ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 35

Document Title:

Weldon Spring Plant – Occupational 

Environmental Dose

Document Number: ORAUT-TKBS-0028-4 

Revision: 00

Effective Date: 06/28/2005

Type of Document: TBD

Supersedes: None 

Subject Experts: Laura McDowell-Boyer and Craig A. Little

Document Owner

Approval: Signature on File Approval Date: 06/22/2005

Robert Meyer, TBD Team Leader

Approval: Signature on File Approval Date: 06/22/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 06/22/2005

Richard E. Toohey, Project Director

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

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

PLEASE DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 2 of 35

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

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

Occupational Environmental Dose. Initiated by Robert Meyer.

04/25/2005 00-B Incorporates internal review comments. Initiated by Robert Meyer.

06/14/2005 00-C Incorporates NIOSH review comments. Initiated by Robert Meyer.

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

Meyer.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 3 of 35

TABLE OF CONTENTS

Section

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

Page

2

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

4.1 Introduction................................................................................................................................. 6

4.1.1 Purpose..............................................................................................................................6

4.1.2 Content ..............................................................................................................................6

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations......................................... 7

4.2.1 Radionuclides of Concern ..................................................................................................7

4.2.1.1 Airborne Particulate Radionuclides ............................................................................ 7

4.2.1.2 Radon......................................................................................................................12

4.2.2 Source Terms for Airborne Radionuclides ........................................................................12

4.2.2.1 Pre-1968 Source Terms (Operational Period).......................................................... 12

4.2.2.2 Post-1974 Source Terms (Monitoring and Remediation Periods)............................. 13

4.2.3 Annual Intake of Radionuclides ........................................................................................13

4.2.3.1 Pre-1968 Intakes (Operational Period)..................................................................... 14

4.2.3.2 Post-1974 Intakes (Monitoring and Remediation Periods) ....................................... 20

4.3 External Exposure to Onsite Atmospheric Radionuclide Concentrations................................... 25

4.3.1 Ambient Exposure during Operational Period (1957-1966)...............................................25

4.3.2 Ambient Exposure during Monitoring and Remediation Periods (1975-2004)................... 26

References ........................................................................................................................................30

Glossary ............................................................................................................................................34
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LIST OF TABLES

Table

4-1 Principal and associated radionuclides in 238U, 232Th, and 235U decay series...............................

Page

8

4-2 Results of screening calculations for 238U, 232Th, and 235U decay series 

radionuclides.............................................................................................................................10

4-3 Screening results for 238U, 232Th, and 235U decay series ............................................................ 11

4-4 Screening results for recycled uranium contaminants ............................................................... 12

4-5 Average airborne particulate and radon concentrations at WSCP, WSRP, and 

WSQ, based on monitoring data ............................................................................................... 15

4-6 Estimated average annual inhalation intake of radioactive air particulates and 

radon at WSCP, WSRP, and WSQ........................................................................................... 17

4-7 Estimated site-wide maximum inhalation intake of radioactive air particulates and 

radon at WSCP, WSRP, and WSQ........................................................................................... 18

4-8 Organ dose factors for alpha-emitting radionuclides of concern at WSP................................... 23

4-9 Normalized concentrations of radionuclides in soil at WSP....................................................... 23

4-10 Largest contributors to dose for the WSRP, WSCP, and WSQ ................................................. 24

4-11 Estimated ambient onsite dose for WSRP, WSCP, and WSQ at Weldon Spring....................... 27

4-12 Estimated maximum site-wide ambient dose at WSP................................................................ 29

LIST OF FIGURES

Figure

4-1 Air particulate, radon, and gamma monitoring stations in WSCP and WSRP............................

Page

21

4-2 Air particulate, radon, and gamma monitoring stations in WSQ................................................. 22

4-3 Topographic map and locations of gamma exposure rate measurements for WSQ .................. 28
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ACRONYMS AND ABBREVIATIONS

α alpha particle

AED Aerodynamic Equivalent Diameter

AMAD Activity Median Aerodynamic Diameter

CDH Colorado Department of Public Health and Environment (previously Colorado 

Department of Health)

DF dose factor

DOE U.S. Department of Energy

EPA U.S. Environmental Protection Agency

ET Extrathoracic

FEMP Fernald Environmental Management Project

ICRP International Commission on Radiological Protection

TBD technical basis document

TLLα total long-lived alpha

U.S.C United States Code

WL Working Level = 1.3 × 105 MeV of alpha energy in 1 liter of air

WLM Working Level Month = Exposure from 1 WL or radon daughters for 170 working hours

WSCP Weldon Spring Chemical Plant

WSP Weldon Spring Plant

WSQ Weldon Spring Quarry

WSRP Weldon Spring Raffinate Pits
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 6 of 35

4.1 INTRODUCTION

4.1.1

Technical basis documents (TBDs) and Site Profile Documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute for Occupational 

Safety and Health in the completion of the individual work required for each dose reconstruction.

Purpose

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

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

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

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

The preparers of this report used available information about the Weldon Spring Plant (WSP) 

environment dating back to 1957. They strongly advise the reader to consider remarks in the text on 

the appropriate use and interpretation of the material.

4.1.2

The term occupational environmental dose refers to the radiation dose received in the course of work 

duties outside plant buildings, but on the WSP site. This TBD considers internal and external 

exposures to radionuclides in the outdoor environment separately in calculating this dose. Dose 

reconstructors can use estimated occupational environmental dose to develop a reliable individual 

dose when a worker was not monitored adequately. 

Content

The periods for which this TBD evaluates environmental doses are those during which the U.S. 

Department of Energy (DOE) or its predecessor agencies (the U.S. Atomic Energy Commission until 

1975 and the Energy Research and Development Administration until 1977) had contractors on some 

or all of the WSP. These include the operational period (1957 to 1967) and the monitoring and 

remediation periods (1975 to present). The monitoring period, from 1975 through 1984, applies only 

to the raffinate pits and quarry areas because DOE did not assume control of the chemical plant until 

1985.

Workers at the Weldon Spring Chemical Plant (WSCP) received environmental doses during the 

operational period as a result of stack effluents from buildings, contamination of soil resulting from 

stack releases, and onsite storage of ore concentrates. Workers at the Weldon Spring Quarry (WSQ) 

received environmental doses during the same period as a result of suspension and resuspension

from contaminated rubble and soil being placed in, or already in, the disposal area. During the 

remediation period, workers received occupational doses due to the presence of contaminated soils 

and structures at the WSQ, the WSCP, and the Weldon Spring Raffinate Pits (WSRP) as a result of 

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 (see Section 2.2.4 of

this Site Profile).

Section 4.2 contains information necessary to estimate internal environmental dose. Section 4.2.1 

identifies radionuclides of concern. Section 4.2.2 discusses source terms (release rates) for 

radionuclides potentially significant to internal environmental dose, available air monitoring data, and 

estimated annual inhalation intake activities of radionuclides.
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Section 4.3 contains information necessary for estimating external environmental dose. It reviews 

and tabulates ambient doses in a format that is usable for dose reconstruction. These doses have 

been reported for the WSRP, WSCP, and WSQ areas in remedial investigations supporting the 

Comprehensive Environmental Response and Liability Act process and in site-wide annual reports, 

and as a result of other radiological surveys. 

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

The focus of this section is internal occupational exposure to radionuclides released from WSP 

facilities during the period of uranium processing from 1957 through 1967 (processing ended in 1966, 

and shutdown procedures occurred in 1967) and the period of monitoring and remediation activities

from 1975 to the present. Internal exposure could have occurred as a result of inhalation of airborne 

radioactive particulates or radon during these periods of work activity. Between 1967 and 1984, the 

site was essentially undisturbed; remedial activities began in 1985 under DOE sponsorship. 

Section 4.2.1 addresses the identification of radionuclides of concern, defined as those contributing to 

95% of the potential dose. Section 4.2.2 describes specific sources of various radionuclides in the 

outdoor environment. Section 4.2.3 calculates activity intakes of radioactivity, in the form of air 

particulates and radon, based on the information reviewed. In addition, Section 4.2.3 discusses 

appropriate assumptions dose reconstructors should make with regard to intake of radioactivity.

4.2.1

4.2.1.1 Airborne Particulate Radionuclides

Radionuclides of Concern

The WSCP processed uranium (

235,238U) and thorium (232Th) ore concentrates and some scrap metal 

during its operational period (see Section 2.2.4 of this Site Profile). The Plant disposed of waste 

materials associated with uranium and thorium processing at the chemical plant and other facilities in 

the quarry (see Section 2.2.3.2 of this Site Profile). The radiological risks associated with the 235U 

decay series are likely to be much lower than those associated with the 238U series because the 

abundance of 235U in natural uranium ore concentrates is low (an activity ratio of 0.046:1 for 235U to 

238U), and WSP processed only a small amount of slightly enriched uranium (see Section 2.2.4 of this 

Site Profile). 

During the operational period (1957–1967), most particulate radioactive contaminants in air 

associated with WSP facilities would have been isotopes of uranium and thorium, along with shortlived daughters that grew in to a significant extent in the 10-year period. A Weldon Spring historical 

offsite dose estimate study by Meshkov et al. (1986) assumed that activities of 230Th, 226Ra, and 210Pb 

in yellowcake were 5%, 1%, and 1%, respectively, of that of 238U. 

During the remediation period (1985 to present), other radioactive daughters will have grown into the

contaminated media (soil around the WSCP, WSRP, and WSQ) to a potentially significant extent. 

Baseline Assessment for the Chemical Plant Area of the Weldon Spring Site (DOE 1992) identified 

the following radionuclides for evaluation: 238U, 235U, 232Th, 230Th, 231Pa, 227Ac, 228Ra, 226Ra, and 210Pb, 

as well as isotopes of radon for separate evaluation. The rationale for selecting these radionuclides 

was given as follows. First, all radionuclides in the 235U, 238U, and 232Th decay series with half-lives 

less than 1 year had established secular equilibrium with longer lived parents in the series because

operations had ended more than 20 years before the baseline assessment. This means that all these 

shorter-lived radionuclides could be assumed to have activities equal to that of the parent. Table 4-1 

indicates the relationship between the longer lived parents and shorter lived daughters in each decay 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 8 of 35

series. Second, the activities of 234U and 238U are equal in natural uranium. Third, the half-life of 228Th

is very short in relation to the period between plant closure and the baseline assessment; thus this 

radionuclide could be assumed to be in transient equilibrium with 228Ra. 

Table 4-1. Principal and associated radionuclides in 

238U, 232Th, and 235U decay series. 

