ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 114

Document Title:

An Exposure Matrix for the Harshaw Chemical 

Company, Cleveland, Ohio

Document Number: ORAUT-TKBS-0022

Revision: 00

Effective Date: 08/17/2007

Type of Document: TBD

Supersedes: None

Subject Expert(s): Janet L. Westbrook, Cindy W. Bloom, and Eugene W. Potter

Site Expert(s): N/A

Approval: Richard E. Merrill Signature on File for Approval Date: 08/16/2007

Eugene W. Potter, Document Owner

Approval: Signature on File Approval Date: 08/09/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/14/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 08/15/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 08/17/2007

James W. Neton, Associate Director for 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. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 2 of 114

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

08/17/2007 00 New Technical Basis Document: An Exposure Matrix for the 

Harshaw Chemical Company, Cleveland, Ohio. First approved 

issue. Incorporates formal internal and NIOSH review comments. 

Adds coworker data, SEC, and Attributions and Annotations section. 

Training required: As determined by the Task Manager. Initiated by 

Eugene W. Potter. 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 3 of 114

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

7

1.0 Introduction ..............................................................................................................................9

1.1 Purpose........................................................................................................................9

1.2 Scope.........................................................................................................................10

2.0 Site Description......................................................................................................................10

2.1 Introduction.................................................................................................................10

2.2 Site Activities ..............................................................................................................11

2.3 Site Processes ........................................................................................................... 14

2.3.1 Uranium Tetrachloride Production................................................................... 14

2.3.2 Uranium Hexafluoride Production ................................................................... 14

2.3.3 Uranium Tetrafluoride Production ................................................................... 14

2.3.4 Uranium Trioxide and Uranium Dioxide Production......................................... 15

2.3.5 Uranyl Nitrate Hexahydrate Production ........................................................... 15

2.3.6 Operations Involving Other Radiological Materials .......................................... 15

2.4 Incidents.....................................................................................................................16

2.5 Summary of Potential Exposures................................................................................ 17

3.0 Occupational Medical Dose.................................................................................................... 17

3.1 Introduction.................................................................................................................17

3.2 Examination Frequencies ........................................................................................... 18

3.3 Equipment and Techniques ........................................................................................ 18

3.4 Organ Dose Calculations............................................................................................ 19

4.0 Occupational Environmental Dose ......................................................................................... 19

4.1 Introduction.................................................................................................................19

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations ........................ 19

4.3 External Dose............................................................................................................. 21

5.0 Occupational Internal Dose.................................................................................................... 21

5.1 Introduction.................................................................................................................21

5.2 Uranium Solubility and Particle Size ........................................................................... 21

5.3 In Vitro Minimum Detectable Activities, Counting Methods, and 

Reporting Protocols .................................................................................................... 22

5.3.1 In Vitro Urine Analysis..................................................................................... 22

5.3.1.1 Early Urine Studies...........................................................................22

5.3.1.2 Routine Urine Program.....................................................................23

5.3.1.3 Minimum Detectable Activities ..........................................................24

5.3.2 In Vitro Methods for Uranium .......................................................................... 24

5.4 Other Bioassay Methods ............................................................................................ 25

5.5 Radon Levels during Operations ................................................................................ 25

5.6 Airborne Radioactive Dust Measurements.................................................................. 26

5.7 Postoperations Internal Dose...................................................................................... 27

5.8 Activity Fractions ........................................................................................................ 28

5.9 Determination of Internal Doses (Instructions to Dose Reconstructors) ...................... 28
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5.9.1 Determining Annual Inhalation Intakes Based on TimeWeighted Daily Average Inhalation Exposure Data (Table B16), 8/14/1942-11/30/1949.............................................................................. 29

5.9.2 Determining Annual Ingestion Intakes Based on TimeWeighted Daily Average Inhalation Exposure Data (Table B17), 8/14/1942-11/30/1949.............................................................................. 30

5.9.3 Estimating Inhalation and Ingestion Intakes by Using 

Time-Weighted Daily Average Exposure Data (Table B-15), 

8/14/1942-11/30/1949..................................................................................... 30

5.9.4 Estimating Annual Radon Exposure by Using Bounding Radon 

Exposure Data (Table B-18), 12/1/1949 on..................................................... 31

5.9.5 Estimating Intakes During the D&D/Postoperations Years by 

Using Calculated Data (Table B-25)................................................................ 31

6.0 Occupational External Dose................................................................................................... 32

6.1 Introduction.................................................................................................................32

6.2 Basis of Comparison .................................................................................................. 32

6.3 Dose Reconstruction Parameters ............................................................................... 33

6.3.1 Site Historical Administrative Practices ........................................................... 33

6.3.2 Site Dosimetry Technology and Calibration..................................................... 35

6.3.3 Workplace Radiation Fields............................................................................. 36

6.3.3.1 Beta/Photon Dosimeter Response....................................................36

6.3.3.2 Uncertainty and Bias for Beta/Photon Dosimeters ............................36

6.3.3.3 Neutron Doses..................................................................................37

6.4 Adjustments to Recorded Dose .................................................................................. 38

6.5 Missed Dose...............................................................................................................38

6.6 Determination of External Doses (Instructions to Dose 

Reconstructors).......................................................................................................... 38

6.7 Determining Exposure During the Operations Years .................................................. 39

6.7.1 Reconstruction of Doses from August 1944 to 1955 ....................................... 39

6.7.2 Reconstruction of Doses Prior to August 1944................................................ 40

6.7.3 Estimating Incidental Dose for Individuals Employed In 

Uranium Processing but Not Involved in Operations ....................................... 40

6.8 Determining Exposure During the Postoperations and D&D Years............................. 41

7.0 Attributions and Annotations .................................................................................................. 41

References ........................................................................................................................................46

Glossary ............................................................................................................................................57

ATTACHMENT A, HARSHAW INTERNAL COWORKER DATA ASSESSMENT............................... 63

ATTACHMENT B, TABLES IMPORTANT TO DOSE RECONSTRUCTION ...................................... 76
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LIST OF TABLES

TABLE TITLE

2-1 Assumed periods of D&D and postoperations use .................................................................

PAGE

13

2-2 Operations, postoperations, and D&D activity ........................................................................ 13

2-3 Assumed fractions of recycled uranium contaminants at metal handling facilities .................. 16

5-1 Air-sampling information......................................................................................................... 22

5-2 Radon levels in the Mallinckrodt plants in the later period of operation................................... 26

5-3 Production of UF4 and UF6 and the start of dust sampling...................................................... 27

5-4 Comparison of RESRAD-calculated and DOE (1984)-measured radon 

concentrations........................................................................................................................28

5-5 Bioassay and alpha air sample activity fractions by period and radionuclide .......................... 28

5-6 Chronic intake rates for types F, M, and S 234U ...................................................................... 29

A-1 Parameters for Harshaw data verification............................................................................... 65

A-2 MIL-STD-105E 4% AQL sampling of Harshaw bioassay data files ......................................... 66

A-3 Uranium mass urinary excretion data..................................................................................... 68

A-4 Uranium activity urinary excretion data................................................................................... 69

A-5 Chronic intake rates for type F 234U ........................................................................................ 70

A-6 Chronic intake rates for type M 234U ....................................................................................... 70

A-7 Chronic intake rates for type S 234U........................................................................................ 70

B-1 Chronology of Harshaw site operations based on available reference material ...................... 77

B-2 Buildings known to have been used at the Harshaw site for uranium processing 

work .......................................................................................................................................79

B-3 Types and quantities of materials used and produced in Harshaw uranium 

processing .............................................................................................................................80

B-4 Functional and process keywords and codes......................................................................... 86

B-5 Measured dose rates ............................................................................................................. 88

B-6 Chest and hand beta doses from ash residue handling, as measured by films....................... 92

B-7 Weekly dose rates for various workers and areas .................................................................. 93

B-8 Annual neutron whole-body doses from the alpha-neutron reaction, various 

uranium forms........................................................................................................................93

B-9 Reserved ...............................................................................................................................94

B-10 Air concentrations in various areas in the green and hex plants............................................. 94

B-11 General area and breathing zone dust measurements in the green and hex 

plants (491 and 492) ..............................................................................................................95

B-12 General area and breathing zone dust measurements in the brown plant (493)..................... 97

B-13 Dust measurements in the green and brown plants................................................................ 99

B-14 Daily DWEs to airborne dust .................................................................................................. 99

B-15 Daily DWEs to airborne dust for various job titles................................................................. 100

B-16 Annual inhalation intakes based on daily weighted average exposures to airborne 

dust for various job titles....................................................................................................... 102

B-17 Annual ingestion intakes based on daily weighted average exposures to airborne 

dust for various job titles....................................................................................................... 103

B-18 Annual radon exposure ........................................................................................................ 104

B-19 General area dust concentrations in the locker rooms and lunchrooms................................ 104

B-20 Average measured dust and urine concentrations by month................................................ 105

B-21 Number of workers............................................................................................................... 105

B-22 Job titles, functions, and appropriate absorption types ......................................................... 106

B-23 Results of a survey of the Harshaw site by Argonne National Laboratory, 1976 to 

1979..................................................................................................................................... 108
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B-24 Source terms used to calculate inhalation and radon doses, D&D/postoperations 

period................................................................................................................................... 109

B-25 Annual inhalation, radon, and ingestion doses, D&D/postoperations period......................... 109

B-26 Annual external dose rates and doses, D&D/postoperations period..................................... 113

LIST OF FIGURES

FIGURE TITLE

A-1 Type F 50th percentile for 12/1/1949 to 12/31/1953 ...............................................................

PAGE

70

A-2 Type M 50th percentile for 12/1/1949 to 12/31/1953 .............................................................. 71

A-3 Type S 50th percentile for 12/1/1949 to 3/31/1950................................................................. 71

A-4 Type S 50th percentile for 4/1/1950 to 12/31/1951................................................................. 72

A-5 Type S 50th percentile for 1/1/1952 to 12/31/1953................................................................. 72

A-6 Type F 84th percentile for 12/1/1949 to 12/31/1953 ............................................................... 73

A-7 Type M 84th percentile for 12/1/1949 to 12/31/1953 .............................................................. 73

A-8 Type S 84th percentile for 12/1/1949 to 3/31/1950................................................................. 74

A-9 Type S 84th percentile for 4/1/1950 to 12/31/1951................................................................. 74

A-10 Type S 84th percentile for 1/1/1952 to 12/31/1953................................................................. 75
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 7 of 114

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANL Argonne National Laboratory

AQL Acceptance Quality Level

AWE atomic weapons employer

cGy centigray

Ci curie

cm centimeter

d day

D&D decontamination and decommissioning

DCF dose conversion factor

DOE U.S. Department of Energy

dpm disintegrations per minute

DWE daily weighted (average) exposure

ECI Export Controlled Information

EEOICPA Energy Employees Occupational Illness Compensation Program Act

F2 fluorine gas

ft foot

FUSRAP Formerly Utilized Sites Remedial Action Program

g gram

gal gallon

GI gastrointestinal

GSD geometric standard deviation

H2 hydrogen gas

HASL Health and Safety Laboratory

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

keV kiloelectron-volt, 1,000 electron-volts

KOH potassium hydroxide

kVp applied kilovoltage; peak kilovoltage

L liter

lb pound

LOD limit of detection

m meter

mCi millicurie

MCW Mallinckrodt Chemical Works

MED Manhattan Engineer District

MeV megavolt-electron, 1 million electron-volts

mi mile
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 8 of 114

min minute

mm millimeter

mo month

mR milliroentgen

mrem millirem

mrep millirep

NG Not Good, aqueous phase of the ether extraction [liquor] that contained a fair 

percentage of uranium

NIOSH National Institute for Occupational Safety and Health

NYOO New York Operations Office (AEC)

oz ounce

PA posterior-anterior

pCi picocurie

PER Program Evaluation Report

ppb parts per billion

Q quarter (used to indicate portion of calendar year in several tables) 

qt quart

R&D research and development

RU recycled uranium

s second

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TBP tributyl phosphate

U.S.C. United States Code

UNH uranyl nitrate hexahydrate

WL working level

WLM working level month

yr year

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 9 of 114

1.0

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of

dose reconstructions at particular sites or categories of sites. They will be revised in the event that 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

INTRODUCTION

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 [AWE] facility" or a "Department of Energy [DOE] facility" as defined in the Energy 

Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. § 7384l 

(5) and (12)]. EEOICPA, as amended, provides for employees who worked at an AWE facility during 

the contract period and/or during the residual contamination period.

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 during which 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

• 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- or U.S. Atomic Energy Commission (AEC)-

related work) must be included in dose reconstructions. Radiation dose received from DOE/AECrelated 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 42 U.S.C. § 

7384n(c)(4), radiation associated with the Naval Nuclear 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 nonweapons-related work, if applicable, will be covered elsewhere.

1.1 PURPOSE

This document establishes the technical basis for the reconstruction of radiation doses to workers at 

the Harshaw Chemical Company's Harvard-Denison Plant at 1000 Harvard Avenue in Cleveland, 

Ohio, which received feed materials from uranium mills throughout the United States and Canada 

(DOE 1984) and refined it to produce various uranium compounds under contract to the U.S.

Government from 1942 to 1955.

A Special Exposure Cohort (SEC) class established for Harshaw includes all AWE employees who 

were monitored or should have been monitored while working at the Harshaw Harvard-Denison Plant 

for a number of workdays aggregating at least 250 workdays from August 14, 1942, through 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 10 of 114

November 30, 1949, or in combination with workdays within the parameters established for one or 

more other classes of employees in the SEC (Elliott 2007). This document also provides guidance for 

EEOICPA-covered employees who participated in Harshaw operations, specifically for non-SEC 

cancers and those presumptive cancer claims for workers who have fewer than 250 workdays under 

this employment or in combination with workdays within the parameters established for other classes 

of employees in the SEC.

1.2 SCOPE

This document covers workers at the Harshaw Harvard-Denison Plant at 1000 Harvard Avenue in 

Cleveland, Ohio, principally in Plant C and several additional buildings listed in Table B-2 (e.g., 

Foundry, K-1, and an Annex facility), which were associated with the chemical conversion and 

production of various uranium compounds. Some early uranium tetrachloride (UCl4) laboratory-scale 

work appears to have been done at a Harshaw laboratory at 1945 East 97th Street (Gamertsfelder ca. 

1944), but little else is known about this facility. (Therefore, while the known exposure rate and 

position information are included in Table B-5, more investigation could be needed about the location 

and use of this facility and it is not included in the scope of this site profile.) In addition, a site 

radiological survey conducted in the 1980s (DOE 1984) indicated that radiological contamination was 

detected inside and outside several other structures at the main plant site.

This document covers the period from the start of contract operations for the AEC and its predecessor 

agency [the Manhattan Engineer District (MED)] through the cessation of operations. In addition, this 

document discusses the periods of decontamination and decommissioning (D&D) for each operational 

area and the period from AEC release of the site to the point where the Formerly Utilized Sites 

Remedial Action Program (FUSRAP) took over decontamination responsibilities for the parts of the 

site at which eligible operations had been performed. Table B-1 lists a detailed site chronology. In 

general, the period of production operations was from January 1942 to about September 1955. 

However, the period of AWE operations started on August 14, 1942, with the establishment of the 

MED. The period of D&D under AEC supervision was from November 1951 to an unspecified point in 

1960 when Harshaw received an unrestricted release from AEC (OEPA 2001). The period of free 

release and non-AEC use until FUSRAP took over is considered to be from 1960 until June 1999

when DOE (which was then in charge of FUSRAP) issued a letter stating that Harshaw was a 

candidate site for remediation and that it would support preliminary characterization by providing 

documentation (Fiore 1999). The latter period includes the time after the purchase of the site by the 

Engelhard Corporation in 1988. Dose reconstructions could be needed from August 14, 1942, to the 

present.

This site profile consists of six sections: (1) Introduction, (2) Site Description, (3) Occupational Medical 

Dose, (4) Occupational Environmental Dose, (5) Occupational Internal Dose, and (6) Occupational 

External Dose. Attachment A describes the assessment of internal coworker data from Harshaw. 

Attachment B contains the tables important to dose reconstruction.

2.0

2.1 INTRODUCTION

SITE DESCRIPTION

Table B-1 lists the chronology of site use. Figure 2-1 shows the buildings that existed at the Harshaw 

site during a radiological survey in 1984 (DOE 1984). Layouts of the uranium tetrafluoride (UF4) and 

uranium hexafluoride (UF6) areas can be found in AEC (1948, 1951a) and that of the ore concentrate

to uranium dioxide (UO2) area in AEC (1951b).
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 11 of 114

Figure 2-1. Harshaw Chemical Company identifying area 

designations during FUSRAP Site Survey (DOE 1984).

Little information about the early history of site use is available. The principal source of information is 

an account by the president of Harshaw of the early work under contracts with the MED or other 

entities that were working for the MED (Harshaw 1945). Harshaw was first approached in 1941 about 

Office of Scientific Research and Development/MED-associated work when it was asked to produce 

UF6 using fluorine gas (F2) based on the company’s experience manufacturing anhydrous hydrofluoric 

acid and laboratory experience in manufacturing F2 (Harshaw 1945). Harshaw had limited experience 

working with uranium, having manufactured uranium-containing ceramic glazes and similar products.

2.2 SITE ACTIVITIES

The entire Harshaw Chemical Company operational area for AEC work was referred to as “Plant C” or 

“Area C” (Velten 1949). Plant C (refinery building also referred to as Building G-1) included all of the 

individual production operations using radiological materials (UO2-to-UF4, UF4-to-UF6 plus distillation, 

and ore-to-UO2 or -UO3). The locations where these three individual production operations were 

performed were also referred to as “plants.” The common usage of the time might have referred to 

the ore-to-UO2 operations as the "brown,” "new brown,” or “493” plant; the UF4-to-UF6 operations as 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 12 of 114

the “hex” or “492” plant; and the UO2-to-UF4 operations as the “green” or “491” plant. In 1945, UF6

operations were expanded in an annex to Building G-1, called the “hex annex.”

Uranium production operations for the MED war effort began in August 1942 when Harshaw produced 

laboratory quantities of UCl4. This work would have been done using ore material. Generally 

speaking, this material would have been in the form of triuranium octaoxide (U3O8), sometimes 

referred to as black oxide or yellow cake. Harshaw also produced UCl4 for the National Bureau of 

Standards. After 1943, other research contracts followed, including a contract to research methods of 

improving uranium capture and recovery (Harshaw 1945).

From March 1942 until 1951, Harshaw produced UF6 (DOE 1997; Stefanec 1951). In 1944, under 

contract to the MED, Harshaw built and operated a full-scale UF6 pilot plant (Harshaw 1945).

In May 1949, the AEC contracted with Harshaw to convert uranium concentrate to uranium trioxide 

(UO3) (also called orange oxide) and then to UO2 (NYOO Medical Division 1949; Mayer and Proschan 

1949). In 1951, AEC discontinued conversion of UO3 to UO2 and production of UO2 ceased. At that 

point, Harshaw began to produce UO3 only from milled ore (DOE 1997). Production of UO3 continued 

until about September 1955, when the plant was shut down.

D&D activities were conducted under AEC supervision starting in November 1951. Decontamination 

of the UF4 and UF6 areas began after early December 1951 (Clarke 1963). Dismantling of equipment 

in the UF6 area and probably the UF4 area began in January 1952 (HCC 1950–1953). In about 

February 1952, an area of the Building G-1 annex, which had held an extension of the UF6 production 

area and also housed some offices and shops, was torn out to make a warehouse, forcing the 

relocation of various offices and shops to other existing areas (Klevin 1952a). AEC equipment was 

removed by September 30, 1955 (NIOSH 2002). AEC D&D surveys of the various production areas 

were performed since November 1951 (Klevin 1953a). The buildings were decontaminated by 

Harshaw in the late 1950s and released from AEC control in 1960 (FUSRAP 2001).

Although Building G-1 and its annex were subjected to D&D from 1952 to 1959 for the older areas 

and 1956 to 1959 for the newer areas (the brown area), it appears that other buildings that had some 

residual contamination – either from work there in the MED years, such as the foundry and Building K1, or from track-in, such as the Boiler House – were not subject to D&D in these time periods. After 

release of the site by the AEC in 1960, Harshaw used all the buildings until the site was purchased by 

the Engelhard Corporation in 1988. Engelhard then did D&D between 1990 and 1997 on all buildings 

except (apparently) Plant C, but did not seem to have used Plant C. Thus, the postoperations years 

would be from 1960 to 1989 for all buildings and the Engelhard D&D years would be 1990 to 1992 for 

those buildings that were demolished (Buildings K-1, M-1, and P-1); 1990 to 1992 for the Boiler House 

(used to store parts of the demolished buildings); and 1990 to 1997 for the remaining buildings other 

than Plant C. Table 2-1 lists the assumed periods of D&D and postoperations use. Note that it is 

assumed that those buildings that were not in use were kept locked and access to them was 

controlled, as is suggested or stated explicitly by all relevant references.

The D&D of the green area, the "old hex" area, and the distillation and recovery areas that took place 

from 1952 to 1959 was not continuous [1]. However, it appears that no operations were performed 

there because the AEC did not release the site until 1960, and it should thus be assumed that this 

period was entirely a D&D period for these areas. The D&D of the "new hex" or annex area that took 

place in 1952 left some residual contamination because the space was to be converted to other uses 

associated with the ore concentrate to UO3 process. This work was true D&D work and the workers 

engaged in it likely had a notation in film badge records such as "decontamination," but the space was 

not completely clean afterward and was used for process-associated work such as warehousing and 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 13 of 114

Table 2-1. Assumed periods of D&D and postoperations use.

Year UF4 area

Old UF6

area

Recovery

/still area

New UF6

/UO3 area 

UO3-Only 

area Lab

Boiler 

house Foundry Garage Warehouse P-1 K-1 M-1

1942-1951

3Q 1944-

3Q 1951

3Q 1944-

4Q 1951

3Q 1944-

4Q 1951

3Q 1944-

4Q 1951 2Q 1949- 3Q 1944-

Start of 

use? 1942-?

Start of 

use?

Converted 

from New 

Hex, 1952?

Start 

of 

use?

Start 

of 

use?

Start 

of 

use?

1952 1Q 1Q 1Q 1Q-3Q

1953-1954

1955 3Q1955 3Q 1955 3Q 1955

1956 4Q 4Q 4Q?

1957-1959

1960-1989

1990-1992 1Q 1Q 1Q 1Q 1Q 1Q 1Q

1993-1997

1998

Shading legend:

In use D&D period Nonuse (locked) Demolished

Q means calendar quarter

storage of contaminated items. Thus, while the post-D&D exposure levels in this area probably did 

not approach the exposure levels of the process area, there was still the potential for non-negligible 

exposure [2].

The period of free release and non-AEC use until FUSRAP took over is considered to be from 1960 

until June 1999. FUSRAP work at the site is not included in the scope of this site profile. Table 2-2 

provides a summary of the operations, postoperations, and D&D history to clarify the application of 

calculated dose estimates.

Table 2-2. Operations, postoperations, and D&D activity.

Area Operations

AEC 

decontamination

Use after 

AEC 

operations 

and/or D&D

Final (Engelhard) 

decontamination

Use after final 

decontamination

UO2-UF4 (491, Green) 1944–1951 1952–1959 1960–1989 ---(a) ---

UF4-UF6 (492, Old hex) 1944–1951 1952–1959 1960–1989 --- ---

Recovery/Still (492, Distillation) 1944–1951 1952–1959 1960–1989 --- ---

UF4-UF6 annex (492, New hex) 1952–1959 1960–1989 --- ---

 UF6 production 1948–1951 1952 --- --- ---

 Storage for UO3 production 1953–1955 1956–1959 1960–1989 --- ---

Laboratory 1944–1955 1956–1959 1960–1989 --- ---

Boiler House 1942?–1955 --- 1956–1989 1990–1997 ---

Foundry 1942–1955 --- 1956–1989 1990–1997 1998 on

Garage 1942?–1955 --- 1956–1989 1990–1997 1998 on

Warehouseb 1953–1955 --- 1956–1989 1990–1997 1998 on

Building K-1 1942–1955 --- 1956–1989 1990–1992 ---

Building M-1 1942?–1955 --- 1956–1989 1990–1992 ---

Building P-1 1942?–1955 --- 1956–1989 1990–1992 ---

a. Table entries that are blank (---) represent periods of inactivity.

b. Assumed to be different from the New hex annex, which was partly converted to extra warehouse space in 1952.

Table B-21 gives information about the number of workers based on AEC dust study reports. Job 

titles should be listed in employment records for individual workers, but these might not correspond to 

job titles or work descriptions in the badge and bioassay records because informal terms or area 

references were often used in the latter. Table B-4 (keywords and codes used) and Table B-22 (job 

titles and functions) should be used to help establish the actual category to be used for an individual 

worker with respect to later tables in this site profile and with respect to interpreting records.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 14 of 114

Most of the Harshaw uranium processing operations were carried out 24 hr/d, 7 d/wk (Lippmann 

1958). AEC dust study reports [e.g., Klevin (1950a) and Lippmann (1958)] indicate that the workday 

for all workers was 8 hours. Klevin (1950a) indicated that there were three shifts per day for every 

type of listed worker except the plant superintendent, general foreman, process engineer, health 

physicist, and laundry workers. The Harshaw operating manual (HCC 1946) stated that the Timken 

Roller Bearing Company plan of rotation was used: there were four crews for each major process or 

activity, with each crew working five 8-hour days for 3 weeks, then working six 8-hour days the fourth 

week, for an average of 42 hr/wk. This was necessary so the plant could operate 24 hr/d, 7 d/wk.

In addition to Harshaw employees, employees from at least two subcontractors were mentioned in 

site documents, M. K. Ferguson (Klevin ca. 1948) and General Welding (AEC 1950a; HCC 1950–

1953).

2.3 SITE PROCESSES

2.3.1

Harshaw shipped its first order of UCl4 to the National Bureau of Standards in March 1942. Harshaw 

began larger scale laboratory production of UCl4 in November 1942; by April 1943, Harshaw was 

producing up to 100 lb of UCl4 daily (Harshaw 1945). Harshaw set up a new production area in 

October 1944. This new production area was initially used to process up to 1 ton per day of UCl4, with 

the production rate continuing to increase monthly thereafter (Harshaw 1945; Ferry 1944a; MED 

1945). In January 1945, MED ordered an additional amount of 65,000 lb of UCl4. This was the final 

order of the material (Simmons 1945). Harshaw stopped production of UCl4 in February 1945 and 

dismantled the UCl4 production area. Parts were shipped to Oak Ridge, Tennessee (Harshaw 1945).

Uranium Tetrachloride Production

2.3.2

In February 1942, Harshaw Chemical Company first produced UF6 and maintained a production rate 

of 5 lb/d throughout 1942. This material was sometimes referred to as “hex” and the processing area 

as the “hex” or “492” plant.” By 1943, Harshaw was producing as much as 50 lb of UF6 per day in a 

pilot plant. Harshaw operated the pilot plant until February 1944, producing a total of 9,000 lb of UF6

(Harshaw 1945).

Uranium Hexafluoride Production

In 1944, Harshaw built a new UF6 production facility containing three units. Harshaw erected 

electrolytic cells to obtain the fluorine that was used to produce UF6 (Quigley 1951a). By July 1944, 

Harshaw was producing as much as 3,300 lb of UF6 per day, and as much as 4,500 lb/d of UF6 by 

April 1945 (AEC 1951c; Harshaw 1945). An auxiliary building known as the “hex annex” or “new hex”

area was added in 1945 to provide additional UF6 production capacity. The hex annex was 

mentioned in some references as an addition to Building G-1 and in others as a separate structure. 

Although this level was not reached on a regular monthly basis, by December 1947, Harshaw was

producing up to 46,000 lb of UF6 per month. In December 1951, Harshaw Chemical Company 

stopped producing UF6 (Sargent 1951).

2.3.3

In 1942, at the request of Standard Oil, Harshaw Chemical Company began to produce UF4 from 

UO2. This material was sometimes referred to as “green salt” and the processing area as the “green” 

or “491” plant. The first UO2 material processed at Harshaw was supplied by Westinghouse Electric 

and Manufacturing Company (HCC ca. 1945). In July 1942, the MED asked Harshaw to produce 

Uranium Tetrafluoride Production
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 15 of 114

1,200 lb of UF4 per day from UO2 produced and supplied by DuPont and Mallinckrodt (Harshaw 

1945).

In September 1942, Harshaw implemented large-scale production using a new facility with a 

production capacity of 50,000 lb (25 tons) of UF4 per month (HCC ca. 1945; AEC 1951c). By 

December 1943, continued improvements in the conversion process increased the production level to 

60 tons per month (Harshaw 1945; AEC 1951c).

In December 1944, Harshaw moved production of UF4 to Building G-1, where the production rate was 

about 3,000 lb/d (HCC ca. 1945). In February 1946, anticipating a later full production of 28,000 lb

(14 tons) per week, the MED authorized an increase to 15,000 lb/wk.

Although discrepancies in production quantities exist and are likely due to the reporting of theoretical 

capacity versus actual production, the final full production level for UF4 (in February 1948) appears to 

have been 81 tons/mo (AEC 1951c). In October 1951, Harshaw stopped producing UF4 (DOE 1997).

2.3.4

In 1947, Harshaw Chemical Company constructed an ore-to-UO3-to-UO2 batch production facility for 

AEC use (Velten 1949). This facility was constructed so Harshaw could produce UO2 on the site, 

alleviating the need to bring in UO2 from other suppliers. “Ore” was typically received as milled ore.

Uranium Trioxide and Uranium Dioxide Production

By July 1949, the ore-to-UO3-to-UO2 batch production facility was operating (AEC 1951c). In 1951, 

through process modification that included a switch to the use of tributyl phosphate-kerosene rather 

than ether, UO2 production increased (AEC 1951c). However, the UO3-to-UO2 portion of the 

operation stopped entirely in 1951 although the AEC implies this occurred as of October 1952 (DOE 

1997; Stefanec 1951; Termini 1952).

The UO3 contractual amount was 200,000 lb/mo from October to December 1952 (Neumann 1952).

Harshaw continued to produce UO3 from ore until August 1953, when UO3 production was placed on 

standby, and the AEC directed Harshaw to end all processing except for a final conversion of all 

leftover feed materials to UO3 (Neumann 1953).

2.3.5

Throughout the 1950-through-1951 timeframe, uranyl nitrate hexahydrate (UNH) [chemical formula, 

UO2(NO3)2∙6 H2O], which is an intermediate liquid produced in the initial processing of ore and 

uranium extraction, was reportedly produced as “research material.” Documentation available to 

NIOSH does not indicate if the UNH was produced for use at Harshaw or elsewhere. Beginning in 

1952, Hanford sent UNH to Harshaw, sometimes via the Brush Beryllium Company, to be converted 

into UO3 (Klevin 1952b; Termini 1952; DOE 2000). Hanford produced UNH using a tributyl phosphate 

chemical process and transported it in tank cars to Harshaw.

Uranyl Nitrate Hexahydrate Production

2.3.6

Between 1943 and 1944, Harshaw Chemical Company manufactured a number of special radiological 

materials, including uranium oxyfluoride (UO2F2), sodium uranate (Na2UO7) at 84%, and uranium

nitrate [U(NO3)2] at 56% (presumably the percentages were of U3O8 equivalent) (Harshaw 1945). 

However, NIOSH has not located documentation describing how Harshaw processed these materials. 

Between February 1947 and August 1950, Harshaw prepared short-lived 234Th (known as UX1) from 

Operations Involving Other Radiological Materials
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 16 of 114

a residue of the UF4-to-UF6 conversion process (Stefanec 1951). Thorium-234 was produced in a 

laboratory in bench quantities (Stefanec 1951).

On at least two occasions, Harshaw Chemical Company processed some low-enriched uranium, in 

the form of UF6, from Hanford (Kelley 1946). In 1945 and 1946, Harshaw was asked to mix natural 

UF6 with UF6 that had been slightly enriched. The resulting slightly enriched UF6, now referred to as 

low-enriched UF6, appears to have been enriched to less than 1% 235U by weight and was shipped to 

K-25 in Oak Ridge, Tennessee (Kelley 1946). The specific activity of 1% enriched uranium is 

0.783 pCi/μg.

In 1952 and 1953, various shipments of UO3 were sent from Hanford to Harshaw for purification 

(Klevin 1952b; BJC and Haselwood Enterprises 2000), which began on 1 October 1952 (Klevin 

1952b). This was the so-called Redox material, which was made from recycled uranium (RU) 

recovered from irradiated uranium fuel at Hanford. At first, Hanford had trouble producing from this 

RU material enough UO3 of adequate purity for refining into UF6 (BJC and Haselwood Enterprises 

2000; DOE 2000), mainly due to the difficulties in removing transuranic elements (such as plutonium 

and neptunium) and fission products, principally 99Tc (BJC and Haselwood Enterprises 2000), and 

hence it was shipped to Harshaw. Klevin (1952b) states that the RU UO3 was shipped in 30-gal

drums and that it had been produced based on depleted uranium. The AEC further indicated that the 

material contained “practically all the elements in the periodic table,” as well as, 9 ppb Pu and 0.64%

of 235U (Klevin 1952b, BJC and Haselwood Enterprises 2000). The purified product was sent to K-25 

by way of Oak Ridge National Laboratory.

UNH was also sent, beginning in 1952, from Hanford to Harshaw (at least sometimes via the Brush 

Beryllium Company) to be converted into UO3 (Termini 1952; Klevin 1952b; DOE 2000). The UNH 

sent from Hanford in 1952 and 1953 appears to have been RU like the UO3 shipped from Hanford at 

about the same time.

To estimate the activity fractions of RU at Harshaw during the period from July 1, 1952, to June 1954, 

the maximum radionuclide mass fractions were used with an assumption of specific activity for 

depleted uranium of 0.4 pCi/μg [3]. Table 2-3 lists the results. These fractions will overestimate the 

activity of RU in the source term for most exposure scenarios [4].

Table 2-3. Assumed fractions of recycled uranium contaminants at 

metal handling facilities.

Recycled uranium contaminant Pu-239 Np-237 Tc-99

Contaminant, ppb of uranium 30 780 12,000

Activity fraction of contaminant in uranium 0.00464 0.00137 0.506

2.4 INCIDENTS

In early 1950, one worker in the green plant and five or six workers in the hex plant showed abnormal 

urine readings (high albumin, variously called albumen-urea, albuminaria, and albuminuria in AEC 

references) (AEC 1950b; Lippmann 1958). At least one worker was immediately removed from 

uranium work (AEC 1950b; Lippmann 1958) and was sent to the hospital at Brookhaven National 

Laboratory for 10 days of study in August and September 1950 (AEC 1950b; Quigley 1950). In 

October 1950, an AEC New York Operations Office (NYOO) doctor met with Harshaw to discuss 

these cases (Sargent 1950a, which lists the workers and their hire dates, work locations, and job 

titles). He emphasized to Harshaw that seven confirmed cases (of 11 reported) for 200 workers was 

far above normal and that tests had eliminated all causes except uranium damage to the kidneys. He 

pointed out that the condition had not occurred in those known to be most heavily exposed, so it could 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 17 of 114

be that the affected workers had a special susceptibility to uranium, or that short intense exposures 

were not being picked up by the existing exposure evaluations. A urologist at Deaconess Hospital 

was unable to find any reason for the condition of the two workers tested there other than their 

uranium exposure (Quigley 1950). A follow-up study initiated by the AEC showed that for those 

followed up, urinary findings were normal within a few months (Lippmann 1958).

2.5 SUMMARY OF POTENTIAL EXPOSURES

Harshaw AWE employees could have received internal and external radiation exposures from 

uranium and nonuranium contaminants, including radium, and thorium in the milled and composite 

materials. The uranium content of mined uranium ores varied based on the quality of the rock being 

mined. Natural conditions resulted in varying degrees of disequilibrium between decay series 

radionuclides within particular ore deposits. A majority of the 226Ra and thorium isotopes would have 

been removed by the milling processes.

While the activity of the 226Ra and thorium isotopes was reduced, much of the potential for external 

exposure to AWE workers at Harshaw was likely due to uranium progeny. Radium-226, a gamma 

emitter, likely produced some of the external whole-body dose received by the Harshaw workers. 

Thorium-234 and 234mPa, both primarily beta emitters, likely produced whole-body skin and extremity 

dose for workers involved in handling the fluorination ash or decontaminating equipment used to 

contain or transport the bed ash (ORAUT 2006a).

Internal exposures would have included alpha radiation resulting from uranium and uranium progeny 

emissions [5]. Workers involved in the UX-1 operation were potentially exposed to alpha emissions 

from thorium. AWE workers were also likely exposed to elevated levels of radon. The concentration 

of radium (and radon) and other progeny present in the ore concentrates, processed uranium, and 

processing residue at any given time depended on various factors, including the concentration of 

uranium in the original ore body; how much uranium progeny remained in the U3O8 product received 

from the mill, the total amount of U3O8 product processed, and how long the U3O8 was stored prior to 

use as feed at Harshaw [6].

Little monitoring information is available on the radiological aspects of the D&D work. No records of 

how the workers were monitored appear to have been made, other than that the monitoring was 

typical of normal operations and would appear among the regular operations records, perhaps with 

the notation "decontamination."

3.0

3.1 INTRODUCTION

OCCUPATIONAL MEDICAL DOSE

Harshaw uranium processing workers were given a preemployment physical that included a chest 

X-ray (Lippmann 1958). They were also given an annual physical that included a chest X-ray (Rauch 

1948; Ferry 1944a). There might have been a hiatus during which annual chest X-rays were not 

given. In March 1947, the AEC specifically approved annual X-rays (Howland 1947a). However, this 

might have represented the AEC agreeing to assume the expense of the annual X-rays rather than 

Harshaw and so might not indicate an interruption of annual chest X-rays.

