ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 45

DOE Review Release 02/05/2013

Document Title:

An Exposure Matrix for Bridgeport Brass: 

Havens Laboratory and Adrian Plant

Document Number: ORAUT-TKBS-0030

Revision: 01

Effective Date: 01/24/2013

Type of Document: TBD

Supersedes: Revision 00

Subject Expert(s): Mutty M. Sharfi

Approval: Signature on File Approval Date: 01/15/2013

Mutty M. Sharfi, Document Owner

Concurrence: Signature on File Concurrence Date: 01/15/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 01/15/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 01/15/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/24/2013

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 2 of 45

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

09/15/2005 00 First approved issue. Training required: As determined by the Task 

Manager. Initiated by Cindy W. Bloom.

01/24/2013 01 Revision initiated to address comments from the Advisory Board 

Subcommittee on Dose Reconstruction. Updated external 

unmonitored dose in accordance with ORAUT-OTIB-0020. Added an 

internal approach to account for remediation efforts in 1976, 1985, 

and 1995. Incorporates formal internal and NIOSH review 

comments. Training required: As determined by the Objective 

Manager. Initiated by Mutty M. Sharfi.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 3 of 45

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 5

1.0 Introduction .............................................................................................................................. 7

1.1 Purpose........................................................................................................................ 8

1.2 Scope........................................................................................................................... 9

2.0 Site Description and Operational History.................................................................................. 9

2.1 Havens Laboratory ....................................................................................................... 9

2.2 Adrian Plant................................................................................................................ 13

3.0 Estimation of Internal Exposure.............................................................................................. 16

3.1 Uranium...................................................................................................................... 17

3.1.1 Uranium Bioassay........................................................................................... 17

3.1.2 Uranium Air Sampling ..................................................................................... 19

3.1.3 Comparison of Uranium Bioassay and Air Concentration Estimates ............... 23

3.1.4 Recycled Uranium........................................................................................... 24

3.2 Thorium ...................................................................................................................... 24

3.3 Miscellaneous Information Related to Internal Dose ................................................... 25

3.4 Occupational Internal Dose Reconstruction assumptions and summary..................... 25

4.0 Estimation of External Exposure ............................................................................................ 27

4.1 Occupationally Required Medical X-Rays................................................................... 28

4.2 Miscellaneous Information Related to External Dose.................................................. 29

4.3 Occupational External Dose Reconstruction Assumptions and Summary................... 29

5.0 Estimation of Dose from Remediation Activities ..................................................................... 30

6.0 Attributions and Annotations .................................................................................................. 32

References ........................................................................................................................................ 33

ATTACHMENT A, GRAPHS SHOWING PREDICTED BIOASSAY FROM AIR 

AND FITS OF COWORKER BIOASSAY ............................................................. 37
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LIST OF TABLES

TABLE TITLE PAGE

3-1 Uranium urinalysis results from coworker data....................................................................... 19

3-2 Inhalation intakes based on coworker data ............................................................................ 20

3-3 Comparison of daily weighted-average air dust measurements at Adrian Plant for 

various quenching and salting strategies................................................................................ 22

3-4 Adrian Plant daily weighted-average air dust concentrations by operator position 

measured January 11 to 13, 1955.......................................................................................... 22

3-5 Adrian Plant daily weighted-average air dust concentrations from January 27 to 29, 

1960.......................................................................................................................................23

3-6 Estimated uranium intake rates based on time-weighted air concentrations........................... 23

3-7 Estimate of contaminant activity fractions in a recycled uranium source term......................... 24

3-8 Estimated thorium intake rates based on uranium intakes...................................................... 24

3-9 Chronic uranium intake assumptions for unmonitored workers .............................................. 26

4-1 External dose for unmonitored workers.................................................................................. 28

4-2 External exposure summary for Havens Laboratory............................................................... 29

4-3 External exposure summary for Adrian Plant ......................................................................... 30

5-1 Annual internal and external exposure during remediation activities ...................................... 31

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Laboratory area at Havens Laboratory................................................................................... 11

2-2 First floor, main building, Havens Laboratory ......................................................................... 12

2-3 Dynapak Area, Havens Laboratory ........................................................................................ 12

2-4 Locations of AEC activity at Adrian Plant ............................................................................... 15

2-5 Area of Adrian Plant involved with AEC activity...................................................................... 15
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AWE atomic weapons employer

BZ breathing zone

CFR Code of Federal Regulations

cm centimeter

d day

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

g gram

GA general area

GM geometric mean

GSD geometric standard deviation

HASL Health and Safety Laboratory

hr hour

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

L liter

LOD limit of detection

m meter

MAC maximum allowable concentration

mg milligram

mm millimeter

mR milliroentgen

mrad millirad

mrem millirem

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

pCi picocurie

POC probability of causation

psi pounds per square inch

R roentgen
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RMI Reactive Metals, Inc.

SRDB Ref ID Site Research Database Reference Identification (number)

t ton

U.S.C. United States Code

yr year

α alpha particle

γ gamma

μCi microcurie

μg microgram

§ section or sections
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 7 of 45

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 sites or categories of sites. They will be revised in the event 

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

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

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

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

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility 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 Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) restrict the “performance of duty” 

referred to in 42 U S. C. § 7384n(b) to nuclear weapons work (NIOSH 2010a).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2010a):

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

conventional structures

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 8 of 45

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

reasons

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 sites or categories of sites. They will be revised in the event 

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

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

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

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

employer [AWE] facility” or a “Department of Energy [DOE] facility” as defined in the Energy 

Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. § 7384I(5) 

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

contract period and/or during the residual period.

Under EEOICPA, employment at an AWE facility is categorized as either (1) during the DOE contract 

period (i.e., when the AWE was processing or producing material that emitted radiation and was used 

in the production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination after the period 

in which weapons-related production occurred). For contract period employment, all occupationally 

derived radiation exposures at covered facilities must be included in dose reconstructions. This 

includes 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 covered period. NIOSH does not consider the following 

exposures to be occupationally derived (NIOSH 2010a):

• 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

For employment during the residual contamination period, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) [i.e., radiation doses received from DOE-related work] must be included in dose 

reconstructions. Doses from medical X-rays are not reconstructed during the residual contamination 

period (NIOSH 2010a). It should be noted that under subparagraph A of 42 U.S.C. § 7384n(c)(4), 

radiation associated with the Naval Nuclear Propulsion Program is specifically excluded from the 

employee’s radiation dose. This exclusion only applies to those AWE employees who worked during 

the residual contamination period. Also, under subparagraph B of 42 U.S.C. § 7384n(c)(4), radiation 

from a source not covered by subparagraph A that is not distinguishable through reliable 

documentation from radiation that is covered by subparagraph A is considered part of the employee’s 

radiation dose. This site profile covers only exposures resulting from nuclear weapons-related work. 

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

1.1 PURPOSE

This document provides an exposure matrix for two Bridgeport Brass Company sites: Havens 

Laboratory (including the Housatonic Pilot Plant) and the Adrian Plant.
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1.2 SCOPE

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

justification, or clarification of the associated information, are presented in Section 6.0.

2.0 SITE DESCRIPTION AND OPERATIONAL HISTORY

The Havens Laboratory and the Adrian Plant were geographically separate facilities, but the 

operations were similar. Havens Laboratory in Bridgeport, Connecticut, was involved primarily with 

uranium extrusion research and development. The other site, referred to in this document as the 

Adrian Plant, was in Adrian, Michigan. The Adrian Plant was also involved in uranium extrusion 

research and development but had a higher production capacity. Both sites processed thorium in 

addition to uranium.

2.1 HAVENS LABORATORY

The information that follows supports an assumed covered period of operations at Havens Laboratory 

and Housatonic Pilot Plant from November 8, 1950, through December 31, 1950, involving 

experimental uranium work, and from June 26, 1952, to August 27, 1962, involving uranium contract 

work. No exposure is assumed for the period from January 1, 1951, to June 25, 1952, which was the 

period between experimental activities and the effective date of the U.S. Atomic Energy Commission 

(AEC) contract. After decontamination in August 1962, the Havens Laboratory was sold and 

converted to a school. No residual radioactive contamination period is assumed for Havens 

Laboratory.

The Havens Laboratory radiological source term consisted primarily of natural uranium metal (a single 

data sheet mentioned depleted uranium), uranium oxides, and natural uranium’s short-lived progeny. 

Long-lived progeny prevent significant ingrowth past 234U in the 238U decay series and beyond 231Th in 

the 235U decay series. The source term also included smaller amounts of thorium.

Havens Laboratory was at Kossuth and Pulaski Streets in Bridgeport, Connecticut. The DOE site 

elimination report states, “the area under consideration consists of one building containing a 

laboratory and associated work areas” (Jones 1987). The Bridgeport Brass Company office was 

several blocks away at 30 Grand Street. An inspection report states that the Pilot Plant was also 

located at 30 Grand Street (Cavanaugh ca. 1954). Some claimants have noted other Bridgeport 

Brass locations as Housatonic Avenue or Grant Avenue. On September 15, 1952, Bethlehem Steel 

was requested to ship six rods to Bridgeport Brass Company, Housatonic Avenue (Belmore 1952). In 

June 1955, AEC surveyed the source and special nuclear controls at the Havens Laboratory and at 

the Housatonic Building (Dowling 1955a). Current maps indicate Housatonic Avenue intersects with 

Grand Street, but is physically separated from the Havens Laboratory by a body of water. Grant 

Avenue does not appear on current maps of Bridgeport, Connecticut.

Before the AEC contracted work at Havens Laboratory in 1952, AEC experimental work was reported 

on November 8 and December 11, 1950 (Klevin 1950; Stroke 1950). The November 8, 1950, 

experiment with uranium rod coating was deemed successful, and it was noted that no air samples 

were collected and, “no health and safety problems appear to exist for this type of operation” (Klevin 

1951). The December 11, 1950, experiment involved the cold drawing of hot rolled rods of uranium, 

pickling to remove the oxide coating, and the drawing of a few unpickled rods. For this December 11 

experimental work, the total number of processed rods was 12, involving up to three passes in the 

drawing process for each rod. Ten air samples were collected, but only the sample taken at the time 

of the pickled rod jam, which resulted in rapid oxidation despite a heavily coated lubricant, exceeded

the maximum allowable concentration (MAC) value (Klevin 1951). The MAC was defined as 

70 α-dpm/m3

. The sample was reported as “128 γ/m3” [the symbol, γ, was shorthand used by the 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 10 of 45

AEC for μg]. General area (GA) air samples taken before and after the drawing operations were 

reported as zero.

Some exposure could have occurred during the drawing operations. The amount of work and the 

limited time involved would clearly limit this exposure. An AEC monthly report states, “the drawing of 

uranium rods was observed and dust samples were collected during the drawing. Of ten dust 

samples, only one showed any uranium material. It is believed that this sample is not representative 

of the operation [because it was collected during the rod jam]. No contamination of the plant was 

found” (AEC 1951, p. 13). There are no other references to work involving uranium or other 

radioactive materials until the beginning of the contract between AEC and Bridgeport Brass in 1952. 

