ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 73

Document Title:

Summary of Extrusion Plant Site Information for 

Dose Reconstruction

Document Number: ORAUT-TKBS-0056

Revision: 00

Effective Date: 04/26/2007

Type of Document: TBD

Supersedes: None

Subject Expert(s): Robert Hysong, Robert Vogel, Cindy W. Bloom, and Shelby Gubin

Site Expert(s): N/A

Approval: Signature on File Approval Date: 04/24/2007

Robert Hysong, Document Owner

Concurrence: Signature on File Concurrence Date: 04/24/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 04/24/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 04/24/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 04/26/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 2 of 73

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 A1. 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. Training required: As determined by the Task 

Manager. Initiated by Robert Hysong.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 3 of 73

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

6

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

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

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

2.0 Site Information........................................................................................................................9

2.1 Process Description and Facilities ..............................................................................11

2.2 Production Quantities .................................................................................................13

2.3 Source Term...............................................................................................................15

 2.3.1 Industrial X-Ray Sources...................................................................................16

2.4 Radiological Safety Program ......................................................................................16

 2.4.1 Area Surveys......................................................................................................19

 2.4.2 Personnel Monitoring.........................................................................................19

 2.4.3 Air Monitoring.....................................................................................................20

2.5 Radiological Incidents.................................................................................................20

2.6 Physical Examinations Including Occupationally Required 

Medical X-Ray Information .........................................................................................20

2.7 Work Periods..............................................................................................................21

2.8 Period After Radiological Production ..........................................................................21

3.0 Internal Dose..........................................................................................................................22

3.1 Uranium Bioassay Data..............................................................................................22

3.2 In Vivo Monitoring.......................................................................................................23

3.3 Intake Assumptions ....................................................................................................25

 3.3.1 Uranium..............................................................................................................25

 3.3.2 Thorium..............................................................................................................25

3.4 Internal Dose Assignment Summary...........................................................................25

4.0 External Dose ........................................................................................................................26

4.1 Penetrating doses.......................................................................................................26

4.2 Nonpenetrating Dose..................................................................................................29

4.3 Neutron Dose .............................................................................................................29

4.4 Extremity Monitoring...................................................................................................29

4.5 Occupationally Required Medical X-Rays...................................................................30

4.6 Miscellaneous Information Concerning External Exposure ......................................... 31

4.7 External Dose Assignment Summary .........................................................................31

5.0 Internal Dose During the Post Production Period ................................................................... 32

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

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

Glossary................................................................................................................................................42
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 4 of 73

ATTACHMENT A, EXTRUSION PLANT BUILDING LAYOUTS....................................................... 47

ATTACHMENT B, EXTRUSION PLANT PROCESSES – N-REACTOR PRODUCTION ................... 68

ATTACHMENT C, EXTRUSION PLANT PROCESSES –

 SAVANNAH RIVER SITE PRODUCTION............................................................ 70

ATTACHMENT D, EXTRUSION PLANT PROCESSES – DOD PENETRATOR PRODUCTION ....... 72

ATTACHMENT E, DETAILED PENETRATING WHOLE BODY DOSES ........................................... 73

LIST OF TABLES

TABLE TITLE

2-1 Primary processing and storage buildings.............................................................................. 11

PAGE

2-2 Uranium receipts for N-Reactor and SRS work ...................................................................... 13

2-3 Annual recycled uranium receipts at RMI ............................................................................... 13

2-4 Uranium-235 weight percentage of uranium handled ............................................................. 14

2-5 Annual receipts of enriched uranium ...................................................................................... 14

2-6 Relative mass ratio of uranium to thorium handled................................................................. 15

2-7 Uranium mixtures, specific activity, and isotopic fractions ...................................................... 15

2-8 Recommended surface contamination limits in 1962.............................................................. 19

3-1 Uranium urinalysis vendors and detection levels.................................................................... 23

3-2 In vivo measurement types and detection levels for various time periods................................24

3-3 In vivo record codes................................................................................................................. 24

3-4 Estimated recycled uranium activity fractions for internal dose reconstruction........................ 25

3-5 Assumptions for intake determinations from uranium bioassay .............................................. 26

4-1 External dosimetry ................................................................................................................. 27

4-2 Penetrating whole body external doses, 1962–1999 .............................................................. 27

4-3 Forearm and chest skin dose monitoring results for personnel who closely handled or 

 inspected uranium billets in 1974........................................................................................... 30

4-4 External dose assumptions .................................................................................................... 31

LIST OF FIGURES

FIGURE TITLE

2-1 Extrusion Plant site map......................................................................................................... 11

PAGE

A-1 Plant Layout...........................................................................................................................47

A-2 Plant Layout – 1957 ...............................................................................................................50

A-3 Plant Layout – 1962 ...............................................................................................................51

A-4 Plant Layout – 1964 ...............................................................................................................52

A-5 Plant Layout – 1965 ...............................................................................................................53

A-6 Plant Layout – 1968 ...............................................................................................................54

A-7 Plant Layout – 1969 ...............................................................................................................55

A-8 Plant Layout – 1979 ...............................................................................................................56

A-9 Plant Layout – 1981 ...............................................................................................................57

A-10 Plant Layout – 1982 ...............................................................................................................58

A-11 Plant Layout – 1983 ...............................................................................................................59

A-12 Plant Layout – 1984 ...............................................................................................................60

A-13 Plant Layout – 1985 ...............................................................................................................61

A-14 Plant Layout – 1986 ...............................................................................................................62

A-15 Plant Layout – 1987 ...............................................................................................................63
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 5 of 73

A-16 Plant Layout – 1988 ...............................................................................................................64

A-17 Plant Layout – 1989 ...............................................................................................................65

A-18 Plant Layout – 1990 ...............................................................................................................66

A-19 Plant Layout – 1991 ...............................................................................................................67
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 6 of 73

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AP anterior-posterior (X-ray)

AWE Atomic Weapons Employer

BZ breathing zone

DOD U.S. Department of Defense

DOE U.S. Department of Energy

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

HEPA High Efficiency Particulate Air

hr hour

IREP Interactive RadioEpidemiological Program

keV kilovolt-electron, 1,000 electron-volts

L liter

lb pound

mg milligram

min minute

mL milliliter

mrad millirad

mrem millirem

MT metric ton

MTU metric tons uranium

NAD nuclear accident dosimeter

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NLO National Lead of Ohio, Inc.

NRC U.S. Nuclear Regulatory Commission

NVLAP National Voluntary Laboratory Accreditation Program

ODH Ohio Department of Health

PA posterior-anterior (X-ray)

pCi picocurie

RCRA Resource Conservation and Recovery Act

RMI Reactive Metals, Inc.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 7 of 73

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year

μCi microcurie

μg microgram

°F degree Fahrenheit

§ section or sections
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 8 of 73

1.0

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

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

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

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

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

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

INTRODUCTION

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

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

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

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

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

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

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

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

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

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

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

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

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

derived:

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 9 of 73

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

1.1 PURPOSE

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

Extrusion Plant (also known as Reactive Metals Inc., or RMI) and to help ensure that all components 

of dose are adequately addressed. The document provides information on the radiological processes 

and source terms and on the radiological controls and monitoring practices. While not meant to 

substitute for a complete site profile, it represents the best understanding of the covered 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 the radiological controls and monitoring 

practices. Sections 3.0 and 4.0 discuss internal and external dose, respectively.

2.0

The DOE Office of Worker Advocacy defines the following information for the Extrusion Plant (DOE 

2006):

SITE INFORMATION

Site Extrusion Plant

(Reactive Metals, Inc.)

Alternative names Reactive Metals, Inc. 

RMI

Location Ashtabula, Ohio

Covered Period 1962 to present

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 and/or closed-die 

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

stocks 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 corporate successor of Bridgeport Brass facilities of Adrian, Michigan, 

and Bridgeport, Connecticut. The work at Adrian, Michigan, 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-ton 

Loewy horizontal extrusion press owned by AEC, was moved from Adrian to Ashtabula in November 

or December 1961 (Haywood 1982; Jeffererson 1961; 1962a) following the cessation of work in 

Adrian and uranium extrusion at Ashtabula began in January 1962 (Koh 1995). The majority of 

material processed at the facility was for AEC and DOE, but non-radioactive metals such as copper, 

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

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

Atomic Energy Act (Koh 1995, p. 2-4) and as an AEC/Nuclear Regulatory Commission (NRC) 

licensee (DOE 2000). The Atomic Energy Act, Section 110 (NRC 2002) states, 

Nothing in this chapter shall be deemed 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 10 of 73

a. 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 91, or for 

the use of such facility by the Department of Defense or a contractor thereof.

Extrusion Plant’s DOE (and predecessor) operations were conducted under a prime contract from 

1962 through August 1987, a subcontract with the Fernald Environmental Management Program 

(FEMP) from September 1987 through November 1992, a prime contract with DOE Oak Ridge 

Operations from December 1992 through March 1993, a prime contract with DOE Chicago Operations 

Office (apparently) from April 1993 through March 1995 and then a prime contract with 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 1999-2004). 

DOE contracts with Extrusion Plant included DE-AC05-760R 01405, administered by the Oak Ridge 

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

Extrusion Plant’s NRC-regulated work was conducted under source material License Number SMB602, first issued in June 1962, which initially allowed the use of 500,000 lbs (230 MT) of uranium and 

thorium; this was changed to 400,000 lbs (180 MT) of uranium and 100,000 lbs (45 MT) of thorium on 

October 31, 1973 (NRC 1962-1999) and to 10,000 lbs (4.5 MT) of natural uranium, 600,000 lbs (270 

MT) of depleted uranium and 10,000 lbs (4.5 MT) of thorium in 1979 (NRC 1979-1980). On June 7, 

1985, the NRC license allowed possession, use, and storage of 5000 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 until September 11, 1997,

when 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; and 

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 and 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-

millicurie sealed 137Cs source used to calibrate a gamma alarm system (RMI 1995a). 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

C.F.R. § 61.55 (RMI 1995a). On March 6, 2004, ODH issued approval for RMI to use sealed sources, 

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

counter, and 63Ni in a gas chromatography instrument. 

DOE work included the extrusion of primarily depleted and enriched uranium for N-Reactor and 

Savannah River. License SMB-602 allowed extrusion of uranium and thorium (NRC 1962-1999); it 

was also RMI’s authorization for use of source material for Department of Defense (DOD) armorpiercing penetrator work from 1974 through 1985 for DOD contractors (DOE 2000).
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 11 of 73

Westinghouse Materials Company of Ohio (1989) noted in their June 1989 review of the RMI program 

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

commercial extrusion work was occurring during the June 1989 review. Westinghouse also stated 

that no uranium was being extruded during the June 1989 review (Westinghouse Materials Company 

of Ohio 1989, p. 15). Other references (DOE 2000; RMI 1995a) report that RMI uranium extrusion for

DOE ceased 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 site profile refers to the radiological production period as January 1, 1962, through October 14,

1991, the period during which uranium was being extruded or during which uranium could have been 

extruded under RMI’s NRC license. The postproduction period is defined in this document as 

October 15, 1991, to the present, the period during which extrusion of radioactive metals ceased 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.

