ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 28

DOE Review Release 03/28/2014

Document Title:

Internal Coworker Dosimetry Data for Area IV 

of the Santa Susana Field Laboratory and the 

De Soto Avenue Facility

Document Number: ORAUT-OTIB-0080

Revision: 00

Effective Date: 03/14/2014

Type of Document: OTIB

Supersedes: None

Subject Expert(s): Matthew G. Arno 

Approval: Signature on File Approval Date: 10/23/2013

Matthew G. Arno, Document Owner

Concurrence: Richard E. Merrill Signature on File for Concurrence Date: 10/22/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 10/22/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 10/22/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/14/2014

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-OTIB-0080 Revision No. 00 Effective Date: 03/14/2014 Page 2 of 28

 PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/14/2014 00 New technical information bulletin to provide internal coworker data 

for SSFL workers. Incorporates formal internal and NIOSH review 

comments. Training required: As determined by the Objective 

Manager. Initiated by Matthew G. Arno.
Document No. ORAUT-OTIB-0080 Revision No. 00 Effective Date: 03/14/2014 Page 3 of 28

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

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

3.0 Data Overview ............................................................................................................................7

3.1 Bioassay Data Selection .................................................................................................7

3.2 Analysis...........................................................................................................................9

3.3 Plutonium ........................................................................................................................9

3.4 Uranium...........................................................................................................................9

3.5 Fission Products............................................................................................................10

4.0 Intake Modeling.........................................................................................................................10

4.1 Assumptions..................................................................................................................10

4.2 Bioassay Fitting.............................................................................................................10

4.3 Plutonium ......................................................................................................................11

4.4 Uranium.........................................................................................................................11

4.5 Fission Products............................................................................................................12

5.0 Assignment of Intakes and Doses ............................................................................................12

5.1 Plutonium ......................................................................................................................12

5.2 Uranium.........................................................................................................................13

5.3 Mixed Fission Products .................................................................................................13

6.0 Attributions and Annotations .....................................................................................................14

References ...........................................................................................................................................15

ATTACHMENT A, COWORKER DATA FIGURES...............................................................................16
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LIST OF TABLES

TABLE TITLE PAGE

5-1 Type M plutonium gross alpha intake rates ..............................................................................13

5-2 Type S plutonium gross alpha intake rates...............................................................................13

5-3 Type F uranium intake rates .....................................................................................................13

5-4 Type M uranium intake rates ....................................................................................................13

5-5 Type S uranium intake rates .....................................................................................................13

5-6 Unadjusted Type F 90Sr intake rates based on mixed fission product bioassay .......................13

A-1 50th- and 84th-percentile urinary excretion rates of plutonium gross alpha, 1968 to 1986 ......16

A-2 50th- and 84th-percentile urinary excretion rates of uranium gross alpha, 1965 to 1988.........16

A-3 50th- and 84th-percentile urinary excretion rates of fission products, 1965 to 1990 ................17

A-4 Type M plutonium intake modeling results................................................................................19

A-5 Type S plutonium intake modeling results ................................................................................22

A-6 Type F uranium intake modeling results ...................................................................................22

A-7 Type M uranium intake modeling results ..................................................................................23

A-8 Type S uranium intake modeling results...................................................................................27

A-9 Type F strontium intake modeling results .................................................................................28

LIST OF FIGURES

FIGURE TITLE PAGE

A-1 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 50th percentile, 1965 to 1968, type M .........................17

A-2 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 50th percentile, 1969 to 1986, type M .........................18

A-3 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 84th percentile, 1965 to 1968, type M .........................18

A-4 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 84th percentile, 1969 to 1986, type M .........................18

A-5 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 50th percentile, all years, type M .................................19

A-6 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 84th percentile, all years, type M .................................19

A-7 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 50th percentile, 1965 to 1968, type S..........................20

A-8 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 50th percentile, 1969 to 1986, type S..........................20

A-9 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 84th percentile, 1965 to 1968, type S..........................20

A-10 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 84th percentile, 1969 to 1986, type S..........................21

A-11 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 50th percentile, all years, type S..................................21

A-12 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with bioassay results, 84th percentile, all years, type S..................................21

A-13 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 50th percentile, all years, type F...........................................22

A-14 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 84th percentile, all years, type F...........................................22
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A-15 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 50th percentile, all years, type M ..........................................23

A-16 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 84th percentile, all years, type M ..........................................23

A-17 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 50th percentile, 1965 to 1968, type S...................................24

A-18 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 50th percentile, 1969 to 1972, type S...................................24

A-19 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 50th percentile, 1973 to 1979, type S...................................24

