ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 31

Document Title:

Paducah Gaseous Diffusion Plant –

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0019-4

Revision: 01

Effective Date: 11/07/2006

Type of Document: TBD

Supersedes: Revision 00

Subject Expert: James East

Approval: Signature on File Approval Date: 10/18/2006

Daniel Mantooth, Document Owner

Approval: Signature on File Approval Date: 10/24/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/24/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 10/23/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 11/07/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 2 of 31

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/24/2004 00 Technical Basis Document for the Paducah Gaseous Diffusion Plant –

Occupational Environmental Dose. First approved issue. Initiated by 

Jay J. Maisler.

11/07/2006 01 Approved Revision 01 initiated to address comments from earlier 

unresolved comments received on TBD. Constitutes a total rewrite of 

document. Incorporates additional information obtained through 

additional data capture for the Paducah Gaseous Diffusion Plant. This 

revision to addresses the Worker Outreach comment: "The use of the 

Site Profile is redundant and does not help claimants in any way. The 

Site Profile only represents incorrect, inaccurate, and incomplete 

information provided by DOE" This revision involved researching

documents to provide a revision which ensured that the most correct, 

accurate, and complete information available was used. Text was 

added or modified in sections 4.1, 4.2, 4.3, 4.4,and in the Acronyms 

and Abbreviations, References, and Glossary. No sections were 

deleted. Incorporates internal, NIOSH and DOL formal review 

comments. This revision results in an increase in assigned dose and a 

PER is required. The increase in assigned dose is based on the 

revision to Table 4-2 that adds transuranic exposures in the 

environment. Training required: As determined by the Task Manager. 

Initiated by Paul A. Szalinski.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 3 of 31

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

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

4.1.1 Purpose........................................................................................................................7

4.1.2 Scope...........................................................................................................................7

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

4.2.1 Preoperational Background Survey ..............................................................................7

4.2.2 Ambient Air Sample Collection Network .......................................................................7

4.2.3 Methodology.................................................................................................................9

4.2.4 Estimation of Ambient Airborne Radionuclide Concentrations .................................... 10

4.2.5 Data for Ambient Airborne Radionuclide Concentrations and Annual 

Airborne Releases......................................................................................................10

4.2.6 Estimation of Annual Intake from Airborne Radionuclides........................................... 12

4.2.7 Summary of Transuranic and Fission Product Exposure ............................................ 14

4.3 External Dose ........................................................................................................................18

4.3.1 Ambient Radiation ......................................................................................................18

4.3.2 Radiological Conditions in Cylinder Storage Yards .....................................................20

4.3.3 Radiological Conditions Inside Buildings ....................................................................20

4.4 Uncertainty.............................................................................................................................20

References ........................................................................................................................................22

Glossary ............................................................................................................................................26

Attachment 4A, OUTDOOR BETA AND ALPHA CONCENTRATIONS.............................................. 29
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 4 of 31

LIST OF TABLES

TABLE TITLE

4-1 Ambient air sampling stations ..................................................................................................

PAGE

8

4-3 Uranium releases, alpha airborne concentrations, and intakes .............................................. 16

4-4 Technetium-99 releases, beta airborne concentrations, and intakes ...................................... 17

4-5 External dose equivalent rates, PGDP external gamma exposure rate of 

mrem/2,000 hr........................................................................................................................19

4A-1 Outdoor beta concentrations .................................................................................................. 30

4A-2 Outdoor alpha concentrations ................................................................................................ 31

LIST OF FIGURES

FIGURE TITLE

4-1 Wind rose from MMES.............................................................................................................

PAGE

9

4-2 Alpha air concentrations......................................................................................................... 11

4-3 Beta air concentrations .......................................................................................................... 11

4-4 Cascade feed in thousands of tons of uranium, 1952 to 1978 ................................................ 13

4-5 2001 TLD locations ................................................................................................................19
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 5 of 31

ACRONYMS AND ABBREVIATIONS

Bq becquerel

CEDE committed effective dose equivalent

cfm cubic feet per minute

CIP Cascade Improvement Program

cm centimeter

cpm counts per minute

CUP Cascade Upgrade Program 

DOE U.S. Department of Energy

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act

FP fission product

ft foot

g gram

GM Geiger-Muller

hr hour

kdpm kilodisintegrations per minute

LMES Lockheed Martin Energy Systems

m meter

mg milligram

MMES Martin Marietta Energy Systems

mR milliroentgen

mrem millirem

NIOSH National Institute for Occupational Safety and Health

PGDP Paducah Gaseous Diffusion Plant

RU recycled uranium

s second

TBD technical basis document

TLD thermoluminescent dosimeter

TRU transuranic

U.S.C. United States Code

yr year

§ section
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 6 of 31

4.1 INTRODUCTION

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

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

documents that provide historic background information and guidance to assist in 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 NIOSH staff in the completion of the individual work required for each dose reconstruction.

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

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

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

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

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)].

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 7 of 31

4.1.1

This TBD provides technical data and other key information, which will serve as the technical basis for 

evaluating environmental radiation dose for EEOICPA claimants who were employed at the Paducah 

Gaseous Diffusion Plant (PGDP).

Purpose

4.1.2

This document discusses the radionuclides potentially encountered by PGDP employees and 

released to the environment during the Plant’s operational history. The PGDP mission was to enrich 

uranium in the form of uranium hexafluoride (UF6) from roughly 0.7% uranium-235 (

235U) (natural 

enrichment) to around 3% 235U for use in domestic and foreign power reactors (DOE 2000a). 

Enrichment operations began in 1952 in the first four process buildings, C-331, C-333, C-310, and C315. From 1953 until 1977, UF6 feed material was produced from uranium trioxide (UO3) at the plant. 

From 1953 to 1964, and again from 1968 to 1977, UF6 was produced from the recycled uranium (RU) 

produced from spent reactor fuel. In May 1977, the feed plants ceased operation and all feed to the 

enrichment process was in the form of UF6 obtained from outside sources. Other chemical 

compounds of uranium were present throughout the Plant’s history including uranyl fluoride, uranium 

fluoride, and UO3.

Scope

ORAUT (2004) contains detailed information on the history of PGDP and the feed conversion and 

enrichment process.

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

4.2.1

The Carbide and Carbon Chemicals Company analyzed samples for airborne uranium in August 1952 

to document the preoperational conditions at PGDP (CCCC 1953). Ten chemically processed 

samples showed 0.00 mg/m3 uranium in the air. Two samples analyzed for alpha activity had results 

of 0.00 and 0.018 cpm/ft3 (approximately 0.01 Bq/m3

). However, what is known about the collection 

process and analysis of samples indicates that these results are not sufficient to conclude that there 

was no preexisting airborne radioactivity in the area. As sampling and analysis techniques improved 

over the years, the presence of measurable radioactivity from sources other than PGDP became 

evident (such as fallout from atomic weapons testing and fly ash from nearby fossil-fuel power plants).

Preoperational Background Survey

4.2.2

By 1958, PGDP had established a network of permanent stations on and off the site to collect 

continuous ambient air samples (UCNC 1959). Sampling data are available in annual environmental 

reports published since 1958 for four perimeter locations (inside the fence) and varying numbers of 

offsite locations. The 1959 environmental monitoring report describes outdoor air sampling:

Ambient Air Sample Collection Network

The outdoor air samples for alpha and beta active particulates were collected 

approximately every eight days at the same locations as the fluoride samples. They 

were 24-hour samples and were collected on Whatman number 40 filter paper at a flow 

rate of 0.5 cubic feet per minute. These samples were counted for alpha activity on 

parallel plate counters equipped with amplifiers and scalers and were counted for beta 

activity with Geiger-Muller tubes equipped with scalers (Brown and Mitchell 1960).
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 8 of 31

An air sampling procedure from September 1962 describes environmental air sampling: “The 

samples are collected by a 110 volt Gast pump using membrane filter type AM-4 2 in. at a flow rate of 

0.3 cfm (11 cm/s). Samples are continuous and of week-long duration” (UCNC 1962a). The 

environmental report for 1960 indicates that the change in monitoring techniques occurred about

mid-1960 (UCNC 1961). A paper from PGDP describes the air sampling equipment used during this 

period (UCNC 1962b).

The principal purpose of the ambient air-monitoring network was to assess if air emissions from 

PGDP affected the air quality in the surrounding area. This would demonstrate compliance with DOE 

Derived Concentration Guidelines or the U.S. Environmental Protection Agency, the State of 

Kentucky, and (as of 1997) the U.S. Nuclear Regulatory Commission regulations for airborne releases 

to the public around PGDP. Therefore, the majority of the monitoring stations are off the site or just 

inside the security fence. There have been a limited number of onsite monitoring stations, but reports 

from these stations are not available. Therefore, in the calculation of worker intake, this TBD 

considers only those results from ambient air-monitoring locations adjacent to or outside the PGDP 

security fence. Table 4-1 lists these monitors and their locations. Publications from Brown and 

Mitchell (1960), Martin Marietta Energy Systems (MMES 1989), and Lockheed Martin Energy 

Systems (LMES 1996) contain maps showing the old and new naming conventions.

Table 4-1. Ambient air sampling stations.

Designation Location

Direction from 

center of plant

PN-6 At security fence between C-535 and C-537 North

PE-8 At security fence near McCaw Road East

PS-2 At security fence approximately 300 meters south of C-100 South

PW-5 At security fence near C-749 West

IN Offsite about 2 km north of PN North

IE Offsite about 1 km east-northeast of PE East

ISE Offsite about 2 km southeast of PS Southeast

IS Offsite approx 1.5 km south-southwest of PS South

IW Offsite approx 1.2 km west-northwest of PW West

BN At the north boundary of property, 0.3 km north of PN North

Several changes to the sampling program occurred over time, primarily involving sample technique 

and equipment. From 1958 to 1993, low-volume samplers were the principal samplers, with multiple 

filter types and flow rates. 

