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External Dose Reconstruction

ORAUT-OTIB-0088 Rev. 01

Effective Date: 10/21/2019

Supersedes: Revision 00 

Subject Expert(s): Mark R. Fishburn and Matthew H. Smith

Document Owner 

Approval: Signature on File

Scott R. Siebert, Objective 3 Manager

Approval Date: 10/17/2019

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 10/17/2019

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 10/21/2019

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 2 of 26

 

PUBLICATION RECORD

EFFECTIVE

DATE

REVISION

NUMBER DESCRIPTION

09/14/2018 00 New document initiated to convert ORAUT-PROC-0006, External 

Dose Reconstruction, to a technical information bulletin. Included 

information on assignment of onsite ambient dose to facilitate 

cancellation of ORAUT-PROC-0060. Updated to current dose 

reconstruction approaches. Incorporated responses to the Advisory 

Board on Radiation and Worker Health Subcommittee on Procedures 

Reviews comments about ICD-9 codes. Incorporates formal internal 

and NIOSH review comments. Training is required. Initiated by 

Mark R. Fishburn.

10/21/2019 01 Revision initiated to correct error in Attachment A, Method for 

Assigning Missed Dose when the Number of Null Results Is 

Unknown, and include a new attachment to provide dates when U.S. 

Department of Energy sites incorporated the recommendations of 

ICRP Publication 60 for neutron weighting factors. Incorporates

formal internal and NIOSH review comments. Constitutes a total 

rewrite of the document. Training is required. Initiated by Mark R. 

Fishburn. 
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TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations.................................................................................................................. 4

1.0 Introduction ................................................................................................................................. 5

1.1 Purpose........................................................................................................................... 5

1.2 Scope.............................................................................................................................. 5

2.0 General Approach....................................................................................................................... 5

2.1 Records and Reconstruction Methods Pertaining to External Radiation 

Exposure......................................................................................................................... 8

2.1.1 External Dosimetry Records............................................................................. 8

2.1.1.1 Worker Was Monitored Adequately............................................... 9

2.1.1.2 Worker Was Not Monitored ........................................................... 9

2.1.1.3 Worker Monitored Inadequately ..................................................10

2.1.2 Occupational Medical X-Ray Dose................................................................. 11

2.1.3 Incident Investigation Reports ........................................................................11

2.1.4 Other Monitoring Records ..............................................................................12

2.2 External Radiation Exposure Types..............................................................................12

2.2.1 Photon (Gamma and X-Ray) Radiation..........................................................12

2.2.2 Neutron Radiation ..........................................................................................14

2.2.3 Electron (Beta) Radiation ...............................................................................16

2.3 Conversion of Reconstructed External Dose to Organ Dose........................................17

2.4 Uncertainty ....................................................................................................................18

References ...........................................................................................................................................19

ATTACHMENT A METHOD FOR ASSIGNING MISSED DOSE WHEN THE NUMBER OF 

NULL RESULTS IS UNKNOWN .......................................................................21

ATTACHMENT B ONSITE AMBIENT DOSE ASSIGNMENT ........................................................23

ATTACHMENT C DOE ADOPTION OF INTERNATIONAL COMMISSION ON 

RADIOLOGICAL PROTECTION (ICRP) PUBLICATION 60.............................26

LIST OF TABLES

TABLE TITLE PAGE

2-1 Hierarchy for accuracy of data sources for external dose reconstruction ................................... 9

2-2 Neutron energy intervals and associated ICRP Publication 60 weighting factors with 

examples of relevant exposures or facilities .............................................................................15

C-1 Site ICRP Publication 60 incorporation dates ...........................................................................26

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Overview of dose reconstruction approaches............................................................................. 7
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ACRONYMS AND ABBREVIATIONS 

AWE atomic weapons employer

CFR Code of Federal Regulations

DCF dose conversion factor

DOE U.S. Department of Energy

DOL U.S. Department of Labor

EALER elevated ambient levels of external radiation

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

GM geometric mean

GSD geometric standard deviation

H*(10) ambient dose equivalent at 10 millimeters depth in tissue

Hp(d) personal dose equivalent at depth d in millimeters in tissue

hr hour

HTML hypertext markup language

ICD International Classification of Diseases

ICRP International Commission on Radiological Protection

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt (1,000 electron-volts)

LOD limit of detection

MeV megaelectron-volt (1 million electron-volts)

mrem millirem

mSv millisievert

NIOSH National Institute for Occupational Safety and Health

ORAUT Oak Ridge Associated Universities Team

ORNL Oak Ridge National Laboratory

POC probability of causation

SRDB Ref ID Site Research Database Reference Identification (number)

TIB technical information bulletin

TLD thermoluminescent dosimeter

U.S.C. United States Code

wk week

wR radiation weighting factor

yr year

§ 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 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 document is to outline external dose reconstruction under Part B of the Energy 

Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA). This document

incorporates direction from NIOSH on the performance of external dose reconstructions as contained 

in OCAS-IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 2007), TIBs, 

approved site profiles, and Oak Ridge Associated Universities Team (ORAUT) procedures.

1.2 SCOPE 

Reconstruction of external dose could involve the calculation and assignment of dose from multiple 

sources (e.g., photons, neutrons, electrons, ambient dose, X-ray, etc.). This document provides 

information about the calculation of external dose and how it should be assigned under EEOICPA.

2.0 GENERAL APPROACH

External dose reconstruction is just one part of the overall dose reconstruction process. Under 

EEOICPA and Executive Order 13179, dose reconstructions are performed using information from the 

U.S. Department of Labor (DOL), DOE, atomic weapons employers (AWEs), and claimants while 

applying the science-based knowledge and practices of dose reconstruction.

During external dose reconstruction, a dose reconstructor assesses the circumstances of exposure, 

the completeness of the supporting data in relation to the employment period, and other information 

relevant to the dose reconstruction approach. The site profiles are reviewed, as necessary, to assist 

the dose reconstructor. Much of the relevant information from site profiles is captured in site-specific 

calculation tools (or generic calculation tools for cases in which site-specific information is unavailable 

or unnecessary), and the dose reconstructor uses this information to assist in the performance of 

external dose reconstruction.

To provide for efficiency in processing claims, overestimating assumptions can be used if a claim is 

initially deemed likely noncompensable. Underestimating assumptions can be used if a claim is 

initially deemed likely compensable. These processing options, which are typically available in the 

calculation tools, are described in ORAUT-PROC-0106, Roadmap to Reconstructing Dose (ORAUT 

2019c). A best-estimate analysis uses all of the available information and analytical processing 

capabilities to assign the most accurate dose possible for a claim. This approach is typically taken 

when a claim is close to the compensability probability of causation (POC). Best-estimate analyses of 

external doses are made using detailed knowledge of specific work locations, job responsibilities, and 

other claim-specific information, as well as the Vose capability process in the site tools to run Monte 

Carlo calculations in order to propagate dose uncertainty. Often a claim determination can be 
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achieved with less detailed knowledge by using assumptions that are favorable to the claimant (i.e., 

overestimate or underestimate). The dose reconstruction approach is presented in Figure 2-1. This 

figure illustrates the manner in which progressively more information is analyzed to complete the 

external dose reconstruction.
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Figure 2-1. Overview of dose reconstruction approaches (DR = dose reconstruction, SP = Site 

Profile, TBD = Technical Basis Document). 

