ORAU Team 

NIOSH Dose Reconstruction Project

Technical Basis Document: Basis for Development of an Exposure 

Matrix for Huntington Pilot Plant 

Document Number: 

ORAUT-TKBS-0004 

Effective Date: 01/16/2004 

Revision No.: 01 

Controlled Copy No.: ______ 

Page 1 of 17

Subject Expert: Jeri L. Anderson 

Document Owner 

Approval: _Signature on File____________ Date: 01/16/2004

 Jeri L. Anderson, TBD Team Leader

Approval: _Signature on File ___________ Date: 01/16/2004 Judson L. Kenoyer Task 3 Manager

Concurrence: _Signature on File_________ Date: 01/16/2004 Richard E. Toohey, Project Director

Approval: _Signature on File____________ Date: 01/16/2004 James W. Neton, OCAS Health Science Administrator 

Supersedes: 

Revision No.: 00

TABLE OF CONTENTS 

Section Page

Record of Issue/Revisions......................................................................................................................3

1.0 Site Description, Operational History, and Process.......................................................................4

2.0 Estimation of Internal Exposure.....................................................................................................5

3.0 Estimation of External Exposure....................................................................................................9

4.0 Estimation of Exposure to Residual Activity.................................................................................12

References ...........................................................................................................................................15

LIST OF TABLES 

Table Page

1 Specifications for nickel-containing barrier material for use in production of nickel 

powder by the carbonyl process....................................................................................................5

2 Composition of uranium isotopes in enriched uranium (maximum enrichment of 39%)................6

3 Ratios of uranium isotopes to nickel in starting material................................................................7

4 Estimates of historical nickel exposures at the International Nickel Company plant in 

Huntington, WV..............................................................................................................................7

5 Estimated annual internal exposure to nickel dust contaminated with 39%-enriched 

uranium..........................................................................................................................................8

6 Annual organ doses due to exposure to surfaces contaminated with enriched uranium.............10

7 Annual organ doses due to exposure to a 5-by-5 array of birdcages ..........................................11

8 Annual organ doses due to the assumed annual diagnostic chest x-ray.....................................12

9 Annual organ doses due to external exposure to residual radioactivity from the end of 

AEC operations until decontamination.........................................................................................13
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 2 of 17 

10 Annual organ doses due to external exposure to residual radioactivity after 

decontamination ..........................................................................................................................14

LIST OF FIGURES 

1 Conceptual diagram of the nickel carbonyl process ..................................................................... 6

2 “Birdcage” container most likely used for storage and shipping of enriched uranium 

at the Huntington Pilot Plant ....................................................................................................... 11
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 3 of 17 

RECORD OF ISSUE/REVISIONS 

ISSUE 

AUTHORIZATION 

DATE 

EFFECTIVE 

DATE 

REV. 

NO. DESCRIPTION 

Draft 07/14/2003 00-A New document to establish the technical basis for 

the development of a radiation exposure matrix 

for Huntington Pilot Plant. Initiated by Jeri L. 

Anderson. 

Draft 10/15/2003 00-B Draft revision to incorporate NIOSH review 

comments. Initiated by Jeri L. Anderson. 

Draft 10/22/2003 00-C Draft revision to incorporate NIOSH review 

comments. Initiated by Jeri L. Anderson. 

10/31/2003 10/31/2003 00 First approved issue. Initiated by Jeri L. 

Anderson. Restricted use – This TBD cannot be 

used to process cases that involve significant 

exposures to electrons (e.g., skin cancers). 

Draft 12/19/2003 01-A Draft revision to resolve issue of exposure to 

electrons and make document applicable to skin 

cancers. Initiated by Jeri L. Anderson. 

Draft 01/13/2004 01-B Draft revision to incorporate NIOSH review 

comments. Initiated by Jeri L. Anderson. 

01/16/2004 01/16/2004 01 Approved Issue of Revision 01. Initiated by Jeri L. 

Anderson. 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 4 of 17 

Basis for Development of an Exposure Matrix for 

HUNTINGTON PILOT PLANT 

Huntington, West Virginia 

Period of Operation: 1951-1963 

1.0 Site Description, Operational History, and Process

According to reports, 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 2002). A 1957 letter supplement also indicates that the 

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

(Humphries 1957). A facility data report (U.S. AEC 1955) and an International Nickel Company 

(INCO) memorandum (Keller 1957) both suggest that the powder was shipped to K-25 for barrier 

production. 

An additional document (DOEIC 2000a) gave details of starting material and barrier 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 survey report (White 1950) indicates that in January 1950, the AEC was considering 

awarding a contract to the International Nickel Company to melt K-25 scrap at their plant in 

Huntington, West Virginia. The work was to be performed in Furnace No. 5 in the Refinery Building. 

According to the survey, a railroad siding adjoined the Refinery Building, which would make it possible 

to unload and weigh the scrap at the siding, place 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 the furnace 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). 

In a letter supplement to Contract No. AT-(40-1)-1092, the AEC wrote to confirm that the INCO was 

willing to produce up to eight tons of metallic nickel powder with the required specifications 

(Humphries 1957). There were two sources of feed material for the nickel powder production. One 

source was uncontaminated nickel sinter. The other source was the above mentioned nickel scrap 

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

maximum enrichment of 4% (by weight) (Clark and Cottrell 1980, Berger 1981). No information could 

be found regarding analyses performed to determine the extent of the uranium contamination. 

