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DOE Review Release: 04/24/2020

Weldon Spring Plant – Occupational 

Environmental Dose

ORAUT-TKBS-0028-4 Rev. 04

Effective Date: 03/27/2020

Supersedes: Revision 03

Subject Expert(s): David P. Harrison

Document Owner 

Approval: Signature on File

David P. Harrison, Document Owner

Approval Date: 03/25/2020

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 03/25/2020

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 03/25/2020

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 03/25/2020

Approval: Signature on File

Timothy D. Taulbee, Associate Director of Science

Approval Date: 03/27/2020

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 2 of 38

 

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

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

Spring Plant – Occupational Environmental Dose. Incorporates 

formal internal and NIOSH review comments. Training is not 

required. Initiated by Robert Meyer.

05/17/2013 01 Revision initiated to revise the TBD as a result of completion of 

Advisory Board Work Group issues. Included information in 

Section 4.2.2.1 about the drying of the raffinate pits and its potential 

for resuspension of pit radionuclide content. Included factors for 

calculation of recycled uranium contaminant activity based on ppb or 

pCi uranium. Revised and implemented according to ORAUTPROC-0031, Rev. 02. Added a table for maximum sitewide median 

intake values to comply with PROC-0031, and also added a table of 

annual median intake values for the WSCP, WSRP, and WSQ. 

Replaced introductory text with updated template language in 

Section 4.1. Revised Section 4.1.1, Purpose, and Section 4.1.2, 

Scope. Revised Sections 4.2.1 and 4.2.2 to more specifically target

the radionuclides of concern and source terms that contribute to 95% 

of the potential internal dose. Revised the approach to determination 

of annual intake of radionuclides (Section 4.2.3.1) during the 

operational period to optimize the use of available site-specific 

monitoring data. Added a table of annual median values for 

occupational external dose to be used as surrogates for onsite 

ambient dose during the operational period. Added Section 4.4 as a 

summary of environmental doses for use by dose reconstructors and 

provided tabulated inhalation intakes and ambient dose default 

values. Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by David P. 

Harrison. 

03/29/2017 02 Revision initiated to refer to ORAUT-TKBS-0017-5, Feed Materials 

Production Center – Occupational Internal Dose, in relation to RU 

contaminants and their mass concentrations. Section 4.2.4 was 

added which includes a table of ratios of RU contaminant intake 

activity per unit activity uranium, Bq/Bq U. Table for maximum 

annual median intakes was changed to include the additional RU 

contaminants. Includes editorial changes. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by David P. Harrison. 

09/08/2017 03 Revision initiated to incorporate SC&A comments regarding the date 

stated for work containing recycled uranium (from 1961 and on). 

Changes were made in Section 4.2.2.1. Updated references in the 

text and in the Reference Section. Added and deleted some 

acronyms and abbreviations from the listing. No sections were 

deleted. No changes were needed as a result of formal internal 

review. Incorporates formal NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

David P. Harrison. 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 3 of 38

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/27/2020 04 Revision initiated to incorporate relevant information from ORAUTPROC-0060 Rev. 01. Rewritten such that only a best-estimate 

approach is used in determining the environmental doses. Radon220 and 228Ra intakes were added for years of thorium operations 

from 1963 to 1966. Table added (Table 4-1) for airborne 

concentrations, which feed the radionuclide intake tables (Tables 4-3 

and 4-4). Intakes based on averages were converted to medians of 

a lognormal distribution using equations in Battelle-TIB-5000, 

resulting in an adjustment factor of 1.828. The following years, 

facilities, and nuclides were adjusted by dividing by 1.828 as they 

were not adjusted in Rev 03: WSRP U-234: 1990–2001; WSRP 

Radon: 1983–2001; WSP U-234: 1985–2001; WSP Th-230: 1985–

2001; WSP Radon: 1985–2001; WSQ Radon: 1983–2001. WSQ U234 values for the years 1990, 1991, 1993–1996, and 1998–2001 

were changed. Rev 03 set these values equal to 0 as the airborne 

concentrations were considered to be insignificant. However, the 

nonzero airborne U-234 concentrations were included in this revision. 

The airborne concentration values for 1992 and 1997 were equal to 

zero thus they were not changed. The RU contaminant intakes from 

1988 to 2001 were changed as the uranium intakes for those years 

were decreased by a factor of 1.828 from converting from an 

arithmetic mean or average to a median of a lognormal distribution. 

Intakes were converted to units of activity per calendar day by 

dividing by an average of 365.25 days per calendar year. Maximum 

intakes for U and Rn added to Table 4-1. Incorporates formal internal 

and NIOSH review comments. Constitutes a total rewrite of the 

document. Training required: As determined by the Objective 

Manager. Initiated by David P. Harrison.
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 4 of 38

 

TABLE OF CONTENTS

SECTION TITLE PAGE

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

4.1 Introduction .............................................................................................................................. 8

4.1.1 Purpose........................................................................................................................ 9

4.1.2 Scope........................................................................................................................... 9

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations.................................... 10

4.2.1 Radionuclides of Concern........................................................................................... 10

4.2.1.1 Airborne Particulate Radionuclides .............................................................. 10

4.2.1.2 Radon .......................................................................................................... 10

4.2.1.3 Recycled Uranium........................................................................................ 10

4.2.2 Source Terms for Airborne Radionuclides .................................................................. 10

4.2.2.1 Operational Period, 1957 to 1966................................................................. 10

4.2.2.2 Postoperational Years, 1967 to 2002 ........................................................... 12

4.2.3 Annual Intake of Radionuclides .................................................................................. 12

4.2.3.1 Operational Period, 1957 to 1966................................................................. 12

4.2.3.2 Postoperational Years, 1967 to 2002 ........................................................... 17

4.2.3.2.1 Airborne Particulate Radionuclides............................................ 17

4.2.3.2.2 Radon ....................................................................................... 20

4.2.4 Radionuclide Intake Derivation ................................................................................... 23

4.3 External Exposure to Onsite Atmospheric Radionuclide Concentrations ................................ 24

4.3.1 Operational Period, 1957 to 1966 ............................................................................... 24

4.3.2 Postoperational Years, 1967 to 2002.......................................................................... 25

4.3.2.1 Weldon Spring Plant .................................................................................... 26

4.3.2.2 Weldon Spring Raffinate Pits ....................................................................... 26

4.3.2.3 Weldon Spring Quarry.................................................................................. 27

4.3.3 Organ Dose Conversion Factors ................................................................................ 27

4.4 Uncertainty............................................................................................................................. 27

4.5 Summary Environmental Dose Default Values....................................................................... 29

4.6 Attributions and Annotations .................................................................................................. 30

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

Glossary ............................................................................................................................................ 37
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 5 of 38

 

LIST OF TABLES

TABLE TITLE PAGE

4-1 Median airborne radioactivity concentrations (WSP, WSRP, WSQ) and sitewide 

maximums (U and Rn) ........................................................................................................... 13

4-2 RU contaminant intakes per unit activity of uranium............................................................... 20

4-3 Daily median intakes for WSP, WSQ, and WSRP.................................................................. 21

4-4 Maximum sitewide daily median intakes................................................................................. 22

4-5 Median values for occupational external gamma dose ........................................................... 25

4-6 Estimated onsite ambient doses............................................................................................. 28

4-7 Estimated maximum sitewide ambient dose at WSP.............................................................. 30

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Perimeter sampling locations and processing facilities........................................................... 15

4-3 Air particulate, radon, and gamma monitoring stations in WSP and WSRP............................ 18

4-4 Air particulate, radon, and gamma monitoring stations in WSQ.............................................. 19

4-5 Topographic map and locations of gamma exposure rate measurements for WSQ ............... 29
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 6 of 38

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

AWE atomic weapons employer

Bq becquerel

Ci curie

cm centimeter

d day

DCAS Division of Compensation Analysis and Support

DCF dose conversion factor

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DR dose reconstructor

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

ft feet

GSD geometric standard deviation

hr hour

IREP Interactive RadioEpidemiological Program

L liter

m meter

mg milligram

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NU natural uranium

ORAU Oak Ridge Associated Universities

pCi picocurie

ppb parts per billion

RU recycled uranium

s second

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

U.S.C. United States Code
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 7 of 38

WLM working level month

WSCP Weldon Spring Chemical Plant

WSP Weldon Spring Plant (also used where it is unnecessary to distinguish between the 

Weldon Spring Plant, Weldon Spring Quarry, and Weldon Spring Raffinate Pits)

WSQ Weldon Spring Quarry

WSRP Weldon Spring Raffinate Pits

yr year

μCi microcurie

μm micrometer

μR microroentgen

§ section or sections
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 8 of 38

 

4.1 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are not official determinations made by 

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

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

dose reconstructions at particular U.S. Department of Energy (DOE) or Atomic Weapons Employer 

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

relevant information is obtained about the affected DOE or AWE facility(ies), such as changing 

scientific understanding of operations, processes, or procedures involving radioactive materials. 

These documents may be used to assist NIOSH staff in the evaluation of Special Exposure Cohort 

(SEC) petitions and the completion of individual dose reconstructions under Part B of the Energy 

Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA).

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

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

“AWE facility” or a “DOE facility.” The term “AWE facility” is defined in EEOICPA to mean “a facility, 

owned by an atomic weapons employer, that is or was used to process or produce, for use by the 

United States, material that emitted radiation and was used in the production of an atomic weapon, 

excluding uranium mining or milling” [42 United States Code (USC) 7384I(5)]. On the other hand, a 

DOE facility is defined as “any building, structure, or premise, including the grounds upon which such 

building, structure, or premise is located—(A) in which operations are, or have been, conducted by, or 

on behalf of, the [DOE] (except for buildings, structures, premises, grounds, or operations ... 

pertaining to the Naval Nuclear Propulsion Program); and (B) with regard to which the [DOE] has or 

had—(i) a proprietary interest; or (ii) entered into a contract with an entity to provide management and 

operation, management and integration, environmental remediation services, construction, or 

maintenance services” [42 USC 7384I(12)]. The DOE determines whether a site meets the statutory 

definition of an AWE facility and the U.S. Department of Labor (DOL) determines if a site is a DOE 

facility and, if it is, designates it as such.

Under EEOICPA, a Part B cancer claim for benefits must be based on an energy employee’s eligible 

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

designated time period and location (i.e., a “covered employee with cancer”). After DOL determines 

that a claim meets the eligibility requirements under Part B of EEOICPA, DOL transmits the claim to 

NIOSH for a dose reconstruction. EEOICPA provides, among other things, guidance on eligible 

employment and the types of radiation exposure to be included in an individual dose reconstruction. 