Receive Claim

Review the following:

• Employment, facilities, 

duration

• Cancer type(s)

• Job title (in NOCTS)

• Monitoring data (if available)

- External – dosimetry/

occupational health 

screening X-rays

- Internal – bioassay

• Telephone interviews (if 

available)

- Job duties

- Work locations

- Special information 

(incidents/accidents, visits to 

work sites other than those 

already verified by DOL)

• DOL initial case file

Select initial/subsequent 

methodology of DR approach: 

overestimate, underestimate, 

or best estimate. (NOTE: 

Typically initial assessments are 

overestimates or 

underestimates)

OVERESTIMATE UNDERESTIMATE

External

Combinations of 

overestimate and 

best estimate of 

measured, missed, 

coworker, ambient, 

medical X-rays, 

incidents/accidents

Internal

Combinations of 

overestimate and 

best estimate of 

measured, missed, 

coworker, 

environmental 

internal, incidents/

accidents

Is estimated POC > 

45%?

Yes

DR complete

Revise 

individual dose 

components to 

best estimates 

until estimated 

POC is 

<45% 

or all dose 

components 

are assessed as 

best estimates

All dose components are 

applied as best 

estimates from the 

outset of dose 

reconstruction. 

Exceptions are only 

permitted if 

documented in SP/TBD 

DR complete

Yes

BEST 

ESTIMATE

No

DR complete

External

Combinations of 

underestimate and 

best estimate of 

measured, missed, 

ambient, medical 

X-rays, incidents/

accidents

Internal

Combinations of 

underestimate and 

best estimate of 

measured, missed, 

coworker, 

environmental 

internal, incidents/

accidents

Sum all calculated 

external and 

internal doses in 

IREP

Estimate POC

Revise individual 

dose components to 

best estimates or 

include additional 

dose components 

until estimated POC 

is ≥ 52% or all dose 

components are 

assessed as best 

estimates

Is estimated POC > 

52%?

Sum all calculated 

external and 

internal doses in 

IREP

Estimate POC

No
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As detailed in OCAS-IG-001, dose reconstruction takes into account only the dose the worker 

received at a covered facility (NIOSH 2007). The dose is sorted by radiation type and energy and is 

reconstructed for a specific target organ, which is determined from the worker’s DOL-verified primary 

cancer (see ORAUT-OTIB-0005, Internal Dosimetry Organ, External Dosimetry Organ, and IREP 

Model Selection by ICD-10 Code (ORAUT 2019a). Only doses that were received before the 

diagnosis of the primary cancer are included in the dose reconstruction. If multiple primary cancers 

have been verified, doses from covered employment are reconstructed for each associated target 

organ until the diagnosis date of the primary cancer. If the worker received medical X-ray 

examinations for occupational health screening and as a condition of employment at a covered site, 

dose reconstruction includes those doses. The following sections summarize key information from 

OCAS-IG-001 about the external dose reconstruction process.

The dose reconstructor ensures that the records supporting the dose reconstruction are included in 

the submittal. The following records are the minimum that should be included in the submittal: 

• Worksheets necessary to recalculate final input parameters for the Interactive 

RadioEpidemiological Program (IREP), 

• IREP Input Spreadsheet, 

• IREP HTML Summary Report of estimated POC, and

• Dose reconstruction report.

2.1 RECORDS AND RECONSTRUCTION METHODS PERTAINING TO EXTERNAL 

RADIATION EXPOSURE

Potentially available radiation monitoring records typically consist of external dosimetry records, 

internal dosimetry records, diagnostic X-ray records, incident investigation reports, and other 

monitoring records. Except for internal dosimetry records, these potentially affect the external dose 

reconstruction.

2.1.1 External Dosimetry Records

Table 2-1 lists the general hierarchy of data sources dose reconstructors should employ for external 

dose reconstruction. Personal dosimeter results are typically the most accurate assessment of a 

worker’s dose. In general, recorded dose data from a personal dosimeter should be used whenever 

possible, and these data are given priority over personal monitors, survey data, or source term data. 

The adequacy and completeness of the dosimeter data should be described in the site profile in terms 

of potential limitations in technology, calibration, workplace radiation fields, and administrative 

practices.

Personal dosimeter records typically include the worker’s entire external dose history at each covered 

facility. These records can consist of annual doses or the individual dosimeter processing results for 

each dosimeter assigned to the worker, or both. If only annual doses are available, the required 

approach is described in Attachment A and is favorable to claimants.

In relation to the availability of personal monitoring data for each claim, three conditions can apply, 

each requiring a different approach to reconstruction of external measured, missed, and unmonitored 

doses, as described below.
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Table 2-1. Hierarchy for accuracy of data sources for external dose reconstruction.

Hierarchy Data source Examples

1 Personal dosimeter Film badge; nuclear track emulsion, type A; thermoluminescent dosimeter

(TLD).

2 Personal monitors Pocket ionization chambers.

3 Coworker data Film badge, TLD, pocket ionization chambers.

4 Area monitoring Workplace radiation surveys, ambient air room monitors, duration of 

exposure.

5 Source term Source nuclide, activity, exposure rate, distance from source, duration of 

exposure, and shielding information.

6 Radiation control 

limits

Generally, workplace posting has been required when the dose rate 

exceeded 0.025 mSv/hr.

2.1.1.1 Worker Was Monitored Adequately

Workers are monitored adequately when the site has a robust dosimetry program. A robust dosimetry 

program involves a demonstrated quality control and quality assurance program for dosimetry 

procedures and policies in terms of who is assigned a dosimeter, the frequency of the monitoring, 

investigation of anomalous results, investigation of missing results, and retrievable records. The use 

of dosimetry should be reviewed for the site across the history of the site and is discussed in the site 

profile.

In general, external monitoring data that has been collected since the implementation of 10 CFR

Part 835, “Occupational Radiation Protection” (issued in 1993 and implemented in the following few 

years), are considered adequate. Monitoring in the “835” era is generally considered adequate 

because the regulations and implementing standards and guides for 10 CFR Part 835 established a 

consistent DOE-complex wide program in line with what NIOSH considers a robust dosimetry 

program.

Accreditation of a site dosimetry program in accordance with the DOE Laboratory Accreditation 

Program or the National Voluntary Laboratory Accreditation Program indicates the site program has 

been reviewed for and has demonstrated quality control and quality assurance for the measurement 

of dose by the dosimeters. In these situations, the dosimetry data can usually be used as the basis 

for external radiation dose reconstruction (i.e., to compute the annual dose for each year of covered 

employment); that is, no adjustments are required other than incorporation of missed dose and 

conversion to organ dose.

For adequately monitored workers, the associated uncertainty should be assumed to be normally 

distributed unless indicated otherwise in the site profile. For overestimate or underestimate 

reconstructions, the dose distribution can be considered a constant (point estimate), which results in 

an IREP input of only one parameter. For conservatism, a multiplicative factor may be applied to the 

assigned dose in lieu of uncertainty for an overestimate and not accounted for in an underestimate.

2.1.1.2 Worker Was Not Monitored

Dose reconstructors determine if the worker should have been monitored. Some unmonitored 

workers were not exposed to occupational external dose, but would receive ambient dose. Other 

unmonitored workers might have been exposed, and dose needs to be assigned by evaluating other 

sources of dose data (see the Table 2-1 data sources).

