ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 94

Document Title:

An Exposure Matrix for Linde Ceramics 

Plant (including Tonawanda Laboratory)

Document Number: ORAUT-TKBS-0025

Revision: 00

Effective Date: 05/31/2005

Type of Document: TBD

Supersedes: None

Subject Expert: Gerald R. Davidson

Document Owner

Approval: Signature on File Approval Date: 05/23/2005

Cindy W. Bloom, TBD Team Leader

Approval: Signature on File Approval Date: 05/24/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 05/24/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 05/31/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

PLEASE DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 2 of 94

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

11/30/2004 00-A New Technical Basis Document: Basis for the Development of an 

Exposure Matrix for Linde Air Products, Tonawanda, New York and 

Buffalo, New York. Initiated by Gerald R. Davidson.

04/07/2005 00-B Major reorganization and some changes in assumptions in response 

to ORAUT and OCAS comments. Removed reference to Linde Air 

Products, Chandler Plant in Buffalo, NY. Added neutron doses and 

residual contamination exposure after 1954. Initiated by Gerald R. 

Davidson and Cindy W. Bloom.

04/12/2005 00-C Reserved internal dose section prior to November 1, 1947 at 

Ceramics Plant and prior to January 1, 1947 at Tonawanda 

Laboratory. Modified beta dose assumptions for the lowest exposure 

categories in some sections and modified residual period external 

exposures. Corrected errors (language, ambiguity, typographical) 

when found. Initiated by Cindy W. Bloom.

05/18/2005 00-D Incorporated issues from the Worker Outreach Meeting on April 18, 

2005 applicable to the site profile. Corrected additional text errors 

found by reviewers. No changes in intent or estimating methods 

were made. Initiated by Cindy W. Bloom.

05/31/2005 00 First approved issue. No training required. Initiated by Cindy W. 

Bloom.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 3 of 94

TABLE OF CONTENTS

Section

Publication Record ..............................................................................................................................

Page

2

Acronyms and Abbreviations ...............................................................................................................8

1.0 Introduction...............................................................................................................................10

2.0 Site Description and Operational History................................................................................... 10

2.1 Facilities..............................................................................................................................11

2.1.1 Buildings A and B....................................................................................................... 11

2.1.2 Building 14 (Tonawanda Laboratory).......................................................................... 11

2.1.3 Building 30 (Ceramics Plant) ...................................................................................... 13

2.1.4 Building 31 (Ceramics Plant) ...................................................................................... 14

2.1.5 Building 37 (Ceramics Plant) ...................................................................................... 14

2.1.6 Building 38 (Ceramics Plant) ...................................................................................... 16

2.2 Source Term — African and Domestic Uranium Ores .........................................................16

2.3 Activities..............................................................................................................................17

2.3.1 Uranium Research and Development ......................................................................... 17

2.3.2 Uranium Production.................................................................................................... 18

2.3.2.1 Step I (Uranium Ore to U3O8) ............................................................................. 18

2.3.2.2 Step II (U3O8 to UO2) .......................................................................................... 20

2.3.2.3 Step III (UO2 to UF4) .......................................................................................... 20

2.3.3 Nickel Production........................................................................................................ 21

2.3.4 Other Radiological Activities ....................................................................................... 21

2.4 Protective Measures ...........................................................................................................22

2.5 Personnel, Job Categories And Workhours.........................................................................23

2.6 Decontamination During the MED/AEC Contract Period .....................................................25

2.7 Post MED/AEC Operations ................................................................................................. 26

3.0 Estimation of Internal Exposure, 1942-1954.............................................................................. 26

3.1 Estimation of Particulate Intakes .........................................................................................27

3.2 Uranium ..............................................................................................................................28

3.2.1 Uranium Urinalysis Data............................................................................................. 29

3.2.2 Uranium Air Concentrations........................................................................................ 30

3.3 Radium ...............................................................................................................................31

3.3.1 Radon Breath Data..................................................................................................... 31

3.4 Uranium Progeny ................................................................................................................31

3.5 Radon ................................................................................................................................. 32

3.5.1 Ceramics Plant Radon Exposures .............................................................................. 32

3.5.2 Tonawanda Laboratory Radon Exposures.................................................................. 34

3.6 Inhalation Intake Estimates of Radioactive Particulates.......................................................34

3.7 Ingestion Intake Estimates ..................................................................................................35

3.8 Consideration of Bioassay Data ..........................................................................................35

3.9 Occupational Internal Dose Reconstruction Assumptions and Summary 

1942-1954...........................................................................................................................36

4.0 Estimation of External Exposure, 1942-1954 ............................................................................ 37

4.1 Ceramics Plant Beta and Gamma Exposure .......................................................................39
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4.1.1 Post-production Radiation in Building 30 .................................................................... 39

4.1.2 External Exposure During Non-production Periods at the Ceramics 

Plant...........................................................................................................................40

4.1.2.1 Pre-production, 1942-1943 ................................................................................ 40

4.1.2.2 Standby, 1946-1947........................................................................................... 41

4.1.2.3 Cleanup, 1949-1954 .......................................................................................... 41

4.1.3 Production, 1943-1946 ............................................................................................... 43

4.1.3.1 Beta ...................................................................................................................43

4.1.3.1.1 Estimation of Worker Doses ...................................................................44

4.1.3.1.2 Outdoor Dose Rate ................................................................................46

4.1.3.1.3 Categories..............................................................................................46

4.1.3.2 Gamma.............................................................................................................. 47

4.1.3.2.1 Estimation of Worker Doses ...................................................................50

4.1.3.2.2 Outdoor Exposure Rate..........................................................................54

4.1.4 Rehabilitation and Production, 1947-1949 .................................................................. 54

4.1.4.1 Beta ...................................................................................................................54

4.1.4.2 Gamma.............................................................................................................. 58

4.2 Tonawanda Laboratory Beta and Gamma Exposure...........................................................60

4.3 Neutron Exposure ...............................................................................................................60

4.4 External Dose Reconstruction Summary, 1942-1954 ..........................................................63

5.0 Occupational Medical Exposure................................................................................................ 65

5.1 Ceramics Plant Employees ................................................................................................. 65

5.1.1 Chest X-rays...............................................................................................................65

5.1.2 Pelvis/Lumbar Spine X-rays ....................................................................................... 69

5.1.3 X-ray Dose Reconstruction Guidelines ....................................................................... 70

5.2 Tonawanda Laboratory Employees .....................................................................................71

6.0 Estimation of Exposures from Residual Contamination After 1954............................................ 71

6.1.1 Internal Exposure ....................................................................................................... 72

6.1.2 External Beta and Gamma Exposure.......................................................................... 73

6.2 Neutron Exposure ...............................................................................................................74

6.3 Dose Reconstruction Summary, 1955 to Present................................................................74

References ........................................................................................................................................75

Glossary ............................................................................................................................................89

A Beta Radiation............................................................................................................

Attachments 

90

B Codes and Special Terminology .................................................................................94
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LIST OF TABLES

Table

1 Key dates...............................................................................................................................

Page

12

2 Department codes for various worker activities ...................................................................... 23

3 Worker information ................................................................................................................. 24

4 Ceramics Plant assumed work schedule, 1942-1954............................................................. 24

5 Intake ratios and absorption types for uranium progeny ......................................................... 31

6 Measured radon concentrations during Ceramics Plant African ore 

processing .............................................................................................................................33

7 Ceramics Plant worker radon exposure rates, 1942-1954...................................................... 33

8 Tonawanda Laboratory radon exposure rates, 1942-1954 ..................................................... 34

9 Assumed airborne concentrations used to estimate intakes ................................................... 34

10 Predicted uranium urinalyses from Ceramics Plant assumed inhalation and 

ingestion chronic uranium intake from November 1, 1947 to December 

31,1954 based on 33 MAC in air............................................................................................ 36

11 Internal exposure summary for the Ceramics Plant, October 1, 1942, through 

December 31, 1954................................................................................................................37

12 Internal exposure summary for the Tonawanda Laboratory, October 1, 1942, 

through December 31, 1954................................................................................................... 38

13 Floor and wall radiation in Building 30 before and after the 1949-1950 

decontamination.....................................................................................................................39

14 Building 30 floor radiation levels in 1976 ................................................................................ 40

15 Pre-production beta and gamma radiation levels ................................................................... 40

16 Ceramics Plant beta and gamma radiation rates during standby............................................ 41

17 Beta and gamma radiation measured during equipment removal, 1948................................. 42

18 External dose reconstruction parameters for cleanup............................................................. 43

19 Beta radiation exposure rates measured at Linde Ceramics January 23, 1944 ...................... 44

20 Dependence of beta dose rate on ore type ............................................................................ 46

21 Assigned 1943-1946 beta dose rates..................................................................................... 47

22 Job categories for 1943-1946 Step I gamma dose analysis ................................................... 49

23 Step I gamma exposure data for African ore processing, 1/31/44-11/12/44 ........................... 49

24 Step I gamma exposure rate parameters for African ore processing, 1/31/44-

11/12/44................................................................................................................................. 50

25 Step I gamma exposure data for cleanup and domestic ore processing, 

11/13/44-2/26/45....................................................................................................................50

26 Step I gamma exposure rate parameters for cleanup and domestic ore 

processing, 11/13/44-2/26/45................................................................................................. 51

27 Estimated maximum exposure rates in selected Building 30 facilities due to 

30 ft x 30 ft x 12 ft pile of 8% ore in southeast corner............................................................. 54

28 Job categories for 1947-1949 Step III film badge analysis...................................................... 56

29 1947-1949 Step III beta dose data ......................................................................................... 56

30 Beta dose rates for Step III, 1947-1949.................................................................................. 57

31 1947-1949 Step III gamma exposure data ............................................................................. 58

32 Gamma exposure rates for Step III, 1947-1949...................................................................... 59

33 Tonawanda Laboratory beta and gamma radiation, 1942-1954.............................................. 60

34 Natural uranium per gram dose rates at 1 foot ....................................................................... 61

35 Neutron dose rates and annual doses from the alpha-neutron sources.................................. 63

36 Summary – annual external exposure from AWE operations, 1942-1954............................... 64

37 Ceramics Plant worker beta and gamma external exposure categories for 

production periods..................................................................................................................66
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38 Highest monthly outdoor airborne radionuclide concentrations at Linde 

Tonawanda site, 2000-2004................................................................................................... 72

39 Annual internal and external exposure to residual radioactivity .............................................. 74

A-1 Maximum beta surface dose rates from various uranium-containing materials....................... 90

A-2 Electron energy released by refined uranium ......................................................................... 90

A-3 Electron energy released by unrefined natural uranium ore ................................................... 91

A-4 Estimated attenuation of beta radiation with distance ............................................................. 92

A-5 Uranium beta dose reduction factors for apparel.................................................................... 92

A-6 Variation of dose rate with body location for Step III process operations ................................ 93

B Codes and special terminology .............................................................................................. 94
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LIST OF FIGURES

Figure

1 Linde Ceramics and Tonawanda Laboratory buildings on Linde Tonawanda 

site.........................................................................................................................................

Page

13

2 Building 30 layout................................................................................................................... 14

3 Building 30 layout................................................................................................................... 15

4 Building 38 layout................................................................................................................... 16

5 Fit to gamma exposure data for a 4 ft x 5 ft x 20 ft pile of 8% African 

pitchblende ore ......................................................................................................................52

6 Calculated exposure rates and analytic fit for a 30 ft x 30 ft x 12 ft pile of 8% 

African pitchblende ore .......................................................................................................... 53
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ACRONYMS AND ABBREVIATIONS

μg microgram

ACE Army Corps of Engineers

AEC Atomic Energy Commission

BNI Bechtel National, Inc.

Bq becquerel

CFR Code of Federal Regulations

Ci curie

d day

DOE Department of Energy

dpm disintegrations per minute

FUSRAP Formerly Utilized Sites Remedial Action Program

GM geometric mean

GSD geometric standard deviation

h hour

hr hour

hrs hours

ICRP International Commission on Radiological Protection

keV kiloelectron volt

L liter

LOD limit of detection

LOOW Lake Ontario Ordnance Works

MAC maximum allowable concentration

m meter

MED Manhattan Engineer District

mm millimeter

mr radiation exposure unit used by MED in the 1940s; interpreted here as mR for photons 

and mrem for beta radiation

mR milliroentgen

mrad millirad

mrem millirem

mrep millirep

NIOSH National Institute for Occupational Safety and Health

NRC Nuclear Regulatory Commission

NYDO New York Directed Operations 

NYOO New York Operations Office 

ORNL Oak Ridge National Laboratory
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PAEC potential alpha energy concentration (a measure of concentration of radon daughters)

PAEE potential alpha energy exposure (a measure of exposure to radon daughters)

pCi picocuries

PL preferred level

R roentgen

R&D research and development

Sv sievert

TBD technical basis document

USBM U.S. Bureau of Mines

U.S.C. United States Code

wd workday

wk week

WL working level

WLM working level month

y year
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 10 of 94

1.0

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

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

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

These documents may be used to assist the National Institute for Occupational Safety and Health 

(NIOSH) 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 facility” or a “Department of Energy facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 [42 U.S.C. Sections 7384l (5) and (12)].

This document provides an exposure matrix for workers at the following facilities of the Linde Air 

Products Company: Tonawanda Laboratory and the Ceramics Plant in Tonawanda, New York.

2.0

The information in this Technical Basis Document supports the assumed operational and residual 

contamination periods listed below. 

SITE DESCRIPTION AND OPERATIONAL HISTORY

Facility Operational and initial cleanup period Residual contamination perioda

Tonawanda Laboratory October 1, 1942 to December 31, 1946 January 1, 1947 to present

Ceramics Plant October 1, 1942 to December 31, 1954

Stand by: August 1, 1946 to September 14, 1947

January 1, 1955 to present

a. Also called the “post-cleanup period” in this document.

When the United States government and its contractors first became interested in uranium, Linde Air 

Products, then a division of Union Carbide and Carbon Corporation, operated Tonawanda Laboratory, 

which had been producing U3O8 that was sold as a coloring agent for ceramics. Because of the great 

interest in obtaining uranium that could be used to create the experimental uranium piles, Linde was 

contracted to develop uranium chemical processes and build a facility that could process large 

amounts of uranium ore. This commissioned facility was called the Linde Ceramics Plant. Linde

worked with two types of radioactive material:

• refined uranium materials and preprocessed (partially refined) domestic ores; 

• unprocessed African ores. 

Processing of the African ores involved exposures to significantly higher levels of 230Th, 226Ra, and 

222Rn. In the other materials, the initial refining process probably removed most non-uranium 

radionuclides. Due to the long half-lives of 230Th and 231Pa, significant ingrowth of these nuclides and 

their progeny in the 238U and 235U decay chains did not occur in the interval between the refining of the 

materials and their processing at Linde.

Linde Air Products Company performed work for the Manhattan Engineer District (MED) and its 

successor the U.S. Atomic Energy Commission (AEC) between 1942 and 1949. Tonawanda 

Laboratory performed research and development on uranium processing for the MED beginning in 

late 1942 and probably ending in 1946. In the early years, it operated pilot plants in order to develop 

procedures used at the Ceramics Plant. The Ceramics Plant produced uranium materials for the 

MED and the AEC from 1943 to 1946 and from 1947 to 1949. In the 1947 to 1949 period (and 

perhaps earlier), Linde received UO2 for processing from Mallinckrodt Chemical Works (AEC 1949a). 

Linde also produced nickel material for the K-25 diffusion barrier. The Ceramics Plant was in standby 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 11 of 94

from mid-1946 to late 1947. The end of production in 1949 was followed by cleanup and 

decontamination and then turnover of the production facilities back to Linde. This turnover probably 

occurred in 1954. 

Uranium processing at the Ceramics Plant was described as involving three steps: 

• Step I, conversion of ore to U3O8 (black oxide); 

• Step II, conversion of U3O8 to UO3 (orange oxide) as an intermediary product and then to UO2

(brown oxide); 

• Step III, conversion of UO2 to UF4 (green salt). 

More detailed histories of Linde facilities, radiation sources and processes related to MED/AEC 

radiation exposures follow. Key dates are noted in Table 1.

2.1 FACILITIES

Tonawanda Laboratory and the Linde Ceramics Plant were located on land owned by Union Carbide 

at East Park Drive and Woodward Avenue in Tonawanda, New York (ORNL 1978, Fig. 3). The site is 

near the intersection of Riverview Boulevard and Woodward Avenue. It is north of Woodward 

Avenue, east of East Park Drive, and west of the Conrail railroad tracks.

Tonawanda site buildings involved in MED/AEC work are shown in Figure 1. Tonawanda Laboratory 

occupied Building 14, and the Ceramics Plant used Buildings 30, 31, 37, 38, and A. Building B 

contained MED offices. Ownership of the Ceramics Plant buildings was transferred to Linde after the 

site cleanup that began with the shutdown of production in 1949. The transfer probably was 

completed in 1954. In the 1990s, the site was acquired by Praxair, Inc. (ACE Buffalo 2003). As of 

this writing (2005), it is owned by Praxair. 

2.1.1

Building A was the Linde Ceramics Plant office building for administrative and support personnel. 

During the MED years, Building B housed the Tonawanda Area Engineer office of the MED (Dupree 

1983a) and probably was later used by AEC personnel. The designations "A" and "B" were adopted 

for this document and may be different from the official Linde or MED designations. Buildings A and B 

were no longer standing in 1978, when the 1976 radiological survey was issued by ORNL (1978). 

Buildings A and B

This figure is based on Figures 3-1 and B-11 of BNI 1982, Figure 3 of Frame et al 1981 and Linde 

Database 1945.

A shelter in which workers were allowed to smoke was located at the south end of Building B (Dupree 

1983a). This was probably for use only by office workers. A separate smoke shelter was provided for 

production workers (Klevin 1949a, data sheets 543 and 546).

2.1.2

Building 14 housed Tonawanda Laboratory. The U.S. Army Corps of Engineers has described its use 

as follows:

Building 14 (Tonawanda Laboratory)
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Table 1. Key dates. 

Assumed date or perioda Event or activity

Ceramics Plant

6/1/43 – 7/31/46b Step I production (U3O8 from ore or sludge)

6/1/43 – 11/30/43 Preprocessed ores (domestic and scrap)

12/1/43 – 11/12/44 African ores

11/13/44 – 1/31/46 Preprocessed domestic ores

2/1/46 – 2/28/46 African ores (48%) and preprocessed domestic ores (52%)

3/1/46 – 6/30/46 African ores

7/1/46 – 7/31/46 African ores and preprocessed ash

4/27/43 – 3/8/44c Step II production (UO2 from U3O8)

7/25/43 – 6/26/46d Step III production (UF4 from UO2)

8/1/46 – 9/14/47 Standby

9/15/47 – 10/31/47e Step III rehabilitation

11/1/47 – 6/30/49f Step III production

10/1/44 – 2/28/46g Production of nickel material for K-25 diffusion barrier

7/1/49 – 3/31/50h Cleanup of Building 30

7/1/49 – 12/31/54i Cleanup of Linde Ceramics Buildings

Prior to 1978 Demolition of Buildings A and B (office buildings on Figure 1)

After 1954 Post-cleanup period

8/31/81j Demolition of Building 37 began

8/31/96j Demolition of Building 38 began

9/5/98k Demolition of Building 30 began

9/30/00j Soil remediation began (scheduled for completion in 2007)j

Tonawanda Laboratory

10/1/42 – 7/31/46l MED-related research and development operations

8/1/46 – 12/31/46m Cleanup period

After 1947 Post-cleanup period

4/30/04j Demolition of Building 14 began

a. Unless more precise information was available, activities were assumed to begin on the first day of 

the start month and to end on the last day of the completion month indicated in the data source, and 

demolition activities were assumed to begin on the last day of the start month in the data source.

b. All Step I dates and data are based on Table B-1 of Aerospace Corporation 1981 except for the 

November 1944 transition date from African to domestic ore. The date was determined from analysis 

of Step I film badge data (see text Section 4.1.3.2).

c. Linde Air Products Company Ceramics Plant 1946b, p. 40.

d. Linde Air Products Company Ceramics Plant 1946c, p. 42.

e. Start date based on Kent 1947, p. 1.

f. Start date per Rennich 1947. End date per AEC 1951, p. 30. 

g. Hickey, Crawford-Brown, and Tankersley 1988, Figure 13, p.2. 

h. Start date assumed to be day after end of Step III production; end date estimated per Eisenbud 1950.

i. Start date assumed to be day after end of Step III production; end date estimated per Harris 1954.

j. Pilon 2004.

k. ACE Buffalo 1998.

l. Start date per Jenness and Ewing 1943. End date assumed to be the end date of 1943-1946 

production at the Ceramics Plant.

m. Start date based on estimated end date of research and development period; end date estimated.

Building 14 was used for laboratory and pilot plant studies for uranium separation in the 

early part of MED operations. Historical drawings indicate that the MED laboratory and 

pilot plant studies were initially confined to the south part of the building. It is unclear 

how extensively the remainder of the building was used for MED operations. However, 

documents indicate that laboratory and pilot plant operations were continued for the 

purpose of experimenting and developing more efficient processing methods, and 

operations appear to have been expanded into most of the building, possibly to support 

larger pilot studies. The available records do not indicate whether or not the use of 

Building 14 ceased before the MED/Atomic Energy Commission (AEC) operations 

were discontinued at Linde.
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Figure 1. Linde Ceramics and Tonawanda 

Laboratory buildings on Linde Tonawanda 

site. 

A 1942 Tonawanda Laboratory directory listed approximately 200 employees (Tonawanda Laboratory 

1942). The locations of most were designated by room numbers. Some employees were said to be 

located in the "Eng. Lab" and others in the "Proving Lab." A 1944 employee list reported some to be 

located in the "Pilot Plant" (Tonawanda Laboratory 1944). The terms "Proving Lab" and "Pilot Plant" 

appear to refer to the same facility.

Tonawanda Laboratory is also referred to by other names such as "Linde Research Laboratory" 

(Dupree 1983a) and "The Laboratory of The Linde Air Products Company" (Jenness and Ewing 

1943).

Contamination in Building 14 was found in a 1976 radiological survey (ORNL 1978). At the time, 

Building 14 housed fabrication facilities, research facilities, and offices generally used by 20 to 30 

employees. Demolition of Building 14 began in April 2004 (Pilon 2004). 

2.1.3

Building 30 — also called the "uranium refinery" — was used for Steps I and II. Figures 2 and 3 show 

the layout. It had two floors. Large pieces of equipment such as digestion tanks were located on the 

ground floor. Process workers operated equipment from the second floor (Dupree 1983a,b). The 

1949 to 1950 decontamination work done by Linde and the results are described in Heatherton 1950. 

As of 1976, the building was used as a shipping and receiving warehouse and occupied by about 20 

to 30 employees. Radioactive contamination was found during a 1976 survey (ORNL 1978). Building 

30 was demolished during the period September 5-19, 1998 (ACE Buffalo 1998).

Building 30 (Ceramics Plant)
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Figure 2. Building 30 layout (from Heatherton 1950).

2.1.4

Building 31 is described as "used in uranium separation process during MED operations" (BNI 1993, 

p. 1-65). A partially-readable 1945 plant drawing indicates that it contained shipping and storage 

areas, maintenance facilities (including a welding shop, a machine shop, a gauge test room, and a 

carpentry area), and a small number of associated offices (Linde Database 1945). Building 31 might 

have been the final location of Linde's Tonawanda nickel processing operations.

Building 31 (Ceramics Plant)

As of 1976, the building was described as housing fabrication facilities, offices and storage areas and 

was occupied by about 12 to 15 employees. Radioactive contamination was found during a 1976 

survey (ORNL 1978). Building 31 was decontaminated in 1997 (ACE Buffalo 2002). As of 2004, it 

was still standing and in use (Pilon 2004).

2.1.5

Building 37 was a small appendage to Building 39, which was located east of Building 38 (Figure 1). 

According to BNI 1993 (p. 1-65), Building 37 was used for Step III. However, no details of this use 

have been found and its small size, approximately 16 ft x 36 ft (BNI 1982 Figure B-11), indicates at 

most a minor role. The 1945 plant drawing labeled it "Test Bldg" (Linde Database 1945). Radioactive 

contamination was found in 1976 (ORNL 1978). Linde demolished Building 37 in 1981 (Pilon 2004).

Building 37 (Ceramics Plant)
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Figure 3. Building 30 layout (from Linde Database Undated A).
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2.1.6

Building 38 was used for Step III. Figure 4 shows its layout. Radioactive contamination was found in 

1976 (ORNL 1978). Building 38 was demolished in 1996 (Pilon 2004).

Building 38 (Ceramics Plant)

Figure 4. Building 38 layout (from Klevin 1949a).

2.2 SOURCE TERM — AFRICAN AND DOMESTIC URANIUM ORES

Linde used two types of starting materials for Step I (ore to U3O8) processing. These are generally 

referred to as "African ore" and "domestic ore" in Linde-related literature. Most of the African ore was 

pitchblende from Africa; some was torbernite (Aerospace Corporation 1981, p. 1). Neither African 

ore underwent chemical preprocessing before being shipped to Linde. Thus, they contained not only 

uranium, but all members of the uranium decay series, including radium and its progeny.

The domestic ore category included preprocessed domestic ores and byproducts of other uranium 

conversion processes. The preprocessed domestic ores derived from tailings from vanadium 

processing. The preprocessing concentrated the uranium and had the effect of removing "the major 

portion of the radium" in the original ore (Aerospace Corporation 1981, p. 5). The preprocessing may 

have also separated uranium from other radionuclides in the ore. If the separation of uranium were 

complete, then the radioactivity in the preprocessed ore delivered to Linde would have consisted 

primarily of the uranium radionuclides U-234, U-235, and U-238 (in their natural abundance) and their 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 17 of 94

short-lived progeny that had grown in after the extraction process. The completeness of the 

separation of uranium from other radioactivity in the vanadium tailings used at Linde is not known. A 

review of the chemical processing concluded only "the radium in the domestic ore was significantly 

less than that in the African ores" (Aerospace Corporation 1981, p. B-1).

The distinction between the two types of ore is important. The primary radiological hazards from an 

ore containing only uranium and its short-lived progeny would be due to alpha and beta emissions. 

However, radium and other progeny in the African ores would produce, in addition, significant gamma 

emissions and elevated levels of radon. In this document, the term "domestic ore" always means 

preprocessed domestic ore, and is sometimes referred to as "refined ore." "African ore" always 

means ore that has not undergone preprocessing.

Enriched uranium (35% or less by mass) might have been used on a limited basis for K-25 barrier 

testing and research and development at Tonawanda Laboratory. Because of the very limited 

quantities in comparison to the ore, only natural-abundance uranium is explicitly considered for this 

site. 

2.3 ACTIVITIES

Linde Air Products Company is reported to have begun research into the processing of uranium ore in 

the late 1930s before it had a contractual relationship with the MED. At the time, uranium was used to 

make colored glazes for ceramic dinnerware. A former MED employee interviewed in 1983 (Dupree 

1983a) said that Linde Research Laboratory in Tonawanda, New York developed a process to obtain 

purified U3O8 from carnotite ore and produced 80 tons of U3O8. Linde's capability to provide 

commercial quantities of U3O8 as of 1941 was confirmed in a letter to atomic weapons physicist Leo 

Szilard (Dana 1941).

On November 16, 1942 Linde Air Products entered into contract W-7401 eng-14 with the MED to 

design and construct facilities at the Ceramics Plant for processing uranium-containing ores and to 

produce U3O8, UO2 and UF4 (Cornell 1942; Marshall 1943). The contract also called for research and 

development regarding the processes that would be used in the Ceramics Plant facilities. 

2.3.1

Available reports indicated two main areas of Tonawanda Laboratory MED uranium work. The first 

was development of methodologies for Ceramics Plant operations. A report on the first year’s work 

covered the period October 1942 to November 1943 (Jenness and Ewing 1943). Most of the work 

was devoted to study of the problems encountered in the design, operation, and improvement of the 

Ceramics Plant.

Uranium Research and Development

Three different large-scale pilot plant programs were described. As of November 20, 1942, one was 

producing 2 to 3 tons of U3O8 per week (Bonsib 1942). The third program, which involved African ore, 

had begun in October 1943 and was expected to conclude in December 1943.

Available documentation from the Laboratory includes three research reports (Brimm 1943a; Skinner 

1944; Wiesendanger 1944) and a final report (Tonawanda Laboratory 1946b). The final report 

discussed work that was done on R-10, a low grade African ore, and Q-20, a natural (i.e., not 

preprocessed) torbernite ore. No dates for this work were provided, but since R-10 and Q-20 

processing did not begin at the Ceramics Plant until February 1946, the R-10 and Q-20 studies at 

Tonawanda Laboratory probably were performed in late 1945.
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The second area of Tonawanda Laboratory work was related to conversion of UF6 process gas to 

UO3. Research reports (Brimm and Schubert 1945a,b) and a final report (Tonawanda Laboratory 

1946a) indicated that the associated pilot plant operations used pounds rather than tons of material.

A later Linde research report on uranium processing dated September 29, 1948 was prepared by 

Ceramics Plant rather than Tonawanda Laboratory personnel (Chapman et al 1948). The absence of 

MED or AEC reports after May, 1946 suggests that MED work by Tonawanda Laboratory ended then

and that there was no later AEC work. For dose reconstruction, it is assumed that there was no 

Tonawanda Laboratory work for the MED or AEC after July 31, 1946, the end date of the initial 

production period at the Ceramics Plant. 