Workers were also given preemployment and annual X-rays of the pelvis for a time, but in January

1944, the AEC recommended that the X-rays of the pelvis be discontinued (Ferry 1944a), and it 

appears that this was done. In 1947, AEC recommended that as part of Harshaw’s health and safety 

program upgrade, an "anterior-posterior radiograph of the thoracic spine" be included in 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 18 of 114

preemployment and termination medical examinations to detect fluorosis (Kelley 1947). No evidence 

has been found to date that indicates Harshaw performed thoracic spine X-rays. It does not appear 

that this was done. However, the X-rays of the pelvis appear to have been resumed at some point 

because in June 1951 an AEC doctor who visited Harshaw recommended discontinuing the annual 

X-rays of the pelvis done on all employees (Quigley 1951a), while in November 1951 another AEC 

doctor recommended (Tabershaw 1951) that even with the contraction in the number of workers, the 

medical program be continued as previously, with X-rays of the pelvis to be continued for new hires 

and terminating employees but not for the annual physical – implying that annual X-rays of the pelvis

were no longer being done.

Due to the corresponding dates, it appears that X-rays of the pelvis were probably resumed (if they 

ever were discontinued) in the second half of 1951 in response to the finding of albuminuria in some 

workers (see Section 2.4). Thus, it would appear that X-rays of the pelvis were done from 1942 to 

1944 and again from about 1950 to 1951; this would be on an annual basis. The rationale for the Xrays of the pelvis was to detect bone effects due to fluoride exposure, so it is likely that only the 

workers in the UF4-to-UF6 process area had X-rays of the pelvis.

No information is available regarding whether all workers (e.g., clerical and other support workers) 

received annual X-rays. In addition, because AEC recommended in October 1948 that Harshaw keep 

a single file of records for each worker and a complete X-ray file (Kelley 1949a) (which implies that 

they had not been doing so), it seems likely that early records of X-ray examinations might be in 

disarray or missing.

No information appears to be available concerning what the practices were during the postoperations

phase, 1956–1999.

3.2 EXAMINATION FREQUENCIES

Therefore, the assumption favorable to claimants is made that all personnel who worked in the areas 

in which work was done for the MED/AEC received initial, annual, and termination chest X-rays [7]. 

Unless there is evidence to the contrary in the employee’s file, the chest X-rays should be assumed to 

be posterior-anterior (PA) radiographic only. Those process area workers who were working in 

MED/AEC work from 1942 to 1944 and from 1950 to 1951 should be assumed to have received 

annual X-rays of the pelvis as well [8].

3.3 EQUIPMENT AND TECHNIQUES

Photofluorography was widely used in the United States for medical screening in the 1930s and was 

used at some sites supporting the Manhattan Project. No evidence has been located that it was used 

for Harshaw workers. Therefore, the assumption is made that only conventional X-rays were used for 

Harshaw.

Because no actual X-ray output measurements or X-ray technique factors are available in Harshaw 

records, default values for entrance kerma appropriate for this period should be used in the 

calculation of organ dose for use in dose reconstruction. Information to be used in dose 

reconstruction for the early years, for which no specific information is available, is provided in ORAUTOTIB-0006, the dose reconstruction project technical information bulletin covering occupational 

medical X-ray procedures (ORAUT 2005a).

Guidance regarding organ doses from X-rays of the pelvis has been included in an update to ORAUT

(2005a). Note that for Harshaw, such contributions would be applicable only to those workers 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 19 of 114

handling or working around fluorides (i.e., workers involved with the UO2 to UF4, UF4 to UF6, and UF6

distillation processes, including workers who handled the hydrogen fluoride supply and storage 

functions). Also, such contributions would be applicable only from 1942 to 1944 and 1950 to 1951, as 

indicated above. Organ doses from thoracic-spine projections should be available in the near future.

3.4 ORGAN DOSE CALCULATIONS

ORAUT-OTIB-0006 (ORAUT 2005a) lists the chest X-ray organ doses to be assumed in default of 

more specific information as approved by the Dose Reconstruction Project. ORAUT (2005a) should 

also be used to assign exposures for individuals who would have had X-ray examinations of the 

pelvis.

4.0

4.1 INTRODUCTION

OCCUPATIONAL ENVIRONMENTAL DOSE

Although atmospheric releases appear to have been significant, especially in the early years, little 

useable information appears to be available.

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

In the production of UF6, most of the uncondensed gas remaining after condensation and distillation 

passed through a device called the "large turbosaturator" (also referred to as a "turbo-agitator" or 

"turbo") to remove solids and then went out a blower on the roof as effluent to the environment. The 

rest of the uncondensed gas passed through the "small turbosaturator." The large turbosaturator 

handled the fumes from four sources: (1) the small vent (exhaust) hood over each individual hex 

reactor; (2) the distillation unit dry ice trap; (3) the bleeds from the distillation unit floor receivers; and 

(4) the bleeds from the dry ice trap receivers. It used a potassium hydroxide (KOH) spray as a 

fume-condensing or capture method and was linked to a packed rooftop tower that created the 

suction and scrubbed the effluent using circulating acidified water. The small turbosaturator handled 

the fumes from the five ice traps serving the hex reactors; it created its own suction and pulled the 

fumes through a well of KOH (Rauch 1948).

The small turbosaturator was highly efficient, but the large turbosaturator was very inefficient during 

most of its process life (Rauch 1948): it was so inadequate that when a receiver was being bled, the 

suction vents sometimes discharged air into the reactor hoods instead of drawing it out (Hunter 

1949a). To solve this problem, Harshaw’s practice was to keep the smaller suction vents closed, 

which negated their effectiveness (Hunter 1949a). In addition, the exhaust system that provided 

suction across the face of the line of reactors was not exhausted through either turbosaturator or any 

other scrubbing or filtration system, but was vented straight to the outside atmosphere (Burman 

1949). This was a different system from the individual hex reactor hood system and was in fact a 

much stronger exhaust than the individual hex reactor hood system (Burman 1949; Hunter 1949a) 

and so actually received more of the contaminants (Hunter 1949a). These systems contained UF4

dust from the loading process and UF6 fumes from the disconnecting of the reactors and receivers 

(Burman 1949). The UF6 fuming was exacerbated when there was a plugging of one of the hex 

headers because in that case the nitrogen purge of the hex reactor prior to disconnection had to be 

skipped and the concentration in the effluent was thus all the higher (Burman 1949).

Releases were highest before the turbosaturator system was installed in about 1947, when Harshaw’s 

neighbors (including homeowners) were threatening lawsuits due to, for example, the etching of glass 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 20 of 114

in car and home windows that took place as a result of the high F2 releases (Ray 1947). The 

turbosaturator system helped a great deal, but levels of both fluorine and uranium dust were still high 

and neighbors still complained (AEC 1949a) until the problems with the large turbosaturator and the 

unfiltered exhaust were finally corrected in 1949 with the installation of two "Buffalo" (Buffalo Forge) 

scrubbers (AEC 1949a; Harris 1949a). These scrubbers were of the same type that ElectroMet had 

found to be effective on its effluent particulates, a "microcrystalline" dust that was very similar to what 

Harshaw was dealing with (Hunter 1949a). The efficiency of the Buffalo scrubbers ranged from 75% 

to 99% (HCC 1950–1953; Stefanec 1951) and they made a significant difference in the Harshaw 

effluent concentration (AEC 1950a) and in, for example, exposure to workers in the shipping and 

receiving area (AEC 1950c). However, even these scrubbers were thought by the AEC to be 

somewhat inadequate with regard to flow rate (Hunter 1949a).

Regarding effluents from the UF4 plant, the roof exhaust vent from the blower had no filtration device 

at all; it was pointed downward and thus much of the particulates accumulated on the gravel of the 

roof (Burman 1949). Harshaw installed a capture device on this vent in about 1949 (Burman 1949). 

Regarding effluents from the brown plant, a micropulverizer between the denitration pots and the 

Rockwell (UO3-to-UO2) furnaces was exhausted through two cyclones and a Hersey-type bag 

collector; the latter discharged directly to the atmosphere of the room it was in and the collected 

contents discharged down a hopper and chute into a drum on the floor below it (Wolf 1948a). In 

addition, exhaust ventilation was installed for the UO3 packaging hood in January 1952 (Klevin 

1952c).

Stack measurements appear to have been done to estimate losses (AEC 1949a) (i.e., usually the 

quantity was expressed in uranium mass lost per hour, per day, or per month). Harris (1949b) 

concluded that the AEC and Harshaw sampling methods were comparable and could be used 

interchangeably, and after that Harshaw appears to have done all its own stack measurements. Few 

measurements are available. The AEC stack effluent samples showed losses of 21 g U/min total from 

the six reactor stacks and 13 g/min from the turbosaturator stack, for an hourly total loss of about 5 lb 

of uranium, which AEC thought was mostly in the form of UF6 (Eisenbud 1949a). Between 100 and 

10,000 ft from the Plant C area, AEC found no sample above 10 μg/m3

, with multiple such samples 

not being above background; the maximum concentration was at about 0.3 mi, where the average 

concentration was 3 μg/m3 (Eisenbud 1949a). The losses, and thus undoubtedly the concentrations, 

decreased significantly with the addition of the Buffalo scrubbers; by July 1949, the hourly loss rate 

was down by a factor of about 2 or 3 (Harris 1949c).

Those not working in process areas had a potential for exposure. Because the work was done mostly 

in one building, the process areas were not always well enclosed or well ventilated, and there was a

considerable loss of material out the various stacks that could be carried by building drafting back into 

the building. Also, contamination appears to have been tracked out of the process areas in the early 

years at Harshaw. For example, this is suggested by the practices discussed in the 1946 Harshaw 

operating manual (HCC 1946), by the observation that doors were left open between areas and there 

was two-way traffic through the one-way turnstiles (Klevin ca. 1948), and by the fact that visitors could 

not be issued cover clothing until the new guardhouse was installed in 1949 (Morgan 1949). Some 

potential for tracking likely existed in later years too, even though revisions of the change room 

operation were intended to prevent this (Ray 1947; Rauch 1948; Kelley 1949a; Eisenbud 1949b; Wolf 

1948a,b).

No information is available about releases, if any, in the postoperations and D&D phase, 1956-1999.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 21 of 114

4.3 EXTERNAL DOSE

Environmental ambient radiation levels have not been recovered for this site.

5.0

5.1 INTRODUCTION

OCCUPATIONAL INTERNAL DOSE

Few radiation measurements or evaluations of dust exposure were made in plants doing MED/AEC 

work in the first few years of operations because it was anticipated that the processing of uranium 

ores, ore concentrates, and compounds would involve little risk of radiation injury. This belief was 

based on the low specific activity of uranium and on what was thought at the time to be the temporary 

nature of the work. In January 1944, the Special Materials Division of the MED Medical Section 

recommended a medical monitoring program to its contractors that included routine physicals, urine 

sampling, and X-ray examinations for worker protection (Ferry 1944a). It is not clear whether MED 

prioritized certain sites to implement these programs, but the same document indicated that several 

contractors had instituted the recommendations while others had not. In addition, both air and dust 

samples had been collected and analyzed at several of the contractor sites, but not initially at 

Harshaw. When the MED/AEC NYOO evaluated the results of these surveys at various sites in light 

of the government intent to continue uranium processing work in the late 1940s, it was determined 

that the hazards were not negligible. Kelley (1949a) listed the requirements for a minimally 

acceptable medical program (including maintaining X-ray records) that it wanted Harshaw to institute 

in addition to what Harshaw might have been doing already.

5.2 URANIUM SOLUBILITY AND PARTICLE SIZE

The uranium processing operations at Harshaw produced some insoluble uranium compounds, such 

as UO2 and U3O8; some moderately insoluble compounds, such as UF4, and UCl4; and some soluble 

compounds, such as UF6 and its byproduct UO2F2. The default absorption types for radioactive 

materials that were likely to have been present at Harshaw can be determined from International 

Commission on Radiological Protection (ICRP) Publication 68 (ICRP 1995). Information on likely 

uranium absorption type by job title is listed in Table B-22.

There is little information on particle sizes at Harshaw except for one 1950 study done by Klevin 

(1950b), when U3O8, UO3, UO2, UF4, and UF6 were all being handled at Harshaw. Table 5-1 lists a 

summary of the mass median particle sizes measured as well as the result of calculations of the 

aerodynamic median activity diameter done for this site profile. In the particle size study, the sizes of 

breathing zone dusts were found to be consistently larger than those in the general air. There was a 

variation in particle size for duplicate impactor runs of up to 0.5 μm for mass median diameters from 

1.28 to 4.7 μm. A "rigorous statistical analysis" showed both impactors to be exactly the same, so it 

was concluded that the size variations were random. A comparison of the total impactor 

concentrations to the filter paper samples showed that 17 out of 22 impactor samples were higher; 

this was found to be statistically significant, although no explanation could be offered for this. When 

adjusted for density, these particle size results are consistent with ICRP Publication 66 (ICRP 1994) 

default particle size distributions. Thus, ICRP (1994) defaults should be assumed, including an 

activity median aerodynamic diameter of 5 μm.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 22 of 114

Table 5-1. Air-sampling information.

Sample location/activity Form Typea

Mass median 

diameter, μmb

Density, 

g/cm3

Activity median 

aerodynamic 

diameter, μmc

Nongeneric

Gulping orange (UO3) UO3 BZ 3.1 7.29 8.37

Near pots, NE corner UO3 GA 1.8 7.29 4.86

Rockwell 1st deck UO2 GA 2.3 10.96 7.61

Rockwell discharge UO2 GA 2.1 10.96 6.95

Dumping 8 brown (UO2) trays UO2 BZ 3.9 10.96 12.9

Brown-green (UO2-UF4) loading platform UO2 GA 4.4 10.96 14.6

20 ft south of laundry UO2 GA 2.6 10.96 9.27

Dumping 8 green (UF4) trays UF4 BZ 4.7 6.7 12.2

Hex area near operator's desk UF6 GA 1.28 4.68 2.77

Removing ice trap UF6 BZ 1.35 4.68 4.22

Center of still area UF6 GA 1.35 4.68 2.92

Generic

Generic UO3 UO3 BZ 3.1 7.29 8.37

UO3 GA 1.3 7.29 3.51

Generic UO2 UO2 BZ 3.9 10.96 12.9

UO2 GA 2.4 10.96 7.95

Generic UF4 UF4 BZ 4.7 6.7 12.2

UF4 GA 4.4 6.7 11.4

Generic UF6 UF6 BZ 2 4.68 4.33

UF6 GA 1.3 4.68 2.81

a. BZ: breathing zone; GA: general area

b. For the top (nongeneric) block: the filter paper concentration is from a single sample, while impactor concentration 

and the mass median particle size are an average of the two impactor samples.

c. Mass median diameter is the term used in the reports but activities not masses were actually measured. The mass 

median aerodynamic diameter is the equivalent of the aerodynamic median activity diameter if the activity is 

considered to be homogenous. The aerodynamic median activity diameter calculation assumed that the particles 

measured could be treated as spherical so that the Stokes diameter is equal to the geometric diameter (measured) 

and the slip correction factor is equal to one.

5.3 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.3.1

5.3.1.1 Early Urine Studies

In Vitro Urine Analysis

Ferry (1944a) stated that Harshaw was one of the sites following the MED-recommended medical 

program, which included urinalysis for uranium done as a "screening experiment" with regard to both 

acute and chronic exposures. HCC (ca. 1945) stated that monthly urine samples for all employees 

exposed to UF6 were sent to the Medical Division of the MED (to be sent on to the University of 

Rochester) for uranium analysis. The 1945 measurements might have been part of a volunteer study 

of 24-hour excretion of uranium; the volunteers, who included the 491/492 plant superintendent, 

attempted to limit voiding to "the hour of arising, noon hour, completion of the day’s work, and on 

retiring" (Mears 1945). In January 1946, MED informed the plant superintendent that its Medical 

Section had decided to terminate all urinalyses for uranium and fluoride content because of the 

contamination of a high percentage of the samples.

In August 1947, AEC told Harshaw that its health program had to be improved, including a routine 

urine testing program for uranium and fluorine to be instituted after a spot check program was 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 23 of 114

completed (Kelley 1947). However, the samples submitted in 1947 again appear to have been 

contaminated on the basis of both the fluorine and the uranium analyses (Howland 1947b).

Because of the questions regarding the validity of sample results prior to December 1949, the 

apparent variations in sample analysis methods, and even who was doing the analyses, the Harshaw 

urinalysis data prior to December 1, 1949 should not be used. It appears that the errors, if any, are in 

the conservative (high) direction and thus would be favorable to claimants [9]. However, urine 

samples are also likely to be rare, necessitating an air sample-based approach.

It has been determined that it is not feasible to perform dose reconstructions from August 14, 1942,

through November 30, 1949, due to the lack of internal dosimetry data for the radionuclides 

associated with uranium for operations at Harshaw [10]. Beginning December 1, 1949, adequate 

information to perform dose reconstructions is available (Elliott 2007).

5.3.1.2 Routine Urine Program

Some routine urinalyses appear to have begun in late 1948 (Harris 1949d). However, there is a 

description of a (new) pilot program to measure uranium in urine in March 1949 [described in Harris 

(1949d) and Eisenbud (1949c)], and Lippmann (1958) stated that the urine sampling program began 

in January 1950. Measurements for the 1948-to-early-1949 period for a few people appear in the 

Harshaw records, and an increasing number of workers appear to have been included in the urine 

sampling program as of late (December) 1949. Sargent (1950b), while it referred to a previous lot of 

urine samples, also requested that Harshaw institute a urine sampling program "on a running basis" to 

sample about 100 workers per month, including occupations that the AEC specified. Thus, the 

January 1950 time point could represent an acceleration of urine sampling rather than its start. In 

1951, an AEC doctor recommended continuing the urinalyses as part of the Harshaw medical 

program even though the green and hex plants were being phased out, but it is not clear if the 

urinalyses were to include a uranium measurement or not (Tabershaw 1951).

Sargent (1950a) stated that 200 workers were subject to urinalysis, which appears to have included 

workers at all three major areas of Plant C. While there was no tally kept by Harshaw of the number 

of people exposed to uranium who had left their employment, in November 1951, Harshaw provided 

the AEC an estimate of several hundred such people who had been exposed for more than a year.

As noted above, the AEC specified at one point the worker categories to be sampled: hex loaders, 

hex operators, still operators, brown and green loaders, orange pot unloaders, Rockwell operators, 

black oxide (ore concentrate) loaders, and shipping and receiving personnel (Turner 1947a). A 

former AEC official stated that samples were taken from "hexafluoride" (UF6) and "nitrate" (UF4) 

workers and that one could see a 10-fold drop in the uranium content between Friday night and 

Monday morning samples for those exposed to soluble forms, but little drop for those exposed to 

insoluble forms (ORAU 1983). Lippman (1958) describes a study of 1950 data to determine the 

correlation between urinalyses and exposure of Harshaw workers in the green and hex plants; he 

mentions workers showing elevated urinary albumin measurements, so this apparently retroactive 

study might have been motivated by that occurrence. Table B-20 is drawn from Lippmann (1958); the 

results are listed along with statistical data calculated for this site profile. Thus, Table B-20 could be 

useful for trend or comparison purposes or for interpreting worker bioassay records.

Before the UF4 and the UF6 production plants were closed in 1951, urine samples were usually taken 

in pairs on a before-and-after weekend basis (Lippmann 1958); these were called the "before 

weekend" or "Friday" sample and the "after weekend" or "Monday" sample, respectively. Samples 

were taken every week but not from every worker or group of workers, so the average was a pair of 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 24 of 114

samples per worker per month (Lippmann 1958). From September 1951 on, the AEC appears to 

have directed Harshaw to discontinue all Friday samples from the 491 area; to obtain Monday 

samples from each 491 operator every 2 weeks; to discontinue urine samples from health physics 

personnel and guards; and to include all process personnel in the urine sampling program (Harris 

1951). The reason for including the last group was that many process area workers had rarely or 

never submitted urine samples, resulting in serious gaps in the data that the AEC had collected 

(Harris 1951). A coworker worker study was conducted to aid in filling these data gaps and is 

described in Section 5.9.

Samples do not appear to have been 24-hour samples. Quigley (1951b) suggests that at least some 

might have been 6- to 8-hour samples. Referring to urine samples sent to the University of Rochester 

from any AEC site, ORAU (1983) stated that samples were collected from workers in 4-oz glass 

bottles with Bakelite caps and shipped offsite for analyses.

It should be noted that many workers, over the years of their employment, worked both at the green 

and brown plants, where the uranium form was fairly or highly insoluble, and at the hex plant, where 

the uranium form could be either soluble or insoluble. Urinalysis data appeared to end in 1953.

5.3.1.3 Minimum Detectable Activities

Uranium fusion photofluorimetry urinalyses performed by the University of Rochester and the AEC 

NYOO were similar to those performed at other AEC facilities. Consistent with the information above, 

the default detection threshold for uranium urinalysis can be assumed to be 10 μg/L based on a 

reported sensitivity of 5 to 10 μg/L for uranium fluorimetry urinalysis in the early years (Wilson 1958). 

Lippman (1958) stated that the analytical precision of the urine samples was about ±10 μg/L for 

samples reading less than 100 μg/L; Howland (1947b) stated that the limit of reliable determination by 

the fluorimetric method was about 0.01 mg/L, or again, about 10 μg/L.

5.3.2

Harshaw urinalysis samples were sent for analysis to the University of Rochester directly (HCC 1950–

1953) or to Health and Safety Laboratory (HASL) (Eisenbud 1975) to be sent on to Rochester. In 

about January 1949, the University of Rochester work in support of the AEC was switched to HASL 

(ORAU 1983) and thus HASL itself was analyzing the Harshaw urine samples. ORAU (1983) stated 

that the HASL urinalysis program ended in 1955 or 1956 and that Harshaw was the last plant for 

which HASL did analyses. However, HCC (1950–1953) indicated that urine samples were sent to 

National Lead Company (Fernald) for analysis late in the plant’s life; this could have been only during 

the decontamination phase.

In Vitro Methods for Uranium

The radiological analysis was apparently only for uranium content (referred to as "X in urine" or 

"uranium-in-urine"). ORAU (1983) stated that all urinalyses done at the University of Rochester used 

the fluorometric method; the urine was dried in a platinum dish, then fluxed with either sodium fluoride 

or a lithium-calcium fluoride mix and counted.

ORAU (1983) also stated the following about samples analyzed at the University of Rochester. The 

samples were run in triplicate, with the results usually being within ±2 μg/L of one another; if this 

turned out not to be the case, it indicated that there had been poor fusion of the flux and the samples 

were re-fused and rerun, which usually corrected the problem. The value recorded was the median 

value of the three. For insoluble uranium, it was considered that 30 μg/L in the urine corresponded to 

an air concentration of 50 μg/m3

. The analysts were confident of readings greater than or equal to 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 25 of 114

5 μg/L, but if more confidence was desired for lower level samples (e.g., for special projects), the 

urine was concentrated either by ion exchange or by extraction, or more aliquots were run.

5.4 OTHER BIOASSAY METHODS

Fecal sample analyses do not appear to have been a routine part of the Harshaw bioassay program. 

No whole-body or lung counts appear to have been performed for Harshaw employees during the 

covered period. A few blood measurements were done in 1950 and 1951. Data have been found in 

Harshaw bioassay records for about 16 people (including one AEC person), with most volumes given 

and with the notation that the blood was assayed using fluorometric analysis [11]. No analytical 

precision has been found for the blood measurements.

5.5 RADON LEVELS DURING OPERATIONS

As indicated above, uranium bearing ore (unprocessed rock) does not appear to have occurred at 

Harshaw, at least in any large quantities. Because the feed material that was received by Harshaw 

had been processed prior to receipt, the activity concentrations of both 226Ra and 230Th would have 

been much lower than that of uranium, and radon concentrations would likewise have been reduced. 

No records of radon measurements made by the AEC or Harshaw have been located for the period 

that the Harshaw facilities are covered by this document.

Because there are no known radon measurements during the period of operations, it is difficult to 

determine what the Harshaw radon levels actually were. It can be inferred that the AEC concluded 

that the Harshaw radon levels were always or nearly always below the radon tolerance level of 

1 × 1010 Ci/L and likely usually below the detectable level of 1 × 10-12 Ci/L due to the low radium 

content of the received material. But to make an upper bound estimate of potential radon levels, 

measurements from the Mallinckrodt site (ORAUT 2005b, Table 25) can be evaluated as a bounding 

set. Table 5-2 lists Mallinckrodt levels from 1954 to 1957, which was after high-radium ores had been 

processed and the feed was mostly MGX (a Belgian Congo ore tailings concentrate that still had a 

high radium level), soda salt, and lower grade ore. These measurements would be higher than the 

levels at Harshaw because of the higher radium content in the received feed and the likely 

comparable residence time of the material in containers. The measurements should be comparable 

also because the Mallinckrodt process included the ore concentrate-to-UO3, UO3-to-UO2, and UO2-toUF4 steps [12].

The results of calculations of working level months (WLMs) are listed in Table B-18. In these 

calculations, the figures in the rightmost column of the table above were used and the following 

assumptions were made [these are the same as those used for Mallinckrodt (ORAUT 2005b)].

1. For ore concentrate storage areas, the equilibrium factor for the radon daughters was 

assumed to be 1.0. For process and maintenance areas, the equilibrium factor was 0.50. For 

nonprocess areas, the equilibrium factor was 0.40.

2. For workers in ore concentrate storage and process areas and in shops and support facilities 

other than laboratories, the occupancy factor was assumed to be 0.75 for the normal work 

area and 0.25 for the break room, locker room, and other low-exposure areas. For 

maintenance workers who visited the process areas, the occupancy factor was assumed to be 

0.50 for the process area, 0.25 for the maintenance shop area, and 0.25 for the break room, 

locker room, and other low-exposure areas. For laboratory workers, the occupancy factor was 

assumed to be 0.88 (i.e., 7 hr/d) in the laboratory area and 0.12 in the break room, locker 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 26 of 114

Table 5-2. Radon levels in the Mallinckrodt plants in the later period of operation, in units of 10-10 Ci/L

of radon.

Area Room or building 1954 1955 1956 1957 Harshawa

Indoor Scalehouse/Ore Storage/Warehouse 0.01 0.01 0.01

Digest/Feed 0.03 0.01 0.03 0.03

Extraction Cells 0.26 0.01 0.01 0.01

Centrifuge Area 0.07 0.01 0.01 0.01

Feinc/Filter/Raffinate/Cloth Storage/Niagara 0.14 0.50 0.07 0.01 0.07

Orange Packing 0.12 0.12

Pot Room 0.02 0.02

Shotgun Lab (UO3 assay) 0.04 0.04

Lab (Research/ Control /X-ray/Radium) 0.01 0.04 0.04

Decontamination Room 0.01 0.01

Nitric Acid House 0.01 0.01

Ether House 0.02 0.02

Receiving (non-ore shipping and receiving) 0.01 0.01

Welding, Millwright, and Electrical Shops 0.01 0.01

Maintenance Shop 0.04 0.04

Smoking (Break) Room, Production Office 0.01 0.01

Outdoors/yards Near Ether House/Acid unloading station 0.01 0.01

a. Mallinckrodt 1956 and 1957 figures correspond to lower grade ores, soda salt, etc. that Harshaw used in its ore concentrate to 

UO2/UO3 processing. These values are recommended for use in dose reconstructions.

room, and other low-exposure areas. For office workers, the occupancy factor was assumed

to be 1.0 in low-exposure areas.

3. The conversion to WLM per year from the measured values in units of pCi/L was assumed to 

be:

Intake (WLM/yr) = 0.12 ∑[Equilibrium factor × Occupancy factor × Radon level (pCi/L)] (5-1)

5.6 AIRBORNE RADIOACTIVE DUST MEASUREMENTS

A formal program of airborne radioactive dust measurements taken by the AEC began when the first 

formal report of Harshaw radioactive dust levels was issued by the AEC (1948), although dust 

samples were said to have been received periodically from Harshaw from perhaps 1943 on (Ferry 

1944a,b). Eisenbud (1975) reported that AEC air sampling was generally done by collecting the dust 

on Whatman #41 filter paper and counting total alphas; a correction for self-absorption in the paper 

was then applied and the results were reported as alpha disintegrations per minute per cubic meter. 

Table 5-3 lists production information for UF4 and UF6 and the start of dust sampling. Additional 

details of the air sampling methods and practices are found in AEC (1950b,c,d), Glauberman and 

Harris (1958), Hayden (1948), Klevin (1950a), and Lippmann (1958).

The results are listed in Tables B-10 to B-15 in disintegrations per minute per cubic meter. Tables 

B-10, B-11, and B-12 list results of instantaneous (spot) measurements for various areas and 

particular jobs, while Table B-13 lists results for a beta count of some air samples taken in the green 

and brown plant. Table B-14 lists results for individual areas, expressed as the daily [time-] weighted 

average exposure (DWE) as calculated by the AEC. Table B-15 lists results for individual 

occupations (job titles), also expressed as the DWE.

Note that Breslin (1958) stated that the DWEs calculated by NYOO from measured data do not 

include any correction for respirator use and should be viewed as (only) potential exposure. Harshaw 

workers were supposed to wear respirators when loading or unloading hex reactors (Ferry 1944b; 

Rauch 1948; HCC 1946) and whenever it was thought that there was a significant potential for 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 27 of 114

Table 5-3. Production of UF4 and UF6 and the start of dust sampling.

a

Date Average, lb/d Notes

UF4 production

Apr–Aug 1942 --- Laboratory production

Sep 1942–Jul 1944 1,650 Production level; Feb 1944: first MED dust samplingb

Aug 1944–Dec 1946 2,650

1947 5,500

1948 5,800 May, Sept 1948: first two AEC dust studies using DWEs

1949 5,000

1950 5,250

Jan–Mar 1951 5,650

UF6 production

To April 1944 400 Pilot Plant; Feb 1944: first MED dust sampling

May–Jun 1944 2,300

Jul–Nov 1944 3,300

Dec 1944–1945 4,500

1946 4,500

1947 5,700

1948 6,800 May, Sept 1948: first two AEC dust studies using DWEs

1949 6,300

1950 6,900

Jan–Mar 1951 6,600

a. Data are from AEC (1951c) supplemented by HCC (ca. 1945).

b. The February 1944 MED dust sampling date represents the earliest date found in records; samples might

have been taken earlier.

exposure to elevated dust or fume concentrations (HCC 1946). However, several documents indicate 

that respirators were not always worn or worn effectively (see Hayes 1947; HCC ca. 1945; HCC 

1950–1953; Kelley 1947; Klevin ca. 1948; Lippmann 1958; Long 1947; Morgan 1949; Rauch 1948;

Sargent 1948; Turner 1947a,b,c,d). Therefore, no credit for respirator use should be taken when 

applying air sample measurements. The contribution of resuspended dusts is assumed to be 

included in all air sample data cited in this site profile.

5.7 POSTOPERATIONS INTERNAL DOSE

To estimate doses to workers from MED/AEC contamination, use was made of the results of the 

various Harshaw contamination and dose rate surveys (i.e., from Blatz 1951; Klevin 1955a; Schoen 

1958; DOE 1984; and FUSRAP 2001). Data taken from these references and used in the calculation 

of exposure are listed in Tables B-23 and B-26; these represent only a small and select subset of data 

from larger sets of many data points although they were representative and conservatively chosen.

The RESRAD-BUILD computer code (ANL 2003) was used to calculate annual exposures from 

inhalation of airborne particulates and radon (and its progeny). Maximum averages of surface 

contamination were used to produce the inhalation and radon source terms that are favorable to 

claimants for RESRAD-BUILD; and the inhalation RESRAD-BUILD results were then used to produce 

the source term for the ingestion calculations. For RESRAD-BUILD parameters other than the source 

term, values favorable to claimants were used when they could be determined; when no specific or 

suitable values could be determined, conservative default values given in the RESRAD-BUILD 

manual (ANL 2003) or other guidance documents were used.

The results of the RESRAD radon calculations for the postoperations years were compared to 

measured radon concentrations given in DOE (1984) as listed in Table 5-4, with the working level 

(WL) values given in DOE (1984) converted to WLM per year. In DOE (1984), the Argonne National 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 28 of 114

Table 5-4. Comparison of RESRAD-calculated and DOE (1984)-measured radon concentrations.

Building

DOE (1984)a RESRAD-BUILDa

Range (WL) Range (WLM/yr) Range-ABGa (WLM/yr) Range (WLM/yr) Mode (WLM/yr)

Bldg G-1 2.5–6.9 0.033–0.093 0.0–0.053 0.0–0.069 0.017

Boiler House 4.7 0.063 0.023 0.0–.00063 0.00031

Foundry 1.7–5.6 0.023–0.075 0.0–0.035 0.0–.027 0.0027

Garage 3.7 0.05 0.01 0.0–.00063 0.00031

Warehouse 3.3–6.5 0.044–0.088 0.004–0.048 0.0–.0021 0.0010

Bldg K-1 1.1–5.7 0.015–0.077 0.0–0.037 0.0–.021 0.0031

Bldg M-1 3.5 0.047 0.007 0.0–.00002 0.00001

Bldg P-1 1.2–3.3 0.015–0.044 0.0–0.004 0.0–0.052 0.015

a. The ANL values include natural background while the RESRAD-BUILD values do not. Thus, for illustrative purposes, 

the "Range - ABG" (Range minus ABG) column gives the ANL range figures minus an average annual indoor radon 

background (ABG), taken to be 0.040 WLM/yr based on 2,000 hr/yr.

Laboratory (ANL)-reported values were based on 100% equilibrium, while the RESRAD-calculated 

values corresponded to the degree of equilibrium present given the assumed room volume, air 

changes per hour, etc. The ANL-measured values also included natural radon background for the 

buildings, which were typically constructed of concrete and brick. Thus, it would be expected that in 

general, the ANL-measured values would be somewhat higher than the RESRAD-calculated values. 

Comparing columns 4 and 6 of the table, it is clear that there is agreement between the two sets of 

values, considering that the radon added by the residual contamination appears to be at or below the 

level of background.

5.8 ACTIVITY FRACTIONS

Harshaw urinalyses measured uranium only. Air samples usually were only analyzed in terms of 

alpha activity. For other radionuclides, Table 5-5 lists the activity fractions relative to uranium activity, 

assuming that the uranium daughter products are in equilibrium [13]. Nearly all of the radium and 

thorium was removed from the milled ore, but no assays of the incoming ore were located. 

Equilibrium fractions were chosen because they are favorable to claimants.

Table 5-5. Bioassay and alpha air sample activity fractions by period and 

radionuclide.

Period Radionuclide

Bioassay activity 

fractions

Alpha air sample 

activity fractions

8/14/1942 – present U-natural 1 0.402

Th-230 0.489 0.196

Ra-226 0.489 0.196

Po-210 0.489 0.196

Pa-231 0.0228 0.00916

Ac-227 0.0228 0.00916

7/1/1952 – present Pu-239 0.00464 0.00464a

Np-237 0.00137 0.00137a

a. Assumed to be the same as bioassay activity fraction.

5.9 DETERMINATION OF INTERNAL DOSES (INSTRUCTIONS TO DOSE 

RECONSTRUCTORS)

Prior to December 1, 1949, bioassay results are not reliable. Therefore dose reconstructors should 

use air-sample based intakes as described in Sections 5.9.1, 5.9.2, and 5.9.3. From December 1, 

1949 on, individual uranium urinalysis results for Harshaw workers should be used to determine 

internal exposure to the individual when they are available. Where individual urinalysis results are not 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 29 of 114

available, illegible, or inadequate, the coworker data included in Attachment A and summarized in 

Table 5-6 are to be used to estimate internal exposures that are favorable to claimants [14]. 

Table 5-6. Chronic intake rates for types F, M, and S 234U.

Dates

Type F Type M Type S

Percentiles

GSDa

Percentiles

GSDa

Percentiles

GSDa

50th

(pCi/d)

84th

(pCi/d)

50th

(pCi/d)

84th

(pCi/d)

50th

(pCi/d)

84th

(pCi/d)

12/1/49–3/31/50 650.3 2,607 4.01 3,934 16,220 4.12 19,910 79,860 4.01

4/1/50–12/31/51 157 650.5 4.14 460.1 1,830 3.98 3,651 18,070 4.95

1/1/52–12/31/53 37.43 86.85 3.00 115.9 201.7 3.00 1,071 2,655 3.00

a. Geometric standard deviations (GSDs) less than 3.0 were assigned a GSD of 3.0.

The Harshaw bioassay data analyzed and assessed to be used as coworker data were verified and

statistically analyzed, and intake estimates were generated to aid dose reconstructors when sample 

data might either be inadequate or unavailable for estimating internal exposures to unmonitored or 

marginally unmonitored workers [14].

The 1948 figures (the earliest DWE figures available) in Table B-15 were used to estimate the 1942 to 

1947 period in Tables B-16 and B-17. There was significant variability in exposure by job, by plant, 

and by year at the Harshaw site. Hence, it is not feasible to calculate a matrix of intakes for all 

occupational types, all locations, and all periods for inclusion in this site profile. The data in 

Tables B-16 and B-17 could then be used for the period including the SEC class, when appropriate, to 

help determine intakes based on the time-weighted daily average exposure level. Note, however, that 

nonuranium doses are not to be assigned for the SEC period, August 14, 1942, through 

November 30, 1949 [10].

