ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 60

DOE Review Release 05/30/2013

Document Title:

Weldon Spring Plant – Occupational Internal 

Dose

Document Number: ORAUT-TKBS-0028-5 

Revision: 02

Effective Date: 05/21/2013

Type of Document: TBD

Supersedes: Revision 01

Subject Expert(s): David P. Harrison and Eugene W. Potter

Approval: Signature on File Approval Date: 05/17/2013

David P. Harrison, Document Owner

Concurrence: Signature on File Concurrence Date: 05/15/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 05/15/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File Concurrence Date: 05/15/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 05/21/2013

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-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 2 of 60

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

06/28/2005 00 First approved issue of new technical basis document for the Weldon 

Spring Plant – Occupational Internal Dose. Incorporated formal 

internal and NIOSH review comments. Training is not required. 

Initiated by Robert Meyer.

03/15/2013 01 Incorporates numerous minor corrections and edits. SRDB numbers 

and page numbers were added. In Section 5.2.1, the specific activity 

for slightly enriched (1%) uranium was adjusted to agree with the 

formula in DOE 2001a. Sections 5.2.2 related to uranium decay 

products were edited to reflect the fact that the early uranium mills 

may not have been effective in removing thorium and to increase the 

amount of Th-230 and daughters to assume in calculations. The 

maximum concentrations for certain decay products at the WSRP 

and WSQ were eliminated. A discussion was added in Section 5.2.3 

on potential intakes from thoron. Site-specific ratios of Th-230 to 

other contaminants were developed in Section 5.6.1.1 for use during 

initial uranium processing. DWA concentrations for thorium dust 

measurements were added as Attachment A, and these values were 

used to create a new table of thorium intakes in Section 5.6.1.2. 

Thoron guidance was also added to this section. A statement was 

added in Section 5.6.1.1 to use Friday urine sampling data statistics 

to avoid underestimating intakes. The estimated annual exposure 

from radon was increased in Section 5.6.1.3, and Section 5.2.4 on 

recycled uranium was updated in response to issues raised in the 

Advisory Board Work Group. Thorium-232 intake rates updated to 

reflect an 8-hour workday normalized to a calendar for each of the 

years for Th-232 operations, and equations were included in the text 

regarding calculation of the median and 95th percentiles of the Th232 intake rates. 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.

05/21/2013 02 Revision initiated per DCAS direction to modify the indoor radon 

equilibrium factor from 0.5 to 0.7 to coincide with Advisory Board 

meeting transcripts from September 19, 2012. This increases the 

radon intake value from 8.8 WLM/year to 12.4 WLM/year. Changes 

were made in Section 5.6.1.3 and Table 5-23. No changes occurred 

as a result of formal internal and NIOSH review. Training required: 

As determined by the Objective Manager. Initiated by David P. 

Harrison.

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Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 3 of 60

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

5.1 Introduction ................................................................................................................................. 7

5.1.1 Purpose...........................................................................................................................8

5.1.2 Scope..............................................................................................................................8

5.1.3 Background on Mallinckrodt Chemical Works and the Weldon Spring Plant..................8

5.1.4 Radiation Protection Practices ......................................................................................12

5.2 Source Term .............................................................................................................................13

5.2.1 Isotopic Composition of Uranium ..................................................................................13

5.2.2 Uranium Decay Products ..............................................................................................13

5.2.3 Natural Thorium and Decay Products ...........................................................................14

5.2.4 Recycled Uranium.........................................................................................................16

5.2.5 Solubility Classification and Absorption Type ...............................................................16

5.2.6 Particle Size ..................................................................................................................18

5.3 In Vitro Measurements..............................................................................................................18

5.3.1 Operational Period (1957 to 1966)................................................................................18

5.3.1.1 Uranium.............................................................................................................18

5.3.1.1.1 Routine Sampling Program................................................................. 18

5.3.1.1.2 Special Sampling Program ................................................................. 18

5.3.1.1.3 Data Reporting Levels and Minimum Detectable Amounts ................19

5.3.1.1.4 Interferences and Uncertainties..........................................................19

5.3.1.1.5 Reporting Formats and Codes............................................................19

5.3.1.1.6 Work Group Data................................................................................21

5.3.1.2 Thorium .............................................................................................................23

5.3.2 Environmental Monitoring Period (1975 to 1984)..........................................................23

5.3.3 Remediation Period (1985 to 2002) ..............................................................................23

5.4 In Vivo Measurements ..............................................................................................................32

5.4.1 Operational Period (1957 to 1966)................................................................................32

5.4.2 Environmental Monitoring Period (1975 to 1984)..........................................................33

5.4.3 Remediation Period (1985 to 2002) ..............................................................................33

5.5 Air Concentration and Dust Measurements ..............................................................................33

5.5.1 Operational Period (1957 to 1966)................................................................................33

5.5.1.1 Maximum Acceptable Concentrations...............................................................34

5.5.1.2 Special Curie for Uranium .................................................................................34

5.5.1.3 Dust Exposure Calculation ................................................................................35

5.5.1.4 Dust Studies ......................................................................................................35

5.5.2 Transfer to the U.S. Army (1967 to 1974).....................................................................37

5.5.3 Environmental Monitoring Period (1975 to 1984)..........................................................37

5.5.4 Remediation Period (1985 to 2002) ..............................................................................37

5.6 Assessment of Intakes..............................................................................................................38

5.6.1 Operational Period (1957 to 1966)................................................................................38

5.6.1.1 Uranium and Related Intakes............................................................................38

5.6.1.2 Thorium-232 Intakes .........................................................................................40

5.6.1.3 Intakes of Other Radionuclides .........................................................................42
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5.6.2 Environmental Monitoring Period (1975 to 1984)..........................................................43

5.6.3 Remediation Period (1985 to 2002) ..............................................................................44

5.7 Summary of Instructions to Dose Reconstructors.....................................................................44

5.8 Attributions and Annotations .....................................................................................................44

References ...........................................................................................................................................46

Glossary................................................................................................................................................50

ATTACHMENT A, DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS .....................54

LIST OF TABLES

TABLE TITLE

5-1 History of dust-generating activities ..........................................................................................

PAGE

10

5-2 Potential internal radionuclide exposure for production buildings.............................................10

5-3 Products or intermediates in production buildings ....................................................................11

5-4 Mass and percent of feed materials processed ........................................................................13

5-5 Thoron and isotopic daughters ................................................................................................. 15

5-6 Solubility class summary...........................................................................................................17

5-7 Cost center codes for workers ..................................................................................................20

5-8 Composite uranium urine data summary ..................................................................................22

5-9 Uranium urine data summary by cost center for 1958 ..............................................................23

5-10 Uranium urine data summary by cost center for 1959 ..............................................................24

5-11 Uranium urine data summary by cost center for 1960 ..............................................................25

5-12 Uranium urine data summary by cost center for 1961 ..............................................................26

5-13 Uranium urine data summary by cost center for 1962 ..............................................................27

5-14 Uranium urine data summary by cost center for 1963 ..............................................................28

5-15 Uranium urine data summary by cost center for 1964 ..............................................................29

5-16 Uranium urine data summary by cost center for 1965 ..............................................................30

5-17 Uranium urine data summary by cost center for 1966 ..............................................................31

5-18 Work group time-weighted average exposure for uranium operations .....................................36

5-19 Time-weighted average 232Th dust concentrations ...................................................................36

5-20 Activity ratios of uranium decay products and other impurities to 230Th activity in 

Weldon Spring raffinate pits......................................................................................................39

5-21 Intakes of uranium decay products and other impurities based on raffinate pit 

measurements ..........................................................................................................................39

5-22 Thorium-232 inhalation and ingestion intake rates in Bq per year............................................41

5-23 Instructions to DRs....................................................................................................................44

A-1 Dust concentrations for thorium production jobs.......................................................................54

LIST OF FIGURES

FIGURE TITLE

5-1 Decay and buildup of 228Th and daughters following processing..............................................

PAGE

15
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

Bq becquerel

BZA breathing-zone air

CEDE committed effective dose equivalent

CFR Code of Federal Regulations

cm centimeter

d day

d/s disintegrations per second

DAC derived air concentration

dpm disintegrations per minute

DR dose reconstructor

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DWA daily weighted average

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

g gram

GSD geometric standard deviation

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

kg kilogram

L liter

lb pound

m meter

MAC maximum allowable concentration

MCW Mallinckrodt Chemical Works

MDA minimum detectable amount

MDC minimum detectable concentrations

MeV megaelectron-volt, 1 million electron-volts

min minute

mL milliliter

MPC maximum permissible concentration

mrem millirem

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NLO National Lead of Ohio

ORAU Oak Ridge Associated Universities
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PAEC potential alpha energy concentration

pCi picocurie

POC probability of causation

ppb parts per billion

PUREX plutonium-uranium extraction

SRDB Ref ID Site Research Database Reference Identification (number)

U.S.C. United States Code

WL working level

WLM working level month

WSCP Weldon Spring Chemical Plant

WSP Weldon Spring Plant

WSQ Weldon Spring Quarry

WSRP Weldon Spring Raffinate Pits

WSSRAP Weldon Spring Site Remedial Action Project

yr year

α alpha particle 

μCi microcurie 

μg microgram

μL microliter

μm micrometer

μs microsecond

§ section or sections
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5.1 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

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 2010b): 

] 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 2010b). 

• 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.
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5.1.1

The purpose of this technical basis document (TBD) is to provide a profile of internal dosimetry 

practices at the Weldon Spring Plant (WSP). This document contains technical information on the 

history, methods, and interpretation of monitoring data for the evaluation of occupational internal dose 

to WSP workers.

Purpose

5.1.2

This TBD covers the methods used to assess internal radiation dose to workers at WSP. The WSP 

includes the Weldon Spring Chemical Plant (WSCP), the Weldon Spring Raffinate Pits (WSRP), and 

the Weldon Spring Quarry (WSQ). Internal radiation dose is the dose to a worker from deposition of 

radionuclides in the body. Such deposition can occur as a result of inhalation of radionuclides in 

airborne dust, incidental ingestion of radionuclides, and intake through intact skin or wounds. 

Because of the nature of the nuclides encountered at WSP, intake through the skin was unlikely. 

Scope

The radionuclides of concern for WSP are the naturally occurring isotopes of uranium (234U, 235U, and 

238U) and their decay products (primarily 230Th and 226Ra). Because the WSP also processed some 

natural thorium, this TBD includes material on 232Th and its decay products (228Ra and 228Th). 

However, due to the small amount of thorium processed at WSP, the primary radionuclides of concern 

for internal radiation dose are the isotopes of uranium.

Section 5.2 provides information on the source term for dosimetrically significant radionuclides at 

WSP. Sections 5.3 and 5.4 describe in vitro and in vivo measurements, respectively. Section 5.5 

discusses air monitoring and dust studies. Section 5.6 details assessment of radionuclide intakes. 

Section 5.7 provides a summary of instructions to dose reconstructors (DRs). Attributions and 

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

clarification of the associated information, are presented in Section 5.8. 

5.1.3

Mallinckrodt Chemical Works (MCW) operated its Uranium Division from 1942 to 1966, initially at its 

downtown location (Destrehan Street), then later at the WSP site. Construction of WSP was 

authorized by the U.S. Atomic Energy Commission (AEC) in 1955, and operations began in 1957. In 

1964, the plant employed 600 individuals, 80 of whom were in technical and managerial positions 

(MCW ca. 1964, p. 4; Mason 1977, p. 29). 

Background on Mallinckrodt Chemical Works and the Weldon Spring Plant

The functions performed at the two MCW facilities were essentially the same. One of the reasons for 

the construction of the new facility was the level of contamination present at the Destrehan Street site 

(Dupree 1979a, p. 3). Worker exposures to airborne uranium were, in some cases, many times the 

tolerance level.

During the 24 years of operation, the MCW Uranium Division processed uranium ores and 

concentrates into uranium trioxide (UO3), uranium hexafluoride (UF6), and uranium metal (Meshkov et 

al. 1986, p. 27). Uranium dioxide (UO2) was an intermediate product in the conversion of UO3 to UF4

(ORAUT 2013c). The feed material at the Destrehan Street plant included high-quality pitchblende (a 

natural uranium ore that includes the decay products generally in equilibrium). The WSP facility 

primarily handled uranium concentrates (e.g., yellowcake) received from uranium-milling facilities with 

lower concentrations of decay products. The historical documents report that WSP did not receive the 

high-quality pitchblende ores (Ingle 1991, pp. 3-7). The WSP processed thorium during the 3 years 

before closure in 1966 (see Section 5.6).
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Production at WSP waned during the later part of its operational period. In 1964, MCW published a 

document intended to encourage diversification and utilization of the existing facilities (MCW ca. 1964, 

pp. 5-29; Mason 1977, p. 29), which reflected that decline. The document described the plant 

facilities and briefly mentioned the operational safety program.

The WSP handled uranium concentrates. The primary concerns for radiation doses to WSP workers 

from uranium operations is internal deposition of uranium and beta dose from the short-lived decay 

products of uranium (234Th and 234Pa). External doses from the short-lived uranium decay products, 

including beta doses, are covered in Weldon Spring Plant – Occupational External Dosimetry

(ORAUT 2013a). This section covers internal deposition of the radionuclides of concern. 

There were four periods in the history of the WSP:

• Operational: 1957 to 1966 

• Transfer to the U.S. Army: 1967 to 1974

• Environmental monitoring: 1975 to 1985

• Remediation: 1985 to 2002 

The WS Quarry was initially transferred from the Army to the AEC in 1958. The operational period 

and the remediation period are of primary concern in relation to internal dosimetry at WSP, with some 

potential applicability of shutdown operations in 1967 before transfer to the U.S. Department of 

Defense in December 1967. The raffinate pits and the quarry were not transferred to the Army. From 

1968 to early 1969 decontamination and dismantling operations commenced to support the herbicide 

production. However, the defoliant project was canceled in February 1969 whereupon the chemical 

plant entered a care and custody status by the Army. The AEC didn’t have any contractors at the WS 

Raffinate Pits or the WS Quarry until August 1975 for environmental monitoring. From 1969 to 1981 

the status of the site did not change, and from 1981 to 1985 the site was placed in caretaker status. 

The WS Chemical Plant was transferred from the Army to the DOE in 1985, and remediation efforts 

began in 1985. EEOICPA does not cover the period when the WSP site was under control of the 

Army, and, therefore, this TBD does not cover that period.

Fuel for the Atomic Age – Completion Report on St. Louis Area Uranium Processing Operations,

1942-1967 (MCW 1967, pp. 154-166) describes the history and elements of the MCW health and 

safety programs. Uranium was initially considered primarily a heavy-metal poison, and the 

radioactivity level was considered low enough so that “small scale, short-term operations would not 

present a radiation problem.” MCW’s health and safety program was based on standard industry 

procedures for handling toxic chemicals. The MCW radiation safety program developed for the 

Destrehan Street site was applied at the WSP site when the transition between the two facilities was 

made.

The principal source of internal deposition of radionuclides for both the Destrehan and WSP sites was 

inhalation of dust generated during the operations, initial cleanup, and maintenance periods. 

Table 5-1 lists the history of dust-generating operations.

Tables 5-2 and 5-3 summarize the operations at WSP from the more detailed discussion in Weldon 

Spring Plant – Site Description (ORAUT 2013c). Ore concentrates containing 60% to 70% 

yellowcake were sampled on receipt at the facility. Some of the material was repackaged in drums 

and some sent directly for processing. The concentrates were digested with nitric acid to produce 

uranium-nitrate solution that was then purified by solvent extraction and denitrated to produce UO3. 

The UO3 was converted to green salt (UF4) in Building 201. Green salt was one of the final products 

of the plant. 
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Table 5-1. History of dust-generating activities.

a

Period Activity

1957–1966 Operation of the uranium feed materials plant

• Uranium concentrates converted to uranium trioxide, uranium tetrafluoride, 

and uranium metal

• Some thorium processing between 1963 and 1966

• Raffinate from processing removed to raffinate pits

• 14,500 metric tons of uranium materials received for processing and sampling 

per year between 1958 and 1964

December 1966 Plant closed

• Hopper and process lines emptied

• Dust collectors cleaned out

January 1967–? Site used as interim storage depot for yellowcake later shipped to other plants for 

refining and processing. 

