Page 1 of 86

DOE Review Release 03/13/2017

Summary of Extrusion Plant Site Information 

for Dose Reconstruction

ORAUT-TKBS-0056 Rev. 01

Effective Date: 03/07/2017

Supersedes: Revision 00

Subject Expert(s): Louise M. Buker

Document Owner 

Approval: Signature on File

Louise M. Buker, Document Owner

Approval Date: 02/14/2017

Concurrence: Wade C. Morris Signature on File for

John M. Byrne, Objective 1 Manager

Concurrence Date: 02/14/2017

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 02/14/2017

Concurrence: Danita Anderson Signature on File for 

Kate Kimpan, Project Director

Concurrence Date: 02/15/2017

Approval: Signature on File

Timothy D. Taulbee, Research Health Scientist

Approval Date: 03/07/2017

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 2 of 86

 

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/26/2007 00 Approved new document to provide information to perform dose 

reconstructions for Extrusion Plant claims. Incorporates formal 

internal review comments. Added additional references and Table 

A-1. Modified to address formal NIOSH review comments and 

typographical errors. Adds a Glossary, and Attributions and 

Annotations section. There is no change to the assigned dose and 

no PER is required. Initiated by Robert Hysong.

03/07/2017 01 Revision initiated to incorporate a rewrite of Section 3.0 - 

Occupational Medical X-rays, based on revisions to ORAUT-OTIB0079, Rev. 01, Guidance on Assigning Occupational X Ray Dose 

Under EEOICPA for X-Rays Administered Off Site, and ORAUTOTIB-0006, Rev. 04, Dose Reconstruction from Occupational 

Medical X-Ray Procedures. Revised Section 4.0 - Unmonitored and 

Environmental Dose, based on 2016 internal ORAUT and NIOSH 

comments for unmonitored and environmental dose. Revised 

Section 5.0 – Internal Dose, for referencing the Fernald Internal 

Dosimetry TBD section for recycled uranium. Revised neutron 

discussion, and resolved NIOSH comments from 2008 and 2016 to 

incorporate the external dosimetry data evaluation in Section 6.0 - 

External Dose. Revised Attachment E – External Data Evaluation of 

Doses, to give the 95th and 50th percentile doses to be used for 

unmonitored workers. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Louise M. Buker. 
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TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

1.0 Introduction .............................................................................................................................. 8

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

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

2.0 Site Information...................................................................................................................... 10

2.1 Process Description and Facilities .............................................................................. 12

2.2 Production Quantities ................................................................................................. 14

2.3 Source Term............................................................................................................... 14

2.3.1 Processed Uranium and Thorium................................................................. 14

2.3.2 Industrial X-Ray Sources.............................................................................. 17

2.4 Radiological Safety Program ...................................................................................... 17

2.4.1 Area Surveys ............................................................................................... 20

2.4.2 Personnel Monitoring ................................................................................... 20

2.4.3 Air Monitoring............................................................................................... 21

2.5 Radiological Incidents................................................................................................. 21

2.6 Physical Examinations Including Occupational Medical X-Rays ................................. 21

2.7 Work Periods.............................................................................................................. 22

2.8 Period After Radiological Production .......................................................................... 22

3.0 Occupational Medical X-Rays ................................................................................................ 23

4.0 Unmonitored and Environmental Dose................................................................................... 24

5.0 Internal Dose.......................................................................................................................... 25

5.1 Uranium Bioassay Data.............................................................................................. 25

5.2 In Vivo Monitoring....................................................................................................... 26

5.3 Intake Assumptions .................................................................................................... 27

5.3.1 Uranium ....................................................................................................... 27

5.3.2 Thorium........................................................................................................ 27

5.4 Internal Dose Assignment Summary........................................................................... 28

5.5 Unmonitored Internal Dose......................................................................................... 29

5.6 Internal Dose During the Postproduction Period ......................................................... 29

6.0 External Dose ........................................................................................................................ 31

6.1 Penetrating Doses...................................................................................................... 31

6.1.1 Gamma Dose............................................................................................... 31

6.1.2 Neutron Dose............................................................................................... 31

6.1.3 Historical Annual Exposures for Penetrating Radiation ................................ 33

6.1.4 Missed Doses .............................................................................................. 35

6.2 Nonpenetrating Dose.................................................................................................. 35

6.3 Extremity Monitoring................................................................................................... 35

6.4 Miscellaneous Information Concerning External Exposure ......................................... 36

6.5 External Dose Assignment Summary ......................................................................... 36

7.0 Attributions and Annotations .................................................................................................. 38

References ........................................................................................................................................ 39
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Glossary ............................................................................................................................................ 49

ATTACHMENT A PLANT LAYOUTS........................................................................................... 53

ATTACHMENT B N-REACTOR PRODUCTION.......................................................................... 75

ATTACHMENT C SAVANNAH RIVER SITE PRODUCTION....................................................... 77

ATTACHMENT D DOD PENETRATOR PRODUCTION.............................................................. 79

ATTACHMENT E EXTERNAL DATA EVALUATION OF DOSES................................................ 80

LIST OF TABLES

TABLE TITLE PAGE

2-1 Site Information......................................................................................................................10

2-2 Primary processing and storage buildings.............................................................................. 13

2-3 Uranium receipts for N-Reactor and SRS work ...................................................................... 14

2-4 Annual recycled uranium receipts .......................................................................................... 15

2-5 Uranium-235 weight percentage of uranium handled ............................................................. 15

2-6 Annual receipts of enriched uranium ...................................................................................... 16

2-7 Masses of DOE uranium and thorium processed and mass ratio of thorium to uranium......... 16

2-8 Uranium mixtures, specific activity, and isotopic fractions ...................................................... 17

2-9 Recommended surface contamination limits in 1962.............................................................. 20

3-1 Frequency of occupational X-ray screening at the Extrusion Plant......................................... 23

4-1 Occupational exposure intakes for uranium dust.................................................................... 24

5-1 Uranium urinalysis vendors and detection levels.................................................................... 26

5-2 In vivo lung measurement types and detection levels for various periods............................... 26

5-3 In vivo record codes...............................................................................................................27

5-4 Occupational exposure intakes from uranium dust................................................................. 29

6-1 External dosimetry .................................................................................................................32

6-2 Penetrating whole-body external doses, 1962 to 1999........................................................... 33

6-3 Upper bound of reported whole-body deep doses, 1974 to 1984 ........................................... 34

6-4 Upper bound of reported whole-body deep doses, 1994 to 1999 ........................................... 34

6-5 Missed external doses ...........................................................................................................35

6-6 Forearm and chest skin dose monitoring results for personnel who closely handled or 

inspected uranium billets in 1974 ........................................................................................... 36

A-1 Key for Figure A-1 ..................................................................................................................55

E-1 Example of analysis of prorating gamma and beta data and normalizing to annual................ 82

E-2 Years with limited data for external data analysis................................................................... 83

E-3 Missed external doses ...........................................................................................................83

E-4 Annual external data evaluation of doses modified to account for missed dose ..................... 85

E-5 Annual CTW dose based on data evaluation of worker dosimetry.......................................... 86
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LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Extrusion Plant site map......................................................................................................... 12

A-1 Plant layout ............................................................................................................................54

A-2 Plant layout, 1957 ..................................................................................................................57

A-3 Plant layout, 1962 ..................................................................................................................58

A-4 Plant layout, 1964 ..................................................................................................................59

A-5 Plant layout, 1965 ..................................................................................................................60

A-6 Plant layout, 1968 ..................................................................................................................61

A-7 Plant layout, 1969 ..................................................................................................................62

A-8 Plant layout, 1979 ..................................................................................................................63

A-9 Plant layout, 1981 ..................................................................................................................64

A-10 Plant layout, 1982 ..................................................................................................................65

A-11 Plant layout, 1983 ..................................................................................................................66

A-12 Plant layout, 1984 ..................................................................................................................67

A-13 Plant layout, 1985 ..................................................................................................................68

A-14 Plant layout, 1986 ..................................................................................................................69

A-15 Plant layout, 1987 ..................................................................................................................70

A-16 Plant layout, 1988 ..................................................................................................................71

A-17 Plant layout, 1989 ..................................................................................................................72

A-18 Plant layout, 1990 ..................................................................................................................73

A-19 Plant layout, 1991 ..................................................................................................................74
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

AP anterior-posterior

AWE Atomic Weapons Employer

BZ breathing zone

CEP Controls for Environmental Pollution

CFR Code of Federal Regulations

cm centimeters

CTW construction trade workers

DOD U.S. Department of Defense

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

ES&H Environment, Safety and Health

FEMP Fernald Environmental Management Project

FMPC Feed Material Production Center

g gram

GM geometric mean

HASL Health and Safety Laboratory

HEPA high-efficiency particulate air

hr hour

ICRP International Commission on Radiological Protection

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volts, 1,000 electron-volts

L liter

lb pound

LOD limit of detection

m meter

MDA minimum detectable activity

MeV megaelectron-volt, 1 million electron-volts

mg milligram

min minute

mL milliliter

mrad millirad

mrem millirem

MT metric ton

MTU metric tons of uranium

NAD nuclear accident dosimeter

nCi nanocurie
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NIOSH National Institute for Occupational Safety and Health

NLO National Lead Company of Ohio

NRC U.S. Nuclear Regulatory Commission

NVLAP National Voluntary Laboratory Accreditation Program

ODH Ohio Department of Health

ORAU Oak Ridge Associated Universities

PA posterior-anterior

pCi picocurie

PER program evaluation report

PNL Pacific Northwest Laboratory

POC probability of causation

RCG radioactivity concentration guide

RCRA Resource Conservation and Recovery Act of 1976

RMI Reactive Metals, Inc.

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

t ton

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

USTC United States Testing Company

yr year

μCi microcurie

μg microgram

°F degree Fahrenheit

§ section or sections
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1.0 INTRODUCTION

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

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

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

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

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

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

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

of the individual work required for each dose reconstruction.

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

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

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

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

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined as “any building, structure, or premise, 

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

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

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (NIOSH 2010):

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

conventional structures

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

reasons
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1.1 PURPOSE

The purpose of this summary document is to provide consistency in dose reconstructions for the 

Extrusion Plant (also known by the primary contractor’s name, Reactive Metals, Inc., or RMI) and to 

ensure that all components of dose are adequately addressed. This document provides information 

on the radiological processes and source terms and on the radiological controls and monitoring 

practices at the Extrusion Plant. While not meant to substitute for a complete site profile, this 

document represents the best understanding of the site at this time and provides assumptions for 

estimating doses when specific dose-related information is not available in individual records.

1.2 SCOPE

Section 2.0 of this document describes the Extrusion Plant and its history, including information about 

the radiological processes and source terms as well as radiological controls and monitoring practices. 

Section 3.0 describes occupational medical dose. Sections 4.0 describes occupational environmental 

dose. Sections 5.0 and 6.0 describe internal and external occupational dose.

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

justification, or clarification of the associated information, are presented in Section 7.0.
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2.0 SITE INFORMATION

The DOE Office of Worker Screening and Compensation Support defines the following information for 

the Extrusion Plant (DOE 2012):

Table 2-1. Site Information.

Site Extrusion Plant

Alternative names Reactive Metals, Inc. (RMI)

Location Ashtabula, Ohio

Covered period 1962 to November 1, 2006

Facility type Department of Energy

The Extrusion Plant received uranium from the Feed Material Production Center (FMPC) and the 

Weldon Spring Plant, as well as lesser quantities from other sites for extrusion, closed-die forging, or 

both (DOE 2000). Most of the uranium arrived in the form of billets, which were extruded into 

feedstocks for fabrication of fuel and target elements for use in nuclear production reactors. In 

addition, smaller quantities of thorium were processed at the plant.

The Extrusion Plant was the successor of the Bridgeport Brass facility in Adrian, Michigan. The work 

at Adrian was very similar to the work at the Extrusion Plant, and the same extrusion press was used 

at both facilities. The press, a 3,850-t Loewy horizontal extrusion press owned by the U.S. Atomic 

Energy Commission (AEC), was moved from Adrian to Ashtabula, Ohio, in November or December 

1961 (Haywood et al. 1982; Jefferson 1961, 1962a) after the end of work in Adrian. Uranium 

extrusion at Ashtabula began in January 1962 (Koh 1997). The majority of material that was 

processed at the facility was for AEC and DOE, but nonradioactive metals such as copper, zirconium, 

titanium, and molybdenum were also extruded for commercial firms (ORAU 1985).

The Extrusion Plant conducted its radiological operations under Section 110, “Exclusions,” of the 

Atomic Energy Act of 1954 (Koh 1997, p. 47) and as a licensee of both AEC and the U.S. Nuclear 

Regulatory Commission (NRC) (DOE 2000). Section 110 of the Act as amended (42 U.S.C. § 2140) 

states:

Nothing in this chapter shall be deemeda. to require a license for (1) the processing, fabricating, or refining of special 

nuclear material, or the separation of special nuclear material, or the separation of 

special nuclear material from other substances, under contract with and for the account 

of the Commission; or (2) the construction or operation of facilities under contract with 

and for the account of the Commission; or

b. to require a license for the manufacture, production, or acquisition by the 

Department of Defense of any utilization facility authorized pursuant to section 2121 of 

this title, or for the use of such facility by the Department of Defense or a contractor 

thereof.

Extrusion Plant operations for DOE and its predecessors were conducted under a prime contract from 

1962 through August 1987, a subcontract under the DOE Fernald Environmental Management 

Program (FEMP) from September 1987 through November 1992, a prime contract under the DOE 

Oak Ridge Operations Office from December 1992 through March 1993, a prime contract under the 

DOE Chicago Operations Office from April 1993 through March 1995, and a prime contract under the 

DOE Ohio Field Office (DOE 2000). On December 22, 2003, the Ohio Department of Health (ODH) 

received notification that the DOE prime contract (DE-AC24-93-CH1055) was terminated (ODH 1999–
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2004). DOE contracts with the Extrusion Plant included DE-AC05-760R 01405, administered by the 

Oak Ridge Operations Office (ORAU 1985) and DE-AC24-93CH10555 (DOE 2000).

The Extrusion Plant’s NRC-regulated work was conducted under source material License Number 

SMB 602, first issued in June 1962, which initially allowed the use of 500,000 lb (230 MT) of uranium 

and thorium. This was changed to 400,000 lb (180 MT) of uranium and 100,000 lb (45 MT) of thorium 

on October 31, 1973 (NRC 1962–1999), and to 10,000 lb (4.5 MT) of natural uranium, 600,000 lb 

(270 MT) of depleted uranium, and 10,000 lb (4.5 MT) of thorium in 1979 (NRC 1979–1980). On 

June 7, 1985, the NRC license allowed possession, use, and storage of 5,000 kg (5 MT) natural 

uranium and 300,000 kg (300 MT) of depleted uranium; thorium was no longer listed.

