ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 24

Document Title:

K-25 Gaseous Diffusion Plant –

Occupational Internal Dose

Document Number: ORAUT-TKBS-0009-5

Revision: 01

Effective Date: 01/06/2006

Type of Document: TBD

Supersedes: Revision 00

Subject Expert: Billy R. Thomas 

Document Owner

Approval: Signature on File Approval Date: 12/22/2005

Jay J. Maisler, TBD Team Leader

Approval: Signature on File Approval Date: 12/22/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 12/28/2005

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/06/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 2 of 24

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

12/21/2004 00 New Technical Basis Document for the K-25 Site – Occupational 

Internal Dose. First approved issue. Initiated by Jay J. Maisler.

01/06/2006 01 Incorporates recent direction from NIOSH to include DOL review 

comments in the Introduction. Incorporates information on bioassay 

codes provided by Task 5 in sections 5.3.2 and 5.4.2; adds 

Attachments A, B, C. Deletes reference to Y-12 bioassay program 

from section 5.3. Approved issue of Revision 01. Training required: 

As determined by the Task Manager. Initiated by Jay J. Maisler.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 3 of 24

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

4

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

5.2 Source Term ............................................................................................................................7

5.3 In Vitro Measurement Methods .............................................................................................. 12

5.3.1 Measurement Types and Detection Levels .................................................................12

5.3.2 Reporting Formats and Codes....................................................................................13

5.3.3 Instructions for Addressing Possible Interferences and Uncertainties......................... 13

5.4 In Vivo Measurement Methods............................................................................................... 14

5.4.1 Measurement Types and Detection Levels .................................................................14

5.4.2 Reporting Formats and Codes....................................................................................14

5.4.3 Instructions for Addressing Possible Interferences and Uncertainties......................... 14

5.4.4 Assessment of Intake for Monitored Employees .........................................................15

5.5 Significant Incidents with Internal Dose Potential ................................................................... 15

References ........................................................................................................................................16

Glossary ............................................................................................................................................19

ATTACHMENT A, URINE BIOASSAY CODES 1948- 1988...............................................................22

ATTACHMENT B, IN VIVO BIOASSAY CODES 1965 TO 1988........................................................23

ATTACHMENT C, IN VIVO BIOASSAY CODES 1988 TO 1991........................................................24

LIST OF TABLES

TABLE TITLE PAGE

5-1 Isotopic fractions for various enrichment percentages.............................................................. 7

5-2 Principal radionuclides found at uranium facilities and gaseous diffusion plants....................... 8

5-3 Absorption classification for some uranium compounds ........................................................... 8

5-4 Source term summary by location ............................................................................................9

5-5 Facility-specific radionuclide conversion factors ..................................................................... 11

5-6 Default isotopic distribution .................................................................................................... 12

5-7 In vitro measurement types and detection levels for various periods...................................... 13

5-8 Nominal (default) detection levels for urine bioassay.............................................................. 14

5-9 In vivo measurement types and detection levels for various periods ...................................... 14
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 4 of 24

ACRONYMS AND ABBREVIATIONS

Bq becquerel

dpm disintegrations per minute

g gram

hr hour

ICRP International Commission on Radiological Protection

IMBA Interactive Modules for Bioassay Analysis

IRF intake retention fraction

KPA kinetic phosphorescence analysis

K-25 Oak Ridge Gaseous Diffusion Plant, or the K-25 Plant

L liter

MDA minimum detectable amount

MDC minimum detectable concentration

mg milligram

ml milliliter

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

ORNL Oak Ridge National Laboratory

pCi picocurie

ppb parts per billion

ppm parts per million

TRU transuranic

U.S.C. United States Code

μCi microcurie

μg microgram

μm micrometer
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 5 of 24

5.1 INTRODUCTION

Site Profiles, which include Technical basis documents (TBDs), are not official determinations made 

by the National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised when additional 

relevant information is obtained about the affected site(s). These documents may be used to assist 

the NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building or group of buildings

that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy facility” as defined in the Energy Employees

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l(5) and (12)). EEOICPA 

defines a DOE facility as “any building, structure, or premise, including the grounds upon which such 

building, structure, or premise is located ... in which operations are, or have been, conducted by, or 

on behalf of, the Department of Energy (except for buildings, structures, premises, grounds, or 

operations ... pertaining to the Naval Nuclear Propulsion Program)” 42 U.S.C. § 7384l(12). 

Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted above, any 

facility that performs or performed DOE operations of any nature whatsoever is a DOE facility 

encompassed by EEOICPA.

For DOE employees with cancer, the DOE facility definition only determines eligibility for a dose 

reconstruction, which is a prerequisite to a compensation decision (except for members of the Special 

Exposure Cohort). The compensation decision for cancer claimants is based on a section of the 

statute entitled “Exposure in the Performance of Duty.” That provision (42 U.S.C. § 7384n(b)) says

that an individual with cancer “shall be determined to have sustained that cancer in the performance 

of duty for purposes of the compensation program if, and only if, the cancer ... was at least as likely 

as not related to employment at the facility [where the employee worked], as determined in 

accordance with the [probability of causation] guidelines established under subsection (c) ....” 42 

U.S.C. § 7384n(b). Neither the statute nor the probability of causation guidelines (nor the dose 

reconstruction regulation) define “performance of duty” for DOE employees with a covered cancer or 

restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” 42 U.S.C. § 7384l(12). 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

(i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”) does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

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

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

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

the Oak Ridge Gaseous Diffusion Plant, also called K-25. The document provides guidance to dose 

reconstructors on input parameters that are specific to employees of the Oak Ridge Gaseous 

Diffusion Plant (K-25), as well as the approach for employees with either missing or no monitoring 

information.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 6 of 24

K-25 began gaseous diffusion enrichment production in 1945. Workers handled mainly uranium 

hexafluoride (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 93% (by weight) of 235U from 

1944 to 1964. After 1964, the highest enrichment was 5%.1

Facilities with significant involvement in RU processing included UF6 feed production facilities, a waste 

ash pulverization and uranium recovery facility, decontamination and uranium recovery facilities, 

facilities fed by waste streams from decontamination facilities, other facilities performing more limited 

uranium recovery and decontamination activities, and the K-25 enrichment cascade, which operated 

in a variety of different configurations over time. In 1985, DOE had an overcapacity of enrichment 

capabilities and placed K-25 on standby status. DOE officially shut down the plant in 1987. 

 The processing of RU, which began in 

1952, affected a broad range of processes and activities in facilities and locations at K-25 (BJC 2000). 

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.

The primary method for monitoring employees for intakes of radionuclides at K-25 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.

Until the mid-1980s, action levels were 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

A review of in-house procedures used to assess the concentration of uranium in urine indicated 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 regarding enrichment, solubility, and the contribution of 

environmental uranium, and because samples were collected at work and during the middle of the 

workweek, meaning that cross-contamination and the inability to unfold soluble from insoluble intake 

fractions contributed to the uncertainty.

. Data are available from 

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

Nonetheless, dose reconstructors can prepare reliable, yet claimant-favorable 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 (see Section 5.8). 

 1 The predominant enrichment level was 3.0%. 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.

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 

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.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 7 of 24

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

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

assessing dose for both monitored and unmonitored employees. Section 5.5 summarizes the existing 

data analysis, and Section 5.6 identifies significant incidents with internal dose potential. 

5.2 SOURCE TERM

The primary mission of K-25 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 93.5% 235U (DOE 2000a). 

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 2000a).

Certain TRU isotopes have been present at K-25 including 237Np and 239Pu. These resulted from the 

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

1969 to 1976, with the exception of 1971 when none of the feed was of reactor origin (Smith 1984, 

p. 9). When this processing occurred, approximately 19% of the feed materials in use at the plant 

were reactor tails (DOE 2000a). The tails included 0.2 ppm of neptunium, and 4.0 ppb of plutonium. 

At K-25, 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, dose reconstructors can use Table 5-1 to derive the isotopic 

fractions associated with each microgram of uranium reported in an analytical result. The default 

isotopic fraction provide in Table 5-1 is related to the typical enrichment used at K-25 over the lifetime 

of the plant operation.

Table 5-1. 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-02 4.90E-03 6.83E-04

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

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

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

a. Sources: ANSI (1995)

Table 5-2 lists the principal radionuclides that comprise the K-25 source term and their absorption 

types (lung solubilities). Table 5-3 provides that information for the various uranium compounds.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 8 of 24

Table 5-2. Principal radionuclides 

found at uranium facilities and 

gaseous diffusion plants.

