ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

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Document Title:

Site Profile for Clarksville Base Weapons 

Storage Area and Modification Center with 

Supplementary Guidance for Medina Base

Document Number: ORAUT-TKBS-0039

Revision: 00

Effective Date: 11/14/2006

Type of Document: TBD

Supersedes: None 

Subject Experts: Ronald J. McConn, Jr., Eva Eckert Hickey, Richard J. Traub, Jerome B. Martin, 

and Donald E. Bihl

Approval: Signature on File Approval Date: 10/17/2006

Donald E. Bihl, Document Owner

Approval: Signature on File Approval Date: 10/18/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/17/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 10/23/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 11/14/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 2 of 69

PUBLICATION RECORD 

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

11/14/2006 00 New Site Profile for Clarksville Base Weapons Storage Area and 

Modification Center with Supplementary Guidance for Medina. First 

approved issue. Incorporated responses to internal formal review 

comments. Added sections 5.2.2 and 6.5 covering guidance for 

1966-67 after operations had ceased. Changes made in response to 

NIOSH formal review comments including: use of “favorable to 

claimants” throughout; removal of HT-to-HTO conversion factor for all 

sections dealing with tritium; Sections 4.0 and 5.1.1 – use of Pantex 

1989 accident doses as surrogate for Clarksville tritium release 

accident; Section 5 – clarification of operation years and 

postoperative years, also added to Glossary; Section 5.2.1 – revised 

DU intakes during weapons disassembly to include room ventilation 

and 8-hr shift; Section 5.2.3 – added DU intake from burning 

contaminated HE; Section 5.3 – added intake from alpha 

contamination on capsules or pits; Section 6.1 – added description of 

roles of SNL, AEC, and military personnel; Section 6.4 – clarified that 

the dose rate from Pu metal is sum of photons and neutrons; Section 

A1.1 – added to description of Medina site and operation; Section 

A2.0 – extended period of medical X-rays to October 1958; Section 

A3.2.2 – clarified dates soil samples were taken; Section A5.0 –

enhanced the description of the radiography source incident. There 

is an increase in assigned dose and a PER is required. Training 

required: As determined by the Task Manager. Initiated by Donald 

E. Bihl.
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 3 of 69

TABLE OF CONTENTS 

SECTION TITLE PAGE

Acronyms and Abbreviations ...............................................................................................................7

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

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

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

2.0 Site Description and Operational History ................................................................................11

2.1 Site Description .......................................................................................................... 11

2.2 Source Terms ............................................................................................................. 14

2.3 Job Descriptions ......................................................................................................... 16

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

3.1 Organ Dose Calculations ............................................................................................ 17

4.0 Occupational Environmental Dose .........................................................................................20

5.0 Occupational Internal Dose ....................................................................................................23

5.1 Tritium Exposure ........................................................................................................ 23

 5.1.1 Operation Years (1949–1965).......................................................................... 23

 5.1.2 Postoperative Years (1966–1967) ................................................................... 25

5.2 Uranium Exposure ...................................................................................................... 25

 5.2.1 Operation Years (1949–1965).......................................................................... 26

 5.2.2 Postoperative Years (1966–1967) ................................................................... 27

 5.2.3 DU from Burning Grounds ............................................................................... 28

5.3 Plutonium or polonium exposure ................................................................................ 29

5.4 Assessing Occupational Internal Dose from Elevated Radon Concentrations ............. 30

 5.4.1 Radon Concentrations ..................................................................................... 30

 5.4.2 Dose from Radon-222 Progeny ....................................................................... 31

5.5 Summary of Intakes .................................................................................................... 32

6.0 Occupational External Dose ...................................................................................................34

6.1 Introduction ................................................................................................................. 34

6.2 External Radiation Dosimeters and Records .............................................................. 34

 6.2.1 Historical Administrative Practices ................................................................... 35

 6.2.2 Dosimetry Technology ..................................................................................... 35

 6.2.2.1 Beta/Photon Dosimeters .................................................................... 36

 6.2.2.2 Neutron Dosimeters ........................................................................... 36

 6.2.3 Dosimeter Calibration Procedures ................................................................... 37

 6.2.3.1 Beta/Photon Dosimeters.................................................................... 37

 6.2.3.2 Neutron Dosimeters .......................................................................... 37

 6.2.4 Workplace Radiation Fields ............................................................................. 37

 6.2.4.1 Depleted Uranium ............................................................................. 37

 6.2.4.2 Photon Radiation ............................................................................... 38

 6.2.4.3 Neutron Radiation ............................................................................. 38

 6.2.5 Dosimeter Response to Radiation Fields ........................................................ 39

 6.2.5.1 Beta/Photon Film Dosimeter Response ............................................. 39
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 6.2.5.2 Neutron Dosimeter Response ........................................................... 39

 6.2.5.3 Neutron-to-Photon Dose Ratios......................................................... 39

 6.2.5.4 Neutron Dose Weighting Factor ........................................................ 40

6.3 Recommendations for Clarksville Worker External Dose Reconstruction – Operative 

Years .......................................................................................................................... 40

 6.3.1 Unmonitored External Dose ............................................................................. 40

 6.3.2 Missed External Dose for Monitored Workers .................................................. 44

 6.3.3 Radiation Dose Fraction .................................................................................. 45

6.4 Organ Dose – Operative Years ................................................................................... 45

6.5 Organ Dose – PostOperative Years............................................................................ 45

References ........................................................................................................................................47

Glossary ............................................................................................................................................52

ATTACHMENT A, GUIDANCE FOR THE MEDINA WEAPONS STORAGE AREA AND 

MODIFICATION CENTER ............................................................................................. ........54

A.1 Site Description and Functions ................................................................................... 55

A.1.1 Site Description .......................................................................................................... 55

A.1.2 Operational History as a Weapons Storage Area........................................................ 59

A.1.3 Operation as a Modification/Disassembly Center........................................................ 60

A.2 Occupational Medical Dose ........................................................................................ 60

A.3 Environmental Dose ................................................................................................... 61

A.3.1 Routine Doses ............................................................................................................ 61

A.3.2 November 13, 1963, Explosion ................................................................................... 61

A.3.2.1Background ..................................................................................................... 61

A.3.2.2Intake Calculation ........................................................................................... 62

A.3.2.3Direct Radiation .............................................................................................. 64

A.3.2.4Inhalation from Resuspension ......................................................................... 65

A.4 Internal Dose .............................................................................................................. 66

A.5 External Dosimetry ..................................................................................................... 67
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LIST OF TABLES 

TABLE TITLE PAGE

2-1 Clarksville Base job categories .............................................................................................. 16

3-1 Organ doses from PA conventional chest X-rays ................................................................... 18

3-2 Organ doses from lumbar spine X-rays .................................................................................. 19

4-1 Summary environmental intake .............................................................................................. 22

4-2 Summary environmental external dose .................................................................................. 22

5-1 Radon concentrations in the 7740 Building ............................................................................ 30

5-2 Radon concentrations in Clarksville structures ....................................................................... 31

5-3 Statistical parameters of radon concentrations ....................................................................... 31

5-4 Working level-months for Clarksville workers ......................................................................... 32

5-5 Occupational intakes .............................................................................................................. 33

6-1 Dosimeter type, period of use, exchange frequency, MRD, and MDL .................................... 35

6-2 Beta, photon, and neutron radiation energies and percentages for Clarksville 

facilities .................................................................................................................................. 38

6-3 Neutron dose energies, percentages and associated ICRP (1991) correction factors ............ 40

6-4 Worker job categories and exposure groups .......................................................................... 41

6-5 Pantex worker photon dose statistics ..................................................................................... 41

6-6 Pantex worker photon statistics for all dosimeters .................................................................. 42

6-7 Dose recommendations for Clarksville Group 1 workers ........................................................ 42

6-8 Dose recommendations for Clarksville Group 2 workers ........................................................ 42

6-9 Dose recommendations for Clarksville Group 3 workers ........................................................ 42

6-10 Potential missed dose for Clarksville workers ......................................................................... 44

6-11 Annual external dose to organs from natural uranium surface contamination, 1966 

to 1967 ................................................................................................................................... 46

A-1 Assumed isotopic mixtures of uranium ................................................................................... 62

A-2 Uranium soil measurements after explosion ........................................................................... 63

A-3 Environmental intakes ............................................................................................................ 66

A-4 Environmental external dose .................................................................................................. 66

A-5 Occupational intakes .............................................................................................................. 66

A-6 Dose recommendations for Medina workers .......................................................................... 68
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LIST OF FIGURES 

FIGURE TITLE PAGE

2-1 General plan view of ABC Structure ....................................................................................... 12

2-2 Plan view of the “A” Structur ................................................................................................... 13

2-3 Cross-sectional view of the “A” Structur ................................................................................. 13

2-4 Plan view of the “B” and “C” Structures in the tunnel comple .................................................. 14

2-5 Typical bird cage .................................................................................................................... 15

6-1 Comparison of Hp(10) for photons with energy responses for sensitive DuPont 502 

emulsion in two-element film badge and sensitive DuPont 555 emulsion in 

multielement film badge ......................................................................................................... 36

6-2 Log probability plot of annual doses received by monitored workers at Clarksville 

Base, 1949 to 1958 ................................................................................................................ 43

6-3 Log probability plot of Pantex dosimetry data that includes missed doses .............................. 44

A-1 San Antonio, Texas, area showing location of Lackland Training Annex (previously 

called Medina Base). .............................................................................................................. 56

A-2 Medina Base .......................................................................................................................... 57

A-3 Log probability plot of recorded annual doses received by monitored workers at 

Medina, 1959 to 1966 ............................................................................................................ 68
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission 

AEHA (U.S. Army) Environmental Hygiene Agency 

AFB Air Force Base 

AIRS Aerometric Information Retrieval System 

AP anterior-posterior 

CATI Computer-Assisted Telephone Interview 

CD compact disk 

Ci curie 

cm centimeter 

d day

DOE U.S. Department of Energy 

dpm disintegrations per minute 

DU depleted uranium 

EEOICPA Energy Employees Occupational Illness Compensation Program Act 

ft foot 

g gram 

gal gallon 

GSD geometric standard deviation 

Gy gray 

HEPA high-efficiency particulate air 

HEU highly enriched uranium 

Hp(d) personal dose equivalent at tissue depth d (d = 10 mm or 0.07 mm) 

hr hour

HVL half-value layer 

IAAP Iowa Army Ammunition Plant 

ICRP International Commission on Radiological Protection 

IMBA Integrated Modules for Bioassay Analysis 

in. inch 

IREP Interactive RadioEpidemiological Program 

keV kiloelectron-volt 

kg kilogram 

L liter 

LAT lateral 

lb pound

LLI lower large intestine 

LLRW low-level radioactive waste 

m meter 

mCi millicurie 

MDL minimum detection level 
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MeV megaelectron-volt, 1 million electron volts 

mGy milligray 

MHSMC Mason and Hanger-Silas Mason Company, Inc. 

min minute 

mL milliliter 

mm millimeter 

mo month 

MRD minimum recordable dose 

mrad millirad 

mrem millirem 

NCRP National Council on Radiological Protection and Measurements 

NIOSH National Institute for Occupational Safety and Health 

NTA Eastman Kodak nuclear track emulsion, type A 

NUREG U.S. Nuclear Regulatory Commission Regulation 

PA posterior-anterior 

PAEC potential alpha energy concentration 

PAEE potential alpha energy exposure 

pCi picocurie 

PM10 particulate matter in the air with an aerodynamic diameter less than or equal to 

10 micrometers 

POC probability of causation 

ppm parts per million 

s second 

SI small intestine 

SNL Sandia National Laboratories 

ST stomach 

TBD technical basis document 

TLD thermoluminescent dosimeter 

TNT trinitrotoluene 

USAPMS U.S. Army Preventive Medicine Service 

U.S.C. United States Code 

UST underground storage tank 

WL working level 

WLM working level-month 

WSA Weapons Storage Area 

yr year 

Z atomic number 

μCi microcurie 

μg microgram 

μm micrometer 

§ section 
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1.0 INTRODUCTION

1.1 PURPOSE 

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

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

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

dose reconstructions at particular sites or categories of sites. The documents will be revised in the 

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

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

reconstruction. 

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA. 

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the POC guidelines (nor the dose 

reconstruction regulation) define “performance of duty” for DOE employees with a covered cancer or 

restrict the “duty” to nuclear weapons work. 

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

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

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

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

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

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

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

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

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

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

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

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

derived: 

 Radiation from naturally occurring radon present in conventional structures 

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

 

1

 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
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1.2 SCOPE 

This Site Profile provides information about U.S. Atomic Energy Commission (AEC) operations at 

Clarksville Base pertaining to radiation exposures for monitored or unmonitored workers. Section 2.0

describes the site and operations pertaining to possible radiation exposures and discusses radiation 

source terms. Section 3.0 provides guidance for determining occupational medical dose. Section 4.0

provides guidance for determining dose to workers outside radiological facilities due to releases of 

radioactive materials to the environment. Section 5.0 provides guidance for determining intakes of 

radionuclides inside facilities. Section 6.0 provides guidance for determining external doses from 

measured doses or for periods when records of measured doses are missing. Because the Medina 

Weapons Storage Area (WSA) and Modification Center were similar in purpose and operation to 

Clarksville Base, Attachment A of this document contains information about Medina. 
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2.0 SITE DESCRIPTION AND OPERATIONAL HISTORY

2.1 SITE DESCRIPTION 

Clarksville Base, at Fort Campbell, Kentucky/Tennessee, was one of 13 WSAs created under the 

Armed Forces Special Weapons Project. The Base was constructed in the mid- to late 1940s (the 

first weapon components arrived in July 1949) and supported by Sandia Corporation [later called 

Sandia National Laboratories (SNL)] for the AEC and the U.S. Navy. SNL, AEC, and the Navy were 

all active at Clarksville Base from 1949 until 1958 performing maintenance and quality assurance on 

nuclear components of weapons. From 1958 until 1965, Mason and Hanger-Silas Mason Company, 

Inc. (MHSMC) operated Clarksville Base for the AEC as a weapons modification and disassembly 

facility (Lamb Associates and Halliburton NUS 1996; McConn 2006; Mitchell 2003). 

Clarksville Base was originally separate from Fort Campbell, which was operated by the U.S. Army. 

In 1965, AEC activities transferred elsewhere, although the AEC was responsible for the Base 

through 1967. The Base was returned to the Army and incorporated into Fort Campbell in 1969 (Last, 

Gilmore, and Bronson 1998). It is unclear what activities occurred between 1967 and 1969 but 

storage of nuclear materials had ceased. 

During the AEC tenure, nuclear weapons and weapon components were stored by the AEC and 

maintained by SNL and military personnel at the WSAs. WSAs consisted of storage buildings that 

housed nuclear capsules, maintenance structures, waste burial sites, and bunkers for storage of 

weapons casings. SNL personnel worked at Clarksville Base under contract to the AEC until early 

1962 (Lamb Associates and Halliburton NUS 1996; Last, Gilmore, and Bronson 1998; McConn 2006). 

Storage of nuclear capsules at Clarksville Base was in an underground complex known as the ABC 

structure. The ABC structure consisted of the “A” structure, which was the nuclear capsule storage 

area that was secured behind a bank-type locking vault door at the end of a 600-ft-long tunnel; the “C” 

structure, which was used for nuclear materials inspection and maintenance; and the “B” structure, 

which was a backup facility for the “C” structure but was used only as a medical wing. “C” structure 

activities involved dismantling the nuclear assembly system, checking the activity of the fissile 

material, and replacing the polonium-beryllium (Po-Be) initiators (Lamb Associates and Halliburton 

NUS 1996; Last, Gilmore, and Bronson 1998; McConn 2006). Figure 2-1 is an overall plan view of 

the ABC structure. Figure 2-2 shows a close-up plan view of the “A” structure and Figure 2-3 shows a 

cross-sectional view. Figure 2-4 shows a plan view of the “B” and “C” structures. [Figures 2-1 

through 2-4 are from Last, Gilmore, and Bronson (1998); they refer to “Fort Campbell” because they 

were inactive facilities on Fort Campbell in 1998; however, at the time of use they were the Clarksville 

WSA.] 

A second “C” structure, constructed in 1957, was an above-ground brick building used only to service 

non-nuclear components. Once this structure opened, the original “C” structure ceased to be used 

due to moisture intrusion. In addition, the Clarksville WSA included a Gravel Gertie that was used for 

weapons maintenance and modification. The newer “C” structure and Gravel Gertie had large 

overhead rails to support weapon subassemblies during maintenance. Standard above-ground igloos 

were used for storage, including the storage of sealed weapons. No nuclear maintenance activities 

occurred in these igloos (Lamb Associates and Halliburton NUS 1996; Last, Gilmore, and Bronson 

1998; McConn 2006). 

Some of the NIOSH Dose Reconstruction Project Computer-Assisted Telephone Interviews (CATIs) 

refer to the “Bird Cage.” This term was used officially to describe the criticality-safe framework built 
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Figure 2-1. General plan view of ABC Structure. 
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Figure 2-2. Plan view of the “A” Structure. 

Figure 2-3. Cross-sectional view of the “A” Structure. 
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Figure 2-4. Plan view of the “B” and “C” Structures in the tunnel complex. 

around a nuclear package (see Figure 2-5); however, the term appears to have been used unofficially 

to refer to the whole Clarksville Modification Center, meaning the fenced, heavily guarded area where 

an AEC Q-level clearance was required (more than just the tunnel complex). 

2.2 SOURCE TERMS 

Early weapons designs were of the in-flight insertable variety. Weapons of this type had removable 

nuclear capsules (also known as the physics package or pit) and were stored in a bird cage. The bird 

cage ensured storage in a criticality-safe manner. The capsules were pressure-sealed. Figure 2-5 

shows a typical bird cage. The bird cages would hold the nuclear weapon, comprised of 
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Figure 2-5. Typical bird cage. 

plutonium, highly enriched uranium (HEU), and Po-Be initiator. Periodic maintenance was required 

on these early weapon pits to exchange the Po-Be initiators due to the short half-life of 210Po. This 

would require disassembly of the weapon pit to remove the initiator, which was at the pit center. 

