ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

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Document Title:

Internal Dose Reconstruction

Document Number: ORAUT-OTIB-0060

Revision: 00

Effective Date: 02/06/2007

Type of Document: OTIB

Supersedes: ORAUT-PROC-0003

Subject Expert(s): Elizabeth M. Brackett 

Site Expert(s): N/A

Approval: Signature on File Approval Date: 01/25/2007

James P. Griffin, Document Owner

Concurrence: Signature on File Concurrence Date: 01/24/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 02/06/2007

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. 00 Effective Date: 02/06/2007 Page 2 of 24

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.
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 3 of 24

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

1.0 Introduction ..............................................................................................................................6

2.0 Purpose ...................................................................................................................................7

3.0 Applicability and Limitations .....................................................................................................7

4.0 Terminology .............................................................................................................................7

4.1 Systemic vs. Nonsystemic ............................................................................................7

4.2 Metabolic vs. NonMetabolic Organs .............................................................................7

5.0 Guidance .................................................................................................................................7

5.1 Selection of Appropriate Organs for Dose Estimate......................................................7

5.2 Selection of Parameter Values .....................................................................................8

5.2.1 Intake Mode ......................................................................................................8

5.2.2 Particle Size Distribution ...................................................................................8

5.2.3 Material Type ....................................................................................................9

5.2.4 Radiation Type..................................................................................................9

5.2.5 Integrated Modules for Bioassay Analysis (IMBA)-Specific 

Parameters ..................................................................................................... 10

5.2.5.1 Data Type......................................................................................... 10

5.2.5.2 Measurement Error........................................................................... 10

5.2.5.3 Error Distribution............................................................................... 11

5.3 Fitting Positive Bioassay Results ................................................................................11

5.3.1 General Philosophy......................................................................................... 11

5.3.2 General Guidelines ......................................................................................... 11

5.4 Assignment of Missed and Unmonitored Dose ...........................................................13

5.4.1 Missed Dose Determination ............................................................................ 15

5.4.2 Coworker Data................................................................................................ 16

5.4.2.1 Application of Coworker Data ........................................................... 16

5.4.2.2 Type S Plutonium ............................................................................. 16

5.4.3 Example Assignment of Coworker and Unmonitored Doses ........................... 17

5.5 Assessment Methods .................................................................................................17

5.5.1 Overestimate .................................................................................................. 18

5.5.1.1 Generic Overestimates..................................................................... 18

5.5.1.2 Individual-Specific Overestimates..................................................... 18

5.5.2 Underestimate................................................................................................. 18

5.5.3 Best Estimate.................................................................................................. 19

5.5.3.1 Performing a Best Estimate Using Bioassay Data ............................ 19

5.6 Guidance Documents .................................................................................................19

5.6.1 OTIB-0001: Maximum Internal Dose Estimates for 

Savannah River Site (SRS) Claims................................................................. 20

5.6.2 OTIB-0002: Maximum Internal Dose Estimates for Certain 

DOE Complex Claims ..................................................................................... 20
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5.6.3 OTIB-0014: Assignment of Environmental Internal Doses 

for Employees Not Exposed to Airborne Radionuclides in 

the Workplace................................................................................................. 20

5.6.4 OTIB-0018: Internal Dose Overestimates for Facilities with 

Air Sampling Programs ................................................................................... 21

5.6.5 OTIB-0033: Application of Internal Doses Based on 

Claimant-Favorable Assumptions for Processing as Best 

Estimates........................................................................................................ 21

5.7 Recycled Uranium ......................................................................................................21

References ........................................................................................................................................22

Attachment A.....................................................................................................................................24

LIST OF TABLES

TABLE TITLE

5-1 The 95% confidence intervals and medians for IREP REF classes ........................................