Nuclide Absorption type

Th-230 M, S

U-234 F, M, S

U-235 F, M, S

U-236 F, M, S

U-238 F, M, S

Pu-238 M, S

Pu-239/240 M, S

Np-237 M

Cm-242 M

Cm-244 M

Am-241 M

Tc-99 F, M

Table 5-3. Absorption classification for some uranium compounds.

a

Compound Chemical notation Absorption classificationb

Uranium hexafluoride UF6 F

Uranyl fluoride UO2F2 F

Uranyl nitrate UO2(NO3)2 F

Uranyl acetate UO2(C2H3O2)2 F

Uranyl chloride UO2Cl2 F

Uranyl sulfate UO2SO4 F

Uranium trioxide UO3 F

Uranium tetrafluoride UF4 M

Uranium oxide U3O8 M

Uranium dioxide UO2 M

Uranium tetroxide UO4 M

Ammonium diuranate (NH4)2 + U2O7 M

Uranium aluminide UAlx S

Uranium carbide UC2 S

Uranium-zirconium alloy Uzr S

High-fired uranium dioxide UO2 S

a. Source: DOE (2000b). 

b. The absorption classification is assumed to be comparable to the ICRP 30 

classification system of D, W and Y. 

If information on the source term to which the claimant was exposed is available, the dose 

reconstructor should use that source term. However, if no source term information is available, the 

values and parameters shown in Table 5-4 provide input to the process. 
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 9 of 24

Table 5-4. Source term summary by location.

