ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 60

DOE Review Release 05/01/2013

Document Title:

Mound Site – Occupational Internal 

Dosimetry

Document Number: ORAUT-TKBS-0016-5

Revision: 02

Effective Date: 04/18/2013

Type of Document: TBD

Supersedes: Revision 01

Subject Expert(s): Donald N Stewart

Approval: Signature on File Approval Date: 04/16/2013

Donald N. Stewart, Document Owner

Concurrence: Signature on File Concurrence Date: 04/15/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 04/15/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 04/15/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/18/2013

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-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 2 of 60

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

09/09/2004 00 First approved issue. New technical basis document for the Mound 

Site – Occupational Internal Dosimetry. Incorporates internal and 

NIOSH review comments. Incorporates additional internal review 

comments. Initiated by Jeffrey S. Vollmer.

03/13/2009 00 PC-1 Approved page change initiated to remove dose reconstruction 

provisions applicable only to Monsanto Chemical Company. Some 

references to Monsanto Chemical Company necessarily remain, but 

dose reconstruction provisions retained are applicable only to the 

Mound Laboratory. These changes occurred on pages 9, 10, 11, 13, 

14, 16 -19, 63 in Sections 5.1, 5.2, 5.3, and Attachment D, 

respectively. The year for the beginning time period changed from 

1947 to 1948 in Table 5-4 on page 15 and in Table 5-20 on page 39. 

NIOSH required language was revised on page 7 in Section 5.1. 

References were updated on pages 42 and 45. Incorporates formal 

internal and NIOSH review comments. No sections were deleted. 

Training required: As determined by Task Manager. Initiated by 

Donald N. Stewart. Approval:

Signature on File 03/04/2009

Donald N. Stewart, Document Owner

Signature on File 03/04/2009

John M. Byrne, Task 3 Manager

Signature on File 03/04/2009

Edward F. Maher, Task 5 Manager

Signature on File 03/11/2009

Kate Kimpan, Project Director

Signature on File 03/13/2009

James W. Neton, Associate Director for Science

12/13/2010 01 Revised to incorporate SEC-00090 and SEC-00171. Revised dose 

reconstruction guidance for Ra-Ac-Th in Section 5.3.2.4, radon dose 

reconstruction guidance for Building SW-19 in Section 5.3.2.1, and 

updated Table 5-19. Incorporates formal internal and NIOSH review 

comments. Training required: As determined by the Objective 

Manager. Initiated by Donald N. Stewart.

04/18/2013 02 Revised to incorporate SEC-00207, as well as comments made on 

the original version by Sanford Cohen & Associates in 2006. 

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: as 

determined by the Objective Manager. Initiated by Donald N. 

Stewart.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 3 of 60

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

5.1 Introduction .............................................................................................................................. 8

5.1.1 Purpose........................................................................................................................ 9

5.1.2 Scope........................................................................................................................... 9

5.1.3 Current Status of Special Exposure Cohort Petitions for Mound................................... 9

5.1.4 Radionuclides of Concern........................................................................................... 10

5.2 Process Operations Involving Primary Radionuclides ............................................................ 11

5.2.1 Polonium-210 ............................................................................................................. 11

5.2.2 Plutonium ................................................................................................................... 11

5.2.2.1 Plutonium-238.............................................................................................. 12

5.2.2.2 Plutonium-239.............................................................................................. 12

5.2.3 Tritium ........................................................................................................................ 13

5.3 Process Operations Involving Secondary Radionuclides........................................................ 13

5.3.1 Radium, Thorium, and Actinium.................................................................................. 13

5.3.2 Thorium ...................................................................................................................... 14

5.3.2.1 Th-232 Source Term.................................................................................... 14

5.3.2.2 Ionium.......................................................................................................... 15

5.3.3 Protactinium-231 ........................................................................................................ 15

5.3.4 Uranium-233............................................................................................................... 15

5.3.5 Reactor Fuels Projects ............................................................................................... 15

5.3.6 Reactor Waste Project................................................................................................ 16

5.3.7 Americium .................................................................................................................. 16

5.3.8 Rare Radioactive Isotope Program............................................................................. 16

5.3.9 Other Radionuclides ................................................................................................... 17

5.3.10 Radon......................................................................................................................... 17

5.4 Internal Dose Monitoring ........................................................................................................ 19

5.4.1 General Description.................................................................................................... 19

5.4.2 In Vitro Monitoring ...................................................................................................... 20

5.4.3 Gross Alpha and Sequential Separation ..................................................................... 21

5.4.4 In Vivo Dose Monitoring.............................................................................................. 22

5.4.5 Personnel Air Sampling Data...................................................................................... 22

5.5 Bioassay Methodologies for Primary Radionuclides ............................................................... 23

5.5.1 Polonium .................................................................................................................... 23

5.5.1.1 Bioassay Methods........................................................................................ 23

5.5.1.2 Polonium Internal Dose Monitoring Records ................................................ 24

5.5.2 Plutonium ................................................................................................................... 25

5.5.3 Tritium ........................................................................................................................ 26

5.6 Secondary Radionuclides Internal Dose Monitoring Programs............................................... 28

5.6.1 Radium-Actinium ........................................................................................................ 28

5.6.2 Thorium ...................................................................................................................... 29

5.6.2.1 Thorium-230................................................................................................. 29

5.6.3 Protactinium-231 ........................................................................................................ 30

5.6.4 Uranium...................................................................................................................... 30
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5.6.5 Americium-241 and Curium-244 Records................................................................... 31

5.6.6 Rare Isotopes ............................................................................................................. 31

5.7 Internal Dose Reconstruction Guidance................................................................................. 31

5.7.1 General Considerations.............................................................................................. 31

5.7.1.1 Use of the Gross Alpha Method ................................................................... 31

5.7.1.2 Internal Dose Estimation Parameters........................................................... 31

5.7.2 Equilibrium Assumptions ............................................................................................ 32

5.8 Reconstructing Dose from Primary Radionuclides Based on Urinalysis Data......................... 34

5.8.1 Polonium-210 ............................................................................................................. 34

5.8.2 Plutonium ................................................................................................................... 35

5.8.2.1 Special Considerations for Plutonium-238.................................................... 36

5.8.2.2 Special Considerations for Plutonium-239.................................................... 39

5.8.3 Tritium ........................................................................................................................ 39

5.9 Reconstructing Dose from Secondary Radionuclides Based on Urinalysis Data .................... 40

5.9.1 Radium-Thorium-Actinium .......................................................................................... 40

5.9.2 Thorium ...................................................................................................................... 40

5.9.2.1 Thorium-232................................................................................................. 40

5.9.2.2 Thorium-230................................................................................................. 41

5.9.3 Protactinium-231 ........................................................................................................ 42

5.9.4 Uranium...................................................................................................................... 43

5.9.5 Reactor Fuels Projects ............................................................................................... 43

5.9.6 Reactor Waste Project................................................................................................ 43

5.9.7 Americium and Curium ............................................................................................... 44

5.9.8 Rare Isotopes ............................................................................................................. 45

5.9.9 Radon......................................................................................................................... 45

5.9.10 Summary of Correction Factors and Excretion Equations........................................... 45

5.9.11 Summary of Analytical Methods and Detection Limits ................................................ 46

5.9.12 Use of Co-Worker Data for Internal Dose Reconstruction........................................... 47

5.10 Attributions and Annotations .................................................................................................. 47

References ........................................................................................................................................ 48

Glossary ............................................................................................................................................ 54

ATTACHMENT A, CODING INSTRUCTIONS FOR URINALYSIS RECORDS ................................. 56

ATTACHMENT B, UNCERTAINTY ASSESSMENT.......................................................................... 57
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LIST OF TABLES

TABLE TITLE PAGE

5-1 Radon isotope survey data by building and period ................................................................. 18

5-2 Bioassay programs and action levels ..................................................................................... 19

5-3 Summary of bioassay methodologies..................................................................................... 20

5-4 Summary of historical plutonium analytical procedures .......................................................... 26

5-5 Reported analytical procedures and MDAs for tritium............................................................. 27

5-6 Recommended equilibrium assumptions for secondary radionuclides.................................... 32

5-7 Radionuclide chemical recoveries for collected fecal samples ............................................... 33

5-8 Retroactive MDAs for fecal samples ...................................................................................... 33

5-9 Reported background count rates and MDAs for 210Po........................................................... 35

5-10 Heat source plutonium feed material characteristics .............................................................. 35

5-11 Weapons-grade plutonium feed material characteristics ........................................................ 35

5-12 Plutonium MDAs ....................................................................................................................36

5-13 Summary of limiting dissolution type versus the duration of a chronic inhalation 

intake of 238Pu ........................................................................................................................37

5-14 Summary of limiting dissolution type versus the time after an acute intake a 

single urine sample is collected.............................................................................................. 38

5-15 IMBA input parameters for types M, S, and the nonmonotonic model L ................................. 39

5-16 Reported analytical procedures and detection limits for tritium............................................... 39

5-17 Natural thorium DLs ...............................................................................................................41

5-18 230Th DLs ...............................................................................................................................42

5-19 Protactinium DLs....................................................................................................................42

5-20 Uranium DLs..........................................................................................................................43

5-21 241Am and 244Cm DLs .............................................................................................................44

5-22 241Am and 244Cm detection limits ............................................................................................ 44

5-23 222Rn WLM assumptions for dose reconstruction ................................................................... 45

5-24 Summary of correction factors and excretion equations ......................................................... 46

5-25 Analytical methods and MDAs for primary radionuclides by period ........................................ 47

A-1 Coding instructions for urinalysis records ............................................................................... 56
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

ANPP Aircraft Nuclear Propulsion Project

Bq becquerel

BZ breathing zone

CFR Code of Federal Regulations

Ci curie

cm centimeter

cph counts per hour

cpm counts per minute

d day

D&D decontamination and decommissioning

DCF dose conversion factor

DE dose equivalent

DL detection limit

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

F fast (absorption type)

GSD geometric standard deviation

hr hour

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

IRF intake retention fractions

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

L liter

LANL Los Alamos National Laboratory

LS liquid scintillation

LSC liquid scintillation counter

M moderate (absorption type)

mCi millicurie

MDA minimum detectable activity

min minute

mL milliliter

mo month
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MPC maximum permissible concentration

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NU natural uranium

ORAU Oak Ridge Associated Universities

pCi picocurie

PERM passive environmental radon monitor

POC probability of causation

PORECON Polonium urinalysis bioassay results database

PURECON Plutonium urinalysis bioassay results database

QF quality factor

RTG radioisotope thermoelectric generator

S slow (absorption type)

SMT stable metal tritide

SNAP Systems for Nuclear Auxiliary Power

SRDB Ref ID Site Research Database Reference Identification (number)

TTA thenoyl trifluoroacetone

U.S.C. United States Code

wk week

WL working level

WLM working level month

yr year

μCi microcurie

μg microgram

μm micrometer

§ section or sections
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5.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

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

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

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

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

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

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

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

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

Occupational Illness Compensation Program Act [EEOICPA; 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 employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) restrict the “performance of duty” 

referred to in 42 U.S.C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

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

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

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

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

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2010):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
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5.1.1 Purpose

This technical basis document (TBD) contains an analysis of the internal doses received by Mound 

Laboratory workers from the period of the initial external dose monitoring on February 21, 1949, 

through the end of operations on the Mound site. The nature of the doses received by Mound 

workers is analyzed, and methods for reconstructing doses are given.

5.1.2 Scope

The information in this TBD facilitates the reconstruction of occupational internal dose to employees of 

the Mound facility. This TBD does not contain guidance to reconstruct doses at Dayton Laboratory 

facilities (other than the Mound Laboratory), or for dates earlier than February 1949.

Section 5.1.3 discusses classes of Mound employees added to the Special Exposure Cohort (SEC). 

Section 5.1.4 identifies the radionuclides of concern to internal dose. Section 5.2 describes process 

operations that involved the primary radionuclides, and Section 5.3 describes operations with 

secondary radionuclides. Section 5.4 discusses internal dose monitoring, Section 5.5 describes 

bioassay for the primary nuclides, and Section 5.6 discusses internal dose monitoring for the 

secondary nuclides. Sections 5.7 to 5.9 provide guidance for reconstruction of internal dose. 

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 5.10.

5.1.3 Current Status of Special Exposure Cohort Petitions for Mound

Classes Added to the SEC

Employees of the Department of Energy (DOE), its predecessor agencies, and DOE 

contractors or subcontractors who worked in any areas at the Mound Plant site from 

October 1, 1949, through February 28, 1959, for a number of work days aggregating at 

least 250 work days or in combination with work days within the parameters 

established for one or more other classes of employees in the SEC (SEC-0090; Leavitt 

2008).

All employees of the Department of Energy (DOE), its predecessor agencies, and its 

contractors and subcontractors who had at least one tritium bioassay sample and 

worked at the Mound Plant in Miamisburg, Ohio from March 1, 1959 through March 5, 

1980, for a number of work days aggregating at least 250 work days, occurring either 

solely under this employment, or in combination with work days within the parameters 

established for one or more other classes of employees in the Special Exposure 

Cohort (SEC-00171; Sebelius 2010).

All employees of the Department of Energy (DOE), its predecessor agencies, and their 

contractors and subcontractors who worked at the Mound Plant in Miamisburg, Ohio, 

from September 1, 1972 through December 31, 1972, or from January 1, 1975 through 

December 31, 1976, for a number of work days aggregating at least 250 work days, 

occurring either solely under this employment or in combination with work days within 

the parameters established for one or more other classes of employees included in the 

Special Exposure Cohort (SEC-00207; NIOSH 2012).

Dose reconstruction guidance in this document for periods: before March 1, 1959; September 1, 1972 

through December 31, 1972; January 1, 1975 through December 31, 1976; or March 1, 1959 through 

March 5, 1980 for the SW (Semi-Works) Building, is presented to provide a technical basis for partial 

dose reconstructions for claims not covered within the SEC classes above. 
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Although NIOSH has found that it is not possible to completely reconstruct radiation doses for the 

SEC classes described in this section, NIOSH intends to use any internal and external monitoring 

data that might become available for an individual claim, and that can be interpreted using existing 

NIOSH dose reconstruction processes or procedures. Therefore, dose reconstructions for employees 

who worked in any areas of the Mound Plant during the periods from October 1, 1949 through 

February 28, 1959, or September 1, 1972 through December 31, 1972, or January 1, 1975, through 

December 31, 1976, or who worked in R or SW Buildings (as demonstrated by records of tritium 

bioassay sampling) during the period from March 1, 1959, through March 5, 1980, but who do not 

qualify for inclusion in the SEC, may be performed using these data as appropriate.

To summarize the classes added to the SEC: 

• The first class of employees was added because the potential for unmonitored intakes of 

radium, actinium, and thorium was assumed to be present for all Mound employees; therefore, 

it was not possible to reconstruct doses or assign maximum doses to all Mound employees 

prior to 1959;

• Because it was not possible to bound radon doses to workers in the SW building, Mound 

workers who worked in R and/or SW buildings were added to the SEC. These workers are 

identified by at least one bioassay sample for tritium in their internal dose records; and

• Since it was found that during some periods that tritium records could not be found, all 

employees were included in the SEC during the periods September-December 1972, and the 

years 1975 and 1976. 

For all of these classes of workers, internal doses are to be reconstructed using case-specific data, if 

present, when they may be estimated using current dose estimation processes and procedures. 

5.1.4 Radionuclides of Concern

Radionuclides are categorized as either primary or secondary based on the potential extent of 

presumptive exposures to Mound employees. Primary radionuclides of concern are discussed in 

Mound internal dosimetry documentation including History of Mound Bioassay Programs (Meyer 

1992) and Mound Site Radionuclides by Location (EG&G 1995). Primary radionuclides assessed for 

occupational exposure are

• Polonium (210Po),

• Plutonium (238Pu, 239Pu, 240Pu, 241Pu, 242Pu), and

• Tritium (3

H).

Secondary radionuclides of concern include the following:

• Radium (

226Ra) ), thorium (228Th), and actinium (227Ac),

• Thorium (

228Th, 230Th, and 232Th),

• Protactinium (231Pa),

• Uranium (238U, 235U, 234U, and 233U),

• Americium (241Am) and curium (244Cm),

• Rare isotopes (

221Fr, 217As, 229Th, 233Th, 233Pa, and 209Po), and

• Radium (228Ra, 226Ra, 224Ra, and 223Ra).
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5.2 PROCESS OPERATIONS INVOLVING PRIMARY RADIONUCLIDES

5.2.1 Polonium-210

Polonium-210 was used in early nuclear weapons as part of the triggering mechanism for the fission 

reaction. The devices, called initiators, were not themselves potent sources of neutrons until they 

were activated; but work with neutron sources was carried on concurrently with the weapons program. 

Mound’s predecessor sites, the Dayton Laboratory facilities (Units III and IV), separated 210Po from 

naturally occurring materials. By the time the Mound facility began polonium processing in 1949, the 

feed material was irradiated bismuth.

At the Dayton Laboratory, the separation and concentration steps of the polonium processing 

operation involved significant releases of airborne contaminants including the most dosimetrically 

significant, 210Po. The polonium facility at the Mound site, T (Technical) Building, was purpose-built to 

entrain the material in process piping and vessels and inside specially designed gloveboxes so that 

the process could be done remotely. Operators controlled the activity from a shielded operations 

gallery (Anthony et al. 1956, BWXT 2002). This had the effect of restricting or eliminating significant 

releases of radioactive materials under routine processing conditions.

Polonium-210 was present in the HH (Hydrolysis House) Building from 1949 to 1959, which generated 

sludge materials with 0.02 Ci/L total alpha activity. Extensive use of 210Po in the R Building involved 

research and electrodeposition studies from 1948 to 1951 and the manufacturing of Po-Be neutron 

source from 1956 to 1961. 

