ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 43

Document Title:

Weldon Spring Plant – Occupational 

External Dosimetry

Document Number: ORAUT-TKBS-0028-6

Revision: 00

Effective Date: 06/24/2005

Type of Document: TBD

Supersedes: None

Subject Experts: James M. Langsted and Craig A. Little

Document Owner

Approval: Signature on File Approval Date: 06/20/2005

Robert Meyer, TBD Team Leader

Approval: Signature on File Approval Date: 06/20/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 06/20/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 06/24/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

PLEASE DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 2 of 43

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

01/14/2005 00-A Technical Basis Document for the Weldon Spring Plant, Section 6, 

Occupational External Exposure. Initiated by Robert Meyer.

03/24/2005 00-B Technical Basis Document for the Weldon Spring Plant, Section 6, 

Occupational External Exposure. Initiated by Robert Meyer.

06/15/2005 00-C Revised in response to OCAS comments. Initiated by Robert Meyer.

06/24/2005 00 First approved issue. Training is not required. Initiated by Robert 

Meyer.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 3 of 43

TABLE OF CONTENTS

Section

Publication Record ..............................................................................................................................

Page

2

Acronyms and Abbreviations ...............................................................................................................5

6.1 Introduction.................................................................................................................................6

6.1.1 Purpose..............................................................................................................................6

6.1.2 Scope.................................................................................................................................6

6.1.3 Dosimetry Overview ...........................................................................................................6

6.1.3.1 Plant Operations Period (1957 to 1966)..................................................................... 7

6.1.3.2 Initial Cleanup Period (1967 to 1969)....................................................................... 12

6.1.3.3 Monitoring and Maintenance Period (1969 to 1985) ................................................ 12

6.1.3.4 Site Remediation Period (1985 to 2000) .................................................................. 12

6.2 Dose Reconstruction Parameters ............................................................................................. 12

6.2.1 Interpreting the External Dosimetry Record......................................................................12

6.2.2 Weldon Spring Historical Administrative Practices............................................................13

6.2.2.1 Recorded Doses...................................................................................................... 13

6.2.2.2 Discrepancies.......................................................................................................... 13

6.2.2.3 Missing Entry........................................................................................................... 13

6.2.2.4 Badge Assignment and Exchange Frequency ......................................................... 14

6.2.2.5 Interpretation of Reported Data ............................................................................... 14

6.2.3 Plant-Wide Dosimetry Results..........................................................................................15

6.2.3.1 Calibration ...............................................................................................................16

6.2.4 Workplace Radiation Fields..............................................................................................16

6.2.4.1 Gamma Dose .......................................................................................................... 17

6.2.4.2 Neutron Dose .......................................................................................................... 19

6.2.4.3 Electron Dose.......................................................................................................... 20

6.2.4.4 Reported Dose-to-Organ-Dose Conversion Factor Units ......................................... 23

6.2.4.5 Limit of Detection..................................................................................................... 23

6.2.4.6 Exchange Frequency............................................................................................... 24

6.2.4.7 Number of Zero Readings........................................................................................ 25

6.2.4.8 Determination of Missed Dose................................................................................. 26

6.2.4.9 Unmonitored Energy Range .................................................................................... 26

6.2.5 Angular Dependence........................................................................................................27

6.2.6 Uncertainty.......................................................................................................................27

6.2.6.1 Film .........................................................................................................................27

6.2.6.2 Thermoluminescent Dosimetry System.................................................................... 27

Glossary ............................................................................................................................................32

Attachment A Example Reports........................................................................................................ 36
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LIST OF TABLES

Table

6-1 Beta and gamma emissions of primary interest .........................................................................

Page

7

6-2 Summary of historical recorded dose practices ....................................................................... 12

6-3 Adjustments to recorded dose ................................................................................................. 13

6-4 Assignment of film badges....................................................................................................... 14

6-5 Badged and nonbadged areas................................................................................................. 14

6-6 Interpretation of reported data ................................................................................................. 15

6-7 Summary of job titles as reported by workers and coded for statistics ..................................... 17

6-8 Annual average gamma and beta dose ................................................................................... 18

6-9 Default photon energy distribution for WSP materials.............................................................. 19

6-10 Energy distribution by building or area..................................................................................... 20

6-11 Photon dose units for use with organ dose conversion factors ................................................ 24

6-12 Electron dose units for use with organ dose conversion factors............................................... 24

6-13 Photon LODs for WSP dosimeters by year.............................................................................. 24

6-14 Potential missed photon dose.................................................................................................. 25

6-15 Electron LODs for WSP dosimeters by year ............................................................................ 25

6-16 Potential missed electron dose................................................................................................ 25

6-17 Tolerance limits at WSP .......................................................................................................... 26

6-18 AEC Standards........................................................................................................................26

6-19 Dose limits (rem) based on exchange frequency ..................................................................... 26

6-20 Bias and uncertainty................................................................................................................28

LIST OF FIGURES

Figure

6-1 Uranium-235 decay series.........................................................................................................

Page

8

6-2 Uranium-238 decay series.........................................................................................................9

6-3 Thorium-232 decay series ....................................................................................................... 10

6-4 Average annual gamma exposure for various job categories................................................... 18

6-5 Average annual beta exposure for various job categories........................................................ 19

6-6 Estimated beta dose rate at surface of uranium metal at various enrichment levels ................ 22

6-7 Shallow dose rate from natural uranium slab........................................................................... 23

A-1 Example of Annual Personnel Internal - External Exposure Report ......................................... 37

A-2 Example of Personnel Internal-External Radiation Summary................................................... 38

A-3 Example of Annual Personnel Internal-External Radiation Exposure Report ........................... 39

A-4 Example of Annual Personnel External Radiation Exposure Report ........................................ 40

A-5 Example employee recorded external and internal exposure record...................................... 421

A-6 Example of Quarterly External Radiation Exposure Report...................................................... 42

A-7 Example of Film Badge Data Summary by Year...................................................................... 43
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

cm centimeter

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act

g gram

keV kilovolt-electron, 1,000 electron volts

LOD limit of detection

MCW Mallinckrodt Chemical Works

MeV megavolt-electron, 1 million electron volts

mg milligram

mm millimeter

mR milliroentgen

mrem millirem

mrep millirep

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ride Associated Universities

R roentgen

TBD Technical Basis Document

TLD thermoluminescent dosimeter

U.S.C. United States Code

wk week

WSP Weldon Spring Plant

yr year

α alpha

β beta

γ gamma
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6.1 INTRODUCTION

This Technical Basis Document (TBD) provides information about documentation of historical

practices at the Weldon Spring Plant (WSP) for evaluation of external exposure data for monitored 

and unmonitored workers to be used as a supplement to or substitute for recorded individual worker 

dose. 

TBDs and Site Profile Documents are general working documents that provide guidance concerning 

the preparation of dose reconstructions at particular sites or categories of sites. They will be revised 

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

may be used to assist the National Institute for Occupational Safety and Health (NIOSH) 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 facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. Sections 7384l(5) and 

(12)].

6.1.1

The purpose of this document is to describe WSP external dosimetry systems and practices. 

Purpose

6.1.2

WSP operations played an important role in the U.S. development of nuclear power and nuclear 

weapons. Operations focused on processing of uranium and thorium from feed stocks to metal and 

intermediate products for use at other facilities. This TBD contains supporting documentation to 

assist in the evaluation of worker dose from WSP operations and processes. External Dose 

Reconstruction Implementation Guideline (NIOSH 2002) provides additional guidance.

Scope

The methods for radiation exposure measurement for workers have evolved since the beginning of 

WSP operations. An objective of this document is to provide supporting technical data to evaluate the 

external occupational dose that can reasonably be associated with WSP worker radiation exposure as 

covered under EEOICPA. The document addresses evaluation of unmonitored and monitored worker 

exposure as well as missed dose. In addition, to the extent possible with available data, this 

document includes information on measurement uncertainties and describes how the uncertainties for 

WSP exposure and dose records are evaluated. 

This TBD is one part of the WSP Site Profile. The Site Profile describes plant facilities and processes, 

historic information about occupational internal and external doses, and environmental data for use if 

recorded individual worker doses are unavailable. To the extent possible, this document provides 

necessary background information and critical data for the dose reconstructor to perform individual 

worker dose reconstructions. 