On October 15, 1991, the NRC license amounts remained the same, but the condition of use was 

changed to possession incidental to site characterization and decommissioning plan preparation. On 

September 11, 1997, the condition of use was changed to “possession incident to decommissioning, 

remediation, restoration, and waste disposal.” On May 26, 1999, the quantities were removed from 

the license and the materials were listed as natural uranium, depleted uranium, enriched uranium, and 

99Tc; the amount was described as “contaminated materials present at the site as of July 1, 1998” 

(NRC 1962–1999). The authority for licensing was later transferred to ODH on August 31, 1999, and 

ODH issued License Number 11900040004, which was similar to SMB-602 (DOE 2000; ODH 1999–

2004). RMI also held NRC License 34-10618-01 (Van Loocke 1979), which was reportedly for a 

1-mCi sealed 137Cs source used to calibrate a gamma alarm system (RMI 1995). As of 1995, RMI 

possessed 171 sealed sources for use in instrument calibrations, including 60Co, 90Sr/Y, 99Tc, 137Cs, 

210Pb, 226Ra, 230Th, natural uranium, 239Pu, 241Am, and 242Am. The quantities were not noted, but it 

was stated that some of the sources to be disposed would be characterized as Class C waste under 

10 CFR. § 61.55 (RMI 1995). On March 6, 2004, ODH issued approval for RMI to use sealed sources 

of 55Fe, 109Cd, and 241Am, in an X-ray fluorescence analyzer, 133Ba in a liquid scintillation counter, and 

63Ni in a gas chromatography instrument.

DOE work at the Extrusion Plant included the extrusion of primarily depleted and enriched uranium for 

the Hanford N-Reactor and the Savannah River Site (SRS). License SMB-602 allowed extrusion of 

uranium and thorium (NRC 1962–1999); it was also RMI’s authorization for use of source material for 

U.S. Department of Defense (DOD) armor-piercing penetrator work from 1974 through 1985 for DOD 

contractors (DOE 2000).

Westinghouse Materials Company of Ohio (WMCO) noted in its June 1989 review of the Extrusion 

Plant program that the most recent DOE extrusion campaign had been completed in September 1988 

and that commercial extrusion work was occurring during the June 1989 review (WMCO 1989). 

Westinghouse also stated that no uranium was being extruded during the June 1989 review (WMCO 

1989, p. 20). Other references (DOE 2000; RMI 1995) report that uranium extrusion for DOE ended 

in September 1988, and all other extrusion operations ceased on October 31, 1990. However, in 

Section 2.2 of this document, it is shown that uranium was received in 1989, 1990, and 1993, albeit in 

quantities much less than during the pre-1989 years. NRC License SMB-602 authorized uranium 

extrusion until the issuance of Amendment 4 on October 15, 1991.

This document refers to the period from January 1, 1962, through October 14, 1991, as the 

“radiological production period” (the time during which uranium was being extruded or during which 

uranium could have been extruded under the NRC license). The “postproduction period” is defined as 

October 15, 1991, to the present, during which no extrusion of radioactive metals occurred and 

predecommissioning and decommissioning activities were underway. Although there appears to be 

receipts for 0.05 MTU in 1993, there is currently no evidence that the material was processed.
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2.1 PROCESS DESCRIPTION AND FACILITIES

The Extrusion Plant was in Ashtabula County, Ohio, slightly east of the city of Ashtabula. The facility 

consisted of 25 buildings occupying 7 acres of a 32-acre site. A floor plan of the buildings is shown in 

Figure 2-1. Table 2-2 lists buildings where uranium was processed or process equipment was stored. 

Attachment A contains detailed site figures by year that show changes to the site over time 

(RMI 1995).

Figure 2-1. Extrusion Plant site map (DOE 1993, p. 13). 

The normal operations of extrusion and forging at the Extrusion Plant can be generally described as a 

metals fabrication process. The plant’s primary function was to change the shape or configuration of 

the received materials and then to ship these reconfigured metals to the receiving sites.

RMI’s primary equipment for handling the uranium and thorium metals included the Loewy extrusion 

press, a runout table, a cooling table, and a cut-off saw. The process components also included a 

straightener, a degreaser, an oil bath, and a cleaning hood. Three gas-fired incinerators were in an 

auxiliary building (RF-3 Butler Building). Two of the incinerators were used to oxidize uranium 

sludges and residues while the third was used to incinerate contaminated combustible materials.

The process steps varied with different materials but generally consisted of heating the metal in a salt 

bath for 1.5 hours after extruding, quenching, a degreasing step, packaging, and weighing. A portion 

of the extruded uranium metal was pickled in a nitric acid solution for the purpose of cleaning the 

material.

The DOE work supported the N-Reactor at the Hanford Site and the reactors at SRS. The Extrusion 

Plant also supported armor-piercing penetrator programs for DOD contractors. The plant produced 

N-Reactor fuel and targets from 1962 to 1988. From 1962 until 1970, uranium processing for 

N-Reactor consisted of receiving uranium, primarily from the FMPC, extruding the metal, and 

returning the extrusions and scrap (again primarily to FMPC). In 1971, RMI began using a forge 

process for N-Reactor fuel, which involved further processing of the extrusions before sending them 

directly to Hanford (DOE 2000).
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Table 2-2. Primary processing and storage buildings.

Building 

number Building namea Description/commentsb

1 Northwest storage building Storage of contaminated process equipment

2 RF-6 Butler Building addition Acid neutralization tanks

3 RF-6 Butler Building Storage of contaminated process equipment, lathes, drill 

press, small extrusion press, warehouse, laboratory, and 

offices; pickling, inspection, machining, and packaging of 

uranium

4 Enclosed ramp None

5 Locker rooms/foreman’s office None

6 Enclosed truck ramp None

7 Dock area None

8 Emergency equipment storage 

building

None

9 RCRA storage building None

10 Billet storage warehouse Storage of incoming and outgoing uranium

11 and

12

Main plant high bay and

Main plant low bay

Uranium processing, 3,850-t extrusion press, transfer table, 

pickling tanks, furnaces, abrasive saw

13 Runout table filter building None

14 Saw filter building None

15 Tool crib Extrusion tooling storage

16 Die head filter building None

17 Switchgear room None

18 Compressor room None

19 Wastewater treatment plant None

20 RF-3 Butler Building Uranium incinerator (oxidizer) and volume reduction 

equipment

21 ES&H Building None

22 Guard house None

23 Sewage disposal plant None

24 Modular laboratory None

25 Modular office None

a. DOE (1993).

b. RMI (1995).

The detailed steps involved in this process are described in Attachment B. Production for SRS 

occurred from 1962 to 1988. The general steps consisted of receiving uranium, extrusion and 

sectioning, and then returning the extrusions and scrap. Some minor changes were made to the 

process over time. These changes, as well as a detailed description of production for SRS, are 

included in Attachment C. Work on armor-piercing penetrators for DOD contractors occurred from 

1974 until 1985. The process was similar to that for SRS except for three process changes: 

1. The extrusions were hydraulically sheared and air-cooled before water quenching.

2. The extrusions were not generally cut on the abrasive saw and were not run through the roll 

straightener.

3. The extrusions were pickled with nitric acid and then rinsed before shipment.

A detailed description of the process is included in Attachment D.

In addition to uranium, thorium metal (with no cladding) was extruded for DOE from 1962 through 

1971 (RMI 1996, pp. 19–29). A small number of clad beryllium ingots for the New Production Reactor 

program were also extruded. This involved a process step described as hand rolling (Breslin and 
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Glauberman 1964, pp. 14–16). Because the ingots were identified as beryllium, they would not have 

been a source of radiation exposure.

2.2 PRODUCTION QUANTITIES 

The quantities of uranium sent to the Extrusion Plant for the N-Reactor and SRS production are listed 

in Table 2-3.

Table 2-3. Uranium receipts for N-Reactor and SRS 

work (MTU).

a

Uranium type 1962–1970 1971–1990

Enriched 13,442 11,829

Normal 4,904 330

Depleted 5,094 30,778

All 23,440 42,937

a. DOE (2000).

Production for the DOD penetrator program was reported as approximately 9,488 MTU of depleted 

uranium from 1974 to 1985 (DOE 2000). According to this inventory information, the total uranium 

sent to the plant was 75,757 MTU. This amount is in basic agreement with the 76,721.78 MTU of 

uranium compiled from annual receipts of recycled uranium at RMI as shown in Table 2-4 

(DOE 2000). 

According to the RMI Environmental, Safety and Health (ES&H) Director in 2006, site documentation 

indicates that the amount of uranium (and presumably thorium) the Extrusion Plant processed and 

handled was larger than the amount it shipped because some of the material was processed more 

than one time (ORAUT 2007a). The total amount of processed uranium, reported as 121,224 MT 

(Saito 1993; RMI 1995) and as 130,712 MT (Koh 1997), provides an upper amount of uranium the 

plant handled. The difference between the latter two numbers is an additional 9,488 MTU from DOD 

work (Koh 1997). Tables 2-5 and 2-6 indicate the enrichment of the uranium handled and received at 

RMI. 

A search of Extrusion Plant records by RMI indicated that thorium was only processed at the plant 

between 1962 and 1971 (RMI 1995). The amounts of uranium the plant processed for DOE and the 

amounts of thorium for SRS, Hanford, Y-12, and Davison Chemical (also known as W. R. Grace) 

between 1962 and 1971 are listed by year in Table 2-7 (Britcher 1992). Consideration of the plant’s 

total uranium process inventory would reduce the Table 2-7 ratios if the thorium inventories are 

complete.

2.3 SOURCE TERM

2.3.1 Processed Uranium and Thorium

Uranium was assumed to be extruded from January 1, 1962, through October 14, 1991, and thorium 

was extruded intermittently from May 1, 1962, through December 31, 1971. Depleted, normal, and 

enriched uranium were processed. Most of the uranium was probably recycled uranium. Uranium 

factors are listed in Table 2-8. As listed in Tables 2-4 and 2-5, the enrichment of uranium the 

Extrusion Plant handled was typically well below 2.1%. For dose reconstruction, the default 

assumption is that the uranium was 2% enriched.
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Table 2-4. Annual recycled 

uranium receipts.

a

Year MTU 

1962 526.85

1963 2,977.63

1964 4,518.48

1965 2,213.02

1966 2,694.82

1967 3,249.29

1968 2,875.46

1969 2,657.32

1970 1,731.69

1971 1,920.31

1972 1,898.09

1973 3,083.36

1974 2,226.30

1975 1,547.26

1976 2,076.36

1977 2,232.35

1978 2,314.72

1979 2,355.17

1980 3,175.80

1981 3,794.46

1982 5,873.91

1983 6,619.13

1984 4,832.58

1985 3,697.74

1986 4,322.51

1987 800.76

1988 496.52

1989 9.67

1990 0.17

1991 0

1992 0

1993 0.05

Total 76,721.78

a. The information comes from DOE 

(2000). It is not clear if this is 

only a listing of DOE-contract 

processed uranium or if it also 

includes NRC-licensed materials.

Table 2-5. Uranium-235 weight percentage of uranium handled (Koh 1997). 

U-235 wt% MTU Percentage of total uranium processed

0.14 and 0.2 64,438 49.3%

0.71 25,178 19.3%

0.86 3,932 3%

0.95 28,115 21.5%

1.02 8,108 6.2%

2.1 941 0.72%

Total 130,712 Not applicable
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Table 2-6. Annual receipts of enriched uranium.

a

Year MTU U-235 MTU

Average % 

enrichment

1962 244.7 2.3 0.947

1963 1,027.3 9.7 0.947

1964 2,503.1 24.2 0.965

1965 1,931.1 18.8 0.972

1966 2,357.3 25.6 1.083

1967 3,102.2 35.3 1.139

1968 1,068.5 18.5 1.732

1969 761.4 7.6 0.995

1970 448.4 4.2 0.946

1971 343.4 3.7 1.064

1972 554.2 5.6 1.005

1973 398.5 3.8 0.962

1974 918.2 9.2 0.999

1975 663.1 6.5 0.986

1976 321.9 3.4 1.046

1977 297.7 2.9 0.980

1978 535.0 5.4 1.009

1979 426.8 4.1 0.970

1980 266.3 2.8 1.048

1981 607.6 6.0 0.982

1982 670.5 6.5 0.970

1983 1,372.3 14.1 1.026

1984 1,324.0 13.1 0.987

1985 1,242.4 12.2 0.982

1986 1,262.3 12.5 0.991

1987 421.6 4.4 1.053

1988 257.7 2.4 0.947

1993 0.002084 0.000025 1.199

Total 25,327.5 264.8 Not applicable

a. DOE (2000).

Table 2-7. Masses of DOE uranium and thorium processed and 

mass ratio of thorium to uranium. 

Year U (MT)a Th (lb)b Th (MT) Th:U

1962 526.9 7,674 3.5 6.64E–03

1963 2,977.6 47,320 21.5 7.22E–03

1964c 4,518.5 4,200 1.9 4.20E–04

1965 2,213.0 0 0 0.00E+00

1966 2,694.8 1,170 0.5 1.86E–04

1967 3,249.3 883 0.4 1.23E–04

1968 2,875.5 0 0 0.00E+00

1969 2,657.3 3,200 1.5 5.64E–04

1970 1,731.7 1,300 0.6 3.46E–04

1971 1,920.3 1,925 0.9 4.69E–04

a. DOE (2000).

b. Britcher (1992).

c. An additional amount of thorium was processed for Hanford’s Project 

A-801-38 (Britcher 1992), but the amount is currently unavailable. It 

consisted of only one extrusion.
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Table 2-8. Uranium mixtures, specific activity, and isotopic fractions.

Uranium mixture and 

radionuclidea

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

Slightly enriched (2%), U-234b 0.7694 0.0002 28.76 0.01

Slightly enriched (2%), U-235b 0.0268 0.02 1 1

Slightly enriched (2%), U-238b 0.2038 0.9798 7.618 48.99

Natural, U-234c 0.4886 5.37E-05 21.4 0.00745

Natural, U-235c 0.0228 7.20E-03 1 1

Natural, U-238c 0.4886 9.93E-01 21.4 138

Depleted, U-234d 0.1546 1.00E-05 14.45 0.00502

Depleted, U-235d 0.0107 1.99E-03 1 1

Depleted, U-236d 0.0005 3.11E-06 0.0467 0.00156

Depleted, U-238d 0.8342 9.98E-01 78 501

a. Although listed as a recycled uranium component, U-236 represents <1% of the dose from exposure 

to uranium (ORAUT 2016a). 

b. Useful factors: Overall activity of 2% enriched uranium is 1.616 pCi/μg and 80.8 pCi/μg U-235. 

c. Useful factors: Overall activity of natural uranium is 0.683 pCi/μg and 94.9 pCi/μg U-235. 

d. Useful factors: Overall activity of depleted uranium is 0.4021 pCi/μg and 202 pCi/μg U-235. 

The Extrusion Plant was licensed by NRC to use thorium from mid-1962 to mid-1985 (NRC 1962–

1999) and might also have processed thorium under an Atomic Energy Act exception (Koh 1997). A 

review of plant records (Britcher 1992) indicated that thorium was not processed by RMI after 1971. 

Table 2-7 shows thorium to uranium mass ratios of less than 1%, and the NRC (1962–1999) licensing 

documents indicate that, for most periods, authorized mass ratio of thorium to uranium (natural and 

depleted) was no more than 25% before 1979, and was less than 2% by 1979. These ratios would be 

lower if consideration was given to the mass of enriched uranium. The low ratios of thorium to 

uranium are also supported by later environmental sampling and characterization surveys that 

identified only uranium and technetium contamination at the site (RMI 1995, 1996, pp. 19–29). 

Thorium-232 (which is the most significant isotope by mass in natural thorium) contamination was not 

detected on site. Thorium-230, a long-lived uranium progeny, was found on site but was 

characterized as consistent with background levels (NRC 1962–1999).