Principal radionuclidea Associated decay chainb

Species Half-lifec Species Half-lifec

U-238 decay series

U-238 4.5E+09 yr Th-234 24 d

Pa-234 1.2 min

U-234 2.4E+05 yr

Th-230 7.5E+04 yr

Ra-226 1.6E+03 yr Rn-222 3.8 d

Po-218 3.1 min

Pb-214 27 min

Bi-214 20 min

Po-214 1.6E+03 μs

Pb-210 22 y Bi-210 5.0 d

Po-210 1.4E+02 d

Th-232 decay series

Th-232 1.4E+10 yr

Ra-228 5.8 yr Ac-228 6.2 hr

Th-228 1.9 yr Ra-224 3.7 d

Rn-220 55 s

Po-216 0.15 s

Pb-212 11 h

Bi-212 61 min

Po-212 (64%)d 0.30 μs

Tl-208 (36%)d 3.1 min

U-235 decay series

U-235 7.0E+08 yr Th-231 26 hr

Pa-231 3.3E+04 yr

Ac-227 22 yr Th-227 (98.6%)d 19 d

Fr-223 (1.4%) 22 min

Ra-223 11 d

Rn-219 4.0 s

Po-215 1.8 ms

Pb-211 36 min

Bi-211 2.1 min

Tl-207 4.8 min

a. Radionuclides with half-lives greater than 1 year. 

b. The chain of decay products of a principal radionuclide 

extending to the next principal radionuclide or stable nuclide. 

c. Half-lives to 2 significant digits from Tuli (2000). 

d. Branching ratios in parentheses (branching ratios less than 1% 

not listed in this table). 

A simplified screening analysis was conducted for this TBD to rank WSP radionuclides in the uranium 

and thorium decay series in order of significance to the total dose from all radionuclides as a result of 

inhalation during the operational and monitoring/remediation periods. Radionuclides selected as 

potentially significant are the highest potential dose producers, contributing up to 95% of the total 

dose. The starting list of radionuclides includes the daughters in the uranium and thorium decay 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 9 of 35

series, listed in Table 4-1 (with the exception of the short-lived daughters of radon isotopes, which are 

discussed in Section 4.2.1.2).

During the operational period, the airborne activity of uranium daughters 230Th, 226Ra, and 210Pb are 

assumed to be 5%, 1%, and 1%, respectively, of the 238U activity for this screening exercise. This is 

consistent with the assumption in Meshkov et al. (1986). In addition to the fact that no discards of 

232Th to the air were measured for this isotope during processing, the activity of 232Th processed in 

any one year was less than 1% of the natural uranium processed (DOE 1986). Thus, 232Th was 

probably not a significant contributor to inhalation dose during the operational period. To assess the 

relative significance of the 238U daughters, this TBD analysis calculated inhalation dose per unit 238U 

activity of each daughter (except 222Rn and its short-lived daughters). A normalized organ dose was 

calculated by multiplying the normalized airborne activities (1.0 for 238U and 234U, and 0.05 for 230

Th, 

and so on) by the extrathoracic (ET) airways, bone surface, and lung dose factors from International 

Commission on Radiation Protection CD1 (ICRP 2001). These organs were identified as those of 

concern based on the fact that, for the list of radionuclides analyzed, the highest dose factors were 

associated with these organs, considering all possible absorption types and assuming inhalation 

particle sizes of 5 μm activity median aerodynamic diameter (AMAD). The analysis evaluated the 

percent contribution of each radionuclide to the summed normalized doses, the results of which are 

listed in Table 4-2. For the operational period, 234U and 238U are estimated to contribute 97% of the ET 

airways and lung dose and 90% of the bone surface dose, while 230Th contributes almost the entire

remaining dose for these organs. 

During the monitoring and remediation period, normalized soil and sludge activity concentrations of 

the radionuclides estimated for the WSRP, WSCP, and WSQ in the early 1990s (DOE 1992) were 

used to assess the potential significance of daughter radionuclides. Most of these estimates are 

based on measurements. Measurements are a better estimate of the relative concentrations because 

radiological decay calculations ignore the differential leaching of the chemical species in the soil and 

sludge. Use of these measured values from the 1990s will overestimate the daughter ingrowth for the 

pre-1990 period. If measured values for particular radionuclides were not available, the estimated 

concentrations were biased high (DOE 1992, p. 2-12). 

The normalized activity concentrations, defined as the measured activity concentration in sludge or 

soil divided by the measured 238U activity concentration in the same media, are in Table 2.3 of DOE 

(1992, p. 2-39). Reasonably assuming that normalized concentrations in air will reflect normalized 

concentrations in soil or sludge due to resuspension processes, this TBD analysis multiplied the value 

for each radionuclide (Bq per Bq of 238U) by the corresponding ICRP (2001) organ dose factors 

(Sv/Bq) to obtain normalized doses. For example, for 230Th in the raffinate pits, the analysis assumed 

that the normalized airborne activity was 25 Bq per Bq of 238U. The corresponding normalized bone 

surface dose was calculated by multiplying this activity by the 230Th organ dose factor of 1.5 ×10-3

Sv/Bq for absorption type S and 5 μm AMAD inhaled particles, to obtain a normalized bone surface 

dose of 3.8 ×10-2 Sv/Bq of 238U inhaled. The percent contribution of each radionuclide to the summed 

normalized dose for each organ and WSP area was calculated. Table 4-2 lists the results. Separate 

calculations were done for each area because the measured activity concentrations were significantly 

different for the WSRP, WSCP, and WSQ. The differences are probably due to the different types of 

wastes deposited in the waste areas and differential leaching mechanisms for the waste areas versus 

the general WSCP site. 

The radionuclides of concern for the operational and monitoring/remediation periods were ranked 

according to potential significance, based on the percent contributions calculated above. Table 4-3 

lists the final ranking for the three areas. The radionuclides listed in bold are those contributing to 

95% of the dose per unit activity of 238U and are potentially significant. (Radionuclides alone 
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Table 4-2. Results of screening calculations for 238U, 232Th, and 235U decay series radionuclides. OPERATIONAL PERIOD (1957-1967)

Radionuclide

Estimated airborne 

activity normalized to U238 activity 

(Bq per Bq U-238)

Percent contribution of each radionuclide to normalized dose

ET airway Bone surface Lungs

U-238 1.0 4.5E+01 1.0E+01 4.4E+01

Th-234 1.0 4.5E-03 3.2E-03 4.8E-02

Pa-234 1.0 5.3E-03 5.8E-05 1.9E-03

U-234 1.0 5.2E+01 1.1E+01 5.3E+01

Th-230 0.05 2.6E+00 7.8E+01 2.6E+00

Ra-226 0.01 9.0E-02 9.3E-02 2.2E-01

Pb-210 0.01 9.7E-04 3.7E-01 1.8E-03

Bi-210 0.01 1.8E-03 1.1E-06 6.3E-03

Po-210 0.01 5.1E-02 9.7E-03 2.2E-01

MONITORING/REMEDIATION PERIOD (1975-PRESENT)

Radionuclide

Estimated airborne 

activity normalized to U238 activity 

(Bq per Bq U-238)a Percent contribution of each radionuclide to normalized dose

WSRP WSCP WSQ 

WSRP WSCP WSQ

ET 

airway

Bone 

surface Lungs

ET 

Airway 

Bone 

surface Lungs

ET 

airway

Bone 

surface Lungs

U-238 decay series

U-238 1.0 1.00 1.00 2.7E+00 1.9E-02 2.4E+00 3.3E+01 1.1E+00 2.8E+01 1.6E+01 3.3E-01 1.1E+01

Th-234 1.0 1.00 1.00 2.7E-04 6.0E-06 2.6E-03 3.3E-03 3.4E-04 3.0E-02 1.6E-03 1.0E-04 1.2E-02

Pa-234 1.0 1.00 1.00 3.2E-04 1.1E-07 1.0E-04 3.9E-03 6.1E-06 1.2E-03 1.8E-03 1.8E-06 4.9E-04

U-234 1.0 1.00 1.00 3.2E+00 2.1E-02 2.9E+00 3.9E+01 1.2E+00 3.4E+01 1.8E+01 3.6E-01 1.3E+01

Th-230 25.0 0.15 1.70 7.8E+01 7.2E+01 7.0E+01 5.7E+00 2.4E+01 4.9E+00 3.0E+01 8.3E+01 2.2E+01

Ra-226 0.62 0.14 0.55 3.4E-01 1.1E-02 7.4E-01 9.4E-01 1.4E-01 2.0E+00 1.7E+00 1.6E-01 3.0E+00

Pb-210 0.62 0.14 1.70 3.6E-03 4.3E-02 6.1E-03 1.0E-02 5.5E-01 1.6E-02 5.8E-02 2.0E+00 7.7E-02

Bi-210 0.62 0.14 1.70 6.8E-03 1.3E-07 2.1E-02 1.9E-02 1.7E-06 5.7E-02 1.1E-01 6.1E-06 2.7E-01

Po-210 0.62 0.14 1.70 1.9E-01 1.1E-03 7.4E-01 5.3E-01 1.4E-02 2.0E+00 3.0E+00 5.2E-02 9.4E+00

Th-232 decay series

Th-232 0.026 0.076 0.13 1.2E-01 7.5E-02 1.4E-01 4.3E+00 1.2E+01 4.8E+00 3.5E+00 6.4E+00 3.2E+00

Ra-228 0.25 0.076 0.48 9.9E-02 1.7E-02 1.1E-01 3.7E-01 3.0E-01 3.9E-01 1.1E+00 5.6E-01 9.6E-01

Ac-228 0.25 0.076 0.48 9.0E-04 4.0E-04 1.4E-03 3.4E-03 6.9E-03 5.0E-03 1.0E-02 1.3E-02 1.2E-02

Th-228 0.25 0.076 0.48 2.2E+00 1.3E-01 3.7E+00 8.2E+00 2.3E+00 1.3E+01 2.4E+01 4.4E+00 3.3E+01

Ra-224 0.25 0.076 0.48 6.3E-02 4.6E-04 3.5E-01 2.3E-01 7.9E-03 1.3E+00 7.0E-01 1.5E-02 3.1E+00

U-235 decay series

U-235 0.046 0.046 0.046 1.3E-01 9.7E-04 1.2E-01 1.6E+00 5.5E-02 1.4E+00 7.7E-01 1.6E-02 5.4E-01

Th-231 0.046 0.046 0.046 4.6E-06 3.3E-08 5.5E-06 5.7E-05 1.9E-06 6.4E-05 2.7E-05 5.5E-07 2.5E-05

Pa-231 0.73 0.038 0.0058 3.4E+00 6.2E+00 3.7E+00 2.1E+00 1.8E+01 2.3E+00 1.5E-01 8.3E-01 1.4E-01

Ac-227 0.58 0.019 0.0029 9.3E+00 2.1E+01 1.1E+01 3.7E+00 3.9E+01 4.2E+00 2.7E-01 1.8E+00 2.5E-01

Th-227 0.58 0.019 0.0029 3.2E-01 4.0E-03 2.6E+00 1.3E-01 7.4E-03 1.0E+00 9.1E-03 3.4E-04 6.0E-02

Fr-223 0.58 0.019 0.0029 2.3E-04 1.2E-06 4.9E-05 9.3E-05 2.3E-06 1.9E-05 6.7E-06 1.0E-07 1.1E-06

Ra-223 0.58 0.019 0.0029 2.9E-01 1.9E-03 1.9E+00 1.2E-01 3.5E-03 7.4E-01 8.4E-03 1.6E-04 4.4E-02

a. Baseline assessment for chemical plant area – when measured values were not available, estimates are biased high according to DOE (1992, p. 2-12).

contributing less than 5% of the dose to any organ were also eliminated from the list of radionuclides 

of potential concern) Thus, these results indicate that 230Th, 234U, and 238U are the radionuclides of 

potential significance for the operational period, and that 227Ac, 231Pa, 210Po, 228Th, 230Th, 232Th, 234U, 

and 238U are the radionuclides of potential significance for the monitoring and remediation periods. 

