ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 124

DOE Review Release 04/07/2010

Document Title:

Site Profile for Battelle Memorial Institute, King 

Avenue and West Jefferson Sites, Columbus, 

Ohio

Document Number: ORAUT-TKBS-0058

Revision: 00

Effective Date: 03/26/2010

Type of Document TBD

Supersedes: None

Subject Expert(s): Vincent A. King, Jack Fix, and Sallie D. Robinson

Site Expert(s): N/A

Approval: Signature on File Approval Date: 03/24/2010

Vincent A. King, Document Owner

Concurrence: Signature on File Concurrence Date: 03/24/2010

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 03/24/2010

Edward F. Maher, Objective 3 Manager

Concurrence: James P. Griffin Signature on File for Concurrence Date: 03/25/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/26/2010

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-0058 Revision No. 00 Effective Date: 03/26/2010 Page 2 of 124

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/26/2010 00 New site profile for the Battelle Memorial Institute sites in Columbus, 

Ohio (King Avenue) and West Jefferson, Ohio (West Jefferson). 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Janet L. Westbrook and Vincent A. King.
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 3 of 124

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

6

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

2.0 Site Description......................................................................................................................11

2.1 Buildings of Concern and the Associated Radioactive Materials 

and Operations...........................................................................................................12

2.2 Access Controls..........................................................................................................13

3.0 Occupational Medical Dose.................................................................................................... 16

4.0 Occupational Internal Dose.................................................................................................... 18

4.1 In Vitro Bioassay.........................................................................................................18

4.1.1 Urinalysis ........................................................................................................ 18

4.1.2 Fecal Sample Analysis.................................................................................... 20

4.1.3 Nose Swabs.................................................................................................... 20

4.2 In Vivo Bioassay.........................................................................................................20

4.3 Lapel Samplers ..........................................................................................................21

4.4 In Vitro Bioassay Detection Limits ..............................................................................21

4.5 Inhalation Absorption Types, Particle Sizes, and Activity 

Fractions ....................................................................................................................24

4.6 Dose Reconstruction Considerations for Fission Products.......................................... 26

4.7 Records and Reports..................................................................................................27

4.8 Internal Dose Assessment for Unmonitored Period (before 

September, 1956).......................................................................................................27

5.0 Occupational External Dose................................................................................................... 28

5.1 Dosimetry processing services ...................................................................................28

5.2 Dosimeter types..........................................................................................................28

5.3 Dosimeter detection capabilities .................................................................................29

5.4 Dosimeter exchange frequencies ...............................................................................30

5.5 Workplace radiation fields...........................................................................................31

5.6 Assessment of Neutron Dose Monitored by NTA Film ................................................32

5.7 Incident reports...........................................................................................................33

5.8 External Dose Assessment for Unmonitored Period (before 

June, 1953) ................................................................................................................33

6.0 Residual Exposure.................................................................................................................34

7.0 Attributions and Annotations .................................................................................................. 35

References ........................................................................................................................................36

Glossary ............................................................................................................................................60

ATTACHMENT A, CHRONOLOGY OF EVENTS AT BATTELLE...................................................... 62

ATTACHMENT B, BUILDINGS AND OTHER FACILITIES AT BATTELLE ....................................... 67
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ATTACHMENT C, HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND 

MATERIAL ..........................................................................................................75

ATTACHMENT D, TYPES AND QUANTITIES OF MATERIAL PRODUCED OR 

USED IN BATTELLE WORK...............................................................................84

ATTACHMENT E, INCIDENTS REPORTED AS OCCURRING AT THE KING 

AVENUE AND WEST JEFFERSON SITES.........................................................90

ATTACHMENT F, RADIONUCLIDES AND CHEMICAL FORMS BY BUILDING .............................. 96

ATTACHMENT G, PHOTON AND ELECTRON ENERGY RANGES .............................................. 101

ATTACHMENT H, NEUTRON ENERGY RANGES......................................................................... 106

ATTACHMENT I, ANALYSIS OF BATTELLE MEASURED NEUTRON AND 

PHOTON DOSE RADIATION SURVEY DATA.................................................. 110

LIST OF TABLES

TABLE TITLE

3-1 Organ doses from chest fluoroscopy......................................................................................

PAGE

17

3-2 Default assumptions for X-ray procedure types and frequencies............................................ 17

4-1 Summary of in vitro bioassay information............................................................................... 19

4-2 Whole body counting analytical sensitivity.............................................................................. 21

4-3 Urinalysis detection limits data to be used in dose reconstruction .......................................... 22

4-4 Isotopic composition data for plutonium work ......................................................................... 25

4-5 Cooling times for fission products at Battelle.......................................................................... 26

5-1 Dosimetry processing services............................................................................................... 28

5-2 Summary of MDLs for dosimeter badges ............................................................................... 29

5-3 Default dosimeter exchange frequency by facility or work group ............................................ 31

5-4 Recommended neutron to photon ratios for dose monitored by NTA film............................... 33

I-1 Battelle reference files uploaded December 3, 2009 containing neutron and 

photon radiation survey data ................................................................................................ 113

I-2 Battelle reference files uploaded February 15, 2010 containing neutron and 

photon radiation survey data ................................................................................................ 113

I-3 SRDB reference files for Battelle information received from OSTI uploaded 

February 26, 2010................................................................................................................ 114

I-4 Mean neutron quality factor and flux corresponding to maximum dose 

equivalent of 1 mSv per 40-hour week ................................................................................. 114

I-5 Recommended lognormal parameters of NP ratios for Building JN-1................................... 116

I-6 Recommended lognormal parameters of NP ratios for Building JN-2................................... 118

I-7 Statistical parameters of JN-3 survey record analysis .......................................................... 120

I-8 Lognormal parameters of NP ratios for Building JN-4........................................................... 122
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LIST OF FIGURES

FIGURE TITLE

2-1 Relative locations of the King Avenue and West Jefferson sites.............................................

PAGE

12

2-2 Plan view of the King Avenue site .......................................................................................... 13

2-3 Plan view of the West Jefferson site....................................................................................... 14

I-1 JN-1 hot cell laboratory NP ratio during 252Cf operations ...................................................... 115

I-2 JN-2 plutonium laboratory NP ratio ...................................................................................... 117

I-3 JN-3 paired neutron and photon dose rate measurements................................................... 119

I-4 JN-3 lognormal probability plot of the measured NP ratio..................................................... 119

I-5 JN-4 historical trend in NP ratio............................................................................................ 121

I-6 JN-4 plutonium laboratory NP ratio ...................................................................................... 121

I-7 JN-4 paired neutron and photon dose rate measurements................................................... 122
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ACRONYMS AND ABBREVIATIONS

AARR Argonne Advanced Research Reactor

AEC U.S. Atomic Energy Commission

AGC Alpha-Gamma Cell

AGCRSP Army Gas-Cooled Reactor Systems Program

AGN Aerojet General Nucleonics

Al aluminum

AI Atomics International (Co)

ANL Argonne National Laboratory

AP anterior-posterior

APPR Army Package Power Reactor 

ARB Army Reactor Branch

ATL-A Advanced Technology Labs (American Standard Corp)

ATR Advanced Test Reactor (INEL)

AWE atomic weapons employer

BCD Battelle Columbus Division

BCL Battelle Columbus Laboratories (i.e., Battelle Memorial Institute)

BCLDP Battelle Columbus Laboratories Decommissioning Project

BMI Battelle Memorial Institute

BRR Battelle Research Reactor

BWR Boiling-water reactor

BZ Breathing zone

CA, CAA Controlled Access Area (at JN-1); the Contaminated Control Area was apparently a 

subsection of the Controlled Access Area

CAL Critical Assembly Facility, i.e., JN-2 in 1957-1963

CAM continuous air monitor

CATI computer-assisted telephone interview

CEMP Columbus Environmental Management Project

CFR Code of Federal Regulations

Ci curie

CEMP Columbus Environmental Management Project

CETR Consolidated Edison Thorium Reactor

Con Ed Consolidated Edison (Co)

CP-6 Same as AARR

CPP Chemical Processing Plant (INEL)

CXR chest X-ray

d day

DCF dose correction factor

DOD U.S. Department of Defense

DOE U.S. Department of Energy

dpm disintegrations per minute

E of C Evidence of contamination (used in badge records)

EDTA Ethylenediaminetetraacetic acid (used experimentally)

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ESD entrance skin dose

ESR Engineering Storage Room in JN-1

ETR Engineering Testing Reactor (INEL)

EU enriched uranium
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FEMP Fernald Environmental Management Project (Fernald site)

ft foot

g gram

GA General Atomics (company)

GCRE Gas-Cooled Reactor Experiment

HAPO Hanford Atomic Power Operations

HC hot cell

HEC High Energy Cell (JN-1)

HEDL Hanford Engineering Development Laboratory

HERLA High-Energy Research Laboratory Area

HEU highly enriched uranium (to 93% or more)

HIP Hot Isostatic Processing (facility)

HLC High Level Cell (JN-1)

HP Health physicist

HTGR High Temperature Gas[-Cooled] Reactor

hr hour

HVL half-value layer

ICRP International Commission on Radiological Protection

in. inch

INEL, INL Idaho National Engineering Laboratory

JN Prefix for some West Jefferson buildings (north part of the site)

JS Prefix for some West Jefferson buildings (south part of the site)

KA King Avenue (site)

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

KPA kinetic phosphorescence analysis

L liter

LAMPF Los Alamos Meson Physics Facility

LAMPRE Los Alamos Molten Plutonium Reactor Experiment

LAT lateral

lb pound

LLC Low Level Cell (JN-1)

LLNL Lawrence Livermore National Laboratory

mCi millicurie

MDL minimum detection level

MED Manhattan Engineer District

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission products

mg milligram

MGCR Maritime Gas-Cooled Reactor

MHW Multihundred watt - a type of plutonium (Pu-238) fuel for RTGs

millirem millirem

mL milliliter

ML-1 Mobile Low-power (reactor) (Army program)

MPBB Maximum permissible body burden

mR milliroentgen
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mrad millirad

MTC Mechanical Test Cell (in JN-1)

MW megawatt

NAC Nuclear Assurance Corporation

NASA National Aeronautics and Space Administration

NCRP National Council on Radiation Protection and Measurement

NIOSH National Institute for Occupational Safety and Health

NLO National Lead Company of Ohio (Fernald)

NMSR Nuclear Merchant Ship Reactor

NPR Nonpower reactor or neutron production reactor

NRC U.S. Nuclear Regulatory Commission

NRTS Same as ETR facility

NSEC Nuclear Science and Engineering Corporation

NTA nuclear track emulsion, type A

NYOO New York Operations Office

OBD Old Back Dock (JN-1)

OID AEC’s Office of Isotopes Development

OMRE, OME Organic-Moderated Reactor Experiment, aka Organic Moderated Solvent Burning 

Experiment (INEL)

ONWI Battelle's Office of Nuclear Waste Isolation

OOA Old Operating Area (JN-1)

ORNL Oak Ridge National Laboratory

PA posterior-anterior

PRTR Plutonium Recycle Test Reactor (Hanford)

PWR Pressurized-water reactor

PyC Pyrolitic carbon

R&D research and development

RAL Radioanalytical Laboratory, which was in the former plutonium lab in JN-2

RCRA Resource Conservation and Recovery Act

RD, AEC-RD, 

AEC-DRD AEC's Research Division

RFI Rocky Flats

RR [Battelle] Research reactor

RTG Radioisotope thermoelectric generator

s second

S&M Surveillance and maintenance (before decontamination)

SAP Sintered aluminum powder

SME Salt-Moderated Experiment (reactor)

SNAP Systems for Auxiliary Nuclear Power (reactor series for space, etc.)

SNM Special Nuclear Material

SRDB Ref ID Site Research Database reference identification number

SS stainless steel (as a material specification); Super S (as a biosolubility class)

T Uranium (e.g., U metal was also called T metal)

TLD thermoluminescent dosimeter

TNT Thorium nitrate tetrahydrate

TREAT Transient Reactor Test Facility (INEL)

TRU transuranic
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U.S.C. United States Code

μCi microcurie

UCRL University of California (Lawrence) Radiation Laboratory

μm micrometer (0.001 meter)

UNH Uranyl nitrate hexahydrate, i.e., UO2(NO3)2 ∙ 6 H2O

μg microgram

VMR Variable Moderator Reactor

w/o weight percent (when used in describing, e.g., alloy forms)

WJ West Jefferson (site)

wk week

WTR Westinghouse Test Reactor (Pennsylvania)

X Uranium (e.g., U metal was also called X metal)

XP-pad Staging area/pad; Not a radiological area, but had RCRA closure significance

ZPR The critical assembly facility (JN-2)

§ section or sections
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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 at particular sites or categories of sites. They will be revised in the event 

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

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

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

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

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

contract period and/or during the residual period.

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

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

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

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

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

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

includes radiation exposure related to the Naval Nuclear Propulsion Program and any radiation 

exposure received from the production of commercial radioactive products that were concurrently 

manufactured by the AWE facility during the covered period. NIOSH does not consider the following 

exposures to be occupationally derived (NIOSH 2007a):

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

conventional structures

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

reasons

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

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

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

period (NIOSH 2007a). It should be noted that under subparagraph A of 42 U.S.C. § 7384n(c)(4), 

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

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

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

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

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

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

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

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

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

The Department of Energy Worker Advocacy Web site (December 26, 2006) gives the following 

information:

SITE DESCRIPTION

Battelle Laboratories – King Avenue

Also known as: Battelle Columbus Laboratories (BCL); Battelle Memorial 

Institute (BMI)

State: Ohio

Location: Columbus

Time period: AWE 1943–1986; DOE 1986–2000

Facility type: Atomic Weapons Employer, Department of Energy

Battelle Laboratories – West Jefferson

Also known as: Battelle Columbus Laboratories (BCL); Battelle Memorial 

Institute (BMI); West Jefferson Plutonium Facilities

State: Ohio

Location: Columbus [West Jefferson]

Time period: AWE 1956–1975; Residual Radiation 1976-–1985; DOE 1986–present 

(remediation)

Facility type: Atomic Weapons Employer, Department of Energy

Facility Description: From 1943 to 1986, BMI performed atomic energy research and 

development [R&D] for DOE and its predecessor agencies. The Battelle Laboratories 

have two separate locations in Columbus: King Avenue and West Jefferson. Battelle's 

research supported the government's fuel and target fabrication program, including 

fabrication of uranium and fuel elements, reactor development, submarine propulsion, 

fuel reprocessing, and the safe e of reactor vessels and piping. The following activities 

were performed at the King Avenue site: processing and machining enriched, natural, 

and depleted uranium and thorium; fabricating fuel elements; analyzing 

radiochemicals; and studying power metallurgy. The following activities were 

performed at the West Jefferson site: operation of a large hot cell facility and a 

research reactor. Reactor operations began in October 1956, and ended in December 

1974. The reactor was defueled and partially dismantled in 1975 and Battelle's license 

was changed to possession-only status. 

In this site summary document, BMI and BCL (the Columbus-based division of BMI operations) are 

referred to as Battelle. Both the Battelle King Avenue and West Jefferson sites are included in this 

site summary. When it is necessary to distinguish between the two sites, they will be referred to as 

King Avenue and West Jefferson, respectively.

Battelle was a research institution and undertook a wide variety of radiologically significant projects as 

part of its Manhattan Engineer District (MED)/AEC/Energy Research and Development Administration 
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(ERDA)/DOE work and its defense and commercial work. These projects ranged from low-level 

biological and industrial radiotracer work to high-level work involving alpha gloveboxes and hot cells.

The relative locations of the King Avenue and West Jefferson sites are shown in Figure 2-1 (BCL 

1974). Individual plan views of the King Avenue and West Jefferson sites are shown in Figures 2-2

and 2-3, respectively (Jenson 2003). Plan views are also available for the Melting Facility at Building 

3 (Kirsch 2000a); the Building 4 Radioisotope Laboratory and various subareas (Sunderman and 

Dickerson 1962; Brown 1966); the Hot Cell Laboratory (JN-1) and subareas (Wastren 2001; 

Sunderman and Dickerson 1962; Dickerson 1956; Brown 1966); the Critical Assembly Laboratory 

(JN-2) (Hogan et al. 1958; Jankowski and Chastain 1958; Kirsch 1987); the Battelle Research 

Reactor (BRR, JN-3) (Chastain et al. 1955; Chastain 1957; Anno and Plummer 1962); the Plutonium 

Laboratory (JN-4) (Flynn et al. 1987; Freas et al. 1971; Rudolph, Kirsch, and Toy 1984); the King 

Avenue site (Sunderman and Dickerson 1962; BMI 1977); and the West Jefferson site (Evans and 

Woodward 1979; BCL 2002).

Figure 2-1. Relative locations of the King Avenue and West Jefferson sites.

2.1 BUILDINGS OF CONCERN AND THE ASSOCIATED RADIOACTIVE MATERIALS AND 

OPERATIONS

Attachment A provides an overview chronology of significant events at Battelle, including the start and 

end of operations. Attachment B is a list of buildings, their major subareas and uses, and the dates of 

their use. Attachment E lists the incidents that were reported at the site. Most of these did not result 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 13 of 124

Figure 2-2. Plan view of the King Avenue site.

in a significant internal or external exposure to any worker, but at times Battelle had to do additional 

monitoring of one or more workers to determine this. The most notable incidents from the point of 

view of potential or actual exposure are those of August 8, 1956, February 10, 1959, August 9, 1961, 

and July 2, 1980, for external exposure and February 10, 1959, May 9, 1974, January 16, 1979, May 

2, 1980, February 27, 1984, and May 1984 for internal exposure.

The abbreviations and definitions should be used as an aid in the interpretation of dosimetry records 

and other relevant documents.

2.2 ACCESS CONTROLS

At the Radioisotope Laboratory in Building 4, the doors from the corridor into the medium- and highlevel laboratories were locked from the outside at all times. Access was otherwise through the locker 

room, which itself was entered through the office area (Sunderman and Dickerson 1962). These 

doors were used only for transporting casks and heavy equipment in and out of the laboratory. 

Permission of the laboratory supervisor was required before the doors could be used.
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Figure 2-3. Plan view of the West Jefferson site.

At the Gamma (60Co irradiation) Facility in Building 6, permission for non-staff members to enter the 

facility had to be obtained from the facility staff (Sunderman and Dickerson 1962). This facility was 

kept locked overnight, on holidays, and on weekends to prevent unauthorized personnel from 

entering.

The Hot Cell Laboratory (JN-1) began operations in 1955 (Wastren 2001). Access was controlled at 

all times and, in compliance with 10 CFR Part 20, the entire building was a restricted area 

(Sunderman and Dickerson 1962; Sunderman and Gates 1965). The locker rooms were divided into 

two areas separated by walk-through showers; one area was used by personnel not in contact with

radioactive material while the other was the dressing area for entry into the contaminated areas 

(Sunderman and Dickerson 1962). Personnel access into the operating and contamination areas was 

normally through the hot locker room, but under special conditions, entry might also be made through 

locked doors directly into the operating area or through an equipment storage area (Sunderman and 

Dickerson 1962; Myers et al. 1994a). Access to the cells was only through locked doors at the rear of 

the cells, reached only through the Controlled Access Area (Sunderman and Dickerson 1962; Myers 

et al. 1994a). Access to the Alpha-Gamma Cell (AGC) operating area and receiving dock area was 

restricted when an AGC dry box was being moved through these areas to the repair station (Gates 

1964). 

At the Critical Assembly Laboratory (JN-2), before operation each day, interlocks were checked, and 

there was a visual check of the facility and the critical assembly room (Hogan et al. 1958, Jankowski 
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et al. 1957). The outer doors and doors between the control room and the critical assembly room 

were required to be locked during operation of the critical assembly; signs were posted before

operation (Hogan et al. 1958). Only personnel engaged in activities directly connected with the 

current operation were allowed to be present in the control room during an experiment. There was a 

fence around the building to exclude personnel and trucks when necessary (Hogan et al. 1958). 

At the Battelle Research Reactor (BRR or JN-3), color-coded badges were used to denote levels of 

access, depending on the level of training; the top level was a free access badge (Hogan et al. 1958, 

Plummer, Anno, and Chastain 1960). Visitors from outside Battelle, transient experimenters, 

maintenance staff, and Battelle non-BRR staff visitors were escorted by reactor staff or by Public 

Relations Office staff. In general, visitors were permitted only in areas free from experimental activity 

(Anno and Plummer 1962); otherwise, permission had to be obtained from the Operating Supervisor 

and personnel monitoring devices such as visitor badges were provided (Anno and Plummer 1962, 

Plummer, Anno, and Chastain 1960).

At the Plutonium Laboratory (JN-4), all outside doors and the doors between the laboratory area 

proper and the office area were kept locked from the inside (BCL 1977a).
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3.0

In 1943, a MED officer reported that no physical or laboratory medical examinations of Battelle 

employees had been done because of their low and intermittent exposures to radioactivity (Ferry 

1943). It is therefore likely that there were no required medical X-ray examinations done in the early 

years. However, in the absence of individual information, a preemployment examination should be 

assumed. A posterior-anterior (PA) chest X-ray examination should be assumed from the beginning 

of site operations through 1951 (at least one preemployment PA chest X-ray was recorded as early as 

1946). Based on a review of site records for Battelle, single chest fluoroscopy examinations were

performed as part of a preemployment examination from 1952 through 1965. After 1965, radiographic 

PA chest X-rays were performed for preemployment examinations. 

OCCUPATIONAL MEDICAL DOSE

In addition to the preemployment examination, annual PA chest X-ray examinations were performed 

for subsequent years of employment during all periods of operations. The records also indicate that a

single additional annual lateral (LAT) chest X-ray examination was performed in addition to the annual 

PA radiographic chest examination (Wright 2002a, b) for the period 1975 through 1980. X-ray 

examinations were occasionally identified as “teleo chest” X-rays, which were performed to evaluate 

heart size. PA chest X-ray doses should be assigned for these examinations because the same 

source-to-image distance and X-ray equipment would have been used for these examinations.

Table 3-1 lists the organ and skin doses for chest fluoroscopy. While ORAUT-OTIB-0006 (ORAUT 

2005a) provides some default assumptions for chest fluoroscopy, new information has been found in 

the modern medical literature for this now obsolete procedure (Marshall 2001). The Battelle claim 

records indicate that the chest fluoroscopy examinations were performed in a private physician’s 

office. Given the era and the private office setting, the fluoroscopic equipment was most likely to have 

been "direct" or non-image intensified fluoroscopy equipment of the same type described by Marshall 

(2001). Entrance air kerma rates for this type of equipment are not commonly found in the literature. 

However, a study of the entrance air kerma rates, half-value layers (HVLs), and estimated fluoroscopy 

exposure times was conducted in Albania and described by Marshall (2001). In this study, the 

exposure time for chest fluoroscopy screening was estimated to be about 20 seconds, with an 

average entrance air kerma rate of about 4 R/min, and an average HVL of about 2.3 mm Al 

equivalent. Therefore, chest fluoroscopy doses for Battelle workers are calculated using the following 

assumptions that are favorable to claimants and based on the Marshall (2001) data: an entrance 

exposure rate of 5 R/min for 30 seconds of fluoroscopy time and International Commission on 

Radiological Protection (ICRP) Publication 34 (ICRP 1982) dose conversion factors (DCFs) for a PA

chest for 2.0 mm Al HVL. For chest fluoroscopy, the field is assumed to include the thoracic organs 

and the liver/gall bladder/spleen, but not the thyroid, gonads, bladder, or colon/rectum, because of the 

short source-to-skin distance (SSD) (approximately 40 cm). However, the beam size was estimated 

to be similar to the beam for a poorly collimated radiographic chest X-ray to account for the fact that 

the field in chest fluoroscopy is dynamic and could expose a larger area of the body than the short 

SSD would initially imply. Doses from chest fluoroscopy examinations in Table 3-1 should be 

assigned as the geometric mean values of a lognormal distribution with a geometric standard 

deviation of 2 to account for the uncertainty in the procedures.

ORAUT-OTIB-0006 (ORAUT 2005a) provides organ doses for PA and LAT radiographic chest 

examinations. 

If individual X-ray records are available, the examination type and frequency should be assigned as 

identified in the records. In the absence of site or individual information, the examination types and 

frequencies in Table 3-2 should be assumed as default values.
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Table 3-1. Organ doses from chest fluoroscopy.

PA chest fluoroscopya,b

DCF mGy/Gy × 1E-3rem-Gy/cGy-mGy

(HVL = 2.0 mm Al eq)

Entrance 

kerma (cGy)

Organ 

dose (rem)

Thyroid 1.50E-01c 2.50E+00 3.75E-01

Eye brain 2.10E-02 2.50E+00 5.25E-02

Ovaries 6.00E-04 2.50E+00 1.50E-03

Liver/gall bladder/spleen 3.55E-01 2.50E+00 8.88E-01

Urinary bladder 6.00E-04 2.50E+00 1.50E-03

Colon rectum 6.00E-04 2.50E+00 1.50E-03

Testes 1.00E-05 2.50E+00 2.50E-05

Lungs 3.55E -01 2.50E+00 8.88E-01

Thymus 3.55E -01 2.50E+00 8.88E-01

Esophagus 3.55E -01 2.50E+00 8.88E-01

Stomach 3.55E -01 2.50E+00 8.88E-01

Bone surfaces 3.55E -01 2.50E+00 8.88E-01

Remainder 3.55E -01 2.50E+00 8.88E-01

Breast 3.20E -02 2.50E+00 8.00E-02

Uterus 7.00E-04 2.50E+00 1.75E-03

Bone marrow 6.90E-02 2.50E+00 1.73E-01

Entrance skin dose (ESD) applies to 

skin of: back and sides of chest to iliac 

crest, back of shoulders, back and 

sides of neck, upper arms, elbows, 

forearms, and palms and backs of 

hands 3.30E+00d

Eye/brain dose applies to skin of face 5.25E-02

Exit skin dose applies to the skin of: 

front of chest, front of shoulders, front 

of upper abdomen (to iliac crest), front 

of neck 6.32E-02e

Outside but near entrance beam dose 

applies to skin of back and sides of 

head, ears, and buttocks. 3.30E-01f

Outside but near exit beam dose 

applies to the skin of front lower 

abdomen from iliac crest to pubis 6.32E-03g

Skin of thighs to knees 8.20E-04

Skin of knees and below 3.00E-04

a. The assumed entrance exposure rate is 5 R/min for 30 seconds.

b. For chest fluoroscopy, the field is estimated to include thoracic organs and the liver/gall bladder/spleen, but not 

the gonads, bladder, or colorectal organs.

c. Used DCF for AP Cervical spine and depth dose correction factor of 0.2 (ORAUT 2005a).

d. ESD is the entrance kerma rate multiplied by a backscatter factor of 1.32 from Table B-8 of National Council on 

Radiation Protection and Measurement (NCRP) Report 102 (NCRP 1989).

e. Exit skin dose is the ESD divided by the absorption factor for a 24-cm chest and 2.0 mm Al HVL, divided by 0.9 to 

account for the uncertainty in the tabulated absorption factors in Table B-7 of NCRP Report 102 (NCRP 1989). 

f. Outside but near entrance beam dose is 10% of the ESD.

g. Outside but near exit beam dose is 10% of the exit skin dose.

Table 3-2. Default assumptions for X-ray procedure types and frequencies.

Period

Preemployment 

PA chest

Preemployment 

fluoroscopic 

chest

Annual PA 

chest LAT chest

Beginning of operations–1951 X X

1952–1965 X X

1966–1974 X X

1975–1980 X X X

1981–present X X
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 18 of 124

4.0

4.1 IN VITRO BIOASSAY

OCCUPATIONAL INTERNAL DOSE

4.1.1

Routine urine samples appear nearly always to have been 24-hour samples, although the bioassay 

records note samples that represent longer or shorter correction periods, for which the results were 

corrected to 24-hour equivalents. A few spot samples were taken, probably as a check or as the 

result of an incident; these appear to be identified as such in the records. In the absence of specific 

recorded information, a urine sample should be considered to represent a 24-hour excretion period. 

Urinalysis

Table 4-1 is a summary of the primary types of routine in vitro bioassay samples and the associated 

sampling periods as observed in claimant analytical reports provided by DOE to the NIOSH project. 

Sampling frequencies are typically monthly, quarterly, or semiannually, depending on the activity in 

which the worker is involved, or driven by follow-up activities to specific incidents. The earliest 

urinalysis mention in records found by the NIOSH Project is of samples submitted in about September 

1956 to a toxicology consultant; these samples were assayed only for uranium and the results were 

given in milligrams. It is unclear who performed these initial urine analyses. In about September 

1957, Battelle began to send the samples to Nuclear Science and Engineering Corporation (NSEC); 

these samples were analyzed for either uranium or gross beta (Edelmann 1958, 1959a) through 

March 18, 1960, then for uranium or gross beta, as requested. By at least September 1959, NSEC 

was analyzing for plutonium (Edelmann 1959b) and by at least February 1962 for uranium, plutonium, 

gross beta, and tritium (Sunderman and Dickerson 1962). Only one type of analysis appears to have 

been done per sample; that is, aliquots were apparently not drawn off the same sample for different 

analyses by NSEC, although this might have been done by Battelle. 

From at least as early as August 1963 to early 1995 (based on inspections of individual records 

obtained by the Project; Sunderson and Gates 1965), the samples were analyzed by the Eberline 

Company; after this, samples were analyzed by Thermo Nutech, of which Eberline was the 

predecessor company. In the first few years, these samples were analyzed for fission products, 

plutonium, polonium, tritium, or uranium (Geiger 1963, 1964a, 1965); from inspection of urinalysis 

reports, the fission products appear to have been analyzed as gross beta. Later, they were analyzed 

for some or all of 3

H, 14C, 32P, 40K, 51Cr, 60Co, 90Sr, 134Cs, 137Cs, 144Ca, 125I, 131I, europium, 234U, 235U, 

238U, 238Pu, 239Pu, 241Am, and/or mixed fission products. Some discussion of recovery for 59Fe, 51Cr, 

and 63Ni is given in Geiger (1964b), indicating these radionuclides may also have been evaluated. 

Finally, a research scientist (probably a chemist) had a one-time 203Hg urinalysis in 1971.

From 1957 on, the samples were collected over a nominal 24 hours (Geiger 1963), although it was 

often noted on records if the sample was actually over a longer period. A few spot samples were 

taken, probably as a check or as the result of an incident; these appear to be identified as such in the 

records. 

It is not known on what general basis workers were selected for bioassay. Based on the dosimetry

records available to the NIOSH Dose Reconstruction Project, on claimant statements, and on Battelle 

site documents, it appears that at King Avenue, workers in at least the following categories were 

included in the urinalysis program:

• Workers in the Building 4 Radioisotope Laboratory complex and in the radiotracer laboratories 

in Buildings A and 1, including office personnel in these facilities
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 19 of 124

Table 4-1. Summary of in vitro bioassay information.

Monitoring type Radionuclide Period

Urine, 24-hour Gross beta 1957–1963

Urine, 24-hour Uranium (total), RUa 1957–1998

Urine, 24-hour U-234, U235, U-238 (SPU)b 1970–1994

Urine, 24-hour Plutonium–total 1961–1998

Urine, 24-hour Pu–isotopic (

238Pu, 239Pu) 1964–1993

Urine, 24-hour H-3 1962–1965

Urine, 24-hour Fission products/MFP 1961–1998

Urine, 24-hour Thorium–isotopic 1981–1994

Urine, 24-hour Thorium–total 1993–1998

Fecal Pu-239, FP, K-40, Ce-141-Ce/Pr-144, Ru-103-Ru/Rh-106, 

Z-95-Nb-95, U-234, U-235, U-38, Am-241, Sr-90

1975–1983

In addition to the information above, Eu, Am-241, Sr-90, Ce-144, Cs-134, Cs-137, Co-60, P-32, I-131, 

Cr-51, and I-125 were all assayed for in urinalyses at different times, as could be seen in the group 

reports from Eberline.

a. When RU is used in the bioassay records, it indicates "radiometric uranium", not recycled uranium.

b. When "SPU" is used in the records, it indicates the three U isotopes 234, 235, and 238, not Pu.

• Workers in areas of Buildings 2, 3, and 5 where uranium and other radioactive materials were 

handled

• Workers in Building 6 laboratories (except the Gamma Laboratory), but possibly only for 

specific projects

It appears that at West Jefferson, workers in at least the following categories were included:

• All workers at JN-1 during all years of operation and cleanup (about 1956 on)

• All workers at JN-2 during its use as a branch plutonium source testing laboratory (1964?-

1970) and possibly during its use as a critical facility (1956-1963)

• All workers at JN-3 during all years of operation and cleanup (1956 on)

• All workers at JN-4 during all years of operation and cleanup (1960 on)

More details are given below. Note that workers in other buildings and facilities than those given 

above or discussed below may also have been included as necessary based on their job functions.

All of the Radioisotopes Laboratory (Building 4 and the two tracer laboratories in Buildings A and 1) 

and JN-1 personnel working with radioactive materials participated in the urinalysis program routinely; 

secretaries and administrative personnel at these facilities were included (Sunderman and Dickerson 

1962). Staff members at JN-4 also submitted urine samples. Most of the workers in nonclerical and 

nonmanagement positions (above the level of supervisors), i.e., who likely were engaged in 

radiological work at least intermittently, were included in urinalysis program. There are urinalysis 

records for a metallurgical technician starting in 1957; this technician is known to have worked with 

uranium in Building 5. Similarly, from dosimetry records and incident reports, technicians in other 

areas of King Avenue and West Jefferson also appear to have had regular bioassay and also special 

bioassays or special attention to regular bioassays when there were incidents.

Staff members at JN-1 submitted urine samples quarterly, while personnel at the Radioisotope 

Laboratory submitted samples semiannually and secretaries and administrative personnel at both 

facilities submitted samples annually (Sunderman and Dickerson 1962). Staff members at JN-4 

submitted samples quarterly and additionally as needed, e.g., in case of an incident (BCL 1977a). In 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 20 of 124

individual dosimetry files accessible to the NIOSH Project, there are multiple notices sent to workers 

regarding their failure to submit urine samples, which are always termed "quarterly" samples. Some 

of the tabulations on the company cards have the notation "To be submitted [every] 6 months" and on 

others the notation "3 months."

A set of urine samples analyzed in 1974 were from workers involved in a May 9, 1974, incident in 

which plutonium was the major airborne contaminant (Geiger 1974); the bioassay records show that 

urine and fecal samples were submitted approximately weekly for some months and that the urine 

samples were analyzed initially for 239Pu, mixed fission products, and 90Sr. Later samples were 

analyzed for 239Pu and fission products. A summary of the bioassay data for the most exposed 

worker is shown in Toy (1975).

4.1.2

The earliest records of fecal samples date from May 1974 and go through at least March 1982, as 

shown in Table 4-1. Fecal sample analyses were all performed by Eberline and measured 40K, 90Sr, 

95Zr95Nb, 103Ru106Rh, 141Ce144Pr, 234U, 235U, 238U, 239Pu, 241Am, and/or mixed fission products, 

depending on the year and the case. The set of samples analyzed in 1974 were from workers 

involved in a May 9, 1974, incident in which plutonium was the major airborne contaminant (Geiger 

1974); the bioassay records show that urine and fecal samples were submitted approximately weekly 

for some months and that the initial fecal samples were assayed for 239Pu, mixed fission products, and 

some specific fission products, including 141Ce144Pr, 103Ru106Ru106Rh, and 95Zr95Nb. Later samples 

were analyzed for 239Pu and fission products.

Fecal Sample Analysis

Fecal samples were taken as appropriate after other incidents (e.g., Langendorfer 1977; Kirsch 

1978a). 

4.1.3

Nose swabs were taken after suspected intakes, as shown in several incident records (e.g., Selander 

1959a; Saling 1963; Kirsch 1978a). But it appears that it was routine or at least advisable to take a 

swab upon exiting from dusty cleanup work in JN-4 even in the absence of a known high airborne 

level, as indicated by an incident report showing that the workers took nose swabs on their own 

initiative (Hicks 1977). Although information from nose swabs typically cannot quantify intake 

amounts for dose reconstruction purposes, it may serve to verify whether or not a positive intake 

occurred after an incident.

Nose Swabs

4.2 IN VIVO BIOASSAY

In vivo whole-body counting is assumed to have begun in August of 1970, because there were no 

records found before that date, and to continue through the most recent operations (latest analysis 

record found was from 1998). In vivo counts performed by Helgeson Scientific included detailed 

documentation of methods, detection levels, counting parameters, and other related information. 