2.3.2 Industrial X-Ray Sources

No Extrusion Plant site documentation indicates that industrial radiography sources were used at the 

Ashtabula facility. Furthermore, the 1965 RMI application for renewal of NRC License SMB-602 

specifically stated that no provision for metallographic laboratory handling activities would be included 

in the application like those previously carried out at the Bridgeport Brass facilities in Seymour, 

Connecticut (Bean 1965a).

2.4 RADIOLOGICAL SAFETY PROGRAM

Information about the Extrusion Plant’s early radiological safety program was described in licensing 

documentation. Other information sources included the Health Protection Rules and Regulations

(RMI 1973); AEC Health and Safety Laboratory (HASL) reports; and health protection, nuclear safety, 

and environmental inspections and appraisals that were conducted periodically throughout the history 

of the Extrusion Plant. In 1985, an independent health, safety, and environmental review at RMI 

resulted in a recommendation that the entire area of industrial hygiene and health physics be 

upgraded starting with a thorough health protection program evaluation and fundamental assistance 

on establishing written procedures, sampling, documentation, and recordkeeping, followed by a 

quality assurance evaluation (Row 1985). Consequently, between July 1985 and 1988, Battelle, the 

Pacific Northwest Laboratory (PNL) contractor, conducted an in-depth review of the RMI radiation 

protection program that included air sample particle size, solubility studies, and recommendations for 

internal and external dose control, contamination control, training, respiratory protection, and 
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associated procedures (Munson 1985). RMI established an As Low as Reasonably Achievable 

Committee, which had its first organizational meeting on April 2, 1985. A Health Physics Appraisal in 

May 1987 indicated that a draft health physics procedure manual was completed on September 25, 

1986, and was formally adopted in July 1989 (RMI 1989a) after DOE uranium extrusion had ceased.

After passing through the guardhouse, hourly production and maintenance workers entered and 

exited the main plant building through the locker room door and performed a complete clothing and 

shoe change (Jefferson 1962a). By 1973, caps were also required by for hourly and production 

personnel (AEC 1973). A storage bin containing clean coveralls and gloves was located in the locker 

room (Jefferson 1962a). Lockers were used to store uncontaminated street shoes and personal 

clothing; contaminated shoes were stored below benches adjacent to the clothing lockers and a drum 

for disposal of contaminated work coveralls and gloves was located just outside the door leading from 

the locker room into the plant production area (Jefferson 1962b). At the end of a work shift, workers 

sat down at a bench in front of their lockers and removed contaminated coveralls and shoes. They 

then put on their shower clogs and deposited their contaminated clothing in the disposal drum as they 

proceeded to the shower area. After showering, a worker would return to his locker, put on personal 

clothing, and exit the plant through the locker room door (Jefferson 1962b).

Salaried workers and visitors were allowed through the main plant office area entrance door, which 

originally required an immediate change into smocks (lab coats) and shoes, or the use of rubber shoe 

covers, for use in contaminated areas (Jefferson 1962a). The location of the rack containing reusable 

smocks was moved a short time later from the shoe change area just inside the main plant office area 

entrance door to just outside the door leading from the main plant office into the main production area 

next to a drum for contaminated smocks (Jefferson 1962b). Entrance to the plant production area 

from the office area required the use of reusable smocks and rubber shoe covers. The entire floor of 

the plant office area was treated as a radiologically controlled area. Spot check surveys of desktops 

and other spots above the floor were used as a means of controlling the spread of contamination to 

desks (Jefferson 1962b).

Contamination monitoring of personnel exiting the operating area was not performed in the early 

years of operation because it was believed that good housekeeping, use of protective clothing, and 

washing of hands before leaving would afford adequate protection against inhalation or ingestion of 

contamination (Jefferson 1962a). Daily surveys of the locker room, office area, and production area 

access points was practiced “rigidly,” and hourly production employees were encouraged to take 

showers at the end of their shifts (Jefferson 1962a). Site documentation implies that contamination 

monitoring was performed routinely in later years. The findings of an independent review of the RMI 

radiation protection program in July 1985 indicated that contamination control boundaries and frisking 

were lax but the nature of materials available to be spread along with existing controls were such that 

the accidental transport of significant quantities of material off site was “unlikely” (Munson 1985). 

Correspondence dated November 7, 1985, indicates that RMI had procured some “needed personnel 

friskers” based on a recommendation from Battelle at PNL and that the radiation worker training 

planned for January 1986 would correspond with the implementation of an improved contamination 

control program (Munson 1985). WMCO (1989) observed that contamination control and monitoring 

practices were being implemented in a June 1989 review of the RMI program. By 1991, personnel 

contamination monitoring when workers exited a contamination control zone was being performed 

with alarming beta-gamma frisker instrumentation. Protective clothing and personal clothing 

contamination limits were 15,000 and 3,700 dpm/100 cm2 beta-gamma (or alpha), respectively (RMI 

1990a), while the skin contamination limit was 3,700 dpm/100 cm2 beta-gamma (RMI 1994).

Before the construction of a new Butler Building sometime in the first quarter of 1965, the plant 

changing facility (locker room) was also used as a lunchroom (Ruch 1965). The dining table in the 

locker room was used at lunch breaks and coffee breaks (three per shift), where plant personnel were 

required to wear blue smocks over contaminated coveralls to prevent the spread of contamination to 
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the table. In addition, the washing of hands was supposed to be “strictly enforced” at the time the 

blue smocks were used during breaks (Jefferson 1962a). The lunchroom and offices were relocated 

to the new Butler Building by March 1965 (Ruch 1965), but alpha contamination survey records from 

1977 and 1978 indicate removable contamination levels up to 1,500 dpm/100 cm2 in the hourly 

employee lunch room as well as elevated fixed contamination (Featsent 1977a). Similar results were 

recorded on 1975 and 1977 alpha contamination survey records from the hourly employee locker 

room (Featsent 1975, 1977b). Contamination surveys of the lunchroom and change room in 

September 1989 indicated total beta-gamma levels of 29,526 and 76,798 dpm/100 cm2

, respectively 

(RRA 1989).

Early site documentation indicates that there were designated eating and smoking areas, but in May 

1985 an independent health, safety, and environmental review at RMI resulted in RMI instituting a nosmoking policy in the production and incineration areas of the plant because “employees were 

smoking throughout the plant” (Row 1985).

Written standard operating procedures that incorporated criticality controls were used for extrusion 

and handling of slightly enriched uranium (Puterbaugh and Van Loocke 1964), and nuclear accident 

dosimeters (NADs) had been placed inside access doors to plant areas shortly before the 1.95%-

enriched uranium (clad with beryllium) New Production Reactor campaign in September 1964 (Bean 

1964). A fourth NAD was installed in the warehouse area of the RF-6 Building in June 1965. By April 

1965, a criticality alarm system had been installed (Bean 1965b).

Ventilation was the primary means used to limit radioactivity in the air at the Extrusion Plant, but 

respirators were provided to personnel for certain operations that caused excessive airborne 

contamination (Jefferson 1962c). HASL conducted dust and ventilation surveys at RMI in June 1962 

and March 1964 that resulted in ventilation system modifications and procedural changes to reduce 

dust levels. Four ventilation systems with hoods were used for the early extrusion process. The 

Loewy extrusion press, die head and runout table, extrusion cooling table, and cut-off saw were each 

serviced by a separate ventilation system. There was also a small hood adjacent to the extrusion 

press used to clean die parts, which had its own fan and exhaust system. Another ventilation system 

serviced three gas-fired incinerators in an auxiliary building (Scrap Building). Two of the incinerators 

were used to oxidize uranium sludges and residues while the third was used for contaminated 

combustible materials (Breslin and Glauberman 1964, pp. 14–16). The gas-fired incinerators were 

equipped with a Type N Rotoclone dust collector (Breslin and Glauberman 1964, pp. 14–16; ORAU 

1985). Additional processes that were served at one time or another by a ventilation or exhaust 

system included the resistance heater-roll straightener, vapor degreaser, acid pickle tanks, lathes, 

forge booths, hand filing and forge area, and the ingot acid etch booth and grinding booths in the 

warehouse portion of the RF-6 Building (Bean 1973). Site documentation indicates that during most 

of the production period of the Extrusion Plant, the primary source of stack emissions was the 

abrasive saw, followed by the scrap incinerator (Ruch 1964; Smith 1973; Hibbitts and Wing 1980; 

Wing 1982). The abrasive saw exhaust stack was not equipped with an emission control system until 

1984 when a precipitator was installed (ORAU 1985).

Isokinetic sampling probes equipped with filter paper discs were periodically used to obtain 

representative airborne particulate samples from each plant stack for alpha counting. One stack 

sample a week was collected by the Safety Officer on a rotating basis so that each exhaust system 

was sampled at least every 7 weeks (AEC 1973). Continuous monitoring of all seven plant stacks, 

along with the emission control improvements for the stacks, serving the abrasive saw, forge area, 

and uranium scrap incinerators was recommended in 1985 (Row 1985). Installation and testing of a 

new abrasive saw ventilation system, which utilized high-efficiency particulate air (HEPA) filters and a 

baghouse filter system, were completed in August 1985, and RMI established a new policy of 

performing continuous stack sampling when the saw was in use. In addition, a ventilation and filter 

(non-HEPA) system was installed on the scrap incinerator sometime between June 1985 and June 
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1987, but isokinetic exhaust sampling was only performed when the ventilation equipment for the 

incinerator facility was operating (Davis 1987). WMCO (1989) noted in a June 1989 review of the RMI 

program that applications had been submitted for permission to install the extrusion press, runout 

table, cooling table, abrasive saw, scrap incinerator, forge area, pickling tanks, tool coating dip tank, 

and lathe emission sources, and that ventilation systems with HEPA filtering and stack emission 

monitoring were in place for each of the sources.

2.4.1 Area Surveys

According to the Extrusion Plant 1962 license application, floor contamination surveys in the locker 

room and plant office areas were to be performed daily and weekly floor surveys were to be 

performed in the production areas associated with the highest contamination levels (including the 

Scrap Building and warehouse portion of the RF-6 Building). Surveys of the surfaces of equipment 

above floor level were to be made based on the judgment of the Safety Officer or whenever 

equipment was shipped or removed from a contaminated area. The routine in-plant clean-up cycles 

were expected to keep the surface contamination limits below the levels listed in Table 2-9. Available 

portable instruments at the start of operations included alpha detectors and Geiger-Müller counters. 

Decontamination procedures generally involved the use of simple detergents, solvents, or steam 

cleaning. Broom sweeping was prohibited; dry vacuuming could be performed using an electrically 

operated industrial vacuum (Jefferson 1962b, 1962c). According to the independent health, safety, 

and environmental review in May 1985, there was no delineation of contaminated zones in areas of 

the plant where uranium was received, handled, processed, machined, and prepared for delivery 

(Row 1985). By June 1985, establishment of contamination control levels were consistent with DOE 

policy, but the total area of the facility was considered a contamination zone. By May 1987, progress 

had been made in reducing the area of contaminated zones to approximately two-thirds of the facility 

(Davis 1987).

Table 2-9. Recommended surface contamination limits in 1962.

a

Location

Removable 

alpha 

(dpm/100 cm2)

Fixed 

alpha 

(dpm/100 cm2)

Removable 

beta-gamma 

(mrad/hr)

Fixed 

beta-gamma 

(mrad/hr)

Hot working areas 3,000 6,000 5 10

Other production plant areas 1,500 3,000 2 5

Locker room floor 250 500 0.2 2.5

Eating table 100 200 Not detectable Not detectable

Plant office 500 1,000 0.3 2.5

Shipping-receiving floor 500 1,000 0.4 2.5

Truck beds (after use) Not detectable 500 Not detectable 0.4

Equipment-in-place (accessible 

areas; above floor level)

10,000 20,000 5 10

Before removal (to other plant 

property sites; above floor level)

2,000 4,000 0.4 2.5

a. Jefferson (1962a, 1962b, 1962c).

2.4.2 Personnel Monitoring

Based on a review of the claim files and available site documentation, it appears that greater than 

50% of the RMI employees were monitored for external and internal radiation. External monitoring for 

most years of RMI operations consisted of the use of a film or thermoluminescent dosimeter (TLD) 

badge worn on the chest to monitor radiation from photons and electrons. Neutron monitoring and 

extremity monitoring studies occurred in later years, and extremity monitoring was routine for some 

employees in later years. Determination of uranium levels in urine began in 1962 for workers believed 

to have the larger potential for internal exposures. Beginning in 1968, in vivo chest counts were also 
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being used to monitor internal exposure from uranium. Beginning in 1979, in vivo chest count results 

sometimes included monitoring for other radionuclides such as thorium and technetium.

2.4.3 Air Monitoring

Early site documentation (Jefferson 1962c) indicates that the Safety Officer periodically performed 

breathing zone (BZ) air sampling at each production worker’s operating position such that each 

position was sampled routinely every 2 months for comparison to the maximum allowable 

concentration of 6 × 10-11 μCi/mL, but it is not clear that specific actions were taken if this level was 

exceeded. BZ sampling appears to have continued throughout the history of the plant. Later 

documentation (Brewer 1990) indicates that workers were required to wear respiratory protection and 

BZ monitoring samplers when air concentrations reached 2 × 10-11 μCi/mL. In addition, at least one 

general area high-volume air sample was collected weekly at different locations around the plant, as 

chosen by the Safety Officer, based on the activities being conducted (Jefferson 1962c). General 

area air sampling appears to have continued throughout the history of the plant. Between 1962 and 

1965, HASL conducted stack, environmental, general area, cyclone, and BZ air sampling at the 

Extrusion Plant. The most comprehensive and complete data report available (Breslin and 

Glauberman 1964, pp. 14–16) has been used as a basis for comparison to airborne exposure at the 

Bridgeport Brass plant. The conclusion that can be reached from this comparison is that, in general, 

exposures at the Extrusion Plant were less than those at the Bridgeport Brass plant. In addition, twostage air-sampling studies that were conducted between 1963 and 1964 at the Extrusion Plant 

indicated that 90% of the air contamination was in a nonrespirable particle size range (Ruch 1964). 

Later studies conducted from 1985 to 1986 indicated that over 90% of the aerosol in the Scrap 

Building was nonrespirable (Munson 1986a) while 66% of the aerosol in the extrusion press area was 

nonrespirable (Munson 1986b).

2.5 RADIOLOGICAL INCIDENTS

There are two notes written beside urine bioassay results that refer to fires in the Scrap Building in 

late June 1965 and in November 1965. No additional information on the 1965 fires has been found. 

A 1995 uranium fire incident involving a drum containing slightly enriched uranium oxide waste (the 

folder that contained the report was labeled “1% Oxide Issue”) occurred inside a glovebox (probably in 

October) (DOE 1995). The initial report indicated a small pie-shaped area (one-sixth the area of the 

drum and a few inches deep) in the bottom of the drum was smoldering like burning embers (DOE 

1995). A later review indicated that the smoldering had changed to flames inside the enclosure, that 

the ventilation remained on during the fire, and that at least one worker in the area was not wearing 

the required gloves (DOE 1995).