1967 Buildings 103 and 105 transferred from the AEC to the Army for herbicide production

March 1968 Army started decontamination and equipment removal

December 1968 Construction of herbicide facility began; project terminated in early 1969 before 

renovation was complete. 

March 1968–June 

1969

Decontamination and equipment removal for Buildings 103 and 105 (see Table 5-2 for 

building operations description)

• About 1,000 metric tons rubble removed to the quarry

• About 2,000 metric tons scrap moved to Tennessee

• About 200 metric tons steel parts moved to Ohio

• About 100 metric tons uranium oxide removed from the buildings

1969–1985 Site remained essentially undisturbed

1985 Remediation initiated by DOE – Weldon Spring Site Remedial Action Project 

(WSSRAP)

October 1986 MK-Ferguson and Jacobs Engineering assumed responsibility for the WSSRAP

a. Adapted from Meshkov et al. (1986, pp. 27-33) and Lesperance, Siegel, and McKinney (1992, pp. 30-36).

Table 5-2. Potential internal radionuclide exposure for production buildings.

a

Bldg. 

no.

Building 

description Building operations Potential radionuclide exposures

101 Sampling Sampling of ore concentrates 

containing 60–70% U 

(yellowcake). Some material 

repackaged in drums – some 

sent to Bldg 103 for processing

U-nat dust, Ra-226, Th-230, Po-210, and Pb-210

Rn-222 and its short-lived decay products

Th-232 and decay products in 1966

103 Digestion Materials digested with nitric 

acid. Uranium bearing solution 

sent to Bldg 105 for 

purification. Materials returned 

for denitration after purification 

and sent to Bldg 201. 

U-nat dust (as yellowcake and UO3), Ra-226, Th-230, 

Po-210, and Pb-210

Rn-222 and its short-lived decay products

Th-232 and decay products starting November 1963

105 Purification Materials were purified by 

solvent extraction and returned 

to Bldg 103 for denitration.

Wet process but some potential for uranium or thorium 

(in 1966) dust exposure. Uranium would have been 

the major internal exposure component but Rn-222 

and decay product exposure possible

108 Acid 

recovery

Recovering and reconcentrating nitric acidRadon gas and its decay products

201 Green salt 

(UF4) plant 

Feed from Bldg 103 (after 

denitration) converted to UF4. 

Potential for uranium exposure as green salt dust or 

natural thorium (1965-66). No significant Th-230, Ra226, or decay product exposure.

301 Metals plant Mg used to convert UF4 to U 

metal. Rotary kiln used to 

convert U metal chips to U3O8. 

U fuel cores produced; 

Potential for uranium exposure as green salt dust and 

U3O8 or natural thorium (1965-66). No significant Th230, Ra-226, or decay product exposure.
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 11 of 60

Bldg. 

no.

Building 

description Building operations Potential radionuclide exposures

acceptable cores shipped to 

reactor sites. Th-232 and decay products starting November 1963

403 Chemical 

pilot plant

Small-scale chemical 

processes

U-nat dust, Ra-226, Th-230, Po-210, and Pb-210

Rn-222 and its short-lived decay products

Th-232 and decay products starting November 1963

404 Metallurgical 

pilot plant

Small-scale metallurgical 

processes

Potential for uranium exposure as green salt dust and 

U3O8 or natural thorium (1965–1966). No significant 

Th-230, Ra-226, or decay product exposure.

407 Analytical 

and 

research 

labs

Small-scale research and 

analytical work on products 

and processes.

U-nat dust, Ra-226, Th-230, Po-210, and Pb-210

Rn-222 and its short-lived decay products

Th-232 and decay products from 1965 to 1966

a. Adapted from ORAUT (2013c). 

Table 5-3. Products or intermediates in production buildings. 

Building no. Products or intermediates present

101 Ore concentrates, yellowcake

103 Yellowcake, UO3, 

Uranium solution in nitric acid

Purified uranium solution

Denitrated uranium solution

105 Uranium solution (UO3) in nitric acid

Purified uranium solution

108 Recovered nitric acid

201 Purified, denitrated uranium solution

UO3, UO2 (intermediate compound)

Green salt (UF4)

301 Green salt (UF4)

Uranium metal

U3O8 

Uranium metal was produced using magnesium to convert UF4 to the metallic form. A rotary kiln was 

used to convert uranium metal chips to U3O8. Uranium fuel cores were produced at WSP and 

shipped directly to reactor sites.

In contrast to the Destrehan Street facility, WSP did not deal with pitchblende ores. From 1946 to 

1955, MCW processed pitchblende ores with concentrations up to 60% uranium (Dupree 1998, p. 5) 

along with very high concentrations of 226Ra. There is no indication in the records that WSP ever 

processed the high-activity concentration pitchblende. Therefore, while inhalation of radium and

radon decay products was a potential contributor to dose at the Destrehan facility, it did not add 

significantly to worker doses at WSP. Dupree (1998, p. 6) stated the judgment that the uranium 

doses of the MCW population were considerably larger than doses from exposure to radon. 

The contributions to internal doses due to inhalation of uranium-bearing dust were similar in kind, if 

not in magnitude, for the MCW Destrehan and WSP workers because the operations at the two 

facilities were essentially the same. In fact, it appears from the worker records that many of the 

Destrehan workers were transferred to WSP. In general, because the Destrehan facility had 

accumulated high levels of surface contamination (Dupree 1979a), the early internal doses (1942 to 

1958) for MCW workers were potentially higher than for individuals who worked at WSP during the 

final years (1958 to 1966) of uranium processing by MCW. 
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5.1.4

The MCW radiation safety program in relation to internal doses evolved during the time when uranium 

ores and concentrates were being processed at the Destrehan Street site in St. Louis. As noted 

above, the initial concern was with the chemical toxicity of uranium rather than the potential radiation 

hazard. 

Radiation Protection Practices

As noted in ORAUT (2013c), the WSP employed approximately 600 workers during full production, of 

whom 300 were likely to have handled uranium. The site processed natural uranium, slightly enriched 

uranium, depleted uranium and, for a short period, natural thorium. The principal radionuclides to 

which workers were exposed were uranium isotopes, thorium isotopes, and 234Pa, the short-lived 

beta-emitting decay product of 238U. Because the site did not process pitchblende ore, 226Ra and its 

decay products were not present in significant quantities. 

Uranium ore concentrates were converted to UO3, UF4, and uranium metal at the WSP by methods

that included acid digestion, solvent extraction, and conversion to the metallic form by reaction with 

magnesium. Table 5-2 lists the specific buildings in which these processes occurred along with the 

likely contaminants in air in those locations.

MCW’s position, even as late as 1965 and as reflected in Summary of Health Protection Practices

(MCW 1965, p. 6), was, “There are no noteworthy health risks from radiation at the Weldon Spring 

Feed Material Plant.” The conclusion was based on the fact that the natural uranium feed materials 

processed at WSP were “essentially free of radioactive daughters which might cause chronic radiation 

exposure to be higher than AEC limits.” However, MCW (1965, p. 6) also stated:

It is the policy of the Weldon Spring Plant to prevent personal injury and to prevent 

harmful exposure of personnel to chemical irritants, to chemical toxicants, to ionizing 

radiation, or to any work conditions which may cause illness, impair health, or reduce 

the effectiveness of employees; to provide suitable monitoring and control programs; to 

obtain the cooperation of each employee in maintaining effective control programs; to 

inform individuals if measured radiation exposure exceeds AEC guide levels.

In line with this policy, urine and air sampling programs were conducted during the operational period 

to monitor the employees’ potential exposures.

No descriptions of internal dosimetry programs during the environmental monitoring period (1975 to 

1984) have been found. 

The remediation program initiated in 1985 involved demolition of the buildings involved in uranium or 

thorium processing, removal of contaminated materials from the WSQ, and stabilization and 

solidification of contaminated sludges. The internal dosimetry programs for the WSQ and WSRP 

were somewhat more rigorous than the program for the buildings that comprised the WSCP because 

the potential for exposure to thorium and radium was not as high in the WSCP. The high-grade 

pitchblende ores processed at the Destrehan Street facility were not present at the WSCP. Thorium 

was processed at WSP for a limited period. The WSP buildings involved in processing thorium were 

subject to a similar, slightly more rigorous, internal dosimetry program as the WSQ and WSRP.

The internal dosimetry programs during site remediation are described in detail in a series of revisions 

to the Internal Dosimetry Program Technical Basis Manual: Weldon Spring Site Remedial Action 

Project from 1991 to 2001 (the WSSRAP Technical Basis Manuals; DOE 1991, 1994, 1997, 

1998a,b,c, 2000a, 2001). The basic elements of the internal dosimetry program were urine and fecal 

bioassay, in vivo lung counting, and air sampling. Section 5.3 describes the urine and fecal bioassay 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 13 of 60

programs, Section 5.4 discusses in vivo lung counting, and Section 5.5 describes air-monitoring 

practices. 

5.2 SOURCE TERM

WSP processed uranium and thorium from feed materials to metal and intermediate products. The 

primary feed material was natural uranium in the form of yellowcake. WSP also processed depleted 

uranium and slightly enriched (up to 1%) uranium as well as natural thorium. Table 2-7 in ORAUT 

(2013c) gives the annual (fiscal year) mass receipts of each of these feed materials. Table 5-4 lists 

the quantity and percent of each type of feed material.

Table 5-4. Mass and percent of feed materials processed. 

Material Mass (kg) Percent of total mass

Natural U 122,015,977 98.43

Depleted U 167,823 0.14

Slightly enriched U 842,585 0.68

Natural thorium 941,347 0.76

Total 123,967,732 100

5.2.1

The isotopic composition and the factors to convert uranium mass to activity are necessary for intake 

and dose assessments. No site-specific isotopic data have been discovered for WSP. It is 

reasonable to assume that the composition of the natural and depleted uranium feed materials at 

WSP were the same as the default compositions for the DOE complex. Therefore, the DR should use 

the default values in the Integrated Modules for Bioassay Analysis (IMBA) program for natural and 

depleted uranium.

Isotopic Composition of Uranium

It is reasonable to assume that the slightly enriched uranium processed at Weldon Spring was 1% 

enriched with a specific activity of 0.783 pCi/μg (DOE 2001a, p. 2-7). Bioassay results in units of 

mass may be converted to activity using this value. After intake rates are calculated, doses are 

assessed assuming that the activity is 100% 234U.

Although uranium with enrichments of less than 1% might have been processed at WSP, it is 

favorable to claimants to assume 1% enrichment for all slightly enriched uranium at WSP.

5.2.2

The materials handled at WSP were uranium concentrates and, to some extent, natural thorium. The 

short-lived decay products of 238U, which are 234Th (24-d half-life) and 234mPa (1.175-min half-life), 

would have built into equilibrium before the material was handled. Thorium-234 and 234mPa emit beta 

particles. The dose from inhaled 234Th is included in the dose from 238U as it builds into equilibrium in 

the body in a relatively short period of time (less than eight months). The 234mPa beta is a high-energy 

beta and contributes to the external dose but, due to its short half-life, does not in itself contribute to 

internal dose. 

Uranium Decay Products

The primary source of decay products (230Th and 226Ra) for the materials processed at WSP would, on 

average, be the residuals in the uranium mill concentrates. These concentrations were not 

considered significant in the design of the radiation protection program at WSP. The Final Generic 

Environmental Impact Statement on Uranium Milling states that the upper range of values for 230Th 

and 226Ra in yellowcake product, based on published reports from the early 1960s, were 5% of the 

238U activity and 0.2% of the 238U activity, respectively (NRC 1980, p. 18). However, it has been 

observed that the early mills did not remove thorium as effectively as indicated by this reference. As 

noted in Weldon Spring Plant – Site Description (ORAUT 2013c), four raffinate pits were constructed 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 14 of 60

between 1958 and 1964 to contain process wastes from the WSP. Measurements of the activity 

concentrations in Raffinate Pits 1, 2, and 3 can be used to determine the relationship between 

thorium-230 and other impurities during the initial uranium processing (feed preparation and sampling)

in Building 101, during the transfer of ore concentrate from Building 101 to Building 103 for nitric acid 

digestion, and transfer of the uranyl nitrate slurry to the aqueous feed tanks in Building 105 before any 

separations occurred, as discussed in Section 5.6. The shorter lived decay products, 210Pb and 210Po, 

for which no raffinate measurements were made should be assumed to be present at the same 

activity as the 226Ra in raffinate pits.

The radionuclide of most concern in the raffinate pits was 230Th due to its high activity concentration 

and its radiotoxicity. The average and maximum activity concentrations are given in the Internal 

Dosimetry Program Technical Basis Manual, Rev. 0 for the WSSRAP (DOE 1991, p. 20).

The AEC acquired the WSQ in 1958 for deposition of thorium residues as well as uranium- and 

radium-contaminated building rubble, equipment, and soils from the Destrehan Street site. These 

materials were deposited at the WSQ between 1959 and 1966.

5.2.3

In contrast to the buildup of decay products of 238U, the conservative (favorable to claimant) 

assumption for natural thorium (232Th) is that the decay products have built up to equilibrium. 

Depending on the time since separation, the 228Ra (5.7-yr half-life) and 228Th (1.9-yr half-life) would be 

significant if not complete. Radium-228 is a beta emitter that decays to 228Th through 228Ac. Thorium228 decays by alpha emission to 224Ra (3.66-d half-life). For assessing dose from 232Th, it is assumed 

that 228Th is in equilibrium with 232Th and 228Ra is added at a ratio of 1-to-2 232Th-to228Ra (NIOSH 

2010a, p. 57). The Weldon Spring Plant – Occupational Environmental Dose (ORAUT 2013b) cites a 

DOE document (DOE 1986) in support of the assumption that 232Th was present during the 

operational period at an activity less than 1% of the natural uranium (ORAUT 2013b). 

Natural Thorium and Decay Products

Thoron (220Rn) is the second daughter product of 228Th, and after a couple of weeks following the 

processing of thorium ores for thorium purification, can be considered to be in full equilibrium with the 

parent, 228Th. Thorium-228 is generally in 40-65% equilibrium with 232Th for materials processed at 

the WSP. The degree of equilibrium is dependent upon both the decay of 228Th (without 

replenishment from the 5.7 yr half-life of 228Ra) after removal of the thorium daughters and the time it 

takes the 228Ra to build into equilibrium with 232Th. Figure 5-1 shows the decay and buildup of 228Th 

and daughters following processing.

Thoron was present and a portion was released during the processing and storage of thorium at the 

WSP and the associated waste storage locations. The thoron, with its subsequent daughter products, 

would act as a potential source of internal exposure in the thorium process buildings and at waste 

storage locations.

There are several aspects of the release and buildup of thorium daughter products that mitigate the 

exposure to thoron in process and storage configurations. These include the following:

• Because of the very short half-life of thoron (56 seconds), much of the isotope decays within 

the material matrix; therefore, the diffusion distance in soils is in the range of approximately 

one inch. Only the quantities of thoron in the first inch of material represent the source term. 

Table 5-5 provides information for thoron and its daughters.

• In process equipment, the release of thoron gas is largely contained within the containment 

system and/or the ventilation system. Essentially, the dose from exposure to thoron comes 

entirely from the daughters 212Pb and 212Bi. A thoron working level (WL) requires 375 pCi/L 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 15 of 60

Figure 5-1. Decay and buildup of 228Th and daughters following processing. 

Table 5-5. Thoron and isotopic daughters. 

Isotope Half life Atoms/7.5 pCi α MeV/atom PAECa MeV per 7.5 pCi

Radon-220 56 sec 23 14.6 335

Polonium-216 0.15 sec <1 b -

Lead-212 10.64 hr 15,476 7.8 1.21E+5

Bismuth-212 60.6 min 1,469 7.8 0.12E+5

Polonium-212 0.3 μs - b -

Thallium-208 3 min 73 no α -

Total 1.33E+5

a. PAEC = potential alpha energy concentration.

b. The alpha energy is emitted only by Po-216 and essentially Bi-212. In reality Bi-212 emits an alpha only 

36% of the time but the other 64% of the time it emits a beta then an alpha almost immediately through the 

decay of Po-212 (304 nsec half-life). The average alpha energy released by a Bi-212 decay is then 7.8 

MeV (6.07 Mev * 0.36 + 8.785 MeV * 0.64). Since Po-216 ultimately decays by two alphas (Po-216 and Bi212) the alpha energy released per Po-216 atom is 14.6 MeV. Pb-212 and Bi-212 both decay with an 

effective alpha energy of 7.8 MeV.

thoron and 216Po to produce measured daughter products 212Pb and 212Bi at 7.5 pCi/L 

assuming a conservative equilibrium factor of 0.02 [1].