It is not clear what happened during the time between the two uranium rod experiments in 1950. In 

addition, it is not clear when the rods left the site, but a date of December 31, 1950, is assumed. The 

daily inhalation intake from November 8 through December 31, 1950, was unlikely to exceed the 

MAC. This site profile assumes that area contamination after this brief operation was limited and so 

only accounts for exposure during the assumed experimental period.

Havens Laboratory conducted laboratory-scale work under AEC contract AT(30-1)-1405, effective 

beginning June 26, 1952, for “research on drawing uranium and related operations” (AEC 1952). The 

contract specifically called for the contractor to:

...perform research work calculated to develop suitable and economic procedures for 

the following:

a. Alpha extrusion [alpha extrusion has to do with the temperature of the metal] of 

uranium

b. Extrusion of zirconium or zirconium alloy tubing

c. Mechanical cladding of uranium with aluminum, zirconium or zirconium alloy 

and the development of slug and closures

d. Investigation of other commercial procedures such as drawing, rolling, rocking, 

annealing, etc. at various temperatures pertinent to the above

e. The execution of such metallographic, thermocycling and X-ray crystallography 

in connection with the above work necessary to evaluate and control the 

products in regard to their suitability for pile operation and such additional work 

as the Commission may require.

The work included cold forming (extrusion) of natural uranium metal and associated cutting, storage, 

and laboratory support. From the beginning of the contract in June 1952, it is assumed that AEC work 

was full-time, although not all work involved radioactive material. This is apparent from the AEC 

portion of Havens Laboratory costs set from 1% to 5% with most AEC costs being allocated at 2% or 

less in the contract. The Plant Laboratory (Department 385-B), Technical Director’s Office 

(Department 385 A), Safety Services (Department 307-D), and Guard Service (Department 314) 

exceeded 2%.

The Housatonic Pilot Plant operation involved “the annealing, cold drawing and swaging of uranium 

and thorium rolled and extruded rods” (Dowling 1955a). The Havens Laboratory performed 

metallography and X-ray crystallography examinations on uranium and thorium samples from the pilot 

plant (Dowling 1955a).
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An inspection report (Cavanaugh ca. 1954) describes the laboratory as, “on the front portion of an old 

Trade School Building.” In addition, there was work with uranium on the first floor of the main 

building. In 1960, a fully equipped machine shop was installed, including a 500-t extrusion press. 

The areas involved in uranium activities are shown in Figures 2-1, 2-2, and 2-3. Because the work 

areas were physically separate, it was necessary to move the uranium from area to area, which 

affected internal and external exposure. For example, an air dust sample from February 20, 1962, 

and labeled “general BZ Billet transfer team” was 1,300 dpm/m3

. The relative physical size of the 

three areas is indicated by the figures, and the ventilated areas of the laboratory are shown.

Figure 2-1. Laboratory area at Havens Laboratory (Survey 1962a). 

A Bridgeport Brass monthly report from February or March of 1954 provides details of some of the 

work at Havens Laboratory (Treco 1954): Ten uranium rods of 1.405-in. average diameter and 20 ft. 

in length were received from Fernald. They were degreased with hydrex and pickled with a solution of 

50% water and 50% concentrated nitric acid. The rods were then cut into lengths of 77 in. (6 ft., 5 in.) 

each. The pickling, which removed the coating that reduced oxidation, and the cutting of the rods 

each contributed to increased air dust concentrations. These rods were then machine-pointed in a 

Paint 

Dryer

X-Ray

Diffract

Chem Hood

Polish Table

Bench

Offices Disc Polisher

Bench

Photomacrography

Room

Furnace

Vac 

Office
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Figure 2-2. First floor, main building, Havens Laboratory (Survey 1962b). 

Figure 2-3. Dynapak Area, Havens Laboratory (Survey 1962b). 

lathe to a diameter of 1.270 in. with a point length of 6 in. After this, the rods were drawn with a 

standard 1.375-in. standard brass rod-drawing die with a calculated area reduction of 4.2%. 

However, subsequent measurements indicated that springback had lowered this reduction to 3.5%. 

The pointing of the rods in the lathe was a source of airborne uranium, and the close work involved in 

measuring the rods provided potential exposure from the surface of the metal. Zirconium tube 

fabrication, extrusion of zirconium and its alloys, cans for cladding, and laboratory technical service 

sections of the report indicate that the uranium work was only part of the AEC activities.

Treco (1954) described the installation of a monorail system for handling heavy billets, which may 

have reduced doses by decreasing manual handling of the billets. In addition, ventilation for newly 

installed pickle and waste tanks was added. White duck (a cotton or linen fabric a bit lighter than 

canvas) coats and coveralls were procured for visitors and workers in the pilot plant. Rubber shoe 

covers were available for visitors.
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AEC and U.S. Department of Energy (DOE) inventory reports use a designation of BBA to identify 

Bridgeport Brass facilities. This designation could include Bridgeport Brass facilities other than 

Havens Laboratory. A designation of Station MBB also referred to Bridgeport Brass (Dowling 1955a).

Source and Special Nuclear Materials Accountability Statements (Dowling 1955a) indicate that 

Havens Lab received about 11,000 kg of uranium between July 1, 1953, and June 30, 1954, and 

received another 50,000 kg by May 31, 1955. During these same periods, Havens received 190 and 

1,570 kg of thorium, respectively, which is less than 5% of the uranium source term by mass. To 

simplify calculations for both Havens Laboratory and Adrian Plant and to account for uncertainty in the 

relative masses of handled uranium and thorium, this analysis assumed that the mass of processed 

thorium was 10% of the mass of processed uranium.

There is mention of X-ray crystallography work in the AEC contract to inspect metal samples, but no 

information about the X-ray machine design or safety precautions was found. Late 1958 to 1960 

biweekly film badge results for areas specified as “X-ray” were usually reported as <10 mrem. In 

addition, film badge records show monitoring of the hospital’s X-ray technologist. (One former 

employee reported that Bridgeport Brass had a small hospital on site.)

Reactive Metals, Inc. (RMI), formerly Bridgeport Brass, noted that it was not possible to obtain the 

names of hourly employees who participated in very early (1953 to 1954) uranium forming 

experiments in Bridgeport, Connecticut (Bean 1967).

In 1962, the Havens Laboratory AEC operation was moved to Seymour Specialty Wire, another 

Bridgeport site in Seymour, Connecticut (DOE ca. 1987). (Seymour Specialty Wire is a covered 

facility.) On August 27, 1962, Bridgeport reported that the cleanup of Havens Laboratory was 

complete and that decontamination was accomplished (Jefferson 1962). The Havens Laboratory was 

transferred to the local Catholic diocese for use as a school. It is not clear whether the Grand 

Avenue/Housatonic Street site was also transferred.

2.2 ADRIAN PLANT

The information that follows supports an assumed covered period of operations at Adrian Plant from 

July 25, 1950, to December 31, 1962. Radiation exposure from residual radioactivity at Adrian Plant 

is assumed to have occurred beginning January 1, 1963.

Like the Havens Laboratory, the Adrian Plant radiological source term also consisted primarily of 

uranium metal, uranium oxides, and their short-lived progeny. Adrian also processed thorium.

Before May 25, 1954, when Bridgeport Brass took over the facility at Adrian, there had been limited 

AEC work done there by previous contractors (the facility was also known as Air Force Plant No. 60 

and the Air Force Experimental Methods Plant). Subcontract No. 31-109-38-313 between GerityMichigan Manufacturing Corporation and Argonne National Laboratory was signed on December 8, 

1950, but was effective before then on July 25, 1950 (Wallo 1985). This was reportedly a services 

contract and was for extrusion work of aluminum alloys on an as-required basis. One memorandum 

suggests the contract may have been for extruding aluminum alloys. The contract terminated on 

March 15, 1951 (Wallo 1985). Negotiations with a new contractor (Reynolds Company) began, but no 

contractual agreement was reached. Mack Industries operated the plant under a maintenance 

agreement for the U.S. Air Force before Bridgeport Brass leased the facility. The DOE Office of 

Worker Advocacy website lists Gerity as a “Beryllium Vendor” and notes that Gerity began work in 

1949 for the AEC (Vogel 2004). In addition, Reynolds Metals Company and Mack Industries were not 

listed as covered facilities. Therefore, it is assumed that no AEC radiological work occurred at the site 

before the AEC-Bridgeport Brass contract modification noted in the following paragraph.
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On May 25, 1954, the Adrian Plant was added to AEC contract AT(30-1)-1405 via Modification 7 to 

work with uranium and thorium (Wallo 1985). This was followed by Modification 9 on February 18, 

1955, which specified research and development work to be done in extrusion development and 

drawing research at Adrian Plant and Havens Laboratory (AEC 1955a). The contract also called for 

Adrian Plant to be able to produce extruded rods or tubes at a semiproduction scale. Modification 9 

required the production of approximately 1,600 extruded rods between October 26, 1954, and April 1, 

1955.

Modification 9 noted that the “Extrusion Development work for Thorium ... shall be effective as of 

October 1, 1954” (AEC 1955a). A July 6, 1955, Bridgeport Brass memorandum indicates that 

because of a curtailment in both the uranium and thorium programs, Bridgeport needed to rethink the 

request for a replacement for the salt bath furnace pot to replace the existing uranium-contaminated 

pot to allow this salt bath to be used for “thorium extrusion as well as other items” (Stearns 1955). On 

July 22, 1955, curtailment of only the thorium program is mentioned in a letter from Bridgeport Brass 

Company to AEC (Treco 1955). On September 8, 1955, Bridgeport Brass stated, “there was no 

production of thorium to report for the month of August at Bridgeport Brass Company – MBA” 

(Schaeffer 1955a). (MBA is the AEC/DOE inventory report designation for the Adrian Plant.) 

On September 15, 1955, a letter from the AEC Feed Materials Division, reported that “approximately 

sixty-five billets of thorium have been delivered to the Adrian, Michigan Plant by National Lead Co. 

[Fernald] and that the extrusion of this metal has been scheduled for September 19 and 20, 1955” 

(Dowling 1955b). The same letter indicates that this work was a special order and was assigned as 

“Production Order No. 3.” In August 1955, thorium production was reported as “none” (Schaeffer 

1955a). In September 1955, 5 tons of thorium ingots were received, and 4 tons of thorium rods and 

1 ton of thorium scrap were produced (Schaeffer 1955b). A November 28, 1955, teletype between 

AEC offices notes that the “Thorium Operating Schedule” effectively meant that the Commission 

would not have facilities for converting thorium nitrate to thorium metal (Karl 1955).