In addition to the radioactive isotopes of the uranium and thorium decay series, dose reconstructors 

must consider the potential presence of other radionuclides associated with recycled uranium. 

Between 1962 and 1966, the WSP might have received some recycled uranium from the Fernald site 

in southwest Ohio, such that trace amounts of residual transuranic elements (including neptunium and 

plutonium isotopes), fission products (such as technetium), and reactor-produced uranium isotopes 

(

236U) might have been present. However, more than 97% of the uranium received by WSP is 

believed to have been natural uranium (DOE 2000, p. C-19). Upper bound estimates on releases of 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 11 of 35

Table 4-3. Screening results for 238U, 232Th, and 235U decay series. 

Radionuclidea 

Highest rank in organ dose contributiona

Raffinate pitsb Chemical plantc Quarry

U-238 decay series

U-238 5 (ET) 2 (2) (ET,L) 4 (ET,L)

Th-234 17 (BS,L) 16 (6) (L) 17 (BS)

Pa-234 18 (ET) 17 (6) (ET) 18 (ET)

U-234 4 ((ET) 1 (1) (ET,L) 3 (ET,L)

Th-230 1 (ET, BS, L) 2 (1) (BS) 1 (BS)

Ra-226 7 (ET) 8 (4) (ET,L) 7 (ET)

Pb-210 6 (BS) 8 (4) (BS) 4 (BS)

Bi-210 15 ((ET,L) 17 (8) (ET,L) 14 (ET,L)

Po-210 9 (L) 8 (5) (L) 5 (L)

Th-232 decay series

Th-232 5 (BS) 4 (BS) 2 (BS)

Ra-228 9 (BS) 9 (BS) 7 (BS)

Ac-228 16 (BS) 15 (BS) 14 (BS)

Th-228 4 (BS,L) 3 (ET) 1 (L)

Ra-224 11 (L) 11 (L) 7 (L)

U-235 decay series

U-235 11 (ET) 8 (ET) 9 (ET)

Th-231 21 (ET,BS,L) 20 (BS,L) 20 (ET,BS,L)

Pa-231 3 (ET,BS,L) 3 (BS) 6 (BS)

Ac-227 2 (ET,BS,L) 1 (BS) 5 (BS)

Th-227 6 (L) 12 (L) 15 (BS,L)

Fr-223 18 (BS) 19 (BS) 21 (ET,BS,L)

Ra-223 8 (L) 15 (L) 16 (BS,L)

a. Radionuclides and rank numbers in bold identify those that are the top contributors 

to 95% of the dose.

b. ET= extrathoracic airways, BS = bone surface, L = lungs.

c. Numbers in parentheses represent ranking for operational period

plutonium isotopes, neptunium, and technetium from WSP are 0.0, 3.0, and 1.2 g per 45 MTU 

released from stacks, respectively (based on DOE 2000, Table ES-7D, p. ES-25), between 1962 and 

1966.

To evaluate the potential impact on dose of contaminants of recycled uranium (239Pu, 240Pu, 237Np, 

99Tc), the maximum estimated total discharge of these contaminants to air for all years in which 

recycled uranium might have been received (1962 to 1967, p. ES-25 in DOE 2000) is assumed to be 

representative. These discharge values are divided by the estimated total discharge of 238U via 

stacks, from DOE (1986) and summarized in DOE (2000), to obtain a normalized release of each 

contaminant to the air (Bq per Bq 238U). 

Assuming that normalized concentrations of recycled uranium contaminants in air will reflect 

normalized releases from the stacks due to atmospheric dispersion processes, the TBD analysis 

multiplied these releases by the corresponding ICRP (2001) organ dose factor (Sv/Bq) for each 

radionuclide to obtain a normalized dose, with units of Sv per Bq 238U. The organs of most concern 

were again identified as ET airways (uranium isotopes), bone surface (plutonium isotopes and 237Np), 

and lungs (99Tc). The percent contribution of each radionuclide to the summed normalized doses for 

each organ was calculated. These estimates apply to all WSP areas, as information did not allow a 

more detailed analysis by location. Table 4-4 lists the results of these calculations for contaminants of 

recycled uranium. The results indicate that 238U contributes approximately 99% of the dose per unit 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 12 of 35

activity, if the contaminants were present to the extent estimated. Thus, this TBD does not consider 

the contaminants of recycled uranium to be potentially significant contributors to dose.

Table 4-4. Screening results for recycled uranium contaminants. 

Radionuclide

Maximum estimated activity released 

to air, normalized to U-238a

(Bq per Bq of U-238)

Percent contribution to 

normalized dose

Radionuclides 

contribute to 95% 

ET airways Bone surface Lungs of total dose?

Pu-239 0.0 0.0E+00 0.0E+00 0.00E+00 No

Pu-240 0.0 0.0E+00 0.0E+00 0.00E+00 No

Np-237 1.40E-04 2.8E-03 9.3E-01 7.0E-03 No

Tc-99 1.36E-03 1.7E-05 4.6E-07 9.6E-05 No

U-238 1.00E+00 1.0E+02 9.9E+01 1.0E+02 Yes

a. Based on DOE (2000). 

To summarize, the most significant radionuclides identified as contributing to 95% (but no less than 

5% each) of the total inhalation dose are listed in bold in Table 4-3. The list includes 227Ac, 231Pa, 

210Po, 228Th, 230Th, 232Th, 234U, and 238U. Section 4.2.1.2 discusses radon and its short-lived daughter 

isotopes.

4.2.1.2 Radon

Isotopes of radon associated with the 235U, 238U, and 232Th decay series can pose a significant dose to 

workers depending on the amount of precursor (radium) in the soil and the length of exposure. The 

dose resulting from inhalation of radon is largely from the short-lived, alpha-particle-emitting radon 

daughters rather than from the gas itself. Three radon isotopes are generated during decay of 235U, 

238U, and 232Th (see Table 4-1) – 219Rn, 222Rn, and 220Rn, respectively. The risks associated with 

219Rn due to its extremely short half-life (4 sec) and small amount of precursor radium (due to the 

limited amount of 235U processed at WSP) were insignificant in comparison to those from 222Rn and 

220Rn, and are not relevant. Estimates made in 1990 of potential exposure to 222Rn and 220Rn at the 

WSQ (DOE 1990) suggested the whole-body effective dose rates resulting from inhalation of these 

isotopes might be slightly less than 1 mrem/hr (at the time and place the measurement occurred). 

Thus, this TBD considers the inhalation intakes for these two isotopes to be potentially significant, as 

discussed in more detail below.

4.2.2

Section 4.2.2.1 describes sources of particulate radionuclides and gaseous radon to air in the WSP

environment. In addition to emissions of 238, 234U and 232Th during ore processing, decay products of 

these three isotopes were measured during the operational and the monitoring/remediation periods. 

Source Terms for Airborne Radionuclides

4.2.2.1 Pre-1968 Source Terms (Operational Period)

During the operational period (1957-1967), drums of uranium concentrate (yellowcake) were emptied 

into the hopper at the top of the receiving and sampling building (Meshkov et al. 1986). 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 

generated in the purification process prior to denitration were bled off in liquid form in the raffinate, 

which was pumped to the pits. The denitration process was the dustiest part of the operation. Thus, 

point sources of uranium emissions from stacks existed during this period. 

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

20%) is released when tied to solid particles, such as those that occur in uranium concentrates. 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 13 of 35

However, the digestion phase of the refinery process would have released the trapped radon during 

the operational period. Offgases containing radon isotopes from this process were conveyed to the 

acid recovery plant at the WSCP. The discharge from the acid recovery plant was the prime source of 

radon emission and perhaps the only point source for radon isotopes.

Wastes disposed of in the raffinate pits were not a significant source of airborne radionuclides, 

particulate or gaseous, during the operational period due 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 when they were being managed, even though Bechtel (1984a) reported that 

Pits 1 and 2 could become dry during the summer months. NLO (1977) concluded that the inherent 

consistency of the raffinate material precluded sufficient drying, and the pits did not pose a significant 

source of airborne contaminants.

Wastes disposed of in the quarry during WSP operations present 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 generated from 226Ra and 224Ra, respectively, in the quarry also present a 

diffuse source of airborne radionuclides. Although most of the material initially dumped in the quarry 

in 1959 was at an elevation below the average natural water level of 457 feet above mean sea level 

(MSL) reported in 1967, later disposals maintained a fill elevation of about 480 feet MSL, such that 

most of the waste was not submerged (Weldon Spring Raffinate Pits and Quarry Task Force Report

1967).

Uranium and thorium effluents from the various stacks were not directly measured. The total amount 

of uranium emitted to the atmosphere has been estimated from monitoring data (Meshkov et al. 1986)

and from a materials balance study (DOE 1986). The reported atmospheric discharges from the 

materials balance study were engineering estimates derived from airflows and other process factors 

along with stack monitoring data, and are summarized in Table 2-8 in Part 2 of this Site Profile. 

Based on these two methods, the estimated amount of uranium activity emitted from the plant ranges 

between approximately 1 and 5 Ci/yr. Meshkov et al. estimated release rates for uranium decay 

products (230Th, 226Ra, 210Pb) as a fixed percent of the estimated amount of 238U released, assuming 

the activities of 230Th, 226Ra, and 210Pb in the original uranium concentrate were 5%, 1%, and 1%, 

respectively, of that of 238U. An estimate of radon release based on the amount of uranium processed 

during the operational period ranged from 12 to 34 Ci/yr (assuming radium activity was 1% of the 

uranium activity, radon was in equilibrium with radium, and all radon was released). 

4.2.2.2 Post-1974 Source Terms (Monitoring and Remediation Periods)

During the monitoring and remediation periods, from 1975 to the present, diffuse emissions 

predominate as the source of radionuclides to the air at the WSCP, WSRP, and WSQ. Beginning in 

1981, annual environmental monitoring reports have provided estimates of air concentrations of 

particulate radionuclides and of radon at the WSCP and WSQ. These estimates reflect emissions 

during the later monitoring period and throughout remediation activities.

4.2.3

As noted in Section 4.2.2, releases of radioactive particulates or radon to the atmosphere have not 

been fully quantified at the WSP for the operational period or the monitoring and remediation periods. 

Meshkov et al. (1986) attempted to quantify point source emissions for the operational period by 

considering monitoring data and the results of a materials balance study. However, in considering 

exposure to radionuclides, material released from stacks provides only a partial picture of how 

radioactivity becomes airborne. Diffuse sources resulting from wind entrainment or other mechanical 

Annual Intake of Radionuclides
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 14 of 35

disturbance of radionuclides deposited on the ground or in the waste pits or quarry dominate 

exposures during the remediation period, and contribute to exposures during the operational period. 