However, a recently declassified document entitled “Appendix ‘C’ – Specifications For Starting 

Material” (DOEIC 2000b) detailed the required minimum and maximum constituents for the nickelcontaining barrier material. These specifications are shown in Table 1. 

The Huntington Pilot Plant was shut down in 1963 and maintained in stand-by condition by Huntington 

Alloys, Inc., a subsidiary of INCO. The Department of Energy (DOE) performed a preliminary 

radiological survey of the site and recommended that the site be decontaminated. The 

decontamination of the facility took place between November 27, 1978 and May 18, 1979. According 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 5 of 17 

Table 1. Specifications for nickel-containing barrier material for 

use in production of nickel powder by the carbonyl process. 

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 -- 

to reports, the classified and contaminated scrap was 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 

site was the compressor building, which is a concrete block building consisting of large open storage 

area, a change room facility with an office upstairs, and a small electrical craft training room. The site 

was surveyed again in May 1979 and recommended for unrestricted release by the DOE. 

A review of the May 1979 survey results by the Health and Safety Research Division of Oak Ridge 

National Laboratory (ORNL) revealed certain areas that required further evaluation. On August 13, 

1980, ORNL performed a preliminary survey at the site to determine the need for additional surveys 

(Clark and Cottrell 1980). ORNL recommended a detailed formal survey that was performed by Oak 

Ridge Associated Universities’ (ORAU) Radiological Site Assessment Program on November 17-19 

and December 3, 1980 (Berger 1981). The results of the formal survey showed that the radiation 

levels at the site were within guidelines for unrestricted release of the property. 

In the 1980 formal survey, radionuclide (U-238, U-235, and Ra-226) and nickel concentrations were 

measured in soil samples collected both onsite and offsite. The average concentrations of U-238 and 

U-235 in the onsite soil samples were 21 and 0.40 pCi/g of soil, respectively. The average 

concentrations of U-238 and U-235 in offsite soil samples were 4.5 and 0.06 pCi/g, respectively. 

Nickel concentrations were also measured in the soil samples. The average nickel concentration 

onsite was 8.2 mg/g of soil and the average concentration offsite was 3.5 mg/g of soil. 

Results of the survey of the compressor building were compared with U.S. NRC release guidelines 

(US NRC 1976) and American National Standard Institute (ANSI) release standards (ANSI 1978) for 

uranium and its decay products. The comparison showed that measured levels were below the 

guidelines for both transferable and total alpha and beta-gamma contamination. External betagamma radiation dose rates at 1 cm from building surfaces were also below NRC guidelines. The 

NRC and ANSI documents provided no guidelines for gamma exposure levels. However, assuming a 

continuous (168 hours/week) exposure to the highest levels measured inside the compressor building 

resulted in an annual dose of less than 305 mrem. The same exposure time to outdoor gamma 

radiation levels would result in an annual dose of less than 200 mrem. 

2.0 Estimation of Internal Exposure

No records were found of radioactivity monitoring during the period the K-25 scrap nickel was melted 

at the Huntington Plant. Thus, it was necessary to make a reasonable estimate of uranium air 

concentration and area contamination in order to estimate exposure to workers during the period that 

the plant operated under contract with the AEC. 

International Nickel Company used the carbonyl process to produce high-purity nickel powders. The 

highly selective process involves exposing the melted nickel containing contaminants to carbon 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 6 of 17 

monoxide gas. This results in the formation of nickel carbonyl gas, which is siphoned off while the 

contaminants remain behind in the residue. The nickel carbonyl gas is decomposed at moderate 

temperatures producing extremely pure nickel, and carbon monoxide that is recycled through the 

process. Figure 1 is a conceptual diagram of the nickel carbonyl process. 

Figure 1. Conceptual diagram of the nickel carbonyl process. (Illustration courtesy of S. Jensen) 

The scrap was supposedly shipped to INCO in cartons that remained sealed during weighing and 

charging of the furnace, the potential for exposure prior to the melting process was considered to be 

low. The potential for external exposure is highest in the area where the melted scrap is stored and 

where the residues from the carbonyl process are disposed of. Information regarding the disposition 

of the carbonyl process residues appears to remain classified. 

According to the required specifications for the starting material, the maximum enrichment allowed 

was approximately 39%. However, most reports indicate that the maximum enrichment was ~4%. To 

be claimant favorable, it was assumed the uranium contaminating the nickel scrap was enriched to 

39% by weight. Table 2 shows the composition of the uranium isotopes in uranium enriched to 39%. 

Table 2. Composition of uranium isotopes in 

enriched uranium (maximum enrichment of 

39%). 

Isotope Weight percent Activity percent 

U-234 0.3471 94.9106 

U-235 39.0000 3.6914 

U-236 0.1794 0.5102 

U-238 60.4735 0.8877 

Table 3 shows the calculated ratio of the uranium isotopes to nickel in the starting material based on 

the required specifications and assuming an enrichment of 39%. 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 7 of 17 

Table 3. Ratios of uranium isotopes 

to nickel in starting material. 

