Battelle Team 

Dose Reconstruction 

Project for NIOSH 

Document Title: 

Site Profiles for Atomic Weapons Employers that 

Refined Uranium and Thorium 

Document Number: Battelle-TBD-6001 

Revision: F0 

Effective Date: 12/13/2006 

Type of Document: TBD 

Supersedes: None 

Subject Expert: R.J. Traub 

Document Owner 

Approval: Signature on File Approval Date: 12/11/2006 

Richard J. Traub, Staff Scientist 

Approval: Signature on File Approval Date: 12/11/2006 

Jay A. MacLellan, Battelle PNWD Task Manager 

Concurrence: Signature on File Concurrence Date: 12/11/2006 

Robert I. Scherpelz, Staff Scientist 

Approval: Signature on File Approval Date: 12/13/2006 

James W. Neton, Associate Director of Science 

New Total Rewrite Revision Page Change 

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION. 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page ii 

Publication Record 

EFFECTIVE 

DATE 

REVISION 

NUMBER DESCRIPTION 

28-April-06 P1 Draft document. No appendices, 

P2 Additions 

14-Aug-2006 P9 Additions to tables and text 

15-Sep_2006 P11 Responses to NIOSH comments 

7-Nov-2006 P12 Added Responses to NIOSH formal comments 

9-Nov-2006 P12a Corrections made to internal dose tables. 

14-Nov-2006 P12b Corrections to external dose tables 

16-Nov-2006 F0 Final modifications from NIOSH comments. 

6-Dec-06 F0 Print for Signature 

75. ii 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page iii 

Contents 

Publication Record ........................................................................................................................................ ii
 

Acronyms and Abbreviations .................................................................................................................... viii
 

1.0 PURPOSE AND SCOPE ................................................................................................................... 1
 

2.0 INTRODUCTION.............................................................................................................................. 2
 

3.0 Uranium Dosimetry ............................................................................................................................ 3
 

3.1 Characteristics of Uranium........................................................................................................ 3
 

3.2 Radiation Emissions .................................................................................................................. 5
 

3.3 External Dosimetry.................................................................................................................... 6
 

3.3.1 Electron Dosimetry ...................................................................................................... 6
 

3.3.2 Other Non-Penetrating ................................................................................................. 8
 

3.3.3 Photon Dosimetry......................................................................................................... 8
 

3.3.4 Neutron Dosimetry....................................................................................................... 8
 

3.4 Source Terms for External Doses ............................................................................................ 10
 

3.4.1 Submersion in Contaminated Air and Exposure to Contaminated Surfaces .............. 10
 

3.4.2 Exposures from Contaminated Surfaces .................................................................... 11
 

3.4.3 Exposures from Contaminated Soil............................................................................ 12
 

3.5 Internal Dosimetry................................................................................................................... 12
 

3.5.1 Inhalation.................................................................................................................... 14
 

3.5.2 Ingestion..................................................................................................................... 14
 

4.0 Process Description .......................................................................................................................... 15
 

4.1 Ore Digestion .......................................................................................................................... 15
 

4.2 Solvent Extraction ................................................................................................................... 16
 

4.3 Boildown and Denitration ....................................................................................................... 16
 

4.4 Oxide Reduction...................................................................................................................... 16
 

4.5 Hydrofluorination .................................................................................................................... 16
 

4.6 Reduction to Metal .................................................................................................................. 17
 

4.7 Recasting ................................................................................................................................. 17
 

4.8 Fluorination ............................................................................................................................. 17
 

4.9 Scrap Recovery........................................................................................................................ 18
 

4.10 Drum Transfer Operation ........................................................................................................ 18
 

4.11 Three Distinct Stages of Site History ..................................................................................... 18
 

5.0 Occupational Medical Dose.............................................................................................................. 20
 

5.1 Chest X-Rays........................................................................................................................... 20
 

5.2 Pelvic X-Rays .......................................................................................................................... 20
 

6.0 Occupational Environmental Dose ................................................................................................... 22
 

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6.1 Environmental External Irradiation ......................................................................................... 22
 

6.1.1 During Operations ...................................................................................................... 22
 

6.1.2 After Operations and Initial Decontamination and Before FUSRAP Remediation ... 23
 

6.1.3 After FUSRAP Remediation ...................................................................................... 23
 

6.2 Environmental Internal Dose Due to Inhalation and Ingestion of Radioactive Material ........ 23
 

6.2.1 During Operations ...................................................................................................... 24
 

6.2.2 After Operations and Initial Decontamination and Before FUSRAP Remediation ... 24
 

6.2.3 After FUSRAP Remediation ...................................................................................... 24
 

6.3 Summary of Environmental Irradiation................................................................................... 24
 

7.0 Occupational External Dose ............................................................................................................. 26
 

7.1 Radiation Doses....................................................................................................................... 26
 

7.1.1 Submersion in Contaminated Air ............................................................................... 26
 

7.1.2 Exposures from Contaminated Surfaces .................................................................... 26
 

7.2 Process Specific Dose Rates.................................................................................................... 27
 

7.3 Drum Doses ............................................................................................................................. 31
 

7.4 Summary of External Doses Received by Workers During Operation ................................... 31
 

8.0 Occupational Internal Dose .............................................................................................................. 38
 

8.1 Uranium Bioassay ................................................................................................................... 38
 

8.2 Process Specific Uranium Air Sampling Data ........................................................................ 38
 

8.2.1 Ore Digestion ............................................................................................................. 38
 

8.2.2 Solvent Extraction ...................................................................................................... 39
 

8.2.3 Boildown and Denitration .......................................................................................... 39
 

8.2.4 Oxide Reduction......................................................................................................... 40
 

8.2.5 Hydrofluorination ....................................................................................................... 41
 

8.2.6 Reduction to Metal ..................................................................................................... 42
 

8.2.7 Recasting .................................................................................................................... 43
 

8.2.8 Fluorination ................................................................................................................ 45
 

8.2.9 Scrap Recovery .......................................................................................................... 47
 

8.2.10 Drum Transfer Operations.......................................................................................... 47
 

8.3 Resuspension During Periods with no Uranium Operations ................................................... 48
 

8.4 Time-Dependent Air Concentration Data ............................................................................... 50
 

8.5 Summary ................................................................................................................................. 54
 

8.5.1 Inhalation Intakes ....................................................................................................... 54
 

8.5.1.1 Stage 1: Operations .................................................................................... 54
 

8.5.2 Resuspension During Periods with no Uranium Operations ...................................... 58
 

8.5.3 Ingestion Intakes ........................................................................................................ 58
 

9.0 References ........................................................................................................................................ 63
 

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10.0 Glossary ............................................................................................................................................ 67
 

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Figures 

Figure 1. Data from Figure 1 of Christofano and Harris (1960) showing the decreasing trend in timeweighted average air concentrations of alpha-emitting dust. 50 

Figure 2. Analysis of time-weighted average air concentrations from Christofano and Harris (1960; Fig. 

1). Variability bars are the geometric mean times or divided by one GSD. 51 

Figure 3. Use of year-specific correction factors to adjust air concentrations for processes over time. 53 

Tables 

Table 3.1 Typical isotopic composition of uranium at different enrichments. 4 

Table 3.2. Assumed activity fraction (Bq contaminant/Bq uranium) of contaminants in recycled uranium. 

5 

Table 3.3. Principal radiation emissions from natural uranium and its short-lived decay products that are 

of concern for external irradiation (not including bremsstrahlung) 6 

Table 3.4. Beta Surface Exposure Rates from Equilibrium Thickness of Uranium Metal and Compounds 

(DOE-STD-1136-2004) 7 

Table 3.5. Neutron yields from spontaneous fission and alpha-neutron reactions for oxides and fluorides. 

8 

Table 3.6 Neutron yields for trace impurities in uranium 9 

Table 3.7 Neutron dose rates and doses from alpha-neutron reaction sources 9 

Table 3.8. Measured beta/gamma dose rates from containers of pitchblende ore. 10 

Table 3.9. Dose conversion factors and daily doses for external dose due to submersion in uraniumcontaminated air. 11 

Table 3.10 Daily dose rates from natural uranium surface contamination (dpm total alpha activity) over 1 

m2 

. 11 

Table 3.11. Dose rates from uranium activity in the soil. 12 

Table 3.12. Some internal dosimetry related characteristics of uranium 13 

Table 5.1 Anatomical locations, reference organs, and organ analogues for occupational X-rays 20 

Table 6.1. Dose rate from submersion in uranium dust and standing on uranium contaminated surfaces.23 

Table 6.2. Environmental daily intakes to workers in non-operational areas of the plant. 24 

Table 6.3. Default environmental dose assumptions for a generic uranium refining site during Stage 1. 

All doses are lognormally distributed with a GSD of 5. 24 

Table 6.4. Default environmental dose assumptions for a generic uranium refining site during Stage 2. 

All doses are lognormally distributed. 25 

Table 6.5. Default environmental dose assumptions for a generic uranium refining site during Stage 3. 

All doses are lognormally distributed. 25 

Table 7.1. Process Specific Dose Rates for Uranium Refining at AWE Sites 28 

Table 7.2 Dose rates from drums of uranium compounds 31 

Table 7.3. Summary external dose rates for uranium refining facilities 32 

Table 8.1. Internal dosimetry parameters for ore digestion 38 

Table 8.2 Dust concentrations during digestion 38 

Table 8.3. Lognormal Distributions of dust concentrations during digestion 39 

Table 8.4. Internal dosimetry parameters for solvent extraction 39 

Table 8.5. Denitration area information and parameters 39 

Table 8.6. Dust Concentrations During Denitration 39 

Table 8.7 Lognormal Distributions of Dust Concentrations during denitration 40 

Table 8.8. Oxide reduction area information and parameters 40 

Table 8.9. Dust concentrations during oxide reduction 40 

Table 8.10. Lognormal Distributions of Dust Concentrations During Oxide Reduction. 41 

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Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page vii 

Table 8.11. Hydrofluorination area information and parameters 41
 

Table 8.12. Dust concentrations during hydrofluorination 41
 

Table 8.13. Lognormal Distribution Parameters for Dust Concentrations During Hydrofluorination 42
 

Table 8.14. Metal reduction information and parameters 42
 

Table 8.15. Dust concentrations during metal reduction 43
 

Table 8.16. Lognormal Distributions of Parameters for Metal Reduction 43
 

Table 8.17. Metal recasting information and parameters 44
 

Table 8.18. Dust concentrations during recasting 44
 

Table 8.19. Lognormal Distributions of Parameters for Metal Recasting 45
 

Table 8.20. Fluorination information and parameters 45
 

Table 8.21. Dust concentrations during fluorination 45
 

Table 8.22. Lognormal Distributions of Parameters for Dust Concentrations During Fluorination 46
 

Table 8.23. Dust concentrations during scrap recovery 47
 

Table 8.24 Lognormal Distributions of Parameters for Air Concentrations During Scrap Recovery 47
 

Table 8.25. Dust concentrations during drum transfer operations 48
 

Table 8.26. Lognormal Distributions of Parameters for Drum Loading Operations. 48
 

Table 8.27. Results of digitizing data from Figure 1. All concentrations (columns 3-7 and 9) are in 

dpm/m3 51
 

Table 8.28. Year-specific correction factors for tabulated process-specific air concentrations, derived from
 

analysis of Christofano and Harris (1960) Fig. 1. 53
 

Table 8.29. Intake rates for inhalation of uranium dusts during Stage 1 – Operations. 54
 

Table 8.30. Intake rates for ingestion of uranium dusts during Stage 1 – Operations. 59
 

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Acronyms and Abbreviations 

ABRWH Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

AMAD Aerodynamic Median Aerosol Diameter 

AP anterior-posterior 

AWE Atomic Weapons Employer 

BZ breathing zone 

CATI Computer-Assisted Telephone Interview 

CV coefficient of variation 

d/m/m3 disintegrations per minute per cubic meter 

DCF dose conversion factor 

DOL Department of Labor 

DR dose reconstruction 

DWA daily weighted average 

Dx diagnosis 

EEOICPA Energy Employees Occupational Illness Compensation Program Act 

FGR-12 Federal Guidance Report 12 

GA general area 

GSD geometric standard deviation 

HP health physicist 

ICD-9 International Classification of Diseases Revision 9 

ICRP International Commission on Radiological Protection 

IH industrial hygienist 

IMBA Integrated Modules for Bioassay Analysis 

IREP Interactive Radioepidemiological Program 

ISO isotropic 

keV kiloelectronvolt 

LAT lateral 

LOGNORM4 computer program 

Max maximum 

MED Manhattan Engineer District 

MeV megaelectronvolt 

Min minimum 

NIOSH National Institute for Occupational Safety and Health 

NOCTS NIOSH Occupational Claims Tracking System 

NORMSDIST() standard Normal distribution function in Microsoft Excel 

NORMSINV() inverse standard Normal distribution function in Microsoft Excel 

OCAS Office of Compensation and Support 

PA posterior-anterior 

ROT rotational 

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SD standard deviation (arithmetic) 

SEC Special Exposure Cohort 

SQRI Site Query Research Interface 

TBD Technical Basis Document 

TIB Technical Information Bulletin 

TLD thermoluminescent dosimeter 

Threshold Limit Value (® American Conference of Governmental Industrial TLV® Hygienists) 

Tn thoron (220Rn) 

TWA time-weighted average 

WLM working level month (a unit of potential alpha energy concentration) 

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Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 1 

1.0 PURPOSE AND SCOPE 

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. They 

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

documents may be used to assist NIOSH in the completion of the individual work required for each dose 

reconstruction. 

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

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

facility” or a “Department of Energy facility” as defined in the Energy Employees Occupational Illness 

Compensation Program Act of 2000 (42 U.S.C. § 7384l(5)) and (12)). 

This technical basis is established for the reconstruction of radiation doses of workers at Atomic Weapons 

Employer (AWE) sites which refined uranium under contract to the United States government from 1942 

to 1958. 

The two principal purposes of this technical basis document are (1) to provide information sufficient to 

enable dose reconstructors to estimate favorable to claimant doses for these workers on an individual 

basis under the provisions of the Energy Employees Occupational Illness Compensation Program Act 

(EEOICPA) and (2) to allow claimants, federal assessors, and others to understand the information 

sources and assumptions on which the dose estimations are based. 

This document covers the generic aspects of uranium refining and the data are useful where no other 

dosimetry data exist. Site specific appendices are included that provide information that apply to only the 

identified site. 

Employment at an AWE facility is categorized as either (1) during the contract period (i.e., when the 

AWE was processing or producing material that emitted radiation and was used in the production of an 

atomic weapon), or (2) during the residual contamination period (i.e., periods that NIOSH has determined 

there is the potential for significant residual contamination outside of the period in which weapons-related 

production occurred). For contract period employment, all occupationally-derived radiation exposures at 

the facility must be included in dose reconstructions. NIOSH does not consider the following exposures 

to be occupationally-derived: 

• radiation from naturally occurring radon present in conventional structures; and 

• radiation from diagnostic X-rays received in the treatment of work-related injuries. 

For residual contamination period employment, only the radiation exposures defined in 42 U.S.C. § 

7384n(c)(4) (i.e., radiation doses received from DOE/AEC-related work) must be included in dose 

reconstructions. Radiation dose received from DOE/AEC-related work includes: (1) radiation from radon 

consistent with NIOSH’s policies for including such radiation in the contract period; and, (2) medical 

screening X-rays, but not diagnostic X-rays for the treatment of work-related injuries. It should be noted 

that: (1) under subparagraph A of § 7384n(c)(4), radiation associated with the Naval Propulsion Program 

is specifically excluded from the employee’s radiation dose; and, (2) under subparagraph B of this 

section, radiation from a source not covered by subparagraph A that cannot be reliably distinguished from 

radiation that is covered by subparagraph A is considered part of the employee’s radiation dose. This site 

profile covers only exposures resulting from nuclear weapons-related work. Exposures resulting from 

non-weapons related work, if applicable, will be covered elsewhere. 

75. 1 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 2 

2.0 INTRODUCTION 

The Manhattan Engineer District (MED), the predecessor agency of the Atomic Energy Commission 

(AEC), asked various companies to assist in the production of uranium metal from ore. In most cases, no 

one contractor had experience in all aspects of the refining process and it was necessary to partition the 

refining process into several steps and apportion the steps among various chemical forms (Christofano 

and Harris 1960). 

In this document, the context for interpretation of the existing records is established, along with the basis 

on which to determine missing doses for periods in which records do not exist. 

Site specific appendices are also presented that may be used for dose reconstruction. Where gaps occur in 

site specific dosimetry information, the dosimetry data of the generic portion of the TBD should be used. 

The primary reference for the generic portion of the TBD is Christofano and Harris (1960). The text and 

data tables were liberally borrowed from Christofano and Harris (1960) unless otherwise stated. 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 3 

3.0 Uranium Dosimetry 

This TBD discusses the radiation doses that could result from work at AWE refinery sites. Most refinery 

work was associated with the production of uranium metal from ore. There was, however, refining for 

thorium metal. This section describes the dosimetry of uranium. 

3.1 Characteristics of Uranium 

Uranium is a heavy metal. There are several isotopes of uranium and, depending on the process in 

question, uranium may have been enriched in some of the isotopes. Most AWE uranium refining sites 

were concerned with natural uranium. 

All isotopes of uranium are radioactive and the decay progeny of uranium are also radioactive, forming a 

long decay chain. For the most part, the radioactive progeny of uranium were separated from uranium 

during the digestion and extraction phases. After those processes were completed only the uranium 

isotopes were present. But some ingrowth of uranium progeny does occur even after a short time (~100 

days) with a dose rate that exceeds the dose rate from pure uranium. 

It is important for the dose reconstructor to determine the source of the ore that was refined. For the 

purposes of this TBD, there were two types of ore; a high grade (high concentration of uranium in the 

mined ore) often refered to as pitchblend, and all other types of ore that were of lesser uranium 

concentration. The reason it is important to know the source of the ore is that the quantity of radioactive 

progeny in the ore would vary according to the source. The relative concentration of radioactive progeny 

might not vary, but the absolute quantity of the radioactive progeny would vary according to the 

concentration of uranium in the ore. The source of uranium is important for only the initial steps of the 

refining process. After the uranium has been extracted from ore, the dose rates from all uranium sources 

are identical. See Table 7.2 to obtain dose rates from drums that contain uranium. Table 3.8 lists 

measured dose rates from drums that contained pitchblend. Because it may not be known which ore was 

processed at a particular facility, the dose rates shown in the External Dosimetry section are for 

pitchblend ore. 

Natural uranium consists of four isotopes, 234U, 235U, 236U, and 238U. The term enrichment refers to the 

extent to which the amount of 235U has been increased relative to naturally occurring uranium. The 

relative amounts of each isotope for different enrichments were obtained from IMBA Expert – OCAS 

Edition and are listed in Table 3.1. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 4 

Table 3.1 Typical isotopic composition of uranium at different enrichments. 

Enrichment Isotope Weight 

Fraction 

Specific Activity 

Bq/mg pCi/mg dpm/mg 

Depleted 234U 0.000010 2.31×1000 6.24×10+01 1.39×10+02 

235U 0.001991 1.59×10-01 4.30×10+00 9.55×10+00 

236U 0.000003 7.44×10-03 2.01×10-01 4.46×10-01 

238U 0.997996 1.24×10+01 3.35×10+02 7.45×10+02 

Total 1.49×10+01 4.02×10+02 8.93×10+02 

Natural 234U 0.000054 1.24×10+01 3.35×10+02 7.44×10+02 

235U 0.007204 5.76×10-01 1.56×10+01 3.46×10+01 

236U 0.000000 0.00×10+00 0.00×10+00 0.00×10+00 

238U 0.992742 1.23×10+01 3.33×10+02 7.41×10+02 

Total 2.53×10+01 6.84×10+02 1.52×10+03 

Low 

Enrichment 

234U 0.000290 6.70×10+01 1.81×10+03 4.02×10+03 

235U 0.034989 2.80×10+00 7.56×10+01 1.68×10+02 

236U 0.000000 0.00×10+00 0.00×10+00 0.00×10+00 

238U 0.964722 1.20×10+01 3.24×10+02 7.20×10+02 

Total 8.18×10+01 2.21×10+03 4.91×10+03 

High 

Enrichment 

234U 0.010606 2.45×10+03 6.62×10+04 1.47×10+05 

235U 0.934636 7.47×10+01 2.02×10+03 4.48×10+03 

236U 0.002075 4.97×10+00 1.34×10+02 2.98×10+02 

238U 0.052683 6.55×10-01 1.77×10+01 3.93×10+01 

Total 2.53×10+03 6.84×10+04 1.52×10+05 

Recycled 

234U 0.000082 1.90×10+01 5.13×10+02 1.14×10+03 

235U 0.009700 7.76×10-01 2.10×10+01 4.65×10+01 

236U 0.000680 1.63×10+00 4.40×10+01 9.77×10+01 

238U 0.989500 1.23×10+01 3.33×10+02 7.38×10+02 

Total 3.37×10+01 9.10×10+02 5.03×10+03 

Table 3.1 includes a reference to recycled uranium. Recycled uranium is uranium that has been irradiated 

in a reactor and from which the plutonium has been extracted. Recycled uranium will contain 236U, 

sparingly found in naturally occurring uranium, due to the processes in a reactor. Note that high 

enrichment uranium contains appreciable amounts of 236U while low enriched uranium contains virtually 

no 236U. The reason for this apparent discrepancy is that high enriched uranium is generally derived from 

recycled uranium and is commonly associated with weapons development whereas low enriched uranium 

is generally associated with the commercial fuel cycle where the feed material for enrichment is natural 

uranium. 

