selected ICD-9 code when appropriate by calculating the doses to each nonmetabolic organ and 

assigning the largest. The following is primarily for informational purposes.

Organs that do not concentrate a radionuclide receive photon exposure because of their proximity to 

the concentrating organs. The newer ICRP biokinetic models consider exposure from beta and alpha 

radiation to these other organs by defining them as a “soft tissue” compartment and describing uptake 

and clearance rates for this compartment. Many of these nonmetabolic organ doses are calculated 

using these techniques. Because these organs are all considered soft tissue, and are therefore all 

similarly exposed, all of the doses are relatively equal. This implies that choosing the largest of these 

doses is favorable to claimants. However, it is possible for one of the organ doses to be much higher 

than the others due to proximity to a concentrating organ that is emitting photon radiation. In this 

instance, the location of the cancer must be evaluated to ensure the estimate is not unrealistically 

large. If it is, the next largest organ dose should be used. For example, the thymus is relatively close 

to the thyroid so would not be included when determining the highest nonmetabolic organ for an 

intake of 131I.

IMBA can be used to determine which organs are not included in the metabolic model. Dose 

reconstructors should:

1. Select the appropriate element (any isotope works) and load the ICRP defaults using the 

button on the toolbar. 

2. Click Biokinetics (near the Close button). 

3. Select Load ICRP Defaults in the window that pops up. 

Several organ names will be highlighted in blue or purple at the top of the window. These are the 

modeled organs for the element (organs that are specifically named in the model as concentrating the 

radionuclide). Therefore, the remaining organs, with the exceptions noted in Section 2.2, are not part 

of the metabolic model and the largest dose among these organs (i.e., the highest nonmetabolic 

organ) is used for assigning doses to organs that are not included in the IMBA calculations.

3.2 SELECTION OF PARAMETER VALUES

Several parameters must be determined or assumed when assessing an internal dose. Specific 

knowledge of an individual claim or site takes precedence (see Section 1.2). However, when values 
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 16 of 52

 

are unknown, default values should be applied. This section provides guidance on parameter 

selection.

3.2.1 Intake Mode

There are five intake routes included in IMBA:

• Inhalation is the most common route for workplace intakes. 

• Ingestion can be associated with an inhalation intake. When an inhalation intake is assigned

based on air monitoring rather than bioassay data, an additional ingestion component must be 

assigned. See OCAS-TIB-009, Estimation of Ingestion Intakes, for guidance (NIOSH 2004).

• Injection is the entry of material directly into the bloodstream; this is sometimes referred to as 

absorption. Part 1 of ICRP Publication 56 treats the inhalation of tritium oxide as injection 

(ICRP 1989).

• Wound is typically characterized by two or more compartments, with a fraction of the material 

absorbed almost immediately into the bloodstream and additional components with longer 

half-lives from material that remains at the wound site. ORAUT-OTIB-0022, Guidance on 

Wound Modeling for Internal Dose Reconstruction (ORAUT 2005a), discusses modeling these 

intakes. Additional information can be found in National Council on Radiation Protection and 

Measurements (NCRP) Report 156 (NCRP 2006). IMBA contains the default models 

described in the NCRP report. Contact the Principal Internal Dosimetrist (PID) for assistance 

with wound modeling.

• Vapor is a specific instance of an inhalation intake; it is defined as the gaseous form of 

substances that are normally in liquid or solid form. Iodine is typically modeled using this 

intake route. Elemental iodine is then selected as the material type using the ICRP Defs Load 

button in IMBA.

In the absence of information about how an intake might have occurred, inhalation is the default 

assumption when starting with bioassay data because this is the most likely route of entry in an 

occupational setting. 

3.2.2 Particle Size Distribution

The particle size distribution dictates the assumed deposition pattern of inhaled material in the various 

regions of the respiratory tract. For occupational exposures, the ICRP Publication 66 default value is 

a 5-μm activity median aerodynamic diameter (ICRP 1994). This value should be used for evaluating 

inhalation intakes in the absence of known information as documented in the site profiles or the claim 

file.

3.2.3 Material Type

3.2.3.1 Standard Types

Material type describes the rate of absorption of deposited material in the respiratory tract into blood. 

ICRP Publication 66 describes three types: F (fast solubilization), M (moderate solubilization), and S 

(slow solubilization) (ICRP 1994). The assignment of an element to one or more of these categories 

is based on the chemical form of the material. The recommendations of ICRP Publication 68 

(ICRP 1995a) are used for this Project because they address worker intakes rather than those of 

members of the public. These recommendations can be found in IMBA. Figure 3-1 shows an 
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example from the IMBA ICRP Defaults selection menu. If material types at a particular site are 

known, they are documented in the site profile and should be applied under the specifications in that 

document. Attachment C contains additional information about material types.

For the majority of claims, it is likely that the material type is unknown or the individual worked in 

multiple areas, which makes exposure to multiple types possible. In such instances, an assessment 

of each type to which the element is assigned in ICRP Publication 68 (ICRP 1995a) should be made, 

and the type that results in the largest dose to the organ of interest should be selected. See 

Attachment B for a complete list of Publication 68 material types for all elements of interest. The 

exception to the use of Publication 68 types is in the case of 241Am in a plutonium matrix. Americium 

is assigned only to type M by ICRP Publication 68 but evidence has shown that when bound in a type 

S plutonium matrix, americium also exhibits type S behavior. Therefore, ICRP Publication 71 

Figure 3-1. Using IMBA to determine Publication 68 default 

material types.

(ICRP 1995b) type S is selected in IMBA or Web CAD (or other assessment method that might be 

used) for 241Am when in a type S plutonium matrix. 

3.2.3.2 Additional Plutonium Types

When assessing 239Pu, type super S could need to be considered in addition to types M and S. This 

type potentially applies to all methods of assigning intakes including coworker doses and efficiency 

methods. ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly Retained in the Lung

(ORAUT 2010), discusses conditions under which this type applies and appropriate assessment

methods. Associated 241Am is also assessed as type super S. Plutonium-238 also has an alternate 

material type that is applied at specific sites under certain conditions. Details of this model and its 

applicability are documented in DCAS-RPT-005, Alternative Dissolution Models for Insoluble Pu-238

(NIOSH 2016).

3.2.3.3 Intakes of Mixtures

When a mixture of nuclides is bound in a single matrix, the contaminants are assumed to be relatively 

tightly bound but not to behave outside of their own models. Therefore, the same material type 

should be applied to the contaminants as that selected for the host radionuclide. If ICRP Publication 

68 does not assign the contaminant nuclide to the chosen host material type, the closest solubility 

should be selected. Examples of these mixtures include thorium and recycled uranium. Table 3-1 
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provides guidance on material types for several elements that may be included as a recycled uranium 

contaminant. Thorium mixtures are similarly assigned, matching the contaminant radionuclide types 

to the dominant thorium as closely as possible. See Section 3.7.4 for additional discussion on 

mixtures.

3.2.3.4 Ingestion Intakes

When ingestion intakes are assigned along with an inhalation mixture, the selected f1 values are 

matched to the f1 values assigned to the absorption types. For example, Figure 3-1 includes the

plutonium f1 values, while Figure 3-2 displays the uranium values. Table 3-2 contains a summary of 

selected f1 values for a recycled uranium mixture.

Table 3-1. Selected absorption types for 

recycled uranium contaminants. 

Uranium 

type

Contaminant and 

absorption type

F Pu M

Np M

Tc F

Th M

Am M

M Pu M

Np M

Tc M

Th M

Am M

S Pu S

Np M

Tc M

Th S

Am M

Figure 3-2. IMBA window showing uranium default material types and f1 values.
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Table 3-2. Selected f1 values for recycled uranium 

contaminants. 

Uranium 

type

Corresponding 

f1 value

Contaminant and 

f1 value

F, M 0.02 Pu 0.0005

Np 0.0005

Tc 0.8

Th 0.0005

Am 0.0005

S 0.002 Pu 0.00001

Np 0.0005

Tc 0.8

Th 0.0002

Am 0.0005

3.2.4 Radiation Type

IREP requires a Radiation Type to be associated with each entered equivalent dose. The radiation 

effectiveness factor (REF) is analogous to relative biological effectiveness in radiobiology. In IREP, 

the equivalent dose to an organ is converted to absorbed dose and modified with the appropriate REF 

as part of the probability of causation (POC) calculation. The selected REF depends on the type of 

radiation, of which there are six:

• Alpha, 

• <15-keV electrons,

• >15-keV electrons,

• <30-keV photons,

• 30- to 250-keV photons, and

• >250-keV photons.

Note that >250-keV photons and >15-keV electrons are the reference radiations and have a point 

value REF of unity (i.e., the categories are interchangeable). The REFs for the other radiation types 

are expressed as distributions (Kocher, Apostoaei, and Hoffman 2002).

Many radionuclides emit multiple types and energies of radiation that encompass more than one REF 

category. Each radionuclide was assigned to a single REF category associated with the majority of 

energy emitted by the radionuclide. Radiation weighting factors were also considered in categorizing 

the radionuclides. Attachment A contains the values to be used for all nuclides available in IMBA and 

the Project-developed tools. 

Radionuclides that do not emit alpha radiation but have progeny that do are assigned to the alpha 

category because it is favorable to claimants. The single exception is 147Pm, which is assigned to the 

“electrons E >15-keV” radiation category because of the extremely long half-life of the 147Sm progeny.

3.2.5 Normalization of Bioassay Data

The regulations for this Project require the use of ICRP models for calculating internal doses and 

specifies the use of certain publications (42 CFR Part 82). The biokinetic models and subsequent 

dose coefficients in those publications are based on male physiology and anatomy. Therefore, when 

urinalysis results are reported in units other than per day, the results are normalized to 24 hours using 

a conversion factor of 1.4 L/d for all individuals, male or female. For fecal samples, the default mass 

is 135 g/d.
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Note that these values are to be used only for instances where a 24-hour sample has not been 

collected. If a 24-hour sample is indicated but the volume is not 1.4 L (or 135 g for fecal samples), do 

not adjust the result. Results must be normalized before IMBA entry. This applies to all data types 

(see Section 3.2.6.1). 

