ORAU TEAM 

Dose Reconstruction 

Project for NIOSH 

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services 

 

Page 1 of 101 

DOE Review Release 05/20/2015 

ORAUT-TKBS-0025 Rev. 04 An Exposure Matrix for Linde Ceramics Effective Date: 05/08/2015 Plant (Including Tonawanda Laboratory) 

Supersedes: Revision 03 

Subject Expert(s): Mutty M. Sharfi 

Document Owner 

Approval: Signature on File 

Mutty M. Sharfi, Document Owner 

Approval Date: 05/05/2015 

Concurrence: Signature on File 

John M. Byrne, Objective 1 Manager 

Concurrence Date: 05/05/2015 

Concurrence: Signature on File 

Edward F. Maher, Objective 3 Manager 

Concurrence Date: 05/05/2015 

Concurrence: Vickie S. Short Signature on File for 

Kate Kimpan, Project Director 

Concurrence Date: 05/05/2015 

Approval: Signature on File 

James W. Neton, Associate Director for Science 

Approval Date: 05/08/2015 

 NewTotal Rewrite Revision Page Change 

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION. 
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 2 of 101

PUBLICATION RECORD

EFFECTIVE REVISION

DATE NUMBER DESCRIPTION

05/31/2005 00 New Technical Basis Document: Basis for the Development of an 

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

Buffalo, New York. First approved issue. Initiated by Cindy Bloom.

01/19/2006 00 PC-1 Modified to incorporate comments from OCAS, a Worker Outreach 

meeting comment and minor editing. The Introduction on page 10 

was updated. Dates were changed in Section 3.0 to incorporate the 

decision to designate Linde Ceramics workers, who received internal 

exposures prior to 11/01/1947, as a Special Exposure Cohort. 

Moved Table 12, an internal exposure table out of Section 4.0. 

Changes were made to page 23 in Section 3, page 34 in Table 7, 

page 35 in Table 8, page 38 in Table 11, and Table 12 on page 39. 

Approved page change revision. No sections were deleted. Training 

required: As determined by the Task Manager. Initiated by Cindy W. 

Bloom.

Signature on File 01/18/2006

Cindy W. Bloom, TBD Team Leader

Signature on File 01/18/2006

Judson L. Kenoyer, Task 3 Manager

Signature on File 01/18/2006

Kate Kimpan, Project Director

Signature on File 01/19/2006

James W. Neton, Associate Director for Science

11/04/2008 01 Approved revision to change from a page change revision (Rev 00 

PC-2-B) to a total rewrite (Rev 01-A) as a result of formal NIOSH 

review. Revised to incorporate: (1) change in facility designation, (2) 

DOL interpretation of applicability of residual period to Ceramics 

Plant, (3) and resolution of Advisory Board Working Group 

comments, (4) clarified the implementation instructions for 

SEC00044 for the period October 1, 1942 through October 31, 1947. 

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Task Manager. Initiated by Joseph S. Guido.

11/10/2009 01 PC-1 Approved page change revision to incorporate additional guidance 

on the use of Attachment E of the document. This change 

constitutes the resolution of Advisory Board Working Group concern 

and clarifies how to proceed in the event of credible indication of 

potential exposure to contaminated burlap bags that were used for 

transport of ore at Linde Ceramics. Changes were made to page 37 

in Section 4.0. No sections were deleted. No changes occurred as a 

result of formal internal review. Incorporates formal NIOSH review 

comments. Training required: As determined by the Objective 

Manager. Initiated by Mutty Sharfi.

Signature on File 11/05/2009

Mutty Sharfi, Document Owner
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 3 of 101

Signature on File 11/04/2009

John M. Byrne, Objective 1 Manager

Signature on File 11/04/2009

Edward F. Maher, Objective 3 Manager

Signature on File 11/04/2009

Kate Kimpan, Project Director

Signature on File 11/10/2009

James W. Neton, Associate Director for Science

07/15/2011 02 Revision to incorporate changes associated with SEC-0107. This 

change constitutes the resolution of SEC-0107. This includes the 

exposure scenarios for the tunnels, the external for the renovation 

period and the post-renovation period. An SEC was approved 

covering internal exposures during the renovation period (1954 to 

1969). Changes were made to Sections 6.0, 6.1, 6.2, and the 

references. In addition, Sections 1.3, 2.3.5, and 6.4 were added. 

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by Mutty M. Sharfi. 

07/26/2012 03 Revision to incorporate changes associated with SEC-0154. This 

change constitutes the resolution of SEC-0154. No new exposure 

scenarios were added. Changes were made to Sections 1.3, 3.4,

3.5, 3.6, 3.8, 4.0, 4.4, 6.0 and 6.1. Minor changes were made to the 

Acronyms and Abbreviations, and Reference Sections. Removed 

former Attachment D (Uranium coworker assessment for November 

1947 to January 1950). Incorporates formal internal and NIOSH 

review comments. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Mutty M. Sharfi.

05/08/2015 04 Revised to correct a footnote in Table 4-24 (the year 1947 

superscript was changed from i to j). Changes made to Section 6.4 

regarding tunnel occupancy time. Previous text of assuming trade 

workers and laborers worked in tunnels doing maintenance for 8 

hr/wd (2 months of the year) and for the other 10 months, a transit 

time of 10 min/wd using the tunnels to get between buildings. For all 

other workers only the transit time of 10 min/wd should be applied 

year-round, was changed to, for Trade Workers & Laborer it is 

assumed that they worked in these tunnels doing maintenance 50% 

of their time. For all other workers, an occupancy factor of 5% was 

assumed. This change did not affect Table 6-12 values. No 

changes occurred as a result of formal internal review. Incorporates 

formal NIOSH review comments. Training required: As determined 

by the Objective Manager. Initiated by Mutty M. Sharfi.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 4 of 101

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

1.1 Purpose...................................................................................................................... 11

1.2 Scope......................................................................................................................... 11

1.3 Special Exposure Cohort............................................................................................ 12

2.0 Site Description and Operational History................................................................................ 13

2.1 Facilities ..................................................................................................................... 15

2.1.1 Building 14 (Tonawanda Laboratory)............................................................ 15

2.1.2 Building 30 (Ceramics Plant)........................................................................ 17

2.1.3 Building 31 (Ceramics Plant)........................................................................ 17

2.1.4 Building 37 (Ceramics Plant)........................................................................ 19

2.1.5 Building 38 (Ceramics Plant)........................................................................ 19

2.2 Source Terms for African and Domestic Uranium Ores .............................................. 19

2.3 Activities ..................................................................................................................... 20

2.3.1 Uranium Research and Development........................................................... 20

2.3.2 Uranium Production ..................................................................................... 21

2.3.2.1 Step I (Uranium Ore to U3O8)................................................... 21

2.3.2.2 Step II (U3O8 to UO2)............................................................... 23

2.3.2.3 Step III (UO2 to UF4) ................................................................ 23

2.3.3 Nickel Production ......................................................................................... 24

2.3.4 Other Radiological Activities......................................................................... 24

2.3.5 Utility Tunnel Complex ................................................................................. 25

2.4 Protective Measures................................................................................................... 28

2.5 Personnel, Job Categories, and Work Hours.............................................................. 28

2.6 Decontamination During the MED/AEC Contract Period............................................. 30

2.7 Post-MED/AEC Operations ........................................................................................ 31

3.0 Estimation of Internal Exposure, November 1, 1947, to December 31, 1953.......................... 31

3.1 Air monitoring for Particulate Intakes .......................................................................... 32

3.2 Uranium...................................................................................................................... 33

3.2.1 Uranium Urinalysis Data............................................................................... 33

3.3 Radium....................................................................................................................... 34

3.4 Uranium Progeny........................................................................................................ 35

3.5 Radon......................................................................................................................... 35

3.5.1 Ceramics Plant Radon Exposures................................................................ 35

3.5.2 Tonawanda Laboratory Radon Exposures ................................................... 36

3.6 Inhalation Intake Estimates of Radioactive Particulates at The Tonawanda 

Laboratory .................................................................................................................. 37

3.7 Ingestion Intake Estimates at The Tonawanda Laboratory ......................................... 37

3.8 Occupational Internal Dose Reconstruction Assumptions and Summary.................... 37

4.0 Estimation of External Exposure, 1942 to December 31, 1953............................................... 38

4.1 Ceramics Plant Beta and Gamma Exposure............................................................... 39

4.1.1 Preproduction, 1942 to 1943 ........................................................................ 39

4.1.2 Production, 1943 to 1946 ............................................................................. 41

4.1.2.1 Beta........................................................................................... 41

4.1.2.1.1 Estimation of Worker Doses ................................ 42
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4.1.2.1.2 Outdoor Dose Rate.............................................. 45

4.1.2.1.3 Categories ........................................................... 45

4.1.2.2 Gamma ..................................................................................... 45

4.1.2.2.1 Estimation of Worker Doses ................................ 49

4.1.2.2.2 Outdoor Exposure Rate ....................................... 53

4.1.3 Standby, 1946 to 1947 ................................................................................. 53

4.1.4 Rehabilitation and Production, 1947 to 1949................................................ 53

4.1.4.1 Beta........................................................................................... 53

4.1.4.2 Gamma ..................................................................................... 57

4.1.5 Cleanup, 1949 to December 31, 1953.......................................................... 58

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

4.3 Neutron Exposure....................................................................................................... 61

4.4 External Dose Reconstruction Summary, October 1, 1942, to December 31, 

1953 ........................................................................................................................... 64

5.0 Occupational Medical Exposure............................................................................................. 67

6.0 Estimation of Exposures from Residual Contamination after 1953 ......................................... 67

6.1 Internal Exposure ....................................................................................................... 67

6.2 External Beta and Gamma Exposure.......................................................................... 69

6.3 Neutron Exposure....................................................................................................... 70

6.4 Utility Tunnel Exposure............................................................................................... 70

7.0 Attributions and Annotations .................................................................................................. 76

References ........................................................................................................................................ 77

Glossary ............................................................................................................................................ 90

ATTACHMENT A, BETA RADIATION............................................................................................... 91

ATTACHMENT B, CODES AND SPECIAL TERMINOLOGY ............................................................ 95

ATTACHMENT C, DATA SOURCES ON URANIUM PROGENY CONCENTRATIONS IN LINDE 

MATERIALS........................................................................................................ 96

ATTACHMENT D, FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM 

ORE BAGS ON THE SITE DURING THE POSTOPERATIONS PERIOD ........... 97

LIST OF TABLES

TABLE TITLE PAGE

2-1 Key dates...............................................................................................................................14

2-2 Department codes for various worker activities ...................................................................... 29

2-3 Worker information.................................................................................................................30

2-4 Ceramics Plant assumed work schedule, 1942 to December 31, 1953 .................................. 30

3-1 Measured radon concentrations during Ceramics Plant African ore processing ..................... 36

3-2 Ceramics Plant worker radon exposure rates, 1947 to 1953 .................................................. 36

3-3 Tonawanda Laboratory radon exposure rates, 1947 to 1953 ................................................. 37

3-5 Radon exposure for the Ceramics Plant, November 1, 1947, to December 31, 1953............. 37

3-6 Radon exposure for the Tonawanda Laboratory, November 1, 1947, to December 31, 

1953.......................................................................................................................................38

4-1 Floor and wall radiation in Building 30 before and after the 1949 to 1950 

decontamination.....................................................................................................................39
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4-2 Building 30 floor radiation levels in 1976 ................................................................................ 40

4-3 Preproduction beta and gamma radiation levels..................................................................... 40

4-4 Beta radiation exposure rates measured at Linde Ceramics, January 23, 1944 ..................... 41

4-5 Dependence of beta dose rate on ore type ............................................................................ 43

4-6 Assigned beta dose rates, 1943 to 1946 ................................................................................ 44

4-7 Job categories for Step I gamma dose analysis, 1943 to 1946............................................... 46

4-8 Step I gamma exposure data for African ore processing, January 31 to November 12, 

1944.......................................................................................................................................47

4-9 Step I gamma exposure rate parameters for African ore processing, January 31 to 

November 12, 1944................................................................................................................48

4-10 Step I gamma exposure data for cleanup and domestic ore processing, November 13, 

1944, to February 26, 1945.................................................................................................... 48

4-11 Step I gamma exposure rate parameters for cleanup and domestic ore processing, 

November 13, 1944, to February 26, 1945............................................................................. 49

4-12 Estimated maximum exposure rates in selected Building 30 facilities due to 30-by 30- by 

12-ft pile of 8% ore in southeast corner.................................................................................. 52

4-13 Ceramics Plant beta and gamma radiation rates during standby............................................ 53

4-14 Job categories for 1947 to 1949 Step III film badge analysis.................................................. 54

4-15 Step III beta dose data, 1947 to 1949..................................................................................... 55

4-16 Beta dose rates for Step III, 1947 to 1949.............................................................................. 56

4-17 Step III gamma exposure data, 1947 to 1949......................................................................... 57

4-18 Gamma exposure rates for Step III, 1947 to 1949.................................................................. 58

4-19 Beta and gamma radiation measured during equipment removal, 1948................................. 59

4-20 External dose reconstruction parameters for cleanup............................................................. 60

4-21 Tonawanda Laboratory beta and gamma radiation, 1942 to 1953.......................................... 60

4-22 Natural uranium per gram dose rates at 1 ft ........................................................................... 62

4-23 Neutron dose rates and annual doses from the alpha-neutron sources.................................. 64

4-24 Summary of annual external exposure from AWE operations, 1942 to 1953.......................... 64

4-25 Ceramics Plant worker beta and gamma external exposure categories for production 

periods...................................................................................................................................66

6-1 Progeny to uranium ratios ...................................................................................................... 68

6-2 Uranium and progeny inhalation intake rates for January 1, 1970, through 2009 ................... 68

6-3 Uranium and progeny ingestion intake rates for January 1, 1970, through 2009 .................... 68

6-4 Radon exposure summary from January 1, 1970, through 2009 ............................................ 69

6-5 Radionuclide fraction calculation using average Linde soil concentrations ............................. 71

6-7 Utility tunnels surface activity ................................................................................................. 72

6-8 Utility tunnels trade worker intake rates.................................................................................. 73

6-9 Utility tunnels other worker intake rates.................................................................................. 73

6-10 Radon parameters for utility tunnels....................................................................................... 74

6-11 Exposure rates from utility tunnel radon contamination .......................................................... 74

6-12 Radon exposure rates for utility tunnels from surrounding soils.............................................. 75

6-13 Utility tunnels external exposure rates.................................................................................... 75

A-1 Maximum beta surface dose rates from various uranium-containing materials....................... 91

A-2 Electron energy released by refined uranium ......................................................................... 91

A-3 Electron energy released by unrefined natural uranium ore ................................................... 92

A-4 Estimated attenuation of beta radiation with distance............................................................. 93

A-5 Uranium beta dose reduction factors for apparel.................................................................... 93

A-6 Variation of dose rate with body location for Step III process operations................................ 94

B-1 Codes and special terminology .............................................................................................. 95
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LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Linde Ceramics Plant and Tonawanda Laboratory buildings on CCCC Tonawanda site ........ 16

2-2 Building 30 layout...................................................................................................................17

2-3 Building 30 layout...................................................................................................................18

2-4 Building 38 layout...................................................................................................................19

4-1 Fit to gamma exposure data for a 4- by 5- by 20-ft pile of 8% African pitchblende ore ........... 50

4-2 Calculated exposure rates and analytic fit for a 30- by 30- by 12-ft pile of 8% African 

pitchblende ore ......................................................................................................................51
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

Bq becquerel

C Celsius

CCCC Carbide and Carbon Chemicals Corporation

CFR Code of Federal Regulations

Ci curie

cm centimeter

d day

D&D decontamination and decommissioning

DCF dose conversion factor

DHHS U.S. Department of Health and Human Services

DOL U.S. Department of Labor

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

F Fahrenheit

ft foot

FUSRAP Formerly Utilized Sites Remedial Action Program

GM geometric mean

GSD geometric standard deviation

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

keV kiloelectron-volt, 1,000 electron-volts

L liter

lb pound

LOD limit of detection

LOOW Lake Ontario Ordnance Works

m meter

MAC maximum allowable concentration

MED Manhattan Engineer District

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mL milliliter

mm millimeter

mR milliroentgen

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

photons and millirem for beta radiation

mrad millirad

mrem millirem

mrep millirep
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MT metric ton (1,000 kg)

NIOSH National Institute for Occupational Safety and Health

NYOO New York Operations Office

OCAS Office of Compensation Analysis and Support

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

PAEC potential alpha energy concentration (a measure of concentration of radon progeny)

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

pCi picocuries

PL preferred level

POC probability of causation

ppm parts per million

R roentgen

R&D research and development

s second

SC&A Sanford Cohen & Associates

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

t ton (2,000 lb)

UCC Union Carbide Corporation

U.S.C. United States Code

USACE U.S. Army Corps of Engineers

USBM U.S. Bureau of Mines

wd workday

wk week

WL working level

WLM working level-month

yr year

α alpha particle

μg microgram

μm micrometer

μR microroentgen

§ section or sections
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1.0 INTRODUCTION

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

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

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

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean, nor should it be equated to, an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7348l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. A DOE facility is defined as “any building, structure, or premise, including 

the grounds upon which such building, structure, or premise is located ... in which operations are, or 

have been, conducted by, or on behalf of, the [DOE] (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program),” and with regard to 

which the DOE has or had a proprietary interest; or “entered into a contract with an entity to provide 

management and operation, management and integration, environmental remediation services, 

construction, or maintenance services.” 42 U.S.C. § 7384l(12). On the other hand, an AWE facility 

means “a facility, owned by an atomic weapons employer, that is or was used to process or produce, 

for use by the United States, material that emitted radiation and was used in the production of an 

atomic weapon, excluding uranium mining or milling.” 42 U.S.C. § 7384l(5). The Department of 

Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time 

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction. 

On the other hand, under EEOICPA, eligible employment at a AWE facility is categorized as 

employment either (1) during “a period when the employer was processing producing, for the use by 

the United States, material that emitted radiation and was used in the production of an atomic 

weapon, excluding uranium mining and milling,” (i.e., the operational period); or (2) during a period 

that NIOSH has determined that “there is a potential for significant residual contamination outside of 

the period in which weapons-related production occurred,” (i.e., the residual contamination period). 

42 U.S.C. § 7384l(3).
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Based on the abovementioned definition for eligible employment during an AWE facility’s operational 

period, NIOSH includes radiation exposures incurred in the performance of duty, such as medical Xrays received as a condition of employment for participating in DOE projects, at an AWE facility in 

dose reconstructions. This may include radiation exposure related to the Naval Nuclear Propulsion 

Program and any radiation exposure received from the production of commercial radioactive products 

that were concurrently manufactured by the AWE facility during the operational period. In contrast, 

only two categories of radiation exposure as defined in 42 U.S.C. § 7384n(c)(4) should be included in 

dose reconstructions for claims involving employment during the residual contamination period. First, 

NIOSH must include exposures to radiological contaminants resulting from activities that had a 

nuclear-weapon nexus or conducted by or on behalf of the DOE (with an exclusion of activities related 

to, among other things, the Naval Nuclear Propulsion Program) that took place during the operational 

period. 42 U.S.C. § 7384n(c)(4)(A). Second, radiation doses from sources not included in the first 

category but which cannot be distinguished through reliable documentation should also be included in 

dose reconstructions. 42 U.S.C. § 7384n(c)(4)(B). Furthermore, because all DOE-related activities 

have ceased during the residual contamination period, NIOSH does not include doses from medical 

X-rays performed during the residual contamination period (NIOSH 2007) in dose reconstructions. 

