Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health

REVIEW OF THE NIOSH SITE PROFILE FOR 

ALLIED CHEMICAL CORPORATION PLANT, 

METROPOLIS, ILLINOIS 

Contract No. 200-2009-25888 

SCA-TR-SP2011-0012 

Prepared by 

Gregory Hofer and Robert Anigstein 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, VA 22182 

September 2011 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board on 

Radiation and Worker Health (ABRWH) to maintain all possible openness in its deliberations. However, 

the ABRWH and its contractor, SC&A, caution the reader that at the time of its release, this report is 

pre-decisional and has not been reviewed by the Board for factual accuracy or applicability within the 

requirements of 42CFR82. This implies that once reviewed by the ABRWH, the Board’s position may 

differ from the report’s conclusions. Thus, the reader should be cautioned that this report is for 

information only and that premature interpretations regarding its conclusions are unwarranted.
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Document No. 

 SCA-TR-SP2011-0012 

Effective Date: 

 Draft – September 29, 2011 

S. COHEN & ASSOCIATES: 

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program Revision No. 

0 (Draft) 

REVIEW OF THE NIOSH SITE PROFILE FOR 

ALLIED CHEMICAL CORPORATION PLANT, 

METROPOLIS, ILLINOIS 

Page 2 of 28 

Task Manager/Lead Author: 

 Date: 

Gregory G. Hofer, P.E. 

Supersedes: 

N/A 

Project Manager: 

 Date: 

John Mauro, PhD 

Peer Reviewer(s): 

 John Mauro, PhD, CHP 

Record of Revisions 

Revision 

Number 

Effective 

Date Description of Revision 

0 (Draft) 09/29/2011 Initial issue 

10/25/2011 Cleared for public release with minimal redactions 
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TABLE OF CONTENTS 

Abbreviations and Acronyms ..........................................................................................................4 

1.0 Introduction..........................................................................................................................5 

2.0 Site Description and Operational History............................................................................7 

2.1 .......................................................................................................................8 Safety

3.0 Internal Exposure...............................................................................................................12 

4.0 External Exposure..............................................................................................................14 

5.0 Review of Exposures during Residual Period ...................................................................17 

5.1 Internal Exposures .................................................................................................17 

5.2 External Exposures................................................................................................18 

6.0 Review of Document Adequacy ........................................................................................19 

6.1 Inadequate Document Research.............................................................................19 

7.0 References..........................................................................................................................20 

Appendix A: Analysis of Uranium Urinalysis Records for the Operational Period (1959–

1976) ..................................................................................................................................22 

Appendix B: Findings on ORAUT-OTIB-0024 ...........................................................................27 
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ABBREVIATIONS AND ACRONYMS 

ACCP Allied Chemical Corporation Plant 

AEC Atomic Energy Commission 

AWE Atomic Weapons Employer 

cm

2 square centimeter 

dpm disintegration per minute 

CATI Computer-Assisted Telephone Interview 

DCAS Division of Compensation Analysis and Support 

DCF dose conversion factor 

DOE U.S. Department of Energy 

FMPC Feed Materials Production Center 

HHS Health and Human Services 

IREP Interactive RadioEpidemiological Program 

keV kiloelectron volt 

MCNP Monte Carlo N-Particle Transport Code 

MeV megaelectron volt 

NOCTS NIOSH OCAS Claims Tracking System 

μg/l microgram per liter 

mrem/hr millirem per hour 

NIOSH National Institute for Occupational Safety and Health 

NRC Nuclear Regulatory Commission 

OCAS Office of Compensation Analysis and Support 

ORAUT Oak Ridge Associated Universities Team 

OSHA Occupational Safety and Health Administration 

pCi/ μg picocurie per microgram 

POC probability of causation 

rem/hr roentgen equivalent man per hour 

SEC Special Exposure Cohort 

SRDB Site Research Database 

TRU transuranics 

UF4 uranium tetrafluoride 

UF6 uranium hexafluoride 

U3 O8 triuranium oxide 
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t 

1.0 INTRODUCTION 

On February 1, 2006, the National Institute for Occupational Safety and Health (NIOSH) issued 

ORAUT-TKBS-0044 (ORAUT 2006a), which provides data and guidance for dose 

reconstruction of workers at the Allied Chemical Corporation Plant (ACCP), in Metropolis, 

Illinois. The facility was also known as the “General Chemical Division,” “Allied Chemical and 

Dye,” and the “Allied Signal Metropolis Plant.” The facility was later purchased by Honeywell. 

Subsequent to authorization to proceed with the review of that document, the NIOSH Division of 

Compensation Analysis and Support (DCAS) issued a revised site profile on October 1, 2007 

(ORAUT 2007). This report presents a review of the revised site profile. 

The U.S. Atomic Energy Commission (AEC) stated the following in one of its 1957 reports 

(AEC 1957, p. 9): 

On February 4, 1957, the Allied Chemical and Dye Corp. announced selection of 

Metropolis Ill., as the site of its plant to process 5,000 tons of U3O8 a year under 

contract with the Commission. The plant will produce uranium-hexafluoride feed 

for the nearby Paducah, Ky., gaseous diffusion plant. 

The period of AEC operations at ACCP in Metropolis, Illinois, from January 1, 1959, to 

December 31, 1976, involved AEC-contracted conversion of uranium ore concentrates (primarily 

U3O8) to uranium hexafluoride (UF6). Commercial uranium conversion operations continued 

after the completion of the AEC contract. However, it is assumed that there was some residual 

contamination at the facility after 1959 remaining from Atomic Weapons Employer (AWE) 

operations, contamination that was difficult to distinguish from contamination resulting from 

commercial activities. Hence, the site profile is concerned with radiation exposures experienced 

by workers during AWE operation, and also from residual radioactivity following the 

termination of AWE operations and the commencement of commercial operations, which 

continue to the present. 

As will be discussed in greater detail below, the uranium ore concentrates that were received at 

ACCP for processing contained variable amounts of residual Ra-226 and Th-230, which were 

not entirely removed from the original ore in the process of separating out the uranium from the 

ore. Because of the variable concentration of these radionuclides in different batches of 

concentrates (also referred to as yellowcake or U3O8) delivered to the facility, and the finding 

that the Ra-226 and Th-230 are separated and concentrated in various uranium chemical 

conversion and purifications steps associated with converting U3O8 to UF6, there were sources 

of residue that contained elevated concentrations of Ra-226 and Th-230, similar to raffinates tha

are produced at facilities that process ore, but at a much reduced level. Nevertheless, the 

presence of the reconcentrated Ra-226 and Th-230 in residue makes it difficult to reconstruct 

worker doses, because bioassay and air sampling programs at that time emphasized uranium and 

not Ra-226 and Th-230. Because of these challenges to dose reconstruction, NIOSH determined 

that internal dose from non-uranium radionuclides cannot be reconstructed with sufficient 

accuracy for ACCP employees who worked at the facility from 1959 through 1976, and a 

Special Exposure Cohort (SEC) for ACCP employees was designated by the Secretary of Health 

and Human Services (HHS) in February 2007 (Leavitt 2007). As such, the site profile and this 
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NOTICE:

review of the site profile is concerned primarily with reconstructing the external and internal 

radiation doses to workers during the AWE operations period from uranium compounds, and 

also the doses to workers during the residual period from all radionuclides. As will be discussed, 

the site profile provides methods for bounding the doses not only from uranium, but also from 

Ra-226 and Th-230 during the residual period. 

