Page 1 of 61

DOE Review Release 03/22/2017

Weldon Spring Plant – Occupational Internal 

Dose

ORAUT-TKBS-0028-5 Rev. 03

Effective Date: 03/14/2017

Supersedes: Revision 02

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

Document Owner 

Approval: Jodie L. Phillips Signature on File for

David P. Harrison, Document Owner

Approval Date: 03/14/2017

Concurrence: Wade C. Morris Signature on File for

John M. Byrne, Objective 1 Manager

Concurrence Date: 03/08/2017

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 03/08/2017

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 03/08/2017

Approval: Signature on File

Stuart L. Hinnefeld, Director DCAS

Approval Date: 03/14/2017

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0028-5 Revision No. 03 Effective Date: 03/14/2017 Page 2 of 61

 

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 230Th and progeny 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. Sitespecific ratios of 230Th 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 232Th 

operations, and equations were included in the text regarding 

calculation of the median and 95th percentiles of the 232Th 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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EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/14/2017 03 Revision initiated in response to revision of ORAUT-TKBS-0017-5, 

Feed Material Production Center – Occupational Internal Dose, in 

relation to recycled uranium contaminants. Updated discussion of 

recycled uranium in Section 5.2.4 in reference to the recycled 

uranium contaminant mass concentrations detailed in the Fernald 

internal dose TBD. Added Section 5.6.1.3.3 for RU including tables 

in units of Bq RU contaminant per g U and in Bq/Bq U to assist the 

DR in calculating the RU contaminant activities. Changed the value 

for the specific activity for 1% enriched uranium from 0.783 pCi/μg 

to 0.973 pCi/μg to match that used in the revised Fernald internal 

dose Technical Basis Document (Rev 02). Includes editorial 

changes. 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.
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TABLE OF CONTENTS

SECTION TITLE PAGE

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

5.1 Introduction .............................................................................................................................. 9

5.1.1 Purpose...................................................................................................................... 10

5.1.2 Scope......................................................................................................................... 10

5.1.3 Background on Mallinckrodt Chemical Works and the Weldon Spring Plant............... 10

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

5.2 Source Term .......................................................................................................................... 15

5.2.1 Isotopic Composition of Uranium ................................................................................ 15

5.2.2 Uranium Decay Products............................................................................................ 15

5.2.3 Natural Thorium and Decay Products......................................................................... 16

5.2.4 Recycled Uranium ...................................................................................................... 18

5.2.4.1 Other Contaminants in Recycled Uranium ................................................... 18

5.2.5 Solubility Classification and Absorption Type.............................................................. 18

5.2.6 Particle Size ............................................................................................................... 19

5.3 In Vitro Measurements ........................................................................................................... 19

5.3.1 Operational Period, 1957 to 1966 ............................................................................... 19

5.3.1.1 Uranium ....................................................................................................... 19

5.3.1.1.1 Routine Sampling Program ....................................................... 19

5.3.1.1.2 Special Sampling Program ........................................................ 21

5.3.1.1.3 Data Reporting Levels and Minimum Detectable Amounts ........ 21

5.3.1.1.4 Interferences and Uncertainties................................................. 21

5.3.1.1.5 Reporting Formats and Codes................................................... 22

5.3.1.1.6 Work Group Data ...................................................................... 24

5.3.1.2 Thorium........................................................................................................ 24

5.3.2 Environmental Monitoring Period, 1975 to 1984 ......................................................... 24

5.3.3 Remediation Period, 1985 to 2002 ............................................................................. 24

5.4 In Vivo Measurements ........................................................................................................... 35

5.4.1 Operational Period, 1957 to 1966 ............................................................................... 35

5.4.2 Environmental Monitoring Period, 1975 to 1984 ......................................................... 36

5.4.3 Remediation Period, 1985 to 2002 ............................................................................. 36

5.5 Air Concentration and Dust Measurements............................................................................ 36

5.5.1 Operational Period, 1957 to 1966 ............................................................................... 36

5.5.1.1 Maximum Acceptable Concentrations .......................................................... 37

5.5.1.2 Special Curie for Uranium ............................................................................ 37

5.5.1.3 Dust Exposure Calculation ........................................................................... 37

5.5.1.4 Dust Studies................................................................................................. 38

5.5.2 Transfer to the Army, 1967 to 1974 ............................................................................ 38

5.5.3 Environmental Monitoring Period, 1975 to 1984 ......................................................... 40

5.5.4 Remediation Period, 1985 to 2002 ............................................................................. 40

5.6 Assessment of Intakes ........................................................................................................... 41

5.6.1 Operational Period, 1957 to 1966 ............................................................................... 41

5.6.1.1 Uranium and Related Intakes....................................................................... 41

5.6.1.2 Thorium-232 Intakes .................................................................................... 42
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5.6.1.3 Intakes of Other Radionuclides .................................................................... 44

5.6.1.3.1 Radon ....................................................................................... 44

5.6.1.3.2 Thoron....................................................................................... 45

5.6.1.3.3 Recycled Uranium..................................................................... 46

5.6.2 Environmental Monitoring Period, 1975 to 1984 ......................................................... 46

5.6.3 Remediation Period, 1985 to 2002 ............................................................................. 47

5.7 Summary of Instructions to Dose Reconstructors................................................................... 47

5.8 Attributions and Annotations .................................................................................................. 48

References ........................................................................................................................................ 49

Glossary ............................................................................................................................................ 53

ATTACHMENT A DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS............... 57

LIST OF TABLES

TABLE TITLE PAGE

5-1 History of dust-generating activities........................................................................................ 12

5-2 Potential internal radionuclide exposure for production buildings ........................................... 13

5-3 Products or intermediates in production buildings .................................................................. 14

5-4 Mass and percent of feed materials processed ...................................................................... 15

5-5 Thoron and isotopic progeny.................................................................................................. 17

5-6 Solubility class summary........................................................................................................ 20

5-7 Cost center codes for workers................................................................................................ 22

5-8 Composite urine statistical summary...................................................................................... 25

5-9 Uranium urine data summary by cost center for 1958 ............................................................ 26

5-10 Uranium urine data summary by cost center for 1959 ............................................................ 27

5-11 Uranium urine data summary by cost center for 1960 ............................................................ 28

5-12 Uranium urine data summary by cost center for 1961 ............................................................ 29

5-13 Uranium urine data summary by cost center for 1962 ............................................................ 30

5-14 Uranium urine data summary by cost center for 1963 ............................................................ 31

5-15 Uranium urine data summary by cost center for 1964 ............................................................ 32

5-16 Uranium urine data summary by cost center for 1965 ............................................................ 33

5-17 Uranium urine data summary by cost center for 1966 ............................................................ 34

5-18 Work group DWA concentrations for uranium operations....................................................... 39

5-19 Time-weighted average 232Th dust concentrations ................................................................. 39

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

pits .........................................................................................................................................42

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

measurements .......................................................................................................................42

5-22 Thorium-232 inhalation intake rates (Bq/yr)............................................................................ 43

5-23 RU contaminant intakes per unit mass of uranium (Bq/g U). .................................................. 46

5-24 RU contaminant intakes per unit activity of uranium (Bq/Bq U). ............................................. 46

5-25 Instructions to DRs.................................................................................................................47

A-1 DWA dust concentrations for thorium production jobs ............................................................ 57
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LIST OF FIGURES

FIGURE TITLE PAGE

5-1 Decay and buildup of 228Th and daughters after processing................................................... 17
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AM arithmetic mean

AMAD activity median aerodynamic diameter

AWE atomic weapons employer

Bq becquerel

BZA breathing-zone air

CEDE committed effective dose equivalent

CFR Code of Federal Regulations

Ci curie

d day

D days (solubility class)