Uranium 

isotope 

Ratio to Ni 

(pCi/mg Ni) 

U-235 0. 425 

U-234 10.901 

U-238 0.102 

Documents giving details of starting material shipped from ORGDP to INCO in 1950-1961 indicated 

that annually between 1,980,000 and 2,587,000 pounds of contaminated nickel scrap could have 

been melted and used to produce nickel powder. Thus, assuming all starting material was 

contaminated with 39% enriched uranium, it is estimated that a residue containing approximately 

0.37-0.49 Ci of U-235, 9.6-13 Ci of U-234, and 0.09-0.12 Ci of U-238 could have been produced per 

year. It is important to note that these quantities for nickel scrap are only the amounts reported by 

ORGDP. It is assumed that both the Paducah and Portsmouth gaseous diffusion plants contributed 

scrap material, although the quantities are not known at this time. 

Workers in nickel smelting operations are exposed to mean nickel aerosol concentrations ranging 

from 0.01 to less than 3.0 mg Ni/m3

. Mean nickel aerosol concentrations for refining operations range 

from 0.003 to about 1.5 mg Ni/m3

 (NiPERA 1997). The nickel concentration in air at the Huntington 

Plant was estimated from exposure information collected during a nickel worker mortality study. Table 

4 shows estimates of nickel exposures at the plant. 

Table 4. Estimates of historicala

 nickel exposures at the International Nickel Company plant in 

Huntington, WV (reproduced from Enterline and Marsh 1982). 

Department 

Exposure 

(mg Ni/m3

) 

Acid reclaim 0.02 

Blacksmith shop 0.02 

Carpenter shopb 0.02 

Chipping and hammer 0.75c

Cold drawing (acid reclaim 1970 on) 0.10c

Combustionb 0.05 

Electricalb 0.10 

Extrusion 0.20c

General officesb 0.01 

Heat treatment 0.02 

Machine shop 0.03c

Mechanicalb 0.01 

Melting and casting 0.25c

Merchant mill 0.30c

Night superintendentb 0.01 

Pipe shopb 0.08 

Polishing 0.10c

Primary mill 0.06 

Reduction pilot plant 0.02 

Department 

Exposure 

(mg Ni/m3

) 

Refinery 5.00c

Research and developmentb 0.05 

Roll grinding 0.03 

Roll turning 0.03 

Sheet mill (acid reclaim to 1970) 0.20c

Shipping 0.01 

Standards or industrial engineeringb 0.01 

Steel shopb 0.12 

Stocks 0.01 

Stores 0.01 

Strip mill 0.20c

Transportationb 0.01 

Vacuum melting 0.15 

Welding products 0.02 

Yardb 0.10 

Warehouse 0.01 

Watchmanb 0.01 

Inspection (unassigned) b 0.05 

a. Based on current measurements except as noted. 

b. Plant-wide work assignment. 

c. Based on historic midget impinger counts. 

These exposure estimates, made by International Nickel Company 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, 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 8 of 17 

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). 

According to Table 4, air concentrations of nickel at the Huntington plant ranged from 0.01 to 5.0 mg 

Ni/m3

 while the air concentration in the Reduction Pilot Plant was estimated at 0.02 mg Ni/m3. These 

measured values appear to be lognormally distributed with a median of 0.05 mg Ni/m3

 and a 

geometric standard deviation (GSD) of 4.3. To be claimant favorable, median (0.05 mg Ni/m3

) of the 

values in Table 4 is used in combination with the ratios of U-235, U-234, and U-238 to nickel 

calculated from the required specifications (see Table 3) to create a lognormal distribution that can be 

used to estimate worker exposure to enriched uranium. 

Table 5 shows the calculation of the annual intake of 

39% enriched uranium from the estimated annual 

internal exposure to nickel dust contaminated with 

enriched uranium. To obtain these estimates, the ratio 

of enriched uranium to nickel is multiplied by the 

median nickel concentration in air to estimate a median 

enriched uranium air concentration. This air 

concentration is multiplied by the breathing rate and the 

number of hours worked to obtain a median annual 

intake. The breathing rate was calculated from the 

volume of air breathed for an adult light worker shown 

in ICRP Publication 66, Table 6 on pg. 23 (ICRP 1994). 

This category assumes an activity distribution of 1/3 

sitting and 2/3 light exercise. 

It was assumed that the worker was exposed to the contaminated nickel dust for a 2000-hour work 

year. Because the exact dates of operation are unknown, to be claimant favorable, it is assumed the 

the plant operated from January 1951 through December 1963. A survey report suggested that the 

melting of the contaminated nickel scrap was scheduled for two weeks per month (White 1950). Also, 

other sources of uncontaminated nickel were used for producing the metallic nickel powder (Clark and 

Cottrell 1980). Thus, this assumption is considered to be claimant favorable. 

In order to make use of the information in Table 5 for estimation of organ dose from internal exposure, 

the annual doses for the organ of interest should be calculated in the IMBA NIOSH program for the 

median intake of U-234 assuming a chronic exposure. The calculated annual organ doses are then 

entered into the NIOSH IREP program as the annual dose due to a chronic exposure to alpha 

radiation using a lognormal distribution with a GSD of 4.3. 

