ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 43

Document Title:

Weldon Spring Plant – Occupational Internal 

Dose

Document Number: ORAUT-TKBS-0028-5

Revision: 00

Effective Date: 06/28/2005

Type of Document: TBD

Supersedes: None

Subject Experts: Janet A. Johnson and Roger B. Falk

Document Owner

Approval: Signature on File Approval Date: 06/21/2005

Robert Meyer, TBD Team Leader

Approval: Signature on File Approval Date: 06/21/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 06/21/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 06/28/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

PLEASE DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 2 of 43

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

03/24/2005 00-A New technical basis document for the Weldon Spring Plant –

Occupational Internal Dose. Initiated by Robert Meyer.

04/28/2005 00-B Incorporates internal review comments. Initiated by Robert Meyer.

06/15/2005 00-C Incorporates OCAS review comments. Initiated by Robert Meyer.

06/28/2005 00 First approved issue. Training is not required. Initiated by Robert 

Meyer.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 3 of 43

TABLE OF CONTENTS

Section

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

Page

2

Acronyms and Abbreviations ...............................................................................................................6

5.1 Introduction.................................................................................................................................8

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

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

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

5.1.4 Radiation Protection Practices..........................................................................................11

5.2 Source Term.............................................................................................................................12

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

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

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

5.2.4 Recycled Uranium............................................................................................................15

5.2.5 Solubility Classification and Absorption Type ...................................................................15

5.2.6 Particle Size .....................................................................................................................16

5.3 In Vitro Measurements.............................................................................................................. 16

5.3.1 Operational Period (1957 to 1967) ...................................................................................16

5.3.1.1 Uranium...................................................................................................................16

5.3.1.1.1 Routine Sampling Program ............................................................................. 17

5.3.1.1.2 Special Sampling Program.............................................................................. 17

5.3.1.1.3 Data Reporting Levels and Minimum Detectable Amounts.............................. 17

5.3.1.1.4 Interferences and Uncertainties ...................................................................... 18

5.3.1.1.5 Reporting Formats and Codes ........................................................................ 18

5.3.1.1.6 Work Group Data............................................................................................ 20

5.3.1.2 Thorium ...................................................................................................................27

5.3.2 Environmental Monitoring Period (1975 to 1984)..............................................................28

5.3.3 Remediation Period (1985 to 2001)..................................................................................28

5.4 In Vivo Measurements .............................................................................................................. 28

5.4.1 Operational Period (1957 to 1967) ...................................................................................28

5.4.2 Environmental Monitoring Period (1975 to 1984)..............................................................29

5.4.3 Remediation Period (1985 to 2001)..................................................................................29

5.5 Air Concentration and Dust Measurements............................................................................... 30

5.5.1 Operational Period (1957 to 1967) ...................................................................................30

5.5.1.1 Maximum Acceptable Concentrations...................................................................... 30

5.5.1.2 Special Curie for Uranium........................................................................................ 30

5.5.1.3 Dust Exposure Calculation....................................................................................... 31

5.5.1.4 Dust Studies ............................................................................................................ 31

5.5.2 Transfer to the U.S. Army (1968 to 1974).........................................................................33

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

5.5.4 Remediation Period (1985 to 2001)..................................................................................33

5.6 Assessment of Intakes.............................................................................................................. 34
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 4 of 43

5.6.1 Operational Period (1957 to 1967) ...................................................................................34

5.6.1.1 Uranium Intakes ...................................................................................................... 34

5.6.1.2 Thorium Intakes....................................................................................................... 35

5.6.1.3 Intakes of Other Radionuclides................................................................................ 36

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

5.6.3 Remediation Period (1985 to 2001)..................................................................................37

References ........................................................................................................................................38

Glossary ............................................................................................................................................41
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 5 of 43

LIST OF TABLES

Table

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

Page

10

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

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

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

5-5 Isotopic uranium data for 1% enriched uranium ........................................................................ 13

5-6 Solubility class summary........................................................................................................... 15

5-7 Cost center codes for workers................................................................................................... 19

5-8 Composite uranium urine data summary................................................................................... 21

5-9 Composite uranium urine data summary................................................................................. 212

5-10 Uranium urine data summary by cost center for 1959 ............................................................... 23

5-11 Uranium urine data summary by cost center for 1960 ............................................................... 24

5-12 Uranium urine data summary by cost center for 1961 ............................................................... 25

5-13 Uranium urine data summary by cost center for 1962 ............................................................... 25

5-14 Uranium urine data summary by cost center for 1963 ............................................................... 26

5-15 Uranium urine data summary by cost center for 1964 ............................................................... 26

5-16 Uranium urine data summary by cost center for 1965 ............................................................... 27

5-17 Uranium urine data summary by cost center for 1966 ............................................................... 27

5-18 Work group time-weighted average exposure for uranium operations....................................... 32

5-19 Work group time-weighted average exposure for thorium operations........................................ 32

5-20 Time-weighted average thorium dust concentrations ................................................................ 33
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 6 of 43

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

Bq becquerel

BZA breathing-zone air

CEDE committed effective dose equivalent

cm centimeter

d day

d/s disintegrations per second

DAC derived air concentration

dpm disintegrations per minute

DR dose reconstructor

DOE U.S. Department of Energy

DWA daily weighted average

EEOICPA Energy Employees Occupational Illness Compensation Program Act

g gram

hr hour

ICRP International Commission on Radiological Protection

IMBA Interactive Modules for Bioassay Analysis

kg kilogram

L liter

m meter

MAC maximum allowable concentration

MCW Mallinckrodt Chemical Works

MDC minimum detectable concentrations

min minute

ml milliliter

MPC maximum permissible concentration

mrem millirem

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

pCi picocurie

ppb parts per billion

U.S.C. United States Code

WL working level

WLM working level month
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 7 of 43

WSP Weldon Spring Plant

WSCP Weldon Spring Chemical Plant

WSQ Weldon Spring Quarry

WSRP Weldon Spring Raffinate Pits

WSSRAP Weldon Spring Site Remedial Action Project

yr year

μCi microcurie

μg microgram

μm micrometer
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 8 of 43

5.1 INTRODUCTION

Technical Basis Documents (TBDs) are general working documents that provide guidance concerning 

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

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

used to assist the National Institute for Occupational Safety and Health in the completion of the 

individual work required for each dose reconstruction.

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

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

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

Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S. C. § 7384l (5) and (12)].

5.1.1

The purpose of this TBD is to provide a profile of internal dosimetry practices at the Weldon Spring 

Plant (WSP). This document contains technical information on the history, methods, and 

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

Purpose

5.1.2

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

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

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

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

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

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

Scope

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

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

natural thorium, this TBD includes material on 232Th and its decay products (

228Ra and 228Th). 

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

for internal radiation dose are the isotopes of uranium.

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

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

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

5.1.3

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

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

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

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

(MCW 1964; Mason c. 1977). 

Background on Mallinckrodt Chemical Works and the Weldon Spring Plant

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

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

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

tolerance level.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 9 of 43

During the 24 yr of operations, the MCW Uranium Division processed uranium ores and concentrates 

into uranium trioxide (UO3), uranium hexafluoride (UF6), and uranium metal (Meshkov 1986). 

Uranium dioxide (UO2) was an intermediate product in the conversion of UO3 to UF4 (ORAU 2005b). 

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

ore that includes the decay products generally in equilibrium). The WSP facility primarily handled

uranium concentrates (e.g., yellowcake) received from uranium-milling facilities with very low 

concentrations of decay products. The historical documents report that WSP did not receive the highquality pitchblende ores (Ingle 1991). The WSP processed thorium during the 2 yr before closure in 

1966 and perhaps during earlier years as well (see Section 5.6).

Production at WSP waned during the later part of its operational period. In 1964, MCW published a 

document intended to encourage diversification and utilization of the existing facilities (MCW 1964;

Mason c. 1977), which reflected that decline. The document described the plant facilities and briefly 

mentioned the operational safety program.

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

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

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

including beta doses, are covered in Technical Basis Document for the Weldon Spring Plant –

Occupational External Dosimetry (ORAU 2005a). This section covers internal deposition of uranium. 

There were four periods in the history of the WSP:

• Operational: 1957 to 1967

• Transfer to the U.S. Army: 1968 to 1974

• Environmental monitoring: 1975 to 1985

• Remediation: 1985 to 2001

The operational period and the remediation period are of primary concern in relation to internal 

dosimetry at WSP, with some potential applicability of operations in 1967 before transfer to the 

U.S. Department of Defense. EEOICPA does not cover the period when the WSP site was under 

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

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

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

Uranium was initially considered primarily a heavy-metal poison, and the radioactivity level was 

considered low enough so that “small scale, short-term operations would not present a radiation 

problem.” MCW’s health and safety program was based on standard industry procedures for handling 

toxic chemicals. The MCW radiation safety program developed for the Destrehan Street site was 

applied at the WSP site when the transition between the two facilities was made.