If a sample volume is reported for a known 24-hour collection period and the results are reported per 

liter, it is appropriate to use the reported sample volume for the adjustment. If a sample volume is 

reported but there is no indication of the sample collection period and the site profile contains no 

additional information, normalize the result when the sample volume is less than those listed above. If 

the sample volume is greater than above, assume a 24-hour collection period and do not normalize 

the values. This is favorable to the claimant because normalization would reduce the reported value.

When determining if a result is positive (i.e., greater than the MDA), like units must be applied. For 

example, if the result is in units of dpm/L it must be compared to an MDA in units of dpm/L. 

3.2.6 IMBA-Specific Parameters and Information

Additional information about IMBA and how it functions can be found in the IMBA documentation 

(under the Help menu in the program, in Documentation/Main documentation), in the IMBA procedure 

(ORAUT 2003b), and on the Project network on the obj3_dr (Q:) drive in the DR Information\Internal 

Dosimetry\IMBA folder. 

3.2.6.1 Data Type

IMBA defines the following data types:

• Real. This type indicates a valid result to be used in the intake calculation. IMBA uses this 

type of result as is. There must be at least one Real result entered for IMBA to perform an 

intake calculation.

• <LOD. This stands for “less than the limit of detection,” and the Project uses this type for 

results less than or equal to the MDA (or reporting level, as appropriate; see Section 2.4.1.3), 

as described in Section 3.4 on fitting positive bioassay data. The sample MDA is entered for 

the Measurement Rate or Measurement Value for results marked as this Data Type. IMBA 

uses a maximum likelihood method for estimating intakes; <LOD results are treated as a 

distribution between zero and the LOD in the intake calculation. These results are displayed in 

brown on the graph.

• Excluded. IMBA ignores all results with this Data Type when calculating the intake. Excluded 

results are displayed in red on the graph.

3.2.6.2 Measurement Error

The inverse square of the Measurement Error is the weighting factor applied to the Measurement 

Rate for the intake fit. The larger the Measurement Error, the smaller the relative weight given to the 

result in the intake calculation. In other words, IMBA attempts to fit the results with the smallest errors 

first, while those with larger errors have less emphasis placed on them. The fit to the data tends to 

move toward the results with the smallest errors on an absolute scale.

IMBA allows the user to generate the measurement error by selecting the type of error relative to the 

measured result (uniform absolute, uniform relative, and square root) and a factor k by which it is 

multiplied. ORAUT-PROC-0002, Use of Integrated Modules for Bioassay Analysis (IMBA) (ORAUT 
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2003b), contains more details on the use of this function. Values entered as <LOD should have an 

error equal to the LOD (this is equivalent to k = 1 if using the Uniform Relative option).

For positive results, use the 1-sigma error associated with the result when available. If there are no 

reported errors with the results, Measurement Error should be calculated using the Uniform Relative 

option with k = 0.3 for all positive (>MDA) results as a starting point. Note that in the case where all 

results are positive (i.e., all are labeled as Real), the value of k is arbitrary; the same intake is 

calculated from any factor, given the same k for all results.

Note also that, when only a single result is used for an intake determination, the value for the 

Measurement Error is not included in the calculation (there is nothing to be weighted) so the intake is 

the same regardless of the value. 

A Uniform Relative error with equal values of k for all results might not be a reasonable estimate if 

there are well-defined peaks or if results vary by more than an order of magnitude. For example, 

given two results with values of 100 and 10, the fit would go closer to the 10 if a Uniform Relative error 

with k = 0.3 was applied because these would be treated as 100 ±30 and 10 ±3. Larger values have 

more precise statistics and might need to be assigned relatively smaller errors to obtain the desired fit 

(e.g., overestimate, underestimate, best estimate). For a best estimate, alternative values for the 

error can be tried if it is not consistent with the results (e.g., the majority of results appear to be 

underpredicted or the larger results are greatly underpredicted). Application of a 10% error to the 

largest results while retaining a 30% error on the smaller positive results might improve the fit. Other 

values can be tried if this does not provide the desired fit. Use of the Uniform Absolute option, with 

the same value entered for all results, yields an unweighted fit (i.e., all results are weighted equally). 

Contact the PID for assistance if a reasonable fit cannot be obtained.

Note that changing the weighting factor (i.e., the Measurement Error) on the results simply moves the 

data fit up and down, it does not change the shape of the curve.

3.2.6.3 Error Distribution

“Normal” is typically selected for the distribution. Bioassay analyses tend to have one primary source 

of error that drives the uncertainty of the analysis. This uncertainty tends to be random about a mean 

value. As such, a normal distribution would be the most appropriate type of error distribution and is 

the assumed distribution type for this program.

3.2.6.4 IMBA Limitations

IMBA is known to calculate inaccurate organ doses for some radionuclide and organ combinations

(Table 3-3). There are two categories of problems: shared versus independent kinetics and very 

short-lived radionuclides. In the case of the former, the current ICRP system models the progeny of a 

radionuclide using the kinetics of the progeny but, because of limitations in the software, IMBA 

employs the older model technique of shared kinetics, where the progeny are assumed to follow the 

behavior of the parent. In the case of very short-lived nuclides, the reason for the problem is unknown 

but IMBA incorrectly calculates the annual doses from chronic intakes.
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Table 3-3. Nuclides for which IMBA calculates 

incorrect organ doses.

Ac-228a Ra-226 Te-133 Th-234

Pb-210 Ra-228 Te-134 U-232

Pr-147a Te-131 Th-228 U-233b

Ra-223 Te-131m Th-229 U-239a

Ra-224 Te-132 Th-232

a. Doses due to chronic intakes only; acute intakes are not 

affected. 

b. Type S only.

For these nuclides, annual dose coefficients have been calculated with the DCAL program (Eckerman 

et al. 2006). These coefficients have been incorporated into Web CAD and verification of the values 

is documented in ORAUT-OTIB-0028, Validation of DCAL Annual Dose Coefficients (ORAUT 2008). 

The IMBA intake calculations are correct, so bioassay results can be input to IMBA for determining 

the intake.

For the radionuclides in Table 3-3, Web CAD must be used to calculate the dose. IMBA cannot be 

used as an over- or underestimate for these nuclides because there is no consistency in the direction 

in which IMBA is incorrect (i.e., it is not always high or low for a given radionuclide).

Contact the PID when a best estimate is needed for time frames that cannot be calculated with Web 

CAD. When sending a claim to the PID for a best estimate, be sure to include:

• Nuclide,

• Intake date(s),

• Material type,

• Intake (acute) or intake rate (chronic) as applicable and including units,

• Organ of interest,

• Date of diagnosis, and

• Claim due date.

3.3 DETERMINATION OF MISSED AND UNMONITORED INTAKE AND DOSE

Although referred to as dose in the following section, much of the discussion is focused on the intake

calculation during the described periods. Doses from those intakes are assigned through the date of 

cancer diagnosis.

3.3.1 Potential for Exposure

The presence of bioassay samples is often an indicator of potential for exposure, but if there are only 

baseline and termination samples (i.e., no other bioassay), they do not necessarily indicate a 

potential. Indicators of potential for internal radiation exposure include:

• Job title,

• Work location, and

• External dose.

Unless it can be shown that the site monitored all those with the potential for an exposure, lack of 

sampling for extended periods is an insufficient reason for assuming a change in exposure potential. 

If the three listed items do not change during an individual’s employment history but there is 

information that indicates a potential for intake at some point (e.g., bioassay data or job title), and no 

information in the site profile to indicate otherwise, a potential for intake must be assumed for the 
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entire employment period. For some claims, the assignment of environmental intakes only is 

appropriate. See Section 4.7 for additional information.

Note: Site-specific information typically takes the form of intakes based on air monitoring or 

source term information and is documented in the site profile.

3.3.2 Missed Dose Determination

A constant, chronic intake is assumed throughout the possible exposure period(s) unless there are 

changes in the MDA as noted below. The date of the last sample result that is less than the MDA in 

the relevant period is used for the calculation. Do not use the date of a positive result and assume 

that no activity was detected. Note that if a fitted dose has been calculated for all periods, there is no 

missed dose and it is not calculated.

If the MDA changes during the intake period, the following must be considered in determining the 

chronic intake (note that “negative” here refers to a result <MDA):

• If the MDA decreases over time and the radionuclide/absorption type reaches equilibrium 

slowly in the compartment of interest (e.g., type M or S plutonium in urine, any type S material 

in the lung), perform the fit using the date of the last negative sample and half of the 

associated MDA and assume a single chronic intake for the entire potential exposure period. 

Only the lowest MDAs need to be considered in this scenario because any assessment of 

early values results in the overestimation of the later, smaller MDA values.

• If the MDA decreases over time for radionuclide/absorption types that reach equilibrium rapidly 

or if the MDA increases over time for any radionuclide/absorption type combination, use IMBA 

to determine chronic intakes applicable to each period. To do this:

– In IMBA, set the number of intakes to the number of periods of different MDAs in which 

the worker has bioassay results (i.e., if the site profile lists a period with a specific MDA 

but the worker has no results during this time, this MDA is not considered in the 

analysis; the intake periods are based on the worker’s sample dates).

– The first chronic intake period begins on the day the exposure began and continues to 

the date of the final sample with a given MDA.

– Each following chronic intake is assessed from the day after the previous period to the 

date of the last sample in the next MDA period, or to the last day of exposure for the 

final exposure period.

– Perform the fit assigning half of the associated MDA to the date of the last sample in 

each period.

• When the MDA oscillates, usually due to samples with individually reported MDAs, selection of 

the sample to use for the missed dose calculation can be claim dependent. For an 

overestimate the use of the largest MDA is appropriate, and conversely, the smallest value 

can be used for an underestimate. A best estimate depends on the pattern of the results, but 

in general a line that runs through the center of the <MDA values would be suitable. Contact 

the PID for assistance if necessary. 

Calculate the annual organ doses from the intakes of each radionuclide and enter the doses into IREP 

using a triangular distribution and:
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1. Set the lower bound (Parameter 1) to 0.