a

Location Form

Mass

% 235U

Concentration 

per gram of U

K-1131 Chemical Plant Stream 1 & 2

(1) RU as UO3 to Feed Hopper

(2) UO3 to UO3 Reduction

UO3 0.64 520 ppb 237Np

4.4 ppb *Pub

7.8 ppm 99Tc

170 ppm 236U

K-1131 Chemical Plant Stream 5 Tower Ash 

Disposal

UF4 0.64 13,000 ppb 237Np

440 ppb *Pub

40 ppm 99Tc

100 ppm 236U

K-1131 Chemical Plant Stream 5 Tower Ash 

Disposal Max case for Pu

UF4 0.64 13,000 ppb 237Np

4000 ppb *Pub

40 ppm 99Tc

100 ppm 236U

K-1131 Chemical Plant Stream 6 UF6 to UF6

Collection

UF6 0.64 393.94 ppb 237Np

0.04 ppb *Pub

7.47 ppm 99Tc

170.71 ppm 236U

K-1131 Chemical Plant Stream 7 UF6 to UF6 to 

Process Vent

UF6 0.64 0.00 ppb 237Np

0.00 ppb *Pub

200 ppm 99Tc

200 ppm 236U

K-1131 Chemical Plant Stream 8 RU as UF6 UF6 0.64 394.34 ppb 237Np

0.04 ppb *Pub

7.280 ppm 99Tc

170.88 ppm 236U

K-1131 Chemical Plant Stream 9 UO3

Reduction to Process Vent

UO2 0.64 0.00 ppb 237Np

0.00 ppb *Pub

10 ppm 99Tc

0.00 ppm 236U

ORGDP Stream 12 

UF6 Cylinder Heel

UF6 0.64 26,000 ppb 237Np

4.00 ppb *Pub

70 ppm 99Tc

100.00 ppm 236U

ORGDP Stream 13 

Hanford & Savannah River UF6 Feed 

Autoclave to Cascade

UF6 0.64 132.79 ppb 237Np

0.00 ppb *Pub

6.64 ppm 99Tc

171.60 ppm 236U

ORGDP Stream 14 

PGDP RU to UF6 Feed Autoclave

UF6 0.65 5.00 ppb 237Np

0.00 ppb *Pub

1.96 ppm 99Tc

0.00 ppm 236U

ORGDP Stream 15 

PGDP UF6 Cylinder Heels

UF6 0.65 5.00 ppb 237Np

0.00 ppb *Pub

40 ppm 99Tc

0.00 ppm 236U

ORGDP Stream 16 

PGDP UF6 to Cascade

UF6 0.65 0.00 ppb Np-237

0.00 ppb Pub

1.75 ppm Tc-99

109.29 ppm U-236

ORGDP Stream 18 

UF6 to Cascade

UF6 0.66 30.66 ppb Np-237

0.00 ppb Pub

2.29 ppm Tc-99

86.79 ppm U-236
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 10 of 24

Table 5-4 (Continued). Source term summary by location.a

Location Form

Mass

% 235U

Concentration 

per gram of U

ORGDP Stream 19 

Depleted UF6

UF6 0.30 0.00 ppb Np-237

0.00 ppb Pub

0.00 ppm Tc-99

39.51 ppm U-236

ORGDP Stream 20 

Enriched UF6 Product

UF6 3.0 50.00 ppb Np-237

0.00 ppb Pub

1.77 ppm Tc-99

394.69 ppm U-236

ORGDP Stream 21 

Cascade Accumulation Stream

UO2F2 1.0 130,000 ppb Np-237

4.00 ppb Pub

7500 ppm Tc-99

100.00 ppm U-236

ORGDP Stream 22 

Purge Cascade Stream

UF6 3.0 0.00 ppb Np-237

0.00 ppb Pub

87 ppm Tc-99

100.00 ppm U-236

ORGDP Stream 24 

99Tc Chem Traps Spent Sorbent Stream

UF6 3.0 5.00 ppb Np-237

0.00 ppb Pub

100,000 ppm Tc-99

100.00 ppm U-236

K-1420 Stream 26 Chem Decon K-1420 to 

Effluent Discharge B & C Ponds Stream

UO2F2 1.0 2.00 ppb Np-237

0.02 ppb Pub

200 ppm Tc-99

100.00 ppm U-236

B & C Ponds Sludge UO2F2 0.7 2.00 ppb Np-237

0.02 ppb Pub

200 ppm Tc-99

100.00 ppm U-236

K-1131 Chemical Plant Stream 10 RU as UF6

UO2 Hydrofluorination to Process Vent

UF4 0.64 0.00 ppb Np-237

0.00 ppb Pub

10 ppm Tc-99

0.00 ppm U-236

a. Source: BJC (2000).

b. Pu = total of all plutonium isotopes.

Table 5-5 provides conversion factors specific to each facility for the activity per unit mass for each 

nonuranium radionuclide of interest.

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 99Tc (Smith 1984, Appendix 12; DOE 2000a).

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 K-25 source term. 

Table 5-6 provides a default isotopic distribution to use when only total uranium results are available 

for a particular measurement. 

For K-25, unless specific information to the contrary is available, this analysis assumes a particle size 

of 5 μm activity median aerodynamic diameter (AMAD), as recommended in International Commission 

on Radiological Protection (ICRP) Publication 68 (ICRP 1994, paragraph 5). If information on the 

source term to which the employee was exposed is available, the dose reconstruction should use that 
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 11 of 24

source term. The default isotopic distribution listed in Table 5-6 is based on typical enrichments used 

at K25 over the course of the operation. The presence of other radionuclides not present in the 

uranium decay series is based on operating experience at Paducah Gaseous Diffusion plant.

Table 5-5. Facility-specific radionuclide conversion factors.

Process description

Activity per unit mass (Bq/μg U)

Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Tc-99 U-236

K-1131 Chemical Plant Stream 1 & 2 

(1) RU as UO3 to Feed Hopper

(2) UO3 to UO3 Reduction

1.36E-05 1.39E-06 9.42E-06 2.21E-06 6.71E-05 3.19E-10 4.89E-03 4.06E-04

K-1131 Chemical Plant Stream 1 & 2 

(1) RU as UO3 to Feed Hopper

(2) UO3 to UO3 Reduction Max case for Pu

1.36E-05 1.27E-05 8.56E-05 2.01E-05 6.10E-04 2.90E-09 4.89E-03 4.06E-04

K-1131 Chemical Plant Stream 3 UO2 to 

UO2 Hydrofluorination

1.36E-05 1.39E-06 9.42E-06 2.21E-06 6.71E-05 3.19E-10 4.89E-03 4.06E-04

K-1131 Chemical Plant Stream 4 UF4 to UF4

Fluorination

1.36E-05 1.39E-06 9.42E-06 2.21E-06 6.71E-05 3.19E-10 4.89E-03 4.06E-04

K-1131 Chemical Plant Stream 5 Tower Ash 

Disposal

3.39E-04 1.39E-04 9.42E-04 2.21E-04 6.71E-03 3.19E-08 2.51E-02 2.39E-04

K-1131 Chemical Plant Stream 5 Tower Ash 

Disposal Max case for Pu

3.39E-04 1.27E-03 8.56E-03 2.01E-03 6.10E-02 2.90E-07 2.51E-02 2.39E-04

K-1131 Chemical Plant Stream 6 UF6 to UF6

Collection

1.03E-05 1.27E-08 8.56E-08 2.01E-08 6.10E-07 2.90E-12 4.68E-06 4.08E-04

K-1131 Chemical Plant Stream 7 UF6 to UF6

to Process Vent

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.25E-01 4.78E-04

K-1131 Chemical Plant Stream 8 RU as UF6 1.03E-05 1.27E-08 8.56E-08 2.01E-08 6.10E-07 2.90E-12 4.56E-06 4.08E-04

K-1131 Chemical Plant Stream 9 UO3

Reduction to Process Vent

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.27E-03 0.00E+00

K-1131 Chemical Plant Stream 10 RU as 

UF6 UO2 Hydrofluorination to Process 

Vent

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.27E-03 0.00E+00

ORGDP 12 ORGDP Stream 12 

UF6 Cylinder Heel

6.79E-04 1.27E-06 8.56E-06 2.01E-06 6.10E-05 2.90E-10 4.39E-02 2.39E-04

ORGDP 13 ORGDP Stream 13 

Hanford & Savannah River UF6 Feed 

Autoclave to Cascade

3.47E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 4.16E-03 4.10E-04

ORGDP 14 ORGDP Stream 14 

PGDP RU to UF6 Feed Autoclave

1.31E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.23E-03 0.00E+00

ORGDP 15 ORGDP Stream 15 

PGDP UF6 Cylinder Heels

1.31E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.51E-02 0.00E+00

ORGDP 16 ORGDP Stream 16 

PGDP UF6 to Cascade

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.10E-03 2.61E-04

ORGDP 18 ORGDP Stream 18 

UF6 to Cascade

8.00E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.44E-03 2.07E-04

ORGDP 19 ORGDP Stream 19 

Depleted UF6

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 9.44E-05

ORGDP 20 ORGDP Stream 20 

Enriched UF6 Product

1.31E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.11E-03 9.43E-04

ORGDP 21 ORGDP Stream 21 

Cascade Accumulation Stream

3.39E-03 1.27E-06 8.56E-06 2.01E-06 6.10E-05 2.90E-10 4.70E+00 2.39E-04

ORGDP 22 ORGDP Stream 22 

Purge Cascade Stream

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.45E-02 2.39E-04

ORGDP 24 ORGDP Stream 24 

99Tc Chem Traps Spent Sorbent Stream

1.31E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.27E+01 2.39E-04

K1420 26K-1420 Stream 26

Chem Decon K-1420 to Effluent Discharge 

B&C Ponds Stream

5.22E-08 6.33E-09 4.28E-08 1.01E-08 3.05E-07 1.45E-12 1.25E-01 2.39E-04

B&C Pond Sludge 5.22E-08 6.33E-09 4.28E-08 1.01E-08 3.05E-07 1.45E-12 1.25E-01 2.39E-04
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 12 of 24

Table 5-6. Default isotopic distribution.

Radionuclide nCi/g U

Pu-239 67.5

Am-241 67.5

U-236 0.93

U-235 43.9

U-234 702

U-238 337.5

Np-237 5.4

Th-230 18.9

Tc-99 0.12

5.3 IN VITRO MEASUREMENT METHODS

From the start of plant operations in 1945, samples of urine from workers involved in enrichment 

operations were analyzed for radionuclide content. As time progressed, other employees were 

included in the monitoring program. Spot urine samples (a single void) were used as a screening 

method for indication of uranium intake. The spot samples were collected often in the early years. By 

1950, 24-hr sample collection was instituted at K-25. Spot urine samples show greater variability than 

24-hr samples (Medley, Kathren, and Miller 1994).