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

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

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

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

Basis Document for Weldon Spring Plant – Site Description (ORAU 2005b). Ore concentrates 

containing 60% to 70% yellowcake were sampled on receipt at the facility. Some of the material was 

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

nitric acid to produce uranium-nitrate solution that was then purified by solvent extraction and 

denitrated to produce UO3.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 10 of 43

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

a

Period Activity

1957-1966 Operation of the uranium feed materials plant

• Uranium concentrates converted to uranium trioxide, uranium tetrafluoride, and uranium metal

• Some thorium processing between 1965 and 1966

• Raffinate from processing removed to raffinate pits

• 14,500 metric tons of uranium materials received for processing and sampling per year between 1958 

and 1964

December 

1966

Plant closed

• Hopper and process lines emptied

• Dust collectors cleaned out

January 

1967-?

Site used as interim storage depot for yellowcake later shipped to other plants for refining and processing. 

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

March 1968 Army started decontamination and equipment removal

December 

1968

Construction of herbicide facility began; project terminated in early 1969 before renovation was complete. 

March 1968-

June 1969

Decontamination and equipment removal for Buildings 103 and 105 (see table 5.2 for building operations 

description)

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

• About 2,000 metric tons scrap moved to Tennessee

• About 200 metric tons steel parts moved to Ohio

• About 100 metric tons uranium oxide removed from the buildings

1969-1985 Site remained essentially undisturbed

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

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

a. Adapted from Meshkov (1986) and Lesperance, Siegel, and McKinney (1992).

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

a

Bldg. 

no.

Building 

description Building operations Potential radionuclide exposures

101 Sampling Sampling of ore concentrates containing 60 

– 70% yellowcake

Some material repackaged in drums – some 

sent to Bldg 103 for processing

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

Rn-222 and its short-lived decay products

Th-232 and decay products in 1966

103 Digestion Materials digested with nitric acid. Uranium 

bearing solution sent to Bldg 105 for 

purification. Materials returned for 

denitration after purification and sent to Bldg 

201. 

U-nat dust (as yellowcake and UO3), Ra-226, Th-230, Po210 and Pb-210

Rn-222 and its short-lived decay products

Th-232 and decay products starting November 1963

105 Purification Materials were purified by solvent extraction 

and returned to Bldg 103 for denitration.

Wet process but some potential for uranium or thorium (in 

1966) dust exposure. Uranium would have been the 

major internal exposure component but Rn-222 and decay 

product exposure possible

108 Acid recovery Recovering and re-concentrating nitric acid Radon gas and its decay products

201 Green salt 

(UF4) plant 

Feed from Bldg 103 (after denitration) 

converted to UF4.

Potential for uranium exposure as green salt dust or 

natural thorium (1965-66). No significant Th-230, 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 uranium exposure as green salt dust and 

U3O8 or natural thorium (1965-66). No significant Th-230, 

Ra-226 or decay product exposure.

Th-232 and decay products starting November 1963

403 Chemical pilot 

plant

Small-scale chemical processes U-nat dust, Ra-226, Th-230, Po-210 and Pb-210

Rn-222 and its short-lived decay products

Th-232 and decay products starting November 1963

404 Metallurgical 

pilot plant

Small-scale metallurgical processes Potential for uranium exposure as green salt dust and 

U3O8 or natural thorium (1965-66) No significant Th-230, 

Ra-226 or decay product exposure.

407 Analytical and 

research labs

Small-scale research and analytical work on 

products and processes.

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

Rn-222 and its short-lived decay products

Th-232 and decay products from 1965 to 1966

a. Adapted from ORAU (2005b).
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 11 of 43

Table 5-3. Products or intermediates in production 

buildings.

Building no. Products or intermediates present

101 Ore concentrates, yellowcake

103 Yellowcake, UO3, 

Uranium solution in nitric acid

Purified uranium solution

Denitrated uranium solution

105 Uranium solution (UO3) in nitric acid

Purified uranium solution

108 Recovered nitric acid

201 Purified, denitrated uranium solution

UO3, UO2 (intermediate compound)

Green salt (UF4)

301 Green salt (UF4)

Uranium metal

U3O8 

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

of the plant. 

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) 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 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) stated the judgment that the uranium doses of the MCW population were 

considerably larger than doses from exposure to radon. 

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

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

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

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

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

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

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

5.1.4

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

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

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

hazard. 

Radiation Protection Practices

As noted in ORAU (2005a), the WSP employed approximately 600 workers during full production, of 

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

uranium, depleted uranium and, for a short period, natural thorium. The principal radionuclides to 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 12 of 43

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

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

decay products were not present in significant quantities.

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

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

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

likely contaminants in air in those locations. 

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

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

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

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

exposure to be higher than AEC limits.” However, MCW (1965) 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.

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

1984) have been found. 

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

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

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

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

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

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

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

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

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

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

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

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

bioassay, in vivo lung counting, and air sampling. Section 5.3 describes the urine and fecal bioassay 

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

practices. 

5.2 SOURCE TERM

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

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

uranium and slightly enriched (up to 1%) uranium as well as natural thorium. Table 2-4 in ORAU 

(2005a) 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.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 13 of 43

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

Material Mass (kg) Percent of total mass

Natural U 122,015,944 98.43

Depleted U 167,823 0.14

Slightly enriched U 842,585 0.68

Natural thorium 941,347 0.76

Total 123,967,712

5.2.1

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

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

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

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

the default values in the IMBA program for natural and depleted uranium.

Isotopic Composition of Uranium

For slightly enriched uranium, it is reasonable to assume that the composition of 1% enriched uranium 

in the Technical Basis Document for the Fernald Environment Management Project – Occupational 

Internal Dosimetry (ORAU 2004a, Table 5-3) is applicable to slightly enriched uranium at WSP. 

Table 5-5 of this document shows those values.

Table 5-5. Isotopic uranium data for 1% enriched uranium.

Isotopes Mass %

Relative 

activity

(pCi/μg) Activity %

Total

pCi/μg

U-234 0.01 0.62 63.7 0.973

U-235 1.0 0.02 2.1

U-236a 0 0 0

U-238 98.99 0.333 34.2

a. U-236 contributes less than 1% activity in 1% enriched uranium.

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

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

5.2.2

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

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

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

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

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

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

internal dose. 

Uranium Decay Products

The buildup of 230Th from 234U is of little consequence because it has an 80,000-yr half-life. Assuming 

the age of the material processed to be no greater than 20 yr, the maximum activity concentration of 

230Th due to decay of 234U would be 0.02% of the 234U activity concentration. Small amounts of 

enriched uranium were apparently processed at the WSP site. That would increase the concentration 

of 234U in natural uranium by a significant amount. Assuming that increase could be as much as a 

factor of 100, the 230Th concentration due to buildup would still be only a very small fraction of the 

uranium concentration. The concentration of 226Ra due to buildup would be even smaller because it 

has a 1,600-yr half-life and would not have had time to build up significantly from 230Th.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 14 of 43

The primary source of decay products (

230Th and 226Ra) for the materials processed at WSP would, on 

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

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

Impact Statement on Uranium Milling (FGEIS) 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). However, later measurements 

indicated that these values, especially for 230Th, were overly conservative. The values assumed for 

the “model mill” that was the subject of the FGEIS, were 0.5% for 230Th and 0.1% for 226Ra. The 

Technical Basis Document for the Weldon Spring Plant – Occupational Environmental Dose (ORAU 

2005c) cites values of 5% for 230Th and 1% for 226Ra based on information from Meshkov et al. (1986). 

The values from ORAU 2005c are claimant-favorable, but very conservative. The shorter lived decay 

products, 210Pb and 210Po should be assumed to be present at the same activity as the 226Ra in 

uranium concentrates.

As noted in Technical Basis Document for the Weldon Spring Plant – Site Description (ORAU 2005b),

four raffinate pits were constructed between 1958 and 1964 to contain process wastes from the WSP. 

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

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

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

maximum activity concentrations provides a claimant-favorable basis for estimating intake. The 

maximum 230Th activity concentration in Pit 1 was the highest at 34000 pCi/g. The estimated 

maximum activity concentrations in Pit 1 for 238U, 226Ra, and 232Th were 600 pCi/g, 1700 pCi/g, and 

110 pCi/g, respectively. The maximum activity concentration of 232Th in Pit 4 was the highest with a 

maximum value of 1400 pCi/g (DOE 1991). These concentrations provide a bounding estimate that 

can be used to calculate claimant-favorable intakes for exposure at the WSRP.

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 to the WSQ between 1959 and 1966. The estimated maximum radionuclide 

concentrations in the WSQ materials were 2400 pCi/g, 6800 pCi/g, and 2780 pCi/g for 238U, 230Th, and 

226Ra respectively (DOE 1991). While the average concentrations were generally lower by at least an 

order of magnitude, the maximum concentrations provide a bounding estimate that can be used to 

calculate claimant-favorable intakes from exposure at the WSQ.

