Page 1 of 59

DOE Review Release 03/10/2017

Mound Site – Occupational Internal Dose

ORAUT-TKBS-0016-5 Rev. 03

Effective Date: 03/03/2017

Supersedes: Revision 02

Subject Expert(s): Donald N. Stewart

Document Owner 

Approval: Signature on File

Donald N. Stewart, Document Owner

Approval Date: 03/02/2017

Concurrence: Wade C. Morris Signature on File for

John M. Byrne, Objective 1 Manager

Concurrence Date: 03/01/2017

Concurrence: Signature on File

Joel Arana, Objective 3 Representative

Concurrence Date: 03/01/2017

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 03/01/2017

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 03/03/2017

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 2 of 59

 

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. 

03/01/2017 03 Revision initiated to incorporate updated guidance in DCAS-RPRT005, Rev 00, Alternative Dissolution Models for Insoluble Pu-238, on 

highly insoluble 238Pu compounds. Minimum detectable amounts 

updated to reflect the latest available values for the period before 

Mound closure 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. 03 Effective Date: 03/03/2017 Page 3 of 59

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 Special Exposure Cohort.............................................................................................. 9

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

5.2 Process Operations Involving Primary Radionuclides ............................................................ 10

5.2.1 Polonium-210 ............................................................................................................. 10

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

5.2.2.1 Plutonium-238.............................................................................................. 11

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

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

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

5.3.1 Radium and Actinium ................................................................................................. 13

5.3.2 Thorium ...................................................................................................................... 13

5.3.2.1 Thorium-232................................................................................................. 14

5.3.2.2 Thorium-230................................................................................................. 14

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 ..................................................................... 22

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

5.4.5 Personnel Air Sampling Data...................................................................................... 23

5.5 Bioassay Methods for Primary Radionuclides ........................................................................ 23

5.5.1 Polonium-210 ............................................................................................................. 23

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

5.5.1.2 Polonium Internal Dose Monitoring Records ................................................ 25

5.5.2 Plutonium ................................................................................................................... 26

5.5.3 Tritium ........................................................................................................................ 27

5.6 Secondary Radionuclides Internal Dose Monitoring Programs............................................... 29

5.6.1 Radium-226, Actinium-227, and Thorium-228 ............................................................ 29

5.6.2 Thorium-230 and -232 ................................................................................................ 30

5.6.2.1 Thorium-230................................................................................................. 30

5.6.2.2 Thorium-232................................................................................................. 30

5.6.3 Protactinium-231 ........................................................................................................ 31
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 4 of 59

5.6.4 Uranium...................................................................................................................... 31

5.6.5 Americium-241 and Curium-244................................................................................. 32

5.6.6 Rare Isotopes ............................................................................................................. 32

5.7 Internal Dose Reconstruction Guidance................................................................................. 32

5.7.1 General Considerations.............................................................................................. 32

5.7.1.1 Use of the Gross Alpha Method ................................................................... 32

5.7.1.2 Internal Dose Estimation Parameters........................................................... 32

5.7.2 Equilibrium Assumptions ............................................................................................ 33

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

5.8.1 Polonium-210 ............................................................................................................. 35

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

5.8.2.1 Use of the Mound Alternate Dissolution Model for Plutonium-238................ 37

5.8.2.2 Special Considerations for Type SS Plutonium-239 ..................................... 38

5.8.3 Tritium ........................................................................................................................ 38

5.9 Reconstructing Dose from Secondary Radionuclides Based on Urinalysis Data .................... 39

5.9.1 Radium-226, Actinium-227 ......................................................................................... 39

5.9.2 Thorium-230 and -232 ................................................................................................ 39

5.9.2.1 Thorium-230................................................................................................. 39

5.9.2.2 Thorium-232................................................................................................. 40

5.9.3 Protactinium-231 ........................................................................................................ 41

5.9.4 Uranium...................................................................................................................... 42

5.9.5 Reactor Fuels Projects ............................................................................................... 42

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

5.9.7 Americium and Curium ............................................................................................... 43

5.9.8 Rare Isotopes ............................................................................................................. 44

5.9.9 Radon......................................................................................................................... 44

5.9.10 Summary of Correction Factors and Excretion Equations........................................... 44

5.10 Attributions and Annotations .................................................................................................. 46

References ........................................................................................................................................ 47

Glossary ............................................................................................................................................ 53

ATTACHMENT A CODING INSTRUCTIONS FOR URINALYSIS RECORDS............................. 55

ATTACHMENT B UNCERTAINTY ASSESSMENT ..................................................................... 56
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 5 of 59

LIST OF TABLES

TABLE TITLE PAGE

5-1 Radon isotope survey data gas concentrations and exposure levels by building and 

period.....................................................................................................................................18

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

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

5-4 In vivo Counting Systems and MDAs ..................................................................................... 22

5-5 Summary of historical plutonium analytical procedures .......................................................... 27

5-6 Reported analytical procedures and MDAs for tritium............................................................. 28

5-7 Recommended equilibrium assumptions for secondary radionuclides.................................... 33

5-8 Retroactive MDAs for fecal samples ...................................................................................... 34

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

5-10 Heat source plutonium feed material percent composition and percent activity by 

effective age...........................................................................................................................36

5-11 Weapons-grade plutonium feed material weight percent composition and percent activity 

by effective age......................................................................................................................36

5-12 Plutonium MDAs ....................................................................................................................37

5-13 Parameters of the Mound alternate dissolution model............................................................ 38

5-14 Reported analytical procedures and MDAs for tritium............................................................. 39

5-15 Th-230 MDAs.........................................................................................................................40

5-16 Natural thorium MDAs............................................................................................................41

5-17 Protactinium MDAs ................................................................................................................41

5-18 Uranium MDAs.......................................................................................................................42

5-19 Am-241 and Cm-244 MDAs ................................................................................................... 43

5-20 Rn-222 WLM assumptions for dose reconstruction ................................................................ 44

5-21 Summary of correction factors and excretion equations ......................................................... 45

A-1 Coding instructions for urinalysis records ............................................................................... 55
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 6 of 59

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

ANPP Aircraft Nuclear Propulsion Project

AWE atomic weapons employer

Bq becquerel

BZ breathing zone

CFR Code of Federal Regulations

Ci curie

cm centimeter

cph counts per hour

cpm counts per minute

d day

DCF dose conversion factor

DL detection limit

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EDE effective dose equivalent

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

F fast (absorption type)

g gram

gal gallon

GSD geometric standard deviation

hr hour

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

kg kilogram

L liter

LANL Los Alamos National Laboratory

LSC liquid scintillation counting

M moderate (absorption type)

mCi millicurie

MDA minimum detectable activity

min minute

mg milligram

mL milliliter

mo month
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 7 of 59

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

PuCd plutonium-cadmium (alloy)

PuCu plutonium-curium (alloy)

PuFe plutonium-iron (alloy)

PuNi plutonium-nickel (alloy)

PuVa plutonium-vanadium (alloy)

PuYt plutonium-yttrium (alloy)

QF quality factor

RTG radioisotopic thermoelectric generator

S slow (absorption type)

SEC Special Exposure Cohort

SMT stable metal tritide

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

SS super slow (absorption type)

TBD technical basis document

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
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 8 of 59

 

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 Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean nor should it be equated to an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined 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 [DOE] (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

of Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and the types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

NIOSH does not consider the following types of exposure as those incurred in the performance of 

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (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
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 9 of 59

 

5.1.1 Purpose

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

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

the end of operations at 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 in the 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 Special Exposure Cohort

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 (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 (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 (Sebelius 2012).

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, partial dose reconstructions for

employees who worked in any Mound area during the above 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 employees who worked in SW (Semi-Works) or R (Research) Buildings (as 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 10 of 59

 

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 class, may be performed using these 

data as appropriate.

To summarize the SEC classes: 

• 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 for 

October 1, 1949, through February 28, 1959; 

• The second class was added because it is not possible to bound radon doses to workers in 

SW and R Buildings from March 1, 1959, through March 5, 1980. These workers are identified 

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

• The third class was added because tritium records for some periods could not be found. This 

includes all employees in all locations from September 1 through December 31, 1972, and 

from January 1, 1975, through December 31, 1976. 