Most AWE workers and some DOE workers were not individually monitored for external radiation 

exposure using assigned personal dosimeters, including some who would have been classified as 

radiation workers by today’s radiation protection standards. Similarly, less-exposed workers who 

were not expected to exceed a significant percentage (e.g., 10% to 30%) of radiation protection 
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standards lacked monitoring; a practice that continues at some sites today and is consistent with 

current regulations. If there was a significant potential for radiation exposure, recorded doses for 

monitored coworkers may be used as surrogates (see ORAUT-OTIB-0020, Use of Coworker 

Dosimetry Data for External Dose Assignment; ORAUT 2011a). However, in the absence of such 

information, workplace radiation measurements (e.g., area monitoring data) could exist that can be 

used to estimate dose. For workers with no significant exposure potential, external radiation dose 

reconstruction is based on ambient dose (see Attachment B). In general, it is expected that 

reconstructed dose to unmonitored workers will be less than dose to monitored workers.

At some facilities, only a small sample of the workforce, or only workers who met certain criteria, were 

monitored to ensure compliance with radiation exposure limits. As an example, although construction 

workers were often unmonitored, it is possible in some instances to use data from representative 

coworkers who received similar exposures, such as radiological control technicians who monitored 

the work activities, to estimate a realistic maximum external dose. When assigning dose to 

construction trade workers, use the guidance in ORAUT-OTIB-0052, Parameters to Consider When 

Processing Claims for Construction Trade Workers (ORAUT 2014).

When no radiation monitoring data are available for a worker or representative coworkers, 

scientifically reasonable estimates of exposure should be developed based on survey data or the 

source term or quantity of radioactive material at the facility. Often a claim determination can be 

achieved with less detailed knowledge by using assumptions that are favorable to the claimant. It 

should be recognized that dose reconstructions based on survey data is probably biased because 

monitoring data tended to be recorded at the highest level to ensure compliance; however, this is an 

acceptable bias in this compensation program. If no survey data are available, the dose should be 

estimated based on the activity of the source term, engineering and administrative controls, and work 

history at the same facility. 

2.1.1.3 Worker Monitored Inadequately

Early workers at DOE and AWE facilities were typically monitored inadequately in comparison with 

current practices. This was due to less capable dosimetry technology, higher control limits, more 

frequent dosimeter exchanges, or monitoring records that were incomplete, missing, or unclear (or 

any combination thereof). Often, routine monitoring of worker neutron exposure was not performed in 

the 1940s and 1950s at some facilities. 

Monitoring data from before 1960 must be evaluated cautiously due to technological shortcomings 

and because monitoring programs were designed to ensure compliance with historically higher 

radiation safety limits. In such cases, external radiation dose reconstruction is based on available 

dosimetry data in combination with available site profile information about site processes, 

radioactivity, radiation fields, and other pertinent information. 

When workers were monitored inadequately, it might have been due to incomplete, missing, or 

unclear monitoring records. In these cases external dose reconstruction must consider options to 

estimate all or a portion of a worker’s dose and the associated uncertainty.

If there are sufficient monitoring records before and after the missing records, the dose to assign can 

be interpolated. The interpolation, or gap fill (which could be a simple average between the 

monitoring periods), is considered reasonable if the work practices, radiological protection measures, 

and administrative and engineering controls did not change. Gap fill can be conducted only if there is 

no indication from the claimant, worker, site radiological, or DOL records that a radiological incident 

resulting in a higher exposure occurred during the period of missing records or that the worker’s job 

duties or the exposure potential changed during the gap period. All documents associated with the 

claim should be reviewed to determine there were no significant changes in the worker’s job duties or 
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exposure. Examples of information that could indicate significant changes in a worker’s job duties or 

exposure include:

• Information (e.g., from telephone interviews, DOL records) that the worker changed jobs 

during the period in question.

• Significant recorded dose changes during the covered periods surrounding the missing 

dosimeter records.

• Information about incidents occurring during the period in question.

The length of the gap fill period is dependent on the surrounding dosimetry exchange frequency, 

whether it was weekly, monthly, quarterly, or semiannually. The gap period must be bounded on both 

ends with reported dosimetry information. This information should only be used to address a gap 

equal to an exchange period that is bounded on both sides by provided dosimetry data. An example 

is a worker with first- and third-quarter dosimeter results, but no dosimeter results for the second 

quarter. If the claim information (e.g., the worker site records, telephone interview) does not indicate 

a change in employment and there is no indication that a higher exposure occurred during the period 

of missing records (as discussed in the previous paragraph), then dose may be assigned for the 

second quarter that aligns with the assigned dose for the first and third quarters.

Site-specific guidance about dosimetry practices should be reviewed because some site guidance 

documents are more prescriptive in their methods for gap fill. In addition, the dose reconstructor 

should evaluate all dosimetry records to ensure that gap filling aligns with those records as 

appropriate (e.g., if an annual result is available, the amount of recorded and gap fill dose must not 

exceed the annual dose). 

The assignment of dose using adjacent dosimetry information as described above should be the first 

method for assigning dose for a missing dosimetry exchange period. If monitoring data is not 

available, coworker data might provide a good option in cases involving missing or incomplete 

records. For records that are unclear, dose reconstruction management should be notified so an 

acceptable approach can be determined, which could involve requesting additional information from 

DOE.

2.1.2 Occupational Medical X-Ray Dose

Records from medical X-ray examinations that identify the date and clinical summary for each 

examination are often available. The site profile is where information about X-ray frequency and type 

is provided and should be the initial document to review when evaluating site X-rays. Dose 

reconstructors should use the approach in ORAUT-OTIB-0079, Guidance on Assigning Occupational 

X-Ray Dose Under EEOICPA for X-Rays Administered Off Site (ORAUT 2017b), and ORAUT-OTIB0006, Dose Reconstruction from Occupational Medical X-Ray Procedures (ORAUT 2019b) to 

determine if offsite examinations are covered and to then assign dose. 

2.1.3 Incident Investigation Reports

Reports about incidents that might have involved the worker are often provided in the DOE records, 

either in separate files or embedded in the external or internal dosimetry records. The dose 

reconstructor should request records if they might exist but are not available in the file (e.g., on 

reading the claimant interview in which a reportable incident involving external dose is described). 

The dose reconstructor needs to determine if the dose from the incident was captured by the 

dosimetry records of the worker and is already included in the dose reconstruction. If not, the dose 
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reconstructor needs to assign the dose from the incident (which might involve a calculation to 

determine the dose to assign) in addition to other dose the worker is assigned.

2.1.4 Other Monitoring Records

Other monitoring records (e.g., area dosimetry results, survey results) that potentially pertain to 

reconstruction of external dose might be available in the DOE files. The dose reconstructor needs to 

determine if the dose from other monitoring records needs to be included because dosimetry 

monitoring was not performed or was inadequate. It is not unusual to have monitoring for one type of 

radiation (e.g., photon) and not be adequately monitored for another radiation type (e.g., neutrons or 

low energy photons). If monitoring was not performed or was inadequate, the dose reconstructor 

needs to assign dose in addition to other dose the worker is assigned. An example of this approach 

that used survey data to calculate and assign neutron dose is available in Section 6.4 of ORAUTTKBS-0006-6, Hanford Site – Occupational External Dose (ORAUT 2010).