Likewise, NIOSH does not consider the following types of exposure as those incurred in the 

performance of duty as a condition of employment for DOE-related activities at a DOE or AWE facility. 

Therefore these exposures are not included in dose reconstructions for covered employees 

(NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

1.1 PURPOSE

This site profile document provides an exposure matrix for workers at the Tonawanda Laboratory and 

Linde Ceramics Plant facilities in Tonawanda, New York.

1.2 SCOPE

DOE has designated the entire Linde Ceramics Plant, which consists of the Tonawanda Laboratory 

plus Buildings 30, 31, 37 and 38, as an AWE facility, while DOL has determined that a portion of the 

Linde Ceramics Plant (Buildings 30, 31, 37 and 38 only) also constitutes a DOE facility (Turcic 2007). 

This document covers both facilities. The information in this site profile supports the assumed 

operational and residual contamination periods listed below. 

Facility Operational period

Residual contamination 

perioda,b 

Linde Ceramics Plant and 

Tonawanda Laboratory

10/01/1942–10/31/1953

1988–1992; 1996 (remediation)

January 1, 1954, to July 2006c

a. Also called the postcleanup period in this document.

b. Coverage during the residual contamination period only applies to the AWE designated facilities on the Linde site.

c. Excluding 1988 to 1992 and 1996.

Section 2.0 describes the site and its operational history. Sections 3.0 and 4.0 describe estimation of 

internal and external exposure from 1942 to December 31, 1953, respectively. Section 5.0 describes 

occupational medical exposure. Section 6.0 provides information on exposures during the residual 

contamination period after 1953. Attributions and annotations, indicated by bracketed callouts and 
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used to identify the source, justification, or clarification of the associated information, are presented in 

Section 7.0.

Attachment A contains data that was used in analyzing exposures of workers to beta radiation. 

Attachment B lists codes and special terminology in the CCCC records. Attachment C shows data 

sources on uranium progeny concentrations, and Attachment D provides an assessment of dose 

consequences from uranium ore bags that were stored on the site during the postoperations period.

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section X.0.

1.3 SPECIAL EXPOSURE COHORT

October 1, 1942, through October 31, 1947

NIOSH has determined, and the Secretary of the U.S. Department of Health and Human Services 

(DHHS) has concurred, that it is not feasible to reconstruct internal radiation dose for (Leavitt 2005):

Atomic Weapons Employees who worked at the Linde Ceramics Plant from October 1, 

1942, through October 31, 1947, and who were employed for a number of work days 

aggregating at least 250 work days either solely under this employment or in combination 

with work days occurring within the parameters (excluding aggregate work day 

requirements) established for other classes of employees included in the SEC [Special 

Exposure Cohort].

Subsequent correspondence (Elliott 2006) confirmed that the Tonawanda Laboratory (as well as all 

other buildings on the Linde site) are included in this SEC class designation. Reconstruction of 

external exposure (including medical X-ray examinations) has been determined to be feasible 

(Leavitt 2005).

For any claim referred to NIOSH regarding an employee,

• Who was employed during the SEC class period but because of limited employment during 

this period is not a member of the SEC, or

• Who is a member of the SEC class and whose cancer is not defined as a specified cancer 

under EEOICPA (and so is not eligible for compensation under EEOICPA without a dose 

reconstruction),

NIOSH will continue to attempt to complete a dose reconstruction for the exposure period based 

solely on external and medical X-ray radiation sources. However, because of the SEC class 

determination that it is infeasible to adequately reconstruct internal dose during the period from 

October 1, 1942, through October 31, 1947 (Leavitt 2005), dose estimates for this period are 

considered partial dose estimates.

November 1, 1947, through December 31, 1953

NIOSH has determined, with concurrence from the Secretary of DHHS (Sebelius 2012), that not all 

aspects of internal doses at the Linde Ceramic Plant can be reconstructed with sufficient accuracy 

from the beginning of 1947 through the end of 1953. For this reason, the following class of Linde 

employees has been added to the SEC (Sebelius 2012):

All Atomic Weapons Employees who worked in any area at the Linde Ceramics Plant in 

Tonawanda, New York, from November 1, 1947 through December 31, 1953, for a 

number of work days aggregating at least 250 work days, occurring either solely under 
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 13 of 101

this employment or in combination with work days within the parameters established for 

one or more other classes of employees included in the SEC.

This SEC class includes all workers during the SEC class period. Because of the identified dose 

reconstruction infeasibility, all dose reconstructions for all workers having employment during the SEC 

class period are considered partial dose reconstructions. If monitoring data are available for workers 

included in the SEC class, dose is to be assigned as appropriate based on such data; however, such 

dose reconstructions are still considered partial dose reconstructions because NIOSH has determined 

that internal exposures during the SEC class period cannot be bounded. 

January 1, 1954, through December 31, 1969

NIOSH has determined, with concurrence from the Secretary of DHHS (Sebelius 2011), that internal 

doses at the Linde Ceramic Plant cannot be reconstructed with sufficient accuracy from the beginning 

of 1954 through the end of 1969. For this reason, the following class of Linde employees has been 

added to the SEC (Sebelius 2011):

All Atomic Weapons Employees who worked at the Linde Ceramics Plant in Tonawanda, 

New York, from January 1, 1954, through December 31, 1969, for a number of work 

days aggregating at least 250 work days, occurring either solely under this employment, 

or in combination with work days within the parameters established for one or more other 

classes of employees in the Special Exposure Cohort.

This SEC class includes all workers during the SEC class period. Because of the identified dose 

reconstruction infeasibility, all dose reconstructions for all workers having employment during the SEC 

class period are considered partial dose reconstructions. If monitoring data are available for workers 

included in the SEC class, dose is to be assigned as appropriate based on such data; however, such 

dose reconstructions are still considered partial dose reconstructions because NIOSH has determined 

that internal exposures during the SEC class period cannot be bounded.

2.0 SITE DESCRIPTION AND OPERATIONAL HISTORY

When the U.S. government and its contractors first became interested in uranium, Carbide and 

Carbon Chemicals Corporation (CCCC) 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 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. The plant worked with two types of 

radioactive material:

• Refined uranium materials and preprocessed (partially refined) domestic ores, and

• 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 nonuranium 

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.

Carbide and Carbon Chemicals Corporation 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 (R&D) on uranium processing for the 

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

develop procedures for the Ceramics Plant. The Ceramics Plant produced uranium materials for the 
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MED and AEC from 1943 to 1946 and from 1947 to 1949. From 1947 to 1949 (and perhaps earlier), 

Linde received UO2 for processing from Mallinckrodt Chemical Works (AEC 1949a). The plant also 

produced nickel material for the K-25 diffusion barrier. The Ceramics Plant was in standby from mid1946 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 involved 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).

The following sections provide more detailed histories of Linde facilities, radiation sources, and 

processes that related to MED/AEC radiation exposures. Key dates are noted in Table 2-1.

Figure 2-1 is based on Figures 3-1 and B-11 of BNI (1982), Figure 3 of Frame et al. (1981), and 

CCCC (1945a).

Table 2-1. Key dates. 

Assumed date or perioda Event or activity

Ceramics Plant

06/01/1943–07/31/1946b Step I production (U3O8 from ore or sludge)

06/01/1943–11/30/1943 Preprocessed ores (domestic and scrap)

12/01/1943–11/12/1944 African ores

11/13/1944–01/31/1946 Preprocessed domestic ores

02/01/1946–02/28/1946 African ores (48%) and preprocessed domestic ores (52%)

03/01/1946–06/30/1946 African ores

07/01/1946–07/31/1946 African ores and preprocessed ash

04/27/1943–03/08/1944c Step II production (UO2 from U3O8)

07/25/1943–06/26/1946d Step III production (UF4 from UO2)

08/01/1946–09/14/1947 Standby

09/15/1947–10/31/1947e Step III rehabilitation

11/01/1947–06/30/1949f Step III production

10/01/1944–02/28/1946g Production of nickel material for K-25 diffusion barrier

07/01/1949–03/31/1950h Cleanup of Building 30

07/01/1949–12/31/1954i Cleanup of Linde Ceramics Buildings

Before 1978 Demolition of Buildings A and B (office buildings on Figure 2-1)

After 1954 Postcleanup period

08/31/1981j Demolition of Building 37 began

08/31/1996j Demolition of Building 38 began

09/05/1998k Demolition of Building 30 began

09/30/2000j Soil remediation began (scheduled for completion in 2007)j

Tonawanda Laboratory

10/01/1942–07/31/1946l MED-related R&D operations

08/01/1946–12/31/1946m Cleanup period

After 1947 Postcleanup period

04/30/2004j 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 in the data source, and demolition activities were assumed to begin on 

the last day of the start month in the data source.
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b. All Step I dates and data are based on Table B-1 of Wallo, Vierzba, and Roberts (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 

Section 4.1.3.2).

c. CCCC (1946a, p. 40).

d. CCCC (1946b, p. 42).

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

f. Start date based on Rennich (1947). End date based on Heatherton (1948a).

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 based on Eisenbud (1950).

i. Start date assumed to be day after end of Step III production; end date estimated based on Harris (1954).

j. Pilon (2004).

k. USACE (1998).

l. Start date based on Jenness and Ewing (1943). End date assumed to be the end date of 1943 to 1946 production at 

the Ceramics Plant.

m. Start date based on estimated end date of R&D period; end date estimated.

2.1 FACILITIES

The Linde Ceramics Plant and Tonawanda Laboratory were on land that was 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 that were involved in MED/AEC work are shown in Figure 2-1. The 

Tonawanda Laboratory occupied Building 14, and the Ceramics Plant used Buildings 30, 31, 37, 38, 

and A. Building B contained MED offices. Some buildings at the Linde Ceramics site have not been 

designated as a DOE or an AWE facility, so they are not described here. Ownership of the Ceramics 

Plant buildings was transferred to UCC after the site cleanup that began with the shutdown of 

production in 1949. The transfer probably was completed in 1957. In the 1990s, the site was

acquired by Praxair (USACE 2003).

2.1.1 Building 14 (Tonawanda Laboratory)

Building 14 housed Tonawanda Laboratory. The U.S. Army Corps of Engineers (USACE) has 

described its use as follows (USACE 2003):
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Figure 2-1. Linde Ceramics Plant and Tonawanda 

Laboratory buildings on CCCC Tonawanda site.

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.

A 1942 Tonawanda Laboratory directory listed approximately 200 employees (CCCC 1942). The 

locations of most were designated by room numbers. Some employees were said to be in the Eng. 

Lab and others in the Proving Lab. A 1944 employee list reported some to be in the Pilot Plant 

(CCCC 1944a). 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).
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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).

2.1.2 Building 30 (Ceramics Plant)

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

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

floor. Process workers operated equipment from the second floor (Dupree 1983a, 1983b). The 1949 

to 1950 decontamination work by CCCC 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 from September 5 to 19, 1998 (USACE 1998).

Figure 2-2. Building 30 layout (Heatherton 1950).

2.1.3 Building 31 (Ceramics Plant)

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

p. 1-63). 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 (CCCC 1945a). Building 31 might have 

been the final location of the Linde Tonawanda nickel processing operations.
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Figure 2-3. Building 30 layout (CCCC undated a).
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As of 1976, the building was described as housing fabrication facilities, offices, and storage areas, 

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

1976 survey (ORNL 1978). Building 31 was decontaminated in 1997 (USACE 2002a). As of 2004, it 

was still standing and in use (Pilon 2004).

2.1.4 Building 37 (Ceramics Plant)

Building 37 was a small appendage to Building 39, which was east of Building 38 (Figure 2-1). 

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

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

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

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

2.1.5 Building 38 (Ceramics Plant)

Building 38 was used for Step III. Figure 2-4 shows its layout. Radioactive contamination was found 

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

Figure 2-4. Building 38 layout (Klevin 1949a).

2.2 SOURCE TERMS FOR 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 
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pitchblende; some was torbernite (Wallo, Vierzba, and Roberts 1981, p. 1). Neither African ore 

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

uranium but also 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 (Wallo, Vierzba, and Roberts 1981, p. 5). The preprocessing 

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

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

consisted primarily of 234U, 235U, and 238U (in their natural abundances) and their 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 Linde used is not known. A review of the chemical 

processing concluded only that “the radium in the domestic ore was significantly less than that in the 

African ores” (Wallo, Vierzba, and Roberts 1981, p. B-1).

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

ore that contained only uranium and its short-lived progeny would be due to alpha and beta 

emissions. In contrast, radium and other progeny in the African ores would produce, in addition to the 

alpha and beta emissions, 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 (up to 35% or less by mass) might have been used on a limited basis for K-25 

barrier testing and R&D 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

Carbide and Carbon Chemicals Corporation 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 t 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, CCCC entered into contract W-7401 eng-14 with the MED to design and 

construct facilities at the Ceramics Plant for processing uranium ores and to produce U3O8, UO2, and 

UF4 (Cornell 1942; Marshall 1943). The contract also called for R&D into processes that would be 

used in the Ceramics Plant facilities.

2.3.1 Uranium Research and Development

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 of work 

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

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

Ceramics Plant.

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

producing 2 to 3 t 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.
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Available documentation from the Laboratory includes three research reports (Brimm 1943a; 

Skinner 1944; Wiesendanger 1944) and a final report (CCCC 1946c). The final report discussed work 

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 because 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.

The second area of Tonawanda Laboratory work related to conversion of UF6 process gas to UO3. 

Two research reports (Brimm and Schubert 1945a, 1945b) and a final report (CCCC 1946d) indicated 

that the associated pilot plant operations used pounds rather than tons of material.

A later Linde research report on uranium processing from 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 Uranium Production

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

III. Step I (production of U3O8) typically involved ores that contained 3% to 20% U3O8 by weight. 

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

were processed, and about 2,300 MT of U3O8 were produced (Wallo, Vierzba, and Roberts 1981, 

Table B-1).

Step I (production of U3O8) 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 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 lb of uranium ore concentrates from Colorado and 8,900 lb of residue, which were the 

“property of Contractor as a result of operations before his present CPFF [cost plus fixed-fee] work for 

the MED (Contract No. W-7401 eng-14).” The transfer was made under contract W-17-028 eng-29 on 

May 4, 1945 (USEO 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.
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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 to 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 most of the remaining 

impurities and left the uranium in solution as sodium uranyl tricarbonate. The slurry was filtered in 

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 to leave 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. Some of the other chemical additives that were 

used at various points were different (CCCC 1946e; Aerospace 1981, p. A-1).

Step I Operations

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

bags. Each ore bag weighed 50 to 80 lb. For African ores, the car doors were left open for 12 to 

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; CCCC 1946e; 

Cranch 1944a).

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 t 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 0.5 lb 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 that were 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 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 that contained 24 rectangular leaves with each leaf approximately 

35 ft2 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-in. 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 
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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 

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, page 3; Olevitch 1944; 

CCCC 1946e; 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,

1983b).

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 either 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 (CCCC 1946a; 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” (CCCC 1946a).

2.3.2.3 Step III (UO2 to UF4)

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

+ →+ UO 4HF UF 2H O 2 4 2

The reaction was carried out at 1,000°F. Gaseous anhydrous HF gas was flowed over the UO2, 

which was placed on magnesium trays stacked on spacer bars inside 9-in.-square tubes 10 ft long 

that were inside a furnace (CCCC 1946b; 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 retreatment. Good 

material was pulverized, blended, sampled, packaged, weighed, and shipped (CCCC 1946b; Hickey, 

Crawford-Brown, and Tankersley 1988, p. 8).

2.3.3 Nickel Production

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

nickel powder, a nonradioactive 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).

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

(USACE 1944) states that Linde was contracted on April 7, 1944, via letter contract W-7418 eng-51 to 

construct and equip a plant to produce 80 t 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 that were used in 

connection with contract W-7401 eng-14 (the contract for the uranium refining work). A drawing that 

was 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. 

Therefore, nickel processing might have occurred in Building 31 (see Figure 2-1). 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 could be that nickel 

processing initially occurred in the Step II area and later was moved to Building 31.

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

processing is found in a 1945 plant drawing (CCCC 1945a). The visible portions of a partially 

obscured label on the drawing are consistent with the interpretation that it says “Nickel Plant.” In 

addition, 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 CCCC (1945b, 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 Other Radiological Activities

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 in 

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).

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 1948b; ORNL 1978; BNI 1982); resuspension would have produced airborne 

radioactivity. One source of ground contamination and airborne radioactivity was the ore unloading 
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process, which involved transporting ore in buggies – sometimes in bulk and sometimes in bags – 

from boxcars to Building 30 (see 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/hr gamma and 

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

materials (Heatherton 1948b).

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 unloaded drums of K-65 product that was shipped by rail from Mallinckrodt in 

St. Louis to the LOOW at Modeltown, New York (Wolf 1949a; Heatherton 1949a; Kelley 1949). Linde 

film badges were worn during this work.

At the April 18, 2005, Worker Outreach Meeting, strontium was mentioned. At the June 27, 2005, 

Worker Outreach Meeting, it was clarified that the strontium was in the form of a 90Sr sealed source. 

This source was reportedly brought to the site in the late 1950s. 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.

2.3.5 Utility Tunnel Complex

Description

The Linde site had a system of utility tunnels, constructed at different times, to carry steam, electricity, 

water, telephone lines, and other utilities from one part of the plant to another (CCCC ca. 1937; UCC 

1957, 1961; Coutts 2005). Documentary evidence shows that the first tunnel section was built in 1937 

and ran from the powerhouse (Building 8) past the Tonawanda Laboratory (Building 14, also called 

the Proving Laboratory) to Building 10. This part of the tunnel complex was not found to be 

contaminated with radium because the pre-World War II uranium-refining operations in Building 14 

used preprocessed ores from the Colorado Plateau vanadium deposits, which did not contain radium. 

Borehole data from the site cited in the 1978 and 1982 Formerly Utilized Sites Remedial Action 

Program (FUSRAP) surveys confirmed that no excess radium was found in the soils near this section 

of tunnel (ORNL 1978; BNI 1982). Another section of tunnel was constructed in 1957 near Buildings 

57, 58, and 31 in the northeastern area of the Ceramics Plant, and an extensive addition to the tunnel 

system was done in 1961 when the 1937 and 1957 tunnels were linked by new tunnels that ran 

between Buildings 30 and 31 and then branched south to Building 8 and west past Buildings 70, 2, 

and 2A. The 1957 and 1961 tunnel sections ran through areas of soil that were contaminated by 

radium-bearing ore and were subject to radon infiltration from this source.

The underground tunnels housed gas, electrical, water, steam, telephone, and other utility lines. They 

had a maximum dimension of about 8 to 10 ft high and 15 ft wide. They also had exhaust fans 

(approximately 6 ft in diameter) in some sections of the tunnel that pulled air through the tunnels 

(Murray 2010a, 2010b, 2010c), and the average air exchange rate in any particular tunnel section was 

estimated by the USACE as approximately once every 10 hours. The tunnels had few stairway 

access points and several locations where covered manways could be used to access the tunnel 

(USACE 2002b). These tunnels were routinely inspected and accessed for maintenance activities. 

However, no uranium process piping was ever installed in these tunnels (USACE 2002b).