The SW Building was used for counting 210Po sources, neutron source measurement, source strength 

verification, and 209,210Po retrieval. The primary processing and separation of 210Po took place in 

T Building from 1949 to 1971. Facilities and activities included neutron source calibration, 

decontamination laboratories, a storage pool for irradiated bismuth slugs, neutron source processing, 

and analytical laboratories. Aqueous wastes that contain 208Po, 209Po, and 210Po have been present in 

the WD and WDA waste disposal buildings from 1949 to the present. Chemical forms included 

polonium nitrate and polonium chloride. There was no cited 210Po activity in the PP (Plutonium 

Processing) or SM (Special Metallurgical) Buildings. A single 210Po incident was reported in the SW

(Semi-Works) Building in 1961, and 11 incidents are associated with the T Building from 1954 to 

1968, where primary polonium separation and processing took place.

The second phase of the polonium project consisted of manufacturing polonium heat sources for 

radioisotope thermoelectric generators (RTGs). The Systems for Nuclear Auxiliary Power (SNAP)-3A 

polonium-powered RTG was built in 1958. However, because the short half-life of 210Po limited the 

useful life of the RTGs, this program was halted after 1969 and cancelled in 1971. Cleanup of 

polonium areas continued through approximately 1974.

5.2.2 Plutonium

Plutonium has been associated with Mound site research activities conducted for the U.S. Atomic 

Energy Commission (AEC) beginning as early as 1949. In addition to ongoing research activities, 

neutron sources that were fabricated using 239Pu and 238Pu for research projects were transferred 

from the Lawrence Livermore National Laboratory to Mound in 1959. As polonium production tailed 

off, 238Pu activities transitioned into a large-scale effort to fabricate initiators, neutron sources, and fuel 

for RTGs and other devices using heat from the alpha decay of the fuel. Some AEC research 

programs carried out at Mound used 239Pu mixtures that are referred to in this document as “weaponsgrade plutonium.” The plutonium for RTGs and other heat sources was primarily composed of 238Pu, 

which is typically called “heat-source” plutonium in Mound historical documents and in this TBD. The 
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 12 of 60

radionuclides and proportions of these mixtures are listed in Tables 5-10 and 5-11 later in this 

document.

5.2.2.1 Plutonium-238

Feed material from the Savannah River Plant as PuO2 powder was enriched in 238Pu and processed 

at Mound to manufacture heat sources. Early 238Pu activities took place primarily in the SM Building. 

The PP Building was built to supplement and replace the 238Pu processing areas in SM.

SM Building activities included production of plutonium microspheres, fabrication of sealed sources, 

and plutonium recovery from high-risk aqueous waste. The microsphere manufacturing process in 

SM Building produced very fine PuO2 particles in the submicrometer to 20-μm range (EG&G 1995).

The microspheres would have been quite stable, and their initial particle size distribution would only 

be affected by depositional processes that remove particles above 10 μm from worker breathing 

zones (BZs). Deposition would therefore tend to increase the inhalation of submicrometer particles. 

Reported solubility varied widely from type S in Room SM-60 to type M in Rooms SM-27 and SM-28. 

Use of a plasma torch process (primarily in SM-39) resulted in PuO2 fumes.

Encapsulation, X-ray inspection, and calibration of Pu-Be neutron sources were conducted in 

T Building from 1955 to 1985. Counting and verification of sources was conducted in various 

laboratories of T Building. Two 238Pu exposure incidents were reported in 1986.

All SM processes were transferred to the PP Building in December 1967. The primary mission in this 

building was to process PuO2 powder from the Savannah River Plant. A number of processes were 

developed to manufacture plutonium microspheres in the PP Building and involved the following:

• PuMo cermet process to coat plutonium particles from 43 to 250 μm with molybdenum from 

1969 to 1973,

• Pressed PuO2 process to fabricate plutonium spheres for encapsulation as heat sources from 

1970 to 1979,

• Milliwatt Generator Program process to seal 105- to 500-μm plutonium shards with yttrium,

• Solid solution process from 1969 to 1974,

• Solution gel process from 1969 to 1971, and

• Microsphere process to produce 50- to 250-μm 239PuO2 ceramic spheres from 1967 to 1970.

The Mound site supplied RTGs for a number of space exploration missions, including Pioneer, 

Voyager, and Apollo. In 1977, production of the PuO2 fuel itself was moved to the Savannah River 

Site, and activities at Mound were oriented around assembly of the RTG capsules (DOE 1993). 

Mound produced more than 500 RTGs from 1961 to 1992. In another 238Pu program, 234U was 

separated from aged 238Pu from 1970 to 1980. The heat source plutonium isotopic mixture listed by 

EG&G (1995) is given in Table 5-10 later in this document.

5.2.2.2 Plutonium-239

Plutonium-239 was separated from neutron-irradiated uranium metal from 1949 to 1952. Reactor fuel 

research was conducted with plutonium from 1956 to 1959. The Neutron Source Group started 

production of heat sources in the R Building using 239Pu in 1956. To date, NIOSH has not determined 
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 13 of 60

the exact mixture of plutonium that made up the weapons-grade material, though one source 

describes it as “more than 95% pure 239Pu with trace amounts of 240Pu and 241Pu” (EG&G 1995). For 

the purposes of dose reconstruction, the mixture resulting in the greatest ingrowth of 241Am was 

selected as a favorable-to-the-claimant assumption, in lieu of specific data. This mixture is listed in 

Table 5-11 based on Hanford plutonium (Battelle 2003).

5.2.3 Tritium

EG&G (1995) reported that tritium (3H) recovery, purification, and use for solid tritiated metal 

compounds took place in 1948 in R building and that tritium, limited to testing and research activities,

was used in T building as early as 1947 (sic2

, EG&G 1995). However, this is the only reference that 

provides information about the presence of tritium on the Mound site before the research program 

described by Meyer (1992) started in 1957. Because no other sources show that tritium was 

processed at Mound before this time, and there is no indication of a bioassay program, for this 

revision of the Mound site profile, no provisions are made to estimate dose from tritium before 1957.

In 1969, the retrieval of tritium from retired weapons systems began, which involved the analysis, 

recycling, separation, recovery, purification, and storage of tritium. In addition, weapons component 

testing and some research activities were conducted with metal tritides in SW Building and with 

tritiated organics in T Building. Tritium was consequently used in all buildings except the PP Building 

and the SM Building. Extensive use of tritium occurred in R, SW, and T Buildings. These activities 

resulted in tritium-contaminated soils under SW Building (EG&G 1995). Eight tritium incidents 

reported by EG&G (1995) occurred in SW Building from 1964 to 1986. Tritium exposures ended in 

1995 as reported by EG&G (1995).

5.3 PROCESS OPERATIONS INVOLVING SECONDARY RADIONUCLIDES

5.3.1 Radium, Thorium, and Actinium

Because of the short half-life of 210Po, the initiators required frequent remanufacture. For this reason, 

the Substitutes program was initiated in the late 1940s to identify longer-lived alpha emitters that 

could be used instead. Through research at Mound and the Los Alamos National Laboratory (LANL), 227Ac was identified as the most promising replacement for 210Po (even though 227Ac is primarily a beta 

emitter, its short-lived progeny are potent alpha sources), and research was conducted on refinement 

and purification. The feed material was neutron-irradiated 226Ra activated as shown below.

226 227 227

88Ra η Ra Ac β 88 89

− +→ → +

Process chemical research projects were carried out at Mound in the late 1940s and early 1950s 

using nonradioactive barium and strontium as chemical surrogates of the radium and actinium. For 

some activities, they were labeled with 140Ba and 90Sr radiotracers.

Once production-scale processing began, the actinium processing operation proved to result in 

unprecedented personnel exposures from airborne alpha emitters and high external dose rates in the 

operation, which was conducted in a containment facility called the “Old Cave” in SW-1 and SW-2. In 

July 1953, the program was deemphasized; large-scale actinium processing was discontinued in early 

1954. Decontamination and decommissioning (D&D) of the actinium facility took place through early 

1959 and ended in February. This activity resulted in large internal doses from radioactive gases and 

airborne particulates associated with the feed material and the process itself. Bioassay was 

performed for the involved individuals, who are identified in Mound records (Meyer 1994), but

 2 T Building was not occupied before February 1948; polonium operations began in September of that year (BWXT 

2002a). 
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uncertainties associated with the method, and the likelihood of exposure of individuals who were not 

monitored under the internal dose monitoring program, led to the addition of a class of Mound 

employees to the SEC.

5.3.2 Thorium

Major processes at Mound that involved thorium were the Monex refinery and the Ionium project. 

Monex was the name of a Mound Laboratory proprietary process to refine natural thorium, 232Th, for 

possible use as a reactor fuel. Ionium activities at Mound were part of a complex-wide project to 

stockpile 230Th (ionium), also for potential uses in reactors. 

5.3.2.1 Th-232 Source Term

The principal 232Th feed material was monazite sludge. This was residue left after monazite sand 

milling. The milling process is believed to have involved treatment with a caustic to solubilize the 

phosphates, decanting the soluble phosphates, and leaching with acids to extract the rare earth 

elements. The caustic treatment was likely to have been a high-temperature process that would have 

resulted in such refractory and highly insoluble thorium compounds as ThO2. The material that was 

received by Mound was sometimes basic and sometimes acidic.

Mound’s activities with 232Th began with a preliminary proposal for a thorium refinery that was 

prepared in 1954 (Monsanto 1954). Mound was authorized to construct the refinery on March 11, 

1955 (Wende 1955). The refinery was anticipated to process approximately 128 kg of thorium per 

hour (Huddleston 1955). In anticipation of this project, Mound received a large quantity of thorium 

sludges (approximately 6,000 55-gallon drums) from United Lead Company of Middlesex, New 

Jersey, between December 10 and 27, 1954 (Meyer 1979). This material came to Mound via rail in 

64 boxcars (McCarthy 1955a). However, doubts about the entire breeder reactor program were being 

expressed in private by AEC as early as April 1955 (McCarthy 1955b), and the Mound refinery project 

was cancelled on May 6, 1955 (Meyer 1955).

This left Mound with the task of disposing of the thorium residues. On arrival at Mound, the drums 

were briefly stored in a warehouse and a garage, but increases in short-lived particulate progeny were 

observed. This prompted the decision to move the drums to an unfrequented warehouse. With the 

exception of 650 drums that had to be protected from the weather, which were briefly stored in a 

Quonset hut (Bradley 1954), most of the drums were eventually stored in an open field in the southern 

part of the plant site (Meyer 1979). Due to the exposed storage conditions and the corrosive nature of 

the sludges, drum deterioration and failure were a recurring problem over the next 12 years and 

required redrumming of the entire lot three times. The redrumming efforts were conducted in the 

outside storage areas during the summer months, and this resulted in contamination of the storage 

fields.

A two-bay storage basin known as Building 21 was completed in 1966, and the drums were emptied 

into this basin. The oxalate sludges (2.23 × 105 kg) occupied a smaller bay that occupied about 25% 

of the storage basin, and hydroxide sludges (1.2 × 106 kg) occupied the larger bay (about 75% of the 

basin).

The thorium residues were sold to General Atomics in 1974 and removed from the site by July 15, 

1975 (Meyer 1979). Removal of the sludge was done under contract, and the contractor provided its 

own health physics oversight and dose monitoring. The final phase of the Building 21 

decontamination was complete in September 1975. The completion of this effort also marked the end 

of the most significant exposure potential to 232Th at the Mound site. Only incidental exposure 

potential to residual contamination existed after this date.
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5.3.2.2 Ionium

A limited ionium research program was begun in early 1956. The program was to research 

separation of 230Th from source material. The program was completed in 1958, with the last bioassay 

sample submitted in September of that year.

Primary feed material for extraction of 230Th was the so-called Airport Cake (Cotter Cake). The 

feedstock was raffinate from extraction of uranium from a hydrofluoric or nitric acid solution using 

diethyl ether followed by treatment with caustic or lime. This feed source was obtained from 

Mallinckrodt and the Cotter Corporation.

5.3.3 Protactinium-231

An AEC program was undertaken in 1955 in the R Building to recover and purify a small amount of 

protactinium from natural sources. This limited research investigated 231Pa for possible use in 

initiators or as source material for RTGs. The separation and processing of 231Pa from Sperry Cake 

raffinates was conducted in the HH Building, Rooms HH-5 and HH-6, from 1955 to 1956. Separation 

of 231Pa from Sperry Cake was also reported in R Building (R-148, R-149, R-167) from 1954 to 1959. 

The program began to wind down in 1958, and was concluded in 1959. Only limited laboratory use 

was reported from 1956 to 1987.

Minor quantities of 231Pa were present in samples that were subjected to spectrographic analysis in 

R-140 and R-142 from 1956 to 1969. Minor concentrations of 231Pa were found in WD/WDA Building, 

which received aqueous wastes from R and SW Buildings. No 231Pa use was reported for SM, SW, T, 

or PP Buildings.

Protactinium-231 extraction and purification was a complex process that is poorly described. There is 

no information to estimate the radioactive compounds likely to have been present. Use of minimal 

quantities of 231Pa in laboratory isotopic studies and during spectrographic analyses of samples after 

1959 is assumed to be insignificant. Dose reconstruction should be conducted using available 

bioassay data from 1954 through 1959.

5.3.4 Uranium-233

A limited research project involving ten researchers took place in 1958 and 1959. Urine sampling 

started July 25, 1958, and ended July 1, 1959.

5.3.5 Reactor Fuels Projects

This program performed basic research on chemical and physical properties of proposed reactor fuels 

for several different types of reactors and reactor projects:

• Aircraft Nuclear Propulsion Project (ANPP). In 1953, Mound started a project to support the 

ANPP. The liquid cycle reactor was a fused salt design.

• Homogenous-Type Reactor. Begun in mid-1950s, this effort worked to identify satisfactory 

fuel systems. Mound measured the physical properties of uranium and thorium solutions and 

slurries at elevated temperatures and pressures. The program started in 1952 and used many 

measurement techniques from the ANPP program.

• Civilian Power Reactor Program. This program, started in 1957, was concerned with 

development of reactors for power and heat for civilian use; it examined the homogenous-type 

reactor using 233U, 235U, and 239Pu as fuel.
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The activities listed above investigated physical attributes and, other than sample preparation, 

involved only the required physical measurements:

• Plutonium fuel for fast breeder reactors. Mound was responsible for acquiring data on 

proposed fuel systems by determining density, viscosity, thermal capacity, thermal 

conductivity, and phase equilibriums of plutonium and plutonium alloys. These were Pu-Fe, 

Pu-Ni, Pu-Cu, Pu-Cd, Pu-Yt, and Pu-Va.

• In 1961 Mound participated in a project to experiment with the fabrication of plutonium-bearing 

glass fibers as a reactor fuel system.

These activities involved chemical processing of plutonium.

5.3.6 Reactor Waste Project

In January 1949, Mound began a study of processes for decontamination of reactor wastes for 

recovery of 239Pu. From 1949 to 1952, three different waste streams were researched: bismuth 

phosphate and tributyl phosphate from the reduction-oxidation process from Hanford, and the wastes 

from the plutonium-uranium extraction process pilot plant at Oak Ridge. The bench-scale research 

was conducted in R-167 using milliliter aliquots of the wastes. Based on the results, a pilot plant was 

constructed in early 1952 in M Building, where cold (nonradioactive) runs were made. When these 

were complete, the process equipment was set up in HH building and became operational by 

November 1, 1952. A volume of approximately 150 to 200 gallons was processed. Some bioassay 

samples for 1951 and 1952 were analyzed for plutonium and likely were the ones taken for this 

program. The program was terminated by February 1953. The pilot plant was dismantled, 

decontaminated, and stored in March 1953 (Meyer 1992, p 378).

5.3.7 Americium

Americium-241 is a constituent of weapons-grade plutonium (

239Pu); however, the percent of gross 

alpha activity in heat source plutonium attributable to 241Am is negligible. Where intakes of 241Am 

occurred as a result of intake of weapons-grade plutonium, the 241Am constituent should be estimated 

using the composition of weapons-grade plutonium in Table 5-11.

When not a constituent of weapons-grade or heat source plutonium, potential doses from 241Am

should be reconstructed separately; this is expected to be necessary for individuals working with 

Americium as part of research projects. This is due to the possibility that it posed a specific intake risk 

from alpha source production, neutron source production, analysis, and vaporization of plutonium 

compounds, 231Pa separation, and the reactor fuel research program.

Americium-241 alpha sources were electroplated in Room R-120 between 1956 and 1962, and it was 

the major isotope of concern in that room. Exposure to 241Am in R-120 could also have occurred 

coincident with intakes of weapons-grade plutonium.

5.3.8 Rare Radioactive Isotope Program

Mound conducted separation and purification of special heavy-element isotopes. The work was on a 

gram- or milligram-scale, and the isotopes were shipped to the Oak Ridge isotope pool. The 

radionuclides included:

• 231Pa.

• 234U obtained from “milking” 238Pu.
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• 229Th obtained from “milking” 233U. This program was to create 229Th to use as a source 

material for 221Fr, 217As, and other members of the 4n + 1 series. Thorium-229 is itself a 

natural product of 233U and was produced in the thorium fuel cycle by neutron irradiation of 

233Th. The 229Th project at Mound started in 1966. Work was performed in the hot cell in the 

SW Building. Mound used 1.5 kg of 233U to produce 30 mg of 229Th. The program concluded 

in early 1970s (Meyer 1992, p. 382).

• 233U from an aged source of irradiated 232Th to obtain 233Pa, which decays to 233U.

