ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 36

DOE Review Release 08/31/2012

Document Title:

Paducah Gaseous Diffusion Plant –

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0019-4

Revision: 03

Effective Date: 08/24/2012

Type of Document: TBD

Supersedes: Revision 02

Subject Expert(s): Jodie L. Phillips

Approval: Signature on File Approval Date: 08/21/2012

Jodie L. Phillips, Document Owner

Concurrence: Signature on File Concurrence Date: 08/21/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 08/21/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 08/21/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/24/2012

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. 03 Effective Date: 08/24/2012 Page 2 of 36

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.

04/23/2007 02 Approved Revision 02 initiated to incorporate Attributions and 

Annotations section. Rearranged tables and corrected the List of 

Tables. Constitutes a total rewrite of the document. Incorporates 

formal internal review comments. This revision results in no change 

to the assigned dose and no PER is required. Training required: As 

determined by the Task Manager. Initiated by Daniel S. Mantooth.

08/24/2012 03 This site profile was predominantly revised to address SC&A 

concerns about the document as identified in SC&A-TR-TASK1-

0016, Review of the NIOSH Site Profile for the Paducah Gaseous 

Diffusion Plant. Wording was revised throughout the document to 

add clarity. Figure 4-5 was added to show onsite air monitoring 

locations. The external dose equivalent rates in Table 4-6 have been 

revised based on available historical data. Guidance provided in 

ORAUT-PROC-0060 has been added. Acronyms and Abbreviations, 

References, and Glossary Sections have been updated. 

Incorporates formal internal and NIOSH review comments. 

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

determined by the Objective Manager. Initiated by Jodie L. Phillips.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 3 of 36

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

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

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

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations...................................... 8

4.2.1 Preoperational Background Survey ..............................................................................8

4.2.2 Ambient Air Sample Collection Network .......................................................................8

4.2.3 Methodology...............................................................................................................10

4.2.4 Estimation of Ambient Airborne Radionuclide Concentrations .................................... 11

4.2.5 Data for Ambient Airborne Radionuclide Concentrations and Annual 

Airborne Releases......................................................................................................11

4.2.6 Estimation of Annual Intake from Airborne Radionuclides........................................... 11

4.2.7 Summary of Transuranic and Fission Product Exposure ............................................ 17

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

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

4.3.2 Radiological Conditions in Cylinder Storage Yards .....................................................21

4.3.3 Radiological Conditions Inside Buildings ....................................................................22

4.4 Uncertainty.............................................................................................................................22

4.5 Attributions and Annotations .................................................................................................. 22

References ........................................................................................................................................25

Glossary ............................................................................................................................................30

LIST OF TABLES

TABLE TITLE PAGE

4-1 Ambient air sampling stations ..................................................................................................9

4-2 Uranium releases, alpha airborne concentrations, and intakes .............................................. 15

4-3 Technetium-99 releases, beta airborne concentrations, and intakes ...................................... 16

4-4 Barton (2006) bounding isotopic contributions in PGDP operations ....................................... 17

4-5 ORAUT (2012a) bounding isotopic contributions in PGDP operations.................................... 17

4-6 External dose equivalent rates, PGDP external gamma exposure rate of 

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

A-1 Outdoor beta concentrations .................................................................................................. 35

A-2 Outdoor alpha concentrations ................................................................................................ 36
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 4 of 36

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Wind rose from MMES........................................................................................................... 10

4-2 Alpha air concentrations......................................................................................................... 12

4-3 Beta air concentrations .......................................................................................................... 12

4-4 Cascade feed in thousands of tons of uranium, 1952 to 1978 ................................................ 14

4-5 Air sampling monitoring stations at PGDP.............................................................................. 20

4-6 TLD locations.........................................................................................................................21
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 5 of 36

