Page 1 of 145

DOE Review Release 07/01/2016

Site Profile for Battelle Memorial Institute, 

King Avenue and West Jefferson Sites, 

Columbus, Ohio

ORAUT-TKBS-0058 Rev. 01

Effective Date: 06/27/2016

Supersedes: Revision 00

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

Document Owner 

Approval: Signature on File

Vincent A. King, Document Owner

Approval Date: 06/17/2016

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 06/16/2016

Concurrence: Signature on File

Edward F. Maher, Objective 3 Manager

Concurrence Date: 06/16/2016

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 06/16/2016

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 06/27/2016

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 2 of 145

 

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.

06/27/2016 01 Revision initiated to include the Special Exposure Cohort class 

determination for the King Avenue facility due to inadequate internal 

radiological exposure monitoring (April 16, 1943, through June 30, 

1956) and inadequate external radiological exposure monitoring 

(April 16, 1943, through February 13, 1951) and to include the class 

determination for the King Avenue facility due to inadequate internal 

monitoring of thorium exposures (July 1, 1956, through December 

31, 1970). Revised Section 3.0 on occupational medical dose to 

incorporate updated Project information from ORAUT-OTIB-0006, 

ORAUT-OTIB-0079 and additional references. Modified Section 4.0 

thorium information for internal dose section to include SEC classes, 

additional detail on chemically separated thorium, and ratios for 

recycled uranium radionuclides. Added information in Section 5.0 on 

occupational external dose section to address interpretation of 

dosimeter results expressed in “arbitrary units.” Added Section 5.8 to 

include assessment of ambient external radiation exposures in dose 

reconstructions. Added Section 6.0 to include assessment of internal 

environmental exposures in dose reconstructions. Incorporates 

formal internal and NIOSH review comments. Constitutes a total 

rewrite of the document. Training required: As determined by the 

Objective Manager. Initiated by Vincent A. King.
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 3 of 145

 

TABLE OF CONTENTS

SECTION TITLE PAGE

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

1.0 Introduction .............................................................................................................................. 9

1.1 Purpose...................................................................................................................... 10

1.2 Scope......................................................................................................................... 10

1.3 Special Exposure Cohort............................................................................................ 11

2.0 Site Description...................................................................................................................... 12

2.1 Buildings of Concern and the Associated Radioactive Materials and Operations........ 14

2.2 Access Controls.......................................................................................................... 15

3.0 Occupational Medical Dose.................................................................................................... 17

4.0 Occupational Internal Dose.................................................................................................... 19

4.1 In Vitro Bioassay......................................................................................................... 19

4.1.1 Urinalysis ..................................................................................................... 19

4.1.2 Fecal Sample Analysis................................................................................. 21

4.1.3 Nose Swabs................................................................................................. 21

4.2 In Vivo Bioassay......................................................................................................... 21

4.3 Lapel Samplers .......................................................................................................... 22

4.4 In Vitro Bioassay Detection Limits .............................................................................. 22

4.5 Inhalation Absorption Types, Particle Sizes, and Activity Fractions ............................ 25

4.6 Dose Reconstruction Considerations.......................................................................... 27

4.7 Records and Reports.................................................................................................. 29

5.0 Occupational External Dose................................................................................................... 30

5.1 Dosimetry Processing Services .................................................................................. 30

5.2 Dosimeter Types ........................................................................................................ 30

5.3 Dosimeter Detection Capabilities................................................................................ 31

5.4 Dosimeter Exchange Frequencies.............................................................................. 33

5.5 Workplace Radiation Fields ........................................................................................ 34

5.6 Assessment of Neutron Dose Monitored by Film ........................................................ 35

5.7 Incident Reports ......................................................................................................... 36

5.8 Onsite Ambient External Dose.................................................................................... 36

6.0 Internal Dose from Onsite Atmospheric Radionuclide Concentrations.................................... 39

6.1 Air Sampling and Release Estimates.......................................................................... 39

6.1.1 King Avenue, 1943 to 1972 .......................................................................... 39

6.1.2 King Avenue, 1973 to 1975 .......................................................................... 39

6.1.3 West Jefferson, 1956 to 1972 ...................................................................... 39

6.1.4 West Jefferson, 1973 to 2005 ...................................................................... 39

6.2 Estimated Annual Intakes from Airborne Radionuclides.............................................. 41

7.0 Residual Exposure................................................................................................................. 44

8.0 Attributions and Annotations .................................................................................................. 45

References ........................................................................................................................................ 46
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Glossary ............................................................................................................................................ 73

ATTACHMENT A CHRONOLOGY OF EVENTS......................................................................... 75

ATTACHMENT B BUILDINGS AND OTHER FACILITIES........................................................... 83

ATTACHMENT C HISTORY OF OPERATIONS BY BUILDING AND MATERIAL ....................... 96

ATTACHMENT D TYPES AND QUANTITIES OF MATERIAL................................................... 111

ATTACHMENT E REPORTED INCIDENTS.............................................................................. 122

ATTACHMENT F ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA ........................................................................ 128

ATTACHMENT G ESTIMATING ONSITE AIR CONCENTRATIONS......................................... 140

LIST OF TABLES

TABLE TITLE PAGE

3-1 Default assumptions for X-ray procedure types and frequencies............................................ 18

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

4-2 Whole-body counting analytical sensitivity ............................................................................. 22

4-3 Urinalysis detection limits data to be used in dose reconstruction .......................................... 23

4-4a Isotopic composition as fraction of total plutonium alpha activity data for plutonium work ...... 26

4-4b Isotopic composition relative to Pu-239 activity for plutonium work ........................................ 26

4-5 Cooling times for FPs at Battelle ............................................................................................ 27

4-6 Activity fractions of contaminant in recycled uranium ............................................................. 29

5-1 Dosimetry processing services............................................................................................... 30

5-2 Summary of MDLs for dosimeter badges ............................................................................... 31

5-3 Default dosimeter exchange frequency by facility or work group ............................................ 33

5-4 Photon and energy ranges for material types......................................................................... 34

5-5 Photon dose units for use with organ dose conversion factors ............................................... 35

5-6 Recommended NP ratios for dose monitored by NTA film...................................................... 35

5-7 Annual environmental dose from external background radiation as measured by TLD........... 38

5-8 Prorated annual dose from external gamma radiation ............................................................ 38

6-1 Air effluent release points at the West Jefferson site in 1975 ................................................. 41

6-2 Occupational environmental inhalation of radionuclides at BCL ............................................. 43

A-1 Chronology of events .............................................................................................................75

B-1 King Avenue site building descriptions ................................................................................... 84

B-2 West Jefferson site building descriptions................................................................................ 89

C-1 History of work operations at the King Avenue site................................................................. 97

C-2 History of operations at the West Jefferson site ................................................................... 104

D-1 Source and production materials.......................................................................................... 112

D-2 Sealed source materials ....................................................................................................... 120

D-3 Waste and remainder materials............................................................................................ 121

E-1 Incidents reported at the King Avenue and West Jefferson sites.......................................... 122

F-1 Buildings JN-2, JN-3, and JN-4 neutron and photon radiation survey data........................... 131

F-2 Buildings JN-2 and JN-4 neutron and photon radiation survey data ..................................... 131

F-3 Building JN-2 hazards summary reports............................................................................... 132

F-4 Mean neutron quality factor and flux corresponding to maximum dose equivalent of 

1 mSv per 40-hour week ...................................................................................................... 133

F-5 Recommended lognormal parameters of NP ratios for Building JN-1 ................................... 134

F-6 Recommended lognormal parameters of NP ratios for Building JN-2 ................................... 136
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F-7 Statistical parameters of JN-3 survey record NP ratio analysis............................................. 137

F-8 Lognormal parameters of NP ratios for Building JN-4 ........................................................... 138

G-1 Intakes of alpha-emitting radionuclides from the West Jefferson site by year....................... 144

G-2 Intakes of mixed fission product radionuclides from the West Jefferson site by year............ 145

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Relative locations of the King Avenue and West Jefferson sites............................................. 13

2-2 Plan view of the King Avenue site .......................................................................................... 14

2-3 Plan view of the West Jefferson site....................................................................................... 15

6-1 Occupational environmental inhalation of alpha emitters at West Jefferson ........................... 40

6-2 Occupational environmental inhalation of MFPs at West Jefferson ........................................ 41

F-1 JN-1 hot cell laboratory NP ratio during 252Cf operations ...................................................... 134

F-2 JN-2 Plutonium Laboratory NP ratio..................................................................................... 135

F-3 JN-3 paired neutron and photon dose rate measurements................................................... 136

F-4 JN-3 lognormal probability plot of the measured NP ratio..................................................... 137

F-5 JN-4 historical trend in NP ratio............................................................................................ 138

F-6 JN-4 Plutonium Laboratory NP ratio..................................................................................... 139

F-7 JN-4 paired neutron and photon dose rate measurements................................................... 139

G-1 Layout of the West Jefferson site, 1973 ............................................................................... 142

G-2 Layout of the King Avenue facilities that had stack emission points, 1973............................ 143
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

AGC Alpha-Gamma Cell

ANL Argonne National Laboratory

AP anterior-posterior

APPR Army Package Power Reactor

ATR Advanced Test Reactor

AWE atomic weapons employer

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

CAA Controlled Access Area (at JN-1)

CAL Critical Assembly Laboratory (JN-2, 1957 to 1963) 

CAM continuous air monitor

CEMP Columbus Environmental Management Project

CFR Code of Federal Regulations

Ci curie

cm centimeter

Con Ed Consolidated Edison

cpm counts per minute

D&D decontamination and decommissioning

d day

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EDTA ethylenediaminetetraacetic acid

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ERDA U.S. Energy Research and Development Administration

ETR Engineering Testing Reactor

EU enriched uranium

FP fission product

ft foot

g gram

gal gallon

GCRE Gas-Cooled Reactor Experiment

GM geometric mean

GSD geometric standard deviation

HAPO Hanford Atomic Power Operations

HEC High Energy Cell (JN-1)

HEDL Hanford Engineering Development Laboratory
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HEPA high-efficiency particulate air

HEU highly enriched uranium

HP health physicist

hr hour

ICRP International Commission on Radiological Protection

in. inch

INL Idaho National 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)

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

KPA kinetic phosphorescence analysis

L liter

LAT lateral

lb pound

LLC Low Level Cell (JN-1)

LLNL Lawrence Livermore National Laboratory

mCi millicurie

MDA minimum detectable amount

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 (type of 238Pu fuel for RTGs) 

mi mile

mm millimeter

mrem millirem

mL milliliter

ML-1 Mobile Low-Power Reactor 1 

MPBB maximum permissible body burden

mph miles per hour

MPC maximum permissible concentration

MPL maximum permissible limit

mR milliroentgen

mSv millisievert

MT metric ton

MW megawatt

NAC Nuclear Assurance Corporation

NASA National Aeronautics and Space Administration

nCi nanocurie

NCRP National Council on Radiation Protection and Measurement

NIOSH National Institute for Occupational Safety and Health

NLO National Lead Company of Ohio

NP neutron-to-photon

NRC U.S. Nuclear Regulatory Commission
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NSEC Nuclear Science and Engineering Corporation

NTA nuclear track emulsion, type A

NU natural uranium

NYOO New York Operations Office

OMR organic-moderated reactor

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

Experiment

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

PA posterior-anterior

pCi picocuries

PRTR Plutonium Recycle Test Reactor

PyC pyrolitic carbon

R roentgen

R&D research and development

RAL Radioanalytical Laboratory

RCRA Resource Conservation and Recovery Act of 1976

RTG radioisotope thermoelectric generator

s second

S&M surveillance and maintenance

SID source-to-image distance

SME Salt-Moderated Experiment

SNAP Systems for Auxiliary Nuclear Power

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 type) 

t ton

TLD thermoluminescent dosimeter

TRU transuranic

U.S.C. United States Code

UNH uranyl nitrate hexahydrate

vol % volume percent

W watt

wt % weight percent

wk week

μCi microcurie

μg microgram

μm micrometer

§ section or sections
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1.0 INTRODUCTION

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

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

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

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

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

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

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

of the individual work required for each dose reconstruction.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

at an AWE facility is categorized as employment either (1) during “a period when the employer was 

processing producing, for the use by the United States, material that emitted radiation and was used 

in the production of an atomic weapon, excluding uranium mining and milling,” (i.e., the operational 

period); or (2) during a period that NIOSH has determined that “there is a potential for significant 

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

residual contamination period). 42 U.S.C. § 7384l(3).

Based on the abovementioned definition for eligible employment during an AWE facility’s operational 

period, NIOSH includes radiation exposures incurred in the performance of duty, such as medical 

X-rays received as a condition of employment for participating in DOE projects, at an AWE facility in 

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

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

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

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

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

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

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

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

period. 42 U.S.C. § 7384n(c)(4)(A). Second, radiation doses from sources not included in the first 
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 10 of 145

 

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

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

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

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

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

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

Therefore these exposures are not included in dose reconstructions for either the operational or 

residual contamination period (NIOSH 2010):

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

conventional structures

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

reasons

1.1 PURPOSE

The purpose of this document is to provide a site profile that contains technical basis information for 

the evaluation of occupational dose for EEOICPA claimants who were employed at the Battelle 

Memorial Institute (BMI) or Battelle Columbus Laboratories (BCL) King Avenue and West Jefferson 

sites in Columbus, Ohio. In this site summary document, BMI and BCL (the Columbus-based division 

of BMI operations) are referred to as Battelle. When it is necessary to distinguish between the two 

sites, they will be referred to as King Avenue and West Jefferson, respectively.

1.2 SCOPE

Section 1.3 describes the Special Exposure Cohort (SEC) class for the King Avenue site. Section 2.0 

describes the King Avenue and West Jefferson sites and their operations. Section 3.0 discusses 

occupational medical dose, and Sections 4.0 and 5.0 detail known information about occupational 

internal and external dose, respectively. Section 6.0 addresses assessment of internal environmental 

exposures in dose reconstructions. Section 7.0 discusses exposure from residual contamination after 

the end of Battelle operations. Attributions and annotations, indicated by bracketed callouts and used 

to identify the source, justification, or clarification of the associated information, are presented in 

Section 8.0. 

Evaluation of drinking water and soil as potential sources of intake for workers was evaluated. Soil 

samples were collected to demonstrate compliance with environmental release requirements and 

results were reported in Battelle annual environmental reports (listed in the Environmental Reports 

section of the References list). Drinking water samples, although not required for environmental 

compliance monitoring, were collected beginning as early as May of 1966 (Kizer 1971). No results 

above regulatory limits were recorded for either soil or drinking water samples, and no potential 

pathway for worker exposure was identified. 

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. Attachments C and D report the history of operations by building and material and the types 

and quantities of material, respectively. Attachment E lists the incidents that were reported at the 

sites. Attachment F provides the analysis of measured neutron and photon dose radiation survey 

data, and Attachment G addresses the estimation of onsite air concentrations.
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1.3 SPECIAL EXPOSURE COHORT

The Secretary of the U.S. Department of Health and Human Services has designated the addition of 

the following class to the Special Exposure Cohort (SEC) for the Battelle King Avenue site (Sebelius 

2013): 

All Atomic Weapons Employees who worked at the King Avenue facility owned by 

Battelle Laboratories in Columbus, Ohio, during the period from April 16, 1943, through 

June 30, 1956, for a number of work days aggregating at least 250 work days, occurring 

either solely under this employment or in combination with work days within the 

parameters established for one or more other classes of employees included in the 

Special Exposure Cohort.

Dose reconstruction guidance in this document for periods through June 30, 1956, is presented to 

provide a technical basis for partial dose reconstructions for claims not compensated under the SEC 

(i.e., nonpresumptive cancers and SEC employment less than 250 days). NIOSH found that it was 

not feasible to reconstruct internal doses for inadequately monitored radionuclides such as uranium, 

thorium, and their progeny through June 30, 1956, nor was it feasible to reconstruct external doses 

from beta, gamma, and neutron radiation through February 13, 1951. Although NIOSH found that it is 

not possible to completely reconstruct radiation doses for the proposed class, it intends to use any 

internal and external monitoring data that might become available for an individual claim (and that can 

be interpreted using existing NIOSH dose reconstruction processes or procedures). Therefore, dose 

reconstructions for individuals employed at the Battelle King Avenue facility during the period from 

October 1, 1943, through June 30, 1956, but who do not qualify for inclusion in the SEC, can be 

performed using these data as appropriate (Sebelius 2013).

In addition, the Secretary of the U.S. Department of Health and Human Services has designated the 

addition of the following class to the Special Exposure Cohort (SEC) for the Battelle King Avenue site 

(Burwell 2016):

All Atomic Weapons Employees who worked at the facility owned by the Battelle 

Laboratories at the King Avenue site in Columbus, Ohio, during the period from July 1, 

1956, through December 31, 1970, for a number of work days aggregating at least 250 

work days, occurring either solely under this employment, or in combination with work 

days within the parameters established for one or more other classes of employees 

included in the Special Exposure Cohort.”. 

Dose reconstruction guidance in this document for the period from July 1, 1956, through December 

31, 1970, is presented to provide a technical basis for partial dose reconstructions for claims not 

compensated under the SEC (i.e., nonpresumptive cancers and SEC employment less than 250 

days). NIOSH found that, due to insufficient personnel and workplace monitoring and source term 

data, it lacked sufficient information necessary to complete individual dose reconstructions with 

sufficient accuracy for internal exposures to thorium and thorium progeny to which these workers may 

have been subjected during the period (Burwell 2016). Although NIOSH found that it is not feasible to 

completely reconstruct radiation doses for this proposed class, NIOSH intends to use any internal and 

external monitoring data that may become available for an individual claim (and that can be 

interpreted using existing NIOSH dose reconstruction processes or procedures). Therefore, dose 

reconstructions for individuals employed at the Battelle Laboratories King Avenue facility during the 

period from July 1, 1956, through December 31, 1970, but who do not qualify for inclusion in the SEC, 

may be performed using these data as appropriate.
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2.0 SITE DESCRIPTION

The DOE Facility List Database provides the following information for the King Avenue and West 

Jefferson sites (DOE 2014a, 2014b):

• Battelle Laboratories - King Avenue

Also Known As: Battelle Columbus Laboratories (BCL); Battelle Memorial Institute (BMI)

State: Ohio; Location: Columbus

Time Period: AWE 1943-1986; BE 1943-1961; DOE 1986-2000 (remediation); Residual 

Radiation 2001 - March 1, 2011

Facility Type: Atomic Weapons Employer/ Beryllium Vendor/ Department of Energy

Facility Description: From 1943 to 1986, Battelle Memorial Institute performed atomic 

energy research and development as well as beryllium work for the Department of 

Energy 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 use of reactor vessels and piping.

The following activities were performed at the King Avenue location: processing and 

machining enriched, natural, and depleted uranium and thorium; fabricating fuel 

elements; analyzing radiochemicals; and studying power metallurgy. Beryllium work was 

conducted from 1943 until at least 1961.

• Battelle Laboratories - West Jefferson

Also Known As: Battelle Memorial Institute (BMI); Battelle Columbus Laboratories 

(BCL); West Jefferson Plutonium Facilities

State: Ohio, Location: Columbus 

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, Battelle Memorial Institute performed atomic 

energy research and development for the Department of Energy and its predecessor 

agencies. The Battelle Laboratories have two separate locations in Columbus - King 

Avenue and West Jefferson. Battelle participated in research on fabrication of uranium 

and fuel elements, reactor development, submarine propulsion, fuel reprocessing, and 

the safe use of reactor vessels and piping.

At the West Jefferson location, Battelle operated 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.
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During the period of residual contamination, as designated by the National Institute for 

Occupational Safety and Health and as noted in the dates above, employees of 

subsequent owners and operators of this facility are also covered under the Energy 

Employees Occupational Illness Compensation Program Act.

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

part of its work for the U.S. Department of Energy and its predecessors1 along with other 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 (Peters 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 1978; BCL 2002a).

Figure 2-1. Relative locations of the King Avenue and West Jefferson sites (BCL 1974).

 

1. Manhattan Engineer District (MED), U.S. Atomic Energy Commission (AEC), and U.S. Energy and Research 

Development Administration (ERDA).
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Figure 2-2. Plan view of the King Avenue site (Jenson 2003).

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 sites.

Most of these did not result 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 for external exposure 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. Those for internal exposure 

are February 10, 1959, May 9, 1974, January 16, 1979, May 2, 1980, February 27, 1984, and May 

1984.
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Figure 2-3. Plan view of the West Jefferson site (Jenson 2003).

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. The 

laboratory supervisor’s permission was required to open the doors.

At the Gamma Facility (Building 6 where Battelle conducted 60Co irradiation), 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 (Peters 2001). Access was controlled at all 

times and, in compliance with 10 CFR Part 20, the entire building was classified as 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 could also be made through 

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

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 full 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 and 

the office area were kept locked from the inside (BCL 1977a).
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3.0 OCCUPATIONAL MEDICAL DOSE

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

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

1943). It was known in 1948 that beryllium was handled at Battelle, and that beryllium workers might 

need to be monitored for adverse health effects (Snapp 1948). An occupational medical program 

appears to have started at Battelle by 1949, which at first only included the beryllium workers. The 

medical program consisted of preplacement, an annual periodic exam, and a termination exam, with 

chest X-rays performed at least during the preplacement and periodic exams. The X-rays were done 

at University Hospital, which was a short distance away from Battelle (Tabershaw 1949). These 

X-rays are not eligible for inclusion in dose reconstruction under EEOICPA, because they were 

performed off site at a noncovered facility (ORAUT 2011a).

During the 1950s, there is evidence that the X-rays of Battelle workers were taken at a private 

physician’s office and included chest fluoroscopy (Wright 2002a, 2002b). Chest fluoroscopy was not 

a widespread screening method for workers (ORAUT 2011b). By 1957, screening by chest 

fluoroscopy was discouraged by the American Trudeau Society (1957, p. 3). Chest fluoroscopy of 

Battelle workers probably ended in 1956 (Wright 2002a; Thomas 2007; Thomas 2010). However, 

because the chest fluoroscopy and other radiographic chest X-rays during this time occurred off site 

at a private physician’s office, the dose from them is not eligible for inclusion in dose reconstruction 

(ORAUT 2011a).

It is not known where X-rays might have been taken during 1957-1968. A few claim file records 

indicate that some X-rays may have occurred offsite at a private physician’s office (Thomas 2010).

However, onsite X-rays cannot be ruled out for this period (Amstein 2008). Therefore, dose 

reconstructors should assign dose from chest X-rays starting in 1957, unless the claim file record 

indicates their X-rays were taken offsite. 

