REPORT TO THE ADVISORY BOARD ON RADIATION AND 

WORKER HEALTH 

National Institute of Occupational Safety and Health 

Audit of Case #PIID* from the Huntington Pilot Plant 

Contract No. 200-2004-03805 Task Order No. 4 

SCA-TR-TASK4-CN005166 

Prepared by 

John Mauro, PhD CHP S. Cohen & Associates 6858 Old Dominion Road, Suite 301 

McLean, Virginia 22101 

February 2005 

NOTICE: This information is protected by Privacy Act 5 USC §552a; disclosure to any third party 

without the written consent of the individual to whom the information pertains is strictly prohibited. 
S. Cohen & Associates: 

Technical Support for the Advisory Board on Radiation 

& Worker Health Review of NIOSH Dose 

Reconstruction Program 

AUDIT OF CASE #PIID* FROM THE 

HUNTINGTON PILOT PLANT 

Task Manager: ___________________ Date: 02/04/05 

Hans Behling, PhD, MPH 

Project Manager: _________________ Date: 02/04/05 

John Mauro, PhD, CHP 

Document No. 

SCA-TR-TASK4-CNPIID* 

Effective Date: February 4, 2005 

Revision No. 1 

Page 2 of 21 

 Supersedes: 

Draft Rev. 0 
 

 

TABLE OF CONTENTS 

Summary ........................................................................................................ 4 

1.0 Introduction ...................................................................................... 6 

2.0 Summary of Doses ............................................................. 7 

3.0 Independent Duplication and Review of Selected NIOSH Derived 

Doses ...................... 10 

3.1Internal Dose from Inhalation ......................................... 10 

3.2 External Dose ................................................................. 12 

3.2.1 External Exposure to Uranium Deposited on Surfaces During 

Operations ...............................................................................13 

3.2.2 External Exposures to Uranium in Birdcages During Operations 

............ 14 

3.2.3 Annual X-Ray Exposures ...................................... 15 

3.2.4 External Dose After Cessation of AEC Operations ..... 15 

References .............................................................................. 17

 Appendix A: Radioactive Contamination in Nickel Feedstock 18 

 Appendix B: External Exposure to Air From Birdcage Arrangement of 

Uranium Cylinders: MicroShield Analysis ................................ 20 
 

SUMMARY 

This report presents the results of an independent audit of a dose reconstruction 

performed by the National Institute of Occupational Safety and Health (NIOSH) for an 

individual who worked at the Huntington Pilot Plant as PIID* from PIID*, to PIID*, 

PIID*, at which time the worker transferred to the Huntington Administration Building.1 

From 1951 to 1963, the Huntington Pilot Plant performed work under contract to the 

Atomic Energy Commission (AEC). The AEC work at Huntington involved the 

processing of scrap nickel to produce refined nickel powder for use in the manufacture of 

gaseous diffusion barriers for gaseous diffusion plants. The feedstock for producing the 

nickel powder was uranium-contaminated nickel that originated from the Oak Ridge 

Gaseous Diffusion Plant (ORGDP). The contaminated nickel scrap metal was offloaded 

by rail at Huntington, where the nickel was loaded into a furnace, and melt refined, and in 

the process, the uranium contamination was separated from the nickel through a special 

step in the process involving the application of carbon monoxide, referred to as the nickel 

carbonyl process. 

As a result of the worker’s employment at the plant, the worker likely experienced 

internal exposures due to the inhalation of airborne particles of enriched uranium, and 

external exposure from working in the vicinity of the nickel scrap metal and the 

separated uranium byproduct. In addition, the worker is believed to have had routine xrays as part of Huntington’s medical surveillance program. The worker was not 

provided with film badge or TLD dosimeters to measure external exposures, nor were 

bioassays performed to estimate internal exposures. As a result, exposures experienced 

by the worker were estimated using the exposure matrix provided in the site profile or 

Technical Basis Document (TBD) prepared by NIOSH for the Huntington plant 

(ORAUT-TKBS-0004, January 16, 2004). 

On PIID*, the employee was diagnosed with cancer of the bladder. Table 1 

summarizes the results of NIOSH’s reconstruction of the doses to the employee’s bladder 

for the purpose of deriving the probability of causation (POC) using IREP. Table 1 also 

presents the results of the audit. The results of the audit are expressed in terms of whether 

we found the exposures to have been derived in a scientifically valid and claimantfavorable manner. 