Radionuclides detected by whole body counting varied with each analysis performed, but typically 

included 40K (naturally occurring), 54Mn, 58Co, 60Co, 110Ag, 131I, 134Cs, and 137Cs. Batch whole body 

counting was performed for a group of workers at a frequency of twice per year (generally in March 

and again in October ), although not all workers may have participated at this frequency. Before

1995, minimum sensitivities were recorded for each semi-annual batch analysis (often designated as 

“Appendix III” of the analytical report provided by Helgeson) and are typically included with the 

individual worker records. If a worker’s record indicates that a whole body count was performed but 

there is no associated batch record of the minimum sensitivity for a radionuclide, the highest reported 

value found in the records is listed in Table 4-2 for each occupationally associated radionuclide and 

may be used as a favorable to claimant assumption for the analysis. 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 21 of 124

Table 4-2. Whole body counting analytical sensitivity.

Radionuclide Minimum Detectable Amount (nCi)

54Mn 4.4

58Co 3.0

60Co 2.0

110Ag 3.1

131I 3.6

134Cs 3.0

137Cs 3.6

Beginning in 1995, details of the whole-body counts (detectors, minimum detection level for each 

radioisotope, etc.) are documented in the individual worker analysis reports provided by the analytical 

laboratory rather than for a batch of workers. 

Although technical documentation included in Helgeson analytical reports described lung counting 

protocols, no evidence has been located in the records indicating that lung counts were performed at 

Battelle.

4.3 LAPEL SAMPLERS

Lapel samplers were used from at least 1975 on (e.g., Hicks 1977; Kirsch 1978a,b; Langendorfer 

1977; BMI 1981).

At JN-4, lapel filters associated with routine operations were removed and analyzed daily, with results 

available before the start of work the next day (BCL 1977a). A new filter was used at the start of any 

operation with a high potential for exposure (e.g., glove changes, bagging in or out, breaches of 

primary containment) and the filter was counted at the end of the operation to associate the exposure 

with the particular operation (BCL 1977a). Lapel samplers were also used as needed at the High 

Energy Cell at JN-1 (i.e., spent fuel), based on data from counting a worker's personal air sampler 

filters from April to August 1981 given in BMI (1981). Although information from bioassay sampling 

should be used rather than from lapel sampling to quantify intake amounts for dose reconstruction 

purposes, lapel sampling information may serve to verify whether or not a positive intake occurred 

after an incident.

4.4 IN VITRO BIOASSAY DETECTION LIMITS

Table 4-3 is a table of the detection limits associated with urinalysis. It should be noted that these 

values are generally derived from analytical records for Battelle workers provided by DOE to the 

NIOSH project; no documentation could be found that specified contractual minimum detectable 

activities (MDAs) in effect for the analytical laboratory, except for the information provided by Waligora 

(2007). The detection limits discussed in this section, other than those provided in Waligora (2007), 

were derived from stated “less than” values as found in the worker records, or from other methods as 

described below. 

Detection limits for urinalysis of a number of individually analyzed radionuclides found in the records 

(

144Ce, 134Cs, 137Cs, 60Co, 131I, 203Hg, and 90Sr) are not listed in Table 4-3 because non-positive results 

for these samples typically include a reported MDA or “less than” value. This is also true of fecal 

sample analyses and in vivo analyses. However, for fecal samples, if the report does not include this 

information, a detection limit of 0.1 pCi/g (ash weight) should be used for routine samples, with an 

optimum value of 0.03 pCi/g for special samples (Waligora 2007).

Detection limits for urinalysis were not routinely included in analytical reports before the 1990s, so the 

information below was derived from records available to NIOSH and information obtained from 

Eberline (Waligora 2007). The limits in Table 4-3 should be used when no information is available in 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 22 of 124

Table 4-3. Urinalysis detection limits data to be used in dose reconstruction.

Radionuclide Isotope Start End

Routine 

detection limit Units

Uranium Total uranium 

(fluorometric analysis

through 1995, KPA 

beginning 1996)

1957 1995 5 μg/L

1996 1998 1 μg/L

U-234, U-235, U-238 1970 1991 0.3 dpm/sample

1992 1994 0.1 pCi/L

Plutonium Total plutonium 1961 1991 1 dpm/24 hrs

1992 1998 0.1 pCi/L

Pu-238, Pu-239 1964 1991 0.2 Dpm/sample or dpm/24 hours

1992 1993 0.06 pCi/L

Tritium 1962 1965 1 μCi/L

Americium Am-241 1974 1980 0.03 dpm/sample

Thorium Th-228, Th-230, 

Th-232

1981 1991 2 dpm/sample

1992 1994 0.2 pCi/L

Total thorium 1993 1998 0.5 pCi/L

Fission products 1957 1992 0.2 dpm/mL

1993 1998 30 pCi/L

the claim records and do not take precedence over actual MDAs found in claim records. Additionally, 

the start and end dates should not be considered absolute if they conflict with information found in the 

claim records. 

Detection limits lower than the routine value shown in Table 4-3 could be achieved for special 

sampling situations by changing analytical count times, sample aliquot volumes, and other parameters

(Waligora 2007). Consequently, analytical results lower than the values listed above which are 

reported as positive results should be considered valid.

Detection limits for the radionuclides listed in Table 4-3 are discussed further below.

Uranium

Fluorometric analyses for total uranium were performed from the beginning of site operations through 

1995. In 1996, kinetic phosphorescence analysis (KPA) became the analytical method to determine 

total uranium in urine. Early sampling results for uranium were reported in units of milligrams per liter

(1956–1964) or micrograms per liter (1964–1998) for total uranium mass. These units have been 

used throughout the operation of the site. Based on claimant records from DOE that were provided to 

the NIOSH project, the minimum reporting level has been 5 μg/L from at least 1964 through 1995 

(although not explicitly identified as a minimum detectable activity (MDA), this is an effective MDA 

because positive results were not reported below this value). After the change to KPA analysis in 

1996, the detection level was reduced to 1 μg/L. Some early records (1956 through approximately 

1969) indicate that occasional sensitivities below 5 μg/L may have been achieved, but in the absence 

of specific information in the records, the 5-μg/L value should be used for dose reconstruction for 

years before 1996, and 1 μg/L should be used for 1996 and later years. Note that these samples 

might have been designated as uranium (or U), fluorometric uranium (FU), total uranium (TU), or 

natural uranium (Nat-U). 

The first records in which reporting of isotopic uranium results (234U, 235U, and 238U) were reported 

were from 1970. No contractual MDA could be identified for isotopic uranium analyses, which were 

reported in claimant records provided by DOE to the NIOSH project in units of dpm per sample or 

dpm per 24 hours before 1992, and in units of pCi per liter (pCi/L) for 1992 and later results. Results 

for 1992 and later, although not explicitly identified in any associated documentation as MDA values, 

were indicated as “<1 pCi/L,” which is consistent with the isotopic MDA provided by Eberline 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 23 of 124

(Waligora 2007). This reference indicates that the later MDA provided by Eberline (Waligora 2007) 

was also applicable for prior periods, which would equate to approximately an MDA of 0.3 dpm per 24 

hour sample. Consequently, an MDA value of 0.3 dpm per 24 hours should be applied for the period 

before 1992 for isotopic uranium samples in the absence of information in the record, which is 

consistent with reported positive values for that period. (This seems to indicate that a change in 

reporting units, rather than a change in analytical sensitivity, occurred in 1992, which is also 

supported by the information supplied by Eberline in Waligora 2007.) 

Note that urinalysis results for uranium from 1976 through 1982 were sometimes designated with the 

abbreviation “RU,” which stood for “radiometric uranium;” this designation should not be confused with 

recycled uranium, for which the abbreviation “RU” may be used at other sites. This designation was 

associated with urinalysis for uranium based on the measurement of gross radioactivity in the 

processed sample as opposed to the measurement of the mass of uranium in the sample by 

fluoroscopy. While no rationale has been identified for this method of sample analysis and no 

detection limits are identified for these samples, it is possible that they were used in an attempt to 

quantify uranium exposures from enriched uranium, for which measurement of uranium mass in urine 

samples would have been ineffective. Another designation, infrequently used, that was associated 

with uranium analysis was “SPU” or “specific uranium,” (not to be confused with plutonium analyses). 

If the uranium enrichment cannot be determined (e.g., from claim files or information in Attachments

C, D, or F of this document), the enrichment associated with fluorometric sampling results (using units 

of milligrams or micrograms) should be assumed to be natural uranium; for isotopic sampling results 

(with units of dpm or microcuries), enriched uranium should be assumed for favorability to claimants. 

Note that if a worker’s activities indicate a potential exposure to enriched uranium when only uranium 

mass measurements are found in the urinalysis records, the Battelle site lead should be consulted for 

further guidance.

Plutonium

Records of urine sample analyses for total plutonium content were found in the 1960s, then again 

beginning in 1989, with a change in reporting units found in later records. The change in detection 

limits and the change in units from dpm/24 hours to pCi/L are assumed to occur in 1992 and are 

based on a general change in reporting units for several other radionuclides (no plutonium urinalysis 

records were located for 1992). Positive reported values before 1992 were often lower than the 

routine detection limit listed in the table, which may be an indication that many of these were possibly 

special samples that incorporated optimum analytical techniques.

Isotopic analyses for individual plutonium isotopes (238Pu and 239Pu) were also performed as early as 

1964. Although the reporting units changed after 1991, the actual detection capability is comparable 

throughout Battelle operations. Later analytical reports (1990s) are more likely to include the actual 

MDA for the analysis performed. 

Tritium

To date, bioassay analyses for tritium have only been found in the records for 1963 and 1965. One of 

the samples was positive; for the remainder of the samples, the listed detection limit for tritium 

urinalysis was 1 μCi/L. Based on the record, it is anticipated that the detection limit will be available in 

the record if tritium analysis is performed for a worker.

Americium

Urine samples were infrequently analyzed for Am-241, with the earliest sample found in the records in 

1977. One sample (in 1978) listed an implied detection limit of 0.03 dpm per sample, while all other 

samples were recorded as zero with an associated error of 0.03 dpm per sample.
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 24 of 124

Thorium

Urine samples have been infrequently analyzed for thorium at Battelle. The first sample analyzed for 

total thorium content found in the records was in 1993. Isotopic thorium analyses for 228Th, 230Th, and 

232Th have been performed since 1981, with an apparent change in detection limit and in reporting 

units from 2 dpm per sample (1981–1991) to 0.2 pCi/L (beginning in 1992). Analytical MDAs are 

typically included in the analytical report for the latter period.

Fission products

Gross beta analysis of urine samples has been performed for a major part of Battelle operations, with 

the earliest result found in the records in 1957. The earliest analyses were for both gross beta (which 

were usually positive due to the contribution of naturally occurring 40K) and for gross beta minus 40K. 

Because the 40K contribution to the gross beta results would typically have masked any measurement 

of occupationally related radionuclides, only the gross beta minus 40K results for the early period, 

assumed to represent fission products, are useful for assessing intakes. Beginning in 1961, analyses 

were designated as FP (fission products); beginning in 1980, they were designated as MFP (mixed 

fission products).

Units were initially reported as dpm/mL, then changed to pCi/L in more recent operations; the year 

listed in the table in which the change in units occurred, 1992, is based on a general change in 

reporting units for several radionuclides and should be considered approximate. (It should be noted 

that a few sample results for the earlier period were recorded in units of dpm/24 hr, which is likely a 

transcription error because the magnitudes of the reported values and associated errors is generally 

the same as for the results reported in dpm/mL. Consequently, units in “dpm/24 hr” for fission product 

or gross beta analysis results before 1992 should be interpreted as dpm/mL unless the record 

indicates otherwise.) 

Detection capabilities for the analyses were seldom recorded in all but the most recent periods of 

Battelle operations. Detection capabilities appear to have been related to sample aliquot size, which 

made them vary from sample to sample. Complicating this were occasional sample results much 

lower, sometimes by an order of magnitude or more, than typical results, possibly as part of a special 

sampling or incident follow-up regime. In general, detection capabilities appeared to improve 

throughout operations. MDAs were not routinely reported until the 1990s. The MDA for the period 

between 1962 and 1991 was derived empirically based on counting parameters from available 

claimant analytical reports provided by DOE to the NIOSH project [1]. The MDA after 1991 is based 

on information from Eberline (Waligora 2007). 

4.5 INHALATION ABSORPTION TYPES, PARTICLE SIZES, AND ACTIVITY FRACTIONS

Attachment F lists radionuclides and the most commonly encountered chemical forms by facility and 

building. Although the list of materials used in each building is fairly complete in Attachment F, there 

may be some materials or forms that are not included; however, these would represent materials or 

forms that were used in very limited periods and areas, such as in an alloy study in which the

materials were studied only briefly. It should also be kept in mind for the reactor that the irradiated 

materials were encapsulated before irradiation and were put into containers as they were removed 

from irradiation, so that the materials that reactor workers and researchers were exposed as far as 

internal dose goes were the normal reactor contamination (fission products and activation products) 

such would be encountered in working in the pool or in removing items from the pool. Similarly, the 

plutonium work in JN-2 involved only encapsulated sources and thus only incidental contamination 

was available for intake.

Table 4-4 shows isotopic ratio data for weapons grade and fuel grade plutonium from ORAUT (2010a) 

and ORAUT (2005b). The data from these references are applicable to the Hanford and Savannah 

River sites but are appropriate for application to Battelle because these sites were the primary 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 25 of 124

producers of plutonium for the DOE complex. These ratios should be applied in the absence of 

specific information in the record. The activity ratios for “fresh” plutonium in Table 4-4 should be used 

for Battelle dose reconstruction if the work involves recently separated weapons or reactor grade 

material. If the age of material is unknown, the 10-year aged ratio should be used as a favorable to 

claimant assumption. For heat source material (as indicated by work activity or bioassay monitoring 

for 238Pu), the activity ratios of all plutonium isotopes other than 238Pu are insignificant, and the 

material may be assumed to be 100% 238Pu (DOE 2006). The activity of 242Pu in these mixtures does 

not contribute significantly to dose and need not be considered.

Table 4-4. Isotopic composition data for plutonium work.

Isotope

Specific activity in mixture (Ci/g)

Weapons grade 

material (fresh)a

Weapons grade 

material (10-year 

aged)a

Reactor grade 

material (fresh)a

Reactor grade 

material (10-year 

aged)a

Pu-238 8.56E-03 7.91E-03 1.71E-02 1.58E-02

Pu-239 5.77E-02 5.77E-02 5.26E-02 5.26E-02

Pu-240 1.36E-02 1.36E-02 2.72E-02 2.72E-02

Pu-241 8.24E-01b 5.09E-01b 3.09E+00b 1.91E+00b

Am-241 0 1.04E-02 0 3.8E-02

a. ORAUT (2005b, 2010a).

b. Pu-241 is primarily a beta emitter with insignificant associated alpha activity.

Various authors note that 241Am ingrowth should be assumed for 239Pu forms and contamination when 

the plutonium contains a few weight percent of 241Pu or more and has aged and that this 241Am should 

be assumed to have the same solubility type as the host matrix (DOE 2006). Am-241 data is included 

in Table 4-4 because Flynn et al. (1987) remarks that for the major research activity at JN-4, i.e., the 

Mixed Nitride Reactor Fuel Program, 241Am was "reportedly associated" with much of the plutonium 

material. 

If the chemical form of material involved in potential intakes for a worker can be obtained from the 

record, ICRP guidance (ICRP 1995) may be used to determine the appropriate choice of the 

inhalation absorption type. If this information cannot be discerned adequately from the claim records, 

favorable to claimant assumptions should be used for these parameters.

ICRP (1995) indicates that plutonium materials encountered in the workplace may be absorption 

Types M or S, depending on the chemical and physical form of the material. Additionally, some 

intakes plutonium oxides have exhibited long-term retention of plutonium in the lung exceeding that 

predicted by the standard Type S model. Because oxidation or high-firing temperatures are 

applicable to some of the plutonium processes at Battelle, Type Super-S (or Type SS) absorption, 

described in ORAUT (2008), must be considered as well for situations in which Type S plutonium 

could be encountered. The exception to this is for workers who were primarily exposed to 238Pu, as 

identified by either work history (e.g., work with heat source material) or bioassay monitoring (e.g., 

specific analysis for 238Pu only), for which Type Super-S is not applicable in accordance with the 

guidelines of ORAUT (2008). 

The dose reconstructor should assume that the route of intake for radionuclides was inhalation and a 

particle size distribution of 5 μm activity median aerodynamic diameter (AMAD), the default ICRP 

Publication 66 particle size is used (ICRP 1994). If specific information is available in the record to 

the contrary, that information should be used.
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4.6 DOSE RECONSTRUCTION CONSIDERATIONS FOR FISSION PRODUCTS

For the particular case of mixed fission products for which bioassay was performed with results as 

only gross beta or gamma measurements, ORAUT-OTIB-0054 (ORAUT 2007b) may be used to arrive 

at appropriate source terms for internal dose reconstruction.

The guidance of ORAUT (2007b) may be applied to urinalyses for workers at the reactor, JN-1 (the 

hot cell facility), and at any King Avenue facility where spent fuel (but not solely fission gases) was 

analyzed. ORAUT (2007b) states that in order to determine if radioiodine intakes need to be used, 

site profile and claim-specific information should be reviewed to determine if chronic iodine intakes 

were feasible because iodine exposures are rarely seen in operations that used adequate ventilation 

and some form of collection, holdup, or filtration; for the workers described above, radioiodine 

exposure in the course of work with spent fuel or spent fuel samples need not be considered. The 

spent fuel cooling (decay) times are listed in Table 4-5 for each activity.

Table 4-5. Cooling times for fission products at Battelle.

Activity (facility) Cooling time

JN-1 (West Jefferson) 1 year

Battelle Research Reactor (West Jefferson) 10 days

Fuel dissolution (King Avenue) 1 year

ORAUT-OTIB-0054 (ORAUT 2007b) is applicable to the hot cell facility because the overwhelming 

bulk of its work was the examination of irradiated reactor fuel specimens and fuel assemblies; this fits 

into the constraints of ORAUT-OTIB-0054 because as Wastren (2001) states, the types of testing 

done on this material "did not involve separation of constituent elements from the fuel" and as 

Pasupathi and Toy (1990) state, operations at JN-1 were almost always dry except for the fuel 

dissolution work, the acid etching and water rinsing that might be done in the metallographic phase of 

the examination work, and the water used with the abrasive grinding wheel in fuel sectioning. 

Because of this, the HEPA filter loads for most cells would consist of particles of spent fuel and 

activated material (e.g., cladding bits). Irradiated fuel samples were also analyzed at King Avenue in 

limited ways, but these were small samples and were usually not highly irradiated, so as to limit the 

external dose rate. Thus a few King Avenue laboratory workers and on occasion a few machine shop 

workers (e.g., ORAUT 2006a) would likely have been exposed to this type of source. Finally, at King 

Avenue, some test dissolution of clad lightly irradiated fuel pins was done (on and off from 1958 to 

1961, as different dissolution processes were tested on a small scale).

It appears that at Battelle, during the period covered by this site profile, all areas where materials 

containing fission products or activation products were handled had adequate ventilation. The reactor 

was the pool type and had exhaust over the pool in addition to the building and other exhausts; at JN1, the spent fuel samples or whole pins or assemblies were opened in the (closed) hot cells, which 

had individual cell exhausts, and there was a building exhaust as well; and at King Avenue such 

samples appear to have been opened and examined in a glovebox (in the Radioisotope Laboratory) 

or a hood, which of course had an exhaust system as did the buildings. None of these facilities had 

iodine filters (except that JN-1 had a charcoal filter from 1978 on to which the exhaust could be 

diverted if an I-131 monitor so indicated). However, HEPA filters were provided for the individual cells 

at JN-1 and for the individual hoods and glovebox at King Avenue, while prefilters and double HEPAs 

were provided for the reactor and JN-1 building exhausts. From at least 1973 (and probably earlier), 

continuous air monitors were used for the West Jefferson building stacks and on the Building 3 

exhaust stack; samples were changed and counted weekly (BCL 1974). The iodine levels were said 

to be well under ERDA standards. Thus it seems likely that these workers had little opportunity to be 

exposed to radioiodine.
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The exception was when the radioiodine was part of a specific study, such as of fission gases, or 

when the iodine may have been used in radiotracer or similar work; in these cases, if the gas sample 

was taken off to a laboratory for analysis there was some potential for exposure although the sample 

container was likely opened in a hood or glovebox. However, the fission gases were generally 

collected off a sample or subassembly in a closed in-reactor sample loop or by vacuum container in a 

hot cell; in both of these cases the gases were counted in the loop or container or they were allowed 

to decay in a container so that the daughters could be trapped and counted. In the reactor, the 

sample or subassembly was removed underwater into a cask in an airtight condition and the gases 

did not escape to the pool (Basham and Rieder 1960). However, this fission gas work is not covered 

under ORAUT (2007b). Urinalyses done specifically for only I-131 (e.g., not also for other fission 

products for the same urine samples) should generally be considered to be part of fission gas, 

radiotracer, or similar work. (A description of the fission gas work is given in a footnote to the table in 

Attachment C.)

ORAUT (2007b) states that "it is generally favorable to claimants to use longer decay times because 

that increases the relative activities of the longer-lived radionuclides, which have higher 50-year dose 

conversion factors." Consequently, the cooling time is assumed to be one year for JN-1. This is 

consistent with Battelle's 1962 accident analysis for JN-1 (Sunderman and Dickerson 1962) that 

assumed a four-month cooling time for an "OMRE" element that was then sent to Battelle for analysis; 

the actual cooling time was usually longer than four months, although it could be as short as 10 days 

for Fermi assemblies taken to JN-1 (Sunderman and Gates 1965). It is also consistent with the fact 

that the personnel exposure would have been after the campaign of which the fuel testing was a part, 

i.e., during cell cleanup (when the cell might have been opened or at a minimum when waste and 

samples were moved out through the ports and transported for packaging or processing and for 

storage). The cooling time for the reactor (BRR) and other facilities handling irradiated samples 

should be assumed to be 10 days, consistent with the short test reactor cycle or the likely 

turnaround time of the sample. Finally, the cooling time for the fuel dissolution at King Avenue should 

be assumed to be 1 year, which seems consistent with the actual case of about 10 months for 

Consolidated Edison test pins irradiated at ORNL and sent to Battelle for analysis (Dayton and Tipton 

1960a).

4.7 RECORDS AND REPORTS

The urinalysis and other bioassay records available for the Battelle sites are of three types. The first 

type is the bioassay report issued by the bioassay analytical service provider; the second type is the 

Battelle (company) summary card on which an individual worker's urinalysis information is tabulated, 

usually over a year; and the third type is the Battelle ad hoc documentation, such as incident reports 

that give information as to potential exposure levels and routes. The principal record to be used in 

dose reconstruction is the report from the bioassay analytical service provider, but the other two 

record types may provide information relevant to missing bioassay readings or to special bioassays 

sampling situations.

4.8 INTERNAL DOSE ASSESSMENT FOR UNMONITORED PERIOD (BEFORE

SEPTEMBER, 1956)

[RESERVED]
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5.0

5.1 DOSIMETRY PROCESSING SERVICES

OCCUPATIONAL EXTERNAL DOSE

It is not clear when badging started at Battelle. The earliest date for a film badge that has been found 

in the available records is June 5, 1953. Badges appear to have been worn thereafter throughout the 

period of operation and the period of residual radioactivity, including during decontamination and 

decommissioning (D&D). Table 5-1 gives summary information about badge service providers. 

Table 5-1. Dosimetry processing services.

Period Processing service

June 1953–March 1956 Unknown

April 1956–1967 Landauer

1968–1969a Eberline

1969a–June 1994 Landauer

July 1994–1996 Siemens/ICN

1997–present Landauer

a. Records indicate that both Landauer and Eberline 

provided processing service in 1969.

Badges appear always to have been processed by a film service, although it is possible that the first 

badges (before 1956) were read by the AEC New York Operations Office (NYOO). Landauer 

performed film dosimeter processing from April of 1956 through 1967. In 1968, processing was 

performed by Eberline. In 1969, records show that both Eberline and Landauer were used as 

processors. Landauer was again the sole processor beginning in 1970 and continued film processing 

through June 1994; TLD dosimeters were added in 1984. It should be noted that some TLD badges 

were in use before 1984, and film badges continued to be used well after this date for some 

individuals. Processing was performed by Siemens/ICN from July 1994 through 1996, then again by 

Landauer beginning in 1997 through the present.

5.2 DOSIMETER TYPES

Film dosimeters were used to measure whole-body dose from the start of monitoring through at least 

1984, with whole-body TLD dosimeters introduced in 1985. During the film badge period, either beta, 

gamma, and X-ray sensitive dosimeters or beta, gamma, X-ray, and Kodak nuclear track emulsion, 

type A (NTA) dosimeters were issued (Author unknown undated). With the introduction of whole-body 

TLD dosimeters, neutron-sensitive TLD materials, and CR-39 neutron dosimeters (chemically etched 

plastic measuring recoil proton tracks from fast neutron interactions) were also implemented. 

Ring or wrist badges were worn routinely by most laboratory, hot cell, machine shop, and similar 

workers who handled hot sources, equipment, or waste with their hands, even if lead gloves were also 

worn (e.g., Ottman 1959, 1962, 1968, 1970; Kirsch 1976; Toy 1981). Ring badges appear to have 

used TLDs beginning in 1970. At JN-1, each person who was temporarily or permanently assigned to 

the facility was issued a ring badge along with a basic film badge (Sunderman and Dickerson 1962;

Sunderman and Gates 1965). At JN-4, as of at least 1977 (the year before D&D began) staff 

members were issued beta-gamma and X-ray TLD ring badges and beta-gamma, X-ray, and neutron 

wrist badges (BCL 1977a); ring badges were required to be worn by all persons who worked with

gloveboxes (BCL 1977a). 

The use of pencil (pocket or self-reading) dosimeters is mentioned in several individual dosimetry 

records. It was noted specifically for the Radioisotopes Laboratory (Building 4), the Hot Cell 

Laboratory (JN-4), and presumably for the Gamma Facility (Building 6) that when staff members 

worked in areas where they might receive an exposure that approached the maximum permissible 

daily dose, they wore pocket ionization chambers to keep track of their whole-body gamma dose 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 29 of 124

(Sunderman and Dickerson 1962; Sunderman and Gates 1965). This apparently included every entry 

into a radiation area with contamination because a pocket dosimeter was listed among the standard 

protective clothing and equipment set for such entries (Sunderman and Dickerson 1962; Sunderman 

and Gates 1965).

5.3 DOSIMETER DETECTION CAPABILITIES

Table 5-2 gives a summary of the minimum detection levels (MDLs) for evaluation of exposures that 

were measured by film badges and TLDs and for assessing missed dose (Landauer 1956–1997). 

Table 5-2. Summary of MDLs (mR through 1958, mrem beginning 1959) for dosimeter 

badges.

Period Photon Beta

Neutron

(fast)

Neutron 

(slow)

3/53–3/56 50 N/A N/A N/A

4/56–3/57 30 50a 60b N/A

4/57–60 20 50 60 N/A

61–63 10 40 50c N/A

64–84 10 40 50 10

85–6/94 10 40 50 10

7/94–96 10 40 50 N/A

97–present 10 40 20 10

a. Generally no beta doses were reported prior to April 1957, although records indicate beta 

measurements might have been made at times for work with uranium. If beta dose monitoring 

is indicated, recommend use of 50 mrem as the MDL.

b. Based on values recorded as “under 10% MPL” (MPL = maximum permissible limit); the MPL 

for the period was 300 mR/wk, so 10% of that value for a two-week period is assumed to be 

60 mR.

c. There is inconsistency in the fast neutron MDL. In early 1964, an MDL of 15 is stated and later 

reports identify 20 mrem, and then 50 mrem during 1994-96. Because it is unlikely that an 

actual decrease in sensitivity would occur, an MDL of 50 mrem is recommended from 1961-

1996.

Film badge reports from before April 1956 list only gamma and X-ray exposure components that are 

tabulated from shielded and unshielded dosimeter readings; units are in milliroentgens, with the MDL

listed as <50 mR.

Beginning in April 1956, neutron exposure was tabulated and specified as "fast neutron" with the 

results in percent of the MPL; the MDL was typically recorded as "<10% of the MPL" with no other 

units listed. Because the MPL for the period (also listed on the reports) was 300 mR and the 

dosimeters were exchanged on a biweekly frequency, it is assumed that the MDL for fast neutrons 

was 60 mR (i.e., 10% of 300 mR/wk for a two-week monitoring period) during this period. The bottom 

of the dose report forms stated that a recorded gamma exposure of zero indicated less than 30 mR 

unless otherwise noted. 

In April 1957, the minimum reported dose value was changed to 20 mR for high-energy X- and 

gamma rays and another category headed "Other" or “millireps if U beta” was added; this column is 

assumed to represent beta dose because in April 1958 this category heading was changed to "Other, 

Arbitrary Units," and Landauer applied this terminology in reporting later Battelle beta dose and in 

reports and memoranda dealing with beta dose (e.g., Selander 1961; Ottman 1959, 1961a). This 

usage was apparently due to the fact that the film badge readout-to-dose conversion factor (typically 

listed as 2.2 for <1.4 MeV energy betas) depended on the calibration for each batch of badges. Units 

of millirep for uranium beta exposures were also noted during this period; the heading "Beta" was not 

routinely included in reports until 1961. Neutron exposures continued to be listed as described for 

1956.
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Beginning in 1961, Landauer reports began listing dose categories of gamma- and X-ray, beta, and 

neutron, with MDLs of 10 mrem for gamma- and high-energy X-rays, 40 mrem for beta, and 15 mrem

for neutrons. It should be noted that from 1961 through the first part of 1964, Landauer listed an MDL 

of 15 mrem for fast neutrons, which was changed to 20 mrem later in 1964 and for subsequent years

through the present. The value of 50 mrem is used as a default value in this site profile as a more 

realistic estimate and because it is unlikely that a decrease in detection sensitivity occurred during the 

later period. The neutron doses were not categorized by energy, so it is assumed that the doses

described fast neutrons, as was the case before 1961. It is also noted that neutron doses changed 

from percentages of the MPL to millirem in 1961.

Beginning in 1963, reports began to distinguish between thermal and fast neutron exposures, with 

stated MDLs of 15 mrem for fast neutrons and 10 mrem for thermal neutrons; as noted above, an 

MDL of 50 mrem is assumed for fast neutrons in this site profile. No changes to the Landauer MDLs 

were noted after 1963; therefore, the values of 10 mrem (gamma- and high-energy X-rays), 40 mrem

(beta), 50 mrem (fast neutrons), and 10 mrem (thermal neutrons) were used for all periods of 

Landauer processing beginning in 1963 through 1996. 

Landauer continued film dosimeter processing until 1967. In 1968, processing was performed by 

Eberline. Contract proposal documentation for Eberline indicates reporting ranges comparable to the 

Landauer MDLs (Eberline 1967). Because no other documentation was found for Eberline that 

related to MDLs, the same values were used during the period of Eberline processing. As noted 

above, both Eberline and Landauer were used as processing services for 1969. Landauer was again 

the sole processor beginning in 1970 and continued film processing through June 1994; TLD 

dosimeters were added to the service in 1984. As noted above, records indicate some TLD badges 

were in use before 1984, and film badges continued to be used well after this date for some 

individuals. 

Processing was performed by Siemens/ICN from July 1994 through 1996, with MDLs as noted in 

Table 5-2, then again by Landauer beginning in 1997 through the present. 

5.4 DOSIMETER EXCHANGE FREQUENCIES

Table 5-3 shows the default dosimeter frequencies that were used to monitor the primary groups of 

workers at Battelle (Eberline undated; BCL 1973; Author unknown undated).

External dose records, computer-assisted telephone interviews, and program documentation from 

Battelle indicated that badges were exchanged at 1-, 2-, and 4-week, monthly, and irregular (“special”) 

intervals dependent on the worker's job (Eberline undated; Yarger 1966; BCL 1973; Author unknown 

undated). Records from Battelle document that the primary exchange frequencies were bi-weekly 

and every 4 weeks, as shown in Table 5-3 (the 4-week frequency was changed to monthly in 1970). 

The exchange frequency was typically stated on Landauer reports that date from at least 1961. The 

account numbers in Table 5-3 were also listed on Landauer reports from 1969 through about 1984. If 

information in claimant records is insufficient to determine the worker’s dosimeter exchange 

frequency, then the dosimeter exchange frequency should be based on the work group that is shown 

in Table 5-3; if the work group cannot be determined, then a bi-weekly exchange frequency should be 

assumed to be favorable to the claimant.

Records show that dosimeter exchange frequencies were specified by Battelle based on work location 

and the type of work, and they might have been different for different work groups during the same 

period or for individual workers over their periods of employment. Records for one individual included 

weekly, biweekly, and monthly monitoring frequencies including two consecutive years where he was 

monitored with all frequencies. Another individual had several ”special” dosimeters that were 

apparently multiple dosimeters (i.e., more than one dosimeter was issued to the individual), but that 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 31 of 124

Table 5-3. Default dosimeter exchange frequency by facility or work group.

Facility or work group

Account No. 

(beginning with 1969 

Landauer reports)

2-week 

exchange 

frequency

4-week exchange 

frequency (before 1970) 

or monthly (after 1970)

Battelle Research Reactor (BRR) 6410 X

Critical Assembly Lab (CAL) 6412A X

Hot cell (HC)a 6408 X

Plutonium Lab 6409 X

Uranium Lab (25 area) 6412B X

Safety -- X

Visitor biweekly 6412Z X

Misc. biweekly 6412D X

Radioisotope Laboratory 2302 X

Analytical chemistry 2366 X

Radiography 2613C, 6412C X

Visitor monthly 6413Y X

Misc. monthly 6412E, 6413E X

a. Hot cell workers might also have been on a weekly exchange frequency (Sunderman and Dickerson 1962).

were issued and retrieved at nonoverlapping periods in 1965 – these might have been in accordance 

with multibadging or planned special exposure procedures (Stewart 1995). Concurrent use of film 

and TLD dosimeters was also noted. The dose reconstructor is responsible for evaluating available 

records to determine the monitoring frequency that is most appropriate to support missed dose 

assessments.

It should be noted that dose monitoring records appear to be missing for short intervals in some 

individual records. This could be due to actual missing records or due to the fact that Battelle based 

the level of monitoring on the exposure potential of the worker and task, so that individuals with little 

or no potential exposure were not monitored (i.e., the lack of monitoring was intentional based on low 

dose potential). The dose reconstructor must evaluate available records and determine whether 

apparent gaps in monitoring require the assessment of unmonitored dose for those periods. Battelle 

records appear to be adequate to support interpolation methods outlined in OCAS-IG-001 (NIOSH 

2007b, Section 3) to determine doses for gaps in records that are deemed to be due to missing 

reports (rather than to the intentional nonmonitoring of workers with low potential for exposure). More 

significant gaps in monitoring data must be evaluated on a case-by-case basis to determine if 

additional requests for dosimetry records should be made. 

5.5 WORKPLACE RADIATION FIELDS

Attachment G summarizes site processes with radiation types and appropriate energy ranges for the 

King Avenue and West Jefferson sites. Note that because the West Jefferson facilities did not start 

up until 1955 or later, exposure to reactor neutrons or plutonium-emitted neutrons was not a 

possibility, except for the limited 239Pu work in a laboratory in Building 6 in 1953 to 1955 and 1961. 

However, the possibility of neutron exposure from the alpha-neutron reaction in appropriate forms, 

especially for UO2 and ThO2, should still be considered in accordance with the guidance of ORAUT 

(2005c). 

Attachment H summarizes the appropriate energy ranges for neutrons. For neutron doses, the 

measured doses should be modified by the ICRP Publication 60 correction factor for the appropriate 

energy range (ICRP 1991), in addition to the energy fractions in Attachment H. These factors 

incorporate the ratio of the Publication 60 weighting factor to the group-averaged NCRP Report 38 

(NCRP 1971) quality factor. These factors are 1.86 for the 10- to 100-keV energy range), 1.91 for the 

0.1- to 2-MeV energy range, and 1.32 for the 2- to 20-MeV energy range (ORAUT 2006b).
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 32 of 124

If information on the specific work processes, periods, or locations can be determined from the record, 

the appropriate energy range fractions in Attachments G and H should be applied. Otherwise, 

assumptions favorable to claimants of 100% 30- to 250-keV photons and 100% 0.1- to 2-MeV 

neutrons (if applicable) should be used; for shallow doses, either 100% >15-keV electrons (for 

nonplutonium-related work) or >30-keV photons (for plutonium-related work) should be used. 

