ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for Paducah Gaseous Diffusion 

Plant – Occupational Internal Dose

Document Number:

ORAUT-TKBS-0019-5

Effective Date: 09/30/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 26

Subject Expert: Carol D. Berger

Document Owner

Approval: Signature on File Date: 09/22/2004

Jay J. Maisler, TBD Team Leader

Approval: Signature on File Date: 09/20/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 09/20/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 09/30/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

3

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

5.1 Introduction.................................................................................................................................5

5.2 Source Term...............................................................................................................................6

5.3 In Vitro Measurement Methods ................................................................................................. 13

5.3.1 Measurement Types and Detection Levels ........................................................................ 14

5.3.2 Reporting Formats and Codes........................................................................................... 15

5.3.3 Instructions for Addressing Possible Interferences and Uncertainties ................................ 15

5.4 In Vivo Measurement Methods.................................................................................................. 18

5.4.1 Measurement Types and Detection Levels ........................................................................ 18

5.4.2 Reporting Formats and Codes........................................................................................... 18

5.4.3 Instructions for Addressing Possible Interferences and Uncertainties ................................ 18

5.4.4 Assessment of Intake for Monitored Employees ................................................................ 19

5.5 Significant Incidents with Internal Dose Potential ...................................................................... 19

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

Glossary ............................................................................................................................................24
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 2 of 26

LIST OF TABLES

Table

5-1 Elemental fractions of uranium, neptunium, and plutonium at specific facilities ........................

Page

6

5-2 Isotopic fractions for various enrichment percentages.............................................................. 7

5-3 Transuranic element concentrations in plant aerosols (1987)................................................... 7

5-4 Default isotopic distribution ......................................................................................................7

5-5 Facility-specific source radionuclides with solubility type and particle size ............................... 8

5-6 In vitro measurement frequencies .......................................................................................... 13

5-7 In vitro measurement types and detection levels for various periods...................................... 14

5-8 Nominal (default) detection levels for urine bioassay.............................................................. 15

5-9 In vitro record codes...............................................................................................................16

5-10 In vivo measurement types and detection levels for various periods ...................................... 18

5-11 In vivo record codes...............................................................................................................19

5-12 Input parameters for significant incidents and events ............................................................. 20
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 11/24/2003 00-A New technical basis document for the Paducah 

Gaseous Diffusion Plant – Occupational Internal 

Dose. Initiated by Jay J. Maisler.

Draft 07/21/2004 00-B Incorporates internal and NIOSH comments and 

Gaseous Diffusion Plant consistency review 

issues. Initiated by Jay J. Maisler.

Draft 08/11/2004 00-C Incorporates internal and NIOSH review 

comments. Initiated by Jay J. Maisler.

Draft 08/30/2004 00-D Incorporates internal and NIOSH review 

comments. Initiated by Jay J. Maisler.

Draft 09/03/2004 00-E Incorporates internal and NIOSH review 

comments. Initiated by Jay J. Maisler.

Draft 09/03/2004 00-F Incorporates internal and NIOSH review 

comments. Initiated by Jay J. Maisler.

09/30/2004 09/30/2004 00 First approved issue. Initiated by Jay J. Maisler.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 4 of 26

ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

Bq becquerel

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

g gram

hr hour

ICRP International Commission on Radiological Protection

KPA kinetic phosphorescence analysis

L liter

MDA minimum detectable amount

MDC minimum detectable concentration

mg milligram

ml milliliter

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

ORNL Oak Ridge National Laboratory

pCi picocurie

PGDP Paducah Gaseous Diffusion Plant

ppm parts per million

TRU transuranic

U.S.C. United States Code

μCi microcurie

μg microgram

μm micrometer
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5.1 INTRODUCTION

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning 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 the National Institute for Occupational Safety and Health 

(NIOSH) 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 [EEOICPA; 42 U.S.C. Sections 7384l(5) and 

(12)].

This document provides a uniform and consistent approach to assessing occupational internal dose at 

the Paducah Gaseous Diffusion Plant (PGDP) for the dose reconstructions for NIOSH in relation to 

the EEOICPA. The document provides guidance to dose reconstructors on input parameters that are 

specific to employees of PGDP, as well as the approach for employees with either missing or no 

monitoring information.

In 1951, the first operator of the facility, Union Carbide, began hiring and construction began. In 1952, 

the first four process buildings began operations. From then until 1977, uranium hexafluoride (UF6) 

feed material was produced from uranium trioxide (UO3) at the plant. From 1953 to 1964, and then 

again from 1968 to 1977, UF6 was produced from the UO3 in spent reactor fuel. In May of 1977, the 

feed plants ceased operation and all feed to the enrichment process was in the form of UF6 obtained 

from outside sources. 

Over the years, workers at PGDP handled mainly UF6 and slightly oxidized forms of uranium. The 

facility processed both virgin feed material and recycled or reprocessed reactor fuel to enrichments of 

up to 5% (by weight) of 235U in the final product.1

The primary method of monitoring employees for intakes of radionuclides at PGDP was urine 

bioassay. Bioassay monitoring was instituted at the start of enrichment operations and has continued 

to the present. However, the focus of the monitoring program in the very early years was the 

detection of excreted soluble uranium. When monitoring for less soluble isotopes of uranium and 

TRU elements was necessary, in vivo methodologies were implemented, primarily whole-body 

counting and chest (lung) counting.

 This processing involved small quantities of 

transuranic (TRU) elements, primarily neptunium and plutonium, present in various workplaces, as 

well as the thorium and protactinium progeny of uranium and the fission product 99Tc.

Until the mid-1980s, action levels were set based on the amount of uranium excreted. Later, intakes 

and doses were assessed based on both in vivo and in vitro monitoring results, using DOSEXPRT, a 

computer program developed by Oak Ridge National Laboratory (ORNL).2

 1 The most predominant enrichment level of the final product was 1.5%. Reprocessed fuel was used as feed from 1952 until 

1976. At that time, the cascade facilities were upgraded and most of the TRU and fission product materials were removed. 

Campaigns involving reprocessed fuel elements ended in the 1980s. In 2000, final products were enriched to the maximum 

value of 5%.

 Data are available from 

1952 to the present for both in vivo and in vitro analysis records and associated interpretations.