The general practice for air samples was to perform gross alpha and beta-gamma counts. The 

activity of gross alpha was associated with the release of uranium, while beta-gamma activity was 

associated with the release of technetium-99 (

99Tc). Overall, the alpha activity was about 7% of the 

beta activity in air samples. 

There have been releases of radionuclides to the atmosphere since the beginning of operations, 

including accidental releases. PGDP release data have been estimated or recorded since operations 

began in 1953. The uncertainty in the amounts released and lack of a complete ambient airmonitoring network for the first 6 years of operation are technical issues that the occupational dose 

reconstruction must address. Therefore, Section 4.2.3 discusses a method to estimate potential 

radionuclide air concentrations for 1953 to 1958 and to account for potential uncertainty in the release 

data.

There are several points of airborne release. These include the stacks for Buildings C-310, C-340, 

C-400, and C-410. The first two stacks are on the east side of the plant, while the remaining release 
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 9 of 31

points are near the center of the plant. General plant ventilation exhausts released lesser 

concentrations over large areas of the building roofs. The many sources of airborne releases, use of 

stacks, and the nearly constant wind contribute to a very effective diffusion of contaminants over the 

small site with no significant terrain features to channel or moderate the wind. Figure 4-1 shows a 

wind rose from the 1993 annual environmental report for PGDP, which shows winds blowing 

predominantly from the south-southwest (MMES 1994). When the wind rose is laid over the plant 

layout, the predominant winds line up with the three cascade buildings.

Figure 4-1. Wind rose from MMES (1994).

4.2.3

The estimation of airborne concentrations at specific locations around the PGDP site using traditional 

transport modeling approaches is limited by several factors:

Methodology

• The numerous release points, which include stacks, vents, and other emission sources

• The characteristics of the release points

• The limited number of air-sampling locations

• The relatively short distances between the release points and the onsite receptor locations

• The density and configurations of buildings at the site

Past estimates of doses to members of the public off the site indicate that the potential internal dose 

from airborne releases to PGDP workers should be relatively low, about 10 mrem committed effective 

dose equivalent (CEDE) or less. Air data from nearby locations and at the security fence are 

consistent from location to location, but fallout levels from weapons testing dominated the results 

through the early 1960s. This collection of data provides reasonable approximations of general 

airborne radioactivity and establishment of trends as a function of time.

Factors other than transport and release rates from PGDP influence the gradual reduction in the air 

concentration over the 50-year history of PGDP. The records show that releases fell by five orders of 
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 10 of 31

magnitude while concentrations decreased by less than three orders of magnitude. These factors 

include: 

• Improved monitoring methods reduced the minimum detectable concentrations.

• Concentrations of naturally occurring radioactive material released from a nearby fossil-fuel 

power plant fell over the years because of improved environmental controls. (Figure 4-5 in 

Section 4.3.1 shows the proximity of the power plant.)

• The effect of atmospheric weapons testing and the radioactive decay of fallout (air-monitoring 

results correspond more with fallout than with plant releases).

The methodology for the intake from onsite atmospheric radionuclide concentrations applies directly 

to available air-sampling data. The maximum site measurement for gross alpha and gross betagamma should be applied for all workers throughout the site. Air-sampling measurement error and 

uncertainty should be accounted for by use of the maximum value reported for the year at any 

location. Available data are limited to annual averages for each location; the limited amount of data 

does not support statistical analysis. 

4.2.4

In years for which air concentration data are unavailable, release data and maximum air concentration 

data for adjacent years were evaluated to generate estimates of the air concentration. The alpha 

concentrations chosen were based on the higher value indicated by releases (1952 to 1956) or by 

air-monitoring trends (1990 and 1994). Similarly, the beta concentrations were chosen to be 

proportional to releases (1952 to 1958 and 1989) and for the later years (1993 and 1995 to 2001) as 

an approximate average of the preceding 10 measurements, which results in higher estimates than 

the downward trend.

Estimation of Ambient Airborne Radionuclide Concentrations

Figures 4-2 and 4-3 show these data and demonstrate that chosen values are reasonable and will not 

underestimate dose. These graphs, one for alpha and one for beta, show release values for each 

year available, measured airborne concentrations, and estimates of assigned values for airborne 

concentrations if actual measurements were not available.

4.2.5

This TBD relied on several data sources for estimating ambient airborne radionuclide concentrations 

and for annual airborne releases by radionuclide. The principal sources were the annual 

environmental reports for PGDP from 1958 to 2001 (see the reference list at the end of this 

document). Measurements from the air-monitoring locations were collected from these documents 

and applied using the method described. Attachment 4A provides an expanded version of this data, 

including additional background monitoring points. If data were not provided or additional information 

was required, these documents were supplemented by information provided by DOE, Bechtel Jacobs 

Company (the current DOE environmental restoration contractor at PGDP), or the United States 

Enrichment Corporation. 

Data for Ambient Airborne Radionuclide Concentrations and Annual Airborne 

Releases
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Uranium Releases and Alpha Air Concentrations

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1950 1955 1960 1965 1970 1975 1980 1985 1990 1995 2000

Alpha Air Concentrations (Bq/m^3)

1.E+05

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

1.E+11

1.E+12

Uranium Releases (Bq/y)

2 PS Alpha 5 PW Alpha 6 PN Alpha 8 PE Alpha Estimate BN BE

IN IE ISE IS IW GR U Releases

Figure 4-2. Alpha air concentrations.

Technetium Releases and Beta Air Concentrations

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1950 1955 1960 1965 1970 1975 1980 1985 1990 1995 2000

Beta Air Concentrations (Bq/m^3)

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

1.E+11

1.E+12

Tc Released (Bq/y)

2 PS Beta 5 PW Beta 6 PN Beta 8 PE Beta

Estimate BN BE IN

IE ISE IS IW

GR Tc99 Releases

Figure 4-3. Beta air concentrations.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 12 of 31

4.2.6

Using the maximum airborne radionuclide concentration in a year (either from actual data or an 

estimated value), annual intakes for the radionuclides were derived with an assumed annual 

respiration rate of 2,400 m3

/yr. 

Estimation of Annual Intake from Airborne Radionuclides

Transuranic (TRU) radionuclides were introduced to the facility with RU feed material beginning in 

1953 and continuing to 1977 (for 5 of the years during this period RU was not included in the feed 

material) (Miners 1990). There were areas in the buildings that contributed to intake, mainly around 

the ash from the fluorination process and the cylinder heels. 

The facility operators soon became aware of the rising TRU presence in the feed plant ash, cylinder 

heels, and to some extent, in the cascade (PACE 2000, DOE 2000b). There was a period when

neptunium-237 (

237

Np) was recovered from feed plant ash and shipped to the Oak Ridge National 

Laboratory (DOE 2000a). 

Workers and safety personnel were used to being able to see contamination. This was due to the low 

specific activity of natural and low-enriched uranium and because enough material had to be present 

to be visible as a stain before the meters of the time would indicate contamination. Management 

became concerned in 1959 when contamination was discovered on a piece of equipment without a 

stain, and soon concluded that the source of the contamination was 50% neptunium by chemical and 

pulse height analysis at 25 kdpm/m2

. Note that the 25 kdpm/m2 was loose contamination determined 

by swipe, and the 50% neptunium was determined by scraping the metal surface of the spool piece. 

Due to the known plating characteristics of neptunium, it is likely that the isotopic distribution from the 

scraping sample might not be representative of the swipe’s isotopic distribution (CCCC 1960).

A paper prepared for the 1964 Bioassay Conference reports these findings (Baker 1964). The paper 

includes a discussion of the dust characteristics, established neptunium controls, and monitoring for 

exposure. The nature of the contamination was studied to characterize particle size and solubility as 

well as retention and distribution of neptunium in laboratory animals. Retention was reported at 2% 

while more than 90% was excreted in feces and 2% in urine. These factors reduce the dose impact of 

an uptake of neptunium.

There were also bioassays performed on the highest exposed individuals. Urinalysis and whole-body 

counting came back negative on these workers. Later studies showed that widespread distribution in 

the cascade was not likely. The surface area consisted of 90% barrier material, and that is where the 

neptunium would have been found. In addition, it would have been removed from the cascade during 

the two upgrade programs that replaced barrier material.

In 1987, Exposure Assessment – Uranium Recycle Material in the Paducah Feed Plant, stated:

From old records and interviews, estimates were made of 20-year average air 

concentrations and the particle size of the radioactive aerosols in the various 

operations areas and of the number of hours operators spent in each job. The fraction 

of the airborne radioactivity due to the various radionuclides was estimated for each 

area. Deposition factors were taken from Figure 5.1 of ICRP Part 1 to provide a 

correction for particle size. External radiation exposure was determined by film badge 

dosimetry. From this the average annual 50-year committed dose equivalent was 

calculated to be 1.54 rem/yr. Of this, TRU and FP's [fission products] contributed 12 

percent. (Baker 1987).
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Further:

In 1962, 15 employees considered to be among those with the highest potential 

exposure to neptunium and uranium in each of three plant groups (237Np recovery, 

converter disassembly, feed plant operations) were sent to Y-12 to be counted. None 

of the employees had detectable 237Np (Baker 1987).

Figure 4-4. Cascade feed in thousands of tons of uranium, 1952 to 1978 (Miners 

1990).