Workers at the West Jefferson site (JN-1, JN-2, JN-3 and JN-4) might have worked with gloveboxes. 

The presence of extremity monitoring (wrist or ring data) can be used as an indicator for glovebox 

work or handling radioactive material [2]. The application of the guidance in NIOSH (2005) appears to 

be appropriate in relation to workers who handled plutonium in gloveboxes at JN-2 (1964 to 1970) and 

JN-4. 

5.6 ASSESSMENT OF NEUTRON DOSE MONITORED BY NTA FILM 

The Battelle West Jefferson Nuclear Sciences Area contains several major buildings with the potential 

for workplace radiation fields (Wallace 2010) as described in Attachment I. Due to the underresponse 

of NTA film dosimeters to low energy neutrons, some neutron exposures may be considered 

unmonitored or inadequately monitored by NTA dosimeters. Consequently, except for exposures to

neutrons from 252Cf at JN-1 (as discussed in Attachment I), neutron doses for the film dosimeter 

period may be assessed based on neutron to photon exposure ratios derived from workplace 

radiation paired (i.e., same time and location) measurements of neutron and photon radiation dose 

rates that have been obtained from selected Battelle instrument survey records. These ratios should 

be applied to measured photon doses to obtain neutron doses for workers for the period when film 

dosimeters were in use (i.e., prior to 1985) and neutron exposures would not have been adequately 

monitored by NTA film. These ratios are not used if records indicate that TLD dosimeters were used 

to for monitoring. Recommended neutron to photon ratios are summarized in Table 5-4 below, based 

on the analysis in Attachment I.

Workplace radiation paired (i.e., same time and location) measurements of neutron and photon 

radiation dose rates were obtained from selected instrument survey records extending over several 

years for Battelle facilities as follows: 

• JN-1 Hot Cell Laboratory (1955-1983). Limited survey data has been received from Battelle 

consisting of 49 measurements for the time period when 252Cf was being processed and handled 

in very large quantities. Neutron to Photon dose (NP) ratio analysis based on this measured data 

presents an upper bound that if applied to claimants over an entire work career would result in a 

significant overestimate of their neutron dose. There are a few claimants who worked during other 

time periods where presently no survey data is available for analysis. A review of potentially 

impacted claims indicated that these claimants have previously been judged to be compensable 

because of high assigned external and internal doses. 

• JN-2 Critical Assembly Laboratory (CAL) (1955-1975). Limited survey data consisting of 15 

measurements has been received from Battelle and for the more recent operational time period

only. No survey data has been received from Battelle for the early years of operation. Fortunately 

at this time, there are no claimants who worked during the early time period. Also, Battelle 

provided information for the plutonium laboratory in this building involved only sealed sources. 

• JN-3 Battelle Research Reactor (BRR) (1960-73). Substantial survey data was received from 

Battelle for the period of 1960 through 1974 consisting of 2,487 paired measurements selected for 

analysis from 3,188 paired measurements identified in the documentation. This is considered to 

be substantial data to develop a robust estimate of the NP ratio for this facility.
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• JN-4 Plutonium Laboratory (1968-76). Survey data was received from Battelle consisting of 67 

paired measurements. This is considered to be sufficient to develop an estimate of the NP ratio 

for this facility.

• JS-1 Naval Reactor Research Laboratory. No paired neutron and photon dose measurements 

were received from Battelle apparently because there were no significant neutron radiation fields 

in this building. The work involved some depleted uranium historically. The building was 

decommissioned in 1990 and based on the type of operation would not be expected to have 

significant neutron radiation fields. 

Results of statistical analyses of the measurement data are summarized in the following:

 Table 5-4. Recommended neutron to photon ratios for dose monitored by NTA film.

Workplace location Years n GM GSD 95th%

JN-1 (Hot Cell Laboratory)

Non252Cf Operations 1955–1983(a) (b)

252Cf Operations 1971–1974 49 2.4 3.9 22.5

JN-2 (Critical Assembly Laboratory)

Critical Experiments 1956–1960 (b)

Plutonium Laboratory 1967–1975 14 1.0 1.3 1.5

JN-3 Battelle Research Reactor

Combined Areas 1956–1974 2487 0.2 4.2 2.1

JN-4 Plutonium Laboratory

Combined Areas 1960–1978 41 0.7 5.0 9.9

JS-1 Naval Reactor Research (c)

a. Neutron exposures probably very limited prior to construction and operation of alpha gamma 

 cells in 1964. 

b. Reserved—additional survey data requested

c. No paired survey data received. Depleted U-238 used in this building with essentially no 

 potential for significant neutron radiation.

Resultant neutron dose is combined with the energy distribution appropriate for the task as shown in 

Attachment H, or should apply an assumed energy range of 100% 0.1 to 2 MeV if the specific task is 

unknown.

5.7 INCIDENT REPORTS

A form for reporting minor incidents such as overexposures, lost badges, etc., was used as early as 

June 1958 (e.g., Selander 1958a,b); the form was also used to document planned overexposures 

(e.g., Swigert 1963). While most of the copies of these forms found in records obtained by the NIOSH 

Project apply to occurrences at the West Jefferson site (mostly at the BRR and the hot cell facility), a 

few apply to occurrences at the King Avenue site (e.g., Kirsch 1976); it is likely that there were fewer 

such reports for King Avenue because of the lower potential for significant exposures.

5.8 EXTERNAL DOSE ASSESSMENT FOR UNMONITORED PERIOD (BEFORE JUNE, 

1953)

[RESERVED]
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6.0

Based on information provided in the Report on Residual Radioactive and Beryllium Contamination at 

Atomic Weapons Employer Facilities and Beryllium Vendor Facilities (NIOSH 2006), Battelle 

Laboratories- West Jefferson, had the potential for residual contamination outside of the period in 

which weapons-related production occurred (1976-1985). 

RESIDUAL EXPOSURE

Individuals who worked at the West Jefferson site in radiological areas were monitored for internal 

and external radiation exposure during the period of potential residual contamination. The use of the 

existing monitoring records in the performance of dose reconstructions is sufficient to account for 

internal and external doses during this period. No additional dose assessment is necessary to 

complete dose reconstructions during the West Jefferson residual contamination period (1976-1985) 

[3].
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 35 of 124

7.0

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction.

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

ATTRIBUTIONS AND ANNOTATIONS

[1] King, Vincent. MJW Corporation. Dose Reconstructor. May 2007. 

The fission product MDA for 1962 through 1991 was determined empirically by Thomas La 

Bone using information in available analytical records from Battelle. 

[2] Robinson, Sallie. Dade Moeller & Associates. Dose Reconstructor. June 2007.

The conclusions in this paragraph were made based on a review of the records provided for 

Battelle Columbus.

[3] Robinson, Sallie. Dade Moeller & Associates. Dose Reconstructor. June 2007. 

The conclusion not to add additional dose for residual contamination is based on the Battelle 

Columbus records that indicate the workers who had the potential for radiological dose were 

monitored. 
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 36 of 124

REFERENCES

AEC (U.S. Atomic Energy Commission), 1947, Report on Research and Development, omnibus 

progress report. [SRDB Ref ID: 9269, p. 7]

AEC (U.S. Atomic Energy Commission), 1949, Health Hazards in NYOO Facilities Producing and 

Processing Uranium, the New York Operations Office Medical Division, New York, New York, 

18 April. [SRDB Ref ID: 3931 or 17068]

AEC (US Atomic Energy Commission), 1950, Miscellaneous air data sheets showing samples taken 

by Battelle and analyzed by NYOO, March-April. [SRDB Ref ID: 8612]

AEC (U.S. Atomic Energy Commission), 1951, The Production of Uranium Feed Materials, AEC 

Director of Production, Washington, D. C., with note by Secretary R. B. Snapp, 22 May. 

[SRDB Ref ID: 4125]

Amstein, T., 2006, No subject, Letter from Battelle Science and Technology International[sic] to M. 

Farris of the US Department of Energy, Columbus, Ohio, July 24. [SRDB Ref ID: 34444]

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Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 60 of 124

GLOSSARY

23 recovery program

The program to recover thorium from various materials, such as monazite sands, using anion 

exchange in a carbonate solution.

25 Area

The uranium receiving/dispensing area in Bldg 3 (Materials Bldg).

Atomic Science Center

Same as Nuclear Sciences Area.

berry bucket/can

One-gallon can used to hold an irradiation or other sample or waste in a hot cell during or after 

an experiment.

coffee cup

Foam coffee cup into which dissolved slugs in liquid form were poured; concrete was added to 

solidify the liquid in the cup.

Contractor Pool

The instrument/experiment end of the BRR pool, also called the Shielding Pool.

Davison

A predecessor or subsidiary of W. R. Grace Co.

Engineering Area

Auxiliary (south) part of West Jefferson site, including Bldgs JS-1 (Hot Isostatic Processing 

Facility for fabricating military reactor fuel) and JS-10 and JS-12 (studies of explosive 

forming/bonding techniques, ballistic studies using nuclear materials).

fissium

Simulated fission-product-containing material as a mixture of nonradioactive isotopes of those 

elements formed as radioactive fission products, for study of chemical and physical behavior 

of fission-product-bearing material.

junior cave

Medium-shielded hot-cell-type room used for handling or storage of lower level sources than a 

fully shielded hot cell or high-level cave or vault would be.

Lamson tube

Pneumatic tube used to move items between the Low Level Cell and Alpha-Gamma Cell 4 

(JN-1).

Metmount

Fuel rod section specimen set in 1.75-in. frame to prepare it for metallurgical examination; 

also, the frame itself.

Myrnalloy

Thorium metal or thorium alloy.

New Plutonium Lab

(Part of) Building JN-4.
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 61 of 124

Nuclear Research Center

Same as Nuclear Sciences Area.

Nuclear Sciences Area

Main (north) part of the West Jefferson site, including Bldgs JN-1 (hot cell facility), JN-2 

(criticality assembly, adjunct plutonium laboratory), JN-3 (reactor facility), and JN-4 (main 

plutonium laboratory).

Old Plutonium Lab

Original section of Building JN-4.

reactor pool

The core end of the BRR pool.

sheep shed

Waste Storage Shed (associated with JN-1).

slug

Burnup analysis specimen waste, after being dissolved, studied, mixed with cement, and 

allowed to solidify in a foam cup.

soda salt

Sodium uranate, Na2U2O7.

target room

A target fabrication area at the King Avenue site.

test cell

Hole in the reactor core to insert test specimens (not a hot cell).

Tu-Be alloy

Uranium-beryllium alloy (Tu was synonymous with uranium).

Tubealloy, tuballoy

Uranium, uranium alloy, or U-Be alloy (Tu was synonymous with uranium).

weasel

Special type of sample capsule, or the BRR fast-flux irradiation facility in which it was used. 

The capsule held an aluminum container, which had ~5 g of powdered UO2 packed into quartz 

tubes. The facility was hydraulically operated and could be positioned in various regions near 

the core.

West Jefferson North

Same as Nuclear Sciences Area.

West Jefferson South

Same as Engineering Area.

White Cross

Hospital in Columbus used by Battelle.
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 62 of 124

Event

Period/ 

start date Description Referencea

Initial work for US Office of Scientific 

Research and Development under 

Contract No. OEMsr-85

1942 Rolling; metallurgical studies of gas 

evolution; consulting on extrusion

Smith et al. 1988; BCL no date a

Experimental and pilot work for MED 

begins under Contract No. W-7405-

ENG-92

16 April 

1943–1944

Done at irregular intervals, mostly on U 

metal: analysis, extrusion into rods, 

hammering, rolling of billets, dipping, 

machining, plating. Some studies of 

uranates, U carbides

Author unknown 1992; Ferry 1943; 

Langsam and Carter 1986; BMI 1945a

R&D and other work with Th metal and 

alloys (except extraction)

1944–1954 Mostly R&D (1944–1946); forming 

ingots into Hanford slugs (1947); 

hydriding (1951); electroplating metals 

on Th (1948-1953); arc-welding Ames 

Th, alloys; iodide process (1951-1954); 

bomb reduction of Th halides

BMI 1945a,b; Udy and Boulger 1951; 

Monroe, Martin, and Voldrich 1952; 

Schwope, Muehlenkamp, and Marsh 

1952; Veigel, Sherwood, and Campbell 

1953; Beach et al. 1953; Goldhoff, 

Ogden, and Jaffee 1953; Berry, Pray, 

and Peoples 1954; Saller, Dickerson, and 

Foster 1955

R&D and pilot programs to extract 

uranium from various materials

1946–1954 Extraction from phosphate rock (Sep 

1946–1950), shales (Sep 1946–1949, 

1952–1954). Mech beneficiation; pyro-, 

electro-, and hydrometallurgical 

methods; shotgun assays

Beyer 1948a,b; Kelley 1948a; BCL no 

date a

Studies of U compounds and alloys, 

including alloying Be with U; start of 

work with UO2

1947 on Properties of U-Be compounds and 

alloys; graphite-U fuel rods, rolling 

techniques for Al-U alloys. In 1947, 

listed as "Tu-Be alloy" and 

"TuO2+BeO"

AEC 1947; Grenell 1947; Randall 1947; 

Keeler and Hare 1954; Gallagher, 

Blosser, and Mann 1955, Vaughan 1957; 

Jankowski and Chastain 1958; 

Cunningham, Seaver, and Edgar 1959; 

Egen at al. 1960

Research on graphite-U piles 1947–? For the reactor at ORNL Author unknown ca. 1987; AEC 1947; 

Boehm and Groner 1986

Process development program for 

extraction of uranium from various 

uranium ores and concentrates

Aug 1947 Carnotite ores (Aug 1947–1949); 

washer slimes/slurries (1950); Western 

ores (1952–1955); low-grade sulfide 

concentrate (1953); U ores (ammonium 

carbonate leaching) (1955–1957)

Kelley 1947a,b, 1948b; Blatz 1949; 

Ewing et al. 1953; Bearse et al. 1953

AEC demonstrates use of radiation 

detectors

Dec 1947 To check U, Th metal, surfaces, 

clothing, etc. Battelle did not have such 

instruments before this date.

Morgan 1947

AEC contract for experimental studies 

on U- and Th-bearing materials, 

including monazite sands

1947–1950 Contract AT-30-1 GEN-228; laboratorylevel work started in Dec 1947Brown 1951; Kelley 1950, 1951; Blatz 

1949; Harris 1952; Kelley 1948b,c

Work for the Air Force's Aircraft 

Nuclear Propulsion Program

1947–1961 Wastren 2001; BCL no date a

Process development program for 

recovering U from phosphate materials

Jan 1948–

1949(?)

Development of practical processes for 

leaching and recovery of U

Beyer 1948a

Active AEC-Battelle contracts for U, Th 

extraction from monazite sands, 

phosphates, shales, and carnotite ore

1949–1950 All ongoing as of Apr 1949: Contract 

GEN-228: extraction of U, Th from 

monazite sands; ENG-27: U from 

phosphate rock; GEN-202: U from 

shales; GEN-258, U from carnotite

Blatz 1949;,Calkins et al. 1950; Calkins 

and Filbert 1950; Wilson et al. 1954

Nuclear submarine fuel element 

prototype development

1948 BCL no date a

Conversion of various forms to UO2 1950 Conversion of pitchblende to UO2

(1950–1951); refining MgX 

(concentrate), V-20 soda salt to UO2

(1950–1951) 

Wesner et al. 1960; Langston, Tangel, 

and Richardson 1950; Ewing, Kiehl, and 

Bearse 1950; Ewing et al. 1950a

R&D of large-batch reduction 1950–1955 Mostly defense-related or reactor 

development work; some processing 

done for government production 

reactors

BMI ca. 1987

ATTACHMENT A

CHRONOLOGY OF EVENTS AT BATTELLE

Page 1 of 5
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Event

Period/ 

start date Description Referencea

R&D and studies of U metal, its alloys, 

U carbides, U nitrides, etc.

1950–1962 Included making and testing fuel forms Schwope, Muehlenkamp, and Marsh 

1952

Battelle said to have provided 10 kg of 

uranium to Oak Ridge firm as ingots

1951 Presumably for Aircraft Nuclear 

Propulsion Program because firm was 

a major contractor for that program

Author unknown ca. 1987

Laboratory, pilot plant study of Th 

extraction from mantle-grade Th nitrate 

tetrahydrate Th(NO3)4)

Aug 1951–

Feb 1952

Solvent extraction pilot plant design by 

Catalytic Construction Co; pilot plant 

started in Jan 1952

Kelley 1951; Brown 1951; Harris 1952; 

Ewing at al. 1952

Completion of tracer studies for a CePr mixture1 Nov 1951 In connection with studies of Th 

extraction from Th nitrate

Kelley 1951

Process support for U refining, metal 

production

1952–1959 For AEC, Mallinckrodt, NLO. mostly 

R&D, process studies

BMI ca. 1987; Author unknown no date; 

Ewing, Kiehl, and Bearse 1950; Schwartz 

and Vaughan 1953; Rengstorff and 

Lownie 1955

Study of fused salt mixtures as liquid 

fuel

1953–1954 U salts Crooks, Snyder, and Clegg 1953; 

Droege, Snyder, and Filbert 1954

Studies of radioisotope tracers 1954 For industrial control Kemp et al. 1955

Apparent first use of enriched U at 

Battelle

1954 Rolling experiments for the APPR Keeler and Hare 1954; Gallagher, 

Blosser, and Mann 1955

King Avenue Buildings 2 through 7 and 

10 through 13 constructed

Completed 

1954–1955

Buildings A and 1 were already in 

existence

BMI 1977

Extraction of Th from Brazilian 

monazite sludge

1954–1955 Wallo 1981; Meeley, Snyder, and Filbert 

1954

Development of APPR 1954–late 

1960s

Wastren 2001; Gallagher, Blosser, and 

Mann 1955; Cunningham, Seaver, and 

Edgar 1959; Peters and Harrison 1998

Extraction of U from various ores and 

concentrates

1950–1957 Washer slimes/slurries (1950); 

Western ores (1952–1955); low-grade 

sulfide concentrate (1953); other U ores 

(ammonium carbonate leaching) 

(1955–1957)

Meadors et al. 1950; Ewing et al. 1950b; 

Bearse at al. 1953; Ewing, Kiehl, and 

Bearse 1955; Van Cleek, Macdonald, 

and Stephens 1956; Langston, 

Macdonald, and Stephens 1957

Start of operation of Bldg JN-1 1955 BCL ca. 1974; BCL no date a DOE 2002

Start of operation of Bldg JN-2 1955 Started as a critical assembly facility BMI ca. 1987; DOE 2000a; Layendecker 

1996; Wastren 2001; Jensen 2003; BCL 

no date a; Kirsch 2000a

Experiments, other work in Bldg JN-1 1955–1960 Mostly small-scale experiments on 

irradiated capsules

BMI ca. 1987; BCL ca. 1974

Work at the critical assembly facility 

(JN-2)

1955–1963 Critical assembly and vault work only Wastren 2001; DOE 2003a; Gallagher, 

Blosser, and Mann 1955; BCL no date a; 

Kirsch 2000a

Continued work in defense 1955–1965 More Naval Reactors research; work on 

water-cooled reactor and the Aircraft 

Nuclear Propulsion Program

BMI ca. 1987; BCL 1997

Subcontract weapons research for 

Sandia, etc

1955–1965 Also for Los Alamos, Rocky Flats, 

Lawrence Livermore

BMI ca. 1987

Postirradiation examination, testing of 

materials in JN-1

1955–1983 Including materials from BRR, other 

research, defense, and commercial 

power reactors

Startup of Bldg JN-3 (BRR) Oct 1956 DOE 2002; Wastren 2001; Anno, 

Plummer, and Chastain 1958; Plummer, 

Arno, and Chastain 1958; BCL no date a; 

Kirsch 2000b

BRR, ETR loop program work in JN-1 

and JN-3

1956–1962 Incl assay of fission gases during 

irradiation of fuel, other materials; BRR 

upgrade to 2 MW, in-pile loop in 1959

Bodnar et al. 1958; Basham and Rieder 

1960

ATTACHMENT A

CHRONOLOGY OF EVENTS AT BATTELLE

Page 2 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 64 of 124

Event

Period/ 

start date Description Referencea

Miscellaneous 

instrument/detector/dosimetry work

1956–1967 E.g., reactor neutron flux; leak 

detection; gamma, neutron effects on 

semiconductors; gamma, neutron 

dosimeters

Diethorne at al. 1959; Smith et al. 1956; 

Klickman et al. 1956; Moody, Kendall, 

and Willardson 1958; Howes, Ellerman, 

and Sunderman 1960; Hedden, Kircher, 

and King 1960; Farkas 1966; Kramer, 

Closser, and Mengali 1966; Kramer 1967

Evaluation and assays for recovery 

processes for spent fuel elements

1957–1964 Zircex, Darex, Thorex, Zirflex, fluoride 

volatility, Niflex processes

Ewing, Brugger, and Sunderman 1961; 

Peterson et al. 1959

BRR upgraded from 1 to 2 MW 1958 Anno, Plummer, and Chastain 1958

R&D on methods of measuring content 

in industrial materials, rates of 

biological/catalytic processes

1958–1963 E.g., Mg, Al, Ca, Fe in cement; 

investigated P-32 labeling, various 

radiotracers, activation analysis

Calkins and Pobereskin 1954; Dayton 

and Dickerson 1962; McFarling and 

Kircher 1963

Defense work in Bldg JN-1and JS-1

(Hot Isostatic Process Facility)

1958–1984 Naval reactors, R&D for Army and other 

defense programs

Wastren 2001

BRR upgraded from 1 to 2 MW Mar 1959 Plummer, Arno, and Chastain 1958; 

Anno, Plummer, and Chastain 1958

BRR beam tube shut down for furnace 

repair

Jun–Dec 

1960

Construction of Building JN-4 (old 

Plutonium Lab)

1960 Wastren 2001; Flynn et al. 1987

R&D: coating of U oxide, new fuel 

forms and materials, containment 

methods

1960s Author unknown ca. 1987; Chastain 1961

AEC license to fabricate U fuel 

elements

1960s Author unknown ca. 1987

Studies involving Pu properties, 

processing, and fabrication; Np target 

rod fabrication for Pu production; 

fabrication of RTGs using Pu and Cm

1960–1977 U recovery process for Pu spent fuel; 

Ta corrosion by liquid Pu alloys; UNPuN fuel materials, cladding for fast 

reactors; PuO2 and other fuel studies; 

Pu, Np recovery from commercially 

irradiated targets; Pu aerosols; 

fabrication of RTGs with Pu-238 (MHW) 

or Cm-244

Author unknown ca. 1987; FUSRAP 

2000a; Swigert 1963; Taylor 1983

Addition of alpha-gamma test cells to 

Bldg JN-1

1964 Wastren 2001; Martin, Storhok, and 

Gates 1966; BCL ca. 1974

Addition to JN-4 1964 Flynn at al. 1987

RTG work with Pm sources 1964–1968 Ritzman et al. 1966

King Avenue: continued defense 

program work

1965–1975 Included fabrication of highly enriched 

fuel elements

BMI ca. 1987

West Jefferson site: materials mockup 

experiments, advanced fuels research

1965–1975 The advanced fuels work was done 

with U nitrides and oxides.

BMI ca. 1987

Further addition to JN-4 1967 Flynn at al. 1987

Addition of Mechanical Test Cell to 

Bldg JN-1

Apr 1967 Wastren 2001, BCL ca. 1974

Research on bioeffects of underground 

explosions at Amchitka Island, Alaska

1967–1974 During/after the Milrow and Cannikin 

tests

Kirkwood 1971, 1974; Burgner, Isakson, 

and Lebednik 1971; Stephan and Mercier 

1972; Williamson and White 1974

JN-2 converted to a plutonium 

laboratory

Late 1960s–

1970

Work done for Lawrence Livermore 

National Laboratory; handled 

encapsulated Pu only. Vault still used

Wastren 2001; Kirsch 2000a

Work for NLO under prime AEC 

contract

~1970? Involved test quantities of radioactive 

metal

Author unknown no date

End of active nuclear work in Bldg JN-2 1970 Wastren 2001

Added High-Energy Cell to JN-1 1972–1973 Allowed for examination of full-sized 

fuel assemblies

DOE 1998; Wastren 2001; BCL ca. 1974, 

1997

Neutron radiography of reactor fuel 1971–1975 Done in reactor pool and an out-of-pool 

facility

Ray 1972

JN-2 converted to a plutonium 

laboratory

Late 1960s–

1970

Work done for Lawrence Livermore 

National Laboratory; handled 

encapsulated Pu only. Vault still used

Wastren 2001; Kirsch 2000a

ATTACHMENT A

CHRONOLOGY OF EVENTS AT BATTELLE

Page 3 of 5
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Event

Period/ 

start date Description Referencea

Production of Cf-252 sources 1972–

1980s?

Sources made from Cf wire in two 

alpha-gamma cells; unclear where wire 

came from or for whom work was done

Wastren 2001; Scotti and Martin 1972

Final shutdown of the Battelle 

Research Reactor

1974 DOE 2002; Layendecker 1996; BCL no 

date a; Kirsch 2000b

Added fuel pool, Washdown Room to 

JN-1

1975 Associated with the High Energy Cell BCL 1997

Partial decommissioning of the BRR to 

NRC criteria, with final survey

1975 Rubadue 2000; DOE 2002, 2003a; 

Layendecker 1996; Wastren 2001, BCL 

no date a; Kirsch 1975, 2000b; Kok 

1975a

Removal of last reactor fuel elements 

from BRR (JN-3), shipment to 

Savannah River Plant

Mar 1975 Layendecker 1996; Kok 1975b

Decontam of Critical Assembly Lab/Pu 

Lab (JN-2)

1975 Wastren 2001; BCL no date a

Work for various defense programs, 

Naval Reactors, DOD (Army, Air 

Force), and NASA

1975–1987 Source materials processed and 

fabricated for DOD activities

BMI ca. 1987

Visit by AEC, ANL personnel to assess 

the need for a full survey and 

decontamination

Jan 1976 FUSRAP ca. 1987

Battelle radiological survey of King 

Avenue facilities used for MED projects

1977 Found to be at background levels, 

except for Buildings A and 1

FUSRAP ca. 1987, 2000b

Decontamination of JN-4 and Buildings 

A and 1

1978–1982 Battelle itself performed the JN-4 

decontamination, using contractors for 

specialty work.

FUSRAP ca. 1987, 2000a; Wissinger 

1978a; Freas and Madia 1982; Rudolph, 

Kirsch, and Toy 1984; Clements 1981; 

Stellrecht, Freas, and Dettorre 1979; 

Madia et al. 1979; Flynn at al. 1987; 

BCLDP 1982

Cask sabotage program work in Bldg 

JN-1

1979–1983 In the pool and the Controlled Area; 

Myers et al. (1994) says 1981–1983.

Wastren 2001; Author unknown 2003; 

Myers et al. 1994; BCL 1997

Further radiation surveys 1980–1982 Probably performed by Battelle FUSRAP 2000a

Research on destruction of chemical 

agents using high-level gamma fields

1981–1982 JN-1: irradiating capsules of chemical 

agents using Co-60 sources, using a 

closed beam port (not in the hot cells)

Wastren 2001

Agreement to form the BCLDP 1984 Apparently not formally implemented 

until 1986

Rubadue 2000

End of nuclear research for DOE; 

formation of CEMP; formal start of 

BCLDP

1986 CEMP DOE 1998, 2002; Langsam and Carter 

1986

Removal of last reactor fuel elements 

from JN-1, shipment to INEL

1986 Wastren (2001) says 1987. Actual 

removal was apparently in 1986, 

license termination in 1987.

DOE 1998; Wastren 2001

Decontam of some King Ave areas by 

Battelle, S&M program for King 

Ave/West Jefferson areas with residual 

contamination

1986–1988 Buildings A and 1 were decontaminated 

under CEMP.

FUSRAP ca. 1987, 2000b; Langsam and 

Carter 1986; BCL 1987, 2002; Author 

unknown 1992

Termination of reactor license 1987 To possess and store fuel elements 

and similar materials

Layendecker 1996

End of use of Building JN-1, shipment 

of last of fuel in JN-1 pool

1987 BCL (no date) says 1988, but this 

appears to be in error.

DOE 1998; Peters 1998a

Preliminary characterization survey of 

15 buildings at the King Avenue, West 

Jefferson sites

1987 By Argonne National Laboratory Smith et al. 1988

D&D begins under the BCLDP 1989 Weaver at al. 2003; Wastren 2001

Verification surveys of Bldgs JS-1, JS10, and JS-11 by ORAU, with release 

by DOE/NRC

1989–1990 DOE 2002; ORAU 2006a; Wastren 2001

ATTACHMENT A

CHRONOLOGY OF EVENTS AT BATTELLE

Page 4 of 5
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Event

Period/ 

start date Description Referencea

Further King Avenue D&D (by 

Battelle?)

1990 Done under DOE’s Surplus Facilities 

Management Program

FUSRAP 2000b

Final verification survey of Bldg 7A 1993 ORAU 2006a

Final verification survey of Bldgs 3 and 

4

1995 ORAU 2006a

Final verification survey of Bldg 5 by 

ORAU

1996 ORAU 2006a

Final verification survey of Bldgs 1, 6, 7 

by ORAU

1997 ORAU 2006a

Completion of D&D of King Avenue site 1998 Except for Building 2? See 1999 entry 

below.

Wastren 2001

Beginning of D&D of Bldg JN-1 1998 Weaver at al. 2003

Final verification survey of Bldg 2 by 

ORAU

1999 ORAU 2006a

Emptying, removal of the JN-1 alphagamma cells1999–2000 DOE 2002

Completion of remediation of King 

Avenue XP pad

Oct 2000 The pad was unrelated to D&D 

activities (BCLDP), but full RCRA 

closure was contingent on closure of 

the pad.

BCL 2002

Completion of D&D of King Avenue 

facilities, final verification survey of 

Bldg A by ORAU

2000 DOE 2002, ORAU 2006a

Removal of the BRR bioshield FY 2001 DOE 2002

Nuclear support, surveillance/ 

maintenance activities; some decon 

conducted at West Jefferson

2001 JN-1, JN-2, and JN-3 were involved. BCL 2002

Completion of RCRA closure 

requirements for the King Avenue site

25 Jul 

2001?

BCL 2002

Surveys of JN-2 and JN-3 by ORAU 2004 ORAU 2006b

a. In addition to the references listed, information was obtained from site descriptions and abstracts given on the OSTI Web site for various 

other Battelle reports.

ATTACHMENT A

CHRONOLOGY OF EVENTS AT BATTELLE

Page 5 of 5
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6/2010 Page 67 of 124

Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

King Avenue site

A Entrance was through 

Bldg 4 or Bldg 11 (BMI 

1977)

Thorium solvent extraction pilot plant (FUSRAP ca. 1987, Jensen 2003), 

ore processing (including carnotite), metallurgical studies (FUSRAP ca. 

1987); encapsulation of highly enriched uranium for ATR fuel elements 

(pilot plant) (Jensen 2003)

First part erected in 1929 as “the King 

Avenue Building” (Boehm and Groner 

1986); added onto before World War II 

(Boehm and Groner 1986); corporate 

offices (Langsam and Carter 1986, 

Jensen 2003). The Safety Office 

(including the health physics base of 

operations) was here in at least 1962 

(Sunderman and Dickerson 1962)

1942–1986

E basement, E 

wing

53 through 57A Not clear that this area was involved in MED projects; Jensen (2003) 

lists only the 1st and 4th floors

Creep Laboratory (BMI 1977)

Basement, E wing, 

S end

58A through 58L, 67 Ore processing, monazite sand solvent extraction After MED/AEC projects, equipment 

dismantled, removed (BMI 1977). 

Unoccupied in 1977 (BMI 1977). Was 

Laser Lab for several years. Residual 

activity found in Room 58 complex in 

late 1980s (FUSRAP ca. 1987)

1st floor Ore processing, monazite sand solvent extraction (Jensen 2003)

S end of 1st floor 

in E wing

163 through 163F Upper part of high bay area of 58 complex Offices

NW basement of 

W wing

6, 6A through 6D, 7 Machine shop operations (BMI 1977) Converted to library facility by 1977 

(BMI 1977)

North basement, S 

of E-W corridor

17 Mechanical Test area (BMI 1977) Laboratory record storage

2nd floor Laboratory used for tracer work (Sunderman and Dickerson 1962)

N end, 2nd floor, S 

of E-W corridor

219, 219C, 219D Metallography sample examination (BMI 1977), storage Reprint and brochure distribution offices

4th floor Pilot plant encapsulation of HEU for ATR fuel elements (Jensen 2003, 

DOE 1996), presumably in the Encapsulation Laboratory; Nuclear & Flow 

Systems Shop, Room A-467, supported the Encapsulation Laboratory 

(Browne 1974). Tracer-level lab (<100 μCi) also located on this floor 

(Sunderman and Dickerson 1962)

Little radioactive material handled 

(Browne 1974). The cafeteria was 

located on the 4th floor also (Boehm 

and Groner 1986)

1 Foundry Foundry (Langsam and Carter 1986, Jensen 2003); melting, cutting, 

grinding facilities (Jensen 2003). Natural U, Th extraction (FUSRAP ca. 

1987, Jensen 2003); ore beneficiation studies (Jensen 2003)

Constructed in about 1937. Source 

material fabrication for DOD, 1975 on 

(BMI ca. 1987)

1942–1986

1st floor (the) Foundry Uranium processing (BMI 1977; DOE 1996) in the tons (Jensen 2003), 

ore beneficiation studies (Jensen 2003)

High bay at S end 

of 2nd floor, 

adjacent area

1223, 1223A, 1223B Uranium production from ore (BMI 1977, DOE 1996); beneficiation 

operations with ores such as carnotite (BMI 1977). Misc physical handling 

of materials in 1223; equipment storage in 1223A, misc lab work in 

1223B

Room 1223 complex found to contain 

residual activity in ~1987 (FUSRAP ca. 

1987), after cleanup

?–1986

Near 2nd floor 

high bay area

1219 and 1219A Ore processing Ore lab

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

Page 1 of 8
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/2

6/2010 Page 68 of 124

Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

4th floor Laboratory used for tracer-level (<100 μCi) work (Sunderman and 

Dickerson 1962)

2 Metalworking Building 

(Langsam and Carter 

1986, Jensen 2003); 

“Tisdale Building”

AEC research, incl electroplating Hanford reactor slugs (Jensen 2003, 

DOE 1996); heat treatment, fabrication of U, Th alloys (Jensen 2003); 

fabrication of U, Th slugs (DOE 1996). Also rolling studies, alloy 

development, and fuel element fabrication (Jensen 2003).

Metal fabricating machinery (hydraulic 

presses, rolling mills, furnaces) removed 

from the building in 1993 (Jensen 2003)

Metalworking Lab Ore processing (Jensen 2003 says “tons”); metalworking (Jensen 2003)

Welding Lab Natural/depleted/enriched U welding and fabrication (Jensen 2003)

2nd floor Misc R&D, heat treatment, electroplating (Jensen 2003) 1954–1986

3 U-235 Processing 

Bldg; Materials Bldg 

(Langsam and Carter 

1986, Jensen 2003)

Research activities using natural, depleted, enriched uranium (Jensen 

2003, DOE 1996). Powder metallurgy, melting, metallographic, ceramics 

research facilities (Jensen 2003). Misc R&D: ceramics research; 

metallographic research with tons of source material (Jensen 2003)

Used as an SNM storage and handling 

point since early times (Carr 1993). 

Source material fabrication for DOD, 

1975 on (BMI ca. 1987).