2 As shown in Eckerman and Ward (1992), DOSEXPRT Version 4.2 was used to analyze intake and dose for PGDP 

personnel for 1991. Version 4.1 was used for analysis of 1990 bioassay data, and Version 3.0 was used for analysis of 1989 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 6 of 26

A review of in-house procedures used to assess the concentration of uranium in urine indicates that a 

variety of quality control steps were an integral part of the process. For example, duplicates were 

consistently run, and comparison of results to known quantities was a critical step. Therefore, the in 

vitro results from in-house processing, typically reported in units of micrograms of uranium per liter, 

can be considered generally reliable. However, interpretation of those results can be difficult, 

primarily because of uncertainty about enrichment, solubility, and the contribution of environmental 

uranium, but also because samples were collected at work and during the middle of the workweek, 

meaning that cross-contamination and the inability to separate soluble from insoluble intake fractions 

contribute to the uncertainty.

Nonetheless, dose reconstructors can prepare reliable estimates of dose from the dates of 

employment, the employment locations, and the urine bioassay results. (In vivo results, because they 

were acquired primarily in response to an incident, are less reliable for assessing routine intakes.) 

Assumptions such as absorption types and the presence or absence of TRU elements can be derived 

from the historical records. 

Section 5.2 provides guidance on selection of source terms. Sections 5.3 and 5.4 involve 

interpretation of in vivo and in vitro measurement results, respectively, each including instructions for 

assessing dose for both monitored and unmonitored employees. Section 5.5 identifies significant 

incidents with internal dose potential. 

5.2 SOURCE TERM

The mission of PGDP was to enrich uranium in the form of UF6 (for use in domestic and foreign 

commercial power reactors) from roughly 0.7% 235U (natural enrichment) to 2.5% 235U (DOE 2000, 

p. 6). In addition, other compounds of uranium were present throughout the plant’s history, including 

UO2F2, UF4, and UO3. The primary radionuclides of concern for the plant are 238U, 235U, and 234U. 

The progeny of dosimetric interest for these radionuclides includes 230Th and 234mPa (DOE 2000, 

p. 24).

Certain TRU isotopes have been a factor at PGDP, including 237Np and 239Pu. These are the result of 

the processing of reactor tails. Reactor tails were fed to the cascade from 1953 to 1964, and again 

from 1969 to 1974 (with the exception of 1971 when none of the feed was of reactor origin) (Smith 

1984, p. 9). When this processing was underway, approximately 19% of the feed material in use at 

the plant was reactor tails (DOE 2000, which included both neptunium and plutonium). Table 5-1 is a 

summary of the percentages of TRU materials as assessed at specific facilities (IT Corporation 1992).

Table 5-1. Elemental fractions of uranium, neptunium, and plutonium at specific 

facilities.

a,b

Plant location Percent uranium Percent neptunium Percent plutonium

Feed plant (general) 100 0 0

Cascade (general) 100 0 0

Bldg. C-333 85.9 11.0 3.3

Bldg. C-720 82.9 16.4 0.7

Bldg. C-337 76.3 22.8 0.98

Bldg. C-400 93.1 6.9 0

Bldg. C-409 76.3 22.0 1.7

a. Elemental fractions, in this case, are presumed to refer to the percentage of total alpha activity.

b. See Table 5-4 for activity concentration of impurities.

 data. The method for computing intake, committed dose, and annual dose did not change between Versions 3.0 and 4.2, 

although a faster algorithm was developed for computation of annual dose for Version 4.1.
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It is important to note that this table presents only a snapshot of TRU percentages over time. 

However, dose reconstructors can apply these percentages to data collected after 1953. 

At the PGDP, monitoring for intakes of uranium, whether in vivo or in vitro, often resulted in reports of 

elemental uranium concentration in urine or the mass of elemental uranium in organs or the whole 

body. However, internal dose assessment requires the use of isotopic concentrations as input to the 

assessment process. Therefore, reconstructors should use Table 5-2 to derive the isotopic fractions 

associated with each microgram of uranium reported in an analytical result. In addition, the table 

provides default values if the specific location where a claimant worked is not available.

Table 5-2. Isotopic fractions for various enrichment percentages.

a

Enrichment

Isotopic fractions

(Bq of isotope/μg total U)

U-234 U-235 U-236 U-238

Natural uranium 1.24E-02 5.50E-04 1.67E-05 1.24E-02

93% enriched feed 2.41E+00 7.38E-03 4.90E-03 6.83E-04

4% enriched feed 6.42E-02 3.17E-03 6.67E-05 1.18E-02

Typical commercial (3% enriched) feed 6.55E-02 2.99E-03 5.00E-05 1.19E-02

3.5% enriched feed 6.67E-02 2.80E-03 5.84E-05 1.20E-02

Low-enrichment (2% enriched) feed 2.52E-02 1.58E-03 3.33E-05 1.21E-02

Typical depleted uranium 2.30E-03 1.59E-04 2.40E-06 1.24E-02

Recycled Uranium Table 5-3 Table 5-3 Table 5-3 Table 5-3

Default 2.52E-02 1.58E-03 3.33E-05 1.21E-02

a. Sources: ANSI (1995); ORAU (2003); IMBA computer software (Birchall et al. 2003); Berger (2004).

Dust samples from a variety of process systems in 1989 were analyzed for their isotopic content. 

Table 5-3 summarizes the results for selected process areas at the plant that dose reconstructors can 

use in the absence of employee-specific information to evaluate the total uranium bioassay data 

collected after 1953 (Baker 1987, pp. 6 to 8). However, reconstructors should use these values 

cautiously as they present only a snapshot of information.

Table 5-3. Transuranic element concentrations in plant aerosols (1987). Pu-239 Np-237 U-238 U-234 Th-230

Location dpm/g U nCi/μg U dpm/g U nCi/μg U dpm/g U nCi/μg U dpm/g U nCi/μg U dpm/g U nCi/μg U

UO3 Powder Area 500 0.225 270 0.122 750,000 337.5 750,000 337.5 42,000 18.9

C-420 Green Salt Plant 500 0.225 270 0.122 750,000 337.5 750,000 337.5 42,000 18.9

Fluorination Tower Area 150,000 67.5 12,000 5.4 750,000 337.5 750,000 337.5 42,000 18.9

Cold Trap and 

Refrigeration Operation

120,000 54 9,000 4.05 750,000 337.5 750,000 337.5 34,000 15.3

C-410 Control Room 

Operations

150,000 67.5 12,000 5.4 750,000 337.5 750,000 337.5 42,000 18.9

The fission product 99Tc has also been present during plant operations, 

particularly during the processing of reactor tails. The available 

documentation indicates that the tails contained from 0.041 to 7.0 ppm 

technetium (Smith 1984, Appendix 12; DOE 2000).