PAGE

10

5-2 Material types for recycled uranium contaminants.................................................................. 21
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ACRONYMS AND ABBREVIATIONS

d day

DOE U.S. Department of Energy

g gram

GI gastrointestinal

GSD geometric standard deviation

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Assessment (computer program)

IREP Interactive RadioEpidemiological Program

LOD limit of detection

MDA minimum detectable activity

MPBB maximum permissible body burden

MPC maximum permissible concentration

nCi nanocurie

pCi picocurie

POC probability of causation

REF radiation effectiveness factor

TIB technical information bulletin

U.S.C. United States Code
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 6 of 24

1.0

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 

historic 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. TIBs may be used to assist the NIOSH staff in the completion of individual dose 

reconstructions.

INTRODUCTION

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 [EEOICPA; 42 U.S.C. § 7384l(5) and (12)].

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 7384l

(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted above, 

any facility that performs or performed DOE operations of any nature whatsoever is a DOE facility 

encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation*

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)].

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at the facility in its dose reconstructions for employees at DOE facilities, including radiation 

exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and external 

dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts are made 

to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ....” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 * The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
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2.0

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.

PURPOSE

3.0

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

case. Options will be dependent on a number of factors, including the employment site, type and 

number of cancers, and availability of monitoring data for the individual. Not all approaches provided 

here will be applicable or appropriate for all cases.

APPLICABILITY AND LIMITATIONS

4.0

4.1 SYSTEMIC VS. NONSYSTEMIC

TERMINOLOGY

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

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

through the inhalation and ingestion of material, and are considered nonsystemic organs. The lymph 

nodes are included in the nonsystemic organs.

4.2 METABOLIC VS. NONMETABOLIC ORGANS

A number of organs are included in the general models provided by the International Commission on 

Radiological Protection (ICRP 1998). However, for a given element, only a specific subset of these 

organs is included in the metabolic or biokinetic modeling. Those organs that are specifically modeled 

are referred to as “metabolic” for the purposes of this Project. Others for which a dose is calculated 

but are not specified by the ICRP element-specific model are called 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) are considered to be metabolic.

5.0

5.1 SELECTION OF APPROPRIATE ORGANS FOR DOSE ESTIMATE

GUIDANCE

Dose must be calculated to the organ in which the cancer originated. ORAUT (2006) correlates 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 

assigned to the organ is the largest dose among the organs reported in IMBA that are not part of the 

ICRP (1998) metabolic model (referred to as the nonmetabolic organs, as described in Section 4.2) 

for the particular radionuclide.

Organs that do not concentrate a radionuclide will 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. Using these techniques, many of these other organ 

doses are calculated. Because these organs are all considered soft tissue and thus are all similarly
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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 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 part of the metabolic model. Select the 

appropriate element (any isotope will work), load the ICRP defaults (using the button on the tool bar), 

and select the Biokinetics button (close to the Close button). Click on 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 4.2, are not part of 

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

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

5.2 SELECTION OF PARAMETER VALUES

5.2.1

There are five intake routes included in IMBA: inhalation, ingestion, injection, wound, and vapor. 

Intake Mode

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

component associated with it. When an inhalation intake is assigned based on air monitoring data 

rather than bioassay, an additional ingestion component must be assigned. See OCAS-TIB-009 

(NIOSH 2004) for guidance.

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

absorption. The ICRP (1990) treats the inhalation of tritium oxide as injection.

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 (2005d) contains guidance on 

modeling these intakes.

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

substances that are normally in liquid or solid form (James 2005). Iodine is typically modeled using 

this intake route.

In the absence of information regarding 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.. 

5.2.2

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 (1994) default value is 5 μm 

activity median aerodynamic diameter. This value is to be used for evaluating inhalation intakes in the 

absence of known information, as documented in the site profiles or the case file.

Particle Size Distribution
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5.2.3

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

ICRP (1994) describes three types, F (fast solubilization), M (moderate solubilization), and S (slow

solubilization); the assignment of an element to one 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, as opposed to members of the public. If material types at a 

particular site are known, they will be documented in the site profile and should be used. However, 

for the majority of cases it is likely that the material type is unknown or the individual might have 

worked in multiple areas, making exposure to multiple types possible. In such cases, an assessment 

of each type to which the element is assigned in ICRP 68 (this is indicated in the IMBA selection 

menu) shall be made, and that which results in the largest dose to the organ of interest shall be 

selected.

