Office of Compensation Analysis and 

Support

Document Number: Battelle-TBD-6001

Appendix D

Effective Date: 04/30/2010

Revision No. 1 

Site Profiles for Atomic Weapons Employers that Refined 

Uranium and Thorium - Appendix D

United Nuclear Corp.

Page 1 of 12

Document Owner: David Allen

Approval:_Signature on file _____ Date: 04/30/2010

 J.W. Neton, Associate Director for Science

Supersedes: 0

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

3/14/2008

EFFECTIVE 

DATE

3/14/2008

REV. NO. 

0 

DESCRIPTION

Appendix to Battelle-TBD-6001 describing 

the use of the TBD for claims at United 

Nuclear Corp.

4/30/2010 4/30/2010 1 Update to reflect additional site specific 

bioassay and external dose data and change 

in facility covered dates.

D.1 Introduction 

This document serves as an appendix to Battelle-TBD-6001, Site Profiles for Atomic 

Weapons Employers that Refined Uranium and Thorium. This appendix describes the 

results of document research specific to this site. Where specific information is lacking, 

research into similar facilities described in the body of this Site Profile is used.

D.2 Site Description

Throughout its history, United Nuclear Corporation (UNC) located in Hematite, 

Missouri, manufactured uranium metal and uranium compounds from natural and 

enriched uranium for use as nuclear fuel. The fuel was manufactured for use by the 

federal government and government contractors and by commercial and research 

reactors. Research and development was also conducted at the plant as well as uranium 

scrap recovery1

. 

Mallinckrodt Chemical Works built the plant which became operational in July 1956. 

The plant initially produced uranium products for use in the naval nuclear fuel program. 

Ownership transferred to United Nuclear Corporation in May of 1961. In 1970 the 

operator became Gulf United Nuclear Fuels Corporation, which was a joint venture 

between UNC and Gulf Nuclear Corporation. The facility was closed in 1973 and sold to 

Combustion Engineering in May of 1974. In 1989 Asea Brown Boveri (ABB) began 
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operating the facility as ABB Combustion Engineering. In April of 2000, Westinghouse 

purchased the nuclear operations of ABB which included the Hematite facility1

. 

Most of the operations at the facility involved manufacturing reactor fuel for the naval 

nuclear program or commercial reactors rather than weapons work. However, the 

Department of Energy has determined that the facility performed uranium scrap recovery 

operations for the weapons program between 1958 and 19732

. Therefore, all sources of 

radiation exposure during this time period must be accounted for, including that from 

commercial and naval reactor work. After 1973 only radiation dose caused by 

contamination remaining from weapons related work must be accounted for (residual 

contamination).3

 

D.3 Occupational Medical Dose

No documentation regarding occupational medical dose specific to UNC was found. 

Information to be used in dose reconstructions for which no specific information is 

available is provided in ORAUT-OTIB-0006, the dose reconstruction project technical 

information bulletin covering diagnostic X-ray procedures. 

D.4 Occupational External Dose

Personnel photon exposures at UNC were directly related to work with enriched uranium, 

primarily from operations in the Red Room, Blue Room, Green Room, and Item Plant. 

These processes are known to have been a routine component of plant operations during 

the entire covered period from 1958-1973 based on a review of plant radiation safety 

manuals and procedures, plant correspondence and operations logs, and film badge 

results. Uranium in various forms and compounds was used. Accordingly, external 

exposures to photon radiation would have resulted from radionuclides in the uranium 

decay chain. Exposure to beta sources at UNC would have resulted principally from 

uranium decay products and was likely encountered during routine scrap recovery 

operations and other production activities at the site. Potential exposure to neutrons 

would have resulted from operations with enriched uranium. 

D.4.1 Co-worker gamma and beta

Dosimetry data for monitored workers at United Nuclear Corporation were evaluated. 

According to the 1962 UNC Health Physics Manual, all personnel assigned to the plant 

were issued a film badge upon entry to the facility. The available personnel monitoring 

data include data that represent the maximally-exposed work group and work scenario 

over the operational period. Film badge data was available for the operational years 

(1958-1973). Data for 1958 through 1960 exists only in reports summarizing exposures. 