A review of file number PIID* (March 26, 2004) indicates some uncertainty as to the period of 

employment of the worker at the Huntington Pilot Plant. It appears that the employee may have worked at 

International Nickel Company from PIID* until PIID* which could have included work at the Huntington 

Pilot Plant from PIID* through PIID*. 
 

 

 

 

 

 

 

 

Table 1. Summary of Internal and External Exposure to the Bladder as estimated 

by NIOSH, Along with the Audit Results 

Exposure Scenario NIOSH Derived 

Annual Doses rem) 

Scientifically 

Valid? 

Claimant 

Favorable? 

Internal exposure from inhalation (alpha) 

during nickel handling and processing 

operations 

0.4 

No. An error 

was made in 

running IMBA 

Yes, with some 

qualifications 

Internal exposures from ingestion (alpha) 

during nickel handling and processing 

operations 

Internal exposure from inhalation (alpha) of 

residual resuspended particles following the 

conclusion of nickel handling and 

processing operations 

Not Addressed 

Not Addressed 

The TIB assumed that the 

worker stopped working when 

AEC operations ceased. This 

does not appear to be the case. 

External exposures during operations 

Ground surface contamination (chronic) 1.42E-2 No No 

Stored enriched uranium (chronic) 1.38E-1 No No 

Submersion in airborne plume (chronic) Not Addressed Yes NA 

Diagnostic x-rays (acute) 7.4E-2 No Yes 

Chronic external exposure to residual 

contamination following the conclusion of 

nickel handling and processing operations in 

PIID* 

Not Addressed 

The TIB assumed that the 

worker stopped working when 

AEC operations ceased. This 

does not appear to be the case. 
1.0 INTRODUCTION 

This report presents the results of an independent audit of a dose reconstruction 

performed by the National Institute of Occupational Safety and Health (NIOSH) for an 

atomic worker employee that worked at the Huntington Pilot Plant in West Virginia. 

This audit is one of several dose reconstruction audits being performed by S. Cohen & 

Associates (SC&A) on behalf of the Advisory Board on Radiation and Worker Health. 

Part one of this audit report presents a summary of our understanding of the doses 

derived by NIOSH, along with a brief description of the basic approach and assumptions 

employed by NIOSH to derive the doses. This material is extracted directly from the final 

dose reconstruction report published by NIOSH for this case, along with supporting 

documentation, including the Technical Basis Document: Basis for Development of a 

Exposure Matrix for Huntington Pilot Plant (ORAUT-TKBS-0004, January 16, 2004). 

This section of the report is provided as a baseline for the actual audit. 

Part two of the audit process consists of an attempt to independently duplicate selected 

doses derived by NIOSH and a discussion regarding the validity of the methods 

employed. The doses selected for duplication are based on the judgment of the auditors, 

as to the importance of the particular doses to the totality of the doses experienced by the 

atomic worker employee. The reason for this step in the audit process is to provide 

NIOSH and the Advisory Board with a level of assurance that the auditors understand 

how NIOSH went about deriving the doses provided in their dose reconstruction report. 

In the process of duplicating the NIOSH derived doses, we also provide a critical review 

of fundamental data, information, models, and assumptions used by NIOSH to perform 

the dose reconstruction. This part of the audit explores the degree to which the data are 

adequate to support the dose reconstruction, and whether the models and assumptions 

adopted by NIOSH to perform the dose reconstruction are scientifically sound and 

claimant favorable. Areas where the methods are found to meet these criteria, or are 

deemed to be inadequate with regard to these criteria, are identified and discussed. The 

report is not exhaustive in the review of these matters, but is limited to those areas of 

inquiry that are judged by the auditors to be potentially significant with respect to the 

dose reconstruction and the derivation of probability of causation (POC). 

Methods employed by NIOSH that are found to be either scientifically inappropriate or 

not necessarily claimant favorable are identified, but no attempt is made to correct these 

deficiencies and redo the dose calculations. It is assumed that the Advisory Board and 

NIOSH will have an opportunity to consider the results of this audit and determine 

whether a revision of the dose reconstruction is needed and, if so, how to go about 

making the necessary revisions. 
 