5.2.3

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

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

time since separation, the 228Ra (6.7-yr half-life) and 228Th (1.9-yr half-life) would be significant if not 

complete. Radium-228 is a beta emitter that decays to 228Th through 228Ac. Thorium-228 decays by 

alpha emission to 224Ra (3.66-d half-life). Therefore, the claimant-favorable assumption for assessing 

dose from 232Th is that all decay products should be assumed to be in equilibrium. The Technical 

Basis Document for the Weldon Spring Plant – Occupational Environmental Dose (ORAU 2005c)

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

operational period at an activity less than 1% of the natural uranium (ORAU 2005c)

Natural Thorium and Decay Products

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

230Th and 226Ra concentrations, thus would contribute only a very small fraction of the potential dose. 

In contrast, the estimated maximum 232Th concentration in materials deposited in the WSQ was 4200 

pCi/g (DOE 1991). This concentration represents a claimant-favorable bounding estimate for 

potential exposures to natural thorium at the WSQ.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 15 of 43

5.2.4

The extent of the processing of recycled uranium at WSP is not well known (ORAU 2005b, 

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

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

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

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

Recycled Uranium

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

(assume 239Pu), neptunium (237Np), and technetium (99Tc). Site-specific data for the mass fractions of 

these contaminants are not available. The DR should consider the factors in Table 5-11 of Technical 

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

(ORAU 2004a). These factors, when multiplied by the assessed uranium gram-value intake, result in 

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

and 9,000 ppb 99Tc.

5.2.5

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

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

worker’s uranium intake. 

Solubility Classification and Absorption Type

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

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

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

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

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

and other ions” depending on the drying temperature (NRC 1980). 

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

with the highest levels of contamination (DOE 2001). The samples were tested for lung solubility 

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

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

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

ICRP Publication 30 designations of D, W, and Y (ICRP 1979). These classes can be assumed to the 

current ICRP solubility classes F, M, and S, respectively (ICRP 1995).

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

Area

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

D W Y D W Y D W Y W Y W Y W Y

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

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

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

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

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

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

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

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

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

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

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

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

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

Q-Sump 0.61 0.52 0.47 0.48 0.58 - 0.42 0.10 0.90 - 1.0 0.10 0.90
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 16 of 43

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

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

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

approximately 90% class Y and 10% class D. The uranium solubility classes for Buildings 406, 408, 

410, and 417 were listed as 100% Class D. These buildings were used for administrative and 

maintenance purposes, and a single composite sample was analyzed to represent the solubility class 

for all non-production buildings. Given the uncertainty and lack of specific data for such areas of the 

WSCP, the most claimant-favorable solubility type should be used. 

The urine bioassay data distribution shows that Friday afternoon samples averaged nearly two times 

the uranium concentration as Monday morning samples. This indicates relatively rapid clearance of 

the previous week’s intake. This lends some support to the assumption that at least a portion of the 

uranium in air at WSP was soluble.

In general, the solubility type to be selected in dose reconstruction for UF4 exposures should be based 

on the cancer site. 

5.2.6

The deposition in the lung of dust particles containing radioactive materials depends on particle size 

and breathing rate. The dose to the lung and other organs from radionuclides in the body depends on 

the solubility of the deposited particles. The excretion rate through the kidneys depends on both 

solubility and particle size. 

Particle Size

DOE assumed a particle size of 1 μm aerodynamic median activity diameter (AMAD) in developing 

derived air concentrations (DACs) for the WSSRAP. ICRP Publication 66 recommends a default 

value of 5 μm AMAD for workplace exposures (ICRP 1994b, paragraph 181). Eisenbud (1975) 

suggests a likely particle size of 5 to 6 μm AMAD based on a mass median diameter of 2 μm and a 

particle density of 9 to 10 g/cm3

. Sanders (1975) studied particle sizes at the Oak Ridge Y-12 plant 

and found that the particle sizes in various locations in the facility ranged from 1 to 3 μm mass median 

diameter, with the bulk of the samples between 2 and 3 μm. The MILDOS program developed in 

1981 to estimate doses to members of the public from particulate emissions from uranium mills uses a 

particle diameter of 1 μm and a density of 8.9 g/cm3 for emissions from yellowcake processing (NRC 

1981). 

Lacking specific information on the particle size for airborne uranium at WSP, use of 5 μm AMAD 

appears to be reasonable. That assumption is supported by the literature. 

5.3 IN VITRO MEASUREMENTS

5.3.1

5.3.1.1 Uranium

Operational Period (1957 to 1967)

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. 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 17 of 43

5.3.1.1.1 Routine Sampling Program

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

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

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

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

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

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

analyzed separately and entered in his summary.

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

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

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

persons are scheduled less frequently to provide a control base.

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

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

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

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

above appears to have been implemented and continued through 1966.

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

5.3.1.1.2 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.

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

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

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

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

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

surrogate 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 (ORAU 2004b) and was based on modern MDA concepts (HPS 1996). The Rocky Flats 

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

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

the surrogate MDA for WSP, although there may be some differences in methods (e.g., grab sample 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 18 of 43

versus 24-hr sample, unknown blank subtraction versus blank subtraction, and unknown volume of 

the aliquot versus 100λ (100 microliters)).

An estimate of the uranium MDA can be derived from Dupree (1979b), 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 surrogate MDA for WSP uranium urine data. 

5.3.1.1.4 Interferences and Uncertainties

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

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

the Dupree (1979b) value of 0.002 ± 0.002 mg/L made during the remediation period on persons in 

the WSP vicinity not occupationally exposed to uranium indicated a geometric mean of 0.05205 μg/L 

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

deviations (e.g., 94.5% of the population) (DOE 2001).” Because these values are very low, DRs

should disregard this source of interference and use the WSP uranium urine data as recorded.

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

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

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

generic to fluorimetric 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)

• 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.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 19 of 43

Table 5-7. Cost center codes for workers. Work group Cost center code

To early 1963 Starting 1963

General Operations

 General Engineering 110 (a)

 Chemical Technology 120 (a)

 Metal Technology 121 (a)

 Plant Services 140 (a)

WSP

 Administration 300 (a)

Production Facilities

 Sampling Plant

 Sampling 310 110

 Repackaging 321 110

 Refinery

 Digest – Raffinate 320 120

 Pot Room 320 120

 Extraction 320 120

 Green Salt Plant

 Green Salt 330 150

 Metal Plant

 Dingot Reduction 360, 340 180

 Extrusion, General 359 200

 Extrusion, Gamma (Dingot Extrusion) 361, 351 210

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

 Scrap Activity 358 280

 Vacuum Outgassing 363 (a)

 Core Fabrication 362, 365 to 369 250

Services Groups or Facilities

 Maintenance 370 370

 Storeroom, Receiving, & Shipping 372 530

 Disposal of Construction Inventory 373 (a)

 Instrument Shop 374 (a)

 Boiler House 376 (a)

 Water Plant 378 (a)

 Warehouse 380 550

 Engineering 390 560

 Plant Protection 392 510

 Laundry 394 (a)

 Custodians 396 520

 Occupancy, Labor 398 (a)

Technical Division

 Administrative 400 (b)

 Metallurgical Development (Building 404) 410 (b)

 Special Projects 460 460

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

 Scrap Plant No Code 290

 Research Laboratory 430 (b)

 Analytical Laboratory 440 (b)

Administration Division

 Uranium Division Management 600 600

 Salaried Personnel Administration 605 (b)

 Accounting 610 610

 Purchasing 620 (b)

 Motor Pool 621 (b)

 Administrative Services 630 (b)

 Production Control 640 640

 Health and Safety 650 (b)

 General Division Expenses 660 (b)

 Data Processing 670 670

 Industrial Relations 680 (b)

 Cafeteria 685 (b)

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

b. Unknown – likely the same as the code to early 1963
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 20 of 43

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

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

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

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

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

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

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

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

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

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

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

5.3.1.1.6 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. In cases 

where there were five or fewer records, the tables contain only the maximum urine bioassay result.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 21 of 43

Table 5-8. Composite uranium urine data summary.

Year Type of analysis

Number of 

records

Urine uranium concentration (mg/L)

Median 95th percentile Maximum

1958 All 1,872 0.005 0.025 0.203a

Routine 1,714 0.005 0.025 0.203

Routine - Monday 321 0.004 0.021 0.047

Routine - Friday 158 0.006 0.023 0.069

1959 All 2,285 0.005 0.024 0.339

Routine 2,090 0.006 0.026 0.339

Routine - Monday 1,124 0.004 0.017 0.041

Routine - Friday 522 0.009 0.048 0.339

1960 All 4,396 0.012 0.038 0.759

Routine 4,246 0.012 0.040 0.759

Routine - Monday 2,602 0.010 0.026 0.088

Routine - Friday 1,347 0.018 0.068 0.759

1961 All 4,184 0.011 0.036 0.344

Routine 4,077 0.011 0.036 0.344

Routine - Monday 2,608 0.010 0.025 0.062

Routine - Friday 1,279 0.018 0.050 0.344

1962 All 3,083 0.010 0.032 0.700

Routine 2,954 0.007 0.024 0.700

Routine - Monday 1,847 0.008 0.020 0.064

Routine - Friday 946 0.013 0.044 0.700

1963 All 3,481 0.014 0.042 0.340

Routine 3,358 0.014 0.041 0.336

Routine - Monday 1,922 0.014 0.029 0.336

Routine - Friday 1,074 0.018 0.056 0.258

1964 All 3,476 0.012 0.052 0.626

Routine 3,264 0.012 0.048 0.626

Routine - Monday 1,767 0.012 0.029 0.282

Routine – Friday 1,049 0.014 0.060 0.626

1965 All 2,980 0.010 0.036 0.812

Routine 2,804 0.009 0.032 0.812

Routine - Monday 1,460 0.009 0.027 0.347

Routine – Friday 869 0.010 0.045 0.318

1966 All 2,145 0.006 0.029 0.459

Routine 1,680 0.007 0.028 0.459

Routine - Monday 882 0.008 0.025 0.066

Routine – Friday 526 0.007 0.039 0.459

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

affect the median or 95th-percentile value.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 22 of 43