While it is clear that thorium was processed at Adrian Plant, research found no thorium bioassay or 

contamination measurement data. In addition, the Adrian Plant thorium source term and production 

period information is limited. The records appear to indicate that the primary focus of the AEC Health 

and Safety Laboratory (HASL) was uranium. A limited review of other DOE and AWE sites seems to 

indicate that thorium was only a minor portion of the source term for most facilities. At Fernald, it was 

noted that thorium represented less than about 5% of the uranium-plus-thorium emissions (by mass) 

from plant processes. Also at Fernald, thorium emissions were reported in 1954 and 1955 but were 

not listed again until 1966. For 1954, the thorium emission percentage at Fernald was about 7% by 

mass. Fernald uranium and thorium workplace air sample results were similar in magnitude (ORAUT 

2004). No available information indicated that Adrian Plant thorium to uranium ratios would have 

been much different from the ratios at Fernald. Based on the reviewed thorium information, as well as 

the relative abundance of uranium information for Adrian Plant, it is assumed that thorium exposure 

could have occurred at Adrian Plant from May 25, 1954, to December 31, 1955, and that the thorium 

source term was not likely to be more than 10% of the uranium source term by mass.

Adrian Plant was located at 1450 East Beecher Street; it occupied several structures on 

approximately 73 acres of land. Only a small fraction of the overall facility was involved in AEC work. 

This consisted of approximately 44,500 ft2 in the main plant and about 2,000 ft2 of office space in the 

main plant. There were a loading dock area and a storage area outside of the main plant (Figure 2-4).

The metal extrusion, cutting, and other support activities were carried out in three bays of the main 

plant (Figure 2-5). The ceiling height varied from 45 to 55 ft. Crane rails, roof drain lines, electrical 

wires and conduits, water pipes, space heaters, and off-gas ducts were supported from a steel 

framework. Blowers were located on the roof for numerous off-gas ducts in the exhaust system.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 15 of 45

Figure 2-4. Locations of AEC activity at Adrian Plant (Haywood et al. Undated). 

Figure 2-5. Area of Adrian Plant involved with AEC activity

(area marked with diagonal lines) (Hill 1976). 

Summary statements of activities at Adrian Plant indicate that depleted, normal, and low-enriched 

uranium were processed. Before 1960, there is no indication that enriched material was processed at 

Adrian Plant. On May 12, 1960, the AEC HASL noted, “The company is planning to extrude, in the 

near future, uranium of 2% enrichment” (AEC 1960). A ventilation system was partially installed in 

Press #7 by the time of the dust survey on December 14, 1960. On December 14, 1960, what appear 

to be the first low-enrichment (0.947%) billets at Adrian Plant were extruded (AEC 1961). A summary 

of the flow of recycled uranium indicates that Adrian Plant shipped and received depleted, normal, 

Loading

Dock

Sp-1

Sp-2

Office

2Δ

3∆

Uranium Extrusion

Operating Area

.S

*
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 16 of 45

and low-enriched (0.947%) uranium to and from Fernald from October 1, 1961, to September 30, 

1962, and the low-enriched uranium mass was more than a 2 times larger than either the normal or 

the depleted mass (DOE 2000) during that period. Because Adrian Plant could have received 

uranium from other facilities and because there is just 1 year of data available, enrichment is assumed 

to be the largest of the reported values, which was 2%.

A wipe test analyzed for 60Co indicates that there was a sealed source in use at Adrian. The source 

was used to check the criticality alarm monitor.

RMI noted that before approximately June 1955 the operations were sporadic and no regular hourly 

press crew was assigned to the AEC work (Bean 1967). Crews were commonly assembled from 

available staff, resulting in many employees being involved in the AEC work.

AEC operations were coming to an end at Adrian Plant in 1961. The operations were relocated to 

Extrusion Plant (Reactive Metals, Inc.), another Bridgeport Brass Company relation in Ashtabula, 

Ohio. Decontamination and close-out work were completed in 1962. The Adrian Plant was still 

named in Contract No. AT-(30-1)-1405, Modification No. 37, which was entered into on November 29, 

1961, for the period from July 1, 1961, to June 30, 1962. In 1961, the Bridgeport Brass Ashtabula 

facility (also known as the Extrusion Plant or Reactive Metals, Inc.) began operations. Employee 

claim information indicates that a number of Adrian Plant employees transferred to Ashtabula and 

bioassay records confirm this beginning in early 1962. A copy of a folder or a divider includes 

Bridgeport Brass Company and the Adrian address and states, “No samples received after 10/5/1961” 

(Author unknown, undated). The Adrian Plant AEC contract work was officially terminated on August 

30, 1963, by Modification No. 42 or 47 [both numbers are cited] (Wallo 1985). This site profile 

assumes that operations might have continued through December 31, 1962, at the Adrian Plant and 

that exposure was due to residual radioactivity after that. Decontamination and survey activities were 

reported in 1976, 1985, and 1995.

3.0 ESTIMATION OF INTERNAL EXPOSURE

The primary source of internal exposure at Havens Laboratory and Adrian Plant was radioactive dust 

from the handling and oxidation of uranium and smaller amounts of thorium during the various 

processes at the facilities. The specific sources of uranium dust were described in what appears to 

be the second of a series of evaluations of occupational exposure at Adrian Plant (AEC 1955b). This 

report outlines the main factors contributing to exposure to dust:

• Fumes and smoke from extrusion press discharge,

• Oxidation of extruded rod surface to oxide that subsequently flaked off on rod movement and 

became airborne,

• Transfer of rod from run-off table to cart due to dust created when loose oxide from rod 

surface was rendered airborne,

• High concentrations at run-off table area due to presence of loose oxide scale on surface of 

rod and rollers, 

• Storage of hot crops in an open area near a tool heating furnace,

• Cutting off the butt ends of extruded rods and from deburring of the die block,

• High GA air concentrations in the vicinity of straighteners due to flaking of loose oxide scale on 

surface of extruded rod on stretching, and
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 17 of 45

• High GA air concentrations at the extrusion exhaust, operators area, run-off table area, and 

crop cooling areas (AEC 1955b).

Individual uranium urinalysis data are available from both Havens Laboratory and Adrian Plant. For 

unmonitored workers or unmonitored periods, an analysis of air monitoring data is provided for use in 

reconstructing internal dose. This document estimates thorium intakes based on uranium intakes.

3.1 URANIUM

Human and animal studies have indicated that oxides of uranium can be very insoluble in the lung 

(ICRP 1995), which indicates absorption type S (0.1% and 99.9% with clearance half-times on the 

order of 10 minutes and 7,000 days, respectively). Other in vitro dissolution studies of compounds at 

uranium facilities have shown that oxides of uranium exhibit moderate solubility (Eidson 1994; 

Heffernan et al. 2001), which suggests absorption type M (10% and 90% with clearance half-times on 

the order of 10 minutes and 140 days, respectively). In vitro dissolution tests on oxides from uranium 

metal during depleted uranium armor penetrator tests have indicated multicomponent dissolution 

rates, with 25% of uranium dissolving with a half-time of less than or equal to 0.14 days and 75% 

dissolving with a half-time of 180 days. Because there was no specific information on the solubility of 

aerosols at Havens Laboratory and Adrian Plant, this analysis assumed that both types M and S were 

available. The selection of absorption type should depend on the organ of interest. Dose 

reconstructions should assume International Commission on Radiological Protection (ICRP) 

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

Havens Laboratory handled uranium of natural enrichment. At Adrian Plant, there is no indication that 

uranium enrichment ever exceeded about 2%, and the records seem to indicate that natural uranium 

was typical. It is favorable to claimants to assume that uranium results reported in mass were 2% 

enriched. After 1952, recycled uranium might have been handled.

3.1.1 Uranium Bioassay

Individual uranium urinalysis results are available for some Havens Laboratory and Adrian Plant 

workers during some periods. Urine samples were not collected from all Bridgeport Brass uranium 

workers, so the lack of bioassay for an individual should not result in a conclusion of no internal 

exposure.

The uranium fusion photofluorimetry urinalyses by the University of Rochester and the AEC New York 

Operations Office were similar to those at other AEC facilities. The default detection threshold for 

uranium urinalysis is assumed to be 10 μg/L based on a reported sensitivity of 5 to 10 μg/L for 

uranium fluorimetry urinalysis in the early years (Wilson 1959). Because bioassays were analyzed by 

photofluorimetry, which is a uranium mass detection method, results at Adrian should be modified to 

account for an enriched source term.

For unmonitored workers or unmonitored periods, this site profile analyzes the bioassay results to 

provide estimates of uranium intakes based on coworker data. 

The first available bioassay samples for Havens Laboratory were dated September 10, 1952: 

urinalyses approached an annual frequency before 1958, when the frequency was increased, 

although no one frequency could be determined from the data. The last available set of sample 

results was reported for March 12, 1962. Samples were collected sometimes in the mornings and 

sometimes in the afternoons, and this is noted on some records. For some sampling periods, the 

analytical laboratory noted that hydrochloric acid was added to some of the containers, but it is not 

clear what the purpose for this was. An incident appears to have occurred in April 1961, based on 

rushed air sample requests and bioassays repeated within a day. For the purpose of coworker intake 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 18 of 45

determinations, it was assumed that the elevated intakes from the incident began on April 15, 1961, 

1 day before the first rushed air sample on a Saturday, and continued through April 21, 1961, the day 

after the first bioassay. In addition, a chronic intake was assumed from June 26, 1952, through 

August 27, 1962.

The first available bioassay samples for Adrian Plant were dated December 14, 1954, and were listed 

as “before operating” results. A few workers had multiple samples in 1955; a few samples were 

collected in 1956. No results were found for 1957, and sampling appeared to be annual in 1958 and 

1959. Beginning in July 1960, sampling appears to have approached a monthly frequency. The last 

available set of sample results was reported for October 9, 1961. Samples were collected sometimes 

in the mornings and sometimes in the afternoons, and this is noted on some records. For some 

sampling periods, the analytical laboratory noted that hydrochloric acid was added to some of the 

containers, but it is not clear what the purpose for this was. Uranium urinalyses appeared to be 

elevated beginning in late 1960 and early 1961 and started to fall again after April 1961. This analysis 

assumes a chronic intake at Adrian Plant from May 24, 1954, through December 31, 1962. An 

additional intake from October 1, 1960, through April 11, 1961 to account for the later elevated 

urinalysis data was assumed.

For each bioassay date, geometric means (GMs) were estimated by ranking the data, determining the 

z-scores, and plotting the respective z-score versus the natural log of the data. A line was fit to the 

data; e raised to the line’s y-intercept value was assumed to be the GM, and e raised to the slope 

value was assumed to be the geometric standard deviation (GSD) of the data. Results that were 

reported as zero were ranked but used only indirectly in the fitting of the line. The 84th percentile was 

estimated as the GM multiplied by the GSD. The number of results for a given date ranged from 1 to 

13 at Havens Laboratory and from 1 to 38 at Adrian Plant. The statistical fit parameter (R2

) results 

ranged from 0.53 (three results) to 0.97 at Havens, from 0.63 to 0.98 at Adrian, and were considered 

adequate for this set of data.

