ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 34

DOE Review Release 09/12/2014

Internal Dose Reconstruction

ORAUT-OTIB-0060 Rev. 01

Effective Date: 09/08/2014

Supersedes: Revision 00

Subject Expert(s): Elizabeth M. Brackett

Document Owner

Approval: Signature on File Approval Date: 08/13/2014

James P. Griffin, Document Owner

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 08/12/2014

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 09/08/2014

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 2 of 34

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

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. 
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 3 of 34

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 5

1.0 Introduction .............................................................................................................................. 7

1.1 Purpose........................................................................................................................ 7

1.2 Scope, Applicability, and Limitations............................................................................. 7

2.0 Terminology ............................................................................................................................. 7

2.1 Systemic Versus Nonsystemic...................................................................................... 7

2.2 Metabolic Versus Nonmetabolic Organs....................................................................... 8

2.3 Positive and Negative Results ...................................................................................... 8

3.0 Guidance ................................................................................................................................. 9

3.1 Selection of Appropriate Organs for Dose Estimate...................................................... 9

3.2 Selection of Parameter Values ................................................................................... 10

3.2.1 Intake Mode ................................................................................................. 10

3.2.2 Particle Size Distribution .............................................................................. 10

3.2.3 Material Type ............................................................................................... 10

3.2.4 Radiation Type............................................................................................. 11

3.2.5 Normalization of Bioassay Data ................................................................... 12

3.2.6 IMBA-Specific Parameters and Information.................................................. 12

3.2.6.1 Data Type.................................................................................. 13

3.2.6.2 Measurement Error ................................................................... 13

3.2.6.3 Error Distribution ....................................................................... 13

3.2.6.4 IMBA Limitations ....................................................................... 14

3.3 Assignment of Missed And Unmonitored Dose........................................................... 15

3.3.1 Missed Dose Determination ......................................................................... 17

3.3.2 Coworker Data ............................................................................................. 18

3.3.3 Example Assignment of Coworker and Unmonitored Doses ........................ 18

3.3.4 Short-Duration Missed Dose ........................................................................ 19

3.4 Fitting Positive Bioassay Results ................................................................................ 19

3.4.1 General Philosophy...................................................................................... 19

3.4.2 General Guidelines ...................................................................................... 20

3.5 Overall Assessment Methods ..................................................................................... 21

3.5.1 Overestimate................................................................................................ 21

3.5.1.1 Generic Overestimates.............................................................. 22

3.5.1.2 Individual-Specific Overestimates.............................................. 22

3.5.2 Underestimate.............................................................................................. 22

3.5.3 Best Estimate............................................................................................... 22

3.5.3.1 Performing a Best Estimate Using Bioassay Data ..................... 23

3.5.3.2 Multiple Cancers........................................................................ 23

3.6 Internal Dosimetry Guidance Documents ................................................................... 23

3.6.1 OCAS-IG-002, Internal Dose Reconstruction Implementation 

Guideline...................................................................................................... 24

3.6.2 ORAUT-OTIB-0001, Maximum Internal Dose Estimates for Savannah 

River Site (SRS) Claims............................................................................... 24

3.6.3 ORAUT-OTIB-0002, Maximum Internal Dose Estimates for Certain 

DOE Complex Claims .................................................................................. 24

3.6.4 ORAUT-OTIB-0011, Tritium Calculated and Missed Dose Estimates........... 24
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3.6.5 ORAUT-OTIB-0014, Assignment of Environmental Internal Doses for 

Employees Not Exposed to Airborne Radionuclides in the Workplace ......... 24

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

Sampling Programs...................................................................................... 24

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

Reconstruction ............................................................................................. 25

3.6.8 ORAUT-OTIB-0033, Application of Internal Doses Based on ClaimantFavorable Assumptions for Processing as Best Estimates ........................... 25

3.6.9 ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly 

Retained in the Lung.................................................................................... 25

3.6.10 ORAUT-OTIB-0054, Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross Gamma Analyses.................. 25

3.6.11 ORAUT-OTIB-0066, Calculation of Dose from Intakes of Special 

Tritium Compounds...................................................................................... 25

3.7 Specific Issues ........................................................................................................... 25

3.7.1 Plutonium Mixtures....................................................................................... 25

3.7.1.1 Background Information ............................................................ 26

3.7.1.2 Assumptions for Use in Dose Reconstruction............................ 26

3.7.1.2.1 Mixture of Material ............................................... 26

3.7.1.2.2 Age of Material .................................................... 26

3.7.2 Methods for Assessing Isotopic Uranium Results......................................... 27

3.7.2.1 Underestimate........................................................................... 27

3.7.2.2 Overestimate............................................................................. 27

3.7.2.3 Best Estimate............................................................................ 27

3.7.3 Assignment of Thoron and Radon Dose....................................................... 28

3.7.4 Assessment of Mixtures of Radionuclides .................................................... 28

4.0 Attributions and Annotations .................................................................................................. 29

References ........................................................................................................................................ 30

