ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 45

DOE Review Release 02/14/2013

Document Title:

Weldon Spring Plant – Occupational External 

Dosimetry

Document Number: ORAUT-TKBS-0028-6

Revision: 01

Effective Date: 02/06/2013

Type of Document: TBD

Supersedes: Revision 00

Subject Expert(s): David P. Harrison and Leo G. Faust

Approval: Signature on File Approval Date: 01/29/2013

David P. Harrison, Document Owner

Concurrence: Signature on File Concurrence Date: 01/29/2013

John M. Byrne, Objective 1 Manager

Concurrence: Scott R. Siebert Signature on File for Concurrence Date: 01/30/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 01/29/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 02/06/2013

Stuart L. Hinnefeld, Director DCAS 

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-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 2 of 45

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

06/24/2005 00 First approved issue of the technical basis document for the Weldon 

Spring Plant, Section 6, Occupational External Exposure. 

Incorporates formal internal and NIOSH review comments. Training 

is not required. Initiated by Robert Meyer.

02/06/2013 01 Revision to remove all surrogate data except the data from MCW. 

Added new material located since the last revision. Included a 

correction factor to account for worker-radiation source geometry to 

be applied to the measured and missed photon dose for operators, 

material handlers and trade workers to avoid an underestimate of 

dose. Organs most affected by this are those in the lower torso 

(stomach, pancreas, ovaries, etc.). DCAS-TIB-0013 was included as 

a reference and used to determine this geometric correction factor of 

2.1. Revision incorporates responses and commitments based on 

SC&A and Advisory Board Working Group issues. Incorporates 

formal internal and NIOSH review comments. Constitutes a total 

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

Objective Manager. Initiated by David P. Harrison.
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 3 of 45

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

6.1.1 Purpose........................................................................................................................ 7

6.1.2 Scope........................................................................................................................... 7

6.2 Dosimetry Overview................................................................................................................. 7

6.2.1 Plant Operations Period (1957 to 1966)...................................................................... 11

6.2.2 Initial Cleanup Period (1967 to 1969).......................................................................... 13

6.2.3 Monitoring and Maintenance Period (1969 to 1985) ................................................... 13

6.2.4 Site Remediation Period (1985 to 2000)..................................................................... 13

6.3 Dose Reconstruction Parameters........................................................................................... 13

6.3.1 Interpreting the External Dosimetry Record ................................................................ 13

6.3.2 Weldon Spring Plant Historical Administrative Practices............................................. 14

6.3.2.1 Recorded Doses .......................................................................................... 14

6.3.2.2 Discrepancies .............................................................................................. 14

6.3.2.3 Missing Entry ............................................................................................... 14

6.3.2.4 Badge Assignment and Exchange Frequency.............................................. 15

6.3.2.5 Interpretation of Reported Data.................................................................... 15

6.3.3 Plant-Wide Dosimetry Results .................................................................................... 17

6.3.3.1 Calibration.................................................................................................... 17

6.3.4 Workplace Radiation Fields ........................................................................................ 18

6.3.4.1 Gamma Dose............................................................................................... 18

6.3.4.2 Neutron Dose............................................................................................... 23

6.3.4.3 Electron Dose .............................................................................................. 24

6.3.5 Reported-Dose-to-Organ-Dose Conversion Factor Units............................................ 25

6.3.6 Limit of Detection........................................................................................................ 26

6.3.7 Exchange Frequency.................................................................................................. 28

6.3.7.1 Number of Zero Readings ............................................................................ 28

6.3.7.2 Determination of Missed Dose ..................................................................... 28

6.3.8 Unmonitored Energy Range ....................................................................................... 29

6.3.9 Angular Dependence.................................................................................................. 29

6.3.10 Uncertainty ................................................................................................................. 29

6.3.10.1 Film.............................................................................................................. 29

6.3.10.2 Thermoluminescent Dosimetry System........................................................ 30

6.3.11 Geometric Correction Factor ...................................................................................... 30

6.4 Attributions and Annotations .................................................................................................. 30

References ........................................................................................................................................ 31

Glossary ............................................................................................................................................ 34

ATTACHMENT A, EXAMPLE REPORTS.......................................................................................... 39
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LIST OF TABLES

TABLE TITLE PAGE

6-1 Beta and gamma emissions of primary interest........................................................................ 7

6-2 Summary of historical recorded dose practices ...................................................................... 14

6-3 Assignment of film badges ..................................................................................................... 15

6-4 Badge and nonbadge areas ................................................................................................... 15

6-5 Interpretation of reported data................................................................................................ 16

6-6 Summary of job titles as reported by workers and coded for statistics.................................... 18

6-7 Annual average gamma and beta dose.................................................................................. 19

6-8 Median values for occupational external dose through the operational period 1957-

1966.......................................................................................................................................20

6-9 Default photon energy distribution for WSP materials ............................................................ 21

6-10 Energy distribution by building or area ................................................................................... 21

6-11 Photon dose units for use with organ dose conversion factors ............................................... 25

6-12 Electron dose units for use with organ dose conversion factors ............................................. 26

6-13 Photon MDLs for WSP dosimeters by year ............................................................................ 27

6-14 Potential missed photon dose ................................................................................................ 27

6-15 Electron MDLs for WSP dosimeters by year........................................................................... 27

6-16 Potential missed electron dose............................................................................................... 27

6-17 Tolerance limits at WSP......................................................................................................... 28

6-18 AEC standards.......................................................................................................................28

6-19 Dose limits based on exchange frequency ............................................................................. 28

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Uranium-235 decay series .......................................................................................................8

6-2 Uranium-238 decay series ..................................................................................................... 10

6-3 Thorium-232 decay series...................................................................................................... 11