DCAS Division of Compensation Analysis and Support

DAC derived air concentration

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

dps disintegrations per second

DR dose reconstructor

DU depleted uranium

DWA daily weighted average

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

F fast (absorption type)

ft foot

g gram

GM geometric mean

GSD geometric standard deviation

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

kg kilogram

L liter

lb pound

M moderate (absorption type)

MAC maximum allowable concentration

MCW Mallinckrodt Chemical Works

MDA minimum detectable amount

MDC minimum detectable concentration

MeV megaelectron-volt, 1 million electron-volts
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mg milligram

min minute

mL milliliter

mo month

MPC maximum permissible concentration

mrem millirem

MT metric ton

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NLO National Lead Company of Ohio

ns nanosecond

NU natural uranium

ORAU Oak Ridge Associated Universities

PAEC potential alpha energy concentration

pCi picocurie

POC probability of causation

PUREX plutonium-uranium extraction

RU recycled uranium

S slow (absorption type)

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TNT thorium nitrate tetrahydrate

U.S.C. United States Code

W weeks (solubility class)

WL working level

WLM working level month

WSP Weldon Spring Plant

WSQ Weldon Spring Quarry

WSRP Weldon Spring Raffinate Pits

WSSRAP Weldon Spring Site Remedial Action Project

Y years (solubility class)

yr year

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 Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

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

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

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

of the individual work required for each dose reconstruction.

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

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

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

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

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined as “any building, structure, or premise, 

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

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

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

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

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

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

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

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

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

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

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

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

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

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (NIOSH 2010a):

• 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
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5.1.1 Purpose

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

practices at the Weldon Spring Plant (WSP), the Weldon Spring Quarry (WSQ), and the Weldon 

Spring Raffinate Pits (WSRP). This document contains technical information on the history, methods, 

and interpretation of monitoring data for the evaluation of occupational internal dose to WSP workers.

5.1.2 Scope

This TBD covers the methods used to assess internal radiation dose to workers at WSP, WSQ, and 

WSRP. Historically, the Weldon Spring Plant has also been called the Weldon Spring Site, Weldon 

Spring Chemical Plant, and the Weldon Spring Feed Materials Plant. The WSP is also known as the 

Chemical Plant, Main Plant, or Main Site. For convenience, WSP is used throughout the remainder of 

this document where it is unnecessary to distinguish between the plant, the Quarry, or the Raffinate 

Pits. 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 at WSP, intake through the skin was unlikely.

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 WSP processed, 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 Background on Mallinckrodt Chemical Works and the Weldon Spring Plant

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. 5).

The operations at the two MCW facilities were similar. One of the reasons for the construction of the 

new facility was the level of contamination 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 tetrafluoride (UF4), and uranium metal (Meshkov et 

al. 1986, p. 27). UO2 was an intermediate product in the conversion of UO3 to UF4 (ORAUT 2013a). 

The feed material at the Destrehan Street plant included high-quality pitchblende [a natural uranium 

(NU) ore that includes the decay products generally in equilibrium]. The WSP facility primarily 

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

concentrations of decay products. Ingle (1991, pp. 3–7) reports that WSP did not receive the highquality pitchblende ores. 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 latter part of its operational period. In 1964, MCW published a 

document to encourage diversification and utilization of the existing facilities (MCW ca. 1964, pp. 5–

29), which reflected that decline. The document described the plant facilities and briefly mentioned 

the operational safety program.

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 ORAUT-TKBS-0028-6, Weldon Spring Plant – Occupational 

External Dosimetry (ORAUT 2013b). 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; and

• Remediation, 1985 to 2002. 

WSQ was initially transferred from the Army to the AEC in 1958. The operational and remediation 

periods are of primary concern in relation to internal dosimetry at WSP, with some potential 

applicability of shutdown operations in 1967 before transfer to the Army 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 herbicide production. However, 

the defoliant project was canceled in February 1969, and Army placed the WSP into care and custody 

status. AEC did not have contractors for environmental monitoring at the WSRP or the WSQ until 

August 1975. From 1969 to 1981 the status of the site did not change, and from 1981 to 1985 the site 

was again placed in caretaker status. WSP 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, so this TBD does not cover that period.

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

1942-1967 describes the history and elements of the MCW health and safety programs (MCW 1967, 

pp. 154–166). 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.” The MCW health and safety program was based on standard industry procedures 

for handling toxic chemicals. The radiation safety program at 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 from 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 ORAUTTKBS-0028-2, Weldon Spring Plant – Site Description (ORAUT 2013a). Ore concentrates of 60% to 

70% uranium (yellowcake) were sampled on receipt at the facility. Some of the material was 

repackaged in drums and some was sent directly for processing. The concentrates were digested 

with nitric acid to produce uranium nitrate solution, which 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 UO3, UF4, and U metal

• Some Th processing between 1963 and 1966

• Raffinate from processing removed to raffinate pits

• 14,500 MT 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–

unknown

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 MT rubble removed to the quarry

• About 2,000 MT scrap moved to Tennessee

• About 200 MT steel parts moved to Ohio

• About 100 MT 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 Company and Jacobs Engineering Group assumed responsibility for 

the WSSRAP

a. Meshkov et al. (1986, pp. 27–33); Lesperance, Siegel, and McKinney (1992, pp. 30–36).

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.

5.1.4 Radiation Protection Practices

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

a

Building

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 

Building 103 for processing.

NU 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. U-bearing 

solution sent to Building

105 for purification. 

Materials returned for 

denitration after purification 

and sent to Building 201. 

NU 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 purified by 

solvent extraction and 

returned to Building 103 for 

denitration.

Wet process but some potential for U or Th (in 1966) 

dust exposure. U would have been the major internal 

exposure component, but Rn-222 and decay product 

exposure possible.

108 Acid 

recovery

Recovering and 

reconcentrating nitric acid.

Rn gas and its decay products.

201 Green salt 

(UF4) plant 

Feed from Building 103 

(after denitration) converted 

to UF4.

Potential for uranium exposure as UF4 dust or natural 

Th (1965–1966). No significant Th-230, Ra-226, 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; acceptable cores 

shipped to reactor sites.

Potential for U exposure as UF4 dust and U3O8 or 

natural Th (1965–1966). No significant Th-230, Ra226, or decay product exposure.

Th-232 and decay products starting November 1963. 

403 Chemical 

pilot plant

Small-scale chemical 

processes. 

NU 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 U exposure as UF4 and U3O8 or natural 

Th (1965–1966). No significant Th-230, Ra-226, or 

decay product exposure.

407 Analytical 

and research 

laboratories 

Small-scale research and 

analytical work on products 

and processes.

NU 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).

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

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

(EU), depleted uranium (DU) 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.
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Table 5-3. Products or intermediates in production buildings. 

Building Products or intermediates present

101 Ore concentrates, yellowcake

103 Yellowcake, UO3,

U solution in nitric acid

Purified U solution

Denitrated U solution

105 UO3 in nitric acid

Purified U solution

108 Recovered nitric acid

201 Purified, denitrated U solution

UO3, UO2 (intermediate compound)

UF4

301 UF4

U metal

U3O8

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 NU feed materials at WSP were 

“essentially free of radioactive progeny 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 that began in 1985 involved demolition of the buildings that were 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 that for the buildings that made up the WSP because the potential 

for exposure to thorium and radium was not as high in the WSP. The high-grade pitchblende ores at 

the Destrehan Street facility were not present at the WSP. Thorium was processed at WSP for a 

limited period. The WSP buildings that were 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,

1998b, 1998c, 2000a, 2001b). 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 programs, Section 5.4 discusses in vivo lung counting, and Section 5.5 describes air

monitoring practices.
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5.2 SOURCE TERM

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

primary feed material was NU in the form of yellowcake. WSP also processed DU and EU (up to 1%) 

as well as natural thorium. Table 2-7 in ORAUT (2013a) 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

NU 122,015,977 98.43

DU 167,823 0.14

EU (1%) 842,585 0.68

Natural Th 941,347 0.76

Totals 123,967,732 100

5.2.1 Isotopic Composition of Uranium

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 NU and DU 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 NU (0.683 pCi/μg) and DU (0.402 

pCi/μg). 