For dose assessment purposes, a nominal distribution of radionuclides 

must be assumed because not all analytical methods were capable of 

detecting many of the radionuclides in the PGDP source term. If only total

uranium results are available for a particular measurement result, Table 

5-4 provides a default isotopic distribution. 

At the PGDP, and unless site-specific information is available, the particle 

size is assumed to be 5 μm aerodynamic median activity diameter 

Table 5-4. Default 

isotopic distribution.

Radionuclide nCi/g U

Pu-239 67.5

Am-241/Pu-241 67.5

U-236 0.93

U-235 43.9

U-234 702.0

U-238 337.5

Np-237 5.4

Th-230 18.9

Tc-99 0.12
Effective Date: 09/

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0/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 

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Table 5-5. Facility-specific source radionuclides with solubility type and particle size.

a,b,c

Facility Compound Radionuclide

Absorption

type

Particle size

(μm AMAD)c

C-331 Process Building UF6 U-234 F 5

U-235 F 5

U-238 F 5

Np-237 M 5

Pu-239 S 5

Tc-99 F 5

C-333 Process Building UF6 U-234 F 5

U-235 F 5

U-238 F 5

Np-237 M 5

Pu-239 S 5

Tc-99 F 5

C-335 Process Building UF6 U-234 F 5

U-235 F 5

U-238 F 5

Np-237 M 5

Pu-239 S 5

Tc-99 F 5

C-337 Process Building UF6 U-234 F 5

U-235 F 5

U-238 F 5

C-310 Process Product 

Withdrawal

UF6 U-234 F 5

U-235 F 5

U-238 F 5

Th-230 S 5

Pa-234m F 5

Tc-99 F 5

C-315 Process Tails Withdrawal UF6 U-234 F 5

U-235 F 5

U-238 F 5

Th-230 S 5

Pa-234m F 5

C-337A Feed Materials Process 

Area

UO2F2 U-234 F, M 5

U-235 F, M 5

U-238 F, M 5
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Table 5-5 (Continued). Facility-specific source radionuclides with solubility type and particle size.a,b,c

Facility Compound Radionuclide

Absorption

type

Particle size

(μm AMAD)c

C-333A Feed Materials Process 

Area

UO2F2 U-234 F, M 5

U-235 F, M 5

U-238 F, M 5

C-720 Maintenance Building UO2F2 U-234 F, M 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 F, M, S 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Th-230 S 5

Pa-234m F 5

C-405 Ash handling U-234 F, M, S 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 F, M 5

Pu-239 S 5

Tc-99 F 5

C-410 UO2, UO3, UF4, UF and UF6 U-234 F, M, S 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 F, M 5

Pu-239 S 5

Tc-99 F 5

C-420 UF4, UO3, UO2, and UF U-234 F, M, S 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 M 5

Pu-239 S 5

Tc-99 F 5

C-409 Decontamination Building UF6, UO2F2 U-234 F, M 5

U-235 F, M 5

U-238 F, M 5

Np-237 F, M 5

Pu-239 S 5
Effective Date: 09/

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Table 5-5 (Continued). Facility-specific source radionuclides with solubility type and particle size.a,b,c

Facility Compound Radionuclide

Absorption

type

Particle size

(μm AMAD)c

C-400 Decontamination Building UF6, UO2F2 U-234 F, M, S 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 F, M, S 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Th-230 S 5

Pa-234m F 5

C-340 Metal facility UF6, UF4, UO2 U-234 M, S 5

U-235 M, S 5

U-238 M, S 5

Th-230 S 5

Pa-234m F 5

C-746Q Waste Building UF6 U-234 F, M 5

U-235 F, M 5

U-238 F, M 5

Np-237 F, M 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Tc-99 F 5

C-746A Waste Building UF6 U-234 F, M 5

U-235 F, M 5

U-238 F, M 5

Np-237 F, M 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Tc-99 F 5
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Table 5-5 (Continued). Facility-specific source radionuclides with solubility type and particle size.a,b,c

Facility Compound Radionuclide

Absorption

type

Particle size

(μm AMAD)c

C-746B Waste Building UF6 U-234 F, M, S 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 M 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Tc-99 F 5

C-404 Burial Ground UF4 U-234 F, M 5

U-235 F, M 5

U-238 F, M 5

C-745A-T Cylinder Yards U-234 F 5

U-235 F 5

U-238 F 5

C-710 Laboratory Neptunium recovery U-234 F, M 5

U-235 F, M 5

U-238 F, M 5

Np-237 M 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Tc-99 F 5

Th-230 S 5

Pa-234m F 5

C-360 Toll Transfer and 

Sampling Building

U-234 F, M, S 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 F, M 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Tc-99 F 5
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Table 5-5 (Continued). Facility-specific source radionuclides with solubility type and particle size.a,b,c

Facility Compound Radionuclide

Absorption

type

Particle size

(μm AMAD)c

Default Not applicable U-234 F, M, S 5

U-235 F, M, S 5

U-238 F, M, S 5

Np-237 F, M, S 5

Pu-238 S 5

Pu-239 S 5

Pu-240 S 5

Am-241 M 5

Th-230 S 5

Pa-234m F 5

Tc-99 F, M 5

a. As of the date of this report, there is little information to support a reduction in the types or forms of radionuclides listed 

in this table on an annual basis, and few effective decommissioning efforts that could have resulted in downgrading 

these radiologically restricted areas. Therefore, it is assumed that all of the radionuclides shown were present in the 

buildings beginning in 1953. Before 1953, reactor tails were not used as feed to the cascade, and the TRU materials 

do not apply.

b. Data and information on source terms and solubility types are from Hill and Strom (1993, Table 16.2), PACE et al. 

(2000, Table 2.4-1), DOE (2000, Table B-1), and BJC (1999, Table 1.7). (These documents also describe a single 

particle size study of unknown duration and procedure that shows particle sizes of 3 μm AMAD for 237Np and certain 

plutonium isotopes in Buildings C-720, C-400, C-476Q, C-746A, C-746B, C-710, and C-360. Because of the 

limitations of the study, the findings were not incorporated into the table.) Where no facility-specific information was 

available, the values are from the recommendations in ICRP (1997).

c. Unless site-specific information is available, the particle size is assumed to be 5 μm AMAD, as recommended in ICRP 

(1994, paragraph 5).
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(AMAD), as recommended by the International Commission on Radiological Protection (ICRP 1994, 

paragraph 5). If information on the source term to which the employee was exposed is available, the 

dose reconstruction should use that source term. However, if no source term information is available, 

the values and parameters in Table 5-5 provide conservative input to the process.