2.0 SUMMARY OF DOSES 

As a subcontractor to the AEC, International Nickel Company (INCO) received 

uncontaminated nickel sinter and uranium-contaminated nickel scrap at their facility in 

Huntington, West Virginia. The material was received at a railroad siding at the facility, 

where it was offloaded, weighed, and placed in buckets while it was still in the steel 

cartons, and then charged into the furnace with the sealed cartons. After the scrap was 

melted, it was transferred to the nickel carbonyl chamber, where carbon monoxide gas 

(CO) was added to the chamber, forming two separate streams, nickel carbonyl gas and 

enriched uranium. 

Because the contaminated nickel remained in its shipping carton until melting, NIOSH 

concluded that the potential for exposure of workers was low prior to melting. 

However, during and following melting, the potential for exposure was assumed to 

exist. 

The atomic worker employee worked at the Huntington Pilot Plant as a PIID* from 

PIID*, to PIID*, at which time the worker transferred to the Huntington Administration 

Building. From 1951 to 1963, the Huntington Pilot Plant performed work under contract 

to the Atomic Energy Commission (AEC). During this time period, the facility was also 

referred to as the Reduction Pilot Plant because of the special nickel processing 

operations the facility performed on behalf of the government, which involved the 

handling and processing of material contaminated with enriched uranium. 

In PIID*, following employment, the atomic worker employee was diagnosed with 

cancer of the bladder. Table 2 presents the results of NIOSH’s reconstruction of the 

doses to the atomic worker employee’s bladder for the purpose of deriving the POC 

using IREP. 

The notations used in Table 2 to present the doses include the year in which the dose was 

received by the organ of interest, the statistical distribution that was used, and the key 

parameters for the distribution. For example, for exposure period number 1 in Table 2 

(1951), a lognormal distribution was employed with a geometric mean of 5.12E-3 rem 

and a geometric standard deviation of 4.3 as the estimate of the internal alpha dose to the 

bladder due to chronic inhalation of airborne uranium particles. A discussion of various 

types of statistical distributions and other parameters used as input to NIOSH-IREP is 

provided in NIOSH (2002). The external and internal doses to the organ of interest were 

determined by NIOSH to be 2.5 and 9.81 rem, respectively. The POC was determined by 

NIOSH to 39.18% at the 99% confidence interval. On this basis, the claim was denied. 
Deletions made for 2 pages – please see hard copies #2 

3.0 INDEPENDENT DUPLICATION AND REVIEW OF SELECTED 

NIOSH-DERIVED DOSES 

This section describes the dose reconstruction methods employed by NIOSH, followed 

by a series of hand and computer calculations that attempt to duplicate selected doses 

derived by NIOSH. In so doing, we will have confirmed that we understand how 

NIOSH performed the dose reconstruction and whether or not the calculations are 

correct, given the models and assumptions employed by NIOSH. In the process of 

attempting to duplicate the doses, we also discuss and critically review the data, 

models, and assumptions employed by NIOSH to reconstruct the doses. 

3.1 INTERNAL DOSE FROM INHALATION 

The starting point that NIOSH used to derive internal doses from inhalation was 

historical air particulate data collected at Huntington, which revealed concentrations 

ranging from 0.01 to5 mg Ni/m3

, depending on location in the plant. These data are 

summarized in Table 4 of the technical basis document (TBD). Using these data, NIOSH 

assumed that the airborne dust loading throughout the plant was lognormally distributed 

with a median of 0.05 mg Ni/m3

 and a geometric standard deviation of 4.3. This 

distribution appears to be claimant friendly, given that the worker was a PIID* PIID*. 

Specifically, Table 4 of the TBD indicates that PIID* worked plant-wide and the 

measured airborne dust loadings throughout the plant were about 0.01 mg Ni/m3 . 

However, we are concerned with two aspects of this analysis. First, the data in Table 4 

includes both historical and more recent measurements. If only the historical data were 

considered applicable to historical dose reconstructions, the median dust loading would 

be 0.2 mg/m3 instead of 0.05 mg/m3. The second concern is that the dust loading in the 

breathing zone of a worker is often higher than the dust loading determined using general 

area air samplers. A discussion of this issue is provided in ICRP Publications 35 and 75. 

It is suggested that NIOSH investigate the degree to which these two issues could impact 

the reconstructed inhalation doses and the associated probability of causation (POC) for 

employees at the Huntington Pilot Plant. 