6.1.3

With few exceptions, the WSP processed uranium, but a small amount of thorium was processed near 

the end of the plant’s operations. Technical Basis Document for the Weldon Spring Plant – Site 

Description (ORAU 2005) contains a chronology of thorium work. Table 6-1 lists the source terms of 

major concern. Figures 6-1 to 6-3 show complete decay chains of 238U, 235U, and 232Th. Pa-234m is 

Dosimetry Overview
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likely the most important contributor to skin dose, because of its frequent high-energy beta emission. 

Pa-234m also emits higher energy gamma rays, albeit less frequently, than other nuclides of concern 

at WSP.

Table 6-1. Beta and gamma emissions of 

primary interest.

a

Radionuclide

Beta energy 

(MeV, max.)

Gamma 

energy (MeV)

U-238 None None

Th-234 0.10 (19%) 0.063 (3.5%)

0.193 (79%) 0.093 (4%)

Pa-234m 2.28 (99%) 0.766 (0.2%)

1.00 (0.6%)

U-235 None

0.144 (11%)

0.163 (5%)

0.186 (54%)

0.205 (5%)

Th-231

0.205 (15%)

0.287 (49%) 0.026 (15%)

0.304 (35%) 0.084 (6.5%)

U-234 None 0.053 (0.1%)

a. Source: Shleien, Slaback, and Birky (1998).

Radiation protection practices and exposures at WSP varied over time. There is no comprehensive 

description of the practices and processes available at this time. Partial descriptions have been 

discerned from several documents as discussed in the following sections. Though contemporary 

references at WSP are limited there is dose information for all years discussed.

6.1.3.1 Plant Operations Period (1957 to 1966) 

A film badge notification memorandum by the Health and Safety Department (MCW 1958) indicates 

that the WSP film badge program began on March 1, 1958. Before that time, dosimetry performed at 

WSP was more than likely provided by the MCW St. Louis plant. A memo from Brandner to Mason 

(Brandner 1956a) states that some St. Louis employees transferred to the Weldon Spring plant 

“where they are no longer being monitored for radiation exposure with film badges.” This agrees with

a footnote from individual film badge data summary sheets in 1966 that states “during start-up at 

Weldon Spring in 1958 and later, some persons were not badged because [they were] not involved in 

radiation work” (an example is shown in fig. 6A-7.)

Each employee, with the exception of “office females,” (Brandner 1956a) wore a combination film 

badge and security badge. The film monitors were changed biweekly or more often as necessary. 

Burr (1959a) indicates that for turret lathe operators, film badges were exchanged weekly on Monday 

night. However, Burr (1959b) states that “monthly exchange of film badges for all plant personnel is 

scheduled for January 30, 1959.” An undated report entitled “Personnel External Radiation 

Monitoring Program” (MCW undated) describes the MCW program. It states that “wage personnel 

film badges are exchanged monthly and salaried personnel film badges quarterly.” A 1965 Summary 

of Health Protection Practices states that “operations badges are exchanged and processed on a 

calendar month schedule, all others on a three-month schedule.” If the exchange frequency cannot 

be explicitly identified, the dose reconstructor should make the claimant-favorable assumption to use 

the most frequent exchange frequency for the period. 
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Figure 6-1. Uranium-235 decay series. (taken from HEW 1970)
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Figure 6-2. Uranium-238 decay series. (taken from HEW 1970)
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Figure 6-3. Thorium-232 decay series. (taken from HEW 1970) 

Brandner (1956b) describes the badges used at MCW as being manufactured by A. M. Samples 

Machine Company of Knoxville, Tennessee. The badge was of stainless-steel construction and held 

both security identification and radiation monitoring film. The front of the badge held the security and 

health identification information and was removable from the badge back. The front of the badge was 

shaped so that a 1-mm thick cadmium shield could be inserted to cover approximately the top 
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 11 of 43

two-thirds of the film. A similar 1-mm cadmium shield was permanently clamped onto the back of the 

badge. The film was DuPont dosimeter type 552 film packets, which contained two dental-size films 

wrapped together in a single wrapper. One of the films was apparently a DuPont type 502 and the 

other a DuPont type 510 film. Brandner deemed the type 502 film more sensitive than “most other 

films in the DuPont dosimeter series,” and three to four times as sensitive as the Eastman type V-120 

film. The net density (with density of the unexposed film deducted) of DuPont type 502 was “nearly 

proportional to the dose of any given type of radiation up to a density of 0.5 on a Welch Densichron.” 

This density supposedly corresponded to a dose of approximately 500 mR of 0.19-MeV gamma 

radiation or approximately 1,000 mrep of beta from aged uranium. These badges were changed once 

every 2 wk. 

A 1965 document summarizes site health protection practices and has the following description of film 

badges in place at the time:

The standard dosimeter is a stainless-steel badge with clip, containing an open window 

to admit soft radiation and integral cadmium shields to exclude soft radiation; single 

film packet having a usable [sic] exposure range from 50 mr [mR] to 200 mr radium 

gamma. For work with enriched uranium, a special badge is used which incorporates 

multiple filters for differential determination of radiation energies (MCW 1965). 

Personnel in operating areas of the plant and in some laboratories were required to wear badges 

continuously at work. Permanent badges were also assigned to those worker who frequently entered 

what were called “badged” areas. Spare badges were provided in available racks for those personnel

who had a casual need to enter a badged area. Fixed location badges were installed in process 

areas to provide reference data about changes in average radiation level. Use of film badges by 

visitors or subcontract personnel was predetermined by the person who authorized entry (MCW

1965).

MCW (1965) also stated that “operations badges are exchanged and processed on a calender [sic] 

month schedule, all others on a three-month schedule.” The term “operations badges” is assumed to 

refer to badges worn by personnel working in the operational, as opposed to administrative, sections 

of the plant. Ingle (1998) states that “film badge results were collected and read on a weekly basis 

until 1959 when the external program adopted a quarterly reading.” Dupree et al. (1999) stated that 

film badges were read weekly from 1945 to 1954 (which would be pre-Weldon Spring), biweekly from 

1955 to 1958 (which includes the initial startup of Weldon Spring), and monthly for production workers 

and quarterly for all other workers from 1959 to 1966. In light of this confusion, it is suggested that if 

the exchange frequency cannot be determined from the claimant file, a client-favorable exchange 

frequency of bi-weekly be assumed for all operations workers through 1958 and that an exchange 

frequency of monthly for production workers and quarterly for all other workers be assumed from 1959 

to 1969.

Belcher (1966a) described monitoring for external radiation exposure as using a “stainless steel 

non-security badge containing a DuPont 555 film.” This badge supposedly had a “useful range” of up 

to 10 R. Beta exposures were measured through the open window (40 mg/cm2

) portion of the badge 

and were compared with a uranium beta calibration curve. Gamma exposures, primarily from 

uranium progeny and thorium, were measured under the cadmium shield and were compared with a 

radium gamma calibration. Mixed beta-gamma exposures were determined by subtraction. 
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6.1.3.2 Initial Cleanup Period (1967 to 1969) 

No information is currently available to describe the external dosimetry program during the initial 

cleanup phase following cessation of operations. There is some anecdotal information to indicate that 

some former WSP workers continued their employment during this period. That being that case, it is 

likely that the same film badge system would have been used.

Belcher (1966b) comments that “any new contractor operations on-site (maintenance, equipment 

removal, etc.) will need a minimum of health protection surveillance.... We do not feel such a 

contractor will need film badge services.” However, it is not clear if this statement refers to a 

continued presence by MCW staff.

6.1.3.3 Monitoring and Maintenance Period (1969 to 1985) 

No information is currently available to describe the external dosimetry monitoring program for this 

period. 

6.1.3.4 Site Remediation Period (1985 to 2000)

During the conduct of the Weldon Spring Site Remedial Action Project, the contractors, MK-Ferguson 

Company and Jacobs Engineering Group, provided personnel with whole-body thermoluminescent 

dosimeters (TLDs) for beta-gamma monitoring. These vendor-provided dosimeters (Landauer Alnor 

Type L-1) were capable of detecting deep and shallow doses to a minimum detection level of 

10 mrem effective dose equivalent (DOE 1994). The dosimeter vendors were participants in the 

National Voluntary Laboratory Accreditation Program (DOE 2000). 