Dose estimates based on coworker exposures were developed for use in the evaluation of 

external dose for certain claimants potentially exposed to workplace radiation, but with 

no or limited monitoring data. The Limit of Detection (LOD) is 40 mrem for beta and 40 

mrem for gamma in accordance with NIOSH technical information bulletins and 

procedures.4,5,6

 

The external dosimetry data for 1958 indicates there was a combination of weekly and 

monthly exchanges.7

 The average reported dose received during 1958 was 80 mrad beta 

and 36 mrad gamma. The maximum reported dose received was 2525 mrad beta and 380 

mrad gamma (for 6 months).7

 A yearly missed dose based on 20 mr (LOD/2) per week 
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was added to the average values to derive a geometric mean (GM) for both the beta dose 

and the gamma dose. The yearly missed dose was also added to the maximum values 

(adjusted to an annual dose) to produce values at the 95th percentile. 

During 1959 all weekly exchanges were changed to monthly. The data for 1959 indicates 

the maximum beta dose received was 240 mrem in a month. The average beta was 90 

mrem in a month. The maximum gamma was 15 mrem in a month.7

 Since it is not clear 

when in 1959 the exchange frequency changed, missed dose was assigned based on a 

weekly exchange frequency for the entire year. Missed dose was added to the maximum 

recorded values and used to represent the 95th percentile dose. Missed dose was added to 

the average beta dose and used to represent the GM beta dose. The GM gamma dose was 

assigned based on the GM beta dose adjusted using the average beta to gamma ratio 

between 1961 and 1965. 

The data for 1960 is limited. One report indicated the maximum gamma dose received 

prior to 1961 was 100 mrem in a week7 while the typical reading was below 50 mrem. 

The maximum value was assumed to occur every week and assigned as the 95th

percentile. The typical reading was also assumed to occur every week and assigned as 

the GM. A Beta to gamma ratio based on 1961 to 1965 data was used to derive the 

annual beta doses. 

The data for 1961 through 1965 was reported as separate beta and gamma readings. This 

data was used to assign gamma and beta dose for each year at the GM and the 95th

percentile. 

For 1966 through 1973 only the total dose (combining beta and gamma) was reported. 

Using the 1961 through 1965 data, a ratio of beta to gamma was calculated for each of 

those years, and the average of that ratio was applied to the 1966 through 1973 data to 

derive annual beta and gamma dose at the GM and the 95th percentile. 

Unmonitored workers that were routinely involved with hands on uranium work will be 

assigned the 95th percentile doses. Unmonitored workers routinely in the area but not 

routinely handling uranium directly (supervisors, laborers, etc.) will be assigned the dose 

distribution assuming a lognormal distribution. Those not routinely in the areas where 

uranium was handled will be assigned the GM of the distribution as a constant. The 

annual external dose values to be applied each year for unmonitored workers are listed in 

Tables D2 and D3. 

D.4.2 Co-worker neutron

A routine external monitoring program for neutron exposure did not exist at UNC, but 

there was a potential for personnel neutron exposures from the uranium-enrichment 

operations.7

 The site received and processed uranium compounds containing fluorine and 

oxygen. The site was authorized by the AEC to receive enriched uranium hexafluoride 

and typically requested quantities of 50 to 100 kilograms (kg) on a quarterly basis with 

enrichments ranging from approximately 20% to 93%. 

The maximum amount of 20% enriched uranium handled at any one time is assumed to 

be 100 kg.7

 The maximum amount of 93% enriched uranium being handled at any one 

time is assumed to be 50 kg due to criticality concerns. Conversion factors from 

ORAUT-OTIB-0024 for naturally enriched uranium without progeny were used to 

estimate the neutron dose rate. The conversion factors were adjusted based on the 
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specific activity of naturally enriched uranium compared to the specific activity of 20% 

and 93% enriched uranium. 

Operators are assumed to be working within 1 foot of the enriched uranium fluoride 1⁄2 

the time (1000 hours per year). Supervisors are assumed to be working within 3 feet of 

the enriched uranium fluoride 1⁄2 the time (1000 hours per year). Both are assumed to be 

working with 20% enriched uranium and 93% enriched uranium for equal amounts of 

time. Others will be assumed to receive an annual dose equal to half the supervisors’ 

dose. Annual dose values are listed in Table D.4. These values represent a bounding 

estimate and should be considered a constant. 

