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DOE Review Release 09/15/2017

Weldon Spring Plant – Occupational 

Environmental Dose

ORAUT-TKBS-0028-4 Rev. 03

Effective Date: 09/08/2017

Supersedes: Revision 02

Subject Expert(s): David P. Harrison

Document Owner 

Approval: Signature on File

David P. Harrison, Document Owner

Approval Date: 09/07/2017

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 09/07/2017

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 09/07/2017

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 09/08/2017

Approval: Signature on File

Stuart L. Hinnefeld, Director, DCAS

Approval Date: 09/08/2017

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 2 of 34

 

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. 
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TABLE OF CONTENTS

SECTION TITLE PAGE

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

4.1 Introduction .............................................................................................................................. 7

4.1.1 Purpose........................................................................................................................ 8

4.1.2 Scope........................................................................................................................... 8

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations...................................... 9

4.2.1 Radionuclides of Concern............................................................................................. 9

4.2.1.1 Airborne Particulate Radionuclides ................................................................ 9

4.2.1.2 Radon ............................................................................................................ 9

4.2.2 Source Terms for Airborne Radionuclides .................................................................... 9

4.2.2.1 Operational Period, 1957 to 1966................................................................... 9

4.2.2.2 Postoperational Years, 1967 to 2002 ........................................................... 11

4.2.3 Annual Intake of Radionuclides .................................................................................. 11

4.2.3.1 Operational Period, 1957 to 1966................................................................. 11

4.2.3.2 Postoperational Years, 1967 to 2002 ........................................................... 15

4.2.4 Radionuclide Intake Derivation ................................................................................... 19

4.3 External Exposure to Onsite Atmospheric Radionuclide Concentrations ................................ 22

4.3.1 Operational Period, 1957 to 1966 ............................................................................... 22

4.3.2 Postoperational Years, 1967 to 2002.......................................................................... 23

4.3.3 Organ Dose Conversion Factors ................................................................................ 26

4.4 Summary Environmental Dose Default Values....................................................................... 26

4.5 Attributions and Annotations .................................................................................................. 27

References ........................................................................................................................................ 28

Glossary ............................................................................................................................................ 33

LIST OF TABLES

TABLE TITLE PAGE

4-1 RU contaminant intakes per unit activity of uranium, Bq/Bq U................................................ 19

4-2 Annual median intakes for WSP, WSRP, AND WSQ ............................................................. 20

4-3 Maximum sitewide (WSP, WSQ, and WSRP) annual median intakes .................................... 21

4-4 Median values for occupational external gamma dose ........................................................... 22

4-5 Estimated ambient onsite doses............................................................................................. 24

4-6 Estimated maximum sitewide ambient dose at WSP.............................................................. 26

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Perimeter sampling locations and processing facilities........................................................... 13
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4-2 WSP and WSRP layout with superimposed 600-m radius circle to demonstrate the 

distance from the plant center to the perimeter ...................................................................... 14

4-3 Air particulate, radon, and gamma monitoring stations in WSP and WSRP............................ 16

4-4 Air particulate, radon, and gamma monitoring stations in WSQ.............................................. 17

4-5 Topographic map and locations of gamma exposure rate measurements for WSQ ............... 25
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 5 of 34

 

ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

AWE atomic weapons employer

Bq becquerel

Ci curie

cm centimeter

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

hr hour

m meter

mg milligram

mrem millirem

MT metric ton

NIOSH National Institute for Occupational Safety and Health

NU natural uranium

ORAU Oak Ridge Associated Universities

pCi picocurie

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

WLM working level month

WSP Weldon Spring Plant

WSQ Weldon Spring Quarry

WSRP Weldon Spring Raffinate Pits

yr year

μCi microcurie
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 6 of 34

 

μm micrometer

μR microroentgen

§ section or sections
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 7 of 34

 

4.1 INTRODUCTION

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

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

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

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

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

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

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

of the individual work required for each dose reconstruction.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

dose reconstructions for covered employees (NIOSH 2010a):

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

conventional structures

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

reasons
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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 as a result of submersion and ambient exposure at WSP.

4.1.2 Scope

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

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

Main Site. 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 the 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.”

Workers at the WSP received environmental doses during the operational period as a result of stack 

effluents from buildings, contamination of soil as a result of stack releases, and onsite storage of ore 

concentrates. Workers at the WSQ received environmental doses during the same period as a result 

of 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 the WSQ, the WSP, and the 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 the 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 summarizes the 

tables that provide environmental dose default values for use by dose reconstructors (DRs). 

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

justification, or clarification of the associated information, are presented in Section 4.5.
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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 4.4 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.

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, and 

their decay products. ORAUT-TKBS-0028-5, Weldon Spring Plant – Occupational Internal Dose

(ORAUT 2017c) 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. The risks associated with 219Rn, due to its extremely short half-life (4 seconds), were 

insignificant in comparison with those from 222Rn and 220Rn and are not relevant. The same is true for 

the small amount of precursor radium due to the limited amount of 235U WSP processed. Therefore, 

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

discussed in more detail below.

