ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 28

Document Title:

Portsmouth Gaseous Diffusion Plant – 

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0015-4 

Revision: 01

Effective Date: 09/08/2006

Type of Document: TBD

Supersedes: Revision 00

Subject Experts: Paul J. Demopoulos, Milton Gorden, and Donald Palmrose

Document Owner

Approval: Signature on File Approval Date: 08/02/2006

Paul J. Demopoulos, TBD Team Leader

Approval: Richard Merrill Signature on File for Approval Date: 08/03/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/02/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 08/07/2006

Kate Kimpan, Project Director

Approval: Brant Ulsh Signature on File for Approval Date: 09/08/2006

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-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 2 of 28

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/17/2004 00 New Technical Basis Document for the Portsmouth Gaseous 

Diffusion Plant - Occupational Environmental Dose. First approved 

issue. Initiated by Mark D. Notich.

09/08/2006 01 Updates document for biennial review. Adds a Purpose and Scope 

subsection to the Introduction (Section 4.1). Changes the ambient 

maximum value to 44.8 mrem in Table 4-6. Adds Table 4-7 

Perimeter road ambient radiological conditions from GM 

measurements and citations. Adds locations 9 and 15 of GM 

monitoring to Figure 4-1. Approved issue of Revision 01. This 

revision results in an increase in assigned dose and a PER is 

required. Training required: As determined by the Task Manger. 

Initiated by Paul J. Demopoulos.
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 3 of 28

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

4

4.1 Introduction ..............................................................................................................................5

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

4.1.2 Scope...........................................................................................................................6

4.2 Internal Dose From Onsite Atmospheric Radionuclide Concentration ...................................... 6

4.2.1 Ambient Air Sample Collection Network ....................................................................... 6

4.2.2 Methodology.................................................................................................................8

4.2.3 Estimation of Ambient Airborne Radionuclide Concentrations Prior to 

1964 .............................................................................................................................8

4.2.4 Data for Ambient Airborne Radionuclide Concentrations and Annual 

Airborne Releases........................................................................................................9

4.2.4.1 Airborne Release Estimation for Neptunium, Plutonium, Uranium 

Daughters, and 99Tc ..................................................................................... 10

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

4.2.6 Contribution of Airborne Radionuclides to External Dose............................................ 11

4.3 External Dose ........................................................................................................................14

4.3.1 Ambient Radiation ...................................................................................................... 17

4.3.1.1 Background Radiation.................................................................................. 18

4.3.1.2 Ambient Radiological Conditions in Cylinder Storage Yards......................... 20

4.3.1.3 Ambient Radiological Conditions Inside Buildings ........................................ 21

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

References ........................................................................................................................................23

Glossary ............................................................................................................................................27

LIST OF TABLES

TABLE TITLE

4-1 Ambient air sampling station locations .....................................................................................

PAGE

7

4-2 Maximum annual intakes and geometric standard deviations (Bq per year)........................... 12

4-3 Instruments used for ambient environmental monitoring ........................................................ 14

4-4 Environmental measurement locations................................................................................... 16

4-5 Site interior ambient radiological conditions ........................................................................... 16

4-6 Perimeter road ambient radiological conditions from GM and TLD measurements ................ 17

4-7 Perimeter road ambient radiological conditions from GM measurements ............................... 18

LIST OF FIGURES

FIGURE TITLE

4-1 Approximate TLD locations ....................................................................................................

PAGE

15

4-2 Cumulative quantity of material by year.................................................................................. 19

4-3 Annual MTU vs. work year doses (2000 hrs/yr)...................................................................... 20
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 4 of 28

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

Bq Becquerel

CAF2:Dy calcium fluoride:dysprosium

Ci curie

cm centimeter

DAC Direct Air Concentration

DCG Derived Concentration Guide

DOE U.S. Department of Energy

DU depleted uranium (uranium with less than 0.7% 235U)

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EPA U.S. Environmental Protection Agency

ERP Extended Range Product

GDP Gaseous Diffusion Plant

GSD Geometric Standard Deviation

HEU highly enriched uranium (over 20% 235U)

HF hydrogen fluoride

LiF lithium fluoride

MTU metric tons of uranium

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

ORAU Oak Ridge Associated Universities

PORTS Portsmouth Gaseous Diffusion Plant

REMS Radiation Exposure Monitoring System

RU recycled uranium

TBD technical basis document

TLD thermoluminescent dosimetry

UF6 uranium hexafluoride 

USEC United States Enrichment Corporation
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 5 of 28

4.1

This Portsmouth Gaseous Diffusion Plant (PORTS) technical basis document (TBD) describes 

the potential for internal dose from the breathing of airborne concentrations of radionuclides 

released on the PORTS site (Section 4.2), the potential for external dose from sources of 

radiation outside and inside the process buildings (Section 4.3), and possible sources of 

uncertainty associated with both these sources of environmental dose (Section 4.4). 

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

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 the facility in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the 

] 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.

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 6 of 28

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

The receptors of concern whom this TBD addresses are unmonitored workers; that is,

employees at PORTS who did not wear personnel dosimetry, and who were not monitored for 

internal dose. At PORTS, there were several periods when all workers, even those with

administrative duties, wore film badges or thermoluminescent dosimeters (TLDs) at all times. 

However, not all film badges were analyzed. For these periods, representative samples for 

various groups of workers were analyzed (ATL 2003a). To provide the basis for estimating the 

environmental dose for years when monitoring did not occur or was not sufficient to apply to 

coworkers, this TBD provides actual and estimated annual intakes and ambient environmental 

doses for every year from 1955 to 2001.

4.1.1

The purpose of this TBD is to describe the environmental radiation released or 

emanating from fixed sources in the environment for the PORTS plant. This document 

discusses historical and current practices in relation to the evaluation of environmental 

external and internal exposure data for PORTS workers.

Purpose

4.1.2

This TBD contains supporting documentation to assist in the evaluation of ambient 

external and internal doses from environmental releases or fixed source terms for

workers at PORTS. An objective of this document is to provide supporting technical 

environmental data or survey information to evaluate, with claimant favorable 

assumptions for worker external and internal dose assessments. 

Scope

4.2

4.2.1

INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATION

In 1964, PORTS established a network of permanent on- and offsite stations to collect 

continuous ambient air samples (DOE 1994). Sampling data beginning in 1964 are

available for five offsite locations (one several miles from PORTS to provide

background measurements) and for four onsite locations beginning in 1974 (Goslow 

1986; GAC 1975). 

Ambient Air Sample Collection Network

The principal purpose of the ambient air monitoring network is to assess whether air 

emissions from PORTS affected air quality in the surrounding area. This would 

demonstrate compliance with either DOE Derived Concentration Guidelines (DCGs), 

and U.S. Environmental Protection Agency (EPA), State of Ohio Environmental 

Protection Agency, and (as of 1997) Nuclear Regulatory Commission (NRC) 

regulations for airborne releases to the general public around PORTS. Thus, most of 

the monitoring stations are off the site, along the site boundary, or next to the 

Perimeter Road. There have been a limited number of onsite monitoring stations. 