2.1 PROCESS DESCRIPTION AND FACILITIES

The Extrusion Plant is 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 2-1 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. 

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

Table 2-1. Primary processing and storage buildings.

Bldg. 

no. 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 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 12 of 73

Bldg. 

no. Building namea Description/commentsb

extrusion press, warehouse, laboratory, and offices; pickling, inspection, 

machining, and packaging of uranium

4 Enclosed ramp

5 Locker rooms/foreman’s office

6 Enclosed truck ramp

7 Dock area

8 Emergency equipment storage building

9 RCRA storage building

10 Billet storage warehouse Storage of incoming and outgoing uranium

11

12

Main plant high bay

Main plant low bay

Uranium processing, 3850 ton extrusion press, transfer table, pickling 

tanks, furnaces, abrasive saw.

13 Runout table filter building

14 Saw filter building

15 Tool crib Extrusion tooling storage

16 Die head filter building

17 Switchgear room

18 Compressor room

19 Wastewater treatment plant

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

21 ES & H Building

22 Guard house

23 Sewage disposal plant

24 Modular laboratory

25 Modular office

a. DOE (1993).

b. RMI (1995a).

The normal operations of extrusion and of forging performed at RMI can be generally described as a 

metals fabrication process. RMI’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 they generally consisted of heating the metal in a 

salt bath for 1.5 hours, followed by extruding and quenching. This was followed by a degreasing step 

and packaging and weighing.

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

cleaning the material.

RMI’s DOE work was done in support of the N-Reactor at the Hanford Site and for the reactors at the

Savannah River Site (SRS). RMI also supported the armor-piercing penetrator programs for DOD 

contractors. RMI produced N-Reactor fuel and targets from 1962 to 1988. From 1962 until 1970, the

uranium processing for N-Reactor consisted of receiving uranium primarily from the FMPC, extruding 

the metal, and returning the extrusions and scrap 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). The detailed steps involved in this process are described in 

Attachment B. SRS production was done at RMI from 1962 to 1988. The general steps consisted of 

receiving uranium, followed by 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 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 13 of 73

description of production for SRS, are included in Attachment C. Work was done on armor-piercing 

penetrators for DOD contractors under RMI’s AEC/NRC license. This was done from 1974 until 1985. 

This process was similar to that for SRS except for three process changes: 

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

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

straightener.

• 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 at RMI from 1962

through 1971 (RMI 1995b). A small number of NPR clad beryllium ingots were also extruded. This 

involved a process step described as hand rolling (Breslin and Glauberman1964). Because the ingots 

were identified as beryllium, they would not have been a source of exposure. 

2.2 PRODUCTION QUANTITIES

The quantities of uranium (listed in MTU) sent to RMI for the N-Reactor and SRS production are listed 

in Table 2-2.

Table 2-2. 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).

RMI production for the DOD penetrator program was reported as approximately MT of depleted 

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

sent to RMI was therefore, 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 shown in Table 2-3 below (DOE 

2000). 

Table 2-3. Annual recycled uranium receipts at RMI.

a

Year MTU Year MTU Year MTU Year MTU

1962 526.85 1970 1,731.69 1978 2,314.72 1986 4,322.51

1963 2,977.63 1971 1,920.31 1979 2,355.17 1987 800.76

1964 4,518.48 1972 1,898.09 1980 3,175.80 1988 496.52

1965 2,213.02 1973 3,083.36 1981 3,794.46 1989 9.67

1966 2,694.82 1974 2,226.30 1982 5,873.91 1990 0.17

1967 3,249.29 1975 1,547.26 1983 6,619.13 1991 0

1968 2,875.46 1976 2,076.36 1984 4,832.58 1992 0

1969 2,657.32 1977 2,232.35 1985 3,697.74 1993 0.05

Total 76,721.78

a. The information comes from reference DOE 2000, and it is not clear if this is only a listing of 

DOE-contract processed uranium or if it also includes NRC licensed materials.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 14 of 73

According to the current (2006) RMI Environmental, Safety and Health Director, site 

documentation indicates that the amount of uranium (and presumably thorium)

processed and handled at the Extrusion Plant was larger than the amount shipped 

because some of the material was processed more than one time at the Plant (Vogel 

2006). The total amount of uranium processed, reported as 121,224 MT (Saito 1993; 

RMI 1995a) and as 130,712 MT (Koh 1995), provides an upper amount of uranium

handled at the Extrusion Plant. The difference between the latter two numbers is 

additional DOD work of 9,488 MTU (Koh 1995). Tables 2-4 and 2-5 indicate the 

enrichment of the uranium handled and received at RMI. On average, the enrichment 

of uranium received and handled at the Extrusion Plant was much lower than the 

highest enrichment ever handled at the Extrusion Plant (2.1%).

Table 2-4. Uranium-235 weight percentage of uranium handled.

a

U-235 % by weight 0.14 & 0.2 0.71 0.86 0.95 1.25 2.1 Total MTU

MTU 64,438 25,178 3,932 28,115 8,108 941 130,712

Percentage of total 

uranium processed

49.3 19.3 3 21.5 6.2 0.72 100

a. Koh (1995).

Table 2-5. 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 .002084 .000025 1.199

Total 25,327.5 264.8

a. DOE (2000).
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 15 of 73

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

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

amounts of thorium processed for SRS, Hanford, the Y-12 Plant, and Davison Chemical (also known 

as W. R. Grace) between 1962 and 1971 are listed by year in Table 2-6 (Britcher 1992). 

Consideration of RMI’s total uranium process inventory would reduce the Table 2-6 ratios if the 

thorium inventories are complete. 

Table 2-6. 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 A801-38 (Britcher 1992), but the amount is currently unavailable. It 

consisted of only one extrusion.

2.3 SOURCE TERM

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 that might be of use are listed in Table 2-7 below. As listed in Tables 2-4 and 2-5, the 

enrichment of uranium handled at the Extrusion Plant was typically well below 2.1%. For dose 

reconstruction, the default assumption is that the uranium was 2% enriched.

Table 2-7. Uranium mixtures, specific activity, and isotopic fractions. 

Slightly enriched (2%)a

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.7694 0.0002 28.76 0.01

U-235 0.0268 0.02 1 1

U-238 0.2038 0.9798 7.618 48.99

U (2%) pCi/μg 1.616

U (pCi): U-235 (μg) 80.8

Useful factors

U (2.0%) pCi/μg 1.616

U (2.0%) U-235 (μg) 80.8

Natural uranium

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.4886 5.37E-05 21.4 0.00745

U-235 0.0228 7.20E-03 1 1

U-238 0.4886 9.93E-01 21.4 138

Useful factors

U-natural pCi/μg 0.683

U (pCi): U-235 (μg) 94.9
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 16 of 73

Depleted

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.1546 1.00E-05 14.45 0.00502

U-235 0.0107 1.99E-03 1 1

U-236 0.0005 3.11E-06 0.0467 0.00156

U-238 0.8342 9.98E-01 78 501

Useful factors

Depleted uranium pCi/μg 0.4021

Depleted uranium (pCi): U-235 (μg) 202

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

to uranium (ORAUT 2004a). 

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 (Koh1995). A review 

of site records (Britcher 1992) indicated that thorium was not processed by RMI after 1971. Table 2-6 

shows Th:U mass ratios of less than 1%, and the NRC (1962-1999) licensing documents indicate that 

for most periods, authorized mass ratio of thorium to U (natural and depleted) was no more than 25% 

prior to 1979 and was less than 2% by 1979, and these ratios would be lower if consideration was 

given to the mass of enriched uranium. The low ratios of Th:U are also supported by later

environmental sampling and characterization surveys that only identified uranium and technetium 

contamination at the site (RMI 1995a, 1995b). Thorium-232 (which is the most significant isotope by 

mass in natural thorium) contamination was not detected on site. Th-230, a long-lived uranium 

progeny, was found on site but was characterized as consistent with background levels (NRC 1962-

1999). 

2.3.1

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

facility. Furthermore, the 1965 RMI application for renewal of 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).

Industrial X-Ray Sources

2.4 RADIOLOGICAL SAFETY PROGRAM 

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

documentation. Other information sources included the RMI (1973) Health Protection Manual; Health 

and Safety Laboratory 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 conducted 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 (ORNL 1985). Consequently, between July 1985 and 1988, Battelle, Pacific Northwest 

Laboratory conducted an in-depth review of the RMI radiation protection program that included air 

sample particle size and solubility studies and recommendations for internal and external dose 

control, contamination control, training, respiratory protection, and associated procedures (Munson 

1985). RMI established an As Low as Reasonably Achievable (ALARA) 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.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 17 of 73

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 and shoes for use in contaminated areas or the 

use of rubber shoe covers (Jefferson 1962a). The location of the rack containing re-usable 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 Plant 

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 area 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 Plant 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, Plant 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; however, 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 material available to be spread along with existing controls were 

such that the accidental transport of significant quantities of material offsite was “unlikely” (Munson 

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

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

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

control program (Munson 1985). Westinghouse Materials Company of Ohio (1989) observed that 

contamination control and monitoring practices were being implemented in their 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 1991), while the skin contamination limit was

3,700 dpm/100 cm2 beta-gamma (RMI 1994).

Before the construction of a new Butler Building sometime around the first quarter of 1965, the Plant 

change 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 

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 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 18 of 73

to the new Butler Building by March 1965 (Ruch 1965); however, 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 dpm/100 cm2 and 76,798, respectively 

(RMI 1989b). 

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

1985, an independent health, safety, and environmental review conducted at RMI resulted in RMI 

instituting a no-smoking policy in the production and incineration areas of the Plant because 

“employees were smoking throughout the plant” (ORNL 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). The Health and Safety Laboratory 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, with 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). The gas fired 

incinerators were equipped with a Type N Rotoclone dust collector (Breslin and Glauberman 1964

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

warehouse portion of the RF-6 Building (Bean 1973). Site documentation (Ruch 1964; Smith 1973;

Hibbitts and Wing 1980; Wing 1982) 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. 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 Plant Safety Officer on a rotating basis so that each exhaust 

system was sampled at least every seven weeks (AEC 1973). Continuous monitoring of all seven 

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

area, and uranium scrap incinerators was recommended in 1985 (ORNL 1985). Installation and 

testing of a new abrasive saw ventilation system, which utilized High Efficiency Particulate Air (HEPA)

filters and a bag house 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 type) system was installed on the scrap incinerator at some time between June 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 19 of 73

1985 and June 1987, but isokinetic exhaust sampling was only performed when the ventilation 

equipment for the incinerator facility was operating (Davis 1987). Westinghouse Materials Company 

of Ohio (1989) noted in their June 1989 review of the RMI program that Permit to Install applications 

had been submitted for 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

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

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 cleanup cycles were expected to keep the 

surface contamination limits below the levels listed in Table 2-8. 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 (ORNL 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).

Area Surveys

Table 2-8. Recommended surface contamination limits in 1962a.