Care must be taken in the case of workers whose work histories show them apparently still doing 

work with radioactive materials after the period of AEC-sponsored decontamination (1956-1960). This 

is because such work during the postdecontamination years was not AEC-contracted work; it is 

uncertain what work might have been done and what materials might have been used; there might be 

no available records covering these years, and such records as there are might reflect doses 

associated with private work, not with residual contamination from AEC activities. Coworker data 

could not be developed for the AEC-sponsored D&D years. Thus, where there are no individual data 

available for this period, reference should be made to the data in Tables B-25 and B-26 [15]. For the 

postdecontamination years, where individual data cannot be located, as is likely, the data of 

Tables B-25 and B-26 should be used, as discussed in Section 5.9.5 below.

5.9.1

Table B-16 lists the DWEs from Table B-15 as converted to effective annual intakes of radioactivity in 

picocuries. The conversion was done by assuming a breathing rate of 1.2 m3

/hr for 2,000 hr/yr (ICRP 

1994). Some ratioing was done to cover job titles for unreported periods or for periods when some 

job titles were lumped together. Also, for the early period when only spot air samples were taken and 

DWEs were not used, measured data from 1948 or (if 1948 data were not available) 1949 data were 

used for 1942 to 1947. The justification for this is as follows. While control of dust and fumes 

undoubtedly improved somewhat from 1942 to 1948, it is clear that administrative and physical control 

measures were often ignored after they were implemented and that increases in production also 

increased dust levels as a general rule. Also, the aging of equipment undoubtedly resulted in more 

leakage and more need for maintenance. The start of intermittent air sampling in about 1944, regular 

air sampling in 1948, and urinalysis in 1949 are correlated with significant UF4 and UF6 production 

Determining Annual Inhalation Intakes Based on Time-Weighted Daily Average 

Inhalation Exposure Data (Table B-16), 8/14/1942-11/30/1949
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 30 of 114

increases (see Table 2-2) or with the peak of production. Hence, it is deemed that the later 

measurements bound the earlier ones.

Table B-16, the table of annual intakes based on time-weighted, daily average inhalation of uranium 

and its daughters, can be used to estimate inhalation intakes on an individual basis if urinalysis and 

related information is unavailable or spotty (e.g., for 1942 to 1948) or to estimate doses for 

comparison to doses calculated from individual urinalysis and other data [16]. Note, however, that 

nonuranium doses are not to be assigned for the SEC period, August 14, 1942, through November 

30, 1949 [10]. For workers whose work history showed job rotation, but the time spent in each job is 

uncertain, a choice of the job title giving the maximum exposure applied through the whole year, will 

be favorable to claimants. The steps are as follows:

1. The job title or work area selection(s) from Table B-22 should be made on the basis of the 

claimant's submitted information, urinalysis records, film badge records (if helpful), 

employment records, and other information. Table B-4 can also be used as an aid.

2. The annual intakes from Table B-16 should be selected to correspond to the job title or work 

area, plants, and periods.

3. Assumptions regarding isotopic content of the radioactivity in the air should be made as 

listed in Table 5-5, above.

5.9.2

Because health physics practices at Harshaw appear to have been substandard, ingestion intakes 

might have been significant. The effects of ingestion on the gastrointestinal (GI) tract are not well 

accounted for by the assumption of inhalation intakes. Thus, if the organ of concern is a GI tract 

organ, chronic ingestion intakes should be included in addition to the inhalation intakes (NIOSH 

2004). Table B-17, the table of annual ingestion intakes based on time-weighted daily average 

inhalation of uranium and its progeny, can be used to determine ingestion intake on an individual 

basis if urinalysis and related information is unavailable or spotty (e.g., for 1942 to 1948) or to 

generate doses for comparison to doses calculated from individual urinalysis and other data [17]. 

Note, however, that nonuranium doses are not to be assigned for the SEC period, August 14, 1942,

through November 30, 1949 [10]. The steps are as given below.

Determining Annual Ingestion Intakes Based on Time-Weighted Daily Average 

Inhalation Exposure Data (Table B-17), 8/14/1942-11/30/1949

1. The job title or work area selection(s) from Table B-22 should be made on the basis of the 

claimant's submitted information, urinalysis records, film badge records (if helpful), 

employment records, etc. Table B-4 can also be used to help make the selection.

2. The annual intake from Table B-17 should be selected to correspond to the job title or work 

area, plants, and periods.

3. Assumptions regarding isotopic content of the radioactivity in the air should be made as 

listed in Table 5-5, above.

5.9.3

Table B-15, the time-weighted daily average exposures for specific job titles, and Table B-14, the 

time-weighted daily average exposures for specific work areas, can be used to estimate intakes using 

Estimating Inhalation and Ingestion Intakes by Using Time-Weighted Daily Average 

Exposure Data (Table B-15), 8/14/1942-11/30/1949
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assumptions different from the standard ones used for Tables B-16 and B-17 [18]. The job title 

selection should be made on the basis of the claimant's submitted information, urinalysis records, film 

badge records (if helpful), employment records, and other information. Tables B-4, B-21, and B-22

should be used to help make the selection [19]. The air concentrations from Table B-15 (or in default 

of information in Table B-15, use Table B-14) should then be selected to correspond to the job title(s), 

work area(s), and periods. Any necessary adjustments should be made for partial years, overtime, 

etc. Note, however, that nonuranium doses are not to be assigned for the SEC period, August 14, 

1942, through November 30, 1949 [10].

The intakes, in picocuries, should be calculated by multiplying the appropriate air concentrations by 

the breathing rate(s) and the hours, and dividing by 2.22 dpm/pCi. Once the inhalation intake has 

been determined, the ingestion intake can be calculated by using the assumptions in NIOSH (2004):

Ingestion intake (pCi/yr) = 0.021 × Inhalation intake pCi/yr (5-2)

5.9.4

Table B-18, the table of annual radon exposures based on Mallinckrodt radon measurements

(ORAUT 2005b), should be used to determine bounding estimates of radon exposures if the organ of 

concern is a respiratory tract organ. Note, however, that nonuranium doses are not to be assigned for 

the SEC period, August 14, 1942, through November 30, 1949 [10]. Table B-18 was based on spot 

measurements usually taken at times of representative or maximum radon emanation, not on daily 

weighted average exposures, which are unavailable.

Estimating Annual Radon Exposure by Using Bounding Radon Exposure Data 

(Table B-18), 12/1/1949 on

1. The job title or work area selection(s) from Table B-22 should be made on the basis of the 

claimant's submitted information, urinalysis records, film badge records (if helpful), 

employment records, and other information. Table B-4 can also be used as needed.

2. The annual radon intake(s) from Table B-18 should then be selected to correspond to the job 

title or work area, plants, and periods.

In general, only workers who spent time in an area where radon might concentrate significantly, such 

as in the ore concentrate storage area, were likely to be exposed to any significant level of radon. 

Nonprocess workers, particularly office workers, can be assumed to have insignificant radon 

exposures.

5.9.5

As stated above, if urinalysis data are available for an individual, they should be used to determine the 

internal exposure to the individual. Little if any urinalysis data are expected to be documented in the 

D&D/postoperations years.

Estimating Intakes During the D&D/Postoperations Years by Using Calculated Data 

(Table B-25)

Because there is not a formal date that can serve as a cutoff for the dose reconstructor to use, 

judgment will have to be applied as to how much of the dose record after 1955 should be counted as 

contributing to dose from AEC decontamination operations [20]. Note that after the operations and 

D&D years, Harshaw received an AEC license to use certain radioactive materials. Care must be 

taken in the case of workers whose work histories show them apparently still doing work with 

radioactive materials after the D&D period; such work during the postdecontamination years was not 

AEC-contracted work; it is uncertain what work might have been done and what materials might have 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 32 of 114

been used, and any intakes found in records from the post-D&D years might be attributable to 

commercial operations with radioactive material, not to residual contamination from AEC work. But it 

is generally acceptable to assume that operations continued until the end of 1955 in the oreconcentrate-to-UO3 (brown) plant and the associated laboratories [21].

Annual intake estimates were calculated from measured and interpolated data listed in Table B-24 

and the results are listed in Table B-25. See the notes following the table for information on how the 

table was developed. For the D&D and postoperations years if individual data cannot be found or 

where it is not clear what the source of the intake data was, the data of Table B-25 should be used

[22]. In using Table B-25, the applicable years of employment for each indicated period should be 

determined and the number of years in the covered period and area should be multiplied by the 

annual value. The annual value can be prorated for partial years.

If the claimant was a process or other worker likely to have spent considerable time in areas of 

significant residual contamination and it is not clear in which building the worker actually spent time, 

the Building G-1 values can be used [23]. Intakes listed in Table B-25 can be ratioed to indicate time 

spent in the contaminated area.

6.0

6.1 INTRODUCTION

OCCUPATIONAL EXTERNAL DOSE

Conditions for external radiation exposure are best summarized by the description in AEC (1949b): 

"Severe exposures to external beta radiation ... exist in this plant." Beta dose and gamma dose to the 

extremities was potentially high for those workers handling hex ash and other residues. Because little 

individual worker monitoring data are available before about 1949, some extrapolation of existing data 

to cover the unmonitored periods is necessary; however, with the significantly lower quantities of 

material handled and produced in the laboratories and pilot plants, the external doses were not likely 

to be greater than they were later. Exposure rate information retrieved from various AEC reports was 

condensed into Tables B-5, B-6, B-7, and B-8 and is to be used to conduct external dose 

reconstructions for periods where personnel exposure data are not available. (A coworker 

assessment was not conducted for the Harshaw external exposure data set.) Individual extremity 

dose data appear to be lacking and must be inferred mostly from measured and/or calculated dose 

rates.

The nominal period of operations was 1942 to 1955. However, as Table 2-2 shows, not all of the 

operating areas started in 1942 or operated until 1955. Thus, for example, an entry of "491 Area" 

(UO2-UF4 area) in 1953 would indicate that D&D was being performed rather than process operations, 

while an entry of "493 Area" (ore-concentrate-to-UO3 area) at the same time would indicate process 

operations. If there is no information to the contrary, the periods of operation and D&D should be 

assumed to be as given in Table 2-2.

6.2 BASIS OF COMPARISON

Since the initiation of the MED in the early 1940s, various radiation dose concepts and quantities have 

been used to measure and record occupational dose. A basis of comparison for dose reconstruction 

is the Personal Dose Equivalent, Hp(d), where d identifies the depth (in millimeters) and represents 

the point of reference for dose in tissue. For weakly penetrating radiation of significance to skin dose, 

d = 0.07 mm and is noted as Hp(0.07). For penetrating radiation of significance to whole-body dose, 

d = 10 mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 33 of 114

recommended for use as the operational quantities to be recorded for radiological protection 

purposes.

Film badge records contain "beta" values that were obtained by subtracting the optical density of the 

film behind the cadmium shield from that behind the open window (Blatz 1950a,b,c). These recorded 

values can be assumed to be equivalent to Hp(0.07). The performance of the MED badge in 

determining Hp(10) under field conditions is less certain. Therefore, an approach favorable to 

claimants is adopted for converting exposure to tissue dose. See Section 6.3.2.

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1

Available film badge data tabulated weekly begin on 25 August 1947 as shown in dose records 

collected in various data captures by the EEOICPA Dose Reconstruction Project; these data are more 

or less continuous until the end of production.

Site Historical Administrative Practices

However, it seems clear that film badging started earlier than that. First, the document titled Harshaw 

Radiation Summary, Aug 44 to Mar 48 (University of Rochester ca. 1948) indicates the character of 

early doses as recorded in film badge results. The earliest results given appear to correspond to a 

badge start date of 29 August 1944. Although the author of this document is not clear, this file is 

similar to a summary prepared for the Mallinckrodt site and is clearly from the same film badge 

service, the University of Rochester (e.g., there are entries termed "Rochester Control" for control 

badges used during processing). The Harshaw summary contains badge results for 187 individuals, 

by name and Social Security Number; the number of weeks of employment for each employee (range 

from 1 to 181); the employee’s total gamma results (range from 0 to 6,890 mrem); the total beta 

results (range from 120 to 139,740 mrem); and the starting and ending dates of monitoring. Weekly 

doses from this list, averaged over all employees, are 2.5 mrem/wk from gamma radiation and 

741 mrem/wk beta equivalent. This corresponds well with Figure 8 of AEC (1949b), which centers 

most individual exposures on an axis of "700 mrep/wk.” Note that Table B-7 of this site profile 

includes data from Figure 8 of AEC (1949b).

Second, there are various mentions of film badging being done prior to 1947, although possibly with 

interruptions. AEC (Hayes 1949) discussed a Harshaw individual who was badged from late 1945 on. 

The MED and AEC (Tybout 1944a, b; Schoen 1958) mentioned beta or gamma measurements using 

film badges worn by workers apparently routinely. Becker (1946) stated that while the plant 

superintendent would go along with the discontinuance of film badge monitoring, he wanted some 

kind of substitute badge for security reasons; Mears (1946a) quoted an MED consultant as 

questioning the advisability of discontinuing film badge monitoring on the basis that the exposure to 

the hands was high and the (chest) film badge thus provided some indication of the extremity dose,

even if the registered (whole-body dose) was below tolerance. Despite this, an MED manager 

advocated discontinuing film badge monitoring on the grounds that the indicated exposures were 

consistent over the 2.5 years and that it should be continuation of monitoring that needed to be 

justified, and not the discontinuation (Mears 1946a). But Mears (1946b) decided that film badging 

was to be continued and every employee was to wear a badge, including supervisors. The issuance 

of film badges to everyone was so that the badge could double as a security badge (Mears 1946b). In 

addition, Blatz (1949a,b) and Eisenbud (1949d,e), dated February or March 1949, discussed a review 

of Harshaw film badge readings, with the details suggesting that badging had been going on for some 

time, and Klevin (ca. 1948) stated in November 1948 that film badges were not always being worn as 

required. Thus, it appears that a significant period of weekly film badge data, from approximately late 
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1945 through 1948, might be missing, although the data for most individuals will likely be found in a 

totalized format in the Harshaw summary discussed above (University of Rochester ca. 1948).

From film badge records, some scapular (shoulder-placed) film badges appear to have been worn in 

May 1949. As discussed in Section 6.3.2, extremity film badges were used only on an experimental 

or study basis due to difficulties (e.g., heat damage).

Film badges, generally, were issued weekly. Badges were turned in at the guardhouse when workers 

left the site and picked up at reentry (HCC 1946). AEC (1951c) stated that over the 13-wk period of 

1 November 1948 to 24 January 1949, 91 badges were worn per week; over the period 4 July 1949 to 

19 June 1950, 180 badges were worn per week.

Most existing records are labeled as being for Plant C, but occasionally, a film badge results card is 

labeled as being for "Plant E." It is unclear what the "E" might stand for. The dose records reflect 

decreasing numbers of badges issued through the early 1950s, to the point that the latter records for 

Plant C have no doses entered for most of the subjects, indicating that most were no longer badged 

and the plant was in shutdown status. At the end of 1954, only six names are listed on the film badge 

results card. This agrees with the statement in Klevin (1954) that in May 1954, there were only five 

workers left in the plant.

Most of the film badge results are listed on forms from the NYOO, with beta and gamma results for all 

monitored workers. Some records also list the optical densities from which the doses were calculated 

and some records list the total dose, computed from adding the beta and gamma together without 

modification. Many of the film badge records are handwritten and, for even the typewritten records, 

personnel names are omitted from some weeks’ results; also, Harshaw dose records list individuals to 

whom badges are issued, apparently, without regard to their week-to-week assignment. However, 

because individuals are associated with consistent badge numbers from week to week, EEOICPA 

Dose Reconstruction Project personnel compiling individual dose histories should have little difficulty 

in assigning doses with no listed names. It should also be noted that badge numbers were initially 

listed serially, with names in alphabetical order, but as employees were added or removed from the 

dose monitoring roster, the badge results continued to be listed in serial order and the names 

departed from strict alphabetical order.

Doses were recorded as "beta" and "gamma." Consistent with the demonstrated practice at the 

University of Rochester for Mallinckrodt (ORAUT 2005b), the result for beta is the open-window result 

with the dose under the shield subtracted. The result for gamma is the dose calculated from the 

optical density under the shield. Confirmation by inference was possible, as it was for Mallinckrodt, 

because there are a few beta results for Harshaw that record beta as less than the limit of detection 

(LOD) and gamma as some number greater than the LOD. 

As the set of monitored individuals shrank, individuals were gradually removed from the roster, often 

by simply lining through their names. The handwritten records often record results less than 

detectable as "X" in the appropriate box, and these are to be interpreted as zeros. As more and more 

individuals returned zero results, the handwritten Xs expanded to cover the appropriate number of 

columns and rows to indicate zero results (i.e., a single X could be written across multiple columns 

and rows). A careful distinction seems to have been made between lined-out entries in the rows and 

crossed-out dose results column. For the purposes of dose reconstruction, X results are to be 

interpreted as individually entered zeros for beta and gamma, while lined-out names are to be 

considered unmonitored personnel for that week. Notations in the film badge records indicate when a 

badge was missing, when a readout was unsuccessful, etc.
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Note that the Rochester Harshaw dose summary for 1944 to 1948 (University of Rochester ca. 1948) 

leaves blanks in the gamma column (though never in the beta column), which are interpreted to be 

zero results. Individual dose report results list many zeros for gamma also, with some also recorded 

as "less than 50" (i.e., they are denoted as 50* with a footnote indicating that the asterisk "denotes 

less than"); some weekly badge records give zeroes as blanks or Xs. All these should be considered 

zero results for the purposes of dose reconstruction [24]. Also, most gamma results in the Rochestergenerated records are rounded to the nearest 10. Those results listed with 1 to 9 in the last column 

might be the result of averaging because a handwritten note at the top of the dose summary file states 

that "these figures include average values inserted where badges were lost or readings missed for

some other reason."

For 1944 to 1946, it is likely that the film badge exchange frequency was weekly, but individual 

records or use of the Rochester summary of dose prior to August 1947 might involve a different 

assumption according to the individual case. The exchange frequency should be assumed to be 

weekly for 1947 on, except for the several years near the end of operation when records indicate the 

consolidation of the last 2 weeks of a year on a single film badge.

6.3.2

No procedures and little other film badge specification data have been found to date (e.g., there is no 

specific calibration information). However, because early badges were processed by Rochester, it is 

very likely that the calibration methods were those of MED/Rochester and later, of the AEC HASL.

Site Dosimetry Technology and Calibration

There is no information to suggest calibration using a phantom, so open-air calibrations were likely 

performed. Thus, it is recommended that the Harshaw-recorded gamma doses be converted using 

dose conversion factors for roentgen-to-HT dose for photons from Appendix B of NIOSH (2006). 

Because exposure to organ dose conversion factors result in a higher organ dose and higher 

probability of causation, given the Radiation Effective Factors of the intermediate energy photons, 

these dose conversion factors will be used to convert recorded film badge doses to organ dose. The 

low-energy component does not seem to be a significant characteristic of the Harshaw spectrum, thus 

no modification is proposed to recorded deep doses, once converted to organ doses.

Extremity dosimeters were not worn routinely at Harshaw. However, some information regarding 

extremity dose measured with films is available. The MED attempted to measure beta dose to the 

hands of hex loaders (from the ash) by using film strips (HCC ca. 1945; Engel 1946; Tybout 1946). In 

February 1946, an MED sergeant was sent to Harshaw to work for a few days as a hex tray loader 

and hex ash handler (Engel 1946). While working, he wore a film strip around each finger and a small 

piece of film in each palm, on each wrist, and on the back of the hand; all of these films were fastened 

on with Scotch® tape. Also, he wore one regular (chest) film badge and an additional chest film 

badge. The regular badge was worn all 4 days, but the hand films and the additional chest films were 

changed every day, except that only new palm badges were worn on the extremities during the ash 

handling operation. Although gloves were worn, perspiration and the heat of the trays rendered some 

of the films unreadable; still, some results were obtained as listed in Table B-6. Because of the heat 

problem, films were not used further as extremity dosimeters.

Although extremity dose was likely to have been high at Harshaw, the proportion of claims requiring 

calculation of extremity dose is unlikely to be high. Thus, this subject is not treated in detail in this site 

profile. Extremity dose estimates, when necessary, should be formulated on a case-by-case basis

[25].
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6.3.3

6.3.3.1 Beta/Photon Dosimeter Response

Workplace Radiation Fields

Both 235U and 238U are primarily alpha-particle emitters. However, 235U does emit a 185-keV photon in 

54% of its decays. Most of the external dose from 238U comes from its short-lived 234Th, 234mPa, and 

234Pa decay products. From an external dose standpoint, the most significant radiations emitted by 

these decay products of 238

U are: (1) the 2.29-MeV beta particle from 234mPa, and (2) the photons 

emitted by 234Pa with energies as large as 1.962 MeV. Photons should be assumed to be in the 30-

to-250 keV energy range, consistent with NIOSH (2006). Beta (electron) radiation should be 

assumed to be in the range greater than 15 keV. Radiation in this range is the primary external dose 

component for Harshaw.

Table B-5 lists measured gamma and total dose rates at various locations and times as reported by 

the AEC and Harshaw. Table B-6 lists chest and hand beta doses from ash residue handling as 

measured by film badge and films taped on the hands. Table B-7 lists weekly doses tabulated for the 

period from August 1944 to January 1949 by the AEC (1949b).

The total dose on film badges was more than 95% beta as noted by the AEC [1949b, 1951c (referring 

to 1948 to 1951)] and as shown by film badge records. The percentage might have been even higher 

for extremities. The AEC was concerned by Harshaw’s lack of close monitoring of beta exposures 

and cited this several times as a failure of the Harshaw health and safety program (e.g., Kelley 1947). 

The AEC (1951c) noted that in 1949, about 25% of the hex plant workers received beta exposures 

higher than tolerance, but by 1951 that figure had dropped to 10%; similarly, in 1949, "hardly a month 

went by" without a number of badges registering over 1,000 mrep/wk, but by early 1951 there was 

only an occasional badge registering that high. This was attributed to better personnel control and to 

the 1950 addition of the central loading/unloading station for hex reactor trays.

Doses registered on film badges worn by people not working directly with the uranium and the 

process and analytical equipment, such as guards and office workers, were more likely from gamma 

exposure than from beta exposure. This is because these workers were usually at some distance 

from the source (the uranium and its progeny). It is true that uranium-containing dust was found 

throughout the plant to varying extents, but that would likely not contribute substantially to the external 

dose rate much in buildings or areas distant from the process areas [26].

In addition to the beta dose rate from the uranium as natural uranium, uranium oxide, etc., two waste 

concentrates produced significant beta dose rates. First, when ether was used with the uranyl nitrate 

to extract the uranium in the ore-to-UO3 production process, 234Th and 234mPa (again, UX1 and UX2,

respectively) were left in the aqueous phase (Eisenbud 1975). When this aqueous solution was 

filtered, the resulting cake(s) contained most of these beta-gamma emitters (Eisenbud 1975). 

Second, the hex reactor ash, as stated above, was highly concentrated in the 234Th and 234mPa from 

the UF4 (AEC 1949b). The highest extremity doses at Harshaw were probably from this source.

6.3.3.2 Uncertainty and Bias for Beta/Photon Dosimeters

There was no quantitative information recovered for the film badge used at Harshaw. Information 

from other sites (Y-12, Hanford) indicates that the uncertainty for a two-element badge of the era 

could be estimated as ±30%. Similarly, information on bias was not located. It is likely that some 

factors such as the calibration techniques could have resulted in recorded doses that were too high 

while other factors such as the angular response and wear location could have resulted in recorded 

doses that were too low for the organ(s) of interest [27].
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6.3.3.3 Neutron Doses

No neutron exposure measurements are available. However, Dupree-Ellis et al. (2000) deemed 

neutron exposures at a similar uranium production facility (the Mallinckrodt site) to be minimal. This 

conclusion seems to be correct for the Harshaw UO3, UO2, and UF4 production processes too, due to 

the similarity of the production processes, and likely for the UF6 production as well. The enriched 

uranium hexafluoride (LEUF6) from Hanford (Kelley 1946) and the Hanford recycle UO3 (BJC and 

Haselwood Enterprises 2000) that Harshaw processed from 1952 to 1954 also do not appear to have 

involved significant levels of neutrons because no extra precautions appear to have been thought 

necessary due to the low enrichment level of the former and the relatively low transuranic content of 

the latter. (See Table B-3 for details of the content of the LEUF6 and the RU.)

In analyzing neutron production by the alpha-neutron reaction, the forms of uranium that would 

produce neutrons at the highest rates were identified as UF4 and UF6, but UO2, UO3, and U3O8 and 

the soda salt form Na2U2O7 were also considered [28]. Little information could be found about UCl4, 

so it was considered in terms of identifying another form that would bound its contribution (i.e., UF4). 

As long as an adequate amount of target material (fluorine, oxygen, sodium, or chlorine) is available 

and it is intimately mixed with the source material (uranium, thorium, or their alpha-emitting progeny), 

as would typically be true in the forms handled at Harshaw, neutron production essentially depends 

on the amount of the source material. The bounding assumption is therefore made that the maximum 

neutron emission occurs (i.e., that there is an adequate amount of target material whatever the form)

[28].

As listed in Table B-3, a fiber or steel drum container of UO3, UO2, or soda salt would weigh 75 lb and 

a steel drum container of U3O8 (ore feed as black oxide) would weigh 100 lb; most of this weight 

would be uranium, so it can reasonably be assumed that the entire weight is uranium. While a larger 

volume could be found in, for example, a digest tank, the liquid and the thick tank wall would provide a 

great deal of shielding. A larger volume could be found in a massed array of containers, but a great 

deal of self-shielding would be involved and a person would likely not spend a great deal of time near 

an array. Thus, it is likely that the dose rate from a single container (being temporarily stored, loaded,

transported, or dumped) would be the typical dose source and this was the source form analyzed. 

The dose rate from a massed array, however, could be estimated by multiplying the single-container 

dose rate by the number of containers.

In the dose rate calculations, assumptions and data from ORAUT (2005c) were used; these included 

the assumptions that there was a point-source geometry, the isotopic composition in the source 

container was that of natural uranium, and the energy of the neutrons produced was 2.0 MeV. The 

resulting whole-body neutron dose rates are listed in Table B-8. To consider the effect of including 

progeny contributions, for full equilibrium of the progeny of natural uranium down to radon (which 

would not be in chemical union with the target and would likely have been vented when the container 

was opened, the form processed, etc.) or for the extreme case to polonium, the additional contribution 

of the alpha-emitting progeny of 238U and 235U was also considered where appropriate. Thus, the 

progeny contribution was included in Table B-8 when there was time for the progeny to build in

significantly (i.e., the ore concentrate and soda that were being sent from various other sites and the 

UO2 and UF4 that were being sent from Mallinckrodt Chemical Works in St. Louis). Note that the UF4

appears to have been sent in significant quantities only from 1945 to 1947, constituting at most 28% 

of the amount used (i.e., Harshaw produced 72% of the UF4 it used during that period and 100% at 

other times).

In the Harshaw-specific calculation of the annual whole-body doses performed for inclusion in this site 

profile, exposure time and distance estimates favorable to claimants were made based on time 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 38 of 114

measurements and estimates given in memoranda, dust studies, etc [29]. Although production 

increased over time, it is assumed that the process worker population grew more or less 

proportionately so that each worker type did the same amount of work on an annual basis (e.g., 

loaded or moved the same number of drums). It was assumed that the process worker-receptor 

spent 1 hr/d at 1 ft from the container with the uranium form in it and 3 hr/d at 3 ft from it, every 

working day for a 2,000-hour workyear [29]. The laboratory worker was assumed to handle much 

smaller sizes of containers (e.g., 2-quart sample jars and the like), taken to contain 10% of the mass 

of the corresponding larger containers, but to spend all of the work time near the containers (e.g., 

sitting on benches or hoods in the laboratory), taken to be 2 hr/d at 1 ft and 6 hr/d at 3 ft from the 

containers [29]. The office/clerical/ management worker was assumed to spend an insignificant 

amount of time near these containers, with their resulting doses being negligible; this is reasonable 

based on the inspection of the process workers and laboratory worker doses [29].

The occupancy time assumptions used in the calculation of Table B-8 should be adjusted for workers 

not likely to have spent considerable time near the uranium forms; for these workers, the doses 

should be ratioed by an appropriate fraction to reflect the time spent near uranium forms in bulk. A 

fraction of 0% is suggested for office workers and shop workers, 5% for higher level managers, and 

25% for maintenance and safety workers who were likely to have spent time in process areas [based 

on engineering judgment, given the information in the various dust studies (including time-and-place 

information), observations in AEC and Harshaw memos and reports, and statements in HCC (1946)]. 

Note that no estimate of neutron dose for postoperations and decontamination work need be made 

because of the small volumetric concentrations of the residual contamination (i.e., mostly surface-type 

deposits on walls, floors, and equipment). Neutron radiation should be assumed to be in the range of 

0.1–2.0 MeV. The neutron doses in Table B-8 should be multiplied by 2 to correct the values 

calculated from ORAUT (2005c) to the ICRP Publication 60 radiation weighting factor for this energy 

range (ICRP 1991).

6.4 ADJUSTMENTS TO RECORDED DOSE

No adjustments to recorded dose are proposed for Harshaw at this time.

6.5 MISSED DOSE

The LOD for beta reported from 9 May 1948 on AEC forms (NYOO 1948) was less than 50 mR for 

both beta and gamma. Since the University of Rochester read the Harshaw films up to May 1948, it 

will be assumed that the LOD was 60 mrep up to May 1948 and 50 mrep from May 1948 on. Thus,

per NIOSH (2006), it should be assumed that beta missed dose should be applied as LOD/2, or a 

lognormal distribution with a mean of 0.030 rem/wk prior to May 1948 and 0.025 rem/wk thereafter, 

with a GSD of 1.52 for Rochester records.

The LOD for gamma dose should be assumed to be 50 mR/wk, based on records generated by the 

University of Rochester; these record gamma results as either zero or "less than 50" if they were less 

than measurable. (It is not clear why they chose to use one or the other.) It should be assumed that 

per NIOSH (2006), gamma missed dose should be applied as LOD/2, or a lognormal distribution with 

a mean of 0.025 rem/wk with a GSD of 1.52.

6.6 DETERMINATION OF EXTERNAL DOSES (INSTRUCTIONS TO DOSE 

RECONSTRUCTORS)

Representative external dose histories can be compiled for employees with work histories beginning 

after about August 1944. When available, individual film badge data should be used to determine the 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 39 of 114

exposure. Some help in interpreting the film badge records can also be found in the urinalysis 

records because the latter could clarify what type of work the worker was doing at a particular time. 

Most workers employed in the early years will have some gaps in monitoring because routine film 

badging did not begin until 1947 (although some data are available from 1944 on) and because there 

were undoubtedly some missed readouts. If individual data are not available, the exposure rate 

information listed in Tables B-5 through B-8 are to be used to reconstruct a claimant’s external 

exposure.

Although some improvements were instituted from 1942 to 1946, the start of routine film badging in 

1947 is correlated with significant UF4 and UF6 production increases or with the peak of production,

and the years after 1946 mostly cover the period of significant external exposure problem reporting. 

Hence, it is deemed that the later measurements are reasonably representative and their use would 

be favorable to claimants if data gaps exist for earlier conditions. To what extent exposures during 

early bench-level operations differed from production-level exposures is not known. However, 

processes were developed on the bench and pilot plant levels and then in many cases quickly scaled 

up to production levels; the production levels then increased repeatedly throughout the wartime and 

early postwar years. So, it can be concluded that the conditions for bench-level and pilot plant 

operations were similar to production level operations, but on a much smaller scale (including 

generally much smaller source quantities) [30].

Note that many workers worked at two or more of the UO2-to-UF4, UF4-to-UF6, and ore-concentrateto-UO2 or -UO3 plants, with their differing potential for external exposure; especially for beta exposure. 

So if it is necessary to determine the maximal job for an individual having several jobs over the course 

of a year, care should be taken to identify which job is indeed the maximal one for the organ of 

interest. Care must also be taken in the case of workers whose work histories show them apparently 

still doing work with radioactive materials after the D&D period; such work during the 

postdecontamination years was not AEC-contracted work, it is uncertain what work might have been 

done and what materials might have been used, and any doses found in records from the post-D&D 

years might be attributable to private operations with radioactive material, not to residual 

contamination from AEC work. But it is generally acceptable to assume that operations continued 

until the end of 1955 in the ore-concentrate-to-UO3 (brown) plant and the associated laboratories. If 

there are no individual data available for this period, Table B-26 should be used, as explained below. 

See the notes following the table for information on how the table was developed.

6.7 DETERMINING EXPOSURE DURING THE OPERATIONS YEARS

6.7.1

As noted, film badging at Harshaw began in at least intermittent fashion in August 1944, although it 

does not appear to have become routine until 1947. The AEC directive to badge all employees who 

might be subject to significant external exposure suggests that it is reasonable to assume that 

exposed employees engaged in MED/AEC work will have at least some film badge results for their 

covered employment, and individuals with no badge results are unlikely to have received anything but 

incidental exposure. Individual dose histories are likely to contain gaps due to missing or damaged 

badges and, especially for earlier periods, "fogged films."

Reconstruction of Doses from August 1944 to 1955

How missing doses can be most accurately estimated is based on what dose data are available. As 

discussed by Watson et al. (1994), the most accurate estimate for a missing annual dose is an 

average of the doses recorded in the years before and after. For the purposes of dose reconstruction, 

this will be assumed to apply to shorter periods as well if the worker clearly worked in the same work 

area and did the same type of work. In such cases, data for a missing year should be filled by 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 40 of 114

averaging the dose for the year before and the year after; data for shorter periods (a missing quarter, 

month, or week) should be filled by averaging appropriate periods before and after the missing period.

If the worker did different work in the years or periods bracketing the missing year or period, the data 

for the missing period should be based on the maximally dose-producing of the two types of work. In 

the case that this method produces clearly adequate results or the method produces inconsistent 

results, exposure rates listed in Table B-5 are to be used with time estimates from dust studies (e.g., 

AEC 1953a) to produce reasonable exposures for comparison.

6.7.2

For workers involved in uranium processing between the beginning of Harshaw work for the MED in 

1942 and the beginning of monitoring in August 1944 who have later dose monitoring results, the 

most accurate estimate of annual doses is likely to come from dose monitoring information for later 

years, if available. As stated in Watson et al. (1994), dose information from the closest period should 

be used to estimate missing dose information. For the early period, the best information is likely in the 

Rochester Radiation Summary (University of Rochester ca. 1948). For individuals listed in that 

summary, the average dose should be applied to each year preceding the period covered by the 

summary. Although application of a simple average will introduce considerable uncertainty, given 

changes in the process and job assignments, it is likely to be an assumption favorable to claimants 

because production levels were steadily increasing over the period (from about 5 lb/d in 1942 to about 

4,200 lb/d in June 1947, shortly before the date of the first routine individual dose monitoring results) 

so the period of the radiation summary covers a time with likely overall higher dose hazards than the 

period before the summary [31].

Reconstruction of Doses Prior to August 1944

For workers involved in uranium processing between the beginning of Harshaw work for the MED in 

1942 and the beginning of monitoring in August 1944 who have no later dose monitoring results (e.g., 

those who might have terminated prior to the start of monitoring), the data in Tables B-5 through B-8 

should be used to establish an exposure estimate. It is unlikely that individuals with no available dose 

monitoring records worked in the uranium operations at Harshaw for more than a short time. 

However, if such a case should present itself – if the evidence indicates that the worker was present 

in the uranium operations and the monitoring results are missing – the data in the tables should be 

used to provide an external exposure estimate that is favorable to claimants.

6.7.3

External dose monitoring was performed for workers directly involved in the uranium processing 

operations, but possibly not always for workers employed in a support capacity. For example, 

process workers, warehouse workers, maintenance workers who entered process areas or received 

process equipment in their shops, safety workers, and all these workers’ supervisors and managers 

were badged; laboratory personnel and process area clerks also were badged [32]. But it is not clear 

that secretaries, nonprocess clerks, and the like were badged. It is reasonable to assume that these 

unmonitored individuals associated with the AEC uranium operations did receive some radiation dose, 

however, due to the possibility that their work locations were in buildings in or near the uranium 

processing area. Dose rates from surveys are listed in Table B-5. The 50th percentile of this data is 

about 7 mrem/hr (presumably 95% or greater is beta dose based on film badge results) [33]. Weekly 

doses for monitored workers are listed in Table B-7, which can serve as an upper bound on dose 

estimates for unmonitored workers. Dose reconstructors should estimate the dose based on the 

information in the individual case. For example, a worker described as a laborer might be assumed to 

have a higher occupancy in radiation areas than a secretary.

Estimating Incidental Dose for Individuals Employed In Uranium Processing but Not 

Involved in Operations
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 41 of 114

6.8 DETERMINING EXPOSURE DURING THE POSTOPERATIONS AND D&D YEARS

Some judgment will have to be applied as to how much of the dose record after 1955 should be 

counted as contributing to dose from AEC decontamination operations. This is particularly true 

because, after the period of covered operations and D&D years, Harshaw received a license from the 

AEC for the use of certain radioactive materials and some workers might have continued to be 

badged [34]. It will be generally acceptable, however, to assume that operations continued until the 

end of 1955 in the ore-concentrate-to-UO3 (brown) plant and the associated laboratories [21].

Because measured dose rates were available, RESRAD-BUILD was not used to calculate external 

exposures. Instead, manual calculations of annual external exposures were performed to estimate 

gamma and beta dose rate values that are favorable to claimants.

Annual external exposure estimates were calculated from measured data and are listed in Table B-26. 