Therefore, this study considers only those ambient air monitoring locations at or within 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 7 of 28

the PORTS site boundary for worker internal dose. Table 4-1 lists these air monitors 

and their locations. 

Table 4-1. Ambient air sampling station locations. 

Designation Location

Direction from 

center of site

Within PORTS boundary (fence line)

A3 South Access Rd., south of DOE property line SSW

A12 McCorkle Rd., north of East Access Rd. ENE

A24 North Access Rd., north of Shyville Rd. N

A29 U.S. Rt. 23 on near leg of old GAT sign W

Onsite air sampling locations (near Perimeter Road)

A10 X-230-J3 west (ditch) environmental monitoring station W

A35 X-230-J2 south (holding pond) environmental monitoring station S

A36 X-611 water treatment plant, southeast corner of facility parking lot ENE

A38 X-230-J environmental storage building ESE

A39 X-760 chemical engineering facility Near mid-point

A40 X-100 administration building penthouse Near mid-point

Source: Table 2.1.8 on page 135 of DOE (1989).

As would be expected, several changes to the sampling program were associated 

principally with sampling technique and equipment; there were at least two major 

changes of monitoring locations. From 1964 to 1993, low-volume samplers were the 

principal type of sampler used, employing several filter types and flow rates. In

general, filters were changed after sampling approximately 200 to 210 cubic meters of 

air. Depending on the set flow rate (1.2 or 0.3 cubic meters per hour), filter exchanges 

were performed weekly or monthly. Major upgrades of the monitoring network 

occurred in 1987 (number of locations and sampling technique) and from 1993 to 1995 

[addition of high-volume samplers and segregation of monitoring responsibilities 

between DOE and United States Enrichment Corporation (USEC)]. 

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

counts. Gross alpha activity was associated with the release of uranium, and betagamma activity was associated with the release of 99Tc and daughter products of 

uranium (Goslow 1986). If the gross counts exceeded established limits of 100 dpm 

alpha or 200 dpm beta-gamma, the filters would be analyzed for specific radionuclides 

(DOE 1994). Only in the last several years have filters been analyzed for specific 

radionuclides. However, no air samples exceeded established limits while they were in 

effect. 

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

operations, including accidental releases. PORTS release data have been estimated 

or recorded since 1954 for uranium, since 1967 for uranium daughter products, and

since 1976 for 99Tc (Goslow 1986). Dose reconstructors must consider several issues, 

such as the accuracy of the amounts released and the fact that there was no ambient 

air monitoring network in operation for the first 10 years of operations and not all 

radionuclides of concern were originally measured in the air effluent release streams. 

Therefore, this TBD discusses a claimant-favorable methodology for estimating

potential radionuclide air concentrations, not only for the period from when operations 

began to 1964 but also to account for potential uncertainty in the release data. The 

assumptions developed to address the missing data will be overly conservative to help 

ensure that the methodology is claimant-favorable.
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 8 of 28

4.2.2

Estimating airborne concentrations at locations around the PORTS site using 

traditional transport modeling approaches is limited by the following 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 release points and onsite receptor locations

• The density and configurations of buildings at PORTS

Therefore, it is unnecessary to develop a complex dispersal model and perform 

transport calculations for the PORTS site because such modeling is not likely to yield 

accurate results nor would it be claimant-favorable.

The methodology for determining internal dose from onsite atmospheric radionuclide 

concentrations applies directly to the available air sampling data. To be claimantfavorable for each year for which data are available, the maximum site measurement 

for gross alpha (uranium) and gross beta-gamma (uranium daughter products and 

99Tc) will be applied for the entire site. Air sampling measurement error and 

uncertainty will be accounted for to provide the 95th percentile (+2σ), so the following 

relationship can determine a geometric standard deviation (GSD) for each 

radionuclide:





 





















= 1.65585

1

Maximum Annual

th percentile GSD 95 (4-1)

The annual intake associated with each radionuclide of concern is based on the 

fraction given in the annual release as appropriately contributing to either gross alpha 

or gross beta-gamma sampling results. Since the gross alpha is only from uranium 

radionuclides, the highest air monitoring concentration is directly applied to the fraction 

of a uranium radionuclide to the total uranium release then multiplied by the assumed 

individual ventilation rate to yield the annual intake for each uranium radionuclide. The 

gross beta/gamma is segregated in a similar manner between 99Tc and the uranium 

daughters where the uranium daughters are further fractioned by 6 percent of the 

uranium daughter fraction for 231Th and the remainder equally split between 234Th and 

234mPa.

4.2.3

Because there are no ambient airborne monitoring data prior to 1964, an alternative 

empirical approach was developed for estimating airborne radionuclide concentrations 

prior to 1964. Based on other dose reconstruction developments (ATL 2003b), an 

empirical relationship of maximum annual air concentration for the entire site with 

respect to uranium release estimates provides a direct relationship of air concentration 

to material release. The empirical relationship is described by Equation 4-2. The 

maximum uranium air concentration measured at the air monitoring station is divided 

by the release estimates for a given year to estimate an annual, site-specific, Chi/Q 

value.

Estimation of Ambient Airborne Radionuclide Concentrations Prior to 1964
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 9 of 28

Re )(

)( )(

3

3

Annual Radionuclide Activity lease Ci

MaximumRadionuclide Air Concentrat Ciion m Chi mQ

−

− = (4-2)

The empirical Chi/Q value can be determined for each year for which air sampling data 

are available. However, due to the significant decrease in annual releases beginning 

in 1994 (from placing the enrichment cascade in cold shutdown), the empirical Chi/Q 

values will be determined only for 1964 to 1993 (i.e., the beginning of air monitoring 

data to the last operational year). To be claimant-favorable, the maximum of all 

empirical Chi/Q values will be used to estimate radionuclide air concentrations for the 

entire PORTS site for years without ambient airborne monitoring data based on annual 

radionuclide release estimates. This approach can estimate annual intake loadings by 

using the estimated air concentrations resulting from the empirical Chi/Q values in the 

same process as using the actual maximum annual airborne radionuclide 

concentration.

4.2.4

This study relied on several data sources for estimating ambient airborne radionuclide 

concentrations and for the annual airborne releases by radionuclide. The two principal 

sources were the annual environmental reports from 1972 and 2001 (the Reference 

List at the end of this TBD includes all DOE site environmental reports from 1973 

through 2002) and a special study of historic radiological releases (Goslow 1986). 

Measurements from the air monitoring locations listed in Table 4-1 were collected from 

these documents and applied to the methodology described above. When data were 

not provided or additional information was required, these core documents were 

supplemented by air monitoring information provided by USEC (ATL 2003c) or by 

estimating the release quantity for specific situations using historical trends or physical 

processes (such as secular equilibrium). 