Later weapon designs did not utilize the in-flight insertable concept or the Po-Be initiator, thereby 

eliminating the need to disassemble the weapon pit for modification. The Po-Be initiator was phased 

out over time, until 1956, and replaced by external neutron generators. The weapon pits are referred 

to as sealed pit designs. These designs included the potential for exposure to tritium. The 

introduction of tritium could have occurred as early as 1954 (McConn 2006; Mitchell 2003). 

Another source of radioactive material used in early nuclear weapons was the spark gap tube. These 

tubes, which were part of the firing circuits, were used to switch large amounts of electrical current. A 

small amount of 137Cs was used in spark gap tubes to stabilize the electrical properties. These tubes, 

which were manufactured of thick glass to prevent breakage, would be a minor exposure pathway 

(McConn 2006). 
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As part of the maintenance activities, Clarksville personnel performed radiographs of the weapon 

components using a large 60Co source. The exact location of this source was not discovered but the 

newer “C” structure is a likely candidate (McConn 2006). 

In summary, the radioactive materials of interest at Clarksville Base are tritium as a gas, weaponsgrade plutonium, HEU, depleted uranium (DU; also used in weapons construction), 210Po in a Po-Be 

neutron generator, a 60Co radiograph source, and small activities of 137Cs. 

2.3 JOB DESCRIPTIONS 

Table 2-1 lists job descriptions described in the claims files or by interviews of former employees (Bihl 

2006a,b). 

Table 2-1. Clarksville Base job categories. 

Title Description

Material handler Moved nuclear devices to and from storage and disassembly areas or 

among magazines; unloaded devices from trucks, railcars, and aircraft 

and drove them to storage areas.

Production operator, operator, operator 

trainee

Assembled/disassembled nuclear devices.

Inspector, quality control specialist, 

quality control inspector

Nuclear components inspectors observed assembly/disassembly, 

recorded condition of components, and ensured correct assembly of 

components; it is possible these workers performed gamma/X-ray 

inspections of devices. Not all inspectors were responsible for nuclear 

components and would have had only incidental exposure to complete 

weapons. Latter should be considered in same category as material 

handlers. 

Warehouseman Received, stored, shipped nuclear devices; conducted inventory of 

nuclear devices. One CATI stated about every 6 mo warehouseman 

worked in storage igloos for 1 week conducting inventory. 

Safety/security inspector Performed security inspection and control of all buildings including 

magazines; probably spent some time in all secure locations but did not 

handle nuclear devices.

Mechanic Repaired equipment, moved nuclear devices; CATI indicates mechanics 

might have been responsible for replacing filters in ventilation systems 

including contaminated filters; spent time in igloos when necessary.

Truck driver, heavy equipment operator Probably involved with transporting nuclear devices to/from railcars, 

airport.

Fireman CATI claims that fireman stood by with fire extinguisher during 

disassemblies.

Sheet metal worker, electrician, 

refrigerator/cooling mechanic, janitor

Probably worked anywhere and might have had some exposure in igloos.

Clerk-typist Worked in offices in Bird Cage; might have entered disassembly areas to 

deliver messages.

Accountant Had office in Bird Cage; occasionally entered igloos.

Bus driver, grounds laborer, power plant 

operator, sewage disposal operator

Probably did not have any exposure except environmental.
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3.0 OCCUPATIONAL MEDICAL DOSE

It is not known if medical X-rays were required for all workers or selected workers as a condition of 

employment. No documentation on X-ray policies, procedures, or equipment has been found. Harold 

Rarrick, an SNL safety engineer familiar with work at WSAs, did not recall X-rays being part of 

medical examinations at any of the WSAs during the period when Sandia was the principal AEC 

contractor (McConn 2006). Records on Clarksville workers from Pantex do not list any X-rays prior to 

1960, even for workers who continued employment into the years MHSMC was the principal AEC 

contractor. A review of the claims information revealed that, for employment between 1960 and 1965, 

53% of the Energy Employees had at least one chest X-ray and 35% had at least one lumbar spine 

X-ray. Most of the X-rays are either labeled as “pre-employment” or occurred in the first year of 

employment. The pattern does not support additional routine X-rays; only four workers had more than 

one chest X-ray and only one had a second lumbar spine X-ray. In terms of job categories, no pattern 

was evident in relation to who had preemployment X-rays and who did not. For instance, the records 

for the two claims for clerk-typists showed one chest X-ray each, whereas some craft workers had no 

record of X-rays. 

Based on the limited information available for workers during 1960 through 1967 with no X-ray 

records, the dose reconstructor should assume one chest X-ray and one lumbar spine examination for 

the entire employment period (not annual). (As described below, a lumbar spine examination is 

assumed to consist of four shots.) The dose from X-rays should be assigned in 1960 or the first year 

of employment after 1960. Do not assign X-rays for employment prior to 1960. 

No information about X-ray equipment manufacturers, models, examination techniques, and exposure 

rates for those techniques has been found. Therefore, assumptions that are favorable to claimants 

and guidance described in Dose Reconstruction from Occupationally Related Diagnostic X-ray 

Procedures (ORAUT 2005a) were used. The actual film was not sent to the Pantex Plant for 

archiving, only the information that an X-ray occurred, the type, and the date; the film size and hence 

type of radiography are not known. For small DOE sites conventional chest X-rays are assumed. 

Lumbar spine X-rays at the Pantex Plant were given only to men. None of the lumbar spine X-rays in 

the Clarksville claim files were given to women; however, the number of women Energy Employees 

among the Clarksville claims is small. Nevertheless, because lumbar spine X-rays were given to 

screen for back injuries that might preclude heavy lifting, and considering that heavy lifting was 

culturally a man’s job in the workplace in 1960–1965, it is reasonable to assign the default lumbar 

spine X-ray only to men. 

3.1 ORGAN DOSE CALCULATIONS 

Guidance in ORAUT (2005a) was used to determine organ doses. For conventional chest X-rays, 

that document recommended a default entrance kerma of 0.2 rem for pre-1970 examinations (ORAUT 

2005a, Table 3-4). In addition, Table 6-5 of ORAUT (2005a) provides factors for converting kerma to 

organ dose for pre-1970 machines that might have had limited collimation. Because nothing is known 

about the type of X-ray machine used, it is favorable to claimants to assume limited collimation. It is 

also assumed that the view is posterior-anterior (PA). The dose conversion factors and organ doses 

from ORAUT (2005a) are listed in Table 3-1. Dose reconstructors should use the most recent update 

of that document. 

According to ORAUT (2005a), two anterior-posterior (AP) shots and two lateral (LAT) shots should be 

assumed for lumbar spine examinations; however, the doses in ORAUT (2005a) include both shots 

per view. The entrance skin exposure for AP was 4.00 rem and for LAT was 10 rem, and it was 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 18 of 69

Table 3-1. Organ doses from PA conventional chest X-raysa. 

Organ

PA dose conversion factor 

(mGy per Gy air kerma) 

HVL 2.5 mm Al minimum 

collimation Organ dose (rem)

Thyroid 174 3.48E-2

Eye/brain 32 6.40E-3

Ovaries NAb2.5E-2

Liver/gall bladder/spleen 451 9.02E-2

Urinary bladder NA 2.5E-2

Colon/rectum NA 2.5E-2

Testes NA 5.0E-3

Lungs (male) 419 8.38E-2

Lungs (female) 451 9.02E-2

Thymus 451 9.02E-2

Esophagus 451 9.02E-2

Stomach 451 9.02E-2

Bone surfaces 451 9.02E-2

Remainder 451 9.02E-2

Breast 49 9.80E-3

Uterus NA 2.5E-2

Bone marrow (male) 92 1.84E-2

Bone marrow (female) 86 1.72E-2

Skinc NA 2.70E-1

a. From ORAUT (2005a, Table 6-5, pre-1970). 

b. NA: not available 

c. Includes backscatter factor of 1.4. 

assumed that the entrance skin exposure in rem is equal to the air kerma in rads. The entrance skin 

exposures were based on data applicable to 1950–1953 but, lacking any other information, are 

assumed to be applicable to the X-rays given at Clarksville Base in 1960–1965 as well. The organ 

doses for lumbar spine X-rays in ORAUT (2005a) per set of two shots are listed in Table 3-2. Dose 

reconstructors should use the most recent update of that document. Enter the values from ORAUT 

(2005a) into the Interactive RadioEpidemiological Program (IREP) as an acute dose due to photons 

with energies between 30 and 250 keV. Assume a normal distribution with a standard deviation of 

±30%. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 19 of 69

Table 3-2. Organ doses from lumbar spine X-rays.a

Organ View Organ dose (rem)

Thyroid AP 8.00E-4

LAT 1.00E-4

Eye/brain AP 8.00E-4

LAT 1.00E-4

Ovaries AP 1.12b

LAT 1.52b

Liver/gall bladder/spleen AP 2.48E-1

LAT 1.00E-1

Urinary bladder AP 1.12b

LAT 1.52b

Colon/rectum AP 1.12b

LAT 1.52b

Testes AP 5.40E-2

b

LAT 1.12E-1

b

Lungs AP 2.48E-1

LAT 1.00E-1

Thymus AP 2.48E-1

LAT 1.00E-1

Esophagus AP 2.48E-1

LAT 1.00E-1

Stomach AP 2.48E-1

LAT 1.00E-1

Bone surfaces AP 2.48E-1

LAT 1.00E-1

Remainder organs AP 2.48E-1

LAT 1.00E-1

Breast AP 7.20E-2

LAT 9.50E-2

Uterus AP 8.68E-1

LAT 2.00E-1

Bone marrow AP 9.60E-2

LAT 1.50E-1

Skin AP 5.28c

LAT 1.32E+1c

a. From ORAUT (2005a). 

b. Based on actual measurements by Lincoln and Gupton (1957).

c. Includes backscatter factor of 1.32. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 20 of 69

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

Occupational environmental dose refers to the dose received by workers outside normal production 

facilities. These doses can be internal or external (e.g., from effluents or scattered radiation through 

building walls or ceilings). No records on environmental releases from Clarksville Base have been 

discovered. Prior to 1959, complete disassemblies were not performed at Clarksville Base, only 

storage, maintenance, and inspection. Maintenance included replacement of major components. 

Plutonium and enriched uranium sources in the weapons were always sealed, as were polonium in 

the initiators and other radionuclides in the radiography sources. There was risk of DU oxide 

contamination in the cells. The underground “C” structure had exhaust vents at the top of the hillside 

under which the tunnel was located (Last, Gilmore, and Bronson 1998). Whether the exhaust passed 

through high-efficiency particulate air (HEPA) filters is not known. According to Harold Rarrick, 

portable gloveboxes with HEPA-filtered exhausts were used to contain oxidized DU when weapons 

underwent inspections, maintenance, and refurbishment during the Sandia years (McConn 2006). 

The DU was cleaned from the nuclear components and deposited as solid waste on cleaning rags. 

Mr. Rarrick mentioned that the DU contamination came from spalling, which produced large 

nonrespirable particles. Environmental intake of DU was insignificant because (1) the likelihood of a 

significant release outside the tunnel was small and (2) the location of the exhaust vent makes it 

unlikely that DU that might have exhausted from the “C” structure would have returned to occupied 

areas of Clarksville Base. 

After tritium reservoirs became part of the weapons, leaks of tritium into the disassembly cell and out 

the cell exhaust duct were possible. Mr. Rarrick was not aware of any releases of tritium from the 

disassembly cells; however, most handling of the tritium reservoirs occurred during the years MHSMC 

operated the facility (McConn 2006). No documentation about tritium releases into the cells or 

through exhaust stacks has been found. Tritium releases from similar work at the Iowa Army 

Ammunition Plant (IAAP) were documented for December 1965 through December 1970, averaging 

26,000 μCi/yr (ORAUT 2005b). Lacking additional information, it was assumed that Clarksville Base 

had similar effluent rates of tritium. Tritium in the reservoirs was in the form of tritiated gas (HT), 

which has essentially no significance for intakes. Tritium gas converts slowly to tritiated water vapor 

as it mingles with humid air. Not all the gas would convert to water vapor during the time workers 

would be exposed (Peterson and Davis 2002); however, it is favorable to claimants to assume 100% 

water vapor. 

To estimate intakes to workers outside facilities if little or no atmospheric information is available, 

techniques recommended by the National Commission on Radiological Protection (NCRP) in 

Volume I of Report 123 (NCRP 1996) were used. The NCRP recommends a graded approach, with 

three screening levels. Level 1 is the most conservative (least dispersion) and requires the least 

amount of input information. The Level 1 method for determining an upper bound air concentration 

can be written as: 

)/(

)/(

)/(

3

3

smV

spCiQxF mpCiX 

(4-1)

where: X = annual average upper-bound air concentration 

F = assumed fraction of time the wind blows in the direction of the subject, assumed to 

be 0.25 

Q = release rate of the radionuclide from the source 

V = volumetric flow rate of the vent (the default value is 0.3 m3/s, typical of hood 

ventilation rates) 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 21 of 69

The model essentially assumes that the subject breathes undiluted effluent from the vent or stack, 

modified by the fraction of time the wind blows in their direction (the factor of 0.25). 

With the NCRP assumptions, the annual average air concentration near the release point should be 

less than: 

 26,000 μCi/yr × (1 × 106 pCi/μCi) × 0.25 / (3.15 × 107 s/yr × 0.3 m3/s) = 688 pCi/m3 (4-2)

Because tritium (as water) can be absorbed through the skin, the tritium inhalation intake is multiplied 

by a factor of 1.5 to obtain the total intake of tritium water. Assuming a breathing rate of 2,400 m3/yr 

and assuming all the tritium is in the form of water results in an estimated annual intake of 

688 pCi/m3 × 1.5 × 2,400 m3 = 2.5 × 106 pCi or 6,800 pCi/d (4-3)

This intake rate results in a dose to any organ of less than 1 mrem, which can be ignored. 

Robert Wells, a Clarksville material handler, described an accident involving a damaged weapon 

returned to Clarksville Base from the military (Bihl 2006a). The accident was corroborated in a 

claimant CATI. At the time of the accident, both individuals were told the damaged weapon was 

leaking tritium. No documentation of the accident or information about the amount of tritium that might 

have leaked has been found. The CATI indicated that the accident occurred in 1962.

Because Mr. Wells was told he had been exposed to tritium while moving the weapon from the airport 

to the storage igloo, it is assumed that the tritium leak had been occurring for some time prior to its 

discovery. The reservoirs are under considerable pressure so most of the contents would have 

leaked prior to arrival at Clarksville; however, there was sufficient leakage still occurring to set off the 

tritium monitors the next day when the weapon was taken into the cell, according to the CATI. 

An accidental release of tritium during a disassembly occurred at the Pantex Plant in 1989 (ORAUT 

2006a). This was a major release causing severe contamination of a cell. Workers were present at 

the initiation of the release and therefore subject to the highest release rate. All of the release 

occurred at Pantex, whereas most of the 1962 leak probably did not occur at Clarksville. It is unlikely 

that the accidental release at Clarksville was worse than the Pantex 1989 release. The Pantex 

Environmental Technical Basis Document (ORAUT 2004) assigns 15-mrem dose from the 1989 

accident to all organs for that year. According the CATI, Clarksville workers stopped the leak within a 

few hours of its discovery, whereas the release at Pantex was so significant no attempt was made to 

stop the leak until it had run its course. Therefore, it is reasonable to conclude that the amount of 

tritium released at Clarksville was much less than the amount released during the 1989 incident at 

Pantex and that the environmental dose to workers was no greater than 15 mrem. All Clarksville 

workers should be given a 15-mrem dose from tritium to any organ for 1962 only. Because the 

Clarksville release was undoubtedly smaller than the Pantex release, the distribution is a constant 

upper bound. 

External radiation dose greater than the ambient rate outside a building where frequent radiography 

was performed might have occurred via direct radiation penetration through walls or from scattered 

radiation. Dose rates are usually quite small in noncontrolled areas near radiography sources. For 

instance, at the Hanford Radiological Calibration Facility, the total annual dose measured by 

thermoluminescent dosimeters (TLDs) on the outside of an interior wall at about 25 ft from a 20-Ci 

137Cs source (662-keV gamma ray) used almost daily was 7 mrem. A 60Co radiography source would 

be expected to have less activity, but the gamma radiation strength of 60Co (1,173- and 1,332-keV 

gamma rays) is about 4 times that of 137Cs. Assuming a smaller activity 60Co source produces about 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 22 of 69

the same annual dose rate outside the radiography building as the 20-Ci 137Cs source and, assuming 

an environmental occupancy factor outside the building of 0.5, dose reconstructors should assign a 

4-mrem whole-body dose per year from external radiation. Assume a 100% 30-to-250-keV photon 

energy category. Calculate organ doses of interest using the external dose reconstruction 

implementation guidelines (NIOSH 2002). Because this dose rate was made by inference from a 

similar situation, rather than by direct measurement, an uncertainty factor of 2 is reasonable. 

Table 4-1. Summary environmental intake. 

Period of 

application Mode Type Radionuclide

Dose to 

any organ Distribution

March 1962 Inhalation and skin absorption Acute Tritium 15 mrem Constant

Table 4-2. Summary environmental external dose. 

Period of application Annual whole-body dose Photon energy category Distribution

July 1949–1965 4 mrem 30-250 keV GSD = 2

1966–1967 None
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 23 of 69

5.0 OCCUPATIONAL INTERNAL DOSE

There is no available reference material that describes actual activities at Clarksville Base. One 

reference source describes “maintenance activities” conducted at Clarksville Base; these activities 

appear to have been primarily to replace Po-Be initiators (Lamb Associates and Halliburton NUS 

1996). Po-Be initiators were replaced with another type of initiator from 1954 to 1957. Although there 

was little risk of intake from the Po-Be initiators themselves, the need to replace them presented the 

need to disassemble and reassemble the components, leading to the potential for uranium 

contamination. 

Based on materials used to assemble weapons and considering that the fissile material was 

encapsulated (plated or sealed), the radionuclides most likely to result in an intake at Clarksville Base 

were DU (D38 or tuballoy) and tritium (3H) starting in about 1954. 

Maintenance activities took place in the “C” structures and never in the “A” structures. Therefore, 

workers who spent most of their time in the “A” structures can be considered to have little or no 

potential for intakes (other than radon) (Lamb Associates and Halliburton NUS 1996). 

“C” structures had a ventilation system designed to prevent release of uranium oxides to the 

atmosphere, indicating preparations for potential airborne hazards. 