The chemical form of the uranium contaminating the nickel dust is unknown. Many details of gaseous 

diffusion technology remain classified such as the chemical effect of UF6 on various components of 

gaseous diffusion plants (US DOE 1994). Prior to the enrichment by gaseous diffusion, the uranium 

compound is converted to UF6, which when exposed to moisture, breaks down to form HF and UO2F2. 

The HF most likely reacts with the nickel to form nickel fluorides. Other possible compounds are U, 

UF3, UF4, and UO2 (Schneider 2003, Zhu 2003). UO2F2 exhibits behavior consistent with Type F 

absorption (ICRP 1995, Schneider 2003). Human and animal studies suggest that UO2 should be 

assigned to Absorption Type S. Uranium fluorides exhibit more complex behavior with some studies 

indicating assignment to Type F and some studies indicating assignment to Type M. The ICRP in 

Publication 71 recommends the use of Absorption Type M in the absence of specific information 

Table 5. Estimated annual internal 

exposure to nickel dust contaminated 

with 39%-enriched uranium. 

 Median 

Ni Air Concentration 

(mg/m3) 

0.05 

EU Air Concentration based on 

Ni Air Concentration (pCi/m3

) 

0.58 

Breathing Rate 

(m3

/h) 

1.2 

Hours Worked 

(h) 

2,000 

Annual Intake 

(pCi)

1,400 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 9 of 17 

(ICRP 1995). The ICRP Publication 66 (ICRP 1994) default values should be used for the deposition 

parameters in the IMBA NIOSH program. 

3.0 Estimation of External Exposure

For estimating maximum external exposure due to submersion in air containing nickel dust 

contaminated with enriched uranium, the maximum air concentration values and work times per year 

from Table 5 were combined with dose coefficients for U-235, U-234, and U-238 and its daughter 

radionuclides Th-234 and Pa-234m from Federal Guidance Report No. 12 (US EPA 1993). The 

median annual organ dose estimates due to submersion in air for 1951 through 1963 were calculated. 

The cumulative maximum organ dose for 1951-1963 is less than 1 mrem and is therefore not included 

in the dose estimation. 

To estimate external dose due to penetrating radiation from contaminated surfaces in the Reduction 

Pilot Plant, the estimated air concentration was multiplied by the indoor deposition velocity and the 

assumed exposure time. The indoor deposition velocity is dependent on the physical properties of the 

room (air viscosity and density, turbulence, thermal gradients, surface geometry) and the particles 

(diameter, shape, density). These characteristics are unknown, thus the terminal gravitational 

settling velocity of 0.002 m/s was used as an estimate of the velocity of deposition to surfaces in the 

building (CRC 1986). This settling velocity is for 5-μm particles with a density of 2.0 g/cm3

 in air at 

25°C and 1 atm. This value falls in the range of deposition velocities (2.7E-06 to 2.7E-03 m/s) 

measured in various studies (US NRC 2002c) and is considered a conservative estimate. Also, room 

air exchange rates, ventilation, and plant housekeeping practices are unknown, so it is assumed that 

there is a steady state air concentration and surface contamination. The amount of surface 

contamination is assumed to be due to a 10-year buildup. 

The estimated activity of the surface contamination is multiplied by the dose coefficients for 

contaminated ground surfaces for U-238, U-234, U-235, Th-234, and Pa-234m from Federal 

Guidance Report No. 12 (US EPA 1993). The annual organ dose estimated by this method is less 

than 1 mrem and is therefore not used for the photon dose estimation. Instead, survey data obtained 

from one of the remaining Huntington Pilot Plant structures (Berger 1981) is used to estimate external 

photon dose. 

According to the 1981 survey report, only the compressor building remained at the site. Gamma 

radiation levels measured at 1 meter above the floor ranged from 9-12 μR/h with a maximum reading 

of 35 μR/h. To give claimant favorable estimate of external photon exposure during operations, it was 

assumed that the median exposure rate was 12 μR/h and the upper 95th percentile was 35 μR/h, 

giving a GSD of 1.9. Thus, to estimate potential external exposure to contaminated surfaces in the 

plant, the median exposure rate was multiplied by the Exposure (R) to Organ Dose (rem) photon dose 

conversion factors from Appendix B of the NIOSH External Dose Reconstruction Implementation 

Guideline (DHHS 2002). The exposure geometry was assumed to be isotropic and the exposure rate 

was divided evenly between the conversion factors for photons with energy between 30 and 250 keV 

and photons with energy greater than 250 keV. Table 6 shows the calculated annual organ doses 

from exposure to contaminated surfaces during plant operations. 

It was also assumed that there was a potential to receive a shallow dose from electrons due to skin 

contamination. The amount of skin contamination was calculated by using a measured deposition 

velocity to skin of 0.012 m/s (Andersson et al. 2002, Fogh et al. 1999). For simplification, it was 

assumed that the material deposited on the skin during an 8-hour period was all deposited at the 

beginning of the shift, and the worker took a shower at the end of the shift. To determine the electron 

shallow dose, the quantity of skin contamination was combined 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 10 of 17 

Table 6. Annual organ doses due to exposure to surfaces 

contaminated with enriched uranium. 