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

Cost center

All samples (mg/L) No. of 

Median 95th percentile Maximum records

300-370 0.002 0.010 0.012 16

400-440 0.000 0.011 0.011 7

600-670 0.003 0.010 0.012 29

Acct 0.003 0.006 0.010 17

Adm 0.004 0.015 0.017 16

Aec 0.003 0.010 0.010 9

Anal Lab 0.004 0.012 0.039 105

Boiler 0.002 0.008 0.008 14

Chem Op 0.004 0.030 0.061 24

Decon 0.003 0.006 0.006 9

Dx 0.012 0.029 0.040 25

Engineering 0.003 0.014 0.020 33

Furnace 0.008 0.019 0.022 24

Gr Salt 0.005 0.018 0.025 30

Guard 0.004 0.013 0.014 41

H&S 0.004 0.007 0.016 28

Instruments 0.002 0.018 0.022 37

Lab 0.004 0.018 0.018 27

Laundry 0.005 0.012 0.024 15

Maint 0.006 0.020 0.098 79

Metals 0.006 0.015 0.101 94

Mtns 0.006 0.025 0.064 137

Office 0.002 0.008 0.016 97

Pilot Plant 0.005 0.014 0.028 16

Pl4 0.010 0.026 0.078 24

Pl6 0.012 0.045 0.048 58a

Pl7 0.016 0.053 0.062 54

Porter 0.005 0.016 0.020 25

Prod Dev. 0.004 0.010 0.011 25

Refinery 0.009 0.041 0.203 127

Research Lab 0.004 0.012 0.022 49

Sampling 0.011 0.040 0.049 31

Stores 0.004 0.009 0.029 16

Warehouse 0.004 0.017 0.051 48

One outlier at 7.8 mg/L was not included.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 23 of 43

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

Cost center

Monday samples (mg/L) Number of 

records

Friday samples (mg/L) Number of 

Median 95th percentile Maximum Median 95th percentile Maximum records

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

300-310 0.002 0.015 0.022 35 0.020 0.048 0.048 12

320 0.006 0.016 0.023 24 0.014 0.088 0.104 40

321-340 0.008 0.018 0.020 25 0.017 0.043 0.067 18

359-369 0.005 0.016 0.034 60 0.012 0.036 0.049 47

370 0.009 0.016 0.019 40 0.014 0.058 0.206 25

372 0.003 0.012 0.017 17 0.004 0.023 0.023 8

374 0.006 0.015 0.018 27 0.006 0.024 0.024 13

376-379 0.002 0.027 0.039 26 0.003 0.035 0.035 13

380 0.004 0.019 0.041 53 0.004 0.071 0.094 32

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

392 0.005 0.016 0.028 38 0.003 0.023 0.026 18

394-396 0.006 0.019 0.019 30 0.006 0.023 0.023 14

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

440 0.003 0.014 0.020 73 0.004 0.012 0.015 16

450 0.004 0.012 0.016 20 0.018 0.079 0.080 15

460 0.005 0.017 0.020 21 0.038 0.156 0.156 14

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

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

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

320 Ref 0.005 0.018 0.028 26 0.020 0.057 0.059 22

330 Gr Salt 0.012 0.020 0.020 20 0.017 0.023 0.023 11

370 Misc 0.006 0.015 0.019 16 0.014 0.051 0.051 12

370 Mtns 0.006 0.019 0.035 164 0.011 0.048 0.083 94

370 Mtns Misc 0.006 0.012 0.014 42 0.011 0.025 0.044 24

Office 0.002 0.009 0.018 17 0.006 0.009 0.009 8

Other 0.003 0.012 0.015 66 0.004 0.022 0.022 28

a. N/A = not applicable (fewer than five records)
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 24 of 43

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

Cost center

Monday samples (mg/L) Number of

records

Friday samples (mg/L) Number of

Median 95th percentile Maximum Median 95th percentile Maximum records

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

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

310 0.015 0.026 0.038 39 0.033 0.071 0.085 29

320 0.012 0.026 0.060 204 0.019 0.062 0.110 129

321-340 0.014 0.025 0.032 76 0.026 0.055 0.080 57

359 0.013 0.022 0.029 41 0.025 0.069 0.161 22

360-369 0.013 0.028 0.061 105 0.016 0.058 0.092 68

370 0.012 0.028 0.064 106 0.015 0.053 0.162 58

378-379 0.012 0.027 0.027 14 0.015 0.082 0.082 7

380-398 0.008 0.020 0.028 120 0.012 0.041 0.088 34

400-430 0.007 0.030 0.071 38 0.018 0.050 0.050 9

450 0.010 0.027 0.058 91 0.027 0.092 0.161 39

460 0.013 0.029 0.060 64 0.026 0.060 0.078 37

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

320 Ref 0.012 0.029 0.079 155 0.032 0.126 0.216 107

330 Gr Salt 0.013 0.029 0.030 43 0.022 0.070 0.092 29

360 Metals 0.013 0.026 0.029 29 0.017 0.036 0.058 24

361 Conv Ext 0.021 0.058 0.058 14 0.036 0.075 0.075 7

362 Cons Fab 0.010 0.021 0.029 18 0.014 0.023 0.030 16

370 Misc 0.010 0.025 0.037 63 0.014 0.064 0.084 36

370 Mtns 0.010 0.026 0.038 375 0.017 0.064 0.228 256

370 Mtns Elec 0.010 0.024 0.028 52 0.013 0.063 0.076 34

370 Mtns Mech 0.010 0.014 0.014 11 0.008 0.012 0.012 6

370 Mtns Misc 0.012 0.028 0.036 42 0.017 0.036 0.038 31

370 Mtns Pipe 0.012 0.030 0.060 31 0.018 0.064 0.069 20

370 Painter 0.015 0.027 0.032 16 0.013 0.162 0.162 11

372 Stores 0.010 0.024 0.032 38 0.012 0.027 0.029 20

374 Inst 0.010 0.024 0.032 61 0.012 0.036 0.060 38

376 Boiler 0.005 0.018 0.020 26 0.006 0.024 0.024 7

380 Warehouse 0.010 0.026 0.034 115 0.012 0.048 0.067 73

392 Plant Pr 0.006 0.017 0.021 41 0.011 0.030 0.030 13

396 Custodian 0.014 0.027 0.088 48 0.014 0.081 0.759 33

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

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

Other 0.010 0.027 0.079 124 0.018 0.066 0.266 45

a. N/A = not applicable (five or fewer records)
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 25 of 43

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

Cost center

Monday samples (mg/L) Number of

records

Friday Samples (mg/L) Number of

Median 95th percentile Maximum Median 95th percentile Maximum records

110-300 0.007 0.015 0.024 61 0.019 0.036 0.036 8

310 0.016 0.028 0.038 35 0.031 0.062 0.064 25

320 0.011 0.027 0.035 205 0.021 0.055 0.140 127

321-350 0.014 0.034 0.052 86 0.025 0.054 0.066 58

359 0.012 0.034 0.039 44 0.022 0.050 0.052 28

360-369 0.011 0.028 0.049 154 0.021 0.040 0.058 83

370 0.011 0.026 0.062 128 0.018 0.048 0.344 78

380 0.009 0.021 0.029 30 0.012 0.030 0.032 25

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

392-398 0.008 0.022 0.039 107 0.011 0.037 0.063 79

400-460 0.011 0.026 0.050 200 0.020 0.050 0.118 66

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

320 Ref 0.012 0.026 0.040 154 0.022 0.081 0.127 93

330 Gr Salt 0.010 0.024 0.030 49 0.024 0.054 0.055 27

370 Mill 0.018 0.025 0.025 10 0.019 0.031 0.031 6

370 Misc 0.009 0.022 0.028 57 0.015 0.064 0.080 39

370 Mtns 0.011 0.025 0.042 341 0.018 0.042 0.058 223

370 Mtns Elec 0.009 0.022 0.034 42 0.019 0.032 0.052 24

370 Mtns Misc 0.010 0.025 0.029 86 0.016 0.044 0.050 54

372 Stores 0.010 0.023 0.027 33 0.010 0.026 0.027 22

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

376 Boiler 0.010 0.020 0.021 26 0.008 0.010 0.010 7

380 Warehouse 0.009 0.022 0.026 105 0.011 0.032 0.036 66

392 Plant Pr 0.009 0.018 0.021 48 0.008 0.014 0.015 15

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

Other 0.008 0.024 0.035 123 0.021 0.061 0.094 43

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

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

Cost center

Monday samples (mg/L) Number of

records

Friday samples (mg/L) Number of

Median 95 records th percentile Maximum Median 95th percentile Maximum

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. NA = not applicable (five or fewer records)
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 26 of 43