4.2.2 Source Terms for Airborne Radionuclides

4.2.2.1 Operational Period, 1957 to 1966

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 

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 (Meshkov et al. 1986, pp. 47–48) assuming (1) 5,000 to 14,500 MT 

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 an upper end estimate), (4) radon was in 

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

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.
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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 the WSP. The discharge from the acid recovery plant was the 

primary source of radon emission.

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. Impurities that were generated in the purification process before denitration were bled 

off in liquid form in the raffinate, which was pumped to the WSRP. 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) 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 Fernald and WSP effectively removed the radium but were not 

effective in removing thorium, specifically 230Th. 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; Sears 1975, p. 143; 1976, p. 4). 

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

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

thorium and radium (Author unknown, 1967, pp. 10–11). Wastes in the 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 

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.

Wastes that were disposed of in the 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 220

Rn. 

Therefore, the radon release before 1963 was negligibly small. 
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In 1963 and 1964, an estimated 38,000 m3 of uranium- and radium-contaminated rubble, equipment, 

and soil were placed in the 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.

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 2017a). For the periods that include RU (i.e., after 

1960), the DR should refer to ORAUT-TKBS-0017-5, Feed Materials Production Center – 

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

the mass concentrations associated with the period from 1961 to 1972 as stated in Table 5-10 of that 

document should be used because the plutonium out of specification in the tower ash (sent to Fernald 

from the Paducah Gaseous Diffusion Plant) was not sent to the WSP (ORAUT 2017b, 

Section 5.3.3.3).

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 the 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. In addition, 

DRs should select the absorption type that yields the highest dose to the organ of interest. Calculated 

annual intakes are based on a 2,400-m3

/yr inhalation rate (ICRP 1994) using an hourly breathing rate 

of 1.2 m3 for light activity and a 2,000-hour work year. Environmental intakes from ingestion are 

considered negligibly small. Intakes can be scaled to a different rate or to consider partial year 

exposures.

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.

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 the 

average airborne particulate concentrations are presented in ORAUT-TKBS-0028-5 (ORAUT 2017c). 

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 for an expanded discussion of 

the use of the special uranium microcurie at WSP. Other radionuclides were not measured.

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 

concentration at two different locations of interest using the Pasquill-Gifford equation for Gaussian 

plume atmospheric dispersion:
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( ) 2 2

2 2

2

1 1

exp 2 2 y y z

X x,y y H

Q πσ σ u σ σ

   

= −+          

(4-1)

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

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

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.

When Equation 4-1 is solved for 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.
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Figure 4-2. WSP and WSRP layout with superimposed 600-m 

radius circle to demonstrate the distance from the plant center 

to the perimeter (modified from Army 1976, p. 176). 

Thorium intake, for those years when it was processed, can be estimated by scaling the annual 

uranium and thorium mass throughput rates 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.

The measured perimeter air concentration data for the 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. 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.

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

perimeter air concentrations were the same as the measured concentrations in 1959, the operational 
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 15 of 34

 

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 2017c), an annual 222Rn intake of 

12.4 WLM/yr 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.

A similar estimate for 220Rn release from the WSP during the operational period was not made. Harris 

(1986, p. 86) states, “Natural thorium was typically received in either a nitrate or oxide form.” In 

addition, Harris (pp. 23, 28) indicates that the amount of processed thorium material ranged from a 

low of 0.05% of the processed natural uranium (NU) in 1964 to a high of 5.4% in 1965. Assuming 

concentrated thorium feed (i.e., 70% thorium) and secular equilibrium among 232Th, 224Ra, and 220Rn, 

the 220Rn release could approach 70% of the 222Rn release. The estimated dose attributable to 220Rn 

and its progeny is 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).

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 were reported in annual reports beginning with a monitoring 

report 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 

the WSP (DOE 1986, pp. 34–35), which was transferred back to DOE at that time. Air particulate 

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

beginning in 1989 at the 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 the WSRP, WSP, and WSQ for periods after 

1974.

Airborne Particulate Radionuclides

Airborne particulate results were reported at the WSP and WSRP beginning in 1987 and at the WSQ 

beginning in 1989 (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).

ORAUT-TKBS-0028-5 (ORAUT 2017c) lists air monitoring data for airborne particulate radionuclides 

during the postoperational years. The data are reported as gross alpha rather than uranium. 