A-20 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 50th percentile, 1980 to 1988, type S...................................25

A-21 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 84th percentile, 1965 to 1968, type S...................................25

A-22 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 84th percentile, 1969 to 1972, type S...................................25

A-23 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 84th percentile, 1973 to 1979, type S...................................26

A-24 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 84th percentile, 1980 to 1988, type S...................................26

A-25 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 50th percentile, all years, type S...........................................26

A-26 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with bioassay results, 84th percentile, all years, type S...........................................27

A-27 Predicted strontium bioassay results calculated using IMBA-derived strontium intake 

rates compared with bioassay results, 50th percentile, all years, type F..................................27

A-28 Predicted strontium bioassay results calculated using IMBA-derived strontium intake 

rates compared with bioassay results, 84th percentile, all years, type F..................................28
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ACRONYMS AND ABBREVIATIONS

CEP Controls for Environmental Pollution, Inc.

d day

DOE U.S. Department of Energy

dpm disintegrations per minute

GM geometric mean

GSD geometric standard deviation

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

L liter

m meter

MDA minimum detectable activity

mL milliliter

NIOSH National Institute for Occupational Safety and Health

OPOS one person–one sample

ORAU Oak Ridge Associated Universities

pCi picocurie

SRDB Ref ID Site Research Database Reference Identification (number)

SSFL Santa Susana Field Laboratory

TIB technical information bulletin

U.S.C. United States Code

μCi microcurie

μm micrometer

§ section or sections
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1.0 INTRODUCTION

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

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

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

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

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

individual dose reconstructions.

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

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

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

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

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

justification, or clarification of the associated information, are presented in Section 6.0.

2.0 PURPOSE

Some employees at DOE sites were not monitored for potential intakes of radioactive material, or the 

records of such monitoring are incomplete or unavailable. In such cases, data from monitored 

coworkers can be used to assign an internal dose to address potential intakes of radioactive material. 

The purpose of this TIB is to provide coworker intake data for Area IV of the Santa Susana Field 

Laboratory (SSFL) and the De Soto Avenue Facility (sometimes referred to as the Energy Technology 

Engineering Center or Atomics International). This document does not apply to the Canoga Avenue 

Facility and Downey Facility because their covered periods end before the periods that are addressed 

in this document.

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment (ORAUT 

2005), describes the general process for analyzing bioassay data for the assignment of doses to 

individuals based on coworker results. ORAUT-PLAN-0014, Coworker Data Exposure Profile 

Development (ORAUT 2004), describes the approach and processes to develop reasonable exposure 

profiles based on available dosimetric information for workers at DOE sites.

The analysis used internal exposure bioassay data from SSFL. A statistical analysis of the data was 

performed according to ORAUT-OTIB-0019 (ORAUT 2005), its implementing procedure, ORAUTPROC-0095, Generating Summary Statistics for Coworker Bioassay Data (ORAUT 2006), and the 

statistical methods in ORAUT-RPRT-0053, Analysis of Stratified Coworker Datasets (ORAUT 2012). 

The results were entered in the Integrated Modules for Bioassay Analysis (IMBA) computer program 

to obtain intake rates for the assignment of dose distributions.

3.0 DATA OVERVIEW

This section provides information on the general selection characteristics of the data and the methods 

of analysis. More detailed radionuclide-specific information is provided in Section 4.0.

3.1 BIOASSAY DATA SELECTION

Urinalysis bioassay data were obtained directly from the SSFL in the form of electronic scans of hardcopy bioassay data records. These records contain data from SSFL internal records and records 

from the analytical laboratories that performed the urinalyses. SSFL received urinalysis data from the

laboratories and transcribed that data onto internal forms, which SSFL referred to as “8X11” and 

“McBee cards.” The methods of recording the data differed between the laboratories and SSFL 
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internal records. In general, data from the laboratories were recorded in a consistent manner with few 

legibility issues. In the 1960s, the laboratory data were recorded based on the date the sample was 

received at the laboratory, which was generally within 2 weeks of the sample collection date. The 

McBee cards and 8X11 forms dated the records based on the sample collection date. In the 1990s 

and after, the laboratory data appear to use the same date as the McBee cards. The data sources 

can be divided into two broad categories, laboratory data and facility data, with the facility data 

consisting of the McBee cards and the 8X11 forms.

The facility data do not contain as much information as the laboratory data. Many positive results 

were indicated as a “+” with no value. Data indicating less than the minimum detectable activity 

(MDA) were recorded either as a zero (earlier years) or as being less than a generic MDA. In 

contrast, the laboratory data report the positive values and commonly report MDAs.