In light of the importance of diffuse sources and the lack of information available for quantifying source 

strengths throughout the site, estimated occupational intakes for the WSP rely solely on available air 

monitoring data for this TBD.

Descriptions of particle size and solubility class are not available for further characterizing the 

information presented below for air particulates. Dose reconstructors should assume that the activity 

intake of air particulates is 100% respirable, and the default aerosol size of 5-μm AMAD. In addition, 

dose reconstructors should base the selection of dose factors on the solubility/clearance class 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), corresponding to an hourly rate of 1.2 m3

, for light activity and a 2,000-hr 

work year. Intakes can be scaled to a different rate or to consider partial-year exposures.

4.2.3.1 Pre-1968 Intakes (Operational Period)

Meshkov et al. (1986) reviewed air monitoring data for airborne particulates in the operational period. 

This TBD analysis used these same data and an onsite measurement made in 1960 (Holt 1960) to 

calculate estimated intakes of radioactive airborne particulates according to the methods described 

below under Airborne Particulate Radionuclides. For radon released during yellowcake processing, 

Meshkov et al. estimated an annual release rate based on the average amount of uranium 

concentrate processed annually, which this analysis used to estimate maximum concentrations in the 

WSCP (see below under Radon). The analysis based estimates of air concentrations of radon 

released from materials deposited in the WSQ during the operational period on measurements made 

between 1979 and 1982. The precursors of radon (226Ra and 224Ra) were not likely to have been 

significantly depleted during the time period between the end of operations and 1979, since this 

element readily sorbs to soil and is not very mobile (ATSDR 1990), and these radium isotopes are 

gradually being replenished from the 238U and 232Th decay series. Although temporal fluctuations in 

radon air concentrations have been noted (Meshkov et al. 1986, MK-Ferguson 1990a, MK-Ferguson 

1992a), likely due to climatic variations, there were no obvious downward trends in concentrations at 

the WSQ until the bulk wastes were removed in the 1990s. (Releases from the WSRP were negligible 

during the operational period; see Section 4.2.2.1).

Airborne Particulate Radionuclides

Meshkov et al. (1986) provides summarized results of uranium air concentration measurements made 

at the WSP perimeter and at two offsite locations from 1959 through 1965. The perimeter data used 

here (Table 4-5) have been converted to units of Bq/m3

. The original data are reported in units of 

special uranium microcuries, defined as the sum of 3.7 × 104 dps from 238U, 3.7 × 104 dps from 234U, 

and 9 × 102 dps from 235U. Converting to units of Bq, the special uranium μCi is multiplied by 3.7 × 

104 Bq/μCi and by a factor of 2.024 to report total uranium activity. Other radionuclides were not 

measured. As indicated in Section 4.2.1.1, only isotopes of 238U, 234U, and 230Th (and their very shortlived daughters) were significant airborne particulate radionuclides during the operational period. 

Intakes of 230Th are estimated from measured airborne uranium concentrations in the manner 

described below.

Although many dust-generating operations occurred inside WSP sampling and refinery buildings, Holt 

(1960) indicated that decontamination of portable hoppers occurred outside Building 103. According 

to Holt, general air samples taken downwind during the cleaning of six 5-ton hoppers ranged from 90 

to 443 dpm/m3

. This corresponds to between 1.5 and 7.4 Bq/m3 of total uranium (assuming the 

measurement was predominantly of uranium isotopes from the dusty operation), or an average of 4.4 

Bq/m3 total uranium activity concentration. 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 15 of 35

Table 4-5. Average airborne particulate and radon concentrations at WSCP, WSRP, and 

WSQ, based on monitoring data. 

Year

WSRP WSCP WSQ

Uraniuma 

(Bq/m3)

Radonb 

(Bq/m3)

Uraniuma

(Bq/m3) Radonb 

(Bq/m3)

Uraniuma

(Bq/m3)

Radonb 

(Bq/m3 Perimeter Hopper cleanout )

Operational period

1957 (c) (c) 1.7E-02 4.4 9.1E+01 (c) (c)

1958 (c) (c) 1.7E-02 4.4 9.1E+01 (c) (c)

1959 (c) (c) 1.7E-02 4.4 9.1E+01 (c) (c)

1960 (c) (c) 1.2E-02 4.4 9.1E+01 (c) (c)

1961 (c) (c) 1.6E-02 4.4 9.1E+01 (c) (c)

1962 (c) (c) 5.3E-03 4.4 9.1E+01 (c) (c)

1963 (c) (c) 8.4E-03 4.4 9.1E+01 2.6E-04 2.6E+01

1964 (c) (c) 1.1E-02 4.4 9.1E+01 1.5E-03 2.6E+01

1965 (c) (c) 3.7E-03 4.4 9.1E+01 1.2E-02 2.6E+01

1966 (c) (c) 3.7E-03 4.4 9.1E+01 1.2E-02 2.6E+01

1967 (c) (c) 3.7E-03 4.4 9.1E+01 1.2E-02 2.6E+01

Monitoring/remediation period

Gross alpha

(Bq/m3)

Radonb

(Bq/m3)

Gross alpha

(Bq/m3)

Radonb

(Bq/m3)

Gross alpha

(Bq/m3)

Radonb

(Bq/m3)

1975-1979 (c) 4.8E+01e (f) (f) (c) 3.3E+01e

1980 (c) 1.2E+02 (f) (f) (c) 2.5E+01

1981 (c) 7.4E+01 (f) (f) (c) 2.6E+01

1982 (c) 5.9E+01 (f) (f) (c) 2.6E+01

1983 (c) 2.1E+01 (f) (f) (c) 2.2E+01

1984 (c) 3.6E+01 (f) (f) (c) 2.5E+01

1985 (c) 2.4E+01 (c) 1.5E+01 (c) 2.5E+01

1986 0g 3.4E+01 0g 2.3E+01 (c) 2.4E+01

1987 0g 2.8E+01 0g 3.0E+01 (c) 4.5E+01

1988 0g 4.4E+01 0g 4.1E+01 0g 6.8E+01

1989 2.0E-04 3.3E+01 1.6E-04 3.8E+01 1.9E-05 4.3E+01

1990 1.0E-04 2.0E+01 9.1E-05 2.4E+01 1.7E-05 3.5E+01

1991 5.0E-05 2.2E+01 5.5E-05 2.0E+01 3.3E-06 3.0E+01

1992 3.9E-05 2.2E+01 4.1E-05 1.6E+01 0d 2.7E+01

1993 4.6E-05 1.0E+01 4.6E-05 7.4E+00 3.1E-06 6.1E+01

1994 5.4E-05 2.8E+01 6.3E-05 1.7E+01 2.4E-05 1.7E+02

1995 4.7E-05 3.4E+01 5.1E-05 2.2E+01 8.0E-06 5.8E+01

1996 6.2E-05 5.7E+01 5.9E-05 2.7E+01 3.3E-06 1.3E+01

1997 7.5E-05 1.3E+02 5.5E-05 2.2E+01 0g 1.1E+01

1998 8.7E-05 2.3E+01 7.7E-05 1.3E+01 2.6E-06 7.4E+00

1999 7.7E-05 3.7E+01 7.6E-05 2.5E+01 3.7E-07 1.7E+01

2000 4.1E-05 2.0E+01 5.1E-05 2.6E+01 3.3E-06 1.3E+01

2001 4.1E-05h 2.0E+01h 5.1E-05h 2.6E+01h 3.3E-06h 1.3E+01h

2002 4.1E-05h 2.0E+01h 5.1E-05h 2.6E+01h 3.3E-06h 1.3E+01h

2003 4.1E-05h 2.0E+01h 5.1E-05h 2.6E+01h 3.3E-06h 1.3E+01h

2004 4.1E-05h 2.0E+01h 5.1E-05h 2.6E+01h 3.3E-06h 1.3E+01h

a. The measured concentrations of uranium during the operational period were expressed as a special uranium μCi/m3, 

which is defined as the sum of 3.7 x 104 dps U-238, 3.7 x 104 dps U-234, and 9 x 102 dps U-235 per m3

. These were 

converted to Bq/m3 of total uranium, by multiplying the special uranium μCi/m3 by 3.7 x 104 Bq/μCi and by the factor 

2.024 to account for activity contributions by all isotopes.

b. Radon refers to both 222Rn and 220Rn, and includes natural background contributions.

c. Blank indicates there is no significant air concentration expected for the corresponding years. 

d. No data reported in 1966 and 1967; assumed equal to the value in 1965.

e. Assumed equal to average of following 10 years prior to remediation activity disturbances.

f. WSCP was not transferred to DOE until 1985.

g. Value of zero indicates the measured value was less than or equal to the background value.

h. Assumed equal to the value in the year 2000, by which time active remediation was essentially complete

Table 4-5 lists estimated average perimeter air concentrations, and onsite air concentrations during 

hopper cleanout, of uranium during the operational period for the WSCP and WSQ. These values

were obtained from the measured uranium concentrations described above. The average offsite 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 16 of 35

background value of 7.4E-04 Bq/m3 (0.02 pCi/m3, Meshkov et al. 1986) was subtracted from the 

values. For 1957 and 1958 at the WSCP, for which measurements are not available, the analysis

assumed that perimeter air concentrations were the same as those measured in 1959, the operational 

year with the highest measured perimeter concentrations. This is probably an overestimate for the 

first 2 years, when uranium receipts were lower than those during the main production years (1960-

1964; Table 2-4 in Part 2 of this Site Profile). Measured concentrations are not available for the WSQ 

until 1961; until then, only drummed thorium wastes were reportedly stored at the quarry (Section 

2.2.3.2 in Part 2 of this Site Profile). Most of the material was dumped at an elevation below the 

present natural water table of the quarry (Weldon Spring Raffinate Pits and Quarry Task Force Report

1967). In 1963 and 1964, an estimated 38,000 m3 of contaminated rubble, equipment, and soil were 

placed in the WSQ, and much of this waste was not submerged. Thus, airborne concentrations of 

uranium and thorium resulting from use of the WSQ were probably negligible before 1963; this 

observation is supported by the measured concentrations of uranium in air in 1961 and 1962. 

The uranium concentrations (Bq/m3

) in Table 4-5 were converted to isotope-specific intakes, in Bq/yr, 

by assuming the ratio of 1:1:0.05 for 238U:234U:230Th activities and multiplying by an appropriate

inhalation rate. For perimeter concentrations, intakes were calculated by multiplying by an inhalation 

rate of 2400 m3

/yr, representing an assumption of continuous exposure over the work year. To 

address intakes associated with the localized contaminated dust arising from the reported hopper 

decontamination operation outside Building 103 (noted above), the analysis assumed that the 

operation took place for 1 hour per workday (12.5% of the work year) every year during the 

operational period, and that unmonitored individuals were at a location in which an average 

concentration of 4.4 Bq/m3 occurred for 5% of their work year (approximately 24 minutes per day). 

This corresponds to a yearly inhalation rate of 120 m3

/yr. The resulting average intakes are presented 

in Table 4-6.