2.2 EXTERNAL RADIATION EXPOSURE TYPES

To arrive at an overall estimate of worker dose and uncertainty, dose reconstructions generally 

address the potential for recorded external dose, internal dose, missed dose, occupational medical 

exposures, and onsite ambient radiation exposures by applying assumptions that are favorable to the 

claimant, as necessary, to minimize the likelihood that dose is underestimated. External radiation 

includes three types of radiation, as described below.

2.2.1 Photon (Gamma and X-Ray) Radiation

There are four basic components of photon radiation dose. The sum of the dose components in each 

calendar year makes up a worker’s annual occupational photon radiation dose Dγ. This is expressed 

as follows:

=++ + D M OM EN (2-1) Dγ DDD D

where

Dγ = total annual photon radiation dose

DD = recorded worker occupational dose typically based on personal dosimeter 

measurements

DM = missed dose (unrecorded or unmeasured due to dosimeter limitations)

DOM = occupational medical monitoring X-ray examination dose

DEN = environmental (onsite ambient) dose, which might or might not have been included in 

the dosimeter dose the site reported

In addition to the calculation above, unmonitored photon dose can be assigned if a worker was not 

monitored for radiation exposure and the dose reconstructor determines that doses exceeding DEN

might have been received. Unmonitored photon doses can be estimated based on adjacent 

monitoring records, coworker doses, applicable dose limits, or area measurement data.

Recorded photon doses DD are adjusted, if appropriate, based on site profile information. If a relevant 

site profile does not exist for a particular case, dose reconstructors should consult ORAUT-OTIB0008, A Standard Methodology for Overestimating External Doses Measured with Thermoluminescent 

Dosimeters (ORAUT 2006a), or ORAUT-OTIB-0010, A Standard Complex-Wide Methodology for 

Overestimating External Doses Measured with Film Badge Dosimeters (ORAUT 2006b), depending 

on whether the measurements were from TLDs or film badges, respectively.
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As discussed in OCAS-IG-001, the prescribed method for evaluating missed photon dose DM is to 

assign a dose equal to the limit of detection (LOD) divided by 2 for each dosimetry measurement that 

is recorded as zero. However, for cases in which multiple badges were issued for a particular 

monitoring period, only one zero measurement should be assigned per monitoring period. These 

doses are then summed for a given year. The LOD/2 method results in a slightly positive bias 

(overestimate) of the true dose in most cases (NIOSH 2007).

Missed doses calculated in this manner have a lognormal distribution for the purpose of calculating 

POC. Specifically, the photon LOD/2 times the number of zero monitoring badges is the central 

estimate or geometric mean (GM) of a lognormal distribution, and the upper 95th-percentile estimate 

is the LOD times the number of zero monitoring badges, which equates to a geometric standard 

deviation (GSD) of 1.52 (NIOSH 2007). As stated in OCAS-IG-001, this approach is for the 

calculation of missed doses and should not be applied to non-missed dose situations.

In addition, OCAS-IG-001 provides guidance on how to assign dose from nonzero dosimetry results 

that are less than the LOD. Dose should be assigned equal to the LOD divided by 2 for each 

dosimetry measurement (i.e., film badge, pocket ionization chamber, or TLD) that is recorded as zero 

or if it is below the LOD divided by 2. Readings greater than or equal to the LOD divided by 2 should 

be used as recorded.

When the number of zero measurements cannot be determined, determining the missed dose 

becomes more complicated. When the records provide only an annual dose, the number of zero 

doses should be estimated based on that dose, the monthly, quarterly, or annual limits for that year, 

and the maximum number of possible zero monitoring intervals (see Attachment A).

External ambient doses DEN are determined using Attachment B of this document. Occupational 

medical X-ray doses DOM are determined using ORAUT-PROC-0061, Occupational Medical X-Ray 

Dose Reconstruction (ORAUT 2017a) and ORAUT-OTIB-0006 (ORAUT 2019b). 

As described in OCAS-IG-001, photon doses are categorized into three input categories for IREP as 

follows (NIOSH 2007):

• <30 keV,

• 30 to 250 keV, and

• >250 keV.

Considering both the organ dose conversion factor (DCF) and the risk associated with photon dose as 

given in IREP, the 30- to 250-keV energy range is typically the photon energy most favorable to the 

claimant if a dose reconstructor is faced with (1) an unknown energy distribution or (2) a desire to 

process a likely noncompensable case using an overestimating approach. One notable exception is 

skin dose from work with or near isotopes of plutonium because the DCF for skin is considered to be 

1 for all energies (ORAUT-OTIB-0017, Interpretation of Dosimetry Data For Assignment of Shallow 

Dose; ORAUT 2005) and because <30-keV photons have a higher risk factor in IREP than do 30- to 

250-keV photons based on information provided in NIOSH-Interactive RadioEpidemiological Program 

(NIOSH-IREP) Technical Documentation (NIOSH 2002). A photon energy division of 25% 30 to 

250 keV and 75% >250 keV can be assumed to be a reasonable minimum approach to process a 

likely compensable case because radiation scattering reasonably precludes the existence of 

exclusively high-energy photons in a workplace environment. Unless there is information to the 

contrary, external ambient doses DEN and medical X-ray doses DOM are always classified as 30 to 250

keV photons as favorable to claimants (Attachment B and ORAUT-PROC-0061; ORAUT 2017a). 

In accordance with OCAS-IG-001, all external photon doses may be assigned as acute or chronic

using parameters of NIOSH-Interactive RadioEpidemiological Program (NIOSH-IREP) Technical 
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Documentation (NIOSH 2002). However, they are typically assigned as acute with the exception for 

external onsite ambient doses, which are always considered chronic (NIOSH 2007).

2.2.2 Neutron Radiation

There are two basic components of neutron radiation dose. The sum of the dose components in each 

calendar year comprises a worker’s annual occupational neutron radiation dose DN. This is 

expressed as follows:

= + (2-2) DDD NDM

where

DN = total annual neutron radiation dose

DD = recorded worker occupational neutron dose, typically based on personal neutron 

dosimeter measurements

DM = missed neutron dose (unrecorded or unmeasured neutron dose due to dosimeter 

limitations)

Because neutron exposures from manmade sources do not generally exist in the environment and 

because neutrons are not used in most diagnostic or occupational medical procedures, ambient and 

medical dose categories are not included in the external neutron dose reconstruction.

Neutron monitoring may be present, but because of neutron energy distribution, dosimeters (e.g., film 

dosimeters) may not capture a portion of the energy spectrum and therefore the individual is not fully 

monitored. In these situations, an adjustment to the recorded dose (e.g., N:P ratio or applying a 

correction factor) would need to be used. 

As is the case with photons, unmonitored neutron dose can be assigned if a worker was not 

monitored or was monitored inadequately for neutron exposure. Unmonitored neutron doses are 

typically determined based on neutron-to-photon ratios from the site profiles or using a quantile 

regression approach. However, many workers would not have received a significant neutron dose, 

and evaluation of the neutron dose component might be unnecessary in such cases (see ORAUTOTIB-0023, Assignment of Missed Neutron Doses Based on Dosimeter Records; ORAUT 2008). 

Therefore, photon dose can often be the only type of external radiation dose that requires evaluation. 