The processing of uranium ore in the 1940s resulted in contamination of the site building surfaces, 

soils, building sumps, and sanitary sewer system. In addition, approximately 55 million gal of liquid 

waste effluent was injected into seven wells that were used for liquid effluent disposal from 
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February 27, 1945, through mid-July 1946. The wells were in two general areas: three near 

Building 8 and four in the area of Buildings 30 and 38. The wells were drilled approximately 150 ft 

(40 m) deep through a clay formation into a gravel and sand layer and the bedrock layer. The liquid 

effluent was injected down into the wells between 27 to 46 m (90 to 150 ft) below the surface at a 

temperature of approximately 60°C (140°F). The effluents primarily contained ions of sodium, sulfate, 

sodium carbonate, sodium bicarbonate, and chloride. Minor concentrations of vanadium, cobalt, 

nickel, molybdenum, uranium, and radium were also present. This liquid had a high pH (greater than 

10) and a high concentration of total dissolved solids (greater than 20,000 ppm). Problems with 

clogging of the wells and fluid refusal that were reported during the injection process were most likely 

to have been caused by the formation of precipitates in the bedrock fractures and pore spaces of the 

sediments. The precipitates were formed when the high-temperature high-pH effluent containing 

dissolved solids became mixed with the cooler-temperature, natural formation water, which resulted in 

the precipitation of most of the dissolved metals (including uranium and radium) from the solution. 

Under high-pH conditions, thorium hydroxide, barium sulfate, calcium carbonate, and calcium sulfate 

would also have precipitated. These precipitates contributed to the pores becoming clogged during 

effluent injection. It has been suggested that runoff water contaminated from these sources entered 

the service tunnels throughout the lifetime of the site and resulted in a constant contamination level 

over time.

Limited radiological characterization of these areas was conducted during the 1978 and 1981 

FUSRAP surveys of the Linde site (ORNL 1978 and BNI 1982). A detailed characterization of these 

tunnels was performed in the summer of 2001 (USACE 2002b).

Area Geology and Hydrology

Sections 3.4 and 3.5 of the 1993 remedial investigation report for the Tonawanda site (BNI 1993, 

p. 230) present information on the Linde site geology including structural features, physical and 

hydrogeologic properties, and the geochemical properties. Section 3.5.5 of the report describes, in 

some detail, the soil interval between the ground surface and the water table, known as the vadose 

zone, and how flow occurs through this layer. The report made a conservative estimate of vertical 

groundwater velocity and applied a retardation factor to account for the slower movement of 

contaminants through the vadose zone due to interactions between the contaminants and the soil 

particles. The report states the uranium contamination in groundwater would migrate 47 times slower 

than the estimated vertical groundwater velocity of 0.09 m/yr (0.3 ft/yr). Because the retardation 

factors for thorium and radium are typically greater than for uranium, thorium and radium would 

migrate at a slower rate.

The Aerospace report (Wallo, Vierzba, and Roberts 1981) indicates the convoluted rock beds in the 

area have a slight southerly slope, about 28 ft/mi. Due to the composition of the formations in the 

area, it is very difficult to predict the direction and flow rate of water at any given point in the 

formations. In carbonate rock, the solution openings that are caused by water dissolving the 

carbonate rock are irregularly distributed both horizontally and vertically.

Tunnel Access

The utility tunnels would have been routinely accessed by inspectors and maintenance personnel 

throughout the site history. There were stairwells from the tunnel up to Buildings 2, 8, 14, and 30 

(Building 30’s access was by means of a ladder at Junction Box 6), but it was not general practice for 

employees to use them to get from one building to the other, and the practice was not condoned by 

the company (Murray 2010c). However as a favorable to claimant assumption, it is assumed that all 

other personnel may have used the tunnels for transit. A transit occupancy time of 10 minutes per 

workday (wd) is assumed.
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Uranium in Liquid Effluent

Uranium losses in the effluents were monitored by Linde. From mid-1944 until 1946, the losses were 

reported on a regular basis in the weekly progress reports and, before that, in memoranda or 

occasional progress reports. The weekly averages of analyzed uranium oxide concentrations in the 

effluents from April 1944 to July 1946 (from progress reports) ranged between 0.011 and 0.064 g of 

uranium oxide per liter of effluent, with the average being about 0.026 g/L (Wallo, Vierzba, and 

Roberts 1981, p. 11). This would equate to an average process loss of about 26 kg of uranium oxide 

per million L (220 lb per million gal) of effluent during the entire period when the wells and storm 

sewer were being used for disposal (February 27, 1945, to mid-July 1946).

Concentrations of uranium oxide in the effluent were somewhat higher during the period when the 

sanitary sewer was used for disposal of the effluent. It is estimated that the concentrations averaged 

0.15 g/L in 1943 and 0.03 g/L during the first 3 months of 1944, or about 1,200 and 250 lb of uranium 

oxide per million gallons, respectively.

Assuming these loss rates and the volume of effluent that was disposed of in the wells (55 million gal), 

approximately 5,400 kg (12,000 lb) of uranium oxide (about 3 Ci of natural uranium) were discharged 

to the wells. The remainder of the process effluents that were discharged to the storm sewer during 

this period (approximately 212 million L or 56 million gal) would have contained approximately 

5,600 kg (12,000 lb) of uranium oxide (again, about 3 Ci of natural uranium). These loss rates are 

consistent with the assumption that Linde maintained uranium losses in the effluents to about 0.5% of 

the total uranium present. If, as reported, 1,812 MT (1990 t) of uranium oxide (at about 97% 

extraction efficiency) were produced from April 1944 to July 1946, then about 9.3 MT (about 10 t) of 

uranium oxide (5 Ci of natural uranium) were lost in the effluents, which is in good agreement with the 

values estimated by using the effluent concentration data. The 1981 survey (Frame et al. 1981) 

estimates the maximum concentration of uranium in the effluent released to the disposal wells and 

storm sewers at 4 × 10-5 μCi/mL.

Radium in the Liquid Effluent

Estimates of the maximum amount of radium that was discharged during the processing of the L-30 

and L-50 ores can be made based on the contractually specified return of at least 95% of the radium 

from the processed ore to African Metals. Processing operations reportedly held losses to less than 

3%, meaning 97% of the radium remained in the residues. Given a total of 986 MT (1,080 t) of U3O8

that was produced from the L-30 and L-50 ores and a uranium extraction efficiency of about 97%, 

then 862 MT (950 t) of uranium, or less than 595 Ci of natural uranium, were in the ore. This equates 

to an estimate of 290 Ci of 238U in the ore. Assuming the 226Ra was in equilibrium with the 238U, the 

maximum effluent losses were calculated to be 8.5 Ci of 226Ra. A similar analysis for the R-10 ore, but 

assuming a 95% extraction efficiency, suggested that a maximum of 2.7 Ci of radium was lost during 

the processing of this ore (Wallo, Vierzba, and Roberts 1981, p. 14).

Based on data in the progress reports, approximately 169 to 173 million L (about 45 million gal) of 

effluent were released during L-30 and L-50 processing, and about 77 million L (about 20 million gal) 

were released during R-10 processing. Based on this volume, the maximum concentration of radium 

in the effluent would be about 4.6 × 10-5 μCi/mL. This value represents the maximum radium content 

in the uranium carbonate solutions and was based on the amount of radium that was not precipitated 

on the Moore filters after the digest circuit. Of the unprecipitated radium at this stage, a portion

probably precipitated during the liquor purification (lead, vanadium, and /or phosphorus removal step), 

but the largest fraction of the remaining radium would be expected to have precipitated with the 

uranium after the addition of the caustic during the soda salt precipitation step. The remainder would 

be discharged with the effluents.

A letter from the University of Rochester to the MED suggested that the radium content of the filtrates 

from the processing of R-10 ore was “on the order of 0.25 × 10-8 grams/liter” (2.5 × 10-6 μCi/mL). This 
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value is less than 1/18 of the maximum value that was estimated in the analysis above, which 

supports the theory that much of the radium that escaped the Moore filters could have been 

precipitated during the liquor purification or the uranium precipitation and therefore was not 

discharged to the wells or storm sewer. The implications are that the radium that was released in the 

effluents from the African ore processing could be on the order of 0.6 Ci rather than 11.2 Ci (Wallo, 

Vierzba, and Roberts 1981, p. 15).

Soil Analysis near Injection Wells

During the remedial investigation (BNI 1993, p. 366) data was collected to determine the nature and 

extent of contamination at the Linde Ceramics facility. Borehole samples were analyzed, including 

borehole B29R23 just to the northwest of Building 30 near a former injection well that was found just 

inside the building. This borehole was sampled at three intervals (depths), and the samples showed 

low-level radium-226 concentration in the first interval (BNI 1993, p. 389), possibly as a result of 

spilled effluent at the site of the injection well. The concentration decreased at the deeper sample 

interval.

2.4 PROTECTIVE MEASURES

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

(CCCC 1945, 1946a, 1946b, 1946e; Dupree 1983a, 1983b). 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 about 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 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 who worked with hydrofluoric acid.

According to interviewees, use of respirators was often at the worker's discretion, but they were 

“usually used” (Dupree 1983b). Inspectors reported mixed compliance with recommendations to wear 

respirators. The author of a June 1944 inspection report commented, “The wearing of respirators 

remains difficult to enforce” (Ferry 1944a). 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. In his report, the inspector noted, 

“The men wore respirators, but [Name redacted] stated that this was the case only because he was 

present” (Tybout 1945a). 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 WORK HOURS 

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

worked there. A September 1942 Laboratory directory (CCCC 1942) lists approximately 
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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 (CCCC 1944a) 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 workers’ jobs. A December 1945 job description specifies the duties of 51 categories of 

workers and lists a department code for each position (CCCC 1945b). Because 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 might help identify the 

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

that were associated with various worker activities as determined in a 1988 study of worker hazards 

(Hickey, Crawford-Brown, and Tankersley, 1988). 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 2-2. 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.

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

describe a day at work. Default assumptions are 250 wd/yr and 8 hr/wd: this results in 2,000 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 integration over calendar days.

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

involving 8-hour shifts 6 d/wk (Dupree 1983a, 1983b, 1983c). A 9-hour workday with a lunch period 

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

(Dupree 1983b, p. 4; MED undated, p.1). An 8-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 6-day 

workweek to a 5-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).

Table 2-3 provides a list of some Linde references that include worker names and sometimes 

additional job information.
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Table 2-3. Worker information. 

Item Reference

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

titles are not stated.

CCCC 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 1948c

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

and stores attendants), 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, August 26, 1949. Klevin 1949b, p. 11

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

operations through December 31, 1953, were assumed to have been as shown in Table 2-4.

Table 2-4. Ceramics Plant assumed work schedule (including lunch and breaks), 1942 to 

December 31, 1953. 

Period Start End hr/wd wd/wk wk/yr

Preproduction 10/01/1942 04/26/1943 9.0 6 50

Production 04/27/1943 07/31/1946 9.0 6 50

Standby 08/01/1946 09/14/1947 8.5 6 50

Rehabilitation and production 09/15/1947 06/30/1949 8.5 6 50

Cleanup 07/01/1949 12/31/1950 8.5 6 50

Cleanupa 01/01/1951 12/31/1953 8.5 5 50

a. Assumed date of transition from 6- to 5-day workweek (based on Dupree 1983b, p. 4). 

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 1948d, 1948e, 1948f). 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 1949c,

1949d).

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

radioactivity in Building 30 to enable its release to Linde for unrestricted use (Heatherton 1950). After 

removal of the bulk of the process equipment, 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.
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Workers who performed surface abrasion activities were provided with U.S. Bureau of Mines (USBM)-

approved rubber hoods 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 respirators or, 

for short periods, USBM respirators approved for all dusts.

An AEC internal memorandum from 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 from April 1954 (Harris 1954) reported levels after decontamination. Average 

readings on categories of equipment to be left at the plant ranged from 0.05 to 0.6 mrep/hr 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 “mrep/hr/ft2

” beta plus gamma. This was considered excessive. The memorandum cited data 

that indicated this could be reduced to 0.065 mrep/hr/ft2 by covering contaminated floor areas with 

asphalt tile. It recommended release of the building to Linde once this occurred.

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

2.7 POST-MED/AEC OPERATIONS

FUSRAP began in 1976. ORNL surveyed the Linde Tonawanda site from October 18 through 

November 5, 1976, to determine if remediation would be required (ORNL 1978). 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 in association with Building 57, and an additional review of the BNI (1993) 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, 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 Plant buildings that were involved in 

MED/AEC work except Building 31. Table 2-1 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 ESTIMATION OF INTERNAL EXPOSURE, NOVEMBER 1, 1947, TO DECEMBER 31, 1953

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

November 1, 1947, until December 31, 1953. 

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 that 

are estimated in this section.
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This document assumes that Tonawanda Laboratory primary internal exposures occurred from 

October 1, 1942, to December 31, 1946. At the 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, 

1953 (operations and cleanup). From October 1, 1942, 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 that remained 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 progeny and radon would have decreased at the 

Ceramics Plant after August 31, 1946, because ore was no longer processed during this period. 

Continued lower level exposures to uranium progeny and to radon were assumed because some 

radioactive waste was disposed of on the 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 post-1946 period.

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

uranium and its progeny.

3.1 AIR MONITORING FOR PARTICULATE INTAKES

During World War II and the time of the 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 in Rochester, New York, 

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 dpm/m3 for natural uranium and was based 

primarily on animal studies. The Medical Division of the AEC 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 (PL) (AEC 1949a). Contractors of the AEC 

often used the term maximum allowable (air) concentration (MAC) interchangeably with PL and 

reported air-sampling results as multiples of the MAC (Stewart 1952; Klevin and Weinstein 1953). 

When considering air concentrations that were 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, CCCC, 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 that were 

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

time-weighted average air concentrations by job category. Up until the time of the 1949 AEC report 

(AEC 1949a), 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 6,250 μg/m3

), 9% 

were exposed at 25 to 125 MAC (1,250 to 6,250 μg/m3

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

(less than 1,250 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 redesign 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 1959).

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 
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not elaborate on internal exposures before its 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 before 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 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.

After the ore processing, Linde began a standby period. 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.

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 before standby. In addition, workers could have worked in multiple uranium process areas. This 

analysis assumed that types F, M, and S were available before 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.

3.2.1 Uranium Urinalysis Data

Medical urinalyses for 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 1944b), but no records are available to indicate the analyses were done. A 

page titled, “Schedule of Examinations, Contract AT-30-1-GEN-165” (Author unknown ca.1947), 

which commenced sometime in 1947 (Rennich 1947), indicates that 60 cm3 of urine from each 

employee was to be sent monthly to the University of Rochester, the location of one of the earliest 

bioassay analysis programs. Individual uranium urinalysis data from November 1947 to 1950 are 

available for some Linde Ceramics workers (CCCC 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 before 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 mg/L, with many of the higher results reported in the November to December 1947 

“preemployment urines.” Because cleanup/preparation activities occurred before collection of these 

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

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

that was 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 results are consistent with intakes of soluble uranium.
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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 2 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 1948). The uranium fusion photofluorimetry urinalyses that were performed 

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

default detection threshold for uranium urinalysis was 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 1959). However, 

the January 1948 uranium urinalysis results AEC transmitted to Linde included a note that indicated 

the minimum detectable amount might have been much higher (Wolf 1948):

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 November 1948 uranium urinalysis results included a similar note, 

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

determination for photofluorometric analysis. The January 1948 determination level of 0.1 mg/L is 

assumed to be 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 

reported numbers.

The AEC 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 3 months (Wolf 1948). Linde’s reply 

requested clarification because it had previously understood that resampling was to occur monthly 

(Heatherton 1948a).

If monitoring data are available for workers, dose is to be assigned as appropriate based on such 

data; however, such dose reconstructions are still considered partial dose reconstructions because 

NIOSH has determined that internal exposures from uranium during the SEC class period cannot be 

bounded (Sebelius 2012).

3.3 RADIUM

DHHS has determined, and NIOSH has concurred, that it is not feasible to reconstruct radium 

exposure that occurred before November 1, 1947 (Leavitt 2005). Information on radon exposure 

before November 1, 1947, is provided only as a basis for extrapolation of later exposures and is not 

intended to be used for this period in which reconstruction of internal dose has been determined to be 

infeasible.

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.

Radon Breath Data

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 from Linde workers between June 12, 1944, and 

January 17, 1945 (CCCC 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, room 

background concentrations were reported respectively as 0.6 and 0.3 pCi/L (Tybout 1944b). The 

measured radon levels were given for some but not all of the samples. Reported radon breath results 
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ranged up to 2.2 pCi/L. The method of deriving the 226Ra level was not stated, but in many cases, the 

reported burden of radium in micrograms was numerically equal to 10% of the breath radon 

concentration in picocuries. Radon breath results are the starting point for dose reconstruction, so it 

might be necessary to back-calculate breath results from either 226Ra burdens or tolerance levels. 

The original records might have sufficient information for this determination. This information has not 

yet been tabulated for inclusion in this site profile.

3.4 URANIUM PROGENY

DHHS has determined, and NIOSH has concurred, that it is not feasible to reconstruct uranium 

progeny exposure that occurred before December 31, 1953 (Leavitt 2005 and Sebelius 2012).

3.5 RADON

DHHS has determined, and NIOSH has concurred, that it is not feasible to reconstruct radon 

exposure that occurred before November 1, 1947 (Leavitt 2005). Information on radon exposure 

before November 1, 1947, is provided only as a basis for extrapolation of later exposures and is not 

intended to be used for this period in which reconstruction of internal dose has been determined to be 

infeasible.

More than 200 measurements of radon concentrations were made during 1942 to 1946 pilot plant and 

production processing of African ore at Linde (CCCC 1944–1946). The early measurements were 

made using liter glass bulbs that were evacuated immediately before sampling to below 50 μm 

(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 222Rn), the 

concentration C in picocuries per liter was converted to potential alpha energy concentration (PAEC) 

in units of working level (WL) using:

PAEC C F =×÷ (100 pCi/L per WL) (3-1)

where the equilibrium factor F was taken as 0.4, a value that is recommended by International 

Commission on Radiological Protection (ICRP) Publication 32 (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-months (WLMs) was obtained from the PAEC and the 

number of months of exposure M using:

PAEE PAEC M = × (3-2)

where a month is assumed to be 170 work hours.

3.5.1 Ceramics Plant Radon Exposures

During Ceramics Plant preproduction 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 from Tonawanda Laboratory work. No direct 

measurement of this was available. An estimate was made based on the lowest measured indoor 

concentrations at the Ceramics Plant during African ore processing. These were viewed as indicating 

the upper limit to the outdoor concentration because outdoor air was drawn indoors for ventilation. 
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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 pCi/L. Therefore, 10 pCi/L was 

taken as the estimated outdoor concentration.

Measured radon concentrations during processing of African ore at the Ceramics Plant are 

summarized in Table 3-1. Based on geometric means (GMs), locations were grouped into three 

categories – high, medium, and low – and the GM and geometric standard deviation (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).

Table 3-1. 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 GM, a <LOD value was assumed to equal the LOD.

b. GSD could not be determined.

After the end of African ore processing, concentration in the main ore processing building, Building 30, 

was assumed to remain at the 10-pCi/L level that was measured during the second period of domestic 

ore processing until the end of cleanup of Building 30. Concentrations in other Ceramics Plant 

buildings were also assumed to be 10 pCi/L until the end of cleanup in those buildings. Because the 

locations of many workers are likely to be unknown, it was assumed that all workers were exposed to 

10 pCi/L of radon from November 1, 1947, through December 31, 1953. Table 3-2 lists 

concentrations and exposure rates.