The first evidence of medical X-ray equipment at Battelle is from 1968 (Buring 1968). This 

memorandum describes the shielding evaluation for a new X-ray machine installation in the health 

service area. The machine is described as a radiographic and fluoroscopic machine. However, there 

is no evidence that chest fluoroscopy continued to be performed after 1956, so it is assumed that only 

the radiographic mode was used for occupational health screening. There is no definitive technique 

factor information in the shielding evaluation documents that can be reliably used to develop sitespecific organ doses for radiography using this machine.

Dose reconstructors should assign dose from screening chest X-rays at Battelle starting in 1957. It 

appears from the available records that the dose from a posterior-anterior (PA) chest X-ray 

examination should be assigned for the preemployment, annual, and termination physicals for 

radiation workers (Wright 2002a, 2002b; BCL 1968, p. 12). The records also indicate that a single 

annual lateral (LAT) chest X-ray was performed in addition to the annual PA radiographic chest X-ray 

for the period from 1975 through 1980. Some of the records use the phrase “teleo chest” X-ray, which 

was performed at the standard 72-in. source-to-image distance (SID) to evaluate heart size. The 

same PA chest X-ray doses should be assigned for these “teleo chest” examinations as for the 

conventional PA chest X-rays because the SID and X-ray equipment would have been the same for 

both. X-rays were performed on site at Battelle from 1968 through 1980 (Amstein 2008).

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 3-1 should be assigned.

Evidence of when the X-ray machine at Battelle was removed has not been found. Therefore, dose 

reconstructors should continue to assign dose from chest X-rays until the end of the covered period, 
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Table 3-1. Default assumptions for X-ray procedure types and frequencies.

Period

Preemployment 

PA chest

Annual PA 

chest LAT chest

Termination PA 

chest

ORAUT-OTIB-0006 

dose

Beginning of 

operations–

1956

All X-rays done off 

site at noncovered 

facility

All X-rays 

done off site 

at 

noncovered 

facility

All X-rays 

done off site 

at 

noncovered 

facility

All X-rays done off 

site at noncovered 

facility

N/A

1957–1967 Yes Yes No Yes Poor collimation 

before 1970

1968-1974 Yes Yes No Yes Proper collimation 

1971-1985

1975–1980 Yes Yes Yes Yes Proper collimation 

1971-1985

1981–2000a Yes Yes No Yes Proper collimation 

1971-1985 or 

Proper collimation 

1986-present

a. Note that the end of the covered period for the West Jefferson facility is “the present”. 

which is 2000 for the King Avenue facility and the present for the West Jefferson facility. 

Occupational medical X-ray doses are not reconstructed for the residual periods at West Jefferson

(1976 through 1985) or King Avenue (2001 to March 1, 2011) in accordance with the guidance of 

NIOSH (2010). 

ORAUT-OTIB-0006, Dose Reconstruction from Occupational Medical X-Ray Procedures (ORAUT 

2011a provides organ doses for PA and LAT radiographic chest examinations. The doses assigned 

from ORAUT-OTIB-0006 should be those for the period of “poor collimation before 1970” for Battelle 

X-rays from 1957 to 1967, and for the period of “proper collimation 1971 through 1985” and “1986 to 

the present”, for Battelle X-rays 1968 to 1985 and 1986 to the present, respectively. 
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4.0 OCCUPATIONAL INTERNAL DOSE

4.1 IN VITRO BIOASSAY

4.1.1 Urinalysis

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

records note samples that represent longer or shorter collection 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.

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 was involved, or driven by follow-up activities to specific incidents. The earliest 

urinalysis mention in records is of samples submitted in July 1956 to a toxicology consultant; these 

samples were assayed only for uranium and the results were given in milligrams (Blauer 1949). 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.

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

a

Monitoring type Radionuclide Period

Urine, 24-hour Gross beta 1957–1963

Urine, 24-hour Uranium (total), radiometric uranium (“RU”)b 1957–1998

Urine, 24-hour U-234, U-235, and U-238, a.k.a. specific uranium (“SPU”)c 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 FPs/MFPs 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

a. In addition to the information above, enriched uranium, Am-241, Sr-90, Ce-144, Cs-134, Cs-137, Co60, P-32, I-131, Cr-51, and I-125 were all assayed for urinalyses at different times, as could be seen in 

the group reports from Eberline.

b. When “RU” is used in the bioassay records, it indicates “radiometric uranium,” not recycled uranium

(see Section 4.6 Uranium discussion).

c. When “SPU” is used in the records, it indicates the three U isotopes U-234, -235, and -238, not Pu

(see Section 4.6 Uranium discussion).

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

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 (FPs), plutonium, polonium, tritium, 

or uranium (Geiger 1963, 1964a, 1965); from inspection of urinalysis reports, the FPs 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 
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products (MFPs). 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 available 

dosimetry records, claimant statements, and site documents, it appears that King Avenue workers 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; 

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

handled; and

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

specific projects. 

It appears that 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 (about 1964

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

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

• 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 might also have been included as necessary based on their job functions. 

Workers from King Avenue facilities appear to have received baseline bioassay samples before 

transferring to West Jefferson facilities (Peterseim 1961). 

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

along with all 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. The program included most 

of the workers in nonclerical and nonmanagement positions (above the level of supervisors), who 

were probably engaged in radiological work at least intermittently. 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 
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individual dosimetry files there are multiple notices sent to workers about 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, MFPs, and 90Sr. Later samples were analyzed for 239Pu and 

FPs. A summary of the bioassay data for the most exposed worker is shown in Toy (1975a).

4.1.2 Fecal Sample Analysis

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 MFPs, depending on the year 

and the case. The set of samples analyzed in 1974 was 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, MFPs, and some specific FPs including 141Ce144Pr, 

103Ru106Ru106Rh, and 95Zr95Nb. Later samples were analyzed for 239Pu and FPs. 

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

1978a).

4.1.3 Nose Swabs

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

1959a; Saling 1963; Kirsch 1978a). In addition, it appears that it was routine or at least advisable to 

take a swab when 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 (BCL 1977b). Although information from nose swabs typically cannot quantify intake 

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

occurred after an incident.

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 was from 1998). In vivo counts 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 might have participated at this frequency. Before 1995, minimum 

sensitivities were recorded for each semiannual 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 minimum 

detectable amount (MDAs) in the records are listed in Table 4-2 for each occupationally associated 

radionuclide and may be assumed as favorable to claimants. 
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Table 4-2. Whole-body counting analytical sensitivity. 

Radionuclide MDA (nCi)

Mn-54 4.4

Co-58 3.0

Co-60 2.0

Ag-110 3.1

I-131 3.6

Cs-134 3.0

Cs-137 3.6

Beginning in 1995, details of the whole-body counts [detectors, minimum detection levels (MDLs) 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 found 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., BCL 1977a; Kirsch 1978a, 1978b; 

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 might serve to verify whether or not a positive intake occurred 

after an incident.

4.4 IN VITRO BIOASSAY DETECTION LIMITS

Table 4-3 lists 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 MDAs in effect for the analytical 

laboratory, except for the information provided by ORAUT (2007). The detection limits discussed in 

this section, other than those provided in ORAUT (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 nonpositive 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 (ORAUT 2007).
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Table 4-3. Urinalysis detection limits data to be used in dose reconstruction. 

Isotope Start End

Routine 

detection limit Units

Total uraniuma 1957 1995 5 μg/L

Total uraniuma 1996 1998 1 μg/L

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

U-234, U-235, U-238 1992 1994 0.1 pCi/L

Total plutonium 1961 1991 1 dpm/24 hr

Total plutonium 1992 1998 0.1 pCi/L

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

Pu-238, Pu-239 1992 1993 0.06 pCi/L

H-3 1962 1965 1 μCi/L

Am-241 1974 1980 0.03 dpm/sample

Th-228, Th-230, Th-232 1981 1991 2 dpm/sample

Th-228, Th-230, Th-232 1992 1994 0.2 pCi/L

Total thorium 1993 1998 0.5 pCi/L

Various FPs 1957 1992 0.2 dpm/mL

Various FPs 1993 1998 30 pCi/L

a. Fluorometric analysis through 1995, KPA beginning 1996.

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 from Eberline 

(ORAUT 2007). The limits in Table 4-3 should be used when no information is available in the claim 

records; they do not take precedence over actual MDAs in claim records. In additions, the start and 

end dates should not be considered absolute if they conflict with information in the claim records.

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

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

2007). Therefore, analytical results lower than the values listed above, but are reported as positive 

results, should be considered valid. Detection limits for the radionuclides 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 milligrams per liter (1956 to 

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

throughout the operation of the site. Based on claimant records from DOE, the minimum reporting 

level was 5 μg/L from at least 1964 through 1995 (although not explicitly identified as an 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 (NU).

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 in dpm/sample or dpm/24 hours before 1992 and in pCi/L for 1992 and 

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

2007) was also applicable for previous periods, which would equate to approximately an MDA of 
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0.3 dpm/24 hr. Consequently, an MDA value of 0.3 dpm/24 hr 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 ORAUT (2007.)

Plutonium

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

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

and the change in units from dpm/24 hr 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 

found for 1992). Positive reported values before 1992 were often lower than the routine detection limit 

in the table, which could 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 was performed for a worker.

Americium

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

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

samples were recorded as zero with an associated error of 0.03 dpm/sample.

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. The first isotopic thorium analyses for 228Th, 

230Th, and 232Th found in the records were in 1981. Battelle internal correspondence from 1981 

indicates that the routine program by that time involved thorium analysis of approximately eight

samples per year and shows that thorium urinalysis was included in the blank and spiked samples in 

the routine quality control reports (Kirsch 1981). Monitoring was performed on frequencies of 6 or 

12 months, depending on the work (BCL ca. 1981). A further indication that thorium bioassay was not 

frequently implemented at Battelle is shown in request forms for individual monitoring, which list 

thorium bioassay as “other” rather than included in typical analysis types (BMI 1981). 

Records indicate an apparent change in detection limit and in reporting units from 2 dpm/sample

(1981 to 1991) to 0.2 pCi/L (beginning in 1992) for isotopic thorium analyses. 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, 
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 25 of 145

 

assumed to represent FPs, are useful for assessing intakes. Beginning in 1961, analyses were 

designated as “FP”; beginning in 1980, they were designated as “MFP”. 

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 FP 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 [1]. The MDA after 1991 is based on information from Eberline 

(ORAUT 2007).

4.5 INHALATION ABSORPTION TYPES, PARTICLE SIZES, AND ACTIVITY FRACTIONS

For the reactor, the irradiated materials were encapsulated before irradiation and were put into 

containers as they were removed from irradiation, so that the materials to which reactor workers and 

researchers were exposed as far as internal dose goes were the normal reactor contamination (FPs

and activation products) such as 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 therefore 

only incidental contamination was available for intake.

Tables 4-4a and 4-4b provide mixture information for weapons-grade and fuel-grade plutonium from 

ORAUT (2012) and ORAUT (2005a). 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 producers of plutonium for the DOE complex. These mixtures should be applied in the 

absence of specific information in the record. The activity ratios for “fresh” plutonium in Tables 4-4a 

and 4-4b should be used for Battelle dose reconstruction if the work involved recently separated 

weapons or reactor-grade material. If the age of material is unknown, the 10-year aged ratio should 

be used as favorable to claimants. 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 4-4a should be applied to results listed as total plutonium activity; Table 4-4b provides results 

for associated radionuclides relative to 239

Pu when isotopic plutonium analyses were performed.
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Table 4-4a. Isotopic composition as fraction of total plutonium alpha activity for plutonium work. 

Isotope

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 1.07E–01 9.98E–02 1.77E–01 1.65E–01

Pu-239 7.23E–01 7.29E–01 5.43E–01 5.50E–01

Pu-240 1.70E–01 1.72E–01 2.81E–01 2.85E–01

Pu-241 1.03E+01b 6.42E+00b 3.19E+01b 2.00E+01b

Am-241 0.00E+00 1.32E–01 0.00E+00 3.98E–01

a. ORAUT (2005a, 2012).

b. Total plutonium alpha activity does not include Pu-241, which is primarily a beta emitter with insignificant associated 

alpha activity.

Table 4-4b. Isotopic composition relative to Pu-239 activity for plutonium work. 

Isotope

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 1.48E–01 1.37E–01 3.25E–01 3.01E–01

Pu-239 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Pu-240 2.35E–01 2.35E–01 5.17E–01 5.18E–01

Pu-241 1.43E+01 8.82E+00 5.88E+01 3.64E+01

Am-241 0.00E+00 1.80E–01 0.00E+00 7.24E–01

a. ORAUT (2005a, 2012).

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

the plutonium contains a few or more weight percent of 241Pu and has aged. This 241Am should be 

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

included in Tables 4-4a and 4-4b 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, 

assumptions should be used for these parameters that are favorable to the claimant. 

ICRP (1995) indicates that plutonium materials in the workplace can be absorption types M or S, 

depending on the chemical and physical form of the material. Some intakes of 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 (2010a), 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 

SS is not applicable in accordance with the guidelines of ORAUT (2010a). For workers primarily 

exposed to 238Pu, absorption type should be evaluated in accordance with the provisions of ORAUT 

(2013c) to determine the appropriate model favorable to the claimant. 

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

the ICRP Publication 66 default particle size distribution of 5-μm activity median aerodynamic 

diameter (ICRP 1994). If specific particle size information is available in the record, that information 

should be used.
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4.6 DOSE RECONSTRUCTION CONSIDERATIONS

Fission Products

For the particular case of MFPs for which bioassay was performed with results as only gross beta or 

gamma measurements, ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal 

Dose-Related Gross Beta and Gross Gamma Analyses (ORAUT 2015), may be used to arrive at 

appropriate source terms for internal dose reconstruction.

The guidance of ORAUT (2015) 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 (2015) states that 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 FPs 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 2015) 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 

high-efficiency particulate air (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. Therefore, a few King Avenue laboratory workers and on occasion a few 

machine shop workers (e.g., BCL 1966–1993) would probably have been exposed to this type of 

source. Finally, at King Avenue, some test dissolution of clad lightly irradiated fuel pins occurred 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 FPs 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 JN-1, 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. 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 131I 

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 HEPA filters 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. Therefore, these workers probably 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 could 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 progeny 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 (2015). Urinalyses done specifically for 131I only (i.e., not along with other 

FPs 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 Table C-1.)

The cooling time is assumed to be 1 year for JN-1. This is consistent with Battelle’s 1962 accident 

analysis for JN-1 (Sunderman and Dickerson 1962) that assumed a 4-month cooling time for an 

Organic-Moderated Reactor Experiment (OMRE) element that was then sent to Battelle for analysis. 

The actual cooling time was usually longer than 4 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 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 the Oak Ridge National Laboratory (ORNL) and sent to Battelle for analysis (Dayton and Tipton 

1960a).

Uranium

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. 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 have been found for these samples, it is possible that they 

were used in an attempt to quantify uranium exposures from enriched uranium (EU), 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 

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

milligrams or micrograms) should be assumed to be natural uranium, using a specific activity of 0.683 

pCi/μg. For sample results expressed in total uranium activity or as isotopic results (with units 

expressed in dpm, microcuries, or becquerel), the measured activity can be assumed to be 234U for 

the purpose of intake and dose assessment in accordance with ORAUT (2014a). 

The extent of the processing of recycled uranium at Battelle is not well known. For all DOE uranium 

after 1952, this analysis assumed the possibility that uranium from refineries was recycled uranium or 

contained recycled uranium. Table 4-6 provides the activity as a fraction of uranium activity that 

should be applied to all uranium intakes after 1952 (NIOSH 2011c).
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Table 4-6. Activity fractions of contaminant in recycled uranium.

Recycled uranium 

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

Activity fraction of 

contaminant in uranium 

0.00246 0.00182 0.379 2.73E-06 2.73E-06

Thorium

Work with thorium metal (forging, rolling, etc.) was performed at King Avenue beginning in 1945 

(Dayton undated, p. 24) through 1952 (Thomas 1952), and records indicate thorium metal and ore on 

site periodically (Chapman 1947; Roth 1948; Saller 1947, 1950; AEC 1951a) through 1957 (Selander 

1957a; Peterseim 1957). This is consistent with thorium inventory records indicating large quantities 

of thorium on site (hundreds of kilograms) through 1957, with significant reductions in inventory after 

1957 (BCL 1988a, pp. 45–127). No record of similar work after this time was found, although 

laboratory research with small quantities of thorium is recorded for later periods through the 1960s 

(Norris 1960; Evans 1963). 

For early work evaluating methods of thorium extraction from monazite sands (1949-1950), thorium 

and progeny radionuclides would have been in equilibrium in the source material. For work with 

thorium at other periods, the degree of chemical separation of thorium from its progeny are generally 

unknown. Consequently, for evaluation of available thorium bioassay results, dose reconstructors 

should use the methods described in ORAUT (2014b) to determine intakes favorable to the claimant

with respect to the degree of disequilibrium of thorium from progeny radionuclides due to possible 

chemical separation.

In accordance with Battelle-SEC-00208 (NIOSH 2012) and Battelle-SEC-00229 (NIOSH 2015), 

internal dosimetry for thorium is considered infeasible from the beginning of site operations through 

December 31, 1970, due to lack of documented bioassay or air sampling records during periods in 

which thorium operations are assumed to have taken place based on records of thorium inventories 

and contamination surveys indicating potential thorium exposure. After this period, NIOSH has not 

identified data that suggest the potential for significant thorium exposure, and the Battelle internal 

dosimetry program had implemented radionuclide-specific bioassay monitoring to a degree so that 

workers performing significant work with thorium would have been monitored.

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 was 

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.
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5.0 OCCUPATIONAL EXTERNAL DOSE

5.1 DOSIMETRY PROCESSING SERVICES

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 in March 1951. 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

March 1951–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; thermoluminescent dosimeters (TLDs) 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 (BCL 1986). 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, 1962b, 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 
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(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 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)

03/1951–03/1956 50 Not available Not available Not available

04/1956–03/1957 30 50a 60b Not available

04/1957–1960 20 50 60 Not available

1961–1963 10 40 50c Not available

1964–1984 10 40 50 10

1985–06/1994 10 40 50 10

07/1994–1996 10 40 50 Not available

1997–present 10 40 20 10

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

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

is indicated, a 50-mrem MDL is recommended. 

b. Based on values recorded as “under 10% MPL,” the MPL for the period was 300 mR/wk, so 

10% of that value for a 2-week period is assumed to be 60 mR.

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

Later reports identify 20 mrem, and then 50 mrem from 1994 to 1996. Because it is unlikely that 

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

to 1996.

In accordance with the SEC described in Section 1.3 (Burwell 2016), NIOSH found that it was not 

feasible to reconstruct external doses from beta, gamma, and neutron radiation through February 13, 

1951. 

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 milliroentgen, 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 maximum permissible limit (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 2-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. Neutron exposures continued to be listed as described for 1956.

As early as May 1956, a category headed “Other” was occasionally included on reports. This 

category is assumed to represent beta dose because in February 1958 the 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., Ottman 1959, 1961). Units of 
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millirep for uranium beta exposures were also noted during this period; the heading “Beta” was not 

routinely included in reports until 1961.

For doses reported during this period as “other” or “arbitrary units” (sometimes abbreviated “a.u.”), 

uncorrected film readings were recorded representing the unshielded portion of the dosimeter reading 

(i.e., due to electrons), which were then converted to millirep of dose using factors derived based on 

assumed beta electron energy (Storm 1951; Landauer 1951). These values were then sometimes 

converted to reported dose to the skin (indicated as “Ds”), by adding the gamma dose to the converted 

electron dose. Battelle records occasionally include notations indicating either the assumed beta 

electron energy, the factor to be applied, or both. For exposures to “old” FPs (defined as greater than 

6 months old), the assumed energy was listed as <0.7 MeV with an associated multiplication factor of 

5.3. For exposures to “new” FPs (defined as less than 6 months old), the assumed energy was listed 

as <1.4 MeV with an associated multiplication factor of 2.2. Other factors were sometimes listed 

depending on the exposure conditions, but no associated assumptions have been found in the 

documentation other than for new or old FP exposures. For the period in which arbitrary units were 

used (i.e., before 1961), dose reconstructors should multiply recorded electron doses by factors as 

documented in the record, if available. For work with uranium, a factor of 1.6 is applied based on 

Landauer (1951) and notations in the records. For work with fission products, if no factor is 

documented in the dose record, but the exposure source is noted as “new fission products,” the dose 

reconstructor should apply a multiplication factor of 2.2 to the recorded value to obtain the shallow 

dose. If no factor or beta energy is documented in the dose record, the approach most favorable to 

the claimant is to assume exposure to old FPs and apply a multiplication factor of 5.3 to obtain the 

shallow dose. Note that the use of arbitrary units was discontinued beginning in 1961 (although 

exposure questionnaire forms continued to include the abbreviation “a.u.” for a few years afterwards). 

Beta doses reported in units of millirem or “millirep if U beta” do not need to be adjusted.

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 (Landauer 1956–1997). 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 1956–1997; SC&A 2016). 

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 ca. 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.
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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 (BMI undated; BCL 1973; Author unknown ca. 1975).

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 Yes No

Critical Assembly Laboratory (CAL) 6412A Yes No

Hot cella 6408 Yes No

Plutonium Laboratory 6409 Yes No

Uranium Laboratory (25 area) 6412B Yes No

Safety -- Yes No

Visitor biweekly 6412Z Yes No

Miscellaneous biweekly 6412D Yes No

Radioisotope Laboratory 2302 No Yes

Analytical chemistry 2366 No Yes

Radiography 2613C, 6412C No Yes

Visitor monthly 6413Y No Yes

Miscellaneous monthly 6412E, 6413E No Yes

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

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 (BMI undated; BMI 1962–1963; BCL 1973; Author unknown 

ca. 1975). Records from Battelle document that the primary exchange frequencies were biweekly 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 biweekly 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 2 consecutive years when 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 

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 
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 34 of 145

 

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 

2007, 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

Table 5-4 summarizes photon energy ranges for materials and their related radiation types at the King 

Avenue and West Jefferson sites (ORAUT 2009, 2010b, 2011b, 2013b). 

Table 5-4. Photon and energy ranges for material types. 

Energy

Natural or 

depleted uranium

Enriched 

uranium

Natural 

thorium, MFP Plutonium

<30 keV 0% 0% 0% 65%

30–250 keV 50% 100% 25% 35%

>250 keV 50% 0% 75% 0%

If information on the specific materials to which a worker was exposed cannot be determined, 100% 

30-250 keV photons should be assumed to provide a deep dose favorable to claimants. 