Material Type

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 Am-241 in a plutonium matrix; americium is assigned 

to Type M by the ICRP (1995a) but when bound in a Type S plutonium matrix, Type S is also 

assumed for the Am-241 and ICRP Publication 71 (ICRP 1995b) Type S is selected in IMBA.

5.2.4

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

effectiveness factor (REF) is analogous to the relative biological effectiveness used in radiobiology. In 

IREP, the equivalent dose to an organ is converted to absorbed dose and modified with the 

appropriate REF as part of the POC calculation. The REF selected depends on the type of radiation, 

of which there are six: 

Radiation Type

• Alpha

• Electrons <15 keV

• Electrons >15 keV

• Photons <30 keV

• Photons 30-250 keV

• Photons >250 keV

Photons >250 keV and electrons >15 keV are the reference radiations and have a point value REF of 

unity. The REFs for the other radiation types are expressed as a distribution. A summary of 2.5, 50, 

and 97.5 percentiles for the REF distributions is shown in Table 5-1. Note that only tritium is included 

in the electron <15 keV category. 

Many radionuclides emit multiple types of radiation. IMBA has an option to split the dose assessment 

into its components for a number of radionuclides, although not all. However, this adds significant 

time to the analysis and yields at least two times the number of entries 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 radiation emitted†

 † The program Radiation Decay (available at 

that, on inspection, would appear to give 

http://www.btinternet.com/~ablumsohn/chemis.htm) was used. For 

radionuclides not listed in Radiation Decay, output from the National Nuclear Data Center (http://www.nndc.bnl.gov/mird/) 

was used.
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Table 5-1. The 95% confidence intervals and medians for IREP 

REF classes.

Radiation 

class 2.5 percentile 50 percentile 97.5 percentile

Alpha

Solid tumors 3.4 18.0 101.0

Alpha

Leukemia 1.0 4.1 42.0

Photon

>250 keV - 1.0 -

Photon

30-250 keV 1.0 1.9 4.7

Photon

<30 keV 1.1 2.4 6.1

Electron

>15 keV - 1.0 -

Electron

<15 keV 1.2 2.4 5.0

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 daughters that do are assigned to the alpha 

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

is assigned to the “electrons E>15keV” radiation category, because of the extremely long half-life of 

the daughter, 147Sm.

5.2.5

5.2.5.1 Data Type

Integrated Modules for Bioassay Analysis (IMBA)-Specific Parameters

• Real: This data type indicates a valid result, to be used in the 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 5.3). The sample 

MDA is entered for the Measurement Rate or Measurement Value. IMBA uses a maximum 

likelihood method for estimating intakes; if the underlying error distribution of the 

measurements is known (or reasonably well-estimated), the probability that the true value is 

between zero and the limit of detection can be calculated as the area under the probability 

distribution curve.

• Excluded: IMBA will ignore all results with this Data Type indicated when calculating the 

intake. Excluded results are displayed in red on the graph. 

5.2.5.2 Measurement Error

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

Rate for the intake fit. For positive results, use the one sigma error associated with the result, when 

available. 

If there are no errors reported with the results, Measurement Error should be calculated using the 

Uniform Relative option, with k = 0.3 as a starting point. Note that in this instance the value of k is 
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somewhat arbitrary; the same intake will result from any percentage, given the same percentage for 

all results. This might not be a reasonable estimate if there are some reasonably well-defined peaks 

or results vary by more than 1 order of magnitude. Larger values will have more precise statistics and 

might need to be assigned relatively smaller errors to obtain a better fit. Alternative values for the 

error should 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 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). If you 

cannot obtain a reasonable fit, contact the Principal Internal Dosimetrist for assistance.

5.2.5.3 Error Distribution

Individual bioassay results are assumed to be normally distributed.

5.3 FITTING POSITIVE BIOASSAY RESULTS

This section describes the process for fitting results to an intake. Specific details of the mechanics of 

using IMBA are included in ORAUT (2003b).