From August 1992 to September 1994, during remediation, extremity doses were measured using 

ring dosimeters. The resultant data demonstrated that extremity dosimetry was not necessary for 

most work during the remediation period with the materials on the site at that time (DOE 2000). 

6.2 DOSE RECONSTRUCTION PARAMETERS

6.2.1

Table 6-2 lists the process used to evaluate the measured film densities and to determine dose. 

Table 6-3 cites the one identified bias correction to be applied to WSP recorded dose values.

Interpreting the External Dosimetry Record

Table 6-2. Summary of historical recorded dose practices.

a

Year Dosimeter measured quantities Compliance dose quantities

Two-element film (photon + electron)b

Plant operations period

1958–1966

SWdensity

OWdensity

OWdensity,beta = OWdensity - (Gdose ÷ CFOW,gamma)

Gdose = SWdensity × CFSW,gamma

Bdose = OWdensity,beta × CFOW,beta

Plant operations period

Special case for enriched uranium

Maintenance period

Landauer

Site remediation period (DOELAP Accredited)

a. Bdose = beta dose (determined dose); CF =calibration factor determined from standard films (dose per unit density); Gdose = gamma 

dose (determined dose); OWdensity = open window (measured density); OWdensity, beta = open window density resulting from beta 

exposure; SWdensity = shielded window (measured density).

b. Source: MCW (1956).
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Table 6-3. Adjustments to recorded dose.

Period Dosimeter Facility Adjustment to reported dose

1957–1966 Two-element film WSP Estimate neutron dose as 10% of reported gamma dose in 

facilities containing UF4 or UF6.

6.2.2

The accuracy of the dosimetry system and recorded doses, and their comparability through time,

depends on administrative practices based on technical, regulatory, and administrative requirements;

dosimetry technologies and calibrations; process technologies; and training programs and practices. 

Weldon Spring Historical Administrative Practices

As mentioned, the use of a dosimeter for production workers was always employed in one form or 

another at WSP. However, exposures have not always been determined for all employees. Female 

workers were not routinely monitored (Mason 1955), at least during the early history of the site. This 

could have been because it was presumed that they would not exceed 10% of the quarterly limit as 

defined by the U.S. Atomic Energy Commission (AEC). 

6.2.2.1 Recorded Doses

WSP recorded both beta (skin) and gamma (deep) doses by determining film densities behind the 

open window and a single filter of approximately 1,000 mg/cm2

. Beta doses were recorded in units of 

millirep, and gamma in units of mR (or mr on some reports). The rep is a historical unit (the word 

derives from roentgen-equivalent-physical), which variously equated to 83 to 95 ergs/g of tissue 

(Parker 1980). In 1956, the MCW Uranium Division considered converting to the rad for both gamma 

and beta dose (Brandner 1956c). In this TBD, a rep is defined as an absorbed dose of 93 ergs/g. It 

does not appear that the conversion to rad was accomplished. It is assumed that the 93-ergs/g rep 

was used throughout the WSP production years.

It is claimant-favorable to assume that roentgens (R or r in the records), rep, and rem are equivalent. 

6.2.2.2 Discrepancies

If the employee’s record contains discrepancies, it is claimant-favorable to use the higher dose in the 

dose reconstruction. Care must be taken to interpret dose numbers properly if units were not 

specified. WSP routinely used milliroentgens or millirep as the unit of dose. Because of the tolerance 

limits in place at WSP, it is highly unlikely that a record would show a dose greater than the quarterly 

or annual limit without an additional record that indicated an overexposure. 

If no activity date is associated with a dose record, it is claimant-favorable to use that dose in the dose 

reconstruction. The dose reconstructor should use best judgment to credit the dose to the most likely 

year.

6.2.2.3 Missing Entry

A missing entry in the dosimetry history probably indicates that the individual missed the dosimeter 

exchange and that the next dosimeter includes the dose from both exchange periods. A less likely 

possibility is that the badge was lost and no dose was assigned for that period. The 

claimant-favorable assumption is that the dosimeter was lost, and dose should be assigned for that 

period using the dosimetry data from before and after that period (consider the approach of Watson et 

al. 1994).
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6.2.2.4 Badge Assignment and Exchange Frequency

Based on reviews of worker files, some individual dosimetry records are available, but the majority of 

results are quarterly totals. It is necessary to estimate the dosimeter exchange frequency from the

available programmatic information. As described in several undated memoranda, personnel whose 

work routinely required them to be in a designated film badge area were assigned permanent film 

badges. “Office females” were not routinely assigned film badges. Table 6-4 summarizes assignment 

of film badges. Table 6-5 lists badged areas and nonbadged areas. 

Table 6-4. Assignment of film badges.

Permanent badges

All MCW Uranium Division wage (hourly) personnel are assigned 

permanent film badges.

All MCW Uranium salaried personnel who regularly work in or routinely 

visit badge areas of the plant are assigned permanent film badges.

Other non-MCW personnel, such as AEC, who regularly work in or 

routinely visit badge areas of the plant are assigned permanent film 

badges.

All MCW Uranium Division personnel who work directly with enriched 

uranium materials are assigned special neutron dosimeter badges, which 

are worn in conjunction with the regular film badges.

Temporary film 

badges

Temporary film badges are provided for the use of other personnel 

(MCW visitors of otherwise) who do not normally work in badge areas but 

find it necessary to enter a badge area for a limited time.

Table 6-5. Badged and nonbadged areas.

Badge area Nonbadged area

Sampling Plant Administration Building

Refinery Service Building

Green Salt Plant Laboratories

Metal Plant Maintenance Stores

Boiler house Parking Lots

Warehouse Water Plant

Pilot Plants

Badges were picked up and returned at the end of the day by the individual workers. Wage personnel 

film badges were exchanged and processed monthly, and salaried personnel film badges were 

exchanged and processed quarterly. Individual film badge data were posted quarterly to the 

employee’s health history file. 

6.2.2.5 Interpretation of Reported Data

Table 6-6 summarizes several different formats in which health personnel recorded external dosimetry 

information for the WSP site. Many of the dosimetry reports did not specify the reporting units, but a 

June 16, 1956, memorandum from K. E. Brandner to J. W. Miller details the change from roentgen 

and rep to rad for both gamma and beta radiation (Brandner 1956c). The memorandum specifies 

that, beginning June 18, 1956, units for both gamma and beta radiation “should be standardized to the 

‘rad’ unit.” This memorandum predates the WSP site and is at odds with reports such as the Annual 

Personnel Internal-External Radiation Exposure Report shown in Figure A-3. It is claimant-favorable 

to assume that all units are rem. Fig. 6A-7 shows a film badge summary report that includes data from 

multiple years. 
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Table 6-6. Interpretation of reported data.

Report Reported quantity

Interpretation of 

zeroes

Interpretation of 

blanks (no data)

Individual 

and annual 

data

Monitored/

unmonitored

Personal Monitoring 

Summary Record

(Figures A-1 - A-6)

Typewritten summary 

of annual dosimetry 

record, including

external and internal 

data 

Annual totals in mR or mrem. 

(Units are not noted.) External γ 

and β + γ reported by year for 

multiyear period. High quarters

of β + γ are noted. Cumulative γ 

and β + γ are noted. Appear to 

be for early period of plant 

operation (1958-1962). 

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. 

Film Badge Data 

Summary. 

(Figures A-1 - A-6) 

Hand-generated 

summary of annual 

and cumulative 

external data. 

Typewritten form.

Annual totals in rad (γ) or rep (β). 

Gamma, beta, and gamma + 

beta (rad) reported by year for 

multiyear period. Form begins in 

1952 and has rows for each year 

through 1966. Form was 

designed to be used for 

Destrehan, W. S., and had notes 

for worker transfer to parent 

company with date noted. 

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. Notation on 

bottom of form indicates 

that “During start-up at 

Weldon Spring in 1958 

and later, some 

persons were not 

badged because not 

involved in radiation 

work.” 

Annual Personnel 

Internal-External 

Radiation Exposure 

Report Health & 

Safety Dept.

