Division of Compensation Analysis and Support

Technical Basis Document for the United Nuclear Corporation

Hematite, Missouri

Document Number:

DCAS-TKBS-0008

Effective Date: 03/21/2011

Revision No.: 0

Page 1 of 15

Subject Expert: David Allen

Approval: Signature on file Date: 03/21/2011

Supersedes:

Battelle-TBD-6001 

James W. Neton, Associate Director for Science Appendix D

TABLE OF CONTENTS 

Section Page

1.0 Introduction .................................................................................................................4

2.0 Site Description and Operational History ...................................................................5

3.0 Internal Dose ...............................................................................................................5

3.1 Uranium Intakes .................................................................................................6

3.2 Internal Dose from Uranium Using Bioassay Results .......................................8

3.3 Thorium Intakes .................................................................................................8

3.4 In Vivo ...............................................................................................................8

3.5 Internal Dose from Residual Contamination .....................................................9

4.0 External Dose ..............................................................................................................9

4.1 Co-worker Gamma and Beta Dose ..................................................................10

4.2 Co-worker Neutron Dose .................................................................................12

4.3 Measured Dose.................................................................................................13

4.3.1 Photons ..................................................................................................13

4.3.2 Electrons ................................................................................................13

4.4 Missed Dose .....................................................................................................13

4.5 External Dose from Residual Contamination ..................................................13

5.0 Occupational Medical Dose ......................................................................................14

6.0 References .................................................................................................................15
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LIST OF TABLES 

Table Page

Table 1 Daily Intakes of Uranium (Solubility Type S) ............................................................7

Table 2 Daily Intakes of Uranium (Solubility Type M) ..........................................................7

Table 3 Daily Intakes of Thorium (Solubility Type M or S) ...................................................8

Table 4 Daily Intakes from Residual Contamination (Solubility Type S) ...............................9

Table 5 Daily Intakes from Residual Contamination (Solubility Type M) .............................9

Table 6 Annual Whole Body External Dose ..........................................................................11

Table 7 Annual Shallow External Dose .................................................................................12

Table 8 Annual Neutron External Dose .................................................................................13

Table 9 Annual Whole Body External Dose from Residual Contamination .........................14

Table 10 Annual Shallow External Dose from Residual Contamination ................................14
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ISSUE

AUTHORIZATION

DATE

EFFECTIVE

DATE REV. NO. DESCRIPTION

Draft 00-A Changes Battelle-TBD-6001 Appendix to 

a standalone document. Change is 

primarily format only. Does not 

incorporate review comments.

Draft 03/17/2011 00-B Internal review comments incorporated.

03/21/2011 03/21/2011 0 Document approved.
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1.0 Introduction 

Technical basis documents and site profile documents are not official determinations made by 

the National Institute for Occupational Safety and Health (NIOSH) but are rather general 

working documents that provide historic background information and guidance to assist in the 

preparation of dose reconstructions at particular sites or categories of sites. They will be revised 

in the event additional relevant information is obtained about the affected site(s). These 

documents may be used to assist NIOSH staff in the completion of the individual work required 

for each dose reconstruction. 

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

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

weapons employer [AWE] facility” or a “Department of Energy [DOE] facility” as defined in the 

Energy Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 

U.S.C. § 7384I(5) and (12)]. EEOICPA, as amended, provides for employees who worked at an 

AWE facility during the contract period and/or during the residual period. 

Under EEOICPA, employment at an AWE facility is categorized as either (1) during the DOE 

contract period (i.e., when the AWE was processing or producing material that emitted radiation 

and was used in the production of an atomic weapon), or (2) during the residual contamination 

period (i.e., periods that NIOSH has determined there is the potential for significant residual 

contamination after the period in which weapons-related production occurred). For contract 

period employment, all occupationally derived radiation exposures received at covered facilities 

must be included in dose reconstructions. This includes radiation exposure related to the Naval 

Nuclear Propulsion Program and any radiation exposure received from the production of 

commercial radioactive products that were concurrently manufactured by the AWE facility 

during the covered period. NIOSH does not consider the following exposures to be 

occupationally derived (NIOSH 2010): 

Background radiation, including radiation from naturally occurring radon present in 

conventional structures 

Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons 

For employment during the residual contamination period, only the radiation exposures defined 

in 42 U.S.C. § 7384n(c)(4) [i.e., radiation doses received from DOE-related work] must be 

included in dose reconstructions. Doses from medical X-rays are not reconstructed during the 

residual contamination period (NIOSH 2007). It should be noted that under subparagraph A of 

42 U.S.C. § 7384n(c)(4), radiation associated with the Naval Nuclear Propulsion Program is 

specifically excluded from the employee’s radiation dose. This exclusion only applies to those 

AWE employees who worked during the residual contamination period. Also, under 

subparagraph B of 42 U.S.C. § 7384n(c)(4), radiation from a source not covered by subparagraph 

A that is not distinguishable through reliable documentation from radiation that is covered by 

subparagraph A is considered part of the employee’s radiation dose. This site profile covers only 
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exposures resulting from nuclear weapons-related work. Exposures resulting from nonweapons-related work, if applicable, will be covered elsewhere. 