1st floor Room 3161 complex, 

the “U-235 area”; “25 

Area/Room”

U-235 processing facility; receiving/distribution point for unirradiated 

sources, SNM, incl enriched material and thorium (BCL 1975, Jensen 

2003, Langsam and Carter 1986, Carr 1993); also for occasional special 

samples (Stickel 1993). Pickling of SNM to remove oxide for use in Bldg 

3, other Battelle areas; solidification of pickling solutions by evaporation; 

salts stored here for later recovery (Carr 1993). U reduction by calcining 

for reprocessing; hydrides, carbides, nitrides also made for use 

elsewhere at Battelle; packaging, storing of wastes from other areas 

(Carr 1993). Metallography lab, ceramics research facility (DOE 1996)

Glovebox, hoods, shower, calciner, 

cutting wheel, prefilter, HEPA filter (Carr 

1993). Decommissioning activities 

completed in 1994, building returned to 

Battelle

1st floor Room 3110 Storage area (AEC 1950) 1954–1975

2nd floor or S end 

of 1st floor

Melting Lab/Facil, 

Room 3178

Melting Lab/facility (Langsam and Carter 1986, BCL 1975, Jensen 2003, 

Kizer 1993). U-235 materials used, 954-1975; tons of source material 

(Jensen 2003). Metallographic research (Jensen 2003)

Had hoods

N end of basement Powder Metallurgy Lab Powder Metallurgy Lab (Jensen 2003, DOE 1996, BCL 1975). Natural U 

storage, source processing (Jensen 2003, BCL 1975)

1954–1986

4 1st floor Radiochemistry 

(Radioisotope) 

Laboratory (Langsam 

and Carter 1986, 

Jensen 2003, 

Sunderman and Gates 

1965)

Entrance to Bldg A was through Bldg 4 (BMI 1977). Radiochemical 

procedures, radiotracer studies, industrial applications of radioisotopes, 

neutron dosimetry, other ops with radioactive materials (Sunderman and 

Gates 1965). Radioisotope Laboratory main facility was divided into lowlevel, medium-level, and high-level rooms; had fission gas laboratory, 

special radiochemistry hoods, junior cave, gloveboxes, shielded 

operating and storage areas, vault, counting room, offices, waste 

disposal facility (Peters 1998a, Sunderman and Dickerson 1962, 

Sunderman and Gates 1965). Also had Radiography Lab with X-ray 

machine, Cs-137 source (Selander 1958c)

The main facilities in Building 4 and the 

associated facilities in Buildings A and 1 

were together called the "Radioisotope 

Laboratory" (Sunderman and Gates 

1965). Hot material transported in cask 

by truck between JN-1 and Building 4 

(High Level Lab in the Radiochemistry 

Lab) (Sunderman and Dickerson 1962). 

Decommissioning completed in 1994

2nd, 4th floors Misc labs on 2nd, 4th floors, incl encapsulation facility for highly enriched 

U, metallography facilities (Jensen 2003, DOE 1996). Had Tribology 

Shop in Rooms 4022 and 2023 during some period before 1994

1954–1986

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

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Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

5 Machine Shop Work for AEC/ERDA/DOE programs: machining (Jensen 2003, DOE 

1996), grinding (Jensen 2003, DOE 1996, Dosim-Serio), milling (Jensen 

2003, DOE 1996). Be machining (Jensen 2003)

Source material fab for DOD, 1975 on 

(BMI ca. 1987). D&D completed in 1995 

(DOE 1996)

Machine shop (Jensen 

2003; DOE 1996; 

Langsam and Carter 

1986)

Machining of depleted, natural, and enriched U and some Th (Jensen 

2003)

Part of the floor of the machine shop 

was replaced at some point to remove 

contamination

Coating Lab Coating (of materials) (Jensen 2003) Years of 

operation

Misc R&D on 1st 

and 2nd floors

1954–1986

6 Misc R&D on 

ground and 1st 

floors

(one) Chemistry 

Building (Langsam and 

Carter 1986, Jensen 

2003)

Analytical chemistry activities in support of the DOE/Navy program, 

including instrumental analyses (Jensen 2003). Alloy studies, chemical 

analyses, and Co-60 irradiation (Jensen 2003)

Ground floor decontamination was 

completed in 1993 (DOE 1996)

1st floor Gamma Laboratory Gamma Facility Co-60 source was 1600 Ci in Feb 1962 (Sunderman and 

Dickerson 1962) or 2100 Ci in Nov 1965 (Sunderman and Gates 1965). 

Source located in belowground pool with 13 ft of water over source, 

stored in lead pig on the floor of the pool (Sunderman and Dickerson 

1962, Sunderman and Gates 1965). 

Source rods were in use from about 

November 1959 on (Sunderman and 

Dickerson 1962)

Ten years 

during 1954–

1986 (?)

7 Misc R&D on 1st, 

2nd, and 3rd floors

(another) Chemistry 

Building (Langsam and 

Carter 1986, Jensen 

2003)

Analytical chemistry activities in support of DOE/Navy program, incl 

instrumental analyses (Jensen 2003), over a (unspecified) period of ten 

years (DOE 1996). Analytical chemistry on source materials and 

corrosion studies (Jensen 2003)

Annex (Building 7A) was 

decontaminated, released to Battelle in 

1993 (DOE 1996); the significant 

radioactive operation in 7A was a 

uranium fluoridation pilot plant (Basham 

1993)

1954–1986

9 Misc R&D on 

ground floor

Mechanical 

Engineering Building 

(Langsam and Carter 

1986, Jensen 2003)

Research programs for AEC/ERDA/DOE: handling natural and depleted 

uranium, krypton studies (Jensen 2003)

Hoods in lab area. The building was 

decontaminated and released to 

Battelle in 1991 (DOE 1996)

1954–1986

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

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Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

West Jefferson site 1955–1986

JN-1 Hot Cell Facility/ 

Laboratory; JN-1A was 

the original bldg, 

added to in the 1960s; 

JN-1B was major 

addition of 1970-1971

Fuel devel research support for AEC (Jensen 2003). Spent fuel 

examination; postirradiation examination, studies of fuel specimens, 

reactor components, other mats/structures; source encapsulation; 

shipping container sabotage (BCL 2002, DOE 1996, Rubadue 2000, 

Jensen 2003); some criticality experiments (Jensen 2003); later, hot cell 

examination of fuel from commercial power reactors "in support of DOE 

programs." All cells had remote manipulation, several operating stations, 

each with lead-glass oil-filled viewing window (DOE 2002). Support 

facilities: cask handling, solid and liquid waste disposal, contam control, 

equipment decontamination, ventilation mezzanines (DOE 2002, 2003a; 

Norris 1961; Sunderman and Dickerson 1962)

Extensively modified in 1960 and 1979 

(DOE 2003a). All cells in JN-1 have 

very high residual radiation fields (DOE 

1998, 2000a). Water in the fuel transfer 

pool was removed in 1996 and 

packaged TRU waste was then stored 

there (DOE 2002). Main offices were on 

the first floor near the laboratories 

(Wastren 2001).

1955–1986

Section JN-1A, 1st 

floor

Controlled Access 

Area (CAA), with small 

spent fuel pool

Support area for hot cell work: provided access to the cells through their 

doors, facilities for moving large items into cells, and space for special 

projects (DOE 2003a). Heat treatment furnace in CAA (Selander 1959c); 

ventilated cubicles for work with small shielded specimens (Sunderman 

and Dickerson 1962); small hot room with pit, sinks, and hot drain for 

equipment assembly/decon, manipulator repair bench/glovebox 

(Sunderman and Dickerson 1962, Myers et al. 1994b). Enclosed drum 

compactor (Myers et al. 1994b). Spent fuel pool used to 1972 (Wastren 

2001, Myers et al. 1994b). Sabotage Program (1981–1983): apparatus 

bolted on top of spent fuel pool, shaped charge shot at small model cask; 

gases collected, analyzed (Wastren 2001, Myers et al. 1994b)

1955–1986

Section JN-1A, 

ground and 1st 

floor (cells), 

basement 

(subcells)

Two large cells with 

their subcells: High 

Level Cell, Low Level 

Cell

The original two large cells (DOE 1998, 2000a, 2002, 2003a; Langsam 

and Carter 1986; Rubadue 2000) designed to provide shielding for 

10,000 Ci and 10 million Ci of 1-MeV gamma emitter, respectively 

(Wastren 2001, BCL no date b, Sunderman and Dickerson 1962). Highdensity concrete walls (ferrophosphorus aggregate at 293 pcf) up to top 

of the manipulators, ordinary concrete after that (DOE 2003a). Three 

operating stations each (BCL 1977a, Sunderman and Dickerson 1962). 

Metmounts cut and ground in the LLC (Wastren 2001), which was also 

used for tensile tests of Co-60 samples (Author unknown 2003). LLC 

also used near end of operations to gamma-scan cans of waste (Myers at 

al. 1994c). Subcells used for creep testing (Author unknown 2003, 

Sunderman and Dickerson 1962)

The LLC was used to declad fuel, 

section Pu-containing fuel rods, and do 

tensile tests on irradiated cobalt 

specimens (Myers at al. 1994c). The 

High Level Cell's primary purpose was 

to cut irradiated fuel sections and defuel 

fuel rod sections (Myers at al. 1994d), 

i.e., the most destructive activities (BCL 

1997). Irradiated Co-60 rods were also 

cut up here (Myers at al. 1994d, BCL 

1997)

Years of 

operation

1955–1986

Section JN-1A, 1st 

floor

Charpy Cell/Room; 

Dry/Radioactive 

Storage Area

Shear test area for nonfuel specimens, using Charpy apparatus (Myers, 

Redd, and Berchtold 1994a); 74 vertical pipe wells in floor, additional 12 

capped storage holes in movable concrete block for canned specimens 

(Wastren 2001, DOE 2003a, Sunderman and Dickerson 1962, Myers, 

Berchtold, and Sands 1995, Myers, Redd, and Berchtold 1994a); rad 

materials stored in steel cans in wells, transported to/from other cells in 

portable lead shield (Sunderman and Dickerson 1962)

Charpy apparatus removed and sent to 

King Avenue in 1988-1989 (Myers, 

Redd, and Berchtold 1994a)

1955–1986

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

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Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

Section JN-1A, 1st 

floor

Transuranic vault Storage 1955–1986

Section JN-1A, 1st 

floor

Evaporator Room Open evaporator tank with exhaust hood and fill pipe coming from 

Evaporator Storage Tank; sludge periodically put into drums (Myers, 

Berchtold, and Tomlin 1994)

1961–1986

Section JN-1A, 1st 

floor

Laboratories Chemistry Laboratory: two hoods, hot sink, hot drain to sump. Counting 

Room: X-ray diffraction, gamma/alpha spectroscopy, radiochemistry 

studies; hot sinks, hood, cave, hot drain to sump. Microprobe Room: 

microprobe equipment, Cs-137 source; isotopic analyses of Metmount 

samples, using X-ray diffraction. (Wastren 2001, Myers et al. 1994e)

1964–1986

Section JN-1A, 1st 

floor

East machine shop (Wastren 2001, Myers et al. 1994e) 1964–1986

Section JN-1A, 

basement

10 small segmented 

alpha-gamma cells 

(Wastren 2001, DOE 

2003a, Rubadue 2000, 

DOE 2002)

Cells 1-9 constructed in 1964 for metallography testing of fuel rod 

specimens; Cell 10 added in early 1970s (Wastren 2001, BCL ca. 1974, 

Myers et al. 1994f). Shared working/access area (BCL no date b), 

individual working box in each cell (Myers et al. 1994f). Cells 1, 2, 3, 10 

for metallography operations: Metmounts ground down in Cell 3; 

polished, washed, acid-etched in Cell 2; hardness-tested, photographed 

with Metallograph in Cell 1. Cell 4 used to port materials to/from CAA, 

other cells; never used for project work. Cells 5, 6 used only to prepare 

Cf-252 sources, mid-1970s on: Cf wire cut, welded into containers, leaktested. Cell 7 used for dissolution, slug solidification of burnup fuel 

specimens. Cell 8: unclad fuel samples heated in furnace, exposed to 

laser beam/detector for analysis. Cell 9: X-ray diffraction testing of 

Metmounts (e.g., Myers, Berchtold, and Sands 1995, Myers et al. 1994f, 

Sunderman and Gates 1965, BCL 1997). Shielding added to wall

between working area and evaporator room, 1964 (Gates 1964)

For handling irradiated fuel 

metallography specimens (BCL no date 

b, 1997). When Cell 10 was built, Cell 1 

equipment abandoned. Cell 3 was the 

most contaminated alpha-gamma cell 

because Metmounts were ground there; 

two in-cell grinders were the most 

contaminated equipment (Wastren 

2001, Myers, Berchtold, and Sands 

1995). Some spills in Cell 5 (Myers et 

al. 1994f). Fumes from acid-fuel 

solutions in Cell 7 contaminated the rear 

of Cells 7, 8 via a faulty connection 

(Wastren 2001, Myers, Berchtold, and 

Sands 1995).

1967–1986

Section JN-1A, 1st 

floor

Mechanical Test Cell Main mechanical test cell (Rubadue 2000, DOE 1998, DOE 2003a, 

Myers, Redd, and Berchtold 1994b) added in 1967 (BCL ca. 1974). 

Three operating stations (BCL no date b). High-density concrete front 

wall(s) (ferro-phosphorus aggregate at 293 pcf) to the top of the 

manipulators, ordinary concrete above that (DOE 2003a, DeMastry, Lusk, 

and Gates 1967). Unspecified thickness of iron plate belatedly added to 

one wall for shielding reasons (DOE 2003a)

Mainly for tensile tests, but also for 

creep, vacuum fusion, burst, radial 

burnup, density, and expanding mandrel 

tests (Myers, Redd, and Berchtold 

1994b, BCL 1997)

Years of 

operation

Section JN-1B, 1st 

floor

Cask Washdown 

Room

For cleaning newly arrived casks, etc. (DOE 2003a, Myers et al. 1994b, 

Myers, Berchtold, and Stickel 1995); also entry-exit area for High-Energy 

Cell (DOE 2003a, Myers, Berchtold, and Stickel 1995). Walls open at top 

for crane access (DOE 2003a, Myers, Berchtold, and Stickel 1995). 

Compactor under hood (Myers, Berchtold, and Sands 1995)

1970–1986

Section JN-1B, 1st 

floor

High Bay/ Operations 

Floor

Connected to the High Energy Cell, the Fuel Storage Pool, and 

associated facilities; designed for handling large casks (DOE 2003a)

1970–1986

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

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Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

Section JN-1B, 1st 

floor

Fuel Storage Pool Storage of fuel assemblies, rod bundles, rod holders, tools (DOE 2003a, 

Myers, Berchtold, and Stickel 1995). Inside and outside sumps to collect 

any pool leakage (Myers, Berchtold, and Stickel 1995). Cask brought in, 

lowered into pool; fuel assemblies or bundles taken out, transferred to 

High Energy Cell through fuel channel running from pool to under the 

High Energy Cell floor (Jensen 2003)

Section JN-1B, 1st 

floor

Pump Room Held wastewater storage tank, deionized water tank, and ion exchange 

tanks used to process the liquid radwaste (Myers, Berchtold, and Sands 

1995, Myers, Berchtold, and Stickel 1995)

1972–1986

Section JN-1B, 1st 

floor

High Energy Cell Largest hot cell (DOE 2003a, DOE 1998, DOE 2002), five operating 

stations (BCL no date b). Power plant fuel element testing, eval (DOE 

1998). Access via door, ceiling plugs, or floor trap door to transfer canal 

(DOE 2003a), then to fuel transfer pool (Rubadue 2000, DOE 2002). 7 

storage holes in floor (DOE 2003a). Two walls barytes (at 220 pcf), two 

ordinary concrete (DOE 2003a). Cask washdown room at entryway (BCL 

no date b, Myers et al. 1994b). Had beam port for Co-60 source inside 

the cell to irradiate test instruments (Myers et al. 1994b), some storage 

holes in floor (Myers et al. 1994b)

Designed to handle entire high-burnup 

fuel assembly (BCL no date b, 1997); 

nondestructive inspection and testing 

(BCL 1997); sectioning of rods (BCL 

1997); fission gas measuring tests done 

here (Myers et al. 1994b, BCL 1997); 

studies and characterization of TMI 

resins (BCL 1997); effects of Co-60 on 

instrumentation (BCL 1997)

Section JN-1B, 1st 

floor

Old Back Dock, Waste 

Control Area

In early years, used to collect and package waste and as the original JN1 dock; later, used for storage of a large number of highly contaminated 

items (Myers at al. 1994g)

1970–1986

Attached to the 

JN-1B section

Waste Storage Shed, 

Sheep Shed, 

Equipment Storage 

Room

Storage of low-level contaminated containerized waste, equipment, other 

material (DOE 2003a,b; Myers and Berchtold 1995). Two internal 

rooms/areas were heavily shielded (Myers and Berchtold 1995)

With JN-1, but not 

attached

Waste Management 

Shed

Storage of highly contaminated containerized waste, equipment, other 

material (DOE 2003a,b)

JN-2: 1955–

1963; 1964–

1970. Vault 

to at least 

1996

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

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Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-2 Critical Assembly 

Laboratory, Critical 

Assembly Facility, 

Critical Assembly 

Building; later also a 

Plutonium Laboratory

Started operation in 1955 (Wastren 2001). 1956–1963: critical assembly 

studies (DOE 2003a, DOE 2000a, Chamberlain ca. 2003, Jankowski et 

al. 1957, Jankowski and Chastain 1958, Egen at al. 1960); zero-power 

organic-moderated critical assembly and other experiments (Jensen 

2003). 1964(?)–1970, used for other nuclear projects, including direct 

conversion concepts/plutonium laboratory (DOE 2000a, 2002, 2003a; 

Jensen 2003; Chamberlain ca. 2003; Kirsch 2000a), irradiation 

experiment assembly (DOE 2003a, Chamberlain ca. 2003), special 

laboratory (Jensen 2003). Had Pu, enriched U storage vault (DOE 

2000a, 2002, 2003a; Jensen 2003; Rubadue 2000; Freas et al. 1971; 

Evans and Woodward 1979; BCL 1977a); control room (Jankowski et al. 

1957, Hogan et al. 1958, Jankowski and Chastain 1958); attached highbay area (DOE 2000a, 2002; Rubadue 2000), machine shop (Hogan et 

al. 1958), offices, labs. Also a fence around the building (separate from 

the site fence) (Hogan et al. 1958)

Built to accommodate zero/low-power 

mockups of prototype reactors (Kirsch 

2000a). Later, a lab installed for work 

with encapsulated Pu for Lawrence 

Livermore Laboratory. 1970: end of 

active nuclear work (Chamberlain ca. 

2003, Wastren 2001). 1975: first 

decontamination (Wastren 2001). Was 

Admin Bldg, had Accountability Lab to 

~1987 (Langsam and Carter 1986). 

RAL in decontaminated former Pu Lab, 

1978 to ~2003 (FUSRAP 2000a; 

Rubadue 2000; DOE 1996,2000a, 

2002, 2003a; Chamberlain ca. 2003; 

Wastren 2001; Jensen 2003); carpenter 

shop in bay moved from JN-3, from 

~1987 on (Langsam and Carter 1986). 

Vault still used in 1996 or later (DOE 

1996). "Minimally contaminated" at final 

decontamination (DOE 1996, 2000a; 

Jensen 2003)

Years of 

operation

1956–1974

JN-3 BRR Building Pool-type research reactor (Jensen 2003); operation began in October 

1956 at 1 MW (Anno, Plummer, and Chastain 1958), was increased to 2 

MW in March 1959 (Plummer, Arno, and Chastain 1958), ended in 

December 1974 (DOE 1996). Shielding facility located at end of thermal 

column (Morgan et al. 1958, Plummer, Arno, and Chastain 1958, Myers 

and Stickel 1995); gas loop facility added 1958 (Bodnar et al. 1958; 

Plummer, Anno, and Chastain 1958, Basham and Rieder 1960, Anno and 

Plummer 1962); fast neutron activation facility added ~1960 (Jung et al. 

1960). Carpenter shop (Langsam and Carter 1986); health physics office 

and instrument laboratory (Langsam and Carter 1986). JN-3 addition: 

well/hole storage in 144 wells/holes, primarily for spent fuel assemblies 

(Myers and Stickel 1995)

Partly decommissioned in 1975 

(Rubadue 2000, DOE 2002, Jensen 

2003, DOE 1996). Carpenter shop, 

Health Physics and Instrument 

Laboratory to ~1987 (Langsam and 

Carter 1986). 1987 on: temporary 

storage of TRU wastes (DOE 2002, 

Jensen 2003), for which a special waste 

storage area was built in the basement 

(Myers and Stickel 1995)

1960–1977 

(FUSRAP 

2000a, 

Rudolph, 

Kirsch, and 

Toy 1984)

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

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Bldg ID

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-4 Main Plutonium 

Laboratory ("old" 

laboratory section and 

its "new" additions)

Facility used in part for production of U-Pu nitride pellets, which were 

packed into fuel pins (Peters and Smith 2000). Development of actinide 

fabrication and joining techniques (Smith 2000). Had Metallography 

Laboratory (BCL 1977a). Room 4118 was used for fabrication of Pu-238 

heat sources (FUSRAP 2000a, Smith 2000). Backup fac for LLL 

weapons assembly (Author unknown 2003, Peters and Smith 2000). 

Some wet chemistry in a glovebox for Allied Chemical Nuclear Services 

project (Smith 2000); some nuclear weapons, test shot (cask sabotage) 

projects (Author unknown 2003). Pu shipped to Battelle was handled 

here (Author unknown 2003, BCL 1977a), initially in the Downdraft Room 

(BCL 1977a).

Support of Pu research, processing 

(Wastren 2001, Freas et al. 1971, 

Rudolph, Kirsch, and Toy 1984, BCL 

1977a); added to in 1964, 1967 (Freas 

et al. 1971, Rudolph, Kirsch, and Toy 

1984, BCL 1977a). ~50% of work was 

Pu alloy R&D to support weapons 

development for Lawrence Livermore, 

Rocky Flats, Army reactors (Author 

unknown 2003, Smith 2000). Two fume 

hoods; 43 gloveboxes in 1971, 28 in 

1978 (Wastren 2001, Freas and Madia 

1982, Rudolph, Kirsch, and Toy 1984). 

Presses, furnaces, grinders, lathes 

(Clements 1981). Old section removed 

by 1982 (BCLDP 1982)

1976–present

JN-6 (New) Guardhouse Security operations (guardhouse)(Rubadue 2000, DOE 2003a) 1965–1975?

JS-1 Used to fabricate military reactor fuel by hot isostatic pressure bonding 

technique (DOE 1996, Boehm and Groner 1986). Hot Isostatic 

Processing Facility (BMI ca. 1987, Langsam and Carter 1986); explosive 

containment (Langsam and Carter 1986); Ballistics Facility (Langsam and 

Carter 1986)

Decommissioned in 1990 (DOE 1996); 

D&D was complete by 2003 ()

?–1988

JS-10 Fabrication of uranium components using explosive forming techniques 

and ballistic experiments with depleted uranium (DOE 1996, Boehm and 

Groner 1986)

Facility characterization in 1998 showed 

that no D&D was necessary (DOE 

1996)

?–1989

JS-12 Operations with radioactive materials in the ballistic tunnel, ductwork, and 

hardware using depleted uranium (DOE 1996, Boehm and Groner 1986)

Decommissioned in 1989 (DOE 1996)

a. An old guardhouse building (with no number) built in 1956 was apparently abandoned when JN-6 was built (DOE 2003b).

ATTACHMENT B

BUILDINGS AND OTHER FACILITIES AT BATTELLE

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Bldga Start End Isotopesb Formc Type(s) of workd

King Avenue Site

A U, Th, 

daughters

DU, NU, EU, Th 6.4 MT (Author unknown 1992) (apparently the total ever worked with in 

this building)

1943

1956

1944;

1956

U, daughters Uranate; U3O8; 

U2O2(NH4)2

Metallurgical R&D, studies of uranates

1943

1950

1958

1972

1944;

1952;

1961; 

1975

U, daughters U carbides, mainly 

UC (nat; enriched 

9-93%)

R&D, studies of U carbides and their alloys

1943

1979

1962;

1979

U, daughters U metal, alloys R&D, studies of work on U metal and its alloys

1946 1957 U, daughters U3O8, UO2, 

(NH4)2U2O7, 

Na2U2O7, other 

forms (UO3?)

U chemical extraction: phosphate rock (1946-1950); shales (1946-1949,

1952-1954); washer slimes/slurries (1950); carnotite ores (1947-1949); 

Western ores (1952-1955); low-grade sulfide concentrate (1953); U ores 

(ammonium carbonate leaching) (1955-1957). Converting pitchblende 

to UO2; refining MgX (concentrate), V-20 soda salt to UO2 (1950-1951)

1950 1971 U, daughters UO2 (nat, depl) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, 

etc.) (a few plates: depleted U)

1953 1954 U, daughters U salt solutions 

(see at right)

Study of fused salt mixtures as liquid fuel: U fluorides, phosphates, 

sulfates, hydrosulfides; solubility of U in fused salts, O2 in uranyl sulfate 

solutions

1953 1959 U, daughters U metal, UO2, UO3, 

UNH

Process support for U metal production for Mallinckrodt, NLO

1953 1959 U, daughters UF4 Process support for U metal production (some work involved UF4); study 

of UF4 as fused salt mixture for liquid fuel (1953-1954)

1956 1959 U, daughters U compounds (see 

at right)

Study of U compounds: U beryllides, borides, silicides, sulfides

1957 1958 U, daughters UCl Study of UCl dispersions (short period only)

1957 1961 U, daughters U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

R&D, studies of U nitrides and their alloys

1958 1960 U, daughters UO2 (enr, mostly 

93%)

R&D on UO2 for fuel of various types (powder, pellet, plate-type, etc.); 

enrichment from few percent to full; bulk of work apparently with 93% 

enr

1944 1961 Th, daughters Th metal, Th alloys Work with Th compounds, alloys (except extraction): R&D with Th metal 

(1944-1946); forging, rolling, machining Th ingots into Hanford slugs 

(1947)

1951 1955 Th, daughters Th(NO3)4, other 

forms

Th extraction from monazite sands (1949-1950). Solvent extraction to 

purify Th(NO3)4 (1951-1952). Extraction from Brazilian monazite sludge 

(1954-1955)

1 U, Th DU, NU, EU, Th 11.6 MT (Author unknown 1992)

1943

1950

1956

1944;

1952;

1962

U, daughters U carbides, mainly 

UC (nat; enr 9-

93%)

R&D, studies of U carbides and their alloys

1943

1956

1944;

1956

U, daughters Uranate; U3O8; 

U2O2(NH4)2

Metallurgical R&D, studies of uranates

1943 1962 U, daughters U metal, alloys R&D, studies of work on U metal and its alloys

1946 1957 U, daughters U3O8, UO2, 

(NH4)2U2O7, 

Na2U2O7, other 

forms (UO3?)

U chemical extraction from phosphate rock (1946-1950); shales (1946-

1949, 1952-1954); washer slimes/slurries (1950); carnotite ores (1948-

1949); Western ores (1952-1955); low-grade sulfide concentrate (1953); 

U ores (ammonium carbonate leaching) (1955-1957). Converting 

pitchblende to UO2 (1950-1951); refining MgX (concentrate), V-20 soda 

salt to UO2 (1950-1951)

1950 1965 U, daughters UO2 (nat, depl) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, 

etc.) (a few plates: depleted U)

1952 1958 U, daughters U metal, UO2, UO3, 

UNH

Process support for U metal production for MED/AEC, Mallinckrodt, NLO

1952 1958 U, daughters UF4 Process support for U metal production (some work involved UF4); study 

of fused salt mixtures as liquid fuel (1953-1954)

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

Page 1 of 9
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Bldga Start End Isotopesb Formc Type(s) of workd

1953 1954 U, daughters U salt solutions 

(see at right)

Study of fused salt mixtures as liquid fuel: U fluorides, phosphates, 

sulfates, hydrosulfides; solubility of U in fused salts, O2 in uranyl sulfate 

solutions

1956 1959 U, daughters U compounds (see 

at right)

Study of U compounds: U beryllides, borides, silicides, sulfides

1956 1962 U, daughters UO2 (enr, mostly at 

93%)

R&D on UO2 for fuel of various types (powder, pellet, plate-type, etc.); 

enrichment from few percent to full; bulk of work apparently with 93% 

enrichment

1957 1961 U, daughters U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

R&D, studies of U nitrides and their alloys

1944 1961 Th, daughters; 

minor U

Th metal, Th 

alloys; Th 

phosphate, halide, 

fluorides, carbides; 

ThN

Th compounds, alloys (except extraction): R&D with Th metal (1944-

1946); forming Th ingots into Hanford slugs (1947); Th hydriding (1951); 

electroplating metals on Th (with Th phosphate) (1948-1953); arcwelding Ames Th, Th alloys; metallurg/mech study of Th; Th prep by 

iodide process (1951-1954); bomb reduction of Th halides; Th ingot prod 

by arc melting, casting (with Th fluorides) (e.g., 1955); Th-5, -10, -15, -

20 w/o U alloys; Th-U alloy made with thorium iodide; ingots cast by 

vacuum induction melting to Th buttons, casting Th-U carbides by arcmelting; ThN prep (1956-1961). Dissolution of Al-canned Th (1955)

1951 1955 Th, daughters Th(NO3)4, other 

forms

Th extraction from monazite sands (1949-1950). Solvent extraction to 

purify Th(NO3)4 (1951-1952). Th extraction from Brazilian monazite 

sludge (1954-1955)

2 U, Th DU, NU, EU, Th 32.6 MT (Author unknown 1992)

1954 1959 U, daughters U metal, UO2, UO3, 

UNH

Process support for U metal production for Mallinckrodt, NLO

1954 1962 U, daughters UF4 Process support for U metal production (some work involved UF4) 

(1954-1958); prep of UO2 crystals from UF4 by vapor deposition (1959-

1962)

1954 1962 U, daughters U metal, alloys R&D, studies of work on U metal and its alloys

1954 1971 U, daughters UO2 (nat, depl) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, 

etc.) (a few plates: depleted U)

1955 1957 U, daughters U3O8, (NH4)2U2O7, 

other forms(?)

U extraction from U ores (ammonium carbonate leaching) (1955-1957)

1956 1956 U, daughters U2O2(NH4)2 Study of prep of ammonium uranate

1956 1959 U, daughters U compounds (see 

at right)

Study of U compounds: U beryllides, borides, silicides, sulfides

1956 1962 U, daughters UO2 (enr, mostly at 

93%)

R&D on UO2 for fuel of various types (powder, pellet, plate-type, etc.); 

enrichment from few percent to full; bulk of work apparently with 93% 

enr

1956 1962 U, daughters U carbides, mainly 

UC (nat; enr 9-

93%)

R&D, studies of U carbides and their alloys

1957

1961

1958;

1961

U, daughters UCl (enr; nat?); 

U3O8 (nat?)

Study of UCl dispersions (enr, short period only, 1957-1958); making 

UCl from U3O8 (1961)

1957 1961 U, daughters U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

R&D, studies of U nitrides and their alloys

1961 1961 U, daughters HEU3O8, HEU Prep of Al-U (HEU) fueled glass fiber fuel plates for irrad

1955

1971

1965;

1972

Th, daughters; 

minor U

Th metal, alloys; 

iodide, phosphate, 

halide, nitride, 

fluoride, 

carbide;ThO2

alloys

Th compounds, alloys (except extraction): bomb reduction of Th halides; 

prod Th ingots by arc melting, casting (with Th fluorides); Th-5, -10, -15, 

-20 w/o U alloys; Th-U alloy made with Th iodide; ingots cast by vacuum 

induction melting as Th buttons, casting Th-U carbides by arc-melting; 

ThN prep (1956-1961); Ta alloy buttons (trace Th) (1959-1960); Th 

carbide powder prep by vapor deposition (1960); corrosion of Th under 

storage, Th compounds in NaK (1960-1961); additives to Th-U (1960-

1965); ThO2 alloys with other mats (1971-1972)

1956 1956 U content Fissium Fissium ("synthetic fissium alloy"): specimen fab, annealed; X-ray diffrac

3 U, Th DU, NU, EU, Th 39.5 MT (Author unknown 1992)

1954 1959 U, daughters U metal, UO2, UO3, 

UNH

Process support for U metal production for Mallinckrodt, NLO

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

Page 2 of 9
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Bldga Start End Isotopesb Formc Type(s) of workd

1954

1979

1960;

1979

U, daughters U metal, alloys R&D, studies of work on U metal; its alloys, incl electroplating; vapor 

deposition of metals on U

1954 1962 U, daughters UF4 Process support for U metal production (some work involved UF4) 

(1954-1958); prep of UO2 crystals from UF4 by vapor deposition (1959-

1962)

1954 1971 U, daughters UO2 (nat, depl) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, 

etc.) (a few plates: depleted U)

1956 1959 U, daughters U compounds (see 

next column)

Study of U compounds: U beryllides, borides, silicides, sulfides

1956 1962 U, daughters UO2 (enr, mostly at 

93%)

R&D on UO2 for fuel of various types (powder, pellet, plate-type, etc.); 

enrichment from few percent to full; bulk of work apparently with 93% 

enrichment

1956 1962 U, daughters U carbides, mainly 

UC (nat; enr 9-

93%)

R&D, studies of U carbides and their alloys

1957 1961 U, daughters U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

R&D, studies of U nitrides and their alloys

1957

1961

1958;

1961

U, daughters UCl (enr; nat?); 

U3O8 (nat?)

Study of UCl dispersions (enr, short period only, 1957-1958); making 

UCl from U3O8 (1961)

1955

1971

1965;

1972

Th, daughters; 

minor U

Th metal, alloys; 

iodide, phosphate, 

halide, nitride, 

fluoride, 

carbide;ThO2

alloys

Th compounds, alloys (except extraction): Th ingot prod by arc melting, 

casting (Th fluorides); Th-5, -10, -15, -20 w/o U alloys; Th-U alloy with 

Th iodide; ingots cast by vacuum induction melting as Th buttons, 

casting Th-U carbides by arc-melting; ThN prep (1956-1961); Ta alloy 

buttons (trace Th) (1959-1960); Th carbide powder prep by vapor 

deposition (1960); corrosion of Th under storage, Th compounds in NaK 

(1960-1961); Th-U additives (1960-1965); ThO2 alloys with other mats 

(1971-1972)

1956 1956 U content Fissium Fissium ("synthetic fissium alloy"): specimen fab, annealed; X-ray diffrac

4 U, Th DU, NU, EU, Th 12.0 MT (Author unknown 1992)

1954 1959 U, daughters U metal, UO2, UO3, 

UNH

Process support for U metal production for Mallinckrodt, NLO

1954

1979

1961;

1979

U, daughters U metal, alloys 

(nat, 10% enr)

R&D, studies of work on U metal and its alloys (nat, 10% enr)

1954 1962 U, daughters UF4 Process support for U metal production (some work involved UF4) 

(1954-1958); for preparation of UO2 crystals from UF4 by vapor 

deposition (1959-1962)

1954 1971 U, daughters UO2 (nat, depl) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, 

etc.) (a few plates: depleted U)

1955 1957 U, daughters U3O8, (NH4)2U2O7, 

other forms(?)

U extraction from U ores (ammonium carbonate leaching) (1955-1957)

1955? 1965? Fission gasese Fission gas analysis in fission gas laboratory (in Radioisotope 

Laboratory)

1956 1962 U, daughters UO2 (enr, mostly at 

93%)

R&D on UO2 for fuel of various types (powder, pellet, plate-type, etc.); 

enrichment from few percent to full; bulk of work apparently with 93% 

enrichment

1956 1956 U, daughters U2O2(NH4)2 Study of preparation of ammonium uranates, including X-ray diffraction

1956 1959 U, daughters U compounds (see 

next column)

Study of U compounds: U beryllides, borides, silicides, sulfides

1956 1962 U, daughters U carbides, mainly 

UC (nat; enr 9-

93%)

R&D, studies of U carbides and their alloys

1957

1961

1958;

1961

U, daughters UCl (enr; nat?); 

U3O8 (nat?)

Study of UCl dispersions (enr, short period only, 1957-1958); making 

UCl from U3O8 (1961)

1957 1961 U, daughters U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

R&D, studies of U nitrides and their alloys

1961 1961 U, daughters HEU3O8, HEU Preparation of Al-U (HEU) fueled glass fiber fuel plates for irradiation

1955 1955 Th, daughters ? Support for Th extraction activities?