Procedures used for maintenance described steps during which there were checks for alpha 

contamination on the outside of containers, indicating knowledge of the potential for contamination. A 

procedure described the use of a Plexiglas glovebox, presumably to minimize or prohibit airborne 

contamination; other procedures described removing the oxidized DU during maintenance (Lamb 

Associates and Halliburton NUS 1996). 

No records were found indicating bioassays were performed or internal dose was assessed for 

workers at Clarksville Base. Mr. Rarrick indicated that some in-house tritium urinalysis was performed 

during the Sandia years (McConn 2006). A letter dated March 22, 1961, indicated that there were 

procedures in place for air sampling in the assembly cell as well as procedures for a radiological 

urinalysis kit (Higgins 1961). However, no data were found for air samples or urinalyses. 

5.1 TRITIUM EXPOSURE 

Tritium intakes could have occurred and probably did occur to a certain extent during disassembly 

and maintenance of weapons at Clarksville Base. Review of CATI documents confirmed that tritium 

monitoring occurred, indicating the understanding that tritium intakes were possible. The actual date 

that tritium arrived at the site is unknown, but it appears to have been between 1954 and 1958. It is 

believed that the tritium reservoirs came from the Savannah River Site, which would make late 1955 

the earliest date for tritium exposure at Clarksville Base (Bebbington 1990, page 52). However, in this 

Site Profile, a date of 1954, which is favorable to claimants, was assumed. 

5.1.1 Operation Years (1949–1965)

The technology of tritium use in nuclear weapons is classified. No air sample or tritium bioassay 

results for Clarksville Base were found. Because MHSMC operated Clarksville Base (from around 

1958 to closing in 1965), Pantex Plant, Medina Base, and IAAP, material and procedures at 

Clarksville are assumed to be similar to those at Pantex and IAAP, except Clarksville did not do 

original assemblies (ORAUT 2005b, 2006a). Hundreds of tritium bioassay results were obtained at 

Pantex in the 1970s and 1980s. The largest internal dose at Pantex from tritium recorded during any 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 24 of 69

year in this period, with the exception of a major accident, was 122 mrem. Using the standard 

calculation for tritium in the 1970s, which used a quality factor of 1.7 [based on International 

Commission on Radiological Protection (ICRP) Publication 10 (ICRP 1968) and explained in U.S. 

Nuclear Regulatory Commission Regulation (NUREG)-0938 (Brodsky 1983)] (1.5-mCi uptake resulted 

in 425 mrem), 122 mrem was indicative of a chronic annual uptake of 430 μCi of tritium. The uptake 

calculated using ICRP Publication 10 methodology accounts for tritium in body fluids from any intake 

mode. 

The 122-mrem maximum dose at the Pantex Plant was recorded in the years when Pantex was 

performing more disassemblies. In only one other year was there a dose over 100 mrem for normal 

operations at Pantex (ORAUT 2006a). This can be assumed to imply that it is a bounding dose. 

Disassembly is similar to maintenance activities that were likely to occur at Clarksville Base. Lacking 

any information on tritium bioassay at Clarksville or additional information on activities at the site, 

122 mrem can be considered a reasonable surrogate dose for production operators from 1954 

through the closing of Clarksville operations in 1965. 

Workers whose job categories were Production Operators, Inspectors, and Quality Control Specialists 

and who worked at Clarksville Base between 1954 and 1965 were the most likely people to receive 

the dose discussed above. 

As discussed in Section 4.0, an accident involving a damaged weapon returned to Clarksville Base 

from the military apparently leaked tritium (Bihl 2006a). No documentation of the accident or any 

information about the amount of tritium that might have leaked has been found. The CATI indicated 

that the accident occurred in 1962. 

For consistency with the assumptions made for environmental doses from the alleged accident, the 

accidental release of tritium during a disassembly in 1989 at the Pantex Plant was used as a basis for 

assigning doses to workers involved in the event at Clarksville Base (ORAUT 2006a). The event at 

Pantex was a major release that was unstopped causing severe contamination of a cell. It is unlikely 

that the leak at Clarksville could have been worse than the estimated release at Pantex of 40,000 Ci 

over 12 d. As noted in Section 4, the CATI indicated that the leak at Clarksville was stopped within 

half a day once the weapon was brought into a cell. The leak rate was probably much slower than 

that of the Pantex leak; otherwise, the tritium would have leaked out before the weapon was returned 

to Clarksville Base from the military. 

Tritium urinalysis samples were collected from workers exposed to the Pantex accident. The 

maximally exposed worker had 460 μCi/L in a sample representing the first 24 hr after intake and 

365.1 μCi/L in a spot sample taken 48 hr after the intake. [Other samples were taken but they were 

affected by medical treatment so were not used in this surrogate calculation.] Using the Integrated 

Modules for Bioassay Analysis (IMBA) software code Version 4.0.9, the acute intake was determined 

to be 1.8 × 1010 pCi and the resulting committed equivalent dose was determined to be 1,200 mrem, 

which is consistent with the Pantex assigned dose of 1,180 mrem. Because the maximally exposed 

worker was the person who loosened the nut that started the leak and had his face nearest the 

source, it is unlikely that anyone at Clarksville received as great an intake. Therefore, the distribution 

should be considered an upper bound. Because all the workers who came in contact with the 

damaged weapon before the leak was stopped are not known, the intake would apply to any 

production operator, quality specialist, or quality control inspector at the site in 1962. 

One operator was sent back into the cell to stop the leak. The CATI indicated that this person was 

wearing protective clothing and respiratory protection appropriate for tritium (Scott air pack). 

Therefore, the above intake is considered an appropriate upper bound intake for that worker as well. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 25 of 69

The material handler who picked up this weapon from the military airport and transported it to a 

storage area was told the next day that he had also been contaminated (Bihl 2006a). Although his 

exposure conditions were different than those in the cell – mostly outdoors with some time near the 

weapon in the storage igloo − the intake for the maximally exposed worker at Pantex is judged to be a 

sufficient upper bound to apply to the specific material handler as well. Do not apply this intake to all 

material handlers. The material handler involved has been identified (Bihl 2006a). 

5.1.2 Postoperative Years (1966–1967)

Last, Gilmore, and Bronson (1998) performed radiological surveys of the 7740 tunnel complex in 

1997, which included the sampling of water in floor drains and sumps. Tritium was detected at 

2.4 pCi/mL in one sump under the central corridor. This concentration is at or maybe slightly above 

fallout levels in surface water in North America. Adjusting for physical decay to 1966 and making the 

gross overestimate of ingestion of 1 L/d at this concentration, the annual dose to any organ is less 

than 1 mrem. Intake by inhalation of water vapor around the sump would be much less, so tritium 

intake in 1966 and 1967 can be disregarded. 

5.2 URANIUM EXPOSURE 

DU oxide was known to be present, but apparently was not considered a significant enough internal 

hazard to warrant bioassay. No records were found indicating that bioassay was performed on 

workers to determine uptake of DU. However, according to Harold Rarrick, portable gloveboxes with 

HEPA-filtered exhausts were used to contain oxidized DU when weapons underwent inspections, 

maintenance, and refurbishment during the Sandia years (McConn 2006, Lamb Associates and 

Halliburton NUS 1996). The DU was cleaned from the nuclear components and deposited as solid 

waste on cleaning rags. Mr. Rarrick mentioned that the DU contamination came from spalling, which 

produced large unrespirable particles. Use of the glovebox indicates knowledge that DU might be in 

the environment and that containment was at least a good practice. Disassembly and maintenance 

activities at Clarksville Base during the MHSMC years were similar to those performed at the Pantex 

Plant and IAAP. 

Intakes of DU might have occurred during disassembly and maintenance of weapons containing 

oxidized DU. DU emits less radiation per gram than natural uranium. By weight, DU is essentially 

pure 238U. Isotopic abundances of 234U and 235U in DU can vary, but those isotopes generally 

contribute less than 10% of the alpha radioactivity. The dose reconstructor can use the isotopic 

abundance for DU in the IMBA computer program.

Concerning the inhalation absorption type, the Pantex Internal Dosimetry Technical Basis and Quality 

Assurance Document states that “the compounds of uranium at Pantex are pure metal or air-oxides; it 

is assumed that all forms encountered will exhibit class Y aerosol behavior” (BWXT Pantex 2001a, 

Section 13.2.1). Table 7.3 of that document indicates that 238U should be considered 20% class D 

and 80% class Y. Uranium contamination at Clarksville Base is assumed to be similar to that at the 

Pantex Plant. The most likely form of uranium at Clarksville would be very insoluble, associated with 

lung absorption type S and a gastrointestinal-tract-to-blood uptake factor, f1, of 0.002. However, 

uranium oxides can exist in many states, and it might be too simplistic to assume a pure absorption 

type when the chemical form is not known for certain. Dose reconstructors should assume either 

type M or type S to maximize the dose to the organ of concern. Exposure to type F uranium at 

Clarksville is not credible. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 26 of 69

5.2.1 Operation Years (1949–1965)

Experience at the Pantex Plant indicated that “about a half of a cup” (118 cm3) of oxidized DU was 

available for resuspension (ORAUT 2006a). The density of UO2 is 11 g/cm3; thus, the estimated 

mass of UO2 is approximately 1,300 g, of which about 1,140 g is DU [(238/270)(1,300g) = 1,140 g]. 

Interviews with people familiar with Clarksville Base never gave the impression that this much oxide 

was present, so using this value should provide an upper bound. No benefit is given for use of the 

glovebox that was mentioned in several interviews and in the site assessment (Lamb Associates and 

Halliburton NUS 1996). 

The airborne release fraction and respirable fraction were obtained from a DOE handbook using the 

scenario described as the free fall spill of UO2 powder from a height of 1 m (DOE 1994). The median 

value for the airborne release fraction was 0.00008, and the respirable fraction was 0.5. The 

assumed volume of air into which the contamination was suspended was 27 m3, a volume judged to 

represent the immediate work area, not the volume of the entire cell. The material was assumed to 

be quickly dispersed into that volume. The immediate air concentration was then: 

(1,140 g)(0.00008)(0.5)/27 m3 = 1.689 x 10-3 g/m3 (5-1)

This air concentration was reduced over time by ventilation. The concentration as a function of time is 

given by: 

C(t) = Cie

(-Qt/V) (5-2)

where Ci is the initial room concentration and C(t) is the concentration at any later time t, Q is the 

ventilation air flow rate, and V is the volume of the contaminated air (Caravanos ca.1990). The air 

exchange rate (a in Equation 5-3), is Q/V, so Equation 5-2 can be written as: 

C(t) = Cie

(-at)

(5-3)

When this is integrated over an 8-hr work shift, the concentration-time exposure time is: 

Concentration-time = Ci[1-e(-8a)]/a (5-4)

The air exchange rate is not known, but there was concern about control of the explosives, for both 

fire hazard and toxicology reasons. Ventilation standards are usually based on certain air flow rates 

per person of occupancy with additional considerations for reducing concentrations of hazardous 

materials. The 1989 standards range from 15 to 30 ft3/min per person for places like bars or dry 

cleaners on the high end to reception areas, hotel lobbies, and supermarkets on the low end 

(ASHRAE 1989). Typical occupancy during disassemblies was 4-5 people2

, so the air flow was 

probably about 150 ft3/min or 9,000 ft3/hr, or maybe more because of the special concerns for the 

explosives. At Pantex the volume of the Gravel Gerties was about 14,000 ft3 and volume of the 

various bays ranged from about 10,000 to 30,000 ft3 (BWXT 2001b). Based on this information, an air 

exchange rate of 0.5/hr was assumed. (Note: the recommended air exchange rate for living areas of 

a house is 0.35/hr (ASHRAE 1989), so an air exchange rate of 0.5/hr for an industrial setting handling 

toxic material seems favorable to claimants.) Therefore, the concentration time for each work day is: 

(0.001689 g/m3)(1-0.01832)/0.5 = 0.003316 g-hr/m3 (5-5)

 

2

 Personal communication, Jerry Martin, former manager of the Radiation Safety Department, 2006. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 27 of 69

Applying the 1.2 m3/hr breathing rate, the total intake was 3.99 × 10-3 g or 1,600 pCi. This applies to 

each weapon disassembled. 

The resulting initial air concentration is an order of magnitude greater than the 1959 uranium air 

concentration limit. Although it is unlikely that the amount of oxidized DU estimated above was 

present in all cases, especially in the early years, the assumption was made that the potential for 

intake of this quantity did exist throughout the covered period at Clarksville Base (1949 to 1965). Not 

all weapon programs exhibited oxidized DU components. According to the Pantex Internal Dosimetry 

TBD, fewer than a third of the weapons disassembled at the Pantex Plant had noticeable oxide and 

only two had significant contamination (ORAUT 2006a). Because NIOSH cannot publicly identify 

programs that had significant DU oxidation, the intake assumption is applied to all maintenance and 

disassembly work over the operational period. 

For dose reconstruction, it was assumed that during the Sandia years, which involved some training, 

inspection, and replacement of Po-Be initiators, on average 50 surveillance/disassembly operations 

might be conducted by a single worker per year with the potential for internal exposures to the 

quantity of DU estimated above. To account for an increased rate of disassembly and modification 

during the MHSMC years, it was assumed that any production operator might have been involved in 

100 operations per year with potential for significant contamination. The intake should be modeled as 

a chronic annual inhalation of 80,000 pCi of DU (219 pCi/d), as a constant upper bound for 1949 

through 1957, and twice that rate for 1958 through 1965. 

Ingestion of DU settling out of the air on drinks or foodstuffs or transfer from hands to cigarettes or 

food cannot be ruled out. Guidance for determining ingestion intakes from air concentration is 

provided in Estimation of Ingestion Intakes (NIOSH 2004). The estimated ingestion intake from air 

concentration is: 

ingestion daily intake = (0.2)(air concentration in pCi/m3) (5-6)

where the air concentration is intended to be the average concentration, not the initial concentration. 

Dividing the concentration-time value in Equation 5-5 by an 8-hr shift gives an average air 

concentration of 4.14 × 10-4 g/m3. Then: 

ingestion daily intake = (0.2)(4.14 × 10-4 g/m3)(4.02 × 105 pCi/g) = 33 pCi (5-7)

where 4.02 × 105 is the specific activity of DU. The material would have been insoluble (f1 = 0.002). 

As for inhalation intakes, the 33 pCi is per weapon. For 1949 to 1957, the daily intake from 

50 weapons per year would have been 4.6 pCi/d; for 1958 to 1965, the intake is 9.1 pCi/d. 

5.2.2 Postoperative Years (1966–1967)

Casual exposure to residual DU contamination after operations ceased is possible. According to Last, 

Gilmore, and Bronson (1998), “[t]he Defense Nuclear Security Agency conducted a closeout 

radiological survey in July 1970 to determine if Clarksville Base had any radioactive contamination. 

The study found elevated levels of indoor radon, but concluded that no radiological health hazards 

existed in Building 318 [numbered 7740 in the Last survey].” The surveys conducted for the Last, 

Gilmore, and Bronson study found no contamination above release criteria. For removable 

contamination of natural uranium, 235U, or 238U, the criterion was less than 1,000 dpm/100 cm2. 

Assuming a passive resuspension factor of 10-6/m, and given that: 

1,000 dpm/100 cm2 = 105 dpm/m2 (5-8)
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 28 of 69

the air concentration of uranium above a surface contaminated just at the release criterion would be: 

(105 dpm/m2)(10-6/m)/2.22 dpm/pCi = 4.5 × 10-2 pCi/m3 (5-9)

After weapons and nuclear material were removed from the site, it is unlikely that workers entered 

potentially contaminated areas more than a few hours a month for minor maintenance or general 

surveillance. An occupancy factor for exposure to airborne contamination of 0.02 was assumed. 

Therefore, a chronic intake of uranium if all surfaces were contaminated just at the release criterion 

would be: 

 intake = (4.5 × 10-2 pCi/m3)(1.2 m3/hr)(2,000 hr/yr)(0.02)/365 d/yr) = 0.0059 pCi/d (5-10)

Because all measurements were reported as simply less than the release criterion, the above 

calculation is analogous with a missed dose calculation; therefore, the intake distribution should be 

considered a triangular distribution with a minimum of zero, a mode of 0.0030 pCi/d, and a maximum 

of 0.0059 pCi/d. 

5.2.3 DU from Burning Grounds 

The 1996 site assessment states that during the years that Clarksville performed modifications and 

disassemblies: 

The high explosive portion of the weapon was a casting that contained paper and felt 

materials. Toluene was used to partially dissolve the high explosive to separate it from 

the paper and felt during the disassembly process. The solvent and explosives mixture 

was burned at a site outside of Clarksville Base. (Lamb Associates and Halliburton 

NUS Corporation 1996)

It is not known if MHSMC employees performed the burns; however, two interviews with claimants 

imply that MHSMC employees did perform the burns, and site operator employees performed the 

burns at IAAP, Pantex, and Medina (ORAUT 2005b, 2006a; Bihl 2006c). Therefore, it is assumed 

that this task was performed by MHSMC employees. 

Usually there was some DU contamination associated with the high explosives. Air sample 

measurements were made for a similar operation at Pantex (ORAUT 2006a). There is no information 

concerning the amount of high explosives or DU contamination burned at Clarksville; therefore, it was 

assumed that the air concentrations and intakes by the burning ground operators were represented 

by the Pantex operation. The assessment of the Pantex operation indicated that the highest 

exposure occurred during the cleanup of the contaminated ash. The estimated intake at Pantex was 

130 pCi/d of DU based on the 95th-percentile air concentration for the cleanup activities. It is 

assumed this intake rate applies as well to Clarksville workers involved in burning high explosives. 

The intake is an upper-bound estimate so the distribution is a constant. Because incomplete 

oxidation of the DU can occur during these operations, dose reconstructors should assume either 

absorption type M or S (ORAUT 2006a). The intake would apply only to 1958 through 1965 and only 

to workers who burned high explosives (not a common task). 