Data for Ambient Airborne Radionuclide Concentrations and Annual Airborne 

Releases

The data provided by Goslow (1986) include accidental releases in addition to routine 

emissions. The largest of such releases occurred in March 1978, when a 14-ton feed 

cylinder fell from its carrier and cracked open, releasing an estimated 4,820 kg of 

uranium to the atmosphere and an additional 680 kg of uranium through the plant 

sewers, for a total activity of less than 3 Ci. Other major accidental releases included 

two valve failures on a tails cylinder in July 1969 and October 1978, and a process 

malfunction in the Side Purge Cascade in December 1983. In addition, a string of 

accidental releases of mostly depleted uranium during the first 5 years of plant 

operation accounted for essentially all the uranium lost to the atmosphere from 1955 

through 1958 and 20% of the losses in 1959. Thus, the variations seen in the uranium 

radionuclide fractions and ultimately in the calculated uranium radionuclide intake 

values are due to the quantity of depleted or enriched material released during a given 

year. 

The uranium daughters identified as a potential of environmental concern were 

Thorium-231 (231Th), Thorium-234 (234Th), and Protactinium-234m (234mPa) (Goslow 

1986). PORTS estimated the fraction of 231Th to be 2.5 percent to 6 percent of all 

uranium daughter radiation measurements; the remaining uranium daughter radiation 

is evenly split between 234Th and 234mPa. For the purposes of estimating the intake of 

specific uranium daughter radionuclides, 6 percent should be from 231Th and 47 

percent for 234Th and 234mPa. Also, since the enrichment cascade was shutdown in 

1993, if the quantity of uranium daughters released were not specifically given in the 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 10 of 28

DOE annual environmental reports after 1993, they will be assumed to be in secular 

equilibrium (i.e., at the same activity as the parent radionuclide).

It is important to note that the release data has values and relationships that one might

believe to be atypical. For example, there are years where the quantity of uranium 

daughters released exceeds the quantity of total uranium released (i.e., 1979, 1981, 

and 1984). The expected situation is for the uranium and its daughters would exist 

together and, in accordance with their respective decay half-lives, in secular 

equilibrium where the activity of uranium daughter would be the same as the uranium

parent. The only time that the uranium daughters can be considered in secular 

equilibrium with the uranium parent is after 1993 when the cascade was shutdown. 

Another example is isotopic quantity of uranium released. The first two years of 

releases (1955 and 1956) were zero for 234U while the same years had as the total 

uranium released as approximately 0.55 and 0.24 Curies. Such atypical data is 

probably due to the nature of the physics of gaseous diffusion, initial enrichment of 

feed at Paducah prior to enrichment at PORTS, and how and where in the enrichment 

cascade the releases were performed.

4.2.4.1

A review of the airborne environmental release data for the PORTS site 

shows a lack of information during the first twenty years of operation for 

radionuclides that later became a source of concern with subsequent 

inclusion in the environmental monitoring program. The principal 

radionuclides added to the environmental monitoring program were the 

uranium daughters in 1967 and 99Tc in 1976. However, there could be a 

concern for releases of transuranics from the processing of recycled 

uranium and for the releases of uranium daughter prior to 1967 and for 99Tc 

prior to 1976. This section investigates the potential for these unknown 

releases and how to include their contribution for the environmental 

occupational dose.

Airborne Release Estimation for Neptunium, Plutonium, Uranium 

Daughters, and 99Tc

Based on the information given in the Recycled Uranium Mass Balance 

Project (BJC 2000), the principal transuranics that were fed into the 

cascade were Np and Pu radionuclides. Both were determined to have

annual inventory levels of less than 100 grams over the entire history of 

PORTS (see Table 5.1-1 and Figure 5.1-1 of BJC 2000). In addition, the Np 

and Pu that entered the cascade soon plates out on cascade components 

(BJC 2000, pg 67). The Np and Pu radionuclides would be removed with 

the uranium deposits when the process equipment was removed and 

decontaminated. The uranium deposits, containing the Np and Pu 

radionuclides, went into the uranium recovery process so a small fraction of 

the transuranics stayed in the recovered uranium as impurities or were

removed into the liquid or solid wastes stream. Therefore, due to the low 

levels of Np and Pu radionuclides that entered the cascade, their plating 

out, and the material staying in liquid and solid product or waste streams, 

airborne releases of Np and Pu are considered to be in trace amounts only 

and not contributors to the environmental occupational doses.

To estimate the potential releases of uranium daughters prior to the 

beginning of release measurements in 1967, a key item to note is the 

tremendous difference in the half-lives of the parent radionuclides (over 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 11 of 28

millions of years) and of the daughters (minutes to days). Such differences 

in half-lives results in the daughters being in what is defined as secular 

equilibrium with the parent radionuclide (i.e., daughter’s activity is at the 

same activity as the parent radionuclide). Therefore, by assuming secular 

equilibrium, 231Th would have the same activity as its parent, 235U and 

234mPa and 234Th would each have the same activity as its parent, 238U. For 

most years, this assumption would be claimant favorable since the average 

uranium daughter released activity was approximately 0.3 of the uranium 

released activity where this value would be closer to 1.0 if both were in 

secular equilibrium.

With the discovery of 99Tc in the east drainage ditch during the first quarter 

of FY1975, this radionuclide was added to those already being monitored 

(uranium and uranium daughters). What is known of the physical migration 

process for 99Tc is that it plates out on surfaces within the cascades and 

slowly migrates through the enriched cascade, and would result in relatively 

higher counts for the gross beta monitoring of the release vents once 

reaching the purge cascade (ATL 2003g). Thus, with the pre-enrichment of 

the recycled uranium at Paducah first removing the 99Tc from the PORTS 

feed material and the slow migration time to reach any release point such 

as the purge cascade, release of 99Tc is considered negligible prior to 1975.

4.2.5

Using the maximum airborne radionuclide concentration in a year (either from a direct 

reading or a calculated value), this study used an assumed individual ventilation rate of 

2,400 cubic meters per year to derive claimant-favorable annual intakes for the 

radionuclides of concern. GSDs associated with these intakes were estimated using 

Equation 4-1. Table 4-2 lists annual intakes and GSDs for each radionuclide of 

concern for each year.