For shallow doses, either 100% >15-keV electrons (for nonplutonium work) or >30-keV photons (for 

plutonium work) should be used.

For neutron doses, a 100% 0.1 to 2 MeV energy should be assumed to provide a dose favorable to 

claimants. 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-OTIB-0024, Estimation of 

Neutron Dose Rates from Alpha-Neutron Reactions in Uranium and Thorium Compounds (ORAUT 

2005b).

For neutron doses, the measured doses should be modified by the ICRP Publication 60 correction 

factor for the energy range (ICRP 1991). This factor incorporates the ratio of the Publication 60 

weighting factor to the group-averaged National Council on Radiation Protection and Measurement 

(NCRP) Report 38 (NCRP 1971) quality factor. This factor is 1.91 for the 0.1 to 2 MeV energy range 

(ORAUT 2006a).

Because little information is available on the exposure geometry for an individual, the standard 

assumption is that all exposures are for AP geometry, which is favorable to the claimant for most 

organs. In accordance with the NIOSH External Dose Reconstruction Implementation Guideline

(NIOSH 2007), rotational and isotropic geometries must also be evaluated in accordance with that 

document for bone surface, red bone marrow, esophagus, or lung exposures. The geometry most 

favorable to the claimant from this comparison should be used for these organs unless workplace 

information indicates that AP geometry is applicable for a specific exposure situation.

Early film dosimeters were calibrated in roentgens. It is assumed that this continued until the 

implementation of TLD dosimeters in 1985, for which the dose equivalent, Hp(10), is the appropriate 

unit for calibration. Table 5-5 lists the dose units to use for organ dose conversion factors.
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Table 5-5. Photon dose units for use with organ dose 

conversion factors.

Years Dose unit

Start of operation–1984 Exposure - R

1985–present Deep dose equivalent - Hp(10)

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 (2011a) 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 FILM

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

for workplace radiation fields (Amstein 2010) as described in Attachment F. Due to the 

underresponse of NTA film dosimeters to low-energy neutrons, some neutron exposures might have 

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

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

dosimeter period should be assessed based on neutron-to-photon (NP) exposure ratios derived from 

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

radiation dose rates from selected Battelle instrument survey records. These ratios should be applied 

to measured and missed photon doses to obtain neutron doses for workers for the period when film 

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

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

used to for monitoring. The recommended NP ratios are summarized in Table 5-6, based on the 

analysis in Attachment F. 

Table 5-6. Recommended NP ratios for dose monitored by NTA film.

Building Location/Work Years n GM GSD 95th%

JN-1 Non-Cf-252 operations 1955–1983a (b) (b) (b) (b)

JN-1 Cf-252 0perations 1971–1974 49 2.4 3.9 22.5

JN-2 Critical Assembly 

Laboratory

Critical experiments 1956–1960 (b) (b) (b) (b)

JN-2 Plutonium Laboratory 1967–1975 14 1.0 1.3 1.5

JN-3 Battelle Research 

Reactor

Combined areas 1956–1974 2,487 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) (c) (c) (c) (c)

a. Neutron exposures probably very limited before construction and operation of alpha gamma cells in 1964.

b. No paired survey data received. 

c. No paired survey data received. Depleted U-238 used in this building with essentially no potential for significant neutron 

radiation.

Paired workplace radiation measurements (i.e., same time and location) 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 to 1983). Limited survey data has been received from Battelle 

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

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

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

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

Battelle-provided information for the Plutonium Laboratory in this building involved only sealed 

sources.

• JN-3 Battelle Research Reactor (BRR) (1960 to 1973). Substantial survey data was received 

from Battelle for the period from 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.

• JN-4 Plutonium Laboratory (1968 to 1976). 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 (DU) 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 Table 5-6. 

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, 1958b); the form was also used to document planned 

overexposures (e.g., Swigert 1963). While most of the copies of these forms found in the records

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.

Attachment E contains a detailed summary of incidents identified at the King Avenue and West 

Jefferson sites.

5.8 ONSITE AMBIENT EXTERNAL DOSE

Annual environmental reports in records obtained from Battelle range from 1973 through 2005, with 

the exception of 1999, for which no report was found (doses for 1999 are addressed as described 

below). These reports are listed in the Environmental Reports section of the References list. 

Evaluation of records indicates that quarterly measurements of external radiation using lithium fluoride 

TLDs began 1978 at the West Jefferson site at 14 monitoring locations within a 0.75-mi radius of the 

site and 25 monitoring locations at the security perimeter. The innermost monitoring locations at the 

security perimeter were used to derive ambient external dose values for the site in this document. 

Monitoring locations were changed to the “recreation area and property boundary line” in 1985 

(15 locations), increasing to 16 locations in 1991. In the fourth quarter of 2004, monitoring locations 

changed from 16 locations around the West Jefferson property boundary to 10 locations around the 

security perimeter of the innermost facility buildings.

Monitoring was also performed during remediation activities at 15 locations at the King Avenue facility 

from the fourth quarter of 1993 through 1998.In reporting potential dose to an individual based on 

environmental monitoring results, Battelle environmental reports indicate that monitoring results were 

consistently below levels expected from naturally occurring background radiation. Reports indicate a 
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terrestrial component of 0.060 rem/yr (based on aerial monitoring and stated to be consistent with the 

national average) and indicate a cosmic ray component for Ohio of 0.050 rem/yr (slightly above the

national average of 0.045 rem/yr). However, because exact contributions from offsite sources cannot

be determined, and to ensure results favorable to the claimant, this document assumes the entire 

reported result from environmental monitoring to be potentially due to site operations. 

For years in which the monitoring locations were at the security perimeter (1978 through 1984, and 

2005), the average annual doses were higher (0.145 rem) than when the monitoring locations were at 

the recreation area and property boundary line (1985 through 2004, with an average annual dose of 

0.110 rem). This suggests that onsite ambient doses to workers are more accurately represented by 

monitoring stations in closer proximity to the facility buildings. For this reason, the higher annual 

average from security perimeter locations was applied in this document for 1985 through 2004 rather 

than the lower annual doses from when the monitoring locations were farther from the facility 

buildings. This higher annual average dose is also considered favorable to claimants for 1999, for 

which no annual environmental report was located.

Results from King Avenue measurements were comparable to, but consistently less than, those from 

the West Jefferson site for the corresponding year. Therefore, reported values for the West Jefferson 

site are applied as bounding doses that are favorable to claimants for the King Avenue site for all 

years, which also accounts for individuals who might have traveled between the two sites.

Annual doses obtained as described above are for an assumed calendar year of 8,760 hours. These 

annual doses are prorated in this document to the dose that would be received by an occupational 

worker using an occupancy factor of 2,000 working hours per year.

From the beginning of operations through 1977, prior to the availability of records for security 

perimeter locations, because no site activities at West Jefferson are evident that would suggest a 

substantial deviation from operations during the monitored period, the highest annual doses occurring 

between 1978 and 2005 may be applied as favorable to claimants. For years after 2005, the average 

annual dose from security perimeter locations is applied. Because King Avenue results were 

consistently lower than those from West Jefferson, and because materials with greater external dose 

potential were consistently in use at West Jefferson, these values may also be applied for the King 

Avenue site.

Table 5-7 provides the reported averages for annual external exposures for the calendar years 

indicated as measured by environmental TLD badges. Table 5-8 provides the annual doses, prorated 

for an occupational year as described above and assumed to be from 30- to 250-keV photons, to be 

applied in dose reconstructions for individuals not monitored for external exposure, applied as 

lognormal distributions with a geometric standard deviation of 3.
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Table 5-7. Annual environmental dose (rem/yr) from external background 

radiation as measured by TLD.

Year West Jefferson King Avenue

Start of operations –

1977

No reports No reports

1978 <0.133 No report

1979 <0.144 No report

1980 <0.143 No report

1981 <0.128 No report

1982 <0.129 No report

1983 <0.130 No report

1984 <0.211 No report

1985 0.120 No report

1986 0.120 No report

1987 0.120 No report

1988 <0.120 No report

1989 <0.120 No report

1990 <0.132 No report

1991 <0.120 No report

1992 <0.120 No report

1993 <0.120 <0.030a

1994 <0.120 <0.030a

1995 0.094 0.090

1996 0.089 0.088

1997 0.091 0.088

1998 0.088 0.081

1999 No report No report

2000 0.109 No report

2001 0.105 No report

2002 0.096 No report

2003 0.104 No report

2004 0.111 No report

2005 0.143 No report

2006–present No reports No reports

a. Represents reported quarterly results, all of which were <0.030 (equivalent to annual 

value of <0.120)

Table 5-8. Prorated annual dose (rem/yr) from external gamma radiation. 

Year West Jefferson and King Avenue

Start of operations–1977 0.048a

1978 0.030

1979 0.033

1980 0.033

1981 0.029

1982 0.029

1983 0.030

1984 0.048

1985–present 0.033b

a. Based on maximum recorded annual value for all reported years.

b. Based on average annual dose from security perimeter locations.
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6.0 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATIONS

6.1 AIR SAMPLING AND RELEASE ESTIMATES

No environmental information is available for either the King Avenue site (operations began in 1943) 

or the West Jefferson Nuclear Sciences Area (operations began in 1956) until 1973. Information used 

to make estimates on inhaled radionuclides for this period is described below.

6.1.1 King Avenue, 1943 to 1972

No information is available for the King Avenue site until 1973. Release information for subsequent 

years is described below. 

6.1.2 King Avenue, 1973 to 1975

Most air emission monitoring and sampling of BCL facilities occurred at the West Jefferson site. King 

Avenue stack sampling results for total released activity are available for 1973 (presumably the sum 

of gross alpha and gross beta measurements). The King Avenue results are less than 4% of West 

Jefferson gross alpha plus gross beta results in 1973; separate gross alpha and gross beta releases 

are available for 1974 and 1975. King Avenue sampling was directed at Building 3, where activities 

with uranium compounds took place (BCL 1974). In June of 1975 the King Avenue stack sampling 

system was placed on standby as the reduced inventory of radioactive materials did not warrant 

continued sampling (BCL 1976). Site boundary air concentrations are reported, and typically 

represent MDAs. The maximum releases reported from the King Avenue facilities between 1973 and 

1975 were 3.18 × 10-8 Ci/y for gross alpha emitters (assumed to be uranium) and 2.46 × 10-6 Ci/y for 

gross beta emitters (assumed to be 137Cs). 

6.1.3 West Jefferson, 1956 to 1972

As noted above, no information is available for the West Jefferson Nuclear Science Area from before

1973. Release information for subsequent years is described below.

6.1.4 West Jefferson, 1973 to 2005

Stack release information for the West Jefferson site became more detailed from 1973 to 1975. In 

1973 only gross alpha and gross beta releases (3.6 and 29 μCi, respectively) were reported for JN-1, 

and 239Pu releases (0.36 μCi) were reported for JN-2 and JN-4. The gross alpha, gross beta, and 

239Pu released activities were similar to 1974 (1.9, 25, and 0.38 μCi, respectively) and 1975 (1.1, 25, 

and 0.36 μCi). Gross alpha and gross beta releases continued to decrease in subsequent years, to 

generally less than 1 μCi gross alpha and 10 μCi gross beta, as radionuclide-specific information 

became more detailed. In 1974 radionuclide-specific information was added for 131I and 85Kr. MFPs

were first identified individually in 1975, and 235U was first reported in 1977. The actual radionuclides 

reported varied from year to year, particularly for FPs; naturally occurring radionuclides such as 40K 

were sometimes included. Uranium-238 and 241Am began to be included in 1988 and 234U in 1996. 

Iodine-131 was not reported after 1980 and 85Kr after 1987. Where applicable, values below the MDA

were reported at that activity. Activities released per year, average in-stack concentrations, and site 

boundary concentrations were typically given for each stack every year (data were not available for 

1999). Figures 6-1 and 6-2 show the calculated intakes of alpha-emitting radionuclides and MFPs, 

respectively, from 1973 through 2005.

Air emissions were continuously monitored using in-stack air samplers on nine exhaust stacks from 

three different West Jefferson facilities during 1975, which was selected as a “typical” year for the 

period, the first for which relatively complete radionuclide release data were available (BCL 1976). 
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These release points, their descriptions, and the analyses conducted are shown in Table 6-1. 

Particulate filters were collected and analyzed weekly, while 131I was analyzed monthly and 85Kr 

releases were monitored continuously. Gamma analysis was conducted monthly on a composite of 

weekly samples from JN-1. MDAs were given as 3 x 10-16 μCi/mL for gross alpha; 2 × 10-16 μCi/mL

for gross beta, 131I, and 239Pu; and 1 × 10-10 μCi/mL for 85Kr.

The year 1975 represents a snapshot in time of West Jefferson operations in relation to air emissions. 

During 1975 the BRR (JN-3) was dismantled without incident after terminating operations December 

31, 19742 (BCL 1987). The number of air emission points changed over subsequent years as 

facilities and operations were changed or retired. By 1980, for example, stack 005 no longer 

appeared for JN-2, and stacks 013 and 014 were added to JN-1 (BCL 1981a). By 1986 stacks 007, 

008, and 009 were gone, while stack 012 had been added to JN-2 (BCL 1987). 

From 1973 to 2005 a total of 4.7 × 106 Bq (0.13 mCi) of alpha emitters other than 238U and 4.1 × 107

Bq (1.1 mCi) of MFPs were released. This release includes 32 of the 33 years during this period 

because data were not available for 1999. Antimony-125 (29%), 60Co (21%), 137Cs (15%) and 152Eu 

(9%) accounted for 73% of released MFP activity. Iodine-131 accounted for nearly 8% of released 

activity with releases in 1974 to 1977 and 1980, with over 90% of the release in 1974 alone. For 

alpha emitters, 241Am accounts for 50% of the activity (no activity before 1988) and 234U accounts for 

44% of the activity with the remainder due to 239Pu. From 1973 until 1988 (when 241Am began to be 

reported) the 235U and 239Pu activity was approximately equal. After that 235U and 241Am activity was

Figure 6-1. Occupational environmental inhalation of alpha emitters at West Jefferson. 

 

2. However, neither the 1973 (BCL 1974) nor 1974 (BCL 1975) annual environmental reports show an air 

emission point for the operating research reactor.
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Figure 6-2. Occupational environmental inhalation of MFPs at West Jefferson. 

Table 6-1. Air effluent release points at the West Jefferson site in 1975. 

Release point Exhaust stack description Analysisa

001 Gross alpha, gross beta, gamma, 85Kr

002 J

JN-1 Hot Cell (old building)

N-1 Hot Cell (new building) Gross alpha, gross beta, gamma, 131I

003 JN-1 Hot Cell (control area) Gross alpha, gross beta, gamma

004 Gross alpha, gross beta, gamma

005 JN-2 Plutonium Laboratory

JN-1 Hot Cell (liquid waste evaporator)

239Pu

006 JN-2 Vault 239Pu

007 JN-4 Old Plutonium Laboratory (east) 239Pu

008 JN-4 Old Plutonium Laboratory (west) 239Pu

009 JN-4 New Plutonium Laboratory 239Pu

a. Weekly for particulates, monthly composite gamma analysis, monthly 131I, continuous 85Kr. 

approximately equal while 239Pu activity was much smaller. Uranium-238 was not included because it 

was not reported until 1988 when most of the BCL activities were already complete; this likely 

represents naturally occurring radioactivity.

6.2 ESTIMATED ANNUAL INTAKES FROM AIRBORNE RADIONUCLIDES

Because the site boundary location is somewhat distant from the West Jefferson facilities (500 m) the 

site boundary concentrations reported in the annual environmental reports are not representative of 

onsite concentrations near the buildings (see the Environmental Reports section at the end of the 

References section). For this site profile, onsite concentrations were calculated using the activities 

released per year (sum from all stacks) and near-field ground-level concentration screening method 

from NCRP Report 123 (NCRP 1996). The details of the method are included in Attachment G. This 

calculation resulted in the derivation of an onsite atmospheric dispersion factor of 4 × 10-3 s/m3

, which 
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compares reasonably and conservatively with site boundary atmospheric dispersion factors reported 

to be around 3 × 10-5 s/m3 for the maximally exposed offsite individual at 500 m (BCL 1974, 1987). 

This conservative, generic atmospheric dispersion factor can also be used for King Avenue facilities.

Workers are assumed to be on site and potentially exposed for 2,000 h/yr. Workers are also 

assumed to breathe at a rate characteristic of “light activity” as described in ICRP Publication 66

(ICRP 1994). This rate is 1.2 m3

/h. 

King Avenue

Using the reported maximum annual releases from Section 6.1.2 above and applying an onsite 

atmospheric dispersion factor of 4 × 10-3 s/m3 derived in Attachment G, a breathing rate of 1.2 m3

/h, 

and an exposure time of 2,000 h/yr, annual intakes of 3.59 × 10-4 Bq/y for 234U and 2.78 × 10-2 Bq/y 

for 137Cs are calculated. At these levels of intake the dose to any organ is less than 0.001 rem/yr. 

Therefore, occupational environmental intakes of inhaled radionuclides do not need to be considered 

for King Avenue facilities for the period from 1973 through 1975. However, intake values for the West 

Jefferson site should be applied for King Avenue site workers who might have traveled between the 

two sites during this period.

West Jefferson

Occupational environmental inhalation of alpha-emitting radionuclides is dominated by 239Pu, 241Am, 

and 235U at various times during the West Jefferson operating history. Annual intakes of alphaemitting radionuclides are shown in Figure 6-1. Based on this information, dose reconstructors should 

apply the annual intakes for all detected radionuclides shown in Table G-1 of Attachment G. Prior to 

the period represented in Table G-1 for which individual alpha-emitting radionuclides are reported 

(beginning in 1973), it is assumed that highest reported intake values from 1993 conservatively 

represent prior operations for 1971 and 1972. 

Inhalation of MFPs varied over the operational period at West Jefferson. Annual intakes of MFPs are 

shown in Figure 6-2. As noted above the main constituents are 125Sb, 60Co, and 137Cs. Based on this 

information, dose reconstructors should apply the annual intakes for all detected radionuclides shown 

in Table G-2 of Attachment G. Prior to the period represented in Table G-2 for which individual fission 

and activation product radionuclides are reported (beginning in 1975), it is assumed that highest

reported intake values from 1977 conservatively represent prior operations from 1971 through 1974. 

The potential organ dose from two other MFPs that were released at West Jefferson was evaluated 

individually because these are not well represented by particulate MFPs. Iodine-131 is volatile and 

inhalation results primarily in dose to the thyroid, while 85Kr is a noble gas and external dose results 

from immersion in a plume of the gas. At the levels of release indicated for West Jefferson the 

potential organ doses from both radionuclides would be less than 0.001 rem/yr. Therefore, these 

radionuclides do not need to be considered separately from MFPs. 

Summary

Table 6-2 summarizes the intakes dose reconstructors should use for occupational environmental 

inhalation of radionuclides at the King Avenue and West Jefferson sites of the Battelle Columbus 

Laboratories. Values for the West Jefferson site should be applied for King Avenue site workers who 

might have traveled between the two sites. Intakes should be applied as lognormal distributions with 

a geometric standard deviation of 3.
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Table 6-2. Occupational environmental inhalation of radionuclides at BCL. 

Site Period Radionuclides Intake (Bq/yr) Distribution

King Avenue 1973–1975 U-234, Cs-137 Negligiblea Not applicable

West Jefferson 1971–1972 Alpha emitters Apply Table G1, 1993 intakes 

for all years

Lognormal

West Jefferson 1973–Present Alpha emitters Table G-1 Lognormal

West Jefferson 1971–1974 Fission/activation products Apply Table G2, 1977 intakes 

for all years

Lognormal

West Jefferson 1975–Present Fission/activation products Table G-2 Lognormal

a. Results in annual doses <0.001 rem. 
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7.0 RESIDUAL EXPOSURE

Based on the Report on Residual Radioactive and Beryllium Contamination at Atomic Weapons 

Employer Facilities and Beryllium Vendor Facilities (NIOSH 2011b), the West Jefferson site had the 

potential for residual contamination from 1976 to 1985 after the period in which weapons-related 

production occurred and the King Avenue site had the potential for residual contamination from 2001 

through the present. 

Individuals who worked at the West Jefferson and King Avenue sites 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 from 1976 to 1985 or the King Avenue residual contamination period from 2001 

through the present [3].
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8.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).

[1] King, Vincent. Oak Ridge Associated Universities (ORAU) Team. Dose Reconstructor. May 

2007.

The FP MDA for 1962 through 1991 was determined empirically by Thomas La Bone using 

information in available analytical records from Battelle.

