Draft

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH

National Institute for Occupational Safety and Health

EVALUATION OF URANIUM DOSE RECONSTRUCTION 

METHODS AT CPP (1953–1962)

Contract No. 211-2014-58081 

SCA-TR-2018-SEC003, Revision 0 

Prepared by

Bob Barton, CHP 

SC&A, Inc.

2200 Wilson Boulevard, Suite 300 

Arlington, VA 22201-3324 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

July 25, 2018 

DISCLAIMER

This is a working document provided by the Centers for Disease Control and Prevention (CDC) technical 

support contractor, SC&A for use in discussions with the National Institute for Occupational Safety and 

Health (NIOSH) and the Advisory Board on Radiation and Worker Health (ABRWH), including its 

Working Groups or Subcommittees. Documents produced by SC&A, such as memorandum, white paper, 

draft or working documents are not final NIOSH or ABRWH products or positions, unless specifically 

marked as such. This document prepared by SC&A represents its preliminary evaluation on technical 

issues.

NOTICE: This document has been reviewed to identify and redact any information that is protected by 

the Privacy Act 5 U.S.C. § 552a and has been cleared for distribution. 
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SC&A, INC.: Technical Support for the Advisory Board on Radiation and 

Worker Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: Evaluation of Uranium Dose Reconstruction Methods at CPP 

(1953–1962)

DOCUMENT NUMBER/

DESCRIPTION:

SCA-TR-2018-SEC003 

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: July 25, 2018

TASK MANAGER John Stiver, MS, CHP [signature on file]

PROJECT MANAGER: John Stiver, MS, CHP [signature on file]

DOCUMENT 

REVIEWER(S):

John Stiver, MS, CHP [signature on file]

Joe Fitzgerald, MS, MPH [signature on file]

Record of Revisions

Revision 

Number

Effective 

Date

Description of Revision

0 (Draft) 7/25/2018 Initial issue
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TABLE OF CONTENTS

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

1 Introduction and Background ................................................................................................. 5

2 Summary of Claimant Intake Evaluations of Uranium Using Gross Beta Urinalysis ............ 8

2.1 Assumptions Used in Intake Calculations ...................................................................... 8

2.1.1 Intake Evaluation Regimes......................................................................................... 9

2.1.2 Intake Mode and Solubility Type................................................................................ 9

2.1.3 Ratio of Uranium to Strontium ................................................................................... 9

2.1.4 Treatment of Zero and Less than Minimum Detectable Activity............................. 10

2.2 Summary of Daily Intake Rates of Sr-90 and Uranium Using Individual Claimant 

Gross Beta Bioassay ..................................................................................................... 11

2.3 Daily Intake Rates of Uranium Using TWOPOS/Coworker Simulation on Claimant 

Gross Beta Bioassay ..................................................................................................... 12

3 Alternate Estimates for Potential Uranium Intake ................................................................ 13

3.1 Uranium Intakes Based on Alpha Air Sampling .......................................................... 13

3.2 Uranium Intakes Based on the Missed Dose from Bioassay ........................................ 13

4 Summary and Conclusion ..................................................................................................... 15

5 References ............................................................................................................................. 16

Attachment A: [Redacted in Full] ................................................................................................. 17

Attachment B: [Redacted in Full] ................................................................................................. 18
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ABBREVIATIONS AND ACRONYMS

Ac actinium

Am americium

ANL-W Argonne National Laboratory – West

Cm curium 

CPP Chemical Processing Plant

DCAS Division of Compensation Analysis and Support 

dpm/d disintegrations per minute per day 

FAP fission and activation products 

ICPP Idaho Chemical Processing Plant

IMBA Integrated Modules for Bioassay Analysis

INL Idaho National Laboratory 

MDA minimum detectable activity

μCi/cm3 microcurie per cubic centimeter

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH/OCAS Claims Tracking System

Np neptunium 

Pa protactinium

pCi picocurie 

Pu plutonium 

SEC Special Exposure Cohort 

Sr strontium 

TBD technical basis document

Th thorium 

TLD thermoluminescent dosimeter

TWOPOS time weighted one-person-one-statistic

U uranium 
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1 INTRODUCTION AND BACKGROUND

As part of the Idaho National Laboratory/Argonne National Laboratory-West (INL/ANL-W) 

