ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for Portsmouth Gaseous 

Diffusion Plant – Occupational Internal Dose 

Document Number:

ORAUT-TKBS-0015-5 

Effective Date: 11/24/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 27

Subject Expert: Paul J. Demopoulos

Document Owner

Approval: Signature on File Date: 11/03/2004

Mark D. Notich, TBD Team Leader

Approval: Signature on File Date: 11/05/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 11/03/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 11/24/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

3

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

5.0 Introduction – Occupational Internal Dose................................................................................... 6

5.1 In Vitro Minimum Detectable Activities (MDAS) and Counting Methods ...................................... 7

5.1.1 In Vitro Urinalysis Records ................................................................................................. 7

5.1.2 In Vitro Methods for Individual Radionuclides ..................................................................... 9

5.1.2.1 In Vitro Bioassay for uranium.....................................................................................9

5.1.2.2 Inductively Coupled Plasma Mass Spectrometry (ICP/MS)........................................ 9

5.1.2.3 Frequency of Urine Bioassays ...................................................................................9

5.1.2.4 Work Restriction ...................................................................................................... 10

5.1.2.5 Uranium MDCs or Reporting Levels ........................................................................ 11

5.1.2.6 Source Term and Isotopic Determination................................................................. 11

5.1.2.7 In Vitro Bioassay for Fission Products (99Tc) ............................................................ 17

5.1.2.8 In Vitro Bioassay for 237Np ....................................................................................... 18

5.1.2.9 In Vitro Bioassay for Trivalent Actinides (Americium)............................................... 18

5.1.2.10 In Vitro Bioassay for Plutonium................................................................................ 18

5.1.2.11 In Vitro Bioassay for Thorium................................................................................... 18

5.1.2.12 In Vitro Bioassay for Protactinium............................................................................ 18

5.2 Correcting for Urinalysis Volume............................................................................................... 19

5.3 In Vivo MDAs, Counting Methods and Reporting Practices....................................................... 19

5.4 Absorption Types......................................................................................................................21
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5.5 Workplace Air Sampling Data as Applicable ............................................................................. 22

5.6 Interference and Uncertainty..................................................................................................... 23

5.6.1 Contamination of Samples ............................................................................................... 23

5.6.2 Uncertainty.......................................................................................................................23

5.7 Assessment of Intake for Monitored and Unmonitored Workers................................................ 24

5.7.1 Monitored Workers with Measurable Intakes .................................................................... 24

5.7.2 Monitored Workers with Nothing Detected in Bioassay..................................................... 24

5.8 In Vitro Urinalysis and In Vivo Lung Counting Data Code Information....................................... 24

References ........................................................................................................................................26

LIST OF TABLES

Table

5.1.1-1 In Vitro Measurement Frequencies, Measurement Types and MDCs for 

Urinanalysis.....................................................................................................................

Page

7

5.1.1-2 Uranium Source Term Information................................................................................... 8

5.1.2.3-1 Recall Limits..................................................................................................................10

5.1.2.6-1 Reactor Returns, RU or Tails Fed to the Cascade......................................................... 11

5.1.2.6-2 Major Uranium 0Facilities at PORTS............................................................................. 12

5.1.2.6-3 Uranium TRU and 99Tc Source Term Information .......................................................... 13

5.1.2.6-4 Lung Absorption Type and Fractional Activity of Radionuclides by Facility .................... 14

5.1.2.12-1 Summary of 99Tc, 237Np, 241Am, 239Pu, 231Th and 234mPa Urinalysis ................................ 19

5.3-1 MMES Chest Counter /1965 –1990 MDAs and Restriction Levels................................. 20

5.3-2 In Vivo Summary (1965 – 1985).................................................................................... 20

5.3-3 Helgesson Chest Counter /1991-1995........................................................................... 20

5.8-1 Internal Dosimetry Record Codes.................................................................................. 25
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 3 of 27

 

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

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

Gaseous Diffusion Plant – Occupational Internal 

Dose. Initiated by Mark D. Notich.

Draft 07/23/2004 00-B Incorporates internal review comments. Initiated 

by Mark D. Notich. 

Draft 09/13/2004 00-C Incorporates NIOSH review comments. Initiated 

by Mark D. Notich.

Draft 10/05/2004 00-D Incorporates NIOSH review comments. Initiated 

by Mark D. Notich.

Draft 10/07/2004 00-E Incorporates NIOSH review comments and Task 

5 comments. Initiated by Mark D. Notich.

Draft 11/02/2004 00-F Incorporates additional NIOSH and Task 5 review 

comments. Initiated by Mark D. Notich.

11/24/2004 11/24/2004 00 First approved issue. Initiated by Mark D. Notich.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 4 of 27

ACRONYMS AND ABBREVIATIONS

241Am Americium-241

237Np Neptunium-237

238,239,240Pu Plutonium-238, -239, -240

234mPa Protactinium-234

99Tc Technetium-99

228,230,231,232,234Th Thorium-228, -230, -231, -232, -234

201Tl Thallium-201

234,235,236,238U Uranium-234, -235, -236, -238

AEC Atomic Energy Commission

amu atomic mass unit

AMAD Activity Median Aerodynamic Diameter

Bq Becquerel

BJ Bechtel Jacobs Corporation

CsI Cesium Iodide

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

DR Dose Reconstructor

ERDA Energy Research and Development Administration

EU enriched uranium

g gram

GAT Goodyear Atomic Corporation Technical Document

HEU highly enriched uranium

hr hour

ICRP International Commission on Radiological Protection

ICP/MS Inductively Coupled Plasma Mass Spectrometry

IL investigation level

KPA kinetic phosphorescence analysis

keV kilovolt electric

LSC liquid scintillation counting

μCi microcurie

μg microgram

MDA minimum detectable activity or, for elemental uranium, minimum detectable amount

MDC minimum detectable concentration

mg milligram

ml milliliter

MMES Martin Marietta Energy Systems
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 5 of 27

M Molar

MTU metric tons uranium

NaI Sodium Iodide

Nat U natural uranium

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health 

NO3 nitrate ion

ORNL Oak Ridge National Laboratory

PAS portable air samples

PGDP Paducah Gaseous Diffusion Plant 

PORTS Portsmouth Gaseous Diffusion Plant

RU recycled uranium

SA specific activity

TBD Technical Basis Document

TRU transuranic

UO2 uranium dioxide

UO3 uranium trioxide

U3O8 uranium oxide

UF4 uranium tetrafluoride 

UF6 uranium hexafluoride

USEC United States Enrichment Corporation
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 6 of 27

5.0

Technical Basis Documents (TBDs) 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.

INTRODUCTION – OCCUPATIONAL INTERNAL DOSE

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. § 7384l (5) and (12)). 

Portsmouth Gaseous Diffusion Plant (PORTS) operations, which involved several processes of the 

nuclear enrichment cycle, played a significant role in the U.S. energy program and the U.S. 

Department of Defense nuclear fuel program. These processes included nuclear fuel enrichment; 

radiochemical separations; refining, finishing, and storing uranium; and handling the associated 

radioactive waste. 

PORTS workers, especially those employed during the production decades of the 1950s and 1960s, 

have been exposed to radiation types, absorption types, different enrichments and radionuclide 

matrixes and energies associated with nuclear energy development processes. PORTS used facility 

and individual worker monitoring methods to measure and control radiation exposures. Evaluations 

are difficult because the extensive scope of facility, process, and worker information relevant to an 

individual worker’s potential dose might involve many years or even decades after employment. 

The Portsmouth site internal dosimetry program started when the site initiated operation in 1954. 

Uranium isotopes were of primary concern. The primary method for monitoring of employees for 

radionuclide intake was urine bioassay. Fluorimetry for elemental uranium analysis and gas flow 

proportional counting for gross alpha counting was the cornerstone of the monitoring methodology for 

PORTS until very recent times. Technetium-99 (99Tc) routine monitoring began in 1965 in vivo and 

1975 in vitro. Transuranics have not been monitored routinely with the exception of Neptunium-237 

(

237Np). 

The radionuclides of concern for internal dosimetry include the following elements and isotopes:

• 237Np

• 99Tc

• Uranium (234U, 235U, 236U, 238U and mixtures) depleted uranium (DU), highly enriched uranium 

(HEU), enriched uranium (EU), natural uranium (Nat U)

• Plutonium (238Pu, 239Pu, 240Pu)

• Thorium ( 228Th, 230Th, 231Th, 232Th, 234Th)

• Protactinium (234mPa)

• Americium (241Am)

These radionuclides are those identified by PORTS internal dosimetry program and the internal dose 

technical basis document (Hill and Strom, 1993). Only uranium, 99Tc and 237Np were monitored 

routinely.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 7 of 27

5.1 IN VITRO MINIMUM DETECTABLE ACTIVITIES (MDAS) AND COUNTING METHODS

An example of probable decision level determination for urinalysis follows. These equations were 

used since about 1992. 