(Because the burn did not occur at Clarksville, no environmental dose to Clarksville workers was 

assigned.) 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 29 of 69

5.3 PLUTONIUM OR POLONIUM EXPOSURE 

Mr. Rarrick believes that during the Sandia years no significant spread of contamination of plutonium 

occurred based on reports from the Sandia safety inspectors from the various WSAs (Rarrick 2006; 

Lamb Associates and Halliburton NUS Corporation 1995). The plutonium was plated and nothing was 

purposely done to compromise the barrier between the plutonium and the workspace. During the 

process of removing the Po-Be initiators, the parts were smeared and counted for alpha 

contamination. In general, the alpha contamination was less than 100 dpm with a few leaks in the 

early days slightly exceeding 100 dpm, but not significantly (Rarrick 2006). Mr. Phillips indicated that 

at Medina and Pantex during the MHSMC years, the incoming nuclear component was promptly 

surveyed for alpha contamination and cleaned before being placed into storage or the cells (Bihl 

2006c). The procedure at Clarksville during the MHSMC years has not been discovered but was 

likely to be similar to the procedure at Medina and Pantex. An assumption was made that workers 

were possibly exposed for a short time to plutonium contamination on the internal parts during 

disassembly. Assuming 100 dpm alpha/100 cm2 (100 cm2 is the industry standard area for smears) 

and a resuspension factor for indoor work of 5 × 10-5/m (IAEA 1970), the air concentration is 

calculated as: 

100 dpm/100 cm2 = 1 dpm/cm2 (5-11)

 air concentration = (1 dpm/cm2)(104 cm2/m2)(5 × 10-5/m) = 0.5 dpm/m3 = 0.225 pCi/m3 (5-12)

The above calculation works for general room concentration from large contaminated surfaces or for 

small volumes of air immediately above small surfaces. Neither condition would have applied to the 

disassembly workers for the whole shift; however, contamination levels might have exceeded 

100 dpm on occasion. On balance, and to be favorable to claimants, the above air concentration was 

assumed to exist for all work hours. The daily inhalation intake was, therefore: 

 intake = (0.225 pCi/m3)(1.2 m3/hr breathing rate)(2,000 hr/yr)/365 d/yr = 1.5 pCi/d (5-13) 

Ingestion intakes resulting from contaminated surfaces and airborne contamination cannot be ruled 

out. For estimating ingestion resulting from contamination inside buildings, NIOSH recommends a 

daily ingestion of 0.2 times the air activity per cubic meter (NIOSH 2004). Therefore, the daily 

ingestion would have been: 

ingestion intake = (0.2)(0.225 pCi/m3) = 0.045 pCi/d (5-14)

These intakes should be applied to production operators and nuclear component inspectors for the 

operative years for in-flight insertable designs (1949 to 1958). Because the calculations ignore the 

prompt cleanup of components upon disassembly and the use of gloveboxes during cleaning (Lamb 

Associates and Halliburton NUS Corporation 1996, Martin 2006a), these intakes should be 

considered constant upper bounds. 

The alpha contamination could have been either plutonium or 210Po. The dose reconstructor should 

use whatever radionuclide is most favorable to the claimant. The plutonium mixture would have been 

weapons grade. The age since original separation from the 241Am would have varied, but an 

assumption of 10-yr-old mixture is reasonable. The plutonium would have been oxidized, so 

inhalation type S and an f1 of 1 × 10-5 should be used (ICRP 1994). 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 30 of 69

If the contamination is assumed to be 210Po, because the chemical form is classified, the dose 

reconstructor should use inhalation type F or M, whichever is favorable to the claimant. For polonium 

the only f1 listed is 0.10 (ICRP 1994). 

No plutonium or 210Po intakes should be assigned during the period of sealed pits or during the 

postoperative years. 

5.4 ASSESSING OCCUPATIONAL INTERNAL DOSE FROM ELEVATED RADON 

CONCENTRATIONS 

Uranium occurs naturally in virtually all soils, with average levels of about 1 ppm. Radium-226 is 

typically in secular equilibrium with 234U and decays to 222Rn with a half-life of 1,600 yr. Therefore, the 

noble gas 222Rn is continuously produced in soil where it can be trapped in the crystalline structure of 

minerals or released to the interstices between solid materials and eventually diffuse out of the soil 

and into the airspace. 

In general, structures that exhaust air to the environment without adequately engineered replacement 

air have higher indoor radon levels than structures that do not do this, and structures that have 

exposed soil (dirt floors, sumps) or exposed minerals (e.g., gravel) tend to have higher radon levels. 

Underground structures have a higher ratio of soil surface to building volume. All other factors being 

equal, an underground building would be likely to have a higher radon concentration than an aboveground building. 

5.4.1 Radon Concentrations

Radon concentrations were measured at Clarksville Base on several occasions from 1971 through 

1986. A listing of all radon measurements up to March 1984 is provided in a memorandum from the 

Chief of the Fort Campbell Preventive Medicine Service to the Deputy Post Commander (Vaeth 

1984). The radon levels in the underground tunnel building (7740) are listed in Table 5-1. 

Table 5-1. Radon concentrations in the 7740 Buildinga. 

Date sampled Agency performing sampling Sample results (pCi/L)

June 7–17, 1971 AEHA 152.48

April 5–6, 1976 USAPMS 14.6

43.4

November 20, 1981 USAPMS 21.4

33.4

a. From Last, Gilmore, and Bronson (1998). 

Another set of radon measurements was made by the U.S. Army Environmental Hygiene Agency 

(AEHA) in 1986 (AEHA 1986). A summary discussion of these measurements is provided in Relative 

Risk Site Evaluation for Buildings 7740 and 7741 Fort Campbell, Kentucky (Last, Gilmore, and 

Bronson 1998). Building 7740 was not sampled in the 1986 study; it was unoccupied and sealed at 

the time of the sampling. Neither how the sampling was conducted nor the time of day of the 

sampling was mentioned in the Vaeth memorandum. However, with the tunnel complex sealed and 

unventilated, it is probable that diurnal fluctuations were dampened. The three sets of sampling were 

conducted in different seasons (November, April, June). Table 5-2 lists results of sampling in other 

Clarksville structures. The AEHA sampling technique in 1986 was a 1-L grab sample of air obtained 

about 1 m above the floor with laboratory analysis of the sample at the Aberdeen Proving Ground. 

Table 5-3 lists statistical parameters from the radon concentration measurements. There is a distinct 

difference between concentrations in the tunnel complex and those in other structures expected to 

have high radon concentrations. The statistics for the tunnel complex are strongly influenced by the 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 31 of 69

Table 5-2. Radon concentrations in Clarksville structures. 

Type of structure Number Date sampled Sample results (pCi/L)

Igloo 7704 June 7–17, 1971 8.60a

November 20, 1981 3.90a

Igloo 7708 June 7–17, 1971 4.53a

October 30, 1981 3.20a

Igloo 7726 June 7–17, 1971 15.86a

October 30, 1981 9.70a

Igloo 7728 February 26-27, 1979 0.60, 0.50, 0.20, 0.10a

Above-ground, concrete ceiling, and 10-ft 

thick concrete walls

7724 September 29, 1981 3.30a

Hallway outside Gravel Gertie 7811 March 3, 1986 8.9 ± 0.3b

Inside Gravel Gertie 7811 March 3, 1986 10 ± 0.4b

a. From Vaeth (1984). 

b. From AEHA (1986). 

 Table 5-3. Statistical parameters of radon concentrations. 

Type of structure Parameter Concentration (pCi/L)

7740 tunnel complex Average 53

Standard deviation 57

Geometric mean 37

GSD 2.4 (unitless)

Igloos, Gravel Gertie, 7724 Bldg Average 5.3

Standard deviation 4.9

Geometric mean 2.5

GSD 5.2 (unitless)

measurement taken in 1971. However, even if that one result is ignored, the median concentration for 

the tunnel complex would be notably higher than the median for the igloos and Gravel Gertie. 

5.4.2 Dose from Radon-222 Progeny

Radon itself produces far less dose to the bronchial epithelium than its progeny. Because radon 

progeny measurements are more difficult to obtain, measurements of radon are often used as a 

surrogate for progeny measurements. Radon progeny concentrations are expressed as the quantity 

potential alpha energy concentration (PAEC), traditionally measured in working levels (WLs). One 

WL is equivalent to 100 pCi/L of radon in equilibrium with its short-lived decay products. Timeintegrated exposures to radon progeny are expressed in the quantity potential alpha energy exposure

(PAEE), which is traditionally measured in working level-months (WLMs) and defined as exposure to 

1 WL for 170 hr or any equivalent concentration and time product. 

The Clarksville measured radon concentrations were converted to equilibrium-equivalent 

concentrations by multiplying the radon concentration by the equilibrium factor F. A value of 0.4 is 

recommended by the ICRP (1981) and the United Nations Scientific Committee on the Effects of 

Atomic Radiation (UNSCEAR 1993) for homes and buildings. However, a higher equilibrium factor 

might be appropriate for poorly ventilated underground rooms or tunnels. Therefore, an equilibrium 

factor of 0.7 was applied to the igloos and the tunnel complex [e.g., see Table XVIII in the DOE 

Standard – Internal Dosimetry (DOE 2003)]. The equilibrium equivalent concentration was divided by 

100 pCi/L/WL to arrive at the PAEC. These operations were combined to create: 

PAEC = C × F/100 pCi/L/WL (5-15)
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 32 of 69

where C is the radon concentration in picocuries per liter and PAEC is in WLs. To determine WLM for 

input to IREP, WLs are multiplied by an occupancy time in terms of hours per month/170 and the 

number of months of exposure per year. In general, material handlers were in and out of igloos and 

similar storage areas for less than a couple of hours per workday, so an occupancy factor of 0.4 was 

used. It was assumed that inspectors and production operators were in the tunnel complex or Gravel 

Gertie full-time, so an occupancy factor of 1.0 was used. For other nonadministrative job categories 

(e.g., crafts, security), some time in the igloos or tunnel complex could not be ruled out, so an 

occupancy factor of 0.1 was used. From claimant interviews, it appears that even administrative 

personnel occasionally entered the tunnel complex or other storage sites; for instance, an accountant 

claimed to spend 1 week a year in the storage areas. For administrative personnel, an occupancy 

factor of 0.02 was assumed. For instance, the WLM for a production operator in the tunnel complex 

would be 

(37 pCi/L)(0.7)(1.0)(12 mo)/100 pCi/L/WL = 3.1 WLM/yr (5-16)

Table 5-4 lists WLMs for job categories. 

Table 5-4. Working level-months for Clarksville workers. 

Job description Assumed location Years

Assumed 

occupancy 

factor

Working level 

mo/yra 

(GSD)

Production operator, any inspector, 

quality control specialist

7740 tunnel complex 1949–1958 1.0 3.1 (3)

Production operator, any inspector, 

quality control specialist

Gravel Gertie 1959–

1965b

1.0 0.21 (5)

Material handler 7740 tunnel complex 1949–1958 0.4 1.2 (3)

Material handler Igloos, Gravel Gertie 1959–1965 0.4 0.084 (5)

Other nonadministrative 7740 tunnel complex 1949–1958 0.1 0.31 (3)

Other nonadministrative Igloos, Gravel Gertie 1959–1965 0.1 0.021 (5)

Administrative 7740 tunnel complex 1949–1958 0.02 0.062 (3)

Administrative Igloos, Gravel Gertie 1959–1965 0.02 0.0042 (5)

Facility surveillance, minor 

maintenance after shutdown

Igloos, Gravel Gertie 1966–1967 0.02 0.0042 (5)

a. Can prorate for partial year by multiplying by x/12 where x is the number of months exposed in a year. 

b. The change in dates approximately coincides with the construction and use of the Gravel Gertie. 

5.5 SUMMARY OF INTAKES 

Table 5-5 summarizes occupational intakes (not including environmental). 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 33 of 69

Table 5-5. Occupational intakes. 

Job category or 

task

Dose calculation parameters IREP input parameters

Period Material Mode

Absorption 

type pCi/d

Distribution 

type 1 2

Production 

operator, nuclear 

components 

inspector, quality 

control specialist

1954–1965 Tritium Chronic 

inhalation 

and skin 

absorption 

(model as 

injection) 

NA 1.3E7 Constant Dose

Production 

operator, nuclear 

components 

inspector, quality 

control specialist

March 1962 Tritium Acute 

inhalation 

and skin 

absorption 

(model as 

injection)

NA 1.8E10 pCi Constant Dose

Production 

operator, nuclear 

components 

inspector, quality 

control specialist

1949–1958 DU Chronic 

inhalation,

Ingestion

M, S

Insoluble

2.2E2

4.6

Constant

Constant

Dose

Dose

Production 

operator, nuclear 

components 

inspector, quality 

control specialist

1959–1965 DU Chronic 

inhalation,

ingestion

M, S

Insoluble 

4.4E2

9.1

Constant 

Constant

Dose

Dose

Production 

operator, nuclear 

components 

inspector, quality 

control specialist

1949-58 Pu 

weapons 

grade, 10 

yr; or Po210

Chronic 

inhalation

ingestion

Pu: S

Po: F or M

Pu: 

insoluble

1.5

0.045

Constant

Constant

Dose

Dose

Production 

operator, any 

inspector, quality 

control specialist

1949–1958 Radon Chronic 

inhalation 

NA NA Lognormal 3.1 WLM/ 

12 mo

3

Production 

operator, any

inspector, quality 

control specialist

1959–1965 Radon Chronic 

inhalation

NA NA Lognormal 0.21 

WLM/ 

12 mo

5

Material handler 1949–1958 Radon Chronic 

inhalation

NA NA Lognormal 1.2 WLM/ 

12 mo

3

Material handler 1959–1965 Radon Chronic 

inhalation

NA NA Lognormal 0.084 5

Individual material 

handler involved in 

1962 damaged 

weapon

March 1962 Tritium Acute 

inhalation 

and skin 

absorption

NA 1.8E10 pCi Constant Dose 

Burning of high 

explosives

1959-1965 DU Chronic M or S 130 Constant Dose

Other 

nonadministrative

1949–1958 Radon Chronic 

inhalation

NA NA Lognormal 0.31 3

Other 

nonadministrative

1959–1965 Radon Chronic 

inhalation

NA NA Lognormal 0.021 5

Administrative 1949–1958 Radon Chronic 

inhalation

NA NA Lognormal 0.062 3

Administrative 1959–1965 Radon Chronic 

inhalation

NA NA Lognormal 0.0042 5

Surveillance after 

shutdown

1966-67 DU Chronic 

inhalation

M, S 0, 3.0E-3, 

5.9E-3

Triangular (a) (a)

Surveillance after 

shutdown

1966-67 Radon Chronic 

inhalation

NA NA Lognormal 0.0042 5

a. Parameter 1 is zero dose; parameter 2 is the mode dose based on the 0.003-pCi/d intake; parameter 3 is the maximum dose based on 

the 0.0059-pCi/d intake. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 34 of 69

6.0 OCCUPATIONAL EXTERNAL DOSE

6.1 INTRODUCTION 

Workers at Clarksville Base were employed by either SNL or MHSMC, but not both. A few SNL 

workers (fewer than 20) performed the maintenance operations involving nuclear components 

between July 1949 and September 1958 (McConn 2006). A similar number of MHSMC workers 

performed inspections and modifications on nuclear weapons from 1959 to September 1965 

(Mitchell 2003). 

Work activities at Clarksville Base undoubtedly varied over time. Analysis of historical information 

showed that maintenance activities at Clarksville began in 1949, which corresponds to the first record 

of personnel monitoring (McConn 2006). The nature of the radiation fields a Clarksville worker could 

have encountered depends on the type of weapon on which work occurred. Nuclear weapons 

components emit alpha, beta, X- and gamma rays, and neutrons; however, doses to workers depend 

strongly on the configuration (i.e., material and shielding) of the source of radiation and the work 

performed (BWXT Pantex 2001c). 

There were three major groupings of workers at Clarksville. AEC employees served primarily in 

oversight positions and generally did not perform hands-on work. Sandia, and later MHSMC, 

employees were responsible for maintenance of the weapons and would be expected to have the 

highest doses. Military personnel would be expected to receive doses that were less than the Sandia 

or MHSMC employees; the military primarily loaded the weapons onto airplanes and performed any 

in-flight operations that were necessary. 

6.2 EXTERNAL RADIATION DOSIMETERS AND RECORDS 

External dosimetry records for Clarksville Base are sparse and the connection between the dose 

record and the worker might be missing. Statistical analysis of doses received at Clarksville cannot 

be performed with the few records found to date. 

Dosimetry data for nine Sandia personnel have been found. These data are analyzed in 

Section 6.3.1. Recorded doses have been found for other workers but they do not contain sufficient 

information to determine which individuals performed a particular task. 

The Pantex Plant maintains a limited database for MHSMC workers at Clarksville Base, containing 

weekly dose information for a few workers from October 1960 to 1965. Although MHSMC began its 

management of Clarksville Base in early 1959, no dosimetry records were found for 1959 through 

September 1960. Annual dose reports supplied to AEC for 1960 to 1965 included individual wholebody dose equivalent from photons and neutrons. At Clarksville, dosimeters were issued to only a 

few workers who had direct contact with nuclear weapon components. 

The first dosimetry records found for MHSMC workers are dated October 1960. Commercial film 

badge service was supplied by Tracerlab from 1960 to 1965. During this period, a small number of 

workers (from 3 to 27) were monitored with mostly negative results (less than the minimum recordable 

dose). Only about 40 positive (nonzero) results were reported of approximately 5,900 individual 

weekly film badges. The highest annual dose to a worker from film data was less than 200 mrem in a 

year. Eastman Kodak nuclear track emulsion, type A (NTA) film was probably added for neutron 

dosimetry in January 1960; the exact date that NTA film was used for neutron dosimetry has not been 

determined. Mr. Rarrick thought the first use of neutron dosimetry might have been 1959. Pantex 

Plant dosimetry data show that NTA film was in use by January 1960 (McConn 2006). Because 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 35 of 69

Clarksville and Pantex were operated by the same contractor, it is reasonable to assume that identical 

dosimetry service was provided for both sites. Only one positive (12-mrem) result was reported out of 

nearly 3,900 individual weekly film badges (Tracerlab 1965). 

6.2.1 Historical Administrative Practices

Clarksville Base started monitoring workers for radiation exposure in July 1949. Dosimeters used at 

that time to measure worker radiation doses were provided by SNL. Table 6-1 summarizes the 

monitoring technique and exchange frequency. The SNL minimum recordable dose (MRD) for 

nonpenetrating skin and penetrating whole-body dose was probably similar to the minimum detection 

levels (MDLs) determined by others (Wilson 1960, 1987; NIOSH 1993; NRC 1989; and Wilson et al. 