The "AEC Decontamination" set should be applied for the D&D of Plant C only; the "PostDecontamination" set should be applied for 1960 to 1989 to Plant C only; and the "Post-AEC 

Operations, Decontamination (Continuing Source Term)" set should be applied for 1960 to 1992 or 

1960 to 1997 to the remaining buildings, as appropriate from Table 2-2. The latter set is applied as if 

the dose rates were continuous (i.e., not being reduced during D&D) due to the lack of dose rate data 

for D&D, an assumption favorable to claimants.

Energy bin assignment should be made as given in Section 6.3.3. The stay time assumptions used 

for Table B-26 (see the text after the table) should be adjusted for workers not likely to have spent 

considerable time in the areas of residual contamination, especially Plant C; the doses listed in 

Table B-26 should then be ratioed by an appropriate fraction to indicate a reasonable amount of time 

spent in the contaminated area. For example, a claimant who was a secretary in the postoperations

years likely did not spend much time in the areas of significant contamination and should be assigned 

only a small fraction of the doses listed in Table B-26. A fraction of 5% is suggested for office 

workers, 10% for higher managers, and 25% for maintenance and safety workers [based on 

engineering judgment, given the information in the various dust studies (including time-and-place 

information), observations in AEC and Harshaw memos and reports, and statements in HCC (1946)].

7.0

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction.

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

ATTRIBUTIONS AND ANNOTATIONS

[1] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This observation was based on several documents. Although hex operations had ended, 

these documents give the impression that the area was not decontaminated immediately. The 

area was instead converted to other uses or held in a standby condition. It is therefore logical 

that it was not completely decontaminated.

[2] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

While this statement is speculative, it is based on the common sense observation that more 

airborne dust would be likely during operations than during D&D. While dust could be raised 
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 42 of 114

during D&D, it would most likely decrease with time unlike the high steady-state levels during 

continuous operations.

[3] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

The percent of 235U by weight in depleted uranium varies according to the reference used but 

is generally in the range of 0.2% to 0.25%. Because of this uncertainty, the specific activity of 

depleted uranium is rounded to one significant figure here.

[4] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

The activity fractions of plutonium, neptunium, and technetium were chosen to be 

conservative. For example, BJC and Haselwood Enterprises (2000) gives the maximum 

fraction of plutonium as 11 and 9 ppb for 1953 and 1954, respectively. Therefore, the 

resultant doses from the RU contaminants are very likely to be overestimates.

[5] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Uranium isotopes and their progeny are primarily alpha emitters. Other emissions are not as 

important for internal dose.

[6] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

An increase in the various factors mentioned would increase the exposure of AWE workers to 

radon.

[7] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

A review of approximately 50% of Harshaw claimant files in 2007 did not find any X-ray 

records. The estimate of an initial, annual, and termination X-ray is based on limited site 

documentation and practices at other sites. It was judged to be more likely that some of these 

X-rays would be missed rather than occur more frequently. Therefore, this assumption is 

favorable to claimants.

[8] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

A review of approximately 50% of Harshaw claimant files in 2007 did not find any X-ray 

records. Assumption of X-rays of the pelvis for process workers in the given date range is 

favorable to claimants.

[9] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

If used in dose reconstructions, contaminated samples would result in a higher intake estimate 

and would therefore be favorable to claimants.

[10] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This statement was added to ensure dose reconstructors do not attempt to use the tables to 

assign nonuranium doses during the SEC period.

[11] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

The examples can be found in several documents in the SRDB. See for example SRDB Ref 

IDs 10503, 10671, 10675, 11162, and 11646.

[12] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This statement is made based on professional judgment because of the similarity of the 

uranium operations at Mallinckrodt.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 43 of 114

[13] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Bioassay activity fractions were calculated assuming that 230Th, 226Ra, and 210Po are in secular 

equilibrium with 238U/

234U and that 231Pa and 227Ac are in secular equilibrium with 235U. The 

specific activities of 238U/

234U and 235U in natural uranium were divided by the specific activity 

of natural uranium to determine the fractions. Alpha air activity fractions were calculated by 

summing the activities of these isotopes in the bioassay activity fraction column and then 

dividing each activity fraction in that column by the total.

[14] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

The coworker study was conducted using methods known to be favorable for uranium sites. 

Details are provided in Attachment A.

[15] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

These tables provide the only information likely to be available for the D&D period. The 

methods used and assumptions are listed after each table.

[16] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Coworker data is not available before December 1949; therefore, Table B-16 is to be used to 

estimate inhalation doses for this period. Values were calculated by multiplying the DWEs in 

Table B-15 by the breathing rate per day and the days worked per year. DWEs were 

averaged when more than one value is given per year in Table B-15.

[17] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Coworker data is not available before December 1949; therefore, Table B-16 is to be used to 

estimate ingestion doses for this period. The values in this table were calculated by 

multiplying the annual inhalation intake by 0.021. This factor is derived from NIOSH (2004).

[18] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This statement is based on the fact that in some cases more specific information might be 

available. In these cases, dose reconstructors can use the specific information rather than the 

Table B-16 and B-17 annual intakes.

[19] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

These tables provide information that could be useful to the dose reconstructor to determine 

individual-specific factors when calculating inhalation and ingestion doses from air-sampling 

data.

[20] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Individual-specific information might or might not be available to the dose reconstructor. If, for 

example, the record indicates that the employee was transferred to an administrative position 

as of a certain date, dose calculations could end as of that date. However, it is favorable to 

claimants to calculate doses to the employment termination date, in lieu of such information.

[21] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This statement is based on information presented in Section 2.0.

[22] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Table B-25 provides the means to assign internal doses in lieu of bioassay, which is unlikely to 

be available for the postoperations years. The assumptions that were used to calculate the 

values are given in the text following the table.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 44 of 114

[23] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

“Building G-1” was the term used to describe the area of most of the operations. The values 

are representative for most workers.

[24] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

A review of the forms indicates that this was the probable meaning of the entries.

[25] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Not enough information is available to formulate a site-specific method for calculating 

extremity dose.

[26] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Levels of contamination were likely to be lower in buildings or areas distant from the process 

areas. See examples of dose rates in areas closer to highly contaminated areas such as 

offices etc. in Table B-5. It is logical that the rates in more distant areas were lower.

[27] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This statement is based on the fact that calibration with a source of higher energy than was 

found in the workplace was common in the early days. This would typically result in an overresponse in the dosimeter. The location and orientation of the dosimeter in relation to the 

radiation sources (above, behind, below), would typically cause an under-response of the 

dosimeter relative to the dose received.

[28] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This is the same approach as in ORAUT (2005c).

[29] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

See Tables B-10 to B-13 for examples of dust studies. The time and distance values here are 

based on professional judgment, and dose reconstructors can change them as appropriate in 

individual cases. See also the times for close contact in Table B-6.

[30] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This is a reasonable default assumption since no information is available.

[31] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

Because production levels were lower during the period before the summary, use of the 

summary data should be favorable to claimants.

[32] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

See Table B-7 for examples.

[33] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

The total dose rates (beta plus gamma) in Table B-5 were analyzed to produce the estimate of 

the 50th-percentile value. The statement that 95% of the total dose on film badges was beta is 

based on the AEC references listed in Section 6.3.3.1.

[34] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This statement is an assumption based on the continued use of radioactive material at 

Harshaw as indicated by the AEC license.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 45 of 114

[35] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

This does not affect the fitting of the data for intake determination because all uranium 

isotopes behave the same biokinetically and the isotopes considered in this analysis have long 

half-lives relative to the assumed intake period. The ICRP Publication 68 dose coefficients 

(also referred to as dose conversion factors) for 234U are larger than those for 235U and 238U.

Because of the isotopic compositions of the source terms, the 234U dose conversion factor will 

overestimate doses.

[36] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. February 2007.

The coworker study for Harshaw was developed using the methods described in ORAUT 

(2005d). The 50th-percentile intakes are assigned as the intake, and the 84th-percentile is 

used to determine the GSD for each intake (GSD = 84th/50th percentile intake for each 

period), with the exception that no GSD less than 3 may be assigned.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 46 of 114

REFERENCES

Some of the cited documents contain Export Controlled Information (ECI); access to those documents 

is limited to paper copies.

AEC (U.S. Atomic Energy Commission), 1948, Occupational Exposure to Radioactive Dust, Harshaw 

Chemical Company, 493 Area, Harshaw-1, November 26.

AEC (U.S. Atomic Energy Commission), 1949a, Air Pollution Survey in the Vicinity of the Harshaw 

Chemical Co., H-1, July 28.

AEC (U.S. Atomic Energy Commission), 1949b, Health Hazards in NYOO Facilities Producing and 

Processing Uranium, April 18.

AEC (U.S. Atomic Energy Commission), 1950a, Monthly Report of Field Activities, October 1950, 

Health and Safety Division, November 6.

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Plant (493), Harshaw Chemical Company, Harshaw-5; October 30.

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AEC (U.S. Atomic Energy Commission), 1951d, Occupational Exposure to Radioactive Dust -- 491-

492 Area, Harshaw Chemical Company, Harshaw-9, December 10.

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Area, Harshaw Chemical Company, Harshaw-10, June 20.

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Area, Harshaw Chemical Company, Harshaw-11, December 5.

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493 Area, Harshaw Chemical Company, Harshaw-13, November 16.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 47 of 114

AEC (U.S. Atomic Energy Commission), 1953b, Occupational Exposure to Radioactive Dust -- 493 

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Blatz, H., 1951, “Visit to Harshaw Chemical Works, Cleveland, Ohio -- November 27, 1951,” 

memorandum to M. Eisenbud, U.S. Atomic Energy Commission, November 28.

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Conference, HASL-58, New York City, U.S. Atomic Energy Commission, pp. 10-15.
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Burman, L. C., 1949, “Visit to Harshaw Chemical Company April 20 and 21 -- Accountability and 

Losses,” memorandum to F. Belmore, U.S. Atomic Energy Commission, April 25.

Chrestia, F. A., 1948, “Monthly Report for Production Branch, January 1948,” memorandum to 

Production Office Files, U.S. Atomic Energy Commission, February 13.

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Inspection by Attributes, Washington, D.C., May 10.

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March 8.

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Commission, New York Operations Office), Harshaw Chemical Company, June 12.

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(U.S. Atomic Energy Commission), Harshaw Chemical Company, July 29.

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memorandum to S. L. Warren, Manhattan Engineer District, January 3.

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Glauberman, H., and W. B., Harris, 1958, “Air Sampling Procedures in Evaluating Exposures,” in 

Symposium on Occupational Health Experience and Practices in the Uranium Industry: 

Proceedings of a United States Atomic Energy Commission Conference, HASL-58, U.S. 

Atomic Energy Commission, New York, New York, pp. 208–211.

Harris, W. B., 1949a, “November 10 Visit to Harshaw Chemical Co.,” memorandum to M. Eisenbud, 

U.S. Atomic Energy Commission, November 16.

Harris, W. B., 1949b, “Stack Losses, Harshaw Chemical Co.,” memorandum to M. Eisenbud, U. S. 

Atomic Energy Commission, December 19.

Harris, W., 1949c, “Stack Effluents, Harshaw Chemical Co.,” memorandum to M. Eisenbud, U.S. 

Atomic Energy Commission, July 1.

Harris, W. B., 1949d, “Harshaw Chemical Co.,” memorandum to M. Eisenbud, U.S. Atomic Energy 

Commission, March 7.
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Harris, W. B., 1951, “Urine Samples -- Harshaw Chemical Company,” memorandum to E. C. 

Sargeant, U.S. Atomic Energy Commission, September 20.

Harshaw, W. J., 1945, Brief History of Work on the Manhattan District Project at the Harshaw 

Chemical Company, Cleveland, Ohio, Harshaw Chemical Company, Cleveland, Ohio, 

September 13.

Hayden, R. E., 1948, “Health Survey of Harshaw Chemical Company,” memorandum report to M. 

Eisenbud, U.S. Atomic Energy Commission, May 4.

Hayes, R., 1947, untitled letter to B. S. Wolf (U.S. Atomic Energy Commission), University of 

Rochester, Rochester, New York, September 19.

Hayes, R., 1949, untitled letter to M. Eisenbud (U.S. Atomic Energy Commission), University of 

Rochester, Rochester, New York, March 30.

HCC (Harshaw Chemical Company), ca. 1945, Medical History of Harshaw Chemical Company.

HCC (Harshaw Chemical Company), 1946, Operating Manual, The Manufacture of Uranium 

Hexafluoride, A-4003, October 22. [Contains ECI]

HCC (Harshaw Chemical Company), 1950–1953, health physics activities, letter reports from A. J. 

Stefanec or J. McKelvey to P. B. Klevin (U. S. Atomic Energy Commission), AEC. Variously 

dated from 26 May 1950 through 22 July 1953.

Hearon, W. M., 1945, untitled memorandum to M. J. Barnett, Manhattan Engineer District, May 24.

Howland, J. W., 1947a, “Medical Program -- Harshaw Chemical Company, Cleveland, Ohio,” 

memorandum to J. B. Quigley, U.S. Atomic Energy Commission, March 28.

Howland, J. W., 1947b, untitled letter to B. S. Wolf (U.S. Atomic Energy Commission), University of 

Rochester, Rochester, New York, September 3.

Hunter, D., 1949a, “Harshaw Recovery System -- Hex Plant,” memorandum to Files, U.S. Atomic 

Energy Commission, June 3.

Hunter, DeK., 1949b, “Record of Negotiations for Supplemental Agreement to Contract W-7408 Eng276, Harshaw Chemical Company,” memorandum to J. C. Clarke, U.S. Atomic Energy 

Commission, September 26.

Hunter, DeK., 1949c, “Harshaw "Hot Ash",” memorandum to H. Blatz, U.S. Atomic Energy 

Commission, November 10.

ICRP (International Commission on Radiological Protection), 1982, Protection of the Patient in 

Diagnostic Radiology, Publication 34, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 51 of 114

ICRP (International Commission on Radiological Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

Kelley, Lt. Col. W. E., 1946, “Disposal of Slightly Enriched C-616,” memorandum to Col. S. L. Brown, 

U.S. Atomic Energy Commission, January 17.

Kelley, W. E., 1947, “Health Program at "Area C",” letter to K. E. Long (Harshaw Chemical Company), 

U.S. Atomic Energy Commission, August 5.

Kelley, W. E., 1948, Monthly Status and Progress Report for May 1948, NYOO, U.S. Atomic Energy 

Commission, June 7. [Contains ECI]

Kelley, W. E., 1949a, untitled letter to W. J. Harshaw (Harshaw Chemical Company), U.S. Atomic 

Energy Commission, January 11.

Kelley, W. E., 1949b, Monthly Status and Progress Report for September 1949, NYOO, October 10.

Klevin, P. B., ca. 1948, “Violation of the Medical Division's Health Program at Harshaw Chemical

Company,” memorandum to W. B. Harris, U.S. Atomic Energy Commission, November 22.

Klevin, P. B., 1950a, Harshaw Chemical Company Occupational Exposure to Radioactive Dust in 

New Brown Plant, U.S. Atomic Energy Commission, Health and Safety Division, March 27.

Klevin, B. M., 1950b, “Particle Size Measurements -- Harshaw Chemical Company,” memorandum to 

W. B. Harris, U.S. Atomic Energy Commission, September 11.

Klevin, P. B., 1950c, “Preparation of UX1 at Harshaw Chemical Company -- Plant C Laboratory -- July 

14, 1950,” memorandum to H. Blatz, U.S. Atomic Energy Commission, August 2.

Klevin, P. B., 1952a, “Harshaw Chemical Company -- Visit of January 24-25, 1952,” memorandum to 

W. B. Harris, U.S. Atomic Energy Commission, February 6.

Klevin, P. B., 1952b, “Harshaw Chemical Company -- Visit of October 8, 1952,” memorandum to W. 

B. Harris, U.S. Atomic Energy Commission, October 29.

Klevin, P. B., 1952c, “Visit to Harshaw Chemical Co., Cleveland, Ohio on April 3, 1952,” memorandum 

to W. B. Harris, U.S. Atomic Energy Commission, April 14.

Klevin, P. B., 1953a, “Visit to Harshaw Chemical Company -- October 26-27, 1953,” memorandum to 

W. B. Harris, U. S. Atomic Energy Commission, November 5.

Klevin, P. B., 1953b, “Harshaw Chemical Company -- Brown Packaging Plant -- Cleveland, Ohio,” 

memorandum to M. Eisenbud, U.S. Atomic Energy Commission, March 11.

Klevin, P. B., 1954, “Harshaw Chemical Co. -- Visit of May 6,1954,” memorandum to W. B. Harris, 

U.S. Atomic Energy Commission, May 11.

Klevin, P. B., 1955a, “Radiation Survey of Harshaw Refinery Equipment and Plant C Premises,” 

memorandum to W. B. Harris, U.S. Atomic Energy Commission, January 10.

Klevin, P. B., 1955b, “Radiation Survey of Harshaw Plant ’C‘ Equipment, May 31 - June 1, 1955,” 

memorandum to W. B. Harris, U.S. Atomic Energy Commission, June 10.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 52 of 114

Lippmann M., 1958, “Correlation of Urine Data and Medical Findings with Environmental Exposure to 

Uranium Compounds,” Symposium on Occupational Health Experience and Practices in the 

Uranium Industry: Proceedings of an Atomic Energy Commission Conference, HASL-58, U.S. 

Atomic Energy Commission, New York, New York, pp. 103–114.

Long, K. E., 1947, untitled letter to W. E. Kelley (U.S. Atomic Energy Commission), Harshaw 

Chemical Company, August 20.

Lynch, D. E., 1949, “Waste Disposal, Harshaw Plant,” memorandum to Files, U.S. Atomic Energy 

Commission, February 9.

Mayer, C. M., and F. Proschan, 1949, “Visit to Harshaw Chemical Company, Cleveland, Ohio, on 

June 23, 1949 -- Preliminary Statistical Survey,” memorandum to F. M. Belmore, U.S. Atomic 

Energy Commission, June 28.

Mears, B. J., 1945, “Visit to Harshaw Chemical Company,” memorandum to J. L. Ferry, Manhattan 

Engineer District, July 27.

Mears, B. J., 1946a, “Film Badge Monitoring at Harshaw Chemical Company,” memorandum to F. A. 

Bryan, Manhattan Engineer District, January 31.

Mears, B. J., 1946b, “Badge Monitoring at Harshaw Chemical Company,” memorandum J. R. Hayes, 

Manhattan Engineer District, April 16.

MED (Manhattan Engineer District), 1945, “Supplemental Agreement No. 1 (to Contract No. W-26-021 

eng-4),” January 1.

MED (Manhattan Engineer District), 1949, “Uranium Metal Production,” memorandum from F. G. 

Stroke (Division of Technical Advisors) to Manhattan Engineer District Files; March 4.

Morgan, J. P., 1949, “Meeting with Harshaw Chemical Company on January 24, 1949,” memorandum 

to Files, U.S. Atomic Energy Commission, February 24.

NCRP (National Council on Radiation Protection and Measurements), 1989, Medical X-Ray, Electron 

Beam and Gamma-Ray Protection for Energies up to 50 MeV (Equipment Design, 

Performance and Use), Report 102, Bethesda, Maryland.

Neumann, A. W., 1952, “Harshaw Contract W-405 ENG-276 Feed and Production Schedule for the 

Quarter Beginning October 1, 1952,” letter to G. R. Fernelius (Harshaw Chemical Company), 

U.S. Atomic Energy Commission, August 1.

Neumann, A. W., 1953, “Termination of Period of Work Performance,” letter to G. R. Fernelius 

(Harshaw Chemical Company), U.S. Atomic Energy Commission, August 17.

NIOSH (National Institutes of Occupational Safety and Health), 2002, “Residual Radioactivity 

Evaluations for Individual Facilities,” Appendix A-2, Progress Report on Residual Radioactive 

and Beryllium Contamination at Atomic Weapons Employer Facilities and Beryllium Vendor 

Facilities, November.
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NIOSH (National Institutes of Occupational Safety and Health), 2004, Estimation of Ingestion Intakes, 

OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, April 

13.

NIOSH (National Institutes of Occupational Safety and Health), 2006, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 2, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August.

NRC (U.S. Nuclear Regulatory Commission), 2002, Re-evaluation of the Indoor Resuspension Factor 

for the Screening Analysis of the Building Occupancy Scenario for NRC’s License Termination 

Rule, NUREG-1720, Washington, D.C., June.

NYOO Radiological Laboratory, 1948, “Film Report, Week Ending 5/9” for Harshaw, New York, New 

York, undated. [SRDB Ref ID: 10705, p. 3 of 224]

NYOO Medical Division, 1949, Health Hazards in NYOO Facilities Producing and Processing 

Uranium, April 18. [SRDB Ref ID: 3931]

OEPA (Ohio Environmental Protection Agency), 2001, “Site Information, Harshaw Chemical 

Company,” information appears to have been written in calendar year 2001,

http://offo2.epa.state.oh.us/DOE/FUSRAP/Harshaw_Chemical.htm.

ORAU (Oak Ridge Associated Universities), 1983, “Interview with (Name Redacted), from NYOO‘s 

HASL, Concerning Radiation Monitoring Done at HASL,” memorandum from E. Dupree et al., 

Oak Ridge, Tennessee, March 25.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Procedures, ORAUT-OTIB-0006, Rev. 03 PC-1, Oak Ridge, 

Tennessee, December 21.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Technical Basis Document: Basis for 

Development of an Exposure Matrix for the Mallinckrodt Chemical Company, ORAUT-TKBS0005, Rev. 01, Oak Ridge, Tennessee, March 10.

ORAUT (Oak Ridge Associated Universities Team), 2005c, Estimation of Neutron Dose Rates from 

Alpha-Neutron Reactions in Uranium and Thorium Compounds, ORAUT-OTIB-0024, Rev. 0, 

Oak Ridge, Tennessee, April 7.

ORAUT (Oak Ridge Associated Universities Team), 2005d, Analysis of Coworker Bioassay Data for 

Internal Dose Assignment, ORAUT-OTIB-0019, Rev. 01, Oak Ridge, Tennessee, October 7.

ORAUT (Oak Ridge Associated Universities Team), 2006a, Site Profile for Allied Chemical 

Corporation Plant, Rev. 00, ORAUT-TKBS-0044, Oak Ridge, Tennessee, February 1.

ORAUT (Oak Ridge Associated Universities Team), 2006b, Generating Summary Statistics for 

Coworker Bioassay Data, Rev. 0, ORAUT-PROC-0095, Oak Ridge, Tennessee, June 5.

ORAUT (Oak Ridge Associated Universities Team), 2007, Harshaw Consolidated Urine Bioassay, 

Oak Ridge, Tennessee, February 20. [SRDB Ref ID: 30119]
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Parke, C. S., 1944, untitled letter to Lt. W. H. Dalton (U.S. Army Corps of Engineers), Harshaw 

Chemical Company, July 8.

Piccot, A. R., 1949, “Harshaw UX1 and UX2 Ash,” memorandum to Files, U.S. Atomic Energy 

Commission, June 13.

Pinkston, J. T. Jr., 1945, “C-616 Production,” memorandum to Major W. E. Kelley, U.S. Army Corps of 

Engineers, May 28.

Quigley, J. A., 1950, “Cases of Nephrosis Due to Exposure to Uranium,” memorandum to Health and 

Safety Division Files, U.S. Atomic Energy Commission, November 1.

Quigley, J. A., 1951a, “Visit to Harshaw Chemical Company, Cleveland, Ohio, on June 6, 1951,” 

memorandum to M. Eisenbud, U.S. Atomic Energy Commission, June 20.

Quigley, J. A., 1951b, untitled letter to E. R. Swanson (consultant to Harshaw Chemical Company), 

U.S. Atomic Energy Commission, July 5.

Rauch, D. D., 1948, “Survey of Operations of Area ’C‘ at the Harshaw Chemical Company,” 

memorandum to Files, U.S. Atomic Energy Commission, October 7.

Ray, W. H., 1947, “Survey of Harshaw Chemical Company Unit at Cleveland, Ohio, July 9-10, 1947,” 

memorandum report to Dr. B. S. Wolf, U.S. Atomic Energy Commission, July 12.

Russell, G. W., 1943, “Shipment of Materials from Mallinckrodt Chemical Works and from Harshaw 

Chemical Company,” letter to W. E. Kelley of the Knoxville Area Engineer, Manhattan 

Engineer District, October 19.

Sargent, E. C., 1948, “Visit to Harshaw -- Week of December 13, 1948,” memorandum to Files, U.S. 

Atomic Energy Commission, December 23.

Sargent, E. C., 1949, “Collection and Storage of Ash from Process Gas Reactors at Harshaw 

Chemical Co.,” letter to G. R. Fernelius (Harshaw Chemical Company), U.S. Atomic Energy 

Commission, September 12.

Sargent, E. C., 1950a, “Medical Meeting at Harshaw, October 4, 1950,” memorandum to J. P. 

Morgan, U.S. Atomic Energy Commission, Cleveland Area Office, October 5.

Sargent, E. C., 1950b, “Urinalysis Program,” letter to G. R. Fernelius (Harshaw Chemical Company), 

U.S. Atomic Energy Commission, January 19.

Sargent, E. C., 1950c, “Filtration of PB-13 Process Liquors,” memorandum to M. Eisenbud, U.S. 

Atomic Energy Commission, June 28.

Sargent, E. C., 1951, “Contract Proposal of Award -- Harshaw Chemical Company Contract W-7405-

ENG-276,” memorandum to W. E. Kelley (AEC New York Operations Office), U. S. Atomic 

Energy Commission, Cleveland Area Office, Cleveland, Ohio, December 29.

Schoen, A., 1958, Radioactive Contamination Survey of Uranium Refinery at Harshaw Chemical 

Company, revised, U.S. Atomic Energy Commission, May 19.
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Simmons, F. W., 1945, “Shipment Security Survey at Mallinckrodt Chemical Works,” memorandum to 

the Officer in Charge, Manhattan Engineer District, February 15.

Stefanec, A. J., 1951, “Health Physics Activities November 15 through December 15, 1951,” letter to 

P. B. Klevin (U.S. Atomic Energy Commission), Harshaw Chemical Company, December 21.

Tabershaw, I. R., 1951, “Contemplated Medical Program at Harshaw Chemical Company,” 

memorandum to M. Eisenbud, U.S. Atomic Energy Commission, November 21.

Termini, J. P., 1952, “Plant Visit: AEC Operations at Harshaw Chemical Company, Cleveland, Ohio,” 

memorandum to Files, U.S. Atomic Energy Commission, September 18.

Turner, R. A. N., 1947a, untitled letter to B. S. Wolf (U.S. Atomic Energy Commission), Harshaw 

Chemical Company, July 29.

Turner, R. A. N., 1947b, untitled letter to B. S. Wolf (U.S. Atomic Energy Commission), Harshaw 

Chemical Company, August 26.

Turner, R. A. N., 1947c, untitled letter to B. S. Wolf (U.S. Atomic Energy Commission), Harshaw 

Chemical Company, September 3.

Turner, R. A. N., 1947d, “Respirators,” letter to B. S. Wolf (U.S. Atomic Energy Commission), 

Harshaw Chemical Company, September 22.

Turner, R. A. N. Jr., 1947e, letter to Dr. B. S. Wolfe (U.S. Atomic Energy Commission with attachment 

“Radiation Readings Taken with Zues [Zeus] in Operating Areas,” Harshaw Chemical 

Company, September 5.

Tybout, R. A, 1944a, “Visit to Harshaw Chemical Company -- 21 September 1944,” memorandum to 

New York Area Engineer, Manhattan Engineer District, September 27.

Tybout, R. A, 1944b, “Visit to the Harshaw Chemical Company -- 16 November 1944,” memorandum 

to New York Area Engineer, Manhattan Engineer District, November 18.

Tybout, R. A., 1945a, “Visit to Harshaw Chemical Co. -- June 19, 1945,” memorandum to W. Bale, 

Manhattan Engineer District, June 22.

Tybout, R. A., 1945b, untitled letter to F. Becker (Harshaw Chemical Company), Manhattan Engineer 

District, March 1.

Tybout, R. A, 1946, “Recommendations for Residue Handling at Harshaw,” memorandum to B. J. 

Mears, Manhattan Engineer District, April 10.

University of Rochester, ca. 1948, Harshaw Radiation Summary, Aug 1944 to Mar 48. [PDN 

010002377]

Velton, E. H., 1949, “Record of Negotiations for Definitive Supplemental Agreement No. 32 to 

Contract No. W-7405 eng-276 -- Harshaw Chemical Company,” memorandum to J. C. Clarke 

(Contract Coordinator), U.S. Atomic Energy Commission.
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Watson, J. E. Jr., J. L. Wood, W. G. Tankersly, and C. M. West, 1994, “Estimation of Radiation Doses 

for Workers without Monitoring Data for Retrospective Epidemiological Studies,” Health 

Physics, volume 67, number 4, pp. 402–405.

Wilson, R. H., 1958, “The Hanford Uranium Bio-Assay Program,” in Symposium on Occupational 

Health Experience and Practices in the Uranium Industry, Proceedings of a United States 

Atomic Energy Commission Conference, HASL-58, U.S. Atomic Energy Commission, New 

York, New York, pp. 77–84.

Wolf, B. S., 1948a, “Harshaw Chemical Company,” memorandum to Files, U.S. Atomic Energy 

Commission, December 27.

Wolf, B. S., 1948b, “Protective Clothing at Harshaw Chemical Company,” memorandum to W. A. 

Taussig, U.S. Atomic Energy Commission, November 1.

Wolf, B. S., 1949 “Survey of Brown Plant -- Harshaw Chemical Co., Cleveland,” memorandum to F. 

M. Belmore, U.S. Atomic Energy Commission, May 27.
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GLOSSARY

air kerma

Air kerma means kerma in a given mass of air. Kerma means the sum of the initial energies of 

all the charged particles liberated by uncharged ionizing particles in a material of given mass. 

Kerma is closely related to the energy absorbed per unit mass (absorbed dose). See rad.

background radiation (also background or natural background)

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

beam quality

A measure of an X-ray beam’s ability to provide useful diagnostic information without 

unnecessary exposure to the patient. Usually expressed in half-value layers of aluminum.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

contamination, radioactive (also residual contamination)

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

daily weighted (average) exposure

The average concentration calculated by summing the products of the concentration 

measured by an air sampler and exposure time (in hours) for each period or task in a day and 

dividing by the total time per day (typically 8 hours).

decontamination

Reduction or removal of radioactive material from a structure, area, object, or person. 

Decontamination can occur through (1) treating the surface to remove or decrease the 

contamination or (2) allowing natural radioactive decay to occur over a period of time.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. As 

examples, spent (used) fuel elements, byproduct tails, residues from uranium isotope 

separation, and some weapons materials contain DU. DU can be blended with highly 

enriched uranium to make reactor fuel or used as a raw material to produce plutonium.
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dose

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual.

dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.

equilibrium factor

A measure of the degree of radioactive equilibrium between radon and its short-lived 

radioactive decay products. The equilibrium factor may be expressed as a fraction or a 

percent. For example, an equilibrium factor of 0.4 (40%) means that the concentration of 

short-lived radioactive decal products is 0.4 of the radon concentration.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

extremity

That portion of the arm extending from and including the elbow through the fingertips, and that

portion of the leg extending from and including the knee and patella through the tips of the

toes.

film

Radiation-sensitive photographic film in a light-tight wrapping.

fission 

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

fission product

(1) Radionuclides produced by fission or by the subsequent radioactive decay of 

radionuclides. (2) Fragments other than neutrons that result from the splitting of an atomic 

nucleus. 

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 
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Gamma photons are identical to X-ray photons of high energy; the difference is that X-rays do 

not originate in the nucleus. 

half-life

Time in which half of a given quantity of a particular radionuclide disintegrates (decays) into 

another nuclear form. During one half-life, the number of atoms of a particular radionuclide

decreases by one half. Each radionuclide has a unique half-life ranging from trillionths of a 

second to billions of years.

half-value layer

Thickness of a specified material (usually aluminum for X-rays) which reduces the exposure 

rate to one-half of its initial value.

irradiate

To expose to ionizing radiation.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties but often have different physical 

properties.

limit of detection (LOD)

The lowest quantity of radiation exposure or dose that can be distinguished from background. 

Measurements below the LOD are usually recorded as zero, but may be noted in some other 

way (for example, “< 50” or blank entries).

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 

microcuries per gram).

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

occupancy factor

The fraction (or percentage) of time that a given area is occupied by workers.

open window

Area on film dosimeter that implies the use of little (i.e., only security credential) shielding over 

the film. Commonly used to label the film response corresponding to the open-window area

on dose reports.

personal dose equivalent [Hp(d)]

Represents the dose equivalent in soft tissue below a specified point on the body at an
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appropriate depth d. The depths selected for personnel dosimetry are 0.07 mm and 10 mm for 

the skin and body, respectively. These are noted as Hp(0.07) and Hp(10), respectively.

photon

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons to high-energy gamma rays. Photons have 

energies between 10 and 100 kiloelectron-volts.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 

radiation

Subatomic particles and electromagnetic rays (photons) that travel from one point to another, 

some of which can pass through or partly through solid materials including the human body.

radioactive

Giving off ionizing radiation such as alpha particles or X-rays.

radioactivity

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei. See 

radionuclide.

radioactive waste

Radioactive solid, liquid, and gaseous materials for which there is no further use. Wastes are 

generally classified as high-level (with radioactivity as high as hundreds of thousands of curies 

per gallon or cubic foot), low-level (in the range of 1 microcurie per gallon or cubic foot), 

intermediate level (between these extremes), mixed (also contains hazardous waste), and 

transuranic. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

reactor 

A container or vessel in which a chemical reaction takes place. Not a nuclear reactor as it is 

used in this document.

recycled uranium

Uranium from spent nuclear fuel from Government reprocessing plants at the Hanford,

Savannah River, and Idaho sites and also at the commercial West Valley site. These plants 

recovered plutonium and uranium from spent nuclear fuel and target material irradiated in 

nuclear reactors. Recycled uranium contains trace amounts of fission products, activation 

products, and transuranic elements.

Redox material

Uranium recovered at Hanford using a solvent chemical extraction separation technique, the 

Reduction-Oxidation process.
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rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural.

rep

An early unit of absorbed radiation dose, which is the amount of energy from any type of 

ionizing radiation deposited in any medium. A dose of 1 rep is equivalent to the absorption of 

93 ergs per gram of absorbing tissue. It is approximately equal to 1 roentgen of 250 kVp Xradiation in soft tissue, or 0.93 rads, or 9.3 milligray. The rep was replaced by the rad.

reprocessing

Normally mechanical and chemical processing of spent nuclear fuel to separate useable 

fissionable products (i.e., uranium and plutonium) from waste material. At Harshaw this term 

applies to recycled uranium material (not spent fuel) received from Hanford for further 

purification.

roentgen (R)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

and negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0° Celsius and standard atmospheric pressure. 

An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher energy photons (generally greater than 100 kiloelectron-volts). 

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

skin dose

Dose equivalent at a depth of 0.007 cm in tissue.

specific activity

A measure of radioactivity per unit mass, such as pCi/g.

spent fuel

Fuel that has been in a reactor long enough to become ineffective because the proportion of 

fissile material has dropped below a certain level.

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The Energy Research and Development Administration 

and the U.S. Nuclear Regulatory Commission assumed separate duties from the AEC in 1974. 

The U.S. Department of Energy succeeded the Energy Research and Development 

Administration in 1979.

waste

See radioactive waste.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 62 of 114

working level (WL)

Any combination of short-lived radon decay products in one liter of air that will result in the

ultimate emission of 130,000 MeV of potential alpha energy. Approximately the total alpha 

energy released from the short-lived decay products in equilibrium with 100 pCi of Rn-222 per 

liter of air.

working level month (WLM)

A unit of exposure used to express the accumulated human exposure to radon decay 

products. 1 WLM = 1 WL exposure for 170 hours.

whole body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1,000 mg/cm2);

however, also used to refer to the recorded dose.

X-ray radiation

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside the nucleus of an atom.
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 63 of 114

TABLE OF CONTENTS

SECTION TITLE

A.1 Data Verification.....................................................................................................................

PAGE

64

A.2 Statistical Analysis ................................................................................................................. 65

A.3 Introduction ............................................................................................................................65

A.4 Intake Assessment................................................................................................................. 68

A.5 Intakes ...................................................................................................................................69

LIST OF TABLES

TABLE TITLE

A-1 Parameters for Harshaw data verification...............................................................................

PAGE

65

A-2 MIL-STD-105E 4% AQL sampling of Harshaw bioassay data files ......................................... 66

A-3 Uranium mass urinary excretion data..................................................................................... 68

A-4 Uranium activity urinary excretion data................................................................................... 69

A-5 Chronic intake rates for type F 234U ........................................................................................ 70

A-6 Chronic intake rates for type M 234U ....................................................................................... 70

A-7 Chronic intake rates for type S 234U........................................................................................ 70

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Harshaw Chemical Company identifying area designations during FUSRAP Site Survey ...... 11

A-1 Type F 50th percentile for 12/1/1949 to 12/31/1953 ............................................................... 70

A-2 Type M 50th percentile for 12/1/1949 to 12/31/1953 .............................................................. 71

A-3 Type S 50th percentile for 12/1/1949 to 3/31/1950................................................................. 71

A-4 Type S 50th percentile for 4/1/1950 to 12/31/1951................................................................. 72

A-5 Type S 50th percentile for 1/1/1952 to 12/31/1953................................................................. 72

A-6 Type F 84th percentile for 12/1/1949 to 12/31/1953 ............................................................... 73

A-7 Type M 84th percentile for 12/1/1949 to 12/31/1953 .............................................................. 73

A-8 Type S 84th percentile for 12/1/1949 to 3/31/1950................................................................. 74

A-9 Type S 84th percentile for 4/1/1950 to 12/31/1951................................................................. 74

A-10 Type S 84th percentile for 1/1/1952 to 12/31/1953................................................................. 75

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Due to the limited availability of bioassay data for use at the Harshaw site, it was necessary to 

conduct a coworker study of all the bioassay data for use in determining intake estimates. In short, 

the data used in this study were transcribed directly from hardcopy (or electronic copy) into 

worksheets. The data in the worksheets were verified (as indicated below), a statistical analysis 

conducted (and verified), and intake assessment conducted (and verified). Each of these processes 

is further described below. The resulting intake tables are provided in the intake assessment section.