5.3 IN VITRO MEASUREMENT METHODS

From the start of plant operations in 1952, samples of urine from workers involved in enrichment 

operations were analyzed for uranium.3

Table 5-6. In vitro measurement frequencies.

a, b

 Over time, other workers were included in the monitoring 

program. In addition, special sampling occurred in response to incidents or issues (i.e., assessments 

of the radiological impacts of TRU elements and technetium); however, those analytical methods were 

typically performed off the PGDP site, at ORNL during the early periods and later at analytical 

services contractor locations). Table 5-6 lists the sampling frequencies and volumes for workers 

assigned to specific buildings during various periods. In addition, the table provides default values if 

the specific location of a claimant is not available.

Period Facility Frequency Volume

1953 to present day C-310 Every 4 weeks Spot

1953 to present day C-315 Every 4 weeks Spot

1953 to present day C-340 Every 4 weeks Spot

1953 to present day C-400 Every 4 weeks Spot

1953 to present day C-410 Every 4 weeks Spot

1953 to present day Remainder Yearly Spot

1959 to present day All Within 0.5 hr of incident Spot

1960 C-410 Every 2 months Spot

1960 C-340 Every 3 months Spot

1960 C-331, C-333, 

C-335, C-337

Every 4 months Spot

1960 C-410 Every 6 months Spot

1991 to present day All Every 4 weeks 24-hr collection or 

simulated 24-hr collection

Default N.A. Every 4 weeks Spot

a. Sources: PACE et al. (2000, Section 7.4); PGDP (2003a).

b. There is evidence that in 1960 the measurement frequency for Building C-410 and C-340 workers could have 

been either every four weeks or every two months (C-410) or four months (C-340). However, there is no 

referenceable indication as to whether there was a special class of worker to which the two-month frequency 

applied. The overlapping values have been left in the table.

At PGDP, routine urine samples were typically single voids collected during the middle of the week. 

Each could have been measured for specific gravity, pH, sugar, and albumin levels, as well as for 

uranium content. During the early years, total uranium concentrations were measured using a 

calibrated fluorimeter with a detection level of 0.005 mg/L. In later years total uranium content was 

assessed by kinetic phosphorescence analysis (KPA). Neither methodology included isotopic 

determinations (PACE et al. 2000, Section 4.2.1). To account for fluid mass balance over a 24-hr 

period in spot samples, the methodology presented by the National Council on Radiation Protection 

and Measurement (NCRP) Report 87 may be used (NCRP 1987). If the specific gravity was 

measured and reported for any given spot sample, the following equation may be used to derive the 

daily urinary excretion rate:

 3 Fecal sampling was occasionally performed for special studies. However, a program for routine or diagnostic monitoring of 

fecal samples was never implemented.
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( 1) g/ mL

(1.024 1) g/ mL

−

− = ×

m

cor m SG

C C (5-1)

where 

Ccor = the corrected radionuclide concentration in the urine sample

Cm = the measured concentration

SGm = the measured specific gravity of that sample

The equivalent 24-hr excretion is then:

24 hr

1.4 L

24 hr L  ×



 



  =



 



 pCi C pCi Au cor (5-2)

In cases where the specific gravity is unknown, the measurement result should be normalized to 

activity per day, as follows, by assuming a reference value V of 1.4 and volume L of 1.0 L:

24 hr L 24 hr

pCi VL C pCi Au m  ×



 



  =



 



 (5-3)

5.3.1

Table 5-7 lists the in vitro measurement types and detection levels during various periods. For 

samples analyzed at the site, the detection level for total uranium in urine was reported as 0.005 mg/L 

(PACE et al. 2000, Section 4.2.1). If referenceable detection levels were not available, those 

specified as typical in ICRP (1988) were used.

Measurement Types and Detection Levels

Table 5-7. In vitro measurement types and detection levels for various periods.

a

Period Measurement type Radionuclide MDCb (mg/L) Recall level (μg/L)

1952 to 1961 In-house fluorimetry Total uranium 0.01 10

1962 to 1976 In-house fluorimetry Total uranium 0.01 60 (single sample)

1962 to 1976 In-house fluorimetry Total uranium 0.01 33 (quarterly average)

1962 to 1969 In-house fluorimetry Total uranium 0.01 12 (three consecutive 

samples)

1977 to 1982 In-house fluorimetry Total uranium 0.01 33 (single sample)

1982 to present KPA

Total uranium (soluble, 

and less than 5% U-235) 0.005 50

1982 to present KPA Total uranium (insoluble) 20

1969 to 1985 In-house Gross beta (dpm L-1) 13 dpm ml-1

1985 to present ORNL Isotopic plutonium 1 pCi/L

1985 to present ORNL beta counting Tc-99 10 dpm/ml

1989 to present Contractor U-234, U-235, and U-238 0.3 pCi/L

1999 to present ORNL Natural uranium 0.06 mg/sample

1999 to present ORNL Tc-99 16 pCi/sample

1999 to present ORNL U-234, U-235, and U-238 0.014 pCi/sample

1999 to present ORNL Th-228, Th-230, and Th232 0.014 pCi/sample

1999 to present ORNL Np-237 0.009 pCi/sample

1999 to present ORNL Pu-238, Pu-239, and Pu2400.009 pCi/sample

1999 to present ORNL Am-241 0.009 pCi/sample

a. Sources: PACE et al. (2000, Table 4.3); PGDP (2003b); BJC (1999, Table 3.3); PGDP (1999, Section 6.1.8).

b. MDC = minimum detectable concentration.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 15 of 26

In some cases, a detection level for a particular radionuclide or analysis method is not available. In 

that case, dose reconstructors should use the nominal detection levels in Table 5-8. In addition, if it is 

not clear from the monitoring records how/where a particular claimant’s sample was analyzed, it 

should be assumed that they were analyzed in-house (i.e., at the PGDP) and the MDC from that 

measurement type used to assess missed dose. Finally, if a record contains a notation of “less than 

X micrograms/L” or “< x pCi/sample”, that value should reflect the MDC for that sample.