Annual organ dose (rem) 

Organ 

Photons 

E=30-250 keV 

Photons 

E>250 keV Total 

Bladder 6.43E-03 7.76E-03 1.42E-02 

Red bone marrow 6.68E-03 7.99E-03 1.47E-02 

Bone surface 1.13E-02 8.17E-03 1.94E-02 

Breast 8.50E-03 8.89E-03 1.74E-02 

Colon 6.18E-03 7.61E-03 1.38E-02 

Esophagus 6.00E-03 7.85E-03 1.38E-02 

Eye 8.90E-03 9.11E-03 1.80E-02 

Ovaries 5.93E-03 7.51E-03 1.34E-02 

Testes 7.58E-03 8.32E-03 1.59E-02 

Liver 6.82E-03 7.98E-03 1.48E-02 

Lung 7.50E-03 8.47E-03 1.60E-02 

Remainder organs 6.66E-03 7.96E-03 1.46E-02 

Skin 8.77E-03 9.11E-03 1.79E-02 

Stomach 6.79E-03 7.97E-03 1.48E-02 

Thymus 7.37E-03 8.18E-03 1.56E-02 

Thyroid 7.67E-03 8.57E-03 1.62E-02 

Uterus 5.82E-03 7.25E-03 1.31E-02 

with electron dose-rate conversion factors for U-238, U-234, U-235, Th-234, and Pa-234m for skin in 

contact with radionuclides (Kocher and Eckerman 1987). This resulted in an annual skin dose of 

4.6E-06 rem. This value is insignificant and is not included in the dose estimation. 

Another potential for exposure existed in a worker’s proximity to the carbonyl process residue, which 

contained an unknown quantity of enriched uranium. This process residue was most likely decanted 

into storage units known as “birdcages.” Figure 2 below is an illustration of a birdcage showing the 

dimensions of the container as well as a 5-by-5 double-stacked array of birdcages. It was assumed 

that basic nuclear safety parameters were utilized at the plant to prevent the occurrence of a criticality 

accident. There are three general methods of control for preventing nuclear accidents: mass control, 

geometry control, and concentration control. According to an AEC report (US AEC 1961), the 

recommended maximum mass of U-235 in a container is 350 g (mass control) and the maximum 

diameter of infinite cylinder is 5.0 inches (geometry control). These parameters are based on 

moderation by water and were most likely used in the event of flooding of the storage area or fire. 

During the carbonyl process, it was possible to control the mass of U-235 by limiting the amount of 

contaminated scrap that was introduced into the system at one time. 

To make a reasonable estimation of the potential deep dose to a worker from this material, it was 

assumed that a worker was located at 1 meter from the 5-by-5 double-stacked array for 2000 hours 

per year. This resulted in an annual exposure of 0.13 R per year. To estimate an upper 95th

percentile exposure, it was assumed that the worker was located 1 foot from the stacked array for 

2000 hours per year resulted in an annual exposure of 0.76 R per year. 

The organ doses were calculated by multiplying the estimated annual exposure of 0.13 R by the 

“Exposure (R) to Organ Dose (HT)” photon dose conversion factors found in Appendix B of the NIOSH 

External Dose Reconstruction Implementation Guideline (DHHS 2002). The exposure geometry was 

assumed to be anterior-posterior (AP) and the dose rate was divided equally between photons with 

E=30-250 keV and photons with E>250 keV. Table 7 below shows the annual organ doses due to the 

potential exposure to the birdcage array. 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 11 of 17 

Figure 2. “Birdcage” container most likely used for storage and shipping of enriched uranium 

at the Huntington Pilot Plant. (Reproduced courtesy of S. Jensen from Paxton 1974) 

Table 7. Annual organ doses due to exposure to a 5-by-5 

array of birdcages. 

Annual organ dose (rem) 

Organ 

Photons 

E=30-250 keV 

Photons 

E>250 keV Total 

Bladder 8.09E-02 5.74E-02 1.38E-01 

Red bone marrow 4.07E-02 4.68E-02 8.75E-02 

Bone surface 7.99E-02 4.97E-02 1.30E-01 

Breast 8.23E-02 6.05E-02 1.43E-01 

Colon 6.89E-02 5.49E-02 1.24E-01 

Esophagus 4.47E-02 4.84E-02 9.31E-02 

Eye 8.03E-02 5.72E-02 1.38E-01 

Ovaries 6.21E-02 5.32E-02 1.15E-01 

Testes 9.32E-02 6.12E-02 1.54E-01 

Liver 6.92E-02 5.49E-02 1.24E-01 

Lung 6.41E-02 5.47E-02 1.19E-01 

Remainder organs 5.71E-02 5.12E-02 1.08E-01 

Skin 5.80E-02 5.43E-02 1.12E-01 

Stomach 8.13E-02 5.75E-02 1.39E-01 

Thymus 9.15E-02 5.80E-02 1.50E-01 

Thyroid 9.36E-02 6.32E-02 1.57E-01 

Uterus 6.57E-02 5.11E-02 1.17E-01 

The organ doses in the second and third columns of Table 7 are entered into the NIOSH IREP 

program assuming a chronic exposure and a lognormal distribution with a GSD of 2.9. The organ 

doses in the second column are attributed to photons with E=30-250 keV and the organ doses in the 

third column are attributed to photons with E>250 keV. 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 12 of 17 