2. Set the mode (Parameter 2) to the annual organ doses as calculated above.

3. Set the upper bound (Parameter 3) to 2 times the mode.

3.3.3 Unmonitored Dose

Unmonitored dose is the potential dose that could have been received by a worker but for which no 

monitoring of the individual was performed or monitoring data are not available. For unmonitored 

periods where there was a potential for intakes, the following priorities are used for assigning dose:

1. Known ratio with other monitored nuclides. For example, the contaminants in recycled 

uranium are not typically monitored directly but can be assessed based on a ratio to the 

calculated uranium intake.

2. Coworker data. Assignment of coworker doses is discussed in Section 3.3.5.

3. Site-specific information. Typically takes the form of default intake values in the site profile.

4. Missed dose. Extension of the missed dose intake rate beyond the last bioassay result. Note 

that this is still based on the actual last bioassay date and result; it does not mean to assume 

there was a later result. By extending this rate, an assumption is made that the intake rate 

was constant throughout the employment period.

3.3.4 Missed Versus Unmonitored Dose

The line between missed and unmonitored dose is not well defined because material from an intake is 

excreted over an extended period depending on the half-life and retention characteristics of the 

nuclide. A long-lived, long-retained nuclide (e.g., plutonium and uranium) can be retained for decades 

with continuous excretion of small amounts. One result after many years of employment can contain 

activity from all previous intakes and provide information for determining an intake amount for all 

previous years and, in such a situation, a lack of bioassay samples for several years would not be 

considered unmonitored because an upper bound can be placed on the intake. This is not true for 

nuclides that are eliminated relatively rapidly from the body (e.g., 137Cs and 3

H). An unmonitored 

period can precede a monitored period for these shorter-retained nuclides.

For both types of nuclides, an individual can be monitored for some period, after which there is an 

unmonitored period. The period after the last bioassay sample is considered unmonitored for both 

long- and short-retained materials.

An individual’s bioassay data always take precedence over other data (e.g., coworker or site default 

values), unless the bioassay has been shown to be flawed or not representative of the individual’s 

exposure.

For long-lived, long-retained radionuclides:

• Missed dose is calculated from the start of the potential intake period through the date of the 

last bioassay sample in the period being assessed. This period is considered to be monitored 

regardless of the date of the first bioassay sample.

• Unmonitored dose is assigned from the day after the last bioassay sample through the end of 

the potential exposure period.
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• Long-lived, long-retained nuclides include all absorption types of plutonium, uranium, and 

americium, unless the only monitoring method is chest counting. Types F and M are not 

retained for significant periods in the lungs and the rules for short-retained radionuclides must 

be followed. 

For short-lived or short-retained radionuclides (including 137Cs, 106Ru, 144Ce, and 90Sr) during potential 

exposure periods: 

• Missed dose is calculated in the intervals where there are bioassay results; other periods are 

considered to be unmonitored. Gaps of greater than 2 years between results are considered 

to be unmonitored.

• Note that for very short-retained materials, notably 3

H and 131I, this period is shorter. Guidance 

on 3

H assessment is contained in ORAUT-OTIB-0011, Tritium Calculated and Missed Dose 

Estimates (ORAUT 2004c). In general, 3

H exposure is assumed only during periods when 

bioassay samples were collected because it is inexpensive, easy, and quick. The primary 

exception to this rule is when a site began using 3

H before the implementation of a bioassay 

program.

• Unmonitored dose is assigned for the period up until 1 year before the first bioassay sample 

for the nuclide of interest.

• Missed dose is calculated from 1 year before the first bioassay result through the date of the 

last bioassay sample.

• If there are more than 2 years between two consecutive samples:

– Missed dose is calculated through the date of the first of these samples.

– Unmonitored dose is assigned from the day after the first sample until 1 year before the 

second sample.

– Missed dose is calculated starting at 1 year before the date of the second result.

3.3.5 Coworker Data

Coworker dose is applied as a best estimate for individuals with a potential for intakes of radioactive 

material but who lack bioassay data or have unmonitored intervals. Data can be lacking because it 

was not available from the site or because monitoring was not performed. Typically, workers with a 

significant potential for intake should be assigned doses at the 95th percentile with a constant 

distribution, while those with less potential are assigned the 50th percentile with a lognormal 

distribution. When coworker analyses do not define how or to whom the intake should apply, 

“significant potential” is subjective, but in general it applies to people who were radiation workers with 

a potential for intakes of radioactive material. The dose reconstructor must make this decision based 

on the worker’s job titles and work locations, as well as other information in the claim file that could 

indicate a potential for intake. ORAUT-OTIB-0014, Assignment of Environmental Internal Doses for 

Employees Not Exposed to Airborne Radionuclides in the Workplace (ORAUT 2004a), provides 

guidance on job categories that are typically most likely to be in the upper end of the distribution. For 

sites that handled multiple independent sources of radionuclides, the site-specific TIBs or site profiles, 

where possible, provide guidance on which nuclides to assign. However, this can be a matter of dose 

reconstructor judgment, again, based on information in the claim file.
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3.3.6 Example Assignment of Coworker and Unmonitored Doses

Employment: 03/1/1957 to 07/12/1989

Job information: Production worker; single work location; no significant fluctuations in external dose 

results. The site profile provided no additional information on potential for 

exposure.

Bioassay: All results <MDA

The MDAs remain constant over the employment period

Plutonium-239 urine samples on 05/4/1960, 12/11/1963, 11/17/1980

Strontium-90 urine samples on 03/12/1965, 09/18/1965, 08/01/1966, 01/05/1978, 

07/12/1978, 04/30/1979

Plutonium-239 dose calculation:

1. Calculate missed dose from 03/1/1957 through 11/17/1980 using 0.5 MDA on 11/17/1980.

2. For the period from 11/18/1980 through 07/12/1989:

a. Assign coworker dose if it is available.

b. If coworker data are not available but the site profile contains default intake values for 

individuals who were potentially exposed, assign the site profile values.

c. If no other information is available, extend the missed dose through this period.

Strontium-90 dose calculation:

1. Assign coworker data or site default values from 03/01/1957 to 03/12/1964.

2. Calculate missed dose from 03/13/1964 to 08/01/1966.

3. Assign coworker data or site default values from 08/02/1966 to 01/05/1977.

4. Calculate missed dose from 01/06/1977 to 04/30/1979.

5. Assign coworker data or site default values from 05/01/1979 to 07/12/1989.

6. If there are other nuclides associated with 90Sr (such as detailed in ORAUT 2015), perform the 

above calculations first and assign the associated radionuclides from the resulting intakes in 

steps 1 through 5.

3.3.7 Short-Duration Missed Dose

The fraction of an element that is absorbed into blood from the small intestine is defined by the 

parameter f1. A very small (i.e., <1 × 10-3) f1 value combined with a long half-life results in small 

fractions of material being excreted via urine. Because of the slow ingrowth to urinary excretion from 

chronic intakes, assessment of missed dose for nuclides with these small f1 values can result in 

implausibly large intake rates when the bioassay result is shortly after the start of intake, particularly 

for those with relatively long half-lives. When calculating a missed dose where the only urine sample 

was collected within 1 year after the start of intake, the coworker intake rate should be assigned as a 

best estimate for nuclides with an f1 value <0.001 and a half-life greater than 50 years.
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The most often encountered nuclides meeting these conditions are all material types of 238Pu and 

239Pu. Note that no uranium material type has an f1 value less than 1 × 10-3. 

3.4 FITTING POSITIVE BIOASSAY RESULTS

This section describes the process for fitting results to known intakes (i.e., positive results). Specific 

details of the mechanics of using IMBA are addressed in ORAUT-PROC-0002, Use of Integrated 

Modules for Bioassay Analysis (IMBA) (ORAUT 2003b).

3.4.1 General Philosophy

The fitting of bioassay data to an intake is a somewhat subjective process, particularly when dealing 

with historical data because intake dates are frequently unknown and additional follow-up sampling is 

not possible. Fits should be as simple as possible; no more complexity than necessary should be 

applied to a given claim. This means if a quick and simple over- or underestimate can be performed 

using the bioassay data (see Section 3.5 for discussion), no further fitting should be tried.

In general, the overall pattern of the data should be fit rather than each individual result. It is not

realistic to develop an intake scenario that yields predicted results that are identical to the measured 

values for all or even most of the measurements because the retention and elimination of radioactive 

materials, as well as the measurement of the material, are stochastic processes that result in 

statistical variations. For urine samples, the concentration varies even within 1 day. In addition, an 

exact match to each measured result is often achieved only through a set of very unrealistic 

assumptions that are often not favorable to claimants. An example of this is fitting each positive result 

to a separate intake. This often requires the assumption that each intake occurred only 1 or 2 days 

before the bioassay sample. If the samples were collected as part of the routine bioassay program 

(as opposed to incident-related samples), it is unlikely that the program caught each unsuspected 

intake immediately upon occurrence. Known intake dates should be used when available.

3.4.2 General Guidelines

Assessment of positive bioassay results is subjective in the absence of known intakes, so the 

following guidelines are provided:

• Use all positive bioassay results, starting with the first positive value.

• For results that are not positive: 

– Enter the MDA value for both the Measurement Result and Measurement Error.

– Include the first negative result after each set of positive results and set Data Type to 

<LOD. 

– If there are several consecutive positive results, include no more than two of the 

immediately following consecutive negative results. For fewer than five consecutive 

positive results, include only the first immediately following negative result. Use of 

additional <LOD results, particularly for chronic exposures, frequently yields a fit that 

appears to underestimate the general trend of the data.

– All other negative results should have a Data Type of Excluded. 

– Note that the presence of a negative result does not mean that a new intake must be 

assigned for the next positive result.
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• Although not necessarily used in fitting, all results (including those that are excluded) should 

be plotted on the graph because they must be consistent with the final fit.

• Fit all of the results simultaneously (i.e., a single IMBA run), even if there are multiple intakes. 

A mix of chronic and acute intakes can be applied, as can a single or multiple chronic intakes. 

A single chronic intake can also be fit when there are only intermittent positive results that are 

relatively small (e.g., within a factor of 2 of the MDA); this could be representative of a lowlevel chronic intake just below the MDA. Note that the limitations on the use of <LOD apply 

here as well.

• For positive results, use the 1-sigma error associated with the result when available. If there 

are no reported errors with the results, see the discussion in Section 3.2.6.2 for guidance.

• If there are several negative results or stretches of employment where all results are negative, 

missed dose is assessed separately. See the discussion in Section 3.3.2. 