Under EEOICPA, eligible employment at a DOE facility includes individuals who are or were 

employed by DOE and its predecessor agencies, as well as their contractors and subcontractors at 

the facility [42 USC 7384I(11)]. Also under EEOICPA, the types of exposure to be included in dose 

reconstructions for DOE employees are those radiation exposures incurred in the performance of 

duty. As such, NIOSH includes all radiation exposures received as a condition of employment at DOE 

facilities in its dose reconstructions for covered employees, which may include radiation exposures 

related to the Naval Nuclear Propulsion Program at DOE facilities, if applicable. This is because 

NIOSH does not determine the fraction of total measured radiation exposure at a DOE facility that is 

contributed by the Naval Nuclear Propulsion Program at the DOE facility during a specified period of 

time for inclusion in dose reconstruction.

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

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

dose reconstructions for covered employees [NIOSH 2010]:

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

conventional structures, and

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

reasons.
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 9 of 38

 

4.1.1 Purpose

The purpose of this technical basis document (TBD) is to describe the potential occupational 

environmental dose to workers at the Weldon Spring Plant (WSP), Weldon Spring Raffinate Pits 

(WSRP), and the Weldon Spring Quarry (WSQ). Occupational environmental exposure refers to 

exposures workers might have received while on the site but outside the facilities from elevated 

ambient radiation, facility effluent releases to the environment, and resuspension of radionuclides in 

soils. Effluent releases can result in internal and external exposures by inhalation of airborne 

radionuclides and by submersion in an effluent. This TBD provides estimated annual intakes for 

inhalation exposure and estimated doses from submersion and ambient exposure at WSP.

4.1.2 Scope

Historically, the WSP has been called the Weldon Spring Site, Weldon Spring Chemical Plant

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

Plant, Main Plant, or Main Site. The facilities covered under EEOICPA are the WSP, WSQ, and the 

WSRP. For convenience, WSP is used throughout the remainder of this document where it is 

unnecessary to distinguish between the plant, the quarry, and the raffinate pits. 

There are four periods for WSP:

• Site acquisition and development, 1954 to 1957;

• Operational, 1957 to 1966;

• U.S. Department of Defense (DOD) control of the WSP, 1967 to 1985; and

• Remediation, 1985 to 2002.

WSP employment is covered under EEOICPA only during the operational and remediation periods 

when the U.S. Atomic Energy Commission (AEC), U.S. Energy Research and Development 

Administration, and DOE had contractors and radioactive materials at WSP. WSQ and WSRP 

employment is covered during those periods and during the DOD control period. Due to the nature of 

site activities and available data, this document refers to the combined DOD control and remediation 

periods as the “postoperational years.” The Weldon Spring Site Remediation Program was effectively 

complete in 2002. A formal review of the remedial action has found that the remedies are “protective 

of human health and the environment” (DOE 2006). Therefore, all environmental doses in 2002 are 

zero. The site has since been transferred for long-term surveillance and maintenance, and it is not 

covered under EEOICPA beyond 2002. 

Workers at WSP received environmental doses during the operational period from stack effluents 

from buildings, contamination of soil from deposition of stack releases, and onsite storage of ore 

concentrates. Workers at WSQ received environmental doses during the same period from 

resuspension from contaminated rubble and soil being placed in (or already in) the disposal area. 

During the remediation period, workers received occupational doses from contaminated soils and 

structures at WSQ, WSP, and WSRP during excavation of contaminated sludges and soils, demolition 

and removal of contaminated structures, and placement of contaminated media in the onsite disposal 

facility in WSRP. For more information, see ORAUT-TKBS-0028-2, Weldon Spring Plant – Site 

Description (ORAUT 2017a).

Section 4.2 contains information for estimation of internal environmental dose and includes 

discussions of airborne particulate and radon concentrations at WSP, WSRP, and WSQ. Section 4.3

contains information for estimation of external environmental dose. Section 4.4 discusses the 

uncertainties that apply to the environmental doses. Section 4.5 summarizes the tables that provide 

environmental dose default values for use by dose reconstructors (DRs). Attributions and 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 10 of 38

 

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

clarification of the associated information, are presented in Section 4.6. 

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATIONS

This section describes internal exposure from the intake of radionuclides. Section 4.2.1 identifies the 

radionuclides of concern. Section 4.2.2 describes specific sources of various radionuclides in the 

outdoor environment. Section 4.2.3 presents the methods that were used to calculate activity intakes 

of radioactivity in the form of air particulates and radon. Section 4.5 provides tabulated values for 

maximum sitewide median inhalation intakes of radioactive air particulates and radon at WSP, WSRP, 

and WSQ.

4.2.1 Radionuclides of Concern

The radionuclides of concern are defined as those that make up 95% of the potential internal dose. 

For the purposes of dose reconstruction, the assumption was made that all uranium processed at 

WSP was natural uranium (NU) from 1957 through 1962, with a specific activity of 683 pCi/mg. After 

1962, all uranium was assumed to be enriched to 1%, with a specific activity of 973 pCi/mg (ORAUT 

2017b). 

4.2.1.1 Airborne Particulate Radionuclides

The radionuclides of concern for dose reconstruction are the naturally occurring isotopes of uranium 

(

234U, 235U, and 238U), their decay products (primarily 230Th and 226Ra), isotopes of natural thorium

(

228Th and 232Th), and their decay products. ORAUT-TKBS-0028-5, Weldon Spring Plant – 

Occupational Internal Dose (ORAUT 2017b) provides additional details for the internal deposition of 

these radionuclides of concern.

4.2.1.2 Radon

Three radon isotopes are generated during the decay of 235U, 238U, and 232Th: 219Rn, 222Rn, and 220Rn, 

respectively. Due to the limited amount of enriched uranium (up to 1%) processed at WSP, there was 

no large source of 223Ra and in turn 219Rn at WSP. The risks associated with 219Rn, due to its 

extremely short half-life (4 seconds) and small source term, were insignificant. Therefore, this TBD 

considers only the inhalation intakes for 222Rn and 220Rn to be potentially significant, as discussed in 

more detail below.

4.2.1.3 Recycled Uranium

The DR should assume, as favorable to claimants, that all of the uranium WSP processed beginning 

in 1961 was recycled uranium (RU) (ORAUT 2017b). For the periods that include RU (i.e., after 

1960), the DR should refer to Section 5.6.1.3.3 of ORAUT-TKBS-0028-5, Weldon Spring Plant – 

Occupational Internal Dose (ORAUT 2017b), in relation to RU contaminant mass concentrations.

4.2.2 Source Terms for Airborne Radionuclides

4.2.2.1 Operational Period, 1957 to 1966

WSP

Uranium and thorium were released from the WSP process building stacks during the operational 

period. No stack monitoring data have been found. Perimeter monitoring data are available in WSP 

environmental reports and a materials balance study (Harris 1986, pp. 10, 33) and form the basis of 

the estimates of uranium and thorium intake rates. The reported atmospheric discharges from the 

materials balance study were engineering estimates that were derived from airflows and other 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 11 of 38

 

process factors (ORAUT 2017a). The estimated amount of uranium activity that was emitted from the 

operating plants ranged from approximately 1 to 5 Ci/yr.

An estimate of radon release based on the amount of processed uranium during the operational 

period ranged from 12 to 34 Ci/yr assuming (1) 5,000 to 14,500 metric tons of uranium materials 

were processed per year, (2) 70% of this was elemental uranium, (3) radium activity was 1% of the 

uranium activity (believed to be conservatively high considering the radium content in yellowcake), 

(4) radon was in equilibrium with radium, and (5) all radon was released through the stack during 

processing (Meshkov et al. 1986, pp. 47–48). 

No raw ores were processed at WSP. Only ore concentrates (e.g., yellowcake) and other highly 

refined uranium and thorium compounds were processed. The ore concentrates were a relatively 

small source of radon because most of the radium (the radon precursor in the ore) was removed in 

the milling process, which occurred elsewhere.

According to the review of source emissions by Meshkov et al. (1986), only a fraction of radon (about 

20%) is released when attached to solid particles, such as those that occur in uranium ore 

concentrates. However, the digestion phase of the refining process would have released the trapped 

radon during the operational period. Off-gases that contained radon isotopes from this process were 

conveyed to the acid recovery plant at WSP. The discharge from the acid recovery plant was the 

primary source of radon emission (Meshkov et al. 1986). 

Drums of yellowcake were emptied into the hopper at the top of the receiving and sampling building 

(ORAUT 2017a). Dust was collected in three large collectors, each with its own stack, and either 

repackaged and shipped off the site or reprocessed. In the refinery plant process, uranium-laden dust 

was generated during material transfers, denitration, reduction and hydrofluorination, and conversion 

to solid metal. The denitration process was the dustiest part of the operation. Therefore, point 

sources of uranium emissions from stacks existed during this period. Stack monitoring concentration 

data for the operational period are not available.

Meshkov et al. (1986, p. 47) and Sears et al. (1975, p. 143) assumed that activity concentrations of 

230Th, 226Ra, and 210Pb in uranium ore concentrates were 5%, 1%, and 1%, respectively, of that of 

238U. However, specific isotopic characterization of raffinates at WSP indicates that the uranium mill 

processes that produced the yellowcake concentrates for both Feed Materials Production Center and 

WSP effectively removed the radium but were not effective in removing thorium, specifically 230Th

(Sears 1976, p. 4; ORAUT 2017c, p. 31). Therefore, this TBD assumes activity ratios for 230Th, 226Ra, 

and 210Pb in uranium ore concentrate of 80%, 1%, and 1%, respectively, of that of 238U (Mason 1958; 

ORAUT 2017c).

WSRP

Impurities that were generated in the uranyl nitrate purification process were bled off in the raffinate, 

which was also pumped to WSRP. The raffinate contained a variety of radionuclides including 

thorium and radium (Author unknown, 1967, pp. 10–11). Wastes in WSRP were not a significant 

source of particulate or gaseous airborne radionuclides during the operational period due primarily to 

the presence of water in the pits. It is unlikely that a significant portion of the sediments became 

sufficiently dry to be considered a source term during this period, even though 2 decades later DOE 

(1984a, p. 14) reported that Pits 1 and 2 could become dry during the summer months. National Lead 

Company of Ohio (NLO 1977, p. 18) stated:

The transport of radiological contamination by air is not considered a problem at the site. 

Good ground cover exists and the raffinate pits are either covered with water or remain 

moist due to the balance between precipitation and evaporation in the area. Even in 
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times of prolonged dry weather, the inherent consistency of the raffinate material 

contained in the pits precludes drying.

If the raffinate pits had been dry, resuspension from the pits should be reflected in the boundary 

station air sampling results (NIOSH 2012a). As discussed below in Section 4.2.3.1, perimeter 

monitoring data were extrapolated to the central portion of the plant for use in occupational 

environmental intake rate calculation.