The workers were also assumed to have had a potential for exposure to electrons that could result in 

a skin dose. This potential existed while transferring the carbonyl process residue containing 

enriched uranium from the reaction chamber to cans, which were then assayed and loaded into the 

birdcages. To estimate this potential exposure, beta dose rates at the surface of uranium metal 

according to percent enrichment were used (US DOE 2001). By combining these data with data 

showing the beta dose rates as a function of distance from the surface of aged yellowcake (US NRC 

2002a), it was possible to extrapolate dose rates from the carbonyl process residue at 20 cm from the 

surface of the residue. The range of enrichment of the uranium in the process residue was 4% to 

40%. The dose rates at 20 cm from the residue ranged from 1.7 mrem/h for 40% enriched, which was 

assumed to be the median, to 6.3 mrem/h for 4% enriched, which was assumed to be the upper 95th

percentile. The workers were assumed to spend 2 hours per day for 250 days per year assaying and 

loading the birdcages. Thus, the annual dose to the skin was estimated to be lognormally distributed, 

with a median of 0.85 rem and a GSD of 2.2. This dose rate is entered into the NIOSH IREP program 

assuming a chronic exposure to electrons with E>15 keV. 

The assumption was also made that workers received an annual occupationally related diagnostic xray. The exposure geometry was assumed to be posterior-anterior (PA) (DHHS 2002). The air kerma 

at skin entrance for the diagnostic chest x-ray was estimated to be 0.108 R (Scalsky 2003). The 

organ doses were calculated by multiplying the air kerma by the “Kerma (Ka) to Organ Dose (HT)” 

photon dose conversion factors for 30-250 keV photons found in Appendix B of the NIOSH External 

Dose Reconstruction Implementation Guideline (DHHS 2002). Table 8 below shows the annual organ 

doses due to the assumed annual diagnostic chest x-ray. 

The values in Table 8 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%. 

Estimation of Exposure to Residual Activity 

After conclusion of the AEC activities at the Huntington Pilot 

Plant sometime in 1963, the plant was shut down and 

maintained in stand-by condition (US DOE 1983). The site 

was then decontaminated between November 1978 and May 

1979. All structures except a compressor building was 

removed and classified and contaminated scrap was buried 

at a classified location at the Portsmouth Plant in Ohio. No 

results of the pre-decontamination survey could be found. 

According to a post-decontamination survey in 1980 (Berger 

1981), exposure rates at 1 meter were between 9 and 12 

μR/h with a maximum of 35 μR/h. 

To estimate potential exposure to residual radioactivity from 

the end of AEC operations until decontamination of the site, 

the maximum exposure rate measured at the site in the 1980 

survey was assumed to be the median exposure rate. The 

maximum exposure level recommended by the NCRP of 500 

mrem per year (250 μR/h for a 2000 hour work year) was assumed to be the upper 95th percentile 

exposure rate at the site. The annual organ doses from contaminated surfaces was calculated by 

multiplying the assumed median annual exposure of 70 mR (35 μR/h for 2000 hours) by the Exposure 

(R) to Organ Dose (rem) photon dose conversion factors from Appendix B of the NIOSH External 

Dose Reconstruction Implementation Guideline (DHHS 2002). The exposure geometry was assumed 

to be isotropic and the annual exposure was divided evenly between the conversion factors for 

Table 8. Annual organ doses due 

to the assumed annual diagnostic 

chest x-ray. 

Organ 

Annual dose 

(rem) 

Bladder 0.074 

Red bone marrow 0.124 

Bone surface 0.178 

Breast 0.059 

Colon 0.095 

Esophagus 0.105 

Eye 0.022 

Ovaries 0.110 

Testes 0.061 

Liver 0.101 

Lung 0.133 

Remainder organs 0.110 

Skin 0.110 

Stomach 0.077 

Thymus 0.048 

Thyroid 0.052 

Uterus 0.096 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 13 of 17 

photons with energy between 30 and 250 keV and photons with energy greater than 250 keV. Table 

9 shows the calculated annual organ doses from external exposure to residual radioactivity after the 

end of AEC operations at the site. 

The organ doses in the second and third columns of Table 9 are entered into the NIOSH IREP 

program assuming a chronic exposure and a lognormal distribution with a GSD of 3.3. The organ 

doses in the second column are attributed to photons with E=30-250 keV and the organ doses in the 

third column are attributed to photons with E>250 keV. The residual contamination exposure period is 

assumed to begin on January 1, 1964 and end on May 31, 1979. 

Table 9. Annual organ doses due to external exposure to 

residual radioactivity from the end of AEC operations until 

decontamination. 

Annual organ dose (rem) 

