ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 26

DOE Review Release 09/16/2009

Document Title:

Portsmouth Gaseous Diffusion Plant – 

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0015-4 

Revision: 02

Effective Date: 08/24/2009 

Type of Document: TBD

Supersedes: Revision 01

Subject Expert(s): Paul J. Demopoulos, Milton Gorden, and Donald Palmrose

Site Expert(s): N/A

Approval: Signature on File Approval Date: 08/13/2009

Susan L. Winslow, Document Owner

Approval: Richard E. Merrill Signature on File for Approval Date: 08/14/2009

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 08/14/2009

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 08/21/2009

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/24/2009

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. 02 Effective Date: 08/24/2009 Page 2 of 26

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.

08/24/2009 02 Changes were made to the document as per OCAS comments. The 

Uranium daughter intake values were corrected in Table 4-2. 

Footnote f was added: “Uranium progeny intakes calculated from 

recorded or estimated equilibrium fractions with the Uranium parent 

radionuclides.” Changed the term, “claimant favorable” to “favorable 

to claimant(s)” as appropriate throughout the document. 

Incorporates formal internal and NIOSH review comments, such as 

revising wording of footnote “e” in Table 4-2. Equation 4-1 for the 

geometric standard deviation (GSD) has been deleted along with 

citations to the GSD in the text and in Table 4-2. This follows 

ORAUT-PROC-0031 guidance. Deleted Table 4-7 due to the data 

being limited and not adding value. Revised Section 4.2.3 to provide 

a better description of the calculations. Training required: As 

determined by the Objective Manger. Initiated by Susan L. Winslow.
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 3 of 26

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.................................................................................................................7

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 .......................................................................................9

4.2.5 Estimation of Annual Intake from Airborne Radionuclides ........................................... 10

4.2.6 Contribution of Airborne Radionuclides to External Dose............................................ 10

4.3 External Dose ........................................................................................................................12

4.3.1 Ambient Radiation ...................................................................................................... 13

4.3.1.1 Background Radiation.................................................................................. 16

4.3.1.2 Ambient Radiological Conditions in Cylinder Storage Yards......................... 19

4.3.1.3 Ambient Radiological Conditions Inside Buildings ........................................ 20

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

4.5 Attributions and Annotations .................................................................................................. 20

References ........................................................................................................................................21

Glossary ............................................................................................................................................25

LIST OF TABLES

TABLE TITLE

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

PAGE

7

4-2 Maximum annual intakes (Bq per year).................................................................................. 11

4-3 Instruments used for ambient environmental monitoring ........................................................ 13

4-4 Environmental measurement locationsa ................................................................................. 15

4-5 Site interior ambient radiological conditionsa .......................................................................... 15

4-6 Perimeter road ambient radiological conditions from GM and TLD measurements ................ 16

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Approximate TLD locations (ATL 2003f).................................................................................

 

14

4-2 Cumulative quantity of material by year.................................................................................. 18

4-3 Annual MTU vs. work year doses (2000 hrs/yr)...................................................................... 18
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 4 of 26

ACRONYMS AND ABBREVIATIONS

Bq Becquerel

CAF2:Dy calcium fluoride:dysprosium

Ci curie

cm centimeter

DOE U.S. Department of Energy

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EPA U.S. Environmental Protection Agency

ft foot

GAT Goodyear Atomic Corporation

g grams

GM Geiger-Muller (tube assembly)

hr hour

LiF lithium fluoride

m3 cubic meters

mo month

mR milliRoentgen

mrem millirem

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

POC probability of causation

REMS Radiation Exposure Monitoring System

TBD technical basis document

TLD thermoluminescent dosimetry

UF6 uranium hexafluoride 

USEC United States Enrichment Corporation

yr year
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 5 of 26

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 facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 [(42 U.S.C. Sections 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

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 2007): 

] 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) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007). 

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

conventional structures

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 6 of 26

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

reasons

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 Portsmouth Gaseous Diffusion Plant (PORTS). This document 

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

internal exposure data for PORTS workers.

Purpose

This 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). 

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 assumptions favorable to the claimant for worker external and internal dose 

assessments. 

Scope

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.

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATION

4.2.1

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

ambient air samples (DOE 1994). Sampling data are available for five offsite locations (one several 

miles from PORTS to provide background measurements) beginning in 1964 and for four onsite 

locations beginning in 1974 (Goslow 1986). 