For the time period from 1944 until 1952, all uranium used by the DOE (and predecessor agencies) was 

derived from natural sources because processes that recover uranium from spent fuel were not available 

(DOE/SO-0003, p 14)(DOE 2003). 

There is the possibility that uranium processed in refineries after 1953 was recycled uranium or contained 

recycled uranium, so for these time periods, in the absence of definitive information about the origin of 

the processed uranium, it should be assumed that the uranium contains the contaminants listed in Table 

3.2 of ORAUT-OTIB-0004, R3 (ORAUT 2005c). The appendices for each site should contain 

information concerning whether recycled uranium was present at the site. 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 5 

Table 3.2. Assumed activity fraction (Bq contaminant/Bq uranium) of contaminants in recycled uranium. 

Recycled Uranium Contaminant Pu-239 Np-237 Tc-99 Th-232 Th-228 

Activity fraction of contaminant in uranium 0.00246 0.00182 0.379 2.73E-06 2.73E-06 

Contaminant in ppb of uranium 10 1,040 9,000 10,888 * 

* assumes same activity as Th-232 

3.2 Radiation Emissions 

Natural uranium emits both beta particles (electrons) and photons (X-ray and gamma photons), as shown 

in Table 3.3, derived from Kocher (1981). The two primordial components of natural uranium are 238U 

and 235U, but some of their decay products grow into equilibrium quickly enough to be hazardous in 

processing metal. 

Uranium-238 transitions by alpha decay to 234Th, emitting traces of weakly penetrating ~ 13 keV L Xrays. However, 234Th transitions primarily to 1.17-minute 234mPa, the metastable state of protactinium-234, 

which in turn transitions to 234U with the emission of a 2.28-MeV beta particle in 98.6% of transitions 

(Kocher 1981). This relatively high-energy beta particle accounts for significant external dose rates to 

skin, lens of the eye, and to a certain extent to shallow portions of tissues such as thyroid, female breast 

and bone marrow. It also produces significant bremsstrahlung X-rays, primarily in the 30-250 keV energy 

range. The photons emitted from 234mPa are responsible for the external doses from uranium progeny. 

Because 234mPa is a decay product of 234Th the dose rate due to 234mPa will decay with an effective half-life 

of 24.1 days. Dose rates from uranium compounds are presented in Sections 7.2 and 7.3. 

Uranium-235 emits alpha particles and gamma photons in about 70% of its transitions, but occurs as 

0.7200 atom % in natural uranium. 

Like 238U, 234U emits alpha particles and traces of weakly penetrating L X-rays. 
 

 

 

 

 

 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 6 

Table 3.3. Principal radiation emissions from natural uranium and its short-lived decay products that are 

of concern for external irradiation (not including bremsstrahlung) 

Photon (x or γ) Energy 

Radionuclide 

238U 

Half-life 

4.468E+9 years 

Beta Energy (MeV Max) 

None 

(MeV) 

x: 0.013 (8.8%) 

234Th 24.1 days 0.096 (25%) x: 0.013 (9.6%) 

0.189 (73%) γ: 0.063 (3.8%) 

234mPa 1.17 minutes 2.28 (98.6%) 

γ: 0.093 (5.4%) 

γ: 0.765 (0.2%) 

235U 7.038E8 years 

~1.4 (1.4%) 

None 

γ: 1.001 (0.6%) 

x: 0.013 (31%) 

x: 0.090-0.105 (9.3%) 

γ: 0.144 (10.5%) 

γ: 0.163 (4.7%) 

γ: 0.186 (54%) 

231Th 25.5 hours 0.206 (15%) 

γ: 0.205 (4.7%) 

x: 0.013 (71%) 

0.288 (49%) γ: 0.026 (14.7%) 

234U 2.445E+5 years 

0.305 (35%) 

None 

γ: 0.084 (6.4%) 

x: 0.013 (10.5%) 

γ: 0.053 (0.2%) 

3.3 External Dosimetry 

External dosimetry refers to the radiation dose received from radioactive material that is outside the body. 

External doses can be associated with immersion in airborne radioactive material, accumulations of 

radioactive material associated with a particular process and radioactive material that contaminates the 

floor or other surfaces. 

Nonpenetrating dose from natural uranium consists primarily of electrons with energies above 15 keV. 

For consistent presentation, exposure or dose is reported as: 

•	 penetrating, assumed to be associated with photons photons having the following energy bin 

structure: 50.5% of the exposure is associated with photons with energies below 30 keV; 28.5% is 

from photons with energies between 30 keV and 250 keV, and the remaining 21.0% is for photons 

with energies greater than 250 keV, and 

•	 nonpenetrating, assumed to be associated with photons of energies less than 30 keV or with electrons. 

3.3.1 Electron Dosimetry 

Uranium metal and compounds emit beta and electron radiation that can irradiate the skin, and to a more 

limited extent, the shallow organs of the body. Table 3.4, taken from DOE-STD-1136-2004 (DOE 2004) 

shows the measured beta surface dose rates from uranium metal and selected uranium compounds. 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 7 

Table 3.4. Beta Surface Exposure Rates from Equilibrium Thickness of Uranium Metal and 

Compounds (DOE-STD-1136-2004) 

Source 

Beta Surface Exposure 

Rate, mrad h-1 

U-Nat metal slab 233 

UO2 207 

UF4 179 

UO2(NO3)26H20 111 

UO3 204 

U3O8 203 

UO2F2 176 

Na2U2O7 167 

a. Beta surface exposure rate in air through a polystyrene 

filter 7mg/cm2 thick. 

The data shown in Table 3.4 show that the dose rates from uranium metal exceed the dose rates from 

other uranium compounds. However, the dose rates from uranium oxides and UF4 are sufficiently similar 

in magnitude to the dose rates from uranium metal so that uranium metal dose rates can be assumed to be 

representative of the dose rates from all uranium compounds. 

Nonpenetrating dose from natural uranium consists primarily of electrons with energies >15 keV. Skin 

doses to uranium refining workers can be estimated using the same methods used to calculate the dose 

from uranium metal. Skin doses (7 mg/cm2

) are estimated for two worker cases: the hands and forearms 

of a worker who handles uranium metal, and the other skin surfaces of a worker who handles the metal. 

Assuming that an operator’s hands are in contact with the uranium metal or compounds for 50% of the 

day would give a non-penetrating dose to the hands and forearms of 1.17 rad for a 10-hour workday (0.93 

rad for an eight-hour workday). This dose estimate assumes that no gloves or protected clothing shielded 

the worker’s hands and forearms, which may have been true for some of the earliest years of uranium 

metal-working, and is an assumption that is favorable to the claimant for other workers. These assumed 

dose rates are also favorable to claimants who worked at uranium refining facilities other than where 

metal reduction occurred. 

For dose to other skin on the worker’s body that is not in direct contact with uranium, but is nearby (for 

example, a worker’s neck and face when the hands are in contact with metal), a dose relation can be used 

that estimates this dose to be 10 times the photon dose rate at 1-foot. This relation, based on a review of 

film badge data, is documented in ORAUT (2005e). The photon dose rate at 1 foot from a slab of 

uranium is 2.08 mrem/h (ORAUT-TKBS-0034) (ORAUT2005e), which would give a non-penetrating 

dose rate of 20.8 mrem/h using this relation. For a worker spending 50% of a 10-hour workday handling 

uranium, this would give a dose rate of 0.10 rad/d (0.083 rad/d for an eight-hour workday). This dose rate 

would be appropriate for areas of the skin other than the hands and forearms, and for tissues and organs 

near the body surface, including breast, penis, testes, skin and eye. 

For a worker who does not directly handle uranium materials, the non-penetrating dose rate would usually 

be zero, but it could be argued that a worker who was very close to the uranium material without actually 

touching it may have been exposed to 0.10 rad/d of non-penetrating radiation, similar to the non-contact 

skin doses for a metal handler described in the previous paragraph. 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 8 

3.3.2 Other Non-Penetrating 

After 1952, small quantities of primarily alpha (Pu-239, Np-237, and Th-232/228) and beta (Tc-99) 

emitting radionuclides found their way into the uranium metal via recycling (see Table 3.2). Because of 

their primarily non-penetrating radiation types, relatively low activities, and relatively low external 

radiation hazard when compared to that of their uranium matrix, their contribution to dose is considered 

adequately addressed by the uranium external dose estimates. A quick check of the relative penetrating 

dose from an overestimate of recycled uranium contaminants in ground surface contamination shows that 

they contribute less than 1% to penetrating dose. The assumed non-penetrating doses from uranium in this 

document are sufficiently large to bound any small contribution from Tc-99. 

3.3.3 Photon Dosimetry 

•	 According to ORAUT-OTIB-0004, Rev. 3 (ORAUT 2005c) the majority of photons from natural 

uranium are in the 30 to 250 keV energy range. Calculations using the program MicroShield indicate 

that about 56% of the photon energy released from natural uranium is above 250 keV. Uranium, even 

when present as chemical compounds, provides considerable shielding of the lower energy photons 

and will tend to harden the spectrum, causing the majority of photons emitted from uranium, to have 

energies greater than 250 keV. While it is recognized that uranium sources will have a somewhat 

hardened photon spectrum, exposure to a thin layer of uranium on a surface will result in a larger 

fraction of exposure to lower energy photons. The analysis described above assumed workers were 

exposed to photon energies in the 30 to 250 keV range, which is favorable to claimants when 

considering both organ dose conversion factors and radiation effectiveness factors. 

3.3.4 Neutron Dosimetry 

There are two sources of neutrons. First, neutrons arise from (α-n) reactions, where the reactant is 

fluorine which occurs during the production of UF4 and UF6 and oxygen that is present in uranium oxides. 

Second, there is a small amount of spontaneous fission by uranium. Table 3.5 shows spontaneous fission 

and alpha-neutron yields for various uranium isotopes and 232Th. 

Table 3.5. Neutron yields from spontaneous fission and alpha-neutron reactions for oxides and 

fluorides. 

Isotope 

Spontaneous 

Fission Yield, 

n s-1 g-1 

Yield in 

Oxides, 

n s-1 g-1 

Yield in 

Fluorides, 

n s-1 g-1 

232Th 2.2E-5 232U 1.3 1.49E4 2.6E+6 233U 8.6E-4 4.8 7.0E+2 234U 5.02E-3 3 5.8E+2 235U 2.99E-4 7.1E-4 0.08 236U 5.49E-3 2.4E-2 2.9 238U 1.36E-2 8.3E-5 0.028 

(DOE-STD-1136-2000) 

The production of neutrons by alpha-neutron interactions in uranium compounds will vary according to 

the light element involved in the interaction. Table 3.6, obtained from DOE-STD-1136-2004 (DOE 

2004), shows the neutron yields for various light elements. The data in this table shows that, for the 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 9 

compounds present in refineries, the greatest yield is from fluorine and the yield from oxygen compounds 

is two orders of magnitude less. 

Table 3.6 Neutron yields for trace impurities in uranium 

Element Neutron Yield 

per 106 Alphas 

at 4.7 MeV (234U) 

Li 0.16 ± 0.04 

Be 44. ± 4 

B 12.4 ± 0.6 

C 0.051 ± 0.002 

O 0.040 ± 0.001 

F 3.1 ± 0.3 

Na 0.5 ± 0.5 

Mg 0.42 ± 0.03 

Al 0.13 ± 0.01 

Si 0.028 ± 0.002 

Cl 0.01 ± 0.01 

(DOE-STD-1136-2004) 

An analysis of neutron dose rates from uranium and thorium oxides and fluorides was performed for the 

Mallinckrodt site (ORAUT-TKBS-0005) (ORAUT 2005a). The calculated neutron dose rates are shown 

in Table 3.7. The calculated neutron dose rates can then be compared to the measured dose rates from 

pitchblende ore, shown in Table 3.8. Pitchblende ore contains a very high concentration of U3O8. The 

measured dose rates from pitchblende ore were divided by 4.43 and 5.91 so that the measured and 

calculated dose rates would represent the dose rate from the same amount of U3O8. Comparison of the 

measured beta/photon dose rates and the calculated neutron dose rates show that the neutron dose rate is 

about 0.07% of the beta/photon dose rate and need not be included in dose rate calculations. For uranium 

metal, the neutron dose rate is even less important. The data in Table 3.5 show that the neutron emission 

rate due to spontaneous fission is much less than the neutron emission rate due to (alpha, n) reactions. 

The calculated neutron dose rates from selected uranium compounds are shown in Table 3.7. 

Table 3.7 Neutron dose rates and doses from alpha-neutron reaction sources 

Form Source Target 

element(s) 

Weight in 

container 

Dose rate at 

1 foot, rem/h 

Dose Rate at 

3 feet, rem/h 

U3O8 U natural mix O 100 lb 1.29E-07 1.44E-08 

U natural mix + daughters O 100 lb 1.22E-07 3.67E-04 

UO3, 

UO2 

U natural mix O 75 lb 9.71E-08 3.24E-05 

U natural mix + daughters O 75 lb 8.27E-07 9.19E-08 

UF4 U natural mix F 75 lb 6.92E-06 7.69E-07 

U natural mix + daughters F 75 lb 5.91E-05 6.57E-06 

Na2U2O7 

(soda salt) 

U natural mix O, Na 75 lb 2.61E-07 2.90E-08 

U natural mix + daughters O, Na 75 lb 2.13E-06 2.37E-07 

Data taken from Table 34 of ORAUT-TKBS-0005 (ORAUT 2005a) 

In the alpha-neutron case, the neutron energy was taken as 1.5 MeV because it is the approximate maximum energy for Th-232, the 

other isotopes also emit neutrons in the range 1.0-2.0, and the flux-to-dose conversion factor varies slowly in this range. 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 10 

The neutron dose rates of Table 3.7 can be compared with the measured beta/gamma dose rates for drums 

that contain pitchblende ore as shown in Table 3.8. Pitchblende ore contains a large fraction of U3O8 and 

allows comparison with the neutron dose rates in Table 3.7. 

Table 3.8. Measured beta/gamma dose rates from containers of pitchblende ore. 

Source Material Exposure 

mrep/h or mR/h 

443 lb 100 lb 75 lb 

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: 60” 2.4 0.55 0.41 

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: 45” 3.6 0.81 0.61 

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: 30” 7.1 1.6 1.2 

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: contact 90 20 15 

Data taken from Table 33 of ORAUT-TKBS-0005 (ORAUT 2005a) . 

The column labeled “443 lb” contains the actual measured dose rates. The column labeled “100 lb” contains dose rates 

that have been reduce by a factor of 4.43 so they can be compared with the neutron dose rates shown in Table 3.7. 

The data shown in Table 3.7 and Table 3.8 indicate that, for U3O8, the neutron dose rate is about 0.3% of 

the beta/gamma dose rate. This fraction is small enough to be neglected in dose calculations. The data in 

Table 3.5, indicate that, for uranium metal, the neutron dose rate will be an even smaller fraction of the 

total dose rate. Dose reconstructors may ignore the neutron dose rates for dose calculations involving 

uranium metal and compounds. 

3.4 Source Terms for External Doses 

External doses in uranium refining facilities typically arise from five different modes of exposure: 

• Submersion in air contaminated with uranium dust, 

• Exposure from contaminated surfaces, 

• Exposure to electrons from the surface of the uranium, particularly if the uranium is metal, 

• Exposure to photons from the surface of uranium, and 

• Exposure to an annual diagnostic X-ray. 

Exposures to diagnostic X-rays are discussed in Section 5.0. The other components of external dose are 

discussed in Sections 3.3. 

3.4.1 Submersion in Contaminated Air and Exposure to Contaminated Surfaces 

When workers are enveloped in a cloud of radioactive dust, they will receive a small amount of external 

dose. External exposure rates from uranium and its radioactive progeny are shown in Table 3.9. The 

doses were calculated using the computer code MicroShield version 6.02 (Grove Engineering 2003). The 

calculated dose rates are for natural uranium and include the dose contribution from the radioactive 

progeny of 238U, 235U, and 234U. In accordance with (ORAUT 2005c) 100 days of radioactive progeny 

ingrowth was assumed for these calculations. 

Notice that the dose quantity calculated for this conversion factor is exposure, in units of milliRoentgen. 

For this particular conversion factor, 50.5% of the exposure is associated with photons with energies 

below 30 keV; 28.5% is from photons with energies between 30 keV and 250 keV, and the remaining 

21.0% is for photons with energies greater than 250 keV. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 11 

The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine 

which of the tabulated organ dose rates should be used for any particular ICD code. 

Table 3.9. Dose conversion factors and daily doses for external dose due to submersion in uraniumcontaminated air. 

External Dose Conversion factor 

Time since separation (mR/h per dpm(α)/m3

) (mR/d per dpm(α)/m3) 

100 d 2.46E-09 1.97E-08 

3.4.2 Exposures from Contaminated Surfaces 

When workers are working on a contaminated surface, they will receive a small amount of external dose. 

External dose rates from uranium and its radioactive progeny are shown in Table 3.10. The dose 

reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine which of 

the tabulated organ dose rates should be used for any particular ICD code. 

The doses were calculated using the computer code MicroShield version 6.02 (Grove Engineering 2003). 

The calculated dose rates are for natural uranium and include the dose contribution from the radioactive 

progeny of 238U, 235U, and 234U. In accordance with (ORAUT 2005c) 100 days of radioactive progeny 

ingrowth was assumed for these calculations. 

Notice that the dose quantity calculated for this conversion factor is exposure, in units of milliRoentgen. 

For this particular conversion factor, 76.7% of the exposure is associated with photons with energies 

below 30 keV; 10.0% is from photons with energies between 30 keV and 250 keV, and the remaining 

13.3% is for photons with energies greater than 250 keV 

The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine 

which of the tabulated organ dose rates should be used for any particular ICD code. 

Table 3.10 Daily dose rates from natural uranium surface contamination (dpm total alpha activity) over 1 

2 m . 

Surface contamination dose conversion factors 

Time since separation (mR/h per dpm(α)/m2) (mR/d per dpm(α)/m2) 

100 d 5.61E-10 4.49E-09 

The quantity of uranium on the floor surface can be obtained from floor survey (wipe) sample 

measurements. Most survey samples were based on a 100 cm2

 sample rather than a 1 meter squared 

sample, in this case the dose values in Table 3.10 should be divided by 10,000. The dose reconstructor 

should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine which of the tabulated 

organ dose rates should be used for any particular ICD code. 

When measured floor contamination rates are not available the contamination on the floor may be 

estimated from measured air concentrations. The floor activity may be computed from the air 

concentrations following the method of ORAUT (2005c) (ORAUT-OTIB -0004, Rev 3, p 15). The level 

of surface contamination was determined by first calculating a terminal settling velocity for 5-μm activity 

mean aerodynamic diameter (AMAD) particles. The calculated terminal settling velocity was 0.00075 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 12 

meters per second. It was assumed that the surface contamination level was due to 365 days of constant 

deposition from the constant air concentration to give a deposition factor of 2.37E4 meters. The floor 

contamination level is then estimated as Floor Concentration (dpm/m2

) = Air Concentration (dpm/m3) × 

2.37E4 meters. This method calculates the surface contamination over 1 m2

 and the dose factors in Table 

3.10 should be used. The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 

(ORAUT 2006) to determine which of the tabulated organ dose rates should be used for any particular 

ICD code. 

Notice that the dose quantity calculated for this conversion factor is exposure, in units of milliRoentgen. 

For this particular conversion factor, 50.5% of the exposure is associated with photons with energies 

below 30 keV; 28.5% is from photons with energies between 30 keV and 250 keV, and the remaining 

21.0% is for photons with energies greater than 250 keV. 

The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine 

which of the tabulated organ dose rates should be used for any particular ICD code. 

3.4.3 Exposures from Contaminated Soil 

Workers at a facility that is no longer performing AWE work will also be exposed to radiation from 

residual contamination. If measurements of the contamination are not available for this time period, the 

doses presented in Table 3.11 can be used as an estimate of the external doses from this dose pathway. 

Table 3.11. Dose rates from uranium activity in the soil. 

Soil contamination dose conversion factor 

Time since separation (mrem/h per g/cm3

) (mrem/d per g/cm3) 

100 d 152.78 1222.22 

15 y 152.85 1222.76 

3.5 Internal Dosimetry 

Internal dosimetry refers to the radiation dose received from radioactive material that is inside the body. 

Radioactive material can enter the body via inhalation of radioactive dusts, ingestion of radioactive dusts 

as may happen due to incidental hand to mouth transfers, and contaminated wounds. Internal doses for 

individuals can be estimated based on airborne concentrations of radioactive materials by using the 

computer code IMBA Expert OCAS Edition that implements the biokinetic models that have been 

developed by the International Commission on Radiological Protection (ICRP). 

IMBA Expert also provides mechanisms for interpretation of bioassay data. Bioassay data can be of two 

kinds; in vivo bioassay and in vitro bioassay. In vivo bioassay consists of placing the person near 

radiation detectors that measure the amount of radiation that exits the body and based on the efficiency of 

the counting process the quantity of radioactive material in the body can be determined. In vitro bioassay 

consists of collecting the excreta (urine and feces) or breath of an individual who has had an intake of 

radioactive material and, by comparing the measured results with the results predicted by biokinetic 

models, the quantity of radioactive material that would be required to be in the body to produce the 

observed excretion is calculated. 