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

available, when those doses 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 extent of potential 

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 (Section 5.2): 

• Polonium (210Po),

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

• Tritium (3

H).

Secondary radionuclides of concern include (Section 5.3): 

• Radium (226Ra) and actinium (227Ac); 

• Thorium (230Th, and 232Th); 

• Protactinium (231Pa); 

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

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

• Rare isotopes (221Fr, 217

As, 229Th, 233Th, 233Pa, and 209Po)

• Radium (226Ra, 224Ra, and 223Ra); and

• Radon (222Rn, 219Rn, and 220Rn).

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 concurrent with the weapons program. Mound’s 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 11 of 59

 

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 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 R Building involved 

research and electrodeposition studies from 1948 to 1951 and manufacture of Po-Be neutron source 

from 1956 to 1961. 

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 

WD (Waste Disposal) and WDA Buildings from 1949 to the present. Chemical forms included 

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

Processing) or SM (Special Metallurgical) Buildings. One 210Po incident was reported in SW Building 

in 1961, and 11 incidents are associated with T Building where primary polonium separation and 

processing took place from 1954 to 1968. 

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

radioisotopic thermoelectric generators (RTGs). The Systems for Nuclear Auxiliary Power-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 239Pu and 238Pu for research projects were transferred from the Lawrence Livermore 

National Laboratory to Mound in 1959. As polonium production tapered 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 at Mound 

used 239Pu mixtures that are referred to in this document as “weapons-grade plutonium.” The 

plutonium for RTGs and other heat sources consisted primarily of 238Pu, which is typically called “heatsource” plutonium in Mound historical documents and in this TBD. Tables 5-10 and 5-11 later in this 

document list the radionuclides and proportions of these mixtures.

5.2.2.1 Plutonium-238

Feed material from the Savannah River Site (SRS) as PuO2 powder was enriched in 238Pu and 

processed at Mound to manufacture heat sources. Early 238Pu activities took place primarily in 

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

SM Building. 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 12 of 59

 

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 

have been affected by depositional processes that removed particles above 10 μm from worker 

breathing zones (BZs). Deposition would therefore have tended 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 PP Building in December 1967. The primary mission in this 

building was to process PuO2 powder from the SRS. A number of processes were developed to 

manufacture plutonium microspheres in 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 in yttrium,

• Solid solution process from 1969 to 1974,

• Solution gel process from 1969 to 1971, and

• Microsphere process to produce 50- to 250-μm 238

PuO2 ceramic spheres from 1967 to 1970.

The Mound site supplied RTGs for space exploration missions including Pioneer, Voyager, and 

Apollo. In 1977, production of the PuO2 fuel itself was moved to the SRS, 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. Table 5-10 later in this document lists the heat source plutonium isotopic mixture (EG&G 

(1995).

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 R Building using 239Pu in 1956. To date, NIOSH has not determined the 

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

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

reconstruction, the mixture that results in the greatest ingrowth of 241Am was selected as favorable to

claimants in lieu of specific data. Table 5-11 lists this mixture based on Hanford plutonium 

(BMI 2003).
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 13 of 59

 

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” [sic1

] (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, this document 

contains no provisions 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 PP Building and 

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 (EG&G 1995).

5.3 PROCESS OPERATIONS INVOLVING SECONDARY RADIONUCLIDES

5.3.1 Radium 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 resulted 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 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 from the feed material and the process itself. Bioassay was performed for the 

involved individuals, who are identified in Mound records (Meyer 1992), but the 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 

 

1 T Building was not occupied before February 1948; polonium operations began in September of that year (BWXT 2002). 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 14 of 59

 

potential 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 Thorium-232

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 activities with 232Th began with a preliminary proposal for a thorium refinery 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-gal 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 by AEC as early as April 1955 

(McCarthy 1955b), and the Mound refinery project was cancelled in May 1955 (McCarthy, 1955c).

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

5.3.2.2 Thorium-230

A limited ionium (

230Th) 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 came from Mallinckrodt 

Chemical Works and the Cotter Corporation.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 15 of 59

 

5.3.3 Protactinium-231

An AEC program was undertaken in 1955 in 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 HH Building (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 ended 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 and WDA 

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

SM, SW, T, or PP Buildings.

Extraction and purification of 231Pa 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 10 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 1952, 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 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.

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 PuFe, 

PuNi, PuCu, PuCd, PuYt, and PuVa.

• Plutonium-bearing glass fibers. In 1961 Mound participated in a project to experiment with 

fabrication of these fibers as a reactor fuel system.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 16 of 59

 

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 gal 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), but 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 who worked with 

americium in research projects. This is due to the possibility that it posed a specific intake risk from 

231Pa separation, the reactor fuel research program, and alpha source production, neutron source 

production, analysis, and vaporization of plutonium compounds. 

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.

• 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 

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.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 17 of 59

 

• 228Th, 227Ac, 209Po, and 226Ra separation began in the mid-1950s and continued until 1985 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 dose reconstruction, external dose is the exposure of concern; 

internal dose is considered negligible.

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

material from SRS. They were also present in aqueous wastes 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). Therefore, the potential to induce 

relatively high doses via the inhalation pathway should be taken into account by dose reconstructors. 

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 very few buildings for limited periods, in relatively minor quantities in comparison with 

the primary and secondary radionuclides. Most of these radionuclides were in sealed sources. 

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

Exposure to these radionuclides is therefore 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, which deposit 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, SW Building had high levels of radon from the 

decay of radioactive materials from 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 222Rn, 

actinon (219Rn), and thoron (220Rn), and an attempt was made to measure the concentration. 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 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 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 Sebelius (2010) and 

Sebelius (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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 18 of 59

 

Mound. For comparison, the reported background concentration of 222Rn ranged from 0.1 to 2.1 pCi/L 

with an average of 0.5 pCi/L.

Table 5-1. Radon isotope survey data gas concentrations and exposure levels by building and period. 

Building

Survey 

dates

Rn-222 

(pCi/L)

Thoron

(Rn-220) 

(pCi/L)

Actinon

(Rn-219) 

(pCi/L) WL WLM/yra

SW Tunnelb 10/12/1979 88,000 28,000 640,000 Not applicable Not applicable

SW19 after vent 03/1980–04/1980 7.7–13.4 No data No data 0.03 0.4

SW19 after vent 12/1989 2.6 Not detected No data No data No data

SW19 after vent 12/1999 1.0–5c No data No data No data No data

SW19 after vent 04/1990 0.7–7.2c No data No data No data No data

SW19 after vent 1998–2000 5.8–12 No data No data No data No data

Building 21 05/1983 17.5–52.8 No data No data 0.4–1.3

0.85 median

4.8–15.6

10.0 median

Building 21 12/1989 34–118c No data No data No data No data

Building 21 04/1990–06/1990 117–161 No data 0.3–0.7 No data No data

125c No data No data No data No data

Building 48 12/1989 3.2 No data No data No data No data

Old SD 12/1989 1.5–2.4 No data No data No data No data

Fire station 12/1989 1.2–1.4 No data No data No data No data

Paint shop 12/1989 0.9–1.2 No data No data No data No data

A 10/1988–02/1989 0.1–1.1c No data No data No data No data

All other buildings 12/1989 0.4–0.9 No data No data No data No data

Outside facilities 05/1983 0.24–0.59 No data No data No data No data

Reported 

background

Not applicable 0.1 to 2.1

Average = 0.5

Not applicable Not applicable Not applicable Not applicable

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

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

c. Concentrations were derived from charcoal canister measurements.

Areas in SW Building had high concentrations of radon, thoron, and actinon 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 divided into and renamed Rooms 

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 (Monsanto 1983a, 1983b, ca. 1983). Radon that was 

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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 19 of 59

 

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. 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 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 are considered to be secondary information. Whole-body counting is not an efficient 

monitoring method for Mound’s primary radionuclides (210Po, 238Pu, and 3

H) or for many secondary 

radionuclides such as 239Pu, 231Pa, those of radium, and those of actinium. 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. 