Location

Alpha (dpm/100 cm2) Beta-gamma (mrad/hr)

Removable Fixed Removable Fixed

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 - -

Plant office 500 1,000 0.3 2.5

Shipping-receiving floor 500 1,000 0.4 2.5

Truck beds (after use) No detectable 500 No detectable 0.4

Process equipment above floor level

Equipment-in-place (accessible areas) 10,000 20,000 5 10

Before removal (to other Plant property sites) 2,000 4,000 0.4 2.5

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

2.4.2

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

Personnel Monitoring
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 20 of 73

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

Early site documentation (Jefferson 1962c) indicates that the Plant 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 at the Plant chosen 

by the Plant 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,

the AEC’s Health and Safety Laboratory 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) has been used as a basis for comparison to airborne exposure at the 

Bridgeport Brass Adrian, Michigan, 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 Adrian 

plant. In addition, two-stage air-sampling studies that were conducted in the 1963 – 1964 timeframe 

at the Extrusion Plant indicated that 90% of the air contamination was in a nonrespirable particle size 

range (Ruch 1964). Later studies conducted in the 1985 – 1986 time frame 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).

Air Monitoring

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 (DOE 1995) 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.

2.6 PHYSICAL EXAMINATIONS INCLUDING OCCUPATIONALLY REQUIRED MEDICAL 

X-RAY INFORMATION

The Extrusion Plant shared medical facilities with a nearby RMI plant (Hibbitts and Pryor 1970), and a 

dispensary was collocated 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, (medical)

urinalyses, and chest X-ray examinations for everyone except female clerical employees. In addition, 

hourly employees had a preemployment lower back (“normal spine”) X-ray examination (Johnson 

1976) because they were more “prone” to back injuries (Hibbitts and Pryor 1970). In 1976, it was 

noted that the back X-ray exam was only for laborers (Johnson 1976). The 1970 appraisal specifically 

stated that X-ray examinations were made off the site at Ashtabula General Hospital. The current 

(2006) RMI Environmental, Safety and Health Director, stated that RMI obtained their own X-ray 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 21 of 73

equipment for chest X-ray examinations in later years, although the specific date of equipment 

acquisition is unknown (Hysong 2006a). Specific X-ray techniques used to perform RMI X-ray 

examinations are not available. X-ray examinations were performed at the RMI dispensary at some 

point after 1976 until 1996. In addition, the practice of taking preemployment back (spinal) X-rays was 

phased out sometime before 1986 (Hysong 2006a). Beginning sometime in 1997, chest X-rays were 

only taken at the physician’s discretion. 

The claim files do not appear to contain any specific information on spinal X-ray examinations. Few 

records noting occupationally required X-ray examinations have been found in the claim files. 

2.7 WORK PERIODS 

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

Health and Safety Laboratory air concentration study (Breslin and Glauberman 1964) 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/day. In general, bioassay and external dosimeters will be used to determine doses, and in 

those cases, the period of exposure will be 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 the radiological production period as January 1, 1962, through October 14, 

1991, during which uranium was being extruded or might have been extruded based on uranium 

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

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

the present). On October 15, 1991, license amendment 4 converted SMB-602 from an Operating 

license to a Possession Only license 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, enabling 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 Environmental, Safety and Health 

Director, starting in January 2004, the only urine bioassay samples submitted by workers were 

preemployment, termination, and “for cause” whenever an intake was suspected (Hysong 2006a). By 

January 2004, building decontamination was completed and work at the site consisted primarily of soil 

and ground-water remediation (Hysong 2006a). In March 2004, decontamination and 

decommissioning activities at the site temporarily ceased until November 2005 when a remediation 

services contractor (LATA-SHARP Remediation Services LLC) was hired to complete 

decontamination and decommissioning of the site (Hysong 2006b). By March 2004, less than 20 RMI 

management and compliance personnel remained at the Extrusion Plant site to provide oversight of 

the remaining decontamination and decommissioning field activities which were likely to be completed 

by December 2006 (Hysong 2006b).

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 2-2 shows receipt of 0.1 MTU in 1993. The following information is from the 

licensing documents (NRC 1962-1999). On October 15, 1991, license amendment 4 converted 

SMB-602 from an operating license to a Possession Only license for possession incident to site 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 22 of 73

characterization and decommissioning plan preparation. License amendment 6 was approved on 

November 9, 1993, allowing for pre-decommissioning activities including surveys, equipment removal, 

waste handling, shipment, and disposal (amendment 5 named a new radiation safety officer). On 

September 11, 1997, the decommissioning plan was approved and incorporated into the license as 

amendment 8. This site profile refers to the radiological “production period” as January 1, 1962,

through October 14, 1991, the period during which uranium was being extruded or could have been 

extruded and the “post production 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). 

3.0

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 shall 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), 99Tc intakes may 

be higher than the intake derived using Table 3-4 below. Consequently, if in vivo monitoring records 

indicate 99Tc lung count results greater than the in vivo counter MDA of 0.5 uCi, then in vivo

monitoring records are used to assess 99Tc intakes. 

INTERNAL DOSE

3.1 URANIUM BIOASSAY DATA

Uranium urinalyses for workers whom 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 conducted by 

AEC’s Health and Safety Laboratory. From 1965 to the early 1970s, uranium urinalyses were

conducted by Tracerlab, Inc. 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 by RMI to have significant 

exposure potential. Starting sometime in the mid-1970s until 1988, total uranium urinalysis was 

conducted by the United States Testing Company, Inc. (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 the present, uranium urinalysis 

has been performed by several vendors including Controls for Environmental Pollution 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 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 23 of 73

in either milligrams or micrograms of uranium per liter. Table 3-1 lists the Extrusion Plant uranium 

urinalysis vendors and detection levels based on a review of claimant records and site documentation. 

Table 3-1. Uranium urinalysis vendors and detection levels.

Analysta Year(s)

Detection threshold 

concentration (μg/L) Reference

Health and Safety Laboratory 1962–1964 2.0 AEC 1964

Tracerlab and National Lead of Ohio 1965–1972 1.0 RMI 1967

USTC Early 1973–1983 3.0 RMI 1974

USTC 1985 5.0 Manninen1988

USTC 1986 0.1 Manninen1988

USTC 1987 0.5 Manninen1988

USTC, Controls for Environmental 

Pollution (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). Although this 

information exists, nothing has been identified directly related to the Extrusion Plant. 

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

weekly, monthly, or quarterly urine samples depending on their work assignment (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 

current RMI Environmental, Safety and Health 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 groundwater remediation. Intakes for personnel who may not have been monitored after 2003 are presented 

in section 5.0. 

3.2 IN VIVO MONITORING

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

determine worker’s 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 AEC/DOE in vivo

mobile counter provided by the Y-12 Plant) 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 3-2 lists general information about the detection capabilities of in vivo lung counting at the 

Extrusion Plant for various time periods. Table 3-3 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 transuranics or thorium until 

1986 (ORAUT 2004a) by which time the Y-12 mobile counter was no longer used at the Extrusion 

Plant. According to correspondence related to the RMI radiation protection program review 

conducted by Battelle Pacific Northwest Laboratory, when previous whole-body counting data were 

evaluated, they revealed “mostly problems and questions” (Munson 1985). 

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) compared to the DOE 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 24 of 73

Mobile Counter (Manninen 1988). March 1990 site correspondence indicates that the lung count 

results from Helgeson indicate 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 are incorrectly 

labeled milligrams of uranium rather than nanocuries of uranium and that the percent of the annual 

limit on intake calculation is 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 3-2. In vivo measurement types and detection levels for various time periodsa

.

Period Equipment

Measurement

type Radionuclide MDAb,c

Action 

level for 

recount

Action level

for work 

restriction

1968-1985 Y-12 Mobile 

Counter

Lung Total uranium 4 mg Unknown Unknown

1968-1985 Y-12 Mobile 

Counter

Lung Enriched uranium 

(2% 235U)

0.1mg Unknown Unknown

1968-1985 Y-12 Mobile 

Counter

Lung Depleted Uranium 4 mg Unknown Unknown

1968-1985 Y-12 Mobile 

Counter

Lung Np-237 200 pCi Unknown Unknown

1968-1985 Y-12 Mobile 

Counter

Lung Tc-99 0.5 uCi Unknown Unknown

1968-1985 Y-12 Mobile 

Counter

Lung Th-232 6 mg Unknown Unknown

1986-1995 Helgeson 

Counter

Lung Total uranium 2 – 4 mg Unknown Unknown

1986-1995 Helgeson 

Counter

Lung Enriched uranium 0.04 – 0.07 

mg

Unknown Unknown

a. Adapted from ORAUT (2006a).

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

c. The Th-232 MDA is based on Scott et. al. (1969) and ORAUT (2006a).

Table 3-3. 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 may 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 National Lead of Ohio, early 

operator of the Fernald facility.

Helgeson Lung n-u handwritten on 

forms with a 

spectral print out

Natural uranium result usually with 

corresponding handwritten result in 

nanocuries with the two 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 two sigma 

uncertainty.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 25 of 73

3.3 INTAKE ASSUMPTIONS

3.3.1

Uranium intakes are assumed to be type M or S. For calculating annual organ doses, the intake (in 

disintegrations per minute) can be assumed to be entirely 234U. Table 3-4 shows an estimate of 

recycled uranium contaminants that might have contributed significantly to internal doses beginning in 

1962. The numbers are based on a review of recycled uranium contaminants at Hanford and FMPC. 

(The relative internal dose contributions from 99Tc, 228Th, and 232Th were low enough to be excluded). 

In addition, the activity fractions assume that the uranium specific activity is based on depleted 

uranium, which increases the proportion of the contaminants by activity. The contaminant levels for 

depleted uranium overestimate the contaminants in uranium of normal enrichment by about 40%. 

Plutonium is assumed to behave as absorption type M or type S material and neptunium as 

absorption type M. If in vivo monitoring records indicate 99Tc lung count results greater than 0.5 uCi,

the 99Tc intake should be assessed assuming a type M intake. 

Uranium

Table 3-4. Estimated recycled uranium activity 

fractions for internal dose reconstruction.

a

Uranium Pu-239 Np-237

1 0.00246 0.00182

a. ORAUT (2005a).

3.3.2

In vitro analyses to estimate RMI thorium exposures are unavailable, and thorium activity determined 

from chest counts was not reported until 1979, which is after the reported thorium processing period. 

Table 2-6 indicates that the thorium received at RMI in any year was less than 1% of the uranium 

received by mass. Because there are no specific process and monitoring records for 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.

Thorium

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 Th:U mass ratio of 1%. This results in a relative 232Th-to-uranium intake fraction by activity of 

0.00161. Further, it is assumed that 232Th is in equilibrium with 228Th, so the 228Th-to-uranium activity 

fraction is also 0.00161. Exposure from 228Ra (half-life of 5.75 years) is assumed to be insignificant 

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

factor is small compared to thorium.

It is further assumed that thoron levels were not significant in the ventilated metal-handling workplace.

3.4 INTERNAL DOSE ASSIGNMENT SUMMARY

Depleted, natural, and slightly enriched uranium were all source terms at RMI, and the type is 

assumed to be M or S. For results measured 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.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 26 of 73

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 mg/L or milligrams per liter and not 

megagram per liter); nevertheless, the results were sometimes recorded in μg/L (micrograms per 

liter).

The assumptions in Table 3-5 can be used to estimate intakes from recycled uranium contaminants 

and thorium once the uranium intake has been calculated from bioassay.