The method used to organize our review is in accordance with the way in which the site profile is 

organized, namely site description and operational history, internal exposure, external exposure, 

and residual radioactivity. The report concludes with a section on overall data adequacy that has 

applicability to all sections of the report.
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2.0 SITE DESCRIPTION AND OPERATIONAL HISTORY 

The site description and operational history is extremely thorough, well researched and 

documented, and well written. However, SC&A has a few observations that are worth 

mentioning. 

2.1 Source Term Description 

The fifth paragraph in Section 2.3 of the ACCP site profile, “Source Term,” states the following: 

The reported concentrate ratio was based on an average of yellowcake feed ratios 

from 33 mills (Perkins 1982, Table II-8). 

However, the statement made in the footnote to Table II-8, “Summary of Environmental Air 

Monitoring Results in 1979,” is as follows (Perkins 1982, p. 24): 

The licensee analyzed only the air samples at stations No. 6, 8, and 11 for 226Ra 

and 230Th. Analysis of air samples showed the ratio of 230Th to natural uranium 

to be much higher than the yellowcake feed average over 33 mills 

(

230Th/U-natural = 0.0052). 

Thus, the Perkins report does not state that the feed into the ACCP facility is from 33 mills, but 

rather that the effluent ratios are higher than the average of 33 mills. 

Observation 1: The statement that the concentrate ratio is based on the average of the 

output from 33 mills is unsubstantiated by the reference (Perkins 1982). 

Notwithstanding the above misstatement, the source term for the ACCP site profile is exclusively 

based on the data presented in the Perkins document. In this document, the data presented are 

from the 1970s and the “Report Objectives” (Perkins 1982, p. 2), which states the following: 

 

As part of the evaluation of effluents/wastes relating to the commercial nuclear 

fuel cycle, the objectives of this report were to determine the process discharge 

streams produced by the UF6 conversion facilities, to determine how these 

streams are presently treated, to collect any publically [sic] available emission 

and monitoring data, to identify the final fate of these wastes, and to assess the 

adequacy of present waste treatment/disposal techniques and available data. 

Thus, the data presented in the report have been processed and summarized for discussing 

environmental control of emissions, not for dose assessment/dose reconstruction. In addition, 

the data are just from the 1970s and may not accurately reflect the ACCP operations in the 1950s 

and 1960s. 

Finally, the Perkins document repeatedly refers back to documents in the Nuclear Regulatory 

Commission (NRC) archives; however, the ACCP site profile does not appear to have reviewed 

these documents from the NRC. 
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Finding 1: The source term is based primarily, if not exclusively, on the reference 

document (Perkins 1982), with little justification why these effluent data are acceptable for 

use in dose reconstruction, and why these data can be extrapolated to reflect operations 

during the 1950s and 1960s. 

2.1 SAFETY 

In the fourth paragraph in Section 2.4 of the ACCP site profile, “Safety,” the following statement 

is made: 

Protective clothing and shoes or shoe covers were provided to employees and 

visitors to ACCP to ensure that personnel were not inadvertently contaminated 

with uranium compounds as of 1982 (Perkins 1982), and probably from the 

earliest days of operations. 

SC&A can find no statement in the Perkins document about protective gear, and it is not clear 

that Perkins ever visited either site discussed in the report. In addition, the above statement 

contains the phrase, “...and probably from the earliest days of operations.” The ACCP site 

profile provides no justification for the assumption of protective clothing availability in the 

1950s through the 1970s. Therefore, this is a back-extrapolation of present-day policies and 

methodologies to those that were in place during the operational period. Thus, this extrapolation 

needs to be better supported. 

Finding 2: The statement with regards to protective clothing is unsubstantiated by the 

reference (Perkins 1982), and neither substantiation nor justification is given for the 

assumption of protective clothing availability in the 1950s through the 1970s. 

Section 2.4.1, “Visual Observation of Contamination;” Section 2.4.2, “Air Activity of 

Contamination;” Section 2.4.3, “Surface Contamination Measurements;” and Section 2.4.4, 

“Decontamination,” discuss procedures and data that are from 1985 or later. While the ACCP 

site profile does not specifically state that these references from 1985 and later are to be used for 

the operating period from the 1950s through the 1970s, it implies that that is the assumption to be 

made. 

Of special note is the third paragraph in Section 2.4.3, “Surface Contamination Measurements,” 

which contains the following statement (Wilkins 1992, p. 3 of 6 and p. 4 of 6): 

One summary of Health Physics data for the last half of 1991 (Wilkins 1992) 

noted that 1.7% of the 2,002 weekly smears exceeded the weekly limit of 

200 dpm/100 cm2

 and that the highest result was 923 dpm/100 cm2 in the 

lunchroom on a table. One of the 570 monthly smears and none of the 163 

quarterly smears exceeded the limit. 

This again gives the impression that the levels of surface contamination were at reasonable levels 

from the 1950s through the 1970s. However, SC&A reviewed the CATI reports for ACCP 
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claimants and the following are some of the conditions that the claimants said existed at ACCP 

(these statements have been paraphrased to assure claimant confidentiality): 

 Yellow dust covered equipment and personnel in sampling lab. 

 Plant was dusty, paint chips everywhere. 

 Majority of the time, you brushed off the ore or green salt and went about your business. 

 Protection wasn’t stressed in the 1950s and 1960s. 

 Contamination levels higher in 1950s and 1960s. 

 Worked around green powdery material (green salt), very dusty. 

 Very dusty all over facility except powerhouse. 

 Feed Building where U was processed was filthy. 

 In early days, plant was notified about NRC (AEC?) inspections, big cleanup effort 

before the visits. 

 No safety procedures in 1950s and 1960s. Safety measures improved in the 1970s. 

 Before OSHA (NRC?) safety wasn’t always a concern. 

Finally, the ACCP site profile made no attempt to access documents from the NRC archives that 

may provide additional information on contamination conditions at the facility. 

Finding 3: Statements of contamination/decontamination are based on procedures and 

surveys that are from 1985 or later. Neither substantiation nor justification is given for the 

assumption that these procedures and surveys are applicable to the facility in the 1950s 

through the 1970s. No references were presented for the contamination condition of the 

ACCP during the 1950s through the 1970s. ACCP claimant statements were not included 

in the site profile. Therefore, the statements in the site profile on contamination/ 

decontamination are questionable. 