ATTACHMENT A, RADIATION TYPES BY NUCLIDE FOR ENTRY INTO IREP .............................. 32

ATTACHMENT B, IMBA DOSE ASSESSMENT LIMITATIONS ........................................................ 33

LIST OF TABLES

TABLE TITLE PAGE

3-1 Selected material types for recycled uranium contaminants ................................................... 11

A-1 Radiation types by nuclide for entry into IREP........................................................................ 32

B-1 IMBA doses relative to the true value..................................................................................... 33
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ACRONYMS AND ABBREVIATIONS

CAD chronic annual dose

CFR Code of Federal Regulations

d day

DCAL Dose and Risk Calculation (program)

DCAS Division of Compensation Analysis and Support

DL decision level

DOE U.S. Department of Energy

ET extrathoracic

g gram

GI gastrointestinal

GSD geometric standard deviation

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

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Assessment

IREP Interactive RadioEpidemiological Program

keV kilovolts-electron, 1,000 electron volts

L liter

LLI lower large intestine

LNET extrathoracic lymph nodes

LNTH thoracic lymph nodes

LOD limit of detection

MDA minimum detectable activity or amount

MDD minimum detectable dose

MPC maximum permissible concentration

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

OBT organically bound tritium

ORAUT Oak Ridge Associated Universities Team

PID Principal Internal Dosimetrist

POC probability of causation

RBM red bone marrow

REF radiation effectiveness factor

SI small intestine

SMT stable metal tritide

SRDB Ref ID Site Research Database Reference Identification (number)

TIB technical information bulletin

U.S.C. United States Code
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 6 of 34

ULI upper large intestine

§ 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, APPLICABILITY, AND LIMITATIONS

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

case. Options are dependent on several factors including the employment site, type and number of 

cancers, and availability of monitoring data for the Energy Employee. Not all of the approaches this 

document provides are applicable or appropriate for all cases.

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 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 claimant file or the site profile. Claimant information also takes precedence over default values in 

the site profile. For example, most site profiles contain tables of minimum detectable activities or 

amounts (MDAs) for urine sample results. If the claimant 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 case rather than the default value from the site profile. Note that this

direction does not apply if the claimant 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. 

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 4.0.

2.0 TERMINOLOGY

2.1 SYSTEMIC VERSUS NONSYSTEMIC

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

circulation.
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Material can be directly deposited in the respiratory and gastrointestinal (GI) tracts through the 

inhalation and ingestion of material, so they are considered nonsystemic organs. The lymph nodes 

are included in the nonsystemic organs. These regions are defined by the International Commission 

on Radiological Protection (ICRP).

The respiratory tract includes the extrathoracic regions (ET1 and ET2), lungs, thoracic lymph node 

(LNTH), extrathoracic lymph node (LNET), and BB (bronchial) region (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 International Commission on Radiological 

Protection (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” organs. 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 (RBM) and bone surfaces, are considered 

to be metabolic.

2.3 POSITIVE AND NEGATIVE RESULTS

Decision level (DL) is defined in ANSI/HPS N13.30-2011 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 practice, the result of an specific analysis is compared to the DL in order to decide if analyte is 

present in that sample.

Minimum detectable amount (MDA) is defined in the standard as being the smallest amount of an 

analyte in a sample that will be detected with a probability β of non-detection while accepting a 

probability α of erroneously deciding that a positive (non-zero) quantity of analyte is present in an 

appropriate blank sample. 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 should not be compared to a specific 

analytical result to make the detection decision.

When discussing bioassay analyses used 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. 

The use of MDA rather than DL for determining if a sample is positive is discussed in section 3.3. 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’s important to understand which is being discussed.

For this Project, a result equal to the MDA is considered to be negative. This definition was chosen 

because many sites recorded the value of the MDA in the employee records for results less than the 
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MDA. If it is clear that this is not the case, a result equal to the MDA should be treated as positive. 

This exception should be documented in the site profile or dose reconstruction report.

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

most common when the nuclide is easily detected, such as 3

H, and a result at the MDA produces a 

very small dose. In such cases, only measurements with values that exceed the reporting level are 

recorded in the employee files. That is, results between the MDA and the reporting level are 

considered less-than values and are recorded as 0 or “<[the reporting level].” In these cases, the 

reporting level becomes the MDA by default and a missed dose should be based on the value of the 

reporting level rather than the MDA.

3.0 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 International Classification of Diseases, 9th Revision (ICD-9) codes with the 

appropriate organs and tissues to be modeled in the Integrated Modules for Bioassay Assessment 

(IMBA) computer program. For cases 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 will select 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 a proximity to a concentrating organ that is emitting photon radiation. In this 

case, 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.

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 will work) and load the ICRP defaults using the 

button on the tool bar;

2. Click Biokinetics (near the Close button); and

3. Click Load ICRP Defaults.

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

modeled organs (organs that are specifically named in the model as concentrating the radionuclide) 

for the element. Therefore, the remaining organs, with the exceptions noted in Section 2.2, are not 
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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 case takes precedence (see Section 1.2). However, when values 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, ingestion, injection, wound, and vapor.