A.1

The Harshaw bioassay data were verified as follows.

DATA VERIFICATION

1. Data were transcribed directly from source documents to spreadsheets by Data Entry 

personnel.

2. The transcribed data entered into eight spreadsheets were evaluated for acceptability using 

the statistical sample procedure of "Sampling by Attributes," which is based on DOE (1989). 

Spreadsheets were deemed acceptable when they passed a completeness and accuracy 

quality control review. These reviews were conducted by comparing the data on the individual 

source documents to the transcribed data.

3. The completeness quality control review was a review to ensure that the transcribed data 

reflected the total amount of data that was available for entry (i.e., the number of individual 

data items on a page of the source document was actually entered onto the spreadsheet for 

that page). If this was acceptable, the spreadsheet accurately reflected the amount of data 

present in the source document.

4. After the completeness quality control review indicated that the spreadsheet was acceptable, a 

quality control review for accuracy was performed. The accuracy quality control review 

ensured that the data entered into the spreadsheet accurately reflected the information from 

the source document. If this was acceptable and the completeness review was acceptable, 

there was reasonable assurance that the data contained in the spreadsheet accurately 

reflected the data from the source document.

5. After the data contained in individual spreadsheets were determined to accurately reflect the 

applicable individual source document information, a review to determine whether duplicate 

entries existed was conducted. The multiple spreadsheets were consolidated into a single 

spreadsheet that included all the data that were reviewed. Data contained in this consolidated 

spreadsheet were then sorted by sample date, last name, and first name. Once this was 

done, the data were reviewed to determine if duplicate data had been entered from source 

documents containing redundant information. As potential duplicate source document pages 

were identified, they were verified by viewing each potential duplicate page simultaneously 

(i.e., they were tiled to allow viewing of both pages). If it was determined that the pages were 

duplicates, they were evaluated for legibility, with the most legible data becoming the "original" 

and the poorer copy becoming the "duplicate." The consolidated spreadsheet was then sorted 

by file name (or Ref ID) and page number; and the information from the previously noted 

"duplicate" page was deleted.

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6. This process was repeated multiple times until each source document had been incorporated 

into the consolidated spreadsheet and no additional duplicate pages were noted.

Table A-1 identifies the parameters used to verify the Harshaw bioassay data in accordance with MILSTD-105E (DOD 1989); Table A-2 lists the details of the values for those parameters for the 4% 

sample.

Table A-1. Parameters for Harshaw data verification.

Values Used for Sampling by Attributes (MIL-STD-105Ea)

Parameter Definition

Value

Completeness review Accuracy review

Batch Size The batch size is the number of items in 

a lot or a batch.

Number of pages in 

source document

Number of lines of data 

entered into spreadsheets

AQL The maximal percent of nonconforming 

items (or the maximal number of 

nonconformities per 100 items), which 

is considered, for inspection purposes, 

as a satisfying process mean.

4%

(Set by Task 3)

4%

(Set by Task 3)

Inspection 

Levelb

The inspection level determines the 

relation between the batch size and 

sample size.

III III

a. DOD (1989).

b. Inspection Levels I, II, and III are general inspection levels:

• Level I requires about half the amount of inspection as level II, and is used when reduced sampling cost are 

required and a lower level of discrimination (or power) can be tolerated. 

• Level II is designated as Normal.

• Level III requires about twice the amount of inspection as level II, and is used when more discrimination (or 

power) is needed.

A.2

The verified data were analyzed according to the requirements within ORAUT-OTIB-0019, Analysis of 

Coworker Bioassay Data for Internal Dose Assignment (ORAUT 2005d) and ORAUT-PROC-0095,

Generating Summary Statistics for Coworker Bioassay Data (ORAUT 2006b). The data analysis 

report follows.

STATISTICAL ANALYSIS

A.3

This is a report on the validation check of the 1947 through 1953 summary statistics for Uranium 

Urine data collected from Harshaw. The summary statistics were compiled annually for 1947 through 

1949; the remaining years were compiled quarterly.

INTRODUCTION

For each period, the geometric mean, GSD, and number of samples were calculated independently 

by two individuals and separate spreadsheets created and results compared. The spreadsheets were 

created in accordance to the methodology in ORAUT (2005d) and ORAUT (2006b).

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Table A-2. MIL-STD-105E (DOD 1989) 4% AQL sampling of Harshaw bioassay data files.

Name of 

spreadsheet

Harshaw 

Chemical CoBio-1949-

1953-SAGJEmlBx15

HarshawUrin

eEarly1950s

HarshawBio-Per1950-51-

CSW-SATMB-KHUrineUrAnal

ysesPeople

Data

Harshaw 

ChemicalBioassay1950-SAEml2Bx29

Harshaw

Dust1948-

50

Harshaw 

Chemical CoBio 1950-GAJEmlBox4MedH

ealth&Safety3

_9-Gasior 

Corresp

Harshaw 

Chemical 

CoBioassayPer-47 to 48-

SA250000064

Harshaw 

Chemical CoBio-Med-Pers48&50-GAJEml2Box24Har

shawChemCo

PlantC_People

Data

HarshawChemic

al & 

MallinckrodtBio-Per-1952 to 

1953-cswEML2Box14Hars

hawChemUrine1

952to55

HARSHAW CHEMICAL 

CO-MiddlesexSylvania-NYOO-VitroMCW-Simonds-BioPer-Ext-1947-1950-

MDE-Eml2BX29

Harshaw Urine & 

Feces 1947-50

Total number of 

pages in source 

document

1,052 155 102 11 9 46 245 245 122

Minimum sample 

size 125 50 32 5 5 20 50 50 32

Total number of 

pages reviewed 

in source 

document during 

the completeness 

review

1,052 50 102 11 9 46 50 245 104

Maximum error 

rate to achieve 

4% AQL

8 3 2 0 0 1 3 3 2

Number of pages 

having incorrect 

number of data 

entry items listed 

for the page

0 0 0 0 0 0 0 0 3

Completeness 

review completed 

successfully?

Yes Yes Yes Yes Yes Yes Yes Yes Yes

Total number of 

lines of data 

entry in the 

spreadsheet

5,935 1,689 8 16 64 8 1,623 1,623 2,831

Minimum sample 

size 315 200 5 16 20 5 200 200 200

Total number of 

lines of data 

entry reviewed 

during the 

accuracy review

315 200 8 16 64 6 200 200 629

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Name of 

spreadsheet

Harshaw 

Chemical CoBio-1949-

1953-SAGJEmlBx15

HarshawUrin

eEarly1950s

HarshawBio-Per1950-51-

CSW-SATMB-KHUrineUrAnal

ysesPeople

Data

Harshaw 

ChemicalBioassay1950-SAEml2Bx29

Harshaw

Dust1948-

50

Harshaw 

Chemical CoBio 1950-GAJEmlBox4MedH

ealth&Safety3

_9-Gasior 

Corresp

Harshaw 

Chemical 

CoBioassayPer-47 to 48-

SA250000064

Harshaw 

Chemical CoBio-Med-Pers48&50-GAJEml2Box24Har

shawChemCo

PlantC_People

Data

HarshawChemic

al & 

MallinckrodtBio-Per-1952 to 

1953-cswEML2Box14Hars

hawChemUrine1

952to55

HARSHAW CHEMICAL 

CO-MiddlesexSylvania-NYOO-VitroMCW-Simonds-BioPer-Ext-1947-1950-

MDE-Eml2BX29

Harshaw Urine & 

Feces 1947-50

Maximum error 

rate to achieve 

4% AQL

18 12 0 1 1 0 12 12 12

Total number of 

lines of data 

entry that were 

found to have at 

least one error

11 2 0 0 0 0 14 2 4

Accuracy review 

completed 

successfully?

Yes Yes Yes Yes Yes Yes No Yes Yes

Reviewer RM EP EP EP EP EP HWJ BF SW

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Methods

Data were supplied in a spreadsheet (ORAUT 2007). All samples without a value for either “sample 

end date” or “sample conc” were excluded. In addition, one value in 1952 that included a “?” was 

excluded.

Independent spreadsheets were created by two individuals to compute the relevant statistics and 

compared.

Results

Periods, Effective Bioassay dates, and units appear to be correct. The comparison of the two 

spreadsheets resulted in a difference of within one-tenth of 1%.

A.4

A lognormal distribution was assumed for the urinary excretion data and the 50th- and 84th-percentile 

uranium (mass) excretion rates were calculated using the method prescribed in ORAUT (2005d) and 

ORAUT (2006b). These excretion rates are listed in Table A-3. The uranium mass excretion rates 

were converted to the uranium activity excretion rates in Table A-4 by applying the specific activity of 

natural uranium (0.68296 pCi/μg) and a urination rate of 1.4 L/d. Bioassay data collected over a 

specified period are analyzed to determine the 50th- and 84th-percentile excretion rates for that 

period. The effective bioassay dates are the midpoints of the periods and they are used in the 

Integrated Modules for Bioassay Analysis (IMBA) software to calculate the intake rates. Data 

collected for 1947 through November 1949 are not considered to be reliable (see Section 5.3.1.1) and 

were not used to determine the intake rates.

INTAKE ASSESSMENT

Table A-3. Uranium mass urinary excretion data 

(mg U/liter).

Period

Effective

bioassay 

date

50th

percentile

(mg/L)

84th

percentile

(mg/L)

1949 Dec 12/15/49 0.1990 0.7163

1950 Q1 2/15/50 0.1373 0.6173

1950 Q2 5/15/50 0.0948 0.4674

1950 Q3 8/15/50 0.0367 0.1221

1950 Q4 11/15/50 0.0406 0.1610

1951 Q1 2/15/51 0.0383 0.1830

1951 Q2 5/15/51 0.0367 0.1327

1951 Q3 8/15/51 0.0428 0.1630

1951 Q4 11/15/51 0.0348 0.1205

1952 Q1 2/15/52 0.0207 0.0403

1952 Q2 5/15/52 0.0124 0.0345

1952 Q3 8/15/52 0.0099 0.0244

1952 Q4 11/15/52 0.0142 0.0296

1953 Q1 2/15/53 0.0139 0.0358

1953 Q2 5/15/53 0.0079 0.0238

1953 Q3 8/15/53 0.0072 0.0185

1953 Q4 11/15/53 0.0058 0.0189

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A.5

All urinary excretion rates were modeled as normally distributed 24-hr urine samples with a uniform 

absolute error of 1 (which results in all results being weighed equally). The excretion data were 

modeled with IMBA Expert ORAU-Edition for multiple chronic intakes of type F, type M, or type S 

uranium. Plots of expected and observed urinary excretion from these fits are shown in Figures A-1 

through A-10. While it may be unlikely for all workers at Harshaw to be chronically exposed to 

uranium, it will approximate a series of acute intakes with unknown intake dates. Intakes were 

assumed to be via inhalation using a default breathing rate of 1.2 m3

/hr and a 5-μm activity median 

aerodynamic diameter particle-size distribution.

INTAKES

Because uranium has a very long half-life and because the type S material is retained in the body for 

long periods, excretion results are not independent. To avoid potential underestimation of intakes for 

people who worked for relatively short periods, each type S intake period was fit independently, using

only the bioassay results from that intake period. This will result in a best estimate of dose if the 

person works in only one period and can result in an overestimate if an individual works in multiple 

periods. 

Table A-4. Uranium activity urinary 

excretion data (pCi/d).

Period

Effective 

bioassay 

date

50th

percentile

(pCi/d)

84th

percentile

(pCi/d)

1949 Dec 12/15/49 190.29 684.86

1950 Q1 2/15/50 131.23 590.22

1950 Q2 5/15/50 90.67 446.92

1950 Q3 8/15/50 35.05 116.73

1950 Q4 11/15/50 38.78 153.92

1951 Q1 2/15/51 36.66 174.96

1951 Q2 5/15/51 35.13 126.90

1951 Q3 8/15/51 40.96 155.87

1951 Q4 11/15/51 33.25 115.19

1952 Q1 2/15/52 19.81 38.49

1952 Q2 5/15/52 11.81 33.00

1952 Q3 8/15/52 9.46 23.32

1952 Q4 11/15/52 13.57 28.28

1953 Q1 2/15/53 13.26 34.19

1953 Q2 5/15/53 7.59 22.73

1953 Q3 8/15/53 6.88 17.64

1953 Q4 11/15/53 5.58 18.07

The intake rates, GSDs, and periods in which they are applicable are listed in Table A-5 for type F 

uranium, Table A-6 for type M uranium, and Table A-7 for type S uranium. The fits to the data are 

shown in Figures A-1 to A-10. The natural uranium is assumed to be 100% 234U for the purpose of 

calculating dose [35].

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Table A-5. Chronic intake rates for type F 234U.

Start

date End date

50th

percentile

(pCi/d)

84th

percentile

(pCi/d) GSD

12/1/1949 3/31/1950 650.3 2,607 4.01

4/1/1950 12/31/1951 157 650.5 4.14

1/1/1952 12/31/1953 37.43 86.85 2.32

Table A-6. Chronic intake rates for type M 234U.

Start

date End date

50th

percentile

(pCi/d)

84th

percentile

(pCi/d) GSD

12/1/1949 3/31/1950 3,934 16,220 4.12

4/1/1950 12/31/1951 460.1 1,830 3.98

1/1/1952 12/31/1953 115.9 201.7 1.74

Table A-7. Chronic intake rates for type S 234U.

Start

date End date

50th

percentile

(pCi/d)

84th

percentile

(pCi/d) GSD

12/1/1949 3/31/1950 19,910 79,860 4.01

4/1/1950 12/31/1951 3,651 18,070 4.95

1/1/1952 12/31/1953 1,071 2,655 2.48

In most cases, doses to be assigned to individuals potentially exposed on a routine basis are

calculated from the 50th-percentile intake rates assuming the solubility type that results in the largest 

probability of causation [36].

Figure A-1. Type F 50th percentile for 12/1/1949 to 12/31/1953.

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Figure A-2. Type M 50th percentile for 12/1/1949 to 12/31/1953.

Figure A-3. Type S 50th percentile for 12/1/1949 to 3/31/1950.

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Figure A-4. Type S 50th percentile for 4/1/1950 to 12/31/1951.

Figure A-5. Type S 50th percentile for 1/1/1952 to 12/31/1953.

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Figure A-6. Type F 84th percentile for 12/1/1949 to 12/31/1953.

Figure A-7. Type M 84th percentile for 12/1/1949 to 12/31/1953.

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Figure A-8. Type S 84th percentile for 12/1/1949 to 3/31/1950. 

Figure A-9. Type S 84th percentile for 4/1/1950 to 12/31/1951.

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Figure A-10. Type S 84th percentile for 1/1/1952 to 12/31/1953.

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LIST OF TABLES

TABLE TITLE

B-1 Chronology of Harshaw site operations based on available reference material ......................

PAGE

77

B-2 Buildings known to have been used at the Harshaw site for uranium processing 

work .......................................................................................................................................79

B-3 Types and quantities of materials used and produced in Harshaw uranium 

processing .............................................................................................................................80

B-4 Functional and process keywords and codes......................................................................... 86

B-5 Measured dose rates ............................................................................................................. 88

B-6 Chest and hand beta doses from ash residue handling, as measured by films....................... 92

B-7 Weekly dose rates for various workers and areas .................................................................. 93

B-8 Annual neutron whole-body doses from the alpha-neutron reaction, various 

uranium forms........................................................................................................................93

B-9 Reserved ...............................................................................................................................94

B-10 Air concentrations in various areas in the green and hex plants............................................. 94

B-11 General area and breathing zone dust measurements in the green and hex 

plants (491 and 492) ..............................................................................................................95

B-12 General area and breathing zone dust measurements in the brown plant (493)..................... 97

B-13 Dust measurements in the green and brown plants................................................................ 99

B-14 Daily DWEs to airborne dust .................................................................................................. 99

B-15 Daily DWEs to airborne dust for various job titles................................................................. 100

B-16 Annual inhalation intakes based on daily weighted average exposures to airborne 

dust for various job titles....................................................................................................... 102

B-17 Annual ingestion intakes based on daily weighted average exposures to airborne 

dust for various job titles....................................................................................................... 103

B-18 Annual radon exposure ........................................................................................................ 104

B-19 General area dust concentrations in the locker rooms and lunchrooms................................ 104

B-20 Average measured dust and urine concentrations by month................................................ 105

B-21 Number of workers............................................................................................................... 105

B-22 Job titles, functions, and appropriate absorption types ......................................................... 106

B-23 Results of a survey of the Harshaw site by Argonne National Laboratory, 1976 to 

1979..................................................................................................................................... 108

B-24 Source terms used to calculate inhalation and radon doses, D&D/postoperations 

period................................................................................................................................... 109

B-25 Annual inhalation, radon, and ingestion doses, D&D/postoperations period......................... 109

B-26 Annual external dose rates and doses, D&D/postoperations period..................................... 113

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Table B-1. Chronology of Harshaw site operations based on available reference material. Event Form Date Brief description Level Amount

Columbia University asks Harshaw to make 

UF6 using F2

UF6 Oct 41 UF6 production request Laboratory

National Bureau of Standards orders 

experimental quantities of UF6

UF6 Dec 41 UF6 production request Laboratory

Production of UCl4 begins UCl4 Jan 42 UCl4 production begins Laboratory

Production of UF6 begins UF6 Feb 42 UF6 production begins Laboratory 5 lb/d (1942 only)

First shipment of UCl4 UCl4 Mar 42 First shipment of UCl4 Laboratory

UF4 production (dry process, for Standard Oil) UF4 11 Mar 1942 UF4 production begins Laboratory Batches: tens of lb

First shipment of UF6 to Columbia Univ (later

to Naval Research Lab., Nat’l Bureau of Stds)

UF6 18 Mar 1942 First shipment of UF6 Laboratory

First shipment of UF4, to Westinghouse Elec. UF4 23 Mar 1942 First shipment of UF4 Laboratory 20 lb

Westinghouse asks Harshaw to make UF4; 

sends special oxide as feed

UF4 14 Apr 1942; 

May 1942

Increased UF4 production Laboratory About 10 lb

Second shipment of UF4, to Westinghouse, UF4 11 May 1942 Second shipment of UF4 Laboratory About 10 lb

Third shipment of UF4, to Princeton University UF4 8 Jun 1942 Third shipment of UF4 Laboratory

MED directs Harshaw to produce UF4. Smallscale production in laboratory, then plant builtUF4 11 Jul 1942 Significant UF4 production 

increase

Laboratory Tens of lb/d, then 

to 1,200 lb/d

Making of UF4 in production-level plant begins UF4 22 Sep 1942 Sustained UF4 production Production 1,200–1,800 lb/d

Laboratory production of UCl4 for University of 

California, with production rate increasing

UCl4 Nov 1942 Sustained UCl4

production begins

Laboratory Up to 100 lb/d

Harshaw constructs pilot plant to produce UF6 UF6 Mar 1943 Sustained UF6 production Pilot plant 50 lb/d

Peak UCl4 production for U. of California UCl4 Apr 1943 Peak UCl4 production Laboratory 100 lb/d

Harshaw manufactures special materials, 

including uranium oxyfluoride

UO2F2; 

Misc

1943-1944 Special material

production

Laboratory

MED and Harshaw agree to build UF6 plant to 

produce 5,000 lb/d

UF6 Jan 1944 UF6 plant planned Production

Harshaw closes the UF6 pilot plant UF6 Feb 1944 Total of 9,000 lb

Two units of new UF6 plant go into operation UF6 Apr 1944 Volume production of UF6 Production 400 lb/d

Harshaw begins work on UF6 analysis, 

recovery (improvements applied in new plant)

UF6 Spring 1944 UF6 yield and recovery 

investigations begin

Laboratory

Building G-1 completed Jul 1944 Building G-1 operation Production

Full-scale production achieved in new UF6

plant (six sets of reactors in Building G-1)

UF6 Jul 1944 Significant UF6 production 

increase

Production Believed to be 

2,000 lb/d

Harshaw begins research on recovery of UF6

from spent carbon trap residues

UF6 Aug 1944 Another UF6 recovery 

investigation begins

Film badging starts Aug 1944 Film badging

New UF6 reactor design plan UF6 Sep 1944 UF6 production changes Production

Harshaw agrees to set up new plant for UCl4 UCl4 15 Oct 1944 UCl4 production plant 1 ton/d or more

Shipment of first ton of UCL4 from UCl4 plant UCl4 30 Oct 1944 UCl4 production

Expansion in new UF6 plant (12 sets reactors) UF6 Nov 1944 UF6 production increase Production Up to 4,600 lb/d

UF4 production moves to Bldg G-1, increases UF4 Dec 1944 UF4 production increase Production 3,000 lb/d

MED orders additional UCl4 by mid-Feb 1945 UCl4 1 Jan 1945 UCl4 production extension Production 65,000 lb (add’l)

UCl4 production work completed in February 

1945, equipment dismantled

UCl4 Feb 1945 UCl4 production ends Production

UF6 production increase UF6 1 Apr 1945 UF6 production increase Production 4,500 lb/d

MED orders shipment to Harshaw of slightly 

enriched UF6, to be mixed with normal UF6

LEUF6 28 May 1945 Blending of normal and 

LEUF6

Limited 

basis

126 lb (after 

blending)

Expansion of new UF6 plant (14 sets reactors) UF6 Jun 1945 UF6 production increase Production 4,600 lb/d

MED orders shipment to Harshaw of slightly 

enriched UF6, to be mixed with normal UF6

LEUF6 17 Jan 1946 Blending of normal and 

LEUF6

Production

MED authorizes UF4 production increase UF4 4 Feb 1946 UF4 production increase 15,000 lb/wk

MED authorizes UF6 production increase UF6 1 Jun 1946 UF6 production increase 4,200 lb/d

Fume recovery system planned for UF6 plant UF6 Jan 1947 Pollution concerns

Central waste disposal system completed Oct 1947

UF4 production expansion UF4 4th Q 1947? UF4 production increase Production 5,600 lb/d

UF6 production expansion UF6 4th Q 1947? UF6 production increase Production 6,300 lb/d

Construction of brown plant Ore-UO2 1948–1949 Construction

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Event Form Date Brief description Level Amount

New laundry begins operation Misc Jan 1949? Improved laundering

The new brown plant begins operation Ore-UO2 May 1949 Brown plant operation Production

AEC air pollution survey in, around Harshaw 

site (measuring radioactivity, fluoride levels)

Jun 1949 Air pollution survey

Continuous green salt reactor use begins UF4 Jan 1950? Continuous UF4 reactor

Central loading station use begins in UF6 area UF6 Jun 1950 Mechanized loading: UF4 Production

Central vacuum system use in the UF6 area UF6 2nd half 1950 Central vacuum system Production

Dust control improvements, ore-to-UO2 area Ore-UO2 Early 1951? Dust control measures

UF4 production shut down UF4 17 Oct 1951 UF4 production ends Shutdown

UO2 production shut down; only UO3

produced in former brown plant

UO2, UO3 4th Q 1951 UO2 production ends Shutdown

AEC contamination/radiation survey of UF4

and UF6 areas (prior to decontamination)

UF4, UF6 Nov 1951 AEC survey D&D

UF6 production shut down, placed on standby UF6 Dec 1951 UF6 production ends Shutdown

Decontamination of UF4 and UF6 areas begins UF4, UF6 After 6 Dec 1951 D&D: UF4, UF6 areas D&D

Equipment dismantling in UF6 area begins UF6 Jan 1952 Equipment dismantling D&D

Mezzanine area of annex (1945 hex bldg) torn 

out, forcing relocation of offices and shops

Jan/Feb 1952? Modification

Final shipment of hex ash to Vitro Apr 1952 Hex ash shipment ends

Dust control improvements in UO3 production Ore-UO3 Jun-Jul 1952 Dust control measures

Hanford ships UNH to Brush Beryllium for 

storage, transfer to Harshaw to make UO3

UNH Mid-1952? UNH processed to UO3 Production

Change in feed materials (black oxide; also 

soda salts, Hanford UNH)

Ore-UO3 1 Oct 1952 Change in feed materials Production 200,000 lb/mo

Capacity expansion in UO3 production plant UO3 Oct 1952 Capacity expansion Production 600,000 lb/mo

Start of processing of Hanford recycled UO3 Recycle U 4th Q 1952 Recycle U processing Production 1,700 tons total

UO3 production shut down, placed on standby UO3 Aug 1953 UO3 production ends Shutdown

AEC contamination/radiation survey, UF6 area UF6 22-23 Oct 1953 AEC survey D&D

Harshaw finishes processing recycled U Recycle U 4th Q 1953 Recycle U processing end Special

AEC check/survey of D&D progress 6 May 1954 D&D

AEC check/survey of D&D progress 22 Jul 1954 D&D

Virtual completion of UF6 plant dismantlement Jul 1954 D&D

AEC contam/radiation survey: UF6, UF4 areas 22-23 Dec 1954 AEC survey D&D

AEC contam/radiation survey: UF6, UF4 areas 31 May – 1 Jun 

1955

AEC survey D&D

AEC directs Harshaw to end processing, 

convert leftover materials to UO3

17 Aug 1955 Completion of processing Production

Contract for removal of AEC equipment ends 30 Sep 1955 Equipment removal

AEC-Harshaw contamination/radiation survey 

to track decontamination progress

14 Nov 1956 AEC-Harshaw D&D 

survey

D&D

AEC-Washington orders AEC area office to 

dismantle Harshaw refinery

17 Dec 1956 Dismantlement of refinery 

– equipment only

D&D

Harshaw license termination (D&D prep) 31 Dec 1956 License termination D&D

AEC contam/radiation survey, ore-UO2 area 21 Nov 1957 AEC survey D&D

Decontamination by Harshaw, AEC oversight 1955(?)-1960 Decontamination D&D

Final AEC contamination/radiation survey 15 Apr 1959 AEC survey D&D

Site released from AEC control 1960 Site release Release

Argonne Nat'l Lab survey of site for DOE 12-20 May 1976 DOE-sponsored survey D&D

Engelhard Corp decontaminates some 

buildings, demolishes others

1990–1997 Further decontamination D&D

NRC releases 3 bldgs for unrestricted use; 

Plant C stays shuttered; Boiler House used to 

store rad materials from demolition

20 Mar 1998 Final disposition of site Release

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Table B-2. Buildings known to have been used at the Harshaw site for uranium processing work.

Building Room or area Use

G1 (Plant C) All levels, partial floors 493: Brown oxide (ore-to-UO2) production

All levels, partial floors 493: Orange oxide (ore-to-UO3) production

1st floor 492: Hex (UF6) production, recovery

2nd floor 491: Green salt (UF4) production

Annex All areas 492: Hex (UF6) production, storage of contaminated equipment and 

black oxide

K1 Some analytical work

Foundry Early processing work

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 4 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 80 of 114

Table B-3. Types and quantities of materials used and produced in Harshaw uranium processing.

a

Material Process or operation Content and form notes Amount

ORES AND OTHER FEEDS

All ores and feeds It does not appear that uranium bearing ores were 

processed at Harshaw, but rather U3O8 was received 

after milling (DOE 2000).

Mar 1953: 1 ton of "contained uranium" of 

each to be shipped to Harshaw: Colorado 

soda salt, Colorado black oxide, Canadian 

black oxide, South African black oxide, 

"Congo precipitate" (Belmore 1953).

Domestic ore and 

tailings

It does not appear that uranium bearing ores were 

processed at Harshaw, but rather U3O8 was received 

after milling (DOE 2000).

Colorado ores (e.g., Uravan, Durango, Grand Junction, and 

Naturita) were carnotite type (Eisenbud 1975); North 

American ore contained less than 1% U3O8.

U3O8 (milled ore or 

black oxide)

Ore arrived at Harshaw in milled or concentrated form, 

as black oxide.

1952: Canadian black oxide at 96% U3O8, Colorado black 

oxide at 70-85% U3O8 (Termini 1952).

4Q 1952: feed included 159 tons Colorado 

black oxide (Neumann 1952).

Sodium diuranate

(soda salt)

Packed in fiber containers (MED 1945). 1952: Vitro soda salt at 75% U3O8 (Termini 1952). Some 

possibly from Anaconda and Durango. 1952: Colorado soda 

salt at 60-72% U3O8 (Termini 1952). 1952: "recycle soda salt" 

also used (Termini 1952). Fiber containers weighed about 75 

lb each when full.

4Q 1952: feed included 43 tons Vitro soda 

salt, 53 tons Colorado soda salt; and 45 

tons Eldorado soda salt (Neumann 1952).

UNH 1952: Hanford transferred UNH to Harshaw via the 

Brush Beryllium Company in Luckey, Oregon, for 

conversion to UO3 (DOE 2000).

Shipped by rail in tanker cars (DOE 2000), at least once via 

the Brush Beryllium Company (DOE 2000). 1952: "recycle 

NG liquor" used (Termini 1952), presumably from Hanford.

1,700 tons total.

REFINING PRODUCTS

UO3

(orange oxide)

Feed digested in nitric acid to convert to nitrate form; 

precipitation of Ra-Pb with sulfuric acid (pitchblende 

ores only); filtration to remove the acid-insolubles; 

sulfate removal with Ba salt addition; centrifuging of 

solution to remove solids; boiling of "liquor"; double 

extraction of U with diethyl ether; purification; water 

wash to remove uranyl nitrate from ether; dewatering in 

Sperry press; boiling of the molten salt to "hex liquor" 

(uranyl nitrate hexahydrate); decomposition in gas-fired 

pots to form UO3; UO3 scooped or "gulped" out of pot 

using vacuum system, packed in fiber containers for 

shipment.

Digestion took 4-8 hr. Solid and liquid wastes, including most 

residues below. Full 2.5-gal fiber containers weighed 75 lb. 

Typical, 1952 (Termini 1952): 4,000-5,000 lb of Canadian and 

Colorado black oxides yielded 900 gal of digested slurry at 

400-500 g U/L; pumped to holding tank and mixed with 600 

gal dilute NG liquor, more acid, recycle soda salt, and Vitro 

soda salt, to produce new slurry at 300 g U/L; after extraction, 

before water wash, the organic liquid saturated to 80-90% 

uranium; after water wash, preboildown OK liquor at 50-75 g 

U/L. UO3 shipped to K-25 in 16-drum lots (800 lb/drum 

gross), 2 lots per shipment; sample taken during drum filling 

was supplied for each lot; 2 lot samples composited, analyzed 

for U content and isotopic ratio at K-25 (BJC and Haselwood 

Enterprises 2000). 

1951: 100 tons/mo capacity (AEC 1951c).

3Q 1952: 55 tons/mo to be sent to K-25 

and about 23 tons/wk to Mallinckrodt 

(Fernelius 1952). 4Q 1952: 200,000 

lb/mo of UO3 to be produced (Neumann 

1952); Sep 1952: 200,000 lb/mo based on 

operation 24 hr/d, 7 d/wk (Termini 1952).

UO2 (brown oxide) 

(received)

Small quantities of UO3 not meeting K-25 acceptance 

criteria for purity were sent to Harshaw for purification 

(DOE 2000).

2.5-gal fiber containers weighed about 75 lb each when full 

(Burman 1949). Shipped by rail from Mallinckrodt (AEC 

1949b). Had to be 97% or more free UO2 (the rest was 

impurities) (HCC 1946).

Mallinckrodt produced 2/3 of US total; 

20% went to Harshaw (MED 1949). In 

1944-1945, 10,000 lb/wk to Harshaw; in 

Sept-Oct 1944; 28,000 lb in Nov; and 

13,000 lb/wk after that (Simmons 1945).

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 81 of 114

Material Process or operation Content and form notes Amount

REFINING PRODUCTS (Cont’d.)

UO2

(brown oxide) 

(produced)

UO3 was transferred from fiber containers onto monel 

trays; reduced with cracked ammonia in batch electric 

(muffle) furnace to form UO2 (AEC 1949b); scooped 

from trays into fiber containers for transfer elsewhere.

In June 1949, 300 lb/drum (Mayer and Proschen 1949). In Oct 1949, 56 tons produced (Kelley 

1949b), design capacity 54 tons/mo (AEC 

1949b, Kelley 1949b). Jul -Dec 1949: avg 

42.5 tons/mo; 1950: avg 53 tons/mo; JanMar 1951: avg 71 tons/mo (AEC 1951c). 

1951, after process mods (e.g., using

TBP), capacity up to 100 tons/mo (AEC 

1951c).

UF4 (green salt) 

(received)

Fiber containers weighed about 75 lb each when full. To 1949, some UF4 from ElectroMet (DOE 

1997) (probably to Harshaw). In 1945, 

8,000 lb/wk were being sent to Harshaw 

from Mallinckrodt (Simmons 1945). 

UF4

(green salt) 

(produced)

UO2 placed on nickel trays (Rauch 1948; HCC 1946), 

loaded into reactor tubes, and placed in 

hydrofluorination reactor (furnace); HF gas passed 

over it to form UF4; UF4 removed from furnace and put 

through pulverizer; UF4 weighed (Mayer and Proschen 

1949) and packed into fiber containers (Simmons 

1945) or 5-gal containers for transfer to another site 

(AEC 1949b). Excess anhydrous HF gas was bled off 

the reactors and sent back to the recovery system 

(Mayer and Proschen 1949).

Four trays per reactor; 9 lb in the top and bottom and 13 in 

each middle tray, for a total charge of 44 lb/reactor (Mayer 

and Proschen 1949; Klevin 1955b; HCC 1946). Oct 1946: 

one bank of 37 reactors, for a total (fresh) charge of 1,893 lb 

per 14-hr cycle; however, reruns of incompletely reacted 

material amounted to 7-10% of the material made (HCC 

1946). UF4 had to be minimum of 97.5% UF4, with less than 

0.3% UO2 and 1.2% UO2F2 (HCC 1946). Unreacted or partly 

reacted material was recharged into trays at 75 lb/tube and 

rerun through the normal 14-hr cycle (HCC 1946). Oct 1948: 

two banks of 37 reactors each operated on 12-hr cycles 

(Rauch 1948). 51 lb UF4 produced per 44 lb UO2 charged 

(HCC 1946). Filled drums contained 200 lb each (Burman 

1949). Gas sent to the recovery system was not analyzed 

(Mayer and Proschen 1949). On a metal basis, from 100% 

charged as UO2, about 0.4% was recovered as scrap and 

0.57% was unaccounted for (HCC 1946). Sent to K-25? 

(DOE 1997). Fiber containers: 75 lb each, full (Simmons 

1945).

Sep 1942: 1,800 lb/d (HCC ca. 1945), avg 

25 tons/mo (AEC 1951c). Dec 1944: 

3,000 lb/d (HCC ca. 1945). Feb 1946: 

18,000 lb/wk for 1Q 1946, then 15,000 

lb/wk (Eisenbud 1949b). Oct 1946: 2 d’ 

inventory normally in process (Gates 

1946), theor. capacity 3,235 lb/d but 

actual production 3,000 lb/d due to reruns 

(HCC 1946). Jul 1944 - Dec 1946: avg 40 

tons/mo (AEC 1951c). 1947: avg 84 

tons/mo; 1948: 88.5; 1949: 76; 1950: 80; 

and Jan-Mar 1951 (projected): 86 (AEC 

1951c). Oct 1948: theor. capacity 7,600 

lb/d (Rauch 1948). Dec 1947, Jan 1948,

Feb 1948: 20.2, 20.0, and (projected) 20.5 

tons produced. respectively (Chrestia 

1948). 4Q 1949, 5,500 lb/d UF4 to be 

produced (Hunter 1949b).

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Material Process or operation Content and form notes Amount

REFINING PRODUCTS (Cont’d.)

UF6 (hex) UF4 placed on nickel trays in sealed boxes; fluorine 

passed over at an elevated temperature; UF6 gas 

condensed in a series of three receivers, with any 

uncondensed gas caught on traps (Gates 1946;

HCC 1946). Completeness of reaction was judged 

by receiver weight, at which time the third reactor 

per unit was reloaded (Rauch 1948). The third 

receiver normally was not filled in each cycle and 

was not removed until full, so was not removed 

every cycle (HCC 1946).

Jul 1944 (HCC ca. 1945): 18 reactors. Nov 1944 (HCC ca. 

1945): 36 reactors. June 1945: 36 reactors (Tybout 1945a) or 42 

reactors (HCC ca. 1945). Oct 1946: each set of reactors 

produced 340 lb/d of crude UF6; there were 14 units (sets of 3 

reactors) in 7 hoods, 14 pairs of fluorine cells, and a 24-hr cycle 

(HCC 1946). Oct 1948: 96 reactors in 16 hoods and 52 fluorine 

cells; three reactors in series formed a unit, with two fluorine 

cells for each unit; 18 reactors without dedicated fluorine cells 

drew on excess fluorine capacity (Rauch 1948). Two reactors

per regular unit operated on 24-hr cycles, the third on 2- to 3-d

cycles; the 18 extra reactors operated on 3- to 4-d cycles (Rauch 

1948). Each reactor had two trays, each charged with 90 (HCC 

1946) or 100 (Rauch 1948) lb UF4; receivers held up to 435 lb of 

condensed UF6. Theoretical output was based on a 100-lb/d

charge, 100% efficiency, 78 reactors (2/3 on a daily cycle, 1/3 on 

3-d cycle), plus 18 (4-d cycle) (Rauch 1948). 1945: UF4 feed 

rate was 4,200 lb/d (HCC ca. 1945).