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 DOE 

Derived Concentration Guidelines, the U.S. Environmental Protection Agency (EPA), the State of 

Ohio Environmental Protection Agency, or (as of 1997) U.S. 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 document considers only those ambient 

air monitoring locations at or within the PORTS site boundary for worker internal dose. Table 4-1 lists 

these air monitors and their locations.
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 7 of 26

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 (Goodyear Atomic Corporation) 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 (1988). 

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 

m3 of air. Depending on the set flow rate (1.2 or 0.3 m3

/hr), 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 beta-gamma 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. 

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 yr of operations and not all radionuclides of 

concern were originally measured in the air effluent release streams. Therefore, this TBD discusses a 

methodology favorable to the claimant 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 favorable to the claimant. 

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
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 8 of 26

• 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. 

The methodology for determining internal dose from onsite atmospheric radionuclide concentrations 

applies directly to the available air sampling data. To be favorable to the claimant for each year for

which data are available, the maximum site measurement for gross alpha (uranium) and gross betagamma (uranium daughter products and 99Tc) will be applied for the entire site. 

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 

at that time, 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 (Goslow

1986). 

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 4-1. The maximum uranium/alpha 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

Re )(

)( )(

3

3

Annual Radionuclide Activity lease Ci

Maximum Radionucli Airde Concentration Ci m Chi mQ

−

− = (4-1) 

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). The available annual 

environmental reports were reviewed and the highest air concentration (alpha or uranium) was used 

to determine the Chi/Q value. The average Chi/Q value and annual radionuclide emissions (Goslow 

1986) were used to calculate the radionuclide concentration for the years where ambient air 

monitoring data was not available. For example following Equation 4-1, the highest alpha air 

concentration for 1974 is 2.56 x 10-13 Ci m-3 (DOE 1975) divided by the annual radionuclide activity 

release for 1974 of 0.023 Ci (Goslow 1986) resulting in 1.11 x 10-11 m-3. This method was used for 

each year where environmental data is available. An average Chi/Q value was then calculated for 

1964 through 1993. This average Chi/Q value was applied to the annual uranium emissions for each 

year prior to 1964 factoring in an assumed individual ventilation rate of 2,400 m3 per year to estimate 

the maximum annual uranium intakes provided in Table 4-2.
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 9 of 26

4.2.4

This document 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 yr 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 environmental concern were 231Th, 234Th, and 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 (Goslow 1986). 

Also, since the enrichment cascade was shutdown in 1993, if the quantity of uranium daughters 

released were not specifically given in the 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 that the 

uranium and its daughters would exist together and, in accordance with their respective decay halflives, 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 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 20 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 

Airborne Release Estimation for Neptunium, Plutonium, Uranium Daughters, and 99Tc
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 10 of 26

the environmental monitoring program were the uranium daughters in 1967 and 99Tc in 1976 (ATL 

2003d). However, there could be a concern for releases of transuranics from the processing of 

recycled uranium and for the releases of uranium daughters prior to 1967 and for 99Tc prior to 1976

(BJC 2000). This section investigates the potential for these unknown releases and how to include 

their contribution for the environmental occupational dose.

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 g 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 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 

favorable to the claimant 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 2003d). 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 document used an assumed individual ventilation rate of 2,400 m3

/yr to derive 

annual intakes that are favorable to the claimant for the radionuclides of concern. Table 4-2 lists 

annual intakes 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 

Contribution of Airborne Radionuclides to External Dose
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 11 of 26

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 conditions most favorable to the 

claimant. 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 (Goslow 1986). Therefore, airborne radioactivity at the PORTS site over all years of 

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

Table 4-2. Maximum annual intakes (Bq per year). 

Year 99Tc Uranium 

progeny(f)

234U 235U 236U 238U Total U

1953 — — — — — — — 

1954 — — — — — — — 

1955(a)

(e)

0.000E+00

6.553E+01 6.011E+01

2.706E+00 1.422E-01 1.174E+02

1.803E+02

1956(a)

(e)