(Figure A-1)

For a single year, gamma 

(mrem), beta (mrem), and 

gamma + beta (mrem) are 

reported by quarter. Cumulative 

for previous year is also reported 

in same units. Number of weeks 

is also reported, but it appears 

that this number represents total 

number of weeks worked since 

initial employment. Internal 

radiation exposure is reported on 

same form.

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. 

Personnel InternalExternal Radiation 

Summary 19xx-xx.

(Figure A-2)

Computer-generated 

form for 2-yr period. 

Quarterly data for gamma and 

« gamma/beta. » No indication 

of units. Gamma/beta 

represents total external 

exposure for the period. 

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. 

Annual Personnel 

Internal-External 

Radiation Exposure 

Report. 

(Figure A-3) 

Computer-generated 

report analogous to 

typewritten report of 

same name. 

For a single year, gamma 

(mrem), beta (mrem), and total, 

noted as « gamma/beta » 

(mrem) are reported by quarter. 

Cumulative for previous year is 

also reported in same units. 

Number of weeks is also 

reported, but it appears that this 

number represents total number 

of weeks worked since initial 

employment date, which is noted 

on print out. The weekly 

average external is also 

reported. 

Internal radiation exposure is 

reported on same form.

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. 

Annual Personal 

External Radiation 

Exposure Report 

Year 19xx.

(Figure A-4) 

Computer-generated 

report.

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. 
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 16 of 43

Table 6-6 (Continued). Interpretation of reported data.

Report Reported quantity

Interpretation of 

zeroes

Interpretation of 

blanks (no data)

Individual 

and annual 

data

Monitored/

unmonitored

Recorded External 

Exposure. 

(Figure A-5) 

Computer-generated 

report showing both 

external and internal 

exposures by quarter. 

Have handwritten 

notations for external 

exposures by month 

for 1966. 

Gamma and beta and gamma by 

year. Lifetime values for each. 

No indications of units, but 

presumably are mrem. 

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. 

Quarterly External 

Radiation Exposure 

Report – Month 

19xx.

(Figure A-6) 

Computer-generated 

list for a given month. 

Includes data for 

several workers on 

same sheet.

For month of the quarter, lists 

beta & gamma and gamma 

alone. Units not specified, but 

presumed to be mrem. Month 

identified numerically. 

Year-to-date beta & gamma and 

gamma values also given for 

each employee. 

Zero likely indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during that 

period. 

6.2.3

Available worker data was analyzed in an attempt to develop a profile of exposure for each type of 

job. Job titles reported in computer-assisted telephone interviews were utilized. As shown in 

Table 6-7, there were over 70 different job titles for workers. These are categorized into nine 

categories that roughly represent the reported job titles. Table 6-8 lists the annual average gamma 

and beta exposures calculated for each category. Figures 6-4 and 6-5 show that the operator 

category received greater exposure to gamma rays in each year. Exposure to beta radiation was 

substantially greater for those in the operator category than for any other job category.

Plant-Wide Dosimetry Results

6.2.3.1 Calibration

The film badges used by MCW at the St. Louis site were calibrated using known exposures given to 

control films (Miller 1955). The same system was used by MCW at WSP. MCW (1965) states that: 

Test and calibration dosimeters are exposed to radium gamma and to uranium beta. 

Density of personal dosimeter film is compared to the calibration film curve, results are 

expressed in mr of gamma (radium equivalent) and mrep of beta (uranium equivalent). 

A direct conversion to mrad is assumed in recording personnel exposure.

6.2.4

No data is readily available to describe the workplace radiation fields at WSP. Summary reports 

indicate that natural uranium was the material that the workers came into contact with most frequently. 

Radiation fields most often consisted of a complex mixture of beta and gamma energies. Neutrons 

were potentially encountered in several buildings as described in Section 6.2.4.2. By reviewing 

personal dosimetry records, the dose reconstructor should be able to determine the relative 

magnitudes of each type of exposure. In many cases, the majority of the exposure would have 

consisted of beta particles, which can deliver substantial doses to bare skin in relatively close 

proximity to the source, but which do not penetrate deeply into the body. 

Workplace Radiation Fields
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 17 of 43

Table 6-7. Summary of job titles as reported by workers and coded for statistics. Coded job 

title Reported worker job titles

Equipment 

Operator

Fork Lift Driver

Fork Lift Operator

Warehouse Fork Lift Operator

Yard Operator

Foreman Foreman

Production Foreman

Manager

Accountant Supervisor

Product Control Supervisor

Supervisor, Plant and Maintenance Scheduler

Nonradiation 

job

Computer Operator

Industrial Nurse

Inventory Control Clerk

Maintenance and Utility Control Clerk

Office Boy/Accounting Clerk

Shipping

Safety, 

security

Production/Safety and Fire Marshall

Safety

Safety and Fire Prevention

Security Guard

Worker

Electrician

General Cleaner

Machinist

Maintenance Electrician

Maintenance & Oiler

Maintenance Electrician

Maintenance, Welder

Maintenance/Rigger

Metal Worker

Millwright

Pipefitter

Tool and Die Maker

Utility Worker

Welder

Welder, Maintenance

Welder/Metal Fabricator

Coded job 

title Reported worker job titles

Operator

Acid Recovery/Loader

Chemical D95 Operator

Chemical Operator, store keeper

Chemical Operator Pot room

Chemical Operator

Chemical Operator/Maintenance

Conversion Green Salt

Foreman-Operator

Machine Operator

Machinist, Operator

Metal Plant, Manufacturer

Operator

Operator, Decontamination, Maintenance

Operator/Labor

Pot Room Worker

Press Operator Refinery 103

Processing Plant

Production

Production Operator

Production Operator A

Refinery Operator

Uranium Processor

Utility Operator

Water Plant, refinery

Chemical and Project Engineer

Engineer

Chemical Engineer

Engineer and Production Control

Mechanical Engineer

Plant Engineer

Process Engineer

Analytical Chemist

Laboratory 

worker

Chemical Technician

Laboratory Technician

Laboratory Technician, Engineer

Research Chemist

6.2.4.1 Gamma Dose

No data have been found to indicate the gamma spectra in WSP work areas. However, nearly all the 

material processed at WSP was natural, slightly enriched, or depleted uranium. It appears from the 

records that depleted and enriched uranium were routinely handled with some shielding, but the type 

and amounts of shielding are not now known. Enriched and depleted uranium are assumed to have 

been relatively fresh with little or no ingrowth of decay products having occurred at the time that the 

material was processed at WSP. 

However, 234mPa is a decay product in the 238U decay chain and emits a 2.29-MeV beta particle. 

Therefore, there are a significant number of photons from bremsstrahlung, and they contribute 

photons of intermediate energy (30 to 250 keV). Bremsstrahlung radiation can contribute up to 40% 

of the photon dose from uranium metal (DOE 2001). This decay product grows in fairly rapidly and is 

present in equilibrium quantities for most depleted uranium that was processed at WSP. It is 

appropriate to use the default assumption for depleted uranium that 50% of the dose is contributed by 

photons in the 30-to-50-keV photon energy range and 50% of the dose is a result of exposure from 

photons in the above-250-keV range. 

Although enriched uranium has significantly less ingrowth of 234mPa, 235U and its decay products emit 

a 185.7-keV photon 57% of the time and a 143.8-keV photon 11% of the time. These photons 

dominate the measured photon energy spectra. Therefore, for enriched uranium, it is appropriate and 
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 18 of 43

Table 6-8. Annual average gamma and beta dose (mR).