The primary focus of the radiological safety programs for thorium was to define the air concentrations 

in the work place, the results of which provided the means of controlling worker exposures to levels 

below the permissible levels. Weldon Spring Plant records do not indicate specific analyses to define 

the concentrations of thoron daughter activities; however, it was standard practice to provide delayed 

counts on the air samples for up to 96 hours to allow the short-lived radon and thoron daughter 

products to decay.

As noted in Section 5.2.2, the 232Th concentrations in the WSRP were relatively small compared to the 

230Th and 226Ra concentrations, and thus would contribute only a very small fraction of the potential 
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dose. In contrast, the estimated maximum 232Th concentration in materials deposited at the WSQ was 

4200 pCi/g (DOE 1991, p. 18).

5.2.4

The extent of the processing of recycled uranium at WSP is not well known (ORAUT 2013c, 

Section 2.2.3). The DR should make the favorable to claimant assumption that all of the uranium 

processed at WSP after 1961 was recycled uranium. This assumption is consistent with that in Ohio 

Field Office Recycled Uranium Recovery Report (DOE 2000b, p. 118), which assumed that all 

uranium receipts at WSP after 1961 were recycled uranium in lieu of better information.

Recycled Uranium

Contaminant radionuclides in recycled uranium that could be dosimetrically significant are plutonium 

(assume 239Pu), neptunium (237Np), and technetium (99Tc). For the periods that included recycled 

uranium, the DR should consider the ppb conversion factors in ORAUT-TKBS-0017-5, Technical 

Basis Document for the Fernald Environmental Management Project – Occupational Internal Dose

(ORAUT 2004). These factors, when multiplied by the calculated uranium gram-value intake, result in 

the activity per gram of uranium of the contaminants at the levels of 100 ppb 239Pu, 3,500 ppb 237Np, 

and 9,000 99Tc. For example, 1 ppb-Pu/U equals 1 × 10-9 gm-Pu/gm-U. For 239Pu the ppb conversion 

factor is 62.89 pCi-Pu/gm-U per ppb, for 237Np the factor is 0.714 pCi-Np/gm-U per ppb, and for 99Tc 

the factor is 17.15 pCi-Tc/gm-U per ppb. 

5.2.5

The WSP handled uranium in several different forms. These forms are listed in Table 5-2 along with 

the facility location (building) in which they were most likely to have contributed significantly to a 

worker’s uranium intake. 

Solubility Classification and Absorption Type

Feed materials likely to have been received at the WSP site came primarily from conventional 

uranium mills and as vanadium-milling wastes in the form of yellowcake (Meshkov et al. 1986). The 

specific uranium compounds received at the WSP have not been identified. However, according to 

the Final Generic Environmental Impact Statement on Uranium Mills, “the yellowcake product from an 

acid leaching plant is a mixture of chemical complexes: diuranates, hydrated oxides, basic uranyl 

sulfate and other ions” depending on the drying temperature (NRC 1980, p. 18). 

The WSP remediation contractor collected samples of materials from the areas in WSCP buildings 

with the highest levels of contamination (DOE 2001b, p. 14). The samples were tested for lung 

solubility class using simulated lung fluid leachate tests. Based on those test results, the contractor 

assumed specific fractions of solubility classes for contaminants in the building sampled. Table 5-6 

lists those fractions for most of the WSCP buildings. The solubility classes in that study were denoted 

using the International Commission on Radiological Protection (ICRP) Publication 30 designations of 

D, W, and Y (ICRP 1979). These classes were generally assigned to the more recent ICRP 

absorption types of F, M and S, respectively (ICRP 1995).

In general, uranium metal dust in Building 404 (Metals Pilot Plant) would most likely have been 

insoluble or type S. While the residues from this building were not analyzed in the solubility study, the 

solubility class for uranium in residues from Building 301 (Metals Plant) was found to be 

approximately 90% class Y and 10% class D.

The information in Table 5-6 is provided for general information only and guidance that says the most 

favorable material solubility type (F, M, or S) should be selected in the absence of definitive 

information for a particular dose reconstruction.
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Table 5-6. Solubility class summary (DOE 2001).

Area

U-234U-235U-238 Th-232 Th-230 Th-228

DWYDWYDWYWYWYWY

Bldg 101 0.41- 0.59 0.28- 0.72 0.41- 0.59 0.20 0.80- 1.0 0.29 0.71

Bldg 103 0.20 0.50 0.30 0.20 0.44 0.36 0.75 0.25-- 1.0- 1.0- 1.0

Bldg 105 0.20 0.50 0.30 0.20 0.44 0.36 0.75 0.25-- 1.0- 1.0- 1.0

Bldg 108 0.19 0.20 0.61 0.14 0.47 0.39 0.19 0.20 0.61- 1.0- 1.0- 1.0

Bldg 201 0.44- 0.56 0.51- 0.49 0.42- 0.58- 1.0- 1.0- 1.0

Bldg 301 0.12- 0.88 0.09- 0.91 0.12- 0.88 0.03 0.97 0.09 0.91 0.03 0.97

Bldg 403 0.19 0.20 0.61 0.14 0.47 0.39 0.19 0.20 0.61- 1.0- 1.0- 1.0

Bldg 406 1.0-- 1.0-- 1.0-- 0.35 0.65 0.68 0.32 0.65 0.35

Bldg 408 1.0-- 1.0-- 1.0-- 0.35 0.65 0.68 0.32 0.65 0.35

Bldg 410 1.0-- 1.0-- 1.0-- 0.35 0.65 0.68 0.32 0.65 0.35

Bldg 417 1.0-- 1.0-- 1.0-- 0.35 0.65 0.68 0.32 0.65 0.35

Pit 3 1.0 1.0-- 1.0--- 1.0- 1.0- 1.0

Q-Bench 0.54 0.12 0.46 0.07 0.47 0.49 0.13 0.38- 1.0- 1.0- 1.0

Q-Sump 0.61 0.52 0.47 0.48 0.58- 0.42 0.10 0.90- 1.0 0.10 0.90
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 18 of 60

5.2.6

Lacking specific information on the particle size for airborne uranium at WSP since there were no 

particle size distribution studies conducted, the ICRP (1994b) value of 5 μm aerodynamic median 

activity diameter (AMAD) should be used (ORAUT 2007b). 

Particle Size

5.3 IN VITRO MEASUREMENTS

5.3.1

5.3.1.1 Uranium

Operational Period (1957 to 1966)

Urine bioassay was the primary method of determining uranium intakes during the production phase. 

The bioassay program was set up in accordance with the general health physics practices of the 

period. Grab (single void) urine samples were collected and analyzed for uranium by 

photofluorimetric analysis. Results of the photofluorimetric analysis were reported as the mass of 

uranium in milligrams, or sometimes micrograms, per liter of urine. 

5.3.1.1.1

MCW (1965) described the uranium urine bioassay program as follows:

Routine Sampling Program

The routine sampling program seeks to have one or more persons from each 

operational group in the plant sample[d] each week. When a person represents his 

group in the sample, he is asked to give samples on (1) Monday a.m., (2) Friday p.m., 

and (3) Monday a.m. The Monday sample tends to show the amount semi-fixed in the 

body, the Friday sample reflects the daily uptake. The sample from each person is 

analyzed separately and entered in his summary.

Each exposed person is scheduled three or more times per year, more frequently if 

there is reason to suspect increased exposure. The rotation of group representation 

tends to show the average level and variation within each plant area. Unexposed 

persons are scheduled less frequently to provide a control base.

A review of urine data in worker files available up to November 2004 indicates that this program was 

implemented in about October 1959. From 1957 to mid-May 1959, urine sampling was apparently 

conducted at random times during the week. From mid-May 1959 through September 1959, sampling 

occurred primarily on Fridays. Starting in October 1959, the routine sampling program described 

above appears to have been implemented and continued through 1966.

Workers also submitted urine samples as part of the hiring and termination processes.

5.3.1.1.2

A repeat sample was required if the result of the Monday morning sample was greater than 100 μg/L 

or if the result of the Friday afternoon sample was greater than 200 μg/L.

Special Sampling Program

Special urine samples were required for known or suspected significant intakes. The DR can identify 

the results of these special samples in worker files either by the code S or by handwritten notes on the 

original urine data cards.
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5.3.1.1.3

No information has been found about the details and quality assurance of the photofluorimetric 

system and data analysis.

Data Reporting Levels and Minimum Detectable Amounts

There were apparently no uranium urine results censored or reported as less than the detection level 

because the recorded data reflect continuous increments of 0.001 mg/L starting at 0.000 mg/L, and no 

less-than values have been observed in the files. It is not known whether a blank was subtracted. If 

so, negative results were reported as 0.000 mg/L. 

Because a site-specific value of the minimum detectable amount (MDA) is not available for WSP, a 

value based on reported MDAs for photofluorimetric systems at other facilities in the 1960s was 

considered. A range of values of 0.001 to 0.014 mg/L has been cited for detection levels of unknown 

pedigree. An MDA value of 0.008 mg/L was determined from original urinalysis data logs at Rocky 

Flats (ORAUT 2007) and was based on modern MDA concepts (HPS 1996, pp. 44-49). The Rocky 

Flats MDA value included the contribution of a blank subtraction and was based on a 100λ (100 μL) 

aliquot from a 24-hour urine sample. This MDA value has a known pedigree and is recommended as 

the MDA for WSP, although there may be some differences in methods (e.g., grab sample versus 24-

hour sample, unknown blank subtraction versus blank subtraction, and unknown volume of the aliquot 

versus 100λ (100 μL)). 

An estimate of the uranium MDA can be derived from Dupree (1979b, p. 26), who cites an average 

value of 0.002 ± 0.002 mg/L for “people off the street.” Based on the consideration that the cited 

standard deviation represents the process standard deviation divided by the calibration factor, the 

estimated MDA is 3.3 times 0.002 mg/L, which is 0.007 mg/L. This value supports the use of 0.008 

mg/L as a reasonable MDA for WSP uranium urine data. 

5.3.1.1.4

It is not known whether the WSP uranium urine data were adjusted for excretions of environmental 

sources of uranium. It is likely that the data were not adjusted. Measurements more sensitive than 

the Dupree (1979b, p. 26) value of 0.002 ± 0.002 mg/L made during the remediation period on 

persons in the WSP vicinity not occupationally exposed to uranium indicated a geometric mean of 

0.05205 μg/L (0.00005205 mg/L), “with variation as high as 0.3016 μg uranium per liter of urine at two 

standard deviations (e.g., 94.5% of the population) (DOE 2001b, p. 23).” Because these values are 

very low, DRs should disregard this source of interference and use the WSP uranium urine data as 

recorded.

Interferences and Uncertainties

Possible contamination of a urine sample from uranium on the hands or clothing cannot be ruled out, 

especially for grab samples after work. Resampling for results over the action levels 

(Section 5.3.1.1.2) would catch cases of excessive contamination. There are numerous uncertainties 

generic to fluorometric uranium measurements. Sample-specific uncertainties were not recorded. 

There is no reason to believe that the quality of WSP measurements was significantly different from 

the quality at other facilities of that era.

5.3.1.1.5

The DR can expect to find one to three reports of uranium urine data in a worker’s files:

Reporting Formats and Codes

• A photocopy of the original, handwritten urine data cards

• A computer printout, Uranium Urine (MCWURWS)

• A computer printout, Uranium Urine (MCWURDES)
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Many WSP workers were assigned to the St. Louis facility before being assigned to WSP. All of the 

urine data reports contain data from both facilities because WSP was a division of MCW. If no other 

information is available, DRs should assume that uranium urine data starting in 1957 is associated 

with WSP. All three reports seem to report the same data, but sometimes in different manners, and 

all units are milligrams per liter. 

Clock No. on the original data cards and Clock Badge on the computer printouts is the worker’s 

employee number. DRs should examine the Clock No. on the photocopies of the original data cards 

to ensure that the numbers are consistent with the worker because investigators have found 

discrepancies in the record. The sample date is recorded and should be interpreted as the date of 

excretion of the grab sample. On some original data cards, this date is listed in the DUE column.

Dept. (on the original data cards) or DEPTJOB (on the computer printouts) is either a cost center 

number, a job or department title, or sometimes a mixture of the two. Table 5-7 lists the cost center 

codes. The DR should be aware that the cost center codes for the work groups in the production 

facilities and services groups (except for Maintenance) changed in early 1963.

Table 5-7. Cost center codes for workers. 

Work group

Cost center code

To early 1963 Starting 1963

General operations

General Engineering 110 (a)

Chemical Technology 120 (a)

Metal Technology 121 (a)

Plant Services 140 (a)

WSP

Administration 300 (a)

Production facilities

Sampling Plant

Sampling 310 110

Repackaging 321 110

Refinery

Digest – Raffinate 320 120

Pot Room 320 120

Extraction 320 120

Green Salt Plant

Green Salt 330 150

Metal Plant

Dingot Reduction 360, 340 180

Extrusion, General 359 200

Extrusion, Gamma (Dingot Extrusion) 361, 351 210

Extrusion, Alpha (Dingot to Rods) 364, 352 220

Scrap Activity 358 280

Vacuum Outgassing 363 (a)

Core Fabrication 362, 365 to 369 250

Services groups or facilities

Maintenance 370 370

Storeroom, Receiving, & Shipping 372 530

Disposal of Construction Inventory 373 (a)

Instrument Shop 374 (a)

Boiler House 376 (a)

Water Plant 378 (a)

Warehouse 380 550

Engineering 390 560

Plant Protection 392 510
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Work group

Cost center code

To early 1963 Starting 1963

Laundry 394 (a)

Custodians 396 520

Occupancy, Labor 398 (a)

Technical Division

Administrative 400 (b)

Metallurgical Development (Building 404) 410 (b)

Special Projects 460 460

Process Development (Building 403) 420, 450 (b)

Scrap Plant No Code 290

Research Laboratory 430 (b)

Analytical Laboratory 440 (b)

Administration Division

Uranium Division Management 600 600

Salaried Personnel Administration 605 (b)

Accounting 610 610

Purchasing 620 (b)

Motor Pool 621 (b)

Administrative Services 630 (b)

Production Control 640 640

Health and Safety 650 (b)

General Division Expenses 660 (b)

Data Processing 670 670

Industrial Relations 680 (b)

Cafeteria 685 (b)

a. Unknown – likely discontinued or merged with another cost center.

b. Unknown – likely the same as the code to early 1963.

The sample result (in milligrams per liter) is reported under columns headed by SCHEDULED, 

RESULTS, or no heading on the original data cards, and under the heading MGUPERL on the 

computer printouts. On the original data cards, the sample result is frequently followed by two dashes 

and a number, which has been assumed to be the pH of the urine sample.

Following the sample result on the original data cards or in the column headed by SAMTYPE on the 

computer printouts, an asterisk indicates a Friday afternoon sample. However, not all Friday samples 

are flagged by the asterisk. If the sample day is important, the DR could determine the actual day 

using the Microsoft Excel WEEKDAY function, which converts a date to a day-of-the-week code 

where 1 is Sunday, 2 is Monday, and so forth. Written on the original data cards after the result or 

under SAMTYPE on the computer printouts are notations that indicate pre-employment samples 

(code P), termination samples (code T), or special samples (code S).

5.3.1.1.6

Urine bioassays were performed routinely as described in Section 5.3.1.1.1. Urine samples were 

obtained weekly from representative individuals in areas of WSP where uranium was handled. The 

data from the representative individuals were intended to be used to assess the intake by coworkers 

so that the work group was continuously monitored. Individual urine bioassay results supplemented 

by contemporaneous data from coworkers could provide the best measure of that person’s uranium 

intake because the sampling for an individual worker could have occurred during quiescent 

operational periods. 