Internal doses are influenced by several properties of the material that is taken into the body. Pertinent 

quantities include the particle size of the aerosol, the shape of individual particles in the aerosol, the 

density of the airborne material, and the solubility of the material. 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 13 

During the refining process, uranium undergoes several chemical reactions and so there are several 

uranium compounds that may have become airborne during various processes. The impact of uranium 

forming compounds is that the specific activity of uranium metal, shown in Table 3.1 will be different 

than the specific activity of the compound. The specific activities of various compounds of natural 

uranium are shown in Table 3.12. 

In the text of this TBD the chemical compound associated with each process will be identified. 

Table 3.12. Some internal dosimetry related characteristics of uranium 

Material 

U weight 

Fraction 

Density Solubility Data Specific Activity 

g/cm3

 Type f1 Bq/mg pCi/mg dpm/mg 

UO2 0.881498 10.96 S 0.002 2.23E+01 6.03E+02 1.34E+03 

UO3 0.832190 7.29 M 0.02 2.11E+01 5.69E+02 1.26E+03 

U3O8 0.848001 8.3 S 0.002 2.15E+01 5.80E+02 1.29E+03 

UF4 0.757999 6.7 M 0.02 1.92E+01 5.18E+02 1.15E+03 

UF6 0.676181 - F 0.02 1.71E+01 4.63E+02 1.03E+03 

UO2(NO3)2 0.604077 - F 0.02 1.53E+01 4.13E+02 9.18E+02 

UO2F2 0.772760 - F 0.02 1.96E+01 5.29E+02 1.17E+03 

UCl4 0.626656 4.87 M 0.02 1.59E+01 4.29E+02 9.53E+02 

Unless specified otherwise, the characteristics of aerosols are: 

• Particle Size is 5 micron AMAD 

• Shape Factor is 1.5 

• Density is 3 

• Lung Solubility Type as given in Table 3.12 

• Absorption factor f1 as given in Table 3.12 

The dose reconstructor should use the default values shown above. The default values of ICRP-66 (ICRP 

1994) should be used. The density values shown in Table 3.12 are for reference only. 

The intake of radioactive material via inhalation depends on the characteristics of the worker, that is their 

breathing rate and the extent to which they breath through the mouth. IMBA does not calculate intakes 

based on air concentrations, rather the dose reconstructor is required to calculate the intake from the 

breathing rate and the air concentration. Unless specified otherwise in the site specific guidance, the 

following worker characteristics should be assumed: 

• Worker type: Light Worker 

• Breathing Characteristics: Nasal Augmentor 

• Breathing Rate: 1.2 m3

/h or 9.6 m3/day (for an 8 h day) 

The intake is the product of the breathing rate, shown above, and the air concentration. The generic air 

concentrations are listed in the Chapter 8. For some processes, the air concentration tables show a DWA 

and in those cases the Daily Weighted Average (DWA) (See Section 11) is the air concentration to use. If 

a DWA air concentration is not available then use the average daily exposure air concentration. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 14 

3.5.1 Inhalation 

Inhaled uranium can have two sources: 

• Uranium dust generated by plant operations. This source will dominate while the plant is operating. 

• Uranium dust generated by resuspension of material on contaminated floors by casual foot traffic. 

Specific information concerning these two sources are discussed in the internal dosimetry section. 

3.5.2 Ingestion 

In the case where inhalation intakes are calculated from air concentrations, ingestion intakes are also to be 

considered. NIOSH (2004) indicates that the ingestion rate, in terms of dpm for an 8-hour workday, can 

be estimated by multiplying the air concentration in dpm per cubic meter by a factor of 0.2. Adjusting for 

a 10-hour day, and converting dpm to pCi, the ingestion intake would be 0.114 times the inhalation 

intake. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 15 

4.0 Process Description 

Refining consists of converting uranium in the form of U3O8 to either metal or UF6. Nine processes were 

associated with the refining of uranium. They are: 

• Ore digestion 

• Solvent extraction 

• Boildown and Denitration 

• Oxide Reduction 

• Hydrofluorination 

• Reduction to Metal 

• Recasting 

• Fluorination 

• Scrap Recovery 

The industrial hygiene aspects of uranium refining varied over time. Initially, the process was spread 

over many facilities and there were few dust control measures in the plants. Packaging and transferring of 

materials between sites often involved some of the highest exposures. With time the processes were 

improved and organized into two parallel integrated DOE plants. The process descriptions are a summary 

of the information provided by Christofano and Harris (1960). 

4.1 Ore Digestion 

Ore digestion was the first step in the refining process. In this process hot uranium ore (often screened to 

20 mesh powder - Christofano and Harris (1960)) or concentrate (without processing) was dissolved in 

hot concentrated nitric acid. Separate systems were maintained for hot ore and concentrate as the hot ore 

required shielding and some remote operation. In the hot ore process, piping became contaminated and 

produced significant amounts of background γ-radiation. If materials were added too rapidly, the 

violence of the reaction would flood the air with uranium-laden nitric fumes. Materials that could become 

airborne were dusts, mist, and nitric oxide fumes. These reactions were kept under a slight negative 

pressure through a nitric acid recovery system. Uncontrolled manual feeding operations had much higher 

exposures than contained and ventilated operations. Highest exposures were associated with the 

maintenance and cleaning of filters used to separate the dissolved uranium solution from the silica (sand). 

Later modification allowed the slurry containing sand to be fed into the solvent extraction process 

removing this high exposure task. 

A byproduct from this process included radium cake (a barium sulfate precipitate) which was separated 

out by centrifuge. Drums of this material had high γ-radiation levels at their surface. 

The uranium ore was a powder, often 20 mesh (Christofano and Harris (1960)). Twenty mesh powder 

has a diameter of 840 μm (Lide 1995, p 15-37). If 840 μm is assumed to be the upper 99%ile of the 

aerosol particle size distribution, the equations of ICRP-66 (ICRP 1994) would imply that the actual 

AMAD of the aerosol is about 54 μm. For this project dusts have an assumed particle size distribution of 

5 micron AMAD. 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 16 

4.2 Solvent Extraction 

This process separates uranium nitrate from nitrates of other, undesirable metals including radioactive 

components. The uranium that exits the solvent extraction process is the purest form of uranium in the 

refining process because ingrowth of uranium progeny that occurs after this time will reduce the purity of 

uranium. The byproduct is called raffinate and may be concentrated and calcined to recover nitric acid or 

simply discarded – presenting a problem of waste storage. The process is continuous, enclosed and 

utilizes only liquids, and thus presents the least exposure of the refining processes. Ethyl ether was used 

early on and replaced by tributyl phosphate in kerosene. The uranium-bearing aqueous solution from the 

digestion process is fed into the top of the extraction column, while the organic solvent is fed into the 

bottom. The organic extract is then scrubbed with hot demineralized water in a similar column. The 

resulting dilute solution of highly purified uranium nitrate is the purest form of uranium in the refining 

process. From here on out contaminants are added by the process. 

4.3 Boildown and Denitration 

Denitration is the process of converting the extracted uranium nitrate to UO3(orange oxide). The dilute 

solution of uranium nitrate is concentrated into uranyl nitrate hexahydrate (UNH) using multi-effect 

evaporators and boildown tanks. If metal is the desired end product, denitration is accomplished by 

controlled heating in gas-fired pots with continuous agitation. First water vapor is driven off, followed by 

oxides of nitrogen (into an acid recovery system), leaving uranium trioxide (UO3). The UO3 is dry and 

very dusty. Hand removal by shovel and scraping resulted in very high exposures (30,000-80,000 

d/m/m3

). This method was later replaced by pneumatic conveyor which considerably reduced exposures 

(560 d/m/m3

). However, when improper heat regulation caused UO3 to build on the pots, it had to be 

chopped out causing the same high exposures as manual removal. A continuous rotary kiln denitration 

process could be used if the product was to be UF6, but introduced too much product variation for metal 

production. This process was used at Paducah and Oak Ridge. Even in these situations, with enclosures 

and ventilation, exposures in the remote control areas exceeded the maximum permissible exposure level 

and required air supplied respirators and rotation of personnel. 

4.4 Oxide Reduction 

The reduction of UO3 to UO2 first requires that the lumpy UO3 be pulverized to a fine powder. It is then 

heated and hydrogen (H2) is passed over it to form UO2 and water vapor. Three methods were employed. 

Two batch methods heated the UO3 in furnaces, one with the UO3 spread thinly on copper trays and the 

other continuously agitating the UO3 with rakes to assure contact with the H2. The later, continuous 

method used screw flites (similar to helical screw conveyors) to move UO3 through horizontal reactors 

with a counterflow of H2. Off gases, which contained excess H2, were burned. The continuous method 

produced a significant amount of uranium salts that had to be filtered from the burned hydrogen stream. 

High exposures were incurred in the batch operations feeding of UO3 and bagging of the fine UO2 

powder. The continuous process reduced exposures by approximately two magnitudes. Manual bagging 

was done in this process only for special circumstances. 

4.5 Hydrofluorination 

Hydrofluorination was used to convert UO2 to UF4 (green salt). The process is similar to the oxide 

reduction process described previously, except that hydrogen fluoride (HF) is passed over the heated UO2, 

with water vapor and excess HF collected by acid recovery units. In the batch mode, HF is passed over 

heated UO2 in shallow trays on an open table. Later this table was enclosed and ventilated, reducing the 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 17 

exposure by about 7 fold (from 10,000 d/m/m3

 to 1,500 d/m/m3). Eventually a continuous method was 

employed using horizontal screw flite reactors identical to those used in the oxide reduction process. In 

some places UO3 was converted directly to UF4 by using a series of these reactors. Exposures during 

continuous processing are negligible, but cleaning and maintenance of the spiral flite reactors presents 

exposure problems. 

UF4 was then used as feedstock for two separate processes. The first process was for production of 

uranium metal. The second process was for the production of UF6 which was used in the enrichment 

process. 

Packaging of uranium powders has presented high exposure problems. High velocity local exhaust 

ventilation around the lip of the drum and total enclosure was tried. However, excessive airborne 

concentrations were still measured in the breathing zones of operators. In addition, empty drums after use 

remained sources of airborne exposure. 

4.6 Reduction to Metal 

Metal is obtained by thermite reduction of the fluoride salt with magnesium in refractory lined steel 

“bombs”. The “bomb” is lined with a refractory material and then filled with a mixture of UF4 and Mg 

chips. The bomb is sealed and heated to start an exothermic reaction that heats the uranium beyond its 

melting point. Uranium metal forms at the bottom of the bomb and a MgF slag floats to the top. After 

the reaction has run to completion, the bomb is opened and the metal is dislodged from the slag and 

refractory lining. This “breakout” task results in the greatest amount of dust for the metal production 

process. 

4.7 Recasting 

Metallic uranium produced by the thermite reaction contains impurities, mostly on the surface of the 

metal, that make the uranium unsuitable for reactor fuel. The metal was vacuum cast into shapes called 

metal derbies. Impurities in the original metal either were volatile and collected on the lid of the vacuum 

casting machine, or floated to the top of the derby. The top of the derby that contained impurities was 

sawed off and the derby was then ready for metal processing. 

4.8 Fluorination 

UF6 is used in the process of enriching the 235U content of uranium. UF6 is produced from UF4 by a 

process known as fluorination. The UF6 product is a gas and can be distilled to remove impurities. 

UF6 is produced by passing a reactive gas, in this case anhydrous fluorine, over a bed of UF4. As UF6 is 

produced it evaporates and is carried away from the reaction. 

Three fluorination methods were used. The first two methods differ only in the tasks to clean the reaction 

vessels. 

For the first and earliest method, UF4 was spread in shallow trays, sealed in reaction tubes and placed in a 

reactor furnace. The reaction was then allowed to run to completion with the evolution of UF6 gas that 

was collected. The process is specific to uranium and so contaminates (ash) would remain in the reaction 

trays. The ash was brushed into a collection barrel. The ash contained 234Th and 234mPa and the external 

dose rates in the vicinity of the collection drum approached 1R/h. The collection drums would be sealed 

and stored for about 6 months at which time the radionuclides in fresh ash would have decayed to 234U 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 18 

with a significantly lower dose rate. Eisenbug (1975) indicates that the dose to the hands from the ash 

could range from 2-3 rad/week. This dose rate is consistent with the dose rates shown in Table 7.1. 

The second method was the same as the first method with the addition of a ventilated enclosure for the 

vacuum and brushing steps. This improvement reduced the radiation exposures of workers. 

The third method was a vertical continuous reactor where the ash is collected in an ash pit and was 

automatically removed. 

4.9 Scrap Recovery 

Uranium was recovered from scrap. The nature of the scrap would vary from plant to plant and, in all 

likelihood, from batch to batch. The basic process was to dry the scrap and then calcine it either in trays 

or multiple hearth furnaces. Metallic uranium is burned to the oxide, generally U3O8, with the evolution 

of some fumes. The steps associated with the scrap recovery process are: 

• Calcine the scrap in a furnace to convert metal and easily oxidized compounds to U3O8. 

•	 Pulverize the oxidized material, digest it with acids, filter the material, and then treat the material 

with hydroxide to precipitate uranium. 

•	 The precipitate is separated from the liquor and purified and concentrated into a diurinate salt which 

can be used as feedstock for the refinery. 

4.10 Drum Transfer Operation 

Purified uranium ore were transferred to the refining facilities in 55 gal drums (see picture in Christofano 

and Harris 1960, p79/443). Dose rates from drums of uranium ore are presented in Table 7.2. The dose 

rates due to neutrons from spontaneous fission are about 0.09% of the photon dose rates. 

4.11 Three Distinct Stages of Site History 

Site history can generally be divided into three distinct stages: 

1. operations through and including cessation of operations and initial decontamination 

2. the period following initial decontamination through and including FUSRAP decontamination 

3. post-FUSRAP decontamination. 

During operations (Stage 1), there may or may not have been personnel monitoring, air sampling, area 

surveys, and contamination monitoring, but this period may include monitored doses, unmonitored doses, 

and missed doses due to occupational medical exposures, environmental exposures, internal irradiation 

due to intakes in the workplace, and external irradiation due to radioactive material in the workplace and 

other AWE-associated radiation sources such as radiography equipment. 

Generally, there was little or no monitoring of workers after initial decontamination (Stage 2). For this 

period of time, exposures to residual contamination include external irradiation and inhalation intakes of 

resuspended contamination. During Stage 2, there is no occupational internal or external irradiation due to 

AWE operations or occupational medical dose. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 19 

During Stage 3, after FUSRAP remediation, it is assumed that persons working on the site are exposed to 

environmental radiation at the limits allowed for a remediated site. During Stage 3, there is no 

occupational internal or external irradiation due to AWE operations or occupational medical dose. 

Clearly, there are exceptions to the default three-stage history. Some sites may not have had any FUSRAP 

cleanup. Others may have been decontaminated several times. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 20 

5.0 Occupational Medical Dose 

In many AWE facilities, workers received a pre-employment X-ray examination, and in many of these 

facilities, the X-ray was also administered annually to employees who worked with uranium. The “preemployment” X-ray was often administered at the start of AEC operations. In the absence of site-specific 

information about medical X-rays required for employment, the assumption should be made that it was 

administered at the beginning of AEC operations (or at the beginning of a worker’s employment) and 

annually during uranium processing. The exposure geometry was assumed to be posterior-anterior (PA) 

(NIOSH, 2002). 

5.1 Chest X-Rays 

Information to be used in dose reconstruction for the early years for which no specific information is 

available is provided in ORAUT-OTIB-0006, the dose reconstruction project technical information 

bulletin covering diagnostic x-ray procedures (ORAUT 2005d). ORAUT (2005d) should be referred to 

regarding the underlying bases, interpretation details, and a sample of a summary data table where actual 

beam data are known. 

Doses for organs not listed in ICRP Publication 34 (ICRP 1982) but specified in the IREP code should be 

determined by analogy with anatomical location as indicated below. Analogues for IREP organs not 

included in ICRP 34. 

Table 5.1 Anatomical locations, reference organs, and organ analogues for occupational X-rays 

Anatomical location ICRP 34 reference organ IREP organ analogues 

Thorax Lung Thymus 

Esophagus 

Stomach 

Bone surface 

Liver/gall bladder/spleen 

Remainder organs 

Abdomen Ovaries Urinary/bladder 

Colon/rectum 

Uterus 

Head and neck Thyroid Eye/Brain 

As ORAUT (2003c) notes, for any individual entrance skin exposure (ESE) or derived organ dose, an 

uncertainty of ± 30% at the one sigma confidence level may be assumed; for further conservatism it may 

be appropriate to assume that errors are all positive and thus only the + 30% should be used. 

5.2 Pelvic X-Rays 

Guidance regarding pelvic x-ray examinations is available in Appendix A of ORAUT-OTIB-0006 

(ORAUT 2005d). Note that pelvic X-rays would be applicable to only some workers, i.e. those working 

with fluorides; presumably this class would consist of workers involved with the UO2-to-UF4 and the 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 21 

UF4-to-derby processes, including workers who handled the hydrogen fluoride supply and storage 

functions. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 22 

6.0 Occupational Environmental Dose
 

In the context of reconstructing doses for Atomic Weapons employees, Occupational Environmental Dose 

includes doses due to 

•	 direct, exernal irradiation from material in process, in storage, in transit, and from radioactive 

contamination on surfaces, in soil and water, and in plumes of radioactive material 

•	 internal irradiation due to intakes of fugitive radioactive material through inhalation, ingestion, 

dermal contact with radioactive material, and through wounds. Generally, dermal contact and wound 

entry are assumed to be negligible in the absence of information to the contrary. 

6.1 Environmental External Irradiation 

Typically, energy employees who were not categorized as radiation workers were not monitored using 

personal dosimeters. However, the work environment for these employees was often routinely monitored 

using area dosimeters or periodically monitored using survey instrumentation to measure the 

“background” environmental radiation levels. At many of these facilities, routine monitoring stations have 

recorded the average photon dose in a general area or at the plant boundaries. At several DOE facilities, 

radioactive emissions from plant stacks have been known to significantly increase the “background” 

radiation levels on the plant site. In general the dose from increased background is rather low. 

6.1.1 During Operations 

Direct, external irradiation of people on site can occur from material in process, in storage, in transit, and 

from radioactive contamination on surfaces, in soil and water, and in plumes of radioactive material. 

Uranium metal emits beta particles (electrons) and to a lesser extent, photon radiation primarily in the 

form of bremsstrahlung X-rays but with some gamma emissions (e.g., the 186 keV photon from 235U). 

Neutron radiation (from spontaneous fission) is negligible, even for massive quantities of uranium (based 

on neutron production calculated by Sources4 (Wilson et al. 1999)). 

During operations, environmental doses were much smaller than those occurring in the actual refining 

portions of the facility, so these doses would be significant only for workers who spent little time in the 

high-dose portions of the facility. For these workers, the exposure scenario is spending eight hours per 

working day in an area with a low level of airborne contamination and a low level of contamination 

deposited on the floor. To estimate external doses resulting from these two pathways, dose factors are 

provided in Table 3.9 and Table 3.10. The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 

02 PC-1 (ORAUT 2006) to determine which of the tabulated organ dose rates should be used for any 

particular ICD code. 

Notice that the dose quantity calculated for this conversion factor is exposure, in units of milliRoentgen. 

For this particular conversion factor, 50.5% of the exposure is associated with photons with energies 

below 30 keV; 28.5% is from photons with energies between 30 keV and 250 keV, and the remaining 

21.0% is for photons with energies greater than 250 keV. 

The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine 

which of the tabulated organ dose rates should be used for any particular ICD code. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 23 

The environmental external dose received by a worker exposed to contaminated air and walking on 

contaminated ground can be found by multiplying the dose factors by the contamination levels, assuming 

a number of hours of exposure per workday. This analysis assumes that during plant operation, the 

worker away from the main operation area is exposed to a uranium concentration in air of 7 dpm/m3 

. 

This level corresponds to 0.1 MAC, and is consistent with general area air sampling results for positions 

on the factory floor away from uranium milling operations. (The results from metal working facilities are 

used for uranium refining operations until better data are found.) Air sampling results for three plants, 

American Machine and Foundry, Medart, and Allegheny Ludlum were examined, comparing air samples 

at the highest concentration areas to air samples at the lowest concentration areas during the same 

operation. For 11 different cases, the ratio of the low to high concentration ranged from 0.002 to 0.029, 

with all but two of those ratios below 0.01. Thus it would be reasonable to use a ratio of 0.01 for areas 

that are further away from the heavy operations, which non-operational areas of the facility would be. In 

similarity with the generic TBD for metal operations, we assume that a heavy operation would produce a 

daily-weighted air concentration of 10 MAC at the operator’s position, so ratioing this by 0.01 gives a 

value of 0.1 MAC or 7 dpm/m3

 in the non-operational areas of the plant. 

For the surface contamination, it is assumed that the uranium deposited on the floor with a deposition 

velocity of 0.00075 m/s for a period of one year without cleanup, then remained at that level of 

contamination for the duration of operations. This would be a contamination level of 1.65x105

 dpm/m2 . 

The worker exposure is estimated as daily doses for 40-, 44-, and 48-hour workweeks. Using these 

assumptions, the daily doses can be calculated, and they are presented in Table 6.1 

Table 6.1. Dose rate from submersion in uranium dust and standing on uranium contaminated surfaces. 