ACRONYMS AND ABBREVIATIONS

Bq becquerel

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

EEOICPA Energy Employees Occupational Illness Compensation Program Act

FP fission product

ft foot

g gram

GM Geiger-Müller

hr hour

kdpm kilodisintegrations per minute

km kilometer

m meter

mg milligram

mR milliroentgen

mrem millirem

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

PGDP Paducah Gaseous Diffusion Plant

RU recycled uranium

s second

TBD technical basis document

TLD thermoluminescent dosimeter

TRU transuranic

UF4 uranium tetrafluoride

UF6 uranium hexafluoride

UO2 uranium oxide

UO2F2 uranyl fluoride

UO3 uranium trioxide

U.S.C. United States Code

yr year

§ section
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 6 of 36

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, 42 C.F.R. Pt. 82) restrict the “performance of duty” 

referred to in 42 U S. C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

The statute also 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 excludes Naval Nuclear Propulsion Facilities from being covered under the Act, 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 occupational radiation exposures are considered valid for inclusion in a 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 (NIOSH 2010):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 7 of 36

4.1.1 Purpose

This TBD provides technical data and other key information that serve as the technical basis for 

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

Gaseous Diffusion Plant (PGDP).

4.1.2 Scope 

This TBD 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% 235U (natural enrichment) to around 3% 235U

for use in domestic and foreign power reactors (BJC 2000). Enrichment operations began in 1952 in 

the first two process buildings C-331 and C-333. 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 plant ceased operation and all feed to the enrichment process was in the form of UF6 from 

outside sources. Other chemical compounds of uranium were present throughout the Plant’s history,

including uranyl fluoride (UO2F2), uranium tetrafluoride (UF4), and uranium oxide (UO2).

ORAUT (2012c) contains detailed information on the history of PGDP and the feed conversion and 

enrichment process.

Sections 4.2 and 4.3 discuss internal and external environmental dose, respectively. Section 4.4 

discusses uncertainty. Attributions and annotations, indicated by bracketed callouts and used to 

identify the source, justification, or clarification of the associated information, are presented in 

Section 4.5.

This document provides supporting technical data in the evaluation of ambient external and internal 

doses from environmental releases or fixed source terms for workers at the PGDP. PGDP is one of 

the original sites that was designated by Congress as part of the Special Exposure Cohort (SEC) 

under EEOICPA [42 U.S.C. § 7384l(14)]. This designation is as follows:

(A) The employee was so employed for a number of work days aggregating at least 

250 work days before February 1, 1992, at a gaseous diffusion plant located in 

Paducah, Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and, during such 

employment—

(i) was monitored through the use of dosimetry badges for exposure at the plant of 

the external parts of employee’s body to radiation; or

(ii) worked in a job that had exposures comparable to a job that is or was monitored 

through the use of dosimetry badges.

Dose reconstruction guidance in this document is presented to provide a technical basis for dose 

reconstructions for nonpresumptive cancers that are not covered in the SEC class through January 

31, 1992. Dose reconstructions for individuals employed at PGDP before February 1, 1992, but who 

do not qualify for inclusion in the SEC, can be performed using this guidance as appropriate.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 8 of 36

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

4.2.1 Preoperational Background Survey

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

document preoperational conditions at PGDP (Whiteaker and Farrar 1953). Ten chemically 

processed samples showed 0.00 mg/m3 of 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) [1].

4.2.2 Ambient Air Sample Collection Network

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

continuous ambient air samples (Baker and Brown 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:

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).

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

mid-1960 (Brown and Baker 1961). A paper from PGDP describes the air sampling equipment used 

during this period (Mitchell undated). A 115 volt electric-powered air sampler continuously collects 

particulate samples on membrane filters, having a pore size of 0.8 micron, at a flow rate of 0.3 cubic 

feet per minute. 

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 U.S. Environmental Protection Agency, Commonwealth of 

Kentucky, and (as of 1997) 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. Brown and Mitchell (1960), Rogers 

and Jett (1989), and Lockheed Martin Energy Systems (LMES 1996) contain maps showing the old 

and new naming conventions.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 9 of 36

Table 4-1. Ambient air sampling stations.