Work Group Data

Because most of the work group urine data summaries have not been discovered, the data have been 

recreated. Approximately 28,000 urine bioassay results were recorded during the operational period 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 22 of 60

(1958 to 1966). Tables 5-8 to 5-17 provide median, 95th-percentile, and maximum concentrations for 

routine urine bioassay samples by year. 

Table 5-8. Composite uranium urine data summary. 

Year Type of analysis

Number of 

records

Urine uranium concentration (mg/L)

Median 95th percentile Maximum

1958 All 1,872 0.005 0.025 0.203a

Routine 1,714 0.005 0.025 0.203

Routine–Monday 321 0.004 0.021 0.047

Routine–Friday 158 0.006 0.023 0.069

1959 All 2,285 0.005 0.024 0.339

Routine 2,090 0.006 0.026 0.339

Routine–Monday 1,124 0.004 0.017 0.041

Routine–Friday 522 0.009 0.048 0.339

1960 All 4,396 0.012 0.038 0.759

Routine 4,246 0.012 0.040 0.759

Routine–Monday 2,602 0.010 0.026 0.088

Routine–Friday 1,347 0.018 0.068 0.759

1961 All 4,184 0.011 0.036 0.344

Routine 4,077 0.011 0.036 0.344

Routine–Monday 2,608 0.010 0.025 0.062

Routine–Friday 1,279 0.018 0.050 0.344

1962 All 3,083 0.010 0.032 0.700

Routine 2,954 0.007 0.024 0.700

Routine–Monday 1,847 0.008 0.020 0.064

Routine–Friday 946 0.013 0.044 0.700

1963 All 3,481 0.014 0.042 0.340

Routine 3,358 0.014 0.041 0.336

Routine–Monday 1,922 0.014 0.029 0.336

Routine–Friday 1,074 0.018 0.056 0.258

1964 All 3,476 0.012 0.052 0.626

Routine 3,264 0.012 0.048 0.626

Routine–Monday 1,767 0.012 0.029 0.282

Routine–Friday 1,049 0.014 0.060 0.626

1965 All 2,980 0.010 0.036 0.812

Routine 2,804 0.009 0.032 0.812

Routine–Monday 1,460 0.009 0.027 0.347

Routine–Friday 869 0.010 0.045 0.318

1966 All 2,145 0.006 0.029 0.459

Routine 1,680 0.007 0.028 0.459

Routine–Monday 882 0.008 0.025 0.066

Routine–Friday 526 0.007 0.039 0.459

a. A recorded value of 7.8 mg/L was deleted from the data set as an outlier. This does not 

affect the median or 95th-percentile value.

The data were analyzed by major work location, cost center or job description, and sample day 

(Monday or Friday). In some cases, cost centers were combined to increase the number of individual 

analyses. The data set for 1958 includes a mixture of WSP and Destrehan Street workers and was 

coded in the original records by job description or work location rather than cost center. A recorded 

value of 7.8 mg/L was deleted from the 1958 data set as an outlier. This does not affect the median 

or 95th-percentile value. In cases where there were five or fewer records, the tables contain only the 

maximum urine bioassay result.
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 23 of 60

Table 5-9. Uranium urine data summary by cost center for 1958. 

Cost center

All samples (mg/L) No. of 

Median 95th percentile Maximum records

300-370 0.002 0.010 0.012 16

400-440 0.000 0.011 0.011 7

600-670 0.003 0.010 0.012 29

Acct 0.003 0.006 0.010 17

Adm 0.004 0.015 0.017 16

Aec 0.003 0.010 0.010 9

Anal Lab 0.004 0.012 0.039 105

Boiler 0.002 0.008 0.008 14

Chem Op 0.004 0.030 0.061 24

Decon 0.003 0.006 0.006 9

Dx 0.012 0.029 0.040 25

Engineering 0.003 0.014 0.020 33

Furnace 0.008 0.019 0.022 24

Gr Salt 0.005 0.018 0.025 30

Guard 0.004 0.013 0.014 41

H&S 0.004 0.007 0.016 28

Instruments 0.002 0.018 0.022 37

Lab 0.004 0.018 0.018 27

Laundry 0.005 0.012 0.024 15

Maint 0.006 0.020 0.098 79

Metals 0.006 0.015 0.101 94

Mtns 0.006 0.025 0.064 137

Office 0.002 0.008 0.016 97

Pilot Plant 0.005 0.014 0.028 16

Pl4 0.010 0.026 0.078 24

Pl6 0.012 0.045 0.048 58a

Pl7 0.016 0.053 0.062 54

Porter 0.005 0.016 0.020 25

Prod Dev. 0.004 0.010 0.011 25

Refinery 0.009 0.041 0.203 127

Research Lab 0.004 0.012 0.022 49

Sampling 0.011 0.040 0.049 31

Stores 0.004 0.009 0.029 16

Warehouse 0.004 0.017 0.051 48

5.3.1.2 Thorium

There is no indication discovered so far that a routine urine-sampling program was implemented for 

thorium. No urine bioassay data for thorium have been found in the worker files.

5.3.2

No personnel bioassay monitoring appears to have been conducted during this period.

Environmental Monitoring Period (1975 to 1984)

5.3.3

An extensive, state-of-the-art bioassay monitoring program was conducted during the 1991 to 2001 

period to detect intakes greater than 100 mrem committed effective dose equivalent (CEDE). This 

program is well defined in the WSSRAP Technical Basis Manual revisions (DOE 1991, 1994, 1997, 

1998a,b,c, 2000a, 2001b). The focus of the program was to conduct bioassay based on workplace 

action levels for air sampling, nasal wipe analysis, and wipe analysis of the inside of respirators at the 

Remediation Period (1985 to 2002)
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 24 of 60

Table 5-10. Uranium urine data summary by cost center for 1959. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

110-140 0.004 0.010 0.010 14 N/Aa N/A 0.012 2

300-310 0.002 0.015 0.022 35 0.020 0.048 0.048 12

320 0.006 0.016 0.023 24 0.014 0.088 0.104 40

321-340 0.008 0.018 0.020 25 0.017 0.043 0.067 18

359-369 0.005 0.016 0.034 60 0.012 0.036 0.049 47

370 0.009 0.016 0.019 40 0.014 0.058 0.206 25

372 0.003 0.012 0.017 17 0.004 0.023 0.023 8

374 0.006 0.015 0.018 27 0.006 0.024 0.024 13

376-379 0.002 0.027 0.039 26 0.003 0.035 0.035 13

380 0.004 0.019 0.041 53 0.004 0.071 0.094 32

390 0.003 0.007 0.012 40 N/A N/A 0.003 3

392 0.005 0.016 0.028 38 0.003 0.023 0.026 18

394-396 0.006 0.019 0.019 30 0.006 0.023 0.023 14

400-430 0.002 0.013 0.020 24 N/A N/A 0.041 5

440 0.003 0.014 0.020 73 0.004 0.012 0.015 16

450 0.004 0.012 0.016 20 0.018 0.079 0.080 15

460 0.005 0.017 0.020 21 0.038 0.156 0.156 14

500-640 0.002 0.014 0.024 79 N/A N/A 0.008 5

650 0.003 0.009 0.011 28 N/A N/A 0.007 3

670-688 0.006 0.013 0.023 27 N/A N/A 0.011 1

320 Ref 0.005 0.018 0.028 26 0.020 0.057 0.059 22

330 Gr Salt 0.012 0.020 0.020 20 0.017 0.023 0.023 11

370 Misc 0.006 0.015 0.019 16 0.014 0.051 0.051 12

370 Mtns 0.006 0.019 0.035 164 0.011 0.048 0.083 94

370 Mtns Misc 0.006 0.012 0.014 42 0.011 0.025 0.044 24

Office 0.002 0.009 0.018 17 0.006 0.009 0.009 8

Other 0.003 0.012 0.015 66 0.004 0.022 0.022 28

a. N/A = not applicable (fewer than five records).
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 25 of 60

Table 5-11. Uranium urine data summary by cost center for 1960. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.001 0.019 0.020 18 N/Aa N/A 0.013 1

300 0.007 0.018 0.019 32 N/A N/A N/A 0

310 0.015 0.026 0.038 39 0.033 0.071 0.085 29

320 0.012 0.026 0.060 204 0.019 0.062 0.110 129

321-340 0.014 0.025 0.032 76 0.026 0.055 0.080 57

359 0.013 0.022 0.029 41 0.025 0.069 0.161 22

360-369 0.013 0.028 0.061 105 0.016 0.058 0.092 68

370 0.012 0.028 0.064 106 0.015 0.053 0.162 58

378-379 0.012 0.027 0.027 14 0.015 0.082 0.082 7

380-398 0.008 0.020 0.028 120 0.012 0.041 0.088 34

400-430 0.007 0.030 0.071 38 0.018 0.050 0.050 9

450 0.010 0.027 0.058 91 0.027 0.092 0.161 39

460 0.013 0.029 0.060 64 0.026 0.060 0.078 37

600-680 0.007 0.018 0.028 114 N/A N/A 0.017 1

320 Ref 0.012 0.029 0.079 155 0.032 0.126 0.216 107

330 Gr Salt 0.013 0.029 0.030 43 0.022 0.070 0.092 29

360 Metals 0.013 0.026 0.029 29 0.017 0.036 0.058 24

361 Conv Ext 0.021 0.058 0.058 14 0.036 0.075 0.075 7

362 Cons Fab 0.010 0.021 0.029 18 0.014 0.023 0.030 16

370 Misc 0.010 0.025 0.037 63 0.014 0.064 0.084 36

370 Mtns 0.010 0.026 0.038 375 0.017 0.064 0.228 256

370 Mtns Elec 0.010 0.024 0.028 52 0.013 0.063 0.076 34

370 Mtns Mech 0.010 0.014 0.014 11 0.008 0.012 0.012 6

370 Mtns Misc 0.012 0.028 0.036 42 0.017 0.036 0.038 31

370 Mtns Pipe 0.012 0.030 0.060 31 0.018 0.064 0.069 20

370 Painter 0.015 0.027 0.032 16 0.013 0.162 0.162 11

372 Stores 0.010 0.024 0.032 38 0.012 0.027 0.029 20

374 Inst 0.010 0.024 0.032 61 0.012 0.036 0.060 38

376 Boiler 0.005 0.018 0.020 26 0.006 0.024 0.024 7

380 Warehouse 0.010 0.026 0.034 115 0.012 0.048 0.067 73

392 Plant Pr 0.006 0.017 0.021 41 0.011 0.030 0.030 13

396 Custodian 0.014 0.027 0.088 48 0.014 0.081 0.759 33

440 Anal Lab 0.010 0.021 0.036 92 N/A N/A N/A 0

650 H&S 0.007 0.017 0.019 31 NA NA 0.003 1

Other 0.010 0.027 0.079 124 0.018 0.066 0.266 45

a. N/A = not applicable (five or fewer records).
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 26 of 60

Table 5-12. Uranium urine data summary by cost center for 1961. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

110-300 0.007 0.015 0.024 61 0.019 0.036 0.036 8

310 0.016 0.028 0.038 35 0.031 0.062 0.064 25

320 0.011 0.027 0.035 205 0.021 0.055 0.140 127

321-350 0.014 0.034 0.052 86 0.025 0.054 0.066 58

359 0.012 0.034 0.039 44 0.022 0.050 0.052 28

360-369 0.011 0.028 0.049 154 0.021 0.040 0.058 83

370 0.011 0.026 0.062 128 0.018 0.048 0.344 78

380 0.009 0.021 0.029 30 0.012 0.030 0.032 25

390 0.005 0.013 0.022 61 N/Aa N/A N/A 0

392-398 0.008 0.022 0.039 107 0.011 0.037 0.063 79

400-460 0.011 0.026 0.050 200 0.020 0.050 0.118 66

600-680 0.007 0.019 0.027 183 N/A N/A 0.013 3

320 Ref 0.012 0.026 0.040 154 0.022 0.081 0.127 93

330 Gr Salt 0.010 0.024 0.030 49 0.024 0.054 0.055 27

370 Mill 0.018 0.025 0.025 10 0.019 0.031 0.031 6

370 Misc 0.009 0.022 0.028 57 0.015 0.064 0.080 39

370 Mtns 0.011 0.025 0.042 341 0.018 0.042 0.058 223

370 Mtns Elec 0.009 0.022 0.034 42 0.019 0.032 0.052 24

370 Mtns Misc 0.010 0.025 0.029 86 0.016 0.044 0.050 54

372 Stores 0.010 0.023 0.027 33 0.010 0.026 0.027 22

374 Inst. Shop 0.008 0.021 0.024 45 0.014 0.040 0.050 28

376 Boiler 0.010 0.020 0.021 26 0.008 0.010 0.010 7

380 Warehouse 0.009 0.022 0.026 105 0.011 0.032 0.036 66

392 Plant Pr 0.009 0.018 0.021 48 0.008 0.014 0.015 15

440 Anal Lab 0.009 0.019 0.029 96 N/A N/A N/A 0

Other 0.008 0.024 0.035 123 0.021 0.061 0.094 43

N/A = not applicable (five or fewer records).
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 27 of 60

Table 5-13. Uranium urine data summary by cost center for 1962. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

120-290 0.009 0.020 0.035 17 0.028 0.071 0.071 11

300-310 0.009 0.021 0.028 63 0.032 0.064 0.090 22

320 0.009 0.022 0.040 138 0.017 0.048 0.700 85

321 0.014 0.034 0.034 11 0.026 0.141 0.141 14

340 0.010 0.027 0.038 34 0.018 0.046 0.046 20

359 0.008 0.020 0.020 29 0.018 0.035 0.036 20

361-369 0.010 0.020 0.024 43 0.014 0.040 0.110 31

360-369 0.009 0.018 0.018 32 0.015 0.039 0.039 24

370 0.010 0.024 0.028 117 0.014 0.051 0.116 92

380 0.006 0.015 0.026 104 0.010 0.024 0.031 74

390 0.006 0.019 0.022 54 N/Aa N/A N/A 0

392 0.005 0.014 0.014 13 N/A N/A 0.010 3

394-398 0.010 0.016 0.026 34 0.011 0.020 0.020 30

396 0.007 0.018 0.018 24 0.014 0.026 0.026 17

400-460 0.008 0.022 0.047 213 0.016 0.040 0.044 44

600-680 0.006 0.015 0.020 152 N/A N/A 0.002 1

320 Ref 0.009 0.022 0.064 77 0.015 0.043 0.056 37

330 Gr Salt 0.009 0.027 0.028 29 0.013 0.028 0.041 17

370 Misc 0.010 0.026 0.046 33 0.013 0.037 0.044 26

370 Mtns 0.009 0.024 0.047 244 0.013 0.038 0.652 190

370 Mtns Elec 0.008 0.015 0.018 26 0.010 0.022 0.068 21

370 Mtns Misc 0.009 0.016 0.027 30 0.015 0.030 0.220 21

372 Stores 0.008 0.015 0.020 37 0.007 0.019 0.027 24

374 Inst. Shop 0.009 0.015 0.029 42 0.014 0.026 0.028 21

376 Boiler 0.007 0.016 0.026 23 0.007 0.012 0.012 6

392 Plant Pr 0.007 0.013 0.017 43 0.008 0.012 0.012 14

Other 0.008 0.024 0.028 92 0.011 0.055 0.181 33

NA = not applicable (five or fewer records).
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 28 of 60

Table 5-14. Uranium urine data summary by cost center for 1963. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.021 0.035 0.259 34 0.027 0.078 0.178 24