Estimation of Annual Intake from Airborne Radionuclides

4.2.6

From the data shown in Table 4-2, the levels of annual intakes during the early years of 

operations at PORTS are two to three orders of magnitude higher than the years after 

1966. High airborne concentrations can have a significant contribution to the direct 

radiation exposure through the pathway of submersion within a cloud of radioactive 

material. With such variations in the air concentrations used to determine the 

estimated annual intakes, a demonstration of the potential occupational impacts would 

show the potential contribution to the external site dose from submersion. Since the 

highest airborne intake is due to uranium daughters based on the 1955 intake, these 

radionuclides and concentration levels would demonstrate the most claimant-favorable 

conditions. The submersion dose was calculated using the air submergence dose 

factors in Federal Guidance Report No. 12 for the uranium daughter radionuclides of 

231Th, 234Th, and 234mPa. As previously stated, the contribution of each radionuclide is 6 

percent of the uranium daughter fraction for 231Th and the remainder equally split 

between 234Th and 234mPa. The potential contribution to the annual external exposure 

from these uranium daughters at the highest airborne concentrations only yields 1.2 

μrem. Therefore, airborne radioactivity at the PORTS site over all years of concern 

would have a negligible impact to the annual occupational external exposure.

Contribution of Airborne Radionuclides to External Dose
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 

09/08/2006 Page 12 of 28

Table 4-2. Maximum annual intakes and geometric standard deviations (Bq per year). 

Year 99-Tc GSD Uranium

daughters GSD U-234 (e) U-235 U-236 U-238 Total U GSD

1953 — — — — — — — — — — 

1954 — — — — — — — — — — 

1955(a) 0.000E+00 0.000E+00 7.322E+03 1.419E+00 6.011E+01 2.706E+00 1.422E-01 1.174E+02 1.803E+02 1.450E+00

1956(a) 0.000E+00 0.000E+00 3.159E+03 1.419E+00 2.598E+01 8.139E-01 6.137E-02 5.108E+01 7.793E+01 1.450E+00

1957(a) 0.000E+00 0.000E+00 2.945E+02 1.419E+00 1.100E+00 1.320E-01 5.721E-03 3.563E+00 4.801E+00 1.450E+00

1958(a) 0.000E+00 0.000E+00 2.436E+03 1.419E+00 3.497E+01 1.540E+00 4.732E-02 3.629E+00 4.019E+01 1.450E+00

1959(a) 0.000E+00 0.000E+00 6.051E+03 1.419E+00 4.289E+01 2.794E+00 1.175E-01 5.349E+01 9.930E+01 1.450E+00

1960(a) 0.000E+00 0.000E+00 2.316E+03 1.419E+00 1.540E+01 1.078E+00 4.498E-02 2.171E+01 3.823E+01 1.450E+00

1961(a) 0.000E+00 0.000E+00 4.645E+03 1.419E+00 3.277E+01 2.222E+00 9.023E-02 4.111E+01 7.620E+01 1.450E+00

1962(a) 0.000E+00 0.000E+00 1.513E+03 1.419E+00 1.628E+01 9.018E-01 2.938E-02 1.214E+01 2.935E+01 1.450E+00

1963(a) 0.000E+00 0.000E+00 2.142E+02 1.419E+00 3.079E+00 2.860E-01 4.160E-03 2.200E-02 3.392E+00 1.450E+00

1964 0.000E+00 0.000E+00 1.865E+02 1.466E+00 1.579E+01 1.283E+00 2.099E-02 9.867E-02 1.776E+01 1.434E+00

1965 0.000E+00 0.000E+00 6.216E+01 1.446E+00 5.751E+01 3.890E+00 8.398E-02 8.288E+00 7.104E+01 1.450E+00

1966 0.000E+00 0.000E+00 8.880E+00 1.520E+00 7.804E+00 5.651E-01 1.050E-02 2.153E-01 8.880E+00 1.520E+00

1967 0.000E+00 0.000E+00 2.664E+01 1.445E+00 7.992E+00 3.552E-01 1.050E-02 4.884E-01 8.880E+00 1.520E+00

1968 0.000E+00 0.000E+00 3.552E+01 1.434E+00 7.400E+00 2.960E-01 1.050E-02 1.184E+00 8.880E+00 1.520E+00

1969 0.000E+00 0.000E+00 3.552E+01 1.458E+00 1.830E+00 2.410E-01 1.050E-02 6.809E+00 8.880E+00 1.520E+00

1970 0.000E+00 0.000E+00 4.618E+01 1.101E+00 3.102E+00 1.432E-01 4.514E-03 5.966E-01 3.818E+00 1.082E+00

1971 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.992E+00 1.834E-01 5.249E-03 1.207E+00 4.440E+00 1.134E+00 

1972 0.000E+00 0.000E+00 0.000E+00 0.000E+00 8.626E-01 6.470E-02 1.785E-03 5.823E-01 1.510E+00 1.097E+00

1973 0.000E+00 0.000E+00 7.193E+00 1.268E+00 2.037E+00 9.960E-02 2.729E-03 9.960E-02 2.309E+00 1.138E+00

1974 0.000E+00 0.000E+00 5.044E+01 1.406E+00 1.779E+01 9.884E-01 2.687E-02 4.448E+00 2.273E+01 1.359E+00

1975 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.009E+01 3.759E-01 1.333E-02 7.170E-01 1.128E+01 1.361E+00

1976 3.609E+01 1.419E+00 1.203E+00 1.419E+00 8.052E+00 3.779E-01 1.039E-02 3.862E-01 8.791E+00 1.351E+00

1977 2.559E+01 1.301E+00 5.214E-01 1.301E+00 2.001E+00 7.773E-02 2.729E-03 2.293E-01 2.309E+00 1.367E+00

1978 2.429E+01 1.308E+00 2.527E+00 1.308E+00 4.778E-01 3.341E-02 1.480E-03 7.396E-01 1.252E+00 1.325E+00

1979 5.479E+00 1.307E+00 4.022E+00 1.307E+00 7.536E-01 2.270E-02 9.553E-04 2.724E-02 8.081E-01 1.344E+00

1980 6.864E+00 1.309E+00 2.638E+00 1.309E+00 7.758E-01 2.442E-02 9.553E-04 5.746E-03 8.081E-01 1.326E+00

1981 6.965E+00 1.252E+00 7.687E+00 1.252E+00 3.472E+00 1.078E-01 4.293E-03 6.386E-02 3.632E+00 1.369E+00

1982 3.181E+01 1.383E+00 2.470E-01 1.383E+00 1.079E+01 5.449E-01 1.365E-02 2.008E-01 1.154E+01 1.386E+00

1983 6.020E+00 1.319E+00 2.672E-01 1.319E+00 6.073E+00 2.943E-01 7.642E-03 9.036E-02 6.465E+00 1.374E+00

1984 5.021E+00 1.221E+00 9.726E-01 1.221E+00 1.666E+00 4.760E-02 2.110E-03 1.190E-01 1.785E+00 1.274E+00
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 