6-4 Average annual gamma exposure for various job categories ................................................. 19

6-5 Average annual beta exposure for various job categories ...................................................... 20

6-6 Estimated beta dose rate at surface of uranium metal at various enrichment levels ............... 24

6-7 Shallow dose rate from natural uranium slab.......................................................................... 25

A-1 Example of Annual Personnel Internal-External Exposure Report.......................................... 39

A-2 Example of Personnel Internal-External Radiation Summary ................................................. 40

A-3 Example of Annual Personnel Internal-External Radiation Exposure Report.......................... 41

A-4 Example of Annual Personnel External Radiation Exposure Report....................................... 42

A-5 Example employee recorded external and internal exposure record ...................................... 43

A-6 Example of Quarterly External Radiation Exposure Report .................................................... 44

A-7 Example of Film Badge Data Summary by year..................................................................... 45
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 5 of 45

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

CFR Code of Federal Regulations

cm centimeter

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act

g gram

hr hour

keV kilovolt-electron, 1,000 electron volts

MCW Mallinckrodt Chemical Works

MDL minimum detectable level

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 Ridge Associated Universities

POC probability of causation

R roentgen

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

wk week

WSP Weldon Spring Plant

yr year

β beta

γ gamma

§ section or sections
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 6 of 45

6.1 INTRODUCTION

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

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

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

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

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

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

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 probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) restrict the “performance of duty” 

referred to in 42 U S. C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

The statute also 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 excludes Naval Nuclear Propulsion Facilities from being covered under the Act, 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 occupational radiation exposures are considered valid for inclusion in a 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 (NIOSH 2010):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 7 of 45

6.1.1 Purpose

The purpose of this document is to describe Weldon Spring Plant (WSP) external dosimetry systems 

and practices.

6.1.2 Scope

WSP operations played an important role in the U.S. development of nuclear weapons through its 

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 2007) provides additional guidance.

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, it 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. The guidance and requirements within this technical basis document 

pertain to covered employment periods for the Weldon Spring Site, which consists of the Weldon 

Spring Plant proper, or the Chemical Plant (1957 – July 1967; October 1985 – 2002), Weldon Spring 

Raffinate Pits (1957 – July 1967; December 1971 – 2002), and Weldon Spring Quarry (July 1960 –

2002). Site remediation at the Weldon Spring Plant proper conducted from January 1, 1968, through 

September 30, 1985, are not covered under EEOICPA.

6.2 DOSIMETRY OVERVIEW

With few exceptions, the WSP processed uranium during the operational period (1957–1966), but a 

small amount of thorium was processed near the end of the plant’s operations (1963–1966). The 

Technical Basis Document for the Weldon Spring Plant – Site Description (ORAUT 2005a, p. 11) 

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 likely the most important 

contributor to skin dose because of its frequent high-energy beta emission. In addition, 234mPa 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%)
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 8 of 45

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).

Figure 6-1. Uranium-235 decay series (HEW 1970).
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Radiation protection practices and exposures at WSP varied over time. While no single documented 

description of the practices and processes were found, good descriptions of the film badge program 

were located (Author unknown undated) and other 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.
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Figure 6-2. Uranium-238 decay series (HEW 1970).
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Figure 6-3. Thorium-232 decay series (HEW 1970).

6.2.1 Plant Operations Period (1957 to 1966)

A film badge notification memorandum by the Mallinckrodt Chemical Works (MCW) Health and Safety 

Department (MCW 1958) indicates that the WSP film badge program began on March 1, 1958. 

Before that time, dosimetry 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 
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 12 of 45

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 Figure A-7 in Attachment A.)

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. An undated report, Film Badge Report 1958 (Author unknown circa 1958) indicates a biweekly 

exchange for dingot forge and chemical operators, but later states the badges were changed monthly. 

Burr (1959b) also states that “monthly exchange of film badges for all plant personnel is scheduled for 

January 30, 1959.” Another undated report, Personnel External Radiation Monitoring Program (MCW 

date unknown), describes the MCW program. It states that “wage personnel film badges are 

exchanged monthly and salaried personnel film badges quarterly.” A Summary of Health Protection 

Practices (MCW 1965) states that “operations badges are exchanged and processed on a calendar 

month schedule, all others [are exchanged] on a three-month schedule.” If the exchange frequency 

cannot be explicitly identified, the dose reconstructor should make the favorable-to-claimant 

assumption to use the most frequent exchange frequency for the period.

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 was shaped so 

that a 1-mm-thick cadmium shield could be inserted to cover approximately the top two-thirds of the 

film. A similar 1-mm cadmium shield was permanently clamped on the back of the badge. The film 

was DuPont dosimeter Type 552 film packets, which contained two dental-size films in a single 

wrapper. One of the films was apparently a DuPont Type 502 film 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 film 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.

A 1965 document (MCW 1965) 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.

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

continuously at work. In addition, permanent badges were assigned to those workers 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. Badges (dosimeters) were located in 

various process/work areas to provide reference data about changes in average radiation levels. Use 

of film badges by visitors or subcontract personnel was predetermined by the person who authorized 

entry and was managed on a self-prescribed basis (MCW 1958).

MCW (1965) also stated that “operations badges are exchanged and processed on a calendar [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 
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 13 of 45

until 1959 when the external program adopted a quarterly reading.” The exception to this policy was 

the weekly exchange period for the turret lathe and dingot operators in the early period of operation 

(MCW 1956). Because of this confusion, it is suggested that if the exchange frequency cannot be 

determined from the claimant file, a favorable-to-claimant exchange frequency of biweekly be 

assumed for all operations workers, with the exception of the turret lathe and dingot operators, which 

would be weekly through 1958, and that an exchange frequency of monthly for production workers 

and quarterly for all other workers be assumed from 1959 to 1966.