2.3.2

The Linde Ceramics Plant engaged in three different uranium production activities — Steps I, II, and 

III. Step I (production of U3O8) typically involved ores containing 3-20% U3O8 by weight. 

Approximately 70% of the ore was African and 30% domestic by weight. About 26,000 metric tons of

ore were processed, and about 2,300 metric tons of U3O8 were produced (Aerospace Corporation 

1981, Table B-1). 

Uranium Production

Step I production took place in Building 30 from June 1943 through July 1946. Step II (production of 

UO2) took place in Building 30 from April 1943 through March 1944. Step III (production of UF4) took 

place in Building 38 from July 1943 to June 1946. Linde sent the U3O8 produced from Step I to 

duPont and Mallinckrodt after Step II production ceased (Gates 1946). Linde received UO2 from other

companies for Step III processing (AEC 1949a). Tonawanda Laboratory might have investigated all 

of these processing steps in its pilot plant. Uranium production was in standby after July 1946. 

Preparations for Step III resumption began in September 1945. Step III operations were resumed 

under contract AT-30-1-GEN-165 in November 1947 (Rennich 1947). Linde ended uranium 

production for the AEC in June 1949.

A MED record-of-negotiations memorandum (Dreveskracht 1945) noted that Linde had offered for 

sale 36,000 pounds of uranium ore concentrates from Colorado and 8,900 pounds of residue, which 

were the "property of Contractor as a result of operations prior to his present CPFF [cost plus fixed 

fee] work for the Manhattan District (Contract No. W-7401 eng-14)." The transfer was made under 

contract W-17-028 eng-29 dated May 4, 1945 (U.S. Engineer Office 1945). Correspondence 

indicates that Linde transferred ownership of the uranium concentrates and residues to the 

government on August 1, 1945 (Martin 1945) and the material was subsequently processed in the 

Ceramics Plant for MED (Dupree 1983a, Robinson 1945).

Detailed descriptions of the typical Ceramics Plant uranium processing methods are presented below. 

These include descriptions of chemical and operational processes.

2.3.2.1 Step I (Uranium Ore to U3O8) 

Step I extracted purified U3O8 (black oxide) from uranium ore. The Step I process varied with time as 

the nature of the ore changed and as Linde gained experience with the procedure. 

Domestic ore was ground and mixed with water to form a slurry. Sulfuric acid and other chemicals 

were added, and the mixture was digested at 90°C for 2-3 hours and then cooled to 60°C. This left 

the uranium in solution as uranyl sulfate along with some of the impurities. Soda ash and sodium 

bicarbonate were added to make the solution basic. This precipitated out most of the remaining 

impurities and left the uranium in solution as sodium uranyl tricarbonate. The slurry was filtered in 
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"Moore filters" (described in the next section). The liquors contained the uranium and some 

objectionable impurities. Vanadium and phosphorous impurities were precipitated by the addition of 

ferrous and ferric sulfates and then removed in a second filtering in plate and frame presses, which 

produced “iron cake.” The liquors were treated with caustic soda to precipitate the uranium as sodium 

diuranate, which was removed as cake in a third filtering process. The sodium diuranate cake was 

treated with sulfuric acid and ammonium sulfate to convert it to an ammonium uranyl sulfate complex. 

This was removed in a fourth filtering process. The resulting “acid leach cake” was fed to a calciner to 

drive out the ammonia, sulfur dioxide and trioxide, and water, leaving the black oxide U3O8. 

A modification of the above procedure was used to treat African ore. The filter cake from the Moore 

filters (first precipitation step) contained the radium in the ore. In the second step, instead of iron 

sulfates, sodium sulfide was added to remove lead. Also, some of the other chemical additives used 

at various points were different (Linde Air Products Company Ceramics Plant 1946a; Aerospace 

1981, p. A-1).

Step I Operations 

Ore was delivered by rail in boxcars. African L-30 pitchblende arrived packaged in burlap bags that 

were inside paper bags. Each ore bag weighed 50-80 lbs. For African ores, the car doors were left 

open for 12-24 hours before unloading to reduce radon concentrations in the car. Either ore bags or 

bulk ore (the practice varied) were manually loaded onto wheeled carts (buggies) by workers called 

Loaders or Movemen and transported to a storage or processing area (Olevitch 1944; Linde Air 

Products Company Ceramics Plant 1946a; Cranch 1944c).

For dry ore, processing began in the dumping room on the receiving platform. Ore was dumped 

through a grate onto a conveyer belt that carried it to the ore storage bin. The storage bin held about 

75 tons of ore. Ore from the bin was fed to a ball mill by a "Feedoweight" (a belt conveyor that

weighed the ore). After milling, the ore was fed to the digestion process.

Wet ore was transported to the second floor using the freight elevator and either dumped into the 

scoop box of the ball mill or directly into digest tanks (skipping the ball mill step). Loaders did not ride 

on the elevator with the ore.

After the dumping, about a half pound of ore remained in each bag. To recover it, the bags were 

shaken in a bag shaker over a conveyor belt onto which freed ore dropped. The bags were then 

washed, centrifuged, and burned in an incinerator. The washed bags contained about 10 to 30% of 

the remnant ore. At times, washed bags awaiting incineration were stored outdoors on the site. A 

July 1944 report noted that 19,000 bags were currently being stored (Olevitch 1944, Appendix A).

For digestion, chemicals appropriate to the ore type were added to the digest tanks, and the mixture 

was heated and cooled to precipitate out components. Precipitates were separated from solutions 

using various filter assemblies. The first filtration in the Step I processing used the Moore filters. A 

Moore filter consisted of a large basket containing 24 rectangular leaves with each leaf approximately 

35 sq. ft. in area. Each leaf was formed by a rectangular perforated pipe frame over which a bag was 

stretched. The filter basket was placed by crane into a neutralized digest tank and held there while 

suction was applied to the pipe. Gelatinous cakes about 1-inch thick formed on the leaves. The 

basket was removed from the digest tank, placed in a wash tank, washed, and then moved to a 

hopper. The suction was then reversed to inflate the bags and cause the cake to fall into the hopper. 

From there the cake was transferred to an open-top truck below the hopper for transport to a storage 

location. After sampling, the filter cake was hauled away by truck. Radium-containing filter cake from 

the African ores was shipped to the Lake Ontario Ordnance Works (LOOW) in Lewiston, New York for 

storage. Filter cake from the domestic ores went to a tailings pile. From some filter presses, the filter 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 20 of 94

cake was manually dumped by the operators into buggies for transport to the next step. From other 

presses, the cake could be fed automatically to the next step (MED Undated A, third page; Olevitch 

1944; Linde Air Products Company Ceramics Plant 1946a; Aerospace 1981, p. A-1).

The tanks used in the process were large, 18 to 31 ft in diameter and one story high (Dupree 

1983a,b).

2.3.2.2 Step II (U3O8 to UO2) 

Step II converted U3O8 to UO2 (brown oxide) with UO3 (orange oxide) as an intermediary product. 

Black oxide (U3O8) was dissolved in nitric acid. Some insoluble impurities were filtered off, and the 

solution was evaporated to molten uranyl nitrate hexahydrate. The next step was based on ether 

extraction of the uranium. The extraction took advantage of the fact that metallic salts are more 

soluble in water than in ether, but the uranyl nitrate is nearly equally soluble in ether and water and 

can be shifted from one medium to the other by changes in volume and concentration ratios. The 

uranyl nitrate hexahydrate was added to cold ether, and the solution was washed with water. The 

impurities went to the water layer and the uranyl nitrate to the ether layer. The purified uranyl nitrate 

was denitrated to powdered uranium trioxide (UO3) by heating in a stainless steel pot furnace. The 

trioxide powder was collected mechanically and flowed by gravity through a micropulverizer into a 

rotary tube reduction furnace. There the powder was reduced to the desired uranium dioxide (UO2) 

using gaseous hydrogen as a reducing agent. The finished UO2 product was packaged, sampled, 

sealed, and stored (Linde Air Products Company Ceramics Plant 1946b; Aerospace 1981, p. A-1).

Step II Operations

U3O8 in buggies from Step I was weighed and hoisted to the second floor. It was then fed into nitric 

acid in digest tanks with scoops. The digest tanks were heated to 90°C. After digestion, insoluble 

impurities were filtered out using a filter press. Liquids were piped from one vessel to another. 

Frequent chemical analyses of samples were required in this process to determine the progress of 

chemical reactions. The processing produced various liquors and some associated cakes that were 

designated "OK" or "NG" (Linde Air Products Company Ceramics Plant 1946b).

2.3.2.3 Step III (UO2 to UF4) 

Step III converted UO2 to UF4 (green salt) using the chemical reaction

UO2 + 4HF  UF4 + 2H2O 

The reaction was carried out at 1000°F. Gaseous anhydrous HF gas was flowed over the UO2, which 

was placed on magnesium trays stacked on spacer bars inside 9” square tubes 10 feet long that were 

inside a furnace (Linde Air Products Company Ceramics Plant 1946c; Hickey, Crawford-Brown, and 

Tankersley 1988, p. 8).

Step III Operations 

UO2 was weighed and hand-troweled into shallow trays inside a hood with a dust collector operating. 

Twelve pounds of UO2 were placed on each tray. The oxide was furrowed when loaded to maximize 

surface area. The trays were transported on buggies and inserted into the furnace. 

After loading, the furnace was sealed, purged of air, and heated to the starting temperature. Then the 

flow of HF gas was started. After the required number of hours, the heat and gas flow were stopped, 

and the furnace was purged of HF gas, cooled, and opened.
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The trays were removed from the furnace and hand-placed onto buggies for transport to and 

placement inside an unloading hood with a dust collector. Good material (light green) was loaded into 

hoppers. Bad material (dark green) was placed in fiber-pack drums for later re-treatment. Good 

material was pulverized, blended, sampled, packaged, weighed, and shipped (Linde Air Products 

Company Ceramics Plant 1946c; Hickey, Crawford-Brown, and Tankersley 1988, p. 8).

2.3.3

After the shutdown of Linde Ceramics Step II in 1944, the Step II equipment was converted for use in

production of nickel powder, a non-radioactive process, which occurred from October 1944 through 

February 1946 (Hickey, Crawford-Brown and Tankersley 1988, pp. 9 and 13). The nickel was for use 

in the K-25 diffusion barrier (Hunter 1949; Dupree 1983b, p. 3). 

Nickel Production

There is discrepant evidence on the location of the nickel processing work. A contractual record (U.S. 

Government 1945) states that Linde was contracted on April 7, 1944 via letter contract W-7418 eng51 to construct and equip a plant to produce 80 tons of special nickel oxide per month and that the 

plant was to be constructed in Tonawanda on the same premises and adjacent to the buildings used 

in connection with Contract W-7401 eng-14, the contract for the uranium refining work. A drawing 

attached to a former Tonawanda Area Engineer Office employee interview record (Dupree 1983a) 

shows the location of the nickel processing to be in a building east of Building 30. Thus, nickel 

processing may have occurred in Building 31 (see Figure 1). However, another record of interview 

(Dupree 1983b) states that when Step II operations in Building 30 ended, the Step II area was used to 

refine nickel. Both recollections could be correct. It may be that nickel processing initially occurred in 

the Step II area and later was moved to Building 31.

Evidence supporting Building 31 as at least the ultimate location for Linde's Tonawanda nickel 

processing is found a 1945 plant drawing (Linde Database 1945). The visible portions of partially 

obscured label on the drawing are consistent with the interpretation that it says "Nickel Plant." Also, 

the drawing shows an ammonia cracking area in Building 31, and other records indicate that ammonia 

was used in the nickel processing (chemical operators in department C2P used ammonia according to 

Linde Air Products Company 1945, p. 17, and department C2P engaged in nickel processing 

according to Hickey, Crawford-Brown, and Tankersley 1988, p. 28).

Nickel processing at the Ceramics Plant has been described as follows (Hickey, Crawford-Brown, and 

Tankersley 1988, p. 9):

Metallic Ni slugs (small ingots) were oxidized with HN03, filtered, and heated to 1200 C 

to produce a nickel oxide (NiO). This was pulverized, mixed with ammonium chloride 

(NH4Cl), and heated to form a porous mass of the oxide, which was reduced to 

powdered nickel by exposure to H2. The powder was pulverized and packaged. ... It is 

not known whether both NiO and powdered Ni or only NiO were produced in quantity at 

this plant.

2.3.4

Early on, radioactive liquid wastes were discharged to the Tonawanda sanitary sewage system. Due 

to the nature of the liquids, this became a problem, and Linde began to dispose of liquid wastes into 

onsite wells that sometimes overflowed. Later still, liquid wastes were discharged to a drainage ditch 

that led to a sewer conduit (BNI 1993, pp. 1-9 to 1-15). 

Other Radiological Activities
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After MED work began at the Tonawanda site, there was potential exposure of workers to radiation 

and radioactivity when outdoors. Portions of the site are known to have been contaminated with 

radioactivity (Heatherton 1948h; ORNL 1978; BNI 1982); resuspension would have produced airborne 

radioactivity. One source of ground contamination and airborne radioactivity was the ore unloading 

process, which involved transporting ore in buggies — sometimes in bulk and sometimes in bags — 

from box cars to Building 30 (see the section on Step I operations in Section 2.3.2.1). Outdoor areas 

of the site were sometimes used for storage of radioactive materials. Olevitch (1944) reports the 

outdoor storage of contaminated ore bags that at times numbered in the thousands. In 1948, 1 mR/h 

gamma and 3,000 α dpm/100 cm2 were measured from the soil in an area formerly used for storage 

of radioactive materials (Heatherton 1948h).

An additional source was the release of liquid effluents either to onsite wells that sometimes 

overflowed, or to an onsite drainage ditch (BNI 1993, pp. 1-9 to 1-15). Airborne effluents from the 

plant were an additional source of outdoor radioactivity.

In 1949, Linde workers were used to unload drums of K-65 shipped by rail to the Lake Ontario 

Ordnance Works at Modeltown, New York (Wolf 1949; Heatherton 1949a). Linde film badges were 

worn during this work.

At the April 18, 2005, Worker Outreach Meeting, strontium was mentioned. Summary information 

from the meeting did not indicate if this was radioactive on nonradioactive strontium, or if this was 

process material or a sealed source. Because no mention of radioactive strontium was found in the 

available Linde documents, it is believed likely that this source might have been a sealed radioactive 

source, nonradioactive material or a small quantity in comparison to the uranium source term. The 

estimates of uranium exposure based on estimates of exposure periods and source term, which were 

based on worst case assumptions when a parameter was not well supported by available information 

would be sufficiently bounding to account for small amounts of radioactive strontium.

2.4 PROTECTIVE MEASURES

Steps were taken to protect against radiological and chemical hazards during processing (Linde Air 

Products Company Ceramics Plant 1946a,b,c; Linde 1945; Dupree 1983a,b). Although some 

protective measures were included early in the Linde program, protective measures were upgraded 

over time in response to both survey findings and newly available information regarding biological 

risks of exposure. These measures included limiting the time a worker was allowed to spend in a 

radiologically hazardous activity, provision and upgrading of ventilation to reduce concentrations of 

radon and radioactive dusts in the air, and good housekeeping to reduce radioactive dust on the floor. 

Respirators to reduce dust and chemical exposure, and wearing of gloves to reduce beta exposure to 

the hands were required for some of the more hazardous tasks, but these safety devices were not 

used in all process areas by all workers. Some workers changed into company-issued clothing when 

they arrived at work and changed back to their own clothing at the end of a shift. Showers were 

available. At some point in Linde operations, showering at the end of a shift became mandatory for 

some workers, and they were "wanded" by a health physicist to check for radioactivity when they left 

the shower.

Chemical protective measures included goggles, respirators, rubber gloves, face shields, rubber or 

plastic coated aprons and sleeves, and head caps. There were more substantial protective measures 

for operators working with hydrofluoric acid. 

According to interviewees (Dupree 1983b), use of respirators was often at the worker's discretion, but 

they were "usually used." Inspectors reported mixed compliance with recommendations to wear 
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respirators. The author of a June 1944 inspection report commented, "The wearing of respirators 

remains difficult to enforce" (Ferry 1944f). The author of an October 1944 inspection report observed, 

"Respirator discipline was about 80% efficient in the dumping room" (Tybout 1944a). In 1945, the 

same inspector witnessed the collection of dust samples by Mr. H. Seemann, Assistant to the Safety 

Engineer. In his report, the inspector noted, "The men wore respirators, but Mr. Seemann stated that 

this was the case only because he was present" (Tybout 1945d). A 1948 dust sample collector 

commented that when samples were collected, individuals on the day shift wore respirators but those 

on the evening shift did not (Hayden 1948).

No credit for safety gear (including respirators) is taken in this document.

2.5 PERSONNEL, JOB CATEGORIES AND WORKHOURS

For Tonawanda Laboratory, two listings of employees may aid in determining whether a claimant 

worked there. A September 1942 Laboratory directory (Tonawanda Laboratory 1942) lists 

approximately 200 employees. Several employees are linked to the Proving Lab, which might be an 

indicator of higher exposures, once the uranium work began. A 1944 employee list (Tonawanda 

Laboratory 1944) contains approximately 120 names and similar information. It is not clear whether 

this list includes all employees.

For the Ceramics Plant, an April 1944 employee list contains nearly 400 names (Neuman 1944) and 

specifies the worker jobs. A December 1945 job description specifies the duties of 51 categories of 

workers and lists a department code for each position (Linde Air Products Company 1945). Since the 

same title (e.g., chemical operator) was sometimes used in different departments in which the nature 

of the work was very different (e.g., Step I and nickel processing), knowing the department may help 

identify the type of activity in which a worker was involved. Table 2 shows Ceramics Plant department 

codes associated with various worker activities as determined in a 1988 study of worker hazards. A 

1949 study reported that there were 139 Step III personnel and listed the number in each job. Names 

of some of the personnel were provided (Klevin 1949a, pp. 6, 80). 

Table 2. Department codes for various worker activities. 

Department Department codea

Administrative Services AB, AD, AL, AM, BA, BB, BT, A/M

Chemical Control, R&D BL, HL

Engineering and Development AW

Maintenance AN

Process Operations: Step I CL

Step II CM

Step III CN

Nickel CP, CR, C2M, C2P

Loading: General AN, AH, LOOW

Janitors AA, GE

Stores and Supplies AS, ASA, A8A, AT

Safety and Security AJ, A9A, A2J

Undetermined Activity CA, CS

Untitled Jobs and Departments — b

a. From Hickey, Crawford-Brown, and Tankersley 1988, p. 27. Based in part on a 

plant document identifying some of the codes and in part on inference from less 

direct information in plant records.

b. Not available.
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Table 3 provides a list of some Linde references that include worker names and sometimes additional 

job information.

Table 3. Worker information. 

Item Reference

Tonawanda Laboratory employees and room numbers dated September 2, 1942. Job 

titles are not stated.

Tonawanda 

Laboratory 1942

Linde employees associated with Contract W-7401-Eng-14, July 20, 1943. Some are 

Tonawanda Laboratory personnel. Some might be Ceramics Plant personnel. Job 

titles are included.

Pew 1943

Ceramics Plant personnel with job titles, April 25, 1944 Neuman 1944

Ceramics Plant Step III process operators, June 14, 1948 Heatherton 1948f

Ceramics Plant Step III process area personnel (including operators, millwrights, and 

stores attendants) included in report published January 25, 1949

Klevin 1949a, p. 

50

Film badge exposure report listing 23 Step III personnel by name and job title, for week 

beginning March 28, 1949

AEC 1949b

List of Ceramics Plant Step III process operators included in report published August 

26, 1949

Klevin 1949b, p. 

11

The term “day” in this document refers to a calendar day. The term “workday” is used to describe a 

day at work. Default assumptions are 250 workdays per year and 8 hours per workday: this results in 

2000 hours of work per year. The distinction between workday and calendar day is especially 

important when considering internal dose rates for use in calculating organ doses, because intakes 

actually occur during the workday, but dosimetry calculations are usually based on integrating over 

calendar days. 

The Ceramics Plant work schedule during the production period (1943-1946) is described as involving 

eight-hour shifts, six days per week (Dupree 1983a,b,c). A nine-hour workday with a lunch period 

included is assumed because there is evidence that many employees worked overtime (Dupree 

1983b, p. 4; MED Undated A, p.1). An eight-hour workday plus a half hour for lunch is known to have 

been in effect for many of the workers in late 1948 (Klevin 1949a). The transition from a six-day 

workweek to a five-day workweek was assumed to have occurred on January 1, 1951 based on a 

report by interviewees that the 48-hour workweek lasted until 1950 (Dupree 1983b, p. 4).

Based on the above, the periods and work schedule at the Ceramics Plant from the onset of MED 

operations through 1954 are assumed to have been as shown in Table 4. 

Table 4 Ceramics Plant assumed work schedule (including lunch and 

breaks), 1942-1954. 

Period Start End h/wd wd/wk wk/y

Pre-production 10/1/42 4/26/43 9.0 6 50

Production 4/27/43 7/31/46 9.0 6 50

Standby 8/1/46 9/14/47 8.5 6 50

Rehabilitation & Production 9/15/47 6/30/49 8.5 6 50

Cleanup 7/1/49 12/31/50 8.5 6 50

1/1/51a 12/31/54 8.5 5 50

a. Assumed date of transition from six-day to five-day workweek (based on Dupree 

1983b, p. 4).
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2.6 DECONTAMINATION DURING THE MED/AEC CONTRACT PERIOD

Cleanup and decontamination activities at the Ceramics Plant facilities during the MED/AEC contract 

period are discussed below. No information was found on cleanup of Tonawanda Laboratory

facilities, but it was noted at the April 18, 2005 Worker Outreach Meeting that Building 14 was 

contaminated and later surveys indicated that residual contamination remained after AEC work (BNI 

1993). 

Cleanup of the Ceramics Plant began before the shutdown of Step III production, which occurred on 

June 30, 1949. Some Step II equipment in Building 30 was removed in March and April of 1948 by or 

with the assistance of a contractor, H. K. Ferguson (Heatherton 1948c,d,e). Dismantling of Step I and 

remaining Step II equipment in Building 30 was under way in May 1949 (Heatherton 1949b). A 

contractor (Kulp Waco) assisted in some of the work (Heatherton 1949d, e). 

Shortly after the shutdown, a comprehensive cleanup effort was undertaken to reduce levels of 

radioactivity in Building 30 in order to enable its release to Linde for unrestricted use (Heatherton 

1950). After removal of the bulk of the process equipment from Building 30, the entire building was 

vacuum cleaned and flushed with water. A systematic radiation survey was conducted to identify 

areas of contamination. Decontamination was accomplished primarily by removing contaminated 

parts of the building (such as portions of the second floor "balcony" on which process operators had 

been stationed) and by abrading surfaces (mostly by sandblasting, although oxygen acetylene torches 

were also used).

After each area was decontaminated, it was again cleaned and flushed, and then a final radiation 

survey was made.

Workers, performing surface abrasion activities, were provided with a U.S. Bureau of Mines (USBM) 

approved rubber hood that protected the face, neck, and shoulders and had a supplied air respirator 

to permit breathing of uncontaminated air. Other personnel in the area wore demand-type respirators 

or, for short periods, USBM respirators approved for all dusts.

An AEC internal memorandum dated March 29, 1950 reported that the decontamination of Building 30 

had been completed and indicated that a decision on release of the building to Linde was to be made 

by April 1, 1950 (Eisenbud 1950).

Decontamination of Building 38 began with a radiation survey in November 1952 (Harris 1952). An 

AEC memorandum dated April 1954 (Harris 1954) reported levels obtained after decontamination. 

Average readings on categories of equipment to be left at the plant ranged from 0.05 to 0.6 mrep/h 

beta plus gamma with the gamma component either undetectable (which meant less than twice 

background) or less than 5% of the total. The memorandum reported "overall" floor and wall levels to 

be 1.01 "mreps/hr/ft2

" beta plus gamma. This was considered excessive. The memorandum cited 

data indicating that this could be reduced to 0.065 (mreps/hr)/ft2 by covering contaminated floor areas 

with asphalt tile. It recommended release of the building to Linde once this was done. The date at 

which Linde met this condition was not found, but since the requirement was modest, it is likely that 

the release to Linde occurred in 1954.

No information was found on cleanup and decontamination activities in Buildings 31 and 37.

This document assumes the end date of the Ceramics Plant cleanup period to be the date of turnover 

of the four Ceramics Plant production buildings to Linde for its use. This date is sometimes stated as 

1953 (see, for example, ACE Buffalo 2004a, Response to Question 4). However, Harris (1954) 
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indicates that the decontamination of Building 38 was not complete as of April 1954. For dose 

reconstruction, it is assumed that turnover did not occur until December 31, 1954.

2.7 POST MED/AEC OPERATIONS

The Formerly Utilized Sites Remedial Action Program (FUSRAP) began in 1976. Oak Ridge National 

Laboratory (ORNL 1978) surveyed the Linde Tonawanda from October 18 through November 5, 1976

to determine if remediation would be required. Radiation and radioactive contamination 

measurements were made inside Buildings 14, 30, 31, 37 and 38; on the Tonawanda property 

outside the buildings; and at nearby offsite locations. Linde employees noted that Building 30 

renovation occurred in the 1960s and could have resulted in elevated employee radiation exposures. 

Notes from the Worker Outreach Meeting on April 18, 2005 mention contamination associated with 

Building 57, and an additional review of the BNI1993 remedial investigation report shows areas of 

residual radioactive contamination were associated with areas in or near Buildings 57, 58 and 90. 

The highest indoor radiation levels were found in the principal production buildings, 30 and 38. Linde 

was designated as a FUSRAP site in 1980. Additional radiological surveys and decontamination 

efforts followed (BNI 1993). These led eventually to demolition of Building 14 and all of the Linde 

Ceramics buildings involved in MED/AEC work except Building 31. Table 1 shows demolition dates. 

As of 2004, Building 31 remained in use, and onsite soil remediation was in progress with completion 

scheduled for 2007 (Pilon 2004).

3.0

This section develops parameters for reconstruction of doses due to internal exposures from October 

1, 1942, the assumed start date of MED work at Linde, until December 31, 1954, the assumed date of 

initial cleanup completion and building turnover from MED/AEC to Linde. 

ESTIMATION OF INTERNAL EXPOSURE, 1942-1954

This section describes information related to internal dose reconstruction including Linde’s uranium 

bioassays, airborne radioactivity concentration measurements, radon breath analyses for 

determination of radium body burdens and radon air concentrations. To expedite preparation of this 

document, the Linde information is considered in conjunction with information from facilities that did 

similar types of uranium processing to establish preliminary estimates of internal intakes and 

exposures. These estimates are considered best estimates, until data can be further considered. It is 

believed that additional analysis of the data will lower at least some of the intakes and exposures 

estimated in this section.

This document assumes that Tonawanda Laboratory’s primary internal exposure occurred from 

October 1, 1942 to December 31, 1946. At Ceramics Plant, four periods were assumed: October 1, 

1942 to July 31, 1946 (operations), August 1, 1946 to September 14, 1947 (standby), September 15, 

1947 to October 31, 1947 (rehabilitation) and November 1, 1947 to December 31, 1954 (operations 

and cleanup). From October 1, 1942 and into 1946 both African and domestic ore were used for 

research, development, and production. Tonawanda Laboratory AEC work is assumed to have 

stopped radiation work after December 31, 1946, although some workers might have visited the 

Ceramics Plant buildings. The exact levels of contamination remaining at the Ceramics Plant and the 

nature of worker activities in areas of residual activity are unknown for the standby and rehabilitation 

periods. Exposures to uranium’s progeny and to radon would have decreased at the Ceramics Plant 

after August 31, 1946, because ore was no longer processed during this later period. Continued 

lower level exposures to uranium progeny and to radon are assumed, because some radioactive 

waste was disposed on site and because initial cleanup was not completed until the end of 1954; 

however, for the Ceramics Plant the uranium exposures would have dominated during the 1947 to 

1954 period. 
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The primary sources of internal radiation exposure at Linde due to MED and AEC work were natural 

uranium and its progeny.

3.1 ESTIMATION OF PARTICULATE INTAKES

During World War II and the time of the Manhattan Engineer District (MED), the permissible level for 

natural uranium air dust concentration was set at 500 μg/m3 for insoluble uranium compounds and 

150 μg/m3 for soluble uranium compounds. After the war, the University of Rochester (Rochester, 

NY) recommended lowering the permissible level to 50 μg/m3 for all uranium compounds on the basis 

of chemical toxicity. This level was also stated as 70 disintegrations per minute per cubic meter (70 

dpm/m3

) for natural uranium and was based primarily on animal studies. The Medical Division of the 

AEC New York Operations Office (NYOO) thought that the “maximum permissible level” should be 

based on human data and was thus unknown. Therefore, the level of 50 μg/m3 was generally referred 

to as the “preferred level” (AEC 1949a). Oftentimes the contractors of the AEC used the term 

“Maximum Allowable (air) Concentration” (MAC) interchangeably with the “preferred level” (PL) and 

reported air-sampling results as multiples of the MAC (NLO 1952, AEC 1953). When considering air 

concentrations reported in multiples of the PL or MAC, the actual assumed value of the PL or MAC 

should be verified.