Note: For the purpose of this discussion, positive means a result greater than the reporting level (this 

could be the MDA, detection level, or some other value that the site used) and negative means less 

than or equal to this value. This definition was chosen because many sites reported the value of the 

MDA for results less than the MDA; if it is clear that is not the case, a result equal to the MDA should 

be treated as positive.

5.3.1

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 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 5.5 for discussion), no further fitting should be tried. 

General Philosophy

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 reported 

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

materials, as well as the measurement of the material, are stochastic processes; varying amounts of 

material will be in the various measured compartments over time. 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, and often not favorable to claimants, 

assumptions. An example of this is fitting each positive result to a separate intake. This often 

requires the assumption that each intake occurred only one or two days prior to the bioassay sample; 

if the samples were collected as part of the routine bioassay program (as opposed to incident follow 

ups), it is unlikely that the program caught each unsuspected intake immediately upon occurrence.

5.3.2

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

following guidelines are provided:

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

• In IMBA, for results < MDA, Measurement Result = MDA value, Data Type = “< LOD.” Include 

the first negative (<MDA) result following each set of positive results. If there are multiple 

positive results, include no more than two 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. 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.

• 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 low 

level chronic intake just below the MDA. Note that the limitations on the use of “<LOD” apply 

here as well.

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

there are no errors reported with the results, Measurement Error should be calculated using 

the Uniform Relative option, with k = 0.3.

Note: This might not be a reasonable estimate if there are some reasonably well-defined 

peaks or results vary by more than an order of magnitude; see the discussion in Section 

5.2.5.2.

• Set Error Distribution = Normal.

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

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

assign intake date at the midpoint between two consecutive samples (this may not be 

reasonable if the results are more than a year or so apart). Intake dates should not be varied 

if there are only a few results for each intake unless projections from the intakes are 

inconsistent with later data (e.g., several <MDA results are predicted to have had detectable 

levels of activity). As discussed in 5.3.1, it is not necessary, nor desirable, to obtain an exact 

fit to each result because variation in excretion rate is to be expected.

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

largest total dose (note that the largest intake does not necessarily correlate to the largest 

dose) to the applicable organ (for the years of interest). If one type provides an unarguably 

better fit, use it. This can generally be shown only in cases where:

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

– There are contemporary (later than 1989) data associated with intakes 25 yr 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 yielding this larger 

intake would be ruled out).
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• 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 to do 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 = Real. Otherwise, enter the MDA for the 

value and mark it <LOD. Note that the issue of measurement error, discussed in Section 

5.2.5.2, also applies here.

5.4 ASSIGNMENT OF MISSED AND UNMONITORED DOSE

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

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

actual bioassay measurements and worker-specific employment information. 

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: 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).

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

will be excreted over an extended period of time, 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., Cs-137, Po-210). 

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 following 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 non-representative of the individual’s 

exposure. 

For long-lived/retained radionuclides:
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• 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 following the last bioassay sample through the 

end of the potential exposure period.

• Long-lived/retained nuclides include: plutonium, uranium, americium.

For short-lived/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 two years between results are considered to be 

unmonitored. Note: For very short retained materials, notably H-3 and I-131, this period would be 

less. Guidance on H-3 assessment is contained in OTIB-0011 (ORAUT 2004d). In general, H-3 

exposure is assumed only during periods where bioassay samples were collected because it is 

cheap, easy, and quick. The primary exception to this rule is when a site began using H-3 prior to the 

implementation of a bioassay program.

• Unmonitored dose is assigned for the period up until one year prior to the first bioassay 

sample for the nuclide of interest.

• Missed dose is calculated from one year prior to the first bioassay result through the date of 

the last bioassay sample.

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

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

o Unmonitored dose is assigned from the day after the first sample until one year prior to 

the second sample.

o Missed dose is calculated starting at one year prior to the date of the second result.

• Short-lived/retained nuclides include: Cs-137, polonium, Ru-106, Ce-144, Sr-90.