Hours worked per 

week 

Submersion Dose 

Rate(mrem/d) 

Contaminated 

Surface Dose 

Rate(mrem/d) 

Total Dose Rate 

(mrem/d) 

40 2.08E-07 3.15E-04 3.15E-04 

44 2.29E-07 3.46E-04 3.47E-04 

48 2.50E-07 3.78E-04 3.78E-04 

6.1.2 After Operations and Initial Decontamination and Before FUSRAP Remediation 

Workers at a facility that is no longer performing AWE work may also be inhaling uranium from residual 

contamination. If measurements of the contamination are not available for this time period, the intake 

presented in the previous section can be used as an estimate of the external doses from this dose pathway. 

6.1.3 After FUSRAP Remediation 

At the conclusion of remediation, exit surveys, such as FUSRAP reports can be used to estimate external 

dose rates. If no exit surveys are available, external doses can be estimated using the doses presented in 

Section 6.1.1. 

6.2	 Environmental Internal Dose Due to Inhalation and Ingestion of 

Radioactive Material 

At several DOE facilities, radioactive emissions from plant stacks have been known to significantly 

increase the “background” radiation levels on the plant site. The estimate of 7 dpm/m3

 used in Section 

6.1.1 can be assumed for the contamination level. Ingestion intakes were found using the equation 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 24 

IIMBA = 3.062 ×10−5 Ah as discussed in Section 8.5.3. Table 6.2 gives the intakes, both from inhalation 

and ingestion, for these conditions. 

6.2.1 During Operations 

Radioactive material aerosolized by various industrial processes in uranium refining can be released 

deliberately or inadvertently to the environment, producing uranium aerosols in the environs of the site. 

An estimate of the intake from the inhalation pathway can be estimated assuming an airborne 

contamination level, a breathing rate, and daily exposure period. The estimate of 1 dpm/m3

 can be 

assumed for the contamination level. A breathing rate of 9.6 m3

/d includes an exposure period of 8 hours 

per day. A conversion factor of 2.22 dpm/pCi must also be employed, to give a daily intake of 4.4 pCi/d 

of uranium. 

Table 6.2. Environmental daily intakes to workers in non-operational areas of the plant. 

Hours worked per 

week 

Daily Intake from 

Inhalation 

(pCi/d) 

Daily Intake from 

Ingestion 

(pCi/d) 

40 20.7 0.429 

44 22.8 0.472 

48 24.9 0.514 

6.2.2 After Operations and Initial Decontamination and Before FUSRAP Remediation 

Workers at a facility that is no longer performing AWE work may also be inhaling uranium from residual 

contamination. If measurements of the contamination are not available for this time period, the intake 

presented in the previous section can be used as an estimate of the internal doses from this dose pathway. 

6.2.3 After FUSRAP Remediation 

At the conclusion of remediation, exit surveys should be available to form a basis for intakes. If no exit 

surveys are available, intakes can be estimated as identical to that given in Section 6.2.1. 

6.3 Summary of Environmental Irradiation 

Environmental doses and intakes for generic sites are summarized in Table 6.3. Each site’s stages 

correspond to dates given in the Appendix for that site. (Site stages are described in Section 4.11). 

Intakes must be adjusted for a 2,400, a 2,200, or a 2,000-hour work year by date as specified in Section 

3.10 of Battelle-TIB-5000 (Strom 2006). 

Default environmental dose assumptions for a generic uranium refining site during Stage 1 are shown in 

Table 6.3. All doses are lognormally distributed. 

Table 6.3. Default environmental dose assumptions for a generic uranium refining site during Stage 1. 

All doses are lognormally distributed with a GSD of 5. 

Type of exposure; radiation type, 

energy, quantity, units, [geometry] 

48-h work week 44-h work week 40-h work week 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 25 

External irradiation, photons, 30<E≤250 

keV, Hp(10) rem/y, AP geometry 

3.78E-4 mrem/d 3.47E-4 mrem/d 3.15E-4 mrem/d 

365-day average daily chronic inhalation 

intake of Type M or Type S 234U, pCi/d 24.9 pCi/d 22.8 pCi/d 20.7 pCi/d 

365-day average daily chronic ingestion 

intake of Type M or Type S 234U, pCi/d 0.514 pCi/d 0.472 pCi/d 0.429 pCi/d 

Table 6.4. Default environmental dose assumptions for a generic uranium refining site during Stage 2. 

All doses are lognormally distributed. 

Type of exposure; radiation type, 

energy, quantity, units, [geometry] 

48-h work week 44-h work week 40-h work week 

External irradiation, photons, 30<E≤250 

keV, Hp(10) rem/y, AP geometry 

NA NA 3.15E-4 mrem/d 

365-day average daily chronic inhalation 

intake of Type M or Type S 234U, pCi/d 

NA NA 20.7 pCi/d 

365-day average daily chronic ingestion 

intake of Type M or Type S 234U, pCi/d 

NA NA 0.429 pCi/d 

Table 6.5. Default environmental dose assumptions for a generic uranium refining site during Stage 3. 

All doses are lognormally distributed. 

Type of exposure; radiation type, 

energy, quantity, units, [geometry] 

48-h work week 44-h work week 40-h work week 

External irradiation, photons, 30<E≤250 

keV, Hp(10) rem/y, AP geometry 

NA NA 3.15E-4 mrem/d 

365-day average daily chronic inhalation 

intake of Type M or Type S 234U, pCi/d 

NA NA 20.7 pCi/d 

365-day average daily chronic ingestion 

intake of Type M or Type S 234U, pCi/d 

NA NA 0.429 pCi/d 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 26 

7.0 Occupational External Dose
 

Occupational external dose can arise from three sources; direct radiation from radioactive material in 

storage or in process, airborne radioactive material, and radioactive material on the surfaces of objects in 

a work area. 

If the workers wore radiation dosimeters and that data are available then the dosimetry data should be 

used for all dose estimates. 

Section 6.0 (Occupational Internal Dose) discusses the air concentrations of uranium dusts due to various 

processes. In some cases the work areas were quite dusty. High air concentrations would also result in 

high contamination levels. It is not known how often work areas were cleaned of dust. We assume that 

areas were cleaned on a weekly basis. 

Film badge usage at AWE sites varied from plant to plant and over time. When film badge data for 

workers are available, these readings are the preferred indicator of the worker’s whole body dose. In 

some cases there may be coworker dosimetry results available that can be used to estimate an unbadged 

worker’s dose. When film badges were not issued, the worker’s dose must be determined from plant 

conditions. This section contains information that can be used to estimate external doses to workers when 

no dosimetry data are available. The external dose rates presented in this section are those for typical 

conditions in the workplace described in this document. The appendices to this document will present 

site-specific information to provide more appropriate estimates where available. If site information is 

insufficient to estimate a dose to a worker, the information in this section can be used. 

7.1 Radiation Doses 

Workers at refining facilities may have been exposed to three types of radiation; beta radiation, gamma 

radiation, and neutron radiation. Detailed information on the dosimetry of uranium was described in 

Chapter 3. 

7.1.1 Submersion in Contaminated Air 

When workers are enveloped in a cloud of radioactive dust, they will receive a small amount of external 

dose. External dose rates from uranium and its radioactive progeny are shown in Table 3.9. 

Notice that the dose quantity calculated for this conversion factor is exposure, in units of milliRoentgen. 

For this particular conversion factor, 50.5% of the exposure is associated with photons with energies 

below 30 keV; 28.5% is from photons with energies between 30 keV and 250 keV, and the remaining 

21.0% is for photons with energies greater than 250 keV. 

The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine 

which of the tabulated organ dose rates should be used for any particular ICD code. 

7.1.2 Exposures from Contaminated Surfaces 

When workers are working on a contaminated surface, they will receive a small amount of external dose. 

External dose rates from uranium and its radioactive progeny are shown in Table 3.9. 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 27 

The doses were calculated using the computer code MicroShield version 6.02 (Grove Engineering 2003). 

The calculated dose rates are for natural uranium and include the dose contribution from the radioactive 

progeny of 238U, 235U, and 234U. In accordance with (ORAUT 2005c) 100 days of radioactive progeny 

ingrowth was assumed for these calculations. 

Notice that the dose quantity calculated for this conversion factor is exposure, in units of milliRoentgen. 

For this particular conversion factor, 50.5% of the exposure is associated with photons with energies 

below 30 keV; 28.5% is from photons with energies between 30 keV and 250 keV, and the remaining 

21.0% is for photons with energies greater than 250 keV. 

The dose reconstructor should refer to ORAUT-OTIB-0005, Rev. 02 PC-1 (ORAUT 2006) to determine 

which of the tabulated organ dose rates should be used for any particular ICD code. 

The quantity of uranium on the floor surface can be obtained from floor survey (wipe) sample 

measurements. Most survey samples were based on a 100 cm2

 sample rather than a 1 meter squared 

sample, in this case the dose values in Table 3.10 should be divided by 10,000. 

When measured floor contamination rates are not available the contamination on the floor may be 

estimated from measured air concentrations. The floor activity may be computed from the air 

concentrations following the method of ORAUT (2005c) (ORAUT-OTIB -0004, Rev 3, p 15). The level 

of surface contamination was determined by first calculating a terminal settling velocity for 5-μm activity 

mean aerodynamic diameter (AMAD) particles. The calculated terminal settling velocity was 0.00075 

meters per second. It was assumed that the surface contamination level was due to 365 days of constant 

deposition from the constant air concentration to give a deposition factor of 2.37E4 meters. The floor 

contamination level is then estimated as Floor Concentration (dpm/m2

) = Air Concentration (dpm/m3) × 

2.37E4 meters. This method calculates the surface contamination over 1 m2

 and the dose factors in Table 

6.1 should be used. 

7.2 Process Specific Dose Rates 

Two references were used to generate the process specific dose rates shown in this section. The first was 

the dose rate information presented by Christofano and Harris (1960). Their data were used primarily to 

estimate the dose rate due to submersion in a dust cloud and due to contamination on the floor. This 

paper has extensive air concentration data. The air concentration data are summarized in chapter 8. 

External dose rates were extracted from the TBD for Mallinckrodt (ORAUT-TKBS-0005) (ORAUT, 

2005a). To compute the daily doses from external sources of radiation it was necessary to estimate the 

exposure duration. This was accomplished by a crude time-and-motion estimate that was primarily an 

educated opinion of how much time was spent near the particular source in a day. 

In the sections that follow, the external dose rate and durations are tabulated. The tabulated values are in 

terms of mR h-1, the conversions to mrem h-1 are performed by the dose reconstruction spreadsheet. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Table 7.1. Process Specific Dose Rates for Uranium Refining at AWE Sites 

Task Duration 

(h/d) 

Penetrating 

(mR/h) 

Non-penetrating 

(hands) 

(mrem/h) 

Non-penetrating 

(other) 

(mrem/h) 

Ore Digestion 

High grade pitchblende ore, 45" 6 6.1 TKBS-0005 [High grade pitchblende ore, 6" 2 66 TKBS-0005 [2

High grade pitchblende ore, contact 2 100 TKBS-0005 [Total 36.6 132 200 

Solvent Extraction 

ether extraction tanks 3 3.1 TKBS-0005 [2

residue dryer 3 75 TKBS-0005 [2

Total 9.3 0 225 

Boildown and Denitration 

ether extraction tanks 3 3.1 TKBS-0005 [2

55 gal drum, 100 day, 1 m 8 3.10 Table 7.2 

55 gal drum, 100 day, 1 cm 2 0.28 Table 7.2 55 gal drum, 100 day, 30 cm 2 4.50 Table 7.2 Total 11.55 9.00 2.65 

Oxide Reduction 

55 gal drum, 100 day, 1 m 8 0.28 Table 7.2 

Handling U material 3 233.00 20.80 Table 3.4 Total 2.25 699.00 62.40 

Hydrofluorination 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Task Duration 

(h/d) 

Penetrating 

(mR/h) 

Non-penetrating 

(hands) 

(mrem/h) 

Non-penetrating 

(other) 

(mrem/h) 

Handling U material 3 233.00 20.80 Table 1.3 

55 gal drum, 100 day, 1 m 8 0.28 Table 5.6 Total 2.25 699.00 62.40 

Reduction to Metal 

Metal (average of all task dose rates) 

8 

13.5 TKBS-0005 [2

loading bomb 2 53 53 TKBS-0005 [2

Contact with unchipped derby 2 116 TKBS-0005 [2

chipping - 1 ft 2 42 TKBS-0005 [2

Total 108 338 190 

Recasting 

Contact with cleaned crucible lid 1 155 TKBS-0005 [

billets 1 180 180 TKBS-0005 [2

Recasting: chest high cleaning top 1 135 TKBS-0005 [Furnace tending 6 665 665 TKBS-0005 [2Total 4170.00 155.00 4305.00 

Fluorination 

55 gal drum, 100 day, 1 m 8 0.28 Table 7.2 

Handling U material 2 233.00 20.80 Table 3.4 Total 2.25 466.00 41.60 

Scrap Recovery Handling U material 4 233 20.8 Table 3.4 

55 gal drum, 100 day, 1 m 8 0.28 Table 7.2 

Total 2.25 932.00 83.20 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Task Duration 

(h/d) 

Penetrating 

(mR/h) 

Non-penetrating 

(hands) 

(mrem/h) 

Non-penetrating 

(other) 

(mrem/h) 

Drum Loading 

55 gal drum, 100 day, 1 m 8 0.28 Table 7.2 

High grade pitchblende ore, 45" 1 6.1 TKBS-0005 [2

Total 8.35 0.00 0.00 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 31 

7.3 Drum Doses 

Uranium compounds were kept in drums of various sizes. The dose rates from drums were calculated 

using MicroShield Version 5.01 for two different situations. The first set of calculations assumed that the 

time of decay was 100 days, which allows the ingrowth of uranium progeny which will increase the dose 

rate. The contents of the drums were modeled as soil at a density of 1.6 (FGR-12 (Eckerman and Ryman 

1993)). This combination of density and elemental composition will result in exaggerated dose rates. 

The calculations did not account for Bremsstrahlung that may have been generated by the interactions of 

beta particles with the contents of the drum. Calculations performed by others (Glover 2006, Anderson 

and Hertel 2005) indicate that the dose rate due to Bremsstrahlung may be equal to the photon dose rate. 

Therefore, the values shown in Table 7.2 are twice the dose rate that was calculated for photons alone. 

The dose rates shown here should be combined with the dose conversion factors for the AP irradiation 

geometry. The dose rates from drums that contain pitchblend were derived from (ORAUT-TKBS-0005) 

(ORAUT, 2005). 

 Table 7.2 Dose rates from drums of uranium compounds 

Drum Size (gal) 

Dose Rates (mR/h) 

1 cm 10 cm 30 cm 100 cm 

100 day decay 

5 3.7 1.4 0.4 0.1 

30 4.4 2.5 1.1 0.2 

55 4.5 2.8 1.3 0.3 

Equilibrium (pitchblende) 

5 80.52 29.85 8.64 1.25 

30 107.68 59.25 25.03 4.72 

55 112.88 68.12 32.11 6.77 

7.4 Summary of External Doses Received by Workers During Operation 

•	 The daily doses received by workers at a plant that handled uranium, according to the five pathways 

described in this section, are given in Table 7.3. Three of the dose pathways are mR/h and are 

converted to dose by the dose assignment spreadsheet. Dose reconstructors should assume that the 

entire photon dose from uranium is due to photons having the following energy bin structure: 50.5% 

of the exposure is associated with photons with energies below 30 keV; 28.5% is from photons with 

energies between 30 keV and 250 keV, and the remaining 21.0% is for photons with energies greater 

than 250 keV. 

 The “hands and forearms” and “other skin” pathways are doses at a skin depth of 0.07 mg/cm2

, caused 

by electrons with energies above 15 keV. 

The daily dose rates shown in Table 7.3 have been adjusted so that the daily dose rates is for calendar 

days rather than working days. The dose reconstructor has to estimate only the number of calendar days 

that elapsed between the start and end of AWE employment to use this table. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Table 7.3. Summary external dose rates for uranium refining facilities Operation Job Title Dose Pathway Geometry 

Daily dose parameters for external do

penetrating radiation and mrem fo

48-h week 44-h week 

Median GSD Median GSD 

Ore Digestion 

Operator 

Submersion ISO 4.58E-07 5.7 4.20E-07 5.7 

Contaminated Floor Underfoot Plane 2.47E-01 5.7 2.27E-01 5.7 

Material handled A-P 8.59E+00 5.0 7.88E+00 5.0 

Non-penetrating hands n/a 3.10E+01 5.0 2.84E+01 5.0 

Non-penetrating, other skin n/a 4.69E+01 5.0 4.30E+01 5.0 

General Laborer 

Submersion ISO 2.29E-07 5.7 2.10E-07 5.7 

Contaminated Floor Underfoot Plane 1.24E-01 5.7 1.13E-01 5.7 

Material handled A-P 4.30E+00 5.0 3.94E+00 5.0 

Non-penetrating hands n/a 1.55E+01 5.0 1.42E+01 5.0 

Non-penetrating, other skin n/a 2.35E+01 5.0 2.15E+01 5.0 

Supervisor 

Submersion ISO 1.14E-07 5.7 1.05E-07 5.7 

Contaminated Floor Underfoot Plane 6.18E-02 5.7 5.67E-02 5.7 

Material handled A-P 2.15E+00 5.0 1.97E+00 5.0 

Non-penetrating hands n/a 7.75E+00 5.0 7.10E+00 5.0 

Non-penetrating, other skin n/a 1.17E+01 5.0 1.08E+01 5.0 

Clerical 

Submersion ISO 1.14E-08 5.7 1.05E-08 5.7 

Contaminated Floor Underfoot Plane 6.18E-03 5.7 5.67E-03 5.7 

Material handled A-P 2.15E-01 5.0 1.97E-01 5.0 

Non-penetrating hands n/a 7.75E-01 5.0 7.10E-01 5.0 

Non-penetrating, other skin n/a 1.17E+00 5.0 1.08E+00 5.0 

Solvent Extraction 

Operator 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 2.18E+00 5.0 2.00E+00 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 5.28E+01 5.0 4.84E+01 5.0 

General Laborer 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 1.09E+00 5.0 1.00E+00 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 2.64E+01 5.0 2.42E+01 5.0 

Supervisor 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 5.46E-01 5.0 5.00E-01 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 1.32E+01 5.0 1.21E+01 5.0 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Operation Job Title Dose Pathway Geometry 

Daily dose parameters for external do

penetrating radiation and mrem fo

48-h week 44-h week 

Median GSD Median GSD 

Clerical 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 5.46E-02 5.0 5.00E-02 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 1.32E+00 5.0 1.21E+00 5.0 

Boildown and Denitration 

Operator 

Submersion ISO 8.73E-07 2.3 8.00E-07 2.3 

Contaminated Floor Underfoot Plane 4.72E-01 2.3 4.32E-01 2.3 

Material handled A-P 2.71E+00 5.0 2.49E+00 5.0 

Non-penetrating hands n/a 2.11E+00 5.0 1.94E+00 5.0 

Non-penetrating, other skin n/a 6.22E-01 5.0 5.70E-01 5.0 

General Laborer 

Submersion ISO 4.36E-07 2.3 4.00E-07 2.3 

Contaminated Floor Underfoot Plane 2.36E-01 2.3 2.16E-01 2.3 

Material handled A-P 1.36E+00 5.0 1.24E+00 5.0 

Non-penetrating hands n/a 1.06E+00 5.0 9.68E-01 5.0 

Non-penetrating, other skin n/a 3.11E-01 5.0 2.85E-01 5.0 

Supervisor 

Submersion ISO 2.18E-07 2.3 2.00E-07 2.3 

Contaminated Floor Underfoot Plane 1.18E-01 2.3 1.08E-01 2.3 

Material handled A-P 6.78E-01 5.0 6.21E-01 5.0 

Non-penetrating hands n/a 5.28E-01 5.0 4.84E-01 5.0 

Non-penetrating, other skin n/a 1.56E-01 5.0 1.43E-01 5.0 

Clerical 

Submersion ISO 2.18E-08 2.3 2.00E-08 2.3 

Contaminated Floor Underfoot Plane 1.18E-02 2.3 1.08E-02 2.3 

Material handled A-P 6.78E-02 5.0 6.21E-02 5.0 

Non-penetrating hands n/a 5.28E-02 5.0 4.84E-02 5.0 

Non-penetrating, other skin n/a 1.56E-02 5.0 1.43E-02 5.0 

Operator 

Submersion ISO 2.50E-06 3.5 2.30E-06 3.5 

Contaminated Floor Underfoot Plane 1.35E+00 3.5 1.24E+00 3.5 

Material handled A-P 5.28E-01 5.0 4.84E-01 5.0 

Non-penetrating hands n/a 1.64E+02 5.0 1.50E+02 5.0 

Non-penetrating, other skin n/a 1.46E+01 5.0 1.34E+01 5.0 

General Laborer 

Submersion ISO 1.25E-06 3.5 1.15E-06 3.5 

Contaminated Floor Underfoot Plane 6.77E-01 3.5 6.21E-01 3.5 

Material handled A-P 2.64E-01 5.0 2.42E-01 5.0 

Non-penetrating hands n/a 8.20E+01 5.0 7.52E+01 5.0 

Non-penetrating, other skin n/a 7.32E+00 5.0 6.71E+00 5.0 

Supervisor 

Submersion ISO 6.26E-07 3.5 5.74E-07 3.5 

Contaminated Floor Underfoot Plane 3.38E-01 3.5 3.10E-01 3.5 

Material handled A-P 1.32E-01 5.0 1.21E-01 5.0 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Operation Oxide Reduction 