• 228Th, 227Ac, 209Po, and 226Ra. Began in the mid-1950s and continued until 1985. Separations 

carried out in SW, SM, and R Buildings.

Meyer (1992) states that routine bioassay programs covered all of these activities, but the bioassay 

method is likely to be based on the parent or more abundant radionuclide, such as 238Pu rather than 

the 234U harvested from it.

5.3.9 Other Radionuclides

Xenon and krypton were present in H Building as a part of the Isotope Separation Program and xenon 

isotope separation activities. For the purpose of dose reconstruction, external dose is the exposure of 

concern and internal dose is considered negligible.

Curium-244 and 237Np were present at the site in minor concentrations in special nuclear heat source 

material from the Savannah River Site. They were also present in aqueous wastes derived from the 

research-scale plutonium recovery operations. Curium-244 and 237Np were also used in metallic or 

oxide form in reactor fuel research and to support the heat source program. Curium-244 has a 

relatively high dose conversion factor (DCF) (less than a factor of 2 below 239Pu and 232Th) and 

therefore the potential to induce relatively high doses via the inhalation pathway should be taken into 

account by the dose reconstructor. However, both 237Np and 244Cm were used in minor quantities and 

were of secondary radiological concern.

Mound engaged in biomedical research with 45Ca, 59Fe, 60Co, 65Zn and other radionuclides. These 

were present in relatively minor quantities in comparison to the primary and secondary radionuclides 

and were used in very few buildings for limited periods. Most of these radionuclides were present as 

sealed sources. Worker exposure was less likely, and bioassay data was not collected for these 

radionuclides. Exposure to these radionuclides can therefore be considered insignificant.

5.3.10 Radon

Radon is a naturally occurring radioactive gas that emanates from materials in the earth’s crust; 

however, when natural levels are enhanced by processing or accumulation of radioactive materials 

that were used, handled, or stored at DOE facilities, the dose must be considered along with other 

potential intakes. Radon is itself a noble gas, but its particulate progeny that are deposited in 

respiratory tract organs result in significant doses. 

Few data are available to use for Mound dose reconstruction before the landmark study that was 

conducted in 1989 (UNC Geotech 1990). However, the SW Building experienced high levels of radon 

from the decay of radioactive materials generated during the radium-actinium processing effort in the 

Old Cave, despite serial remediation activities. A study occurred in the late 1970s to identify the 

proportions of radon, actinon, and thoron, (alternative names for 222Rn, 219Rn, and 220Rn, each deriving 

from the predecessor radionuclide), and an attempt was made to measure the concentration. It has 

been determined that these data do not allow the calculation of potential doses with sufficient 

accuracy to estimate a maximum dose. For this reason, NIOSH is unable to determine an upper 
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bound for the radon exposures in SW-19 until the point at which a ventilation system was installed in 

early 1980 (before March 5, 1980) to vent SW Building air. The inability to bound the concentration in 

SW-19 extends to the contiguous air spaces in the SW and R Buildings. The lack of sufficient 

information to bound the dose from radon and its progeny formed the bases for the SEC class 

definitions in SEC-00171 (Sebelius 2010) and SEC-00207 (NIOSH 2012).

Thorium sludge from 232Th extraction and purification operations was stored outside in drums from 

1954 to 1966. In 1966, thorium sludge was stored in 4,914 drums in Building 21. All thorium sludge 

drums were removed from Mound in 1975. In 1976, 1,258 drums of Cotter concentrate were stored to 

support purification of 231Pa and 230Th. Table 5-1 shows the results of various radon surveys at 

Mound.

Table 5-1. Radon isotope survey data by building and period.

Building

Survey 

dates

Gas concentration (pCi/L) WL WLM/yra

Radon Thoron Actinon

SW Tunnelb 10/12/1979 88,000 28,000 640,000

SW19 after vent 03/1980–04/1980 7.7–13.4 0.03 0.4

12/1989 2.6 Not detected

12/1999 1.0–5.c

04/1990 0.7–7.2

1998–2000 5.8–12

Building 21 05/1983 17.5–52.8

12/1989 34–118 0.3–0.7 0.4–1.3

0.85 median

4.7–15.3

04/1990–06/1990 117–161 10.0 median

125

Building 48 12/1989 3.2

Old SD 12/1989 1.5–2.4

Fire station 12/1989 1.2–1.4

Paint shop 12/1989 0.9–1.2

A 10/1988–02/1989 0.1–1.1

All other buildings 12/1989 0.4–0.9

Outside facilities 05/1983 0.24–0.59

Reported 

background

0.1 to 2.1

Average = 0.5

a. WLM/yr values were determined as WL (2,000 hr/yr) (1/170 hr/WLM).

b. The tunnel under SW was not an occupied workspace; concentration provided for information only.

c. Bolded concentrations were derived from charcoal canister measurements.

Areas in the SW Building experienced high concentrations of radon, thoron, and actinon that 

emanated from residual radium, thorium, and actinium from the remediated SW-1 and SW-2 Rooms; 

some materials had been disposed in place and a new floor built above them. The area was renamed 

SW-17, -18, and -19. In addition to contaminated components and fill material, indications were that 

concentrations of radium, thorium, and actinium were high in the soil underneath the concrete flooring. 

Noble gases were transported through an underground tunnel into SW-17, -18, and -19. A ventilation 

system was installed in early 1980 to vent SW-17 tunnel air to an SW-18 exhaust.

The installed ventilation system reduced 222Rn concentrations in SW-19 by a factor of 10 or more, to 

8.2 pCi/L in March 1980. This result was confirmed with a passive environmental radon monitor

(PERM) measurement of 15.4 pCi/L and RDT310A continuous measurements from 7.7 pCi/L to 

13.4 pCi/L. A working level (WL) measurement by the employee’s desk was 0.03 WL or 0.03 

(12 mo/yr) = 0.4 working level month (WLM) in comparison with an occupational limit of 4 WLM/yr.

Radon surveys in Building 21 were conducted with PERMs in 1983, and the results ranged from 

17.5 to 52.8 pCi/L with WLs from 0.40 to 1.3. WLMs exceeded the occupational limit of 4 WLM with 

calculated values from 4.8 to 15.6 WLM (Author unknown, 1983a, 1983b, 1983c). Radon that was 
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measured during the radon study with charcoal canisters ranged from 34 to 118 pCi/L in 1989 (UNC 

Geotech 1990). Building 21 was an unoccupied silo for storage of thorium feed material.

A DOE radon study was conducted from December 12 to 15, 1989, to measure radon in various 

Mound buildings (UNC Geotech 1990). The majority of buildings had radon concentrations below 

1.0 pCi/L 222Rn, except SW and Old SD buildings. In June 2000 a radon study summary report was 

issued based on 1990 and 1999 measurements (Doody 2000). Mound site radon background was 

reported to be 0.5 pCi/L with a range of 0.1 to 2.1 pCi/L. The SW Building, specifically room SW-19, 

was identified as the only area of potential occupational exposure to 222Rn and 220Rn.

Thoron and actinon measurement results were very limited despite the high concentrations that were 

observed in the underground tunnel. The only actinon measurements were made in Building 21 at 

0.3 to 0.7 pCi/L, but the holding time before analysis was excessive in comparison with the half-life of 

219Rn (3.96 seconds). The only thoron measurement in SW-19 was below detection.

5.4 INTERNAL DOSE MONITORING

5.4.1 General Description

Mound site bioassay results consist of in vitro urine data and limited fecal sample results. The 

primary source of sampling data for Mound dose reconstruction is in vitro results in employee records. 

Nose swipe data were considered to be secondary information. Whole-body counting is not an 

efficient monitoring method for Mound’s primary radionuclides (tritium, 210Po, and 238Pu) or for many

secondary radionuclides such as 239Pu, 231Pa, Ra, or Ac. Table 5-2 summarizes Mound urinalysis 

bioassay programs with reported radionuclide action levels, monitored worker populations, and 

monitoring frequency.

Table 5-2. Bioassay programs and action levels.

Bioassay program Period Action levelsa Monitored workers Frequency

In vitro monitoring

Po-210 processing 1945–1974 10 cpm/50 mL Operations personnel Weekly spots

Weekly 24 hr

Ac-227, Ra-226, Th-228 1951–1960 8.3 cpm/24 hr Research personnel Weekly 24 hr

Th-232 sludge, redrumming 1955–1975 2.8 cpm/24 hr Th refinery personnel Monthly 24 hr

Th-230 (ionium) research 1956–1959 3.5 cpm/24 hr Research personnel Bi-weekly 24 hr

Pa-231 extraction 1956–1960 2.2 cpm/24 hr Research personnel Monthly 24 hr

U-233 research 1958–1960 20 cpm/1,500 mL 10 research personnel Weekly 24 hr

Tritium (hot gas) production 1957–present 10 μCi/L Recovery personnel Weekly spots

Weekly 24 hr

Pu-239 neutron source 1957–present 3.5 cpm/24 hr Processing personnel Weekly spots

Weekly 24 hr

Pu-238 heat source 1960–present 3.5 cpm/24 hr Processing personnel Weekly spots

Weekly 24 hr

Am-241 7/1989–

present

0.023 pCi/Sampleb Potentially exposed 

personnel

Daily

Cm-244 7/1989–

present

0.009pCi/Sampleb Potentially exposed 

personnel

Daily

In vivo monitoring

Whole-body counting 1960–1970 Not available Operations personnel Annual

Chest counting 1970–present Variablec Operations personnel Annual

a. Observed cpm for 50 mL or 24-hour samples that result in a worker dose of 300 mrem/wk to the target organ.

b. Barton (1995); for results prior to this date, dose reconstructors should base MDA assumptions on actual data or refer to 

the ‘Table of Other Radionuclides’ (Author unknown 1955-1988).

c. Late MDAs at the Mound site ranged from 0.1 nCi for Am-241 to 130 nCi for Pu-239.
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Routine bioassay monitoring included collection of spot and 24-hour urine samples depending on the 

radionuclide to which specific employees were assigned. Monitoring frequencies were described by 

Meyer (1992) and inferred from personnel sampling records. For programs that used spot sampling, 

such as 210Po and 238Pu, individuals suspected of exposure during incidents, or who had urinalysis 

results in excess of the reporting limits, were required to submit additional urine 24-hour samples.

Administrative personnel were not monitored and monitoring was discontinued for operational 

personnel if their work on specific projects was no longer needed.

Mound used the action levels in Table 5-1 to evaluate and control worker doses to a 300-mrem/wk 

target organ dose limit. Results in excess of action levels triggered resampling and placement of the 

worker on the “hot roster” to control further exposures if necessary. Reporting limits for americium 

and curium were 0.0, and any detectable activity was reported and triggered resampling. The 

dosimetry controls in place by 1949 when Mound operations were moved from the Dayton Laboratory 

facilities were consistent with the limits stated in Handbook 52 published in 1953 (NBS 1953).

5.4.2 In Vitro Monitoring

In vitro monitoring programs were established for each activity at the Mound site as described in 

Meyer (1992) and summarized in Table 5-3. Although Mound maintained the ability to analyze 

samples for specific radionuclides, many analyses were conducted using the gross alpha procedure. 

Monitoring protocols for each program are listed below in the appropriate paragraphs; because the 

gross alpha procedure applied to many Mound bioassay programs, it is described, in general, in the 

section below.

Table 5-3. Summary of bioassay methodologies.

Nuclide Method/description Period MDA

H-3

(hot gas)

Vibrating reed electrometer 1957–1961 1.0 μCi/L

Liquid scintillation (LS) 1962–1995 0.01 μCi/L

Spontaneous electrodeposition from 

raw urine, low-background 

proportional counting

1944–1963 Spot samples

6.0 dpm

168 dpm/d

Po-210 Spontaneous electrodeposition from 

acidified urine, low-background 

proportional counting.

1964–1973 Spot samples

0.7 dpm

20 dpm/d

24-hr samples

1.3 dpm/24 hr

Radium extraction and activity 

inferred from 227Th or 223Ra dpm 

assuming secular equilibrium

08/17/1955–1956 0.3 dpm 24-hr samples

0.15 dpm/24 hr

Pa-231 Alkaline earth phosphate 

coprecipitation, Pu separation with 

cerium carrier gross alpha 

proportional counting

1957–1958 0.02 dpm

1959–12/07/1959 0.08 dpm

Before 1957 (0.85) pCi/sample

Pu-238, 

Pu-239

Above with sample acidification 1957 (0.94) pCi/sample

1958 (0.6) pCi/sample

1959 (0.5) pCi/sample

1960 (0.5) pCi/sample

1961–1966 0.05 pCi/sample

Above with anion exchange, 

electrodeposition, gross alpha 

proportional counting

1967–05/1981 0.11 pCi/sample

Above with alpha spectroscopy for 

Pu-238 and Pu-239/240, Pu-242 

tracer, blanks counted, DLs quantified

06–12/1981 0.066, 0.044 pCi/sample
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Nuclide Method/description Period MDA

Radium separation, intakes and 

excreted Ra-226 and Th-230 

assumed to be in equilibrium.

1982 0.063, 0.042 pCi/sample

1983 0.064, 0.043 pCi/sample

1984 0.066, 0.045 pCi/sample

1985 0.047, 0.031 pCi/sample

1986 0.042, 0.028 pCi/sample

1987 0.022, 0.011 pCi/sample

1988 0.017, 0.01 pCi/sample 

01–07/1989 0.019, 0.008 pCi/sample

08/1989–12/1991 0.018, 0.008 pCi/sample

1992–present 0.018, 0.009 pCi/sample

Before 01/1994 0.3 dpm

Th-230

(Ionium)

Th-230 extraction, alpha 

spectroscopy.

01/1994–1995 0.17 dpm

Thorium separation, proportional 

counting.

Before 12/1987 0.3 dpm

Th-natural Thorium separation, alpha 

spectroscopy.

12/1987–1994 0.02 (Th-232) dpm

01/1994–1995 0.08 (Th-228) dpm

LiF fusion, fluoroscopy. 1950s–1985 0.4 μg/L

Uranium Uranium separation, proportional 

counting.

Before 1985 0.3 dpm

Uranium separation, alpha 

spectroscopy.

05/1985–1995 0.07 (U-234) dpm

0.03 (U-235) dpm

0.09 (U-238) dpm

Americium or curium separation, 

proportional counting.

1950–1989 0.2 pCi

Am-241, 

Cm-244

Americium or curium separation, 

alpha spectroscopy.

Before 07/1989 0.3 pCi

07/1989–1995 0.05 Ci

5.4.3 Gross Alpha and Sequential Separation

A procedure was developed at the Mound Laboratory for extraction of alpha emitters from human 

urine to assess the concentrations of radium, actinium, and thorium in MLM-1003 (Kirby and 

Brodbeck 1954). This procedure was later used for or adapted to a number of radionuclides over the 

operating period of the internal dose monitoring program at Mound.

The gross alpha procedure method involved initial concentration of a 24-hour urine sample by 

coprecipitation with alkaline earth phosphates and the addition of ammonium hydroxide; alkaline earth 

phosphates were separated with a cerium phosphate carrier. Alpha counts of the mounted cerium 

phosphate fraction resulted in a gross alpha count that could be assumed to be composed of the 

radionuclide an individual worked with, and this is typically what occurred over the long history of the 

program.

As an alternative after the cerium carrier step, sequential chemical separations could isolate 

plutonium, americium, neptunium, and uranium, respectively, in subsequent steps. Thorium, curium, 

transplutonic radionuclides, strontium, and polonium were not absorbed on the ion exchange column 

and would be present in the rinse operation that was conducted as step 2 of the plutonium separation.

The procedure itself was simple and provided consistent results; processing through the cerium 

carrier step gave acceptable results for plutonium or uranium as long as other radionuclides were not 

present. The procedure specific to plutonium and isotopes of thorium, uranium, americium, and 

others was published in a formal analytical procedures manual, MD-20736, in 1968 (Meyer1992,

p. 335). The gross alpha method was preferable due to its simplicity, and it was typically used for

individuals who worked only with 238Pu, as all alpha activity could be assumed to result from 238Pu. 
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5.4.4 In Vivo Dose Monitoring

Whole-body counting has most likely been performed since 1960 when the Whole Body Counting 

Facility was completed. However, whole-body counting to assess accidental lung deposition of 238Pu, 

239Pu, and 241Am began in 1969. Chest counting began in the early 1970s using phosphor sandwich 

(phoswich) detectors.

Extensive modifications were made in the counting room to reduce natural background radiation and 

prevent influx of airborne contaminants using activated charcoal filtration. By 1982, dual 5-in.-

diameter thallium-activated phoswich scintillation detectors were employed to detect radionuclides in 

the lungs. In addition, a single 4- by 8-in. sodium iodide detector was mounted under the counting 

bed in 1982 for in vivo counting of high-energy photon sources. All monitored persons were 

decontaminated before counting for a 2,000-second period. Detectors were placed directly on the 

upper chest while the person was lying flat on the counting bed. Chest wall thickness was determined 

with an ultrasonic transducer and used to quantify lung burdens using phantom calibration curves, but 

the documentation does not state when chest wall thickness measurements began. In vivo count 

records are not kept in a central electronic database; each count is maintained as a single hard-copy 

record. A whole-body count record and a chest count record on the same day are listed on a single 

hard-copy record.

Reported minimum detectable activities (MDAs) in 1990 ranged from 69.6 to 73.7 nCi for plutonium 

and 0.18 nCi for americium based on four female and four male subjects (Author unknown 1990). 

Personnel records from 1994 reported MDAs associated with chest counting of selected individuals 

and 171 counts (Author unknown 1993–1994).