Because more than 99% of the material WSP received was NU, it is likely that the gross alpha data 

represents the NU concentration but may be assumed to be enriched uranium (EU) for dose 

calculation purposes. Figures 4-3 and 4-4 show the locations of monitoring stations for the WSRP 
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 16 of 34

 

and WSP and for the WSQ, respectively. Radioactive air particulates were not monitored before the 

beginning of remediation activities at the site in 1985. In 1987 and 1988, monitoring of particulates at 

a few perimeter locations for the WSRP and WSP and in 1989 at the WSQ indicated that measured 

gross alpha concentrations were statistically indistinguishable from background (DOE 1988, p. 81; 

1989a, p. 92; 1990a, p. 133). Therefore, the concentrations in 1987 and 1989 at the WSRP and WSP

and in 1989 at the WSQ are zero. Because all areas of the WSP were essentially undisturbed 

between 1975 and the beginning of remediation activities in 1985, this analysis assumed that 

radioactive air particulate concentrations before 1987 were the same as those measured in 1987 

through 1989. Therefore, ORAUT-TKBS-0028-5 lists particulate concentrations from 1975 through 

1986 as insignificant (ORAUT 2017c). 

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

WSP and WSRP (modified from DOE 1999, p. 69).
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 17 of 34

 

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

DOE 1999, p. 70).

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 

indistinguishable from background. The measurements occurred at perimeter locations for the 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 in ORAUT-TKBS-0028-5 (ORAUT 2017c).

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 the WSP for most years of interest. The 

lack of refinement is probably because (1) the radionuclides of interest are alpha emitters and (2) the 

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

seldom showed air concentration measurements higher than those at offsite background monitoring 

stations. Some isotope-specific data appear in annual reports in the late 1990s but, because the ratio 

of isotopes probably would be highly variable by location due to the variability in isotopic composition 

of the contaminated soil, the utility of these data is limited. 

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. The Weldon Spring 
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 18 of 34

 

Site Remediation Program was effectively complete in 2002. The site has since been transferred for 

long-term surveillance and maintenance, and it is not covered under EEOICPA beyond 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 are zero beginning in 2002.

Table 4-2 lists the annual median intakes for WSP, WSRP, and WSQ. Table 4-3 lists the maximum 

sitewide median 234U intakes for the remediation period. These intakes correspond to the highest 

value, by year, for the WSP, WSRP, and WSQ. The intakes were obtained by multiplying the 234U 

concentrations (Bq/m3

) by an assumed inhalation rate of 2,400 m3/yr (ORAUT 2013). The RU 

contaminant intake activities were calculated as stated in Section 4.2.4 by multiplying the 234U value, 

either that stated in Table 4-2 or Table 4-3, by the ratio stated in Table 4-1.

Radon

Radon measurements in the 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.

During the postoperational years, outdoor radon was measured at the 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 the 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. 

At the 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 WSP and 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 each of the two areas (WSP and WSRP). ORAUT-TKBS-0028-5 (ORAUT 

2017c) 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.

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

when monitors were placed inside the perimeter down to the quarry floor. The ratio of the average 

perimeter radon concentration to the average interior 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 
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 19 of 34

 

bulk waste in 1993, 1994, and 1995. This TBD assumes the radon concentration during this period of 

active remediation 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 683 pCi/mg. After 1962, all uranium 

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

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

RU from the Hanford Site, which was recycled from weapons-grade plutonium. Therefore, a 6% 

weapons-grade mixture that had been chemically purified in 1961 was selected for dose calculations. 

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 2017c). The 

following table lists the RU contaminant intakes per unit activity of uranium.

Table 4-1. 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.

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

periods outside the operational period from 1957 through 1966. The WLM/yr intake per Bq/m3 for 

222Rn was calculated as:

222 4

3

WLM/yr 0 3 2 000 Rn 9 54 10

Bq/m 37 000 170

. , . ,

× − = = ×

×

(4-2)

Table 4-2 lists the annual median radon intake values for the WSP, WSRP, and WSQ. Table 4-3 lists 

maximum sitewide median intakes across WSP to be represented in the Interactive 

RadioEpidemiological Program as lognormal distributions with a geometric standard deviation of 3.

The purpose of Table 4-2 is to assign the WSP worker an estimate of annual intake for various 

radionuclides for the three areas. This table can also be used to determine environmental internal 

doses specific to the worker’s work site. The values in Table 4-1 can be used with the uranium values 

in Table 4-2 to determine the RU contaminant activities associated with the three WSP areas 

(raffinate pits, chemical plant, quarry). The purpose of Table 4-3 is to give the DR the maximum 

median intake for a specific year for use in determining the worker’s dose. Table 4-3 would most 

likely be used by the DR because the calculated dose is favorable to the claimant and expedites 

completion of the dose reconstruction.
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 20 of 34

 

Table 4-2. Annual median intakes for WSP, WSRP, and WSQ.