Inhalation is the most common route for workplace intakes, although there can be an ingestion

component associated with it. 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 1990).

Intake via a 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 2005d), discusses modeling these intakes. 

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

(NCRP) Report 156 (NCRP 2006). 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 (Type F) is 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 case 

file.

3.2.3 Material Type

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 
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members of the public. 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. Additional nonstandard 

material types should be applied in some cases, type Super S plutonium being the primary example.

For the majority of cases, 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 cases, an assessment of 

each type to which the element is assigned in ICRP Publication 68 (ICRP 1995a) (this is indicated in 

the IMBA selection menu) should be made, and the type that results in the largest dose to the organ 

of interest should be selected. The exception to this rule is in the case of a radionuclide where trace 

atoms are bound in a matrix of another nuclide. The primary example of this is 241Am in a plutonium 

matrix. Americium is assigned to type M by ICRP Publication 68 but, when bound in a type S 

plutonium matrix, type S is also assumed for the 241Am. Therefore, ICRP Publication 71 

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

used). This also applies to natural thorium where the progeny (e.g., 228Ac, 228Ra, 224Ra) are assumed 

to be the same material type as the thorium parent.

For contaminants in a recycled uranium mixture, the nonuranium nuclides are assumed to be 

relatively tightly bound but not to behave outside of their own models. This is because the materials 

are primarily contaminants that have been mixed in with the uranium rather than progeny, so they are 

not as intimately bound. Therefore, the same material type should be applied to the contaminants as 

that selected for the uranium. If ICRP Publication 78 (ICRP 1998) does not assign the nuclide to the 

chosen uranium material type, the closest solubility should be selected. Table 3-1 provides guidance 

on material types for several elements that could be included as a recycled uranium contaminant.

Table 3-1. Selected material types for 

recycled uranium contaminants.

Uranium 

type Contaminant

Contaminant 

types 

(unmonitored)

F Pu M

Np M

Tc F

Th M

M Pu M

Np M

Tc M

Th M

S Pu S

Np M

Tc M

Th S

In the case of 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 adjustment 

factors to the intake and dose.

3.2.4 Radiation Type

The Interactive RadioEpidemiological Program (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 
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 12 of 34

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 2002).

Many radionuclides emit multiple types of radiation. IMBA Expert ORAU-Edition has an option to split 

the dose assessment into its components for many but not all radionuclides. However, this adds 

significant time to the analysis and yields at least two times the number of entries that are required for 

IREP. Because of these concerns, and the IREP limit on number of doses that can be input, each 

radionuclide is assigned to a single radiation type. Attachment A contains the values to be used for all 

nuclides available in IMBA and the Project-developed tools. The radionuclides were assigned to a 

REF category based on the type and energy of emitted radiation that, on inspection, would appear to 

give the highest absorbed dose to the source organ. Consideration was also given to radiation types 

that are assigned a larger weighting factor.

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 lists 

specific publications to be used (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 to be 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.

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 are normalized prior to entry to the IMBA program. This applies to all 

data types (see 3.2.6.1). 

If there is no indication as to the sample collection period and the site profile contains no additional 

information, normalize the result when the sample volume is less than that 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.

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 DR Folders\DR 

Information\Internal Dosimetry\IMBA.
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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 will use 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 is used for results less than the 

MDA, as described in the section on fitting positive bioassay data (Section 3.4). 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 will 

be treated as a distribution between zero and the LOD in the intake calculation. These results 

are displayed in brown on the graph.

• Excluded – IMBA will ignore all results with this Data Type indicated 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. The fit to the data will tend to move toward those results assigned the 

smallest errors on an absolute scale.

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 will result 

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

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 will 

have more precise statistics and might need to be assigned relatively smaller errors to obtain a better 

fit. For a best estimate, alternative values for the error can be tried if a reasonable fit is not obtained 

(e.g., the majority of results appear to be underpredicted or the larger results are 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 a satisfactory fit. Use

of the Uniform Absolute option, with the same value entered for all results, will yield 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 factors on the results simply moves the data fit up and down, it will 

not change the shape of the curve.

3.2.6.3 Error Distribution

Individual bioassay results are assumed to have a normal distribution.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 14 of 34

3.2.6.4 IMBA Limitations

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

There are two categories of problems: shared versus independent kinetics and very short-lived 

radionuclides. In the case of the former issue, the current ICRP system models the progeny of a 

radionuclide using its kinetics, 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.

• Shared versus Independent Kinetics – The nuclides this issue affects are 210Pb, 223Ra, 224Ra, 

226Ra, 228Ra, 131Te, 131mTe, 228Th, 232Th, 234Th, 232U, and 233U. The value IMBA reports might be 

larger or smaller than the value that is calculated using independent kinetics, depending on the 

nuclide and organ, and not all organs are affected for all nuclides. A summary of the 

differences is presented in Table B-1.