Designation Location

Direction from 

center of Plant

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

PE-8 At security fence near McCaw Road East

PS-2 At security fence about 300 m south of Building C-100 South

PW-5 At security fence near C-749 West

IN Off site about 2 km north of PN North

IE Off site about 1 km east-northeast of PE East

ISE Off site about 2 km southeast of PS Southeast

IS Off site approx 1.5 km south-southwest of PS South

IW Off site approx 1.2 km west-northwest of PW West

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

See figure 4-5 for air sampling station locations.

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 99Tc. Overall, the alpha activity was about 7% of the beta activity in air 

samples [2]. 

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 air

monitoring 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 

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 cascade buildings.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 10 of 36

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

4.2.3 Methodology

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

transport modeling approaches is limited by several factors:

• 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 or less [3]. 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 [4]. 

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

over the 50-year history of PGDP. The records show that releases fell by 5 orders of magnitude while 

concentrations decreased by less than 3 orders of magnitude [5]. 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).
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 11 of 36

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 [6]. 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 Estimation of Ambient Airborne Radionuclide Concentrations

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 [7]. 

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 Data for Ambient Airborne Radionuclide Concentrations and Annual Airborne 

Releases

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 (the TBD text cites some of these reports; the 

References section lists all of them). Measurements from the air monitoring locations were collected 

from these documents and applied using the method described. Attachment A provides an expanded 

version of these 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, or the United States Enrichment Corporation. 

4.2.6 Estimation of Annual Intake from Airborne Radionuclides

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. 
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 12 of 36

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. 03 Effective Date: 08/24/2012 Page 13 of 36

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) (Ritter, Trowbridge, and Meiners 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, the cascade (PACE and University of Utah 2000; DOE 2000). There was 

a period when 237Np was recovered from feed plant ash and shipped to Oak Ridge National 

Laboratory (BJC 2000). 

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 [8]. 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

. 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 undated).

A paper prepared for the 1964 Bioassay Conference reports these findings (Baker 1962). 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 Cascade Improvement Program (CIP) and Cascade Upgrade Program (CUP) that replaced barrier 

material.

Baker (1987) 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 [International 

Commission on Radiological Protection] 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. 

Further:

In 1962, 14 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 [in Oak Ridge, 

Tennessee] to be counted. None of the employees had detectable 237Np (Baker 1962).
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 14 of 36

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

Trowbridge, and Meiners 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 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 Ritter, Trowbridge, and Meiners (1990): 

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 

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.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 15 of 36

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 (Ritter, 

Trowbridge, and Meiners 1990).

Table 4-2 lists uranium releases, alpha airborne concentrations, and intakes. Table 4-3 lists 99Tc 

releases, beta airborne concentrations, and intakes.

Table 4-2. 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
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 16 of 36

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

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

1996 1.1E+08

Table 4-3. 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)

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. 03 Effective Date: 08/24/2012 Page 17 of 36

4.2.7 Summary of Transuranic and Fission Product Exposure

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

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

99Tc as the fission product of concern. Tables 4-4 and 4-5 are derived from Barton (2006) and 

ORAUT (2012a), respectively. 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 (BJC 2000; PACE and University of Utah 

2000).

Table 4-4. Barton (2006) 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 

post-1983b

Tc/Np 

recovery 

operations, 

C-400c

Balance of 

plant

pre-1983c

Balance of 

plant

post-1983b

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).

Table 4-5. ORAUT (2012a) 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 

post-1983b

Tc/Np 

recovery 

operations, 

C-400c

Balance of 

plant

pre-1983c

Balance of 

plant

post-1983b

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 4-5 converted from picocuries per gram of uranium (2% enriched) to pCi/pCi of uranium (natural). Conservative 

assumption when uranium basis by weight is 2% enriched because the activity is higher, however when uranium activity 

basis is used, the lower specific activity of natural uranium should be used.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 18 of 36

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- Müller (GM) counter 3 ft 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 [9]. 