The presence of bioassay samples are often an indicator of potential for exposure, but baseline 

and/or termination samples alone (i.e., no bioassay other than a single baseline and/or a single 

termination measurement) do not necessarily indicate a potential. Indicators of potential for internal 

radiation exposure include the following: 

• Job title

• Work location

• External dose

• By itself, lack of sampling for extended periods is an insufficient reason for assuming a change 

in exposure potential. If the three items listed above do not change during an individual’s 

employment history but there is information, such as bioassay data or job title, that indicates a 
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 15 of 24

potential for intake at some point, a potential for intake must be assumed for the entire 

employment period.

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.

5.4.1

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

and associated dosimetry program information.

Missed Dose Determination

The MDA and date associated with the last sample result are used for the calculation. When

calculating a missed dose to compare to a fitted dose (for best estimate), the MDA and date 

associated with 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/

transuranics or Type S uranium), perform the fit using the date of the last sample and half of 

the associated detection threshold, assuming a single chronic intake for the entire potential 

exposure period.

• 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. 

– 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 the 

Interactive RadioEpidemiological Program (IREP) using a triangular distribution.

Set the lower bound (parameter 1) to zero. 

Set the mode (parameter 2) to the annual organ doses calculated above.
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 16 of 24

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

5.4.2

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

Weapons Employer sites. ORAUT (2004c) provides a generic discussion on the development of 

these data sets while ORAUT (2005c) contains more specific information on the derivation of internal 

dosimetry data parameters. Site-specific TIBs and site profiles are available for some sites and 

provide assumptions for intakes based on coworker dosimetry analyses.

Coworker Data

5.4.2.1 Application of Coworker Data

Coworker dose is applied as a best estimate to unmonitored individuals (or workers with unmonitored 

intervals) with a significant potential for intakes of radioactive material, or as an overestimate for those 

unlikely to have had intakes. When coworker analyses do not define how or to whom the intake 

should apply, “significant potential” is subjective but in general applies to people who were radiation 

workers. The Dose Reconstructor must make this decision based on the worker’s job titles and work 

locations, as well as any other information in the file that could indicate a potential for intake. 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, but this could be a matter of Dose 

Reconstructor judgment, again based on information in the file.

5.4.2.2 Type S Plutonium

Because of the interdependence between individual bioassay results and because detection 

sensitivities and radiation controls have improved over time, it is generally not possible to fit Type S 

plutonium to the coworker data in a manner that would be representative of all individuals for all 

periods. In general, intakes based on the early results greatly overpredict the later values. Therefore, 

only a minimizing intake has been calculated for Type S plutonium in most of the coworker TIBs. In 

such cases, to apply coworker data to sites with plutonium:

Use the Type M plutonium intakes for all systemic organs.

For nonsystemic organs, do the following:

• Run the Type M intakes. If this results in a POC > 50%, no further assessment is necessary. 

If this does not result in a POC > 50%,

• Run the minimizing Type S intake. If this still does not yield a POC > 50%,

• Manually fit the coworker bioassay data (these data are listed in a table in the site-specific TIB; 

electronic versions of the information are available through the Principal Internal Dosimetrist)

for the time frame of interest for the employee, using the assumption of Type S material. Use 

standard fitting techniques to fit the plutonium urinalysis. Acute or chronic intakes can be 

assigned, depending on the patterns in the data. Both the 50th - and 84th -percentile data 

must be fit. If a lognormal distribution is assumed, the 50th-percentile intakes are assigned as 

the intake and the 84th-percentile is used to determine the geometric standard deviation 

(GSD) for each intake (GSD = 84th/50th percentile intake for each period). No GSD less than 

3 may be assigned (i.e., if the calculated GSD<3, enter 3 in IREP).
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 17 of 24

5.4.3

Employment: 3/1/1957 to 7/12/1989

Example Assignment of Coworker and Unmonitored Doses

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

dose results 

Bioassay: Plutonium urine sample on 5/4/1960, 12/11/1963, 11/17/1970

Fission product urine samples 3/12/1965, 9/18/1965, 8/1/1966, 1/5/1978, 7/12/1978, 

4/30/1979

If no coworker data or site default values exist, calculate missed dose from 3/1/1957 through 

7/12/1989 using 0.5 MDA on 11/17/1970.