120 0.016 0.035 0.085 135 0.024 0.078 0.143 66

150 0.012 0.023 0.023 14 0.014 0.035 0.035 9

180 0.014 0.027 0.034 33 0.020 0.034 0.036 22

200 0.012 0.028 0.033 18 0.030 0.120 0.120 14

250-290 0.019 0.043 0.052 50 0.027 0.078 0.100 26

300-310 0.014 0.027 0.048 34 0.037 0.054 0.083 16

320 0.014 0.026 0.052 148 0.027 0.112 0.185 72

321-340 0.014 0.037 0.046 42 0.025 0.068 0.137 28

350-369 0.014 0.032 0.336 75 0.016 0.078 0.205 47

370 0.016 0.040 0.059 115 0.021 0.054 0.122 75

372-378 0.010 0.026 0.035 97 0.013 0.036 0.042 43

380-396 0.012 0.023 0.029 87 0.016 0.035 0.039 39

400-460 0.012 0.029 0.048 115 0.014 0.056 0.081 47

500-510 0.008 0.021 0.025 43 0.012 0.024 0.045 22

520-530 0.013 0.027 0.028 51 0.012 0.035 0.041 25

550 0.013 0.025 0.027 46 0.012 0.027 0.041 36

600-690 0.008 0.021 0.065 91 0.008 0.023 0.027 33

330 Gr Salt 0.013 0.027 0.029 46 0.020 0.042 0.044 21

370 Elec 0.012 0.023 0.025 28 0.018 0.059 0.059 13

370 Mill 0.016 0.029 0.029 31 0.027 0.058 0.258 25

370 Misc 0.015 0.027 0.091 91 0.016 0.043 0.111 59

370 Mtns 0.014 0.028 0.041 178 0.019 0.052 0.095 117

370 Pipe 0.014 0.029 0.045 22 0.029 0.058 0.058 14

380 Warehouse 0.012 0.027 0.029 69 0.012 0.025 0.035 50

392 Plant Pr 0.010 0.026 0.027 50 0.014 0.027 0.034 31

Other 0.012 0.029 0.070 95 0.015 0.036 0.087 55
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 29 of 60

Table 5-15. Uranium urine data summary by cost center for 1964. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.020 0.066 0.282 84 0.032 0.103 0.232 75

120 0.012 0.029 0.090 197 0.016 0.071 0.182 115

150 0.012 0.026 0.028 57 0.014 0.050 0.067 36

180 0.017 0.043 0.054 32 0.018 0.056 0.226 29

200 0.019 0.053 0.078 32 0.018 0.088 0.114 25

290 0.012 0.038 0.054 32 0.013 0.070 0.106 16

310-321 0.014 0.036 0.106 72 0.018 0.063 0.118 53

350-360 0.011 0.073 0.104 48 0.008 0.085 0.358 26

370 0.012 0.025 0.062 138 0.017 0.066 0.626 86

372-380 0.012 0.022 0.026 44 0.010 0.025 0.027 25

380-390 0.010 0.026 0.029 66 0.012 0.040 0.084 34

392 0.008 0.023 0.026 41 0.008 0.018 0.040 25

400-460 0.012 0.027 0.068 94 0.013 0.048 0.165 34

500-510 0.008 0.020 0.024 72 0.006 0.022 0.024 44

520 0.010 0.021 0.029 64 0.012 0.033 0.044 35

530 0.010 0.026 0.064 28 0.008 0.017 0.052 16

550-569 0.010 0.022 0.029 118 0.008 0.026 0.031 75

600-690 0.008 0.017 0.027 73 0.010 0.016 0.024 14

370 Elec 0.012 0.024 0.029 40 0.011 0.050 0.052 28

370 Mill 0.014 0.026 0.028 33 0.024 0.052 0.081 22

370 Misc 0.012 0.025 0.028 99 0.009 0.046 0.058 63

370 Mtns 0.017 0.062 0.125 102 0.013 0.029 0.064 57

370 Pipe 0.013 0.029 0.053 37 0.013 0.052 0.090 27

Other 0.012 0.027 0.042 95 0.011 0.050 0.059 52
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 30 of 60

Table 5-16. Uranium urine data summary by cost center for 1965. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.018 0.034 0.042 41 0.018 0.094 0.152 19

120 0.011 0.025 0.047 156 0.014 0.072 0.318 102

150 0.011 0.035 0.347 61 0.011 0.036 0.049 35

180 0.020 0.045 0.052 45 0.024 0.102 0.146 26

200 0.015 0.024 0.028 19 0.016 0.069 0.088 16

290 0.010 0.015 0.021 26 0.010 0.025 0.039 19

310-321 0.011 0.042 0.065 29 0.012 0.024 0.024 16

350 0.007 0.018 0.027 36 0.005 0.012 0.024 19

370-379 0.008 0.022 0.045 165 0.010 0.055 0.193 120

380-390 0.010 0.026 0.050 46 0.011 0.026 0.033 27

392 0.004 0.010 0.015 28 0.006 0.015 0.016 18

400-460 0.009 0.039 0.078 85 0.017 0.045 0.058 25

500-510 0.006 0.015 0.016 60 0.005 0.012 0.016 34

520 0.007 0.018 0.020 57 0.008 0.022 0.030 33

530 0.006 0.018 0.022 19 0.012 0.022 0.022 7

550-569 0.007 0.018 0.027 109 0.007 0.014 0.026 73

600-690 0.006 0.014 0.016 65 0.005 0.011 0.011 5

370 Elec 0.008 0.024 0.027 28 0.008 0.027 0.031 22

370 Mill 0.012 0.030 0.048 35 0.012 0.048 0.060 22

370 Misc 0.010 0.024 0.057 88 0.008 0.025 0.039 63

370 Mtns 0.011 0.025 0.069 69 0.012 0.029 0.061 48

370 Pipe 0.012 0.020 0.024 37 0.011 0.027 0.042 26

Other 0.010 0.020 0.075 85 0.009 0.033 0.055 43
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 31 of 60

Table 5-17. Uranium urine data summary by cost center for 1966. 

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.015 0.030 0.039 19 0.025 0.053 0.053 12

120 0.009 0.030 0.052 130 0.015 0.048 0.100 83

150 0.006 0.018 0.019 25 0.011 0.054 0.054 13

180 0.012 0.036 0.036 20 0.025 0.048 0.048 9

200 0.006 0.027 0.027 17 0.011 0.024 0.024 9

290 0.006 0.024 0.024 22 0.006 0.036 0.036 12

310-321 0.009 0.021 0.022 20 0.016 0.459 0.459 11

350 0.002 0.008 0.008 13 0.002 0.009 0.009 15

370-379 0.006 0.027 0.046 107 0.008 0.050 0.052 61

380-390 0.007 0.021 0.025 26 0.008 0.029 0.029 13

400-460 0.006 0.025 0.035 53 0.008 0.032 0.045 24

500-510 0.002 0.008 0.011 26 0.004 0.008 0.010 35

520 0.006 0.025 0.018 34 0.004 0.024 0.024 25

530 0.005 0.019 0.021 24 0.004 0.022 0.022 12

550-569 0.005 0.012 0.020 72 0.004 0.012 0.028 38

610-690 0.006 0.011 0.012 18 0.004 0.008 0.008 4

370 Elec 0.012 0.027 0.027 13 0.009 0.036 0.036 10

370 Mill 0.008 0.024 0.027 32 0.008 0.034 0.071 16

370 Misc 0.008 0.021 0.028 53 0.010 0.033 0.036 29

370 Mtns 0.009 0.022 0.066 46 0.009 0.050 0.055 25

370 Pipe 0.006 0.015 0.024 22 0.015 0.088 0.088 13

Other 0.006 0.019 0.027 58 0.004 0.025 0.213 37
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 32 of 60

end of each day they were used. These action levels triggered fecal sampling, urine sampling, and in 

vivo measurements, as appropriate.

There are no Technical Basis Manuals available to document the bioassay monitoring program during 

the early part of the remediation period (1985 to 1990). 

If no individual bioassay data and no applicable co-worker bioassay data are available for a claimant 

for the 1985 to 1990 time period, the environmental data for 1985 to 1990 described in the Weldon 

Spring Plant – Occupational Environmental Dose (ORAUT 2013b) can be used to estimate intake. 

Alternatively, individual or co-worker bioassay data for the subsequent year (1991) may be used to 

estimate worker intakes. 

Routine uranium urine sampling also occurred monthly for at-risk workers. Uranium MDAs were 

reported as 1 μg/L in 1991 (DOE 1991, p. 29) using laser fluorimetry and as 0.1 μg/L in 1994 to 1998 

(DOE 1994, p. 31; 1997, p. 22; 1998a, p. 21) and 0.0524 μg/L in September 1998 to 2001 (DOE 

1998b, p. 20; 1998c, p. 20; 2000a, p. 20; 2001b, p. 22) using kinetic phosphorescence analysis. 

Uranium results of 0.2 μg/L or greater were considered positive for occupational uranium intakes in 

1997 (DOE 1997, p. 23), and results of 0.3 μg/L or greater (DOE 1998c, p. 21; 2000a, p.21, 2001b, p. 

23) were considered positive for occupational uranium intakes from 1998 through 2001.

5.4 IN VIVO MEASUREMENTS

5.4.1

There is no indication that WSP had an in vivo measurement program or performed any in vivo

measurements for uranium, but there is an indication that in vivo measurements were performed on 

some WSP workers for thorium (natural thorium or 232Th) in 1966:

Operational Period (1957 to 1966)

From July 11 through July 27, 1966, Y-12 personnel visited the Weldon Spring plant 

and set up the portable Whole Body Counter for in vivo thorium counting to quantify 

body burden deposition and the risk inherent with using the current Atomic Energy 

occupational air concentration limits (3.7E-11 μCi/ml). During this period of testing, 

200 measurements were made in the monitoring of 148 persons. The determination of 

workers to be monitored was done on a strictly voluntary basis. A good cross 

representation of workers volunteered. The interpretation of the result is as follows:

1. Workers who showed net counts less than 60 counts per 20 minutes had less 

than detectable amounts of thorium in their lungs and were therefore given a 

‘negative’ result.

2. Workers showing net counts in excess of 60 counts per 20 minutes but less 

than 204 were interpreted as a ‘trace’ of thorium.

3. Net counts in excess of 204 counts for 20 minutes were considered as ‘positive’ 

evidence of thorium lung burdens. A person who showed 204 counts for 20 

minutes was considered to have at least one lung burden.

The overall results showed workers involved in areas 101, 103, 301, 403, 

Maintenance, and Health and Safety, which were principal exposure positions, had a 

more frequent occurrence of ‘trace’ detections. No workers monitored showed a 

‘positive’ designation. (Ingle 1991, p. 6)
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 33 of 60

The reports of these measurements observed in the worker files are titled “Thallium 208 in Vivo 

Results.” This indicates that 208Tl was measured as a marker for thorium. Thallium-208 is in 

approximate equilibrium (with a branching ratio of 0.337) with 228Th, which might not have been in 

equilibrium with the thorium series parent 232Th following chemical purification of the natural thorium 

feed materials. The quantification of thorium depositions from these in vivo 208Tl measurements is, 

therefore, uncertain without knowledge of the degree of equilibrium of the thallium with the 232Th 

parent. The record only gives a qualitative indication, background or trace, of the detection of 208Tl as 

a marker for thorium.

5.4.2

No personnel in vivo monitoring appears to have been conducted during this period.

Environmental Monitoring Period (1975 to 1984)

5.4.3

An in vivo measurement program was included in the design of the WSSRAP internal dosimetry 

program to evaluate intakes of 238U and 232Th. Because the number of radiological workers exposed 

to airborne radioactivity at the WSP site was expected to be small, WSSRAP could not justify the 

expense of having its own in vivo measurement system. Instead, the program was initially based on 

detection sensitivities provided by the Helgeson Scientific Services mobile counting laboratory. The 

lower limit of detection cited for that system was 74 Bq (2 nCi) for natural uranium and 37 Bq (1 nCi) 

for 232Th in the lung (DOE 1991, p. 28). 

Remediation Period (1985 to 2002)

Later, the program was based on detection sensitivities of the in vivo measurement system at the 

Fernald Environmental Management Project. The sensitivities (of unstated pedigree) cited for that 

system were 2.0 nCi for 238U and 1.2 nCi for 228Ac, assumed to be in secular equilibrium with 232Th 

(DOE 2001b, p. 20). Revision 7 (DOE 2001b, p. 20) of the WSSRAP Technical Basis Manual states:

This assumption [of secular equilibrium] will not necessarily be true in an actual worker 

intake.

and

It is important to note that these ‘typical’ detection limits are highly dependent upon the 

individual worker’s physical features such as height and chest size. The Weldon 

Spring site has sent individuals to the Fernald site for lung counts, and detection limits 

were 2.5 times higher than the typical values due to the individual’s physical features.

In vivo lung measurements could have been performed as a special bioassay measurement following 

a suspected or actual intake. Revision 7 (DOE 2001b, p. 30) states that such measurements were 

normally reserved for “those incidents where the intake was suspected to exceed 500 mrem 

CEDE ....” Revision 7 (DOE 2001b, p. 30) also states that “requiring lung counts for affected workers” 

was one of the actions “never used.”

5.5 AIR CONCENTRATION AND DUST MEASUREMENTS

5.5.1

Two types of air sampling were conducted by MCW at the WSP. A daily weighted-average (DWA) 

concentration index was calculated based on a combination of breathing-zone air (BZA) samples and 

general air samples (MCW 1965, p. 16). 

Operational Period (1957 to 1966)
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Semi-fixed location general air samplers were located in each process building. The results of the 

samples were used to assess changes in plant air concentrations due to equipment malfunction or 

incorrect operation. However, these data were considered to have no direct value in assessing 

individual intakes or doses.

Samples were analyzed for gross alpha activity only. The action level for uranium dust exposure was 

700 alpha dpm/d (MCW 1965, p. 16). As noted above, this was a time-weighted average exposure. 

This was derived from the guideline concentration of 70 alpha dpm/m3 and the assumption that the 

average breathing rate for workers is 10 m3

/d for an 8-hour workday.

BZA samples were collected at WSP during specific times, and the data was applied to specific jobs. 

5.5.1.1 Maximum Acceptable Concentrations

The time-weighted average maximum allowable concentration (MAC) for uranium was 70 alpha 

dpm/m3 (3.2 × 10-11 μCi/cm3

) or 50 μg/m3 (74 alpha dpm/m3). The 70 alpha dpm/m3 MAC was 

apparently derived by rounding down the radioactivity equivalent of the mass concentration MAC.

The AEC standards for radiation protection applicable in 1965 listed the allowable concentrations for 

uranium in air for occupational exposure as 7 × 10-11 μCi/cm3 for soluble forms and 6 × 10-11 μCi/cm3

for insoluble forms. The WSP guideline of 70 alpha dpm/m3 (3.2 × 10-11 μCi/cm3

) was more restrictive 

than the AEC standard.

There are no indications in the radiation protection program documents that the WSP MAC was 

adjusted in areas where thorium was processed. The values for soluble and insoluble natural 

uranium from the 1959 ICRP II convention for maximum permissible concentrations (MPCs) were 

7 × 10-11 μCi/cm3 and 6 × 10-11 μCi/cm3

, respectively (HPS 1960, Volume 3, p. 114). As noted above, 

these values were adopted by the AEC. The ICRP II MPCs for soluble and insoluble natural thorium 

were much more restrictive at 2 × 10-12 μCi/cm3 and 4 × 10-12 μCi/cm3

, respectively (HPS 1960, 

Volume 3, p. 112). However, a provisional level of 3 × 10-11 μCi/cm3

 was recommended. This is the 

value that was adopted in Annex 1 of AEC Manual Chapter 0524, “Standards for Radiation Protection” 

(AEC 1963). This provisional level for the MPC for natural thorium is similar to the time-weighted 

average MAC for uranium used at WSP.

Based on the information available, it is reasonable to assume that the gross alpha MAC of 70 dpm 

alpha/m3 was applied across the WSP.

5.5.1.2 Special Curie for Uranium

The air-sampling data were reported in alpha particle disintegrations per minute rather than curies 

because the former was an unambiguous designation. Until 1973 when the National Council on 

Radiation Protection and Measurements (NCRP) discouraged its use, the special curie was 

commonly employed for natural uranium. The special curie was defined in 1959 as follows (NCRP 

1973, p. 3):

Special curie = 3.7 × 1010 d/s 238U + 3.7 × 1010 d/s 234U + 9 × 108 d/s 235U 

= 7.49 × 1010 d/s

The definition was altered slightly in 1963 to use 1.7 × 109 d/s for 235U. It is important to understand 

the use of the term special curie when the data are reviewed. In addition, MCW used the ratio 

between the measured DWA and the guideline or standard as an index of exposure. 
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5.5.1.3 Dust Exposure Calculation

The total dust exposure worksheets used to record data for the MCW St. Louis site have a provision 

for entering the dust concentration, but the applicable worksheet column does not list the units. The 

values are most likely the index (i.e., the DWA divided by the guideline), and there is no indication of 

the source of the data. The intake calculated from urine bioassay and the intake calculated from 

measured dust concentrations were averaged with equal weight given to each source. There is no 

evidence to show when use of these forms was discontinued and no indication that they were used for 

WSP employees. 