In vitro monitoring

Bioassay program Period Action levelsa Monitored workers Frequency

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 07/1989–

present

0.023 pCi/sampleb Potentially exposed 

personnel

Daily

Cm-244 07/1989–

present

0.009 pCi/sampleb Potentially exposed 

personnel

Daily

In vivo monitoring

Bioassay program Period Action levelsa Monitored workers Frequency

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 before this date, dose reconstructors should base MDA assumptions on actual data or refer to 

the “Table of Other Radionuclides” (Author unknown 1955–1988).

c. Later MDAs at the Mound site ranged from 0.1 nCi for Am-241 to 130 nCi for Pu-239.

Routine bioassay monitoring included collection of spot and 24-hour urine samples depending on the 

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

Meyer (1992) and inferred from personnel sampling records. For programs that used spot sampling, 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 20 of 59

 

such as 210Po and 238Pu, individuals who were 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 5-2 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 National Bureau of Standards Handbook 52 

(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 methods. 

Primary radionuclides of concern

Nuclide Method/description Period and MDA

H-3 (hot gas) Vibrating reed electrometer 1957–1961, 1.0 μCi/L

H-3 (hot gas) Liquid scintillation counting (LSC) 1962–1995, 0.01 μCi/L

Po-210

(postum)

Spontaneous electrodeposition from raw 

urine, low-background proportional 

counting

1944–1963, Spot samples,6.0 dpm,

a

168 dpm/d

24-hr samples, 1.3 dpm/24 hra

Po-210 Spontaneous electrodeposition from 

acidified urine, low-background 

proportional counting.

1964–1973, Spot samples, 0.72 dpm,

a

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/sample,b

1957, (0.94) pCi/sample,

b

1958, (0.6) pCi/sample,

b

1959–1960, (0.5) pCi/sample,

b

Pu-238, 

Pu-239

Above with sample acidification 1961–1966, 0.05 pCi/sample

Pu-238, 

Pu-239

Above with anion exchange, 

electrodeposition, gross alpha proportional 

counting

1967–05/1981, 0.11 pCi/samplec
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 21 of 59

 

Nuclide Method/description Period and MDA

Pu-238, 

Pu-239

Above with alpha spectroscopy for Pu-238 

and Pu-239/240, Pu-242 tracer, blanks 

counted, detection limits (MDAs) quantified

06–12/1981, 0.066 pCi/sample Pu-238,

0.044 pCi/sample Pu-239,

1982, 0.063 pCi/sample Pu-238,

0.042 pCi/sample Pu-239, 

1983, 0.064 pCi/sample Pu-238,

0.043 pCi/sample Pu-239, 

1984, 0.066 pCi/sample Pu-238,

0.045 pCi/sample Pu-239, 

1985, 0.047 pCi/sample Pu-238,

0.031 pCi/sample Pu-239, 

1986, 0.042 pCi/sample Pu-238,

0.028 pCi/sample Pu-239, 

1987, 0.022 pCi/sample Pu-238,

0.011 pCi/sample Pu-239, 

1988, 0.017 pCi/sample Pu-238,

0.01 pCi/sample Pu-239, 

01–07/1989, 0.019 pCi/sample Pu-238,

0.008 pCi/sample Pu-239, 

08/1989–12/1991, 0.018 pCi/sample Pu-238,

0.008 pCi/sample Pu-239, 

1992–present, 0.018 pCi/sample Pu-238,

0.009 pCi/sample Pu-239

Secondary radionuclides of concern

Nuclide Method/description Period and MDA

Pa-231 Radium extraction and activity inferred 

from Th-227 or Ra-223 dpm assuming 

secular equilibrium

08/17/1955–1956, 0.3 dpm,

1957–1958, 0.02 dpm,

1959–12/07/1959 0.08 dpm

Th-230

(ionium)

Radium separation, intakes and 

excreted Ra-226 and Th-230 assumed 

to be in equilibrium.

Before 01/1994, 0.3 dpm

Th-230

(ionium)

Th-230 extraction, alpha spectroscopy. 01/1994–1995, 0.17 dpm

Th-natural Thorium separation, proportional 

counting.

Before 12/1987, 0.3 dpm,

12/1987–1994, 0.02 dpm Th-232,

Th-natural Thorium separation, alpha 

spectroscopy.

01/1994–1995, 0.08 dpm Th-228

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

Uranium Uranium separation, proportional 

counting.

Before 1985, 0.3 dpm

Uranium Uranium separation, alpha 

spectroscopy.

05/1985–1995, 0.07 dpm U-234,

0.03 dpm U-235, 

0.09 dpm U-238

Am-241, 

Cm-244

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

a. Polonium-210 spot and 24-hour samples were first normalized assuming 85% extraction efficiency, then corrected for a 

10% chemical recovery from 1944 to 1963. A chemical extraction efficiency of 64% was assumed for 1964 to 1973.

b. Plutonium MDAs from 1957 to 1960 were normalized assuming 85% extraction efficiency, then corrected for a 10% 

chemical recovery during electroplating.

c. Plutonium MDAs from MCP (2004).

In general, the MDAs in Table 5-3 should be used only when no value accompanies the result. MESH 

records, available starting in 1991, typically list the MDA for each result, which are to be used for 

results after the dates in the table. The Mound bioassay program collected samples as late as 2006.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 22 of 59

 

5.4.3 Gross Alpha and Sequential Separation

A procedure was developed at 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 consist 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, which was 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 (Meyer 1992, 

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. 

5.4.4 In Vivo Dose Monitoring

An effort to upgrade the Mound internal dose monitoring program with an in vivo counting facility 

began in 1964, and the “reasonably standard“ Mound Body Counting Facility was completed with a 

Packard Instrument Company steel room in 1969. Detectors installed at that time were two standard 

sodium iodide detectors, 4 inches in diameter (Tomlinson et al. 1974a). Once personnel counting 

began, problems were identified, and later in that year, the system became operational with a single 

phoswich detector (ORAUT 2007b); in December, a second phoswich detector was added, and the 

modified system became fully operational in January 1970 (ORAUT 2008). Another source states 

that the dual-phoswich counter was operational in May of 1970 (Tomlinson et al. 1974a). A switch 

was made to dual five-inch phoswich detectors in October of 1972 (Tomlinson et al. 1974a). During 

this period, in accordance with guidance in the National Bureau of Standards Handbook 80, the MDA 

was determined to be the 3-sigma value of the background count (Tomlinson et al. 1974b). Table 5-4 

summarizes the in vivo counting systems and associated MDAs for specified time periods. 

Table 5-4. In vivo Counting Systems and MDAs. 

Dates Detector Types MDAs

1969 Dual NaI(Tl) detectors, 10.2 cm 11 nCi Pu-238 (claimed)a

July 1969 Single Phoswich, 10.2 cm Not listed

January or May 1970 Dual Phoswich, 10.2 cm 5.1 nCi Pu-238 (claimed)a

October 1972 Dual Phoswich 12.7 cm 4 nCi Pu-238 (claimed)a

1990 Dual Phoswich 12.7 cm 69.6 to 73.7 nCi Pu-238, 0.18 nCi Am-241b

a. Tomlinson et al. (1974a). 

b. Tomlinson (1990). 

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 (Barton, 1995, pp 325 – 327). All monitored 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 23 of 59

 

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. 

Tomlinson et al. (1974b) discuss chest wall thickness measurements at the Mound site. 

Measurements were made using a Hewlett-Packard ultrasonic sounding instrument. Two methods for 

ultrasonic measurements were investigated, along with a chest prediction method based on various 

body measurements. Though the ultrasonic methods proved to be most accurate, Mound limited the 

measurements to counts to four to six per week to conserve time spent (Tomlinson et al. 1974b). In 

vivo count records are not kept in a central electronic database; each count is maintained as a hardcopy 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 (Tomlinson 1990). Some 

records observed of those available (extant logbooks from 1969–1993) report MDAs for each count; 

however many do not. An example of a lung count record from 1976 shows that measurementspecific MDAs were to be reported with the counting result based on ultrasound chest-thickness 

measurements (Tomlinson 1976). 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. 