 

DL = 

4.65

TREVA (Equation 5-1)

 2 X (CBackground)

1/2 + 3

Where

DL = decision level (dpm/L)

CBackground = background counts in the region of interest

T = count time (min)

R = recovery fraction

E = average detector efficiency

V = sample volume (L), and

A = the alpha abundance for the radionuclide in question

The decision level is the level at which activity is considered present in a sample with a 95% 

confidence level.

MDA = 

TREVA (Equation 5-2)

 4.65 X (CBackground)

1/2 + 3

5.1.1

The analysis records have been stored in several data electronic bases over PORTS site history. The 

databases include baseline (new hire), routine, special, recall and termination measurements. See 

section 5.8 for a description of database codes. The current database contains all of the information 

from the previous databases from 1954 to the present. The MDAs for the isotopes analyzed over the 

time frames are listed in Table 5.1.1-1. Activity fractions for uranium isotopes are listed in Table 5.1.1-

2.

In Vitro Urinalysis Records 

MDA values, flags or investigation levels (ILs) were used to determine the recorded values. The recall 

counts were counted for the same amount of time as the other counts. No change in MDA values 

would have occurred for these counts. 

Table 5.1.1-1. In Vitro Measurement Frequencies, Measurement Types and MDCs for Urinalysis

Period Frequency Radionuclide MDC

Record 

Level

Urinalysis 

Method Volume

1954-

3/31/1995

Weekly to monthly, weekly 

for suspected exposures > 

IL, monthly or bi monthly 

routine by work loc, 

function, etc.

Uranium Total 0.005 mg/L 0.01 mg/L Fluorimetry. Spot or 24-hr 

Simulated

1954-

3/31/1995

Weekly to monthly, weekly 

for suspected exposures > 

IL, monthly or bi monthly 

routine by work loc, 

function etc. 

Alpha Total 10 dpm/L

 

10 dpm/L Ion Exchange/ 

Proportional 

Counter

Spot or 24-hr 

Simulated

4/1/95-

Present 

Weekly to monthly Did not 

measure total alpha 

anymore. ICP/MS replaced 

IC/Prop as well. Performed 

U235, 

U238

0.047 μg/L

0.300 μg/L

0.1 pCi/L 

(all RNs)

ICP/MS 24-hr 

Simulated
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 8 of 27

Period Frequency Radionuclide MDC

Record 

Level

Urinalysis 

Method Volume

Bechtel Jacobs 

Corporation (BJ) personnel 

until about 1998 when BJ 

decided to send samples 

to Wash or Oak Ridge 

National Laboratory 

(ORNL).

10/0/98-

6/30/01

Weekly to monthly

Quanterra/BJ employees

24-hr sample.

Uranium Total 0.06 μg/L Kinetic 

Phosphoresce

nce Analysis 

(KPA)

24-hour 

Simulated

7/1/01-

Present 

Weekly to monthly

ORNL/BJ employees

24-hour sample.

U234, 

U235 &

U238

0.02 dpm/L, 

0.02 dpm/L, 

0.02 dpm/L 

Alpha 

Spectrometry

24-hour 

Simulated

1/1/1994-

Present 

Weekly to monthly

BJ & United States 

Enrichment Corporation 

(USEC).

237Np 0.02 dpm/1 

liter ORNL

Alpha 

Spectroscopy

24-hour 

Simulated

1/1/1994-

Present

Weekly to monthly

BJ & USEC.

241Am 0.02 dpm/1 

liter ORNL

Alpha 

spectroscopy

24-hour 

Simulated

1/1/1994-

Present

Weekly to monthly 

BJ & USEC.

239Pu 0.02 dpm/1 

liter ORNL

Alpha 

spectroscopy

24-hour 

Simulated

1/1/1994-

Present 

Weekly to monthly

BJ & USEC.

231Th 8000 dpm/1 

liter 

Gas flow 

Proportional & 

LSC 

24-hour 

Simulated

1/1/1994-

Present 

Weekly to monthly

BJ & USEC.

234mPa 8000 dpm/1 

liter 

Gas flow 

Proportional & 

LSC 

24-hour 

Simulated

Source: BJC, 1999; Mayfield, 1995; Hill and Strom, 1993; GAT, 1979; GAT, 1966a; GAT, 1956. 

Table 5.1.1-2. Uranium Source Term Information

Uranium Source 

Term Reference

Specific Activity

pCi/μg

Activity Fractions 234U 235U 236U 238U

Natural Uranium IMBAa 0.683 0.489 0.023 - 0.489

93.% IMBAa 68.1 0.968 0.030 0.002 0.0003

3.5% Typical 

PORTS

IMBAa 2.20 0.818 0.034 - 0.147

2% HPSb 1.20 0.648 0.041 0.0009 0.311

Typical DU IMBAa 0.402 0.155 0.011 0.0005 0.834

Recycled Uranium Hanfordc 0.910 0.563 0.023 0.048 0.365

Uranium Source 

Term Reference

Specific Activity

pCi/μg

Specific Constituent Activity in Mixture 

(uCi/g, nCi/mg, or pCi/ug)

Natural Uranium IMBAa 0.683 0.334 0.016 - 0.334

93.% IMBAa 68.1 65.9 2.04 0.136 0.020

3.5% Typical 

PORTS

IMBAa 2.20 1.80 0.075 - 0.323

2% HPSb 1.20 0.778 0.049 0.001 0.373

Typical DU IMBAa 0.402 0.062 0.004 0.0002 0.335

Recycled Uranium Hanfordc 0.910 0.563 0.023 0.048 0.365

a. IMBA computer software, and World Information Service on Energy (WISE) Uranium Project.

b.ANSI HPS N13.22-1995, Bioassay Programs for Uranium, (An American National Standard), October, 1995.

c. Hanford TBD Table 5.2.5-1 ORAUT-TKBS-0006-5.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 9 of 27

5.1.2

5.1.2.1 In Vitro Bioassay for uranium

In Vitro Methods for Individual Radionuclides

Urinalysis for total alpha activity has been conducted since PORTS began operation in 1954. Urine 

samples were evaporated to dryness with an excess of concentrated nitric acid and ignited over a 

blast burner. The white salt residue was dissolved in distilled water and a double calcium precipitation 

was made with 0.4 M ammonium oxalate. The uranium was electroplated onto nickel disks from the 

combined supernates. The plated disks were ignited and the alpha activity was measured on a 

proportional counter in α counts/min / 100 ml urine {1 dpm/100 ml minimum detectable concentration 

(MDC)}. 100 ml of urine was needed for this technique. A result of 5 dpm/100 ml was considered the 

reporting level for the time-period of 1954–1993 (GAT, 1955; GAT, 1985a; Hill and Strom, 1993). 

Review of claim records reveals that sometimes values less than 5 dpm were recorded, probably 

down to the MDC of 1 dpm/100 ml. 

To determine total uranium mass per unit volume a fluorimetry procedure was used. A fusion of 

uranium salts with sodium fluoride gives a characteristic yellow green fluorescence under UV light. 

The intensity of the fluorescence is proportional to the concentration of the uranium in the fused disk. 

The fluorescence is measured with a fluorimeter that has been calibrated with a series of standard 

uranium disks. Results exceeding 0.01 mg U/liter were re-run. This 0.01 mg/liter was considered a 

positive result from the time-period of 1954–1993 (GAT, 1956; Hill and Strom, 1993). 

The calculation was as follows:

Voltmeter reading x fluor scale x 1000

5 x 0.2 x 1000

 = mg uranium/ liter (Equation 5-3)

5.1.2.2 Inductively Coupled Plasma Mass Spectrometry (ICP/MS)

As of 3/31/95, ICP/MS is being utilized for the analysis of 235U and 238U in urine. Urine is digested and 

wet oxidized with strong nitric and hydrochloric acids to solubilize uranium and to destroy the organic 

matter. Uranium is selectively separated from the chloride salts by an anion exchange resin and is 

extracted with dilute nitric acid. The uranium isotopes are then measured by ICP/MS.

The MDA for this process is 0.1 pCi/liter. The method can analyze directly for 235U and 238U. 234U 

cannot be directly measured because of the 1 amu resolution of ICP/MS but can be estimated based 

upon knowledge of the enrichment. A 0.2 μg, standard with 3.0% by weight 235U is counted daily. 

Typical results from ICP/MS are a 3.2% enrichment determination with a 1.8% relative standard 

deviation (Mayfield, 1995).