In Section 2.5, “Incidents,” the ACCP site profile makes the following statement: 

The claims include some information about incidents, but details are few and 

there are no dates or references to particular incident reports. Occasional 

leaking valves or inadequate packing resulted in material releases. 

However, SC&A reviewed the CATI reports for ACCP claimants, and the following are some of 

the incidents that the claimants said happened at ACCP (these statements have been paraphrased 

to assure claimant confidentiality): 

 1960s: bought a cornfield and an orchard due to contamination. 

 New process (learning curve) in the 1950s and 1960s equipment malfunctions caused 

frequent releases. 

 Spills/releases occurred frequently in 1950s and 1960s (sometimes up to once a month). 
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 Spills/releases would cause evacuation of buildings. 

 Spill shut down traffic in a 5-mile radius. 

 Spill contaminated car windshields. Allied replaced the windshields of several cars. 

 UF6 releases were quite often. 

 UF6 releases minor, 10–12 per year; major, once every 2–3 years. 

 UF6 release like a thick fog, workers trapped in basement. 

 UF4 (green salt) releases. 

 Drums stacked outdoors. Some drums would burst open and material would leak into the 

ground. 

 At home—saw yellow cloud coming from the plant. 

 Loud boom, became covered in green salt (UF4), washed off UF4 from body and washed 

out UF4 from mouth. 

 U ore dust so thick you could barely see the lights. 

 If release was bad enough (i.e., major), the siren was turned on to warn Metropolis. 

 When building was evacuated, the NRC would come to investigate. 

 NRC would not come for a “puff” release (assumed to mean when detaching a filled UF6

cylinder – some UF6 remained in the line, and although vacuum hoses were used, some 

UF6 would escape when the cylinder is disconnected); they would only come for an 

actual release. 

 Nearly all areas had leaks (UF6, UF4) except powerhouse. 

 UF6, UF4, U ore, yellowcake spills/releases. 

 UF6 fogs so thick, could barely see the UF6 building. 

The site profile needs to explore these incidents and conditions and determine if the internal and 

external dosimetry data can be used to reconstruct exposures associated with these incidents and 

conditions. 

As pointed out in Section 3.2 of the site profile, “Notations on Bioassay Records,” some bioassay 

samples are labeled “S,” which the site profile assumes means special. According to the CATI 

reports, additional urine samples were taken after exposure incidents and if a routine bioassay 

was high. It is likely that such urine samples were labeled “S,” and as a result, information about 

incidents might exist in the bioassay records and should be investigated. 

In addition, the site profile does not access files at the state/local government level. This source 

of information might be useful, because if releases could be seen from outside the facility, the 

plant siren was activated, and/or roads had to be closed, the state and local governments may 

have some records of the incidents. Also, such incidents may be documented in the local 

newspapers. 
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Finally, the ACCP site profile made no attempt to access documents from the NRC archives that 

may provide additional information on events/incidents at the facility. 

Finding 4: Based on the CATI reports and the “S” bioassay samples, documentation 

searches of the NRC archive, state/local government records, and local newspaper records 

need to be undertaken to assess the incidents that have taken place at the ACCP facility. 

With respect to Section 2.6, “Physical Examinations—X-Rays,” SC&A concurs with the site 

profile that the assumption that x-ray examinations happened annually is claimant favorable. 

In Section 2.7 of the site profile, “Summary Operational Period Assumptions, Workdays, Work 

Hours, Work Categories,” it was decided that determining workdays, work hours, and work 

categories was unnecessary because of the availability of bioassay and film badge data. While 

SC&A finds this acceptable for photon and U uptake, it is not apparent that such an approach is 

acceptable for neutron exposures, which need to be estimated from workdays, work hours, and 

work categories. 

SC&A reviewed the CATI reports for ACCP claimants and the following are some of the 

information that claimants provided with respect to work hours at ACCP (these statements have 

been paraphrased to assure claimant confidentiality): 

 40+ hours per week were routine, with some workers working up to 80 hours per week. 

 Since operation was 24/7, during strikes, non-union personnel would live on-site and run 

the plant. 

Work hours will be further discussed in Section 5.0 of this review with respect to neutron 

exposure. 
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3.0 INTERNAL EXPOSURE 

In Section 3.0, “Estimating Internal Exposure,” the ACCP site profile reiterates the information 

from Section 1.0 with regards to the SEC and thus states that only uranium will be evaluated 

with respect to internal exposure. SC&A concurs with this assessment. 

Section 3.1, “Uranium,” discusses isotopic fractions and lung absorption types. SC&A concurs 

with this discussion. 

The site profile also explains that virtually all workers during the covered period were under a 

bioassay program that included periodic urine analyses of all workers using fluorometric 

analyses, including some chest count data that can be used to verify the results of the urine 

analyses. Because of the comprehensive nature of the bioassay program, the site profile states 

that it was not necessary to develop a coworker model. In order to confirm this statement, we 

tried to access the data in the Site Research Database (SRDB); however, we were unable to find 

an electronic database for Allied Chemical. As a result, we reviewed the statements made by 

NIOSH by performing a semi-random sample of claimant hardcopy records (see 

Appendix A). Overall, SC&A examined the records of 62 claims (60 completely random, and 2 

extra to assure that the major job types were covered). The results are quite favorable, as follows: 

 

 Only one worker had no uranium analyses; however, this worker was only on site for 

6 months at the very beginning of the operational period. 

 The sampling schedule appears to be monthly for most years and workers; however, if 

there was a generally high result (arbitrarily chosen by SC&A as greater than 40 μg/l), 

the sampling schedule was shortened to around 2 weeks. If the sample was exceptionally 

high, re-sampling often occurred within a few days. 

 Over 95% of the worker-years examined had uranium analyses; over 77% of the 

claimants examined had uranium monitoring in every single year they were employed 

during the operational period. 

 

SC&A concludes that the uranium bioassay monitoring was comprehensive and had a focus on 

workers with higher exposure potential. 

The site profile also states that only natural and perhaps depleted recycled uranium was received 

and processed at the facility during the covered period. As a result, the specific activity of 

0.683 pCi/μg of uranium was employed. We concur with this value, which applies to natural 

uranium. However, the site profile mentions the possibility that a small amount of recycled 

uranium might have been shipped to ACCP from Rocky Flats. Given the potential for high 

levels of contamination with transuranics (TRU), with very high dose conversion factors (DCFs), 

at least a semi-quantitative assessment of the possible impact of such contamination should have 

been performed. 