Work Group’s review of the National Institute of Occupational Safety and Health (NIOSH) 

evaluation report for Special Exposure Cohort (SEC) Petition SEC-000219 (NIOSH 2017), 

SC&A was tasked to investigate potential SEC issues related to other areas and time periods 

apart from the proposed class at the Chemical Processing Plant (CPP), which is defined as 

follows: 

All employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Idaho National Laboratory 

(INL) in Scoville, Idaho, and (a) who were monitored for external radiation at the 

Idaho Chemical Processing Plant (CPP) (e.g., at least one film badge or TLD 

dosimeter from CPP) between January 1, 1963 and February 28, 1970; or (b) 

who were monitored for external radiation at INL (e.g., at least one film badge or 

TLD dosimeter) between March 1, 1970 and December 31, 1974 for a number of 

work days aggregating at least 250 work days, occurring either solely under this 

employment, or in combination with work days within the parameters established 

for one or more other classes of employees in the Special Exposure Cohort. 

[NIOSH 2017] 

As part of SC&A’s review tasking, a focused review of the CPP prior to 1963 was performed. In 

a joint effort by the Work Group, NIOSH, and SC&A, several data captures and interviews with 

former INL workers were conducted between January and December 2016. These activities 

included: 

• January 25, 2016, to January 27, 2016: In-person interviews with former INL workers

• January 28, 2016: INL onsite data capture 

• February 16, 2016: Telephone interviews with former INL workers 

• March 15, 2016, to March 16, 2016: INL onsite data capture 

• April 5, 2016: Telephone interviews with former INL workers

• November 8, 2016, to November 10, 2016: In-person interviews with former INL 

workers

• December 15, 2016: Telephone interviews with former INL workers 

In July 2017, SC&A delivered its evaluation of dose reconstruction feasibility at CPP prior to 

1963, which focused on the ability to reconstruct doses to alpha-emitting radionuclides during 

this period (SC&A 2017). That report contained five findings and five observations related to the 

ability to reconstruct doses to alpha-emitting material (primarily isotopes of uranium) that was 

no longer comingled with fission and activation products (FAP), which are beta/gamma-emitting 

material. This is of particular import because current methods for reconstructing doses to alphaemitting material for workers who were not directly monitored for alpha contaminants is to use a 

ratio to the individual worker’s intake based on gross beta and/or gross gamma urinalysis. The 
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summary conclusion of SC&A report, as stated in the Executive Summary of that document, is 

as follows: 

Summary Conclusion: SC&A identified several example locations and time 

periods for which alpha contamination was identified and was not directly 

comingled with FAP. Reconstruction of internal exposures to alpha material by 

ratioing to calculated intakes of FAP material would not be technically 

appropriate for at least some workers, activities, and locations within CPP. 

[SC&A 2017, p. 7]

In April 2018, NIOSH provided a white paper response (NIOSH 2018a) to SC&A’s evaluation 

of alpha exposures at CPP during the pre-1963 era. That white paper provided detailed responses 

to each of SC&A’s five findings and five observations and came to the following conclusion: 

In summary, based on further evaluation of the findings and observations from 

Draft Review of Internal Alpha Exposure Potential at CPP Prior to 1963, DCAS 

still believes it can reconstruct doses with sufficient accuracy for workers at CPP 

before 1963. This conclusion was largely reached due to the comprehensive 

Health Physics at CPP, which included tight contamination controls, airborne 

radioactivity monitoring, attention to changes in source term based on process 

knowledge with appropriate monitoring, and a large in-vitro bioassay program. 

There was a small subset of the CPP workforce that was on a routine uranium 

bioassay program due to internal exposure potential to uranium without fission 

products present. Internal monitoring for most of the CPP workforce for alphaemitting material was limited, as such monitoring was typically incident-based, as 

the actual exposure potential was likely restricted to certain operations and 

analytical laboratory personnel. Special bioassay monitoring was identified in 

interviews with former CPP workers, as well as in bioassay records and incident 

reports. [NIOSH 2018a, p. 6]

As there was no discussion in NIOSH 2018a of alternate dose reconstruction methods for 

exposure to alpha-emitting material that was not comingled with FAP material, the logical 

implication is that the Division of Compensation Analysis and Support (DCAS) plans to 

continue to use the ratio method to derive bounding intakes of alpha-emitting material when 

uranium or other bioassay data are not available for an individual energy employee. 