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

Cost center

Monday samples (mg/L) Number of

records

Friday samples (mg/L) Number of

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.021 0.035 0.259 34 0.027 0.078 0.178 24

120 0.016 0.035 0.085 135 0.024 0.078 0.143 66

150 0.012 0.023 0.023 14 0.014 0.035 0.035 9

180 0.014 0.027 0.034 33 0.020 0.034 0.036 22

200 0.012 0.028 0.033 18 0.030 0.120 0.120 14

250-290 0.019 0.043 0.052 50 0.027 0.078 0.100 26

300-310 0.014 0.027 0.048 34 0.037 0.054 0.083 16

320 0.014 0.026 0.052 148 0.027 0.112 0.185 72

321-340 0.014 0.037 0.046 42 0.025 0.068 0.137 28

350-369 0.014 0.032 0.336 75 0.016 0.078 0.205 47

370 0.016 0.040 0.059 115 0.021 0.054 0.122 75

372-378 0.010 0.026 0.035 97 0.013 0.036 0.042 43

380-396 0.012 0.023 0.029 87 0.016 0.035 0.039 39

400-460 0.012 0.029 0.048 115 0.014 0.056 0.081 47

500-510 0.008 0.021 0.025 43 0.012 0.024 0.045 22

520-530 0.013 0.027 0.028 51 0.012 0.035 0.041 25

550 0.013 0.025 0.027 46 0.012 0.027 0.041 36

600-690 0.008 0.021 0.065 91 0.008 0.023 0.027 33

330 Gr Salt 0.013 0.027 0.029 46 0.020 0.042 0.044 21

370 Elec 0.012 0.023 0.025 28 0.018 0.059 0.059 13

370 Mill 0.016 0.029 0.029 31 0.027 0.058 0.258 25

370 Misc 0.015 0.027 0.091 91 0.016 0.043 0.111 59

370 Mtns 0.014 0.028 0.041 178 0.019 0.052 0.095 117

370 Pipe 0.014 0.029 0.045 22 0.029 0.058 0.058 14

380 Warehouse 0.012 0.027 0.029 69 0.012 0.025 0.035 50

392 Plant Pr 0.010 0.026 0.027 50 0.014 0.027 0.034 31

Other 0.012 0.029 0.070 95 0.015 0.036 0.087 55

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

Cost center

Monday samples (mg/L) Number of

records

Friday samples (mg/L) Number of

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.020 0.066 0.282 84 0.032 0.103 0.232 75

120 0.012 0.029 0.090 197 0.016 0.071 0.182 115

150 0.012 0.026 0.028 57 0.014 0.050 0.067 36

180 0.017 0.043 0.054 32 0.018 0.056 0.226 29

200 0.019 0.053 0.078 32 0.018 0.088 0.114 25

290 0.012 0.038 0.054 32 0.013 0.070 0.106 16

310-321 0.014 0.036 0.106 72 0.018 0.063 0.118 53

350-360 0.011 0.073 0.104 48 0.008 0.085 0.358 26

370 0.012 0.025 0.062 138 0.017 0.066 0.626 86

372-380 0.012 0.022 0.026 44 0.010 0.025 0.027 25

380-390 0.010 0.026 0.029 66 0.012 0.040 0.084 34

392 0.008 0.023 0.026 41 0.008 0.018 0.040 25

400-460 0.012 0.027 0.068 94 0.013 0.048 0.165 34

500-510 0.008 0.020 0.024 72 0.006 0.022 0.024 44

520 0.010 0.021 0.029 64 0.012 0.033 0.044 35

530 0.010 0.026 0.064 28 0.008 0.017 0.052 16

550-569 0.010 0.022 0.029 118 0.008 0.026 0.031 75

600-690 0.008 0.017 0.027 73 0.010 0.016 0.024 14

370 Elec 0.012 0.024 0.029 40 0.011 0.050 0.052 28

370 Mill 0.014 0.026 0.028 33 0.024 0.052 0.081 22

370 Misc 0.012 0.025 0.028 99 0.009 0.046 0.058 63

370 Mtns 0.017 0.062 0.125 102 0.013 0.029 0.064 57

370 Pipe 0.013 0.029 0.053 37 0.013 0.052 0.090 27

Other 0.012 0.027 0.042 95 0.011 0.050 0.059 52
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 27 of 43

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

Cost center

Monday samples (mg/L) Number of

records

Friday samples (mg/L) Number of

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.018 0.034 0.042 41 0.018 0.094 0.152 19

120 0.011 0.025 0.047 156 0.014 0.072 0.318 102

150 0.011 0.035 0.347 61 0.011 0.036 0.049 35

180 0.020 0.045 0.052 45 0.024 0.102 0.146 26

200 0.015 0.024 0.028 19 0.016 0.069 0.088 16

290 0.010 0.015 0.021 26 0.010 0.025 0.039 19

310-321 0.011 0.042 0.065 29 0.012 0.024 0.024 16

350 0.007 0.018 0.027 36 0.005 0.012 0.024 19

370-379 0.008 0.022 0.045 165 0.010 0.055 0.193 120

380-390 0.010 0.026 0.050 46 0.011 0.026 0.033 27

392 0.004 0.010 0.015 28 0.006 0.015 0.016 18

400-460 0.009 0.039 0.078 85 0.017 0.045 0.058 25

500-510 0.006 0.015 0.016 60 0.005 0.012 0.016 34

520 0.007 0.018 0.020 57 0.008 0.022 0.030 33

530 0.006 0.018 0.022 19 0.012 0.022 0.022 7

550-569 0.007 0.018 0.027 109 0.007 0.014 0.026 73

600-690 0.006 0.014 0.016 65 0.005 0.011 0.011 5

370 Elec 0.008 0.024 0.027 28 0.008 0.027 0.031 22

370 Mill 0.012 0.030 0.048 35 0.012 0.048 0.060 22

370 Misc 0.010 0.024 0.057 88 0.008 0.025 0.039 63

370 Mtns 0.011 0.025 0.069 69 0.012 0.029 0.061 48

370 Pipe 0.012 0.020 0.024 37 0.011 0.027 0.042 26

Other 0.010 0.020 0.075 85 0.009 0.033 0.055 43

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

Cost center

Monday samples (mg/L) Number of

records

Friday samples (mg/L) Number of

Median 95th percentile Maximum Median 95th percentile Maximum records

110 0.015 0.030 0.039 19 0.025 0.053 0.053 12

120 0.009 0.030 0.052 130 0.015 0.048 0.100 83

150 0.006 0.018 0.019 25 0.011 0.054 0.054 13

180 0.012 0.036 0.036 20 0.025 0.048 0.048 9

200 0.006 0.027 0.027 17 0.011 0.024 0.024 9

290 0.006 0.024 0.024 22 0.006 0.036 0.036 12

310-321 0.009 0.021 0.022 20 0.016 0.459 0.459 11

350 0.002 0.008 0.008 13 0.002 0.009 0.009 15

370-379 0.006 0.027 0.046 107 0.008 0.050 0.052 61

380-390 0.007 0.021 0.025 26 0.008 0.029 0.029 13

400-460 0.006 0.025 0.035 53 0.008 0.032 0.045 24

500-510 0.002 0.008 0.011 26 0.004 0.008 0.010 35

520 0.006 0.025 0.018 34 0.004 0.024 0.024 25

530 0.005 0.019 0.021 24 0.004 0.022 0.022 12

550-569 0.005 0.012 0.020 72 0.004 0.012 0.028 38

610-690 0.006 0.011 0.012 18 0.004 0.008 0.008 4

370 Elec 0.012 0.027 0.027 13 0.009 0.036 0.036 10

370 Mill 0.008 0.024 0.027 32 0.008 0.034 0.071 16

370 Misc 0.008 0.021 0.028 53 0.010 0.033 0.036 29

370 Mtns 0.009 0.022 0.066 46 0.009 0.050 0.055 25

370 Pipe 0.006 0.015 0.024 22 0.015 0.088 0.088 13

Other 0.006 0.019 0.027 58 0.004 0.025 0.213 37

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.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 28 of 43

5.3.2

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

Environmental Monitoring Period (1975 to 1984)

5.3.3

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

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

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

1998a,b,c, 2000a, 2001). 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.

Remediation Period (1985 to 2001)

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

the early part of the remediation period (1985 to 1990). If no individual bioassay data and no 

applicable co-worker bioassay data are available for a claimant for the 1985 to 1990 time period, the 

environmental data for 1985 to 1990, described in the Technical Basis Document for the Weldon 

Spring Plant – Occupational Environmental Dose (ORAU, 2005c) can be used to estimate intake. 

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

estimate worker intakes. 