Refer to Table 5-2 of ORAUT (2016) for isotopic mass fractions and relative activities for the uranium 

isotopes associated with depleted uranium, natural uranium, and enriched uranium.

DOE (1986 pp. 17 & 27; 2000c p. 35) indicates that the enriched uranium shipped to the WSP was 

equal to or less than 1%. Thus, it is reasonable to assume that the EU at Weldon Spring was 1% 

enriched with a specific activity of 0.973 pCi/μg. Bioassay results in units of mass can 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 EU at WSP.

5.2.2 Uranium Decay Products

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

short-lived decay products of 238U, which are 234Th (24-day half-life) and 234mPa (1.175-minute halflife), 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 (less than 8 months). The 234mPa beta is a highenergy beta and contributes to the external dose but, due to its short half-life, does not in itself 

contribute to internal dose.

The primary source of decay products (230Th and 226Ra) for the materials 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. 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 ORAUT-TKBS-0028-
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2 (ORAUT 2013a), four raffinate pits were constructed between 1958 and 1964 to contain process 

wastes. 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 slurry to the 

aqueous feed tanks in Building 105 before separation or uranium purification (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 Revision 0 of the 

Internal Dosimetry Program Technical Basis Manual 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 Natural Thorium and Decay Products

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-year half-life) and 228Th (1.9-year half-life) 

would be significant if not complete. Radium-228 is a beta emitter that decays to 228Th through 228Ac. 

Thorium-228 decays by alpha emission to 224Ra (3.66-day 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 2010b, p. 57). ORAUT-TKBS-0028-2 (ORAUT 2013a) 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 NU. 

Thoron (220Rn) is the second progeny of 228Th. Within a few weeks of processing of thorium ores for 

thorium purification, it can be considered to be in full equilibrium with the parent, 228Th. Thorium-228 

is generally in 40% to 65% equilibrium with 232Th for materials at the WSP. The degree of equilibrium 

is dependent on both the decay of 228Th (without replenishment from the 5.7-year half-life of 228Ra) 

after removal of the thorium progeny and the time it takes the 228Ra to build into equilibrium with 232Th. 

Figure 5-1 shows the decay and buildup of 228Th and progeny after processing.

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

and the associated waste storage locations. The thoron, with its subsequent progeny, 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 progeny 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 

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

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

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

system, the ventilation system, or both. The dose from exposure to thoron essentially comes 

entirely from the progeny 212Pb and 212Bi. A thoron working level (WL) requires 375 pCi/L 

thoron and 216Po to produce measured progeny 212Pb and 212Bi at 7.5 pCi/L assuming a 

conservative equilibrium factor of 0.02 [1].
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Figure 5-1. Decay and buildup of 228Th and progeny after processing. 

Table 5-5. Thoron and isotopic progeny. 

Isotope Half life

Concentration 

(atoms/7.5 pCi)

Alpha energy 

(MeV/atom)

PAECa

(MeV/7.5 pCi)

Ra-220 56 s 23 14.6 335

Po-216 0.15 s <1 (b) Not applicable

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

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

Po-212 0.3 μs Not applicable (b) Not applicable

Tl-208 3 min 73 No alpha Not applicable

Total Not applicable Not applicable Not applicable 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-ns 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). Because Po-216

ultimately decays by two alphas (Po-216 and Bi-212) the alpha energy released per Po-216

atom is 14.6 MeV. Lead-212 and Bi-212 both decay with an effective alpha energy of 7.8 MeV.

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. WSP records do not indicate specific analyses to define the 

concentrations of thoron progeny activities, but 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 progeny to decay.

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

with the 230Th and 226Ra concentrations, and would therefore contribute only a very small fraction of 

the potential dose. In contrast, the estimated maximum 232Th concentration in materials deposited at 

the WSQ was 4,200 pCi/g (DOE 1991, p. 18).
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5.2.4 Recycled Uranium

Recycled uranium (RU) is uranium that has been recovered from irradiated production reactor fuel 

and plutonium production fuels and returned (hence, recycled) to the enrichment and/or feed 

production facilities to fabricate new fuels for nuclear reactors and various uranium compounds and 

metals. The RU sent from chemical separation facilities contained trace amounts of transuranic 

elements (including neptunium and plutonium) and fission products (such as technetium). The extent 

of the processing of RU at WSP is not well known (ORAUT 2013c, Section 2.2.3). However, the Ohio 

Field Office Recycled Uranium Project Report (DOE 2000b) indicates that Weldon Spring received 

relatively small quantities of uranium materials that might have contained RU constituents. Therefore, 

WSP was identified as a minor receiver and processor of RU. The DR should make the assumption 

that all of the uranium processed at WSP after 1961 was RU; this is favorable to claimants. This 

assumption is consistent with that in the DOE report, which assumed that all uranium receipts at WSP 

after 1961 were RU. This was based on the fact that WSP received most of the RU from the Feed 

Materials Production Center, which in turn received its first major shipment of RU from Hanford in 

Production Order A500 in February 1961.

5.2.4.1 Other Contaminants in Recycled Uranium

RU was assumed to have been shipped to WSP from Fernald beginning in 1961, so none of the 

associated contaminants was present before 1961. Therefore, uranium intakes in 1961 and later 

should include default factors to account for exposure to the RU contaminants. NU is assumed from 

1961 to 1962, and EU (1%) is assumed from 1963 to 1966. ORAUT (2016) indicates that recycled 

uranium began to be received at Fernald in 1961, and that 1% EU was assumed from the start of 

plant operations through 1964 and 2% EU was assumed thereafter. However, it is evident that the 

enriched uranium sent to WSP from Fernald did not exceed 1% uranium-235 (DOE 1986, p. 17 & 27; 

DOE 2000c p. 35). 

After 1966, the ending material balance was 0 kg in fiscal year 1967 for both EU and NU at WSP, but 

an estimated 152 MT NU versus 2.8 MT EU were sent as discards to the raffinate pits (DOE 1986). 

This indicates exposure to NU during monitoring and remediation efforts was much more likely. 

However, EU is assumed to continue after 1966 throughout the remainder of the covered periods. 

Weapons-grade plutonium mix (6% 240Pu) is assumed to determine the 241Pu and 241Am activities, with 

plutonium alpha assumed to be 100% 239Pu. The age of the plutonium mix (i.e., the time after 241Am 

was separated from the mix) was chosen in accordance with ORAUT (2016). Other contaminants 

were added in accordance with ORAUT (2016) and Section 5.6.1.3.3.

Accounting for the age of plutonium mix, start of RU receipts, and change from NU to EU, the periods 

used to calculate RU contaminant activities are 1961–1962; 1963–1965; 1966–1975; 1976–1985; and 

1986–2002. RU contaminant activities are stated in Tables 5-23 and 5-24.

5.2.5 Solubility Classification and Absorption Type

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.