An additional concern that does not apply to this case, but may apply to other Huntington 

cases, is that the TBD provides direction that the median dust loading is appropriate for 

use in all cases that may arise from the Huntington Pilot Plant. We believe that a more 

claimant-favorable approach is to assume that, lacking information regarding a worker’s 

specific responsibilities at the plant, it should be assumed that claimants were members of 

the critical population groups at the facility, where the assumed average or median 

exposures were toward the high end of the distribution, perhaps the 90th

 or 95th percentile 

value. At the 90th percentile level, the dust load is estimated to be 5.57 mg Ni/m3 and the 

95th percentile level is 7.15 mg/m3 using the data in Table 4 of the TBD. It is suggested 
 

 

 

that NIOSH investigate the degree to which these two issues could impact the 

reconstructed inhalation doses and the associated POC for employees at the Huntington 

Pilot Plant. 

The level of uranium contamination on the nickel starting material, as reported in the 

TBD, is as follows: 

Uranium Isotope Ratio to Ni (pCi/mg) 

U-235 0.425 U-234 10.901 U-238 

0.102 

This level of contamination is based on the assumption that the residual 

contamination of uranium on the nickel was 39% enriched and had the following 

composition: 

Isotope Weight Percent Activity Percent 

U-234 0.3471 94.9106 

U-235 39 3.6914 

U-236 0.1794 0.5102 

U-238 60.4735 0.8877 

In the first draft of this audit report, SC&A expressed some concerns with the 

assumptions employed in the TBD that the nickel scrap received from the gaseous 

diffusion plants was contaminated only with uranium isotopes and their progeny. 

Various literature sources have documented that nickel scrap generated in the same 

general time frame was also contaminated with isotopes of Tc, Np and Pu. Appendix A 

presents a description of the basis for this concern. 

During the factual review meeting with NIOSH on January 12, 2005, NIOSH concurred 

that issues regarding recycled uranium need to be explicitly addressed in these types of 

analyses. However, NIOSH also pointed out that, in this particular case, NIOSH 

employed the assumption that all the uranium was 39% enriched in U-235, when, in fact, 

the enrichment level was much lower. This assumption, coupled with the fact that the 

starting point for the dose reconstruction was measured airborne dust loadings expressed 

in units of mass, the reconstructed doses were derived in a highly conservative manner, 

which more than accounted for any small incremental doses associated with trace levels 

of other radionuclides that may have been present in recycled uranium. We concur with 

NIOSH with regard to this matter. Nevertheless, it is suggested that in the future, NIOSH 

investigate whether isotopes of Tc, Np, and Pu may have also been present in the 

feedstock, and if so, evaluate the potential significance of this additional contamination 

on the reconstructed doses. NIOSH could then provide the rationale for not explicitly 

addressing this particular contribution to worker exposures. 

Accepting the uranium contamination levels at face value, the median airborne 

concentration of uranium dust is estimated in the TBD as 0.58 pCi/m3

. The annual dose 

to the bladder was derived by NIOSH using IMBA, and assuming a breathing rate of 1.2 

m 3/hr, 2,000 hours per year, (or 1,400 pCi/yr as the inhalation rate), Absorption Type M, 

and 5 micron AMAD. As indicated in Table 2, beginning in 1951, the exposure rate to 
the bladder continually increased as the body burden increased. The exposures leveled 

off at about 0.4 rem/yr and then, after 1962, the time when processing of the 

contaminated nickel ceased, the exposures gradually declined due to the clearance of the 

uranium from the body. 

An independent check on this calculation confirms the uranium inhalation rate of 1,400 

pCi/yr, with the following isotopic distribution; U-234 = 1308 pCi/yr, U-235 = 51 

pCi/yr, U-238 = 12.24 pCi/yr, and U-236 = 7.1 pCi/yr. Using this inhalation rate for 

enriched uranium, we were not able to duplicate the annual doses to the bladder of 

about 0.4 rem/yr as reported in the TBD. Our dose estimate using IMBA and another 

internal dosimetry code resulted in substantially lower doses (i.e., <1 mrem/yr versus 

400 mrem/yr to the bladder). Inspection of the IMBA file contained in the claimant file 

package indicates that an error was made in the input for the uranium inhalation rate. 