The following information from the Department of Energy’s Office of Health, Safety and 

Security EEOICPA Find Facilities webpage defines the EEOICPA covered periods for the 

United Nuclear Corporation.

Site: United Nuclear Corporation

Location: Hematite, Missouri 

Covered Period: 1958-1973, Residual Radiation 1974-October 2009 

This document contains a summary of the description of the site as well as the Atomic Energy 

Commission activities performed there, and provides the technical basis to be used to evaluate 

the occupational radiation doses for EEOICPA claims. 

2.0 Site Description and Operational History 

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 recovery (Westinghouse 2003).

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 operating the facility as ABB Combustion 

Engineering. In April of 2000, Westinghouse purchased the nuclear operations of ABB which 

included the Hematite facility (Westinghouse 2003).

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 1973 (DOE 2009). 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) (NIOSH 2010a). 

3.0 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 alpha-emitting 

materials via inhalation and ingestion of airborne uranium and thorium (and progeny). 
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3.1 Uranium Intakes 

There is urinalysis data for workers at UNC available for most years from 1958 through 1973. If 

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

unmonitored workers, a co-worker intake estimate was developed. 

The urinalysis data, however, generally does not separate people into job categories. A subset of 

this data that does separate people into job categories was used to estimate intakes. For this 

exposure estimate, the workers were broken up into three categories; Operators, 

Supervisor/Laborers, and Others. The “Operator” category is for personnel that routinely 

handled uranium or operated uranium processing equipment. The “Supervisor/Laborer” 

category is for personnel that were routinely in the uranium handling areas and handled uranium 

or uranium processing equipment, but not on a routine basis. Examples of people fitting this 

category are foreman, guards, maintenance personnel, and janitors. Laboratory technicians and 

chemists also fit this category because they routinely handled uranium, but not in production 

quantities. The “Other” category is for personnel that did not routinely enter the uranium 

production areas. Examples of personnel fitting this category are office personnel and store 

room attendants. 

Coworker internal dose from uranium urinalysis samples, starting in late 1962, revealed some 

“Operators” had higher intakes than previously believed based on air sampling (NIOSH 2010b). 

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 “non-operators” 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 Tables 1 and 2 below. 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/Laborer” intake will be used for people categorized as “Others”. 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. 

This intake estimate was compared to the larger data set. The larger data set was 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 United Nuclear intake 

calculations. 
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The predicted urine concentrations for “Operators” overestimated most annual results for both 

type M and type S intakes. The predicted concentrations for “Supervisors/Laborers” 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/Laborers” and “Others” 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. 

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

sample data and urinalysis data exist (ORAUT 2006a). The geometric mean of the assumed 

intakes under-predicted 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 Tables 1 and 2 below, for type S material were considerably 

higher than the air sample results for these personnel. 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, 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. 

Table 1: Daily Intakes of Uranium (Solubility Type S) 

Job Category Years Operation 

Phase

Radionuclide Inhalation 

(dpm/day)

GSD

Operators 1/01/58 - 6/13/63 Operations U-234 12590 3.29

Supervisors/Laborers 4784 3.0

Others 4784 Constant

Operators 6/14/63-12/31/73 7662 3.0

Supervisors/Laborers 2311 3.0

Others 2311 Constant

Table 2: Daily Intakes of Uranium (Solubility Type M) 

Job Category Years Operation 

Phase

Radionuclide Inhalation 

(dpm/day)

GSD

Operators 1/01/58 - 6/13/63 Operations U-234 871.9 3.29

Supervisors/Laborers 331.2 3.0

Others 331.2 Constant

Operators 6/14/63-12/31/73 560.2 3.0

Supervisors/Laborers 169 3.0

Others 169 Constant
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3.2 Internal Dose from Uranium Using Bioassay Results 

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

3.3 Thorium Intakes 

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 (NIOSH 2010b). Thorium dioxide powder 

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

assemblies for breeder reactors (NIOSH 2010b). 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 other personnel 

who routinely entered the processing area but did not routinely handle thorium will be assigned 

the full distribution of intakes. “Others”, personnel who did not routinely enter the area, will be 

assigned the geometric mean of the distribution. 