2.6 PHYSICAL EXAMINATIONS INCLUDING OCCUPATIONAL MEDICAL X-RAYS

Section 3.0 provides information about what is known about the occupational health program at RMI, 

including the X-rays taken for screening for occupational disease. The Extrusion Plant shared 

medical facilities with a nearby RMI plant (Hibbitts and Pryor 1970), and a dispensary was co-located 

with the main guardhouse. The medical program, described in 1970 and 1976 health and nuclear 

safety appraisals (Hibbitts and Pryor 1970; Johnson 1976), consisted of preemployment, annual, and 

termination physicals including blood tests, audiograms, immunizations, urinalyses, and chest X-ray 

examinations for everyone except female clerical employees. In addition, laborers had a 

preemployment lower back (“normal spine”) X-ray examination (Johnson 1976) because they were 

more “prone” to back injuries (Hibbitts and Pryor 1970).
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2.7 WORK PERIODS

For much of the operational period at RMI, there were at least two shifts: day and night. The 1964 

HASL air concentration study (Breslin and Glauberman 1964, pp. 14–16) used 480 minutes for the 

total exposure time per day on the site, but assigned 40 minutes for lunch with a morning break and 

10 minutes in the locker room, which would reduce the time engaged in work to 430 min/d. In 

general, bioassay and external dosimeters should be used to determine doses and, in those cases, 

the period of exposure is irrelevant. To account for exposures that were not monitored by bioassay or 

dosimeters, this summary assumes a default exposure of 2,000 hr/yr.

2.8 PERIOD AFTER RADIOLOGICAL PRODUCTION

This summary refers to January 1, 1962, through October 14, 1991, as the “radiological production” 

period, during which uranium was being extruded or might have been extruded based on uranium 

receipts. The “postproduction period” is that during which extrusion of radioactive metals ceased and 

predecommissioning and decommissioning activities were underway (from October 15, 1991, to 

December 2006). On October 15, 1991, License Amendment 4 converted NRC License SMB-602 

from an operating license to a license only for possession incident to site characterization and 

decommissioning plan preparation. On May 14, 1993, License Amendment 5 was approved to 

designate a new Radiation Safety Officer and contact person for RMI. License Amendment 6 was 

approved on November 9, 1993, to enable predecommissioning activities that included surveys, 

equipment removal, waste handling, shipment, and disposal. On September 11, 1997, the 

decommissioning plan was approved and incorporated into the license as Amendment 8. Internal and 

external monitoring continued throughout the postproduction period until 2004 when internal 

monitoring practices changed. According to the current RMI ES&H Director, starting in January 2004, 

the only urine bioassay samples submitted by workers were preemployment, termination, and “for 

cause” whenever an intake was suspected (ORAUT 2007b). By January 2004, building 

decontamination was completed and work at the site consisted primarily of soil and ground-water 

remediation (ORAUT 2007b). In March 2004, decontamination and decommissioning activities at the 

site temporarily ceased until November 2005 when a remediation services contractor (LATA-SHARP 

Remediation Services) was hired to complete decontamination and decommissioning (ORAUT 

2007b). By March 2004, less than 20 RMI management and compliance personnel remained at the 

Extrusion Plant to provide oversight of the remaining decontamination and decommissioning field 

activities, which were likely completed by December 2006 (ORAUT 2007b).

Uranium extrusion for DOE at RMI ceased in September 1988, and all other extrusion operations 

reportedly ceased on October 31, 1990 (DOE 2000), although the inventory receipt information 

compiled in Table 2-3 shows receipt of 0.1 MTU in 1993. This document refers to the radiological 

production period as January 1, 1962, through October 14, 1991, and the postproduction period as 

the period during which extrusion of radioactive metals is believed to have ceased (despite the later 

receipt of uranium in 1993, which appears to not have been extruded based on the license change in 

1991).
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3.0 OCCUPATIONAL MEDICAL X-RAYS

The medical program that is described in 1970 and 1976 health and nuclear safety appraisals 

consisted of preemployment, annual, and termination physicals including a chest X-ray for everyone 

except female clerical employees (Hibbitts and Pryor 1970; Johnson 1976). The female clerical 

employees did not appear to receive the annual physicals, only the preemployment and termination 

physicals (Hibbitts and Pryor 1970, p. 8). In addition, laborers received one preemployment lumbar 

spine X-ray (Hibbitts and Pryor 1970; Johnson 1976).

The 1970 appraisal specifically states that X-ray examinations were made off the site at Ashtabula 

General Hospital; claim data support this statement. The RMI ES&H Director stated in an interview in 

2006 that RMI obtained its own X-ray equipment for chest X-ray examinations in later years, although 

the specific date of equipment acquisition is unknown (ORAUT 2007b). Based on information in the 

claim file records, it appears that X-rays were taken off site at Ashtabula General Hospital through 

1980 [1]. According to the guidance in ORAUT-OTIB-0079, Guidance on Assigning Occupational X 

Ray Dose Under EEOICPA for X-Rays Administered Off Site (ORAUT 2016b), the dose from X-rays 

performed off site at a noncovered facility (such as the Ashtabula General Hospital) should not be 

included in dose reconstruction.

X-ray examinations were performed on site at the Extrusion Plant starting in 1981. The practice of not 

providing annual physicals to female clerical employees was an early one (Hibbitts and Pryor 1970, 

p. 8), and might not have been retained into the 1980s. The practice of taking preemployment lumbar 

spine X-rays was phased out sometime before 1986 and, beginning in 1997, chest X-rays were only 

taken at the physician’s discretion (ORAUT 2007b). There is no evidence collected to date that 

establishes when the X-ray equipment might have been removed from the site, so it is assumed that 

X-rays were taken on site from 1981 to 2006, when the site was decommissioned. What is known 

about RMI’s occupational X-ray screening frequency is summarized in Table 3-1.

Table 3-1. Frequency of occupational X-ray screening at the Extrusion Plant. 

Period Workers X-ray projections Frequency

Before 1981 

(off site)

None X-rays not included in dose 

reconstructiona

Not applicable.

1981–1986 All workers PA chest Preemployment, annual, and termination.

1981–1986 Laborers and 

trade workers 

(use job title)

AP and AP spot, lateral,

and lateral spot lumbar 

spineb

Preemployment only.

1987–2006 All workers PA chest Per physician direction and as contained 

in the claim file records. Default to 

preemployment, annual, and termination if 

no records in the claim file.

a. ORAUT (2016b), Hibbitts and Pryor (1970), Johnson (1976).

b. ORAUT-OTIB-0006 (ORAUT 2011).

No information is currently available about the type of X-ray equipment used at the Extrusion Plant. 

Specific X-ray techniques for the examinations have not been found. Some X-ray technique factors 

have been collected from RMI records, but they are for examinations that would not have been 

performed in an industrial setting (such as pediatric and upper and lower gastrointestinal 

examinations) (Heinlein 1983), so it is believed that these might have actually come from another 

facility such as the Ashtabula General Hospital. Therefore, only doses for X-rays on site at RMI from 

1981 through 2006 should be assigned in accordance with ORAUT-OTIB-0006, Dose Reconstruction 

from Occupational Medical X-Ray Procedures (ORAUT 2011).
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4.0 UNMONITORED AND ENVIRONMENTAL DOSE

An extensive data evaluation of externally monitored workers was conducted and the results are given 

in Attachment E. In accordance with ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for 

External Dose Assignment, unmonitored employees with low potential for exposure (those not 

working in hot areas) can be assigned the values in Table E-4 at the 50th percentile for gamma as a 

constant distribution from 1962 to 2004. 

Also, a site characterization report (Koh 1997) concluded that, based on actual site conditions in 

1992, the doses that were received by the workers and the public were below NRC and DOE limits 

and did not pose a significant health risk. The mean external worker dose was 0.0039 rem/yr, which 

is below the reasonable maximum dose of 0.006 rem/yr for a worker at the RMI site. The 

Environmental Assessment of the Safety Evaluation for the license amendment of 1995 indicated that 

most outside areas in the main plant area were at ambient background levels.

The study by Breslin and Glauberman (1964, pp. 14–16) specifically indicates that the internal 

environmental exposure for operating personnel such as draftsmen and secretaries (see Table 4-1) 

was less than 0.1 of the radioactivity concentration guide (RCG), which was <28 dpm/m3

. Although 

the report shows that the average air concentrations from uranium dust were less than 10% of the 

RCG, an intake rate based on 10% of the RCG is considered bounding and should be assigned to 

nonradiation workers. For the International Commission on Radiological Protection (ICRP) 

Publication 66 breathing rate of 1.2 m3

/hr for an 8-hour workday (ICRP 1994), this is 184 dpm per 

calendar day. 

Table 4-1. Occupational exposure intakes for uranium dust. 

Job description or location

Air concentration

(dpm/m3)

Intake rate

(dpm/workday)

Intake rate

(dpm/calendar day)

Secretary 8.5 81.6 55.9

Draftsman 11 105.6 72.3

Office 8.5 81.6 55.9

Locker room for lunch period 20 192 131.5

Locker room for clothes changes 18 172.8 118.4

10% of the RCG 28 268.8 184.1
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5.0 INTERNAL DOSE

The source term at RMI consisted of depleted uranium, normal uranium, slightly enriched uranium, 

recycled uranium contaminants, and thorium. Uranium urinalyses were performed, but it appears that 

urinalysis for thorium was never performed at the Extrusion Plant. Some in vivo chest counts are 

available beginning in 1968. Uranium intakes should be assessed based on bioassay data. Thorium 

and recycled uranium contaminant intakes are derived from the uranium intakes. Because of the 

tendency of technetium to become airborne more readily than uranium (DOE 2004), chronic 99Tc 

intakes might have higher than the intake that would be derived using Table 5-5 later in this section. 

Therefore, if in vivo monitoring records indicate 99Tc lung count results greater than the in vivo counter 

minimum detectable activity (MDA) of 0.5 μCi, then in vivo monitoring records should be used to 

assess 99Tc intakes.

5.1 URANIUM BIOASSAY DATA 

Uranium urinalyses for workers who RMI judged to be exposed were performed quarterly at the 

Extrusion Plant beginning in 1962. By 1963, urine sampling was performed quarterly for production 

and maintenance workers and semiannually for salaried personnel. Special studies involving urine 

sampling were performed when new processes were started. The RMI site investigation level for 

uranium in urine was 50 μg/L until 1985 when it was changed to 15 μg/L. Work restrictions were 

implemented if urinalysis results exceeded 30 μg/L and continued until repeated analysis indicated 

levels less than 15 μg/L. By 1997, the investigation level was decreased to 1 μg/L (Henderson 1997).

From 1962 through 1964, bioassay consisted of uranium photofluorimetry urinalyses by HASL. From 

1965 to the early 1970s, uranium urinalyses were by Tracerlab of Waltham, Massachusetts, also 

using photofluorimetry methods. For most of the 1960s, three spot urine samples (the first submitted 

on a Monday morning before work, the second on the following Friday morning before work, and the 

third on the following Monday morning before work) were obtained every 3 months from personnel 

judged to have significant exposure potential. Starting sometime in the mid-1970s until 1988, total 

uranium urinalysis was conducted by the United States Testing Company (USTC) of Richland, 

Washington, using a mass-based uranium measurement technique. Site documentation suggests 

that single urine voidings were obtained at an onsite restroom on two consecutive Mondays every 

3 months until 1986 (Manninen 1986). Beginning in 1986, workers were provided sample containers 

to take home during the weekend instead of providing urine samples in the potentially contaminated 

restroom on return to work on Monday morning. Also beginning in 1986, workers provided a 

“simulated 12 hour [sic] sample” that consisted of all urine passed by the employee between 2 hours 

before bedtime and 0.5 hours after waking the next morning (Manninen 1986). From 1988 to 2006, 

uranium urinalysis was performed by several vendors including Controls for Environmental Pollution 

(CEP) of Santa Fe, New Mexico; Quanterra and Severn Trent Laboratories in Richland, Washington; 

and General Engineering Laboratories in Charleston, South Carolina. Based on available information 

in claim files, all urinalysis vendors used a mass-based uranium measurement technique, and results 

are presented in either milligrams or micrograms of uranium per liter. Table 5-1 lists the Extrusion 

Plant uranium urinalysis vendors and detection levels based on a review of claimant records and site 

documentation.

According to the 1990 RMI license renewal application, RMI employees were required to submit 

weekly, monthly, or quarterly urine samples depending on work assignments (Marsh 1990). 

According to the 1992 RMI urine bioassay program description, bargaining unit personnel were 

sampled monthly and salaried personnel were sampled quarterly (Gammon 1992). According to the 

RMI ES&H Director, starting in January 2004, the only urine bioassay samples submitted by workers 

were preemployment, termination, and “for cause” whenever an intake was suspected. By January 

2004, work at the site consisted primarily of soil and ground-water remediation. Intakes for personnel 

who might not have been monitored after 2003 are presented in Section 5.5.
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Table 5-1. Uranium urinalysis vendors and detection levels. 

Analysta Year(s)

Detection threshold 

concentration (μg/L) Reference

HASL 1962–1964 2.0 AEC 1964

Tracerlab and NLO 1965–1972 1.0 RMI 1967

USTC Early 1973–1983 3.0 RMI 1974

USTC 1985 5.0 Manninen 1988

USTC 1986 0.1 Manninen 1988

USTC 1987 0.5 Manninen 1988

USTC, CEP, Quanterra, Severn Trent, 

and General Engineering Laboratories 

1988–2003 0.5 or less CEP 1991 and 

claim file records

a. CEP was the RMI vendor from 1989 through 1993. In April 1994, Sandia National Laboratory stopped using CEP’s 

bioassay services because quality control testing had raised questions about the reliability of CEP urinalysis results 

(NRC 1994). CEP results are considered invalid for the purpose of this Project, and CEP analysis should not be used 

for any year.

5.2 IN VIVO MONITORING

Beginning in 1968, RMI began using the results of either in vivo lung counting or uranium urinalysis to 

determine internal exposures, although site documentation indicates a much greater reliance on lung 

counts for demonstrating regulatory compliance. Annual in vivo chest counts for total uranium and 

enriched uranium were performed by an unidentified vendor (probably the Y-12 Plant in vivo mobile 

counter) in 1968, 1969, and 1971 through 1985. From 1968 through 1978, lung count results are 

reported in micrograms of 235U, milligrams of total uranium, grams of potassium, and nanocuries of 

137Cs on an “Invivio [sic] Radiation Monitoring Report” for individual workers at the plant. There is no 

evidence of occupational intakes of 137Cs at the Extrusion Plant, so no dose of record should be 

associated with these measurement results in any year. Table 5-2 lists general information about the 

detection capabilities of in vivo lung counting at the Extrusion Plant for various periods. Table 5-3 lists 

codes with their interpretations. A review of available records indicates that from 1979 through 1985 

some in vivo lung-counting records also contain results for thorium activity inferred from the 228Ac 

and/or 212Pb lung activity. However, the Y-12 mobile counter did not provide adequate detection 

sensitivity for either transuranic nuclides or thorium until 1986 (ORAUT 2016a), by which time it was 

no longer used at the Extrusion Plant. According to correspondence about the RMI radiation 

protection program review by PNL, evaluation of the previous whole-body counting data revealed 

“mostly problems and questions” (Munson 1985). The action level for recounts and work restrictions 

could not be determined.