MDAs were relatively high for the plutonium isotopes due to weak X-ray emissions and a half-value 

layer in tissue of only 0.6 cm for 239Pu (Langham 1964). In vivo chest counting was therefore 

conducted as a secondary indication of worker contamination in conjunction with primary routine 

bioassay urinalysis. MDA data have been found to be limited or unavailable for periods before 1990. 

Dose reconstruction should therefore be based on urine bioassay results to assess missed worker 

doses. Chest-counting data should only be used to assess intake when counting was conducted as a 

follow-up for a known worker intake; it is not expected, based on experience with Mound claims, that 

this will be a common occurrence. Dose reconstructors needing to reconcile chest count results with 

urinalysis results should contact the lead dose reconstructor for the Mound site when MDA data are 

not listed with the in-vivo count results. 

5.4.5 Personnel Air Sampling Data

A complex array of particulates, aerosols, dusts, and fumes was present during Mound operations. 

Representative area or personnel air sampling of these airborne materials would require sophisticated 

sampling equipment, procedures, and calibration efforts. In addition, area air sampling cannot be 

used effectively in some workplace scenarios and can greatly underestimate actual worker inhalation 

intakes. Obtaining air sample data would probably entail a time-consuming search of archived site 

records.

Personnel BZ air sample data should be used in preference to area air sampling data in instances 

where intakes cannot be estimated from bioassay data and no other means of estimating dose is 

possible. Area air sampling data should only be used to assign dose if it can be demonstrated to be 

representative of BZ concentrations. In addition, the DOE claims data do not generally provide air 

concentration data for nuclides other than tritium.
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5.5 BIOASSAY METHODOLOGIES FOR PRIMARY RADIONUCLIDES

5.5.1 Polonium

5.5.1.1 Bioassay Methods

All Mound employees who worked in polonium operations in T, HH, WD, H, B (Biological), and R 

Buildings were required to submit weekly spot urine samples (Meyer 1992). Monitoring reports 

indicated that the number of monitored workers increased steadily from 1959 to 1963. Samples were 

usually collected on Monday and Wednesday, or three samples were collected on Monday, 

Wednesday, and Friday. Samples were initially treated without HNO3 digestion. Polonium in raw 

urine samples was spontaneously deposited on nickel disks and counted using a low-background 

proportional counter.

Using an improved method with acid digestion, spot urine samples of 50 mL were collected each 

week and preserved with 50 mg of sulfamic acid before analysis as reported in MLM-460 (Spoerl 

1950a). Preserved polonium samples were digested in the plating cell with nitric acid or hydrochloric 

acid to produce a 1 N final solution before a 2-hour electrodeposition plating time. Electrodeposition 

was conducted primarily on 1-in. copper disks within a few hours of collection. Recovery was 

reported in 1950 as 86.1 ±11.9% for 50-mL samples in Spoerl (1950a) and Meyer (1992). Spoerl 

(1950a) reported that 95% of all recoveries would fall between 62% and 110% based on 310 samples 

and the assumption of normality. However, three outliers were discarded from the original dataset 

and the complete dataset distribution exhibited nonnormality. Spoerl (1950a) also reported that 

samples with activities in excess of 200 cpm/50 mL had reduced recoveries from 56% to 64%. 

However, urine sample activity above 200 cpm/50 mL is limited to only 0.5% of the total polonium 

bioassay records in the PORECON database. The procedure stated that 210Po losses might occur 

during the time between sample collection and sample assay (Spoerl 1950a). No specific detection 

limit (DL) was reported for this method of analysis in Meyer (1992).

All urine bioassay results were reported as cpm in 50 mL of urine and were converted to dpm per 

50 mL using an assumption of 2π.

Chemical Recovery Correction

A study conducted in 1989 determined that significant recovery losses from raw undigested urine 

were observed for metabolized 210Po due to plating and/or formation of precipitates (Fellman et al. 

1989; Meyer 1992). This study was conducted using urine samples from primates by investigators at 

the New York Medical Center and by Spitz and Robinson at Mound and published in Health Physics

(Fellman et al. 1989). Metabolized polonium was shown to have a low average chemical recovery of 

10% without complete acid digestion or oxidation by ashing the sample. An average ratio of 10.5 was 

reported for ashed to unashed urine sample results (Fellman et al. 1989). Recovery of the added 

unmetabolized polonium tracer typically ranged from 75 to 85% (Fellman et al. 1989) as previously 

observed at Mound (Spoerl 1950a). Mound had originally assumed that metabolized polonium was 

plated with the same efficiency as the unmetabolized tracer. Actual chemical recoveries at Mound 

were therefore potentially only 10% rather than the assumed 86%. Reported bioassay urinalysis 

results would have to be corrected by a factor of 10 up until 1964 when complete acid digestion of 

urine samples was required (Meyer 1992).

Dr. Spitz and Dr. Skrable accepted the 10% chemical recovery and applied this correction to evaluate 

occupational doses from 210Po in a retrospective study in 1989 (Spitz 1989a; Skrable 1989; Themalis 

1989). In addition, DOE requested that Mound use the 10% chemical recovery to conduct 210Po 

dosimetry (Spitz 1989b). A peer review committee was established in 1987 to study polonium dose 

methodology. This committee agreed in 1991 that the 10% recovery should be adopted for intake 

assessment (Spitz 1991).
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 24 of 60

MJW Corporation conducted a subsequent evaluation of the chemical recovery of polonium (MJW 

2002a, Appendix B). The position paper concluded that the 10% recovery observed by Fellman et al. 

(1989) was not appropriate for correction of Mound worker bioassay results due to differing sample 

storage times after collection but before analysis. The majority of Mound sample analyses were 

thought to have occurred within a 4-hour timeframe after collection (Spoerl 1950a). In addition, MJW 

stated that the 10% chemical recovery for polonium was based solely on primate data and possibly 

not representative of recoveries observed for human urine samples. MJW therefore decided to use 

the original recovery that was reported by Spoerl (1950a) of 86%.

However, the data that was reported by Fellman et al. (1989) corroborated early recoveries that were 

reported by Mound in 1950 for undigested urine samples and demonstrated similar estimates of 

urinary 210Po excretion for primate and human subjects using undigested urine sample procedures. 

Differences in sample storage times between the original Mound analyses and the Fellman study did 

not have any significant effect on reported recoveries. Fellman et al. reported recoveries that ranged 

from 79% to 85% for all human and primate undigested urine samples in comparison to Spoerl 

(1950a) at Mound with an average recovery of 86 ±11.9%. In addition, the Fellman et al. study is in 

good agreement with a previous Argonne National Laboratory study that was published in 1971 about 

seven individuals who were exposed to inhalation of 210Po (Sedlet and Robinson 1971). This Argonne 

study also observed that in contrast to the added 209Po tracer, “only a small fraction of the metabolized 

and excreted polonium could be deposited directly from acidified urine without preliminary oxidation.”

Dose reconstruction is performed using the favorable-to-the-claimant assumptions described in 

Section 5.8.1. 

Reported Half-Life

The effective half-life of 210Po was investigated in 1950 and reported by Spoerl (1950b) to be 36 days. 

Dose calculations that were used at Mound assumed this effective half-life from 1950 to the end of 

polonium monitoring. Various quality factors (QFs) that ranged from 10 in the 1950s to 20 in the 

1980s were used to determine dose after polonium exposure. International Commission on 

Radiological Protection (ICRP) Publication 30, Part 1, reported a 50-day biological half-life (ICRP 

1979). However, Sheehan (ca. 1951) reported a much shorter effective half-life of only 24 days for 

exposure that resulted from a puncture wound involving a platinum wire contaminated with 210Po at 

Mound.

5.5.1.2 Polonium Internal Dose Monitoring Records

Polonium-210 bioassay data were documented on several different types of records at Mound as 

discussed by Meyer (1992). Weekly Reports of urinalysis data were kept for workers who were 

grouped by location with entries in cpm per specified volume and 24-hour samples identified. Weekly 

results were transcribed onto Polonium Cards for individual workers with entries in cpm per 50 mL

unless otherwise stated. Twenty-Four Hour Reports were also kept and used to convert individual 50-

mL spot samples that were collected on a single day to a volume-weighted average 24-hour excretion 

in cpm. Finally, Special and Unscheduled Sample Forms were kept when any additional urine 

samples were required to document possible worker exposures or to confirm previous results with all 

entries in cpm and duplicate indicated. Bioassay results in activity per day are reported as becquerels 

per day in the PURECON database under the heading of BQ_DAY. Sample results are also reported 

in cpm and dpm in this database.

Bioassay data for 210Po have been placed in the PORECON database by MJW by evaluating 

individual dosimetry files from original logbooks (MJW 2002b). This database contains individual 
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 25 of 60

sample results from 1944 to 1973, which encompasses polonium operations at the Dayton 

Laboratory3 facilities as well as the Mound Laboratory.4

Primary and secondary records exist in most Mound claim files for polonium bioassay. The dose 

reconstructor should typically use the bioassay results listed in the PORECON database as most 

convenient. However, it is important to review all claim records to ensure the database reflects all 

listed bioassay results in the primary data.

5.5.2 Plutonium

In 1952, an attempt was made to electroplate plutonium onto copper plates by following the procedure 

for 210Po. This method proved to be inconsistent and resulted in poor recoveries due to the inability to 

extract plutonium. An improved method was developed in 1954 and documented in MLM-1003 (Kirby 

and Brodbeck 1954). This newer method used an alkaline earth phosphate coprecipitation process 

followed by separation of plutonium with a cerium carrier and counting with a low-background 

proportional detector (Kirby and Brodbeck 1954; Meyer 1992). Recoveries for this method were 

reported to be 90% or greater. However, erratic recoveries from 6% to 85% were observed for 

metabolized plutonium in urine due to plating. This method continued from 1954 to 1960. A chemical 

recovery of 10% (based on the analysis of reduced polonium recoveries) should be assumed for 

plutonium bioassay data that was collected from 1954 to 1960.

Starting in 1961, urine samples were acidified to a pH of 2 with nitric acid, which resulted in recoveries 

for metabolized plutonium of 90% ±3% (Sheehan, Wood, and Kirby 1963). Sample preservation was 

modified to treat raw urine samples with alkaline phosphates to coprecipitate plutonium and prevent 

plating of metabolized plutonium present as organic complexes or colloids (Sheehan, Wood, and 

Kirby 1963). This method was largely continued until 1980 when anion exchange was permanently 

introduced to separate plutonium from the other alpha emitters. Anion exchange followed by 

electrodeposition was first introduced in 1966 with the MD-20736 procedure (Monsanto 1968). This 

procedure was modified in 1978 as MD-80030 to include use of a 242Pu tracer, counting of blanks to 

quantify DLs, and alpha spectroscopy (Phillips 1988). However, gross alpha results for bioassay were 

reported as plutonium for all 238Pu workers from July 1966 to 1971. Specific isotopic results for 

plutonium were not reported from 1971 to 1979 until the introduction of alpha spectrometry. Detected 

alpha activity from 1952 to 1980 was therefore assumed to be plutonium. These results might be 

overestimates of the true plutonium urine results due to the possible presence of uranium.

Table 5-4 summarizes the MDAs for plutonium isotopes in 24-hour bioassay urine samples over time 

(MJW 2002b). The lower limit of detection recommended by the Health Physics Society (HPS 1985) 

for accreditation of bioassay laboratories was 0.06 pCi/L for plutonium as reported in the Quarterly 

Environmental Science Reports in Meyer (1992). This standard was met in all years of analysis for 

238Pu and 239,240Pu except where MDAs in parentheses were corrected for a 10% chemical recovery.

From 1981 to 1983 average recovery of 85% and counting efficiency of 30% were reported for 

3,000-minute counts, and a lower recovery of well below 85% was reported for 1,000-minute counts 

(MCP 2004). Approximately 90% of all samples were counted for 1,000 minutes. 

 3 The polonium bioassay program described below was also used in Units I, III, and IV of the Monsanto Chemical 

Company. Although this document provides dose reconstruction provisions only for the Mound Laboratory, polonium 

dose may be calculated for the earlier facilities by using the assumptions listed below for the period before 1964. 

4 Possible exposures to 210Po in 1943 are assumed to be insignificant based on limited work in a single laboratory and 

reported administrative activities before initiation of bioassay in 1944 (Silverman 1962). Some bioassay data from 1944 

through the first half of 1945 was reported missing (Spitz 1989a).
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 26 of 60

Table 5-4. Summary of historical plutonium analytical procedures.

Period Reporting levela Analytical method

1954–1960 Kirby and Brodbeck 1954

0.05 dpm/1.5

0.5 dpm/1.5 L (corrected DL)b

Alkaline earth phosphate coprecipitation; plutonium separation 

with cerium carrier; gross alpha for plutonium; proportional 

counting; recoveries 6%–85% due to colloidal plating of 

metabolized Pu.

1961–1967 Kirby and Brodbeck 1954

0.1 dpm/1.5 L

Sample acidification; alkaline earth phosphate coprecipitation; 

plutonium separation with cerium carrier; gross alpha for 

plutonium; proportional counting; recoveries 90% ±3%. 

Uranium and plutonium results reported separately only in 

1966 using anion exchange methods.

1968–1977 MD-20736 (Monsanto 1968)

0.1 dpm/1.5 L

Alkaline earth phosphate coprecipitation; plutonium separation 

with cerium carrier; anion exchange; electrodeposition; gross 

alpha reported for plutonium; proportional counting. 

1978–1979 MD-80030 (Phillips 1988)

0.1 dpm/1.5 L

Alkaline earth phosphate coprecipitation; plutonium separation 

with cerium carrier; anion exchange; electrodeposition; alpha 

spectroscopy for Pu-238 and Pu-239/240; Pu-242 tracer.

1980–present MD-80030 (Phillips 1988)

0.07 dpm/1.5 L for routine 

samples

Alpha spectroscopy for Pu-238 and Pu-239/240; anion 

exchange; Pu-242 tracer; blanks counted for decision level 

determination.

a. The 24-hour reporting levels are based on a standard sample volume of 1.5 L. Standard Man in ICRP Publication 2 

(ICRP 1959) and Reference Man in ICRP Publication 23 (ICRP 1975) both have a urine volume of 1.4 L. Actual 

collected urine volumes varied.

b. DL was corrected for a low chemical recovery of 10% due to colloidal plating of plutonium.

Plutonium Internal Dose Monitoring Records

Primary and secondary records exist in many Mound claim files for plutonium bioassay. The dose 

reconstructor should typically use the listed bioassay results in the PURECON database as most 

convenient. However, it is important to ensure that the database reflects all bioassay results that are 

listed in the primary data. The dose reconstructor will also need to ensure that PURECON reflects 

earlier measurements using the gross alpha methodology and ensure that the gross alpha samples 

were taken as a plutonium intake monitoring program rather than another radionuclide that used the 

gross alpha bioassay method.

5.5.3 Tritium

No urinalysis data were available from 1948 to 1956; according to Meyer (1992), tritium urinalysis for 

workers started in 1957. A vibrating Reed Electrometer method of analysis was officially adopted 

from a Savannah River Plant procedure in 1958 and continued into the early 1960s. This method 

used 20 mL of urine mixed in a calcium chloride solution to liberate tritium gas, then analyzed the gas 

by passing an acetylene carrier gas from the urine sample into a 1-L metal ionization chamber. The 

measured current was therefore proportional to the tritium-acetylene mixture. The reported MDA for 

this method was 1 μCi/L in urine in Recommended Procedures for Tritium (Meyer 1992). Estimates of 

accuracy were 1 to 2 μCi/L at 1 μCi/L, 9.5 to 10.5 μCi/L at 10 μCi/L, and 99 to 101 μCi/L at 100 μCi/L

(Meyer 1992).

Liquid scintillation (LS) counting began in 1961 followed by an approved procedure in 1968. This 

methodology used a 3-mL urine sample placed in an Ultimate Gold cocktail solution (Butler 1961). A 

quench curve was used to correct results for counting efficiency, which was affected by chemical and 

physical factors that inhibit photon detection. In addition, many liquid scintillation counters (LSCs) 

used external 133Ba or 226Ra sources to automatically determine the degree of quench for each 

sample. Duplicate analyses were conducted to support quality evaluation. The reported MDA for this 

method was improved from early ionization detection methods to 0.01 μCi/L of tritiated water vapor 

(HTO) (Phillips 1988). Tritium results were always reported in units of activity per liter of urine. Table

5-5 lists the reported analytical procedures and DLs for tritium.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 27 of 60

Table 5-5. Reported analytical procedures and MDAs for tritium.

Period Analytical method MDA (μCi/L)

1957–1961 Vibrating reed electrometer 1.0

1962–1995 LS 0.01

Tritium Dose Calculations

A method is presented in this section to convert recorded tritium doses based on LANL dosimetry that 

was used at Mound to current ICRP Publication 67 and 68 dosimetry (ICRP 1994a, 1995). Correction 

from ICRP Publication 30 (ICRP 1979) to Publication 67 is not needed due to an overall correction 

factor of 1.0 as shown in Equation 5-3 below.

A dosimetry method developed at LANL was used for acute tritium dose calculations from 1957 

through 1988 (Lawrence 1957). This methodology assumed a maximum permissible body burden of 

3.7 mCi, a relative biological equivalent of 1.7, an effective half-life of 12 days, and a body water 

compartment of 43.4 kg. This results in a maximum permissible concentration (MPC) of 85 μCi/L for a 

target dose limit of 300 mrem/wk under the LANL system of dosimetry (Lawrence 1957; DOE 1988). 