WSQ

Wastes that were disposed of in WSQ during the operational period presented a diffuse source of 

particulate emissions to the air as a result of entrainment of contaminated, wind-exposed surface soil 

or rubble. Radon-222 and 220Rn (from 226Ra and 224Ra, respectively) in the quarry also presented a 

diffuse source of airborne radionuclides. Before 1963, the quarry contained only drummed thorium 

wastes that were probably submerged (Author unknown 1967, pp. 17–19) and not a significant source 

of radon because water is an effective barrier to radon release. Each 25-cm increase in the depth of 

a water column reduces the 222Rn concentration by approximately 50% (Usman, Spitz, and Weisman 

2005). The attenuation of 220Rn flux is expected to be even larger due to the shorter half-life of 220Rn. 

Therefore, the radon release before 1963 was negligibly small. 

In 1963 and 1964, an estimated 38,000 m3 of uranium- and radium-contaminated rubble, equipment, 

and soil were placed in WSQ after demolition of the Mallinckrodt Destrehan Street site. A majority of 

this waste was not submerged and was a potential source of radon exposure (Author unknown 1967, 

pp. 17–20). The emission rate from the quarry is assumed to be the same from 1963 until it was 

measured during the postoperational years.

4.2.2.2 Postoperational Years, 1967 to 2002

Diffuse (i.e., non-point-source) emissions predominated as the source of radionuclides in the air at the 

WSP, WSRP, and WSQ. Annual environmental monitoring reports provided estimates of air 

concentrations of particulate radionuclides and of radon at WSP and WSQ beginning in 1979. These 

estimates reflect emissions during the later years (1975 to 1985) of the DOD control period and 

throughout the remediation period (1985 to 2002).

4.2.3 Annual Intake of Radionuclides

Descriptions of particle size and absorption type are not available for radionuclides of concern. DRs 

should assume the default aerosol size of 5-μm activity median aerodynamic diameter (AMAD) (ICRP 

1994; ORAUT 2018a). In addition, DRs should select the absorption type that yields the highest dose 

to the organ of interest. Calculated intakes in units of activity per calendar day (becquerel per day, 

Bq/d) are based on a 2,400-m3

/yr inhalation rate using a breathing rate of 1.2 m3/hr for light activity

(Section 9.3.1.1 of ICRP 1994) and a 2,000-hour work year, adjusting for an average of 365.25 days 

in a calendar year (ORAUT 2020). Intakes can be scaled to a different rate or to consider partial year 

exposures. Environmental intakes from ingestion are considered negligibly small and therefore have 

not been included in the tabulated results. Airborne radioactivity concentrations that were reported as 

an arithmetic mean were converted to a median of a lognormal distribution using the equations in 

Table 2.2 of the Battelle-TIB-5000 (BT 2007), and a geometric standard deviation of 3 (ORAUT 

2018a, p. 14; ORAUT 2012, p. 17). 

4.2.3.1 Operational Period, 1957 to 1966

This TBD analysis used the available perimeter monitoring data to calculate estimated intakes of 

radioactive airborne particulates and radon according to the methods described in this section.
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Airborne Particulate Radionuclides

Uranium air concentration measurements at the WSP (and WSRP) perimeter from 1959 through 1965 

were evaluated. The perimeter data were converted to units of becquerels per cubic meter and 

adjusted by the dilution factor of 30 (discussed below), and the median airborne particulate 

concentrations are presented in Table 4-1. The original data were reported in units of “special 

uranium microcuries,” which is a modification of the “special curie” as defined in DOE (2009) and 

stated in the glossary section of this TBD. Converting to units of becquerel, the special uranium 

microcurie is multiplied by 3.7 × 104 Bq/μCi and by a factor of 2.024 to report total uranium activity. 

Refer to ORAUT-TKBS-0028-5 (ORAUT 2017b) for an expanded discussion of the use of the special 

uranium microcurie at WSP. Other radionuclides were not measured.

Table 4-1. Median airborne radioactivity concentrations (WSP, WSRP, WSQ) and sitewide

maximums (U and Rn).

a

 

Year

WSRP

U-234

(Bq/m3)

WSRP

Radon 

(Bq/m3)

WSP 

U-234 

(Bq/m3)

WSP

Th-230 

(Bq/m3)

WSP

natural 

Thorium 

(Bq/m3)

WSP 

Radon 

(Bq/m3)

WSQ 

U-234 

(Bq/m3)

WSQ 

Radon 

(Bq/m3)

Max. U-234 

(Bq/m3)

Max. 

Radon 

(Bq/m3)

1957–1959 0.00E+00 0.00E+00 2.83E–01 2.26E–01 0.00E+00 (b) 0.00E+00 0.00E+00 2.83E–01 (b)

1960 0.00E+00 0.00E+00 5.77E–01 4.61E–01 0.00E+00 (b) 0.00E+00 0.00E+00 5.77E–01 (b)

1961 0.00E+00 0.00E+00 9.90E–01 7.92E–01 0.00E+00 (b) 1.45E–03 0.00E+00 9.90E–01 (b)

1962 0.00E+00 0.00E+00 2.44E–01 1.95E–01 0.00E+00 (b) 9.83E–04 0.00E+00 2.44E–01 (b)

1963 0.00E+00 0.00E+00 4.42E–01 3.54E–01 0.00E+00 (b) 1.06E–03 (b) 4.42E–01 (b)

1964 0.00E+00 0.00E+00 3.38E–01 2.70E–01 2.62E–04 (b) 2.46E–03 (b) 3.38E–01 (b)

1965 0.00E+00 0.00E+00 1.84E–01 1.47E–01 4.86E–03 (b) 1.64E–02 (b) 1.84E–01 (b)

1966 0.00E+00 0.00E+00 1.84E–01 1.47E–01 1.59E–02 (b) 1.64E–02 (b) 1.84E–01 (b)

1967–1974 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1975–1976 0.00E+00 4.26E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 1.99E+01 1.64E–02 4.26E+01

1977–1978 0.00E+00 6.07E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 2.02E+01 1.64E–02 6.07E+01

1979 0.00E+00 1.11E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 1.11E+01 1.64E–02 1.11E+01

1980 0.00E+00 6.88E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 1.82E+01 1.64E–02 6.88E+01

1981 0.00E+00 4.51E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 2.33E+01 1.64E–02 4.51E+01

1982 0.00E+00 9.11E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 2.65E+01 1.64E–02 2.65E+01

1983 0.00E+00 5.74E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 1.20E+01 1.64E–02 1.20E+01

1984 0.00E+00 9.84E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.64E–02 1.37E+01 1.64E–02 1.37E+01

1985 0.00E+00 6.56E+00 7.47E–04 5.97E–04 0.00E+00 1.23E+02 1.64E–02 1.37E+01 1.64E–02 1.23E+02

1986 0.00E+00 9.30E+00 7.47E–04 5.97E–04 0.00E+00 1.89E+02 1.64E–02 1.31E+01 1.64E–02 1.89E+02

1987 0.00E+00 7.66E+00 7.47E–04 5.97E–04 0.00E+00 2.46E+02 1.64E–02 2.46E+01 1.64E–02 2.46E+02

1988 0.00E+00 1.20E+01 7.47E–04 5.97E–04 0.00E+00 3.36E+02 0.00E+00 3.72E+01 7.47E–04 3.36E+02

1989 0.00E+00 9.02E+00 7.47E–04 5.97E–04 0.00E+00 3.12E+02 0.00E+00 2.35E+01 7.47E–04 3.12E+02

1990 2.73E–05 5.47E+00 7.47E–04 5.97E–04 0.00E+00 1.97E+02 9.30E–06 1.91E+01 7.47E–04 1.97E+02

1991 1.37E–05 6.02E+00 4.51E–04 3.61E–04 0.00E+00 1.64E+02 1.80E–06 1.64E+01 4.51E–04 1.64E+02

1992 1.07E–05 6.02E+00 3.36E–04 2.69E–04 0.00E+00 1.31E+02 0.00E+00 1.48E+01 3.36E–04 1.31E+02

1993 1.26E–05 2.73E+00 3.77E–04 3.02E–04 0.00E+00 6.07E+01 1.70E–06 3.34E+01 3.77E–04 6.07E+01

1994 1.48E–05 7.66E+00 5.17E–04 4.13E–04 0.00E+00 1.39E+02 1.31E–05 9.30E+01 5.17E–04 1.39E+02

1995 1.29E–05 9.30E+00 4.18E–04 3.35E–04 0.00E+00 1.80E+02 4.38E–06 3.17E+01 4.18E–04 1.80E+02

1996 1.70E–05 1.56E+01 4.84E–04 3.87E–04 0.00E+00 2.21E+02 1.80E–06 7.11E+00 4.84E–04 2.21E+02

1997 2.05E–05 3.55E+01 4.51E–04 3.61E–04 0.00E+00 1.80E+02 0.00E+00 6.02E+00 4.51E–04 1.80E+02

1998 2.38E–05 6.29E+00 6.32E–04 5.05E–04 0.00E+00 1.07E+02 1.42E–06 4.05E+00 6.32E–04 1.07E+02

1999 2.11E–05 1.01E+01 6.23E–04 4.99E–04 0.00E+00 2.05E+02 2.02E–07 9.30E+00 6.23E–04 2.05E+02

2000 1.12E–05 5.47E+00 4.18E–04 3.35E–04 0.00E+00 2.13E+02 1.80E–06 7.11E+00 4.18E–04 2.13E+02

2001 1.12E–05 5.47E+00 4.18E–04 3.35E–04 0.00E+00 2.13E+02 1.80E–06 7.11E+00 4.18E–04 2.13E+02

2002 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

a Source: ORAUT (2019b). 

b Radon concentrations were calculated for the WSP during the operational period and for the WSQ from 1963 through 1966. However, 

these concentrations were not used in determining the radon intake for the operational period. The value of 12.4 WLM/yr (or 3.39E-02 

WLM/d) was used from ORAUT (2017b, p. 44) as stated below in Table 4-3. 

Perimeter monitoring data were extrapolated to the central portion of the plant for use in occupational 

environmental intake rate calculation. The extrapolation was based on the ratio of the calculated 

ground-level air concentration at two different locations of interest (100 meters and 750 meters

discussed further below). The ground-level air concentrations were calculated at these two downwind 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 14 of 38

 

distances using the Pasquill-Gifford equation for Gaussian plume atmospheric dispersion (shown 

below) assuming an emission rate of 1 Bq/s and the following parameters: 

(4-1)

( )    

= −+        

2 2

2 2

1 1

exp 2 2 y z z z

χ x,y y H

Q πσ σ u σ σ

where

χ(x,y) = ground-level concentration at point (x,y) (Bq/m3

).

x = downwind distance on plume centerline (m). In this case the values were 100 m 

and 750 m.

y = crosswind distance (m). In this case, y = 0 m.

Q = emission rate (Bq/s), which is set to unity in this case.