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

Page 3 of 9
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Bldga Start End Isotopesb Formc Type(s) of workd

1955

1971

1965;

1971

Th, daughters; 

minor U

Th metal, alloys; 

iodide, phosphate, 

halide, nitride, 

fluoride, 

carbide;ThO2

alloys

Th compounds, alloys (except extraction): bomb reduction of Th halides; 

Th ingot prod by arc melting, casting (with Th fluorides); Th-5, -10, -15, -

20 w/o U alloys; Th-U alloy made with Th iodide; ingots cast by vacuum 

induction melting as Th buttons, casting Th-U carbides by arc-melting; 

ThN prep (1956-1961); Ta alloy buttons (trace Th) (1959-1960); Th 

carbide powder prep by vapor deposition (1960); corrosion of Th under 

storage conditions, Th compounds in NaK (1960-1961); additives to ThU (1960-1965); ThO2 alloys (1971-1972)

1956 1956 U content Fissium Fissium ("synthetic fissium alloy"): specimen fab, annealed; X-ray diffrac

1954 Radioisotopes in industrial control (tracer methods)

1958 1960 Rad chemistry of inclusion compounds: nitration of hydrocarbons, urea 

compounds

1967 1974 Research on bioeffects of underground explosions at Amchitka Island, 

Alaska during/after Milrow, Cannikin tests (samples counted at Battelle?)

1982 1982? Rad surveillance for Piqua Nuclear Power Facility: annual rad survey, 

soil sampling. Apparently done by Battelle (samples counted at 

Battelle?)

5 U, Th DU, NU, EU, Th 14.7 MT (Author unknown 1992)

1954 1959 U, daughters Only U metal? Process support for U metal production for Mallinckrodt, NLO, including 

dingot metal

1954

1979

1960;

1979

U, daughters U metal, alloys R&D, studies of work on U metal and its alloys, including extrusion 

cladding, roll-bonding, fuel element fabrication

1954 1971 U, daughters UO2 (nat, depl) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, 

etc.) (a few plates: depleted U)

1956 1962 U, daughters UO2 (enr, mostly at 

93%)

R&D on UO2 for fuel of various types (powder, pellet, plate-type, etc.); 

enrichment from few percent to full; bulk of work apparently with 93% 

enrichment

1956 1962 U, daughters U carbides, mainly 

UC (nat; enr 9-

93%)

R&D, studies of U carbides and their alloys, including fuel element fab 

and coating

1957 1961 U, daughters U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

R&D, studies of U nitrides and their alloys, including bonding, welding

1956 1959 U, daughters U compounds (see 

at right)

Study of U compounds: U beryllides, borides, silicides, sulfides. 

Included ball-milling, fabrication

1955

1971

1965;

1972

Th, daughters; 

minor U

Th metal, alloys; 

iodide, phosphate, 

halide, nitride, 

fluoride, 

carbide;ThO2

alloys

Th compounds, alloys (except extraction work): production of Th ingots; 

Th-5, -10, -15, -20 w/o U alloys; Th-U alloy made with Th iodide; Th, ThU carbides; ThN prep (1956-1961); Ta alloy buttons (trace Th) (1959-

1960); prep Th carbide powders (1960); additives to Th-U (1960-1965); 

ThO2 alloys with other mats (1971-1972)

1958 1960 P-32, Ba-140, 

La-140, Y-91

Tracers in cement Develop activation analysis, other methods to measure Mg, Ca, Al, Fe in 

cement; e.g., P-32 labeling, EDTA titration to precipitate tagged Ag*IO

6/7 U, Th DU, NU, EU, Th 7.9 MT in Bldg 6, 8.7 MT in Bldg 7 (Author unknown 1992)

1954 1959 U, daughters U metal, UO2, UO3, 

UNH

Process support for U metal production for Mallinckrodt, NLO

1954

1979

1961;

1979

U, daughters U metal, alloys R&D, studies of work on U metal and its alloys, including corrosion 

studies

1954 1962 U, daughters UF4 Process support for U metal production (some work involved UF4) 

(1954-1958); for preparation of UO2 crystals from UF4 by vapor 

deposition (1959-1962)

1954 1971 U, daughters UO2 (nat, depl) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, 

etc.) (a few plates: depleted U)

1955 1957 U, daughters U3O8, (NH4)2U2O7, 

other forms(?)

Chemical extraction of U from U ores (by ammonium carbonate 

leaching, as feed to Grand Junction pilot plant) (1955-1957)

1956 1956 U, daughters U2O2(NH4)2 Study of preparation of ammonium uranates

1956 1959 U, daughters U compounds (see 

at right)

Study of U compounds: U beryllides, borides, silicides, sulfides

1956 1962 U, daughters UO2 (enr, mostly at 

93%)

R&D on UO2 for fuel of various types (powder, pellet, plate-type, etc.); 

enrichment from few percent to full; bulk of work apparently with 93% 

enrichment

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

Page 4 of 9
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Bldga Start End Isotopesb Formc Type(s) of workd

1956 1962 U, daughters U carbides, mainly 

UC (nat; enr 9-

93%)

R&D, studies of U carbides and their alloys

1957 1961 U, daughters U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

R&D, studies of U nitrides and their alloys

1957

1961

1958;

1961

U, daughters UCl (enr, nat); 

U3O8 (nat?)

Study of UCl dispersions (enr, short period only, 1957-1958); making 

UCl from U3O8 (1961)

1961 1961 U, daughters U3O8 Preparation of single U3O8 crystals, examination

1955 1955 Th, daughters Support for Th extraction activities?

1955

1971

1965;

1972

Th, daughters; 

minor U

Th metal, ThO2

alloys; iodide, 

phosphate, halide, 

nitride, fluoride, 

carbide

Th compounds, alloys (except extraction): Th-5, -10, -15, -20 w/o U 

alloys; Th-U alloy made with Th iodide; ThN prep (1956-1961); solubility 

of U in Th, corrosion of Th under storage, Th compounds in NaK (1960-

1961); Th-U additives (1960-1965); ThO2 alloys (1971-1972). 

Dissolution of Al-canned Th (1955)

1960? 1972? Pu, U Spent fuel (small 

quantities?); Pu 

alloys, UN-PuN

Study of Pu in pyrometallurgical process for U recovery from spent fuel 

elements, with Al-10 w/o U-5 w/o Pu; corrosion of Ta, Ta alloys by liquid 

Pu alloys (1960?). Study of UN-PuN fuel mat, cladding for fast reactors

1956 1956 U content Fissium Fissium (synthetic fissium alloy): specimen fab, annealed; X-ray

diffraction

1954

1958

1954;

1962

Fe-59, Mn-54

1956 1961 HEUO2, depl UO2; 

other?

Miscellaneous neutron dosimetry/monitor work, including wire-activation 

method for reactor flux profile data; fast neut effects on semiconductors; 

neut activation of ion chamber walls; ceramic tubes of Al2O3-Mo-Si2-UO2

with enriched U for fission-power elements, depleted U for heating 

elements (1959-60)

1958 1960 Rad chemistry of inclusion compounds by nitration of hydrocarbons and 

their urea (nitration) compounds

1958 1960 P-32, Ba-140, 

La-140, Y-91

Tracers in cement Develop activation analysis, other methods to measure Mg, Ca, Al, Fe in 

cement; radiometry of samples. P-32 labeling, EDTA titration with 

tagged Ag*IO

1959 1960 Fission 

gases?e

Isotopic exchange leak detection system (an Ag-Br column in 

conjunction with delayed-neutron monitor)

1960 1962 H-3 Triated cytidine, 

trace levels

Miscellaneous biological effects studies, including nucleic acid 

metabolism in HeLa S3 cells after X-ray-induced mitotic delay (using 

tritiated cytidine)

1962 1963 MFP; UO2 UO2 particles Internal rad effects on catalyst activity: 5 w/o ceramic-coated UO2

particles added to catalyst; catalyst efficiency tested; neut irradiation

1967 1974 Research on bioeffects of underground explosions at Amchitka Island, 

Alaska during/after Milrow, Cannikin tests (samples counted at Battelle?)

1982 1982? Rad surveillance for Piqua Nuclear Power Facility: annual rad survey, 

including soil sampling, by Battelle (samples counted at Battelle?)

9 U, Th DU 0.06 MT (Author unknown 1992)

1958 1962? Kryptone Krypton, UO2 Analysis of gas from SME, OMRE, and OMR fuel specimens; limited 

laboratory-level work with depleted and natural U

West Jefferson site

JN-1 

(Hot 

cells)

U, Th, AP, FP, 

Pu

DU, NU, EU, Th, 

AP, FP, Pu

8.9 MT (Author unknown 1992)

1954 1962 U, MFP, AP U metal, alloys 

(nat, 10% enr)

Postirradiation studies/examinations of clad nat U element; fuel 

elements, specimens irradiated elsewhere; U alloys (nat or 10% enr); UNb alloys

1954 1972 U, MFP, APe UO2 (nat) Postirradiation studies of corrosion of clad U metal, alloys and elements 

irradiated elsewhere; uncoated, coated fuel particles; fission gas release 

from UO2 crystals, UO2 in cladding; sweep capsules; cleavage faces for 

recoils, tracks; cladding contacted with UO2 foil (MFP penetration 

analysis). Fission product diffusion and recoil studies, including 

sectioning; rad damage studies in ceramics, Al by contact with UO2; 

iodine-retaining coatings

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

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Bldga Start End Isotopesb Formc Type(s) of workd

1962 1972 U, MFP, AP UO2 (nat) Study of Cr foils impregnated with fission product conc of 2 x 109

recoils/cm2 to measure Xe-133 release and Ce-141, Ru-103, and Nb-95 

distribution

1956 1961 Th, U, MFP, 

Cs-137e

U-Th alloys Studies of U-Th alloys: fission gas retention study, other postirradiation 

analyses and tests, including burnup disk sectioning, Cs-137 release

1956 1962 U, MFP, APe UO2 (enr 25-93%) BRR, ETR loop programs (25-50% enrichment); postirradiation 

examination of HEUO2 dispersed in SS; UO2-SS fuel (mostly at 93%); 

spherical fueled elements (93% enrichment); high burnup fuel (some at 

30% enrichment, some at 93%); encapsulated UO2-SS fuel (at 93% 

enrichment), UO2 pin-type fuel (at 35%); pellets (at 25-50% enrichment) 

and compacts for ML-1. Miscellaneous studies of fuel; UO2 dispersions 

in various matrices, claddings (20% enrichment); coated particles, 

including crushing to release fission gases; high-burnup irradiation 

stability study of Zr2O2-UO2, similar fuels

1957

1965

1961;

1971

U, MFP U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

Postirradiation testing of UN dispersion, cermet forms; fission bubble 

growth kinetics, migration; irradiation behavior of UN fuel elements

1958 1962 U, MFPe U carbides, mainly 

UC (nat; enr 10-

93%)

Postirradiation examination of UC (some natural, some enriched) 

capsules; dispersed UC, UC2 (nat); other UC, UC2 specimens (at 30% 

enrichment, some at 93%), further specimens (at 10% enrichment) to 

high burnup; included section/seal/heat in tubes for fission gas 

examination, postirradiation heat/capture fission gases on charcoal traps 

or by freezing, e.g., coated nat UC2 particles; fission product deposition 

in out-of-pile loop system (UC2); fission bubble growth kinetics, 

migration; irradiation behavior of UC elements

1962

1965

1962;

1966

U, MFPe U silicides, 

hydrides

Postirradiation studies of U2Si2; deposition behavior of fission products in 

an out-of-pile gas loop system, with U hydride as fuel

1957 1964 U, Th; MFPf, 

AP

Spent fuel in solid, 

liquid form (mostly 

non-rad)

Recovery (dissolution) of spent fuel: Zircex, Darex, Thorex, Zirflex, 

fluoride volatility, Niflex processes; mostly non-rad Zirflex: U fuel 

elements; Thorex: Th forms. Examine ORNL Fluoride Volatility Pilot 

Plant vessels, Hanford Purex evap steam coil sections. 1961 on: mostly 

fluoride volatility process studies 

1958 1966 AP, some 

incidental MFP

Al plate; Nb, Zr, SS 

as cladding, matrix, 

or alloy component

Postirradiation examination of GCRE Mark 1 fuel plate Al; elements with 

Nb, Zr cladding or alloys; irradiation surveillance program on SS; seals, 

gaskets, etc.; non-rad core, structural materials, especially cladding (SS, 

Inconel, Zircaloy-2, etc.), sintered Al particles, often with postirradiation 

heating; Ta alloy damage, high-burnup boron compound dispersions; 

materials, electrical components for SNAP-8

1964 1983 MFP, AP Al plate; Nb, Zr, SS 

as cladding, matrix, 

or alloy component

Postirradiation examination of items from power and defense reactors: 

Shippingport Atomic Power 1B generator (1964-65); Fermi Reactor 

failed fuel (1967?); PM-3A Type I, Serial 2 (defective) core and Serial 3 

core (1967-68); Vermont Yankee 304 SS Sch 80 reducer (1974); 

Arkansas Nuclear Plant 1 reactor bldg spray system SS 304 recirc

piping (≤1975); H. B. Robinson, Point Beach Zircaloy (1975-1976); fuel 

rods (unspecified, 1977); Point Beach 1, Dresden 3 fuel rods, 

Oskarshamn 1 fuel rod sections (1978); Connecticut Yankee failed fuel 

assemblies (1981); oxide thickness on spent fuel rods (unspecified, 

1983); support of TMI cleanup: characterization of highly loaded 

EPICOR II prefilter liners #3 and #16 by sampling, analyzing liner 

contents, examining liner (1981-1983)

1979 1983? MFP Spent fuel Shipping cask sabotage source term investigation; used five irradiated 

fuel rods from the H. B. Robinson reactor (1979)

1964 1968 Pm-147 Pm oxide RTG work: hot cell examination of SD-4 thermionic diode (1964-1968)

1964 1977 Pu-238, U, Th, 

MFP; Nd-150, 

N-15 as tracers

PuO2, UO2, UO2-

PuO2, UN-PuN

Rad effects on irradiated PuO2 (1967); in-pile creep of PuO2, UO2-PuO2

(1970-1972). UN-PuN capsule postirradiation examination (1970?-

1971); PuN-UN burnup with Nd-150 tracer, UN-PuN burnup with N-15 

tracer (1970). After Pu-238 prod in commercial power reactor: 

postirradiation examination, Pu recovery (1974?-1975). Experiments, 

irradiation effects on fuel mat aerosols (UO2, PuO2) (1975?-1977)

1974? 1975 Np, Pu Pu fuel? After production of Pu-238 in commercial power reactor: postirradiation 

examination, Np recovery (1974?-1975)

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

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Bldga Start End Isotopesb Formc Type(s) of workd

1972 1980s

?

Cf-252 Cf-252 wire Processing and encapsulating Cf-252 (produced elsewhere)

1956

1966

1961;

1967

HEUO2, depl UO2; 

other?

Miscellaneous neutron dosim/monitor work, including fast neut effects 

on semiconductors; neut activation of ion chamber walls; proportional 

flow counter for neut activation studies of MGCR specimen fission gas 

release (including in-op degassing) (1959). Ceramic tubes of Al2O3-MoSi2-UO2 with enrichment U, depl U (1959-1960)

1958 1960 Develop activation analysis, other methods to measure Ca, Al, Fe in 

cement

1961 1962 Evaluation of stresses, other test parameters on sealed radiation 

sources

1962 1963 MFP; U UO2 Internal radiation effects on catalyst activity: 5 without ceramic-coated 

UO2 particles added to catalyst; catalyst efficiency measured after neut 

irradiation. Postirradiation beta-gamma activity followed as function of 

decay time. Catalysts with up to 65 mCi/g of fission product beta activity 

used in dehydration of cyclohexanol

JN-2 U, Pu NU, EU, Pu 1.7 MT (Author unknown 1992)

(Critical 

Assembly Lab, 

Pu Lab)

1955 1961 U metal foil (nat, 

depl); UO2 (2-93% 

enr)

APPR criticality experiment (1955, enriched U foil). Evaluation reflected 

heterog BWR, including critical assembly experiments (U foil) (1957-

1958). Crit experiment with movable water reflector; core: Alpolyethylene-U core (1957-?). Crit assembly experiment for GCRE-1 

(UO2) (1958). Plastic-moderated crit assembly: polyethylene, 

polystyrene (1958-1961). Mockup ML-1-1B reactor, GCRE crit 

assembly elements rewrapped to average 303 g U-235 per element, 61 

element positions; 19 pins/element, with 7 at 35 vol%, 12 at 70 vol% 

UO2 (1959-1960). Ditto, but 48 elements with total of 31.1 kg U-235 

(1961). Crit assembly experiments with UO2 pellets, pins (1.8% 

enrichment); depleted U foil (0.04 w/o U-235), natural U foil to calibrate 

Battelle counting facilities in activity per fission in U-235, U-238 (1959-

1960)

1964 1968 Pm-147 Pm oxide RTGs: purify multicurie quantities of Pm-147 (oxide) (1964-1966). Fab, 

test SD-4 thermionic diode (1964-1968), segmented thermoelectric 

modules (1966)

1964

1974

1968;

1978

Pu-238, Pu, U, 

Th, MFP; Nd150, N-15 (as 

tracers)

Pu and U-Pu forms 

(see at right)

Study PuN as fuel; Pu-Al, Pu-Fe, U-Pu-fissium, U-Pu-Mo, PuO2-UO2; 

solid-solution alloys (Th-Pu, U-Pu-Nb, etc.); liquid-metal fuels (Pu 

alloys); ceramic fuels (PuC, PuC-UC mixtures, PuN, Pu2Si3, PuSi2); 

cermet fuels (PuC, PuO2); cast dispersion alloys (Pu-Fe alloys); PuB; 

fab, casting, etc. (1964-1968). Prepare U, Pu microspheres as intrinsic 

thermocouple flux probes (PuO2) (1966-1968). Study UN-PuN fuel mat, 

cladding for fast reactors: prepare UCN-PuCN by arc melting, casting; 

vaporization of Pu nitride, carbide fuels; UN-PuN with liquid Na or with 

SS, Nb-Zr, Inconel, 

Incaloy cladding (1966-1968?). Fab, examine UN-PuN pellets, EBR II 

pellet-type fuel, UN-PuN pin-type elements, irrad capsules (1968-1971). 

1970: Study of PuN-UN burnup using Nd-150 tracer, N diffusion in PuN, 

UN-PuN with N-15 as tracer (1970). Devel Na-bonded SS-clad PuN, 

capsule (1972?). Design, fab RTGs with Pu-238 (PuO2) (1974-1978)

1966 1974? Np, Pu NpO2 Prepare Np microspheres as thermocouple flux probes (Np oxide) 

(1966-1967?); fab NpO2-ZrO2, NpO2-CaO-ZrO2 target rods for Pu prod 

(1971?-1974?)

1974 1978 Cm-244 Cm oxide? Design, fab RTGs with Cm-244 (1974-1978)

JN-3 U NU, EU 0.1 MT (Author unknown 1992)

(BRR) 1955 1965 U, MFP, APe UO2 (nat) Irradiation/damage study of: graphite; UO2 rods in graphite; uncoated, 

coated fuel particles; ceramics. Foils in recoil studies. Irradiate to study 

fission gas diffusion/release from foils, coated powders, clad forms; 

releases from sweep capsules, fueled graphite in in-pile loop. UO2-

contacted Al damage study

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

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Bldga Start End Isotopesb Formc Type(s) of workd

1956 1962 U, MFP, APe UO2 (enr 25-93%) BRR, ETR loop programs (25-50% enrichment); evaluate spherical 

fueled elements (93% enrichment); fission gas release from high-burnup 

irrad (some at 30% enrichment, some at 93%); irradiation of UO2-SS 

capsules, other fuels (at 93% enrichment), including gas loop 

measurements; irrad coated particles 

1957 1957 U, MFP, AP U metal (nat), 

alloys (nat, 10% 

enr)

Irradiation of Zr-clad natural U element; irradiation study of U alloys 

(natural or 10% enrichment)

1958 1962 U, MFP, APe U carbides, mainly 

UC (nat; enr 10-

93%)

Irradiation of UC (natural, enriched) capsules; PyC-coated UC2 particles 

(natural); UC, UC2 (natural) dispersion specimens, other UC, UC2

specimens (at 30% enrichment, some at 93%); high-burnup specimens 

(at 10% enrichment); coated spheres in sweep capsules, including inpile fission gas analysis. Fission product deposition in out-of-pile loop 

system (UC2); fission bubble growth kinetics, migration; UC fuel irrad 

behavior

1958 1972 AP See at right Irradiation effects on seals, gaskets, etc.; non-rad core, structural mats, 

especially cladding, sintered Al particles. Irradiation of SNAP-8 mats, 

electrical components

1965 1971 U, MFP U nitrides, mainly 

UN (depl; nat; enr 

5-93%)

Postirradiation testing of UN dispersion, cermet forms; fission bubble 

growth kinetics, migration; irradiation behavior of UN fuel elements

1956 1961 U, Th, MFP U-Th alloys U-Th alloy studies: irradiation to measure fission gas retention, other 

postirradiation data

1964 1971 Pu, U, MFP; 

Nd-150, N-15 

(tracers)

PuO2, PuN, UNPuN; Nd-150, N-15 

(as tracers)

Rad effects on PuO2 (1967); PuO2, UO2-PuO2 in-pile creep studies 

(1970-1972). Study of UN-PuN fuel mats: irradiation of capsule(s) 

(1970-1971). PuN-UN burnup study with Nd-150 tracer; UN-PuN with N15 tracer (1970). Irradiation, experiments re fuel material aerosol 

properties (UO2, PuO2) (1975?-1977)

1964 1968 Pm-147 Pm oxide RTG work: irradiation, in-pile testing of SD-4 thermionic diode (1964-

1968)

1956

1966

1961;

1967

U, MFP, AP UO2 (enr, depl), 

other forms?

Miscellaneous neutron dosimetry/monitor work, including the following. 

Wire-activation for reactor flux profile data; fast neut effects on 

semiconductors; thermal neut flux meter, sensor; "tactical" neut dosim; 

fast neut dosim system; neut activation of ion chamber walls. 

Proportional flow counter for neut activation studies of MGCR specimen 

fission gas release (1959). Thermal neut flux monitoring system for 

Hanford; ceramic tubes of Al2O3-Mo-Si2-UO2 with enriched U, depleted 

U (1959-60). Study of 1-50K rad semiconductor fast-neut dosimeter 

(1966-1967) 

1956 1961 Miscellaneous gamma dosimetry/monitor work, including the following. 

Gamma effects on semiconductors, tin oxide films; effects of highintensity rad bursts on electrometer-type ion chambers; gamma testing 

of photovoltaic gamma dosimeter; high-level gamma dosimetry 

research, including glasses for high-level gamma dosimeters

1958 1960 Ba-140, La140, Y-91Tracers in cement Develop activation analysis to measure Mg, Ca, Al, Fe in cement

1958 1961 Fission gasese Installation, use of in-pile apparatus (loop) to study fission gas release 

(in beam tube facility) (1958-1960). Beam tube furnace for coated 

particle fuel irradiation, including collection, analysis of released fiss 

products; develop methods to study fission gases with gamma spec 

analysis, charcoal traps (1961)

1959 1960 Fission 

gases?e

Develop fuel element leak detector

1960 AP? Facility to study fast neutron activation using the BRR spectrum

1962 1963 U, daughters UO2 Internal rad effects on catalyst activity: 5 w/o ceramic-coated UO2

particles added to catalyst, irradiation

1964 N-17/O-17 Develop N-17 (O-17) decay neutron monitor to track reactor power

1971 1975 AP? Likely HEUO2 Neut radiography of reactor fuel: done in reactor pool and in out-of-pool 

facility

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

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Bldga Start End Isotopesb Formc Type(s) of workd

JN-4 

(part of 

this 

was 

called 

the 

"Old" 

Pu Lab)

1960? 1968 Pu, Th, U Pu, PuO2, PuO3, 

Th-Pu and U-Pu 

alloys

Study Pu in pyromet process for U recovery from spent fuel, with Al-10 

w/o U-5 w/o Pu; corrosion of Ta, Ta alloys by liquid Pu alloys (1960). 

Study PuO2 pellets in SS, Inconel, etc.: 10-15 vol% Pu2O3 presintering to 

PuO2 (1960-1962). Pu alloys, 5-28 w/o Pu; Th-Pu; heat, hot roll, 

metallog examination (1960-1962). Evaluate PuN fuel; Pu-Al, Pu-Fe, UPu-fissium, U-Pu-Mo, PuO2-UO2; solid-solution alloys (Th-Pu, U-Pu-Nb); 

liquid-metal fuels (Pu alloys); ceramics (PuC, PuC-UC mixtures, PuN, 

Pu silicides); cermets (PuC, PuO2); dispersion Pu-Fe alloys; PuB; fab, 

cast (1962-1968). Prep U, Pu microspheres (PuO2) (1967-1968). Study 

UN-PuN fuel mat: prep UCN-PuCN by arc melting, casting; vaporize Pu 

nitride, carbide fuels; UN-PuN with liquid Na or with SS, Inconel 

cladding; Na wetting of UN-PuN fuels (1966-1968?). Fabricate, examine 

UN-PuN pellets, EBR II pellet fuel, UN-PuN pin-type elements, irradiated 

capsules (1968-1971)

1964 1968 Pm-147 Pm oxide RTG: purify multicurie quantities of Pm-147 (oxide) (1964-1966)? 

Fabricate, test SD-4 thermionic diode (1964-1968). Develop segmented 

thermoelectric modules (1966)

1966 1967? Np Prepare Np microspheres as thermocouple flux probes (Np oxide)

1978 1980 Pu, Th, U, 

daughters

Residual 

radioactivity

Decontamination of JN-4

JS-1 1960s

?

Early 

1980s

?

U, Th DU, NU, EU, Th 0.9 MT (Author unknown 1992)

JS-10, 

JS-12

1960s

?

Early 

1980s

?

U DU, EU 0.3 MT for JS-10, 0.002 MT for JS-12 (Author unknown 1992)

INEL 1958 1960 U, MFP, fission 

gases

UC (nat, enr) Irradiate UC (natural, enriched) fission product release; in-pile 

irradiation, postirradiation examination of UC (SRE specimens): at 

MTR(?), Xe-133 assay during postirradiation heating with charcoal traps, 

gamma spec; also Kr-85. Battelle representative observed data-taking 

at MTR (INEL) 

Hanford 1961 1962 Research at Hanford (HAPO) on reactor materials in NPR; on PRTR 

operation

Power 

plants, 

ANL, 

INEL

1970 1981 Work at commercial nuclear power plants, etc. The 1970 work at 

Connecticut Yankee might have been part of program to use commercial 

reactors to make Pu-238 from Np target rods.

a. Bldgs 4, 6, and 7 might have been used for some work even where not listed, because it is not always clear what associated 

laboratory-type tests might have been done; also, X-ray diffraction of most samples and specimens should be assumed to have been 

performed (usually in Building 4 if the specimen was not hot; however, plutonium was not handled at Building 4).

b. MFP: mixed fission products; FP: fission products; AP: activation products (for the materials in question).

c. The U form was probably natural uranium when not specified; enriched uranium appears not to have been used before 1954. 

Sometimes only the final form was indicated by the relevant reference, so the possibility of a different initial or intermediate form than 

listed above must be kept in mind.

d. X-ray diffraction was said always to have been on a powder sample, so diffraction samples or specimens of a solid material had to be 

ground or milled before the test could be performed. Thus, at least small amounts of the material should be assumed to have been 

ground to powder form.

e. Fission gases could be generated in samples or rods irradiated elsewhere or in the Battelle reactor. Fission gases could be counted in 

the in-reactor loop at the Battelle reactor, in a JN-1 hot cell when a rod or sample was opened, or (if the total fission product content 

was less than 1 Ci) at the fission gas laboratory in the Building 4 Radioisotope Laboratory at King Avenue (Sunderman and Dickerson 

1962). If chemical separations or special counting had to be done, the sample was opened in the hot cell or at the fission gas 

laboratory. In both such cases, the sample could be heated in a quartz tube to drive off the gases, which were then collected. Townley 

et al. (1960) indicate that in the analysis of fission gases from samples irradiated at the Battelle reactor, some or all of the Kr-89, Xe137, Xe-140, and Xe-141 fission product chains were studied, with the respective down-the-chain daughters S-89, Cs-137, Ba-140, 

and Ce-141 being analyzed in some studies by being trapped on stainless steel mesh and with I-131 and I-133 also being analyzed 

when they deposited on the traps; the traps were dissolved in acid and the various daughters and radioiodine were then chemically 

separated and counted. 

f. One report stated that the principal radioisotope after a 10-month cooling of the fuel was Co-60 (~100% of gamma activity in the eluted 

fraction) (Dayton and Tipton 1960a). However, another report stated that at high burnup, Co-60 was a minor contributor to the total 

activity of decladding solutions and that the MFPs were in the same proportion as in the fuel dissolved except for cesium (Ewing, 

Brugger, and Sunderman 1961).

ATTACHMENT C

HISTORY OF WORK DONE AT BATTELLE, BY BUILDING AND MATERIAL

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Material Process or operation Content and form notesa Period and amountb References

Source and production materials

U metal, natural Production from phosphate, etc. (see 

other entries below); metal refining 

process support; development of fuel 

rods as metal in a matrix; e.g., U in 

graphite; studies, fabrication of U metal, 

U alloys

1943-1979 overall. 1946-1957 

(extraction). 1952-1958 (metal refining 

process support, e.g., casting large 

ingots); largest ingot melt ~1,200 lb, 170 

lb typical (Rengstorff and Lownie 1955)]. 

1951 [10 kg of U ingots to ORNL (Author 

unknown ca. 1987)]

Author unknown ca. 1987; FUSRAP 

2000b, Ferry 1943, DOE 2000a, Dayton 

and Tipton 1959a, Rengstorff and 

Lownie 1955

U metal, depleted Used as foil in critical assembly work; 

machined, rolled 

1957-1960, possibly other years also FUSRAP 2000b, DOE 2000a

U metal, enriched R&D on fuel elements; processing and 

machining

10-93% enrichment 1954-1962; 1979. One critical 

experiment: critical mass of HEU foil was 

4.935 kg (Dayton and Tipton 1957).

Author unknown ca. 1987, FUSRAP 

2000b, DOE 2000a, Dayton and Tipton 

1957

U as recycle 

uranium

Apparently used for testing of recovery 

methods, assay

Depleted, normal, or enriched 1962-1980, sent by Battelle to FEMP or 

RFI: 168 kg depleted U, 5,135 kg normal 

U, and 29 kg enriched U (including 1.45 

kg U-235 for enriched). 1962-1982, 

received by Battelle: 1,616 kg depleted 

U, 2846 kg normal U, and 0.06 kg 

enriched U (including 1 g U-235 in the 

enriched). More than half of the total kg 

U transfers were in 1964

DOE 2000b

UO3 Small amount in metal refining process 

support; some made from UNH (1957)

Natural 1954-1959 Vaughan 1957

UO2 Extensive research, testing: production

from pitchblende, MgX, soda salt; 

metallurgical, fab operations, including 

fuel element production; metal refining 

process support; other uses, e.g., in 

instruments

Depleted, nat, enr; much as HEUO2; 

apparently all natural for refining process 

support. As powder, pellets, or plates. 

For 1.7 g UO2 powder, standard capsule: 

activation to 1.5 Ci at 10 s 

postirradiation, 0.23 Ci at 2 d

1946-1972 overall. 1952-1958 (metal 

refining process support). 1956-1961 

(instrument work). Neutron radiography 

of HEUO2 reactor fuel in the BRR in 

1971-1975. Not enriched before about 

1956

Author unknown ca. 1987, FUSRAP 

2000b, Chastain 1961

U carbides R&D, study of U carbides and alloys, 

mainly for fuel apps

Natural and enriched, up to 93% 1943-1944; 1950-1952; 1956-1961;

1972-1975

Litton 1960, Hanson 1961

U nitrides R&D, study of U nitrides and alloys, 

mainly for fuel apps

Depleted, natural, and enriched, up to 

93%

1945; 1957-1961 BMI 1945b

Uranate, UNH 

forms

Uranate studies; metal refining support; 

extraction products, byproducts; 

producing fueled glass fiber plates; 

producing UCl from U3O8

U3O8; HEU3O8 (fueled glass); 

U2O2(NH4)2; Na2U2O7. Possibly some 

enriched forms

1943-1944, 1956 (uranate studies). 

1952-1959 (metal refining process 

support: UNH, uranates). 1946-1957 (U 

extraction from various forms). 1961 

(fueled glass, UCl prod)

U salts, including 

UF4

Study fused salt mixtures as liquid fuel; 

refining process support (UF4); vapor 

deposition conversion of UO2 to UF4

U fluorides, hydrosulfides, phosphates, 

sulfates

1953-1954 (fused salt study). 1952-

1959 (metal refining process support). 

1959-1962 (conversion of UO2 to UF4)

ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL PRODUCED OR USED IN BATTELLE WORK

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Material Process or operation Content and form notesa Period and amountb References

Other U compounds Studies of miscellaneous U compounds; 

postirradiation studies of some; study, 

production of UCl

Natural and enriched, up to 93%. U 

beryllides, borides, UCl (dispersion), 

silicides

1945: Al with 1-30% U, 10-15-lb melts 

forming ingots ~6 in. × 10 in. × 1.5 in. 

(BMI 1945b). Feb 1947: 14 lb “TU-Be 

alloy” at 90% Be; 13 lb at 30% Be; 72 lb 

Tu-Be dross at 2% Be; 59 lb TuO2 +BeO 

at 33% Be (Grenall 1947). 1956-1959 

(misc compds). 1962, 1965-1966 

(postirradiation studies). 1957-1958 (UCl 

study); 1961 (UCl production)

BMI 1945b, Grenell 1947

Thorium metal and 

alloys

R&D with Th metal; forging, rolling, 

machining ingots into slugs for Hanford; 

arc-welding Ames Th, Th alloys; 

mech/metallurg study of Th; making ThU alloy with Th iodide; casting Th 

buttons; casting Th-U carbides by arcmelting

1944-1961 overall. 1944-1946 (R&D). 

1947 (slugs for Hanford)

BMI 1945a,b; Author unknown ca. 1987

Th nitrate and other 

Th salts; Thcontaining materials

Production of Th nitrate from various 

feed materials (see below); Th solvent 

extraction pilot plant, to purify Th(NO3)4

and produce Th metal

November 1949: Th product with 33 

parts U3O8 per million of ThO2 produced 

from carbonate solution initially having 

38,000 parts U3O8 per million of ThO2

1949-1950 (extraction from monazite 

sands). 1951-1952 (R&D, pilot plant). 

1954-1955 (extraction from Brazilian 

monazite sludge)

FUSRAP ca. 1987, 2000b; Author 

unknown ca. 1987; Brown 1951; Hunter 

1949, Kelley 1949, 1951; DOE 2000a

Other Th 

compounds

Th hydriding; electroplating metals on Th 

(Th phosphate); Th prep by iodide 

process; bomb reduction of Th halides; 

Th ingot prod by arc melting, casting 

(with Th fluorides); making Th-U alloy 

with Th iodide; ThN prep; Th carbide 

powder prep by vapor deposition; prep of 

Th-U, ThO2 with various additives

Th carbides, fluorides, halides, iodide, 

nitride, phosphate; ThO2. Most often to 

produce Th metal or a Th alloy. Th-U 

alloys: 5-10 without U

1951 (hydriding). 1948-1953 

(electroplating metals on Th). 1951-

1954 (Th prep by iodide process). 1954-

1955 (electrodeposition of Th to form 

ingots). 1956-1961 (ThN prep). 

Gallagher, Blosser, and Mann 1955

ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL PRODUCED OR USED IN BATTELLE WORK

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Material Process or operation Content and form notesa Period and amountb References

Plutonium and other 

transuranics

Study of Pu in U spent fuel recovery; Ta 

corrosion by Pu alloys; studies of UNPuN fuel, PuO2 pellets, Th-Pu and U-Pu 

alloys, liquid metal fuels, ceramic/cermet 

fuels; R&D on Pu processing, fabrication; 

fabrication of Pu-238 heat sources; 

breeder experiments; Pu production from 

Np targets fabricated at Battelle 

(irradiated elsewhere), Np/Pu recovery; 

Pu microspheres; aerosol experiments

Some transuranics as a result of breeder 

experiments. Usually Pu-238 (e.g., for 

RTGs: powder transferred, coated, 

pressed, sintered to pellets, welded into 

capsules), but Pu-239/240 also handled; 

Np-237 in microspheres. U-Pu spent 

fuel: 10 without U, 5 without Pu. 

Ceramic, cermet, and alloy fuels 

included Pu carbides, nitrides, oxide, and 

silicides and various U-Pu compounds. 