Table 5-8. Nominal (default) detection levels for urine bioassay.

a

Radionuclide Analytical method Detection level (pCi/L)

Th-228 Radiochemical separation and alpha spectrometry 0.27

Th-232 Radiochemical separation and alpha spectrometry 0.27

U-234 Alpha spectrometry 0.27

U-235 Alpha spectrometry 0.27

U-238 Alpha spectrometry 0.27

Pu-238 Alpha spectrometry 0.27

Pu-239 Alpha spectrometry 0.27

Pu-240 Alpha spectrometry 0.27

Am-241 Alpha spectrometry 0.27

a. Source: ICRP (1988).

5.3.2

A variety of codes occurs on various urine bioassay records for PGDP. Table 5-9 lists a summary of 

those known as of the date of this report, along with their interpretations.

Reporting Formats and Codes

5.3.3

The practice of offsite collection of samples that takes place 24 to 48 hr after leaving the plant not only 

minimizes the possibility of sample cross-contamination, but it ensures that samples are collected 

after the transfer of the rapid clearance component. Some PGDP employees were asked to collect 

samples after 1 or 2 days off from work; if so, that collection instruction was sometimes noted on the 

analytical record.

Instructions for Addressing Possible Interferences and Uncertainties

Urine samples were typically collected in the workplace at PGDP, predominantly on Wednesdays.4

Dietary intakes of uranium pose a potential problem in interpreting urine bioassay results for PGDP 

workers. Because studies of the average daily uranium excretion on Paducah residents do not 

appear to have been performed, it is not possible to make corrections for the contribution of 

nonoccupational intakes of uranium to a given urine sample result. However, to put a given result into 

perspective, a nominal daily (24 hr) urinary excretion rate for uranium of 0.43 μg (environmental 

decision level at 95% confidence) can be used (BJC 1999). No correction for environmental levels of 

uranium is required for samples analyzed by fluorimetry or KPA because the MDC is larger than the 

correction. 

 

Therefore, contamination of samples from the worker’s hands or clothing cannot be ruled out as a 

contributor to any given result. If a second analysis was performed and if that result was negative, it 

is reasonable to assume the first result was a false positive due to sample contamination or laboratory 

error.

 4 At some point, the sample collection day changed to Monday. However, the date of that procedural change is unclear.
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0/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 16 of 26

Table 5-9. In vitro record codes.

a, b, c, d

Form identifier

Measurement 

type Column identifier Code Interpretation

WCP-455 Urine bioassay Reason for Visit 33, 35 Industrial health recheck

WCP-455 Urine bioassay Reason for Visit OB, 39, 35-1 Recall sample requested following an elevated sample

WCP-455 Urine bioassay Reason for Visit MM recall Monday morning recall sample; requested after days off 

work

WCP-455 Urine bioassay Reason for Visit 32, 33, Term Termination samples

WCP-455 Urine bioassay Reason for Visit 30, 22 Rehire

WCP-455 Urine bioassay Reason for Visit 37, 18, Per. Periodic physical; confirmatory samples were collected 

during routine physicals; this typically did not pertain to 

those on a routine monitoring program.

WCP-455 Urine bioassay Reason for Visit 26 Pre-employment

WCP-455 Urine bioassay Reason for Visit 005, 07, 60, 

Exposure, 

Special, Release

Samples collected following an exposure or potential 

exposure in a uranium release or spill

WCP-455 Urine bioassay Bottle No. Permanent sample number

PGDP_HISTORICAL_URINE Urine bioassay Results There is no distinction between positive results and 

detection limits.

PGDP_HISTORICAL_URINE Urine bioassay Sample Type Physical Routine physicals included the collection of a bioassay 

sample (random sampling program); this typically did not 

pertain to those on a routine monitoring program.

WCP-455 Urine bioassay Reason for Visit I.H.R. or IHR Industrial health recheck (associated with routine 

physicals)

WCP-455 Urine bioassay Top of Card “A” Refers to the shift worked (“A” = day shift)

WCP-455 Urine bioassay Top of Card “B” Refers to the shift worked (“B” = evening shift)

WCP-455 Urine bioassay Top of Card “O” Refers to the shift worked (“O” = midnight shift)

WCP-455 Urine bioassay F, HG and OTHER “B” (128) Indicates the shift and the hours worked

PGDP_ANALIS_URINE Urine bioassay Results Results are given in μg/L.

5EA HPINT - Bioassay Results Report Urine bioassay Units 4 dpm/L

5EA HPINT - Bioassay Results Report Urine bioassay Units 1 dpm/ml

5EA HPINT - Bioassay Results Report Urine bioassay Units 2 dpm/day

5EA HPINT - Bioassay Results Report Urine bioassay Units 3 dpm/sample

5EA HPINT - Bioassay Results Report Urine bioassay Units 5 μg/ml

5EA HPINT - Bioassay Results Report Urine bioassay Units 6 Bq/L

5EA HPINT - Bioassay Results Report Urine bioassay Units 7 Bq/day

5EA HPINT - Bioassay Results Report Urine bioassay Reason 3 Routine

5EA HPINT - Bioassay Results Report Fecal analysis Units 3 dpm/sample

5EA HPINT - Bioassay Results Report Fecal analysis Units 7 Bq/day

5EA HPINT - Bioassay Results Report Breath analysis Units 2 dpm/day

5EA HPINT - Bioassay Results Report Breath analysis Units 3 dpm/sample

5EA HPINT - Bioassay Results Report Breath analysis Units 4 dpm/L

5EA HPINT - Bioassay Results Report Breath analysis Units 6 Bq/L

5EA HPINT - Bioassay Results Report Breath analysis Units 7 Bq/day
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0/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 17 of 26

Table 5-9 (Continued). In vitro record codes (Cont’d.).a. b, c, d

Form identifier

Measurement 

type Column identifier Code Interpretation

WCP-885 Urine bioassay Schedule First digit refers to day of week; second digit(s) refer to 

type of analysis

WCP-885 Urine bioassay Schedule A Day sample is to be taken (1=Monday; 2=Tuesday; 

3=Wednesday; 4=Thursday)

WCP-885 Urine bioassay Schedule B Type analysis (1=Uranium; 2=Fluoride; 3=Mercury)

WCP-885 Urine bioassay Location Shifts A, B, O and D

Frequency Uranium #1 = 1 a month; Fluoride #2 = 1 a year; Mercury 

#3 = Blank

Permanent log (sample collection log) Urine bioassay Number Permanent 

sample number

Numbers are consecutive from top to bottom of page 

and continue from one page to the next.