σy, σz = horizontal and vertical standard deviations of the contaminant concentration in the 

plume (m). At 100 m, σy = 17.6 and σz = 10.9. At 750 m, σy = 109 and σz = 80.

u = mean wind speed at level of plume centerline (m/s), which in this case is an annual 

average of 4.6 m/s.

H = effective release height (m), which is assumed to be 10 m to represent the stack of a 

large one-story industrial building.

The values of σy and σz were determined from equations the U.S. Nuclear Regulatory Commission 

uses in the RACHET2 computer program (Napier and Ramsdell 2005).

The wind speed of 4.6 m/s was based on the average wind frequency data from Lambert Field in 

St. Louis for 1961 through 1990 (ISCO 2009). The Lambert Field wind speed data were used since it 

is likely that it is the closest facility to the Weldon Spring Plant to which reliable wind speed 

information could be obtained. The same data were used to select an atmospheric stability class. 

Stability class B was selected for this analysis to represent a relatively turbulent flow that would 

increase the dilution between the center of the plant and the plant perimeter where the measurements 

were made. Stability class B is characterized as an unstable or turbulent atmosphere with surface 

wind speed in the range of 3 to 5 m/s and moderate incoming solar radiation. The only stability class 

that would be more turbulent – and therefore more favorable to the claimant – would be class A, but 

the wind frequency data do not support that choice.

The distances for the location of the perimeter monitoring and the onsite worker were based on 

inspection of a map of WSP sampling locations near the fence line (see Figure 4-1). Figure 4-1 does 

not include a distance scale, but Figure 4-2 includes a distance scale and a circle superimposed near 

the fence line. The radius of the circle is nominally 600 m, so this value was considered to represent 

the distance from the release point to the sampling location.

Further inspection shows that the processing area is on the east side of the plant, and a better 

representation of the greatest distance from the center of the processing area to a point on the 

perimeter is 750 m. The value of 100 m for the exposed worker location is based on the limitations of 

the atmospheric dispersion model and is because it is likely to be a reasonable average estimate of 

the distance of any individual on the ground from a stack on the roof of a process building.
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 15 of 38

 

Figure 4-1. Perimeter sampling locations and processing facilities (lacking distance scale) 

(Meshkov et al. 1986, p. 123).

When Equation 4-1 is solved for the ground-level air concentration (i.e., a unit emission rate) at both 

100 m and 750 m, the ratio of results defines a dilution factor that can be used to extrapolate the data 

that were measured at the perimeter to the concentration outdoors in the processing area. The 

calculated dilution factor is 30. For example, a measured air concentration of 0.01 Bq/m3 at the 

perimeter can be extrapolated to a concentration of 0.3 Bq/m3 near the center of the operating area of 

the plant.

Thorium intake (natural thorium or 232Th), for those years when it was processed (1963 to 1966), can 

be estimated by scaling the annual uranium and thorium mass throughput rates (2017a, p. 11) and 

apportioning that fraction of the intake rate to thorium instead of uranium. This is justified by the fact 

that similar or identical processing facilities and environmental controls were used for both uranium 

and thorium operations, so a portion of the environmental release could be expected to be thorium. 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 16 of 38

 

Thorium-228 is assumed to be in equilibrium with 232Th, so the 228Th intakes and activities equal those 

for 232Th. Radium-228 is added at a ratio of 1-to-2 232Th-to228Ra (NIOSH 2010b, p. 57).

The measured perimeter air concentration data for WSP and WSQ can be found in quarterly and 

semiannual environmental monitoring reports (MCW 1961a, p. 36, 53–54; 1961b, p. 6; 1961c, pp. 9, 

11; 1962a, p. 14; 1962b, p. 44, p. 65; 1962c, pp. 85–86; 1964a, p. 14; 1964b, pp. 10–11; 1965, 

pp. 13–14; 1966, pp. 10–11), and in the summary report by Meshkov et al. (1986, pp. 101, 103–104). 

In some instances, the same data were reported in more than one document; in those cases the 

contemporary environmental reports were preferred as the primary data source. Daily intake rates 

can be calculated based on the extrapolated air concentrations and an assumed breathing rate of 1.2 

m3

/hr for 2,000 hr/yr (Section 9.3.1.1 of ICRP 1994), adjusted for an average of 365.25 days in a 

calendar year (ORAUT 2020). 

For 1957 and 1958 at WSP, for which measurements are not available, this analysis assumed that 

perimeter air concentrations were the same as the measured concentrations in 1959, which is the 

operational year with the highest measured perimeter concentrations. This is considered an 

overestimate for the first 2 years when uranium receipts were lower than those during the main 

production years (1960 to 1964; see ORAUT 2017a).

Radon

Measured ground-level air concentrations of radon during the operational period were not reported in 

the available references, but Meshkov et al. (1986) estimated an annual release rate of 222Rn in the 

range of 12 to 34 Ci. This radon was released from the acid recovery building stack. Using the radon 

intake calculation as stated in ORAUT-TKBS-0028-5 (ORAUT 2017b), a daily 222Rn intake of 

3.39E-02 working level months/day (WLM/d), or 12.4 WLM/yr and adjusting for 365.25 days in a 

calendar year, should be assigned to WSP personnel during the operational period (NIOSH 2012b, 

pp. 112–116). This TBD analysis estimated radon concentrations for the WSQ during the operational 

period from measurements in the vicinity from 1977 through 1982 (Weidner and Boback 1982, p. 50; 

Bechtel 1983a, p. 25; 1983b, pp. 24, 27; Meshkov et al. 1986, p. 101). This approach is justified 

because the activities of 226Ra and 224Ra were not significantly depleted in the quarry over time due to 

limited leaching and continuous production of these isotopes from precursors in the waste. These 

years reasonably represent radon emanations for the WSQ during the operational period. These 

measurements were applied only to the period from 1963 to 1966. Before that, drummed thorium 

waste was most likely submerged and no significant source of radon existed. There were no radon 

intakes assigned for the WSRP during the operational period since the source term was small and the 

contents of the pits were covered with water.

Thoron

Although the estimated dose attributable to 220Rn and its progeny is relatively insignificant in relation 

to 222Rn and its progeny because 222Rn progeny have a higher estimated equilibrium factor and higher 

associated dose factors than the 220Rn thoron progeny (DOE 2001a, pp. 31–35), an estimate of the 

220Rn release during the operational period was made. Natural thorium was processed on a batch 

basis in the refinery, and oxide production and firing systems at WSP from November 1963 through 

September 1966. Using the thoron intake calculation as stated in ORAUT-TKBS-0028-5 (ORAUT 

2017b), a daily 220Rn intake of 7.94E–06 WLM/d (2.9E–03 WLM/yr divided by 365.25 d/yr) should be 

assigned to WSP personnel from 1963 to 1966. This has been prorated for the partial years of 

thorium operations in 1963 and 1966 as shown in Tables 4-3 and 4-4. Similar to the rationale stated 

above, this thoron intake activity is applied to the WSQ as well. There were no thoron intakes 

assigned for the WSRP during the operational period since the source term was small and the 

contents of the pits were covered with water.
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 17 of 38

 

4.2.3.2 Postoperational Years, 1967 to 2002

DOE had no contractors on the site until 1975, when air monitoring occurred in the WSRP and WSQ 

areas. Measurements of radon in air appear in annual reports beginning with monitoring reports for 

1979 and 1980 (Weidner and Boback 1982, pp. 28–30, 33, 50), which were specific to the raffinate 

pits and quarry. In October 1985, the radon monitoring program was expanded to include WSP (DOE

1986, pp. 34–35), which was transferred back to DOE at that time. Radioactive airborne particulates 

were not monitored before the beginning of remediation activities at the site in 1985. Air particulate 

monitoring was added beginning in 1987 at WSP and WSRP (DOE 1988, pp. 77–83) and beginning in 

1989 at WSQ (DOE 1990a, pp. 125–133). Air sampling data from the annual monitoring reports from 

1985 to 2000 (when remediation was nearly complete), were the basis for the following estimates of 

air concentrations and intakes at WSRP, WSP, and WSQ for periods after 1974.

4.2.3.2.1 Airborne Particulate Radionuclides

WSP

Airborne particulate results were reported at WSP beginning in 1987 (DOE 1988, pp. 77–83; 1989a, 

pp. 90–92; 1990a, pp. 125–133; 1991, pp. 131–135; 1992a, pp. 98–101; 1993, pp. 128–132; 1994, 

pp. 106–111; 1995, pp. 104–108; 1996, pp. 96–99; 1997, pp. 92–97; 1998a, pp. 78–81; 1999, pp. 86–

90; 2000, pp. 81–85; 2001b, pp. 79–83). From 1987 to 1989, monitoring of particulates at a few 

perimeter locations for WSP indicated that measured gross alpha concentrations were statistically 

indistinguishable from background (DOE 1988, p. 81; 1989a, pp. 90–92; 1990a, pp. 125–133). 

Although a significant source term was not expected, the concentrations from 1985 to 1989 were set 

equal to the 1990 concentration value because it was the highest nonzero value for the adjacent year.

WSRP

Airborne particulate results were reported at WSRP beginning in 1987 (DOE 1988, pp. 77–83; 1989a, 

pp. 90–92; 1990a, pp. 125–133; 1991, pp. 131–135; 1992a, pp. 98–101; 1993, pp. 128–132; 1994, 

pp. 106–111; 1995, pp. 104–108; 1996, pp. 96–99; 1997, pp. 92–97; 1998a, pp. 78–81; 1999, pp. 86–

90; 2000, pp. 81–85; 2001b, pp. 79–83). From 1987 to 1989, monitoring of particulates at a few 

perimeter locations for WSRP indicated that measured gross alpha concentrations were statistically 

indistinguishable from background (DOE 1988, p. 81; 1989a, pp. 90–92; 1990a, pp. 125–133). 

Therefore, the concentrations from 1987 to 1989 at WSRP are zero. Because all areas of WSRP 

were essentially undisturbed between 1975 and the beginning of remediation activities in 1985, and 

because no significant airborne concentration was likely because the source term was small and the 

contents of the pits were covered with water, this analysis assumed that radioactive airborne

particulate concentrations before 1987 were the same as those measured in 1987 through 1989. 

Therefore, the particulate concentrations from 1975 through 1989 were zero.

WSQ

The airborne concentrations from 1975 to 1987 were set equal to the value for 1965 because there 

were no data for those years and the 1965 value is the closest measured year. Airborne particulate 

results were reported at WSQ beginning in 1988 (DOE 1988, pp. 77–83; 1989a, pp. 90–92; 1990a, 

pp. 125–133; 1991, pp. 131–135; 1992a, pp. 98–101; 1993, pp. 128–132; 1994, pp. 106–111; 1995, 

pp. 104–108; 1996, pp. 96–99; 1997, pp. 92–97; 1998a, pp. 78–81; 1999, pp. 86–90; 2000, pp. 81–

85; 2001b, pp. 79–83). In 1988 and 1989 at WSQ, monitoring of particulates at a few perimeter 

locations indicated that measured gross alpha concentrations were statistically indistinguishable from 

background (DOE 1989a, p. 92; DOE 1990a, p. 133). Therefore, the concentrations in 1988 and

1989 at WSQ were assumed to be zero.