Instead of inputting the inhalation rate in pCi/day, as required by IMBA, the value that 

was input into the pCi/day field appears to be the total quantity of uranium inhaled 

by the worker for the full duration of employment. This error corresponds to about a 

4,000-fold overestimate in assigned internal dose. It would appear that, at least in this 

case, there was a breakdown in quality control that needs to be investigated. 

The TBD indicates that exposures terminated after AEC nickel processing operations 

ceased in 1963, and no additional intake occurred between 1963 and when the employee 

stopped work in 1972. Apparently, as indicated in File No. PIID* (March 26, 2004), the 

employee continued to work at the Huntington Pilot Plant until PIID*. The TBD 

assumed that he stopped working at the plant in PIID*. This does not appear to be the 

case. If he did in fact continue to work at Huntington from PIID* to PIID*, he may have 

experienced additional internal and external exposures associated with residual 

radioactivity at the facility because, according to the TBD, the facility was not 

decontaminated until 1978. As a result, the worker may have experienced additional 

exposures that are not accounted for in his dose reconstruction. During our meeting with 

NIOSH on January 12-13, 2005, NIOSH indicated that they are re-evaluating the 

potential that this individual continued employment at the Huntington Pilot Plant through 

PIID*. 

3.2 EXTERNAL DOSE 

Enriched uranium contains U-238, U-234, U-235, and U-236, along with their short-lived 

progeny, which grew in after the uranium was separated from the ore. The short-lived 

progeny include primarily Th-234 (T1/2 = 24 days) and Pa-234m (T1/2 = 1.17 minutes) 

from the U-238 series, and Th-231 from U-235. These radionuclides emit mostly weak 

gammas, x-rays, and beta particles with their associated bremsstrahlung radiation. In 

addition, Pa-234m decays to Pa-234, which emits a strong gamma with a branching 

fraction of 0.16%. 

The TBD evaluated several sources of potential exposure to external radiation, including 

(1) exposure to surfaces contaminated with deposited uranium dust, (2) exposure due to 
 

 

 

 

 

proximity to the carbonyl process residue, and (3) exposure to contaminated skin. For 

the employee, however, beta doses to skin and near-surface organs and tissue were 

ignored, because the betas (and associated bremsstrahlung) do not have sufficient 

penetrating power to expose the bladder. 

Table 3 presents the external exposures experienced by the organ of interest as 

reported in the TBD. 

Table 3. External Exposure to the Bladder 

External Exposure Scenario Annual Dose to Bladder (rem) 

E= 30-250 keV E>250 keV 

During Ni Processing (1951-1963) 

Ground surface contamination (chronic) 6.43E-3 7.76E-3 

 Birdcages (chronic) 8.09E-2 5.74E-2 

Diagnostic x-rays (acute) 7.4E-2

 Submersion Negligible Negligible 

Chronic residual exposures following decon (after 

PIID*) 6.43E-3 7.76E-3 

3.2.1 External Exposure to Uranium Deposited on Surfaces During Operations 

Two separate methods were used to estimate the external exposures due to uranium 

deposited on surfaces. The first method involved multiplying the airborne dust loading 

described above by the deposition velocity. The deposition velocity was assumed to be 

0.002 m/s, as cited in CRC 1986 for 5 micron particles with a density of 2 g/cm3 in air at 

25oC and 1 atm. This deposition rate was assumed to continue for 10 years without any 

removal due to building ventilation, cleanup, or natural weathering. Based on this 

buildup, the external dose rate was derived using the dose conversion factors provided in 

Federal Guidance Report No. 12. This approach resulted in an external dose rate of <1 

mrem/yr, which was low when compared to the actual external dose rates measured in a 

survey performed in 1981. Hence, the 1981 survey results were used to derive external 

exposures during and following operations. 

Though the deposition velocity approach was not used in this case for deriving external 

exposures from deposited activity, it is worth noting that actual measurements reveal that 

this approach is not claimant favorable, at least not in this case. The reason for bringing 

this up is, in other cases, NIOSH may have employed this method for estimating 

exposures from residual deposited activity, believing it to be claimant favorable. The 

implications of the survey results are that the deposition velocity approach may not be 

claimant favorable. We suspect that the reason for this is residual contamination on 

surfaces may have come about from spills and from the deposition of particles much 

larger than 5 micron, which may have had a much higher deposition velocity than those 

that apply to 5 micron particles. 

