ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Portsmouth Gaseous Diffusion 

Plant – Occupational Environmental Dose

Document Number:

ORAUT-TKBS-0015-4 

Effective Date: 03/17/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 26

Subject Expert: Donald Palmrose

Document Owner

Approval: Signature on File Date:03/16/2004 Mark Notich, TBD Team Leader

Approval: Signature on File Date: 03/16/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 03/17/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 03/17/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................4

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

4.2 Internal Dose From Onsite Atmospheric Radionuclide Concentration ......................................... 5

4.2.1 Ambient Air Sample Collection Network .......................................................................... 5

4.2.2 Methodology....................................................................................................................6

4.2.3 Estimation of Ambient Airborne Radionuclide Concentrations Prior to 1964.................... 7

4.2.4 Data for Ambient Airborne Radionuclide Concentrations and Annual 

Airborne Releases...........................................................................................................7

4.2.4.1 Airborne Release Estimation for Neptunium, Plutonium, Uranium 

Daughters, and 99Tc.............................................................................................8

4.2.5 Estimation of Annual Intake from Airborne Radionuclides ............................................... 9

4.2.6 Contribution of Airborne Radionuclides to External Dose .............................................. 12

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

4.3.1 Ambient Radiation ......................................................................................................... 16

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.............................................. 19

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

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

Glossary ............................................................................................................................................25
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LIST OF TABLES

Table

4.2.1-1 Ambient Air Sampling Station Locations..........................................................................

Page

6

4.2.5-1 Maximum Annual Intakes and Geometric Standard Deviations (Bq per year)................ 10

4.3-1 Instruments Used for Ambient Environmental Monitoring .............................................. 13

4.3-2 Environmental Measurement Locations......................................................................... 13

4.3-3 Site Interior Ambient Radiological Conditions ................................................................ 14

4.3.1-1 Perimeter Road Ambient Radiological Conditions ......................................................... 16

LIST OF FIGURES

Figure

4.3-1 Approximate TLD Locations ..........................................................................................

Page

15

4.3.1.1-1 Cumulative Quantity of Material by Year........................................................................ 18

4.3.1.1-2 Annual MTU vs. Work Year Doses (2000 hrs/yr) ........................................................... 18
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 10/30/2003 00-A New Technical Basis Document for the 

Portsmouth Gaseous Diffusion Plant - 

Occupational Environmental Dose. Initiated by 

Mark D. Notich.

Draft 12/31/2003 00-B Incorporates NIOSH review comments. Initiated 

by Mark D. Notich.

Draft 02/19/2004 00-C Incorporates additional NIOSH review comments. 

Initiated by Mark D. Notich.

03/17/2004 03/17/2004 00 First approved issue. Initiated by Mark D. Notich.
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 4 of 26

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

Bq Becquerel

CAF2:Dy calcium fluoride:dysprosium

Ci curie

cm centimeter

DAC Direct Air Contact

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
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4.1 INTRODUCTION

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). 

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 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 

U.S. Department of Energy (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 the PORTS site boundary for worker 

internal dose. Table 4.2.1-1 lists these air monitors and their locations. 

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 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 
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 6 of 26

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.

Table 4.2.1-1. 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).

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 claimant-favorable 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:
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 7 of 26





 





















= 655851

1

Maximum Annual

95th percentile GSD .

(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 241Th 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 direction 

relationship of air concentration to material release. The empirical relationship is described by 

Equation 4-2. The annual average uranium air concentration measured at the air monitoring station

with the maximum sampling measurement 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

)(

)( )( Annual Radionuclide ReActivity lease CiMaximum Radionuclide AirConcentrat mCiion mQChi

3

3

−

− = (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 4.2.1-1 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
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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 potentially 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 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 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 

parent uranium 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 Airborne Release Estimation for Neptunium, Plutonium, Uranium Daughters, and 99Tc

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.

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 radonuclides would be 
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 9 of 26

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 

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 (outfall 002) 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 4.2.5-1 lists annual intakes and GSDs 

for each radionuclide of concern for each year.

Estimation of Annual Intake from Airborne Radionuclides
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Table 4.2.5-1. 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

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
Effective Date: 

03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 11 of 26

Table 4.2.5-1. (Continued)

Year 99-Tc GSD

Uranium 

daughters GSD U-234 U-235 U-236 U-238 Total U GSD

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 (USEC 1996, 1997, 1998, 1999)

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

e. For the years of 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.
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 12 of 26

4.2.6

From the data shown in Table 4.2.5-1, 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 241Th, 234Th, and 234mPa. As previously 

stated, the contribution of each radionuclide is 6 percent of the uranium daughter fraction for 241Th 

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

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 and solid-state scalar tubes 3 feet above ground level. Beginning in September 1981, TLD

readings of gamma radiation were taken and compared to GM measurements from earlier in the year. 

TLD readings were considered more accurate than GM measurements because the GM device takes 

more discrete readings and could miss fluctuations in gamma radiation levels (DOE 1982). 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). Lacking the 

technical specifications for the GM tube, this value can be used as a claimant-favorable minimum 

detection limit. 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 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 4.3-1 summarizes the history of the type of instruments used for ambient environmental 

monitoring.