Plutonium dose assignments:

If there are coworker data or site default values:

1. Calculate missed dose from 3/1/1957 through 11/17/1970 using 0.5 MDA on 

11/17/1970. (Note that “real” dose could also be assigned in this time frame if there 

were positive bioassay results).

2. For the period from 11/18/1970 through 7/12/1989:

Assign coworker dose if it’s available.

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

individuals who were potentially exposed, assign the TBD values.

Check the TBD (or coworker OTIB) for the nuclide to be assumed when running FP results. In most 

cases these are Sr-90 or Cs-137, and in general they will be relatively short-lived/retained.

Fission product assignment:

1. Assign coworker data or site default values from 3/1/1957 to 3/12/1964.

2. Calculate missed dose (or real dose, if there are positive bioassay results) from 

3/13/1964 to 8/1/1966.

3. Assign coworker data or site default values from 8/2/1966 to 1/5/1977.

4. Calculate missed dose (or real dose, if there are positive bioassay results) from 

1/6/1977 to 4/30/1979.

5. Assign coworker data or site default values from 5/1/1979 to 7/12/1989

6. If there are mixtures associated with the fission products, perform the above 

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

steps 1 through 5.

5.5 ASSESSMENT METHODS

There are typically a number of approaches that can be applied to any 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 while retaining an accurate compensation decision. 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 best estimate is needed.
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 18 of 24

5.5.1

An overestimate, for an internal dose assessment, is assignment of an intake or dose that exceeds 

the possible exposure of the worker. If the resulting probability of causation (POC), including all 

sources of potential exposure, is less than 45% (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.

Overestimate

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 employee-specific information.

5.5.1.1 Generic Overestimates

Several methods have been developed and documented in technical information bulletins. 

Summaries and general applicability of these are described in Section 5.6; specific details are in the 

documents. 

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). 

Date limitations apply to several of the overestimate methods. These time restrictions can be relaxed 

for individual cases where it can be shown that the appropriate limits or monitoring assumed in the 

basis of the method were in place at the facility in question.

5.5.1.2 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) vs. 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.

5.5.2

An underestimate is the assignment of an intake or dose to a worker that is less than the intake or 

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

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

Underestimate
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 19 of 24

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.

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.

5.5.3

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 

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 given these data. 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. 

Best Estimate

5.5.3.1 Performing a Best Estimate Using Bioassay Data

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

data can be fit at once, as discussed in 5.3.2), as follows:

1. Ignore positive data (this means that the date of the last result less than the MDA is used for 

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

in Section 5.4.1. If multiple material types are possible, select the one yielding the largest total 

dose to the applicable organ for the years of interest.

2. Fit the positive data in accordance with Section 5.3.

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

the IREP annual dose distribution type based on missed or “positive” dose assignment:

• For years in which the dose determined in 1 is larger than that from 2, use the triangular 

distribution, where Min = 0, Mode = Annual dose, and Max = Mode*2.

• For years in which the dose from 2 is equal to or larger than that from 1, use the lognormal 

distribution, where Median = Annual dose and GSD = 3.

5.6 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 
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 20 of 24

only; refer to the individual documents for the current revision and for applicability and limitations of 

each.

5.6.1

This method (see ORAUT 2003a) uses the largest recorded intakes in the history of the SRS to 

assign an overestimated intake for workers. An average of the largest five intakes for each monitored 

nuclide was used; although it is conceivable that a given individual had an intake in excess of one of 

the values, it is very unlikely that a single individual 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.