Recorded neutron doses DD are adjusted, if appropriate, based on site profile information. When 

using film dosimetry, track fading and angular dependence might be an issue. If needed, the site 

profile should provide guidance. The neutron dose for each energy category is adjusted to the 

International Commission on Radiological Protection (ICRP) Publication 60 weighting factors (ICRP 

1991) as generally described in OCAS-IG-001 (NIOSH 2007). For example, since the 1950s a quality 

factor of 10 has generally been applied to fast neutron exposures; however, it has varied from 5 to 20 

across facilities and times. Attachment C lists numerous sites and the date when they implemented

ICRP Publication 60 neutron weighting factors. Contact the Principal External Dosimetrist if 

information about the implementation of the ICRP Publication 60 values is needed for an unlisted site. 

Table 2-2 lists the ICRP Publication 60 neutron radiation weighting factors for specific energy ranges.

Neutron monitoring was not fully implemented until the late 1950s or was generally inadequate. As a 

result, missed or unmonitored neutron doses DM have the potential to contribute significantly to the 

annual occupational dose, especially in the early years of the DOE weapons complex.

If the monitoring data and methods were considered adequate according to site profile information, a 

neutron missed dose should be evaluated using the same method discussed for photons. 
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 15 of 26

 

Specifically, the neutron LOD/2 times the number of zero monitoring badges is the central estimate of 

a lognormal distribution, and the upper 95th-percentile estimate is the LOD times the number of zero 

monitoring badges. However, a zero in the records for neutron dose does not necessarily mean that 

there was a potential for neutron exposure, so exceptions apply (see ORAUT-OTIB-0023; ORAUT 

2008). Use the site profile information to reconstruct unmonitored neutron dose for workers who had 

unmonitored exposures. This dose is generally reconstructed using neutron-to-photon ratios based 

on the reconstructed measured and missed photon doses. An additional method involving quantile 

regression, described in ORAUT-RPRT-0087, Applications of Regression in External Dose 

Reconstruction (ORAUT 2018), can also be used, which uses monitoring results from dosimeters that 

are sensitive to both photons and neutrons and models that relationship to determine possible neutron 

dose when only a photon result is available.

Table 2-2. Neutron energy intervals and associated ICRP Publication 60 weighting factors with 

examples of relevant exposures or facilities. 

Neutron 

energy 

(MeV)

Publication 60 radiation 

weighting factor 

(wR)

a Typical exposure scenario

<0.01 5 Low-energy neutron exposures include thermal neutrons commonly 

found around nuclear reactors or moderated neutron sources. More 

prevalent around heavy-water reactors.

0.01–0.10 10 Intermediate-energy neutron exposures can also result from operation 

around nuclear reactors as high-energy neutrons are moderated to 

thermal energies.

0.10–2.00 20 Commonly called fission spectrum neutrons, this is the most typical 

energy range from operation of light-water or graphite-moderated 

reactors.

2.0–20.0 10 Reactions between alpha particles from materials such as plutonium 

or polonium and light materials such as beryllium can result in the 

production of neutrons. These reactions are commonly called alphaneutron (α,n) reactions. This neutron energy interval also includes 14 

MeV neutrons from fusion reactions.

>20.0 5 Exposures to neutrons greater than 20 MeV can result from work 

around accelerators.

a. Source: ICRP (1991).

As described in OCAS-IG-001, neutron doses are categorized into five IREP input categories as 

follows (NIOSH 2007):

• <10 keV,

• 10 to 100 keV,

• 0.1 to 2 MeV,

• 2 to 20 MeV, and

• >20 MeV.

Considering the neutron weighting factor adjustment, organ DCF, and the risk associated with neutron 

dose as give in IREP, the 0.1 to 2 MeV energy range is typically the most favorable to the claimant, 

but the dose reconstructor should use the guidance in the site profile for neutron dose distribution.

In accordance with OCAS-IG-001, all neutron doses should be entered as chronic in IREP

(NIOSH 2007).
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2.2.3 Electron (Beta) Radiation

In general, external electron radiation dose is significant only for exposures to the surface skin tissue 

of the body. The exposure to skin can originate from an unshielded electron source, such as 90Sr/90Y 

or uranium decay products, or from skin contamination with beta/gamma emitters. Other organs for 

which external electron exposure is relevant include the breast, testes, and lip; see ORAUT-OTIB0017, Interpretation of Dosimetry Data for Assignment of Shallow Dose for detailed information 

(ORAUT 2005).

There are three basic components of skin radiation dose. The sum of these components in each 

calendar year is a worker’s annual occupational electron radiation dose DE. This is expressed as 

follows:

=++ (2-3) DDDD E DMS

where

DE = total annual electron radiation dose

DD = recorded worker occupational electron skin dose typically based on personal dosimeter 

measurements

DM = unrecorded or unmeasured electron dose commonly referred to as the missed electron 

or skin dose

DS = dose from skin contamination by beta/gamma-emitting nuclides; this dose poses a 

unique exposure scenario that should be evaluated in skin cancer cases

As is the case with photons, unmonitored electron dose can be assigned if a worker was not 

monitored or was monitored inadequately for electron exposure.

All components of electron dose (DD, DM, and DS) must be calculated based on the guidance in 

ORAUT-OTIB-0017 (ORAUT 2005). This involves an understanding of the open window result values 

(e.g., electron or low-energy photon), site dosimetry reporting schemes, limit of detection issues, 

dosimetry shielding by security credentials, and personal protective equipment. Recognizing these 

complicating factors and using the approach provided in ORAUT (2005) is necessary to determine the 

dose to assign. When calculating the electron dose, separate the nonpenetrating doses and 

determine if they result from >15 keV electrons (corrected for attenuation, if applicable) include any 

applicable missed dose. As discussed in ORAUT (2005), a DCF of 1 is assumed for determining skin 

dose.

As described in OCAS-IG-001, electron doses are categorized into two IREP input categories as 

follows (NIOSH 2007):

• ≤15 keV, and

• >15 keV.

However, only the >15 keV category is considered to be a source of external radiation. As described 

in OCAS-IG-001, Section 1.2, electrons ≤15 keV do not have sufficient energy to penetrate the 

epidermal layer of the skin and are therefore not considered an external radiation hazard (NIOSH 

2007). 

In accordance with OCAS-IG-001, external electron doses should be entered as acute in IREP 

(NIOSH 2007).
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 17 of 26

2.3 CONVERSION OF RECONSTRUCTED EXTERNAL DOSE TO ORGAN DOSE

For external dose reconstruction under EEOICPA, the organ or tissue that developed the cancer is 

the primary organ of interest. International Classification of Diseases (ICD) codes are used to 

determine general information regarding corresponding organ for which external dose should be 

calculated. Detailed information on ICD codes and the appropriate external organ is provided in 

ORAUT-OTIB-0005 (ORAUT 2019a). 

Film badges and TLDs were typically worn on the upper front torso of the worker’s body. Depending 

on the monitoring era, workplace radiation fields, and site, these devices were calibrated to a selected 

radiation quantity as follows:

• Exposure,1

1 Exposure is used to measure gamma and X-ray radiation interaction only with air and was historically measured in 

roentgens: 1 R = 2.58 × 10-4 coulomb per kilogram. It is a measure of the ionizations of the molecules in a mass of air. 

The quantity is easy to measure directly and was used historically to calibrate radiation protection instruments and 

dosimeters.