To evaluate whether the ratio method would indeed bound potential intakes of alpha-emitting 

material (specifically uranium), SC&A compiled the gross beta and gross gamma urinalysis data 

for the 32 claimants discussed in Appendix A of SC&A 2017. These 32 claims were identified in 

SC&A 2017 as working in CPP and having job titles most likely to be associated with laboratory 

areas and the final product stages where alpha-emitting material not comingled with FAP 

material would be present. These job titles included , 

, , , , and 

. SC&A then used the gross beta and gross gamma bioassay for these claims to 

calculate intakes of FAP material, which can be used in conjunction with the ratios presented in 

the INL/ANL-W internal dose technical basis document (TBD), ORAUT-TKBS-0005-5, 
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Revision 03 (NIOSH 2010a1), to arrive at intakes of uranium. The intakes of uranium using the 

ratio method can then be compared to alternate methods of calculating uranium intakes to 

determine if they are indeed bounding. 

1 It should be noted that the ratios presented in NIOSH 2010a are currently under review by the INL/ANL-W Work 

Group and are the subject of the SC&A white paper, SC&A’s Evaluation of the NIOSH Evaluation Report Proposed 

Use of Fission-Activation Product Bioassay Indicator Radionuclides (in Conjunction with ORAUT-OTIB-0054 and 

ORAUT-TKBS-0007-5) for Assessment of Fission-Activation Product and Actinide Intakes at Idaho National 

Laboratory (SC&A 2015). 

Section 2 of this report presents the evaluation of uranium intakes using the ratio method and 

available claimant bioassay compiled from the 32 claimants discussed in SC&A 2017. Both the 

individual claimant intakes are evaluated (see Section 2.2), as well as a simulated coworker 

approach in which all of the data for the 32 individuals are combined (see Section 2.3). Section 3 

presents alternate methods of calculating intakes of uranium based on air sampling data 

(Section 3.1), and uranium urinalysis (Section 3.2). Section 4 summarizes the comparison of 

derived uranium intakes using the different methods discussed in the previous two sections and 

provides SC&A’s summary recommendation. 

 
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2 SUMMARY OF CLAIMANT INTAKE EVALUATIONS OF URANIUM 

USING GROSS BETA URINALYSIS 

As noted in the introduction to this report, NIOSH’s currently proposed method for 

reconstructing doses to uranium involves evaluating the energy employee’s available gross beta 

and/or gross gamma urinalysis data and applying a ratio to obtain intakes to alpha-emitting 

material. A full description of this proposed dose reconstruction method can be found in 

Section 5.5.2 of ORAUT-TKBS-0005-5, Revision 03 (NIOSH 2010a). 

Appendix A of SC&A 2017 identified 32 claimants who worked at CPP during the period of 

interest and had job titles or other information to indicate work in and around the CPP laboratory 

facilities or final product staging areas. These are the plant areas most likely to contain source 

terms in which alpha-emitting uranium material could be found separated from the fission and 

activation products that were considered waste material. To characterize the magnitude of 

assigned uranium intakes based on the currently proposed dose reconstruction methodology, 

SC&A used the Integrated Modules for Bioassay Analysis (IMBA) program to evaluate the 

available gross beta urinalysis data2 for these claims and assuming the activity was entirely 

strontium-90 (Sr-90). Fifteen of the 32 claimants had sufficient gross beta urinalysis to 

effectively model a chronic intake of Sr-90. The derived daily intake rates of strontium were then 

used with ratios developed in the INL internal dose TBD to establish an associated uranium 

intake rate. 

2 Gross beta urinalysis was chosen for analysis over gross gamma because there were over three times more gross 

beta urinalysis samples among the 32 reviewed claims.

In addition to using IMBA to model the chronic intakes for each of the individual workers, 

SC&A simulated the development of a coworker model for the 32 claims using all available 

gross beta urinalysis for those claims (20 of 32 had at least some gross beta urinalysis results). 

Section 2.1 describes the assumptions that SC&A used in both the individual and simulated 

coworker intake calculations. Section 2.2 summarizes the results for the individual chronic intake 

calculations, and Section 2.3 discusses the simulated coworker results. For more detailed 

information on the claimant-specific intake calculations, see Attachment A.