Apr 1944: 400 lb/d UF6 (HCC ca. 1945). 

Sept 1944: new semicontinuous process 

reactor to produce 800 lb/d UF6 (per set) 

(Tybout 1944a), 3,300 lb/d total (AEC 

1951c). Nov 1944: 4,500 lb/d (AEC 

1951c). Nov 1944 (HCC ca. 1945), Jun 

1945 (Tybout 1945a): capacity 4,000 lb/d.

Feb 1946: 4,200 lb/d to Jun 1946, then 

4,500 lb/d (Eisenbud 1949b). Oct 1946: 

theor. capacity 4,600-4,700 lb/d or about 

14.5 lb/hr of crude UF6, actual production 

3,000 lb UF4 and 4,600 lb UF6 per day 

(HCC 1946). Dec 1946: 6,300 lb/d; Nov 

1947: 9,000 lb/d (AEC 1951c). By Mar 

1948: expansion to produce 5,600 lb/d or 

121,000 lb/mo (Chrestia 1948). Oct 1948: 

theor. capacity 14,600 lb/d (assumes total 

UF4 input of 13,000 lb/d, more than 

Harshaw itself produced) (Rauch 1948). 

1950: production cut back to 6,300 lb/d

(AEC 1951c).

Distilled UF6 (hex) Crude UF6 purified by distillation: most 

noncondensible gases were drawn off via floor 

bleeders (HCC 1946); the HF, the rest of the 

condensables, and some UF6 were drawn off via 

scale bleeders (HCC 1946); the rest of the UF6 was 

distilled into nickel or steel cylinders for shipment 

(Gates 1946).

1946: 5% by weight of crude UF6 was bled off to a scale bleeder 

(HCC 1946). The rest was distilled into nickel or steel shipping 

containers, with nickel container capacity of 462 lb UF6 (HCC 

1946). Sample cylinders held 5 lb UF6, with ~4.5 lb left after 7-oz 

sample draw (HCC 1946). On a metal basis, from 100% of U 

charged as UF4, 0.69% was recovered as scrap metal and 

1.82% unaccounted for. Cylinders sent to K-25 (Gates 1946;

AEC 1949b) by government truck (AEC 1949b). Oct 1943: 5 lb 

UF6 to be shipped to Oak Ridge (Clinton Engineer Works) 

(Russell 1943). Destination of this UF6 was the S-50 project 

(1,050 lb/d, May - Nov 1945, total of 500,000 lb by mid-Oct 

1945) (Hearon 1945).

May 1945: capacity 4,600 lb/d (Pinkston 

1945). Normal in-process inventory, Oct 

1946: 1 week's production (Gates 1946). 

Dec 1947, Jan 1948, Feb 1948 (Chrestia 

1948): 45,800, 46,400, and 42,500 lb 

produced, respectively. By Mar 1948: 

expansion to 6,300 lb/d, 136,000 lb/mo 

(Chrestia 1948). Jun 1948: 7,600 lb/d

produced (Kelley 1948). Oct 1948: theor. 

capacity 14,600 lb/d (required UF4 input 

beyond what Harshaw produced) (Rauch 

1948). Apr 1949: 90 tons/mo to K-25 

(AEC 1949b). 4Q 1949, 6,000 lb/d UF6

planned (Hunter 1949b).

UCl4 Assumed production 7 d/wk (Belmore 1953; AEC 1949b). For 

Jan 1945 contract extension, feed materials to be "Product 306, 

Product B, Product G, and other similar materials" (MED 1945).

Oct 1944: new plant capacity of 2,000 lb/d

by Nov 1944, 80,000 lb to be supplied by 

1 Jan 1945 (Belmore 1953; AEC 1949b). 

Jan 1945: 65,000 lb by 15 Feb 1945 

(MED 1945).

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 83 of 114

Material Process or operation Content and form notes Amount

SPECIAL PROCESS MATERIALS AND SAMPLING

Slightly enriched 

UF6

Shipped from Hanford (Kelley 1946). After blending in May 1945: .3% enriched (Kelley 1946). In May 1945, 126 lb shipped from 

Harshaw after blending (Kelley 1946).

RU Shipped from Hanford (BJC and Haselwood 

Enterprises 2000).

Oct 1952-Jun 1953: RU from Harshaw to K-25, ~1,403,000 

kg UO3 at 0.666%; Jul 1953-Dec 1953, ~300,000 kg at

0.671%; in this total of ~1,702,000 kg, there were 5.96 g Pu, 

1328 g Np, and 11,900 g Tc (BJC and Haselwood Enterprises

2000).

Sampling 

Harshaw-produced 

UO3

20-qt jar of UO3 unloaded into grinder; ground UO3

loaded into 2-qt jar and put into blender, which was 

then collected in 2-qt jars.

1953: tests on Harshaw UO3 were performed daily (AEC 

1953b).

Sampling of 

Hanford-produced 

UO3

In lab, UO2 scooped from cans into blenders; blended; 

scoop-unloaded into grinder; ground; and collected in 

2-quart cans. Shotgun tests: samples of ground UO3

weighed out; loaded into a die; pressed in hydraulic 

press; and used in the assay process (AEC 1953b).

1953: tests on Hanford UO3 were performed about 3 times 

every 2 wk (AEC 1953b). Shotgun samples consisted of ten 

45-g aliquots, apparently per 2-quart jars (AEC 1953b).

RESIDUES, OTHER WASTES, AND RECOVERED MATERIALS

UF4-to-UF6

fluorination ash

Vacuum-conveyed from fluorination trays to a bag dust 

collector (AEC 1949b; Rauch 1948) on the roof (Piccot 

1949), where it was removed by turning a crank that 

allowed the ash to fall through a star valve into a drum.

The drum was filled about every 2 wk, at which time it 

was removed, lidded, and conveyed by a truck to the 

first floor (Piccot 1949). It was then stored temporarily 

for decay (Mayer and Proschen 1949; HCC 1946); 

from 1946 or before, until at least 1949, the storage 

time was 6 mo (HCC 1946; Eisenbud 1949f).

Ash: 0.1% of the original mass of the uranium, but practically 

all of the UX1-UX2; beta activity of ~1 mCi/g (AEC 1949b). 

May 1948: 100 g of ash had volume of 100 cm3

; spread out in 

flat dish 10 cm in diameter, gave a reading corresponding to 

15 mCi. Sep 1949: ash being collected (and probably shipped 

in) 15-gal drums (Sargent 1949; Piccot 1949) lined with 1/16

in. or 3/32 in. of lead on top, bottom, and sides (Piccot 1949). 

April 1949 (AEC 1949b): shipped to Oak Ridge or Lake 

Ontario Ordnance Works for storage; (Mayer and Proschen 

1949, Jun 1949): being weighed and sent to Vitro. (Hunter 

1949c, Nov 1949): 2,250 lb of Harshaw ash in storage at Oak 

Ridge, with newest well over 1 yr old (i.e., no longer being 

sent to Oak Ridge); this ash supposedly contained 10-13% 

uranium. Jun 1949 (Piccot 1949): ash was about 1 g/cc and 

contained about 18% metal.

In Jul 1944, about 5 lb/d of ash produced 

(Parke 1944); in Apr 1949, 2,500 g/d (AEC 

1949b). In Apr 1949, 2,000 gal/d

apparently shipped for storage (AEC 

1949b).

UO2 loading dust Vacuum-conveyed from fluorination trays to a bag dust 

collector (AEC 1949b; Rauch 1948).

Ash from 

miscellaneous 

combustible waste

Produced from burning UO2 containers, floor 

sweepings ("black ash"), and scrap from the recovery 

room (Rauch 1948).

Shipped to Vitro (Mayer and Proschen 

1949).

Floor drainage Floor drainage sent through filter press, producing a 

filtrate and a press cake (AEC 1949b).

Press cake shipped to Vitro for processing (AEC 1949b); 

filtrate sent to the Cuyahoga River (AEC 1949b).

Filtrate: about 150 gal/d in April 1949 

(AEC 1949b).

UF6 condenser 

water

Radionuclide concentration normally none (AEC 1949b).

Condenser 

(receiver) tail gas

Passed through a dry ice trap to remove residual UF6

(AEC 1949b).

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 84 of 114

Material Process or operation Content and form notes Amount

RESIDUES, OTHER WASTES, AND RECOVERED MATERIALS (Cont’d.)

Floor spillage UF4 mostly washed to the sewer in 1945 (Anderson 

1945).

Some washed out to Cuyahoga River, some lodged in sewer 

(Anderson 1945).

Up to 40 lb/d of UF4 (Anderson 1945).

Scale and floor 

bleeds

Crude UF6 in the receivers, UF6 on the associated dry 

ice traps: "bled" to remove volatile gases. The former 

bleed sent via floor and scale bleeders (HCC 1946) to 

large turbosaturator on the roof and the latter to the 

small one (Rauch 1948). Floor receivers in distillation 

unit and dry ice trap receivers bled directly to large 

turbosaturator, bypassing scale and floor bleeders and 

the dry ice trap (Hunter 1949a). When full, floor and 

scale bleeders were processed like receivers, 

minimizing loss (HCC 1946).

The large turbosaturator and tower handled the fumes from 

the small vent hoods over each individual reactor, the dry ice 

trap for the distillation unit, and the bleed from the floor 

receivers in the distillation unit and the dry ice trap receivers; 

the last two of these were bled directly to the large 

turbosaturator, thereby bypassing the scale and floor bleeders 

and the dry ice trap (Hunter 1949a). Condensate from the dry 

ice traps contained up to 40% HF (HCC 1946).

Up to 10% of the original quantity was 

bled off (Rauch 1948).

UF4-to-UF6 reactor 

hood exhaust

Exhausted to the large turbosaturator on roof, then to 

the atmosphere (Rauch 1948).

Direct vents April 1949: area in front of UF4 reactor hoods vented

directly to atmosphere; no provision for collection of 

dust; mostly UF4 being loaded but also UF6 fuming 

from reactors (Burman 1949; Hunter 1949a). UF6

process area ceiling exhaust also vented directly to 

atmosphere (Hunter 1949a).

Sump liquid Mainly washings from process areas, especially the 

hex reactor area (Burman 1949). This was filtered; the 

filtrate was sent to the sewer (Lynch 1949) and the 

press cake was weighed and sent to Vitro (Burman 

1949; Mayer and Proschen 1949; Lynch 1949).

Floor drainage and the water from the turbosaturator were 

collected and treated in a recovery process, producing press 

cake and ~150 gal/d of filtrate (Lynch 1949). April 1948: 

Cuyahoga River samples showed that U concentration to be 

0.00-0 .006 μg/cm3 in water and 0.00-0.58 μg/g in mud; said 

to be about normal background (Lynch 1949).

Digestion process 

residues

Slurry from digestor was sent through Niagara filter;

insoluble cake was removed, treated again with nitric 

acid, filtered; U adhering to cake after 2nd acid 

digestion and filtration was precipitated with strong 

caustic. Slurry was filtered; filtrate was sent to sewer 

and cake to Vitro in 1949 (Mayer and Proschen 1949).

Filtrate sent to sewer was not checked for uranium 

content, but the final cake was analyzed (Mayer and 

Proschen 1949).

1952 (Termini 1952): raffinate from the final extraction tank 

was collected in holding tanks for analysis; if > 0.1 g U/L,

raffinate was recycled or treated with caustic to precipitate 

soda salt (latter removed as a cake by filtration); if < 0.1 g U/L, 

raffinate was discarded. Average weight of cake from good 

(70%) black oxide feed, was 504 lb per digest batch; 1,975 lb

when operating on a 70%-30% batch (Fernelius 1950).

In Oct 1949, accumulations of 70,000 lb of 

press cake in 55-gal drums and 35,000 lb 

of Bird (centrifuge) residue were being 

stored temporarily at Harshaw (Blatz 

1949c).

Metal scrap Harshaw received Hanford metal scrap (DOE 2000); 

also generated its own metal scrap (AEC 1949b).

In April 1949, scrap stored at Harshaw; might have been 

disposed of in unknown ways previously (AEC 1949b).

Probably to Vitro.

30-, 55-gal drums Empty drums, stored as contaminated waste.

Turbosaturator 

slurry

The slurry from both turboagitators was filter-pressed 

to remove the uranium precipitates. The filtrate was 

sent to the sewer and the cake was weighed and sent 

to Vitro (Mayer and Proschen 1949).

Floor drainage and water from fume scrubbing (turbo-sat) was 

collected and treated in a recovery process, producing a press 

cake and 150 gal/d of filtrate (Lynch 1949). Apr 1948: 

samples from Cuyahoga River showed that U concentration 

was 0.00-0.006 μg/cm3 in water and 0.00-0.58 μg/g in mud, 

said to be about normal background (Lynch 1949).

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 9 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 85 of 114

Material Process or operation Content and form notes Amount

RESIDUES, OTHER WASTES, AND RECOVERED MATERIALS (Cont’d.)

Stack discharges Sep 1948: discharge, 40 lb/d U (Harris 

1949c). Jun 1949: 21 g/min total from 6 

reactor stacks, 13 g/min from 2 

turbosaturator stacks, 5 lb/hr from main 

sources (likely UF6) (Eisenbud 1949a); in 

outdoor air, no samples >10 μg/m3 at 100-

10,000 ft from plant; average conc at 0.3 

mi was 3 μg/m3 (Eisenbud 1949a). 

Samples up to Jul 1949: 22 lb/d (Harris 

1949c). 15 Nov -15 Dec 1951: losses of 

530 lb for UF6 area, 14 lb for UO3 area 

(Stefanec 1951).

Condenser cooling 

water

Normally not contaminated (AEC 1949b); sent to the 

Cuyahoga River (Lynch 1949; AEC 1949b).

April 1948: samples taken from the Cuyahoga River showed 

that the uranium concentration in the water was 0.00-0.006 

μg/cm3 and the mud concentration 0.00-0.58 μg/g, which were 

said to be of the order of normal background (Lynch 1949).

Aqueous tails from 

the ether 

extraction

The water fraction from the ether extraction was heated 

to boil off ether, which was recycled. The de-etherized 

solution was partly neutralized to precipitate impurities.

The cake was given a second acid leach and filtered. 

The filtrate was returned to the process flow; the cake 

was treated with caustic and filtered. The final cake 

was sent to Vitro. Caustic was added to the filtrate; 

after U precipitated, the slurry was filtered; the filtrate 

was alkalinized, refiltered. and discarded (Mayer and 

Proschen 1949).

The de-etherized solution had a low percentage uranium 

content (Mayer and Proschen 1949).

a. The text describes process details; Table B-4 lists other code numbers and terms.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 10 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 86 of 114

Table B-4. Functional and process keywords and codes. Process Keyword Notes

UF4-UF6, UF6 Distill 1st Level First floor of Bldg G1: all UF4-UF6, distillation processes located there

192 MED code number for the Harshaw laboratory

2 Generic MED code for soluble uranium forms

2048 Generic MED code for radiation

216 Fluorine or hydrofluoric gas

UF4-UF6 2nd Level Second floor of Bldg G1: all UO2-UF4 processes located there

32 Generic MED code for handling of (uranium) metal

36 MED code number for the Harshaw production plant

4 Generic MED code for insoluble uranium forms

UF4-UF6 490 Usage suggests this is an early synonym for 492

UO2-UF4 491 The UO2-to-UF4 (green salt) operation or area

UF4-UF6 492 The UF4-to-UF6 operation or area (hex area), including still ops

Ore-UO2, Ore-UO3 493 The ore-to-UO2 or -to-UO3 operation or area (brown plant)

516 UCl4

UF4-UF6, Distill 616 UF6 (and sometimes its ash residue)

64 Generic code for UF6

All Acid Usage suggests this is an early synonym for 492

UF4-UF6 Ash Left after fluorination to UF6 and removal of UF6

Ore-UO3 Batch In batches (i.e., as opposed to a continuous or semicontinuous process)

Ore-UO3 BDT & DEET Boildown tank and de-etherizing extraction tanks

Ore-UO3 Bird Centrifuge used in the ore-to-UO3 production process

Ore-UO3 Black oxide U3O8

Ore-UO3 Boildown Reduction used in the ore-to-UO3 production process

Ore-UO2, UO2-UF4 Brown UO2

UO2-UF4 Brown loader Job title for loader of UO2; also mechanical loader

All Buffalo Gas scrubber

C-216 Fluorine or hydrofluoric gas (HF)

C-516 UCl4

C-616 UF6

Ore-UO3 Centrifuge Centrifuge used in the ore-to-UO3 production process

All Cleanup Usually refers to general area cleanup (e.g., the vacuuming or washdown of floors)

UO2-UF4 Continuous furnace Used in the UO2-to-UF4 production process

CPM Chief process man (lead operator)

Ore-UO3 CR-15 UO3

UF6 Distillation Cylinder Container for UF6 storage and shipment

UO3-UO2 Deck In general, a localized or intermediate partial deck or floor for process access

All Decontamination Cleaning of equipment or areas beyond normal housekeeping by operators

All Develop, Dev, Dev Engr Development of engineering-type modifications, research

Ore-UO3 Digester Digest tank

UF6 Distillation Distillation Distillation of crude UF6 into specification-grade UF6

Ore-UO3 Ether Used in extraction process

Ore-UO3 Extraction Ether extraction of uranium in the ore-to-UO3 process

Ore-UO3 Feed Input to the digestion process (e.g., U3O8 or soda salt)

Ore-UO3 Frame Frame-and-press-type filter, producing a cake

UO3-UO2 Furnace (1) Rockwell furnace, (2) brown-to-green furnace

All G1 Building G1 (i.e., Plant C, where uranium processing took place)

UO2-UF4, UF4-UF6 Green UF4

Ore-UO3 Grinding Grinding orange oxide (UO3) lumps

Ore-UO3, Ore-UO2 H-32 Milled ore (black oxide)

Ore-UO3 HE Code name for Hanford

Ore-UO3 HE-33 Hanford-produced UO3 (sent to Harshaw for reprocessing)

UF4-UF6 Hex Area Loader Worker who loaded UF4 and (usually) who unloaded ash from the hex reactors

HGE Code name for Hanford, used as a prefix to identify Hanford-produced UO3 being 

cold-pressed prior to analysis at Harshaw

UO2-UF4 HL-7, HL7 UF4, the UO2-to-UF4 production process

UF4-UF6 Hopper For loading green salt

JH-6 Harshaw-produced UO3 made from recycle UO3 from Hanford

K1 Building K1, where analytical work was done

KoZ Unclear

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 11 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 87 of 114

Process Keyword Notes

UF4-UF6 Lift truck Used for loading green salt

UO2-UF4, UF4-UF6 Loader Usually, worker who loaded UO2 or UF4

Ore-UO3 NG Liquor NG = UNH (including the nitric acid?); also used as name of tanks in which extraction was 

done

Ore-UO3 Niagara Filter used in the ore-to-UO3 process

O.G. Fluorine (F2) gas

Ore-UO3, UO3-UO2 Orange UO3

Ore-UO3, UO3-UO2 PB-13, PB13 UO2; the ore-to-UO2 or –to-UO3 operation or area (brown plant)

P.G. UF6

PH Process helper

PH-30 Unclear

Pilot Plant Any of several small-scale facilities set up to produce a U form for a limited period

All Plant C The Harshaw uranium processing plant; usually, Bldg. G1

PM Process man (operator)

POB Unclear; it might be associated with the term "Process Man".

Ore-UO3 Pot Denitration pot

Ore-UO3, UF4-UF6 Press cake Residue containing trace uranium as potassium uranate (or typical product of a filter press, 

e.g., for sump washings or extraction fluids)

UF6 distillation Purification Synonym for distillation

Q-2-X A type of ore feed (U3O8)

Ore-UO3 Raffinate Postdigestion residue

UF4-UF6 Reactor Hex reactor (i.e., the reactor in which UF6 forms)

UF4-UF6 Receiver Container into which UF6 is drawn as it forms

All Recovery Recovery of wastes, for the ore-UO2 processes, or recovery of UF6 and associated vapors, 

for the UF4-UF6 and UF6 distillation processes

UF4-UF6, UF6 Distill Recovery Room Fume recovery room

UO3-UO2 Rockwell Brown process production furnace

UF4-UF6 RT-12, RT12 UF6, the UF4-UF6 production process

RX-10 UF6

S-15 Soda salt

All Scrubber Gas scrubber

Ore-UO3 Soda salt Sodium diuranate (Na2U2O7), a feed form

SS-20 It is unclear what this is. It may be a nonradioactive process material.

UF6 Distillation Still Distillation still used in the processing of crude UF6 to specification UF6

T Code synonym for U (e.g., TCl4, TF4 for UCl4, UF4)

Ore-UO3 TBP Ether form used for extraction

UF4-UF6, UF6 Distill Trap Usually, dry ice trap for UF6; could also be carbon or other trap for UF6

UO2-UF4 Tube Reaction container for UO2

UF4-UF6, UF6 Distill Turbosaturator Scrubber used to collect acid vapors and U gas forms

All Vacuum Portable vacuum cleaners or material transfer devices; the central vacuum system

WE-22 Probably ash

WE-22S Probably scrap material

UO2-UF4 WE-61 UF4 or its processing

X-32 Same as H-32?

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 12 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 88 of 114

Table B-5. Measured dose rates.

a

Area or source item Position/distance

Gamma 

dose rate, 

mR/hr

Beta dose 

rate, 

mrep/hr

Total dose 

rate, 

mrep/hr Date Reference

UCl4 production laboratory ("1945 East 97th Street")

Near shelves with bottles of UCl4 Highest rate 1.3 Circa 1944 Gamertsfelder ca. 

1944

Storeroom Highest rate 1.8 Circa 1944 Gamertsfelder ca. 

1944

In fluorinat'g hood, not being used 0.8 Circa 1944 Gamertsfelder ca. 

1944

UF4 plant ("1000 Harvard Avenue")

Dust-covered UF4 canning table 2 in. away 3.3 Circa 1944 Gamertsfelder ca. 

1944

At worker position 1.2

2 ft from table, 4.5 ft from 

floor

0.4

Man's shirt, worn in lab 5 d Contact with 

electroscope

0.6 Circa 1944 Gamertsfelder ca. 

1944

Storeroom, Fe can with 140 lb UF4 Top of can 4 Circa 1944 Gamertsfelder ca. 

1944

UO2-to-UF4 (Green Area, 2nd level)

Corridor of UO2 storage area, avg 43 in. above floor 1.5 1.1 2.6 May 1948 Hayden 1948

UO2 carton by Furnace #41 Top, 1 in. 10 Aug 1947 Turner 1947e

Put-up bench, layer of fine UO2 Top, 1 in. 20 Aug 1947 Turner 1947e

Under put-up bench, r sweepings Floor, 1 in. 2 Aug 1947 Turner 1947e

Under furnace loading stand Floor, 1 in. <2 Aug 1947 Turner 1947e

Furnace purging rack Floor, 1 in. <2 Aug 1947 Turner 1947e

Under furnaces Floor, 1 in. ND-<2 Aug 1947 Turner 1947e

Rear of furnaces Floor, 1 in. ND Aug 1947 Turner 1947e

Burned trays Top, 1 in. <2 Aug 1947 Turner 1947e

In front of panel board Floor, 1 in. 5 Aug 1947 Turner 1947e

Metal cans near furnace Side, 1 in. <2 Aug 1947 Turner 1947e

Repair bench, fine UF4 layer Top, 1 in. 2 Aug 1947 Turner 1947e

Worker's right shoe top, UF4 on it Top, 1 in. 17 Aug 1947 Turner 1947e

Fence (inside) in SW area Floor, 1 in. ND Aug 1947 Turner 1947e

At south wall, over storage boxes Wall, 1 in. 5 Aug 1947 Turner 1947e

By window upstairs (green area?) Floor, 1 in. ND Aug 1947 Turner 1947e

Stairs between UF4, Hex areas:

Top landing Floor, 1 in. <2 Aug 1947 Turner 1947e

Middle of stairs Floor, 1 in. 10

Bottom of stairs Floor, 1 in. 10.5

Hoods 43 in. above floor 0.6 4.4 4.9 May 1948 Hayden 1948

General hood area 43 in. above floor 1 1.3 2.3 May 1948 Hayden 1948

At a reactor tube rack 43 in. above floor 0.5 1.2 1.7 May 1948 Hayden 1948

At reactor tube cooling rack 43 in. above floor 0.4 1.9 2.3 May 1948 Hayden 1948

Product container cleanout area 43 in. above floor 1 7.8 8.8 May 1948 Hayden 1948

UF4-to-UF6 (Hex Area, 1st level)

Catch pans on scale, UF4 in pan 1 in. 100-240 Aug 1947 Turner 1947e

Above a UF4 tray 1 in. 170 Aug 1947 Turner 1947e

Above a UF4 tray 3 in. 240 Aug 1947 Turner 1947e

Near tray, some UF4 salt and ash Floor, 1 in. 15 160 Aug 1947 Turner 1947e

Loading rack w/ UF4 and hex ash Surface, 1 in. <2 52 Aug 1947 Turner 1947e

Loading rack hood end: UF4, ash Surface, 1 in. 60 1,200 Aug 1947 Turner 1947e

Loading rack Top, 1 in. 18 240 Aug 1947 Turner 1947e

Loading rack Top, 1 in. 240 Aug 1947 Turner 1947e

2 ft in rear of cell rooms Floor, 1 in. ND Aug 1947 Turner 1947e

Inside cell room Floor, 1 in. ND Aug 1947 Turner 1947e

Instrument panel for Cells #18, 18A Floor, 30 in. 10 Aug 1947 Turner 1947e

In front of Panel Board #5 Floor, 1 in. 5 Aug 1947 Turner 1947e

2 ft S of Hood #3, UF4 on wet floor Floor, 1 in. 10 135 Aug 1947 Turner 1947e

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 13 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 89 of 114

Area or source item Position/distance

Gamma 

dose rate, 

mR/hr

Beta dose 

rate, 

mrep/hr

Total dose 

rate, 

mrep/hr Date Reference

Various points around hoods Floor, 1 in. 15-40 Aug 1947 Turner 1947e

Top of reactor rod rack, Crew #3 Floor, 1 in. 20 Aug 1947 Turner 1947e

Various pts betw. cells, reactors Floor, 1 in. 22-25 Aug 1947 Turner 1947e

At hoods, 1 ft from receivers Floor, 1 in. 5-7 25-55 Aug 1947 Turner 1947e

Near receivers opposite cells Floor, 1 in. 7.5 20-60 Aug 1947 Turner 1947e

6 in. from receiver opposite a cell Floor, 30 in. <1 25 Aug 1947 Turner 1947e

4 ft in front of receivers opp. a cell Floor, 30 in. 20 Aug 1947 Turner 1947e

Main entrance to building (Annex) Floor, 1 in. ND Aug 1947 Turner 1947e

Doorway opp. Furnace Bank #20 Floor, 1 in. 10 Aug 1947 Turner 1947e

Various pts around furnace banks Floor, 1 in. 5-20 Aug 1947 Turner 1947e

Above a tray containing hex ash 1 in. 600 (3 in.) 6,400 (1 in.) Aug 1947 Turner 1947e

Floor drain betw. Cells #9A, #10 Floor, 1 in. 22 Aug 1947 Turner 1947e

Floor drain in front of Hood #2 Floor, 1 in. 12.5 Aug 1947 Turner 1947e

Drain in front of hex work table Floor, 1 in. 5 Aug 1947 Turner 1947e

Dolly of Hood #6 Top, 1 in. 2 25 Aug 1947 Turner 1947e

Dolly of Hood #1 Floor, 1 in. 5 50 Aug 1947 Turner 1947e

"Cap dollie" (receiver cap dolly?) Floor, 1 in. 50 Aug 1947 Turner 1947e

Around various ice traps Floor, 1 in. 10-20 Aug 1947 Turner 1947e

Around ice traps, hydrolized UF6 on 

floor

Floor, 1 in. 2-5 20-25 Aug 1947 Turner 1947e

Top of Ice Trap #69 Top, 1 in. 3 Aug 1947 Turner 1947e

Hoods 43 in. above floor 2.6 24.5 27.1 May 1948 Hayden 1948

Near loading hood 43 in. above floor 0.5 4.4 4.9 May 1948 Hayden 1948

Near UF4 loading rack 43 in. above floor, 5 ft 

away

0.3 2.6 2.9 May 1948 Hayden 1948

UF4 weighing scale 43 in. above floor 1 6 7 May 1948 Hayden 1948

Control aisle (panel boards) 43 in. above floor 0.93 6.57 7.5 May 1948 Hayden 1948

Reactor area 3 Apr 1952 HCC 1950–1953

Dry ice trap 43 in. above floor 2.2 14.9 17.1 May 1948 Hayden 1948

Walls with hydrolized UF6 Wall, 1 in. 110-150 Aug 1947 Turner 1947e

Near steam table Floor, 1 in. 110 Aug 1947 Turner 1947e

Wall E of steam table, 4 ft over floor Wall, 1 in. 80 Aug 1947 Turner 1947e

Wall W of clean'g table, 4 ft over flr Wall, 1 in. 10 120 Aug 1947 Turner 1947e

Near hot water container Floor, 1 in. 240 Aug 1947 Turner 1947e

Opposite ice box Floor, 1 in. <2 Aug 1947 Turner 1947e

Empty receiver storage area Floor, 1 in. <2 Aug 1947 Turner 1947e

Under off-gas vent line 3 ft S of 

east wall

Floor, 1 in. 30 Aug 1947 Turner 1947e

Door, N2 trailer/cold H2O machine Floor, 1 in. <2 Aug 1947 Turner 1947e

Opposite CuF2 drums Floor, 1 in. 70 Aug 1947 Turner 1947e

Line unplugging area (UF6 lines) 43 in. above floor 8.7 104.3 113 May 1948 Hayden 1948

Line cleaning vat (UF6 lines) 43 in. above floor 6.4 35.4 41.8 May 1948 Hayden 1948

Line repair area 43 in. above floor 1.2 6.1 7.3 May 1948 Hayden 1948

Empty used cylinder to be painted Floor, 1 in. 5 Aug 1947 Turner 1947e

Full cylinder on storage rack Floor, 1 in. <1 -2 Aug 1947 Turner 1947e

Cylinder work table Top, 1 in. 2 Aug 1947 Turner 1947e

Between, behind stills, trap scale Floor, 1 in. ND - 2 Aug 1947 Turner 1947e

2" in front of still room Floor, 1 in. <2 Aug 1947 Turner 1947e

Outside still area Floor, 1 in. ND Aug 1947 Turner 1947e

Still area (UF6 distillation) 43 in. above floor BG BG BG May 1948 Hayden 1948

1 ft south of Recovery Hood #16 Floor, 1 in. <2 20 Aug 1947 Turner 1947e

1 in, S of Recovery Hood #20, UF4

on floor

Floor, 1 in. 2 25 Aug 1947 Turner 1947e

6 ft S of recovery hoods, ash on flr Floor, 1 in. 2-130 50-2,400 Aug 1947 Turner 1947e

Asbestos gloves, used, hex crew Surface, 1 in. 20 Aug 1947 Turner 1947e

Asbestos gloves on flr, hex panel Surface, 1 in. 140 Aug 1947 Turner 1947e

Rubber gloves worn by operator Surface, 1 in. 5 Aug 1947 Turner 1947e

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 14 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 90 of 114

Area or source item Position/distance

Gamma 

dose rate, 

mR/hr

Beta dose 

rate, 

mrep/hr

Total dose 

rate, 

mrep/hr Date Reference

Tray welding, cutting (subcontr.) 40 Oct 1950 HCC 1950–1953

Residue storage room Center of room 21 Nil 21 Sep 1944 Tybout 1944a

Outside door to ash collector rm 43 in. above floor 7.7 6.2 13.9 May 1948 Hayden 1948

Ash collector room, average At bin, 43 in. above floor 42 44 86 May 1948 Hayden 1948

Tray of fresh ash (before putting 4 in. above tray 4,000, 

2,000

Feb 1946 Tybout 1946

 into can) 3 in. above tray 3,000 Feb 1946 Tybout 1946

Smaller container (can) of ash Side, contact 11,000 Feb 1946 Tybout 1946

 (1/32-in, steel wall) Side[?], 1.5 ft 2,000 Feb 1946 Tybout 1946

Ash can from center separ. (filter) Touching the top 3,300 Aug 1947 Turner 1947e

Ash collector drum on roof Contact >20 Apr 1949 Wolf 1949

Drum of roof asha No lid , 2 in. above ash 300 Jun 1949 Piccot 1949

Side, contact 164[?]

Side, 8 in.[?] Illegible

Side, 12 in. Illegible

Side, 24 in.[?] Illegible

Smaller container (can) of ash Side, contact 110 (2.5") Jun 1949 Piccot 1949

 (1/8-in. steel wall) Side, 8 in.[?] 35

Side, 12 in. 11

Side, 18 in.[?] 3

Side, 24 in.[?] 3

Fresh ash in large beaker 6 in. 160 2,300 >2,500 α May 1950 Klevin 1950c

Fresh ash in large beaker covered 

with a watch glass

6 in. 12.5 350 410 α May 1950 Klevin 1950c

Fresh ash, fines 6 in. 650 >2,500 >2,500 α May 1950 Klevin 1950c

Fresh ash, coarse residue 6 in. 54 950 1,100 α May 1950 Klevin 1950c

Service corridor, west end 43 in. above floor 0.04 0.9 0.9 May 1948 Hayden 1948

Service corridor, center 43 in. above floor 0.12 0.8 0.9 May 1948 Hayden 1948

Macadam floor (hex area?) Contact 17 May 1952 Blatz 1952

Ore-to-UO2 or -UO3 (Brown/Orange Area)a

Extraction tanks NG-1, NG-2 Contact 3.5 Apr 1949 Wolf 1949

Extract. boildown tanks (4a, 4b, 5) Contact 1.2 Apr 1949 Wolf 1949

Extraction tanks NG-1, NG-2 Contact 4 4 Oct 1949 Blatz 1952

 (higher of the two readings is 6 in. 2 2

 shown) 12 in. 1 1

18 in. 0.9 0.9

24 in. 0.6 0.6

36 in. 0.5 0.5

Bird centrifuge 2 ft? 1 Apr 1949 Wolf 1949

Niagara and miscellaneous filters Half-full, contact 1 Apr 1949 Wolf 1949

Niagara filters 1, 2 Contact 0.4, 0.3 20, 18 May 1950 Sargent 1950c

6 in. 0.25, 0.2 9, 7.5

12 in. 0.2, 0.1 4, 3.5

Plate-and-frame filter Contact 1.6 May 1950 Sargent 1950c

6 in. 1.4

8 in. 20

12 in. 1 14

24 in. 8 May 1950 Sargent 1950c

Postdigestion, Hanford HG-33: Mar 1953 Klevin 1953b

 Digest tank (empty) 1 in. 2.5

 NG liquor (extraction) 1 in. 2.5

 Raffinate liquor 1 in. 10.5

 Strip water 1 in. 0.05

 Stripped solvent 1 in. 0.25 - 2.0

 Recovered acid 1 in. 0.25

 Tank #9 recovered acid 1 in. 0.5

 Bi-Carbonate scrub 1 in. 2

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 15 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 91 of 114

Area or source item Position/distance

Gamma 

dose rate, 

mR/hr

Beta dose 

rate, 

mrep/hr

Total dose 

rate, 

mrep/hr Date Reference

 Raffinate cake 1 in. 2.5 - 5.0

 Discharge vapor: stack exhaust

(mixer-settler extractor fumes)

1 in. 7-9

 Vapors off strip water exhaust 1 in. 0.02

 Nitric acid recovery exhst stack 1 in. 0.02

 Waste liquor to sewer 1 in. 10.5

Digester tank 1-2 ft above liquid 1 5-9 May - Jul 1953 HCC 1950–1953

Raffinate tank 2 in.-3 ft above liquid 0-0.2 4-12 May - Jul 1953 HCC 1950–1953

Press cake filtrate, in process 3 in. above liquid 0.05 4 May - Jul 1953 HCC 1950–1953

Bi-Carb scrub tank 1 in.-6 ft above liquid 0 0-0.7 May - Jul 1953 HCC 1950–1953

Strip water tank 2 in.-3 ft above liquid 0 0-0.8 May - Jul 1953 HCC 1950–1953

Strip solvent tank 2 ft above liquid 0.2 May - Jul 1953 HCC 1950–1953

Recovery acid tank 1 in.-4 ft from liquid 0 0-2 May - Jul 1953 HCC 1950–1953

Press cake filtrate, in process 3 in. above liquid 0.05 4 May - Jul 1953 HCC 1950–1953

Sewer, 0.01-0.02 g/L raff. expelled 2-3 in. from liquid 0 6-11 May - Jul 1953 HCC 1950–1953

Digest tank exhaust Stack, discharge 0 0.8 May - Jul 1953 HCC 1950–1953

Strip water exhaust Stack, discharge 0 0-0.1 May - Jul 1953 HCC 1950–1953

Extractor fume exhaust Stack, discharge 0 0-3 May - Jul 1953 HCC 1950–1953

Spilled cake from frame/press filt. 2-3 ft from spill 2 >20 Sep 1949 Blatz 1949a,b

Group 55-gal drums, press cake Contact? 1 8 Sep 1949 Blatz 1949a, 1949b

Group 55-gal drums, Bird residue Contact? 3 Sep 1949 Blatz 1949a,b

Group 55-gal drums, Bird residue 2 ft 2 Apr 1949 Wolf 1949

Pot, during denitration 1 5 May - Jul 1953 HCC 1950–1953

Pot, gulping ~3 in. above 0.5-1 5-6 May - Jul 1953 HCC 1950–1953

Pot, opened for repair 2 in. above bottom 0 2 May - Jul 1953 HCC 1950–1953

Other areas within Plant C

Wooden pallets in storage area Surface, 1 in. 10 Aug 1947 Turner 1947e

Floor betw. stock room, elevator Floor, 1 in. ND Aug 1947 Turner 1947e

Stacks of UO3 cardboard Contact 1.5 Apr 1949 Wolf 1949

containers, being stored General area 1

UO3 containers to be shipped Contact 1.5 May 1952 HCC 1950–1953

Laboratory Floor, 1 in. ND Aug 1947 Turner 1947e

Laboratory, sample & other tables Top, 1 in. ND Aug 1947 Turner 1947e

Work bench, Maintenance Shop Top, 1 in. ND Aug 1947 Turner 1947e

Maintenance room Floor, 1 in. ND-<2 Aug 1947 Turner 1947e

Leather glove, Maint Shop, left 

hand, still being used

Surface, 1 in. 10 Aug 1947 Turner 1947e

Leather glove, Maint Shop, right 

hand, still being used

Surface, 1 in. 60 Aug 1947 Turner 1947e

Locker Rm #1: shower, lav., area Floor, 1 in. <2 Aug 1947 Turner 1947e

Lunchrm #1 (cafeter.), kitchen, rec 

room

Floor, 1 in. ND Aug 1947 Turner 1947e

Lunchroom #2, floor and benches Floor, 1 in. <2 Aug 1947 Turner 1947e

Lunchroom #2, table Top, 1 in. ND Aug 1947 Turner 1947e

Lunch, locker, shower rooms 43 in. above floor BG BG BG May 1948 Hayden 1948

Superintendent's desk in office Top, 1 in. <1 <1 Aug 1947 Turner 1947e

1 ft in front of lavat. in Supt's office Floor, 1 in. <1 <1 Aug 1947 Turner 1947e

1 ft in front of clerk's desk in office Floor, 1 in. <1 5 Aug 1947 Turner 1947e

1 ft in front of filng cabinet in office Floor, 1 in. <1 <1 Aug 1947 Turner 1947e

Aisle near hex area, 15 ft W of 

office

Floor, 1 in. 2 Aug 1947 Turner 1947e

Office, superintendent's 43 in. above floor BG (.2) BG BG May 1948 Hayden 1948

Drinking fountain 43 in. above floor 3.4 1.8 5.2 May 1948 Hayden 1948

Roof and yard areasb

Annex roof, 20 ft fr WE-61 [vent?]