09/08/2006 Page 13 of 28

Year 99-Tc GSD Uranium

daughters GSD U-234 (e) U-235 U-236 U-238 Total U GSD

1985 5.445E+00 1.216E+00 6.818E-01 1.216E+00 8.426E-01 4.837E-02 1.291E-03 2.177E-01 1.092E+00 1.180E+00

1986 6.895E+00 1.145E+00 1.594E+00 1.145E+00 1.154E+00 6.145E-02 1.719E-03 7.044E-01 2.098E+00 1.240E+00

1987 1.333E+01 1.300E+00 1.716E-01 1.300E+00 2.575E+00 8.454E-02 1.214E-04 2.226E-02 2.664E+00 1.335E+00

1988 2.113E+01 1.202E+00 2.752E-01 1.202E+00 5.328E+00 1.998E-01 1.998E-03 1.079E-01 5.594E+00 1.299E+00

1989 1.603E+01 1.327E+00 1.374E+00 1.327E+00 1.089E+00 4.156E-02 1.056E-03 2.284E-02 4.085E+00 1.353E+00

1990 2.356E+01 1.391E+00 2.018E+00 1.391E+00 3.113E-01 8.963E-03 4.881E-05 1.248E-02 1.776E+00 1.206E+00

1991 6.022E+01 1.302E+00 2.737E+00 1.302E+00 4.724E+00 1.384E-01 4.724E-04 2.092E-01 5.062E+00 1.299E+00

1992 1.332E+01 1.298E+00 8.880E-02 1.298E+00 1.228E+00 3.589E-01 1.133E-03 2.078E-01 1.322E+01 1.326E+00

1993 1.434E+02 1.316E+00 1.355E-01 1.316E+00 2.123E+00 5.586E-02 2.383E-04 1.080E-01 2.309E+00 1.266E+00

1994 5.329E+00 1.190E+00 8.737E-02 1.190E+00 5.231E-01 1.655E-02 1.606E-04 9.989E-03 7.104E-01 1.214E+00

1995 1.411E+01 1.000E+00 5.664E-01 1.000E+00 1.313E+00 3.847E-02 3.889E-04 3.459E-02 1.386E+00 1.198E+00

1996(b) 1.045E+01 1.000E+00 1.161E+00 1.000E+00 2.999E+00 1.077E-01 2.154E-03 2.468E-01 3.356E+00 1.208E+00

1997(b) 1.607E+01 1.316E+00 1.785E+00 1.316E+00 2.121E+00 9.193E-02 1.051E-03 6.963E-01 2.910E+00 1.219E+00

1998(b) 1.576E+02 1.000E+00 1.751E+01 1.000E+00 8.890E+00 3.186E-01 1.058E-02 6.138E-01 9.833E+00 1.361E+00

1999(c) 3.692E+01 1.421E+00 4.103E+00 1.421E+00 1.966E+01 7.011E-01 1.336E-02 2.266E+00 2.264E+01 1.304E+00

2000(c) 1.120E+01 1.000E+00 1.245E+00 1.000E+00 4.167E+00 1.486E-01 2.831E-03 4.801E-01 4.798E+00 1.287E+00

2001(d) 3.552E-01 1.318E+00 3.552E-02 1.318E+00 2.042E-01 8.702E-03 1.954E-03 7.637E-02 2.913E-01 — 

a. Average empirical Chi/Q value from 1964 through 1993 used to estimate the annual intake.

b. Applied release amounts as given in annual NESHAP reports 1995-1998 (ATL 2003h). 

c. Applied average fractions for isotopic releases as given in annual NESHAP reports 1995-1996 (ATL 2003h) 

d. Air sample measurements by radionuclide, not gross alpha or gross beta-gamma.

e. For the years 1955 and 1956, the potential intake of U-234 is based on an average fraction of 50% for U-234 activity toward the total uranium activity during the years 

1957 through 1962.
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 14 of 28

4.3

Workers are subjected to external doses from ambient radiation levels. Until September 1981, 

external gamma radiation levels were measured with a calibrated open-shield Geiger-Muller 

(GM) tube 3 feet above ground level. Beginning in September 1981, TLD readings of gamma 

radiation were taken. The minimum detection limit for the GM tube assembly is not known; 

however, in the period of 1972 through 1981, the minimum dose rate detected was 0.0048 

mrem per hour (DOE 1974). Information is not available on the TLD detection limit.

EXTERNAL DOSE

Between 1981 and 1992, TLDs used to measure onsite environmental gamma radiation levels 

contained calcium fluoride: dysprosium (CaF2:Dy) chips. Commercially available CaF2-based 

environmental dosimeters have a detection limit of 5 mR per month that the dose reconstructor 

can use for the detection limit of the TLDs used in the early 1990s (ICN 2003). These TLDs 

were used because of their sensitivity. 

Beginning in late 1992, PORTS began using TLDs containing lithium fluoride (LiF) chips to 

monitor ambient conditions. These TLDs had the same technology used to monitor worker 

doses (DOE 1996). TLDs used from 1981 through the mid-to-late 1990s were not calibrated 

to measure neutrons. 

TLDs were upgraded in the mid-to-late 1990s to be able to measure beta, gamma, and 

neutron radiation. These TLDs can differentiate dose between the different forms of radiation 

and between shallow and deep dose. Neutron dose is included in the deep dose (DOE 

1999a). The minimum limit of detection is 10 mrem for beta-gamma and 10 mrem for neutrons 

(ICN 2003).

Table 4-3 summarizes the history of the type of instruments used for ambient environmental 

monitoring.

Table 4-3. Instruments used for ambient environmental monitoring. 

Instrument Period of use Minimum detection level

GM tube 1955-1981 4.8 × 10-3 mR per hour

TLD (CAF2:Dy chip) 1981-1992 10 mR (based on current TLDs with CAF2:Dy chip)

TLD (LiF chip) 1992 – mid to late 1990s Unknown

ICN 760 TLD (LiF chip) Mid to late 1990s – present 10 mrem for beta-gamma

10 mrem for neutrons

Throughout the history of the site, environmental radiological measurements were taken at 

outdoor locations around Perimeter Road and at offsite locations to monitor public dose

(locations 3, 12, 24, 29, A36, A39, A40, and 874). Since 1998, ambient radiological conditions 

at six major facilities (X-7725, X-326, X-345, X-744G, X-745C, and X-745E) at the interior of 

the site have been recorded, in addition to those at the Perimeter Road locations, to assist in 

monitoring worker dose. Figure 4-1 shows the locations of the public and worker monitoring 

points. Table 4-4 includes a description of each location. The TLDs used to monitor facilities 

at the interior of the site are outside the buildings.

Data collected by TLDs used to monitor public dose were used to establish ambient 

environmental doses for most areas and most years because these data are all that is 

available to establish ambient environmental conditions inside the site boundary. Ambient 

environmental doses for workers were not monitored consistently until 1998 (see Table 4-5) 

because of the emphasis on monitoring public dose and because throughout much of PORTS 

history most workers were monitored with personal dosimetry. The ambient environmental 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 15 of 28

data for workers, listed in Table 4-5, do not provide enough information and are too specific to 

the buildings listed in the table for application over all years and all outdoor areas.