Table 3-2. Ceramics Plant worker radon exposure rates, 1947 to 1953. 

Period/work location

Time-weighted concentration 

(pCi/L)

Exposure rate

(WLM/yr)

11/01/1947–12/31/1953

All workers 10.0 0.480

3.5.2 Tonawanda Laboratory Radon Exposures

Few measurements were available for the Tonawanda Laboratory, 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 occurred 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 

before the beginnings of the two African ore processing periods at the Ceramics Plant. For dose 

reconstruction, it was assumed that African ore processing at Tonawanda Laboratory occurred during 
 

 

 

 

 

 

 

 

 

 

 

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the entire MED period, but that the peak concentrations of radon were equal to the medium values at 

the Ceramics Plant. It was 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 cleanup, radon exposure in the Laboratory was based on 

the highest GM PAEC that was determined for a Tonawanda site building from measurements in 

1981. That value was 1.68 × 10-02 WL for Building 31 based on analysis of data in BNI (1982, 

Table B-3). Table 3-3 summarizes the assumed radon concentrations and resultant exposures. 

Table 3-3. Tonawanda Laboratory radon exposure rates, 1947 to 1953. 

Period Time-weighted concentration 

(pCi/L) 

Exposure rate 

Start End (WLM/yr) 

Postcleanup 

11/01/1947 12/31/1953 — 0.202 

Radon and Radon Progeny 

The radon exposure rates that are provided here are annual PAEE rates in WLM/yr. 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 3-1). 

3.6	 INHALATION INTAKE ESTIMATES OF RADIOACTIVE PARTICULATES AT THE 

TONAWANDA LABORATORY 

DHHS has determined, and NIOSH has concurred, that it is not feasible to reconstruct uranium 

exposure that occurred before December 31, 1953 (Leavitt 2005 and Sebelius 2012). 

3.7	 INGESTION INTAKE ESTIMATES AT THE TONAWANDA LABORATORY 

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/wd. To adjust this to ingestion intake per calendar day, 0.685 dpm/wd was multiplied by 

250 wd/yr and divided by 365 d/yr, which equals 0.469 dpm/d. In accordance with NIOSH (2004), the 

f1-value used for inhalation dose calculations is to be used for ingestion dose calculations. 

3.8	 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND SUMMARY 

DHHS has determined, and NIOSH has concurred, that it is not feasible to reconstruct internal 

exposures for uranium and its progeny that occurred before December 31, 1953 (Leavitt 2005 and 

Sebelius 2012) at the Ceramic Plant. 

Intake estimates and radon exposures that were estimated for Linde workers are shown in Tables 3-4 

and 3-5. 

Table 3-4. Radon exposure for the Ceramics Plant, November 1, 1947, to December 31, 1953. 

Radionuclide Start End 

Intake 

route 

Absorption 

type Intake or exposure Units 

Rn-222 11/01/1947 12/31/1953 Inhalation - 0.48E+00 WLM/yr 
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Table 3-5. Radon exposure for the Tonawanda Laboratory, November 1, 1947, to December 31, 

1953. 

Radionuclide Start End

Intake

route

Absorption

type Intake or exposureb Units

Rn-222 11/01/1947 12/31/1953 Inhalation - 2.021E-01 WLM/yr

a. Choose same f1-value as used for inhalation per NIOSH (2004).

b. The dose distribution is assumed to be lognormal with a GSD of 3.43.

Uranium bioassays are available for some workers during the period from November 1947 through 

January 1950. If uranium bioassays are available for a worker, they should be used to reconstruct an 

individual’s dose. 

4.0 ESTIMATION OF EXTERNAL EXPOSURE, 1942 TO DECEMBER 31, 1953

Because of the SEC determinations (Leavitt 2005, Sebelius 2012) that it is infeasible to adequately 

reconstruct internal dose during the period from October 1, 1942, through December 31, 1953, dose 

estimates for this period are considered partial dose estimates.

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, 

1953. In some parts of this section, measurements and parameters for the period after 1953 are 

analyzed because they provide information used to estimate external exposures before 1954. 

Occupational medical exposures are treated in Section 5.0. Dose reconstruction for the period after 

1953 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.

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, or 

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 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 from the Ceramics Plant (see Table A-6).

For the purpose of calculation of organ dose, all exposure geometries are assumed to be anteriorposterior (AP).

Burlap Bag Exposures

Based on the assessment in Attachment D of this TBD, NIOSH’s review of historical records indicates 

that the period during which the burlap bags were staged and burned is within the current SEC period. 

NIOSH has found no other credible evidence, other than a former employee affidavit that indicated 

that two pallets of filled burlap bags were in Building 30 at the Ceramics Plant, to support a potential 

burlap bag exposure issue outside of the SEC period. If credible information is presented in a 

claimant interview or documentation related to an individual claim, to establish the worker’s 

occupational proximity to the burlap bags, dose reconstructors should use the external dose estimates 

in Attachment D to assess this external exposure to support assessment of this external dose for the 

purpose of individual dose reconstruction.
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4.1 CERAMICS PLANT BETA AND GAMMA EXPOSURE

4.1.1 Preproduction, 1942 to 1943

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- by 15-ft 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 in millirep per hour was measured at contact at the square’s center 

and corners and at 3 ft 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 4-1 display the results of a statistical analysis of the 

data.

Another comprehensive survey of Building 30 was made in 1976 (ORNL 1978). Beta plus gamma 

levels in millirad per hour were measured at 1 cm from the surface, and gamma levels microroentgen 

per hour were measured at 1 m. Table 4-2 displays the results of a statistical analysis of the data. 

The measured levels were similar to the postdecontamination results of 1950, which indicated little 

change in conditions. Because 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 in Table 4-1 for the pre- and postdecontamination results.

The measurements just before decontamination in 1949 and 1950 were made after the building had 

undergone vacuum cleaning and flushing. Brief, semiquantitative reports were available for two 1-day 

Table 4-1. Floor and wall radiation in Building 30 before and after the 1949 to 1950 decontamination. 

Time Location

Measured beta plus gammaa Estimated

Median (mrep/hr) GSD Beta (mrem/hr) Gamma (mR/hr)

Before vacuum 

cleaning and flushing

Contact (b) (b) 2.03c 0.131c

3 ft (b) (b) 0.676c 0.131c

Predecontamination 

(just before 

decontamination and 

after vacuum cleaning 

and flushing)

Contact 0.719 1.67 0.675d 0.0438d

3 ft 0.240 1.67 0.225d 0.0438e

After decontamination Contact 0.311 2.04 0.292d 0.0189d

3 ft 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/hr was stated). These were used with the computer program LOGNORM4 (Strom and 

Stansbury 2000) to estimate the median and GSD. For measurements at 3 ft, 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 predecontamination values. The factor three is based on April 1949 data; 

see discussion in text.

d. Estimates based on beta and gamma percentages in Table 4-2.

e. Assumed to equal contact gamma exposure rate.
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Table 4-2. 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/hr 3.47 (b)

Gamma at 1 mc 0.0154 mR/hr 1.69 (d)

Estimated gamma at 1 cme 0.0154 mR/hr (d) 6.09%

Estimated beta at 1 cmf 0.238 mrad/hr (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 from the external gamma values. Outdoor external 

background in the Tonawanda area was reported to be 8 to 15 μR/hr (ORNL 1978, p. 15).

d. Not applicable.

e. Estimated as equal to gamma at 1 m.

f. Estimated as 0.253 (beta plus gamma) less 0.0154 (gamma).

surveys in April 1949 before the vacuum cleaning and flushing (Blatz 1949; Wolf 1949b). Typical 

levels of measured gamma radiation at 3 ft on April 19 and at contact on April 22 were similar and 

averaged about 0.18 mR/hr, about 4 times higher than the estimated median contact gamma level 

before decontamination. Typical levels of measured beta at contact on April 22 averaged about 

0.87 mrep/hr, about 1.3 times higher than the estimated median beta level before decontamination. 

The April results were the basis of the estimate in Table 4-1 that beta and gamma radiation levels 

before vacuum cleaning and flushing were 3 times higher than the values that were measured 

afterward.

In the 1976 survey, Building 30 was the most contaminated building on the site (ORNL 1978). The 

levels for Building 30 were assumed to apply to all buildings on the site.

For the preproduction period, it was assumed that there was no significant external beta or gamma 

exposure indoors for Ceramics Plant workers because the buildings were under construction or new. 

Some exposure while outdoors was possible because the transport of ore to Tonawanda Laboratory 

(Building 14) and the work there could have produced some site contamination. Resultant outdoor 

radiation levels are very likely to have been less than floor and wall radiation levels inside Building 14 

because 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 4-1 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 hr/d. Table 4-3 shows estimated 

annual exposure rates.

Table 4-3. Preproduction beta and gamma radiation levels.

Beta (rem/yr)a Gamma (R/yr)a GSDb

1.01E-01 1.97E-02 3

a. Based on Table 4-1 beta and gamma levels at 3 ft before vacuum cleaning and flushing. 

Assumed exposure period: 0.5 hr/wd, 6 wd/wk, 50 wk/yr. The beta and gamma rates are for 

the whole body.

b. Estimated on the basis of the GSD values in Tables 4-1 and 4-2.

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 Production, 1943 to 1946

4.1.2.1 Beta

No beta film badge data were found for 1943 to 1946. There were beta film badge data for 1947 to 

1949, but none of the data pertained to Steps I or II, as these processes were no longer in operation. 

For 1943 to 1946, one report of workplace beta exposure rates was found. Solid samples from 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 1944c) and are reproduced here in the first five columns of Table 4-4. The roentgen is 

technically defined only for photon radiation; beta dose should be expressed in absorbed dose units 

such as rad or rem. However, because 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 

rem.

Table 4-4. 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/yr)d Sample # Sample location

Beta radiation 

(R/8-hr day)b

Allowable time of 

exposure per 

worker (hr/wk)c 

Actual time of 

exposure per 

worker (hr/wk)

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 1944c).

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/d and 3.0 R per 6-day 

week for beta radiation.

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 text. 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/8hr) hr/wk hr/yr rem/yr

Hands in tailings 1.4 10 500 88

Within range of radiation 0.7 40 2,000 175

Total 263

From the fact that no distance values were 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 (see Table A-1).

The fourth and fifth columns of Table 4-4 contain the allowable times of exposure based on a 

3.0-R/wk 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 hr/wk, much greater than the 

allowable exposure time of 18 hours. A MED memorandum (Hadlock 1944) questioned this 

discrepancy. It brought the following MED response (Thomas 1944):
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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....

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 4-4 shows worker doses derived from the data. These range from 2 to 

263 rem/yr.

It is striking that the 1943 to 1946 dose rates in Table 4-4 are much higher than the 1947 to 1949 

rates Section 4.1.4.1 discusses. For 1943 to 1946 Step I process operators, the beta dose rate was 

estimated as 263 rem/yr. For 1947 to 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/yr (approximate value of annual DD + DM; see Table 4-16 later in this section). Despite the high 

estimates, the 1943 to 1946 data does not appear unreasonable. The dose rates in Table 4-4 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 

6-day week (Ferry 1944c) or 150 R/yr. Therefore, the estimates in Table 4-4 are considered a valid 

basis for dose reconstruction.

4.1.2.1.1 Estimation of Worker Doses

Beta doses for specific jobs are estimated below and summarized in Table 4-24 at the end of this 

section.

1943 to 1946 Loaders, Step I Process Operators, Step II Process Operators, Ball Mill Operators, 

and Weighmasters

The calculated dose in Table 4-4 for sample 13 was assumed to apply to a loader. The doses 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 same dose as a 

Step I process operator because the ball mill operators 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 CCCC 1945b). Reductions for shielding by apparel and containers were not taken into account in 

estimating doses because of insufficient information. The dose rates in Table 4-4 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 Hands and Forearms Dose in 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 CCCC 1945b) 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 hr/d at a location 4 ft from a pile of ore bags. The data in Table 4-4 

and Table A-1 indicate that processed materials and byproducts had about 15 times the beta dose 
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rate of ore. Therefore, the weighmaster was assigned the same beta exposure rate as a Step I 

process operator.

The dose rates in Table 4-4 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 prerefined ore (which the domestic ore used at Linde might 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.

To use the data in Table 4-4 to estimate time-averaged beta dose rates applicable to the entire 1943 

to 1946 production period at Linde, the ratio of the average dose rates to those in Table 4-4 was 

estimated. Table 4-5 documents the determination of the ratio. The grade range of the L-30 ore that 

was used at Linde was 8% to 12% (Wallo, Vierzba, and Roberts 1981, Table B-1). To obtain the 

highest ratio, it was assumed that the measurements in Table 4-4 were made on the lowest L-30 ore 

grade, 8%. It was also assumed that beta dose was proportional to electron energy that is released 

per decay and that 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 that would be 

predicted by using the data in Table 4-4. Therefore, the estimated dose rates in Table 4-4 were 

multiplied by 0.84 to obtain time-averaged dose rates for 1943 to 1946 production. The results are in 

Table 4-6, which summarizes the results of all beta dose rate estimates in this section and divides the 

job categories into three groups (high, medium, and low).

1943 to 1946 Workers with Jobs Analogous to 1947 to 1949 Jobs

Workers in 1943 to 1946 who held jobs analogous to ones in 1947 to 1949 were assigned 3 times the 

1947 to 1949 beta dose rates. The factor of 3 increase accounts for potential exposure to radiation 

from waste products from unrefined uranium ore and for the possibility that procedures in 1943 to 

1946 did not involve as much radiological protection. (Tables A-2 and A-3 indicate unrefined uranium 

materials release approximately 2.6 times as much electron energy per uranium decay as refined 

uranium materials. In addition, the plant had a health physicist in 1947 to 1949 but not in 1943 to 

1946.) Workers with jobs analogous to those assigned to the medium beta exposure category for 

1947 to 1949 were assigned dose rates of 17.6 rem/yr to the hands and forearms and 5.85 rem/yr to 

the remainder of the body.

Workers with jobs analogous to those assigned to the low beta exposure category for 1947 to 1949 

were assigned a beta dose rate of 3 rem/yr to the whole body. Table 4-6 shows the assignments.

Table 4-5. 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

Massweighted 

average beta 

dose rate 

relative to 

8% L-30d

06/1943 to 

11/1943

Not 

stated

Domestic 

& scrap 0.447 15.0 0.72 1,000 0.028

12/1943 to 

10/1944 L-30 African 1.157 10.8 1.35 8,504 0.447

10/1944 to 

11/1944 L-50 African 1.157 6.7 0.84 1,486 0.048

12/1944 to 

02/1946 L-19 Domestic 0.447 14.0 0.68 6,102 0.161
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Mass

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

weighted 

average beta 

dose rate 

relative to 

8% L-30d

02/1946 to 

07/1946 R-10 African 1.157 3.5 0.44 8,492 0.145

06/1946 to 

07/1946 Q-20 African 1.157 17.7 2.21 82 0.007

Sum: 25,666 0.84

a. Source: Wallo, Vierzba, and Roberts (1981), Table B-1. MT = metric ton (1,000 kg).

b. Per Table A-2. nt = nuclear transformation.

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 to 1946 ore beta dose to the ore dose due to 

8% L-30 ore.

Table 4-6. Assigned beta dose rates, 1943 to 1946. 

Category/job 

Median dose rate (rem/yr)

Hands and 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 and 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 Section 4.1.2.1.3.

1943 to 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/yr to the hands and forearms and 5.85 rem/yr to the remainder of the 

body.
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4.1.2.1.2 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 4.1.1. Therefore, the outdoor beta rate was estimated as 0.676 mrem/hr (based on beta dose 

rate at 3 ft before vacuum cleaning and flushing in Table 4-1). For 0.5 hr/wd exposure, 6 wd/wk, and 

50 wk/yr, the average worker exposure would have been 0.1 rem/yr. This is negligible in comparison 

with the indoor doses, given the approximate nature of the estimates, and was ignored.

4.1.2.1.3 Categories

Table 4-6 summarizes assigned beta dose rates for the 1943 to 1946 production period. All are 

considered to be median values of a lognormal distribution. The dose rates that were derived from 

the 1943 to 1946 data in Table 4-4 were judged to be based on near-maximum exposure parameters. 

They were assigned a GSD of 1.52, which corresponds to a distribution in which the ratio of 95thpercentile value to median is equal to a factor of 2. The other dose rates were considered to be more 

uncertain. They were assigned a GSD of 2.65, which corresponds to a distribution in which the ratio 

of 95th-percentile value to median is equal to a factor of 5.

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 dose reconstruction, it should be 

obtained as follows:

• For the low and medium categories, the values in Table 4-6 should be used.

• For the ball mill operator, Step I process operator, and weighmaster, the value in Table 4-6 

should be used.

• For the Step II process operator, the dose to the hands and forearms should be taken as 

3 times the remainder-of-the-body dose for the high category or 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 or 74 rem.

4.1.2.2 Gamma

Gamma film badge data are available for Step I process workers for a portion of the 1943 to 1946 

production period. The data covers the period from January 31, 1944, through February 26, 1945, 

except for a 3-week gap from April 18 to May 5, 1944, for which data were not obtained due to a film 

handling error (Ferry 1944d). Badges were usually worn for about a week, but during the first

3 months some of the badging periods were longer (2 to 4 weeks). In the data analysis, each 

multiweek measurement was treated as a set of individual weekly results with each equal to the 

weekly average that was determined from the multiweek measurement. For example, a 4-week result 

of 300 mR was treated as four individual 1-week results of 75 mR.
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About 20 workers were typically 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. Film badges were provided by 

the Medical Section of the MED (presumably the University of Rochester).

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). 

After March 27, 1944, 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/8-hr day and also as 0.5 R/wk and 0.6 R/wk (Ferry 1944e; Tybout 1945c). Linde employees are 

reported to have worked a 6-day 48-hour workweek (Dupree1983a, 1983b). For analysis of the 

gamma data, a 6-day workweek was assumed. To obtain roentgen per week, results in roentgen per 

8-hour day 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/wk. In the analysis of the data, 44 mR/wk was taken as the LOD, and 

results reported as “<X” or zero were considered to be <LOD results. However, all nonzero results, 

even if below 44 mR/wk, 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 (Wallo, Vierzba, and Roberts 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 

2 weeks of November with the low point 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 from January 31 to November 12, 1944 (41 weeks) and November 13, 1944 to 

February 26, 1945 (15 weeks).

The data identified each worker's job activity. The job activities were grouped into categories. Each 

category consisted of jobs that were judged to have similar exposure potential based on job 

descriptions and the data. Table 4-7 lists the job activities and the category to which each was 

assigned.

African Ore

Table 4-8 displays statistical characteristics of the film badge data for the period from January 31, 

1944 to November 12, 1944, 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 LOD,

• fM = missed fraction = the fraction of results in a year below the LOD,

• mD = GM of all results above the LOD, and

• The GSD of all results above the LOD based on the assumption that their distribution was 

lognormal.

Table 4-7. Job categories for Step I gamma dose analysis, 1943 to 1946. 

Job activitya Category

Ball mill Ball mill operator

Ball mill operator Ball mill operator
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Job activitya Category

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 4-8. Step I gamma exposure data for African ore processing, January 31 to November 12, 

1944. 