2.1 ASSUMPTIONS USED IN INTAKE CALCULATIONS

The following subsections describe the assumptions SC&A employed in developing the intake 

calculations using IMBA. These assumptions include: 

• Development of intake evaluation regimes that reflect the assumed starting and ending 

dates for the modeled chronic intake period (see Section 2.1.1) 

• Intake mode and solubility type (see Section 2.1.2) 

• Ratio used to arrive at an intake of uranium relative to the modeled intake of Sr-90 (see 

Section 2.1.3) 

• Treatment of urinalysis results that were less than the limit of detection, including results 

that had a reported activity of zero (see Section 2.1.4) 

 
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2.1.1 Intake Evaluation Regimes 

The start date for each intake evaluation was assumed to be the first covered employment date 

during the period of interest, which is defined as February 1, 1953, through December 31, 1962. 

The start date of the period of interest is based on the fact that the first hot run at CPP occurred in 

February 1953 (NIOSH 2010b, p. 25). If the start of covered employment for a given claim 

occurred prior to February 1, 1953, then the start of the evaluated intake period was assumed to 

coincide with the first hot run at CPP. The end date for each intake evaluation was assumed to be 

the date of the final gross beta urinalysis result during the period of interest (February 1, 1953–

December 31, 1962). 

2.1.2 Intake Mode and Solubility Type 

For the purposes of calculating an intake from the available gross beta urinalysis, it was assumed 

that the material was strontium Type F and the material was inhaled. SC&A evaluated Type F 

strontium because of the following statement in the INL internal TBD:

Because the available information on the INEL Site does not indicate that 

strontium titanate (SrTiO3) was ever present at the site and because strontium 

titanate was an uncommon strontium compound, strontium only needs to be 

assessed as type F material. [NIOSH 2010a, p. 16] 

Therefore, only Type F Sr-90 was considered in this evaluation. 

2.1.3 Ratio of Uranium to Strontium 

Section 5.5.2 of the INL internal TBD (NIOSH 2010a) provides ratios of various actinide 

contaminants to gross beta and gross gamma urinalysis. SC&A’s review of the internal 

dosimetry files for the 32 claimants described in Appendix A of SC&A 2017 determined that 

gross beta urinalysis was the most common and routinely found bioassay method during the 

period of interest; therefore, this is the bioassay method analyzed in this section. Table 5-22 of 

NIOSH 2010a provides the actinide-to-Sr-90 ratios for use in dose reconstruction; these ratios 

are shown in Table 1 below for convenience. 

Specific to the Chemical Processing Plant, Section 5.5.2 of NIOSH 2010a states the following:

For the ICPP, the actinide ratios for the aluminum fuels likely provide a 

reasonable overestimate of the actinides present at the ICPP before 1971, 

because the non-aluminum fuels that were reprocessed at the ICPP before 1971 

had much lower burnups than what was assumed for the aluminum fuels. 

[page 41]

Therefore, SC&A assumed the Sr-90-to-uranium ratio for aluminum reactor fuels as shown in 

italicized in red in Table 1. 
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Table 1. Actinide-to-Sr-90 Ratios Presented in NIOSH 2010a (Table 5-22) 

Actinide

Aluminum 

Reactor Fuel 

Type Ratio 

(Isotope to Be 

Used)

Zirconium 

Reactor Fuel 

Type Ratio 

(Isotope to Be 

Used)

Stainless-Steel 

Reactor Fuel 

Type Ratio 

(Isotope to Be 

Used)

Maximum 

Reactor Fuel 

Type Ratio 

(Isotope to Be 

Used)

Actinium (Ac) 8.0E-12

(Ac-227)

1.3E-11

(Ac-227)

2.3E-10

(Ac-227)

2.3E-10

(Ac-227)

Thorium (Th) 2.4E-08

(Th-228)

6.4E-08

(Th-228)

2.3E-07

(Th-228)

2.3E-07

(Th-228)

Protactinium (Pa) 1.2E-10

(Pa-231)

1.1E-10

(Pa-231)

3.8E-09

(Pa-231)

3.8E-09

(Pa-231)

Uranium (U) 5.6E-05

(U-234)

6.2E-06

(U-236)