The monitoring methodology during the early years of K-25 operations is unknown as of the date of 

this TBD, however existing data show results from a variety of analytical techniques that patterned 

those performed at the Y-12 plant during the same time periods. Because of the plant’s relative 

proximity to the Y-12 plant, and because professional interactions between the two organizations 

were likely, for the purposes of this report a monitoring program history similar to that found at Y-12 is 

assumed. 

As site-specific information becomes available, this assumption may change. In the interim, however, 

it is assumed that fluorometric analysis of urine was the monitoring method of choice beginning in the 

mid-1940’s and continuing until the late 1980’s (ORAU, 2004, Sect. 5.2.2). In addition, uranium 

analysis by electrodeposition and alpha counting may also have been performed in order to assess 

intakes of enriched uranium. After 1989, uranium analysis by alpha spectrometry was the 

predominant in vitro monitoring method.

5.3.1

Table 5-7 lists the in vitro measurement types and detection levels assumed to be applicable during 

various time periods at K-25. The expected intake pattern in most cases is acute. At K-25, airborne 

and surface contamination was typically controlled to prevent intakes, so most would have been the 

result of unexpected releases. It is possible that small, intermittent releases occurred that were not 

immediately detected, so an individual could have had multiple acute intakes. 

Measurement Types and Detection Levels

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

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

Likewise, if a monitoring result does not match with the time periods shown in Table 5-7, the detection 

level given in Table 5-7 for the applicable monitoring method and for the closest time period should be 

used.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 13 of 24

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 K-25) and the MDC from 

that measurement type should be used. 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.

5.3.2

A variety of codes occurs on various urine bioassay records for K-25. A summary of the codes for 

1948 to 1988 is given in Attachment A (Wallace, 2005a.)

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 K-25 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 K-25. 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, sample cross-contamination could 

have occurred during the first collection.

Dietary intakes of uranium and diurnal fluctuations in metals excretion pose a potential problem in 

interpreting urine bioassay results for K-25 workers. Because studies of the average daily uranium 

excretion on Oak Ridge 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. 

No correction for environmental levels of uranium is required for samples analyzed by fluorimetry or 

KPA because the MDC is larger than the correction. 

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

a

Period Measurement type Radionuclide MDCb (mg/L)

1945 to 1957 In-house fluorimetry Total uranium 10 microgram/literc

1958 to 1989 In-house fluorimetry Total uranium 5 microgram/liter

1950 to 1965 Gross alpha counting Enriched uranium 3.3 dpm/ 100 milliliter

1965 to 1989 Gross alpha counting Enriched uranium 1.8 dpm/100 milliliter

1989 to present Alpha spectrometry Isotopic uranium 0.01 dpm/100 milliliter

a. ORAU, 2004, Section 5.2.1.3.

b. MDC = minimum detectable concentration.

c. No non-zero values below a reporting level of 0.01 milligrams of uranium per liter appear in the 

reported data until 1958. Therefore, while the methodology may have been capable of detecting 5 

micrograms of uranium per liter prior to that time, the nominal MDC was rounded up to the 

applicable significant figure. 
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 14 of 24

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).

5.4 IN VIVO MEASUREMENT METHODS

Whole-body counting and other in vivo methods were implemented after 1965. This measurement 

method was used primarily in response to incidents and for assessing the magnitude of insoluble 

material intakes (ORNL, 1981). Even when routine whole-body counting was instituted for certain 

K-25 employees in the late 1960s, the counting frequency was sporadic and seldom greater than 

once per year. A comprehensive listing of scheduled whole-body or lung counts was not available.

5.4.1

At K-25 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 5-9 lists general 

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

Measurement Types and Detection Levels

Table 5-9. In vivo measurement types and detection levels for various periods.

a

Period Equipment

Measurement 

type Radionuclide

MDA

(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 (2000b); Hill and Strom (1993); BJC (1999, Tables 3.2 and 3.5); Scott and West (1967); Bassett (1985, p. 27).

5.4.2

A variety of codes occurs on various in vivo bioassay records for K-25. Summaries of the in vivo

codes for 1965 to 1988 and 1989 to 1991 are shown on Attachments B and C (Wallace, 2005b and 

2005c.)

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 

Instructions for Addressing Possible Interferences and Uncertainties
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 15 of 24

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

should be assumed. 

Radon progeny and medical diagnostic or therapeutic procedures involving radionuclides have 

caused interferences to 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.