Finding 5. The site profile should include a discussion of the impact of any recycled 

uranium that may have been processed at the facility. 
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The site profile goes on to state that, since the facility worked with U3O8, UF4, and UF6, it is 

possible the chemical form of the uranium intake could have been Type S, M, or S, and, as a 

result, the dose reconstructor is to assume the form that is limiting for the organ of concern. We 

concur with this claimant-favorable strategy.
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4.0 EXTERNAL EXPOSURE 

Section 4.0 of the ACCP site profile, “Estimating External Exposure,” discusses penetrating 

exposure (photon energies of 30 keV or greater), non-penetrating exposure (photon energies of 

less than 30 keV or electrons), and neutron exposure (neutron energies of 0.1 to 2 MeV). 

The site profile explains that, during the AWE period, all workers were badged for photon 

exposures and non-penetrating exposures using film badge dosimeters. SC&A reviewed the film 

badge data and concurs that most if not all workers were badged, and that the assumed energy 

ranges of exposure and the badge limits of detection are reasonable. However, SC&A has some 

concerns with the neutron exposure discussion. 

With respect to neutron exposures, the site profile explains that the doses are based on 

calculations of the alpha,n reaction associated with UF4 and UF6, which, in principle, is a 

scientifically valid strategy for estimating potential neutron exposures. In the last paragraph of 

Section 4.0, the site profile explains and makes the following statement about the derived 

neutron doses: 

A reasonable neutron dose estimation which is favorable to the claimant is based 

on a 500-pound drum that contains 75% by mass of natural uranium. It is 

assumed that the alpha particles collide with a fluorine target. Using the method 

described in [ORAUT 2005a], the neutron dose is 3.28E-04 rem/hr, calculated at 

1 foot from a drum. A worker who is 1 foot from such a drum for 2,000 hr/yr 

would be exposed to a dose rate of 6.56E-01 rem/yr. This document considers 

this estimated neutron dose rate to be an upper bound (maximum estimate) of the 

ACCP neutron dose rate. The mode neutron dose rate is estimated as 

5.47E-03 rem/yr, which is the result of applying an occupancy of 3 hours/week 

over 50 weeks to the estimated dose rate at 3 feet from a drum [5]. The minimum 

neutron dose rate is estimated as 0 rem/yr. These estimated minimum, mode and 

maximum neutron dose rates are to be used only for reconstructing doses during 

the listed operational period, and applied in IREP as a triangular distribution. 

During the residual period, the weapons-related source materials (uranium, 

fluorine targets) needed to produce neutrons are not presumed to be present in 

amounts significant enough to warrant the estimation of neutrons. 

This exposure rationale is based on ORAUT-OTIB-0024 (ORAUT 2005a) (see Appendix B for a 

summary of SC&A’s findings associated with the review of this procedure) and on callout [5], 

which refers to the following statement that is provided in Section 6.0 of the site profile, 

“Attributions and Annotations:” 

Olsen, Bernard M. MJW Corporation. Senior Health Physicist. August 2007. 

Surveys of UF6 storage cylinders more enriched than those at Allied Chemical, 

and studies of work practices in moving those cylinders, have been made at the 

Portsmouth Gaseous Diffusion Plant, as described in ORAUT-2006d

[ORAUT-TKBS-0015-6 (ORAUT 2006b)]. It was estimated in ORAUT-2006d 

that a worker would take on average 3 hours/week, for 50 weeks in a year, to 
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 the 

ificant 

transport these cylinders, and that the average neutron dose rate to the worker 

would be that at approximately 1 meter from the cylinder surface. The upper 

bound neutron dose rate developed in Section 4.0 is 3.28E-04 rem/hr at 1 foot 

from the cylinder. At 3 feet from the cylinder, the neutron dose rate would be 

estimated at 3.28E-04 rem/hr [(1 ft)2

/(3 ft)2], or 3.64E-05 rem/hr. The average 

neutron dose rate, applying the 150 hr/year occupancy methodology of ORAUT2006d (3 hr/wk for 50 wk/yr), is 5.47E-03 rem/yr. 

The neutron dose rates cited above are not scientifically correct, for the following reasons:. 

 In an earlier report, SC&A (2007) calculated the dose rate at a distance of 1 m from a 

55-gal drum filled with UF4 to be 0.015 mrem/h. We wish to compare this calculated 

dose rate to the dose rate of 3.64E-05 rem/h (0.036 mrem/h) at 3 ft, listed in the site 

profile. If we were to employ the inverse-square law, the methodology in the site profile, 

the value of 0.036 mrem/h at 3 ft extrapolates to 0.030 mrem/h at 1 m. Hence, the value 

reported in the site profile is twice the value calculated by SC&A (2007) (i.e., 

0.030 mrem/h vs. 0.015 mrem/h at 1 meter). We note that SC&A (2007) employed the 

SOURCES-4C code developed by Los Alamos National Laboratory to estimate the 

neutron spectrum, and the MCNP code to calculate the dose rate based on that spectrum, 

both state-of-the-art physics-based computer codes. Furthermore, we note that the 

inverse-square law applies only to point sources in a vacuum; the dose rate near a large 

object, such as the drum, decreases more slowly with distance. Thus, the discrepancy 

between the SC&A calculation and the site profile is even greater. The SC&A (2007) 

calculation was based on a drum containing ~1,500 lb UF4, rather than 500 lbs that was 

cited by NIOSH, indicating an even greater overestimate in the site profile. 

 Using the inverse-square to extrapolate the dose rate at 1 ft to 3 ft is incorrect for a drum 

at such close distances, and would lead to a gross underestimate at the greater distance, 

assuming that the dose rate at 1 ft was correct. 

The site profile makes the assumption that the work practices at Portsmouth (ORAUT 2006b) are 

equivalent to the work practices at Allied Chemical (ORAUT 2007). There are reasons to 

believe that the practices differed, as follows. Essentially, Portsmouth receives UF6, enriches 

UF6, ships enriched UF6, and ships/stores depleted UF6. Portsmouth does not do any chemical 

processing of UF6, while Allied Chemical was responsible for turning U3O8 into UF6. 

Therefore, while it is probably acceptable to just consider the neutron doses emanating from

UF6 cylinders for Portsmouth workers, it may not be appropriately applied to Allied Chemical, 

where UF4 (green salt) was first manufactured in vessels, then the green salt was converted to 

UF6 in other vessels, and then transferred to the UF6 cylinders. That said, the neutron flux from 

UF6 is ~2.5% higher than from UF4, per unit mass of the compound, which is not a sign

difference. 

The site profile assumes 3 hours per week neutron exposure, which is inadequate for Allied 

Chemical workers. A review of the CATI documents shows that Allied Chemical workers 

routinely worked 40+ hours per week in the manufacture of UF4 and UF6, with some working as 

much as 80 hours per week. Many ACCP workers could have been exposed to neutrons for their 
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entire shift, depending on their work assignments and/or locations. Also, in the event of a strike 

(and there were several at the facility), salaried workers stayed on site at the facility 24/7 in order 

to maintain UF6 production. Thus, during a strike, these workers received a larger neutron dose. 

Finding 6: The assumptions for neutron exposure do not appear to be claimant favorable. 