Year

WSRP 

U-234a 

(Bq/yr)

WSRP 

radonb 

(WLM/yr)

WSP 

U-234 

(Bq/yr)

WSP 

Th-230 

(Bq/yr)

WSP 

natural Th 

(Bq/yr)

WSP 

radonb 

(WLM/yr)

WSQ 

U-234 

(Bq/yr)

WSQ 

radonb 

(WLM/yr)

1957–

1959

0.00E+00c 0.00E+00c 6.78E+02 5.43E+02 0.00E+00 1.24E+01d 0.00E+00e 0.00E+00e

1960 0.00E+00c 0.00E+00c 1.38E+03 1.11E+03 0.00E+00 1.24E+01d 0.00E+00e 0.00E+00e

1961 0.00E+00c 0.00E+00c 2.37E+03 1.90E+03 0.00E+00 1.24E+01d 3.47E+00 0.00E+00e

1962 0.00E+00c 0.00E+00c 5.85E+02 4.68E+02 0.00E+00 1.24E+01d 2.36E+00 0.00E+00e

1963 0.00E+00c 0.00E+00c 1.06E+03 8.49E+02 0.00E+00 1.24E+01d 2.56E+00 1.24E+01d

1964 0.00E+00c 0.00E+00c 8.11E+02 6.49E+02 6.28E–01 1.24E+01d 5.90E+00 1.24E+01d

1965 0.00E+00c 0.00E+00c 4.42E+02 3.54E+02 1.17E+01 1.24E+01d 3.93E+01 1.24E+01d

1966 0.00E+00c 0.00E+00c 4.42E+02 3.54E+02 3.82E+01 1.24E+01d 3.93E+01 1.24E+01d

1967–

1974

0.00E+00f 0.00E+00f 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00f 0.00E+00f

1975–

1976

0.00E+00c 4.06E–02 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 1.90E–02

1977–

1978

0.00E+00c 5.79E–02 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 1.93E–02

1979 0.00E+00c 1.06E–02 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 1.06E–02

1980 0.00E+00c 6.56E–02 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 1.74E–02

1981 0.00E+00c 4.30E–02 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 2.22E–02

1982 0.00E+00c 8.69E–03 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 2.53E–02

1982 0.00E+00c 1.00E–02 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 2.10E–02

1984 0.00E+00c 1.72E–02 0.00E+00g 0.00E+00g 0.00E+00g 0.00E+00g 3.93E+01 2.38E–02

1985 0.00E+00c 1.14E–02 3.28E+00 2.62E+00 0.00E+00h 2.15E–01 3.93E+01 2.38E–02

1986 0.00E+00c 1.62E–02 3.28E+00 2.62E+00 0.00E+00h 3.29E–01 3.93E+01 2.29E–02

1987 0.00E+00 1.34E–02 3.28E+00 2.62E+00 0.00E+00h 4.29E–01 3.93E+01 4.29E–02

1988 0.00E+00 2.10E–02 3.28E+00 2.62E+00 0.00E+00h 5.87E–01 0.00E+00 6.49E–02

1989 0.00E+00 1.57E–02 3.28E+00 2.62E+00 0.00E+00h 5.44E–01 0.00E+00 4.10E–02

1990 1.20E–01 9.54E–03 3.28E+00 2.62E+00 0.00E+00h 3.43E–01 0.00E+00 3.34E–02

1991 6.00E–02 1.05E–02 1.98E+00 1.58E+00 0.00E+00h 2.86E–01 0.00E+00 2.86E–02

1992 4.68E–02 1.05E–02 1.48E+00 1.18E+00 0.00E+00h 2.29E–01 0.00E+00 2.58E–02

1993 5.52E–02 4.77E–03 1.66E+00 1.32E+00 0.00E+00h 1.06E–01 0.00E+00 5.82E–02

1994 6.48E–02 1.34E–02 2.27E+00 1.81E+00 0.00E+00h 2.43E–01 0.00E+00 1.62E–01

1995 5.64E–02 1.62E–02 1.84E+00 1.47E+00 0.00E+00h 3.15E–01 0.00E+00 5.53E–02

1996 7.44E–02 2.72E–02 2.12E+00 1.70E+00 0.00E+00h 3.86E–01 0.00E+00 1.24E–02

1997 9.00E–02 6.20E–02 1.98E+00 1.58E+00 0.00E+00h 3.15E–01 0.00E+00 1.05E–02

1998 1.04E–01 1.10E–02 2.77E+00 2.22E+00 0.00E+00h 1.86E–01 0.00E+00 7.06E–03

1999 9.24E–02 1.76E–02 2.74E+00 2.19E+00 0.00E+00h 3.58E–01 0.00E+00 1.62E–02

2000 4.92E–02 9.54E–03 1.84E+00 1.47E+00 0.00E+00h 3.72E–01 0.00E+00 1.24E–02

2001 4.92E–02 9.54E–03 1.84E+00 1.47E+00 0.00E+00h 3.72E–01 0.00E+00 1.24E–02

2002 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00h 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 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.
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Table 4-3. Maximum sitewide (WSP, WSQ, and WSRP) annual median intakes.