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

ambient radiation levels. TLD readings are 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 [10]. 

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 [11]. The highest security 

fence observation was 224 mrem deep dose equivalent for 2,000 hours (BJC 1999). This occurred in 

1998 near the UF6 cylinder storage yards. 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 [12]. 

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 

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

operations. An assumed deep dose equivalent rate of 224 mrem/yr for all years would be reasonable, 

and deficiencies in earlier measurement techniques thereby become immaterial. 

4.3.1 Ambient Radiation

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 (the TBD text cites some of these reports;

the References section lists all of them). Ambient radiation includes natural background and that from 

the facility. Data in these documents (see the References section) include GM counter and TLD 

radiation measurements. Table 4-6 lists the GM counter and TLD results. 

Figure 4-5 shows the locations of air sample stations where the GM counter readings were taken and 

Figure 4-6 shows the locations of the TLDs.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 19 of 36

Table 4-6. External dose equivalent rates, PGDP external gamma exposure rate of mrem/2,000 hr.

Location PN PE PS PW IN IE IS IW ISE

Reported 

background

1962 40 60 40 40 40 40 40 40

1963 40 60 40 40 40 40 40 40 40

1965 40 40 40 40 40 40 40 40 40

1966 40 40 40 40 40 40 40 40 40

1967 40 40 40 40 40 40 40 40 40

1968 40 40 40 40 40 40 40 40 40

1969 40 100 40 40 40 40 40 40 40

1970 40 80 40 40 40 40 40 40 40

1971 40 80 40 40 40 40 40 40 40

1982 15 17 33 35 15 15 17 14 17

1984 16 29 45 38 15 16 16 15 21

1986 11 22 24 24 11 0 11 12 12

1987 12 78 38 26 66 122 196 162 156

1988 12 187 139 37 8 11 11 14 14

1989 9 109 108 27 6 5 9 9 9

1991 12 18 88 38 10 10 13 14 16

1992 12 55 81 38 10 10 14 13 13

1993 10 10 97 28 8 10 10 12 12

Location T1 T2 T3 T47 T48 T50 T51 T52 T53

1998 114 102 224 26

1999 97 167 173 25 23

2000 97 171 107 28 42 24

2001 115 186 99 53 33 39 37 30 68 26

The ambient radiation measured by GM counter and TLDs near the security fence included natural 

background radiation, fallout, and cosmic radiation. The GM counter readings and the TLD results 

provided an indication of worker exposure levels in the general proximity of the security fence, but not 

inside the buildings. Table 4-6 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. 

Observations reported in the early environmental reports (before 1982), based on surveys with 

portable instruments, reported 0.02 mrem/hr at all locations, with the exception of air sampling station 

PE, which is near the UF6 cylinder storage yards. For a 2,000-hour 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 [13]. 

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

literature values for State of Kentucky 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 [14]. 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.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 20 of 36

Figure 4-5. Air sampling stations at PGDP (MMES 1990).
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 21 of 36

Figure 4-6. TLD locations.

4.3.2 Radiological Conditions in Cylinder Storage Yards

Observations reported in the early environmental reports (before 1982), based on surveys with 

portable instruments, reported increased dose rates at air sampling station, PE which is near the UF6 

cylinder storage yards. Several fenceline TLDs were adjacent to the 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 [15]. 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 significant time 

in the area to zero [16]. 

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

(ORAUT 2012b.)
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 22 of 36

Given this information, an ambient annual radiation dose equivalent of 224 mrem should be applied to 

unmonitored workers working near the cylinder yards [17]. Coworker dose is recommended for an 

unmonitored worker who worked primarily in the cylinder yards and should have been monitored.

4.3.3 Radiological Conditions Inside Buildings

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 [18]. 