D.4.3 Measured Dose

If external dosimetry data is available for the employee, the data should be used to

reconstruct external dose received. The external dosimetry data was reported as a mix of 

monthly and yearly exposures. For dose reconstruction purposes, unless otherwise 

indicated, exchange frequencies are considered to be weekly for 1958 and 1959 and 

monthly for all other years. 

 Photons

Dose Reconstructors should include an uncertainty factor of 1.3 multiplied by the 

measured photon dose, assume a constant distribution, and utilize the Exposure ( R ) to 

Organ Dose (HT), dose conversion factors from IG-001.8

 The photon radiation should be 

considered 50% in the 30-250 keV energy range and 50% in the greater than 250 keV 

energy range. 

Electrons

From 1961 through 1965 beta and photon exposures were measured separately. 

Beginning in 1966 the photon dose and the beta dose were reported as a sum of the beta 

and photon dose. For combined doses, 25% should be considered photon dose and 75% 

considered beta dose. This is based on the ratio of photon to beta dose from the 1961-

1965 data. The guidance contained in ORAUT-OTIB-00175 for assignment of skin dose 

from penetrating and nonpenetrating radiation should be followed. The nonpenetrating 

radiation is predominantly in the E >15 keV energy range. 

D.4.4 Missed dose

Missed dose applies to workers who were monitored but had results below the Limit of 

Detection (LOD) of their personal radiation monitor. An acceptable, favorable to 

claimant estimate of missed dose is one-half the LOD multiplied by the number of zero 

dose results for dose results below one-half the LOD. A GSD of 1.52 should be applied 

to missed dose. For summary data, recorded dose should be assumed to have occurred 

during only one monitoring period and missed dose assigned for the remaining periods.

 

D.5 Occupational Internal Dose

Bioassay and air monitoring data from UNC identified the potential for internal radiation 

exposures to workers. The primary source of this exposure was deposition of alphaemitting materials via inhalation and ingestion of airborne uranium and thorium (and 

progeny). 
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 D.5.1 Uranium 

There is urinalysis data for workers at UNC available for most years from 1958 through 

1973. This data, however, generally does not separate people into job categories. The 

available samples were analyzed on an annual basis to determine the parameters of a 

lognormal distribution. The geometric mean for each year was compared to the urine 

concentrations predicted from the original intake calculations in revision 0 of this 

appendix. The predicted urine concentrations for operators overestimated most annual 

results for both type M and type S intakes. The predicted concentrations for supervisors 

were at the low end of the annual results for both solubility types. This is consistent with 

data that has known job categories. Results from the operators are consistently higher 

than those from the supervisors and other job categories. Since the original intake 

estimate was able to separate operators from others and appears to be consistent with 

additional data, the original intake estimate, as described below, will be used.

If bioassay data is available for the employee, the data should be used to estimate intakes. 

Coworker Internal Dose from Uranium

Urinalysis samples starting in late 1962 revealed some operators had higher intakes than 

previously believed based on air sampling7

. Several improvements were made and the 

samples were reduced by mid 1963. Therefore, the urinalysis results were broken into 

four categories. The categories were operators prior to June 1963 and after June 1963 as 

well as “others” prior to June 1963 and after. All the samples in each group were used to 

determine parameters of a lognormal distribution of urine samples. From these 

distributions, intake rates were determined assuming first type M solubility material and 

then type S material for each of the four categories. In determining the intakes, a 

constant chronic intake was assumed from 1/1/1958 (the first day of the covered work) to 

6/13/1963 (the day before the sample date showing a significant decrease in results). The 

second period assumed a constant chronic intake from 6/14/1963 to 12/31/1973 (the last 

day of the contract period). 

The resulting intakes are the geometric mean of a lognormal distribution. The intake 

rates along with the geometric standard deviations (GSD) are listed in Table D.1. When 

the calculated GSD was less than three, the value of 3.0 was substituted in order to 

account for the uncertainty in the biokinetic models used in dose estimation. The 

geometric mean of the supervisor intake will be used for people categorized as other. 

Since this category is reserved for people that do not routinely enter the production areas, 

this value represents a bounding estimate and should be considered to be a constant.