Table 3-5. Assumptions for intake determinations from uranium bioassay.

Radionuclide Type

Intake activity

ratio compared 

to uranium Applicable period Note

U M, S 1 January 1, 1962–present Can assume U-234 for 

dose calculations

Pu-239 M, S 0.00246 January 1, 1962–present

Np-237 M 0.00182 January 1, 1962–present

Th-228 M, S 0.00161 January 1, 1962–December 31, 1971

Th-232 M, S 0.00161 January 1, 1962–December 31, 1971

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 prior to uranium extrusions, it appears favorable to 

the claimant to assume that residual contamination resulted in thorium intakes in 1965 and 1968 when 

no thorium (232Th and 228

Th) processing was recorded.

4.0

RMI employees were exposed to radiations 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.

EXTERNAL DOSE

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

containing 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 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) and later Form NRC 5.

4.1 PENETRATING DOSES

Table 4-1 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.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 27 of 73

Routine film badge results included whole-body beta and gamma monitoring results. Table 4-1 

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

Neutron doses appear to have been monitored for most RMI employees in 1986, 1987, and 1988.

Table 4-1. External dosimetry

Period

External dosimetry 

vendor

Exchange frequency/

monitoring method Detection threshold reported in records

1962–1966 Controls for Radiation,

Cambridge, MA

Monthly/

Film

A zero indicates less than minimum detectable dose 

of 5 mrem for X-ray and gamma <175 keV or 10 

mrem for hard X-ray and gamma and beta.

1967–1987 USTC,

Richland, WA

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. 

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 117-keV

photons. 

A 1983 film badge report indicates film detection limits 

of 10 mrem for photons and 20 mrem for beta. 

1986 and 1987 TLD badge reports indicate detection 

limits of 10 mrem for X-ray, gamma, beta, and 

neutron radiation.

1988–1991 USTC,

Richland, WA, and 

Westinghouse Materials 

Company of Ohio at 

FEMP

Monthly/

TLD

1988 and 1989 TLD badge reports from USTC 

indicate detection limits of 10 mrem for X-ray, gamma 

ray, beta, and neutron radiation. The minimum 

detection limit for FEMP TLDs was 30 mrad beta and 

gamma from 1989 to 1992. 

1992– 2004 Westinghouse Materials 

Company of Ohio at 

FEMP, TMA Eberline, 

ThermoNUtech, Landauer, 

ICN, and Global Dosimetry 

Solutions

Probably quarterly/

TLD

The minimum detectable level for FEMP TLDs was 30 

mrad beta and gamma from 1989 to 1992. Other

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 all TLD vendors 

were DOE Laboratory Accreditation Program or 

National Voluntary Laboratory Accreditation Program

certified by 1992. 

Table 4-2 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/or temporary workers beginning in the 1980s. The doses in Table 4-2 do not include 

nonpenetrating and extremity doses and Table 4-2 only includes NCRP 38 (1971) neutron doses 

measured on TLD badges in 1986, 1987 and 1988. Attachment E contains a more detailed 

breakdown of doses for years in which they are available beginning in 1974.

Table 4-2. Penetrating whole-body external doses, 1962–1999

Year

Number not 

monitored

Number 

monitored

Dose (rem)

0–1 1–2 Reference

1999 Unknown 1,202 1,202 Eberline 2000

1998 Unknown –a – –

1997 Unknown 290 290 TN 1997

1996 Unknown 285 285 TN 1996

1995 Unknown 255 255 TMA 1995

1994 Unknown 332 332 TMA 1994

1993 Unknown – – –

1992 Unknown – – –

1991 Unknown – – –

1990 Unknown 282 282 Rizzi 1991
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 28 of 73

Year

Number not 

monitored

Number 

monitored

Dose (rem)

0–1 1–2 Reference

1989 Unknown – – –

1988 Unknown 920 920 Author unknown

1987 Unknown 822 822 Brewer 1988

1986 Unknown 134 130 4 Manninen 1988

1985 Unknown 124 124 Manninen 1988

1984 Unknown 117 117 Schaeffer 1985

1983 Unknown 118 118 Schaeffer 1984

1982 Unknown 116 116 Schaeffer 1983

1981 Unknown 89 89 Van Looke 1982

1980 Unknown 80 80 Schaeffer 1981

1979 Unknown 80 80 Van Looke 1980

1978 Unknown 68 68 Schaeffer 1979

1977 Unknown 62 62 Schaeffer 1978

1976 Unknown 58 58 Schaeffer 1977

1975 Unknown 62 62 Heiser 1976

1974 Unknown 69 69 Heiser 1975

1973 27 53 53 Travis 1974

1972 20 46 46 Travis 1973

1971 24 46 46 Lenhard 1972

1970 25 47 47 Lenhard 1971

1969 33 41 41 Smith 1970

1968 31 62 62 Hibbits 1969

1967 28 65 65 Lenhard 1968

1966 27 57 57 Sapirie 1967

1965 8 66 66 Sapirie 1966

1964 14 81 81 Sapirie 1965

1963 9 71 70 1 Sapirie 1964

1962 7 52 51 1 Sapirie 1963

a. – = No data available.

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 Plant operation, a strip of indium metal foil was 

placed in badges for criticality monitoring. In 1986, 1987, and 1988, neutron monitoring was 

performed in addition to beta, X-ray, and gamma monitoring based on a recommendation from the 

Battelle Pacific Northwest Laboratory RMI Radiation Protection Program Review. Dosimetry badges 

were exchanged monthly from 1962 through 1966 and quarterly or monthly thereafter. 

Film badge detection thresholds are assumed to be 40 mrem for beta and gamma (ORAUT 2004b). 

The least sensitive beta-gamma TLD badge vendor appears to be from the FEMP with a detection 

threshold of 30 mrad beta and gamma (ORAUT 2004c). The sensitivity of the USTC neutron TLD 

badges was reported as 10 mrem in 1986, 1987, and 1988. 

The overall uncertainty in recorded dose is dependent upon: 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 (2005d). The uncertainty 

associated with beta-gamma, film badge results from 1962 through 1985 are estimated to be + 60%,

and the uncertainty associated with beta-gamma, TLD badge results from 1986 through 2004 doses 

are estimated to be + 30%. The uncertainty associated with neutron TLD badge results from 1986, 

1987 and 1988 are estimated to be +100%. 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 29 of 73

4.2 NONPENETRATING DOSE

The nonpenetrating dose has not been tabulated for this site. At RMI’s predecessor operation, the 

Bridgeport Brass Adrian Plant, the ratio of nonpenetrating to penetrating dose, 7.87, can be 

determined from the information in ORAUT (2005b). This ratio can be used to estimate 

nonpenetrating dose when no monitoring is available for workers in the plant area and workers whose

locations are unknown.

4.3 NEUTRON DOSE

No records have been located to indicate the source of neutron exposure. According to the current 

RMI Health Physics Manager, neutron monitoring was performed based on a recommendation during 

the 1985 Battelle Pacific Northwest Laboratory review of RMI’s radiation protection program. 

Therefore, the neutron dose must be attributable to spontaneous fission neutrons and, to a much 

lesser extent, from alpha-neutron reactions. Of the 379 available quarterly TLD badge results from 

1986 and 1987, 38 were reported at or above the recording level of 10 mrem with the highest 

quarterly neutron dose reported as 90 mrem and the highest annual dose reported as 140 mrem (no 

missed dose was considered). A maximum quarterly dose of 90 mrem can be used to account for 

individuals with just some unmonitored quarters, and a maximum annual dose of 230 mrem (140 

mrem, the maximum dose for three quarters, plus 90 mrem to account for the unmonitored quarter) 

can be used to account for unmonitored neutron dose in areas such as the Billet Storage Warehouse, 

Main Plant extrusion area and the RF-6 and RF-3 Butler Buildings, where large amounts of source 

material were handled or stored from 1962 through 1991. For neutrons generated by spontaneous 

fission neutrons and alpha-neutron reactions, all of the neutron doses should be assumed to result 

from the 0.1 – 2 MeV energy group. It is not known for certain how the TLDs for neutron dose were 

calibrated in 1986 to 1988. For dose reconstruction purposes it is assumed that the badges 

measured the NCRP 38 (1971) dose which must be converted to ICRP 60 (1990) neutron dose using 

a factor of 1.91 (ORAUT 2006b). 

4.4 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 annually. 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 performed from July 1 to December 31, 1974, are listed in Table 4-3. According to 

correspondence from RMI to the Energy Research and Development Administration (a DOE 

predecessor agency) in January 1977, ring badge dosimeters had not yet been used at RMI 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 Extrusion Plant (Van Loocke 1977).
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 30 of 73

Table 4-3. Forearm and chest skin dose 

monitoring results for personnel who closely 

handled or inspected uranium billets in 1974.

a

Job/task

Wrist badge 

(mrem/year)

Chest badge 

(mrem/year)

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. Fletcher (1976).

A study was performed in the November 1985 timeframe 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 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% low compared to 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 underresponded (Munson and Stacy 1985).

Even though 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; however, by June 4, 1985, this was corrected by placing the badges into a properly

sized heat-sealed plastic holders (ORNL 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 be about a factor of 5. The ORAUT

(2005a) methodology can be used to estimate a maximum extremity dose.

4.5 OCCUPATIONALLY REQUIRED MEDICAL X-RAYS

The medical program described in 1969 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. In addition, hourly employees (laborers) are assumed to have received one single 

preemployment back lumbar spine X-ray from 1962 through 1975 and laborers are assumed to have 

received pre-employment lumbar spine X-rays prior to 1986 (Hibbitts and Pryor 1970; Johnson 1976). 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 31 of 73

No information is currently available regarding the type of X-ray equipment and the techniques used 

for RMI exams. With the exception of female clerical personnel, this document assumes that 

preemployment, annual, and termination posterior-anterior (PA) chest X-ray examinations were 

performed on all personnel from 1962 to the present and that one preemployment lumbar-spine X-ray 

series, assumed to consist of two anterior-posterior (AP) and two lateral views in accordance with 

OCAS 2005c, was performed on hourly personnel before 1986. ORAUT-OTIB-0006 (ORAUT 2005c 

or current version) can be used to determine the dose from occupationally required medical X-ray 

procedures. There is no Extrusion Plant documentation that indicates that photofluorography was 

performed. 

4.6 MISCELLANEOUS INFORMATION CONCERNING EXTERNAL EXPOSURE

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 RMI Ashtabula facility. Consideration should be given to assigning the other industrial 

radiation dose found in ORAUT (2005a) to unmonitored workers whose claims indicate analytical 

laboratory work.

4.7 EXTERNAL DOSE ASSIGNMENT SUMMARY 

Many RMI employees were monitored for beta, X-ray, and gamma external exposures. The data in 

Table 4-2 indicates that the annual 95th percentile dose falls into the 0 to 1 rem dose range. For 

periods when no gamma monitoring results are available for an individual, it is favorable to the 

claimant to assign an annual penetrating exposure of 1 rem.

ORAUT (2005b) external dose methodology can be used to reconstruct dose for unmonitored workers 

or unmonitored periods when another method is not specified.

Table 4-4. External dose assumptions.