In Table 4-6, the “

238U and 234U” columns for the operational period represent the sum of 238U and 234U 

intakes. Since natural uranium accounted for more than 97% of the nuclear materials throughput 

(Sect. 2.2.2.2 of this SPD), it is reasonable to assume that the activities of these two isotopes are 

equal in the measured concentrations, as they are approximately equal in natural uranium ore 

concentrates. Thorium-230 intake is assumed to be 5% of uranium isotope (in this case, either 238U or 

234U) intake, consistent with the assumptions by Meshkov et al. (1986), which are based on estimates 

for yellowcake composition. During the 10-year operational period, the amount of 230Th likely to be 

present on contaminated surfaces and in air is more reflective of the amount present in the original 

material being processed than of the daughter ingrowth due to decay of 234U. This assumption is 

applicable to post-1962 intakes for the WSQ, although the ratio is less certain for waste deposited in 

the quarry, which has not been well-characterized radiologically. 

Although the intake values in Table 4-6 for the operational period can be attributed largely to the one 

measured value of uranium downwind of the hopper cleanout operation in 1966, it is notable that the 

average value of 555 Bq/yr for total uranium intake (corresponding to 277 Bq/yr for 234U intake) 

compares closely with the average estimated intake of 366 Bq/yr for 234U at the Feed Materials Plant 

in Fernald, Ohio (ORAU 2004), for the same years of operation. The Fernald plant received 

yellowcake in similar quantities to those received by the WSP between 1957 and 1966, and used 

similar operations in processing the receipts for ultimate shipment. Table 4-7 lists site-wide maximum

intakes, representing the maximum intake each year for the areas listed in Table 4-6.

Radon

Measured air concentrations of radon during the operational period are not reported in the literature 

available for this TBD. Therefore, the analysis used a very simplistic screening-level model for 

estimating air concentrations at the WSCP, the location of the acid recovery plant, to estimate air 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 17 of 35

Table 4-6. Estimated average annual inhalation intake of radioactive air particulates and radon at 

WSCP, WSRP, and WSQ (assume respirable fraction of sampled particles = 1.0, and based on 2,400 

m3

/yr inhalation rate). 

Year

WSRP WSCP WSQ

U-238 and 

U-234

(Bq/yr)

Th-230

(Bq/yr)

Radon

(WLM/yr)

U-238 and 

U-234a

(Bq/yr)

Th-230

(Bq/yr)

Radon

(WLM/yr)

U-238 and

U-234a 

(Bq/yr)

Th-230

(Bq/yr)

Radon

(WLM/yr)

Operational period

1957 Not significant Not significant Not significant 5.7E+02 1.4E+01 8.7E-02 Not significant Not significant Not significant

1958 Not significant Not significant Not significant 5.7E+02 1.4E+01 8.7E-02 Not significant Not significant Not significant

1959 Not significant Not significant Not significant 5.7E+02 1.4E+01 8.7E-02 Not significant Not significant Not significant

1960 Not significant Not significant Not significant 5.6E+02 1.4E+01 8.7E-02 Not significant Not significant Not significant

1961 Not significant Not significant Not significant 5.7E+02 1.4E+01 8.7E-02 Not significant Not significant Not significant

1962 Not significant Not significant Not significant 5.4E+02 1.3E+01 8.7E-02 Not significant Not significant Not significant

1963 Not significant Not significant Not significant 5.5E+02 1.4E+01 8.7E-02 6.2E-01 3.1E-02 2.5E-02

1964 Not significant Not significant Not significant 5.5E+02 1.4E+01 8.7E-02 3.6E+00 1.8E-01 2.5E-02

1965 Not significant Not significant Not significant 5.4E+02 1.3E+01 8.7E-02 2.9E+01 1.5E+00 2.5E-02

1966 Not significant Not significant Not significant 5.4E+02 1.3E+01 8.7E-02 2.9E+01 1.5E+00 2.5E-02

1967 Not significant Not significant Not significant 5.4E+02 1.3E+01 8.7E-02 2.9E+01 1.5E+00 2.5E-02

Monitoring/remediation period

Year

WSRP WSCP WSQ

Gross alpha

(Bq/yr)

Radon

(WLM/yr)

Gross alpha

(Bq/yr)

Radon

(WLM/yr)

Gross alpha

(Bq/yr)

Radon

(WLM/yr)

1975-1979 0 4.5E-02 Not under DOE Not under DOE 0 3.1E-02

1980 0 1.1E-01 Not under DOE Not under DOE 0 2.4E-02

1981 0 7.1E-02 Not under DOE Not under DOE 0 2.5E-02

1982 0 5.6E-02 Not under DOE Not under DOE 0 2.5E-02

1983 0 2.0E-02 Not under DOE Not under DOE 0 2.1E-02

1984 0 3.5E-02 Not under DOE Not under DOE 0 2.4E-02

1985 0 2.3E-02 0 1.5E-02 0 2.4E-02

1986 0 3.3E-02 0 2.2E-02 0 2.3E-02

1987 0 2.6E-02 0 2.8E-02 0 4.3E-02

1988 0 4.2E-02 0 3.9E-02 0 6.5E-02

1989 4.9E-01 3.2E-02 3.9E-01 3.6E-02 4.6E-02 4.1E-02

1990 2.5E-01 1.9E-02 2.2E-01 2.3E-02 4.0E-02 3.3E-02

1991 1.2E-01 2.1E-02 1.3E-01 1.9E-02 8.0E-03 2.9E-02

1992 9.3E-02 2.1E-02 9.9E-02 1.5E-02 0 2.6E-02

1993 1.1E-01 9.6E-03 1.1E-01 7.1E-03 7.5E-03 5.8E-02

1994 1.3E-01 2.7E-02 1.5E-01 1.6E-02 5.8E-02 1.6E-01

1995 1.1E-01 3.3E-02 1.2E-01 2.1E-02 1.9E-02 5.5E-02

1996 1.5E-01 5.4E-02 1.4E-01 2.6E-02 8.0E-03 1.2E-02

1997 1.8E-01 1.2E-01 1.3E-01 2.1E-02 0 1.1E-02

1998 2.1E-01 2.2E-02 1.8E-01 1.3E-02 6.2E-03 7.1E-03

1999 1.9E-01 3.5E-02 1.8E-01 2.4E-02 8.9E-04 1.6E-02

2000 9.9E-02 1.9E-02 1.2E-01 2.5E-02 8.0E-03 1.2E-02

2001 9.9E-02 1.9E-02 1.2E-01 2.5E-02 8.0E-03 1.2E-02

2002 9.9E-02 1.9E-02 1.2E-01 2.5E-02 8.0E-03 1.2E-02

2003 9.9E-02 1.9E-02 1.2E-01 2.5E-02 8.0E-03 1.2E-02

2004 9.9E-02 1.9E-02 1.2E-01 2.5E-02 8.0E-03 1.2E-02

a. Represents the sum of 238U and 234U activities. Assume 238U and 234U are equal in activity.

concentrations of radon during this period. The model is from Volume 1 of National Council on 

Radiation Protection and Measurements Report 123 (NCRP 1996). 

The NCRP (1996) screening model adopted for these calculations is a simple dispersion model that 

accounts for potential increases in exposure due to building wake effects for close-in receptors (i.e., 

those less than 100 m from the source location). Air concentration (C) of radon was calculated from:

C = (f Q)/(π u h K) Ci/m3 (4-1)

where:

C = concentration at receptor (Ci/m3

) 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 18 of 35

Table 4-7. Estimated site-wide maximum 

inhalation intake of radioactive air particulates 

and radon at WSCP, WSRP, and WSQ

(assume respirable fraction of sampled particles 

= 1.0, and based on 2,400 m3

/yr inhalation rate).

Year

WSP (unspecified location)

U-238 and U-234a

(Bq/yr)

Th-230

Bq/yr

Radon

(WLM/yr)

Operational period

1957 5.7E+02 1.4E+01 8.7E-02

1958 5.7E+02 1.4E+01 8.7E-02

1959 5.7E+02 1.4E+01 8.7E-02

1960 5.6E+02 1.4E+01 8.7E-02

1961 5.7E+02 1.4E+01 8.7E-02

1962 5.4E+02 1.3E+01 8.7E-02

1963 5.5E+02 1.4E+01 8.7E-02

1964 5.5E+02 1.4E+01 8.7E-02

1965 5.4E+02 1.3E+01 8.7E-02

1966 5.4E+02 1.3E+01 8.7E-02

1967 5.4E+02 1.3E+01 8.7E-02

Year

WSP (unspecified location)

Gross alpha

(Bq/yr)

Radon

(WLM/yr)

Monitoring and remediation period

1975-1979 0 4.5E-02

1980 0 1.1E-01

1981 0 7.1E-02

1982 0 5.6E-02

1983 0 2.1E-02

1984 0 3.5E-02

1985 0 2.4E-02

1986 0 3.3E-02

1987 0 4.3E-02

1988 0 6.5E-02

1989 4.9E-01 4.1E-02

1990 2.5E-01 3.3E-02

1991 1.2E-01 2.9E-02

1992 9.3E-02 2.6E-02

1993 1.1E-01 5.8E-02

1994 1.5E-01 1.6E-01

1995 1.2E-01 5.5E-02

1996 1.5E-01 5.4E-02

1997 1.8E-01 1.2E-01

1998 2.1E-01 2.2E-02

1999 1.9E-01 3.5E-02

2000 1.2E-01 2.5E-02

2001 1.2E-01 2.5E-02

2002 1.2E-01 2.5E-02

2003 1.2E-01 2.5E-02

2004 1.2E-01 2.5E-02

a. Represents the sum of 238U and 234U activities. Assume 238U 

and 234U are equal in activity. 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 19 of 35

Q = stack or building vent release rate (Ci/s)

f = wind frequency

u = wind speed (m/s)

h = height of effluent release (m)

K = constant (m)

Values used for the independent factors f and K correspond to the defaults in NCRP (1996), such that 

f is assumed equal to 0.25 (the maximum frequency for any compass point), and K is 1 m. The 

average wind speed at the WSCP from 1951 to 1970 was 4.2 m/s from the south (Weidner and 

Boback 1982). The height of effluent release from the acid recovery plant is not known, but is 

assumed to be 10 m.

The estimate by Meshkov et al. (1986) of an average annual release of radon (222Rn) between 12 and 

34 Ci/yr (1.4 × 104 to 4.0 × 104 Bq/s) is based on the following: (1) 5,000 to 14,500 MT of uranium 

materials were processed per year; (2) 70% of this was 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 escaped during processing. Using this value in Equation 4-1, the estimated average 

radon concentration within 100 m of the source is between approximately 30 and 80 Bq/m3 (0.7 and 

2 pCi/L). Because the radon concentrations reported here include background radon contributions, 

the average value of 11 Bq/m3 (0.3 pCi/L) for the Weldon Spring area (MK-Ferguson 2001a) is added 

to the calculated estimates, such that the average concentration ranges from 41 to 91 Bq/m3

. This 

TBD uses the upper estimate, which is a claimant-favorable assumption.