Organ 

Photons 

E=30-250 keV 

Photons 

E>250 keV Total 

Bladder 1.88E-02 2.26E-02 4.14E-02 

Red bone marrow 1.95E-02 2.33E-02 4.28E-02 

Bone surface 3.28E-02 2.38E-02 5.67E-02 

Breast 2.48E-02 2.59E-02 5.07E-02 

Colon 1.80E-02 2.22E-02 4.02E-02 

Esophagus 1.75E-02 2.29E-02 4.04E-02 

Eye 2.60E-02 2.66E-02 5.25E-02 

Ovaries 1.73E-02 2.19E-02 3.92E-02 

Testes 2.21E-02 2.43E-02 4.64E-02 

Liver 1.99E-02 2.33E-02 4.32E-02 

Lung 2.19E-02 2.47E-02 4.66E-02 

Remainder organs 1.94E-02 2.32E-02 4.26E-02 

Skin 2.56E-02 2.66E-02 5.22E-02 

Stomach 1.98E-02 2.32E-02 4.31E-02 

Thymus 2.15E-02 2.39E-02 4.54E-02 

Thyroid 2.24E-02 2.50E-02 4.74E-02 

Uterus 1.70E-02 2.11E-02 3.81E-02 

The results of the 1980 post-decontamination survey of the remaining compressor building were used 

to estimate potential exposure to residual activity after decontamination operations were completed at 

the site. To estimate the annual organ doses from contaminated surfaces in the plant, the median 

annual exposure of 24 mR (12 μR/h for 2000 hours) was multiplied by the Exposure (R) to Organ 

Dose (rem) photon dose conversion factors from Appendix B of the NIOSH External Dose 

Reconstruction Implementation Guideline (DHHS 2002). The exposure geometry was assumed to be 

isotropic and the annual exposure was divided evenly between the conversion factors for photons with 

energy between 30 and 250 keV and photons with energy greater than 250 keV. Table 10 shows the 

calculated annual organ doses from external exposure to residual radioactivity after decontamination 

of the site. 

The organ doses in the second and third columns of Table 10 are entered into the NIOSH IREP 

program assuming a chronic exposure and a lognormal distribution with a GSD of 1.9. The organ 

doses in the second column are attributed to photons with E=30-250 keV and the organ doses in the 

third column are attributed to photons with E>250 keV. The residual contamination exposure period is 

assumed to begin on June 1, 1979 and end on the employee’s last day of work at the site. 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 14 of 17 

Table 10. Annual organ doses due to external exposure to 

residual radioactivity after decontamination. 

Annual organ dose (rem) 

Organ 

Photons 

E=30-250 keV 

Photons 

E>250 keV Total 

Bladder 6.43E-03 7.76E-03 1.42E-02 

Red bone marrow 6.68E-03 7.99E-03 1.47E-02 

Bone surface 1.13E-02 8.17E-03 1.94E-02 

Breast 8.50E-03 8.89E-03 1.74E-02 

Colon 6.18E-03 7.61E-03 1.38E-02 

Esophagus 6.00E-03 7.85E-03 1.38E-02 

Eye 8.90E-03 9.11E-03 1.80E-02 

Ovaries 5.93E-03 7.51E-03 1.34E-02 

Testes 7.58E-03 8.32E-03 1.59E-02 

Liver 6.82E-03 7.98E-03 1.48E-02 

Lung 7.50E-03 8.47E-03 1.60E-02 

Remainder organs 6.66E-03 7.96E-03 1.46E-02 

Skin 8.77E-03 9.11E-03 1.79E-02 

Stomach 6.79E-03 7.97E-03 1.48E-02 

Thymus 7.37E-03 8.18E-03 1.56E-02 

Thyroid 7.67E-03 8.57E-03 1.62E-02 

Uterus 5.82E-03 7.25E-03 1.31E-02 

The maximum internal exposure from residual radioactivity during both post-AEC time periods was 

estimated by assuming that the facility was uniformly contaminated at the level of maximum smear 

result (considered removable contamination) of 140 dpm/100 cm2

. This value was multiplied by a 

resuspension factor of 1E-06 m-1 (US NRC 2002b). This resulted in an estimated maximum residual 

air concentration of 0.006 pCi/m3

. Assuming a breathing rate of 1.2 m3/h and a 2000-h work year 

results in a possible annual inhalation intake of 14 pCi. This value is considered negligible as it 

results in an annual dose of less than 1 mrem to the maximally exposed organ and is not included in 

the dose reconstruction. 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 15 of 17 

REFERENCES 

American National Standards Institute (ANSI), Draft American National Standard, Control of 

Radioactive Surface Contamination on Materials, Equipment, and Facilities to be Released for 

Uncontrolled Use, ANSI 13.12-1978, New York, ANSI, 1978 

Andersson, K.G., Fogh, C.L., Byrne, M.A., Roed, J., Goddard, A.J.H., Hotchkiss, S.A.M., “Radiation 

dose implications of airborne contaminant deposition to humans,” Health Physics, 82(2), pp. 

226-232, 2002 

Berger, J.D., Radiological Assessment of the Former Reduction Pilot Plant, Huntington Alloys, Inc., 

Huntington, WV, Final Report Prepared by Oak Ridge Associated Universities for Office of 

Operational Safety, U.S., DOE, July 1981 

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

Huntington, WV, ORNL, October 1980 

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

Press, Inc., Boca Raton, FL, 1986 

Department of Energy Information Center (DOEIC), 6/9/00 RHTG 86249 OR0014842 Oak Ridge 

Gaseous Diffusion Plant, Detail of Nickel Ingot Transactions, Supplemental Information 

Requested for FY 1961 Budget, 6/30/61, sanitized 9/25/99 (DOEIC 2000a) 

Department of Energy Information Center (DOEIC), 7/13/00 RHTG113949 Appendix ‘C’ Specifications 

for Starting Material – Nickel-containing Barrier, Declassified 6/28/99 (DOEIC 2000b) 