[2] Robinson, Sallie. ORAU Team. 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. ORAU Team. 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.
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REFERENCES

Cited References

AEC (U.S. Atomic Energy Commission), 1947, Report on Research and Development. [SRDB Ref 

ID: 9269]

AEC (U.S. Atomic Energy Commission), 1949, Health Hazards in NYOO Facilities Producing and 

Processing Uranium (A Status Report - April 1, 1949), New York Operations Office, New York, 

New York, April 18. [SRDB Ref ID: 3931]

AEC (U.S. Atomic Energy Commission), 1950, collection of air sample analysis reports. [SRDB Ref 

ID: 8612]

AEC (U.S. Atomic Energy Commission), 1951a, Survey of Accounting Control over Source and 

Fissionable Materials, Battelle Memorial Institute, Columbus, Ohio, Survey #16, Chicago 

Operations Office, Chicago, Illinois, April 16. [SRDB Ref ID: 95610]

AEC (U.S. Atomic Energy Commission), 1951b, The Production of Uranium Feed Materials, AEC

194/3, AEC Director of Production, New York Operations Office, New York, New York, May 

22. [SRDB Ref ID: 4125]

American Trudeau Society, 1957, The Chest X-ray and Chest X-ray Surveys related to X-ray 

Radiation Effects and Protection from Radiation Exposure, Report of the Executive Committee 

of the American Trudeau Society, October 15. [SRDB Ref ID: 11325]

Amstein, T., 2008, Battelle Response to Questions about X-Ray Screening Program. [SRDB Ref ID: 

47125, pg. 4]

Amstein, T, 2010, untitled letter to C. Wallace (Stoller Corporation/Source One Management), 

Battelle, Columbus, Ohio, February 8. [SRDB Ref ID: 79065]

Anno, J. N., and A. M. Plummer, 1962, “The Battelle Research Reactor,” Research Reactor Journal, 

volume 2, number 2. [SRDB Ref ID: 24148]

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Radiological and Nonradiological Parameters, BCLDP-081699, Battelle Columbus 

Laboratories, Columbus, Ohio, August 16. [SRDB Ref ID: 102127]

BCL (Battelle Columbus Laboratories), 2001, Site Environmental Report for Calendar Year 2000 on 

Radiological and Nonradiological Parameters, BCLDP-100101, Battelle Columbus 

Laboratories, Columbus, Ohio, October 1. [SRDB Ref ID: 102128]
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 72 of 145

 

BCL (Battelle Columbus Laboratories), 2002b, Site Environmental Report for Calendar Year 2001 on 

Radiological and Nonradiological Parameters, BCLDP-090102, Battelle Columbus 

Laboratories, Columbus, Ohio, September 1. [SRDB Ref ID: 6731]

BCL (Battelle Columbus Laboratories), 2003, Site Environmental Report for Calendar Year 2002 on 

Radiological and Nonradiological Parameters, BCLDP-093003, Battelle Columbus 

Laboratories, Columbus, Ohio, September 30. [SRDB Ref ID: 102129] 

BCL (Battelle Columbus Laboratories), 2004, Site Environmental Report for Calendar Year 2003 on 

Radiological and Nonradiological Parameters, Battelle Columbus Laboratories, Columbus, 

Ohio, October 15. [SRDB Ref ID: 27736]

BCL (Battelle Columbus Laboratories), 2006, Site Environmental Report for Calendar Year 2004 on 

Radiological and Nonradiological Parameters, Battelle Columbus Laboratories, Columbus, 

Ohio, June 1. [SRDB Ref ID: 28407]

BCL (Battelle Columbus Laboratories), 2007, Columbus Closure Project Site Environmental Report 

for Calendar Year 2005 on Radiological and Nonradiological Parameters, Battelle Columbus 

Laboratories, Columbus, Ohio, February 15. [SRDB Ref ID: 102130]
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 73 of 145

 

GLOSSARY

23 recovery program

Program to recover thorium from various materials (e.g., monazite sands) using anion 

exchange in a carbonate solution.

25 Area

Uranium receiving and dispensing area in Building 3 (Materials).

Atomic Science Center

See Nuclear Sciences Area.

berry bucket or can

One-gallon can for holding 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 and experiment end of the Battelle Research Reactor pool, also called the 

Shielding Pool.

Davison

Predecessor or subsidiary of W. R. Grace and Company.

Engineering Area

Auxiliary (south) part of West Jefferson site, including Buildings JS-1 (Hot Isostatic Processing 

Facility for fabricating military reactor fuel) and JS-10 and JS-12 (studies of explosive forming 

and bonding techniques, ballistic studies using nuclear materials).

fissium

Simulated fission-product-containing material as a mixture of nonradioactive isotopes of those 

elements that form as radioactive fission products for study of chemical and physical behavior 

of fission-product-bearing material.

junior cave

Medium-shielded room similar to a hot cell 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 for moving items between the Low Level Cell and Alpha-Gamma Cell 4 

(JN-1).

Metmount

Fuel rod section specimen set in 1.75-inch frame to prepare it for metallurgical examination. 

Also the frame itself.

Myrnalloy

Thorium metal or thorium alloy.
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New Plutonium Laboratory

Part of Building JN-4.

Nuclear Research Center

See Nuclear Sciences Area.

Nuclear Sciences Area

Main (north) part of the West Jefferson site, including Buildings JN-1 (Hot Cell Laboratory), 

JN-2 (criticality assembly, adjunct Plutonium Laboratory), JN-3 (Battelle Research Reactor), 

and JN-4 (main Plutonium Laboratory).

Old Plutonium Laboratory

Original section of Building JN-4.

reactor pool

The core end of the Battelle Research Reactor pool.

sheep shed

Waste Storage Shed (associated with JN-1).

slug

At Battelle, 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).

Tuballoy, tuballoy

Uranium, uranium alloy, or Uranium-beryllium alloy (Tu was synonymous with uranium).

weasel

Special type of sample capsule, or the Battelle Research Reactor fast-flux irradiation facility in 

which it was used. The capsule held an aluminum container that had about 5 grams of 

powdered uranium dioxide (UO2) packed into quartz tubes. The facility was hydraulically 

operated and could be positioned in various regions near the core.

West Jefferson North

See Nuclear Sciences Area.

West Jefferson South

See Engineering Area.

White Cross

Hospital in Columbus used by Battelle.
 

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ATTACHMENT A 

CHRONOLOGY OF EVENTS

Table A-1. Chronology of events. 

Event

Period/

start 

date Description Referencea

Initial work for U.S. 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 undated 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, 1945b; 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). Mechanical beneficiation; 

pyro-, electro-, and hydrometallurgical 

methods; shotgun assays

Beyer 1948a, 1948b; Kelley 1948a; BCL undated 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, p. 7; Grenell 1947; Randall 1947; Keeler and Hare 

1954; Gallagher, Blosser, and Mann 1955, Vaughan, Bridge, and 

Schwartz 1957; Jankowski and Chastain 1958; Cunningham, 

Seaver, and Edgar 1959; Egen et al. 1960

Research on graphite-U piles 1947–? For the reactor at ORNL Author unknown ca. 1987; AEC 1947, p. 7 ; 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, 1947b, 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
ATTACHMENT A

CHRONOLOGY OF EVENTS (continued) 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 76 of 145

Event

Period/

start 

date Description Referencea

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 December 1947Brown 1951; Kelley 1950, 1951; Blatz 1949; Harris 1952; Kelley 

1948b, 1948c

Work for the Air Force’s 

Aircraft Nuclear Propulsion 

Program

1947–

1961

Work for the Air Force’s Aircraft Nuclear 

Propulsion Program

Peters 2001; BCL undated 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 Nuclear submarine fuel element 

prototype development

BCL undated 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. 1950; 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

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 mantlegrade 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 Ce-Pr mixture

1 Nov 

1951

In connection with studies of Th 

extraction from Th nitrate

Kelley 1951
ATTACHMENT A

CHRONOLOGY OF EVENTS (continued) 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 77 of 145

Event

Period/

start 

date Description Referencea

Process support for U refining, 

metal production

1952–

1959

For AEC, Mallinckrodt, NLO, mostly 

R&D, process studies

BMI ca. 1987; Author unknown ca. 1975; 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 EU at 

Battelle

1947 Development of EU fuel elements and 

assemblies

Saller 1947

King Avenue Buildings 2 

through 7 and 10 through 13 

constructed

Complet

ed 

1954–

1955

Buildings A and 1 were already in 

existence

BMI 1977

Extraction of Th from Brazilian 

monazite sludge

1954–

1955

Extraction of Th from Brazilian monazite 

sludge

Wallo 1981; Meeley, Snyder, and Filbert 1954

Development of APPR 1954–

late 

1960s

Development of APPR Peters 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)

Meaders et al. 1950; Ewing et al. 1950b; Bearse et al. 1953; 

Ewing, Kiehl, and Bearse 1955; Van Cleek, Macdonald, and 

Stephens 1956; Langston, Macdonald, and Stephens 1957

Start of operation of JN-1 1955 Start of operation of JN-1 BCL ca. 1974; BCL undated a, DOE 2003a

Start of operation of JN-2 1955 Started as a critical assembly facility BMI ca. 1987; DOE 2000a; Layendecker 1996; Peters 2001; 

Jensen 2003; BCL undated a; Kirsch 2000a

Experiments, other work in JN1 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 Peters 2001; DOE 2003b; Gallagher, Blosser, and Mann 1955; 

BCL undated 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 1997a

Subcontract weapons 

research for Sandia, etc.

1955–

1965

Also for Los Alamos, Rocky Flats, LLNL BMI ca. 1987

Postirradiation examination, 

testing of materials in JN-1 

1955–

1983

Including materials from BRR, other 

research, defense, and commercial 

power reactors

None
ATTACHMENT A

CHRONOLOGY OF EVENTS (continued) 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 78 of 145

Event

Period/

start 

date Description Referencea

Startup of JN-3 (BRR) Oct 1956 Startup of JN-3 (BRR) DOE 2003a; Peters 2001; Anno, Plummer, and Chastain 1958; 

Plummer, Anno, and Chastain 1960; BCL undated a; Kirsch 

2000b

BRR, ETR loop program work 

in JN-1 and JN-3 

1956–

1962

Including 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

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. 1958; 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 JN-1and JS-1 

(Hot Isostatic Process Facility)

1958–

1984

Naval reactors, R&D for Army and other 

defense programs

Peters 2001

BRR upgraded from 1 to 2 

MW

Mar 

1959

BRR upgraded from 1 to 2 MW Plummer, Anno, and Chastain 1960; Anno, Plummer, and 

Chastain 1958

BRR beam tube shut down for 

furnace repair

Jun–Dec 

1960

BRR beam tube shut down for furnace 

repair

None

Construction of Building JN-4 

(old Plutonium Laboratory)

1960 Construction of Building JN-4 (old 

Plutonium Laboratory)

Peters 2001; Flynn et al. 1987

R&D: coating of U oxide, new 

fuel forms and materials, 

containment methods

1960s R&D: coating of U oxide, new fuel forms 

and materials, containment methods

Author unknown ca. 1987; Chastain 1961

AEC license to fabricate U fuel 

elements

1960s — Author unknown ca. 1987
ATTACHMENT A

CHRONOLOGY OF EVENTS (continued) 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 79 of 145

Event

Period/

start 

date Description Referencea

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; UN-PuN 

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; DOE 2000b; Swigert 1963; Taylor 1983

Addition of alpha-gamma test 

cells to JN-1

1964 Addition of alpha-gamma test cells to JN1Peters 2001; Martin, Storhok, and Gates 1966; BCL ca. 1974

Addition to JN-4 1964 Addition to JN-4 Flynn at al. 1987

RTG work with Pm sources 1964–

1968

RTG work with Pm sources Ritzman et al. 1966

King Avenue: continued 

defense program work

1965–

1975

Included fabrication of HEU 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 Further addition to JN-4 Flynn at al. 1987

Addition of Mechanical Test 

Cell to JN-1

Apr 1967 Addition of Mechanical Test Cell to JN-1 Peters 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 LLNL; handled 

encapsulated Pu only. Vault still used

Peters 2001; Kirsch 2000a

Work for NLO under prime 

AEC contract

~1970? Involved test quantities of radioactive 

metal

Author unknown ca. 1975

End of active nuclear work in 

JN-2

1970 — Peters 2001

Added High-Energy Cell to JN11972–

1973

Allowed for examination of full-sized fuel 

assemblies

Peters 2001; BCL ca. 1974, 1997a

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 LLNL; handled 

encapsulated Pu only. Vault still used

Peters 2001; Kirsch 2000a
ATTACHMENT A

CHRONOLOGY OF EVENTS (continued) 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 80 of 145

Event

Period/

start 

date Description Referencea

Production of Cf-252 sources 1972–

1980s?

Sources made from Cf wire in two alphagamma cells; unclear where wire came 

from or for whom work was done

Peters 2001; Scotti and Martin 1972

Final shutdown of the Battelle 

Research Reactor

1974 Final shutdown of the Battelle Research 

Reactor

DOE 2003a; Layendecker 1996; BCL undated a; Kirsch 2000b

Added fuel pool, Washdown 

Room to JN-1

1975 Associated with the High Energy Cell BCL 1997a

Partial decommissioning of the 

BRR to NRC criteria, with final 

survey

1975 Partial decommissioning of the BRR to 

NRC criteria, with final survey

Rubadue 2000; DOE 2003a, 2003b; Layendecker 1996; Peters 

2001, BCL undated a; Kirsch 1975, 2000b; Kok 1975a

Removal of last reactor fuel 

elements from BRR (JN-3), 

shipment to Savannah River 

Plant

March 

1975

Removal of last reactor fuel elements 

from BRR (JN-3), shipment to Savannah 

River Plant

Layendecker 1996; Kok 1975b

Decontamination of CAL/Pu 

Laboratory (JN-2)

1975 Decontamination of CAL/Pu Laboratory

(JN-2)

Peters 2001; BCL undated 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 Visit by AEC, ANL personnel to assess 

the need for a full survey and 

decontamination

DOE 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

DOE ca. 1987, DOE 2000c

Decontamination of JN-4 and 

Buildings A and 1

1978–

1982

Battelle itself performed the JN-4 

decontamination, using contractors for 

specialty work.

DOE ca. 1987, DOE 2000b; Wissinger 1978; Freas and Madia 

1982; Rudolph, Kirsch, and Toy 1984; Clements 1981; Stellrecht, 

Freas, and Dettorre 1979; Madia et al. 1979; Flynn at al. 1987; 

BCL ca. 1982

Cask sabotage program work 

in JN-1

1979–

1983

In the pool and the Controlled Area; 

Myers et al. (1994) says 1981–1983.

Peters 2001; Author unknown 2003; Myers et al. 1994a; BCL 

1997a

Further radiation surveys 1980–

1982

Probably performed by Battelle DOE 2000b

Research on destruction of 

chemical agents using highlevel 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)

Peters 2001
ATTACHMENT A

CHRONOLOGY OF EVENTS (continued) 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 81 of 145

Event

Period/

start 

date Description Referencea

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 2003a; Langsam and Carter 1986

Removal of last reactor fuel 

elements from JN-1, shipment 

to INL

1986 Peters (2001) says 1987. Actual 

removal was apparently in 1986, license 

termination in 1987.

Peters 2001

Decontamination 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.

DOE ca. 1987, DOE 2000c; Langsam and Carter 1986; 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 (undated a) says 1988, but this 

appears to be in error.

Peters 1998

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 D&D begins under the BCLDP Weaver et al. 2003; Peters 2001

Verification surveys of 

Buildings JS-1, JS-10, and JS11 by ORAU, with release by 

DOE/NRC

1989–

1990

Verification surveys of Buildings JS-1, 

JS-10, and JS-11 by ORAU, with release 

by DOE/NRC

DOE 2003a; ORAU 2006a; Peters 2001

Further King Avenue D&D (by 

Battelle?)

1990 Done under DOE’s Surplus Facilities 

Management Program

DOE 2000c

Final verification survey of 

Building 7A

1993 Final verification survey of Building 7A ORAU 2006a

Final verification survey of 

Buildings 3 and 4

1995 Final verification survey of Buildings 3 

and 4

ORAU 2006a

Final verification survey of 

Building 5 by ORAU

1996 Final verification survey of Building 5 by 

ORAU

ORAU 2006a

Final verification survey of 

Buildings 1, 6, 7 by ORAU

1997 Final verification survey of Buildings 1, 6, 

7 by ORAU

ORAU 2006a
ATTACHMENT A

CHRONOLOGY OF EVENTS (continued) 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 82 of 145

Event

Period/

start 

date Description Referencea

Completion of D&D of King 

Avenue site

1998 Except for Building 2? See 1999 entry 

below.

Peters 2001

Beginning of D&D of JN-1 1998 Beginning of D&D of JN-1 Weaver et al. 2003

Final verification survey of 

Building 2 by ORAU

1999 Final verification survey of Building 2 by 

ORAU

ORAU 2006a

Emptying, removal of the JN-1 

alpha-gamma cells

1999–

2000

Emptying, removal of the JN-1 alphagamma cellsDOE 2003a

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 2002a

Completion of D&D of King 

Avenue facilities, final 

verification survey of Building 

A by ORAU

2000 Completion of D&D of King Avenue 

facilities, final verification survey of 

Building A by ORAU

DOE 2003a, ORAU 2006a

Removal of the BRR bioshield FY 2001 Removal of the BRR bioshield DOE 2003a

Nuclear support, surveillance/ 

maintenance activities; some 

decontamination conducted at 

West Jefferson

2001 JN-1, JN-2, and JN-3 were involved. BCL 2002a

Completion of RCRA closure 

requirements for the King 

Avenue site

25 Jul 

2001?

Completion of RCRA closure 

requirements for the King Avenue site

BCL 2002a

Surveys of JN-2 and JN-3 by 

ORAU

2004 Surveys of JN-2 and JN-3 by ORAU 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.
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 83 of 145

 

ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES

LIST OF TABLES

TABLE TITLE PAGE

B-1 King Avenue Site building descriptions................................................................................... 84

B-2 West Jefferson Site building descriptions ............................................................................... 89
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 84 of 145

Table B-1. King Avenue site building descriptions. 

Building

Location or 

subarea

Room, area, 

or alternate 

name Description and MED/AEC use Later uses and status

Years of 

operation

A Entrance was 

through 

Building 4 or 

Building 11 

(BMI 1977)

Thorium solvent extraction pilot plant (DOE ca. 1987, 

Jensen 2003), ore processing (including carnotite), 

metallurgical studies (DOE ca. 1987); encapsulation of 

HEU 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

A 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) Not 

applicable

A 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 

Laboratory for several years. 

Residual activity found in Room 58 

complex in late 1980s (DOE ca. 

1987)

Unknown

A 1st floor Not applicable Ore processing, monazite sand solvent extraction 

(Jensen 2003)

Unknown Unknown

A S end of 1st 

floor in E 

wing

163 through 

163F

Upper part of high bay area of 58 complex Offices Unknown

A NW 

basement of 

W wing

6, 6A through 

6D, 7

Machine shop operations (BMI 1977) Converted to library facility by 

1977 (BMI 1977)

Unknown

A North 

basement, S 

of E-W 

corridor

17 Mechanical Test area (BMI 1977) Laboratory record storage Unknown

A 2nd floor Not applicable Laboratory used for tracer work (Sunderman and 

Dickerson 1962)

Unknown Unknown
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 85 of 145

Building

Location or 

subarea

Room, area, 

or alternate 

name Description and MED/AEC use Later uses and status

Years of 

operation

A N end, 2nd 

floor, S of EW corridor

219, 219C, 

219D

Metallography sample examination (BMI 1977), storage Reprint and brochure distribution 

offices

Unknown

A 4th floor Not applicable Pilot plant encapsulation of HEU for ATR fuel elements 

(Jensen 2003, DOE 1999), presumably in the 

Encapsulation Laboratory; Nuclear & Flow Systems 

Shop, Room A-467, supported the Encapsulation 

Laboratory (Browne 1974). Tracer-level laboratory (<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)

Unknown

1 Not 

applicable

Foundry Foundry (Langsam and Carter 1986, Jensen 2003); 

melting, cutting, grinding facilities (Jensen 2003). NU, Th 

extraction (DOE 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

1 1st floor (the) Foundry Uranium processing (BMI 1977; DOE 1999) in the tons 

(Jensen 2003), ore beneficiation studies (Jensen 2003)

Unknown Unknown

1 High bay at S 

end of 2nd 

floor, 

adjacent 

area

1223, 1223A, 

1223B

Uranium production from ore (BMI 1977, DOE 1999); 

beneficiation operations with ores such as carnotite (BMI 

1977). Miscellaneous physical handling of materials in 

1223; equipment storage in 1223A, miscellaneous

laboratory work in 1223B

Room 1223 complex found to 

contain residual activity in ~1987 

(DOE ca. 1987), after cleanup

?–1986

1 Near 2nd 

floor high bay 

area

1219 and 

1219A

Ore processing Ore laboratory Unknown

1 4th floor Not applicable Laboratory used for tracer-level (<100 μCi) work 

(Sunderman and Dickerson 1962)

Unknown Unknown

2 Not 

applicable

Metalworking 

Building 

(Langsam and 

Carter 1986, 

Jensen 2003); 

“Tisdale 

Building”

AEC research, including electroplating Hanford reactor 

slugs (Jensen 2003, DOE 1999); heat treatment, 

fabrication of U, Th alloys (Jensen 2003); fabrication of 

U, Th slugs (DOE 1999). 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)

Unknown

2 Not

applicable

Metalworking 

Laboratory

Ore processing (Jensen 2003 says “tons”); metalworking 

(Jensen 2003)

Unknown Unknown
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, 

or alternate 

name Description and MED/AEC use Later uses and status

Years of 

operation

2 Not 

applicable

Welding 

Laboratory

NU, DU, EU welding and fabrication (Jensen 2003) Unknown Unknown

2 2nd floor Not applicable Miscellaneous R&D, heat treatment, electroplating 

(Jensen 2003)

Unknown 1954–1986

3 Not 

applicable

U-235 

Processing 

Buildings; 

Materials 

Building 

(Langsam and 

Carter 1986, 

Jensen 2003)

Research activities using NU, DU, EU (Jensen 2003, 

DOE 1999). Powder metallurgy, melting, metallographic, 

ceramics research facilities (Jensen 2003). 

Miscellaneous R&D: ceramics research; metallographic 

research with tons of source material (Jensen 2003)

Used as an SNM storage and 

handling point since early times. 

Source material fabrication for 

DOD, 1975 on (BMI ca. 1987).

3 1st floor Room 3161 

complex, the 

“U-235 area”; 

“25 

Area/Room”

U-235 processing facility; receiving/distribution point for 

unirradiated sources, SNM, including enriched material 

and thorium (BCL 1975, Jensen 2003, Langsam and 

Carter 1986); also for occasional special samples (Stickel 

1993). Pickling of SNM to remove oxide for use in 

Building 3, other Battelle areas; solidification of pickling 

solutions by evaporation; salts stored here for later 

recovery. U reduction by calcining for reprocessing; 

hydrides, carbides, nitrides also made for use elsewhere 

at Battelle; packaging, storing of wastes from other 

areas. Metallography laboratory, ceramics research 

facility (DOE 1999)

Glovebox, hoods, shower, 

calciner, cutting wheel, prefilter, 

HEPA filter. Decommissioning 

activities completed in 1994, 

building returned to Battelle

Unknown

3 1st floor Room 3110 Storage area (AEC 1950) Unknown 1954–1975

3 2nd floor or S 

end of 1st 

floor

Melting 

Laboratory/ 

Facility, Room 

3178

Melting Laboratory/Facility (Langsam and Carter 1986, 

BCL 1975, Jensen 2003, Kizer 1971). U-235 materials 

used, 1954–1975; tons of source material (Jensen 2003). 

Metallographic research (Jensen 2003)

Had hoods Unknown

3 N end of 

basement

Powder 

Metallurgy 

Laboratory

Powder Metallurgy Laboratory (Jensen 2003, DOE 1999, 

BCL 1975). NU storage, source processing (Jensen 

2003, BCL 1975)

Unknown 1954–1986
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, 

or alternate 

name Description and MED/AEC use Later uses and status

Years of 

operation

4 1st floor Radiochemistry 

(Radioisotope) 

Laboratory 

(Langsam and 

Carter 1986, 

Jensen 2003, 

Sunderman 

and Gates 

1965)

Entrance to Building A was through Building 4 (BMI 

1977). Radiochemical procedures, radiotracer studies, 

industrial applications of radioisotopes, neutron 

dosimetry, other operations with radioactive materials 

(Sunderman and Gates 1965). Radioisotope Laboratory 

main facility was divided into low-level, 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 1998, Sunderman 

and Dickerson 1962, Sunderman and Gates 1965). Also 

had Radiography Laboratory 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 Laboratory in the 

Radiochemistry Laboratory) 

(Sunderman and Dickerson 1962). 