• Absorbed dose in air,

• Ambient dose equivalent, or

• Penetrating dose at a selected depth in tissue (i.e., similar to current personal dose equivalent,

Hp(d), where d = 0.07 millimeter for shallow dose and d = 10 millimeters for deep dose.

The precise radiation quantity a site historically used to measure and record dose to workers is 

difficult to evaluate retrospectively. Under most circumstances, it is known that calibration of early film 

dosimeters using comparatively high-energy radium or 137Cs gamma sources generally resulted in an 

overestimation of the dose to a worker in the comparatively lower-energy photon radiation fields 

typical of the workplace because of the film overresponse to lower-energy photons from shielding and 

scattering of the radiation. In view of these considerations, the corresponding uncertainties, and the 

desire to conduct consistent evaluations, organ DCFs are selected from OCAS-IG-001 based on the 

most favorable geometry (NIOSH 2007). For most organs this is an anterior-posterior exposure, but 

for some organs (i.e., lung, esophagus, red bone marrow, bone surfaces) both anterior-posterior and 

rotational geometries must be evaluated and the geometry that results in the higher POC should be 

selected. When evaluating claims with multiple cancers, to be consistent, whichever geometry is 

determined to have the highest overall POC should be applied to all cancers.

Certain exceptions exist to the use of the DCFs in OCAS-IG-001 (NIOSH 2007). For likely 

noncompensable cases, the calculation of measured doses involves rounding organ DCFs up to 1 

unless the DCF exceeds 1, in which case the actual DCF is applied. This practice was established 

(1) to help ensure favorability to claimants and (2) to avoid, if possible, reporting a reconstructed

organ dose less than the dose of record.

Other exceptions include: 

• Reconstruction of dose to the skin, where a DCF of 1 is assumed in accordance with ORAUTOTIB-0017 (ORAUT 2005),

• Reconstruction of external ambient dose (ambient dose equivalent [H*(10)] DCF in OCAS-IG001 (NIOSH 2007) for an isotropic exposure geometry as indicated in Attachment B), and

• Reconstruction of organ doses from occupational medical X-rays. Organ-specific doses are

presented in the respective site profiles as described in ORAUT-PROC-0061 (ORAUT 2017a).
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2.4 UNCERTAINTY

The uncertainties in the measured dosimeter dose and the occupational medical dose are assumed to 

have a normal distribution, while the uncertainties in the missed dose and the ambient onsite dose are 

assumed to have a lognormal distribution. The uncertainty in the organ DCF is assumed to have a 

triangular distribution as provided in NIOSH (2007). 

Assessment of the overall external dose uncertainty is dependent on the dose reconstruction 

approach. For likely compensable claims (underestimates), uncertainty might not need to be applied. 

For likely noncompensable claims (overestimates), the uncertainty can be incorporated by increasing 

the assigned dose components and assigning them as constants. For claims where the a priori 

compensability is unknown (best estimates), each dose component is treated as a distribution and 

Monte Carlo sampling of the distributions is employed to calculate the overall uncertainty of the dose 

estimate.

Attachment B and ORAUT (2018a) provide details on the assessment of uncertainties for external 

ambient and occupational medical X-ray doses, respectively.
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 19 of 26

 

REFERENCES 

BT (Battelle Team), 2007, Default Assumptions and Methods for Atomic Weapons Employer Dose 

Reconstructions, Battelle-TIB-5000, Rev. 00, Richland, Washington, April 2. [SRDB Ref ID: 

32016]

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England. [SRDB Ref ID: 11517]

NIOSH (National Institute for Occupational Safety and Health), 2002, NIOSH-Interactive 

RadioEpidemiological Program (NIOSH-IREP) Technical Documentation, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, June 18. [SRDB Ref ID: 22398]

NIOSH (National Institute for Occupational Safety and Health), 2007, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21. [SRDB Ref ID: 38864]

ORAUT (Oak Ridge Associated Universities Team), 2005, Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev. 01, Oak Ridge, Tennessee, 

October 11. [SRDB Ref ID: 19434]

ORAUT (Oak Ridge Associated Universities Team), 2006a, A Standard Methodology for 

Overestimating External Doses Measured with Thermoluminescent Dosimeters, ORAUTOTIB-0008, Rev. 01, Oak Ridge, Tennessee, May 12. [SRDB Ref ID: 29950]

ORAUT (Oak Ridge Associated Universities Team), 2006b, A Standard Complex-Wide Methodology 

for Overestimating External Doses Measured with Film Badge Dosimeters, ORAUT-OTIB0010, Rev. 01, Oak Ridge, Tennessee, June 5. [SRDB Ref ID: 29953]

ORAUT (Oak Ridge Associated Universities Team), 2008, Assignment of Missed Neutron Dose 

Based on Dosimeter Records, ORAUT-OTIB-0023, Rev. 01, Oak Ridge, Tennessee, May 14. 

[SRDB Ref ID: 43937]

ORAUT (Oak Ridge Associated Universities Team), 2010, Hanford Site – Occupational External 

Dose, ORAUT-TKBS-0006-6, Rev. 04, Oak Ridge, Tennessee, January 7. [SRDB Ref ID: 

77581]

ORAUT (Oak Ridge Associated Universities Team), 2011a, Use of Coworker Dosimetry Data for 

External Dose Assignment, ORAUT-OTIB-0020, Rev. 03, Oak Ridge, Tennessee, 

November 14. [SRDB Ref ID: 104029]

ORAUT (Oak Ridge Associated Universities Team), 2011b, DOE Adoption of 1990 ICRP 

Recommendations on Neutron Weighting Factors, Oak Ridge, Tennessee, December 20. 

[SRDB Ref ID: 106069]

ORAUT (Oak Ridge Associated Universities Team), 2014, Parameters to Consider When Processing 

Claims for Construction Trade Workers, ORAUT-OTIB-0052, Rev. 02, Oak Ridge, Tennessee, 

July 24. [SRDB Ref ID: 133862]

ORAUT (Oak Ridge Associated Universities Team), 2017a, Occupational Medical X-Ray Dose 

Reconstruction, ORAUT-PROC-0061, Rev. 04, Oak Ridge, Tennessee, June 6. [SRDB Ref 

ID: 166910]
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 20 of 26

 

ORAUT (Oak Ridge Associated Universities Team), 2017b, Guidance on Assigning Occupational XRay Dose Under EEOICPA for X-Rays Administered Off Site, ORAUT-OTIB-0079, Rev. 02, 

Oak Ridge, Tennessee, June 15. [SRDB Ref ID: 166967]

ORAUT (Oak Ridge Associated Universities Team), 2018, Applications of Regression in External 

Dose Reconstruction, ORAUT-RPRT-0087, Rev. 00, Oak Ridge, Tennessee, March 13. 