The Y-12 counter had been calibrated specifically for 237Np. In addition, “Deposition of 237Np in the 

lung from exposure to dust from cascade equipment disassembly and modification was lower than 

predicted by air sampling because of the large size of the dispersed particulates” (Baker 1987).

There were comments about some very high neptunium alpha fractions, especially during the 

maintenance on converters during the Cascade Improvement Program and Cascade Upgrade 

Program (CIP/CUP) years. Some fractions were as high as 90%. Little is known about how these 

fractions were determined in the early years. As assay techniques improved, such high alpha 

fractions have not been found.

From Neptunium Experience at PGDP:

Neptunium, as discussed in the chemistry section of this report, is relatively immobile. 

A survey of equipment removed from the cascade during the more recent upgrade 

program showed Np concentrated in the vicinity of the historical feed points for RU, 

several years after it had been fed to those locations in quantities sufficient to account 
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 14 of 31

for the material found. On the other hand, a small proportion of product cylinders in the 

late 1970s showed ppb levels of Np. Thus, there may be a very slight tendency to 

mobility on a time scale of decades. Most likely, the Np fed to the cascade is still in the 

equipment to which it was fed. Some of the converters, however, were physically 

relocated within the cascade, and a large number had their barrier and other cascade 

components removed during upgrade programs (Miners 1990).

In addition: 

The 1974 study examined, among other things, soil contamination levels in drainage 

ditches that had been used during early years of the plant for discharge of liquid 

streams potentially containing radionuclides. From these analyses it was estimated 

that less than 4 grams of Np were present in the soil of these drainage ditches (Miners

1990).

4.2.7

To determine the general risk from TRU and fission product exposures, an analysis was performed to 

determine the relative intake of TRU materials and 99Tc to uranium. PACE (2000) reports 99Tc as the 

fission product of concern. Tables 4-2a and 4-2b are derived from Barton (2006) and ORAUT 

(2006a). In this case, natural uranium is used because its lower specific activity gives a more 

favorable to claimant ratio and because reports describe releases of UF6 from building C-410 during 

the fluorination process prior to enrichment (DOE 2000a; PACE 2000).

Summary of Transuranic and Fission Product Exposure

Table 4-2a. Bounding isotopic contributions in PGDP operations (pCi/g U).

Radionuclide

1 2 3 4 5 6

Pulverizer, 

ash 

handling, 

green salt 

C-410 

Pre-1983a

Converter 

Salvage 

Line 

Pre-1983a

Converter 

Salvage 

Line 

Post1983b

Tc/Np 

recovery 

Operations, 

C-400c

Balance 

of plant 

Pre1983c

Balance of 

plant Post1983b

Np-237 3.55E+04 1.61E+06 3.81E+04 1.76E+07 1.67E+03 3.67E+00

Pu-239/240 9.00E+05 3.24E+04 7.09E+04 4.42E+06 4.11E+01 1.95E+00

Pu-238 1.94E+05 7.01E+03 1.53E+04 9.54E+05 8.89E+00 4.21E-01

Pu-242 4.46E+01 1.61E+00 3.52E+00 2.19E+02 2.04E-03 9.67E-05

Pu-241 3.51E+07 1.27E+06 2.76E+06 1.72E+08 1.61E+03 3.38E-01

Am-241 1.56E+05 5.62E+03 1.85E+03 7.65E+05 7.13E+00 4.21E+00

Th-230 1.18E+06 1.67E+05 3.64E+05 2.27E+07 2.11E+02 1.00E+01

Tc-99 1.20E+05 1.20E+05 1.20E+05 3.81E+07 1.20E+05 1.20E+05

U-234 6.81E+05 6.81E+05 6.81E+05 6.81E+05 6.81E+05 6.81E+05

U-235 4.27E+04 4.27E+04 4.27E+04 4.27E+04 4.27E+04 4.27E+04

U-238 3.27E+05 3.27E+05 3.27E+05 3.27E+05 3.27E+05 3.27E+05

U-236 9.00E+02 9.00E+02 9.00E+02 9.00E+02 9.00E+02 9.00E+02

a. PACE and University of Utah (2000)

b. Hightower et al (2000)

c. BJC (2000)
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 15 of 31

Table 4-2b. Bounding isotopic contributions in PGDP operations (pCi/pCi U)a.

Radionuclide

1 2 3 4 5 6

Pulverizer, 

ash 

handling, 

green salt 

C-410 

Pre-1983a

Converter 

Salvage 

Line 

Pre-1983a

Converter 

Salvage 

Line 

Post1983b

Tc/Np 

recovery 

Operations, 

C-400c

Balance 

of plant 

Pre1983c

Balance 

of plant 

Post1983b

Np-237 3.38E-02 1.53E+00 3.63E-02 1.67E+01 1.59E-03 3.49E-06

Pu-239/240 8.56E-01 3.08E-02 6.75E-02 4.21E+00 3.91E-05 1.86E-06

Pu-238 1.85E-01 6.67E-03 1.46E-02 9.08E-01 8.46E-06 4.01E-07

Pu-242 4.24E-05 1.53E-06 3.35E-06 2.08E-04 1.94E-09 9.20E-11

Pu-241 3.34E+01 1.21E+00 2.63E+00 1.64E+02 1.53E-03 3.22E-07

Am-241 1.48E-01 5.35E-03 1.76E-03 7.28E-01 6.79E-06 4.01E-06

Th-230 1.12E+00 1.59E-01 3.46E-01 2.16E+01 2.01E-04 9.52E-06

Tc-99 1.14E-01 1.14E-01 1.14E-01 3.63E+01 1.14E-01 1.14E-01

U-234 4.88E-01 4.88E-01 4.88E-01 4.88E-01 4.88E-01 4.88E-01

U-235 2.17E-02 2.17E-02 2.17E-02 2.17E-02 2.17E-02 2.17E-02

U-238 4.88E-01 4.88E-01 4.88E-01 4.88E-01 4.88E-01 4.88E-01

U-236 1.31E-03 1.31E-03 1.31E-03 1.31E-03 1.31E-03 1.31E-03

a. Table 5-2 converted from pCi/g U (2% enriched) to pCi/pCi U (Natural). Conservative assumption when U basis by weight is 2% 

enriched since the activity is higher, however when U activity basis is used, the lower specific activity of natural Uranium should be used.

Table 4-3 lists the uranium releases, alpha airborne concentrations, and intakes from 1952 to 2001. 

Table 4-4 lists the 99Tc releases, beta airborne concentrations, and intakes for the same period.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 16 of 31

Table 4-3. Uranium releases, alpha airborne concentrations, and intakes.

Year

Uranium 

release 

(Bq/yr)

Outdoor alpha concentrations (Bq/m3) at certain monitoring locations Annual 

uranium 

intake

(Bq/yr)