Job Title Dose Pathway Geometry 

Daily dose parameters for external do

penetrating radiation and mrem fo

48-h week 44-h week 

Median GSD Median GSD 

Non-penetrating hands n/a 4.10E+01 5.0 3.76E+01 5.0 

Non-penetrating, other skin n/a 3.66E+00 5.0 3.36E+00 5.0 

Clerical 

Submersion ISO 6.26E-08 3.5 5.74E-08 3.5 

Contaminated Floor Underfoot Plane 3.38E-02 3.5 3.10E-02 3.5 

Material handled A-P 1.32E-02 5.0 1.21E-02 5.0 

Non-penetrating hands n/a 4.10E+00 5.0 3.76E+00 5.0 

Non-penetrating, other skin n/a 3.66E-01 5.0 3.36E-01 5.0 

Hydrofluorination 

Operator 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 5.28E-01 5.0 4.84E-01 5.0 

Non-penetrating hands n/a 1.64E+02 5.0 1.50E+02 5.0 

Non-penetrating, other skin n/a 1.46E+01 5.0 1.34E+01 5.0 

General Laborer 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 2.64E-01 5.0 2.42E-01 5.0 

Non-penetrating hands n/a 8.20E+01 5.0 7.52E+01 5.0 

Non-penetrating, other skin n/a 7.32E+00 5.0 6.71E+00 5.0 

Supervisor 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 1.32E-01 5.0 1.21E-01 5.0 

Non-penetrating hands n/a 4.10E+01 5.0 3.76E+01 5.0 

Non-penetrating, other skin n/a 3.66E+00 5.0 3.36E+00 5.0 

Clerical 

Submersion ISO 0.00E+00 1.0 0.00E+00 1.0 

Contaminated Floor Underfoot Plane 0.00E+00 1.0 0.00E+00 1.0 

Material handled A-P 1.32E-02 5.0 1.21E-02 5.0 

Non-penetrating hands n/a 4.10E+00 5.0 3.76E+00 5.0 

Non-penetrating, other skin n/a 3.66E-01 5.0 3.36E-01 5.0 

Operator 

Submersion ISO 2.10E-06 4.5 1.92E-06 4.5 

Contaminated Floor Underfoot Plane 1.13E+00 4.5 1.04E+00 4.5 

Material handled A-P 2.54E+01 5.0 2.32E+01 5.0 

Non-penetrating hands n/a 7.93E+01 5.0 7.27E+01 5.0 

Non-penetrating, other skin n/a 4.46E+01 5.0 4.09E+01 5.0 

General Laborer 

Submersion ISO 1.05E-06 4.5 9.62E-07 4.5 

Contaminated Floor Underfoot Plane 5.67E-01 4.5 5.20E-01 4.5 

Material handled A-P 1.27E+01 5.0 1.16E+01 5.0 

Non-penetrating hands n/a 3.97E+01 5.0 3.64E+01 5.0 

Non-penetrating, other skin n/a 2.23E+01 5.0 2.04E+01 5.0 

Submersion ISO 5.25E-07 4.5 4.81E-07 4.5 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Operation Metal Reduction 

Job Title Dose Pathway Geometry 

Daily dose parameters for external do

penetrating radiation and mrem fo

48-h week 44-h week 

Median GSD Median GSD 

Contaminated Floor Underfoot Plane 2.83E-01 4.5 2.60E-01 4.5 

Material handled A-P 6.34E+00 5.0 5.81E+00 5.0 

Non-penetrating hands n/a 1.98E+01 5.0 1.82E+01 5.0 

Non-penetrating, other skin n/a 1.11E+01 5.0 1.02E+01 5.0 

Clerical 

Submersion ISO 5.25E-08 4.5 4.81E-08 4.5 

Contaminated Floor Underfoot Plane 2.83E-02 4.5 2.60E-02 4.5 

Material handled A-P 6.34E-01 5.0 5.81E-01 5.0 

Non-penetrating hands n/a 1.98E+00 5.0 1.82E+00 5.0 

Non-penetrating, other skin n/a 1.11E+00 5.0 1.02E+00 5.0 

Recasting 

Operator 

Submersion ISO 1.86E-06 2.1 1.71E-06 2.1 

Contaminated Floor Underfoot Plane 1.01E+00 2.1 9.22E-01 2.1 

Material handled A-P 9.79E+02 5.0 8.97E+02 5.0 

Non-penetrating hands n/a 3.64E+01 5.0 3.34E+01 5.0 

Non-penetrating, other skin n/a 1.01E+03 5.0 9.26E+02 5.0 

General Laborer 

Submersion ISO 9.30E-07 2.1 8.53E-07 2.1 

Contaminated Floor Underfoot Plane 5.03E-01 2.1 4.61E-01 2.1 

Material handled A-P 4.89E+02 5.0 4.49E+02 5.0 

Non-penetrating hands n/a 1.82E+01 5.0 1.67E+01 5.0 

Non-penetrating, other skin n/a 5.05E+02 5.0 4.63E+02 5.0 

Supervisor 

Submersion ISO 4.65E-07 2.1 4.26E-07 2.1 

Contaminated Floor Underfoot Plane 2.51E-01 2.1 2.30E-01 2.1 

Material handled A-P 2.45E+02 5.0 2.24E+02 5.0 

Non-penetrating hands n/a 9.10E+00 5.0 8.34E+00 5.0 

Non-penetrating, other skin n/a 2.53E+02 5.0 2.32E+02 5.0 

Clerical 

Submersion ISO 4.65E-08 2.1 4.26E-08 2.1 

Contaminated Floor Underfoot Plane 2.51E-02 2.1 2.30E-02 2.1 

Material handled A-P 2.45E+01 5.0 2.24E+01 5.0 

Non-penetrating hands n/a 9.10E-01 5.0 8.34E-01 5.0 

Non-penetrating, other skin n/a 2.53E+01 5.0 2.32E+01 5.0 

Operator 

Submersion ISO 2.27E-06 3.3 2.08E-06 3.3 

Contaminated Floor Underfoot Plane 1.23E+00 3.3 1.13E+00 3.3 

Material handled A-P 5.28E-01 5.0 4.84E-01 5.0 

Non-penetrating hands n/a 1.09E+02 5.0 1.00E+02 5.0 

Non-penetrating, other skin n/a 9.76E+00 5.0 8.95E+00 5.0 

General 

Submersion ISO 1.14E-06 3.3 1.04E-06 3.3 

Contaminated Floor Underfoot Plane 6.15E-01 3.3 5.63E-01 3.3 

Material handled A-P 2.64E-01 5.0 2.42E-01 5.0 

Non-penetrating hands n/a 5.47E+01 5.0 5.01E+01 5.0 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Operation Fluorination 

Job Title 

Laborer 

Dose Pathway Geometry 

Daily dose parameters for external do

penetrating radiation and mrem fo

48-h week 44-h week 

Median GSD Median GSD 

Non-penetrating, other skin n/a 4.88E+00 5.0 4.48E+00 5.0 

Supervisor 

Submersion ISO 5.69E-07 3.3 5.21E-07 3.3 

Contaminated Floor Underfoot Plane 3.07E-01 3.3 2.82E-01 3.3 

Material handled A-P 1.32E-01 5.0 1.21E-01 5.0 

Non-penetrating hands n/a 2.73E+01 5.0 2.51E+01 5.0 

Non-penetrating, other skin n/a 2.44E+00 5.0 2.24E+00 5.0 

Clerical 

Submersion ISO 5.69E-08 3.3 5.21E-08 3.3 

Contaminated Floor Underfoot Plane 3.07E-02 3.3 2.82E-02 3.3 

Material handled A-P 1.32E-02 5.0 1.21E-02 5.0 

Non-penetrating hands n/a 2.73E+00 5.0 2.51E+00 5.0 

Non-penetrating, other skin n/a 2.44E-01 5.0 2.24E-01 5.0 

Scrap Recovery 

Operator 

Submersion ISO 6.96E-07 5.0 6.38E-07 5.0 

Contaminated Floor Underfoot Plane 3.76E-01 5.0 3.45E-01 5.0 

Material handled A-P 5.28E-01 5.0 4.84E-01 5.0 

Non-penetrating hands n/a 2.19E+02 5.0 2.01E+02 5.0 

Non-penetrating, other skin n/a 1.95E+01 5.0 1.79E+01 5.0 

General Laborer 

Submersion ISO 3.48E-07 5.0 3.19E-07 5.0 

Contaminated Floor Underfoot Plane 1.88E-01 5.0 1.72E-01 5.0 

Material handled A-P 2.64E-01 5.0 2.42E-01 5.0 

Non-penetrating hands n/a 1.09E+02 5.0 1.00E+02 5.0 

Non-penetrating, other skin n/a 9.76E+00 5.0 8.95E+00 5.0 

Supervisor 

Submersion ISO 1.74E-07 5.0 1.59E-07 5.0 

Contaminated Floor Underfoot Plane 9.40E-02 5.0 8.62E-02 5.0 

Material handled A-P 1.32E-01 5.0 1.21E-01 5.0 

Non-penetrating hands n/a 5.47E+01 5.0 5.01E+01 5.0 

Non-penetrating, other skin n/a 4.88E+00 5.0 4.48E+00 5.0 

Clerical 

Submersion ISO 1.74E-08 5.0 1.59E-08 5.0 

Contaminated Floor Underfoot Plane 9.40E-03 5.0 8.62E-03 5.0 

Material handled A-P 1.32E-02 5.0 1.21E-02 5.0 

Non-penetrating hands n/a 5.47E+00 5.0 5.01E+00 5.0 

Non-penetrating, other skin n/a 4.88E-01 5.0 4.48E-01 5.0 

Operator 

Submersion ISO 3.56E-06 2.5 3.27E-06 2.5 

Contaminated Floor Underfoot Plane 1.93E+00 2.5 1.77E+00 2.5 

Material handled A-P 1.96E+00 5.0 1.80E+00 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 0.00E+00 5.0 0.00E+00 5.0 

Submersion ISO 1.78E-06 2.5 1.63E-06 2.5 

Contaminated Floor Underfoot Plane 9.63E-01 2.5 8.83E-01 2.5 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 

Operation Drum Loading 

Job Title 

General Laborer 

Dose Pathway Geometry 

Daily dose parameters for external do

penetrating radiation and mrem fo

48-h week 44-h week 

Median GSD Median GSD 

Material handled A-P 9.80E-01 5.0 8.98E-01 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 0.00E+00 5.0 0.00E+00 5.0 

Supervisor 

Submersion ISO 8.91E-07 2.5 8.16E-07 2.5 

Contaminated Floor Underfoot Plane 4.81E-01 2.5 4.41E-01 2.5 

Material handled A-P 4.90E-01 5.0 4.49E-01 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 0.00E+00 5.0 0.00E+00 5.0 

Clerical 

Submersion ISO 8.91E-08 2.5 8.16E-08 2.5 

Contaminated Floor Underfoot Plane 4.81E-02 2.5 4.41E-02 2.5 

Material handled A-P 4.90E-02 5.0 4.49E-02 5.0 

Non-penetrating hands n/a 0.00E+00 5.0 0.00E+00 5.0 

Non-penetrating, other skin n/a 0.00E+00 5.0 0.00E+00 5.0 
 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 38 

8.0 Occupational Internal Dose 

In this section the internal dosimetry parameters are discussed. For each process, there is a table of 

internal dosimetry parameters – particle size and solubility information. The tabulated values are for 

reference only, the ICRP (1994) default values are to be used for all internal dose calculations (see 

Section 3.5.) Also included for each process is a table of default dust air concentrations that can be used 

to estimate intakes when site specific information is lacking. 

8.1 Uranium Bioassay 

Bioassays were performed on workers at many sites, and urinalysis was the most effective method of 

determining an individual worker’s uranium intake. When bioassay data are available for individual 

workers, this should be the primary source of dose reconstruction for internal exposure. 

8.2 Process Specific Uranium Air Sampling Data 

The data shown in this section are from Chrostofano and Harris (1960). 

8.2.1 Ore Digestion 

Internal dosimetry parameters for dusts in ore digestion are shown in Table 8.1. 

Table 8.1. Internal dosimetry parameters for ore digestion 

Time Period All times 

Particle Parameters 5 micron AMAD; 8.3 g/cm3 

Input Material U3O8 

Output Material U nitrate solution 

Default air concentrations of U3O8 are shown in Table 8.2. 

Table 8.2 Dust concentrations during digestion 

Operations 

Dust Concentration, d/min/m3 

Ore d/m/m3 Concentrate d/m/m3 

min max average min max average 

BZ Reaming ore chute 350 8,000 1,000 

BZ drum dumping 

 Uncontrolled 2,500 1,000 6,000 2,400

 Ventilated 0 220 90

 Remote 6 44 30 

BZ lidding and delidding 

drums 

600 1,700 1,200 

GA Digest area 6 330 150 0 75 30 

GA Ore room 90 2,600 1,000 

Average daily exposure 7 350 110 17 100 40 

The lognormal distributions parameters for digestion are shown in Table 8.3. 
 

 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 39 

Table 8.3. Lognormal Distributions of dust concentrations during digestion 

Operations 

Dust Concentration, d/m/m3 

Ore d/m/m3 Concentrate d/m/m3 

Median GSD Median GSD 

BZ Reaming ore chute 1673 0.357 

BZ drum dumping 

 Uncontrolled 685 5 2449 0.960 

Ventilated 25 5 

 Remote 16 3.409 

BZ lidding and delidding drums 1010 1.412 

GA Digest area 44 11.364 8 5 

GA Ore room 484 4.274 

Average daily exposure 49 4.939 41 0.941 

8.2.2 Solvent Extraction 

Internal dosimetry parameters for dusts during solvent extraction are shown in Table 8.4. 

Table 8.4. Internal dosimetry parameters for solvent extraction 

Time Period All times 

Particle Parameters 5 micron AMAD; 8.3 g/cm3 

Input Material U nitrate solution8 

Output Material U-nitrate solution 

The airborne material is assumed to be U3O8. The air concentration table for solvent extraction is in 

development. 

8.2.3 Boildown and Denitration 

Internal dosimetry parameters for dusts of boildown and denitration are shown in Table 8.5. 

Table 8.5. Denitration area information and parameters 

Time Period All times 

Particle Parameters 5 micron AMAD; 7.3 g/cm3 

Input Material U nitrate solution 

Output Material UO3 

Default air concentrations of dusts during denitration are shown in Table 8.6. 

Table 8.6. Dust Concentrations During Denitration 

Operations 

Dust Concentration, d/min/m3 

min max average 

GA denitration area 7 750 140 

GA UO3 packaging area 15 160 63 

BZ Scrape and shovel UO3 from pots 29,000 82,000 48,000 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 40 

BZ Pneumatically empty UO3 pot 70 1,800 560 

BZ Pour UNH liquor into pots 45 63 56 

Average weighted exposure up to 1949 4,200 32,000 15,000 

Average weighted exposure since 1949 31 234 130 

Table 8.7 shows lognormal distributions values for denitration operations. 

Table 8.7 Lognormal Distributions of Dust Concentrations during denitration 

Operations 

Dust Concentration, d/min/m3 

Median GSD 

GA denitration area 72 3.733 

GA UO3 packaging area 49 1.654 

BZ Scrape and shovel UO3 from pots 48765 0.969 

BZ Pneumatically empty UO3 pot 355 2.489 

BZ Pour UNH liquor into pots 53 1.106 

Average weighted exposure up to 1949 11593 1.674 

Average weighted exposure since 1949 85 2.330 

8.2.4 Oxide Reduction 

Internal dosimetry parameters for dusts found in oxide reduction are found in Table 8.8. 

Table 8.8. Oxide reduction area information and parameters 

Time Period All times 

Particle Parameters 5 micron AMAD; 7.3 g/cm3

 (UO3), 10.96 g/cm3 (UO2) 

Input Material UO3 

Output Material UO2 

The air concentrations of uranium for the three methods of UO3 reduction are shown in Table 8.9. Note 

that the introduction of the continuous flow method (method 3) resulted in a significant reduction in the 

air concentration of powder to which the workers were exposed. 

Table 8.9. Dust concentrations during oxide reduction 

Operations 

Dust Concentration, d/min/m3 

min max average 

BZ Load UO3 trays 3,200 61,000 26,000 

BZ Load UO3 horizontal reactor 70 240 170 

BZ Load UO3 multiple hearth furnace 160 2,700 1,400 

BZ Unload UO2 from trays 4,000 125,000 60,000 

BZ Package UO2 from trays 25,000 115,000 80,000 

BZ Package UO2 from multiple hearth 16,500 35,000 23,000 
 

 

 

 

 

 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 41 

GA Tray furnace area 78 3,300 1,800 

GA Multiple hearth furnace area 245 440 340 

GA Horizontal reactor area 33 120 60 

DWA Tray furnace operations 9,800 32,000 20,000 

DWA Multiple hearth operations 31 4,200 700 

DWA Horizontal reactor operations 45 234 140 

Lognormal distribution parameters for oxide reductions are shown as Table 8.10. 

Table 8.10. Lognormal Distributions of Dust Concentrations During Oxide Reduction. 

Operation 

Dust Concentration, d/min/m3 

Median GSD 

BZ Load UO3 trays 13971 3.463 

BZ Load UO3 horizontal reactor 130 1.720 

BZ Load UO3 multiple hearth furnace 657 4.537 

BZ Unload UO2 from trays 22361 7.2 

BZ Package UO2 from trays 53619 2.226 

BZ Package UO2 from multiple hearth 24031 0.916 

GA Tray furnace area 507 12.578 

GA Multiple hearth furnace area 328 1.072 

GA Horizontal reactor area 63 0.909 

DWA Tray furnace operations 17709 1.276 

DWA Multiple hearth operations 361 3.763 

DWA Horizontal reactor operations 103 1.861 

8.2.5 Hydrofluorination 

Internal dosimetry parameters for hydrofluorination are shown in Table 8.11. 

Table 8.11. Hydrofluorination area information and parameters 

Time Period All times 

Particle Parameters 5 micron AMAD; 10.96 g/cm3

 (UO2), 6.7 g/cm3 (UF4) 

Input Material UO2 

Output Material UF4 

As with oxide reduction, more than one method was used for this process. In similarity to the Oxide 

reduction process, the first methods involved manually loading the reactant onto a tray, the tray was 

placed in a reactor, and after the reaction was complete the product was manually scooped out of the tray. 

The later method was a continuous process where the UO2 was exposed to HF in a counter current 

process similar to that for oxide reduction. Default dust air concentrations for hydrofluorination are 

shown in Table 8.12. 

Table 8.12. Dust concentrations during hydrofluorination 

Dust Concentration, d/min/m3 
 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 42 

Operations 

Trays Reactor 

min max average min max average 

Weighted Average Exposures 

UO2 loaders 260 8,900 3,300 - - -

Furnace Operator 50 1,200 500 30 40 30

 UF4 package 110 4,400 1,300 1 57 20 

Breathing Zone operations 

Loading UO2 730 15,000 5,000 - - -

Unload UF4 400 53,000 14,000 - - -

 Interchanging Trays 70 3,300 1,000 - - -

Package UF4 500 115,000 23,000 10 390 110 

The lognormal distribution parameters for hydrofluorination are shown in Table 8.13. 

Table 8.13. Lognormal Distribution Parameters for Dust Concentrations During Hydrofluorination 

Operations 

Dust Concentration, d/min/m3 

Trays Reactor 

Median GSD Median GSD 

Weighted Average Exposures 

UO2 loaders 1334 6.118 - -

Furnace Operator 245 4.167 8 5

 UF4 package 696 3.492 8 7.081 

Breathing Zone operations 

Loading UO2 3309 2.283

 Unload UF4 4604 9.245

 Interchanging Trays 481 4.329

 Package UF4 7583 9.200 60 3.361 

8.2.6 Reduction to Metal 

Internal dosimetry parameters for reduction to metal are shown in Table 8.14. 

Table 8.14. Metal reduction information and parameters 

Time Period All times 

Particle Parameters 5 micron AMAD; 6.7 g/cm3

 (UF4) 

Input Material UF4 

Output Material U metal 

The dust concentrations in air for the metal reduction process are shown in Table 8.15. The modification 

from manual handling and remote handling occurred between 1949 and 1951. The dose reconstructor is 

urged to determine if manual or remote handling occurred when the worker under consideration was 

working. 
 

 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 43 

Table 8.15. Dust concentrations during metal reduction 

Operations 

Dust Concentration, d/min/m3 

Before Modification Remote Handling 

min max average min max average 

Bomb Preparation 

 BZ Load bomb 700 320,000 90,000 1 220 45

 BZ Jolting 12,000 110,000 50,000 10 200 36

 GA Preparation Area 80 5,900 2,650 1 311 21

 Weighted Ave. Exp. 300 2,300 875 7 130 30 

Reduction Operations 

 BZ Change bombs 300 2,700 1,200 0 144 27

 GA Reduction furnace area 80 2,200 520 0 32 7

 Weighted Ave. Exp. 130 970 300 7 20 11 

Bomb Breakout 

 BZ Clean and unload shell 370 5,000 1,600 0 146 43

 BZ Chipping Derby 400 6,000 2,000 0 300 70

 GA Breakout Area 26 2,100 530 0 75 25 

Table 8.16 shows the lognormal distribution parameters for metal reduction. 