Due to the very high MDAs, missed dose should be based on urine bioassay results for dose 

reconstruction. Chest-counting data would best be used to assess intake when counting was 

conducted as a follow-up for a known worker intake and records are available to the dose 

reconstructor. It is not expected, based on experience with Mound claims, that chest counts results 

will be commonly available. 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

Mound operations produced a complex array of particulates, aerosols, dusts, and fumes. 

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 is unlikely to accurately estimate actual worker 

inhalation intakes in most situations 

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.

5.5 BIOASSAY METHODS FOR PRIMARY RADIONUCLIDES

5.5.1 Polonium-210

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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 24 of 59

 

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

Wednesday, and Friday. Samples were initially treated without HNO3 digestion. Polonium-210 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

1950). 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 (Spoerl 1950; Meyer 1992). Spoerl (1950) 

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 (1950) 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 1950). 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 1989 study determined that significant recovery losses from raw undigested urine were observed for 

metabolized 210Po due to plating or formation of precipitates (Fellman et al. 1989; Meyer 1992). This 

study used urine samples from primates by investigators at the New York Medical Center and by Spitz 

and Robinson at Mound (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 1950). 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 through 1963, after 

which complete acid digestion of urine samples was required (Meyer 1992).

The 10% chemical recovery correction was applied to evaluate occupational doses from 210Po in a 

retrospective study in 1989 (Spitz 1989a; Themelis 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 method. 

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 1950). 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 (1950) of 86%.

However, the data in 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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 25 of 59

 

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 a 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 (1950) 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 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 should be performed using the assumptions in Section 5.8.1, which are favorable 

to claimants. 

Reported Half-Life

The effective half-life of 210Po was investigated in 1950 and reported by Spoerl (1951) 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 per 24-hour sample. Weekly results 

were transcribed onto Polonium Cards for individual workers with entries in cpm/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 additional urine samples were 

required to document possible worker exposures or to confirm previous results with all entries in cpm 

and duplicates 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 evaluating individual 

dosimetry files from original logbooks (MJW 2002b). This database contains individual sample results 

from 1944 to 1973, which encompasses polonium operations at the Dayton Laboratory2 facilities as 

well as the Mound Laboratory.3

Most Mound claim files contain primary and secondary records for polonium bioassay. As a general 

procedure, conduct dose reconstruction using the primary records contained in the claimant files. The 

dose reconstructor may also use the databases as convenient listings of the claimant information but 

 

2 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 listed assumptions for the period before 1964. 

3 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. 03 Effective Date: 03/03/2017 Page 26 of 59

 

reconcile their use with the primary records. If variations from the general procedure occur, validate – 

to the extent practical, all information used in dose reconstruction calculations.

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 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 MDAs, 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 5-5 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. 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 27 of 59

 

Table 5-5. 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 

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

Plutonium Internal Dose Monitoring Records

Most Mound claim files contain primary and secondary records for plutonium bioassay. As a general 

procedure, conduct dose reconstruction using the primary records contained in the claimant files. The 

dose reconstructor may also use the databases as convenient listings of the claimant information but 

reconcile their use with the primary records. If variations from the general procedure occur, validate – 

to the extent practical, all information used in dose reconstruction calculations.

The dose reconstructor also needs to ensure that PURECON reflects earlier measurements using the 

gross alpha method and that the gross alpha samples were taken for 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 an SRS 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 counting (LSC) began in 1961 followed by an approved procedure in 1968. This 

method 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 scintillation counters 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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 28 of 59

 

1988). Tritium results were always reported in units of activity per liter of urine. Table 5-6 lists the 

reported analytical procedures and MDAs for tritium.

Table 5-6. Reported analytical procedures and MDAs for tritium. 

Period Analytical method MDA (μCi/L)

1957–1961 Vibrating reed electrometer 1.0

1962–2006a LSC 0.01

a. No conclusive data have been identified supporting tritium MDAs past 1995, 

so this value is assumed through the end of Mound operations. 

Tritium Dose Calculations

This section presents a method 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 necessary due to an overall correction 

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

A dosimetry method from LANL was used for acute tritium dose calculations from 1957 through 1988 

(Lawrence 1957). This method assumed a maximum permissible body burden of 3.7 mCi, a relative 

biological effectiveness (RBE) 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 (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

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

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

Reported doses on 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 reported doses using Equation 5-2.

Current dosimetry for tritium uses a quality factor (QF) of 1.0 in comparison with 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 in the LANL calculations (ICRP Publications 67 and 68; ICRP 

1994a, 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 0 34

68 8 1 7

. t/ . t/

. t/

. e . e dt

ICRP68 . Ratio . LANL . . e dt

(5-2)

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

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

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

1994a). The ratio of 1.0 confirms that no dose corrections are necessary for doses from 1989 to 

1995.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 29 of 59

 

( ) 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 ICRP68 Ratio

ICRP30 e dt

∞ − × − ×

∞ − ×

× +×  = × =    

∫

∫ (5-3)

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. The MDAs in Table 5-6 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-226, Actinium-227, and Thorium-228

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 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 

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.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 30 of 59

 

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).

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-230 and -232

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 an EDE 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).

5.6.2.2 Thorium-232

Monitoring for 232Th began in 1955 to support thorium refining; preliminary research began in 

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.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 31 of 59

 

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 

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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 32 of 59

 

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. Only one 

document (Author unknown undated a, p. 2) lists a DL in mass units for natural uranium (NU). There 

is no evidence that this is a Mound document and, because the bioassay methods listed 90Sr and 

147

Pr, 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

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 by NIOSH (2016). Guidance has been developed to estimate doses to high-fired forms of 

238Pu, which can under some circumstances exhibit nonmonotonic clearance rates (NIOSH 2016). A 

Mound-specific alternative dissolution 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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 33 of 59

 

this model. The alternative dissolution is modeled using specific IMBA input parameters in Table 5-13

later in this document.

Exposure Dates

The assumptions described in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2014), 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 5-7 lists 

equilibrium assumptions that are favorable to claimants to infer primary radionuclide activity in urine 

based on detected alpha activity.

Table 5-7. 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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 34 of 59

 

after the first year. All pre-1946 210Po intake estimates (at Monsanto Chemical Company facilities) 

should therefore assume equilibrium with 210Pb.

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 2014). 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 (Barton 1995, p. 4).

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

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 claims for which dose 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 35 of 59

 

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 0 85 1 400 mL - 0 50 0 10

Database cpm . , Po 210 Daily Excretion . . Aliquot mL

= ×× (5-4) 

1964 onward 0 85 1 400 mL - 0 50 0 63

Database cpm . , Po 210 Daily Excretion . . Aliquot mL

= ×× (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. 

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

Period

Default 

sample 

volume (mL)

Detector 

efficiency

(%)

Recovery

efficiency

(%)

Reported

background

(cpm/24 hr)

Spot sample

MDA 

(dpm/50 mL)

24-hr 

sample MDA 

(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.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 36 of 59

 

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 are to 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 percent 

composition and percent activity by effective age.

a,b

Nuclide % mass

% activity 

at 1.0 yr

(pCi/g)

% activity 

at 3.2 yr

(pCi/g)

% activity 

at 10 yr

(pCi/g)

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 Am-241 activity is zero.

A review was performed of well-documented weapons-grade plutonium mixtures for DOE and its 

predecessor agencies. It was determined that the Hanford mixture was bounding with respect to the 

proportion of 241Am, and this mixture is to be used to estimate dose to Mound workers (BMI 2003). 