The target dose limit of 300 mrem/wk was based on an organ dose limit of 15 rem (FRC 1960). The 

LANL dose relationship used at Mound is shown in Equation 5-1:

Dt B ( ) 0.00873 = (5-1)

where

B = measured acute body concentration in urine or body water (μCi/L)

D(t) = the dose commitment at time t after exposure (rem)

Effective T1/2 = 12 days

Reported doses on the Radioactive Gas Exposure Evaluation forms confirm that Mound used 

Equation 5-1 for tritium dose calculation and that 6 μCi/L corresponds to 52 mrem as entered on the 

forms. However, this equation should not be used to reconstruct dose and only provides the 

confirmed basis to correct doses reported by Mound using Equation 5-2.

Current dosimetry for tritium uses a quality factor (QF) of 1.0 in comparison to an assumed LANL 

relative biological effectiveness of 1.7, a two-compartment elimination model, and a target mass of 

68.8 kg in comparison with 43.4 kg used in the LANL calculations [ICRP Publication 67 (1994a); 

Publication 68 (1995)]. Equation 5-2 provides a ratio of Publication 68 dose in comparison with LANL 

dose, which results in a ratio of 0.34:

0.693 /10 0.693 /40

0

0.693 /12

0

(0.97 0.03 ) 43.4 (I / ) 0.34

68.8 1.7

t t

t

e e dt

Ratio CRP 68 LANL

e dt

∞ − × − ×

∞ − ×

× +× = × =

×

∫

∫ (5-2)

Beginning in 1989, tritium doses were reported in terms of effective dose equivalent (DE) under ICRP 

Publication 30 methodology (ICRP 1979). Equation 5-3 determines the ratio of Publication 68 (ICRP 

1995) dose in comparison Publication 30 using the Publication 67 target mass of 68.8 kg (ICRP 

1994a). The ratio of 1.0 confirms that no dose corrections would be required for doses reported from 

1989 to 1995.

0.693 /10 0.693 /40

0

0.693 /10

0

(0.97 0.03 ) 63 ( ) 1.0

68.8

t t

t

e e dt

Ratio ICRP 68 / ICRP 30

e dt

∞ − × − ×

∞ − ×

× +× = × = ∫

∫ (5-3)
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 28 of 60

Tritium Reporting Limits

Mound reporting limits for tritium affected what doses were reported for monitored workers. Doses in 

rem or microcuries were reported as blanks on Mound Tritium Exposure Reports when urine 

concentrations were less than 6 μCi/L (Meyer 1992). Reporting limits of 3 μCi/L as discussed by 

James (1997) or 6 μCi/L were both reported on the Radioactive Gas Exposure Evaluation Forms as 

actual bioassay results. 

Values of 0.0 were used for nondetection results, which biased the doses below the true estimate of 

the average dose. MDAs in Table 5-5 should therefore be used to estimate intakes at or below 

detection.

Stable Metal Tritides

Limited open information is available on stable metal tritides (SMTs) to which workers could have 

been exposed. Uranium and lithium were most likely used to trap tritium gas in closed tritium handling 

systems. In general, these materials are reactive with air and water, especially when they are in a 

finely divided state. Uranium tritide reacts with air to produce HTO. A DOE report notes that lithium 

tritide reacts with water vapor to produce mostly HT (DOE 1994).

5.6 SECONDARY RADIONUCLIDES INTERNAL DOSE MONITORING PROGRAMS

5.6.1 Radium-Actinium

Section 2 of Meyer (1992) provides a description of bioassay procedures used for 226Ra, 227Ac, and 

228Th. A detailed description of the initial procedure is entitled Indirect Radiochemical Urinalysis for 

Actinium, Thorium and Radium Isotopes (Meyer 1992). Additional detail was found in MLM-773 

(Kirby 1952), MLM-1003 (Kirby and Brodbeck 1954), and MD-20736 (Monsanto 1968). Radium was 

extracted from collected 24-hour urine bioassay samples. Intakes of 227Ac and 228Th were thought to 

be approximately in equilibrium with their radium progeny, subject to the caution that the irradiated 

226Ra was allowed to age for several weeks before being processed in the SW Building.

Radium extracted from urine samples was subjected to a series of gross alpha counts for various 

periods of up to 30 or 35 days. The resultant count data for a single sample was manipulated to form 

a series of simultaneous linear equations that could be solved for individual count rates due to 226Ra, 224Ra, and 223Ra. Each bioassay urine sample of interest was apparently evaluated in this manner. 

Interpretation of these bioassay data requires some rather involved calculations.

Due to the large uncertainties implicit in the reconstruction of dose based on these procedures, 

NIOSH has determined that it cannot estimate internal exposures to radium, actinium, and thorium 

from the arrival of K-65 sludge in October 1949 through February 28, 1959, when the related area of 

work was decontaminated and decommissioned and sufficient monitoring was in place. The lack of 

sufficient information to bound the dose from these exposures, even for workers with specific 

bioassay monitoring results in their records, formed the basis for the SEC class definition in SEC00090 (Leavitt 2008).

Radium-226, Actinium-227, and Thorium-228 Records

Although dose from this bioassay method may not be calculated for partial dose reconstructions, the 

records are described to allow the dose reconstructor to discriminate between these records and 

records for other bioassay protocols.

Primary bioassay data for these radionuclides were found in four brown spiral notebooks labeled 

RA#1a and RA#1c and on forms O-318. Manila folders also contained primary bioassay data with 

collected counts, assumptions, equations, theoretical curve computations, and a written analysis of 

the observed data (Meyer 1992).
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Secondary urinalysis results were written on hand-drawn forms and placed in the “J. B. Black binder.” 

Each form had an employee name, sample identification number, and the 223Ra, 224Ra, 226Ra, and 

227Ac results. Secondary data also appeared in various forms such as weekly reports, and on forms 

O-756 and O-634. The Site Research Database (SRDB) contains available secondary bioassay 

records for the Radium Irradiation project, which are typically included in the DOE claim files. As 

stated above, these must be reviewed to separate results from the radium-actinium program, which 

may not be reconstructed, to other thorium results. 

5.6.2 Thorium

5.6.2.1 Thorium-230

Twenty-four hour samples were taken for the urinalysis program for the ionium activities. A thenoyl 

trifluoroacetone (TTA) extraction procedure was used to identify the 230Th. There is some evidence 

that the MLM-1003 procedure was used as well, and thorium separation was performed (Meyer 1992, 

p. 270).

Thorium-230 bioassay procedures are described briefly in MLM-MV-93-93-0003, History of Mound 

Bioassay Programs, Sections 1 and 2 (Meyer 1992). Thorium-230 bioassay was based on 24-hour

urine specimens. The objective of the urine bioassay sampling was to detect an MPC that 

corresponded to a 300-mrem/wk radiation exposure to the critical organ per the metabolic model in 

use at the time. A count rate of 3.5 cpm for a 24-hour urine sample corresponded to a DE of 300 

mrem/wk. Based on Meyer (1992, Section 2), three different urine bioassay procedures were tried for 

thorium. A TTA urine extraction was the initial procedure (Kirby and Brodbeck 1954). Starting with 

sample 6507-1I, nine samples were analyzed by the “Kirby process-thorium fraction.” This apparently 

refers to the thorium separation and isolation procedure in MLM-1003 from August 1954 (Kirby and 

Brodbeck 1954, p. 14).

Urine bioassay was subsequently changed to the strategy of radium separation in September 1954 

(Meyer 1992). This approach was based on a pair of speculative assumptions. First, Mound 

appeared to have assumed that intakes of 230Th were in equilibrium with 226Ra. The average ratio of 

the 230Th to 226Ra in the materials that were handled by Mound workers is not known. However, it is 

well known that sulfuric acid leaching of uranium ore materials results in a raffinate solution that is 

enriched in thorium and depleted in radium, and in tails that are enriched in radium and depleted in 

thorium. The assumption of equilibrium might not be correct. The assumption of secular equilibrium 

should therefore not be made as discussed and recommended by Meyer (1992). Second, Mound 

appears to have assumed that the excreted 226Ra was in equilibrium with 230Th. This assumption can 

be compensated for by using ratios of the incremental urine excretion curves from Potter (2002). 

Excretion curves for acute and chronic exposures can also be readily calculated using IMBA. ICRP 

Publication 68 assigns radium to absorption type M, but thorium can be either type M or S (ICRP 

1995).

Thorium-232

Monitoring for 232Th began in 1955 to support thorium refining; preliminary research began in the R 

Building. Exposures to 232Th and progeny could have continued until thorium bulk storage areas were 

cleaned out (by July 15, 1975). Several rooms in Buildings 38, PP, WD/WDA, HH, SW, and SM had 

activities that could have resulted in exposure to 232Th.

Thorium-232 Records

Thorium-232 records are diverse due to programs that were conducted over many years. Primary 

232Th bioassay data were entered into a small brown spiral notebook marked “Radium-Thorium” and 

“Radium-Thorium Separation from 8/15/1955 to 2/2/1959” (Meyer 1992). Additional primary 232Th 

bioassay data were recorded in a large hardcover record book. However, the first 38 pages from this 
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 30 of 60

record book, for July 6, 1959, to January 9, 1961; July 19, 1964, to November 15, 1964; and May 30, 

1965, to June 6, 1965, were removed. Secondary records in weekly reports contained 232Th results 

as cpm per 24-hour sample beginning March 17, 1958. In August 1959, secondary results were 

reported on form O-756. The available data is a table of secondary 232Th bioassay data extracted 

from other primary records (Author unknown 1955–1988). Meyer (1992) states that bioassay results 

for this operation will always have both radium and thorium entries. The corresponding primary 232Th 

records from the 1950s carry a sample number that ends in “T.” This analysis assumed the “T” result 

is from the thorium (which includes 232Th and its progeny 228Th) fraction and the “R” indicates that 

results are from the radium fraction for a given sample.

5.6.3 Protactinium-231

Urinalysis was started for workers to monitor separation and purification program activities in 

R Building on August 17, 1955 (Meyer 1992). Bioassay data for 231Pa concluded on December 7, 

1959 (Meyer 1992). Bioassay was limited to collection of 24-hour samples. A significant volume of 

urine was required to detect an MPC target dose of 300 mrem/wk due to the very slow elimination of 

231Pa.

Samples were subjected to various precipitation and dissolution procedures to extract radium, 

thorium, and protactinium fractions (Kirby 1952; Kirby and Brodbeck 1954; Kirby 1959). Purification of 

231Pa from natural sources was discussed by Kirby and Figgins (1964), Kirby (1959), and Salutsky et 

al. (1956). Residues were initially filtered from aqueous raffinate solutions after a diethyl ether 

extraction of uranium. Filtrates were generally dissolved in HCl saturated with NaCl, precipitated with 

a silicate or cerium carrier and NaOH or NH4OH, digested in 9 N HCl or HNO3 solutions, and either 

counted the precipitate directly or isolated by anion exchange.

The thorium fraction results in cph were converted to cpm, corrected for a 50% counting efficiency, 

and entered directly into dpm in the database Author unknown (1955–1988). Thorium fraction activity 

results were then assumed to be 231Pa with a small contribution from ingrowth of decay products. 

Decay products present in average-age composite solutions at Mound were reported to contribute 

only 0.05% of the total alpha count (Kirby and Figgins 1964). Analyses were later conducted in 

accordance with procedure MD-20736 (Monsanto 1968).

5.6.4 Uranium

For uranium, samples were processed using the MLM-1003 procedure (Kirby and Brodbeck 1954). 

Although Meyer’s statement is not definitive, it appears that most routine samples were spot samples 

that were processed through the cerium-carrier stage. For investigative samples, 24-hour samples 

were collected and uranium was extracted using the sequential separation protocol (Meyer 1992, 

p. 321).

Both spot and 24-hour samples were collected from workers for bioassay. The maximum permitted 

concentration of uranium in urine bioassay samples was 20 cpm in a 1,500-mL urine sample. 

According to Section 2 of Meyer (1992), the uranium gross count rate for spot samples was scaled up 

to 1,500 mL on the primary record to determine a 24-hour sample result.

Site-specific values for background uranium dietary intake at Mound have not been listed for Mound 

workers. Typical intakes of uranium attributable to background should be on the order of 4.9 Bq/yr 

238U, 4.9 Bq/yr 234U, 0.2 Bq/yr 235U (UNSCEAR 1993), and 12 Bq/yr for total uranium (NCRP 1992). 

ICRP Publication 23 (ICRP 1975) reports that uranium background urinary excretion rates for 

Reference Man range from 0.07 to 0.7 dpm/d in the urine and 2.1 to 2.7 dpm/d in the feces. A single 

reference (Author unknown undated a) lists a detection level in mass units for natural uranium (NU). 

This is listed in a 488-page file that is a collection of references with the title of “Historical MDAs at the 
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 31 of 60

Mound site,” and the page in question has “Unknown Document 1” hand-written at the top and is titled 

“Appendix A, Detectable Amounts for Urinalysis.” There is no evidence that this is a Mound document 

and, because the bioassay methods listed 90Sr and 147Pr, neither of which was ever performed at 

Mound, it is not likely that it refers to Mound specifically.

5.6.5 Americium-241 and Curium-244 Records

Documentation for 241Am and 244Cm is very limited. Primary records for 241Am and 244Cm are 

unknown. Secondary data for 241Am and 244Cm are found in Author unknown (1955–1988), which 

indicate that 244Cm results are related to the plutonium logbooks. All 241Am and 244Cm urinalysis data 

are in units of activity (picocuries). MJW (2002b) assumed that 241Am and 244Cm data in Mound 

dosimetry records such as forms O-756 are in units of activity excreted per 24-hour period. There are 

no aliquot data in the database.

5.6.6 Rare Isotopes

For most rare isotopes, the parent or more abundant radionuclide is likely to present the greatest dose 

consequence based on the small amounts of the radionuclides processed. For example, in the 

production of small amounts of 234U from a 238Pu “cow,” it is the plutonium that represents the highest 

potential for dose.

5.7 INTERNAL DOSE RECONSTRUCTION GUIDANCE

Most estimates of internal dose are based solely on urinalysis data. For urinalysis, standard 

parameters are used rather than site-specific values.

5.7.1 General Considerations

5.7.1.1 Use of the Gross Alpha Method

Gross alpha analyses were conducted at Mound and, until 1980, activity was assigned to specific 

isotopes of plutonium and other alpha emitters based on sequential gross alpha counting or chemical 

separation followed by gross alpha counting. Alpha spectrometry capable of identifying specific alpha 

emitting isotopes was not introduced until 1980.

5.7.1.2 Internal Dose Estimation Parameters

Particle Size

This project assumes a particle size of 5 micrometers activity median aerodynamic diameter (AMAD).

Absorption Type

Rather than identifying the most likely absorption type based on the form of the chemical compounds 

of the radionuclides worked with, the dose reconstruction project makes a determination of the 

appropriate absorption type based on a fit to the data in the case or based on efficiency measures 

that assign the absorption type that results in the higher or highest dose for overestimates or the lower 

or lowest dose when underestimates of internal dose are appropriate. Note that an absorption type 

specific to Mound has been identified based on historical internal dose measurements at Mound and 

at LANL. Guidance has been developed to estimate doses to high-fired forms of 238Pu, which can

under some circumstances exhibit nonmonotonic clearance rates (ORAUT 2013a). The type L model 

developed from Mound site data is unlikely to be typical of Mound site exposures to 238Pu, but each 

potential exposure should be evaluated for applicability of this model. Type L is modeled using 

specific IMBA input parameters in Table 5-15 later in this document.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 32 of 60

Exposure Dates

The assumptions described in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2007a),

are used to establish appropriate estimated intake dates when case data do not suggest an 

appropriate date to use.

5.7.2 Equilibrium Assumptions

Gross alpha counting was conducted on bioassay urine samples using a procedure of thorium or 

radium chemical extraction. Individual radium bioassay samples were counted three times in a lowbackground proportional counter over a period of weeks. Then the individual radium isotope activities 

were estimated by solving a system of simultaneous equations. This procedure requires equilibrium 

assumptions to properly infer activity associated with individual decay chain radionuclides. For 

example, sequential gross alpha counting was conducted on bioassay urine samples after radium 

extraction by coprecipitation to quantify 223Ra, 224Ra, and 226Ra activity. Evaluation of alpha activity 

with decay time allowed the activities of individual radium isotopes to be estimated. Table 5-6 lists 

equilibrium assumptions that are favorable to claimants to infer primary radionuclide activity in urine 

based on detected alpha activity.

Table 5-6. Recommended equilibrium assumptions for secondary radionuclides.

Radionuclide Method of inference Recommendations

Th-230 Gross alpha counting of extracted 

thorium in urine starting with 

sample 6507-1I in August 1954. 

Gross alpha counting of extracted 

radium in urine beginning with 

sample 6507-10I from September 

1954 to September 9, 1958. 

Assume Th-232 and Th-228 have minimal 

radiological impact to workers from uranium feed 

material. Assume equilibrium and equal intakes for 

Th-230 and Ra-226 for exposure to feed materials. 

Assume MDA for negative bioassay radium result for 

demonstrated exposure to only thorium compounds.

Th-232b Gross alpha counting of extracted 

thorium and radium in urine for 

monazite refinery workers.

Assume Th-228, Th-232, and Ra-224 are all in 

equilibrium for exposure to feed materials.a Assume 

minimum possible ratio of 0.42 for Th-228/Th-232 

and equilibrium between Ra-224 and Th-228 for 

exposure to only thorium compounds for doses 

based on Ra-224 bioassay data.a

 Assume Th-228 

and Th-232 are in equilibrium for doses based on 

gross alpha bioassay data.

Uranium isotopes: 

U-233, EUc, 

U-234

Gross alpha counting of specific 

radionuclides.

No assumptions required. Assume all activity is 

attributable to U-233 in absence of specific 

radionuclide data. Use U-234 for EU.