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

reporting levels, dose reconstructors 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 K-25 workers could have 

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

foodstuffs). There is evidence neither of 137Cs in the source term at K-25 nor of occupational intakes 

of 137Cs among K-25 workers; therefore, no dose of record should be associated with these 

measurement results.

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

At this time, no information is available about incidents where there was a potential for internal dose. 

If such information should become available, this section will be revised in subsequent versions of this 

document.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 16 of 24

REFERENCES

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

HPS N13.22-1995, Washington, D.C.

Cited References

Bassett, A. C., 1985, In vivo Monitoring Logistics at Goodyear Atomic Corporation, GAT-S-51, 

Goodyear Atomic Corporation, Portsmouth, Ohio.

Bechtel Jacobs, 2002, Internal Dosimetry Technical Basis Document, EH-4501/R2, 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, Tennessee.

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

Gaseous Diffusion Plant Site Report, BJC/OR-584, Oak Ridge, Tennessee.

DOE (U. S. Department of Energy), 2000a, 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, D.C.

DOE (U. S. Department of Energy), 2000b, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2000, Washington, D.C.

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), 1974, Reference Man, Publication 23, 

Pergamon Press, Oxford, England.

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.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 17 of 24

Medley, D. W., R. L. Kathren, and A. G. Miller, 1994, “Diurnal Urinary Volume and Uranium Output in 

Uranium Workers and Unexposed Controls,” Health Physics, volume 67, number 2, pp. 

122-133.

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 for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

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

Cincinnati, Ohio.

ORAU (Oak Ridge Associated Universities), 2004, Technical Basis Document for the Y-12 National 

Security Complex - Occupational Internal Dose, ORAUT-TKBS-0014-5, Revision 00, Oak 

Ridge, Tennessee.

ORNL (Oak Ridge Associated Universities), 1952, Fluorphotometric Analysis of Urine for Uranium, 

Health Physics Progress Report, Y-940, Oak Ridge, Tennessee.

ORNL (Oak Ridge Associated Universities), 1981, Methods and Procedures for Internal Dosimetry at 

ORNL, ORNL/TM-7923, 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.

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, pp. 594-789.

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

Uranium Processing Plant" Union Carbide Corporation, Y-KB-77, Oak Ridge, Tennessee.

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

Carbide Corporation, Paducah, Kentucky. 

Wallace, P, 2005a, K-25 Urinalysis Codes (definitions), 1948 – 1988, personal communication with 

Union Carbide.

Wallace, P, 2005b, K-25 Whole Body Counter Codes (definitions), 1965 – 1988, personal 

communication with Union Carbide.

Wallace, P, 2005c, K-25 Whole Body Counter Codes (definitions), 1989 – 1991, personal 

communication with Union Carbide.

WISE (World Information Service on Energy), 2002, Internet resource affiliated with World Information 

Service on Energy, http://www.antenna.nl/wise/uranium/. 
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 18 of 24

ANSI (American National Standards Institute), 1996, Performance Criteria for Radiobioassay, HPS 

ANSI N13.30, Washington, D.C.

Other References

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

Paducah, Kentucky.

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

Permissible Dose for Internal Radiation, Publication 2, Pergamon Press, Oxford, England.

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

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

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

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

Tennessee.

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

Guide 8.11, Washington, D.C.

ORNL (Oak Ridge Associated Universities), 1961, Procedures and Practices for Radiation Protection, 

Health Physics Manual, Oak Ridge, Tennessee.

Sula, M. J., E. H. Carbaugh, and D. E. Bihl, 1989, Technical Basis for Internal Dosimetry at Hanford, 

PNL-6886, U.S. Department of Energy, Pacific Northwest Laboratory, Richland, Washington.

Schultz, N. B., and A. F. Becher, 1963, “Correlation of uranium alpha surface contamination, air-borne 

concentrations, and urinary excretion rates,” Health Physics, volume 9, 

pp. 901-909.Tockstein, D. J., 1996, Technical Basis for Routine Bioassay Participation, 

Lockheed Martin Energy Systems, Oak Ridge, Tennessee.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 19 of 24

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 in a short period and generally considered instantaneous for 

dose reconstruction purposes. 

bioassay

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

biological material excreted or removed from the body; also called 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 over a long period.