In 1949, the Medical Division of the NYOO published a report on the health hazards at seven facilities 

that produced and/or processed uranium for the AEC. These facilities included Mallinckrodt Chemical 

Works, Harshaw Chemical Company, Linde Air Products, Electro Metallurgical Company, and Vitro 

Manufacturing Company. The AEC used the information on work tasks with measured air 

concentrations in breathing zones, general areas and process areas to determine average air 

concentrations weighted by exposure times and summed these time-weighted air concentrations to 

determine daily time-weighted average air concentrations by job categories. Up until the time of the 

1949 AEC report, surveys by the NYOO indicated that out of 648 exposed workers at these plants, 

9% were exposed to uranium air concentrations greater than 125 MAC (greater than 6250 μg/m3

), 9% 

were exposed at 25-125 MAC (1250-6250 μg/m3

), and 82% were exposed to less than 25 MAC (less 

than 1250 μg/m3

). Linde’s maximum time weighted exposure during this period was 33 or 32 MAC

(the data in the report’s text and its graph differed). As a result of the NYOO report, significant 

improvements were made in operational conditions such as re-design of ventilation systems, 

enclosing some processes, and using remote controls (AEC 1949a). By the end of 1949, exposure 

levels were significantly reduced at these larger plants even though production levels increased 

(Mason 1958).

Indications are that some of the higher routine (versus episodic) exposures occurred at the uranium 

ore processing facilities. The AEC did not exist until January 1, 1947, and its report (AEC 1949a) did 

not elaborate on internal exposures prior to their tenure. A general review of air concentration data, 

safety reports, and production and progress reports from the early 1940s through 1946 indicates that 

there were significant reductions in exposures due to improved engineering, process, workplace and 

administrative controls brought into effect prior to the AEC report. This indicates that the exposures at 

the ore and waste processing plants in the earliest years of operations would likely have been higher 

than the exposures that occurred during AEC tenure. Indeed, the “tolerance” air concentration was 10 

times higher for insoluble compounds during the early years and 3 times higher for soluble 

compounds.

As of this writing, the pre-1947 operational period intakes are reserved. Therefore, the pre-1947 

information is provided only as a description of what the likely upper bound exposures might have 

been, and is not currently planned for use in Linde dose reconstruction. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 28 of 94

For the pre-1947 period, the MAC would have been assumed to be based on inclusion of uranium’s 

alpha emitting progeny. Although short-term exposures might have exceeded 300 MAC, it is very 

unlikely that long-term exposures would have. A review of the predicted urinalyses, kidney burdens 

and lung burdens, indicate that it is highly unlikely that an individual would have sustained exposures 

like these for any length of time. Evidence of sustained exposure to the more soluble uranium 

compounds might have shown up in the medical urinalyses, as increases in proteins and glucose in 

the urine (note that other conditions can also account for these increases). The assumption of air 

concentrations at 300 MAC seems adequate to provide a quick estimate of exposure, and although 

the type F uranium bioassay results are high, they do not seem inconceivable for some workers 

during this early period. However, it is also likely that Linde workers were exposed to a mixture of 

uranium absorption types. The analysis of radium exposures in Section 3.8 is partially based on the 

assumption of alpha activity air concentrations of 300 MAC during Linde’s ore processing period. 

After the ore processing, Linde began a standby period. It was initially and arbitrarily assumed that 

exposures decreased to 1 MAC during the standby period at the Ceramics Plant, and that exposures 

decreased to 0.1 MAC at the Tonawanda Laboratory after cleanup in 1946 until the end of cleanup at 

the Ceramics Plant in 1954. Based on reviews of later air concentrations at Linde, and reviews of air 

concentration data from other sites, it is believed that most workers’ exposures would have been

much lower during these periods. 

The standby period at Linde Ceramics was assumed to end on September 14, 1947. Rehabilitation of 

the Step III process was assumed to begin on September 15, 1947 and continue through October 31, 

1947. Intakes from the standby and rehabilitation periods are reserved. Beginning November 1, 1947 

at Linde Ceramics, workers were assumed to be exposed to 33 MAC and it was assumed this 

exposure continued through cleanup in 1954. Uranium progeny are not included in this later period, 

because only refined uranium was used and because the dose from intakes of contamination left from 

earlier work would have been insignificant compared to the dose to uranium during operations.

To simplify calculations, it assumed that the workweek was 40 hours long during all years, although it 

is likely that the workweek for many was in excess of 40 hours especially during the earlier years. 

The assumed air concentrations are sufficiently large to account for any differences in actual hours 

exposed.

Dose reconstructions should assume International Commission on Radiological Protection (ICRP) 

Publication 66 default parameters for particle deposition (ICRP 1994).

3.2 URANIUM

During uranium research and the processing of uranium ores at Tonawanda Laboratory and the 

Ceramics Plant, workers might have been exposed to a variety of uranium chemical forms that 

encompass all lung absorption types: F, M and S. Although some process steps might have had 

more or less exposure to a given lung absorption type, it is not clear how well separated these areas 

were prior to standby. In addition, workers may have worked in multiple uranium process areas. This 

analysis assumed that types F, M and S were available prior to standby. After operations ceased in 

1946, it is more likely that remaining uranium contamination would have been oxidizing and becoming 

less soluble. Beginning with the rehabilitation period for restart of Step III operations in 1947, it is 

assumed that only type M and S materials would be available. The selection of absorption type 

should depend on the organ of interest. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 29 of 94

3.2.1

Medical urinalyses (specific gravity, proteins, sugars, etc.) were a requirement at Linde from the 

beginning of MED work, but there is no evidence that urine was analyzed for radioactivity until 1947. 

MED noted that tentative arrangements had been made for uranium urinalyses of certain groups of 

employees at Linde (Ferry 1944a), but no records are available indicating the analyses were done. A 

page (no author, circa 1947) titled, “Schedule of Examinations, Contract AT-30-1-GEN-165,” which 

commenced sometime in 1947 (Rennich 1947), indicates that 60 cc of urine from each employee was 

to be sent monthly to University of Rochester, the location of one of the earliest bioassay analysis 

programs. 

Uranium Urinalysis Data

Individual uranium urinalysis data from November 1947 to 1950 are available for some Linde 

Ceramics workers (Linde Ceramics Plant Urinalysis Data 1947-1950). The period covers both Step III 

production in Building 38 and decontamination in Building 30. Worker radiation exposures during this 

period were likely to be lower than during the production period prior to standby, because African ore 

was no longer being processed, Step I and II processes had ceased, and engineered safety controls 

had been improved.

The reported results of Linde uranium urinalyses during Step III production and later cleanup ranged 

from 0 to 3.10, with many of the higher results reported in the November to December 1947 “preemployment urines.” Because cleanup/preparation activities were occurring prior to collection of 

these samples, it is possible the elevated urinalysis results represented exposures received either 

during rehabilitation of the Step III equipment or perhaps from the less soluble component of uranium 

taken in during earlier operations. 

In November 1948, two uranium urinalysis results, 3.10, and 0.33 mg/L, were noted as high, and 

Linde was requested to consider the possibility of contamination. A fluoride result from the individual

with the highest uranium results was also elevated. No information was found to indicate whether 

these were contaminated samples, but the individuals’ results are consistent with an intake of soluble 

uranium.

Some samples might have been collected over the weekend: one set of samples indicated that 

collection bottles were taken home on a Friday because there was no work for two days, and the 

bottles were returned on Monday morning at 8:00 a.m. A note on the January 1950 uranium 

urinalysis results states, “please add approx[imately] 1% of conc[entrated] HCl to urine as a 

preservative instead of toluene.” 

It appears that the University of Rochester and the AEC NYOO were performing the uranium 

urinalyses for Linde (Wolf 1948g). The uranium fusion photofluorimetry urinalyses performed by the 

University of Rochester and the AEC NYOO were similar to those performed at other AEC facilities. 

The default detection threshold for uranium urinalysis is 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). 

However, the January 1948 uranium urinalysis results transmitted by the AEC to Linde (Wolf 1948) 

included a note that indicates the MDA might have been much higher.

The uranium content of those samples listed as being 0.1 mg U/l or less is below the limit of accurate 

quantitative determination by the photofluorometric method of analysis. 

February, May, June, August, and November1948 uranium urinalysis results include a similar note, 

but 0.01 mg U/L was reported as the limit of accurate quantitative determination or reliable 

determination for photofluorometric analysis. Based on this information, the assumed detection 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 30 of 94

threshold through January 31, 1948 is 0.1 mg/L for Linde uranium photofluorimetry urinalyses. 

Beginning February 1, 1948 the assumed uranium urinalysis detection threshold is 0.01 mg/L. Note 

that it is possible that the January 1948 determination level of 0.1 mg/L is a typographical error, 

because this is the same as the determination level reported for (nonradioactive) fluoride urinalysis, 

and because there seems to be no change in the format of the numbers reported. 

The AEC (Wolf 1948) requested “check samples” on all uranium urinalyses that exceeded 0.1 mg/L, 

and noted in February 1948 that it would be appropriate to resample in about three months. Linde’s 

reply (Heatherton 1948g) requested clarification, because it was their previous understanding that 

resampling was to occur monthly.

3.2.2

Information regarding particulate air concentration measurements at Linde is provided in this section. 

This information might be useful in future analyses, however, this site profile currently uses an 

estimate of Linde’s largest time weighted air concentration reported by AEC (1949a) to estimate 

intakes during uranium operations. Other Linde particulate air concentration data were reviewed and 

considered, but are not discussed here.

Uranium Air Concentrations

1943-1944

Air concentrations were measured at Linde Ceramics at selected locations during 1943-1946. These 

early results were reported in units of μg/m3 (Linde Ceramics Plant Airborne Dust Uranium Data 1943-

1946). Laboratory data sheets are available for measurements made in 1943. These indicate that 

each sample was collected in a tube. The mass of “X material” (uranium) in the tube was determined 

by multiplying the mass of uranium in a standard by the ratio of instrument response measured for the

sample to that measured for the standard, and dividing by the volume of air sampled. The instrument 

response was in units of volts/sec, but no description of the instrument was found. In a discussion of 

1945 results that also involved determination of the radioactivity of a dust sample in a tube, the 

methodology was characterized as involving “radioactive methods” and “radioactivity measurements” 

(Bryan 1945; Tybout 1945c), although another document referring to sampling of the nickel operations

seemed to imply that perhaps not all samples included radioactivity analyses (Rehm 1944b). 

Although there is the suggestion that radiation from air samples was measured, it is possible that 

some results reflect the mass of all materials suspended in the sampled air rather than the 

radioactivity in air. If this were the case, the reported air concentrations would be an overestimate of 

the uranium mass-based air concentration. In addition, if the uranium standard contained only refined 

uranium, then the reported results might overestimate uranium concentrations from African ore, 

because the progeny would also be counted as uranium. It is likely that the early measurements did 

account for radionuclides (or particulate) other than uranium, so although the early measurements are 

reported in terms of uranium mass, it is assumed that they are actually radiation-based values, which 

can be apportioned to different uranium progeny. 

1947-1949

Airborne dust radioactivity concentrations were measured in general areas and breathing zones 

associated with various locations and tasks in 1947-1949. Air particulate samples were collected by 

pumping air through filter paper disks, and an alpha counter was used to measure the radioactivity on 

a filter (Klevin 1949a). Results were reported in units of α dpm/m3 or of “preferred level” (noted then 

as 70 α dpm/m3

). The AEC combined the results with data on the amounts of time workers spent at 

various locations and tasks in order to determine daily time-weighted average radioactivity 

concentrations in the air breathed by workers in various jobs (Klevin 1949a, b). 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 31 of 94

3.3 RADIUM

All radium compounds are lung absorption type M. Radon breath analyses have been used to 

provide information on the amount of radium in the body and are available for some Linde workers. 

Assignment of radium exposures when radon breath analyses are not available or cannot be 

interpreted is addressed below in Section 3.4. 

3.3.1

The level of radon in exhaled breath can provide information on the amount of radium in the body. 

Radon breath data were found for 45 samples collected from Linde workers between June 12, 1944 

and January 17, 1945 (Linde Ceramics Plant Breath Radon Data 1944-1945). The records indicate 

that Dr. R. D. Evans in Cambridge, Massachusetts analyzed some of the samples. Room background 

concentrations of radon were sometimes subtracted from the results and sometimes not. In July and 

August 1944 (Tybout 1944b), room background concentrations were reported respectively as 0.6 and 

0.3 pCi/L (a picocurie equals a micro-microcurie). The measured radon levels were given for some but 

not all of the samples. Reported radon breath results ranged up to 2.2 pCi/L. The method of deriving 

the Ra-226 level was not stated, but in many cases, the reported burden of radium in μg was 

numerically equal to 10% of the breath radon concentration in pCi. Radon breath results are the 

starting point for dose reconstruction, so it might be necessary to back-calculate breath results from 

either Ra-226 burdens or tolerance levels. The original records might have sufficient information for 

this determination. This information has not yet been tabulated for inclusion in the site profile.

Radon Breath Data

3.4 URANIUM PROGENY

Ceramics Plant 1943-1946 Production, and Tonawanda Laboratories

In the absence of data on exposures to uranium progeny, their intake rates are determined by 

assuming secular equilibrium. Table 5 lists equilibrium-based ratios for uranium progeny of particular 

interest in dose reconstruction. Absorption types for their likely chemical forms are also shown. The 

intake ratios provide reasonably realistic estimates of intakes of progeny due to dust from African ore. 

The uranium activity fractions overestimate relative intakes of most progeny when the dust is from 

preprocessed domestic ore. They may underestimate intakes of progeny when the dust is from filter 

cakes or waste products that contain uranium progeny but very little uranium. The ratios in Table 5 

are for use for the entire 1943-1946 production period for all workers even though only about 70% of 

the ore processed was African ore (see Section 2.3.2) and many workers handled only refined 

uranium materials. This along with the claimant-favorable assumptions made in the estimation of 

worker dust exposures is judged to provide sufficient overestimation to balance any underestimation 

associated with the handling of waste products.

Note that the uranium fractions are applied when the activity of uranium is known. The activity 

fractions for gross alpha are applied to data measured as alpha activity.

Table 5. Intake ratios and absorption types for uranium progeny. 

Nuclide

Uranium

activity fractions

Gross alpha

activity fractions Absorption type

U-natural 1 4.02E-01 F, M, S

Th-230 4.89E-01 1.96E-01 M, S

Ra-226 4.89E-01 1.96E-01 M

Po-210 4.89E-01 1.96E-01 F,M

Pa-231 2.28E-02 9.16E-03 M, S

Ac-227 2.28E-02 9.16E-03 F, M, S
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 32 of 94

Ceramics Plant 1947-1949 Step III Production, and Subsequent Initial Cleanup

During this period, refined uranium materials were handled. None of the progeny listed in Table 5 

would have been present in significant quantities compared to the uranium at the Ceramics Plant. 

3.5 RADON

More than 200 measurements of radon concentrations were made during 1942-1946 pilot plant and 

production processing of African ore at Linde (Linde Ceramics Plant Radon Data 1944-1946). The 

early measurements were made using liter glass bulbs that were evacuated immediately before 

sampling to below 50 microns (Skinner 1944). It is likely that the later measurements during 

operations were made in the same way. Early samples were analyzed at the Massachusetts Institute 

of Technology; later analyses appear to have been performed at the University of Rochester. The 

tolerance value for radon, sometimes abbreviated as TV in the old records, was 100 pCi/L.

To determine exposure due to a specified radon concentration (assumed to be Rn-222), the 

concentration C in pCi/L was converted to potential alpha energy concentration (PAEC) in units of 

working level (WL) using: 

PAEC = C * F / (100 pCi/L per WL) 

where the equilibrium factor F was taken as 0.4, a value recommended by the ICRP (1981) and the 

United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR 1993). Potential 

alpha energy exposure (PAEE) in units of working level month (WLM) was obtained from PAEC and 

the number of months of exposure M using: 

PAEE = PAEC * M 

where a month is assumed to be 170 work-hours. 

3.5.1

During Ceramics Plant pre-production and initial production (which involved only domestic ore 

processing), the only source of radon was African ore processing at Tonawanda Laboratory. The 

indoor and outdoor radon concentrations to which Ceramics Plant workers were exposed were

assumed to equal the outdoor concentration resulting from Tonawanda Laboratory work. No direct 

measurement of this was available. An estimate was made based on the lowest indoor 

concentrations measured at the Ceramics Plant during African ore processing. These were viewed as 

indicating the upper limit to the outdoor concentration since outdoor air is drawn indoors for 

ventilation. Approximately 20% of the measurements in the Ceramics Plant ore processing building 

yielded results of 10 pCi/L or less, with most of these results at or near 10. Therefore, 10 pCi/L was 

taken as the estimated outdoor concentration.

Ceramics Plant Radon Exposures

Radon concentrations measured during processing of African ore at the Ceramics Plant are 

summarized in Table 6. Based on geometric means, locations were grouped into three categories — 

high, medium, and low, and the geometric mean and GSD of all results in each category were 

determined. Some results were reported as “less than” values and these are denoted as less than the 

limit of detection (<LOD). 

To simplify, this analysis assumes that workers, who were likely to spend the majority of their time in 

process areas, or in boxcars (where some of the highest radon levels were measured, about 200 

times tolerance), or whose jobs were unknown were exposed to 99.3 pCi/L of radon for 2040 hours 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 33 of 94

Table 6. Measured radon concentrations during Ceramics Plant African 

ore processing. 

Location category/location

# of 

Samples

GM

(pCi/L)a GSD

GM

(pCi/L)a 

GSD

High 99.3 3.43

Scales 5 153 2.50

Ore box car 17 125 3.92

Ore sampling room 24 100 3.11

Ore dumping grill 18 88.6 2.91

Ore storeroom 26 88.5 4.49

Ore bin/conveyor 7 83.3 9.37

Medium 42.5 2.72

Ore bag tumbler 17 60.2 2.53

Ore bag wash room 8 45.1 2.78

ball mill & classifier 24 32.5 2.74

Low 22.4 3.17

Step 1 digestion & 

filtration 55 23.7 3.17

Loading dock 1 18.0 —b

receiving desk 5 12.9 2.43

a. In computing the geometric mean, a <LOD value was assumed to equal the LOD.

b. GSD could not be determined.

(12 work-months) per year prior to standby. Workers who did not work or have their offices in the 

process buildings are assumed to have been exposed to 22.4 pCi/L of radon prior to standby. 

Because a job in current times might not be in or near a process area, does not mean the same held 

true 60 years ago. Nurses, some stenographers, launderers and seamstresses and some clerical 

workers had jobs or locations that put them in contact with the uranium and progeny (Homes 1944b). 

The initial period of African ore processing was followed by a second period of domestic ore 

processing. Thirteen measurements of radon concentration during the domestic ore processing were 

available. The geometric mean of the measurements, assuming the <LOD values were equal to the 

LOD, was 9.1 pCi/L. To estimate exposure during this domestic ore processing period, both indoor 

and outdoor radon concentration were assumed to be 10 pCi/L. 

After the end of African ore processing, concentration in the main ore processing building, Building 30, 

was assumed to remain at the level measured during the second period of domestic ore processing, 

10 pCi/L, until the end of its cleanup. Concentrations in other Ceramics Plant buildings were also 

assumed to be 10 pCi/L until the end of their cleanups. Since the locations of many workers are likely 

to be unknown, it was assumed that all workers were exposed to 10 pCi/L of radon until the assumed 

end of the Ceramics Plant building cleanup in 1954.

Table 7. Ceramics Plant worker radon exposures rates, 1942-

1954. 

Period/work location

Time-weighted

concentration

(pCi/L)

Exposure

rate 

(WLM/y)

10/1/1942 to 7/31/1946

In process and research areas 99.3 4.76

Not in process and research areas 22.4 1.08

8/1/1946 to 12/13/1954

All workers 10.0 0.480
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3.5.2

Few measurements were available, so estimates of radon concentrations were based on Ceramics 

Plant data. African ore work at Tonawanda Laboratory appears to have occurred just before each 

period of African ore processing at the Ceramics Plant. The precise dates of African ore research at 

Tonawanda Laboratory are not known. Based on the available reports, it appears to have been 

ongoing only about half of the time that the Laboratory was engaged in MED work, with most of the 

African ore work concentrated in the periods immediately preceding the beginnings of the two African 

ore processing periods at the Ceramics Plant. For dose reconstruction, it is assumed that African ore 

processing at Tonawanda Laboratory occurred during its whole MED period, but that the peak 

concentrations of radon were equal to the medium values at the Ceramics Plant. It is assumed that 

after MED research and initial cleanup at the Laboratory ended, the radon concentration dropped to 

10 pCi/L, the level in the Ceramics Plant after it switched from African ore processing to domestic ore 

processing, and remained there until the end of the cleanup at the Laboratory. After the end of its 

cleanup, radon exposure in the Laboratory was based on the highest geometric mean PAEC 

determined for a Tonawanda site building from measurements made in 1981, 1.68E-02 WL for 

Building 31 (based on analysis of data in BNI 1982, Table B-3). 

Tonawanda Laboratory Radon Exposures

Table 8 summarizes the assumed radon concentrations and resulting exposures.

Table 8. Tonawanda Laboratory radon exposure rates, 1942-1954

Period Time-weighted concentration

pCi/L

Exposure rate

Start End (WLM/y)

R&D and cleanup

10/01/42 12/31/46 42.5 2.04

Post-cleanup

01/01/47 12/31/54 — 0.202

Radon and Radon Daughters

Radon exposure rates provided here are annual PAEE rates in WLM/y. Each value is assumed to be 

the median of a lognormal distribution with a GSD of 3.43. This GSD is based on the location 

category having the highest GSD (3.43 per Table 6). 

3.6 INHALATION INTAKE ESTIMATES OF RADIOACTIVE PARTICULATES

Airborne concentrations were assumed in Section 3.1, for the purpose of estimating intakes for Linde 

workers. This analysis assumes that a worker was chronically exposed to alpha airborne 

concentrations as stated in Section 3.1 and summarized in Table 9. 

Table 9. Assumed airborne concentrations used to estimate intakes. 

Start End Activity description # MAC alpha dpm/m3 Source

Ceramics plant

10/1/1942 7/31/1946 Uranium ore processing Reserved Reserved Uranium and progeny

8/1/1946 9/14/1947 Standby Reserved Reserved Uranium and progeny

9/15/1947 10/31/1947 Rehabilitation Reserved Reserved Reserved

11/1/1947 12/31/1954 Step III processing 33 2,310 Uranium

Tonawanda plant

10/1/1942 12/31/1946 Uranium ore research Reserved Reserved Uranium and progeny

1/1/1947 12/31/1954 Post 0.1 7 Uranium and progeny

This analysis assumes Linde workers were exposed to the given air concentrations for eight hours per 

day, five days per week, 50 weeks per year. Time-weighted exposure studies of even the larger 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 35 of 94

plants like Mallinckrodt showed that the majority of workers were exposed at lower levels than those 

estimated for the early 1942 to 1946 operational period. No credit is taken for breaks or working at 

tasks where radioactive material intakes would be much lower or nonexistent, or for the use of 

facemasks; on the other hand no consideration was given to the longer workweeks and work hours in 

the early years. In addition, because workers were exposed to multiple absorption types, but the 

single absorption type that produces the larger dose is used to estimate organ dose, it is believed that 

this estimate adequately accounts for internal exposure. 

Because the early air concentration measurements would have included detection of radiations from 

progeny, the air concentrations that are listed in Table 9 as including progeny are apportioned based 

on the assumption of full equilibrium among uranium and its progeny, using the alpha ratios of 

uranium and progeny of internal dosimetric importance. 

For example, the annual uranium inhalation intake due to chronic exposure at 0.1 MAC is estimated 

by multiplying the air concentration of 7 dpm/m3

 by the alpha fraction of uranium, 0.402; the ICRP 66 

(ICRP 1994) recommended breathing rate of 1.2 m3

/h; and the assumed 2000 work-hours per 

calendar year. This results in an annual chronic inhalation intake of 6.75E+03 dpm, which is equal to 

a daily intake rate of 18.5 dpm/day. For the assumed exposure at 33 MAC, no alpha activity is 

apportioned to progeny so the daily uranium intake would be 1.52E+04 dpm/day.

Inhalations of the uranium progeny are estimated by substituting the alpha fraction of the progeny for 

the alpha fraction of the uranium. 

3.7 INGESTION INTAKE ESTIMATES

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

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

workday, can be estimated by multiplying the air concentration in dpm per cubic meter by a factor of 

0.2, so the uranium ingestion rate based on an air concentration of 7 alpha dpm/m3 would be 0.563

dpm/workday. To adjust this to ingestion intake per calendar day, 0.563 dpm/workday is multiplied by 

250 workdays per year and divided by 365 days per year, which equals 0.385 dpm/day. For the 

assumed exposure at 33 MAC, no alpha activity is apportioned to progeny so the daily uranium intake 

would be 316 dpm/day. In accordance with NIOSH 2004, the f1-value used for inhalation dose 

calculations is to be used for ingestion dose calculations.

3.8 CONSIDERATION OF BIOASSAY DATA

Predicted uranium urinalysis results, provided in Table 10, were calculated for the last day of 

assumed chronic intake periods of 30 and 60 days, 0.5 years, 1 year and extended annually 

thereafter through the end of operations, assuming the estimated inhalation and ingestion intakes of 

natural uranium were based on a uranium air concentration of 33 MAC. A cursory review of the 

highest uranium urinalysis data from facilities that handled uranium in large quantities (Mallinckrodt, 

Harshaw, Hanford, ORNL, K-25, Paducah, and Portsmouth) indicates that results exceeding 10 mg/L 

are rare and that most results are less than 1 mg/L. At the Ceramics Plant, where the first Linde 

uranium bioassays were performed after standby, two of the available urinalysis results exceeded 1 

mg/L. Subsequent results from these individuals were much lower. From November 1947 through 

January 1950, most Linde uranium urinalyses (about 95%) were less than 0.1 mg/L, but it is notable 

that exposures would likely have been lower during this period than in the earlier days of operations. 

The predicted results in Table 10 do not seem inconsistent with the limited Linde urinalyses.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 36 of 94

Table 10. Predicted uranium urinalyses from 

Ceramics Plant assumed inhalation and 

ingestion chronic uranium intake from 

November 1, 1947 to December 31, 1954

based on 33 MAC in air. 

Bioassay 

date

Type M Type S

dpm/d mg/L dpm/d mg/L

12/1/1948 566 0.3 18 0.01

12/31/1948 661 0.3 20 0.01

5/2/1949 853 0.4 28 0.01

11/1/1948 961 0.5 36 0.02

11/1/1949 1,013 0.5 48 0.02

11/1/1950 1,022 0.5 57 0.03

11/1/1951 1,026 0.5 64 0.03

11/1/1952 1,028 0.5 70 0.03

11/1/1953 1,031 0.5 74 0.03

11/1/1954 1,033 0.5 77 0.04

12/31/1954 1,033 0.5 78 0.04

*Mass results assume natural uranium exposure

Given a chronic exposure to uranium and its alpha emitting progeny at 300 MAC, the activity fraction 

of Ra-226 would be 0.196, which means that the chronic inhalation rate would be 2.7E+04 dpm/d. 

This gives a whole body activity of 2.6E+05 dpm at one year, and about 4.0E+05 dpm at 4 years

(calculated using IMBA Expert (OCAS), Version 3.2.20). The Ra-226 body activity was estimated 

using the largest breath radon result found for Linde, 2.2 pCi/L, by multiplying the radon result by a 

conversion factor of 2.52E+05 pCi/(pCi/L) (ORAUT 2005). This gives a body activity of 5.5 E+05 pCi, 

which is equal to 1.2 E+06 dpm, and is within a factor of 3 of the estimated intake from a 4-year 

chronic exposure to 300 MAC. Because other Linde radon breath analyses are lower and because a 

chronic exposure scenario may not best represent a worker’s exposure pattern, the assumption of 

300 MAC chronic exposure was believed to be adequate for reconstructing doses in the pre-1947 

research and production period, but at this time this period is reserved. 

3.9 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY 1942-1954

Intake estimates and radon exposures estimated for Linde workers are shown in the Tables 11 and 

12. Uranium bioassays are available for some workers during the period November 1947 through 

January 1950. If uranium bioassays are used to reconstruct an individual’s dose, additional intakes 

from uranium progeny might need to be added for Tonawanda Laboratories and for intakes assumed 

to occur before November 1947 at the Ceramics Plant. Table 5 lists the activity ratio of the progeny to 

uranium.

For workers, whose work location is considered indeterminate, intakes for the Ceramics Plant are 

assumed. 

For dose calculations, uranium intakes can be assumed to be U-234. Absorption types are listed in 

the table.