5.1.2.3 Frequency of Urine Bioassays

The frequency of urine bioassays could vary considerably at PORTS. As stated in one version of the 

urine bioassay procedure (circa 1971): “All individuals who may come in contact with toxic materials 

are placed on a routine urinary program and scheduled on a 1 week, 1 month, 3 month or 6 month 

frequency depending on the following“ (shown below). A number of categories of urinalysis include 

routine, recall and a supervision requested analysis. Exact criteria were not established. It was based 

upon the coordinated judgment of the employee’s supervision and the Industrial Hygiene and Health 

Physics Departments (GAT, 1966a, p2).
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 10 of 27

Type I - Routine

A. Potential Exposure

(a) High Daily – 1 week

(b) Daily – 1 month

(c) Weekly – 3 months

(d) Occasional – 6 months

B. Concentration involved

(a) Planned and Unplanned High Airborne – 1 month

(b) High (or high assay) – 1 month

(c) Moderate – 3 months

(d) Low – 6 months

C. Past Medical History

(a) Several Positive Exposures – 1 month

(b) Few Exposures – 3 months

(c) Infrequent or None – 6 months (GAT, 1966a, p3)

Type II - Recall

Recall samples were obtained to verify negative results, to replace a sample that was misplaced, 

spilled, or lost due to laboratory error, to replace an inadequate sample or whenever results indicate 

values above the recall limits: 

Table 5.1.2.3-1. Recall Limits

Recall in Uranium

(mg/l)

Alpha

(dpm/100 ml)

1 month 0.01 5.0

1 week 0.02 9.0

Immediately 0.06 30.0

 Source: GAT, 1966a, p4.

Type III – Supervision’s request

Any time an individual is exposed to high concentrations of materials such as purge gas release, etc., 

a special urine sample should be submitted. 

5.1.2.4 Work Restriction

Whenever a single sample or a series of samples indicate positive results equaling 0.3 mg/liter or 45 

dpm/100 ml followed by a Monday sample of 0.06 mg U/l or 9 dpm/100 ml, the individual was 

considered for a work restriction. All individuals were then asked to submit daily samples until urinary 

values fell below 0.06 mg U/l and 9.0 dpm/100ml for two consecutive Mondays (GAT, 1966a, p 9-10). 

A new list of personnel required to receive a urinalysis was compiled every January. Routine samples 

were submitted on Monday of every week and recorded on form A-551. A special sample was given 4 

hours after exposure and one voiding for suspected inhalation incidents. Total uranium analysis 

required 2 ml and a total alpha analysis required a 100-ml volume. Until around 1995 spot samples 

were the norm at PORTS. One out of ten samples were controls.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 11 of 27

5.1.2.5 Uranium MDCs or Reporting Levels

There has been a variety of methods used historically to analyze for uranium at PORTS. These 

methods and their associated detection capabilities are summarized in Table 5.1.1-1.

5.1.2.6 Source Term and Isotopic Determination

In order to use either the total alpha or elemental uranium results, use Table 5.1.1-2 to determine the 

actual radionuclide isotopic values. For gaseous diffusion enriched uranium, the approximate specific 

activity (SA) of a given enrichment over a 1-50% range is (DOE, 2000c):

Specific Activity of Enriched Uranium = (0.4 + 0.38E + 0.0034E2

) x 10-6 Ci/g (Equation 5-4)

Where E = % 235U by weight enrichment > or = 0.72

HEU production at PORTS ended in 1978. However, reintroduction of HEU from various sources to 

the cascade occurred up to 1998 (see Table 5.1.2.6-1 for further information). The increase in SA is 

mostly due to 234U not 238U replacement with 235U. The above equation is not valid for recycled 

uranium (RU). (DOE, 2000c). 

Table 5.1.2.6-1. Reactor Returns, RU or Tails Fed to the Cascade

Fiscal Year

Amount 

fed (MTU)

Enrichment 

(%235U) Source Remarks

1955 105.8 0.64 – 0.68 Paducah May – Sept. 1955

1956 54.5 0.64 – 0.68 Paducah

1956 293.4 0.64 – 0.68 Oak Ridge

1957 6.2 0.64 – 0.68 Paducah

1958 64.2 0.64 – 0.68 Paducah

1970 168.1 0.64 – 0.68 Paducah Oct. & Nov. 1969

1974 398.8 0.64 – 0.68 Paducah Jan. 1974

1974 – 1978 1.86 2 – 50 PORTS Oxide Conversion

1968 – 1977 0.15 78 – 80 Division of International Affairs

1977 – 1998 0.15 78 – 97 Babcock & Wilcox

1969 – 1993 0.07 78 Atomic Energy Commission (AEC) Office of 

Safeguards & Materials Management

1997 – 1998 1.10 56 – 82 France

Source: Table 2.2.2.5-1 (BJC, 2000, p. 22). 

PORTS mission was to enrich uranium in the form of UF6 for use in domestic and foreign commercial 

power reactors from slightly enriched uranium of roughly 2.5% 235U to about 5% enrichment of 235U. 

Up until about 1992 PORTS enriched uranium to about 93% enrichment of 235U. PORTS also 

accepted used or recycled uranium from domestic and foreign sources for enrichment. Compounds of 

uranium included UO2, UO3, U3O8, UF4 and UF6. The primary nuclides of concern for the plant are 

238U, 235U and 234U. The progeny of domestic interest for these radionuclides includes 230Th and 

234mPa. The major facilities where uranium was processed are listed in Table 5.1.2.6-2. 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 12 of 27

Table 5.1.2.6-2. Major Uranium Facilities at PORTS

Building 

No. Name

Dates of 

operation Activities

X-326 Gaseous Diffusion Process Bldg. 1954–1991 High Assay Product 

X-330 Gaseous Diffusion Process Bldg. 1954–2001 Intermediate Process & Tails Withdrawal

X-333 Gaseous Diffusion Process Bldg. 1954–2001 Initial Enrichment & Reactor Product

X-342A Fixed Feed Facility 1954–2001 Feed UF6 to Process Line

X-342 Fluorine Generation Facility 1954–2001 Generate Elemental F2 for Converters

X-343 Fixed Feed Facility 1954–2001 Feed UF6 to Process Line

X-344 UF6 Feed Manufacturing Plant 1958–1962 Conversion of UF4 to UF6

X-345 Special Nuclear Material Storage 1978–2003 Highly Enriched Uranium Storage

X-700 Maintenance Building 1954–2003 Large Component Repairs

X-705 Decontamination & Cleaning Bldg. 1954–2003 Equipment Wash & Uranium Recovery

X-705E Oxide Conversion Plant 1957–1978 Convert U3O8 to UF6

X-710 Labs, Electrical and I&C Shops 1954–2003 Testing, Calibration, & Repair

X-720 Compressor Shop 1954–2003 Disassembly & Repair of Compressors

X-744G Smelter & Aluminum Recovery 1954–1978 Recover Aluminum from Scrap

Source: DOE, 2000b, p. 16.

Because of excretion of uranium from natural sources, the MDAs for natural or depleted uranium 

listed above should not be used to determine occupational intakes. Based on the Savanah River TBD 

report and an upper limit at 99% confidence, a 0.15 dpm/L DL was set based on analytical noise and 

background. (ORAUT-TKBS-0003-05, rev.1) 

Hanford studies from 1985 and 1990 indicate a uranium urinary excretion of 0.05 to 0.5 μg/day in the 

Hanford area. The 1985 study (Sula et al, 1991) indicated that at the 99.9th percentile the daily 

uranium output was 0.2 μg/day (ORAUT-TKBS-0006-05, rev.1, p70). Any result greater than this 

value was considered to contain occupationally derived uranium. PORTS/Paducah Gaseous Diffusion 

Plant (PGDP) adopted this value according to the 1993 Internal Dosimetry PORTS/PDGP TBD (Hill 

and Strom, 1993). Uranium source term or isotopic breakdown information is presented in Table 

5.1.1-2. 

Because studies of the average daily uranium excretion on Portsmouth residents do not appear to be 

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 

activity 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, KPA, 

or total alpha because the MDAs were larger than the correction.

No record exists of attempts to characterize the size of particles at PORTS. For dose reconstruction 

purposes the default 5-μm AMAD particle size should be used. 

