DOE Review 12/11/2013

Division of Compensation Analysis and Support

Technical Basis Document for the Huntington Pilot Plant, 

Huntington, West Virginia

Document Number:

DCAS-TKBS-0004

Effective Date: 12/12/2013

Revision No.: 01

Page 1 of 32

Subject Experts: Tom Tomes, Sam Glover, Christine Corwin

Approval: Signature on file Date: 12/12/2013

Supersedes:

OCAS-TKBS-004

Revision 00 James W. Neton, Associate Director for Science

TABLE OF CONTENTS

Section Page

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

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

3.0 Process Description ................................................................................................................8

4.0 Uranium Concentration and Enrichment ................................................................................9

5.0 Internal Dose during Operations...........................................................................................10

5.1 Uranium Intakes...............................................................................................10

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

6.0 External Dose during Operations..........................................................................................14

6.1 Deep Dose...................................................................................................................14

6.2 Shallow Dose ..............................................................................................................15

7.0 Dose during Remediation .....................................................................................................15

8.0 Internal and External Dose Summary...................................................................................16

9.0 Occupational Medical Exposures.........................................................................................18

10.0 References ............................................................................................................................20

Attachment A.....................................................................................................................................24
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LIST OF TABLES

Table Page

1 Specifications for Starting Material.................................................................................. 9

2 Estimates of Historical Nickel Exposures....................................................................... 10

3 Annual Intakes of Uranium............................................................................................. 13

4 Annual Intakes of Recycled Uranium Contaminants...................................................... 14

5 Summary of Inhalation and Ingestion Intakes................................................................ 17

6 Summary of External Doses ........................................................................................... 18

A1 Radionuclides in Drums of Residue ............................................................................... 26

A2 Gamma Ray Spectra Used for Analysis of Dose Rate.................................................... 27

A3 Beta Particle Emission Rates.......................................................................................... 28

A4 Dose Rate from Gamma Sources.................................................................................... 29

LIST OF FIGURES

Figure Page

A1 Energy Distribution for Photon Emissions..................................................................... 29

A2 Distribution of Dose from Bremsstrahlung..................................................................... 30

A3 Photon Dose at 77.9 cm above Array of Drums............................................................. 31

A4 Bremsstrahlung Dose at 77.9 cm above Array of Drums............................................... 32
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RECORD OF ISSUE/REVISIONS

ISSUE

AUTHORIZATION

DATE

EFFECTIVE

DATE REV. NO. DESCRIPTION

Draft 7/18/2008 00-A Draft document for review to replace 

ORAUT-TKBS-004 as the technical 

basis for reconstructing radiation doses

for former workers from the Huntington 

Reduction Pilot Plant. Newly available 

references have been incorporated with 

additional internal and external dose 

evaluations.

Draft 7/30/2008 00-B Incorporated formal review comments.

8/13/2008 00 Approved for use.

Draft 11/21/2013 01-A Added values of Am-241, Th-230, and 

Tc-99 to Tables 4 and 5. Changed

(corrected) the header of the fourth 

column of Table A-2 to “Photons per 

hour per Ci 238U.” Corrected three errors 

in the text of Section 6.2 describing dose 

and dose rates (actual doses reported for 

dose reconstructions in Table 6 were 

correct).

12/02/2013 12/12/2013 01-B Incorporated formal review comments.
Effective Date:

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

Employees Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) 

and (12)]. EEOICPA defines a DOE facility as “any building, structure, or premise, including 

the grounds upon which such building, structure, or premise is located ... in which operations 

are, or have been, conducted by, or on behalf of, the Department of Energy (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion 

Program)” [42 U.S.C. § 7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear 

Propulsion Program noted above, any facility that performs or performed DOE operations of any 

nature whatsoever is a DOE facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only 

determines eligibility for a dose reconstruction, which is a prerequisite to a compensation 

decision (except for members of the Special Exposure Cohort). The compensation decision for 

cancer claimants is based on a section of the statute entitled “Exposure in the Performance of 

Duty.” That provision [42 U.S.C. § 7384n(b)] says that an individual with cancer “shall be 

determined to have sustained that cancer in the performance of duty for purposes of the 

compensation program if, and only if, the cancer ... was at least as likely as not related to 

employment at the facility [where the employee worked], as determined in accordance with the 

POC [probability of causation1] guidelines established under subsection (c) ...” [42 U.S.C. § 

7384n(b)]. Neither the statute nor the probability of causation guidelines (nor the dose 

reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 

7384l(12)]. While this definition excludes Naval Nuclear Propulsion Facilities from being 

covered under the Act, the section of EEOICPA that deals with the compensation decision for 

covered employees with cancer [i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the 

Performance of Duty”] does not contain such an exclusion. Therefore, the statute requires 

NIOSH to include all occupationally-derived radiation exposures at covered facilities in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the 

Naval Nuclear Propulsion Program. As a result, all internal and external occupational radiation 

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
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12/12/2013 Revision No. 01 Document No.DCAS-TKBS-0004 Page 5 of 32

exposures are considered valid for inclusion in a dose reconstruction. No efforts are made to 

determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2007):

 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

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 

Huntington Pilot Plant. 

Site: Huntington Pilot Plant

Alternate Names: Reduction Pilot Plant

Location: Huntington, West Virginia

Covered Period: 1951 – 1963; 1978 – 1979

Facility Type: Department of Energy

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 bases to be used to evaluate 

the occupational radiation doses for EEOICPA claims.

2.0 Site Description and Operational History

The Huntington Pilot Plant (also known as the Reduction Pilot Plant) was built by the Atomic 

Energy Commission (AEC) in 1951 to supply nickel powder used to make gaseous diffusion 

barriers for the gaseous diffusion plants in Paducah, Kentucky, and Portsmouth, Ohio (Clark and 

Cottrell 1980, Berger 1981, US DOE 2007). A 1957 letter supplement also indicates that the 

AEC’s operating contractor receiving the nickel powder was Union Carbide Nuclear Company 

(AEC 1957a). A facility data report (AEC 1955) and an International Nickel Company (INCO) 

memorandum both suggest that the powder was shipped to K-25 for barrier production (AEC

1957d). INCO was the operating contractor of the plant.

A memorandum survey report (AEC 1950) indicates that in January 1950 the AEC was 

considering awarding a contract to the International Nickel Company to produce nickel powder 

by melting K-25 scrap at their plant in Huntington, West Virginia. The proposal was to utilize 

Huntington Works Furnace No. 5 in the Refinery Building. According to the survey report, a 

railroad siding adjoined the Refinery Building, which would make it possible to unload and 

weigh the scrap at the siding, place it in buckets while it was still in the sealed cartons, and 

charge the furnace with the sealed cartons. The memorandum also suggested that the intended 

schedule was to use furnace No. 5 for melting K-25 scrap for two weeks each month and that the 

melt period took eight hours. At the time of the memorandum survey report, it was estimated 

that about one million pounds of nickel scrap was waiting to be melted (White 1950). There are 
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no records indicating the proposed plan to process scrap nickel in the existing Huntington Works 

Refinery Building was pursued. 

A January 1951 AEC memo detailed a meeting between the AEC and INCO regarding the 

construction of a new facility to produce nickel powder for the AEC from nickel oxide. INCO 

was to purchase property adjoining their existing Huntington Works site. The AEC would then 

purchase the property from INCO and construct the new building. INCO would then lease the 

building from the AEC and perform the AEC nickel operations (AEC 1951).