• Use known information about intakes where available (e.g., intake date, material type, and 

particle size distribution). For unknown parameters, begin with default values where possible. 

These can be adjusted as necessary, but there must be sufficient justification when doing so.

For an unknown intake date, the default is the midpoint between the date of the positive result 

and that of the previous sample measuring the same radionuclide from the same bioassay 

type. Intake dates should not be varied if there are only a few results for each intake unless 

projections from the intakes are inconsistent with later data (e.g., several negative results are 

predicted to have had detectable levels of activity). As discussed in Section 3.4.1, it is neither 

necessary nor desirable to obtain an exact fit to each result because variation in excretion rate 

is to be expected.

• If the material type is unknown, perform a fit for each possible type. Select the one that yields 

the largest total dose to the applicable organ for the years of interest. Note that the largest 

intake does not necessarily correlate to the largest dose. In some instances, it might be 

possible to rule out a material type based on later, overpredicted sample results or 

disagreement with other measurement types. If one type provides an unarguably better fit, 

use it. This can generally be shown only where:

– A single intake has many (more than 10) consecutive positive results,

– There are contemporary data (later than 1989) associated with intakes 25 years or 

more earlier (depending on the nuclide and its associated half-life), or

– Results from other bioassay methods cannot be reconciled with the larger dose 

determinations (e.g., the intake determined from urine samples predicts detectable 

activity in a chest count but all results are <MDA; in this case, the material type that 

yields this larger intake would be ruled out).

• If the majority of results are positive and scattered throughout the intake period (with no more 

than a few consecutive negative results), use all results for the intake assessment. For those 

few results that are negative, enter the MDA for the value and use Data Type <LOD. Note that 

the issue of measurement error (Section 3.2.6.2) also applies here.
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3.5 OVERALL ASSESSMENT METHODS

There are typically several approaches that can be applied to a given claim. The best approach is 

that which takes the least amount of time while still producing the correct decision. Many claims do 

not require a detailed, accurate dose assessment; efficiency methods can be used to expedite claim

completion with sufficient levels of precision to allow the U.S. Department of Labor (DOL) to arrive at 

correct compensation decisions. There are two general types of expediting methods that can be 

applied: overestimates and underestimates. When neither of these can be applied to a claim, a more 

refined best-estimate assessment is needed.

3.5.1 Overestimate

An overestimate is the assignment of an intake or dose that exceeds the possible exposure of the 

worker. If the resulting POC, including all sources of potential exposure, is less than 45% (note that 

this value is determined by Project Management and NIOSH and is subject to change), further 

refinement is not necessary because it would only lower the assigned dose.

This method is typically appropriate to cancers of nonmetabolic organs because the radioactive 

material does not concentrate in such organs. Therefore, relatively large intakes can yield small 

doses. The method also lends itself to the development of generic values that can be used for many 

individuals. Individual overestimates can also be made using individual-specific information.

3.5.1.1 Generic Overestimates

Several methods have been developed and documented in TIBs. Summaries and general 

applicability of these are described in Section 4.0; specific details are in each document.

Most of the overestimating methods are applicable to individuals with no positive bioassay results. 

However, this can be extended to individuals with positive results as long as the positive results are 

taken into account (i.e., it is shown that the assigned intake yields larger projected values than those 

reported or the positive results are assessed separately and the subsequent dose is added to the 

efficiency method results).

3.5.1.2 Individual-Specific Overestimates

Overestimates can sometimes be applied to individuals with positive bioassay data. In such cases, 

most of the bioassay results should be overpredicted by the selected intake. This can be done by 

running a chronic intake assessment using only the largest bioassay result; all others should be 

plotted but excluded from the fit. If there are several large results, use of the earliest value to perform 

the fit typically yields the largest intake. After calculating the intake, review the measured 

(Measurement Rate) versus predicted (Theoretical Rate) results to determine if most results have 

been overpredicted (this can be done quickly with the graph).

If there are later results that are underpredicted, determine the ratio of the measured result to the 

predicted result, multiply the intake by this ratio, then run the Intakes-to-Bioassay calculation to 

demonstrate that all bioassay results have been overpredicted.

A similar method can also be used when there is an acute intake. Start by using only one result, and 

adjust the intake as necessary to obtain an overestimate of all of the results associated with the 

intake.

The constant distribution is assigned in IREP because this is an upper bound of the dose 

(NIOSH 2002).
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3.5.2 Underestimate

An underestimate is the assignment of a dose to a worker that is less than the dose that would 

potentially be assigned under this program. If the resulting POC is greater than 52% (note that this 

value is determined by Project Management and NIOSH and is subject to change), further refinement 

is not necessary because it would only increase the assigned dose. An underestimate is typically 

performed in the form of a partial assessment of dose, such as reconstruction of a single incident, 

missed dose only, or the underprediction of all or most positive bioassay results. The assigned 

distribution in IREP depends on the type of dose (e.g., missed or fitted) that is calculated.

An underestimate is most likely to be successful when applied to metabolic organs, particularly in 

claims where the detection level for the nuclide is large. This is frequently the case with actinides in 

the earlier decades of the complex. In such instances, a missed dose calculation alone might be 

adequate for determining compensability.

Because this method is dependent on an individual’s bioassay data, the details are claim-specific and 

do not lend themselves to a generic approach that can be documented in a TIB.

3.5.3 Best Estimate

A best estimate is required when an efficiency method results in a decision that is incompatible with 

the assumptions (i.e., an underestimate yields a POC less than 52% and an overestimate yields a 

POC greater than 45%; as noted previously, the specific values are subject to change). The purpose 

of this Project is to provide dose reconstructions with sufficient levels of precision to allow the DOL to 

arrive at correct compensation decisions. A best estimate is based on all available data and is the 

most realistic assessment that can be performed with these data and the requirements of the Energy 

Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA). It can include 

some parameter values that are under- or overestimated if the outcome is consistent with those 

assumptions. That is, the POC is greater than 52% when underestimating parameters have been 

included or less than 45% where overestimates were used. 

When information for a particular parameter value is unknown or there are multiple options, the choice 

that is favorable to claimants (i.e., the one resulting in the largest POC) is selected.

3.5.3.1 Performing a Best Estimate Using Bioassay Data

A best estimate uses all available information. Both missed and fitted dose are included but are 

assessed separately and compared. Fitting them simultaneously results in the inappropriate 

combining of distributions (lognormal and triangular) in a single fit and possible underestimation of 

both the individual fitted and missed doses. The exception to this is when the vast majority of results 

are positive, as discussed in Section 2.5.1 and the final bullet of Section 3.4.2. In such instances, 

there is no missed dose because all of the results and intake periods are taken into account in the 

fitted dose.

1. Calculate the missed dose: 

Ignore positive data (this means that the date of the last result <MDA is used for the missed 

dose calculation) and perform a missed dose (mode only) calculation as described in 

Section 3.3.2. If multiple material types are possible, select the one that yields the largest total 

dose to the applicable organ for the years of interest. If there are no (or very few, as noted 

above) results <MDA, no missed dose is calculated.
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2. Calculate the fitted dose: 

Fit the positive data in accordance with Section 3.4. Unless known intake dates are 

documented or bioassay results are indicated to be special rather than routine, it is not 

considered a best estimate to assign all intake dates at 1 to 2 days before the date of a 

positive bioassay sample. As noted above, this is not a realistic scenario and in most cases is 

not favorable to the claimant. If this is the only way a fit can be obtained, it is likely that an 

inappropriate material type is being applied or too much effort is being made to fit every result 

exactly.

3. Assigned annual dose for a given year is the maximum value from step 1 or 2. Choose the 

IREP annual dose distribution type based on missed or fitted dose assignment:

a. For years in which the dose determined in step 1 is larger than that from step 2, use 

the triangular distribution, where Min = 0, Mode = annual dose, and Max = Mode*2.

b. For years in which the dose from step 2 is equal to or larger than that from step 1, use 

the lognormal distribution, where Median = annual dose and the GSD = 3. 

3.5.3.2 Multiple Cancers

Consistent assumptions must be made for all cancers when performing a best estimate. For 

example, the material type for a given intake of a radionuclide must be the same for all organs. Note 

that this does not preclude the assignment of multiple material types for a given nuclide. This can 

happen when one type is maximizing for a fitted dose while a different type is more favorable to the 

claimant for the missed dose. However, a consistent material type must be used across all organs for 

each fitted dose for a given nuclide and similarly, a consistent material type must be used for the 

missed dose calculation for all organs. 

Selection of the material type is based on that which yields the largest overall probability of causation. 

An exception to this rule is the assessment of doses using ORAUT-OTIB-0054, Fission and Activation 

Product Assignment for Internal Dose-Related Gross Beta and Gross Gamma Analyses (ORAUT 

2015). A number of scenarios comparing the largest and smallest doses to various organs from the 

ORAUT-OTIB-0054 mixture were reviewed. Because the differences were not large and because of 

the massive amounts of computing time required to make material types consistent across organs, it 

was determined to be acceptable to run each organ separately for ORAUT-OTIB-0054 claims with 

multiple cancers, using the maximum material types for each organ. 

3.6 WORKING WITH MULTIPLE BIOASSAY TYPES

3.6.1 Fitting Multiple Bioassay Types

While IMBA allows the simultaneous fitting of multiple bioassay types, this option must be applied with 

extreme caution. Oftentimes different bioassay techniques yield results that are of different orders of 

magnitude, so one of them can completely dominate the fit regardless of how sensitive or accurate 

the relative techniques are. The preferred approach for most claims, particularly those where most or 

all results from one technique are ≤MDA, is to fit one of the bioassay types and project the resulting 

intake to the dates of the second technique to make sure the two are not in disagreement. The most 

appropriate scenario for assessing them simultaneously is the case of a known acute intake with 

primarily positive results for all assessment techniques.
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Multiple bioassay types are commonly seen for sites with plutonium mixtures, where plutonium is 

measured directly in the urine and 241Am is measured in the lungs to determine the plutonium intake. 

In this case, because two different nuclides are measured, they cannot be combined into a single 

assessment. Section 3.7.1.2.1 provides additional information on plutonium assessments using urine 

samples and chest counts. Multiple bioassay methods are also often seen at uranium facilities where, 

in addition to routine urine samples, chest counts are sometimes used to determine if there is longterm buildup in the lungs.