The Annual Personnel Internal-External Radiation Exposure Report form, apparently in use by 1959 

for WSP, includes a section for average dust concentration in disintegrations per minute per cubic 

meter by calendar quarter. None of the exposure reports reviewed had any data in that section. This 

indicates that the dust concentration was not routinely recorded. However, because the average dust 

concentration, when recorded, is in units of disintegrations per minute per cubic meter, the average 

daily intake can be calculated by assuming a breathing rate of 10 m3

/d for typical light work for an 8-

hour workday or by using a job-specific value.

No specific in-plant air monitoring analysis sheets were found, but samples of the forms for reporting 

perimeter air sample data were available. These forms could also have been used for in-plant 

measurements. The forms include information on the sampling rate, time, and the gross alpha 

activity. The samples were analyzed for alpha and beta activity. The results of the dust studies are 

summarized in the following section.

5.5.1.4 Dust Studies

A study of specific areas and jobs in Building 301 was conducted in 1961 (MCW 1961). Timeweighted average concentrations were calculated based on the number of work hours at various 

positions. The measured concentrations were reported in microcuries per cubic centimeter using the 

special curie unit and in micrograms per cubic meter. The data were used primarily as a basis for 

recommending actions to reduce concentrations. There is no indication that the data were used to 

assess intake.

An undated document titled Summaries of Dust Concentrations at Production Jobs (MCW ca. 1966, 

pp. 5-46) provides data on time-weighted average dust concentrations for various work areas for the 

period from 1958 to 1966. The data were summarized for historic use in evaluating worker dust 

exposures. The dust samples were collected on open-face Whatman No. 41 or membrane filters with 

areas ranging from 3 to 5 cm2

. The membrane filters had a pore size of 0.8 μm. The flow rate ranged 

from 10 to 20 L/min. The report notes that the samples were taken either as fixed general air samples 

or as “hand held breathing zone type.” 

The filters were analyzed for gross alpha by scintillation counters. Samples from uranium areas were 

counted after a minimum 24-hour delay to allow for the decay of radon progeny. Samples from 

thorium areas were counted after a minimum delay of 100 hours. 

For uranium work areas, a work group index was calculated based on an average time-weighted 

exposure for various job titles. The jobs were rotated among the workers in each work group so the 

work group index could provide a reasonable time-weighted average concentration in milligrams 

uranium per cubic meter. The activity on the filter was converted to mass concentration using a factor 

of 1.5 alpha dpm/μg U (0.68 pCi/μg U). Table 5-18 summarizes the work group concentrations.
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 36 of 60

Table 5-18. Work group time-weighted average exposure for uranium operations. 

Work group title

Daily weighted average concentration (μg U/m3)

1958 1959 1960 1961 1962 1963 1964 1965 1966

Sampling and repackaging cost 

center 110

65 55 45 55 32 31 20 60

Digest-raffinate, cost center 120 17 7 11 8 7

Extraction, cost center 120 4 3 5 2 2

Denitration (pot room), CC 120 117 70 22 45 16

Green salt, cost center 150 40 37 37 53 53a 16a 16 22

Dingot reduction, cost center 180 115 65 150a 150a 78 50

Metal – other than reduction

cost centers 200, 210, 220, 280

44 52 63 40b 55e

Core area, cost center 250 20

Pilot/scrap plants, cost center 290 50c 50c 50c 50c 50c

Special projectsd

a. Estimate by extension to other years or averaged over other years.

b. 50% of effort on uranium jobs; also worked with thorium.

c. Overall average - not weighted.

d. No averages given.

e. 40% of effort on uranium jobs; also worked with thorium

Thorium (natural thorium, 232Th) airborne dust concentrations were routinely measured and evaluated 

throughout the operating period. These data, in the form of DWA concentrations, were summarized 

for the thorium processing period 1963 through 1966 (MCW ca. 1966, pp 15-47) and are reproduced 

in Attachment A. Dust concentrations were also measured for individual operations for specific years. 

These operations were intermittent and did not represent the total work year. The time-weighted 

average concentrations as a fraction of the MACs were calculated. The MAC for thorium at the time 

was 70 dpm/m3

. Table 5-19 summarizes these calculations. 

Table 5-19. Time-weighted average 232Th dust concentrations. 

Operation Bldg. Year

Fraction 

of MAC Total effort

Repackaging ThO2 feeds 101 1966 0.1 10 shifts

Hopper packaging 101 1966 0.2 10 shifts

Conveyors at repackaging 101 1966 0.1 10 shifts

Oven drying ThO2 solution 101 1965–66 <0.1 2 person d

Oven drying ThO2 solution 101 1965 0.2 Intermittent 

Hopper feed and digestion 103 1966 0.1 No information

Raffinate 103 1966 <0.1 No information

Misc. operations 103 1966 <0.1 Non-routine job

ThO2 repackaging 103 1964 3.7 Short term–Airline masks used in hood

Extraction 105 1966 <0.1 No information

Crystal denitration 103 1963–64 1.25 Four months–Airline masks used

Crystal denitration 103 1964–65 0.9 Airline masks prescribed

TNTa liquor denitration 103 1965 0.7 Eight months–Airline masks prescribed

Solution drying–vacuum 

unload

103 1965–66 0.6 10 months–“Comfo” respirators prescribed

Pot denitration 103 1965 2 One month–Airline mask prescribed for some 

tasks

High firing at recast furnace 301 1963–64 0.3 Eight months–Airline masks for some tasks

High firing in billet heaters 301 1964–66 0.15 No information

TNTa repackaging 301 1964–65 0.04 No information

ThO2 repackaging 301 1965 30 One month–Airline masks and cover clothing 

used

ThO2 repackaging 301 1965 0.8 Airline masks used for some tasks

Kiln calcining sump cake 301 1966 0.3 “Comfo” respirators used for some tasks
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Operation Bldg. Year

Fraction 

of MAC Total effort

TNT repackaging 403 1963–64 0.2 Eight months

ThO2 repackaging 403 1963–64 3.3 Eight months–Airline masks used on some 

operations

ThO2 repackaging 403 1964–65 0.3 13 months

Fluid bed denitration 403 1965 0.6 Three months

Fluid bed denitration 403 1965–66 0.4 No information

a. TNT = thorium nitrate tetrahydrate. 

5.5.2

The site was transferred to the Army in 1967 for use in herbicide production. Some areas of Buildings 

101 and 103 were partially cleaned up in preparation for the new mission. However, the project was 

terminated in early 1969, and no further cleanup or construction occurred. No dust measurement 

data were found for that period.

Transfer to the U.S. Army (1967 to 1974)

5.5.3

With the exception of the WSQ and WSRP, the remainder of the site including the WSCP was turned 

over to the U.S. Army Corp of Engineers in 1969. No major activities took place from 1969 to 1985. 

There were no AEC contractors on site from 1969 to 1975, after which the AEC contracted with 

National Lead of Ohio (NLO) to perform environmental monitoring around the WSQ and WSRP. 

Bechtel National took over management of the WSRP and WSQ in 1981, and the WSCP site was 

transferred from the Army to DOE in 1985.

Environmental Monitoring Period (1975 to 1984)

No monitoring data were available in relation to personnel dosimetry programs for this period. 

5.5.4

DOE contracted with MK-Ferguson to manage the remediation project in 1986. The WSRP and the 

WSQ were put on the U.S. Environmental Protection Agency’s National Priorities List in 1987 and 

1989, respectively. Site cleanup began in earnest in 1990 as described in ORAUT-TKBS-0028-2,

Weldon Spring Plant – Site Description (ORAUT 2013c). 

Remediation Period (1985 to 2002)

No specific air-sampling records were available for this period. However, the internal dosimetry 

programs were well documented in the WSSRAP Technical Basis Manuals (DOE 1991, 1994, 1997, 

1998a,b,c, 2000a, 2001). The air-sampling program consisted of area samples and BZA samples. 

Area sample data were not used routinely for dose assessment, but were used (only) if no BZA 

sample data were available and bioassay measurements could not confirm potential intakes.

At first, area samples were taken with both high-volume (1,100 L/min) and low-volume (40 L/min) 

samplers. The filters were analyzed by an offsite laboratory for isotopic content. The MDA on the 

filters was 1 pCi for each nuclide. The calculated minimum detectable concentrations (MDCs) for a 6-

hour sampling period were 2.7 × 10-15 μCi/mL and 6.9 × 10-14 μCi/mL for the high- and low-volume 

samplers, respectively. The gross alpha MDC for the high-volume sampler was 1.5 × 10-14 μCi/mL. 

By 2001, area air sampling was normally performed with low-volume samplers. The samples were 

analyzed for long-lived gross alpha activity. The results were compared to an area-specific effective 

DAC developed based on isotopic concentrations in the dust. The MDC for the system as it was 

routinely used (sampling and counting times, etc.) was normally less than 1.2 × 10-14 μCi/mL. 

BZA samples were routinely used with bioassay results to assess worker intakes. The BZA samples 

had higher MDCs, generally below 2.1 × 10-13 μCi/mL (1991) and 1.9 × 10-13 μCi/mL (2001). To 
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assess worker intakes, the gross alpha concentrations were apportioned to specific radionuclides 

based on isotopic analysis of area samples. 

Area air sampling was required in all areas where a worker could have an annual intake greater than 

2% of the annual limit of intake. Based on the various revisions to the WSSRAP Internal Dosimetry 

Technical Basis Manual, the requirements for breathing-zone samples varied slightly over time. At 

first, every worker spending at least 1 hour a day in the WSQ or WSRP airborne radioactivity areas 

was required to wear a BZA. Later requirements were somewhat less strict. Revision 1 of the 

Technical Basis Manual (DOE 1994, pp. 45, 49) required BZA samples for one in five individuals 

spending more than 1 hour a day in areas with concentrations greater than 10% of the DAC and one 

in three individuals in the WSQ or WSRP areas. As of Revision 2 (DOE 1997, p. 32), one in four 

workers in areas with concentrations greater than 2% of the DAC were required to wear BZA 

samplers.

The internal dosimetry program required that routine monitoring of environmental 222Rn (radon) and 

220Rn (thoron) and their decay products be instituted when an individual was likely to receive an 

annual intake of 10% or more of the annual limit of intake. According to Revision 7 of the Technical 

Basis Manual (DOE 2001b, p. 32), that threshold was never exceeded. Therefore, routine monitoring 

data for radon gas or short-lived decay products are not available. Environmental radon 

measurements that were taken periodically are provided in the Weldon Spring Plant – Occupational 

Environmental Dose (ORAUT 2013b). These measurements provide a basis for estimating worker 

exposure to short-lived radon decay products. 

5.6 ASSESSMENT OF INTAKES 

5.6.1

5.6.1.1 Uranium and Related Intakes

Operational Period (1957 to 1966)

The worker’s urine bioassay data are the primary data available to the DR to quantify the uranium 

intake for the worker who is the subject of the claim. These data can be supplemented by work group 

monitoring data because essentially continuous bioassay monitoring for a worker was simulated by at 

least one worker in the group being sampled each week with Monday – Friday – Monday sampling for 

potentially exposed workers (Section 5.3.1.1.1). The work group data have been reconstructed from 

urine data for all WSP workers by cost center code. Tables 5-8 to 5-17 (Section 5.3.1.1.6) list the 

median, the 95th percentile, and the maximum values of the uranium urine data per year for Monday 

samples and Friday samples by cost center and calendar year. The worker’s urine data reports 

provide the cost center (Section 5.3.1.1.5). The statistical data associated with Friday urine sampling 

should be used in dose reconstructions to avoid underestimating uranium intakes.

If specific information is not available in the worker’s file, the DR should consider the following default 

uranium source terms:

• Natural uranium, before 1961

• Natural uranium, recycled, 1961 to 1962

• Enriched (1%) uranium, recycled, 1963 to 1966

Because the feed uranium and uranium during processing were purified to some degree, it is 

reasonable to assume that the contributions of the long-lived uranium progeny, i.e., 230Th and 226Ra, 

were small in most areas of the WSP. Measurements of the activity concentrations in Raffinate Pits 1, 

2, and 3 can be used to determine the relationship between 230Th and other impurities during the initial 

uranium processing (feed preparation and sampling) in Building 101, during the transfer of ore 

concentrate from Building 101 to Building 103 for nitric acid digestion, and transfer of the uranyl nitrate 
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slurry to the aqueous feed tanks in Building 105 before any separations occurred. The shorter-lived 

decay products for which no raffinate measurements were made (e.g., 210Pb and 210Po) can be 

assumed to be present at the same activity as their 226Ra parent in the mill concentrate feeds. The 

other uranium streams (e.g., recycled uranium) had been previously processed and contained 

essentially no thorium. Table 5-20 gives the results of a statistical analysis of raffinate pit

measurements. The data were taken from the 1989 Waste Assessment Radiological Characterization 

of the Weldon Spring Site Raffinate Pits (MK-Ferguson 1989, pp. 72-75).

Table 5-20. Activity ratios of uranium decay products and other impurities to 

230Th activity in Weldon Spring raffinate pits. 

Isotope

Pits 1 Pit 2 Pit 3 Pit 4

Mean 95th % Mean 95th % Mean 95th % Mean 95th %

Ra-226 0.090 0.367 0.021 0.032 0.021 0.034 0.197 0.585

Ra-228 0.004 0.013 0.005 0.006 0.004 0.011 0.251 1.003

Th-228 0.004 0.011 0.004 0.006 0.006 0.010 0.326 1.508

Th-232a 0.006 0.016 0.006 0.009 0.009 0.016 0.326 1.545

a. Measurements of thorium-232 in Raffinate Pits 1, 2, and 3 were not used due to alpha 

peak interference from thorium-230 (MK-Ferguson 1989, p. 75). Thorium-232 was 

calculated assuming that thorium-228 was in 65% equilibrium, except for Pit 4 where 

reliable measurements were available.

The mean values in Table 5-20 represent the amount of each isotope that as a fraction of thorium-230 

an employee would have been exposed to over the long term. It is unlikely that any employee would 

have been continuously exposed to fractions greater than the 95th percentile; therefore, the 95th

percentile of the ratios to thorium-230 in Table 5-20 can be used to bound the intakes from the other 

potentially important radionuclides. The data for Pit 4 were not used in this analysis since natural 

thorium wastes were deposited there. The activity concentrations for the uranium decay products and 

impurities are determined by multiplying the 95th-percentile ratios in Table 5-20 by their 

corresponding 230Th concentration value for Pits 1, 2, and 3 stated in Table 5-21. 

Table 5-21. Intakes of uranium decay products and other impurities 

(nCi/mg-U) based on raffinate pit measurements. 

Isotope

Intakes based 

on Pit 1

Intakes based 

on Pit 2

Intakes based 

on Pit 3 

Bounding 

intakesa

Pb-210b 0.125 0.007 0.011 0.125

Po-210b 0.125 0.007 0.011 0.125

Ra-226 0.125 0.007 0.011 0.125

Ra-228 0.005 0.002 0.002 0.005

Th-228 0.004 0.001 0.002 0.004

Th-230c 0.340 0.340 0.340 0.340

Th-232 0.006 0.002 0.003 0.006

a. Indicates the higher value of intakes based on Pits 1, 2, and 3.

b. Measurements assumed to be equal to the intake of Ra-226.

c. Th-230 is assumed to be equal to the U-234 activity, or one-half the total natural 

uranium activity. 

Because site-specific particle size data are not available, the default value of 5 μm AMAD should be 

used.