The daily uranium excretion in urine was calculated by multiplying the results in milligrams per liter by 

reference man’s daily urine output (1.4 L/d) (ICRP 1975). Because Adrian used slightly enriched 

uranium, those results were multiplied by 1,616 pCi/mg to obtain a daily uranium concentration in 

picocuries per day. The Havens uranium was assumed to be unenriched, and a factor of 683 pCi/mg 

was used to convert to activity per day. Attachment A shows the bioassay results that were used in 

the intake analyses. Table 3-1 shows a summary of the estimated GM, 84th-percentile, and 

maximum uranium urinalyses that were used to derive intakes from the chronic inhalation intake 

regimes. Graphs showing the fits of these intake regimes are provided in Attachment A. Additional 

intakes and alternate periods were tried, but fits were not more satisfactory than those chosen. When 

intakes are estimated from bioassay data, the mode of intake should usually be assumed to be 

inhalation, unless there is information that indicates that other modes of intake were more likely. 

When using bioassay data, the inhalation intake model assumes that some of the intake behaves as 

ingested material. In general, intakes from bioassay will be larger when an inhalation rather than an 

ingestion intake is assumed.

The intakes were determined with the Integrated Modules for Bioassay Analysis (IMBA) assuming an 

absolute uniform error of 1 and normal error distributions for each bioassay result. Data noted as less 

than the limit of detection (<LOD) is shown on the graph but was not used directly in the fitting. The 

GSDs for the intakes were calculated by dividing the intake from the 84th-percentile regime by the 

intake from GM intake regime. Table 3-2 shows the inhalation intake distributions from the analyses 

of the Havens Laboratory and Adrian Plant uranium urinalysis data, assuming that either a type M or a 

type S (but not both) intake occurred. Two intake scenarios are shown for each site. One scenario 

excludes some of the elevated data and assumes that workers were exposed chronically throughout 

their employment: this scenario should be used only for workers whose employment period does not 

overlap the period of assumed elevated exposure. For workers whose employment coincided with the 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 19 of 45

period of elevated exposure, the appropriate site two-intake scenario should be assumed if there are 

no available bioassay data.

To estimate intakes for the purpose of dose reconstruction, the following approach has been 

prescribed: the GM of the intake is adjusted to the 95th percentile using the larger of a GSD of 3 or 

the GSD of the distribution itself, and this 95th percentile is assigned as a constant distribution in the 

Interactive RadioEpidemiological Program (IREP). This document adjusts the intakes in Table 3-2 to 

95th-percentile intakes by multiplying the GMs by a GSD of 3 raised to the 1.645 power. The 

resulting intakes are summarized in Table 3-8. 

For Havens Laboratory an additional intake is estimated for 1950 based on air sampling data.

3.1.2 Uranium Air Sampling

Air samples were collected at both Havens Laboratory and Adrian Plant. Because samples were 

measured by alpha activity detection systems, enrichment did not affect the results.

Table 3-1. Uranium urinalysis results from coworker data.

a 

Havens Laboratory Adrian Plant

Date # GM 

(pCi/d)

84th 

(pCi/d)

GM 

(mg/L)

84th 

(mg/L) Date # GM (pCi/d)

84th 

(pCi/d)

GM 

(mg/L)

84th 

(mg/L)

9/10/1952 5 2.07 4.83 0.002 0.005 <LOD 12/14/1954 7 <22.62 <22.62 <0.01 <0.01 <LOD

9/11/1952 1 3.82 3.82 0.004 0.004 <LOD 1/12/1955 6 <22.62 <22.62 <0.01 <0.01 <LOD

9/1/1953 5 1.1 1.38 0.001 0.001 <LOD 4/6/1955 5 <22.62 <22.62 <0.01 <0.01 <LOD

10/1/1953 2 0.96 0.96 0.001 0.001 <LOD 8/3/1956 4 <22.62 <22.62 <0.01 <0.01 <LOD

2/4/1954 3 3.13 8.55 0.003 0.009 <LOD 8/22/1958 10 10 21.28 0.004 0.009

6/1/1954 8 1.94 3.35 0.002 0.004 <LOD 10/10/1958 1 <22.62 <22.62 <0.01 <0.01 <LOD

2/27/1956 6 2.81 6.59 0.003 0.007 10/16/1959 18 3.43 18.53 0.002 0.008

10/7/1957 7 8.49 19.85 0.009 0.021 11/13/1959 15 <22.62 <22.62 <0.01 <0.01 <LOD

1/27/1958 8 3.6 8.77 0.004 0.009 7/14/1960 16 14.5 30.74 0.006 0.014

7/3/1958 2 2.34 3.16 0.002 0.003 <LOD 8/23/1960 11 14.69 42.4 0.006 0.019

9/15/1958 8 1.93 3.49 0.002 0.004 <LOD 9/6/1960 10 14.99 45 0.007 0.02

1/19/1959 5 2.45 7.27 0.003 0.008 9/12/1960 10 24.11 91.22 0.011 0.04

2/25/1959 4 1.84 6.42 0.002 0.007 <LOD 9/19/1960 11 12.13 31.67 0.005 0.014

3/16/1959 3 2.41 10.1 0.003 0.011 9/22/1960 18 6.21 18.13 0.003 0.008

8/28/1959 0 0 0 0 0 <LOD 10/14/1960 9 10.59 19.91 0.005 0.009

10/19/1959 4 0.61 1.59 0.001 0.002 11/18/1960 8 12.84 25.38 0.006 0.011

2/5/1960 0 0 0 0 0 <LOD 12/19/1960 12 7.45 22.56 0.003 0.01

5/26/1960 5 3.14 4.99 0.003 0.005 <LOD 1/13/1961 19 8.63 21.21 0.004 0.009

10/26/1960 7 17.26 51.38 0.018 0.054 1/23/1961 15 6.33 15.62 0.003 0.007

1/9/1961 8 2.23 4.51 0.002 0.005 <LOD 1/30/1961 16 13.29 27.41 0.006 0.012

4/1/1961 3 5.59 7.13 0.006 0.007 <LOD 2/10/1961 15 51.24 126.3 0.023 0.056

4/20/1961 10 61.23 228.78 0.064 0.239 2/28/1961 14 42.27 63.51 0.019 0.028

4/21/1961 9 12.59 44.29 0.013 0.046 3/13/1961 19 43.18 61.34 0.019 0.027

4/24/1961 8 9.4 18.03 0.01 0.019 3/27/1961 28 38.45 56.46 0.017 0.025

5/22/1961 7 16.29 35.44 0.017 0.037 4/10/1961 32 26.55 105.79 0.012 0.047

5/26/1961 1 13.39 13.39 0.014 0.014 4/24/1961 35 6.02 13.96 0.003 0.006

9/25/1961 6 0.72 2.68 0.001 0.003 <LOD 5/8/1961 37 11.61 21.31 0.005 0.009

9/26/1961 7 1.1 3.55 0.001 0.004 <LOD 5/22/1961 38 5.96 14.43 0.003 0.006

3/3/1962 13 2.32 5.2 0.002 0.005 6/5/1961 37 6.63 16.26 0.003 0.007

3/9/1962 13 2.07 4.37 0.002 0.005 <LOD 6/26/1961 19 26.63 54 0.012 0.024

3/12/1962 11 1.3 2.05 0.001 0.002 <LOD 7/25/1961 24 16.49 30.74 0.007 0.014

7/28/1961 30 21.87 68.56 0.01 0.03

7/31/1961 16 4.96 18.22 0.002 0.008

10/2/1961 35 4.72 16.09 0.002 0.007

10/6/1961 32 7.43 33.95 0.003 0.015

10/9/1961 35 4.24 11.41 0.002 0.005

a. Assumes 1.4 L/d and 683 pCi/mg for Havens and 1,616 pCi/mg for Adrian. 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 20 of 45

Table 3-2. Inhalation intakes based on coworker data. 

Havens coworker intake scenario choices

Scenario Start End

Type M Type S

GM 

(pCi/d) GSD

GM 

(pCi/d) GSD

Chronic or 06/26/1952 08/27/1962 1.03E+02 2.29 1.43E+03 2.28

Chronic plus incident 06/26/1952 08/27/1962 9.59E+01 2.14 1.32E+03 2.16

04/15/1961 04/21/1961 1.13E+03 3.62E+04

Adrian coworker intake scenario choices

Scenario Start End

Type M Type S

GM 

(pCi/d) GSD

GM 

(pCi/d) GSD

One chronic or 05/24/1954 12/31/1962 1.69E+02 2.76 2.31E+03 2.76

Two chronics 05/24/1954 12/31/1962 1.32E+02 2.96 1.62E+03 2.87

10/01/1960 04/11/1961 3.03E+02 9.79E+03

For the assumed experimental uranium work period at Havens Laboratory from November 8, 1950, 

through December 31, 1950, the daily intake was derived by assuming that the median air 

concentration was equal to the MAC of 70 dpm/m3

. This is likely to overestimate actual exposures, 

but is believed to account adequately for any uranium exposure that might have occurred during this 

2-month period based on the air sample results noted in Section 2.1. The MAC was chosen to 

estimate this intake based on the fact that 10 air samples were collected on one uranium workday and 

it was noted that the average of the 10 samples was essentially zero (Klevin 1951). One sample was 

about 2.5 times the MAC. Because the uranium operations were believed to be very limited and 

intermittent during this period both in terms of time and source material, one MAC is used to provide 

an upper estimate of a chronic intake for this 2-month period. The daily inhalation intake rate can be 

determined by multiplying the air concentration by the breathing rate per work year (2,400 m3

/yr), 

adjusting from disintegrations per minute to picocuries, and dividing by the number of calendar days in 

a year. The result is 207 pCi/d.

Adrian Plant work was evaluated for air dust levels near the very beginning of operations. Data from 

these evaluations were documented in a series of HASL documents beginning with HASL-B-BRA-2 in 

March 1955 (AEC 1955b), the first available document in the series. This 1955 report consisted of 

three dust studies performed October 26 to 27, 1954; December 14 to 16, 1954; and January 11 to 

13, 1955. Those studies were followed by air sample studies on February 16 and 24, 1956 (AEC 

1956); January 27 to 29, 1960 (AEC 1960); and December 14, 1960 (AEC 1961). The purposes of 

these studies were:

• To evaluate and document occupational exposure to alpha-emitting dust,

• To identify sources of uranium air contamination and recommend corrective actions where 

necessary, and

• To appraise the effectiveness of the ventilation system and existing control measures.

The air samples included radioactive particulate concentration measurements of breathing zones 

(BZs), GA, processes, and effluents. The air sample results were matched with information about 

worker categories, locations, tasks, and the time at each location or task. Daily weighted-average 

exposures were then determined for job categories.

The April 1956 report (AEC 1956) noted progress in exposure control since the March 1955 report 

(AEC 1955b). Progress included installation of ventilation equipment at the extrusion press discharge 

and crop waste discharge areas, which effectively reduced airborne uranium concentrations in these 

areas. “For example, process air concentrations of 343 dpm/m3

, 1200 dpm/m3, and 128 dpm/m3 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 21 of 45

obtained during the survey of January, 1955, at the extrusion press discharge, die head operator 

position and 6 feet south of the extrusion press, respectively, were reduced to 80 dpm/m3

, 

107 dpm/m3

, and 18 dpm/m3 during the present survey” (AEC 1956).