Annual average gamma exposure by year

Job description 1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Engineer 110 102 43 121 83 75 94 244 170 94

Equipment Operator 135 - 135 168 154 151 176 177 164 278

Foreman 233 85 150 183 190 186 154 130 192 102

Laboratory worker 220 89 71 71 134 175 204 285 575 155

Manager 52 - 37 517 102 41 142 112 78 64

Non-rad job 125 - 48 96 105 72 93 226 181 33

Operator 305 129 151 234 369 371 640 516 413 298

Safety, security 168 75 265 59 65 83 415 273 119 183

Unknown 411 48 85 141 170 223 184 294 292 248

Worker 240 132 67 142 140 216 317 303 198 177

Annual average beta exposure by year

Job description 1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Engineer 228 23 191 579 486 572 277 308 186 90

Equipment Operator 190 1 878 1170 710 788 283 279 514 190

Foreman 328 45 492 1293 804 919 1336 735 637 257

Laboratory worker 382 133 343 463 796 724 367 327 340 278

Manager 107 35 199 1160 1099 403 204 128 91 181

Non-rad job 248 135 123 378 538 204 224 243 142 50

Operator 761 274 1122 2642 2695 1648 1309 1018 884 297

Safety, security 198 60 94 170 301 463 107 98 197 91

Unknown 524 128 297 890 1427 1146 338 401 453 205

Worker 697 93 472 923 1070 1107 780 354 473 317

Gamma exposure by year for nine job types

0

100

200

300

400

500

600

700

1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Year

Gamma exposure, mR

Engineer Equipment Operator Foreman Laboratory worker Manager

Non-rad job Operator Safety, security Worker

Figure 6-4. Average annual gamma exposure for various job categories. 
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 19 of 43

Beta exposure by year for nine job types

0

500

1000

1500

2000

2500

3000

Year

Beta exposure, mR

Engineer Equipment Operator Foreman Laboratory worker Manager

Non-rad job Operator Safety, security Unknown Worker

Figure 6-5. Average annual beta exposure for various job categories.

claimant-favorable to use the default assumption that the entire photon dose is a result of exposure in 

the 30-to-250-keV photon energy range. Table 6-9 shows the default assumptions. Table 6-10 lists 

energy distributions for WSP buildings.

Table 6-9. Default photon energy distribution for WSP materials.

Energy

Natural 

uranium

Depleted 

uranium

Slightly enriched 

uranium

Natural 

thorium

<30 keV 0% 0% 0%

30-250 keV 50% 100% 25%

>250 keV 50% 0% 75%

6.2.4.2 Neutron Dose

Although no neutrons were anticipated or measured with the WSP film badge, it is possible that 

neutrons from the alpha, neutron reaction from UF4 and UF6 could have contributed dose to WSP

workers. The analysis performed for the similar situation at Fernald (ORAU 2004a) is appropriate and 

will be used here.

Using the results of gamma and neutron dose rate measurements performed on depleted and lowenriched UF4 drums, a neutron-to-gamma ratio was developed. Natural uranium was addressed as 

well. The results of this analysis were that a neutron-to-gamma ratio of 0.1, lognormally distributed 

with a geometric standard deviation of 1.71 and an upper 95% ratio limit of 0.23, should be applied in 

those areas where there is the potential for neutron dose from uranium fluoride compounds. 

Table 6-10 lists energy distributions for WSP buildings.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 20 of 43

Table 6-10. Energy distribution by building or area. Building Description Radiation Energy Percentage

101 Sampling Plant Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

Area 102 A&B Refinery Tank Farm Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

103 Digestion and Denitration Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

Natural Th

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

25%

75%

Slightly enriched U

Electron

Photon

>15 keV

30 - 250 keV

100%

100%

104 Lime Storage None

105 Extraction Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

Natural Th

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

25%

75%

Slightly enriched U

Electron

Photon

>15 keV

30 - 250 keV

100%

100%

106 Refinery sewer sampling Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

108 Nitric acid plant Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

109, 110 West Drum Storage, East Drum 

Storage

Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

201 Green Salt Building Natural U

Electron

Photon

Neutron

>15 keV

30 - 250 keV

>250 keV

0.1 - 2 MeV

100%

50%

50%

100%

Natural Th

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

25%

75%

Slightly enriched U

Electron

Photon

>15 keV

30 – 250 keV

100%

100%

202 A&B Green Salt Tank Farm None

6.2.4.3 Electron Dose

Beta radiation fields are usually the dominant external radiation hazard in facilities that involve contact 

work with unshielded forms of uranium. This was the case at WSP for natural and depleted uranium 

work. The most common exposure at WSP was to natural uranium, but depleted uranium was also 

present at the site on an intermittent basis. Slightly enriched uranium (less than 1% 235U by weight)

was also present at times in the form of scrap metal or residues.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 21 of 43

Table 6-10 (Continued). Energy distribution by building or area.

Building Description Radiation Energy Percentage

301 Metals Building Natural U

Electron

Photon

Neutron

>15 keV

30 – 250 keV

>250 keV

0.1 – 2 MeV

100%

50%

50%

100%

Natural Th

Electron

Photon

>15 keV

30 – 250 keV

>250 keV

100%

25%

75%

Slightly enriched U

Electron

Photon

>15 keV

30 – 250 keV

100%

100%

302 Magnesium Building None

Pad 303 Material Storage Pad None

401 Steam Plant None

403 Chemical Pilot Plant Natural U

Electron

Photon

Neutron

>15 keV

30 – 250 keV

>250 keV

0.1 – 2 MeV

100%

50%

50%

100%

Depleted U

Electron

Photon

Neutron

>15 keV

30 – 250 keV

>250 keV

0.1 – 2 MeV

100%

40%

60%

100%

404 Metallurgical Pilot Plant Natural U

Electron

Photon

Neutron

>15 keV

30 – 250 keV

>250 keV

0.1 – 2 MeV

100%

50%

50%

100%

Depleted U

Electron

Photon

Neutron

>15 keV

30 – 250 keV

>250 keV

0.1 – 2 MeV

100%

40%

60%

100%

405A & B Pilot Plant Maintenance Natural U

Electron

Photon

Neutron

>15 keV

30 – 250 keV

>250 keV

0.1 – 2 MeV

100%

50%

50%

100%

406 Warehouse Natural Th

Electron

Photon

>15 keV

30 – 250 keV

>250 keV

100%

25%

75%

407 Laboratory Natural U

Electron

Photon

Neutron

>15 keV

30 - 250 keV

>250 keV

0.1 - 2 MeV

100%

50%

50%

100%

408 Maintenance and Stores None

409 Administration None

410 Services Building None

412 Electrical Substation None

413 Cooling Tower and Pump House None

414 Salvage Building Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

415 Process Incinerator Natural U

Electron

Photon

>15 keV

30 - 250 keV

>250 keV

100%

50%

50%

417 Paint Shop None

426 Water Tower None

427 Primary Sewage Treatment Plant none

428 Fuel Gas Plant None
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 22 of 43

Table 6-10 (Continued). Energy distribution by building or area.

Building Description Radiation Energy Percentage

429 Water Reserve Facilities None

430 Ambulance Garage None

43l Laboratory Sewer Sampler Natural U

Electron

Photon

Neutron

>15 keV

30 - 250 keV

>250 keV

0.1 - 2 MeV

100%

50%

50%

100%

432 Main Sewer Sampler None

437 Records Retention Building None

439, 443 Fire Training and Storage 

Building

None

441 Cylinder Storage None

Figure 6-6 shows estimated beta dose rates from a semi-infinite slab of uranium metal at various 

enrichment levels. For uranium enrichments up to 30%, the beta radiation field is dominated by 

contributions from 238U decay products. Therefore, for depleted uranium, the most energetic 

contributor to the beta exposure is the 2.29-MeV (maximum energy) beta particle from 234mPa.

Figure 6-6. Estimated beta dose rate at surface of uranium metal at 

various enrichment levels (DOE 2001).

Processes that separate and sometimes concentrate beta-emitting uranium progeny are not 

uncommon in U.S. Department of Energy (DOE) uranium facilities. Surface beta dose rates on the 

order of 1 to 20 rad/hour have been observed at some DOE facilities. Exposure control is 

complicated by the fact that considerable contact work takes place in facilities that process uranium 

metal. At MCW, and presumably WSP, chronic overexposure of workers’ hands was a serious 

problem (Mason 1955). Many operations required contact between the hands and the radioactive 

materials, and the glove program was “sketchy and inadequate” (Mason 1955).

The beta spectrum from uranium is highly dependent on the quantity of progeny in the uranium, which 

in turn is dependent on the enrichment level. Depleted uranium progeny grow into secular equilibrium 

relatively quickly (about 30 days); it is conservative to assume that progeny would have been present 
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 23 of 43

at these levels. Figure 6-7 shows the relative dose rate in relation to energy. Depleted uranium 

would be similar to the natural uranium used for this experiment.