According to the 1981 survey report, gamma radiation readings ranged from 9-12 μR/hr, 
 

with a maximum of 23 μR/hr at 1 meter above the floor. The TBD assumed that the 

exposure rates were lognormally distributed, with a median of 12 μR/hr and a geometric 

standard deviation of 1.9 (which corresponds to a 95% confidence level of 35 μR/hr). 

To convert the exposure rate to the dose rate to the organ of interest, the Exposure (R) to 

Organ Dose (HT) photon dose conversion factors provided in Appendix B of NIOSH 

External Dose Reconstruction Implementation Guideline (NIOSH 2002a) were used. 

The exposure geometry was assumed to be isotropic, and the dose rate was divided 

equally between photons with E=30-250 keV and photons with E>250 keV. 

Since the external gamma readings were obtained in 1981, which was several years after 

AEC operations ceased and after decontamination of the facility took place in 1978, the 

use of the 1981 survey data for determining exposures to deposited radioactivity during 

operations is highly questionable. It is suggested that NIOSH revisit this issue using a 

strategy for estimating exposures to deposited radioactivity that more realistically 

represents the conditions at the time of operations. One approach may be to obtain data 

on the surface contamination of nickel dust at other nickel-melting operations. 

3.2.2 External Exposures to Uranium in Birdcages During Operations 

Following melting and the carbonyl separation process, the separated uranium was 

decanted into containers referred to as “birdcages.” Birdcages are made up of 5 by 5 

arrays of cylinders 4.25 inches in diameter and 9.75 inches high, each containing no 

more than 350 grams of U-235. These 5 by 5 arrays were stacked in no more than 2 

layers to avoid criticality. It was assumed that a worker was located 1 meter from the 5 

by 5 double-stacked array for 2,000 hours per year, resulting in an annual exposure rate 

of 0.13 R. The upper 95th percentile value of 0.76 R/y was calculated assuming 2,000 

hours per year of exposure at 1 foot away. To convert the exposure rate to the dose rate 

to the organ of interest, the Exposure (R) to Organ Dose (HT) photon dose conversion 

factors provided in Appendix B of NIOSH External Dose Reconstruction 

Implementation Guideline (NIOSH 2002a) were used. The exposure geometry was 

assumed to be anterior-posterior and the dose rate was divided equally between photons 

with E=30-250 keV and photons with E>250 keV. In order to confirm these values, we 

first estimated the quantity of uranium that may have been produced per year, and 

compared that quantity to the quantity of uranium housed in the birdcages. We then 

independently estimated the air exposure rate using MicroShield. Appendix B presents 

a description of our MicroShield analysis.2 

The birdcages are 5 by 5 double-stacked 

containers, each containing a maximum of 350 grams of U-235. Accordingly, the total 

inventory of enriched uranium in the 50 containers is 44.9 kg. According to the TBD, 

up to 2,587,000 pounds of nickel were shipped from ORGDP to INCO per year from 

1950 to 1961. Given that the starting material contained 11.428 pCi U/mg Ni (i.e., about 

5e-7 g U/mg Ni), the annual production rate of enriched uranium was about 582 kg per 

year, or enough to fill up about 13 arrays per year. 

Using MicroShield, we estimated the air exposure rate 1 meter and 1 foot from a 5 by 5 

double array to be 0.91 R/y and 2.38 R/y, respectively. These air exposure rates 

correspond to a bladder dose of 0.97 rem/y at 1 meter and 2.53 rem/yr, respectively. The 

dose to the bladder (at 1 meter away) reported in the TBS is 0.138 rem/y. The difference 
between our derived bladder doses and those reported in the TBD are large enough to 

raise some question regarding the accuracy of the values in the TBD. 2

 The MicroShield 

analysis described in Appendix B is considered to be a reasonable approximation of the exposure. A more 

precise analysis would involve the use of MCNP. 

3.2.3 Annual X-Ray Exposures 

NIOSH also assumed that the workers received annual chest x-ray examinations as a 

part of routine medical surveillance of workers. The air kerma skin dose, which was 

assumed to be 0.108 R for each diagnostic chest x-ray (Scalsky 2003), was multiplied 

by the Kerma (Ka) to Organ Dose (Ht) photon dose conversion factors for 30-250 

keV photons provided in Appendix B of DHHS 2002 to obtain the annual dose to 

the bladder of 0.074 rem. Unlike the other exposures, these exposures were assumed 

to be acute and the uncertainty in the x-ray exposures was assumed to be normally 

distributed, with a standard deviation of 30%. 