2 PS 

alpha

5 PW 

alpha

6 PN 

alpha

8 PE 

alpha

Estimate for 

missing 

years

Maximum 

reported

Applied 

concentration

1952 7.4E+08 7.3E-04 7.3E-04 1.7E+00

1953 9.3E+09 9.1E-03 9.1E-03 2.2E+01

1954 8.9E+10 8.7E-02 8.7E-02 2.1E+02

1955 1.6E+11 1.5E-01 1.5E-01 3.7E+02

1956 6.0E+10 5.9E-02 5.9E-02 1.4E+02

1957 4.1E+10 4.0E-02 4.0E-02 9.6E+01

1958 4.0E+10 3.6E-02 4.0E-02 3.6E-02 3.6E-02 4.0E-02 4.0E-02 9.5E+01

1959 3.4E+10 8.0E-03 1.0E-02 1.7E-02 9.2E-03 1.7E-02 1.7E-02 4.0E+01

1960 7.4E+10 6.1E-03 6.2E-03 1.1E-02 7.0E-03 1.1E-02 1.1E-02 2.6E+01

1961 8.9E+10 2.7E-03 2.4E-03 4.8E-03 2.5E-03 4.8E-03 4.8E-03 1.2E+01

1962 4.8E+10 1.9E-03 1.7E-03 3.0E-03 2.4E-03 3.0E-03 3.0E-03 7.1E+00

1963 4.8E+10 1.9E-03 1.7E-03 3.0E-03 2.4E-03 3.0E-03 3.0E-03 7.1E+00

1964 2.2E+10 2.6E-03 2.2E-03 3.7E-03 3.7E-03 3.7E-03 3.7E-03 8.9E+00

1965 7.4E+08 7.4E-04 7.4E-04 1.1E-03 7.4E-04 1.1E-03 1.1E-03 2.7E+00

1966 7.4E+08 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.8E+00

1967 7.4E+08 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.8E+00

1968 1.1E+10 7.4E-04 7.4E-04 1.5E-03 7.4E-04 1.5E-03 1.5E-03 3.6E+00

1969 3.7E+10 7.4E-04 7.4E-04 1.5E-03 1.1E-03 1.5E-03 1.5E-03 3.6E+00

1970 1.9E+10 7.4E-04 7.4E-04 7.4E-04 1.5E-03 1.5E-03 1.5E-03 3.6E+00

1971 2.6E+10 7.4E-04 7.4E-04 1.1E-03 1.5E-03 1.5E-03 1.5E-03 3.6E+00

1972 2.6E+10 7.4E-04 7.4E-04 1.1E-03 1.1E-03 1.1E-03 1.1E-03 2.7E+00

1973 3.0E+10 7.4E-04 8.1E-04 3.9E-03 1.1E-03 3.9E-03 3.9E-03 9.4E+00

1974 2.2E+10 3.3E-04 4.1E-04 1.5E-03 4.1E-04 1.5E-03 1.5E-03 3.5E+00

1975 2.6E+10 6.7E-04 1.0E-03 3.2E-03 8.5E-04 3.2E-03 3.2E-03 7.6E+00

1976 3.3E+10 1.0E-03 1.0E-03 3.2E-03 5.2E-04 3.2E-03 3.2E-03 7.6E+00

1977 1.5E+10 4.1E-04 5.2E-04 2.2E-03 7.4E-04 2.2E-03 2.2E-03 5.3E+00

1978 1.5E+09 3.3E-04 5.2E-04 1.0E-03 1.5E-04 1.0E-03 1.0E-03 2.4E+00

1979 7.4E+08 1.5E-04 1.9E-04 1.0E-03 5.2E-04 1.0E-03 1.0E-03 2.4E+00

1980 3.7E+07 2.2E-04 2.2E-04 7.8E-04 3.3E-04 7.8E-04 7.8E-04 1.9E+00

1981 1.9E+09 1.1E-04 1.9E-04 3.0E-04 3.0E-04 3.0E-04 3.0E-04 7.1E-01

1982 4.8E+09 1.1E-04 1.5E-04 4.1E-04 1.9E-04 4.1E-04 4.1E-04 9.8E-01

1983 1.7E+08 2.2E-04 3.0E-04 2.2E-04 2.6E-04 3.0E-04 3.0E-04 7.1E-01

1984 7.0E+07 3.0E-04 1.9E-04 4.8E-04 3.3E-04 4.8E-04 4.8E-04 1.2E+00

1985 1.4E+08 4.4E-04 2.2E-04 4.1E-04 4.8E-04 4.8E-04 4.8E-04 1.2E+00

1986 1.3E+07 2.6E-04 3.0E-04 3.0E-04 2.6E-04 3.0E-04 3.0E-04 7.1E-01

1987 1.1E+07 1.3E-04 1.3E-04 1.4E-04 1.3E-04 1.4E-04 1.4E-04 3.4E-01

1988 2.2E+06 1.6E-04 1.5E-04 1.5E-04 2.0E-04 2.0E-04 2.0E-04 4.7E-01

1989 1.1E+07 9.3E-05 1.1E-04 1.3E-04 8.9E-05 1.3E-04 1.3E-04 3.2E-01

1990 1.2E+06 1.2E-04 1.2E-04 3.0E-01

1991 2.5E+05 5.2E-05 9.3E-05 7.8E-05 4.8E-05 9.3E-05 9.3E-05 2.2E-01

1992 7.8E+07 5.6E-05 5.9E-05 4.8E-05 5.2E-05 5.9E-05 5.9E-05 1.4E-01

1993 1.2E+08 8.0E-05 8.0E-05 1.9E-01

1994 7.0E-05 1.3E-04 2.0E-04 9.3E-05 2.0E-04 2.0E-04 4.8E-01

1995

1996 1.1E+08

1997

1998

1999

2000

2001
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 17 of 31

Table 4-4. Technetium-99 releases, beta airborne concentrations, and intakes.

Year

Tc-99 

release 

(Bq/yr)

Outdoor beta concentrations (Bq/m3) at certain monitoring stations

Tc-99 intake 

2 PS beta 5 PW beta 6 PN beta 8 PE beta Estimate Maximum Applied (Bq/yr)

1952

1953

1954 2.4E-02 2.4E-02 5.7E+01

1955 3.7E+10 2.4E-02 2.4E-02 5.7E+01

1956 3.7E+10 3.8E-02 3.8E-02 9.1E+01

1957 9.6E+10 2.0E-01 2.0E-01 4.8E+02

1958 9.6E+10 2.0E-01 2.0E-01 4.8E+02

1959 1.8E+11 3.7E-01 3.7E-01 8.9E+02

1960 2.3E+11 4.9E-01 4.9E-01 1.2E+03

1961 1.9E+11 5.1E-02 7.6E-02 1.0E-01 7.1E-02 1.0E-01 1.0E-01 2.4E+02

1962 1.5E+11 4.0E-02 3.2E-02 9.8E-02 8.1E-02 9.8E-02 9.8E-02 2.4E+02

1963 1.6E+11 9.6E-02 9.6E-02 1.6E-01 1.0E-01 1.6E-01 1.6E-01 3.9E+02

1964 1.5E+11 1.8E-01 2.0E-01 3.2E-01 2.6E-01 3.2E-01 3.2E-01 7.6E+02

1965 1.6E+11 1.8E-01 2.0E-01 3.2E-01 2.6E-01 3.2E-01 3.2E-01 7.6E+02

1966 2.0E+11 3.3E-02 3.3E-02 8.5E-02 6.7E-02 8.5E-02 8.5E-02 2.0E+02

1967 1.6E+11 7.4E-03 1.1E-02 1.5E-02 1.1E-02 1.5E-02 1.5E-02 3.6E+01

1968 3.7E+09 5.6E-03 5.9E-03 1.2E-02 7.0E-03 1.2E-02 1.2E-02 2.9E+01

1969 3.7E+09 4.4E-03 5.2E-03 4.8E-03 8.5E-03 8.5E-03 8.5E-03 2.0E+01

1970 3.7E+09 5.6E-03 7.4E-03 3.5E-02 6.7E-03 3.5E-02 3.5E-02 8.4E+01

1971 3.7E+09 1.0E-02 1.9E-02 5.9E-02 3.5E-02 5.9E-02 5.9E-02 1.4E+02

1972 1.2E+11 6.7E-03 1.1E-02 2.4E-02 1.4E-02 2.4E-02 2.4E-02 5.7E+01

1973 1.1E+11 1.5E-02 1.9E-02 1.0E-01 2.6E-02 1.0E-01 1.0E-01 2.5E+02

1974 3.7E+09 1.1E-02 1.9E-02 8.9E-02 3.0E-02 8.9E-02 8.9E-02 2.1E+02

1975 1.3E+11 8.5E-03 1.6E-02 1.5E-01 6.1E-02 1.5E-01 1.5E-01 3.6E+02

1976 2.2E+11 7.6E-03 8.3E-03 1.3E-02 8.1E-03 1.3E-02 1.3E-02 3.2E+01

1977 3.0E+10 6.7E-03 7.8E-03 1.8E-02 7.8E-03 1.8E-02 1.8E-02 4.3E+01

1978 3.7E+09 5.2E-03 5.9E-03 1.4E-02 7.0E-03 1.4E-02 1.4E-02 3.5E+01

1979 3.7E+09 5.2E-03 5.2E-03 1.7E-02 9.6E-03 1.7E-02 1.7E-02 4.1E+01

1980 2.2E+09 7.8E-03 7.4E-03 1.0E-02 9.6E-03 1.0E-02 1.0E-02 2.4E+01

1981 2.2E+09 5.6E-03 5.6E-03 7.0E-03 5.6E-03 1.3E-02 1.3E-02 3.1E+01

1982 2.0E+09 5.9E-03 5.9E-03 9.6E-03 7.4E-03 9.6E-03 9.6E-03 2.3E+01

1983 2.2E+08 1.1E-02 1.3E-02 1.4E-02 1.3E-02 1.4E-02 1.4E-02 3.3E+01

1984 3.7E+08 7.8E-03 7.8E-03 1.1E-02 7.4E-03 1.1E-02 1.1E-02 2.8E+01

1985 1.1E+08 7.8E-03 1.0E-02 1.2E-02 1.3E-02 1.3E-02 1.3E-02 3.0E+01

1986 1.3E+09 1.2E-02 9.6E-03 1.6E-02 1.2E-02 1.6E-02 1.6E-02 3.7E+01

1987 5.7E+08 8.5E-03 7.4E-03 1.1E-02 8.1E-03 1.3E-02 1.3E-02 3.2E+01

1988 3.3E+08 6.7E-03 4.8E-03 6.3E-03 5.2E-03 7.3E-03 7.3E-03 1.7E+01

1989 3.3E+07 3.3E-03 2.9E-03 3.7E-03 3.4E-03 4.4E-03 4.4E-03 1.1E+01

1990 1.4E+08 1.2E-03 1.7E-03 1.4E-03 1.4E-03 1.7E-03 1.7E-03 4.0E+00

1991 1.3E+08 1.2E-03 1.3E-03 1.3E-03 1.1E-03 1.3E-03 1.3E-03 3.2E+00

1992 1.4E+07 1.1E-03 1.1E-03 1.2E-03 1.2E-03 1.2E-03 1.2E-03 2.9E+00

1993 1.1E+08 9.3E-04 9.3E-04 1.0E-03 8.5E-04 1.1E-03 1.1E-03 2.8E+00

1994 7.6E+06 7.4E-04 5.6E-04 7.4E-04 7.0E-04 8.5E-04 8.5E-04 2.0E+00

1995 1.2E+09 3.0E-03 3.0E-03 7.2E+00

1996 1.0E-03 1.2E-03 1.3E-03 1.0E-03 1.3E-03 1.3E-03 3.1E+00

1997 3.0E-03 3.0E-03 7.2E+00

1998 1.5E+09 3.0E-03 3.0E-03 7.2E+00

1999 1.1E+08 3.0E-03 3.0E-03 7.2E+00

2000 3.0E-03 3.0E-03 7.2E+00

2001 3.0E-03 3.0E-03 7.2E+00
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 18 of 31

4.3 EXTERNAL DOSE

Workers received external dose from ambient radiation levels. Until September 1981, external 

gamma radiation levels were measured with a calibrated Geiger-Muller (GM) counter three feet above 

the ground. The limit of sensitivity for the GM tube was 1.0 × 10-4 mR/hr (the basis for energy 

calibration is not known). The frequency of the GM exposure rate surveys varied; sometimes they 

were weekly and sometimes they were monthly. 

