DRAFT 

ADVISORY BOARD ON RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health 

COMPARISON OF SC&A’S BLIND DOSE RECONSTRUCTION 

TO NIOSH’S DOSE RECONSTRUCTION OF CASE 

[REDACTED] FROM THE ALLIED CHEMICAL AND DYE 

CORPORATION, NORTH CLAYMONT, DELAWARE

Contract No. 211-2014-58081

SCA-TR-DRC2014-CN[REDACTED], Rev. 0

Prepared by 

Kathleen Behling 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

December 2014 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board on 

Radiation and Worker Health (ABRWH) to maintain all possible openness in its deliberations. However, 

the ABRWH and its contractor, SC&A, caution the reader that at the time of its release, this report is predecisional and has not been reviewed by the Board for factual accuracy or applicability within the 

requirements of 42 CFR 82. This implies that once reviewed by the ABRWH, the Board’s position may 

differ from the report’s conclusions. Thus, the reader should be cautioned that this report is for 

information only and that premature interpretations regarding its conclusions are unwarranted.
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S. Cohen & Associates:

Technical Support for the Advisory Board on

Radiation & Worker Health Review of

NIOSH Dose Reconstruction Program

Document No. 

SCA-TR-DRC2014-CN[REDACTED]

Effective Date: 

Draft –December 19, 2014

Comparison of SC&A’s Blind Dose 

Reconstruction to NIOSH’s Dose 

Reconstruction of Case [Redacted] from Allied 

Chemical and Dye Corporation, North 

Claymont, Delaware

Page 2 of 20

Revision No:

0 (Draft)

Task Manager:

________________________ Date: ___________

Douglas Farver, CHP

Supersedes:

N/A

Project Manager:

________________________ Date: ___________

John Stiver, MS, CHP

Reviewer:

 U. Hans Behling

 Douglas Farver

 John Stiver

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

0 (Draft) 12/19/2014 Initial issue
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TABLE OF CONTENTS 

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

1.0 Statement of Purpose ...........................................................................................................6

2.0 Comparison of Methodology/Doses used by SC&A and NIOSH for Case 

[REDACTED]......................................................................................................................8

2.1 Occupational External Dose Calculations .............................................................10

2.1.1 Photon Doses during the Operational Period .............................................10

2.1.2 Photon Doses during the Residual Period ..................................................11

2.1.3 Occupational Medical Doses .....................................................................12

2.2 Occupational Internal Doses ..................................................................................12

2.2.1 Inhalation Doses during Operational Period ..............................................13

2.2.2 Inhalation Doses during the Residual Period .............................................15

2.2.3 Ingestion Doses ..........................................................................................16

2.2.4 Radon Doses ..............................................................................................16

3.0 Summary Conclusions .......................................................................................................18

4.0 References ..........................................................................................................................20

LIST OF TABLES 

Table 1-1. Comparison of SC&A’s Blind Dose Reconstruction to NIOSH’s Dose 

Reconstruction for Case [Redacted] ..............................................................................7

Table 2-1. Comparison of Data and Assumptions Used by NIOSH and SC&A ............................9

Table 2-2. SC&A’s ‘Method A’ Residual Photon Doses .............................................................11

Table 2-3. Comparison of Modeled External Photon Doses .........................................................12

Table 2-4. Comparison of Occupational Medical Doses ..............................................................12

Table 2-5. SC&A’s ‘Method A’ Relative Radionuclide Concentrations and Intakes ..................14

Table 2-6. SC&A’s ‘Method A’ Adjusted Intakes for Uranium and Thorium 

during the Residual Period ...........................................................................................15

Table 2-7. Comparison of Internal Doses .....................................................................................17

Table 3-1. Comparison of Total External and Internal Doses to the Lung ...................................18
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ABBREVIATIONS AND ACRONYMS 

ACP Allied Chemical Plant 

Advisory Board Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

Bq Becquerel 

CADW Chronic Annual Dose Workbook 

CATI Computer-Assisted Telephone Interview 

CF conversion factor 

DCF dose conversion factor

DOE (U.S.) Department of Energy 

DOL (U.S.) Department of Labor 

DR dose reconstruction 

EE Energy Employee 

EPA (U.S.) Environmental Protection Agency 

FIPR Florida Institute of Phosphate Research 

GM geometric mean 

GSD geometric standard deviation

hr hour 

IREP Interactive RadioEpidemiological Program 

keV kiloelectron volts 

l liter 

m

3 cubic meter

m/s meter per second 

mrem millirem

NIOSH National Institute for Occupational Safety and Health 

ORAUT Oak Ridge Associated Universities Team 

pCi picocuries 

POC probability of causationrem Roentgen equivalent man

SC&A S. Cohen and Associates (SC&A, Inc.) 

sec second 

SRDB Site Research Database 

TBD technical basis document 
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TENORM Technologically Enhanced Naturally Occurring Radioactive Materials 

U3O8 triuranium octoxide 

WLM/yr working level month per year

yr year
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1.0 STATEMENT OF PURPOSE 