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

non-security badge containing a DuPont 555 film.” This badge 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.

6.2.2 Initial Cleanup Period (1967 to 1969)

Information has not been found that describes the external dosimetry program during the initial 

cleanup and shutdown 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 

the case, it is likely that the same film badge system would have been used and any dose received 

during this period would be in the claimant file.

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 personnel.

6.2.3 Monitoring and Maintenance Period (1969 to 1985)

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

period.

6.2.4 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).

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1 Interpreting the External Dosimetry Record

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

1957–1966

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).

6.3.2 Weldon Spring Plant Historical Administrative Practices

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.

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).

6.3.2.1 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 doses in units of mR (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); however it appears that the conversion to rad was never 

accomplished. It is assumed that the 93-ergs/g rep was used throughout the WSP production years 

and it is favorable to claimants to assume that roentgens (R or r in the records), rep, and rem are all 

equivalent.

6.3.2.2 Discrepancies

If the employee’s record contains discrepancies, it is favorable to claimants 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 favorable to claimants to use that dose in the dose reconstruction. 

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

6.3.2.3 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 
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 15 of 45

possibility is that the badge was lost and no dose was assigned for that period. The favorable-toclaimant assumption is that the dosimeter was lost, and dose should be assigned for that period using 

the dosimetry data from the previous and following periods, providing the job remained the same 

(consider the approach of Watson et al. 1994).

6.3.2.4 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. These data are found in Burr 

(1959c) and Mason (1955). Table 6-3 summarizes assignment of film badges during the operational 

period. Table 6-4 lists badge areas and nonbadge areas.

Table 6-3. Assignment of film badges.

a

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 or 

otherwise) who do not normally work in badge areas but find it necessary to enter a badge 

area for a limited time.

a. Source: MCW (1958, p. 62).

Table 6-4. Badge and nonbadge areas.

Badge area Nonbadge 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. Wage personnel film 

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

badges were exchanged and processed monthly and quarterly, with both dependent on the period as 

given in Section 6.2.1 above. Posting of individual film badge data to the employee’s health history 

file varied dependent on the exchange period, but the frequency was no greater than quarterly.

6.3.2.5 Interpretation of Reported Data

Table 6-5 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 on 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 favorable to 
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 16 of 45

claimants to assume that all units are rem. Figure A-7 shows a film badge summary report that 

includes data from multiple years.

Table 6-5. Interpretation of reported data.

Report Reported quantity

Interpretation of 

zeroes

Interpretation of 

blanks (no data)

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 indicates a 

monitored 

exposure reported 

as zero and 

should be treated 

as < MDL.

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, β, and γ + β (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 

facility, WSP, and had notes for 

worker transfer to parent company 

with date noted. 

Zero indicates a 

monitored 

exposure reported 

as zero and 

should be treated 

as < MDL.

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, γ (mrem), β 

(mrem), and γ + β (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 indicates a 

monitored 

exposure reported 

as zero and 

should be treated 

as < MDL.

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 γ and « γ / β. » No 

indication of units. Gamma/ β 

represents total external exposure for 

the period. 

Zero indicates a 

monitored 

exposure reported 

as zero and 

should be treated 

as < MDL.

Blank indicates 

unmonitored during 

that period. 

Annual Personnel 

Internal-External 

Radiation Exposure 

Report

(Figure A-3) Computergenerated report 

analogous to typewritten 

report of same name. 

For a single year, γ (mrem), β 

(mrem), and total, noted as « γ / β » 

(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 

printout. The weekly average 

external is also reported. 

Internal radiation exposure is 

reported on same form.

Zero indicates a 

monitored 

exposure reported 

as zero and 

should be treated 

as < MDL.

Blank indicates 

unmonitored during 

that period. 

Annual Personal 

External Radiation 

Exposure Report Year 

Zero indicates a 

monitored 

exposure reported 

Blank indicates 

unmonitored during 

that period. 
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 17 of 45

Report Reported quantity

Interpretation of 

zeroes

Interpretation of 

blanks (no data)

19xx

(Figure A-4) Computergenerated report.

as zero and 

should be treated 

as < MDL.

Recorded External 

Exposure

(Figure A-5) Computergenerated report showing 

both external and internal 

exposures by quarter. 

Have handwritten 

notations for external 

exposures by month for 

1966. 

Gamma and β and γ by year. 

Lifetime values for each. No 

indications of units, but presumably 

are mrem. 

Zero indicates a 

monitored 

exposure reported 

as zero and 

should be treated 

as < MDL.

Blank indicates 

unmonitored during 

that period. 

Quarterly External 

Radiation Exposure 

Report – Month 19xx

(Figure A-6) Computergenerated list for a given 

month. Includes data for 

several workers on same 

sheet.

For month of the quarter, lists β and γ 

and γ alone. Units not specified, but 

presumed to be mrem. Month 

identified numerically.

Year-to-date β and γ and γ values 

also given for each employee. 

Zero indicates a 

monitored 

exposure reported 

as zero and 

should be treated 

as < MDL. 

Blank indicates 

unmonitored during 

that period. 

6.3.3 Plant-Wide Dosimetry Results

Available worker data were 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 6-6, there were more than 70 different job titles for workers. These are categorized in nine 

categories that roughly represent the reported job titles. Table 6-7 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 most years. Exposure to beta radiation was 

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

In the rare instance when a worker was unmonitored for external dose and should have been 

monitored based on his occupation and/or work location, the annual average gamma and beta dose 

values associated with the worker’s occupation stated in Table 6-7 can be used to assign a 

reasonable estimate of dose. If a worker’s occupation is unknown, the median dose values in Table 

6-8 can be used to assign an unmonitored dose which is considered to be favorable to the claimant.