Figure 4-1. Approximate TLD locations. 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 16 of 28

Table 4-4. Environmental measurement locations. 

Designation

(Other designations used over the 

years for measurements taken in 

this vicinity)

Location

3 (39, X-230-J2, A35) Generally, south side of site, near south holding pond.

12 (PP933) Power pole 933 opposite East Access Road.

24 (PP906) Power pole 906 opposite North Access Road.

29 (10, PP862) Power pole 862 opposite main access road.

A36 (X-611) Vicinity of water treatment plant.

A39 (PP722) Power pole 722 at south end of Pike Ave. between X-231 oil 

biodegradation plots.

A40 (35, X-100) Vicinity of Guard headquarters.

PP518 Power pole 518 near X-104.

PP1406 (PP1404A) Near X-7725 and warehouses on west side of site.

874 On power pole 874 at northwest corner of X-745C Depleted Uranium 

Storage Yard.

X-7725 Waste Storage Facility.

X-326 Process Building.

X-330 Process Building.

X-333 Process Building.

X-344 Containment Building. 

X-345 Special Nuclear Material (SNM) Storage Building.

X-705 Decontamination Building.

X-720 Maintenance and Stores Building.

X-744G Bulk Storage Building.

X-745C, X-745E Depleted Uranium Cylinder Storage Yards.

Table 4-5. Site interior ambient radiological conditions. 

Mean annual dose by location(a), (b)

(mrem/year)

Year

X-7725 X-326 X-345 X-744G X-745C X-745E

Avg. 

deep 

dose

Avg. 

shallow 

dose

Avg. 

deep 

dose

Avg. 

shallow 

dose

Avg. 

deep 

dose

Avg. 

shallow 

dose

Avg. 

deep 

dose

Avg. 

shallow 

dose

Avg. 

deep 

dose

Avg. 

shallow 

dose

Avg. 

deep 

dose

Avg. 

shallow 

dose

1998 21 36 2 4 19 26 — — 35 39 20 21

1999 5 10 1 4 1 3 23 19 37 30 56 47

2000 14 — 0 — 0 — 285 — 122 — 178 — 

2001 23 — 0 — 0 — 1,056 — 142 — 175 — 

a. 2,000-hour work-year, prorated from an 8,736-hour year.

b. Includes beta, gamma, and neutron. Neutrons included in deep dose.

— No data.

The Perimeter Road TLD locations changed over the years. For some of these locations (described

in the previous paragraph), TLD readings might have been taken off the site during certain years, but 

are attributed to the particular location as the nearest point to the industrial area. For example, in the 

early years location 29 on the west side of the site was at the U.S. Highway 23 exit ramp of the main 

access road to the site. This location was later moved to the Perimeter Road/main access road 

intersection. This TBD assumes that all public dose readings were at the Perimeter Road.
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 17 of 28

4.3.1

The environmental radiological profile has been developed for PORTS for use by dose 

reconstructors when personal dosimetry or bioassay program participation was not 

required. Annual site environmental reports, health physics surveys, and other reports 

were reviewed for data that would be useful in reconstructing ambient radiation levels. 

Data in these documents (see reference section for citations) include TLD radiation 

measurements. Tables 4-5 and 4-6 list the GM and TLD results. The following 

sections discuss these tables further.

Ambient Radiation

Table 4-6. Perimeter road ambient radiological conditions from GM and TLD measurements. 

Mean annual dose by location(a), (b)

(mrem/year)

Year 874 PP518 

29

(10, PP862)

24 

(PP906) 

12

(PP933) PP1406

3 

(39,

X-230-J2, A35)

A40

(35,

(X104)

A36

(X611)

A39 

(PP722

) 

1954 24.0 24.0 24.0 24.0 24.0 24.0 24.0 24.0 24.0 24.0

1972 — — 23.6 23.8 22.4 — 24.0 — — — 

1973 — — 27.8 26.4 25.6 — 27.2 — — — 

1974 — — 28.2 27.0 26.0 — 26.8 — — — 

1975 — — 19.2 18.8 18.4 — 18.8 — — — 

1976 — — 20.6 20.4 20.0 — 20.0 — — — 

1977 — — 23.8 24.0 24.2 — 24.4 — — — 

1978 — — 21.8 21.4 20.6 — 21.2 — — — 

1979 — — 21.8 21.2 22.0 — 20.2 — — — 

1980 — — 21.4 20.8 20.8 — 21.6 — — — 

1981 — — 21.0 22.6 20.2 — 19.8 — — — 

1982 — — 14.4 15.6 14.6 — 16.2 17.4 — — 

1983 — — 15.6 12.8 13.4 — 15.2 15.8 — — 

1984 — — 12.4 6.4 12.8 — 20.4 11.4 — — 

1985 — — 13.8 13.8 13.8 — 13.8 13.8 — — 

1986 239.3 — 20.3 16.4 19.1 — 16.9 16.4 20.6 19.1

1987 267.5 — 21.6 20.0 21.4 — 18.9 19.4 24.5 22.8

1988 260.6 — 19.6 17.4 19.4 — 20.8 16.6 23.2 21.6

1989 257.2 — 22.7 20.8 20.7 — 20.7 17.4 22.5 22.7

1990 250.7 — 23.7 17.9 24.7 — 20.0 19.6 21.0 — 

1991 261.0 16.7 21.5 17.0 21.7 21.7 20.1 NR 20.9 — 

1992 246.5 29.1 36.6 17.4 19.9 44.8 33.7 NR 25.6 — 

1993 150.4 24.8 25.4 24.0 23.6 12.2 35.9 26.0 26.9 — 

1994 27.8 22.9 24.6 30.3 24.6 22.1 29.8 19.9 32.3 — 

1995 90.0 20.0 16.3 16.8 17.4 17.3 18.1 15.5 19.4 — 

1996 112.0 14.0 22.1 26.9 28.4 13.4 15.7 10.5 29.7 — 

1997 131.4 13.8 21.5 25.9 27.5 13.9 15.2 10.4 28.5 — 

1998 135.0 22.0 23.3 23.3 24.3 23.1 24.7 20.4 24.3 — 

1999 131.4 18.5 19.2 16.9 20.6 20.8 20.4 15.8 20.1 — 

2000 149.2 24.5 22.0 20.6 29.5 22.0 24.0 19.7 23.8 — 

2001 150.4 19.1 23.2 16.8 27.2 19.2 20.8 15.7 19.5 — 

2002 142.4 18.4 23.3 15.7 27.1 18.8 20.4 14.8 18.4 — 

a. 2,000-hour work-year, prorated from an 8,736-hour year.

b. Data for all years are from PORTS annual environmental reports with the exception of 1993 – 2002 for locations PP518, PP862, 

PP906, PP933, PP1406, A35, A40, A36 and 874, which were obtained from USEC by separate communication (ATL 2003f).