The AEC Site consisted of 3.2 acres located east of International Nickel Company’s “Huntington 

Works” plant. The property was bounded on the north by the Chesapeake and Ohio Railroad, on 

the east by Cole Street, on the south by Altizer Avenue, and on the west by the “Huntington 

Works” site (Smith 1957). “Huntington Works” referred to the INCO site that was owned by 

INCO and not involved with AEC work. The AEC site, which was totally enclosed by a chain 

link fence, was referred to as the Reduction Pilot Plant. The facility is also referred to as the 

Huntington Pilot Plant.

The AEC contracted with Rust Engineering to construct the new AEC building, which was 

called the Reduction Pilot Plant (RPP). Construction included necessary support buildings, the 

installation of the equipment and utilities, the construction of a railroad spur, access road and 

security fence. The scheduled completion date for the construction of the Reduction Pilot Plant 

was December 31, 1951. Due to the inability of a vendor to furnish a satisfactory gas separation 

plant, the completion date was pushed back to January 31, 1952 (AEC 1952). 

As discussed above, large quantities of nickel scrap were owned by the AEC. Records indicate 

that the scrap nickel was contaminated with uranium, and trace amounts of transuranic elements 

that are associated with recycled uranium from having been used as barrier in the gaseous 

diffusion cascades (AEC 1958c, INCO 1958). There are several documents that indicate INCO 

received relatively small quantities of nickel scrap from the AEC starting in February 1956. The 

scrap was to be processed in the RPP with the AEC receiving a credit against current invoices for 

nickel powder purchases. The purpose of this work was to determine the feasibility of 

converting larger quantities of such scrap material to reduced metal. In February 1956 the AEC 

approved INCO processing “10,000 pounds of government owned nickel scrap material for the 

purpose of determining the feasibility of converting larger quantities of such scrap material to 

reduced metal” (AEC 1956a). Another letter on April 6, 1956, indicates that the AEC authorized 

processing an additional 20,000 pounds of nickel scrap (AEC 1956b). Then on April 27, 1956, 

the AEC authorized processing of 22,000 pounds of scrap and approximately 3,000 pounds per 

day for 60 days (AEC 1956c). In November 1956 the AEC authorized an additional 12,000 

pounds of scrap to be processed (AEC 1956c). All of the authorizations for scrap processing 

required INCO to keep material segregated to the extent practical and required them to deliver 

both the reduced metal and scrap residues to the AEC in labeled drums. 

In June 1956 the AEC authorized expenditures for installation of some new equipment to process 

scrap, but did not fully fund modifications of the RPP for processing scrap until fiscal year 1957 

(AEC 1956d). The AEC issued a Directive on February 14, 1957, for INCO to modify the 

Reduction Pilot Plant to permit the processing of scrap barrier into nickel powder. A progress 
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report dated December 9, 1957, indicated the plant should be ready to use the new material by

January 31, 1958 (AEC 1957b, AEC 1957e).

In a January 1957 letter from the AEC to INCO, the AEC discussed the maximum uranium 

contamination that was expected for the scrap material. The letter states that INCO had visited 

Oak Ridge in September of 1956 to observe and learn of the health and safety problems that may 

result from the processing of the contaminated scrap material. The AEC expected levels of 300 

ppm at an average assay of 1-2% U-235. The AEC suggested monitoring the process through 

the various stages with a suitable radiation detection instrument. Union Carbide (the AEC’s 

operating contractor for the K-25 Gaseous Diffusion Plant) had provided INCO the radiation 

detection instrumentation (AEC 1957c). 

Two December 1957 documents discussed the “new starting material” and the associated 

radiation hazards of the material (INCO 1957a) (INCO 1957b). The documents also state that 

operations with the new starting material will begin shortly.

In February of 1958 the AEC and Union Carbide began regular shipments of scrap to INCO for 

processing. An AEC letter documented that “clean” scrap barrier was to be shipped to INCO the 

weeks of February 2 and 9 for processing (150,000 lbs. total), and that “contaminated” scrap was 

to be shipped the weeks of February 23, March 2 and March 9 for processing (240,000 lbs. total). 

Details of shipping arrangements were made for the material moving between Huntington and 

Oak Ridge (AEC 1958b). A 1960 AEC document stated that nine million pounds of scrap 

barrier material had been processed by INCO (AEC 1960a).

An additional document (AEC 1961) gave details of starting material transactions of the Oak 

Ridge Gaseous Diffusion Plant (ORGDP) for fiscal years 1959 through 1961. According to this 

document, during this time period ORGDP shipped between 1,980,000 and 2,587,000 pounds of 

starting material to INCO per year. A narrative attached to the document indicates that these 

quantities were the required amounts that satisfied ORGDP’s portion of the total quantity of 

starting material that was to be shipped to INCO by the three gaseous diffusion plants. The 

amounts required by INCO were based on the assumption of receipts of powder from INCO at 

the rate of 390,000 pounds per month. 

A memorandum route slip indicates that the Reduction Pilot Plant was shut down on Tuesday, 

December 11, 1962, (AEC 1962a) and maintained in stand-by condition by Huntington Alloys, 

Inc., a subsidiary of INCO. This was accomplished by discontinuing shipments of starting 

material to the plant and processing the inventory reserves. All classified material was removed 

from the facility. Production systems were then purged, drained and lubricated. Equipment 

instrumentation was removed and put in storage. Routine facility and equipment maintenance as 

well as security inspections would be performed while the facility was in stand-by condition 

(INCO 1962a). 

The Department of Energy (DOE) performed a preliminary radiological survey of the site and 

recommended that the site be decontaminated. The decontamination and demolition of classified 

and contaminated equipment at the facility took place between November 27, 1978, and May 18, 

1979 (Clark and Cottrell 1980). According to reports, the classified and contaminated scrap was 
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transported to the Portsmouth Plant in Ohio in 59 truckloads and 4 railcar loads. There were 138 

truckloads of clean scrap that were removed by the Cleveland Wrecking Company (Clark and 

Cottrell 1980). The only original building remaining on the AEC site was the Compressor 

Building.

3.0 Process Description 

There were two sources of feed material used for the nickel powder production for the AEC. 

One source was nickel oxide, and the other source was barrier scrap supplied by the AEC.

According to process descriptions documented by the AEC, nickel oxide was processed using 

carbonyl to produce high-purity nickel powder. This process was conducted in the Reduction 

Pilot Plant. The process involved reducing the NiO to nickel with hydrogen in a reduction 

vessel. The following process description was provided in an AEC memorandum: “The nickel 

is then volatized by combining it with carbon monoxide in a pressure vessel at approximately 

300 lbs. PSI, thus forming a gaseous nickel carbonyl. This gaseous carbonyl is passed through a 

condenser to change it to a liquid status. The liquid carbonyl is then purified by selective 

distillation, removing the iron oxide and other impurities. The purified carbonyl then passes

through the decomposer (a vertical vessel approximately 5’ in diameter x 13’ high) and, by 

controlled temperature, the pure carbonyl is broken down, dropping out of the bottom of the 

vessel as metal powder. The metal powder then passes over screens to remove lumps, etc. After 

screening operations are completed, the metal powder is packed and ready for shipment” (AEC 

1951). 

The contaminated scrap nickel barrier material that was processed in the Reduction Pilot Plant on 

a pilot scale starring in 1956 was expanded into a routine operation in 1958. A modification was 

made to the nickel carbonyl process in order to process the scrap barrier. The process

modification was described by McAlduff (AEC 1958a). The scrap nickel furnished by the AEC 

contained only small quantities of NiO; therefore, the reduction step previously utilized for the 

NiO starting material was eliminated. The scrap nickel was activated with hydrogen in 4000 lb. 

batches in two kilns. The material was then dumped through an air tight seal into two storage 

vessels, which were previously used as the reduction and activation reactors for NiO. The 

process from this step forward was the same as the process described above for NiO starting 

material. 