Table 8.16. Lognormal Distributions of Parameters for Metal Reduction 

Operations 

Dust Concentration, d/min/m3 

Before Modification Remote Handling 

Median GSD Median GSD 

Bomb Preparation 

 BZ Load bomb 24650 5 15 9.205

 BZ Jolting 36332 1.894 10 5

 GA Preparation Area 687 14.878 18 2.894

 Weighted Ave. Exp. 240 5 8 5 

Reduction Operations 

 BZ Change bombs 900 1.778 7 5

 GA Reduction furnace area 143 5 2 5

 Weighted Ave. Exp. 82 5 3 5 

Bomb Breakout 

 BZ Clean and unload shell 438 5 12 5

 BZ Chipping Derby 1549 1.667 19 5

 GA Breakout Area 234 5.145 7 5 

8.2.7 Recasting 

Internal dosimetry parameters for recasting are shown in Table 8.17. 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 44 

Table 8.17. Metal recasting information and parameters 

Time Period All times 

Particle Parameters 5 micron AMAD; 6.7 g/cm3

 (UF4) 

Input Material UF4 

Output Material U metal 

Default air concentrations of uranium dusts for recasting are shown in Table 8.18. 

Table 8.18. Dust concentrations during recasting 

Operations 

Dust Concentration, d/min/m3 

Before 1951 After 1951 

min max average min max average 

BZ Load crucible 28 2,700 1,050 64 870 250 

GA Crucible loading area 6 77 25 5 87 32 

Weighted Average – crucible 

loading 

25 77 51 15 86 36 

BZ Open furnace and clean 

furnace 

20 77,000 32,000 130 4,600 2,100 

BZ Remove crucible 6,600 10,000 8,400 800 1,700 1,300 

GA Recasting furnace area 11 3,400 580 0 290 34 

Weighted Average – 

recasting 

110 4,100 1,100 11 83 47 

GA Crucible burnout area 17 140 55 0 250 32 

Weighted average – crucible 

burnout 

25 100 50 7 360 70 

BZ Billet removal 1,200 3,800 2,700 78 310 180 

BZ Clean Billet with chisel 4,900 6,300 5,600 -­ -- -­

BZ Degrease and weigh 

billet 

46 70 59 45 1,200 130 

Weighted Average – billet 

cleaning 

14 80 47 14 49 28 

BZ Assemble crucible 56 2,800 1,300 12 270 81 

BZ Clean crucible parts 160 360 250 92 1,000 320 

GA Crucible assembly area 17 140 80 13 280 81 

Weighted average – crucible 

assembly 

100 540 360 31 81 55 

Table 8.19 shows the lognormal distributions parameters for recasting. 
 

 

 

 

 

 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 45 

Table 8.19. Lognormal Distributions of Parameters for Metal Recasting 

Operations 

Dust Concentration, d/min/m3 

Before 1951 After 1951 

Median GSD Median GSD 

BZ Load crucible 275 14.583 69 5 

GA Crucible loading area 7 5 21 2.354 

Weighted Average – crucible loading 44 1.351 10 5 

BZ Open furnace and clean furnace 8763 5 773 7.375 

BZ Remove crucible 2300 5 1166 1.243 

GA Recasting furnace area 193 8.995 9 5 

Weighted Average – recasting 672 2.683 30 2.419 

GA Crucible burnout area 15 5 9 5 

Weighted average – crucible burnout 14 5 19 5 

BZ Billet removal 2135 1.599 155 1.340 

BZ Clean Billet with chisel 1534 5 - -

BZ Degrease and weigh billet 57 1.081 

Weighted Average – billet cleaning 33 1.972 8 5 

BZ Assemble crucible 396 10.788 57 2.025 

BZ Clean crucible parts 68 5 88 5 

GA Crucible assembly area 49 2.689 60 1.802 

Weighted average – crucible assembly 232 2.4 50 1.205 

8.2.8 Fluorination 

Table 8.20. Fluorination information and parameters 

Time Period All times 

Particle Parameters 5 micron AMAD; 6.7 g/cm3

 (UF4); 

Input Material UF4 

Output Material UF6 

Table 8.21. Dust concentrations during fluorination 

Operations 

Dust Concentration, d/min/m3 

min max average 

Hex Loaders

 BZ Remove ash 1,000 14,000 5,800

 BZ Loading tubes into furnace 400 26,000 8,200

 BZ Connect line to receiver 1,000 35,000 2,100

 DWA Hex loaders 670 7,300 2,600 

Fluorination Operators 

 GA Cell room and steam room 10 90 55 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 46

 GA Fluorination area 210 7,500 1,500

 DWA Fluorination Operators 150 5,600 1,100 

Redistillation Operators 

BZ Make or break line connections 35 2,000 530

 BZ Remove receivers 55 450 190

 GA Distillation Area 66 1,300 450

 DWA Still operators 65 910 350 

Central Loaders 

 BZ Dump UF4 to central hopper 1,900 6,300 4,400

 BZ Vacuum ash from tray 1,200 1,400 1,350

 BZ Load UF4 at central loading 190 370 240

 GA Central loading room 100 430 220

 DWA Central loaders 220 1,100 550 

The lognormal distribution parameters for dust concentrations during fluorination are shown in Table 

8.22. 

Table 8.22. Lognormal Distributions of Parameters for Dust Concentrations During Fluorination
 

Operations 

Dust Concentration, d/min/m3 

Median GSD 

Hex Loaders

 BZ Remove ash 3742 2.403

 BZ Loading tubes into furnace 3225 6.465

 BZ Connect line to receiver 575 5

 DWA Hex loaders 2212 1.382 

Fluorination Operators 

 GA Cell room and steam room 30 3.361

 GA Fluorination area 1255 1.429

 DWA Fluorination Operators 624 3.103 

Redistillation Operators 

BZ Make or break line connections 265 4.013

 BZ Remove receivers 157 1.459

 GA Distillation Area 293 2.360

 DWA Still operators 243 2.071 

Central Loaders 

 BZ Dump UF4 to central hopper 3460 1.617

 BZ Vacuum ash from tray 1296 1.085

 BZ Load UF4 at central loading 66 5

 GA Central loading room 60 5

 DWA Central loaders 469 1.375 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 47 

8.2.9 Scrap Recovery 

Table 8.23. Dust concentrations during scrap recovery 

Operations 

Dust Concentration, d/min/m3 

Trays Calciner 

average average 

Furnace operators’ average exposures up to 1952 3,000 --

Furnace operators’ average exposures since 1952 300 1,000 

GA Furnace area 900 200 

BZ Dump scrap into furnace 23,000 135,000 

BZ Rake charge 2,500 2,000 

BZ Remove material from furnace 7,500 4,000 

Digest operator average exposure 2,500 2,000 

Filtration operator average exposure 7,500 4,000 

Digest operator average exposure 28 

Filtration operator average exposure 10 

The lognormal distribution parameters for scrap recovery are listed in Table 8.24. 

Table 8.24 Lognormal Distributions of Parameters for Air Concentrations During Scrap Recovery
 

Operations 

Dust Concentration, d/min/m3 

Trays Trays 

Median GSD Median GSD 

Furnace operators’ average exposures up to 1952 822 5 -- --

Furnace operators’ average exposures since 1952 82 5 274 5 

GA Furnace area 246 5 55 5 

BZ Dump scrap into furnace 6299 5 36971 5 

BZ Rake charge 685 5 548 5 

BZ Remove material from furnace 2054 5 1095 5 

median GSD 

Digest operator average exposure 8 5 

Filtration operator average exposure 3 5 

8.2.10 Drum Transfer Operations 

Several chemical forms of uranium were involved in drum transfer operations. The default internal 

dosimetry parameters to be used should be in agreement with the data in Table 3.12. 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 48 

Table 8.25. Dust concentrations during drum transfer operations 

Operations 

Dust Concentration, d/min/m3 

min max average 

 BZ Fill uranyl nitrate drum 300 2,000 1,100

 BZ Fill UO3 drums from pots 31 29,000 600

 BZ Fill UO2 drums – manual 170 20,000 10,000

 BZ Fill UO2 drums – remote 26 950 220

 BZ Fill UO2 cartons 1,000 175,000 36,000

 BZ Fill UO2 cartons – manual 3,800 40,000 21,000

 BZ Fill UO2 cartons – remote 120 980 400

 BZ Open UO2 cartons for remote dump 3,300 31,000 11,000 

Package UF4

 BZ Weigh and seal full drum 110 115,000 21,000

 BZ Adjust final drum weight 100 6,400 1,600

 GA Packaging area before 1950 110 960 500

 GA Packaging area since 1950 1 180 43 

The lognormal distribution parameters for drum loading operations are shown in Table 8.26. 

Table 8.26. Lognormal Distributions of Parameters for Drum Loading Operations. 

Operations 

Dust Concentration, d/min/m3 

median GSD 

 BZ Fill uranyl nitrate drum 775 2.017

 BZ Fill UO3 drums from pots 164 5

 BZ Fill UO2 drums – manual 2739 5

 BZ Fill UO2 drums – remote 157 1.960

 BZ Fill UO2 cartons 13229 7.406

 BZ Fill UO2 cartons – manual 12329 2.901

 BZ Fill UO2 cartons - remote 110 5

 BZ Open UO2 cartons for remote dump 3012 5 

Package UF4

 BZ Weigh and seal full drum 5,751 5

 BZ Adjust final drum weight 800 4.000

 GA Packaging area before 1950 325 2.367

 GA Packaging area since 1950 13 10.272 

8.3 Resuspension During Periods with no Uranium Operations 

There was a potential for internal exposure to resuspended material from the AEC work during non-AEC 

operations. To estimate exposure from resuspended materials, this analysis assumed that surfaces in the 

building became contaminated by deposition of uranium dust during the uranium operations. NIOSH 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 49 

(2005) estimates that for uranium operations, a reasonable maximum time-weighted average air 

concentration would be 7000 dpm/m3

 during AEC operations. 

The level of contamination was determined by multiplying the air concentration of 7000 dpm/m3 

(ORAUT-OTIB-0004) by the indoor deposition velocity and the assumed deposition time, which for 

uranium was 20 hr per operating day. The indoor deposition velocity is dependent on the physical 

properties of the room (air viscosity and density, turbulence, thermal gradients, surface geometry, etc.). It 

is also dependent on the physical properties of the aerosol particles (such as diameter, shape, and density). 

These characteristics are not known, so the terminal settling velocity was calculated for an aerosol with 

the default particle size distribution of 5-μm activity median aerodynamic diameter. The calculated 

terminal settling velocity was 7.5 × 10-4 m/s, which is within the range of deposition velocities (2.7 × 10-6 

to 2.7 × 10-3 m/s) measured in various studies (NRC 2002a). 

The calculated surface contamination level created from airborne dusts during a 1-year period of uranium 

metal-working operations would be 3.44×107

 pCi/m2 . This level of surface contamination assumes that 

all uranium deposited on the floor was present for the entire period of AEC operations. Therefore, using a 

resuspension factor of 1×10-6/m (NRC 2002b), the air concentration due to resuspension would be 34.4 

pCi/m3 

. 

The annual inhalation intake received from resuspension of deposited material, assuming 10-hour 

workdays and the worst-case air concentrations for a one-year metal-working operation would be 413 

pCi/day. 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 50 

8.4 Time-Dependent Air Concentration Data 

The air concentration in refining plants varied with time. The following information was extracted from 

Strom (2006). Christofano and Harris (1960) showed that there was a large reduction over the years in 

time-weighted average air concentrations of uranium in various refining plants (Figure 1). 

Figure 1. Data from Figure 1 of Christofano and Harris (1960) showing the decreasing trend in timeweighted average air concentrations of alpha-emitting dust. 

The image in Figure 1 was digitized1

 and the data analyzed for each year. The average value (mean) was 

computed from the digitized data for each year, and a lognormal was fit to the data for each year by 

computing natural logs, averaging them, and taking their standard deviation as described in BTIB-5000. 

The geometric mean and geometric standard deviation (GSD) was calculated for each year, and an upper 

95th percentile. These results are shown in Table 8.27 and Figure 2. Vertical bars are “times or divided 

by” one GSD. 

1 UnGraph V5, Biosoft, http://www.biosoft.com, PO Box 1013, Great Shelford, Cambridge, CB2 5WQ United 

Kingdom. tel: +44 1223 841700 fax: +44 1223 841802 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 51 

Table 8.27. Results of digitizing data from Figure 1. All concentrations (columns 3-7 and 9) are in 

dpm/m3 

No. Geometric Geometric 

Data Standard Geometric Mean Mean Upper 

Year Points Mean Deviation Mean ×GSD ÷GSD GSD 95%ile 

1948 17 7398 12417 2061 10792 393 5.24 31391 

1949 9 964 1119 540 1902 153 3.52 4286 

1950 25 349 951 101 434 24 4.29 1112 

1951 8 521 569 219 1062 45 4.85 2940 

1952 15 124 133 79 207 30 2.62 385 

1953 31 71 121 40 102 16 2.53 186 

1954 15 94 85 60 170 21 2.84 334 

1955 8 63 27 57 91 36 1.58 122 

1956 9 85 109 43 144 13 3.35 315 

Total 137 "Weighted Exposures " (Concentration [dpm/m 3 ])

100,000 

10,000 

1,000 

100 

10 

Mean 

Geometric Mean 

Upper 95%ile 

1947 1948 1949 1950 1951 1952 1953 1954 1955 1956 1957 

Year 

Figure 2. Analysis of time-weighted average air concentrations from Christofano and Harris (1960; Fig. 

1). Variability bars are the geometric mean times or divided by one GSD. 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 52 

Clearly, the mean concentration drops rapidly from 1948 through 1950, from 7,400 dpm/m3

 to 350 

dpm/m3

 (a factor of over 20), as engineered workplace controls were installed at the dustiest locations. 

Alpha-emitting dust concentrations for the years 1953-1957 are roughly 100 times lower than they were 

in 1948. 

The first difficulty in interpreting Table 8.27 is that it combines data from different processes, and we do 

not know which processes these are, that is, which are the dustiest or cleanest in a given year. 

The second difficulty in interpreting the data in Table 8.27 and the analysis in Figure 2 is that data for the 

various processes may have been measured one or more times, and we do not know which times these 

were. So, while we know that workplace air became dramatically cleaner on the average, it is necessary to 

develop a defensible method to apply this knowledge to the tabulated values for the various operations 

that have been derived from the 1960 Christofano and Harris paper. 

One approach to using these data would be to ignore process-specific information and use the upper 95th 

percentile value in a given year for all process workers, and the median value for non-process workers. 

This assumption may be favorable to the claimant, but it ignores information that can be used. 

A preferred approach to using these data is to analyze all of the averages and create a year-specific scaling 

or correction factor for the tabulated values for the various operations that have been derived from the 

1960 Christofano and Harris paper. The time-dependent correction factors that result from this method are 

shown in Figure 3. The average of these factors is 1.00, so that a worker who was present for all 9 years 

would have a predicted intake that is unchanged by the application of this method. In this sense, the time 

correction method is unbiased. A worker who was present only for 1948 would have a nearly 7-fold 

higher intake, while workers who were present after 1953 would have dramatically lower intakes. The 

1948 value would be used for all years prior to 1948, and the 1956 value would be used for all years after 

1956. These factors are presented in Table 8.28. 

The annual intake is thus computed using 

I = (1.2m3/h) × (actual or default hours per year) annual 

× (process - specific concentration) × (year - specific correction factor), 

which is a modification of Eq. 21 in BTIB-5000 (Strom 2006). 

1 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 53 

10

Annual Mean/Overall Mean 

1 

0.1 

0.01 

1947 1948 1949 1950 1951 1952 1953 1954 1955 1956 1957 

Year 

Figure 3. Use of year-specific correction factors to adjust air concentrations for processes over time. 

Table 8.28. Year-specific correction factors for tabulated process-specific air concentrations, derived 

from analysis of Christofano and Harris (1960) Fig. 1. 

YearSpecific 

Correction 

Year Factor 

1948 or before 6.89 

1949 0.897 

1950 0.325 

1951 0.485 

1952 0.116 

1953 0.0656 

1954 0.0875 

1955 0.0582 

1956 or after 0.0792 

Average 1 

The tabulated process-specific data reported by Christofano and Harris (1960) include measurements 

made over the duration of the survey. They state, “In substantially every case, the first survey conducted 

at any plant disclosed the highest exposures and the most recent measurements were lowest. In reporting 

occupational dust exposures, we have presented the range of exposures and also the numerical average of 

all such evaluations made during the 10 years from 1948 through 1957” (Christofano and Harris 1960, p. 
 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 54 

77)2

. The inclusion of measurements made at different times as ventilation and engineered controls 

improved means that some of the broader lognormal distributions include non-random variability due to 

temporal improvements in workplace controls, as well a component of random variability due to other 

factors such as process differences between plants and air sample representativeness. Using the yearspecific correction factors to average air concentrations improves the precision of the estimates of air 

concentrations. 

8.5 Summary 

This section provides summary internal dosimetry information for the dose reconstructor. 

8.5.1 Inhalation Intakes 

The algorithm for populating this table was to use the TWA or DWA values for the operators who were 

assumed to be exposed the entire working day. General laborers were assumed to be exposed to 50% of 

the operators. Supervisors were assumed to be exposed for one-fourth of the working day. Clerical 

workers (who represent all other workers), were assumed to have intakes one-tenth that of a supervisor. 

The inhalation intake rate is the air concentration times the breathing rate for a light worker, 9.6 m3

/d. 

This intake rate is the actual daily intake rate experienced by the individual. Internal doses are calculated 

using the IMBA codes and, for chronic intakes, it is necessary to provide a daily intake rate that 

corresponds to intakes that occur even during times when the individual was not at work. The intake rate, 

from inhalation of known air concentrations, that accommodates the IMBA code is 

IIMBA = 3.288×10−3 Ah , where IIMBA is the IMBA chronic intake in pCi/d, A is the air concentration 

(pCi/m3

) and h is the number of working hours in a year. If the measured air concentration has units of 

dpm/m3

 then divide the constant by 2.22 (dpm/pCi). 

8.5.1.1 Stage 1: Operations 

The daily inhalation intake rates for the operations phase of the plant are shown in Table 8.29. These 

intake rates have been calculated for use in the IMBA code, as described above. 

[Solvent Extraction Data is Reserved] 

Table 8.29. Intake rates for inhalation of uranium dusts during Stage 1 – Operations. 

Operation Job Title Years 

Inhalation Rates (pCi/d) for three work week durations. 

48-h/week 44-h/week 40-h/week 

median GSD median GSD median GSD 

Ore Digestion 

Operator 

All 

1.74E+02 4.94 1.60E+02 4.94 1.45E+02 4.94 

general 

laborer 8.71E+01 4.94 7.98E+01 4.94 7.26E+01 4.94 

Supervisor 4.35E+01 4.94 3.99E+01 4.94 3.63E+01 4.94 

Clerical 4.35E+00 4.94 3.99E+00 4.94 3.63E+00 4.94 

Solvent Extraction 

Operator 

start – 

1949 

N/A N/A N/A N/A N/A N/A 

general 

laborer N/A N/A N/A N/A N/A N/A 

Supervisor N/A N/A N/A N/A N/A N/A 

Clerical N/A N/A N/A N/A N/A N/A 

Operator N/A N/A N/A N/A N/A N/A 

2

 There are no data from 1957 in Figure 1, so this analysis goes only through 1956. 
 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 55 

general 

laborer 

1950 -

end N/A N/A N/A N/A N/A N/A 

Supervisor N/A N/A N/A N/A N/A N/A 

Clerical N/A N/A N/A N/A N/A N/A 

Boildown and 

Denitration 

Operator 

Start – 

1949 

4.12E+04 1.67 3.78E+04 1.67 3.43E+04 1.67 

general 

laborer 2.06E+04 1.67 1.89E+04 1.67 1.72E+04 1.67 

Supervisor 1.03E+04 1.67 9.44E+03 1.67 8.59E+03 1.67 

Clerical 1.03E+03 1.67 9.44E+02 1.67 8.59E+02 1.67 

Operator 

1950 – 

End 

3.02E+02 2.33 2.77E+02 2.33 2.52E+02 2.33 

general 

laborer 1.51E+02 2.33 1.38E+02 2.33 1.26E+02 2.33 

Supervisor 7.55E+01 2.33 6.92E+01 2.33 6.29E+01 2.33 

Clerical 7.55E+00 2.33 6.92E+00 2.33 6.29E+00 2.33 

Oxide Reduction: 

Tray Furnace 

Operations 

Operator 

All 

6.29E+04 1.28 5.77E+04 1.28 5.25E+04 1.28 

general 

laborer 3.15E+04 1.28 2.89E+04 1.28 2.62E+04 1.28 

Supervisor 1.57E+04 1.28 1.44E+04 1.28 1.31E+04 1.28 

Clerical 1.57E+03 1.28 1.44E+03 1.28 1.31E+03 1.28 

Oxide Reduction: 

Multiple Hearth 

Operations 

Operator 

All 

1.28E+03 3.76 1.18E+03 3.76 1.07E+03 3.76 

general 

laborer 6.42E+02 3.76 5.88E+02 3.76 5.35E+02 3.76 

Supervisor 3.21E+02 3.76 2.94E+02 3.76 2.67E+02 3.76 

Clerical 3.21E+01 3.76 2.94E+01 3.76 2.67E+01 3.76 

Oxide Reduction: 

Horizontal 

Reactor 

Operations 

Operator 

All 

3.66E+02 1.86 3.36E+02 1.86 3.05E+02 1.86 

general 

laborer 1.83E+02 1.86 1.68E+02 1.86 1.53E+02 1.86 

Supervisor 9.15E+01 1.86 8.39E+01 1.86 7.63E+01 1.86 

Clerical 9.15E+00 1.86 8.39E+00 1.86 7.63E+00 1.86 

Hydrofluorination: 