These assumed intakes were compared with a sampling of the individuals for which both 

air sample data and urinalysis data exist4

. The geometric mean of the assumed intakes 

underpredicted urinalysis results for the people with higher sample results but 

overpredicted those with lower results. This indicates good agreement with the 

distribution. However, the calculated intakes in Table D.1 for type S material were 

considerably higher than the air sample results for these people. The intakes calculated 

for type M material did, however, match the air sample values reasonably well. The 

reports containing these values imply the most significant difficulty they had in 

controlling intakes involved the metal reduction area. This area would contain both type 

M and type S material and it is likely the highest urinalysis results came from type M 

exposures. However, it is not possible to determine if the urine sample results from 

1962-1964 involved only type M material or less soluble type S material. Therefore, 
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Appendix D

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since both were available at the site, intakes rates for both were determined and the dose 

reconstructor should apply the more favorable of the two. 

Internal Dose from Uranium Using Bioassay Results

Beginning in 1958, UNC operating personnel submitted urinalysis samples on a three- to 

six-month basis dependent on specific work assignments. Sample analysis incorporated 

gross activity counting methods to determine uranium activity levels. UNC evaluated 

bioassays by different analytical protocols, including enriched and fluorometric uranium 

techniques. Total uranium concentrations were measured and reported with a minimum 

detectable concentration (MDC) of 10 dpm per liter. 

D.5.2 Thorium

Approximately nine tons of natural thorium was on site during the covered period for a 

single, specific project conducted in 1964 in the Pellet Plant.7

 Thorium dioxide powder 

was blended with uranium dioxide powder to produce Th-U fuel pellets to be used in fuel 

assemblies for breeder reactors.7

 Each pellet consisted of a mixture of 97 % ThO2 and 

3% UO2, compressed and sintered (heated) in a furnace. A total of 210 air samples, 

designated specifically as “ThO2” (thorium dioxide), were collected throughout the Pellet 

Plant project in 1964 and the results were reported in uCi/ml. These samples were used 

to determine parameters of a lognormal distribution. From this distribution, intake rates 

were determined. Operators who routinely handled thorium or operated thorium 

processing equipment will be assigned the 95th percentile of this distribution. 

Supervisors, laborers and others who routinely entered the processing area but did not 

routinely handle thorium will be assigned the full distribution of intakes. Others who did 

not routinely enter the area, will be assigned the geometric mean of the distribution. 

The Technical Information Bulletin, “Estimation of Ingestion Intakes”11 was also utilized 

to determine additional internal intake due to ingestion with the f1 value the same as that 

used for inhalation. The intake rates along with the geometric standard deviations (GSD) 

are listed in Table D.1. For internal dose estimates, the thorium will be considered to be 

type M or S solubility, whichever produces the highest dose. No thorium intakes will be 

assigned after 1964 since indications are that a cleanup of residual thorium dioxide was 

conducted following the project. 

D.5.3 In vivo

Direct (in vivo) whole-body counting (WBC) was conducted during the years 1963, 1964, 

and 1965 for several workers, using total body counters available at the Y-12 plant in 

Tennessee.7 Emphasis was placed primarily on several workers who had received 

elevated internal exposures in the Red Room (exceeding 10 C.F.R. pt. 20 limits). The 

analysis was performed to assay the amount (micrograms) of U-235 in the workers’ 

lungs. Other employees were also assayed for uranium at Y-12 during these years but not 

on a consistent basis. In vivo WBC was also conducted during the years 1968-1973. 

UNC contracted Helgeson Nuclear Services, Inc., to conduct WBC for evidence of U-235 

intakes.7

 

D.6 Residual Contamination

Residual contamination from weapons related work may have been present at United 

Nuclear after weapons work ended in 1973. Only exposure from weapons related work is 

addressed after 1973 even though additional exposure from non-weapons related work 

may have been received.3
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Appendix D

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In order to estimate this residual contamination, the highest intake rate from table D.1 

was converted to an air concentration and assumed to settle to the floor and accumulate 

for an entire year. The surface contamination resulting from this was then assumed to 

expose an individual for 2000 hours per year. Both external exposure and intakes from 

resuspension were calculated based on this value. 