Under Contract No. 200-2009-28555, SC&A was tasked by the Advisory Board on Radiation 

and Worker Health (Advisory Board) to perform six blind dose reconstructions (DRs) at the May 

21, 2013, DR Subcommittee meeting. SC&A was provided all of the Department of Energy 

(DOE) dosimetry records; the Department of Labor (DOL) correspondence, forms, and medical 

records; and the Computer-Assisted Telephone Interview (CATI) Reports that were made 

available to the National Institute for Occupational Safety and Health (NIOSH) for 

reconstructing doses in behalf of these cases. SC&A used two independent approaches to 

reconstruct occupational external and internal doses for the cases. Both approaches used the 

available dosimetry records and current guidance from NIOSH. The first approach, referred to 

as DR–Method A, used the spreadsheets and other tools developed by NIOSH to calculate the 

doses, whereas the second approach, referred to as DR–Method B, manually calculated the 

doses. 

One of the six draft blind DR reports [Blind Dose Reconstruction of Case [Redacted] from the 

Allied Chemical Plant (SC&A 2014)], was submitted to the Advisory Board and NIOSH on 

February 21, 2014. In this report, SC&A presents a comparison between SC&A’s and NIOSH’s 

DR methodologies, doses, and resultant probability of causation (POC) values for Case 

[Redacted]. Table 1-1 summarizes the external and internal occupational doses calculated by 

SC&A (using two independent methods) and the NIOSH-assigned dose for Case [Redacted]. A 

detailed comparison of the three methodologies used to calculate doses in behalf of this case is 

presented in Section 2. Section 3 of this report provides Summary Conclusions. 

It should be noted that an explanation is provided regarding the differences in doses and why 

they occurred; however, SC&A does not make any value judgments regarding which are the 

more correct approaches. It is our position that further discussions are best addressed by the DR 

Subcommittee. 
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Table 1-1. Comparison of SC&A’s Blind Dose Reconstruction to NIOSH’s Dose 

Reconstruction for Case [Redacted] 

SC&A–Method A SC&A–Method B NIOSH

External Dose (Occupational):

 ▪ Operations Dose (rem)

 - Photons 30–250 keV 0.462 NA 0.209

 - Photons >250 keV 0.579 NA 0.172

 ▪ Residual Dose (rem)

 - Photons 30–250 keV 0.086 NA 0.056

 - Photons >250 keV 0.108 NA 0.047

 ▪ Occupational Medical Dose (rem)

 - Photons 30–250 keV 1.886 NA 1.592

Internal Dose (rem):

 - Uranium/Thorium (Operational) 93.679 NA 15.106

 - Uranium/Thorium (Residual) 24.635 NA 0.088

Total 121.435 NA 17.271

Total Radon 0.812 WLM 2.115 WLM 0.214 WLM

POC 85.4% 64.1% 45.90%

 NA = not applicable
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2.0 COMPARISON OF METHODOLOGY/DOSES USED BY SC&A AND 

NIOSH FOR CASE [REDACTED] 

Case [Redacted] represents an energy employee (EE) who worked at the Allied Chemical and 

Dye Company in North Claymont, Delaware [referred to in this report as the Allied Chemical 

Plant (ACP)], from [redacted] to [redacted], and [redacted] to [redacted]. This employment 

period spans the operational period of 1950 through 1969, when the ACP was involved in a 

small-scale pilot operation recovering uranium from a phosphoric acid plant, as well as the 

residual contamination period from 1970 through 1977. According to the DOL records and the 

CATI, the EE worked as a [redacted] during both employment periods. No employee 

monitoring records or site survey records were found. The EE was diagnosed with oat cell 

carcinoma of the lung (ICD-9 Code 162.9) in [redacted].

Since no monitoring records exist for the EE and there are no technical basis documents (TBDs) 

or survey information for ACP, both SC&A and NIOSH used surrogate data for reconstructing 

external doses. Key guidance documents used by SC&A and/or NIOSH include the following: 

 ORAUT-OTIB-0043, Rev. 00, Characterization of Occupational Exposure to Radium 

and Radon Progeny during Recovery of Uranium from Phosphate Materials (ORAUT 

2006).

 Battelle-TBD-6000, Rev. 1, Technical Basis Document: Site Profiles for Atomic 

Weapons Employers that Worked Uranium Metals (Battelle 2011).

 ORAUT-OTIB-0070, Rev. 01, Dose Reconstruction during Residual Radioactivity 

Periods at Atomic Weapons Employer Facilities, (ORAUT 2012). 

 DCAS-TKBS-0002, Rev. 3, Technical Basis Document for Atomic Energy Operations at 

Blockson Chemical Company, Joliet, Illinois (DCAS 2010).

 Florida Institute of Phosphate Research (FIPR), Evaluation of Exposure to 

Technologically Enhanced Naturally Occurring Radioactive Materials (TENORM) in the 

Phosphate Industry, Publication 05-046-155 (FIPR 1998).