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 listed in Table 4-6 were adjusted to reflect a 2,000-hour work 

year. The early data were originally reported in site environmental reports as GM counter dose rates 

in millirem per hour. The TLD data were originally reported 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 [19]. Therefore, a 

2,600-hour work year (multiplication factor of 1.3) should be applied as a maximizing assumption, 

while a 2,500-hour work year (multiplication factor of 1.25) should be assumed for best estimates. An 

uncertainty of 30% should also be applied to the assigned doses (ORAUT 2006). 

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 [20]. 

4.5 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section with information that identifies the source and 

justification for each item. Conventional references are provided in the next section that link data, 

quotations, and other information to documents available for review on the Oak Ridge Associated 

Universities (ORAU) Team servers.

[1] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

This wording demonstrates the emerging science of environmental monitoring; as 

instrumentation improved, more was learned about the environment. State-of-the-art 

instrumentation was crude by today’s standards and procedures were not clearly defined.

[2] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

Summary statement of sampling methods, and general results.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 23 of 36

[3] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

Summary statement based on environmental monitoring report data represented in 

Figures 4-2 and 4-3 that cover the life span of the facility.

[4] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

These data are presented in this way to demonstrate that the sampling program was recording 

changes in the airborne activity that were primarily from sources other than the facility. These 

data provide the basis for airborne intake for the unmonitored worker. The primary source 

term in this case originates from offsite sources.

[5] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

This is a summary statement that is based on environmental reports and represented in 

Figures 4-2 and 4-3 (see the description of reference sources in Section 4.2.5). In addition, 

the purpose is to establish basis for estimates of airborne concentrations in years for which

applicable data are not a part of the record.

[6] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

There are a limited number of observations and not of wide distribution. It is clear that 

determination of the location of the unmonitored worker is difficult. Using the maximum 

observed concentration is favorable to claimants.

[7] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

Estimates were required for missing years. The method of selection is presented.

[8] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

The referenced document (CCCC undated) states that high alpha activity was found on metal 

without a visible uranium stain (low specific activity). As is often the case with uranium, the 

visible stain indicates its presence. Minor activity can be found without a stain (especially with 

modern instruments), but activity to the extent observed was outside the experience of the 

survey technicians and Health Physics staff.

[9] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

This summary statement is based on environmental operating reports. The surveys 

conducted by GM are “snapshots.” The continuous monitoring by TLDs provides a more 

complete picture. No documents have been found to distinguish the environmental TLD chip 

set from the chips used in personnel monitoring.

[10] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

This summary statement is based on environmental operating reports. The surveys 

conducted by GM are “snapshots.” The continuous monitoring by TLDs provides a more 

complete picture. No documents have been found to distinguish environmental TLD chips as 

different than the chips used in personnel monitoring.

[11] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

Two factors contribute to the buildup of external doses around the cylinders. It is well 

understood that the buildup of progeny increases the gamma field around stored uranium. 

Because depleted uranium is a byproduct of the enrichment process, the storage of depleted 

uranium increased proportionally to production. Some depleted uranium was removed from 

onsite storage yards, but is not seen as a significant factor in this case.

[12] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

Two factors contribute to the buildup of external doses around the cylinders. It is well 

understood that the buildup of progeny increases the gamma field around stored uranium. 
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 24 of 36

Because depleted uranium is a byproduct of the enrichment process, the storage of depleted 

uranium increased proportionally to production. Some depleted uranium was removed from 

onsite storage yards, but is not seen as a significant factor in this case.

[13] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

This is a summary to form support for the numbers chosen. Later instrumentation provides 

better data, and the stability and knowledge of the source term add to confidence.

[14] East, James E. PrSM Corporation. Senior Health Physicist. July 2006. 

Environmental photon surveys were consistent across the lower Ohio Valley, including most 

monitoring points around PGDP, with the exception of areas within a few meters of the 

depleted uranium storage cylinders.

[15] East, James E. PrSM Corporation. Senior Health Physicist. July 2006. 