Ideally, for a given bioassay sample, there should be both a laboratory record and a facility record that 

provide the results of the analysis. However, this is not always the case. In addition, it is not possible 

to establish a 100%-certain one-to-one correlation between facility and laboratory data. It is possible, 

especially with reported positive results in the facility data, to individually correlate records by hand 

based on professional judgment. Examination of the different data sources does reveal that neither 

source can be considered complete. There are laboratory data for which no facility data can be found 

and vice versa. This is most easily demonstrated by observed instances where a given tracking 

number (a unique number for an individual) appears only in one of the data sources for a given year. 

Therefore both data sources were used with identified duplicates excluded as discussed below.

For most of the 1960s, both data sources are available and the number of records from each source 

is comparable. Only facility data are available for February 14, 1968, through July 19, 1989. From 

July 20, 1989, forward, there are more laboratory data available than facility data.

The sample receipt date on the laboratory data form is usually a few days after the sample collection 

date on the facility form. The bioassay data records were reviewed by a Project health physicist and 

in those instances where a reasonable correlation between the facility data and the laboratory data 

could be made, the facility data was excluded as duplicate, especially because the facility data 

typically contain less detail about a given sample.

During the review, multiple results from the same date (based on sample collection date) with the 

same result or where a zero and a less-than result were reported were excluded to remove duplicates 

and leave only one result for statistical analysis. In addition, questionable reported activity units were 

encountered, such as results in units of μCi/L when pCi/L would be consistent with contemporaneous 

data or dpm/mL instead of dpm/sample. The reported units were altered to be consistent with the 

contemporaneous reporting units. As part of this, results marked as “ND,” “NDA,” or “Neg” were set to 

zero, illegible results were excluded, and the date the sample was received by the laboratory was 

used as the sample date if no sample date was available.

Data from before January 1, 1965 (the end of the Special Exposure Cohort period) or after December 

31, 2005 (the last year with a complete dataset) were not used. Other records were excluded from 

the analysis for the following reasons:

• Results with volume or mass units of grams were assumed to be fecal samples.

• Controls for Environmental Pollution, Inc. (CEP)-derived data, including all data from August 4, 

1991, through June 1, 1993, and on June 20, 1993, were suspect due to quality issues with 

the laboratory. This period encompasses all CEP-derived data.

• Data with a blank or “?” for analysis type were not usable.
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Some of the records have individual fields that are blank or illegible but contain sufficient information 

to proceed with the analysis. The following rules were used to adjust for blank or unusable fields [1]:

• Volumes were assumed to be 1,500 mL, and

• Units were assumed to be dpm/sample.

In addition, all sample results were adjusted from a daily urinary excretion of 1,500 mL to 1,400 mL.

3.2 ANALYSIS

Bioassay data were analyzed by year or multiyear span depending on the amount of data available for 

each radionuclide during a given period and the expected biokinetics of each radionuclide. A 

lognormal distribution was assumed. After log-transforming the data, the 50th and 84th percentiles 

were determined for each period through the use of the methods described in ORAUT-RPRT-0053 

(ORAUT 2012).

In ORAUT-OTIB-0075, Use of Claimant Datasets for Coworker Modeling (ORAUT 2009), arguments 

are presented to support the practice of treating a claimant dataset as a simple random sample from 

the population of all monitored workers. A potential problem in using a claimant dataset is that the 

workers who were involved in incidents usually submitted more samples than workers who submitted

only routine (not incident-related) samples. This is problematic because a small number of workers 

who were involved in incidents can dominate the claimant sample in a given year through the sheer 

number of samples and because the samples in the dataset are no longer independent of each other. 

To compensate for the unequal number of samples from the workers, the one person–one sample 

(OPOS) technique was used, in which only one result is used for each person for each radionuclide 

for a given year. The OPOS statistic was calculated using the maximum possible mean methodology 

in ORAUT (2012).

3.3 PLUTONIUM

Two analytical techniques were used for plutonium bioassay – autoradiography (labeled “PUA”) and 

gas-flow proportional counting (labeled “PUB”). The two techniques were distinguished beginning in 

1974. Previous results were only indicated as “PU,” which was presumed to be indicative of 

autoradiography based on the earlier use of this procedure. It was assumed the A and B designations 

were added with the introduction of the gas-flow proportional counting procedure. Analysis type 

designations that were used for this study were “PUA” (autoradiography), “PU,” and “Pu” (unknown 

analytical technique, presumed to be autoradiography). No results were identified with an analysis

type of “PUB.” Sufficient bioassay data were available to perform a statistical analysis for 1965 

through 1986. Although plutonium bioassay data are available through 1994, there were insufficient 

data to perform a statistical analysis after 1986.