Radiation 

type

Assumed energy 

and exposure 

assumptions Monitoring method

Assumed 

detection 

threshold Notes

Photons 30 to 250 keV, 

acute, 100% AP

Film badge (1962–1985)

TLD (1986–present)

40 mrem

30 mrem

Default dose for the unmonitored 

worker is either the 95th percentile 

dose in Attachment E or 1 rem/y. 

Interactive RadioEpidemiological 

Program (IREP) distribution is 

constant for default assumption.

Electrons >15 keV, acute Film badge (1962–1985)

TLD (1986–Present)

40 mrem

30 mrem

Default dose is 7.87 times the 

photon dose.

Extremities Film (1974–1985)

TLD (1986–2006)

40 mrem

30 mrem

Multiply chest badge dose by 3.

Fingers Multiply extremity dose by 1.7 or 

chest badge dose by 5.

PA chest 

X-ray

Use assumptions 

in ORAUT 2005c

or current version

Initial plus one examination per year 

plus a termination examination is the 

default assumption. Female clerical 

employees are assumed to have had 

an initial and termination chest X-ray.

2 AP and 2 

lateral 

lumbar 

spinal X-rays

Use assumptions 

in ORAUT 2005c 

or current version

One initial exam for hourly laborers 

from January 1, 1962, through 1985.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 32 of 73

Radiation 

type

Assumed energy 

and exposure 

assumptions Monitoring method

Assumed 

detection 

threshold Notes

Neutron 0.1 – 2 MeV TLD (1986–1988) 10 mrem Default neutron dose is 90 mrem per 

quarter, but no more than 230 mrem 

per year (439 mrem per year ICRP 

60 dose) for unmonitored periods 

from January 1, 1962, through 

October 14, 1991. IREP distribution 

is constant for default assumption.

5.0

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

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

intakes occurred from residual uranium contamination on site structures and components and on site 

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 

contractor (LATA-SHARP Remediation Services LLC), 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 completed and work at the site consisted primarily of soil and ground-water 

remediation. 

INTERNAL DOSE DURING THE POSTPRODUCTION PERIOD

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

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

from locations within the main plant area and the property located 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 1995b). Using a mass loading factor of 2 x 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 recycle uranium contaminants would be an inconsequential component. The 

mass loading factor of 2 x 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 resulting in the daily ingestion of 

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

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

that the soil concentrations are lognormally distributed and that the average uranium soil 

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

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

95th percentile (RMI 1995b). The resulting geometric standard deviation is 3.4. 

6.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 33 of 73

REFERENCES

AEC (U.S. Atomic Energy Commission), 1964, “Form NY-99”, Health and Safety Laboratory, New 

York Operations Office, October 24. [p.12 of EmlBx7ReacMetalsIncMultiSampTypes1964.pdf] 

AEC (U.S. Atomic Energy Commission), 1973, “Source Material License,” License Number SMB-602, 

Amendment 1, for RMI Company, with Supplemental Sheets, August 31. [pp. 48-57 of 

12388.pdf] 

Author unknown, “RMI Extrusion Plant, Ashtabula, Ohio, Supplement to DOE 5484.1, Year - 1988.”

[pp. 120–123 of 030012380.pdf]

Aldridge, R., 1990, “Statistical Method for Calculating Internal Exposure,” interdepartmental 

memorandum to G. Gammon, Ashtabula, Ohio, March 28. [pp. 2–4 of 030013110. pdf]

Bean, R. L., 1964, “Action on Recommendation to Report of Annual Health Protection Review –

Reactive Metals, Inc. Extrusion Plant – 1964,” letter to J. W. Ruch (U.S. Atomic Energy 

Commission), Reactive Metals, Inc., Ashtabula, Ohio, October 10. [pp. 3–4 of 240000874.pdf]

Bean, R. L., 1965a, “Application for Renewal of Source Material License SMB-602,” letter to U.S. 

Atomic Energy Commission, Washington, D.C., April 27. [pp. 27–28 of 030012388.pdf]

Bean, R. L., 1965b, “Report on the Annual Health Protection Review of Reactive Metals, Inc., 

Extrusion Plant – 1965,” letter to J. W. Ruch (U.S. Atomic Energy Commission), Reactive 

Metals, Inc., Ashtabula, Ohio, June 2. [pp. 10–11 of 240000701.pdf] 

Bean, R. L., 1973, “Application for License Renewal - Source Material License No. SMB-602 Docket 

No. 40-2384,” letter to James C. Malero (U.S. Atomic Energy Commission), Reactive Metals, 

Inc., Ashtabula, Ohio, August 31. [pp. 37–47 of 030012381. pdf]

Breslin, A. J., and H. Glauberman, 1964, Reactive Metals Inc., Ashtabula, Ohio, Occupational Health 

and Dust Control Ventilation Survey, Occupational Exposure to Airborne Contaminants, 

Technical Memorandum 64-17, U.S. Atomic Energy Commission, New York Operations Office, 

Health and Safety Laboratory, October 12. [pp. 5–55 of 240000876.pdf] 

Brewer, N. F., 1987, “Annual Occupational Radiation Exposure Report – 1986,” letter to M. R. Theisen 

(U.S. Department of Energy), Reactive Metals, Inc., July 14. [pp. 2–5 of 030012380.pdf]

Brewer, N. F., 1988, “Annual Occupational Radiation Exposure Report – 1987,” letter to L. C. Bogar 

(Westinghouse Materials Corporation), August 22. [pp. 102–106 of 030012380.pdf]

Brewer, N. F., 1990, letter requesting renewal of RMI’s radioactive materials license to G.M. McMann 

[sic McCann] (U.S. Nuclear Regulatory Commission), Reactive Metals, Inc., March 6. [pp. 254

of 030012388.pdf]

Britcher, L. J., 1992, “Research of Thorium Extrusion and Processing Procedures at RMI,” 

memorandum to J. W. Henderson (Reactive Metals, Inc.), December 4. [pp. 2–3. of 

030012417.pdf]

CEP (Controls for Environmental Pollution, Inc.), 1991, no title – bioassay sample results, Sante Fe, 

New Mexico. [pp 30-41 of 030013204.pdf]
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 34 of 73

Davis, S. L., 1987, Health Physics Appraisal of RMI Company, Extrusion Plant, Nuclear Safety 

Branch, Oak Ridge, Tennessee, May 18–22. [pp. 4-17 of 030012412. pdf]

DOE (U.S. Department of Energy), 1993, Environmental Assessment for Decommissioning and 

Remedial Activities at the RMI Titanium Company Extrusion Plant Ashtabula, Ohio, Chicago, 

Illinois, September. [p. 13 of 030012399. pdf]

DOE (U.S. Department of Energy), 1995, Uranium Oxide Fire Incident Review, RMI Decommissioning 

Project Ashtabula, Ohio, November 2-3, 1995, Ohio Field Office, Miamisburg, Ohio. 

[030013178.pdf]

DOE (U.S. Department of Energy), 2000, DOE Ohio Sites Recycled Uranium Project Report, DOEOH-00-0001, Washington D.C., May 15. [p. 43, p. 91, pp. 260-307, and pp. 815–828 of 

010001306\Disk 1 (Ohio Field Office)\DOE-OH-00-0001 Ohio Field Office Recycled 

Uranium.pdf]

DOE (Department of Energy), 2004, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities. DOE-STD-1136-2004, December. 

[http://www.hss.energy.gov/NuclearSafety/techstds/standard/std1136/doe-std-1136-2004.pdf]

DOE (U.S. Department of Energy), 2006, Office of Worker Advocacy Web site,

http://www.eh.doe.gov/advocacy/faclist/showfacility.cfm, accessed October 11, 2006.

Eberline Dosimetry Services, 2000, March 22. [p. 2 of 030012370.pdf]

Featsent, N., 1975, “Alpha Contamination Survey Data Sheet,” Reactive Metals, Inc., Ashtabula, 

Ohio, October 28. [p. 18 of 030013125.pdf]

Featsent, N., 1977a, “Alpha Contamination Survey Data Sheet,” Reactive Metals, Inc., Ashtabula, 

Ohio, May 6. [p. 26 of 030013124.pdf]

Featsent, N., 1977b, “Alpha Contamination Survey Data Sheet,” Reactive Metals, Inc., Ashtabula, 

Ohio, May 6. [p. 20 of 030013126.pdf]

Fletcher, H. D., 1976, “Annual Health Protection, Nuclear Criticality and Transportation Safety 

Appraisal of RMI, 1976,” letter to R. D. Heiser (Reactive Metals, Inc.), U.S. Energy Research 

and Development Administration, June 18. [pp. 27–53 of 030012407.pdf]

Gammon, G. A., 1992, RMIEP Urine Bioassay Program, RMI-L-60.031, Reactive Metals, Inc., 

Environmental, Safety, and Health Department, Ashtabula, Ohio, May 1. [030013137.pdf]

Haywood, F. F., et al, 1982, Radiological Survey of the Former Bridgeport Brass Company, Special 

Metals Extrusion Plant, Adrian, Michigan, Oak Ridge, Tennessee, April. [Ref ID 9852, p. 10)

Helgeson Scientific Services, 1997, no title - report of in vivo counting, Pleasanton, California, October 

14. [030012382. pdf]

Heiser, R. D., 1975, “Annual Occupational Radiation Exposure Report – 1974,” letter to H. D. Fletcher 

(U.S. Energy Research and Development Administration), Reactive Metals, Inc., Ashtabula, 

Ohio, May 28. [pp. 174–176 of 030012380.pdf]
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 35 of 73

Heiser, R. D., 1976, “Annual Occupational Radiation Exposure Report – 1975,” letter to H. D. Fletcher 

(U.S. Energy Research and Development Administration), Reactive Metals, Inc., Ashtabula, 

Ohio, March 19. [pp. 172–173 of 030012380.pdf] 

Henderson, J. W., 1997, letter containing RMI’s responses to NRC comments on RMI’s 

Decommissioning Plan to R. Uleck (US Nuclear Regulatory Commission), RMI, Ashtabula, 

Ohio, April 15. [pp. 2–4 of 030010820.pdf]

Hibbitts, H. W., 1969, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1968],” Form AEC-190, U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee. [p. 92 of 030001115.pdf]

Hibbits, H. W., and W. A. Pryor, 1970, Health and Nuclear Safety Appraisal Reactive Metals Inc. 

Extrusion Plant, Oak Ridge, Tennessee, May. [030013097.pdf.]

Hibbits, H. W., and J. F. Wing, 1980, “USDOE – ORO Environmental Management Appraisal of RMI 

Company,” March. [pp. 7–15 of 030012393.pdf] 

Hysong, R., 2006a, telephone conversation with Eric Marsh, October 20. 

Hysong, R., 2006b telephone conversation with Eric Marsh, October 30. 

ICRP, (1990) 1990 Recommendations of the International Commission on Radiological Protection, 

ICRP Publication 60, Annals of the ICRP Vol. 21, Pergamon Press, Oxford, England.

Jefferson, D. R., 1961, “Application for Source Material License,” letter to U.S. Atomic Energy 

Commission (Washington, D.C.), Bridgeport, Connecticut, December 11. [Ref ID: 25415, p. 