A letter to the AEC4 describes some of the measures taken in 1960 to reduce airborne 

concentrations. These measures included wet mopping at frequent intervals not only the 

floors but overhead, piping and duct work. It also describes various type of ventilation 

used to remove airborne contaminants. Therefore, the assumption that all the airborne 

contamination settled out and was never removed is considered a bounding estimate and 

will be used as a constant in dose estimates.

D.7 References

1. RefID 30008 (pp. 16-17) Westinghouse, 2003, Hematite Former Fuel Cycle Facility 

Decommissioning, Historical Site Assessment, May 20, 2003 

2. DOE Office of Health, Safety and Security, EEOICPA web site. 

http://www.hss.energy.gov/healthsafety/fwsp/advocacy/faclist/findfacility.cfm

3. NIOSH, 2007, Radiation exposures covered for Dose Reconstructions under Part B 

of the Energy Employees Occupational Illness Compensation Program Act, Rev. 0, 

OCAS-IG-003, National Institute for Occupational Safety and Health, Office of 

Compensation Analysis and Support, Cincinnati, Ohio. 

4. ORAUT (Oak Ridge Associated Universities Team), ORAUT- OTIB-0010, 

Technical Information Bulletin: A Standard Complex-Wide Methodology for 

Overestimating External Doses Measured with Film Badge Dosimeters, Rev 01, June 

5, 2006. 

5. ORAUT (Oak Ridge Associated Universities Team), ORAUT- OTIB-0017, Technical 

Information Bulletin: Interpretation of Dosimetry Data for Assignment of Shallow 

Dose, Rev 01, October 11, 2005. 

6. ORAUT (Oak Ridge Associated Universities Team), ORAUT-PROC-0006, External 

Dose Reconstruction, Rev 01, June 5, 2006. 

7. NIOSH, SEC Petition Evaluation Report for the United Nuclear Corporation, 

Petition SEC-00116, Rev 1, January 28, 2010. 

8. NIOSH (2007) External Dose Reconstruction Implementation Guideline, Rev 3, 

OCAS-IG-001, National Institute for Occupational Safety and Health, Office of 

Compensation Analysis and Support, Cincinnati, Ohio. 

9. RefID 11724 (pp. 22-25) J. A. Lindberg, United Nuclear Corporation, 1963, Letter to 

Charles Keller, U. S. Atomic Energy Commission, October 28, 1963. 

10. RefID 17606 Frank Zeitlin, United Nuclear Corporation, 1960, Letter to H. L. Price, 

U. S. Atomic Energy Commission, December 16, 1960 
Document No. Battelle-TBD-6001; 

Appendix D

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11. OCAS (Office of Compensation Analysis and Support), OCAS-TIB-009, Estimation 

of Ingestion Intakes, Rev 0, April 13, 2004. 
Document No. Battelle-TBD-6001; Appendix D Revision No. 1 Effective Date: 4/30/2010 Page 9 

Table D.1 INTERNAL DOSE PATHWAYS - Inhalation of Airborne Radionuclides 

Assumptions:

Intakes are in dpm per calendar day. 

Dose entered into IREP as alpha radiation 

Intake value is the geometric mean of a lognormal distribution except where GSD is listed as constant. Those intake values are constants (no 

distribution) 

 

Job Category

 

Year

Operation

Phase

Relevant

Nuclide

Solubility 

Type

Intake

(dpm/d) GSD

 

TBD Reference or 

Research Justification

Operators 1/1/1958 - 6/13/1963 Operations U234 S 12590 3.29 Bioassay results

Supervisors 1/1/1958 - 6/13/1963 Operations U234 S 4784 3.0 Bioassay results

Other 1/1/1958 - 6/13/1963 Operations U234 S 4784 Constant Bioassay results

Operators 6/14/1963 - 12/31/1973 Operations U234 S 7662 3.0 Bioassay results

Supervisors 6/14/1963 - 12/31/1973 Operations U234 S 2311 3.0 Bioassay results

Other 6/14/1963 - 12/31/1973 Operations U234 S 2311 Constant Bioassay results

Operators 1/1/1958 - 6/13/1963 Operations U234 M 871.9 3.29 Bioassay results

Supervisors 1/1/1958 - 6/13/1963 Operations U234 M 331.2 3.0 Bioassay results

Other 1/1/1958 - 6/13/1963 Operations U234 M 331.2 Constant Bioassay results

Operators 6/14/1963 - 12/31/1973 Operations U234 M 560.2 3.0 Bioassay results

Supervisors 6/14/1963 - 12/31/1973 Operations U234 M 169 3.0 Bioassay results

Other 6/14/1963 - 12/31/1973 Operations U234 M 169 Constant Bioassay results

All 1/1/1974 – 07/31/2006 Residual U234 S 10.34 Constant Bioassay results

All 1/1/1974 – 07/31/2006 Residual U234 M 0.72 Constant Bioassay results

 