The Technical Information Bulletin, “Estimation of Ingestion Intakes” (NIOSH 2004) 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 3 below. 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. 

Table 3: Daily Intakes of Thorium (Solubility Type M or S) 

Job Category Years Operation 

Phase

Radionuclide Inhalation 

(dpm/day)

Ingestion

(dpm/day)

GSD

Operators 1964 Operations Th-232 4130 126 Constant

Supervisors/Laborers 1964 165 5.02 7.09

Others 1964 165 5.02 Constant

3.4 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 

(NOSH 2010b). Emphasis was placed primarily on several workers who had received elevated 
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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 (NIOSH 2010b). 

3.5 Internal Dose from 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 

(NIOSH 2010a).

In order to estimate this residual contamination, the highest intake rate from Table 3 above 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. The internal intake rates along with the geometric standard deviations 

(GSD) are listed in Tables 4 and 5 below.

A letter to the AEC (UNC 1960) 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.

Table 4: Daily Intakes from Residual Contamination (Solubility Type S) 

Job Category Years Operation 

Phase

Radionuclide Inhalation 

(dpm/day)

GSD

All 1/01/74-7/31/06 Residual U-234 10.34 Constant

Table 5: Daily Intakes from Residual Contamination (Solubility Type M) 

Job Category Years Operation 

Phase

Radionuclide Inhalation 

(dpm/day)

GSD

All 1/01/74-7/31/06 Residual U-234 0.72 Constant

4.0 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 
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during routine scrap recovery operations and other production activities at the site. Potential 

exposure to neutrons would have resulted from operations with enriched uranium. 

4.1 Co-worker Gamma and Beta Dose 

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 (ORAUT 2005b, 2006a, 2006b). 

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

exchanges (NIOSH 2010b). 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) (NIOSH 2010b). A yearly missed dose based on 20 mr (LOD/2) per 

week 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 exchanges. 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 (NIOSH 2010b). 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 week (NIOSH 2010b) 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 
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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. 

 

For this exposure estimate, the workers were broken up into three categories; Operators, 

Supervisor/Laborers, and Others. The “Operator” category is for personnel that routinely 

handled uranium or operated uranium processing equipment. The “Supervisor/Laborer” 

category is for personnel that were routinely in the uranium handling areas and handled uranium 

or uranium processing equipment, but not on a routine basis. Examples of people fitting this 

category are foreman, guards, maintenance personnel, and janitors. Laboratory technicians and 

chemists also fit this category because they routinely handled uranium, but not in production 

quantities. The “Other” category is for personnel that did not routinely enter the uranium 

production areas. Examples of personnel fitting this category are office personnel and store 

room attendants. 

“Operators” are assigned the 95th percentile of the annual dose. “Supervisors/Laborers” are 

assigned the full distribution of doses. The “Others” category is assigned the 50

th percentile of 

the distribution. The annual external dose values to be applied each year for unmonitored 

workers are listed in Tables 6 and 7 below. 

Table 6: Annual Whole Body External Dose 

Operation 

Phase

Year Others

(rem/year)(a)

Operators

(rem/year)(a)

GSD Supervisors

(rem/year)(a)

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

(a) Applied as Photons 30-250 keV. Whole body photon doses are to be converted to organ doses using the Exposure to Organ 

Dose Conversion Factors (US DHHS 2007). 
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Table 7: Annual Shallow External Dose

Operation 

Phase

Year Others

(rem/year)(a)

Operators

(rem/year)(a)

GSD Supervisors

(rem/year)(a)

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

(a) Applied as Electrons > 15 keV. 

4.2 Co-worker Neutron Dose 

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 (NIOSH 

2010b). 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 (2010b). 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 

(ORAUT 2005a) for naturally enriched uranium without progeny were used to estimate the 

neutron dose rate. The conversion factors were adjusted based on the specific activity of 

naturally enriched uranium compared to the specific activity of 20% and 93% enriched uranium.

 

“Operators” are assumed to have worked within 1 foot of the enriched uranium fluoride 1⁄2 the 

time (1000 hours per year). “Supervisors/Laborers” are assumed to have worked within 3 feet of 

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

worked with 20% enriched uranium and 93% enriched uranium for equal amounts of time. 