UO2 Loaders 

(Trays) 

Operator 

All 

4.74E+03 6.12 4.35E+03 6.12 3.95E+03 6.12 

general 

laborer 2.37E+03 6.12 2.17E+03 6.12 1.98E+03 6.12 

Supervisor 1.19E+03 6.12 1.09E+03 6.12 9.88E+02 6.12 

Clerical 1.19E+02 6.12 1.09E+02 6.12 9.88E+01 6.12 

Hydrofluorination: 

Furnace Operator 

(Trays) 

Operator 

All 

8.71E+02 4.17 7.98E+02 4.17 7.26E+02 4.17 

general 

laborer 4.35E+02 4.17 3.99E+02 4.17 3.63E+02 4.17 

Supervisor 2.18E+02 4.17 2.00E+02 4.17 1.81E+02 4.17 

Clerical 2.18E+01 4.17 2.00E+01 4.17 1.81E+01 4.17 

Hydrofluorination: 

Package UF4 

(Trays) 

Operator 

All 

2.47E+03 3.49 2.27E+03 3.49 2.06E+03 3.49 

general 

laborer 1.24E+03 3.49 1.13E+03 3.49 1.03E+03 3.49 

Supervisor 6.18E+02 3.49 5.67E+02 3.49 5.15E+02 3.49 

Clerical 6.18E+01 3.49 5.67E+01 3.49 5.15E+01 3.49 

Hydrofluorination: 

Furnace Operator 

(Reactor) 

Operator 

All 

1.07E+02 5.00 9.78E+01 5.00 8.89E+01 5.00 

general 

laborer 5.33E+01 5.00 4.89E+01 5.00 4.44E+01 5.00 

Supervisor 2.67E+01 5.00 2.44E+01 5.00 2.22E+01 5.00 

Clerical 2.67E+00 5.00 2.44E+00 5.00 2.22E+00 5.00 

Hydrofluorination: 

Package UF4 

(Reactor) 

Operator 

All 

2.84E+01 7.08 2.61E+01 7.08 2.37E+01 7.08 

general 

laborer 1.42E+01 7.08 1.30E+01 7.08 1.18E+01 7.08 

Supervisor 7.11E+00 7.08 6.52E+00 7.08 5.92E+00 7.08 

Clerical 7.11E-01 7.08 6.52E-01 7.08 5.92E-01 7.08 

Metal Reduction: 

Bomb Preparation 

Operator 

Before 

1952 

8.53E+02 5.00 7.82E+02 5.00 7.11E+02 5.00 

general 

laborer 4.27E+02 5.00 3.91E+02 5.00 3.55E+02 5.00 

Supervisor 2.13E+02 5.00 1.96E+02 5.00 1.78E+02 5.00 

Clerical 2.13E+01 5.00 1.96E+01 5.00 1.78E+01 5.00 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 56 

Metal Reduction: 

Reduction 

Operations 

Operator 

Before 

1952 

2.91E+02 5.00 2.67E+02 5.00 2.43E+02 5.00 

general 

laborer 1.46E+02 5.00 1.34E+02 5.00 1.21E+02 5.00 

Supervisor 7.29E+01 5.00 6.68E+01 5.00 6.07E+01 5.00 

Clerical 7.29E+00 5.00 6.68E+00 5.00 6.07E+00 5.00 

Metal Reduction: 

Bomb Breakout 

Operator 

Before 

1952 

8.32E+02 5.00 7.62E+02 5.00 6.93E+02 5.00 

general 

laborer 4.16E+02 5.00 3.81E+02 5.00 3.47E+02 5.00 

Supervisor 2.08E+02 5.00 1.91E+02 5.00 1.73E+02 5.00 

Clerical 2.08E+01 5.00 1.91E+01 5.00 1.73E+01 5.00 

Metal Reduction: 

Bomb Preparation 

Operator 

1952 and 

later 

2.84E+01 5.00 2.61E+01 5.00 2.37E+01 5.00 

general 

laborer 1.42E+01 5.00 1.30E+01 5.00 1.18E+01 5.00 

Supervisor 7.11E+00 5.00 6.52E+00 5.00 5.92E+00 5.00 

Clerical 7.11E-01 5.00 6.52E-01 5.00 5.92E-01 5.00 

Metal Reduction: 

Reduction 

Operations 

Operator 

1952 and 

later 

1.07E+01 5.00 9.78E+00 5.00 8.89E+00 5.00 

general 

laborer 5.33E+00 5.00 4.89E+00 5.00 4.44E+00 5.00 

Supervisor 2.67E+00 5.00 2.44E+00 5.00 2.22E+00 5.00 

Clerical 2.67E-01 5.00 2.44E-01 5.00 2.22E-01 5.00 

Metal Reduction: 

Bomb Breakout 

Operator 

1952 and 

later 

2.49E+01 5.00 2.28E+01 5.00 2.07E+01 5.00 

general 

laborer 1.24E+01 5.00 1.14E+01 5.00 1.04E+01 5.00 

Supervisor 6.22E+00 5.00 5.70E+00 5.00 5.18E+00 5.00 

Clerical 6.22E-01 5.00 5.70E-01 5.00 5.18E-01 5.00 

Recasting: 

Crucible Loading 

Operator 

Prior to 

1951 

1.56E+02 1.35 1.43E+02 1.35 1.30E+02 1.35 

general 

laborer 7.82E+01 1.35 7.17E+01 1.35 6.52E+01 1.35 

Supervisor 3.91E+01 1.35 3.58E+01 1.35 3.26E+01 1.35 

Clerical 3.91E+00 1.35 3.58E+00 1.35 3.26E+00 1.35 

Recasting: 

Recasting 

Operator 

Prior to 

1951 

2.39E+03 2.68 2.19E+03 2.68 1.99E+03 2.68 

general 

laborer 1.19E+03 2.68 1.09E+03 2.68 9.95E+02 2.68 

Supervisor 5.97E+02 2.68 5.47E+02 2.68 4.98E+02 2.68 

Clerical 5.97E+01 2.68 5.47E+01 2.68 4.98E+01 2.68 

Recasting: 

Crucible Burnout 

Operator 

Prior to 

1951 

4.98E+01 5.00 4.56E+01 5.00 4.15E+01 5.00 

general 

laborer 2.49E+01 5.00 2.28E+01 5.00 2.07E+01 5.00 

Supervisor 1.24E+01 5.00 1.14E+01 5.00 1.04E+01 5.00 

Clerical 1.24E+00 5.00 1.14E+00 5.00 1.04E+00 5.00 

Recasting: 

Billet Cleaning 

Operator 

Prior to 

1951 

1.17E+02 1.97 1.08E+02 1.97 9.78E+01 1.97 

general 

laborer 5.87E+01 1.97 5.38E+01 1.97 4.89E+01 1.97 

Supervisor 2.93E+01 1.97 2.69E+01 1.97 2.44E+01 1.97 

Clerical 2.93E+00 1.97 2.69E+00 1.97 2.44E+00 1.97 

Recasting: 

Crucible 

Assembly 

Operator 

Prior to 

1951 

8.25E+02 2.40 7.56E+02 2.40 6.87E+02 2.40 

general 

laborer 4.12E+02 2.40 3.78E+02 2.40 3.44E+02 2.40 

Supervisor 2.06E+02 2.40 1.89E+02 2.40 1.72E+02 2.40 

Clerical 2.06E+01 2.40 1.89E+01 2.40 1.72E+01 2.40 

Recasting: 

Crucible Loading 

Operator 

1951 and 

later 

3.55E+01 5.00 3.26E+01 5.00 2.96E+01 5.00 

general 

laborer 1.78E+01 5.00 1.63E+01 5.00 1.48E+01 5.00 

Supervisor 8.89E+00 5.00 8.15E+00 5.00 7.41E+00 5.00 

Clerical 8.89E-01 5.00 8.15E-01 5.00 7.41E-01 5.00 

Recasting: 

Recasting 

Operator 

1951 and 

later 

1.07E+02 2.42 9.78E+01 2.42 8.89E+01 2.42 

general 

laborer 5.33E+01 2.42 4.89E+01 2.42 4.44E+01 2.42 

Supervisor 2.67E+01 2.42 2.44E+01 2.42 2.22E+01 2.42 

Clerical 2.67E+00 2.42 2.44E+00 2.42 2.22E+00 2.42 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 57 

Recasting: 

Crucible Burnout 

Operator 

1951 and 

later 

3.20E+01 5.00 2.93E+01 5.00 2.67E+01 5.00 

general 

laborer 1.60E+01 5.00 1.47E+01 5.00 1.33E+01 5.00 

Supervisor 8.00E+00 5.00 7.33E+00 5.00 6.66E+00 5.00 

Clerical 8.00E-01 5.00 7.33E-01 5.00 6.66E-01 5.00 

Recasting: 

Billet Cleaning 

Operator 

1951 and 

later 

2.84E+01 5.00 2.61E+01 5.00 2.37E+01 5.00 

general 

laborer 1.42E+01 5.00 1.30E+01 5.00 1.18E+01 5.00 

Supervisor 7.11E+00 5.00 6.52E+00 5.00 5.92E+00 5.00 

Clerical 7.11E-01 5.00 6.52E-01 5.00 5.92E-01 5.00 

Recasting: 

Crucible 

Assembly 

Operator 

1951 and 

later 

1.78E+02 1.21 1.63E+02 1.21 1.48E+02 1.21 

general 

laborer 8.89E+01 1.21 8.15E+01 1.21 7.41E+01 1.21 

Supervisor 4.44E+01 1.21 4.07E+01 1.21 3.70E+01 1.21 

Clerical 4.44E+00 1.21 4.07E+00 1.21 3.70E+00 1.21 

Fluorination: 

Hex Loaders 

Operator 

All 

7.86E+03 1.38 7.21E+03 1.38 6.55E+03 1.38 

general 

laborer 3.93E+03 1.38 3.60E+03 1.38 3.28E+03 1.38 

Supervisor 1.97E+03 1.38 1.80E+03 1.38 1.64E+03 1.38 

Clerical 1.97E+02 1.38 1.80E+02 1.38 1.64E+02 1.38 

Fluorination: 

Operators 

Operator 

All 

2.22E+03 3.10 2.03E+03 3.10 1.85E+03 3.10 

general 

laborer 1.11E+03 3.10 1.02E+03 3.10 9.24E+02 3.10 

Supervisor 5.55E+02 3.10 5.08E+02 3.10 4.62E+02 3.10 

Clerical 5.55E+01 3.10 5.08E+01 3.10 4.62E+01 3.10 

Fluorination: 

Still Operators 

Operator 

All 

8.64E+02 2.07 7.92E+02 2.07 7.20E+02 2.07 

general 

laborer 4.32E+02 2.07 3.96E+02 2.07 3.60E+02 2.07 

Supervisor 2.16E+02 2.07 1.98E+02 2.07 1.80E+02 2.07 

Clerical 2.16E+01 2.07 1.98E+01 2.07 1.80E+01 2.07 

Fluorination: 

Central Loaders 

Operator 

All 

1.67E+03 1.38 1.53E+03 1.38 1.39E+03 1.38 

general 

laborer 8.34E+02 1.38 7.64E+02 1.38 6.95E+02 1.38 

Supervisor 4.17E+02 1.38 3.82E+02 1.38 3.47E+02 1.38 

Clerical 4.17E+01 1.38 3.82E+01 1.38 3.47E+01 1.38 

Scrap Recovery: 

Furnace Operator 

(Trays) 

Operator 

1951 and 

earlier 

2.92E+03 5.00 2.68E+03 5.00 2.43E+03 5.00 

general 

laborer 1.46E+03 5.00 1.34E+03 5.00 1.22E+03 5.00 

Supervisor 7.30E+02 5.00 6.70E+02 5.00 6.09E+02 5.00 

Clerical 7.30E+01 5.00 6.70E+01 5.00 6.09E+01 5.00 

Scrap Recovery: 

Furnace Operator 

(Trays) 

Operator 

1952 and 

later 

2.91E+02 5.00 2.67E+02 5.00 2.43E+02 5.00 

general 

laborer 1.46E+02 5.00 1.34E+02 5.00 1.21E+02 5.00 

Supervisor 7.29E+01 5.00 6.68E+01 5.00 6.07E+01 5.00 

Clerical 7.29E+00 5.00 6.68E+00 5.00 6.07E+00 5.00 

Scrap Recovery: 

Furnace Operator 

(Calciner) 

Operator 

1952 and 

later 

9.74E+02 5.00 8.93E+02 5.00 8.12E+02 5.00 

general 

laborer 4.87E+02 5.00 4.46E+02 5.00 4.06E+02 5.00 

Supervisor 2.43E+02 5.00 2.23E+02 5.00 2.03E+02 5.00 

Clerical 2.43E+01 5.00 2.23E+01 5.00 2.03E+01 5.00 

Scrap Recovery: 

Digest Operator 

Operator 

All 

2.84E+01 5.00 2.61E+01 5.00 2.37E+01 5.00 

general 

laborer 1.42E+01 5.00 1.30E+01 5.00 1.18E+01 5.00 

Supervisor 7.11E+00 5.00 6.52E+00 5.00 5.92E+00 5.00 

Clerical 7.11E-01 5.00 6.52E-01 5.00 5.92E-01 5.00 

Scrap Recovery: 

Filtration Operator 

Operator 

All 

1.07E+01 4.94 9.78E+00 4.94 8.89E+00 4.94 

general 

laborer 5.33E+00 4.94 4.89E+00 4.94 4.44E+00 4.94 

Supervisor 2.67E+00 4.94 2.44E+00 4.94 2.22E+00 4.94 

Clerical 2.67E-01 4.94 2.44E-01 4.94 2.22E-01 4.94 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 58 

Drum Loading 

Operations 

Operator 

All 

N/A N/A N/A N/A N/A N/A 

general 

laborer N/A N/A N/A N/A N/A N/A 

Supervisor N/A N/A N/A N/A N/A N/A 

Clerical N/A N/A N/A N/A N/A N/A 

8.5.2 Resuspension During Periods with no Uranium Operations 

There was a potential for internal exposure to resuspended material from the AEC work during non-AEC 

operations following soon after the actual operations. To estimate exposure from resuspended materials, 

this analysis assumed that surfaces in the building became contaminated by deposition of uranium dust 

during operations similar to metal-working operations. NIOSH (2005) estimates that for uranium metalworking operations, a reasonable maximum time-weighted average air concentration would be 7000 

dpm/m3

 (equal to 100 MAC) during AEC operations. The estimation of time-weighted average air 

concentrations for refining operations are still in preparation and therefore the metal working data will be 

used. 

The level of contamination was determined by multiplying the air concentration of 7000 dpm/m3

 by the 

indoor deposition velocity and the assumed deposition time, which for uranium was 20 hr per operating 

day. The indoor deposition velocity is dependent on the physical properties of the room (air viscosity and 

density, turbulence, thermal gradients, surface geometry, etc.). It is also dependent on the physical 

properties of the aerosol particles (such as diameter, shape, and density). These characteristics are not 

known, so the terminal settling velocity was calculated for an aerosol with the default particle size 

distribution of 5-μm activity median aerodynamic diameter. The calculated terminal settling velocity was 

7.5 × 10-4 m/s, which is within the range of deposition velocities (2.7 × 10-6 to 2.7 × 10-3 m/s) measured in 

various studies (NRC 2002a). 

The calculated surface contamination level created from airborne dusts during a 1-year period of uranium 

metal-working operations would be 3.44×107

 pCi/m2 . This level of surface contamination assumes that 

all uranium deposited on the floor was present for the entire period of AEC operations. Therefore, using a 

resuspension factor of 1×10-6/m (NRC 2002b), the air concentration due to resuspension would be 34.4 

pCi/m3 

. 

The annual inhalation intake received from resuspension of deposited material, assuming 10-hour 

workdays and the worst-case air concentrations for a one-year working operation, would be 413 pCi/day. 

8.5.3 Ingestion Intakes 

Ingestion Intakes are determined using the OCAS method (NIOSH 2004) and are shown in Table 8.30. 

Briefly, the daily ingestion rate from food contamination is given by I = 0.0985A , where I is the daily 

ingestion rate in pCi/d, A is the average air concentration (pCi/m3

). The ingestion rate I must be adjusted 

for the fact that IMBA assumes chronic intakes, even during weekends, and that the number of hours 

worked in a year changed over time (Strom 2006). The adjustments result in an IMBA chronic intake rate 

IIMBA = 3.373×10−5 Ah , where IIMBA is the daily intake (pCi/d), A is the median air concentration 

(pCi/m3

), and h is the number of hours in a working year. 

Making similar adjustments to Neton’s (NIOSH 2004) equations for incidental hand-to-mouth ingestion 

the chronic IMBA daily intake rate is IIMBA = 3.425×10−5 Ah , where IIMBA is the daily chronic intake 

rate (pCi/d), h is the number of hours in a work year, and A is the median dust concentration (pCi/m3

). 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 59 

The total ingestion rate is the sum of the food contamination and incidental hand-to-mouth ingestion rates 

and is I IMBA = 6.798×10−5 Ah where the variables are the same as previously defined. 

Table 8.30. Intake rates for ingestion of uranium dusts during Stage 1 – Operations. 

Operation Job Title Years 

Ingestion Rates (pCi/d) for three work week durations. 

48-h/week 48-h/week 48-h/week 

median GSD median GSD median GSD 

Ore Digestion 

Operator 

All 

1.79E+00 4.94 1.64E+00 4.94 1.49E+00 4.94 

general 

laborer 8.93E-01 4.94 8.18E-01 4.94 7.44E-01 4.94 

Supervisor 4.46E-01 4.94 4.09E-01 4.94 3.72E-01 4.94 

Clerical 4.46E-02 4.94 4.09E-02 4.94 3.72E-02 4.94 

Solvent Extraction 

Operator 

start – 

1949 

N/A N/A N/A N/A N/A N/A 

general 

laborer N/A N/A N/A N/A N/A N/A 

Supervisor N/A N/A N/A N/A N/A N/A 

Clerical N/A N/A N/A N/A N/A N/A 

Operator 

1950 -

end 

N/A N/A N/A N/A N/A N/A 

general 

laborer N/A N/A N/A N/A N/A N/A 

Supervisor N/A N/A N/A N/A N/A N/A 

Clerical N/A N/A N/A N/A N/A N/A 

Boildown and 

Denitration 

Operator 

Start – 

1949 

4.22E+02 1.67 3.87E+02 1.67 3.52E+02 1.67 

general 

laborer 2.11E+02 1.67 1.94E+02 1.67 1.76E+02 1.67 

Supervisor 1.06E+02 1.67 9.68E+01 1.67 8.80E+01 1.67 

Clerical 1.06E+01 1.67 9.68E+00 1.67 8.80E+00 1.67 

Operator 

1950 – 

End 

3.10E+00 2.33 2.84E+00 2.33 2.58E+00 2.33 

general 

laborer 1.55E+00 2.33 1.42E+00 2.33 1.29E+00 2.33 

Supervisor 7.74E-01 2.33 7.10E-01 2.33 6.45E-01 2.33 

Clerical 7.74E-02 2.33 7.10E-02 2.33 6.45E-02 2.33 

Oxide Reduction: 

Tray Furnace 

Operations 

Operator 

All 

6.45E+02 1.28 5.91E+02 1.28 5.38E+02 1.28 

general 

laborer 3.23E+02 1.28 2.96E+02 1.28 2.69E+02 1.28 

Supervisor 1.61E+02 1.28 1.48E+02 1.28 1.34E+02 1.28 

Clerical 1.61E+01 1.28 1.48E+01 1.28 1.34E+01 1.28 

Oxide Reduction: 

Multiple Hearth 

Operations 

Operator 

All 

1.32E+01 3.76 1.21E+01 3.76 1.10E+01 3.76 

general 

laborer 6.58E+00 3.76 6.03E+00 3.76 5.48E+00 3.76 

Supervisor 3.29E+00 3.76 3.01E+00 3.76 2.74E+00 3.76 

Clerical 3.29E-01 3.76 3.01E-01 3.76 2.74E-01 3.76 

Oxide Reduction: 

Horizontal 

Reactor 

Operations 

Operator 

All 

3.75E+00 1.86 3.44E+00 1.86 3.13E+00 1.86 

general 

laborer 1.88E+00 1.86 1.72E+00 1.86 1.56E+00 1.86 

Supervisor 9.38E-01 1.86 8.60E-01 1.86 7.82E-01 1.86 

Clerical 9.38E-02 1.86 8.60E-02 1.86 7.82E-02 1.86 

Hydrofluorination: 

UO2 Loaders 

(Trays) 

Operator 

All 

4.86E+01 6.12 4.46E+01 6.12 4.05E+01 6.12 

general 

laborer 2.43E+01 6.12 2.23E+01 6.12 2.03E+01 6.12 

Supervisor 1.22E+01 6.12 1.11E+01 6.12 1.01E+01 6.12 

Clerical 1.22E+00 6.12 1.11E+00 6.12 1.01E+00 6.12 

Hydrofluorination: 

Furnace Operator 

(Trays) 

Operator 

All 

8.93E+00 4.17 8.18E+00 4.17 7.44E+00 4.17 

general 

laborer 4.46E+00 4.17 4.09E+00 4.17 3.72E+00 4.17 

Supervisor 2.23E+00 4.17 2.05E+00 4.17 1.86E+00 4.17 

Clerical 2.23E-01 4.17 2.05E-01 4.17 1.86E-01 4.17 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 60 

Hydrofluorination: 

Package UF4 

(Trays) 