0.000E+00

2.832E+01 2.598E+01

8.139E-01 6.137E-02 5.108E+01

7.793E+01

1957(a) 0.000E+00 1.745E+00 1.100E+00 1.320E-01 5.721E-03 3.563E+00 4.801E+00

1958(a) 0.000E+00 1.461E+01 3.497E+01 1.540E+00 4.732E-02 3.629E+00 4.019E+01

1959(a) 0.000E+00 3.609E+01 4.289E+01 2.794E+00 1.175E-01 5.349E+01 9.930E+01

1960(a) 0.000E+00 1.389E+01 1.540E+01 1.078E+00 4.498E-02 2.171E+01 3.823E+01

1961(a) 0.000E+00 2.769E+01 3.277E+01 2.222E+00 9.023E-02 4.111E+01 7.620E+01

1962(a) 0.000E+00 1.067E+01 1.628E+01 9.018E-01 2.938E-02 1.214E+01 2.935E+01

1963(a) 0.000E+00 1.233E+00 3.079E+00 2.860E-01 4.160E-03 2.200E-02 3.392E+00

1964 0.000E+00 6.454E+00 1.579E+01 1.283E+00 2.099E-02 9.867E-02 1.776E+01

1965 0.000E+00 2.582E+01 5.751E+01 3.890E+00 8.398E-02 8.288E+00 7.104E+01

1966 0.000E+00 3.227E+00 7.804E+00 5.651E-01 1.050E-02 2.153E-01 8.880E+00

1967 0.000E+00 3.108E-01 7.992E+00 3.552E-01 1.050E-02 4.884E-01 8.880E+00

1968 0.000E+00 1.480E-01 7.400E+00 2.960E-01 1.050E-02 1.184E+00 8.880E+00

1969 0.000E+00 1.696E-01 1.830E+00 2.410E-01 1.050E-02 6.809E+00 8.880E+00

1970 0.000E+00 2.029E-01 3.102E+00 1.432E-01 4.514E-03 5.966E-01 3.818E+00

1971 0.000E+00 1.614E+00 2.992E+00 1.834E-01 5.249E-03 1.207E+00 4.440E+00

1972 0.000E+00 5.486E-01 8.626E-01 6.470E-02 1.785E-03 5.823E-01 1.510E+00

1973 0.000E+00 4.980E-02 2.037E+00 9.960E-02 2.729E-03 9.960E-02 2.309E+00

1974 0.000E+00 1.977E-01 1.779E+01 9.884E-01 2.687E-02 4.448E+00 2.273E+01

1975 0.000E+00 4.098E+00 1.009E+01 3.759E-01 1.333E-02 7.170E-01 1.128E+01

1976 3.609E+01 8.216E-05 8.052E+00 3.779E-01 1.039E-02 3.862E-01 8.791E+00

1977 2.559E+01 7.057E-01 2.001E+00 7.773E-02 2.729E-03 2.293E-01 2.309E+00

1978 2.429E+01 3.535E-02 4.778E-01 3.341E-02 1.480E-03 7.396E-01 1.252E+00

1979 5.479E+00 1.133E+00 7.536E-01 2.270E-02 9.553E-04 2.724E-02 8.081E-01

1980 6.864E+00 2.898E-01 7.758E-01 2.442E-02 9.553E-04 5.746E-03 8.081E-01

1981 6.965E+00 4.757E+00 3.472E+00 1.078E-01 4.293E-03 6.386E-02 3.632E+00

1982 3.181E+01 3.090E+00 1.079E+01 5.449E-01 1.365E-02 2.008E-01 1.154E+01

1983 6.020E+00 1.654E-01 6.073E+00 2.943E-01 7.642E-03 9.036E-02 6.465E+00

1984 5.021E+00 2.927E+00 1.666E+00 4.760E-02 2.110E-03 1.190E-01 1.785E+00
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 12 of 26

Year 99Tc Uranium 

progeny(f)

234U 235U 236U 238U Total U

1985 5.445E+00 6.007E-01 8.426E-01 4.837E-02 1.291E-03 2.177E-01 1.092E+00

1986 6.895E+00 1.409E+00 1.154E+00 6.145E-02 1.719E-03 7.044E-01 2.098E+00

1987 1.333E+01 1.288E-01 2.575E+00 8.454E-02 1.214E-04 2.226E-02 2.664E+00

1988 2.113E+01 2.811E-01 5.328E+00 1.998E-01 1.998E-03 1.079E-01 5.594E+00

1989 1.603E+01 3.079E-01 1.089E+00 4.156E-02 1.056E-03 2.284E-02 4.085E+00

1990 2.356E+01 1.811E-01 3.113E-01 8.963E-03 4.881E-05 1.248E-02 1.776E+00

1991 6.022E+01 6.749E-01 4.724E+00 1.384E-01 4.724E-04 2.092E-01 5.062E+00

1992 1.332E+01 7.745E-01 1.228E+00 3.589E-01 1.133E-03 2.078E-01 1.322E+01

1993 1.434E+02 2.756E-01 2.123E+00 5.586E-02 2.383E-04 1.080E-01 2.309E+00

1994 5.329E+00 4.305E-02 5.231E-01 1.655E-02 1.606E-04 9.989E-03 7.104E-01

1995 1.411E+01 8.431E-02 1.313E+00 3.847E-02 3.889E-04 3.459E-02 1.386E+00

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

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

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

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

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

2001(d) 3.552E-01 3.552E-02 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 2003e). 