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

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

uranium compounds 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).
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The site’s remediation contractor collected samples of materials from the areas in WSP 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 sampled building. Table 5-6 lists those 

fractions for most of the WSP 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 1994a).

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 (DOE 2001b). 

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

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

information for a particular dose reconstruction.

5.2.6 Particle Size

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

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

diameter (AMAD) should be used (ORAUT 2014a).

5.3 IN VITRO MEASUREMENTS 

5.3.1 Operational Period, 1957 to 1966

5.3.1.1 Uranium

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 Routine Sampling Program

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 uranium urine sampling program described below appears to 

have been implemented and continued through 1966 (MCW 1965):

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.
 

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Table 5-6. Solubility class summary (DOE 2001b).

Area

U-234 

D

U-234 

W

U-234 

Y

U-235 

D

U-235 

W

U-235 

Y

U-238 

D

U-238 

W

U-238 

Y

Th-232 

W

Th-232 

Y

Th-230 

W

Th-230 

Y

Th-228 

W

Th-228 

Y

Building 101 0.41 NAa 0.59 0.28 NA 0.72 0.41 NA 0.59 0.20 0.80 NA 1.0 0.29 0.71

Building 103 0.20 0.50 0.30 0.20 0.44 0.36 0.75 0.25 NA NA 1.0 NA 1.0 NA 1.0

Building 105 0.20 0.50 0.30 0.20 0.44 0.36 0.75 0.25 NA NA 1.0 NA 1.0 NA 1.0

Building 108 0.19 0.20 0.61 0.14 0.47 0.39 0.19 0.20 NA NA 1.0 NA 1.0 NA 1.0

Building 201 0.44 NA 0.56 0.51 NA 0.49 0.42 NA 0.58 NA 1.0 NA 1.0 NA 1.0

Building 301 0.12 NA 0.88 0.09 NA 0.91 0.12 NA 0.88 0.03 0.97 0.09 0.91 0.03 0.97

Building 403 0.19 0.20 0.61 0.14 0.47 0.39 0.19 0.20 0.61 NA 1.0 NA 1.0 NA 1.0

Building 406 1.0 NA NA 1.0 NA NA 1.0 NA NA 0.35 0.65 0.68 0.32 0.65 0.35

Building 408 1.0 NA NA 1.0 NA NA 1.0 NA NA 0.35 0.65 0.68 0.32 0.65 0.35

Building 410 1.0 NA NA 1.0 NA NA 1.0 NA NA 0.35 0.65 0.68 0.32 0.65 0.35

Building 417 1.0 NA NA 1.0 NA NA 1.0 NA NA 0.35 0.65 0.68 0.32 0.65 0.35

Pit 3 1.0 NA NA 1.0 NA NA 1.0 NA NA NA 1.0 NA 1.0 NA 1.0

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

Q-Sump 0.61 NA NA 0.52 0.47 0.48 0.58 NA 0.42 0.10 0.90 NA 1.0 0.10 0.90

a. NA = not available
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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.

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

5.3.1.1.2 Special Sampling Program

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 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.

5.3.1.1.3 Data Reporting Levels and Minimum Detectable Amounts

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

system and data analysis.

MCW apparently did not report uranium urine results as censored or 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 2014b) 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 100-μL (100λ) 

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

the MDA for WSP, although there could have been 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 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 Interferences and Uncertainties

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

of uranium. It is likely they were not. Measurements more sensitive than the Dupree (1979b, p. 26) 

value of 0.002 ±0.002 mg/L during the remediation period on persons in the WSP vicinity not 

occupationally exposed to uranium indicated a geometric mean (GM) 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.
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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 have revealed 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 Reporting Formats and Codes

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

• A photocopy of the original, handwritten urine data cards;

• A computer printout, Uranium Urine (MCWURWS); and

• A computer printout, Uranium Urine (MCWURDES).

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.a 

General operations

Work group To early 1963 Starting 1963

General Engineering 110 (b)

Chemical Technology 120 (b)

Metal Technology 121 (b)

Plant Services 140 (b)

WSP

Work group To early 1963 Starting 1963

Administration 300 (b)

Production facilities

Work group To early 1963 Starting 1963 

Sampling Plant, Sampling 310 110

Sampling Plant, Repackaging 321 110

Refinery, Digest – Raffinate 320 120

Refinery, Pot Room 320 120

Refinery, Extraction 320 120

Green Salt Plant, Green Salt 330 150

Metal Plant, Dingot Reduction 360, 340 180
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Work group To early 1963 Starting 1963 

Metal Plant, Extrusion, General 359 200

Metal Plant, Extrusion, Gamma (Dingot Extrusion) 361, 351 210

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

Metal Plant, Scrap Activity 358 280

Metal Plant, Vacuum Outgassing 363 (b)

Metal Plant, Core Fabrication 362, 365 to 369 250

Services groups or facilities

Work group To early 1963 Starting 1963

Maintenance 370 370

Storeroom, Receiving, & Shipping 372 530

Disposal of Construction Inventory 373 (b)

Instrument Shop 374 (b)

Boiler House 376 (b)

Water Plant 378 (b)

Warehouse 380 550

Engineering 390 560

Plant Protection 392 510

Laundry 394 (b)

Custodians 396 520

Occupancy, Labor 398 (b)

Technical Division

Work group To early 1963 Starting 1963

Administrative 400 (c)

Metallurgical Development (Building 404), 410 (c)

Metallurgical Development, Special Projects 460 460

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

Process Development, Scrap Plant No code 290

Research Laboratory 430 (c)

Analytical Laboratory 440 (c)

Administration Division

Work group To early 1963 Starting 1963

Uranium Division Management 600 600

Salaried Personnel Administration 605 (c)

Accounting 610 610

Purchasing 620 (c)

Motor Pool 621 (c)

Administrative Services 630 (c)

Production Control 640 640

Health and Safety 650 (c)

General Division Expenses 660 (c)

Data Processing 670 670

Industrial Relations 680 (c)

Cafeteria 685 (c)

a. Source: MCW, ca. 1961.

b. Unknown; likely discontinued or merged with another cost center.

c. 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.
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An asterisk after the sample result on the original data cards, or in the column headed by SAMTYPE 

on the computer printouts, 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 a 

spreadsheet program function such as WEEKDAY(), which converts a date to a day-of-the-week 

value where 1 is Sunday, 2 is Monday, and so forth. Notations written on the original data cards after 

the result or under SAMTYPE on the computer printouts indicate preemployment samples (code P), 

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

5.3.1.1.6 Work Group Data

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.

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 

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

routine urine bioassay samples by year.

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.

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 Environmental Monitoring Period, 1975 to 1984

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

5.3.3 Remediation Period, 1985 to 2002

An extensive, state-of-the-art bioassay monitoring program was conducted from 1991 to 2001 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, 1998b,

1998c, 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 end of 

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

measurements, as appropriate.
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Table 5-8. Composite urine statistical summary. 