Table 5-11. Weapons-grade plutonium feed material weight percent composition and 

percent activity by effective age.

a,b

Nuclide wt%

% activity at 0 

yr (fresh) (Ci/g)

% activity at 

5 yr (Ci/g)

% activity at 

10 yr (Ci/g)

% activity at 

15 yr (Ci/g)

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: BMI (2003).

b. Effective age is the age when Am-241 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 1950). This 

correction is not required for later samples. These corrections are summarized in Equations 5-7 and 

5-8: 

1957–1960 0 85 - 0 10

Pu database dpm . Pu 238,239 Daily Excretion .× = (5-7) 

1960–2006 Pu-239,239 Daily Excretion=Pu databasedpm (5-8) 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 37 of 59

 

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 when the MDA is not listed with the 

result (MESH results, from 1991, typically list result-specific MDAs). 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,

tc = 1,000 min

Pu-238,

tc = 3,000 min

Pu-239/240,

tc = 1,000 min

Pu-239/240,

tc = 3,000 min

Before 1957 (0.85)b Not applicable (0.85)b Not applicable

1957 (0.94)b Not applicable (0.94)b Not applicable

1958 (0.6)b Not applicable (0.6)b Not applicable

1959 (0.5)b Not applicable (0.5)b Not applicable

1960 (0.5)b Not applicable (0.5)b Not applicable

1961–1966 0.05 Not applicable 0.05 Not applicable

1967–05/1981 0.11 Not applicable 0.11 Not applicable

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 Not applicable 0.011 Not applicable

1988 0.017 Not applicable 0.01 Not applicable

01–07/1989 0.019 Not applicable 0.008 Not applicable

08/1989–12/1991 0.018 Not applicable 0.008 Not applicable

1992–1995 0.018a Not applicable 0.009a Not applicable

a. tc = count time; plutonium MDAs from before 1957 to 1960 were 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. Based on Mound (1996).

b. MDAs in parentheses were corrected for a 10% chemical recovery.

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).

5.8.2.1 Use of the Mound Alternate Dissolution Model for Plutonium-238 

Workers who are exposed to airborne intakes of 238

Pu may exhibit urinary excretion patterns indicative 

of a variety of dissolution rates of plutonium in the lung. Guilmette et al. (1994) described the urinary 

excretion resulting from exposures to 238Pu at three facilities, including the Mound site, and the 

patterns were observed to be very different. This is of concern from a dose reconstruction point of 

view, as the standard dissolution model of ICRP Publication 66 (ICRP 1994b) may, in some 

situations, result in underestimates of dose. 

The complete dataset for Mound urine bioassay was reviewed, as described in Alternate Dissolution 

Models for Insoluble Pu-238 (NIOSH 2016). The review included results from 2,082 workers. Of 

these, records of nine workers were shown both to exhibit non-monotonic excretion rates 

characterized in Guilmette et al. (1994) and to be suitable for modeling. 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 38 of 59

 

A model was formulated for an alternate dissolution model for Mound, and specific parameters for use 

in IMBA are listed in Table 5-13. Mound site dose reconstructions assigning dose from 238Pu must 

evaluate this alternate model, the Mound highly-insoluble (MHI) plutonium-238 model, along with the 

standard absorption types (NIOSH 2016). 

Table 5-13. Parameters of the Mound alternate dissolution model 

(NIOSH 2016).

Parameter fr sr ss f1

Value -8.46818E-02 2.91697E-02 2.6468E-03 1E-08

The general steps to enter the alternate dissolution model into IMBA are listed below.

• Select 238Pu as indicator radionuclide.

• Select IRCP-68 Type S default parameters.

• Select “Absorption” button.

• Select “Alternative Representation” tab.

• Select “User Defined” button.

• Enter Table 5-13 values for fr, sr, ss, and f1.

5.8.2.2 Special Considerations for Type SS 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 

1992); 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 SS 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.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 39 of 59

 

Bioassay results that are reported at or below detection or as zeros should be corrected using the 

applicable MDAs in Table 5-14. 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 5-14. Reported analytical procedures and MDAs for tritium

(μCi/L). 

Period Analytical method MDA

1948–1956 No urinalysis conducted Use 95% dosea

1957–1961 Vibrating reed electrometer 1.0

1962–1995 Liquid scintillation 0.01

a. The upper 95th-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 method described in ORAUT (2007a). 

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 2013). 

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-226, Actinium-227 

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-230 and -232

5.9.2.1 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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 40 of 59

 

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 5-9 and entries from 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 = × (5-9) 

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 5-15 lists available 230Th MDAs. Where no aliquot size is given in Author unknown (1955–1988) 

assume the MDA is in dpm excreted per 24-hour sample.

Table 5-15. Th-230 MDAs (dpm).

a

Period MDA Reference

Before 01/1994 0.3 MDA (MJW 2002b)a

01/1994–present 0.17 MDA (Barton 1995)

a. MDA recommended for all secondary radionuclides (MJW 2002b). 

Correction for aliquot size should be made if available.

5.9.2.2 Thorium-232

Bioassay for 232Th was based on 24-hour urine samples and followed Mound bioassay procedures 

MLM-460 (Spoerl 1950) 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 always contain both thorium and radium results. It is assumed that 228Th, 232Th, and 

224Ra are in equilibrium in feed materials, but 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 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 EDE 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 and detection efficiency and represents a 24-hour 

urine excretion. If an aliquot volume is given, 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, a 

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 5-16 lists 232Th MDAs by period. These MDAs have units of dpm excreted per 24-hour when no 

aliquot size is given. However, MDA units should be dpm per aliquot, if an aliquot volume was 

reported.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 41 of 59

 

Table 5-16. Natural thorium MDAs (dpm). 

Period MDA 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 5-10: 

Daily Excretion Ra-224 RaDatabasedpm % = ÷ 90 (5-10) 

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 5-11: 

Daily ExcretionTh-228 Th-232 ThDatabasedpm % += ÷ 92 (5-11)

Equations 5-10 and 5-11 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 

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 (ICRP 1975). Dose reconstructors should assume that 

background natural thorium in the diet of monitored workers resulted in urinary excretion of 0.05 pCi/d 

(ICRP 1975).

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 5-17 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 5-17. Protactinium MDAs.a

Period

Background

count rate (cpm)

MDA

(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 MDA if available.

b. A MDA of 0.3 dpm should be used when there are no bioassay 

data available.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 42 of 59

 

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 data 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 1,400 mL and a modal aliquot volume of 120 mL:

or 1,400 mL - ×

or 120 mL

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 only uranium, 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), use fluorometric or gross alpha 

determination.

• Assume NU ratios for 238U, 234U, and 235U or use IMBA default ratios for NU and DU.

Table 5-18 lists uranium MDAs. Equation 5-13 provides correction for aliquot volume, if necessary:

Table 5-18. Uranium MDAs. 

Period MDA Reference

Before 05/1985 0.3 dpm MDA (MJW 2002b)b

05/1985–present 0.07 dpm U-234

0.03 dpm U-235

0.09 dpm U-238

MDA (Barton 1995)

a. Correction for aliquot size should be made for the MDA 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, p. 3). 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 IMBA.

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, claim information 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 43 of 59

 

includes bioassay records 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.

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 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 MDAs 

Dose reconstructors should assume that all data in the ORAU database are in pCi excreted per 

24 hours and that no additional corrections are required. The MDA of 0.3 recommended by MJW 

(2002b) is consistent with the MDA of 0.2 for the period from the 1950s to 1989. The 0.3 MDA should 

therefore be used as the estimate of the MDA from the 1950s to July 1989 as it is more favorable to 

claimants. Table 5-19 lists MDAs by period for 241Am and 244Cm.

Table 5-19. Am-241 and Cm-244 MDAs (dpm/24 hours). 

Period

MDA

(dpm/d) Reference

Before 07/1989 0.3 MDA (MJW 2002b)a,b

07/1989–present 0.05 MDA (Barton 1995)

a. Correction for aliquot size should be made for the MDA if available.

b. Statistical basis is unknown and therefore recommend using 0.3 dpm/d as 

reported by MJW (2002b). 

Daily excretion should be determined using Equation 5-14: 

Daily Excretion Am-241 or Cm-244 Database pCi = (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. Table 5-18 provides the MDAs for 241Am and 244Cm. The 0.3 MDA should be used as 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 44 of 59

 

the estimate of the MDA from the 1950s to July 1989 as more favorable to claimants than a lower 

value reported anecdotally for this period.

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 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) overestimates dose from the rare radionuclide of interest. The same is true under the 

standard assumption that uranium mixtures were 100% 234U. Dose may not be calculated for 

presumptive exposures without specific claim information.

5.9.9 Radon

NIOSH is unable to determine an upper bound for 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 Sebelius (2010) and Sebelius (2012). The recommended 

WLM values in Table 5-20 should be used to reconstruct doses from radon exposures for the 

buildings and periods shown.

Table 5-20. Rn-222 WLM assumptions for dose reconstruction. 