Figure 6-7. Shallow dose rate from natural uranium slab (DOE 2001).

Although depleted uranium, slightly enriched uranium, and natural thorium were present in the waste 

stream processing buildings, the dose from these materials would be small in comparison to natural 

uranium because of the predominance of the latter (more than 97%) that was processed at the plant.

Table 6-10 lists energy distributions for WSP buildings.

6.2.4.4 Reported Dose-to-Organ-Dose Conversion Factor Units

The roentgen was the unit of calibration. It is reasonable to assume that this continued throughout the 

life of the WSP film dosimetry system. Little is known about the dosimetry system between plant 

shutdown and the remediation period. Calibration of the dosimetry system consistent with the DOE 

Laboratory Accreditation Program (DOELAP) was utilized during the remediation period. Thus, the 

personal dose equivalent [Hp(10)] is the appropriate unit to use for the remediation period. Tables 6-

11 and 6-12 show these units.

6.2.4.5 Limit of Detection

Miller (1955) describes an investigation of calibration data. The badge was very similar to that used 

throughout the early weapons program and it was likely the same as that used at Hanford and 

Fernald. A Pacific Northwest National Laboratory study of this two-element dosimeter identified a 

detection level of about 40 mR at the upper 95% confidence level for radium gamma radiation (ORAU 

2004b). The Fernald TBD (ORAU 2004a) cites a minimum detection limit of 30 to 40 mrem, but it 

does not give a source for that information. The MCW St. Louis Plant TBD (ORAU 2003a) cites 

records in which gamma dose results are shown as "50*" where the asterisk refers to a footnote that 
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 24 of 43

Table 6-11. Photon dose units for use with organ dose conversion factors.

Year Unit Year Unit Year Unit Year Unit Year Unit

1957 R 1967 R 1977 1987 1997 Hp(10)

1958 R 1968 R 1978 1988 1998 Hp(10)

1959 R 1969 R 1979 1989 1999 Hp(10)

1960 R 1970 1980 1990 2000 Hp(10)

1961 R 1971 1981 1991

1962 R 1972 1982 1992 Hp(10)

1963 R 1973 1983 1993 Hp(10)

1964 R 1974 1984 1994 Hp(10)

1965 R 1975 1985 1995 Hp(10)

1966 R 1976 1986 1996 Hp(10)

Table 6-12. Electron dose units for use with organ dose conversion factors.

Year Unit Year Unit Year Unit Year Unit Year Unit

1957 rad 1967 rad 1977 1987 1997 H'(0.07)

1958 rad 1968 rad 1978 1988 1998 H'(0.07)

1959 rad 1969 rad 1979 1989 1999 H'(0.07)

1960 rad 1970 1980 1990 2000 H'(0.07)

1961 rad 1971 1981 1991

1962 rad 1972 1982 1992 H'(0.07)

1963 rad 1973 1983 1993 H'(0.07)

1964 rad 1974 1984 1994 H'(0.07)

1965 rad 1975 1985 1995 H'(0.07)

1966 rad 1976 1986 1996 H'(0.07)

reads, "indicates less than." Values of 60 and 80 with asterisks are sometimes found in the beta 

column. Based on this information, it is reasonable to adopt a claimant-favorable LOD for the WSP 

film dosimeter of 50 mR gamma and 80 mrep beta. Landauer (the manufacturer) typically quotes a 

minimum detection level of 10 mrem and does not report doses less than this level (DOE 1994). 

Tables 6-13 to 6-16 show these data.

Table 6-13. Photon LODs for WSP dosimeters by year.

Year LOD Year LOD Year LOD Year LOD Year LOD

1957 50 mR 1967 1977 1987 1997 10 mrem

1958 50 mR 1968 1978 1988 1998 10 mrem

1959 50 mR 1969 1979 1989 1999 10 mrem

1960 50 mR 1970 1980 1990 2000 10 mrem

1961 50 mR 1971 1981 1991

1962 50 mR 1972 1982 1992

1963 50 mR 1973 1983 1993

1964 50 mR 1974 1984 1994 10 mrem

1965 50 mR 1975 1985 1995 10 mrem

1966 50 mR 1976 1986 1996 10 mrem

6.2.4.6 Exchange Frequency

Based on the historical evidence discussed in Section 6.1.3.1, it is claimant-favorable to assume that 

dosimeters were exchanged biweekly through 1958 and then monthly for operations workers and 

quarterly for all other workers during the WSP operational and initial cleanup periods.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 25 of 43

Table 6-14. Potential missed photon dose.

Period 

of use Dosimeter LOD

Exchange 

frequency

Max. annual 

missed dosea

1957-1958 Two-element film 50 mR Weekly (n=52) 1,300 mR

Biweekly (n=24) 600 mR

1959-1969 Two-element film 50 mR Monthly (n=12) 300 mR

Quarterly (n=4) 100 mR

1975-1988 Monthly (n=12)

Quarterly (n=4)

1989-2000 Landauer Alnor Type L-1 10 mR Monthly (n=12) 60 mrem

Quarterly (n=4) 20 mrem

a. Maximum annual missed dose calculated using 

(minimum detection limit × exchange frequency) ÷ 2, from OCAS-IG-001 (NIOSH 2002).

Table 6-15. Electron LODs for WSP dosimeters by year.

Year LOD Year LOD Year LOD Year LOD Year LOD

1957 80 mrep 1967 1977 1987 1997 10 mrem

1958 80 mrep 1968 1978 1988 1998 10 mrem

1959 80 mrep 1969 1979 1989 1999 10 mrem

1960 80 mrep 1970 1980 1990 2000 10 mrem

1961 80 mrep 1971 1981 1991

1962 80 mrep 1972 1982 1992

1963 80 mrep 1973 1983 1993

1964 80 mrep 1974 1984 1994 10 mrem

1965 80 mrep 1975 1985 1995 10 mrem

1966 80 mrep 1976 1986 1996 10 mrem

Table 6-16. Potential missed electron dose.

Period 

of use Dosimeter LOD

Exchange 

frequency

Max. annual 

missed dosea

1957-1966 Two-element film 80 mrep

Weekly (n=52) 2,080 mrep

Semimonthly (n=24) 960 mrep

Monthly (n=12) 480 mrep

1975-1988

Semimonthly (n=24)

Monthly (n=12)

Quarterly (n=4)

1989-2000 Landauer Alnor Type L-1 10 mrem Monthly (n=12) 60 mrem

a. Maximum annual missed dose calculated using 

(minimum detection limit × exchange frequency) ÷ 2, from OCAS-IG-001 (NIOSH 2002).

6.2.4.7 Number of Zero Readings

If an individual’s job assignment cannot be determined, the dose reconstructor should use the most 

frequent dosimeter exchange rate used during that year, which is claimant-favorable.

Table 6-17 lists tolerance limits in use at MCW, and presumably WSP, according to Mason (1955). 

The goal was to keep each individual's cumulative exposure to no greater than one-half the tolerance 

limit when aggregated over a 3-month period. Table 6-18 lists AEC standards for protection from 

external radiation that were in effect during the period of WSP operations (AEC 1963). Table 6-19 

divides these Federal dose limits into the badge exchange period. Reconstructors should use 

dosimetry records, if available, to determine or estimate the exchange frequency. Using the 

methodology of NIOSH (2002), it is possible to develop a claimant-favorable estimate of the number 

of zeros and ultimately the missed dose.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 26 of 43

Table 6-17. Tolerance limits at WSP.

Type of exposure

WSP tolerance 

limit per week

Beta to whole or partial body 500 mrep/wk

Gamma to whole or partial body 300 mR/wk

Beta & gamma to whole or partial body 500 mrep/wk

Hands and forearms 1500 mrep/wk

Table 6-18. AEC standards.

Type of exposure Period of time Dose (rem)

Whole body, head and trunk, active bloodforming organs, gonads, or lens of eye

Accumulated dose 5 * (N – 18)

Calendar quarter or 

13 consecutive wk 3

Skin of whole body and thyroid

Year 30

Calendar quarter or 

13 consecutive wk 10

Hands and forearm, feet and ankles

Year 75

Calendar quarter or 

13 consecutive wk 25

Table 6-19. Dose limits (rem) based on exchange frequency.