A similar estimate for thoron release from the WSCP during the operational period was not made, in 

part because the type of thorium material processed is not sufficiently described in accessible 

documentation. DOE (1986) states that “natural thorium was typically received in either a nitrate or 

oxide form,” with no mention of it as a concentrated feed material. In addition, DOE (1986) indicates 

that the amount of thorium material processed ranged from a low of 0.05% of the natural uranium 

material processed in 1964 to a high of 7% in 1965. According to Wallo (1981), the ThO2 mass 

content, and thus the 232Th content, of natural thorium materials (like monazite sand) is approximately 

5%. Assuming this and a secular equilibrium between 232Th and 224Ra and 220Rn, it follows that the 

estimated release of thoron according to the procedure above for radon is between 0.04% and 5% of 

the radon release. If, however, the thorium material is assumed to be concentrated similarly to 

uranium (i.e., 70% 232Th), the thoron release could range from 0.5% and 70% of the radon release. 

Either way, the estimated dose attributable to thoron and its progeny is insignificant under these 

assumptions with respect to radon and its progeny because radon progeny have a higher estimated 

equilibrium factor and higher associated dose factors than thoron progeny (MK-Ferguson 2001b). 

This TBD analysis estimated radon concentrations for the WSQ during the operational period from 

measurements made in the vicinity from 1979 through 1982 (Meshkov et al. 1986; MK-Ferguson 

1989a). Because the activities of 226Ra and 224Ra are not significantly depleted in the quarry over 

time, due to limited leaching and continuous production of these isotopes from precursors present in 

the waste (Section 4.2.3.1), these years reasonably represent radon emanations for the WSQ during 

the years of operation of the plant. These measurements are applied to the 1963-1967 period only. 

Prior to that, drummed thorium waste was probably submerged and no significant source of radon

probably existed. Table 4-5 lists the average measured concentrations for this area.

Measurements of radon in 1979 through 1982 did not distinguish between 222Rn and 220Rn. 

Therefore, the analysis made a dose-maximizing assumption that all the radon measured was 222Rn. 

Because most of the dose from inhaled radon is due to alpha-emitting, short-lived daughters, the 

working level month (WLM) unit is often a preferred method of reporting inhalation intake exposure to 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 20 of 35

radon to specify the degree of equilibrium of radon daughters assumed in calculating exposure. For 

222Rn, 1 working level (WL) = 3,700 Bq/m3 (100 pCi/L). The calculation of WLM for 222Rn assumes a 

2,000-hr/yr occupational exposure period and an outdoor environmental radon daughter equilibrium 

factor of 30%, which is the upper limit of measured factors for the WSP (MK-Ferguson 2001b). The 

WLM/yr intake per Bq/m3 for 222Rn is calculated from

4

3

222 .9 1054

1703700

3.0 2000

/

/ − ×=

×

× = Bq m

WLM yr Rn (4-2)

The radon concentrations in Table 4-5 were converted to WLM/yr by multiplying activity 

concentrations by the factor derived in equation 4-2; the results are in Table 4-6, which lists inhalation 

intakes for relevant WSP areas. Table 4-7 lists site-wide maximum intakes.

4.2.3.2 Post-1974 Intakes (Monitoring and Remediation Periods)

After 1974, 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), which was specific to the raffinate pits and quarry. In October 1985, the monitoring program 

was expanded to include the WSCP (Bechtel 1986), which transferred back to DOE at that time. 

Beginning in 1987, air particulate monitoring was added (MK-Ferguson 1988) at the WSCP and 

WSRP, and at the WSQ in 1989 (MK-Ferguson 1990b). These data, reported in the annual 

monitoring reports for 1979 to 2000 (Weider and Boback 1982; Bechtel 1983a,b, 1984b, 1985a, 1986; 

MK-Ferguson 1987, 1988, 1989b, 1990b, 1991, 1992a, 1993, 1994, 1995, 1996, 1997, 1998a, 1999, 

2000b, 2001a), when remediation was complete, are the basis for the following estimates of air

concentrations and intakes at the WSRP, WSCP, and WSQ for the post-1974 periods.

Airborne Particulate Radionuclides

Table 4-5 lists air monitoring data for airborne particulate radionuclides during the monitoring and 

remediation periods according to year and WSP location. Figures 4-1 and 4-2 show the locations of 

monitoring stations for the WSRP and WSCP and for the WSQ, respectively. Radioactive air 

particulates were not monitored prior to the beginning of remediation activities at the site. In 1987 and 

1988, monitoring of particulates at a few perimeter locations for the WSRP and WSCP, and in 1989 at 

the WSQ, indicated that measured gross alpha concentrations were statistically indistinguishable from 

background (MK-Ferguson 1988, 1989b, 1990b). Therefore, the concentrations listed in Table 4-5 for

1987 and 1989 at the WSRP and WSCP 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 (which 

began in 1990 at the WSQ), this analysis assumed that radioactive air particulate concentrations prior 

to 1987 are the same as those measured in 1987 through 1989; therefore, Table 4-5 lists particulate 

concentrations from 1975 through 1986 as insignificant. There is no reason to believe that the late 

1980s data do not adequately represent differences in vegetative cover or wind events during the 

unmonitored period.

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

locations identified in Figures 4-1 and 4-2, even though many of the concentrations were statistically 

indistinguishable from background. The measurements occurred at perimeter locations for the 

WSRP, WSCP, and WSQ. To consider the possibility of higher onsite concentrations, the analysis 

used the ratio of onsite radon concentration to perimeter radon concentration to scale the perimeter 

particulate data. As explained in Section 4.2.3.2.2, the maximum ratio of average onsite to average 

perimeter radon concentrations is 2.0. It is reasonable to assume that this ratio represents the ratio 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 21 of 35

of average onsite to average perimeter air particulate concentrations because the source of the 

airborne radioactivity is the same. Therefore, the measured perimeter concentrations of airborne 

Figure 4-1. Air particulate (P), radon (R), and gamma (G) 

monitoring stations in WSCP and WSRP. 

particulate radionuclides were multiplied by a factor of 2 to represent the average onsite airborne 

concentrations in Table 4-5.

Section 4.2.1.1 identifies 227Ac, 231Pa, 210Po, 228Th, 230Th, 232Th, 234U, and 238U 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 WSRP, WSCP, and WSQ, representing 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 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 22 of 35

demonstrated a dose to the public of greater than 1 mrem (MK-Ferguson 2000a). Therefore, this TBD 

assumes that radioactive air particulate concentrations measured in 2000 are a reasonable estimate 

of the air concentrations in subsequent years to the present.

Figure 4-2. Air particulate (P), radon (R), and gamma (G) monitoring stations in WSQ. 

Table 4-6 lists estimated annual gross alpha activity intakes of airborne particulate radionuclides in 

terms of gross alpha (Bq/yr) between 1975 and 2004. The table does not include estimates for the 

WSCP from 1975 to 1984 because this facility did not transfer to DOE until 1985. The intakes were 

obtained by multiplying the respective gross alpha concentrations (Bq/m3

) in Table 4-5 by the 

assumed inhalation rate of 2,400 m3

/yr. 

Table 4-7 lists site-wide maximum gross alpha intakes for the remediation period. These intakes 

correspond to the highest value, by year, for any location listed in Table 4-6.

Dose reconstructors must assign the gross alpha activity intakes to a radionuclide or radionuclides 

likely to produce most of the organ dose. In lieu of this information, it is reasonable to consider the 

radionuclides capable of maximizing the calculated dose. Table 4-8 lists the ICRP (2001) organ dose 

factors for 238U, 234

U, 230Th, 210Po, 232Th, 228Th, 231Pa, and 227Ac. Only organs for which one or more 

isotopic dose factors exceed 2 × 10-5 Sv/Bq are listed. This lower limit was selected based on

consideration of the highest site-wide maximum gross alpha intake reported in Table 4-6, which is 

0.49 Bq/yr. This intake would correspond to an organ dose of at least 10-5 Sv/yr (1 mrem/yr). The 

highest organ dose factor for each organ or tissue is shown in bold type in Table 4-8. Except for 

dose to kidneys, the highest dose factor is associated with 227Ac. However, because airborne 

particulates in the postoperational period probably arise from contaminated soils, it is reasonable to 

also consider the relative prevalence of these radionuclides in soil. Table 4-9 lists the activity 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 23 of 35

concentration ratios estimated for these radionuclides in raffinate sludge (which became a potential 

source of air contamination during remediation), WSCP soil, and quarry soil (DOE 1992). Based on 

the information in Tables 4-8 and 4-9, Table 4-10 was constructed to identify the most important 

radionuclide contributing to the gross alpha based on consideration of organ dose factors and relative 

concentrations in soil for the three areas (WSRP, WSCP, and WSQ). The percentages in 

parentheses show the estimated percent contribution to total dose for the radionuclide listed, 

assuming the relative concentrations in Table 4-9 are representative of the air concentrations 

encountered. From Table 4-10, it is evident that 230Th is generally the most important contributor to 

dose for the WSRP and WSQ, which is a reasonable result considering the elevated concentration of 

this radionuclide relative to other alpha-emitters present (Table 4-9). Ac-227 and uranium isotopes 

are more important at the WSCP, where the 230Th concentrations are not found to be as relatively 

significant. Table 4-10 indicates that small contributors to activity may be relatively large contributors 

to dose.

Table 4-8. Organ dose factors for alpha-emitting radionuclides of concern at WSP. 

Radionuclide

Critical organ 50-year committed dose factora,b,c (Sv/Bq)

Bone 

surface Kidneys Liver

Red 

marrow

Ovary/

testesd

ET 

airway Lung

U-238 1.0E-05 3.6E-06 1.3E-06 1.1E-06 3.4E-07 6.5E-05 3.4E-05

U-234 1.1E-05 4.0E-06 1.5E-06 1.1E-06 3.9E-07 7.5E-05 4.1E-05

Th-230 1.5E-03 2.2E-05 1.9E-05 5.2E-05 1.2E-05 7.4E-05 4.0E-05

Po-210 9.4E-07 7.5E-06 3.9E-06 1.6E-06 1.6E-07 7.3E-06 1.7E-05

Th-232 1.5E-03 2.3E-05 2.3E-05 5.7E-05 1.3E-05 1.1E-04 7.7E-05

Th-228 2.8E-04 1.2E-05 7.7E-07 2.2E-05 2.4E-06 2.1E-04 2.1E-04

Pa-231 4.4E-03 4.3E-07 2.8E-07 3.6E-04 8.2E-09 1.1E-04 7.2E-05

Ac-227 1.9E-02 9.5E-08 4.3E-03 1.5E-03 2.3E-04 3.8E-04 2.2E-04

a. Critical organ is defined here as the organ associated with the highest committed dose over all

absorption classes, for particles of 5-μm AMAD. 

b. Dose factors from ICRP (2001). 

c. For ET Airways and Lung, the least soluble form (absorption type S) is associated with the highest dose 

factor; for all other organs, the most soluble form listed in ICRP (2001) (absorption type M or F) is 

associated with the highest dose factor. 

d. The highest dose factor for ovary and testes is listed (in most cases, the value is the same for both 

organs).