Po-210 Electrodeposition followed by lowbackground proportional counting.Assume equilibrium with Pb-210 to determine worker 

intakes.

Uranium isotopes: 

NUd, DUe

Fluorometric or gross alpha 

determination.

Assume NU ratios for U-238, U-234, and U-235 or 

use IMBA default ratios for NU and DU.

a. IMBA does not directly calculate excretion of progeny due to intake of parent radionuclides with progeny.

b. Th-232 is also identified as natural thorium and Th-230 as ionium.

c. Uranium enriched in U-234 and U-235. The degree of enrichment was not reported. These uranium isotopes would not 

be present as isotopically pure materials.

d. Uranium with naturally occurring mass and activity ratios.

e. Uranium depleted in U-234 and U-235.

Mound did bench-scale reworking of K-65 uranium extraction residue that was used at Mallinckrodt, 

Weldon Springs, and Fernald as a potential source of radium. There was also an initial attempt to 

produce 210Po initiators by reworking radium and uranium tails. This process was quickly abandoned 

after the first year. All pre-1946 210Po intake estimates (at Monsanto Chemical Company facilities) 

should therefore assume equilibrium with 210Pb.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 33 of 60

Extraction operations using Mound uranium cake involved possible exposure to all thorium isotopes 

from this particular source term. However, 232Th and 228Th had very low activity percentages (0.4%) in 

comparison with 230Th. Therefore, these thorium isotopes would have resulted in a minimal 

radiological impact to workers. Because of the relatively low activity abundance of 232Th and 228Th in 

this source term, it is reasonable to assume for dose reconstruction purposes that all of the activity is 

from 230Th. Depending on solubility type, this could result in a slight underestimate of dose of less 

than 2%. Both 232Th and 228Th can therefore be ignored when estimating the intake and doses 

associated with 230Th.

In vivo data

In vivo records for Mound employees are sparse; the site lagged behind in the use of chest counts for 

alpha emitters, and whole-body counts were performed only in the later history of the internal dose 

monitoring program. Some claims do contain whole-body count records, but these often carry a 

notation saying “Practice Exam” and may not be considered valid records for this reason. Due to low 

gamma emitter use at the Mound site, valid whole-body count records should only be used to 

estimate a missed dose in cases where case data imply a clear exposure to beta-emitting 

radionuclides.

Chest count results may only be used to verify urinalysis results.

Fecal sample data

Fecal samples were also infrequent during most of the history of the Mound site’s internal dose 

monitoring program. When fecal data are present in claim records, they should be reconciled with 

urinalysis data in estimating the internal dose. Guidance for this is given in ORAUT-OTIB-0060

(ORAUT 2007a). MDAs for fecal data are listed in the table below.

Average percent recoveries for plutonium, uranium, thorium, and americium for fecal samples were 

determined from an analysis of fecal sample data collected from July to October 1989 Oak Ridge 

Associated Universities (ORAU) Team files (Author unknown undated b). Table 5-7 summarizes 

these chemical recoveries as determined using 242Pu, 232U, 229Th, and 243Am as tracers. Specific 

MDAs for these fecal analyses could not be quantified without reported background sample counts. 

However, the MDAs appeared to range from 0.01 to 0.04 pCi/sample for 238Pu and from 0.001 to 

0.008 for 239Pu, assuming that reported error terms associated with the sample activities were 1 

sigma.

Table 5-7. Radionuclide chemical recoveries for collected 

fecal samples.

Nuclide Samples Average Minimum Maximum

Pu-242 27 0.77 0.32 1.02

U-232 9 0.69 0.13 1.05

Th-229 2 0.89 0.74 1.03

Am-243 2 0.77 0.71 0.82

For dose reconstruction, resolution of fecal samples may be performed using the MDAs listed in 

Table 5-8 below. These MDAs were reported as part of a retrospective project to establish MDAs for 

Mound bioassay procedures conducted in 1995 (Author unknown undated).

Table 5-8. Retroactive MDAs for fecal samples.

Radionuclide

MDA 

(pCi/sample) Earliest date

Pu-238 0.012 December 1989

Pu-239 0.010 December 1989

Am-241 0.023 July 1989

U-234 0.27 January 1990
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U-238 0.28 January 1990

5.8 RECONSTRUCTING DOSE FROM PRIMARY RADIONUCLIDES BASED ON 

URINALYSIS DATA

This section provides specific information to dose reconstructors for estimating internal dose to Mound 

workers from the primary radionuclides.

The goal of the internal dose monitoring program at Mound was to conduct bioassay for all programs 

that could result in internal deposition of radioactive material (Meyer 1992). For this reason, for most 

claims urinalysis data are abundant for primary radionuclides and the chances that an individual 

experienced significant intakes in the absence of bioassay data are remote. However, for individuals 

who, based on case-specific information, can be presumed to have experienced intakes that cannot 

be based on bioassay data, coworker intakes have been developed as documented in ORAUT-OTIB0061, Internal Coworker Doses for the Mound Site (ORAUT 2012a). For those claims that dose 

reconstructors determine that the use of coworker dose is necessary, justification should be included 

in the dose reconstruction report.

5.8.1 Polonium-210

Polonium-210 sample results in the PORECON database are typically complete, but primary data are 

also present in claim files and should be reviewed to ensure the accuracy and completeness of the 

PORECON results. Note that PORECON includes results as far back as 1944; results from before

1949 would be for the Dayton Laboratory facilities. Although dose estimation for Dayton Laboratory 

facilities is outside the scope of this TBD, doses from potential intakes of polonium may be calculated 

using the dose estimation recommendations in this section.

Urinalysis data for 210Po collected before 1964 should be normalized assuming 85% extraction 

efficiency, then corrected for a 10% chemical recovery. Polonium-210 data that was reported after 

1963 should be normalized assuming 85% extraction efficiency and then corrected for a 63% 

chemical recovery based on the analysis by Fellman et al. (1989). Polonium intakes were controlled 

using cpm per sample, and were not corrected for aliquot sizes. Sample volumes as reported should 

always be used when available. Default sample volumes are 50 mL before July 13, 1964, and 100 

mL on and after July 13, 1964. Spot samples and 24-hour samples were analyzed using the same 

analysis procedure. The required corrections are summarized in Equations 5-4 and 5-5:

1944–1963 (5-4)

1964 onward (5-5)

All urine bioassay results were reported as cpm in 50 mL of urine and were converted to dpm per 

50 mL using an assumption of 2π geometry and Equation 5-6:

2 dpm cpm

mL mL

= × (5-6)

Missed doses should be estimated using MDAs recorded with the results; however, when MDAs are 

not listed with results, dose reconstructors should use the derived MDAs in Table 5-9.

0.85 1400

0.50 0.10 

210 Database cpm mL Po Daily Excretion Aliquot mL

= ××

 cpm 0.85 1400

0.50 0.63

210 Database mL Po Daily Excretion Aliquot mL

= ××
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 35 of 60

Table 5-9. Reported background count rates and MDAs for 210Po.

Period

Default sample 

volume (mL)

Detector 

efficiency

(%)

Recovery

efficiency

(%)

Reported

background

(cpm/24 hr)

MDAs

Spot samples

(dpm/50 mL)

24-hr samples

(dpm/24 hr)

1944–1963 50 50 10 0.2 6.0 dpm, 168 dpm/d 1.3

1964–1973 100 50 63 0.1 0.72 dpm, 20 dpm/d 0.16

5.8.2 Plutonium

In most cases, presumptive exposures of Mound employees to plutonium can be assumed to result 

from heat source plutonium. However, claim data must be reviewed to ensure that this is indeed the 

case, and that the worker was not also (or instead) exposed to potential intakes of weapons-grade 

material.

The dose reconstructor should keep in mind that both gross alpha and alpha spectrometric programs 

were used to detect 238Pu, and that the PURECON database results should be verified against the 

primary data in claim files, even though PURECON results are typically listed as gross alpha results 

when this is the case. Isotopic results that list both 238Pu and 239Pu begin to appear in 1984. This 

should not typically be presumed to represent exposures to both types of plutonium. Rather, this 

indicates that the spectrographic analysis described in Section 5.5.2 above was used.

The radionuclides associated with both plutonium source terms are listed in Tables 5-10 and 5-11 for 

heat-source plutonium and weapons-grade material, respectively.

Table 5-10. Heat source plutonium feed material 

characteristics.

a

Nuclide % Mass

% Activity by effective ageb

1.0 yr 3.2 yr 10 yr

Pu-236 0.0001 0.0 0.0 0.0

Pu-238 80.2 95.6 95.9 96.8

Pu-239 15.9 0.1 0.1 0.1

Pu-240 3.0 0.0 0.0 0.1

Pu-241 0.6 4.3 4.0 3.0

Pu-242 0.1 0.0 0.0 0.0

Am-241 0.0 0.0 0.0 0.1

a. Source: EG&G (1995).

b. Effective age is the age when 241Am activity is zero.

Plutonium urinalysis data from 1954 to 1960 should be normalized assuming 85% extraction 

efficiency and then corrected for a 10% chemical recovery due to plating of organic complexes or 

colloidal plutonium before sample analysis (Sheehan, Wood, and Kirby 1963; Spoerl 1950a). This 

correction is not required for later samples. These corrections are summarized in Equations 5-7 and 

5-8:

Table 5-11. Weapons-grade plutonium feed material characteristics.

a

Nuclide wt%

% Activity by effective ageb (Ci/g)

Fresh (0 yr) 5 yr 10 yr 15 yr

Pu-238 0.05 8.56E-03 8.23E-03 7.91E-03 7.6E-03

Pu-239 93.1 5.77E-02 5.77E-02 5.77E-02 5.77E-02

Pu-240 6 1.36E-02 1.36E-02 1.36E-02 1.36E-02

Pu-241 0.8 8.24E-01 6.48E-01 5.09E-01 4.00E-01

Pu-242 0.05 1.97E-06 1.97E-06 1.97E-06 1.97E-06

Am-241 0.00 0 5.83E-03 1.04E-02 1.39E-02

a. Source: PNNL (2003).

b. Effective age is the age when 241Am activity is zero.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 36 of 60

1957–1960 (5-7)

1960–2006 (5-8)

From 1981 to 1983, average recovery of 85% and counting efficiency of 30% were reported for 

3,000-minute counts, and a lower recovery of well below 85% was reported for 1,000-minute counts 

(MCP 2004). Approximately 90% of all samples were counted for 1,000 minutes.

The MDAs in Table 5-12 should be used to estimate missed dose. If the sample count time is 

unknown, then the 1000-minute MDAs should be used to reconstruct doses. Daily excretion should 

be determined using Equations 5-7 and 5-8 as appropriate.

Table 5-12. Plutonium MDAs (pCi/sample of urine).a

Period

Pu-238 MDAs Pu-239/240 MDAs

tc = 1,000 min tc = 3,000 min tc = 1,000 min tc = 3,000 min

Before 1957 (0.85) (0.85)

1957 (0.94) (0.94)

1958 (0.6) (0.6)

1959 (0.5) (0.5)

1960 (0.5) (0.5)

1961–1966 0.05 0.05

1967–05/1981 0.11 0.11

06–12/1981 0.066 0.027 0.044 0.020

1982 0.063 0.023 0.042 0.017

1983 0.064 0.024 0.043 0.018

1984 0.066 0.026 0.045 0.019

1985 0.047 0.017 0.031 0.012

1986 0.042 0.015 0.028 0.011

1987 0.022 0.011

1988 0.017 0.01

01–07/1989 0.019 0.008

08/1989–12/1991 0.018 0.008

1992–1995 0.018a 0.009a

a. tc = count time; plutonium MDAs from before 1957 to 1960 were normalized assuming 85% 

extraction efficiency and then corrected for 10% chemical recovery due to plating of organic 

complexes or colloidal plutonium before sample analysis. Based on Mound (1996).

Table 5-12 summarizes the MDAs for plutonium isotopes in 24-hour bioassay urine samples over time 

(MJW 2002b). The lower limit of detection recommended by American National Standards Institute 

(HPS 1985) for accreditation of bioassay laboratories was 0.06 pCi/L for plutonium as reported in the 

Quarterly Environmental Science Reports in Meyer (1992). This standard was met in all years of 

analysis for 238Pu and 239,240Pu except where MDAs in parentheses were corrected for a 10% chemical 

recovery.

5.8.2.1 Special Considerations for Plutonium-238

Plutonium-238 has a higher specific activity than 239Pu; for this reason its solubility is typically higher 

than weapons-grade plutonium mixtures. However, based on data from an incident at Los Alamos 

National Laboratory (LANL), and on an analysis of several results at Mound, some exposures to heat 

0.85

0.10

238,239 Pu database dpm Pu Daily Excretion × =

238,239Pu Daily Excretion Pu database dpm =
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 37 of 60

source material can exhibit nonmonotonic excretion curves. The dose implications of the complex 

excretion pattern are described by ORAUT (2013a). 

A dissolution model that better describes the unusual 238Pu urinary excretion patterns that have been 

observed at Mound was derived from five Mound intake cases reported by Guilmette, Griffith, and 

Hickman (1995), and described in ORAUT (2013a). The Mound-specific dissolution type is called 

type L to distinguish it from type J, a LANL-specific model. The type L dissolution model uses custom 

dissolution functions within the framework of the standard ICRP Publication 66 respiratory tract model 

(ICRP 1994b). The use of such custom dissolution functions is the approach recommended by the 

ICRP and allows the models to be implemented in IMBA.

There are certain intake scenarios for which the assumption of standard types M and S plutonium 

gives higher organ dose estimates than the assumption of type L plutonium, and there are scenarios 

where the assumption of type L plutonium gives higher dose estimates. The dissolution types that 

give the highest organ doses for typical dose reconstruction scenarios are identified in Tables 5-13 

and 5-14. For Table 5-13, the urine sample is assumed to be collected at the end of the listed chronic 

intake period.

Table 5-13. Summary of limiting dissolution type versus the duration of a 

chronic inhalation intake of 238Pu. 

Organ or tissue

Chronic intake period (d)

10,000 1,000 400 100 40 10

Adrenals S S S S L L

Urinary bladder S S S S L L

Brain S S S S L L

Breast S S S S L L

Gall bladder S S S S L L

Heart wall S S S S L L

Kidneys S S S S L L

Liver S S S S L L

Muscle S S S S L L

Ovaries S S S S L L

Pancreas S S S S L L

Testes S S S S L L

Thyroid S S S S L L

R.B.M. S S S S L L

Bone surface S S S S L L

Stomach S S S S L L

S.I. S S S S L L

U.L.I. S S S S L L

L.L.I. S S S S L L

Skin S S S S L L

Spleen S S S S L L

Thymus S S S S L L

Uterus S S S S L L

ET S S S S S S

Lung S S S S S S

Colon S S S S L L

ET1 S S S S S L

ET2 S S S S S S

LN(ET) S S S S S S

BBsec S S S S S L

BBbas S S S S S L

bb S S S S S S
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 38 of 60

Organ or tissue

Chronic intake period (d)

10,000 1,000 400 100 40 10

AI S S S S S S

LN(TH) S S S S S S

Esophagus S S S S L L

Table 5-14. Summary of limiting dissolution type versus the time after an acute 

intake a single urine sample is collected.

Organ or tissue

Day sample collected after acute intake (d)

10,000 1,000 400 100 40 15 10 2

Adrenals M S S S S S L L

Urinary bladder M S S S S S L L

Brain M S S S S S L L

Breast M S S S S S L L

Gall bladder M S S S S S L L

Heart wall M S S S S S L L

Kidneys M S S S S S L L

Liver M S S S S S L L

Muscle M S S S S S L L

Ovaries M S S S S S L L

Pancreas M S S S S S L L

Testes M S S S S S L L

Thyroid M S S S S S L L

R.B.M. M S S S S S L L

Bone surface M S S S S S L L

Stomach M S S S S S L L

S.I. M S S S S S L L

U.L.I. M S S S S S L L

L.L.I. M S S S S S L L

Skin M S S S S S L L

Spleen M S S S S S L L

Thymus M S S S S S L L

Uterus M S S S S S L L

ET S S S S S S S L

Lung S S S S S S S L

Colon M S S S S S L L

ET1 S S S S S S S L

ET2 S S S S S S S L

LN(ET) S S S S S S S S

BBsec S S S S S S S L

BBbas S S S S S S S L

bb S S S S S S S L

AI S S S S S S S S

LN(TH) S S S S S S S S

Esophagus M S S S S S L L

The overall conclusion is that standard plutonium dissolution types (M and S) should be assumed for 

cases where urine samples are collected more than two weeks after an acute intake and for chronic 

intakes in excess of three months in duration. Under these conditions types M and S will give the 

highest organ doses for all organs (i.e., type M or S plutonium is limiting). The type L dissolution type 

should be used for cases that do not meet these criteria. Therefore, in practice, type L should be 

added to the evaluation regimen and applied in the same fashion as types M and S, selecting 

whichever gives the largest dose for the particular case.

Parameters that relate to the type L model are reproduced in Table 5-15 along with those for types M 

and S.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 39 of 60

Table 5-15. IMBA input parameters for types M, S, and the

nonmonotonic model L.

Rate constant

Absorption type

M S L

kp,t 90 100 0.02

kt,blood 0.005 0.0001 0.003341

kp,blood 10 0.1 1.00 E-06

kr 100 100.1 0.02

ks 0.005 0.0001 0.003341

fr 0.09995 0.001 -0.20048

5.8.2.2 Special Considerations for Plutonium-239

It might be necessary for the dose reconstructor to infer exposure to one or the other of the two 

plutonium source terms based on claim-specific information. For the most part, 239Pu exposures 

would be to workers in laboratory settings. During part of the time, R Building results were noted as 

such in bioassay logs and represented presumptive exposures to weapons-grade materials (Meyer

1994); SM Building results were assumed to represent potential exposures to heat-source material.