Certain transuranic (TRU) radionuclides are present at PORTS including 239Pu and 237Np. These 

come from reactor tails or RU that were processed at PORTS. Refer to Table 5.1.2.6-1 “Reactor 

returns, RU or tails fed to the cascade”. Only about 2% of the total feed was reactor tails. Table 

5.1.2.6-3 gives the TRU and 99Tc activities in dpm per gram of reactor tail uranium per year. This data 

is based upon the PORTS mass balance report (BJC, 2000). Note the information does not break 

down the TRU or 99Tc information by facility or building. This information is presented in Table 5.1.2.6-

4, “activity fraction” column and is based on air sampling results from 1992–1998, and thus is only a 

snap shot of the total time period of the PORTS operations. 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 13 of 27

Table 5.1.2.6-3. Uranium TRU and 99Tc Source Term Information

Year

237Np 99Tc 238Pu 239Pu 240Pu

dpm/gRU dpm/gRU dpm/gRU dpm/gRU dpm/gRU

1955 1.20E+02 5.16E+01 0.00E+00 0.00E+00 0.00E+00

1956 5.54E+01 2.08E+01 0.00E+00 0.00E+00 0.00E+00

1957 1.08E+02 3.38E+01 0.00E+00 0.00E+00 0.00E+00

1958 9.52E+01 4.16E+01 0.00E+00 0.00E+00 0.00E+00

1959 9.52E+01 5.35E+01 0.00E+00 0.00E+00 0.00E+00

1960 9.52E+01 5.95E+01 0.00E+00 0.00E+00 0.00E+00

1961 9.52E+01 6.54E+01 0.00E+00 0.00E+00 0.00E+00

1962 9.52E+01 7.14E+01 0.00E+00 0.00E+00 0.00E+00

1963 9.52E+01 8.33E+01 0.00E+00 0.00E+00 0.00E+00

1964 9.52E+01 8.92E+01 0.00E+00 0.00E+00 0.00E+00

1965 9.52E+01 9.52E+01 0.00E+00 0.00E+00 0.00E+00

1966 9.52E+01 1.01E+02 5.95E-02 1.77E-01 5.26E-01

1967 9.52E+01 1.07E+02 5.95E-02 1.77E-01 5.26E-01

1968 1.04E+02 1.13E+02 5.95E-02 1.77E-01 5.25E-01

1969 1.09E+02 6.31E+01 5.74E-02 8.24E-02 1.18E-01

1970 1.09E+02 6.60E+01 5.74E-02 8.24E-02 1.18E-01

1971 1.09E+02 6.74E+01 8.61E-02 1.24E-01 1.77E-01

1972 1.09E+02 6.89E+01 8.61E-02 1.23E-01 1.77E-01

1973 1.09E+02 7.45E+01 8.59E-02 1.23E-01 1.76E-01

1974 6.44E+01 9.16E+01 1.00E-01 1.44E-01 2.06E-01

1975 6.58E+01 9.30E+01 1.00E-01 1.43E-01 2.05E-01

1976 6.76E+01 9.02E+01 3.24E-01 4.57E-01 6.44E-01

1977 6.90E+01 8.45E+01 3.24E-01 4.56E-01 6.43E-01

1978 7.04E+01 8.17E+01 3.24E-01 4.56E-01 6.43E-01

1979 7.04E+01 8.03E+01 3.24E-01 4.56E-01 6.43E-01

1980 7.04E+01 7.89E+01 3.24E-01 4.56E-01 6.43E-01

1981 7.04E+01 7.75E+01 3.24E-01 4.56E-01 6.43E-01

1982 7.04E+01 7.61E+01 3.24E-01 4.56E-01 6.43E-01

1983 7.04E+01 7.46E+01 2.25E-01 3.17E-01 4.47E-01

1984 6.90E+01 7.32E+01 2.25E-01 3.17E-01 4.47E-01

1985 6.90E+01 6.90E+01 2.25E-01 3.17E-01 4.47E-01

1986 6.90E+01 6.76E+01 1.83E-01 2.58E-01 3.63E-01

1987 6.90E+01 6.62E+01 1.83E-01 2.58E-01 3.63E-01

1988 6.90E+01 6.48E+01 1.83E-01 2.58E-01 3.63E-01

1989 6.90E+01 6.34E+01 1.83E-01 2.58E-01 3.63E-01

1990 6.90E+01 6.34E+01 1.83E-01 2.58E-01 3.63E-01

1991 6.90E+01 6.34E+01 1.83E-01 2.58E-01 3.63E-01

1992 6.90E+01 6.20E+01 1.83E-01 2.58E-01 3.63E-01

1993 6.90E+01 6.20E+01 1.83E-01 2.58E-01 3.63E-01

1994 6.90E+01 6.20E+01 1.83E-01 2.58E-01 3.63E-01

1995 6.76E+01 6.06E+01 1.83E-01 2.58E-01 3.63E-01

1996 6.62E+01 6.06E+01 1.69E-01 2.38E-01 3.35E-01

1997 6.48E+01 4.93E+01 1.69E-01 2.38E-01 3.35E-01

1998 6.48E+01 4.79E+01 1.41E-01 1.98E-01 2.79E-01

Based upon data from BJC, 2000, Table 3.2-1 and Figure 5.1-1.

If specific source term information to which the employee has been exposed is available, the dose 

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

values and parameters in Table 5.1.2.6-4 provide an adequate input to the process. Note that the 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 14 of 27

activity fraction and contaminant ratio to uranium columns are based upon 1993 through 1999 PORTS 

air sampling data and is therefore a snapshot of the temporal radionuclide matrix of the facilities.

If isotopic data are available, it can be used to determine if a uranium urinalysis result represents an 

occupational uptake. The following rules apply:

• If 235U or 236U are detected, then 234U must also be detected.

• If both 234U and 238U are detected, the uranium should be considered from occupational 

sources if the 234U/238U ratio is outside a range from 1 to 3, regardless of the total uranium 

excretion rate.

• If 238U is detected, the uranium should be considered from an occupational source if the ratio 

of the 234U critical level to the 238U result is less than 1 and the total uranium excretion is 

greater than 0.43 μg/d.

• If 234U is detected, the uranium should be considered from an occupational source if the ratio 

of the 234U result to the 238U critical level is greater than 3 (ORAUT-TKBS-0003-05, rev.1, p71).

Table 5.1.2.6-4. Lung Absorption Type and Fractional Activity of Radionuclides by Facility

Facility Compound RN Suggested 

Absorption Typed 

Absorption Type 

Range

Activity 

Fractionf,h

Intake Activity 

Fraction Relative 

to Total Uraniumj 

X-326 UF6, U3O8, and uranyl 

fluoride (UO2F2) are 

likely (Hill and Strom 

1993, p14.7).

234Ua F F-S 9.12 × 10-1

235 Ua F F-S 3.16 × 10-2

236Ua F F-S 2.64 × 10-3

238Ua F F-S 5.33 × 10-2

99Tcg F F-M 6.43 × 10-3 6.43 × 10-3

237Np M M 9.58 × 10-5 9.58 × 10-5

238Pub S M-S 1.79 × 10-5

239Pub S M-S 1.79 × 10-5 5.37 × 10-5

240Pub S M-S 1.79 × 10-5

241Amb S M-S 1.79 × 10-5 1.79 × 10-5

228Thi S F-S 1.15 × 10-4 1.15 × 10-4

230Th S F-S 3.47 × 10-4 3.47 × 10-4

231Th S F-S Trace 232Thi S F-S 3.6 × 10-6 3.6 × 10-6

234Thc M F-S Trace 2134mPac M M-S (In equilibrium with 234Th)

X-330 UF6, U3O8, and uranyl 

fluoride (UO2F2) are 

likely (High and 

Strom 1993, p14.7).

234Ua F F-S 7.44 × 10-1

235 Ua F F-S 3.33 × 10-2

236Ua F F-S 2.46 × 10-3

238Ua F F-S 2.20 × 10-1

99Tcg F F-M 3.70 × 10-3 3.70 × 10-3

237Np M M 8.22 × 10-5 8.22 × 10-5

238Pub S M-S 1.03 × 10-5

239Pub S M-S 1.03 × 10-5 3.09 × 10-5

240Pub S M-S 1.03 × 10-5

241Amb S M-S 1.03 × 10-5 1.03 × 10-5

228Thi S F-S 1.51 × 10-5 1.51 × 10-5

230Th S F-S 8.36 × 10-5 8.36 × 10-5

231Th S F-S Trace 232Thi S F-S 1.66 × 10-5 1.66 × 10-5

234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-333 UF6, U3O8, and uranyl 

fluoride (UO2F2) are 

likely (High and 

234Ua F F-S 5.03 × 10-1

235 Ua F F-S 2.83 × 10-2

236Ua F F-S 1.74 × 10-3
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 15 of 27

Facility Compound RN Suggested 

Absorption Typed 

Absorption Type 

Range

Activity 

Fractionf,h

Intake Activity 

Fraction Relative 

to Total Uraniumj 

Strom 1993, p14.7). 238Ua F F-S 4.67 × 10-1

99Tcg F F-M 1.54 × 10-2 1.54 × 10-2

237Np M M 6.81 × 10-5 6.81 × 10-5

238Pub S M-S 4.29 × 10-5

239Pub S M-S 4.29 × 10-5 1.29 × 10-4

240Pub S M-S 4.29 × 10-5

241Amb S M-S 4.29 × 10-5 4.29 × 10-5

228Thi S F-S 4.42 × 10-5 4.42 × 10-5

230Th S F-S 8.23 × 10-5 8.23 × 10-5

231Th S F-S Trace 232Thi S F-S 1.04 × 10-5 1.04 × 10-5

234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-342A 

and X-342

UF6 (High and Strom, 

1993, p14.7).