Average BZ concentrations of 7,600 dpm/m3 obtained during diehead operations of January 1955 

were reduced to 95 dpm/m3

. These improvements were the result of mechanical changes in the 

process as well as such changes in the process components as the use of salt as a dust suppressant. 

In addition, quench water was used at varying flow rates to reduce dust with results somewhat 

dependent on the type of extrusion.

These safety gains were offset by unimplemented recommendations and facility modifications from 

the March 1955 report (AEC 1955b): 

2. Use better housekeeping techniques to prevent the accumulation of loose oxide on 

the run-off conveyers, table and general operating area. These areas should be 

thoroughly vacuumed at the end of each 12-ingot heat if possible. Eliminate broom 

sweeping.

3. Provide dust catch pans underneath all roller conveyers to reduce the spread of 

contamination and maintenance time.

4. Unless the floors can be kept clean, provide steel matting on the operating floor 

areas to reduce the spread of both airborne and direct surface contamination.

These recommendations were reiterated by reference in the April 1956 report (AEC 1956).

The significance of broom sweeping is that it was an unauthorized activity noted as contributing to 

elevated dust concentrations. However, elevated levels from sweeping were unlikely to have been 

directly included in the HASL surveys of airborne radioactivity.

The effect of not implementing recommendations 2 and 3 was noted (AEC 1956):

Visible loose oxide concentration found was on the discharge conveyor after the 

passage of each extruded rod. Average air concentrations obtained in the conveyor 

area during extrusion 25 ft and 50 ft south of the extrusion press discharge were 368 

dpm/m3 and 463 dpm/m3 respectively. Alpha radiation measurements made in this 

area with a Juno Survey Meter (SIC 17C) showed surface contamination of 

15,000-40,000 α-dpm/100 cm2

 ... 

and

All the uncontrolled sources of dust which have been pointed out contributed to raising 

the general air concentration throughout the plant including such areas as the 

operating and tool heating sites. Excessive air concentrations found in these areas 

were primarily responsible for the undesirable exposures reported for the Extrusion 

Press Operator, Die Head Man, Foreman and Project Engineer.

The 1960 HASL report (AEC 1960) again restated the need to implement Recommendations 2 and 3 

from the March 1955 report, so it appears that these recommendations had not been implemented 

after 5 years.

There are some air sampling data for Havens Laboratory, and the results seem generally consistent 

with the Adrian Plant results. However, the data are not organized into weighted daily averages by 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 22 of 45

job category. Because the Havens Laboratory and Adrian Plant activities were very similar, it seemed 

reasonable to apply the Adrian uranium air sampling data to estimations of employee exposures at 

both Adrian Plant and Havens Laboratory.

Air dust measurements are very affected by location, ongoing processes and their variations (salt, no 

salt, quench rate, as well as type of extrusion), and ventilation. Table 3-3 shows the process-effect of 

applying salt coatings and quenching uranium.

Table 3-3. Comparison of daily weighted-average air dust measurements at 

Adrian Plant (dpm/m3

) for various quenching and salting strategies.

a

Individual

2-psi quench

tube extrusion

Salting

plate extrusion, no quench17.5-psi quenchtube extrusion

Die head man 2,000 1.6 131

a. Source: AEC (1956).

A report from Adrian Plant (AEC 1960) states that before November 1957 the use of salt to coat hot 

uranium before and after extrusion was effective in suppressing air dust concentrations. When salt 

coatings were used, no average time-weighted dust concentrations in excess of the MAC were 

reported. Sometime after November 1957, it was determined that the use of salt coatings induced 

undesirable properties in the finished product. Therefore, this method of dust suppression was 

abandoned. Measurement of air dust concentrations after salt was no longer used indicated that 8 of 

the 17 studied employees were exposed to average time-weighted dust concentrations in excess of 

the MAC. It was clear that the use of salt to coat hot uranium reduced air dust concentrations. The 

role of salt in the extrusion process at Havens Laboratory is unclear. Use of salt coatings and quench 

water in the extrusion process could explain the measurement fluctuations of air dust concentrations.

An example of air concentrations is shown in Table 3-4 where air dust levels had previously been 

reduced through facility and process modification. The data are daily weighted averages 

(incorporating time spent in each activity as well as time spent away from the workplace, such as at 

lunch or in the washroom) rather than raw air dust values.

Table 3-4. Adrian Plant daily weighted-average 

air dust concentrations by operator position 

measured January 11 to 13, 1955. 

Job category

Concentration

(dpm/m3)

Salt bath operator 54

Salt bath helper 51

Extrusion press operator 97

Die head operator 2,000

Run out and winch operator 28

Weigh man 120

Lube man and winch operator 280

Foreman 83

Project engineer 87

The GM of the Table 3-4 daily weighted averages is 115 dpm/m3, and the GSD is 3.5. The daily 

weighted average from 1955 can be compared to similarly measured daily weighted-average 

exposures from HASL-85 (AEC 1960) in Table 3-5.

The GM of the Table 3-5 individual daily weighted-average air concentration is 250 dpm/m3

, and the 

GSD of the exposures is 1.9 (there are 17 concentrations associated with the workers). However, the 

GSD of the daily weighted-average concentration (there are nine concentrations) is 2.2.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 23 of 45

Table 3-5. Adrian Plant daily weighted-average air dust 

concentrations from January 27 to 29, 1960. 

Operator

Number of 

workers

Average daily weighted 

exposures (dpm/m3)

Heater 2 200

1st general helper 2 200

2nd general helper 2 200

Press operator 2 420

Die head man 2 470

Finished saw man 2 330

Inspector 2 310

Foreman 2 270

Machinist 1 32

Total 17

Table 3-5 has the largest GM of the reported daily weighted-average uranium air concentrations at 

Adrian, and its GM was used to calculate intakes for unmonitored Havens Laboratory and Adrian 

Plant employees beginning in 1952.

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

to be considered. NIOSH (2004a) states that the daily ingestion rate in picocuries can be estimated 

by multiplying the daily air concentration in picocuries per cubic meter by a factor of 0.2, so the daily 

uranium ingestion rate based on an air concentration of 250 dpm/m3 would be 23 pCi/workday. The 

ingestion rate associated with an air concentration of 70 dpm/m3 would be 6.3 pCi/workday.

A summary of estimated uranium intake rates based on air concentrations is shown in Table 3-6. 

Although these intakes were based on an upper estimate of air exposure (constant distribution), the 

metabolic models were assumed to have an uncertainty associated with a lognormal distribution and 

a GSD of 3. The Table 3-6 intakes for the period November 8, 1950, to December 31, 1950, are 

adjusted to 95th-percentile intakes for the purpose of dose reconstruction by multiplying by a factor of 

3 raised to the 1.645 power. The resulting values are summarized in Table 3-8 later in this section. 

Table 3-6. Estimated uranium intake rates based on time-weighted air concentrations.

Start End

Intake 

mode Radionuclide

Absorption 

type

Exposure rate

(pCi/d)

Havens

Laboratory

11/08/1950 12/31/1950 Inhalation U-natural M, S 2.07E+02

11/08/1950 12/31/1950 Ingestion U-natural (a) 4.32E+00

06/26/1952 08/27/1962 Inhalation U-natural M, S 7.40E+02

06/26/1952 08/27/1962 Ingestion U-natural (a) 1.54E+01

Adrian Plant 05/25/1954 12/31/1962 Inhalation U-234 M, S 7.40E+02

05/25/1954 12/31/1962 Ingestion U-234 (a) 1.54E+01

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

3.1.3 Comparison of Uranium Bioassay and Air Concentration Estimates

Except for 1950, summary estimates of uranium intakes shown in Section 3.4 are based on 

Bridgeport Brass bioassay data.

Estimates of total intakes derived from urinalysis data and air concentrations appear to be similar. 

Differences in the values of intake estimates from air and bioassay data are likely due to a multitude 

of factors, but one of the more significant factors is the choice of uranium absorption type. For 

interpretation of both the air and bioassay data, intake pattern assumptions were simplified based on 

the limited information. If the time patterns of intake are assumed reasonable, it appears reasonable 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 24 of 45

to conclude that workers were not exposed to a source term that was clearly pure type M or pure 

type S.

3.1.4 Recycled Uranium

Recycled uranium might have been processed at Havens Laboratory and Adrian Plant after 1952. An 

estimate of contaminants that might contribute the most to internal doses, based on Site Profiles for 

Atomic Weapons Employers that Worked Uranium Metals (NIOSH 2011), is shown below in Table 3-

7. It is unlikely that recycled uranium would constitute the entire source term. In addition, the activity 

fractions assume that the uranium specific activity is based on depleted uranium, which increases the 

proportion of the contaminants by activity. The contaminant levels for depleted uranium overestimate 

the contaminants in uranium of normal enrichment by about 40%. The contaminants are assumed to 

be oxides. Plutonium oxides are assumed to be type M or S. All chemical forms of neptunium are 

assumed to be type M.

Table 3-7. Estimate of contaminant activity fractions in a recycled uranium source term (pCi 

contaminant per pCi uranium). 

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

1 0.00246 0.00182 0.379 2.73E-06 2.73E-06

3.2 THORIUM

It is not clear when thorium work started at Havens Laboratory. The Adrian Plant records indicate that 

extrusion of thorium most likely started on May 25, 1954. Records indicate that thorium work might 

have slowed down or ceased after 1955, but no inventory records were found after this date. AEC 

records indicate that there was continuing interest in thorium after 1955. It is favorable to claimants to 

assume that thorium processing continued throughout the AEC work periods. To date, no records of 

thorium air monitoring or bioassay for either site have been found. 

To account for unmonitored thorium exposures at Havens Laboratory and Adrian Plant, it is assumed 

that the thorium intake is equal to 10% of the uranium intake by mass for the same period. Natural 

uranium has a lower specific activity than enriched uranium, so it is favorable to claimants to assume 

natural uranium when determining the relative activity of thorium. To determine the relative activities 

of uranium to thorium, the specific activity of 232Th is divided by the specific activity of natural uranium 

and multiplied by 10%. This results in a relative 232Th -to-uranium intake fraction by activity of 0.0161. 

Further, it is assumed that 232Th is in equilibrium with 228Th, so the 228Th to uranium activity fraction is 

also 0.0161. Exposure from 228Ra (half-life of 5.75 years) is assumed to be insignificant because the 

thorium was likely to have been recently produced and because the dose conversion factor is small 

compared to thorium. Thorium intakes are summarized in Table 3-8 below.

Table 3-8. Estimated thorium intake rates based on uranium intakes. 

Start End

Intake 

mode Radionuclide

Absorption 

type

Exposure rate

(pCi/d)

Havens

Laboratory

11/08/1950 12/31/1950 Inhalation Th-232 & Th-228 M, S 3.33E+00

11/08/1950 12/31/1950 Ingestion Th-232 & Th-228 (a) 6.96E-02

06/26/1952 08/27/1962 Inhalation Th-232 & Th-228 M, S 1.19E+01

06/26/1952 08/27/1962 Ingestion Th-232 & Th-228 (a) 2.48E-01

Adrian 

Plant

05/25/1954 12/31/1962 Inhalation Th-232 & Th-228 M, S 1.19E+01

05/25/1954 12/31/1962 Ingestion Th-232 & Th-228 (a) 2.48E-01

a. Choose same f1-value as used for inhalation per NIOSH (2004a).
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 25 of 45

3.3 MISCELLANEOUS INFORMATION RELATED TO INTERNAL DOSE

This section includes internal dose information that could be of value for specific dose reconstructions. 