Roof, main building Roof, 1 in. ND Aug 1947 Turner 1947e

Roof S of UO2-to-UF4 tube rack 43 in. above floor 0.64 5.1 5.2 May 1948 Hayden 1948

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 16 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 92 of 114

Area or source item Position/distance

Gamma 

dose rate, 

mR/hr

Beta dose 

rate, 

mrep/hr

Total dose 

rate, 

mrep/hr Date Reference

Surface, can of UF6 cylinder heel Inside asbestos gloves 2.5 Jun 1945 Tybout 1945a

Inside leaded glove 136.3

Roof, main building Roof, 1 in. ND Aug 1947 Turner 1947e

Drum 2/3 full: heel + water washed 

from cylinders

Contact 13.8 Jun 1945 Tybout 1945a

80-90 horiz. cylinders, at 1 lb UF6 6 in. above cylinder pile 2.0-2.5 Jun 1945 Tybout 1945a

Two stacks cylinders: 4 ft high, 20

ft long, 1.5 ft apart, 6 in. from 

ground

Between the two stacks 0.5 Jun 1945 Tybout 1945a

Pile of cylinders 10 ft from bottom 0 Jun 1945 Tybout 1945a

Pile of cylinders 3 ft from broadside end 0.5-.75 Jun 1945 Tybout 1945a

Line of 25 full cylinders of UF6, laid 

singly on ground

6 in. above 0.3 Jun 1945 Tybout 1945a

Cylinder storage area (yard?) General area 3 Apr 1949 Wolf 1949

Outside wall of guard shack near 

cylinders

Contact, 2 ft off ground 2 Jun 1945 Tybout 1945a

Badge board at guard shack 1 ft in front 0.5 Jun 1945 Tybout 1945a

Railcar floors Contact? 1 0.0-0.5 1.5 Apr 1951 HCC 1950–1953

Fence opposite trailer, outside Surface, 1 in. ND Aug 1947 Turner 1947e

2 ft down inside sewer Inside, 1 in. <2 <2 Aug 1947 Turner 1947e

a. Question marks indicate that the reference was illegible or that the information was incomplete or ambiguous. For example, "Contact?" indicates that the 

distance was not specified but the likely measurement point was at contact.

b. Ash from a faulty collector blew over the roof at times. The above roof ash readings were not considered reliable because the roof and surroundings were 

contaminated.

c. Cylinders are UF6 cylinders, empty except as noted.

Table B-6. Chest and hand beta doses from ash residue handling, as measured 

by films.

a

Operation Dose, R Dose rate, R/hr

Hand (film) dose during ash container transfer 0.4 2.4

Tray handling, day 1 (0.35 hr)

Chest (regular) film badge, worn 1 d 0.21 0.60

Right wrist 0.225 0.64

Right back of hand 0.2 0.57

Left wrist 0.2 0.57

Left back of hand 0.15 0.43

Tray handling, day 2 (1.05 hr)

Chest (regular) film badge, worn 1 d 0.28 0.27

Tray handling, day 3 (1.75 hr)

Chest (regular) film badge, worn 1 d 0.35 0.20

Right wrist 0.45 0.26

Right back of hand 0.3 0.17

Left wrist 0.4 0.23

Tray handling, day 4 (1.75 hr)

Chest (regular) film badge, worn 1 d 0.35 0.20

Right wrist 0.45 0.26

Right back of hand 0.4 0.23

Left wrist 0.4 0.23

Left back of hand 0.3 0.17

Chest (regular) film badge, worn all 4 d (5.07 hr total) 1.4 0.28

a. These films were exposed at the same time as the ash residue dose rate measurements in Table B-5 for 

February 1946 were taken (Tybout 1946; Engel 1946). Dose rate measurements were taken with an ion 

meter and did not take into account the shielding provided by the cotton and asbestos gloves worn by the 

tray handlers. The hand films listed above were taped on the skin under the gloves. Note that units of 

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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roentgen were used by the reference, although the measurement was specifically beta. Missing 

measurements above are due to the films’ being damaged by perspiration and heat.

Table B-7. Weekly dose rates for various workers and areas (from AEC 1949b, Figures 8 and 10).

Position No.

Average dose rate 

(mrep/wk)

Median dose 

rate (mrep/wk)

Minimum dose 

rate (mrep/wk)

Maximum dose 

rate (mrep/wk)

August 1944 – January 1949a

Hex area operator 24 600 630 345 1,145

Hex area loader 25 970 905 270 2,000

Still operator 8 350 250 165 840

Brown loader, reactor operator 16 265 240 170 650

Maintenance 4 280 315 120 595

Recovery 1 360 --- --- ---

Supervisor 5 200 220 195 335

Foreman 3 220 210 190 275

Miscellaneous and unclassified 5 190 195 110 335

Clerks 1 100 --- --- ---

Plant manager 1 80 --- --- ---

May 1948b Beta-gamma 

(mrep/wk at 4 in.)

Beta-gamma 

(mrep/wk at 43 in.)

UO2 tray loaders 6? 340 70

UO2 tube handlers 6? 145 80

UF4 Loaders, etc. (day shift) 9 1,640 463

UF4 Loaders, etc. (day shift) 9 885 250

Still area 6? 137 38

Recovery room 1 830 243

Office 4 93 33

a. The figures given for August 1944 to January 1949 are individual averages based on the total dose accumulated over the period, from film 

badge data. The maximum dose rate is the average dose rate to the most exposed person (in terms of total dose) in the position; the minimum 

is the average dose rate to the least exposed person.

b. The exposures given for May 1948 appeared to arise mostly from UX1 and UX2. Daily dose rates were multiplied by 5 to get weekly dose rates 

(the probable original units).

Table B-8. Annual neutron whole-body doses from the alpha-neutron reactiona, various uranium 

forms.

Worker type Form Sourceb Target

Weight of 

form in 

containerc Applicable years 

Dose rate, 

1 ft,

rem/hr

Dose rate, 

3 ft,

rem/hr

Annual 

dose, 

remd

Ore/digestion U3O8 (covers Na2U2O7) NU+D O (Na,O) 100 (75) lb 1949–1954 4.74E-06 5.27E-07 1.58E-03

UO3, UO2

load/unloader

UO3, UO2 NU O 75 lb 1949–1954 2.70E-07 3.00E-08 9.00E-05

UF4

load/unloader

UF4 NU F 75 lb 1942–1944; 

1948–1951

2.26E-05 2.51E-06 7.53E-03

NU+D 1945–1947 4.04E-04 4.49E-05 1.35E-01

UF4-UF6

loading

UF4 NU F 75 lb 1942–1944; 

1948–1951

2.26E-05 2.51E-06 7.53E-03

NU+D 1945–1947 4.04E-04 4.49E-05 1.35E-01

UF6 (receiver) NU 435 lb 1942–1951 1.31E-04 1.46E-05 4.37E-02

UF6 distillation UF6 (cylinder) NU F 462 lb 1942–1951 1.39E-04 1.54E-05 4.63E-02

Shipping & 

Receiving

UF6 (cylinder) NU F 462 lb 1942–1948 1.39E-04 1.54E-05 4.63E-02

U3O8 (covers Na2U2O7) NU+D O (Na,O) 100 (75) lb 1949–1951 4.74E-06 5.27E-07 1.58E-03

UO3 NU O 75 lb 1952–1954 2.70E-07 3.00E-08 9.00E-05

UO2 NU+D 75 lb 1942–1951 4.74E-06 5.27E-07 1.58E-03

Laboratory Ore/HGE-33/other feed NU+D O (Na,O) --- 1949–1954 4.74E-07 5.27E-08 3.16E-04

UF4 (from Harshaw) NU F --- 1949–1954 2.26E-06 2.51E-07 1.51E-03

UF6 NU --- 1942–1951 2.26E-06 2.51E-07 1.51E-03

UO3, UO2 (from others) NU+D O --- 1942–1953 4.74E-07 5.27E-08 3.16E-04

UO3, UO2 (fr/ Harshaw) NU --- May 1949-1953 2.70E-08 3.00E-09 1.80E-05

a. Dose rates were calculated using data and assumptions from ORAUT (2005c). The neutron doses in Table B-8 should be multiplied by 2 to correct the 

values calculated from ORAUT (2005c) to the ICRP Publication 60 radiation weighting factor for the energy range (ICRP 1991).

b. NU: natural uranium mix, uranium only; NU+D: natural uranium mix with daughters through polonium (i.e., full secular equilibrium).

c. Production data are from Gates (1946), Rauch (1948), HCC (1946), Tybout (1944a), Tybout (1945a), Simmons (1945), and AEC (1951c).

d. In calculating annual dose, it was assumed for all but the laboratory worker and the office/clerical/manager worker that 4 hr/d was spent handling the 

container: 1 hour at 1 foot and 3 hours at 3 feet. For the laboratory worker, source quantities were much smaller but close access time was greater; thus 

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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the dose rates were multiplied by 0.1 to allow for smaller quantities, while for the doses the times spent per day were doubled. For the 

office/clerical/manager worker, the dose was assumed to be negligible (by inspection of the more exposed worker doses and by consideration of the 

likelihood of exposure). In all cases, the work year was taken to be 2,000 hours.

Table B-9. Reserved.

a

Table B-10. Air concentrations in various areas in the green and hex plants (Ferry 

1944b; Tybout 1945b).

Location X dust (μg/m3)

a Uranium dust exposure (dpm/m3

)

b

Green (UF4) Plant UO2/UF4 exposure (dpm/m3

)

Center of room 69 in. high 230 322

In front of hood, unloading 240 336

Charging in front of hood by scale 200 280

Front center of hood 350 490

In front of tube rack 69 in. high 320 448

Next to record desk 240 336

Hex (UF6) Plant UF6 exposure (dpm/m3

)

East end of corridor 30 42

Center of corridor 89 125

West end of corridor 40 56

East end of Hood 9 370 518

Center of Hood 9 30 42

West end of Hood 9 9,130 12,782

a. All μg/m3 values are ±10%, except for the last entry, which corresponds to ±20%.

b. The tolerance level is given as 150 μg/m3 of "X dust" (uranium dust) and 40 μg/m3 for UF6. For this table, it 

was assumed that 50 μg of uranium was equal to 70 dpm.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Table B-11. General area and breathing zone dust measurements in the green and hex plants (491 and 492), in alpha dpm/m3.

Ferry 

1944b

Tybout 

1945b

Hayden 

1948

Harris 

1949d, 

AEC 

1951d

AEC 

1950d

AEC 

1950e, 

AEC 

1950c

AEC 

1950c

HCC 1950–

1953

AEC 

1951b

AEC 

1951d

AEC 

1953a

AEC 

1953a

AEC 

1953a

AEC 

1953a

MED MED AEC AEC Harshaw AEC AEC Harshaw AEC AEC Harshaw AEC Harshaw AEC

Feb

1944 Sep 1944

May 

1948 Sep 1948

Feb

1949 Sep 1949

Jul

1950

Oct

1950

Jan

1951 Sep 1951 Sep 1952

Jan

1953

Feb

1953

Aug 

1953

General Area Samples

491 general area, postoperations 3 26 7 1

492 general area, postoperations 0 0 11 2

Reactor furnace area 260 532 63 336 126 238 63

HF tank scale area 469 42 399 49 133 203

Brown loading room area 336 18,000 441 1,610 525 140 350 77

Still area 1,300 406 280 602 210 1,274 63

SW corner of the still area 294 77 658 280

Hex area 125 175 1,260 511 7,490 252 217 189

Line cleaning area 1,000

Central loading room 21,000 98 434 140

Lunchroom No. 1 150 126 77 56 42 91 56

Cell-Steam Room area 273 91 7 14 42 21

Locker Room No. 1 150 413 70 49 56 329 98

Lunchroom No. 2 150 42 42 91 70 28 70

Locker Room No. 2 150 84 42 63 63 21 35

Lunchroom No. 3 150 21 70 14 7

Locker Room No. 3 150 63 119 84 35

Maintenance shops 21 98 322 196 28

Maintenance office 77 28 42 49 14

Cylinder storage area 84 210 161 196 224

Cylinder wash area 70 553 35 63 714

Respirator cleaning area 1.7

Shipping & Receiving area 2.0 3.5 83.0 0.4 0.9 0.3

Laboratory area 1.0 3.8 1.0 1.0 0.9

Guard office 0.8 0.1 0.3 0.2 0.3

Health Physics 2.1 2.3 0.5 0.9

Recovery area 18.5 0.8 2.3 1.0

Recovery Room 35,400 180

Production Office 160 77 119 56 28 21

Breathing Zone Samples

Reactor furnace area

Disconnect HF lines 910 0.0 371 371 651 595

Remove hot tube to rack 3,290 0.0 1,687 497 98 70

Connect HF lines 910 637 2,660 294 259

Clean up area 8,680

Still operations

Break receiver connections 672 35 203 77

Installing receivers 7,000 210

Removing receivers 1,610 56

Take used receivers to storage, put in 

new ones

448 84 168 126

Reconnect receiver lines 931 1,960 294 406

Brown oxide loading

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 96 of 114

Ferry 

1944b

Tybout 

1945b

Hayden 

1948

Harris 

1949d, 

AEC 

1951d

AEC 

1950d

AEC 

1950e, 

AEC 

1950c

AEC 

1950c

HCC 1950–

1953

AEC 

1951b

AEC 

1951d

AEC 

1953a

AEC 

1953a

AEC 

1953a

AEC 

1953a

MED MED AEC AEC Harshaw AEC AEC Harshaw AEC AEC Harshaw AEC Harshaw AEC

Feb

1944 Sep 1944

May 

1948 Sep 1948

Feb

1949 Sep 1949

Jul

1950

Oct

1950

Jan

1951 Sep 1951 Sep 1952

Jan

1953

Feb

1953

Aug 

1953

Open UO2 carton 95,200 57,050 37,100

Carry tubes from storage to loading 

hood

490 1,680 1,330 469 126 98 70

Open tubes 13,300 11,900 651 7,770 6,790 6,090

Unload UF4 336 24,500 30,100 53,200 10,430 9,170 2,310

Load trays with UO2, put into tubes 280 14,700 5,600 3,080 5,950 7,840 1,890

Prepare and seal tubes 6,720 9,450 5,180 9,100 4,060 3,710

Weigh, seal, tag UF4 drum 35,000 6,790 115,500 35,910 2,240 10,500

Clean up area 122,500 1,820 5,180 3,080 1,274 4,760

Take tubes to UF4 storage area 448 1,680 1,330 469 126 98 70

Hex loading, before installation of the central loading system

Break lines 28,000 3430 14,000

Remove hot tube to rack 25,200 2660 7,700

Load UF4 into trays 4,448 210,000 25,900 12,250

Place trays in furnace 72,100 2,100 12,250

Secure furnace, connect lines 13,650 980 3,500

Pull ice traps 56,000 37,800 12,250

Clean up area 46,900 490 2,100

Remove ash from tube 15,750

Special operations (by subcontractor)

Cutting supports of trays 158

Welding 64

Press brake operator, welding 36

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 97 of 114

Table B-12. General area and breathing zone dust measurements in the brown plant (493), in alpha dpm/m3 a. Klevin 1950a, 

AEC 1950e AEC 1950e AEC 1951a AEC 1952a AEC 1953b AEC 1953b AEC 1953b AEC 1953b AEC 1953a AEC 1953a

AEC AEC AEC AEC Harshaw AEC Harshaw AEC Harshaw AEC

Mar 1950 Aug 1950 Mar 1951 Jan 1952 Jun 1952 Jul 1952 Sep 1952 Jan 1953 Feb 1953 Aug 1953

General Area Samples

X-Tank area 29 62 16 59 22 78 8

Y-Tank area 51 50 10 14 29 122 19

Bird centrifuge 126 133 35

Digestion area 112 42 28 22 82 <1 12 33 49 45

TBP panel board area 70 31 <1 13 8 49 5

Denitration pot area 112 42 63 72 69 77 7.3 41 123 21

Niagara, frame filters (3rd deck) 259 49 21

Digestion and recovery tank area 336 28 56

Ether extraction area 42 91 21

Nitric acid makeup area 140 28 28

Rockwell area 567 441 315

Discharge of Rockwells 48 68 8 42 58 93 49

2nd deck over Rockwells 256 114 39 36 100 43 22

3rd deck over Rockwells 1,030 61 110 363 161 75 19

Orange packing area 56 49 157 15 19 125 23

Room 1-A 33 79 86 30 20 41 1.0

Room 1-B 18 66 <1 14 5.6 28 1.0

Room 1-C 1 34 13 12 2.5 17 2.0

Room 2-A 11 78 55 8.6 17 69 4.0

Room 2-B 15 67 360 9.6 4.8 54 25

Room 3-A 17 39 43 9.7 82 34 7.0

Room 3-B (filter press) 17 18 12 14 116 28 15

Room 3-C 9

Brown production area 546 161 140

Recovery area 11 3 8.8 9.1 6 1

Production office 189 14 49 8 4 <1 1.6 6.3 12 1

Locker Room No. 3-C, D 49 21 84 27 0 1.5 6.3 11 6 6

Lunchroom No. 3 56 7 42 7 3 0 3.0 11 23 18

Locker Room No. 2-C, D 41 15 17 11 22 23 51

Lunchroom No. 2 32 19 18 12 2.8 33 8

Locker Room No. 1-C, D 30 4 40 2.6 48 19 2

Lunchroom No. 1 63 7 17 3 35 28 2

Laundry - dirty 126 42 84 0 11 27 4 2

Laundry - clean 14

Laboratory grinding room 5 8.6 650 92 6

Center of main laboratory 8 5 95 23 18

Laboratory office 0 11 0 7 3

Balance Room 0 7.6 2.8 0 4

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 22 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 98 of 114

Klevin 

1950a, 

AEC 1950e AEC 1950e AEC 1951a AEC 1952a AEC 1953b AEC 1953b AEC 1953b AEC 1953b AEC 1953a AEC 1953a

AEC AEC AEC AEC Harshaw AEC Harshaw AEC Harshaw AEC

Mar 

1950

Aug 

1950

Mar 

1951 Jan 1952

Jun 

1952 Jul 1952

Sep 

1952 Jan 1953

Feb 

1953

Aug 

1953

Old laboratory 0 3.6 8.3 8 3

Guardhouse 0 16 11 12 1

Maintenance Office 0 2.8 16 4 4

Maintenance area 0 5.6 7.7 11 2

Stockroom 6 191 29 3 5

Health Physics Office 147

Shipping & Receiving Office 4.2 4.3 27 21 1

Research & Development Office 0 30 22 8

Research & Development Laboratory 0 7 12 6

Breathing Zone Samples

Dump drum of U3O8 into digest hoppera 9,800 161 210 0.0 0.0 49 133 126 1,750 56

Clean empty drums after dumping 196

Provide makeup of nitric acid 140

Pour liquor into pots 63 42 63

Gulp UO3 from pots to collector 2,730 392 161 196 406 161 77 910 980 203

Clean pots, chip UO3 from paddle 43,400 (b)

Unload UO3 from pot into drum, break lumps 29,400 (b)

Shovel lumps into UO3 crusher 7,700 (b)

Feed orange from pots to Rockwell (3rd deck) 2,730

Feed orange from pots to Rockwell (vacuum system)

Discharge UO2 from Rockwell into can, weigh, store 31,500 16,520 17,220

Discharge drum UO3 into 30-gal drum, store 420 161 1,820 686 2,870 105 581

Shake the bag of the UO3 collector 14,980

Change drum under UO3 collector 3,570 616 1,540 350 2,030 4,270 2,380

Shake bag of floor sweepings 3,570

Clean and dump filter press 350 525 581 539 287 175 399 203 483 546

Grind UO3

a 140 119 21

Transfer UO3 from blender to sample jara 280

Transfer UO3 from sampling can to blendera 3,500

Transfer UO3 from blender to grindera 1,050

Fill UO3 into weighing bottles from grindera 3,500 <70 21

Determine density of UO3

a 133 <70 14

Clean blender with vacuum cleanera 378 56

Build Bird filter cake 182 119 49

Plow off filter cake 182 392 91

Load press cake into drums 42

a. The grinding and related operations were performed infrequently, per AEC (1953b). In AEC (1953a) and AEC (1953b), the UO3 was referred to as JH-6 or HGE, which appear to denote 

Hanford forms. It appears that most of the UO3 was "gulped" (removed by vacuum system) from the pots, but that lumps too large to suction had to be removed with scoops ("Unload UO3

from pot. ...").

b. These operations were new in January or February 1953, per AEC (1953a) and AEC (1953b).

c. These operations were supposed to have been discontinued prior to this survey, but per AEC (1951a), Harshaw still found it necessary to do some lump crushing.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 99 of 114

Table B-13. Dust measurements in the green and brown plants, in beta dpm/m3

.

a

Occupation Operation

Beta 

concentration, 

dpm/m3

Alpha 

concentration, 

dpm/m3

Ratio, beta 

to alpha

Brown oxide loader 

No. 1

Opening cartons of UO2, dumping into 

barrel

46,000 95,250 0.48

Brown oxide loader 

No. 2

Cleaning green dumping counter 59,000 122,500 0.48

Hex loader (new area) Dumping 300-lb barrel of UF4 into hopper of 

mechanical loader

232,000 397,000 0.58

Hex area loader Loading tray with UF4 134,000 207,440 0.65

Hex area loader Cleaning tray (ash and unreacted UF4) 57,000 23,300b 2.45

Hex area hood Pulling ice trap 39,000 56,000 0.70

a. Beta data are from AEC (1948). The corresponding alpha data is from Table B-10.

b. It appears that the alpha value was not averaged in the typed record, as can be seen from the original sample values 

given in the handwritten sample record. The correct average has therefore been put into the table above.

Table B-14. Daily DWEs to airborne dust, in alpha dpm/m3.

a

Hayden 1948 Klevin 1950a AEC 1950a AEC 1950a

491/492 493 493 493

May 1948 Mar 1950 Mar 1950 Aug 1950

Boildown tank and de-etherizer <70 <70

Ether extraction 42 <70 1.4

Bird centrifuge/filter 126 1.4 2.8

Niagara and frame filter 259 4.2 1.4

No. 1 Digester (batch) and nitric acid 

recovery

336 4.2 1.4

Nitric acid makeup area 140

Rockwell furnace 567 67 17

Denitration pots 112 84 210b

General 493 production area 546

Hoods, UF4-to-UF6 21,000

Control aisle, UF4-to-UF6 175

Still area 1,300

Recovery Room 36,400

Line cleaning area 1,000

Hoods, UO2-to-UF4 18,000

General, UO2-to-UF4 260

Clean laundry room (East) 14

Dirty laundry room (West) 126

Lunch/locker-shower rooms 150 56/49

Office 160

Production Office (493) 189

Health Physics Office 147

Average of daily weighted exposures 22 2.8

a. From the references, the data in this table represent daily weighted averages for partial time spent by a typical worker in 

an area, without identification of any job title. Thus, these data can apparently be used for workers with time split on a 

daily or weekly basis between areas in nonroutine ways, such as when substituting for a sick coworker.

b. Some major dusty parts of this operation are not included in this average.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 100 of 114

Table B-15. Daily DWEs to airborne dust for various job titles, in alpha dpm/m3.

a

Hayden 

1948

AEC 

1950b

AEC 

1950b

AEC 

1950b

Klevin 

1950a, 

AEC 

1950a

AEC 

1950b

AEC 

1950c, 

AEC 

1950e

AEC 

1951b

AEC 

1951a

AEC 

1951d

AEC 

1952a

Klevin 

1952b

AEC 

1952b

AEC 

1953a

AEC 

1953b

AEC 

1953a

AEC 

1953a

491/492 491/492 491/492 491/492 493 491/492 493 491/492 493 491/492 493 493 493 493 493 493 493

AEC AEC AEC AEC AEC AEC AEC AEC AEC AEC AEC Harshaw AEC Harshaw AEC Harshaw AEC

Operation or Job

May 

1948 Sep 1948

Feb 

1949

Sep 

1949

Mar 

1950

Jul 

1950

Aug 

1950

Feb 

1951

Mar 

1951

Sep 

1951

Jan 

1952 Jun 1952

Jul 

1952

Sep 

1952

Jan 

1953

Jun 

1953

Aug 

1953

Brown oxide loaders 11,480 11,480 5,880 9,800 2,660 2,800 1,540

Cleaners 987 294 1,729 70

Foremen and supervisors (491/492) 840 2,506 147 168 126

General foreman (491/492) 420 2,520 133 182 84

Hex area loaders 5,411 23,800 3,780 7,280 2,660 1,610 672

Hex area operators 1,050 308 5,600 196 175 196

Laboratory personnel 105 210 69 62 53.9

Process engineer (491/492) New 2,520

Reactor furnace (incl tube handlers) 245 1,190 98 539 189 203 112

Recovery operator 6,965 15,120 Not in 

operation

67 168 70

Shift foreman 140 175 147

Still operators 1,015 1,274 231 497 238 910 64.4

Analyst 4.9 46.2 23.1 16.1

Assistant superintendent (493) 399 63 21.7 11.2 7.7 9.8 11.9

Chief chemist 11.2 14 13.3 7.7

Chief/lead process man (493) 399 140 112 70 46.2 56 18.9 23.1 98 30.8

Foremen and supervisors (493) 2,520 147 105 55.3 25.2 7.7 2.8 13.3 35 6.3

General foreman (493) 399 67

Process engineer (493) 399 119 63 31.5 23.1 25.2 23.1 30.1 7.7

Process helper/man: ether extraction 49 84 32 60.2 33.6 7.7 18.2 13.3 35 7

Process helper/man: Niagara, frame 

filters

203 91 77

Process helper: Bird filter/centrifuge 140 98 44

Process helper: boildown tank, deetherizer49 49 98 73.5

Process helper: general 119 91 67.2 77 49.7 105 77

Process man: Bird centrifuge 203 140

Process man: boildown tank 91 56 77 53.9 49.7

Process man: denitration pots 4,200 175 105 105 98 91 30.8 189 189 58.8

Process man: digester, nitric acid 

recovery

210 70 52 20.3 44.1 43.4 16.8 35.7 301 39.9

Process man: Rockwell furnace 3,220 728 840

Research and development 2.8 14 13.3 7.7

Sampler (KoZ)b 651 9.8

Clerk 168 98 73 73 51.8

Guards 35 7 35 42 30.8

Health physicist 1,820 154 161 126

Laundry man 1,680 66 154 154 88 9.8 25.2 9.1 4.9

Maintenance and repair 210 910 217 196 84 20.3 23.1 21 7

Maintenance foreman New title 910 147 126 37.8

Office 182

Shipping and Receiving 133 2,450 62 93 168 35 35 30.8 9.1

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 25 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 101 of 114

Hayden 

1948

AEC 

1950b

AEC 

1950b

AEC 

1950b

Klevin 

1950a, 

AEC 

1950a

AEC 

1950b

AEC 

1950c, 

AEC 

1950e

AEC 

1951b

AEC 

1951a

AEC 

1951d

AEC 

1952a

Klevin 

1952b

AEC 

1952b

AEC 

1953a

AEC 

1953b

AEC 

1953a

AEC 

1953a

491/492 491/492 491/492 491/492 493 491/492 493 491/492 493 491/492 493 493 493 493 493 493 493

AEC AEC AEC AEC AEC AEC AEC AEC AEC AEC AEC Harshaw AEC Harshaw AEC Harshaw AEC

Operation or Job

May 

1948 Sep 1948

Feb 

1949

Sep 

1949

Mar 

1950

Jul 

1950

Aug 

1950

Feb 

1951

Mar 

1951

Sep 

1951

Jan 

1952 Jun 1952

Jul 

1952

Sep 

1952

Jan 

1953

Jun 

1953

Aug 

1953

Stock Room 84 18.9 9.1 4.9

Superintendent (491/492/493) 420 609 469 133 59 42 16.1 4.2

Average weighted exposures 4,690 2,730 2,380 1,155 441 182 560 154 266 63.7 51.1 42.7 23.1 45.5 60.2 22.4

a. Improvements were made between September 1948 and February 1949, hence the change in levels for certain jobs (AEC 1948, 1949b).

b. The job of sampler was new as of January 1952.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 26 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 102 of 114

Table B-16. Annual inhalation intakes based on daily 

weighted average exposures to airborne dust for various job 

titles, in pCi. 491/492 (UF4/UF6) workers 1942-1947 1948 1949

Brown oxide loaders 1.25E+07 1.25E+07 8.55E+06

Cleaners 1.08E+06 1.08E+06 1.10E+06

Foremen and supervisors (491/492) 9.16E+05 9.16E+05 2.73E+06

General foreman (491/492) 4.58E+05 4.58E+05 2.75E+06

Hex area loaders 5.90E+06 2.60E+07 6.03E+06

Hex area operators 1.15E+06 1.15E+06 3.22E+06

Hex central loadersa --- --- ---

Laboratory personnel 1.15E+05 1.15E+05 2.29E+05

Process engineer (491/492) --- 2.75E+06 2.75E+06

Reactor furnace (incl tube handlers) (hex) 2.67E+05 1.30E+06 3.48E+05

Recovery operator 7.60E+06 1.65E+07 8.25E+06

Shift foreman 8.18E+05 8.18E+05 2.30E+06

Still operators 1.11E+06 1.39E+06 3.97E+05

Generic 491 (UF4) process worker 1.25E+07 1.25E+07 8.55E+06

Generic 492 (UF6) process worker 3.20E+06 9.26E+06 3.65E+06

493 (UO2/UO3) workers 1949b

Analyst ---

Assistant superintendent (493) 4.35E+05

Chief chemist 2.52E+05

Chief/lead process man (493) 4.35E+05

Foremen and supervisors (493) 2.75E+06

General foreman (493) 4.35E+05

Process engineer (493) 4.35E+05

Process helper/man: ether extraction 5.35E+04

Process helper/man: Niagara/frame filters 2.21E+05

Process helper: Bird filter/centrifuge 1.53E+05

Process helper/man: boildown tank/de-etheriz. 5.35E+04

Process helper: general 5.03E+06

Process man: Bird centrifuge 2.21E+05

Process man: denitration pots 4.58E+06

Process man: digester, nitric acid recovery 2.29E+05

Process man: Rockwell furnace 3.51E+06

Research and development ---

Sampler (KoZ) ---

Generic 493 (UO2/UO3) process worker 1.45E+06

Support and management workersc 1942-1947 1948 1949

Clerk (production) 1.83E+05 1.83E+05 1.07E+05

Guard 3.82E+04 3.82E+04 7.64E+03

Health physicist 4.80E+05 4.80E+05 2.87E+06

Laundry man --- --- 1.83E+06

Maintenance and repair 2.29E+05 2.29E+05 9.93E+05

Maintenance foreman 2.29E+05 2.29E+05 9.93E+05

Office 1.78E+05 1.78E+05 2.57E+05

Shipping and Receiving 1.45E+05 1.45E+05 2.67E+06

Stock Room 6.44E+04 6.44E+04 1.18E+06

Superintendent (491/492/493) 4.58E+05 4.58E+05 6.64E+05

Overall Average Weighted Exposures 1942-1947 1948 1949

491/492 basis 2.53E+06 5.12E+06 2.79E+06

493 basis --- --- ---

a. This job began in mid-1950, so this figure should be applied during only the 

second half of 1950.

b. UO2/UO3 plant production started in May 1949, so this column is applicable 

only for May–December 1949.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 27 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 103 of 114

Table B-17. Annual ingestion intakes based on daily 

weighted average exposures to airborne dust for various job 

titles, in pCi. 491/492 (UF4/UF6) workers 1942-1947 1948 1949

Brown oxide loaders 2.63E+05 2.63E+05 1.80E+05

Cleaners 2.27E+04 2.27E+04 2.31E+04

Foremen and supervisors (491/492) 1.92E+04 1.92E+04 5.73E+04

General foreman (491/492) 9.62E+03 9.62E+03 5.78E+04

Hex area loaders 1.24E+05 5.46E+05 1.27E+05

Hex area operators 2.42E+04 2.42E+04 6.76E+04

Hex central loadersa --- --- ---

Laboratory personnel 2.42E+03 2.42E+03 4.81E+03

Process engineer (491/492) --- 5.78E+04 5.78E+04

Reactor furnace (incl tube handlers)(hex) 5.61E+03 2.73E+04 7.31E+03

Recovery operator 1.60E+05 3.47E+05 1.73E+05

Shift foreman 1.72E+04 1.72E+04 4.83E+04

Still operators 2.33E+04 2.92E+04 8.34E+03

Generic 491 (UF4) process worker 2.63E+05 2.63E+05 1.80E+05

Generic 492 (UF6) process worker 6.72E+04 1.94E+05 7.67E+04

493 (UO2/UO3) workers 1949b

Analyst ---

Assistant superintendent (493) 9.14E+03

Chief chemist 5.29E+03

Chief/lead process man (493) 9.14E+03

Foremen and supervisors (493) 5.78E+04

General foreman (493) 9.14E+03

Process engineer (493) 9.14E+03

Process helper/man: ether extraction 1.12E+03

Process helper/man: Niagara/frame filters 4.64E+03

Process helper: Bird filter/centrifuge 3.21E+03

Process helper/man: boildown tank/de-etheriz. 1.12E+03

Process helper: general 1.06E+05

Process man: Bird centrifuge 4.64E+03

Process man: denitration pots 9.62E+04

Process man: digester, nitric acid recovery 4.81E+03

Process man: Rockwell furnace 7.37E+04

Research and development ---

Sampler (KoZ) ---

Generic 493 (UO2/UO3) process worker 3.05E+04

Support and Management Workersc 1942-1947 1948 1949

Clerk (production) 3.84E+03 3.84E+03 2.25E+03

Guard 8.02E+02 8.02E+02 1.60E+02

Health physicist 1.01E+04 1.01E+04 6.03E+04

Laundry man --- --- 3.84E+04

Maintenance and repair 4.81E+03 4.81E+03 2.09E+04

Maintenance foreman 4.81E+03 4.81E+03 2.09E+04

Office 3.74E+03 3.74E+03 5.40E+03

Shipping and Receiving 3.05E+03 3.05E+03 5.61E+04

Stock Room 1.35E+03 1.35E+03 2.48E+04

Superintendent (491/492/493) 9.62E+03 9.62E+03 1.39E+04

Overall Average Weighted Exposures 1942-1947 1948 1949

491/492 basis 5.31E+04 1.08E+05 5.86E+04

493 basis --- --- ---

a. This job began in mid-1950, so this figure should be applied during only the second half 

of 1950.

b. UO2/UO3 plant production started in May 1949, so this column is applicable only for 

May–December 1949.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 28 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 104 of 114

Table B-18. Annual radon exposure, in WLM/yr.

a

Area or worker type

Main work area Break and other low-exposure areas

Total, 

pCi/L Rn Eq factor WLM/yr

Occ 

factor WLM/yr pCi/L Rn Eq factor

Occ 

factor WLM/yr

Ore Storage/Warehouse 1.0 1.0 0.75 0.09 1.0 0.4 0.25 0.01 0.10

Digest/Feed 3.0 0.5 0.75 0.14 1.0 0.4 0.25 0.01 0.15

Extraction Cells 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Centrifuge 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Feinc/Filter/Raffinate/Niagara 7.0 0.5 0.75 0.32 1.0 0.4 0.25 0.01 0.33

Pot Room 2.0 0.5 0.75 0.09 1.0 0.4 0.25 0.01 0.10

UO3 packing area 12 0.5 0.75 0.54 1.0 0.4 0.25 0.01 0.55

UO3-UO2 areab 2.0 0.5 0.75 0.09 1.0 0.4 0.25 0.01 0.10

Green (UO2-UF4) areab 2.0 0.5 0.75 0.09 1.0 0.4 0.25 0.01 0.10

Hex (UF4-UF6) areab 2.0 0.5 0.75 0.09 1.0 0.4 0.25 0.01 0.10

Distillation (UF6) and recoveryb 2.0 0.5 0.75 0.09 1.0 0.4 0.25 0.01 0.10

Decontamination (normal oper. cleanup) 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Nitric acid storage area 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Ether storage area 2.0 0.5 0.75 0.09 1.0 0.4 0.25 0.01 0.10

Shipping & Receiving (non-ore) 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Welding Shop 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Millwright Shop 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Electrical Shop 1.0 0.5 0.75 0.05 1.0 0.4 0.25 0.01 0.06

Maintenance Shop 4.0 0.5 0.75 0.18 1.0 0.4 0.25 0.01 0.19

Shotgun Lab (UO3 assay) 4.0 0.5 0.88 0.21 1.0 0.4 0.12 0.01 0.22

Other Lab (Research, Analytical, etc.) 1.0 0.5 0.88 0.05 1.0 0.4 0.12 0.01 0.06

Office workers 1.0 0.4 0.88 0.04 1.0 0.4 0.12 0.01 0.05

Yards and other outdoor areas 1.0 --- --- --- 1.0 --- --- --- ---

a. See Section 5.5 for assumptions for this table.

b. Exposures in these areas were assumed to be bounded by those in the Pot Room.