Table 5-2. In vivo lung measurement types and detection levels for various 

periods.

a

Period Equipment Radionuclide MDAb,c

1968–1985 Y-12 mobile counter Total uranium 4 mg 

1968–1985 Y-12 mobile counter Enriched uranium (2% U-235) 0.1 mg

1968–1985 Y-12 mobile counter Depleted uranium 4 mg 

1968–1985 Y-12 mobile counter Np-237 200 pCi

1968–1985 Y-12 mobile counter Tc-99 0.5 μCi

1968–1985 Y-12 mobile counter Th-232 6 mg

1986–1995 Helgeson counter Total uranium 2–4 mg

1986–1995 Helgeson counter Enriched uranium 0.04–0.07 mg

a. Adapted from ORAUT (2012).

b. The Tc-99 MDA is based on McDougal (1980).

c. The Th-232 MDA is based on Scott et al. (1969) and ORAUT (2012).

Beginning in 1986, Helgeson Scientific was contracted to perform annual in vivo lung counts for 

natural uranium and 235U because of their greatly improved sensitivity (2.5 nCi) in comparison with the 

Y-12 mobile counter (Manninen 1988). March 1990 site correspondence indicates that the lung count 
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 27 of 86

 

results from Helgeson indicated system bias and that a background level would need to be 

established to test the data for significance at a 95% confidence level (Aldridge 1990). 

Correspondence containing corrected lung count data (dated October 14, 1997) from Helgeson 

Scientific to RMI indicated that Helgeson reports from 1986 to September 1996 have headings that 

were incorrectly labeled milligrams of uranium rather than nanocuries of uranium, and that the percent 

of the annual limit on intake calculation was low by a factor of 1.4 (Helgeson 1997). A review of 

several claim files indicates that these corrections were made to records in the available claim files. 

From a review of claim files, it appears that annual lung counting was discontinued around 2002 when 

Helgeson ceased operation.

Table 5-3. In vivo record codes. 

Form 

identification

Measurement 

type Code Interpretation

In Vivo Radiation 

Monitoring Report

Lung F/B Ratio This is a measure of how close to the front or 

back the internal contamination is. A ratio of 

greater than 1 could indicate external 

contamination.

In Vivo Radiation 

Monitoring Report

Lung A. Enriched 

Uranium

The maximum U-235 enrichment at the Extrusion 

Plant was 2.1%.

In Vivo Radiation 

Monitoring Report

Lung J. NLO Uranium Refers to the special spectrum region of interest 

for NLO, early operator of the FMPC.

Helgeson Lung n-u handwritten on 

forms with a 

spectral print out

Natural uranium result usually with corresponding 

handwritten result in nanocuries with 2-sigma 

uncertainty.

Helgeson Lung U-235 handwritten 

on forms with a 

spectral print out

U-235 (enriched uranium) result usually with 

corresponding handwritten result in micrograms 

with the 2-sigma uncertainty.

5.3 INTAKE ASSUMPTIONS

5.3.1 Uranium

Uranium intakes are assumed to be type M or S because the processes at the Extrusion Plant were 

similar to those at the Adrian Plant and receipt of material in metal or billet form. The predominant 

form of uranium that was processed at the site is uranium metal, and no uranium hexafluoride or 

uranium nitrate materials (type F) were received. Further, the in vitro dissolution studies of 

compounds at uranium facilities have shown that oxides of uranium can exhibit moderate solubility, 

which indicates absorption type M based on Eidson (1994). 

For calculating annual organ doses, the uranium intake (in disintegrations per minute) can be 

assumed to be entirely 234U. Because most of the uranium came from Fernald, the ratio of the 

recycled uranium contaminants to uranium is based on Section 5.5.1.4 of ORAUT-TKBS-0017-5, 

Feed Materials Production Center – Occupational Internal Dose (ORAUT 2016a). Recycled uranium

intakes should be assigned by applying the contaminants and ratios from the table containing 

contaminant intakes per unit activity of uranium. The selected material types for the contaminants

should be assigned using the direction in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 

2014b). 

5.3.2 Thorium

In vitro analyses to estimate thorium exposures at the Extrusion Plant are unavailable, and thorium 

activity determined from chest counts was not reported until 1979, after the reported thorium 

processing period. Table 2-7 indicates that the thorium received by RMI in any year was less than 1% 

of the uranium received by mass. Because there are no specific process and monitoring records for 
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thorium use, this document assigns a U:Th mass intake ratio as 0.01 during the thorium production 

period, which is assumed to be from January 1, 1962, through December 31, 1971. See Table 2-7. 

The specific activity of normal uranium is a factor of 1.7 larger than that of depleted uranium, but on 

the order of a factor of 2.4 smaller than that of 2%-enriched uranium. As a consequence, it appears 

to be favorable to claimants to assume natural uranium when determining the relative activity of 

thorium based on the relative mass comparisons. 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 a 232Th:U mass intake ratio of 1%. This results in a relative 232Th:U activity intake ratio of 0.00161. 

It is assumed that the thorium is natural because this the dose from this form is more favorable to 

claimants. Therefore, it is assumed that the 232Th, 228Ra, 228Ac, 228Th, 224Ra are in secular 

equilibrium and an equivalent intake should be assigned for each. This 1% factor overestimates 

232Th:U mass ratio for the production years and could be an order of magnitude higher for some 

years (ORAUT 2016c). Therefore, exposures from thoron are accounted for with the application of 

this 1% factor for thorium. 

Thorium was not processed in 1965, 1968, or after 1971 (Britcher 1992). Site documentation 

indicates that RMI cleaned up after thorium extrusions to prevent commingling of thorium and 

recyclable uranium (Jefferson 1962a), but it is uncertain if the clean-up included all areas of the plant. 

Even though the thorium was cleaned up before uranium extrusions, it is favorable to the claimant to 

assume that residual contamination resulted in thorium intakes in 1965 and 1968 when no thorium 

processing was recorded. No thorium exposure should be assumed for periods after 1971.

5.4 INTERNAL DOSE ASSIGNMENT SUMMARY

Depleted, natural, and slightly enriched uranium were all source terms at the Extrusion Plant and are 

assumed to be types M or S. For results in mass units, it is favorable to claimants to assume 

exposures to slightly enriched uranium (2.0% 235U by mass) if no better information is available. If 

uranium urinalysis measurements were made in both mass and activity units, the activity units should 

be used to calculate intakes and 234U can be assumed for uranium dose calculations.

Uranium urinalysis, when available, should be used to estimate uranium intakes; the default detection 

thresholds are 3 μg/L for samples analyzed between 1962 and 1984, 5 μg/L in 1985, and 0.5 μg/L 

from 1986 to 2003. It should be noted that the forms used to record uranium urinalysis results at the 

Extrusion Plant had a column labeled “Mg/L” (which meant milligrams per liter rather than megagrams 

per liter); nevertheless, the results were sometimes recorded in micrograms per liter.

It should be noted that bioassay data from 1989 to 1993 analyzed by CEP are not to be used in any 

internal dose assessment. For individuals with potential intakes during that period, the following 

priorities should be used for assessing uranium intake rates:

1. Urinalysis data after 1993. These can be used to bound the earlier years.

2. Body counts during 1989 to 1993.

3. Urinalysis data before 1989 (extension of the dose calculation beyond the last bioassay 

result).

4. Data from an Extrusion Plant coworker with the same job (following the same priorities as 

above).

5. Unmonitored uranium intakes from a similar site (Bridgeport Brass, Adrian facility).
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A 232Th:U mass intake ratio of 1% is assumed. This results in a relative 232Th-to-uranium intake 

activity ratio of 0.00161. Based on the assumption that the thorium is natural, the 232Th, 228Ra, 228Ac, 

228Th, 224Ra are in secular equilibrium and an equivalent intake should be assigned for each.

Thorium was not processed in 1965, 1968, and after 1971 (Britcher 1992); site documentation 

indicates that RMI cleaned up after thorium extrusions to prevent commingling of thorium and 

recyclable uranium (Jefferson 1962a), but it is uncertain whether the clean-up included all areas of the 

plant. Even though the thorium was cleaned up before uranium extrusions, it appears favorable to the 

claimant to assume that residual contamination resulted in thorium intakes in 1965 and 1968 when no 

thorium processing was recorded. No thorium exposure should be assumed after periods after 1971.

5.5 UNMONITORED INTERNAL DOSE

Breslin and Glauberman (1964, pp. 14–16) reported the daily average exposures of operating 

personnel. Air samples were obtained at locations selected to characterize workers’ exposures, and 

both breathing zone and general air samples were collected. Table 5-4 lists the various job

categories that were assessed. The intake rates are based on the daily average exposures (dpm/m3

) 

for operating personnel (assuming a breathing rate of 9.6 m3

/d for an 8-hour day). 

Table 5-4. Occupational exposure intakes from uranium dust.

a

Job description

Air concentration

(dpm/m3)

Intake rate

(dpm/workday)

Intake rate

(dpm/calendar day)

Heater 40 384 263.0

Heater Helper 40 384 263.0

Press Operator 63 604.8 414.2

Die Head Man 83 796.8 545.8

Extrusion Puller 150 1,440 986.3

Saw Man 280 2,688 1,841.1

Stamper 410 3,936 2,695.9

Inspector (Rod) 1,200 11,520 7,890.4

Oil Bath and Roll Straightener 96 921.6 631.2

Billet Inspector 46 441.6 302.5

Packer 54 518.4 355.1

Jeep Driver 60 576 394.5

Tool Crib 83 796.8 545.8

Scale Clerk 29 278.4 190.7

Scrap Handler 81 777.6 532.6

Maintenance 110 1,056 723.3

Engineering 33 316.8 217.0

Foreman 100 960 657.5

Head Foreman 60 576 394.5

Health and Safety 48 460.8 315.6

a. (ORAUT 2016d). 

5.6 INTERNAL DOSE DURING THE POSTPRODUCTION PERIOD

During the postproduction period at the Extrusion Plant (after October 14, 1991), uranium extrusion 

activities, along with the handling of bulk uranium metal for extrusion, had ceased. Worker exposures 

and intakes occurred from residual uranium contamination on structures and components and in soil 

during characterization surveys and sampling, equipment dismantlement and removal, 

decontamination, demolition, and waste packaging and shipping activities. Section 2.8 contains a 

summary of worker monitoring changes that occurred during the postproduction period at the 

Extrusion Plant (after October 14, 1991), but based on a review of available site documentation such 

as personnel dosimetry files and claim files, it is not evident that site personnel from the remediation 
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 30 of 86

 

contractor (LATA-SHARP Remediation Services), or even RMI, participated in an internal or external 

monitoring program after 2003. As stated in Section 2.8, by January 2004 building decontamination 

was complete and work at the site consisted primarily of soil and groundwater remediation.

To calculate the intakes from soil to unmonitored site personnel during soil remediation activities that 

began in 2004, this analysis assumed that uranium exposure was to the average soil concentration 

from locations within the main plant area and the property to the east of the main plant at a level of 

350 pCi/g total uranium and to the maximum 99Tc concentration of 49 pCi/g within the main plant area 

(RMI 1996, pp. 19–29). Using a mass loading factor of 2 × 10-4 g/m3 and a breathing rate of 

2,400 m3

/yr, the calculated annual uranium inhalation intake was 168 pCi and the annual 99Tc intake 

was 24 pCi. The daily inhalation intakes for total uranium and 99Tc are 0.460 pCi and 0.065 pCi. It is 

assumed that other recycled uranium contaminants would be a small fraction (see Table 5-4). The 

mass loading factor of 2 × 10-4 g/m3 takes into account short periods of high mass loading and 

sustained periods of normal activity on a typical farm (Yu et al. 2001).

According to NIOSH (2004), 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, which results in the daily 

ingestion of 0.014 pCi of uranium and 1.96 × 10-3 pCi 99Tc. 

An estimate of the uncertainty that is associated with soil resuspension intakes has been made by 

assuming (1) that the soil concentrations are lognormally distributed, (2) that the average uranium soil 

concentration (350 pCi/g) can be used to underestimate the 50th-percentile concentration, and 

(3) that the maximum uranium soil concentration ever measured at the plant (2,600 pCi/g) represents 

the upper 95th percentile (RMI 1996). The resultant geometric standard deviation is 3.4.
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6.0 EXTERNAL DOSE

RMI employees were exposed to radiation from uranium, thorium, and their short-lived progeny. This 

document assumes that photon energies were in the 30- to 250-keV range, which is favorable to 

claimants when considering both organ dose conversion factors and radiation effectiveness factors. 

Shallow or open-window dose is assumed to be from electrons with energies greater than 15 keV.

From 1962 to the mid-1980s, claim files contain a handwritten Film Badge Assignment Sheet 

containing a name, badge number (typically between 1 and 80 assigned by RMI), and notes or 

comments such as lost film, damaged film, inverted film, quit, terminated, retired, vacation, wore two 

badges in monitoring period, and issue date (if not on the first day of the month or quarter). The next 

page(s) contain the dosimeter vendor report with the badge number corresponding to the Film Badge 

Assignment Sheet handwritten next to the appropriate results. Beginning in 1986, RMI personnel 

names are specifically provided in the vendor reports, and there either is no Film Badge Assignment 

Sheet or there is a database printout (1987 through 1996) that contains personnel names and 

external dosimetry data. Some claim files contain a Terminated Employee Radiation Exposure 

Report containing skin and whole-body dose for the employment period. In addition, some claim files 

contain Form AEC-5 (beginning in the early 1960s), later Form NRC 5. All available external 

dosimetry data, was evaluated for and stripped of duplicates, and is summarized in several tables in 

Attachment E. An evaluation of the almost 10,000 lines of data at the 95th and 50th percentiles was 

also given for gamma and beta doses.

6.1 PENETRATING DOSES

6.1.1 Gamma Dose

Table 6-1 lists available information on badges used from 1962 to 2004; whole-body external 

dosimetry results appear to be present in claimant files. There are also some extremity badge results 

beginning in 1974.

Routine film badge results included whole-body beta and gamma monitoring results. Table 6-1 

provides some of the available information on the monitoring methods, detection limits, and reporting. 

6.1.2 Neutron Dose

Neutron measurements began with the use of the five-element TLD in January of 1986 (badges were 

received in late December of 1985). Neutron doses appear to have been monitored for most RMI 

employees in 1986, 1987, and through the third quarter of 1988. No basis was given for starting or 

ending the neutron measurements. There were no requests for area neutron surveys and none were 

conducted. A Pacific Northwest National Laboratory audit of the RMI health physics program 

(Munson 1985) did not suggest or indicate there was a concern about neutron exposure. It appears 

the monitoring of neutron dose was initiated because of the ability of the new five-element TLD. 

The neutron measurements appear to be invalid, or the results are spurious due to crosscontamination of the TLD chips in the dosimetry badge by other contaminated badges. See Summary 

for Extrusion Plant (RMI) Research, High 1986 Results (ORAUT 2015), which describes the research 

that was conducted and which concluded that the neutron results were not valid and likely the result of 

improper handling of the badges, which caused them to be contaminated. In 1987, contamination 

control practices were used to place the badges in separate thin plastic bags. The photon dose 

results then dropped back to the 1985 levels. Another reason for not considering the neutron 

measurements valid is the lack of the sensitivity algorithm that was not added until 1988. The lack of 

this added sensitivity feature might have allowed for erroneous neutron and gamma results to go 

unchecked, possibly contributing to the neutron dose signal that was measured by the algorithm.
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Table 6-1. External dosimetry 

Period

External 

dosimetry vendor

Exchange frequency/ 

monitoring method Detection threshold reported in records

1962–

1966

Controls for 

Radiation, 

Cambridge, 

Massachusetts

Monthly/film A zero indicates less than minimum detectable 

dose of 5 mrem for X-ray and gamma <175 keV 

or 10 mrem for >175 keV X-ray and gamma and 

beta.

a

1967–

1987

USTC, Richland, 

Washington

Quarterly/film until 

TLDs in 1986

1972 and 1976 film badge reports indicate film 

detection limits of 10 mrem for X-ray, 20 mrem 

for gamma ray, and 20 mrem for beta.

b 

According to a 1977 film badge report 

calibrations for gamma performed with Cs-137, 

beta performed with Sr-90, and X-ray performed 

with 16- to 17-keV photons.