From 1981 to the present, thermoluminescent dosimeters (TLDs) have been used to determine 

ambient radiation levels. TLD readings are considered more accurate than GM measurements 

because the GM counter takes more discrete readings and could miss fluctuations in gamma levels. 

TLD locations include those previously monitored by GM counters. TLDs offer greater sensitivity and 

represent an integration of data rather than the discrete results associated with a GM counter. 

Information is not available on the environmental TLD detection limit.

Over the years, various combinations of TLD chip configurations and monitoring locations were used. 

It is clear from security fence data that external exposure rates increased over the life of the site. Two 

factors contributed to this trend: 1) the improved monitoring of the newer TLDs and, more significantly, 

2) the increasing inventory of depleted uranium cylinders. The highest security fence observation was 

194 mrem deep dose equivalent for 2,000 hours. This occurred in 2001, the most recent year during 

which results were from the modern monitoring techniques.

During the early period of PGDP operations, significant quantities of depleted uranium cylinders would

not have been stored at the site. As production continued, the inventory of depleted uranium and 

associated direct radiation levels increased.

Unmonitored workers in the early years did not have significant inventories of depleted uranium to 

contribute to external dose. Later, unmonitored workers would not spend their entire work year at the 

depleted cylinder storage yards and would therefore not reach the maximum dose recorded by fence 

line monitoring. No other significant sources of external exposure are associated with the PGDP 

operations. An assumed deep dose equivalent rate of 200 mrem/yr for all years would be reasonable, 

and deficiencies in earlier measurement techniques thereby become immaterial.

4.3.1

The environmental radiological profile has been developed for PGDP for use by dose reconstructors 

when personal dosimetry or bioassay program participation was not performed or required. Site 

annual environmental reports, health physics surveys, and other reports were reviewed for data that 

would be useful in reconstructing ambient radiation levels. Ambient radiation includes natural 

background and that from the facility. Data in these documents (see the reference list for citations) 

include TLD radiation measurements. Table 4-5 lists TLD results and Figure 4-5 shows the locations 

of TLDs at PGDP.

Ambient Radiation

The ambient radiation measured by TLDs near the security fence included natural background 

radiation, fallout, and cosmic radiation. The TLDs provided an indication of worker exposure levels in 

the general proximity of the security fence, but not inside the buildings. Table 4-5 lists annual 

exposure levels for locations near the security fence. These data have been adjusted to be 

representative of exposure for a 2,000-hour work year. 
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 19 of 31

Table 4-5. External dose equivalent rates, PGDP external gamma exposure rate of mrem/2,000 hr.

TLD 

location 1 2 3 4 5 6 7 8 9 10 11 12 13 25 30 47 48 50 51 52 53

Reported 

background

1961 42 42 42 63

1986 31 25 25 23 14

1989 112 10 9 5 8 6 10 28 114

1990 98 14 16 17 18 15 13 35 15

1991 92 12 14 10 14 11 13 40 19

1992 84 12 14 10 9 11 13 40 58

1993 110 21 20 21 20 18 21 39 20 10

1998 120 106 234 28

1999 101 174 180 26 24

2000 101 179 112 23 24 22 27 18 25 21 23 22 26 29 20 25

2001 120 194 103 26 27 24 29 20 26 23 25 24 28 23 56 35 41 39 31 71 27

Figure 4-5. 2001 TLD locations.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 20 of 31

Observations reported in the early environmental reports, based on surveys with portable instruments, 

reported 0.02 mrem/hr at all locations. For a 2,000-hr work year, this indicates a dose of 40 mrem. 

The instrumentation was less advanced by today’s standards and did not provide continuous 

monitoring, so the results produced are overshadowed by recent, more reliable observations.

PGDP personnel have annually compared these data with TLD data from offsite locations and 

literature values for State and regional exposure levels. The determination has always been that 

onsite ambient radiological conditions as measured at the security fence are not significantly different 

from offsite, State, and regional annual exposure levels. This is attributed to the geology of the region 

around PGDP. Exceptions to this observation have been monitoring locations near depleted uranium

cylinder storage yards in recent years. These locations show an increase in external exposure as the 

inventory of depleted uranium increased. External exposure in this TBD has not had background 

environmental radiation subtracted.

4.3.2

Several fenceline TLDs were adjacent to the UF6 cylinder storage yards and, due to their proximity 

(less than 100 m), represent within a certain amount of confidence the dose rate ranges near the 

storage yards. Workers performed activities in other facilities and probably did not work in the 

cylinder yards 2,000 hr/yr without dosimetry. During recent years, this area has been posted as a 

radiological area, which has reduced the number of unmonitored workers spending any significant 

time in the area to zero.

Radiological Conditions in Cylinder Storage Yards

Missed and unmonitored neutron dose is discussed in the External Dosimetry Section for this site

(ORAUT 2006b.)

Given this information, an ambient annual radiation dose equivalent of 200 mrem should be applied to 

unmonitored workers. 

4.3.3

Because all workers were badged and monitored throughout much of the PGDP operating history, 

coworker exposure data can be used to assign dose to unmonitored workers for those periods in 

which all workers were not badged or were badged but not analyzed. Other sources of information 

that describe potential radiation exposures can be used as backup to the coworker data.

Radiological Conditions Inside Buildings

4.4 UNCERTAINTY

The locations of the monitoring points add uncertainty to the results. The monitoring points, as stated 

above, have been located around the PGDP perimeter and off the site to monitor public exposures. 

Before 1998, external environmental exposures were not monitored in relation to workers; that is, 

monitoring stations were not normally placed at the interior of the site among the process buildings. 

Because of data availability, this analysis used public exposure information for worker environmental 

exposures. The maximum value of environmental exposure is recommended for years when data are 

unavailable to compensate for lack of worker-specific environmental dose information.

All external environmental dose data were adjusted to reflect a 2,000-hour work year. The data were 

originally reported in site environmental reports as representative of an employee who worked at the 

site 24 hr/d, 365 d/yr. Using an employee permanently on the site, however, is an unrealistic 

assumption that would clearly overstate onsite environmental exposures. 
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 21 of 31

Uncertainty related to internal exposures presents similar concerns. The highest internal exposures 

occurred during the early years and lessened over time with increased controls and better equipment. 

Assumption of the maximum uptake for all years reduces the need to include an uncertainty factor 

(other than default values) for intakes.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 22 of 31

REFERENCES

Cited References

Baker, R. C., 1964, “Neptunium-237 Contamination Controls in a Gaseous Diffusion Plant,” draft for 

presentation at Bioassay Conference, Union Carbide Nuclear Company, Paducah Gaseous 

Diffusion Plant, Paducah, Kentucky.

Baker, R. C. 1987, “Exposure Assessment-Uranium Recycle Materials in the Paducah Feed Plant,” 

attachment to letter to J. W. Parks (DOE), Martin Marietta Energy Systems, Paducah Gaseous 

Diffusion Plant, Paducah, Kentucky, October 9.

Barton, C. B., 2006, "Verifying and Expanding Paducah Occupational Internal Dose Document 

Tables," memorandum to D. S. Mantooth, SRDB Ref ID: 24017, Dade Moeller & Associates, 

Richland, Washington, May.

Brown E. G., and K. K. Mitchell, 1960, Environmental Monitoring Summary for the Paducah Plant for 

1959, KY-332, Union Carbide Nuclear Company, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

CCCC (Carbide and Carbon Chemicals Company), 1953, Water, Mud & Air Survey at the Paducah 

Plant, KYL-11 (LB09905-0221), Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

CCCC (Carbide and Carbon Chemicals Company), 1960, The Neptunium Problem at the Paducah 

Plant, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

DOE (U.S. Department of Energy), 2000a, Recycled Uranium Mass Balance Project Paducah

Gaseous Diffusion Plant Report, Rev. 0, BJC-PGDP-167, Paducah, Kentucky.

DOE (U. S. Department of Energy), 2000b, Phase II, Independent Investigation of the Paducah 

Gaseous Diffusion Plant, Environment, Safety and Health Practices, 1952-1990, SRDB Ref ID: 

12040, Office of Oversight, Environment, Safety and Health, Washington, D.C., February. 

Miners, S. E., 1990, Neptunium Experience at PGDP, K/ETO-30, Martin Marietta Energy Systems,

Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

LMES (Lockheed Martin Energy Systems), 1996, Paducah Annual Site Environmental Report for 

1994, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1989, Paducah Gaseous Diffusion Plant Site Environmental 

Report for 1988, ES/ESH-8/V3, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1994, Paducah Site Annual Environmental Report for 1993, 

Martin Marietta Energy Systems, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

ORAUT (Oak Ridge Associated Universities Team), 2004, Technical Basis Document for the Paducah 

Gaseous Diffusion Plant – Site Description, ORAUT-TKBS-0019-2, Revision 00, Oak Ridge, 

Tennessee, May 7.

ORAUT (Oak Ridge Associated Universities Team), 2006a, PGDP –Occupational Internal Dose, 

ORAUT-TKBS-0019-5, Rev. 01, Oak Ridge, Tennessee.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 23 of 31

ORAUT (Oak Ridge Associated Universities Team), 2006b, PGDP –Occupational External Dose, 

ORAUT-TKBS-0019-6, Rev. 02, Oak Ridge, Tennessee.

PACE (Paper, Allied Industrial, Chemical and Energy Workers International Union) and University of 

Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion Plant, SRDB 

Ref ID: 10870, December. 