234Ua F F-S 8.46 × 10-1

235 Ua F F-S 3.48 × 10-2

236Ua F F-S 4.69 × 10-3

238Ua F F-S 1.14 × 10-1

99Tcg F F-M 4.77 × 10-3 4.77 × 10-3

237Np M M-S 3.35 × 10-4 3.35 × 10-4

238Pub S M-S 1.33 × 10-5

239Pub S M-S 1.33 × 10-5 3.99 × 10-5

240Pub S M-S 1.33 × 10-5

241Amb S M-S 1.33 × 10-5 1.33 × 10-5

228Thi S F-S 1.05 × 10-4 1.05 × 10-4

230Th S F-S 1.06 × 10-4 1.06 × 10-4

231Th S F-S Trace 232Thi S F-S Trace 234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

(In equilibrium with 234Th)

X-343 UF6 (Hill and Strom, 

1993, p14.7).

234Ua F F-S 8.06 × 10-1

235 Ua F F-S 3.45 × 10-2

236Ua F F-S 2.31 × 10-3

238Ua F F-S 1.57 × 10-1

99Tcg F F-M 2.25 × 10-2 2.25 × 10-2

237Np M M 8.78 × 10-5 8.78 × 10-5

238Pub S M-S 6.26 × 10-5

239Pub S M-S 6.26 × 10-5 1.88 × 10-4

240Pub S M-S 6.26 × 10-5

241Amb S M-S 6.26 × 10-5 6.26 × 10-5

228Thi S F-S 2.49 × 10-5 2.49 × 10-5

230Th S F-S 1.95 × 10-4 1.95 × 10-4

231Th S F-S Trace 232Thi S F-S Trace 234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-344 UF6 (High and Strom 

1993, p14.7) 1958-

1962 UF6, U3O8, and 

UO2F2 1962-2001.

234Ue S F-S 7.44 × 10-1

235 Ue S F-S 3.20 × 10-2

236Ue S F-S 4.9 × 10-4

238Ue S F-S 2.23 × 10-1

99Tcg F F-M 1.25 × 10-3 1.25 × 10-3

237Np M M 2.37 × 10-5 2.37 × 10-5

238Pub S M-S 3.48 × 10-6

239Pub S M-S 3.48 × 10-6 1.04 × 10-5

240Pub S M-S 3.48 × 10-6

241Amb S M-S 3.48 × 10-6 3.48 × 10-6

228Thi S F-S 9.95 × 10-6 9.95 × 10-6

230Th S F-S 1.42 × 10-4 1.42 × 10-4

231Th S F-S Trace 232Thi S F-S Trace
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 16 of 27

Facility Compound RN Suggested 

Absorption Typed 

Absorption Type 

Range

Activity 

Fractionf,h

Intake Activity 

Fraction Relative 

to Total Uraniumj 

234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-345 UF6, UO3, and UO2F2 

.

234Ua F F-S 9.78 × 10-1

235 Ua F F-S 1.68 × 10-2

236Ua F F-S 1.19 × 10-3

238Ua F F-S 2.16 × 10-3

99Tcg F F-M Trace 237Np M M 2.04 × 10-3 2.04 × 10-3

238Pub S M-S Trace 239Pub S M-S Trace 240Pub S M-S Trace 241Amb S M-S Trace 228Thi S F-S Trace 230Th S F-S 1.2 × 10-1 1.2 × 10-1

231Th S F-S Trace 232Thi S F-S Trace 234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-700 UF6, U3O8, and 

UO2F2 .

234Ue S F-S 5.97 × 10-1

235 Ue S F-S 3.14 × 10-2

236Ue S F-S 2.55 × 10-3

238Ue S F-S 3.64 × 10-1

99Tcg F F-M 1.88 × 10-2 1.88 × 10-2

237Np M M 1.37 × 10-3 1.37 × 10-3

238Pub S M-S 5.24 x 10-5

239Pub S M-S 5.24 x 10-5 1.57 x 10-4

240Pub S M-S 5.24 x 10-5

241Amb S M-S 5.24 x 10-5 5.24 x 10-5

228Thi S F-S 3.74 x 10-4 3.74 x 10-4

230Th S F-S 3.17 × 10-3 3.17 × 10-3

231Th S F-S Trace 232Thi S F-S 3.23 x 10-4 3.23 x 10-4

234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-705 &

X-705E + 

H 

UF6, U3O8, and 

UO2F2 Includes MgF2

traps, TRUs mostly in 

tower ash (705E 

1957-1978).

234Ue S F-S 8.87 × 10-1

235 Ue S F-S 3.43 × 10-2

236Ue S F-S 1.86 × 10-2

238Ue S F-S 5.88 × 10-2

99Tcg F F-M 3.16 × 10-2 3.16 × 10-2

237Np M M 1.60 × 10-4 1.60 × 10-4

238Pub S M-S 8.79 × 10-5

239Pub S M-S 8.79 × 10-5 2.64 × 10-4

240Pub S M-S 8.79 × 10-5

241Amb S M-S 8.79 × 10-5 8.79 × 10-5

228Thi S F-S 1.30 × 10-4 1.30 × 10-4

230Th S F-S 6.66 × 10-4 6.66 × 10-4

231Th S F-S Trace 232Thi S F-S 3.37 × 10-5 3.37 × 10-5

234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-710 UF6, UO3, and UO2F2 

. 

234Ua F F-S 9.18 × 10-1

235 Ua F F-S 2.70 × 10-2

236Ua F F-S 1.75 × 10-2

238Ua F F-S 4.69 × 10-2

99Tcg F F-M 2.60 × 10-1 2.60 × 10-1

237Np M M 1.80 × 10-3 1.80 × 10-3

238Pub S M-S 7.25 × 10-4

239Pub S M-S 7.25 × 10-4 2.18 × 10-3
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 17 of 27

Facility Compound RN Suggested 

Absorption Typed 

Absorption Type 

Range

Activity 

Fractionf,h

Intake Activity 

Fraction Relative 

to Total Uraniumj 

240Pub S M-S 7.25 × 10-4

241Amb S M-S 7.25 × 10-4 7.25 × 10-4

228Thi S F-S 3.49 × 10-4 3.49 × 10-4

230Th S F-S 3.94 × 10-3 3.94 × 10-3

231Th S F-S Trace 232Thi S F-S 4.61 × 10-5 4.61 × 10-5

234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-720 UF6, UO3, and UO2F2. 234Ua F F-S 7.77 × 10-1

235 Ua F F-S 3.45 × 10-2

236Ua F F-S 8.9 × 10-4

238Ua F F-S 1.86 × 10-1

99Tcg F F-M 1.11 × 10 -1 1.11 × 10 -1

237Np M M Trace 238Pub S M-S 3.08 × 10-4

239Pub S M-S 3.08 × 10-4 9.24 × 10-4

240Pub S M-S 3.08 × 10-4

241Amb S M-S 3.08 × 10-4 3.08 × 10-4

228Thi S F-S 5.64 × 10-4 5.64 × 10-4

230Th S F-S 4.68 × 10-4 4.68 × 10-4

231Th S F-S Trace 232Thi S F-S Trace 234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

X-745G UF6, UO3, and UO2F2. 234Ua F F-S 9.44 × 10-1

235 Ua F F-S 3.26 × 10-2

236Ua F F-S 2.57 × 10-3

238Ua F F-S 1.88 × 10-2

99Tcg F F-M 7.93 × 10-2 7.93 × 10-2

237Np M M 4.61 × 10-4 4.61 × 10-4

238Pub S M-S 2.21 × 10-4

239Pub S M-S 2.21 × 10-4 6.63 × 10-4

240Pub S M-S 2.21 × 10-4

241Amb S M-S 2.21 × 10-4 2.21 × 10-4

228Thi S F-S 3.08 × 10-4 3.08 × 10-4

230Th S F-S 1.04 × 10-3 1.04 × 10-3

231Th S F-S Trace 232Thi S F-S 4.92 × 10-5 4.92 × 10-5

234Thc M F-S Trace 234mPac M M-S (In equilibrium with 234Th)

a. UF6 and UO2F2 are considered type F materials (Hill and Strom, 1993, p. 14.7). U3O8 is considered type S (Hill and Strom, 1993, p. 14.7). Most material 

should be UF6 feed. Most bioassay results indicate that uranium acts as a type F material (Hill and Strom, 1993, p. 14.5). 

b. Americium and plutonium not distinguished in analysis; use 239Pu for the intake.

c. 234Th/234mPa are found in equilibrium and are type M for GDPs. 

d. Always use bioassay information to determine absorption class, when available.

e. Absorption type for this facility has indicated type S in many bioassay results. 

f. Based upon air sampling data from 1993-1999 obtained by PORTS HP department

g. Based upon 99Tc to 239Pu ratio from Table 5.1.2.6-3.

h. Note some fractions may add to greater than 1 since the 99Tc values (footnote g) were added to the air sampling data (footnote f).

i. 228Th exceeds the value of 232Th because averages are taken and the positive results for thorium analysis is near the LLD of the radionuclides. 

j. Multiply the total uranium intake by this fraction to obtain the intake activity of the contaminant.