1983, 1984,

c and 1985d film badge reports 

indicates film detection limits of 10 mrem for 

photons and 20 mrem for beta.

1986 and 1987e TLD badge reports indicate 

detection limits of 10 mrem for X-ray, gamma, 

beta, and neutron radiation.

1988–

1991

USTC and WMCO

at FEMPf 

Monthly/TLD 1988 and 1989 TLD badge reports from USTC 

indicate detection limits of 10 mrem for X-ray, 

gamma ray, and beta radiation. The minimum 

detection limit for FEMP TLDs was 5 mrem 

gamma and 10 mrem beta from 1989 to 1992.g,h

1992–

2004

WMCO at FEMP, 

TMA Eberline, 

ThermoNutech, 

Landauer, ICN, and 

Global Dosimetry 

Solutions

Probably 

quarterly/TLD

TLD badge reports typically indicate detection 

limits of 10 mrem for X-ray, gamma ray, and 

beta, but information was not found for every 

vendor’s badge type. Site documentation 

implies that TLD vendors were DOE Laboratory 

Accreditation Program or National Voluntary 

Laboratory Accreditation Program certified from 

1989 to 1992.

a. CR (1964) 

b. USTC (1974–1982).

c. USTC (1984).

d. USTC (1985).

e. USTC (1986), Mason (1988).

f. Mason (1988).

g. ORAUT (2014a) 

h. ORAUT (2013)

Further, a review of the sources and source term at the Extrusion Plant did not indicate there were 

any sealed sources for neutrons or neutron generating equipment (ORAUT 2015). No records have 

been found that indicate the source of neutron exposure. In addition, Bridgeport Brass the 

predecessor to the Extrusion Plant conducted a coworker study and considered the neutron 

exposures minimal or negligible. Neutron exposures at the Extrusion Plant are also assumed to be 

minimal or spurious for those years when neutrons were monitored. The neutron doses were added 

into the photon or deep penetrating doses, these practices are considered favorable to claimants 

(ORAUT 2013; 2016e). Therefore, no additional neutron doses should be assigned based on the 

neutrons being included into the dosimetry reports for penetrating dose.
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6.1.3 Historical Annual Exposures for Penetrating Radiation

Table 6-2 summarizes RMI-reported external penetrating doses for 1962 through 1999 and provides 

a general sense of the program size, although the number of persons monitored appears to include 

visitors and temporary workers beginning in the 1980s. The doses in Table 6-2 do not include 

nonpenetrating and extremity doses. The term “no data” means there were no reported doses for 

that period. For 1962 through 1973, the bounding dose for unmonitored workers is assumed to be 1 

R/yr based on Table 6-2. 

Table 6-2. Penetrating whole-body external doses, 1962 to 1999.

Year

Number not 

monitored

Number 

monitored

Dose (rem)

0–1

Dose (rem)

1–2 Reference

1962 7 52 51 1 Sapirie 1963

1963 9 71 70 1 Sapirie 1964

1964 14 81 81 No data Sapirie 1965

1965 8 66 66 No data Sapirie 1966

1966 27 57 57 No data Sapirie 1967

1967 28 65 65 No data Lenhard 1968

1968 31 62 62 No data Hibbitts 1969

1969 33 41 41 No data Smith 1970

1970 25 47 47 No data Lenhard 1971

1971 24 46 46 No data Lenhard 1972

1972 20 46 46 No data Travis 1973

1973 27 53 53 No data Travis 1974

1974 Unknown 69 69 No data Van Loocke 1975

1975 Unknown 62 62 No data Heiser 1976

1976 Unknown 58 58 No data Schaeffer 1977

1977 Unknown 62 62 No data Schaeffer 1978

1978 Unknown 68 68 No data Schaeffer 1979

1979 Unknown 80 80 No data Van Loocke 1980

1980 Unknown 80 80 No data Schaeffer 1981

1981 Unknown 89 89 No data Van Loocke 1982

1982 Unknown 116 116 No data Schaeffer 1983

1983 Unknown 118 118 No data Schaeffer 1984

1984 Unknown 117 117 No data Schaeffer 1985

1985 Unknown 124 124 No data Manninen 1988

1986 Unknown 134 130 4 Manninen 1988

1987 Unknown 822 822 No data Brewer 1988

1988 Unknown 920 920 No data RMI 1989b

1989 Unknown No data No data No data Not applicable

1990 Unknown 282 282 No data Rizzi 1991

1991 Unknown No data No data No data Not applicable

1992 Unknown No data No data No data Not applicable

1993 Unknown No data No data No data Not applicable

1994 Unknown 332 332 No data TMA 1995

1995 Unknown 255 255 No data TMA 1996

1996 Unknown 285 285 No data TN 1997

1997 Unknown 290 290 No data TN 1998

1998 Unknown No data No data No data Not applicable

1999 Unknown 1,202 1,202 No data Eberline 2000

Dosimeters consisted of film badges for beta, X-ray, and gamma radiation from 1962 through 1985. 

TLDs replaced film in 1986. In the early period of Extrusion Plant operations, a strip of indium metal 

foil was placed in badges for criticality monitoring. In 1986, 1987, and 1988, neutron monitoring was 
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 34 of 86

 

performed in addition to beta, X-ray, and gamma monitoring. Dosimetry badges were exchanged 

monthly from 1962 through 1966 and quarterly or monthly thereafter.

Film badge detection thresholds for beta and gamma are based on information provided by film badge 

vendors in Table 6-1 and ORAUT-OTIB-0010, A Standard Complex-Wide Correction Factor for 

Overestimating External Doses Measured with Film Badge Dosimeters (ORAUT 2006a). The 0.010 

rem given by the film badge vendor from 1962 to 1966 is not in line with guidance from OTIB-0010 

and therefore, to be consistent with the values given later, the limit of detection for the earliest period 

of operation is 0.030 rem and from 1967 to 1985 was estimated to be 0.020 rem per exchange period. 

The beta-gamma TLD from Fernald has a detection threshold of 10 mrad beta and 5 mrad gamma 

(ORAUT 2014a). Therefore, the value was set at 0.010 rem for both beta and gamma through 2004. 

The overall uncertainty in recorded dose is dependent on (1) administrative practices, (2) dosimetry 

technology, (3) calibration, and (4) workplace radiation fields. The Extrusion Plant used a variety of 

dosimetry vendors from 1962 to 2004, and the precise details of dosimeter type and calibrations are 

not available; consequently, uncertainty estimates are based on ORAUT (2005b). The uncertainty 

associated with beta-gamma film badge results from 1962 through 1985 are estimated to be ±30%, 

and the uncertainty associated with beta-gamma TLD badge results from 1986 through 2004 are also 

estimated to be ±30%.

Tables 6-2 through 6-4 are for information purposes and are based on the upper bound of reported 

doses received at the Extrusion Plant. The doses are listed in Table 6-3 for 1974 to 1984 and Table 

6-4 for 1994 to 1999.

Table 6-3. Upper bound of reported whole-body deep doses, 1974 to 1984. 

Year

Number 

monitored

No 

measurable 

data

<0.100 

(rem)

0.100–

0.249 

(rem)

0.250–

0.499 

(rem)

0.500–

0.749 

(rem) Reference

1974 69 22 27 17 3 No data Heiser 1975

1975 62 19 31 11 1 No data Heiser 1976

1976 58 30 20 8 No data No data Schaeffer 1977

1977 62 26 36 No data No data No data Fletcher 1978, p. 167

1978 68 16 21 26 5 No data Schaeffer 1979

1979 80 14 45 19 2 No data Van Loocke 1980

1980 80 16 32 28 4 No data Schaeffer 1981

1981 89 34 40 15 No data No data Fletcher 1978, p. 167 

1982 116 20 51 34 10 1 Schaeffer 1983

1983 118 32 39 32 15 No data Schaeffer 1984

1984 117 23 48 33 13 No data Schaeffer 1985, p. 134

1985 124 38 58 22 4 2 Theisen 1986 

1987 129 30 74 21 3 1 Theisen 1988

1988 117 45 68 4 No data No data Theisen 1989

Table 6-4. Upper bound of reported whole-body deep doses, 1994 to 1999. 

Year

Number 

monitored

<0.010 

(rem)

0.01–0.099 

(rem)

0.100–0.249 

(rem)

95th 

percentile Reference

1994 330 291 38 1 0.099 TMA 1996

1995 255 220 35 No data 0.099 TN 1996

1996 285 241 44 1 0.099 TN 1997

1997 290 271 19 No data 0.099 TN 1998

1998 No data No data No data No data No data Not applicable

1999 1,202 1,170 32 No data <0.010 Eberline 2000
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Table E-4 or E-5 should be used to assign external doses for either workers or construction trade 

workers (CTW) for the rare occasion when they are unmonitored CTW.

6.1.4 Missed Doses

Missed dose occurs when the dose of record is zero because the interpreted dose had a negative 

bias, dosimeter response was less than the minimum detectable level, or there was no dose of record 

for an assigned badge for a badge cycle. Estimates for missed dose can be assigned from Table 6-5 

considering that the maximum potential annual missed dose is the 95th-percentile and missed dose is 

reported at the median (NIOSH 2007).

Table 6-5. Missed external doses (rem).

Monitoring 

period

Penetrating and 

nonpenetrating LOD

Exchange 

frequency

Maximum potential annual missed 

penetrating and nonpenetrating dose

1962–1966 0.030 Monthly 0.360

1967–1985 0.020 Quarterly 0.080

1986–1988 0.020 Quarterly 0.080

1989–1991 0.010 Monthly 0.120

1992–2003 0.010 Quarterly 0.040

6.2 NONPENETRATING DOSE

Because film badge and or TLD data are available for most of the period from 1962 through 1991, 

dose should usually be assigned based on the dose of record. Nonpenetrating dose associated with 

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

For those workers who were not monitored or with unexplained gaps in their monitoring, the dose 

reconstructor should assign doses based on Table E-4 or E-5 for construction trade workers. 

6.3 EXTREMITY MONITORING

The earliest extremity badge results were for wrist badge assignments to 12 personnel from July to 

September 1, 1974. Quarterly wrist badge results were also noted in December 1983 in two claim 

files. Finger rings appear to have replaced wrist badges in 1986. Monthly finger ring results (left and 

right) were present for 1986. Extremity doses were also noted in database output forms in the claim 

files beginning in 1987 and on some AEC-5 Forms in claim files.

According to the 1974 annual health protection appraisal, RMI was advised to begin immediate 

monitoring of extremity radiation exposures to workers who closely handled or inspected uranium 

billets to determine if continuous extremity monitoring would be required. This was in anticipation that 

the AEC would soon revise the radiation standard for forearm exposures to conform to a National 

Council on Radiation Protection and Measurements recommendation to decrease allowable 

exposures from 75 to 30 rem/yr. Several employees were expected to exceed 10% of the anticipated 

30-rem limit that would require monitoring (Jelinek 1974). The use of wrist film badges was initiated in 

July 1974. The annualized results of a 6-month study of skin doses to the forearm and chest from 

July 1 to December 31, 1974, are listed in Table 6-6. According to correspondence from RMI to the 

U.S. Energy Research and Development Administration (a DOE predecessor agency) in January 

1977, ring badge dosimeters had not yet been used at the Extrusion Plant to assess hand doses, but 

RMI estimated that hand exposures would run as high as 10 to 15 rem/yr for four to six workers at the 

plant (Van Loocke 1977).

A study was performed in November 1985 to determine the ratio between the fingertip dose and the 

TLD ring dose on workers whose inspection tasks required them to run their bare fingers across the 
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 36 of 86

 

Table 6-6. Forearm and chest skin dose monitoring 

results for personnel who closely handled or 

inspected uranium billets in 1974.

a

Job/task

Wrist badge

(mrem/yr)

Chest badge

(mrem/yr)

Lathe Operator 100 1,120

Billet Inspector 2,460 3,860

Lathe Operator 680 1,120

Forge Helper 5,120 5,420

Forge Helper 2,080 2,600

Roll Operator 360 2,500

Runout Table 820 1,900

Salt-Bath Operator 1,320 460

Forge Inspector 8,900 4,660

Saw Operator 820 1,960

Extrusion Inspector 3,020 4,580

Forge Inspector 3,640 5,580

a. Johnson (1976).

surface of uranium metal. Calibrated tissue-equivalent dosimeters were exposed on a hand phantom 

on a uranium billet at the plant. The results indicated that a properly worn finger ring would respond a 

maximum of 60% lower in comparison with a fingertip in contact with a billet. The study showed that a 

finger ring dosimeter worn on the inside of the little finger was as accurate as one worn on the inside 

of the finger ring. Finger rings worn on the outside of the hand significantly under-responded (Munson 

and Stacy 1985).

For most work categories, if extremity monitoring is not available, it is reasonable to estimate 

extremity entrance dose as being equal to the shallow dose measured by the chest badge result. For 

workers whose job category was inspector, the factor of 3 (derived from the ratio of the 1974 Salt Bath 

Operator’s annual chest badge exposure to his annual wrist badge exposure), which is more 

favorable to the claimant, should be applied. Based on the Munson and Stacy (1985) study, finger 

doses could be a factor of 1.67 larger or more, which would result in a whole-to-finger ratio of 5.0. 

6.4 MISCELLANEOUS INFORMATION CONCERNING EXTERNAL EXPOSURE

Early documentation from the Extrusion Plant indicates that whole-body film badges were to be 

displayed in the open and not obscured (RMI 1973), the results of a May 1985 independent health, 

safety, and environmental review at the Extrusion Plant indicated that film badges were 

inappropriately being placed into oversized plastic wrap to prevent them from becoming 

contaminated. By June 4, 1985, this was corrected by placing the badges into a properly sized heatsealed plastic holder (Row 1985).

Personal contamination surveys were nonexistent or lax until 1986.

No Extrusion Plant site documentation was found that indicates industrial radiography sources were 

used at the facility.

6.5 EXTERNAL DOSE ASSIGNMENT SUMMARY

Many Extrusion Plant employees were monitored for beta, X-ray, and gamma external exposures. 

The dose reconstructor should assign the gamma doses as 30-250 keV photons, acute, 100% 

anterior posterior geometry, unless to lung, esophagus, or red bone marrow in which case isotropic 

and rotational geometries must be evaluated. All electron or beta doses are assigned as greater than 
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 37 of 86

 

15 keV. All monitored and unmonitored doses are assigned in IREP as a constant distribution. Missed 

doses for monitored workers are assigned as a lognormal distribution in IREP.

For those employees not monitored sufficiently or unmonitored, the data in Attachment E summarizes 

the dose assignment values for unmonitored workers and construction trade workers. Dose 

reconstructors should assign dose to unmonitored workers with minimal potential for radiation 

exposure from RMI operations using the 50th-percentile dose. Electron (beta) doses are all >15 keV. 