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

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

1994, 1997, 1998a) and 0.0524 μg/L in September 1998 to 2001 (DOE 1998b,c, 2000a, 2001) using 

kinetic phosphorescence analysis. Uranium results of 0.2 μg/L or greater were considered positive for 

occupational uranium intakes in 1997 (DOE 1997), and results of 0.3 μg/L or greater (DOE 1998c, 

2000a, 2001) were considered positive for occupational uranium intakes from 1998 through 2001.

5.4 IN VIVO MEASUREMENTS

5.4.1

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

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

some WSP workers for thorium in 1966:

Operational Period (1957 to 1967)

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.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 29 of 43

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

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

minutes was considered to have at least one lung burden.

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

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

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

‘positive’ designation. (Ingle 1991)

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

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

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

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

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

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

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

a marker for thorium.

5.4.2

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

Environmental Monitoring Period (1975 to 1984)

5.4.3

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

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

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

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

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

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

for 232Th in the lung (DOE 1991). 

Remediation Period (1985 to 2001)

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

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

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

(DOE 2001). Revision 7 (DOE 2001) of the WSSRAP Technical Basis Manual states:

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

intake.

and

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

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

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

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

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

a suspected or actual intake. Revision 7 (DOE 2001) states that such measurements were normally 

reserved for “those incidents where the intake was suspected to exceed 500 mrem CEDE ....” 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 30 of 43

Revision 7 (DOE 2001) also states that “requiring lung counts for affected workers” was one of the 

actions “never used.”

5.5 AIR CONCENTRATION AND DUST MEASUREMENTS

5.5.1

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

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

general air samples (MCW 1965). 

Operational Period (1957 to 1967)

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). 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 eight-hour workday.

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

5.5.1.1 Maximum Acceptable Concentrations

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

dpm/m3 (3.2 × 10-11 μCi/cm3

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

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

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

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

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

) was more restrictive 

than the AEC standard.

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

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

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

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

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

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

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

, respectively. However a provisional 

level of 3 × 10-11 μCi/cm3 was recommended. This is the value that was adopted in AEC Appendix 

0524 Annex 1, Standards for Radiation Protection (AEC 1963). This provisional level for the MPC for 

natural thorium is similar to the time-weighted average MAC for uranium used at WSP.

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

alpha/m3 was applied across the WSP.

5.5.1.2 Special Curie for Uranium

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

because the former was an unambiguous designation. Until 1973 when the National Council on 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 31 of 43

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

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

1973):

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

= 7.49 × 1010 d/s

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

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

between the measured DWA and the guideline or standard as an index of exposure. 

5.5.1.3 Dus t Expos ure Calculation

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

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

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

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

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

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

WSP employees. 

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

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

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

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

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

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

/d for typical light work for an 

eight-hour work day or by using a job-specific value.

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

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

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

activity. The samples were analyzed for alpha and beta activity. 

5.5.1.4 Dus t Studies

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

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

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

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

assess intake.

An undated document titled Summaries of Dust Concentrations at Production Jobs (MCW c. 1966) 

provides data on time-weighted average dust concentrations for various work areas for the period 

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

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

ranging from 3 to 5 cm2

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

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

as “hand held breathing zone type.”
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 32 of 43

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

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

areas were counted after a minimum delay of 100 hr.

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

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

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

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

of 1.5 alpha dpm/μg U (0.68 pCi/μg U). Table 5-18 summarizes the work group concentrations.

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

Work group title

Daily weighted average concentration (μg U/m3)

1958 1959 1960 1961 1962 1963 1964 1965 1966

Sampling and Repackaging 

Cost center 110

65 55 45 55 32 31 20 60

Digest-Raffinate, Cost center 120 17 7 11 8 7

Extraction, Cost center 120 4 3 5 2 2

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

Green Salt, Cost center 150 40 37 37 53 53a 16a 16 22

Dingot Reduction, Cost Center 180 115 65 150a 150a 78 50

Metal – other than reduction

Cost centers 200, 210, 220, 280

44 52 63 40b 55e

Core area, Cost center 250 20

Pilot/Scrap Plants, Cost center 290 50c 50c 50c 50c 50c

Special Projectsd

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

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

c. Overall average - not weighted.

d. No averages given.

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

Thorium was used intermittently in various operations at WSP from November 1963 to 1966. 

Table 5-19 lists the thorium dust concentration data by year.

Table 5-19. Work group time-weighted average 

exposure for thorium operations.

Work group title

Daily weighted average 

concentration (dpm/m3)

1963 1964 1965

Pot Room 78a 50b

Metals other than reduction 17b 113c

Scrap plant 180d 117c 28e

a. 40% of the effort on thorium at this index.

b. 50% of the effort on thorium was at this index.

c. 60% of the effort on thorium jobs.

d. 17% of the effort on thorium.

e. 80% of the effort on thorium.

The dust concentration data can be used to indicate general levels of airborne radionuclides. The 

percentage of effort stated in the Table 5-19 indicates the fraction of the time the work involved 

thorium. As noted previously, there is no indication that the dust concentration data were used in 

calculating doses to individual workers at WSP.

Dust concentrations were also measured for individual operations for specific years. These 

operations were intermittent and did not represent the total work year. The time-weighted average 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 33 of 43

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

70 dpm/m3

. Table 5-20 summarizes these calculations.

Table 5-20. Time-weighted average thorium dust concentrations.

Operation Bldg. Year

Fraction

of MAC Total effort

Repackaging ThO2 feeds 101 1966 0.1 10 shifts

Hopper packaging 101 1966 0.2 10 shifts

Conveyors at repackaging 101 1966 0.1 10 shifts

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

Oven drying ThO2 solution 101 1965 0.2 Intermittent 

Hopper feed and digestion 103 1966 0.1 No information

Raffinate 103 1966 <0.1 No information

Misc. operations 103 1966 <0.1 Non-routine job

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

Extraction 105 1966 <0.1 No information

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

Crystal denitration 103 1964-65 0.9 Airline masks prescribed

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

Solution drying – vacuum unload 103 1965-66 0.6 10 months – “Comfo” respirators prescribed

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

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

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

TNTa repackaging 301 1964-65 0.04 No information

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

ThO2 repackaging 301 1965 0.8 Airline masks used for some tasks

Kiln calcining sump cake 301 1966 0.3 “Comfo” respirators used for some tasks

TNT repackaging 403 1963-64 0.2 Eight months

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

ThO2 repackaging 403 1964-65 0.3 13 months

Fluid bed denitration 403 1965 0.6 Three months

Fluid bed denitration 403 1965-66 0.4 No information

a. TNT = trinitrotoluene.

5.5.2

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

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

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

data were found for that period.

Transfer to the U.S. Army (1968 to 1974)

5.5.3

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

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

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

National Lead of Ohio to perform environmental monitoring around the WSQ and WSRP. Bechtel 

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

from the Army to DOE in 1985.

Environmental Monitoring Period (1975 to 1984)

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

5.5.4

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

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

1989, respectively. Site cleanup began in earnest in 1990 as described in ORAU (2005b).

Remediation Period (1985 to 2001)
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 34 of 43

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

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

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

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

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

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

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

filters was 1 pCi. The calculated minimum detectable concentrations (MDCs) for a 6-hr 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 hr/d in the WSQ or WSRP airborne radioactivity areas was 

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

Basis Manual (DOE 1994) required BZA samples for one in five individuals spending more than 1 hr/d

in areas with concentrations greater than 10% of the DAC and one in three individuals in the WSQ or 

WSRP areas. As of Revision 2 (DOE 1997), one in four workers in areas with concentrations greater 

than 2% of the DAC were required to wear BZA samplers.

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

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

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

Basis Manual (DOE 2001), that threshold was never exceeded. Therefore routine monitoring data for 

radon gas or short-lived decay products are not available. Environmental radon measurements that 

were taken periodically are provided in the Technical Basis Document for the Weldon Spring Plant –

Occupational Environmental Dose (ORAU, 2005c). These measurements provide a basis for 

estimating worker exposure to short-lived radon decay products. 

5.6 ASSESSMENT OF INTAKES

5.6.1

5.6.1.1 Uranium Intakes

Operational Period (1957 to 1967)

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

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

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

least one worker in the group being sampled each week with Monday – Friday – Monday sampling for 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 35 of 43

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 5-7 to 5-16 (Section 5.3.1.1.5), list the 

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

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

provide the cost center (Section 5.3.1.1.5).

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

uranium source terms:

• Natural uranium, before 1961

• Natural uranium, recycled, 1961 to 1962

• Enriched (1%) uranium, recycled, 1963 to 1967

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. However, for workers in Building 101 where the uranium 

concentrates were initially processed, the concentrations of 230Th and 226Ra should be assumed to be

5% of the 238U activity and 1% of the 238U activity, respectively as described in Section 5.2.2.

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 claimant-favorable.

5.6.1.2 Thorium Intakes

No quantitative in vitro or in vivo bioassay results have been observed for thorium (Sections 5.3.1.2 

and 5.4.1). However, dust studies for thorium operations (Section 5.5.1.4) indicate that airborne 

thorium dust concentrations exceeded the MAC for several of the operations studied. 