Decommissioning completed in 

1994

Unknown

4 2nd, 4th 

floors

Not applicable Miscellaneous labs on 2nd, 4th floors, including

encapsulation facility for HEU, metallography facilities 

(Jensen 2003, DOE 1999). Had Tribology Shop in 

Rooms 4022 and 2023 during some period before 1994

Unknown 1954–1986

5 Not 

applicable

Machine Shop Work for AEC/ERDA/DOE programs: machining (Jensen 

2003, DOE 1999), grinding (Jensen 2003, DOE 1999), 

milling (Jensen 2003, DOE 1999). Be machining 

(Jensen 2003)

Source material fab for DOD, 1975 

on (BMI ca. 1987). D&D 

completed in 1995 (DOE 1999)

Unknown

5 Not 

applicable

Machine shop 

(Jensen 2003; 

DOE 1999; 

Langsam and 

Carter 1986)

Machining of DU, NU, and EU and some Th (Jensen 

2003)

Part of the floor of the machine 

shop was replaced at some point 

to remove contamination

Unknown

5 Not 

applicable

Coating 

Laboratory

Coating (of materials) (Jensen 2003) Unknown Unknown

5 Miscellaneou

s R&D on 1st 

and 2nd 

floors

Not applicable Not applicable Unknown 1954–1986
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, 

or alternate 

name Description and MED/AEC use Later uses and status

Years of 

operation

6 Miscellaneou

s 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 1999)

Unknown

6 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 Miscellaneou

s 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, including instrumental analyses (Jensen 2003), 

over a (unspecified) period of ten years (DOE 1999). 

Analytical chemistry on source materials and corrosion 

studies (Jensen 2003)

Annex (Building 7A) was 

decontaminated, released to 

Battelle in 1993 (DOE 1999); the 

significant radioactive operation in 

7A was a uranium fluoridation pilot 

plant (Basham 1993)

1954–1986

9 Miscellaneou

s R&D on 

ground floor

Mechanical 

Engineering 

Building 

(Langsam and 

Carter 1986, 

Jensen 2003)

Research programs for AEC/ERDA/DOE: handling NU

and DU uranium, krypton studies (Jensen 2003)

Hoods in laboratory area. The 

building was decontaminated and 

released to Battelle in 1991 (DOE 

1999)

1954–1986
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 89 of 145

Table B-2. West Jefferson site building descriptions

Building

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-1 Not 

applicable

Hot Cell 

Facility/ 

Laboratory; 

JN-1A was 

the original 

building, 

added to in 

the 1960s; 

JN-1B was 

major addition 

of 1970–1971

Fuel development 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 2002a, DOE 1999, 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 leadglass oil-filled viewing window (DOE 2003a). Support 

facilities: cask handling, solid and liquid waste disposal, 

contamination control, equipment decontamination, 

ventilation mezzanines (DOE 2003a, 2003b; Norris 1961; 

Sunderman and Dickerson 1962)

Extensively modified in 1960 and 

1979 (DOE 2003b). All cells in 

JN-1 have very high residual 

radiation fields (DOE 2000a). 

Water in the fuel transfer pool was 

removed in 1996 and packaged 

TRU waste was then stored there 

(DOE 2003a). Main offices were 

on the first floor near the 

laboratories (Peters 2001).

1955–1986

JN-1 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 2003b). Heat 

treatment furnace in CAA (Selander 1959b); ventilated 

cubicles for work with small shielded specimens 

(Sunderman and Dickerson 1962); small hot room with pit, 

sinks, and hot drain for equipment 

assembly/decontamination, 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 (Peters 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 (Peters 2001, Myers et al. 

1994b)

Unknown 1955–1986
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-1 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 2000a, 2003a, 2003b; 

Langsam and Carter 1986; Rubadue 2000) designed to 

provide shielding for 10,000 Ci and 10 million Ci of 1-MeV 

gamma emitter, respectively (Peters 2001, BCL undated b, 

Sunderman and Dickerson 1962). High-density concrete 

walls (ferrophosphorus aggregate at 293 lb/ft2) up to top of 

the manipulators, ordinary concrete after that (DOE 2003b). 

Three operating stations each (BCL 1977a, Sunderman and 

Dickerson 1962). Metmounts cut and ground in the LLC 

(Peters 2001), which was also used for tensile tests of Co60 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 1997a). Irradiated 

Co-60 rods were also cut up here 

(Myers at al. 1994d, BCL 1997a) 

1955–1986

JN-1 Section 

JN-1A, 1st 

floor

Charpy 

Cell/Room; 

Dry/Radioacti

ve 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 (Peters 

2001, DOE 2003b, Sunderman and Dickerson 1962, Myers, 

Berchtold, and Sands 1995, Myers, Redd, and Berchtold 

1994a); radioactive 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

JN-1 Section 

JN-1A, 1st 

floor

Transuranic 

vault

Storage Unknown 1955–1986

JN-1 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)

Unknown 1961–1986

JN-1 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. (Peters 2001, Myers et al. 

1994e)

Unknown 1964–1986
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-1 Section 

JN-1A, 1st 

floor

East machine 

shop

(Peters 2001, Myers et al. 1994e) Unknown 1964–1986

JN-1 Section 

JN-1A, 

basement

10 small 

segmented 

alpha-gamma 

cells (Peters 

2001, DOE 

2003b, 

Rubadue 

2000, DOE 

2003a)

Cells 1-9 constructed in 1964 for metallography testing of 

fuel rod specimens; Cell 10 added in early 1970s (Peters 

2001, BCL ca. 1974, Myers et al. 1994f). Shared 

working/access area (BCL undated 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, 

leak-tested. 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 

undated b, 1997a). 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 

(Peters 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 (Peters 

2001, Myers, Berchtold, and 

Sands 1995).

1967–1986

JN-1 Section 

JN-1A, 1st 

floor

Mechanical 

Test Cell

Main mechanical test cell (Rubadue 2000, DOE 2003b, 

Myers, Redd, and Berchtold 1994b) added in 1967 (BCL ca. 

1974). Three operating stations (BCL undated b). Highdensity concrete front wall(s) (ferrophosphorus aggregate at 

293 lb/ft2) to the top of the manipulators, ordinary concrete 

above that (DOE 2003b, DeMastry, Lusk, and Gates 1967). 

Unspecified thickness of iron plate belatedly added to one 

wall for shielding reasons (DOE 2003b)

Mainly for tensile tests, but also for 

creep, vacuum fusion, burst, radial 

burnup, density, and expanding 

mandrel tests (Myers, Redd, and 

Berchtold 1994b, BCL 1997a) 

1967– 1986

JN-1 Section 

JN-1B, 1st 

floor

Cask 

Washdown 

Room

For cleaning newly arrived casks, etc. (DOE 2003b, Myers 

et al. 1994b, Myers, Berchtold, and Stickel 1995); also 

entry-exit area for High-Energy Cell (DOE 2003b, Myers, 

Berchtold, and Stickel 1995). Walls open at top for crane 

access (DOE 2003b, Myers, Berchtold, and Stickel 1995). 

Compactor under hood (Myers, Berchtold, and Sands 1995)

Unknown 1970–1986
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-1 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 2003b)

Unknown 1970–1986

JN-1 Section 

JN-1B, 1st 

floor

Fuel Storage 

Pool

Storage of fuel assemblies, rod bundles, rod holders, tools 

(DOE 2003b, 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)

Unknown Unknown

JN-1 Section 

JN-1B, 1st 

floor

Pump Room Held wastewater storage tank, deionized water tank, and ion 

exchange tanks used to process the liquid radioactive waste

(Myers, Berchtold, and Sands 1995, Myers, Berchtold, and 

Stickel 1995)

Unknown 1972–1986

JN-1 Section 

JN-1B, 1st 

floor

High Energy 

Cell

Largest hot cell (DOE 2003b, DOE 2003a), five operating 

stations (BCL undated b). Power plant fuel element testing, 

evaluation (DOE 2000a). Access via door, ceiling plugs, or 

floor trap door to transfer canal (DOE 2003b), then to fuel 

transfer pool (Rubadue 2000, DOE 2003a). Seven storage 

holes in floor (DOE 2003b). Two walls barytes (at 220 

lb/ft2), two ordinary concrete (DOE 2003b). Cask washdown 

room at entryway (BCL undated 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 highburnup fuel assembly (BCL 

undated b, 1997a); nondestructive 

inspection and testing (BCL 1997); 

sectioning of rods (BCL 1997a); 

fission gas measuring tests done 

here (Myers et al. 1994b, BCL 

1997a); studies and 

characterization of Three Mile 

Island resins (BCL 1997a); effects 

of Co-60 on instrumentation (BCL 

1997a)

Unknown

JN-1 Section 

JN-1B, 1st 

floor

Old Back 

Dock, Waste 

Control Area

In early years, used to collect and package waste and as 

the original JN-1 dock; later, used for storage of a large 

number of highly contaminated items (Myers at al. 1994g)

Unknown 1970–1986

JN-1 Section 

JN-1B, 1st 

floor

Old Back 

Dock, Waste 

Control Area

In early years, used to collect and package waste and as 

the original JN-1 dock; later, used for storage of a large 

number of highly contaminated items (Myers at al. 1994g)

Unknown 1970–1986

JN-1 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 2003b, 2003c; Myers and 

Berchtold 1994). Two internal rooms/areas were heavily 

shielded (Myers and Berchtold 1994)

Unknown Unknown
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-1 With JN-1, 

but not 

attached

Waste 

Management 

Shed

Storage of highly contaminated containerized waste, 

equipment, other material (DOE 2003b, 2003c)

Unknown JN-2: 1955–

1963; 1964–

1970. Vault 

to at least 

1996

JN-2 Not 

applicable

CAL, Critical 

Assembly 

Facility, 

Critical 

Assembly 

Building; later 

also a 

Plutonium 

Laboratory

Started operation in 1955 (Peters 2001). 1956–1963: 

critical assembly studies (DOE 2003b, DOE 2000a, 

Chamberlain ca. 2003, Jankowski et al. 1957, Jankowski 

and Chastain 1958, Egen at al. 1960); zero-power organicmoderated critical assembly and other experiments (Jensen 

2003). 1964(?)–1970, used for other nuclear projects, 

including direct conversion concepts/Plutonium Laboratory

(DOE 2000a, 2003a, 2003b; Jensen 2003; Chamberlain ca. 

2003; Kirsch 2000a), irradiation experiment assembly (DOE 

2003b, Chamberlain ca. 2003), special laboratory (Jensen 

2003). Had Pu, enriched U storage vault (DOE 2000a, 

2003a, 2003b; Jensen 2003; Rubadue 2000; Freas et al. 

1971; Evans and Woodward 1978; BCL 1977a); control 

room (Jankowski et al. 1957, Hogan et al. 1958, Jankowski 

and Chastain 1958); attached high-bay area (DOE 2000a, 

2003a; 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/lowpower mockups of prototype 

reactors (Kirsch 2000a). Later, a 

laboratory installed for work with 

encapsulated Pu for LLNL. 1970: 

end of active nuclear work 

(Chamberlain ca. 2003, Peters 

2001). 1975: first decontamination 

(Peters 2001). Was Admin 

Building, had Accountability 

Laboratory to ~1987 (Langsam 

and Carter 1986). RAL in 

decontaminated former Pu 

Laboratory, 1978 to ~2003 (DOE 

2000b; Rubadue 2000; DOE 1999, 

2000a, 2003a, 2003b; 

Chamberlain ca. 2003; Peters 

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 1999). 

“Minimally contaminated” at final 

decontamination (DOE 1999, 

2000a; Jensen 2003)

1956–1974
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

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Building

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JN-3 Not 

applicable

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, 

Anno, and Chastain 1960), ended in December 1974 (DOE 

1999). Shielding facility located at end of thermal column 

(Morgan et al. 1958, Plummer, Anno, and Chastain 1960, 

Myers and Stickel 1995); gas loop facility added 1958 

(Bodnar et al. 1958; Plummer, Anno, and Chastain 1960, 

Basham and Rieder 1960, Anno and Plummer 1962); fast 

neutron activation facility added ~1960 (Jung, Epstein, and 

Chastain 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 2003a, 

Jensen 2003, DOE 1999). 

Carpenter shop, Health Physics 

and Instrument Laboratory to 

~1987 (Langsam and Carter 

1986). 1987 on: temporary 

storage of TRU wastes (DOE 

2003a, Jensen 2003), for which a 

special waste storage area was 

built in the basement (Myers and 

Stickel 1995)

1960–1977 

(DOE 

2000b, 

Rudolph, 

Kirsch, and 

Toy 1984)

JN-4 Not 

applicable

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 

(DOE 2000b, Smith 2000). Backup facility for LLNL

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 (Peters 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 LLNL, Rocky 

Flats, Army reactors (Author 

unknown 2003, Smith 2000). Two 

fume hoods; 43 gloveboxes in 

1971, 28 in 1978 (Peters 2001, 

Freas and Madia 1982, Rudolph, 

Kirsch, and Toy 1984). Presses, 

furnaces, grinders, lathes 

(Clements 1981). Old section 

removed by 1982 (BCL ca. 1982)

1976–

present

JN-6 Not 

applicable

(New) 

Guardhouse

Security operations (guardhouse)(Rubadue 2000, DOE 

2003b)

None 1965–1975?
ATTACHMENT B 

BUILDINGS AND OTHER FACILITIES (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 95 of 145

Building

Location or 

subarea

Room, area, or 

alternate name Description and MED/AEC use Later uses and status

Years of 

operation

JS-1 Not 

applicable

Not applicable Used to fabricate military reactor fuel by hot isostatic 

pressure bonding technique (DOE 1999, 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

1999); D&D was complete by 2003 

?–1988

JS-10 Not 

applicable

Not applicable Fabrication of uranium components using explosive forming 

techniques and ballistic experiments with DU (DOE 1999, 

Boehm and Groner 1986)

Facility characterization in 1998 

showed that no D&D was 

necessary (DOE 1999)

?–1989

JS-12 Not 

applicable

Not applicable Operations with radioactive materials in the ballistic tunnel, 

ductwork, and hardware using DU (DOE 1999, Boehm and 

Groner 1986)

Decommissioned in 1989 (DOE 

1999)

Unknown
Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 96 of 145

 

ATTACHMENT C 

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL

LIST OF TABLES

TABLE TITLE PAGE

C-1 History of work operations at the King Avenue site................................................................. 97
ATTACHMENT C

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 97 of 145

Table C-1. History of operations at the King Avenue site. 

Buildinga Period Isotopesb Formc Type(s) of workd

A 1943–1979 U, Th, progeny DU, NU, EU, Th 6.4 MT (Author unknown 1992) (apparently the total ever worked with in this 

building)

A 1943–1944;

1956–1956

U, progeny Uranate; U3O8; 

U2O2(NH4)2

Metallurgical R&D, studies of uranates

A 1943–1944;

1950–1952;

1958–1961;

1972–1975

U, progeny U carbides, mainly UC 

(NU; EU 9%–93%)

R&D, studies of U carbides and their alloys

A 1943–1962;

1979–1979

U, progeny U metal, alloys R&D, studies of work on U metal and its alloys

A 1946–1957 U, progeny 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)

A 1950–1971 U, progeny UO2 (NU, DU) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, etc.) (a 

few plates: DU)

A 1953–1954 U, progeny 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

A 1953–1959 U, progeny U metal, UO2, UO3, UNH Process support for U metal production for Mallinckrodt, NLO

A 1953–1959 U, progeny UF4 Process support for U metal production (some work involved UF4); study of UF4

as fused salt mixture for liquid fuel (1953–1954)

A 1956–1959 U, progeny U compounds (see at 

right)

Study of U compounds: U beryllides, borides, silicides, sulfides

A 1957–1958 U, progeny UCl Study of UCl dispersions (short period only)

A 1957–1961 U, progeny U nitrides, mainly UN 

(DU; NU; EU 5%–93%)

R&D, studies of U nitrides and their alloys

A 1958–1960 U, progeny UO2 (EU, 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% EU

A 1944–1961 Th, progeny 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)

A 1951–1955 Th, progeny 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 1943–1962 U, Th DU, NU, EU, Th 11.6 MT (Author unknown 1992)
ATTACHMENT C

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 98 of 145

Buildinga Period Isotopesb Formc Type(s) of workd

1 1943–1944;

1950–1952;

1956–1962

U, progeny U carbides, mainly UC 

(NU; EU 9%–93%)

R&D, studies of U carbides and their alloys

1 1943–1944;

1956–1956

U, progeny Uranate; U3O8; 

U2O2(NH4)2

Metallurgical R&D, studies of uranates

1 1943–1962 U, progeny U metal, alloys R&D, studies of work on U metal and its alloys

1 1946–1957 U, progeny 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)

1 1950–1965 U, progeny UO2 (NU, DU) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, etc.) (a 

few plates: DU)

1 1952–1958 U, progeny U metal, UO2, UO3, UNH Process support for U metal production for MED/AEC, Mallinckrodt, NLO

1 1952–1958 U, progeny UF4 Process support for U metal production (some work involved UF4); study of 

fused salt mixtures as liquid fuel (1953–1954)

1 1953–1954 U, progeny 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

1 1956–1959 U, progeny U compounds (see at 

right)

Study of U compounds: U beryllides, borides, silicides, sulfides

1 1956–1962 U, progeny UO2 (EU, 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

1 1957–1961 U, progeny U nitrides, mainly UN 

(DU; NU; EU 5%–93%)

R&D, studies of U nitrides and their alloys

1 1944–1961 Th, progeny; 

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); arc-welding Ames Th, Th 

alloys; metallurgical/mechanical 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 wt % U alloys; Th-U alloy made 

with thorium iodide; ingots cast by vacuum induction melting to Th buttons, 

casting Th-U carbides by arc-melting; ThN prep (1956–1961). Dissolution of Alcanned Th (1955)

1 1951–1955 Th, progeny 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 1954–1972 U, Th DU, NU, EU, Th 32.6 MT (Author unknown 1992)
ATTACHMENT C

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 99 of 145

Buildinga Period Isotopesb Formc Type(s) of workd

2 1954–1959 U, progeny U metal, UO2, UO3, UNH Process support for U metal production for Mallinckrodt, NLO

2 1954–1962 U, progeny UF4 Process support for U metal production (some work involved UF4) (1954–1958); 

prep of UO2 crystals from UF4 by vapor deposition (1959–1962)

2 1954–1962 U, progeny U metal, alloys R&D, studies of work on U metal and its alloys

2 1954–1971 U, progeny UO2 (NU, DU) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, etc.) (a 

few plates: DU)

2 1955–1957 U, progeny U3O8, (NH4)2U2O7, other 

forms(?)

U extraction from U ores (ammonium carbonate leaching) (1955–1957)

2 1956–1956 U, progeny U2O2(NH4)2 Study of prep of ammonium uranate

2 1956–1959 U, progeny U compounds (see at 

right)

Study of U compounds: U beryllides, borides, silicides, sulfides

2 1956–1962 U, progeny UO2 (EU, 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% EU

2 1956–1962 U, progeny U carbides, mainly UC 

(NU; EU 9%–93%)

R&D, studies of U carbides and their alloys

2 1957–1958;

1961–1961

U, progeny UCl (EU; NU?); U3O8

(NU?)

Study of UCl dispersions (EU, short period only, 1957–1958); making UCl from 

U3O8 (1961)

2 1957–1961 U, progeny U nitrides, mainly UN 

(DU; NU; EU 5%–93%)

R&D, studies of U nitrides and their alloys

2 1961–1961 U, progeny HEU3O8, HEU Prep of Al-U (HEU) fueled glass fiber fuel plates for irradiation

2 1955–1965;

1971–1972

Th, progeny; 

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 wt % 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)

2 1956–1956 U content Fissium Fissium (“synthetic fissium alloy”): specimen fab, annealed; X-ray diffraction

3 1954–1972 U, Th DU, NU, EU, Th 39.5 MT (Author unknown 1992)

3 1954–1959 U, progeny U metal, UO2, UO3, UNH Process support for U metal production for Mallinckrodt, NLO

3 1954–1960;

1979–1979

U, progeny U metal, alloys R&D, studies of work on U metal; its alloys, including electroplating; vapor 

deposition of metals on U

3 1954–1962 U, progeny UF4 Process support for U metal production (some work involved UF4) (1954–1958); 

prep of UO2 crystals from UF4 by vapor deposition (1959–1962)

3 1954–1971 U, progeny UO2 (NU, DU) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, etc.) (a 

few plates: DU)
ATTACHMENT C

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 100 of 145

Buildinga Period Isotopesb Formc Type(s) of workd

3 1956–1959 U, progeny U compounds (see next 

column)

Study of U compounds: U beryllides, borides, silicides, sulfides

3 1956–1962 U, progeny UO2 (EU, 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

3 1956–1962 U, progeny U carbides, mainly UC 

(NU; EU 9%–93%)

R&D, studies of U carbides and their alloys

3 1957–1961 U, progeny U nitrides, mainly UN 

(DU; NU; EU 5%–93%)

R&D, studies of U nitrides and their alloys

3 1957–1958;

1961–1961

U, progeny UCl (EU; NU?); U3O8

(NU?)

Study of UCl dispersions (EU, short period only, 1957–1958); making UCl from 

U3O8 (1961)

3 1955–1965;

1971–1972

Th, progeny; 

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 wt % 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)

3 1956–1956 U content Fissium Fissium (“synthetic fissium alloy”): specimen fab, annealed; X-ray diffraction

4 1954–1982 U, Th DU, NU, EU, Th 12.0 MT (Author unknown 1992)

4 1954–1959 U, progeny U metal, UO2, UO3, UNH Process support for U metal production for Mallinckrodt, NLO

4 1954–1961;

1979–1979

U, progeny U metal, alloys (NU, 

10% EU)

R&D, studies of work on U metal and its alloys (NU, 10% EU)

4 1954–1962 U, progeny 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)

4 1954–1971 U, progeny UO2 (NU, DU) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, etc.) (a 

few plates: DU)

4 1955–1957 U, progeny U3O8, (NH4)2U2O7, other 

forms(?)