1.4E-03

(U-234)

1.4E-03

(U-234)

Neptunium (Np) 3.4E-06

(Np-237)

3.7E-06

(Np-237)

6.8E-07

(Np-237)

3.7E-06

(Np-237)

Plutonium (Pu) 8.7E-03

(Pu-238)

1.5E-02

(Pu-238)

3.7E-03

(Pu-239)

1.5E-02

(Pu-238)

Americium (Am) 1.4E-04

(Am-241)

3.9E-06

(Am-241)

9.0E-08

(Am-241)

1.4E-04

(Am-241)

Curium (Cm) 4.9E-05

(Cm-244)

1.8E-06

(Cm-244)

1.1E-10

(Cm-242)

4.9E-05

(Cm-244)

2.1.4 Treatment of Zero and Less than Minimum Detectable Activity 

Section 5.2.1 of NIOSH 2010a lists the minimum detectable activity (MDA) for gross beta 

urinalysis as 24,000 disintegrations per minute per day (dpm/d) for the years 1951–1953 and 

26,000 dpm/d for the years 1954 through 1960. Of the 15 claimants included in this internal dose 

assessment, only 5 (or 33%) had gross beta bioassays with reported results above 1⁄2 of the 

applicable MDA. Only 7 of the 293 gross beta bioassay samples (~2.4%) included in the analysis 

were greater than 1⁄2 the MDA. Only a single bioassay out of the 293 evaluated samples had 

results above the full MDA. To account for this, SC&A used two different analytical approaches 

to estimate the intake rate of Sr-90 from gross beta urinalysis for the individual workers. These 

analytical methods can be summarized as follows:

1. Fitted Sr-90 Intake Rate: Each non-zero analytical result was included in the IMBA 

fitting calculation regardless of whether it was greater than 1⁄2 of the MDA. Bioassay 

values that were reported as zero were excluded from the calculation but can be seen in 

the IMBA-generated urinalysis figures shown in Attachment A.

2. Adjusted Sr-90 Fitted Intake Rate: all bioassay results that were greater than 1⁄2 of the 

MDA were used “as is” in the IMBA fitting calculation. If the first and/or last bioassay 

results in the intake evaluation were less than 1⁄2 of the MDA, they were set to 1⁄2 of the 

MDA. All other bioassay results were excluded. 

For the coworker simulation presented in Section 2.3, it was necessary to first calculate the timeweighted one-person-one-statistic (TWOPOS) value for each year under evaluation in 
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accordance with the instructions in ORAUT-RPRT-0053, Revision 02, Analysis of Stratified 

Coworker Datasets (NIOSH 2014). Each of these TWOPOS values was non-zero and thus no 

adjustments were made in the coworker simulation.

2.2 SUMMARY OF DAILY INTAKE RATES OF SR-90 AND URANIUM USING INDIVIDUAL 

CLAIMANT GROSS BETA BIOASSAY 

Table 2 presents the results of the IMBA calculations for 15 claimants with extensive gross beta 

bioassay results during the period of interest. In addition to the IMBA-derived estimates of Sr-90 

intake, Table 2 shows the resulting intake rate of uranium based on the ratio method in NIOSH 

2010a. As seen in the table, estimates of individual uranium intake rates using non-zero claimant 

bioassay data (analytical method 1 as described in the previous section) ranged from 

0.25 picocuries per day (pCi/d) to 1.08 pCi/d with an average value of 0.57 pCi/d. Adjusting the 

IMBA calculation to only include values greater than 1⁄2 of the MDA and assuming the starting 

and ending bioassay was equal to 1⁄2 of the MDA (analytical method 2 as described in the 

previous section) yielded slightly larger estimates of the daily intake rate ranging from 1.36 to 

1.96 pCi/d with an average of 1.48 pCi/d. 