3.4 URANIUM

Uranium data are presented as fluorometric (mass) and radiometric (activity) results. Samples 

analyzed fluorometrically were also analyzed radiometrically. Only the data that were reported in 

activity units were used. The analysis type designation that was used was “UR.” Sufficient bioassay 

data were available to perform a statistical analysis for 1965 through 1988. Although uranium 

bioassay data are available through 1993, there were insufficient data to perform a statistical analysis 

after 1988.
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3.5 FISSION PRODUCTS

Beta emitter bioassay data were used to represent mixed fission products. These bioassay data 

should be evaluated in accordance with the latest revision of ORAUT-OTIB-0054, Fission and 

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

(ORAUT 2014) and, specifically, the data should be considered to be chemically processed beta 

samples in accordance with Table 7-2 of that document. This is consistent with some of the analytical 

techniques that were used by SSFL or is favorable to claimants for the other analytical techniques that 

were used by SSFL. Because data from the different techniques were merged for analysis, the 

categorization of the different analytical techniques most favorable to claimants was used for 

application of ORAUT-OTIB-0054.

The analysis type designations that were used were “MFP” (mixed fission product, chemical 

separation of alkaline earths and rare earths including strontium) if the method type was “B” for beta 

counting (which excludes gamma-counting data), “MFPB,” and “MFP(B),” and “FP” if the method type 

was “3A” (mixed fission products less cesium and volatiles, assumed to indicate strontium). Sufficient 

bioassay data were available to perform a statistical analysis for 1965 through 1991. Although fission 

product bioassay data is available through 1993, there was insufficient data to perform a statistical 

analysis after 1991.

4.0 INTAKE MODELING

This section discusses intake modeling assumptions, intake fitting, and intake materials.

4.1 ASSUMPTIONS

Each result that was used in the intake calculations was assumed to have a normal distribution. A 

uniform absolute error of 1 was applied to all results, assigning the same weight to each result. 

Because of the nature of work at SSFL, intakes could have been chronic or acute. However, a series 

of acute intakes can be approximated as a chronic intake. Therefore, intakes were assumed to be 

chronic and to occur through inhalation with a 5-μm activity median aerodynamic diameter particle 

size distribution.

For intake modeling, all plutonium activity was assumed to be 239Pu. This assumption did not affect 

the fitting of the data for intake determination because all plutonium isotopes have the same biokinetic 

behavior and the isotopes this analysis considered all have long half-lives in relation to the assumed 

intake period. 

For intake modeling, all uranium activity was assumed to be 234U. This assumption did not affect the 

fitting of the data for intake determination because all uranium isotopes have the same biokinetic 

behavior and the isotopes this analysis considered all have long half-lives in relation to the assumed 

intake period. International Commission on Radiological Protection (ICRP) Publication 68 dose 

coefficients (also referred to as dose conversion factors) for 234U are 7% to 31% larger than the dose 

coefficients for 235U, 236U, and 238U (ICRP 1995). Therefore, the assumption that the intake was 100% 

234U provides a result that is favorable to claimants.

4.2 BIOASSAY FITTING

IMBA was used to fit the bioassay results to a series of inhalation intakes. Data for each radionuclide 

were fit as a series of chronic intakes. The intake assumptions were based on observed patterns in 

the bioassay data. Periods with constant chronic intake rates were chosen by the selection of periods 

in which the bioassay results were similar. A new chronic intake period was started if the data 
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indicated a significant and sustained change in the bioassay results. By this method, the years were 

divided into multiple chronic intake periods for each radionuclide.

4.3 PLUTONIUM

Because the plutonium isotopes at SSFL have very long radiological half-lives, and because the 

material is retained in the body for long periods, excretion results are not independent. For example, 

an intake in the 1960s could contribute to urinary excretion in the 1980s and later. To avoid potential 

underestimation of intakes for people who worked at SSFL for relatively short periods, each chronic 

intake was fit independently using only the bioassay results from the single intake period for types M 

and S solubility. This method resulted in an overestimate of intakes for exposures that extended 

through multiple assumed intake periods; however, these intake rates are to be considered best 

estimates. Only the results in the intake period were selected for use in the fitting for each period. 

Excluded results are shown in light gray or red in the figures in Attachment A. Included results are 

dark gray or blue. The results of the plutonium statistical analysis that was used to calculate the 

intakes are provided in Table A-1.

Insufficient bioassay data were available for 1965 and 1966 to perform a statistical analysis. The 

Nuclear Materials development facility began operation in 1967, which led to an increase in potential 

plutonium exposure and therefore plutonium urinalysis data. The 1967-to-1968 data was used to 

back-extrapolate for 1965 and 1966, when the potential plutonium exposure was less. This 

assumption is favorable to claimants.