3.6.2 Projections

As described in the previous section, when there are two bioassay types available for a given nuclide, 

the assessment is typically calculated using one bioassay type followed by a projection to the other 

type to ensure that the results are not in disagreement. 

When missed dose is being calculated, the projection should be compared to half of the MDA of the 

second bioassay type. That is, the intake calculated from the first bioassay type does not disagree 

with the second type if the projected values are less than or equal to half of the MDA.

For fitted doses, projections are compared to the full MDA because the fits are based on values 

exceeding the MDA.

3.7 SPECIFIC ISSUES

3.7.1 Plutonium Mixtures

Plutonium-239 is found in various mixtures depending on the purpose of the material. This typically 

includes several plutonium isotopes as well as 241Am from 241Pu decay. A given bioassay technique 

does not necessarily measure all of the components; different methods can be used to measure the 

different nuclides. There are two primary complications in assessing intakes of these mixtures: 

(1) 241Am activity increases over time while the plutonium activities are decreasing, which means that 

the ratios are not constant and makes the age of the material a factor; and (2) an assumption that is 

favorable to the claimant for one technique might not be for another technique. Dose reconstructors 

must therefore be sure to take all information into account.

Note that if the material is from a plutonium heat source, the primary plutonium isotope is 238Pu and 

this discussion is not applicable.

3.7.1.1 Background Information

Plutonium mixtures are characterized by their 240Pu content; they are referred to by its weight 

percentage. Weapons-grade mixtures are 6% by weight 240Pu while fuel-grade plutonium is 12% by 

weight 240Pu.

Americium-241 builds up from near zero at the end of irradiation; however, it is removed during 

separation of the plutonium product and begins to build up again as the 241Pu remaining in the product 

decays. Therefore, the ratio of 241Pu to 239+240Pu decreases from the time of the end of irradiation 

because of decay (241Pu has a half-life of only 14.4 years); whereas the ratio of 241Am to 239+240Pu 

increases from the time of the last separation of the 241Am from the plutonium.

Because 241Am emits a gamma ray that can be detected more readily than the emissions from any of 

the plutonium isotopes, chest counting for its presence is sometimes performed and used as an 

indicator of plutonium intake. Interpretation of the result is complicated by the simultaneous 
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production (from the decay of 241Pu) and decay of 241Am. See Section 3.7.1.3 for guidance on using 

IMBA for this assessment.

3.7.1.2 Assumptions for Use in Dose Reconstruction

As with all dose reconstructions, when there is known information about an intake it should be used 

regardless of default assumptions. In this instance, if the plutonium mixture or age of the material is 

known, it should be used rather than the information below.

3.7.1.2.1 Mixture of Material

If the mixture is unknown and the intake is being calculated from urine sample results, the 

12% mixture may be used as a default assumption that is favorable to claimants.

Because there is less 241Am in the 6% mixture than the 12% mixture, use of a 6% mixture is the 

default starting point for limiting doses based on chest counts.

If both types of data are available, it is necessary to compare them. For likely noncompensable 

claims, it is acceptable to overpredict one of the sets of monitoring results. However, if the claim is 

likely compensable, the selected intake scenario must not contradict any of the worker’s monitoring 

data (urine or chest counting). For example, if the intake is based on urine samples with fuel-grade 

plutonium (12% mixture), run a prediction to the dates of the chest counts based on the intake (using 

the Intakes to Bioassay tab in the IMBA Bioassay Calculations window). If the predicted values are 

greater than the measured values of 241Am in the lung including ingrowth from 241Pu, the intake can be 

used for a likely noncompensable claim but not for a likely compensable claim. In the latter case, the 

intake would then be determined using the 6% mixture assumptions to fit the chest count results. 

Although unlikely, a prediction from this intake to the urine sample results is then necessary to ensure 

that they are not overpredicted.

3.7.1.2.2 Age of Material

Until the fifth year of site operation, assume fresh plutonium. For years 5 through 9, assume a 5-yearold plutonium mixture. After these times, 10-year-old plutonium should be assumed. These 

assumptions can be used for either urinalysis or chest counting.

3.7.1.3 Americium Ingrowth to Plutonium Mixture

As noted above, interpretation of chest count results for 241Am in a plutonium mixture is complicated 

because it is continually produced by the decay of 241Pu. IMBA has a module that allows this to be 

taken into account.

Initial set-up steps:

1. Open new session of IMBA.

2. In the Indicator Radionuclide box select 241Am.

3. In the same box, change the Number of Associated Radionuclides to 1.

4. In the Associated Radionuclides box, select 241Pu.

5. In the same box, set the Abundance for the associated radionuclide.

a. This value is 100 times the ratio of activity of 241Pu to activity of 241Am in the mixture.
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b. For example, for the Savannah River Site (SRS) 10-year-aged 12% mixture, this value is 

4910. 

6. From the toolbar at the top left of the screen, select Advanced, Advanced Dosimetry 

Options. 

a. Select Bioassay tab.

b. Check Allow ingrowth of Am-241 from Pu-241. 

c. Select Pu/Am ratio fixed at start of each intake regime (this should already be 

selected as the default setting).

d. Click OK. 

7. In the Units block:

a. Select appropriate units for Intake.

b. Set the Reference Date in the Specify Time As section.

NOTE: This date must be identical to the specified date for the beginning of the 

intake regime.

c. Click the Date radio button.

8. Select Acute or Chronic as appropriate in the Intake Regimes block.

9. Set the Start Date for IR1 (must be identical to the Reference Date entered in step 7b).

10. Set the End Date for IR1 (if the intake is chronic).

11. Load the ICRP Default f1 value and Absorption Type. For a type S plutonium mixture, you 

must select the ICRP71 type S option for 241Am.

To project 241Am in lung count at any time after intake with a known initial intake of 241Am and 241Pu:

1. Enter the 241Am intake activity on the Main page (IMBA automatically calculates and includes 

the associated 241Pu for this intake per the Associated Radionuclide abundance set above).

2. Click Bioassay Calculations. 

3. Set up a table for lung data and enter the date(s) for the required projected lung counts (green 

area for projections).

4. Use the Intakes to Bioassay tab to calculate the lung count value on the specified dates.

5. Use the projected values to compare to actual lung count data.

To calculate the 241Am and 241Pu intake from lung count data:

1. Click Bioassay Calculations on the Main screen.

2. Set up a table with the appropriate lung count data (blue area for data).
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3. On the Bioassay to Intake tab, check Lungs. 

4. Click Start Calculation. 

5. Go back to the Main screen to see the calculated 241Am intake.

6. Use the 241Am intake value and the activity ratio of 241Pu to 241Am to determine the 241Pu 

intake.

7. Use the activity ratios for 238Pu and 239Pu to 241Am to calculate the 238Pu and 239Pu intake 

values for the mixture. 

3.7.2 Methods for Assessing Isotopic Uranium Results

Note: Isotopic results are from samples that were analyzed using alpha spectroscopy with

individual results reported for 234U, 235U, 238U, and sometimes 236U. ln the case of inductively 

coupled plasma–mass spectrometry (ICP-MS), typically only 235U and 238U are reported.

When a urine sample is analyzed for uranium using a gross alpha technique, the measured activity is 

representative of the total uranium activity because all isotopes are alpha emitters of relatively similar 

energies. When uranium is analyzed isotopically via alpha spectroscopy or ICP-MS, each isotope is 

measured separately and the individual components must all be accounted for in the dose 

assessment because this is a known mixture of material. That is, assessing only one or two of the 

components, even if they are >MDA and the others are not, results in an underestimate because at a 

minimum 234U, 235U, and 238U are present in a mixture; 236U is also present in some recycled uranium 

mixtures. On the other hand, for a best estimate it is not appropriate to assess each component 

individually and sum the results unless all isotopes are >MDA because the isotopes are not present in 

equal concentrations for any mixture. For example, 235U contributes only 2% of the total natural 

uranium activity so assuming it was present at the same activity as 234U and 238U (which would be the 

case if all isotopes were reported as <MDA and the MDA was used for each) would yield an 

overestimate, as noted below in Section 3.7.2.1.2. 

There are several options for assessing an isotopic uranium urine result. The selected method is 

dependent on claim details.

3.7.2.1 Alpha Spectroscopy

The general rule of thumb for all claims is to start with the isotope with the highest activity. For 

depleted uranium this is 238U; for all others it is 234U.

3.7.2.1.1 Underestimate

Assessment of a single isotope can be used as an underestimate for a likely compensable claim. The 

appropriate IREP distribution is dependent on the type of intake that was calculated:

• Triangular for missed dose, and

• Lognormal for fitted dose.

3.7.2.1.2 Overestimate

For an overestimate, the activity from all three isotopes can be summed and assessed as 234U. For 

isotopes <MDA, add the MDA value (all three MDAs can be added for an overestimated missed 

dose).
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Run the intake and dose estimates using the MDA values rather than 0.5 MDA.

Use a constant distribution for the doses in IREP.

3.7.2.1.3 Best Estimate

For a best estimate, all isotopes of uranium must be included in the appropriate proportion for the 

site’s enrichment. Default enrichments are documented in the site profile. Before an assessment can 

be done, each isotopic result needs to be adjusted to a total uranium activity value: 

1. If all isotopes for a given sample are >MDA

a. Sum the individual results.

b. No further adjustments are needed

2. If 234U and 238U >MDA:

a. Sum the 234U and 238U results.

b. Adjust the result to obtain total intake activity by summing the fractions (based on the 

appropriate site values) that 234U and 238U contribute to the total uranium activity and 

dividing the result by this value. 

For example, natural uranium consists of 48.86% 234U, 48.86% 238U, and 2.28% 235U. 

Given 0.5 dpm 234U and 0.5 dpm 238U in a sample with 235U <MDA, the total activity 

would be:

(3-1) ( ) 0.5 0.5 dpm 1.029 dpm 0.486 0.486

+ = +

3. If only a single isotope is >MDA (for alpha spectroscopy results this could be 234U or 238U;), use 

the positive result to determine the total uranium activity. The calculation is the same as the 

example in method 2 above but based on the single positive isotope values.