A summary of how these documents were used by each DR method, as well as other 

assumptions and dose parameters, is provided in Table 2-1: 
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Table 2-1. Comparison of Data and Assumptions Used by NIOSH and SC&A 

Dose Element NIOSH SC&A’s DR–Method A SC&A’s DR–Method B

External Dose:

 Operations 

Assumed 22 mrem/yr, which represents 

10% of upper-bound external dose value 

of 220 mrem/yr from Table 4-1 of 

ORAUT-OTIB-0043.

Lung Exposure (R) DCF (mode) values 

from OCAS-IG-001.

Photon energy range = 50% 30–250 keV; 

50% >250 keV.

Dose distribution = constant.

Used the geometric mean external 

dose value of 70 mrem/yr from 

Table 4-1 of ORAUT-OTIB-0043.

Lung Hp(10) DCF (mode) values 

from OCAS-IG-001.

Photon energy range = 50% 30–250 

keV; 50% >250 keV.

Dose distribution = lognormal with 

GSD of 2.0.

Not considered

Residual Period

Used same photon dose parameters as 

above for the operations. However, 

calculated residual contamination based on 

guidance in Battelle-TBD-6000 and 

ORAUT-OTIB-0043.

Used same photon dose parameters 

as above for the operations.

However, calculated residual 

contamination based on data in 

ORAUT-OTIB-0070.

Not considered

Occupational 

 Medical

Assumed an annual x-ray for each year of 

employment during operations 

([redact]–[redact], (redact]–[redact])

based on doses in ORAUT-OTIB-0006.

Dose distribution = normal with 30% 

uncertainty.

Assumed annual x-ray for each year 

of employment (both operational 

and residual periods; ([redact]–

[redact], ([redact]–[redact]) based 

on doses in ORAUT-OTIB-0006.

Dose distribution = normal with 

30% uncertainty.

Not considered

Onsite Ambient Not considered Not considered Not considered

Internal Dose:

 Inhalation

Although not explicitly stated in the DR, it 

was determined via personal 

communications (Allen 2014) and data in 

EE’s file that NIOSH used 10% of the 

maximum U-238/Th-232 (and progeny) 

values from Table 4-3 of ORAUT-OTIB0043 for the operations period

(([redact]–[redact]).

Intake for the residual period (([redact]–

[redact]) was based on operational data. 

Assumed contamination settled for one 

year and was resuspended based on a 

resuspension factor of 1E-6 m-1.

Used the uranium intake value of 

44.0 pCi/day for U-238 and 

0.605 pCi/day Th-232 from 

Table 4-3 of ORAUT-OTIB-0043, 

plus the ratio values for associated 

radionuclides from the Blockson 

TBD and ORAUT-OTIB-0043 to 

derive the potential intakes during 

the operational period.

For the residual period, these same 

intake values were adjusted for 

depletion according to ORAUTOTIB-0070.

Not considered

Ingestion

Ingestion intake based on an inhalation 

conversion factor of 0.2 from OCAS-TIB009.Not considered Not considered

Radon

Used 10% of maximum radon intake 

value (i.e., 0.0112 WLM/yr) from Table 4-

4 of ORAUT-OTIB-0043 for operations 

period.

Calculated intake for residual period using 

operations data and same assumptions as 

inhalation, as described above. 

Used best-estimate radon intake 

value of 0.036 WLM/yr from 

Table 4-4 of ORAUT-OTIB-0043 

to assign yearly radon intakes for 

the operational years, and the same 

value, adjusted for depletion rate 

according to ORAUT-OTIB-0070, 

for the residual period.

Assumed radon intake of 

4 pCi/l in 50% 

equilibrium, which 

corresponds to 

0.235 WLM/yr. The value 

of 4 pCi/l is the EPA 

guideline for radon as 

quoted in FIPR 1998, and 

was considered a 

minimizing approach.
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2.1 OCCUPATIONAL EXTERNAL DOSE CALCULATIONS

2.1.1 Photon Doses during the Operational Period 

The EE was employed for all of the operational years (i.e., 1950–1969) at ACP except for 

[redacted]. As indicated above, there were no dosimetry records or ACP-specific radiological 

survey data available for estimating the EE’s dose. Therefore, both NIOSH and SC&A’s 

Method A used generic guidance provided in ORAUT-OTIB-0043, Rev. 00, Characterization of 

Occupational Exposure to Radium and Radon Progeny During Recovery of Uranium from 

Phosphate Materials (ORAUT 2006). Table 4-1 of ORAUT-OTIB-0043 provides DRs with the 

option of assuming (1) an upper-bound exposure for plant workers of 220 mrem/yr entered into 

the Interactive RadioEpidemiological Program (IREP) as a constant value, or (2) the geometric 

mean (GM) exposure rate of 70 mrem/yr with a geometric standard deviation (GSD) of 2.0.