Category

# weekly

results

Dosimeter

fraction fD

Missed

fraction fM

For results in dosimeter fraction

GM mD (mR/wk) 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 subscripts D and M denote dosimeter dose and missed dose, respectively, as defined by NIOSH 

(2007a, pp. 30 and 32). For the current set of data, the value of fD was estimated as the number of 

≥LOD results divided by the number of measurements, and fM was estimated as the number of <LOD 

results 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.)

The estimated median annual dosimeter dose DD for a 50-week year is shown in Table 4-9. This was 

calculated as follows:

= 50 D fm D DD (4-1)
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Table 4-9. Step I gamma exposure rate parameters for African ore processing, 

January 31 to November 12, 1944. 

Category

DDa

(R/yr) GSDb 

DMa

(R/yr) 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 4-8 is assumed to apply to all categories.

For simplicity and because a GSD value based on small number of measurements might not be valid, 

the highest GSD in Table 4-8 was assigned to every job category.

In accordance with NIOSH (2007a), 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:

= 50 ( / 2) D f LOD M M (4-2)

= 95 50 ( ) D f LOD M M (4-3)

A GSD of 1.52 was calculated from DM and DM95. The results for DM are in Table 4-9.

Domestic Ore

The analysis of exposure data for the period November 13, 1944, to February 26, 1945, 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 4-10 and Table 4-11. 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 4-10. Step I gamma exposure data for cleanup and domestic ore processing, November 13, 

1944, to February 26, 1945. 

Category

# Weekly

results

Dosimeter

fraction fD

Missed

fraction fM

For results in dosimeter fraction

GM mD (mR/wk) GSD

Ball mill operator 0 (a) (a) (a) (a)

Cleanupb 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 because 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 because only one result was ≥LOD.

d. Not available because 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.
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Table 4-11. Step I gamma exposure rate parameters for cleanup and domestic 

ore processing, November 13, 1944, to February 26, 1945. 

Category

DDa

(R/yr) GSD

DMa

(R/yr) GSD DD + DM

Relative

DD + DMb

Ball mill operatorc 0.32 1.27 0.90 1.52 1.22 28%

Cleanupd 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.

4.1.2.2.1 Estimation of Worker Doses

African Ore

Step I Process Workers

The annual gamma exposures of the monitored Step I workers (Table 4-9) 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/yr, the highest value of DD + DM in Table 4-9, and a GSD of 

2.61, the highest GSD in Table 4-9.

Other Workers

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.

One source of gamma exposure, however, was more dependent on worker location than on worker 

task. This was radiation from large quantities of ore that were stored in ore piles. A report of radiation 

surveys on March 2 and 3, 1944 (Ferry 1944f) 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- by 30- by 12-ft pile of 8% ore at 

an unspecified location (percentages specify weight % of U3O8). Based on an analysis of typical 

pitchblende ore at Linde, the ore density was estimated as 3.25 g/cm3 (Wallo, Vierzba, and Roberts 

1981, Table C-1). Therefore, the 30- by 30- by 12-ft pile contained approximately 2 million lb 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 2-3). 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 1944a), and a letter from the Ceramics Plant on May 8, 

1944, indicated compliance with this recommendation (Rehm 1944b).
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In Ferry (1944f), an exposure rate of 0.23 R/8-hr day was reported at 1.5 ft from the 30- by 30- by 

12-ft 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 from Linde pilot plant studies for a 4- by 5- by 20-ft pile of 8% ore (Skinner 1944; 

Wiesendanger 1944). The fit was obtained using the MicroShield computer program (MicroShield 

Team 2003), which is based on the point kernel methodology. The values from the program were 

multiplied by a single scaling factor that was chosen to provide a best fit to the measured data. The 

adjustment compensated for simplifications in the modeling. Figure 4-1 shows the data and the fit. 

Gamma exposure rates measured during pilot plant studies at Linde Ceramics (Skinner 1944) are 

shown by the diamond symbols. The rectangles show the fit to the data with distance assumed to be 

measured along the perpendicular bisector of a 4- by 20-ft. face. The ore grade (8%) was assumed to 

be that determined in the analyses reported by Wiesendanger (1944, p. 18).The data covers the 

range of 0 to 10 ft from the 4- by 20-ft 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 

30- by 30- by 12-ft pile by more MicroShield calculations that were scaled with the same factor.

Figure 4-1. Fit to gamma exposure data for a 4- by 5- by 20-ft pile 

of 8% African pitchblende ore. 
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Figure 4-2 shows the MicroShield results and an analytic function fit to them. 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 12- by 30-ft face. The values were 

calculated using MicroShield (MicroShield Team 2003). The line is a plot of the analytic function

obtained by a least-squares fit to the MicroShield results. The function is:

× × × + −× [ ] 2 (1/ ) exp exp(- ) ( ) r a br cdr (4-4)

where a = 1.906, b = 3.126×10-02, c = 7.660, and d = 1.521×10-03. The fit agreed with the calculated 

values to within ±3.0% over the range of 16 to 2,000 ft from the center of the pile.

Figure 4-2. Calculated exposure rates and analytic fit for a 30- by 30- 

by 12-ft pile of 8% African pitchblende ore. 

The function facilitates calculation of 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, which was 1.5 ft from the face. The result was 0.19 R/8 hr, which is 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.

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
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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. In addition, the calculations are based on distance along the perpendicular bisector 

of a pile face. Exposure rate at any distance is lower for a point not on the perpendicular bisector.

To estimate the impact on the other workers in Building 30, who did not wear dosimeters, exposure 

rates due to a 30- by 30- by 12-ft pile 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 4-12. They range from 0.28 R/yr (for the 

laboratory) to 4.9 R/yr (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.

Table 4-12. Estimated maximum exposure rates in selected Building 30 

facilities due to 30-by 30- by 12-ft pile of 8% ore in southeast corner. 

Facility Distance (ft)a Exposure rate (R/yr)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 2,550 hr/yr exposure.

Although few of the other workers spent the major portion of their work time near the southeast corner 

of Building 30, some might have worked near other ore piles, near tailings piles, or near large 

quantities of ore in process. Because the locations and quantities of these other sources are not 

known in sufficient detail, any worker who spent a substantial amount of time in Building 30 had a 

potential for significant gamma exposure. Such workers included not only the production personnel 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 lunchroom, and office 

personnel who might have had to regularly visit the building (e.g., product accountants, industrial 

relations representatives, the fire marshal, and engineers). Last, 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 examinations states that it would not be appropriate to 

reduce the examination 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 1944a). In view of the likely difficulty of 

establishing that a particular worker did not spend a substantial portion of 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 were assigned a gamma dose of 5.35 R/yr for the whole body, based 

on the results for the most exposed group of monitored process workers (Table 4-9).

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 at Linde (Section 2.2). The value of DD + DM for Step I process operators 

during domestic ore processing (Table 4-11) is 42% of the value during African ore processing (Table 

4-9). 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 to 1946 

production period, the gamma exposure rate of 5.35 R/yr was assumed to apply for the whole period.
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4.1.2.2.2 Outdoor Exposure Rate

The average outdoor gamma exposure rate to which workers could 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 4.1.1. Therefore, the outdoor gamma rate was estimated as 0.131 mR/hr based on the 

level before vacuuming and flushing in Table 4-1. For 0.5 hr/wd exposure, 6 wd/wk, and 50 wk/yr, the 

average worker exposure would have been 0.020 R/yr. This is negligible in comparison with the 

indoor exposures given the approximate nature of the estimates and was ignored.

4.1.3 Standby, 1946 to 1947

Little information is available about the status of activities during the standby period. It is likely that 

the onsite staff 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 entire standby period, each worker's allocation of time was assumed to have been as 

indicated by the occupancy factors in Table 4-13.

Table 4-13. Ceramics Plant beta and gamma radiation rates during standby.

Parameter

Category Time-weighted radiation ratea

Office Production Outdoors Beta (rem/yr) Gamma (R/yr)

Beta (mrem/hr) 0.000 0.676 0.676 (b) (b)

Gamma (mR/hr) 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 hr/d exposure, 6 d/wk, 50 wk/yr. 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.

Gamma radiation levels were measured at six locations in Building 30 on October 22, 1946. 

Measurements were made at 1 in. from the surface of interest. The results were reported as 0 R/8 hr 

for four of the locations and 0.005 R/8 hr (0.625 mR/hr) for the other two locations (each near an ore 

dumping grill) (Howland 1946). Because 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 4-1 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 preproduction period. The gamma and beta radiation rates in an office 

building were assumed to be zero.

Table 4-13 summarizes the calculation of annual radiation rates based on the above parameters.

Because 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.

4.1.4 Rehabilitation and Production, 1947 to 1949 

4.1.4.1 Beta

For part of the period from 1947 to 1949, weekly film badge measurements of beta and gamma 

exposure were available (CCCC undated b). The data had been “de-identified” by removal of 
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personnel names (Wallace 2003), but many of the records were labeled with a job title. The data file 

contained about 6,000 records from January 7, 1948 through December 12, 1949. 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 before March 15, 1948, badges 

were processed by the University of Rochester (Heatherton 1948g, Osinski 1948a); thereafter, they 

were processed by the AEC New York Directed Operations Office 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 labels to indicate 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 1949d, 1949e). 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 

March 15, 1948, a beta value below the limit of detection was reported as “<80.” After March 15, 

1948, 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 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 that were specified in the 

records were binned into categories that combined jobs that were judged to have had similar 

exposure potential. Table 4-14 displays the job titles in each category.

Table 4-14. Job categories for 1947 to 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
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Job title Category Job title Category

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 4-15 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 to 1946 gamma dosimetry data (Section 4.1.2.2). 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, which is 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.

Table 4-15. Step III beta dose data, 1947 to 1949. 

Job # Badges

Dosimeter 

fraction fD

Missed 

fraction fM

For results ≥LOD

GM mD (mrem/wk) 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.

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 to 1946 gamma data 

(Equations 4-1, 4-2, and 4-3 in Section 4.1.2.2). Because badges would not have been worn until 

workers arrived indoors, an outdoor beta dose rate of 0.10 rem/yr (from Section 4.1.2.1.2) 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 4-16 displays the results.

For most jobs, the largest contribution to DD + DM + OD was missed dose DM. The production workers 

in 1947 to 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, lunchroom, nurses’ 

office, and chemistry laboratory; 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 
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was estimated (Section 4.1.1) as 0.676 mrem/hr (value before vacuuming and flushing in Table 4-1), 

which is equivalent to 34.5 mrem/wk and 1.7 rem/yr for a person spending 8.5 hr/d, 6 d/wk, and 50 

wk/yr in the building. This is comparable to the film badge LOD for beta radiation (35 mrem/wk) and 

so might have been a substantial portion of the missed dose. Based on this estimate of 1.7 rem/yr of 

possible dose and the lowest film badge DD + DM value in Table 4-16 of approximately 1 rem/yr for 

workers, who were regularly badged, it was assumed that all workers, even office workers, received a 

beta dose of at least 1 rem/yr. Office workers were included because it is unknown which office 

workers spent substantial portions of their time in production buildings. 

Table 4-16. Beta dose rates for Step III, 1947 to 1949. 

Job

DDa DM + ODa DD + DM +OD

(rem/yr) GSD (rem/yr) b (rem/yr) 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 4-15 was assumed for all categories for which a GSD can be assigned.

c. No result ≥LOD.

Some workers who might have had offices in the office building might also have spent much of their 

workday in the production area (e.g., engineers, health physicist, and fire marshal), and some clerical 

workers might have actually had work locations in the production area [discussion near the end of 

Section 4.1.3.2.1 and Holmes (1944a)].

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/yr. 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 to 1946 for 

truck drivers to work as loaders when not driving a truck (CCCC 1945b, p. 10); it was assumed that 

the same practice applied from 1947 to 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/yr. 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.

Rehabilitation

Beta dose rates during rehabilitation were assumed to be the same as during production to account 

for exposures while cleaning contaminated facilities, setting up, and testing equipment.
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Hands and Forearms Dose

Film badges were worn on the chest (Heatherton 1948h). 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 dependent on the operation and the area of the body 

(Heatherton 1948h). The results are shown in Table A-6. Based on these data, it was estimated for 

this site profile that the average dose to the hands and forearms of a worker handling radioactive 

material was 3 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 studied 

operations with the result rounded up to the nearest integer. No credit was taken for wearing gloves. 

For dose reconstruction, a hands and forearms dose of 5.85 rem/yr was assumed for the medium 

category. Workers in the low category were assumed to have a hands and forearms dose of 1 rem/yr, 

which is 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 to 1949 production were available (CCCC 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. Results reported as “0,” “*,” or “**” were interpreted as meaning less than 

the LOD. For records before March 15, 1948, a gamma value below the LOD was reported as “0” or 

“<50.” For records after March 15, 1948, the limit of gamma detection was not explicitly stated but 

was deduced to be 45 mR from an examination of the data. A few scattered reports of nonzero 

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 that for the analysis of the beta data (Section 4.1.4.1). Table 4-17 summarizes the statistical 

characteristics of the gamma data.

Table 4-17. Step III gamma exposure data, 1947 to 1949. 

Category # Badges

Dosimeter

fraction fD a 

Missed

fraction fM b 

For results ≥LOD

GM mD (mR/wk) 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.

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 to 1949 beta data 

(Section 4.1.4.1). Because badges would not have been worn until workers arrived indoors, an 

outdoor gamma exposure rate of 0.02 R/yr (Section 4.1.2.2.2) was added to the DM values. Because 
 

 

 

 

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the outdoor contribution was not the major portion of any of the sums, the GSD was taken as not 

changed by the addition. Table 4-18 displays the gamma exposure rates that were determined. 

As in the case of 1947 to 1949 beta dose (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/hr (value before 

vacuuming and flushing in Table 4-1), which is equivalent to 6.7 mR/wk and 0.34 R/yr for a person 

who spent 8.5 hr/d, 6 d/wk, and 50 wk/yr in the building. 

Table 4-18. Gamma exposure rates for Step III, 1947 to 1949. 

Job 

D a D D a M + OD DD + DM + OD 

(R/yr) GSD (R/yr) b (R/yr) 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 4-17 was assumed for all categories for which a GSD can be assigned. 

c. No results ≥LOD. 

This is below the film badge LOD for gamma radiation (45 mR/wk). Any clerical personnel who 

worked in production buildings (see Holmes 1944a) were assumed to have received this estimated 

floor and wall radiation plus the 0.14 R/yr that was received by the occasionally badged office worker 

for a total of 0.48 R/yr. Because it might not be easy to identify which clerical personnel worked in a 

production building, all unbadged or infrequently badged personnel were assumed to have received 

0.48 R/yr. 

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 to 1949 beta category assignment 

(Categories discussion in Section 4.1.4.1). Jobs in the medium gamma category were assigned an 

exposure rate of 1.61 R/yr for the whole body. Jobs in the low category were assigned an exposure 

rate of 0.48 R/yr 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 while cleaning contaminated facilities, setting up, and testing 

equipment. 

4.1.5 Cleanup, 1949 to December 31, 1953 

Cleanup of the Ceramics Plant is discussed in Section 2.6. Floor and wall radiation levels, which 

were measured at the start and end of the 1949 to 1950 decontamination of Building 30, were 

available and are summarized in Table 4-1. In addition, results were available from film badge 

measurements of beta and gamma radiation during the 1948 removal of equipment from Building 30. 
 

 

 

 

 

 

 

 

 

 

 

	 

	 

	 

	 

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These results are summarized in Table 4-19. Because the results in Table 4-19 are film badge 

measurements, they include floor and wall radiation in addition to radiation from contaminated 

equipment. 

Table 4-19. Beta and gamma radiation measured during equipment 

removal, 1948. 

Beta dose rate 

DD (rem/yr)a GSD DM (rem/yr)a GSD 

1.99 1.53 0.52 1.52 

Gamma exposure rate 

DD (R/yr)a GSD DM (R/yr)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 (2007a), pp. 30 and 32. 

For external dose reconstruction, each worker is classified in one of the following categories: 

•	 Cleanup worker, 

•	 Cleanup support worker, and 

•	 Non-cleanup worker. 

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 sandblaster 

operators) and laborers who were involved in moving and cleaning activities. A cleanup support 

worker is one who spent a substantial portion of the workday 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 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: 

(4-5) 

where DD and DM were taken from Table 4-1

D + OD	 D + DM 

9. The parameter OD is the outdoor dose rate. As for 

the preproduction and standby periods, it was estimated as equal to the estimated indoor floor and 

wall level at 3 ft before vacuum cleaning and flushing in Table 4-1. 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( D 	 (4-6) D + D ) + OD	 M 

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 workers after 1946 (numbers were not adjusted for exposure time). 

The radiation level was estimated as: 

(4-7) 0.05(DD + DM) + OD	 	 

for gamma radiation. The beta radiation values specified are for all parts of the body. Table 4-20 

presents the results. 
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The beta dose rates in Table 4-20 are based primarily on film badge measurements. The film badges 

were worn on the chest (Heatherton 1948h). 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, and scrubbing contaminated surfaces). For dose reconstruction, the beta dose 

rate to the hands and forearms of a cleanup worker was taken as 3 times that to the remainder of the 

body. The factor three was based on measurements discussed in Hands and Forearms Dose in 

Section 4.1.4.1.

Table 4-20. External dose reconstruction parameters for cleanup. 

Period/category

Betaa Gammab

Median (rem/yr)c GSDd Median (R/yr)c GSDd

07/01/1949–12/31/1950, 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

01/01/1951–12/31/1953, 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 3 times higher. See text.

b. Gamma exposure rates are for the whole body.

c. Annual rates based on 8.5 hr/wd indoors, 0.5 hr/wd outdoors, 50 wk/yr, 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 95thpercentile value is 10 times the median.

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 might occasionally have visited the research 

facilities. It includes secretaries, clerks, draftspersons, and high-level managers. Three periods 

(R&D, cleanup, and postcleanup) 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 later measurements 

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 4-21 shows the assumed 

annual radiation rates and their bases.

Table 4-21. Tonawanda Laboratory beta and gamma radiation, 1942 to 1953.

Period

Beta

rem/yr Gamma

Start End Job

Hands &

forearms

Remainder

of body GSD Basisa R/yrb GSD Basisa

R&D

10/01/1942 07/31/1946 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

08/01/1946 12/31/1946 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
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Period 

Beta 

rem/yr Gamma 

Hands & Remainder 

Start End Job forearms of body GSD Basisa R/yrb GSD Basisa 

Postcleanup 

01/01/1947 12/31/1953 All 3.26E-01c 3.26E-01c (d) E 6.80E-02 (d) E 

a.	 Basis: 

1.	 One-half of value for most exposed Ceramics Plant process worker during 1943 to 1946 production (Step I process 

operator). The factor of 2 reduction allows for the lower intensity of research work in comparison with production. 

For example, 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. 

2.	 Based on Ceramics Plant office worker during 1943 to 1946 production. 

3.	 Cleanup worker during Ceramics Plant decontamination. 

4.	 Non-cleanup worker during Ceramics Plant decontamination. 

5.	 Building 30 postcontamination level (Table 4-1) for 8.5 hr/wd plus outdoor level on Tonawanda site for 0.5 hr/wd. 

Outdoor level assumed equal to Building 30 level before vacuum cleaning and flushing (Table 4-1). A 6-day week 

was assumed. 