4. If no isotopes >MDA, this calculation is the same as that in method 3 but is based on a single 

MDA. For depleted uranium, use the 238U MDA and fraction; for all other enrichments use 234U 

values.

Once all results have been adjusted, they are assessed as usual. The total activity is assessed in 

IMBA and Web CAD as 234U, with the exception of depleted uranium, which is assessed as 238U. 

Results that had any component exceeding the MDA are treated as positive values and those with all 

components less than the MDA are treated as <MDA.

3.7.2.2 Inductively Coupled Plasma-Mass Spectrometry

This analytical technique measures the mass of a nuclide rather than activity. This is a relatively new 

technique that is inexpensive and quick for analyzing uranium so has been implemented at some 

uranium facilities in the last 20 or so years. Only 235U and 238U are typically reported because 234U 

contributes a very small mass to all uranium mixtures, which means it is not likely to be detectable 

(this is in contrast to activity, where 234U is most often the largest contributor). However, in all cases, 

including depleted uranium, 234U contributes a significant fraction of the total activity so it must be 
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 37 of 52

 

included in the assessment. This also means that a uranium enrichment assumption is necessary for 

assessing all claims, with the possible exception of an underestimate.

The site profile should include a description of the measurement techniques used at the site. If no 

methods are specified and reported uranium results include only 235U and 238U then it is likely that 

ICP-MS was performed.

Because only two isotopes are reported, the assessment techniques are less complicated than those 

for alpha spectroscopy results and there is no need for simplifying efficiency methods. However, in 

order to assess the dose, conversion to total activity based on uranium enrichment will be needed. 

Once the appropriate enrichment has been determined, convert the result to total uranium activity.

Uranium-238 will account for the majority of the mass except for the case of highly enriched uranium, 

so it is typically used to perform the conversion. Use the 238U result (or MDA, if the result is negative) 

to obtain the total uranium activity. 

Using natural uranium as an example once again, the composition by mass is 0.0054% 234U, 99.27% 

238U, and 0.72% 235U. Given a 238U result of 6 μg, the total activity in the sample is:

(3-2) 6 μg 7 6 6 83 10 μCi/μg 4 13 10 μCi 0 9927 . . .

− − ×× =×

Where 6.83 × 10-7 μCi/μg is the specific activity of natural uranium. 

Once all results are converted to total uranium activity, they are assessed as usual and can be 

combined with other sample results that are also in units of total activity. 

3.7.3 Assignment of Thoron and Radon Dose

For lung cancers, IREP requires 222Rn exposures to be entered in units of working level months 

(WLMs). Radon-220 (also known as thoron) exposures are also frequently recorded in these units, 

but because the decay products have characteristics that are sufficiently different from 222Rn, the 

exposure model is not applicable to thoron. In these circumstances, the reported thoron values must 

be converted to dose. DCAS-TIB-0011, Lung Dose Conversion Factor for Thoron WLM (NIOSH 

2013), contains conversion factors from working level months (WLMs) to dose for respiratory tract 

segments for 220Rn and 219Rn. Because the exposure model applies specifically to the lung for 

radon, it also contains conversion factors for ET1 and ET2 for 222Rn.

3.7.4 Assessment of Mixtures of Radionuclides

When dealing with mixtures of materials, such as those discussed in Section 3.2.3.3, all components 

must be considered and summed for the dose comparison when determining the type that is favorable 

to the claimant.

When several nuclides might have been present at a site, monitoring and reporting was frequently 

limited to the one or two primary dose contributors. Dose from additional radionuclides is assigned 

based on ratios to these primary nuclides. Principal examples of these ratios include weapons-grade 

plutonium, recycled uranium, and mixed fission and activation products. Site profiles contain details 

of the plutonium and uranium mixtures specific to the site, as well as other mixtures that are sitespecific. ORAUT-OTIB-0054 (ORAUT 2015) provides guidance on the assignment of mixed fission 

and activation products.
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 38 of 52

 

For a best estimate, review the nuclides that are included in a ratio method and do not assign any that 

were directly monitored because doing so would account for the same nuclide twice. For example, 

recycled uranium contains plutonium and some sites that handled recycled uranium also handled 

plutonium. A bioassay sample for plutonium would account for any plutonium dose regardless of 

whether it came from a uranium or plutonium mixture.

4.0 ADDITIONAL INTERNAL DOSIMETRY GUIDANCE DOCUMENTS

This section summarizes several documents that provide efficiency methods or additional information 

for making decisions on exposures to assign to a particular worker. This is intended as an overview 

only; refer to the individual documents for the current revision and for applicability and limitations of 

each.

4.1 OCAS-IG-002, INTERNAL DOSE RECONSTRUCTION IMPLEMENTATION GUIDELINE

This document provides implementation guidance on methods and approaches that can be used to 

reconstruct occupational radiation dose from internally deposited radionuclides in support of 

EEOICPA (NIOSH 2002).

4.2 OCAS-TIB-009, ESTIMATION OF INGESTION INTAKES

Guidance for estimating associated ingestion intakes when inhalation intakes are assessed through 

methods other than bioassay (NIOSH 2004). 

4.3 DCAS-TIB-011, LUNG DOSE CONVERSION FACTOR FOR THORON WLM

This TIB provides the conversion factor for calculating lung dose from 220Rn decay products measured 

in working level months (WLM). It also provides conversion factors for 222Rn and 219Rn (NIOSH 2013). 

4.4 DCAS-RPT-005, ALTERNATIVE DISSOLUTION MODELS FOR INSOLUBLE PU-238

This report provides guidance on the evaluation of intakes for workers who were exposed to insoluble 

forms of 238Pu. This guidance applies to Los Alamos National Laboratory, the Mound Site, SRS, and 

the Nuclear Materials and Equipment Corporation (NUMEC) (NIOSH 2016). 

4.5 ORAUT-OTIB-0001, MAXIMUM INTERNAL DOSE ESTIMATES FOR SAVANNAH RIVER 

SITE (SRS) CLAIMS

This method uses the largest recorded intakes in the history of SRS to assign an overestimated intake 

for workers (ORAUT 2003a). An average of the largest five intakes for each monitored nuclide was 

used. Although it is conceivable that an individual had an intake in excess of one of the values, it is 

very unlikely that the worker had large undocumented intakes of all radionuclides present on the site. 

Tritium is not included; if the worker had the potential for significant or chronic intakes of tritium, 

additional dose must be included.

4.6 ORAUT-OTIB-0011, TRITIUM CALCULATED AND MISSED DOSE ESTIMATES 

IMBA does not directly calculate intakes from 3

H urine data, so an alternative tool was developed. 

This TIB provides documentation of the method for estimating tritium missed and calculated doses 

from urine data (ORAUT 2004c).
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 39 of 52

 

4.7 ORAUT-OTIB-0014, ASSIGNMENT OF ENVIRONMENTAL INTERNAL DOSES FOR 

EMPLOYEES NOT EXPOSED TO AIRBORNE RADIONUCLIDES IN THE WORKPLACE

Internal radiation doses to some employees were limited to doses from inhalation of airborne 

radionuclides in the ambient environment from site operations or contamination rather than from 

localized airborne radionuclides from uncontained radioactive materials in the workplace. For these 

employees, assignment of environmental dose only is appropriate. ORAUT-OTIB-0014 (ORAUT 

2004b) provides guidance for determining such instances based on:

• Job description,

• Work location,

• Time frame, and

• Presence or absence of internal monitoring data.

This document also provides information on job categories that can be helpful in determining 

“significant potential” when assigning coworker intakes. 

4.8 ORAUT-OTIB-0018, INTERNAL DOSE OVERESTIMATES FOR FACILITIES WITH AIR 

SAMPLING PROGRAMS

This method is based on limiting air concentrations (ORAUT 2005c). A chronic exposure to the 

maximum permissible concentration (MPC) throughout employment is assumed. Because of this, the 

method applies only to sites that controlled exposure to intakes based on rigorous air sampling 

programs. While it is possible for a worker to have been occasionally exposed to levels exceeding the 

MPC, it is very unlikely that an individual was continuously exposed at such levels for 40 hr/wk 

throughout the employment period. An additional conservatism is achieved by assuming that the 

airborne activity consisted of the single nuclide, in each year of intake, that results in the largest dose 

to the organ of interest rather than assigning a mixture of radionuclides.

Application of this overestimating technique is limited to specific sites and organs; refer to the TIB and 

the tool documentation for applicability.

4.9 ORAUT-OTIB-0022, GUIDANCE ON WOUND MODELING FOR INTERNAL DOSE 

RECONSTRUCTION

This document focuses on how to use IMBA to evaluate intakes of plutonium by wound, although the 

concept can be applied to other radionuclides (ORAUT 2005a). Guidance on initial parameters is 

provided.

4.10 ORAUT-OTIB-0033, APPLICATION OF INTERNAL DOSES BASED ON CLAIMANTFAVORABLE ASSUMPTIONS FOR PROCESSING AS BEST ESTIMATES

This method (ORAUT 2005d) applies a graded approach to internal dose overestimates and unites 

the application of ORAUT-OTIB-0014 (ORAUT 2004a), ORAUT-OTIB-0018 (ORAUT 2005c), and 

coworker dose. Some judgment is needed to apply these values; guidance is provided based on:

• The period of employment,

• The processes at the site,

• The job category and work location, and

• The results of bioassay measurements for the energy employee.
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 40 of 52

 

4.11 ORAUT-OTIB-0049, ESTIMATING DOSES FOR PLUTONIUM STRONGLY RETAINED IN 

THE LUNG

A handful of accidental intakes of plutonium oxides have exhibited long-term retention of plutonium in 

the lung exceeding that predicted by the standard type S model. This TIB provides assessment 

methods for calculating a best estimate of the annual organ doses for intakes of type super S 

plutonium and describes the conditions for applicability of this method (ORAUT 2010).

4.12 ORAUT-OTIB-0054, FISSION AND ACTIVATION PRODUCT ASSIGNMENT FOR 

INTERNAL DOSE-RELATED GROSS BETA AND GROSS GAMMA ANALYSES

Reactor operations can produce a wide assortment of radionuclides, but routine bioassay monitoring 

typically includes only those that predominate the mixture or are reported as a gross measurement 

that could contain an assortment of nuclides. This TIB provides guidance on the assignment of 

radionuclide-specific intakes of mixed fission and activation products when air sampling or urinalysis 

data associated with reactors or reactor fuels are available only as gross or total beta activity or gross 

or total gamma activity (ORAUT 2015). The derived ratios can also be applied to claims where the 

activity of a single radionuclide in the mixture is known, such as 137Cs from a whole-body count result 

or 90Sr in a urine sample.