NIOSH’s Method

For calculating photon doses during the operational period, NIOSH assumed 10% of upperbound external dose value of 220 mrem/yr from Table 4-1 of ORAUT-OTIB-0043, or 

22 mrem/yr. This value was multiplied by an Exposure (R) to Organ (HT) dose conversion factor 

(DCF) for the lung of 0.986 for photon energies 30–250 keV and 0.842 for energies >250 keV 

from Appendix A of OCAS-IG-001, External Dose Reconstruction Implementation Guideline

(OCAS 2007). External dose values were assigned assuming 50% 30–250 keV photon energies 

and 50% >250 keV energies, as specified in Table 4-1 of ORAUT-OTIB-0043, and entered into 

IREP as a constant.

This resulted in the calculation of photon doses for each of the operational years (1950–1969), as 

shown below:

Photon 30–250 keV = Annual dose (10% upper-bound) × DCF × Percent of Energy Range

= 0.022 rem × 0.986 × 0.5

= 0.011 rem

Photon >250 keV = Annual dose (10% upper-bound) × DCF × Percent of Energy Range

= 0.022 rem × 0.842 × 0.5

= 0.009 rem

SC&A’s ‘Method A’ 

Note: Only SC&A’s ‘Method A’ calculated photon doses; SC&A’s ‘Method B’ did a partial 

dose assessment considering only the radon exposure and did not estimate any external doses.

SC&A’s ‘Method A’ also used values cited in Table 4-1 of ORAUT-OTIB-0043 for estimating 

photon doses. However, this method selected the GM value of 0.070 rem/yr, which was entered 

into IREP as a lognormal distribution with a GSD of 2.00. The Hp10 organ DCFs from 

Appendix A of OCAS-IG-001 were used to convert dose to the lung. As specified in ORAUTOTIB-0043, SC&A also used an energy distribution of 50% 30–250 keV and 50% >250% keV.
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Using these data, annual photon doses during the operational years were calculated as follows:

Photon 30–250 keV = Annual dose (GM value) × DCF × Percent of Energy Range

= 0.070 rem × 0.695 × 0.5

= 0.024 rem

Photon >250 keV = Annual dose (GM value) × DCF × Percent of Energy Range

= 0.070 rem × 0.870 × 0.5

= 0.030 rem

2.1.2 Photon Doses during the Residual Period

The residual period begins in 1970, after the completion of operations, and is assumed to last 

until 1977. The EE was employed during the residual period from [redacted] through 

[redacted]. Photon doses were calculated by NIOSH and SC&A’s ‘Method A’ as described 

below. 

NIOSH’s Method

For the residual period, NIOSH used the same assumptions for calculating photon doses that 

were used during the operational period. Therefore, for the residual years [redacted] through 

[redacted], the annual doses are identical to those calculated above during operations. For the 

year [redacted], the EE only worked [redacted] and the dose was prorated accordingly.

SC&A’s ‘Method A’

SC&A’s ‘Method A’ calculated the EE’s exposure to residual contamination using the guidance 

found in ORAUT-OTIB-0070, Dose Reconstruction during Residual Radioactivity Periods at 

Atomic Weapons Employer Facilities (ORAUT 2012), from 1970 until the EE’s last year of 

employment in [redacted]. 

ORAUT-OTIB-0070 provides guidance and adjustment factors to account for depletion of the 

source term during the residual period based on an average depletion rate of 0.00067 per day. 

This resulted in the estimate of photon doses as shown in Table 2-2. These values were entered 

into IREP as a GM with a GSD of 2.0.

Table 2-2. SC&A’s ‘Method A’ Residual Photon Doses

Year Adjustment Factor Photon Dose (rem)

30–250 keV photons >250 keV photons

[redacted] 1.000 0.024 0.030

[redacted] 0.783 0.019 0.024

[redacted] 0.613 0.015 0.019

[redacted] 0.480 0.012 0.015

[redacted] 0.376 0.009 0.011

[redacted] 0.294 0.007 0.009

A summary of modeled photon doses derived by NIOSH and SC&A for the operational and 

residual periods is presented in Table 2-3.
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Table 2-3. Comparison of Modeled External Photon Doses

Modeled External Doses SC&A-Method A 

(rem)

SC&A-Method B

(rem)

NIOSH

(rem)

Lung Dose during 

Operational Period

0.462 (30–250 keV)

0.579 (>250 keV)

1.041 Total

Not considered

0.209 (30–250 keV)

0.172 (>250 keV)

0.381 Total

Lung Dose during 

Residual Period

0.086 (30–250 keV)

0.108 (>250 keV)

0.194 Total

Not considered

0.056 (30–250 keV)

0.047 (>250 keV)

0.103 Total

The operational period photon dose calculated by SC&A’s ‘Method A’ is 2.7 times higher than 

the NIOSH-assigned photon dose. Although both methods used dose data cited in Table 4-1 of 

ORAUT-OTIB-0043, SC&A’s ‘Method A’ assumed the EE’s exposure rate was more likely 

represented by the GM of 70 mrem/yr to workers located at gypsum stacks, while NIOSH 

assumed a lower exposure represented by 10% of the upper-bound value of 0.220 rem/yr.