The suggested approach to assess natural thorium intakes is to use the approach in Section 5.3 of 

the Fernald TBD for internal dosimetry (ORAU 2004a). Application of the Fernald approach to WSP is 

based on the assumption that the thorium operations at Fernald are reasonably applicable to thorium 

operations at WSP. Assumptions and information used for the Fernald approach that also appear to 

be valid for WSP are:

1. Few, if any, in vitro analyses exist in worker files. The in vivo results that exist are not 

quantified for thorium, and there is not enough information available to quantify the thorium 

deposition from the in vivo results.

2. Although respiratory protection equipment was available and its use required, it cannot be 

assumed that the respirators were always used or that there was a tight seal to the face when 

they were used. No respiratory protection factor is assumed.

3. The MAC of 100 dpm/m3 (4.5 × 10-11 μCi/cm3

) was used at Fernald for control purposes. The 

MAC of 70 dpm/m3 was used at WSP for control purposes, or a factor of 0.7 less than the 

control level at Fernald.

4. Workers could have been exposed to airborne thorium above the MAC level. It is assumed 

that 100 hr/yr at 10 MAC accounts for intermittent exposures to high levels of airborne thorium.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 36 of 43

5. Normal operations would reasonably be controlled to levels well under the MAC. An intake of 

0.1 MAC for 500 hr/yr was assumed for normal operations. 

Under these assumptions, the claimant-favorable assumption results in an annual exposure of 

735 MAC-hr (in comparison to 1,050 MAC-hr for Fernald) or an intake of approximately 40 nCi Th/yr

(in comparison to 60 nCi Th/yr for Fernald).

The Fernald approach further considers the isotopic composition and daughter contributions:

Therefore, in the absence of monitoring data a claimant-favorable default intake is 

30 nCi/y (164 pCi/d) each of Th-232 and Th-228 (the alpha emitting isotopes detected 

on the air samples) plus a 60 nCi/y (164 pCi/d) intake of Ra-228. The Ac-228 beta 

emitter adds about 1% to the effective dose and therefore can be ignored in the dose 

calculations. Although an equal intake of Ra-224 might have occurred, because of its 

alpha emissions, it is adequately accounted for in the thorium intake assumption, which 

is based on detection of all alpha emitters on the air samples (ORAU 2004a).

The corresponding default intake values for WSP are 20 nCi/yr each of 232Th and 228Th (the 

alpha-emitting isotopes detected on the air samples) plus a 40-nCi/yr intake of 228Ra. This approach 

should be applied to those who worked directly with thorium or in the immediate vicinity of thorium 

processes (e.g., chemical or metal operators and maintenance workers as well the Health and Safety 

staff who monitored the operations).

Thorium operations were mainly in the pilot plants (Building 403) and the refinery (Building 103) (DOE 

1986). The MCW Four-Plant Study considered workers in any of the following locations to have been 

exposed to thorium dust:

Bldg. 101 – sampling and repackaging

Bldg. 103 – digestion, denitration, packaging

Bldg. 105 – organic extraction, purification

Bldg. 201 – interim storage

Bldg. 301– high temperature firing, packaging

Bldg. 403 – packaging, fluid bed denitration

Bldg. 404 – No thorium operations but some exchange of personnel with 403.

Maintenance – All crafts involved in repair of thorium handling machinery.

Janitors – Clean up of contaminated equipment and laundry.

Warehousing – storage, shipping (Ingle 1991).

It is claimant-favorable to apply the default approach to any worker in these buildings or professions in 

the period of thorium processing (November 1963 to 1966). For workers not in these buildings or 

professions during this period, thorium intakes should be assessed as environmental intakes in 

accordance with Technical Basis Document for the Weldon Spring Plant – Occupational 

Environmental Dose (ORAU 2005c) unless there is information in the worker’s file that indicates

involvement in thorium operations.

5.6.1.3 Intakes of Other Radionuclides

Intakes of other radionuclides, such as radon, should be assessed as environmental intakes in 

accordance with the Technical Basis Document for the Weldon Spring Plant – Occupational 

Environmental Dose (ORAU, 2005c). There are no personal dosimetry data or air concentration 

measurements available for radon in the WSCP during the operational period.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 37 of 43

The annual average environmental radon concentrations for the Weldon Spring Chemical Plant 

(WSCP), Weldon Spring Raffinate Ponds (WSRP), and Weldon Spring Quarry (WSQ) are provided in 

Table 4.5 of ORAU (2005c). The site-wide maximum annual exposures to radon decay products, by 

year, are provided in Table 4-7 of that same document. These estimates are based on environmental 

radon gas releases. The calculation in ORAU (2005c) was based on the assumption that all radon 

gas generated by decay of Ra-226 was released during processing. The environmental 

concentrations were calculated for the areas within 100 meters of the assumed release point, the acid 

recovery plant stack. 

Occupational exposures to radon decay products in Working Level Months (WLM) were calculated 

assuming an equilibrium factor of 0.3, a value that is typical for outdoor environments. The estimated 

annual radon decay product exposure for the years 1957 to 1967 was 8.7E-2 WLM. The equilibrium 

factor for indoor environments, thus the exposure for a given 222Rn concentration, may be higher than 

for outdoor environments, depending on the general building ventilation. 

The Weldon Spring Historic Dose Estimate (Meshkov et al., 1986) states that the prime source of 

radon emissions was the acid recovery plant. The denitration process that Meshkov (1986) describes 

as the “dustiest part of the whole operation” was carried out inside large hoods. Therefore, it is 

reasonable to assume that most of the radon released during processing went out the stack and only 

a small fraction was released into the room. However, the radon released to the environment and 

present in the area would have been drawn back into the building through the ventilation systems. 

Under those circumstances, the radon concentration inside the building would have been 

approximately equal to the radon concentration outdoors. If a typical indoor equilibrium factor of 0.5 

is assumed rather than the outdoor factor of 0.3, the estimated annual exposure would have been 

1.5E-1 WLM for 100 percent occupancy for 2000 hours per year.

5.6.2

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

(2005c).

Environmental Monitoring Period (1975 to 1984)

5.6.3

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

remediation that was designed to detect and evaluate intakes of 100 mrem CEDE or more for “all 

occupational radionuclide intakes in a year (other than radon, thoron, and their progeny)” (DOE 2001), 

it is reasonable to consider that the worker’s file will contain a detailed report of the pertinent data 

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

be used by the DR as found. If needed, a detailed description of the internal dosimetry program is 

available in the series of WSSRAP Technical Basis Manuals (DOE 1991, 1994, 1997, 1998a,b,c, 

2000a, 2001). 

Remediation Period (1985 to 2001)

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

accordance with ORAU (2005c).
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 38 of 43

REFERENCES

ACJ & Associates. 2004. User Manual for IMBA Expert TMOCAS-Edition (Version 3.2). Richland WA. 

April.

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

Appendix 0524 Annex 1. Washington D. C. August 12.

DOE (U.S. Department of Energy), 1986, Historical Nuclear Materials Balance Report for the Former 

AEC-Owned Weldon Spring Chemical Plant, DOE/OR-872, Weldon Spring, Missouri, July.

DOE (U.S. Department of Energy), 1991, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 0, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, November.

DOE (U.S. Department of Energy), 1994, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 1, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, September.

DOE (U.S. Department of Energy), 1997, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 2, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, January.

DOE (U.S. Department of Energy), 1998a, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 3, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, January.

DOE (U.S. Department of Energy), 1998b, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 4, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, September, Oak Ridge Operations Office, Oak 

Ridge, Tennessee.

DOE (U.S. Department of Energy), 1998c, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 5, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, November.

DOE (U.S. Department of Energy), 2000a, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 6, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, January.

DOE (U.S. Department of Energy), 2000b, Ohio Field Office Recycled Uranium Recovery Report,

Ohio Field Office.

DOE (U.S. Department of Energy), 2001, Internal Dosimetry Program Technical Basis Manual: 

Weldon Spring Site Remedial Action Project, Rev. 7, DOE/OR/21548-241, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, December.

Dupree, E. A., 1979a, Preliminary Report on Mortality Study of MCW Workers, Review Committee, 

Uranium Division, Mallinckrodt Chemical Works, July 25. 
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 39 of 43

Dupree, E. A., 1979b, “Mallinckrodt Chemical Works – Visit from Mont Mason, August 21-24, 1979,” 

internal memorandum to A. P. Polednak, September 5, Oak Ridge Associated Universities, 

Oak Ridge, Tennessee.

Dupree, E. E., 1998, Mallinckrodt Expanded Cohort Study, Proposed Work Plan for Internal Radiation 

Dose Assessment, attachment to Letter to D. Utterback (National Institute for Occupational 

Safety and Health (NIOSH)), July 1998, Oak Ridge Associated Universities (ORAU). Oak 

Ridge, Tennessee.

Eisenbud, M., 1975, “Early Occupational Exposure Experience with Uranium Processing,” pp. 9–24 of

Conference on Occupational Health Experience with Uranium, Arlington, Virginia, 28-30 April 

1975, Energy Research and Development Administration.

HPS (Health Physics Society), 1960, Report of the ICRP Committee II on Permissible Dose for 

Internal Radiation (1959), Universities Press, Belfast, Ireland.