Job 

Category

 

Year

Operation

Phase

Relevant

Nuclide

Solubility 

Type

Intake

(dpm/d)

Ingestion

(dpm/d) GSD

 

TBD Reference or 

Research Justification

Operators 1964 Operations Th-232 M or S 4130 126 Constant Air sample data

Supervisors 1964 Operations Th-232 M or S 165 5.02 7.09 Air sample data

Other 1964 Operations Th-232 M or S 165 5.02 Constant Air sample data
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Table D.2 EXTERNAL DOSE PATHWAYS - Whole Body 

Based on monthly film badge results and 12 months per year of work. 

Values are the geometric mean of a lognormal distribution except where GSD is listed as constant. Those intake values are constants (no distribution)

Operation 

Phase

Year Others

(rem/year)

Operators

(rem/year)

GSD Supervisors

(rem/year)

GSD

Operations 1958 1.016 1.74 Constant 1.016 1.390 

1959 0.680 1.16 0.680 1.384 

1960 0.620 1.240 0.620 1.524

1961 0.346 0.550 0.346 1.325 

1962 0.511 1.349 0.511 1.805 

1963 0.467 1.070 0.467 1.655 

1964 0.538 1.255 0.538 1.674 

1965 0.769 3.494 0.769 2.509 

1966 0.181 0.382 0.181 1.576 

1967 0.160 0.271 0.160 1.376 

1968 0.177 0.386 0.177 1.605 

1969 0.181 0.371 0.181 1.546 

1970 0.179 0.514 0.179 1.897 

1971 0.134 0.176 0.134 1.179 

1972 0.151 0.390 0.151 1.779 

1973 0.172 0.385 0.172 1.633 

Operation

Phase

Year Whole body

(mr/yr)

Job Category GSD

Residual 1/1/1974 – 07/31/2006 11.6 All Constant 
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Table D.3 EXTERNAL DOSE PATHWAYS - Skin

Assumptions:

Based on monthly film badge results and 12 months per year of work 

Doses are entered into IREP as Electron >15 keV

These doses are in addition to the Whole Body doses in Table D.2

Values are the geometric mean of a lognormal distribution except where GSD is listed as constant. Those intake values are constants (no 

distribution) 

Operation 

Phase

Year Others

(rem/year)

Operators

(rem/year)

GSD Supervisors

(rem/year)

GSD

Operations 1958 1.060 6.030 Constant 1.060 2.877

1959 2.060 3.860 2.060 1.465 

1960 1.880 3.760 1.880 1.524 

1961 0.840 3.169 0.840 2.242 

1962 2.166 10.902 2.166 2.671 

1963 3.276 9.670 3.276 1.931 

1964 2.566 7.413 2.566 1.906 

1965 0.904 3.102 0.904 2.115 

1966 0.548 1.158 0.548 1.576 

1967 0.485 0.820 0.485 1.376 

1968 0.537 1.170 0.537 1.605 

1969 0.549 1.123 0.549 1.546 

1970 0.543 1.556 0.543 1.897 

1971 0.407 0.534 0.407 1.179 

1972 0.458 1.181 0.458 1.779 

1973 0.520 1.166 0.520 1.633 

Operation

Phase

Year Skin

(mr/yr)

Job Category GSD

Residual 1/1/1974 – 07/31/2006 186 All Constant 
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Table D.4 EXTERNAL DOSE PATHWAYS - Neutron 

Assumptions:

Doses are entered into IREP as Neutrons 100 keV – 2 MeV

Values are constant (no distribution) 

Operation Phase Year Others

(rem/year)

Operators

(rem/year)

Supervisors

(rem/year)

GSD

Operations 1958 - 1973 0.114 2.06 0.228 Constant