• Very Short-Lived Nuclides – The affected radionuclides are 228Ac, 147Pr, 131Te, and 239U. For 

chronic intakes, the doses IMBA reports are smaller than the actual values. The longer the 

intake, the larger the underestimate. Acute intakes are unaffected.

Table B-1 shows the radionuclide and organ combinations for which IMBA doses are high or low 

relative to the true value. In these cases, annual dose coefficients have been calculated with the 

Dose and Risk Calculation (DCAL) program (Eckerman et al. 2006). These coefficients have been 

incorporated into the CAD Workbook and verification of the values is documented in ORAUT-OTIB0022, 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 all radionuclide and organ combinations in Table B-1 that are high or low in IMBA, CAD must be 

used to calculate the dose (IMBA can be used for all other dose calculations). IMBA cannot be used 

as an overestimate for the high cases or as an underestimate for the low cases because the 

comparison is based on the 50-year committed dose and the annual doses can vary from this. In 

addition, in some instances, IMBA is high for one material type but low for another, but this is not 

indicated in the table.

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

When sending a case 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.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 15 of 34

3.3 ASSIGNMENT OF MISSED AND UNMONITORED DOSE

Although referred to as dose in the following section, intake is technically being assessed during the 

described periods. Doses from those intakes are assigned through the date of cancer diagnosis.

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 the following:

• Job title,

• Work location, and

• External dose.

By itself, 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), a potential for intake must be assumed for the entire employment period. In some cases, the 

assignment of environmental intakes only is appropriate. See section 3.6.5 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.

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

but, because of limitations in the monitoring system, was undetected. Missed dose is assigned using 

actual bioassay measurements and worker-specific employment information. In an operational 

setting, the application of dosimetric and biokinetic models to the MDA is the minimum detectable 

dose (MDD). The MDD is used for selecting bioassay programs to meet given monitoring 

requirements and is not assigned to workers. 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 the early years of the weapons complex, most sites used the MDA to determine if activity had been 

detected, so the “<” values reported by the site are used in the missed dose assessment. In more 

recent years, sites have used DL to make the decision regarding 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.

Unmonitored dose is the potential dose that could have been received by an Energy Employee but for 

which no monitoring of the individual was performed or monitoring data are not available. For 

unmonitored periods, 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.

3. Site-specific information. Note that this information typically takes the form of default intake 

values documented in the site profile.

4. Missed dose (extension of the missed dose calculation beyond the last bioassay result).
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 16 of 34

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, 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, 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, site-specific 

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. 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.

• Long-lived, long-retained nuclides include 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.

For short-lived or short-retained radionuclides:

• 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 

would be less. 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 cheap, 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.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 17 of 34

– 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.

• Short-lived, short-retained nuclides include 137Cs, 106Ru, 144Ce, and 90Sr.

3.3.1 Missed Dose Determination

Missed dose is assigned using actual bioassay sample parameters (e.g., specific dates and MDAs) 

and associated dosimetry program information. As noted in 3.3, the true missed dose falls between 0 

and the dose that would result in bioassay values equal to the MDA. This range is described by the 

triangular distribution, with a minimum of 0, a mode equal to the dose that yields bioassay results 

equal to half of the MDA, and a maximum that is twice that, or where the bioassay results are equal to 

the MDA. Therefore, when calculating missed dose, a result equal to half of the MDA is used in the 

calculation. The resulting dose is assigned as the mode of the triangular distribution. Note that in 

some instances the reporting level will be used in place of the MDA; see section 2.3 for information.

At a minimum, the MDA and date of the last sample result in the relevant period are used for the 

calculation. When calculating a missed dose to compare to a fitted dose (for best estimate), the MDA 

and date of the last result less than the MDA are used.

To calculate a missed dose, a chronic intake throughout the possible exposure period is assumed. 

The specific dates can vary depending on the bioassay method’s MDA over time.

If the detection threshold changes through the intake period, the following must be considered in 

determining the chronic intake:

• If the detection threshold decreases over time and the radionuclide/absorption type reaches 

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

transuranic elements or type S uranium), perform the fit using the date of the last sample and 

half of the associated detection threshold 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 will result in the overestimation of the later, smaller 

MDA values.

• 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 case 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 will be dependent on the pattern of the 

results but in general a line that runs through the center of the values would be suitable. 

Contact the Principal Internal Dosimetrist if you need assistance. 

• If the detection threshold decreases over time for radionuclide/absorption types that reach 

equilibrium rapidly, or if the detection threshold increases over time, use IMBA to determine 

chronic intakes applicable to each period (note that this is applicable only if there is a bioassay 

result in the period). To do this:

– In IMBA, set the number of intakes to the number of periods of different detection 

thresholds in which the Energy Employee has bioassay results.

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

date of the final sample in that detection threshold period.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 18 of 34

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

date of the last sample in the next detection threshold period, or to the last day of exposure 

for the final exposure period.

– Perform the fit assigning half of the associated detection threshold to the date of the last 

sample in each period.