— No data.
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 18 of 28

Some ambient radiation levels for several locations for 1964-1966 are listed in Table 4-

7 from GM measurements (GAT 1965a, 1965b and 1967). TLD measurements are 

considered more accurate because the GM device takes more discrete readings and 

could miss fluctuations in gamma radiation levels (DOE 1982). 

Table 4-7. Perimeter road ambient 

radiological conditions from GM 

measurements. 

Mean annual dose by location(a)

(mrem/year)

Year 9 15

1964 20.6 18.6

1965 43.4 39.8

1966 48.0 70.0

a. 2,000-hour work-year. (GAT 1965a, 1965b 

and 1967). 

4.3.1.1

The ambient radiation measured by TLDs near Perimeter Road included 

natural background radiation, nuclear weapons testing fallout, and cosmic 

radiation (DOE 1994). These TLDs provided an indication of worker dose 

levels in the general proximity of Perimeter Road (with the exception of the 

cylinder storage yards), but not inside buildings. Table 4-6 lists annual dose

levels for locations near Perimeter Road. These data, available for 1954 

and 1972-2002, are representative of dose for a 2000-hour work-year. 

Background Radiation

Because of a lack of environmental dosimetry data prior to the year 1972, 

an extrapolation will be assumed based on the available data. To support 

the assumptions, it is necessary to determine if there is a correlation 

between either the production levels (in metric tons uranium (MTU)) and/or 

the accumulation of material (i.e., tails in the cylinder yards) at the PORTS 

site, and the available environmental dosimetry. The annual production 

levels are available in Appendix IV of the “Recycled Uranium Mass Balance 

Project, Portsmouth, Ohio Site Report” (BJC 2000). However, annual data 

on the accumulation of tails material in the cylinder yards does not appear 

to be publicly available. Instead, an assumption is made that a significant 

majority of the feed material processed at PORTS would eventually leave 

the enrichment cascade as tails material and placed into long-term storage 

in one of the site’s cylinder yards. Therefore, the cumulative quantity of 

feed material by year based on the data given in Appendix IV of the 

Recycled Uranium Mass Balance Project Report would be proportional to 

the accumulation of the tails material.

Figure 4-2 shows the cumulative quantity of feed material by year at 

PORTS versus the available environmental dosimetry data previously 

presented in Tables 4-5 and 4-6. This figure demonstrates that no direct 

relationship exists between the quantity of material accumulated at the 

PORTS site and the environmental dosimetry readings for the years 

available (after 1972). Figure 4-3 compares the same environmental 

dosimetry data to the annual MTU processed by PORTS from 1955 to 1997. 

This result would show if there is a potential for “building shine” as a 

contributor to external environmental radiation levels within the central area 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 19 of 28

of PORTS. It is also important to compare the site readings to distant 

reading locations (i.e., background radiation levels). If there is an effect 

caused by the quantity of material processed then it should be an additive 

value to the site’s background radiation levels. As shown in Figure 4-3, the 

environmental dosimetry data fluctuates more with the background radiation 

levels than the quantity of material processed. Therefore, the amount of 

feed stock processed per year (i.e., building shine or general radiation 

levels) does not significantly contribute to the environmental dose that all 

site workers would be expected to receive. It can therefore be assumed 

that the data presented in Tables 4-5 and 4-6 are representative of ambient 

radiological conditions for the entire period of 1954 to 2002.

In the past, personal dosimetry devices worn by site workers were stored in 

racks at the guard gates. Based on a 1989 study to review the practice of 

using storage racks, some employees were allowed to leave PORTS with 

their personal dosimetry. Beginning in 1999, all workers were allowed to 

leave PORTS with their personal dosimetry. Procedures require an 

exchange of dosimetry if the worker believes the dosimetry might have been 

inadvertently exposed away from PORTS due to dose from a medical exam, 

airport security system, or other sources (ATL 2003d).

Exposure to neutrons is not a factor in PORTS outdoor areas except in the 

uranium hexafluoride (UF6) cylinder storage yards (see Section 4.3.1.2). 

Other areas where neutron dose might be of concern are indoors (Cardarelli 

1997). 

 

0.00

50,000.00

100,000.00

150,000.00

200,000.00

250,000.00

300,000.00

350,000.00

195519571959196119631965196719691971197319751977197919811983198519871989199119931995199719992001

MTU

0.00

50.00

100.00

150.00

200.00

250.00

300.00

mrem

Cumulative MTU

X-7725

X-326

X-345

874

PP518

PP862

PP906

PP933

PP1406

A35

A40

A36

X-745C

X-745E

Bkgrd. Loc.

Cylinder Yards

Bkgrd. And Onsite Locations

Figure 4-2. Cumulative quantity of material by year 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 20 of 28 

0

5000

10000

15000

20000

25000

195519571959196119631965196719691971197319751977197919811983198519871989199119931995199719992001

MTU/yr.

0.00

50.00

100.00

150.00

200.00

250.00

300.00

mrem

X-7725

X-326

X-345

Total MTU/Yr.

874

PP518

PP862

PP906

PP933

PP1406

A35

A40

A36

X-745C

X-745E

Bkgrd. Loc.

Cylinder Yards

Figure 4-3. Annual MTU vs. work year doses (2000 hrs/yr) 

If background is not to be subtracted, the maximum value of 44.8 mrem 

from Table 4-6 can be used to assign annual ambient environmental dose 

to workers in areas near Perimeter Road such as the general employee 

parking lots; the guard gates on the outer perimeter of the security area; the 

switchyards; warehouses X-744S, T, and U; process buildings, and 

wastewater facility X-611. (The values in Table 4-6 for location 874 are 

much higher, but are specific to the UF6 cylinder storage yards and not to 

the rest of the facility.) If background is to be subtracted, the ambient 

radiation for these areas should be assigned a value of 0 mrem because of 

the argument presented above. 

The data in Table 4-5 are representative of ambient radiological conditions 

for the areas immediately surrounding X-7725, X-326, X-345, and 744G. 

These data are applicable only to 1998 through 2001, as specified in the 

table. It is difficult to apply the data in Table 4-5 for other years because of 

its variability and limited amount. The value of 44.8 mrem per year should 

be used for 1954 to 1997 outside these four buildings. For 744G, an outdoor 

ambient dose of more than 1,000 mrem for 2001 (2,000-hour work-year) is 

expected because of the increasing inventory of nuclear material stored in 

this facility (ATL 2003e). The data in Table 4-5 associated with X-745C and 

X-745E cylinder storage yards are discussed in Section 4.3.1.2.