1990). Dosimeters were supplied by Tracerlab from October 1960 to 1965 for MHSMC workers. No 

dosimetry records were found for 1959 through September 1960. The MRDs reported by Tracerlab 

were 30 and 10 mrem, respectively (Tracerlab 1965). Actual MDLs are typically higher because of 

additional uncertainty in field use and the use of dose recording thresholds. Table 6-1 lists 

reasonable MDLs for most applications for film dosimeters based on Wilson (1960, 1987), NIOSH 

(1993), NRC (1989), and Wilson et al. (1990). MRDs varied with time and processor, as listed in 

Table 6-1.

Table 6-1. Dosimeter type, period of use, exchange frequency, MRD, and MDL (Martin 2005). 

Dosimeter type–provider Period

Exchange 

frequency

MRD (mrem) MDL (mrem)

Skin

β/γ 

deep Neutron Skin Deep Neutron

βγ film–SNL 7/1949–1958 Monthly 40 40 40a40a

βγ film–Tracerlab 10/1960–1962 Weekly

b

30c10c40a40a

βγ film and NTA film–Tracerlab 1962–1965 Weeklyb30

c

10

c

15b

40a40a(d)

a. Estimated MDL typical of film dosimeter capabilities (Wilson 1960, 1987; NIOSH 1993; NRC 1989; Wilson et al. 1990). 

b. The weekly exchange frequency was established from dosimetry reports. No dosimetry reports for July 1949 through 

1958 or 1959 though October 1960 have been found. 

c. Based on minimum doses recorded on dosimetry reports (Tracerlab 1965). 

d. For years of NTA film use, between 1960 and 1965, the reconstructed neutron dose is calculated using the adjusted 

photon dose and a neutron-to-photon dose ratio. 

The routine practice at Clarksville Base appears to have required assigning dosimeters to personnel 

designated as radiation workers who could receive an external radiation dose greater than 10% of the 

Radiation Protection Guidelines in effect. Dosimeters were exchanged on a routine schedule. 

However, during the 1960–1965 period when Tracerlab provided film badges, individual worker 

names were not recorded with specific film badge numbers. In addition, dose components appear to 

be missing for some workers based on such designations as blanks or “damaged film” in records. 

These missing components can be reconstructed from other recorded dosimeter data by using 

recommended methods described later in this Site Profile. 

6.2.2 Dosimetry Technology

SNL radiation workers were monitored by film badges provided by SNL. Initially the “film badge” 

consisted of a piece of dental X-ray film in a plastic pouch with a pin for fastening to clothing. A lead 

filter was later added to the plastic pouch. The Oak Ridge metal film badge holder with three filters 

was used from 1957 through 1958. NTA film for neutron dosimetry was added in 1959 or 1960. 

Results from film badges were reported on “cardex” dosimetry records. However, records from these 

dosimeters have not been found in the SNL archives (McConn 2006). Mr. Rarrick stated that the 

maximum reported radiation dose was as high as 1 rem/yr from 1949 to 1952 (McConn 2006). After 

1952, maximum radiation doses were about 100 mrem/yr, according to Mr. Rarrick.
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 36 of 69

The first commercial dosimeter used by MHSMC at Clarksville Base was a two-element film badge 

supplied by Tracerlab for measuring beta, X-ray, and gamma exposures (Tracerlab 1965). Beginning 

in July 1962, Clarksville used a multielement film badge that incorporated NTA film to measure beta, 

X-rays, gamma rays, and fast neutrons (Tracerlab 1965). 

6.2.2.1 Beta/Photon Dosimeters 

Figure 6-1 shows the response of a film badge to photon radiation of different energies; it also shows 

the Hp(10) response. The figure shows two responses for film badges: one for a sensitive DuPont 

502 emulsion in a two-element badge (Pardue, Goldstein, and Wollan 1944), and one for a sensitive 

DuPont 555 emulsion in the multielement badge (Thornton, Davis, and Gupton 1961). The response 

of the sensitive Eastman Type 2 film in a multielement film badge is similar to that of the sensitive 

DuPont 555 emulsion. The film badges show an over-response at photon energies around 100 keV, 

due primarily to relatively (compared to tissue) high atomic numbers (Z) [silver (47) and bromine (35)] 

in the film emulsions. The film badges under-respond to lower energy photons, but the relative 

response of the two-element film badge to 60-keV photons from 241Am is nearly unity. The 

multielement film badge typically over-responds to 60-keV photons. 

10-210-110 |0

Photon Energy (MeV)

0.1

1

10 Relative Response

Hp(10)

Two-Element Film Badge

Multi-Element Film Badge

Figure 6-1. Comparison of Hp(10) for photons with energy 

responses for sensitive DuPont 502 emulsion in two-element film 

badge (Pardue, Goldstein, and Wollan 1944) and sensitive 

DuPont 555 emulsion in multielement film badge (Thornton, Davis, 

and Gupton 1961). 

6.2.2.2 Neutron Dosimeters 

The response of film dosimeters to neutron radiation was not good. NTA film was added to the holder 

used for the Clarksville beta/gamma dosimeter in 1958 and from July 1962 through 1965 (Martin 

2005). In general, the response of the NTA film decreases with decreasing neutron energies greater 

than a minimum threshold energy for laboratory studies, estimated to be about 500 keV (IAEA 1990; 

ORAUT 2006b). The minimum threshold energy for routine use in Clarksville mixed photon and 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 37 of 69

neutron radiation fields is probably about 1 MeV. Results reported at the first AEC Neutron Dosimetry 

Workshop indicated that laboratory dose measurements made with NTA film were about one-half to 

one-fourth of those measured with other methods, including the neutron TLD (Vallario, Hankins, and 

Unruh 1969). See Table 6-7 in Section 6.3 for information concerning methods to obtain estimates of 

neutron dose. The results of the NTA film should not be used for dose calculations because NTA is 

too unreliable. 

6.2.3 Dosimeter Calibration Procedures

6.2.3.1 Beta/Photon Dosimeters 

Clarksville film badges were originally calibrated with 226Ra sources, with exposure measured by 

Victoreen R chambers. Deliberately irradiated film badges were sent periodically to SNL and 

Tracerlab, and reported doses were compared with measured doses for calibration (McConn 2006). 

6.2.3.2 Neutron Dosimeters 

An account of the historical aspects of the calibration of Clarksville neutron dosimeters is not 

available. 

6.2.4 Workplace Radiation Fields

The main workplace radiation fields at Clarksville Base arose from the handling of nuclear weapon 

components containing plutonium, HEU, Po-Be initiators, and DU. The highest dose rates were 

encountered from the handling of bare pits and Po-Be initiators. The nuclides in the sealed nuclear 

weapon component pits emit alpha, beta, X-, gamma, and neutron radiation. From an external 

dosimetry perspective, the radiations of concern are beta particles, photons (X- and gamma rays), 

and neutrons. Radiation exposure to workers depends significantly on processes used in the 

preparation, design, and construction of the weapons. 

With few exceptions, the following sections show that, for external dose reconstruction purposes, all 

beta radiation fields are greater than 15 keV, all photon radiation fields are between 30 and 250 keV, 

and all neutron fields are between 0.1 and 2 MeV. Presuming that 100% of the radiation fields are 

within these ranges is a simplifying, conservative assumption that is generally favorable to claimants. 

Table 6-2 summarizes the radiation energy categories. 

6.2.4.1 Depleted Uranium 

Clarksville workers handled DU (primarily 238U) during assembly and disassembly of weapon 

components and during maintenance operations. An important progeny nuclide for potential worker 

exposure in 238U decay is 234mPa with a half-life of 24 d. In a few months after purification, DU 

components have 234mPa activities nearly equal to that of 238U. Protactinium-234m emits beta 

radiation 98.6% of the time when it changes to its ground state with a maximum energy of 2.28 MeV 

and an average energy of 0.825 MeV (Shleien, Slayback, and Birky 1998; ICRP 1973). An additional 

source of exposure in the Clarksville workplace was from bremsstrahlung produced in high-Z

materials from interactions with higher energy beta particles. Beta particles emitted by 234mPa excite 

both bremsstrahlung and characteristic X-rays in DU or 238U.

Beta radiation from DU could contribute to extremity and skin dose to workers unless precautions 

were taken to protect workers from the radiation. Protective clothing and gloves provide a protection 

factor of 2 or more, depending on the thickness. A bare slab source of natural uranium contributes an 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 38 of 69

Table 6-2. Beta, photon, and neutron radiation energies and percentages for Clarksville facilities. 

Process/

buildings Description

Operations

period

Radioactive

material

Radiation

type

Energy 

selection

Percent

(notes)

Bays

Cells

Tunnel complex

Assembly/ disassembly 

of nuclear weapons

1949–1965 DU Beta >15 keV 100a

Photons 30-250 keV 100b

1958–1965 Tritium Beta <15 keV 100c

1949–1965 Plutonium, HEU Photons 30-250 keV 100

Neutrons 0.1-2 MeV 100d

Casual surveillance, 

minor maintenance

1966–1967 DU Beta >15 keV 100

1966–1967 DU Photons 30-250 keV 100

Igloos Staging of weapons 

and plutonium pits

1949–1965 Plutonium, HEU Photons 30-250 keV 100

Neutrons 0.1-2 MeV 100d

Casual surveillance, 

minor maintenance

1966–1967 DU Beta >15 keV 100

1966–1967 DU Photons 30-250 keV 100

Transportation Movement of weapons 1949–1965 DU, HEU, 

plutonium

Photons 30-250 keV 100b

Neutrons 0.1-2 MeV 100d

Warehouse Packaging 

Components

1949–1965 Weapon 

components

Beta >15 keV 100a

Photons 30-250 keV 100b

Neutrons 0.1-2 MeV 100d

1954–1965 Tritium Beta <15 keV 100c

a. Workplace beta radiation has energy greater than 15 keV. 

b. Most photons from DU have energies greater than 30 keV; some have energies greater than 250 keV. If shielding 

materials are present, fewer photons are in the categories less than 30 keV or greater than 250 keV. The simplifying, 

conservative assumption that 100% of the photons from DU are between 30 and 250 keV is recommended as generally 

favorable to claimants. 

c. Beta particles from tritium are classified in the less-than-15-keV category. 

d. The energy of neutrons in the workplace is predominately in one of two ranges: Between 0.1 and 2 MeV or between 2 

and 20 MeV. In some cases, with significant moderating materials, some neutrons are less than 0.1 MeV. However, 

the simplifying, conservative assumption that 100% of the neutrons are between 0.1 and 2 MeV is recommended as 

generally favorable to claimants. 

Hp(0.07) dose of approximately 230 mrad/hr at the surface compared to an Hp(10) dose at 1 ft of 

approximately 2 mrad/hr (ORAUT 2005c). However, significant beta exposures to Clarksville workers 

were rarely detected by film badges, based on a review of shallow and deep dosimetry data. 

6.2.4.2 Photon Radiation 

Photon radiation in the workplace could have been readily measured at Clarksville Base with available 

dosimeter technology during all years of operation. It is assumed that all photons at Clarksville are 

within the 30- to 250-keV range, similar to data from the Pantex Plant. 

6.2.4.3 Neutron Radiation 

The in-flight-insertable design of nuclear weapons required a neutron initiator source. The first 

initiator sources were 210Po mixed with beryllium (McConn 2006). The average energy of the 

neutrons was 4.2 MeV (Shleien, Slayback, and Birky 1998). The ratio of neutron dose to photon dose 

produced by the α/n reaction was approximately 4 (Shleien, Slayback, and Birky 1998). The energies 

of the photons and neutrons were 0.8 and 4.45 MeV, respectively (Shleien and Terpilak 1984). 

Unfortunately, the half-life of 210Po is only 138.4 d (Shleien, Slayback, and Birky 1998), so the initiator 

sources had to be exchanged frequently (McConn 2006). 

Until 1957, the primary radiological task was periodically changing the Po-Be initiators. Between 1954 

and 1956, Po-Be initiators were gradually replaced with a newer type of sealed neutron generator that 

did not require routine replacement. Maintenance activities were reduced to annual disassembly of 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 39 of 69

capsules to verify the integrity of fissile materials, and radiation exposures to workers were reduced 

(Lamb Associates and Halliburton NUS 1996). 

6.2.5 Dosimeter Response to Radiation Fields

6.2.5.1 Beta/Photon Film Dosimeter Response 

The dosimeters used after 1957 contained an open window with little filtration, a lower energy window 

for allowing beta particles and lower energy photons to enter a film area with a plastic filter, and a film 

area with a metal (usually aluminum) filter. The open window enabled measurement of beta particles 

and lower energy photons. The plastic filter enabled measurement of intermediate energy photons 

and the metal filter enabled measurement of higher energy photons (1-cm depth). 

Tracerlab provided commercial dosimetry services from October 1960 through 1965. The AEC tested 

film badges provided by Tracerlab with exposures to 40-, 70-, and 210-keV X-rays and 60Co gamma 

rays, and mixed-energy exposures of all four radiations (AEC 1955). The film badges generally 

responded well “with a tendency to interpret most exposures too high.” The over-response (in the 

100- to 200-keV region) tended to yield conservatively high results. 

The film badge dosimetry reports provided by Tracerlab are less than adequate because no individual 

names are recorded in relation to specific film badge numbers. Radiation doses to workers were 

probably low during this period and many positive doses less than the minimum recordable dose 

would have been recorded as zero. The weekly film badge exchange frequency increased the 

probability that low doses were truncated to zero. 

6.2.5.2 Neutron Dosimeter Response 

The neutron doses of record at Clarksville are unreliable, and dose reconstructors should not use 

them.

6.2.5.3 Neutron-to-Photon Dose Ratios 

Neutron-to-photon dose ratios were calculated from Pantex post-1993 dosimeter data because the 

work performed at Clarksville Base and the Pantex Plant were similar. The Pantex data were 

analyzed by Strom (2004), and neutron-to-photon dose ratios were determined where the neutron and 

photon doses were greater than 50 mrem/yr. The median ratio is 0.7915 (rounded to 0.8) and the 

95th-percentile ratio from this distribution is 1.603 (rounded to 1.6). These data represent radiation 

workers who were exposed to photons and neutrons emitted from nuclear weapon components, 

primarily bare pits. 

Although the annual neutron-to-photon dose ratios have varied over the decades, the earlier annual 

neutron-to-photon dose ratio should be bounded by the 95th-percentile value of 1.6 derived from the 

analysis of data from dosimeters accredited by the DOE Laboratory Accreditation Program. Applying 

an annual neutron-to-photon dose ratio of 1.6 provides a method for reconstructing Clarksville worker 

neutron doses that is favorable to claimants. 

Clarksville radiation workers accumulated photon doses from a variety of workplace sources, 

including full weapon assemblies, partially shielded pits, and bare pits. During the 1949–1956 period 

when Po-Be initiators were exchanged, the maximum neutron-to-photon dose ratio was 4 (Shleien, 

Slayback, and Birky 1998); use of this ratio is recommended. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 40 of 69

Workplace records and measurements have shown that most neutron doses after 1957 were received 

during the handling of bare pits. A neutron-to-photon dose ratio of 1.6 should be applied to this 

period. 

6.2.5.4 Neutron Dose Weighting Factor 

The recommended neutron-to-photon ratios for Clarksville were based on Pantex dosimeter readings. 

At Pantex, thermoluminescent neutron dosimeters were calibrated with measurements based on 

fluence-to-dose conversion factors and quality factors similar to those from ICRP Publication 21 

(ICRP 1973) and NCRP Report 38 (NCRP 1971). It is necessary to adjust the neutron dose to 

account for the change in neutron quality factors between historic and current scientific guidance, as 

discussed in NIOSH (2002). Table 6-3, from ORAUT (2006c), lists the correction factor to use. 

Table 6-3. Neutron dose energies, percentages, and associated ICRP (1991) correction factors. 

Process Description

Neutron 

energy

(MeV)

Default 

dose

fractiona 

(%)

ICRP (1991)/ 

NCRP (1971)

correction 

factor

Nuclear weapons 

component assembly

Neutron exposure associated with weapons 

assembly and disassembly activities

0.1-2 100 1.91

a. From Table 6-2; assuming all neutron energies are between 0.1 and 2 MeV is favorable to claimants. 

6.3 RECOMMENDATIONS FOR CLARKSVILLE WORKER EXTERNAL DOSE 

RECONSTRUCTION – OPERATIVE YEARS 

Dose reconstruction for Clarksville workers is based on the foregoing information, which requires 

assessment of dose to be added to the assumed photon dose from three primary causes: 

 Adjustments to assumed photon dose for dosimeter uncertainty 

 Calculated neutron dose using a neutron-to-photon dose ratio 

 Multiplication of the calculated neutron dose by an ICRP (1991) neutron weighting factor 

adjustment of 1.91 for neutron energies between 0.1 and 2 MeV 

6.3.1 Unmonitored External Dose

At Clarksville Base, the concept of “unmonitored worker” will have to be expanded to include 

“monitored but records not found.” Few dosimetry records have been found for Clarksville; those that 

have been found do not always identify the person receiving the radiation dose. Therefore, it was 

necessary to estimate the radiation doses that Clarksville workers might have received. To perform 

these estimates, four exposure groups were identified (Table 6-4). Exposure Group 1 consists of 

individuals who worked with nuclear devices on a daily basis; these were considered full-time 

radiation workers who received the highest doses. Exposure Group 2 consists of individuals who 

routinely entered radiation work areas but were not in close contact with nuclear devices or were not 

exposed full time; they were assumed to have received half of the dose received by Group 1. 

Exposure Group 3 consists of individuals who were only occasionally exposed and were not in close 

contact with nuclear components. They were assumed to have received one-quarter of the dose 

received by Group 1. Exposure Group 4 consists of individuals who did not enter radiation areas; 

they were assumed to have received only environmental dose. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 41 of 69

Table 6-4. Worker job categories and exposure groups. 