The solid contaminants that resulted from the purification and distillation of the nickel starting 

material were removed from the system as a residue. As part of a radiological safety evaluation 

in 1957, INCO described how the residues were handled and stored. The residues contained the 

majority of the radiological contaminants and were concentrated in a much smaller mass than 

were present in the received scrap. The residues were placed into 20 gallon drums and then 

stored on the ground floor of the plant spaced over a 40 ft. by 105 ft. area until shipped back to 

the AEC (INCO 1957b). All of the residues from a single 4,000 lb. batch of scrap were stored in 

a single 20 gal. drum (UCNC 1958c). 
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4.0 Uranium Concentration and Enrichment

The nickel scrap sent to the RPP from the AEC included scrap barrier material that was 

contaminated with uranium from having been used in the uranium enrichment process. Records 

also indicate that some of the scrap was not contaminated (AEC 1958b). The contaminated scrap 

was decontaminated at the gaseous diffusion plants prior to shipment to the RPP (AEC 1960a). 

However, the scrap still contained small quantities of uranium and associated radionuclides. 

The AEC provided specifications for the material shipped to the RPP (AEC 1958d). The 

specifications are shown in Table 1. The values were noted to be guaranteed minimum nickel 

and maximum contaminate levels. All other documentation and actual data indicates lower 

levels of uranium.

Table 1. Specifications for Starting Material

Constituent Maximum Minimum

Nickel 98.0%

Water 0.50% --

Iron 0.40% --

Uranium-235 0.0875 grams per pound of 

starting material average in any 

one drum

--

Total Uranium 500 ppm --

1. Reproduced from AEC 1958d.

Reports also stated that nickel scrap from the K-25 plant was contaminated with uranium with a 

maximum enrichment of 4% (by weight) (Clark and Cottrell 1980, Berger 1981). In late 1956 

the AEC and Union Carbide evaluated criticality concerns prior to the approval of modifications 

of the RPP to allow processing scrap barrier. The evaluation concluded that the AEC would be 

sending barrier scrap containing a maximum of 300 ppm with an average enrichment of 1-2% U235, noting that some degree of uncertainty can be expected (AEC 1957c). However, the AEC 

agreed to make reasonable efforts to deliver the starting material at the lowest practicable 

enrichment. 

In January of 1958 Union Carbide sent a letter to INCO stating that INCO will be receiving two 

samples of starting material that differ in uranium content and U-235 assay, indicating that the 

uranium content of these samples was higher than normal and it was agreed that special written 

notice would be given in such instances. The letter stated that the samples will contain 0.78% 

uranium and 3.5% U-235 assay. The scrap with the 3.5% enrichment uranium contained only 24 

ppm total uranium compared to the 165 ppm total uranium concentration of the 0.78% material

(AEC 1958c). 

Another reference provides average uranium concentration and enrichment. A 1960 AEC 

document stated that over 9,000,000 pounds of AEC scrap barrier had been processed at the 

RPP. It was reported that composite samples indicated the residues resulting from the nickel 

recovery at the RPP had “a total uranium content of less than one percent with a U-235 content 

of about 0.90 %.”
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5.0 Internal Dose during Operations

5.1 Uranium Intakes

There are no records available of airborne radioactivity monitoring during the period that 

contaminated scrap nickel was processed at the Reduction Pilot Plant, although there are several 

available documents of radiation safety evaluations before the routine scrap operations began in 

1958. Records indicate that the changes made to the RPP for scrap processing included 

engineered controls for transferring contaminated residues to drums and a vacuum system for 

handling contaminated residues (AEC 1957e). 

In May 1958 the AEC asked Union Carbide to perform uranium bioassay services for two rounds 

of uranium urinalyses. The request by the AEC to Union Carbide was prompted by an INCO 

request to the AEC. INCO did not expect an exposure problem but wanted the analyses to 

evaluate the effectiveness of their controls (AEC 1958e). However, no other documentation on 

the bioassay request or bioassay records has been located. 

Airborne uranium concentrations have been estimated from an evaluation of personnel exposure 

to airborne nickel at the Huntington Works. The results of that evaluation are provided in Table 

2 below.

Table 2. Estimates of Historical Nickel Exposures 

(reproduced from Enterline and Marsh 1982)

Department Exposure

(mg Ni/m3)

Acid reclaim 0.02

Blacksmith shop 0.02

Carpenter shopb0.02

Chipping and hammer 0.75c

Cold drawing (acid reclaim 1970 on) 0.10c

Combustionb0.05

Electricalb0.10

Extrusion 0.20c

General officesb0.01

Heat treatment 0.02

Machine shop 0.03c

Mechanicalb0.01

Melting and casting 0.25c

Merchant mill 0.30c

Night superintendentb0.01

Pipe shopb0.08

Polishing 0.10c

Primary mill 0.06

Reduction pilot plant 0.02

Refinery 5.00c
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Department Exposure

(mg Ni/m3)

Research and developmentb0.05

Roll grinding 0.03

Roll turning 0.03

Sheet mill (acid reclaim to 1970) 0.20c

Shipping 0.01

Standards or industrial engineeringb0.01

Steel shopb0.12

Stocks 0.01

Stores 0.01

Strip mill 0.20c

Transportationb0.01

Vacuum melting 0.15

Welding products 0.02

Yardb0.10

Warehouse 0.01

Watchmanb0.01

Inspection (unassigned) b0.05

aBased on current measurements except as noted.

b

Plant-wide work assignment.

CBased on historic midget impinger counts.

These exposure estimates, made by INCO for each of the various departments, were intended to 

represent average airborne concentrations of nickel in all forms over an 8-hour period. 

Generally, modern exposure data were used to estimate historic exposures and, whenever 

possible, were adjusted on the basis of process changes and environmental controls that were 

implemented over the years. For some departments, historical sample data were available that 

were obtained by the midget impinger particle counting technique. These data were then 

converted to the modern gravimetric expression (Enterline and Marsh 1982).

The data in Table 2 includes a result for average worker exposure in the Reduction Pilot Plant. 

However, the single value has no reported uncertainty. To determine a bounding airborne nickel 

exposure at the Reduction Pilot Plant, the Huntington Works department exposure values in 

Table 2 have been ranked and fit to a lognormal distribution, which results in a median value of 

0.046 mg Ni/m3 with a geometric standard deviation of 3.9. The 95th percentile value of the 

distribution is 0.44 mg Ni/m3. The 95th percentile value will be used as the upper bound 

concentration of airborne nickel in the RPP. Intakes of uranium and associated radionuclides are 

then derived from the nickel concentration.

The residues resulting from processing contaminated scrap nickel contained the most 

concentrated source of uranium in the RPP. As discussed above, the enrichment was expected to 

be between 1% and 2% U-235 with some variations expected (AEC 1957c). The AEC reported 

in 1960 that the uranium concentration in the residues were less than 1% total uranium from 

processing over 9,000,000 pounds of scrap with a U-235 enrichment of about 0.90% (AEC 
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1960a). Thus 2% U-235 enrichment is assumed for the residues to allow for variation and to 

bound uranium radioactivity based on mass to radioactivity conversion. 