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

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

e. For the years 1955 and 1956, the potential intake of 234U is based on a calculated fraction of 0.5 of 234U compared to 

theTotal Uranium activity between 1957 through 1962. The Total U was then increased by 50% in order to maintain the 

concentrations of 235U and 238U.

f. Uranium progeny intakes calculated from recorded or estimated equilibrium fractions with the Uranium parent 

radionuclides. 

4.3 EXTERNAL DOSE

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 ft 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/hr (DOE 1974). Information is 

not available on the TLD detection limit.

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 10 mR/mo, however, no particular reference for 

Portsmouth CaF2 TLDs identifying the detection level have been identified. These TLDs were used 

because of their sensitivity (ICN 2003). 

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).
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 13 of 26

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/hr

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 (X7725, 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 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.

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.

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
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 14 of 26

Figure 4-1. Approximate TLD locations (ATL 2003f).
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 15 of 26

Table 4-4. Environmental measurement locationsa. 

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.

a. Source: ATL (2003f).

Table 4-5. Site interior ambient radiological conditionsa. 

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

(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. Sources: DOE (1999a), DOE (2001a), DOE (2001b), and DOE (2002).

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

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

 — No data.
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 16 of 26

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,

(X-104)

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.

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-hr work-year. 

Background Radiation
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 17 of 26

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 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 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, 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 2003g).

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). 
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 18 of 26

 

0.00

50,000.00

100,000.00

150,000.00

200,000.00

250,000.00

300,000.00

350,000.00

19551957

1959

1961

196319651967

1969

1971197319751977197919811983198519871989199119931995199719992001

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 

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 
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 19 of 26

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-hr work-year) is expected because of the increasing inventory of nuclear material stored in this 

facility (ATL 2003h). 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 average dose from 

1986 to 2002 was 1,116 mrem/yr with a maximum of 1,172 mrem over 24 hr and 365 d/yr. This is 

equivalent to an average of 255 mrem and a maximum of 267 mrem for a 2,000-hr work-year as 

indicated in Table 4-6 for TLD 874. An approximate 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 rate at the surface and would quickly decrease to background radiation levels due to 

distance (i.e., 1/r2 rule) (Cardarelli 1997). 

Ambient Radiological Conditions in Cylinder Storage Yards

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 yr 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 document, the average radiation dose to cylinder yard workers from 1990 to 1995 ranged 

from 55 to 196 mrem/yr 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 hr each year.

Given the above information, an ambient radiation dose of 267 mrem (2,000 hr work-year), which is 

favorable to the claimant, 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 favorable to the 

claimant. 
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 20 of 26

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 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/yr. The highest median gamma measurement was 160 mrem/yr. 

Both of these measurements occurred in X-344 (CDC 1987). The 160-mrem median gamma 

measurement can be used to 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 the Radiation Exposure Monitoring 

System (REMS) (Section 4.3.1.2).

4.4 UNCERTAINTY

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

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

years is an assumption favorable to the claimant. 

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-hr work-year. The data were 

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

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

assumption that might overstate onsite environmental doses.

4.5 ATTRIBUTIONS AND ANNOTATIONS 

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

section of this document.
Document No. ORAUT-TKBS-0015-4 Revision No. 02 Effective Date: 08/24/2009 Page 21 of 26

REFERENCES

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Maryland, September 2. [SRDB Ref ID: 8477]

ATL (Advanced Technologies and Laboratories International, Inc.), 2003c, “Portsmouth High Volume 

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DOE (U.S. Department of Energy), 1975, Portsmouth Gaseous Diffusion Plant Environmental 

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Monitoring Report for Calendar Year 1983, GAT-4012, Goodyear Atomic Corporation, Piketon, 

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DOE (U.S. Department of Energy), 1988, Portsmouth Gaseous Diffusion Plant Site Environmental 

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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

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 

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

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. 02 Effective Date: 08/24/2009 Page 26 of 26

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.