6.3.3.1 Calibration

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

control films (Miller 1955). That same system was used 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 curves with 

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

equivalent). A direct conversion to mrad is assumed in recording personnel exposure.
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 18 of 45

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

Coded job 

title Reported worker job titles

Coded job 

title Reported worker job titles

Equipment 

operator

Fork Lift Driver Operator Acid Recovery/Loader

Fork Lift Operator Chemical D95 Operator

Warehouse Fork Lift Operator Chemical Operator, Storekeeper

Yard Operator Chemical Operator Pot Room

Foreman Foreman Chemical Operator

Production Foreman Chemical Operator/Maintenance

Manager Accountant Supervisor Conversion Green Salt

Product Control Supervisor Foreman-Operator

Supervisor, Plant and Maintenance 

Scheduler

Machine Operator

Non-radiation 

job

Computer Operator Machinist, Operator

Industrial Nurse Metal Plant, Manufacturer

Inventory Control Clerk Operator

Maintenance and Utility Control 

Clerk

Operator, Decontamination, 

Maintenance

Office Boy/Accounting Clerk Operator/Labor

Shipping Pot Room Worker

Safety, 

security

Production/Safety and Fire Marshal Press Operator Refinery 103

Safety Processing Plant

Safety and Fire Prevention Production

Security Guard Production Operator

Worker Electrician Production Operator A

General Cleaner Refinery Operator

Machinist Uranium Processor

Maintenance Electrician Utility Operator

Maintenance & Oiler Water Plant, Refinery

Maintenance Electrician Engineer Chemical and Project Engineer

Maintenance, Welder Chemical Engineer

Maintenance/Rigger Engineer and Production Control

Metal Worker Mechanical Engineer

Millwright Plant Engineer

Pipefitter Process Engineer

Tool and Die Maker Laboratory 

worker

Analytical Chemist

Utility Worker Chemical Technician

Welder Laboratory Technician

Welder, Maintenance Laboratory Technician, Engineer

Welder/Metal Fabricator Research Chemist

6.3.4 Workplace Radiation Fields

No data were found that describe the workplace radiation fields at WSP. Summary reports indicate 

that natural uranium was the material the workers came in contact with most frequently. Radiation 

fields most often consisted of a complex mixture of beta and gamma energies. 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 relatively close to the source, but these 

particles do not penetrate deeply into the body.

6.3.4.1 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 uranium, with small amounts of slightly enriched (< 1%), and 
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 19 of 45

Table 6-7. Annual average gamma and beta dose (mR).

Job description

Annual average gamma exposure by year

1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Engineer 116 102 60 164 106 143 148 330 195 94

Equipment operator 135 — 135 168 154 151 176 148 164 278

Foreman 361 80 134 181 171 176 132 143 171 113

Laboratory worker 280 78 70 74 130 176 203 255 531 155

Manager 52 — 37 517 102 41 142 112 78 64

Nonradiation job 114 — 48 90 94 72 93 226 198 33

Operator 298 122 159 250 402 383 648 533 417 295

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

Worker 240 132 70 150 189 257 400 370 279 195

Job description

Annual average beta exposure by year

1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Engineer 250 23 203 605 513 561 239 288 181 60

Equipment operator 190 344 1,167 835 800 668 324 430 338 171

Foreman 368 118 476 1,342 794 931 1,199 633 561 242

Laboratory worker 549 143 328 449 885 904 342 318 289 278

Manager 107 35 199 1,160 1,099 403 204 128 91 181

Nonradiation job 145 135 108 351 541 182 224 243 134 34

Operator 730 247 1,160 2,828 2,908 1,722 1,541 1,157 874 287

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

Worker 697 93 514 1,216 1,440 1,326 1,113 549 724 354

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

0

100

200

300

400

500

600

700

1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Gamma Exposure, mR

Year

Gamma exposure by year for nine job types

Engineer Equipment operator Foreman Laboratory worker Manager

Nonradiation job Operator Safety, security Worker
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 20 of 45

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

Table 6-8. Median values for occupational external dose through the 

operational period 1957-1966.

a

Year 50th-percentile gamma

(mrem)b

50th-percentile beta

(mrem)

1957 161 305

1958 81 116

1959 78 317

1960 126 679

1961 136 859

1962 139 735

1963 220 314

1964 187 293

1965 151 226

1966 123 114

a. These annual dose values represent the median dose across all worker categories stated 

above in Table 6-7.

b. Calculated as the 50th-percentile value of a lognormal distribution in accordance with BattelleTIB-5000 (BMI 2007).

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 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 2009, p. 155). This decay product grows in fairly rapidly 

0

500

1000

1500

2000

2500

3000

3500

1957 1958 1959 1960 1961 1962 1963 1964 1965 1966

Beta Exposure, mR

Year

Beta exposure by year for nine job types

Engineer Equipment operator Foreman Laboratory worker Manager

Nonradiation job Operator Safety, security Worker
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 21 of 45

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 

favorable to claimants 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% 0%

30-250 keV 50% 50% 100% 25%

>250 keV 50% 50% 0% 75%

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

Building Description Radiation Energy Percentage

101 Sampling Plant Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

Area 102 

A&B

Refinery Tank Farm Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

103 Digestion and Denitration Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

Natural Th

Electron >15 keV 100

Photon 30–250 keV 25

>250 keV 75

Slightly enriched U

Electron >15 keV 100

Photon 30–250 keV 100

104 Lime Storage None

105 Extraction Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

Natural Th

Electron >15 keV 100

Photon 30–250 keV 25

>250 keV 75

Slightly enriched U

Electron >15 keV 100

Photon 30–250 keV 100

106 Refinery sewer sampling Natural U

Electron >15 keV 100

Photon 30–250 keV 50
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 22 of 45