If the absorption type of the uranium to which the worker was exposed cannot be discerned from the 

data in the worker’s file, the DR should use the absorption type that is the most favorable to claimants.
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5.6.1.2 Thorium-232 Intakes

No quantitative in vitro bioassay results have been observed for thorium (Sections 5.3.1.2 and 5.4.1). 

Late in the thorium processing period a limited amount of in vivo counting was performed. The in vivo

data should be considered if available. However, dust studies for thorium operations (Section 5.5.1.4 

and Appendix A) were routine (MCW 1966). The dust studies are presented in terms of daily 

weighted activity (DWA) or daily weighted exposure (DWE). The DWA data provide a clear picture of 

when (on a monthly timeframe) and where thorium processing and packaging operations occurred. 

The original air sampling data used to generate the DWA value for each job have not been located, so 

an alternate approach to characterizing the variability of the exposure in the workplace is necessary. 

Using Atomic Weapon Employer data, Davis and Strom (2008) demonstrated that the point estimate 

DWA value will favorably characterize the exposure potential by assuming that the DWA value is the 

arithmetic mean of a lognormal distribution with a geometric standard deviation (GSD) of 5. Air 

concentrations for each thorium operation are obtained from DWA summary tables (Attachment A). 

Thus, for example, in March 1965 the DWA value assumed to represent the arithmetic mean (AM) of 

exposure to thorium workers performing the task “Oven drying ThO2 sol-pan transfer” was 3 alpha 

dpm per m3

. It is assumed that the reported DWA value is the AM from a lognormal distribution with a 

geometric standard deviation (GSD) of 5. Equation 2 from Strom and Stansbury (2000) defines the 

relationship between GSD and σ

σ = lln(GSD), 

where σ is the standard deviation of the natural logarithm of the observations in a lognormal 

distribution. Equation 13 from Strom and Stansbury (2000) is used to compute the median which is 

equal to the geometric mean (GM) in a lognormal distribution:

GWW = AWW ∗ yy−σ2/2

Table 2-2 in Strom (2007) includes the following formula for μ, the natural logarithm of the GM:

μ = lln(GWW)

Equation 5 in Strom (2007) defines the 95th-percentile value of the lognormal distribution: 

95th ppyyyycyyntppllyy = yy(μ+1.645σ)

Thus, a lognormal distribution with AM = 3 and GSD = 5 has a GM of 0.82 and a 95th percentile value 

of 11.6 alpha dpm/m3

. 

Inhalation intake rates based on the median and 95th percentile DWA 232Th air concentration

(calculated using the equations stated above and the DWA values from Attachment A) are calculated 

assuming a breathing rate of 1.2 m3 per hour and an 8-hour workday. This 8-hour workday is 

normalized to a calendar day based on the number of workdays in a year. The number of workdays 

in 1963 was 43 (21 for November and 22 for December), there were 250 workdays per each year in 

1964 and 1965, and there were 195 workdays in 1966 since thorium operations ceased in September 

1966. The ingestion intake rate is based on guidance in NIOSH (2004) and ORAUT (2012). The 

ingestion intake rates, which are stated in Table 5-22, were derived from the inhalation intake rates in 

Table 5-22 by multiplying the inhalation rate by 0.02. 

The original version of this site profile cautioned that Work Unit data at Weldon Spring may not be a 

reliable indicator of work assignment. This caution was based on conditions that existed at the 

Mallinckrodt Destrehan Street facility, the predecessor to the Mallinckrodt-operated Weldon Spring 

Plant. An interview with a former worker established that Work Unit designation at Weldon Spring 
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Table 5-22. Thorium-232 inhalation and ingestion intake rates (Bq per year). 

Plant name

and building number

Work group title

and cost center

Inhalation (Bq/year) Ingestion (Bq/year)

1963 1964 1965 1966 1963 1964 1965 1966

Sampling - 101 Sampling - 0402 0 0 125 304 0 0 2 6

Other occupants of Sampling Plant 0 0 9 22 0 0 0 0

Refinery - 103 Digest-Raffinate - 0501 196 4810 5100 2320 4 96 102 46

Other occupants of Refinery 14 341 360 165 0 7 7 3

Extraction - 105 Extraction 0 52 0 0 0 1 0 0

Other occupants of Extraction Plant 0 4 0 0 0 0 0 0

Metals - 301 Metals - other than Reduction 0702 47 1010 21550 1595 1 20 431 32

Other occupants of Metals Plant 3 72 1525 113 0 1 31 2

Scrap Plant - 403 Scrap Plant - 0601 645 14700 2535 0 13 294 51 0

Other occupants of Scrap Plant 46 1040 180 0 1 21 4 0

Undetermined location Any worker not otherwise specified above 

whose work location is unknown 46 1040 1525 165 1 21 31 3 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 42 of 60

was reliable (ORAUT 2009). Consequently, the Work Unit (also called Work Group or Job Code) 

designation can be used as an indicator of potential thorium exposure. 

The 232Th inhalation intake rates were calculated on a month-by-month basis for the work tasks 

associated with each location where thorium was processed or handled. The maximum monthly 

intake values were determined by facility, which were then summed for each year of thorium 

operations and summarized in Table 5-22. When more than one task occurred in a location during a 

month, the task yielding the highest intake rate was used in the calculation. Operators in Work 

Groups identified as being directly involved in the process are assigned intakes based on the 95thpercentile value of the lognormal DWA distribution (the intake value in the top line for each group in 

Table 5-22) for the most highly exposed task in each plant. Other occupants of the same plant are 

assigned intakes based on the 50th-percentile value, which is the intake value in the bottom line for 

each group in Table 5-22. For workers in job categories, for example maintenance or janitorial 

services, who may have been exposed but whose location cannot be reliably established, an intake 

rate based on the 50th-percentile inhalation value for the plant yielding the highest intake value during 

that year is recommended. For example, the highest median value (50th percentile) for 1963 is 46 Bq 

per year from the Scrap Plant. Workers with undetermined work locations would be assigned an 

inhalation intake of 46 Bq per year 232Th and an ingestion intake of 1 Bq per year 232Th for 1963. 

Thorium-228 should be assumed to be in equilibrium with 232Th and thus the activities of 228Th are 

equal to the 232Th activities stated in Table 5-22. Radium-228, which is a beta emitter and would not 

have been detected on the alpha air sample, is added at a ratio of 1-to-2 232Th-to228Ra (NIOSH 

2010a, p. 57). For operational periods before or after the time frame of Table 5-22, no occupational 

internal dose from 232Th is recommended. It is appropriate to recognize that the DWA data often are 

annotated with statements indicating that airline or half-mask respiratory protection equipment was 

used or prescribed in some of the thorium operations. 

For purposes of calculating the 232Th intake rates, the use of respiratory protection is ignored because 

the quality of the respiratory protection equipment and diligence in the use of the equipment cannot be 

established. Nevertheless, the routine use of respiratory protection equipment during thorium 

operations suggests that the thorium intake rates are quite likely to be biased in favor of the claimant.

For workers not in these buildings or professions during this period, thorium intakes should be 

assessed as environmental intakes in accordance with Weldon Spring Plant – Occupational 

Environmental Dose (ORAUT 2013b) unless there is information in the worker’s file that indicates 

involvement in thorium operations.

5.6.1.3 Intakes of Other Radionuclides

The annual median environmental radon concentrations for the WSCP, WSRPs, and WSQ are 

provided in Table 4-1 of ORAUT (2013b). The maximum site-wide annual median exposures to radon 

decay products, by year, are provided in Table 4-2 of that same document. These estimates are 

based on environmental radon gas releases. The calculation in ORAUT (2013b) was based on the 

assumption that all radon gas generated by decay of Ra-226 was released during processing. There 

are no personal dosimetry data or air concentration measurements available for radon in the WSCP 

during the operational period.

The Weldon Spring Historic Dose Estimate (Meshkov et al. 1986, p. 47) states that the prime source 

of radon emissions was the acid recovery plant. The 222Rn activity released annually was estimated 

to be 34 Ci per year (Meshkov et al. 1986, pp. 47–48), and this was assumed to be released from 

Building 103 ventilation stacks. However, the radon released to the environment and present in the 

area would have been drawn back into the building through the ventilation systems. Under those 

circumstances, the radon concentration inside the building would have been approximately equal to 

the radon concentration outdoors. The 222Rn equilibrium concentration is calculated by dividing the 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 43 of 60

radon influx (222Rn activity/pCi/hour) by the product of the number of air changes per hour (ach) and 

the volume of the ventilated area. Using the Building 103 volume of 2.6 × 104

 m3 (2.6 × 107 L) and 1 

ach, the 222Rn equilibrium concentration is calculated to be 150 pCi/L. A working level (WL) is defined 

as 100 pCi/L 222Rn in full equilibrium with its short-lived alpha-emitting progeny. However, because 

full equilibrium is never achieved in occupied spaces, a favorable to claimant indoor equilibrium factor 

of 0.7 is assumed (NIOSH 2012). Using the number of work hours in a month of 170 in defining the 

WLM and a 2,000-hour working year, the estimated annual exposure is calculated to be 12.4 WLM for 

100% occupancy for 2,000 hours per year as shown in this equation:

WWWWWW

yyyyyyyy = 150 ppppppll ×

ll ∙ WWWW

100 pppppp

×

170 h ×

2000 h

yyyyyyyy

× 0.7 = 12.4

As described in Section 5.2.3, there are several aspects of the release and buildup of the daughter 

products that mitigate the exposure to thoron in process and storage configurations.

From November 1963 through September 1966, natural thorium was processed on an intermittent 

batch basis in the refinery and oxide production/firing systems at the Weldon Spring Plant; therefore, 

NIOSH’s evaluation of operational exposures to thoron at Weldon Spring is confined to those years. 

During 1966 (the maximum production year), the Weldon Spring Plant processed approximately 

100,000 kg of thorium per month for 6 months, or up to 5,000 kg of 232Th per day. Typical thorium 

processing at other facilities averaged <1,000 kg of thorium per day. NIOSH recognizes that the 

thorium processing was not continuous, but rather was operated more on a batch basis. Given the 

specific activity of thoron in thorium feed materials, and assuming the feed materials were received 

with at least a 1-year delay since processing, it is possible to calculate the amount of thoron in 

process per day (approximately 0.35 Ci thoron [2]) during the period of the maximum production rate. 

By assuming a conservative equilibrium factor for a plant configuration with large buildings and 

engineered ventilation (0.02) [3], it is possible to determine the concentration of thoron and its 

daughters to achieve 1 WL. The release fraction of the thoron from thorium can be calculated by 

comparing the particulate thorium in the working environment of the process equipment to the 

inventory amounts in process. The thoron releases are expected to be less than the particulate 

materials, since the gaseous state will be more easily captured by the ventilation systems.

Potential exposure to thoron is mitigated by several factors, as discussed in Section 5.2.1.6, including 

the very short half-life of thoron and the resulting limitations on the diffusion distance within the 

material matrix, and the effective shielding provided by the thorium processing apparatus. Thus, 

thoron is not a potential source of exposure to be broadly accounted for across the evaluated class of 

Weldon Spring workers. Using the distribution of values in Table 5-19, the 50th percentiles and 95th 

percentiles of DWA were estimated as 21 and 250 alpha dpm/m3

, respectively, over the entire thorium 

processing period. Assuming that this activity consisted solely of 232Th, 228Th, and 224Ra, release 

fractions for particulates of approximately 7.6 × 10-7 and 9.0 × 10-6 were calculated for the 50th and 

95th percentiles, respectively. As discussed previously, these would be conservative estimates for 

the thoron gas, which is more easily captured by the ventilation. A volume of 100,000 ft3 (100 by 100 

by 10 ft) was assumed for the processing area [4]. Using the equilibrium factor of 0.02, estimates of 

2.9 × 10-3 and 3.5 × 10-2 WLM/yr for the thoron daughters were obtained for the 50th and 95th 

percentiles, respectively. These are conservative since they are based on the peak production rate of 

5,000 kg 232Th per day. DRs can apply these values to the 3-year period of thorium processing, 

depending on whether it appears that a worker was occasionally or routinely exposed to thorium 

operations.

5.6.2

All intakes in this period should be assessed as environmental intakes in accordance with ORAUT 

(2013b). 

Environmental Monitoring Period (1975 to 1984)
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 44 of 60

5.6.3

Because WSSRAP conducted an extensive, state-of-the-art internal dosimetry program during 

remediation that was designed to detect and evaluate intakes of 100 mrem CEDE or more for “all 

occupational radionuclide intakes in a year (other than radon, thoron, and their progeny)” (DOE 

2001b, p. 36), it is reasonable to consider that the worker’s file will contain a detailed report of the 

pertinent data used for the assessment. It is also reasonable to expect that the data will be selfexplanatory and can be used by the DR as found. If needed, a detailed description of the internal 

dosimetry program is available in the series of WSSRAP Technical Basis Manuals (DOE 1991, 1994, 

1997, 1998a,b,c, 2000a, 2001). 

Remediation Period (1985 to 2002)

In addition to assessing specific intakes, the DR should assess the environmental intakes in 

accordance with ORAUT (2013b). 

5.7 SUMMARY OF INSTRUCTIONS TO DOSE RECONSTRUCTORS

Table 5-23 contains a summary of the instructions to DRs.

Table 5-23. Instructions to DRs. 

Intake Data/Information to be used

Operational Period (1957–1966)

Uranium

(U-234, U-235, U-238)

Primary: individual worker’s uranium bioassay data.

Supplemental: workgroup data (Tables 5-8 through 5-17). 

Default isotopic assumptions:

• Natural uranium, before 1961

• Natural uranium, recycled, 1961 to 1962

• Enriched (1%) uranium, recycled, 1963 to 1966, 

specific activity of 0.783 pCi/μg

• Use Section 5.2.4 for recycled uranium 

contaminants

Uranium contaminants and decay products 

(less radon)

Bldg. 101, Bldg. 103, Bldg. 105 prior to separations: 

assigned based on uranium bioassay (Table 5-21). 

Others: NA, assume pure uranium after separations.

Thorium (Th-232 and Th-228) and Ra-228 

(1963–1966)

Use calculated intake rates based on DWA estimates for 

workers involved in thorium processing (Table 5-22). 

Th-228 activity equal to Th-232 activity. Ra-228 activity 

equal to two times Th-232 activity.

Radon 12.4 WLM/yr for 100 percent occupancy for 2000 hours.

Thoron (1963–1966) 2.9E-03 and 3.5E-02 WLM/yr, for 50th and 95th 

percentiles, respectively.

Environmental Monitoring Period (1975–1984)

Any All intakes in this period should be assessed as 

environmental intakes using ORAUT-TKBS-0028-4.

Remediation Period (1985–2002)

Any The worker’s file will contain a detailed report of the 

pertinent data to be used for the assessment.

5.8 ATTRIBUTIONS AND ANNOTATIONS

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

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

These callouts are listed here in the Attributions and Annotations section, with information to identify 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 45 of 60

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

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

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

[1] Rich, Bryce. Oak Ridge Associated Universities (ORAU) Team. Health Physicist. August 

2010.

This statement is based on experience.

[2] Potter, Eugene. ORAU Team. Health Physicist. August 2010.

This statement assumes equilibrium with 228

Th, which is in 65% equilibrium with 232Th.

[3] Rich, Bryce. ORAU Team. Health Physicist. August 2010.

This statement is based on experience.

[4] Potter, Eugene. ORAU Team. Health Physicist. August 2010.