The assumptions are based on ORAUT-OTIB-0024 (ORAUT 2005a), which NIOSH has 

determined needs to be revised to reflect a work-hour exposure rate that, at a minimum, is 

an order of magnitude too low, and workplace conditions that do not match the conditions/ 

configurations/processes of the ACCP workplace. Since this plant produces UF4 and UF6, 

and these materials are in significant quantities throughout the facility, the validity of any 

dose reconstruction that results in a POC that is under 50% is questionable until an 

acceptable neutron dose can be incorporated into the ACCP site profile and into any 

affected dose reconstruction. 

SC&A concurs with the discussion in Section 4.1, “Occupationally Required Medical X-Ray.” 

SC&A’s review of the CATI reports provides sufficient evidence that it is reasonable to assume 

that workers received annual required medical x-ray exams. SC&A further agrees that the use of 

ORAUT-TIB-0006 (ORAUT 2005b) is claimant favorable. 

In Section 4.2, “Miscellaneous Information about External Dose,” the site profile discusses the 

possibility that those workers who handled fluorination bed ash could receive significantly 

higher doses on their extremities than measured by the dosimeters carried on the trunk of their 

bodies. However, the site profile provides no guidance on how the dose reconstructor is to 

calculate this exposure. Without guidance, inconsistent methodologies could be used that could 

result in dose reconstructions that are not claimant favorable. 

Finding 7: No guidance is provided for the dose reconstructor to calculate extremity 

exposure for those workers who handled fluorination bed ash. Without guidance, 

inconsistent methodologies could be used that could result in dose reconstructions that are 

not claimant favorable. 

With the exception of the neutron exposure, which has been previously discussed, SC&A 

concurs with the summary presented in Section 4.3 of the site profile, “Occupational External 

Dose Reconstruction Assumptions and Summary.”
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5.0 REVIEW OF EXPOSURES DURING RESIDUAL PERIOD 

The residual period extended from January 1, 1997, to the present, a time period when workers 

experienced exposures associated with commercial uranium processing and exposures associated 

with residual radioactivity from AWE operations. It is noteworthy that Section 5 of the site 

profile states that exposures to residual uranium, Ra-226, and Th-230 from AWE operations can 

be reconstructed during the residual period based on (1) personnel monitoring during the residual 

period, (2) knowledge of the radionuclide activity ratios in relation to uranium based on reported 

activities in ACCP concentrates, wastes, air, and effluents, as provided in Table 2-1 of the site 

profile and in Attachment A, “Application Of Residual Dose Factors,” of the site profile, and 

(3) assumptions regarding the rate at which AWE residual contamination declined during the 

residual period. 

5.1 INTERNAL EXPOSURES 

Since uranium bioassay data are available for all workers in 1977, it appears that the site profile 

assumes that the intake of uranium for 1977, as derived using the bioassay data for individual 

workers in 1977, is all due to the inhalation of resuspended uranium that was residual from the 

AWE period. The intake rate of all other radionuclides is based on the ratios in Table 1. 

Table 1. Source Term Activity Ratios Relative to Uranium 

U-Natural 1 

Th-232 0.076 

Th-230 0.167 

Th-228 0.076 

Ra-226 0.00846 

Pb-210 0.00846 

Po-210 0.00846 

Note: Based on the largest reported ratio in relation to uranium in Table 2-1 

of the site profile, excerpted directly from Table 5-1 of the site profile 

Table 5-1 of the site profile omits Ra-228, the first daughter of Th-232. Since Th-228 is 

assumed to be in equilibrium with Th-232, Ra-228 should be included as well. Otherwise, 

SC&A has determined that this basic strategy for deriving the intake rates of radionuclides for 

1977, the first year of the residual radioactivity period, is scientifically sound and claimant 

favorable. We recognize that one could reasonably question whether the intake rate of nonuranium radionuclides can be derived in this manner. However, it would seem that the ratios of 

these radionuclides as observed in various waste streams is a reasonable approximation of what 

one would expect in the residue during the residual period. SC&A acknowledges that this is 

very much a judgment call, but since the site profile elected to use the highest of these observed 

ratios, we find this strategy appropriately claimant favorable. Notwithstanding this conclusion, 

we refer the reader to Section 6 of this report, where we discuss the fact that extensive data are 

available from the Atomic Energy Commission (AEC) docket that might provide additional 

useful information on the mix of radionuclides and should be consulted. 
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SC&A does have concerns with respect to the rate at which these intake rates are assumed to 

decline with time during the residual period. Specifically, as indicated in Table 5-2 of the site 

profile, in 1978, the intake rates of these radionuclides are assumed to decline by a factor of 0.03 

compared to 1977, and for all subsequent years, the intake rates for these radionuclides are 

0.0007 the 1977 intake rate. The site profile cites a number of papers that describe the rate at 

which resuspension factors have been observed to decline and concludes that, for indoor 

environments, the guidance adopted in ORAUT-OTIB-0004 (ORAUT 2006c)is applicable here; 

specifically, it is assumed that the rate at which the resuspension factor declines is 1% per day. 

This issue, i.e., the rate at which exposures associated with the residual period decline, has been 

discussed extensively by the Subcommittee on Procedures Review as part of SC&A’s review of 

ORAUT-OTIB-0070 (ORAUT 2008) and the Norton site profile. Agreement has been achieved, 

in principle, with NIOSH, the Subcommittee on Procedures Review, and SC&A that the 1% per 

day assumption needs to be revised, since there is empirical evidence that the rate of decline of 

residual radioactivity is much slower and may get progressively slower with time. 

Finding 8: NIOSH should revisit the rate of decline factors provided in Table 5-1 of the 

site profile in light of the new information under development by NIOSH regarding the 1% 

per day issue. 

5.2 EXTERNAL EXPOSURES 

The site profile explains that external exposures during the residual period are based on the 

results of external dosimetry performed on all workers during that time period. This is an 

extremely claimant-favorable (to the extent that it might not be plausible) assumption, because it 

includes exposures from residual contamination plus exposures from ongoing commercial 

operations. A more plausible approach would be to assume that the external exposures 

associated with 1976 dosimetry decline at a realistic rate for normal depletion of deposited 

radioactivity. One could argue that the current approach is certainly claimant favorable, but it 

also could be argued that such claimant favorability should also be granted for the residual 

period at other sites. 

Finding 9: The external exposures assigned to workers during the residual period appear 

to be implausibly high. 
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6.0 REVIEW OF DOCUMENT ADEQUACY 

In reviewing the ACCP site profile, SC&A has determined that the document has several general 

deficiencies, which are discussed in this section. 