Exposure 

group Conditions Group members

1 Extensive work with pits; full-time exposure 

(2,000 hr/yr)

Production operator, operator, operator trainee, 

nuclear inspector, nuclear quality control 

inspector, nuclear quality control specialist

2 Entered radiation areas but did not handle 

pits; exposure equivalent to 500 hr/yr

Material handler, warehouseman, safety/ 

security inspector, fireman, inspectors not 

associated with nuclear components

3 Infrequent entry into radiation areas; 

exposure equivalent to 200 hr/yr

All job categories not explicitly listed in this table

4 Did not enter radiation areas; exposure from 

environmental sources only

Bus driver, grounds laborer, power plant 

operator, sewage disposal operator

Operations at Clarksville Base between 1959 and 1965 were similar to those at the Pantex Plant, and 

MHSMC operated both facilities. Therefore, statistical information from the Pantex External 

Dosimetry TBD (ORAUT 2006c) was used to provide guidance for unmonitored workers at Clarksville 

for the MHSMC years. Table 6-5 summarizes the respective lognormal probability statistical 

parameters for the period from 1952 to 1965 for Pantex annual dose results that are equal to or 

exceed a gamma dose of 50 mrem. The statistics in Table 6-5 are based only on nonzero dose 

results; thus, they represent measured annual doses. What is not known is the number of zero badge 

readings that were included in the reported annual doses. The dose data reported at Pantex were 

reanalyzed with zero dose readings replaced by MDL/2 for the monthly period (see Table 6-6). For 

1960 and later, the assumed photon dose received by Exposure Group 1 (Table 6-7) was equal to the 

median photon dose for the Pantex Plant for the year the worker was employed at Clarksville Base, 

where the median is assumed to be the greater of the measured 50th-percentile dose or the 50thpercentile dose including potential missed dose. The year 1960 was chosen because dose data from 

only four Pantex workers were available for 1959, and this population size was too small to form the 

basis of dose estimates for Clarksville Base. For those same years, assumed Exposure Group 2 

photon doses were one-half of the median photon dose received at the Pantex Plant for the year the 

worker was employed at Clarksville Base (Table 6-8); and one quarter of the median doses were 

assumed to apply to Exposure Group 3 (Table 6-9). 

Table 6-5. Pantex worker photon dose statistics. 

Year

Annual recorded photon dose dataaLognormal fit

No. of workers reported 

photon dose >50 mrem

Dose (mrem) Dose (mrem)

Mean Maximum Median 95% GSD

1952–58 (b)

1959 4 36.3 40 36.0 45 1.15

1960 8 69.4 170 58.0 160 1.86

1961 33 55.7 190 50.1 103 1.55

1962 58 55.5 210 50.1 101 1.53

1963 186 65.7 513 49.6 141 1.88

1964 581 120.0 1,820 74.9 306 2.35

1965 380 101.0 2,950 64.3 231 2.18

a. Individual dosimeter records analyzed only if photon dose was equal to or greater than 

50 mrem. 

b. All recorded doses were less than 50 mrem. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 42 of 69

Table 6-6. Pantex worker photon statistics for all dosimeters. 

Year

No. of

dosimetersa

Monthly and annual photon 

dose data (mrem)

Monthly Annual

50th 95th 50th 95th

1952-58 227 20 20 240 240

1959 246 86.7 86.7 1,040 1,040

1960 220 86.7 86.7 1,040 1,040

1961 614 20 35 240 420

1962 585 20 45 240 540

1963 919 20 60 240 720

1964 2,653 20 140 240 1680

1965 3,448 40 60 480 720

a. The dosimetry data contained no identifiers, so it was not possible to 

determine how many workers were represented. 

Table 6-7. Dose recommendations for Clarksville Group 1 workers. 

Period Dose type Records Dose if no information

7/1949–1956 Photon None 1,040 mrem/yr (constant upper bound)

1957–1960 Photon None 1,040 mrem/yr (constant upper bound)

1961–1965 Photon Missing Median from Table 6-6, with GSD from Table 6-5

7/1949–1956 Neutron Any Neutron dose = 4 × photon dose

1957–1965 Neutron Any Neutron dose = 1.6 × photon dose

7/1949–1965 Neutron Any Multiply assigned neutron dose by 1.91a

a. ICRP (1991) weighting adjustments 

Table 6-8. Dose recommendations for Clarksville Group 2 workers. 

Period Dose type Records Dose if no information

7/1949–1956 Photon None 520 mrem/yr (constant upper bound)

1957–1960 Photon None 520 mrem/yr (constant upper bound)

1961–1965 Photon Missing 1⁄2 median from Table 6-6, with GSD from Table 6-5

7/1949–1956 Neutron Any Neutron dose = 4 × photon dose

1957–1965 Neutron Any Neutron dose = 1.6 × photon dose

7/1949–1965 Neutron Any Multiply assigned neutron dose by 1.91a

a. ICRP (1991) weighting adjustments 

Table 6-9. Dose recommendations for Clarksville Group 3 workers. 

Period Dose type Records Dose if no information

7/1949–1956 Photon None 260 mrem/yr (constant upper bound)

1957–1960 Photon None 260 mrem/yr (constant upper bound)

1961–1965 Photon None 1/4 median from Table 6-6, with GSD from Table 6-5

7/1949–1956 Neutron Any Neutron dose = 4 × photon dose

1957–1965 Neutron Any Neutron dose = 1.6 × photon dose

7/1949–1965 Neutron Any Multiply assigned neutron dose by 1.91a

a. ICRP (1991) weighting adjustments 

For the Sandia years, 1949 to 1958, workers handled components rather than performing intimate 

handling of pits, so doses would be expected to be smaller than those for MHSMC operations. This is 

mostly consistent with Mr. Rarrick’s recollections in that he indicated that maximum doses of 1 rem 

occurred up to 1952 and that all doses were less than 100 mrem after that (McConn 2006). A small 

number of external dose data have been found for 1949 through 1957. Of those dose records, 

28 worker-years of data have been found for nine Sandia workers who did surveillance on nuclear 

capsules. There were no results for 1949, only two for 1950, and only one each for 1956 and 1957. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 43 of 69

The data are in the form of annual photon doses to the body and wrist. Dose units were not indicated 

on the data sheets but were assumed to be rem per year. Because there are insufficient data to 

perform a year-by-year analysis of the doses, the data for body and wrist were analyzed for all years. 

For the analysis, all zero dose values were replaced by 5 mrem, which was one-half of the lowest 

recorded dose. 

Figure 6-2 shows the results for the body dose data. The data indicate that the 50th-percentile dose 

is 50 mrem and the 95th-percentile dose is 680 mrem. For the wrist, the measured doses are slightly 

smaller with a 50th-percentile dose of 50 mrem and a 95th-percentile dose of 560 mrem. The only 

item of note is that, for one case, the maximum body dose of 1,510 mrem exceeds the maximum dose 

recalled by Mr. Rarrick (McConn 2006). The data do not indicate how many dosimeter readings were 

zero, and this analysis does not include consideration of the number of exchange periods. However, 

the data do provide a measure of confidence that the Pantex and Clarksville doses for comparable 

years are compatible. Therefore, the highest 95th-percentile annual dose from Table 6-5 or 6-6 for 

1959 to 1960 was judged to apply to Clarksville Base for the Sandia years, 1949 to 1958. This dose, 

1,040 mrem, which is favorable to claimants, was assumed to be an upper bound and exceeds the 

95th-percentile annual dose shown in Figure 6-2. Based on the above sources of information and 

assumptions, guidance on assigning photon doses to unmonitored or records-missing workers are 

listed in Tables 6-7, 6-8, and 6-9 for Groups 1, 2, and 3, respectively. 

1 5 10 50 100 500 1000

.01

.02

.05

.1

.2

.3

.4

.5

.6

.7

.8

.9

.95

.98

.99

Lognormal Probability Plot

Point plotting method = nelson

Clarksville Annual Body Doses (mrem) Lognormal Probabilities

50

680.5

Figure 6-2. Log probability plot of annual doses received by monitored workers at 

Clarksville Base, 1949 to 1958. 

For all workers, neutron doses should be assigned as indicated in Tables 6-7, 6-8, or 6-9. Figure 6-2 

shows a plot of all data available for workers at Pantex for 1960; zero doses have been replaced by 

the MDL/2 dose. The data in Figure 6-3 show that worker dosimetry data do not follow a lognormal 

distribution; they also do not follow a normal distribution. For dose reconstruction, the GSD listed in 

Table 6-5 can be used. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 44 of 69

2 2.5 3 3.5 4 4.5 5

0.001

0.003

0.01 

0.02 

0.05 

0.10 

0.25 

0.50 

0.75 

0.90 

0.95 

0.98 

0.99 

0.997

0.999

ln (Monthly dose)

Probability

Pantex: All Doses - 1960

Figure 6-3. Log probability plot of Pantex dosimetry data that includes missed 

doses. 

For all years, it appears that the 50th-percentile dose is equal to the MDL/2 times the number of 

exchange periods. In many years, the 95th-percentile dose is equal to the 50th-percentile dose. The 

dose reconstructor should use the dose data listed in Tables 6-7, 6-8, and 6-9. 

6.3.2 Missed External Dose for Monitored Workers

If external dose data are found in a worker’s file, the dose reconstructor should assign a missed 

photon dose based on the MDL/2 method and the number of exchange periods (NIOSH 2002) listed 

in Table 6-10 for the respective dosimetry systems. 

Table 6-10. Potential missed dose for Clarksville workers.a

Dosimeter Period

Exchange

frequencyb

MDL (mrem)

Missed annual mean 

dose (mrem)

Skin Deep Neutron Skin Deep Neutron

βγ film–SNL 7/1949–1958 Monthly 40c40c(d) 240 240

βγ film 1/1958–12/1959 Weekly 40c40c(d) 1,040 1,040

βγ film–NTA film 1/1960–3/1961 Weekly 40 40 (d) 1,040 1,040 (e)

4/1961–9/1964 Monthly 40 40 (d) 240 240 (e)

10/1964–12/1965 2/month 40 40 (d) 520 520 (e)

a. Data for 1958 and later are assumed identical to Pantex data (ORAUT 2006c). 

b. Exchange frequencies were established from dosimetry reports and the Rarrick interview (McConn 2006). The weekly 

exchange frequency was established with Tracerlab in October 1960 (Tracerlab 1965). 

c. Estimated MDL typical of film dosimeter capabilities (Wilson 1960, 1987; NIOSH 1993; NRC 1989; Wilson et al. 1990). 

d. The MDL for neutron doses was unreliable. 

e. The reconstructed neutron dose is calculated using the adjusted photon dose and a neutron-to-photon dose ratio. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 45 of 69

6.3.3 Radiation Dose Fraction

Table 6-2 summarizes the recommended fractions for Clarksville dose according to facilities, 

processes or activities, and energy categories required by IREP. 

6.4 ORGAN DOSE – OPERATIVE YEARS 

Once the Hp(10) adjusted doses have been calculated for each year, the values are used to calculate 

organ doses of interest using the external dose reconstruction implementation guidelines (NIOSH 

2002). Consistent with NIOSH Office of Compensation Analysis and Support/Oak Ridge Associated 

Universities Team agreements, dose reconstructors should assume the use of the AP (front-to-back) 

geometry for the irradiation geometry and for conversion to organ dose. Multiply the calculated 

neutron doses by the neutron deep dose equivalent organ dose conversion factors for AP irradiation 

from Appendix B of NIOSH (2002). For photons applicable to Clarksville Base (i.e., film badge era), 

use the conversion factor from exposure to organ dose. 

Some workers at Clarksville Base, generally production operators, operators, or operator trainees, 

might have handled unshielded pits with gloved hands. The dose rate at the bare surface of 

plutonium metal is approximately 4 rem/hr for the total of both photon and neutron radiation. This 

dose rate was based on calculations of the deep dose Hp(10) (Traub, Sherpelz, and Taulbee 2005) 

and modified at low photon energies to account for shallow dose, Hp(0.07). The dose rate to the 

hands will decrease due to the use of protective clothing such as gloves. If the individual has skin 

cancer on the hands and handled unshielded pits, dose reconstructors should perform case-specific 

dose rate calculations to the hands. 

6.5 ORGAN DOSE – POSTOPERATIVE YEARS 

External dose might have been received by persons doing minor maintenance or surveillance of the 

site after the facility was shut down in 1966–1967, due to possible residual DU contamination on 

floors or surfaces. As mentioned in Section 5.2.2, a radiological survey in 1997 found no 

contamination above release criteria (Last, Gilmore, and Bronson 1998). The release criterion for 

combined fixed and removable contamination was 5,000 dpm/100 cm2. Organ dose rates were 

calculated assuming the following: 

 A geometry of standing on a contaminated floor 

 5,000 dpm/100 cm2 of natural uranium, which would be favorable to claimants in relation to DU 

 Progeny radionuclides after 5-yr ingrowth 

Calculations were based on dose factors provided in the compact disk (CD) supplement to Federal 

Guidance Report No. 13 (Eckerman et al. 1999). The dose rates to the skin are averages over the 

entire phantom used for the calculations; no credit for protective clothing or shoes was applied in the 

calculations. It was assumed that time spent in proximity to contaminated surfaces was limited to a 

few hours a month so an occupancy factor of 0.02 was applied. The annual organ doses are listed in 

Table 6-11.

Because all measurements were reported as simply less than the release criterion, the above 

calculations are analogous with missed dose calculations; therefore, a triangular distribution should be 

applied with a minimum of zero, a mode based on the surface contamination at one-half of the release 

criterion, and a maximum based on the surface contamination at the release criterion. However, the 

doses are all less than 1 mrem/yr and can be ignored. They are provided in case a dose 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 46 of 69

Table 6-11. Annual external dose to organs 

from natural uranium surface contamination, 

1966 to 1967. 

Organ

Annual dose from surface 

contamination at or below 

release criterion (rem)

Mode Maximum

Adrenals 7.28E-07 1.46E-06

Bladder wall 7.98E-07 1.60E-06

Bone surface 2.08E-06 4.16E-06

Brain 7.58E-07 1.52E-06

Breasts 1.06E-06 2.12E-06

Esophagus 6.73E-07 1.35E-06

ST wall 7.95E-07 1.59E-06

SI wall 7.46E-07 1.49E-06

ULI wall 7.68E-07 1.54E-06

LLI wall 7.74E-07 1.55E-06

Kidneys 8.07E-07 1.61E-06

Liver 7.96E-07 1.59E-06

Lungs 8.44E-07 1.69E-06

Muscle 1.00E-06 2.00E-06

Ovaries 7.41E-07 1.48E-06

Pancreas 7.10E-07 1.42E-06

Red marrow 8.27E-07 1.65E-06

Skin 2.80E-04 5.60E-04

Spleen 8.01E-07 1.60E-06

Testes 1.05E-06 2.10E-06

Thymus 7.86E-07 1.57E-06

Thyroid 8.78E-07 1.76E-06

Uterus 7.42E-07 1.48E-06

reconstructor believes it appropriate to use a larger occupancy factor. No neutron doses should be 

applied. 
Document No. ORAUT-TKBS-0039 Revision No. 00 Effective Date: 11/14/2006 Page 47 of 69

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Last, G. V., T. J. Gilmore, and F. J. Bronson, 1998, Relative Risk Site Evaluation for Buildings 7740 

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McConn, R. J., 2006, “Interview with Harold Rarrick Concerning Operation and Worker Exposures at 

Clarksville on December 29, 2005, and Subsequent Electronic Mail Messages,” letter to H. 

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Computer Codes, NUREG/CR-6697, U.S. Nuclear Regulatory Commission, Washington, D.C. 
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GLOSSARY 

U.S. Atomic Energy Commission (AEC) 

Original agency established for nuclear weapons and power production; a predecessor to the 

U.S. Department of Energy. 

bird cage 

(1) Criticality-safe framework around a nuclear capsule in the in-flight insertable design; 

(2) common expression for Clarksville secured area.

bremsstrahlung 

Secondary photon radiation emitted by a charged particle under acceleration such as X-rays 

emitted by an electron that is scattered by a nucleus. From German meaning brake radiation. 

depleted uranium (DU) 

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

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

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

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

equilibrium factor (F)

In relation to the potential alpha energy of radon and its progeny in air, the ratio of the 

equilibrium equivalent concentration to the actual activity concentration of radon. See 

potential alpha energy concentration. 

Gravel Gertie 

Facility covered with crushed gravel to contain radioactive contamination from the potential 

accidental explosion of a nuclear weapon during assembly or disassembly, characterized by a 

cable-supported roof of wire mesh and a thick gravel overburden. Also called Gertie. 

igloo 

Earth-covered storage area for items that might explode, such as ammunition, high explosives, 

bombs, or bomb parts. 

nuclear emulsion 

Thick photographic coating in which the tracks of various fundamental particles show as black 

traces after development. The number of tracks in a given area is a measure of the dose from 

that radiation. See nuclear track emulsion, type A. 

nuclear track emulsion, type A (NTA) 

Film made by Eastman Kodak sensitive to fast neutrons. The developed image has tracks 

caused by neutrons that are visible under oil immersion with about 1,000-power magnification. 

operational years 

The period of AEC responsibility during which the site had an active mission that involved 

handling or storing radioactive materials. 