Year

U-234a

(Bq/yr)

Th-228

(Bq/yr)

Th-230 

(Bq/yr)

Th-232 

(Bq/yr)

Pu-239

(Bq/yr)

Pu-241

(Bq/yr)

Am-241

(Bq/yr)

Np-237

(Bq/yr)

Tc-99

(Bq/yr)

Ru-106

(Bq/yr)

Zr-95

(Bq/yr)

Nb-95

(Bq/yr)

Sr-90

(Bq/yr)

Radonb 

(WLM/yr)

1957–1959 6.78E+02 0.00E+00 5.43E+02 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 1.24E+01

1960 1.38E+03 0.00E+00 1.11E+03 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 1.24E+01

1961 2.37E+03 2.67E–03 1.90E+03 3.82E–03 2.77E+00 2.86E+01 1.84E–03 9.58E–01 3.54E+02 3.83E+02 5.74E+01 5.74E+01 1.53E+01 1.24E+01

1962 5.85E+02 6.60E–04 4.68E+02 9.42E–04 6.85E–01 7.06E+00 4.54E–04 2.36E–01 8.74E+01 9.44E+01 1.42E+01 1.42E+01 3.78E+00 1.24E+01

1963 1.06E+03 8.39E–04 8.49E+02 1.20E–03 8.70E–01 8.98E+00 5.77E–04 3.01E–01 1.11E+02 1.20E+02 1.80E+01 1.80E+01 4.80E+00 1.24E+01

1964 8.11E+02 6.28E–01 6.49E+02 6.28E–01e 6.66E–01 6.87E+00 4.42E–04 2.30E–01 8.50E+01 9.19E+01 1.38E+01 1.38E+01 3.68E+00 1.24E+01

1965 4.42E+02 1.17E+01 3.54E+02 1.17E+01e 3.63E–01 3.74E+00 2.41E–04 1.25E–01 4.63E+01 5.01E+01 7.51E+00 7.51E+00 2.00E+00 1.24E+01

1966 4.42E+02 3.82E+01 3.54E+02 3.82E+01e 3.60E–01 2.31E+00 4.72E–02 1.25E–01 4.63E+01 5.01E+01 7.51E+00 7.51E+00 2.00E+00 1.24E+01

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

1975 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.20E–02 2.06E–01 4.20E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 4.06E–02

1976 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 4.06E–02

1977 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 5.79E–02

1978 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 5.79E–02

1979 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 1.06E–02

1980 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 6.56E–02

1981 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 4.30E–02

1982 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 2.53E–02

1982 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 2.10E–02

1984 3.93E+01 3.11E–05 0.00E+00d 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 2.38E–02

1985 3.93E+01 3.11E–05 2.62E+00 4.44E–05 3.17E–02 1.27E–01 6.70E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 2.15E–01

1986 3.93E+01 3.11E–05 2.62E+00 4.44E–05 3.15E–02 7.88E–02 8.20E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 3.29E–01

1987 3.93E+01 3.11E–05 2.62E+00 4.44E–05 3.15E–02 7.88E–02 8.20E–03 1.11E–02 4.12E+00 4.45E+00 6.68E–01 6.68E–01 1.78E–01 4.29E–01

1988 3.28E+00 2.60E–06 2.62E+00 3.71E–06 2.63E–03 6.57E–03 6.84E–04 9.30E–04 3.44E–01 3.72E–01 5.57E–02 5.57E–02 1.49E–02 5.87E–01

1989 3.28E+00 2.60E–06 2.62E+00 3.71E–06 2.63E–03 6.57E–03 6.84E–04 9.30E–04 3.44E–01 3.72E–01 5.57E–02 5.57E–02 1.49E–02 5.44E–01

1990 3.28E+00 2.60E–06 2.62E+00 3.71E–06 2.63E–03 6.57E–03 6.84E–04 9.30E–04 3.44E–01 3.72E–01 5.57E–02 5.57E–02 1.49E–02 3.43E–01

1991 1.98E+00 1.57E–06 1.58E+00 2.24E–06 1.59E–03 3.97E–03 4.13E–04 5.62E–04 2.08E–01 2.24E–01 3.36E–02 3.36E–02 8.97E–03 2.86E–01

1992 1.48E+00 1.17E–06 1.18E+00 1.67E–06 1.19E–03 2.97E–03 3.09E–04 4.20E–04 1.55E–01 1.68E–01 2.52E–02 2.52E–02 6.71E–03 2.29E–01

1993 1.66E+00 1.31E–06 1.32E+00 1.88E–06 1.33E–03 3.33E–03 3.46E–04 4.71E–04 1.74E–01 1.88E–01 2.82E–02 2.82E–02 7.52E–03 1.06E–01

1994 2.27E+00 1.80E–06 1.81E+00 2.57E–06 1.82E–03 4.55E–03 4.74E–04 6.44E–04 2.38E–01 2.57E–01 3.86E–02 3.86E–02 1.03E–02 2.43E–01

1995 1.84E+00 1.46E–06 1.47E+00 2.08E–06 1.48E–03 3.69E–03 3.84E–04 5.22E–04 1.93E–01 2.08E–01 3.13E–02 3.13E–02 8.34E–03 3.15E–01

1996 2.12E+00 1.68E–06 1.70E+00 2.40E–06 1.70E–03 4.25E–03 4.42E–04 6.01E–04 2.22E–01 2.40E–01 3.60E–02 3.60E–02 9.61E–03 3.86E–01