Building Description Radiation Energy Percentage

>250 keV 50

108 Nitric acid plant Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

109, 110 West Drum Storage, East 

Drum Storage

Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

201 Green Salt Building Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

Neutron 0.1–2 MeV 100

Natural Th

Electron >15 keV 100

Photon 30–250 keV 25

>250 keV 75

Slightly enriched U

Electron >15 keV 100

Photon 30 – 250 keV 100

301 Metals Building Natural U

Electron >15 keV 100

Photon 30 – 250 keV 50

>250 keV 50

Neutron 0.1 – 2 MeV 100

Natural Th

Electron >15 keV 100

Photon 30 – 250 keV 25

>250 keV 75

Slightly enriched U

Electron >15 keV 100

Photon 30 – 250 keV 100

302 Magnesium Building None

Pad 303 Material Storage Pad None

401 Steam Plant None

403 Chemical Pilot Plant Natural U

Electron >15 keV 100

Photon 30 – 250 keV 50

>250 keV 50

Depleted U

Electron >15 keV 100

Photon 30 – 250 keV 40

>250 keV 60

404 Metallurgical Pilot Plant Natural U

Electron >15 keV 100

Photon 30 – 250 keV 50

>250 keV 50

Depleted U

Electron >15 keV 100

Photon 30 – 250 keV 40

>250 keV 60

405A & B Pilot Plant Maintenance Natural U

Electron >15 keV 100

Photon 30 – 250 keV 50
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 23 of 45

Building Description Radiation Energy Percentage

>250 keV 50

406 Warehouse Natural Th

Electron >15 keV 100

Photon 30 – 250 keV 25

>250 keV 75

407 Laboratory Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

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 >15 keV 100

Photon 30–250 keV 50

>250 keV 50

415 Process Incinerator Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

417 Paint Shop None

426 Water Tower None

427 Primary Sewage 

Treatment Plant

None

428 Fuel Gas Plant None

429 Water Reserve Facilities None

430 Ambulance Garage None

43l Laboratory Sewer 

Sampler

Natural U

Electron >15 keV 100

Photon 30–250 keV 50

>250 keV 50

432 Main Sewer Sampler None

437 Records Retention 

Building

None

439, 443 Fire Training and Storage 

Building

None

441 Cylinder Storage None

202 A&B Green Salt Tank Farm None

6.3.4.2 Neutron Dose

Because the WSP processed very small amounts of slightly enriched uranium (<1% 235U and 0.68% 

of total throughput and no UF6), the exposure to neutrons was miniscule (ORAUT 2010a, p. 33). The 

fact that the uranium was in the form of UF4 and other nonproducing neutron compounds resulted in 

the total absence of recordable neutron doses even though neutron dosimeters were worn by those 

employees working with the enriched uranium (Author unknown undated). The slightly enriched 

uranium was processed in Buildings 103,105, 201, and 301, so employees assigned to these facilities 

during the processing of this material received neutron dosimeters. Studies as reported in ORAUT 

2010a (p. 33 and pp. 59-61) provide adequate evidence that there is no technical reason to expect 

any measurable neutron doses and, therefore, no reported results.
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 24 of 45

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

neutrons from the alpha-neutron reaction from UF4 could have contributed dose to WSP workers. The 

analysis performed for the similar situation at Fernald (ORAUT 2004) 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. 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 a 95th percentile ratio of 0.23, should be applied in those areas where there is the potential for 

neutron dose from uranium fluoride compounds (ORAUT 2004, p. 20; NIOSH 2011a, p. 97). 

6.3.4.3 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 [0.14% of 

total mass (ORAUT 2010a, p. 23)] was also present at the site on an intermittent basis. Slightly 

enriched uranium [less than 1% 235U by weight and 0.68% of total mass (ORAUT 2010a, p. 23)] was 

also present at times in the form of scrap metal or residues.

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 2009, p. 152).

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

uncommon in DOE uranium facilities. Surface beta dose rates on the order of 1 to 20 rad/hr 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. Many operations required contact 

between the hands and the radioactive materials, and the glove program was “sketchy and 

inadequate” (Mason 1955).
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 25 of 45

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 

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 2009, p. 154).

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 with 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.

6.3.5 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.

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 Hp(10) 1997 Hp(10)

1958 R 1968 R 1978 1988 Hp(10) 1998 Hp(10)

1959 R 1969 R 1979 1989 Hp(10) 1999 Hp(10)

1960 R 1970 1980 1990 Hp(10) 2000 Hp(10)

1961 R 1971 1981 1991 Hp(10) 2001 Hp(10)

1962 R 1972 1982 1992 Hp(10) 2002 Hp(10)

1963 R 1973 1983 1993 Hp(10)
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 26 of 45

1964 R 1974 1984 1994 Hp(10)

1965 R 1975 1985 Hp(10) 1995 Hp(10)

1966 R 1976 1986 Hp(10) 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 H'(0.07) 1997 H'(0.07)

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

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

1960 rad 1970 1980 1990 H'(0.07) 2000 H'(0.07)

1961 rad 1971 1981 1991 H'(0.07) 2001 H'(0.07)

1962 rad 1972 1982 1992 H'(0.07) 2002 H'(0.07)

1963 rad 1973 1983 1993 H'(0.07)

1964 rad 1974 1984 1994 H'(0.07)

1965 rad 1975 1985 H'(0.07) 1995 H'(0.07)

1966 rad 1976 1986 H'(0.07) 1996 H'(0.07)

6.3.6 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 or very similar as that used at 

Hanford (ORAUT 2010b, pp. 58) and the Savannah River Site (ORAUT 2005b, p. 91), which both 

state a minimum detectable level (MDL) of 0.040 mrem. A Hanford study of two-element dosimeters 

identified a detection level of about 40 mR at the upper 95% confidence level for radium gamma 

radiation (ORAUT 2010b, p. 41) and is repeated here:

Therefore, in the Hanford studies, dose results with these different dosimeter systems 

could be evaluated as described in Wilson et al. (1990). Wilson (1960b) described 

three changes in 1960 that led to a lower detection level of about 15 mrem at the 90% 

confidence level involving: (1) elimination of nonisotropic effect of calibration source, 

(2) automated film processing and (3) change to the more sensitive DuPont 508 film. 