Residues were not the only potential source of uranium exposure in the RPP, but are assumed to 

have the highest dose potential based on source term and process information. Residues had 

significantly higher uranium concentration than both the incoming contaminated nickel and the 

RPP nickel produced from contaminated scrap (AEC 1960a). A 1959 letter indicates that the 

AEC had requested INCO to consider blending residues with future scrap nickel. This was 

apparently an inquiry to determine if the unreacted nickel reporting to the residues from the 

nickel carbonyl process at the RPP could be recovered. INCO rejected the proposal because they 

feared that such a new operation could result in dust problems. INCO also performed laboratory 

scale tests of reprocessing residues, the results of which indicated that it was feasible. INCO

proposed a few trial runs to determine the feasibility of using the residues as starting material

(INCO 1959). The 1960 report (AEC 1960a) on processed material does not indicate residues as 

a source. Also, the AEC was considering shutting down the RPP in 1960 and performed cost 

studies on nickel processing. That December 1960 report indicated the AEC had only two 

agreements with INCO: lease of the facility, and the contract for nickel production from recovery 

of nickel from scrap. The cost of contract obligations was a consideration in keeping the plant 

open past 1960 (AEC 1960b). Also, there was no indication that the AEC had re-evaluated the 

basis for criticality that would likely be required to reprocess residues. There is no 

documentation indicating the residues were ever used as a source of starting material other than 

the small scale tests mentioned above. Worker exposures from the tests are considered 

insignificant in comparison to exposure from processing large quantities of contaminated scrap.

The 95th percentile airborne nickel concentration of 0.44 mg Ni/m3is considered to be a 

bounding estimate for concentration in the RPP. The AEC reported that the residues were less 

than 1% uranium. The residues were known to contain nickel as well as other contaminants that 

may have concentrated in the residues during nickel refining. The ratio of nickel to uranium in 

the residue is estimated by using the upper bound 1% uranium concentration (AEC 1960a) and 

assuming that 1% of the nickel in the starting material reported to the residues. Starting material 

was at least 98% nickel (AEC 1958d), thus each 4,000 pound batch of processed starting 

material would have resulted in at least 39.2 pounds of nickel in the drum of residue, assuming 

1% loss. The mass of residue in each batch was estimated to be 50 pounds (UCNC 1958c). For

a uranium concentration of less than 1%, the mass of uranium in each drum of residue would be 

less than 0.5 pounds. Thus, the uranium to nickel ratio in the residue is 0.013 lb U per pound Ni 

(0.5 lb/39.2 lb). This estimate accounts for 40 of the 50 pounds of residue per drum. Other 

contaminates, including iron, are assumed to account for the remainder.

Airborne uranium concentration based on the 95th percentile nickel airborne concentration is 

estimated to be:

3 3

0.013 0.00572 *0.44

m

mg U

mg Ni

mg U

m

mg Ni



Mass of 2% enriched uranium is estimated to be 1200 pCi/mg (DOE 2004). Uranium isotopic 

radioactivity concentrations are estimated to be 68% U-234, 28% U-238, and 4% U-235. 
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Maximally exposed workers are assumed to have been exposed to the bounding concentration 

for 2,000 hours per work year at the breathing rate of 1.2 m3/hr resulting in the following annual 

inhalation intake of total uranium:

yr

E pCi U

yr

m

mg U

E pCi U

m

mg U 2400 1.6 4

*

1.2 3

*

0.00572 3

3



Daily ingestion of uranium is estimated to be 0.2 times the airborne activity per cubic meter 

(OCAS 2004). Therefore, exposure to the bounding airborne concentration for 250 work days 

per year also results in an annual ingestion intake of 340 pCi total uranium per year. 

Production workers are assumed to be maximally exposed. Workers in the Production category 

include production operators, laborers, supervisors and maintenance personnel. Intakes for those 

workers are based on the 95th percentile values discussed above. Other workers in the RPP are 

assumed to have been exposed at the median value, which results in an annual intake equal to 

10.5% of the maximally exposed worker. Workers in the Administrative category include 

managers, secretaries and other administrative personnel.

Annual intakes are summarized in Table 3 below.

Table 3. Annual Intakes of Uranium

Worker 

Category Radionuclide

Airborne 

Concentration, 

pCi/m3

Inhalation,

pCi/yr

Ingestion,

pCi/yr

Production Total U 6.87 1.6E+4 340

Administrative Total U 0.72 1.7E+3 36

5.2 Recycled Uranium

Recycled uranium may have been present with the contaminated scrap material that was 

processed beginning in 1956 (AEC 1958c). Recycled uranium is uranium that has been 

irradiated in a reactor and from which the uranium has been extracted from the plutonium, 

fission products and other elements. However, small quantities of some radioactive elements are 

still present in the material and may have been present on the contaminated scrap barrier shipped 

to the RPP. Pu-239 and Np-237 are the contaminants that are likely to have contributed to 

significant dose at the RPP.

Plutonium and neptunium were present in the recycled uranium that was fed into the gaseous 

diffusion plants. The recycled uranium comprised a significant portion of the feed to the gaseous 

diffusion plants starting in 1952. The AEC had accumulated large quantities of scrap nickel 

prior to the introduction of recycled uranium into the AEC complex. Additionally, the scrap 

nickel sent to INCO consisted of both contaminated scrap (from used barrier material) and 

uncontaminated nickel (nickel not used in the enrichment cascades) (AEC 1961). Assuming all 

scrap was contaminated with both uranium and recycled uranium contaminants would result in 

an overestimate of worker exposure.
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The operational barrier material used in the gaseous diffusion plant equipment was located in the 

converters. Much of the plutonium and neptunium was removed by other processes at the 

gaseous diffusion plants prior to introduction of UF6 into the converters. However, some plating 

of the contaminants likely occurred due to the presence of trace quantities in the gas. The default 

isotopic ratios from the Technical Basis Document K-25 Gaseous Diffusion Plant - Occupational 

Internal Dose (ORAUT 2006) are applied to the uranium intakes estimated for the RPP. Table 4 

below contains the relative activities and annual intakes for inhalation and ingestion based on the 

values for uranium in Table 3 above. 

Table 4. Annual Intakes of Recycled Uranium Contaminants

Worker 

Category Contaminant Contaminant Activity1

Inhalation,

pCi/yr

Ingestion,

pCi/yr

Production

Pu-239 0.062 1.0E+3 2.1E+1

Np-237 0.0050 8.0E+1 1.7E+0

Am-241 0.062 1.0E+3 2.1E+1

Th-230 0.017 2.8E+2 5.9E+0

Tc-99 0.00011 1.8E+0 3.8E-2

Administrative

Pu-239 0.062 1.1E+2 2.2E+0

Np-237 0.0050 8.5E+0 1.8E-1

Am-241 0.062 1.1E+2 2.2E+0

Th-230 0.017 3.0E+1 6.3E-1

Tc-99 0.00011 1.9E-1 4.0E-3

1. Relative to total uranium activity.

6.0 External Dose during Operations

6.1 Deep Dose

Penetrating photon dose at the RPP has been modeled based on continuous exposure to the 

residues, which contained the highest concentration of radioactive constituents in the plant (AEC 

1960a). The process residue from each 4,000 lb. batch of scrap was discharged from the nickel 

carbonyl reaction chamber into a 20 gallon drum. The drums of residue were stored on the 

ground floor and spaced over a 40’ by 105’ area. Residue from as much as 150,000 lbs. of 

starting material (contaminated scrap) would be stored in this area at a given time (INCO 1957b, 

UCNC 1958c). 

External whole body dose rates at 30 cm and at 100 cm have been estimated from source term 

information using MCNP. A summary of the evaluation is in Attachment A. Doses were 

modeled assuming a worker was standing in front of a row of drums that were stored with no 

space between them. The total whole body deep dose rates are estimated to be 3.24E-5 rad/hr 

and 1.70E-5 rad/hr at distances of 30 cm and 100 cm, respectively.