1997 1.98E+00 1.57E–06 1.58E+00 2.24E–06 1.59E–03 3.97E–03 4.13E–04 5.62E–04 2.08E–01 2.24E–01 3.36E–02 3.36E–02 8.97E–03 3.15E–01

1998 2.77E+00 2.19E–06 2.22E+00 3.13E–06 2.22E–03 5.55E–03 5.78E–04 7.86E–04 2.90E–01 3.14E–01 4.71E–02 4.71E–02 1.26E–02 1.86E–01

1999 2.74E+00 2.17E–06 2.19E+00 3.10E–06 2.20E–03 5.49E–03 5.72E–04 7.77E–04 2.87E–01 3.10E–01 4.66E–02 4.66E–02 1.24E–02 3.58E–01

2000 1.84E+00 1.46E–06 1.47E+00 2.08E–06 1.48E–03 3.69E–03 3.84E–04 5.22E–04 1.93E–01 2.08E–01 3.13E–02 3.13E–02 8.34E–03 3.72E–01

2001 1.84E+00 1.46E–06 1.47E+00 2.08E–06 1.48E–03 3.69E–03 3.84E–04 5.22E–04 1.93E–01 2.08E–01 3.13E–02 3.13E–02 8.34E–03 3.72E–01

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 1.00E+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 and includes natural background contribution.

c. The WSP was not a covered facility under EEOICPA and there were no AEC contractors at the 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.
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4.3 EXTERNAL EXPOSURE TO ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

External ambient exposure at the 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 the WSRP and WSQ.

A 1975 aerial radiological survey of the WSP (Jobst 1976, p. 32) indicated that, for most of the areas, 

the normal terrestrial gamma exposure rate was 3 to 6 μR/hr at 1 m above ground level and the 

average cosmic exposure rate was approximately 4 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 the WSRP and the WSQ, which indicated the presence of manmade 

changes from the natural radioisotopes (NIOSH 2010b, pp. 61–62). Section 4.3.1 identifies ambient 

exposures during the operational period, and Section 4.3.2 provides ambient exposures during the 

postoperational years. 

4.3.1 Operational Period, 1957 to 1966

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

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 the 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.

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 the WSP has 

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

the occupational external dose (Table 4-4) for monitored workers during the operational period as 

favorable to claimants. 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.

Table 4-4. Median values for occupational 

external gamma dose.

Year

50th-percentile gamma 

(mrem)a 

1957 161

1958 81

1959 78

1960 126

1961 136

1962 139

1963 220

1964 187

1965 151

1966 123

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

lognormal distribution in accordance with 

Battelle-TIB-5000 (BT 2007).
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4.3.2 Postoperational Years, 1967 to 2002

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

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

many perimeter locations around the 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 the WSP and the WSRP were monitored, but the results did not 

indicate a significant difference in exposure rate from the perimeter stations. Figures 4-3 and 4-4 

show 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.

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

close to waste in the 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 exposure rates from 116 to 164 μR/hr with decreasing 

levels from 11.8 to 31.8 μR/hr at approximately 1,000 ft from the pits. These measurements represent 

terrestrial radiation only. The average cosmic exposure rate of 4 μR/hr should be added to the

isopleths to compute the total exposure rate for personnel within the site boundary (Jobst 1976, 

p. 32).

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. A comparison of the average survey-based 

value (201 mrem/yr) 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 4-5 

represent twice the average perimeter values for 1984 through 2000 and are recommended for 

environmental dose reconstruction. From 2001 to 2002, after the completion of remediation of most 

of the WSP (DOE 2001b, p.79), the measured dose rate result has not been factored. For 1975 

through 1983, the value of 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 similar radiological survey of the 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 the 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 
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Table 4-5. Estimated ambient onsite doses.

a

Year

WSRP average 

(mrem/yr)b 

WSP average 

(mrem/yr)c 

WSQ average 

(mrem/yr)d 

Background 

average (mrem/yr)

1957–1966 (e) (e) Not significant 69

1975–1981 201 (f) 97 69

1982 201 (f) 85 69

1983 201 (f) 95 72

1984 245 (f) 127 110

1985 228 199 123 114

1986 183 157 97 82

1987 139 145 95 78

1988 120 126 73 58

1989 131 140 81 64

1990 127 133 77 59

1991 140 150 84 67

1992 135 137 73 67

1993 116 125 67 56

1994 118 122 65 55

1995 144 140 66 62

1996 165 135 62 60

1997 186 135 61 66

1998 139 126 57 57

1999 123 124 55 55

2000–2001 63g 111g 48g 52

2002 0 0 0 0

a. Based on 8,760 hr/yr exposure; 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 4-3 or evaluate on a case-by-case basis. 

Workers assigned to these areas were monitored; ambient dose was accounted for as part of occupational 

dose.

f. WSP did not transfer to DOE until 1985.