[SRDB Ref ID: 170100]

ORAUT (Oak Ridge Associated Universities Team), 2019a, Internal Dosimetry Organ, External 

Dosimetry Organ, and IREP Model Selection by ICD-10 Code, ORAUT-OTIB-0005, Rev. 06, 

Oak Ridge, Tennessee, September 27. [SRDB Ref ID: 178311] 

ORAUT (Oak Ridge Associated Universities Team), 2019b, Dose Reconstruction from Occupational 

Medical X-Ray Procedures, ORAUT-OTIB-0006, Rev. 06, Oak Ridge, Tennessee, September 

27. [SRDB Ref ID: 178310] 

ORAUT (Oak Ridge Associated Universities Team), 2019c, Roadmap to Reconstructing Dose,

ORAUT-PROC-0106, Rev. 01, Oak Ridge, Tennessee, June 7. [SRDB Ref ID: 176807]
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 21 of 26

 

ATTACHMENT A 

METHOD FOR ASSIGNING MISSED DOSE

WHEN THE NUMBER OF NULL RESULTS IS UNKNOWN

The basis for these instructions is Section 2.1.2.3 of OCAS-IG-001, External Dose Reconstruction 

Implementation Guideline (NIOSH 2007).

Overestimating Approach (Appropriate for Likely Noncompensable Cases)

Instruction: Calculate the maximum number of potential zeros Nmax, taking into consideration the 

published exchange frequency and applicable dose limits. This number multiplied by the LOD 

(i.e., Nmax × LOD) is considered the 95th percentile of a lognormal distribution. Multiply by LOD/2 

(i.e., Nmax × LOD/2) to determine the GM. These values are assumed to have a lognormal 

distribution with a GSD of 1.52.

Example: Assume the case and site profile information has provided a 1-rem reported annual dose, 

weekly exchanges, LOD of 40 mrem, 300-mrem/wk dose limit for the era, and a 52-wk work year: 

• The maximum number of zeros is 48 (e.g., 300 mrem/wk for 3 wk, 100 mrem in 4th wk, zero 

the rest).

• The missed dose assigned is 48 × 0.040 rem/2 = 0.960 rem (lognormal, 1.52 in IREP 

Parameter 2).

Underestimating Approach (Appropriate for Likely Compensable Cases)

Instruction: Calculate the median between the maximum and minimum number of potential zeros 

Nmed, taking into consideration the published exchange frequency, applicable dose limits, and LOD. 

When multiplied by LOD/2, Nmed × LOD/2 is the GM. This is considered an underestimating approach 

in comparison with the guidance in OCAS-IG-001 (NIOSH 2007). If the arithmetic mean is a partial 

number round up (can’t have a fraction of a dosimeter and rounding up is slightly claimant favorable). 

For example, the median between 27 and 48 dosimeters is 37.5 dosimeters. Use 38 dosimeters in 

the calculation.

Example: Using the same assumptions as above: 

• The maximum number of zeros is 48.

• The minimum number of zeros is 27 (e.g., 40 mrem in 25 wk, zero the rest).

• The median is 38. 

• The missed dose assigned is 38 × 0.040 rem/2 = 0.760 rem (lognormal, 1.52 in IREP 

Parameter 2). 

Best-Estimate Approach (Appropriate for Cases in Which Overestimating Approach Results in 

POC >45% and Underestimating Approach Results in POC <52%)

Instruction: Calculate the median (Nmed) between the maximum (Nmax) and minimum number of 

potential zeros, taking into consideration the published exchange frequency, applicable dose limits, 

and LOD. When multiplied by LOD/2 (i.e., Nmed × LOD/2), the resultant value is considered the GM of 

a lognormal distribution. When multiplied by the LOD (i.e., Nmax × LOD), the resultant value is 
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 22 of 26

ATTACHMENT A

METHOD FOR ASSIGNING MISSED DOSE

WHEN THE NUMBER OF NULL RESULTS IS UNKNOWN (continued)

considered the 95th percentile of a lognormal distribution, and the GSD must be calculated 

accordingly using the 95th- and 50th-percentile dose. The GSD is calculated as:

(A-1)

      

=    

1

95th percentile 1.64485

GSD

50th percentile

Example: Assumptions are the same as above.

• The maximum number of zeros is 48.

• The minimum number of zeros is 27.

• The median is 38. 

• The missed dose is 38 × 0.040/2 = 0.760 rem (lognormal). 

• 95th percentile dose is 48 × 0.040 = 1.920 rem; 

      

= =    

1

1.920 1.64485

GSD 1.757

0.760

• The missed dose assigned = 0.760 rem (calculated in previous bullet), lognormal, 1.757 in 

IREP parameter 2.
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 23 of 26

 

ATTACHMENT B 

ONSITE AMBIENT DOSE ASSIGNMENT 

This attachment provides guidance for the assignment of external onsite ambient dose. This 

attachment is intended to provide general information about assignment of this dose; site-specific 

guidance might be provided in the site profile.

B.1 GENERAL

As described in OCAS-IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 

2007), doses from elevated background radiation as a result of DOE or AWE activities must be 

included in dose reconstructions. This requirement is complicated by site reporting practices, fallout 

from atmospheric weapons testing, and worker location in relation to site monitoring data. Because 

these exposures are a concern for workers who were not monitored or who worked at a site where 

elevated background radiation from DOE or AWE activities might have been subtracted from 

dosimeter results, reconstruction of doses must rely on information in site-specific site profiles and 

other published health physics resources.

External film dosimeters or TLDs have been used for occupational radiation monitoring since the 

1940s. To account for background radiation levels that are not traditionally included in occupational 

radiation dose records, control dosimeters have been used from the outset. Good radiation protection 

practice dictates that during shipment a control dosimeter accompanies each batch of dosimeters 

issued to workers. Between manufacture or annealing and issuance, and between retrieval and 

processing, each shipment of dosimeters is irradiated by natural cosmic and terrestrial radiation 

sources and potentially inadvertently irradiated by other sources. The function of the control 

dosimeter is to measure all nonoccupational radiation exposure to the batch of dosimeters. When 

processed, the reading from the control dosimeter is subtracted from the reading of each of the other 

dosimeters in the batch, which yields a result for each dosimeter that is solely due to occupational 

radiation exposure. Note that the subtraction could occur with raw data, such as optical density 

readings for film or glow curves, or with transformed data, such as exposures in roentgens, absorbed 

doses in rad or grays, or dose equivalents in rem or sieverts.

Determination as to whether control dosimeters were exposed to elevated ambient levels of external 

radiation (EALER) is generally associated with where the control dosimeters were stored. From the 

intended use of control dosimeters, it is clear that controls should be subjected to exactly the same 

nonoccupational radiation exposure as the issued dosimeters and differ only in the occupational 

component. The implementation of this intention (that is, procedures for issuance and retrieval of 

dosimeters) likely differed over time at a given facility and certainly differed among DOE and AWE 

sites. For example, at large facilities, controls might have been kept at a central dosimeter location or 

distributed with batches of dosimeters to remote identification (ID) badge or dosimeter exchange 

buildings such as guard stations near reactors, reprocessing facilities, or manufacturing facilities. 