The maximally exposed worker dose was estimated by assuming that the worker was located 30 

centimeters from the center of a row of drums for 2000 hours per year. This results in an annual 

exposure of 0.065 rad per year. This value applies to those Production personnel who may have 
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been continually working with the drums of material, such as operators, laborers, supervisors and 

clerks. Other workers at the RPP are assumed to have been exposed at one half the value of the 

maximally exposed worker, or 0.033 rad per year. This value applies to Administrative 

personnel such as managers, secretaries, and other office workers.

The annual whole body doses are to be converted to organ doses by multiplying the estimated air 

kerma doses discussed above by the “Kerma (Ka) to Organ Dose (HT)” photon dose conversion 

factors found in Appendix B of the NIOSH External Dose Reconstruction Implementation 

Guideline (US DHHS 2007). The exposure geometry was assumed to be anterior-posterior (AP) 

and the whole body dose rate is to be divided equally between photons with E=30-250 keV and 

photons with E>250 keV and applied as a constant distribution. 

6.2 Shallow Dose

Shallow dose from nonpenetrating radiation is assumed to have occurred from exposure to 

uncontained residues. Sealed drums of residue effectively shield emitted beta particles. For dose 

reconstruction purposes the shallow dose from uranium is considered to be from electrons 

greater than 15 keV. 

The AEC evaluated beta dose rates at the RPP. They estimated an upper bound dose rate from 

the residues at 20 mrep/hour on contact (UCNC 1958c). Workers would not normally handle 

uncontained residues. However, some direct contact would be possible due to maintenance 

activities or incidental contact by those who directly handled the material and equipment. It is 

assumed that some workers could have had direct contact with the residues for fifty hours per 

year resulting in a shallow dose to the hands and forearms of 1.0 rem per year. This dose would 

only be applied to the equipment operators and maintenance workers with potential for direct 

contact with the residues. 

Shallow dose to other tissues are determined by estimating the dose rate at 30 cm from the 

surface of the residues. The beta dose rate from aged yellowcake drops by a factor of 

approximately 75 from contact to 30 cm (NRC 2002). The bounding contact dose rate estimated 

for the residues is thus divided by 75 to determine beta dose rates at 30 cm from the surface of 

the residues, resulting in a beta dose rate of 0.27 mrem per hour. Production Operators are 

assumed to have been exposed at that rate for 2,000 hours per year resulting in a maximum 

shallow dose of 0.54 rem per year. Workers who only had intermittent exposure to uncontained 

material are assumed to have been exposed at the 30 cm rate for 1,000 hours resulting in a beta 

dose rate of 0.27 rem per year, which is applied to Production Workers other than Operators, 

which would include laborers and supervisors. Administrative personnel who did not work with 

equipment or open containers are assumed to have had no significant beta exposure. 

7.0 Dose during Remediation

The RPP ceased production in 1962 and was placed in a standby status in 1963. Procedures were 

developed in 1962 to remove material and chemicals, and to purge all systems and place the 

plant in an acceptable standby condition. The residue areas were also required to be completely 

cleaned. The plant was never restarted. It was demolished from 1978 through 1979. There are 
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no available records of radiation monitoring during the demolition period. Survey results are 

available for the area after the plant was demolished. The only remaining structure was a 

Compressor Building that was located adjacent to the plant (INCO 1962, Berger et al., AEC 

1962b).

Internal and external doses during the remediation period in 1978 through 1979 are to be 

assigned at the same level as estimated for the operational period in Sections 5 and 6. 

8.0 Internal and External Dose Summary

Table 5 summarizes the intake data contained in Section 5.0 for the purposes of reconstructing

internal doses at the Huntington Reduction Pilot Plant. The values have been normalized to 

intake per calendar day. These values should be applied as a continuous chronic intake. Intakes 

should be considered constants for dose reconstruction purposes because the methods used to 

derive intakes were based on bounding assumptions. 
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Table 5. Summary of Inhalation and Ingestion Intakes

Production Workersc

StartbEndb Radionuclide Intake Mode Absorption 

Type

Intake Rate 

pCi/day

01/01/1956

01/01/1978

12/31/1963

12/31/1979

Total U

Inhalation M or S 4.4E+1

Ingestion (a) 9.3E-1

Pu-239

Am-241

Inhalation M or S 2.7E+0

Ingestion (a) 5.8E-2

Np-237

Inhalation M 2.2E-1

Ingestion (a) 4.6E-3

Th-230

Inhalation M or S 7.6E-1

Ingestion (a) 1.6E-2

Tc-99

Inhalation F or M 4.8E-3

Ingestion (a) 1.0E-4

Administrative Workersc

Start End Radionuclide Intake Mode Absorption 

Type

Intake Rate 

pCi/day

01/01/1956

01/01/1978

12/31/1963

12/31/1979

Total U

Inhalation M or S 4.7

Ingestion (a) 0.099

Pu-239

Am-241

Inhalation M or S 0.30

Ingestion (a) 6.0E-3

Np-237

Inhalation M 2.3E-2

Ingestion (a) 4.9E-4

Th-230

Inhalation M or S 8.1E-2

Ingestion (a) 6.3E-1

Tc-99

Inhalation F or M 1.9E-1

Ingestion (a) 4.0E-3

(a) The f1 absorption value for ingestion should be chosen to be consistent with the inhalation material type.

(b) Intakes are normalized to calendar days and are to be assigned based on actual start and stop dates during the 

listed covered period. 

(c) See Section 5 for determination of Production or Administrative Workers.
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Table 6 summarizes the external doses contained in Section 6 for the purposes of reconstructing

external doses at the Huntington Reduction Pilot Plant. The dose values are normalized per 

calendar year. These values should be applied as a chronic exposure. The external doses should 

be considered constants for dose reconstruction purposes because the methods used to derive 

intakes were based on bounding assumptions. 

Table 6. Summary of External Doses

Production Workersc

StartaEnda Dose Type Annual dose, radb Radiation and energy

01/01/1956

01/01/1978

12/31/1963

12/31/1979 Deep Dose 0.065

photons

30 - 250 keV 50%

> 250 keV 50%

01/01/1956

01/01/1978

12/31/1963

12/31/1979

Shallow Dose

0.540 (Operators)

0.270 (other 

Production 

Workers)

electrons

> 15 keV

Shallow Dose 

(hands and 

forearms)

1.000

(Operators and 

Maintenance only)

electrons

> 15 keV

Administrative Workersc

Start End Dose Type Annual dose, radb Radiation and energy

01/01/1956

01/01/1978

12/31/1963

12/31/1979 Deep Dose 0.033

photons

30 - 250 keV 50%

> 250 keV 50%

01/01/1956

01/01/1978

12/31/1963

12/31/1979 Shallow Dose none

(a) Doses are normalized to calendar days and are to be assigned based on actual start and stop dates of covered

employment during the listed covered period. 

(b) Whole body photon doses are to be converted to organ doses using the Kerma to Organ Dose conversion 

factors (US DHHS 2007).

(c) See Section 6 for determination of Production or Administrative Workers.

9.0 Occupational Medical Exposures

No information regarding occupational medical dose specific to Huntington Reduction Pilot 

Plant was found. Information to be used in dose reconstructions for which no specific 

information is available is provided in ORAUT-OTIB-0006, the technical information bulletin 

covering diagnostic X-ray procedures. The assumption is made that workers received an annual 

occupationally related diagnostic X-ray beginning with the start of the AEC contract in 1951 and 

continuing through 1963 when AEC operations ended. X-rays are also assumed for 1978 and 

1979 during the remediation period. Annual PA chest X-rays should be applied in accordance 

with the current revision of ORAUT-OTIB-0006, Technical Information Bulletin: Dose 
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Reconstruction from Occupationally Related Diagnostic X-Ray Procedures (ORAUT 2003). 