4.13 ORAUT-OTIB-0066, CALCULATION OF DOSE FROM INTAKES OF SPECIAL TRITIUM 

COMPOUNDS

This document provides guidance on how to use urine bioassay data to calculate best estimates of 

doses for intakes of organically bound tritium and stable metal tritides.

5.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 41 of 52

 

REFERENCES

Eckerman, K. F., R. W. Leggett, M. Cristy, C. B. Nelson, J. C. Ryman, A. L. Sjoreen, and R. C. Ward, 

2006, User’s Guide to the DCAL System, ORNL/TM-2001/190, UT-Battelle, Oak Ridge 

National Laboratory, Oak Ridge, Tennessee, August. [SRDB Ref ID: 43944]

HPS (Health Physics Society), 2011, An American National Standard – Performance Criteria for 

Radiobioassay, ANSI/HPS N13.30-2011, Health Physics Society, McLean, Virginia, 

December 16. [SRDB Ref ID: 168975] 

ICRP (International Commission on Radiological Protection), 1979, Limits for the Intake of 

Radionuclides by Workers, Part 1, Publication 30, Part 1, Pergamon Press, Oxford, England. 

[SRDB Ref ID: 22733]

ICRP (International Commission on Radiological Protection), 1989, Age-dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 1, Publication 56, Pergamon Press, 

Oxford, England. [SRDB Ref ID: 135823]

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22732]

ICRP (International Commission on Radiological Protection), 1995a, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22731]

ICRP (International Commission on Radiological Protection), 1995b, Age-dependent Doses to 

Members of the Public from Intakes of Radionuclides: Part 4 Inhalation Dose Coefficients, 

Publication 71, Pergamon Press, Oxford, England. [SRDB Ref ID: 22729]

ICRP (International Commission on Radiological Protection), 1998, Individual Monitoring for Internal 

Exposure of Workers, Replacement of ICRP Publication 54, Publication 78, Pergamon Press, 

Oxford, England. [SRDB Ref ID: 132761]

Kocher, D.C., A. I. Apostoaei, and F. O. Hoffman, 2002, Radiation Effectiveness Factors (REFs) for 

Use in Calculating Probability of Causation in Radiogenic Cancers, SENES Oak Ridge, Center 

for Risk Analysis, Oak Ridge, Tennessee, June 17. [SRDB Ref ID: 34278]

NCRP (National Council on Radiation Protection and Measurements), 1987, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Bethesda, Maryland, December 30. [SRDB Ref ID: 22736]

NCRP (National Council on Radiation Protection and Measurements), 2006, Development of a 

Biokinetic Model for Radionuclide-Contaminated Wounds for Their Assessment, Dosimetry 

and Treatment, Report 156, Bethesda, Maryland, December 14. [SRDB Ref ID: 44742]

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August. [SRDB Ref ID: 22402]

NIOSH (National Institute for Occupational Safety and Health), 2004, Estimation of Ingestion Intakes, 

OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

April 13. [SRDB Ref ID: 22397]
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 42 of 52

 

NIOSH (National Institute for Occupational Safety and Health), 2013, Lung Dose Conversion Factor 

for Thoron WLM, DCAS-TIB-011, Rev. 4, Office of Compensation Analysis and Support, 

Cincinnati, Ohio, May 1. [SRDB Ref ID: 126391] 

NIOSH (National Institute for Occupational Safety and Health), 2016, Alternative Dissolution Models 

for Insoluble Pu-238, DCAS-RPT-005 Rev. 00, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, June 30. [SRDB Ref ID: 161248]

ORAUT (Oak Ridge Associated Universities Team), 2003a, Technical Information Bulletin: Maximum 

Internal Dose for Savannah River Site (SRS) Claims, ORAUT-OTIB-0001, Rev. 00, 

Oak Ridge, Tennessee, July 15. [SRDB Ref ID: 19407] 

ORAUT (Oak Ridge Associated Universities Team), 2003b, Use of Integrated Modules for Bioassay 

Analysis (IMBA), ORAUT-PROC-0002, Rev. 01, Oak Ridge, Tennessee, August 14. [SRDB 

Ref ID: 168571] 

ORAUT (Oak Ridge Associated Universities Team), 2004a, Assignment of Environmental Internal 

Doses for Employees Not Exposed to Airborne Radionuclides in the Workplace, ORAUTOTIB-0014, Rev. 00, Oak Ridge, Tennessee, June 22. [SRDB Ref ID: 19432] 

ORAUT (Oak Ridge Associated Universities Team), 2004b, Coworker Data Exposure Profile 

Development, ORAUT-PLAN-0014, Rev. 00, Oak Ridge, Tennessee, November 24. [SRDB 

Ref ID: 166559] 

ORAUT (Oak Ridge Associated Universities Team), 2004c, Tritium Calculated and Missed Dose 

Estimates, ORAUT-OTIB-0011, Rev. 00, Oak Ridge, Tennessee, June 29. [SRDB Ref ID: 

19430] 

ORAUT (Oak Ridge Associated Universities Team), 2005a, Guidance on Wound Modeling for Internal 

Dose Reconstruction, ORAUT-OTIB-0022, Rev. 00, Oak Ridge, Tennessee, November 18. 

[SRDB Ref ID: 19443] 

ORAUT (Oak Ridge Associated Universities Team), 2005b, Analysis of Coworker Bioassay Data for 

Internal Dose Assignment, ORAUT-OTIB-0019, Rev. 01, Oak Ridge, Tennessee, October 7. 

[SRDB Ref ID: 19438] 

ORAUT (Oak Ridge Associated Universities Team), 2005c, Internal Dose Overestimates for Facilities 

With Air Sampling Programs, ORAUT-OTIB-0018, Rev. 01, Oak Ridge, Tennessee, August 9. 

[SRDB Ref ID: 19436] 

ORAUT (Oak Ridge Associated Universities Team), 2005d, Application of Internal Doses Based on 

Claimant-Favorable Assumptions for Processing as Best Estimates, ORAUT-OTIB-0033, 

Rev. 00, Oak Ridge, Tennessee, April 20. [SRDB Ref ID: 19457] 

ORAUT (Oak Ridge Associated Universities Team), 2008, Validation of DCAL Annual Dose 

Coefficients, ORAUT-OTIB-0028, Rev. 02, Oak Ridge, Tennessee, July 28. [SRDB Ref ID: 

46673] 

ORAUT (Oak Ridge Associated Universities Team), 2010, Estimating Doses for Plutonium Strongly 

Retained in the Lung, ORAUT-OTIB-0049, Rev. 01 PC-2, Oak Ridge, Tennessee, 

November 29. [SRDB Ref ID: 90666] 
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ORAUT (Oak Ridge Associated Universities Team), 2012, Internal Dosimetry Organ, External 

Dosimetry Organ, and IREP Model Selection by ICD-9 Code, ORAUT-OTIB-0005, Rev. 05, 

Oak Ridge, Tennessee, December 20. [SRDB Ref ID: 121336] 

ORAUT (Oak Ridge Associated Universities Team), 2015, Fission and Activation Product Assignment 

for Internal Dose-Related Gross Beta and Gross Gamma Analyses, ORAUT-OTIB-0054, 

Rev. 04, Oak Ridge, Tennessee, August 27. [SRDB Ref ID: 146884] 

PMIC (Practice Management Information Corporation), 2007, International Classification of Diseases, 

9th Revision, Clinical Modification, Sixth Edition (Color Coded 2008), ICD-9-CM Sixth Edition, 

Los Angeles, California. [SRDB Ref ID: 159208]
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 44 of 52

ATTACHMENT A 

RADIATION TYPES BY NUCLIDE FOR ENTRY INTO IREP

Table A-1. Radiation types by nuclide for entry into IREP.