5.1.2.7 In Vitro Bioassay for Fission Products (99Tc)

In vitro assessment of 99Tc began in 1965, although until 1998 the analysis was not specific for 99Tc. 

From at least 1965 to 1985, an addition of a liquid scintillator to the urine sample was used for the 

subsequent determination of total beta and alpha activity. Any urine sample greater than 1000 

dpm/100 ml beta activity was re-analyzed to subtract any thorium/uranium complement. The 

remaining activity was reported as 99Tc. There was at least one occurrence when 99Tc was confused 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 18 of 27

with Thallium-201 (201Tl) (GAT-S-52, 1986, p7. The Dose Reconstructor should use the 99Tc to 

uranium ratios as indicated in Table 5.1.2.6-4 for determining intakes to 99Tc.

Urinalysis for 99Tc is the current bioassay method employed at PORTS. The stated 99Tc bioassay 

MDA is 5,000 dpm L-1 (or about 80 Bq L-1) (Hill and Strom, 1993).

As of 10/01/1998, Bechtel Jacobs (DOE) sent 24-hour samples to an outside lab yielding an MDA of 

33 dpm/sample for Quanterra and 200 dpm/sample for ORNL. 

5.1.2.8 In Vitro Bioassay for 237Np

Although transuranics were suspected as a potential internal dosimetry issue routine monitoring was 

not conducted at PORTS. Alpha spectroscopy was performed on product analysis early in the site’s 

history but the MDA of this system was considered too high for bioassay applications. In 1994, alpha 

spectroscopy was utilized for some suspect urine samples. A 0.5 pCi/liter MDC was reported. ICP/MS 

was available since 1995 but not calibrated for neptunium monitoring. In 1998, Bechtel Jacobs sent 

their samples offsite to ORNL or offsite vendors. ORNL had a 0.02 dpm/sample MDA for 237Np. Np 

was analyzed in vivo at PORTS since about 1965. 

5.1.2.9 In Vitro Bioassay for Trivalent Actinides (Americium)

Routine monitoring for 241Am was not conducted. In 1994 alpha spectroscopy was utilized for some 

suspect urine samples. A 0.5 dpm/liter MDC was reported. In 1998, Bechtel Jacobs sent their

samples offsite to ORNL or offsite vendors. ORNL had a 0.02 dpm/sample MDA for 241Am.

5.1.2.10 In Vitro Bioassay for Plutonium

Routine monitoring for 238,239,240Pu was not conducted. In 1994, alpha spectroscopy was utilized for 

some suspect urine samples. A 0.5 dpm/liter MDC was reported. In 1998, Bechtel Jacobs sent their 

samples offsite to ORNL or offsite vendors. ORNL had a 0.02 dpm/sample MDA for 238,239,240Pu.

According to the Mass Balance report for PORTS (BJC, 2000, p 65), 0.23 grams of plutonium was 

processed from RU during the site’s lifetime. Small amounts of TRU were also processed. Air sample 

analysis in the site indicated presence of TRUs. The site was broken up into two categories:

1. Process buildings with TRUs and thorium at < 0.05% alpha activity.

2. Non-process buildings with TRU =0.5% (default) to 2%. Only two facilities met this criteria – 

705 & 710.

5.1.2.11 In Vitro Bioassay for Thorium

No bioassay for 231Th was conducted at PORTS. The bioassay analysis for 234Th has a MDC of 8000 

dpm/liter. Urinalysis was performed for emergency situations (Hill and Strom 1993, p15.33).

5.1.2.12 In Vitro Bioassay for Protactinium

The bioassay analysis for 234mPa has a MDC of 8000 dpm/liter. Urinalysis was performed on 

emergency situations (Hill and Strom, 1993, p 15.37). Table 5.1.2.12 summarizes the MDC and urine 

analysis methods used for technetium, uranium daughters and the transuranics.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 19 of 27

Table 5.1.2.12-1. Summary of 99Tc, 237Np, 241Am, 239Pu, 231Th and 234mPa Urinalysis

Period Personnel 

Organization Radionuclide MDC Urinalysis Urinalysis Method

1/1/1965-Present PORTS USEC 99Tc 5000 dpm/liter Gas Flow Proportional & 

LSC

10/1/1998-Present PORTS BJ 99Tc 200 dpm/liter ORNL

33 dpm/liter Quanterra

Chemical Separation and 

LSC 

1/1/1998-Present BJ & 237Np 0.02 dpm/1 liter ORNL Alpha Sspectroscopy

1/1/1994-12/31/97 BJ + USEC 237Np 0.5 pCi/1 liter ORNL Alpha Spectroscopy

1/1/1998-Present BJ & 241Am 0.02 dpm/1 liter ORNL Alpha Spectroscopy

1/1/1994-12/31/97 BJ + USEC 241Am 0.5 pCi/1 liter ORNL Alpha Spectroscopy

1/1/1998-Present BJ & 239Pu 0.02 dpm/1 liter ORNL Alpha Spectroscopy

1/1/1994-12/31/97 BJ + USEC 239Pu 0.5 pCi /1 liter ORNL Alpha Spectroscopy

1/1/1994-Present BJ & USEC 234Th 8000 dpm/1 liter Gas flow Proportional & LSC 

1/1/1994-Present BJ & USEC 234mPa 8000 dpm/1 liter Gas flow Proportional & LSC 

5.2 CORRECTING FOR URINALYSIS VOLUME

At PORTS all on-site urine sample analysis are from spot samples collected at the end of shift on 

Friday (Hill and Strom, 1993, p 15.4). The samples, which have been sent off site since 1998, are 24-

hour samples. All samples can be normalized to the reference volume of 1.4 liters/day by knowing 

either the period or volume collected. Normalization by volume would be appropriate for the spot 

samples and might be appropriate for the 24-hour samples if the volume is not within the range of 

normal variability for 24-hour samples.

Note that some of the urine samples are utilized for chemical monitoring: 50 ml for fluorides, 50 ml for 

mercury and 50 ml for other chemicals.

5.3 IN VIVO MDAS, COUNTING METHODS AND REPORTING PRACTICES

The whole body counting system was routinely started in 1965. The Y-12 mobile in vivo laboratory 

was operated by Martin Marietta Energy Systems (MMES) and utilized under contract with the 

Goodyear Atomic Corporation Technical Document. The mobile unit operated on the GAT site for 

between 12 to 16 weeks per year and was operated by Y-12 employees (GAT, 1985b, p1).

The in vivo unit employed two large 9” NaI (Tl) detectors for the detection of gamma and 

bremsstrahlung radiation in the chest cavity. Non-transportable or insoluble compounds such as U3O8

are retained in the lungs for extended periods of time. Therefore, this in-vivo monitoring technique is 

used for detecting these types of compounds. Monitoring for type M 99Tc activity in the lung using 

bremsstrahlung radiation from the 85 keV (ave.) beta was also attempted. Soluble or transportable 

compounds such as UO2F2 and 99Tc (Type F) are monitored via urinalysis (GAT, 1985b). The MDAs, 

recount values and restriction levels are summarized in Table 5.3-1. The results are given in mass or 

activity units.
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 20 of 27

Table 5.3-1. MMES Chest Counter /1965 –1990 MDAs and Restriction Levels

Measurement Category Enriched Uranium 

(

235U)

Depleted 

Uranium (238U)

Total Uranium

(

238U + 235U) Tc-99 Np-237

Quantity Below Which 

Detection is Impossible

100 μg 4 mg

1.46 nCi

4 mg 1 μCi 0.2 nCi

Quantity at Which Recount 

is Required to Confirm Data

100 μg 4 mg 4 mg 1 μCi 1.7 nCi

Restrictive Value 240 μg 37 mg 27 mg 9 μCi 17 nCi

Source: GAT, 1985b, p 27.

A summary of in vivo monitoring at PORTS from 1965–1985 is presented in Table 5.3-2. The total 

number of counts above the MDA is given for each year. As can be seen, PORTS had an active in 

vivo monitoring program.

Table 5.3-2. In Vivo Summary (1965–1985)

Year Total Counts 

Taken 

Number of 

Counts >100 

μg 235U

Year Total Counts 

Taken

Number of 

Counts >100 

μg 235U

1965 27 18 1976 411 58

1966 30 14 1977 971 96

1967 236 28 1978 542 29

1968 364 39 1979 497 15

1969 393 73 1980 924 4

1970 147 32 1981 868 2

1971 179 56 1982 910 1

1972 157 36 1983 632 3

1973 392 26 1984 613 0

1974 521 65 1985 798 4

1975 684 92

Source: BJC, 2000, p 70.