Some vaporization was done for fuel 

production. RTGs, aerosols: as PuO2

Work apparently began in 1960 

(recovery, corrosion studies). 1960-1962 

(PuO2 pellets; e.g., ~1 kg UO2-PuO2 

fuel pellets fabricated into test fuel in one 

campaign); nitride fuel studies. 1962-

1968 (various fuel studies; >20 kg of Pu 

metal fabricated into classified shapes). 

1964-1971 (test fuel irradiations). 1966-

1968 (Pu microspheres as probes). 

1968-1971 (fabrication of UN-PuN 

pellets, EBR II pellet fuel, UN-PuN pins, 

irradiation specimens; ~20 kg of Pu 

metal powder processed into PuN or 

PuN-UN pellets/test fuel)). 1974-1975 

(Pu recovery from targets). 1974-1978 

(RTG work: ~4 kg of Pu-238-O2 handled 

in fuel development studies, heat source 

fabrication. 1975-1977 (aerosol 

experiments). Some Pu sent from INEL 

in 1969-70 (form unknown).

FUSRAP 2000a, Toy 1986, DOE 2000a, 

Freas and Madia 1982, BCL (circa 1980)

Phosphate rock, 

superphosphate 

material

Extraction of U via leaching and other 

chemical processes

Florida phosphate rock, 0.009% U, 

extracted at 87-94% efficiency; 

superphosphate, 0.006% U, at 55% 

efficiency

1946-1950 Beyer 1948a, Blatz 1949

Shales Extraction of U via leaching and other 

chemical processes

Tennessee shales: 0.005-0.009 % U 1946-1949, 1952-1954 Beyer 1948b, Blatz 1949, Bearse et al. 

1948c

Monazite sands Extraction of U via leaching, other 

chemical processes; extraction of Th via 

solvent extraction, with production of 

ThO2 or Th metal

5-7% ThO2 for Brazilian, Indian sand 

concentrates, 3-4% for Idaho sand 

concentrates. 1949: 6 lb U + 100 lb Th 

from each ton of sand, i.e., 5-6% Th and 

0.3% U. 5-10 mR/hr contact with sack

1948-1950. In 1949, expected to 

process 30-40 lb of sand per week 

eventually (Blatz 1949)

Hunter 1949, Wallo 1981, Blatz 1949

Colorado carnotite 

ore; "Western" ores; 

other ores

Carnotite ore: ground to minus 10 mesh; 

sep of minus 325 mesh fraction; mixing 

with 10% salt; roasting. No details: 

Colorado, Western ores. Other ores: 

ammonium carb leaching

Colorado carnotite: 0.25-0.3% U as ore; 

after grinding, the 5% that was the minus 

325 fraction assayed at 0.7-1% U

1948-1949 (carnotite). 1952-1955 

(Western ores). 1955-1957 (other ores)

Willigman 1949, Author unknown ca. 

1987, FUSRAP 2000b, Blatz 1949

ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL PRODUCED OR USED IN BATTELLE WORK

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Material Process or operation Content and form notesa Period and amountb References

Pitchblende ore Probably same as Mallinckrodt process, 

including grinding; converted to UO2

Pitchblende ore, July 1944 on, at 300 mg 

Ra per ton (Mason 1977); to 65% (DOE 

1997, AEC 1949) or 70% (Dupree-Ellis 

et al. 2000) U3O8 by wt; to 100 mg 

Ra/ton (Dupree-Ellis et al. 2000);

average 135 mg Ra/ton (AEC 1949); 

0.185 ppm equil Ra for Q-11 (60%) ores 

(AEC 1949); 0.1 Ci/ton, to 70% U, avg U 

conc >25%, ~100 mg Ra/ton ore for 25% 

U (Eisenbud 1975)

1950-1951 Wesner et al. 1950, Langston, Tangel, 

and Richardson 1950, Mason 1977, 

DOE 1997, AEC 1949, Dupree-Ellis et al. 

2000, Eisenbud 1975

MgX, other sludge 

forms

MgX concentrate (Mg uranate prepared 

in Africa from low-grade ore tailings) 

converted to UO2. Other forms: washer 

slimes/slurries, low-grade sulfide 

concentrate

Pitchblende ore tailings, from which MgX 

was made, were 30-50% U3O8 (AEC 

1951).

1950 (washer slimes). 1950-1951 

(MgX). 1953 (sulfide concentrate)

Meaders et al. 1950; Ewing, Kiehl, and 

Bearse 1950; Ewing et al. 1950a, 1951; 

AEC 1951; Wheeler, Langston, and 

Stephens 1955; Van Kleeck, Macdonald, 

and Stephens 1956; Langston, 

Macdonald, and Stephens 1957

V-20 soda salt Converted to UO2 Probably sodium uranate (Na2U2O7). 

1952: Vitro soda salt assayed at 75% 

U3O8, Colorado soda salt at 60-72% 

U3O8 (Termini 1952)

1950-1951 Termini 1952

Intact fuel elements, 

assemblies; fuel 

samples

Unirradiated, irradiated, or spent fuel 

material: postirradiation studies, 

including fission gases (assayed in an inpile loop)

Irradiated/spent fuel specimens (usually 

enriched), mixed fission products; when 

cladding/additives present, could also 

have substantial level of activation 

products. Fission progeny on one loop 

filter: Ba-140, Ce-141, Cs-137, Sr-89 (I 

already decayed)

1955-~1983 DOE 2000a

Fuel elements and 

dissolution products 

(Th, U, activation 

products, fission 

products)

Dissolution of simulated and actual spent 

fuel elements to test recovery processes: 

Sulfex, Zircex, Darex, Thorex, Zirflex, 

fluoride volatility, Niflex processes. Con 

Ed pins were dissolved one at a time, 

with the solids from previous dissolutions 

added to each new dissolution. 

Co-60 was principal isotope after 10-

month fuel cooling (~100% of gamma 

activity in eluted fraction) (Dayton and 

Tipton 1960a); pin core had 44 g of 

95.5% ThO2-UO2 by weight, cladding 

had 25 g of Type 304 SS (Ewing, 

Brugger, and Sunderman 1961). But: at 

high burnup, Co-60 a minor contributor 

to total activity of decladding solutions, 

MFPs in same proportion as in fuel, 

except for Cs (Ewing, Brugger, and 

Sunderman 1961).

1957-1964 overall. 1955 (dissolution of 

Al-canned Th); 1959 (dissolution of 25 

ThO2-UO2 Con Ed pins, each 7.4 in. long 

by 0.31 in. in diameter; dissolution 

products were returned to ORNL for 

analysis)

Ewing, Brugger, and Sunderman 1961, 

Peterson et al. 1959

Miscellaneous 

activated material

Material irradiated at BRR or elsewhere, 

studied in hot cells during and after 

irradiation

Typically Al, SS, Nb, Zr. Often highly 

activated

1955-~1983

ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL PRODUCED OR USED IN BATTELLE WORK

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Material Process or operation Content and form notesa Period and amountb References

Fission plate Used in the BRR shielding facility for 

fission spectrum/energy studies, etc.

Plate: 3.5 kg U-235 (as HEU), clad with 

~25 mils 2S Al (Anno and Plummer 

1962, Morgan et al. 1958). Ratio of 

thermal to epithermal neutrons at plate 

was 67; average gamma 2 MeV; thermal 

neutron/gamma flux ratio 16 (Morgan et 

al. 1958)

Plate was 28 in. in diameter (Morgan et 

al. 1958)

Morgan et al. 1958

Fission productcontaining materialUsed to dope catalysts to test efficacy of 

catalysts

Up to 65 mCi per gram of catalyst At least 1962 Dayton and Dickerson 1962

Pm oxide Purification of multicurie quantities of 

Pm-147; fabrication of diodes, RTGs

Pm-147 1964-1966 (purification). 1964-1968 

(diodes). 1966-1968 (RTGs)

Ritzman et al. 1966

Np Production of Pu from Np targets 

fabricated at Battelle (irrad elsewhere), 

Np recovery; study, preparation of U, Pu, 

Np microspheres as flux probes

Np-237 in microspheres as Np oxide 1960-1974 overall? 1960 (recovery, 

corrosion studies). 1966-1967 

(microsphere work). 1974-1975 (Np 

recovery from irradiated targets)

FUSRAP 2000a

Cm RTG work Cm-244 (Cm oxides?) 1974-1978

Cf-252 material Cf wire ported into alpha-gamma cell, 

cut, weighed, put into containers. These 

were sealed by welding, leak-tested, 

washed. In second cell, welded into 

larger container, leak-tested, washed

Cf-252 Started in the mid-1970s Wastren 2001, Scotti and Martin 1972

Miscellaneous 

radioisotopes

Studies using radiotracers or labeled 

compounds; hydrocarbon rad chemistry 

studies; study using Cr foil impregnated 

with fission products to study Xe-133 

release, other fission product distribution; 

alpha cell direct-conversion generator 

develop

P-32, Ag-?, H-3, I-125 as labels; Nd-150, 

N-15 as tracers for burnup. Cr foil 

impregnated with 2E9 recoils/cm2 of 

unspecified fission products (to see Ce141, Ru-103, and Nb-95 distribution). 

Po-210 layer as cathode in alpha cell 

direct conversion generator

1954 (tracers for industrial process 

controls). 1958-1960 (P-32, Ag used for 

labeling). 1958-1960 (rad chemistry of

hydrocarbons). 1960-1962 (tritiated 

cytidine for bio studies). 1962-1970 (Cr 

foil during small part of this period). 

1963-1964 (alpha cell generator devel). 

1970 (burnup study). 1976 (at least) (I125 labeling)

Calkins and Pobereskin 1954; McFarling 

et al. 1962; Plummer et al. 1964

"Fissium" (faux 

irradiated fuel 

material)

Used to study dissolution, interactions of 

fuel materials

Radioactive content unclear: faux fission 

product mix, but might have had actual U 

content

1956 Dayton and Tipton 1956

Sealed sources

Gamma sources Used in nondestructive testing laboratory 

for radiography; used in Co-60 facility for 

irradiation effects research, destruction 

of chemical agents. Some sources 

stress-tested for standards development

Radiog: Co-60, Cs-137; at hot cells, 100 

mCi Co-60 source in well, other large 

sealed Co-60 sources, 100 mCi Ra-226 

source; Ir-192 (to 30 Rhm), Tm-170 (to 8 

Rhm) studied at King Ave (Rhoten 1958)

All of operating period (radiog). 1961-

1962 (stress eval). 1981-1982 (chemical 

agent destruction). Large Co-60, Ra-226 

sources: instrument calibration. ~1982: 

Co-60 pellets put in spent fuel cask to 

make 10K-Ci source for calibrating highlevel detectors for Three Mile Island

BMI 1977, Rhoten 1958, DeArmond 

2006

ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL PRODUCED OR USED IN BATTELLE WORK

Page 5 of 6
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/2

6/2010 Page 89 of 1

Material Process or operation Content and form notesa Period and amountb References

Beta sources Studies of reactor coolant channel voids; 

U penetration into Al cladding; Po-210 

alpha cell generators

Sr-90 source for uranium penetration into 

Al cladding (backscattering method); CePr source for alpha cell generator 

voltage checks

1960 (reactor coolant channel voids, U 

penetration into cladding); 1964 (Ce-Pr 

source)

Dayton and Tipton 1960b, 1961a; 

Plummer et al. 1964

Neutron sources Radiographic source(s) in testing 

laboratory; startup source in critical 

assembly facility (JN-2); activation 

source for production of short-lived 

isotopes (unspecified location

Critical assembly: Po-Be source ("small") 

(Hogan et al. 1958, Jankowski and 

Chastain 1958). Activation source: 10-Ci 

Po-Be source (Sunderman 1962, Dayton 

and Tipton 1961)

All of operating period for radiography 

(?); 1956-1970 for the startup source; 

Sept 1961 on for the activation source

Hogan et al. 1958, Jankowski and 

Chastain 1958, Ray 1972, Dayton and 

Tipton 1961, Sunderman 1962

U3O8 Small amount present with the metal in 

most processes

Ferry 1943

Residue, scrap from 

carnotite ore 

processing

70-200 mg U per L of liquid residue; 

0.34% U in cake residue; some scrap at 

0.2% U

Blatz 1949

Wastes and remainders

Leftover samples, 

specimens

Some were highly irradiated and 

activated

Activated parts, 

mats

Activation products DOE 2000a

a. X-ray diffraction was done on samples of many and probably most specimens. For this test, the material had to be in powdered form; therefore even larger solid specimens would have a 

small amount of powder associated with them, probably produced by grinding of the specimen or a specimen from the same lot. The age of forms obtained from elsewhere (e.g., metal 

from a refinery) is generally not available, so the material might have had significant progeny.

b. Information on quantities is not available in most cases. However, except for fuel production contracts and fuel handled in the BRR and the hot cells, the amounts handled at any given 

time appear to have been relatively small (laboratory or pilot-plant scale).

ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL PRODUCED OR USED IN BATTELLE WORK

Page 6 of 6
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 90 of 124

Description of incident Location Date Reference

When Co-60 wires were being cut and cast in the High Level Cell, the 

dose rates were so high that two windows cracked. (The dust level was 

high, so it is not clear if the dose rate was from the sources themselves or 

from surface dust on the windows.)

JN-1 Between 

1955 and 

1987

BCL 1997

A worker was preparing to expose a solution using a 5-Ci Cs-137 source. 

The arm of the camera (which locks the source in place when the top plug 

is screwed into its lowest position) was placed in backwards; this released 

the source and allowed it to fall out of the camera and into the solution 

when she opened the lower plug to create the beam. She tried to recover 

the source and replace in the camera using a threaded rod, but she was 

unable to do so after several attempts. She left the laboratory and sat in 

the office (where films in a drawer registered the dose rate there from the 

loose source) until a coworker came in from lunch. He helped replace the 

source in the camera. The source was later transferred from the 

defective camera to another camera by the assistant division chief. The 

first worker's film badge registered 9,600 mR; the second worker forgot to 

put on his film badge before attending to the source, so his dose was 

estimated to be 3,200 mR; the assistant chief also forgot to put on a 

badge, but no estimate of his dose is available. The workers were 

allowed to continue to work in the lab, but were restricted to non-rad work. 

Building 4, 

Radiography 

Lab

8/8/1956 Selander 1956, 

Petersheim 1956

Minor corrosion of the heat exchanger tube bundle resulted in 

contamination of the primary demineralized-water system by the 

undemineralized secondary-system water. Activation of the dissolved 

salts in the pool water caused radioactive contamination of the water. 

The tube bundle and the primary water were replaced. No exposures 

noted.

BRR Late 

October 

1957

Anno and 

Plummer 1962

In studies of the thermal conductivity of U and UO2, a leak developed in 

the hot cell measuring apparatus during analysis of hot uranium.

JN-1 Mar 1958

During SRE fuel materials program postirradiation studies, for one 

capsule, there was a significant air leak at the capsule-punch seal, so the 

capsule puncturing apparatus had to be modified. No exposures noted.

JN-1 Jun 1958 Dayton and Tipton 

1958

Irradiated capsule containing graphite-coated UO2 spheres was opened 

with power hacksaw (at ends) and a remote milling machine (lengthwise 

split). "During milling, the capsule slipped in the vise, allowing the middle 

tool to penetrate the middle can containing the two spheres fabricated at 

Battelle. The halves of the capsule pulled apart, freeing the inner cans ..., 

invalidating the fission gas measurements." After the canister gases 

were sampled, the canisters were opened by slitting the sides and ends 

with a milling machine. The spheres were recovered along with the 

graphite "flour" used to fill the annulus around the spheres and neutron 

dosimeter wires.

JN-1 ~12/1/1958 Dayton and Tipton 

1959b

A worker punctured his finger with contaminated tweezers while cleaning 

the low-level cell. After he removed his outer clothing, he tried to clean 

the wound, but after several attempts, the dose rate at the wound was 0.4 

mR/hr. He was taken to the emergency room at White Cross Hospital, 

where the doctor was not able to clean the contamination out of the 

wound. The surface of the skin around the wound was excised, which 

reduced the reading to background level. The HP accompanied him to 

the hospital and performed the radiation monitoring; the HP checked all 

surgical equipment and saw that all materials used in the 

decontamination were properly disposed of. (Fission products)

JN-1 2/10/1959 Selander 1959b

ATTACHMENT E

INCIDENTS REPORTED AS OCCURRING AT THE KING AVENUE 

AND WEST JEFFERSON SITES

Page 1 of 6
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 91 of 124

Description of incident Location Date Reference

A worker punctured his finger with a piece of metal containing radioactive 

dust, while cutting up the pan from beneath the cutoff wheel in the highlevel cell at the Hot Cell Building. This operation also caused a dust 

cloud that highly contaminated his face and clothing. After he removed 

his outer clothing, the radiation reading at 6 inches in front of his face was 

> 5 mR/hr. After decontamination, a survey showed that he had 

contamination in his nose and around his eyes. Contamination in his 

nose was 6,000 cpm beta-gamma on a Q-Tip. Readings around his eyes 

were about 0.5 mR/hr. Readings at the puncture location on the back of 

his fourth finger, right hand, were 5 mR/hr. (Fission products)

JN-1 2/12/1959 Selander 1959a

A worker technician was overexposed (greater than 3 rem in a quarter) 

while removing a Weasel capsule from the reactor pool. It was shipped 

from Bldg 4, Room 4005, via the Battelle mail car to the BRR. It had ~1 g 

of alumina powder and 1.7 g of UO2 powder packed into three quartz 

tubes, which in turn were in a sealed hollow aluminum container inside 

the Weasel capsule. The capsule was being irradiated in the pool at the 

end of a long wire. The technician was raising the wire when the CAM 

alarmed. He and a researcher left the area. Fission product activity was 

calculated to be 1.5 Ci at 10 s after exposure of the capsule, 0.23 Ci 2 d 

later. The capsule was placed in a cask. See Table 5-5 for dose rates 

related to this. The technician's dose to the lens of the eye was 7.75 rem 

(beta plus gamma); to the gonads, 13.55 rem; to the extremities, "within 

tolerance limits," and the whole-body dose, 13.2 rem gamma. (Probably 

the same incident as the entry immediately above.)

BRR 8/9/1961 Chastain 1961

A worker got a hand overexposure in cleaning highly contaminated 

equipment in a hot cell. He did not wear leaded gloves during many parts 

of the operation because they were too cumbersome and he thought he 

would not go over the limit.

JN-1 1Q62 Saling 1962

A worker opened each of four small lead containers, removed the X-ray 

diffraction mounts from the lead plugs, and placed each mount on a 

length of angle iron. A second worker pushed the angle iron with the 

mount through a port into the hot cell, removed the mount, and extracted 

the angle iron for the next mount. After all four mounts had been 

transferred into the cell, the first worker discovered that he was 

contaminated. The insides of both his lead gloves were contaminated, 

indicating that the contamination was on the outside of the lead 

containers because these had been handled with bare hands and all the 

work after the small containers were opened were done with the lead 

gloves on. The second worker checked himself and found that he too 

was contaminated; some contamination appeared to be inside his nose.

JN-1 11/14/1963 Saling 1963

A technician was overexposed (greater than 3 rem in a quarter). The 

overexposure was discovered when the film badge reports came back 

from Landauer. The cause was said to be the failure of a supervisor to 

keep proper track of radiation exposure.

JN-1 1Q64 Saling 1964

A worker's film badge for March 14-27, 1966, read 1,800 mrem. This was 

thought to have resulted from maintenance work in the stall area of the 

reactor pool on March 19. Although the dose rate in that general area, 

where he spent 5 minutes, was 2.5 R/hr before the reactor flanges were 

removed, the radiation level immediately in front of the beam tube was 

10-20 times higher and he might have passed close to the tube. The 

worker also handled a beam tube plug reading 10 R/hr for about half a 

minute. His pocket dosimeter had been dropped before this and had 

been thought to be in working order, but on March 21 it was discovered to 

be malfunctioning, so it did not register the exposure.

BRR 3/19/1966

ATTACHMENT E

INCIDENTS REPORTED AS OCCURRING AT THE KING AVENUE 

AND WEST JEFFERSON SITES

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Description of incident Location Date Reference

An unspecified spill occurred in the plutonium laboratory. During the 

cleanup, coveralls, rubber gloves, booties, paper suits, masks, and both 

body and ring film badges were worn. Nevertheless, one worker's 

badges were contaminated and his dose had to be estimated from past 

experience.

JN-4 9/9/1968 Ottman 1968

A worker exceeded 3 rem for 2Q73 by the 6/3/1973 film badge report. 

The exposure was primarily associated with control area and cell entries 

that involved cask unloading operations and experiment setups. ( Typical 

of various such reports from 1960 on.)

JN-1 2Q73 Lowry and Kirsch 

1973

Over many years, about 55 gal annually of hydraulic fluid leaked from the 

glands/ seals of the rams onto the Hydraulic Door Room floor to a depth 

of ~5 in. The fluid was pumped out and the seal gaskets were replaced in 

about 1974, but the leakage continued. At one time, the walls of the room 

were sandblasted and some sand was left in the room, so it was 

suspected that the rams were scarred and that was why leakage 

continued even after the seal gasket replacement. Washwater from many 

CAA cleanings was also thought to have drained into the opening 

between the door and the CAA floor, adding 1-3 in. of washwater. There 

was also an opening between the door and the floor of the cell interior. 

Fuel/cladding dust and fragments, etc., and even small items fell through 

the opening between the door and the HLC, producing a layer of sludge.

JN-1 Before

1974

Myers and 

Berchtold 1994a

A worker was exposed to Pu-239 via inhalation. Urine samples collected 

during the first week after exposure indicated excretion rates from 0.09 to 

0.21 dpm/24 hr. Urine samples collected during the fifth and sixth weeks 

after exposure indicated excretion rates below 0.03 dpm/24 hr. A sample 

collected July 20-21 with an unusually high 4,000 mL of urine showed a 

rate of 0.25 ±0.11 dpm/24 hr; however, this was considered a dubious 

result because of the volume. Urine, fecal, and whole-body counting data 

are also available for September-October 1974 for this worker, who was 

involved in another unspecified potential intake incident during the week 

of September 9-13, 1974. Fecal samples indicated that the lung burden 

of Pu-239 was unlikely to be more than a few percent of the MPBB. Two 

other workers exposed in the same incident showed some positive Pu239 results, but the urine and feces indicated an even lower intake for 

them, likely less than 1-2% of the MPBB.

JN-4 5/9/1974 Geiger 1974

A worker was exposed to an average airborne concentration of 1.0E-11 

μCi/mL of Pu-239, 7.17E-10 μCi/mL of Sr-90, and 7.37E-12 μCi/mL of U238 while engaged in operations associated with transfer operations in 

the control area.

JN-1 9/9-13/74 Kirsch 1974a

One worker was exposed to an airborne concentration of 3.60E-10 

μCi/mL of soluble Pu-239 and 4.31E-10 μCi/mL of soluble Sr-90 for a 90-

min period while engaged in transferring waste containers from the lowlevel cell to a shipping hopper in the control area at the Hot Lab Facility. 

Data were obtained from a lapel BZ sample. This was a planned 

operation: airborne contamination was anticipated and the worker was 

wearing protective apparel, including a full-face respirator. The respirator 

smears were taken immediately and the results indicated that

contamination did not enter the respirator. Urine and fecal samples were 

also negative. It was concluded that the worker "did not inhale any of the 

contaminant."

JN-1 Week of 

10/27/75-

11/2/75

Toy 1975

ATTACHMENT E

INCIDENTS REPORTED AS OCCURRING AT THE KING AVENUE 

AND WEST JEFFERSON SITES

Page 3 of 6
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 93 of 124

Description of incident Location Date Reference

Two electricians were tearing out electrical wiring connected to the 

ducting in the old part of the building. One apparently thought that 

airborne levels were high, so they both took nose swabs. The swabs 

were negative, but the second worker's lapel air filter read positive (80 

dpm of mostly Pu-239, 0.13% soluble). His gloves had been found to be 

contaminated and had been changed several times during the work, but 

his coveralls read clean. He submitted 24-hour urine and fecal 

specimens over the next 2 days and a smear survey of his work area was 

done.

JN-4 2/25/1977 Hicks 1977; 

Langendorfer 

1977; BCL 1977b 

(air sample 

report), 1977c 

(smear survey 

report); Farmelo 

and Helwagen 

1977 (lapel 

sampler report)

Core samples taken around the holding tanks and autoclave during the 

1978-1982 decontamination showed that contamination had occurred 

"sometime during the 16-year use."

JN-4 Between 

1960 and 

1977

Rudolph, Kirsch, 

and Toy 1984

Several minor spills occurred in the Plutonium Laboratory. JN-4 Between 

1960 and 

1977

Rudolph, Kirsch, 

and Toy 1984

At the end of the working day on February 25, 1977, it was found that 

during the removal of electrical wiring from the Old Laboratory area, the 

lapel sampler filter of a technician exceeded the 40-hr MPC for soluble 

Pu-239, 2.0E-12 μCi/mL. The concentration of February 25 was 

calculated at 3.51E-12 μCi/mL; after a 64-hr decay it showed 2.84E-12. 

The technician was removed from laboratory activities and placed on a 

bioassay program. His nasal swabs indicated no activity. A smear 

survey of the areas in which he had been working indicated activity; it was 

cleaned up.

JN-4 2/25/1977 Langendorfer 

1977, Farmelo and 

Helwagen 1977, 

Hicks 1977, BCL 

1977a,b

A release took place during a manipulator removal operation. The 

highest measurable contamination was 138 dpm/100 cm2 beta-gamma; 

no airborne contamination was detected.

JN-1 4/7/1977 Peters and 

Harrison 1998

A worker was repairing a contaminated transient tube burst apparatus, 

wearing what was later found to be a contaminated lead apron. The 

alpha air concentration inferred from his lapel air sampler was found to be 

10% percent greater than the 40-hr control guide for soluble Pu-239 (the 

soluble control guide was applied until a solubility analysis was done). 

Subsequent analyses of the air filter showed that the material was 

insoluble Pu-239, giving a concentration of 5.5% of the insoluble Pu-239 

control guide . Nasal swabs results were negative, as were bioassays for 

Pu, U, and Sr-90 in the urine and feces. Gamma isotopic analyses were 

also performed on the bioassay specimens.

JN-1 3/16/1978 Kirsch 1978a 

(includes most of 

the bioassay data)

A worker entered the Pu-238 laboratory with two other workers; they wore 

lapel samplers but he did not because he intended to stay only a minute 

or two. However, he became involved in moving gloveboxes (#33 and 

#44) with them over a 2-hr work period. His involvement became known 

only several days later. Because of the elevated readings from the lapel 

samplers, urine and fecal specimens were collected from him starting 

October 26. Because of the delay, the readings were essentially negative 

and he was assigned the same dose as that to another of the workers, 

196 mrem as the 50-yr accumulated lung dose.

JN-4 10/17/1978 Kirsch 1978b

ATTACHMENT E

INCIDENTS REPORTED AS OCCURRING AT THE KING AVENUE 

AND WEST JEFFERSON SITES

Page 4 of 6
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 94 of 124

Description of incident Location Date Reference

A routine count of a nasal swab for a worker after a control area entry 

prompted follow-up bioassay sampling. Counts for both a follow-up nasal 

swab and a smear of the worker's respirator interior were negative. Fecal 

samples were collected 1, 2, and 3 days later to assess the early 

clearance phase; a urine sample was collected. Results for all samples 

were negative for Pu, Am, and fission products, while the feces also 

showed normal concentrations of uranium. It was concluded that an 

inhalation intake had not occurred and that the original nasal swab 

probably was contaminated by contact with a contaminated surface in the 

hot change room.

JN-1 11/9/1978 Kirsch 1978c 

(includes most 

bioassay data)

The Hot Lab received an empty NAC shipping cask under the HEDL 

program. Workers were not aware that the cask was pressurized when 

the lid was removed in the JN-1N High Bay area in preparation for loading 

an HEDL fuel element. Airborne radioactivity was released. Workers 

involved in the cask operation were wearing respiratory equipment, but 

another worker not involved but in the area was not. All workers 

evacuated immediately. There were no CAM alarms. The immediate 

surrounding area was contaminated by airborne radioactive debris. 

Bioassay data was normal for the worker not wearing a respirator. 

Calculations indicated that DOE and NRC airborne concentration guides 

were not exceeded; concentrations measured in the area were an alpha 

max of 3.14E-14 μCi/mL and a beta-gamma max of 1.36E-11 μCi/mL. 

The floor was contaminated to 5-10 dpm/100 cm2 alpha and 153,000 

dpm/100 cm2 beta-gamma; equipment surfaces were contaminated to <5 

dpm/100 cm2 alpha and 42.3K dpm/100 cm2 beta-gamma. [Timeline 

given in Author unknown 1979; air sampling results in BCL 1979]

JN-1 1/16/1979 Author unknown 

1979, BCL 1979 (a 

timeline is given in 

Author unknown 

1979, air sampling 

results in BCL 

1979), Peters and 

Harrison 1998, 

Ray 1990

A depleted uranium fire occurred in an autoclave in Building JS-1, where 

military fuel was fabricated. The fire was brought under control and 

moved to another part of the building where a controlled burn continued 

until it self-extinguished. Then removable contamination was cleaned up. 

Personnel were not exposed to airborne contaminations above [NRC?] 

guidelines.

JS-1 9/22/1979 Smith et al. 1988, 

Peters and 

Harrison 1998, 

Ray 1990

In the Hot Cell Facility (JN-1), during unloading operations of a failed 

spent fuel assembly from the Connecticut Yankee Atomic Power 

Company, there was a release of airborne radioactivity and subsequent 

surface contamination from what one source described as an explosion 

when unloading a “superheated” fuel element. Exposure was said to be 

for from one to several minutes. At least one worker was not wearing a 

respirator. This was reported to NRC. Bioassay procedures, including in 

vivo counting, showed that exposures were below 10 CFR 20 limits and 

that pulmonary depositions were <1% of the permissible body burden. 

[Probably same incident as entry below]

JN-1 5/2/1980 or 

5/3/1980

Jensen 2003, 

DOE 2003a, Ray 

1990, Kirsch 1987

In a second NAC cask incident, a cask was opened above the storage 

pool, which resulted in the release of widespread radioactive 

contamination within the high bay area up to the ceiling. There was no 

release from the building; personnel exposures were well below 10 CFR 

20 limits. The cleanup and disposal of radioactive waste were completed.

JN-1 5/2/1980 Peters and 

Harrison 1998, 

Pasupathi and Toy 

1990, Kirsch 

2000b

ATTACHMENT E

INCIDENTS REPORTED AS OCCURRING AT THE KING AVENUE 

AND WEST JEFFERSON SITES

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Description of incident Location Date Reference

A worker received a hand dose of 22.48 rem in 2Q80. Of this, 20.560 

rem were received during an operation that involved the collection of 

scrapings from the interior wall of a spent fuel cask. The personnel 

monitoring device was a finger ring TLD. The operation involved two 

tasks: the collection of the sample and the manipulation of a plastic 

shroud attached between the cask lid and the cask as the lid was lowered 

onto the top of the cask. The shroud was used to minimize the migration 

of airborne particulate radioactive material from the cask interior. A small 

opening was made in the shroud and a scoop attached to a 3-ft rod was 

inserted through the opening to collect the sample. The sample was then 

removed and transferred to a container held by the employee in the right 

hand. There was a deviation [unspecified] in how this was done from how 

it was planned to be done. [Apparently it was flashed, unknown to HP.]

JN-1 7/2/1980 Toy 1981

A worker was cut in an unspecified fashion. The wound was smeared at 

the exit from the contaminated area and again several times after 

decontamination of the wound; all smears were negative. The Battelle 

wound monitor (detector) was also used to check the cut, again with 

negative results. It was concluded that no uptake had likely occurred. 

(Fission products)

JN-1 3/17/1981 BCL 1981a

A fire that resulted in $100,000 of damage occurred at Building 7. It is not 

clear if any radioactive material was involved.

Building 7 7/15/1982 ORAUT 2006a

A worker was burned in an unspecified fashion at Building 10 [sic]. The 

material was depleted uranium. Special urine monitoring was done every 

2 weeks (see undated entry below) in May 1984 due to what was said to 

be a potential inhalation in Building 2 and that could be related to this 

burn incident. (Note that Building 10 was not found to be contaminated in

characterization surveys (e.g., Jensen 2003) and was not listed in any 

document as being the site of radioactive work. Because Building 10 

adjoined Building 5, it is likely that the incident actually occurred in 

Building 5.)

"Building 10" 27 Feb 

1984

BCL 1984 

(includes wound 

survey results); 

dosimetry records

A worker was subject to special biweekly urine monitoring in May 1984 

due to a potential inhalation of uranium in Building 2. At this time, it was 

recommended that he discontinue rad work. When he was found still to 

be excreting low levels of uranium in October 1984, it was recommended 

that filtered exhaust ventilation be used if he were needed to perform 

work rolling uranium or the like.

Building 2 May 1984 McKown 1984a,b; 

dosimetry records

Co-60 wires were cut and cast into slugs in the High Level Cell, creating 

radioactive Co-60 dust. The radiation levels in the cell were so high that 

two windows cracked.

JN-1 Not stated Sunderman and 

Gates 1965

Air currents from a heating fan in the ceiling of the HEC fuel pool (high 

bay) disturbed the surface currents in the Washdown Room and caused a 

release into the pool area.