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

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 4-5 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). For Table 4-5, the 

average value as presented in annual environmental reports is considered the best value for dose 

reconstruction.

A 1987 radiologic characterization of the 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 the WSRP and WSP. Correlation of those counts with a pressurized ionization 

chamber allowed estimates of exposure rate in μR/hr. Exposure rates ranged from 9 to 287 μR/hr (at 

1 m), which corresponds to a dose rate 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. 

Assuming 10% of the locations were at 85 μR/hr and the remaining 90% were at approximately 

11 μR/hr produces an average exposure rate of approximately 18 μR/hr. This is a reasonable 

assumption for the WSP 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 
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Figure 4-5. Topographic map and locations of gamma exposure rate 

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

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 the WSP was essentially 

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

rate. Table 4-5 lists values reflecting this assumption.

Measured exposure rates are not available for 1975 through 1981 for the WSQ or for 1975 through 

1982 for the WSRP. It is reasonably assumed that the average of the exposure rates from 1982 

through 1989 for the WSQ and from 1983 through 1989 for the WSRP adequately represent these 

earlier years, when the entire site was essentially undisturbed. Table 4-5 lists these averaged rates. 

Table 4-6 lists sitewide maximum external exposure rates for WSP during the remediation period.
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Table 4-6. Estimated maximum sitewide 

ambient dose at WSP.a,b

Year

Maximum for WSRP, WSP, 

and WSQ (mrem/yr)

1957–1966 (c)

1975–1983 201

1984 245

1985 228

1986 183

1987 145

1988 126

1989 140

1990 133

1991 150

1992 137

1993 125

1994 172

1995 144

1996 165

1997 186

1998 139

1999 124

2000–2001 111d

2002 0

a. Maximum of annual recommended average value 

for each area (from Table 4-4).

b. Based on 8,760-hr/yr exposure.

c. Apply the 50th-percentile occupational external 

dose from Table 4-3 or evaluate on a case-bycase basis.

d. Not monitored after 2000; assumed to be equal 

to the sitewide maximum for the WSP for 2000.

4.3.3 Organ Dose Conversion Factors

The measured photon dose is used with the dose conversion factors (DCFs) to calculate annual 

organ doses of interest using OCAS-IG-003, External Dose Reconstruction Implementation Guideline

(NIOSH 2007), for an isotropic exposure geometry. For the operational period (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. For the period after 1983, the 

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

4.4 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 2017b) in relation to the 

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

• Table 4-2 lists the annual median intake values for the WSP, WSRP, and WSQ. 

• Table 4-3 lists maximum sitewide estimated median inhalation intakes of radioactive air 

particulates and radon. 

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

workers during the operational period. 
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• Table 4-5 lists estimated average ambient onsite doses for WSRP, WSP, and WSQ. 

• Table 4-6 provides the maximum sitewide estimates of ambient dose. 

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.

4.5 ATTRIBUTIONS AND ANNOTATIONS

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

section of this document.
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REFERENCES

Army (U.S. Department of the Army), 1976, Assessment of Weldon Spring Chemical Plant in 

St. Charles, County, Missouri, Office of the Deputy Assistant Project Manager for Chemical 

Demilitarization and Installation Restoration, Aberdeen Proving Ground, Maryland, March 1. 

[SRDB Ref ID: 32709]

Author unknown, 1967, Weldon Spring Raffinate Pits and Quarry Task Force Report, June 1. [SRDB 

Ref ID: 14763]

Bechtel (Bechtel National), 1983a, Weldon Spring Site (WSS) Environmental Monitoring Report, 

Calendar Year 1982, Oak Ridge, Tennessee, May. [SRDB Ref ID: 13376]

Bechtel (Bechtel National), 1983b, Weldon Spring Site (WSS) Environmental Monitoring Report, 

Calendar Year 1981, Oak Ridge, Tennessee, May. [SRDB Ref ID: 13338]

BT (Battelle Team), 2007, Default Assumptions and Methods for Atomic Weapons Employer Dose 

Reconstructions, Battelle-TIB-5000, Rev. 00, Richland, Washington, April 2. [SRDB Ref ID: 

32016]

DOE (U.S. Department of Energy), 1984a, Weldon Spring Site Environmental Monitoring Report, 

Calendar Year 1983, DOE/OR/20722-16, Oak Ridge, Tennessee, June. [SRDB Ref ID: 

12530]

DOE (U.S. Department of Energy), 1984b, Radiological Survey Report for the Weldon Spring 

Raffinate Pits Site, DOE/OR/20722-7, Oak Ridge, Tennessee, August. [SRDB Ref ID: 12540]

DOE (U.S. Department of Energy), 1985a, Weldon Spring Site Environmental Monitoring Report, 

Calendar Year 1984, DOE/OR/20722-58, Oak Ridge, Tennessee, July. [SRDB Ref ID: 

12532]

DOE (U.S. Department of Energy), 1985b, Radiological Survey Report for the Weldon Spring Quarry, 

DOE/OR/20722-70, Oak Ridge, Tennessee, September. [SRDB Ref ID: 3566]

DOE (U.S. Department of Energy), 1986, Weldon Spring Site Annual Site Environmental Monitoring 

Report, Calendar Year 1985, DOE/OR/20722-101, Oak Ridge, Tennessee, September. 