The dose distribution for particulate intakes is assumed to be constant. The dose distribution for 

radon is assumed to be lognormal with a GSD of 3.43. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 37 of 94

Table 11. Internal exposure summary for the Ceramics Plant, October 1, 1942, through 

December 31, 1954 

Radionuclide Start End

Intake

route

Absorption

type

Intake or 

exposure Units

U-234 10/1/1942 7/31/1946 Inhalation F, M, S Reserved dpm/day

10/1/1942 7/31/1946 Ingestion (a) Reserved dpm/day

8/1/1946 9/14/1947 Inhalation F, M, S Reserved dpm/day

8/1/1946 9/14/1947 Ingestion (a) Reserved dpm/day

11/1/1947 12/31/1954 Inhalation M, S 1.52E+04 dpm/day

11/1/1947 12/31/1954 Ingestion (a) 3.16E+02 dpm/day

Th-230 10/1/1942 7/31/1946 Inhalation M, S Reserved dpm/day

10/1/1942 7/31/1946 Ingestion (a) Reserved dpm/day

8/1/1946 9/14/1947 Inhalation M, S Reserved dpm/day

8/1/1946 9/14/1947 Ingestion (a) Reserved dpm/day

Ra-226 10/1/1942 7/31/1946 Inhalation M Reserved dpm/day

10/1/1942 7/31/1946 Ingestion (a) Reserved dpm/day

8/1/1946 9/14/1947 Inhalation M Reserved dpm/day

8/1/1946 9/14/1947 Ingestion (a) Reserved dpm/day

Po-210 10/1/1942 7/31/1946 Inhalation F, M Reserved dpm/day

10/1/1942 7/31/1946 Ingestion (a) Reserved dpm/day

8/1/1946 9/14/1947 Inhalation F, M Reserved dpm/day

8/1/1946 9/14/1947 Ingestion (a) Reserved dpm/day

Pa-231 10/1/1942 7/31/1946 Inhalation M, S Reserved dpm/day

10/1/1942 7/31/1946 Ingestion (a) Reserved dpm/day

8/1/1946 9/14/1947 Inhalation M, S Reserved dpm/day

8/1/1946 9/14/1947 Ingestion (a) Reserved dpm/day

Ac-227 10/1/1942 7/31/1946 Inhalation F, M, S Reserved dpm/day

10/1/1942 7/31/1946 Ingestion (a) Reserved dpm/day

8/1/1946 9/14/1947 Inhalation F, M, S Reserved dpm/day

8/1/1946 9/14/1947 Ingestion (a) Reserved dpm/day

Rn-222 10/1/1942 7/31/1946 Inhalation - Reservedb WLM/y

10/1/1942 7/31/1946 Inhalation - Reservedc WLM/y

8/1/1946 12/31/1954 Inhalation (d) 0.48E+00 WLM/y

a. Choose same f1-value as used for inhalation per NIOSH (2004).

b. Assign to workers who were who were likely to spend the majority of their time in process areas, or in 

boxcars or whose jobs were unknown.

c. Assign to workers who did not work or have their offices in the process buildings. 

d. For the year 1947, assume that 048 WLM/y applies to the partial exposure period.

4.0

This section develops parameters for reconstruction of doses due to external exposures due to work 

activities from October 1, 1942, the assumed start date of MED work at Linde, until December 31, 

1954, the assumed date of completion of initial cleanup at the Ceramics Plant and its turnover to 

Linde. In some parts of this section, measurements and parameters for the period after 1954 are 

analyzed because they provide information used to estimate external exposures before 1955. 

Occupational medical exposures are treated in Section 5.0. Dose reconstruction for the period after 

1954 is discussed in Section 6.0. Film badges were worn by some workers during some periods. For 

unmonitored workers and periods, doses are estimated from source term information, workplace 

measurements, and from available dosimetry results for workers.

ESTIMATION OF EXTERNAL EXPOSURE, 1942-1954
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Table 12. Internal exposure summary for the Tonawanda Laboratory, October 1, 1942, 

through December 31, 1954 

Radionuclide Start End

Intake

route

Absorption

type

Intake or 

exposure Units

U-234 10/1/1942 12/31/1946 Inhalation F, M, S Reserved dpm/day

10/1/1942 12/31/1946 Ingestion (a) Reserved dpm/day

1/1/1947 12/31/1954 Inhalation M, S 1.85E+01 dpm/day

1/1/1947 12/31/1954 Ingestion (a) 3.85E-01 dpm/day

Th-230 10/1/1942 12/31/1946 Inhalation M, S Reserved dpm/day

10/1/1942 12/31/1946 Ingestion (a) Reserved dpm/day

1/1/1947 12/31/1954 Inhalation M, S 9.04E+00 dpm/day

1/1/1947 12/31/1954 Ingestion (a) 1.88E-01 dpm/day

Ra-226 10/1/1942 12/31/1946 Inhalation M Reserved dpm/day

10/1/1942 12/31/1946 Ingestion (a) Reserved dpm/day

1/1/1947 12/31/1954 Inhalation M 9.04E+00 dpm/day

1/1/1947 12/31/1954 Ingestion (a) 1.88E-01 dpm/day

Po-210 10/1/1942 12/31/1946 Inhalation F, M Reserved dpm/day

10/1/1942 12/31/1946 Ingestion (a) Reserved dpm/day

1/1/1947 12/31/1954 Inhalation F, M 9.04E+00 dpm/day

1/1/1947 12/31/1954 Ingestion (a) 1.88E-01 dpm/day

Pa-231 10/1/1942 12/31/1946 Inhalation M, S Reserved dpm/day

10/1/1942 12/31/1946 Ingestion (a) Reserved dpm/day

1/1/1947 12/31/1954 Inhalation M, S 4.22E-01 dpm/day

1/1/1947 12/31/1954 Ingestion (a) 8.79E-03 dpm/day

Ac-227 10/1/1942 12/31/1946 Inhalation F, M, S Reserved dpm/day

10/1/1942 12/31/1946 Ingestion (a) Reserved dpm/day

1/1/1947 12/31/1954 Inhalation F, M, S 4.22E-01 dpm/day

1/1/1947 12/31/1954 Ingestion (a) 8.79E-03 dpm/day

Rn-222 10/1/1942 12/31/1946 Inhalation - Reserved WLM/y

1/1/1947 12/31/1954 Inhalation - 2.021E-01 WLM/y

a. Choose same f1-value as used for inhalation per NIOSH (2004).

Throughout this document, it is assumed that 

1 mR = 1 mrad(air) = 1 mrad(tissue) = 1 mrem = 1 mrep

Beta Radiation

In developing the beta dose rate estimates in this section, no attenuation due to shoes, apparel, and 

protective items (e.g., gloves, face shields) was considered. Attenuation due to apparel is variable 

and sometimes not very significant (see Table A-5). The types of protective items (e.g., gloves) 

issued to workers varied with time, and the extent to which workers used them is uncertain.

Workers who frequently handled significant quantities of beta-emitting materials were assumed to 

have had higher beta doses to the hands and forearms than to the remainder of the body. The 

relationships assumed between contact dose rate and dose rates to other parts of the body were 

based in part on data collected at the Ceramics Plant (see Table A-6). 
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4.1 CERAMICS PLANT BETA AND GAMMA EXPOSURE

4.1.1

Little information was available on radiation levels in Ceramics Plant buildings during periods of nonproduction. Estimates for these periods were based on measurements made after the end of 

production in Building 30, the main processing building.

Post-production Radiation in Building 30

A systematic characterization of radiation levels in Building 30 was made in 1949 in conjunction with 

its decontamination (Heatherton 1950). A survey scheme called “restricted randomization” was used. 

The floor area was divided into 15’x15’ squares each of which in turn was divided into 100 small 

squares. In each large square, four of the small squares were selected at random and surveyed. 

Total beta plus gamma radiation (mrep/h) was measured at contact at the square’s center and 

corners and at 3’ above the floor. A similar method was used to measure wall contamination. Before 

the initial survey, the building was vacuumed and flushed with water. After the initial survey, heavily 

contaminated portions of the building were removed or cleaned, mostly by sandblasting. Then the 

building was again vacuumed and flushed, and a final survey was taken. Heatherton (1950) reported 

the number of measurements and the lowest, mean, and highest values for each of several major 

plant areas. The middle columns of Table 13 display the results of a statistical analysis of the data. 

Table 13. Floor and wall radiation in Building 30 before and after the 1949-1950 

decontamination. 

Time Location

Measured beta plus gammaa Estimated

Median (mrep/h) GSD Beta (mrem/h) Gamma (mR/h)

Before vacuum cleaning and 

flushing

Contact —b —b 2.03c 0.131c

3’ —b —b 0.676c 0.131c

Pre-decontamination (just before 

decontamination and after vacuum 

cleaning and flushing)

Contact 0.719 1.67 0.675d 0.0438d

3' 0.240 1.67 0.225d 0.0438e

After decontamination

Contact 0.311 2.04 0.292d 0.0189d

3' 0.0940 2.04 0.0883d 0.0189e

a. Based on data in Heatherton 1950 Tables III and IV. The data were assumed to be characterized by a lognormal distribution. For the 

contact measurements, the mean value and a percentile value were available (the percent of results less than 1 mrep/h was stated). 

These were used with the computer program LOGNORM4 (Strom and Stansbury 2000) to estimate the median and GSD. For 

measurements at 3 feet, the mean was available but not the percentile value. The GSD determined from the contact data was 

assumed to apply and used to estimate the median.

b. Not available.

c. Assumed to be three times higher than the pre-decontamination values. The factor three is based on April 1949 data; see discussion 

in text.

d. Estimates based on beta and gamma percentages in Table 14. 

e. Assumed to equal contact gamma exposure rate. 

Another comprehensive survey of Building 30 was made in 1976 (ORNL 1978). Beta plus gamma 

levels (mrad/h) were measured at one centimeter from the surface, and gamma levels (μR/h) were 

measured at one meter. Table 14 displays the results of a statistical analysis of the data. The levels 

found were similar to the post-decontamination results of 1950, indicating little change in conditions. 

Since gamma measurements were reported, the beta and gamma percentages of the radiation could 

be estimated. These were used to estimate the beta and gamma components shown in Table 13 for 

the pre- and post-decontamination results.

The measurements just before decontamination in 1949-1950 were made after the building had 

undergone vacuum cleaning and flushing. Brief, semi-quantitative reports were available for two one- 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 40 of 94

Table 14. Building 30 floor radiation levels in 1976. 

Parameter

GMa

GSDa 

Composition of 

Value Units radiation at 1 cm

Beta plus gamma at 1 cm 0.253 mrad/h 3.47 —b

Gamma at 1 mc 0.0154 mR/h 1.69 —d

Estimated gamma at 1 cme 0.0154 mR/h —d 6.09%

Estimated beta at 1 cmf 0.238 mrad/h —d 93.9%

a. GM and GSD were calculated from data in Table 3 of ORNL 1978.

b. Estimate made in last two lines of the table.

c. Background was not subtracted out from the external gamma values. Outdoor external 

background in the Tonawanda area was reported to be 8-15 μR/h (ORNL 1978, p. 15).

d. Not applicable.

e. Estimated as equal to gamma at 1 meter. 

f. Estimated as 0.253 (beta plus gamma) – 0.0154 (gamma). 

day surveys taken in April 1949, before the vacuum cleaning and flushing (Blatz 1949; Wolf 1949). 

Typical levels of gamma radiation measured at 3’ on April 19 and at contact on April 22 were similar 

and averaged about 0.18 mR/h, about four times higher than the estimated median contact gamma 

level before decontamination. Typical levels of beta measured at contact on April 22 averaged about 

0.87 mrep/h, about 1.3 times higher than the estimated median beta level before decontamination. 

The April results were the basis of the estimate in Table 13 that beta and gamma radiation levels 

before vacuum cleaning and flushing were three times higher than the values measured afterward.

In the 1976 survey, Building 30 was the most contaminated building on the site. The levels found for 

Building 30 were assumed to apply to all buildings on the site. 

4.1.2

4.1.2.1 Pre-production, 1942-1943

External Exposure During Non-production Periods at the Ceramics Plant

For the pre-production period, it was assumed that there was no significant external beta or gamma 

exposure indoors for Ceramics Plant workers since their buildings were under construction or new. 

Some exposure while outdoors was possible since the transport of ore to Tonawanda Laboratory 

(Building 14) and the work there could have produced some site contamination. Resulting outdoor 

radiation levels are very likely to have been less than floor and wall radiation levels within Building 14 

since the indoor and outdoor sources would have been comparable but the outdoor sources could 

spread over a larger area and were subject to dispersal by the weather. The radiation levels before 

vacuuming and flushing that are estimated in Table 13 are taken as an estimate of the levels in 

Building 14 and also as an upper limit estimate of site outdoor radiation levels. Ceramics Plant 

workers were assumed to have been exposed to these levels for 0.5 h/d. Table 15 shows estimated 

annual exposure rates. 

Table 15. Pre-production beta and gamma radiation levels. 

Beta (rem/y)a Gamma (R/y)a GSDb

1.01E-01 1.97E-02 3

a. Based on Table 13 beta and gamma levels at 3’ before vacuum 

cleaning and flushing. Assumed exposure period: 0.5 h/wd, 6 

wd/wk, 50 wk/y. The beta and gamma rates are for the whole 

body.

b. Estimated on the basis of the GSD values in Tables 13 and 14. 

The beta dose rate was assumed to apply to the whole body with no added dose to the hands and 

forearms because it is unlikely that there was any significant handling of radioactive materials by 

Ceramics Plant workers in this period.
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4.1.2.2 Standby, 1946-1947

Little information is available about the status of activities during the standby period. It is likely that 

the staff on site consisted primarily of a small number of management and janitorial personnel, both of 

whom worked primarily in an office environment, and guards. For dose reconstruction, each worker 

during standby was classified as either a guard or a general worker, and worker time was assumed to 

have been spent in an office building, in production buildings, and outdoors. Averaged over the whole 

standby period, each worker's allocation of time is assumed to have been as indicated by the 

occupancy factors in Table 16. 

Gamma radiation levels were measured at six locations inside Building 30 on October 22, 1946. 

Measurements were made at one inch from the surface of interest. The results were reported as 

0 R/8 hours for four of the locations and 0.005 R/8 hours (0.625 mR/h) for the other two locations 

(each near an ore dumping grill) (Howland 1946). Since the dumping grill was one of the most 

contaminated spots in the plant, the exposure rate there was not considered typical of the conditions 

that would have been encountered upon occasional entry during standby. Instead, the indoor gamma 

and beta levels for a production building were taken as the values in Table 13 before vacuum cleaning 

and flushing. Outdoor gamma and beta levels were taken as equal to the indoor rates based on the 

reasoning used above in the discussion of the pre-production period. The gamma and beta radiation 

rates in an office building were assumed to be zero.

Table 16 summarizes the calculation of annual radiation rates based on the above parameters.

Table 16. Ceramics Plant beta and gamma radiation rates during standby. 

Parameter

Category Time-weighted radiation ratea

Office Production Outdoors beta (rem/y) gamma (R/y)

Beta (mrem/h) 0.000 0.676 0.676 —b —b

Gamma (mR/h) 0.000 0.131 0.131 —b —b

Occupancy Factor

General Worker 0.833 0.111 0.056 3.04E-01 5.91E-02

Guard 0.756 0.111 0.133 4.46E-01 8.67E-02

a. Based on 9.0 h/d exposure, 6 d/wk, 50 wk/y. Based on the underlying data and judgment, a GSD of 3 is assigned. 

The beta and gamma rates are for the whole body.

b. Not applicable.

Since there would have been little need for direct handling of radioactive materials by Ceramics Plant 

workers in this period, beta dose rate to the hands and forearms was taken as equal to the beta dose 

rate to the remainder of the body.

4.1.2.3 Cleanup, 1949-1954

Cleanup of the Ceramics Plant is discussed in Section 2.6. Floor and wall radiation levels, measured 

at the start and end of the 1949-1950 decontamination of Building 30, were available and are 

summarized in Table 13. In addition, results were available from film badge measurements of beta 

and gamma radiation during the 1948 removal of equipment from Building 30. These results are 

summarized in Table 17. Since the results in Table 17 are film badge measurements, they include 

floor and wall radiation in addition to radiation from contaminated equipment.

For external dose reconstruction, each worker is classified in one of the following categories:

• Cleanup Worker

• Cleanup Support Worker

• Non-cleanup Worker
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Table 17. Beta and gamma radiation measured during 

equipment removal, 1948. 

Beta dose rate

DD (rem/y)a GSD DM (rem/y)a GSD

1.99 1.53 0.52 1.52

Gamma exposure rate

DD (R/y)a GSD DM (R/y)a GSD

0.89 1.70 0.94 1.52

a. DD and DM are dosimeter dose and missed dose, 

respectively, as defined in NIOSH (2002), pp. 30 and 32.

A Cleanup Worker is defined as a worker who was directly engaged in the removal of radioactive 

equipment, components, or contamination (e.g., by sandblasting, vacuuming, or washing) for a 

substantial portion of the workday. This category includes equipment operators (such as a 

sandblaster operator) and laborers involved in moving and cleaning activities. A Cleanup Support 

Worker is one who spent a substantial portion of the work day in the building or area being 

decontaminated to support the cleanup activities but was not actively engaged in the removal 

activities. Examples are the plant health physicist and a stores worker stationed in a building being 

cleaned. All remaining plant personnel are considered Non-cleanup Workers. For a cleanup worker, 

the radiation estimate for both beta and gamma was taken as: 

DD + DM + OD

where DD and DM were taken from Table 17. The parameter OD is the outdoor dose rate. As was 

done for the pre-production and standby periods, it was estimated as equal to the estimated indoor 

floor and wall level at 3’ before vacuum cleaning and flushing as given in Table 13. A cleanup support 

worker was assumed to have only half as much exposure to indoor radiation. The radiation level 

estimate was taken as 

0.5*(DD + DM) + OD

for both beta and gamma. Finally, a non-cleanup worker was assumed to have only 5% of the 

exposure to gamma radiation of a cleanup worker and to have a beta exposure equal to that 

calculated for Tonawanda Plant worker’s after 1946 (numbers were not adjusted for exposure time). 

The radiation level was estimated as 

0.05*(DD + DM) + OD

for gamma radiation. The beta radiation values specified are for all parts of the body. Table 18

presents the results. 

The beta dose rates in Table 18 are based primarily on film badge measurements. The film badges 

were worn on the chest (Heatherton 1948a). The hands and forearms would have been closer than 

the chest to radioactive materials during some of the steps involved in cleanup (e.g., dismantling and 

removing equipment, scrubbing contaminated surfaces). For dose reconstruction, the beta dose rate 

to the hands and forearms of a cleanup worker was taken as three times that to the remainder of the 

body. The factor three was based on measurements discussed under the heading “Hand and 

Forearm Dose” at the end of Section 4.1.4.1.
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Table 18. External dose reconstruction parameters for cleanup. 

Period/Category

Betaa Gammab

Median (rem/y)c GSDd Median (R/y)c GSDd

7/1/49-12/31/50, 6-day week 

Cleanup Worker 2.61E+00 4.04 1.85E+00 4.04

Cleanup Support Worker 1.36E+00 4.04 9.34E-01 4.04

Non-Cleanup Worker 3.26E-01 3.00 1.11E-01 4.04

1/1/51-12/31/54, 5-day week

Cleanup Worker 2.18E+00 4.04 1.54E+00 4.04

Cleanup Support Worker 1.13E+00 4.04 7.78E-01 4.04

Non-Cleanup Worker 3.26E-02 3.00 9.26E-02 4.04

a. For the Cleanup Support and Non-cleanup workers, the indicated beta rate is to the 

whole body. For the Cleanup Worker, the rate is to all parts of the body except the 

hands and forearms. For these, the rate is three times higher. See text.

b. Gamma exposure rates are for the whole body.

c. Annual rates based on 8.5 h/wd indoors, 0.5 h/wd outdoors, 50 wk/y, except as noted for 

the non-cleanup worker.

d. The GSD for the beta value for the non-cleanup worker is the estimated GSD for outdoor 

beta dose. The other GSD values are based on the assumption that for indoor exposure 

the 95th percentile value is 10 times the median.

4.1.3

4.1.3.1 Beta 

Production, 1943-1946

No beta film badge data were found for 1943-1946. There were beta film badge data for 1947-1949; 

however, none of the data pertained to Steps I or II, as these processes were no longer in operation. 

For 1943-1946, one report of workplace beta exposure rates was found. Solid samples collected at 

the Ceramics Plant were sent on January 23, 1944 to laboratories of the Medical Section of the MED 

for measurement of beta radiation. Results were reported as “Beta Radiation Roentgens/8 Hr Day" 

(Ferry 1944b) and are reproduced here in the first five columns of Table 19. Technically, the unit 

roentgen (R) is defined only for photon radiation; beta dose should be expressed in absorbed dose 

units such as rad or rem. However, since there is an approximate numerical equivalence between 

these units, it is assumed here that the "Roentgen" unit used is equivalent to shallow dose at 0.07 mm 

in units of rem.

From the fact that no distance values are reported with the measurements and from later discussion 

of the measurements in MED correspondence (Thomas 1944), it is apparent that contact exposure

rates were reported. This is supported by the similarity of the results to contact dose rates for various 

natural uranium materials (Table A-1). 

The fourth and fifth columns of Table 19 contain the allowable times of exposure based on a 3.0 

R/week tolerance level and the actual hours of exposure per week as stated in the MED report. For 

the L-30 tailings sample, the actual exposure time was stated to be 50 hours per week, much greater 

than the allowable exposure time of 18 hours. An MED memorandum (Hadlock 1944) questioned this 

discrepancy. It brought the following MED response (Thomas 1944):

No action has been taken on the Captain Ferry's recommendation that the exposure 

time to L-30 tailings be limited to 18 hours. The primary reason for this is the fact that 

the estimated fifty hours is based on the time which a man is actually within range of 

the radiation whereas the Medical Section's allowable exposure time is based on 

contact of the tailings directly with the hands....
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 44 of 94

Table 19. Beta radiation exposure rates measured at Linde Ceramics January 23, 1944. 

From MED report of data during processing of L-30 African orea

Estimated worker 

dose rate (rem/y)d Sample # Sample location

Beta radiation

(roentgen 

/8-hr day)b

Allowable time of 

exposure per 

man (hrs/week)c 

Actual time of 

exposure per man 

(hrs/week)

13 L-30 Ore, Step I 0.1 150 50 31

9 L-30 Tailings, Step I 1.4 18 50 263

12 Barium Cake, Step I 0.8 30 0.33 2

14 A. L. Cake, Step I 1.3 18 9 73

15 PbS Cake, Step I 0.2 120 2 3

16 Soda Salt, Step I 0.5 48 9 28

10 Iron Cake, Step II 0.8 30 3 15

11 N.G. Cake, Step II 4.2 6 6 158

a. Results reported by the MED for measurements made on solid samples sent from Linde Ceramics January 23, 1944 (Ferry 1944b). 

b. Units are those stated in the MED report.

c. The MED allowable time of exposure was based on the then-effective tolerance levels of 0.5 R/day and 3.0 R per six-day week for beta 

radiation, where R denotes one "Roentgen."

d. Except for Sample #9 (L-30 tailings), the dose rate is based on an assumed 50-week year and the dose rate and actual hours of 

exposure per week stated by the MED. For the worker handling L-30 tailings, the dose rate is estimated based on information in the 

MED letter (Thomas 1944) cited in the current text section. The dose rate when the worker is "within range of the radiation" is 

estimated as half of the contact dose rate. If the source is sufficiently large and there is no shielding, a distance of 1.6 m from the 

tailings would be required to reduce the dose rate by a factor of 2 (see Table A-4). The dose estimate calculation is as follows:

Location (rem/8h) hr/wk hr/y rem/y

Hands in tailings 1.4 10 500 88

Within range of radiation 0.7 40 2,000 175

Total 263

The Linde operating and safety sections have estimated that the average time a worker 

spends with his hands in the tailings is six to eight hours one week. The maximum 

time is never over ten hours one week and even then heavy rubber gloves are worn. 

The last column of Table 19 shows worker doses derived from the data. These range from 2 to 263 

rem/y.

It is striking that the 1943-1946 dose rates indicated in Table 19 are much higher than the 1947-1949 

rates discussed in Section 4.1.4.1. For 1943-1946 Step I Process Operators, the beta dose rate is 

estimated as 263 rem/y. For 1947-1949, the job with the highest beta dose rate was that of a Step III 

Process Operator, and film badge data indicate that the beta dose rate was only about 2 rem/y 

(approximate value of annual DD+DM; see Table 30). Despite the high estimates, the 1943-1946 data 

does not appear unreasonable. The dose rates in Table 19 are typical of contact dose rates for 

uranium materials (see Table A-1). The worker exposure times were debated and reviewed within the 

MED. In the MED’s interpretation of the data, with allowances for the protective measures (e.g., 

gloves) the rates were within the limit in effect at the time, 3.0 R per six-day week (Ferry 1944b) or 

150 R/y. Therefore, the estimates in Table 19 are considered a valid basis for dose reconstruction.

4.1.3.1.1 Estimation of Worker Doses

Beta doses for specific jobs are estimated below and summarized in Table 36. 

1943-1946 Loaders, Step I Process Operators, Step II Process Operators, Ball Mill Operator, 

and Weighmaster

The calculated dose shown in Table 19 for sample 13 was assumed to apply to a Loader. The doses 

shown in Table 19 for samples 9 and 11 were assumed to be the doses to the most highly exposed 

Step I and Step II Process Operators, respectively. The dose for the most highly exposed Process 

Operator was assigned to all Process Operators in a step. A Ball Mill Operator was assigned the 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 45 of 94

same dose as a Step I Process Operator because the Ball Mill Operator performed some of the tasks 

of Process Operators when not operating the ball mill (see job description for CL-3, Chemical 

Operator C on p. 13 of Linde Air Products Company 1945). Reductions for shielding by apparel and 

containers were not taken into account in estimating doses because of insufficient information. The 

dose rates in Table 19 are based on contact measurements, so they represent dose rates to the 

hands and forearms. For all but the Loader, dose rates to the remainder of the body were taken to be 

one-third of the contact dose rates. This reduction was based on data in Table A-6 (see discussion 

under the heading “Hand and Forearm Dose” at the end of Section 4.1.4.1). For the Loader, dose 

rates for the remainder of the body were assumed to equal the contact dose rates because Loaders 

frequently hand carried ore bags, pushed carts of ore, and worked in the close vicinity of large ore 

piles.

The Weighmaster was responsible for verifying weights of all materials (incoming, product, and 

byproduct; see p. 12 of Linde Air Products Company 1945) and so had frequent close contact with 

these materials. Furthermore, a radiological survey data sheet from March 1944 indicates that at that 

time the Weighmaster was stationed for 8 hours a day at a location 4 ft from a pile of ore bags. The 

data in Table 19 and Table A-1 indicate that processed materials and byproducts had about 15 times 

the beta dose rate of ore. Therefore, the Weighmaster was assigned the same beta exposure rate as 

a Step I Process Operator. 

The dose rates in Table 19 were based on measurements on a particular batch of African L-30 ore. 

For a given type of ore, beta dose rates to ore handlers in Step I would be approximately proportional 

to ore grade (the weight percent of U3O8 in the ore). Ore type (African or domestic) would also matter. 

For the fully pre-refined ore (which the domestic ore used at Linde may have approximated) the 

electron energy release per uranium decay is only 39% of that for African ore (compare Table A-2 and 

Table A-3). Beta dose rates in Steps II and III would be independent of ore grade. Dependence on 

ore type should have been weak because most radioactive impurities would have been removed in 

Step I.

In order to use the data in Table 19 to estimate time-averaged beta dose rates applicable to the entire 

1943-1946 production period at Linde, the ratio of the average dose rates to those measured in Table 

19 was estimated. Table 20 documents the determination of the ratio. The grade range of the L-30 

ore used at Linde was 8-12% (Aerospace Corporation 1981, Table B-1). To obtain the highest ratio, it 

was assumed that the measurements in Table 19 were made on the lowest L-30 ore grade, 8%. It 

was also assumed that beta dose was proportional to electron energy released per decay and the 

worker doses were proportional to the mass of ore processed. With these assumptions, it was 

estimated that average doses would have been 0.84 of the doses predicted by using the data in Table 

19. Therefore, the estimated dose rates in Table 19 were multiplied by 0.84 to obtain time-averaged 

dose rates for 1943-1946 production. The results are in Table 21, which summarizes the results of all 

beta dose rate estimates in this section and groups the job categories into three groups (high, 

medium, and low).

1943-1946 Workers with Jobs Analogous to 1947-1949 Jobs

Workers in 1943-1946 who held jobs analogous to ones in 1947-1949 were assigned three times the 

1947-1949 beta dose rates. The factor three increase accounts for potential exposure to radiation 

from waste products from unrefined uranium ore and for the possibility that procedures in 1943-1946 

did not involve as much radiological protection. (Per Tables A-2 and A-3, unrefined uranium materials 

release approximately 2.6 times as much electron energy per uranium decay as refined uranium 

materials. The plant had a health physicist in 1947-1949 but not in 1943-1946.) Workers with jobs 

analogous to those assigned to the medium beta exposure category for 1947-1949 were assigned 

dose rates of 17.6 rem/y to the hands and forearms and 5.85 rem/y to the remainder of the body. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 46 of 94

Table 20. Dependence of beta dose rate on ore type. 