A second chest counting unit, the Helgesson Phoswich detector system that was comprised of a 5” 

diameter, 1⁄2” NaI detector and a 1 1⁄2” CsI thick detector, was used from 1991 through 1995. The chest 

counter had four phoswich detectors, two located over the whole body, two located over the thighs 

(background subtraction). Its detection capabilities are listed in Table 5.3-3.

The Helgesson system was used as a Quality Assurance check on the urine analysis and air 

sampling programs at PORTS and not as a routine bioassay system. The MDAs and restriction levels 

are summarized in Table 5.3-3.

Table 5.3-3. Helgesson Chest Counter /1991-1995

Compound Enriched Uranium Total Uranium

Quantity Below Which Detection is Impossible (235U) 40 - 70 μg 2 – 4 mg

Quantity at Which Recount is Required to Confirm Data 40 - 70 μg 2 – 4 mg

Restrictive Value 240 μg 27 mg

137Cs is reported in the chest counting database. The National Council on Radiation Protection and 

Measurements Report 94 (NCRP 94, p 160) reports body burdens in the United States from fallout. In 

the 1960s and 1970s, fallout contributed to the body burdens with the most peaking at 19 nCi in 1964. 

137Cs burdens would likely not have been due to occupational exposures at PORTS and can be 

ignored.
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The uranium isotope 235U decays with a physical half-life of 7.0 x 108 years to 231Th by alpha emission. 

231Thorium, with a half-life of 25.5 hours, decays by beta emission and grows rapidly into equilibrium 

with 235U. There is no significant in-growth of other daughters. The gamma emissions from 235U that 

can be used for in vivo counting are 144 keV (10.5%), 186 keV (54.0%), and 200 keV (5.0%) (Hill and 

Strom, 1993, p 15.17).

The uranium isotope 238U decays with a physical half-life of 4.5 x 109

 year to 234Th by alpha emission. 

Thorium 234, with a half-life of 24 days, decays by beta emission and grows into equilibrium with 238U. 

The two low energy photons emitted from 238U, namely 13.0 keV (2.96%), and 15.1 keV (4.47%), 

cannot be used for in vivo counting. In actuality the 234Th photons are used to indirectly estimate the 

quantification of 238U. The photons used are 63.3 keV (3.81%), 92.3 keV (2.73%) and 93.8 keV 

(2.69%) (Hill and Strom, 1993, p 15.29).

The uranium isotope 234U decays with a physical half-life of 2.4 x 105 years to 230Th by alpha emission. 

The long half-life of 230Th prevents significant in growth of radioactive daughters. The gamma 

emissions cannot be utilized for in vivo counting. The isotope is quantified indirectly based on the 

amount of 235U or 238U present (Hill and Strom, 1993, p 15.17).

The uranium isotope 236U decays with a physical half-life of 2.4 x 105 years to 232Th by alpha emission. 

The long half-life of 232Th precludes the in growth of radioactive daughters. The 13.0 keV (3.36%) and 

16.1 keV (5.08%) gamma emissions are likely too low to be utilized for in vivo counting (Hill and 

Strom, 1993, p15.29).

In summary, 235U was probably counted directly using the 186 kev photon; 238U was probably counted 

via the 63.3 keV photon from 234Th assuming equilibrium with 238U; and 234U and 236U were not 

counted in vivo. 

234mPa could be monitored through the gamma emissions of 234mPa for in vivo counting. This is not 

currently done at PORTS. 

99Tc is detected by Bremsstrahlung production by in-vivo counting for insoluble forms. (ORAUTTKBS-0014-5, 5.3.2.3) However, urine analysis is the routine monitoring bioassay system at PORTS. 

237Np decays with a physical half-life of 2.14 x 106 years to 233Pa by alpha emission. 233mPa, with a 

half-life of 27 days, decays by beta emission and grows into equilibrium with 237Np. The low energy 

photons from 237Np that can be used for in vivo counting are primarily the 86.5 keV (12.6%) and from 

233Pa, the 94.7 keV (10.1%), 98.4 keV (15.3%) and the 312 keV (36.0%) photons (Hill and Strom, 

1993, p 15.43). Naturally-occurring radon daughters provide for a false positive for 237Np.

5.4 ABSORPTION TYPES

Radioactive materials classified as F, M, or S (fast, medium or slow) depends on their retention time 

in the pulmonary region. These designations are similar to the classes D, W and Y, but refer strictly to 

the rate of absorption from the respiratory tract to the blood.

International Commission on Radiological Protection (ICRP) Publication 68 lists UF6, UO2F2 and UO2 

(NO3) (uranyl nitrate) as inhalation type F, UF4 and UO3 as type M and U3O8 and UO2 as type S. UO2 

could be considered type S (ICRP 68, 1994, p 83).

The chemical form and the enrichment have varied over time at PORTS. From about 1975–1978 most 

of the HEU up to about 93% enrichment was produced at PORTS. Oxide conversion produced the 
Effective Date: 11/24/2004 Revision No. 00 Document No. ORAUT-TKBS-0015-5 Page 22 of 27

potential for insoluble uranium compound exposures in building X-705E from 1957-1978. Recovery 

and decontamination processes still evoke the potential of insoluble uranium exposures. In building X344 a UF4 to UF6 conversion facility operated from 1958–1962. This also produced insoluble forms of 

uranium. Current air sampling analysis (1992-2003) and bioassays has indicated type S material in 

these two locations at PORTS. All other areas indicate a Type F form of uranium. Table 5.1.2.6-4 lists 

solubility types for the separate facilities and indicates a range of type F-S for all buildings with the 

exception of X-344 and X705 (E). Most urine analysis cases at PORTS indicated soluble uranium with 

a demonstrated 1.25-day effective half-life. Insoluble uranium with a 120-day effective half-life was 

used for dose calculations as indicated in the site literature (GAT, 1966a, p 1 supplement).

99Tc is an inhalation hazard. The tendency of the 99Tc is to accumulate at the higher stage levels at 

purge and vent areas. Detection levels at PORTS have been limited to about 5000 dpm/liter via urine 

analysis. 99Tc has appeared to act like a type F material. Oxides, hydroxides, halides, and nitrates of 

technetium are assigned to inhalation type M; all others are assigned to type F (ICRP 68, 1994, p 82). 

According to Hill and Strom (1993), Tc at PORTS exists as Tc pertechnetate so type F is suggested. 

Classifying to type M should be done if the organs of interest are the ET airways, LLI wall, Colon or 

lungs, and classifying to type F should be done if the organs of interest are ST wall (ICRP 68, 1994). 

237Np is type M for all compounds (ICRP 68, 1994, p 83). 

Plutonium oxides, carbides, and hydroxides are type S, nitrates and other compounds are type M. 

Where plutonium is a small contaminant in a uranium matrix, the plutonium might behave the same as 

the host uranium, so the exact absorption type for Pu at PORTS is not well known. The DR can 

assume either type M or S to maximize the dose to the organ of concern. 

Americium is type M for all compounds; however, as for plutonium, trace amounts in a host matrix 

may behave the same as the host compound. Therefore, the DR can assume either type M or S. 

Thorium oxides, carbides, and hydroxides are type S, nitrates and other compounds are assigned to 

type M. Because of the host matrix effect mentioned above, the DR can assume either M or S. 

5.5 WORKPLACE AIR SAMPLING DATA AS APPLICABLE 

PORTS has an extensive air sampling program. All process areas and most other work areas are 

sampled by low to high volume air samplers. Grab samples have also been taken from the beginning 

of operations. 50 μg/m3 for uranium was the limit utilized from 1955 and the air concentration limits 

have gone through transitions. Despite the changes from AEC to the Energy Research and 

Development Administration (ERDA) to the U.S. Department of Energy (DOE) regulations the overall 

limit to uranium for both soluble and insoluble forms has remained about the same. For example, the 

calculated Derived Air Concentration based upon on-site air sampling and DOE 10 CFR 835 for Class 

Y uranium was equal to 1.9 x 10-11

μCi/ml (BJC, 2000, p 101) and on the older ERDA regulations 6 x 

10-11

μCi/cc. 

Portable air samples (PAS) have also been taken, but primarily limited to larger maintenance work. 