U extraction from U ores (ammonium carbonate leaching) (1955–1957)

4 1955?–

1965?

Fission gasese Fission gases Fission gas analysis in fission gas laboratory (in Radioisotope Laboratory)

4 1956–1962 U, progeny UO2 (EU, 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

4 1956–1956 U, progeny U2O2(NH4)2 Study of preparation of ammonium uranates, including X-ray diffraction

4 1956–1959 U, progeny U compounds (see next 

column)

Study of U compounds: U beryllides, borides, silicides, sulfides
ATTACHMENT C

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 101 of 145

Buildinga Period Isotopesb Formc Type(s) of workd

4 1956–1962 U, progeny U carbides, mainly UC 

(NU; EU 9%–93%)

R&D, studies of U carbides and their alloys

4 1957–1958;

1961–1961

U, progeny UCl (EU; NU?); U3O8

(NU?)

Study of UCl dispersions (EU, short period only, 1957–1958); making UCl from 

U3O8 (1961)

4 1957–1961 U, progeny U nitrides, mainly UN 

(DU; NU; EU 5%–93%)

R&D, studies of U nitrides and their alloys

4 1961–1961 U, progeny HEU3O8, HEU Preparation of Al-U (HEU) fueled glass fiber fuel plates for irradiation

4 1955–1955 Th, progeny ? Support for Th extraction activities?

4 1955–1965;

1971–1971

Th, progeny; 

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 wt % 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 Th-U (1960–1965); ThO2 alloys (1971–1972)

4 1956–1956 U content Fissium Fissium (“synthetic fissium alloy”): specimen fab, annealed; X-ray diffraction

4 1954 Tracers Tracers Radioisotopes in industrial control (tracer methods)

4 1958–1960 Not applicable Not applicable Radiochemistry of inclusion compounds: nitration of hydrocarbons, urea 

compounds

4 1967–1974 Unknown unknown Research on bioeffects of underground explosions at Amchitka Island, Alaska 

during/after Milrow, Cannikin tests (samples counted at Battelle?)

4 1982–1982? Not applicable Not applicable Radiation surveillance for Piqua Nuclear Power Facility: annual survey, soil 

sampling. Apparently done by Battelle (samples counted at Battelle?)

5 1954–1972 U, Th DU, NU, EU, Th 14.7 MT (Author unknown 1992)

5 1954–1959 U, progeny Only U metal? Process support for U metal production for Mallinckrodt, NLO, including dingot 

metal

5 1954–1960;

1979–1979

U, progeny U metal, alloys R&D, studies of work on U metal and its alloys, including extrusion cladding, 

roll-bonding, fuel element fabrication

5 1954–1971 U, progeny UO2 (NU, DU) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, etc.) (a 

few plates: DU)

5 1956–1962 U, progeny UO2 (EU, 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

5 1956–1962 U, progeny U carbides, mainly UC 

(NU; EU 9%–93%)

R&D, studies of U carbides and their alloys, including fuel element fab and 

coating

5 1957–1961 U, progeny U nitrides, mainly UN 

(DU; NU; EU 5%–93%)

R&D, studies of U nitrides and their alloys, including bonding, welding
ATTACHMENT C

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 102 of 145

Buildinga Period Isotopesb Formc Type(s) of workd

5 1956–1959 U, progeny U compounds (see at 

right)

Study of U compounds: U beryllides, borides, silicides, sulfides. Included ballmilling, fabrication

5 1955–1965;

1971–1972

Th, progeny; 

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 wt % U alloys; Th-U alloy made with Th iodide; Th, Th-U 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)

5 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 1954–1982 U, Th DU, NU, EU, Th 7.9 MT in Building 6, 8.7 MT in Building 7 (Author unknown 1992)

6/7 1954–1959 U, progeny U metal, UO2, UO3, UNH Process support for U metal production for Mallinckrodt, NLO

6/7 1954–1961;

1979–1979

U, progeny U metal, alloys R&D, studies of work on U metal and its alloys, including corrosion studies

6/7 1954–1962 U, progeny 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)

6/7 1954–1971 U, progeny UO2 (NU, DU) R&D on UO2 used in fuel of various types (powder, pellet, plate-type, etc.) (a 

few plates: DU)

6/7 1955–1957 U, progeny 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)

6/7 1956–1956 U, progeny U2O2(NH4)2 Study of preparation of ammonium uranates

6/7 1956–1959 U, progeny U compounds (see at 

right)

Study of U compounds: U beryllides, borides, silicides, sulfides

6/7 1956–1962 U, progeny UO2 (EU, 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

6/7 1956–1962 U, progeny U carbides, mainly UC 

(NU; EU 9%–93%)

R&D, studies of U carbides and their alloys

6/7 1957–1961 U, progeny U nitrides, mainly UN 

(DU; NU; EU 5%–93%)

R&D, studies of U nitrides and their alloys

6/7 1957–1958;

1961–1961

U, progeny UCl (EU, NU); U3O8

(NU?)

Study of UCl dispersions (EU, short period only, 1957–1958); making UCl from 

U3O8 (1961)

6/7 1961–1961 U, progeny U3O8 Preparation of single U3O8 crystals, examination

6/7 1955–1955 Th, progeny Unknown Support for Th extraction activities?
ATTACHMENT C

HISTORY OF OPERATIONS BY BUILDING AND MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 103 of 145

Buildinga Period Isotopesb Formc Type(s) of workd

6/7 1955–1965;

1971–1972

Th, progeny; 

minor U

Th metal, ThO2 alloys; 

iodide, phosphate, 

halide, nitride, fluoride, 

carbide

Th compounds, alloys (except extraction): Th-5, -10, -15, -20 wt % U alloys; ThU 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)

6/7 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 wt % U-5 wt % Pu; corrosion of Ta, Ta alloys by liquid Pu 

alloys (1960?). Study of UN-PuN fuel mat, cladding for fast reactors

6/7 1956–1956 U content Fissium Fissium (synthetic fissium alloy): specimen fab, annealed; X-ray diffraction

6/7 1954–1954;

1958–1962

Fe-59, Mn-54 Metal Unknown

6/7 1956–1961 U HEUO2, DUO2; other? Miscellaneous neutron dosimetry/monitor work, including wire-activation method 

for reactor flux profile data; fast neutron effects on semiconductors; 

6/7 — U Al2O3-Mo-Si2-UO2, EU, 

DU

Neutron activation of ion chamber walls; ceramic tubes of Al2O3-Mo-Si2-UO2

with EU for fission-power elements, DU for heating elements (1959-60)

6/7 1958–1960 Not applicable Not applicable Radiation chemistry of inclusion compounds by nitration of hydrocarbons and 

their urea (nitration) compounds

6/7 1958–1960 P-32, Ba-140, 

La-140, Y-91

Tracers in cement Develop activation analysis, other methods to measure Mg, Ca, Al, and Fe in 

cement; radiometry of samples, P-32 labeling, EDTA titration with tagged Ag*IO

6/7 1959–1960 Fission gases?e Fission gases? Isotopic exchange leak detection system (an Ag-Br column in conjunction with 

delayed-neutron monitor)

6/7 1960–1962 H-3 Tritiated cytidine, trace 

levels

Miscellaneous biological effects studies, including nucleic acid metabolism in 

HeLa S3 cells after X-ray-induced mitotic delay (using tritiated cytidine)

6/7 1962–1963 MFP; UO2 UO2 particles Internal radiation effects on catalyst activity: 5 wt % ceramic-coated UO2

particles added to catalyst; catalyst efficiency tested; neutron irradiation

6/7 1967–1974 Not applicable Not applicable Research on bioeffects of underground explosions at Amchitka Island, Alaska 

during/after Milrow, Cannikin tests (samples counted at Battelle?)

6/7 1982–1982? Not applicable Not applicable Radiation surveillance for Piqua Nuclear Power Facility: annual radiation survey, 

including soil sampling, by Battelle (samples counted at Battelle?)

9 1958–1962? U, Th DU 0.06 MT (Author unknown 1992)

9 1958–1962? Kryptone Krypton, UO2 Analysis of gas from SME, OMRE, and OMR fuel specimens; limited laboratorylevel work with DU and NU

a. Buildings 4, 6, and 7 might have been used for some work even when not listed because it is not always clear what associated laboratory-type tests might have been 

done. In addition, 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. AP = activation products in this table only (for the materials in question).

c. The U form was probably NU when not specified; EU appears not to have been used before 1947. 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.
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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. Therefore, 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 FP 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, Xe-137, Xe-140, and 

Xe-141 FP chains were studied, with the respective down-the-chain progeny 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 progeny and 

radioiodine were then chemically separated and counted.

Table C-2. History of operations at the West Jefferson site. 

Building Period Isotopesa,b Formc Type(s) of workd

JN-1 (Hot 

cells)

1954–1983 U, Th, AP, FP, 

Pu

DU, NU, EU, 

Th, AP, FP, 

Pu

8.9 MT (Author unknown 1992)

JN-1 1954–1962 U, MFP, AP U metal, alloys 

(NU, 10% EU)

Postirradiation studies/examinations of clad NU element; fuel elements, specimens 

irradiated elsewhere; U alloys (NU or 10% EU); U-Nb alloys

JN-1 1954–1972 U, MFP, AP UO2 (NU) 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). FP diffusion and recoil studies, including sectioning; 

radiation damage studies in ceramics, Al by contact with UO2; iodine-retaining coatings

JN-1 1962–1972 U, MFP, AP UO2 (NU) Study of Cr foils impregnated with FP concentration of 2 x 109 recoils/cm2 to measure Xe133 release and Ce-141, Ru-103, and Nb-95 distribution

JN-1 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

JN-1 1956–1962 U, MFP, AP UO2 (EU 

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% to 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

JN-1 1957–

1961;

1965–1971

U, MFP U nitrides, 

mainly UN 

(DU; NU; EU 

5%–93%)

Postirradiation testing of UN dispersion, cermet forms; fission bubble growth kinetics, 

migration; irradiation behavior of UN fuel elements
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Building Period Isotopesa,b Formc Type(s) of workd

JN-1 1958–1962 U, MFPe U carbides, 

mainly UC 

(NU; EU 10%–

93%)

Postirradiation examination of UC (some NU, some EU) capsules; dispersed UC, UC2

(NU); 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 NU UC2 particles; FP deposition in out-of-pile loop system (UC2); fission 

bubble growth kinetics, migration; irradiation behavior of UC elements

JN-1 1962–

1962;

1965–1966

U, MFPe U silicides, 

hydrides

Postirradiation studies of U2Si2; deposition behavior of FPs in an out-of-pile gas loop 

system, with U hydride as fuel

JN-1 1957–1964 U, Th; MFP,f

AP

Spent fuel in 

solid, liquid 

form (mostly 

nonradioactive)

Recovery (dissolution) of spent fuel: Zircex, Darex, Thorex, Zirflex, fluoride volatility, Niflex 

processes; mostly non-radioactive Zirflex: U fuel elements; Thorex: Th forms. Examine 

ORNL Fluoride Volatility Pilot Plant vessels, Hanford Purex evaporator steam coil 

sections. 1961 on: mostly fluoride volatility process studies

JN-1 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-radioactive

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

JN-1 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 (1964–1965); Fermi Reactor failed fuel (1967?); PM-3A Type 

I, Serial 2 (defective) core and Serial 3 core (1967-68); Vermont Yankee 304 SS Schedule

80 reducer (1974); Arkansas Nuclear Plant 1 reactor building spray system SS 304 

recirculation 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 Three Mile Island cleanup: characterization of highly 

loaded EPICOR II prefilter liners #3 and #16 by sampling, analyzing liner contents, 

examining liner (1981–1983)

JN-1 1979–

1983?

MFP Spent fuel Shipping cask sabotage source term investigation; used five irradiated fuel rods from the 

H. B. Robinson reactor (1979)

JN-1 1964–1968 Pm-147 Pm oxide RTG work: hot cell examination of SD-4 thermionic diode (1964–1968)

JN-1 1964–1977 Pu-238, U, 

Th, MFP; Nd150, N-15 as 

tracers

PuO2, UO2, 

UO2-PuO2, 

UN-PuN

Radiation 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)
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Building Period Isotopesa,b Formc Type(s) of workd

JN-1 1974?–

1975

Np, Pu Pu fuel? After production of Pu-238 in commercial power reactor: postirradiation examination, Np 

recovery (1974?–1975)

JN-1 1972–

1980s?

Cf-252 Cf-252 wire Processing and encapsulating Cf-252 (produced elsewhere)

JN-1 1956–

1961;

1966–1967

HEUO2, DU 

UO2; other?

Miscellaneous neutron dosimetry and monitoring work, including fast neutron effects on 

semiconductors; neutron activation of ion chamber walls; proportional flow counter for 

neutron activation studies of MGCR specimen fission gas release (including in-operation

degassing) (1959). Ceramic tubes of Al2O3-Mo-Si2-UO2 with E U, DU (1959–1960)

JN-1 1958–1960 Not 

applicable

Not applicable Develop activation analysis, other methods to measure Ca, Al, Fe in cement

JN-1 1961–1962 Not 

applicable

Not applicable Evaluation of stresses, other test parameters on sealed radiation sources

JN-1 1962–1963 MFP; U UO2 Internal radiation effects on catalyst activity: 5 without ceramic-coated UO2 particles added 

to catalyst; catalyst efficiency measured after neutron irradiation. Postirradiation betagamma activity followed as function of decay time. Catalysts with up to 65 mCi/g of FP 

beta activity used in dehydration of cyclohexanol

JN-2 (CAL, Pu 

Laboratory)

1955–1978 U, Pu NU, EU, Pu 1.7 MT (Author unknown 1992)

JN-2 1955–1961 U U metal foil 

(NU, DU); UO2

(2%–93% EU)

APPR criticality experiment (1955, EU foil). Evaluation reflected heterogeneous BWR, 

including critical assembly experiments (U foil) (1957–1958). Critical experiment with 

movable water reflector; core: Al-polyethylene-U core (1957-?). Critical assembly 

experiment for GCRE-1 (UO2) (1958). Plastic-moderated critical assembly: polyethylene, 

polystyrene (1958–1961). Mockup ML-1-1B reactor, GCRE critical 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). Critical assembly experiments with UO2 pellets, pins (1.8% 

enrichment); DU foil (0.04 wt % U-235), NU foil to calibrate Battelle counting facilities in 

activity per fission in U-235, U-238 (1959–1960)

JN-2 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)

JN-2 1964–

1968;

1974–1978

Pu-238, Pu, 

U, Th, MFP; 

Nd-150, 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
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Building Period Isotopesa,b Formc Type(s) of workd

JN-2 1966–1978 See Types of 

work at right

See Types of 

work at right

Incaloy cladding (1966–1968?). Fabricate, examine UN-PuN pellets, EBR II pellet-type 

fuel, UN-PuN pin-type elements, irradiation capsules (1968–1971). 1970: Study of PuNUN burnup using Nd-150 tracer, N diffusion in PuN, UN-PuN with N-15 as tracer (1970). 

Develop Na-bonded SS-clad PuN capsule (1972?). Design, fab RTGs with Pu-238 (PuO2) 

(1974–1978)

JN-2 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?)

JN-2 1974–1978 Cm-244 Cm oxide? Design, fab RTGs with Cm-244 (1974–1978)

JN-3(BRR) 1955–1975 U NU, EU 0.1 MT (Author unknown 1992)

JN-3 1955–1965 U, MFP, APe UO2 (NU) 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

JN-3 1956–1962 U, MFP, APe UO2 (EU 

25%–93%)

BRR, ETR loop programs (25%–50% enrichment); evaluate spherical fueled elements 

(93% enrichment); fission gas release from high-burnup irradiation (some at 30% 

enrichment, some at 93%); irradiation of UO2-SS capsules, other fuels (at 93% 

enrichment), including gas loop measurements; irradiation coated particles

JN-3 1957–1957 U, MFP, AP U metal (NU), 

alloys (NU, 

10% EU)

Irradiation of Zr-clad NU element; irradiation study of U alloys (NU or 10% enrichment)

JN-3 1958–1962 U, MFP, APe U carbides, 

mainly UC 

(NU; EU 10%–

93%)

Irradiation of UC (NU, EU) capsules; PyC-coated UC2 particles (NU); UC, UC2 (NU) 

dispersion specimens, other UC, UC2 specimens (at 30% enrichment, some at 93%); 

high-burnup specimens (at 10% enrichment); coated spheres in sweep capsules, including 

in-pile fission gas analysis. FP deposition in out-of-pile loop system (UC2); fission bubble 

growth kinetics, migration; UC fuel irradiation behavior

JN-3 1958–1972 AP See at right Irradiation effects on seals, gaskets, etc.; non-radioactive core, structural mats, especially 

cladding, sintered Al particles. Irradiation of SNAP-8 mats, electrical components

JN-3 1965–1971 U, MFP U nitrides, 

mainly UN 

(DU; NU; EU 

5%–93%)

Postirradiation testing of UN dispersion, cermet forms; fission bubble growth kinetics, 

migration; irradiation behavior of UN fuel elements

JN-3 1956–1961 U, Th, MFP U-Th alloys U-Th alloy studies: irradiation to measure fission gas retention, other postirradiation data

JN-3 1964–1971 Pu, U, MFP; 

Nd-150, N-15 

(tracers)

PuO2, PuN, 

UN-PuN; Nd150, N-15 (as 

tracers)

Radiation 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 N-15 tracer (1970). Irradiation, experiments re fuel 

material aerosol properties (UO2, PuO2) (1975?–1977)

JN-3 1964–1968 Pm-147 Pm oxide RTG work: irradiation, in-pile testing of SD-4 thermionic diode (1964–1968)
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Building Period Isotopesa,b Formc Type(s) of workd

JN-3 1956–

1961;

1966–1967

U, MFP, AP UO2 (EU, DU), 

other forms?

Miscellaneous neutron dosimetry/monitor work, including the following. Wire-activation for 

reactor flux profile data; fast neutron effects on semiconductors; thermal neutron flux 

meter, sensor; “tactical” neutron dosimetry; fast neutron dosimetry system; neutron

activation of ion chamber walls. Proportional flow counter for neutron activation studies of 

MGCR specimen fission gas release (1959). Thermal neutron flux monitoring system for 

Hanford; ceramic tubes of Al2O3-Mo-Si2-UO2 with EU, DU (1959–1960). Study of 1 to 

50,000 rad semiconductor fast neutron dosimeter (1966–1967)

JN-3 1956–1961 Not 

applicable

Not applicable Miscellaneous gamma dosimetry/monitor work, including the following. Gamma effects on 

semiconductors, tin oxide films; effects of high-intensity radiation bursts on electrometertype ion chambers; gamma testing of photovoltaic gamma dosimeter; high-level gamma 

dosimetry research, including glasses for high-level gamma dosimeters

JN-3 1958–1960 Ba-140, La140, Y-91Tracers in 

cement

Develop activation analysis to measure Mg, Ca, Al, Fe in cement

JN-3 1958–1961 Fission 

gasese

Fission gases 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 fission products; develop methods to study fission gases 

with gamma spec analysis, charcoal traps (1961)

JN-3 1959–1960 Fission 

gases?e

Fission gases Develop fuel element leak detector

JN-3 1960 AP? AP? Facility to study fast neutron activation using the BRR spectrum

JN-3 1962–1963 U, progeny UO2 Internal radiation effects on catalyst activity: 5 wt % ceramic-coated UO2 particles added 

to catalyst, irradiation

JN-3 1964 N-17/O-17 Not applicable Develop N-17 (O-17) decay neutron monitor to track reactor power

JN-3 1971–1975 AP? Likely HEUO2 Neutron radiography of reactor fuel: done in reactor pool and in out-of-pool facility

JN-4 (part of 

this was called 

the “Old” Pu 

Laboratory) 

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 wt % U-5 wt % 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 wt % Pu; ThPu; heat, hot roll, metallographic examination (1960–1962). Evaluate PuN fuel; Pu-Al, PuFe, U-Pu-fissium, U-Pu-Mo, PuO2-UO2; solid-solution alloys (Th-Pu, U-Pu-Nb); liquidmetal 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)
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Building Period Isotopesa,b Formc Type(s) of workd

JN-4 (part of 

this was called 

the “Old” Pu 

Laboratory)

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)

JN-4 (part of 

this was called 

the “Old” Pu 

Laboratory)

1966–

1967?

Np NpO Prepare Np microspheres as thermocouple flux probes (Np oxide)

JN-4 (part of 

this was called 

the “Old” Pu 

Laboratory)

1978–1980 Pu, Th, U, 

progeny

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)

INL 1958–1960 U, MFP, 

fission gases

UC (NU, EU) Irradiate UC (NU, EU) FP 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 

(INL)

Hanford 1961–1962 Not 

applicable

Not applicable Research at Hanford (HAPO) on reactor materials in Non-Power Reactor; on PRTR 

operation

Power plants, 

ANL, INL

1970–1981 Not 

applicable

Not applicable 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. AP = activation products in this table only (for the materials in question).

b. The U form was probably NU when not specified; EU appears not to have been used before 1947. 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.

c. The U form was probably NU when not specified; EU appears not to have been used before 1947. 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. Therefore, 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 FP 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, Xe-137, Xe-140, and 
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Xe-141 FP chains were studied, with the respective down-the-chain progeny 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 progeny 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).
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ATTACHMENT D 

TYPES AND QUANTITIES OF MATERIAL

LIST OF TABLES

TABLE TITLE PAGE

D-1 Source and production materials.......................................................................................... 112

D-2 Sealed source materials....................................................................................................... 120

D-3 Waste and remainder materials............................................................................................ 121
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TYPES AND QUANTITIES OF MATERIAL (continued)

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Table D-1. Source and production materials. 

Material Process or operation Content and form notesa Period and amountb Reference

U metal, NU 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

None 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; 

DOE 2000c, Ferry 1943, 

DOE 2000a, Dayton and 

Tipton 1959a, Rengstorff and 

Lownie 1955

U metal, DU Used as foil in critical 

assembly work; machined, 

rolled 

None 1957–1960, possibly other 

years also

DOE 2000c, DOE 2000a

U metal, EU R&D on fuel elements; 

processing and machining

10%–93% enrichment 1947–1962; 1979. One 

critical experiment: critical 

mass of HEU foil was 4.935 

kg (Dayton and Tipton 1957).