Plutonium Type M: The solid lines in Figures A-1 to A-4 in Attachment A show the individual fits to 

the 50th- and 84th-percentile excretion rates, respectively, for type M materials. Figures A-5 and A-6 

show the predicted 50th- and 84th-percentile excretion rates, respectively, from all type M intakes. 

Table A-4 lists the 50th- and 84th-percentile intake rates determined from statistical analysis of the 

plutonium urinalysis data along with the associated geometric standard deviations (GSDs).

Plutonium Type S: The solid lines in Figures A-7 to A-10 in Attachment A show the individual fits to 

the 50th- and 84th-percentile excretion rates, respectively, for type S materials. The same intake 

periods were applied for both percentiles because the values followed a similar pattern. Figures A-11 

and A-12 show the predicted 50th- and 84th-percentile excretion rates, respectively, from all type S 

intakes. Table A-5 lists the 50th- and 84th-percentile intake rates determined from statistical analysis 

of the plutonium urinalysis data along with the associated GSDs.

4.4 URANIUM

Because the uranium isotopes at SSFL have very long radiological half-lives, and because the 

material is retained in the body for long periods, excretion results are not independent. For example, 

an intake in the 1950s could contribute to urinary excretion in the 1980s and later. To avoid potential 

underestimation of intakes for people who worked at SSFL for relatively short periods, each chronic 

intake was fit independently using only the bioassay results from the single intake period for type S 

solubility. This method results in an overestimate of intakes for exposures that extended through 

multiple assumed intake periods; however, these intake rates are to be considered best estimates. 

Only the results in the intake period were selected for use in the fitting for each period. Excluded 

results are shown in light gray or red in the figures in Attachment A. Included results are dark gray or 

blue. For types M and F solubility, this approach was not used. In 1974, the excretion rates from the 

statistical analysis were inconsistent with the contemporaneous years and were excluded to be 

favorable to claimants. In 1984 and 1985, the results are zero (statistical analysis not possible) and 

were similarly excluded. The results of the uranium statistical analysis that was used to calculate the 

intakes are provided in Table A-2.
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Uranium Type F: The solid lines in Figures A-13 and A-14 in Attachment A show the fit to the 50thand 84th-percentile excretion rates, respectively, for type F materials. Table A-6 lists the 50th- and 

84th-percentile intake rates determined from statistical analysis of the uranium urinalysis data along 

with the associated GSDs.

Uranium Type M: The solid lines in Figures A-15 and A-16 in Attachment A show the fit to the 50thand 84th-percentile excretion rates, respectively, for type M materials. Table A-7 lists the 50th- and 

84th-percentile intake rates determined from statistical analysis of the uranium urinalysis data along 

with the associated GSDs.

Uranium Type S: The solid lines in Figures A-17 to A-24 in Attachment A show the individual fits to 

the 50th- and 84th-percentile excretion rates, respectively, for type S materials. The same intake 

periods were applied for both percentiles because the values followed a similar pattern. Figures A-25 

and A-26 show the 50th- and 84th-percentile predicted excretion rates, respectively, from all type S 

intakes. Table A-8 lists the 50th- and 84th-percentile intake rates determined from statistical analysis 

of the uranium urinalysis along with the associated GSDs.

4.5 FISSION PRODUCTS

The fission product bioassay data is a beta emitter analysis and was assumed to be 90Sr for the 

purposes of intake modeling. The results of the statistical analysis of fission product bioassay data 

that were used to calculate the intakes are provided in Table A-3. The solid lines in Figures A-27 and 

A-28 in Attachment A show the individual fits to the 50th- and 84th-percentile excretion rates, 

respectively, from all type F intakes. Table A-9 lists the 50th- and 84th-percentile intake rates the 

determined from statistical analysis of the beta-emitter urinalysis data along with the associated 

GSDs.

5.0 ASSIGNMENT OF INTAKES AND DOSES

This section describes the derived intake rates and provides guidance for assigning doses. For the 

calculation of doses to individuals from bioassay data, a minimum GSD of 3 was used to account for 

biological variation and uncertainty in the models. It was considered inappropriate to assign a value 

less than 3 for the coworker data. Therefore, a GSD of at least 3 was assigned for each of the intake 

periods. The 95th-percentile values were based on the adjusted GSD for the intake period. For 

cases in which there is justification that the individual might have had larger intakes than the 50thpercentile intake rates, dose reconstructors should use the 95th-percentile intake rates input into 

IREP as a constant. The original GSDs are provided in the tables for each radionuclide and solubility 

type in Attachment A.