WSP, WSRP, and WSQ

From 1989 to 2000, concentrations of radioactive air particulates were measured and reported at the 

locations in Figures 4-3 and 4-4, even though many of the concentrations were statistically 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 18 of 38

 

indistinguishable from background. The measurements occurred at perimeter locations for WSP, 

WSRP, and WSQ. These perimeter data were extrapolated to represent concentrations at the center 

of the operating area using the method in Section 4.2.3.1. Therefore, the measured perimeter 

concentrations of airborne particulate radionuclides were multiplied by a factor of 30 to represent the 

average onsite airborne concentrations.

Figure 4-3. Air particulate (P), radon (R), and gamma (G) monitoring stations in 

WSP and WSRP (modified from DOE 1999, p. 69).

Section 4.2.1.1 identifies the radionuclides of concern as the naturally occurring isotopes of uranium 

(

234U, 235U, and 238U) and their decay products (primarily 230Th and 226Ra) as potentially significant 

isotopes of concern during the remediation period. However, air concentrations of these isotopes 

were measured in terms of gross alpha concentrations at WSP for most years of interest. The lack of 

refinement is probably because (1) the radionuclides of interest are alpha emitters and (2) the 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 19 of 38

 

perimeter data for WSP, WSRP, and WSQ, which are all that are available for air particulates, seldom 

showed air concentration measurements higher than those at offsite background monitoring stations. 

Figure 4-4. Air particulate (P), radon (R), and gamma (G) monitoring stations in WSQ (modified from 

DOE 1999, p. 70).

Air particulate monitoring ended at the end of 2000 for all WSP areas because radioactive waste 

handling activities were essentially complete and no critical receptor air monitoring data had 

demonstrated a dose to the public of greater than 1 mrem (DOE 2001b, pp. 95, 230–237; DOE 2003, 

pp. 33, 44). Because there was still some movement of material at the site and workplace monitors 

were used for the brief period of placement of waste material in the Disposal Cell in 2001 (DOE 2002, 

pp. 8–9), the intakes for 2001 were assumed to be equal to the intakes for 2000.

Table 4-2 lists RU contaminant intakes per unit activity of uranium. Table 4-3 lists the median intakes 

for WSP, WSRP, and WSQ, and Table 4-4 lists the maximum sitewide median 234U intakes for the 

remediation period. These intakes correspond to the highest value, by year, for WSP, WSRP, and 

WSQ. The daily intakes were obtained by multiplying the 234U concentrations (Bq/m3

) by an assumed 

inhalation rate of 2,400 m3

/yr (Section 9.3.1.1 of ICRP 1994), adjusted by dividing by an average of 

365.25 day in a calendar year (ORAUT 2020). The RU contaminant intake activities were calculated 

as stated in Section 4.2.4 by multiplying the ratio stated in Table 4-2 by the 234U value, either that 

stated in Table 4-3 or Table 4-4. As stated earlier, 232Th was estimated for 1964–1966 by scaling 

the annual uranium and thorium mass throughput rates and apportioning that fraction of the intake
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 20 of 38

 

rate to thorium. Thorium-228 and 228Ra intake rates were included for those years at ratios stated

earlier.

Table 4-2. RU contaminant intakes per unit activity of uranium, Bq/Bq U (ORAUT 2017c).

Radionuclide 1961–1962 1963–1965 1966–1975 1976–1985 1986–2001c

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

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

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

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

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

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

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

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

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

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

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

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

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

weapons-grade plutonium mixture.

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

c. The year 2002 is not listed, although a covered period, because all environmental doses in 2002 are zero.

4.2.3.2.2 Radon

WSP

During the postoperational years, outdoor radon was measured at WSP, WSRP, and WSQ (Weidner 

and Boback 1982, pp. 28–30, 33, 50; Bechtel 1983a, pp. 22–25; 1983b, pp. 22-27; DOE 1984a, 

pp. 23–26; 1985a, pp. 24–48; 1986, pp. 34–41; 1987, p. 58–63; 1988, pp. 66–73; 1989b, pp. 143–

147; 1989a, pp. 80–84; 1990b, pp. 8–17; 1990a, pp. 110–119; 1991, pp. 105, 115–127; 1992a, 

pp. 75–90; 1992b, pp. 276–277, 431; 1993, pp. 114–124; 1994, pp. 92–103; 1995, pp. 87–10; 1996, 

pp. 75–91; 1997, pp. 70–88; 1998a, pp. 58–73; 1998b, pp. 32–36; 1999, pp. 67–82; 2000, pp. 63–78; 

2001b, pp. 64–76). At WSP, only perimeter measurements occurred except in 1999 and 2000, when 

some measurements occurred inside the perimeter where the Disposal Cell received much of the 

remediation wastes. Because there was still some movement of material at the site and workplace 

monitors were used for the brief period of placement of waste material in the Disposal Cell in 2001 

(DOE 2002, pp. 8, 9), the intakes for 2001 were assumed to be equal to the intakes for 2000.

This TBD assumes the average measured radon concentrations at the perimeter stations were 

underestimates of the average radon concentration over the WSP onsite areas during the remediation 

period. Therefore, just as for particulates, the measured concentrations were multiplied by 30 to 

provide an average net radon concentration for the WSP. Table 4-3 lists the derived concentrations, 

which include contributions of radon from naturally occurring 226Ra and 228Ra. These concentrations 

reflect total radon (222Rn and 220Rn) as measured by F-type alpha track detectors.
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4 Effective Date: 03/27/2020 Page 21 of 38

Table 4-3. Daily median intakes for WSP, WSQ, and WSRP.

Year

WSRP 

U-234a 

(Bq/d)

WSRP 

radonb 

(WLM/d)

WSP 

U-234 

(Bq/d)

WSP 

Th-230 

(Bq/d)

WSP 

natural Th 

(Bq/d)

WSP

Rn-220

(WLM/d)

WSP 

Rn-222b

(WLM/d)

WSQ 

U-234 

(Bq/d)

WSQ

Rn-220

(WLM/d)

WSQ 

Rn-222b

(WLM/d)

1957–1959 0.00E+00c 0.00E+00c 1.86E+00 1.49E+00 0.00E+00 0.00E+00 3.39E–02d 0.00E+00 0.00E+00 0.00E+00

1960 0.00E+00c 0.00E+00c 3.79E+00 3.03E+00 0.00E+00 0.00E+00 3.39E–02d 0.00E+00 0.00E+00 0.00E+00

1961 0.00E+00c 0.00E+00c 6.51E+00 5.20E+00 0.00E+00 0.00E+00 3.39E–02d 9.53E–03 0.00E+00 0.00E+00

1962 0.00E+00c 0.00E+00c 1.60E+00 1.28E+00 0.00E+00 0.00E+00 3.39E–02d 6.46E–03 0.00E+00 0.00E+00

1963 0.00E+00c 0.00E+00c 2.90E+00 2.33E+00 0.00E+00 1.33E–06 3.39E–02d 6.97E–03 1.33E–06 3.39E–02d

1964 0.00E+00c 0.00E+00c 2.22E+00 1.77E+00 1.72E–03 7.94E–06 3.39E–02d 1.62E–02 7.94E–06 3.39E–02d

1965 0.00E+00c 0.00E+00c 1.21E+00 9.66E–01 3.19E–02 7.94E–06 3.39E–02d 1.08E–01 7.94E–06 3.39E–02d

1966 0.00E+00c 0.00E+00c 1.21E+00 9.66E–01 1.04E–01 5.94E–06 3.39E–02d 1.08E–01 5.94E–06 3.39E–02d

1967–1974 0.00E+00f 0.00E+00f 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 0.00E+00 0.00E+00f 0.00E+00f

1975–1976 0.00E+00c 1.11E–04 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 5.20E–05

1977–1978 0.00E+00c 1.59E–04 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 5.28E–05

1979 0.00E+00c 2.90E–05 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 2.90E–05

1980 0.00E+00c 1.80E–04 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 4.75E–05

1981 0.00E+00c 1.18E–04 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 6.09E–05

1982 0.00E+00c 2.38E–05 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 6.92E–05

1983 0.00E+00c 1.50E–05 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 3.13E–05

1984 0.00E+00c 2.57E–05 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00 0.00E+00g 1.08E–01 0.00E+00 3.58E–05

1985 0.00E+00c 1.71E–05 4.91E–03 3.92E–03 0.00E+00h 0.00E+00 3.21E–04 1.08E–01 0.00E+00 3.58E–05

1986 0.00E+00c 2.43E–05 4.91E–03 3.92E–03 0.00E+00h 0.00E+00 4.94E–04 1.08E–01 0.00E+00 3.42E–05

1987 0.00E+00 2.00E–05 4.91E–03 3.92E–03 0.00E+00h 0.00E+00 6.42E–04 1.08E–01 0.00E+00 6.42E–05

1988 0.00E+00 3.13E–05 4.91E–03 3.92E–03 0.00E+00h 0.00E+00 8.78E–04 0.00E+00 0.00E+00 9.72E–05

1989 0.00E+00 2.36E–05 4.91E–03 3.92E–03 0.00E+00h 0.00E+00 8.15E–04 0.00E+00 0.00E+00 6.14E–05

1990 1.79E–04 1.43E–05 4.91E–03 3.92E–03 0.00E+00h 0.00E+00 5.14E–04 6.11E–05 0.00E+00 4.99E–05

1991 9.00E–05 1.57E–05 2.96E–03 2.37E–03 0.00E+00h 0.00E+00 4.28E–04 1.18E–05 0.00E+00 4.28E–05

1992 7.03E–05 1.57E–05 2.21E–03 1.77E–03 0.00E+00h 0.00E+00 3.42E–04 0.00E+00 0.00E+00 3.87E–05

1993 8.28E–05 7.13E–06 2.48E–03 1.98E–03 0.00E+00h 0.00E+00 1.59E–04 1.12E–05 0.00E+00 8.72E–05

1994 9.72E–05 2.00E–05 3.40E–03 2.71E–03 0.00E+00h 0.00E+00 3.63E–04 8.61E–05 0.00E+00 2.43E–04

1995 8.48E–05 2.43E–05 2.75E–03 2.20E–03 0.00E+00h 0.00E+00 4.70E–04 2.88E–05 0.00E+00 8.28E–05

1996 1.12E–04 4.07E–05 3.18E–03 2.54E–03 0.00E+00h 0.00E+00 5.77E–04 1.18E–05 0.00E+00 1.86E–05

1997 1.35E–04 9.27E–05 2.96E–03 2.37E–03 0.00E+00h 0.00E+00 4.70E–04 0.00E+00 0.00E+00 1.57E–05

1998 1.56E–04 1.64E–05 4.15E–03 3.32E–03 0.00E+00h 0.00E+00 2.79E–04 9.33E–06 0.00E+00 1.06E–05

1999 1.39E–04 2.64E–05 4.09E–03 3.28E–03 0.00E+00h 0.00E+00 5.35E–04 1.33E–06 0.00E+00 2.43E–05

2000 7.36E–05 1.43E–05 2.75E–03 2.20E–03 0.00E+00h 0.00E+00 5.56E–04 1.18E–05 0.00E+00 1.86E–05

2001 7.36E–05 1.43E–05 2.75E–03 2.20E–03 0.00E+00h 0.00E+00 5.56E–04 1.18E–05 0.00E+00 1.86E–05

2002 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

a. Data are reported as gross alpha rather than uranium. NU is assumed through 1962, and EU is assumed thereafter.