Table 4-9. Normalized concentrations of radionuclides in soil at WSP. 

Radionuclide

Raffinate sludge 

activity normalized to 

U-238 concentrationa 

Chemical plant soil 

activity normalized to 

U-238 concentrationa 

Quarry soil activity 

normalized to 

U-238 concentrationa

U-238 1.00E+00 1.00E+00 1.00E+00

U-234 1.00E+00 1.00E+00 1.00E+00

Th-230 2.50E+01 1.50E-01 1.70E+00

Po-210 6.20E-01 1.40E-01 1.70E+00

Th-232 2.60E-02 7.60E-02 1.30E-01

Th-228 2.50E-01 7.60E-02 4.80E-01

Pa-231 7.30E-01 3.80E-02 5.80E-03

Ac-227 5.80E-01 1.90E-02 2.90E-03

a. Most measurements made during remediation period in 1989 and 1990 (DOE 1992). 

Based on the information in these tables, the following recommendations are made for selecting 

representative radionuclides for the gross alpha intakes. For the raffinate pits area (WSRP), it is 

recommended that it be assumed that the combined 238,234U activity is 10% of the total, 230Th is 85% of 

the total, 228Th is 1% of the total, and 227Ac is 4% of the total activity. For the chemical plant area 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 24 of 35

(WSCP), it is recommended that it be assumed that the combined 238,234U activity is 85% of the total, 

230Th is 10% of the total, 228Th is 4% of the total, and 227Ac is 1% of the total activity. For the quarry 

area (WSQ), it is recommended that it be assumed that the combined 238,234U activity is 35% of the 

Table 4-10. Largest contributors to dose for the WSRP, WSCP, 

and WSQ.

a

Organ of interest WSRP WSCP WSQ

Bone surface Th-230 (72%) Ac-227 (40%) Th-230 (85%)

Kidneys Th-230 (97%) U-234/238 (52%) Th-230 (56%)

Liver Ac-227 (84%) Ac-227 (91%) Th-230 (56%)

Red marrow Th-230 (53%) Ac-227 (49%) Th-230 (75%)

Ovary/testes Th-230 (69%) Ac-227 (54%) Th-230 (82%)

ET airway Th-230 (79%) U-234/238 (75%) Th-230 (32%)

Lung Th-230 (76%) U-234/238 (67%) Th-228 (36%)

a. Based on organ dose factors in Table 4-8, and relative contributions to soil 

concentration listed in Table 4-9.

total, 230Th is 54% of the total, 228Th is 10% of the total, and 227Ac is 1% of the total activity. These 

recommendations are summarized as:

Percent activity to assume for gross alpha activity

WSRP WSCP WSQ

238/234U (total) 10 85 35

230Th 85 10 54

228Th 1 4 10

227Ac 4 1 1

Radon 

During the monitoring and remediation periods, outdoor radon was measured at the WSCP, WSRP, 

and WSQ (Bechtel 1986, MK-Ferguson 1987, 1988, 1989a,b, 1990b, 1991, 1992a,b, 1993, 1994, 

1995, 1996, 1997, 1998a,b, 1999, 2000b, 2001a). At the WSCP, 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. The ratio of radon concentrations in the 

disposal cell area to the WSCP perimeter ranged from 1.4 to 5.0 for 1999 and 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 (1998-2000). The maximum ratio of average radon 

concentrations at the interior locations to the average at the perimeter was 2.0.

This TBD assumes that the average radon concentration measured at the perimeter stations might

underestimate the average radon concentration over the WSCP and WSRP onsite areas during the 

remediation period by a factor of 2. Therefore, the measured concentrations were multiplied by 2 to 

provide an average net radon concentration for each of the two areas (WSCP and WSRP). Table 4-5 

lists the derived concentrations, which include contributions of radon from naturally occurring 226Ra 

and 228Ra. Again, these concentrations reflect total radon (222Rn and 220Rn) as measured by F-type 

alpha track detectors.

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 3.6. This is a significant 

increase; however, the interior of the quarry is fairly inaccessible, with a quick drop in elevation and an 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 25 of 35

abundance of vegetation. Thus, 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. The TBD assumes the radon concentration during this period of active 

remediation to be 3.6 times the concentration measured at the perimeter, as reflected in Table 4-5 

values.

The TBD analysis used equation 4-2 to convert the radon concentrations in Table 4-5, and Table 4-6 

lists the results, including inhalation intakes for relevant areas of the WSP. Table 4-7 lists site-wide 

maximum intakes.

4.3 EXTERNAL EXPOSURE TO ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

The external dose-producing radionuclides of concern for the WSP are those associated with the 235U, 

238U, and 232Th decay series, and include short-lived daughter isotopes of radium. External ambient 

exposure at the WSP is a result of gamma and X-ray radiation emitted from radionuclides in the ore 

concentrate stored on the site during the operational period, and from radionuclides in the raffinate 

pits and quarry. An aerial radiological survey of the WSP (Jobst 1976) indicated that for most of the 

site the normal terrestrial gamma exposure rate is 3 to 6 μR/hr at 1 meter above ground level, and the 

average cosmic exposure rate is 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 raffinate pits and the quarry, indicating the presence of manmade 

changes from the natural radioisotopes. 

4.3.1

Available documents do not contain monitoring data describing 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 primarily due to gamma radiation from short-lived decay products of 238U in 

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

46). Exposure to radionuclides accumulated in the raffinate pits and quarry would be mitigated 

somewhat by the shielding and physical barrier provided by water covering the pits entirely and some 

of the waste in the quarry for much of the year.

Ambient Exposure during Operational Period (1957-1966)

In lieu of site-specific data for the WSCP during the operational period, the TBD analysis used 

ambient dose rates for the Feed Materials Plant in Fernald, Ohio, as reported in the Occupational 

Environmental Dose TBD for the Fernald Environmental Management Project (FEMP; ORAU 2004). 

FEMP received yellowcake for processing similar to that received at the WSP, in similar quantities. A 

comparison of the natural uranium processed (shipped) between 1957 and 1966 by the WSP and 

FEMP indicated a nearly identical yearly average of approximately 11.5 × 106 kg uranium (DOE 1986; 

RAC 1995). From the Fernald TBD, ambient dose rates reported for the early operational period 

(prior to 1976) are not based on measurements made during those years, but were scaled from 

measured rates in later production years, according to production level. The reported ambient rate is 

averaged over all FEMP areas except the K-65 Silo area, which is infrequently occupied, and which 

does not have a physical counterpart at WSP. The net (background subtracted) average dose rate for 

1956 to 1970 was 0.18 mrem/hr, or 1,576 mrem/yr for continuous exposure. Assuming this is a 

reasonable dose rate estimate for the WSP during the operational period, the analysis added the 

WSP site background rate of 99 mrem/yr (Bechtel 1986) to derive an ambient dose rate for 

continuous exposure of 1,675 mrem/yr. This corresponds to a 2,000-hr/yr occupational exposure of 

382 mrem.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 26 of 35

4.3.2

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

many perimeter locations around the WSCP, the WSRP, and the WSQ (Bechtel 1983b, 1984a, 

1985a, 1986; MK-Ferguson 1987, 1988, 1989a,b, 1990b, 1991, 1992a,b, 1993, 1994, 1995, 1996, 

1997, 1998a,b, 1999, 2000b, 2001a). In 1986, locations in the WSCP and the WSRP were 

monitored, although the results did not indicate a significant difference in exposure rate from the 

perimeter stations. Figures 4-1 and 4-2 show the locations of monitoring stations. 

Ambient Exposure during Monitoring and Remediation Periods (1975-2004)

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 values reported 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 TBD 

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 radiation exposure rates existed prior to 

remediation in areas of the WSP that are close to waste in the raffinate pits or quarry, as recognized 

in Jobst (1976). A comprehensive radiological survey of the WSRP in 1982 and 1983 (Bechtel 1984b) 

indicated an average onsite gamma exposure rate of 23 μR/hr for that site, 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 prior to remediation. A comparison of the average survey-based 

value (200 mrem/yr) to the average perimeter value in 1983 of 88 mrem/yr (Bechtel 1984a) 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 rate in Table 4-11 

represents twice the average perimeter values for 1984 through 2005, and is recommended for 

environmental dose reconstruction. For 1975 through 1983, the value of 200 mrem/yr in Table 4-11 is 

the average survey value for 1982-1983, from Bechtel (1984b), as is considered the best value to use 

for those years. 

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

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

over the quarry floor, where contamination was greatest (Bechtel 1985b). These exposure rates 

correspond to annual doses between 70 and 2,500 mrem/yr for continuous exposure (8,760 hr/yr). 

An average value was not provided, nor were the original data available, so the utility of this 

information is limited. The extent of the high exposure rate area is not defined. The survey report did 

state, however, that characterization of the WSQ was “extremely difficult because of the rough 

terrain”, and that “the area was densely vegetated.” Figure 4-3 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. Thus, the perimeter exposure rates are appropriate for 

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

and 1995 (MK-Ferguson 1994, 1995, 1996). This TBD assumes that the exposure rate during the 

years the quarry bulk wastes were being excavated to be approximately 2,500 mrem/yr for continuous 

(8,760-hr/yr) exposure, in lieu of other relevant data. It is not reasonable to assume that an individual 

spent 24 hours a day at this location, and the dose reconstructor should make a special effort to 

define accurately the potential time of exposure inside the WSQ perimeter when calculating the best 

estimate dose. Table 4-11 lists the appropriate value to assume for external environmental exposure

at the WSQ.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 27 of 35

Table 4-11. Estimated ambient onsite dose for WSRP, WSCP, 

and WSQ at Weldon Spring. 

Year

WSRP averagea

(mrem/yr)b 

WSCP averagec

(mrem/yr)b 

WSQ averaged

(mrem/yr)b 

1957-1967 Not significant 1,675 Not significant

1975-1979 200 (e) 97

1980 200 (e) 97

1981 200 (e) 97

1982 200 (e) 85

1983 200 (e) 95

1984 236 (e) 125

1985 223 198 123

1986 184 155 96

1987 139 145 95

1988 126 120 73

1989 140 131 81

1990 133 132 77

1991 150 140 84

1992 137 135 73

1993 125 118 2,500

1994 120 118 2,500

1995 140 144 2,500

1996 135 165 62

1997 135 185 61

1998 126 139 57

1999 124 124 55

2000 61 113 48

2001 61f 113f 48f

2002 61f 113f 48f

2003 61f 113f 48f

2004 61f 113f 48f

a. Prior to 2000, average value is twice the average measured at perimeter 

locations for the WSRP.

b. Based on 8,760 hr/yr exposure; includes contributions from natural 

background.

c. Prior to 2001, average value is twice the average measured at perimeter 

locations for the WSCP.

d. Except for 1993-1995, values are the average measured at perimeter 

locations for the WSQ.

e. WSCP did not transfer to DOE until 1985.

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

location.