UCNC (Union Carbide Nuclear Company), 1959, Environmental Monitoring Summary for the Paducah 

Plant for 1958, KY-273, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

UCNC (Union Carbide Nuclear Company), 1961, Environmental Monitoring Summary for the Paducah 

Plant for 1960, KY-371, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

UCNC (Union Carbide Nuclear Company), 1962a, “Air Monitoring Procedure,” PGDP EM&EF 

Records Transmittal List, 9/1/62 (H-00013-0575), Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

UCNC (Union Carbide Nuclear Company), 1962b, Continuous Outdoor Air Sampler, Paducah 

Gaseous Diffusion Plant, Paducah, Kentucky.

Additional Environmental Reports

Baker, R. C., 1964, Environmental Monitoring Summary for the Paducah Plant for 1962 and 1963, 

KY-458, Union Carbide Nuclear Company, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

Baker, R. C., and A. W. Russell, 1965, Environmental Monitoring Summary for the Paducah Plant for 

1964, KY-484, Union Carbide Corporation, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

Baker, R. C., and K. A. Davis, 1967, Environmental Monitoring Summary for the Paducah Plant for 

1965 & 1966, KY-543, Union Carbide Corporation, Paducah Gaseous Diffusion Plant, 

Paducah, Kentucky.

BJC (Bechtel Jacobs Company), 1998, Paducah Site 1997 Annual Environmental Report, Paducah 

Gaseous Diffusion Plant, Paducah, Kentucky.

BJC (Bechtel Jacobs Company), 1999, Paducah Site 1998 Annual Environmental Report, Paducah 

Gaseous Diffusion Plant, Paducah, Kentucky.

BJC (Bechtel Jacobs Company), 2001, Paducah Site Annual Site Environmental Report for 1999, 

Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

Brown, E. G., and R. C. Baker, 1962, Environmental Monitoring Summary for the Paducah Plant for 

1961, KY-419, Union Carbide Nuclear Company, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

CDM Federal Services, 2001, Paducah Site Annual Site Environmental Report for Calendar Year 

2000, BJC/PAD-237, Paducah Gaseous Diffusion Plant, Kevil, Kentucky, October.

Davis, K. A., 1971, Environmental Monitoring Summary for the Paducah Plant for 1970, KY-629, 

Union Carbide Corporation, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 24 of 31

Davis, K. A., and E. G. Brown, 1969, Environmental Monitoring Summary for the Paducah Plant for 

1967 and 1968, KY-582, Union Carbide Corporation, Paducah Gaseous Diffusion Plant, 

Paducah, Kentucky.

LMES (Lockheed Martin Energy Systems), 1997a, Paducah Site Annual Report for 1995, KY/EM-176, 

Kevil, Kentucky.

LMES (Lockheed Martin Energy Systems), 1997b, Paducah Site 1996 Annual Environmental Report, 

Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1984, Environmental Monitoring Report, Paducah Gaseous 

Diffusion Plant, Calendar Year 1983, KY-742, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

MMES (Martin Marietta Energy Systems), 1985, Environmental Monitoring Report, United States 

Department of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1984, KY-747, 

Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1986, Environmental Monitoring Report, United States 

Department of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1985, KY-755, 

Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1990, Paducah Gaseous Diffusion Plant Environmental 

Report for 1989, ES/ESH-13/V3, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1991, Paducah Environmental Report for 1990, ES/ESH18/V3, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1992, Paducah Gaseous Diffusion Plant Environmental 

Report for 1991, ES/ESH-22/V3, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

MMES (Martin Marietta Energy Systems), 1993, Paducah Gaseous Diffusion Plant Environmental 

Report for 1992, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

Oakes, T. W., D. L. Ashburn, and F. M. O'Hara, 1987, Environmental Surveillance of the U.S. 

Department of Energy Paducah Gaseous Diffusion Plant and Surrounding Environs During 

1986, ES/ESH-1/V3, Martin Marietta Energy Systems, Paducah Gaseous Diffusion Plant, 

Paducah, Kentucky, April.

Rogers, J. G., and D. L. Ashburn, 1988, Environmental Surveillance of the U.S. Department of Energy 

Paducah Gaseous Diffusion Plant and Surrounding Environs During 1987, ES/ESH-4/V3, 

Martin Marietta Energy Systems, Paducah Gaseous Diffusion Plant, Paducah, Kentucky, April.

UCC (Union Carbide Company), 1980, Environmental Monitoring Report, United States Department 

of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1979, Y/UB-14, Paducah 

Gaseous Diffusion Plant, Paducah, Kentucky.

UCC (Union Carbide Company), 1981, Environmental Monitoring Report, United States Department 

of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1980, KY-717, Paducah 

Gaseous Diffusion Plant, Paducah, Kentucky.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 25 of 31

UCC (Union Carbide Company), 1982, Environmental Monitoring Report, United States Department 

of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1981, KY-724, Paducah 

Gaseous Diffusion Plant, Paducah, Kentucky.

UCC (Union Carbide Company), 1983, Environmental Monitoring Report, United States Department 

of Energy, Paducah Gaseous Diffusion Plant for Calendar Year 1982, KY-731, Paducah 

Gaseous Diffusion Plant, Paducah, Kentucky.

UCC (Union Carbide Corporation), 1970, Environmental Monitoring Summary for the Paducah Plant 

for 1969, KY-624, Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

UCC (Union Carbide Corporation), 1972, Environmental Monitoring Report, United States Atomic 

Energy Commission, Paducah Gaseous Diffusion Plant, Calendar Year 1971, UCC-ND-222, 

Paducah Gaseous Diffusion Plant, Paducah, Kentucky.

UCC (Union Carbide Corporation), 1973, Environmental Monitoring Report, United States Atomic 

Energy Commission, Paducah Gaseous Diffusion Plant, Calendar Year 1972, UCC-ND-245,

Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1974, Environmental Monitoring Report, United States Atomic 

Energy Commission, Paducah Gaseous Diffusion Plant, Calendar Year 1973, UCC-ND-279, 

Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1975, Environmental Monitoring Report, United States Energy 

Research and Development Administration, Paducah Gaseous Diffusion Plant, Calendar Year 

1974, UCC-ND-303, Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1976, Environmental Monitoring Report, United States Energy 

Research and Development Administration, Paducah Gaseous Diffusion Plant, Calendar Year 

1975, Y/UB-5, Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1977, Environmental Monitoring Report, United States Department 

of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1976, Y/UB-7, Oak Ridge, 

Tennessee.

UCC (Union Carbide Corporation), 1978, Environmental Monitoring Report, United States Department 

of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1977, Y/UB-9, Oak Ridge, 

Tennessee.

UCC (Union Carbide Corporation), 1979, Environmental Monitoring Report, United States Department 

of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1978, Y/UB-11, Oak Ridge, 

Tennessee.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 26 of 31

GLOSSARY

alpha radiation

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

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

charge of +2. Alpha particles have low penetrating power and a short range (a few 

centimeters in air). Outside the body, the most energetic alpha particle generally fails to 

penetrate the dead layers of cells covering the skin or a sheet of paper. Alpha particles 

emitted by radionuclides inside the body are a more significant health risk. 

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

becquerel (Bq)

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

37 billion (3.7 × 1010) Bq. 

beta radiation

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

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

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

deep dose equivalent

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose.

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays. Various terms narrow the type of dose, 

and some are additive: 

• Absorbed dose is the amount of energy deposited in a substance by ionizing radiation.

• Collective dose is the sum of the doses to a specific population.

• Committed dose is the dose over time (usually 50 years for workers) to a specific organ 

or tissue from an intake of radioactive material.

• Cumulative dose is the sum of all doses to the same portion of the body or to the whole 

body over time.

• Deep dose is the dose at a 1-centimeter depth in tissue (1,000 milligrams per square 

centimeter).
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 27 of 31

• Effective dose is the sum of the equivalent doses in the principal tissues and organs of 

the body, each weighted by a tissue weighting factor that accounts for the probabilities 

of fatal and nonfatal cancers according to severity and the average length of life lost 

due to an induced cancer. It indicates the biological effect of the radiation exposure in 

that tissue. 

• Equivalent dose or dose equivalent is the absorbed dose in a tissue or organ multiplied 

by a weighting factor for the particular type of radiation. 

• Organ dose is the dose to a specific organ.

• Penetrating dose is that from higher energy photon (gamma and X-ray) radiation and 

neutron radiation that penetrates the outer layers of the skin. Nonpenetrating dose is 

that from beta and lower energy photon radiation. 

• Personal dose equivalent is the dose equivalent in soft tissue below a specified point 

on the body at a specified depth. 

• Shallow dose is the dose at a 0.07-centimeter depth in tissue (7 milligrams per square 

centimeter).

• Skin dose is the dose to the skin.

• Whole-body dose is the dose to the entire body excluding the contents of the 

gastrointestinal tract, urinary bladder, and gall bladder.

dosimetry

Measurement and calculation of internal and external radiation doses.

exposure

1) In general, the act of being exposed to ionizing radiation. 2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

radioactivity

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural.

thermoluminescent dosimeter (TLD)

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

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

light provides a measurement of absorbed dose. 
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 28 of 31

uranium hexafluoride (UF6) cylinder storage yard

Site for maintenance of cylinders containing depleted UF6. The cylinders typically weigh 10 

and 14 tons. The depleted UF6 is primarily in a solid form. There are storage yards at PGDP 

and the Oak Ridge K-25 site.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 29 of 31

ATTACHMENT 4A

OUTDOOR BETA AND ALPHA CONCENTRATIONS

Page 1 of 3

Table 4A-1 lists outdoor beta concentrations, and Table 4A-2 lists outdoor alpha concentrations.
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 30 of 31

ATTACHMENT 4A

OUTDOOR BETA AND ALPHA CONCENTRATIONS

Page 2 of 3

Table 4A-1. Outdoor beta concentrations.