PM10

Particles less than 10 micrometers in aerodynamic median diameter that include both fine and 

coarse dust particles; essentially particles of respirable size. 
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postoperative years 

The period of AEC responsibility during which no active functions were being performed that 

involved handling or storing radioactive materials. The period after the site was shut down and 

major radioactive materials were removed, but was still under AEC jurisdiction. 

potential alpha energy concentration (PAEC) 

Kinetic energy in units of working levels potentially released in a unit volume of air by alpha 

particles emitted by the short-lived radioactive progeny of 222Rn (218Po, 214Pb, 214Bi, and 214Po) 

and 220Rn (216Po, 212Pb, 212Bi, 212Po). See potential alpha energy exposure and working level. 

potential alpha energy exposure (PAEE) 

Average potential alpha energy concentration to which a worker is exposed multiplied by the 

time of exposure in working months of 170 hours (units of working level months). PAEE is the 

potential alpha energy concentration multiplied by time. See potential alpha energy 

concentration and working level-month. 

progeny 

Nuclides that result from decay of other nuclides. In the case of 226Ra, for example, nine 

successive radioactive progeny occur in a decay chain. The chain ends with 206Pb (lead), 

which is a stable nuclide. Also called decay products and formerly called daughter products. 

radon (Rn) 

Radioactive gaseous element with atomic number 86. Radon is a decay product (progeny) of 

other radioactive elements such as thorium and radium. 

working level (WL) 

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 

radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 130,000 MeV of alpha energy; 1 WL equals 

2.083 × 10-5 joules per cubic meter. See potential alpha energy concentration. 

working level-month (WLM) 

Unit of exposure to radon progeny defined as exposure for 1 working month (170 working 

hours) to a potential alpha energy concentration from of 1 WL; 1 WLM equals 1 WL times 

170 hours, which is 0.00354 joule-hours per cubic meter. See potential alpha energy 

exposure and working level. 
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TABLE OF CONTENTS 

SECTION TITLE PAGE

A.1 Site Description and Functions ............................................................................................... 55

A.1.1 Site Description .......................................................................................................... 55

A.1.2 Operational History as a Weapons Storage Area........................................................ 59

A.1.3 Operation as a Modification/Disassembly Center........................................................ 60

A.2 Occupational Medical Dose .................................................................................................... 60

A.3 Environmental Dose ............................................................................................................... 61

A.3.1 Routine Doses ............................................................................................................ 61

A.3.2 November 13, 1963, Explosion ................................................................................... 61

A.3.2.1 Background .................................................................................................... 61

A.3.2.2 Intake Calculation ........................................................................................... 62

A.3.2.3 Direct Radiation ............................................................................................. 64

A.3.2.4 Inhalation from Resuspension ........................................................................ 65

A.4 Internal Dose .......................................................................................................................... 66

A.5 External Dosimetry ................................................................................................................. 67

LIST OF TABLES 

TABLE TITLE PAGE

A-1 Assumed isotopic mixtures of uranium ................................................................................... 62

A-2 Uranium soil measurements after explosion ........................................................................... 63

A-3 Environmental intakes ............................................................................................................ 66

A-4 Environmental external dose .................................................................................................. 66

A-5 Occupational intakes .............................................................................................................. 66

A-6 Dose recommendations for Medina workers .......................................................................... 68

LIST OF FIGURES 

FIGURE TITLE PAGE

A-1 San Antonio, Texas, area showing location of Lackland Training Annex (previously 

called Medina Base). .............................................................................................................. 56

A-2 Medina Base .......................................................................................................................... 57

A-3 Log probability plot of recorded annual doses received by monitored workers at 

Medina, 1959 to 1966 ............................................................................................................ 68

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A.1 SITE DESCRIPTION AND FUNCTIONS

Medina Base, located on about 3,700 acres of Lackland Air Force Base southwest of San Antonio, 

Texas, was one of 13 former WSAs created under the Armed Forces Special Weapons Project. 

Medina Base was constructed by the U.S. Air Force and the AEC between 1953 and 1955, with the 

first weapons components arriving in 1955. It was supported by Sandia Corporation (later SNL) for 

the AEC and the Air Force. SNL, AEC, and the Air Force were all active at Medina from 1955 until 

early 1959, performing maintenance and quality assurance on nuclear components of weapons 

(Mitchell 2003). 

The approved dates for applicability under the EEOICPA for this site are 1958 through 1966 

only. 

During the AEC tenure, nuclear weapons and weapon components were stored by the AEC and 

maintained by SNL and Air Force personnel at the Medina WSA. WSAs were comprised of storage 

buildings that housed nuclear capsules, maintenance structures, waste burial sites, and bunkers used 

for storage of weapons casings. SNL personnel worked at Medina under contract to the AEC until 

early 1959 (Martin 2006a). 

In 1958, MHSMC was chosen by AEC to manage Medina Base, and construction of new facilities was 

undertaken. Three Gravel Gertie cells were constructed along with other specialized facilities that 

comprised the Medina Modification Center (Mitchell 2003). From April 1959 until 1966, MHSMC 

operated Medina for the AEC as a weapons modification and disassembly facility. The mission was 

to perform stockpile surveillance, modifications, retrofits, and weapon retirements (Carr ca. 1992). 

This work included inspections for corrosion and replacement of tritium reservoirs. Medina was 

operated by MHSMC until January 1966, when its mission was transferred to the Pantex Plant, and 

Medina Base was transferred back to the Air Force. 

A.1.1 Site Description

Lackland Air Force Base (AFB) is within the San Antonio metropolitan area in Bexar County, Texas 

(Figure A-1). Commercial and residential developments border Lackland AFB on the north, west, and 

south sides, and Kelly AFB borders it on the east. The western portion of Lackland AFB was the 

Medina Base, which is now designated the Lackland Training Annex. The Medina Base was 

comprised of four main areas designated as the 200 Area (general shops and stores), 300 Area 

(operations and storage), 400 Area (Plants 1 and 2, main production operations), and 500 Area 

(igloos and storage facilities). Two additional areas were the Burning Ground and the Railhead 

(Figure A-2) (Lamb Associates and Halliburton NUS 1995). 

There were seven main types of structures at the Medina Base, including “A” Structures, “C” 

Structure, Base Spares Warehouse, assembly/maintenance buildings, “S” Structure, storage igloos, 

and the modification/disassembly plants (Lamb Associates and Halliburton NUS 1995). The site also 

included low-level radioactive waste (LLRW) disposal areas and emergency underground holding (or 

storage) tanks (USTs). Each of these structures is briefly described in the following paragraphs. 

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Figure A-1. San Antonio, Texas, area showing location of Lackland Training Annex (previously called 

Medina Base). Source: Rand McNally Road Atlas 2006. 

“A” Structures included Buildings 301, 402, 403, 404, 552, 556, 562, 571, and 585 in the 300, 400, 

and 500 Areas. “A” Structures were used to store nuclear capsules for weapon systems (Lamb 

Associates and Halliburton NUS 1995). The buildings, though massive concrete structures, had only 

four small storage rooms, each approximately 10 ft wide, 13 ft deep, and 9 ft high. Each room had a 

capacity for approximately 30 nuclear capsules that were stored in critically safe bird cage containers 

(Figure 2-5). Each room had a bank-vault type door equipped with dual combination locks. The 10-ftthick walls and massive berms around the rooms were designed to protect the nuclear capsules from 

external attacks, rather than as containment of possible accidental detonations within the buildings. 

Maintenance activities always took place in the “C” Structure, never in the vault where the capsules 

were stored. Therefore, no nuclear material was ever exposed in an “A” Structure and there was little 

or no potential for a release of radioactive material with these buildings. Activities in the “A” 

Structures ended in 1960 (Lamb Associates and Halliburton NUS 1995). 

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Figure A-2. Medina Base. Source: Lamb Associates and Halliburton NUS 1995. 

Building 307, a “C” Structure, was used as a nuclear materials inspection laboratory and maintenance 

building for the nuclear weapons stored at Medina Base. The “C” Structure provided bench space to 

perform required maintenance operations, storage for neutron calibration and assay sources, and 

support facilities that included a change room and storage areas. Nuclear capsules destined for 

maintenance were transported in their bird cages to the “C” Structure; when maintenance was 

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completed, the capsules were placed back in their bird cages, sealed, and returned to one of the “A” 

Structure buildings (Lamb Associates and Halliburton NUS 1995). 

Workrooms in the “C” Structure were constructed with floor drains that were connected to an 

emergency UST in the building. If there had been an accident, the UST would have collected the 

decontamination washwater and any plutonium released. No accidental releases of plutonium are 

known to have occurred at Medina Base during the 1955–1966 period (Lamb Associates and 

Halliburton NUS 1995). 

The “C” Structure had an extensive ventilation system that prevented the release of uranium oxides to 

the atmosphere. Based on information reviewed from other environmental reports, there were no 

indications that any spills or releases of radioactive material occurred in the “C” Structure during its 

operational lifetime (Lamb Associates and Halliburton NUS 1995). 

The Base Spares Warehouse was used to store spare weapon components for maintenance 

purposes (Lamb Associates and Halliburton NUS 1995). 

The Assembly/Maintenance Buildings were also called Plants 1 and 2; each plant consisted of two 

buildings used for maintaining non-nuclear components of weapons stored at Medina. The buildings 

contained several bays, and activities included inspection, testing, and assembly of non-nuclear 

mechanical and electrical systems. The buildings featured heavy blast doors and earthworks that 

would have deflected the effects of an accidental explosion upward. The design is typical for facilities 

in which large amounts of high explosives are handled (Lamb Associates and Halliburton NUS 1995). 

An “S”, or Surveillance, Structure (Building 444) was used for inspections and testing of weapons in 

the stockpile. The “S” Structure separated quality assurance activities from the routine maintenance 

and assembly functions performed at Plants 1 and 2. Building 444 contained electrical and 

mechanical bays and a calibration room, but did not have a floor drain tied to a UST. The building 

was modified in 1959 for the modification/disassembly mission managed by MHSMC (Lamb 

Associates and Halliburton NUS 1995). 

One hundred storage igloos were constructed in the 500 Area for the storage of weapon components, 

assembled weapons, and weapon casings. All nuclear materials stored in the igloos were sealed in 

the weapons. No maintenance activities took place in the igloos and, therefore, there was little or no 

potential for any release of radioactive or hazardous materials (Lamb Associates and Halliburton NUS 

1995). 

In 1959, three Gravel Gertie cells were built for modifying and disassembling weapons. Highexplosive shells were removed from nuclear assemblies in these structures. Several tons of gravel 

were located above the ceiling of each structure for containment of fissile material in the event of an 

accidental detonation of the high-explosive system. The Gravel Gerties were used between 1959 and 

1965 (Lamb Associates and Halliburton NUS 1995). 

Three sites at Medina (RW-15, RW-17, and RW-19) were designed for the collection of dry and liquid 

LLRW. RW-15 was a landfill used for the disposal of LLRW, and is suspected of having received 

classified limited-life components disposed of during weapons modification and disassembly 

operations between 1959 and 1965. The LLRW was excavated in 1965 and transferred in CONEX 

containers to Pantex. The RW-15 site was cleared by AEC as being decontaminated when the site 

was closed in 1965 (Lamb Associates and Halliburton NUS 1995). Dry LLRW generated in the “C” 

Structure was packed in cardboard boxes and disposed in the RW-17 site. RW-17 was an unlined pit 

that was fenced and considered a classified waste landfill. RW-19 was a small gravel leaching area 
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behind Building 444. The site reportedly received intermittent LLRW wastewater discharges from 

Building 444. Gravel and soil from the area were excavated in 1965 and transferred in CONEX 

containers to Pantex (Lamb Associates and Halliburton NUS 1995). 

Three emergency USTs were designated RW-16, RW-18, and RW-20. The USTs were intended for 

the collection of liquid LLRW in the event of an accidental release; however, no evidence exists that 

any of the USTs were ever used (Lamb Associates and Halliburton NUS 1995). RW-16 and RW-18 

are 1,000-gal steel tanks; RW-20 is a 5,000-gal steel tank. 

A.1.2 Operational History as a Weapons Storage Area

The early nuclear weapons used Po-Be initiators to generate neutrons during the explosion sequence. 

Because 210Po has a half-life of only 138 d, the initiators had to be replaced periodically. According to 

former SNL personnel (Martin 2006a), these devices were maintained following precise quality control 

methods that required maintenance personnel to: 

1. Release pressure from the bird cage container through a filter and check the filter for alpha 

activity. If no activity was found, remove the capsule from the container using a handling tool. 

2. Place the capsule on a table top with an alpha probe at one end. (The table was covered with 

a large piece of butcher paper to contain any spalling of uranium oxides.) 

3. Place a Plexiglas glovebox over the capsule. 

4. Disassemble the capsule parts and check the integrity of the coatings. 

5. Remove the glovebox. 

6. Remove uranium oxide deposits from the threads using a small cloth or paper swipe and 

trichloroethylene. Wipe the threads with ethyl alcohol to dry the components. 

7. Use acetone to remove previous markings made with blue machinist’s dye and make new 

markings. (Later components had serial numbers etched on their surfaces.)

8. Check the activity of the fissile material using beta and gamma radiation measurements. 

9. Assay the nuclear material by accurately weighing it and perform subcritical multiplication 

measurements using external neutron sources. 

10. Replace the Po-Be initiators. (These were later replaced with nonradioactive initiators.) 

11. Reassemble the capsule. 

12. Place the capsule and a sack of desiccant in the bird cage container. 

13. Screw on the bird cage container top. Repressurize and wire seal the bird cage container 

(positive pressure was maintained to ensure dryness and keep O-rings in place) (Lamb 

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Associates and Halliburton NUS 1995). [Later designs used a vacuum rather than 

overpressure (Bihl 2006c).]

Contaminants generated in the maintenance process were chemicals mentioned above and uranium 

oxides. Polonium-210 waste was not generated (Lamb Associates and Halliburton NUS 1995). 

(However, intakes of 210Po are modeled; see Section 5.3.) Used (decayed) initiators were sent to Los 

Alamos National Laboratory for storage, regeneration, or disposal. The spalled uranium oxides, 

swipes contaminated with solvents, lead-wire seals, and gloves were wrapped in butcher paper and 

placed in 18-by-18-by-24-in. cardboard boxes for disposal in the dry LLRW disposal area (currently 

designated RW-17) (Lamb Associates and Halliburton NUS 1995). 

A source safe (a cylindrical apparatus located below the floor surface with a polyethylene neutron 

absorber at the top) was located in a corner of the laboratory room in the “C” Structure. Below the 

neutron absorber, the source safe had a tray for storing check sources used to verify the activity of the 

fissile material in the weapons (Lamb Associates and Halliburton NUS 1995). 

Between 1954 and 1957, the Po-Be initiators and plated pits used in the in-flight-insertable design 

were phased out and replaced with sealed pits and a newer type of sealed neutron generator that did 

not require routine replacement (Lamb Associates and Halliburton NUS 1995, Mitchell 2003). Thus, 

some SNL workers handled Po-Be initiators and plated pits prior to 1957, but no MHSMC workers 

were exposed to the initiators or plated pits. After 1957, maintenance activities were reduced to 

annual disassembly of capsules to determine their condition and to verify the integrity of the fissile 

materials. Maintenance of the newer capsules generated the same types of waste, but in smaller 

quantities because of the less frequent maintenance schedule. By 1960, nuclear capsules had been 

phased out of the stockpile and maintenance activities no longer involved any exposed nuclear 

material (Lamb Associates and Halliburton NUS 1995). 

A.1.3 Operation as a Modification/Disassembly Center

Between 1959 and 1966, the mission of MHSMC workers was to perform stockpile surveillance, 

modifications, retrofits, and weapon retirements (Carr ca. 1992). Typical modifications involved 

disassembly and reassembly with some modified components and replacement of tritium reservoirs. 

Weapon retirements involved complete disassembly and return of nuclear components to other DOE 

sites. Some damaged weapons were returned to Medina during these years. Their ultimate 

disposition has not been determined, but they were no longer at Medina after shutdown. During this 

period, the MHSMC workers at Medina were exposed to the same types of radiation and levels of 

contamination as Pantex workers because their work activities were nearly identical (except there 

were no hydroshot operations at Medina) (Mitchell 2003). 

A.2 OCCUPATIONAL MEDICAL DOSE 

Similar to Clarksville Base, the Pantex Plant has the X-ray histories of Medina workers during the 

MHSMC years. Only written histories, not film, are available. Because essentially the same work was 

done by the same contractor, it is assumed that the X-ray guidance for Clarksville Base should be 

applied to Medina workers for the MHSMC years, 1958 to 1966. (There is evidence in the claims files 

that MHSMC gave some preemployment X-rays in October 1958). No X-rays should be assigned for 

the Sandia years (i.e., prior to October 1958) unless there is clear evidence in the claim files. 
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A.3 ENVIRONMENTAL DOSE 

A.3.1 Routine Doses

Insufficient information has been obtained about Medina Base to develop site-specific considerations 

for environmental releases. The only environmental doses assigned for the Pantex Plant are from 

radon and a large accidental release of tritium. Similarly, environmental doses at Clarksville Base 

resulted from a tritium accident modeled after the Pantex accident and from direct radiation from use 

of radiography sources. Throughputs, in terms of number of disassemblies, for instance, of the four 

assembly/disassembly sites (Pantex, IAAP, Clarksville, Medina) are classified, so comparison of 

effluents by relationships to throughput is not possible. Medina had three Gravel Gerties compared to 

one at Clarksville. 

The average tritium release from the Pantex Plant from 1981 through 1988, when a large number of 

disassemblies was performed, was 70,000 μCi/yr. This compares to a range of 6,000 to 26,000 μCi 

during a few years at IAAP (ORAUT 2005b). Assuming that Medina Base was a larger operation than 

Clarksville Base and approximated Pantex operations, and using the same formulas as those for the 

Clarksville tritium releases, an annual environmental tritium intake of 18,300 pCi/d is estimated 

[(70,000/26,000)(6,800 pCi/d)]. The annual dose to any organ from this intake is less than 1 mrem 

and can be ignored. 

There is no reason not to believe that Medina Base had the same radiography sources as Clarksville 

Base. Dose reconstructors should assign Medina workers the same 4-mrem external dose from 

radiography as that for Clarksville workers (Section 4.0 and Table 4-2.) 

A.3.2 November 13, 1963, Explosion

A.3.2.1 Background 

On November 13, 1963, an explosion involving 50,500 kg (123,000 lb) of chemical high-explosive 

components of nuclear weapons occurred at the Medina site. Workers were placing subassemblies 

from dismantled atomic bombs into storage Igloo 572. The subassemblies, which were being stored 

for further processing and disposal, contained chemical high explosives, aluminum, natural uranium, 

and DU. They were handled by a three-man crew – two forklift operators who moved them from a 

straddle carrier into the igloo and one man on the carrier. 

Most of the load was in the igloo when, at about 10:24 a.m., the explosive in one of the 

subassemblies ignited. Seeing the flash, the drivers sprinted for cover, alerting the workers outside. 

For about 45 sec the explosive burned, then it detonated with a force of over 60 tons of TNT. The first 

explosion set off other subassemblies in the igloo and those still on the carrier. The igloo disappeared 

in a cloud of smoke and dust, leaving a crater some 20 ft deep. A large cloud of dust was seen near 

the ground moving downwind of the event. 

In the 45 sec between ignition and detonation, the three workers got away. Their injuries were minor. 

Adjacent igloos were not disturbed. The only known radioactive material involved was a mixture of 

DU and natural uranium. Amounts involved are not available. Fissile materials (enriched uranium or 

plutonium) were not involved in the accident. 