Annual organ doses are entered into the NIOSH-IREP 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%. 
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10.0 References

AEC 1950, Memo Survey Report on International Nickel Company, Inc. from C.V. White, AEC, 

January 24, 1950 [SRDB Ref ID 22128].

AEC 1951, Memo from F. C. Hungerford, AEC, to N. Williams, Acting Director of Production 

and Engineering, AEC, January 31, 1951 [SRDB Ref ID 19713].

AEC 1952, Memo from F. R. Dowling, Acting Director of Production and Engineering, to S. R. 

Sapirie, Manager, January 4, 1952 [SRDB Ref ID 19717].

AEC 1955, Facility Data Report on International Nickel Company, Inc., February 17, 1955

[SRDB Ref ID 22126].

AEC 1956a, Letter from John R. Moore, Contracting Officer AEC, to the International Nickel 

Company, Inc., February 10, 1956 [SRDB Ref ID 21998].

AEC 1956b, Letter from R. G. Humphries, Contracting Officer AEC, to the International Nickel 

Company, Inc., April 6, 1956 [SRDB Ref ID 21999 pp. 2-3].

AEC 1956c, Letter from John R. Moore, Contracting Officer AEC, to the International Nickel 

Company, Inc., April 27, 1956 [SRDB Ref ID 21999, pp. 4-5].

AEC 1956d, Letter from S. R. Sapirie, Manager, Oak Ridge Operations, to the International 

Nickel Company, Inc., June 29, 1956 [SRDB Ref ID 21999, p. 6].

AEC 1956e, Letter from John R. Moore, Contracting Officer, to the International Nickel 

Company, Inc., November 23, 1956 [SRDB Ref ID 21999, pp. 7-8].

AEC 1957a, Letter Supplement from R.G. Humphries, Contracting Officer AEC, to the 

International Nickel Company, Inc., June 13, 1957 [SRDB Ref ID 22106].

AEC 1957b, Memo from W. J. Hayes, Director Construction Division AEC, to N.A. Shearon, 

Director, Budget and Reports Division AEC, December 9, 1957 [SRDB Ref ID 19710].

AEC 1957c, Letter from R. C. Armstrong, Director, Production Division, Oak Ridge Operations, 

to G. K. Crosby, The International Nickel Company, Inc., January 15, 1957 [SRDB Ref ID 

19726].

AEC 1957d, Memo from C.A. Keller, Director Production Division, to F.F. Gallagher, Security 

Division, September 11, 1957 [SRDB Ref ID 22125].

AEC 1957e, Memo from E. A. Wende, Deputy Manager Oak Ridge Operations, to Director, 

Production Division, Oak Ridge Operations, February 14, 1957 [SRDB Ref ID 22000]. 
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AEC 1958a, Memo from H. J. McAlduff, Deputy Director Inspection Division AEC, to C.A. 

Keller, Director Production Division AEC, March 5, 1958 [SRDB Ref ID 18870].

AEC 1958b, Memo from R. S. Kendall, Assistant Chief, Physical Security Branch AEC, to F.P. 

Callaghan, Director, Security Division AEC, February 3, 1958 [SRDB Ref ID 18866].

AEC 1958c, Letter from S. R. Sapirie, Manager, Oak Ridge Operations, to the International 

Nickel Company, Inc., February 28, 1958 [SRDB Ref ID 18869].

AEC 1958d, Appendix “C” – Specifications for Starting Material, undated c. 1958 [SRDB Ref 

ID 22105].

AEC 1958e, Letter from C. A. Keller, Director of Production Division, AEC, to A. P. Huber, 

Plant Superintendent, ORGDP, Union Carbide Nuclear Company, May 26 [SRDB Ref ID 

18811]. 

AEC 1960a, Memo from G. F. Quinn, Director, Division of Production, to S. R. Sapirie, 

Manager, Oak Ridge Operations, August 29 [SRDB Ref ID 19657].

AEC 1960b, Memo from R. G. Armstrong, Assistant Manager for Operations, to Charles A. 

Keller, Director, Production Division, December 7 [SRDB Ref ID 19655].

AEC 1961, Supplemental Information Required for Fiscal Year 1961 Budget [SRDB Ref ID 

22104].

AEC 1962a, Memo from John R. Moore, Assistant Manager for Administration, Oak Ridge 

Operations, AEC, to Leo W. Meyer, Division of Contracts, Washington D. C., AEC, et al., 

December 14, 1962 [SRDB Ref ID 21999, p. 24].

AEC 1962b, Memo to file by C. F. Fullam, Assistant Chief, Audit Branch, Shutdown of INCO 

Plant, August 17 [SRDB Ref ID 21999, p. 23].

Andersson, K.G., et al, Radiation Dose Implications of Airborne Contaminant Depositions to 

Humans, Health Physics, 82(2), pp. 226-232, 2002.

Battelle 2006, Site Profiles for Atomic Weapons Employers that Worked Uranium and Thorium 

Metals, Battelle-TBD_6000, Rev F0, PNWD-3738 Rev 0, December 13.

Berger, J.D., et al, Oak Ridge Associated Universities, Radiological Assessment of the Former 

Reduction Pilot Plant Huntington Alloys, Inc., Huntington, West Virginia, July 2, 1981 [SRDB 

Ref ID 9018].

BJC (Bechtel Jacobs Company) 2000, Recycled Uranium Mass Balance Project, Oak Ridge 

Gaseous Diffusion Plant Site Report, BJC/OR-584, June [SRDB Ref ID 16497].
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Browne 1986, Table of Radioactive Isotopes, E. Browne and R. Firestone, John Wiley and Sons, 

New York.

Clark, C., and Cottrell, D., Preliminary Radiological Survey of the Former Reduction Pilot, 

Huntington, West Virginia 1980 [SRDB Ref ID 9018].

CRC, Handbook of Chemistry and Physics, 67th Edition, R.C. Weast, Ph.D., Editor-in-Chief, 

CRC Press, Inc., Boca Raton, FL, 1986.

DOE 2004, DOE Standard, Guide of Good Practices for Occupational Radiological Protection 

in Uranium Facilities, DOE-STD-1136-2004, Washington, D.C., December. 

Enterline P.E, Marsh, G.M., Mortality Among Workers in a Nickel Refinery and Alloy 

Manufacturing Plant in West Virginia, JNCI 1982; 6:925-933

Fogh, C.L., et al, Quantitative Measurement of Aerosol Deposition on Skin, Hair and Clothing 

for Dosimetric Assessment – Final Report, Roskilde, Denmark: Riso National Laboratory, RisoR-1075(EN), 1999.

INCO 1957a, Letter from C.B. Goodrich, Section Head, Special Projects INCO, to N. Williams, 

Deputy Assistant Manager for Operations AEC, December 11, 1957 [SRDB Ref ID 19709].

INCO 1957b, Letter from C.B. Goodrich, Section Head, Special Projects INCO, to H.H. Stoner, 

Union Carbide, December 11, 1957 [SRDB Ref ID 18862].

INCO 1958, Letter from R.R. Clappier to S. R. Sapirie, Manager, Oak Ridge Operations, May 

15, 1958 [SRDB Ref ID 18874].

INCO 1959, Letter from C. B. Goodrich, Section Head – Technical Services, INCO, to C. A. 

Keller, Director, Production Division, AEC, July 6 [SRDB Ref ID 23564].

INCO 1962a, Letter from R. R. Clappier, INCO, to N. Williams, Deputy Assistant Manager for 

Operations, AEC, May 3 [SRDB Ref ID 21999, pp. 13-22].