The depleted cylinders are arranged such that unmonitored workers would not spend a lot of 

time in the area. Some maintenance of fence, lighting, and grounds is to be expected. The 

assumption of 2,000 hours of exposure within 100 m of the cylinders is judged to not 

underestimate the dose to unmonitored workers.

[16] East, James E. PrSM Corporation. Senior Health Physicist. July 2006. 

There are two reasons for this statement: (1) During the earlier years, the depleted uranium 

was young, and it takes decades of storage for the gamma-emitting progeny to ingrow, and 

(2) a series of changes in directives, rules, and regulations elevated the importance and 

compliance of radiological postings.

[17] East, James E. PrSM Corporation. Senior Health Physicist. July 2006. 

This summary statement provides a simple method of accounting for ambient radiation.

[18] East, James E. PrSM Corporation. Senior Health Physicist. July 2006. 

Specific studies of area dose rates and incident investigations could be used to refine dose 

estimates. However, based on available documentation, such studies are unlikely to yield 

dose estimates to an unmonitored worker that are greater than the external doses proposed in 

this section.

[19] East, James E. PrSM Corporation. Senior Health Physicist. July 2006. 

To be consistent with Project guidance, only expected work-hours onsite are included.

[20] East, James E. PrSM Corporation. Senior Health Physicist. July 2006.

There are a limited number of data points for each year. The assignment of the maximum 

uptake is favorable to the claimant.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 25 of 36

REFERENCES

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

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Kentucky, May 22. [SRDB Ref ID: 27035]

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

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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, 

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30636]

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

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Hightower, J. R., L. R. Dole, D. W. Lee, G. E. Michaels, M. I. Morris, D. G. O'Conner, S. J. Pawel, R. 

A. Schmoyer, L. D. Trowbridge, and V. S. White, 2000, Strategy for Characterizing 

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1994, Paducah Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12622]
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for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October. 

ORAUT (Oak Ridge Associated Universities Team), 2006, Occupational Onsite Ambient Dose 

Reconstruction for DOE Sites, ORAUT-PROC-0060, Rev. 01, Oak Ridge, Tennessee, 

June 28.

ORAUT (Oak Ridge Associated Universities Team), 2012a, Paducah Gaseous Diffusion Plant –

Occupational Internal Dose, ORAUT-TKBS-0019-5, Rev. 03, Oak Ridge, Tennessee, August 

24.

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Occupational External Dose, ORAUT-TKBS-0019-6, Rev. 04, Oak Ridge, Tennessee, August 

24.

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Description, ORAUT-TKBS-0019-2, Rev. 03, Oak Ridge, Tennessee, August 22.

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. [SRDB Ref ID: 10870]

Ritter, R. L., L. D. Trowbridge, and S.E. Meiners, 1990, Neptunium Experience at PGDP, K/ETO-30, 

Martin Marietta Energy Systems, Uranium Enrichment Organization, Oak Ridge, Tennessee, 

September. [SRDB Ref ID: 30497]

Rogers, J. G., and T. G. Jett, 1989, Paducah Gaseous Diffusion Plant Site Environmental Report for 

1988, ES/ESH-8/V3, Martin Marietta Energy Systems, Paducah Gaseous Diffusion Plant, 

Paducah, Kentucky. [SRDB Ref ID: 39298]

Whiteaker, B. H., and R. B. Farrar, 1953, Water, Mud & Air Survey at the Paducah Plant, KYL-11, 

Carbide and Carbon Chemicals Company, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky. [SRDB Ref ID: 27034]

Environmental Reports and Related Documents 

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

KY-458, Union Carbide Corporation, Nuclear Division, Paducah Gaseous Diffusion Plant, 

Paducah, Kentucky, July 1. [SRDB Ref ID: 12648]

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

1964, KY-484, Union Carbide Corporation, Nuclear Division, Paducah Gaseous Diffusion 

Plant, Paducah, Kentucky. [SRDB Ref ID: 12650]
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Baker, R. C., and K. A. Davis, 1967, Environmental Monitoring Summary for the Paducah Plant for 