Year Type of analysis

Number of 

records

Median

(mg/L)

95th percentile

(mg/L)

Maximum

(mg/L)

1958 All 1,872 0.005 0.025 0.203a

1958 Routine 1,714 0.005 0.025 0.203

1958 Routine–Monday 321 0.004 0.021 0.047

1958 Routine–Friday 158 0.006 0.023 0.069

1959 All 2,285 0.005 0.024 0.339

1959 Routine 2,090 0.006 0.026 0.339

1959 Routine–Monday 1,124 0.004 0.017 0.041

1959 Routine–Friday 522 0.009 0.048 0.339

1960 All 4,396 0.012 0.038 0.759

1960 Routine 4,246 0.012 0.040 0.759

1960 Routine–Monday 2,602 0.010 0.026 0.088

1960 Routine–Friday 1,347 0.018 0.068 0.759

1961 All 4,184 0.011 0.036 0.344

1961 Routine 4,077 0.011 0.036 0.344

1961 Routine–Monday 2,608 0.010 0.025 0.062

1961 Routine–Friday 1,279 0.018 0.050 0.344

1962 All 3,083 0.010 0.032 0.700

1962 Routine 2,954 0.007 0.024 0.700

1962 Routine–Monday 1,847 0.008 0.020 0.064

1962 Routine–Friday 946 0.013 0.044 0.700

1963 All 3,481 0.014 0.042 0.340

1963 Routine 3,358 0.014 0.041 0.336

1963 Routine–Monday 1,922 0.014 0.029 0.336

1963 Routine–Friday 1,074 0.018 0.056 0.258

1964 All 3,476 0.012 0.052 0.626

1964 Routine 3,264 0.012 0.048 0.626

1964 Routine–Monday 1,767 0.012 0.029 0.282

1964 Routine–Friday 1,049 0.014 0.060 0.626

1965 All 2,980 0.010 0.036 0.812

1965 Routine 2,804 0.009 0.032 0.812

1965 Routine–Monday 1,460 0.009 0.027 0.347

1965 Routine–Friday 869 0.010 0.045 0.318

1966 All 2,145 0.006 0.029 0.459

1966 Routine 1,680 0.007 0.028 0.459

1966 Routine–Monday 882 0.008 0.025 0.066

1966 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.
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Table 5-9. Uranium urine data summary by cost center for 1958. 

Cost center

Median

(mg/L)

95th percentile

(mg/L)

Maximum

(mg/L)

No. of 

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

Accounting 0.003 0.006 0.010 17

Administrative 0.004 0.015 0.017 16

AEC 0.003 0.010 0.010 9

Analytical Laboratory 0.004 0.012 0.039 105

Boiler House 0.002 0.008 0.008 14

Chemical Operator 0.004 0.030 0.061 24

Decontamination 0.003 0.006 0.006 9

Dx (possibly Dingot 

Extrusion) 0.012 0.029 0.040 25

Engineering 0.003 0.014 0.020 33

Furnace 0.008 0.019 0.022 24

Green Salt 0.005 0.018 0.025 30

Guard 0.004 0.013 0.014 41

Health & Safety 0.004 0.007 0.016 28

Instrument Shop 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

Porter 0.005 0.016 0.020 25

Product Development 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
 

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

Cost center

Monday 

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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

Cost center

Monday 

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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 N/A N/A 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).
 

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

Cost center

Monday 

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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

a. N/A = not applicable (five or fewer records).
 

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

Cost center

Monday 

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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

a. N/A = not applicable (five or fewer records).
 

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

Cost center

Monday 

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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
 

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

Cost center

Monday 

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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
 

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

Cost center

Monday 

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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
 

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

Cost center

Monday

median

(mg/L)

Monday 

95th percentile

(mg/L)

Monday 

maximum

(mg/L)

Number of 

records

Friday 

median

(mg/L)

Friday 

95th percentile

(mg/L)

Friday 

maximum

(mg/L)

Number of 

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
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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 coworker bioassay data are available for a worker for 

1985 to 1990, the environmental data for 1985 to 1990 in ORAUT-TKBS-0028-4, Weldon Spring Plant 

– Occupational Environmental Dose (ORAUT 2013b), can be used to estimate intake. Individual or

coworker bioassay data for the subsequent year (1991) may also 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 from 1994 to 

1998 (DOE 1994, p. 31; 1997, p. 22; 1998a, p. 21) and 0.0524 μg/L from 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 Operational Period, 1957 to 1966

There is no indication 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 (Ingle 1991, p. 6):

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. 

The reports of these measurements 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 

parent 232Th after 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 of the detection of 208Tl as a marker for thorium. 
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5.4.2 Environmental Monitoring Period, 1975 to 1984

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

5.4.3 Remediation Period, 1985 to 2002

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 who

would be exposed to airborne radioactivity at the 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 from the Helgeson Scientific Services mobile counting laboratory. 

The lower limits of detection for that system were 74 Bq (2 nCi) for NU and 37 Bq (1 nCi) for 232Th in 

the lung (DOE 1991, p. 28).

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

Feed Materials Production Center. The sensitivities (of unstated pedigree) 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 of the WSSRAP technical basis manual states (DOE 2001b, p. 20):

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 after a 

suspected or actual intake. Revision 7 states that such measurements were normally reserved for 

“those incidents where the intake was suspected to exceed 500 mrem CEDE ...”; and that “requiring 

lung counts for affected workers” was one of the actions “never used” (DOE 2001b, p. 30). 

5.5 AIR CONCENTRATION AND DUST MEASUREMENTS

5.5.1 Operational Period, 1957 to 1966

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

concentration index (also called daily weighted exposure) was calculated based on a combination of 

breathing-zone air (BZA) samples and general air samples (MCW 1965, p. 16).

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 were applied to specific jobs.
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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 therefore 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 NU from the 

1959 ICRP Publication 2 for maximum permissible concentrations (MPCs) were 7 × 10-11 μCi/cm3 and 

6 × 10-11 μCi/cm3

, respectively (ICRP 1959, Volume 3, p. 114). As noted above, these values were 

adopted by the AEC. The Publication 2 MPCs for soluble and insoluble natural thorium were much 

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

, respectively (ICRP 1959, 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 Chapter 0524, “Standards for Radiation Protection,” of the AEC Manual

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

average MAC for uranium 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 NU. The special curie was defined in 1959 as follows (NCRP 1973, p. 3):

Special curie = 3.7 × 1010 dps 238U + 3.7 × 1010 dps 234U + 9 × 108 dps 235U (5-1) 

= 7.49 × 1010 dps 

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

the use of special curies 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.

5.5.1.3 Dust Exposure Calculation

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

for WSP, includes a section for average dust concentration in dpm/cm3 by calendar quarter. None of 

the reviewed exposure reports had 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 dpm/cm3

, 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 
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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 1960–1961). 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, 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 historical 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 micrograms

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.

Natural thorium (

232Th) airborne dust concentrations were routinely measured and evaluated 

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

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

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

years. These operations were intermittent and did not represent the total work year. The timeweighted 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.

5.5.2 Transfer to the Army, 1967 to 1974

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.
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Table 5-18. Work group DWA concentrations for uranium operations (μg U/m3). 

Work group title 1958 1959 1960 1961 1962 1963 1964 1965 1966

Sampling and repackaging cost 

center 110

65 55 45 55 32 31 20 60 NAa

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

Extraction, cost center 120 4 NA 3 NA 5 NA 2 2 NA

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

Green salt, cost center 150 40 37 37 53 53b 16b 16 22 NA

Dingot reduction, cost center 180 NA NA 115 65 NA 150b 150b 78 50

Metal – other than reduction 

cost centers 200, 210, 220, 280

NA NA NA 44 52 63 40c 55d NA

Core area, cost center 250 NA NA 20 NA NA NA NA NA NA

Pilot/scrap plants, cost center 290 NA NA 50e NA 50e 50e 50e 50e NA

Special projectsf NA NA NA NA NA NA NA NA NA

a. NA = not available.

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

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

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

e. Overall average; not weighted.

f. No averages given.