6.	 Based on most exposed Ceramics Plant process worker during 1943 to 1946 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 in the compounds that 

were processed at Linde are presented in Table 4-22. 

The following assumptions were made about the conversion from neutron production rate from the 

alpha-neutron reaction to an annual dose at the receptor point: 

1.	 Point source geometry was used to produce a nominal ambient neutron dose rate. 

2.	 Point source strength was based on the daily rate of U3O8 handling or UF4 production. Selfshielding within the source-target compound matrix was assumed to be negligible. 

3.	 A 238U:235U composition of 99.3% to 0.711% by weight for natural uranium was assumed. 

4.	 	 The entire mass of uranium compounds was conservatively assumed to be attributable to the 

uranium (source) content. 

5.	 	 Average neutron energy from alpha-neutron reactions was 2 MeV. 

6.	 	 Dose equivalent rate-to-fluence rate conversion factor for 2 MeV was 1.3 × 10-4 rem/hr per 

neutron/cm2

-s. 

7.	 	 The work year consisted of 8 hr/d, 6 d/wk, and 50 wk/yr. 
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Table 4-22. Natural uranium per gram dose rates at 1 ft.

Chemical form

Dose ratea

(rem/hr-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

a. ORAUT-OTIB-0024 (ORAUT 2005) data used to calculate the alpha-neutron dose rates is from Shleien, Slaback, and 

Birky (1998), Salmon and Hermann (1992), Reilly et al. (1991), and DOE (2000).

During the Step I and Step II conversion processes in Building 30 (Ceramics Plant) the bounding 

neutron dose rate would be attributable to operations that involved handling of African Ore with alphaemitting progeny in secular equilibrium through 226Ra. This arises from the fact that the African ore 

feed material contained alpha-emitting progeny that contributed to the alpha-neutron production rate. 

As indicated in Table 4-22, the African ore form produced 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 after 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. As a consequence, the dose that 

was received by loaders (or “Movemen”) would be bounding for all personnel who were engaged in 

Step I and Step II processing activities. As indicated in the Table 4-22, during the Step III conversion 

process in Building 38, the neutron dose rate from UF4 is almost 85 times greater than from the UO2

feed material. Therefore, 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 MT (26 million kilograms) of ore over a period of 37 months beginning in June 1943 and 

ending in July 1946. Assuming a 6-day workweek for 50 wk/yr, and a uniform rate of ore processing, 

28,111 kg/wd (5,622 kg of uranium) were handled at Linde. The assumption favorable to claimants 

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 assumption favorable 

to claimants that all of the ore was 20% U3O8 by weight has also been made although the Linde site 

literature indicates that the ore contained 3% to 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 1949a) or 3,147 kg/d.

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 one-tenth of the daily production amount of U3O8 or UF4 at a distance of 

1 ft. This quantity of material is an estimate that is favorable to claimants 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 one-third 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 who worked 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 ft away 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:

• R&D for Steps I, II and III. This work occurred primarily from October 1942 to December 1943 

(Jenness and Ewing 1943). The largest-scale efforts were pilot plant projects. The Step I pilot 
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plant produced 2 to 3 t of U3O8 per week in November 1942 (Bonsib 1942). The Step III 

equipment could convert up to 90 lb of UO2 to UF4 in a single run (CCCC 1946c, p. 84).

• Conversion of UF6 to UO3. This work appears to have been done after November 1943 

(because it is not mentioned in the progress report for work from October 1942 to November 

1943; Jenness and Ewing 1943). Interim reports were issued from July 1944 through July 

1945, and a final report was issued in May 1946 (CCCC 1946d, 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 (CCCC 1946c, p. 89). The work involved processing 

2,750 lb 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 t of U3O8 was processed per week, which corresponds to a daily processing 

rate of 454 kg from October 1, 1942, through February 29, 1944.

• Exposures were to 100 lb of UF4 or UF6 per week daily from October 1, 1942, to July 31, 1946. 

This accounts for the Step III and UF6 to UO3 work.

• Exposures were to 3,000 lb of UF4 for 2 months at some point between October 1, 1942, and 

July 31, 1946. This accounts for the UF4 grinding research. The neutron dose in a year is 

equivalent to that which would be produced by continuous exposure to 500 lb of UF4 for a 

year. Because the dates of the work are uncertain, it was assumed that there was continuous 

exposure to 500 lb of UF4 from October 1, 1942, to July 31, 1946.

To reflect all types of UF4 and UF6 work, a daily source term of 600 lb (272 kg) was assumed. All

source terms (both oxide and fluoride) were divided by 3 to account for the fact that no worker would 

be within close range of the source material at all times. This yielded 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 were assumed to be exposed to these source amounts at a distance of 1 ft. 

Table 4-23 summarizes the model parameters and presents calculated dose rates.
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Table 4-23. Neutron dose rates and annual doses from the alpha-neutron sources.

Daily Dose rate 

Alpha Target source at 1 foot Dose rate 

Location/time period Process source atom term (g) (rem/hr) (rem/yr)

Building 30 Steps I and II African ore Oxygen 5.62E+6 1.15E-5 2.77E-2

06/01/1943–07/31/1946 20% U by 

weight

Building 38 Step III UF4 Fluorine 2.76E+6 2.08E-4 5.0E-1

07/25/1943–08/31/1946a

09/15/1947–09/30/1949a

Tonawanda Laboratory Steps I and II U3O8 Oxygen 1.51E+5 3.10E-6 7.44E-3

(Building 14)

10/01/1942–02/29/1944

Tonawanda Laboratory Step III; UF4 or UF6 Fluorine 9.07E+4 6.01E-5 1.44E-1

(Building 14) UF4 grinding;

10/01/1942–07/31/1946 UF6 process 

gas to UO3

a. The period of exposure was extended beyond the end of the production period to account for inventory left on hand. 

UF4 that was 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 1949b) and data on production at Electromet 

(DOE undated, pp. 23–24).

4.4 EXTERNAL DOSE RECONSTRUCTION SUMMARY, OCTOBER 1, 1942, TO DECEMBER 

31, 1953

This section summarizes guidelines and parameters for reconstruction of doses due to external 

radiation during October 1, 1942, to December 31, 1953 except for doses due to occupational medical 

exposure, which are addressed in Section 5.0. The parameters that are provided in this section are 

for use when individual worker data are unavailable or inadequate. Because of the SEC 

determinations (Leavitt 2005, Sebelius 2012) that it is infeasible to adequately reconstruct internal 

dose for the period from October 1, 1942, through December 31, 1953, dose estimates for this period 

are considered partial dose estimates.

For dose reconstruction, exposures for different jobs have been grouped into categories. Table 4-24 

specifies the categories and the annual radiation doses or exposures assigned to each. For the 

production periods at the Ceramics Plant, categories were designated high, medium, and low. In 

other cases, designations more descriptive of the work activity are used.

Table 4-24. Summary of annual external exposure from AWE operations, 1942 to 1953.

Yeara Work category

Beta (rem)a

Gamma (R)a Neutron (rem) Hands & forearms Rest of body a,b

Ceramics Plant (Buildings 30, 31, 37, and 38)

1942c All workers 2.55E–02 2.55E–02 4.97E–03 (d)

1943e

High 1.51E+02f5.05E+01

3.65E+00 3.41E–01g 5.05E+01h

Medium 1.20E+01 3.97E+00

Low 2.08E+00 2.08E+00

1944

1945 High 2.21E+02f

7.40E+01

5.35E+00 5.00E–01g 7.40E+01h

Medium 1.76E+01 5.85E+00

Low 3.00E+00 3.00E+00

1946i

High 1.28E+02f4.32E+01

3.11E+00 3.33E–01g 4.32E+01h

Medium 1.04E+01 3.59E+00

Low 1.93E+00 1.93E+00

1947j Medium 2.04E+00 8.91E–01 5.37E–01 1.48E–01g
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Yeara Work category

Beta (rem)a

Gamma (R)a Neutron (rem) Hands & forearms Rest of body a,b

Low 6.10E–01 6.10E–01 2.03E–01

1948 Medium 5.85E+00 1.95E+00 1.61E+00 5.00E–01g

Low 1.00E+00 1.00E+00 4.80E–01

1949 Medium/lowk

6.85E+00 2.28E+00 1.73E+00 3.74E–01g

Cleanup (d) Non-cleanupl 6.60E–01 6.60E–01 2.94E–01

1950

1951

1952

1953

Cleanupm 7.83E+00 2.61E+00 1.85E+00

(d) Non-cleanupn 3.26E–01 3.26E–01 1.11E–01

Tonawanda Laboratory (Building 14)

1942c 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

1946 Researcho 6.78E+01 2.26E+01 3.88E+00 8.36E–02p

Officeq 1.78E+00 1.78E+00 3.15E+00

1947

1948

1949

1950

1951

1952

1953

All workers 3.26E–01 3.26E–01 6.80E–02 (d) 

a. Total annual exposure (dose) for the designated year. Prorated based on calendar year and applicable notations below.

b. 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.

c. Exposure for the period from October 1 through December 31, 1942 only.

d. Neutron dose rate was negligible.

e. Values prorated: For January 1 through April 26, 1943 (preproduction period), applicable values from Table 4-1 applied; 

for April 27 to December 31, 1943, applicable 1944 to 1945 values applied.

(Example calculation: 1943 high-ball mill operator = 0.315 × 1.01E-01 + 0.685 × 2.21E+02 = 1.51E+02).

f. Based on 221 rem/yr for ball mill operator, Step I and Step II process operators, and weighmaster.

g. The Building 38 neutron dose rate for Step III processing was assumed to apply from April 27, 1943, to August 31, 1946, 

and from September 15, 1946, to September 30, 1949. The neutron dose rate was negligible from September 1, 1946, 

to September 14, 1949 (standby), and after September 30, 1949 (cleanup and postcleanup). The period of neutron 

exposure extended beyond the end of production in 1946 and 1949 due to remaining inventory of UF4. 

h. Based on 74 rem/yr for loader per Section 4.1.2.1.3.

i. Values prorated: For January 1 to July 31, 1946, applicable 1944 to 1945 values applied; for August 1 to December 31, 

1946 (standby period), applicable values from Table 4-4 (guard) applied.

j. Values prorated: For January 1 to September 14, 1947 (standby period), applicable values from Table 4-4 (guard) 

applied; for September 15 to December 31, 1947, applicable 1948 values applied.

k. Values prorated: For January 1 to June 30, 1949 (Step III production), applicable 1948 medium values applied; for 

July 1 to December 31, 1949, 1950 to 1953 cleanup values applied.

l. Values prorated: For January 1 to June 30, 1949 (Step III production), applicable 1948 low values applied; for July 1 to 

December 31, 1949, 1950 to 1953 non-cleanup values applied.

m. All cleanup workers and cleanup support workers as defined in Section 4.1.5 are assigned to the cleanup exposure 

category. Parameters are those of the cleanup worker for a 6-day week in Table 4-20.

n. All non-cleanup workers as defined in Section 4.1.5 are assigned to the non-cleanup exposure category. Parameters 

are those of the non-cleanup worker for a 6-day week in Table 4-20.

o. Values prorated: For January 1 to July 31, 1946, applicable 1944 to 1945 values applied; for August 1 to December 31, 

1946, applicable values from Table 4-21 (cleanup-R&D scenario) applied.

p. Includes neutron exposures through July 31, 1946.

q. Values prorated: For January 1 to July 31, 1946, applicable 1944 to 1945 values applied; for August 1 to December 31, 

1946, applicable values from Table 4-21 (cleanup-office scenario) applied.
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Where the dose or exposure estimates were based on dosimetry parameters (dosimeter dose DD and 

missed dose DM) plus possible outdoor dose OD, the dose was estimated as the sum of the 

parameters (DD +DM +OD). Each estimate is considered to be the median of a lognormal distribution. 

A GSD of 3 was assigned to all beta and gamma dose estimates based on the typical GSD levels 

estimated for the underlying data; that is, beta and gamma doses were assigned as lognormal 

distributions with GSDs of 3 and an acute exposure rate. Neutron doses were assigned as a constant 

distribution and a chronic exposure rate.

Prorating of exposures over less than a year is not straightforward for all periods. If doses are to be 

prorated, the footnotes to Table 4-24 should be checked to determine how the dose should be 

distributed throughout the year. Beta doses were assigned as electrons with energy >15 keV. 

Gamma doses were assigned as photons with energy from 30 to 250 keV, which is favorable to 

claimants. Neutrons are assigned as neutrons of 0.1- to 2-MeV energy.

Table 4-25 shows how Ceramics Plant production jobs were assigned to the various categories. For 

1943 to 1946, the assignments were based on Table 4-25 and the discussion in Section 4.1.2.1.3. 

For 1947 to 1949, the assignments were based on Tables 4-16 and 4-18 and the discussion in 

Section 4.1.4.2. In each case, some job titles that reflect subcategories of titles in the earlier tables 

were added.

For the Ceramics Plant cleanup period, the cleanup category includes both cleanup workers and 

cleanup support workers as defined in Section 4.1.5. For Tonawanda Laboratory, the research and 

office categories 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.

Table 4-25. 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 marshal 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
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1943–1946 production 1947–1949 production

Job

Beta

category Job

Beta and gamma

categorya

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.

5.0 OCCUPATIONAL MEDICAL EXPOSURE

NIOSH is required to account for dose from medical X-rays performed on an EEOICPA-covered site 

(either the covered site where the AWE work was being performed or a covered site where medical 

X-rays were performed as a service). Linde X-rays were performed off-site at a non-covered 

commercial facility (Holmes 1944b; Van Horn 1943c; Author unknown 1948, 1950). Therefore, the

dose from medical X-rays is not accounted for in the overall estimated dose calculation 

(ORAUT 2011). 

6.0 ESTIMATION OF EXPOSURES FROM RESIDUAL CONTAMINATION AFTER 1953

This section develops parameters for reconstruction of doses due to internal and external exposures 

at the Ceramics Plant starting January 1, 1954.

NIOSH has determined, with concurrence from the Secretary of DHHS, that internal doses at the 

Linde Ceramic Plant cannot be reconstructed with sufficient accuracy from the beginning of 1954 

through the end of 1969 (Sebelius 2011). If monitoring data are available for workers who are 

included in the SEC class, dose is to be assigned as appropriate based on such data. However, such 

dose reconstructions are still considered partial dose reconstructions because NIOSH has determined 

that internal exposures during the SEC class period (1954 through 1969) cannot be bounded.

6.1 INTERNAL EXPOSURE

There are no radiological survey data available for the period during which Building 30 underwent 

remodeling. However, air samples were collected during decontamination of Building 30 

(Heatherton 1950) and Building 31 (AEC 1954). Based on the nature of the likely activities during 

remodeling, the maximum air concentration from removing concrete with a pneumatic hammer 

(surfaces previously cleaned by sand blasting) was determined to be most representative. The 

maximum air concentration that was reported in the 1950 NYOO report (Heatherton 1950) for 

removing concrete with a pneumatic hammer (surfaces previously cleaned by sand blasting) was 

2.3 MAC (where 1 MAC = 70 dpm/m3

) or 161 dpm/m3. 

Comprehensive surveys of the Tonawanda site buildings were made in 1976 (ORNL 1978) and 1981 

(BNI 1982). Building 30 was found to be the most contaminated building on the site. The indoor 

airborne uranium concentration was measured as 1.9 × 10-2 pCi U/m3 (ORNL 1978). This air 

concentration was used to evaluate personnel internal exposures to uranium for the residual period.

Therefore, an exponential interpolation was made between the uranium air concentrations that were 

assumed for the remediation period (161 dpm/m3

) and the levels that were measured in the 1976 

survey (4.22 × 10-2 dpm/m3

). After 1976, the 1981 survey indicates a more consistent level of 
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contamination. Therefore, the air concentrations after 1976 are assumed to be constant at the 1976 

levels.

During this period, refined uranium materials were handled. None of the uranium progeny would have 

been present in significant quantities in the refined uranium materials but, to account for uranium 

progeny potentially present from past activities and resuspended during decontamination and 

decommissioning (D&D) activities, data from the postoperations period was reviewed to determine 

bounding activity ratios (Attachment C). Table 6-1 presents bounding indoor uranium progeny ratios 

for use for dose reconstruction for the period. The values in this table were the highest observed 

values from the indoor and storm sewer sampling locations.

Table 6-1. Progeny to uranium ratios.

Progeny/U (total) Ratio to uranium

Th-230/U 0.26

Ra-226/U 0.21

Po-210/Ua 0.21

Ac-227/U 0.29

Pa-231/U 0.01

a. Po-210 activity not reported, assumed to be the same as parent 

(Ra-226)

The inhalation and ingestion rates are provided in Tables 6-2 and 6-3. The ingestion intake rates in 

Table 6-3 were estimated using the steps in Section 3.7.

Table 6-2. Uranium and progeny inhalation intake rates (dpm/d) for January 1, 1970,

through 2009.

Years U-234a Th-230a Ra-226a Po-210a Pa-231a Ac-227a

1970 1059 275 222 222 10.6 307

1971 268 69.7 56.3 56.3 2.68 77.7

1972 67.8 17.6 14.2 14.2 0.678 19.7

1973 17.1 4.45 3.59 3.59 0.171 4.96

1974 4.33 1.13 0.909 0.909 0.043 1.26

1975 1.10 0.285 0.230 0.230 0.011 0.318

1976–2009 0.277 0.072 0.058 0.058 0.003 0.080

a. Values listed are assigned as a constant distribution.

Table 6-3. Uranium and progeny ingestion intake rates (dpm/d) for January 1, 1970,

through 2009.

Years U-234a Th-230a Ra-226a Po-210a Pa-231a Ac-227a

1970 22.1 5.73 4.63 4.63 0.221 6.40

1971 5.58 1.45 1.17 1.17 0.056 1.62

1972 1.41 0.367 0.297 0.297 0.014 0.410

1973 0.356 0.093 0.075 0.075 0.004 0.103

1974 0.090 0.023 0.019 0.019 0.001 0.026

1975 0.023 0.006 0.005 0.005 0.0002 0.007

1976–2009 0.006 0.002 0.001 0.001 0.0001 0.002

a. Listed values are assigned as a constant distribution.

Radon Progeny

There are no radiological radon survey data available for the remodeling period (1954 to 1969). 

Therefore the radon air concentrations for the end of the remodeling period are assumed to be the 

same as the previous period (0.480 WLM/yr) as described in Section 3.5.1.
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Fifty-five measurements of radon progeny concentration were made between 1976 (ORNL 1978) and 

1981 (BNI 1982) in the Tonawanda site buildings. The highest radon concentration (from Building 31 

in 1981) was used to evaluate radon exposures. No correction was made for natural background 

radon, although these levels were comparable to the levels in Building 31 in 1981.

A fit was made between the levels that were used for the remediation period (0.48 WLM/yr) to the 

1981 results (0.201 WLM/yr). During the remediation of Building 31 between 1995 and 1998 

(McKenzie and Uziel 1998), radon concentration levels were measured. These levels were consistent 

with the levels that were measured in the 1976 and 1981 surveys. Therefore, the exposure rates after 

1981 for radon are assumed to be constant at the 1981 levels. Table 6-4 lists these rates.

Table 6-4. Radon exposure summary from 

January 1, 1970, through 2009 (WLM/yr). 

Years Exposure ratea

1970 0.480

1971 0.444

1972 0.410

1973 0.379

1974 0.350

1975 0.323

1976 0.299

1977 0.276

1978 0.255

1979 0.235

1980 0.218

1981–2009 0.201

a. Values listed are assigned as a constant distribution.