The following subsections list the intake rates that should be used for each radionuclide and the 

period of applicability of each intake rate.

5.1 PLUTONIUM

Tables 5-1 and 5-2 list the plutonium intakes and associated GSDs to be used for each year of 

potential plutonium exposure. All data derive from urinalysis bioassay results and should be treated 

as plutonium gross alpha intakes, which can be considered to be 100% 239Pu. If evaluation of Type 

Super S solubility is necessary, ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly 

Retained in the Lung (ORAUT 2010a), should be applied accordingly. The 1986 intake rates can be 

extended past 1986 as a measure favorable to claimants.
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Table 5-1. Type M plutonium gross alpha intake rates (dpm/d).

Start End 50th percentile GSD 95th percentile

1/1/1965 12/31/1968 13.12 5.15 195

1/1/1969 12/31/1986 3.018 3.00 18.4

Table 5-2. Type S plutonium gross alpha intake rates (dpm/d).

Start End 50th percentile GSD 95th percentile

1/1/1965 12/31/1968 329 5.15 4,880

1/1/1969 12/31/1986 41.62 3.00 254

5.2 URANIUM

Tables 5-3 to 5-5 list the uranium intakes and associated GSDs to be used for each year of potential 

uranium exposure. The 1988 intake rates can be extended past 1988 as a measure favorable to 

claimants.

Table 5-3. Type F uranium intake rates (dpm/d).

Start End 50th percentile GSD 95th percentile

1/1/1965 12/31/1968 17.6 3.00 107

1/1/1969 12/31/1972 8.91 3.00 54.3

1/1/1973 12/31/1979 18.87 3.00 115

1/1/1980 12/31/1988 5.666 3.00 34.5

Table 5-4. Type M uranium intake rates (dpm/d).

Start End 50th percentile GSD 95th percentile

1/1/1965 12/31/1968 75.91 3.00 463

1/1/1969 12/31/1972 34.92 3.00 213

1/1/1973 12/31/1979 78.02 3.00 475

1/1/1980 12/31/1988 21.92 3.00 134

Table 5-5. Type S uranium intake rates (dpm/d).

Start End 50th percentile GSD 95th percentile

1/1/1965 12/31/1968 1,592 3.00 9,701

1/1/1969 12/31/1972 744.4 3.00 4,536

1/1/1973 12/31/1979 1,266 3.00 7,714

1/1/1980 12/31/1988 389 3.00 2,370

5.3 MIXED FISSION PRODUCTS

Intake rates in Table 5-6 were calculated from the fission product urine sample results, assuming a 

90Sr biokinetic model. Because the urine results were based on gross activity analyses, a number of 

mixed fission products were included in the analysis; therefore, these intake rates must be adjusted to 

account for the fraction of activity attributable to 90Sr. The intake rates should be treated as samples 

following major chemical processing. The resulting intake rates should then be used to assign intakes 

for all mixed fission products in accordance with ORAUT-OTIB-0054 (ORAUT 2014). The 1991 intake 

rate can be extended past 1991 as a measure favorable to claimants.

Table 5-6. Unadjusted Type F 90Sr intake rates (dpm/d) based on 

mixed fission product bioassay.

Start End 50th percentile GSD 95th percentile

1/1/1965 12/31/1991 61.05 3.00 372
Document No. ORAUT-OTIB-0080 Revision No. 00 Effective Date: 03/14/2014 Page 14 of 28

6.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).

[1] Arno, Matthew. ORAU Team. Dose Reconstructor. January 21, 2011.

These rules were developed based on examination of the data for trends in how the data was 

recorded, review of the internal dose technical basis document for SSFL (ORAUT 2010b) to 

determine default, and/or assumed analytical techniques and reporting conventions.
Document No. ORAUT-OTIB-0080 Revision No. 00 Effective Date: 03/14/2014 Page 15 of 28

REFERENCES

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

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22731]

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

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

ORAUT (Oak Ridge Associated Universities Team), 2005, 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), 2006, Generating Summary Statistics for 

Coworker Bioassay Data, ORAUT-PROC-0095, Rev. 00, Oak Ridge, Tennessee, June 5.

ORAUT (Oak Ridge Associated Universities Team), 2009, Use of Claimant Datasets for Coworker 

Modeling, ORAUT-OTIB-0075, Rev. 00, Oak Ridge, Tennessee, May 25.

ORAUT (Oak Ridge Associated Universities Team), 2010a, Estimating Doses for Plutonium Strongly 

Retained in the Lung, ORAUT-OTIB-0049, Rev. 01 PC-2, Oak Ridge, Tennessee, November 

29.