[SRDB Ref ID: 12533]

DOE (U.S. Department of Energy), 1987, Weldon Spring Site Annual Environmental Monitoring 

Report, Calendar Year 1986, DOE/OR/21548-T1, Oak Ridge Operations Office, Oak Ridge, 

Tennessee. [SRDB Ref ID: 12535]

DOE (U.S. Department of Energy), 1988, Weldon Spring Site Annual Environmental Monitoring 

Report, Calendar Year 1987, DOE/OR/21548-015, Oak Ridge Operations Office, Oak Ridge, 

Tennessee. [SRDB Ref ID: 12536]

DOE (U.S. Department of Energy), 1989a, Annual Environmental Monitoring Report, 1988, for the 

Weldon Spring Site Remedial Action Project, DOE/OR/21548-079, Rev. 0, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, June. [SRDB Ref ID: 13170]

DOE (U.S. Department of Energy), 1989b, Remedial Investigations for the Quarry Bulk Wastes, 

DOE/OR/21548-066, Rev. 1, Oak Ridge Operations Office, Oak Ridge, Tennessee, 

December. [SRDB Ref ID: 12348]
Document No. ORAUT-TKBS-0028-4 Revision No. 03 Effective Date: 09/08/2017 Page 29 of 34

 

DOE (U.S. Department of Energy), 1990a, Annual Site Environmental Report, 1989, for the Weldon 

Spring Site Remedial Action Project, DOE/OR/21548-129, Rev. 1, Oak Ridge Operations 

Office, Oak Ridge, Tennessee, November. [SRDB Ref ID: 14901]

DOE (U.S. Department of Energy), 1990b, Evaluation of Radon Emissions and Potential Control 

Requirements, Rev. 2, DOE/OR/21548-082-Rev. 2, Oak Ridge Operations Office, Oak Ridge, 

Tennessee, January. [SRDB Ref ID: 13513]

DOE (U.S. Department of Energy), 1991, Annual Site Environmental Report for Calendar Year 1990, 

Weldon Spring Site Remedial Action Project, Weldon Spring, Missouri, DOE/OR/21548-193, 

Rev. 1, Oak Ridge Operations Office, Oak Ridge, Tennessee, September. [SRDB Ref ID: 

14902]

DOE (U.S. Department of Energy), 1992a, Weldon Spring Site Environmental Report for Calendar 

Year 1991, DOE/OR/21548-283, Oak Ridge Operations Office, Oak Ridge, Tennessee, 

August 20. [SRDB Ref ID: 12350]

DOE (U.S. Department of Energy), 1992b, Remedial Investigation for the Chemical Plant Area of the 

Weldon Spring Site, Volume I, DOE/OR/21548-074, Oak Ridge Operations Office, Oak Ridge, 

Tennessee, November. [SRDB Ref ID: 13545]

DOE (U.S. Department of Energy), 1993, Weldon Spring Site Environmental Report for Calendar 

Year 1992, DOE/OR/21548-372, Rev. 0, Oak Ridge Field Office, St. Charles, Missouri, June. 

[SRDB Ref ID: 13175]

DOE (U.S. Department of Energy), 1994, Weldon Spring Site Environmental Report for Calendar 

Year 1993, DOE/OR/21548-436, Rev. 0, Oak Ridge Operations Office, Oak Ridge, 

Tennessee, May. [SRDB Ref ID: 14990]

DOE (U.S. Department of Energy), 1995, Weldon Spring Site Environmental Report for Calendar 

Year 1994, DOE/OR/21548-512, Rev. 0, Oak Ridge Operations Office, Oak Ridge, 

Tennessee, May. [SRDB Ref ID: 14991]

DOE (U.S. Department of Energy), 1996, Weldon Spring Site Environmental Report for Calendar 

Year 1995, DOE/OR/21548-592, Rev. 0, Oak Ridge Operations Office, Oak Ridge, 

Tennessee, June. [SRDB Ref ID: 14992]

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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.

beta radiation

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

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

beta particle is a positron.

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. See acute exposure and chronic 

exposure. (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.

gamma ray, particle, or photon

See gamma radiation.

ionizing radiation

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

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

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

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

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

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

radiation. 

isotope

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

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

Isotopes have very nearly the same chemical properties.
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neutron radiation

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

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

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

radioactive by absorbing free neutrons.

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.

photon radiation 

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

waves to gamma rays.

radiation

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

matter through various mechanisms that involve energy transfer. See ionizing radiation. 

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 radioactivity and nuclide. 

raffinate

Waste from uranium or thorium extraction as well as solids from the neutralization of this 

waste.

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 (TLD)

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

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

a measurement of absorbed dose.

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.