JN-1 Not stated Myers et al. 1994b

ATTACHMENT E

INCIDENTS REPORTED AS OCCURRING AT THE KING AVENUE 

AND WEST JEFFERSON SITES

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Bldg Compound or alloya

Large 

amounts 

handled? RNb

A U metal Yes U, daughters

U alloys Yes U, daughters

U3O8 Yes U, daughters

UO3 (part of production support R&D from UNH) U, daughters

UO2 (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

UNH U, daughters

UO2CO3·2(NH4)2CO3 (uranyl ammonium carbonate) U, daughters

U2O7(NH4)2 (ammonium diuranate) U, daughters

U2O7Na2 (soda salt) Yes U, daughters

U beryllides U, daughters

U borides U, daughters

U carbides, mainly UC (natural; enriched 9-93%) Yes NU/EU, daughters

UCl (enriched) EU, daughters

UF4 U, daughters

U fluorides (except UF4) (salt solution) U, daughters

U hydrosulfides (salt solution) U, daughters

U nitrides, mainly UN (natural; depleted; enriched 5-93%) Yes U, daughters

U phosphates (salt solution) Yes U, daughters

U silicides U, daughters

U sulfates (salt solution) U, daughters

U sulfides U, daughters

Th metal (including casting by vacuum induction) Yes Th, daughters

Th alloys (except with U) Yes Th, daughters

Th(NO3)4 Yes Th, daughters

Th phosphate Yes Th, daughters

Tracers in cement Ba-140, La-140

Tracers in cement Y-91

Tracers in cement: tagged Ag*IO3 Ag-110

1 U metal Yes U, daughters

U alloys Yes U, daughters

U3O8 Yes U, daughters

UO3 (part of production support R&D from UNH) U, daughters

UO2 (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

UNH U, daughters

UO2CO3·2(NH4)2CO3 (uranyl ammonium carbonate) Yes U, daughters

U2O7(NH4)2 (ammonium diuranate) U, daughters

U2O7Na2 (soda salt) Yes U, daughters

U beryllides U, daughters

U borides U, daughters

U carbides, mainly UC (natural; enriched 9-93%) Yes NU/EU, daughters

UF4 U, daughters

U fluorides (except UF4) (salt solution) U, daughters

U hydrosulfides (salt solution) U, daughters

U nitrides, mainly UN (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

U phosphates (salt solution) U, daughters

U silicides U, daughters

U sulfates (salt solution) U, daughters

U sulfides U, daughters

Th metal (including casting by vacuum melting) Yes Th, daughters

Th alloys (except with U) Yes Th, daughters

Th carbides Th, daughters

Th fluorides Th, daughters

Th halide Th, daughters

Th iodide (used to make Th-U alloy) Yes Th, daughters

ATTACHMENT F

RADIONUCLIDES AND CHEMICAL FORMS BY BUILDING

Page 1 of 5
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Bldg Compound or alloya

Large 

amounts 

handled? RNb

ThN Yes Th, daughters

Th(NO3)4 Yes Th, daughters

Th phosphate Yes Th, daughters

Th-U alloys (5, 10, 15, 20 weight percent U) Yes Th, some U, daughters

Th-U carbides (casting by arc melting) Th, some U, daughters

Tracers in cement P-32

Tracers in cement Ba-140, La-140

Tracers in cement Y-91

Tracers in cement: tagged Ag*IO3 Ag-110

2 U metal Yes U, daughters

U alloys Yes U, daughters

U3O8 (depleted; natural; enriched, including HEU) Yes DU/NU/EU, daughters

UO3 (part of production support R&D from UNH) U, daughters

UO2 (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

UNH U, daughters

UO2CO3·2(NH4)2CO3 (uranyl ammonium carbonate) Yes U, daughters

U2O7(NH4)2 (ammonium diuranate) U, daughters

U2O7Na2 (soda salt) Yes U, daughters

U beryllides U, daughters

U borides U, daughters

U carbides, mainly UC (natural; enriched 9-93%) Yes NU/EU, daughters

UCl (enriched) U, daughters

UF4 U, daughters

U nitrides, mainly UN (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

U silicides U, daughters

U sulfides U, daughters

Th metal Yes Th, daughters

Th alloys (except with U) Yes Th, daughters

ThO2 alloys (except with U) Yes Th, daughters

Th carbides Th, daughters

Th fluorides Th, daughters

Th halide Th, daughters

Th iodide (used to make Th-U alloy) Yes Th, daughters

ThN Yes Th, daughters

Th phosphate Yes Th, daughters

Th-U alloys (5, 10, 15, 20 weight percent U) Yes Th, minor U, daughters

Th-U carbides (casting by arc melting) Yes Th, some U, daughters

Fissium ("synthetic fissium alloy") (unirrad'd U) EU, daughters

3 U metal Yes U, daughters

U alloys Yes U, daughters

U3O8 (depleted; natural; enriched, including HEU) Yes DU/NU/EU, daughters

UO3 (part of production support R&D from UNH) U, daughters

UO2 (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

UNH U, daughters

U beryllides U, daughters

U borides U, daughters

U carbides, mainly UC (natural; enriched 9-93%) Yes NU/EU, daughters

UCl (enriched) U, daughters

UF4 U, daughters

U hydrides Yes U, daughters

U nitrides, mainly UN (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

U silicides U, daughters

U sulfides U, daughters

Th metal Yes Th, daughters

Th alloys (except with U) Yes Th, daughters

ThO2 alloys (except with U) Th, daughters

ATTACHMENT F

RADIONUCLIDES AND CHEMICAL FORMS BY BUILDING

Page 2 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 98 of 124

Bldg Compound or alloya

Large 

amounts 

handled? RNb

Th carbides Th, daughters

Th fluorides Th, daughters

Th halide Th, daughters

Th iodide (used to make Th-U alloy) Yes Th, daughters

ThN Yes Th, daughters

Th phosphate Yes Th, daughters

Th-U alloys (5, 10, 15, 20 weight percent U) Yes Th, minor U, daughters

Th-U carbides (casting by arc melting) Yes Th, some U, daughters

Fissium ("synthetic fissium alloy") (unirrad'd U) EU, daughters

4 U metal U, daughters

U alloys (natural; enriched 10%) NU/EU, daughters

U3O8 (depleted; natural; enriched, including HEU) DU/NU/EU, daughters

UO3 (part of production support R&D from UNH) U, daughters

UO2 depleted; (natural; enriched 5-93%) Yes DU/NU/EU, daughters

U fuel specimens, irradiated (usually EU, probably usually 

UO2)

Yes NU/EU, daughters, some MFP

UNH U, daughters

U2O7(NH4)2 (ammonium diuranate) U, daughters

U beryllides U, daughters

U borides U, daughters

U carbides, mainly UC (natural; enriched 9-93%) Yes NU/EU, daughters

UCl (enriched) EU, daughters

UF4 U, daughters

U nitrides, mainly UN (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

U silicides U, daughters

U sulfides U, daughters

Gaseous fission product, daughter counting Kr, Xe, I, daughters

Th metal Th, daughters

Th alloys (except with U) Th, daughters

ThO2 alloys (except with U) Th, daughters

Th carbides Th, daughters

Th fluorides Th, daughters

Th halide Th, daughters

Th iodide (used to make Th-U alloy) Th, daughters

ThN Th, daughters

Th(NO3)4 Th, daughters

Th phosphate Th, daughters

Th-U alloys (5, 10, 15, 20 weight percent U) Th, minor U, daughters

Th-U carbides (casting by arc melting) Th, some U, daughters

Fissium ("synthetic fissium alloy") (unirrad'd U) DU, daughters

5 U metal Yes U, daughters

U alloys (natural; enriched 10%) Yes NU/EU, daughters

UO2 (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

U fuel, irradiated (usually EU, probably usually UO2) Yes NU/EU, daughters, some MFP

U beryllides U, daughters

U borides U, daughters

U carbides, mainly UC (natural; enriched 9-93%) Yes NU/EU, daughters

U nitrides, mainly UN (depleted; natural; enriched 5-93%) Yes DU/NU/EU, daughters

U silicides U, daughters

U sulfides U, daughters

Th metal Yes Th, daughters

Th alloys (except with U) Yes Th, daughters

ThO2 alloys (except with U) Yes Th, daughters

Th carbides Yes Th, daughters

Th fluorides Th, daughters

Th halide Th, daughters

Th iodide (used to make Th-U alloy) Th, daughters

ThN Yes Th, daughters

ATTACHMENT F

RADIONUCLIDES AND CHEMICAL FORMS BY BUILDING

Page 3 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 99 of 124

Bldg Compound or alloya

Large 

amounts 

handled? RNb

Th phosphate Th, daughters

Th-U alloys (5, 10, 15, 20 weight percent U) Yes Th, minor U, daughters

Th-U carbides (casting by arc melting) Yes Th, some U, daughters

6, 7 U metal Yes U, daughters

U alloys (natural; enriched 10%) Yes NU/EU, daughters

U3O8 (depleted; natural; enriched, including HEU) DU/NU/EU, daughters

UO3 (part of production support R&D from UNH) U, daughters

UO2 (enriched, mostly at 93%) Yes EU, daughters

UO2 particles, irradiated (enriched?) Yes EU; MFP

U2O7(NH4)2 (ammonium diuranate) U, daughters

U beryllides U, daughters

U borides U, daughters

U carbides, mainly UC (natural; enriched 9-93%) NU/EU, daughters

UCl (enriched) EU, daughters

UF4 (probably only Building 7) Yes U, daughters

U nitrides, mainly UN (depleted; natural; enriched 5-93%) DU/NU/EU, daughters

U silicides U, daughters

U sulfides U, daughters

Th metal Yes Th, daughters

Th alloys (except with U) Yes Th, daughters

ThO2 alloys (except with U) Yes Th, daughters

Th carbides Th, daughters

Th fluorides Yes Th, daughters

Th halide Th, daughters

Th iodide (used to make Th-U alloy) Th, daughters

ThN Yes Th, daughters

Th(NO3)4 Th, daughters

Th phosphate Th, daughters

Th-U alloys (5, 10, 15, 20 weight percent U) Yes Th, minor U, daughters

Th-U carbides (casting by arc melting) Th, some U, daughters

Spent fuel in solid or liquid form (recovery studies; mostly 

unirradiated)

Yes U, Th; MFP6, AP

Pu alloys, UN-PuN (Building 6 only) Yes Pu-239, HEU

Pu recovery from U spent fuel Yes Pu-239, HEU 

PuO2 volatility (very limited benchtop study) Pu-239

Fissium ("synthetic fissium alloy") (unirrad'd U)

Tracers: CePr (rare-earth study) Yes Probably Ce/Pr-144

Radiotracer study of activation (Mg sulfate solution) Fe-59, Mn-54

Radiotracer study of activation (Fe56(n,p)Mn56) Mn-56

Tracers in cement P-32

Tracers in cement Ba-140, La-140

Tracers in cement Y-91

Tracers in cement: tagged Ag*IO3 Ag-110

Triated cytidine, trace levels in biological studies H-3

Unspecified labeling (Building 7) I-125

9 U, form unknown (depleted; natural) DU/NU, daughters

Th, form unknown (minor amounts found in D&D) Th, daughters

Krypton gas studies (lab-level amounts) Kr-85

West Jefferson site

JN-1 U metal (natural; enriched) Yes NU/EU, MFP

U alloys (natural; enriched 10%) Yes NU/EU, MFP

U3O8 (depleted; natural; enriched, including HEU) DU/NU/EU, MFP

UO2 (enriched 25-93%, mostly at 93%) Yes EU, MFP

UO2 (one specific project) Yes U; MFP up to 6 mCi/g

U carbides, mainly UC (natural; enriched 10-93%) Yes NU/EU, MFP

U hydrides U, MFP

U nitrides, mainly UN (depleted; natural; enriched 5-93%) Yes U, MFP

U silicides U, MFP

ATTACHMENT F

RADIONUCLIDES AND CHEMICAL FORMS BY BUILDING

Page 4 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 100 of 124

Bldg Compound or alloya

Large 

amounts 

handled? RNb

U sulfides U, MFP

Spent U fuel in solid or liquid form in recovery studies Yes U; MFP, AP

U-Th alloys Yes Th, U, MFP, Cs-137

Spent U-Th fuel in solid or liquid form in recovery studies Yes U, Th; MFP, AP

Al, Nb, Zr, SS as cladding, matrix, or alloy component Yes AP, incidental MFP

Miscellaneous spent fuel forms Yes U, MFP

PuO2, UO2-PuO2 Pu-239, U, MFP

UN-PuN Pu-239, U, MFP

Pu, Np recovery from Np rods irradiated in commercial 

reactor

Np, Pu-238, MFP

Pm oxide (RTG work) Pm-147

Cf-252 wire Yes Cf-252

Co-60 research (probably as the metal and/or in activation 

materials)

Yes Co-60

JN-2 U metal foil (depleted; natural) (used in zero power 

experiments)

Yes DU/NU, MFP

UO2 (2-93% enrichment) (used in zero power experiments) Yes EUO2, MFP

JN-2 Pu source study (encapsulated form only, probably as 

PuO2)

Yes Pu-239

JN-3 U metal (natural) Yes U, MFP, AP

U alloys (natural; enriched 10%) Yes NU/EU, MFP, AP

UO2 (natural; enriched 25-93%) Yes NU/EU, MFP, AP

U carbides, mainly UC (natural; enriched 10-93%) Yes NU/EU, MFP, AP

U nitrides, mainly UN (depleted; natural; enriched 5-93% ) Yes DU/NU/EU, MFP, AP

Encapsulated U fuel (probably enriched; in in-pile gas loop 

facility)

Yes Fission gases

Nonfuel reactor structural and associated components 

(radiation effects studies)

Yes AP (steel, Al, etc.)

Th-U alloys U, Th, MFP

PuO2 (encapsulated form only) Pu, U, MFP

PuN, UN-PuN (all encapsulated form only; probably all U 

was enriched)

Pu, EU, MFP

Pm oxide (encapsulated only, as irradiation test of RTG) Pm-147

JN-4 Pu metal (e.g., >20 kg of Pu metal fabricated into various 

shapes for classified work)

Yes Mostly Pu-239

PuO2 (e.g., 1 kg of UO2-PuO2 fabricated into fuel pellets) Yes Mostly Pu-239

PuO2 (~4 kg of PuO2 made into pellets, sintered, welded into 

capsules for use as heat sources)

Yes 80% Pu-238; probably 16% Pu239, 1.2 ppm Pu-236

PuO3 Mostly Pu-239

Pu hydride (mostly formed as intermediate product, but 

some scrap formed)

Pu-239

PuN, PuUN (e.g., ~20 kg of Pu metal powder processed into 

fuel pins: Pu or U crushed, hydrated, nitrated, and reducing 

to nitride; PuN/UN pressed into pellets, loaded into fuel pins)

Yes 90% w/o Pu-239; 93% w/o U-235

U-Pu alloys (probably enriched U) Yes Pu-239, EU, daughters

Th-Pu alloys Pu-239, Th, daughters

Pu reprocessing study : Pu as 1 kg of Pu nitrate solution Yes Pu-239, Pu-238

Pm oxide Yes Pm-147

Np microsphere preparation Yes Np

>1 kg of NpO2 fabricated into NpO2-ZrO2 and NpO2-CaOZrO2 target rodsYes Np

JS-1 DU, NU, EU (assumed as UO2) Yes DU/NU/EU daughters

Th (assume as ThO2) Th, daughters

JS-10 DU, EU (assume as UO2) Yes DU/EU

JS-12 DU, EU (assume as UO2) DU/EU

a. The U form was probably natural U when not specified; enriched U appears not to have been used before 1954. 

Sometimes only the final form was indicated by the relevant reference; thus, the possibility of a different initial or 

intermediate form than listed above must be kept in mind.

b. MFP: mixed fission products; FP: fission products; AP: activation products (for the materials in question).

ATTACHMENT F

RADIONUCLIDES AND CHEMICAL FORMS BY BUILDING

Page 5 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 101 of 124

Bldga Descriptionb

Operations Photon Electron

Reference

Location 

Begin End in Ref

Energy, 

keV %

Energy, 

keV %

A Metallurgical R&D on U, including 

alloyingc

1943 [1962] 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

A R&D, fabrication of nat/depl U as 

pellets, powder, or plates

1950 1971 30-250 100 >15 100 ORAUT 2006b p. 23

A R&D, fabrication of enriched U as 

pellets, powder, or plates

1954 1962 30-250 100 >15 100 ORAUT 2006b p. 23

A Extracting U from phosphates, slimes, 

slurries (except 1951)

1946 1954 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

A Extracting U from carnotite ores 1947 1949 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

A Extracting U from Western ores 1952 1955 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

A Extracting U by ammonium carbonate 

leaching

1955 1957 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

A Converting pitchblende to UO2 1950 1951 30-250

>250

50

50

>15 100 ORAUT 2004 Table 6-9

A Refining MgX (concentrate), V-20 soda 

salt to UO2

1950 1951 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

A Process support for U metal production 

(including casting)

1953 1959 30-250

>250

70

30

>15 100 ORAUT 2006b p. 23

A Metallurgical R&D on Th, including 

alloying

1944 1961 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

A Extracting Th from monazite sands, etc. 

(except 1953)

1949 1955 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

A Radiotracer work - industrial (certain 

labs) (e.g., P-32, Ba-140, La-140, Y-91, 

and Ca-45 in cement; Fe-59, Mn-54 in 

industrial control; tagged AgIO)

1953 1960 30-250 100 >15 100

1 Rolling, machining, fabrication of U 

metal, other forms

1943 1980 30-250 100 >15 100 ORAUT 2009b; 

2006b

p. 37; 23

1 Casting and other heating and melting 

operations on U metal 

1943;

1979

1962;

1979

30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

1 Metallurgical R&D on U, including 

alloying

1943 [1962] 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

1 Extracting U from phosphates, slimes, 

slurries (except 1951)

1946 1954 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

1 Extracting U from carnotite ores 1947 1949 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

1 Extracting U from Western ores 1952 1955 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

1 Extracting U by ammonium carbonate 

leaching

1955 1957 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

1 Refining MgX (concentrate), V-20 soda 

salt to UO2

1950 1951 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

1 Process support for U metal production 

(including casting)

1953 1959 30-250

>250

70

30

>15 100 ORAUT 2006b p. 23

1 Rolling, machining, casting, fabrication 

of Th metal, other forms

1944 1980 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

1 Metallurgical R&D, including alloying Th 1944 1961 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

1 Radiotracer work - industrial (certain 

labs) (e.g., P-32, Ba-140, La-140, Y-91, 

and Ca-45 in cement; Fe-59, Mn-54 in 

industrial control; tagged AgIO)

1953 1960 30-250 100 >15 100

2 Rolling, machining, fabrication of U 

metal, other forms

1950 1980 30-250 100 >15 100 ORAUT 2009a, 

2009b

p. 23; 37

2 Casting and other heating and melting 

operations on U

1950 1980 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

ATTACHMENT G

PHOTON AND ELECTRON ENERGY RANGES

Page 1 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 102 of 124

Bldga Descriptionb

Operations Photon Electron

Reference

Location 

Begin End in Ref

Energy, 

keV %

Energy, 

keV %

2 Metallurgical R&D on U, including 

alloying

1950 [1962] 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

2 Metallurgical R&D on U, including 

alloying

1979 1979 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

2 R&D, fabrication of nat/depl U as 

pellets, powder, or plates

1950 1971 30-250 100 >15 100 ORAUT 2006b p. 23

2 R&D, fabrication of enriched U as 

pellets, powder, or plates

1954 1962 30-250 100 >15 100 ORAUT 2006b p. 23

2 Extracting U from phosphates, slimes, 

slurries (except 1951)

1950 1954 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

2 Extracting U from Western ores 1952 1955 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

2 Extracting U by ammonium carbonate 

leaching

1955 1957 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

2 Refining MgX (concentrate), V-20 soda 

salt to UO2

1950 1951 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

2 Process support for U metal production 1953 1959 30-250

>250

70

30

>15 100 ORAUT 2006b p. 23

2 Rolling, machining, casting, fabrication 

of Th metal, other forms

1950 1980 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

2 Metallurgical R&D on Th, including 

alloying

1950 1961 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

2 Metallurgical R&D on Th, including 

alloying

1971 1972 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

2 Preparing Th in Th iodide process 1951 1954 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

2 R&D on Th-U alloys 1960 1965 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

3 Metallurgical R&D on U, including 

alloying

1950 [1962] 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

3 Metallurgical R&D on U, including 

alloying

1979 1979 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

3 Production of U forms for other facilities, 

including melting and powder formation

1954 1975 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

3 R&D, fabrication of nat/depl U as 

pellets, powder, or plates

1950 1971 30-250 100 >15 100 ORAUT 2006b p. 23

3 R&D, fabrication of enriched U as 

pellets, powder, or plates

1954 1962 30-250 100 >15 100 ORAUT 2006b p. 23

3 Calcining of U samples and waste for 

recovery

1954 1975? 30-250

>250

70

30

>15 100 ORAUT 2006b p. 23

3 R&D on Th-U alloys 1960 1965 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

3 Storage and dispensing of SNM, incl EU 

and Thd,e

1950 1975 30-250 100 >15 100 ORAUT 2009b; 

2006b

p. 37; 23

4 R&D, fabrication with enriched UO2 in 

pellet, powder, and plate forms (not the 

Radioisotope Lab)

1954 1962 30-250 100 >15 100 ORAUT 2006b p. 23

4 High-level radiochemistry (mostly fission 

products)

1953 1975 30-250

>250

75

25

>15 100

4 Low-level radiochemistry (mostly fission 

products)

1953 1975 30-250

>250

75

25

>15 100

4 Gaseous fission product counting, 

radiochemistryf

1956 1966 30-250 100 >15 100

4 Radiotracer work - industrial 1953 1960 30-250 100 >15 100

4 Radiography of material samples (Cs137, Co-60)1953 1975 30-250

>250

25

75

>15 100 ORAUT 2004 Table 6-9

4 Activation using Po-Be source 1953? 1975 30-250

>250

25

75

>15 100 ORAUT 2004 Table 6-9

ATTACHMENT G

PHOTON AND ELECTRON ENERGY RANGES

Page 2 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 103 of 124

Bldga Descriptionb

Operations Photon Electron

Reference

Location 

Begin End in Ref

Energy, 

keV %

Energy, 

keV %

5 Rolling, machining, fabrication of U 

metal, other forms

1953 1980b 30-250 70 >15 100 ORAUT 2009b; 

2006b

p. 37; 23

5 Casting and other heating and melting 

operations on U

1953 1980b 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

5 R&D, fabrication of nat/depl U as 

pellets, powder, or plates

1953 1971 30-250 100 >15 100 ORAUT 2006b p. 23

5 R&D, fabrication of enriched U as 

pellets, powder, or plates

1954 1962 30-250 100 >15 100 ORAUT 2006b p. 23

5 Process support for U metal production 

(including casting)

1953 1959 30-250

>250

70

30

>15 100 ORAUT 2006b p. 23

5 Rolling, machining, casting, fabrication 

of Th metal, other forms

1953 1980 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

5 R&D on Th-U alloys 1960 1965 30-250

>250

25

75

>15 100

6 Metallurgical R&D on U, including 

alloying

1954 1962 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

6 Metallurgical R&D on Th, including 

alloying

1954 1961 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

6 Metallurgical R&D on Th, including 

alloying

1971 1972 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

6 R&D on Th-U alloys 1960 1965 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

6 Low-level radiochemistry (mostly fission 

products)

1954 1975 30-250

>250

75

25

>15 100

6 Co-60 irradiationsg 1959 1975 30-250

>250

25

75

--- --- ORAUT 2010b Table 6-2

6 R&D on Pu-239 alloys and other Pu 

forms

1954 1955 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photon

100 ORAUT 2009b p. 37

7 Metallurgical R&D on U, including 

alloying

1954 1962 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

7 (7A) Research, pilot U fluoridation 1954 1954 30-250

>250

50

50

>15 100 ORAUT 2005d Table 6-10

7 Metallurgical R&D on Th, including 

alloying

1954 1961 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

7 Low-level radiochemistry (mostly fission 

products)

1954 1975 30-250

>250

75

25

>15 100

7 Radiotracer work - biological/chemical 

(H-3, tritiated cytidine), C-14, P-32, I125)

1954? 1975? 30-250 100 >15 100

9 Examination of krypton from test fuel 

rods (gases extracted at JN-1) or other 

samplesf

1958 1962? 30-250 100 >15 100

JN-1 Cell preparation, cleaning, maintenance: 

irradiated U fuel specimens

1956 1971 30-250

>250

75

25

>15 100 ORAUT 2010b Table 6-2

JN-1 Cell preparation, cleaning, maintenance: 

spent fuel sectioning

1956? 1980? 30-250

>250

75

25

>15 100 ORAUT 2010b Table 6-2

JN-1 Cell preparation, cleaning, maintenance: 

examination of irradiated reactor nonfuel parts

1958 1975 30-250 100 >15 100

JN-1 Cell preparation, cleaning, maintenance: 

irradiated U-Th fuel specimens (>90% 

Th)

1956 1961 30-250

>250

25

75

>15 100 ORAUT 2005d Table 6-10

JN-1 Cell preparation, cleaning, maintenance: 

sectioning of Co-60 rods

1971 1982? 30-250

>250

25

75

>15 100 ORAUT 2010b Table 6-2

ATTACHMENT G

PHOTON AND ELECTRON ENERGY RANGES

Page 3 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 104 of 124

Bldga Descriptionb

Operations Photon Electron

Reference

Location 

Begin End in Ref

Energy, 

keV %

Energy, 

keV %

JN-1 Cell preparation, cleaning, maintenance: 

fabrication of Cf-252 sources (two AGC 

cells)

1972 1974? <30 100 >15 100

JN-1 Co-60 irradiations (HEC) 1972? 1982 30-250

>250

25

75

--- --- ORAUT 2010b Table 6-2

JN-1 Cask movements 1955 1983 30-250

>250

25

75

--- ---

JN-1 Cleaning/maintenance in non-Cf cells 1956 1983 30-250

>250

75

25

>15 100 ORAUT 2010b Table 6-2

JN-1 Cell preparation, cleaning, maintenance: 

irradiated Pu, U-Pu forms (except 1968-

1969)

1967 1972 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photon

100 ORAUT 2009b p. 37

JN-1 Cell prep, cleaning, maintenance: 

irradiated Pu fuel rods, recovery of Pu238, Np-237 (Low Level Cell)

1974? 1975 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photon

100

JN-1 Study of irradiated U, Pu fuel aerosols 

(Pu-239?)

1975 1977 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photon

100 ORAUT 2009b p. 37

JN-1 Sabotage project work in HEC, pool 

(spent fuel)

1979 1983 30-250

>250

75

25

>15 100 ORAUT 2010b Table 6-2

JN-1 Waste handling (mostly MFPs) 1955 1983 <30

30-250

25

75

>15 100 ORAUT 2010b; 

2007a

Table 6-2; 

6-9

JN-2 Critical assembly (zero power facility) 

operations

1955 1963 30-250

>250

50

50

>15 100 ORAUT 2010b Table 6-2

JN-2 Research, fabrication with encapsulated 

Pu-239 (direct conversion devices) for 

LLNL

1964 1970 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photonf

100f ORAUT 2009b p. 37

JN-2 Storage and dispensing of SNM, incl 

EU, The

1955? 1959 30-250 100 >15 100 ORAUT 2009b; 

2006b

p. 37; 23

JN-2 Storage and dispensing of SNM, incl 

EU, Pu, Thd

1960 1996 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photon

100 ORAUT 2009b; 

2006b

p. 37; 23

JN-3 Routine/typical reactor operator 

activities

1956 1974 30-250

>250

25

75

>15 100 ORAUT 2009b, 

2010b, 2007a

JN-3 Reactor maintenance activities (reactor 

down)

1956 1974 30-250

>250

25

75

>15 100

JN-3 Reactor irradiation/activation work 1956 1974 30-250

>250

25

75

--- ---

JN-3 Reactor beam area work 1956 1974 30-250

>250

25

75

--- ---

JN-3 Reactor thermal column/shielding plate 

work

1956 1974 30-250

>250

25

75

--- ---

JN-3 Reactor shielding pool, maintenance 

work in

1956 1974 30-250

>250

25

75

>15 100

JN-4 Alloying/mixing Pu with various 

materials, incl U, Th

1960 1977 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photong

100

g

ORAUT 2009b p. 37

ATTACHMENT G

SELECTION OF BETA AND PHOTON RADIATION ENERGIES AND PERCENTAGES

Page 5 of 6

ATTACHMENT G

PHOTON AND ELECTRON ENERGY RANGES

Page 4 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 105 of 124

Bldga Descriptionb

Operations Photon Electron

Reference

Location 

Begin End in Ref

Energy, 

keV %

Energy, 

keV %

JN-4 Other research with Pu-239 (corrosion, 

etc.)

1960 1977 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photong

100

g

ORAUT 2009b p. 37

JN-4 Preparation of UCN-PuCN by arc 

melting, casting

1967 1967 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photong

100

g

ORAUT 2009b p. 37

JN-4 Fabrication of UN-PuN (Pu-239) pellets 

and pins, including heat treatment and 

rolling

1968 1971 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photong

100

g

ORAUT 2009b p. 37

JN-4 Research, fabrication of RTGs with Pu2381974 1977 <30

30-250

65

35

>15g 100

g

JN-4 R&D on Po-210 alpha-cell direct 

conversion generator

1963? 1964 <30

30-250

50

50

>15g 100

g

ORAUT 2004 Table 6-9

JN-4 Purification of multicuries of Pm-147 for 

RTGs

1964 1966 --- --- >15g 100

g

JN-4 Research, fabrication of RTGs with Cm2441964 1968 <30

30-250

>250

25

50

25

>15g 100

g

ORAUT 2010b Table 6-2

JN-4 Preparation of Np oxide microspheres 1966 1967? <30

30-250

80

20

>15g 100

g

JN-4 Fabricating NpO2 targets for Pu-238 

production (Np-237)

1971? 1974? <30

30-250

80

20

>15g 100

g

JN-4 Pu waste handling (mostly Pu-239) 1960 1977 <30

30-250

65

35

>15 keV 

electron 

or <30 

keV 

photon

100 ORAUT 2009b p. 37

JS-1 Production of fuel using the hot isostatic 

pressure bonding process (HEU plates)

1965? 1980 30-250 100 >15 100 ORAUT 2006b,e p. 23; 

Table 6-13

JS-10, 

JS-12

Explosive forming, ballistic studies 

(mostly DU, some EU)

1979? 1988? <30

30-250

90

10

>15 100 ORAUT 2007a Table 6-9

a. Only Buildings A and 1 existed before the 1950s, when Buildings 2 through 9 were built. It is known that Buildings 2 and 3 were in use 

in 1950. Thus, Buildings 2 and 3 should not be assumed to have been in use before 1950, Buildings 4 and 5 before about 1953, and 

Buildings 6, 7, and 9 before about 1954. It is not clear when JS-1 was built, but it likely was not in operation before 1965.

b. Where it is stated that a year or period is excepted, it means that no work of this type was done during that year or period.

c. Per ORAUT (2006b p. 23), during casting, any U-238 decay products float to the top surface of the molten metal and condense as 

surface residues. The Pa-234 has high beta and photon energies; Pa-234 has 93% of its photons at >250 keV, while U-238 in 

equilibrium with Th-234m, Pa-234m, and Pa-234 has 18% of its photons at >250 keV. Thus, in processing that involves melting, a 

higher percentage of photons with energies greater than 250 keV should be assumed.

d. Some parts of Building 3, probably the vault, were in use until 1991; some intermittent machining of uranium forms was done from 1981 

to 1991 in the Building 5 machine shop.

e. The vault in Building 3 probably contained more EU than anything else, as probably did the vault in JN-2 until the critical assembly 

facility shut down in 1963; after that, the JN-2 vault likely contained more Pu than any other material.

f. See footnote to table in Attachment C for a description of the fission gas analysis work.

g. These figures are assumed to be applicable to a bare or wall-shielded source. However, at Battelle, these Co-60 rods were actually 

covered by water during irradiations, producing a minimal dose rate at the worker position. It is assumed that most of the Battelle dose 

resulted from movements or changes of the rods, in which less water covered the sources, but some adjustments might still need to be 

made for the water shielding.

g. This was glovebox work, hence betas should not be included except for periods when the glovebox was assumed to be opened, gloves 

were changed, or wastes were removed. Such periods should be assumed to total <10% of work time.

ATTACHMENT G

PHOTON AND ELECTRON ENERGY RANGES

Page 5 of 5
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 106 of 124

Bldg Descriptiona

Operations Neutron

Reference

Location 

Begin End in Ref b Energy %

A Metallurgical R&D on U, including alloying 1943 [1962] For alpha-neutron 

reactions, UO2 may be 

taken to be typical

A R&D, fabrication of nat/depl U as pellets, powder, or 

plates

1950 1971 --- ---

A R&D, fabrication of enriched U as pellets, powder, or 

plates

1954 1962 --- ---

A Extracting U from phosphates, slimes, slurries (except 

1951)

1946 1954 --- ---

A Extracting U from carnotite ores 1947 1949 --- ---

A Extracting U from Western ores 1952 1955 --- ---

A Extracting U by ammonium carbonate leaching 1955 1957 --- ---

A Converting pitchblende to UO2 1950 1951 0.1-2.0 MeV

2-20 MeV

50

50

ORAUT 2004 Table 6-9

A Refining MgX (concentrate), V-20 soda salt to UO2 1950 1951 --- ---

A Process support for U metal production (including 

casting)

1953 1959 --- ---

A Metallurgical R&D on Th, including alloying 1944 1961 --- ---

A Extracting Th from monazite sands, etc. (exc 1953) 1949 1955 --- ---

A Radiotracer work - industrial (certain labs only) 1953 1960 --- ---

1 Rolling, machining, fabrication of U metal, other forms 1943 1980 --- ---

1 Casting and other heating and melting operations on 

U

--- ---

1 Metallurgical R&D on U, including alloying 1943 [1962]

1 Extracting U from phosphates, slimes, slurries (exc 

1951)

1946 1954 --- ---

1 Extracting U from carnotite ores 1947 1949 --- ---

1 Extracting U from Western ores 1952 1955 --- ---

1 Extracting U by ammonium carbonate leaching 1955 1957 --- ---

1 Refining MgX (concentrate), V-20 soda salt to UO2 1950 1951 --- ---

1 Process support for U metal production (including 

casting)

1953 1959 --- ---

1 Rolling, machining, casting, fabrication of Th metal, 

other forms

1944 1980 --- ---

1 Metallurgical R&D, including alloying Th 1944 1961 --- ---

1 Radiotracer work - industrial (certain labs) 1953 1960 --- ---

2 Rolling, machining, fabrication of U metal, other forms 1950 1980 --- ---

2 Casting and other heating and melting operations on 

U

1950 1980 --- ---

2 Metallurgical R&D on U, including alloying 1950 [1962] For alpha-neutron 

reactions, UO2 may be 

taken to be typical

2 Metallurgical R&D on U, including alloying 1979 1979 For alpha-neutron 

reactions, UO2 may be 

taken to be typical

2 R&D, fabrication of nat/depl U as pellets, powder, or 

plates

1950 1971 --- ---

2 R&D, fabrication of enriched U as pellets, powder, or 

plates

1954 1962 --- ---

2 Extracting U from phosphates, slimes, slurries (except 

1951)

1950 1954 --- ---

2 Extracting U from Western ores 1952 1955 --- ---

2 Extracting U by ammonium carbonate leaching 1955 1957 --- ---

2 Refining MgX (concentrate), V-20 soda salt to UO2 1950 1951 --- ---

2 Process support for U metal production 1953 1959 --- ---

2 Rolling, machining, casting, fabrication of Th metal, 

other forms

1950 1980 --- ---

ATTACHMENT H

NEUTRON ENERGY RANGES

Page 1 of 4
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 107 of 124

Bldg Descriptiona

Operations Neutron

Reference

Location 

Begin End in Ref b Energy %

2 Metallurgical R&D on Th, including alloying 1950 1961 --- ---

2 Metallurgical R&D on Th, including alloying 1971 1972 --- ---

2 Preparing Th in Th iodide process 1951 1954 --- ---

2 R&D on Th-U alloys 1960 1965

3 Metallurgical R&D on U, including alloying 1950 [1962] For alpha-neutron 

reactions, UO2 may be 

taken to be typical

3 Metallurgical R&D on U, including alloying 1979 1979 For alpha-neutron 

reactions, UO2 may be 

taken to be typical

3 Production of U forms for other facilities, including 

melting and powder formation

1954 1975 --- ---

3 R&D, fabrication of nat/depl U as pellets, powder, or 

plates

1950 1971 --- ---

3 R&D, fabrication of enriched U as pellets, powder, or 

plates

1954 1962 --- ---

3 Calcining of U samples and waste for recovery 1954 1975? --- ---

3 R&D on Th-U alloys 1960 1965

3 Storage and dispensing of SNM, incl EU and Thb,d 1950 1975 0.1-2 MeV

2-14 MeV

79

21

ORAUT 2006b p. 27

4 R&D, fabrication with enriched UO2 in pellet, powder, 

and plate forms (not the Radioisotope Lab)

1954 1962 --- ---

4 High-level radiochemistry (mostly fission products) 1953 1975 --- ---

4 Low-level radiochemistry (mostly fission products) 1953 1975 --- ---

4 Gaseous fission product counting, radiochemistrye 1956 1966 --- ---

4 Radiotracer work - industrial 1953 1960 --- ---

4 Radiography of material samples (Cs-137, Co-60) 1953 1975 --- ---

4 Activation using Po-Be source 1953? 1975 0.1-2.0 MeV

2-20 MeV

25

75

ORAUT 2004; 

2007a

Table 6-9; 

6-16

5 Rolling, machining, fabrication of U metal, other forms 1953 1980c --- ---

5 Casting and other heating and melting operations on 

U

1953 1980c --- ---

5 R&D, fabrication of nat/depl U as pellets, powder, or 

plates

1953 1971 --- ---

5 R&D, fabrication of enriched U as pellets, powder, or 

plates

1954 1962 --- ---

5 Process support for U metal production (including 

casting)

1953 1959 --- ---

5 Rolling, machining, casting, fabrication of Th metal, 

other forms

1953 1980 --- ---

5 R&D on Th-U alloys 1960 1965

6 Metallurgical R&D on U, including alloying 1954 1962 For alpha-neutron 

reactions, UO2 may be 

taken to be typical

6 Metallurgical R&D on Th, including alloying 1954 1961 --- ---

6 Metallurgical R&D on Th, including alloying 1971 1972 --- ---

6 R&D on Th-U alloys 1960 1965 --- ---

6 Low-level radiochemistry (mostly fission products) 1954 1975 --- ---

6 Co-60 irradiations 1959 1975 --- ---

6 R&D on Pu-239 alloys and other Pu forms (taken to 

be Pu processing not in a glovebox

1954 1955 10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

7 Metallurgical R&D on U, including alloying 1954 1962 For alpha-neutron 

reactions, UO2 may be 

taken to be typical

7 (7A) Research, pilot U fluoridation (consider to be as for 

UF4 (green salt))

1954 1954 0.1-2 MeV 100 ORAUT 2005d Table 6-10

7 Metallurgical R&D on Th, including alloying 1954 1961 --- ---

ATTACHMENT H

NEUTRON ENERGY RANGES

Page 2 of 4
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 108 of 124

Bldg Descriptiona

Operations Neutron

Reference

Location 

Begin End in Ref b Energy %

7 Low-level radiochemistry (mostly fission products) 1954 1975 --- ---

7 Radiotracer work - biological/chemical (assume H-3, 

C-14, P-32, I-125)

1954? 1975? --- ---

9 Examination of krypton) from test fuel rods (gases 

extracted at JN-1) and other samplese

1958 1962? --- ---

JN-1 Cell preparation, cleaning, maintenance: irradiated U 

fuel specimens

1956 1971 0.1-2 MeV 100

JN-1 Cell preparation, cleaning, maintenance: spent fuel 

sectioning (nondestructive examinations performed on 

the assemblies, bundles, rods; rods (encased in 

cladding) were removed; fission gases were collected 

and analyzed from a hole drilled in the cladding. 