6.2 EXTERNAL BETA AND GAMMA EXPOSURE

As noted, Building 30 was found in 1976 to be the most contaminated building on the site (ORNL 

1978). Measured floor and wall radiation levels in Building 30 in 1976 were compared with similar 

measurements from 1950 immediately after its decontamination (see Section 4.1.1 and 

Heatherton 1950). The results were similar, but the 1950 values were slightly higher.

Outdoor gamma radiation levels at 1 m 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/hr (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 nonuniform and varying from background up to 250 μR/hr. Regions with readings 

above about 20 μR/hr 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/hr. BNI (1982, p. B-12) stated that the points of maximum radiation that were determined in 

the ground surveys were slightly displaced from but in general agreement with those that were 

determined in a 1979 aerial radiological survey. The total number of reported readings ≥25 μR/hr was 

16. The net readings (after subtraction of 8 μR/hr to correct for background) had a GM of 94 μR/hr 

and a GSD of 3.95. This was taken as an estimate of worker exposure rate when outdoors. This 

estimate was assumed to apply starting January 1, 1954. No credit was taken for the soil remediation 

at the site that began in 2000 because the remediation has not been completed (Pilon 2004). The 

estimate is probably an overestimate for the following reasons: measurements were reported only for 

areas that were 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.
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The beta dose rate at 3 ft above the ground that corresponded to the adopted value of gamma 

exposure rate was estimated as 4.38 × 10-1 mrem/hr, which is 4.66 times the gamma rate where 4.66 

is the ratio of beta mrem/hr at 3 ft to gamma mR/hr at 3 ft for the floors and walls in Building 30 after 

decontamination (Table 4-1). 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 exposures after cleanup, based on Building 30 contamination levels, were 

used to estimate the external exposure rate. No adjustments for changes in work hours were made. 

The radiation energy distributions were assumed to be same as those during the operational period 

(lognormal distribution with a GSD of 3). This results in an exposure rate of 0.068 rem/yr penetrating 

dose and 0.326 rem/yr nonpenetrating dose.

6.3 NEUTRON EXPOSURE

Because only small quantities of radioactive material were on the site in the period after cleanup, 

there were no significant neutron exposures.

6.4 UTILITY TUNNEL EXPOSURE

Tunnel Contamination

There are two pathways for the radiological contamination of the tunnels: (1) transport of the effluent 

through the ground formations into the tunnels directly, and (2) stormwater runoff contaminated by the 

waste effluent overflow (possibly remaining in the soil following well overflow) into the tunnels during 

flooding conditions.

The scenario of effluent transport through the ground formations and groundwater transport into the 

tunnels has multiple challenges. The effluent was pumped into the bedrock level at depths between 

90 and 150 ft. Documentation of the greatest depth of the tunnel has not been found as yet, but it is 

likely that the tunnel would not have been nearly that deep given that access could be accomplished 

through covered manways. The information NIOSH has about accessing the tunnel for maintenance 

activities implies the tunnel would not have exceeded 20 to 30 ft in depth. Consideration of the cost 

associated with digging in bedrock to build the utility tunnel, and the associated cost of running the 

utilities back up to the buildings they would supply, supports this conclusion. If the tunnel did have a 

depth of 20 to 30 ft from the surface, upward transport of the radiological contamination would be 

required for the effluent to directly contaminate the tunnel level. The remedial investigation report 

(BNI 1993) describes, in some detail, the soil interval between the ground surface and the water table, 

known as the vadose zone, and how flow occurs through this layer. The report made a conservative 

estimate of vertical groundwater velocity and applied a retardation factor to account for the slower 

movement of contaminants through the vadose zone due to interactions between the contaminants 

and the soil particles. The report states the uranium contamination in groundwater would migrate 

47 times slower than the estimated vertical groundwater velocity of 0.09 m/yr (0.3 ft/yr). The 

retardation factors for thorium and radium are typically greater than for uranium, so thorium and 

radium would migrate at a slower rate. Radiological contaminants from the effluents injected into the 

contact-zone aquifer underneath the Linde site appear to be immobilized in the particulate phase 

(BNI 1993, p. 5–17).

Another challenge to overcome in terms of tunnel contamination directly from the effluent is the 

construction of the tunnel. The Linde utility tunnel is constructed of reinforced concrete. There is 

evidence of degradation in several older sections where the concrete is brittle and cracked. Such 

locations, along with cracks and seams, would be the best pathways for contaminated soil and 

groundwater to infiltrate the tunnel.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 71 of 101

The assumption that the infiltration of the tunnel was due to contaminated stormwater runoff is also 

due consideration. It is possible that stormwater passing through contaminated soils and flooding the 

tunnel could leave contamination behind. It is unlikely that any soluble uranium contamination in the 

soil around the injection wells would remain available for contamination of the tunnel for very long 

beyond the end of use and overflow of the wells. Natural weather action would be likely to have 

scoured the soluble components from the soil long before the timeframe under evaluation (1954 to

2006; the final injections occurred in mid-July 1946), leaving insoluble uranium to contaminate the 

tunnel.

The limited distribution of the contamination in the tunnel, as demonstrated by the 2001 survey 

(USACE 2002b), is difficult to reconcile with the idea of contamination washing into the tunnel during 

flooding. The dose assessment performed by the USACE concludes the contamination in the utility 

tunnel is a result of operations associated with Building 14 and contaminated solid and in-leakage 

near Building 31. This conclusion is supported by the limited distribution of contamination in the 

tunnels. Had the tunnels been contaminated by stormwater runoff flooding as has been suggested, 

the distribution of the tunnel contamination would be expected to be more uniform because there is no 

evidence of any remediation of the tunnels. The action of pumping the stormwater from the tunnel 

would have further distributed and diluted any contamination that came in via floodwaters. This is not 

consistent with the distribution seen on the survey map.

From review of the locations of the former injection wells near Building 8 and the section of the utility 

tunnel closest to those wells, the tunnel is approximately 25 ft to the north of the injection wells at the 

closest point. The nearest contaminated (greater than 5,000 beta/m/100cm2

) tunnel areas (identified 

by USACE 2002b) are approximately 50 to 100 ft northwest of the closest injection well. The second 

set of injection wells, in the vicinity of Building 30, are approximately 80 ft, 180 ft, 280 ft, and 400 ft 

from the closest contaminated section of utility tunnel (which runs north to south just to the west of 

Building 31).

Source Term

The detailed radiological characterization performed in 2001 was based on beta surface 

contamination levels. Six to eight circumferential measurements were taken in 1-m increments along 

the tunnel. The maximum result was then recorded. In addition, hot spot readings were collected.

The beta contamination levels were then converted to individual isotopic activities based on gamma 

spectroscopy analysis of soil samples that were taken during the site remediation. The analysis of 

this data is presented in Table 6-5.

Table 6-5. Radionuclide fraction calculation using average Linde soil concentrations

(USACE 2002b). 

Isotope

Soil 

concentratio

n gamma 

spec (pCi/g)

Uranium 

fraction 

f(U)

Soil 

concentratio

n isotopic 

(pCi/g)

Source 

fractio

n (fs)

Beta 

emission 

(β/dis)

Beta 

fractio

n (fβ)

Isotopic 

abundance 

per beta 

(SA)

Radium-226 7.46 – 7.46 0.080 1.82 0.170 0.094

Thorium-230 8.49 – 8.49 0.092 – – 0.107

Uranium-238 36.33 0.473 36.33 0.392 1.98 0.830 0.419

Uranium-235 – 0.022 1.69 0.018 – 0.020

Uranium-234 – 0.505 38.79 0.418 – – 0.447

Total – – 92.76 1.000 3.80 – –
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 72 of 101

Using the data in Table 6-5, the following equations can be used to convert the gross beta surface 

contamination levels to isotopic activities:

226Ra (dpm/m2

) = 9.4 × [(β/min)/100 cm2

]

230Th (dpm/m2

) = 10.7 × [(β/min)/100 cm2

]

238U (dpm/m2

) = 41.9 × [(β/min)/100 cm2

]

235U (dpm/m2

) = 2.0 × [(β/min)/100 cm2

]

234U (dpm/m2

) = 44.7 × [(β/min)/100 cm2

]

The 1937 tunnels were not found to be contaminated with radium because the pre-World War II 

uranium-refining operations in Building 14 used preprocessed ores from the Colorado Plateau 

vanadium deposits, which did not contain radium. Therefore, for these tunnels, 100% of the beta is 

assumed to be from 238U. This results in the following equations to convert the gross beta surface 

contamination levels to isotopic activities:

226Ra (dpm/m2

) = 50.5 × [(β/min)/100 cm2

]

230Th (dpm/m2

) = 2.3 × [(β/min)/100 cm2

]

238U (dpm/m2

) = 53.8 × [(β/min)/100 cm2

]

An analysis of all of the gross beta survey data was performed and is summarized in Table 6-6.

Table 6-6. Gross beta survey data for utility tunnels.

50th percentile 6,430 (β/min)/100 cm2

GSD 2.10

95th percentilea 30,870 (β/min)/100 cm2

a. The 95th percentile result is the measured 95th not the fitted since 

the tail of the curve bent above the projected fit.

Table 6-7 provides isotopic contamination levels that were calculated using the equations above and 

the 95th-percentile gross beta surface contamination level.

Table 6-7. Utility tunnels surface activity (dpm/m2

). 

Isotope

1954–1956

surface activity

1957 on

surface activity

Radium-226 - 2.90E+05

Thorium-230 - 3.30E+05

Uranium-238 1.56E+06 1.29E+06

Uranium-235 7.10E+04 6.17E+04

Uranium-234 1.66E+06 1.38E+06

These contamination levels were assumed to be uniformly present on all surfaces of the tunnels for all 

years. This source term is considered to be favorable to the claimant for the following reasons:

The beta measurements are based on the highest observed measurements.

1. The beta measurements include both fixed and removable contamination (it is normally 

estimated that contamination is only 10% removable and therefore a contributor to airborne 

contamination. Including both fixed and removable contamination overestimates the 

contamination available for resuspension.).

2. Hot spot contamination levels were treated as uniformly distributed contaminated surfaces and 

included in the distribution.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 73 of 101

3. The tunnels were reportedly a damp environment, which would have the effect of preventing 

resuspension of material.

Internal

Based on the 95th-percentile beta surface contamination level (Table 6-6), an air concentration 

activity was calculated based on guidance provided in ORAUT-OTIB-0070, Dose Reconstruction 

during Residual Radioactivity Periods at Atomic Weapons Employer Facilities (ORAUT 2012). The 

following assumptions were made:

• Breathing rate: 1.2 m3

/hr

• Resuspension factor: 1 × 10-6 m-1

It is assumed that trade workers and laborers worked in these tunnels doing maintenance 50% of the 

time (ATL 2012). For all other workers, an occupancy factor of 5% was assumed (ATL 2012). This 

results in the intakes in Tables 6-8 and 6-9.

Table 6-8. Utility tunnels trade worker intake rates (dpm/yr).

Trade workers and laborers

Radionuclide

1954–1956 1957 on

Inhalation rate Ingestion rate Inhalation rate Ingestion rate

Radium-226 - - 348 7.3

Thorium-230 - - 396 8.3

Uranium-238 1871 39.0 1552 32.3

Uranium-235 85.2 1.8 74.1 1.5

Uranium-234 1993 41.5 1656 34.5

Table 6-9. Utility tunnels other worker intake rates (dpm/yr).

All other workers

Radionuclide

1954–1956 1957 on

Inhalation rate Ingestion rate Inhalation rate Ingestion rate

Radium-226 - - 34.8 0.7

Thorium-230 - - 39.6 0.8

Uranium-238 187 3.9 155 3.2

Uranium-235 8.5 0.2 7.4 0.2

Uranium-234 199 4.2 166 3.4

Radon

Based on Tunnel Contamination

Radon is the decay product of 226Ra. The level of 226Ra contamination in becquerels per square 

centimeter multiplied by the contaminated area provides the total 226Ra contamination in becquerels 

(disintegrations per second). Because each disintegration of a 226Ra atom produces a 222Rn atom, 

this is then the production rate of 222Rn in atoms per second. Multiplying by the 222Rn decay constant 

converts the production rate into becquerels per second. Dividing this by the volume of air into which 

the radon is emitted produces the rate that the air concentration is increased.

The air concentration is decreased by two factors. The first is due to the radioactive decay of the 

radon itself, and the second is the rate that ventilation removes the radon from the tunnel. This can 

be expressed mathematically by the following equation:

λ λ λ× × = −− t

t Vt

dC S A C C

dt V (6-1)
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 74 of 101

where:

Ct = 222Rn concentration in the tunnel (Bq/cm3

)

S = 226Ra contamination value (Bq/cm2

)

A = surface area of the contaminated tunnel walls/ceiling and floor (cm2

)

V = air volume in the tunnel (cm3

)

λ = decay constant of 222Rn gas (1/s)

λV = air exchange rate in the tunnel due to ventilation (1/s)

At steady state, the derivative equals zero and the equation can be rearranged to find the steady-state 

radon concentration in the tunnel, which can be expressed as:

( )

λ

λ λ

× × = × + t

v

S A C

V (6-2)

The area of the tunnel surfaces (walls, ceiling, and floor) can be determined on a per-unit length 

basis. The volume of air associated with the unit length can be similarly determined. Because the 

area is divided by the volume, the per-unit basis cancels out. Using the parameters in Table 6-10 and 

converting units, the equilibrium radon concentration in the tunnel due to surface contamination can 

be calculated to be 18.35 pCi/L.

Table 6-10. Radon parameters for utility tunnels.

Parameter Value Units

S - Radium-226 contamination value (based on 290,000 dpm/m2) 0.483 Bq/cm2

A – area per linear cm (based on 200 cm square cross-section) 800 cm2

V – volume per linear cm 40000 cm3

λ - Radon-222 decay constant 2.1E-6 1/s

λV - Air exchange rate 2.78E-5 1/s

For Trade Workers & Laborer it is assumed that they worked in these tunnels doing maintenance 50% 

of their time (ATL 2012). For all other workers, an occupancy factor of 5% was assumed (ATL 2012). 

This results in the exposure rates provided in Table 6-11. Because radium was not present in the 

tunnels built before 1957, these exposure rates only apply after 1957.

Table 6-11. Exposure rates from utility tunnel radon 

contamination (WLM/yr).

Radionuclide Trade workers All other workers

Radon-222 0.540 0.054

Based on Surrounding Soil

The 222Rn concentration in the tunnels at the Linde Ceramics Plant (Rntunnels) was estimated using the 

following equation:

= × tunnels Erie Co basements Linde Erie Co Rn Rn Ra Ra / (6-3)

where:

RnErie Co basements = 222Rn concentration reported for basements in Erie County.

RaLinde = 226Ra concentration in soils at Linde Ceramics.

RaErie Co = 226Ra concentration in soils in Erie County.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 75 of 101

The Linde site is just south of the county line in Erie County. The Erie County statistics for radon in 

basements have a GM of 1.68 pCi/L and a GSD of 3.95 pCi/L (NYDOH 2010). The Erie County 226Ra 

soil concentration is taken to be 0.636 pCi/g and is applied as a constant value. This number was 

derived from the number for Niagara County that Sanford Cohen & Associates (SC&A) calculated, 

using 1.81 ppm of uranium, and then proportionally increasing the result using the Erie County figure 

of 1.89 ppm (LBNL 2011).

The 226Ra concentration in soils at the Ceramics Plant was taken from 51 soil sample results from an 

aggregate 59 ft of soil samples. Discrete probabilities were assigned for each foot of soil by assuming 

the concentration reported for each foot of soil had a 0.0169 (1/59) probability of occurrence.

The computer program @Risk was used to solve the equation for radon concentration in the tunnels. 

Input parameters were applied as the distribution described above. To solve the equation, 10 

simulations were run with 10,000 iterations each, with different random seed numbers. The 

95th-percentile value of 99.31 pCi/L provides a plausible upper-bound estimate for radon exposure in 

the Linde utility tunnels for all years at the Linde site beginning from the first MED production through 

the present.

For Trade Workers & Laborer it is assumed that they worked in these tunnels doing maintenance 50% 

of their time (ATL 2012). For all other workers, an occupancy factor of 5% was assumed (ATL 2012). 

This results in the exposure rates provided in Table 6-12. Because radium was not present in the 

tunnels built before 1957, these exposure rates only apply after 1957.

Table 6-12. Radon exposure rates for utility tunnels from 

surrounding soils (WLM/yr). 

Radionuclide Trade workers All other workers

Radon-222 2.921 0.292

External

External contamination and submersion dose coefficients factors (DCFs) were calculated for effective 

whole-body and skin dose based on information in Federal Guidance Report No. 12, External 

Exposure to Radionuclides in Air, Water, and Soil (Eckerman and Ryman 1993). These DCFs are 

based on a single-plane exposure model. However, the fact that the 95th-percentile surface 

contamination level (including hotspot activities in the distribution) is being applied to the entire length 

of the tunnel is considered to bound any dose that would be associated with contamination on the 

ceiling and walls of the tunnels.

It is assumed that trade workers and laborers worked in these tunnels doing maintenance 50% of the 

time (ATL 2012). For all other workers, an occupancy factor of 5% was assumed (ATL 2012). 

Using these DCFs, occupancy factors, and the calculated surface activity and air concentrations from 

the internal dose section, an external dose potential was estimated. These potential external annual 

dose rates are listed in Table 6-13. An assumed energy distribution of 100% 30- to 250-keV photons 

for the whole-body dose and 100% >15-keV electrons for the skin dose is recommended as an 

assumption favorable to claimants. Since these estimates are based on source term data, external 

ambient DCFs (H*(10)) should be used to assign organ dose.

Table 6-13. Utility tunnels external exposure rates (rem/yr). 

Radionuclide

1954–1956 1957 on

Gamma Beta Gamma Beta

Trade workers 0.001a 0.093 0.003 0.099

All others workers 0.001a 0.009 0.001a 0.010

a. Dose less than 0.001 rem. 
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7.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 77 of 101

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Office, Kenmore, New York, September 4. [SRDB Ref ID: 12382, p. 20]

Van Horn, E. L., 1945, untitled letter to A. R. Holmes (Linde Air Products Company), U.S. Engineer 

Office, Kenmore, New York, February 22. [SRDB Ref ID: 21817]

Wallace, P., 2003, “Notes on Linde data,” Center for Epidemiologic Research, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, August 18. [SRDB Ref ID: 4107]

Wallo, A., III, E. A. Vierzba, and G. N. Roberts, 1981, Evaluation of the 1943-to-1946 Liquid Effluent 

Discharge From the Linde Air Products Company Ceramics Plant, ATR-82 (7963-04)-2, The 

Aerospace Corporation, Germantown, Maryland, December. [SRDB Ref ID: 8829]
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 89 of 101

Warren, S. L., 1944, “Routine Pelvic Films on Special Materials Employees,” memorandum to Area 

Engineer, EIDMM-3, Manhattan District, New York, New York, January 11. [SRDB Ref ID: 

5815, p. 4]

Wiesendanger, R., 1944, Pilot Plant Operations on African Ore, Tonawanda Laboratory, Linde Air 

Products Company, Tonawanda, New York, M-4541, U.S. Atomic Energy Commission, 

May 26. [SRDB Ref ID: 3757]

Wilson, R. H., 1959, “The Hanford Uranium Bio-assay Program,” Symposium on Occupational Health 

Experience and Practices in the Uranium Industry, Held in New York City, October 15–17, 

1958, HASL-58, U.S. Atomic Energy Commission, New York Operations Office, New York, 

New York, September, p. 77. [SRDB Ref ID: 7886, p. 89]

Wolf, B. S., 1948, “Results of Uranium and Fluoride Analyses,” letter to R. C. Heatherton (Linde Air 

Products Company), U.S. Atomic Energy Commission, New York, New York, February 27. 