In general for the dose reconstruction project, when an exposure to weapons-grade plutonium is 

presumed, the chance of exposure to highly insoluble plutonium is a possibility that cannot be 

ignored, and the Mound site is no different in this respect. Dose reconstructions for Mound should 

evaluate the potential for exposure to type super S plutonium in accordance with ORAUT-OTIB-0049, 

Estimating Doses for Plutonium Strongly Retained in the Lung (ORAUT 2010). As stated above, such 

exposures were likely to be uncommon at the Mound site because work with the 239Pu isotope was a 

small part of the work that was conducted. If presumptive exposures to weapons-grade material are 

assumed, the basis should be documented in the dose reconstruction report. 

5.8.3 Tritium

Meyer describes the initiation of tritium research as beginning in 1957, and bioassay records begin in 

that year. Information in EG&G (1995) implies that research began in 1947 in R Building, but that was

before construction of R or T Buildings and there are no specific records that suggest that the material 

was actually used, handled, or stored at Mound during this period. If claim data suggest tritium 

exposure (positive tritium doses in MESH), dose may be assigned before this time. Zero tritium doses 

in the MESH database for years before the beginning of tritium bioassay (before 1957) are artifacts 

from the recordkeeping system and are not to be assumed to be presumptive exposures for which 

missed dose is required to be assigned.

For 1957 onward the listed and zero doses may be used, with equation 5-2, to assign a dose only 

when dose overestimates are appropriate. If they are not appropriate, it is necessary for the dose 

reconstructor to review tritium logbooks for the employee’s name for the years in question and to 

assign dose from tritium bioassay, or missed dose based on these records instead.

Bioassay results that are reported at or below detection or as zeros should be corrected using the 

applicable MDAs in Table 5-16. Although a class of Mound employees has been added to the SEC to 

include 1972, 1975 and 1976, tritium logbook data are available for portions of the years 1972 and 

1976.

Table 5-16. Reported analytical procedures and detection limits for 

tritium.

Period Analytical method Detection limit (μCi/L)

1948–1956 No urinalysis conducted Use 95% dosea

1957–1961 Vibrating reed electrometer 1.0
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 40 of 60

1962–1995 Liquid scintillation 0.01

a. The upper 95 percentile dose of 340 mrem should be used to reconstruct doses.

Stable Metal Tritides

Work was performed with SMTs at Mound, as well as with elemental tritium and HTO. Due to the 

concern that urinalysis bioassay samples for tritium, together with the metabolic model that was used, 

are unlikely to generate accurate dose estimates for the less soluble tritides, an evaluation was made 

of other data sources. 

For reconstruction of doses, data from urinalysis bioassay are to be used for the small number of 

personnel performing actual “hands-on” work with SMT. These individuals can be identified by name 

and, in this case, doses should be reconstructed using the methodology described in ORAUT

(2007b). Dose reconstructors should contact the lead dose reconstructor for Mound when 

determining whether specific individuals worked with special tritium compounds.

For other workers, doses are likely to be small for all organs, as described in ORAUT-RPRT-0057, A 

Method for Estimating Stable Metal Tritide Exposures to Tritium Workers Based on Swipe Data in 

Rooms SW-9, SW-13, SW-150, and R-108 at the Mound Laboratory, 1968 to 1989 (ORAUT 2013b). 

This report describes a method for estimating doses to incidentally exposed workers. 

5.9 RECONSTRUCTING DOSE FROM SECONDARY RADIONUCLIDES BASED ON 

URINALYSIS DATA

5.9.1 Radium-Thorium-Actinium

For the actinium refinery project in the Old Cave, it has been determined that it is not possible to 

estimate dose or a maximum dose with sufficient accuracy. Therefore, these doses may not be 

reconstructed, and only partial dose reconstructions are possible, in accordance with the wording of 

the designation of the class listed in (Leavitt 2008).

5.9.2 Thorium

5.9.2.1 Thorium-232

Bioassay for 232Th was based on 24-hour urine samples and followed Mound bioassay procedures 

MLM-460 (Spoerl 1950a) and MD-20736 (Monsanto 1968). Mound conducted alpha counting of 

extracted thorium and radium in urine for monazite refinery workers. This was the gross alpha 

procedure, with specific analysis for radium and thorium. Meyer (1992) states that for these workers, 

bioassay results will always contain both thorium and radium results. It is assumed that 228Th, 232Th,

and 224Ra are in equilibrium in feed materials; however, an examination of the ratios of radium to 

thorium in these results shows that the proportions vary. To estimate doses from these results, dose 

reconstructors should: 

• Calculate dose from 224Ra in the mixture based on the radium result;

• Assume the thorium result is composed of 228Th and 232Th in equal proportions.

Meyer (1992, Section 1) indicates that 2.8 cpm on a 24-hour urine sample corresponded to an intake 

that would result in a 300-mrem/wk DE to the critical organ under the dosimetry system in effect when 

the sample was collected. Where no aliquot volume is given in the claimant data, assume that the 

dpm value has been corrected for background, detection efficiency, and represents a 24-hour urine 

excretion. If an aliquot volume is given, then assume that the bioassay result is in dpm per aliquot. 

No other corrections are required for urinalysis results in dpm. If urinalysis results are in cpm, then a 
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 41 of 60

correction for 50% detector efficiency and 92% extraction efficiency should be made to determine 

dpm results unless it is clear from primary records that these corrections have already been made.

Table 5-17 lists 232Th DLs by period. These DLs have units of dpm excreted per 24-hour when no 

aliquot size is given. However, DL units should be dpm per aliquot, if an aliquot volume was reported.

Table 5-17. Natural thorium DLs (dpm).

Period DL Reference

Before 12/1987 0.3 MDA (MJW 2002b)

a

12/1987–1994 0.02 (Th-232) MDA (Barton 1995)

01/1994–present 0.08 (Th-228) MDA (Barton 1995)

a. Correction for aliquot size should be made if available. Most 

database entries do not have reported aliquot volumes.

Thorium-232 bioassay results in Author unknown (1955–1988) contain results for radium, thorium, or 

both. If a radium result is provided for a 232Th urine bioassay, the primary record sample identification 

number usually ends in “R” and refers to 224Ra. From these results, the daily excretion of 224Ra can 

be obtained using Equation 5-9:

Daily Excretion Ra 224 Ra Database dpm - = ÷ 90% (5-9)

A primary record identification number that ends in “T” corresponds to a natural thorium result that 

includes 232Th and 228Th. The daily excretion of natural thorium can be calculated using 

Equation 5-10:

Daily Excretion Th Th Th Database dpm − +− = 228 232 ÷ 92% (5-10)

Equations 5-9 and 5-10 include correction factors of 90% and 92%, respectively, for radiochemical 

yield because database results for radium and thorium have not been corrected for chemical 

extraction.

Local background thorium urinary excretion rate data is not available. However, ICRP Publication 23 

(ICRP 1975) reports that background excretion rates for natural thorium range from 0.05 to 1.0 dpm/d, 

and that Reference Man excretes 0.05 dpm/d. For dose reconstruction, assume that background 

natural thorium in the diet of monitored workers resulted in urinary excretion of 0.05 pCi/d (ICRP 

1975).

5.9.2.2 Thorium-230

Gross alpha counting of extracted thorium in urine started with sample 6507-1I in August 1954. Gross 

alpha counting of extracted radium in urine began with sample 6507-10I in September 1954 and 

continued until September 9, 1958. Dose reconstructors should assume:

• 232Th and 228Th from uranium feed material have minimal radiological impact to workers.

• The MDA for negative bioassay radium result for demonstrated exposure to thorium 

compounds only.

When reconstructing a particular dose history, careful attention must be paid to the reporting basis 

that Mound used because several different bioassay strategies were attempted for 230Th. If the 

reporting basis for an ionium bioassay result is ambiguous, assume that 230Th was extracted and 

counted. Excretion rates in dpm/d should be calculated using Equation 5-11 and entries from the 
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Author unknown (1955–1988) that contain information on volume, aliquot size, and dpm results, 

unless contradicted by another bioassay record:

- Sample mL Daily ExcretionTh 230 Database dpm Aliquot mL = × (5-11)

If no aliquot volume is given and no definitive primary record is available, then assume that the 

sample volume is equal to the aliquot volume. No other corrections should be made for urinalysis 

results in dpm. If urinalysis results are in cpm, then a correction for 50% detector efficiency and 90% 

extraction efficiency should be made to determine dpm results.

Table 5-18 lists available 230Th DLs. Where no aliquot size is given in Author unknown (1955–1988)

assume the DL is in dpm excreted per 24-hour sample.

Table 5-18. 230Th DLs (dpm).

a

Period DL Reference

Before 01/1994 0.3 MDA (MJW 2002b)

a

01/1994–present 0.17 MDA (Barton 1995)

a. DL recommended for all secondary radionuclides (MJW 2002b). 

Correction for aliquot size should be made if available.

5.9.3 Protactinium-231

The primary program for 231Pa was completed in 1959. For partial dose reconstructions, doses from 

potential intakes of 231Pa may be included only when records in the case show that bioassay was 

conducted.

Intakes to workers should be estimated only for exposures in HH, R, and SW Buildings. Activity data 

in dpm should only be used as taken from Author unknown (1955–1988) for the thorium fraction. 

MDAs in Table 5-19 should be used to reconstruct missing results at or below detection. An MDA of 

0.3 dpm should be assumed for all exposures that occurred from 1954 to August 17, 1955. Daily 

excretion should be estimated using Equation 5-12. If no aliquot size is listed, the default aliquot size 

is 120 mL.

Table 5-19. Protactinium DLs (dpm).

a

Period

Background

count rate (cpm)

DL

(dpm)

1954–08/17/1955 No bioassay 0.3b

08/17/1955–1956 0.3 0.3

1957–1958 0.2 0.02

1959–12/07/59 0.15 0.08

a. Correction for aliquot size should be made for the DL if available.

b. A DL of 0.3 dpm should be used when there are no bioassay data 

available.

Reported bioassay results in cph were generally corrected for a 78% extraction efficiency, 50% 

detector efficiency, and background counting rate. However, no corrections were made for aliquot 

volumes in primary records. An evaluation of the 231Pa in Author unknown (1955–1988) determined 

that some sample volumes were not reported and the modal aliquot was 120 mL. If aliquot or sample 

volumes are not reported, a modal aliquot of 120 mL and a Standard Man sample volume of 1,400 mL

should be assumed to correct reported results in terms of daily excretion. Protactinium-231 daily 

excretion should only be inferred from the thorium fraction dpm from Author unknown (1955–1988), as 

shown in Equation 5-12, using actual sample and aliquot volumes or by assuming a standard man 

sample volume of 1400 mL and a modal aliquot volume of 120 mL:
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 43 of 60

or 1400

- ×

or 120

Sample mL Daily Excretion Pa 231 Th Fraction Database dpm Aliquot mL = (5-12)

5.9.4 Uranium

The screening procedure for uranium is likely to be simply the gross alpha procedure through the 

cerium carrier step, and the results were assumed to be uranium based on the worker’s job 

assignment. Sequential separation would isolate uranium only, so the original results should be 

checked in the logbook if an exposure to uranium is suspected. Dose reconstruction assumptions are 

the same for both contingencies:

• All activity is attributable to 233U in absence of specific radionuclide data. 

• Use 234U for EU. For NU or depleted uranium (DU), fluorometric or gross alpha determination.

• Assume NU ratios for 238U, 234U, and 235U or use IMBA default ratios for NU and DU.

Table 5-20 lists uranium DLs. Equation 5-13 provides correction for aliquot volume, if necessary:

Table 5-20. Uranium DLs (dpm).

Period DL Reference

Early 1950s–1985 0.4 (NU) μg/L Author unknown 

undated aa

Before 05/1985 0.3 MDA (MJW 2002b)

b

05/1985–present 0.07 (U-234)

0.03 (U-235)

0.09 (U-238)

MDA (Barton 1995)

a. Use for fluorometric results if reported. Statistical basis for this 

MDA in urine is unknown.

b. Correction for aliquot size should be made for the DL if available.

1400 mL U Daily Excretion Database dpm Aliquot mL = × (5-13)

In addition to gross alpha and specific analysis for uranium, one information source lists an MDA of 

0.4 μg/L for uranium fluorometry (Author unknown undated a). However, no evidence could be found 

that this bioassay method was ever used at Mound. Dose reconstructors should, if they find records 

of fluorometry (using mass instead of activity units), assume the value above as the DL. The source 

lists this DL for NU, and a specific activity of 683 pCi/mg for NU should be used to convert the value to 

activity for use in the IMBA code.

5.9.5 Reactor Fuels Projects

Most projects involved physical testing of fuel forms and compounds, which did not involve the 

possibility of significant intakes. For activities that did involve potential intakes, bioassay records will

be present in claim information for internal dose monitoring for plutonium, uranium, or thorium. Dose 

may not be calculated for presumptive exposures without specific claim information.

5.9.6 Reactor Waste Project

Processing of milliliter aliquots of reactor waste in laboratory conditions is unlikely to have resulted in 

significant intakes of radioactive material. However, bioassay samples for uranium and plutonium 

may be included in claim files. Dose may not be calculated for presumptive exposures without 

specific claim information.
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5.9.7 Americium and Curium

Mound had the ability to selectively monitor for exposure to americium using the gross alpha and 

sequential separation protocols described above. If a specific exposure to americium is determined to 

be likely, as opposed to a potential intake of americium as a constituent of a plutonium mixture, it is 

important to keep in mind that a specific bioassay for americium might have been coded in 

PURECON as a plutonium result; an analysis of the bioassay log for that year should give insight into 

the actual result.

Americium present as part of a mixture of plutonium is included in the intake of 239Pu. For heat source 

plutonium the mixture is assumed to be composed of 100% 238Pu.

Americium-241 alpha sources were electroplated in Room R-120 between 1956 and 1962, and it was 

the major isotope of concern in that room. Exposure to 241Am in R-120 could also have occurred 

coincident with intakes of weapons-grade plutonium.

Bioassay Procedures and DLs

Dose reconstructors should assume that all data in the ORAU database are in units of pCi excreted 

per 24 hours and that no additional corrections are required. The DL of 0.3 recommended by MJW 

(2002b) is consistent with the DL of 0.2 for the period from the 1950s to 1989. The 0.3 DL should 

therefore be used as the estimate of the DL from the 1950s to July 1989 as more favorable to 

claimants. Table 5-21 lists DLs by period for 241Am and 244Cm.

Table 5-21. 241Am and 244Cm DLs (dpm excreted per 24 

hours).

Period DL (dpm/d) Reference

Early 1950s–1989 0.2 Author unknown 

undated aa

Before 07/1989 0.3 MDA (MJW 2002b)

b

07/1989–present 0.05 MDA (Barton 1995)

a. Statistical basis is unknown and therefore recommend using 

0.3 dpm/d as reported by MJW (2002).

b. Correction for aliquot size should be made for the DL if available.

Daily excretion should be determined using Equation 5-14.

Daily Excretion Am 241 Cm 244 Database pCi - or - = (5-14)

Americium-241 and Curium-244 Recommendations

Exposure to 241Am in heat source plutonium should be considered negligible. Intakes of 241Am from 

exposure to weapons-grade plutonium should be estimated using the percent activities in Table 5-12. 

Dose reconstructors should assume the bioassay data in the ORAU database are for personnel who 

were working with 241Am and 244Cm source materials (Author unknown 1955–1988).

All data in the ORAU database are in units of pCi excreted per 24 hours, and no additional corrections 

are required. The 0.3 DL should be used as the estimate of the DL from the 1950s to July 1989 as 

more favorable to claimants than a lower value reported anecdotally for this period.

Table 5-22. 241Am and 244Cm detection limits (dpm/24 

hours).

Period Detection limit Reference

Before 7/1989 0.3 MDA (MJW 2002)a

July 1989–present 0.05 MDA (Barton 1995)

a. Correction for aliquot size should be made for the DL if available.
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5.9.8 Rare Isotopes

Very few workers were involved in this program, which yielded small quantities of the rare isotopes of 

interest over the years of its operation. Internal dose monitoring was likely to reflect the progenitor 

radionuclides of 238Pu or 233U. For 238Pu-derived radionuclides, the standard assumption that the 

exposure was to 100% 238Pu (neglecting the presence of the constituent radionuclides resulting in 

lower doses) will overestimate dose from the rare radionuclide of interest. The same is true when the 

standard assumption that exposure to uranium mixtures were composed of 100% 234U. Dose may not 

be calculated for presumptive exposures without specific claim information.

5.9.9 Radon

NIOSH is unable to put an upper bound on the radon exposures in SW-19 until the point at which a 

ventilation system was installed in early 1980 (before March 5, 1980) to vent SW Building tunnel air. 

The inability to bound the concentration in SW-19 extends to the contiguous air spaces in the SW and 

R Buildings. The lack of sufficient information to bound the dose from radon and its progeny formed 

the bases for the SEC class definitions in SEC-00171 (Sebelius 2010) and SEC-00207 (NIOSH 

2012). The recommended WLM values in Table 5-23 should be used to reconstruct doses from radon 

exposures for the buildings and periods shown.

Table 5-23. 222Rn WLM assumptions for dose reconstruction.

Building conditions Period

Radon 

(pCi/L)

Radon 

(WL)

Radon 

(WLM/y)

SW-19 postproduction and after addition 

of ventilation

03/06/1980–1995 0.7–13.4 0.03 0.4

Building 21 1966–1995 17.5–125 0.85 10.0

Building 48 1980–1995 3.2 0.01 0.2

R Building Ra-Ac processing 1949–1955 2.1 0.009 0.1

All other buildings 1949–1995 0.9 0.004 0.05

Background–indoors 1949–1995 0.5 0.002 0.03

All WLMs should be assumed to be median values of lognormal distributions with a geometric 

standard deviation (GSD) of 3.0.