Table B-19. General area dust concentrations in the locker rooms and lunchrooms, in 

alpha dpm/m3

.

Survey

Locker rooms Lunchrooms

1 2 3 1 2 3

AEC, 9/1948 (AEC 1948) 411 84 15 126 40 30

Harshaw(?), 2-5/1949 (AEC 1950d) 70 42 77 42

AEC, 9/1949 (AEC 1950d) 49 63 63 56 91 21

Harshaw, 11/1949: Position 1/19C (Klevin 

1950a)

78 51 194 35

Harshaw, 11/1949: Position 2/19D (Klevin 

1950a)

109 72

Harshaw, 11/1949: Position 3 (Klevin 1950a) 37

AEC, 3/1950 (Klevin 1950a) 50 50 34 55 55 21

AEC, 7/1950 (AEC 1950c) 56 63 119 42 70 70

AEC, 8/1950(AEC 1950e) 21 7

AEC, 1/1951 (Blatz 1950a,b,c) 329 21 84 91 28 14

AEC, 3/1951 (AEC 1951a) 84 42

AEC, 9/1951 (AEC 1951d) 98 35 35 56 70 7

AEC, 1/1952 (AEC 1952a) 30 41 27 63 32 7

Harshaw, 6/1952 (AEC 1952b) 4 15 0 7 19 3

AEC, 7/1952 (AEC 1952b) 40 17 1.5 17 18 0

Harshaw, 9/1952 (AEC 1953b) 2.6 11 6.3 3 12 3

AEC, 1/1953 (AEC 1953b): Position 1/C 31 21 22 35 2.8 11

AEC, 1/1953 (AEC 1953b): Position 2/D 64 22 0 35 2.8 11

Harshaw, 2/1953 (AEC 1953a) 19 23 6 28 33 23

AEC, 8/1953 (AEC 1953a) 2 51 6 2 8 18

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 29 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 105 of 114

Table B-20. Average measured dust and urine concentrations by month (Lippmann 1958).a

Soluble uranium: UF4-to-UF6 Insoluble uranium: UO2-to-UF4

Hex operators Still operators UO2 loaders/UF4 unloaders Reactor furnace operatorsb

Air sample Air sample Air sample Urine sample Air sample Urine sample

Month Conc (μg/m3) Conc (μg/m3) Conc (μg/m3) B, μg/L A, μg/L B/A ratio Conc (μg/m3) B, μg/L A, μg/L B/A ratio

1950

January 440 620 9,700 160 68 2.4 170

February 370 640 110 40 2.8 150 49 18 2.7

March 9,700 120 48 2.5 110 35 3.1

April 170 420 9,500 140 33 4.2 130 7 18.6

May 200 40 5.0 71 13 5.5

June 200 320 8,100 150 70 2.1 230 67 21 3.2

July 140 170 4,400 210 45 4.7 110 65 22 3.0

August 260 3,700 50 28 1.8 33 18 1.8

September 45 8 5.6

October 170 2,600 81 49 1.7 140 30 17 1.8

November 470 160 2,500 61 35 1.7 140 30 17 1.8

December 81 310 2,800 51 15 3.4 180 21 11 1.9

1951

January 88 500 2,200 40 20 2.0 160 40 8 5.0

February 180 940 2,600 42 21 2.0 110 30 13 2.3

March 300 160 3,600 49 29 1.7 120 12 6 2.0

April 130 120 3,600 49 29 1.7 120 12 6 2.0

May 240 1,800 54 25 2.2 53 15 3.5

June 140 90 1,800 54 25 2.2 200 75 22 3.4

July 1,300 50 12 4.2

August 100 250 18 67 11 6.1

September 130 150 1,300 21 88 18

October 61 91

November UF4 production shut down in mid-October 1951

December 160

Summary data

Median 160 250 2,800 61 29 2.2 140 49 14 3.1

Std deviation 0.58 0.70 0.67 0.58 0.43 0.36 0.27 0.65 0.48 0.59

Sample periods 17 17 17 17 19 17 13 19 20 19

a. Each data point is the average of worker samples. B = before-weekend sample, usually Friday; A = after-weekend sample (i.e., 

Monday morning). Statistical analysis of this data was done for this site profile. A lognormal distribution was assumed.

b. The reactor furnace operators were the process men in the green (UO2-to-UF4) area.

Table B-21. Number of workers. Date Plant No. of workersa Reference 

September 1948 Green/Hex 100 AEC 1948

September 1949 Green/Hex 121 AEC 1950d

July 1950 Green/Hex 131 AEC 1950c

January 1951 Green/Hex 127 AEC 1951b

September 1951 Green/Hex 125 AEC 1951d

March 1950 Brown 42 Klevin 1950a

August 1950 Brown 33 AEC 1950e

March 1951 Brown 39 AEC 1951a

January 1952 Brown (Orange) 34 AEC 1952a

July 1952 Brown (Orange) 34 AEC 1952b

September 1952 Brown (Orange) 68 AEC 1953b

January 1953 Brown (Orange) 69 AEC 1953b

February–June 1953 Brown (Orange) 69 AEC 1953a

August 1953 Brown (Orange) 69 AEC 1953a

May 1954 Brown (Orange) 5 Klevin 1954

a. Figures shown represent all workers dedicated to the plant listed, including management 

and clerks, but do not include support workers from other (non-AEC/MED) areas of the 

Harshaw site.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 30 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 106 of 114

Table B-22. Job titles, functions, and appropriate absorption types.

Job title or classification Notes

Absorption types 

appropriate for 

work area

491 Area (UO2-to-UF4, or green plant)

2nd-level tray loaders Load UO2 onto trays and the trays into tubes; empty trays of UF4 M, S

2nd-level tube handlers Move tubes of UO2 to the UO2-to-UF4 furnace and to rack M, S

Brown (oxide) loader Load UO2 onto trays and the trays into tubes M, S

Chief process man Control acid flow, check process and leaks, remove hot tubes from furnace; 

may include the term Brown

M, S

Cleaner General area cleanup M, S

Foreman May include the term Green, Maintenance, or Assistant M, S

General foreman For both 491 and 492 F, M, S,

Green area (worker) Generic for process man, operator, or the like; use geometry values for the 

appropriate such title or the generic values at right

M, S

HL-7 (worker) See Green area (worker) M, S

Loader Loaded UO2 onto trays and the trays into tubes M, S

Process man, PM Assist chief process man M, S

Reactor (furnace) operator UO2-UF4 reactor tender M, S

Reaction operator See Reactor (furnace) operator M, S

Recovery operator Fume recovery room/turbosaturator recovery room F, M, S

Shift foreman M, S

Supervisor Might include the term green M, S

Tray loader Load UO2 onto trays and the trays into tubes; empty trays of UF4 M, S

Tube handler Moved tubes of UO2 to the UO2-to-UF4 furnace and to rack M, S

WE-61 Function unclear M, S

492 Area (UF4-to-UF6 and UF6 distilling, or hex plant)

1st-Level loader Loader of UF4 in the UF4-to-UF6 process F, M

Chief process man, CPM Check cells and all hex operations, add acid F, M

Cleaner Cleaned areas and respirators F, M

Cyl cleaning Cylinder cleaning: normally done by Shipping and Receiving, but possibly 

done by others; see Shipping and Receiving or appropriate other title

F, M

Engineer Plant, process, production, dev(elopment); apparently not same as 493 eng F, M, S

Foreman Might include the term hex, maintenance, or assistant F, M

Fume recovery room operator See Recovery operator F, M, S

General foreman For both 491 and 492 F, M, S

Hex area (central) loader Loader of UF4 in the UF4-to-UF6 process, newer loading area F, M

Hex area operator Loader of UF4 in the UF4-to-UF6 process F, M

Laborer Might include the term maintenance or recovery F, M, S

Loader Load trays of UF4, put into furnaces, replace full receivers, clean ice traps F, M

Meter reader F, M

Office, office clerk F, M, S

Operator Same as process man? F, M

Process helper, PH Connect and disconnect lines, unplug (clear) lines F, M

Process man, PM Assist chief process man, connect lines, unplug (clear) lines F, M

Recovery operator Fume recovery room/turbosaturator recovery room F, M

Shift foreman F, M

Still operator UF6 distillation process F, M

Supervisor Might include the term hex F, M

Tray loader Load UF4 into trays, load trays into hex reactors, unload trays, receivers F, M

Welder Contract welding F, M

493 Area (ore-to-UO2 and ore-to-UO3, or brown plant)

Analyst Lab work M, S

Batch makeup operator Add ore and nitric acid to digester M, S

Brown area (worker) Generic for process man, process helper, or the like M, S

Chemist Lab work M, S

Chief chemist Lab work M, S

Chief process man, CPM Work at the TBP panel board and tanks, change drums and bags M, S

Engineer, process Apparently separate from the 492 engineer M, S

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 31 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 107 of 114

Job title or classification Notes

Absorption types 

appropriate for 

work area

493 Area (continued)

Foreman Might include the term brown, maintenance, or assistant M, S

Laboratory material handling Help dump presses, make up caustic solution, handle feed material for 

batch makeup

M, S

Lead man See chief process man M, S

PH Process helper M, S

Process helper: Bird filter/centrifuge Run centrifuge, draw sample and weigh, pump liquor to tanks M, S

Process helper: boildown tank & deetherizersSend filtrate from Niagaras to extraction tank, de-etherize hex liquor, pump 

to concentrators for second boildown; send concentrate to pots

M, S

Process helper: filter/filter press Same as process helper: Niagara filters? M, S

Process helper: Niagara filters Precoat, filter batches through Niagaras, pump filtrate to boildown tanks, 

dump cake into process tanks, clean filter; might include the term frame filter

M, S

Process man: batch makeup, digester, 

acid recovery

Make up batches of feed and acid, check digestion instruments, check 

recovery towers, pump acid from towers

M, S

Process man: denitrating pots Fill pots with hex liquor, check furnace instruments, remove, delump UO3 M, S

Process man: extractor (ether) Measure hex liquid, add ether; add water layer; transfer washings; reextract 

liquor, check instruments and ether storage

M, S

Process man: filter/filter press Similar to process helper: Niagara filters? M, S

Process man, PM Similar to process helper M, S

Process man: Rockwell Keep feed hoppers filled, check feed table, package product in drums, 

check and record temperatures

M, S

Process man: tank See process man: boildown tank M, S

Recovery operator Recovery of nitric acid; also cleaned cake off filters M, S

Research and development Work in R&D (chemistry and process) lab, office areas M, S

Sampler Lab prep: UO3, HGE, and JH-6 grinding and blending, weigh JH-6 and UO3, 

cold-press UO3 in die; might contain the term KOZ

M, S

Spec main Special maintenance? M, S

Statistician Assay work M, S

Stock room M, S

Supervisor May include the term brown M, S

Generic or All Areas or "490"

Accounting clerk In Plant C office, but not in the plant itself F, M, S

AEC Employee of AEC F, M, S

Analyst Chemical analyses? F, M, S

Captain of guards For all of Harshaw site; less than 1/3 of time spent in Plant C F, M, S

Chemist F, M, S

Clerk For production clerks, some production area access F, M, S

Dev maint Unclear: might be maintenance in pilot plant and lab areas F, M, S

Electrician F, M, S

Engineer Less than 1/3 of time in plant F, M, S

File clerk In the Plant C office, but not in the plant itself F, M, S

Fork truck operator F, M, S

Guard Less than 1/3 of time in plant F, M, S

Health physicist Health physics office was in the 493 area, but served all areas F, M, S

Instrument man Presumably serviceman for instruments F, M, S

Janitor If office janitor, did not enter plant F, M, S

Laboratory assistant F, M, S

Laboratory personnel Generic F, M, S

Laborer Labor, maintenance, recovery F, M, S

Laundry man The two laundries were in the 493 area, but served all areas F, M, S

Maintenance and repair Worked out of maint. office, served all areas; might include term Devel(op) F, M, S

Maintenance foreman Worked out of the maintenance office, served all areas F, M, S

Mechanic F, M, S

Medical technician Not in plant F, M, S

Miscellaneous and unclassified Unclear what was meant by this AEC-used term F, M, S

Office Generic for worker in office; might or might not have access to process 

areas

F, M, S

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 32 of 39
Document No. ORAUT-TKBS-0022 Revision No. 00 Effective Date: 08/17/2007 Page 108 of 114

Job title or classification Notes

Absorption types 

appropriate for 

work area

Generic (continued)

Office manager Less than 1/3 of time in plant F, M, S

Operator Generic usage when otherwise unclear -- see "process man" and like titles F, M, S

Plant clerk Same as production clerk? Less than 1/3 of time in plant F, M, S

Plant manager Generally in charge of all Plant C areas, although at one time there 

appeared to be different 491/492 and 493 managers

F, M, S

Porter Unclear if there was production area access F, M, S

Production clerk Some plant access F, M, S

Production office The Production Office was in the 493 area F, M, S

Property clerk Less than 1/3 of time in plant F, M, S

Research and development Working in the R&D lab, apparently in the 493 area F, M, S

Shipping and receiving Besides handling materials, washed the UF6 cylinders F, M, S

Security See Guard F, M, S

Spec maint Special maintenance, assumed to be the same as maintenance F, M, S

Stenographer In office, but not in the plant F, M, S

Stockroom (worker) Apparently handled uranium forms or spent time in or near process areas F, M, S

Sweeper Assumed to be the same as Cleaner (see 492) F, M, S

Superintendent See plant manager F, M, S

Table B-23. Results of a survey of the Harshaw site by Argonne National Laboratory, 1976 to 1979

(DOE 1984).

Building Area

Max (avg high)

alpha 

(dpm/100 

cm2

)

a

Typ max 

alpha 

(dpm/100 

cm2

)

Max b-g 

(dpm/100 

cm2)

Typ max 

b-g 

(dpm/100 

cm2)

Max 

contact

b-g 

(mR/hr)

Avg max 

contact

b-g 

(mR/hr)

a

Avg 

contact

b-g 

(mR/hr)

a

Max 3-ft 

b-g 

(mR/hr)

c

Radon 

(pCi/L)

Radon,

10-3

WLb

Plant C 1st floor, new --- --- 20,000 --- --- --- --- --- --- ---

1st floor, main 21,000 (13,000) 1,300 1,100,000 455,000 3 1.8 0.16 0.1 --- ---

2nd floor 30,000 7,360 1,100,000 870,000 1.4 1.2 0.6 0.07 --- ---

Unspecified --- --- --- --- --- --- --- --- .25 - .69 2.5 - 6.9

Boiler house 300 0 150,000 0 3 0.0 0 --- 0.47 4.7

Foundry 13,333 1,300 333,333 0.3 3 2.3 0.3 --- .17 - .56 1.7 - 5.6

Garage 0 0 0 0 0 0.0 0 --- 0.37 3.7

Warehouse Lab and office 1,000 1000 6,000 0 0.05 0.05 0 --- .33 - .60 3.3 - 6.0

Second floor 0 0 2,000 0 0.02 0.02 0 --- .33 - .60 3.3 - 6.0

H-1 0 0 0 0 0.03 0 0 --- 0.5 5

H-2 0 0 0 0 0.03 0 0 --- 0.59 5.9

K-1 10,000 1,500 50,000 17,000 1.5 0.80 0.4 --- .11 - .57 1.1 - 5.7

M-1 10 0 50,000 12,000 0.5 0.07 0.0 --- 0.35 3.5

P-1 1st floor 25,000 7,150 200,000 67,000 0.7 0.6 0.2 --- 0.33 3.3

2nd floor 0 0 20,000 6,400 0.2 0.2 0.07 --- 0.12 1.2

a. The "average" designation above indicates the average of the reported measurements for spots where at least one type of reading was 

greater than background (as interpreted from DOE 1984). The vast majority of measurements were at background and thus the true 

average was very much lower than is indicated here.

b. This assumes radon daughter equilibrium (DOE 1984). The U.S. Environmental Protection Agency limit at the time was given as 0.02 

WL (DOE 1984).

c. The ambient penetrating radiation level (by Eberline PRM-7 μR meter) was found to be 40 times less than the contact Geiger-Mueller

radiation level (DOE 1984).

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Table B-24. Source terms used to calculate inhalation and radon doses, D&D/postoperations period.

a

Building

Dose 

potential

Alpha, dpm/

100 cm2

Principal 

reference Building

Dose 

potential

Alpha, dpm/ 

100 cm2

Principal 

referencea

AEC decontamination Postops, predecontamination; Post-AEC decontamination

Old UF6 area (492) Low 2.00E+03 Schoen 1958 Boiler house Low 0.00E+00 DOE 1984

(excl. Recov, Still) Moderate 5.90E+03 Klevin 1955a Moderate 1.50E+02 DOE 1984

High 6.60E+04 As per Recovery High 3.00E+02 DOE 1984

Old UF6 area (492) Low 2.00E+03 Schoen 1958 Foundry Low 0.00E+00 DOE 1984

(Recov, Still only) Moderate 1.35E+04 Schoen 1958 Moderate 1.30E+03 DOE 1984

High 1.50E+05 Schoen 1958 High 1.30E+04 DOE 1984

New UF6 area (annex) Low 2.00E+03 Schoen 1958 Garage Low 0.00E+00 DOE 1984

(492/493) Moderate 9.00E+03 Schoen 1958 Moderate 1.50E+02 DOE 1984

High 2.60E+04 Schoen 1958 High 3.00E+02 DOE 1984

UO2/UO3 area (493) Low 2.00E+03 Schoen 1958 Warehouse, incl. lab, Low 0.00E+00 DOE 1984

(incl. Pilot Plant) Moderate 7.25E+03 Schoen 1958 offices Moderate 5.00E+02 DOE 1984

High 2.50E+04 Schoen 1958 High 1.00E+03 DOE 1984

UF4 area (491) Low 2.00E+03 Schoen 1958 P-1 Low 0.00E+00 DOE 1984

Moderate 2.00E+04 Schoen 1958 Moderate 7.20E+03 DOE 1984

High 1.25E+05 Schoen 1958 High 2.50E+04 DOE 1984

Laboratory area --- 5.00E+03 Schoen 1958 K-1 Low 0.00E+00 DOE 1984

Locker area --- 2.00E+03 Schoen 1958 Moderate 1.50E+03 DOE 1984

Postdecontamination High 1.00E+04 DOE 1984

Building G-1: 1st floor Low 0.00E+00 DOE 1984 M-1 Low 0.00E+00 DOE 1984

Moderate 1.30E+03 DOE 1984 Moderate 5.00E+00 DOE 1984

High 2.10E+04 DOE 1984 High 1.00E+01 DOE 1984

Building G-1: 2nd floor Low 0.00E+00 DOE 1984

Moderate 7.40E+03 DOE 1984

High 3.00E+04 DOE 1984

a. FUSRAP (2001) was another source of information in some cases.

Table B-25. Annual inhalation, radon, and ingestion doses, D&D/postoperations period.

a

AEC decontamination

(Building G-1)

1952-1959 1955-1959

UF4 area Old UF6 area

Recovery/

still area

New UF6

area

UO3/UO2

areas Lab area Locker area

Inhalation Min 1.33E+02 1.53E+02 1.53E+02 1.53E+02 2.43E+02 --- ---

 (pCi/yr) Mode 1.33E+03 4.51E+02 1.03E+03 6.87E+02 8.83E+02 6.09E+02 1.53E+02

Max 8.31E+03 5.04E+03 1.15E+04 1.99E+03 3.04E+03 --- ---

Radon Min 4.12E-03 4.10E-03 4.10E-03 4.10E-03 4.04E-03 --- ---

 (WLM/yr) Mode 4.12E-02 1.21E-02 2.77E-02 1.85E-02 1.46E-02 1.01E-02 4.10E-03

Max 2.57E-01 1.35E-01 3.08E-01 5.33E-02 5.05E-02 --- ---

Ingestion Min 2.79E+00 3.21E+00 3.21E+00 3.21E+00 5.10E+00 --- ---

 (pCi/yr) Mode 2.79E+01 9.47E+00 2.16E+01 1.44E+01 1.85E+01 1.28E+01 3.21E+00

Max 1.75E+02 1.06E+02 2.42E+02 4.18E+01 6.38E+01 --- ---

Postdecontamination

1960-1997

Building G-1,

1st floor

Building G-1,

2nd floor

Inhalation Min 0.00E+00 0.00E+00

 (pCi/yr) Mode 6.55E+00 3.73E+01

Max 1.06E+02 1.51E+02 1st floor: UF6/ UO3 area (in general)

Radon Min 0.00E+00 0.00E+00 2nd floor: UF4/ UO2 area (in general)

 (WLM/yr) Mode 2.97E-03 1.69E-02

Max 4.80E-02 6.86E-02

Ingestion Min 0.00E+00 0.00E+00

 (pCi/yr) Mode 1.38E-01 7.83E-01

Max 2.23E+00 3.17E+00

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Post-AEC operations, 

Predecontamination

1955-1989 1959-1989

Boiler house Garage Warehouse Bldg M-1 Foundry Bldg P-1 Bldg K-1

Inhalation Min 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

 (pCi/yr) Mode 2.46E+00 2.46E+00 8.21E+00 8.21E-02 2.13E+01 1.18E+02 2.46E+01

Max 4.92E+00 4.92E+00 1.64E+01 1.64E-01 2.13E+02 4.10E+02 1.64E+02

Radon Min 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

 (WLM/yr) Mode 3.13E-04 3.13E-04 1.04E-03 1.04E-05 2.71E-03 1.50E-02 3.13E-03

Max 6.27E-04 6.27E-04 2.09E-03 2.09E-05 2.71E-02 5.22E-02 2.09E-02

Ingestion Min 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

 (pCi/yr) Mode 5.17E-02 5.17E-02 1.72E-01 1.72E-03 4.47E-01 2.48E+00 5.17E-01

Max 1.03E-01 1.03E-01 3.44E-01 3.44E-03 4.47E+00 8.61E+00 3.44E+00

Post-AEC 

decontamination

1990-1992 1990-1992

Boiler house Garage Warehouse Bldg M-1 Foundry Bldg P-1 Bldg K-1

Inhalation Min 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

 (pCi/yr) Mode 3.03E+01 3.03E+01 1.01E+02 1.01E+00 2.63E+02 1.46E+03 3.03E+02

Max 6.05E+01 6.05E+01 2.02E+02 2.02E+00 2.63E+03 5.05E+03 2.02E+03

Radon Min 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

 (WLM/yr) Mode 3.03E-04 3.03E-04 1.01E-03 1.01E-05 2.62E-03 1.45E-02 3.03E-03

Max 6.06E-04 6.06E-04 2.02E-03 2.02E-05 2.62E-02 5.05E-02 2.02E-02

Ingestion Min 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

 (pCi/yr) Mode 6.36E-01 6.36E-01 2.12E+00 2.12E-02 5.52E+00 3.07E+01 6.36E+00

Max 1.27E+00 1.27E+00 4.24E+00 4.24E-02 5.52E+01 1.06E+02 4.24E+01

a. Data in the table above were calculated using RESRAD-BUILD from information in Schoen (1958), Klevin (1955a), DOE (1984), and FUSRAP 

(2001). The periods of application above are based on the AEC-overseen decontamination of Plant C (UF4, UF6, and related areas, 1952–

1959; UO3/UO2 and related areas, 1955–1959); the subsequent years of use of Plant C (1960–1997) before FUSRAP agreed to recharacterize 

the radiological status of the plant; the years of post-AEC use of the other buildings, without significant decontamination (1955–1989); and the 

company-sponsored decontamination of some buildings, not under AEC/U.S. Energy Research and Development Administration oversight but 

done to NRC standards (1990-1992). See text below regarding assumptions made, etc.

The principal assumptions made for the RESRAD-BUILD calculations (Table B-25) are given below.

1. The inhalation and radon source terms were derived on the basis of the average maximum, 

high average, and typical surface contamination levels, respectively, in each plant area or 

building regardless of surface location. However, ceiling and overhead levels were not 

considered because the overheads were generally found to be far less contaminated than the 

walls and floor and were likely not to be frequently accessed; thus, they would contribute 

negligibly to the total dose. The three surface contamination levels used for each plant area or 

building are given as "maximum,0" "mode" (average of the high readings), and "minimum" 

(typical; i.e., the median for all the readings), with one case run for each. The room model was 

assumed to have the respective surface concentration over all wall and floor surfaces, which 

were taken to be concrete. These sources are listed in Table B-24.

2. Source contamination was measured as gross alpha and as either total beta and total gamma 

separately or total beta-gamma together, depending on the survey. Because the degree of 

equilibrium of the uranium with its daughters was not known, it was assumed for the inhalation 

case that there was 100% equilibrium; this appears to be reasonable because of the length of 

time between the cessation of operations and the beginning of intrusive decontamination 

operations. The source terms were then determined by assuming that the alphas were being 

emitted by 238U, 234U, 230Th, and 226Ra (together yielding 97.8% of the total alpha emissions

and by 235U, 231Pa, and 227Ac (together yielding 2.2% of the total alpha

3. The breathing rate was taken to be 1.2 m3/hr.

emissions).

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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4. The worker was assumed to spend his entire work time (8 hr/workday) in the modeled room

(i.e., in the contaminated work area of the given plant area or building). The worker was 

assumed to spend 2,000 hr/yr in one location. The takedown of a building might have been on 

the order of weeks and decontamination on the order of months; however, continuous 

decontamination and demolition work over the course of a 50-week year (2,000 hours) was 

assumed for such workers.

5. The room size was taken to be about 10 by 20 by 10 ft high (3 by 6 by 3 m). There were many 

process areas that were larger, but they were often partitioned and they were undoubtedly 

decontaminated in sections. Thus, assuming a smaller room would be conservative in terms 

of concentrating or confining the contamination in the ventilated space and, therefore,

exposure estimates would be expected to be favorable to claimants.

6. One air change per hour was assumed. While only limited information is available regarding 

the ventilation systems at Harshaw, it was clear that process areas had forced ventilation. 

These were perhaps not always used during significantly dusty work (e.g., checking ducting), 

but in those cases respirators were likely to have been worn by the workers. Thus, it is 

reasonable to assume that either the normal forced ventilation was used in the general area, in 

which case one air change per hour is a rate favorable to claimants, or vented enclosures 

were used, in which case the air change rate would have been far higher. The worker would 

likely have been wearing a respirator, and the calculated intakes would represent a marked 

overestimate of the likely actual intakes.

7. For the period that decontamination took place, the resuspension factor for the transferable 

contamination was assumed to be 1 × 10-4

; for the postdecontamination years case, it was 

conservatively assumed to be 1 × 10-6

. The latter value is based on NRC (2002) and the 

former is taken to be a value favorable to claimants for nonrespirator work, as is consistent 

with the discussion and tables in the RESRAD-BUILD manual (ANL 2003).

8. The deposition (settling) velocity was taken to be 0.00075 m/s, a reasonable value for 5-μm 

sized particles, as shown in Figure J.3 of ANL (2003).

9. The removable fraction for the decontamination years case was assumed to be 30%, based 

on the fact that some early decontamination was done at the end of operations (rinsing out 

process vessels, cleaning floors, etc.). The removable fraction for the postdecontamination 

period was assumed to be 10%. This should be reasonable because the postdecontamination 

period followed an extensive decontamination. Default erosion, radon emanation, and 

associated values were used because they are favorable to claimants.

10. The removable/erosion portion of the source was assumed to be removed completely over the 

modeled time, linearly. However, the fraction of radon assumed to be released to the air (as it 

was formed) was 1.0.

11. Because the inhalation and radon calculations did not depend on the position of the receptor in 

the room model and because isotope proportions were taken to be the same at the beginning 

of the calculation, one wall of the maximally contaminated plant was modeled with the D&D 

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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sources in RESRAD-BUILD; the same was done with the post-D&D sources. The results were 

then ratioed for the entire wall and floor area and for other area and building cases.

For use in dose reconstruction, the annual inhalation and radon doses that were the results of 

computations in RESRAD-BUILD had to be converted back to activity units, in this case to picocuries 

and WLM of intake, respectively. The RESRAD family of codes uses the dose conversion factors for 

inhalation given in Eckerman, Wolbarst, and Richardson (1988), as also listed in the RESRAD-BUILD 

manual (ANL 2003). The radon conversion is also from the RESRAD-BUILD manual (ANL 2003). 

Because the conversion factors are applied at the end of the RESRAD-BUILD calculation, it is 

appropriate to reverse the conversion using the factors. The converted results are listed in 

Table B-25.

RESRAD-BUILD has a limitation on how many yearly printouts can be made. So, inhalation and 

radon exposures were calculated for each year for the D&D case, but only for the first few years and 

every 5 years thereafter for the post-D&D case. This is appropriate because as the output data show,

the values change little from year to year. So, although multiple years might be indicated in the 

column headings in Table B-25, the figures below them are for each

Annual inhalation intakes derived from the RESRAD-BUILD inhalation dose results were used as a 

basis for calculating ingestion doses according to the methodology of NIOSH (2004), as follows.

year, not the sum for the 

indicated years.

Ingestion intake (pCi/yr) = 0.2 × concentration (pCi/m3

) × 250 d/yr

Ingestion intake (pCi/yr) = 0.021 × Inhalation intake pCi/yr

The RESRAD-BUILD main output gives output only in dose units, while supplementary output does 

not give total picocuries per cubic meter or disintegrations per minute per cubic meter. But the 

supplementary output (the .diag files) shows that because equilibrium is assumed between the 

uranium isotopes and their progeny down to radon and because linear resuspension is assumed, 

there is a constant conversion factor from inhalation dose to air concentration. Thus, doses reported 

in the main output can be converted back to concentrations and, thus, to ingestion doses. The results 

of applying this conversion to obtain the ingestion are listed in Table B-25.

The principal assumptions made for the Annual external dose rates and doses, D&D/postoperations

period (Table B-26) are given below.

Direct radiation levels for gamma and beta were based on the maximum dose rates in each plant area 

or building, regardless of location. The maximum was the absolute maximum reading for an area or 

building, with one case for each level and type of radiation. The source terms are listed in Table B-26. 

Ceiling and overhead levels were not considered because of the lower activity levels in these areas 

and the infrequency of access. Without regard to the actual correspondence of gamma dose rate, 

beta dose rate, and surface contamination levels in particular rooms, the room model was assumed to 

contain the highest spot and average gamma and beta dose rates found anywhere in the respective 

area or building. Separate manual calculations were performed for gamma and beta radiation. Other 

assumptions and data manipulation details for the external dose calculations are given below.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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Table B-26. Annual external dose rates and doses, D&D/postoperations period.

a

Gamma (mR/hr) Beta (mrep/hr) Gamma (mR/yr) Beta (mrep/yr)

Building or area Min Mode Max Min Mode Max Min Mode Max Min Mode Max

Old Hex (492) (excl Recov, Still) 0.020 0.034 0.09 0.9 1.6 7.6 40 69 180 1760 3150 15200

Old Hex (492) (Recov, Still only) 0.010 0.015 0.04 0.04 0.07 0.32 20 30 80 80 133 640

New Hex (annex) (492/493) 0.01 0.015 0.04 0.11 0.19 0.93 20 30 80 220 387 1860

Brown (493) (incl Pilot Plant) 0.012 0.023 0.06 2.3 4.2 20 23 46 120 4620 8310 40000

Green (491) 0.012 0.023 0.06 0.29 0.52 2.5 23 46 120 577 1040 5000

Plant C: 1st floor (UF6/UO3 area) 0.075 0.15 0.39 0.35 0.62 3.0 150 294 771 693 1250 6000

Plant C: 2nd floor (UF4/UF6 area) 0.035 0.069 0.18 0.16 0.29 1.4 70 137 360 323 582 2800

Boiler House 0.075 0.15 0.39 0.35 0.62 3.0 150 294 771 693 1250 6000

Foundry 0.08 0.15 0.39 0.35 0.62 3.0 150 294 771 693 1250 6000

Garage 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0

Warehouse: lab, offices 0.0013 0.0024 0.006 0.006 0.010 0.05 2.5 4.9 13 12 21 100

Building P-1 0.018 0.034 0.09 0.08 0.15 0.7 35 69 180 162 291 1400

Building K-1 0.04 0.08 0.21 0.17 0.31 1.5 80 157 411 346 624 3000

Building M-1 0.013 0.024 0.06 0.06 0.10 0.5 25 49 129 115 208 1000

a. Data in the table are based on Blatz (1951) (predecontamination 491-492 dose rates), Schoen (1958) (491-492-493 

predecontamination beta dose rates for all but main hex), Klevin (1955a) (hex area predecontamination dose rates), and DOE (1984)

(postdecontamination dose rates). The maximum measured dose rate(s) in each area was used to calculate the annual whole-body 

doses. Assumptions and details are shown below.

1. Because exposure (dose) rates were used, the source terms did not have to be translated into 

activity units. However, while some surveys reported gamma and beta dose rates separately, 

others gave reported combined beta-gamma dose rates. These were ratioed using 

information from the surveys to produce separate gamma and beta dose rates. The resulting 

assumed source terms are listed in Table B-26.

2. In the surveys, the measurement point for betas or mixed beta-gamma radiation was usually a 

contact or near-contact dose rate for both walls and floors, while for gammas, it was most 

often a contact or near-distance dose rate for walls and a 3-ft measurement for floors. It was 

thus assumed that the measured beta dose rate and the measured mixed beta-gamma dose 

rate represented all-beta radiation emanating from a wall surface to a receptor point at 1 cm 

from a wall surface; similarly, the measured gamma dose rate was assumed to represent 

gamma radiation emanating from a wall surface to a receptor point at 1 m from a wall surface.

3. A preliminary set of calculations was done to see what size of source (e.g., point, small radius, 

large radius) was most appropriate for the measured data for each type of radiation. For both 

beta and gamma, a large-radius source was found to be most appropriate and favorable to 

claimants. For the gamma case, it was assumed that the source was of infinite radius 

because that was not a very large increase from a 4-m radius source and the room could thus 

be assumed to be on the order of the room used for the RESRAD-BUILD calculation (i.e., 

about 3.3 × 6.7 m for the wall lengths). For the beta case, the source was assumed to be of 

essentially infinite radius (i.e., 8.5 m, the range of the most energetic beta emitted from the 

uranium-daughter source mix).

4. Dose rates per Item 2 above were used to determine the area source strength for beta and 

gamma separately and then these source strengths were used to calculate the dose rates at 

1 ft and 1 m for beta and at 1 cm and 1 ft for gamma as needed. Per NIOSH direction, the 

respective dose rates at 1 cm were then considered to be the maximum dose rates, the dose 

rates at 30 cm (1 ft) the most likely dose rates, and the dose rates at 1 m (3 ft) the minimum 

dose rates.

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

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5. The receptor was assumed to stay at 1 ft from the source for 2 hr/workday and at 3 ft from the 

source for 6 hr/workday, for a total time of 2,000 hr/yr. This ignored break time, which was 

usually spent in areas of very low or no contamination.

The stay time assumptions in Item 5 should be adjusted for workers not likely to have spent 

considerable time in the areas of residual contamination, especially Plant C; the doses listed in Table 

B-26 should then be ratioed by an appropriate fraction to indicate a reasonable amount of time spent 

in the contaminated area. For example, a claimant who was a secretary in the postoperations years 

likely did not spend much time in the areas of significant contamination and should be assigned only a 

small fraction of the doses listed in Table B-26. A fraction of 5% is suggested for office workers, 10% 

for higher managers, and 25% for maintenance and safety workers [based on engineering judgment, 

given the information in the various dust studies (including time-and-place information), observations 

in AEC and Harshaw memos and reports, and statements in HCC (1946)].

ATTACHMENT B

TABLES IMPORTANT TO DOSE RECONSTRUCTION

Page 39 of 39