6.1 INADEQUATE DOCUMENT RESEARCH 

From its inception, ACCP was a commercial facility licensed by the AEC (later the NRC). Thus, 

ACCP was required to apply for a materials license. Submittals and correspondence to/from 

ACCP to the AEC/NRC are part of the public record; this record resides in the NRC archives and 

is accessible. The documents for any licensed facility are filed by Docket Number, which for 

ACCP is 40-3392. Even though this docket number appeared in several references of the Perkins 

report (Perkins 1982, p. 30); in NRC License SUB-526, Amendment 10 (NRC 1993); and in 

NRC License SUB-526, Amendment 15 (NRC 2003), there is no apparent attempt to obtain 

additional information from the NRC. 

In addition, based on many incidents outlined in the ACCP CATI reports, not only should the 

NRC archives have been accessed, but also state and local government files and the local 

newspaper archives for additional information. 

Finding 10: The document research for the ACCP site profile is inadequate and needs to 

be reinstituted to provide additional documentation and data for the site profile, including 

data available from the AEC docket for this facility. 

Observation 2: In the references for the ACCP site profile (ORAUT 2007), the reference 

“NRC 1993” erroneously labeled as “Amendment 15,” is actually Amendment No. 10. 

Finding 11: The site profile needs to be revised to provide a more thorough exploration of 

incidents cited in the CATI reports. 

The figures provided are of poor quality. Figure 2-1 from the Perkins report (Perkins 1982, p. 3) 

is partially illegible and a half page in the site profile (it’s a full page in the Perkins report). In 

addition, the copy in the SRDB is of poor quality. SC&A went to the Department of Energy 

(DOE) Information Bridge (http://www.osti.gov/bridge/), typed in the Los Alamos document 

number LA-9397-MS (Perkins 1982, Cover Page) and a much more legible copy of the Perkins 

report was available for download. Finally, Perkins most likely obtained this plot plan from one 

of the NRC documents referenced in the report. Thus, an even better quality figure is available. 

When the site profile is revised, a much cleaner plot plan should be provided and it should take 

an entire page. 

Figure 2-2 is from the NRC license (NRC 2003, p. 9-18), and since this is a simple diagram, it 

could have been redrawn for better clarity; and again, in the revised site profile, it should take an 

entire page. 

Observation 3: Any revision of the site profile should contain more legible figures.
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7.0 REFERENCES 

AEC 1957. Major Activities in the Atomic Energy Programs, January-June 1957 – Production 

of Special Nuclear Materials, U.S. Atomic Energy Commission, Washington, DC. July 1957. 

[SRDB Ref ID: 9511] 

ICRP 1997. ICRP Publication 74: Conversion Coefficients for use in Radiological Protection 

against External Radiation Annals of the ICRP Volume 26/3. International Commission on 

Radiological Protection. March 1997. 

Leavitt, M.O., 2007. HHS Designation of Additional Members of the Special Exposure Cohort 

under The Energy Employees Occupational Illness Compensation Program Act of 2000, 

Designating a Class of Employees from Allied Chemical Corporation Plant, Metropolis, Illinois, 

Allied Chemical Corp. February 1, 2007. [SRDB Ref ID: 54600] 

NCRP 1983. Operational Radiation Safety – Training. National Council on Radiation 

Protection and Measurements (NCRP). Bethesda, Maryland: NCRP Publications; NCRP 

Report 71; 1983. 

NRC 1993. Materials License, License Number SUB-526, Amendment No. 10, with excerpts 

from Allied Chemical Application for Renewal of Source Materials License SUB-526, Docket 

40-3392, U.S. Nuclear Regulatory Commission, Washington, DC. August 26, 1993. [SRDB 

Ref ID: 12377] 

NRC 2003. Materials License, License Number SUB-526, Amendment No. 15, with excerpts 

from Honeywell Application for Renewal of Source Materials License SUB-526, Docket 40-3392, 

U.S. Nuclear Regulatory Commission, Washington, DC. January 30, 2003. [SRDB Ref ID: 

12380] 

ORAUT 2005a. Estimation of Neutron Dose Rates from Alpha-Neutron Reactions in Uranium 

and Thorium Compounds, ORAUT-OTIB-0024, Rev. 00, Oak Ridge Associated Universities 

Team (ORAUT), Cincinnati, Ohio. April 7, 2005. [SRDB Ref ID: 19445] 

ORAUT 2005b. Dose Reconstruction from Occupationally Related Diagnostic X-Ray 

Procedures, ORAUT-OTIB-0006, Rev. 03 PC-1, Oak Ridge Associated Universities Team 

(ORAUT), Cincinnati, Ohio. December 21, 2005. [SRDB Ref ID: 20220] 

ORAUT 2006a. Site Profile for Allied Chemical Corporation Plant, ORAUT-TKBS-0044, 

Rev. 00, Oak Ridge Associated Universities Team (ORAUT), Cincinnati, Ohio. February 1, 

2006. [SRDB Ref ID: 22273] 

ORAUT 2006b. Portsmouth Gaseous Diffusion Plant - Occupational External Dose, 

ORAUT-TKBS-0015-6, Rev. 01, Oak Ridge Associated Universities Team (ORAUT), 

Cincinnati, Ohio. November 7, 2006. [SRDB Ref ID: 30055] 
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NOTICE:

ORAUT 2006c. Estimating the Maximum Plausible Dose to Workers at Atomic Weapons 

Employer Facilities, ORAUT-OTIB-0004, Rev. 03 PC-2, Oak Ridge Associated Universities 

Team (ORAUT), Cincinnati, Ohio. December 6, 2006. 

ORAUT 2007. Site Profile for Allied Chemical Corporation Plant, ORAUT-TKBS-0044, 

Rev. 01, Oak Ridge Associated Universities Team (ORAUT), Cincinnati, Ohio. October 1, 2007. 

[SRDB Ref ID: 35196] 

ORAUT 2008. Dose Reconstruction during Residual Radioactivity Periods at Atomic Weapons 

Employer Facilities, ORAUT-OTIB-0070, Rev. 00, Oak Ridge Associated Universities Team 

(ORAUT), Cincinnati, Ohio. March 10, 2008. 

Perkins, B.L., 1982. Evaluation of Environmental Control Technologies for Commercial 

Nuclear Fuel Conversion (UF6) Facilities, LA-9397-MS, University of California, Los Alamos 

National Laboratory, Los Alamos, New Mexico. October, 1982. [SRDB Ref ID: 12379] 

SC&A 2007. “Review of the Feed Materials Production Center (FMPC) Special Exposure 

Cohort (SEC) Petition-00046 and the NIOSH SEC Petition Evaluation Report,” SCA-SECTASK5-0056, SC&A, Inc., Vienna, Virginia, and Saliant, Inc., Jefferson, Maryland. May 31, 

2007. 

Wilkins, C.W., 1992. “Semiannual Health Physics Data Review (July 1991–December 1991).” 