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

Operation Building 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 1965–1966 <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

Miscellaneous operations 103 1966 <0.1 Nonroutine job

ThO2 repackaging 103 1964 3.7 Short term; airline masks used in hood

Extraction 105 1966 <0.1 No information

Crystal denitration 103 1963–1964 1.25 4 mo; airline masks used

Crystal denitration 103 1964–1965 0.9 Airline masks prescribed

Thorium nitrate tetrahydrate

(TNT) liquor denitration

103 1965 0.7 4 mo; airline masks prescribed

Solution drying–vacuum 

unload

103 1965–1966 0.6 4 mo; Comfo respirators prescribed

Pot denitration 103 1965 2 1 mo; airline mask prescribed for some 

tasks

High firing at recast furnace 301 1963–1964 0.3 8 mo; airline masks for some tasks

High firing in billet heaters 301 1964–1966 0.15 No information

TNT repackaging 301 1964–1965 0.04 No information

ThO2 repackaging 301 1965 30 1 mo; 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

TNT repackaging 403 1963–1964 0.2 8 mo

ThO2 repackaging 403 1963–1964 3.3 8 mo; airline masks used on some 

operations

ThO2 repackaging 403 1964–1965 0.3 13 mo

Fluid bed denitration 403 1965 0.6 3 mo

Fluid bed denitration 403 1965–1966 0.4 No information
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5.5.3 Environmental Monitoring Period, 1975 to 1984

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

over to the 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 U.S. Energy Research and 

Development Administration (and later DOE) contracted with National Lead Company 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 WSP was transferred from the Army to DOE in 

1985.

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

5.5.4 Remediation Period, 1985 to 2002

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 

(ORAUT 2013a).

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

programs were well documented in the revisions of the WSSRAP technical basis manual (DOE 1991, 

1994, 1997, 1998a, 1998b, 1998c, 2000a, 2001b). The air sampling program consisted of area 

samples and BZA samples. Area sample data were not used routinely for dose assessment unless

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 

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 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 (DOE 1994, pp. 45, 49). As of Revision 2, one in four workers in areas with 

concentrations greater than 2% of the DAC were required to wear BZA samplers (DOE 1997, p. 32). 
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The internal dosimetry program required routine monitoring of environmental 222Rn (radon) and 220Rn 

(thoron) and their decay products 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, that 

threshold was never exceeded (DOE 2001b, p. 32). 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 ORAUT-TKBS-0028-4 (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 Operational Period, 1957 to 1966

5.6.1.1 Uranium and Related Intakes

Urine bioassay data are the primary data available to the DR to quantify uranium intake. 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, 95th-percentile, and maximum values of the 

uranium urine data per year for Monday samples and Friday samples by cost center and calendar 

year. 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:

• NU, before 1961; 

• NU, recycled, 1961 to 1962; and

• EU (1%), 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 in Building 101 (feed preparation and sampling), 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. 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., RU) 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).

The mean values in Table 5-20 represent the amount of each isotope that, as a fraction of 230Th, 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 230Th 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 because natural 

thorium wastes were deposited there. The activity concentrations for the uranium decay products and 
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Table 5-20. Activity ratios of uranium decay products and other impurities to 

230Th activity in raffinate pits. 

Isotope

Pit 1 

Mean

Pit 1 

95th %

Pit 2 

Mean

Pit 2 

95th %

Pit 3 

Mean

Pit 3 

95th %

Pit 4 

Mean

Pit 4 

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 Th-232 in Raffinate Pits 1, 2, and 3 were not used due to alpha peak 

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

assuming that Th-228 was in 65% equilibrium, except for Pit 4 where reliable 

measurements were available.

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 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 NU

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 the 

claimant.

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 were 

routine (MCW 1966; see Section 5.5.1.4 and Appendix A). The dust studies are presented in terms of 

DWAs. The data provide a clear picture of when (on a monthly timeframe) and where thorium 

processing and packaging operations occurred. The original air sampling data MCW used to 

generate the DWA value for each job have not been found, so an alternate approach to characterizing 

the variability of the exposure in the workplace is necessary. Using AWE 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). Therefore, for example, in March 1965 the DWA value 

that is 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 
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DWA value is the AM from a lognormal distribution with a GSD of 5. Equation 2 from Strom and 

Stansbury (2000) defines the relationship between GSD and σ: 

σ ln GSD = ( ) (5-2)

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 GM in a lognormal distribution:

(5-3) − = × 2 σ / 2 GM AM e

Table 2-2 in Battelle Team (2007) and Table 1 in Strom and Stansbury (2000) includes the following 

formula for μ the natural logarithm of the GM:

μ = ln(GM ) (5-4)

Equation 5 in Battelle Team (2007) defines the 95th-percentile value of the lognormal distribution:

(5-5) 1 645 95th ( μ. σ ) percentile e + =

Therefore, 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 concentrations 

(calculated using the above equations and the DWA values from Attachment A) are calculated 

assuming a breathing rate of 1.2 m3

/hr 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) since thorium operations began in November 1963, there 

were 250 workdays per each year in 1964 and 1965, and there were 195 workdays in 1966 because 

thorium operations ceased in September 1966. Based on guidance in NIOSH (2004) and ORAUT 

(2012), the ingestion intake rates in Table 5-22 were derived from the inhalation intake rates by 

multiplying the inhalation rate by 0.02.

Table 5-22. Thorium-232 inhalation and ingestion intake rates (Bq/yr). 

Plant name and 

building number

Work group title

and cost center

Inh. 

1963

Inh. 

1964

Inh. 

1965

Inh. 

1966

Ing. 

1963

Ing. 

1964

Ing. 

1965

Ing. 

1966

Sampling–101 Sampling–0402 0 0 125 304 0 0 2 6

Other occupants 0 0 9 22 0 0 0 0

Refinery–103 Digest-Raffinate–0501 196 4,810 5,100 2,320 4 96 102 46

Other occupants 14 341 360 165 0 7 7 3

Extraction–105 Extraction 0 52 0 0 0 1 0 0

Other occupants 0 4 0 0 0 0 0 0

Metals–301 Metals–other than 

Reduction 0702

47 1,010 21,550 1,595 1 20 431 32

Other occupants 3 72 1,525 113 0 1 31 2

Scrap Plant–403 Scrap Plant–0601 645 14,700 2,535 0 13 294 51 0

Other occupants 46 1,040 180 0 1 21 4 0

Undetermined 

location

Any worker not otherwise 

specified above whose 

work location is unknown

46 1,040 1525 165 1 21 31 3

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 MCW

Destrehan Street facility, the predecessor to WSP. An interview with a former worker established that 
Document No. ORAUT-TKBS-0028-5 Revision No. 03 Effective Date: 03/14/2017 Page 44 of 61

 

Work Unit designation at Weldon Spring was reliable (ORAUT 2009). Therefore, the Work Unit (also 

called Work Group or Job Code) designation can be used as an indicator of potential thorium 

exposure.

The 232

Th 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 (as 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 should be assigned intakes based on the 

95th-percentile 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 

should be 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 who might have been exposed but 

whose location cannot be reliably established, for example maintenance or janitorial services, 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/yr from the Scrap Plant. Workers with undetermined work locations would be assigned an 

inhalation intake of 46 Bq/yr 232Th and an ingestion intake of 1 Bq/yr 232Th for 1963. Thorium-228 

should be assumed to be in equilibrium with 232Th, so the activities of 228Th are equal to the 232Th 

activities in Table 5-20. Radium-228, which is a beta emitter and would not have been detected from 

alpha analysis of an air sample, is added at a ratio of 1-to-2 232Th-to228Ra (NIOSH 2010b, 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 favorable to claimants. 

For workers not in these buildings or professions during this period, thorium intakes should be 

assessed as environmental intakes in accordance with ORAUT-TKBS-0028-4 (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 WSP, WSRP, and WSQ are provided 

in Table 4-2 of ORAUT-TKBS-0028-4 (ORAUT 2013b). The maximum site-wide annual median 

exposures to radon decay products, by year, are provided in Table 4-3 of that same document. 