11]

Jefferson, D. R., 1962b, “Supplementary Information to Source Material License Application –

Bridgeport Brass Company,” letter to U.S. Atomic Energy Commission (Washington, D.C.), 

May 18. [pp. 18–22 of 030012388.pdf]

Jefferson, D. R., 1962a, “Supplementary Information to Source Material License Application –

Bridgeport Brass Company,” letter to U.S. Atomic Energy Commission (Washington, D.C.), 

May 8. [pp. 12–17 of 030012388.pdf]

Jefferson, D. R., 1962c, “Supplementary Information to Source Material License Application -

Bridgeport Brass Company,” letter to U.S. Atomic Energy Commission (Washington, D.C.), 

February 23. [pp. 2–8 of 030012388.pdf]

Jelinek, T. M., 1974, “USAEC – ORO Health Protection Appraisal of RMI Company,” June. [pp. 22–

26 of 030012407.pdf] 

Johnson, W. E., 1976, US ERDA – ORO Annual Health Protection, Nuclear Criticality and 

Transportation Safety Appraisal, Oak Ridge, Tennessee, April. [pp. 28–44 of 030012407.pdf

Koh (B. Koh & Associates), 1995, Site Characterization Report for the RMI Titanium Company 

Extrusion Plant, Owings Mills, Maryland, April. [030010711.pdf]
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 36 of 73

Lenhard, J. A., 1968, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1967],” letter to D. M. Ross (U.S. Atomic Energy 

Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, April 15. [pp. 65–66 of 

030001114.pdf]

Lenhard, J. A., 1971, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1970],” letter to M. B. Biles (U.S. Atomic Energy 

Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 18. [pp. 40–41 of 

030001117.pdf]

Lenhard, J. A., 1972, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1971],” letter to M. B. Biles (U.S. Atomic Energy 

Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 9. [pp. 52–53 of 

030001118.pdf]

Manninen, B. S., 1986, “Bioassay protocol [sic],” letter to General Distribution, Ashtabula, Ohio, 

November 6. [p. 823 of 030013203. pdf]

Manninen, B. S., 1988, “Investigation of Increased Dose Rate in 1986,” memorandum to R. L. Mason, 

Ashtabula, Ohio, January 28. [pp. 109-111 of 030012380.pdf]

Marsh, E. P., 1990, untitled letter [request to renew RMI Extrusion plant’s radioactive materials license 

SMB-602] to George McMann [sic McCann] (Chief, Material Licensing Section, U.S. Nuclear 

Regulatory Commission, Region III), May 31. [030013199.pdf]

McDougal, B. E., 1980, memorandum to R. C. Baker, “In Vivo Radiation Monitoring Results, Second 

Run – 1979,” Paducah, Kentucky, March 24. [Ref ID 18774, pp. 3-9]

Munson, L. F., 1985, “Upgrading the RMI Radiation Protection Program,” letter to F. Van Loocke 

(Reactive Metals, Inc.), Battelle, Pacific Northwest Laboratory, Richland, Washington, 

November 7. [pp. 12–16 of 030010721.pdf]

Munson, L. F., 1986a, untitled letter regarding a particle size study to B. Manninen (Reactive Metals, 

Inc.), Battelle, Pacific Northwest Laboratory, Richland, Washington, November 10. [pp. 2–5 of 

030010719.pdf]

Munson, L. F., 1986b, untitled letter regarding a particle size study to B. Manninen (Reactive Metals, 

Inc.), Battelle, Pacific Northwest Laboratory, Richland, Washington, March 24. [pp. 6–18 of 

030010719.pdf]

Munson, L. F., and L. C. Stacy, 1985, letter to F. Van Loocke (Reactive Metals, Inc.), Battelle, Pacific 

Northwest Laboratory, Richland, Washington, December 6. [pp. 17–20 of 030010721.pdf]

NCRP, (1971) National Council on Radiation Protection and Measurements. Protection Against 

Neutron Radiation, NCRP Report 38, National Council on Radiation Protection and 

Measurements, Bethesda, Maryland.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 37 of 73

NIOSH (National Institute for Occupational Safety and Health), 2004, Estimation of Ingestion Intakes,

Technical Information Bulletin, OCAS-TIB-009, Revision 0, Office of Compensation Analysis 

and Support, Cincinnati, Ohio, April 13. 

NRC (U. S. Nuclear Regulatory Commission), 1962-1999, licensing documentation for Bridgeport 

Brass (later known as Reactive Metals, Incorporated.; RMI Company, Extrusion Plant; RMI 

Titanium Company, Extrusion Plant; RMI Environmental Services; and Earthline 

Technologies), License Number SMB-602. [Ref ID: 24109]

NRC (U. S. Nuclear Regulatory Commission), 1994, “Information Notice 94-81” 

http://www.nrc.gov/reading-rm/doc-collections/gen-comm/info-notices/1994/in94081.html

NRC (U. S. Nuclear Regulatory Commission), 1979-1980, licensing documentation for RMI Company, 

Extrusion Plant, License Number SMB-602. [pp 294-298 of 030012388.pdf]

NRC (U. S. Nuclear Regulatory Commission), 2002, Nuclear Regulatory Legislation, 107th Congress; 

1st Session, NUREG-0980, Vol. 1, No.6, Office of the General Counsel, Office of the General 

Counsel, Washington, DC 20555–0001, June. [http://www.nrc.gov/reading-rm/doccollections/nuregs/staff/sr0980/ml022200075-vol1.pdf#pagemode=bookmarks&page=14]

ODH (Ohio Department of Health), 1999-2004, licensing documentation for the RMI Environmental 

Services (later known as Earthline Technologies), License Number 11900040004. [Ref ID: 

24109]

ORAU (Oak Ridge Associated Universities), 1985, Environmental Program Review of the RMI 

Company Extrusion Plant, Ashtabula, Ohio, Oak Ridge, Tennessee, August. [030010428.pdf]

ORAUT (Oak Ridge Associated Universities Team), 2004a, Technical Basis Document for the Fernald 

Environmental Management Project (FEMP) – Occupational Internal Dose, ORAUT-TKBS0017-5, Revision 00, Oak Ridge, Tennessee, May 28.

ORAUT (Oak Ridge Associated Universities Team), 2004b, Technical Information Bulletin: A 

Standard Complex-Wide Correction Factor for Overestimating External Doses Measured with 

Film Badge Dosimeters, ORAUT-OTIB-0010, Revision 00, Oak Ridge, Tennessee, January 

12.

ORAUT (Oak Ridge Associated Universities Team), 2004c, Technical Basis Document for the Fernald 

Environmental Management Project (FEMP) – Occupational External Dose, ORAUT-TKBS0017-6, Revision 00, Oak Ridge, Tennessee, April 20.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Estimating the Maximum Plausible Doses 

to Workers at AWE Facilities, ORAUT-OTIB-0004, Rev. 02, Oak Ridge, Tennessee, 

August 12.

ORAUT (Oak Ridge Associated Universities Team), 2005b, An Exposure Matrix for Bridgeport Brass: 

Havens Laboratory and Adrian Plant, ORAUT-TKBS-0030, Oak Ridge, Tennessee, 

September 15.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 38 of 73

ORAUT (Oak Ridge Associated Universities Team), 2005c, Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Procedures, ORAUT-OTIB-0006, Rev. 03 PC-1, Oak Ridge, 

Tennessee, December 21.

ORAUT (Oak Ridge Associated Universities Team), 2005d, Technical Basis Document for the 

Savannah River Site, ORAUT-TKBS-0003, Revision 03, Oak Ridge, Tennessee, April 5.

ORAUT (Oak Ridge Associated Universities Team), 2006a, Technical Basis Document for the 

Paducah Gaseous Diffusion Plant – Occupational Internal Dose, ORAUT-TKBS-0019-5, 

Revision 01, Oak Ridge, Tennessee, October 25.

ORAUT (Oak Ridge Associated Universities Team), 2006b, Technical Basis for the Conversion from 

NCRP Report 38 Neutron Quality Factors to ICRP Publication 60 Radiation Weighting Factors 

for Respective IREP Input Neutron Energy Ranges, ORAUT-OTIB-0055, Revision 00, Oak 

Ridge, Tennessee, June 5.

ORNL (Oak Ridge National Laboratory), 1985, Report on Martin Marietta Energy Systems May 15, 

1985 Site Visit of RMI Extrusion Plant, Oak Ridge, Tennessee, June 4. [pp. 2-31 of 

030012413.pdf]

Puterbaugh, J. F., and F. G. Van Loocke, 1964, Standard Operating Procedure for Primary Extrusion 

of 1.95% Enriched Uranium for the NPR Program, ASHTA-263, Reactive Metals, Inc., 

Ashtabula, Ohio, proposed September 4. [pp. 5–46 of 240000877.pdf]

RMI (Reactive Metals, Inc.), 1967, Liquid Waste – Bioassay Analysis Report. Report Nos. U122-67 

and U123-67 [pp. 28-29 of 030013201.pdf]

RMI (Reactive Metals, Inc.), 1973, Health Protection Rules and Regulations, RMI-L-60, revised 

December, Ashtabula, Ohio. [pp. 2–19 of 030013172. pdf]

RMI (Reactive Metals, Inc.), 1974, Liquid Waste – Bioassay Analysis Report. Report No. U118-74 [p. 

232 of 030013210.pdf]

RMI (Reactive Metals, Inc.), 1989a, RMI Company Extrusion Plant Multi-Discipline Review Response, 

Ashtabula, Ohio, October 23. [pp. 26–37 of 030013179.pdf]

RMI (Reactive Metals, Inc.), 1989b, Radiation Survey of Interoffice Airspaces, Ashtabula, Ohio, 

September 16. [030013181.pdf]

RMI (Reactive Metals, Inc.), 1991, RMI Company Extrusion Plant Health Physics Manual, Ashtabula, 

Ohio, February. [030010714.pdf]

RMI (Reactive Metals, Inc.), 1994, The Reporting/Documenting of External Radioactive 

Contamination of Personnel and/or Their Belongings, Environmental Services Division, 

Ashtabula, Ohio, March 3. [030013138.pdf]

RMI (Reactive Metals, Inc.), 1995a, Decommissioning Plan for RMI Titanium Company Extrusion

Plant, Environmental Services Division, Ashtabula, Ohio, April. [030010710.pdf]
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 39 of 73

RMI (Reactive Metals, Inc.), 1995b, Environmental Report for Decommissioning Activities at the RMI

Titanium Company Extrusion Plant, Ashtabula, Ohio, RDP-ESH-009, Rev. 0, Environmental 

Services Division, Ashtabula, Ohio, April 28. [030010572.pdf p.19-29]

Rizzi, J. N., 1991, “Annual Occupational Radiation Exposure Report – 1990,” letter to W. Millet (EG&G 

Idaho), Reactive Metals, Inc., Ashtabula, Ohio, August 27. [pp. 1–5 of 030013182.pdf]

Ruch, W. R., 1964, “Report of the Annual Health Protection Review of Reactive Metals, Inc., Extrusion 

Plant-1964,” letter to R. L. Bean (Reactive Metals, Inc.), U.S. Atomic Energy Commission, Oak 

Ridge Operations Office, Oak Ridge, Tennessee, August 10. [pp. 5–9 of 240000874.pdf]

Ruch, W. R., 1965, “Report of the Annual Health Protection Review of Reactive Metals, Inc., Extrusion 

Plant-1965,” letter to R. L. Bean (Reactive Metals, Inc.), U.S. Atomic Energy Commission, Oak 

Ridge Operations Office, Oak Ridge, Tennessee, April 21. [pp. 12–15 of 240000701.pdf]

Saito, E. F., 1993, “Thorium-232 Contamination,” memorandum to E. P. Marsh, Reactive Metals, Inc., 

Ashtabula, Ohio, June 24. [p.4 of 030012417]

Sapirie, S. R., 1963, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1962],” letter to N. H. Woodruff (U.S. Atomic 

Energy Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 13. [pp. 41–42 of 

030001109.pdf]

Sapirie, S. R., 1964, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1963],” letter to N .H. Woodruff (U.S. Atomic 

Energy Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 4. [pp. 100–104 of 

030001110.pdf]

Sapirie, S. R., 1965, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1964],” letter to N. H. Woodruff (U.S. Atomic 

Energy Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 15. [pp. 64–66 of 

030001111.pdf]

Sapirie, S. R., 1966, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1965],” letter to P. A. Morris (U.S. Atomic Energy 

Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 22. [pp. 70–72 of 

030001112.pdf]

Sapirie, S. R., 1967, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1966],” letter to M. B. Biles (U.S. Atomic Energy 

Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, April 6. [pp. 84–85 of 

030001113.pdf]

Schaeffer, M. R., 1977, “Annual Occupational Radiation Exposure Report – 1976,” letter to H. D. 