“Others” were assumed to have received an annual dose equal to half the “Supervisors/Laborers” 

dose. Annual dose values are listed in Table 8. These values represent a bounding estimate and 

should be considered a constant. 
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Table 8: Annual Neutron External Dose 

Operation 

Phase

Years Others

(rem/year)(a)

Operators

(rem/year)(a)

Supervisors

(rem/year)(a)

GSD

Operations 1958-1973 0.114 2.06 0.228 Constant

(a) Applied as Neutrons 100 keV - 2 MeV. 

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. 

4.3.1 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 (NIOSH 2007). The photon radiation should be considered 

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

4.3.2 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-0017 (ORAUT 2005b) for assignment of skin dose from penetrating and nonpenetrating radiation should be followed. The non-penetrating radiation is predominantly in the 

E >15 keV energy range. 

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.

4.5 External Dose from 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 

(NIOSH 2010a). 
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In order to estimate this residual contamination the highest intake rate from Table 1 above 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. Annual dose values are listed in Tables 9 and 10 below. These values 

represent a bounding estimate and should be considered a constant. 

Table 9: Annual Whole Body External Dose from Residual Contamination 

Operation Phase Years Whole Body

(mrem/year)(a)

Job Category GSD

Residual 1/01/74 - 7/31/06 11.6 All Constant

(a) Applied as Photons 30-250 keV. Whole body photon doses are to be converted to organ doses using the Exposure to Organ 

Dose Conversion Factors (US DHHS 2007). 

Table 10: Annual Shallow External Dose from Residual Contamination 

Operation Phase Years Shallow Dose

(mrem/year)(a)

Job Category GSD

Residual 1/01/74-7/31/06 186 All Constant

(a) Applied as Electrons > 15 keV. 

5.0 Occupational Medical Dose

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

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

provided in ORAUT-OTIB-0006, Technical Information Bulletin: Dose Reconstruction from 

Occupationally Related Diagnostic X-Ray Procedures (ORAUT 2005c). The assumed frequency 

in this document is pre-employment, annual, and termination PA chest X-rays which will be 

assigned between the years 1958 and 1973. Annual organ doses are entered into the NIOSHIREP program as the annual dose due to an acute exposure to photons (E=30-250 keV). The 

distribution is assumed to be normal with a standard deviation of 30%. 
Effective Date: 

03/21/2011

Revision No.

0

Document No.

DCAS-TKBS-0008

Page 15 of 15

6.0 References 

DOE 2009, Office of Health, Safety and Security, EEOICPA web site. 

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

NIOSH (National Institute for Occupational Safety and Health) 2004, Estimation of Ingestion 

Intakes, Rev 0, OCAS-TIB-0009, National Institute for Occupational Safety and Health, Office 

of Compensation Analysis and Support, Cincinnati, Ohio. 

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. 

NIOSH 2010a, Radiation Exposures Covered for Dose Reconstructions Under Part B of the 

Energy Employees Occupational Illness Compensation Program Act, Rev 1, OCAS-IG-003, 

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

Support, Cincinnati, Ohio. 

NIOSH 2010b, SEC Petition Evaluation Report for the United Nuclear Corporation, Petition 

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

ORAUT (Oak Ridge Associated Universities Team) 2005a, ORAUT-OTIB-0024, Estimation of 

Neutron Dose Rates from Alpha-Neutron Reactions in Uranium and Thorium Compounds, Rev 

00, April 7, 2005. 

ORAUT 2005b, ORAUT- OTIB-0017, Technical Information Bulletin: Interpretation of 

Dosimetry Data for Assignment of Shallow Dose, Rev 01, October 11, 2005. 

ORAUT 2005c, ORAUT- OTIB-0006, Technical Information Bulletin: Dose Reconstruction 

from Occupationally Related Diagnostic X-Ray Procedures, Rev 03 PC-1, December 21 2005 

ORAUT 2006a, ORAUT- OTIB-0010, Technical Information Bulletin: A Standard ComplexWide Methodology for Overestimating External Doses Measured with Film Badge Dosimeters, 

Rev 01, June 5, 2006. 

ORAUT 2006b, ORAUT-PROC-0006, External Dose Reconstruction, Rev 01, June 5, 2006.

UNC 1960, Letter from Frank Zeitlin to H. L. Price, U. S. Atomic Energy Commission, 

December 16, 1960 [Ref ID 17606]. 

UNC 1963, Letter from J. A. Lindberg, United Nuclear Corporation, to Charles Keller, U. S. 

Atomic Energy Commission, October 28, 1963 [Ref ID 11724 (pp. 22-25)]. 

Westinghouse 2003, Hematite Former Fuel Cycle Facility Decommissioning, Historical Site 

Assessment, May 20, 2003 [Ref ID 30008 (pp. 16-17)].