Building conditions Period

Radon 

(pCi/L)

Radon 

(WL)

Radon 

(WLM/y)

SW-19 postproduction and after addition 

of ventilation

03/06/1980–2006a 0.7–13.4 0.03 0.4

Building 21 1966–2006 17.5–125 0.85 10.0

Building 48 1980–2006 3.2 0.01 0.2

R Building Ra-Ac processing 1949–1955 2.1 0.009 0.1

All other buildings 1949–2006 0.9 0.004 0.05

Background–indoors 1949–2006 0.5 0.002 0.03

a. End-date figures are currently unavailable; the year 2006 is assumed for the purpose of the table, but should be used 

through the end of covered employment past this date. 

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 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 of 10% before 1964 and 63% for 1964 and after.

Table 5-21 summarizes corrections for all radionuclides. The primary data types are listed with 

corrections necessary to estimate intake. In addition, equations are provided to calculate daily 

excretion. In the records, most radionuclide results are in dpm and do not require detector or 

extraction efficiency corrections except the radium and actinium data in cph. The radium and actinium 

data and the 241Am244Cm data do not require aliquot corrections. The uranium and 230Th data require 
 

Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 

03/03/2017 Page 45 of 59

Table 5-21. Summary of correction factors and excretion equations. 

Nuclide Perioda

Primary

data type

Aliquot

correction

Detector

efficiencyb

Extraction

efficiencyc Equation

Po-210d 1944–1963 cpm Yes Yes 10% Daily Excretion Po-210 = Po cpm/0.5 x 0.85/0.1 x 1400 

mL/Aliquot mL

Po-210c 1964–1973 cpm Yes Yes 63% Daily Excretion Po-210 = Po cpm/0.5 x 0.85/0.1 x 1400 

mL/Aliquot mL

Pu-238\

Pu-239

1957–1960 dpm No No 10% Daily Excretion Pu = Pu Database dpm x 85% ÷ 10%

Pu-238\

Pu-239

1961–2006 dpm No No No Daily Excretion Pu = Pu Database dpm

Pa-231 08/17/1955–

12/1959

dpm Yes No No - = × Sample mL Daily Excretion Pa 231 Thorium Database dpm

Aliquot mL

Th-230 01/1994–2006 dpm Yes No No - Sample mL Daily Excretion Th 230 Database dpm Aliquot mL = ×

Th-232 Before 12/87-

2006

dpm Yes No 92% Daily Excretion Ra-224 = Ra Database dpm ÷ 90%

Daily Excretion Natural Thorium = Th Database dpm ÷ 92%

Uranium 1950s–2006 dpm Yes No No 1 400 mL , Daily Excretion U Database dpm Aliquot mL = ×

Am-241/

Cm-244

1950–2006 pCi No No No Daily Excretion Am-241or Daily Excretion Cm-244 = 

Database pCi

a. Ending dates set to an assumed date of 2006. 

b. Detector efficiencies of 50% should be used to correct data in cpm or cph

c. Assume intakes of Ra-223 and Ra-224 were in equilibrium with Ac-227 and Th-228, respectively.

d. Po-210 database values must be normalized for existing 86% extraction and then corrected to 10% and 63%.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 46 of 59

 

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.

For the most part, employees with the potential for intakes of radioactive material were monitored 

during their employment at the Mound site. Coworker data are available to assign doses from 

assumed exposures to primary radionuclides at the Mound site, if case-specific information suggests 

an unmonitored exposure to potential intakes (ORAUT 2012). 

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. 03 Effective Date: 03/03/2017 Page 47 of 59

REFERENCES

Anthony, D. S., J. E. Campbell, G. W. Charles, R. N. Cowden, G. C. Cox, R. K. Davis, R. R. Deem, R. 

W. Endebrock, P. M. Engle, L. B. Gnagey, J. M. Goode, P. M. Hamilton, W. L. Hood, F. M. 

Huddleson, E. F. Joy, F. C. Mead, Jr., H. V. Moyer, G. D. Nelson, J. L. Richmond, A. J. 

Rogers, C. E. Shoemaker, E. Spoerl, J. S. Stanton, J. W. Wright, and R. E. Zipf, 1956, 

Polonium, H. V. Moyer, editor, U.S. Atomic Energy Commission, Technical Information Service 

Extension, Oak Ridge, Tennessee, July. [SRDB Ref ID: 12703]

Author unknown, 1955–1988, “Internal Dose Data - Table of Other Radionuclides,” spreadsheet. 

[SRDB Ref ID: 8755]

Author unknown, undated a, “Appendix A. Detectable Amounts for Urinalysis.” [SRDB Ref ID: 8641, 

p. 2]

Barton, B. D., 1995, “Review of Historical MDAs,” memorandum to T. Vaughn et al., EG&G Mound 

Applied Technologies, Mound Facility, Miamisburg, Ohio, May 11. [SRDB Ref ID: 8641, p. 3]

BMI (Battelle Memorial Institute), 2003, Methods and Models of the Hanford Internal Dosimetry 

Program, PNNL-MA-860, Pacific Northwest National Laboratory, Richland, Washington, 

January 31. [SRDB Ref ID: 11521]

Bradley, J. E., 1954, “Removal of Drums from Quonset,” memorandum to E. C. McCarthy, Monsanto 

Chemical Company, Mound Laboratory, Miamisburg, Ohio, March 23. [SRDB Ref ID: 42049]

Butler, F. E., 1961, “Determination of Tritium in Water and Urine: Liquid Scintillation Counting and 

Rate-of-Drift Determination,” Analytical Chemistry, volume 33, number 3, pp. 409–414. [SRDB 

Ref ID: 36897]

BWXT (BWXT of Ohio), 2002, White Paper, T Building Structural History and Process History 

Summary Background Document, EC&AS Department, Mound Facility, Miamisburg, Ohio, 

November. [SRDB Ref ID: 25941]

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, Order DOE 

5480.11, Washington, D.C., December 21. [SRDB Ref ID: 8073]

DOE (U.S. Department of Energy), 1993, Operable Unit 9, Site Scoping Report: Volume 7 - Waste 

Management, Final (Rev. 0), EG&G MOUND-29-01-01-07-07-9502080001, Albuquerque Field 

Office, Albuquerque, New Mexico, February. [SRDB Ref ID: 11877].

DOE (U.S. Department of Energy), 1994, Primer on Tritium Safe Handling Practices, DOE-HDBK1079-94, Washington, D.C., December. [SRDB Ref ID: 22724]

Doody, K. R., 2000, “Radon Study Summary (REV. 1),” memorandum to R. Daily, BWXT of Ohio, 

Mound Facility, Miamisburg, Ohio, August 7. [SRDB Ref ID: 48445, p. 3]

EG&G (EG&G Mound Applied Technologies), 1995, Mound Site Radionuclides by Location, MD22153, Issue 1, Mound Facility, Miamisburg, Ohio, June 22. [SRDB Ref ID: 3240]

Fellman, A., L. Ralston, D. Hickman, L. Ayers, N. Cohen, H. Spitz, and B. Robinson, 1989, “The 

Importance of Acid Digestion of Urine Prior to Spontaneous Deposition of 210Po,” Health 

Physics, volume 57, number 4, pp. 615–621. [SRDB Ref ID: 36893]
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 48 of 59

FRC (Federal Radiation Council), 1960, Background Material for the Development of Radiation 

Protection Standards, Report 1, Washington, D.C., May 13. [SRDB Ref ID: 8731]

Guilmette, R. A., W. C. Griffith, and A. W. Hickman, 1994, “Intake Assessment for Workers Who Have 

Inhaled Pu-238 Aerosols,” Radiation Protection Dosimetry, volume 53, numbers 1–4, pp. 127–

131. [SRDB Ref ID: 120904]

HPS (Health Physics Society), 1985, An American National Standard, Performance Criteria for 

Radiobioassay, ANSI N13.30, American National Standards Institute, McLean, Virginia. 