Allied Chemical Corp. March 30, 1992. [SRDB Ref ID: 12374, p. 58]
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APPENDIX A: ANALYSIS OF URANIUM URINALYSIS RECORDS FOR 

THE OPERATIONAL PERIOD (1959–1976) 

In order to determine the scope of uranium monitoring at the Allied Chemical Corporation Plant, 

SC&A compiled uranium urinalysis data for 62 claimants who were employed during the 

operational period (1959–1976). This represents over one-third of the total number of claimants 

contained in the NIOSH OCAS Claims Tracking System (NOCTS) for the site. Of the 62 

claimants, 60 were chosen purely at random; the additional 2 claimants were chosen to assure 

that all major job classifications were captured in the sample set. The breakdown by job 

classification for the group of 62 claimants is shown in Table A-1, along with the total number of 

workers in each category contained in NOCTS. As seen in Table 1, the job classifications of 

‘Operator,’ ‘Chemist/Lab Tech,’ and ‘Maintenance/Mechanic’ made up the majority of claimants 

surveyed, so the analysis likely covers the workers with the highest exposure potential. 

Table A-1. Overview of the Job Classification for Claimants Selected 

as Part of the Test Sample 

Classification # Claimants in 

NOCTS 

# Claimants 

Compiled 

(% of Total) 

% of the Total 

Number of 

Claimants in 

Category 

Laborer [redacted] 2 22% 

Maintenance/Mechanic 58 23 40% 

Operator 54 16 30% 

Chemist/Lab Tech 23 10 43% 

Foreman [redacted] 2 33% 

Miscellaneous or Unknown [redacted] 1 17% 

Administrative 11 6 55% 

Health Physics [redacted] 1* 50% 

Instrument Tech [redacted] 1* 33% 

Total 172 62 36% 

*Claimants specifically chosen to assure all job classifications were covered in the test sample

The 62 selected test cases were analyzed to determine how frequently workers were sampled 

during their employment; the results are presented by year in Table A-2. The table displays the 

number of test claimants that were employed in a given year, the % of those workers sampled, 

and the number of records per worker per year (presented as the arithmetic mean, median, and 

geometric mean). As shown, the percentage of workers sampled for uranium by year was 

generally greater than 95% and displayed 100% coverage. The exception to this is in 1965 and 

1966, when the coverage dropped to around 50% of the employed claimant test population; this 

also coincided with a general decrease in the number of test claimants employed during these 

years. This coincides with the operational shutdown described in Section 2 of the site profile. 

 

For most years, the number of records per worker was right around 12, which would indicate a 

monthly sampling schedule. In 1964 and 1967, uranium monitoring appears to have switched to 

a bimonthly schedule, though this might also be the result of the plant shutdown. There was a 
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much lower number of records per worker in 1965 and 1966, but this is likely also the result of 

the operational shutdown previously mentioned. 

Table A-2. Uranium Monitoring by Year for the Test Claimant Subpopulation 

Number of Records per Worker per 

Year Year 

# of 62 Sampled 

Claimants 

Employed in year 

% of Workers 

Sampled for 

Uranium in Year Arithmetic 

Mean Median Geometric Mean 

1959 23 95.7% 14.0 13 13.1 

1960 26 96.2% 14.0 13 13.8 

1961 25 96.0% 13.0 13 13.1 

1962 25 100.0% 15.1 13 13.2 

1963 25 100.0% 10.5 11 10.3 

1964 24 95.8% 6.5 7 6.4 

1965 17 52.9% 1.2 1 2.1 

1966 17 41.2% 3.6 0 8.0 

1967 24 100.0% 6.6 7.5 5.8 

1968 40 100.0% 10.6 12 9.7 

1969 35 97.1% 11.1 12 11.1 

1970 34 97.1% 11.5 12 11.5 

1971 33 97.0% 11.9 12 12.2 

1972 33 97.0% 10.7 11 10.9 

1973 36 100.0% 10.0 11 9.1 

1974 37 100.0% 10.3 10 10.0 

1975 42 100.0% 17.0 17.5 15.8 

1976 43 100.0% 19.5 23 17.7 

SC&A also analyzed the dosimetry records to determine the number of samples above an 

assumed threshold value of 40 μg/l.1

 The duration between the sample reported as above 40 μg/l 

and the next sample was compiled in order to characterize any special sampling practices. 

The overview for each test claimant is shown in Table A-3, which displays the following: 

 A randomly generated reference number assigned to the test claimants 

 Number of years employed during operational period 

 Total number of records during employment 

 Average number of records per year 

 Percentage of employed years with no uranium monitoring records 

 Number of records with results greater than 40 μg/l 

 Average time between sample greater than 40 μg/l and the next sample 

 1

 40 μg/l was chosen by SC&A, based on the sampled claimant records, as representing a significant 

positive value for the purposes of analyzing special monitoring practices such as re-sampling. 
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The claimants shown in Table A-3 are sorted by the number of results that were reported as 

greater than 40 μg/l. As shown in Table A-3, 42 of the 62 claimants tested (~77.4%) had 

uranium monitoring records for every year they were employed during the operational period. 

One worker had no uranium monitoring associated with Allied Chemical Corporation Plant; 

however, this worker was only employed for approximately 6 months at the very beginning of 

the operational period (1959). The final two columns of Table A-3 show the number of samples 

that were greater than 40 μg/l, and the average duration between these higher samples and the 

next uranium sample. 

One general trend that can be noticed is that workers with a large number of samples above 

40 μg/l were (on average) sampled approximately 2 weeks after the high sample was reported. 

This would likely represent a special sample, since the normal schedule was monthly for most 

years. Workers with five or fewer records above 40 μg/l often showed no change to their normal 

monthly monitoring schedule. These two observations would indicate that any special sampling 

was directed at the workers with the highest exposure potential (as evidenced by a greater 

number of samples above 40 μg/l). 

Overall, the claimant test population was missing uranium monitoring records in less than 5% of 

the worker-years compiled. The population averaged ~12 records per year, and the average time 

between a high sample (above 40 μg/l) and the next sample was about 11 days. It can also be 

qualitatively stated that the very highest samples observed were often followed up within a few 

days of the high result. 