IBM Report Cards Urine bioassay A Name

IBM Report Cards Urine bioassay B Badge

IBM Report Cards Urine bioassay C Date (date shown on log)

IBM Report Cards Urine bioassay D Code for this service

IBM Report Cards Urine bioassay E Analysis results in boxes labeled Uranium, Fluorides, 

and Mercury

UCN-5242, “Sample Analysis (Medical).” Urine bioassay Same as for NCP455Same as for NCP455This form number was in use from 1969 to 1970.

a. Sources: Maisler (2003); Eckerman and Ward (1992); PGDP (2003b), ), Tomes (2004)

b. Form WCP-885 is referred to as “NCP-885” in Eckerman and Ward (1992), however they appear to refer to identify the same form.

c. Around June 1956 form WCP-455 was modified with additional columns added. Individual bioassay records reviewed indicates the exact date the newly expanded form was 

used varies a little from person to person, but was in the middle of 1956 for all records reviewed. 

d. The bioassay records we have starting in 1977 (exact dates vary from person to person) are not copies of results recorded on bioassay cards--they are database printouts, e.g., 

“PGDP_Historical_Urine.” The uranium results are in units of ug/L with results records to the nearest integer, i.e., 0, 1, 2, .etc. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 18 of 26

5.4 IN VIVO MEASUREMENT METHODS

Whole-body counting and other in vivo methods were implemented beginning in the early 1950s. 

Until the present day (2003), this measurement method was used primarily in response to incidents, 

or for assessing the magnitude of insoluble material intakes. Even when routine whole-body counting 

was instituted for certain PGDP employees in the late 1960s, the counting frequency was sporadic 

and seldom greater than once per year. 

5.4.1

At the PGDP, whole-body counting was performed using a mobile counter provided by the Y-12 plant 

(sometimes referred to as the MMES Counter) and at other facilities. Table 5-10 lists general 

information about the detection capabilities of this counting system for various periods.

Measurement Types and Detection Levels

Table 5-10. In vivo measurement types and detection levels for various periods.

a

Period Equipment

Measurement 

type Radionuclide

MDAb

(units of 

record)

Action level 

for recount

Action level 

for work 

restriction

1958 ORNL Lung Pu-239 0.04 μCi Not specified Not specified

1960-1967 Y-12 Whole body Np-237 0.5 μCi Not specified Not specified

1968-1980 Y-12 mobile counter Whole body U-235 83 μg Not specified Not specified

1968-1980 Y-12 mobile counter Whole body U-238 4 mg Not specified Not specified

1968-1980 Y-12 mobile counter Whole body Np-237 0.017 μCi Not specified Not specified

1965-1991 Y-12 mobile counter Lung Total uranium 4 mg 4 mg 27 mg

1965-1991 Y-12 mobile counter Lung Enriched uranium (2%) 100 μg 100 μg 240 μg

1965-1991 Y-12 mobile counter Lung Depleted uranium 4 mg 4 mg 37 mg

1965-1991 Y-12 mobile counter Lung Np-237 0.2 nCi 1.7 nCi 17 nCi

1991-1995 Helgesson counter Lung Total uranium 2 – 4 mg 2 – 4 mg 27 mg

1991-1995 Helgesson counter Lung Enriched uranium 40 – 70 μg 40 – 70 μg 240 μg

> 1995 No counting performed.

a. Sources: DOE (2000); Hill and Strom (1993); BJC (1999, Tables 3.2 and 3.5); Scott and West (1967).

b. MDA = minimum detectable activity or amount. The MDAs shown for uranium, while given in units of mass, are presumed to 

have been based on measurement of Th-234 activities along with an assumed isotopic ratio. It is presumed that the results for 

enriched uranium are based on measurement of the measurement of U-235 activities and an assumption of enrichment. 

However, these presumptions cannot be confirmed thus cautious use of these MDA values is recommended.

5.4.2

A variety of codes and reporting formats appear in the in vivo bioassay records. Table 5-11 lists the 

known codes with their interpretations.

Reporting Formats and Codes

5.4.3

For in vivo measurements, contamination could have occurred as external to the body or, in the case 

of chest counting, as external to the lung. If a follow-up in vivo count (the same day or within a few 

days) showed a dramatic decrease in activity or no detectable activity, then external contamination 

should be assumed. 

Instructions for Addressing Possible Interferences and Uncertainties

Radon progeny and medical diagnostic or therapeutic procedures involving radionuclides have 

caused interferences among in vivo measurements, especially when sodium-iodide detectors were 

used. However, unless the count was invalidated or noted as being influenced by such interferences,

the results should be used as recorded.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 19 of 26

Table 5-11. In vivo record codes.

a

Form identifier

Measurement 

type Column identifier Code Interpretation

In Vivo Radiation Monitoring 

Report

Whole body count Surface contamination Check mark, yes The surface contamination on the 

subject was checked. If measurable 

activity was found, it was so noted on the 

card.

In Vivo Radiation Monitoring 

Report

Whole body count Analysis Sequence These align with the Output - Analysis 

Sequence Results listed at the bottom of 

the card; they are not relevant to the 

dose reconstruction process.

In Vivo Radiation Monitoring 

Report

Whole body count Output - Analysis 

Sequence Results

A. Enriched 

Uranium

The maximum U-235 enrichment was 

2% until 2000, after which it was 5%.

In Vivo Radiation Monitoring 

Report

Whole body count Output - Analysis 

Sequence Results

J. NLO Uranium Refers to a special spectrum region of 

interest for National Lead of Ohio, an 

early operator of the Fernald facility.

In Vivo Radiation Monitoring 

Report

Whole body count Hand-written notes No Np noted The presence or absence of Np-237 was 

qualitatively evaluated.

In Vivo Radiation Monitoring 

Report (with boxes for 

data entry)

Whole body count Hand-written notes BFD Initials of whole body counter operator 

(individual who filled out the card)

5EA HPINT - Bioassay 

Results Report

In vivo records Units M μCi

5EA HPINT - Bioassay 

Results Report

In vivo records Units N nCi

5EA HPINT - Bioassay 

Results Report

In vivo records Units P pCi

5EA HPINT - Bioassay 

Results Report

In vivo records Units D dpm

5EA HPINT - Bioassay 

Results Report

In vivo records Units B Bq

5EA HPINT - Bioassay 

Results Report

In vivo records Units U μg

a. Sources: Maisler (2003); Eckerman and Ward (1992); PGDP (2003a).