OTIB-0001: Maximum Internal Dose Estimates for Savannah River Site (SRS) Claims

5.6.2

This method (see ORAUT 2004a) is based on the ICRP Publication 2 control value of maximum 

permissible body burden (MPBB) (ICRP 1959). It assumes that an intake resulting in 10% MPBB 

would be noticed and that there would be some sort of indication in the worker’s records. The 

assumed source term includes the more significant radionuclides expected to be found on a variety of 

sites:

OTIB-0002: Maximum Internal Dose Estimates for Certain DOE Complex Claims

• Reactor sites

• Nonreactor sites

• Uranium sites

A tritium dose must be assigned separately (i.e., it is not included in this method) if there was a 

significant potential for intake. Because tritium intakes are very easy to monitor, an individual was 

likely to have been monitored should this be the case.

A single acute intake of all of the radionuclides listed for a particular site type is assumed, and is 

assigned on the first day of employment. The method allows for nuclides not present on a given site 

to be excluded from assignment, but this must be used with caution because part of the basis for the 

overestimate is that no single individual would have had undetected intakes of a large number of 

radionuclides. Application is limited to specific organs because the intakes were maximized for 

systemic organs.

5.6.3

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

radionuclides in the ambient environment resulting from site operations or contamination, as opposed 

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

employees, assignment of environmental dose only is appropriate. This document (ORAUT 2004b) 

provides guidance for determining such instances. Guidance is based on:

OTIB-0014: Assignment of Environmental Internal Doses for Employees Not 

Exposed to Airborne Radionuclides in the Workplace

• Job description

• Work location

• Time frame

• Presence/absence of internal monitoring data
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 21 of 24

5.6.4

This method (see ORAUT 2005a) is similar to that presented in OTIB-0002 (ORAUT 2004a), but is 

based on limiting air concentrations rather than MPBB. 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 employment. An additional conservatism is achieved by assuming that the airborne 

activity was comprised of the single nuclide, in each year of intake, resulting in the largest dose to the 

organ of interest, rather than assigning a mixture of radionuclides. 

OTIB-0018: Internal Dose Overestimates for Facilities with Air Sampling Programs

5.6.5

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

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

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

OTIB-0033: Application of Internal Doses Based on Claimant-Favorable 

Assumptions for Processing as Best Estimates

• 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.

5.7 RECYCLED URANIUM

When assigning dose due to contaminants in recycled uranium, the same material type should be 

applied to the contaminants as that selected for the uranium. If the ICRP (1998) does not assign the 

nuclide to the chosen uranium material type, the closest solubility should be selected. Table 5-2 

contains material types for several elements that could be included as a recycled uranium 

contaminant.

Table 5-2. Material types for recycled uranium 

contaminants.

Uranium type

F M S

Contaminant types:

(unmonitored)

Pu M Pu M Pu S

Np M Np M Np M

Tc F Tc M Tc M

Th M Th M Th S
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 22 of 24

REFERENCES

ICRP (International Commission on Radiological Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, Publication 2, Pergamon Press, Oxford, England.

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.

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.

ICRP (International Commission on Radiological Protection), 1995a, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

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.

ICRP (International Commission on Radiological Protection), 1998, Individual Monitoring for Internal 

Exposure of Workers, Publication 78, Pergamon Press, Oxford, England.

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

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

13.

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.

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 – Maximum 

Internal Dose Estimates for Certain DOE Complex Claims, ORAUT-OTIB-0002, Rev. 01 PC-2, 

Oak Ridge, Tennessee, May 7.

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

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

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

Estimates, ORAUT-OTIB-0011, Rev. 00, Oak Ridge, Tennessee, June 29.
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 23 of 24

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. 02, Oak Ridge, Tennessee, November 18.

ORAUT (Oak Ridge Associated Universities Team), 2006, External Dosimetry Organ, Internal 

Dosimetry Organ, and IREP Model Selection by ICD-9 Code, ORAUT-OTIB-0005, Rev. 02 

PC-1, Oak Ridge, Tennessee, February 10.

NBS (National Bureau of Standards), 1953, Maximum Permissible Amounts of Radioisotopes in the 

Human Body and Maximum Permissible Concentrations in Air and Water, Handbook 52, U.S. 