He noted in this report a detection level of 40 mrem at the 95% confidence level for the 

Hanford two-chip dosimeter system with the DuPont 502 film used prior to these 

changes. An important consideration in this analysis concerned the level of potential 

missed dose. Wilson describes the analysis of 49 batches of Hanford routine 

calibration results that indicated a 25% standard deviation at the 30-mrem calibration 

level based on the optical density readings. Based on an analysis of the capabilities of 

the densitometer used to process the film, he estimated a likelihood of 0.33 (1/3) that a 

dose of 15 mrem would not be detected. The likelihood that this would occur for each 

successive monthly exchange for an entire year would be (0.33)12 or about 1 in a 

million. Based on the 13 exchanges during the year at that time, he estimated a 

maximum potential missed dose of 195 mrem (i.e., 15 × 13). Conversely, Wilson 

estimated that about 8% of the time a positive dose would be recorded for dosimeters 

that received no exposure. A similar analysis could be performed for the dosimeter 

used prior to 1960 with an estimate that about 30 mrem would be detected one-third of 

the time.

The MCW St. Louis Plant Site Profile (ORAUT 2009, p. 108) cites records in which gamma dose 

results are shown as "50*" where the asterisk refers to a footnote that 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 more favorable-to-claimant MDL for the WSP film dosimeter of 

50 mR gamma and 80 mrep beta. Landauer (the dosimeter manufacturer) typically quotes a 
Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 27 of 45

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 MDLs for WSP dosimeters by year.

Year MDL Year MDL Year MDL Year MDL Year MDL

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

Table 6-14. Potential missed photon dose.

Period 

of use Dosimeter MDL

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 (MDL × exchange frequency) ÷ 2, from 

OCAS-IG-001 (NIOSH 2007).

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

Year MDL Year MDL Year MDL Year MDL Year MDL

1957 80 mrepa 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

a. 1 rep (Roentgen-equivalent-physical) is equal to about 83 ergs of energy absorbed per gram tissue. It 

is favorable to the claimant to assume that rep = rad = rem.

Table 6-16. Potential missed electron dose.

Period 

of use Dosimeter MDL

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
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a. Maximum annual missed dose calculated using 

(minimum detection limit × exchange frequency) ÷ 2, from OCAS-IG-001 (NIOSH 2007).

6.3.7 Exchange Frequency

Based on the historical evidence discussed in Section 6.1.3.1, it is favorable to claimants 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. The one 

exception is that turret lathe and dingot operators had weekly exchanges through 1958.

6.3.7.1 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 favorable to the claimant. 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. Dose reconstructors should use dosimetry 

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

NIOSH (2007), it is possible to develop a favorable-to-claimant estimate of the number of zeros and 

ultimately the missed dose.

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 1,500 mrep/wk

Table 6-18. AEC standards.

Type of exposure Period Dose (rem)

Whole body, head and trunk, active bloodforming organs, gonads, or lens of eyeAccumulated dose 5 × (N – 18)

Calendar quarter or 

13 consecutive wks

3

Skin of whole body and thyroid Year 30

Calendar quarter or 

13 consecutive wks

10

Hands and forearm, feet and ankles Year 75

Calendar quarter or 

13 consecutive wks 

25

Table 6-19. Dose limits 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

6.3.7.2 Determination of Missed Dose

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

discussed in Section 2.1.2.2 of NIOSH (2007). If the number of zero readings is indeterminate, it can 

be estimated under the assumption that prorated dose limits were not exceeded.
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6.3.8 Unmonitored Energy Range

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

River Site TBD (ORAUT 2005b, pp. 92-95) 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.

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 

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

appropriate for this dosimetry system.

6.3.9 Angular Dependence

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 (ORAUT 2010b). This factor is included in the overall bias provided elsewhere in this TBD.

6.3.10 Uncertainty

6.3.10.1 Film

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. The Miller study provided 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 (ORAUT 2010b, p. 58). The factors 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)
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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 (ORAUT 2010b, p. 58) is 

appropriate to use for the WSP dosimeter as well.

6.3.10.2 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%.

6.3.11 Geometric Correction Factor

Consideration should be given to geometry when performing dose reconstruction for uranium facility 

workers who worked with uranium metals, powders, or residues or for workers who worked on 

equipment contaminated with uranium. An underestimation of the measured and missed photon 

doses could occur if the energy employee wore their dosimeter on the upper chest or lapel and not in 

the central area of the chest or on the waist. The organs located in the lower torso region are most 

affected. These include, but are not limited to, the stomach, liver, kidney, ureter, gall bladder, 

pancreas, small intestine, large intestine, rectum, ovaries, uterus, urinary bladder, and prostate. 