4.3.1.2

Several TLDs were near the UF6 cylinder storage yards, so elevated 

ambient radiation levels are likely near the storage yards in comparison to 

other outdoor areas of the site. TLD 874, on a power pole on Perimeter 

Road at the X-745C cylinder storage yard, monitors public dose. The 

Ambient Radiological Conditions in Cylinder Storage Yards
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 21 of 28

average dose from 1986 to 2002 was 1,116 mrem per year with a maximum 

of 1,172 mrem over 24 hours and 365 days per year. This is equivalent to 

an average of 255 mrem and a maximum of 267 mrem for a 2,000-hour 

work-year, as indicated in Table 4-6 for TLD 874. An approximately 40% 

drop in dose readings at this location in 1993 was probably due to a change 

in the type of TLD monitor in late 1992 (this cannot be confirmed but it 

corresponds to a period in which TLDs were upgraded). The TLDs used to 

monitor this dose were not calibrated to detect neutrons until 1998. 

Because the TLD on the power pole was probably not within several feet of 

a UF6 cylinder but was used to monitor public dose, neutrons are not an 

exposure factor; the 10-ton cylinders produced radiation levels of 0.5 

mrem/hr neutron dose equivalent (DE) rate at the surface and would quickly 

decrease to background radiation levels due to distance (i.e., 1/r2 rule) 

(Cardarelli 1997). 

In comparison, Table 4-5 lists results for a different set of TLDs in cylinder 

yards X-745C and X-745E from 1998 to 2001. These TLDs are in the 

cylinder yards, not on the perimeter of the yards, to focus on worker dose. 

They were calibrated to measure neutrons, which are included in the values 

for deep dose equivalent. The average dose for the two cylinder yards over 

4 years was 130 mrem (deep plus shallow dose), with a maximum of 178

mrem. These values are less than those measured on the power pole.

In another study, the average radiation dose to cylinder yard workers from 

1990 to 1995 ranged from 55 to 196 mrem/year, based on personal 

dosimetry (DOE 1999b). These workers conducted activities in other site 

facilities wearing the same TLDs and probably did not work in the cylinder 

yards 2,000 hours each year.

Given the above information, an ambient radiation dose of 267 mrem (2,000 

hour work-year), which is claimant favorable, should be applied to cylinder 

yard workers. The portion of this dose attributable to neutrons can be 

assumed to be 178 mrem per year. This value for the annual neutron dose 

is obtained from Table 4-5 as the maximum annual deep dose equivalent 

reported in the period 1998 to 2001 for the cylinder storage yards X-745C 

and X-745E. This assumes that the neutron dose is the only contributor to 

the deep dose equivalent. All of these assumptions are claimant-favorable. 

4.3.1.3

Because all workers were badged and monitored throughout much of the 

PORTS operating history, coworker dose data can be used to assign dose 

to unmonitored workers for the periods during which all workers were not 

badged, or were badged but not analyzed. Other sources of information 

that describe potential radiation doses can be used as backup to coworker 

data.

Ambient Radiological Conditions Inside Buildings

In 1983, a survey by the Centers for Disease Control and Prevention (CDC) 

recorded area beta and gamma measurements for buildings X-326, X-330, 

X-333, X-344, X-705 and X-720. The highest gamma measurement was 

4,000 mrem per year. The highest median gamma measurement was 160 

mrem per year. Both of these measurements occurred in X-344 (CDC

1987). The 160-mrem median gamma measurement can be used to 
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 22 of 28

indicate conservative ambient conditions inside all buildings that contain

radiological materials or contamination. The highest gamma measurement 

of 4,000 mrem is not an external dose reading that could be considered 

reflective of overall operations and should not be assumed over an 

exposure period because personal dosimetry results do not indicate this 

type of exposure level at PORTS, according to the information from REMS 

(Section 4.3.1.2). 

4.4

As discussed in the previous sections, estimates of annual intakes employed conservative 

(i.e., claimant-favorable) assumptions. For example, applying maximum annual dose to all 

years is a claimant-favorable assumption. 

UNCERTAINTY

The locations of the monitoring points from which data are summarized in this TBD add 

uncertainty to the results. These points, as stated above, were around the site perimeter and 

off the site to monitor public dose. Until 1998, external environmental exposures were not 

monitored in relation to workers; that is, monitoring locations were not normally at the interior 

of the site among the process buildings. Because of data availability, however, public dose 

information had to be used for worker environmental doses. The maximum value of 

environmental dose is recommended for years for which data are unavailable to compensate 

for lack of worker-specific environmental dose information.

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

were originally reported in site environmental reports as representative of an employee who 

works at the site 24 hours per day, 365 days per year. Using a permanently located employee 

at the site, however, is an unrealistic assumption that might overstate onsite environmental 

doses.
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 23 of 28

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GLOSSARY 

alpha radiation

Radiation consisting of charged particles of the isotope helium-4, consisting of two 

protons and two neutrons. Alpha radiation is common for heavier nuclei.

background radiation

Background radiation is the radiation received that is not associated with a worker’s 

occupation. This includes cosmic, terrestrial, and man-made sources.

Becquerel

The derived SI unit of radioactivity equal to one disintegration per second.

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) of the 

direct fission products emit beta radiation. Physically, the beta particle is identical to 

an electron moving at high velocity.

deep dose equivalent

The dose equivalent at the respective depth of 1.0 cm in tissue.

derived concentration guidelines (DCG)

A calculated concentration that would result in an annual dose of 100 millirem.

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, 

etc., from external and/or internal sources of radiation.

exposure

As used in the technical sense, exposure refers to a measure expressed in roentgens 

(R) of the ionization produced by photon radiation (i.e., gamma and X-rays) in air.

film badge

In general, a “film packet” that contains one or more pieces of film in a light-tight 

wrapping. The film when developed has an image caused by radiation that can be 

measured using an optical densitometer.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying 

many nuclear reactions (e.g., fission, radioactive decay, and neutron capture). 

Physically, gamma rays are identical to X-rays of high energy, the only essential 

difference being that X-rays do not originate in the nucleus.

neutron

A basic particle that is electrically neutral, having nearly the same mass as the 

hydrogen atom.

radiation

Alpha, beta, neutron, and photon radiation.

radiation exposure monitoring system (REMS)

An online DOE database that contains records of personal dosimetry information.
Document No. ORAUT-TKBS-0015-4 Revision No. 01 Effective Date: 09/08/2006 Page 28 of 28

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, 

and neutrons from unstable nuclei.

rem

A unit of dose equivalent, which is equal to the product of the number of rad absorbed 

and the “quality factor.”

shallow dose equivalent

The dose equivalent at the respective depth of 0.07 mm of the skin or an extremity.

thermoluminescent dosimeter (TLD)

A holder containing solid chips of material that when heated release the stored energy 

as light. The measurement of this light provides a measurement of absorbed dose.

uranium hexafluoride (UF6) cylinder storage yard

Storage yards at the Paducah and Portsmouth Gaseous Diffusion Plants and Oak 

Ridge K-25 site. These yards maintain cylinders containing depleted UF6. The 

cylinders, which typically weigh 10 and 14 tons, contain depleted UF6 primarily in a 

solid form.