Year

Release (Bq/yr)

Tc-99

Outdoor beta concentrations (Bq/m3) at each monitoring location Bq/yr

2 PS beta 5 PW beta 6 PN beta 8 PE beta BN BE IN IE ISE IS IW GR Estimate Maximum Applied Uptake

1952

1953

1954 2.4E-02 2.4E-02 5.7E+01

1955 3.7E+10 2.4E-02 2.4E-02 5.7E+01

1956 3.7E+10 3.8E-02 3.8E-02 9.1E+01

1957 9.6E+10 2.0E-01 2.0E-01 4.8E+02

1958 9.6E+10 2.0E-01 2.0E-01 4.8E+02

1959 1.8E+11 3.7E-01 3.7E-01 8.9E+02

1960 2.3E+11 4.9E-01 4.9E-01 1.2E+03

1961 1.9E+11 5.1E-02 7.6E-02 1.0E-01 7.1E-02 1.0E-01 1.0E-01 2.4E+02

1962 1.5E+11 4.0E-02 3.2E-02 9.8E-02 8.1E-02 9.8E-02 9.8E-02 2.4E+02

1963 1.6E+11 9.6E-02 9.6E-02 1.6E-01 1.0E-01 8.1E-02 8.5E-02 1.2E-01 8.9E-02 1.6E-01 1.6E-01 3.9E+02

1964 1.5E+11 1.8E-01 2.0E-01 3.2E-01 2.6E-01 2.0E-01 1.8E-01 1.9E-01 1.8E-01 3.2E-01 3.2E-01 7.6E+02

1965 1.6E+11 1.8E-01 2.0E-01 3.2E-01 2.6E-01 2.0E-01 1.8E-01 1.9E-01 1.8E-01 3.2E-01 3.2E-01 7.6E+02

1966 2.0E+11 3.3E-02 3.3E-02 8.5E-02 6.7E-02 3.3E-02 3.3E-02 3.3E-02 3.3E-02 8.5E-02 8.5E-02 2.0E+02

1967 1.6E+11 7.4E-03 1.1E-02 1.5E-02 1.1E-02 1.1E-02 1.1E-02 1.1E-02 1.1E-02 1.5E-02 1.5E-02 3.6E+01

1968 3.7E+09 5.6E-03 5.9E-03 1.2E-02 7.0E-03 6.3E-03 6.7E-03 7.0E-03 6.3E-03 1.2E-02 1.2E-02 2.9E+01

1969 3.7E+09 4.4E-03 5.2E-03 4.8E-03 8.5E-03 5.6E-03 5.2E-03 6.7E-03 6.7E-03 3.3E-03 8.5E-03 8.5E-03 2.0E+01

1970 3.7E+09 5.6E-03 7.4E-03 3.5E-02 6.7E-03 7.8E-03 6.7E-03 6.7E-03 6.7E-03 7.4E-03 3.5E-02 3.5E-02 8.4E+01

1971 3.7E+09 1.0E-02 1.9E-02 5.9E-02 3.5E-02 8.9E-03 6.7E-03 8.5E-03 8.5E-03 7.4E-03 5.9E-02 5.9E-02 1.4E+02

1972 1.2E+11 6.7E-03 1.1E-02 2.4E-02 1.4E-02 8.5E-03 1.7E-02 7.4E-03 7.4E-03 7.4E-03 2.4E-02 2.4E-02 5.7E+01

1973 1.1E+11 1.5E-02 1.9E-02 1.0E-01 2.6E-02 1.5E-02 1.1E-02 1.1E-02 1.1E-02 7.4E-03 1.0E-01 1.0E-01 2.5E+02

1974 3.7E+09 1.1E-02 1.9E-02 8.9E-02 3.0E-02 1.1E-02 7.4E-03 7.4E-03 7.4E-03 7.4E-03 8.9E-02 8.9E-02 2.1E+02

1975 1.3E+11 8.5E-03 1.6E-02 1.5E-01 6.1E-02 1.6E-02 9.3E-03 5.6E-03 4.8E-03 3.7E-03 4.8E-03 1.5E-01 1.5E-01 3.6E+02

1976 2.2E+11 7.6E-03 8.3E-03 1.3E-02 8.1E-03 8.9E-03 7.6E-03 7.2E-03 7.9E-03 7.4E-03 7.4E-03 1.3E-02 1.3E-02 3.2E+01

1977 3.0E+10 6.7E-03 7.8E-03 1.8E-02 7.8E-03 8.1E-03 7.8E-03 7.4E-03 8.1E-03 6.7E-03 6.7E-03 1.8E-02 1.8E-02 4.3E+01

1978 3.7E+09 5.2E-03 5.9E-03 1.4E-02 7.0E-03 7.0E-03 5.6E-03 6.3E-03 5.9E-03 5.6E-03 5.6E-03 1.4E-02 1.4E-02 3.5E+01

1979 3.7E+09 5.2E-03 5.2E-03 1.7E-02 9.6E-03 7.4E-03 4.8E-03 5.9E-03 4.8E-03 5.2E-03 5.2E-03 1.7E-02 1.7E-02 4.1E+01

1980 2.2E+09 7.8E-03 7.4E-03 1.0E-02 9.6E-03 8.9E-03 8.1E-03 9.6E-03 7.8E-03 7.8E-03 7.0E-03 1.0E-02 1.0E-02 2.4E+01

1981 2.2E+09 5.6E-03 5.6E-03 7.0E-03 5.6E-03 5.9E-03 5.6E-03 5.2E-03 5.2E-03 5.9E-03 5.6E-03 5.6E-03 1.3E-02 1.3E-02 1.3E-02 3.1E+01

1982 2.0E+09 5.9E-03 5.9E-03 9.6E-03 7.4E-03 6.3E-03 7.0E-03 5.6E-03 5.6E-03 5.2E-03 6.3E-03 4.8E-03 5.2E-03 9.6E-03 9.6E-03 2.3E+01

1983 2.2E+08 1.1E-02 1.3E-02 1.4E-02 1.3E-02 1.2E-02 1.3E-02 1.1E-02 9.6E-03 1.0E-02 1.2E-02 8.1E-03 1.3E-02 1.4E-02 1.4E-02 3.3E+01

1984 3.7E+08 7.8E-03 7.8E-03 1.1E-02 7.4E-03 7.8E-03 6.7E-03 7.8E-03 8.5E-03 6.3E-03 7.4E-03 6.7E-03 8.5E-03 1.1E-02 1.1E-02 2.8E+01

1985 1.1E+08 7.8E-03 1.0E-02 1.2E-02 1.3E-02 1.0E-02 1.1E-02 7.8E-03 6.7E-03 1.0E-02 1.0E-02 8.5E-03 8.5E-03 1.3E-02 1.3E-02 3.0E+01

1986 1.3E+09 1.2E-02 9.6E-03 1.6E-02 1.2E-02 1.2E-02 1.1E-02 1.1E-02 1.3E-02 7.4E-03 8.5E-03 1.1E-02 1.3E-02 1.6E-02 1.6E-02 3.7E+01

1987 5.7E+08 8.5E-03 7.4E-03 1.1E-02 8.1E-03 1.0E-02 9.3E-03 9.6E-03 5.9E-03 1.3E-02 8.9E-03 1.3E-02 1.0E-02 1.3E-02 1.3E-02 3.2E+01

1988 3.3E+08 6.7E-03 4.8E-03 6.3E-03 5.2E-03 5.3E-03 7.1E-03 7.3E-03 5.7E-03 6.1E-03 4.6E-03 6.4E-03 6.4E-03 7.3E-03 7.3E-03 1.7E+01

1989 3.3E+07 3.3E-03 2.9E-03 3.7E-03 3.4E-03 3.4E-03 3.6E-03 4.1E-03 4.1E-03 4.4E-03 3.3E-03 3.5E-03 3.7E-03 4.4E-03 4.4E-03 1.1E+01

1990 1.4E+08 1.2E-03 1.7E-03 1.4E-03 1.4E-03 1.1E-03 1.3E-03 1.5E-03 1.3E-03 1.3E-03 1.5E-03 1.5E-03 1.4E-03 1.7E-03 1.7E-03 4.0E+00

1991 1.3E+08 1.2E-03 1.3E-03 1.3E-03 1.1E-03 1.3E-03 1.3E-03 1.2E-03 1.1E-03 1.1E-03 1.2E-03 1.1E-03 1.1E-03 1.3E-03 1.3E-03 3.2E+00

1992 1.4E+07 1.1E-03 1.1E-03 1.2E-03 1.2E-03 1.1E-03 1.1E-03 1.1E-03 1.2E-03 1.1E-03 1.1E-03 1.1E-03 1.2E-03 1.2E-03 1.2E-03 2.9E+00

1993 1.1E+08 9.3E-04 9.3E-04 1.0E-03 8.5E-04 8.1E-04 8.5E-04 1.1E-03 7.8E-04 9.3E-04 9.3E-04 1.0E-03 7.8E-04 1.1E-03 1.1E-03 2.8E+00

1994 7.6E+06 7.4E-04 5.6E-04 7.4E-04 7.0E-04 8.5E-04 6.7E-04 8.1E-04 8.5E-04 8.1E-04 8.5E-04 8.1E-04 7.8E-04 8.5E-04 8.5E-04 2.0E+00

1995 1.2E+09 3.0E-03 3.0E-03 7.2E+00

1996 1.0E-03 1.2E-03 1.3E-03 1.0E-03 1.1E-03 1.2E-03 1.2E-03 1.1E-03 1.2E-03 1.1E-03 1.3E-03 1.3E-03 1.3E-03 3.1E+00

1997 3.0E-03 3.0E-03 7.2E+00

1998 1.5E+09 3.0E-03 3.0E-03 7.2E+00

1999 1.1E+08 3.0E-03 3.0E-03 7.2E+00

2000 3.0E-03 3.0E-03 7.2E+00

2001 3.0E-03 3.0E-03 7.2E+00
Document No. ORAUT-TKBS-0019-4 Revision No. 01 Effective Date: 11/07/2006 Page 31 of 31

ATTACHMENT 4A

OUTDOOR BETA AND ALPHA CONCENTRATIONS

Page 3 of 3

Table 4A-2. Outdoor alpha concentrations.