Fletcher (U.S. Energy Research and Development Administration), Reactive Metals, Inc., 

Ashtabula, Ohio, April 15. [pp. 37–40 of 030012380.pdf]
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 40 of 73

Schaeffer, M. R., 1978, “Annual Occupational Radiation Exposure Report – 1977,” letter to H. D. 

Fletcher (U.S. Energy Research and Development Administration), Reactive Metals, Inc., 

Ashtabula, Ohio, April 3. [pp. 37–40 of 030012380.pdf]

Schaeffer, M. R., 1979, “Annual Occupational Radiation Exposure Report – 1978,” letter to H. D. 

Hickman (U.S. Department of Energy), Reactive Metals, Inc., Ashtabula, Ohio, March 23. 

[pp. 32–36 of 030012380.pdf]

Schaeffer, M. R., 1981, “Annual Occupational Radiation Exposure Report – 1980,” letter to H. D. 

Hickman (U.S. Department of Energy), Reactive Metals, Inc., Ashtabula, Ohio, March 26. 

[pp. 19–21 of 030012380.pdf]

Schaeffer, M. R., 1983, “Annual Occupational Radiation Exposure Report – 1982,” letter to M. R. 

Theisen (U.S. Department of Energy), Reactive Metals, Inc., Ashtabula, Ohio, March 14. 

[pp. 10–13 of 030012380.pdf]

Schaeffer, M. R., 1984, “Annual Occupational Radiation Exposure Report – 1983,” letter to M. R. 

Theisen (U.S. Department of Energy), Reactive Metals, Inc., Ashtabula, Ohio, March 9. 

[pp. 6-9 of 030012380.pdf]

Schaeffer, M. R., 1985, “Annual Occupational Radiation Exposure Report – 1984,” letter to M. R. 

Theisen (U.S. Department of Energy), Reactive Metals, Inc., Ashtabula, Ohio, March 18. 

[pp. 2–5 of 030012380.pdf]

Scott, L.M., 1969, “Design and Development of a Mobile In Vivo Radiation Monitoring Laboratory,” 

American Industrial Hygiene Association Journal, Volume 30, March-April. [Ref ID 8495]

Smith, R. D., 1973, “USAEC – ORO Health Protection Appraisal of RMI Company,” July. [pp. 17–20 

of 030012407.pdf] 

Smith, R. D., 1970, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1969],” Form AEC-190, Atomic Energy

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee. [p. 43 of 030001116.pdf]

Tauche, R.M., 1998, Dosimetry Reports. Thermo NUtech. Albuquerque, New Mexico.

TMA (TMA Eberline), 1994a, Current TLD Occupational Radiation Exposure Report-Summary. 

[030012361.pdf]

TMA (TMA Eberline), 1994, Current TLD Occupational Radiation Exposure Report-Summary. 

Albuquerque, New Mexico. [p. 2 of 030012361.pdf]

TMA (TMA Eberline), 1996, Current TLD Occupational Radiation Exposure Report-Summary. 

Albuquerque, New Mexico. [p. 89 of 030012378.pdf]

TN (Thermo Nutech), 1996, untitled, Albuquerque, New Mexico. [p. 2 of 030012369.pdf]

TN (Thermo Nutech), 1997, untitled, Albuquerque, New Mexico. [p. 2 of 030012377.pdf]

Travis, W. H., 1974, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1973],” letter to M. B. Biles (U.S. Atomic Energy 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 41 of 73

Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 4. [pp. 43–44 of 

030001120.pdf]

Travis, W. H., 1973, “Summary of Whole Body Radiation Exposure to External Penetrating Radiation 

Exposure Accumulated During the Year [CY-1972],” letter to M. B. Biles (U.S. Atomic Energy 

Commission, Division of Operational Safety, Washington, D.C.), U.S. Atomic Energy 

Commission, Oak Ridge Operations Office, Oak Ridge, Tennessee, March 6. [pp. 42–43 of 

030001119.pdf]

USTC (United States Testing Company, Inc.), 1974-1982, Current Radiation Report, final quarter 

reports. [030012355.pdf]

USTC (United States Testing Company, Inc.), 1983, Current Radiation Report, final quarter report. 

[030012356.pdf]

USTC (United States Testing Company, Inc.), 1984, Current Radiation Report, final quarter reports. 

[030012357.pdf]

USTC (United States Testing Company, Inc.), 1985, Current Radiation Report, final quarter reports. 

[030012358.pdf]

Van Loocke, F. G., 1977, “Recommendation Regarding Reduction of the Occupational Exposure Limit 

for the Hands,” letter to H. D. Fletcher (U.S. Nuclear Regulatory Commission), Reactive 

Metals, Inc., Ashtabula, Ohio, January 3. [p. 3 of 030013122.pdf]

Van Loocke, F. G., 1979, Annual Whole Body Exposure Report”, letter to U.S. Nuclear Regulatory 

Commission, Reactive Metals, Inc., Ashtabula, Ohio, March 27. [p. 161 of 030012380.pdf]

Van Loocke, F. G., 1980, “Annual Whole Body Exposure Report,” letter to U.S. Nuclear Regulatory 

Commission, Reactive Metals, Inc., Ashtabula, Ohio, April 8. [pp. 25–27 of 030012380.pdf]

Van Loocke, F. G., 1982, “Annual Whole Body Exposure Report- 1981,” letter to U.S. Nuclear 

Regulatory Commission, Reactive Metals, Inc., Ashtabula, Ohio, February 25. [pp. 16–18 of 

030012380.pdf]

Vogel, R., 2006, telephone conversation with Eric Marsh, September 14. 

Westinghouse Materials Company of Ohio, 1989, RMI Extrusion Plant Multi-Discipline Review, June 

6-8, 1989, Quality and Safety Department, FMPC Restoration Department, Cincinnati, Ohio, 

September 8. [pp. 5-25 of 0300013179.pdf] 

Wing, J. F., 1982, “USDOE – ORO Health Protection Appraisal of RMI Company Extrusion Plant,” 

July. [pp. 3–11 of 030012408.pdf] 

Yu, C., 2001, User’s Manual for RESRAD Version 6, Argonne National Laboratory, Argonne Report 

ANL/EAD-4, Argonne, Illinois, July.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 42 of 73

GLOSSARY

accreditation

Recognition that a dosimeter system has passed the performance criteria of the DOE 

Laboratory Accreditation Program (DOELAP) standard in specified irradiation categories. 

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. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

contamination, radioactive

Radioactive material in an undesired location including air, soil, buildings 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.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. As 

examples, spent (used) fuel elements, byproduct tails, residues from uranium isotope 

separation, and some weapons materials contain DU. DU can be blended with highly 

enriched uranium to make reactor fuel or used as a raw material to produce plutonium.

deep dose equivalent

The dose equivalent at the respective depth of 1.0 cm in tissue.

DOELAP

The DOE Laboratory Accreditation Program accredits DOE site dosimetry programs based on 

performance testing and onsite reviews on a 2-year cycle.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 43 of 73

dose

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See thermoluminescent dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.

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

Radiation-sensitive photographic film in a light-tight wrapping. 

fission 

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 

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

In glass. Outside the living body and in an artificial environment. Typically used for bioassay 

of a contaminant in excreta, such as fecal or urine samples. 

In vivo

In the living. In the living body of a plant or animal. Bioassay counting analysis of 

radionuclides in the human body. 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 44 of 73

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 but often have different physical 

properties.

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 

microcuries per gram).

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

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

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons to high-energy gamma rays. Photons have 

energies between 10 and 100 kiloelectron-volts.

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

Historical value assigned to reflect the average effectiveness of a particular kind of radiation in 

producing biological effects in humans, now called radiation weighting factor. The quality 

factor multiplied by the absorbed dose yields the dose equivalent.

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 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 45 of 73

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 (alpha, beta, neutron) and electromagnetic rays (photons) that travel from 

one point to another, some of which can pass through or partly through solid materials 

including the human body.

radioactive

Giving off ionizing radiation such as alpha particles or X-rays.

radioactivity

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei. See 

radionuclide.

radioactive waste

Radioactive solid, liquid, and gaseous materials for which there is no further use. Wastes are 

generally classified as high-level (with radioactivity as high as hundreds of thousands of curies 

per gallon or cubic foot), low-level (in the range of 1 microcurie per gallon or cubic foot), 

intermediate level (between these extremes), mixed (also contains hazardous waste), and 

transuranic. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural.

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° Celsius and standard atmospheric pressure. 

An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher energy photons (generally greater than 100 kiloelectron-volts). 

thermoluminescent dosimeter (TLD)

A device 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

Radioisotopes of nuclides having an atomic number greater than 92.

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 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 46 of 73

for military and civilian applications. The Energy Research and Development Administration 

and the U.S. Nuclear Regulatory Commission assumed separate duties from the AEC in 1974. 

The U.S. Department of Energy succeeded the Energy Research and Development 

Administration in 1979.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 centimeters; also used to 

refer to the dose recorded.

X-ray radiation

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside the nucleus of an atom. 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 47 of 73

Figure A-1. Plant Layout.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 1 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 48 of 73

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 2 of 20

Table A-1. Key for Figure A-1.