Kocher, D.C., and Eckerman, K.F., Electron Dose-Rate Conversion Factors for External 

Exposures of the Skin from Uniformly Deposited Activity on the Body Surface, Health Physics, 

53(2), pp. 135-141, 1987.

NIOSH (National Institute for Occupational Safety and Health) 2007, Radiation Exposures 

Covered for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, OCAS-IG-003, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November [SRDB Ref ID 35987].

NiPera (Nickel Producers Environmental Research Association), Safe Use of Nickel in the 

Workplace, 1997, http://www.nipera.org/index.cfm/ci_id/13034/la_id/1.htm#APPENDIX%20A
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NRC 2002, U. S. Nuclear Regulatory Commission Regulatory Guide 8.30, Health Physics 

Surveys in Uranium Recovery Facilities, Rev 1, Figure 2, Beta Dose Rate from Yellowcake 

Separated from Ore for More Than 100 Days, May 2002.

OCAS (NIOSH Office of Compensation Analysis and Support) 2004, Estimation of Ingestion 

Intakes, OCAS-TIB-009, Rev 0, Cincinnati, OH, April 13.

ORAUT 2006, K-25 Gaseous Diffusion Plant - Occupational Internal Dose, ORAUT-TKBS0009-5, Rev 01 PC-1, October 4.

Smith, R.M., Final Report of Annual Fire Survey of Reduction Pilot Plant, Huntington, West 

Virginia, July 2-3, 1957 [SRDB Ref ID 18812].

UCNC (Union Carbide Nuclear Company) 1958a, Letter from W.E. Tewes, Technical Division 

Union Carbide, to C.B. Goodrich, INCO, January 31, 1958 [SRDB Ref ID 18865].

UCNC 1958b, Memo from H.F. Henry, Safety, Fire and Radiation Control to E.C. Bollinger, 

January 22, 1958 [SRDB Ref ID 18063]. 

UCNC 1958c, H. F. Henry, to E. C. Bollinger, Oak Ridge Gaseous Diffusion Plant, Study of 

INCO Radiation Problem, January 22, SRDB Ref ID 18863.

U.S. DOE (United States Department of Energy) Worker Advocacy Web Site Facility List, 

November 2007, http://tis.eh.doe.gov/advocacy/faclist/findfacility.cfm.

U.S. DHHS (United States Department of Health and Human Services), External Dose 

Reconstruction Implementation Guideline, Rev. 3, Appendix B, National Institute of 

Occupational Safety and Health, 2007.
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Attachment A

Determination of Dose from Nickel Scrap at the 

Huntington Pilot Plant

Background:

Huntington Pilot Plant (or Reduction Pilot Plant) produced nickel barrier material for the 

Atomic Energy Commission (AEC). One of the raw materials was contaminated nickel scrap

that was contaminated with low levels of low enriched uranium (AEC 1957c). The feed was 

called starting material (SM). The contaminated SM was processed in no more than 4,000 pound 

batches, and all residues from a batch were stored in a 20 gallon drum (INCO 1957b). 

Information from a variety of sources was used to develop source term information for the 

residues for the purpose of modeling the external dose to workers from these drums.

Geometry:

The uranium-bearing residues were stored in "20-gallon drums spaced over a 40 ft. by 

105 ft. area" (INCO 1957b) which implies a spaced configuration of the drums. For external 

dose modeling purposes it was assumed that the drums of residues are stored in a row without 

spacing, and that workers were exposed to the center of a row of drums containing the uranium 

bearing residues. 

Uranium Source term:

The material being considered was nickel scrap from the gaseous diffusion plants. 

Reference indicates that controls were in place to process no more than 4,000 pounds of material 

at one time. That was based on an assumption of an always safe U-235 mass of 350 grams 

(UCNC 1958c). Specifications included a maximum U-235 concentration in the feed of 0.0875 

gram per pound of SM (AEC 1958d). The UCNC (1958c) reference indicates that all the 

residues from a 4,000 pound batch would be placed in a 20 gallon drum, and that the maximum 

U-235 content in the drum was less than 350 grams, the always safe concentration. That 

reference also assumed the maximum total uranium concentration in the starting material was 

200 ppm.

Another reference provided guaranteed specifications (see Section 4.0) that the total 

uranium would always be less than 500 ppm. This would result in a limiting amount of 2 pounds 

of uranium per 4,000 pound batch of starting material feed. If all the contained uranium in a 

4,000 pound batch was collected as residue in a 20 gallon drum, the 2 pounds of total uranium 

per batch can be used to bound the uranium concentration of the residue. The maximum 

enrichment of received starting material is assumed to be 4% U-235. However, as discussed in 

Section 4.0, the average uranium concentration and enrichment were lower than these bounding 

assumptions used to model dose. Based on these values the maximum uranium content of each 

20 gallon drum was 908 grams of 4% enriched U per an evaluation done by Oak Ridge in 1958 

(UCNC 1958c).
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Drum contents:

The source term (908 grams of total uranium in 50 pounds of residue) was modeled for 

the determination of external dose using a fixed source term. The mass of residue was a variable 

for which we have no specific data. The Oak Ridge evaluation report (on radiation exposure 

from the drums) stated their evaluation was done with conservative assumptions. 

For the purposes of this evaluation, it was determined that a 50 pound source term and a 

container 3⁄4 full would be modeled. Since distance was a factor in estimating the dose, the 

drums were assumed to be 3/4 full (and hence contained 15 gallons of residue). Test runs of the 

process had been completed, hence a 20 gallon drum provided likely excess capacity per batch. 

For 50 pounds this results in a density of 0.4 g/cc. But, for example, if we assume a 2 g/cc 

density, the mass of the 15 gallons of residue would be 250 pounds, which is more than indicated 

by the references. Use of the lower density for a fixed source term was determined to be 

claimant favorable because of the reduced shielding implied.

For a constant radiological source term, an increase in density will result in a lower dose 

rate. So, we can bound dose rates by assuming 50 lb. of residue at a 0.4 g/cc density, which 

results in a height of approximately 14 inches (from assumed drum dimensions of 18.25 inches 

in diameter by 19.25 inches high). The contaminated nickel scrap had a specified 98% minimum 

nickel concentration (see Section 4.0). 

The residues contained nickel that was not fully recovered in the process (at one point the 

AEC was considering blending the residues with other Ni feed to recover more of the Ni at RPP)

(INCO 1959). This external dose analysis is based on the 98 % minimum nickel concentration in 

the feed and the assumption the nickel carbonyl process achieved 99% efficiency for nickel 

recovery, which would result in 40 pounds of nickel lost to the drummed residue. Based on the

limit for iron in the specifications, iron may have been present in trace quantities. For dose 

modeling purposes the residues are assumed to contain nickel and uranium. Other radiological 

impurities are assumed to be present based on a function of the uranium content. 

Radioactivity in a 20-gallon drum: Table A1 below contains radionuclides of concern including 

recycled U contaminates. 
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Table A1. Radionuclides in Drums of Residuea,b

Mass fraction Specific Act., 

μCi/g μCi/(g total U) Total in drum,μCi

U-234 negligible - 1.56 1420

U-235 0.04 2.16 0.0864 78

U-238 0.96 0.338 0.324 295

Th-234 295

Pa-234m 295

Pa-234 0.471

Th-231 78

Pu-239 0.00486

Np-237 0.00359

Tc-99 0.748

a. Specific concentration of 4% enriched uranium=1.97 μCi/g of uranium

b. Total mass of uranium in drum=908 g

Dose Calculations

Content of drums

Assume the material is 908 g of uranium, and the rest of the 50 pounds (22,680 grams) per drum 

was nickel (22,680-908 = 21,773 g of nickel). The material used for this analysis consisted of:

mass %

Total 

Mass (g)

Isotopic 

mass (g)

Mass % of 

Mixture

U-235 4.00% 907 36.28 0.16%

U-238 96.00% 907 870.72 3.84%

natural nickel 100.00% 21773 21773 96.00%

total 22680

Gamma dose rate

Results were determined using MCNPX 2.6f. 77.9 cm was chosen as the height of the testes and 

as an overestimate for many organs. 8 drums were modeled in a close packed geometry and the 

dose rate determined from the center of the drums.