Author unknown ca. 1987, 

DOE 2000c, DOE 2000a, 

Dayton and Tipton 1957; 

Saller 1947

U as recycle uranium Apparently used for testing of 

recovery methods, assay

DU, NU, or EU 1962–1980, sent by Battelle 

to Feed Materials Production 

Center or Rocy Flats Plant: 

168 kg DU, 5,135 kg NU, and 

29 kg EU (including 1.45 kg 

U-235 for enriched). 1962–

1982, received by Battelle: 

1,616 kg DU, 2846 kg normal 

U, and 0.06 kg EU (including 

1 g U-235 in the EU). More 

than half of the total kg U 

transfers were in 1964

DOE 2000d

UO3 Small amount in metal 

refining process support; 

some made from UNH (1957)

NU 1954–1959 Vaughan, Bridge, and 

Schwartz 1957
ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL (continued)

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Material Process or operation Content and form notesa Period and amountb Reference

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

DU, NU, EU; much as 

HEUO2; apparently all NU 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, 

DOE 2000c, Chastain 1961

U carbides R&D, study of U carbides 

and alloys, mainly for fuel 

apps

NU and EU, 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

DU, NU, and EU, 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)

None

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)

None

Other U compounds Studies of miscellaneous U 

compounds; postirradiation 

studies of some; study, 

production of UCl

NU and EU, 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 “TuBe alloy” at 90% Be; 13 lb at 

30% Be; 72 lb Tu-Be dross at 

2% Be; 59 lb TuO2 +BeO at 

33% Be (Grenell 1947). 

1956–1959 (miscellaneous

compounds). 1962, 1965–

1966 (postirradiation 

studies). 1957–1958 (UCl 

study); 1961 (UCl production)

BMI 1945b, Grenell 1947
ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL (continued)

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Material Process or operation Content and form notesa Period and amountb Reference

Thorium metal and 

alloys

R&D with Th metal; forging, 

rolling, machining ingots into 

slugs for Hanford; arcwelding Ames Th, Th alloys; 

mechanical/metallurgical

study of Th; making Th-U 

alloy with Th iodide; casting 

Th buttons; casting Th-U 

carbides by arc-melting

None 1944–1961 overall. 1944–

1946 (R&D). 1947 (slugs for 

Hanford)

BMI 1945a, 1945b; 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)

DOE ca. 1987, DOE 2000c; 

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 ThU, 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 (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 115 of 145

Material Process or operation Content and form notesa Period and amountb Reference

Plutonium and other 

TRU elements 

Study of Pu in U spent fuel 

recovery; Ta corrosion by Pu 

alloys; studies of UN-PuN 

fuel, PuO2 pellets, Th-Pu and 

U-Pu alloys, liquid metal 

fuels, ceramic/cermet fuels; 

R&D on Pu processing, 

fabrication; fabrication of Pu238 heat sources; breeder 

experiments; Pu production 

from Np targets fabricated at 

Battelle (irradiated 

elsewhere), Np/Pu recovery; 

Pu microspheres; aerosol 

experiments

Some TRU elements as a 

result of breeder 

experiments. Usually Pu-238 

(e.g., for RTGs: powder 

transferred, coated, pressed, 

sintered to pellets, welded 

into capsules), but Pu239/240 also handled; Np237 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 INL in 1969-70 (form 

unknown).

DOE 2000b, Toy 1986, DOE 

2000a, Freas and Madia 

1982, BCL (ca. 1983)

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. 1948
ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 116 of 145

Material Process or operation Content and form notesa Period and amountb Reference

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 to 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; separation of 

minus 325 mesh fraction; 

mixing with 10% salt; 

roasting. No details: 

Colorado, Western ores. 

Other ores: ammonium 

carbon 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, DOE 

2000c, Blatz 1949

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 weight; to 100 mg 

Ra/ton (Dupree-Ellis et al. 

2000); average 135 mg 

Ra/ton (AEC 1949); 0.185 

ppm equilibrium Ra for Q-11 

(60%) ores (AEC 1949); 0.1 

Ci/ton, to 70% U, average U 

concentration >25%, ~100 

mg Ra/t 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
ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL (continued)

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Material Process or operation Content and form notesa Period and amountb Reference

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 1951b).

1950 (washer slimes). 1950–

1951 (MgX). 1953 (sulfide 

concentrate)

Meaders et al. 1950; Ewing, 

Kiehl, and Bearse 1950; 

Ewing et al. 1950a, 1952; 

AEC 1951b; 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 in-pile loop)

Irradiated/spent fuel 

specimens (usually 

enriched), MFPs; when 

cladding/additives present, 

could also have substantial 

level of activation products. 

Fission progeny on one loop 

filter: Ba-140, Ce-141, Cs137, Sr-89 (I already 

decayed)

1955 to about 1983 DOE 2000a
ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 118 of 145

Material Process or operation Content and form notesa Period and amountb Reference

Fuel elements and 

dissolution products 

(Th, U, activation 

products, FPs) 

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 to about 1983 None

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

FP-containing 

material

Used 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
ATTACHMENT D

TYPES AND QUANTITIES OF MATERIAL (continued)

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Material Process or operation Content and form notesa Period and amountb Reference

Np Production of Pu from Np 

targets fabricated at Battelle 

(irradiation 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)

DOE 2000b

Cm RTG work Cm-244 (Cm oxides?) 1974–1978 None

Cf-252 material Cf wire ported into alphagamma cell, cut, weighed, 

put into containers. These 

were sealed by welding, leaktested, washed. In second 

cell, welded into larger 

container, leak-tested, 

washed

Cf-252 Started in the mid-1970s Peters 2001, Scotti and 

Martin 1972

Miscellaneous 

radioisotopes

Studies using radiotracers or 

labeled compounds; 

hydrocarbon radiation

chemistry studies; study 

using Cr foil impregnated 

with FPs to study Xe-133 

release, other FP 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 

FPs (to see Ce-141, Ru-103, 

and Nb-95 distribution). Po210 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 

(radiation 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 development). 

1970 (burnup study). 1976 

(at least) (I-125 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 FP mix, but might have 

had actual U content

1956 Dayton and Tipton 1956

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 so 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 (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 120 of 145

Table D-2. Sealed source materials. 

Material Process or operation Content and form notesa Period and amountb Reference

Gamma sources Used in nondestructive testing 

laboratory for radiography; used in 

Co-60 facility for irradiation effects 

research, destruction of chemical 

agents. Some sources stresstested for standards development

Radiography: 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 

(radiography). 1961–1962 (stress 

evaluation). 1981–1982 (chemical 

agent destruction). Large Co-60, 

Ra-226 sources: instrument 

calibration. ~1982: Co-60 pellets 

put in spent fuel cask to make 10KCi source for calibrating high-level 

detectors for Three Mile Island

BMI 1977, Rhoten 1958, ORAUT

2006b 

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); Ce-Pr 

source for alpha cell generator 

voltage checks

1960 (reactor coolant channel voids, 

U penetration into cladding); 1964 

(Ce-Pr source)

Dayton and Tipton 1960b, 1961; 

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

None Unknown Ferry 1943

Residue, scrap 

from carnotite ore 

processing

Not applicable 70–200 mg U/L of liquid residue; 

0.34% U in cake residue; some 

scrap at 0.2% U

Unknown Blatz 1949

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 (continued)

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Table D-3. Waste and remainder materials. 

Material Process or operation

Content and form 

notesa Period and amountb Reference

Leftover 

samples, 

specimens

Not applicable Some were highly 

irradiated and activated

Unknown None

Activated 

parts, mats

Not applicable Activation products Unknown 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).
 

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 122 of 145

ATTACHMENT E 

REPORTED INCIDENTS

Table E-1. Incidents reported at the King Avenue and West Jefferson sites. 

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 1997a

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 the worker opened the lower plug to create the beam. The worker tried to recover the source 

and replace in the camera using a threaded rod, but was unable to do so after several attempts. The 

worker 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. The coworker helped replace the source 

in the camera. The source was later transferred from the defective camera to another camera by the 

[Title Redacted]. The first worker’s film badge registered 9,600 mR; the second worker forgot to put 

on the film badge before attending to the source, so second workers dose was estimated to be 3,200 

mR; the [Title Redacted] also forgot to put on a badge, but no estimate of the dose is available. The 

workers were allowed to continue to work in the laboratory, but were restricted to non-radioactive

work. 

Redacted 1956 Selander 1956, 

Peterseim 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 March 1958 None

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 June 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 ~12/01/195

8 

Dayton and Tipton 

1959b
ATTACHMENT E

REPORTED INCIDENTS (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 123 of 145

Description of incident Location Date Reference

A worker punctured his finger with contaminated [Item Redacted] 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 [Location Redacted, 

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. (FPs)

Redacted 1959 Selander 1959c

A worker punctured his finger with a piece of metal containing radioactive dust, while cutting up the 

pan from beneath the [Redacted] in the high-level cell at the [Redacted]. 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. (FPs)

Redacted 1959 Selander 1959a

A worker technician was overexposed (greater than 3 rem in a quarter) while removing a [Redacted]

from the [Redacted]. It was shipped from [Redacted], via the Battelle mail car to the [Redacted]. 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 [Redacted]. The [Redacted] was being 

irradiated in the [Redacted] at the end of a long wire. The technician was raising the wire when the 

CAM alarmed. He and a researcher left the area. FP activity was calculated to be 1.5 Ci at 10 s 

after exposure of the [Redacted], 0.23 Ci 2 d later. The [Redacted] was placed in a cask. See Table 

5-6 for dose rates related to this. The technician’s dose to the [Redacted] was 7.75 rem (beta plus 

gamma); to the [Redacted], 13.55 rem; to the [Redacted], “within tolerance limits,” and the wholebody dose, 13.2 rem gamma. (Probably the same incident as the entry immediately above.)

Redacted 1961 Chastain 1961

A worker got a [Redacted] overexposure in cleaning highly contaminated equipment in a hot cell. He 

did not wear leaded [Redacted] during many parts of the operation because they were too 

cumbersome and he thought he would not go over the limit.

Redacted 1962 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 [Redacted] were contaminated, indicating 

that the contamination was on the outside of the lead containers because these had been handled 

with [Redacted] and all the work after the small containers were opened were done with the lead 

[Redacted] on. The second worker checked himself and found that he too was contaminated; some 

contamination appeared to be inside his [Redacted].

Redacted 1963 Saling 1963
ATTACHMENT E

REPORTED INCIDENTS (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 124 of 145

Description of incident Location Date Reference

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 First quarter 

1964

Saling 1964

A worker’s film badge read 1,800 mrem. This was thought to have resulted from maintenance work 

in the [Location and Date Redacted. 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 [Redacted] was 10-20 times higher and he might have passed close to the [Redacted]. 

The worker also handled a [Redacted] 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 [Date Redacted] it 

was discovered to be malfunctioning, so it did not register the exposure.

Redacted Redacted None

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 09/09/1968 Ottman 1968

A worker exceeded 3 rem for the second quarter of 1973 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 Second 

quarter 

1973

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. Wash water

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 wash water. 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 High Level Cell, producing a layer of sludge.

JN-1 Before 

1974

Myers and Berchtold 

1994

A worker was exposed to Pu-239 via [Redacted]. 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 [Date Redacted] 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 [Dates Redacted] for this worker, who was involved 

in another unspecified potential intake incident during the week of [Date Redacted]. 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 Pu-239 results, but the urine 

and feces indicated an even lower intake for them, likely less than 1%–2% of the MPBB.

Redacted Redacted 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 U-238 while engaged in operations associated with 

[Redacted] in the [Redacted] area.

Redacted 1974 Kirsch 1974
ATTACHMENT E

REPORTED INCIDENTS (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 125 of 145

Description of incident Location Date Reference

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 [Redacted]. 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.”

Redacted 1975 Toy 1975b

Two [Redacted] were tearing out [Redacted] 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.

Redacted 1977 BCL 1977b; 

Langendorfer 1977; 

Wissinger and 

Swindall 1977 (air 

sample report), 

Wissinger 1977 

(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 [Redacted], it was found that during the removal of [Redacted] from 

the [Redacted], the lapel sampler filter of a technician exceeded the 40-hr MPC for soluble Pu-239, 

2.0E-12 μCi/mL. The concentration was calculated at 3.51E-12 μCi/mL; after a 64-hr decay it 

showed 2.84E-12. The technician was removed from [Redacted] 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.

Redacted Redacted Langendorfer 1977, 

Farmelo and 

Helwagen 1977, BCL 

1977a, BCL 1977b, 

Wissinger and 

Swindall 1977

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 04/07/1977 Peters and Harrison 

1998

A worker was repairing a contaminated [Redacted], wearing what was later found to be a 

contaminated [Redacted]. The alpha air concentration inferred from his lapel air sampler was found 

to be 10% 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.

Redacted Redacted Kirsch 1978a 

(includes most of the 

bioassay data)
ATTACHMENT E

REPORTED INCIDENTS (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 126 of 145

Description of incident Location Date Reference

A worker entered the [Redacted] laboratory with 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 

[Redacted] 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. 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.

Redacted Redacted Kirsch 1978b

A routine count of a [Redacted] for a worker after a control area entry prompted follow-up bioassay 

sampling. Counts for both a follow-up [Redacted] and a smear of the worker’s respirator interior 

were negative. [Redacted] 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 FPs, while the [Redacted] also showed normal concentrations of uranium. It was concluded that 

an inhalation intake had not occurred and that the original [Redacted] probably was contaminated by 

contact with a contaminated surface in the [Redacted].

Redacted Redacted 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 to 

t10 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 (BCL 1979; timeline 

and air sampling results).

JN-1 01/16/1979 BCL 1979 (timeline 

and air sampling 

results), Peters and 

Harrison 1998, Ray 

1990

A DU 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 09/22/1979 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 Part 20 limits and that pulmonary 

depositions were <1% of the permissible body burden. [Probably same incident as entry below]

JN-1 05/02/1980 

or 

05/03/1980

Jensen 2003, DOE 

2003b, Ray 1990, 

Kirsch 1987
ATTACHMENT E

REPORTED INCIDENTS (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 127 of 145

Description of incident Location Date Reference

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 Part 20 limits. The 

cleanup and disposal of radioactive waste were completed.

JN-1 05/02/1980 Peters and Harrison 

1998, Pasupathi and 

Toy 1990, Kirsch 

2000b

A worker received a [Redacted] of 22.48 rem in the second quarter of 1980. Of this, 20.560 rem 

were received during an operation that involved the collection of [Redacted]. The personnel 

monitoring device was a finger ring TLD. The operation involved two [Redacted] tasks. 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.]

Redacted Redacted Toy 1981

A worker was [Redacted] 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. (FPs)

Redacted 1981 Swindall 1981

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 07/15/1982 BCL 1966–1993

A worker was [Redacted] in an unspecified fashion at [Redacted]. The material was DU. Special 

urine monitoring was done every 2 weeks (see undated entry below) in [Redacted] due to what was 

said to be a potential inhalation in [Redacted] and that could be related to this [Redacted] incident. 

(Note that [Redacted] 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 [Redacted]

adjoined [Redacted], it is likely that the incident actually occurred in [Redacted].)

Redacted Redacted Swindall 1984 

(includes wound 

survey results); 

dosimetry records

A worker was subject to special [Redacted] monitoring in [Redacted] 1984 due to a potential 

inhalation of [Redacted] in [Redacted]. At this time, it was recommended that he discontinue 

radiation work. When he was found still to be excreting low levels of [Redacted] in [Redacted] 1984, 

it was recommended that [Redacted] be used if he were needed to perform work rolling [Redacted]

or the like.

Redacted 1984 McKown 1984a, 

1984b; 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
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ATTACHMENT F 

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

F.1 Introduction .......................................................................................................................... 129

F.1.1 Building JN-1, Hot Cell Building ................................................................................ 129

F.1.2 Building JN-2, Critical Assembly Laboratory ............................................................. 129

F.1.3 Building JN-3, Battelle Research Reactor................................................................. 130

F.1.4 Building JN-4, Plutonium Laboratory......................................................................... 130

F.1.5 Building JS-1 ............................................................................................................ 130

F.2 Available Data collections from Battelle................................................................................ 130

F.3 Methods ............................................................................................................................... 132

F.4 Results................................................................................................................................. 132

F.4.1 Building JN-1, Hot Cell Laboratory............................................................................ 132

F.4.2 Building JN-2, Critical Assembly Laboratory ............................................................. 133

F.4.3 Building JN-3, Battelle Research Reactor................................................................. 136

F.4.4 Building JN-4, Plutonium Laboratory......................................................................... 137

LIST OF TABLES

TABLE TITLE PAGE

F-1 Buildings JN-2, JN-3, and JN-4 neutron and photon radiation survey data........................... 131

F-2 Buildings JN-2 and JN-4 neutron and photon radiation survey data ..................................... 131

F-3 Building JN-2 hazards summary reports............................................................................... 132

F-4 Mean neutron quality factor and flux corresponding to maximum dose equivalent of 

1 mSv per 40-hour week ...................................................................................................... 133

F-5 Recommended lognormal parameters of NP ratios for Building JN-1 ................................... 134

F-6 Recommended lognormal parameters of NP ratios for Building JN-2 ................................... 136

F-7 Statistical parameters of JN-3 survey record NP ratio analysis............................................. 137

F-8 Lognormal parameters of NP ratios for Building JN-4 ........................................................... 138

LIST OF FIGURES

FIGURES TITLE PAGE

F-1 JN-1 hot cell laboratory NP ratio during 252Cf operations ...................................................... 134

F-2 JN-2 Plutonium Laboratory NP ratio..................................................................................... 135

F-3 JN-3 paired neutron and photon dose rate measurements................................................... 136

F-4 JN-3 lognormal probability plot of the measured NP ratio..................................................... 137

F-5 JN-4 historical trend in NP ratio............................................................................................ 138

F-6 JN-4 Plutonium Laboratory NP ratio..................................................................................... 139

F-7 JN-4 paired neutron and photon dose rate measurements................................................... 139
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

F.1 INTRODUCTION

This analysis examines radiation survey data from the West Jefferson site, which contains several 

major buildings with the potential for workplace radiation fields (Amstein 2010), in relation to neutron 

and photon dose.

F.1.1 Building JN-1, Hot Cell Building

This facility contains hot cells that were used to provide research and technical assistance in the 

areas of power reactor fuel performance evaluations, postirradiation examination of nuclear materials 

and components, radiation source encapsulation, and physical and mechanical property studies on 

irradiated materials and structures.

The original section of this facility, referred to as JN-1A since 1973, was built in 1955. JN-1A 

contained offices, laboratory 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 shut down and put under a surveillance and maintenance program 

before initiation of decontamination and decommissioning activities.

The primary purpose of the entire facility was for destructive and nondestructive 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 ad hoc 

research experiments using sealed neutron sources (e.g., usually 252

Cf) 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 the kinds of neutron sources for 

instrument calibration that were stored periodically for subsequent transfer off site.

F.1.2 Building JN-2, Critical Assembly Laboratory

JN-2 was designed and constructed for use as a CAL and was used for critical experiments. This 

building was constructed in 1955 and housed a vault used to store special nuclear materials, a 

radioanalytical 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 about 1956 through about 1960 and involved the use of very low

levels of uranium foils. After the cessation of critical assembly experiments, the building was used for 

several nuclear projects, including direct conversion concepts, irradiation experiment assemblies, and 

special nuclear materials storage and dispensing. The storage area contained a thick-walled vault on 

the ground floor.

A small plutonium laboratory constructed in 1966 (noted in survey records as JN-2 Pu Lab) 

was used for research conducted for what became known as the “Lawrence 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 
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

radiological samples associated with onsite facilities, including analyses of various health 

physics samples and environmental media (BCL ca. 2010).

F.1.3 Building JN-3, Battelle Research Reactor

BRR operations began in October 1956 and ended December 1974. Based on Battelle licensing 

submittals to the NRC (Dingee and Chastain 1961), several critical assembly systems have been 

studied including beryllium-moderated, a plastic-moderated, two gas-cooled water-moderated, and a 

uranium dioxide-fueled, organic-and-water moderated reactors. This facility was selected because 

the shielding research area was equipped with instruments to measure fast neutron, thermal neutron, 

and gamma radiation dose rates and spectra (Klingensmith, Epstein, and Chastain 1959). BRR 

defueling and partial dismantlement was completed in 1975.

F.1.4 Building JN-4, Plutonium Laboratory

Built in 1960 with additions in 1964 and 1967, the Plutonium Laboratory housed activities for 

plutonium research and processing. The facility was operated from 1960 to 1978 when it was shut 

down for D&D. The facility contained one small laboratory known as the 238Pu laboratory with a 

neutron exposure potential. It contained several gloveboxes 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.

F.1.5 Building JS-1 

This facility was used to develop alloy and fabrication processes, for corrosion chemistry studies, and 

to conduct engineering analyses for the U.S. Navy nuclear reactor program. Records indicate that 

uranium (depleted in 235U) was used in this facility (Cotten and Payne 1990). The building was 

decommissioned in 1990.

F.2 AVAILABLE DATA COLLECTIONS FROM BATTELLE

The Project has the following data:

• Buildings JN-2, JN-3, and JN-4. The Project obtained 25 documents with results of 

radiological surveys during the period from 1957 to 1977, primarily associated with operation 

of the BRR in Building JN-3. The pre-1970 neutron dose survey data were in units of neutron 

flux (i.e., n/cm2

/s). Often survey results were provided for thermal and fast neutrons. The 

instruments for the surveys were typically listed on the respective documents. Selander 

(1957b) identified the MPL (i.e., assumed to be 5 rem/yr) for fast neutrons as being 50 n/cm2

/s 

for 40 hours 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 NP ratio. 

Twenty of the files were identified as containing useful survey data to analyze the NP dose 

ratio and these were selected for data entry. The files along with their SRDB Reference 

Identification (Ref ID) numbers are listed in Table F-1.

• Buildings JN-3 and JN-4. The Project obtained 16 documents of supplemental information

with results of radiological surveys during the period from 1971 to 1974, primarily associated 

with operation of the JN-2 and JN-4 facilities but including information from JN-3. Fourteen of 

the files were identified as containing useful survey data to analyze the NP dose ratio and 

these were selected for data entry. As noted in Amstein (2010), the Battelle monthly health 
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

physics program reports for plutonium facilities contain a section for neutron flux surveys on. 

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 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 in JN-2. Note that the reference to the “238 lab” under the same section was the 

238Pu laboratory 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 F-2 with the associated SRDB reference.

• Building JN-2. Two hazard summary reports were found that described the relative neutron 

and photon hazards at the JN-2 facility (Table F-3).

Table F-1. Buildings JN-2, JN-3, and JN-4 

neutron and photon radiation survey data. 