[SRDB Ref ID: 135490]

Huddleston, F. M., 1955, “Thorium Refinery Waste Disposal Problem,” memorandum to F. C. Mead, 

Monsanto Chemical Company, Miamisburg, Ohio, April 13. [SRDB Ref ID: 42070]

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

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

[SRDB Ref ID: 11106]

ICRP (International Commission for Radiological Protection), 1975, Report of the Task Group on 

Reference Man, Publication 23, Pergamon Press, Oxford, England. [SRDB Ref ID: 22735]

ICRP (International Commission for Radiological Protection), 1979, Limits for Intakes of Radionuclides 

by Workers, Publication 30, Part 1, Pergamon Press, Oxford, England. [SRDB Ref ID: 22733]

ICRP (International Commission for Radiological Protection), 1994a, Age-Dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 2 Ingestion Dose Coefficients, 

Publication 67, Pergamon Press, Oxford, England. [SRDB Ref ID: 36901]

ICRP (International Commission for Radiological Protection), 1994b, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22732]

ICRP (International Commission for Radiological Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22731]

ICRP (International Commission for Radiological Protection), 1998, Individual Monitoring for Internal 

Exposure of Workers, Publication 78, Pergamon Press, Oxford, England. [SRDB Ref ID: 

132761]

James, B., 1997, “View on Responses to DOE Contractor DL/MDA Use,” memorandum to R. Weaver, 

Mound Facility, Miamisburg, Ohio, July 10. [SRDB Ref ID: 8641, p. 242]

Kirby, H. W., 1952, Tentative Procedure for the Analysis of Mixtures Containing Radium-226, 

Actinium-227, and Thorium-228, MLM-773 (Rev.), Monsanto Chemical Company, Mound 

Laboratory, Miamisburg, Ohio, October 15. [SRDB Ref ID: 8641, p. 55]

Kirby, H. W., 1959, The Radiochemistry of Protactinium, NAS-NS 3016, National Academy of 

Sciences, National Research Council, Washington, D.C., December. [SRDB Ref ID: 41602]

Kirby, H. W., and R. M. Brodbeck, 1954, Determination of Radium-226, Actinium-227, and Thorium228 in Human Urine (Ad Interim Report), MLM-1003, Monsanto Chemical Company, Mound 

Laboratory, Miamisburg, Ohio, August 20. [SRDB Ref ID: 8641, p. 34]
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 49 of 59

Kirby, H. W., and P. E. Figgins, 1964, Purification of Protactinium from Natural Sources, Monsanto 

Research Corporation, Mound Laboratory, Miamisburg, Ohio, December. [SRDB Ref ID: 

1962, p. 302]

Langham, W. H., 1964, “Physiological Properties of Plutonium and Assessment of Body Burdens in 

Man,” pp. 565–581, Assessment of Radioactivity in Man, Volume II, proceedings, International 

Atomic Energy Agency, Vienna, Austria. [SRDB Ref ID: 135825]

Lawrence, J. N. P., 1957, Estimation of Whole Body Dose (Rem) from Tritium in Body Water, LA2163, University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. 

[SRDB Ref ID: 959]

Leavitt, M. O. 2008, HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees from Mound Plant, Miamisburg, Ohio, U.S. Department of Health and 

Human Services, Office of the Secretary, Washington, D.C., March 3. [SRDB Ref ID: 80513]

MCP (Miamisburg Closure Project), 2004, Technical Manual, Internal Dosimetry Technical Basis 

Document, MD-10435, Issue 7, August 10. [SRDB Ref ID: 82022]

McCarthy, E. C., 1955a, “BTS from Middlesex,” MLM-CF-55-1-41, memorandum to J. E. Wiesler et 

al., Monsanto Chemical Company, Mound Laboratory, Miamisburg, Ohio, January 13. [SRDB 

Ref ID: 42042]

McCarthy, E. C., 1955b, untitled memorandum to N. N. T. Samaras, Monsanto Chemical Company, 

Mound Laboratory, Miamisburg, Ohio, April 7. [SRDB Ref ID: 42045]

McCarthy, E.C., 1955b, Mound Monthly Report for May, 1955, MLM-1051, Monsanto Chemical 

Company, Mound Laboratory, Miamisburg, Ohio, May 31, 1955. (SRDB Ref ID: 24645]

Meyer, H. E., 1979, Thorium-232 Program - Mound Facility, Mound Facility, Miamisburg, Ohio, 

October 3. [SRDB Ref ID: 42084]

Meyer, H. E., 1992, History of Mound Bioassay Programs, Volume 1, MLM-MV-93-93-0003, EG&G 

Mound Applied Technologies, Mound Facility, Miamisburg, Ohio, September. [SRDB Ref ID: 

1962]

MJW (MJW Corporation), 2002a, Pre-1989 Dose Assessment Project, Phase II Final Report, 

Williamsville, New York, June. [SRDB Ref ID: 8738]

MJW (MJW Corporation), 2002b, Pre-1989 Dose Assessment Project, Phase I Final Report, 

Volume I, Williamsville, New York, April. [SRDB Ref IDs: 8739, 8741, 8745]

Monsanto (Monsanto Chemical Company), 1954, Preliminary Proposal Thorium Sludge Processing 

Facility, MLM-CF-54-10-61, Mound Laboratory, Miamisburg, Ohio. [SRDB Ref ID: 42038]

Monsanto (Monsanto Research Corporation), 1968, Bioassay Group Analytical Procedures, Technical 

Manual MD-20736 Issue 1 (TM-20736), Mound Laboratory, Miamisburg, Ohio, November 19. 

[SRDB Ref ID: 4097]

Monsanto (Monsanto Research Corporation), 1983a, collection of PERM measurements inside and 

outdoors near Building 21, Mound Facility, Miamisburg, Ohio. [SRDB Ref ID: 8714]
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 50 of 59

Monsanto (Monsanto Research Corporation), 1983b, Building 21 Air Sample #B, Mound Facility, 

Miamisburg, Ohio, May 4. [SRDB Ref ID: 8716]

Monsanto (Monsanto Research Corporation), ca. 1983, Radon/Radon Progeny Assessment in the 

Thorium Storage Building 21, Mound Facility, Miamisburg, Ohio, . [SRDB Ref ID: 8699]

NBS (National Bureau of Standards), 1953, Maximum Permissible Amounts of Radioisotopes in the 

Human Body and Maximum Permissible Concentrations in Air and Water, Handbook 52, U.S. 

Department of Commerce, Washington, D.C., March 20. [SRDB Ref ID: 12332]

NCRP (National Council on Radiation Protection and Measurements), 1992, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Bethesda, Maryland. [SRDB Ref ID: 22736]

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

Implementation Guidelines, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 9.

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 01, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5.

NIOSH (National Institute for Occupational Safety and Health), 2016, Alternative Dissolution Models 

for Insoluble Pu-238, DCAS-RPT-005, Rev. 00, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, June 30.

ORAUT (Oak Ridge Associated Universities Team), 2007a, Calculation of Dose from Intakes of 

Special Tritium Compounds, ORAUT-OTIB-0066, Rev. 00, Oak Ridge Tennessee, April 26.

ORAUT (Oak Ridge Associated Universities Team), 2007b, SEC ER Related Interview Questions –

Mound (February 1, 1949-August 14, 2007), [Redacted] November 8, 2007. [SRDB Ref ID:

036878]

ORAUT (Oak Ridge Associated Universities Team), 2008, Interview with [Redacted], March 18, 2008. 

[SRDB Ref ID: 069248].

ORAUT (Oak Ridge Associated Universities Team), 2010, Estimating Doses for Plutonium Strongly 

Retained in the Lung, ORAUT-OTIB-0049, Rev. 01 PC-2, Oak Ridge Tennessee, 

November 29.

ORAUT (Oak Ridge Associated Universities Team), 2012, Internal Dosimetry Coworker Data for the 

Mound Site, ORAUT-OTIB-0061, Rev. 02, Oak Ridge Tennessee, April 13.

ORAUT (Oak Ridge Associated Universities Team), 2013, A Method for Estimating Stable Metal 

Tritide Exposures to Tritium Ancillary Workers Based on Swipe Data in Rooms SW-8, SW-13, 

SW-150 and R-108 at the Mound Laboratory, 1968 to 1989, ORAUT-RPRT-0057, Rev. 00, 

Oak Ridge Tennessee, March 7.