HPS (Health Physics Society), 1996, An American National Standard - Performance Criteria for 

Radiobioassay, HPS N13.30-1996, McLean, VA.

Ingle, J. N., 1991, “Mallinckrodt Chemical Works: Four Plant Study Classifications of Radium, Radon 

and Thorium Exposure,” ORAU Memorandum NIPF AU 005101, September 16, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee. 

ICRP (International Commission on Radiological Protection), 1979, Limits for Intakes of Radionuclides 

by Workers, ICRP Publication 30, Part 1. Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1994a, Dose Coefficients for Intakes of 

Radionuclides by Workers, ICRP Publication 68, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1994b, Human Respiratory Tract Model 

for Radiological Protection, ICRP Publication 66, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1995, Age-Dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, 

ICRP Publication 71, Pergamon Press, Oxford, England.

Lesperance, A. M., M. R. Siegel, and M. D. McKinney, 1992, A Review of Occupational Safety and 

Health Issues Relevant to the Environmental Restoration Program: Selected Case Histories 

and Associated Issues, PNL-8184; DE93 003287, Pacific Northwest Laboratory, Richland, 

Washington, October.

MCW (Mallinckrodt Chemical Works), 1961, Uranium Process Dust Studies from 1960 and 1961,

Weldon Spring, Missouri. 

MCW (Mallinckrodt Chemical Works), 1964, Diversification of the Weldon Spring Feed Material Plant,

Weldon Spring, Missouri. 

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

Department, Weldon Spring, Missouri.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 40 of 43

MCW (Mallinckrodt Chemical Works), c. 1966, Summaries of Dust Concentrations at Production Jobs, 

Life of Operation — from Year 1957 through 1966, Weldon Spring, Missouri.

Mallinckrodt Chemical Works, 1967, Fuel for the Atomic Age – Completion Report on St. Louis Area 

Uranium Processing Operations, 1942-1967, St. Louis, Missouri. September.

Mason, M. G., c. 1977, handwritten note on a copy of MCW (1964).

Meshkov, N., P. Benioff, J. Wang, and Y. Yuan, 1986, Weldon Spring Historical Dose Estimate, 

ANL/EES-TM-308; DE86 014085, Argonne National Laboratory, Energy and Environmental 

Services Division, Argonne, Illinois, July.

NCRP (National Council on Radiation Protection and Measurements), 1973, Specification of Units for 

Natural Uranium and Natural Thorium, NCRP Statement No. 4, December 28, Bethesda, 

Maryland.

NRC (Nuclear Regulatory Commission)1980. Final Generic Environmental Impact Statement on 

Uranium Milling (Appendix G). NUREG-0708, Vol. III. Washington D. C. September.

NRC (Nuclear Regulatory Commission)1981. MILDOS – A Computer Program for Calculating 

Environmental Radiation Doses from Uranium Recovery Operations, NUREG/CR-2011. 

Washington D. C. April.

ORAU (Oak Ridge Associated Universities), 2004a, Technical Basis Document for the Fernald 

Environmental Management Project – Occupational Internal Dose, ORAUT-TKBS-0017-5,

Rev. 00, Oak Ridge, Tennessee, May.

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

Plant – Occupational Internal Dose, ORAUT-TKBS-0011-5, Rev. 00, Oak Ridge, Tennessee, 

April.

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

Plant – Occupational External Dosimetry, ORAUT-TKBS-0028-6, Rev. 00, Oak Ridge, 

Tennessee.

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

Plant – Site Description, ORAUT-TKBS-0028-2, Rev. 00-B Oak Ridge, Tennessee.

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

Plant – Occupational Environmental Dose, ORAUT-TKBS-0028-4, Rev. 00, Oak Ridge, 

Tennessee.

Sanders, M., 1975, “Analysis of Long-Term Data on Uranium in Air at Y-12,” pp. 364–386 of

Conference on Occupational Health Experience with Uranium, Arlington, Virginia, 28-30 April 

1975, Energy Research and Development Administration.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 41 of 43

GLOSSARY

activity median aerodynamic diameter (AMAD)

The 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)

A unit of radioactivity equal to one nuclear disintegration per second.

bioassay

Measurement of amount or concentration of radioactive material either in the body or in 

biological material excreted or removed from the body. Another word for radiobioassay.

bioassay procedure

A 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

The quantity of radioactive material in an individual’s body at a particular point in time.

chronic

Low-level intakes received on a continuous basis.

class

The respiratory tract classification scheme in ICRP Publication 30 for inhaled material 

according to its rate of clearance from the pulmonary region of the lung. Materials are 

classified as D (days), W (weeks), or Y (years), according to how fast they clear the lungs; the 

half-times are class D in less than 10 days, class W in 10 to 100 days, class Y in more than 

100 days. Recent recommendations in ICRP Publication 66 modified the lung model; instead 

of class D, W, and Y, the standard nomenclature is now lung absorption types F (fast), M 

(moderate), and S (slow).

curie

A special unit of activity. One curie equals exactly 3.7 × 1010 nuclear transitions per second.

depleted uranium

Uranium having a percentage of 235U smaller than 0.7% found in natural uranium.

dose

A general term for absorbed dose, dose equivalent, effective dose equivalent, committed dose 

equivalent, committed effective dose equivalent, or total effective dose equivalent.

dose equivalent (H)

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

Dose equivalent is expressed in units of rem (or sievert), where 1 rem equals 0.01 sievert.

dosimetry

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

from external or internal sources of radiation.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 42 of 43

enriched uranium

Uranium that has been processed to contain a higher concentration of the isotope 235U than in 

natural uranium.

exposure

The general condition of being subjected to ionizing radiation, such as by exposure from 

external sources or to sources inside the body.

insoluble

A term loosely used to describe the relative degree of solubility of a material in body fluids. 

Recognizing that no material is absolutely insoluble, the terms low solubility or poorly soluble

are preferable.

intake

The amount of radionuclide taken into the body by inhalation, absorption through intact skin, 

injection, or ingestion or through wounds. Depending on the radionuclide involved, intakes 

can be reported in units of mass, activity, or potential alpha energy.

internal dose or exposure

The dose equivalent received from radioactive material taken into the body (i.e., internal 

sources).

internal dose assessment

An assessment of the intake and associated internal radiation dose to workers based on 

measurements taken in the work environment or from individual bioassay measurements.

in vitro

Outside the living body and in an artificial environment. Typically used for bioassay of a 

contaminant in excreta, such as in fecal or urine samples. The word derives from Latin 

meaning in glass. 

in vivo

In the living body of a plant or animal. Bioassay counting analysis of radionuclides in the 

human body. The word derives from Latin meaning in [something] alive.

isotope

Nuclides having the same number of protons in the nuclei (same atomic number) but having 

differing numbers of neutrons (different mass number).

lung solubility type (F, M, or S)

A classification scheme for inhaled material according to its rate of clearance from the 

pulmonary region of the lung.

minimum detectable amount (MDA)

The smallest amount (activity or mass) of an analyte in a sample that will be detected with a 

probability of nondetection (Type II error) while accepting a probability of erroneously deciding 

that a positive (nonzero) quantity of analyte is present in an appropriate blank sample (Type I 

error).

minimum detectable concentration (MDC)

The minimum detectable amount expressed in units of concentration.
Document No. ORAUT-TKBS-0028-5 Revision No. 00 Effective Date: 06/28/2005 Page 43 of 43

monitoring (personnel)

The measurement of radioactivity in the whole body, in a region of the body, in material 

eliminated from the body, or in the air for reasons related to the estimation of intake of 

radioactive material. The term includes interpretation of the measurements.

natural uranium

Uranium occurring in the natural state that has not been through a 235U enrichment process.

occupational dose

An individual’s ionizing radiation dose (external and internal), which results from that 

individual’s work assignment. Occupational dose does not include doses received as a 

medical patient or doses from background radiation or participation as a subject in medical 

research programs.

rad

A unit of absorbed dose equal to 100 ergs per gram of irradiated material.

radiation

Ionizing radiation: Alpha particles, beta particles, gamma rays, X-rays, neutrons, high-speed 

electrons, high-speed protons, and other particles capable of producing ions. Radiation, as 

used in this part, does not include non-ionizing radiation, such as radio waves, microwaves, or 

visible, infrared, or ultraviolet light.

radionuclide

A radioactive species of an atom characterized by the constitution of its nucleus specified by 

the total of the number of protons and neutrons.

rem

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

factor, where 1 rem equals 0.01 sievert. The word derives from roentgen equivalent in man.

routine monitoring

Monitoring carried out at regular intervals during normal operations.

type

The rate of material absorption from the lung to the blood; includes types F (fast), M 

(moderate), and S (slow).

working level (WL)

That concentration of short-lived 222Rn decay products (218Po, 214Pb, 214Bi, and 214Po) or shortlived 220Rn decay products (216Po, 212Pb, 212Bi, 212Po) in one liter of air that will result in the 

eventual emission of 1.3 x 105 MeV of alpha energy. 

working level month (WLM)

Exposure to a concentration of radon decay products in air equal to 1.0 working level (WL) for 

170 hours.