Given that the bladder is outside the primary beam of a chest x-ray, it is questionable 

whether the high bladder dose assumed in the TBD is scientifically appropriate. A more 

appropriate approach would have been to employ the method described in ORAUTOTIB-006, Technical Information Bulletin: Dose Reconstruction from Occupationally 

Related Diagnostic X-ray Procedures, December 29, 2003. Secondly, in accordance with 

the guidance and lacking information to the contrary, it would have been appropriate to 

assume that at least some examinations included photofluorography, because such 

examinations were a matter of standard practice prior to 1960. Table 4.0-1 of the 

guideline recommends a dose to the bladder of 0.025 rem for each photofluorography. 

Using this approach, as opposed to the approach used in the TBD, the medical exposures 

would be lower by about a factor of three. 

3.2.4 External Dose After Cessation of AEC Operations 

NIOSH assumed that the worker terminated employment at the facility in PIID*. 

However, as described above, it appears that employment at the Huntington Pilot Plant 

continued until PIID*. As a result, external exposures may have continued after PIID* 

due to residual radioactivity at the facility, until employment was terminated in PIID*. 

These exposures should be added to the worker’s reconstructed doses if the worker did, 

in fact, continue to work at the facility after PIID*. 

The TBD provides estimates of post-operational exposures due to residual 

radioactivity, which persisted until decontamination of the facility in November 1978. 

NIOSH was unable to find any pre-decontamination survey data. However, as 

described above, post-decontamination surveys revealed exposure rates of about 9 to 12 

μR/hr, with a maximum of 35μR/hr. As a means to estimate the pre-decontamination 
exposure rates, NIOSH assumed that the median exposure rate was 35 μR/hr and the 

upper 95% confidence level was 250 μR/hr, which corresponds to the radiation 

exposure limit of 500 mrem/yr recommended by the NCRP at that time. The annual 

dose rate to the bladder was then derived using the Exposure (R) to Organ Dose (rem) 

photon dose conversion factors from Appendix B of NIOSH 2002a. The exposure 

geometry was assumed to be isotropic, and the dose rate was divided equally between 

photons with E=30-250 keV and photons with E>250 keV. 

This approach to deriving the post-PIID* external exposure rates appears to be 

somewhat arbitrary. There is no reason to believe the highest air exposure rate 

observed post-decontamination bears any relationship to the exposure rate prior to 

decontamination. NIOSH may want to revisit this issue and develop a dose 

reconstruction methodology that has a better scientific foundation. 

REFERENCES 

NIOSH 2002, “NIOSH-Interactive Radioepidemiological Program (NIOSH-IREP) 

Technical Documentation,” Final Report. June 18, 2002. 

NIOSH 2002a, “External Dose Reconstruction Implementation Guideline,” OCAS-IG0001, Rev 1, August 2002. 

ORAUT-TKBS-0004, “Technical Basis Document: Basis for Development of an 

Exposure Matrix for Huntington Pilot Plant,” January 16, 2004. 

ORAUT-OTIB-0006, Kathren, R.L., Shockley, V.E., Thomas, E.M., Taulbee, T.D., 

Technical Information Bulletin: Dose Reconstruction from Occupationally Related 

Diagnostic X-ray Procedures, 2003. 

APPENDIX A: RADIOACTIVE CONTAMINATION IN NICKEL 

FEEDSTOCK 

Radioactivity in the gaseous diffusion plants consists of U-234, U-235, and U-238, which 

occur naturally in uranium ores,1

 their short-lived progeny, and Tc-99, U-236, and trace 

quantities of Np-237 and Pu-239. This contamination was introduced into the plants with 

reprocessed uranium fuel from plutonium production reactors at Hanford and Savannah 

River (NAS 1996). Eighty-eight percent of the reprocessed uranium was shipped to and 

introduced into the uranium enrichment facilities at the PGDP, and 12% initially went to 

the K-25 Site. Of the contamination in the reprocessed uranium feed shipped to K-25, 

only about 85% of the Tc-99 and about 25%t of the Np-237 and Pu-239 actually entered 

the cascade, because variable yields of different elements occurred when the uranium (as 