The geometric correction factor calculated for application to the general population of workers is 2.1 

and applied as a constant in IREP as determined in NIOSH (2010b). This geometric correction factor 

should be applied to the measured and missed photon doses for operators, material handlers, and 

trade workers including chemical operators, production operators, uranium material handlers, 

pipefitters, carpenters, welders, sheet metal workers, electricians, foremen, etc. Guidance in DCASTIB-0013, Selected Geometric Exposure Scenario Considerations for External Dose Reconstruction 

at Uranium Facilities (NIOSH 2010b), should be used to adjust the measured and missed photon 

dose to Weldon Spring workers. For glovebox workers, the geometric correction factor has a 

lognormal distribution with a geometric mean of 2.19 and a geometric standard deviation of 1.34 as 

stated in NIOSH (2011b).

6.4 ATTRIBUTIONS AND ANNOTATIONS

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

section of this document.
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REFERENCES

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

Washington, DC, August 12. [SRDB Ref ID: 1499]

Author unknown, circa 1958, Film Badge Report 1958. [SRDB Ref ID: 15871]

Author unknown, undated, “Personnel External Radiation Monitoring Program”, Mallinckrodt Chemical 

Works, Uranium Division, St. Louis, Missouri. [SRDB Ref ID: 15864, pp. 61-64]

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, 

Atomic Energy Commission, St. Louis Area Office, Missouri, February 11. [SRDB Ref ID: 

11806, pp. 5-10]

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

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

Commission, St. Louis Area Office, Missouri, November 2. [SRBD Ref ID: 11806, pp. 2-4]

BMI (Battelle Memorial Institute), 2007, Technical Information Bulletin: Default Assumptions and 

Methods for Atomic Weapons Employer Dose Reconstructions, Rev 00, Battelle-TIB5000/PNWD-3741, April 2. [SRDB Ref ID: 32016]

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

Weldon Spring,” memorandum to M. G. Mason, Mallinckrodt Chemical Works, Uranium 

Division, St. Louis, Missouri, October 26. [SRDB Ref ID: 15578, pp. 3-4] 

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, May. [SRDB Ref ID: 15578, pp. 5-16]

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

Chemical Works, Uranium Division, St. Louis, Missouri, June 16. [SRDB Ref ID: 15565] 

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

Mallinckrodt Chemical Works, Uranium Division, St. Louis, Missouri, January 26. [SRDB Ref 

ID: 15864, p. 41]

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

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

[SRDB Ref ID: 15864, p. 42]

Burr, R.H.,1959c, Weldon Spring Film Badge Program, Mallinckrodt Chemical Works, Uranium 

Division, St. Louis, Missouri. [SRDB Ref ID: 15864, pp. 27-31]

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. [SRDB Ref ID: 12294]

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

Site Remedial Action Project, DOE/OR/21548-441, Rev. 0, Oak Ridge Operations Office, Oak 

Ridge, Tennessee, April. [SRDB Ref ID: 79446]
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DOE (U.S. Department of Energy), 2000, External Dosimetry Technical Basis Manual: Weldon Spring 

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

Ridge Tennessee, January. [SRDB Ref ID: 32723]

DOE (U.S. Department of Energy), 2009, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2009, Washington, D.C. [SRDB Ref ID: 

97381]

Dupree-Ellis, E., J. Watkins, J. Ingle, and J. Phillips, 2000, “External Radiation Exposure and Mortality 

Among a Cohort of Uranium Processing Workers,” American Journal of Epidemiology, volume 

152, number 1, pp. 91-95. [SRDB Ref ID: 9486]

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

(Revised Edition), Public Health Service, Food and Drug Administration, Bureau of 

Radiological Health, Rockville Maryland, January. [SRDB Ref ID: 75017, pp. 110, 112-113]. 

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

Dupree (Oak Ridge Associated Universities), Delta-21 Resources, Inc., February 4. [SRDB 

Ref ID: 13176] 

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

Mallinckrodt Chemical Works, St. Louis, Missouri, February 21. [SRDB Ref ID: 13180]

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

Division, St. Louis, Missouri, November 1. [SRDB Ref ID: 15864, pp. 2-10]

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

memorandum to W. L. Utnage, Uranium Division, Health and Safety Department, Weldon 

Spring, Missouri. [SRDB Ref ID: 15864, p. 32]

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

Uranium Division, St. Louis, Missouri. [SRDB Ref ID: 14998]

MCW (Mallinckrodt Chemical Works), date unknown, Personnel External Radiation Monitoring 

Program, Uranium Division, St. Louis, Missouri. [SRDB Ref ID: 15864, pp. 61-64]

Miller, J. W., 1955, “Investigation of Film Badge Standards Calibration Data,” memorandum to M. G. 

Mason, Mallinckrodt Chemical Works, St. Louis, Missouri, February 3. [SRDB Ref ID: 6513, 

pp. 58-64] 

NIOSH (National Institute for Occupational Safety and Health), 2007, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio. [SRDB Ref ID: 38864]

NIOSH (National Institute for Occupational Safety and Health), 2010a, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio. [SRDB Ref ID: 88929]

NIOSH (National Institute for Occupational Safety and Health), 2010b, Technical Information Bulletin: 

Selected Geometric Exposure Scenario Considerations for External Dose Reconstruction at 

Uranium Facilities, Rev 1, Division of Compensation Analysis and Support, DCAS-TIB-0013, 

November 23. [SRDB Ref ID: 90694]
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NIOSH (National Institute for Occupational Safety and Health), 2011a, Advisory Board on Radiation 

and Worker Health Work Group on Weldon Spring meeting. September 13. [SRDB Ref ID: 

120526 pp. 136-139]

NIOSH (National Institute for Occupational Safety and Health), 2011b, Technical Information Bulletin: 

Best Estimate External Dose Reconstruction for Glovebox Workers, Rev 4, Division of 

Compensation Analysis and Support, DCAS-TIB-0010, November 28. [SRDB Ref ID: 

104815]

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

National Academy Press, Washington D.C. [SRDB Ref ID: 1905]

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

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

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

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

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

June 24. 