Operator 

All 

2.54E+01 3.49 2.32E+01 3.49 2.11E+01 3.49 

general 

laborer 1.27E+01 3.49 1.16E+01 3.49 1.06E+01 3.49 

Supervisor 6.34E+00 3.49 5.81E+00 3.49 5.28E+00 3.49 

Clerical 6.34E-01 3.49 5.81E-01 3.49 5.28E-01 3.49 

Hydrofluorination: 

Furnace Operator 

(Reactor) 

Operator 

All 

1.09E+00 5.00 1.00E+00 5.00 9.11E-01 5.00 

general 

laborer 5.46E-01 5.00 5.01E-01 5.00 4.55E-01 5.00 

Supervisor 2.73E-01 5.00 2.50E-01 5.00 2.28E-01 5.00 

Clerical 2.73E-02 5.00 2.50E-02 5.00 2.28E-02 5.00 

Hydrofluorination: 

Package UF4 

(Reactor) 

Operator 

All 

2.91E-01 7.08 2.67E-01 7.08 2.43E-01 7.08 

general 

laborer 1.46E-01 7.08 1.34E-01 7.08 1.21E-01 7.08 

Supervisor 7.29E-02 7.08 6.68E-02 7.08 6.07E-02 7.08 

Clerical 7.29E-03 7.08 6.68E-03 7.08 6.07E-03 7.08 

Metal Reduction: 

Bomb Preparation 

Operator 

Before 

1952 

8.74E+00 5.00 8.02E+00 5.00 7.29E+00 5.00 

general 

laborer 4.37E+00 5.00 4.01E+00 5.00 3.64E+00 5.00 

Supervisor 2.19E+00 5.00 2.00E+00 5.00 1.82E+00 5.00 

Clerical 2.19E-01 5.00 2.00E-01 5.00 1.82E-01 5.00 

Metal Reduction: 

Reduction 

Operations 

Operator 

Before 

1952 

2.99E+00 5.00 2.74E+00 5.00 2.49E+00 5.00 

general 

laborer 1.49E+00 5.00 1.37E+00 5.00 1.24E+00 5.00 

Supervisor 7.47E-01 5.00 6.85E-01 5.00 6.22E-01 5.00 

Clerical 7.47E-02 5.00 6.85E-02 5.00 6.22E-02 5.00 

Metal Reduction: 

Bomb Breakout 

Operator 

Before 

1952 

8.53E+00 5.00 7.81E+00 5.00 7.10E+00 5.00 

general 

laborer 4.26E+00 5.00 3.91E+00 5.00 3.55E+00 5.00 

Supervisor 2.13E+00 5.00 1.95E+00 5.00 1.78E+00 5.00 

Clerical 2.13E-01 5.00 1.95E-01 5.00 1.78E-01 5.00 

Metal Reduction: 

Bomb Preparation 

Operator 

1952 and 

later 

2.91E-01 5.00 2.67E-01 5.00 2.43E-01 5.00 

general 

laborer 1.46E-01 5.00 1.34E-01 5.00 1.21E-01 5.00 

Supervisor 7.29E-02 5.00 6.68E-02 5.00 6.07E-02 5.00 

Clerical 7.29E-03 5.00 6.68E-03 5.00 6.07E-03 5.00 

Metal Reduction: 

Reduction 

Operations 

Operator 

1952 and 

later 

1.09E-01 5.00 1.00E-01 5.00 9.11E-02 5.00 

general 

laborer 5.46E-02 5.00 5.01E-02 5.00 4.55E-02 5.00 

Supervisor 2.73E-02 5.00 2.50E-02 5.00 2.28E-02 5.00 

Clerical 2.73E-03 5.00 2.50E-03 5.00 2.28E-03 5.00 

Metal Reduction: 

Bomb Breakout 

Operator 

1952 and 

later 

2.55E-01 5.00 2.34E-01 5.00 2.13E-01 5.00 

general 

laborer 1.28E-01 5.00 1.17E-01 5.00 1.06E-01 5.00 

Supervisor 6.38E-02 5.00 5.84E-02 5.00 5.31E-02 5.00 

Clerical 6.38E-03 5.00 5.84E-03 5.00 5.31E-03 5.00 

Recasting: 

Crucible Loading 

Operator 

Prior to 

1951 

1.60E+00 1.35 1.47E+00 1.35 1.34E+00 1.35 

general 

laborer 8.02E-01 1.35 7.35E-01 1.35 6.68E-01 1.35 

Supervisor 4.01E-01 1.35 3.67E-01 1.35 3.34E-01 1.35 

Clerical 4.01E-02 1.35 3.67E-02 1.35 3.34E-02 1.35 

Recasting: 

Recasting 

Operator 

Prior to 

1951 

2.45E+01 2.68 2.24E+01 2.68 2.04E+01 2.68 

general 

laborer 1.22E+01 2.68 1.12E+01 2.68 1.02E+01 2.68 

Supervisor 6.12E+00 2.68 5.61E+00 2.68 5.10E+00 2.68 

Clerical 6.12E-01 2.68 5.61E-01 2.68 5.10E-01 2.68 

Recasting: 

Crucible Burnout 

Operator 

Prior to 

1951 

5.10E-01 5.00 4.68E-01 5.00 4.25E-01 5.00 

general 

laborer 2.55E-01 5.00 2.34E-01 5.00 2.13E-01 5.00 

Supervisor 1.28E-01 5.00 1.17E-01 5.00 1.06E-01 5.00 

Clerical 1.28E-02 5.00 1.17E-02 5.00 1.06E-02 5.00 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 61 

Recasting: 

Billet Cleaning 

Operator 

Prior to 

1951 

1.20E+00 1.97 1.10E+00 1.97 1.00E+00 1.97 

general 

laborer 6.01E-01 1.97 5.51E-01 1.97 5.01E-01 1.97 

Supervisor 3.01E-01 1.97 2.76E-01 1.97 2.50E-01 1.97 

Clerical 3.01E-02 1.97 2.76E-02 1.97 2.50E-02 1.97 

Recasting: 

Crucible 

Assembly 

Operator 

Prior to 

1951 

8.45E+00 2.40 7.75E+00 2.40 7.04E+00 2.40 

general 

laborer 4.23E+00 2.40 3.87E+00 2.40 3.52E+00 2.40 

Supervisor 2.11E+00 2.40 1.94E+00 2.40 1.76E+00 2.40 

Clerical 2.11E-01 2.40 1.94E-01 2.40 1.76E-01 2.40 

Recasting: 

Crucible Loading 

Operator 

1951 and 

later 

3.64E-01 5.00 3.34E-01 5.00 3.04E-01 5.00 

general 

laborer 1.82E-01 5.00 1.67E-01 5.00 1.52E-01 5.00 

Supervisor 9.11E-02 5.00 8.35E-02 5.00 7.59E-02 5.00 

Clerical 9.11E-03 5.00 8.35E-03 5.00 7.59E-03 5.00 

Recasting: 

Recasting 

Operator 

1951 and 

later 

1.09E+00 2.42 1.00E+00 2.42 9.11E-01 2.42 

general 

laborer 5.46E-01 2.42 5.01E-01 2.42 4.55E-01 2.42 

Supervisor 2.73E-01 2.42 2.50E-01 2.42 2.28E-01 2.42 

Clerical 2.73E-02 2.42 2.50E-02 2.42 2.28E-02 2.42 

Recasting: 

Crucible Burnout 

Operator 

1951 and 

later 

3.28E-01 5.00 3.01E-01 5.00 2.73E-01 5.00 

general 

laborer 1.64E-01 5.00 1.50E-01 5.00 1.37E-01 5.00 

Supervisor 8.20E-02 5.00 7.51E-02 5.00 6.83E-02 5.00 

Clerical 8.20E-03 5.00 7.51E-03 5.00 6.83E-03 5.00 

Recasting: 

Billet Cleaning 

Operator 

1951 and 

later 

2.91E-01 5.00 2.67E-01 5.00 2.43E-01 5.00 

general 

laborer 1.46E-01 5.00 1.34E-01 5.00 1.21E-01 5.00 

Supervisor 7.29E-02 5.00 6.68E-02 5.00 6.07E-02 5.00 

Clerical 7.29E-03 5.00 6.68E-03 5.00 6.07E-03 5.00 

Recasting: 

Crucible 

Assembly 

Operator 

1951 and 

later 

1.82E+00 1.21 1.67E+00 1.21 1.52E+00 1.21 

general 

laborer 9.11E-01 1.21 8.35E-01 1.21 7.59E-01 1.21 

Supervisor 4.55E-01 1.21 4.17E-01 1.21 3.80E-01 1.21 

Clerical 4.55E-02 1.21 4.17E-02 1.21 3.80E-02 1.21 

Fluorination: 

Hex Loaders 

Operator 

All 

8.06E+01 1.38 7.39E+01 1.38 6.72E+01 1.38 

general 

laborer 4.03E+01 1.38 3.69E+01 1.38 3.36E+01 1.38 

Supervisor 2.01E+01 1.38 1.85E+01 1.38 1.68E+01 1.38 

Clerical 2.01E+00 1.38 1.85E+00 1.38 1.68E+00 1.38 

Fluorination: 

Operators 

Operator 

All 

2.27E+01 3.10 2.08E+01 3.10 1.89E+01 3.10 

general 

laborer 1.14E+01 3.10 1.04E+01 3.10 9.47E+00 3.10 

Supervisor 5.68E+00 3.10 5.21E+00 3.10 4.74E+00 3.10 

Clerical 5.68E-01 3.10 5.21E-01 3.10 4.74E-01 3.10 

Fluorination: 

Still Operators 

Operator 

All 

8.85E+00 2.07 8.12E+00 2.07 7.38E+00 2.07 

general 

laborer 4.43E+00 2.07 4.06E+00 2.07 3.69E+00 2.07 

Supervisor 2.21E+00 2.07 2.03E+00 2.07 1.84E+00 2.07 

Clerical 2.21E-01 2.07 2.03E-01 2.07 1.84E-01 2.07 

Fluorination: 

Central Loaders 

Operator 

All 

1.71E+01 1.38 1.57E+01 1.38 1.42E+01 1.38 

general 

laborer 8.54E+00 1.38 7.83E+00 1.38 7.12E+00 1.38 

Supervisor 4.27E+00 1.38 3.92E+00 1.38 3.56E+00 1.38 

Clerical 4.27E-01 1.38 3.92E-01 1.38 3.56E-01 1.38 

Scrap Recovery: 

Furnace Operator 

(Trays) 

Operator 

1951 and 

earlier 

2.99E+01 5.00 2.75E+01 5.00 2.50E+01 5.00 

general 

laborer 1.50E+01 5.00 1.37E+01 5.00 1.25E+01 5.00 

Supervisor 7.49E+00 5.00 6.86E+00 5.00 6.24E+00 5.00 

Clerical 7.49E-01 5.00 6.86E-01 5.00 6.24E-01 5.00 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 62 

Scrap Recovery: 

Furnace Operator 

(Trays) 

Operator 

1952 and 

later 

2.99E+00 5.00 2.74E+00 5.00 2.49E+00 5.00 

general 

laborer 1.49E+00 5.00 1.37E+00 5.00 1.24E+00 5.00 

Supervisor 7.47E-01 5.00 6.85E-01 5.00 6.22E-01 5.00 

Clerical 7.47E-02 5.00 6.85E-02 5.00 6.22E-02 5.00 

Scrap Recovery: 

Furnace Operator 

(Calciner) 

Operator 

1952 and 

later 

9.98E+00 5.00 9.15E+00 5.00 8.32E+00 5.00 

general 

laborer 4.99E+00 5.00 4.58E+00 5.00 4.16E+00 5.00 

Supervisor 2.50E+00 5.00 2.29E+00 5.00 2.08E+00 5.00 

Clerical 2.50E-01 5.00 2.29E-01 5.00 2.08E-01 5.00 

Scrap Recovery: 

Digest Operator 

Operator 

All 

2.91E-01 5.00 2.67E-01 5.00 2.43E-01 5.00 

general 

laborer 1.46E-01 5.00 1.34E-01 5.00 1.21E-01 5.00 

Supervisor 7.29E-02 5.00 6.68E-02 5.00 6.07E-02 5.00 

Clerical 7.29E-03 5.00 6.68E-03 5.00 6.07E-03 5.00 

Scrap Recovery: 

Filtration Operator 

Operator 

All 

1.09E-01 4.94 1.00E-01 4.94 9.11E-02 4.94 

general 

laborer 5.46E-02 4.94 5.01E-02 4.94 4.55E-02 4.94 

Supervisor 2.73E-02 4.94 2.50E-02 4.94 2.28E-02 4.94 

Clerical 2.73E-03 4.94 2.50E-03 4.94 2.28E-03 4.94 

Drum Loading 

Operations 

Operator 

All 

N/A N/A N/A N/A N/A N/A 

general 

laborer N/A N/A N/A N/A N/A N/A 

Supervisor N/A N/A N/A N/A N/A N/A 

Clerical N/A N/A N/A N/A N/A N/A 
 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 63 

9.0 References 

AEC (U.S. Atomic Energy Commission). 1949. Simonds Saw & Steel Co. Summary Report of Three 

Surveys (Oct. 27, 1948 – Feb. 15, 1949). 1949. [pp. 249–254 of Simonds Saw and Steel Company.pdf; 

Reference ID #11996] 

Anderson, JL and Hertel, NE. 2005. Bremsstrahlung doses from natural uranium ingots. Radiation 

Protection Dosimetry 115(1-4:298-301. 

Strom DJ. 2006. Default Assumptions and Methods for Atomic Weapons Employer Dose Reconstruction. 

Battelle TIB-5000 rev. 00-A.14, Battelle, Richland, Washington. 

Christofano E and WB Harris. 1960. The Industrial Hygiene of Uranium Refining. Env Health 1(5):74­

96. 

DOE (U.S. Department of Energy). 2003. Recycled Uranium: United States Production, Enrichment, 

and Utilization. DOE/SO-0003. Available on the web at: 

http://www.ssa.doe.gov/sp70/documents/Recycled%20Uranium%20May%202003.pdf . 

DOE (U.S. Department of Energy). 2004. Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities. DOE-STD-1136-2004, December. 

Eckerman, KF and JC Ryman. 1993. External exposure to radionuclides in air, water, and soil. Federal 

Guidance Report No. 12. EPA-402-R-93-081. Environmental Protection Agency. Washington, DC. 

Eisenbud M and JA Quigley. 1956. Industrial Hygiene of Uranium Processing. AMA Arch Ind Health 

14(1):12-22. 

Eisenbud, M. 1975. Early Occupational Exposure Experience with Uranium Processing. In Health 

Experience with Uranium. Proceedings of a United States Atomic Energy Commission conference. 

Arlington, Virginia: ERDA-93; 1975: 9-24 

EPA (Environmental Protection Agency). 2002. Federal Guidance Report 13. Cancer Risk Coefficients 

for Environmental Exposure to Radionuclides: CD Supplement. EPA-402-C-99-001, Rev. 1. U.S. 

Environmental Protection Agency, Washington, DC. 

Grove Engineering. 2003. MicroShield Software Program, Version 6.02. Copyright 1992-2003. 

ICRP (International Commission on Radiological Protection). 1996. Conversion Coefficients for use in 

radiological protection against external radiation. ICRP Publication 74. Annals of the ICRP 26(3-4). 

ICRP (International Commission on Radiological Protection). 1994. Human Respiratory tract model for 

radiological protection. ICRP Publication 66. Annals of the ICRP 24(1-3). 

Kocher DC. 1981. Radioactive Decay Data Tables. DOE/TIC-11026, National Technical Information 

Service, Springfield, Virginia. 

Lide, DR. 1995. CRC Handbook of Chemistry and Physics, 75th edition. CRC Press, Boca Raton, FL 
 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 64 

NIOSH (National Institute for Occupational Safety and Health). 2002. External Dose Reconstruction 

Implementation Guideline. OCAS-IG-001, Rev. 1, Office of Compensation Analysis and Support, 

Cincinnati, Ohio. 

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

Technical Information Bulletin. OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and Support, 

Cincinnati, Ohio, April 13. 

NIOSH (National Institute for Occupational Safety and Health), 2005. Estimating the Maximum 

Plausible Dose to Workers at Atomic Weapons Employer Facilities. Technical Information Bulletin, 

OCAS-TIB-004, Rev. 2, Office of Compensation Analysis and Support, Cincinnati, Ohio, November 18. 

NRC (U.S. Nuclear Regulatory Commission), 2002a. Technical Basis for Calculating Radiation Doses 

for the Building Occupancy Scenario Using the Probabilistic RESRAD-BUILD 3.0 Code. NUREG/CR6755, Washington, D.C. 

NRC (U.S. Nuclear Regulatory Commission). 2002b. Re-Evaluation of the Indoor Resuspension Factor 

for the Screening Analysis of the Building Occupancy Scenario for NRC’s License Termination Rule. 

NUREG-1720; Draft Report for Comment, Washington, D.C. 

ORAUT (Oak Ridge Associated Universities Team). 2004a. Technical Basis Document: Basis for 

Development of an Exposure Matrix for Bethlehem Steel Corporation, Lackawanna, New York: Period of 

Operation: 1949-1952. ORAUT-TKBS-0001, Revision 01, Cincinnati, Ohio. 

ORAUT (Oak Ridge Associated Universities Team). 2004b. Technical Basis Document for the Fernald 

Environmental Management Project (FEMP) – Occupational Internal Dose. ORAUT-TKBS-0017-5, 

Revision. 00, Cincinnati, OH. 

ORAUT (Oak Ridge Associated Universities Team). 2004c. Technical Basis Document for the Fernald 

Environmental Management Project – Occupational External Dosimetry. ORAUT-TKBS-0017-6, 

Revision 00, June 2004. 

ORAUT (Oak Ridge Associated Universities Team). 2004d. Technical Basis Document for Rocky Flats 

Plant – Occupational External Dosimetry. ORAUT-TKBS-0011-6, Revision 00, January 2004. 

ORAUT (Oak Ridge Associated Universities Team). 2005a. Basis for development of an exposure 

matrix for the Mallinckrodt Chemical Company St. Lewis downtown site and the St. Lewis airport site, 

St. Louis, Missouri, period of operation: 1942-1958. ORAUT-TKBS-0005, Revision 01. Cincinnati, 

Ohio. 

ORAUT (Oak Ridge Associated Universities Team). 2005b. Site Profile for Simonds Saw and Steel. 

ORAUT-TKBS-0032, July 2005. 

ORAUT (Oak Ridge Associated Universities Team). 2005c. Estimating the Maximum Plausible Dose to 

Workers at Atomic Weapons Employer Facilities. ORAUT-OTIB -0004, Revision 3, August 2005. 

ORAUT (Oak Ridge Associated Universities Team). 2005d. Dose Reconstruction from Occupationally 

Related Diagnostic X-ray Procedures. ORAUT-OTIB-0006, Rev. 3. 
Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 65 

ORAUT (Oak Ridge Associated Universities Team). 2005d. An exposure matrix for Superior Steel, 

Carnegie, Pennsylvania, period of operation: January 1, 1952 through December 31, 1957. ORAUTTKBS-0034 Revision 00 PC-1. 

ORAUT (Oak Ridge Associated Universities Team). 2006. Internal Dosimetry Organ, External 

Dosimetry Organ, and IREP Model Selection by ICD-9 Code. ORAUT-OTIB-0005, Rev. 02 PC-1. 

Pelowitz, DB (editor). 2005. MCNPXTM Users Manual, Version 2.5.0. LA-CP-05-0369, Los Alamos 

National Laboratory, Los Alamos, NM (OUO). 

Rich BL, SL Hinnefeld, CR Lagerquist, WG Mansfield, LH Munson and ER Wagner. 1988. Health 

Physics Manual of Good Practices for Uranium Facilities. EGG-2530. Idaho National Engineering 

Laboratory, Idaho Falls, ID. 

Strom, DJ. 2006. Default Assumptions and Methods for Atomic Weapons Employer Dose 

Reconstructions. Battelle-TIB-5000 Rev. 00-A.14 (draft). Battelle, Pacific Northwest Division, Richland 

WA. 

Wilson, WB., RT Perry, WS Charlton, TA Parish, GP Estes, TH Brown, ED Arthur, M Bozoian, TR 

England, DG Madland and JE Stewart. 1999. Sources 4A: A code for Calculating (α-n), Spontaneous 

Fission, and Delayed Neutron Sources and Spectra. LA-13639-MS, Los Alamos National Laboratory, 

Los Alamos, NM. 

Thompson, J. 2003. Recycled Uranium: Establishing Confidence in Uranium Production and Shipment 

Data. DOE/SO-0005. Available on the web at: 

http://www.ssa.doe.gov/sp70/documents/Recycled%20Uranium%20July%202003.pdf . 
 

 

 

 

Document No. Battelle-TBD-6001 Revision No. F0 Effective Date: 12/13/2006 page 66 

10.0 Glossary 

Black Oxide Common name for U3O8 

Brown Oxide Common name for UO2 

Green Salt Common name for UF4 

Orange Oxide Common name for UO3 

Yellow Cake Common name for U3O8 

BZ Breathing Zone samples 

GA General Area samples 

TWA Time Weighted Average. The concentration in ppm or mg/cm3

 of a chemical component 

multiplied by the time of each individual sampling period, summed for all samples taken during an 

interval and divided by the total sampling time. 

DWA Daily Weighted Average – the sum of the products of the air concentration and the 

duration of all tasks performed as part of a job divided by the duration of the shift. 

UX1	 234Th 

234m UX2 Pa