This analysis did consider the information generically, but it should also be considered in dose 

reconstructions based on individual dosimetry analysis.

Air samples labeled “rush” were dated April 16 and 17, 1961, for Havens Laboratory. These 

combined with the collection of multiple urine bioassay samples from workers on April 20 and 21, 

1961, are indicative of an incident that may have resulted in internal exposures.

3.4 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

The assumed uranium photofluorimetry detection threshold is 10 μg/L. Intakes of neptunium, 

plutonium, and thorium are derived from intakes of uranium. The solubility types for uranium and its 

contaminants should be determined per guidance in the Internal Dose Reconstruction Technical

Information Bulletin (ORAUT 2007).

At Havens Laboratory, intakes are assumed to be from natural uranium (0.683 pCi/μg). Thorium 

intakes should be assumed beginning June 26, 1952. Recycled uranium contaminants are included 

after 1952. There are two periods of operational exposure:

• November 8, 1950, to December 31, 1950; and

• June 26, 1952, to August 27, 1962. 

The 1950 intake is assigned to all Havens Laboratory workers, whose covered work period 

overlapped the 1950 intake period. For unmonitored work periods or workers, two exposure 

scenarios for Havens are shown in Table 3-9. The first can be used for unmonitored internal 

exposures that did not include work during the period from April 15 to 21, 1961. The second scenario 

accounts for the higher exposures during the period from April 15 to 21, 1961, and should be used for 

unmonitored exposures that overlapped this period.

At Adrian Plant, intakes are likely to be from natural (0.683 pCi/μg) or low-enriched (0.973 or 1.616 

pCi/μg) uranium and thorium. Only natural uranium enrichment is likely before 1960. The default 

enrichment assumption for 1960 through 1962 is 2% (1.616 pCi/μg).

There is one period of operational exposure: May 25, 1954, to December 31, 1962.

At Adrian Plant, thorium and recycled uranium contaminant intakes should be assumed for the entire 

operational exposure period.

For unmonitored work periods or workers, two exposure scenarios for Havens are shown in Table 3-9. 

The first can be used for unmonitored internal exposures that did not include work during the period 

from October 14, 1960, to April 11, 1961. The second scenario accounts for the higher exposures 

during the period from October 14, 1960, to April 11, 1961, and should be used for unmonitored 

exposures that overlapped this period.

Table 3-7 can be used for both Havens Laboratory and Adrian Plant to estimate intakes of the other 

radionuclides from intakes of uranium, such as when intakes are derived from uranium bioassay or 

from uranium air concentrations after 1952. 

For unmonitored workers or unmonitored periods, Table 3-9 lists uranium intake rate assumptions. 

The intakes are assumed to be chronic. The dose distributions are assumed to be constant.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 26 of 45

Table 3-9. Chronic uranium intake assumptions for unmonitored workers. 

Assumptions for 

unmonitored workers Scenarios Start End

Intake 

mode

Intake 

type

Radionucl

ide Type

Uranium type Md

(pCi/d)

Uranium type Sd

(pCi/d)

Havens Laboratory

Choose one of the two 

bordered scenarios. Choose 

the second scenario if work 

overlapped the period from 

April 15 to 22, 1961.

Choose intakes based on 

either type M or S uranium, 

not both.

1 

11/08/1950 12/31/1950 Inhalation Chronic U-234 (a) 1.26E+03 1.26E+03

11/08/1950 12/31/1950 Ingestion Chronic U-234 (b) 2.63E+01 2.63E+01

06/26/1952 08/27/1962 Inhalation Chronic U-234 (c) (a) 6.29E+02 8.71E+03

2 

11/08/1950 12/31/1950 Inhalation Chronic U-234 (a) 1.26E+03 1.26E+03

11/08/1950 12/31/1950 Ingestion Chronic U-234 (b) 2.63E+01 2.63E+01

06/26/1952 08/27/1962 Inhalation Chronic U-234 (c) (a) 5.84E+02 8.02E+03

04/15/1961 04/21/1961 Inhalation Chronic U-234 (c) (a) 6.91E+03 2.20E+05

Adrian Plant

Choose one of the two 

bordered scenarios. Choose 

the second scenario if work 

overlapped the period from 

October 1, 1960, to April 11, 

1961.

Choose intakes based on 

either type M or S uranium, 

not both.

1 05/25/1954 12/31/1962 Inhalation Chronic U-234 (c) (a) 1.03E+03 1.41E+04

2 

05/25/1954 12/31/1962 Inhalation Chronic U-234 (c) (a) 8.03E+02 9.85E+03

10/01/1960 04/11/1961 Inhalation Chronic U-234 (c) (a) 1.85E+03 5.97E+04

a. Choose intake rates from the appropriate column based on the assumption of either a type M or a type S (not both) uranium intake. For each dose 

 reconstruction, intake rates should be chosen from only one column, not from multiple columns.

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

c. Assign recycled uranium contaminants per Table 3-7 and using ORAUT (2007) guidance on solubility.

d. Calculated at the 95th percentile using a GSD of 3 and assigned as a constant.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 27 of 45

4.0 ESTIMATION OF EXTERNAL EXPOSURE

Individual film badge results for Havens Laboratory and Adrian Plant are reported from late 1958 

through early 1961.

Because film badge data are available for about a 2-year period for at least some workers, this 

document does not attempt to address worker external exposures based on workplace data. When 

film badge results are available for a worker, the individual’s dosimeter results can be used to 

estimate dose. For unmonitored workers or unmonitored periods, this site profile provides an 

estimate of external dose based on analysis of coworker film badge dosimetry records.

The majority of photons from natural uranium metals are in the 30- to 250-keV energy range. Solid 

uranium objects provide considerable shielding of the lower energy photons and harden the spectrum, 

causing the majority of photons from a solid uranium object, such as a billet or a rod, to have energies 

greater than 250 keV. While it is recognized that solid uranium sources have a hardened photon 

spectrum, exposure to a thin layer of uranium on a surface results in a larger fraction of exposure to 

lower energy photons. This analysis assumed workers were exposed to photon energies in the 30- to 

250-keV range, which is favorable to claimants. Nonpenetrating dose from natural uranium consists 

primarily of electrons with energies above 15 keV. For consistent presentation, exposure or dose is 

reported as:

• Penetrating, assumed to be from photons of energies 30 keV or greater, and

• Nonpenetrating, assumed to be from photons of energies less than 30 keV or from electrons.

Adrian Plant initially used film badges (detecting beta, X-ray, and gamma radiations) for some workers 

on the AEC contract during a trial 26-week period from about November 3, 1958, through May 3, 

1959. The badge wear periods were nominally 2 weeks. The film badge data were tabulated using 

the end date on the badge report, but this sometimes differs by several days from the badge 

assignment records.

Contamination of the badges was a problem especially before 1959. During the period ending about 

February 22, 1959, the badge assignment sheet indicates that several of the badges were used to 

monitor the hacksaw, abrasive saw, induction heater, billet storage area, and salt bath for 24 hours, 

and then used by the workers. In later periods, some badges were assigned to work areas rather 

than people. Some badges might not have been used when there were no records of assignment, but 

elevated doses on some unassigned badges indicate that the badges were in areas of elevated 

radiation. No assignment has been found for badges dated between about May 4 and August 23, 

1959. No record indicated that monitoring occurred between September 17 and 25, 1961. No results 

were found for the badges reportedly assigned from September 25 through October 23, 1961. 

Results for badges reportedly worn from October 1 through 31, 1961, only included names for 

individuals who did not return the badges. It is possible that the names for the October 23, 1961, 

period should be associated with the overlapping October 31 period. It appears this last set of badges 

might have been worn for more than 2 weeks, but the doses are consistent with a 2-week period.

Film badges were supplied to Adrian Plant by Controls for Radiation in Cambridge, Massachusetts. 

No details on the dosimeter design or the film type are currently available. It is possible that two types 

of badges or badge holders were used based on a note about the January 8, 1961, badges that 

indicated the films were inserted improperly in the badge holders and were not read. A similar note 

was included with the February 5, 1961, badges, but results were reported for this period. The 

reporting format also changed about the same time in 1961. Instead of reporting X-ray or gamma and 

beta results as <10 mR or mrad, the results were reported as 0 mrem with a footnote that 0 “indicates 

less than minimum detectable dose – 5 mrem for X and γ <175 keV; 10 mrem for hard X, γ and β, 60 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 28 of 45

mrem for neutrons.” Eighteen of the 35 photon results for the period ending August 6, 1961, and one 

less-than result for the period ending October 18, 1959, were illegible.

Control badge doses were reported for most periods. These were generally reported as <10 mrem, 

but ranged up to 167 mrem for both photon and beta doses. For some periods, the photon and beta 

control badge doses differed from each other, but sometimes they were reported to be the same.

Neutron doses were reported for periods ending May 1 through September 4, 1960 and from 

December 25, 1960, through September 17, 1961 (except the week ending March 5, 1961). It was 

noted that calculated neutron doses were based on the assumption of a fast neutron source term, and 

that 1 rem equaled 14 × 106 neutrons/cm2

. In the early periods, results were reported as less than 

0.8 × 106 neutrons/cm2 which, using the conversion above, is consistent with the 60-mrem reporting 

threshold. Neutron dosimeters had an unshielded portion and a cadmium-shielded portion. All 938 

reported results for the shielded portion of the dosimeter were less than 60 mrem. Five of the 938 

results for the unshielded portion of the dosimeter equaled or exceeded the detection threshold (one 

other result reported neither as nonzero or “less than” was 0.3 × 106 neutrons/cm2

), and the maximum 

result was 1 × 106 neutrons/cm2

, which would equal about 100 mrem. Neutron dosimeters were 

calibrated with a polonium-beryllium source. This analysis concludes that the reported neutron dose 

results are consistent with the assumption of no significant neutron exposures, and the 0.5% rate of 

positive results is not necessarily indicative of workplace neutron exposures.

Badges were assigned to 14 to 37 workers per monitoring period. Historical review of film badge 

detection limits for this era indicated that although detection limits were reported as 10 mR or 

10 mrad, other documents indicates an LOD of 40 mrem for penetrating and shallow doses that were 

measured with film badge dosimeters in this era (ORAUT 2006). The exposure (R) organ dose 

conversion factors should be used to assess dose during this period.

To estimate doses for periods when dosimetry data are unavailable, the approach described in 

ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for External Dose Assignment (ORAUT 2011a) 

was used. The cycle data was converted to annual for all workers. For totals based on less than 

1 year of cycle data, the partial year doses were prorated to an annual dose. Cycled data below the 

LOD/2 were treated as missed dose. The 50th and 95th percentiles are in Table 4-1 below.