Although two different approaches were used to calculate doses during the residual period, the 

NIOSH-assigned and SC&A total doses differed by less than a factor of 2.

2.1.3 Occupational Medical Doses

NIOSH and SC&A’s ‘Method A’ calculated an occupational medical dose from diagnostic x-ray 

procedures required as a condition of employment, even though DOE records contained no 

diagnostic x-ray records from ACP. In the absence of records, both methods used guidance 

provided in the Technical Information Bulletin: Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Procedures, Rev. 04 (ORAUT 2011). NIOSH assumed an annual 

medical x-ray procedure for each year of employment during the Atomic Energy Commission 

(AEC) contract period. SC&A assumed the EE received an annual x-ray exam for all years of 

employment, which included the operational and residual periods.

A comparison of medical doses derived by the SC&A and NIOSH methods is presented in 

Table 2-4.

Table 2-4. Comparison of Occupational Medical Doses

Occupational Medical Doses SC&A-Method A 

(rem)

SC&A-Method B

(rem)

NIOSH

(rem)

Lung Dose 1.886 Not considered 1.592

SC&A’s ‘Method A’ doses are slightly higher than those calculated by NIOSH, strictly because 

NIOSH did not assign medical doses during the residual period (i.e., [redacted]– [redacted]). 

2.2 OCCUPATIONAL INTERNAL DOSES

No records of bioassay monitoring results were found for the EE’s employment at ACP. 

However, both NIOSH and SC&A’s ‘Method A’ assumed that the EE was chronically exposed 

to material during the operational period of uranium recovery and from residual contamination, 

as discussed below.
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2.2.1 Inhalation Doses during Operational Period

NIOSH’s Operational Inhalation Dose

For calculating inhalation doses during the operational period, a professional judgment was made 

by NIOSH, which assumed 10% of the maximizing intake value of 8.17 pCi/hr for U-238 in 

equilibrium with its daughters, and 0.112 pCi/hr for Th-232 and daughters from Table 4-3 of 

ORAUT-OTIB-0043. The basis for this assumption was provided by David Allen (Allen 2014):

. . . The primary issue is that we assumed 10% of the exposure associated with 

OTIB-43 rather than 100%. This is due to the fact that this was a bench scale 

operations (produced a few pounds of U) and OTIB-43 was based on large scale 

production. This bench scale operation tried to extract uranium from phosphoric 

acid but there is no indication the phosphoric acid was made there. SRDB 

#99079 (pg 4 of 4) indicates the No. Claymont plant was part of the General 

Chemical division while the Nitrogen Division is the one that produced a wide 

line of fertilizers so it is very possible the phosphoric acid was brought into the 

lab rather than created there.

Using the 10% assumption, NIOSH calculated inhalation from U-238 and Th-232 during 

recovery operations as detailed below:

 U-238:

8.17 pCi/hr × 2,000 hr/yr × 0.1 (factor for bench scale vs. production scale)/365 d/yr 

= 4.478 pCi per calendar day

 Th-232:

0.112 pCi/hr × 2,000 hr/yr × 0.1 (factor for bench scale vs. production scale)/365 d/yr 

= 0.0613 pCi per calendar day

All U-238 and daughter radionuclides, with the exception of Ra-226, and Th-232 and daughter 

radionuclides were assumed to be Type S solubility. The most claimant-favorable solubility type 

for Ra-226 was Type M. The above-cited intakes were entered into the Chronic Annual Dose 

Workbook (CADW) and resulted in a dose due to inhalation of 15.105 rem.

SC&A’s ‘Method A’ Operational Inhalation Dose

Since there are no ACP survey records or EE dosimeter records, SC&A determined that the wet 

chemical phosphoric acid treatment process used at ACP is similar to the process used in 

Building 55 at the Blockson Chemical Company site. Therefore, using the Blockson Site Profile 

(DCAS 2010) and ORAUT-OTIB-0043 (ORAUT 2006), the following assumptions were used to 

calculate an inhalation dose during the operational period.

(1) 85% of U reports to phosphoric acid (DCAS 2010); ORAUT 2006 cites “approximately 

86%.”

(2) 4% of Ra-226 reports to acid (DCAS 2010).
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(3) Thorium reports to the acid in same proportion as uranium (DCAS 2010, ORAUT 2006).

(4) U-238:Th-232 radioactivity ratio in Blockson’s rock was 30:1. However, ORAUT 2006

uses a U-238/Th-232 ratio of 72:1 based on material averages from several facilities. 

Since the exact composition of the Allied material is unknown, the U-238/Th-232 ratio of 

72:1 was used. Th-232 progeny are assumed to be in equilibrium. Although most of the 

Ra-228 would have been separated and removed with the phosphogypsum, it is assumed 

to be in equilibrium with Th-232 for dose modeling to allow for ingrowth over the 

operational and residual contamination period (DCAS 2010).

(5) Pb-210 and Po-210 assumed to report to the acid the same as U-238 (DCAS 2010, 

ORAUT 2006).