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

using a triangular distribution and:

• Set the lower bound (Parameter 1) to 0.

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

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

3.3.2 Coworker Data

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 2005c), 

contains more specific information on the derivation of internal dosimetry data parameters. Sitespecific TIBs and site profiles are available for some sites and provide assumptions for intakes based 

on coworker dosimetry analyses.

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. 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 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 could be a matter of dose reconstructor 

judgment, again, based on information in the file.

3.3.3 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

Bioassay: All results <MDA

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
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 19 of 34

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 2014), perform the 

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

steps 1 through 5.

3.3.4 Short-Duration Missed Dose

The fraction of an element that is absorbed into blood from the small intestine (SI) 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 for a period of less than 1 

year, 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.

3.4 FITTING POSITIVE BIOASSAY RESULTS

This section describes the process for fitting results to an intake. 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 case. 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 will even vary throughout the 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 
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 20 of 34

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.

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 <MDA:

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

– Include the first negative (<MDA) result after each set of positive results and set Data Type 

equal to “<LOD.” 

– If there are multiple positive results, include no more than two consecutive negative 

results. For fewer than five consecutive positive results, include only one 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 <MDA results should have a Data Type of “Excluded.”

– Note that the presence of a result less than the MDA does not mean that a new intake 

must be assigned for the next result greater than the MDA.

• 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 many results less than the MDA or stretches of employment where all results are 

less than the MDA, missed dose will be assessed separately. See the discussion in 

Section 3.3.1.

• 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. 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 
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 21 of 34

(e.g., several <MDA 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 cases it may 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 in cases 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 <MDA results), use all results for the intake assessment. If the data 

are not censored (results <MDA are recorded as measured rather than as a “<” value or as the 

MDA), enter the result as recorded with a Data Type of Real. Otherwise, enter the MDA for 

the value and mark it <LOD. Note that the issue of measurement error (Section 3.2.6.2) also 

applies here.

3.5 OVERALL ASSESSMENT METHODS

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

which takes the least amount of time while still producing the correct decision. Many cases do not 

require a detailed, accurate dose assessment; efficiency methods can be used to expedite case 

completion with sufficient levels of precision to allow the U.S. Department of Labor 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 case, a more refined 

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 will 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.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 22 of 34

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 3.6; 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 will typically yield 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 a single result, 

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

intake.

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 will 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 will depend on the type of dose (e.g., missed and fitted) that is calculated.

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

cases 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 cases, 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 case-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 
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 23 of 34

U.S. Department of Labor 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. It 

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

those assumptions. 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 (unless all 

data can be fit at once, as discussed in Section 3.4.2) as follows:

1. 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 3.3.1. 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 results <MDA, no missed 

dose is calculated.

2. 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:

– 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.

– 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 geometric standard deviation 

(GSD) = 3. 

3.5.3.2 Multiple Cancers

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

that the same material type must be used for all cancers for a given nuclide and intake combination 

for a best estimate. If there are exceptions to this rule, they will be documented in TIBs specific to the 

circumstance.

3.6 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.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 24 of 34

3.6.1 OCAS-IG-002, Internal Dose Reconstruction Implementation Guideline

DCAS general guidance document on methods and approaches that can be used to reconstruct 

occupational radiation dose from internally deposited radionuclides in support of EEOICPA.

3.6.2 ORAUT-OTIB-0001, Maximum Internal Dose Estimates for Savannah River Site (SRS) 

Claims

This method uses the largest recorded intakes in the history of the Savannah River Site 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.

3.6.3 ORAUT-OTIB-0002, Maximum Internal Dose Estimates for Certain DOE Complex 

Claims

This document has been withdrawn and is no longer to be used.

3.6.4 ORAUT-OTIB-0011, Tritium Calculated and Missed Dose Estimates

IMBA will not directly calculate intakes from 3H urine data, so an alternative tool was developed. This 

OTIB provides documentation of the method for estimating tritium missed and calculated doses from 

urine data (ORAUT 2004c).

3.6.5 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.

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

Programs

This method is based on limiting air concentrations (ORAUT 2005a). 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 hours per 

week throughout the employment period. An additional conservatism is achieved by assuming that 

the airborne activity was comprised 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.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 25 of 34

3.6.7 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 2005d). Guidance on initial parameters is 

provided.

3.6.8 ORAUT-OTIB-0033, Application of Internal Doses Based on Claimant-Favorable 

Assumptions for Processing as Best Estimates

This method (ORAUT 2005b) applies a graded approach to internal dose overestimates and unites 

the application of ORAUT-OTIB-0014 (ORAUT 2004a), ORAUT-OTIB-0018 (ORAUT 2005a), and 

coworker dose. Some judgment is needed to apply these values; guidance is provided based on:

• The period during which the Energy Employee worked,

• The processes conducted at the site at which the Energy Employee worked,

• The job category and work location of the Energy Employee, and

• The results of bioassay measurements for the Energy Employee.