Enterline, P.E. and March, G.M., “Mortality among workers in a nickel refinery and alloy manufacturing 

plant in West Virginia, J National Cancer Inst, 68, p. 925-933, 1982 

Frey, H.C., and Cullen, A.C., “Distribution Development for Probabilistic Exposure Assessment,” 

Paper No. 95-94.02, Proceedings of the 88th Annual Meeting of the Air & Waste Management 

Association, San Antonio, Texas, June 18-23, 1995 

Fogh, C.L. Byrne, M.A., Andersson, K.G., Bell, K.F., Roed, J., Goddard, A.J.H., Vollmair, D.V., 

Hotchkiss, S.A.M., Quantitative measurement of aerosol deposition on skin, hair and clothing 

for dosimetric assessment—final report, Roskilde, Denmark: Riso National Laboratory, RisøR-1075(EN), 1999 

Humphries, R.O., Letter to R.R. Clappier of the International Nickel Company, Inc., Letter Supplement 

to Contract No. AT-(40-1)-1092, U.S. Atomic Energy Commission, June 13, 1957 

International Commission on Radiological Protection, “Age-dependent Doses to Members of the 

Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients,” ICRP Publication 71, 

1995 

International Commission on Radiological Protection, “Human Respiratory Tract Model for 

Radiological Protection,” ICRP Publication 66, 1994 

Keller, Charles A., Director, Production Division, Memorandum to F.F. Callaghan, Director, Security 

Division, International Nickel Company, Inc., September 11, 1957 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 16 of 17 

Kocher, D.C., and Eckerman, K.F., “Electron dose-rate conversion factors for external exposure of the 

skin from uniformly deposited activity on the body surface,” Health Physics, 53(2), pp. 135-

141, 1987 

Nickel Development Institute (NiDI), Health Studies of High Nickel Alloy Workers in the U.S., 1997; 

http://www.nidi.org/index.cfm/ci_id/224.htm

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

Workplace,” 1997; http://www.nipera.org/safe_use_guide_3.cfm

Oak Ridge Gaseous Diffusion Plant, Detail of Nickel Ingot Transactions, Supplemental Information 

Requested for FY 1961 Budget, (1961b) 

Paxton, H.C., Criticality Experiments and Data, from Nuclear Criticality Safety, TID26286, 

Proceedings of a Short Course held at the D.H. Lawrence Ranch near Taos, NM on May 7-11, 

1973, published by the U.S. AEC, 1974 

Scalsky, E.D., Technical Basis Document for the Savannah River Site To Be Used for EEOICPA 

Dose Reconstructions, ORAUT-TKBS-0003, Rev. 01, Effective Date August 21, 2003 

Schleien, B., Birky, B., & Slaback, L., Handbook of Health Physics and Radiological Health, Third 

Edition, Lippincott, Williams & Wilkins, Baltimore, January 1998 

Schneider, Alfred, Ph.D., Personal Communication, April 2003 

U.S. AEC, Facility Data Report, International Nickel Company Reduction Pilot Plant, Huntington, West 

Virginia, February 17, 1955 

U.S. AEC, Nuclear Safety Guide, Rev. 1, a U.S. Atomic Energy Commission Report prepared by the 

Goodyear Atomic Corporation, TID-7016, 1961 

U.S. DHHS, External Dose Reconstruction Implementation Guide, Rev. 1, Appendix B, National 

Institutes of Occupational Safety and Health, August 2002 

U.S. DOE, Draft Public Guideline to DOE Classification of Information, Office of Declassification, U.S. 

Department of Energy, Washington, D.C. 20545, June 27, 1994 

U.S. DOE Worker Advocacy Web Site Facility List, December 2002, 

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

U.S. DOE, DOE Standard: Health Physics Manual of Good Practices for Uranium Facilities, (DOESTD-1136-2000, Change Notice No. 3, December 2001). 

http://tis.eh.doe.gov/techstds/standard/std1136/STD11362000.pdf

U.S. EPA, External Exposure to Radionuclides in Air, Water, and Soil; Federal Guidance Report No. 

12, September 1993 

U.S. NRC, Guidelines for Decontamination of Facilities and Equipment Prior to Release for 

Unrestricted Use or Termination of Licenses for By-Product, Source, or Special Nuclear 

Material, U.S. Nuclear Regulatory Commission, Division of Fuel Cycle and Material Safety, 

Washington, D.C., 20555, November 1976 
Effective Date: 01/16/2004 Revision No. 01 Document No. ORAUT-TKBS-0004 Page 17 of 17 

U.S. NRC, Health Physics Surveys in Uranium Recovery Facilities, Revision 1, Regulatory Guide 

8.30, May 2002 (2002a) 

U.S. NRC, Re-evaluation of the Indoor Resuspension Factor for the Screening Analysis of the 

Building Occupancy Scenario for NRC’s License Termination Rule, Draft Report for Comment, 

NUREG-1720, June 2002 (2002b) 

U.S. NRC, Technical Basis for Calculating Radiation Doses for the Building Occupancy Scenario 

Using the Probabilistic RESRAD-BUILD 3.0 Code, NUREG/CR-6755, February 2002 (2002c) 

White, C.V., Memorandum Survey Report-International Nickel Corporation, Huntington, WV, January 

24, 1950 

Zhu, Joseph, PhD., Personal communication, May 2003