This is a reasonable estimate for the size of the processing area. 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 46 of 60

REFERENCES 

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Dose Assessment, attachment to Letter to D. Utterback (National Institute for Occupational 

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1975, U.S. Energy Research and Development Administration. [SRDB Ref ID: 4613]

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Radiobioassay, HPS N13.30-1996, McLean, VA. [SRDB Ref ID: 23441]

Ingle, J. N., 1991, “Mallinckrodt Chemical Works: Four Plant Study Classifications of Radium, Radon 

and Thorium Exposure,” memorandum, NIPF AU 005101, Oak Ridge Associated Universities, 

Oak Ridge, Tennessee, September 16. [SRDB Ref ID: 12400] 

ICRP (International Commission on Radiological Protection), 1979, Limits for Intakes of Radionuclides 

by Workers, ICRP Publication 30, Part 1, Pergamon Press, Oxford, England. [SRDB Ref ID: 

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Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

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for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1995, Age-Dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, 

Publication 71, Pergamon Press, Oxford, England. [SRDB Ref ID: 22729]

Lesperance, A. M., M. R. Siegel, and M. D. McKinney, 1992, A Review of Occupational Safety and 

Health Issues Relevant to the Environmental Restoration Program: Selected Case Histories 

and Associated Issues, PNL-8184, DE93 003287, Pacific Northwest Laboratory, Richland, 

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Weldon Spring, Missouri. [SRDB Ref ID: 12363] 
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MCW (Mallinckrodt Chemical Works), ca. 1964, Weldon Spring: An Introduction, Mallinckrodt 

Chemical Works, Weldon Spring, Missouri. [SRDB Ref ID: 13179]

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Department, Weldon Spring, Missouri. [SRDB Ref ID: 14998]

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Jobs, Life of Operation — from Year 1957 through 1966, Weldon Spring, Missouri. [SRDB Ref 

ID: 15000]

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Area Uranium Processing Operations, 1942-1967, St. Louis, Missouri. September. [SRDB 

Ref ID: 6538]

Mason, M. G., 1977, handwritten note, May 22. [SRDB Ref ID: 13179, p. 29] 

Meshkov, N., P. Benioff, J. Wang, and Y. Yuan, 1986, Weldon Spring Historical Dose Estimate, 

ANL/EES-TM-308, DE86 014085, Argonne National Laboratory, Energy and Environmental 

Services Division, Argonne, Illinois, July. [SRDB Ref ID: 13117]

MK-Ferguson, 1989, Waste Assessment Radiological Characterization of the Weldon Spring Site 

Raffinate Pits, Rev. 0, DOE/OR/21548-062; MK-Ferguson Company and Jacobs Engineering 

Group; August. [SRDB Ref ID: 12541]

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Natural Uranium and Natural Thorium, NCRP Statement No. 4, Bethesda, Maryland, 

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OCAS-TIB-009, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, April 

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Petition Evaluation Report Petition SEC-00143, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, April 16.

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

Compensation Program Act, DCAS-IG-003, Rev. 0, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, November 5.

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and Worker Health Work Group 86th Meeting. September 19. [SRDB Ref ID: 123610 pp. 

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Assessments,” pp. 13–93, Final Generic Environmental Impact Statement on Uranium Milling, 

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Dose, ORAUT-TKBS-0011-5, Rev. 02, Oak Ridge, Tennessee, August 17.

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GLOSSARY

absorption type

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization). Also called solubility type. See inhalation class. 

activity median aerodynamic diameter (AMAD)

Diameter of a unit density sphere with the same terminal settling velocity in air as that of the 

aerosol particle whose activity is the median for the entire aerosol.

becquerel (Bq)

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals 

37 billion (3.7 × 1010) Bq.

bioassay

Measurement of amount or concentration of radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body. Also called radiobioassay. 

bioassay procedure

Procedure used to determine the kind, quantity, location, and retention of radionuclides in the 

body by direct (in vivo) measurements or by in vitro analysis of material excreted or removed 

from the body.

body burden

Amount of radioactive material in an individual's body at a particular point in time.

chronic exposure

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years).

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

depleted uranium (DU)

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

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, DE)

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. 
Document No. ORAUT-TKBS-0028-5 Revision No. 02 Effective Date: 05/21/2013 Page 51 of 60

dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium (EU)

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7% by mass. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade 

uranium contains greater than 90% 235U.

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.

inhalation class

Former respiratory tract inhalation classification scheme developed by the International 

Council on Radiological Protection for inhaled material according to its rate of clearance from 

the pulmonary region of the lung. Materials were classified as D (days, half-life less than 10 

days), W (weeks, 10 to 100 days), or Y (years, more than 100 days). See absorption type, 

which superseded this concept.

insoluble

Having very low solubility. No material is absolutely insoluble. See absorption type. 

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

internal dose or exposure

Dose received from radioactive material in the body.

internal dose assessment

Estimation of an intake of radioactive material and the consequent radiation dose based on 

bioassay or other measurements in the work environment.

in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

Measurements of radioactive material in the human body utilizing instrumentation that detects 

radiation emitted from the radioactive material in the body.

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.

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. See element. 
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lung solubility type

See absorption type. 

minimum detectable activity or amount (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

minimum detectable concentration (MDC)

Minimum detectable activity (or amount) in units of concentration. See minimum detectable 

activity. 

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

natural uranium (U, U-nat, NU)

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram).

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose.

radionuclide

Radioactive nuclide. See 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.

routine monitoring

Monitoring carried out at regular intervals during normal operations.

working level (WL)

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 
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radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 130,000 MeV of alpha energy; 1 WL equals 

2.083 × 10-5 joules per cubic meter.

working level month (WLM)

Unit of exposure to radon progeny defined as exposure for 1 working month (170 working 

hours) to a potential alpha energy concentration from of 1 WL; 1 WLM equals 1 WL times 

170 hours, which is 0.00354 joule-hours per cubic meter.
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Table A-1. Dust concentrations for thorium production jobs. 

Plant 

name & 

no. Materials

Work 

group title 

& cost 

centera Job titleb Periodc

Workers/

day

Daily weighted average 

(DWA) concentrations

(alpha dpm/m3)

1963 1964 1965 1966 Notes

Sampling

101

Extracted 

thoriumnatural

Sampling

0402

Oven drying ThO2 solpan transferMar 1965 2 - - 3 - Test date: Mar 30, 1965. Five 

ovens operated on a 24-hour 

cycle at 400 lbs/oven. 

Experimental, intermittent 

operation.

Oven drying ThO2 solvacuum unloadAug 1965 2 - - 16 - Test date: Aug 9, 1965. 

Experimental, intermittent 

operation.

Repackaging ThO2 feeds 

for digestion-drum 

dumping

Apr–Aug 

1966

1 - - - 9 Total project operating time 

estimated at 1 week or 10 

shifts intermittent operation.

Repackaging ThO2 feeds 

for digestion-hopper 

packaging

Apr–Aug 

1966

1 - - - 12 Total project operating time 

estimated at 1 week or 10 

shifts.

Repackaging ThO2 feeds 

for digestion-outgoing 

drum conveyor

Apr–Aug 

1966

1 - - - 7 Total project operating time 

estimated at 1 week or 10 

shifts (empty drums).

Refinery

103

Extracted 

& reextracted 

thorium232 & 

daughters

DigestRaffinate

0501

Hopper feed & digestion Apr–May 

1966

3 - - - 6 Test date: Apr 22, 1966. One 

hopper (15,000 pounds) made 

1 tank batch, 20,000 lbs/day 

product on first cycle; recycle 

is liquid feed. 

Raffinate treatment & 

disposal

May–Sep 

1966

3 - - - 2 -

Misc. digestion of 

drummed scrap-wet feed

May–Sep 

1966

2 - - - 4 Test date: May 3, 1966. 

Material was damp to wet; 

floor sweepings were used as 

feed. Non-routine job.

ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Page 1 of 7 
Document No. ORAUT-TKS-0028-5 Revision No.

0

2 Effective Date: 05/21/2013 Page 55 of 60

Plant 

name & 

no. Materials

Work 

group title 

& cost 

centera Job titleb Periodc

Workers/

day

Daily weighted average 

(DWA) concentrations

(alpha dpm/m3)

1963 1964 1965 1966 Notes

Refinery 

Pot Room

103

Extracted 

thoriumnatural

Pot Room

0502d

Pot denitration & sol 

drying-hand transfers

Nov–Dec 

1963; 

May–Aug 

1964

6 88 88 - - Test dates: Nov 1963 and 

May 1964. Airline masks used 

to enter pot sections.

Pot denitration (crystals) 

& sol drying-vacuum 

unload

Aug–Dec 

1964

6 - 60 - - Test samplings for vacuuming 

and packaging collected from 

Sep 1964 to Oct 1965. Airline 

masks prescribed for entrance 

into pot or packaging sections. 

Small crucibles packaged.

Pot denitration (liquor) & 

sol drying-vacuum 

unload

Jan–Aug 

1965

6 - - 48 - Airline masks prescribed for 

packaging and unloading; halfface dust respirators (Comfo) 

for misc. attending duties in 

pot sections. Small crucibles 

packaged.

Sol drying (fluid bed sol)-

vacuum unload

Aug 1965–

Jan 1966; 

Jun–Sep 

1966

6 - - 38 38 Packaging in small crucibles 

(500 lb). Comfo respirators 

prescribed in pot sections.

Pot denitration-hand 

unload & bucket dump to 

sol tank

Oct–Nov 

1965

6 - - 144 - Pot denitration used for special 

project and to maintain 

production at scheduled rate 

when fluid bed was down. 

Airline masks prescribed for 

dumping and scooping. 

Comfo respirators were used 

for vacuuming and packaging 

activities.

ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Page 2 of 7 
Document No. ORAUT-TKS-0028-5 Revision No.

0

2 Effective Date: 05/21/2013 Page 56 of 60

Plant 

name & 

no. Materials

Work 

group title 

& cost 

centera Job titleb Periodc

Workers/

day

Daily weighted average 

(DWA) concentrations

(alpha dpm/m3)

1963 1964 1965 1966 Notes

ThO2 repackaging-multipurpose hoodAug 1964 2 - 260 - - Short-term operation. Airline 

masks used in hood.

DWA index for the Pot 

room work group

Apr 1964–

Dec 1965

- - 78 50 - For Apr-Dec 1964, 

approximately 45% of the 

effort (manpower) was on

thorium jobs. For 1965, 

approximately 50% of the 

effort was on thorium jobs. 

Results are yearly indices and 

represent DWAs for 3 job 

titles.

Extraction

105

Extracted 

thorium 

and 

thorium232 

daughters

Extraction Pumper decanter, pulse 

column, strippers, and 

NOK (product liquor)

Jun–Sep 

1966

9 - - - 2 -

Metals 

301

Extracted 

thoriumnatural

Metals-other 

than 

Reduction

0702

High-firing 1,000 lb 

crucibles at recast 

(furnace)

Nov–Dec 

1963; 

May–Sep 

1964

6 21 21 - - Test dates included Dec 1963, 

May 1964, and Oct 1964. 

Airline masks used for 

entrance into enclosure for 

thermocouple change and 

inspection; operation 

performed 7 days/week.

ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Page 3 of 7 
Document No. ORAUT-TKS-0028-5 Revision No.

0

2 Effective Date: 05/21/2013 Page 57 of 60

Plant 

name & 

no. Materials

Work 

group title 

& cost 

centera Job titleb Periodc

Workers/

day

Daily weighted average 

(DWA) concentrations

(alpha dpm/m3)

1963 1964 1965 1966 Notes

TNT repackaging Oct–Dec 

1964

2 - 3 - - Test date: Oct 5, 1964. For 

part of this period, crushing 

was done externally by electric 

grabs and transferred to the 

refinery without repacking 

TNT.

High-firing 500 lb 

crucibles in billet heaters

Sep 1964–

Jan 1966; 

Jun–Sep 

1966

3 - 11 11 11 Test date Nov 4, 1964. Four 

heaters were installed, each 

handling a 500-lb crucible on a 

24-hourcycle.

Repackaging ThO2 in 

recast enclosure

Oct–Dec 

1965;

Jan 1966;

Jun–Sep 

1966

2 - - 2060 

(55)

55 The DWA result of 2,060 

dpm/m3 is for Oct 1965; airline 

masks and cover clothing were 

used inside the transfer 

enclosure, which exhausted to 

the dust collector. 

Test dates: Sep 16, 1965 

(experimental batch) and Sep 

27, 1965- improvements were 

recommended. 

The DWA result of 55 dpm/m3

is for Nov 1965 and later. 

Test dates: Feb 24, 1966 and 

Jul 14, 1966. Airline masks 

were used for all scooping 

operations.

ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Page 4 of 7 
Document No. ORAUT-TKS-0028-5 Revision No.

0

2 Effective Date: 05/21/2013 Page 58 of 60

Plant 

name & 

no. Materials

Work 

group title 

& cost 

centera Job titleb Periodc

Workers/

day

Daily weighted average 

(DWA) concentrations

(alpha dpm/m3)

1963 1964 1965 1966 Notes

Kiln calcining of NLO 

sump cake

Apr–Jul 

1966

3 - - - 20 Test dates: Apr 21, 1966; Jun 

8-9, 1966. Comfo (MSA) dust 

respirators used on dumping 

platform and at packaging 

stations. Feed was normally 

wet, ranging from damp to a 

surface water layer.

DWA index for the 

Metals (other than 

Reduction) work group

May 1964–

Dec 1965

- - 17 113 - For 1964, approximately 50% 

of the (manpower) effort was 

on thorium jobs, but none (on 

thorium) during the first 4 

months. For 1965, 

approximately 60% of effort

was on thorium jobs. Results 

are yearly indices and 

represent DWAs for 2 to 3 job 

titles.

ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Page 5 of 7 
Document No. ORAUT-TKS-0028-5 Revision No.

0

2 Effective Date: 05/21/2013 Page 59 of 60

Plant 

name & 

no. Materials

Work 

group title 

& cost 

centera Job titleb Periodc

Workers/

day

Daily weighted average 

(DWA) concentrations

(alpha dpm/m3)

1963 1964 1965 1966 Notes

Scrap 

Plant

403

Extracted 

thoriumnatural

Scrap Plant

0601 e

TNT crystalsrepackagingNov–Dec 

1963;

May–Sep 

1964

6 15 15 - - Equivalent production rate of 1 

ton of ThO2/day.

ThO2 repackaging-from 

recast crucibles

Nov–Dec 

1963;

May–Sep 

1964

6-8 290 290 - - After Sep 1964, smaller 

crucibles were used. Airline 

masks were used on all 

scooping operations; 

operations conducted inside 

multi-purpose walk-in hoods. 

Exhaust was filtered by autoair-mat (paper rolls). 

Manpower for 1964 was 

variable (5 in May to 15 in Jul); 

normal manpower was 6 to 8. 

ThO2 repackaging included 

screening, blending, and 

product packaging.

ThO2 repackaging-from 

billet heater crucibles

Sep 1964–

Oct 1965

4 - 20 20 - Test date: May 24, 1965. 

Airline masks used on all 

scooping operations. Shallow 

crucibles could be scooped 

with breathing zone more 

distant from the opening than 

with recast crucibles (lower 

concentrations observed). 

Long-handles scoops were 

used in the blender drum.

ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Page 6 of 7 
Document No. ORAUT-TKS-0028-5 Revision No.

0

2 Effective Date: 05/21/2013 Page 60 of 60

Plant 

name & 

no. Materials

Work 

group title 

& cost 

centera Job titleb Periodc Workers/day

Daily weighted average 

(DWA) concentrations

(alpha dpm/m3)

1963 1964 1965 1966 Notes

Fluid bed denitrationproduct to drumsMay–Jul 

1965

6 - - 44 - Test date Jun 1965. 

Experimental use of NLO 

liquor during May.

Fluid bed denitration-wet 

feed & product

Aug 1965–

Jan 1966;

Jun–Sep 

1966

6 - - 29 29 Liquor feed and sol product at 

1 ton/day production rate.

DWA index for Scrap 

Plant work group

Nov 1963–

Dec 1965

- 180 117 28 - The 1963 index was for a 6-

man group (Nov-Dec) with 

17% of manpower on thorium. 

The 1964 index was for a 6-

man group with 60% of 

manpower on thorium. The 

1965 index was for a 12-man 

group (Jan-Jun) and a 6-man 

group (Jul-Dec) with 80% of 

manpower on thorium. 

Results are yearly indices and 

represent DWAs for 2 to 3 job 

titles.

a. “Cost Center” refers to an employer charge code. Cost center codes are published in some TBDs and can be used for identifying employee work areas. 

b. “Job Title” is the term used in the referenced report; this term more accurately represents a task description rather than an employee job title. 

c. “Time Period” was recorded from the air sampling data sheets and was estimated if date ranges were incomplete. 

d. Manpower in this group ranged from 18 to 21 men during the years 1963 through 1965; base production rate was 1 ton ThO2/day. 

e. Average production rate was 1 ton ThO2/day.

ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Page 7 of 7