Approximate 

processing 

perioda Ore IDa Ore typea

Electron energy 

released per 

decay (MeV/nt)b Average gradea 

Average 

beta dose 

rate relative 

to 8% L-30c 

Mass 

processed 

(MT)a 

Mass weighted 

average beta 

dose rate relative 

to 8% L-30d 

6/43 to 11/43 Not stated Domestic & scrap 0.447 15.0% 0.72 1,000 0.028

12/43 to 10/44 L-30 African 1.157 10.8% 1.35 8,504 0.447

10/44 to 11/44 L-50 African 1.157 6.7% 0.84 1,486 0.048

12/44 to 2/46 L-19 Domestic 0.447 14.0% 0.68 6,102 0.161

2/46 to 7/46 R-10 African 1.157 3.5% 0.44 8,492 0.145

6/46 to 7/46 Q-20 African 1.157 17.7% 2.21 82 0.007

Sum: 25,666 0.84

a. Per Aerospace Corporation 1981, Table B-1. A metric ton is 1000 kilograms.

b. Per Table A-2. Nuclear transformation is abbreviated nt.

c. (Electron Energy Released per Decay)*(Average Grade)/8%.

d. The value for each ore ID is (Mass Processed)*(Average Beta Dose Rate Relative to 8% L-30)/(Total Mass Processed). The sum of 

the values is the approximate ratio of the average 1943-1946 ore beta dose to the ore dose due to 8% L-30 ore.

Workers with jobs analogous to those assigned to the low beta exposure category for 1949-1949 were 

assigned a beta dose rate of 3.00 rem/y to the whole body. Table 21 shows the assignments made. 

1943-1946 Workers Ore Sampler

The Ore Sampler was considered to have an exposure potential similar to that of a chemist and was 

assigned dose rates of 17.6 rem/y to the hands and forearms and 5.85 rem/y to the remainder of the 

body.

4.1.3.1.2 Outdoor Dose Rate

The average outdoor beta dose rate to which workers would have been exposed during production 

was assumed to be at most equal to the indoor level in Building 30 based on the reasoning in Section 

4.1.2.1. Therefore, the outdoor beta rate was estimated as 0.676 mrem/h (based on beta dose rate at 

3’ before vacuum cleaning and flushing given in Table 13). For 0.5 h/wd exposure, 6 wd/wk, and 50 

wk/y, the average worker exposure would have been 0.10 rem/y. This is negligible compared to the 

indoor doses given the approximate nature of the estimates and was ignored.

4.1.3.1.3 Categories

Table 21 summarizes beta dose rates assigned for the 1943-1946 production period. All are 

considered to be median values of a lognormal distribution. The dose rates derived from the 1943-

1946 data in Table 19 were judged to be based on near maximum exposure parameters. They were 

assigned a GSD of 1.52, which corresponds to a ratio of 95th percentile value to median of two. The 

other dose rates were considered to be more uncertain. They were assigned a GSD of 2.65, which 

corresponds to a ratio of 95th percentile value to median of five. 

To simplify dose reconstruction and take into account the uncertainties of the estimates, jobs were 

grouped into three categories — high, medium, and low — based on beta dose to the “remainder of 

the body” (all parts of the body except the hands and forearms). For all jobs in each category, the 

beta dose to the remainder of the body was assumed to be the highest value among the jobs in the 

category. 

Hands and Forearms Guidelines

If beta dose to the hands and forearms is needed for a particular job, it should be obtained as follows:

• For the low and medium categories, the values in Table 21 should be used.
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Table 21. Assigned 1943-1946 beta dose rates. 

Category/job 

Median dose rate (rem/y)

Hands & forearms Remainder of body GSD

High Variesa 74

Ball Mill Operator 221 74 1.52

Step I Process Operator 221 74 1.52

Weighmaster 221 74 1.52

Step II Process Operator 158 53 1.52

Loader 26 26 1.52

Medium 17.6 5.85

Chemist/Lab Technician 17.6 5.85 2.65

Engineer 17.6 5.85 2.65

Janitor 17.6 5.85 2.65

Laundry Worker 17.6 5.85 2.65

Maintenance Worker 17.6 5.85 2.65

Ore Sampler 17.6 5.85 2.65

Seamster, Seamstress 17.6 5.85 2.65

Step III Process Operator 17.6 5.85 2.65

Tool Crib Worker 17.6 5.85 2.65

Low 3.00 3.00

Draftsman 3.00 3.00 2.65

Fire Inspector 3.00 3.00 2.65

Guard 3.00 3.00 2.65

Nickel Operator 3.00 3.00 2.65

Nurse 3.00 3.00 2.65

Office Worker 3.00 3.00 2.65

Plant Superintendent, Asst Supt 3.00 3.00 2.65

Shipping & Receiving Clerk 3.00 3.00 2.65

Storekeeper 3.00 3.00 2.65

Tank Farm Operator 3.00 3.00 2.65

a. See text of Section 4.1.3.1.3.

• For the Ball Mill Operator, Step I Process Operator, and Weighmaster, the value in Table 21

should be used.

• For the Step II Process Operator, the dose to the hands and forearms should be taken as 

three times the remainder-of-the-body dose for the high category, i.e., as 221 rem.

• For the Loader, the dose to the hands and forearms should be taken as equal to the 

remainder-of-the-body dose for the high category, i.e., as 74 rem.

4.1.3.2 Gamma

Gamma film badge data are available for Step I process workers for a portion of the 1943-1946 

production period. The data covers the period January 31, 1944 through February 26, 1945 except 

for a 3-week gap (4/18/44-5/8/44) for which data were not obtained due to a film handling error (Ferry 

1944g). Usually, the badges were worn for about a week, but during the first three months, some of 

the badging periods were longer (two to four weeks). In the data analysis, each multi-week 

measurement was treated as a set of individual weekly results with each equal to the weekly average 

determined from the multi-week measurement. For example, a four-week result of 300 mR was 

treated as four individual one-week results of 75 mR.

Typically, about 20 workers were badged in each period — about 10 to 12 Loaders and 7 to 9 

processors. It is not known whether all Step I personnel were badged or only some.
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From January 31, 1944 through March 27, 1944, results were in units of roentgen per 8-hour day (the 

designations "Roentgen," "R," and "r" in different reports of the data were interpreted as synonymous). 

Afterwards, results were reported in tolerance units (designated "fractions of tolerance" or "times 

tolerance"). The MED tolerance value for gamma radiation in this period was stated as 0.1 R per 8-

hour day and also as 0.5 R/week and 0.6 R/week (Ferry 1944e; Tybout 1945a). Linde employees are 

reported to have worked a six-day, 48-hour workweek (Dupree1983a, b). For analysis of the gamma 

data, a six-day workweek was assumed. To obtain R/week, results in R/8h were multiplied by 6 and 

results in tolerance units were multiplied by 0.6.

Some of the gamma results were reported in the form "<X," where the value of X varied. The average 

of the X values was 44 mR/week. In the analysis of the data, 44 mR/week was taken as the limit of 

detection (LOD), and results reported as "<X" or zero were considered to be "<LOD results." 

However, all non-zero results, even if below 44 mR/week, were treated as "≥LOD results."

During the first part of the measuring period, the plant processed radium-containing L-30 African 

pitchblende ore. In an analysis of plant operations (Aerospace Corporation 1981, Table B-1), it was 

estimated that processing of the African ore began in December 1943 and continued through 

November 1944 and that processing of domestic ore with minimal radium content began in December

1944. The film badge data showed a sharp drop in the number of ≥LOD results during the first two 

weeks of November with the low point reached during the week ending November 12, so this was 

taken as the date when processing of the African ore ended. Exposures to workers were averaged 

separately for the periods 1/31/44 – 11/12/44 (41 weeks) and 11/13/44 – 2/26/45 (15 weeks).

The data identified each worker's job activity. The job activities were grouped into categories. Each 

category consisted of jobs judged to have similar exposure potential based on job descriptions and 

the data. Table 22 lists the job activities and the category to which each was assigned.

African Ore

Table 23 displays statistical characteristics of the film badge data for the period 1/31/44 – 11/12/44, 

when African ore was processed. For each category, a lognormal distribution was assumed and the 

following parameters were determined:

• the number of film badge measurements;

• fD = dosimeter fraction = the fraction of results in a year at or above the limit of detection 

(LOD);

• fM = missed fraction = the fraction of results in a year below the LOD;

• mD = geometric mean of all results above the LOD;

• the geometric standard deviation (GSD) of all results above the LOD based on the assumption 

that their distribution was lognormal.

The subscripts D and M denote "dosimeter dose" and "missed dose," respectively, as defined in the 

NIOSH 2002 (pp. 30 and 32). For the current set of data, the value of fD was estimated as the number 

of results ≥LOD divided by the number of measurements, and fM was estimated as the number of 

results <LOD divided by the number of measurements. (Case-by-case consideration is required when 

the number of measurements is very low. Such situations are dealt with in Sections 4.1.4.1 and 

4.1.4.2.)
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Table 22. Job categories for 1943-1946 Step I 

gamma dose analysis

Job activitya Category

Ball Mill Ball Mill Operator

Ball Mill Operator Ball Mill Operator

Cleaning up plant Cleanup

Loader Loader

Loader (handling ore) Loader

Loader Foreman Loader

Ore Sampler Ore Sampler

Sampler Ore Sampler

Barreling of tails from Moore Filter Process Operator

Chief Operator Process Operator

Digest Process Operator

Digest & V.P. Process Operator

Eimco & Prod. Precipitation Process Operator

Foreman Process Operator

Lead Removal Process Operator

Making ore digest in Pachuca Tanks Process Operator

Moore Filter Operator Process Operator

Moore Operator Process Operator

Moore tailings Process Operator

Moore Tailings/Moore Operator Process Operator

Moores Process Operator

Pachuca Digest Process Operator

Pachuca Tanks Process Operator

Product Precipitation Process Operator

V. P. Press Process Operator

V.P. Removal Process Operator

Weighmaster Weighmaster

a. Job activity descriptions are from the film badge records.

Table 23. Step I gamma exposure data for African ore processing, 1/31/44-11/12/44. 

Category

# 

Weekly

results

Dosimeter

fraction

fD

Missed

fraction

fM

For results in dosimeter fraction

Geometric mean mD

(mR/week) GSD

Ball Mill Operator 17 0.9412 0.0588 93 1.81

Loader 507 0.8383 0.1617 123 2.00

Ore Sampler 36 0.5000 0.5000 109 2.61

Process Operator 202 0.6980 0.3020 71 1.93

Weighmaster 31 0.8387 0.1613 100 2.15

The estimated median annual dosimeter dose DD for a 50-week year is shown in Table 24. This was 

calculated as follows:

DD = 50 fD mD (2)

For simplicity and because a GSD value based on small number of measurements might not be valid, 

the highest GSD in Table 23 was assigned to every job category. 

In accordance with NIOSH 2002, the annual missed dose was assumed to have a lognormal 

distribution with median DM and the 95% confidence limit DM95. For a 50-week year, these were 

calculated as follows:

DM = 50 fM (LOD/2) (3)
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DM95 = 50 fM LOD (4)

A GSD of 1.52 was calculated from DM and DM95. The results for DM are in Table 24.

Table 24. Step I gamma exposure rate parameters for 

African ore processing, 1/31/44-11/12/44. 

Category

DD

a

(R/y) GSDb 

DM

a

(R/y) GSD DD+DM

Ball Mill Operator 4.37 2.61 0.06 1.52 4.44

Loader 5.17 2.61 0.18 1.52 5.35

Ore Sampler 2.72 2.61 0.55 1.52 3.27

Process Operator 2.47 2.61 0.33 1.52 2.80

Weighmaster 4.21 2.61 0.18 1.52 4.39

a. Each dose value represents the median of a lognormal distribution.

b. Highest GSD for DD values in Table 23 is assumed to apply to all 

categories.

Domestic Ore

The analysis of exposure data for the period 11/13/44 – 2/26/45, when domestic ore with minimal 

radium content was being processed, was similar to that made for the preceding period. The results 

are displayed in Table 25 and Table 26. For routine processing of domestic ore, the Weighmaster 

and Ball Mill Operator were assigned the highest annual dose. Most of this dose is missed dose.

Table 25. Step I gamma exposure data for cleanup and domestic ore processing, 

11/13/44-2/26/45. 

Category

# 

Weekly

results

Dosimeter

fraction

fD

Missed

fraction

fM

For results in dosimeter fraction

Geometric mean mD

(mR/week) GSD

Ball Mill Operator 0 — a — a — a — a

Cleanup b 15 0.3333 0.6667 30.0 1.00

Loader 164 0.0061 0.9939 34.2 — c

Ore Sampler 11 0.0000 1.0000 0.0 — d

Process Operator 74 0.1486 0.8514 32.9 1.17

Weighmaster 11 0.1818 0.8182 35.5 1.27

a. Not available as no weekly data was reported.

b. Cleanup was an occasional short-term activity that was not part of routine domestic ore 

processing.

c. Not available as only one result was ≥LOD.

d. Not available as no results were ≥LOD.

The results obtained for this period also apply to the processing of scrap and residues from uranium 

processing. The uranium processing would have removed most of the content of radium and radium 

progeny.

4.1.3.2.1 Estimation of Worker Doses

African Ore

The annual gamma exposures of the monitored Step I workers (Table 24) were considered to be 

sufficiently similar that all could be assigned the same value. For dose reconstruction purposes, all 

were assigned an exposure rate of 5.35 R/y, the highest value of DD+DM in Table 24, and a GSD of 

2.61, the highest GSD in Table 24.

Step I Process Workers
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Table 26. Step I gamma exposure rate parameters for cleanup 

and domestic ore processing, 11/13/44-2/26/45. 

Category

DD

a

(R/y) GSD

DM

a

(R/y) GSD DD+DM

Relative

DD+DM

b

Ball Mill Operator c 0.32 1.27 0.90 1.52 1.22 28%

Cleanup d 0.50 1.27 0.73 1.52 1.23 — e

Loader 0.01 1.27 1.09 1.52 1.10 21%

Ore Sampler 0.00 1.27 1.10 1.52 1.10 34%

Process Operator 0.24 1.27 0.94 1.52 1.18 42%

Weighmaster 0.32 1.27 0.90 1.52 1.22 28%

a. Each dose value represents the median of a lognormal distribution.

b. DD+DM for domestic ore processing as percent of its value for same category for 

African ore processing.

c. There were no measured data available for the Ball Mill Operator. This category 

was assigned the same dose as Weighmaster because of the similarity of their 

exposure potential and because of the possibility that processing of some types of 

domestic ore might require handling by the ball mill operator.

d. Cleanup was an occasional short-term activity that was not part of routine 

domestic ore processing.

e. Not available.

For other workers, most of the task-related sources of gamma exposure were probably comparable to 

or weaker in effect than they were for Step I process workers. Step II and Step III process workers 

were handling refined uranium materials with much lower gamma emission rates because much of the 

radium and radium progeny had been removed. Nickel process workers were handling nonradioactive materials. In general, support personnel — such as laboratory, maintenance and janitorial 

personnel — had less frequent exposure to radium-containing materials or dealt with smaller 

quantities or worked at greater distances from the materials. They also worked in less dusty 

environments and so had less exposure to airborne radioactivity and to floor and wall contamination. 

Other Workers

One source of gamma exposure, however, was more dependent on worker location than on worker 

task. This was radiation from large quantities of ore stored in ore piles. A report of radiation surveys 

conducted March 2 and 3, 1944 (Ferry 1944c) identified two 200,000-lb piles of 10% ore and one 

300,000-lb pile of 10% ore "in the receiving room" as well as one 30 ft x 30 ft x 12 ft pile of 8% ore at 

an unspecified location (percentages specify weight % of U3O8). Based on an analysis of typical 

pitchblende ore used at Linde (Aerospace Corporation 1981, Table C-1), the ore density was 

estimated as 3.25 g/cc. Therefore, the 30x30x12 pile contained approximately 2,000,000 lbs of ore 

and was the largest of the four piles.

The probable location of the ore was the shipping and receiving platform at the south end of Building 

30 (Figure 3). This would have placed the ore near the "grizzly enclosure" through which ore was 

dumped onto the conveyor belt at the start of Step I processing. A letter to Linde reporting on a 

March 2, 1944 inspection recommended cleaning up and rearranging the platform so that ore would 

be stored in its southeast corner (Cranch 1944c), and a letter from the Ceramics Plant dated May 8, 

1944 indicated compliance with this recommendation (Rehm 1944a).

In Ferry 1944c, an exposure rate of 0.23 R/8-hr day was reported at 1.5 ft from the 30x30x12 pile. 

Potential exposures at more distant locations were estimated by mathematical analysis. For the 

region very close to an ore pile, the dependence of exposure rate on distance was determined by 

fitting data obtained in Linde pilot plant studies for a 4 ft x 5 ft x 20 ft pile of 8% ore (Skinner 1944; 

Wiesendanger 1944). The fit was obtained using the MicroShield computer code (MicroShield Team 

2003), which is based on the point kernel methodology. The values calculated by the code were 

multiplied by a single scaling factor chosen to provide a best fit to the measured data. The adjustment 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 52 of 94

compensated for simplifications in the modeling. Figure 5 shows the data and the fit obtained. The 

data covers the range 0 ft to 10 ft from the 4x20 face. The adjusted MicroShield output fit the data to 

within ±10% between 3 ft and 10 ft from the face. The results were extended to larger distances and 

to a 30x30x12 pile by more MicroShield calculations scaled with same factor. Figure 6 shows the 

MicroShield results and an analytic function fit to them. The function facilitates calculating exposures 

at arbitrary distances. As a check on the methodology, the analytic function was used to calculate the 

exposure rate at a point where it had been measured, 1.5 ft from the face. The result was 0.19 R/8 

hours, 17% below the measured value. This degree of difference is consistent with the uncertainties 

typical of this type measurement and the approximations inherent in the MicroShield code.

Exposure Rates Near 4 ft x 5 ft x 20 ft Ore Pile

0

5

10

15

20

25

30

35

40

45

0 2 4 6 8 10

Distance from 4x20 Face (ft)

Exposure Rate (mR/h)

Data

Fit

Figure 5. Fit to gamma exposure data for a 4 ft x 5 

ft x 20 ft pile of 8% African pitchblende ore. 

Gamma exposure rates measured during pilot 

plant studies at Linde Ceramics (Skinner 1944) are 

shown by the diamond-shaped symbols. The 

rectangles show the fit obtained to the data by the 

method described in the text (Section 4.2.1.2) with 

distance assumed measured along the 

perpendicular bisector of a 4x20 face. The ore 

grade (8%) was assumed to be that determined in 

the analyses reported on p. 18 of Wiesendanger 

1944. 

The calculated results are expected to overestimate radiation levels well away from the pile. The 

calculations account for air attenuation, but not for attenuation due to objects such as walls, floors, 

and equipment. Also, the calculations are based on distance along the perpendicular bisector of a 

pile face. Exposure rate at any distance will be lower for a point not on the perpendicular bisector.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 53 of 94

Calculated Exposure Rate 

Due to 30 ft x 30 ft x 12 ft Ore Pile

1.000E-05

1.000E-04

1.000E-03

1.000E-02

1.000E-01

1.000E+00

1.000E+01

1.000E+02

1 10 100 1,000 10,000

Distance from Pile Center (ft)

mR/h

Data

Fit

Figure 6. Calculated exposure rates and analytic fit for 

a 30 ft x 30 ft x 12 ft pile of 8% African pitchblende ore. 

The diamond-shaped symbols show calculated values 

of exposure rate as a function of distance from the pile 

center. Distance is measured along the perpendicular 

bisector of a 12x30 face. The values were calculated 

using MicroShield (MicroShield Team 2003). The line 

is a plot of the analytic function obtained by a leastsquares fit to the MicroShield results. The function is 

(1/r2

)*exp[a*exp(-b*r) + (c-d*r)], where a = 1.906,

b = 3.126E-02, c = 7.660, and d = 1.521E-03. The fit 

agreed with the calculated values to within ±3.0% over 

the range 16 ft to 2,000 ft from the center of the pile.

To estimate impact on the other workers in Building 30, who did not wear dosimeters, exposure rates 

due to a 30x30x12 pile located in the southeast corner of Building 30 were calculated for various 

facilities in the building. In each case, the exposure rate was calculated for the point closest to the 

center of the pile. The calculated rates are in Table 27. They range from 0.28 R/y (for the 

Laboratory) to 4.9 R/y (for the Lunchroom). The results indicate a potential for exposures to the other 

workers in Building 30 comparable to the exposures to Step I process workers.

Although few of the other workers spent the major portion of their work time near the southeast corner 

of Building 30, some may have worked near other ore piles, near tailings piles, or near large quantities 

of ore in process. Since the locations and quantities of these other sources are not known in sufficient 

detail, any worker who spent a substantial of work time in Building 30 had a potential for significant 

gamma exposure. Such workers included not only the production personnel who worked in the 

building, but also, nurses (whose office was there), maintenance personnel (who came there to 

maintain, repair, or refurbish), janitors, personnel who used locker rooms or the lunch room, and office 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 54 of 94

Table 27. Estimated maximum exposure rates in 

selected Building 30 facilities due to 30 ft x 30 ft x 12 

ft pile of 8% ore in southeast corner. 

Facility

Distance

(ft)a 

Exposure rate

(R/y)b 

Bag Washer Room 73 1.09E+00

Laboratory 128 2.81E-01

Locker & Wash Rooms 43 4.46E+00

Lunchroom 42 4.89E+00

Sampling Room 57 2.10E+00

a. Distance is the distance from the pile center to the closest 

point that is inside the facility and 3 ft above the floor.

b. Based on 2550 h/y exposure.

personnel who may have had to regularly visit the building (e.g., product accountants, industrial 

relations representatives, the fire marshal, engineers). Finally, it appears that there were office 

personnel who were permanently stationed in the building. A February 1944 letter from the plant 

addressing the frequency of required medical exams, states that it would not be appropriate to reduce 

the exam frequency for all office personnel because “we have some stenographers and clerks, stores 

clerks, etc., who work in the main building and are from time to time exposed to the same hazards as 

the operating personnel” (Holmes 1944b). In view of the likely difficulty of establishing that a 

particular worker did not spend a substantial portion of work time in Building 30, for dose 

reconstruction it is assumed that all plant personnel had a significant potential for gamma exposure. 

All Ceramics Plant employees are assigned a gamma dose of 5.35 R/y for the whole body, based on

the results for the most exposed group of monitored process workers (Table 24).

Domestic Ore

During domestic ore processing, the gamma exposure rates were lower because the ore contained 

lower proportions of radium and other uranium series progeny than African ore. However, it is difficult 

to estimate how much lower because of uncertainty as to how much African ore radioactivity remained 

in the preprocessed ores used at Linde (see Section 2.2). The value of DD+DM for Step I Process 

Operators during domestic ore processing (Table 26) is 42% of the value during African ore 

processing (Table 24). This difference is judged not to be large enough to justify the complication of 

using lower doses for periods of domestic ore processing. Therefore, for dose reconstruction during 

the 1943-1946 production period, the gamma exposure rate of 5.35 R/y is assumed to apply for the 

whole period.

4.1.3.2.2 Outdoor Exposure Rate

The average outdoor gamma exposure rate to which workers would have been exposed during 

production was assumed to be at most equal to the indoor level in Building 30 based on the reasoning 

in Section 4.1.2.1. Therefore, the outdoor gamma rate was estimated as 0.131 mR/h based on the 

level before vacuuming and flushing in Table 13. For 0.5 h/wd exposure, 6 wd/wk, and 50 wk/y, the 

average worker exposure would have been 0.020 R/y. This is negligible compared to the indoor 

exposures given the approximate nature of the estimates and was ignored.

4.1.4

4.1.4.1 Beta

Rehabilitation and Production, 1947-1949

For part of 1947-1949, weekly film badge measurements of beta and gamma exposure were available 

(Linde Database Undated B). The data had been "de-identified" by removal of personnel names 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 55 of 94

(Wallace 2003), but many of the records were labeled with a job title. The data file contained about 

6000 records dated from 1/7/48 through 12/12/49. By comparison with a copy of a laboratory film 

badge report (AEC 1949b), it was determined that the date identified the beginning of the week when 

the badge was worn. For records dated before 3/15/48, badges were processed by the University of 

Rochester (Heatherton 1948b, Osinski 1948a); thereafter, they were processed by the U.S. AEC 

NYDO Radiological Laboratory (Osinski 1948b; AEC 1949b). When use of the badges was initiated, 

the intention was for them to "be available for use by all entering a contaminated area" (Heatherton 

1947).

The computer file had three columns of data but no label indicating units or type of radiation. The 

third column was obviously the total radiation dose. The first column was determined to be beta 

radiation and the second gamma radiation by comparison with a laboratory film badge report (AEC 

1949b). The report stated both the beta and gamma units to be "exposure, mr." Progress reports by 

the Linde health physicist referred to the film badge results sometimes as "mr" and sometimes as 

"mrep" (Heatherton 1949c,d). The beta units are designated here as mrem and are assumed for 

purposes of dose reconstruction to be equivalent to shallow dose at 0.07 mm, Hp(0.07). Before 

3/15/48, a beta value below the limit of detection was reported as "<80." After 3/15/48, the limit of 

beta detection was not explicitly stated, but it was deduced to be 35 mrem from an examination of the 

data. Some of the results were reported as "0", "*" or "**". These were interpreted as meaning "less 

than the LOD." 

The beta film badge data was analyzed in order to determine dose rates for various job categories. 

Approximately one-third of the badge records were labeled with the job title NA. These were ignored 

because it could not be determined which had been worn by a worker and which had been used for 

other purposes (e.g., as controls or to measure process dose). Job titles specified in the records were 

binned into categories that combined jobs judged to have had similar exposure potential. Table 28

displays the job titles in each category. 

Table 29 displays statistical characteristics of the beta film badge data. The data were assumed to 

have a lognormal distribution. The parameters fD, fM, and mD were defined as in the analysis of the 

1943-1946 gamma dosimetry data (Section 4.1.3.2). For categories in which there were 50 or more 

badge results, fD was taken as equal to the number of results at or above the LOD divided by the total 

number of badge measurements, and fM was taken as (1 – fD). Categories with fewer than 50 results 

were examined on a case-by-case basis. If the workers in the category were rarely badged, it was 

assumed that they only occasionally entered the Linde radiation area. The value assigned to fD was 

the number of results at or above the LOD divided by 50, the number of workweeks per year. The 

missed fraction fM was taken as (1/50) times the highest number of results below the LOD for any 

member of the category. In the present case, this alternative method was applied to the categories 

Office and Superintendent.

The estimated median annual dosimeter dose DD and the median annual missed dose DM for a 50-

week year and their GSDs were determined in the same way as for the 1943-1946 gamma data (see 

Equations 2, 3, and 4 in Section 4.1.3.2). Since badges would not have been worn until workers 

arrived indoors, an outdoor beta dose rate of 0.10 rem/y (from Section 4.1.3.1.2) was added to the DM

values. For the office worker and superintendent, the outdoor contribution was larger than the indoor 

DM value, so the outdoor GSD of 3 was assigned. For the other categories, the outdoor contribution 

was not the major portion of the sum, so the GSD was taken as not changed by the addition. Table 

30 displays the results.
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Table 28. Job categories for 1947-1949 Step III film badge analysis. 

Job title Category Job title Category

AEC Office Maint G L Maintenance

Asst Engineer Engineer Maint Group Leader Maintenance

Asst Supt Superintendent Maintenance Maintenance

Carpenter Maintenance Millwright A Maintenance

Chem Oper Process Operator Millwright C Maintenance

Chem Oper A Process Operator Moveman Loader

Chem Oper B Process Operator Painter Maintenance

Chem Oper C Process Operator Painter B Maintenance

Chem Oper G L Process Operator Personnel Office

Chemist Chemist Pipefitter Maintenance

Control Engr Engineer Process Foreman Process Operator

Electrician Maintenance Prod Accounting Office

Electrician Hlpr Maintenance Property Dept Office

Fire Chief Engineer Safety Engr Engineer

Foreman (Unspecified Type) — a Seamster Seamster

Group Leader (Unspecified Type) — a Security Agent Office

Health Physicist Engineer Shift Storekpr Storekeeper

Janitor Janitor Shift Stores Att Storekeeper

Labor Foremanb Loader Shipping & Rec Storekeeper

Laundry Helper Laundry Worker Storekeeper Storekeeper

Laundry Worker Laundry Worker Superintendent Superintendent

Loader Loader Tool Crib Keeper Storekeeper

Loader Foreman Loader Trades Helper Maintenance

Mail Girl Office Truck Driver Truck Driver

Maint Maintenance Weighmaster Weighmaster

Maint Foreman Maintenance Welder Maintenance

a. There were two records with the job title Foreman and one with the job title Group Leader. They were not included in the gamma dose 

analysis because the title was not descriptive enough to identify the work activity.

b. From available data (Heatherton 1949b; AEC 1949b), it was deduced that the worker with the job title Labor Foreman was supervising 

loaders.

Table 29. 1947-1949 Step III beta dose data. 

Job

# 

Badges

Dosimeter 

fraction

fD

Missed 

fraction

fM

For results ≥ LOD

Geometric mean mD

(mrem/week) GSD

Chemist 183 0.0055 0.9945 95 — a

Engineer 243 0.0535 0.9465 113 2.12

Janitor 261 0.0421 0.9579 132 2.17

Laundry 115 0.1565 0.8435 106 1.65

Loader 304 0.0559 0.9441 78 1.47

Maintenance 913 0.0526 0.9474 146 1.62

Office 9 0.0000 0.0600 — b — b

Process operator 1,814 0.1527 0.8473 145 1.81

Seamster 44 0.0909 0.9091 141 1.31

Storekeeper 207 0.0048 0.9952 120 — a

Superintendent 3 0.0000 0.0400 — b — b

Truck driver 18 0.0000 1.0000 — b — b

Weighmaster 61 0.0492 0.9508 136 1.42

a. Only one result was ≥ LOD; GSD could not be determined.

b. No results ≥ LOD; GM and GSD could not be assigned.