Table 2. Summary of IMBA Fitted Intake Rates and Resulting Uranium Intakes Using the 

Ratio Method 

Case 

ID**

Total Gross 

Beta 

Bioassay 

IMBA Fitted 

Sr-90 Type F

Intake Rate 

(pCi/d)

Resulting 

Uranium 

Intake 

(pCi/d)

Adjusted IMBA 

Fitted Sr-90 

Type F Intake 

Rate (pCi/d)

Adjusted 

Uranium 

Intake Rate 

(pCi/d)

A [redacted] [redacted] [redacted] [redacted] [redacted]

B [redacted] [redacted] [redacted] [redacted] [redacted]

C [redacted] [redacted] [redacted] [redacted] [redacted]

D [redacted] [redacted] [redacted] [redacted] [redacted]

E [redacted] [redacted] [redacted] [redacted] [redacted]

F [redacted] [redacted] [redacted] [redacted] [redacted]

G [redacted] [redacted] [redacted] [redacted] [redacted]

H [redacted] [redacted] [redacted] [redacted] [redacted]

I [redacted] [redacted] [redacted] [redacted] [redacted]

J [redacted] [redacted] [redacted] [redacted] [redacted]

K [redacted] [redacted] [redacted] [redacted] [redacted]

L [redacted] [redacted] [redacted] [redacted] [redacted]

M [redacted] [redacted] [redacted] [redacted] [redacted]

N [redacted] [redacted] [redacted] [redacted] [redacted]

O [redacted] [redacted] [redacted] [redacted] [redacted]

Average [redacted] [redacted] [redacted] [redacted] [redacted]

* Due to a break in the covered employment, these intake rates represent the weighted average of two evaluated 

intake periods. 

** Case ID is an arbitrary designation that does not have any direct connection to the claimant. Attachment B 

provides the actual NOCTS claim numbers associated with each Case ID designation for reference.

While only Type F Sr-90 intakes were evaluated due to the instructions in NIOSH 2010a and 

discussed in Section 2.1.2 of this report, it should be noted that scoping calculations indicate that 
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Type M and Type S Sr-90 would increase the associated intakes by an average factor of

approximately 3.5 and 47, respectively. 

2.3 DAILY INTAKE RATES OF URANIUM USING TWOPOS/COWORKER SIMULATION ON 

CLAIMANT GROSS BETA BIOASSAY

In addition to the individual claimant intake assessment presented in Section 2.2, SC&A also 

analyzed all gross beta urinalysis data among the group of 32 claimants originally identified in 

Appendix A of SC&A 2017. The gross beta urine data were analyzed to calculate the TWOPOS 

for each year in a simulation of how coworker data would be treated per the instructions in 

ORAUT-RPRT-0053 (NIOSH 2014). Table 3 displays the results of the TWOPOS calculation. 

Due to the relatively low number of TWOPOS results by year, SC&A did not attempt to fit the 

data to a distribution and simply used the arithmetic average TWOPOS value for intake analysis. 

For the IMBA calculation, each annual TWOPOS value was assumed to occur on July 1 of the 

year of analysis, and the inhalation intake period was assumed to extend from January 1, 1953, 

through December 31, 1960. Based on these assumptions, the daily intake rate of Type F Sr-90 

was calculated to be 6.5×103

 dpm/d with a corresponding uranium intake of 0.164 pCi/d. 

Table 3. Calculated Gross Beta TWOPOS Results by Year 

Year

Assumed 

Date of 

Sample

Number of 

TWOPOS 

Results

Minimum 

TWOPOS 

Value (dpm/d)

Maximum 

TWOPOS 

Value (dpm/d)

Average 

TWOPOS Value 

Used in Intake 

Calculation 

(dpm/d)

1953 7/1/1953 10 2,970 13,109 6,615

1954 7/1/1954 12 2,007 14,178 6,021

1955 7/1/1955 14 0 5,224 2,125

1956 7/1/1956 15 0 6,290 2,526

1957 7/1/1957 14 0 7,197 3,218

1958 7/1/1958 13 779 3,596 2,226

1959 7/1/1959 12 221 3,213 1,048

1960 7/1/1960 10 0 3,360 1,124
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3 ALTERNATE ESTIMATES FOR POTENTIAL URANIUM INTAKE 

The following two subsections demonstrate alternate methods that could be used to determine 

uranium intakes that do not utilize the ratio method described in Section 5.5.2 of the INL internal 

TBD (NIOSH 2010a). These two methods are:

1. Use of alpha air sampling data found in the Product Bottle Room and laboratory 

Room 216 

2. Calculation of uranium intakes based on the assumed uranium MDA (also known as a 

“missed dose” calculation) 

The intake rates described below can be compared to the intake rates derived using the ratio 

method described in Sections 2.2 and 2.3 above. 