3.6.9 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 OTIB provides adjustment 

factors 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).

3.6.10 ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal DoseRelated 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 OTIB 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 2014). The derived ratios can also be applied to cases 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.

3.6.11 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 (OBT) and stable metal tritides (SMT).

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 
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 26 of 34

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 (its half-life is 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.

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 noncompensable claims, it is 

acceptable to overpredict one of the sets of monitoring results. However, if the claim is compensable 

the selected intake scenario must not contradict any of the claimant’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 lung 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 

noncompensable case but not for a compensable case. 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 whether the bioassay being used is urine or chest counting.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 27 of 34

3.7.2 Methods for Assessing Isotopic Uranium Results

Note: Isotopic results are from samples that were analyzed using alpha spectroscopy;

individual results were reported for 234U, 235U, 238U, and sometimes 236U.

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, 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, will result in an underestimate because at a minimum 234U, 235U, and 238U will be 

present in a mixture. 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 will 

not be present in equal concentrations for any mixture.

There are several options for assessing an isotopic uranium urine result. The selected method will be 

dependent on case details.

The general rule of thumb for all cases 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 Underestimate

Assessment of a single isotope can be used as an underestimate for a compensable claim. The 

appropriate IREP distribution will be dependent on the type of intake that was calculated:

• Triangular for missed dose, and

• Lognormal for fitted dose.

3.7.2.2 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).

Run the intake and dose estimates using the MDA values rather than 0.5 MDA.

Use a constant distribution for the doses in IREP.

3.7.2.3 Best Estimate

If all isotopes are >MDA:

• Sum the individual results.

• Run the intake and dose as 234U.

• Use a lognormal distribution and GSD = 3 in IREP.

If 234U and 238U >MDA:

• Sum the 234U and 238U results.

• Run the intake assessment as 234U.

• Divide the result by 0.965 to obtain total intake activity.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 28 of 34

Note: Uranium-235 activity content typically only varies between 1% and 3.5% of the total 

activity regardless of the enrichment (see the uranium mixes in IMBA for examples). The 

factor here assumes the maximum amount; if it is overestimated, it will have little impact on 

the total dose because it is a small fraction. However, if the enrichment is known for a given 

claim, the known value should be used.

• Run the intake calculated in the previous step as 234U for the dose assessment.

• Use a lognormal distribution and GSD = 3 in IREP.

If only 234U or only 238U >MDA:

• Run the intake assessment for the positive result.

• Determine the appropriate enrichment assumption.

• Calculate the intakes of the remaining isotopes from the ratios in an assumed enrichment.

• Sum the intake rates for all isotopes and assess the dose as 234U.

• Use a lognormal distribution and GSD = 3 in IREP.

If no isotopes >MDA (note that this is now simply a missed dose assessment):

• Determine the appropriate enrichment assumption.

• Assess 234U (238U for depleted uranium) intake using 0.5 MDA for the result.

• Calculate the intakes of the remaining isotopes from the ratios in an assumed enrichment.

• Sum the intake rates for all isotopes and assess the dose as 234U.

• Use a triangular distribution in IREP with:

– Min = 0.

– Mode = calculated dose.

– Max = 2 × calculated dose.

Note: The substitution of 234U for all isotopes does not apply to mixtures that contain 232U.

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 

(WLM). 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 WLM 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, all components 

must be considered and summed for the dose comparison when determining the type that is favorable 

to the claimant.

When several nuclides may have been present at a site, the one or two primary dose contributors 

were frequently monitored and reported. 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 
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 29 of 34

plutonium and uranium mixtures specific to the site, as well as other mixtures that are site-specific. 

ORAUT-OTIB-0054 (ORAUT 2014) provides guidance on the assignment of mixed fission and 

activation products.

When a best estimate is being performed, 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 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. 01 Effective Date: 09/08/2014 Page 30 of 34

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.

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.

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), 1990, Age-dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 1, Publication 56, Pergamon Press, 

Oxford, England.

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, Publication 78, Pergamon Press, Oxford, England. [SRDB Ref ID: 

36902]

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. [SRDB Ref ID: 44742]

NIOSH (National Institute for Occupational Safety and Health), 2004, Estimation of Ingestion Intakes, 

OCAS-TIB-009, Rev. 00, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

April 13.

NIOSH (National Institute for Occupational Safety and Health), 2013, Lung Dose Conversion Factor 

for Thoron WLM, DCAS-TIB-0011, Rev. 4, Office of Compensation Analysis and Support, 

Cincinnati, Ohio, May 1.

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.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 31 of 34

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.

ORAUT (Oak Ridge Associated Universities Team), 2004a, Technical Information Bulletin: 

Assignment of Environmental Internal Doses for Employees Not Exposed to Airborne 

Radionuclides in the Workplace, ORAUT-OTIB-0014, Rev. 00, Oak Ridge, Tennessee, June 

22.

ORAUT (Oak Ridge Associated Universities Team), 2004b, Coworker Data Exposure Profile 

Development, ORAUT-PLAN-0014, Rev. 00, Oak Ridge, Tennessee, November 24.