Year

Release 

(Bq/yr)

U

Outdoor alpha concentrations (Bq/m3) at each monitoring location (Bq/yr)

2 PS 

alpha

5 PW 

alpha

6 PN 

alpha

8 PE 

alpha BN BE IN IE ISE IS IW GR

Estimate 

for missing 

years

Maximum 

reported

Applied 

concentration

Annual 

uranium 

intake

1952 7.4E+08 7.3E-04 7.3E-04 1.7E+00

1953 9.3E+09 9.1E-03 9.1E-03 2.2E+01

1954 8.9E+10 8.7E-02 8.7E-02 2.1E+02

1955 1.6E+11 1.5E-01 1.5E-01 3.7E+02

1956 6.0E+10 5.9E-02 5.9E-02 1.4E+02

1957 4.1E+10 4.0E-02 4.0E-02 9.6E+01

1958 4.0E+10 3.6E-02 4.0E-02 3.6E-02 3.6E-02 4.0E-02 4.0E-02 9.5E+01

1959 3.4E+10 8.0E-03 1.0E-02 1.7E-02 9.2E-03 1.7E-02 1.7E-02 4.0E+01

1960 7.4E+10 6.1E-03 6.2E-03 1.1E-02 7.0E-03 1.1E-02 1.1E-02 2.6E+01

1961 8.9E+10 2.7E-03 2.4E-03 4.8E-03 2.5E-03 4.8E-03 4.8E-03 1.2E+01

1962 4.8E+10 1.9E-03 1.7E-03 3.0E-03 2.4E-03 3.0E-03 3.0E-03 7.1E+00

1963 4.8E+10 1.9E-03 1.7E-03 3.0E-03 2.4E-03 2.7E-03 2.6E-03 2.7E-03 2.7E-03 3.0E-03 3.0E-03 7.1E+00

1964 2.2E+10 2.6E-03 2.2E-03 3.7E-03 3.7E-03 1.7E-03 1.7E-03 1.7E-03 1.4E-03 1.4E-03 3.7E-03 3.7E-03 8.9E+00

1965 7.4E+08 7.4E-04 7.4E-04 1.1E-03 7.4E-04 1.7E-03 1.7E-03 1.7E-03 1.4E-03 1.4E-03 1.1E-03 1.1E-03 2.7E+00

1966 7.4E+08 7.4E-04 7.4E-04 7.4E-04 7.4E-04 2.0E-03 2.5E-03 2.8E-03 1.9E-03 2.8E-03 7.4E-04 7.4E-04 1.8E+00

1967 7.4E+08 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.8E+00

1968 1.1E+10 7.4E-04 7.4E-04 1.5E-03 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.5E-03 1.5E-03 3.6E+00

1969 3.7E+10 7.4E-04 7.4E-04 1.5E-03 1.1E-03 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.5E-03 1.5E-03 3.6E+00

1970 1.9E+10 7.4E-04 7.4E-04 7.4E-04 1.5E-03 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.5E-03 1.5E-03 3.6E+00

1971 2.6E+10 7.4E-04 7.4E-04 1.1E-03 1.5E-03 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.5E-03 1.5E-03 3.6E+00

1972 2.6E+10 7.4E-04 7.4E-04 1.1E-03 1.1E-03 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.1E-03 1.1E-03 2.7E+00

1973 3.0E+10 7.4E-04 8.1E-04 3.9E-03 1.1E-03 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 3.9E-03 3.9E-03 9.4E+00

1974 2.2E+10 3.3E-04 4.1E-04 1.5E-03 4.1E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 7.4E-04 1.5E-03 1.5E-03 3.5E+00

1975 2.6E+10 6.7E-04 1.0E-03 3.2E-03 8.5E-04 1.8E-03 7.4E-04 7.4E-04 7.4E-04 3.3E-04 3.3E-04 3.2E-03 3.2E-03 7.6E+00

1976 3.3E+10 1.0E-03 1.0E-03 3.2E-03 5.2E-04 4.1E-04 3.3E-04 3.3E-04 3.3E-04 3.3E-04 3.3E-04 3.2E-03 3.2E-03 7.6E+00

1977 1.5E+10 4.1E-04 5.2E-04 2.2E-03 7.4E-04 1.0E-03 6.7E-04 1.0E-03 1.0E-03 6.7E-04 6.7E-04 2.2E-03 2.2E-03 5.3E+00

1978 1.5E+09 3.3E-04 5.2E-04 1.0E-03 1.5E-04 8.1E-04 1.2E-03 1.3E-03 1.5E-03 1.2E-03 1.0E-03 1.0E-03 1.0E-03 2.4E+00

1979 7.4E+08 1.5E-04 1.9E-04 1.0E-03 5.2E-04 7.4E-04 3.0E-04 3.7E-04 2.6E-04 3.7E-04 3.3E-04 1.0E-03 1.0E-03 2.4E+00

1980 3.7E+07 2.2E-04 2.2E-04 7.8E-04 3.3E-04 5.2E-04 3.3E-04 3.3E-04 3.3E-04 3.3E-04 3.3E-04 7.8E-04 7.8E-04 1.9E+00

1981 1.9E+09 1.1E-04 1.9E-04 3.0E-04 3.0E-04 3.3E-04 3.3E-04 1.5E-04 1.5E-04 1.5E-04 1.5E-04 3.3E-04 1.5E-04 3.0E-04 3.0E-04 7.1E-01

1982 4.8E+09 1.1E-04 1.5E-04 4.1E-04 1.9E-04 2.2E-04 2.2E-04 2.2E-04 2.2E-04 2.2E-04 2.2E-04 2.2E-04 2.2E-04 4.1E-04 4.1E-04 9.8E-01

1983 1.7E+08 2.2E-04 3.0E-04 2.2E-04 2.6E-04 1.5E-04 1.5E-04 1.5E-04 1.1E-04 1.1E-04 1.9E-04 1.1E-04 1.5E-04 3.0E-04 3.0E-04 7.1E-01

1984 7.0E+07 3.0E-04 1.9E-04 4.8E-04 3.3E-04 1.9E-04 1.9E-04 1.9E-04 1.1E-04 1.1E-04 1.5E-04 1.1E-04 1.5E-04 4.8E-04 4.8E-04 1.2E+00

1985 1.4E+08 4.4E-04 2.2E-04 4.1E-04 4.8E-04 2.2E-04 1.5E-04 1.9E-04 4.1E-04 1.5E-04 1.5E-04 1.9E-04 1.9E-04 4.8E-04 4.8E-04 1.2E+00

1986 1.3E+07 2.6E-04 3.0E-04 3.0E-04 2.6E-04 2.6E-04 2.6E-04 2.6E-04 3.3E-04 1.5E-04 1.5E-04 2.2E-04 2.2E-04 3.0E-04 3.0E-04 7.1E-01

1987 1.1E+07 1.3E-04 1.3E-04 1.4E-04 1.3E-04 3.3E-04 2.6E-04 3.0E-04 1.9E-04 7.4E-05 1.9E-04 3.3E-04 3.7E-04 1.4E-04 1.4E-04 3.4E-01

1988 2.2E+06 1.6E-04 1.5E-04 1.5E-04 2.0E-04 3.3E-04 3.0E-04 3.0E-04 3.0E-04 2.6E-04 2.6E-04 2.6E-04 2.6E-04 2.0E-04 2.0E-04 4.7E-01

1989 1.1E+07 9.3E-05 1.1E-04 1.3E-04 8.9E-05 1.1E-04 1.7E-04 1.7E-04 1.3E-04 2.1E-04 1.0E-04 1.1E-04 1.3E-04 1.3E-04 1.3E-04 3.2E-01

1990 1.2E+06 1.1E-04 1.1E-04 2.3E-04 1.3E-04 1.6E-04 1.6E-04 1.4E-04 1.6E-04 1.2E-04 1.2E-04 3.0E-01

1991 2.5E+05 5.2E-05 9.3E-05 7.8E-05 4.8E-05 1.1E-04 1.0E-04 8.1E-05 1.2E-04 7.8E-05 1.2E-04 8.9E-05 1.2E-04 9.3E-05 9.3E-05 2.2E-01

1992 7.8E+07 5.6E-05 5.9E-05 4.8E-05 5.2E-05 9.6E-05 1.1E-04 1.2E-04 9.6E-05 1.2E-04 1.2E-04 1.1E-04 9.3E-05 5.9E-05 5.9E-05 1.4E-01

1993 1.2E+08 7.8E-05 4.1E-05 1.1E-04 4.8E-05 8.1E-05 4.8E-05 7.4E-05 4.8E-05 8.0E-05 8.0E-05 1.9E-01

1994 7.0E-05 1.3E-04 2.0E-04 9.3E-05 5.9E-05 4.4E-05 9.6E-05 5.6E-05 4.8E-05 4.8E-05 5.6E-05 7.4E-05 2.0E-04 2.0E-04 4.8E-01

1995

1996 1.1E+08 8.0E-05 9.6E-05 9.9E-05 8.5E-05 1.0E-04 7.7E-05 1.1E-04 1.0E-04

1997

1998

1999

2000

2001