Year Limit

Exchange period

Biweekly Monthly Quarterly

1957-1958 500 mrep/wk 1 rep

1959-1966 500 mrep/wk 2.167 rep 6.5 rep

1967-2000 5 rem/yr 0.417 rem 1.25 rem

6.2.4.8 Determination of Missed Dose

Determination of missed dose is performed using LOD/2 times the number of zero readings, as 

discussed in Section 2.1.2.2 of NIOSH (2002). If the number of zero readings is indeterminate, it can 

be estimated under the assumption that prorated dose limits were not exceeded.

6.2.4.9 Unmonitored Energy Range

The two-element film dosimeter used at WSP was similar to those used at other sites. The Savannah 

River Site TBD (ORAU 2003b) discussed the response of this dosimeter. The dosimeter (shielded 

window) was calibrated with radium photons. The penetrating dose was evaluated by the response 

behind the cadmium metal filter. This heavy-metal filter attenuated the lower energy photons and 

should have resulted in an underestimated response behind that filter for measured dose and Hp(10). 

Because most, but not all, penetrating radiations are above 30 keV, it is suggested that adjustments 

are necessary to satisfy dose reconstruction criteria of recorded penetrating whole-body doses due to 

the contribution to Hp(10) from low-energy photons, which include the L-X-rays from both uranium 

and thorium. It is estimated that a correction equal to 10% of the less-than-250 keV values be added 

to the Hp(10) dose due to the contribution of these low-energy photons to penetrating dose that would 

have been absorbed by the thick filter. This is the same approach taken in the Fernald Environmental 

Management Project Occupational External Dose TBD (ORAU 2004a).

The DOELAP accreditation of the Landauer dosimeter system was based on a range of DOELAP 

exposure categories (DOE 1994). The response of the dosimeter was evaluated in relation to these 

exposures. Therefore, the Landauer dosimeter system is unlikely to have missed photon dose in an 
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 27 of 43

energy range to which workers could have been exposed. No correction for missed dose is 

appropriate for this dosimetry system.

6.2.5

The film dosimeter used at WSP had variant angular response. Dosimeters were not always exposed 

perpendicularly, which resulted in varying responses in relation to actual worker exposure. This 

dependence was considered as one factor when the response of the Hanford film badge was 

evaluated. This factor is included in the overall bias provided elsewhere in this TBD.

Angular Dependence

6.2.6

6.2.6.1 Film

Uncertainty

MCW used film to measure photons between 1957 and 1966. The film was DuPont dosimeter type 

552 film packets, which contained a DuPont type 502 film and a DuPont type 510 film. DuPont 502 

film had a useful range from 10 or 20 mR up to approximately 10 R (NRC 1989). 

A limited review of the calibration data developed from standard films developed with each batch was 

performed at MCW’s St. Louis plant (Miller 1955). It is reasonable to assume that similar variability 

existed in the film badge processing at WSP. This study provides an estimate of the laboratory 

random error associated with processing the film badges; it cited a ±50 mR maximum error at a 

125-mR gamma calibration exposure. Therefore, a 40% error (95% upper bound) is assigned for the 

random uncertainty.

Hanford performed an evaluation of the two-element film dosimeter in a variety of exposure 

environments (ORAU 2004b). The factors considered included:

• Exposure geometry

• Energy response

• Mixed fields

• Missed dose

• Environmental effects

The exposure environment most appropriate to WSP is the fuel fabrication facility, in which workers 

were exposed to beta and gamma radiation from uranium. The identified bias factor [ratio of Hp(10)

to recorded whole-body photon dose] ranges from 0.5 to 1.6. These are multiplicative factors 

[reported dose × bias factor = Hp(10)] and are appropriate to use for WSP doses. The midpoint of

this bias range is close to 1, and it is therefore not appropriate to apply a bias based on these factors.

The systematic uncertainty factor determined for the Hanford dosimeter (ORAU 2004a) is appropriate 

to use for the WSP dosimeter as well.

6.2.6.2 Thermoluminescent Dosimetry System

The Landauer TLD dosimetry system used during the remediation period was accredited by DOELAP. 

To meet accreditation requirements, the system passed performance testing consistent with DOE 

(1986). This standard allows a total error (precision + accuracy) of no more than 30%. Therefore, the 

worst would be a total bias of 30% or a total accuracy error of 30% (see Table 6-20).
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 28 of 43

Table 6-20. Bias and uncertainty.

Site-specific dosimetry system

Bias magnitude and range Uncertainty factors

Overall bias Range in bias Systematic Randoma

Two-element film 1957-1969 (photon) 1.0 0.5 - 1.6 1.2 1.4b

Two-element film 1957-1969 (electron) 1.0 0.5 - 1.6 1.2

Landauer 1.0 NA 1.3c 1.3c

a. 95% upper (or lower) bound on normal distribution.

b. From Miller (1955).

c. Based on DOELAP performance standard.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 29 of 43

AEC (U.S. Atomic Energy Commission), 1963, Standards for Radiation Protection, Appendix 0524, 

Washington, DC, August 12.

REFERENCES

Belcher, F. H., 1966a, “Report of Annual Health Protection Review at Mallinckrodt Chemical Works, 

Uranium, Division, Weldon Spring, 1965,” memorandum to H. M. Roth (Director, Research and 

Development Division, Oak Ridge Operations), with attached report dated October 1965, 

February 11, Atomic Energy Commission, St. Louis Area Office, Missouri.

Belcher, F. H., 1966b, “Weldon Spring Health Protection Discussion,” memorandum to H. M. Roth 

(Director, Research and Development Division, Oak Ridge Operations), November 2, Atomic 

Energy Commission, St. Louis Area Office, Missouri.

Brandner, K. E., 1956a, “Film Badge Procedures for Transfers to Weldon Spring and New Hires for 

Weldon Spring,” memorandum to M. G. Mason, October 26, 1956, Mallinckrodt Chemical 

Works, Uranium Division, St. Louis, Missouri.

Brandner, K. E., 1956b, “Summary of the Mallinckrodt Chemical Works Film Badge Program,” 

memorandum to J. W. Miller, May, Mallinckrodt Chemical Works, Uranium Division, St. Louis, 

Missouri.

Brandner, K. E., 1956c, “Conversion to the "rad" Unit,” memorandum to J. W. Miller, June 16, 1956, 

Mallinckrodt Chemical Works, Uranium Division, St. Louis, Missouri. 

Burr, R. H, 1959a, “Exchange of Turret Lathe Operators Film Badges,” memorandum to W. L. Utnage, 

January 26, 1959, Mallinckrodt Chemical Works, Uranium Division, St. Louis, Missouri.

Burr, R. H, 1959b, “Exchange of Film Badge Boards on January 30, 1959,” memorandum to W. L. 

Utnage, January 29, 1959, Mallinckrodt Chemical Works, Uranium Division, St. Louis, 

Missouri.

DOE (U.S. Department of Energy), 1986, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE/EH-0027, Office of Environment, Safety and 

Health, Washington, D.C., December.

DOE (U.S. Department of Energy), 1994, External Dosimetry Technical Basis Manual: Weldon Spring 

Site Remedial Action Project, DOE/OR/21548-44-1, Revision 0, Oak Ridge Operations Office, 

Oak Ridge, Tennessee, April.

DOE (U.S. Department of Energy), 2000, External Dosimetry Technical Basis Manual: Weldon Spring 

Site Remedial Action Project, Revision 6, DOE/OR/21548-441, Oak Ridge Operations Office, 

Oak Ridge Tennessee, January.

DOE (U.S. Department of Energy), 2001, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136, Change Notice No. 3, Washington, D.C.

Dupree, E. E., J. P. Watkins, J. N. Ingle, and J. A. Phillips, 1999. External Radiation Exposure and 

Mortality Among a Cohort of Uranium Processing Workers, Dupree Ellis, E. Watkins, J. P., 

Ingle, J. N., and Phillips, J. A., Oak Ridge Associated Universities, Oak Ridge, Tennessee, 

undated.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 30 of 43

HEW (U.S. Department of Health, Education and Welfare), 1970, “Radiological Health Handbook,” 

Public Health Service, Food and Drug Administration, Bureau of Radiological Health, Rockville 

Maryland.