ORAUT (Oak Ridge Associated Universities Team), 2014, Technical Information Bulletin: Internal 

Dose Reconstruction, ORAUT-OTIB-0060, Rev. 01, Oak Ridge Tennessee, September 8.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 51 of 59

Phillips, C. A., 1988, Tritium in Urine, Technical Manual PD-80030, Issue 1 (TM-80030), Monsanto 

Research Corporation, Mound Facility, Miamisburg, Ohio, May 20. [SRDB Ref ID: 47166]

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

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

volume 83, number 5, pp. 594–789. [SRDB Ref ID: 12373]

Salutsky, M. L., K. Shaver, A. Elmlinger, and M. L. Curtis, 1956, “Separation of Protactinium from 

Uranium Residues,” Journal of Inorganic Nuclear Chemistry, volume 3, pp. 289–295. [SRDB 

Ref ID: 48045]

Sebelius, K., 2010, HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees from the Mound Plant, Miamisburg, Ohio, U.S. Department of Health 

and Human Services, Office of the Secretary, Washington, D.C., July 14. [SRDB Ref ID: 

165206]

Sebelius, K., 2012, HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees, Mound Plant, Miamisburg, Ohio, U.S. Department of Health and 

Human Services, Office of the Secretary, Washington, D.C., December 7. [SRDB Ref ID: 

165207]

Sedlet, J., and J. J. Robinson, 1971, “Abstract,” Elimination of Polonium-210 Following Accidental 

Inhalation of Neutron-Irradiated Bismuth, March. [SRDB Ref ID: 89281]

Sheehan, W. E., ca. 1951, Effective Half-Life of Polonium, Monsanto Research Corporation, Mound 

Laboratory, Miamisburg, Ohio. [SRDB Ref ID: 8641, p. 365]

Sheehan, W. E., W. R. Wood, and H. W. Kirby, 1963, “Urinalysis for Metabolized Plutonium,” 

Proceedings of the Ninth Annual Conference on Bioassay and Analytical Chemistry, October 

10-11, 1963, San Diego, California, TID-7696, U.S. Atomic Energy Commission, Washington, 

D.C. [SRDB Ref ID: 1962, p. 713]

Silverman, L. B., 1962, “A Short History of the Health and Safety Program at the Dayton Polonium 

Project During the World War II Period,” letter to S. L. Warren, April 10. [SRDB Ref ID: 

11769]

Spitz, H. B., 1989a, Retrospective Analysis of Polonium-210 Urinalysis Data Collected as Part of a 

Routine Occupational Monitoring Program Between 1943 through 1965, Draft, EG&G Mound 

Applied Technologies, Mound Facility, Miamisburg, Ohio, February 28. [SRDB Ref ID: 62520, 

p. 5]

Spitz, H. B., 1989b, “Statement of Parameter Values for Po-210 Dosimetry,” memorandum to B. 

Robinson, EG&G Mound Applied Technologies, Mound Facility, Miamisburg, Ohio, March 20. 

[SRDB Ref ID: 43691]

Spoerl, E. S., 1950, Urine Assay Procedure at the Mound Laboratory (Final Report), MLM-460, 

Monsanto Chemical Company, Mound Laboratory, Miamisburg, Ohio, April 14. [SRDB Ref ID: 

1962, p. 117]
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 52 of 59

Spoerl, E. S., 1951, “Problem Title: A Derived Biological Half-Life of Polonium in Humans,” Report for 

Biological Research, June 18, 1951 to September 17, 1951, MLM-626, Monsanto Chemical 

Company, Mound Laboratory, Miamisburg, Ohio, October 22. [SRDB Ref ID: 8641, p. 470]

Tomlinson, F. K., R. Brown, H. Anderson, and R. Robinson, 1974a, Application of Phoswich Detectors 

for Lung Counting Plutonium-238, pp. 999–1004, Proceedings of the Third International 

Congress of the International Radiation Protection Association held at Washington, D.C., 

September 9-14, U.S. Atomic Energy Commission, Oak Ridge Operations Office, Oak, Ridge, 

Tennessee, February 1. [SRDB Ref ID: 47562]

Tomlinson, F. K., R. Brown, H. Anderson, and R. Robinson, 1974b, Chestwall Tissue Measurements 

for Lung Counting Applications, pp. 1005–1010, Proceedings of the Third International 

Congress of the International Radiation Protection Association held at Washington, D.C., 

September 9-14, U.S. Atomic Energy Commission, Oak Ridge Operations Office, Oak, Ridge, 

Tennessee, February 1. [SRDB Ref ID: 14082]

Tomlinson, F. K., 1976, Calibration for Plutonium-238 Lung Counting at Mound Laboratory, MLM2339 (OP), Proceedings of the Workshop on Measurement of Heavy Elements In Vivo, 

Battelle Memorial Institute, Richland, Washington. [SRDB Ref ID: 47114] 

Tomlinson, F. K., 1990, “1-8 Tests of MDA Calculations Using Estimated Backgrounds and Chest 

Wall Thicknesses,” Monsanto Research Corporation, Mound Laboratory, Miamisburg, Ohio, 

August 6. [SRDB Ref ID: 8641, p. 280]

Themelis, J. G., 1989, “Review of Mound Po-210 Internal Dose Assessment Effort,” memorandum to 

J. A. Morley, U.S. Department of Energy, Albuquerque, New Mexico, June 15. [SRDB Ref ID: 

8641, p. 462]

UNC Geotech, 1990, Results of the U.S. Department of Energy Indoor Radon Study, Volume 1, 

DOE/ID/12584-75 Vol. 1, CNG/GJ-TP-1, Grand Junction, Colorado, August. [SRDB Ref ID: 

12287]

UNSCEAR (United Nations Scientific Committee on the Effects of Atomic Radiation), 1993, 

“Exposures from Natural Sources of Radiation,” Annex A, Sources and Effects of Ionizing 

Radiation, New York, New York. [SRDB Ref ID: 22742]

Wende, E. A., 1955, “Directive for Thorium Refinery, Mound Laboratory Project 2-24X-5604,” 

memorandum to Area Manager, Dayton Area, U.S. Atomic Energy Commission, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, March 11. [SRDB Ref ID: 42044]
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 53 of 59

 

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 insoluble in the 

respiratory tract (slow solubilization).

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 undesired locations including air, soil, buildings, animals, and persons.

curie (Ci)

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.

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. 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 54 of 59

 

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.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 55 of 59

 

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. 03 Effective Date: 03/03/2017 Page 56 of 59

 

ATTACHMENT B 

UNCERTAINTY ASSESSMENT

TABLE OF CONTENTS

SECTION TITLE 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...................................................................................................... 57

B.5 Particle Size Distributions....................................................................................................... 58

B.6 Solubility and Lung Clearance Type....................................................................................... 58

B.7 Radionuclide Use by Location................................................................................................ 58

B.8 Analytical Procedures ............................................................................................................ 59
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 57 of 59

ATTACHMENT B 

UNCERTAINTY ASSESSMENT

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 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 

estimate and considered favorable to claimants for all categories of radiation workers. Uncertainty of 

extrapolating doses from the period from 1959 to1962 to that from 1947 to 1956 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 
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 58 of 59

ATTACHMENT B 

UNCERTAINTY ASSESSMENT

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 

absorption 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 absorption 

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 20% 

larger than corresponding dose coefficients for 1-μm particles. Therefore, uncertainty in dose 

coefficients in relation to particle size distribution is dependent on absorption 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). The absorption type of a 

radionuclide is one of the most important parameters that affects internal dose (NIOSH 2002). Lung 

absorption 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 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 the primary radionuclides of 

concern (Table 5-1). There is little uncertainty associated with documentation for 210Po, the plutonium 

isotopes, and tritium along with 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 SRS source 

material.
Document No. ORAUT-TKBS-0016-5 Revision No. 03 Effective Date: 03/03/2017 Page 59 of 59

ATTACHMENT B 

UNCERTAINTY ASSESSMENT

Information on the location and use of radioactive gases was primarily available only 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 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 Mound analytical procedures 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 

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 that was determined with low-background proportional counting to specific alphaemitting 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 MDAs.