Dose reconstructors should exercise judgment about the energy employee’s work location and duties 

to determine the probability of extremity dose. For example, administrative personnel would be 

unlikely to receive extremity dose.
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 38 of 86

 

7.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

[1] Elyse Thomas. Oak Ridge Associated Universities (ORAU) Team. Principal Medical 

Dosimetrist. May 2014.

Review of claim file records from Extrusion Plant.
Document No. ORAUT-TKBS-0056 Revision No. 01 Effective Date: 03/07/2017 Page 39 of 86

 

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[SRDB Ref ID: 160888]
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Schaeffer, M. R., 1981, “Annual Occupational Radiation Exposure Report – 1980,” letter to H. D. 

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GLOSSARY

accreditation

For external dosimetry, the assessment of whether or not a personnel dosimetry system meets 

specific criteria. The assessment includes dosimeter performance and the associated quality 

assurance and calibration programs.

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

decontamination

Reduction or removal of radioactive material from a structure, area, object, or person. 

Decontamination can occur through (1) treating the surface to remove or decrease the 

contamination or (2) allowing natural radioactive decay to occur over a period of time.

deep dose equivalent

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose. 

depleted uranium

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium.

DOE Laboratory Accreditation Program (DOELAP)

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rad, rep, or grays.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See thermoluminescent dosimeter. 
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dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7% by mass. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade 

uranium contains greater than 90% 235U.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. (2) X-ray film.

fission

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X rays do not originate in the nucleus.

half-life

Time in which half of a given quantity of a particular radionuclide disintegrates (decays) into 

another nuclear form. During one half-life, the number of atoms of a particular radionuclide 

decreases by one half. Each radionuclide has a unique half-life ranging from trillionths of a 

second to billions of years.

in vitro bioassay 

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay 

Measurements of radioactive material in the human body utilizing instrumentation that detects 

radiation emitted from the radioactive material in the body.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.
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natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram).

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

nucleus

Central core of an atom, which consists of positively charged protons and, with the exception

of ordinary hydrogen, electrically neutral neutrons. The number of protons (atomic number) 

uniquely defines a chemical element, and the number of protons and neutrons is the mass 

number of a nuclide. The plural is nuclei.

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

proton

Basic nucleic particle with a positive electrical charge and mass slightly less than that of a 

neutron. There are protons in the nuclei of every atom, and the number of protons is the 

atomic number, which determines the chemical element.

quality factor

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent. See dose. 

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation. 

radioactive

Of, caused by, or exhibiting radioactivity.

radioactive waste

Radioactive solid, liquid, and gaseous materials for which there is no further use. Wastes are 

generally classified as high-level (with radioactivity as high as hundreds of thousands of curies 

per gallon or cubic foot), low-level (in the range of 1 microcurie per gallon or cubic foot), 
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intermediate level (between these extremes), mixed (also contains hazardous waste), and 

transuranic.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. See radionuclide. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

roentgen

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

and negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0 degrees Celsius and standard atmospheric 

pressure. An exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher energy photons (generally greater than 100 kiloelectron-volts).

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose.

transuranic elements

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium.

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the Energy Research and 

Development Administration in 1979.

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose. 

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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ATTACHMENT A 

PLANT LAYOUTS

LIST OF TABLES

TABLE TITLE PAGE

A-1 Key for Figure A-1 ..................................................................................................................55

LIST OF FIGURES

FIGURES TITLE PAGE

A-1 Plant layout ............................................................................................................................54

A-2 Plant layout, 1957 ..................................................................................................................57

A-3 Plant layout, 1962 ..................................................................................................................58

A-4 Plant layout, 1964 ..................................................................................................................59

A-5 Plant layout, 1965 ..................................................................................................................60

A-6 Plant layout, 1968 ..................................................................................................................61

A-7 Plant layout, 1969 ..................................................................................................................62

A-8 Plant layout, 1979 ..................................................................................................................63

A-9 Plant layout, 1981 ..................................................................................................................64

A-10 Plant layout, 1982 ..................................................................................................................65

A-11 Plant layout, 1983 ..................................................................................................................66

A-12 Plant layout, 1984 ..................................................................................................................67

A-13 Plant layout, 1985 ..................................................................................................................68

A-14 Plant layout, 1986 ..................................................................................................................69

A-15 Plant layout, 1987 ..................................................................................................................70

A-16 Plant layout, 1988 ..................................................................................................................71

A-17 Plant layout, 1989 ..................................................................................................................72

A-18 Plant layout, 1990 ..................................................................................................................73

A-19 Plant layout, 1991 ..................................................................................................................74
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-1. Plant layout (RMI 1990b). 
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ATTACHMENT A

PLANT LAYOUTS (continued)

Table A-1. Key for Figure A-1. 

No. Description

1 Guard House - Contractor Emergency Assembly 

Area

2 Emergency Assembly Area

3 Hazardous Waste Storage Building

4 RF-3 Building

4A RF-3 Building Addition

5 Equipment Cleaning Area

6 Electrical Substation

7 Sewage Disposal

8 Firehouse

9 Truck Ramp Enclosure

10 Shipping and Receiving Back

11 Northeast Billet Storage Warehouse

12 Accumulator Station

13 Air Compressor

14 Substation

15 3850 Ton Loewy Horizontal Extrusion Press

16 Die Head

17A Runout Table (Movable Section)

17B Powered Runout Table

18 Horizontal Extrusion Quench Tank

19 Extrusion Cooling and Transfer Table

20 Rotating NPR Extrusion Cooling Table

21 Extrusion Transfer Conveyor To Campbell Saw

22 Abrasive Campbell Saw

23 Extrusion Transfer Conveyor From Campbell Saw

24 Inspection After Sawing

25 Transfer Table To Roll Straightener

26 Roll Straightener Entrance Conveyor

27 Roll Straightener

28 Roll Straightener Exit Conveyor

29 In Process Transfer and Storage Cable

30 Salt Bath

31 Vertical Extrusion Quench Tank

32 Sunbeam Furnace

33 FSI Furnace

34 Salt Bath Loading Area

35 Receipt Inspection Area

36 MK-31 Rinse Tank

37A Commercial Pickle Tank

37B Commercial Rinse Tank

38 Extrusion Wash Tank

39 Extrusion Pickle Tank

40 Inspection Weighing and Packing Table

41 Container Preheat Furnace

42 Toll Crib

43 Gauge Room

44 Tool Preheat Furnace

45 Prefill Tank

46 Floor Scale

47 Area Heating Furnace

No. Description

53 Women's Clean Locker Room

54 Enclosed Rampway

55 Storeroom

56 Lunchroom

57 Electrical Shop

58 Salary Change Room

59 Lab Tool Room

60 Wet Chemical Room

61 Metalographical Room

62 Conference Room

63 Computer Room

64 Chemist/Technicians Office

65 Print Room

66 Dark Room

67 Analytical Room

68 Development Lab and Quality Control Area

69 350 Ton Lombard Horizontal Extrusion 

Press and Ventilation

70 #1 Giddings and Lewis CNC Lathe

71 #2 Giddings and Lewis CNC Lathe

72 #3 Giddings and Lewis CNC Lathe

73 Small Monarch Lathe

74 Lodge and Shipley Lathe

75 Large Monarch Lathe

76 Rockwell Drill Press

77 Racine Saw

78 Cincinnati Mill Machine

79 Stanley Grinder

80 Do All Saw

81 U.S. Electrical Tool Co. Grinder

82 Cincinnati Gilbert Drilling Machine

83 Stairway Down To Boiler Room

84 Acid Neutralization Tank

85 Engleberg Belt Sander

86 Evaporator

87 Mandrel Quench Tank

88 Hazardous Waste Emergency Equipment 

Building

89 Toolmex Lathe

90 Lathe Ventilation

91 Cooley Furnace

92 Substation

93 Storeroom Area

94 RF -- 6 Cold Storage Warehouse

95 Northwest Storage Building

96 Stack #1A Press Exhaust

97 Outdoor Substation Addition

98 Stack #3A Filter Building

99 North Gisholt Lathe

100 Fan -- Stack #5A

101 Sandblaster

102 Emergency Generator
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ATTACHMENT A

PLANT LAYOUTS (continued)

No. Description

48 Engineering Office Area

49 Foremen's Office Area

50 Men's Hourly Dirty Locker Room

51 Men's Hourly Clean Locker Room

52 Women's Dirty Locker Room

108 Auxiliary Storage Area

109 South Gisholt Lathe

110 K. R. Wilson Hydraulic Press

111 Chicago Shear

112 Lift Truck Hoist

113 Stack #4Afilter Building

114 Caustic Tanks

115 Startrite Band Saw

116 Main Plant Acid, Stack #7

117 Health-Safety Technician's Office

118 Plate and Frame Filter Press

119 Wastewater Treatment Building

120 Fire Hydrant

109 South Gisholt Lathe

110 K. R. Wilson Hydraulic Press

111 Chicago Shear

112 Lift Truck Hoist

113 Stack #4Afilter Building

114 Caustic Tanks

115 Startrite Band Saw

116 Main Plant Acid, Stack #7

117 Health-Safety Technician's Office

118 Plate and Frame Filter Press

119 Wastewater Treatment Building

120 Fire Hydrant

121 Chip Chopper

122 Thermal Oxidation Tank #1

123 Thermal Oxidation Tank #2

124 Baghouse

125 HEPA Filter Housing

126 95% and HEPA Filter Housing

127 Cartridge Filter Housing

128 Fan -- Stack #1A

129 Cartridge Filter Housing

130 95% and HEPA Filter Housing

131 Fan -- Stack #3A

132 Moisture Separator

133 Cartridge Filter Housing

No. Description

103 Mezzanine

104 Forge Area Stack #8

105 Outside Air Sampler

106 Storage Area

107 Extrusion Storage Area

134 95% and HEPA Filter Housing

135 Fan -- Stack #4A

136 Clear Well

137 Pressure Filters

138 Detention Tank #2

139 Detention Tank #1

140 Processed Tank #1

141 Processed Tank #2

142 Processed Tank #3

143 Sludge Pump #2

144 Sludge Pump #1

145 Pressure Filter

146 Backwash Pumps

147 H. M. C. Furnace

148 Lumber Rack

149 Tool Coating Vent and Stack

150 Tool Grinding Booth

151 Holding Tank

152 D. C. Exciters

153 Receiver

154 Water Heater Room

155 Project Storage Cage

156 Maintenance Tool Storage

157 Incoming Water and Water Meter Pit

158 Incoming Natural Gas and Gas Meter

159 Still Pumps

160 Abrasive Saw

161 Salt Pots

162 Sheldon Lathe

163 Fisher Scientific Iso-Temp Lab Refrigerator

164 Ross Temp Icemaker

165 Cooley Furnace

166 Neytech Furnace

167 Walker-Turner Drill Press

168 Delta Band Saw

169 Pneumotive Air Compressor

170 RF -- 6 Sump

– –
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-2. Plant layout, 1957 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-3. Plant layout, 1962 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-4. Plant layout, 1964 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-5. Plant layout, 1965 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-6. Plant layout, 1968 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-7. Plant layout, 1969 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-8. Plant layout, 1979 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-9. Plant layout, 1981 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-10. Plant layout, 1982 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-11. Plant layout, 1983 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-12. Plant layout, 1984 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-13. Plant layout, 1985 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-14. Plant layout, 1986 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-15. Plant layout, 1987 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-16. Plant layout, 1988 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-17. Plant layout, 1989 SRDB (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-18. Plant layout, 1990 (RMI 1990b).
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ATTACHMENT A

PLANT LAYOUTS (continued)

Figure A-19. Plant layout, 1991 (RMI 1990b).
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ATTACHMENT B 

N-REACTOR PRODUCTION

According to RMI (1995), “the N-Reactor Production consisted of a primary extrusion process followed 

by a forging process.” The typical steps follow RMI (1995):

1. Ingots were received from FMPC.

2. Ingots were transferred to storage. 

Note: All onsite transfer of ingots, billets, forgings, and extrusions was accomplished 

by use of forklifts, fixtures, and overhead cranes.

3. Ingots were transferred to inspection.

4. Inspection was performed using an overhead monorail, scale, and inspection stand. 

Inspection included weighing and a dimensional and visual inspection.

5. Ingots were transferred back to storage.

6. Ingots were transferred to the salt baths and heated in molten salt to approximately 1,180°F 

for 1.5 hours minimum, 6 hours maximum.

7. Ingots were transferred to the press.

8. Ingots were extruded through the press into a heavy walled tube. This process included 

lubrication of press tooling to reduce friction during high-pressure extrusion.

9. The extrusion exited through the die head onto a runout table, and was placed on a rotating 

table and left for approximately 2.5 minutes.

10. The extrusion was transferred and lowered vertically into a water-filled quench tank for cooling 

for a minimum of 3 minutes.

11. The extrusion was transferred to a transfer table.

12. The extrusion was lifted horizontally and:

a. From 1962 to the mid-1960s, was placed into a trichloroethylene vapor degreaser tank to 

clean the extrusion.

b. From the mid-1960s, was placed in a nitric acid pickle tank to clean the extrusion.

13. The extrusion was transferred to a water rinse tank, rinsed, and transferred to the packing 

station for inspection. Then:

a. Until the 1960s, extrusions were packed, stored, and shipped back to FMPC for further 

processing.

b. In the late 1960s, further processing was performed at as described below.

14. The extrusion was cut into sections (billets).
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ATTACHMENT B 

N-REACTOR PRODUCTION (continued)

15. Billets were nitric acid pickled, inspected, machined, and reinspected before a closed die 

forging process at the press.

16. Billets were transferred to the storage area to await forging.

17. Billets were transferred to the salt baths and heated in molten salt. Billets were heated to 

approximately 1,170°F for 1-hour minimum, 6-hour maximum.

18. Billets were transferred to the press.

19. Billets were forged (closed die process) in the press. This process shaped the billet to the 

approximate dimensions of the final size and included lubrication of press tooling.

20. After forging, billets were lifted directly from the die head and lowered into a water-filled 

quench tank for cooling for a minimum of 3 minutes.

21. Billets were transferred to a transfer table for post forging inspection.

22. Billets were nitric acid pickled, inspected, machined (if required, repickled and reinspected), 

and packed for shipment.

23. Packaged billets were stored for shipment to Richland, Washington.

24. Packaged billets were shipped.

Note: Residues and metal turnings generated throughout the production process were 

processed (dried, sampled, oxidized, etc.) and returned to FEMP.

Note: According to Breslin and Glauberman (1964, pp. 14–16), after extrusion, hand rolling 

was performed for 5 minutes.
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ATTACHMENT C 

SAVANNAH RIVER SITE PRODUCTION

The typical steps for uranium processing for the SRS reactors follow (RMI 1995).

1. Ingots were received from FMPC.

2. Ingots were transferred into storage. 

Note: All onsite transferring of ingots, billets, forgings, and extrusions was 

accomplished by the use of forklifts, fixtures, and overhead cranes.

3. Ingots were transferred to inspection.

4. Inspection was performed using an overhead monorail, scale, and inspection stand. 

Inspection included weight and a dimensional and visual inspection.

5. Ingots were transferred back to storage.

6. Ingots were transferred to the salt baths and heated in molten salt to approximately 1,160°F 

for 75 minutes minimum.

7. Ingots were transferred to the press.

8. Ingots were extruded through the press into tubing. This process included lubrication of press 

tooling to reduce friction during high-pressure extrusion.

9. The extrusion exited through the press into tubing. This process included lubrication of press 

tooling to reduce friction during high-pressure extrusion.