ORAUT (Oak Ridge Associated Universities Team), 2005b, Savannah River Site, ORAUT-TKBS0003, Rev. 03, Oak Ridge, Tennessee, April 5. 

ORAUT (Oak Ridge Associated Universities Team), 2010, Basis for Development of an Exposure 

Matrix for the Mallinckrodt Chemical Company St. Louis Downtown Site and the St. Louis 

Airport Site, St. Louis, Missouri, ORAUT-TKBS-0005, Rev. 03, Oak Ridge, Tennessee, 

November 22. 

ORAUT (Oak Ridge Associated Universities Team), 2010a, SEC Petition Evaluation Report for the 

Weldon Spring Plant, Petition SEC-00143, Rev 00, Oak Ridge, Tennessee, April 16.

ORAUT (Oak Ridge Associated Universities Team), 2010b, Hanford Site – Occupational External 

Dosimetry, ORAUT-TKBS-0006-6, Rev. 04, Oak Ridge, Tennessee, January 7. 

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. [SRDB Ref 

ID: 22737]

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. [SRDB Ref ID: 8601]
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GLOSSARY

beta dose

Designation (i.e., beta) on some records for external dose from beta and less-energetic X-ray 

and gamma radiation, often for shallow dose or dose to the lens of the eye. 

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

bremsstrahlung 

Electromagnetic radiation released as a result of inelastic scattering of a moving charged 

particle within the nucleus of an atom. X-rays produced in a typical medical X-ray tube 

frequently originate from inelastic scattering of accelerated electrons in the anode material. 

deep absorbed dose

Absorbed dose in units of rem or sievert at a depth of 1 centimeter (1,000 milligrams per 

square centimeter). See dose.

deep dose equivalent [Hp(10)] 

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose.

depleted uranium

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

DOE Laboratory Accreditation Program (DOELAP) 

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs. 

dose 

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

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

units of roentgens, rads, reps, or grays.

dose equivalent (H)

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dosimeter

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

received by an individual. See film dosimeter and thermoluminescent dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.
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dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results. 

electron 

Basic atomic particle with negative charge and a mass 1/1,837 that of a proton. Electrons 

surround the positively charged nucleus of the atom.

exchange period 

Period (weekly, biweekly, monthly, etc.) for routine exchange of dosimeters. Also called 

exchange frequency.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

extremity

The portion of the arm from and including the elbow through the fingertips and the portion of 

the leg from and including the knee and patella through the toes. 

field calibration

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

environment.

film

In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter.

film density

See optical density.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fission

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

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

two or three neutrons.

fissionable

Capable of undergoing fission by capturing neutrons, including fast neutrons. Uranium-238 is 

fissionable. Fissionable indicates both spontaneous and induced fission. 

gamma radiation 

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

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

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus. 
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ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray radiation.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. See element.

neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.

nuclide 

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

open window

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). See film dosimeter.

operating area

Usually refers to a major operational work area at a site.

optical density

Measure of the degree of opacity of photographic or radiographic film defined as OD = log10

(I0/I), the base-10 logarithm of the ratio of the reference light intensity I0 (without film) to the 

transmitted light intensity (through the film). Also called film density and density reading. 

personal dose equivalent Hp(d)

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz. 
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photon radiation 

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays. 

quality factor, Q

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent. See dose.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.

radioactive 

Of, caused by, or exhibiting radioactivity.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., carbon-14) of 

spontaneously emitting energetic particles (electrons or alpha particles) by the disintegration of 

their atomic nuclei. See radionuclide. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

rep 

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 83 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical.

roentgen

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts). 

shallow absorbed dose (Ds)

Absorbed dose at a depth of 0.07 millimeters (7 milligrams per square centimeter) in a material 

of specified geometry and composition.

shallow dose equivalent (Hs)

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.
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shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

sievert (Sv) 

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

skin dose

See shallow dose equivalent.

thermoluminescent dosimeter (TLD) 

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose. 

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose.

X-ray radiation 

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom. 
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LIST OF FIGURES 

FIGURE PAGE

A-1 Example of Annual Personnel Internal-External Exposure Report.......................................... 39

A-2 Example of Personnel Internal-External Radiation Summary ................................................. 40

A-3 Example of Annual Personnel Internal-External Radiation Exposure Report.......................... 41

A-4 Example of Annual Personnel External Radiation Exposure Report....................................... 42

A-5 Example employee recorded external and internal exposure record ...................................... 43

A-6 Example of Quarterly External Radiation Exposure Report .................................................... 44

A-7 Example of Film Badge Data Summary by year..................................................................... 45

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

ATTACHMENT A

EXAMPLE REPORTS

Page 1 of 7
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Figure A-2. Example of Personnel Internal-External Radiation Summary. 

ATTACHMENT A

EXAMPLE REPORTS

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Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 41 of 45

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

ATTACHMENT A

EXAMPLE REPORTS

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Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 42 of 45

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

ATTACHMENT A

EXAMPLE REPORTS

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Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 43 of 45

Figure A-5. Example employee recorded external and internal 

exposure record.

ATTACHMENT A

EXAMPLE REPORTS

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

ATTACHMENT A

EXAMPLE REPORTS

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Document No. ORAUT-TKBS-0028-6 Revision No. 01 Effective Date: 02/06/2013 Page 45 of 45

Figure A-7. Example of Film Badge Data Summary by year.

ATTACHMENT A

EXAMPLE REPORTS

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