5.9.10 Summary of Correction Factors and Excretion Equations

Extraction efficiencies should be used to correct data that are reported in cpm or cph. 

Protactinium-231, 230Th, and uranium data have been corrected for chemical recovery in Author 

unknown (1955–1988). However, if any data for these radionuclides are reported in cpm or cph, 

extraction efficiencies of 78%, 90%, and 80% should be used for 231Pa, 230Th, and uranium, 

respectively. PORECON database entries for 210Po must first be normalized assuming 86% 

extraction efficiency and then corrected for a chemical recovery 10% before 1964 and 63% for 1964 

and after.

A summary of corrections for all radionuclides is provided below in Table 5-24. The primary data 

types are listed with corrections necessary to estimate intake. In addition, equations are provided to 

calculate daily excretion. Most radionuclides are in units of dpm and do not require detector or 

extraction efficiency corrections except the radium-actinium data in cph. The radium-actinium and 

241Am244Cm data do not require aliquot corrections. The uranium and 230Th data require aliquot 

corrections when an aliquot volume is available, and MDAs should also be corrected when necessary. 

If no aliquot volume is reported, MDAs should be assumed to be dpm/24-hour sample.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 46 of 60

Table 5-24. Summary of correction factors and excretion equations.

Nuclide Period

Primary

data type

Aliquot

correction

Detector

efficiencya

Extraction

efficiencyb

Po-210c

1944–1963 cpm or Bq/d no no 10

1964–1973 cpm or Bq/d no no 63

Pu-238

Pu-239

1957–1960 dpm no no 10

1961–1966 dpm no no no

1967–5/81 dpm no no no

06/1981–1995 dpm no no no

Pa-231

08/17/1955–

1956 dpm yes no no

1957–1958 dpm yes no no

1959–12/1959 dpm yes no no

Th-230

Before 01/1994 dpm yes no no 01/1994–1995

Th-232

Prior 12/87

12/87–1994 dpm yes no 92

1/94–1995

Uranium

1950s–1985

Prior 1985 dpm yes no no

5/85–1995

Am-241

Cm-244

1950–1989

Prior 7/89 pCi no no no

7/89–1995

a. Detector efficiencies of 50% should be used to correct any data in cpm or cph.

b. Assume that intake of Ra-223 and Ra-224 were in equilibrium with Ac-227 and Th-228, respectively.

c. Po-210 database values must be normalized for existing 86% extraction and then corrected to 10% and 

63%.

5.9.11 Summary of Analytical Methods and Detection Limits

This section presents information required to ensure proper evaluation of bioassay data as well as a 

means to assign missed dose. Table 5-25 summarizes radionuclide analytical methods and MDAs by 

period for urinalysis results for primary Mound radionuclides.

Daily Excretion Po 210 Po Database Bq d - = / 86% 10% (1944 to 1963) × ÷

Daily Excretion Po 210 Po Database Bq d - = / 86% 63% (1964 to 1973) × ÷

Daily Excretion Pu Pu Database dpm = × ÷ 85% 10% (1957 to 1960)

Daily Excretion Pu Pu Database dpm = (1961 to present)

- Sample mL Daily Excretion Pa 231 Thorium Database dpm

Aliquot mL = ×

- Sample mL Daily Excretion Th 230 Database dpm Aliquot mL = ×

Daily Excretion Ra 224 Ra Database dpm - = ÷ 90%

Daily Excretion Natural Thorium Th Database dp 92% = m ÷

1400 mL U Daily Excretion Database dpm Aliquot mL = ×

Daily Excretion Am 241 Cm 244 Database pCi - or - =
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 47 of 60

Table 5-25. Analytical methods and MDAs for primary radionuclides by period.

Nuclide Method/description Period MDA

H-3

(hot gas)

Vibrating reed electrometer 1957–1961 1.0 μCi/L

LS 1962–1995 0.01 μCi/L

Po-210

(Postum)

Spontaneous electrodeposition from raw 

urine, low-background proportional 

counting.

1944–1963 Spot samples

6.0a dpm

168 dpm/d

24-hr 

samples

1.3 

dpm/24 

hra

Spontaneous electrodeposition from 

acidified urine, low-background 

proportional counting.

1964–1973 Spot samples

0.72a dpm

20 dpm/d

24-hr 

samples

0.15 

dpm/24 

hra

Pu-238, 

Pu-239

Alkaline earth phosphate coprecipitation, 

Pu separation with cerium carrier gross 

alpha proportional counting.

Before 1957 (0.85) pCi/sampleb

1957 (0.94) pCi/sampleb

1958 (0.6) pCi/sampleb

1959 (0.5) pCi/sampleb

1960 (0.5) pCi/sampleb

Above with sample acidification. 1961–1966 0.05 pCi/sample

Above with anion exchange, 

electrodeposition, gross alpha 

proportional counting.

1967–05/1981 0.11 pCi/samplec

Above with alpha spectroscopy for Pu238 and Pu-239/240, Pu-242 tracer, 

blanks counted, DLs quantified.c

06–12/1981 Pu-238 Pu-239

(pCi/sample)

0.066 0.044

1982 0.063 0.042

1983 0.064 0.043

1984 0.066 0.045

1985 0.047 0.031

1986 0.042, 0.028

1987 0.022 0.011

1988 0.017 0.01

01–07/1989 0.019 0.008

08/1989–12/1991 0.018 0.008

a. Polonium-210 spot and 24-hour samples were first normalized assuming 85% extraction efficiency, then corrected 

 for a 10% chemical recovery from 1944-1963. A chemical extraction efficiency of 64% was assumed for

 1964-1973.

b. Plutonium DL’s from the period 1957-1960 were normalized assuming 85% extraction efficiency, then corrected for

 a 10% chemical recovery during electroplating. 

c. Plutonium MDA’s from MCP (2004). 

5.9.12 Use of Co-Worker Data for Internal Dose Reconstruction

For the most part, employees potentially exposed to intakes of radioactive material were monitored 

during their employment at the Mound site. Co-worker data are available to assign doses from 

assumed exposures to primary radionuclides at the Mound site, should case-specific information 

suggest an unmonitored exposure to potential intakes (ORAUT 2012a). 

5.10 ATTRIBUTIONS AND ANNOTATIONS

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

section of this document.
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 48 of 60

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for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

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Support, Cincinnati, Ohio, October 5.

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Human Services, Office of the Secretary, Washington, D.C., July 14.

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Sheehan, W. E., W. R. Wood, and H. W. Kirby, 1963, “Urinalysis for Metabolized Plutonium,” 

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GLOSSARY

absorption type

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

curie

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

dosimetry

Measurement and calculation of internal and external radiation doses.

fission

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

in vitro bioassay

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 bioassay

Measurements of radioactive material in the human body utilizing instrumentation that detects 

radiation emitted from the radioactive material in the body.
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ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei.

radionuclide 

Radioactive nuclide. See radioactivity and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.
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ATTACHMENT A

CODING INSTRUCTIONS FOR URINALYSIS RECORDS

The following coding instructions are summarized for urinalysis records that were reported on 

Urinalysis Cards, Weekly Reports, and 24-hour reports at Mound.

Table A-1. Coding instructions for urinalysis 

records.

Code Description

All dates MM/DD/YY or MM/DD/YYYY

Am Am-241 results

CC Container believed to be contaminated

Cm Cm-244 results

CSU Contamination source unknown

DBS Destroyed by survey

D/M dpm

DM-24 dpm/24 hours

H Highest result

H-24 Highest 24-hour sample result

HS Home sample

I, Io Ionium (Th-230) results

INS Insufficient

LATE Arrived too late to process

LOST Lost in processing

NO No sample

NS New sample

OS Old sample

R Radium fraction results

R-24 Recount 24-hour sample

S&W Soap and water

SPIL Spills

T Natural thorium results

U Uranium results

V/I Vacation, illness, out of town

24 HR 24-hour sample

1AS First aid sample

? ? After cpm entry
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 57 of 60

TABLE OF CONTENTS

SECTION PAGE

B.1 Contamination of Samples ..................................................................................................... 57

B.2 Assignment of Gross Alpha Results as Plutonium.................................................................. 57

B.3 Missing Bioassay Records ..................................................................................................... 57

B.4 Assumption of Equilibrium...................................................................................................... 58

B.5 Particle Size Distributions....................................................................................................... 58

B.6 Solubility and Lung Clearance Type....................................................................................... 58

B.7 Radionuclide Use by Location................................................................................................ 59

B.8 Analytical Procedures ............................................................................................................ 59

Uncertainties in estimating worker dose commitments from bioassay data result from uncertainty in 

monitoring measurements and assessment of intake (ICRP 1998). Uncertainties for Mound bioassay 

measurements were not stated in the records.

B.1 CONTAMINATION OF SAMPLES

Cross contamination of samples was not specifically discussed in the historical information, and 

verifiable quality control laboratory procedures were implemented at Mound. This possible source is 

therefore considered a minor contributor to overall uncertainty.

B.2 ASSIGNMENT OF GROSS ALPHA RESULTS AS PLUTONIUM

Before alpha spectrometry, gross alpha activity based on chemical extractions was administratively 

assigned to 238Pu or 239Pu based on the employee workplace. If the gross alpha result for bioassay 

urine samples was due to any other alpha emitter, then worker dose could have been inaccurately 

estimated and improperly allocated among the organs.

B.3 MISSING BIOASSAY RECORDS

The record of urinalysis data is generally complete for the Mound site. In addition, some data are 

available for the Monsanto Chemical Company Dayton Laboratory facilities before the beginning of 

operations at the Mound Laboratory. Reportable worker exposures in the early years could also have 

been undetected due to low recovery that was caused by plating on sample containers before sample 

analysis. Intakes for missing bioassay results should be estimated using the MDAs in Table 5-9 for 

spot and 24-hour samples.

Tritium bioassay data were not reported from 1948 to 1956. Dose reconstructors have been advised 

to use a value of 1,000 mrem corrected to 340 mrem ICRP Publication 68 dosimetry (ICRP 1995) for 

unmonitored workers. This value was assumed to be the approximate upper 95th-percentile dose 

ATTACHMENT B

UNCERTAINTY ASSESSMENT

Page 1 of 4
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estimate and considered favorable to claimants for all categories of radiation workers. Uncertainty of 

extrapolating doses from the 1959- to-1962 period to the 1947-to-1956 period are considered minor 

and compensated by use of the upper 95th-percentile dose estimate. Doses should not be assigned 

to administrative or other nonradiological workers.

B.4 ASSUMPTION OF EQUILIBRIUM

Dose uncertainty in relation to degree of equilibrium is potentially important when assigning dose from 

bioassay for secondary radionuclides. For example, there are indications that 230Th doses were 

assigned based on the presumption of equilibrium of 230Th with 226Ra in some instances. This could 

result in errors of up to a factor of 10 when acute intakes of unspecified feed materials such as 

raffinate, tailings, and various production residues occurred. For instance, raffinate from sulfuric acid 

leaching of ore would have a relatively low concentration of radium in comparison with thorium due to 

radium sulfate precipitation. Therefore, the inferred 230Th intake could be greatly underestimated. 

The assumption of equilibrium is therefore largely dependent on the true nature of feed materials to 

which workers were exposed and the actual degree of observed equilibrium. In many instances, the 

origin and radiological and chemical composition of feed materials was unknown.

Errors associated with the assumption of equilibrium between 228Th and 224Ra, and between 227Ac and 

223Ra, are less than the 230Th case because (1) the 238U decay chain and possible cases of 

disequilibria are not involved, and (2) both 223Ra and 224Ra have short half-lives in days and therefore 

quickly achieve equilibrium with their respective parent radionuclides. 

B.5 PARTICLE SIZE DISTRIBUTIONS

Microsphere processing, plutonium metal production, and plasma torch operations at Mound all 

produced particle distributions of plutonium that potentially included submicrometer sizes dependent 

on the specific processes. These processes were used in buildings, and reconstruction necessitates 

deviation from the default ICRP Publications 66 and 68 assumption of an AMAD of 5 μm for 

occupational exposure (ICRP 1994b, 1995). Alpha emitters with the larger 5-μm particles and lung 

clearance types S or M tend to have smaller inhalation effective dose coefficients than do smaller 

1-μm particles. Use of a smaller AMAD of 1 μm increases worker dose after inhalation of type M 

239Pu by a factor of 1.5.

Of the secondary radionuclides, the difference in DCFs for 1-μm versus 5-μm particles with clearance 

types S or M was greatest for 231Pa by a factor of 2. The opposite trend was seen for type F alpha 

emitters, where 5-μm particles of uranium or 227Ac have effective inhalation dose coefficients that are 

20% larger than corresponding dose coefficients for 1-μm particles. Therefore, uncertainty in dose 

coefficients in relation to particle size distribution is dependent on lung clearance type and is 

comparatively small based on data in ICRP Publication 68 (ICRP 1995). Uncertainties in dose to a 

specific organ would likely be somewhat greater than the uncertainty in the effective dose.

B.6 SOLUBILITY AND LUNG CLEARANCE TYPE

Uncertainties in ICRP Publication 68 dose coefficients due to lung clearance kinetics are much 

greater than uncertainties that relate to particle size (ICRP 1995). Solubility of a radionuclide is one of 

the most important parameters that affects internal dose (NIOSH 2002). Lung clearance types should 

therefore be assigned on a case-by-case basis if there is sufficient bioassay data for radionuclides to 

surmise lung clearance behavior for acute exposures. Assigning a radionuclide to type S when it is 

ATTACHMENT B

UNCERTAINTY ASSESSMENT

Page 2 of 4
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actually M, or the opposite, can lead to a potential factor of 3 error in the dose coefficient. Assigning a 

radionuclide to type F when it is actually M, or the reverse, can lead to similar error.

B.7 RADIONUCLIDE USE BY LOCATION

The uncertainty associated with radionuclide use by location was almost entirely documented by 

EG&G (1995) and was considered minor in comparison with other sources of uncertainty. 

Radionuclide use by location has been sufficiently documented for primary radionuclides (Table 5-1). 

There is little uncertainty associated with documentation for tritium, 210Po, and the plutonium isotopes 

and their processes by building. Protactinium-231 had limited use, and there is therefore less 

documentation judged to be sufficient to evaluate use by location and potential worker exposures.

Secondary radionuclide documentation was generally less extensive, and some building production 

and research processes are unknown or poorly described. Isotopic uranium data were limited in 

comparison with those for NU. Isotopic radium and thorium data were also limited. Almost no 

information was available for 244Cm and 237Np other than as a contaminant in the Savannah River 

Plant source material.

Information on the location and use of radioactive gases was primarily only available for 222Rn (radon) 

from 1979 to the present. Very limited information was available for 220Rn (thoron) and 219Rn 

(actinon), and the measurements were questionable due to excessive decay times before analysis. 

However, measurements in the underground tunnel in 1979 confirmed the presence of 222Rn at 

88,000 pCi/L, 220Rn at 28,000 pCi/L, and 219Rn at 640,000 pCi/L. These concentrations clearly 

indicate the potential for indoor worker exposure to these inert radioactive gases.

The lack of data for the presence of thoron and actinon inside buildings at Mound could cause a large 

degree of uncertainty in actual worker exposures. Radon, thoron, and actinon in particular could have 

caused substantial worker exposure inside buildings before implementation of proper ventilation 

controls in 1980. Lung dose commitments from actinon could have been significant due to a very 

short actinon half-life of less than 4.0 seconds and direct deposition of actinon progeny on respiratory 

tract tissues.

Uncertainty from lack of radioactive gas information could underestimate worker doses by a factor of 

at least 10. However, use of an assumed GSD of 3 compensates for this uncertainty.

B.8 ANALYTICAL PROCEDURES

Uncertainties associated with analytical procedures Mound used throughout the history of bioassay 

measurements are significant. Early analytical procedures resulted in large errors and MDAs in 

relation to required regulatory dose limits. Reductions in regulated body burdens for 210Po required 

modifications to improve analytical sensitivities. Positive bioassay detections would therefore 

correspond to significant worker dose commitments at the lower end of detection in urine for some 

radionuclides.

Systematic errors occurred in analytical procedures in the early years due primarily to ineffective 

plating techniques and low recoveries. Low recoveries were due largely to colloidal plating of 

metabolized 238,239Pu and 210Po from raw urine onto sample containers. Polonium-210 recovery in the 

early years before sample acidification could have resulted in an order of magnitude correction for 

ATTACHMENT B

UNCERTAINTY ASSESSMENT

Page 3 of 4
Document No. ORAUT-TKBS-0016-5 Revision No. 02 Effective Date: 04/18/2013 Page 60 of 60

210Po urinalysis results. Use of an alkaline earth phosphate coprecipitation and separation of 

plutonium with a cerium carrier appeared effective throughout the history of bioassay analyses.

Additional errors occurred in the early years as a result of the inability to accurately assign gross 

alpha activity determined with low-background proportional counting to specific alpha-emitting 

radionuclides. Gross alpha activity was therefore assigned to particular radionuclides based on 

known work locations. This situation was resolved at Mound and within the DOE complex with the 

advent of alpha spectrometry and anion exchange capabilities. Accuracy in early years was also 

affected by the lack of tracers to determine recoveries. In addition, historical documents indicate that 

blanks were not counted before 1978, which precluded rigorous quantification of DLs.

ATTACHMENT B

UNCERTAINTY ASSESSMENT

Page 4 of 4