3.1 URANIUM INTAKES BASED ON ALPHA AIR SAMPLING

Section 3.0 and Section 5.0 of SC&A 2017 noted several instances where long-lived alpha air 

sampling was noted. Figure 5 of SC&A 2017 presents an air sample in the Product Bottle Room 

of 5×10-11 microcuries per cubic centimeter (μCi/cm3) alpha. However, as noted in NIOSH 

2018a, that sample was recounted to correct for radon/thoron progeny, so the actual long-lived 

alpha activity was likely closer to 3×10-11 μCi/cm3. NIOSH 2018a also notes that this 

corresponds to the “general permissible concentration” for alpha emitters based on 1952 

radiological guidelines. Assuming a standard breathing rate of 1.2 m3/hr and a typical 8-hour 

work day, the daily intake of uranium can be estimated by the following formula: 

3×10-11 μCi/cm3 × 1003 cm3/m3 × 1.2 m3/hr × 8 hr/d × 106 pCi/μCi = 288 pCi/d 

Per the information in Table 7-7 of the SEC-00219 evaluation report (NIOSH 2017), the 

permissible airborne conditions were lowered to 1.7×10-11 μCi/cm3

 based on a 1954 CPP 

operating manual (NIOSH 2017). This air concentration limit would result in a daily intake rate 

of 163.2 pCi/d. Section 5.0 of SC&A 2017 noted several long-lived alpha activity air samples 

from Room 216 that ranged from 1.2×10-13 μCi/cm3

 to 9.6×10-12 μCi/cm3. The upper end of 

these samples would correspond to a daily intake rate of approximately 92 pCi/d. 

3.2 URANIUM INTAKES BASED ON THE MISSED DOSE FROM BIOASSAY 

SC&A used IMBA to calculate a hypothetical missed dose for a worker who was chronically 

exposed from the assumed start of the assumed period of interest (February 1, 1953) to the end of 

the assumed period of interest (December 31, 1962). In accordance with ORAUT-OTIB-0060, 

Revision 02, Internal Dose Reconstruction (NIOSH 2018b), the intake rate was evaluated 

assuming that a bioassay sample was submitted on the final day of the exposure and was 

assumed to be 1⁄2 of the MDA. The assumed MDA for uranium was taken from Table 5-14 of the 

INL internal TBD and is equal to 1×10-5 grams of uranium per liter (or 14 micrograms of 

uranium per day). Assuming natural uranium3 with a specific activity of 6.8×102 (pCi/mg) 

 

3 Although natural uranium was analyzed for this report, assuming low or highly enriched uranium would only 

increase the intake estimates based on the uranium urinalysis, which is on a per mass basis.
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results in an activity-based urinary excretion rate of 9.56 pCi/d at the MDA (4.78 pCi/d at 1⁄2 the 

MDA).

Using this MDA for uranium, SC&A performed a basic missed dose calculation in which a 

worker was chronically exposed via inhalation to natural uranium for the period from 

February 1, 1953, through December 31, 1962. A single bioassay measurement at 1⁄2 the MDA 

was assumed to have occurred on the last day of exposure (December 31, 1962). Based on these 

assumptions, IMBA calculated the following uranium intake rates for each relevant solubility 

type:

• Type F solubility natural uranium intake rate: 17.3 pCi/d 

• Type M solubility natural uranium intake rate: 70.5 pCi/d 

• Type S solubility natural uranium intake rate: 869 pCi/d 
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4 SUMMARY AND CONCLUSION 

Table 4 displays the calculated daily uranium intake rates using the methods described in 

Sections 2 and 3 of this report. As seen in the first 3 row entries, the current method for deriving 

uranium intakes, which involves using a ratio to gross beta intakes (assumed to be Sr-90), results 

in a daily intake rate ranging from 0.16 pCi/d to 1.48 pCi/d. The other two methods for 

determining uranium intake, which are based on either air samples or the available uranium 

bioassay limits, are at least an order of magnitude higher. Notably, the daily intake rates based on 

the missed dose from uranium urinalysis and assuming Type S natural uranium was over four 

orders of magnitude higher than the uranium intakes derived when simulating a coworker 

approach in conjunction with the ratio method. 