(6) The daily 8-hour U-238 intake (in equilibrium with progeny) and Th-232 intake (in 

equilibrium with progeny) are 44.0 pCi/day (1.63 Bq/day) and 0.605 pCi/day 

(0.0224 Bq/day), respectively (using best-estimate hourly intake values from Table 4-3 of 

ORAUT 2006).

(7) The U3O8 product produced from wet phosphoric acid by filtering the precipitated 

uranium most closely corresponds to the clearance rate associated with Type M uranium 

material (DCAS 2010).

(8) Thorium could have been Type M or Type S, and polonium could have been F or M. 

Therefore, the thorium and polonium solubility types were selected based on the types 

that deliver the largest dose to the target organ (DCAS 2010). For this case, the solubility 

types of thorium and polonium are S and M, respectively. Pb-210 is Type F.

These assumptions result in the following ratios and intakes:

Table 2-5. SC&A’s ‘Method A’ Relative Radionuclide Concentrations and Intakes

Radionuclide Ratio to U-238 Daily Intake Solubility 

pCi/day Bq/day Type

U-238 1 44.0 1.628 M

U-234 1 44.0 1.628 M

Th-230 1 44.0 1.628 S

Po-210 1 44.0 1.628 M

Pb-210 1 44.0 1.628 F

Ra-226 4% = 0.040 1.76 0.0765 M

Th-232 1/72 = 0.014 0.605 0.0224 S

Th-228 1/72 = 0.014 0.605 0.0224 S

Ra-228 1/72 = 0.014 0.605 0.0224 M

Using the CADW and the daily intake values and solubility types cited in Table 2-5, SC&A 

calculated an inhalation dose of 93.68 rem for the operational period. Annual internal doses 

were entered as lognormal distributions with an uncertainty of 1.270.
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2.2.2 Inhalation Doses during the Residual Period

NIOSH’s Residual Inhalation Dose

Residual period inhalation doses were based on operational period intake values. Calculations 

for inhalation of U-238 are shown below.

 4.478 pCi/calendar day × 365 days/1.2 m3per hr/2,000 hrs = 0.681 pCi/m3

The 0.681 pCi/m3 was assumed to settle for a full year at a rate of 0.00075 m/s to produce 

16,107.7 pCi/m2of contamination (0.681 pCi/m3 × 0.00075 m/s × 365 days × 24 hrs ×

3,600 sec.). It was assumed that the contamination was resuspended using a resuspension factor 

of 1E-6 m-1to get 0.0161 pCi/m3. This resulted in the following inhalation rate for U-238 and 

daughters:

0.0161 pCi/m3 × 1.2 m3/hr × 2,000 hrs/365 days = 0.106 pCi/calendar day

Using the Th-232 operational intake of 0.0613 pCi/calendar day and the above-cited settling and 

resuspension factors, NIOSH calculated a residual period inhalation rate of 

1.45E-03 pCi/calendar day for Th-232 and daughters. 

The NIOSH-calculated uranium and thorium intake values were entered into the CADW with the 

solubility Type S for all radionuclides except Ra-226, which was entered with solubility Type M. 

This resulted in an inhalation dose of 0.088 rem.

SC&A’s ‘Method A’ Residual Inhalation Dose

SC&A’s ‘Method A’ calculated the EE’s internal exposure to residual contamination using the 

guidance found in ORAUT-OTIB-0070, Dose Reconstruction during Residual Radioactivity 

Periods at Atomic Weapons Employer Facilities (ORAUT 2012), from 1970 until the EE’s last 

year of employment in [redacted]. ORAUT-OTIB-0070 provides adjustment factors based on 

an average depletion rate of 0.00067 per day. Table 2-6 shows how the factors were applied to 

the U-238 and Th-232 intakes; these adjustment factors were also applied to the daughter

radionuclides.

Table 2-6. SC&A’s ‘Method A’ Adjusted Intakes for Uranium and Thorium

during the Residual Period

Year

Adjustment 

Factor

Adjusted Intake (pCi/d)

U-238 Th-232

[redacted] 1.000 44.0 0.014

[redacted] 0.783 34.4 0.011

[redacted] 0.613 27.0 0.009

[redacted] 0.480 21.1 0.007

[redacted] 0.376 16.5 0.005

[redacted] 0.294 12.9 0.004

Using the CADW and the daily intake values cited in Table 2-6, along with solubility types cited 

in Table 2-5, SC&A calculated an inhalation dose of 24.634 rem for the residual period. Annual 

internal doses were entered as lognormal distributions with an uncertainty of 1.270.
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2.2.3 Ingestion Doses

Only NIOSH calculated doses associated with the ingestion pathway. Their calculation method 

used guidance in OCAS-TIB-009, Estimation of Ingestion Intakes, which states that daily 

ingestion dose should be calculated by assuming 0.2 times the activity per cubic meter of air and 

10% of the maximum intake values cited in Table 4-3 of ORAUT-OTIB-0043. Shown below are 

the calculations used for deriving U-238 and Th-232 intakes during the operational and residual 

periods.