During some periods, dosimeters were incorporated into ID badges to ensure that no one entered 

without a dosimeter, and these ID badges were picked up at the entrance station at the beginning of 

each shift and returned there at the end of each shift. If control dosimeters were kept at remote 

exchange facilities, they would have recorded EALER at those facilities. Such doses from EALER 

recorded by the controls would subsequently have been subtracted from each worker’s dosimeter 

reading. However, if control dosimeters were kept at a distant central badging facility where ambient 

radiation levels were lower than in the work areas, each worker’s dosimeter would have recorded not 

only his or her occupational exposure, but also his or her exposure to EALER. In the latter case, no 

adjustment for occupational environmental radiation levels is needed because it would have been 

included in the worker’s occupational measurements.
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 24 of 26

ATTACHMENT B 

ONSITE AMBIENT DOSE ASSIGNMENT (continued)

For the early days of reactor operation, one important component of external environmental dose 

arises from submersion in, or irradiation at a distance from, a plume of 41Ar (with a radiological half-life 

of 1.83 hours), which formed when naturally occurring 40Ar nuclei absorbed neutrons near operating 

reactors. The emissions from 41Ar are primarily a 1.2-MeV (maximum) beta particle and a 1.3-MeV 

photon. A wooden badge exchange building would likely provide very little shielding or attenuation of 

the photons and, if air exchange rates at the control dosimeter storage point were high, even the beta 

component could have approached outdoor levels. In addition to radiation from airborne releases of 

radioactive materials, other components of EALER could arise from:

• Scattered radiation from waste trenches, storage facilities, etc.; 

• Terrestrial contamination; and

• Skyshine (radiation scattered to the ground from air over nuclear or high-energy accelerator 

facilities).

However, these components are unlikely to have been the same at dosimeter exchange facilities as 

they were on the rest of the site, so control dosimeters that were stored at remote exchange facilities 

would not have recorded this component.

Processes with potential for significant EALER include:

• Operating production reactors,

• Fuel reprocessing,

• Other radiochemical processing facilities,

• Atmospheric nuclear weapons testing,

• Underground nuclear weapons testing with significant venting of fission gasses,

• Accidental airborne releases of radioactive materials, and

• Certain high-energy accelerators (in the early years).

Large sites might have had inhomogeneous EALER, and control dosimeters that were distributed with 

batches intended for particular areas might have been able to measure significant EALER that the 

workers’ dosimeters might have missed.

In general, there is a point in time at a particular site after which there is no need to assess specific 

EALER to add to the worker’s dose because (1) levels were so low they would not significantly affect 

the POC, and (2) control dosimeters were kept under controlled conditions and EALER would not 

have been subtracted. As environmental monitoring programs matured, environmental TLD 

measurements ruled out significant EALER values. This point in time can only be established for a 

particular site by reviewing site and worker monitoring practices.

External dosimetry results account for both occupational and environmental penetrating radiation 

exposures if control dosimeters were not exposed to EALER due to operations. Not all DOE and 

AWE sites experienced the problem of missed EALER.

All external ambient doses are assigned in IREP as follows:

• Exposure rate: Chronic, and

• Radiation type: Photons E = 30 to 250 keV.
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 25 of 26

ATTACHMENT B 

ONSITE AMBIENT DOSE ASSIGNMENT (continued)

Due to the variations in site geography, monitoring practices, reporting practices, facilities, and 

operations, a best estimate cannot be generated for a site that does not have a complete site profile. 

To assess doses in recent years that might not be covered in the site profiles, the most recent onsite 

ambient doses can be assumed to apply.

B.2 BEST-ESTIMATE METHOD

Most dose reconstructions, for which data are provided in the site profile, use a Monte Carlo method 

in which the external environmental dose with its uncertainty distribution is multiplied by the 

appropriate organ DCF distribution to calculate the best-estimate ambient dose. A best estimate must 

take into account all available records pertinent to determination of work location. For workers who 

worked in multiple areas of the site, if the worker’s records and claimant interview do not provide 

enough information to determine specific work locations, the site GM dose, with GSD, is appropriate 

for best estimates. Battelle-TIB-5000 (BT 2007) provides guidance on the relationship between the 

arithmetic average and GM (and GSD).

Dose reconstructors should use site-specific guidance on the calculation of environmental dose when 

provided (e.g., number of hours, area dose rates, etc. for calculation of the ambient dose assignment). 

If site-specific guidance is not available, the assumption should be made that the worker worked 

50 hr/wk and 50 wk/yr (to reasonably account for holiday and/or vacation time) or a total of 2,500 work 

hours per year when assigning ambient dose as a best estimate. Partial years of employment should 

be scaled accordingly. In addition, if the claimant interview indicates that fewer or more hours were 

worked, then this information should be used. For example, if the worker was off the site half of the 

time where only natural background radiation levels existed, only 25 hours of exposure per week 

should be assumed.

The calculated annual dose using site information is multiplied by the appropriate exposure-(R)-toorgan DCF in OCAS-IG-001 (NIOSH 2007) for an isotropic exposure geometry for most sites to 

determine the dose to be assigned for each year. Some sites provide environmental data based on 

survey measurements or calculations rather than film or TLD dosimeter results. For sites that provide 

the data, the calculated annual dose should be multiplied by the appropriate ambient dose equivalent 

[H*(10)] DCF in OCAS-IG-001 for an isotropic exposure geometry. A DCF of 1 for onsite ambient 

dose is applied for cancers where the skin is used to calculate external dose.

If supporting documentation for all of the elements necessary for a best-estimate dose reconstruction 

does not exist, conservative assumptions should be applied in relation to work location and area 

conditions, erring in favor of the claimant.
Document No. ORAUT-OTIB-0088 Revision No. 01 Effective Date: 10/21/2019 Page 26 of 26

 

ATTACHMENT C DOE ADOPTION OF INTERNATIONAL COMMISSION ON RADIOLOGICAL 

PROTECTION (ICRP) PUBLICATION 60

DOE sites provided information about their incorporation of the ICRP Publication 60 recommendations 

on neutron radiation weighting factors, which were published in 1990 (ORAUT 2011b). This 

information is presented below:

Table C-1. Site ICRP Publication 60 incorporation dates.

Site Date ICRP Publication 60 incorporated

Argonne National Laboratory (East and West) 06/18/2010

Brookhaven National Laboratory 01/01/2010

Fermi National Accelerator Laboratory 07/01/2010

Hanford Site 01/01/2011

Idaho Cleanup Project 04/10/2010

Idaho National Laboratory – Battelle Energy Alliance CWI dosimeters – 04/01/2010

Landauer – 06/13/2010

Kansas City Plant No neutron dose to assign since 1997

Lawrence Berkeley National Laboratory 07/01/2010

Lawrence Livermore National Laboratory 01/01/2009

Los Alamos National Laboratory 01/01/2009

Mound Site After 2006, under DOE Legacy Management

Nevada Test Site 01/01/2010

Oak Ridge National Laboratories (ORNL)

University of Tennessee – Battelle ORNL

DOE ORNL

Bechtel Jacobs Company (UCOR)

TRU Project

Los Alamos Technical Associates – Kentucky

Oak Ridge Institute for Science and Education

ORNL – Wackenhut

01/01/2010

Paducah Gaseous Diffusion Plant

Uranium Disposition Services

Paducah Remediation Service

01/01/2010

Paducah Gaseous Diffusion Plant – Swift & Staley 2011 (calendar year only)

Pantex Plant 01/01/2010

Portsmouth Gaseous Diffusion Plant

Uranium Disposition Services 

Los Alamos Technical Associates/Parallax

Fluor-Babcock & Wilcox

01/01/2010

Sandia National Laboratories (Albuquerque and California) 01/01/2010

Savannah River Site 01/01/2010

Stanford Linear Accelerator Center 07/01/2010

Thomas Jefferson National Accelerator Laboratory 07/01/2010

Waste Isolation Pilot Project 06/01/2010

West Valley Demonstration Project Not implemented

Y-12 Plant 01/01/2010