ORAUT (Oak Ridge Associated Universities Team), 2004c, Tritium Calculated and Missed Dose 

Estimates, ORAUT-OTIB-0011, Rev. 00, Oak Ridge, Tennessee, June 29.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Internal Dose Overestimates for Facilities 

With Air Sampling Programs, ORAUT-OTIB-0018, Rev. 01, Oak Ridge, Tennessee, August 9.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Application of Internal Doses Based on 

Claimant-Favorable Assumptions for Processing as Best Estimates, ORAUT-OTIB-0033, 

Rev. 00, Oak Ridge, Tennessee, April 20.

ORAUT (Oak Ridge Associated Universities Team), 2005c, Analysis of Coworker Bioassay Data for 

Internal Dose Assignment, ORAUT-OTIB-0019, Rev. 01, Oak Ridge, Tennessee, October 7.

ORAUT (Oak Ridge Associated Universities Team), 2005d, Guidance on Wound Modeling for Internal 

Dose Reconstruction, ORAUT-OTIB-0022, Rev. 00, Oak Ridge, Tennessee, November 18.

ORAUT (Oak Ridge Associated Universities Team), 2008, Validation of DCAL Annual Dose 

Coefficients, ORAUT-OTIB-0028, Rev. 02, Oak Ridge, Tennessee, July 28.

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.

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.

ORAUT (Oak Ridge Associated Universities Team), 2014, Fission and Activation Product Assignment 

for Internal Dose-Related Gross Beta and Gross Gamma Analyses, ORAUT-OTIB-0054, 

Rev. 02, Oak Ridge, Tennessee, March 6.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 32 of 34

ATTACHMENT A

RADIATION TYPES BY NUCLIDE FOR ENTRY INTO IREP

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

Nuclide Radiation typea Nuclide Radiation typea Nuclide Radiation typea

Ac-227 Alpha I-133 Electrons E >15 keV Rn-222 Radon

Ac-228 Alpha I-134 Photons E >250 keV Ru-103 Electrons E >15 keV

Ag-110m Photons E >250 keV I-135 Photons E >250 keV Ru-106 Electrons E >15 keV

Am-241 Alpha Ir-192 Photons E >250 keV S-35 Electrons E >15 keV

Am-243 Alpha La-140 Photons E >250 keV Sb-124 Photons E >250 keV

As-74 Photons E >250 keV Lu-174 Photons E = 30–250 keV Sb-125 Photons E >250 keV

As-76 Electrons E >15 keV Mn-54 Photons E >250 keV Sc-46 Photons E >250 keV

Au-194 Photons E <30 keV Mn-56 Electrons E >15 keV Sm-151 Electrons E >15 keV

Ba-133 Photons E >250 keV Mo-99 Electrons E >15 keV Sn-113 Electrons E >15 keV

Ba-140 Photons E >250 keV Na-22 Photons E >250 keV Sr-85 Photons E >250 keV

Bk-249 Alpha Na-24 Photons E >250 keV Sr-89 Electrons E >15 keV

C-14 Electrons E >15 keV Nb-94 Photons E >250 keV Sr-90 Electrons E >15 keV

Ca-45 Electrons E >15 keV Nb-95 Electrons E >15 keV Sr-91 Electrons E >15 keV

Ce-139 Photons E = 30–250 keV Ni-63 Electrons E >15 keV Ta-182 Photons E >250 keV