Uncertainties in the bioassay measurements were not stated in the records. For results near or at the 

reporting levels, dose reconstructions should apply the prescribed standard deviation of 0.3 times the 

MDA or reporting level (NIOSH 2002).

On occasion, in vivo measurement results included 137Cs. However, those PGDP workers could have 

had body burdens of 137Cs from nonoccupational sources (e.g., fallout and consumption of specific 

foodstuffs). There is no evidence of occupational intakes of 137Cs at the PGDP, thus no dose of 

record should be associated with these measurement result, if any.

5.4.4

In general, available urine results should be considered the primary method of dose reconstruction. 

The in vivo measurements, especially in the earlier years of operation, were not used for routine 

monitoring purposes. However, those results can and should be used to verify assessments of dose 

based on urine bioassay results, in determining likely absorption types, or to provide upper and lower 

limits to the range of possible doses. 

Assessment of Intake for Monitored Employees

5.5 SIGNIFICANT INCIDENTS WITH INTERNAL DOSE POTENTIAL

During operations at PGDP, a number of incidents occurred that increased the potential for intakes of 

radioactive materials. If a claimant (or employee) recalls involvement in one or more of those 

incidents, the information in Table 5-12 can be used as input to an incident-specific dose assessment.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 20 of 26

Table 5-12. Input parameters for significant incidents and events.

a

Incident date Incident description Facility Other information

1952–1990 Exposure to UF4, UO, and process 

dust during guard patrolling

All buildings

1952–1980 Exposure to uranium metal

July 1953 First use of reactor tails

November 1956 Fire C-310

1957–1977 Green salt, black oxides on floors 

and other surfaces

C-340

December 1962 Fire C-337

March 1962 Explosion and fire C-340 One fatality

April 1968 Worker overexposure Unknown Two workers overexposed

January 1978 Fire C-315

1958 to 1962 Cascade improvement program C-331, C-333, C-335 and C-337

1974 to 1982 Cascade improvement program C-331, C-333, C-335 and C-337

Neptunium production C-400

1980 to 1982 Exposure to UF4 and uranium dust 

during drum crushing

C-746

a. Source: PACE et al. (2000).
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 21 of 26

REFERENCES

ANSI (American National Standards Institute), 1995, Bioassay Programs for Uranium, HPS 

N13.22-1995, Washington, DC.

Cited References

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

Berger, C. D., 2004, “Information from World Information Service on Energy website,” memorandum 

to file, Oak Ridge Associated Universities, Oak Ridge, Tennessee.

Birchall, A., T. James, J. March, D. Dorrian, K. Davis, T. Smith, A. Phipps, T. Smith, N. Jarvis, T. 

Riddell, M. Oeoce, F. Smith, M. Puncher, and T. Ward, 2003, Integrated Modules for Bioassay 

Analysis (IMBA) Expert, OCAS Edition, Version 3.0.63, ACJ & Associates, Richland, 

Washington. 

BJC (Bechtel Jacobs Company, LLC), 1999, Internal Dosimetry Technical Basis Document for 

Bechtel Jacobs Company LLC, BJC/OR-258, Oak Ridge National Laboratories, Oak Ridge, 

Tennessee.

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

Gaseous Diffusion Plant - Environment, Safety and Health Practices, 1952-1990, Office of 

Oversight, Environment, Safety and Health, Washington, DC.

Eckerman K. F., and R. C. Ward, 1992, DOSEXPRT - A Bioassay Dosimetry Code for Martin Marietta 

Energy Systems, Inc., ORNL/TM-11857, Oak Ridge National Laboratories, Oak Ridge, 

Tennessee.

Hill, R. L., and D. J. Strom, 1993, Internal Dosimetry Technical Basis Manual for Portsmouth and 

Paducah Gaseous Diffusion Plants, PNL-8723, Pacific Northwest Laboratories, Richland, 

Washington.

ICRP (International Commission on Radiological Protection), 1988, Individual Monitoring for Intakes of 

Radionuclides by Workers: Design and Interpretation, Publication 54, Pergamon Press, 

Oxford, England.

ICRP (International Commission on Radiological Protection), 1994, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1997, Individual Monitoring for Internal 

Exposure of Workers, Publication 78, Pergamon Press, Oxford, England.

IT Corporation, 1992, Personnel Exposure Potential to Transuranic Materials at the Paducah 

Gaseous Diffusion Plant, IT/NS-91-109.

Maisler, J., 2003, “Paducah Gaseous Diffusion Plant discussion with K. Duncan, September 29-30, 

2003”, memorandum to file, November 20, 2003, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 22 of 26

NCRP (National Council on Radiation Protection and Measurements), 1987, Use of Bioassay 

Procedures for Assessment of Internal Radionuclide Deposition, Report 87, Bethesda, 

Maryland.

NIOSH (National Institute of Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guide, OCAS-IG-002, Office of Compensation Analysis and Support, 

Cincinnati, Ohio.

ORAU (Oak Ridge Associated Universities), 2003, Technical Basis Document for the Hanford Site –

Occupational Internal Dose, ORAUT-TKBS-0006-6, Revision 00, Oak Ridge, Tennessee.

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

U.S. Department of Energy, 2000, Exposure Assessment Project at the Paducah Gaseous 

Diffusion Plant, Utah.

PGDP (Paducah Gaseous Diffusion Plant), 1999, PGDP Urinalysis Program for Detection of Intakes 

of Radionuclides, Procedure CP2-HP-DS1030, Revision 2, Paducah, Kentucky.

PGDP (Paducah Gaseous Diffusion Plant), 2003a, “Urinalysis Program,” undated and unnumbered 

document obtained by J. E. East in September 2003, Paducah, Kentucky.

PGDP (Paducah Gaseous Diffusion Plant), 2003b, “Uranium in Urine Analysis,” undated and 

unnumbered document obtained by J. E. East in September 2003, Paducah, Kentucky.

Scott, L. M., and C. M. West, 1967,Health Physics Application of In Vivo Gamma Spectrometry in a 

Uranium Processing Plant, Y-KB-77, Union Carbide Corporation.

Smith, R. F., 1984, Historical Impact of Reactor Tails on the Paducah Cascade, KY/L-1239, Union 

Carbide Corporation.

Tomes, T., 2004, ORAU Team Dose Reconstruction Project for NIOSH Document Review Comment 

(ORAUT-TKBS-0019-5 Rev. 00-E), September 10, 2004.

ATSD (Agency for Toxic Substances and Disease Registry), 2001, Public Health Assessment, 

Paducah Gaseous Diffusion (USDOE), Paducah, Kentucky.