 U-238Operations = 8.17 pCi/hr/1.2 m3/hr × 0.2 (TID-9 CF) × 0.1 (frac. process mat’l)

= 0.136 pCi/calendar day

 Th-232Operations = 0.112 pCi/hr/1.2 m3/hr × 0.2 (TID-9 CF) × 0.1 (frac. process mat’l)

= 0.0019 pCi/calendar day

 U-238Residual = 1.61E-02 pCi/m

3 × 0.2 (TID-9 CF) = 3.22E-03 pCi/ calendar day

 Th-232Residual = 2.21E-04 pCi/m

3 × 0.2 (TID-9 CF) = 4.42E-05 pCi/ calendar day

Using the CADW, the above-cited intakes, and the same solubility types as for the inhalation 

pathway, NIOSH calculated an ingestion dose that was less than 0.001 rem. Therefore, this dose 

was not included in the IREP input sheet.

2.2.4 Radon Doses

All DR methods (i.e., NIOSH, SC&A’s ‘Method A’ and SC&A’s ‘Method B’) calculated dose 

from exposure to radon and daughters. Their calculation methods are summarized below.

NIOSH’s Radon Exposure Estimates

NIOSH assumed the EE was exposed to radon during the operational and residual periods. 

Doses were calculated based on 10% of maximizing WLM/yr values cited in Table 4-4 of 

ORAUT-OTIB-0043. This resulted in the assignment of 0.0112 WLM/yr for operational years 

of 1950 through 1969. The WLM/yr values for the residual period years of 1970 through 

[redacted] were derived by applying settling and source term adjustment factors and resulted in

2.65E-04 WLM/yr. A total of 0.214 WLM of lung exposure to radon progeny was assigned for 

the operational and residual periods.

SC&A’s ‘Method A’ Radon Exposure Estimate

Radon exposure was also assessed by SC&A’s ‘Method A’ using guidance from ORAUT-OTIB0043, Section 4.2. The best-estimate value of 0.036 WLM/yr cited in Table 4-4 was selected for 

assigning exposure during the operational period. In the absence of any residual radon 

information, the adjustment factors shown in Table 2-6 were applied to the annual radon 

exposure of 0.036 WLM/yr for years [redacted]– [redacted]. The EE’s total radon exposure is 

0.812 WLM.
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SC&A’s ‘Method B’ Radon Exposure Estimates

SC&A’s ‘Method B’ elected to derive a minimized dose to the lung by only considering radon

exposures at a very low level, i.e., 4 pCi/L, which translates to 0.235 WLM/yr at 50% 

equilibrium. ‘Method B’ based this approach on a 1998 report by the FIPR on dose at Florida 

phosphate plants that processed phosphate rock, and a 1978 U.S. Environmental Protection 

Agency (EPA) report of exposure to workers at an Idaho phosphate plant (FIPR 1998; EPA 

1978). The annual 0.235 WLM value was entered into IREP for only 9 years of employment 

(i.e., [redacted]– [redacted], [redacted]– [redacted]). Without considering other internal and 

external exposures, this radon exposure was sufficient to result in a POC of greater than 50%.

A summary of the internal doses calculated by NIOSH and SC&A’s ‘Method A’ and ‘Method B’ 

is shown in Table 2-7.

Table 2-7. Comparison of Internal Doses

Internal Doses SC&A ‘Method A’ 

(rem)

SC&A ‘Method B’

(rem)

NIOSH

(rem)

Total Uranium/Thorium for 

Operational and Residual Periods 118.314 NA 15.194

Radon Exposure 0.812 WLM 2.115 WLM 0.214 WLM

A comparison of total internal dose values for uranium and thorium derived by NIOSH and 

SC&A’s ‘Method A’ varies significantly, with SC&A’s ‘Method A’ doses more than 7.7 times 

higher than the NIOSH-assigned values. As described above, both DR methods used guidance in 

ORAUT-OTIB-0043; however, NIOSH assumed 10% of the maximum internal intakes, while 

SC&A used the best-estimate values provided in Table 4-3.

In addition, radon exposures estimated by all three DR methods resulted in significantly different 

WLM values. Although both NIOSH and SC&A’s ‘Method A’ used guidance in ORAUTOTIB-0043, ‘Method A’ calculated a total WLM value that is 3.8 times higher than the NIOSH 

value. This variance in exposure estimates can once again be explained by the difference in 

professional judgment used by the two approaches. Namely, NIOSH assumed 10% of the 

maximum radon exposure values cited in Table 4-4 of ORAUT-OTIB-0043, and SC&A assumed 

best-estimate values were appropriate. Using radon exposure data provided by FIPR and EPA,

SC&A’s ‘Method B’ derived a total WLM value that was nearly 10 times higher than the value 

assigned by NIOSH.
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3.0 SUMMARY CONCLUSIONS

A comparison of the total external and internal doses and resultant POCs calculated by SC&A’s

‘Method A’ and ‘Method B,’ and NIOSH in behalf of Case [Redacted] is presented in Table 3-1.