Ingle, N., 1998, “Mallinckrodt Chemical Works Dose Estimate Methodology,” memorandum to Dr. 

Betsy (E.E.) Dupree, Oak Ridge Associated Universities, from N. Ingle, February 4, Delta-21 

Resources, Inc., location unknown. 

MCW (Mallinckrodt Chemical Works), 1958, “Weldon Spring Plant Film Badge Program,” 

memorandum to W. L. Utnage, no date, Uranium Division, Health and Safety Department, 

Weldon Spring, Missouri.

MCW (Mallinckrodt Chemical Works), 1956, Film Badge, Technique and Calibration, November 1, 

Uranium Division, St. Louis, Missouri.

MCW (Mallinckrodt Chemical Works), 1965, Summary of Health Protection Practices (Revised 1965), 

Uranium Division, St. Louis, Missouri.

MCW (Mallinckrodt Chemical Works), undated, Personnel External Radiation Monitoring Program 

Uranium Division, St. Louis, Missouri.

Mason, M. G., 1955, Mallinckrodt Chemical Works Health Program: Policies, Status and Summary,

February 21.

Miller, J. W. 1955. Investigation of Film Badge Standards Calibration Data. Memo to M. G. Mason. 

Mallinckrodt Chemical Works, St. Louis, Missouri, February 3, 1955. 

Miller, J. W., 1956, “Summary of the Mallinckrodt Chemical Works Film Badge Program,” 

memorandum to K. E. Brandner, June 16, Mallinckrodt Chemical Works, Uranium Division, St. 

Louis, Missouri.

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 1, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

NRC (National Research Council), 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, 

National Academy Press, Washington D.C.

ORAU (Oak Ridge Associated Universities), 2003a, Technical Basis Document: Basis for 

Development of an Exposure Matrix for the Mallinckrodt Chemical Company St. Louis 

Downtown Site, St. Louis Missouri, Period of Operation: 1942-1958, ORAUT-TKBS-0005, 

Rev. 00, Oak Ridge, Tennessee, October.

ORAU (Oak Ridge Associated Universities), 2003b, Technical Basis Document for the Hanford Site –

Occupational External Dosimetry, ORAUT-TKBS-0006-6, Rev. 01, Oak Ridge, Tennessee, 

January 9.

ORAU (Oak Ridge Associated Universities), 2004b, Technical Basis Document for the Savannah 

River Site, ORAUT-TKBS-0004, Rev. 02, Oak Ridge, Tennessee, October 29.
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ORAU (Oak Ridge Associated Universities), 2004a, Technical Basis Document for the Fernald 

Environmental Management Project – Occupational External Dose, ORAUT-TKBS-0017-6, 

Rev. 00, Oak Ridge, Tennessee, April 20.

ORAU (Oak Ridge Associated Universities), 2005, Technical Basis Document for the Weldon 

Spring Plant – Site Description, ORAUT-TKBS-0028-2, Revision 00, Oak Ridge, 

Tennessee.

Parker, H. M., 1980, “Health Physics, Instrumentation and Radiation Protection, Hanford Engineer 

Works,” Health Physics, volume 38, pp. 957–996.

Shleien, B., L. A. Slaback, Jr., and B. K. Birky, 1998, Handbook of Health Physics and Radiological 

Health, 3rd Edition, Lippincott, Williams and Wilkins, Philadelphia, Pennsylvania. 

Watson, J. E., J. L. Wood, W. G. Tankersley, and C. M. West, 1994, "Estimation of Radiation Doses 

for Workers without Monitoring Data for Retrospective Epidemiologic Studies," Health Physics, 

volume 67, number 4, pp. 402–405.
Document No. ORAUT-TKBS-0028-6 Revision No. 00 Effective Date: 06/24/2005 Page 32 of 43

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 and the Energy Research and Development Administration.

beta dose

A designation (i.e., beta) on some external dose records referring to the dose from lessenergetic beta, X-ray, and/or gamma radiation (see open window, or shallow dose). 

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) of the direct 

fission products emit beta radiation. Physically, the beta particle is identical to an electron 

moving at high velocity.

deep absorbed dose

The absorbed dose at the depth of 1.0 cm in a material of specified geometry and 

composition.

dose equivalent (H)

The product of the absorbed dose D, the quality factor Q, and any other modifying factors. 

The special unit is the rem. When D is in gray, H is in sieverts, where 1 sievert is 100 rem.

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See film dosimeter.)

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation. 

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, or 

extremities. This includes the fabrication, assignment, and processing of dosimeters as well 

as interpretation and documentation of the results.

exchange period (frequency)

Period (weekly, semimonthly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

As used in the technical sense, exposure refers to a measure expressed in roentgens of the 

ionization produced by photons (i.e., gamma and X-rays) in air. 

extremity

That portion of the arm extending from and including the elbow through the fingertips and that 

portion of the leg extending from and including the knee and patella through the tips of the 

toes.
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field calibration

Dosimeter calibration based on radiation types, intensities, and energies in the work 

environment.

film

In general, a packet that contains one or more pieces of film in a light tight wrapping. When 

developed, the film has an image caused by radiation that can be measured using an optical 

densitometer.

film density

See optical density.

film dosimeter

A small packet of film within a holder that attaches to a wearer.

fission

The splitting of a heavy atomic nucleus accompanied by the release of energy.

fissionable

Material capable of undergoing fission.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

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

physically identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus. 

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

isotope

Elements having the same atomic number but different atomic weights; identical chemically 

but having different physical and nuclear properties.

neutron

A basic particle that is electrically neutral and weighs nearly the same as the hydrogen atom.

open window

Designation on film dosimeter reports that implies the use of little (i.e., only security credential) 

shielding. Commonly used to label the film response corresponding to the open window area. 

operating area

Designation of major onsite operational work areas.

optical density

The quantitative measurement of photographic blackening; density defined as D = Log10 (Io/I).
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personal dose equivalent Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 and 10 millimeters

for the skin and body, respectively. These are noted as Hp(0.07) and Hp(10), respectively. 

photon

A unit or "particle" of electromagnetic radiation consisting of X- or gamma rays. 

photon X-ray

Electromagnetic radiation of energies between 10 and 100 kilovolts-electron whose source 

can be an X-ray machine or radioisotope.

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

radiation

Alpha, beta, neutron, and photon radiation. 

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, or

neutrons from unstable nuclei.

radionuclide

A radioactive isotope of an element, distinguished by atomic number, atomic weight, and 

energy state.

rem

A unit of dose equivalent equal to the product of the number of rad absorbed and the quality 

factor.

roentgen

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X-) rays that produces a total charge of 2.58 × 104 coulomb in 1 kilogram of dry air. 

An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher (about 100 kilovolts-electron or more) energy photons.

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.007 centimeter in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.007 centimeter in tissue.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation.

skin dose

Absorbed dose at a tissue depth of 7 milligrams per square centimeter.
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whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 centimeter (1,000 milligrams 

per square centimeter); however, also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin.
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ATTACHMENT A

Example Reports

LIST OF FIGURES

Figure

A-1 Example of Annual Personnel Internal - External Exposure Report .........................................

Page

37

A-2 Example of Personnel Internal-External Radiation Summary................................................... 38

A-3 Example of Annual Personnel Internal-External Radiation Exposure Report ........................... 39

A-4 Example of Annual Personnel External Radiation Exposure Report ........................................ 40

A-5 Example employee recorded external and internal exposure record...................................... 421

A-6 Example of Quarterly External Radiation Exposure Report...................................................... 42

A-7 Example of Film Badge Data Summary by Year...................................................................... 43
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Figure A-1. Example of Annual Personnel Internal - External Exposure Report.
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Figure A-2. Example of Personnel Internal-External Radiation Summary.
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Figure A-3. Example of Annual Personnel Internal-External Radiation Exposure Report.
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Figure A-4. Example of Annual Personnel External Radiation Exposure Report.
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Figure A-5. Example employee recorded external and internal exposure 

record.
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Figure A-6. Example of Quarterly External Radiation Exposure Report.
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Figure A-7 Example of Film Badge Data Summary by Year.