For most jobs, the largest contribution to DD+DM+OD was missed dose DM. The production workers in 

1947-1949 were probably scattered among three buildings — Building 38, the location of the Step III 

processing facilities; Building 30, which contained a loading dock, locker rooms, lunch room, nurses’ 

office, and chemistry lab; and Building 31, which contained shop facilities. None of the buildings had 

been decontaminated. The average beta radiation level in Building 30 for this period is estimated in 
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Table 30. Beta dose rates for Step III, 1947-1949. 

Job

DD

a DM+ODa DD+DM +OD

(rem/y) GSD (rem/y) b (rem/y) GSD

Process operator 1.11 2.17 0.84 1.52 1.95

Laundry 0.83 2.17 0.84 1.52 1.67

Seamster 0.64 2.17 0.90 1.52 1.54

Maintenance 0.38 2.17 0.93 1.52 1.31

Weighmaster 0.33 2.17 0.93 1.52 1.26

Engineer 0.30 2.17 0.93 1.52 1.23

Janitor 0.28 2.17 0.94 1.52 1.22

Loader 0.22 2.17 0.93 1.52 1.15

Chemist 0.03 2.17 0.97 1.52 1.00

Storekeeper 0.03 2.17 0.97 1.52 1.00

Truck Driver — c — c 0.98 1.52 0.98

Office — c — c 0.15 3.00 0.15

Superintendent — c — c 0.14 3.00 0.14

a. Each DD and DM value represents the estimated median of a lognormal 

distribution. OD = estimated outdoor dose rate. 

b. The highest GSD from Table 29 is assumed for all categories for which a GSD 

can be assigned.

c. No result ≥ LOD. 

Section 4.1.1 as 0.676 mrem/h (value before vacuuming and flushing in Table 13), which is equivalent 

to 34.5 mrem/wk and 1.7 rem/y for a person spending 8.5 hours per day, 6 days per week, and 50 

weeks per year in the building. This is comparable to the film badge LOD for beta radiation (35 

mrem/wk) and so may have been a substantial portion of the missed dose. Based on this estimate of 

1.7 rem/y of possible dose and the lowest film badge DD+DM value in Table 32 of approximately 1 

rem/y for workers, who were regularly badged, it was assumed that all workers, even office workers, 

received a beta dose of at least 1 rem/y. Office workers were included because it is unknown which 

office workers spent substantial portions of their time in production buildings. Some workers who may 

have had offices in the office building may have spent much of their work day in the production area 

(e.g., engineers, health physicist, fire marshal) and some clerical workers may have actually had work 

locations in the production area (see discussion near the end of 4.1.3.2.1 and Holmes 1944b).

Categories

For dose reconstruction, the jobs were grouped into two categories — medium and low. The medium 

category was assigned a dose rate of 1.95 rem/y. Workers, whose jobs might have required frequent 

handling of radioactive materials, were placed in the medium category. The Tool Crib Attendant was

assigned to the medium category because of the possible handling of contaminated tools. The Truck 

Driver was placed in the medium category because it was the practice at the plant in 1943-1946 for 

Truck Drivers to work as Loaders when not driving a truck (Linde Air Products Company 1945, p. 10); 

it was assumed that the same practice applied in 1947-1949. Engineers and the plant health physicist 

(considered a type of engineer) were placed in the medium category because of their roles in closely 

monitoring and troubleshooting production activities. 

The low category was assigned a dose rate of 1.00 rem/y. Assigned to it were most office workers, 

the superintendent, and storekeepers. It also included guards and the fire marshal. 

A list of jobs with the category assignment of each is provided in Section 4.4. 
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Rehabilitation

Beta dose rates during rehabilitation were assumed to be the same as during production in order to 

account for exposures incurred while cleaning contaminated facilities, setting up, and testing 

equipment. 

Hand and Forearm Dose

Film badges were worn on the chest (Heatherton 1948a). Measurements during various Step III 

process operations showed that dose rates to other areas of the body differed from dose rates at the 

film badge location with the difference depending on the operation and the area of the body 

(Heatherton 1948a). The results are shown in Table A-6. Based on these data, it was estimated for 

this TBD that the average dose to the hands and forearms of a worker handling radioactive material 

was three times the chest dose and that average dose to the remainder of the body was equal to the 

chest dose. For each body area, the estimate was based on the average ratio for the operations 

studied with the result rounded up to the nearest integer. No credit was taken for wearing gloves. For 

dose reconstruction, a hands and forearm dose of 5.85 rem/y was assumed for the medium category. 

Workers in the low category were assumed to have a hands and forearm dose of 1 rem/y, the same 

as the dose assigned to the remainder of the body.

4.1.4.2 Gamma

As discussed in Section 4.1.4.1, weekly film badge measurements of beta and gamma exposure 

during 1947-1949 production were available (Linde Database Undated B). As noted, the 

measurement units were specified as "mr" or "mrep." It is assumed here that for gamma radiation the 

units were equivalent to milliroentgen (mR). Results reported as "0", "*" or "**" were interpreted as 

meaning "less than the LOD." For records date before 3/15/48, a gamma value below the LOD was 

reported as "0" or "<50." For records dated 3/15/48 or later, the limit of gamma detection was not 

explicitly stated, but it was deduced to be 45 mR from an examination of the data. A few scattered 

reports of non-zero exposures below 45 mR were judged not to reflect routine measuring capabilities.

The procedure for analyzing the gamma data and determining annual gamma exposures was the 

same as employed in the analysis of the beta data (see Section 4.1.4.1). Table 31 summarizes the 

statistical characteristics of the gamma data.

Table 31. 1947-1949 Step III gamma exposure data. 

Category

# 

Badges

Dosimeter

fraction

fD a 

Missed

fraction

fM b 

For results ≥ LOD

Geometric mean mD

(mR/week) GSD

Chemist 183 0.0164 0.9836 82 1.41

Engineer 243 0.0247 0.9753 145 2.13

Janitor 261 0.0115 0.9885 62 1.27

Laundry 115 0.0783 0.9217 81 1.46

Loader 304 0.0789 0.9211 141 2.14

Maintenance 913 0.0164 0.9836 87 1.38

Office 9 0.0200 0.0400 80 — c

Process Operator 1,814 0.0309 0.9691 86 1.52

Seamster 44 0.0000 1.0000 — d — d

Storekeeper 207 0.0048 0.9952 65 — c

Superintendent 3 0.0200 0.0400 50 — c

Truck Driver 18 0.0000 1.0000 — d — d

Weighmaster 61 0.0000 1.0000 — d — d

a. fD = fraction of measurements ≥ LOD.

b. fM = fraction of measurements < LOD.

c. Only one result was ≥ LOD; GSD could not be determined.

d. No results ≥ LOD: GM and GSD could not be assigned.
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The estimated median annual dosimeter dose DD and median annual missed dose DM for a 50-week 

year and their GSDs were determined in the same way as for the 1947-1949 beta data (see Section 

4.1.4.1). Since badges would not have been worn until workers arrived indoors, an outdoor gamma 

exposure rate of 0.020 R/y (from Section 4.1.3.2.2) was added to the DM values. Because the 

outdoor contribution was not the major portion of any of the sums, the GSD was taken as not changed 

by the addition. Table 32 displays the gamma exposure rates that were determined.

Table 32. Gamma exposure rates for Step III, 1947-1949. 

Job

DD

a DM+ODa

DD+DM +OD

(R/y)

(R/y) GSDb (R/y) GSD

Loader 0.56 2.14 1.06 1.52 1.61

Laundry 0.32 2.14 1.06 1.52 1.37

Engineer 0.18 2.14 1.12 1.52 1.30

Process Operator 0.13 2.14 1.11 1.52 1.24

Maintenance 0.07 2.14 1.13 1.52 1.20

Chemist 0.07 2.14 1.13 1.52 1.19

Janitor 0.04 2.14 1.13 1.52 1.17

Storekeeper 0.02 2.14 1.14 1.52 1.15

Seamster — c — c 1.14 1.52 1.14

Truck Driver — c — c 1.14 1.52 1.14

Weighmaster — c — c 1.14 1.52 1.14

Office 0.08 2.14 0.06 1.52 0.14

Superintendent 0.05 2.14 0.06 1.52 0.11

a. Each DD and DM value represents the estimated median of a 

lognormal distribution. OD = estimated outdoor dose rate.

b. The highest GSD from Table 31 is assumed for all categories for which 

a GSD can be assigned.

c. No results ≥ LOD. 

As in the case of 1947-1949 beta dose (see end of Section 4.1.4.1), it is to be expected that there was 

a contribution to gamma dose due to floor and wall contamination in the buildings. The average 

gamma radiation level in Building 30 for this period is estimated in Section 4.1.1 as 0.131 mR/h (value 

before vacuuming and flushing in Table 13), which is equivalent to 6.7 mR/wk and 0.34 R/y for a 

person spending 8.5 hours per day, 6 days per week, and 50 weeks per year in the building. This is 

below the film badge LOD for gamma radiation (45 mR/wk). Any clerical personnel who worked in 

production buildings (see Holmes 1944b) are assumed to have received this estimated floor and wall 

radiation plus the 0.14 R/y received by the occasionally badged office worker for a total of 0.48 R/y. 

Because it may not be easy to identify which clerical personnel worked in a production building, all 

unbadged or infrequently badged personnel are assumed to have received 0.48 R/y. 

Categories

For dose reconstruction, the jobs were grouped into two categories — medium and low — with the 

gamma category assignment for each job the same as its 1947-1949 beta category assignment (see 

subsection “Categories” in Section 4.1.4.1). Jobs in the medium gamma category were assigned an 

exposure rate of 1.61 R/y for the whole body. Jobs in the low category were assigned an exposure 

rate of 0.48 R/y for the whole body. 

Rehabilitation

Gamma exposure rates during rehabilitation were assumed to be the same as during production in 

order to account for exposures incurred while cleaning contaminated facilities, setting up, and testing 

equipment.
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4.2 TONAWANDA LABORATORY BETA AND GAMMA EXPOSURE

For external dose reconstruction, workers at Tonawanda Laboratory are classified into two categories 

— research and office. The research category includes all personnel who performed “hands on” work 

in research facilities (laboratories, fabrication facilities, or pilot plants) or who provided some type of 

support for these facilities that involved working in or very close to them. It includes scientists, 

technicians, and shop, maintenance, stores, and janitorial personnel. The office category includes all 

personnel who had primarily desk jobs although they may occasionally have visited the research 

facilities. It includes secretaries, clerks, draftspersons, and high-level managers. Three periods 

(research and development, cleanup, and post-cleanup) are considered for Tonawanda Laboratory

external dose reconstruction. Data on worker doses were not available for Tonawanda Laboratory, 

but survey results indicated that radiation levels during pilot plant operations were similar to those 

found later at the Ceramics Plant. Therefore, radiation level estimates were based to a large extent 

on the estimates for workers who did similar work at the Ceramics Plant. Table 33 shows the 

assumed annual radiation rates and their bases.

Table 33. Tonawanda Laboratory beta and gamma radiation, 1942-1954.

Period

Beta

rem/y Gamma

GSD Basisa Start End Job 

hands &

forearm

remainder

of body R/yb GSD Basisa

R&D

10/01/42 07/31/46 Research 1.11E+02 3.70E+01 1.52 A 5.35E+00 2.61 F

Office 3.00E+00 3.00E+00 2.65 B 5.35E+00 2.61 B

Cleanup

08/01/46 12/31/46 Research 7.83E+00 2.61E+00 4.04 C 1.85E+00 4.04 C

Office 1.01E-01 1.01E-01 3.00 D 1.11E-01 4.04 D 

Post-cleanup

01/01/47 12/31/54 All 3.26E-01c 3.26E-01c —d E 6.80E-02 —d E

a. Bases:

A. One half of value for most exposed Ceramics Plant process worker during 1943-6 production (Step I Process 

Operator). The factor 2 reduction allows for the lower intensity of research work in comparison with production, 

e.g., the amount of material processed and the day-to-day hands on radiation work would have been significantly 

lower than for the Step I to III operators.

B. Based on Ceramics Plant office worker during 1943-6 production.

C. Cleanup Worker during Ceramics Plant decontamination.

D. Non-cleanup Worker during Ceramics Plant decontamination.

E. Building 30 post-contamination level (Table 13) for 8.5 h/wd plus outdoor level on Tonawanda site for 0.5 h/wd. 

Outdoor level assumed equal to Building 30 level before vacuum cleaning and flushing (Table 13). A 6-day week is 

assumed.

F. Based on most exposed Ceramics Plant process worker during 1943-6 production (Step I Process Operator).

b. Where the gamma exposure rate estimate was based on dosimetry data for the Ceramics Plant (available as values of 

DD and DM) plus possibly also outdoor dose OD, the gamma exposure rate for Tonawanda Laboratory was 

approximated as the sum of the parameters (e.g., DD+DM+OD). Gamma exposure rates are for the whole body.

c. The increased beta radiation level for office workers after the end of cleanup is due in part to allowing for the possibility 

that any worker could have been stationed anywhere in the facility once it was all considered clean.

d. Not estimated.

4.3 NEUTRON EXPOSURE

No neutron exposure measurements are available for the Linde site. Neutron production by means of 

the alpha-neutron reaction would have resulted in a relatively small neutron dose component during 

Step I conversion of African ore to purified U3O8; Step II conversion of U3O8 to UO2; and Step III 

conversion of UO2 to UF4. Neutron dose rates per gram of natural uranium present in the compounds 

processed at Linde are presented in Table 34 below. 
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Table 34. Natural uranium per gram dose rates at 1 foot.

Chemical form

Dose rate 

(rem/h-gram)

African Ore (U3O8) with alpha emitting progeny in secular equilibrium through Ra-226 2.05E-11

Purified UO3 and UO2 with no alpha emitting progeny 7.91E-12

UF4/UF6 without alpha emitting progeny 6.62E-10

ORAUT-OTIB-0024 Data used to calculate the alpha-neutron dose rates is from Shleien et al (1998), Salmon and Hermann 

(1992) NRC (1991) and DOE 2000. 

The following assumptions were made regarding the conversion from neutron production rate from the 

alpha-neutron reaction to an annual dose at the receptor point. 

1. Point source geometry has been used to produce a nominal ambient neutron dose rate.

2. Point source strength is based on the daily rate of U3O8 handling or UF4 production. Selfshielding within the source-target compound matrix is assumed to be negligible.

3. A U-238/U-235 composition of 99.3% to 0.711% by weight for natural uranium.

4. The entire mass of uranium compounds is conservatively assumed to be attributable to the 

uranium (source) content. 

5. Average neutron energy from alpha-neutron reactions of 2.0 MeV. 

6. Dose equivalent rate-to-fluence rate conversion factor for 2.0 MeV of 1.3 x 10-4 rem/hour per 

neutron/cm2

-s. 

7. The work year consisted of 8 hours per day, 6 days per week, and 50 weeks per year.

During the Step I and Step II conversion processes in Building 30 (Ceramics Plant) the bounding 

neutron dose rate would be attributable to operations involving the handling of African Ore with alpha 

emitting progeny in secular equilibrium through Ra-226. This arises from the fact that the African ore 

feed material contains alpha emitting progeny that contribute to the alpha-neutron production rate. As 

indicated in Table 34, the African ore form produces neutron dose rates that are 2.6 times greater 

than the purified UO3 and UO2 with no alpha emitting progeny. In addition, during Step I processing, 

water, a neutron moderator, was added to the digestion process following ore milling. Similarly, 

during Step II processing, uranyl nitrate hexahydrate was added to ether, an organic hydrocarbon, 

and the solution was washed with water. The presence of these neutron moderators in the Step I and 

Step II processes would decrease the average neutron energy and consequently reduce the dose 

equivalent rate-to-fluence rate conversion factor. Consequently, the dose received by “Loaders” or 

“Movemen” would be bounding for all personnel engaged in Step I and Step II processing activities. 

As indicated in the Table 34, during the Step III conversion process in Building 38, the neutron dose 

rate from UF4 is almost eighty-five times greater than from the UO2 feed material. Consequently, the 

dose contribution from handling UO2 feed material can be neglected. 

As discussed in section 2.3.2, the Linde Ceramics Plant (Building 30) processed approximately 

26,000 metric tons (26 million kilograms) of ore over a period of 37 months beginning June 1943 and 

ending July 1946. Assuming a six-day workweek for 50 weeks per year, and a uniform rate of ore 

processing, 28,111 kg per workday (5,622 kg of uranium) were handled at Linde. The claimant 

favorable assumption that all of the ore was African ore has been made although 30% of the ore was 

known to be preprocessed and would contain significantly less alpha emitting progeny. The claimant 

favorable assumption that all of the ore was 20% U3O8 by weight has also been made although the 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 62 of 94

Linde site literature indicates that the ore contained 3-20% U3O8 by weight. According to the available 

Linde literature, the highest weekly production rate (Step III) of UF4 at the Linde Ceramics Plant 

(Building 38) was 41,624 pounds (Kent 1949d) or 3,147 kg per day.

The only potentially significant source of neutron exposure at Linde would have been neutrons 

produced by the alpha-neutron reaction in materials where uranium was mixed with elements of low 

atomic number such as fluorine and oxygen. Ceramics Plant (Buildings 30 and 38) personnel are 

assumed to be exposed to 1/10 of the daily production amount of U3O8 or UF4 at a distance of 1 foot. 

This quantity of material is a claimant-favorable estimate of the time-averaged amount of material 

likely to have been close to the maximally exposed worker during a work shift. The factor of 10 

reduction takes into account several factors: that the plant operated around the clock so that each 

shift dealt with only 1/3 of a day’s throughput; that many workers were involved in each type of 

operation so that each worked closely with only a portion of a shift’s throughput; that an individual 

working in the vicinity of a large quantity of material (e.g., barrels of finished product) would on the 

average have been much more than 1 foot distant because of the large volume it would have 

occupied; and that even a worker in a job that involved being close to large quantities of material also 

had other activities at larger distances from the source term. 

Major uranium-related projects at Tonawanda Laboratory were as follows:

• Research and development for Steps I, II and III. 

This work occurred primarily in the period October 1942 to December 1943 (Jenness and 

Ewing 1943). The largest scale efforts were pilot plant projects. The Step I pilot plant was 

producing 2 to 3 tons of U3O8 per week in November 1942 (Bonsib 1942). The Step III 

equipment could convert up to 90 lbs of UO2 to UF4 in a single run (Tonawanda Laboratory 

1946b, p. 84).

• Conversion of UF6 to UO3

This work appears to have been done after November 1943 (since it is not mentioned in the 

progress report describing work from October 1942 to November 1943; see Jenness and 

Ewing 1943). Interim reports were issued from July 1944 through July 1945, and a final report 

was issued in May 1946 (Tonawanda Laboratory 1946a, p. 111). Typical runs involved use of 

a 150 lb cylinder of UF6 and produced a few pounds of product. 

• Grinding UF4

The laboratory conducted research on grinding UF4 at an unknown time between October 

1942 and May 1946 (Tonawanda Laboratory 1946b, p. 89). The work involved processing 

2750 lbs of UF4. The description of the program suggests that its duration was short (a few 

weeks).

Based on the above and allowing for date uncertainties and the storage of materials before and after 

processing or research, the following model was used to estimate neutron doses at Tonawanda 

Laboratory:

• Ore containing 3 tons of U3O8 processed per week corresponding to a daily processing rate of 

454 kg from 10/1/42 through 2/29/44.

• Exposure to 100 lbs of UF4 or UF6 per week daily from 10/1/42 to 7/31/46.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 63 of 94

• This accounts for the Step III and UF6 to UO3 work.

• Exposure to 3000 lbs of UF4 for 2 months at some point between 10/1/42 and 7/31/46.

This accounts for the UF4 grinding research. The neutron dose in a year is equivalent to that 

which would be produced by exposure to 500 lbs of UF4 continuously for a year. Since the 

dates of the work are uncertain, it is assumed that there was continuous exposure to 500 lbs 

of UF4 from 10/1/42 to 7/31/46.

To reflect all types of UF4/UF6 work, a daily source term of 600 lbs (272 kg) is assumed. All source 

terms (both oxide and fluoride) are divided by 3 to account for the fact that no worker would be within 

close range of the source material at all times. This yields a daily processing rate of 151 kg of U3O8

and a continuous exposure to 90.7 kg of UF4 or UF6. Tonawanda Laboratory (Building 14) personnel 

are assumed to be exposed to these source amounts at a distance of 1 foot.

Table 35 summarizes the model parameters and presents calculated dose rates. 

Table 35. Neutron dose rates and annual doses from the alpha-neutron sources.

Location/time period Process Alpha source

Target 

atom

Daily source 

term (g)

Dose rate at 1 

foot (rem/h)

Dose rate 

(rem/y)

Building 30

6/1/43 – 7/31/46

Steps I and II African ore 20% U 

by weight

Oxygen 5.62E+6 1.15E-5 2.77E-2

Building 38

7/25/43 – 8/31/46a 

9/15/47 – 9/30/49a 

Step III UF4 Fluorine 2.76E+6 2.08E-4 5.0E-1

Tonawanda Laboratory 

(Building 14)

10/1/42 – 2/29/44

Steps I & II U3O8 Oxygen 1.51E+5 3.10E-6 7.44E-3

Tonawanda Laboratory

(Building 14)

10/1/42 – 7/31/46

Step III;

UF4 grinding;

UF6 process gas to UO3

UF4 or UF6 Fluorine 9.07E+4 6.01E-5 1.44E-1

a. The period of exposure is extended beyond the end of the production period to account for inventory left on hand. UF4 produced at Linde was 

shipped to Electromet. The dates when shipping of all Linde inventory was completed were estimated from data in a Linde progress report 

(Kent 1949e) and data on production at Electromet (DOE undated, pp. 23-4).

4.4 EXTERNAL DOSE RECONSTRUCTION SUMMARY, 1942-1954

This section summarizes guidelines and parameters for reconstruction of doses due to external 

radiation during 1942-1954 except for doses due to occupational medical exposure, which are 

addressed in Section 5.0. The parameters provided in this section are for use when individual worker 

data are unavailable or inadequate.

For dose reconstruction, exposures for different jobs have been grouped into categories. Table 36

specifies the categories and the annual radiation doses or exposures assigned to each. For the 

production periods at the Ceramics Plant, categories are designated high, medium, and low. In other 

cases, designations more descriptive of the work activity are used.

Where the dose or exposure estimates were based on dosimetry parameters (dosimeter dose DD and 

missed dose DM) plus possibly also outdoor dose OD, the dose was estimated as the sum of the 

parameters (e.g., DD +DM +OD). Each estimate is considered to be the median of a lognormal 

distribution. A GSD of 3.0 is assigned to all beta and gamma dose estimates based on the typical 

GSD levels estimated for the underlying data, i.e., beta and gamma doses are assigned as lognormal 

distributions with GSDs of 3, and an acute exposure rate. Neutron doses are assigned as a constant 

distribution and a chronic exposure rate.
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Table 36. Summary – annual external exposure from AWE operations, 1942-1954.

Yeare Work category

Beta (rem)e

Gamma (R)e Neutron (rem)e,l Hands & forearms Rest of body

Ceramics Plant (Bldgs 30, 31, 37, 38)

1942f All workers 2.55E-02 2.55E-02 4.97E-03 —a

1943g

High 1.51E+02b1

5.05E+01

3.65E+00 3.41E-01 5.05E+01 m b2

Medium 1.20E+01 3.97E+00

Low 2.08E+00 2.08E+00

1944

1945

High 2.21E+02b1

7.40E+01

5.35E+00 5.00E-01 7.40E+01 m b2

Medium 1.76E+01 5.85E+00

Low 3.00E+00 3.00E+00

1946h

High 1.28E+02b1

4.32E+01

3.11E+00 3.33E-01 4.32E+01 m b2

Medium 1.04E+01 3.59E+00

Low 1.93E+00 1.93E+00

1947i Medium 2.04E+00 8.91E-01 5.37E-01 1.48E-01m

Low 6.10E-01 6.10E-01 2.03E-01

1948 Medium 5.85E+00 1.95E+00 1.61E+00 5.00E-01m

Low 1.00E+00 1.00E+00 4.80E-01

1949j

Medium/Low 6.85E+00 2.28E+00 1.73E+00 3.74E-01m

Cleanup —a

Non-cleanup 5.47E-01 5.47E-01 2.94E-01

1950

1951

1952

1953

1954

Cleanupc 7.83E+00 2.61E+00 1.85E+00

—a

Non-cleanupd 3.26E-01 3.26E-01 1.11E-01

Tonawanda Laboratory (Building 14)

1942f Research 2.80E+01 9.33E+00 1.35E+00 3.63E-02 Office 7.56E-01 7.56E-01

1943 Research 1.11E+02 3.70E+01 5.35E+00 1.44E-01 Office 3.00E+00 3.00E+00

1944

1945

Research 1.11E+02 3.70E+01 5.35E+00 1.44E-01 Office 3.00E+00 3.00E+00

1946k Research 6.78E+01 2.26E+01 3.88E+00 8.36E-02n Office 1.78E+00 1.78E+00 3.15E+00

1947

1948

1949

1950

1951

1952

1953

1954

All workers 3.26E-01 3.26E-01 6.80E-02 —a 

a. Neutron dose rate was negligible.

b. 1. Based on 221 rem/y for Ball Mill Operator, Step I and Step II Process Operators, and Weighmaster

2. Based on 74 rem/y for Loader, per Section 4.1.3.1.3.

c. All Cleanup Workers and Cleanup Support Workers as defined in Section 4.1.2.3 are assigned to the Cleanup 

exposure category. Parameters are those of the Cleanup Worker for a six-day week in Table 18. 

d. All Non-cleanup workers as defined in Section 4.1.2.3 are assigned to the Non-cleanup exposure category. 

Parameters are those of the Non-cleanup Worker for a six-day week in Table 18. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 65 of 94

Table 36 (Continued). Summary – annual external exposure from AWE operations, 1942-1954.

e. Total annual exposure (dose) for the designated year. Prorated based on calendar year and applicable notations 

below.

f. Exposure for the period of 10/1/42 through 12/31/42 only.

g. For pre-production period of 1/1/43 through 4/26/43, assigned exposures of 3.21E-02 rem, 3.21E-02 rem and 

6.26E-03 R, respectively for the dose categories in the table.

h. For the standby period of 8/1/46 through 12/31/46, assigned beta and gamma exposures of 1.87E-01 rem, 1.87E01 rem and 3.63E-02 R, respectively for the dose categories in the table. 

i. For the standby period of 1/1/47 through 9/14/47; assigned all workers beta and gamma exposures of 3.14E-01 

rem, 3.14E-01 rem and 6.10E-02 R, respectively for the dose categories in the table. 

j. Includes exposures from the Step III period of 1/1/49 through 6/30/49. High/ Medium category worker exposures of 

2.90 rem, 9.67E-01 rem and 7.98E-01 R and Low category workers exposures of 4.96E-01 rem, 4.96E-01 rem and 

2.38E-01 R, respectively for the dose categories in the table. 

k. Includes exposures from the Cleanup period of 8/01/46 through 12/31/46; assigned beta and gamma exposures for 

Research workers – 3.28 rem, 1.09 rem and 7.75E-01 R, respectively and Office workers – 4.23E-02 rem, 4.23E-02 

rem and 4.65E-02 R, respectively for the dose categories in the table.

l. Because of the possible difficulty in determining whether a worker was working with oxide or fluoride materials, 

each worker was assigned the larger neutron dose due to fluorides.

m. The Building 38 neutron dose rate estimated for Step III processing was assumed to apply from 4/27/43 to 8/31/46 

and from 9/15/46 to 9/30/49. The neutron dose rate was negligible from 9/1/46 to 9/14/49 (standby) and after 

9/30/49 (cleanup and post-cleanup). The period of neutron exposure extended beyond the end of production in 

1946 and 1949 due to remaining inventory of UF4

n. Includes neutron exposures through 7/31/46.

Prorating of exposures that were less than a year is not straightforward for all periods. If doses are 

prorated, the footnotes to Table 36 should be checked to determine how the dose is distributed 

throughout the year. Beta doses are assigned as electrons, >15 keV. Gamma doses are assigned as 

photons, 30 to 250 keV, which is claimant favorable. Neutrons are assigned as neutrons, 0.1 to 2 

MeV.

Table 37 shows how Ceramics Plant production jobs are assigned to the various categories. For 

1943-1946, the assignments are based on Table 21 and the discussion in Section 4.1.3.1.3. For 

1947-1949, the assignments are based on Tables 30 and 32 and the discussion in Section 4.1.4.2. In 

each case, some job titles reflecting subcategories of titles in the earlier tables have been added.

For the Ceramics Plant cleanup period, the category “cleanup” includes both cleanup workers and 

cleanup support workers as defined in Section 4.1.2.3. For Tonawanda Laboratory, the categories 

“research” and “office” are defined in Section 4.2. 

If the exact job of a worker is not listed, dose reconstruction should be based on the most similar job.

5.0

5.1 CERAMICS PLANT EMPLOYEES

OCCUPATIONAL MEDICAL EXPOSURE 

Pre-employment, repeat and termination chest x-rays were required for at least some of the workers 

at Linde Ceramics. On certain occasions, pelvis x-rays were also required.