ORAUT (Oak Ridge Associated Universities Team), 2010b, Area IV of the Santa Susana Field 

Laboratory, the Canoga Avenue Facility, the Downey Facility, and the De Soto Avenue Facility 

(sometimes referred to as Energy Technology Engineering Center [ETEC] or Atomics 

International) - Occupational Internal Dose, ORAUT-TKBS-0038-5, Rev. 01, Oak Ridge, 

Tennessee, April 26.

ORAUT (Oak Ridge Associated Universities Team), 2012, Analysis of Stratified Coworker Datasets, 

ORAUT-RPRT-0053, Rev. 01, Oak Ridge, Tennessee, July 16.

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

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

02, Oak Ridge, Tennessee, March 6.
Document No. ORAUT-OTIB-0080 Revision No. 00 Effective Date: 03/14/2014 Page 16 of 28

Table A-1. 50th- and 84th-percentile urinary excretion 

rates of plutonium gross alpha, 1968 to 1986 (dpm/d).

Effective

bioassay date

50th

percentile

84th

percentile

No. of

employees

1/1/1968 0.064 0.328 34

1/1/1970 0.026 0.097 31

1/1/1974 0.013 0.040 28

1/1/1976 0.037 0.049 48

1/1/1980 0.022 0.034 30

1/1/1984 0.040 0.053 48

1/1/1986 0.020 0.031 53

Table A-2. 50th- and 84th-percentile urinary excretion 

rates of uranium gross alpha, 1965 to 1988 (dpm/d).

Effective

bioassay date

50th

percentile

84th

percentile

No. of

employees

7/1/1965 2.90 8.33 470

7/1/1966 4.52 10.83 298

7/1/1967 5.47 14.69 338

7/1/1968 6.22 14.05 269

7/1/1969 2.50 8.60 153

7/1/1970 3.41 7.79 164

7/1/1971 2.35 5.09 135

7/1/1972 1.68 5.67 122

7/1/1973 4.99 7.09 88

7/1/1974 5.95 7.53 86

7/1/1975 0.23 1.13 135

7/1/1976 5.77 8.03 166

7/1/1977 4.52 6.68 95

7/1/1978 4.89 14.14 147

7/1/1979 5.14 6.52 124

7/1/1980 1.76 3.66 109

7/1/1981 1.34 2.08 104

7/1/1982 0.22 0.86 75

7/1/1983 3.17 3.41 57

7/1/1986 2.05 3.18 34

7/1/1987 1.48 2.29 59

7/1/1988 1.44 2.24 63

ATTACHMENT A

COWORKER DATA FIGURES

Page 1 of 13
Document No. ORAUT-OTIB-0080 Revision No. 00 Effective Date: 03/14/2014 Page 17 of 28

Table A-3. 50th- and 84th-percentile urinary excretion 

rates of fission products, 1965 to 1990 (dpm/d).

Effective

bioassay date

50th

percentile

84th

percentile

No. of

employees

7/1/1965 26.06 58.88 321

7/1/1966 21.04 38.26 216

7/1/1967 20.99 44.65 221

7/1/1968 19.63 30.48 154

7/1/1969 10.07 24.71 116

7/1/1970 19.34 24.40 125

7/1/1971 16.54 28.26 84

7/1/1972 14.67 23.58 75

7/1/1973 12.90 24.28 66

7/1/1974 4.65 22.27 55

7/1/1975 12.90 20.51 84

7/1/1976 18.72 29.14 67

7/1/1977 6.98 31.58 75

7/1/1978 14.25 30.33 118

7/1/1979 9.06 22.97 84

7/1/1980 5.54 17.07 82

7/1/1981 17.75 33.77 55

7/1/1982 14.67 33.51 65

7/1/1983 15.42 23.90 45

7/1/1984 3.59 14.29 65

7/1/1985 15.89 26.69 67

7/1/1986 16.15 27.12 78

7/1/1987 21.92 38.70 63

7/1/1988 24.52 29.45 65

7/1/1989 22.79 37.83 50

7/1/1990 7.67 45.77 33

3/1/1991 8.08 35.30 27

Figure A-1. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 50th percentile, 1965 to 1968, type M.

ATTACHMENT A

COWORKER DATA FIGURES

Page 2 of 13
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Figure A-2. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 50th percentile, 1969 to 1986, type M.

Figure A-3. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 84th percentile, 1965 to 1968, type M.

Figure A-4. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 84th percentile, 1969 to 1986, type M.

ATTACHMENT A

COWORKER DATA FIGURES

Page 3 of 13
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Figure A-5. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 50th percentile, all years, type M.