[SRDB Ref ID: 16979. p. 96]

Wolf, B. S., 1949a, untitled letter to E. O. Kent (Linde Air Products Company), U.S. Atomic Energy 

Commission, New York, New York, March 16. [SRDB Ref ID: 16316, p. 17]

Wolf, B. S., 1949b, “Linde Air Products Co., Division of Union Carbide & Carbon Corp, Ceramics 

Plant, Tonawanda, N.Y. – Decontamination of Main Building,” memorandum to F. M. Belmore, 

U.S. Atomic Energy Commission, New York, New York, May 12. [SRDB Ref ID: 16316, p. 13]
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 90 of 101

GLOSSARY

carnotite

Mineral of the form K2(UO2)2(VO2)2·

3

(H2O) found in Colorado and Utah and mined for its 

uranium and vanadium content. Much of the domestic ore processed at Linde Ceramics Plant 

derived from tailings from carnotite ore. Preprocessing of this material before shipment is 

believed to have removed much of its radium content (Wallo, Vierzba, and Roberts 1981).

day (d)

As used in this Linde document, 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 (wd) refers to a day that is assigned for work. The length of a workday depends on 

the amount of time spent on the job. Default assumptions are 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 higher than the defaults for the early 

years.

Linde database

Set of records of Linde Ceramics Plant operations available to the National Institute for 

Occupational Safety and Health dose reconstruction project.

N.G. or NG Cake

Filter cake produced in Step II processing at Linde. There were two types of cakes produced 

called OK and N.G.

pitchblende

Mineral containing uranium oxide (UOX) of variable composition ranging between UO2 and 

U3O8 as well as radium. Raw ore from Africa was processed for its uranium content at Linde 

Ceramics Plant the raw ore was not prerefined and so contained all of the natural uranium 

radioactive progeny including radium.

Step I, Step II, and Step III

Sequential uranium production processes at Linde:

• Step I, conversion of 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

Mineral in the form of Cu(UO2)2(PO4)2·8-12(H2O) processed for its uranium content. Like the 

African pitchblende processed at Linde, torbernite was not prerefined before its arrival at Linde 

and so contained all of the natural uranium radioactive progeny including radium.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 91 of 101

ATTACHMENT A

BETA RADIATION

TABLE OF CONTENTS

SECTION TITLE PAGE

A-1 Maximum beta surface dose rates from various uranium-containing materials....................... 91

A-2 Electron energy released by refined uranium ......................................................................... 91

A-3 Electron energy released by unrefined natural uranium ore ................................................... 92

A-4 Estimated attenuation of beta radiation with distance............................................................. 93

A-5 Uranium beta dose reduction factors for apparel.................................................................... 93

A-6 Variation of dose rate with body location for Step III process operations................................ 94

This attachment contains data that was 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/hra,b rad/8 hr

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)a

Electron energy

per decay of parent

(MeV/nt)a % 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

a. nt = nuclear transformation.
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ATTACHMENT A

BETA RADIATION (continued)

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

Principal 

nuclides

Electron energy

per decay of nuclide

(MeV/nt)a

Electron energy

per decay of parent

(MeV/nt)a % 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

a. nt = nuclear transformation.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 93 of 101

ATTACHMENT A

BETA RADIATION (continued)

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 

byproductsc

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 (e) 0.012

12.6f (e) 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 centimeters. 

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 was estimated from the rule of thumb that the half thickness of a beta 

absorber is one-eighth of the range of the beta rays (Cember 1983, p. 97). For natural uranium ore and 

byproducts with beta emitters other than Bi-214, the actual falloff with distance would be greater 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. No radiation at this distance.

f. Range of 3.27-MeV beta radiation per Cember (1983, p. 100).

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

Itema

Fraction of beta

dose 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).
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ATTACHMENT A

BETA RADIATION (continued)

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

Operation

(mR/hr)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)

Radiation level relative to chestc

Radiation level 

relative 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 (1948h).

b. No data.

c. Results in lower half of the table are based on the data in the upper half.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 95 of 101

ATTACHMENT B

CODES AND SPECIAL TERMINOLOGY

Special codes and terminology were used in correspondence and documents that 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 might have had multiple meanings. 

Table B-1 lists apparent definitions deduced during preparation of this document.

Table B-1. Codes and special terminology. 

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% to 16.5% 

U3O8 (Aerospace 1981, Table B-1)

L-30 African pitchblende ore; estimated to contain 8% to 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 
ATTACHMENT C

DATA SOURCES ON URANIUM PROGENY CONCENTRATIONS IN LINDE MATERIALS

Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 9

6 of 101

calculated value

<MDA, mda shown

SRDB Ref ID Date Location Description U238 U234 U235 totalU Th230 Ra226 Th232 Th228 Ac227 Pa231 units Th230/U Ra226/U Th232/U Ac/U Pa/U

16294 1995 bldg 14 dry valve pit (dust) 1038 1068 53.8 2159.8 354.7 15.3 2.3 1.7 pCi/g 0.16 0.01 0.00

sump (dust) 1.4 1.7 0.07 3.17 0.72 0.52 0.2 0.22 pCi/g 0.23 0.16 0.06

corridor overhead (dust) 369 378.8 11.6 759.4 60.2 0.14 5 0.67 pCi/g 0.08 0.00 0.01

corridor wall (terra cotta block) 3.2 3.2 0.14 6.54 1.7 1.4 1.4 1.4 pCi/g 0.26 0.21 0.21

14620 1978 bldg 30 air samples during D&D 1.90E-08 2.10E-09 1.10E-09 pCi/ml 0.11 0.06 0.00

9009 1981 sediment, onsite and offsite Ellicott Creek 0.82 0.05 1.69 0.6 0.55 0.7 pCi/g 0.36 0.33 0.41

Creek1 0.95 0.05 1.95 0.7 0.7 0.8 pCi/g 0.36 0.36 0.41

Twomile Creek - upstream 4.3 0.1 8.7 0.92 0.69 0.01 pCi/g 0.11 0.08 0.00

Twomile Creek - linde discharge 0.71 0.06 1.48 0.02 0.52 0.02 pCi/g 0.01 0.35 0.01

Twomile Creek - downstream 1.5 0.05 3.05 0.96 0.59 0.48 pCi/g 0.31 0.19 0.16

Storm Sewer 6.47 0.19 13.13 1.4 1.35 0.62 pCi/g 0.11 0.10 0.05

Storm Sewer 99 4.57 202.57 18 6.93 0.51 pCi/g 0.09 0.03 0.00

Storm Sewer 13 0.52 26.52 2 1.59 0.65 pCi/g 0.08 0.06 0.02

Storm Sewer 116 4.1 236.1 9.9 0.89 0.34 pCi/g 0.04 0.00 0.00

Storm Sewer 4.5 0.17 9.17 0.2 0.64 0.39 pCi/g 0.02 0.07 0.04

Sanitary Sewer 362 13 737 1.33 1.94 0.11 pCi/g 0.00 0.00 0.00

Sanitary Sewer 0.51 0.05 1.07 0.34 0.38 0.21 pCi/g 0.32 0.36 0.20

9026 1990 Linde Soils area 1 - mean 11.2 22.4 7.8 4.3 1.6 pCi/g 0.35 0.19 0.07

area 2 - mean 12.7 25.4 5.7 3.4 1.4 pCi/g 0.22 0.13 0.06

area 3 - mean 17.1 34.2 24.4 9.4 1.4 pCi/g 0.71 0.27 0.04

area 4 - mean 46.8 93.6 30.7 9.8 1.4 pCi/g 0.33 0.10 0.01

8828 1981 Linde Soil and Sediment near disposal well - subsurface (loc 11) 24.05 0.84 48.94 5.9 5.53 0.92 14.25 0.73 pCi/g 0.12 0.11 0.02 0.29 0.01

test well debris (loc 13) 26.4 1.09 53.89 3.53 0.82 0.51 2.1 0.29 pCi/g 0.07 0.02 0.01 0.04 0.01

Sanitary Sewer (loc 15) 362 12.93 736.93 1.33 1.14 0.11 5.54 0.95 pCi/g 0.00 0.00 0.00 0.01 0.00

Storm Sewer (loc 19) 99.2 4.57 202.97 17.7 6.93 0.51 14.29 1.14 pCi/g 0.09 0.03 0.00 0.07 0.01

Storm Sewer (loc 21) 116 4.1 236.1 3.89 0.89 0.34 3.07 0.39 pCi/g 0.02 0.00 0.00 0.01 0.00
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 97 of 101

ATTACHMENT D

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD

Original Issue 

During the SC&A-conducted interview in Buffalo, Linde site experts and past workers 

indicated that there were many thousands of used burlap bags stacked up in the open 

bay area behind Building 30 (see Attachment 3 of this review report). These bags were 

used for transporting uranium ore to Linde for processing. After the end of the operation 

period, the contaminated burlap bags were stored behind Building 30 awaiting disposal. 

Many Linde workers, operation staff, and administrative personnel sat on these 

contaminated bags during breaks and lunch periods. This practice continued for many 

years, exposing many people at close distances to beta and gamma radiation sources 

left over in the uranium-contaminated burlap bags. The site profile does not estimate 

the missed beta and gamma doses to workers resulting from sitting on or standing next 

to those burlap bags (SC&A 2006).

Subsequent Analysis

During the March 26, 2007, Linde Working Group meeting the burlap bag issue was discussed by 

members of the working group, NIOSH, and Sanford Cohen & Associates (SC&A). The following task 

was defined by the working group (based on SC&A 2006):

NIOSH to investigate details of used burlap bags. Which bags (formerly containing 

African or domestic ore) were stored at which locations and during which periods of time. 

This may affect both internal and external exposures. Even though African ores were 

processed only during the SEC [Special Exposure Cohort] period (pre 10/31/1947), 

empty bags, that had contained African ore, may have been around longer (i.e., after 

1950).

NIOSH to determine whether there was an on-site incinerator to burn used burlap bags 

and, if so, the possible effects on internal and external exposures.

In November 2007, an analysis of the burlap bag issue was conducted by NIOSH and submitted to 

the working group along with responses to the other issues that had been identified in the site profile 

review. Table 5-1 of the response document summarizes the issue and is reproduced below 

(NIOSH 2007b).

Date Data Source Data

Undated Medical Section, Manhattan 

District, Data Sheet for Industrial 

Hazard Rating

(MED undated)

• Step I material (ore) arrives in burlap bags, 

covered in paper sacks

• Step III material (UO2) arrives in 75 lb drums

1943–1946 Linde Construction, Operations 

Report

(CCCC 1946c)

• Empty bags and drums are held until they can be 

burned or scrapped

• Bags from dry ore are shaken in the bag shaker 

... the bags are next taken to the laundry room 

for washing. All bags are then burned in the 

incinerator. Ashes from the incinerator are 

returned to the grinding circuit for processing.
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ATTACHMENT D

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD (continued)

Date Data Source Data

08-25-1943 “Safety Precautions, Ceramics 

Plant”

(Murphy 1943) 

• Bags in which low-grade ore is received are to be 

sent back to the original supplier

• Bags in which African ore is delivered are to be 

burned to recover the elements

07-04-1944 “Interview with [Name redacted; 

Privacy Act Information], a Former 

Linde Employee,” April 10, 1981

(Dupree 1981)

• Washing of burlap bags in which the ore was 

received began

07-15-1944 Step I - L-30 Ore

(Olevitch 1944)

• Empty ore bags contain about 1⁄2 lb of ore

• Washing burlap bags recovers 70% of material

• After washing, the bags are removed, 

centrifuged, and stored out in the backyard of the 

plant. To date about 19,000 bags have been 

washed and stored.

10-10-1944 “Interview with [Name redacted; 

Privacy Act Information], a Former 

Linde Employee,” April 10, 1981

(Dupree 1981)

• Burning of burlap bags in which the ore was 

received began

10-20-1944 “Visit to the Tonawanda Area, 18 

and 19 October 1944”

(Tybout 1944a)

• The contractor has decided not to use the bag 

washer for the bags from L-50 (the low grade 

ore)

• The incinerator which will be used for burning the 

bags was inspected. At the time, it had been in 

operation only one or two days.

10-??-1944 Interview with [Name redacted; 

Privacy Act Information], a former 

Linde employee, April 10, 1981

(Dupree 1981)

• The burlap bags had to be disposed of. To 

control the ore dust that remained in the bags 

after they were dumped and prevent dust from 

escaping during their disposal, a series of 

mechanical devices were installed. First a bag 

shaker was added to the disposal process to 

shake out ore dust left behind after dumping. 

Than a bag washer was installed to wash any 

material remaining in the bag after it had been 

shaken. Finally, a bag incinerator was installed 

to burn the bags after they had been washed. 

This entire process was completely installed by 

October, 1944.

02-22-1945 E. L. Van Horn (USACE) to A. R. 

Holmes (CCCC) 

(Van Horn 1945)

• ... It is desired, however, to continue film 

monitoring on employees engaged in burning 

L-30 ore bags ... 

06-20-1946 “Interview with [Name redacted; 

Privacy Act Information], a Former 

Linde Employee,” April 10, 1981

(Dupree 1981)

• Burning of burlap bags in which the ore was 

received ends

Just before the January 8 working group meeting, a review of NIOSH (2007b) from SC&A was 

received (SC&A 2008): 

Section 5.0 of the NIOSH response discusses the “burlap bag issue.” NIOSH concludes 

that “Based on the reviewed historical records, and considering the fact that the period 
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ATTACHMENT D

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD (continued)

during which the burlap bags were staged and burned is within the current SEC period, 

a revision to the current dose reconstruction methodology is not warranted” (NIOSH 

2007[c], Section 5.0). This, however, does not adequately respond to the site expert 

interview assertion that thousands of burlap bags were still stacked behind Building 30 

after 1950; as stated in the SC&A site profile review: “During the MED period, they 

stacked all the contaminated burlap bags in storage area of Building 30. These 

contaminated burlap bags were kept in there until they were removed to be burned in 

the incinerator in the late 1950s. Many of the people working in Building 30, including 

operation personnel, secretaries, and maintenance workers, would sit on those bags 

resting or eating their lunch. This went on for many years.” (SC&A 2006, Attach. 3, p. 

112)

During the January 8 working group meeting an attempt was made to resolve the burlap bag issue, 

considering both NIOSH (2007b) and SC&A (2006). A consensus on the subject could not be 

reached. The conclusion of the working group was to have a technical conference that focused on 

the burlap bag issue. SC&A agreed to review the interview notes that were associated with the issue. 

As a result, a conference call was held on February 13, 2008. The former employee that reported the 

burlap bag issue was able to participate on this call.

Based on clarification of the burlap bag issue from the former employee who made the original 

assertion, it was determined that the burlap bags in question were inside Building 30:

... there were two pallets of bags containing some material in Building 30 in the time 

period around August 1951. He and his friends would get drinks, stand by the pile of 

bags, and place their cups on the bags. He was told by one co-worker that the bags 

were left over from the Manhattan project. He was asked about the appearance of the 

bags and he replied that they looked like large canvas sandbags.

An affidavit that describes the burlap bags was submitted by this former employee along with the 

Linde SEC Petition Documents that were submitted on March 19, 2008.

Based on this newly clarified description of the burlap bags, NIOSH was tasked with an assessment 

to compare calculated doses from an assumed pile of bags (two pallets of bags 2 to 3 ft high, for both 

ore and concentrate) to currently assigned doses.

Comparison of currently assigned dose to potential exposure from ore bags

Currently assigned dose (1950 to 1954)

1.85 R/yr (GM) with GSD of 4.04 from Table 4-20. (note: 95th-percentile value would be 18.5 R/yr).

Dose assessment for the maximizing case:

Measurements take in 1944 on African ore (Skinner 1944)

Location Gamma dose rate

Ore sampling room

15 ore bags in center on ore 0.019 R/hr

1 ft from ore bags 0.0062 R/hr

Ore box car

Center of car on ore 0.017 R/hr
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 100 of 101

ATTACHMENT D

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD (continued)

Location Gamma dose rate

Ore storage area (pile 4 by 5 by 20 ft)

Center of 4- by 20-ft face on ore 0.040 R/hr

Center of 4- by 20-ft face on ore 0.026 R/hr

1 ft from 4- by 20-ft face 0.013 R/hr

2 ft from 4- by 20-ft face 0.013 R/hr

3 ft from 4- by 20-ft face 0.010 R/hr

5 ft from 4- by 20-ft face 0.0061 R/hr

10ft from 4- by 20-ft face 0.0028 R/hr

Based on the description of the exposure conditions from the former employee, the following 

exposure conditions and assumptions can be inferred:

• A dose rate of 0.0062 R/hr based on 1 ft distance from 15 ore bags, and

• An exposure duration of 1 hr/wd, 250 wd/yr.

These result in a total annual exposure of 1.5 R/yr

Summary

Based on the weight of the available evidence (tabulated below), it is unlikely that two pallets of 

uranium ore (which was last processed at Linde in 1946) would have been in Building 30 in 1951 

(5 years after the cessation of processing of uranium ore). The current external exposure model for 

the period in question incorporates uncertainty in the external dose assignment by application of a 

lognormal distribution with a GM of 1.85 and a GSD of 4.04. This assumed distribution (with a 

95th-percentile value of 18.5 R/yr) accounts for possible deviation of the actual worker exposure of 

the magnitude that would result from the assumption that two pallets of uranium ore were in 

Building 30 in 1951.
Document No. ORAUT-TKBS-0025 Revision No. 04 Effective Date: 05/08/2015 Page 101 of 101

ATTACHMENT D

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD (continued)

Information that contradicts the presence of 

2 pallets of uranium ore in Building 30 in 1951

Information that supports the presence of 

2 pallets of uranium ore in Building 30 in 1951

1. CCCC documentation outlines the processing of 

uranium ore including the receipt of uranium ore 

in bags, handling of the ore bags, and methods 

that were used to recover material held up in 

bags after they were dumped (shaking, washing, 

incineration, and recycling of incinerator ash).

2. CCCC documentation establishes that uranium 

ore was received in bags and that ore 

concentrates were received in drums.

3. CCCC documentation establishes end of 

uranium ore processing as 1946.

4. Affidavit (1981) from a former employee 

substantiates the description of the handling of 

the uranium ore bags in the CCCC

documentation.

5. Surveys of Building 30 in 1950 and the balance 

of the site in 1952 did not identify elevated 

radiation levels consistent with two pallets of 

uranium ore.

1. Affidavit (2008) from a former Linde employee 

identifies the presence of 2 pallets of bagged 

material in Building 30 in 1951.

2. Affidavit (2008) from a former Linde employee 

identifies the material as uranium ore based on 

statement from a coworker.

It should be noted that the uranium intake value in this revision of the site profile (15,200 dpm/d) 

would bound any internal exposure scenario from two pallets of uranium ore bags in the stated 

configuration (bagged with plastic lining) and conditions (storage).