1965 & 1966, KY-543, Union Carbide Corporation, Nuclear Division, Paducah Gaseous 

Diffusion Plant, Paducah, Kentucky, September 1. [SRDB Ref ID: 12652]

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

Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12637]

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

Paducah Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12646]

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, July 20. [SRDB Ref ID: 12642]

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

2000, BJC/PAD-237, Kevil, Kentucky, October. [SRDB Ref ID: 12649]

CDM Federal Services, 2002, Paducah Site Annual Site Environmental Report for Calendar Year 

2001, BJC/PAD-219, Kevil, Kentucky, September. [SRDB Ref ID: 12647]

Davis, K. A., 1970, Environmental Monitoring Summary for the Paducah Plant for 1969, KY-624, 

Union Carbide Corporation, Nuclear Division, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky, November 2. [SRDB Ref ID: 12656]

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

Union Carbide Corporation, Nuclear Division, Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky, April 1. [SRDB Ref ID: 12657]

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

1967 and 1968, KY-582, Union Carbide Corporation, Nuclear Division, Paducah Gaseous 

Diffusion Plant, Paducah, Kentucky, May 9. [SRDB Ref ID: 12655]

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

Environmental Management Division, Kevil, Kentucky, January. [SRDB Ref ID: 12314]

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

1996, KY/EM-206, Environmental Management Division, Kevil, Kentucky, December. [SRDB 

Ref ID: 12316]

MMES (Martin Marietta Energy Systems), 1984, Environmental Monitoring Report, United States 

Department of Energy, Paducah Gaseous Diffusion Plant, Calendar Year 1983, KY-742, 

Paducah Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12594]

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. [SRDB Ref ID: 12595]

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

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

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

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

[SRDB Ref ID: 12615]

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Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12616]

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12618]

Oakes, T. W., 1983, “PGDP TLD Results,” memorandum to R. C. Baker, Union Carbide Corporation, 

Nuclear Division, February 23. [SRDB Ref ID: 37840]

Oakes, T. W., 1985, “PGDP TLD Results,” memorandum to R. C. Baker, Martin Marietta Energy 

Systems, Paducah, Kentucky, March 7. [SRDB Ref ID: 37842]

Oakes, T. W., 1985, “PGDP TLD Results,” memorandum to R. C. Baker, Martin Marietta Energy 

Systems, Paducah, Kentucky, February 27. [SRDB Ref ID: 37843]

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

Department of Energy Paducah Reservation and Surrounding Environs During 1986, ES/ESH1/V3, Martin Marietta Energy Systems, Oak Ridge, Tennessee, and Paducah, Kentucky, April. 

[SRDB Ref ID: 4084, pp. 6-267]

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4084, pp. 269-683]

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near the Paducah Plant, Report for the First Half of 1971, Paducah Gaseous Diffusion Plant, 

Paducah, Kentucky. [SRDB Ref ID: 18510]

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

Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12677]

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

Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12568]

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

Paducah Gaseous Diffusion Plant, Paducah, Kentucky. [SRDB Ref ID: 12570]

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

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

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GLOSSARY

air sampling 

Collection of samples of the ambient atmosphere to detect or measure the presence of 

radioactive material in the air.

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.

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.

byproduct 

Material left over from a nuclear, physical, or chemical process designed to produce a 

particular substance. Examples include the tailings or wastes from the extraction or 

concentration of uranium or thorium from ore.

committed effective dose equivalent (CEDE)

Sum of the effective dose equivalents to various tissues or organs in the body each multiplied 

by the appropriate tissue weighting factor and committed for a 50-year period following an 

acute intake or the onset of chronic intake. It does not include contributions from external 

dose. See dose.

contamination 

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

counter 

Radiation detection meters that detect and measure radiation in which one count indicates a

detected ionization event.