b. Radon refers to both Rn-220 and Rn-222, represented as 100% Rn-222, and includes natural background contribution.

c. No significant airborne concentration was likely because the source term was small and the contents of the pits were covered with water.

d. Assumed radon intake value for the operational period at the WSP.

e. Before 1963, the quarry contained only drummed waste that was typically submerged in water and was therefore a negligible contributor.

f. There were no AEC contractors on site during this period. The National Lead Company of Ohio began environmental monitoring at the WSRP and WSQ in August 1975.

g. DOD control period.

h. No source term for natural thorium.
Document No. ORAUT-TKBS-0028-4 Revision No. 0

4 Effective Date: 03/27/2020 Page 22 of 38

Table 4-4. Maximum sitewide daily median intakes.

Year

U-234a

(Bq/d)

Th-228 

(Bq/d)

Ra-228

(Bq/d)

Th-230 

(Bq/d)

Th-232 

(Bq/d)

Pu-239 

(Bq/d)

Pu-241 

(Bq/d)

Am-241 

(Bq/d)

Np-237 

(Bq/d)

Tc-99 

(Bq/d)

Ru-106 

(Bq/d)

Zr-95 

(Bq/d)

Nb-95 

(Bq/d)

Sr-90 

(Bq/d)

Rn-220

(WLM/d)

Rn-222b

(WLM/d)

1957–1959 1.86E+00 0.00E+00 0.00E+00 1.49E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.39E–02

1960 3.79E+00 0.00E+00 0.00E+00 3.03E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.39E–02

1961 6.51E+00 7.34E–06 0.00E+00 5.20E+00 1.05E–05 7.61E–03 7.85E–02 5.05E–06 2.63E–03 9.72E–01 1.05E+00 1.58E–01 1.58E–01 4.20E–02 0.00E+00 3.39E–02

1962 1.60E+00 1.81E–06 0.00E+00 1.28E+00 2.58E–06 1.88E–03 1.93E–02 1.24E–06 6.48E–04 2.40E–01 2.59E–01 3.88E–02 3.88E–02 1.04E–02 0.00E+00 3.39E–02

1963 2.90E+00 2.30E–06 4.60E–06 2.33E+00 3.28E–06 2.38E–03 2.46E–02 1.58E–06 8.24E–04 3.04E–01 3.29E–01 4.94E–02 4.94E–02 1.32E–02 1.33E–06 3.39E–02

1964 2.22E+00 1.72E–03f 3.44E–03f 1.77E+00 1.72E–03e 1.82E–03 1.88E–02 1.21E–06 6.30E–04 2.33E–01 2.52E–01 3.77E–02 3.77E–02 1.01E–02 7.94E–06 3.39E–02

1965 1.21E+00 3.20E–02f 6.41E–02f 9.66E-01 3.20E–02e 9.93E–04 1.02E–02 6.59E–07 3.43E–04 1.27E–01 1.37E–01 2.05E–02 2.05E–02 5.48E–03 7.94E–06 3.39E–02

1966 1.21E+00 1.05E–01f 2.09E–01f 9.66E-01 1.05E–01e 9.84E–04 6.32E–03 1.29E–04 3.43E–04 1.27E–01 1.37E–01 2.05E–02 2.05E–02 5.48E–03 5.94E–06 3.39E–02

1967–1974 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c 0.00E+00c

1975 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.77E–05 5.64E–04 1.15E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 1.11E–04

1976 1.08E–01 8.54E–08 0.00E+00 0.00E+00d 1.22E–07 8.70E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.89E–04 0.00E+00 1.11E–04

1977 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 1.59E–04

1978 1.08E–01 8.54E–08 0.00E+00 0.00E+00d 1.22E–07 8.70E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.89E–04 0.00E+00 1.59E–04

1979 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 2.90E–05

1980 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 1.80E–04

1981 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 1.18E–04

1982 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 6.92E–05

1983 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 3.13E–05

1984 1.08E–01 8.53E–08 0.00E+00 0.00E+00d 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 3.58E–05

1985 1.08E–01 8.53E–08 0.00E+00 3.92E–03 1.22E–07 8.69E–05 3.49E–04 1.84E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 3.21E–04

1986 1.08E–01 8.53E–08 0.00E+00 3.92E–03 1.22E–07 8.64E–05 2.16E–04 2.25E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 4.94E–04

1987 1.08E–01 8.53E–08 0.00E+00 3.92E–03 1.22E–07 8.64E–05 2.16E–04 2.25E–05 3.06E–05 1.13E–02 1.22E–02 1.83E–03 1.83E–03 4.88E–04 0.00E+00 6.42E–04

1988 4.91E–03 3.88E–09 0.00E+00 3.92E–03 5.55E–09 3.93E–06 9.84E–06 1.02E–06 1.39E–06 5.15E–04 5.56E–04 8.34E–05 8.34E–05 2.22E–05 0.00E+00 8.78E–04

1989 4.91E–03 3.88E–09 0.00E+00 3.92E–03 5.55E–09 3.93E–06 9.84E–06 1.02E–06 1.39E–06 5.15E–04 5.56E–04 8.34E–05 8.34E–05 2.22E–05 0.00E+00 8.15E–04

1990 4.91E–03 3.88E–09 0.00E+00 3.92E–03 5.55E–09 3.93E–06 9.84E–06 1.02E–06 1.39E–06 5.15E–04 5.56E–04 8.34E–05 8.34E–05 2.22E–05 0.00E+00 5.14E–04

1991 2.96E–03 2.35E–09 0.00E+00 2.37E–03 3.35E–09 2.38E–06 5.94E–06 6.18E–07 8.41E–07 3.11E–04 3.36E–04 5.04E–05 5.04E–05 1.34E–05 0.00E+00 4.28E–04

1992 2.21E–03 1.75E–09 0.00E+00 1.77E–03 2.50E–09 1.77E–06 4.42E–06 4.61E–07 6.26E–07 2.31E–04 2.50E–04 3.75E–05 3.75E–05 1.00E–05 0.00E+00 3.42E–04

1993 2.48E–03 1.96E–09 0.00E+00 1.98E–03 2.80E–09 1.99E–06 4.96E–06 5.17E–07 7.03E–07 2.60E–04 2.81E–04 4.21E–05 4.21E–05 1.12E–05 0.00E+00 1.59E–04

1994 3.40E–03 2.69E–09 0.00E+00 2.71E–03 3.84E–09 2.72E–06 6.81E–06 7.09E–07 9.64E–07 3.56E–04 3.85E–04 5.77E–05 5.77E–05 1.54E–05 0.00E+00 3.63E–04

1995 2.75E–03 2.17E–09 0.00E+00 2.20E–03 3.11E–09 2.20E–06 5.50E–06 5.73E–07 7.79E–07 2.88E–04 3.11E–04 4.67E–05 4.67E–05 1.24E–05 0.00E+00 4.70E–04

1996 3.18E–03 2.52E–09 0.00E+00 2.54E–03 3.60E–09 2.55E–06 6.37E–06 6.64E–07 9.02E–07 3.33E–04 3.60E–04 5.40E–05 5.40E–05 1.44E–05 0.00E+00 5.77E–04

1997 2.96E–03 2.35E–09 0.00E+00 2.37E–03 3.35E–09 2.38E–06 5.94E–06 6.18E–07 8.41E–07 3.11E–04 3.36E–04 5.04E–05 5.04E–05 1.34E–05 0.00E+00 4.70E–04

1998 4.15E–03 3.29E–09 0.00E+00 3.32E–03 4.69E–09 3.33E–06 8.32E–06 8.66E–07 1.18E–06 4.35E–04 4.70E–04 7.06E–05 7.06E–05 1.88E–05 0.00E+00 2.79E–04

1999 4.09E–03 3.24E–09 0.00E+00 3.28E–03 4.63E–09 3.28E–06 8.20E–06 8.54E–07 1.16E–06 4.29E–04 4.64E–04 6.96E–05 6.96E–05 1.86E–05 0.00E+00 5.35E–04

2000 2.75E–03 2.17E–09 0.00E+00 2.20E–03 3.11E–09 2.20E–06 5.50E–06 5.73E–07 7.79E–07 2.88E–04 3.11E–04 4.67E–05 4.67E–05 1.24E–05 0.00E+00 5.56E–04

2001 2.75E–03 2.17E–09 0.00E+00 2.20E–03 3.11E–09 2.20E–06 5.50E–06 5.73E–07 7.79E–07 2.88E–04 3.11E–04 4.67E–05 4.67E–05 1.24E–05 0.00E+00 5.56E–04

2002 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.00E+00 0.00E+00 0.00E+00 0.00E+00

a. Data are reported as gross alpha rather than uranium. NU is assumed through 1962, and EU is assumed thereafter.

b. Radon refers to both Rn-220 and Rn-222, represented as 100% Rn-222, and includes natural background contribution.

c. WSP was not a covered facility under EEOICPA and there were no AEC contractors at WSRP or WSQ during this period.

d. DOD control period.

e. Values pertain to processed natural thorium. All other Th-232 values from 1961 to 2001 are RU contaminant values.

f. Values are associated with the processed Th-232 values at ratios of 1:1 for Th-232:Th-228 and 2:1 for Ra-228:Th-232. 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 23 of 38

 

WSRP

Radon measurements in WSRP in the late 1970s and early 1980s (before remediation) averaged 

34 ±37 Bq/m3

, which indicates that the pits were not a major source of radon (Meshkov et al. 1986, 

p. 101). This value is limiting for the operational period because the amount of 226Ra in the pits was at 

its maximum at the end of operations and it would not have decayed or leached out of the pits 

significantly before the measurements were made.