SRDB Ref IDa File

76975 1960 JN-3 redacted

76976 1961 JN-3 redacted

76977 1962 JN-3 redacted

76978 1963 JN-3 redacted

76979 1964 JN-3 redacted

76980 1965 JN-3 redacted

76982 1966 JN-3 redacted

76983 1967 JN-2 redacted

76984 1967 JN-3 redacted

76985 1968 JN-3 redacted

76986 1968 JN-4 redacted

76987 1969 JN-3 redacted

76988 1969 JN-4 redacted

76989 1970 JN-3 redacted

76990 1970 JN-4 redacted

76991 1971 JN-3 redacted

76993 1973 JN-3 redacted

76994 1974 JN-4 redacted

76995 1975 JN-4 redacted

76996 1976 JN-4 redacted

Table F-2. Buildings JN-2 and JN-4 

neutron and photon radiation survey data. 

SRDB Ref ID File

79055 1971 JN-3 Reactor

79067 1972 (Apr) JN-4

79071 1972 (Aug) JN-4

79076 1974 (Jan) JN-4

79079 1972 (Jun) JN-4

79081 1974 (Mar) JN-4

79083 1972 (Oct) JN-4

79057 1974

79069 1974 (Apr) JN-4

79074 1974 (Feb) JN-4

79078 1972 (Jul) JN-4

79080 1972 (Mar) JN-4

79082 1972 (May) JN-4

79084 1972 (Sep) JN-4
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

Table F-3. Building JN-2 hazards summary reports.

Reference Title Document number Year

Jankowski et al. 1956a Hazards Summary Report for the PWAR-3 Critical 

Assembly Experiments

BMI-ACRS-611 1956

Jankowski et al. 1956b Hazards Summary Report for the APB ReflectorControl Critical-Assembly ExperimentsBMI-ACRS-615 1956

F.3 METHODS

The respective records of photon and neutron dose measurement data were categorized according to 

the general workplace or type of exposure radiation from information in the survey documentation. 

The data were analyzed to eliminate incomplete data in terms of 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 milliroentgen per hour. 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

-s) and 

beginning in 1970 as a measured dose rate in millirem per hour depending on the instrument. 

Thermal and fast neutron components were identified although most data 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 I-1 as follows:

 2 5 mrem/hr

20 680

fast neutron flux thermal neutron flux neutron dose .   =   +   (I-1)

This equation is based on NCRP Report 38 guidance on values to use to convert neutron flux to dose 

rate (NCRP 1971) for identified neutron energies as shown in Table F-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 to result in a higher NP ratio, which 

is favorable to claimants.

F.4 RESULTS

The available data for each building have been analyzed and presented in the following sections.

F.4.1 Building JN-1, Hot Cell Laboratory

The primary mission of the JN-1 Hot Cell facility was nondestructive 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 

transuranic materials that have a potential for neutron exposure. These cells were added to the 

basement of JN-1 in 1964 (Peters 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. From 1971 

through 1974, potential for neutron exposure was quite large as the quantities of 252Cf being worked 

with were in the milligram range. Other operations involving transuranic isotopes appear to have 

occurred periodically from startup in 1955 through 1988, but the potential for neutron exposure would 

be much lower than during the 252Cf operations.
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

Table F-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)

Quality 

factor

Neutron flux density

(n/cm2/s)

2.5 × 10-3 (thermal) 2 680

1 × 10-7 2 680

1 × 10-6 2 560

1 × 10-5 2 560

1 × 10-4 2 580

1 × 10-3 2 680

1 × 10-2 2.5 700

1 × 10-1 7.5 115

5 × 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

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 NP ratio can be described as a lognormal 

distribution with a geometric mean (GM) of 2.4 (2.42) and a geometric standard deviation (GSD) of 

3.9 (3.93) as shown in Figure F-1. The R2 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 a Pearson correlation coefficient of 0.54 and a Spearman rank correlation 

coefficient of 0.56.

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 with the appropriate adjustments (NIOSH 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 before assigning a neutron 

dose based on the NP ratio listed in Table F-5 and illustrated in Figure F-1. The data in Figure F-1 

are considered the upper bound of the NP ratio because the NP ratio for 252Cf is probably higher than 

other possible exposure scenarios. However, it is probably 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 the 95th-percentile ratio of 20:1, as listed in Table F-5, is certainly bounding, but 

the worker could not have worked with 252Cf until at least 1971. 

F.4.2 Building JN-2, Critical Assembly Laboratory

According to Battelle (BCL ca. 2010), the JN-2 CAL was built in 1955. From 1956 through the early 

1960s, the facility conducted various critical experiments using EU. Hazards summary reports were 
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

Figure F-1. JN-1 hot cell laboratory NP ratio during 252Cf operations. 

Table F-5. Recommended lognormal parameters of NP ratios for Building JN-1. 

Workplace location Years n GM GSD 95th%

Non-252Cf Operations 1955–1983a (b) (b) (b) (b)

252Cf Operations 1971–1974 49 2.4 3.9 22.5

a. Neutron exposures probably very limited before construction and operation of alpha/gamma 

cells in 1964.

b. No paired survey data received. 

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. To date, records of 

these early surveys have not been located.

For interim use, an NP ratio was developed based on the hazards 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-W range would result in a photon dose 

that was one-eighth 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 about one-quarter 

of the weekly tolerance dose of 300mrem/wk or 75 mrem/wk. This fast neutron dose corresponds to a 

dose rate of approximately 1.88 mrem/hr. Based on the available information (i.e., same time and 

location), the expected 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 JN-1 Hot Cell Facility and the JN-3 BRR. 

The dosimetry reports have been redacted for other workers at the facility, but at least 30 workers per 

2-week cycle were monitored for neutron exposure. Most of the fast neutron results were below the 
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

30-mrem detection limit, but occasionally the dose would be positive, usually less than 50 mrem. 

Assuming that some of the neutron badges were issued to workers in the JN-2 facility, the measured 

neutron dose rates are well below the conservative dose estimates from the hazards summary report. 

As a result the dose rates in the hazards summary report are considered reasonable and bounding.

According to Amstein (2010) and BCL (ca. 2010), in 1966 a small Plutonium Laboratory was 

constructed at the JN-2 facility. Radiological surveys for this laboratory have been obtained. The 

earliest survey obtained to date was conducted in September 1967. Additional surveys from 1972 

through 1974 have been added to the analysis for the development of an NP ratio for the facility 

during the Plutonium Laboratory operations starting around 1967. The NP ratio can be described as a 

lognormal distribution with a GM of 1.0 (0.97) and a GSD of 1.3 (1.28) (Figure F-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 F-2. JN-2 Plutonium Laboratory NP ratio. 

The NP ratio for the vault area of the JN-2 facility was significantly lower than that for the Plutonium 

Laboratory. The arithmetic mean of the NP ratio for the vault area was calculated from nine data 

points to be 0.29 or approximately one-third of the NP ratio for the Plutonium Laboratory. 

Because it is virtually impossible to place individual workers within the JN-2 facility (i.e. vault or 

Plutonium Laboratory), the approach most favorable to claimants 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 
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

from 1956 through 1966, an NP ratio of 2:1 should be used. For the later years when the JN-2 

Plutonium Laboratory was in, an 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 GM of 1.0 (0.97) and a GSD of 1.3 (1.28) 

as shown in Table F-6.

Table F-6. Recommended lognormal parameters of NP ratios for Building JN-2. 

Workplace location Years n GM GSD 95th%

Critical Experiments 1956–1960 (a) (a) (a) (a)

Plutonium Laboratory 1967–1975 14 1.0 1.3 1.5

a. No paired survey data received. 

F.4.3 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 from 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 evaporators were combined. After 

categorization, data associated with 252Cf was eliminated based on the understanding that the 252Cf 

source was present in the reactor for about 1 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 paired neutron and photon dose rate 

measurements is presented in Figure F-3. The data illustrate a higher relative measured photon dose 

in comparison with the measured neutron dose rate.

Figure F-3. JN-3 paired neutron and photon dose rate measurements. 
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

A lognormal probability plot of the data in Figure F-3 is presented in Figure F-4. The data are best 

represented by a lognormal distribution with a GM of 0.2, a GSD of 3.3 (3.34) and a 95th percentile of 

1.4 (1.383).

Figure F-4. JN-3 lognormal probability plot of the measured NP ratio. 

Each of the workplace categories was analyzed individually; the results in Table F-7 illustrate the 

variation in the NP ratio. The recommended option is to use the lognormal parameters for all of these 

work areas collectively, as shown in the last line of Table F-7, because exposed workers would be 

expected to move throughout the reactor facility.

Table F-7. Statistical parameters of JN-3 survey record NP ratio analysis. 

General location No. 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

F.4.4 Building JN-4, Plutonium Laboratory

The JN-4 facility (Plutonium Laboratory) began operations in the early 1960s. According to Battelle 

(BCL ca. 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 
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

50 boxes of records. It is understood from Battelle that as many as 700 boxes of similar 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 data by year in Figure F-5 

illustrate that the NP ratio remained relatively constant over the operational period.

Figure F-5. JN-4 historical trend in NP ratio. 

The 41 records of paired measurements were analyzed and are best described by a lognormal 

distribution with a GM of 0.7 (0.71) and a GSD of 5.0 (4.99) (Figure F-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 F-8.

Table F-8. Lognormal parameters of NP ratios for Building JN-4. 

Workplace location Years n GM GSD 95th%

Combined areas 1960–1978 41 0.7 5.0 9.9

The relatively large GSD is attributed to the different operations within the facility. Many of the paired 

NP ratio measurements greater than 1 were from the 238Pu 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 F-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.
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ATTACHMENT F

ANALYSIS OF MEASURED NEUTRON AND PHOTON DOSE 

RADIATION SURVEY DATA (continued)

Figure F-6. JN-4 Plutonium Laboratory NP ratio. 

Figure F-7. JN-4 paired neutron and photon dose rate measurements. 
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ATTACHMENT G 

ESTIMATING ONSITE AIR CONCENTRATIONS

LIST OF TABLES

TABLE TITLE PAGE

G-1 Intakes of alpha-emitting radionuclides from the West Jefferson site by year....................... 144

G-2 Intakes of mixed fission product radionuclides from the West Jefferson site by year............ 145

LIST OF FIGURES

FIGURES TITLE PAGE

G-1 Layout of the West Jefferson site, 1973 ............................................................................... 142

G-2 Layout of the King Avenue facilities that had stack emission points, 1973............................ 143
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ATTACHMENT G

ESTIMATING ONSITE AIR CONCENTRATIONS (continued)

An atmospheric dispersion factor is a measure of the average atmospheric dispersion for long-term 

(chronic) atmospheric releases using Gaussian dispersion plume modeling, with units of s/m3

. For a 

given point or location at some distance from the source, it represents the average air concentration in 

μCi/m3 divided by the release rate (μCi/s):

 over 

c χ Q' Q' = (J-1)

where

c = annual average concentration at some location (μCi/m3

) 

Q' = radionuclide release rate (μCi/s) 

Radionuclide release rate information is available or can be readily determined for the West Jefferson 

facilities, but onsite air concentration data are not. However, air concentrations or atmospheric 

dispersion factors can be estimated using conservative atmospheric dispersion assumptions. The 

approach taken was to use Equation 2.9 from NCRP Report 123, Screening Models for Release of 

Radionuclides to Atmosphere, Surface Water, and Ground (NCRP 1996). This equation is applicable 

for situations described in Section 2.2.4 of Report 123: “Source and Receptor Not on Same Building 

Surface, x ≤2,5 (AG)

0.5 or x ≤ 100m.” This equation is:

fQ C

πuhK = (J-2)

where

C = concentration (μCi/m3)

f = frequency the wind blows toward the receptor (default = 0.25)

Q = release rate (μCi/s)

u = wind speed (assumed at 2 m/s) 

h = smaller of building height or width (assumed at 10 m)

K = constant (1 m)

Based on a unit release rate of 1 μCi/s, the concentration at some location less than 100-m distance 

is calculated using this equation as 0.004 μCi/m3

. Substituting into the first equation above, the 

representative onsite atmospheric dispersion factor is 0.004 s/m3

. 

This value would decrease for greater wind speeds and larger buildings. For comparison, the 1973 

annual environmental report (BCL 1974) states a dispersion factor of 2.87 × 10-5 s/m3

 for the 

maximally exposed individual, and the 1986 annual report (BCL 1987) states a factor of 3.35 × 10-5

s/m3 for the MEI at a distance 500 meters southeast. These comparisons make the onsite factor 

seem reasonably conservative.

The National Oceanic and Atmospheric Administration reports the annual average wind speed for 

Columbus, Ohio to be 8.3 mph, or 3.7 m/s, based on 53 years’ worth of data 

(http://lwf.ncdc.noaa.gov/oa/climate/online/ccd/avgwind.html), accessed 20 January 2012. The 1986 

annual report documents an average wind speed of 4.5 m/s, so use of the 2 m/s wind speed in the 

second equation above is conservative.
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ATTACHMENT G

ESTIMATING ONSITE AIR CONCENTRATIONS (continued)

Ten meters was used as a representative industrial building height; because emissions are from 

stacks, the heights could be even greater. The buildings at the West Jefferson and King Avenue sites 

are clearly longer than 10 m, as can be seen in Figures G-1 and G-2. The 1986 annual report uses a 

stack height of 18 m and an effective release height of 24.2 m, so 10 m is conservative. Because

total emissions were considered but actually come from four different stacks on JN-1 and nine stacks 

total, the frequency of 0.25 is probably even more conservative than intended as a default value. A 

dispersion factor of 0.004 s/m3 also compares reasonably and conservatively to the site boundary 

dispersion factor of 0.0000287 s/m3

, being 140 times larger.

Therefore, an atmospheric dispersion factor of 0.004 s/m3 is considered a reasonably conservative 

representation of onsite atmospheric dispersion at the West Jefferson site, and is applicable and 

conservative at the King Avenue site as well. The atmospheric dispersion factor can be used with 

site-specific radionuclide release data and worker breathing rates to make estimates of inhaled 

radionuclide activity, as shown in Tables G-1 and G-2. 

Figure G-1. Layout of the West Jefferson site, 1973.
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ATTACHMENT G

ESTIMATING ONSITE AIR CONCENTRATIONS (continued)

Figure G-2. Layout of the King Avenue facilities that had stack emission points, 1973. 
ATTACHMENT G

ESTIMATING ONSITE AIR CONCENTRATIONS (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 144 of 145

Table G-1. Intakes of alpha-emitting radionuclides from the West Jefferson site by year (Bq).

Year U-238 U-235 Pu-238 Pu-239 Am-241 U-234

1973 Not detected Not detected Not detected 3.96E-03 Not detected Not detected

1974 Not detected Not detected Not detected 4.27E-03 Not detected Not detected

1975 Not detected Not detected Not detected 3.96E-03 Not detected Not detected

1976 Not detected Not detected Not detected 7.01E-03 Not detected Not detected

1977 Not detected 2.13E-03 Not detected 6.70E-03 Not detected Not detected

1978 Not detected Not detected Not detected 1.28E-02 Not detected Not detected

1979 Not detected 2.26E-04 Not detected 3.96E-03 Not detected Not detected

1980 Not detected 2.04E-03 Not detected 8.23E-03 Not detected Not detected

1981 Not detected 1.13E-02 Not detected 1.80E-03 Not detected Not detected

1982 Not detected 5.79E-03 Not detected 5.79E-04 Not detected Not detected

1983 Not detected 2.83E-03 Not detected 9.45E-03 Not detected Not detected

1984 Not detected 1.80E-03 Not detected 4.88E-03 Not detected Not detected

1985 Not detected 2.07E-02 Not detected 6.70E-03 Not detected Not detected

1986 Not detected 2.99E-02 Not detected 4.57E-03 Not detected Not detected

1987 Not detected Not detected Not detected 6.70E-03 Not detected Not detected

1988 1.80E-01 1.86E-02 1.13E-04 4.57E-04 2.96E-02 Not detected

1989 1.55E-01 1.13E-02 1.13E-04 4.57E-04 1.92E-02 Not detected

1990 1.01E-01 1.98E-02 1.13E-04 Not detected 1.43E-02 Not detected

1991 1.95E-02 3.35E-02 4.57E-04 1.13E-04 2.01E-02 Not detected

1992 1.49E+00 1.28E-01 3.35E-04 1.13E-04 2.38E-01 Not detected

1993 2.77E+00 2.47E-01 2.26E-04 1.13E-04 1.80E-01 Not detected

1994 1.22E-01 5.79E-02 1.22E-04 5.49E-05 1.37E-02 Not detected

1995 1.04E-01 5.18E-02 Not detected Not detected 1.55E-02 Not detected

1996 1.10E-04 8.23E-06 3.02E-05 5.18E-05 1.10E-02 1.22E-04

1997 1.68E-06 1.43E-05 1.65E-05 2.32E-05 1.46E-02 1.86E-04

1998 2.29E-04 1.62E-05 1.52E-05 1.40E-05 1.68E-02 2.47E-04

1999a 2.00E-04 3.17E-05 1.10E-05 9.60E-06 1.60E-02 2.10E-04

2000 1.71E-04 1.55E-05 6.70E-06 5.18E-06 1.52E-02 1.74E-04

2001 1.65E-04 1.25E-05 1.10E-05 7.62E-06 1.49E-02 1.68E-04

2002 1.52E-04 1.28E-05 1.43E-05 8.53E-06 1.65E-02 1.62E-04

2003 1.40E-04 6.70E-05 2.44E-05 2.26E-05 1.40E-02 1.37E-04

2004 1.31E-04 1.01E-05 1.22E-04 4.27E-05 7.92E-02 1.34E-04

2005b 2.53E-04 2.26E-05 4.57E-04 1.34E-04 9.45E-03 2.29E-04

a. No report was located for 1999; values are based on average of adjacent years. 

b. Intakes for 2005 may be applied to subsequent years through the present.
ATTACHMENT G

ESTIMATING ONSITE AIR CONCENTRATIONS (continued)

Document No. ORAUT-TKBS-0058 Revision No. 01 Effective Date: 06/27/2016 Page 145 of 145

Table G-2. Intakes of mixed fission product radionuclides from the West Jefferson site by year (Bq).

Year Co-57 Co-60 Sr-90 Sb-125 Cs-134 Cs-137 Eu-152 Eu-154

1975 Not detected 4.27E-02 Not detected 7.92E-03 5.79E-03 1.22E-02 Not detected Not detected

1976 1.01E-03 5.79E-02 Not detected 1.43E+00 5.49E-03 5.49E-03 Not detected Not detected

1977 4.57E-04 1.58E+00 Not detected 3.96E-01 1.46E-03 1.46E-02 9.14E-03 5.18E-03

1978 2.41E-02 3.02E-01 Not detected 1.10E-01 2.83E-03 4.88E-02 Not detected Not detected

1979 1.25E-03 2.83E-02 Not detected 6.10E-03 3.35E-04 1.01E-03 Not detected 1.13E-04

1980 5.79E-04 5.49E-02 Not detected 5.18E-01 1.01E-04 7.01E-03 Not detected 1.13E-05

1981 1.13E-04 3.35E-03 3.35E-02 9.75E-02 2.86E-02 8.53E-02 1.25E-03 Not detected

1982 1.13E-04 5.79E-02 3.35E-06 5.18E-02 7.01E-03 4.27E-02 4.88E-03 5.18E-03

1983 Not detected 1.98E-02 5.79E-07 5.18E-02 1.13E-04 2.44E-02 Not detected Not detected

1984 Not detected 8.84E-03 3.35E-07 2.59E-02 Not detected 3.35E-03 Not detected Not detected

1985 Not detected Not detected 3.96E-03 5.79E-02 Not detected 2.65E-02 Not detected Not detected

1986 Not detected Not detected 4.88E-03 5.79E-02 1.65E-02 3.35E-02 Not detected Not detected

1987 Not detected 2.26E-02 1.46E-03 4.27E-02 1.95E-02 4.57E-02 4.27E-02 5.79E-02

1988 1.74E-02 3.66E-02 1.34E-03 7.31E-02 2.59E-02 3.35E-02 7.31E-02 9.75E-02

1989 5.49E-03 1.52E-02 1.71E-03 2.56E-02 9.75E-03 1.46E-02 2.56E-02 3.96E-02

1990 4.88E-03 1.40E-02 2.38E-03 2.86E-02 1.77E-02 2.77E-02 2.44E-02 3.66E-02

1991 8.23E-03 1.71E-02 1.46E-03 3.66E-02 1.19E-02 1.65E-02 3.05E-02 4.88E-02

1992 3.02E-01 9.14E-02 3.96E-03 1.62E-01 7.92E-02 1.25E-01 4.88E-01 3.05E-02

1993 2.38E-02 6.40E-02 2.26E-04 1.46E-01 5.18E-02 5.79E-02 2.13E-01 4.88E-02

1994 7.31E-03 1.49E-02 7.31E-03 3.96E-02 1.31E-02 1.43E-02 2.04E-02 1.43E-02

1995 5.79E-03 1.28E-02 1.71E-04 3.66E-02 1.13E-02 1.19E-02 1.65E-02 1.16E-02

1996 5.79E-03 1.16E-02 1.19E-04 3.05E-02 1.04E-02 1.16E-02 1.68E-02 1.16E-02

1997 6.10E-03 1.31E-02 1.10E-04 3.35E-02 1.16E-02 1.25E-02 1.77E-02 1.25E-02

1998 Not detected 1.40E-02 1.40E-04 3.05E-02 1.22E-02 1.34E-02 1.77E-02 1.25E-02

1999a No report 1.33E-02 1.87E-04 3.20E-02 1.17E-02 1.30E-02 1.71E-02 8.63E-03

2000 Not detected 1.25E-02 2.35E-04 3.35E-02 1.13E-02 1.25E-02 1.65E-02 1.16E-03

2001 Not detected 1.31E-02 3.05E-04 3.05E-02 1.10E-02 1.22E-02 1.65E-02 1.16E-02

2002 Not detected 1.28E-02 2.90E-04 3.35E-02 1.13E-02 1.25E-02 1.68E-02 1.83E-02

2003 Not detected 1.25E-02 3.05E-04 3.35E-02 1.10E-02 1.16E-02 1.62E-02 1.13E-02

2004 Not detected 8.84E-02 8.53E-04 2.59E-02 8.84E-03 1.98E-01 1.25E-02 8.84E-03

2005b Not detected 1.13E-02 5.49E-03 3.05E-02 1.07E-02 1.01E+00 1.62E-02 1.13E-02

a. No report was located for 1999; values are based on average of adjacent years. 

b. Intakes for 2005 may be applied to subsequent years through the present.