No. Description No. Description

1 Guard House - Contractor Emergency Assembly Area 53 Women's Clean Locker Room

2 Emergency Assembly Area 54 Enclosed Rampway

3 Hazardous Waste Storage Building 55 Storeroom

4 RF-3 Building 56 Lunchroom

4A RF-3 Building Addition 57 Electrical Shop

5 Equipment Cleaning Area 58 Salary Change Room

6 Electrical Substation 59 Lab Tool Room

7 Sewage Disposal 60 Wet Chemical Room

8 Firehouse 61 Metalographical Room

9 Truck Ramp Enclosure 62 Conference Room

10 Shipping and Receiving Back 63 Computer Room

11 Northeast Billet Storage Warehouse 64 Chemist/Technicians Office

12 Accumulator Station 65 Print Room

13 Air Compressor 66 Dark Room

14 Substation 67 Analytical Room

15 3850 Ton Loewy Horizontal Extrusion Press 68 Development Lab and Quality Control Area

16 Die Head 69 350 Ton Lombard Horizontal Extrusion Press and Ventilation

17A Runout Table (Movable Section) 70 #1 Giddings and Lewis CNC Lathe

17B Powered Runout Table 71 #2 Giddings and Lewis CNC Lathe

18 Horizontal Extrusion Quench Tank 72 #3 Giddings and Lewis CNC Lathe

19 Extrusion Cooling and Transfer Table 73 Small Monarch Lathe

20 Rotating NPR Extrusion Cooling Table 74 Lodge and Shipley Lathe

21 Extrusion Transfer Conveyor To Campbell Saw 75 Large Monarch Lathe

22 Abrasive Campbell Saw 76 Rockwell Drill Press

23 Extrusion Transfer Conveyor From Campbell Saw 77 Racine Saw

24 Inspection After Sawing 78 Cincinnati Mill Machine

25 Transfer Table To Roll Straightener 79 Stanley Grinder

26 Roll Straightener Entrance Conveyor 80 Do All Saw

27 Roll Straightener 81 U.S. Electrical Tool Co. Grinder

28 Roll Straightener Exit Conveyor 82 Cincinnati Gilbert Drilling Machine

29 In Process Transfer and Storage Cable 83 Stairway Down To Boiler Room

30 Salt Bath 84 Acid Neutralization Tank

31 Vertical Extrusion Quench Tank 85 Engleberg Belt Sander

32 Sunbeam Furnace 86 Evaporator

33 FSI Furnace 87 Mandrel Quench Tank

34 Salt Bath Loading Area 88 Hazardous Waste Emergency Equipment Building

35 Receipt Inspection Area 89 Toolmex Lathe

36 MK-31 Rinse Tank 90 Lathe Ventilation

37A Commercial Pickle Tank 91 Cooley Furnace

37B Commercial Rinse Tank 92 Substation

38 Extrusion Wash Tank 93 Storeroom Area

39 Extrusion Pickle Tank 94 RF -- 6 Cold Storage Warehouse

40 Inspection Weighing and Packing Table 95 Northwest Storage Building

41 Container Preheat Furnace 96 Stack #1A Press Exhaust

42 Toll Crib 97 Outdoor Substation Addition

43 Gauge Room 98 Stack #3A Filter Building

44 Tool Preheat Furnace 99 North Gisholt Lathe

45 Prefill Tank 100 Fan -- Stack #5A

46 Floor Scale 101 Sandblaster

47 Area Heating Furnace 102 Emergency Generator

48 Engineering Office Area 103 Mezzanine

49 Foremen's Office Area 104 Forge Area Stack #8

50 Men's Hourly Dirty Locker Room 105 Outside Air Sampler

51 Men's Hourly Clean Locker Room 106 Storage Area

52 Women's Dirty Locker Room 107 Extrusion Storage Area
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 49 of 73

No. Description No. Description

108 Auxiliary Storage Area 134 95% and HEPA Filter Housing

109 South Gisholt Lathe 135 Fan -- Stack #4A

110 K. R. Wilson Hydraulic Press 136 Clear Well

111 Chicago Shear 137 Pressure Filters

112 Lift Truck Hoist 138 Detention Tank #2

113 Stack #4Afilter Building 139 Detention Tank #1

114 Caustic Tanks 140 Processed Tank #1

115 Startrite Band Saw 141 Processed Tank #2

116 Main Plant Acid, Stack #7 142 Processed Tank #3

117 Health-Safety Technician's Office 143 Sludge Pump #2

118 Plate and Frame Filter Press 144 Sludge Pump #1

119 Wastewater Treatment Building 145 Pressure Filter

120 Fire Hydrant 146 Backwash Pumps

109 South Gisholt Lathe 147 H. M. C. Furnace

110 K. R. Wilson Hydraulic Press 148 Lumber Rack

111 Chicago Shear 149 Tool Coating Vent and Stack

112 Lift Truck Hoist 150 Tool Grinding Booth

113 Stack #4Afilter Building 151 Holding Tank

114 Caustic Tanks 152 D. C. Exciters

115 Startrite Band Saw 153 Receiver

116 Main Plant Acid, Stack #7 154 Water Heater Room

117 Health-Safety Technician's Office 155 Project Storage Cage

118 Plate and Frame Filter Press 156 Maintenance Tool Storage

119 Wastewater Treatment Building 157 Incoming Water and Water Meter Pit

120 Fire Hydrant 158 Incoming Natural Gas and Gas Meter

121 Chip Chopper 159 Still Pumps

122 Thermal Oxidation Tank #1 160 Abrasive Saw

123 Thermal Oxidation Tank #2 161 Salt Pots

124 Baghouse 162 Sheldon Lathe

125 HEPA Filter Housing 163 Fisher Scientific Iso-Temp Lab Refrigerator

126 95% and HEPA Filter Housing 164 Ross Temp Icemaker

127 Cartridge Filter Housing 165 Cooley Furnace

128 Fan -- Stack #1A 166 Neytech Furnace

129 Cartridge Filter Housing 167 Walker-Turner Drill Press

130 95% and HEPA Filter Housing 168 Delta Band Saw

131 Fan -- Stack #3A 169 Pneumotive Air Compressor

132 Moisture Separator 170 RF -- 6 Sump

133 Cartridge Filter Housing
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 50 of 73

Figure A-2. Plant Layout – 1957.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 3 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 51 of 73

Figure A-3. Plant Layout – 1962.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 4 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 52 of 73

Figure A-4. Plant Layout – 1964.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 5 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 53 of 73

Figure A-5. Plant Layout – 1965.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 6 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 54 of 73

Figure A-6. Plant Layout – 1968.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 7 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 55 of 73

Figure A-7. Plant Layout – 1969.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 8 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 56 of 73

Figure A-8. Plant Layout – 1979.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 9 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 57 of 73

Figure A-9. Plant Layout – 1981.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 10 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 58 of 73

Figure A-10. Plant Layout – 1982.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 2 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 59 of 73

Figure A-11. Plant Layout – 1983.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 3 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 60 of 73

Figure A-12. Plant Layout – 1984.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 4 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 61 of 73

Figure A-13. Plant Layout – 1985.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 14 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 62 of 73

Figure A-14. Plant Layout – 1986.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 5 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 63 of 73

Figure A-15. Plant Layout – 1987.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 6 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 64 of 73

Figure A-16. Plant Layout – 1988.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 7 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 65 of 73

Figure A-17. Plant Layout – 1989.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 8 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 66 of 73

Figure A-18. Plant Layout – 1990.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 9 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 67 of 73

Figure A-19. Plant Layout – 1991.

ATTACHMENT A

EXTRUSION PLANT BUILDING LAYOUTS

Page 20 of 20
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 68 of 73

ATTACHMENT B 

EXTRUSION PLANT PROCESSES – N-REACTOR PRODUCTION

Page 1 of 2

According to RMI (1995a), ”the N-Reactor Production consisted of a primary extrusion process 

followed by a forging process.” The typical steps follow (RMI 1995a).

1. Ingots were received from FEMP.

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 hour minimum, 6 hour 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.

a. Until the 1960s, extrusions were packed, stored, and shipped back to FEMP for further 

processing.

b. In the late 1960s, further processing was performed at RMI as described below.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 69 of 73

ATTACHMENT B 

EXTRUSION PLANT PROCESSES – N-REACTOR PRODUCTION

Page 2 of 2

14. The extrusion was cut into sections (billets).

15. Billets were nitric acid pickled, inspected, machined, and reinspected prior to 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 postforging 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), after extrusion, hand rolling was performed for 5 

minutes. 
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 70 of 73

ATTACHMENT C

EXTRUSION PLANT PROCESSES – SAVANNAH RIVER SITE PRODUCTION

Page 1 of 2

The typical steps for uranium processing for the Savannah River reactors follow (RMI 1995a).

1. Ingots were received from FEMP.

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. Prior to this, extrusions were air-cooled.

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 prior to the roll straightener. This 

process continued for approximately 1 year. After this, extrusions were not heated prior to 

straightening.

13. Extrusions were stored on the process table.

14. Extrusions were lifted horizontally and

a. From 1962 to approximately 1964, were placed in a vapor degreaser tank to clean the 

extrusion.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 71 of 73

ATTACHMENT C

EXTRUSION PLANT PROCESSES – SAVANNAH RIVER SITE PRODUCTION

Page 2 of 2

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 FEMP.

16. Packaged extrusions were stored prior to 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 FEMP.
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 72 of 73

ATTACHMENT D

EXTRUSION PLANT PROCESSES – DOD PENETRATOR PRODUCTION

Page 1 of 1

The typical steps for source material processing for DOD Penetrators follow (RMI 1995a).

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

a. Extrusion press

7. Postextrusion 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

(1) Stored in main plant before 1984.

(2) Stored in main plant and northeast warehouse after 1984.

i. Shipment
Document No. ORAUT-TKBS-0056 Revision No. 00 Effective Date: 04/26/2007 Page 73 of 73

ATTACHMENT E

DETAILED PENETRATING WHOLE BODY DOSES

Page 1 of 1

Year

Number 

monitored

Whole body deep dose (rem) Assumed 

neutron dose 

<0.010 0.01–0.099 ICRP 60 (rem) Reference

0.100–

0.249

95th 

percentile

1999 1,202 1,170 32 <0.010 0 Eberline Dosimetry 

Services 2000

1998 –a _ _ _ _ 0

1997 290 271 19 0.099 0 ThermoNUtech 1998

1996 285 241 44 1 0.099 0 Tauche, R.M. 1998

1995 255 220 35 0.099 0 ThermoNUtech 1996

1994 330 291 38 1 0.099 0 TMA Eberline 1995

a. – = No data available.

Year

Number 

monitored

Whole body beta-gamma dose (rem) Assumed 

neutron dose 

None <0.100 ICRP 60 (rem) Reference

0.100–

0.249

0.250–

0.499

0.500–

0.749

95th 

percentile

1984 117 23 48 33 13 0.499 0.439 Schaeffer 1985

1983 118 32 39 32 15 0.499 0.439 Schaeffer 1984

1982 117 20 54 32 10 1 0.499 0.439 Schaeffer 1983

1981 –a _ _ _ _ _ _ 0.439

1980 80 16 32 28 4 0.249 0.439 Schaeffer 1981

1979 80 14 45 19 2 0.249 0.439 Van Looke 1980

1978 68 16 21 26 5 0.499 0.439 Schaeffer 1979

1977 _ _ _ _ _ _ 0.439 Schaeffer 1977

1976 58 30 20 8 0.249 0.439

1975 62 19 31 11 3 0.249 0.439 Heiser 1976

1974 69 22 27 17 1 0.249 0.439 Heiser 1975

a. – = No data available.