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, 
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 13 of 26

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 4.3-1 

shows the locations of the public and worker monitoring points. Table 4.3-2 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 4.3-3) 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 4.3-3, 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.

Table 4.3-1. Instruments Used for Ambient Environmental Monitoring. 

Instrument Period of use Minimum detection level

GM tube 1955-1981 1.0 × 10-4 mR per hour

TLD (CAF2:Dy chip) 1981-1992 5 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

Table 4.3-2. 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.
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 14 of 26

Designation

(Other designations used over the years for 

measurements taken in this vicinity) Location

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 4.3-3. 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.
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 15 of 26

Figure 4.3-1. Approximate TLD Locations. 

3 σ 

12 σ 

24 σ 

24 σ 

A36 σ 

A39 σ 

A40 σ 

PP518 σ 

PP1406 σ

874 σ 

φ

φ

φ

φ

φ

φ

X-344

X-705

X-720

X-744G
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 16 of 26

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

when personal dosimetry or bioassay program participation is 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 4.3-3 and 4.3.1-1 list the TLD results. The following 

sections discuss these tables further.

Ambient Radiation

Table 4.3.1-1. Perimeter Road Ambient Radiological Conditions. 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 

(X-611)

A39 

(PP722)

1954 23.8 23.8 23.8 23.8 23.8 23.8 23.8 23.8 23.8 23.8

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 Background Radiation

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 4.3.1-1 lists annual dose levels for 
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 17 of 26

locations near Perimeter Road. These data, available for 1954 and 1972-2002, are representative of 

dose for a 2000-hour work-year. 

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 4.3.1.1-1 shows the cumulative quantity of feed material by year at PORTS versus the 

available environmental dosimetry data previously presented in Tables 4.3-3 and 4.3.1-1. 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 

4.3.1.1-2 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 then it should be an additive value to the site’s 

background radiation levels. As shown in Figure 4.3.1.1-2, 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 4.3-3 and 4.3.1-1 

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). 

If background is not to be subtracted, the maximum value of 35.9 mrem from Table 4.3.1-1 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 4.3.1-1 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. 
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 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

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 4.3.1.1-1. 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 4.3.1.1-2. Annual MTU vs. Work Year Doses (2000 hrs/yr) 
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 19 of 26

The data in Table 4.3-3 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 4.3-3 for other years 

because of its variability and limited amount. The value of 35.9 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 4.3-3 associated with X-745C and X745E cylinder storage yards are discussed in Section 4.3.1.2.

4.3.1.2 Ambient Radiological Conditions in Cylinder Storage Yards

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 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 4.3.1-1 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 4.3-3 lists results for a different set of TLDs in cylinder yards X-745C and X745E 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) should 

be applied to cylinder yard workers as claimant-favorable. 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 4.3-3 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 Ambient Radiological Conditions Inside Buildings

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.
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 20 of 26

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 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 UNCERTAINTY

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. 

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.
Effective Date: 03/17/2004 Revision No.: 00 Document No. ORAUT-TKBS-0015-4 Page 21 of 26

REFERENCES

ATL (Advanced Technologies and Laboratories International, Inc.), 2003a, “Film Badge Program 

Procedures,” Memorandum to File, Germantown, Maryland, October 22.

ATL (Advanced Technologies and Laboratories International, Inc.), 2003b, “Application of Empirical 

Chi/Q When Ambient Air Samples Are Not Available,” Memorandum to File, Germantown, 

Maryland, September 2.

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

Ambient Air Sampling Data for 1993 through 2000,” Memorandum to File, Germantown, 

Maryland, October 20.

ATL (Advanced Technologies and Laboratories International, Inc.), 2003d, “Dosimetry Badging During 

Non-Work Hours,” Memorandum to File, Germantown, Maryland, October 20.

ATL (Advanced Technologies and Laboratories International, Inc.), 2003e, “Environmental Dose at 

744-G and Treatment of Background,” Memorandum to File, Germantown, Maryland, October 

20.

ATL (Advanced Technologies and Laboratories International, Inc.), 2003f, “Quarterly External Gamma 

Radiation Levels for 1993 to 2002,” Memorandum to File, Germantown, Maryland, October 21.

ATL (Advanced Technologies and Laboratories International, Inc.), 2003g, “Historical Information 

Concerning the Release of 99Tc from PORTS Prior to 1975,” Memorandum to File, 

Germantown, Maryland, December 1.

ATL (Advanced Technologies and Laboratories International, Inc.), 2003h, “USEC NESHAP Reports 

for the Years 1995 through 2002,” Memorandum to File, Germantown, Maryland, December 1.

BJC (Bechtel Jacobs Company LLC) 2000, Recycled Uranium Mass Balance Project, Portsmouth, 

Ohio Site Report, BJC/PORTS 139/R1, U. S. Department of Energy, Office of Environmental 

Management, Washington, DC

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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.
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radiation exposure monitoring system (REMS)

An online DOE database that contains records of personal dosimetry information.

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.