5.1.1

The earliest record found of a requirement for chest x-rays at Linde is in a report of a conference on 

worker safety at Linde held March 31, 1943 (Brimm and Neumann 1943). The conference report 

called for a pre-employment medical examination that included "fluoroscopic examination.” A 

September 4, 1943 letter from the Medical Section of the MED recommended initial and repeat 

examinations that included chest x-rays (Van Horn 1943a). Available medical records indicate that 

Chest X-rays
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Table 37 Ceramics Plant worker beta and gamma external exposure categories for 

production periods

1943-1946 production 1947-1949 production

Job

Beta

category Job

Beta and gamma

categorya

Ball Mill Operator High A Operator Medium

Chemist Medium B Operator Medium

Draftsman Low C Operator Medium

Engineer Medium Chemist Medium

Fire Inspector Low Fire Marshall Low

Guard Low First Aid Nurse Low

Janitor Medium Foreman, Labor Medium

Lab Technician Medium Foreman, Loader Medium

Laundry Worker Medium Foreman, Maintenanceb Medium

Loader/Moveman High Foreman, Step III Medium

Maintenance Workerb Medium Group Leader, Maintenanceb Medium

Nickel Operator Low Group Leader, Step III Medium

Nurse Low Guard Low

Office Workerc Low Janitor Step III Medium

Ore Sampler Medium Janitor, outside Step III Low

Plant Superintendent, Asst Supt Low Lab Technician Medium

Seamster, Seamstress Medium Laundry Man Medium

Shipping & Receiving Clerk Low Loader/Moveman Medium

Step I Process Operator High Maintenanceb Medium

Step II Process Operator High Office employee, AEC Low

Step III Process Operator Medium Office employee, Lindec Low

Storekeeper/Stock Clerk Low Shipping & Receiving Clerk Low

Tank Farm Operator Low Storekeeper/Stock Clerk Low

Tool Crib Attendant Medium Timekeeper Low

Truck Driver/Operator High Tool Crib Attendant Medium

Weighmaster High Truck Driver/Operator Medium

Weighmaster Medium

a. The category assignments are the same for beta and gamma radiation.

b. Maintenance personnel include all craftspersons who constructed, fabricated, repaired, or refurbished (e.g., 

carpenter, electrician, instrument repairman, lathe operator, mason, millwright, oiler, painter, pipe fitter, sheet 

metal worker, trades helper, and welder).

c. Office workers include accountant, bookkeeper, clerk, comptometer operator, draftsman, industrial relations 

representative, mail person, plant superintendent and assistant superintendent, product accountant, property 

department worker, secretary, security agent, stenographer, telephone operator, and typist. 

chest x-rays were regularly administered to workers in 1943 and later years. Based on the 

correspondence and the indications of plant practice found in the medical records, the assumptions 

below are provided for use in reconstruction of doses due to chest x-rays.

Linde Ceramics X-ray Assumptions

1. Applicability: These parameters apply to all

2.

job categories at the Ceramics Plant.

X-ray Category and Frequency

a.

: 

Pre-employment

b.

: Taken prior to starting employment at Linde Ceramics. These x-rays 

were also required for Linde employees transferring to Linde Ceramics from other parts of 

the company.

(1) Before August 13, 1943, taken at three-month intervals.

Repeat

(2) From August 13, 1943 through December 31, 1944, taken at six-month intervals.
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(3) From January 1, 1945 on, taken at 12-month intervals.

c. Termination

3.

: Taken upon termination of employee participation in MED/AEC work at Linde 

(either due to employee leaving or the work ending).

Methodology

a. Before July 16, 1943, by fluoroscopy of the lungs.

: 

b. On or after July 16, 1943, by taking an 11"x17" radiographic film x-ray.

4. Period

Bases of Assumptions

: The chest x-ray program is assumed to have started on November 16, 1942 and 

ended on December 31, 1954.

1. Applicability

2.

: A MED letter (Van Horn 1943a) recommended chest x-rays for “those persons 

working in direct contact with special materials," where the term "special materials" appears to 

have denoted materials containing uranium or uranium and radium (Ferry 1944a). When the 

medical examination program at the Ceramics Plant was established, it was decided that all 

employees should receive the same physical examinations and rechecks (Cranch 1944a). By 

January 1944, after reports from the early exams did not indicate any noticeable medical 

hazard, discussions were under way on reducing the frequency of examinations for those 

clerical workers, who had no occasion to go into operating areas (Cranch 1944a). The 

suggestion was made that the examination frequency be reduced for office personnel working 

in the office annex building but not for certain clerical help (described as "some stenographers 

and clerks, stores clerks, etc."), who worked in the main building, and were "from time to time 

exposed to the same hazards as operating personnel" (Holmes 1944b). A reduced frequency 

for guards was also suggested. By November of 1944, all employees were to have a preemployment examination, which included a chest x-ray. Employees "working in the plant or 

who have occasion to come in contact with any portion of the process or the materials" were 

given periodic repeat exams and a termination exam. However, employees "not having 

occasion to contact the process in any way" were exempt from the periodic and termination 

exams (Holmes 1944c). From the foregoing, it is clear that some employees were not subject 

to repeat and termination x-rays. However, it is not clear how those employees can be 

identified. Therefore, the general assumption for dose reconstruction is that all employees 

were subject to the same chest x-ray imaging requirements.

X-ray Category and Frequency

3.

: Pre-employment, repeat and termination x-rays are discussed 

in various 1943 correspondence. A three-month repeat interval was stipulated in the report of 

March 31, 1943 conference on Linde Step II safety (Brimm and Neuman 1943). A six-month 

interval was specified in a letter distributed August 13, 1943 (Cranch 1943). No medical 

records exemplifying a three-month interval were found, but the available medical records for 

1943 appear to be incomplete. Available worker files in which the first indicated chest x-ray 

date is July 1943 or later are consistent with a repeat interval of approximately six months for 

x-rays in 1944 and a repeat interval of 12 months after 1944. 

X-ray Type: The March 31, 1943 safety conference stipulated that the pre-employment exam 

should include "fluoroscopic examination." A medical record indicates chest examination by 

"fluoroscope, lungs" in a pre-employment exam on March 31, 1943 (Claim File number 

redacted, DOE Response number redacted, p. 23). A comment in the notes of a safety 

conference held May 7, 1943 indicates that the fluoroscopic examinations did not involve the 

use of film. The conference participants recommended the use of x-ray film in order to have a 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 68 of 94

permanent record of the chest condition (Brimm 1943b). The earliest chest x-ray found in 

available worker files was a radiographic film dated July 16, 1943 (Claim File number 

redacted, DOE Response number redacted, p. 28). The type of film x-rays taken at Linde 

(radiographic rather than photofluorographic) and their size (14"x17") was determined from 

several pieces of evidence:

(a) Some of the claimant files contain data sheets from the Black Rock Clinic, where Linde xrays were taken, that state the size to be 14"x17" (e.g., Claim File number redacted, DOE 

Response number redacted, p. 61).

(b) Approximately 60 arbitrarily selected Linde Air Products x-ray films in the ORO (Oak Ridge 

Organization) vaults were pulled and viewed. All were 14"x17" (Beck 2004).

(c) In 1949, Linde requested proposals for microfilming various records on cards and x-ray 

films. A response to this request indicated that 3000 14"x17" x-ray films were involved and 

mentions no other size (Medhurst 1949).

One piece of evidence suggesting the use of photofluorographic x-ray techniques was found, a 

purchase requisition dated November 4, 1944 for three gross of Type F X-ray film, 4.5"x10" 

(Linde Database 1944). The size of this film was appropriate for photofluorographic 

stereoscopic chest x-rays (Cardarelli et al 2002, p. 492). However, the following factors 

suggest it was not used for such x-rays:

(1) No photofluorographic film images were found in the Linde claim files.

(2) The purchase requisition was issued by Linde and called for delivery of the film to the 

"Laboratory." If the film were for chest x-rays, it probably would have been issued by the 

Black Rock Clinic, the only location for taking x-rays indicated in the Linde medical files 

that were examined, and delivered there.

(3) The purchase requisition stated the film was for "research on army contract #W-7405 Eng. 

26." This contract supported research at Linde's Tonawanda Laboratory (Brimm and 

Schubert 1945b) suggesting that the film was for use in such research. It is unlikely that 

the laboratory had its own worker x-ray program as other correspondence indicates a 

common program for the Ceramics Plant and the Proving Laboratory, the portion of 

Tonawanda Laboratory involved in developing uranium processing techniques (Cranch 

1943).

4. Period: November 16, 1942 is the date on which Linde entered into contract W-7401 eng-14 

for its work for the MED (Section 2.1.2). It is likely that the medical examination requirements 

were not implemented until some time after this date, but exactly how long after is not known. 

A letter to the Linde Safety Steering Committee dated August 13, 1943 indicated that preemployment exams were already a regular practice (Cranch 1943). The earliest document 

found that referred to the x-ray requirements was the report of the March 31, 1943 safety 

conference, but the program may have been established before that date. Production work at 

the Ceramics Plant is assumed to have ended on June 30, 1949; cleanup work is assumed to 

have ended on December 31, 1954. It is likely that employees who participated in the cleanup 

work were required to have chest x-rays. In the absence of specific information, it is assumed 

that other employees were also required to have chest x-rays.
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5.1.2

MED’s Medical Section’s September 4, 1943 letter to Linde recommended initial and annual pelvis xrays for workers "who will be exposed to fluorine or fluorine compounds" (Van Horn 1943a). 

However, on January 11, 1944, the MED recommended that pelvis x-rays be omitted from both preemployment and follow-up examinations (Warren 1944). To assess company practice, Linde medical 

records were checked. The review was limited in its scope because only a small portion of the 

medical records for the period were available — records of EEOICPA claimants and records in 

general Linde files because of other compensation claims that had been filed against the company. 

Eight records were found which showed x-rays during the period 9/4/43 – 1/11/44. The records 

indicated that all of the subjects had had chest x-rays, but only one had also had a pelvis x-ray (Linde 

Workmen’s Compensation File 1950, p. 67). The name of the worker for whom the pelvis x-ray was 

indicated was found in a Ceramics Plant employee list dated April 1944 (Neumann 1944). The list 

indicated that the employee worked in Step II, an activity that did not involve handling of fluorine or 

fluorine compounds.

Pelvis/Lumbar Spine X-rays

Additional information was found in Linde correspondence files. When the recommendation to end 

pelvis x-rays was relayed to the Ceramics Plant, the plant administration advised the corporate office 

“that we have never made any pelvic x-rays of any of the people working in the Ceramics Plant” 

(Holmes 1944a). The corporate office checked with the medical doctor in its Industrial Toxicology 

Department who had oversight over medical procedures at the Ceramics Plant. The doctor wrote

It is my understanding that no arrangement had as yet been made for the taking of 

pelvic x-rays of employees at the Ceramics plant. When this was discussed some time 

ago it was decided that in view of the possible fluorine exposure it might be advisable 

later, perhaps at the WCX plant, to include pelvic x-rays when our facilities would make 

this more feasible. For this reason, a space was provided in the form for “laboratory 

examinations.” ... The opinion at present is that there would be no need at the 

Ceramics Plant to include pelvic x-rays in the physical examinations. (Cranch 1944b)

These records indicate that Linde did not institute routine pelvis x-ray examination of its employees in 

response to the 9/4/43 recommendation from the MED. The indication of a single pelvis x-ray found 

in the medical records may have been related to a special situation concerning the employee. While 

dose reconstruction for the particular employee should reflect what is in the employee’s medical 

record, this particular record is not interpreted as evidence that Linde instituted a general policy of 

taking pelvis x-rays of a particular category of employees in response to the September 4, 1943 MED 

letter.

In the review of employee medical records, it was noticed that the record of a female employee had 

the notations “LMP 2/10/44” and “LMP 3/4/44” in the row on the report form reserved for recording 

pelvis x-ray results. Other pages of the employee’s record had similar monthly entries, but they were 

in a section that began in the succeeding row of the report that was designated for recording results of 

“Other Tests.” This was interpreted as indicating that the two entries on the pelvis x-ray row had been 

entered on that row in error. The entries were interpreted as recording the date of the last monthly 

period in compliance with suggestions made by the MED (Van Horn 1943b). 

Pelvis x-rays of Linde employees were taken at a later time. The Ceramics Plant progress report for 

the week ending December 21, 1947 stated, "Complying with instructions received from the health 

authorities of the Atomic Energy Commission, pelvis X-rays are being taken of all Ceramics Plant 

employees. It is expected that this program will be completed before the end of the year" (Linde 

Ceramics Plant 1948). Employee medical records confirm that x-rays of the pelvic region were taken. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 70 of 94

Images of x-ray films with the records indicate that 14"x17" films were taken and that the long 

direction of the film was parallel to the length of the body, an orientation like that used for standard 

lumbar spine x-rays and different from the 17"x14" orientation (14" parallel to the length of the body) 

use for standard pelvis x-rays. In each of the available files having copies of pelvis x-ray films, there 

were two images of the pelvic region. It was difficult to determine whether these were images of two 

portions of one x-ray exposure or whether two x-ray exposures had been taken at one session. To 

avoid underestimation of worker doses, for dose reconstruction it is assumed that two exposures were 

taken on each pelvis x-ray occasion. 

In June 1949 as the plant was nearing shutdown, it wrote the AEC to ask whether a pelvis x-ray 

examination should be included in the termination medical exam of all workers or only of workers 

whether or not they had access to contaminated areas (Kent 1949). A statement of the decision in 

this matter was not found. Employee medical records indicated that x-rays of the pelvic region were 

taken upon termination (around the end of June 1949), but the number of records is not sufficient to 

indicate whether they were taken of all employees. The x-ray film images available had 

characteristics similar to those taken at the end of 1947.

Based on the foregoing, the assumptions below are provided for use in reconstruction of doses due to 

x-rays of the pelvic region.

Assumptions

1. Applicability: These parameters apply to all

2.

job categories at the Ceramics Plant.

Occasions

a.

: Pelvis X-rays were taken at the following times:

Near the end of 1947

b.

: A pelvis x-ray was taken of every Ceramics Plant employee in 

November or December 1947.

Upon employee termination:

3.

 A pelvis x-ray was taken of every Ceramics Plant employee 

terminating after November or December 1947.

Number

4.

: Two exposures were taken at each occasion.

Methodology

5.1.3

: A 14"x17" radiographic film was taken (17" parallel to the length of the body), a 

format similar to that for lumbar spine x-rays.

Dose reconstruction should be based on information specific the subject to the extent that it is 

available and adequate. The guidelines in this section are for use when the records for an individual 

worker are not available or are incomplete. The guidelines are for use only to the extent that they are 

not inconsistent with the worker's records. For example, if the medical records are complete and 

indicate a lower or higher examination frequency than stated in the assumptions provided above, the 

data in the medical records should be used.

X-ray Dose Reconstruction Guidelines

X-ray doses shall be determined in accordance with the latest revision of the project technical 

information bulletin "Dose Reconstruction from Occupationally Related Diagnostic X-ray Procedures" 

(current version is ORAU Team 2003) when applicable. Before using the guidelines therein, dose 

reconstructors shall confirm that the assumptions on which they are based are applicable.

In relation to the above, attention is called to the following:
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 71 of 94

• For many employees, medical records are available that show the dates and types of the xrays taken.

• Often, images of the actual x-rays are also available in claimant files.

• Some chest x-rays in Linde claimant files show portions of the chin. This indicates that 

collimation was not as good as in current practice.

• X-rays of the pelvis thus far examined in Linde claimant files had the format of lumbar spine xrays. The lumbar spine x-ray doses in the x-ray technical information bulletin might be more 

suitable for these cases than the pelvis x-ray doses. 

5.2 TONAWANDA LABORATORY EMPLOYEES

The medical x-ray requirements for Ceramics Plant employees appear to have also applied to 

Tonawanda Laboratory employees who worked in pilot plant projects (Cranch 1943; Brimm 1943c). 

The extent to which the requirements were extended to other Tonawanda Laboratory employees is 

not known. In the absence of this information and in view of the likely difficulty of establishing whether 

or not a particular worker was involved in handling radioactive materials, it is assumed for dose 

reconstruction that chest x-rays were required of every Tonawanda Laboratory employee in the period 

of MED work unless it can be shown that the employee was not involved in MED projects. 

The following assumptions are made regarding the frequency of x-rays of Tonawanda Laboratory 

employees:

1. Applicability: These parameters apply to all

2.

job categories at Tonawanda Laboratory. 

X-ray Category and Frequency

a.

: 

Initial x-ray

b.

when the worker started working on an MED contract.

Repeat x-rays

c. 

taken at the same frequency as specified for Linde Ceramics.

Termination x-ray

3.

taken upon termination of employee participation in MED/AEC work at 

Linde.

Methodology:

4.

 Same as for Linde Ceramics.

Period:

Dose reconstruction guidelines are the same as for Linde Ceramics employees. 

 MED-related work at Tonawanda Laboratory is assumed to have started on October

1, 1942 and ended on December 31, 1946. 

6.0

This section develops parameters for reconstruction of doses due to internal and external exposures 

of Ceramics Plant and Tonawanda Laboratory workers after December 31, 1954, the assumed 

completion date of cleanup at the Ceramics Plant. Both facilities were on Linde’s Tonawanda, New 

York site

ESTIMATION OF EXPOSURES FROM RESIDUAL CONTAMINATION AFTER 1954
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Initial cleanup of Tonawanda Laboratory is assumed to have been completed on December 31, 1946. 

Tonawanda Laboratory workers’ radiation exposures from January 1, 1947 to December 31, 1954 are 

discussed in Sections 3.0 and 4.0. The assumed Ceramics Plant initial cleanup date is December 31, 

1954.

Beginning on January 1, 1955, It is assumed that Linde employees could have been exposed to 

residual contamination for 2000 hours per year. 

6.1.1

Airborne Radioactivity

Internal Exposure

In recent years, outdoor air radioactivity has been measured in conjunction with the site soil 

remediation program (ACE Buffalo 2004b). Table 38 displays the highest monthly average air 

concentrations of Ra-226, Th-230 and U-238 for July 2000 through June 2004. Assuming that the 

measured U-238 is part of a natural uranium source term, the total uranium concentration would be 

the U-238 concentration divided by the fraction of activity due to U-238, 0.4886, or 4.3E-04 pCi/m3

. 

The activity ratios of Th-230 and Ra-226 to uranium are 0.84 and 1.7, respectively. 

Table 38. Highest monthly outdoor 

airborne radionuclide concentrations 

at Linde Tonawanda site, 2000-2004.

Nuclide

Highest monthly 

concentration (pCi/m3)

a 

Ra-226 7.5E-04

Th-230 3.6E-04

U-238 2.1E-04

a. Per analysis of data from ACE 

Buffalo 2004b.

Building 30 was found in 1976 to be the most contaminated building on the site (ORNL 1978). The 

indoor airborne uranium concentration in 1976 was measured as 1.90E-02 pCi U/m3 (ORNL 1978, 

Table 14). This larger uranium air concentration is used to estimate a chronic intake. To maintain 

the consistent intake units and estimate the annual intake, the uranium air concentration of 1.90E-02 

pCi/m3 is multiplied by 2.22 dpm/pCi, a breathing rate of 1.2 m3

/h and 2000 work-hours/y. The annual 

intake rate is estimated at 100 dpm/y for all Tonawanda employees beginning in 1955. Dividing by 

365 days/y give a daily uranium inhalation intake rate of 0.277 dpm/day. The Th-230 and Ra-226 

daily inhalation intake rates would be 0.233 and 0.471 dpm/d, respectively. Ingestion intake rates are 

estimated using the steps described in Section 3.7. The summarized intake results are shown in 

Table 39. It is unlikely that uranium would be in a chemical form consistent with type F absorption 

during the residual contamination period, so it is assumed that only types M and S would be inhaled 

after 1954. Th-230 could be type M or S, and Ra-226 is type M. 

Workers noted that Building 30 renovations occurred in the 1960s that could have influenced air 

concentrations. Specific details of the renovation, including the actual period of renovation, dust 

control measures, location of work and occupancy of areas are not available. It is reasonable to 

assume that renovations could have resulted in elevated airborne radioactivity. Because a maximum 

value of measured air concentration is used to estimate the intakes during the residual exposure 

period, a factor of 14 is assumed to be reasonable to describe the uncertainty at the 95th percentile 

associated with the possibility of elevated intakes during Building 30 renovations. So for organ doses, 

the distribution is assumed to be lognormal with a GSD of 5. 

Other progeny are partially accounted for by the assigned radon exposure. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 73 of 94

Radon Daughters

Fifty-five measurements of radon daughter concentration were made in 1976 and 1981 in the 

Tonawanda site buildings used in MED/AEC work (ORNL 1978, pp. 17 and 84; BNI 1982, p. B-24). 

The 1981 survey was more comprehensive and yielded significantly higher concentrations, so the 

1976 results were ignored. Building 31 had the highest radon daughter concentration. The geometric 

mean PAEC was 1.68E-02 WL with a GSD of 1.89 (based on analysis of data in BNI 1982, Table B3). No correction was made for natural background radon, as its value was unknown. Based on 12 

work-months per year, the radon exposure is estimated as 2.01E-01 WLM/y, and is listed in summary 

Table 39.

6.1.2

As noted, Building 30 was found in 1976 to be the most contaminated building on the site (ORNL 

1978). Floor and wall radiation levels measured in Building 30 in 1976 were compared with similar 

measurements made in 1950 immediately after its decontamination (see Section 4.1.1 and 

Heatherton 1950). The results were similar, but the 1950 values were slightly higher. 

External Beta and Gamma Exposure

Outdoor gamma radiation levels at one meter above the surface on and near the Tonawanda site 

were measured in 1976 and 1981. Natural background gamma radiation levels in the Tonawanda 

area were said to be 8 to 15 μR/h (ORNL 1978, p. 15; BNI 1982, p. B-12). ORNL 1978 (p. 15) 

reported only isolated areas with levels significantly higher than background; readings in these areas 

were described as very non-uniform and varying from background up to 250 μR/h. Regions with 

readings above about 20 μR/h were indicated in a figure in the report (Fig. 26). Figure B-5 of BNI 

1982 summarized the combined results of the two studies. Locations and values were shown for 

readings ≥25 μR/h. BNI 1982 (p. B-12) stated that the points of maximum radiation determined in the 

ground surveys were slightly displaced from but in general agreement with those determined in a 

1979 aerial radiological survey. The total number of readings ≥25 μR/h reported by BNI was 16. The 

net readings (after subtraction of 8 μR/h to correct for background) had a geometric mean of 94.0

μR/h and a GSD of 3.95. This was taken as an estimate of worker exposure rate when outdoors. 

This estimate was assumed to apply from January 1, 1955 to the present (2005). No credit was taken 

for the soil remediation at the site that began in 2000, as the remediation has not been completed

(Pilon 2004). The estimate is probably an overestimate for the following reasons: measurements 

were reported only for areas identified as having above-background radiation levels; these areas 

occupied much less than half of the open area of the site (based on Fig. B-5 of BNI 1982); and the 

reported measurements were not averages over these areas but only readings at a few hot spots. 

The beta dose rate at 3’ above the ground corresponding to the adopted value of gamma exposure 

rate was estimated as 4.38E-01 mrem/h, 4.66 times the gamma rate, where 4.66 is the ratio of beta 

mrem/h at 3’ to gamma mR/h at 3’ for the floors and walls in Building 30 after decontamination (per 

Table 13). It is assumed that the GSD for the beta dose rate would be the same as for the photon 

dose rate.

Because the radiation levels seemed to remain fairly constant, and because the levels were fairly low, 

the Tonawanda Laboratory post-cleanup exposures, based on Building 30 contamination levels, were 

used to estimate the external exposure rate. No adjustments for changes in work-hours were made.

Table 39 summarizes the results. The radiation energy distributions are assumed to be same as those 

during the operational period.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 74 of 94

6.2 NEUTRON EXPOSURE

Because only small quantities of radioactive material were on the site in the post-cleanup period, 

there were no significant neutron exposures. 

6.3 DOSE RECONSTRUCTION SUMMARY, 1955 TO PRESENT

This section summarizes guidelines and parameters for reconstruction of doses due to internal and 

external exposure beginning in 1955. 

Table 39. Annual internal and external exposure to residual radioactivity. 

Internal Startb Exposure Absorption type Intake (dpm/d) IREP distribution

U-234 1/1/1955 Inhalation M, S 2.77E-01 Lognormal GSD 5

1/1/1955 Ingestion (a) 5.78E-03 Lognormal GSD 5

Th-230 1/1/1955 Inhalation M, S 2.32E-01 Lognormal GSD 5

1/1/1955 Ingestion (a) 4.84E-03 Lognormal GSD 5

Ra-226 1/1/1955 Inhalation M 4.84E-01 Lognormal GSD 5

1/1/1955 Ingestion (a) 1.01E-02 Lognormal GSD 5

WLM/y

Rn-222 1/1/1955 Inhalation - 2.01E-01 Lognormal GSD 1.89

External Start Exposure Basis

1/1/1955 Penetrating Survey instrument 0.068 R/y Lognormal GSD 3

1/1/1955 Non-penetrating Survey instrument 0.326 rem/y Lognormal GSD 3

a. Choose same f1-value as used for inhalation per NIOSH (2004).

b. No end date is set, because there is no evidence that the site is restricted or has been remediated. 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 75 of 94

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1949, ORAU 88/A-16, Oak Ridge Associated Universities, January [Linde Air and Ceramics, 

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Processing at Linde (written 1988)-1943-1949.pdf].
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 81 of 94

Holmes, A. R. 1944a, untitled letter to S. P. Murphy, Linde New York office, from Linde Air Products 

Company, Ceramics Plant, Tonawanda, NY, January 26 [Linde Air and Ceramics, NY\05-25-

04 - Data Capture\25000090 p. 11].

Holmes, A. R. 1944b, "Physical Examination Program – Ceramics Plant," letter to S. P. Murphy, Linde 

New York office, from Linde Air Products Company, Ceramics Plant, Tonawanda, NY, 

February 10 [Linde Air and Ceramics, NY\05-25-04 - Data Capture\25000090 p. 7].

Holmes, A. R. 1944c, untitled letter to E. L. Van Horn, Area Engineer, Tonawanda Area, United States 

Engineer's Office, P. O. Box 8, Kenmore, NY, from Linde Air Products Company, Ceramics 

Plant, Tonawanda, NY, November 14 [Linde Air and Ceramics, NY\05-25-04 - Data 

Capture\25000090 p. 5].

Howland, J. W., 1946, "Possible Hazards at Linde Ceramics and Electro-Metallurgical Plants," 

memorandum to the Area Engineer, Madison Square Area, New York, NY, from U.S. Engineer 

Office, Rochester Area, Rochester, NY, November 19 [Linde Air and Ceramics, NY\01-6-04 

Data Capture\240000580].

Hunter, D., 1949, "Record of Negotiations for Supplemental Agreement to Contract W-7405 Eng-276, 

Harshaw Chemical Company," memorandum to J. C. Clarke, September 26 [Linde Air and 

Ceramics, NY\Linde Air Products.pdf, p. 13].

ICRP (International Commission on Radiological Protection), 1981, Limits for Inhalation of Radon 

Daughters by Workers, Publication 32, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, ICRP Publication 66, Pergamon Press, Oxford, England.

Jenness, L. G. and R. Ewing, 1943, Summary of Research, Development, and Pilot Plant Operation 

at the Laboratory of the Linde Air Products Company, Tonawanda, New York, October 1942 to 

November 1943, Linde Tonawanda Laboratory, transmitted by letter from J. F. Eversole to P. 

B. Pew dated November 11 [Linde Air and Ceramics, NY\08-07-03 - Data 

Capture\010001218\000001, pp. 29-33].

Kent, E. C., 1947, "Plant Progress Report, September – 1947," letter to A. R. Holmes, from Ceramics 

Plant, Tonawanda, NY, October 1 [Linde Air and Ceramics, NY\10-16-03 - Data 

capture\2400153-A.pdf, pp. 212-220].

Kent, E. C., 1949, untitled letter to U.S. Atomic Energy Commission, Tonawanda Sub-Office, 

Kenmore NY, for attention of F. J. Epp, from Linde Air Products Company Ceramics Plant, 

October 10 [Linde Air and Ceramics, NY\05-25-04 - Data Capture\25000091, p. 20].

Kent, E.C., 1949d, “Ceramics Plant Progress Report for Week Ending March 27, 1949,” Linde Air 

Products Ceramics Plant, March 28 [\Linde Air and Ceramics, NY\Linde Ceramics NY.pdf. p. 

146].

Kent, E.C., 1949e, “Ceramics Plant Progress Report for Week Ending July 24, 1949,” Linde Air 

Products Ceramics Plant, July 26 [\Linde Air and Ceramics, NY\Linde Ceramics Plant.pdf, p. 

32].
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 82 of 94

Klevin, P. B., 1949a, Occupational Exposure to Radioactive Dust, Linde Air Products, Medical 

Division, New York Operations Office, U.S. Atomic Energy Commission, January 25 [Linde Air 

and Ceramics, NY\07-22-03 - Data Capture 2-Linde\ 

Eml2Bx24LindeAirProdCoOccExpRadioDustJan1949.pdf].

Klevin, P. B. 1949b, Occupational Exposure to Radioactive Dust, Linde Air Products, Medical Division, 

New York Operations Office, U.S. Atomic Energy Commission, August 1 [Linde Air and 

Ceramics, NY\07-22-03 - Data Capture 2-Linde\ 

Eml2Bx24LindeAirProdCeramicsPlOccExpRadDustAug1949.pdf].