UO3) is converted to UF6, and when the UF6 is vaporized prior to being fed into the 

cascade (Smith 1984). (For example, during feedstock vaporization, PuF6 and NpF6 

would tend to remain in the storage cylinders.) Similarly, at K-25, only 25% of the Np­
237 and 1.5% of the Pu-239 entered the cascade. Smith (1984) estimated that 539 kg of 

Tc-99, 4.6 kg of Np-237, and 0.1 g of Pu-239 were fed into the Paducah cascades from 

1953 until September 11, 1975, when use of reactor return feeds was discontinued. Much 

of the contamination that had been deposited in the uranium enrichment facilities prior to 

1975, when this feed source was eliminated, was removed during the cascade 

improvement and upgrade programs (CIP/CUP) in the 1980s (NAS 1996). Some of this 

contamination is also present at the Portsmouth plant. 

1 

Natural uranium contains 0.712% U-235, 99.282% U-238, and trace amounts of U-234 (0.006%). 
 

 

 

 

 

 

 

 

 

REFERENCES 

NAS 1996. National Academy of Sciences/National Research Council, “Affordable 

Cleanup? Opportunities for cost reduction in the decontamination and decommissioning 

of the nation’s uranium enrichment facilities.” National Academy Press, Washington 

DC. 

Smith, R.F 1984. “Historical Impact of Reactor Tails on the Paducah Cascade,” KY/L1239, Paducah Gaseous Diffusion Plant, March 19, 1984. 

APPENDIX B: EXTERNAL EXPOSURE TO AIR FROM BIRDCAGE 

ARRANGEMENT OF URANIUM CYLINDERS: MICROSHIELD 

ANALYSIS 

In order to simulate the “birdcage” container (described in Figure 2 of the Huntington 

Pilot Plant TBD) for MicroShield, we assumed that the amount of uranium in the 50 

cylinders that make up a birdcage is evenly distributed within the volume of the entire 

container. The volume of the container is calculated to be 5,243,860 cm3. Using the 

maximum mass control of 350 g of U-235, along with a maximum of 39% by weight of 

U-235, the mass of enriched uranium in 50 cylinders is calculated to be 44.9 kg. The 

density of the material is therefore 8.55E-3 g/cm3. The activities of the uranium isotopes 

and their short-lived progeny are calculated based on the weight percentages presented in 

Table 2 of the TBD and the maximum mass control of 350 g of U-235. The short-lived 

progeny is assumed to be in equilibrium with its parents. Those calculations are presented 

Table B-1. 

Enriched Uranium in 50 cylinders (birdcage) 

Table B-1. 

Characteristics of 

the Enriched 

Uranium in the 

Birdcages Nuclide 

Weight % in 

EU 

Weight 

(g) pCi/g of EU Ci in birdcage 

U234 0.3471% 156 2.16E+07 9.71E-01 

U235 39% 17500 8.43E+05 3.78E-02 

Th231 3.78E-02 

U236 0.1794% 81 1.14E+05 5.21E-03 

U238 60.4735% 27136 2.02E+05 9.14E-03 

Th234 9.14E-03 

Pa234m 9.14E-03 

Pa234 1.46E-05 

Total EU 44872 2.28E+07 1.02 

The geometry dimensions used to simulate the external dose from the birdcage are 

presented in the figure below. Simulations were performed at both 1 foot and 1 meter 

away. 
 

 

 

 

 

 

Figure B-1. Schematic of the Exposure Geometry 

Huntington Pilot Plant Case #005166 20 S. Cohen & Associates 

The analysis was also performed with and without the short-lived progeny and with and 

without shielding. The shield is assumed to be the equivalent of a 55-gallon carbon steel 

drum, with a wall thickness of 1.5189 mm (ORNL No. 100-1A-2-0006) and a density of 

7.8 g/cm3 

(Handbook of Health Physics and Radiation, 3rd ed., Table 5.4). The calculated 

activities and the described geometries were used as input into MicroShield. Table B-2 

presents the results. 

Table B-2. MicroShield Results 

Distance from source 

Air Exposure with short-lived progeny (R/yr 

assuming 2000 hrs/yr exposure) 

No shield with shield 

1 meter 1.08 0.91 

1 foot 2.93 2.38 

Distance from source 

Air Exposure without short-lived progeny (R/yr 

assuming 2000 hrs/yr exposure) U-234, U-235, 

U-236, U-238 only 

No shield with shield 

1 meter 0.90 0.78 

1 foot 2.46 2.0