Figure A-6. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 84th percentile, all years, type M.

Table A-4. Type M plutonium intake modeling results (dpm/d).

Year(s) GM 84th percentile GSD Adj. GSD 95th percentile

1965–1968 13.12 67.61 5.15 5.15 195

1969–1986 3.018 4.959 1.64 3.00 18.4

ATTACHMENT A

COWORKER DATA FIGURES

Page 4 of 13
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Figure A-7. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 50th percentile, 1965 to 1968, type S.

Figure A-8. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 50th percentile, 1969 to 1986, type S.

Figure A-9. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 84th percentile, 1965 to 1968, type S.

ATTACHMENT A

COWORKER DATA FIGURES

Page 5 of 13
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Figure A-10. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 84th percentile, 1969 to 1986, type S.

Figure A-11. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 50th percentile, all years, type S.

Figure A-12. Predicted plutonium bioassay results calculated using IMBA-derived plutonium 

intake rates (line) compared with bioassay results (dots), 84th percentile, all years, type S.

ATTACHMENT A

COWORKER DATA FIGURES

Page 6 of 13
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Table A-5. Type S plutonium intake modeling results (dpm/d).

Year(s) GM 84th percentile GSD Adj. GSD 95th percentile

1965–1968 329 1,695 5.15 5.15 4,880

1969–1986 41.62 64.15 1.54 3.00 254

Figure A-13. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 50th percentile, all years, type F.

Figure A-14. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 84th percentile, all years, type F.

Table A-6. Type F uranium intake modeling results (dpm/d).

Year(s) GM 84th percentile GSD Adj. GSD 95th percentile

1965–1968 17.6 44.09 2.51 3.00 107

1969–1972 8.91 24.35 2.73 3.00 54.3

1973–1979 18.87 30.03 1.59 3.00 115

1980–1988 5.666 8.652 1.53 3.00 34.5

ATTACHMENT A

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Page 7 of 13
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Figure A-15. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 50th percentile, all years, type M.

Figure A-16. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 84th percentile, all years, type M.

Table A-7. Type M uranium intake modeling results (dpm/d).

Year(s) GM 84th percentile GSD Adj. GSD 95th percentile

1965–1968 75.91 190.7 2.51 3.00 463

1969–1972 34.92 93.84 2.69 3.00 213

1973–1979 78.02 124.1 1.59 3.00 475

1980–1988 21.92 33.04 1.51 3.00 134

ATTACHMENT A

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Page 8 of 13
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Figure A-17. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 50th percentile, 1965 to 1968, type S.

Figure A-18. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 50th percentile, 1969 to 1972, type S.

Figure A-19. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 50th percentile, 1973 to 1979, type S.

ATTACHMENT A

COWORKER DATA FIGURES

Page 9 of 13
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Figure A-20. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 50th percentile, 1980 to 1988, type S.

Figure A-21. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 84th percentile, 1965 to 1968, type S.

Figure A-22. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 84th percentile, 1969 to 1972, type S.

ATTACHMENT A

COWORKER DATA FIGURES

Page 10 of 13
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Figure A-23. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 84th percentile, 1973 to 1979, type S.

Figure A-24. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 84th percentile, 1980 to 1988, type S.

Figure A-25. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 50th percentile, all years, type S.

ATTACHMENT A

COWORKER DATA FIGURES

Page 11 of 13
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Figure A-26. Predicted uranium bioassay results calculated using IMBA-derived uranium intake 

rates (line) compared with bioassay results (dots), 84th percentile, all years, type S.

Table A-8. Type S uranium intake modeling results (dpm/d).

Year(s) GM 84th percentile GSD Adj. GSD 95th percentile

1965–1968 1,592 3,930 2.47 3.00 9,701

1969–1972 744.4 1,998 2.68 3.00 4,536

1973–1979 1,266 2,101 1.66 3.00 7,714

1980–1988 389 581 1.49 3.00 2,370

Figure A-27. Predicted strontium bioassay results calculated using IMBA-derived strontium 

intake rates (line) compared with bioassay results (dots), 50th percentile, all years, type F.

ATTACHMENT A

COWORKER DATA FIGURES

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Figure A-28. Predicted strontium bioassay results calculated using IMBA-derived strontium 

intake rates (line) compared with bioassay results (dots), 84th percentile, all years, type F.

Table A-9. Type F strontium intake modeling results (dpm/d).

Year(s) GM 84th percentile GSD Adj. GSD 95th percentile

1965–1991 61.05 124.3 2.04 3.00 372

ATTACHMENT A

COWORKER DATA FIGURES

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