Table 4-1. External dose (rem) for unmonitored workers. 

Percentile

Havens Adrian

Gamma Beta Gamma Beta

50th 0.520 0.798 0.596 1.495

95th 1.225 2.932 1.221 5.832

If applicable, adjustments to the whole-body dose should be made for the extremities (e.g., hand and 

forearms) to account for geometry issues using the guidance in DCAS-TIB-0013, Selected Geometric 

Exposure Scenario Considerations for External Dose Reconstruction at Uranium Facilities (NIOSH 

2010b).

4.1 OCCUPATIONALLY REQUIRED MEDICAL X-RAYS 

Information about occupationally required medical X-ray examinations at Havens Laboratory and 

Adrian Plant is unavailable. AEC usually, but not always, required pre-employment and periodic 

(annual) medical examinations of workers who were involved in the larger uranium processing 

programs. The term pre-employment as used here means before performing AEC-contracted 

radiological work. The typical AEC medical program included a pre-employment chest X-ray 

examination with annual examinations thereafter. The type and frequency of X-ray examination 

should be based on current Oak Ridge Associated Universities (ORAU) Team guidance. Organ 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 29 of 45

doses can be obtained from the current revision of ORAUT-OTIB-0006, Dose Reconstruction from 

Occupational Medical X-Ray Procedures (ORAUT 2011b).

4.2 MISCELLANEOUS INFORMATION RELATED TO EXTERNAL DOSE

This section includes external dose information that could be of value for specific dose 

reconstructions. This analysis did not consider such information generically because of its limited 

applicability or because of limited details.

Havens Laboratory performed X-ray crystallography and had medical X-ray equipment.

Adrian used a 60Co source to check criticality monitors.

4.3 OCCUPATIONAL EXTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

Limited individual film badge results are available to determine doses. The detection limit is assumed 

to be 40 mR for penetrating radiation and 40 mrad for nonpenetrating radiation. Table 4-2 

summarizes annual external doses. The annual exposures can be used to estimate doses for 

unmonitored workers or unmonitored periods.

Table 4-2. External exposure summary for Havens Laboratory. 

Site

Exposure 

category

Exposure 

type Basis Year

50thpercentile 

annual 

exposure

95thpercentile 

annual 

exposure

IREP 

distribution

Havens 

Laboratory

Penetratinga

Analysis of 

Havens 

film badge 

results

1950

1952

1953

1954

1955

1956

1957

1958

1959

1960

1961

1962

0.076

0.268

0.520

0.520

0.520

0.520

0.520

0.520

0.520

0.520

0.520

0.520

0.178

0.631

1.225

1.225

1.225

1.225

1.225

1.225

1.225

1.225

1.225

1.225

Constant

Nonpenetrating

Analysis of 

Havens 

film badge 

results

1950

1952

1953

1954

1955

1956

1957

1958

1959

1960

1961

1962

0.116

0.411

0.798

0.798

0.798

0.798

0.798

0.798

0.798

0.798

0.798

0.798

0.427

1.510

2.932

2.932

2.932

2.932

2.932

2.932

2.932

2.932

2.932

2.932

Constant

Medical X-ray See ORAUT-OTIB-0006 (ORAUT 2011b)

a. The exposure (R) organ dose conversion factors should be used to assess dose during this period.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 30 of 45

Table 4-3. External exposure summary for Adrian Plant. 

Site

Exposure 

category

Exposure 

type Basis Year

50thpercentile 

annual 

exposure

95thpercentile 

annual 

exposure

IREP 

distribution

Adrian 

Plant

Penetratinga

Analysis of 

Adrian film 

badge 

results

1954

1955

1956

1957

1958

1959

1960

1961

1962

0.360

0.596

0.596

0.596

0.596

0.596

0.596

0.596

0.596

0.738

1.221

1.221

1.221

1.221

1.221

1.221

1.221

1.221

Constant

Nonpenetrating

Analysis of 

Adrian film 

badge 

results

1954

1955

1956

1957

1958

1959

1960

1961

1962

0.904

1.495

1.495

1.495

1.495

1.495

1.495

1.495

1.495

3.525

5.832

5.832

5.832

5.832

5.832

5.832

5.832

5.832

Constant

Medical X-ray See ORAUT-OTIB-0006 (ORAUT 2011b)

a. The exposure (R) organ dose conversion factors should be used to assess dose during this period.

5.0 ESTIMATION OF DOSE FROM REMEDIATION ACTIVITIES

Havens Laboratory

The areas shown in Figures 2-1 to 2-3 were the areas being used for uranium activities when 

operations ceased at Havens Laboratory, and the figures show the locations of sampling for each 

area (Survey 1962a and Survey 1962b). The conclusion of the post decontamination survey was that 

decontamination was successful.

Havens Laboratory was sold to the local Catholic diocese for use as a school. It was subsequently 

resold to the City of Bridgeport Board of Education for use as a high school and later as an 

educational center. A report describes the results of an Oak Ridge National Laboratory radiological 

survey of the site (MMES 1985, p. 48). An Office of Remedial Action and Waste Technology report 

stated that the site required no remedial action and would not be included in the Formerly Utilized 

Sites Remedial Action Program (DOE ca. 1987). A letter to the Superintendent of Schools for 

Bridgeport restated this conclusion (Fiore 1987). 

According to the NIOSH Residual Radioactivity survey (NIOSH 2004b), there is little potential for 

significant residual contamination outside the period of weapons-related work at the site. Therefore, 

for the purpose of dose reconstruction, potential doses from residual radioactivity at Havens 

Laboratory are not included.

Adrian Plant

In 1961, work was transferred from Adrian Plant to RMI at the Extrusion Plant in Ashtabula, Ohio. At 

that time, one large extrusion press was shipped to Ashtabula and put in operation there. Other 

equipment formerly used at Adrian Plant was dismantled and scrapped. The whereabouts of this 

material is unknown. Bridgeport Brass completed decontamination and close out in 1962. 

The Adrian Plant was sold to Martin Marietta in the early 1960s. It was used by that company until 

1974 when it was sold to General Motors Chevrolet Manufacturing Division. In a subsequent 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 31 of 45

decontamination effort in July of 1976, 866 kg of dust containing 3.9 kg of uranium were removed 

from roof supports. The air samples that were taken during the cleanup indicated that the air 

concentration did not exceed 8 × 10-12 μCi/cm3 (Hill 1976). Multiplying this air concentration by an air 

intake rate of 1.2 m3 per hour and an exposure period 176 hours (for the entire month of July) results 

in a calculated uranium annual inhalation intake of 1,690 pCi. Using the method described in Section 

3.0, the calculated ingestion intake was 33.8 pCi. 

In 1995, remediation of the service pits and removal of some drainage pipes took place from April to 

July. The maximum measured air concentration was 1.7 × 10-12 μCi/cm3 (TMA/Eberline 1995). 

Multiplying this air concentration by an air intake rate of 1.2 m3 per hour and an exposure period of 

704 hours (for the entire months of April through July) results in a calculated uranium annual 

inhalation intake of 1,436 pCi. Using the method described in Section 3.0, the calculated ingestion 

intake was 28.7 pCi.

To estimate external exposure during the remediation efforts in 1976, surveys indicate that the 

maximum exposure rate at 1 cm was 10 mR/hr beta-gamma. This dose rate extrapolate to 1 m 

(assuming source covers a radius of 5.64 cm or approximately an area of 100 cm2

), the dose rate at 1 

m would be 0.009 mR/hr. Assuming an exposure period for the entire month of July (176 hours); a 

total exposure of about 1.6 mR would have occurred. The ratio of nonpenetrating to penetrating 

exposure rates for external exposures at Adrian during the operational exposure period (Table 4-1) 

was 4.8. Therefore, this results in a gamma dose of 0.3 mR and a beta dose of 1.3 mR.

The post remediation survey report concludes that both indoor and outdoor average gamma radiation 

levels cannot be distinguished from background at this site (Wallo 1985). Therefore no further dose is 

assessed.

The exposure organ dose conversion factors should be used to assess dose during this period.

Table 5-1. Annual internal and external exposure during remediation activities. 

Internal Start End Exposure 

Absorption 

type

Intake

(pCi/d) IREP distribution

 U-234

07/01/1976 07/31/1976 Inhalation M, S 5.45E+01 Constant

07/01/1976 07/31/1976 Ingestion (a) 1.09E-00 Constant

04/01/1995 07/31/1995 Inhalation M, S 1.18E+01 Constant

04/01/1995 07/31/1995 Ingestion (a) 2.35E-01 Constant

External Start End Exposure Basis Exposure/yr IREP distribution

07/01/1976 07/31/1976

Penetratingb Survey 

instrument and 

an operational 

beta:gamma 

ratio

0.0003 R Constant

07/01/1976 07/31/1976 Nonpenetrating Survey 

instrument and 

an operational 

beta:gamma 

ratio

0.0013 R Constant

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

b. The ambient (H*10) organ dose conversion factors should be used to assess dose during this period.

The estimated annual penetrating and nonpenetrating external exposures to residual radioactivity 

from AEC operations at the site in Table 5-1 were calculated by assuming that workers were exposed 

for 2,000 hours per year.

Recycled uranium contaminants and thorium were not included in the assumed intakes for the 

residual period because the overestimate of uranium intake should be large enough to account for 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 32 of 45

dose from all intakes. Other assumptions about residual exposures should be consistent with 

assumptions from the operational period.

6.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 33 of 45

REFERENCES

AEC (U.S. Atomic Energy Commission), 1951, Monthly Report of Field Activities, December 1950 - 

January 1951, New York Operations Office, New York, New York, March 16. [SRDB Ref ID: 

60352]

AEC (U.S. Atomic Energy Commission), 1952, Research on Drawing Uranium and Related 

Operations, Contract AT-30-1-1405 with Bridgeport Brass Company, New York Operations 

Office, New York, New York. [SRDB Ref ID: 10988]

AEC (U.S. Atomic Energy Commission), 1955a, Supplemental Agreement, Contract AT(30-1)-1405, 

Mod. No. 9, with Bridgeport Brass Company, Oak Ridge Office, Oak Ridge, Tennessee. 