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.

depleted uranium (DU) 

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. 
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 31 of 36

depleted uranium

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. 

dose 

In general, the specific amount of energy from ionizing radiation that is 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.

dose equivalent

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dosimeter 

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter and thermoluminescent dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

effective dose

Sum of the equivalent doses in the principal tissues and organs of the body, each weighted by 

a tissue weighting factor that represents the relative contributions of that tissue or organ to the 

total detriment from uniform irradiation of the body. 

effective dose equivalent

Average of the dose equivalents weighted for the susceptibility of harm to different tissues or 

organs in units of rem or sievert. See dose. 

enriched uranium (EU) 

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7% by mass. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade

uranium contains greater than 90% 235U.

enrichment 

Isotopic separation process that increases the percentage of a radionuclide in a given amount 

of material above natural levels. For uranium, enrichment increases the amount of 235U in 

relation to 238U. Along with the enriched uranium, this process results in uranium depleted in 

235U. At PGDP this involves a process that occurs as UF6 passes through barriers in 

converters allowing isotopes of lower molecular weight to pass through.

environmental occupational dose 

Dose received from radiation site-related activities (i.e., above normal background levels) 

while on a site, which is often recorded by monitoring stations in specific areas or along the 

boundaries of facilities (e.g., plant stack emissions). 

exposure

In general, the act of being exposed to ionizing radiation.

film dosimeter 

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 32 of 36

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fission product (FP) 

(1) Radionuclides produced by fission or by the subsequent radioactive decay of 

radionuclides. (2) Fragments other than neutrons that result from the splitting of an atomic 

nucleus. 

gaseous diffusion enrichment 

Process by which uranium hexafluoride (UF6) is filtered through a series of semipermeable 

molecular barriers to separate the lighter 235U from the heavier, more naturally abundant 238U. 

gaseous diffusion plant 

Facility where uranium hexafluoride (UF6) gas is filtered to enrich the 235U and separate it from 

238U. The process requires enormous amounts of electric power and results in an increase in 

235U enrichment from 1% to about 3%. 

Geiger-Müller (G-M, GM) counter 

Most common radiation detection and measuring instrument, usually known simply as a 

Geiger counter, it is a gas-filled tube containing electrodes between which there is a voltage 

potential but no current flow. When ionizing radiation passes through, a short, intense pulse of 

current passes from one electrode to the other. The number of pulses per second (or counts 

per minute) indicates the rate of ionizing events in the tube. 

monitoring 

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements. 

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram).

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. 

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.

roentgen (R)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 
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exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts). 

Special Nuclear Material (SNM) 

Plutonium or uranium enriched to a higher-than-natural assay including 239Pu, 233U, uranium 

containing more than the natural abundance of 235U, or any material artificially enriched in one 

of these isotopes.

spent nuclear fuel

Fuel that has been in a reactor long enough to become ineffective because the proportion of 

fissile material has dropped below a certain level. Spent nuclear fuel contains fission and 

activation products.

tails 

Unwanted remains of a process such as mining, milling, chemical separation, or another 

refinement method. 

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. 

transuranic (TRU) elements 

Elements with atomic numbers above 92 (uranium). Other examples include plutonium and 

americium. 

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 

Paducah Gaseous Diffusion Plant and the Oak Ridge K-25 site.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 34 of 36

ATTACHMENT A

OUTDOOR BETA AND ALPHA CONCENTRATIONS

Page 1 of 3

Table A-1 lists outdoor beta concentrations, and Table A-2 lists outdoor alpha concentrations.
Document No. ORAUT-TKBS-0019-4 Revision No. 03 Effective Date: 08/24/2012 Page 35 of 36

ATTACHMENT A

OUTDOOR BETA AND ALPHA CONCENTRATIONS

Page 2 of 3

Table A-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

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. 03 Effective Date: 08/24/2012 Page 36 of 36

ATTACHMENT A

OUTDOOR BETA AND ALPHA CONCENTRATIONS

Page 3 of 3

Table A-2. Outdoor alpha concentrations.

Year

Release 

(Bq/yr)

U

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

uranium 

intake

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

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

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