Table 3-1. Comparison of Total External and Internal Doses to the Lung

Total Doses SC&A-Method A 

(rem)

SC&A-Method B

(rem)

NIOSH

(rem)

External Lung Doses:

 - Operational Period

 - Residual Period

 - Occupational Medical

1.041

0.194

1.886

NA

NA

NA

0.381

0.103

1.592

Internal Lung Doses:

 - Operational Period

 - Residual Period

93.679

24.635

NA

NA

15.106

0.088

Total Lung Dose 121.435 NA 17.271

Total Radon 0.812 WLM 2.115 WLM 0.214 WLM

POC 85.4% 64.1% 45.90%

NA = Not applicable.

As shown in Table 3-1, the three DR methods resulted in doses and radon exposure values that 

are substantially different and, for the two SC&A methods, produced POC values great than 

50%. Primary differences in the doses involve dose reconstructor decisions associated with the 

following selection of model parameters and assumptions:

 Assumptions regarding external dose values cited in Table 4-1 of ORAUT-OTIB-0043

 – NIOSH used 10% of the “upper bound for exposures to plant workers”

 – SC&A’s ‘Method A’ selected the GM dose associated with “exposure from work 

located at gypsum stacks”

 Selection of organ DCF values cited in External Dose Reconstruction Implementation 

Guideline (OCAS-IG-001)

 – NIOSH used DCF values associated with the lung Exposure (R)

 – SC&A’s ‘Method A’ used the Hp(10) to Organ DCF for the lung

 Assignment of occupational medical doses 

 – NIOSH assigned an annual x-ray exam for operational years only 

 – SC&A’s ‘Method A’ assigned an annual x-ray exam for all years of employment

 Assumptions regarding internal dose values cited in Table 4-3 of ORAUT-OTIB-0043

 – NIOSH used 10% of the maximizing DR approach intake values for U-238/Th-232

 – SC&A’s ‘Method A’ used the intake values associated with the best-estimate DR 

approach
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 Assumptions regarding radon exposure values cited in Table 4-4 of ORAUT-OTIB-0043

 – NIOSH used 10% of the maximizing DR approach WL values

 – SC&A’s ‘Method A’ used the best-estimate WL values
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4.0 REFERENCES

Allen, D., 2014. Email from David Allen (NIOSH) To: Kathleen Behling (SC&A)

RE: Blind/OTIB-82. July 8, 2014.

Battelle 2011. Technical Basis Document: Site Profiles for Atomic Weapons Employers that 

Worked Uranium Metals, Battelle-TBD-6000, Rev. 1. June 17, 2011.

DCAS 2010. Technical Basis Document for Atomic Energy Operations at Blockson Chemical 

Company, Joliet, Illinois. DCAS-TKBS-0002, Rev. 03, Division of Compensation Analysis and 

Support, Cincinnati, Ohio. December 20, 2010.

EPA 1978. Radiological Surveys of Idaho Phosphate Ore Processing – The Wet Process Plant, 

U.S. Environmental Protection Agency (EPA), Technical Note ORP/LV-78-1, Las Vegas, 

Nevada, April 1978. (SRDB Ref ID 30886)

FIPR 1998. Evaluation of Exposure to Technologically Enhanced Naturally Occurring 

Radioactive Materials (TENORM) in the Phosphate Industry, Florida Institute of Phosphate 

Research (FIPR), Publication 05-046-155, 1998.

OCAS 2004. Estimation of Ingestion Intakes, Rev. 0, National Institute for Occupational Safety 

and Health, Office of Compensation Analysis and Support, Cincinnati, Ohio. April 13, 2004.

OCAS 2007. External Dose Reconstruction Implementation Guideline, OCAS-IG-001, Rev. 3, 

National Institute for Occupational Safety and Health, Office of Compensation Analysis and 

Support, Cincinnati, Ohio. November 21, 2007.

ORAUT 2006. Characterization of Occupational Exposure to Radium and Radon Progeny 

during Recovery of Uranium from Phosphate Materials, ORAUT-OTIB-0043, Rev. 00, Oak 

Ridge Associated Universities Team, Cincinnati, Ohio. January 6, 2006.

ORAUT 2011. Technical Information Bulletin: Dose Reconstruction from Occupationally 

Related Diagnostic X-ray Procedures, ORAUT-OTIB-0006, Rev. 04, Oak Ridge Associated 

Universities Team, Cincinnati, Ohio. June 20, 2011.

ORAUT 2012. Technical Information Bulletin: Dose Reconstruction during Residual 

Radioactivity Periods at Atomic Weapons Employer Facilities, ORAUT-OTIB-0070, Rev. 01, 

Oak Ridge Associated Universities Team, Cincinnati, Ohio. March 5, 2012.

SC&A 2014. Blind Dose Reconstruction of Case [Redacted] from the Allied Chemical Plant, 

SCA-TR-BDR2014-CN[redacted], Rev. 0. SC&A, Inc., Vienna, Virginia. February 21, 2014.