Other References

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

Gaseous Diffusion Plant Site Report, BJC/PGDP-167, Paducah Gasseous Diffusion Plant, 

Paducah, Kentucky.

DOE (U. S. Department of Energy), 1999, Phase I - Independent Investigation of the Paducah 

Gaseous Diffusion Plant - Environment, Safety and Health Issues, Office of Oversight; 

Environment, Safety and Health, Washington, DC.

DOE (U. S. Department of Energy), 2001, A Preliminary Review of the Flow and Characteristics of 

Recycled Uranium Throughout the DOE Complex 1952 – 1999 Project Overview and Field 

Site Reports, F-001-001.

ICRP (International Commission on Radiological Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, Publication 2, Pergamon Press, Oxford, England.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 23 of 26

ICRP (International Commission on Radiological Protection), 1974, Reference Man, Publication 23, 

Pergamon Press, Oxford, England.

NRC (U.S. Nuclear Regulatory Commission), 1974, Applications of Bioassay for Uranium, 1974, 

Regulatory Guide 8.11, Washington, DC.

PGDP (Paducah Gaseous Diffusion Plant), 1991, PGDP Monthly Urinalysis Results, March 1989 to 

March 1991, Paducah, Kentucky.

PGDP (Paducah Gaseous Diffusion Plant), 2003c, “Notice to Health Physics & Hygiene Department 

Personnel,” undated and unnumbered, refers to “revised SPP-41", which is a 4-plant 

committee recommendation on uranium-in-urine procedures, Paducah, Kentucky.

Potter, C. A., 2002, “Intake Retention Fractions Developed from Models Used in the Determination of 

Dose Coefficients Developed for ICRP Publication 68 - Particulate Inhalation,” Health Physics, 

volume 83, number 5, November.

Tockstein, D. J., 1996, Technical Basis for Routine Bioassay Participation, Lockheed Martin Energy 

Systems.

Travers, W. D., 1998, Report to Congress on the Gaseous Diffusion Plants Located Near Paducah, 

Kentucky and Portsmouth, Ohio, SECY-98-275, U. S. Nuclear Regulatory Commission, 

Washington, DC.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 24 of 26

GLOSSARY

activity median aerodynamic diameter (AMAD)

The diameter of a unit density sphere with the same terminal settling velocity in air as that of 

the aerosol particle whose activity is the median for the entire aerosol.

acute

Pertaining to intakes received “acutely,” i.e., within a short period. 

bioassay

Measurement of amount or concentration of radioactive material either in the body or in 

biological material excreted or removed from the body. Another word for radiobioassay.

bioassay procedure

A procedure used to determine the kind, quantity, location, and retention of radionuclides in 

the body by direct (in vivo) measurements or by in vitro analysis of material excreted or 

removed from the body.

body burden

The quantity of radioactive material contained in the individual's body at a particular point in 

time.

chronic

Pertaining to low-level intakes received on a continuous basis.

dose 

A general term for absorbed dose, dose equivalent, effective dose equivalent, committed dose 

equivalent, committed effective dose equivalent, or total effective dose equivalent.

dose equivalent (H)

The product of absorbed dose (D) in rad (or gray) in tissue, a quality factor (Q), and other 

modifying factors (N). Dose equivalent is expressed in units of rem (or sievert) (1 rem = 0.01 

sievert).

exposure

The general condition of being subjected to ionizing radiation, such as by exposure to ionizing 

radiation from external sources or to ionizing radiation sources inside the body. In this 

document, exposure does not refer to the radiological physics concept of charge liberated per 

unit mass of air.

insoluble material

A term loosely used to describe the relative degree of solubility of a material in body fluids. 

Recognizing that no material is absolutely insoluble, the terms low solubility or poorly soluble 

are preferable.

intake

The amount of radionuclide taken into the body by inhalation, absorption through intact skin, 

injection, ingestion, or through wounds. Depending on the radionuclide involved, intakes may 

be reported in units of mass, activity, or potential alpha energy.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 25 of 26

internal dose or exposure

The dose equivalent received from radioactive material taken into the body (i.e., internal 

sources).

internal dose assessment

An assessment of the intake and associated internal radiation dose to workers based on 

measurements taken in the work environment or from individual bioassay measurements.

In vitro measurement

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

In vivo measurement

The measurement of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body.

lung solubility type (F, M, or S)

A classification scheme for inhaled material according to its rate of clearance from the 

pulmonary region of the lung.

minimum detectable amount (MDA)

The smallest amount (activity or mass) of an analyte in a sample that will be detected with a 

probabilityof non-detection (Type II error) while accepting a probability of erroneously 

deciding that a positive (non-zero) quantity of analyte is present in an appropriate blank 

sample (Type I error).

minimum detectable concentration (MDC)

The minimum detectable amount, MDA, expressed in units of concentration.

monitoring (personnel)

The measurement of radioactivity in the whole body, in a region of the body, in material 

eliminated from the body or in the air for reasons related to the estimation of intake of 

radioactive material. The term monitoring includes interpretation of the measurements.

occupational dose

An individual's ionizing radiation dose (external and internal) resulting from that individual's 

work assignment. Occupational dose does not include doses received as a medical patient or 

doses resulting from background radiation or participation as a subject in medical research 

programs.

radiation

Ionizing radiation: alpha particles, beta particles, gamma rays, X-rays, neutrons, high-speed 

electrons, high-speed protons, and other particles capable of producing ions. Radiation, as 

used in this part, does not include nonionizing radiation, such as radio- or microwaves, or 

visible, infrared, or ultraviolet light.

Reactor Tails

Recycled uranium (typically UO3) from reactor operations (typically Savannah River and 

Hanford) that contains traces of transuranics that was not removed during prior chemical 

processing (i.e., REDOX and PUREX processes).
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-5 Page 26 of 26

recording level

A value below which data or results were considered to be too low to record and thus may not 

have been maintained.

rem

A special unit for dose equivalent. One rem is equal to 0.01 sievert.

routine monitoring

Monitoring carried out at regular intervals during normal operations.

sievert

The special name for the International System unit of dose equivalent. One sievert equals 

1 joule per kilogram, which equals 100 rem.

special monitoring

Monitoring carried out in actual or suspected abnormal conditions (i.e., measurements 

performed to estimate the amount of radionuclide deposited in a person when an intake is 

known or is suspected to have occurred). 

spot sample

A single void of urine.