Ce-141 Electrons E >15 keV Np-237 Alpha Tb-160 Electrons E >15 keV

Ce-143 Electrons E >15 keV Np-239 Alpha Tc-99 Electrons E >15 keV

Ce-144 Electrons E >15 keV P-32 Electrons E >15 keV Te-131 Photons E >250 keV

Cf-249 Alpha P-33 Electrons E >15 keV Te-131M Electrons E >15 keV

Cf-252 Alpha Pa-231 Alpha Th-228 Alpha

Cl-36 Electrons E >15 keV Pa-233 Alpha Th-230 Alpha

Cm-242 Alpha Pa-234 Alpha Th-232 Alpha

Cm-243 Alpha Pb-210 Alpha Th-234 Alpha

Cm-244 Alpha Pm-147 Electrons E >15 keV Tl-201 Photons E = 30–250keV

Co-57 Photons E = 30–250 keV Po-208 Alpha Tl-202 Photons E >250 keV

Co-58 Photons E >250 keV Po-209 Alpha Tl-204 Electrons E >15 keV

Co-60 Photons E >250keV Po-210 Alpha Tm-170 Electrons E >15 keV

Cr-51 Photons E >250 keV Pr-143 Photons E >250 keV U-232 Alpha

Cs-134 Electrons E >15 keV Pr-147 Electrons E >15 keV U-234 Alpha

Cs-137 Electrons E >15 keV Pu-236 Alpha U-235 Alpha

Eu-152 Photons E >250 keV Pu-238 Alpha U-236 Alpha

Eu-154 Electrons E >15 keV Pu-239 Alpha U-238 Alpha

Eu-155 Electrons E >15 keV Pu-240 Alpha U-239 Alpha

Eu-156 Electrons E >15 keV Pu-241 Alpha Y-88 Photons E >250 keV

Fe-55 Photons E <30 keV Pu-242 Alpha Y-90 Electrons E >15 keV

Fe-59 Electrons E >15 keV Ra-220 Alpha Y-91 Electrons E >15 keV

H-3 Electrons E <15 keV Ra-223 Alpha Yb-169 Photons E >250 keV

Hf-181 Electrons E >15 keV Ra-224 Alpha Zn-65 Photons E >250 keV

I-125 Photons E <30 keV Ra-226 Alpha Zr-95 Electrons E >15 keV

I-129 Photons E = 30–250 keV Ra-228 Alpha

I-131 Electrons E >15 keV Rn-220 Alpha

a. E = energy; the 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. 01 Effective Date: 09/08/2014 Page 33 of 34

ATTACHMENT B

IMBA DOSE ASSESSMENT LIMITATIONS

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

See Section 3.2.6.4 for a complete discussion of the issues. Table B-1 provides a summary of the 

differences between IMBA-calculated doses and the true value; those indicated to be high or low 

cannot be assessed with IMBA. These are general trends and may not apply to all material types for 

the listed organ or for any given year of annual dose.

Table B-1. IMBA doses relative to the true value (calculated using DCAL).

a

Organ

Ac-228 

(chronic 

intakes 

only) Pb-210

Pr-147 

(chronic 

intakes 

only) Ra-223 Ra-224 Ra-226 Ra-228 Te-131 Te-131m

Adrenals Low High Low – Low – Low (b) High

Bladder Low High Low – Low – Low (b) –

Bone surf Low – Low – High – Low (b) High

Brain Low High Low – Low – Low (b) –

Breast Low High Low – Low – Low (b) High

Esophagus Low High Low – Low – Low (b) –

Kidneys Low Low Low Low Low Low Low (b) High

Liver Low High Low – Low – Low (b) High

Muscle Low High Low – Low – Low (b) –

Ovaries Low High Low – Low – Low (b) High

Pancreas Low High Low – Low – Low (b) High

RBM Low Low Low – High – Low (b) High

Skin Low High Low – Low – Low (b) High

Spleen Low Low Low – Low Low Low (b) High

Testes Low High Low – Low – Low (b) High

Thymus Low High Low – Low – Low (b) –

Thyroid Low High Low – Low – Low (b) Low

Uterus Low High Low – Low – Low (b) High

LNET Low High Low – – – – (b) –

LNTH Low High Low – – – – (b) High

ET1 Low High Low – – – – (b) –

ET2 Low High Low – – – – (b) –

Lungs Low High Low – – – – (b) –

Stomach Low High Low – Low – Low (b) High

SI Low High Low – – – Low (b) –

ULI Low High Low – – High Low (b) –

LLI Low High Low – – High Low (b) –

Colon Low High Low – – High Low (b) –

a. A dash (-) indicates that the IMBA value is less than 10% different from the DCAL values and can be used for a dose 

assessment.

b. Because Te-131 falls into both categories, it is not possible to cover all possible scenarios. There are two competing 

issues: (1) overestimates for a number of organs because of the kinetics issue and (2) underestimates of all organs for 

chronic intakes, the magnitude of which increases as the length of intake increases. Therefore, IMBA cannot be used to 

calculate doses from Te-131.
Document No. ORAUT-OTIB-0060 Revision No. 01 Effective Date: 09/08/2014 Page 34 of 34

ATTACHMENT B

IMBA DOSE ASSESSMENT LIMITATIONS (continued)

Organ Th-228 Th-229 Th-232 Th-234 U-232

U-233 

(type S 

inh. only)

U-239 

(chronic 

intakes only)

Adrenals High High High – Low – Low

Bladder High High High – Low – Low

Bone surf High High High High Low Low Low

Brain High High High – Low – Low

Breast High High High – Low – Low

Esophagus High High High – Low – Low

Kidneys High High High – Low – Low

Liver – – High – Low Low Low

Muscle High High High – Low – Low

Ovaries High High High – Low Low Low

Pancreas High High High – Low – Low

RBM High High High – Low Low Low

Skin High High High – Low – Low

Spleen High High High – Low – Low

Testes High High High – Low Low Low

Thymus High High High – Low – Low

Thyroid High High High – Low – Low

Uterus High High High – Low – Low

LNET – High High – – – Low

LNTH – – High – – – Low

ET1 – – High – – – Low

ET2 – – High – – – Low

Lungs – – High – – – Low

Stomach High High High – Low – Low

SI High High High – Low – Low

ULI Low High High – Low – Low

LLI Low High High – Low – Low

Colon Low High High – Low – Low

a. A dash (-) indicates that the IMBA value is less than 10% different from the DCAL values and can be used

 for a dose assessment.