Table 4. Summary of Daily Uranium Intake Rates (pCi/d) using the Gross Beta Ratio 

Method, Alpha Air Sampling, and Uranium Urinalysis Missed Intake Calculations 

Uranium Intake 

Method Description

Daily Intake 

Rate of 

Uranium 

(pCi/d)

Ratio to Sr-90 

IMBA-calculated bioassay fit using coworker simulation 

approach by calculating TWOPOS annual excretion rates (see 

Section 2.3)

0.16 

Ratio to Sr-90 

Average of individual claimant assessments using IMBA and 

only considering non-zero gross beta urinalysis results (see 

Section 2.2)

0.57 

Ratio to Sr-90 

Average of individual claimant assessments using IMBA and 

assuming 1⁄2 of the MDA with the inclusion of any positive 

values above 1⁄2 of the MDA (see Section 2.2)

1.48 

Missed Dose from 

Uranium Urinalysis Solubility Type F (see Section 3.2) 17.3 

Missed Dose from 

Uranium Urinalysis Solubility Type M (see Section 3.2) 70.5 

Alpha Air Sampling Maximum Room 216 air sample from SC&A 2017: 9.6×10-12

μCi/cm3 (see Section 3.1) 92 

Alpha Air Sampling 

Permissible airborne levels of natural uranium (1.7×10-11

μCi/cm3) from a 1954 CPP operating manual (NIOSH 2017) 

(see Section 3.1)

163 

Alpha Air Sampling 

Product Bottle Room air sample from SC&A 2017 assumed to 

have decayed to the 1952 general permissible concentration of 

3×10-11 μCi/cm3 (see Section 3.1)

288 

Missed Dose from 

Uranium Urinalysis Solubility Type S (see Section 3.2) 869 

Summary Recommendation: Evidence suggests that the current dose reconstruction method for 

reconstructing uranium intakes based on a ratio of uranium intakes to intakes of FAP may not be 

a sufficiently bounding approach for workers who may have been exposed to uranium that was 

not comingled with FAP material. It would be beneficial for NIOSH to explore alternate methods 

of dose reconstruction for workers who were not directly monitored for uranium. 
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5 REFERENCES

NIOSH 2010a. Idaho National Laboratory and Argonne National Laboratory – West –

Occupational Internal Dose, ORAUT-TKBS-0007-5, Revision 03, National Institute for 

Occupational Safety and Health, Cincinnati, OH. March 2, 2010.

NIOSH 2010b. Idaho National Laboratory and Argonne National Laboratory-West – Site 

Description, ORAUT-TKBS-0007-2, Revision 04, National Institute for Occupational Safety and 

Health, Cincinnati, OH. August 2, 2010.

NIOSH 2014. Analysis of Stratified Coworker Datasets, ORAUT-RPRT-0053, Revision 02,

National Institute for Occupational Safety and Health, Cincinnati, OH. October 8, 2014.

NIOSH 2017. SEC Petition Evaluation Report – Petition SEC-00219 – Idaho National 

Laboratory, Revision 2, National Institute for Occupational Safety and Health, Cincinnati, OH. 

February 22, 2017.

NIOSH 2018a. Response to SCA-TR-2017-SEC008, “Draft Review of Internal Alpha Exposure 

Potential at CPP Prior to 1963,” National Institute for Occupational Safety and Health,

Cincinnati, OH. April 18, 2018.

NIOSH 2018b. Internal Dose Reconstruction, ORAUT-OTIB-0060, Revision 02, National 

Institute for Occupational Safety and Health, Cincinnati, OH. April 20, 2018.

SC&A 2015. SC&A’s Evaluation of the NIOSH Evaluation Report Proposed Use of FissionActivation Product Bioassay Indicator Radionuclides (in Conjunction with ORAUT-OTIB-0054 

and ORAUT-TKBS-0007-5) for Assessment of Fission-Activation Product and Actinide Intakes at 

Idaho National Laboratory, SCA-SEC-2015-0074-E1, Revision 0, SC&A, Inc., Vienna, VA, and 

Saliant, Inc., Jefferson, MD. October 26, 2015.

SC&A 2017. Draft Review of Internal Alpha Exposure Potential at CPP Prior to 1963, SCATR-2017-SEC008, Revision 0, SC&A, Inc., Arlington, VA. July 10, 2017.
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