These estimates are based on environmental radon gas releases. The calculation in ORAUT-TKBS0028-4 was based on the assumption that all radon gas from decay of 226Ra was released during 

processing. There are no personal dosimetry data or air concentration measurements available for 

radon in the WSP during the operational period.

5.6.1.3.1 Radon

Weldon Spring Historic Dose Estimate states that the prime source of radon emissions was the acid 

recovery plant (Meshkov et al. 1986, p. 47). The 222Rn activity released annually was estimated to be 

34 Ci/yr (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, 
Document No. ORAUT-TKBS-0028-5 Revision No. 03 Effective Date: 03/14/2017 Page 45 of 61

 

the radon concentration inside the building would have been approximately equal to the radon 

concentration outdoors. The 222Rn equilibrium concentration was calculated by dividing the radon 

influx (222Rn activity/pCi/hour) by the product of the number of air changes per hour and the volume of 

the ventilated area. Using the Building 103 volume of 2.6 × 104

 m3 (2.6 × 107 L) and 1 air change per 

hour, the 222Rn equilibrium concentration was 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, pp. 114–115). Using the number of work hours in 

a month of 170 in defining the working level-month (WLM) and a 2,000-hour working year, the 

estimated annual exposure was calculated to be 12.4 WLM for 100% occupancy for 2,000 hr/yr as 

shown in this equation:

(5-6) 2000 150 0 7 12 4

100 170

WLM pCi l WL M h . . year l pCi h year⋅ = × × × ×=

5.6.1.3.2 Thoron

As described in Section 5.2.3, there are several aspects of the release and buildup of the progeny 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 and firing systems at WSP. Therefore, the evaluation 

of operational exposures to thoron at Weldon Spring is confined to those years. During 1966 (the 

maximum production year), WSP 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 progeny 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, because the gaseous state is more easily captured 

by ventilation systems.

As discussed in Section 5.2.1.6, potential exposure to thoron is mitigated by several factors including 

its very short half-life (with the resulting limitations on the diffusion distance within the material matrix) 

and the effective shielding provided by the thorium processing apparatus. Therefore, 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 5-21, the 50th and 95th percentiles of the DWAs 

were estimated as 21 and 250 alpha dpm/m3

, respectively, over the entire thorium processing period. 

Assuming 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 are conservative estimates for the thoron gas, which is more easily 

captured by 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 progeny were obtained for the 50th and 95th percentiles, respectively. These 

are conservative because 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.
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5.6.1.3.3 Recycled Uranium

Starting in 1961, RU components should be added to uranium intakes. Intakes are assigned by 

applying the ratios in Table 5-23 (ratio to uranium mass) or Table 5-24 (ratio to uranium activity) to the 

calculated uranium intakes. These values are based on Table 5-10 of ORAUT (2016), using only the 

values associated with the 1961–1972 column since all other values are not applicable to WSP. The 

selected material types for RU contaminants are assigned using the direction in ORAUT (2014a). The 

contaminants are added to all uranium intakes regardless of monitoring method (urine, chest counts, 

BZ air samples, etc.).

Table 5-23. RU contaminant intakes per unit mass of uranium (Bq/g U). 

Radionuclide 1961–1962 1963–1965 1966–1975 1976–1985 1986–2002

Uranium 2.53E+04 3.60E+04 3.60E+04 3.60E+04 3.60E+04

Pu–alphaa 2.96E+01 2.96E+01 2.93E+01 2.90E+01 2.89E+01

Pu–241 3.05E+02 3.05E+02 1.88E+02 1.17E+02 7.22E+01

Am–241 1.96E–02 1.96E–02 3.85E+00 6.14E+00 7.51E+00

Np–237 1.02E+01 1.02E+01 1.02E+01 1.02E+01 1.02E+01

Tc–99 3.77E+03 3.77E+03 3.77E+03 3.77E+03 3.77E+03

Th–232 4.07E–02 4.07E–02 4.07E–02 4.07E–02 4.07E–02

Th–228 2.85E–02 2.85E–02 2.85E–02 2.85E–02 2.85E–02

Ru–103/106b 4.08E+03 4.08E+03 4.08E+03 4.08E+03 4.08E+03

Zr–95 6.12E+02 6.12E+02 6.12E+02 6.12E+02 6.12E+02

Nb–95 6.12E+02 6.12E+02 6.12E+02 6.12E+02 6.12E+02

Sr–90 1.63E+02 1.63E+02 1.63E+02 1.63E+02 1.63E+02

a. The plutonium alpha mixture is assessed as 100% Pu-239. Am-241 and Pu-241 are assessed based on 6% weaponsgrade plutonium mixture.

b. Ru-103/106 is assumed to be 100% Ru-106 due to its longer half-life.

Table 5-24. RU contaminant intakes per unit activity of uranium (Bq/Bq U). 

Radionuclide 1961–1962 1963–1965 1966–1975 1976–1985 1986–2002

Uranium 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Pu–alphaa 1.17E–03 8.21E–04 8.14E–04 8.07E–04 8.02E–04

Pu–241 1.21E–02 8.47E–03 5.23E–03 3.24E–03 2.00E–03

Am–241 7.76E–07 5.45E–07 1.07E–04 1.71E–04 2.09E–04

Np–237 4.04E–04 2.84E–04 2.84E–04 2.84E–04 2.84E–04

Tc–99 1.49E–01 1.05E–01 1.05E–01 1.05E–01 1.05E–01

Th–232 1.61E–06 1.13E–06 1.13E–06 1.13E–06 1.13E–06

Th–228 1.13E–06 7.91E–07 7.91E–07 7.91E–07 7.91E–07

Ru–103/106b 1.61E–01 1.13E–01 1.13E–01 1.13E–01 1.13E–01

Zr–95 2.42E–02 1.70E–02 1.70E–02 1.70E–02 1.70E–02

Nb–95 2.42E–02 1.70E–02 1.70E–02 1.70E–02 1.70E–02

Sr–90 6.46E–03 4.53E–03 4.53E–03 4.53E–03 4.53E–03

a. The plutonium alpha mixture is assessed as 100% Pu-239. Am-241 and Pu-241 are assessed based on 6% weaponsgrade plutonium mixture.

b. Ru-103/106 is assumed to be 100% Ru-106 due to its longer half-life.

5.6.2 Environmental Monitoring Period, 1975 to 1984

All intakes in this period should be assessed as environmental intakes in accordance with the 

environmental dose TBD (ORAUT 2013b).
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5.6.3 Remediation Period, 1985 to 2002

Because WSSRAP conducted an extensive, state-of-the-art internal dosimetry program during 

remediation 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 expect a worker’s file to contain a detailed report of the pertinent data used for the 

assessment. It is also reasonable to expect the data will be self-explanatory and can be used by the 

DR as found. If necessary, a detailed description of the internal dosimetry program is available in the 

revisions of the WSSRAP technical basis manuals (DOE 1991, 1994, 1997, 1998a, 1998b, 1998c, 

2000a, 2001b). 

In addition to assessing specific intakes, the DR should assess the environmental intakes in 

accordance with the environmental dose TBD (ORAUT 2013b).

5.7 SUMMARY OF INSTRUCTIONS TO DOSE RECONSTRUCTORS

Table 5-25 contains a summary of the instructions to DRs.

Table 5-25. Instructions to DRs. 

Operational period, 1957 to 1966

Intake Data/Information to be used

U-234, U-235, U-238 Primary: Individual uranium bioassay data.