At WSRP, radon measurements occurred at the perimeter monitoring stations between 1985 and 

2000, at one location inside the perimeter during 1985 and 1986, and at several interior locations 

during the years of active remediation (1998 to 2000). This TBD assumes the average measured 

radon concentrations at the perimeter stations were underestimates of the average radon 

concentration over the WSRP onsite areas during the remediation period. Therefore, just as for 

particulates, the measured concentrations were multiplied by 30 to provide an average net radon 

concentration for the WSRP. Table 4-3 lists the derived concentrations, which include contributions of 

radon from naturally occurring 226Ra and 228Ra. These concentrations reflect total radon (222Rn and 

220Rn) as measured by F-type alpha track detectors.

WSQ

Measurements in the quarry area in the late 1970s and early 1980s (before remediation) averaged 

24 ±15 Bq/m3

, which indicates the quarry was not a major source of radon (Meshkov et al. 1986, 

p. 101).

At WSQ, radon measurements occurred only at the perimeter monitoring stations except in 1989, 

when monitors were placed on the quarry floor. The ratio of the average perimeter radon 

concentration to the average quarry floor concentration was 30. This is a significant increase, but the 

interior of the quarry is fairly inaccessible, with a quick drop in elevation and an abundance of 

vegetation. Therefore, this TBD assumes that the perimeter concentrations are reasonably 

representative of exposure concentrations except during the active removal of the quarry bulk waste 

in 1993, 1994, and 1995. During this period of active remediation, this TBD assumed the radon 

concentration to be 30 times the concentration measured at the perimeter (ORAUT 2017c).

4.2.4 Radionuclide Intake Derivation

For the purposes of dose reconstruction, the assumption was made that all uranium processed at 

WSP was NU from 1957 through 1962, with a specific activity of 683 pCi/mg. After 1962, all uranium 

was assumed to be enriched to 1%, with a specific activity of 973 pCi/mg (ORAUT 2017b). Most of 

the RU at WSP came from the Feed Materials Production Center, which in turn received most of its 

RU from the Hanford Site, which was recycled from uranium that had been irradiated to produce

weapons-grade plutonium (ORAUT 2017c). Therefore, a 6% weapons-grade mixture that had been 

chemically purified in 1961 was selected for dose calculations (ORAUT 2017b; ORAUT 2017c). For 

dose reconstruction, the plutonium alpha dose from the plutonium mixtures in RU is assumed to be 

from 100% 239Pu. Plutonium mixture ratios for fresh, 10-, 20-, and 30-year-aged intervals after 

purification were used to estimate the 241Pu and 241Am contaminant levels (ORAUT 2017b). Table 4-2 

lists the RU contaminant intakes per unit activity of uranium.

The TBD analysis used Equation 4-2 to convert the radon concentrations into intake values for 

periods outside the operational period (i.e., after 1966). The WLM/d intake per Bq/m3 for 222Rn was 

calculated as:

(4-2) 222 6

3

/ 0 3 2 000 2 61 10

/ 3,700 170 365 25

WLM d . , Rn . Bq m .× − = = ×

× ×
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 24 of 38

 

where 0.3 is the equilibrium factor (DOE 2001a, p. 31), 2,000 is the number of hours in a working 

year, 3,700 is the conversion factor of Bq/m3 per 100 pCi/L, 170 is the number of hours in a working 

month, and 365.25 is the average number of days in a calendar year. One Working Level is equal to 

100 pCi/L of 222Rn. This is adapted from ORAUT (2017b). 

Table 4-1 lists the median airborne radioactivity concentrations used to calculate the radionuclide 

intake activities. Table 4-2 lists the intake activity ratios of the RU contaminants to uranium. 

Table 4-3 lists the median intake values for WSP, WSRP, and WSQ, with radon represented as 222Rn. 

Table 4-4 lists maximum sitewide median intakes across WSP to be represented in the Interactive 

RadioEpidemiological Program (IREP) as lognormal distributions with a geometric standard deviation 

(GSD) of 3. The intakes for Tables 4-3 and 4-4 are stated in units of activity per calendar day

(ORAUT 2020). 

The purpose of Table 4-3 is to assign the WSP worker an estimate of intake for various radionuclides 

for the three areas (raffinate pits, chemical plant, and quarry). This table can also be used to 

determine environmental internal doses specific to the worker’s work site. The values in Table 4-2 

can be used with the uranium values in Table 4-3 to determine the RU contaminant activities 

associated with the three WSP areas. The purpose of Table 4-4 is to give the DR the maximum 

median intake for a specific year for use in determining the worker’s dose when the specific work 

location is not known. Table 4-4 would most likely be used by the DR because the calculated dose is 

favorable to the claimant and expedites completion of the dose reconstruction.

4.3 EXTERNAL EXPOSURE TO ONSITE ATMOSPHERIC RADIONUCLIDE

CONCENTRATIONS

External ambient exposure at WSP was a result of gamma and X-ray radiation from radionuclides in 

the ore concentrate that were stored on the site during the operational period and from radionuclides 

in WSRP and WSQ. Section 4.3.1 identifies ambient exposures during the operational period, and 

Section 4.3.2 provides ambient exposures during the postoperational years.

Onsite ambient doses at WSP are to be calculated as best estimates. DRs use a Monte Carlo 

technique where the uncertainty distributions associated with the onsite ambient dose, either 

lognormal or normal, is multiplied by the appropriate organ dose conversion factor (DCF), which is a 

triangular distribution. It is assumed that the worker worked 2,500 hours during the year, 50 hours per 

week and 50 weeks per year, prorated for partial year employment and scaled as necessary for 

periods of no exposure potential as determined from the telephone interview or employment records 

(ORAUT 2018b). The 2,500 work hours per year assumption is also based on a computer-assisted 

telephone interview query using the first 10 claims which included employment at any of the three 

facilities (Weldon Spring Plant, Quarry, and Raffinate Pits) and the 20 most recent claims as of July 

16, 2019. Nineteen of the selected energy employees indicated overtime work, while the remaining 

11 did not.

4.3.1 Operational Period, 1957 to 1966

Available documents do not contain monitoring data that describe the ambient exposure rate at WSP 

during the operational period. During this period, ambient dose rates in excess of natural background 

radiation would have been due primarily to gamma radiation from short-lived decay products of 238U in 

the vicinity of the storage pad for the drums that contained the ore concentrate (Meshkov et al. 1986, 

p. 46). Exposure to radionuclides that accumulated in WSRP and WSQ would be mitigated 

somewhat by the shielding and physical barrier that was provided by water that covered the pits 

entirely and some of the waste in the quarry for much of the year.
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Employees who worked in radiological areas were monitored, and their exposures are accounted for 

in their normal dosimetry results. Because all employees in operational areas were required to wear 

film dosimeters, there should not have been unmonitored employees subject to exposure in those 

areas (NIOSH 2010b, pp. 61–64). Because no site-specific ambient exposure data for WSP has been 

discovered for the operational period, DRs should consider applying the 50th-percentile value

(equivalent to median or geometric mean) of the occupational external dose (Table 4-5) for monitored 

workers during the operational period as the onsite ambient dose for unmonitored workers (ORAUT 

2019a). Potential ambient environmental dose for unmonitored workers, such as unbadged 

administrative personnel, is bounded by the 50th-percentile value of the data for monitored 

individuals. There are no median values for 1967 because there were no measured external doses 

associated with the pool of WSP claimants. DRs should adjust the dose by prorating for partial year 

employment periods and/or partial year external dose monitoring. Although these dose values could 

be considered favorable to the claimant, in lieu of site-specific ambient exposure information during 

this period, they are assigned as best estimates.

Table 4-5. Median values for occupational 

external gamma dose.

Year

Median gamma dose

(rem)a GSD

1957 0.161 2.63

1958 0.081 2.15

1959 0.078 2.44

1960 0.126 2.27

1961 0.136 2.94

1962 0.139 2.80

1963 0.220 2.92

1964 0.187 3.14

1965 0.151 3.71

1966 0.123 3.16

a. Calculated as the 50th-percentile value of a 

lognormal distribution in accordance with 

Battelle-TIB-5000 (BT 2007; ORAUT 2019a). 

4.3.2 Postoperational Years, 1967 to 2002

DOE had no contractors on the site until 1975 when it began air monitoring at WSRP and WSQ. The 

results of monitoring for ambient exposure from 1975 to 1981 are not available, so the analysis 

estimated ambient doses for the years from 1975 to 1981 from later data as explained below.

A 1975 aerial radiological survey of WSP (Jobst 1976, p. 32) indicated that the average background 

gamma exposure rate was 7 to 10 μR/hr. For continuous exposure over a year, this represents an 

annual ambient dose of 61 to 88 mrem/yr. However, the aerial survey found elevated terrestrial 

exposures over WSRP and WSQ, which indicated the presence of manmade changes from the 

natural radioisotopes (NIOSH 2010b, pp. 61–62). 

Ambient exposure rates in excess of natural background rates existed before remediation in areas 

close to waste in WSRP or WSQ (Jobst 1976, p. 32–33). Based on the results of the aerial survey, 

gamma radiation exposure rate isopleths were drawn in the immediate vicinity of the plant and 

centered on the raffinate pits. This indicated total exposure rates from 120 to 168 μR/hr with 

decreasing levels from 15.8 to 35.8 μR/hr at approximately 1,000 ft from the pits (Jobst 1976, p. 32).
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4.3.2.1 Weldon Spring Plant

Between 1982 and 2000, ambient exposure was monitored using thermoluminescent dosimeters at 

many perimeter locations around WSP, WSRP, and WSQ (Bechtel 1983a, pp. 26–27; DOE 1984a, 

pp. 27–28; 1985a, pp. 28–30; 1986, pp. 40–44; 1987, pp. 64–66, 58–61; 1988, pp. 73–76; 1989b, pp. 

147–149; 1989a, pp. 85–90; 1990a, pp. 119–125; 1991, pp. 127–131; 1992a, pp. 90–98; 1992b, pp. 

276–277, 432; 1993, pp. 125–128; 1994, pp. 103–106; 1995, pp. 100–104; 1996, pp. 92–95; 1997, 

pp. 89–92; 1998a, pp. 74–78; 1998b, pp. 32–36; 1999, pp. 82–86; 2000, pp. 78–81; 2001b, pp. 76–

79). In 1986, locations in WSP and WSRP were monitored, but the results did not indicate a

significant difference in exposure rate from the perimeter stations. Figure 4-3 shows the locations of 

monitoring stations.

Reported ambient exposure values included background except for 1985 when an average 

background rate of 99 mrem/yr was subtracted from the total measured values at the monitoring 

stations. Many of the reported values for 1985 were zero, representing a net value (exposure rate 

minus background rate) of either zero or less than zero. To use these numbers for 1985, the analysis 

added the background value of 99 to the reported net value. Therefore, the average ambient 

exposures for 1985 might be a slight overestimate due to the addition of background to a zero value 

that actually might have been less than zero.