Nuclide Radiation typea

Ac-227 Alpha

Ac-228 Alpha

Ag-110m Photons E >250 keV

Am-241 Alpha

Am-243 Alpha

As-74 Photons E >250 keV

As-76 Electrons E >15 keV

Au-194 Photons E <30 keV

Ba-133 Photons E >250 keV

Ba-140 Photons E >250 keV

Bk-249 Alpha

C-14 Electrons E >15 keV

Ca-45 Electrons E >15 keV

Ce-139 Photons E = 30–250 keV

Ce-141 Electrons E >15 keV

Ce-143 Electrons E >15 keV

Ce-144 Electrons E >15 keV

Cf-249 Alpha

Cf-252 Alpha

Cl-36 Electrons E >15 keV

Cm-242 Alpha

Cm-243 Alpha

Cm-244 Alpha

Co-57 Photons E = 30–250 keV

Co-58 Photons E >250 keV

Co-60 Photons E >250keV

Cr-51 Photons E >250 keV

Cs-134 Electrons E >15 keV

Cs-137 Electrons E >15 keV

Eu-152 Photons E >250 keV

Eu-154 Electrons E >15 keV

Eu-155 Electrons E >15 keV

Eu-156 Electrons E >15 keV

Fe-55 Photons E <30 keV

Fe-59 Electrons E >15 keV

H-3 Electrons E <15 keV

Hf-181 Electrons E >15 keV

I-125 Photons E <30 keV

I-129 Photons E = 30–250 keV

I-131 Electrons E >15 keV

Nuclide Radiation typea

I-133 Electrons E >15 keV

I-134 Photons E >250 keV

I-135 Photons E >250 keV

Ir-192 Photons E >250 keV

La-140 Photons E >250 keV

Lu-174 Photons E = 30–250 keV

Mn-54 Photons E >250 keV

Mn-56 Electrons E >15 keV

Mo-99 Electrons E >15 keV

Na-22 Photons E >250 keV

Na-24 Photons E >250 keV

Nb-94 Photons E >250 keV

Nb-95 Electrons E >15 keV

Ni-63 Electrons E >15 keV

Np-237 Alpha

Np-239 Alpha

P-32 Electrons E >15 keV

P-33 Electrons E >15 keV

Pa-231 Alpha

Pa-233 Alpha

Pa-234 Alpha

Pb-210 Alpha

Pm-147 Electrons E >15 keV

Po-208 Alpha

Po-209 Alpha

Po-210 Alpha

Pr-143 Photons E >250 keV

Pr-147 Electrons E >15 keV

Pu-236 Alpha

Pu-238 Alpha

Pu-239 Alpha

Pu-240 Alpha

Pu-241 Alpha

Pu-242 Alpha

Ra-220 Alpha

Ra-223 Alpha

Ra-224 Alpha

Ra-226 Alpha

Ra-228 Alpha

Rn-220 Alpha

Nuclide Radiation typea

Rn-222 Radon

Ru-103 Electrons E >15 keV

Ru-106 Electrons E >15 keV

S-35 Electrons E >15 keV

Sb-124 Photons E >250 keV

Sb-125 Photons E >250 keV

Sc-46 Photons E >250 keV

Sm-151 Electrons E >15 keV

Sn-113 Electrons E >15 keV

Sr-85 Photons E >250 keV

Sr-89 Electrons E >15 keV

Sr-90 Electrons E >15 keV

Sr-91 Electrons E >15 keV

Ta-182 Photons E >250 keV

Tb-160 Electrons E >15 keV

Tc-99 Electrons E >15 keV

Te-131 Photons E >250 keV

Te-131m Electrons E >15 keV

Th-228 Alpha

Th-230 Alpha

Th-232 Alpha

Th-234 Alpha

Tl-201 Photons E = 30–250keV

Tl-202 Photons E >250 keV

Tl-204 Electrons E >15 keV

Tm-170 Electrons E >15 keV

U-232 Alpha

U-234 Alpha

U-235 Alpha

U-236 Alpha

U-238 Alpha

U-239 Alpha

Y-88 Photons E >250 keV

Y-90 Electrons E >15 keV

Y-91 Electrons E >15 keV

Yb-169 Photons E >250 keV

Zn-65 Photons E >250 keV

Zr-95 Electrons E >15 keV

a. E = energy; the radiation effectiveness factors (REFs) for >15-keV electrons and >250-keV photons are equal. To

minimize the number of lines in IREP, these two categories can be entered with either energy type and combined if the

same distribution type.
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 45 of 52

ATTACHMENT B 

ICRP PUBLICATION 68 DEFAULT MATERIAL TYPES

When modeling inhalation intakes, the rate at which material leaves the respiratory tract and is 

transported to the systemic organs is of importance. The ICRP (ICRP 1994) uses material type (also 

called absorption type) to describe this rate. These designations apply only to material leaving the 

lung; they provide no indication of the total amount of time the material clears from the rest of the 

body. These types and their approximate half-times in the lung are:

• V (vapor). 100% is absorbed instantaneously.

• F (fast). 100% is absorbed from the lung with a half-time of 10 minutes.

• M (moderate). 10% at 10 minutes and 90% at 140 days.

• S (slow). 0.1% at 10 minutes, 99.9% at 7,000 days.

Table B-1 lists the ICRP default material types for all elements available for assessment in IMBA.

Table B-1. Publication 68 default material typesa

Atomic 

number Element Symbol

Material 

types

1 Hydrogen H V

4 Beryllium Be M, S

6 Carbon C V

11 Sodium Na F

12 Magnesium Mg F, M

13 Aluminum Al F, M

14 Silicon Si F, M, S

15 Phosphorus P F, M

16 Sulphur S F, M, V

17 Chlorine Cl F, M

19 Potassium K F

20 Calcium Ca M

21 Scandium Sc S

22 Titanium Ti F, M, S

23 Vanadium V F, M

24 Chromium Cr F, M, S

25 Manganese Mn F, M

26 Iron Fe F, M

27 Cobalt Co M, S

28 Nickel Ni F, M

29 Copper Cu F, M, S

30 Zinc Zn S

31 Gallium Ga F, M

32 Germanium Ge F, M

33 Arsenic As M

34 Selenium Se F, M

35 Bromine Br F, M

37 Rubidium Rb F

38 Strontium Sr F, S

39 Yttrium Y M, S

40 Zirconium Zr F, M, S

41 Niobium Nb M, S

42 Molybdenum Mo F, S

43 Technetium Tc F, M

44 Ruthenium Ru F, M, S

45 Rhodium Rh F, M, S
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 46 of 52

ATTACHMENT B

ICRP PUBLICATION 68 DEFAULT MATERIAL TYPES (continued)

Atomic 

number Element Symbol

Material 

types

46 Palladium Pd F, M, S

47 Silver Ag F, M, S

48 Cadmium Cd F, M, S

49 Indium In F, M

50 Tin Sn F, M

51 Antimony Sb F, M

52 Tellurium Te F, M

53 Iodine I F, V

55 Cesium Cs F

56 Barium Ba F

57 Lanthanum La F, M

58 Cerium Ce M, S

59 Praseodymium Pr M, S

60 Neodymium Nd M, S

61 Promethium Pm M, S

62 Samarium Sm M

63 Europium Eu M

64 Gadolinium Gd F, M

65 Terbium Tb M

66 Dysprosium Dy M

67 Holmium Ho M

68 Erbium Er M

69 Thulium Tm M

70 Ytterbium Yb M, S

71 Lutetium Lu M, S

72 Hafnium Hf F, M

73 Tantalum Ta M, S

74 Tungsten W F

75 Rhenium Re F, M

76 Osmium Os F, M, S

77 Iridium Ir F, M, S

78 Platinum Pt F

79 Gold Au F, M, S

80 Mercury Hg F, V

81 Thallium Tl F

82 Lead Pb F

83 Bismuth Bi F, M

84 Polonium Po F, M

85 Astatine At F, M

87 Francium Fr F

88 Radium Ra M

89 Actinium Ac F, M, S

90 Thorium Th M, S

91 Protactinium Pa M, S

92 Uranium U F, M, S

93 Neptunium Np M

94 Plutonium Pu M, S

95 Americium Am M

96 Curium Cm M

97 Berkelium Bk M

98 Californium Cf M
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 47 of 52

ATTACHMENT B

ICRP PUBLICATION 68 DEFAULT MATERIAL TYPES (continued)

Atomic 

number Element Symbol

Material 

types

99 Einsteinium Es M

100 Fermium Fm M

101 Mendelevium Md M

a. Source: ICRP (1995a).
Document No. ORAUT-OTIB-0060 Revision No. 02 Effective Date: 04/20/2018 Page 48 of 52

ATTACHMENT C 

RETENTION OF MATERIAL IN THE LUNGS

LIST OF FIGURES

FIGURE TITLE PAGE

C-1 Lung retention following an acute inhalation intake of 1 Bq of type F material........................ 49

C-2 Lung retention following an acute inhalation intake of 1 Bq of type M material ....................... 50

C-3 Lung retention following an acute inhalation intake of 1 Bq of type S material........................ 50

C-4 Lung retention of a 365-d chronic inhalation intake of a type F material ................................. 50

C-5 Lung retention of a 365-d chronic inhalation intake of a type M material ................................ 51

C-6 Lung retention of a 365-d chronic inhalation intake of a type S material................................. 51

C-7 Lung retention of a 1,000-d chronic inhalation intake of a type M material ............................. 51

C-8 Lung retention of a 1,000-d chronic inhalation intake of a type S material.............................. 52
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ATTACHMENT C 

RETENTION OF MATERIAL IN THE LUNGS (continued)

As noted in Section 3.2.3, material type, also referred to as absorption type, describes the rate of 

absorption of deposited material in the respiratory tract into blood. Once in the blood the material is 

subsequently transferred to other organs and tissues, so it is not an indication of the total amount of 

time that the material is retained in or eliminated from the total body. 

The very rapid clearance of type F material following an acute inhalation intake is shown in 

Figure C-1. The material is transferred out of the lungs in less than a tenth of a day. Figures C-2 and 

C-3 show the clearance of types M and S, respectively, following an acute inhalation intake. The 

same scale is used for both for comparison. 

Figures C-4 through C-6 display the retention and clearance of the three types during and following a 

365-day chronic inhalation intake of 1 Bq/d. The type F material rapidly reaches equilibrium in the 

lungs, although a very small fraction of the intake is retained even during the intake period. The 

material is equally rapidly removed from the lungs once the intake stops. Types M and S both 

continue to increase in the lungs over a 365-day chronic intake, with the type M activity reaching 

about half that of the type S activity. Following the cessation of intake, type M is removed more 

rapidly but not immediately. 

Chest counting is most often used for detecting the long-term buildup of type S material in the lungs. 

Going to a longer intake of 1000 days, it can be seen in Figure C-7 that the type M material reaches 

an equilibrium state in the lungs shortly after about a year of chronic intake while the activity of type S, 

plotted in Figure C-8, continues to increase throughout the period. 

These plots demonstrate that chest counting cannot be used for assessing a type F material because 

at any time during or following an intake little to no activity is expected to be present in the lungs. 

Chest counting can be used to detect either type M or S in the lungs but at any given time urine 

sampling may be more sensitive (i.e., able to detect smaller intakes) depending on the time since 

intake and the relative sensitivities of the two bioassay methods.

Figure C-1. Lung retention following an acute inhalation intake of 1 Bq of type F material.
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ATTACHMENT C 

RETENTION OF MATERIAL IN THE LUNGS (continued)

Figure C-2. Lung retention following an acute inhalation intake of 1 Bq of type M material.

Figure C-3. Lung retention following an acute inhalation intake of 1 Bq of type S material.

Figure C-4. Lung retention of a 365-d chronic inhalation intake of a type F material.
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ATTACHMENT C 

RETENTION OF MATERIAL IN THE LUNGS (continued)

Figure C-5. Lung retention of a 365-d chronic inhalation intake of a type M material.

Figure C-6. Lung retention of a 365-d chronic inhalation intake of a type S material. 

Figure C-7. Lung retention of a 1,000-d chronic inhalation intake of a type M material.
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ATTACHMENT C 

RETENTION OF MATERIAL IN THE LUNGS (continued)

Figure C-8. Lung retention of a 1,000-d chronic inhalation intake of a type S material.