Supplemental: Workgroup data (Tables 5-8 through 5-17).

Default isotopic assumptions:

• NU, before 1961; 

• NU, recycled, 1961 to 1962; 

• EU (1%), recycled, 1963 to 1966, specific activity 

of 0.973 pCi/μg; and

• RU contaminants (see Section 5.2.4 and 

5.6.1.3.3).

Uranium impurities and decay products 

(less radon)

Building 101, Building 103, Building 105 before

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 to 1984

Intake Data/Information to be used

Any All intakes in this period should be assessed as 

environmental intakes using ORAUT-TKBS-0028-4 

(ORAUT 2013b).

Remediation period, 1985 to 2002

Intake Data/Information to be used

Any The worker’s file will contain a detailed report of the 

pertinent data to be used for the assessment.
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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 

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 228Th, 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. 03 Effective Date: 03/14/2017 Page 49 of 61

 

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Environmental Dose, ORAUT-TKBS-0028-4, Rev. 01, Oak Ridge, Tennessee, May 17.

ORAUT (Oak Ridge Associated Universities Team), 2014a, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 01, Oak Ridge, Tennessee, September 8.

ORAUT (Oak Ridge Associated Universities Team), 2014b, Rocky Flats Plant – Occupational Internal 

Dose, ORAUT-TKBS-0011-5, Rev. 03, Oak Ridge, Tennessee, September 30.

ORAUT (Oak Ridge Associated Universities Team), 2016, Feed Materials Production Center – 

Occupational Internal Dose, ORAUT-TKBS-0017-5, Rev. 02, Oak Ridge, Tennessee, 

September 30.

Strom, D. J., and P. S. Stansbury, 2000, “Determining Parameters of Lognormal Distributions from 

Minimal Information,” American Industrial Hygiene Association Journal, volume 61, pp. 877–

880. [SRDB Ref ID: 13722]
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GLOSSARY

absorption type

Categories for materials according to their rates 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).

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 rad, rep, or grays.

dose equivalent

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. 
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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

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 using 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.
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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 (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.

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 

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 1.30 x 105 MeV of alpha energy; 1 WL 

equals 2.083 × 10-5 joules per cubic meter.
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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 of 1 WL; 1 WLM equals 1 WL times 

170 hours, which is 0.00354 joule-hours per cubic meter.
 

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ATTACHMENT A 

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS

Table A-1. DWA dust concentrations for thorium production jobs (alpha dpm/m3

). 

Sampling Plant, Building 101, extracted Th-natural, 

Sampling work group, cost center 0402

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

Oven drying ThO2 sol-pan transfer Mar 1965 2 N/Aa N/A 3 N/A Test date: Mar 30, 1965. Five ovens 

operated on a 24-hour cycle at 400 

lb/oven. Experimental, intermittent 

operation.

Oven drying ThO2 sol-vacuum 

unload

Aug 1965 2 N/A N/A 16 N/A Test date: Aug 9, 1965. Experimental, 

intermittent operation.

Repackaging ThO2 feeds for 

digestion-drum dumping

Apr–Aug 1966 1 N/A N/A N/A 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 N/A N/A N/A 12 Total project operating time estimated at 1 

week or 10 shifts.

Repackaging ThO2 feeds for 

digestion-outgoing drum conveyor

Apr–Aug 1966 1 N/A N/A N/A 7 Total project operating time estimated at 1 

week or 10 shifts (empty drums).

Refinery, Building 103, extracted and reextracted Th-232 and progeny,

Digest-Raffinate work group, cost center 0501

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

Hopper feed & digestion Apr–May 1966 3 N/A N/A N/A 6 Test date: Apr 22, 1966. One hopper 

(15,000 pounds) made 1 tank batch, 

20,000 lb/d product on first cycle; recycle 

is liquid feed. 

Raffinate treatment & disposal May–Sep 1966 3 N/A N/A N/A 2 None.

Misc. digestion of drummed scrapwet feedMay–Sep 1966 2 N/A N/A N/A 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 (continued)

Document No. ORAUT-TKBS-0028-5 Revision No. 03 Effective Date: 03/14/2017 Page 58 61

Refinery Pot Room, Building 103, extracted Th-natural, 

Pot Room work group, cost center 0502e

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

Pot denitration & sol drying-hand 

transfers

Nov–Dec 1963; 

May–Aug 1964

6 88 88 N/A N/A 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 N/A 60 N/A N/A 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 N/A N/A 48 N/A Airline masks prescribed for packaging 

and unloading; half-face 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 N/A N/A 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 N/A N/A 144 N/A 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.

Pot denitration & sol drying-hand 

transfers

Nov–Dec 1963; 

May–Aug 1964

6 88 88 N/A N/A Test dates: Nov 1963 and May 1964. 

Airline masks used to enter pot sections.

ThO2 repackaging-multi-purpose 

hood

Aug 1964 2 N/A 260 N/A N/A Short-term operation. Airline masks used 

in hood.

DWA index for the Pot room work 

group

Apr 1964–Dec 

1965

N/A N/A 78 50 N/A 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.
ATTACHMENT A

DUST CONCENTRATIONS FOR THORIUM PRODUCTION JOBS (continued)

Document No. ORAUT-TKBS-0028-5 Revision No. 03 Effective Date: 03/14/2017 Page 59 61

Extraction Plant, Building 105, extracted Th and Th-232 progeny, 

Extraction work group, no cost center code

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

Pumper decanter, pulse column, 

strippers, and NOK (product 

liquor)

Jun–Sep 1966 9 N/A N/A N/A 2 None

Metals Plant, Building 301, extracted Th-natural, 

Metals-other than Reduction work group, cost center 0702

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

High-firing 1,000 lb crucibles at 

recast (furnace)

Nov–Dec 1963; 

May–Sep 1964

6 21 21 N/A N/A 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.

TNT repackaging Oct–Dec 1964 2 N/A 3 N/A N/A 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 N/A 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 N/A N/A 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 (continued)

Document No. ORAUT-TKBS-0028-5 Revision No. 03 Effective Date: 03/14/2017 Page 60 61

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

Kiln calcining of NLO sump cake Apr–Jul 1966 3 N/A N/A N/A 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

N/A N/A 17 113 N/A 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.

Scrap Plant, Building 403

extracted Th-natural, Scrap Plant work group, cost center code 0601f

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

TNT crystals-repackaging Nov–Dec 1963;

May–Sep 1964

6 15 15 N/A N/A Equivalent production rate of 1 ton of 

ThO2/day.

ThO2 repackaging-from recast 

crucibles

Nov–Dec 1963;

May–Sep 1964

6–8 290 290 N/A N/A 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 auto-airmat (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 N/A 20 20 N/A 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 (continued)

Document No. ORAUT-TKBS-0028-5 Revision No. 03 Effective Date: 03/14/2017 Page 61 61

Job descriptionc Periodd

Workers per 

day 1963 1964 1965 1966 Notes

Fluid bed denitration-product to 

drums

May–Jul 1965 6 N/A N/A 44 N/A 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 N/A N/A 29 29 Liquor feed and sol product at 1 ton/day 

production rate.

DWA index for Scrap Plant work 

group

Nov 1963–Dec 

1965

N/A 180 117 28 N/A 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. N/A = not applicable.

b. 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. 

c. Job Title is the term used in the referenced report; this term more accurately represents a task description rather than an employee job title. 

d. Time Period was recorded from the air sampling data sheets and was estimated if date ranges were incomplete. 

e. Manpower in this group ranged from 18 to 21 workers during 1963 through 1965; base production rate was 1 ton ThO2 per day. 

f. Average production rate was 1 ton ThO2 per day.