Table 3. Overview of the 62 Test Claimants 

Reference 

Number* 

# Years of 

Employment 

During 

Operational 

Period 

Total 

Number of 

Records 

Average # 

Records per 

Year 

Percentage of 

Employment 

Years with No 

Records 

# Records 

Greater than 

40 μg/l 

Average Time 

Between Sample 

Greater than 40 μg/l 

and the Next Sample

1 18 286 15.9 0.0% 111 6.1 

2 6 104 17.3 0.0% 52 11.4 

3 12 151 12.6 0.0% 44 6.8 

4 18 210 11.7 11.1% 33 15.2 

5 18 207 11.5 0.0% 26 13.3 

6 16 187 11.7 0.0% 22 11.0 

7 17 197 11.6 11.8% 18 6.4 

8 6 76 12.7 0.0% 18 17.3 

9 18 227 12.6 5.6% 17 9.0 

10 18 208 11.6 11.1% 17 14.9 

11 17 199 11.7 11.8% 16 10.4 

12 8 75 9.4 0.0% 12 17.3 

13 18 211 11.7 5.6% 12 17.3 

14 17 175 10.3 11.8% 7 10.6 

15 15 195 13.0 0.0% 7 11.4 
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Table 3. Overview of the 62 Test Claimants 

Reference 

Number* 

# Years of 

Employment 

During 

Operational 

Period 

Total 

Number of 

Records 

Average # 

Records per 

Year 

Percentage of 

Employment 

Years with No 

Records 

# Records 

Greater than 

40 μg/l 

Average Time 

Between Sample 

Greater than 40 μg/l 

and the Next Sample

16 6 58 9.7 0.0% 7 14.7 

17 2 60 30.0 0.0% 5 3.0 

18 16 162 10.1 0.0% 5 20.6 

19 6 66 11.0 0.0% 5 26.0 

20 9 114 12.7 0.0% 4 4.3 

21 4 38 9.5 0.0% 4 9.0 

22 5 59 11.8 0.0% 4 21.3 

23 18 177 9.8 11.1% 4 23.0 

24 9 115 12.8 0.0% 4 34.0 

25 4 74 18.5 0.0% 3 1.0 

26 18 180 10.0 11.1% 3 29.0 

27 2 21 10.5 0.0% 3 29.7 

28 2 33 16.5 0.0% 2 1.5 

29 18 185 10.3 0.0% 2 15.0 

30 9 122 13.6 0.0% 2 15.0 

31 6 66 11.0 0.0% 2 24.0 

32 4 23 5.8 25.0% 1 1.0 

33 9 91 10.1 0.0% 1 1.0 

34 9 123 13.7 0.0% 1 5.0 

35 3 55 18.3 0.0% 1 9.0 

36 9 79 8.8 33.3% 0 NA 

37 17 129 7.6 23.5% 0 NA 

38 9 94 10.4 11.1% 0 NA 

39 10 99 9.9 0.0% 0 NA 

40 10 129 12.9 0.0% 0 NA 

41 10 111 11.1 0.0% 0 NA 

42 9 99 11.0 0.0% 0 NA 

43 9 110 12.2 0.0% 0 NA 

44 9 110 12.2 0.0% 0 NA 

45 9 136 15.1 0.0% 0 NA 

46 9 96 10.7 0.0% 0 NA 

47 9 105 11.7 0.0% 0 NA 

48 9 102 11.3 0.0% 0 NA 

49 4 59 14.8 0.0% 0 NA 

50 4 63 15.8 0.0% 0 NA 

51 3 22 7.3 0.0% 0 NA 
Effective Date: 

September 29, 2011 

Revision No. 

 0 (Draft)

Document No. 

SCA-TR-SP2011-0012

Page No. 

 26 of 28

 This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Table 3. Overview of the 62 Test Claimants 

Reference 

Number* 

# Years of 

Employment 

During 

Operational 

Period 

Total 

Number of 

Records 

Average # 

Records per 

Year 

Percentage of 

Employment 

Years with No 

Records 

# Records 

Greater than 

40 μg/l 

Average Time 

Between Sample 

Greater than 40 μg/l 

and the Next Sample

NOTICE:

52 2 13 6.5 0.0% 0 NA 

53 2 4 2.0 0.0% 0 NA 

54 2 41 20.5 0.0% 0 NA 

55 2 7 3.5 0.0% 0 NA 

56 2 33 16.5 0.0% 0 NA 

57 2 38 19.0 0.0% 0 NA 

58 2 15 7.5 0.0% 0 NA 

59 2 47 23.5 0.0% 0 NA 

60 1 15 15.0 0.0% 0 NA 

61 1 4 4.0 0.0% 0 NA 

62 1 0 0.0 100.0% 0 NA 

Average Value for All Test Claimants 12.2 4.8% 7.8 11.1 

* All numbers are randomly assigned. 
Effective Date: 

September 29, 2011 

Revision No. 

 0 (Draft)

Document No. 

SCA-TR-SP2011-0012

Page No. 

 27 of 28

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

APPENDIX B: FINDINGS ON ORAUT-OTIB-0024 

Procedure ORAUT-OTIB-0024 (ORAUT 2005a) was reviewed by SC&A for the Advisory 

Board on Radiation and Worker Health. The review produced seven findings. These findings 

and NIOSH’s initial responses are provided below: 

Item 1

Finding: 

The dose rates are expressed as per gram of source isotopes, rather than per gram of compound. 

Initial Response: 

ORAUT-OTIB-0024 will be revised using a modern computer code and dose rates will be 

expressed appropriately. 

Item 2

Finding: 

The OTIB limits the neutron generation to the (,n) reaction, and overlooks the contribution 

from spontaneous fission. 

Initial Response: 

ORAUT-OTIB-0024 will be revised using a modern computer code taking into account the 

neutron dose contribution from spontaneous fission. 

Item 3

Finding: 

The doses are truncated at 226Ra for the 238U decay series, and at 223Ra for the 235U decay series. 

Initial Response: 

ORAUT-OTIB-0024 will be revised using a modern computer code to calculate the neutron dose 

rate contribution from all of the alpha emitting progeny in the decay chains. 

Item 4

Finding: 

The overriding issue with the OTIB is its reliance on outdated experimental results collected 

from secondary or even tertiary sources; also, it overlooks a current computer code, 

SOURCES 4C. 

Initial Response: 

ORAUT-OTIB-0024 will be revised using a modern computer code. 
Effective Date: 

September 29, 2011 

Revision No. 

 0 (Draft)

Document No. 

SCA-TR-SP2011-0012

Page No. 

 28 of 28

 This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

NOTICE:

Item 5

Finding: 

The validity of portions of the calculation are questioned, e.g., extrapolating to higher alpha 

energies, use of 1971 NCRP rather than ICRP Publication 74 quality factors and fluence per unit 

dose equivalent, interpolation of interpolated data, etc. 

Initial Response: 

ORAUT-OTIB-0024 will be revised using a modern computer code and ICRP Publication 74 

quality factors and fluence per unit dose equivalent data. 

Item 6

Finding: 

Presenting dose rates at 1 foot and 3 feet from a point source does not appear to be a realistic 

representation of actual working conditions. 

Initial Response: 

ORAUT-OTIB-0024 will be revised using a modern computer code and more realistic exposure 

scenarios. 

Item 7

Finding: 

The average dose per neutron was calculated from a neutron spectrum that does not embody 

source nuclides with E greater than the maxima in the level branching data for the given target 

nuclide. The neutrons from the entire decay chain would have a somewhat different spectrum. 

Initial Response: 

ORAUT-OTIB-0024 will be revised using a modern computer code to accurately determine the 

neutron energy spectrum. 

The status of all seven findings is “In Abeyance.” There is no schedule as to when these seven 

findings will be resolved.