Area and PAS data have been used to establish airborne areas as well as characterizing the 

percentage of TRU in the air. If a positive bioassay occurs and work in a specified area can be 

established, more specific detail of the radionuclide make-up and chemical form can be determined by 

utilizing Table 5.1.2.6-4. In addition, removable contamination surveys have been useful in 

characterizing uptakes in the past at PORTS.
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However, the air sampling data is very extensive and prone to larger bias and systemic errors. The air 

sample data from 1993 through 1999 was utilized to create the activity fraction column for the tables 

for each building in Table 5.1.2.6-4. In addition, site-specific bioassay and contamination monitoring 

aided in characterizing these areas. No formal attempt i.e. in the literature has been developed for the 

PORTS facility unlike other DOE sites to document workplace air monitoring results or practices. Air 

sample records can probably be located but are not currently stored in a centralized location. 

5.6 INTERFERENCE AND UNCERTAINTY

5.6.1

Excreta samples were collected off-site. There still may have been a possibility for crosscontamination from workers’ hands to the samples. Laboratory contamination issues, sample count 

failures or mix-ups were possible, but not recurrent. Quality controls had been placed upon sample 

collection, data transferal and equipment operation since the start of operations at PORTS.

Contamination of Samples

If there was a high sample count, a recall sample was requested and counted. Other samples may 

have been re-counted to ensure that possible laboratory error could be eliminated. If a sample 

seemed to be a true high count further urine samples were requested and an in vivo count requested 

and obtained. The solubility type of the radionuclide could then most likely be determined.

Contamination to the whole body could interfere with the lung counting. Workers were asked to 

shower and change prior to the chest count. If contamination was suspected, the worker would repeat 

the shower. A section in the database and on the data form indicates a “FB Ratio”. This ratio was 

used to determine the possibility of external contamination as opposed to internal deposition. If the 

ratio was greater than one, the measured activity was closer to the surface and probably due to 

external contamination. A recount would be made after another shower. Subsequent re-counts and 

urine bioassays would be given to validate the positive chest count or urine count, or both.

5.6.2

Uncertainties for bioassay measurements were not stated in the records or database. 

Uncertainty

PORTS had a practice of submitting routine urine samples at most within two days after the last work 

day. The sample was to be taken at the medical center within two hours of clocking in to minimize the 

possibility of inadvertent contamination from loose material on a person’s clothing. If the individual 

comes to the hospital after the two hour period, the receptionist would reschedule the specimen for 

the next Monday. Non-routine samples had a greater possibility of cross contamination because the 

specimens were given either immediately or within anytime during the work day. (GAT, 1966a) 

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

workers. Because studies of the average daily uranium excretion on PORTS 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-hour) urinary excretion rate for uranium of 0.43 μg (environmental decision level at 

95% confidence) can be used (BJC, 1999).
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5.7 ASSESSMENT OF INTAKE FOR MONITORED AND UNMONITORED WORKERS

5.7.1

PORTS had an extensive bioassay program from the beginning of its operations. Urine analysis 

started out in 1954 with the addition of chest counting in 1965. Although the earlier techniques had 

their sensitivity limitations, the detection sensitivity seemed to keep pace with the fast paced 

regulatory and safety changes. Seldom did workers achieve or surpass the body burden of the 

radionuclides of concern, as a whole. Some assumptions that a Dose Reconstructor can make:

Monitored Workers with Measurable Intakes

• Time of intake – If the date or time of intake to a particular bioassay result is unknown, the 

intake occurred mid-way between prior and the current bioassay result. (ICRP 78, 1997).

• Radionuclide, particle size and solubility type- Tables 5.1.1-2, 5.1.2.6-3, and 5.1.2.6-3 list the 

values. If the work location of the employee is not known, default assumptions for radionuclide, 

particle size and solubility type are equal to those in the most closely related facility. 

• When an intake of uranium has been determined, add intakes of the other radionuclide 

contaminants in proportions established in Table 5.1.2.6-4. If there are specific bioassay 

results for any of the contaminants that indicate a higher intake than determined using the ratio 

to uranium, use the specific bioassay results to determine the intake for that radionuclide.

• Uranium enrichment – 3.5% unless HEU of 93% is suspected.

• Uranium intake activity can be assumed to be 234U.

• Route of intake – Intake is by inhalation. Note ingestion is possible.

5.7.2

For monitored workers with no positive results, a triangular distribution is used, with the mode 

determined using half of the MDC value and the maximum using the MDC as input into the dosimetry 

codes.

Monitored Workers with Nothing Detected in Bioassay

5.8 IN VITRO URINALYSIS AND IN VIVO LUNG COUNTING DATA CODE INFORMATION

The codes utilized for PORTS records and databases have been consistent over most of the history 

of the site. Since the 1950’s both the industrial hygiene bioassay and health physics bioassay 

information was entered on the same form. The industrial hygiene information is not needed for 

radiation dose reconstruction. Table 5.8-1 lists a summary of known codes along with their 

interpretations. 
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Table 5.8-1. Internal Dosimetry Record Codes

Measurement 

Type Column Identifier Code Interpretation

Urine Bioassay Alpha (dpm/100 ml) Total alpha counts per 100 ml of urine.

Urine Bioassay Beta(Tc) (dpm/100ml) Total beta counts per 100 ml of urine (all counts assumed 

beta).

Urine Bioassay Column listed “A” 0 Not tested.

1 Tested.

2 Recall or above a flag.

4 Restriction.

Blank < MDA.

Urine Bioassay Alpha (dpm/dL) Total alpha counts per 100 ml of urine (dL = deci-liter).

Urine Bioassay URAN Uranium in mg/liter by fluorometry.

Urine Bioassay ALPHA Total alpha counts per 100 ml of urine (dL = deci-liter) by Gas 

flow proportional. 

Urine Bioassay TECH Total Beta counts per 100 ml of urine (dL = deci-liter) by Gas 

flow proportional.

Urine Bioassay FLUOR, MERC, 

LEAD, ZINC & CAD

Fluorine, Mercury, lead, zinc and cadmium Chemical analysis 

not to be included in dose reconstruction.

Urine Bioassay IV IV Insufficient urine volume.

Urine Bioassay OE OE Operator error.

Lung Bioassay FB Ratio FB This is a measure of how close to the front or back the internal 

contamination is. A ratio of greater than one may indicate 

external contamination.

Lung Bioassay Negative value in TotU columnA negative value indicates less than the 4 mg MDA for total 

Uranium.
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REFERENCES

ANSI HPS N13.22-1995, Bioassay Programs for Uranium, (An American National Standard), October, 

1995.

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

Jacobs Company LLC, BJC/OR-258.

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

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

Management, Washington, DC.

DOE (U. S. Department of Energy) 2000b, Independent Investigation of the Portsmouth Gaseous 

Diffusion Plant, Volume 2, “Current Environment, Safety, and Health Practices”, Office of 

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

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

Protection in Uranium Facilities, DOE Standard DOE-STD-1136-2000, DOE ES&H Technical 

Information Services, August 2000.

GAT-2-D-10-0, Voss, Frank S., “Urine Analysis for Uranium”, August 10, 1956.

GAT-2-D-11-0, Voss, Frank S., “Urine Analysis for Alpha Activity,” September 14, 1955.

GAT-2-J-10-0, Trivissous, “Alpha Activity of Uranium in Urine,” March 19, 1979.

GAT-S-45, Ruggles, D. J., “Urinalysis Parameters,” July 16, 1985a.

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

17, 1985b.

GAT-S-52, Bassett, A. C., “In Vivo Monitoring Summary -Industrial Hygiene and Health Physics”, 

January 16, 1986. 

GAT Urinary Bioassay Program, D13-FR55 067 265, circa 1966a.

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

and Paducah Gaseous Diffusion Plants, PNL-8723 (UC-607), Pacific Northwest Laboratories, 

Battelle Memorial Institute. 

ICRP 68, “Dose coefficients for intakes of radionuclides to Workers,” International Commission on 

Radiological Protection, Pergamon Press, 1994.

ICRP 78, “Individual Monitoring for Internal Exposure for Workers,” International Commission on 

Radiological Protection, Pergamon Press, 1997. 

Mayfield, Charles K., “Analysis of Urine for 235U and 238U by Inductively Coupled Plasma – M ass 

Spectrometry,” XP4-TS-RL7800, rev. 0, March 31, 1995. 

NCRP 94, “Exposure of the Population in the United States and Canada from Natural Background 

Radiation,” Bethesda, MD, 12/30/1987.
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ORAUT-TKBS-0003-05, rev 01, (Oak Ridge Associated Universities), 2003a, Technical Basis 

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ORAUT-TKBS-0006-05, rev 01, (Oak Ridge Associated Universities), 2003b, Technical Basis 

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ORAUT-TKBS-0014-5, rev.1, “Technical Basis Document for the Y-12 Site to be used for EEOICPA 

Dose Reconstructions,” August 21, 2003.

Sula, M. J., Carbaugh, E. H., and Bihl, D. E., Rev. 1, “Technical Basis for Internal Dosimetry at 

Hanford,” PNL-6866, US DOE, July 1991.