Linde, 1945, Linde Industrial Hazard Bulletin, March 7 [Linde Air and Ceramics, NY\10-16-03 - Data 

capture\240000328, p. 16].

Linde Air Products Company, 1945, Job Descriptions Ceramics Plant, December 7 [Linde Air and 

Ceramics, NY\08-07-03 - Data Capture\010001218, p. 115].

Linde Air Products Company Ceramics Plant, 1946a, Step I Construction, Process and Operation 

Report, Contract W-7401-eng-14, October 1 [Linde Air and Ceramics, NY\08-07-03 - Data 

Capture\010001220\000001]

Linde Air Products Company Ceramics Plant, 1946b, Step II Construction, Process and Operation 

Report, Contract W-7401-eng-14, October 1 [Linde Air and Ceramics, NY\08-07-03 - Data 

Capture\010001220\000004]

Linde Air Products Company Ceramics Plant, 1946c, Step III Construction, Process and Operation 

Report, Contract W-7401-eng-14, October 1 [Linde Air and Ceramics, NY\08-07-03 - Data 

Capture\010001220\000005]

Linde Ceramics Plant, 1948, Ceramics Plant Progress Reports for the weeks ending February 29, 

1948 through April 11, 1948 [Linde Air and Ceramics, NY\Linde Ceramics NY.pdf, pp. 35-57].

Linde Ceramics Plant Airborne Dust Uranium Data, 1943-1946, miscellaneous reports of dust uranium 

concentrations at Linde Ceramics Plant: 

Linde Air and Ceramics, NY\08-07-03 - Data Capture\010001218, pp. 83-85;

Linde Air and Ceramics, NY\10-16-03 - Data capture\240000280, pp. 3-30;

Linde Air and Ceramics, NY\10-20-03 - Data Capture\240000379, pp. 6-7;

Linde Air and Ceramics, NY\10-21-03 - Data Capture\240000226, p.4;

Linde Air and Ceramics, NY\10-21-03 - Data Capture\240000230, p.4; 

Linde Air and Ceramics, NY\10-21-03 - Data Capture\240000253, p.10; 

Linde Air and Ceramics, NY\10-21-03 - Data Capture\240000255, pp. 6-8.

Linde Ceramics Plant Breath Radon Data, 1944-1945, miscellaneous reports of breath radon 

measurements for Linde Ceramics' workers:

Linde Air and Ceramics, NY\10-20-03 - Data Capture\240000371, p. 3; 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 83 of 94

Linde Air and Ceramics, NY\10-20-03 - Data Capture\240000378, p. 3, 5;

Linde Air and Ceramics, NY\10-22-03 - Data Capture\240000348, pp, 3-8;

Linde Air and Ceramics, NY\10-22-03 - Data Capture\240000357, pp. 3-5;

Linde Air and Ceramics, NY\01-6-04 Data Capture\240000582, pp. 3-7.

Linde Ceramics Plant Radon Data, 1944-1946, miscellaneous reports of radon concentrations at 

Linde Ceramics plant: 

Linde Air and Ceramics, NY\01-6-04 Data Capture\240000583, pp. 17, 20, 22, 31, 32, 34-36, 

38, 39, 42, 44, 45;

Linde Air and Ceramics, NY\01-6-04 Data Capture\240000584 pp. 23, 24, 33, 34, 39;

Linde Air and Ceramics, NY\10-20-03 - Data Capture\240000370, pp. 5-7; 10, 11, 13

Linde Air and Ceramics, NY\10-20-03 - Data Capture\240000378, pp. 7, 10, 11, 13, 14;

Linde Air and Ceramics, NY\10-20-03 - Data Capture\240000379, pp. 2-13;

Linde Air and Ceramics, NY\10-20-03 - Data Capture\240000404, pp. 13-14;

Linde Air and Ceramics, NY\10-21-03 - Data Capture\240000253, pp. 5, 8, 13, 16, 19, 22, 26;

Linde Air and Ceramics, NY\10-21-03 - Data Capture\240000255, p. 7.

Linde Ceramics Plant Urinalysis Data, 1947-1950, miscellaneous reports of uranium and fluorine 

urinalysis data for Linde Ceramics' workers 

Linde Air and Ceramics, NY\03-09-04 - Data Capture\260000120, pp. 1-29; 

Linde Air and Ceramics, NY\01-22-04 - Data Capture\260000009, pp. 1-48.

Linde Database Undated A, layout of Building 30 showing Linde Step I and II areas [Linde Air and 

Ceramics, NY\07-22-03 - Data Capture 2-Linde\Eml2Box25LindeAirProducts46_50, p. 16].

Linde Database Undated B, file linde_fb_sep84 [Linde Air and Ceramics, NY\08-20-03 - Data 

Capture\010001335].

Linde Database, 1944, Requisition for Material, Requisition Number 434898, November 6 [Linde Air 

and Ceramics, NY\10-16-03 - Data capture\240000141, p. 3]

Linde Database, 1945, Property Location Plan, Ceramics Plant Area, Drawing AMX-1__ (rest of 

designation not readable), April 10 [Linde Air and Ceramics, NY\03-05-04 - Data Capture Air 

Products\260000048 p. 38]

Linde Workmen’s Compensation File, 1950, file containing medical and other records related to a 

workmen’s compensation claim against Linde [Linde Air and Ceramics, NY\07-22-03 - Data 

Capture 2-Linde\Linde Dosimetry Data\EmlBox2LindseyAirProducts\000001.pdf].
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 84 of 94

Marshall, J. C., 1943, "Contractor: The Linde Air Products Company, New York, NY; Contract for: 

Design, Engineering, Construction, Equipping, and Operation of a plant for the production of 

various products," Corps of Engineers, Contract No. W-7401 eng-14, January 18 [Linde Air 

and Ceramics, NY\Linde Ceramics Plant.pdf p. 345].

Martin, J. L., 1945, untitled letter, to E. W. Moscript (U.S. Army Engineer, Madison Square Area, New 

York, NY), from Linde Air Products Company, November 16 [Linde Air and Ceramics, 

NY\Linde Ceramics Plant.pdf p. 300].

MED (Manhattan Engineer District), Undated A, Data Sheets for Industrial Hazard Rating, Linde 

Ceramics Plant, Substance: Ore, Medical Section [Linde Air and Ceramics, NY\10-16-03 - 

Data capture\240000279 pp. 3-45].

MicroShield Team, 2003, MicroShield Version 6 User's Manual, Grove Engineering, Rockville, MD 

20850.

Medhurst, A. E., 1949, untitled letter, to M. Trost, Linde Air Products Company, Kenmore, NY, from 

Langtry Photo Service, Buffalo, NY, August 12 [Linde Air and Ceramics, NY\05-25-04 - Data 

Capture\25000091 p. 28-29]

National Lead Company of Ohio (NLO), 1952, Production Report on the Rolling of Two Hundred and 

Eighteen Uranium Billets at Bethlehem Steel Corporation’s Lackawanna Plant on March 15, 

1952, Contract No. AT(30-1)-1156, March 21.

NCRP (National Council on Radiation Protection and Measurements), 1971, Protection Against 

Neutron Radiation. NCRP Report No. 38, Washington, D.C. 

Neumann, F. L., 1944, untitled letter (listing employees holding various jobs) to W. G. Thomas 

(Operations Unit, Kenmore NY), from Linde Air Products Company, Ceramics Plant, 

Tonawanda NY, April 26 [Linde Air and Ceramics, NY\10-16-03 - Data capture\240000151].

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 1, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August.

NIOSH (National Institute for Occupational Safety and Health), 2004, Estimation of Ingestion Intakes, 

OCAS-OTIB-009. Revision 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

April 13.

NRC, (Nuclear Regulatory Commission), 1991, Passive Nondestructive Assay of Nuclear Materials.

NUREG/CR-5550. Washington, D.C. [http://www.nis5.lanl.gov/panda_manual.htm accessed 

on 12/16/04] 

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; Draft Report for Comment, Washington, D.C.

Olevitch, E., 1944, Processing L-30 Ore in Step I, internal report, Linde Ceramics Plant, July 15 [Linde 

Air and Ceramics, NY\08-07-03 - Data Capture\010001220\000003 p. 40].
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 85 of 94

ORAU (Oak Ridge Associated Universities) Team, 2003, Technical Information Bulletin: Dose 

Reconstruction from Occupationally Related Diagnostic X-ray Procedures, NIOSH Dose 

Reconstruction Project, ORAUT-OTIB-0006, Rev. 02. 

ORAUT (Oak Ridge Associated Universities Team), 2005, Technical Information Bulletin: Estimation 

of Radium-226 Activity in the Body from Breath Radon-222 Measurements, NIOSH Dose 

Reconstruction Project, ORAUT-OTIB-0025, Rev. 00, April 5.

ORNL (Oak Ridge National Laboratory), 1978, Formerly Utilized MED/AEC Sites Remedial Action 

Program, Radiological Survey of the Former Linde Uranium Refinery, Tonawanda, New York, 

Final Report, DOE/EV-0005/5 UC-70, May. [Linde Air and Ceramics, NY\12-06-04- Data 

Capture\030001717.pdf]

Osinski, V., 1948a, “Ceramics Plant Progress Report for the Week Ending on January 4, 1948,” Linde 

Ceramics Plant, January 6 [Linde Air and Ceramics, NY\Linde Ceramics NY.pdf, p. 13].

Osinski, V., 1948b, “Ceramics Plant Progress Report for the Week Ending March 21, 1948,” Linde 

Ceramics Plant, March 22 [Linde Air and Ceramics, NY\Linde Ceramics NY.pdf, p. 44].

Peterson, H. T., 1983, "Terrestrial and Aquatic Food Chain Pathways," Chapter 5 of Radiological 

Assessment, by J. E. Till and H. R. Meyer, U.S. Nuclear Regulatory Commission, Washington, 

D.C., NUREG/CR-3332, ORNL-5968, September.

Pew, P. B., 1943, untitled letter to G. Vanden Buick, War Department, Manhattan District, U.S. 

Engineer Office, New York, NY," July 20 [Linde Air and Ceramics, NY\08-07-03 - Data 

Capture\010001221\000014, pp. 1-5]

Pilon, R. L., 2004, untitled attachment to "RE: Request or Information on Linde Site Cleanup and 

Remediation," e-mail to Gerald R. Davidson, from U.S. Army Engineer District – Buffalo, 

Buffalo, NY, October 8. [Linde Air and Ceramics, NY\12-06-04- Data Capture\030001725.pdf]

Rehm, C. W., 1944a, untitled letter to S. P. Murphy, Linde New York office, from Linde Ceramics 

Plant, May 6 [Linde Air and Ceramics, NY\08-07-03 - Data Capture\010001228\000007, p. 1].

Rehm, C. W., 1944b, untitled letter to E. L. Van Horn, Area Engineer, Tonawanda Area, Kenmore, 

NY, from Linde Ceramics Plant, June 19 [Linde Air and Ceramics, NY\08-07-03 - Data 

Capture\010001218, pp. 83-85].

Rennich, G., 1947, memorandum to E. M. Velten (Production Division), “Uranium-Thorium Branch 

Semi-Annual Report to Congress,” AECPR-21 NY, U.S. Atomic Energy Commission, 

December 1 [Linde Ceramics Plant.pdf, pp. 8-11].

Robinson, B. W., 1945, "Sampling and Weighing of Material Purchased from LAP Under Contract W17-028 eng 29," memorandum to files, Manhattan Engineering District, August 20 [Linde Air 

and Ceramics, NY\Linde Ceramics Plant.pdf p. 297].

Salmon, R., and O.W. Hermann, 1992, ALPHN: A Computer Program for Calculating (α, n) Neutron 

Production in Canisters of High-Level Waste, ORNL/TM-12239; October. 

[http://www.ornl.gov/info/reports/1992/3445603684010.pdf].
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 86 of 94

Shleien B., L. A. Slaback Jr., and B. K. Birky, 1998, Handbook of Health Physics and Radiological 

Health, Third Edition, Baltimore: Williams & Wilkins.

Skinner, G. M., 1944, Radiation Measurements in the Proving Laboratory During Processing of 

African Pitchblende Ore, Tonawanda Laboratory, Linde Air Products Company, Research 

Notes No. 275, February 11 [Linde Air and Ceramics, NY\10-16-03 - Data 

capture\240000137].

Strom, D. J. and P. S. Stansbury, 2000, "Determining Parameters of Lognormal Distributions from 

Minimal Information," American Industrial Hygiene Association Journal, volume 61, pp. 877-

880 [____General Information____\08-23-04 - Data Capture\030000999].

Thomas, W. G., 1944, "Radiation Hazards," EIDM-T-12, memorandum to office of the Area Engineer, 

Madison Square Area (Attention C. Hadlock), from office of the Area Engineer, Tonawanda, 

March 8 [Linde Air and Ceramics, NY\10-21-03 - Data Capture\ 240000255, p. 24].

Tonawanda Laboratory, 1942, Laboratory Directory, Tonawanda Laboratory, Tonawanda, NY, 

September 2 [Linde Air and Ceramics, NY\08-07-03 - Data Capture\010001222, p. 23]

Tonawanda Laboratory, 1944, “MD Personnel,” Tonawanda Laboratory, Tonawanda, NY, September 

20 [Linde Air and Ceramics, NY\10-16-03 - Data Capture\240000142].

Tonawanda Laboratory, 1946a, Final Report on Conversion of K-25 Process Gas to Uranium Trioxide, 

for Contract No. W-7405 Eng. 26, Sub No. 2, Linde Air Products Company, May 3 [Linde Air 

and Ceramics, NY\08-07-03 - Data Capture\010001220\000009].

Tonawanda Laboratory, 1946b, Final Report in Accordance with the Terms of Contract No. W-7401–

eng-14 – Ceramics Section, Linde Air Products Company, May 10 [Linde Air and Ceramics, 

NY\10-16-03 - Data capture\240000145].

Tybout, R. A., 1944a, "Visit to Tonawanda Area – 18 and 19 October 1944," memorandum to the Area 

Engineer, Madison Square Area, New York, NY (Attention: Lt. J. Howland), October 20 [Linde 

Air and Ceramics, NY\10-21-03 - Data Capture\240000255, p. 21].

Tybout, R. A., 1944b, memorandum to the District Engineer (U.S. Engineer Office, Oak Ridge, 

Tennessee), "Results of Breath Sample Analyses," Corps of Engineers, October 20 [Linde Air 

and Ceramics, NY\10-22-03 - Data Capture\240000348, p. 6 of 9].

Tybout, R. A., 1945a, "Industrial Health Control Tests," memorandum to the Area Engineer, 

Tonawanda Area, Tonawanda, NY from U.S. Engineer Office, Manhattan District, Rochester, 

NY, March 12 [Linde Air and Ceramics, NY\01-6-04 Data Capture\240000584 pp. 42-43].

Tybout, R. A., 1945b, "Industrial Health Control Tests," memorandum to the Area Engineer, 

Tonawanda Area, Tonawanda, NY from U.S. Engineer Office, Manhattan District, Rochester, 

NY, EIDMM-4, March 13 [Linde Air and Ceramics, NY\10-22-03 - Data Capture\240000221 pp. 

5-6].

Tybout, R. A., 1945c, "Dust Sample Results from Chandler Plant," memorandum to the District 

Engineer, U.S. Engineer Office, Oak Ridge, TN (Attention: Captain Fred A. Bryan), from U.S. 

Engineer Office, Manhattan District, Rochester, NY, EIDMM-4c, EIDMM-2b, MD723.1, May 4 

[Linde Air and Ceramics, NY\Linde Ceramics Plant.pdf p. 678].
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 87 of 94

Tybout, R. A., 1945d, "Dust Exposure to Loaders," memorandum to the District Engineer, U.S. 

Engineer Office, Oak Ridge, TN (Attention: Major J. L. Ferry), from U.S. Engineer Office, 

Rochester Area, Rochester, NY, May 23 [Linde Air and Ceramics, NY\10-20-03 - Data 

Capture\240000373 p. 3].

UNSCEAR (United Nations Scientific Committee on the Effects of Atomic Radiation), 1993, Sources 

and Effects of Ionizing Radiation. 

U.S. Engineer Office, 1945, “Contract to Linde Air Products Company, New York, NY,” W-17-028 eng29, May 4 [Linde Air and Ceramics, NY\Linde Ceramics Plant.pdf p. 266].

U.S. Government, 1944, “Supplemental Agreement No. 4 to Contract No. W-7401 eng-14, with Linde 

Air Products Company,” August 16 [Linde Air and Ceramics, NY\Linde Ceramics Plant.pdf p. 

335].

Van Horn, E. L., 1943a, untitled letter, #T-1, to Linde Air Products Company, Tonawanda, New York, 

Attention: Mr. T. J. Coleman, from Area Engineer, September 4 [Linde Air and Ceramics, 

NY\10-16-03 - Data capture\240000136 p. 5].

Van Horn, E. L., 1943b, untitled letter, T-1, to Linde Air Products Company, Tonawanda, New York, 

Attention: Dr. E. L. McCandless, with enclosure entitled “Suggestions for Examiners,” from 

Area Engineer, September 9 [Linde Air and Ceramics, NY\10-16-03 - Data 

capture\240000163.pdf p. 10].

Wallace, P., 2003, "Notes on Linde Data," Center for Epidemiologic Research, Oak Ridge Associated 

Universities, Oak Ridge, TN, August 18 [Linde Air and Ceramics, NY\08-20-03 - Data 

Capture\010001334].

Warren, S. L., 1944, "Routine Pelvic Films on Special Materials Employees," EIDMM-3, memorandum 

to Area Engineer, Tonawanda Area (Tonawanda, NY), from Manhattan District, New York, NY, 

January 11 [Linde Air and Ceramics, NY\10-21-03 - Data Capture\240000286, p. 4].

Wiesendanger, R., 1944, Pilot Plant Operations on African Ore, Tonawanda Laboratory, Linde Air 

Products Company, Tonawanda, NY, U.S. Atomic Energy Commission, M-4541, May 26 

[Linde Air and Ceramics, NY\08-07-03 - Data Capture\010001220\000006]. 

Wilson, R. H., 1958, “The Hanford Uranium Bio-Assay Program,” Symposium on Occupational Health 

Experience and Practices in the Uranium Industry, HASL-58, U. S. Atomic Energy 

Commission, New York, New York, October 15 -17 [p. 87 of 001760.pdf] 

Wolf, B. S., 1948, “Results of Uranium and Fluoride Analyses,” letter to R. C. Heatherton (Linde Air 

Products, Ceramics Plant), U.S. Atomic Energy Commission, New York, February 27 [Linde 

Air and Ceramics, NY\Linde Ceramics Plant.pdf, pp. 96-98]

Wolf, B. S., 1949, untitled letter to E. O. Kent (Linde Air Products, Inc., Ceramics Division), Medical 

Section, New York office, U.S. Atomic Energy Commission, March 16 [Linde Air and 

Ceramics, NY\07-22-03 - Data Capture 2-Linde\Eml2Box25LindeAirProducts47_52, p. 17].

Wolf, B. S., 1949, “Linde Air Products Co., Division of Union Carbide & Carbon Corp, Ceramics Plant, 

Tonawanda, N.Y. – Decontamination of Main Building,” memorandum to F. M. Belmore, 

Production Division, from Medical Section, New York office, U.S. Atomic Energy Commission, 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 88 of 94

May 12 [Linde Air and Ceramics, NY\07-22-03 - Data Capture 2-Linde\ 

Eml2Box25LindeAirProducts47_52.pdf, p. 13].
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 89 of 94

GLOSSARY

carnotite

K2(UO2)2(VO4)2·3(H2O); a mineral found in Colorado and Utah that was mined for its uranium 

and vanadium content. [Much of the domestic ore processed at Linde Ceramics derived from 

tailings from carnotite ore. Preprocessing of this material prior to its shipment to Linde is 

believed to have removed much of its radium content (Aerospace Corporation 1981).] 

day

As used in this report a day refers to a calendar day, unless otherwise described. A calendar 

day is any 24-hour day in a year. There are at least 365 calendar days in a year. A workday 

refers to a day that is assigned for work. The length of a workday depends on the amount of 

time spent on the job. A default assumption is that there are 8 hours in a workday, and 250 

workdays in a year. At Linde, the number of hours per workday and the number of workdays 

per year for some employees were larger than the default for the early years. 

Linde database

Records of Linde operations that were included in NIOSH computer-stored data and were 

available as this document was being prepared.

N. G. (or NG) Cake

Filter cake produced in Step II processing at Linde. Two types of cakes were produced; they 

were called "OK" and "NG."

pitchblende

A mineral containing uranium oxide of variable composition ranging between UO2 and U3O8. 

(Pitchblende ore from Africa was processed at Linde Ceramics for its uranium content. It was 

not pre-refined before its arrival at Linde and so contained all of the natural uranium 

radioactive progeny including radium.) 

Step I, Step II and Step III

The three uranium production processes at Linde. 

Step I, conversion from ore to U3O8 (black oxide)

Step II, conversion of U3O8 to UO3 (orange oxide) to UO2 (brown oxide)

Step III, conversion of UO2 to UF4 (green salts)

torbernite

Cu(UO2)2(PO4)2·8-12(H2O); a mineral in ore processed at Linde Ceramics for its uranium 

content. (Like the African pitchblende processed at Linde, torbernite was not pre-refined 

before its arrival at Linde and so contained all of the natural uranium radioactive progeny, 

including radium.) 
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 90 of 94

ATTACHMENT A 

BETA RADIATION

LIST OF TABLES

Table

A-1 Maximum beta surface dose rates from various uranium-containing materials.......................

Page

90

A-2 Electron energy released by refined uranium ......................................................................... 90

A-3 Electron energy released by unrefined natural uranium ore ................................................... 91

A-4 Estimated attenuation of beta radiation with distance............................................................. 92

A-5 Uranium beta dose reduction factors for apparel.................................................................... 92

A-6 Variation of dose rate with body location for Step III process operations ................................ 93

This attachment contains data used in analyzing exposures of workers to beta radiation.

Table A-1. Maximum beta surface dose 

rates from various uranium-containing 

materials. 

Source

Beta surface dose rate

mrad/ha,b rad/8 h

Slab of U metal 233 1.86

UO2 207 1.66

UO3 204 1.63

U3O8 203 1.62

UF4 179 1.43

a. From Table 2-7 on p. 2-18 of DOE 2000.

b. Beta surface dose rate in air through a 

polystyrene filter 7 mg/cm2 thick.

Table A-2. Electron energy released by refined uranium. 

Principal 

nuclides

Electron energy

per decay of nuclide

(MeV/nt)

Electron energy

per decay of parent

(MeV/nt) % of total

U-238 series U-238 0.010 0.005 1.1

Th-234 0.060 0.029 6.6

Pa-234m 0.822 0.402 89.8

U-234 0.013 0.006 1.4

U-235 series U-235 0.049 0.001 0.2

Th-231 0.165 0.004 0.8

Total: 0.447
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 91 of 94

Table A-3. Electron energy released by unrefined natural uranium ore. 

Principal 

nuclides

Electron energy

per decay of nuclide

(MeV/nt)

Electron energy

per decay of parent

(MeV/nt) % of total

U-238 series U-238 0.010 0.005 0.4

Th-234 0.060 0.029 2.5

Pa-234m 0.822 0.402 34.7

U-234 0.013 0.006 0.5

Th-230 0.015 0.007 0.6

Ra-226 0.004 0.002 0.2

Rn-222 0.000 0.000 0.0

Po-218 0.000 0.000 0.0

Pb-214 0.293 0.143 12.4

Bi-214 0.659 0.322 27.8

Po-214 0.000 0.000 0.0

Pb-210 0.038 0.019 1.6

Bi-210 0.389 0.190 16.4

Po-210 0.000 0.000 0.0

U-235 series U-235 0.049 0.001 0.1

Th-231 0.165 0.004 0.3

Pa-231 0.065 0.001 0.1

Ac-227 0.016 0.000 0.0

Th-227 0.053 0.001 0.1

Ra-223 0.076 0.002 0.1

Rn-219 0.006 0.000 0.0

Po-215 0.000 0.000 0.0

Pb-211 0.456 0.010 0.9

Bi-211 0.010 0.000 0.0

Tl-207 0.493 0.011 1.0

Total: 1.157
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 92 of 94

Table A-4. Estimated attenuation of beta radiation with distance. 

Distance from

surface (m)

Relative dose ratea

Refined natural

uraniumb 

Unrefined natural uranium 

ore or ore by-productsc 

0.0 1.00 1.00

0.3 0.80 0.88

0.94 0.50 0.66

1.0 0.48 0.64

1.6 0.31 0.50

2.0 0.23 0.42

3.0 0.11 0.27

5.0 0.025 0.11

8.5d 0.000 0.024

10.0 — f 0.012

12.6e — f 0.000

a. Only exponential attenuation due to air is considered. Additional reduction 

would come from geometric dispersion, which is ignored here.

b. Dose rate assumed to vary with distance as exp(-0.0074x), where x is 

distance from the surface in cm. This was derived from a fit to data in 

Coleman, Hudson and Plato 1983.

c. Calculated falloff rate based on the assumption that all of the beta radiation is 

due to the 3.27 MeV beta ray from Bi-214. The dose rate is assumed to vary 

with distance as exp(-0.0044x), where x is distance from the surface in cm. 

The attenuation factor 0.0044 cm-1 was estimated from the rule of thumb that 

the half thickness of a beta absorber is 1/8 of the range of the beta rays 

(Cember 1983, p. 97). For natural uranium ore and by-products containing 

beta emitters other than Bi-214, the actual falloff with distance would be faster 

because all other significant beta emissions have lower energies than 3.28 

MeV.

d. Range of 2.28 MeV beta radiation per Cember 1983, p. 99.

e. Range of 3.27 MeV beta radiation per Cember 1983, p. 100.

f. No radiation at this distance.

Table A-5. Uranium beta dose reduction factors 

for apparel. 

Itema Fraction of betadose remaining

Leather, medium weight 0.62

White cotton gloves 0.89

"Tyvek" coveralls 0.98

65% Dacron/35% cotton lab coat 0.91

a. Selected from Table 6-11 on p. 6-23 of DOE 2000.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 93 of 94

Table A-6. Variation of dose rate with body location for Step III process operations. 

Operation

(mR/h)a

Hands, bare Hands, gloved Face Chest Legs Feet

Handling empty trays 4 3 1.5 4 —b —b

Loading trays (brown) 25 13 2 4.5 4.5 2

Handling loaded trays(brown) 25 13 5 16 —b —b

Handling loaded trays (green) 25 16 7.5 25 —b —b

Unloading operation (green) 25 —b 7.5 7.5 4 2

Blending 15 10 7 8.5 12 —b

RL relative to chestc RL rel to hands, barec

Hands, bare Face Chest Legs Feet Hands, gloved

Handling empty trays 1.00 0.38 1.00 —b —b 0.75

Loading trays (brown) 5.56 0.44 1.00 1.00 0.44 0.52

Handling loaded trays(brown) 1.56 0.31 1.00 —b —b 0.52

Handling loaded trays (green) 1.00 0.30 1.00 —b —b 0.64

Unloading operation (green) 3.33 1.00 1.00 0.53 0.27 —b

Blending 1.76 0.82 1.00 1.41 —b 0.67

Maximum 5.56 1.00 1.00 1.41 0.44

a. Data from Heatherton 1948a.

b. No data.

c. RL = radiation level. Results in lower half of the table are based on the data in the upper half.
Document No. ORAUT-TKBS-0025 Revision No. 00 Effective Date: 05/31/2005 Page 94 of 94

ATTACHMENT B 

CODES AND SPECIAL TERMINOLOGY

Special codes and terminology were used in correspondence and documents related to Linde's work 

for the MED and AEC in place of common names or to specify special forms of materials with 

particular characteristics. Over the course of time, some terms may have had multiple meanings. 

The table below lists apparent definitions deduced during preparation of this document. 

Code or term Apparent meaning

C-103 UO2 from DuPont; a feed material used for Linde Step III

C-306 UO2 from Mallinckrodt; a feed material used for Linde Step III

C-316 UO2 from Mallinckrodt; a feed material used for Linde Step III

F-29 U3O8 made at Linde

K-25 UF6 gas containing various concentrations of U-235

K-65 Residues containing Ra and Pb that were byproducts of the ore to UO3 conversion process at 

Mallinckrodt Chemical Works

L-19 A domestic ore processed at Linde; estimated to contain mostly 10 – 16.5% U3O8 (Aerospace 

1981, Table B-1)

L-30 African pitchblende ore; estimated to contain 8 – 12% U3O8 (Aerospace 1981, Table B-1)

L-50 African pitchblende ore; estimated to contain 6.7% U3O8 (Aerospace 1981, Table B-1)

Mx, MX Uranium or U3O8

MX-308 U3O8 (in ore)

My Radium

Mz, MZ Radon

O-71 UF4

P-65 UO2 produced by Linde Step II; a feed material used for Linde Step III

Product 65 P-65

Q-20 Torbernite ore processed at Linde; estimated to contain 17.7% U3O8 (Aerospace 1981, Table B-1)

R-10 African pitchblende ore processed at Linde; estimated to contain 3.5% U3O8 (Aerospace 1981, 

Table B-1)

T Uranium

Tubealloy dioxide UO2

Tubealloy tetrafluoride UF4

X Uranium or uranium ore