10. The extrusions were lowered horizontally into a water-filled quench tank.

a. Water quenching was begun in approximately 1966. Before this, extrusions were aircooled.

11. The extrusions were cut into sections on an abrasive saw.

12. The extrusions were transferred to, and run through, the roll straightener.

a. From 1962 until approximately 1964, the extrusions were lowered horizontally.

b. In approximately 1964, the hot oil bath was removed and an induction heater was installed 

in its place. Extrusions were run through the induction heater before the roll straightener. 

This process continued for approximately 1 year. After this, extrusions were not heated 

before straightening.

13. Extrusions were stored on the process table.

14. Extrusions were lifted horizontally, then:

a. From 1962 to approximately 1964, were placed in a vapor degreaser tank to clean the 

extrusion.
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ATTACHMENT D 

DOD PENETRATOR PRODUCTION (continued)

b. After approximately 1964, the vapor degreaser tank was no longer used and extrusions 

were transferred to a water rinse tank, rinsed, and transferred to the packing station.

15. Extrusions were inspected, weighed, and packed for shipment to FMPC.

16. Packaged extrusions were stored before shipment.

17. Packaged extrusions were shipped.

Note: Residues and metal turnings generated throughout the production process were 

processed (dried, sampled, oxidized, etc.) and returned to FMPC. 
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ATTACHMENT D 

DOD PENETRATOR PRODUCTION

The typical steps for source material processing for DOD penetrators follow (RMI 1995).

1. Ingots were received.

2. Ingots were transferred to storage:

a. Stored in main plant before 1984.

b. Stored in main plant and northeast warehouse after 1984.

3. Ingot inspection:

a. Inspected at inspection station.

b. Inspected at floor scale.

4. Storage:

a. Stored in main plant before 1984.

b. Stored in main plant and northeast warehouse after 1984.

5. Heating:

a. Salt baths.

b. Sunbeam furnace (preheat only).

c. IFSI furnace (preheat only).

d. Other electric furnaces in, or transferred to the main plant salt bath area.

6. Extrusion on the extrusion press.

7. Postextrusion, which included numerous different processes:

a. Water quench.

b. Air cool.

c. Saw.

d. Transfer.

e. Pickle.

f. Water rinse.

g. Inspection and weighing at packing station or floor scale.

h. Storage:

i. Stored in main plant before 1984.

ii. Stored in main plant and northeast warehouse after 1984.

i. Shipment.
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ATTACHMENT E 

EXTERNAL DATA EVALUATION OF DOSES

TABLE OF CONTENTS

SECTION TITLE PAGE

E.1 Purpose ................................................................................................................................. 81

E.2 Background............................................................................................................................ 81

E.3 General Approach.................................................................................................................. 81

E.4 Applications and Limitations................................................................................................... 81

E.5 External Data Development ................................................................................................... 82

E.6 Adjustment for Missed Dose .................................................................................................. 83

E.7 Annual Data Evaluation of Dose Summaries.......................................................................... 83

LIST OF TABLES

TABLE TITLE PAGE

E-1 Example of analysis of prorating gamma and beta data and normalizing to annual................ 82

E-2 Years with limited data for external data analysis................................................................... 83

E-3 Missed external doses ........................................................................................................... 83

E-4 Annual external data evaluation of doses modified to account for missed dose ..................... 85

E-5 Annual CTW dose based on data evaluation of worker dosimetry.......................................... 86
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ATTACHMENT E 

EXTERNAL DATA EVALUATION OF DOSES (continued)

E.1 PURPOSE

The purpose of this attachment is to provide information to allow dose reconstructors to assign doses 

that based on worker dosimetry data for Extrusion Plant workers who have no or limited monitoring 

data. In addition, the data in this attachment should be used to assign doses for gaps in dosimetry 

records. The data are to be used either in addition to dosimetry data from the site or when monitoring 

data is lacking for covered employment at the site.

E.2 BACKGROUND

The ORAU Team has evaluated the external dosimetry records for the Extrusion Plant to permit dose 

reconstructors to complete certain cases for which external or internal monitoring data are unavailable 

or incomplete. Cases that do not have complete monitoring data could fall into one of several 

categories:

• The worker was unmonitored and, even by today’s standards, did not need to be monitored 

(e.g., a nonradiological worker).

• The worker was unmonitored but, by today’s standards, would have been monitored.

• The worker might have been monitored, but the data are not available to the dose 

reconstructor.

• Partial information is available, but it is insufficient to facilitate a dose reconstruction.

Some cases without complete monitoring data can be processed based on assumptions and 

methodologies of evaluating the data that existed for the Extrusion Plant and given at the 50th

percentile. For example, many cases in the first category can be processed by the assignment of 

external and internal doses based on information in the relevant site technical basis documents.

E.3 GENERAL APPROACH

The general approach to the development of this data evaluation for cases without external monitoring 

data is to assign either 50th- or 95th-percentile doses with the intent that the assigned doses 

represent, but do not underestimate, the doses that would have been assigned had the worker been 

monitored. 

E.4 APPLICATIONS AND LIMITATIONS

Some Extrusion Plant workers could have worked at one or more other sites in the DOE complex 

during their employment histories. Therefore, the data in this attachment must be used with caution to 

ensure that, for likely noncompensable cases, unmonitored external doses from multiple site 

employments have been overestimated. This typically requires the availability of the recorded doses 

or information for external coworker dosimetry data for all relevant sites.

The data in this attachment address penetrating gamma radiation and beta or nonpenetrating electron 

and/or low-energy photon radiation. These doses are to be assigned with a constant distribution in 

IREP.
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ATTACHMENT E 

EXTERNAL DATA EVALUATION OF DOSES (continued)

E.5 EXTERNAL DATA DEVELOPMENT

Dosimetry Data

The Extrusion Plant database contains four kinds of dosimetry data: gamma, beta, shallow, and 

deep. There is also a field for neutron dose that contains a minimal amount of data, which was 

insufficient for analysis. Gamma and deep are assumed to be equivalent measurements, and shallow 

dose was assumed to be combination of penetrating gamma and electron dose. The database 

contains 9,663 records.

All four dosimetry components could apply to one individual in a given year, and the periods could 

overlap. To perform the analysis, dose for a given person and year was prorated by summing all 

gamma or beta dose divided by the fraction-of-year the worker was monitored. Values for individuals 

that were found to be duplicates were removed.

Table E-1 contains an example calculation. This example has a gamma reading and beta reading 

that covers part of the year and the method for prorating the dose.

Table E-1. Example of analysis of prorating gamma and beta 

data and normalizing to annual, 330 days of monitoring. 

Value Gamma Dose Beta dose

Sum of dose within a year 50 100

Calculation = 50/(330/365) = 100/(330/365)

Prorated dose 53.94 107.88

Personal Identifiers

The personal identifiers in the database include Social Security Number, last name, and first name. A 

combination of the first and last name was chosen to distinguish individuals. This is believed to 

represent the best possible method of separating individuals given the available data. This analysis 

found that 68 records did not have a last name or a first name; they were excluded from the analysis. 

Several of the 68 records that were not used did have a Social Security Number but no positive data 

was recorded for these records. 

Limited Data

Adjustments were made to years with gaps or limited data. Table E-2 shows the years that were 

affected and the number of data points available for the original analysis. To compensate for the 

shortfall, data from surrounding years were added to the original data. 
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ATTACHMENT E 

EXTERNAL DATA EVALUATION OF DOSES (continued)

Table E-2. Years with limited data for external data 

analysis. 

Year Data points Years combined

1965 0a 1964–1966

1966 1b 1965–1967

1969 2c 1968–1971

1970 0c 1968–1971

1984 9d 1983–1985

1995 0e 1994–1996

2000 2f 1999–2001

a. Gamma data from 1964 and 1966 to give gamma data for 1965.

b. Beta results were not useable because film badge data was 

inverted for 1966. Gamma results were useable for 1966. Used beta 

results from 1964 and 1967.

c. Beta results from 1968 and 1971 to give 1969 and 1970 beta dose.

d. Gamma and beta values from 1983 and 1985 were used for 1984.

e. Gamma and beta values from 1994 and 1996 were used for 1995

because there was inadequate data for 1995. 

f. Gamma and beta values from 1999 and 2001 were used for 2000.

E.6 ADJUSTMENT FOR MISSED DOSE

According to OCAS-IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 2007b), 

missed doses are to be assigned for reported zero readings for each monitoring cycle to account for 

the possibility that doses were received but either not recorded by the dosimeter or not reported by 

the site. In addition, reported dose values less than one-half the applicable MDLs are to be assigned 

as missed dose. Annual maximum potential missed doses are calculated by multiplying the number 

of zero or unrecorded badge readings by the reported dosimeter limit of detection (LOD) and 

summing the results. These values are used as the 95th-percentile values of a lognormal distribution 

to calculate the probability of causation, which is determined by DOL. Therefore, in the Interactive 

RadioEpidemiological Program (IREP), Parameter 1 is equal to the calculated maximum annual 

missed doses multiplied by 0.5, and Parameter 2 is equal to 1.52. These values represent the GM 

and GSD, respectively, for each year of analysis. Table E-3 lists the maximum annual missed dose 

by monitoring period.

Table E-3. Missed external doses (rem) 

Monitoring 

period

Gamma and Beta 

LOD

Exchange 

frequency

Maximum potential 

annual missed gamma 

and beta dose

1962–1966 0.03 Monthly 0.360

1967–1985 0.02 Quarterly 0.080

1986–1988 0.02 Quarterly 0.080

1989–1992 0.01 Monthly 0.120

1993–2003 0.01 Quarterly 0.040

E.7 Annual Data Evaluation of Dose Summaries

Based on the described information and approaches, Extrusion Plant annual external dosimetry 

summaries were developed from available data for use in the evaluation of external gamma and beta 

dose for certain workers who were potentially exposed to workplace radiation but for whom there is no 

or limited monitoring data. 
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ATTACHMENT E 

EXTERNAL DATA EVALUATION OF DOSES (continued)

These summaries were developed using the following steps:

Step 1. The reported gamma dose was modified for each worker to account for partial years of 

employment. This permits the dose reconstructor to assign an appropriate prorated dose to 

account for partial years of employment or potential exposure.

Step 2. One-half of the maximum potential annual missed doses in Table E-3 were added to the 

reported annual doses from Step 1 (with the exception of reported positive doses, in which 

case the maximum missed dose was reduced by the dose that corresponded to one badge 

exchange because it is not possible that all individual badge results were zero if a positive 

annual dose was reported).

Step 3. The 50th- and 95th-percentile annual data evaluation of gamma doses were derived from 

the doses from Step 2 by ranking the data into cumulative probability curves and extracting 

the 50th- and 95th-percentile doses for each year.

Step 4. Table E-4 lists the results of the external dosimetry data analysis. These percentile doses 

should be used for Extrusion Plant workers with no or limited monitoring data. In general, 

the 50th-percentile dose can be used as a best estimate of a worker’s dose when 

professional judgment indicates that the worker was probably exposed to intermittent low 

levels of external radiation. The 50th-percentile dose should generally not be used for 

workers who were routinely exposed. For routinely exposed workers (i.e., workers who were 

expected to have been monitored and routinely exposed), the 95th-percentile dose should 

be applied. For instance, for cases in which routine monitoring data exist but there is a gap 

in the data, then Table E-4 or E-5 dose can be used to supplement missing quarters or 

years. The percentile dose should be the one that is consistent with the recorded doses 

unless there is reason to believe that the worker’s job or location in that year differed 

significantly from the job or location during the years dose was recorded. For workers who 

are unlikely to have been exposed, dose reconstructors should assign external dose 

equivalent to the 50th percentile of Table E-4. 

Step 5. Table E-5 lists gamma and beta dose values (as described in the steps above) that have 

been adjusted using the guidance in Section 8.0 of ORAUT-OTIB-0052, Parameters to 

Consider When Processing Claims for Construction Trade Workers (ORAUT 2014c). This 

guidance is applicable for workers who meet the criteria for CTW.
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ATTACHMENT E 

EXTERNAL DATA EVALUATION OF DOSES (continued)

Table E-4. Annual external data evaluation of doses modified to account for 

missed dose (mrem).

Year

Gamma

95th percentile

Gamma

50th percentile

Beta

95th percentile

Beta

50th percentile

1962 811 285 3,314 486

1963 656 376 4,316 1,438

1964 405 240 4,880 1,175

1965 402 225 5,639 1,148

1966 381 205 6,002 1,156

1967 517 167 6,056 1,029

1968 345 110 4,474 750

1969 723 271 3,330 661

1970 539 90 3,330 661

1971 504 185 2,247 576

1972 190 110 1,057 471

1973 297 105 4,422 587

1974 334 110 4,787 1,033

1975 263 90 3,590 541

1976 190 70 2,977 540

1977 321 80 2,451 541

1978 296 150 1,760 882

1979 208 95 1,552 536

1980 231 60 2,553 491

1981 391 150 2,792 692

1982 473 190 4,131 1,324

1983 403 150 2,447 732

1984 297 110 2,222 511

1985 231 110 2,276 391

1986 905 151 3,229 419

1987 238 50 1,023 91

1988 517 90 1,017 40

1989 105 75 60 60

1990 75 60 60 60

1991 75 60 60 60

1992 60 60 60 60

1993 40 40 40a 40

1994 53 40 74 40

1995 54 40 74 40

1996 94 40 40 40

1997 40 40 115 40

1998 94 51 85 55

1999 76 40 46 40

2000 70 40 40 40

2001 65 40 40a 40

2002 40 40 40 40

2003 56 40 40a 40

2004 55 35 40a 40

a. 95th percentile calculated lower, set equal to 50th percentile.
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ATTACHMENT E 

EXTERNAL DATA EVALUATION OF DOSES (continued)

Table E-5. Annual CTW dose based on data evaluation of worker dosimetry

(mrem).

Year

Gamma

95th percentile

Gamma

50th percentile

Beta

95th percentile

Beta

50th percentile

1962 1,069 333 4,573 614

1963 853 460 5,977 1,948

1964 501 270 6,766 1,579

1965 497 249 7,829 1,541

1966 467 221 8,337 1,552

1967 712 221 8,466 1,429

1968 470 142 6,251 1,038

1969 1,000 367 4,649 913 

1970 743 114 4,649 913

1971 693 248 3,134 795

1972 255 142 1,468 648

1973 404 135 6,178 809

1974 455 142 6,690 1,434

1975 356 114 5,014 746

1976 254 86 4,156 744

1977 437 100 3,419 746

1978 402 198 2,452 1,223

1979 279 121 2,160 739

1980 311 72 3,562 676

1981 535 198 3,896 957

1982 651 255 5,772 1,841

1983 552 198 3,413 1,013

1984 403 142 3,099 704

1985 311 142 3,175 535

1986 1,255 199 4,509 575

1987 321 58 1,420 115

1988 711 115 1,412 40

1989 125 83 60 60

1990 82 60 60 60

1991 83 60 60 60

1992 60 60 60 60

1993 40 40 40a 40

1994 62 40 92 40

1995 63 40 92 40

1996 119 40 40 40

1997 40 40 148 40

1998 119 60 107 64

1999 95 40 53 40

2000 86 40 44 36

2001 79 40 40a 40

2002 44 40 40 40

2003 67 44 40a 40

2004 65 37 40a 40

a. 95th percentile calculated lower, set equal to 50th percentile.