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 are 

reported in units of mass, activity, or potential alpha energy.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 20 of 24

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 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 using 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 nondetection (Type II error) while accepting a probability, ∀, of erroneously 

deciding that a positive (nonzero) 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 that relate 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.

recording level

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

have been maintained.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 21 of 24

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 suspected). 

spot sample

A single void of urine.
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 22 of 24

ATTACHMENT A

URINE BIOASSAY CODES 1948- 1988

Data name Value Description

Payroll W Weekly

Payroll M Monthly

Payroll H Hourly

SamType R Routine

SamType F Follow-up

SamType M Medical

SamType S Supervisory

SamType D Delayed

Recall A First recall

Recall B Second recall

Recall C Third recall

Recall D Fourth recall

Recall T Termination

Freq M Monthly

Freq Q Quarterly

Freq S Semi-Annual

Freq 1 1 month sampling

Freq 2 2 month sampling

Freq 3 3 month sampling

Freq 6 6 month sampling

Freq S Special

Atype 01 Uranium analysis type..aka..contamination code

Atype 02 Fluoride

Atype 03 Mercury

Atype 04 Alpha Activity

Atype 05 Beryllium

Atype 06 Lead

Atype 07 Plutonium-239

Atype 08 Trichloracetic Acid

Atype 09 Nickel

Atype 10 Strontium-90

Atype 11 Neptunium-237

Atype 12 Cadmium

Atype 13 Barium

Atype 14 Carbon-14

Atype 15 Radon

Atype 16 Technetium-99

Atype 17 Copper

Atype 18 Phenol

Atype 19 Lead (blood)

Atype 20 Chromium

Atype 21 Iron

Atype 22 Benzene (phenol)

Atype 23 Magnesium

Atype 24 Calcium

Atype 25 Molybdenum

Atype 26 Manganese

Atype 27 Cobalt

Atype 28 Zinc

SamCode I Industrial hygiene

SamCode H Health physics

Exc 1 Used when result exceeded size of format. Note(only occurred once)

Pdate Prior date

Ptime Prior time

Vsam Volume of sample

Tsam 1 AM time sample was taken

Tsam 2 PM

Tsam 0 Not used
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 23 of 24

ATTACHMENT B

IN VIVO BIOASSAY CODES 1965 TO 1988

Data 

name Value Description

Surface 1 No surface contamination if in first position

Surface 2 or 3 Surface contamination if in first position

Surface Second through sixth are meter readings

Atype 1 Uranium 235

Atype 3 Thorium

Atype 4 U-238

Atype 6 Thorium and U-238

Atype 7 All codes

Mtype 01 Unknown or Mixed U enriched

FTB Front to back ratio one decimal place

U235LB U-235 lung burden in microgram

TLB Thorium Lung Burden in milligram

CC Cesium count in nanocuries

PC Potassium (K-40) count in grams

U238LB U238 Lung Burden in milligrams

AC Actinium count in nanocuries(one decimal place)

LC Lead 212 count in nanocuries(one decimal place)

TC Technetium 99 count in microcuries (one decimal place)
Document No. ORAUT-TKBS-0009-5 Revision No. 01 Effective Date: 01/06/2006 Page 24 of 24

ATTACHMENT C

IN VIVO BIOASSAY CODES 1988 TO 1991

Name of 

field Value Description

Plant Plant

Plant K K25

Plant Y Y12

Plant X X10

DOB Date of birth

CountDate Date of count

Dept_No Department number

EAC_Code EAC Code

HP_Area_Code HP Area Code

ReasonCode Reason Code

ReasonCode 1R Incident routine

ReasonCode 1S Incident acute

ReasonCode 2R Follow-up routine

ReasonCode 2S Follow-up acute

ReasonCode 3 Routine

ReasonCode 4 Background

ReasonCode 5 Other

ReasonCode 6 Control

ReasonCode 7 Random

ReasonCode 8 Start date

ReasonCode 9 Stop date

ReasonCode 0 Acute

ReasonCode 1 Old value:::Incident

ReasonCode 2 Old value:::Followup

SampleReason Reason for sample

WBC_Type Type of whole body count

WBC_Type LUN Lung count

WBC_Type O Other

WBC_Type RBC Red blood cells

WBC_Type THY Thyroid count

WBC_Type WBL Whole body lung

WBC_Type WBC Whole body count