Department of Commerce, Washington, D.C., March 20. [SRDB Ref ID: 12332]
Document No. ORAUT-OTIB-0060 Revision No. 00 Effective Date: 02/06/2007 Page 24 of 24

ATTACHMENT A

Radionuclide Radiation type* Radionuclide Radiation type* Radionuclide Radiation type*

Ac-227 Alpha I-133 Electrons E>15 keV Rn-220 Alpha

Ac-228 Alpha I-134 Photons E>250 keV Rn-222 Radon

Ag-110m Photons E>250 keV I-135 Photons E>250 keV Ru-103 Electrons E>15 keV

Am-241 Alpha Ir-192 Photons E>250 keV Ru-106 Electrons E>15 keV

Am-243 Alpha La-140 Photons E>250 keV S-35 Electrons E>15 keV

As-74 Photons E>250 keV Lu-174 Photons E=30-250 keV Sb-124 Photons E>250 keV

Ba-133 Photons E>250 keV Mn-54 Photons E>250 keV Sb-125 Photons E>250 keV

Ba-140 Photons E>250 keV Mo-99 Electrons E>15 keV Sm-151 Electrons E>15 keV

Bk-249 Alpha Na-22 Photons E>250 keV Sn-113 Electrons E>15 keV

C-14 Electrons E>15 keV Na-24 Photons E>250 keV Sr-85 Photons E>250 keV

Ca-45 Electrons E>15 keV Nb-94 Photons E>250 keV Sr-89 Electrons E>15 keV

Ce-139 Photons E=30-250 keV Nb-95 Electrons E>15 keV Sr-90 Electrons E>15 keV

Ce-141 Electrons E>15 keV Ni-63 Electrons E>15 keV Sr-91 Electrons E>15 keV

Ce-143 Electrons E>15 keV Np-237 Alpha Ta-182 Photons E>250 keV

Ce-144 Electrons E>15 keV Np-239 Alpha Tb-160 Electrons E>15 keV

Cf-249 Alpha P-32 Electrons E>15 keV Sm-151 Electrons E>15 keV

Cf-252 Alpha P-33 Electrons E>15 keV Tc-99 Electrons E>15 keV

Cl-36 Electrons E>15 keV Pa-231 Alpha Te-131 Photons E>250 keV

Cm-242 Alpha Pa-233 Alpha Te-131M Electrons E>15 keV

Cm-243 Alpha Pa-234 Alpha Th-228 Alpha

Cm-244 Alpha Pb-210 Alpha Th-230 Alpha

Co-57 Photons E=30-250 keV Pm-147 Electrons E>15 keV Th-232 Alpha

Co-58 Photons E>250 keV Po-208 Alpha Th-234 Alpha

Co-60 Photons E>250keV Po-209 Alpha Tl-202 Photons E>250 keV

Cr-51 Photons E>250 keV Po-210 Alpha Tl-204 Electrons E>15 keV

Cs-134 Electrons E>15 keV Pr-143 Photons E>250 keV Tm-170 Electrons E>15 keV

Cs-137 Electrons E>15 keV Pr-147 Electrons E>15 keV U-232 Alpha

Eu-152 Photons E>250 keV Pu-236 Alpha U-234 Alpha

Eu-154 Electrons E>15 keV Pu-238 Alpha U-235 Alpha

Eu-155 Electrons E>15 keV Pu-239 Alpha U-236 Alpha

Eu-156 Electrons E>15 keV Pu-240 Alpha U-238 Alpha

Fe-55 Photons E<30 keV Pu-241 Alpha U-239 Alpha

Fe-59 Electrons E>15 keV Pu-242 Alpha Y-88 Photons E>250 keV

H-3 Electrons E<15 keV Ra-220 Alpha Y-90 Electrons E>15 keV

Hf-181 Electrons E>15 keV Ra-223 Alpha Y-91 Electrons E>15 keV

I-125 Photons E<30 keV Ra-224 Alpha Yb-169 Photons E>250 keV

I-129 Photons E=30-250 keV Ra-226 Alpha Zn-65 Photons E>250 keV

I-131 Electrons E>15 keV Ra-228 Alpha

* The radiation effectiveness factors (REF) for electrons > 15 keV and photons > 250 keV 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.