A 1987 radiologic characterization of the WSCP and WSRP (Marutzky et al. 1988) used portable 

scintillation counters to record external radiation, in counts per minute, at 880 locations across the 

WSRP and WSCP; cross-correlation with a pressurized ionization chamber allowed estimates of 

exposure rate, in μR/hr. Exposure rates ranged from 9 to 287 μR/hr, which corresponds to a dose 

rate between 79 and 2,500 mrem/yr for continuous exposure (8,760 hr/yr). One location had an 

exposure rate of 287 μR/hr; the next highest value was 85 μR/hr. Fewer than 10% of the 880 

locations surveyed had exposure rates measuring 15 μR/hr or above. 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 WSCP area 

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

areas are probably not those in which occupational exposures would normally occur. The elevated 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 28 of 35

Figure 4-3. Topographic map and locations of gamma 

exposure rate measurements for WSQ (from Bechtel 1985b). 

exposure rates were in the areas surrounding the buildings, in onsite dump areas, in Ash Pond, and in

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 WSCP perimeter exposure rate was 72 mrem/yr. Thus, for the 

WSCP, the TBD analysis assumed that until remediation was essentially completed in 2000, the 

average onsite exposure rate was approximately twice the measured perimeter rate. Table 4-11 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 averaged 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 4-11 lists these averaged rates.

Table 4-12 lists site-wide maximum external exposure rates for WSP during the remediation period.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 29 of 35

Table 4-12. Estimated maximum site-wide

ambient dose at WSP. 

Year

Maximum for WSRP, WSCP, 

and WSQa (mrem/yr)b 

1957-1967 1,675

1975-1979 200

1980 200

1981 200

1982 200

1983 200

1984 236

1985 223

1986 184

1987 145

1988 126

1989 140

1990 133

1991 150

1992 137

1993 2,500

1994 2,500

1995 2,500

1996 165

1997 185

1998 139

1999 124

2000 113

2001 113c

2002 113c

2003 113c

2004 113c

a. Maximum of annual recommended average value 

for each area (from Table 4-11).

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

c. Not monitored after 2000; assumed to be equal to 

the site-wide maximum for the WSP for 2000.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 30 of 35

REFERENCES

ATSDR (Agency for Toxic Substances and Disease Registry), 1990, Toxicological Profile for Radium, 

U.S. Department of Health and Human Services, Public Health Service, Atlanta, GA.

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

Calendar Year 1982, Oak Ridge, Tennessee, May.

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

Calendar Year 1981, Oak Ridge, Tennessee, May. 

Bechtel (Bechtel National, Inc.) 1984a, Weldon Spring Site Environmental Monitoring Report, 

Calendar Year 1983, DOE/OR/20722-16, Oak Ridge, Tennessee, June.

Bechtel (Bechtel National, Inc.), 1984b, Radiological Survey Report for the Weldon Spring Raffinate 

Pits Site, DOE/OR/20722-7, Oak Ridge, Tennessee, August. 

Bechtel (Bechtel National, Inc.), 1985a, Weldon Spring Site Environmental Monitoring Report, 

Calendar Year 1984, DOE/OR/20722-58, Oak Ridge, Tennessee, July. 

Bechtel (Bechtel National, Inc.), 1985b, Radiological Survey Report for the Weldon Spring Quarry, 

DOE/OR/20722-70, Oak Ridge, Tennessee, August. 

Bechtel (Bechtel National, Inc.), 1986, Weldon Spring Site Annual Site Environmental Monitoring 

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

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

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

Ridge Operations Office, Oak Ridge, Tennessee, July. 

DOE (U.S. Department of Energy), 1990, Baseline Risk Evaluation for Exposure to Bulk Wastes at the 

Weldon Spring Quarry, Weldon Spring, Missouri, DOE/OR/21548-065, Oak Ridge Operations 

Office, Oak Ridge, Tennessee, July. 

DOE (U.S. Department of Energy), 1992, Baseline Assessment for the Chemical Plant Area of the 

Weldon Spring Site, DOE/OR/21548-091, Oak Ridge Operations Office, Oak Ridge, 

Tennessee, July. 

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

Holt, D. L., 1960, “Decontamination of Five-Ton Portable Hoppers for Shipment,” letter report to W. L. 

Utnage, Mallinckrodt Chemical Works, Uranium Division, Weldon Spring, Missouri, 8 April.

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

for Radiological Protection, Publication 66, Annals of the ICRP, Vol. 24/1-3, Pergamon Press, 

Oxford. 

ICRP (International Commission on Radiological Protection), 2001, ICRP CD1: Database of Dose 

Coefficients: Worker and Members of the Public, Elsevier B.V.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 31 of 35

Jobst, J. E., 1976, Aerial Radiological Survey of the Weldon Spring Chemical Plant (St. Charles, 

Missouri), Date of Survey: 22 November 1975, EGG-1183-1688, EG&G Inc., Las Vegas 

Operations, Las Vegas, Nevada, April. 

Marutzky, S. J., 1988, Radiologic Characterization of the Weldon Spring, Missouri, Remedial Action 

Site, DOE/ID/12584-22, UNC Geotech, Incorporated, Grand Junction, Colorado, February. 

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

ANL/EES-TM-308, Argonne National Laboratory, Argonne, Illinois, July. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1987, Weldon Spring Site Annual 

Environmental Monitoring Report, Calendar Year 1986, DOE/OR/21548-T1, St. Charles, 

Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1988, Weldon Spring Site Annual 

Environmental Monitoring Report, Calendar Year 1987, DOE/OR/21548-015, St. Charles, 

Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1989a, Remedial Investigations for the 

Quarry Bulk Wastes, Revision 1, DOE/OR/21548-066, St. Charles, Missouri.

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group),1989b, Annual Environmental 

Monitoring Report, 1988, for the Weldon Spring Site Remedial Action Project, DOE/OR/21548-

079, St. Charles, Missouri.

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1990a, Evaluation of Radon Emissions 

and Potential Control Requirements, DOE/OR/21548-082-Rev.2, St. Charles, Missouri.

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1990b, Annual Environmental 

Monitoring Report, 1989, For the Weldon Spring Site Remedial Action Project,

DOE/OR/21548-129, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1991, Annual Environmental 

Monitoring Report, 1990, For the Weldon Spring Site Remedial Action Project,

DOE/OR/21548-193, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1992a, Weldon Spring Site 

Environmental Monitoring Report for Calendar Year 1991, Revision 1, DOE/OR/21548-283, St. 

Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1992b, Remedial Investigation for the 

Chemical Plant Area of the Weldon Spring Site, Volume 1, DOE/OR/21548-074, Vol. I, St. 

Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1993, Weldon Spring Site 

Environmental Report for Calendar Year 1992, DOE/OR/21548-372, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1994, Weldon Spring Site 

Environmental Report for Calendar Year 1993, DOE/OR/21548-436, St. Charles, Missouri. 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 32 of 35

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1995, Weldon Spring Site 

Environmental Report for Calendar Year 1994, DOE/OR/21548-512, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1996, Weldon Spring Site 

Environmental Report for Calendar Year 1995, DOE/OR/21548-592, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1997, Weldon Spring Site 

Environmental Report for Calendar Year 1996, DOE/OR/21548-676, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1998a, Weldon Spring Site 

Environmental Report for Calendar Year 1997, DOE/OR/21548-730, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1998b, Remedial Investigation for the 

Quarry Residuals Operable Unit of the Weldon Spring Site, Weldon Spring, Missouri, 

Volume I, DOE/OR/21548-587, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 1999, Weldon Spring Site 

Environmental Report for Calendar Year 1998, DOE/OR/21548-773, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 2000a, External Dosimetry Technical 

Basis Manual, Rev. 6, Weldon Spring Site Remedial Action Project, DOE/OR/21548-441, 

Weldon Spring, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 2000b, Weldon Spring Site 

Environmental Report for Calendar Year 1999, DOE/OR/21548-845, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 2001a, Weldon Spring Site 

Environmental Report for Calendar Year 2000, DOE/OR/21548-886, St. Charles, Missouri. 

MK-Ferguson (MK-Ferguson and Jacobs Engineering Group), 2001b, Internal Dosimetry Program 

Technical Basis Manual, Rev. 7, Weldon Spring Site Remedial Action Project, 

DOE/OR/21548-241, Weldon Spring, Missouri. 

NCRP (National Council on Radiation Protection and Measurements), 1996, Screening Models for 

Releases of Radionuclides to Atmosphere, Surface Water and Ground, Report 123, Vol. 1, 

Bethesda, Maryland. 

NLO (National Lead of Ohio, Inc.), 1977, Study of Radioactive Waste Storage Areas at ERDA-Weldon 

Spring Site, NLCO-1144 (Special), Cincinnati, Ohio, April. 

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

Environmental Management Project - Occupational Environmental Dose, ORAUT-TKBS-0017-

4, Oak Ridge, Tennessee.

RAC (Radiological Assessments Corporation), 1995, The Fernald Dosimetry Reconstruction Project, 

Tasks 2 and 3, Radionuclide Source Terms and Uncertainties, Reports CDC-5, Neeses, South 

Carolina. 

Tuli, Jagdish K., 2000, Nuclear Wallet Cards, 6th Edition, National Nuclear Data Center, Brookhaven 

National Laboratory, Upton, New York, January. 
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 33 of 35

Wallo, A., III, 1981, “Preliminary History of AEC Thorium Research and Development Operations” 

attached to letter to William E. Mott (U.S. Department of Energy), Aerospace Corporation, 

2 March. 

Weidner, R. B., and M. W. Boback, 1982, Weldon Spring Storage Site Environmental Monitoring 

Report for 1979 and 1980, NLCO-1176, NLO, Inc., Cincinnati, Ohio. 

Weldon Spring Raffinate Pits and Quarry Task Force Report, 1967, DOE Germantown Records 

Holding Area, June 1, 1967.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 34 of 35

GLOSSARY

alpha particles

Positively charged particles of discrete energies emitted by certain radioactive materials; alpha 

particles usually expend their energy in short distances and will not usually penetrate the outer 

layer of skin; they are a significant hazard only when taken into the body where their energy Is 

absorbed by tissues.

becquerel

A special unit of activity. One Becquerel equals 1 nuclear transition per second.

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) of the direct 

fission products emit beta radiation. Physically, the beta particle is identical to an electron 

moving at high velocity.

curie

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

exposure

As used in the technical sense, a measure expressed in roentgens (R) of the ionization 

produced by photons (i.e., gamma and X-rays) in air. 

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus. 

isotope

Elements having the same atomic number but different atomic weights; identical chemically 

but having different physical and nuclear properties. 

radiation

Alpha, beta, neutron, and photon radiation. 

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

radionuclide

A radioactive isotope of an element, distinguished by atomic number, atomic weight, and 

energy state. 

raffinate

A term loosely applied to waste from the uranium or thorium extraction step. Solids that result

from the neutralization of this are also called “raffinates”.
Document No. ORAUT-TKBS-0028-4 Revision No. 00 Effective Date: 06/28/2005 Page 35 of 35

rem

A unit of dose equivalent equal to the product of the number of rad absorbed and the quality 

factor.

thermoluminescent dosimeter

A holder containing solid chips of material that when heated will release stored energy as light. 

The measurement of this light provides a measurement of absorbed dose.