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Radiation surveys on the Base and downwind were performed immediately after the accident. No 

metallic fragments were identified. Soil samples were collected and analyzed for uranium content, 

reported as 238U in micrograms uranium per gram of soil. Concentrations elevated above background 

were found in the downwind direction, west-southwest of the site of the explosion; no areas of 

elevated concentration were found off the Base. It is not known if areas of higher contamination were 

removed. 

Based on the historical description, the only radionuclides of concern are natural uranium and DU. 

The isotopic mixtures of these in terms of activity are listed in Table A-1.

Table A-1. Assumed isotopic mixtures of 

uranium, % activity.a

Type of uranium U-234 U-235 U-238

Natural 48.9 2.25 48.9

Depleted 15.4 1.06 83.6

a. From Rademacher (2000). 

Because the initial masses of uranium in the igloo are not reported, and only soil concentrations of 

238U are reported (it is hypothesized that total uranium activities were reported as 238U), it is favorable 

to claimants for external dose rate analyses to assume the mixture is natural uranium and, for 

inhalation analyses, that the activity is all 234U.

A.3.2.2 Intake Calculation 

The available historical soil measurements, circa November – December 1963, are listed in Table A-2 

(from Rademacher 2000). The basis for the coordinate system is unknown; however, assuming that it 

is a Cartesian grid in feet, the area of elevated concentration is an ellipse larger than 4,000 ft by 

2,000 ft. The highest concentrations approach 2,400 μg 238U/g of soil, equivalent to about 800 pCi/g. 

It is assumed that all the uranium is reported as 238U. The depth of soil sampled in these 

measurements is not known. If it is assumed that a relatively shallow sample (conservative because it 

assumes little mixing into deeper soil) of 5 cm was taken, with a soil density of 1.5 g/cm3, this enables 

the calculation of an areal contamination of: 

800 pCi/g × 104cm2/m2 × 5 cm × 1.5 g/cm3 ×10-6 μCi/pCi = 60 μCi/m2(A-1)

[To get an estimate of the total amount of uranium involved in the explosion, if we assume from the 

tabulated measurements that the major contaminated area is about 4,500 ft by 600 ft (1,500 m by 

200 m) at an average concentration of about 400 pCi/g, the total deposition would be about 

400 pCi/g × (60 μCi/m2/800 pCi/g) × 1,500 m × 200 m = 9,000,000 μCi (A-2)

or, with a specific activity of 0.33 μCi/g for 238U, around 30 tons of uranium. This is somewhat 

comparable to the amount of high explosive involved, so it seems reasonable.] 
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Table A-2. Uranium soil measurements after explosion, circa November-December 

1963. 

The areal concentration enables direct estimation of the external dose rate and indirect estimation of 

the total time-integrated atmospheric concentration. It is assumed that the uranium was in particulate 

form, either as tiny fragments of material or condensed or agglomerated onto dust from the explosion. 

Ten-μm (respirable size) particles with density of about 10 g/cm3 have a deposition velocity of about 

0.03 m/sec (Sehmel 1984). Models (HOTSPOT, GENII) indicate that particles of this size, for the 

energetic event described and for a source of this magnitude, would result in a deposition pattern 

approximately the same as that seen in the measurements. Using a deposition velocity of 

0.03 m/sec, and the maximum measured deposition level of about 60 μCi/m2, the time-integrated air 

concentration is 

(6.0 × 107 pCi/m2) ÷ 0.03 m/s = 2.0 × 109 pCi-s/m3(A-3)

This can be assumed to represent the maximum likely exposure. The plume drifted to the westsouthwest of Igloo 572. There appear to be no regularly occupied facilities or structures in this 

direction from the point of the explosion for about 10,000 ft (3,000 m), at which point dispersion 

modeling indicates that the concentration would be less than one-tenth of this value. 

A person breathing at a rate of 1.2 m3/hr at the maximum point would inhale 

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 (2 × 109 pCi-s/m3)(1.2 m3/hr) ÷ 3,600 sec/hr = 6.7 × 105 pCi 238U or 1.4 × 106 pCi U nat (A-4)

Because the material was a mixture of DU and natural uranium, assuming 100% natural uranium is 

favorable to claimants. This would be an acute inhalation of type S material for the three workers who 

were closest to the explosion or to anyone who reported having been in the plume immediately 

following the explosion. For general workers at Medina in the nearest occupied area, the maximum 

intake was 1.4 × 105 pCi. Use 10-μm particle size. The geometric standard deviation for 10-μm 

particles is 2.499526 (to be used in IMBA when changing the particle size; not to be confused with a 

GSD of 3 for dose for input into IREP). A spreadsheet (intake doses Medina explosion.xls) providing 

annual doses to organs from this acute intake has been developed for use by dose reconstructors. 

Following the accident, there were two potential sources of dose: the external dose rate from direct 

radiation from the soil and inhalation of resuspended radioactive materials. 

A.3.2.3 Direct Radiation 

Assuming an infinite plane source and the distribution of natural uranium (which has a higher 

compounded dose rate than DU) in equilibrium with its short-lived progeny, external doses can be 

calculated. Dose rates can be calculated for each organ using organ-specific, dose-rate conversion 

factors such as those in Federal Guidance Report No. 12 (Eckerman and Ryman 1993) or the CD 

supporting Federal Guidance Report No. 13 (Eckerman et al. 1999). 

Using the factors from Federal Guidance Report No. 13 CD (Eckerman et al. 1999), the rate of 

effective dose per μCi/m2 would be 

238U + 234Th + 234mPa (0.489)(49.5 + 876 + 12,600) × 10-6 rem/yr per μCi/m2

235U + 231Th (0.0225)(16,400 + 1,720) × 10-6 rem/yr per μCi/m2

234U (0.489)(68.4) × 10-6 rem/yr per μCi/m2

= 0.025 rem/yr = 25 mrem/yr per μCi/m2 for full-time occupancy (A-5)

However, this value is somewhat misleading; Federal Guidance Report No. 13 factors include a 

tissue-weighting factor for skin dose. Using Federal Guidance Report No. 13 factors but omitting the 

contribution from the weighted skin dose, the gamma dose rate, which more closely approximates the 

average organ dose, is 

238U + 234Th + 234mPa (0.489)(40.7 + 856 + 1,600) × 10-6 rem/yr per μCi/m2

235U + 231Th (0.0225)(16,200 + 1,820) × 10-6 rem/yr per μCi/m2

234U (0.489)(57.8) × 10-6 rem/yr per μCi/m2

= 0.0017 rem/yr = 1.7 mrem/yr per μCi/m2 for full-time occupancy (A-6)

For an occupational exposure of 2,000 hr/yr, this would be 0.39 mrem/yr per μCi/m2. The highest 

onsite location, 60 μCi/m2, might have resulted in 23 mrem/yr. For regularly occupied areas 

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downwind, a value of one-tenth of that, or 2.3 mrem/yr, would be appropriate. Most of the site is 

upwind of the footprint left by the plume, so the external dose from uranium in the soil would be 

negligible. However, at present there is no way to discriminate between workers in offices upwind 

versus downwind of the plume footprint. Therefore, dose reconstructors should apply a 2.3-mrem/yr 

external dose (photon 30–250 keV) for all organs except skin. 

The skin dose, using Federal Guidance Report No. 13 values (Eckerman et al. 1999), could be as 

high as: 

238U + 234Th + 234mPa (0.489)(867 + 2,010 + 1,100,000) × 10-6 rem/yr per μCi/m2

235U + 231Th (0.0225)(22,700 + 10,000) × 10-6 rem/yr per μCi/m2

234U (0.489) (1,060) × 10-6 rem/yr per μCi/m2

= 0.54 rem/year = 540 mrem/year per μCi/m2 for full-time occupancy (A-7)

For an occupational exposure of 2,000 hr/year, this would be 123 mrem/yr per μCi/m2. The highest 

onsite location, 60 μCi/m2, might have resulted in a skin dose of 7,400 mrem/yr. For regularly 

occupied areas downwind, a value of one-tenth of that, or 740 mrem/yr, would be appropriate. For 

areas upwind (most of the site), a negligible value would apply; however, dose reconstructors should 

use the 740-mrem/yr value if it is not possible to tell if the work location was outside the plume. 

A.3.2.4 Inhalation from Resuspension 

Material deposited on the ground can be resuspended by wind or mechanical stresses. If it is 

assumed that the annual average concentration of dust in the air is 50 μg/m3 and that it is all 

generated locally, an upper bound on the amount inhaled can be developed. [The RESRAD manual 

(Yu et al. 2000) presents a distribution from the Environmental Protection Agency’s Aerometric 

Information Retrieval System (AIRS). The RESRAD distribution, verified by download from the AIRS 

Web site (http://www.epa.gov/airs/airs.html), indicates that the average U.S. concentration of PM10 is 

only about 23 μg/m3. Recent data from a particulate sampler in the Hanford 200 Area for February 

2001 through June 2002 is available (Napier and Snyder 2002, Appendix C). The mean air 

concentration of PM10 in the 200 Areas, in an outdoor area influenced by the Hanford 24 Command 

wildfire in 2000, was 21 μg/m3. The 95th-percentile daily value was 36.5 μg/m3. These data show 

that 50 μg/m3is an upper bound value.] For continuous occupancy in the most contaminated location, 

about 2,400 μg 238U per g soil, the amount inhaled per calendar day would be approximated as:

 (50 μg soil/m3air)(1.2 m3/hr)(2,000 hr/yr)(2,400 μg 238U/g soil)(1 × 10-6 g soil/μg soil)/365 cal d/yr 

= 0.79 μg 238U/d (A-8)

For natural or depleted uranium, the 238U accounts for essentially all of the mass (more than 99%), so 

the total uranium mass is also 0.79 μg/d. If a specific activity of 0.7 pCi/μg natural uranium is used, 

the daily intake is 0.55 pCi/d. However, because there are only igloos in the vicinity, it is unlikely that 

a worker would have continuously occupied the area around the most contaminated soil. Therefore, 

an occupancy factor of 0.1 was assumed. The resulting daily intake from resuspension is 0.055 pCi/d 

of natural uranium (modeled as 234U.) The absorption type is S; the distribution is a constant. 

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Tables A-3 and A-4 summarize environmental intakes and external doses, respectively. 

Table A-3. Environmental intakes. 

Period of 

application Applied to Mode Type Radionuclide Intake Distribution

November 1963 In plume close to Igloo 572 Inhalation Acute U-234 type Sa1.4E6 pCi Constant

November 1963 In plume at nearest occupied 

facility

Inhalation Acute U-234 type Sa1.4E5 pCi Constant

November 1963 Facilities not in plume No intake

November 1963–

1967

In plume footprint at nearest 

occupied facility or working 

outdoors near crater

Inhalation Chronic U-234 type Sb5.5E-2 

pCi/d

Constant

a. Use 10-μm particle size and GSD of 2.499526 (IMBA only, not IREP). 

b. Use standard 5-μm particle size. 

Table A-4. Environmental external dose. 

Period of 

application Applied to

Annual wholebody dose Photon energy 

category Distribution

1958–1966 All 4 mrem 30-250 keV Lognormal GSD = 2

1964–1966 All 2.3 mrem 30-250 keV Lognormal GSD = 2

1964–1966 All Skin dose: 740 Beta Lognormal GDS = 2

A.4 INTERNAL DOSE 

Occupational intakes of tritium and DU by Medina workers are assumed to be the same as those for 

Clarksville workers during the MHSMC years, with the exception of the accidental tritium release at 

Clarksville Base. See Table A-5.

Table A-5. Occupational intakes. 

Job category

Dose calculation parameters IREP input parameters

Period Material Mode

Absorption 

type pCi/d

Distribution 

type 1 2

Production operator, 

nuclear components 

inspector, quality control 

specialist

1958–Jan. 

1966

Tritium Chronic inhalation and 

skin absorption

NA 1.3E7 Constant Dose

Production operator, 

nuclear components 

inspector, quality control 

specialist

1958–Jan. 

1966

DU Chronic inhalation, 

Ingestion

M, S

Insoluble 

4.4E2

9.1

Constant 

Constant

Dose

Dose

Production operator, any 

inspector, quality control 

specialist

1958–1966 Radon Chronic inhalation NA NA Lognormal 0.072 

WLM/ 

12 mo. 

3

Material handler 1958–1966 Radon Chronic inhalation NA NA Lognormal 0.029 5

Other nonadministrative 1958–1966 Radon Chronic inhalation NA NA Lognormal 0.0072 3

Burning of high 

explosives

1958-1966 DU Chronic inhalation M or S 130 Constant Dose

No information about radon levels in Medina facilities has been found. Mr. Rarrick indicated that 

radon was an issue at Medina (Martin 2006a). The Pantex Plant is in an area rated moderate for 

radon by the U.S. Geological Survey, whereas Medina is in an area rated low and Clarksville is in a 

boundary area between low and moderate. However, radon levels at Medina would probably be more 

similar to those at Pantex than to those at Clarksville because the covered structures were similar. 

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Therefore, workers who spent essentially full time in covered buildings such as Gravel Gerties should 

be assigned 0.072 WLM per 12-mo exposure. Material handlers, who often entered igloos or other 

storage areas, should be assigned an occupancy factor of 0.4, and workers who occasionally entered 

Gravel Gerties, cells, or igloos (e.g., security or crafts staff) should be assigned an occupancy factor 

of 0.1 (ORAUT 2006a).

A.5 EXTERNAL DOSIMETRY 

The MHSMC dosimetry records for Medina began on April 27, 1959, and extended through 

January 24, 1966 (Pantex Plant 1959–1966). It is possible that some earlier dosimetry records exist 

for Sandia personnel, but none has been found to date. Exposures to the few (less than 10) Sandia 

personnel between 1955 and 1959 did not involve exchange of Po-Be initiators, and Mr. Rarrick 

estimated maximum exposures at 100 mrem per year (Martin 2006a). However, the approved 

EEOICPA dates for this site are 1958 through 1966 only. MHSMC utilized weekly dosimetry services 

provided by Tracerlab from April 1959 until July 1964 (Tracerlab 1965). Weekly dosimeters were 

provided by Nuclear Service Laboratories of Knoxville, Tennessee, from July 1964 until January 1966 

(NSL 1964–1966). Work at Medina was completed in January 1966, and dosimetry services were 

terminated on January 24, 1966. Most of the MHSMC workers left the site in January and February 

1966. Many of the workers transferred to the Pantex Plant in Amarillo, Texas. 

The clerical maintenance of the dosimetry records was not rigorous. For example, individual names 

were consistently assigned to only 11 film badge numbers, and some numbers were reused when an 

individual left the program. More than 100 other film badges were assigned to different people each 

week and the individual names, for nonzero results only, were recorded on the film badge dose report 

after it was received. In some weeks, the task of recording names associated with nonzero dose 

results was not completed, so it is impossible to reconstruct who received nonzero doses. Because 

more than one individual was assigned to a given film badge number, the quarterly and annual totals 

maintained by the film badge processor were not useful. 

Therefore, the dose records that were retrieved from Pantex were carefully examined, corrected, and 

summarized in a letter to the Project file, “Medina Dose Records, 1959-1966” (Martin 2006b). Dose 

reconstructors should disregard any incomplete dose records provided with the claims and use Martin 

(2006b). 

Figure A-3 shows a log probability plot of all recorded annual doses received by monitored workers at 

Medina Base for 1959 to 1966. This plot can be compared with the plot in Figure 6-2 for Clarksville 

Base. The 50th- and 95th-percentile doses shown in Figure A-3 are slightly higher than the same 

percentile doses for the Clarksville data. 

Because the doses received by Medina and Clarksville workers appear similar in magnitude within the 

uncertainties in how annual doses were calculated, the recommendations for Medina are the same as 

those for Clarksville. The exception is that it appears the exchange frequency at Medina was weekly 

for all years of operation. Because the work performed at Medina was similar to that performed at 

Clarksville, and the distribution of annual doses received at Medina, as seen in Figure A-3, is similar 

to that for Clarksville, it is reasonable to use the same dose assignment recommendations for Medina 

as those described for Clarksville Base. A summary of dose recommendations for Medina is listed in 

Table A-6.
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1 5 10 50 100 500 1000

.001

.005

.01

.02

.05

.1

.2

.3

.4

.5

.6

.7

.8

.9

.95

.98

.99

.995

.998

Lognormal Probability Plot

Point plotting method = nelson

Medina Annual Body Doses (mrem) Lognormal Probabilities

100

813

Figure A-3. Log probability plot of recorded annual doses received by monitored 

workers at Medina, 1959 to 1966 (Pantex 1959–1966). 

Table A-6. Dose recommendations for Medina workers. 

Monitored status Dose type Recommendation

Monitored Photon Maximum of recorded annual dose or 20 mrem × number of 

weeks worked during year

Neutron Neutron dose = 1.6 × photon dose

Neutron Multiply assigned neutron dose by 1.91

Not monitored Exposure categories As listed in Table 6-4

Photon As listed in Tables 6-7 to 6-9

Neutron Neutron dose = 1.6 × photon dose

Neutron Multiply assigned neutron dose by 1.91

Radiation Incident at Medina on September 7, 1959

On September 7, 1959, there was a radiation exposure incident at Medina that involved a radiography 

source (Pantex Plant 1959–1966). The details were provided in CATIs, the most detailed one 

associated with claim number 5800. A vendor came to the site to demonstrate a new type of portable 

radiography source/shield that was air operated (apparently referred to as a Puff camera). The device 

could pneumatically transfer a source from the safe shielded position, through a 30-ft tube, to a 

radiography exposure location. The end of the tube was located so that it could take an X-ray of an 

electronic part. The vendor representative first performed the transfer of the capsule out to the end of 

the tube and back into the lead shield using a dummy (nonradioactive) capsule. The transfer was 

successfully demonstrated several times and the Medina radiography staff also successfully used the 

equipment with the dummy capsule in a training exercise. While the vendor representative was 

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leaving the site, he realized he did not have the dummy capsule with him and there could be a 

problem. An investigation revealed that the dummy capsule was still in the lead shield and when the 

air flow was reversed, the 30-Ci 192Ir source could not enter the shield because the dummy capsule 

was in the shield. When the air flow stopped, the 192Ir source fell back down the tube to an area 

where the radiographers were present. As a result of this incident, three individuals received 

unusually high radiation exposures, as follows: Radiographer A – 6,600 mrem gamma, Radiographer 

B – 3,000 mrem gamma, and Radiographer C – 2,370 mrem gamma. These exposures are included 

in the summarized records (Martin 2006b).