[SRDB Ref ID: 9838]

AEC (U.S. Atomic Energy Commission), 1955b, Bridgeport Brass Company, Aluminum Division, 

Adrian, Michigan, Occupational Exposure to Airborne Contaminants, HASL-B-BRA-2, New 

York Operations Office, New York, New York, March 8. [SRDB Ref ID: 10898]

AEC (U.S. Atomic Energy Commission), 1956, Bridgeport Brass Company, Aluminum Division, 

Adrian, Michigan, Occupational Exposure to Airborne Contaminants, HASL-B-BRA-3, New 

York Operations Office, Health and Safety Laboratory, New York, New York, April 24. [SRDB 

Ref ID: 13210]

AEC (U.S. Atomic Energy Commission), 1960, Bridgeport Brass Company, Aluminum Division, 

Adrian, Michigan, Occupational Exposure to Dust during Extrusion of Uranium Rods, HASL85, B-BRA-5, New York Operations Office, New York, New York, May 12. [SRDB Ref ID: 

9870]

AEC (U.S. Atomic Energy Commission), 1961, Bridgeport Brass Company, Adrian, Michigan, 

Occupational Exposure to Radioactive Dust during Extrusion of Uranium Rods, HASL-107, BBRA-6, New York Operations Office, New York, New York, February 8. [SRDB Ref ID: 14418

Author unknown, undated, “No samples received after 10/5/61.” [SRDB Ref ID: 9873, p. 508]

Bean, R. L., 1967, “Contract AT-(30-1)-1405 Employees,” letter to C. A. Keller (U.S. Atomic Energy 

Commission, Oak Ridge Operations Office), Reactive Metals, Inc., Extrusion Plant, Ashtabula, 

Ohio, August 7. [SRDB Ref ID: 12435, p. 4]

Belmore, F. M., 1952, “Alpha Rolled Rods for Bridgeport Brass Company,” memorandum to H. J. 

Hershman (Tonawanda Sub-Office), U.S. Atomic Energy Commission, New York Operations 

Office, New York, New York, September 15. [SRDB Ref ID: 63808]

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Connecticut, U.S. Atomic Energy Commission. [SRDB Ref ID: 9608]

DOE (U.S. Department of Energy), ca. 1987, Elimination Report for Bridgeport Brass Company, 

Havens Laboratory (Reactive Metals, Inc.), Kossuth and Pulaski Streets, Bridgeport, 

Connecticut, Office of Nuclear Energy, Washington, D.C. [SRDB Ref ID: 9774, p. 1]

DOE (U.S. Department of Energy), 2000, DOE Ohio Sites Recycled Uranium Project Report, DOEOH-00-0001, Ohio Field Office, Miamisburg, Ohio, May 15. [SRDB Ref ID: 3644]
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Dowling, F. R., 1955a, “Accountability Survey Report No. 68,” letter with enclosure to R. M. Treco 

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Ridge, Tennessee, September 30. [SRDB Ref ID: 14414, p. 9]

Dowling, F. R., 1955b, “Special Order Work,” letter to R. M. Treco (Bridgeport Brass Company), U.S. 

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[SRDB Ref ID: 16529, p. 18]

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Health Physics, volume 67, number 1, pp. 1–14. [SRDB Ref ID: 22418]

Fiore, J. J., 1987, untitled letter to J. A. Connelly (Superintendent of Schools, Bridgeport, 

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9780 p. 1]

Haywood, F. F., H. W. Dickson, W. D. Cottrell, W. H. Shinpaugh, J. E. Burden, D. R. Stone, R. W. 

Doane, and W. A. Goldsmith, Undated, Abstract, Radiological Survey of the Former Bridgeport 

Brass Company Special Metals Extrusion Plant, Adrian, Michigan. [SRDB Ref ID: 112247 p. 

6] 

Heffernan, T. E., J. C. Lodwick, H. Spitz, J. Neton, and M. Soldano, 2001, “Solubility of Airborne 

Uranium Compounds at the Fernald Environmental Management Project,” Health Physics, 

volume 80, number 3, pp. 255–262. [SRDB Ref ID: 7924]

Hill, R. F., 1976, “Radiation Survey of the Chevrolet Plant in Adrian, Michigan”, General Motors 

Corporation, Warren, Michigan, October 26. [SRDB Ref ID: 14421, p. 55-62] 

Hill, R. L., and E. W. Schneider, ca. 1986, “Removal of a Radioactive Drain Line from the CPC Adrian 

Plant,” General Motors Research Laboratories, Warren, Michigan. [SRDB Ref ID: 15381, p. 

6]

ICRP (International Commission on Radiological Protection), 1975, Report of the Task Group on 

Reference Man, Publication 23, Pergamon Press, Oxford, England. [SRDB Ref ID: 22735]

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for Radiological Protection, Publication 66, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22732]

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Members of the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, 

Publication 71, Pergamon Press, Oxford, England. [SRDB Ref ID: 22729]

Jefferson, D. R., 1962, “Health-Safety Inspection of Seymour Site and Decontamination of Havens 

Laboratory,” letter to J. W. Ruch (U.S. Atomic Energy Commission, Oak Ridge Operations 

Office), Bridgeport Brass Company, August 27. [SRDB Ref ID: 9588, p. 369]

Jones, S. E., 1987, “Final Elimination Reports and Site Summaries,” letter to A. Wallo III (U.S. 

Department of Energy), The Aerospace Corporation, Washington, D. C., July 28. [SRDB Ref 

ID: 14538, p. 4]

Karl, C. L., 1955, no title, teletype Nov 5528/1703Z to S. R. Sapirie (Oak Ridge Operations Office), 

U.S. Atomic Energy Commission, Fernald Area Office, Ross, Ohio, November 28. [SRDB Ref 

ID: 16529, p. 39] 
Document No. ORAUT-TKBS-0030 Revision No. 01 Effective Date: 01/24/2013 Page 35 of 45

Klevin, P. B., 1950, “Visit to Bridgeport Brass Company, Bridgeport, Connecticut - November 8, 1950,” 

memorandum to W. B. Harris, U.S. Atomic Energy Commission, November 27. [SRDB Ref ID: 

9588, p. 235]

Klevin, P. B., 1951, “Visit to Bridgeport Brass Co. - December 11, 1950,” memorandum to W. B. 

Harris, U.S. Atomic Energy Commission, January 10. [SRDB Ref ID: 9588, p. 234]

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Havens Plant of the Bridgeport Brass Company, Bridgeport, Connecticut, ORNL/RASA-85/6, 

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OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, April 

14.

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Radioactive and Beryllium Contamination at Atomic Weapons Employer Facilities and 

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for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

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Support, Cincinnati, Ohio, October 5.

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Scenario Considerations for External Dose Reconstruction at Uranium Facilities, DCAS-TIB0013, Rev. 1, Division of Compensation Analysis and Support, Cincinnati, Ohio, November 23.

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for Overestimating External Doses Measured with Film Badge Dosimeters, ORAUT-OTIB0010, Rev. 01, Oak Ridge Associated Universities Team, Cincinnati, Ohio, June 5.

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Schaeffer, M. R., 1955a, letter to F. R. Dowling (U. S. Atomic Energy Commission), Bridgeport Brass 

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Graphs showing the GMs (or detection thresholds in some cases) of the coworker uranium urinalysis 

data are shown on the following pages. Data in black were considered for fitting intakes, data in red 

were excluded from the fitting. The x-axis is in units of days, and the y-axis is in units of picocuries 

per day of uranium excreted in urine. For the Havens Laboratory graphs, day 0 is June 26, 1952, and 

day 3,714 is August 27, 1962. For the Adrian Plant graphs, day 0 is May 24, 1954, and day 3,142 is 

December 31, 1962. Some graphs end on the last day of bioassay results rather than the last day of 

the intake period.

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The following set of graphs showing predicted bioassay results from the estimated air intakes, superimposed on the GM and 84thpercentile bioassay results.

At Havens Laboratory the assumed early excretion (estimated from air samples) from intake rates of 740 pCi/d inhalation and 15.4 pCi/day 

for a type M assumption overestimate most of the GM and 84th-percentile data are shown starting on day –596 (November 8, 1950). For a 

type S assumption, the prediction is a little low for the GM data and underestimates most of the 84th-percentile data.

Havens Laboratory, type M, GM. Havens Laboratory, type S, GM.

Havens Laboratory, type M, 84th. Havens Laboratory, type S, 84th.

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At Adrian Plant, the intake rates of 740 pCi/d inhalation and 15.4 pCi/d ingestions estimated from air samples and a type M assumption 

overestimate most of the GM data and the early 84th-percentile data. For a type S assumption, the air concentration that was determined 

from the intake rate is a low for the GM data and underestimates all of the 84th-percentile data.

Adrian Plant, type M, GM. Adrian Plant, type S, GM.

Adrian Plant, type M, 84th. Adrian Plant, Type S, 84th.

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This set of graphs shows fitting of coworker bioassay data to one chronic inhalation intake. These data fits were not used to assign 

intakes.

For Havens Laboratory it can be seen that the incident data cause the overprediction of the uranium urinalysis results for the early years.

Havens Laboratory, type M, GM. Intake rate 174.3 pCi/d. Havens Laboratory, type S, GM. Intake rate 2,390 pCi/d.

Havens Laboratory, type M, 84th. Intake rate 514.0 pCi/d. Havens Laboratory, type S, 84th. Intake rate 7,063 pCi/d.

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For Adrian Plant, it can be seen that the elevated results in later years cause a slight overprediction in the majority of uranium urinalysis 

results.

Adrian Plant, type M, GM. Intake rate 235.9 pCi/d. Adrian Plant, type S, GM. Intake rate 3,223 pCi/d.

Adrian Plant, type M, 84th. Intake rate 564.8 pCi/d. Adrian Plant, type S, 84th. Intake rate 7,718 pCi/d.

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Graphs showing fitting of coworker bioassay data with two inhalation intakes. The first listed intake rate is for the period June 26, 1952, to 

August 27, 1962. The second listed intake rate is for the period April 15 to 21, 1961.

Havens Lab, type M, GM. Intake rates 95.88 and 1,134 pCi/d. Havens Lab, type S, GM. Intake rates 1,316 and 36,170 pCi/d.

Havens Lab, type M, 84th. Intake rates 201.5 and 4,519 pCi/d. Havens Lab, type S, 84th. Intake rates 2,729 and 146,000 pCi/d.

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Graphs showing fitting of coworker bioassay data with two inhalation intakes. The first listed intake rate is for the period May 24, 1954, to 

December 31, 1962. The second listed intake rate is for the period October 1 to April 11, 1961.

Adrian Plant, type M, GM. Intake rates 131.7 and 303.4 pCi/d. Adrian Plant, type S, GM. Intake rates 1,617 and 9,791 pCi/d.

Adrian Plant, type M, 84th. Intake rates 418.0 and 427.7 pCi/d. Adrian Plant, type S, 84th. Intake rates 5,540 and 13,260 pCi/d.

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The following graphs show the fitting of the bioassay data when data that seems to be associated with an incident are excluded from the 

analyses.

For Havens Laboratory additional data from April 22 to 24, 1961 were excluded from the analysis.

Havens Laboratory, type M, GM. Intake rate 103.3 pCi/d. Havens Laboratory, type S, GM. Intake rate 1,430 pCi/d.

Havens Laboratory, type M, 84th. Intake rate 236.6 pCi/d. Havens Laboratory, type S, 84th. Intake rate 3,259 pCi/d.

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For Adrian Plant, additional data from October 14, 1960, through April 10, 1961, were excluded from the analysis.

Adrian Plant, type M, GM. Intake rate 169.0 pCi/d. Adrian Plant, type S, GM. Intake rate 2,307 pCi/d.

Adrian Plant, type M, 84th. Intake rate 466.3 pCi/d. Adrian Plant, type S, 84th. Intake rate 6,367 pCi/d.

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