A 1987 radiologic characterization of WSP and WSRP (Marutzky, Colby, and Cahn 1988, pp. 159–

168) used portable scintillation counters to record external radiation in counts per minute at 

880 locations across WSRP and WSP. Correlation of those counts with a pressurized ionization

chamber allowed estimates of exposure rate in microroentgen per hour. Exposure rates ranged from 

9 to 287 μR/hr (at 1 m), which correspond to dose rates between 79 and 2,500 mrem/yr for 

continuous exposure for 8,760 hr/yr. Fewer than 10% of the 880 surveyed locations had exposure 

rates of 15 μR/hr or greater. The 95th percentile of the exposure rates is approximately 18 μR/hr. 

This is reasonable because few of the exposure rates over 15 μR/hr exceed 20 μR/hr, and many of 

the contaminated areas were probably not those in which occupational exposures would normally 

have occurred. The elevated exposure rates were in the areas around the buildings, onsite dump 

areas, the Ash Pond, and drainages of Frog Pond, but it is not possible to sort out exposure rates as a 

function of location in the reports. An exposure rate of 18 μR/hr represents an annual dose of 

158 mrem/yr for continuous exposure. In 1987, the reported WSP perimeter exposure rate was 

72 mrem/yr. Therefore, the TBD analysis assumed that, until remediation of WSP was essentially 

complete in 2000, the average onsite exposure rate was approximately twice the measured perimeter 

rate.

4.3.2.2 Weldon Spring Raffinate Pits

Measured exposure rates are not available for 1975 through 1982 for WSRP. It is reasonable to 

assume that the average of the exposure rates from 1983 through 1989 for WSRP adequately 

represent these earlier years, when the entire site was essentially undisturbed. 

A comprehensive radiological survey of the WSRP in 1982 and 1983 indicated an average onsite 

gamma exposure rate of 23 μR/hr for that site (DOE 1984b, p. 22), which gives a dose rate of 

approximately 200 mrem/yr for continuous exposure (8,760 hr/yr). This is higher than the perimeter 

monitors reported for the WSRP before remediation. For 1975 through 1983, 201 mrem/yr is the 

average survey value for 1982 to 1983 from DOE (1984b, p. 22) because it is considered the best 

value to use for those years. A comparison of the average survey-based value (201 mrem/yr, 

adjusted to 57 mrem/yr for a 2,500-hour work year) to the average perimeter value in 1983 

(88 mrem/yr) (DOE 1984a, p. 27) suggests the area inside the WSRP perimeter is more appropriately 

characterized by an exposure rate approximately twice the average perimeter values. The modified 

ambient dose rates in Table 4-6 represent twice the average perimeter values for 1984 through 2000 
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 27 of 38

 

and are recommended for environmental dose reconstruction. Figure 4-3 shows the locations of 

monitoring stations. From 2001 to 2002, after the completion of remediation of most of WSP (DOE 

2001b, p.79), the measured dose rate result has not been factored. 

4.3.2.3 Weldon Spring Quarry

Measured exposure rates are not available for 1975 through 1981 for WSQ. It is reasonable to 

assume that the average of the exposure rates from 1982 through 1989 for WSQ adequately 

represent these earlier years, when the entire site was essentially undisturbed. 

A similar radiological survey of WSQ in 1984 and 1985, before remediation, indicated gamma 

exposure rates from 8 μR/hr (similar to the overall area background rate) to 286 μR/hr over the quarry 

floor where contamination was greatest (DOE 1985b, p. 29). This single maximum result is provided 

to indicate the upper measured exposure rate in the quarry. The survey report stated that 

characterization of WSQ was “extremely difficult because of the rough terrain,” and that “the area was 

densely vegetated” (DOE 1985b). Figure 4-5 includes a topographic map of the WSQ area, along 

with locations of gamma exposure rate measurements. The survey included many of the relatively 

inaccessible areas of the quarry, where significant exposure was unlikely except during excavation of 

quarry bulk wastes. Therefore, the perimeter exposure rates in Table 4-6 are appropriate for 

estimating exposures except during the excavation period, which occurred during most of 1993, 1994, 

and 1995 (DOE 1994, pp. 44–45; 1995, p. 48; 1996, pp. 50–51). Figure 4-5 shows the locations of 

monitoring stations. In Table 4-5, the average value as presented in annual environmental reports is 

considered the best value for dose reconstruction.

Table 4-6 lists the average onsite ambient dose rates when specific work locations are known. 

Table 4-7 lists the sitewide maximum external exposure rates among the three sites and should be 

used when the work locations are either unknown or when the energy employee worked at multiple 

facilities.

4.3.3 Organ Dose Conversion Factors

The calculated annual dose should be multiplied by the appropriate DCFs in OCAS-IG-001, External 

Dose Reconstruction Implementation Guideline (NIOSH 2007) or ORAUT-OTIB-0088, External Dose 

Reconstruction (ORAUT 2018b), for an isotropic exposure geometry. For the operational period from 

(1957 to1966) and initial cleanup period (1967 to 1969), the exposure (R)-to-organ DCF is used. For 

the period from 1975 to 1983, the ambient dose equivalent-to-organ DCF is used because those data 

are based on survey measurements or calculations (ORAUT 2018b). For the period after 1983, the 

exposure (R)-to-organ DCF is used.

4.4 UNCERTAINTY

For external doses and the operational period from 1957 through 1966, the median of a lognormal 

distribution and its associated GSD are assigned from Table 4-5 as Parameter 1 and 2 in the 

Interactive RadioEpidemiological Program, respectively. For the postoperational external doses, the 

onsite ambient dose is represented as a normal distribution with a standard deviation of 30% (Section 

6.7.1 of ORAUT 2012).

For internal doses, the environmental internal doses are assigned as lognormal distributions with a 

GSD of 3.
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Table 4-6. Estimated onsite ambient doses.

a 

Year

WSRP average 

(mrem/yr)b 

WSP average 

(mrem/yr)c 

WSQ average 

(mrem/yr)d 

1957–1966 (e) (e) Not significant

1975–1981 57 (f) 28

1982 57 (f) 24

1983 57 (f) 27

1984 70 (f) 36

1985 65 57 35

1986 52 45 28

1987 40 41 27

1988 34 36 21

1989 37 40 23

1990 36 38 22

1991 40 43 24

1992 39 39 21

1993 33 36 19

1994 34 35 19

1995 41 40 19

1996 47 39 18

1997 53 39 17

1998 40 36 16

1999 35 35 16

2000 18 32 14

2001 18g 32g 14g

2002 0 0 0

a. Based on 8,760 hr/yr exposure; adjusted for 2,500-hour work year; includes 

contributions from natural background.

b. From 1984 to 2000, average value is twice the average of the measured values 

at perimeter locations for the WSRP.

c. Before 2001, average value is twice the average of the measured values at 

perimeter locations for the WSP.

d. Values are the average of the measured values at perimeter locations for the 

WSQ.

e. Apply the 50th-percentile occupational external dose from Table 4-5 or evaluate 

on a case-by-case basis. 

f. WSP did not transfer to DOE until 1985.

g. Not monitored after 2000; assumed equal to average for 2000 for respective 

location.
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Figure 4-5. Topographic map and locations of gamma exposure rate 

measurements for WSQ (DOE 1985b, pp. 14, 17).

4.5 SUMMARY ENVIRONMENTAL DOSE DEFAULT VALUES 

This section summarizes the tables that list the recommended default values for inhalation intakes 

and ambient dose. Section 4.2.2.1 refers to ORAUT-TKBS-0017-5 (ORAUT 2017c) in relation to the 

RU contaminants and their mass concentrations to be applied beginning in 1961. 

• Table 4-1 lists the median airborne concentrations for WSP, WSRP, and WSQ.

• Table 4-2 lists the RU contaminant-to-uranium intake ratios. 

• Table 4-3 lists the daily median intake values for WSP, WSRP, and WSQ.
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• Table 4-4 lists maximum sitewide estimated daily median inhalation intakes of radioactive air 

particulates and radon. 

• Table 4-5 lists the 50th-percentile values for external environmental dose for unmonitored 

workers during the operational period. Although potentially favorable to claimants, they are 

considered best estimates.

• Table 4-6 lists estimated average ambient onsite doses for WSRP, WSP, and WSQ.

• Table 4-7 provides the maximum sitewide estimates of ambient dose, which should be used 

for a best estimate when the work locations are either unknown or include multiple locations.

Table 4-7. Estimated maximum sitewide ambient 

dose at WSP.a,b

Year

Maximum for 

WSRP, WSP, and WSQ 

(mrem/yr)

1957–1966 (c)

1975–1983 57

1984 70

1985 65

1986 52

1987 41

1988 36

1989 40

1990 38

1991 43

1992 39

1993 36

1994 35

1995 41

1996 47

1997 53

1998 40

1999 35

2000 32

2001 32d

2002 0

a. Maximum of annual recommended average value for 

each area (from Table 4-5). 

b. Based on 8,760-hr/yr exposure, adjusted for a 2,500-

hour work year.

c. Apply the 50th-percentile occupational external dose 

from Table 4-5 or evaluate on a case-by-case basis.

d. Not monitored after 2000; assumed to be equal to the 

sitewide maximum for WSP for 2000.

DRs should consider the naturally occurring isotopes of uranium (234U, 235U, and 238U) and their 

progeny (primarily 230Th and 226Ra) when calculating and assigning doses to unmonitored workers 

from the occupational environmental isotopic air species.

4.6 ATTRIBUTIONS AND ANNOTATIONS

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

section of this document.
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GLOSSARY

alpha particle

See alpha radiation.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

becquerel (Bq)

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals 

37 billion (3.7 × 1010) Bq.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

exposure

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

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

gamma radiation

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

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

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

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

isotope

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

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

Isotopes have very nearly the same chemical properties.

nuclide

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

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

unstable nuclide.

radiation

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

matter through various mechanisms that involve energy transfer. 

radioactivity

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

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

nuclei. See radionuclide.

radionuclide

Radioactive nuclide. See nuclide.

raffinate

A product that has had a component or components chemically removed. In this case, waste 

from uranium or thorium extraction as well as solids from the neutralization of this waste.
Document No. ORAUT-TKBS-0028-4 Revision No. 04 Effective Date: 03/27/2020 Page 38 of 38

rem

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

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

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

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

special uranium microcurie

The sum of 3.7 × 104 disintegrations per second from 238U, 3.7 × 104 disintegrations per 

second from 234U, and 9 × 102 disintegrations per second from 235U.

thermoluminescent dosimeter

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

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

a measurement of absorbed dose. Thermoluminescent dosimeters replaced film dosimeters 

at essentially all U.S. Department of Energy sites beginning in the 1960s.

X-ray radiation

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

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

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

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