Photons per decay of 238U

The number and energy of photons per unit decay of 238U was compiled from ENSDF files from 

NUDAT (11/28/1005) located at http://www.nndc.bnl.gov/nudat2/indx_dec.jsp. These 

emissions were binned by emission probability as summarized in the following table (normalized 

per decay of 238U). The resulting total photon emission probability per decay of 238U was 0.7728 

for the energy range of 0.001 to 2.2 MeV.
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Table A2: Gamma Ray Spectra Used for Analysis of Dose Rate 

(normalized per 238U decay)

Lower, 

MeV

Upper, 

MeV

Photons per 

decay 238U

Photons per 

hour per Ci 

238U

Relative 

flux

0.001 0.01 0.000087252 1.162E+10 0.01%

0.01 0.015 0.307244729 4.092E+13 39.76%

0.015 0.02 0.001287628 1.715E+11 0.17%

0.02 0.03 0.0373962 4.981E+12 4.84%

0.03 0.04 0.000142811 1.902E+10 0.02%

0.04 0.05 0.001174059 1.564E+11 0.15%

0.05 0.06 0.007275764 9.691E+11 0.94%

0.06 0.08 0.049808992 6.635E+12 6.45%

0.08 0.1 0.109028516 1.452E+13 14.11%

0.1 0.15 0.048335431 6.438E+12 6.25%

0.15 0.2 0.168322232 2.242E+13 21.78%

0.2 0.3 0.019368187 2.580E+12 2.51%

0.3 0.4 0.000392679 5.230E+10 0.05%

0.4 0.5 0.000169829 2.262E+10 0.02%

0.5 0.6 0.000370926 4.941E+10 0.05%

0.6 0.8 0.005213862 6.945E+11 0.67%

0.8 1 0.006946372 9.253E+11 0.90%

1 1.2 0.008687465 1.157E+12 1.12%

1.2 1.4 0.000162724 2.167E+10 0.02%

1.4 1.6 0.000488166 6.502E+10 0.06%

1.6 1.8 0.000369724 4.925E+10 0.05%

1.8 2 0.000539191 7.182E+10 0.07%

2 2.2 6.56E-08 8.738E+06 0.00%

Beta particle emission per decay of 238U

The number and energy of beta particle emissions per unit decay of 238U in aged uranium was 

compiled from the RADAR (Health Physics 83(4):471-475, 2002) and supplemented as 

necessary using data from the Table of Radioactive Isotopes (Browne 1986). These emissions 

were binned by emission probability as summarized in the following table (normalized per decay 

of 238U). The resulting total beta emission probability per decay of 238U was 2.20.
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Table A3: Beta Particle Emission Rates

Lower, 

MeV

Upper, 

MeV

Betas per 

decay 238U

Betas per 

second per 

Ci of 238U

Relative

flux

0.001 0.01 0.183809669 2.448E+13 8.35%

0.01 0.015 0.010281587 1.370E+12 0.47%

0.015 0.02 0.150193582 2.001E+13 6.82%

0.02 0.03 0.142229865 1.895E+13 6.46%

0.03 0.04 0.118989077 1.585E+13 5.41%

0.04 0.05 0.100401518 1.337E+13 4.56%

0.05 0.06 0.084726918 1.129E+13 3.85%

0.06 0.08 0.137967375 1.838E+13 6.27%

0.08 0.1 0.089276301 1.189E+13 4.06%

0.1 0.15 0.191455688 2.550E+13 8.70%

0.15 0.2 0.03381993 4.505E+12 1.54%

0.2 0.3 0.078860315 1.050E+13 3.58%

0.3 0.4 0.072597099 9.670E+12 3.30%

0.4 0.5 0.077502183 1.032E+13 3.52%

0.5 0.6 0.08059423 1.074E+13 3.66%

0.6 0.7 0.081581916 1.087E+13 3.71%

0.7 0.8 0.080614842 1.074E+13 3.66%

0.8 0.9 2.37416E-05 3.162E+09 0.00%

0.9 1 0.077835735 1.037E+13 3.54%

1 1.1 0.073435665 9.782E+12 3.34%

1.1 1.2 0.067673436 9.014E+12 3.07%

1.2 1.3 0.060809325 8.100E+12 2.76%

1.3 1.4 0.053143 7.079E+12 2.41%

1.4 1.5 0.0449251 5.984E+12 2.04%

1.5 1.65 0.03647 4.858E+12 1.66%

1.65 1.8 0.0281167 3.745E+12 1.28%

1.8 1.95 0.0331863 4.420E+12 1.51%

1.95 2.1 0.0071157 9.478E+11 0.32%

2.1 2.3 0.003223461 4.294E+11 0.15%

The dose rate fields in the plane of the 77.9 cm height (Figure A3 and A4) shows the dose rates

to be uniform in the center of the array. The following tallies were obtained:

 Tally 5: dose rate in rad/hr 30 cm from front edge of center barrel array, 77.9 cm above 

ground.

 Tally 15: dose rate in rad/hr 100 cm from front edge of center barrel array, 77.9 cm 

above ground.

 Tally 25: photon flux in photons/cm2/hr 30 cm from front edge of center barrel, 73 cm 

above ground.
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Table A4. Dose Rate from Gamma Sources

Distance from 

center drum

Tally Gamma Dose 

rate (rad/hr)

Brem. Dose 

Rate (rad/hr)

Total Dose 

Rate (rad/hr)

30 cm 5 2.56E-05 6.76E-6 3.24E-5

100 cm 15 1.34E-05 3.55E-6 1.70E-5

Approximately a 50/50 split for 0-250 keV/>250 keV for dose is supported by evaluating the 

cumulative dose graph for the dose from photon emissions and bremsstrahlung (Figures A1 and 

A2). Figures A3 and A4 show the dose rate distribution for photon dose and bremsstrahlung 

dose, respectively, in the horizontal plane (77.9 cm in height) which lies above the drums. The 

dose rate (rad/hr) levels are shown above the 8 drums configuration (at 77.9 cm above ground) 

from photon emissions. This corresponds to the dose rate in the plane (similar to a table top 77.9 

cm off the ground) above this array of drums. These figures provide further evidence in the 

claimant favorable nature of the location used for the dose evaluation.

cumulative fraction of dose from gamma emitters by energy Figure A1: Energy Distribution for Photon Emissions

Photon Energy (MeV)

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

Cumulative dose from photons

0.0

0.2

0.4

0.6

0.8

1.0

1.2

250 keV
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cumulative fraction of dose from beta emitters by energy (brem) Figure A2: Distribution of Dose from Bremsstrahlung

Photon Energy (MeV)

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

Cumulative bremsstrahlung dose from beta emitters

0.0

0.2

0.4

0.6

0.8

1.0

1.2

250 keV
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Figure A3: Photon Dose at 77.9 cm above Array of Drums (rad/hr)
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Figure A4: Bremsstrahlung Dose at 77.9 cm above Array of Drums (rad/hr)