These examinations: ~ 90 percent of the work 

performed in the HEC)

1956? 1980? 0.1-2 MeV 100 Wastren 2001

JN-1 Cell preparation, cleaning, maintenance: examination 

of irradiated reactor non-fuel parts

1958 1975 --- ---

JN-1 Cell preparation, cleaning, maintenance: irradiated UTh fuel specimens (>90% Th) (93.5 w/o ThO2-6.5 w/o 

UO2 (U was 92.7% enriched); core had 44 g of 95.5 

w/o ThO2-4.5 w/o UO2)

1956 1961 0.1-2 MeV 100 Dayton and Tipton 

1959; Peterson et 

al. 1959

JN-1 Cell preparation, cleaning, maintenance: sectioning of 

Co-60 rods

1971 1982? --- ---

JN-1 Cell preparation, cleaning, maintenance: fabrication of 

Cf-252 sources (two AGC cells)

1972 1974? 0.1-2 MeV

2-14 MeV

23

77

ORAUT 2006b; 

2007a

p. 29; 

Table 6-16

JN-1 Co-60 irradiations (HEC) 1972? 1982 --- ---

JN-1 Cask movements 1955 1983 --- ---

JN-1 Cleaning/maintenance in non-Cf cells 1956 1983 0.1-2 MeV 100

JN-1 Cell preparation, cleaning, maintenance: irradiated 

Pu, U-Pu forms (except 1968-1969)

1967 1972 10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

JN-1 Cell preparation, cleaning, maintenance: irradiated Pu 

fuel rods, recovery of Pu-238 and Np-237 (Low Level 

Cell)

1974? 1975 10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

JN-1 Study of irradiated U, Pu fuel aerosols (Pu-239?) 1975 1977 10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

JN-1 Sabotage project work in HEC, pool (spent fuel) 1979 1983 --- ---

JN-1 Waste handling (mostly MFPs) 1955 1983 0.1-2 MeV

2-20 MeV

90

10

ORAUT 2007 Table 6-3

JN-2 Critical assembly (zero power facility) operations 1955 1963 10-100 keV

0.1-2 MeV

2-20 MeV

3.2

59

38

ORAUT 2009b Table 6-21

JN-2 Research, fabrication with encapsulated Pu-239 

(direct conversion devices) for LLNL

1964 1970 10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

JN-2 Storage and dispensing of SNM, incl EU, Thd 1955? 1959 0.1-2 MeV

2-14 MeV

79

21

ORAUT 2006b Table F-3

JN-2 Storage and dispensing of SNM, incl EU, Pu, Thd 1960 1996 0.1-2 MeV

2-20 MeV

80

20

ORAUT 2010c Table 6-25

JN-3 Routine/typical reactor operator activities 1956 1974 0.1-2 MeV 100 ORAUT 2009b; 

2006c

Table 6-21; 

6-3

JN-3 Reactor maintenance activities (reactor down) 1956 1974 --- ---

JN-3 Reactor irradiation/activation work 1956 1974 0.1-2 MeV 100 ORAUT 2009b; 

2006c

Table 6-21; 

6-3

JN-3 Reactor beam area work 1956 1974 0.1-2 MeV 100 ORAUT 2009b; 

2006c

Table 6-21; 

6-3

JN-3 Reactor thermal column/shielding plate work 1956 1974 0.1-2 MeV 100 ORAUT 2009b; 

2006c

Table 6-21; 

6-3

JN-3 Reactor shielding pool, maintenance work in 1956 1974 --- ---

ATTACHMENT H

NEUTRON ENERGY RANGES

Page 3 of 4
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 109 of 124

Bldg Descriptiona

Operations Neutron

Reference

Location 

Begin End in Ref b Energy %

JN-4 Alloying/mixing Pu with various materials, incl U, Th 

(taken to be Pu processing (not machining, etc.)f

1960 1977 10-100 keV

0.1-2 MeV

2-20 MeV

10

50

40

ORAUT 2009b Table 6-21

JN-4 Other research with Pu-239 (corrosion, etc.) (taken to 

be Pu processing (not machining, etc.)f

1960 1977 10-100 keV

0.1-2 MeV

2-20 MeV

10

50

40

ORAUT 2009b Table 6-21

JN-4 Preparation of UCN-PuCN by arc melting, casting 

(taken to be Pu processing (not machining, etc.)f

1967 1967 10-100 keV

0.1-2 MeV

2-20 MeV

10

50

40

ORAUT 2009b Table 6-21

JN-4 Fabrication of UN-PuN (Pu-239) pellets and pins, 

including heat treatment and rolling (taken to be Pu 

processing (not machining, etc.)f

1968 1971 10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

ORAUT 2009b Table 6-21

JN-4 Research, fabrication of RTGs with Pu-238f 1974 1977

JN-4 R&D on Po-210 alpha-cell direct conversion 

generatorf

1963? 1964 0.1-2 MeV

2-20 MeV

25

75

ORAUT 2004 Table 6-9

JN-4 Purification of multicuries of Pm-147 for RTGsf 1964 1966 --- ---

JN-4 Research, fabrication of RTGs with Cm-244f 1964 1968 0.1-2.0 100 ORAUT 2010b Table 6-3

JN-4 Preparation of Np oxide microspheresf 1966 1967? 10-100 keV

0.1-2 MeV

2-20 MeV

10

50

40

ORAUT 2009b Table 6-21

JN-4 Fabricating NpO2 target rods for Pu-238 productionf 1971? 1974? 10-100 keV

0.1-2 MeV

2-20 MeV

10

50

40

ORAUT 2009b Table 6-21

JN-4 Pu waste handling (mostly Pu-239) (taken to be 

manual)

1960 1977 10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

ORAUT 2009b Table 6-21

JS-1 Production of fuel using the hot isostatic pressure 

bonding process (HEU plates)

1965? 1980 --- ---

JS-10, 

JS-12

Explosive forming, ballistic studies (mostly DU, some 

EU)

1979? 1988? --- ---

a. Only Buildings A and 1 existed before the 1950s, when Buildings 2 through 9 were built. It is known that Buildings 2 and 3 were in use 

in 1950. Thus, Buildings 2 and 3 should not be assumed to have been in use before 1950, Buildings 4 and 5 before about 1953, and 

Buildings 6, 7, and 9 before about 1954. It is not clear when JS-1 was built, but it probably was not in operation before 1965.

b. Where it is stated that a year or period is excepted, it means that no work of this type was done during that year or period.

c. Some parts of Building 3, probably the vault, were in use until 1991; some intermittent machining of uranium forms was done from 

1981-1991 in the Building 5 machine shop.

d. The vault in Building 3 probably contained more EU than anything else, as probably did the vault in JN-2 until the critical assembly 

facility shut down in 1963; after that, the JN-2 vault likely contained more Pu than any other material.

e. See footnote to table in Attachment C for a description of the fission gas analysis work.

f. Glovebox work (hence betas are not included).

ATTACHMENT H

NEUTRON ENERGY RANGES

Page 4 of 4
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 110 of 124

CONTENTS

SECTION TITLE

1.0 Introduction ..........................................................................................................................

PAGE

111

2.0 Methods ............................................................................................................................... 113

3.0 Results................................................................................................................................. 115

4.0 References .......................................................................................................................... 123

TABLES

I-1 Battelle reference files uploaded December 3, 2009 containing neutron and 

photon radiation survey data ................................................................................................ 113

I-2 Battelle reference files uploaded February 15, 2010 containing neutron and 

photon radiation survey data ................................................................................................ 113

I-3 SRDB reference files for Battelle information received from OSTI uploaded 

February 26, 2010................................................................................................................ 114

I-4 Mean neutron quality factor and flux corresponding to maximum dose 

equivalent of 1 mSv per 40-hour week ................................................................................. 114

I-5 Recommended lognormal parameters of NP ratios for Building JN-1................................... 116

I-6 Recommended lognormal parameters of NP ratios for Building JN-2................................... 118

I-7 Statistical parameters of JN-3 survey record analysis .......................................................... 120

I-8 Lognormal parameters of NP ratios for Building JN-4........................................................... 122

FIGURES

I-1 JN-1 hot cell laboratory NP ratio during 252Cf operations ...................................................... 115

I-2 JN-2 plutonium laboratory NP ratio ...................................................................................... 117

I-3 JN-3 paired neutron and photon dose rate measurements................................................... 119

I-4 JN-3 lognormal probability plot of the measured NP ratio..................................................... 119

I-5 JN-4 historical trend in NP ratio............................................................................................ 121

I-6 JN-4 plutonium laboratory NP ratio ...................................................................................... 121

I-7 JN-4 paired neutron and photon dose rate measurements................................................... 122

ATTACHMENT I

ANALYSIS OF BATTELLE MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA

Page 1 of 15
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 111 of 124

1.0 INTRODUCTION

On April 16, 1943, Battelle Memorial Institute entered into contract with the Manhattan Engineering 

District to perform atomic energy research and development activities (Battelle 2002). Since that time, 

Battelle has continuously performed research and developmental work at its facilities for the 

Department of Energy (DOE) and its predecessor agencies. Work was done at Battelle King Avenue 

and West Jefferson facilities. The Battelle West Jefferson Nuclear Sciences Area contains several 

major buildings with the potential for workplace radiation fields (Wallace 2010) as follows:

• Building JN-1 Hot Cell Building

The original section of this facility, referred to as JN-1A since 1973, was built in 1955. JN-1A 

contained offices, lab space, hot cells, and subterranean alpha gamma cells. In 1973, JN-1B was 

added to the building as a high-bay structure containing a large hot cell, adjacent fuel storage 

pool, and truck bay designed to off-load casks containing irradiated fuel assemblies into the pool 

for subsequent transfer underwater in to the cell. Both JN-1A and JN-1B were in operation until 

1983. In 1983 both JN-1A and JN-1B were shutdown and put under a surveillance and 

maintenance program prior to initiation of decontamination and decommissioning activities.

contains hot cells that were used to provide research and 

technical assistance in the areas of power reactor fuel performance evaluations, post-irradiation 

examination of nuclear materials and components, radiation source encapsulation, and physical 

and mechanical property studies on irradiated materials and structures. 

The primary purpose of the entire facility was for destructive and non-destructive examination of 

irradiated reactor fuel, cladding materials and associated reactor components. The facility was 

never a source of routine neutron radiation exposures. Occasionally, over the years, several adhoc research experiments using sealed neutron sources (e.g., usually 252Cf) were performed for 

relatively short periods. This facility was used for these experiments because of its inherent 

shielding and radiation source handling capabilities. The only other neutron exposure potential is 

associated with storage of these sealed sources and also a small number of instrumentcalibration-type neutron sources stored periodically for subsequent transfer offsite.

• Building JN-2 CAL Building

A small plutonium laboratory constructed in 1966 (noted in survey records as JN-2 Pu Lab, 

Amstein 2010) was used for research conducted for what became known as the “Lawrence 

was designed and constructed for use as a critical assembly 

laboratory and was used for critical experiments. This building was constructed in 1955 and 

housed a vault used to store special nuclear materials, a radio-analytical laboratory used to assay 

routine health physics samples, and capabilities to perform low-level radioactivity and 

environmental sample analyses. From inception of use, the building housed administrative 

offices and a radiation detection instrument calibration laboratory that used various radiation 

sources including at least one or more sealed neutron calibration sources. The high bay of 

the building housed a facility for reactor related-critical assembly experiments. The 

experiments were conducted from ~ 1956 through ~ 1960 and involved the use of very lowlevels of uranium foils. After the cessation of critical assembly experiments, the building was 

used for several nuclear-related projects, including direct conversion concepts, irradiation 

experiment assemblies, and special nuclear materials storage and dispensing. The storage 

area contained a thick-walled vault located on the ground floor.

ATTACHMENT I

ANALYSIS OF BATTELLE MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA

Page 2 of 15
Document No. ORAUT-TKBS-0058 Revision No. 00 Effective Date: 03/26/2010 Page 112 of 124

Livermore National Laboratory” involving only sealed sources of plutonium contained in glove 

boxes. This lab was decommissioned in 1975. After decommissioning, this small plutonium lab 

was converted into a radioanalytical laboratory (RAL) dedicated to analyses of low-level 

radiological samples associated with onsite facilities, including analyses of various health physics 

samples and environmental media.

• Building JN-3

•

, Battelle Research Reactor (BRR) operations began in October 1956 and ended 

December 1974. Based on Battelle licensing submittals to the Nuclear Regulatory Commission 

(Dingee and Chastain, Jr., 1961) several critical assembly systems have been studied including a 

beryllium-moderated, a plastic-moderated, two gas-cooled, water –moderated, and a uranium 

dioxide-fueled, organic-and-water moderated. This facility was selected because the shieldingresearch area was equipped with instruments to measure fast-neutron, thermal-neutron, and 

gamma radiation dose rates and spectra (Klingensmith, Epstein, and Chastain, Jr., 1959). BRR 

defueling and partial dismantlement was completed in 1975. 

Building JN-4, Plutonium Laboratory

•

was built in 1960, with additions constructed in 1964 and 

1967, to house activities for plutonium research and processing. The facility was operated from 

1960 to 1978 when it was shut down for decontamination and decommissioning (D&D). The 

facility contained a neutron exposure-potential area in one small lab within the building known as 

the Pu-238 laboratory. This laboratory contained several glove boxes for research studies 

involving the use of 238Pu. The laboratory portion of the facility was dismantled in 1985. 

Thereafter a hazardous materials research facility was located in Building JN-4 that involved only 

nonradioactive hazardous materials. 

Building JS-1

Data collections from Battelle were received as follows:

was used to develop alloy and fabrication processes, for corrosion Chemistry 

studies, and to conduct engineering analyses for the Naval Reactor Program. Records indicate 

that uranium (depleted in U-235) was used in this Facility (Cotten and Payne 1990). The building 

was decommissioned in 1990. 

• December 3, 2009. ORAUT staff uploaded 25 documents to the NIOSH database. These 

documents contained results of radiological surveys during the period of 1957-1977, primarily 

associated with operation of the Battelle Research Reactor (BRR) in Building JN-3. The pre-1970 

neutron dose survey data were in units of neutron flux (i.e., neutrons/cm2

/sec). Often survey 

results were provided for thermal and fast neutrons. The instruments used in the surveys were 

typically listed on the respective documents. A memorandum (Selander 1957) identified the 

maximum permissible limit (i.e., assumed to be 5 rem per year) for fast neutrons as being 

50n/cm2

/sec for 40 hrs for the JN-2 critical assembly. The primary focus of this effort was to 

collect and analyze paired photon and neutron radiation measurements for use in analyzing the 

neutron-to-photon dose (NP) ratio. Twenty of the files were identified as containing useful survey 

data to analyze the neutron to photon dose ratio and these were selected for data entry. The 

respective files along with the Site Research Database (SRDB) reference are listed in Table I-1.

• February 15, 2010. ORAUT staff uploaded 16 documents of supplemental information requested 

(1/7/10), approved by DOE (1/20/10) and received (Wallace 2010) from Battelle to the NIOSH 

database. These documents contained results of radiological surveys during the period of 1971-

1974, primarily associated with operation of the JN-2 and JN-4 facilities. Fourteen of the files 

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Table I-1. Battelle reference files uploaded December 3, 2009 

containing neutron and photon radiation survey data.

SRDB* File SRDB File

76975 1960 JN-3 redacted 76986 1968 JN-4 redacted

76976 1961 JN-3 redacted 76987 1969 JN-3 redacted

76977 1962 JN-3 redacted 76988 1969 JN-4 redacted

76978 1963 JN-3 redacted 76989 1970 JN-3 redacted

76979 1964 JN-3 redacted 76990 1970 JN-4 redacted

76980 1965 JN-3 redacted 76991 1971 JN-3 redacted

76982 1966 JN-3 redacted 76993 1973 JN-3 redacted

76983 1967 JN-2 redacted 76994 1974 JN-4 redacted

76984 1967 JN-3 redacted 76995 1975 JN-4 redacted

76985 1968 JN-3 redacted 76996 1976 JN-4 redacted

*Site Research Database (SRDB) Reference identification.

were identified as containing useful survey data to analyze the neutron to photon dose ratio and 

these were selected for data entry. As noted in Wallace (2010) The Battelle Monthly Health 

Physics Program Reports for Plutonium Facilities contain a section for “Neutron Flux Surveys” on 

page 2 of the report. The reference to the “Old Lab” refers to the original section of the JN-4 

plutonium facility. The “New Lab” refers to the plutonium facility located in the section of JN-4 

constructed in 1967. The “Office Area” refers the office area in JN-4. The JN-2 “Pu Lab” refers to 

the small plutonium laboratory located in JN-2. Also note that the reference to the “238 lab” under 

the same section was the plutonium-238 laboratory located in the JN-4 “New Lab.” The New Lab 

was also referred to periodically as the “Large Lab.” The respective files are listed in Table I-2 

with the associated SRDB reference.

Table I-2. Battelle reference files uploaded February 15, 

2010 containing neutron and photon radiation survey data.

SRDB* File SRDB File

79055 1971 JN-3 Reactor 79057 1974

79067 1972 (Apr) JN-4 79069 1974 (Apr) JN-4

79071 1972 (Aug) JN-4 79074 1974 (Feb) JN-4

79076 1974 (Jan) JN-4 79078 1972 (Jul) JN-4

79079 1972 (Jun) JN-4 79080 1972 (Mar) JN-4

79081 1974 (Mar) JN-4 79082 1972 (May) JN-4

79083 1972 (Oct) JN-4 79084 1972 (Sep) JN-4

*Site Research Database (SRDB) Reference identification.

• February 28, 2010. ORAUT staff received from OCAS (2/26/10) documents provided by OSTI for 

scanning and data entry. Although significant information was found describing the operations at 

several facilities at the West Jefferson location, no survey data was found. Two hazard summary 

reports were found that described the relative neutron and photon hazards at the JN-2 facility 

(Table I-3). 

2.0 METHODS 

The respective records of photon and neutron dose measurement data were categorized according to 

the general workplace or type of exposure radiation per information in the survey documentation. The 

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Table I-3. SRDB reference files for Battelle information received from OSTI uploaded February 26,

2010. 

SRDB* Title

Document 

Number Year

79448 Hazards Summary Report for the PWAR-3 Critical Assembly Experiments BMI-ACRS-611 1956

79450 Hazards Summary Report for the APB Reflector-Control Critical-Assembly 

Experiments

BMI-ACRS-615 1956

*Site Research Database (SRDB) Reference identification.

data were analyzed to eliminate incomplete data with respect to dates, missing photon or missing 

neutron measured doses, less-than photon doses and those results with a calculated NP ratio of zero. 

Beta/gamma survey instrument data were typically recorded in units of mR/hr. The results were 

typically for photon radiation only but when it was clear that beta radiation significantly contributed to 

the reading, the gamma component only was obtained if available for use in the analysis. Neutron 

survey results were recorded during earlier years in terms of flux (n/cm2

-sec) and beginning in 1970 

as a measured dose rate (mrem/hr) depending upon the instrument. Thermal and fast neutron 

components were identified although most data results used in the analysis were identified as fast 

neutron doses only. The thermal and fast neutron flux was converted to a neutron dose rate using 

equation 1 as follows:

Neutron dose = 2.5 (Fast Neutron Flux/20 + Thermal neutron flux/680) mrem/hr (Eq 1)

This equation is based on NCRP Report 38 guidance regarding values to use to convert neutron flux 

to dose rate for identified neutron energies as shown in Table I-4. The selection of 20 (i.e., neutron 

energy of approximately 1 MeV) to convert the workplace measured fast neutron flux to dose 

equivalent tends to increase the measured neutron dose rate and will tend to result in a higher NP 

ratio which is favorable to claimants. 

Table I-4. Mean neutron quality factor and flux 

corresponding to maximum dose equivalent of 1 mSv 

(100 mrem) per 40-hour week (NCRP 1971).

Neutron Energy,

MeV QF

Neutron Flux Density,

cm-2sec-1

2.5 X 10-3 (thermal) 2 680

1 x 10-7 2 680

1 x 10-6 2 560

1 x 10-5 2 560

1 x 10-4 2 580

1 x 10-3 2 680

1 x 10-2 2.5 700

1 x 10-1 7.5 115

5 x 10-1 11 27

1 11 19

2.5 9 20

5 8 16

7 7 17

10 6.5 17

14 7.5 12

20 8 11

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3.0 RESULTS

The available data for each Battelle Building have been analyzed and presented in the following 

sections.

Building JN-1, Hot Cell Laboratory

The primary mission of the JN-1 Hot Cell facility was non-destructive and destructive testing of 

irradiated reactor fuels. Most of the cells appear to have been used for these testing operations. 

There was one area that contained alpha-gamma cells that were specifically designed for work with 

transuranium materials which have a potential for neutron exposure. These cells were added to the 

basement of the JN-1 facility in 1964 (Wastren 2001). Some of these cells contained water windows 

for neutron shielding. Cells 5 and 6 were primarily used beginning in 1971 for work with 252Cf. During 

the time period 1971 through 1974, potential for neutron exposure was quite large as the quantities of 

252Cf being worked with were in the mg range. Other operations involving transuranium isotopes 

appear to have occurred periodically from startup in 1955 through 1988; however, the potential for 

neutron exposure would be much lower than during the 252Cf operations. 

For the development of an NP ratio, several radiological surveys during the 252Cf operations from 

1971 through 1974 were collected from a small sampling of records boxes. In total, 49 paired neutron 

and photon dose rate measurements were evaluated. The neutron to photon ratio can be described 

as a log normal distribution with a geometric mean of 2.4 (2.42) and a geometric standard deviation of 

3.9 (3.93) as shown in Figure I-1. The r

2 value of the fitted distribution was 0.97 indicating a robust fit 

to the dataset. The paired neutron and photon measurements also exhibited a reasonable degree of 

correlation with the Pearson correlation coefficient of 0.54 and the Spearman Rank correlation 

coefficient of 0.56.

Figure I-1. JN-1 hot cell laboratory NP ratio during 252Cf operations.

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Because NTA film would have adequately monitored neutron energies associated with 252Cf, for dose 

reconstruction if the worker was monitored for neutron exposure and worked in the JN-1 facility, the 

recorded neutron doses should be used for dose reconstruction with the appropriate adjustments 

(OCAS 2007) for missed dose. If the worker worked in the JN-1 facility and was not monitored for 

neutron exposure, consideration should be given as to the potential for neutron exposure prior to 

assigning a neutron dose based on the NP ratio listed in Table I-5 and illustrated in Figure I-1. The 

data presented in Figure I-1 are considered the upper bound of the NP ratio, since the NP ratio for 

252Cf is likely higher than other possible exposure scenarios. However, it is likely not appropriate to 

assign this ratio for a best estimate case for a long term employee who was not monitored for neutron 

exposure. For example, assigning 95th percentile ratio of 20:1, as listed in Table I-5, is certainly 

bounding however the worker could not have worked with 252Cf until at least 1971. Additional survey 

data for exposure scenarios for earlier years has been requested in order to provide a more accurate 

estimate of potential neutron dose. 

Table I-5. Recommended lognormal parameters of NP ratios for Building 

JN-1.

Workplace location Years n GM GSD 95th%

JN-1 (Hot Cell Laboratory)

Non252Cf Operations 1955–1983(a) (b)

252Cf Operations 1971–1974 49 2.4 3.9 22.5

a. Neutron exposures probably very limited prior to construction and operation of alpha 

 gamma cells in 1964. 

b. Reserved—additional survey data requested

Building JN-2, Critical Assembly Laboratory

According to Battelle (BCL 2010), the JN-2 Critical Assembly Laboratory was built in 1955. From 

1956 through the early 1960s, the facility conducted various critical experiments using enriched 

uranium. Hazard summary reports were written documenting the radiological controls in place to 

prevent high level exposures (Jankowski et al., 1956a and 1956b). Neutron survey results presented 

in Selandar (1957) show significant neutron dose rates associated with operation of a critical 

assembly. These reports indicated that radiation surveys through the building would be conducted 

during the early stages of operations. During the small sampling of radiological survey boxes (<50 

boxes of over 700 boxes), Battelle has not yet located these early surveys. 

In the interim without these surveys, a neutron to photon ratio is developed based upon the Hazard 

Summary Reports. According to Jankowski et al. (1956a) the calculated maximum neutron dose rate 

in the hallway outside the critical assembly during operations with power levels in the 1 to 10 watt 

range would result in a photon dose that was 1/8 the tolerance dose of 300 mR/wk. This corresponds 

to a photon dose rate of approximately 0.94 mR/hr. In addition, the calculated fast neutron dose 

would be on the order of 1⁄4 the weekly tolerance dose of 300mrem/week or 75 mrem/week. This fast 

neutron dose corresponds to a dose rate of approximately 1.88 mrem/hr. Based on the information 

the expected neutron to photon ratio for the JN-2 facility is approximately 2:1. 

Based on a review of the radiological records provided by Battelle for claimants, multiple workers at 

the West Jefferson facility were monitored for neutron exposure as early as 1956. Unfortunately to 

date all of the claimant data is for individuals who worked in the JN-1 (Hot Cell Facility) and the JN-3 

(Research Reactor). The dosimetry reports have been redacted for other workers at the facility 

however at least 30 workers per 2 week cycle were monitored for neutron exposure. Most of the fast 

neutron results were below the 30 mrem detection limit, however, occasionally the dose would be 

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positive but usually less than 50 mrem. Assuming that some of the neutron badges were issued 

workers in the JN-2 facility the measured neutron dose rates are well below the conservative dose 

estimates used in the Hazard Summary Report. As a result the dose rates in the Hazard summary 

report are considered reasonable and bounding.

According to Amstein (2010) in the late 1960’s a small plutonium laboratory was constructed at the 

JN-2 facility where the former radioanalytical laboratory was located. Radiological surveys for this 

laboratory have been obtained. The earliest survey obtained to date was conducted in September 

1967. Additional surveys conducted from 1972 through 1974 have been added to the analysis for the 

development of a NP ratio for the facility during the plutonium laboratory operations starting around 

1967. The neutron to photon ratio can be described as a log normal distribution with a geometric 

mean of 1.0 (0.97) and a geometric standard deviation of 1.3 (1.28) (Figure I-2). The arithmetic mean 

of the distribution was 1.01. Although the r2 value for the regression was only 0.73, this ratio is 

considered reasonable given the limited data and the relatively low GSD. The Pearson correlation 

coefficient for the 14 paired neutron and photon dose rate measurements was 0.94 and the Spearman 

Rank correlation coefficient was 0.91 indicating a very high degree of correlation between the 

measurements. 

Figure I-2. JN-2 plutonium laboratory NP ratio.

The NP ratio for the vault area of the JN-2 facility was significantly lower than the plutonium 

laboratory. The arithmetic mean of the NP ratio for the vault area was calculated from 9 data points to 

be 0.29 or approximately on third of the NP ratio for the plutonium laboratory. 

Since it is virtually impossible to place individual workers within the JN-2 facility (i.e. vault or Pu 

laboratory), the favorable to claimant approach is to use the higher of the two NP ratios for the JN-2 

Facility. Therefore for the early period during the critical assembly experiments from 1956 through 

1966, a NP ratio of 2:1 should be used. For the later years when the JN-2 plutonium laboratory was 

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in use a NP ratio of 1:1 should be used. Analysis results of the JN-2 survey data were best described 

by a lognormal distribution with a geometric mean of 1.0 (0.97) and a GSD of 1.3 (1.28) as shown in 

Table I-6. 

Table I-6. Recommended lognormal parameters of NP ratios for Building JN-2.

Workplace location Years n GM GSD 95th%

JN-2 (Critical Assembly Laboratory)

Critical Experiments 1956–1960 (a)

Plutonium Laboratory 1967–1975 14 1.0 1.3 1.5

a. Reserved–additional survey data requested

Building JN-3, Battelle Research Reactor

A total of 3,188 NP ratios were obtained from the Battelle documents. These were categorized 

according to general work location or type of exposure per information in the respective survey 

documentation. For example, all of the vacuum cleaner samples whether for dust or sludge were 

combined. Similarly, all survey data associated with evaporator were combined. Following 

categorization, data associated with 252Cf was eliminated based on the understanding that the 252Cf 

source was present in the reactor for about one year. The neutron dose rates associated with 252Cf 

were all less than detectable (i.e., 1 mrem/hr). The data with less-than photon doses were also 

eliminated. This reduced the data set to 2,487 values. A scatter plot of the neutron and photon dose 

rate paired measurements is presented in Figure I-3. The data illustrate a higher relative measured 

photon dose compared to the measured neutron dose rate. 

A lognormal probability plot of the data shown in Figure I-3 is presented in Figure I-4. The data are 

best represented by a lognormal distribution with a geometric mean of 0.2, a geometric standard 

deviation of 3.3 (3.34) and a 95th percentile of 1.4 (1.383).

Each of the workplace categories were analyzed individually and the results are listed in Table I-7 

illustrating the variation in the NP ratio and the variability. The recommended option is to utilize the 

lognormal parameters for all of these work areas collectively as shown in the last line of Table I-7 as 

exposed workers would be expected to move throughout the reactor facility. 

 

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Figure I-3. JN-3 paired neutron and photon dose rate measurements.

Figure I-4. JN-3 lognormal probability plot of the measured NP ratio.

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Table I-7. Statistical parameters of JN-3 survey record analysis.

General location No.

NP ratio

Median 84th % GSD 95th %

Radioactive dust and sludge 251 0.125 0.050 4.00 1.000

Evaporator 290 0.050 0.167 3.34 0.500

Hot area (3rd floor) 174 0.250 0.667 2.67 1.000

Hot area (basement) 144 0.153 0.667 4.36 1.000

Reactor (beam tubes) 1,116 0.250 0.876 3.50 2.527

Reactor (thermal column) 366 0.222 0.667 3.00 1.028

Shielding tank 146 0.250 0.501 2.00 4.717

Combination of all areas 2,487 0.2 0.8 4.2 2.1

Building JN-4, Plutonium Laboratory

The JN-4 facility (Plutonium Laboratory) began operations in the early 1960s. According to Battelle 

(BCL 2010), the facility was modified in 1964 and 1967. A sampling of radiological survey records for 

the entire West Jefferson plant resulted in a total of 93 records of paired neutron and photon dose rate 

measurements from 1968 through 1976 for the JN-4 facility. These were obtained from less-than 50 

boxes of records. It is understood from Battelle that as many as 700 boxes of similar type records are 

available. Of the 93 records of paired measurements, the neutron measurements were greater than 

the detection limit in 41 records. The neutron dose rate detection limit varied somewhat from 1968 

through 1976 but was generally in the range of 0.2 to 0.3 mrem/hr. The neutron to photon ratio by 

year in Figure I-5 illustrates that the NP ratio remained relatively constant over the operational time 

period.

 
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Figure I-5. JN-4 historical trend in NP ratio.

Figure I-6. JN-4 plutonium laboratory NP ratio.

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Table I-8. Lognormal parameters of NP ratios for Building JN-4.

Workplace location Years n GM GSD 95th%

JN-4 Plutonium Laboratory

Combined areas 1960–1978 41 0.7 5.0 9.9

The 41 records of paired measurements were analyzed and are best described by a lognormal 

distribution with a geometric mean of 0.7 (0.71) and a GSD of 5.0 (4.99) (Figure I-6). The 95th 

percentile of the fitted distribution was 5.0 (5.1) and the r2 value was 0.94 indicating a strong 

lognormal distribution. The arithmetic mean of the distribution was 1.71 and the Pearson Correlation 

Coefficient was 0.68 while the Spearman Rank correlation coefficient was 0.73. The NP ratio is

summarized in Table I-8.

The relatively large geometric standard deviation is attributed to the different operations within the 

facility. Many of the paired NP ratio measurements greater than 1 were from the Pu-238 laboratory. 

This is likely due to the moderately high neutron production rates due to the high specific activity and 

the alpha interactions with light materials including oxygen. Due to the relatively large GSD, 

uncertainty should be considered in most dose reconstructions. A scatter plot of the paired 

measurements illustrates that the measured neutron and photon dose rates are highly correlated 

(Figure I-7). The NP ratio 95th percentile of 9.9 is shown on the graph and illustrates that this bounds

the majority of the measured data as would be expected. 

Figure I-7. JN-4 paired neutron and photon dose rate

measurements.

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4.0 REFERENCES

Amstein, T, Battelle Columbus Laboratories, EEOICPA POC, Letter to Ms. Wallace, dated February 8, 

2010, entitled “Battelle Survey Data,” Battelle Columbus Laboratories, Columbus, OH. [SRDB 

Ref ID: 79065] 

Anno, J. N., B. P. Fairand, and J. W. Chastain, Jr., 1959, “Neutron-Flux Measurements in a 

Concentric-Cylinder Fuel Element,” Battelle Columbus Laboratories, Columbus, OH, January 

29. [SRDB Ref ID: 79557]

Battelle Columbus Laboratories (BCL), 2002, “Site Environmental Report for Calendar Year 2001 on 

Radiological and Nonradiological Parameters,” BCLDP-090102, Battelle Columbus 

Laboratories, Columbus, OH, September 1. [SRDB Ref ID: 6731]

Battelle Columbus Laboratories (BCL), 2010, “Additional Information Regarding Neutron Surveillance

For Battelle West Jefferson Buildings JN-1, JN-2, and JN-4,“ Battelle Columbus Laboratories, 

Columbus, OH, June 15. [SRDB Ref ID: 79264]

Chastain, Jr., J. W., R. F. Redmond, A. Klickman, J. Anno, and S. L. Fawcett, 1955, “Hazards 

Summary Report for the Battelle Research Reactor, “ BMI-ACRS-601 (Rev), Battelle 

Columbus Laboratories, Columbus, OH, April 4. [SRDB Ref ID: 26862]

Cotten, P. R., and J. L. Payne, 1990, Verification Survey of the JS-1 Building,” Battelle Columbus 

Laboratory, Columbus, OH, May [SRDB Ref ID: 67047]

Dingee, D. A., and J. W. Chastain, Jr., 1961, “Hazards Summary Report for the Battelle Plastic 

Reactor Facility,” BMI-ACRS-659, Battelle Columbus Laboratories, Columbus, OH, April 4, 

June 15. [SRDB Ref ID: 79563]

Jankowski, F. J., R.F. Redmond, J. W. Chastain, and S. L. Fawcett, 1956a, “Hazards Summary 

Report for the PWAR-3 Critical Assembly Experiments, “ BMI-ACRS-611 (Rev), Battelle 

Columbus Laboratories, Columbus, OH, May 28. [SRDB Ref ID: 79448]

Jankowski, F. J., W. S. Hogan, R.F. Redmond, J. W. Chastain, and S. L. Fawcett, 1956b, “Hazards 

Summary Report for the APB Reflector-Control Critical-Assembly Experiments, “ BMI-ACRS615 (Rev), Battelle Columbus Laboratories, Columbus, OH, July 31. [SRDB Ref ID: 79450]

Klingensmith, R. W., H. M. Epstein, and j. W. Chastain, Jr., 1959, “Shielding Studies on Salt Slabs, 

Gadolinium, and Water,” Battelle Columbus Laboratories, Columbus, OH, October 7. [SRDB 

Ref ID: 79570]

NCRP (National Council on Radiation Protection and Measurements), 1973, Protection Against 

Neutron Radiation, NCRP Report 38, NCRP Publications, Washington, D.C.

Selandar, C. L., 1957, Memorandum to F. D. Peterseim, dated April 1, 1957 entitled “Fast Neutron 

Levels in the Critical Assembly Building,” Battelle Memorial Institute, Columbus, Ohio [SRDB 

Ref ID: 76974]

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Yarger, L. L., and A. M. Plummer, 1964, “Irradiation of Neutron Detectors and Electrical Components,” 

BMI-X-10101, Battelle Columbus Laboratories, Columbus, OH, October 2. [SRDB Ref ID:

79558]

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