ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 66

DOE Review Release 04/30/2009

Document Title:

Mound Site – Occupational Internal Dosimetry

Document Number: ORAUT-TKBS-0016-5 

Revision: 00 PC-1 

Effective Date: 03/13/2009

Type of Document TBD

Supersedes: Revision 00

Subject Expert(s): Jere Millard

Site Expert(s): N/A

Approval: Signature on File Approval Date: 03/04/2009

Donald N. Stewart, Document Owner

Concurrence: Signature on File Concurrence Date: 03/04/2009

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 03/04/2009

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 03/11/2009

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/13/2009

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 2 of 66

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.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 3 of 66

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

5.1 Introduction ..............................................................................................................................7

5.1.1 Mound Feed Materials..................................................................................................8

5.1.2 Radionuclides of Concern.............................................................................................8

5.1.3 Internal Dose Control Program .....................................................................................9

5.1.4 Gross Alpha Analysis ................................................................................................. 11

5.1.5 Detection Limit and Minimum Detectable Amount Quantification................................ 11

5.2 Building Use and Site Processing Activities ...........................................................................11

5.2.1 Particle Size Distributions ........................................................................................... 12

5.2.2 Equilibrium Assumptions ............................................................................................ 13

5.3 Primary Bioassay Detection Data...........................................................................................13

5.3.1 Primary Radionuclides................................................................................................ 13

5.3.1.1 Tritium.......................................................................................................... 14

5.3.1.2 Polonium-210............................................................................................... 17

5.3.1.3 Protactinium-231.......................................................................................... 22

5.3.1.4 Plutonium-238 and Plutonium-239 ............................................................... 23

5.3.2 Secondary Radionuclides ........................................................................................... 28

5.3.2.1 Radon Isotopes............................................................................................ 28

5.3.2.2 Thorium-230................................................................................................. 30

5.3.2.3 Natural Thorium ........................................................................................... 31

5.3.2.4 Radium-226, Actinium-227, and Thorium-228.............................................. 33

5.3.2.5 Uranium Isotopes......................................................................................... 34

5.3.2.6 Americium-241 and Curium-244................................................................... 35

5.3.2.7 Americium-241 and Curium-244 Recommendations .................................... 36

5.3.3 Summary of Correction Factors and Excretion Equations........................................... 36

5.3.4 Summary of Analytical Methods and Detection Limits ................................................ 37

5.4 Secondary Bioassay Detection Data ......................................................................................37

5.4.1 Fecal Sample Analysis ............................................................................................... 39

5.4.2 Whole Body Counting................................................................................................. 40

5.4.3 Personnel Air Sampling Data...................................................................................... 41

References ........................................................................................................................................42
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ATTACHMENT A, LUNG SOLUBILITY RECOMMENDATIONS FOR SITE 

PROCESS ACTIVITIES ......................................................................................47

ATTACHMENT B, SECONDARY RADIONUCLIDES BY LOCATION AND 

PERIOD ..............................................................................................................50

ATTACHMENT C, CODING INSTRUCTIONS FOR URINALYSIS RECORDS ................................. 62

ATTACHMENT D, UNCERTAINTY ASSESSMENT..........................................................................63

D.0 UNCERTAINTY ASSESSMENT ............................................................................................ 63

D.1 CONTAMINATION OF SAMPLES .........................................................................................63

D.2 ASSIGNMENT OF GROSS ALPHA RESULTS AS PLUTONIUM .......................................... 63

D.3 MISSING BIOASSAY RECORDS ..........................................................................................63

D.4 ASSUMPTION OF EQUILIBRIUM .........................................................................................64

D.5 PARTICLE SIZE DISTRIBUTIONS ........................................................................................64

D.6 SOLUBILITY AND LUNG CLEARANCE TYPE ......................................................................64

D.7 RADIONUCLIDE USE BY LOCATION...................................................................................65

D.8 ANALYTICAL PROCEDURES ...............................................................................................65

LIST OF TABLES

TABLE TITLE

5-1 Bioassay programs and action levels .....................................................................................

PAGE

10

5-2 Radionuclide use by building.................................................................................................. 12

5-3 Recommended equilibrium assumptions for secondary radionuclides.................................... 14

5-4 Reported analytical procedures and detection limits for tritium............................................... 15

5-5 Numbers of workers in reported dose ranges by year ............................................................ 17

5-6 Reported background count rates and derived minimum detectable 

activities for 210Po................................................................................................................... 19

5-7 Protactinium detection limits................................................................................................... 23

5-8 Heat source plutonium feed material characteristics .............................................................. 24

5-9 Weapons-grade plutonium feed material characteristics ........................................................ 24

5-10 Plutonium compound usage in the PP Building...................................................................... 25

5-11 Summary of historical plutonium analytical procedures .......................................................... 26

5-12 Plutonium for reported periods ............................................................................................... 27

5-13 Radon isotope survey data by building and period ................................................................. 28

5-14 Radon-222 WLM assumptions for dose reconstruction .......................................................... 29

5-15 Thorium-230 detection limits .................................................................................................. 31

5-16 Natural thorium detection limits .............................................................................................. 32

5-17 Uranium isotope detection limits............................................................................................. 35

5-18 Americium-241 and curium-244 detection limits ..................................................................... 36

5-19 Summary of correction factors and excretion equations ......................................................... 37

5-20 Analytical methods and detection limits by time period........................................................... 39

5-21 Radionuclide chemical recoveries for collected fecal samples ............................................... 40

5-22 Reported 1994 chest counting MDAs ..................................................................................... 40

A-1 Recommended lung clearance types for site process activities .............................................. 47

B-1 Secondary radionuclides by location and period .................................................................... 51

C-1 Coding instructions for urinalysis records ............................................................................... 62
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 5 of 66

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AL Action Limit

ALI annual limit of intake

AMAD activity median aerodynamic diameter

Bq Becquerel

Ci curie

cm centimeter

cph counts per hour

cpm counts per minute

d day

DOE U.S. Department of Energy

DL Detection Limit

dpm disintegrations per minute

EDTA ethylene diamine tetraacetic acid 

hr hour

HTO tritiated water vapor

GSD Geometric Standard Deviation

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in inch

keV kilovolt-electron, 1,000 electron volts

kg kilogram

kV kilovolt

L liter

mCi millicurie

MDA minimum detectable activity

min minute

ml milliliter

mo month

MPBB maximum permissible body burden

MPC maximum permissible concentration

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NR Not Reported

pCi picocurie

PERM Passive Environmental Radon Monitor

SNAP System for Nuclear Auxiliary Power
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 6 of 66

wk week

WL working level

WLM working level month

yr year

μCi microcurie

μg microgram

μm micrometer
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 7 of 66

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 historic background information and guidance to assist in the preparation of 

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

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

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

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

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

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

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

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

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

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

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

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

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2007):

] guidelines established under 

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

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

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

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

conventional structures

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

reasons

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 8 of 66

5.1.1

The following process information for Mound feed materials was extracted from Oak Ridge Associated 

Universities (ORAU) files, Processes and Characteristics of Major Isotopes Handled at Mound (ORAU 

2003a):

Mound Feed Materials

• Airport Cake (Cotter Cake): Primary feed material for extraction of Thorium-230 (230Th). The 

feedstock was raffinate produced by extraction of uranium from a hydrofluoric or nitric acid 

solution using diethyl ether followed by treatment with caustic or lime. This feed source was 

obtained from Mallinckrodt and the Cotter Corporation.

• Brazilian Monazite Sludge: Primary feed material for the thorium refinery project, treated with 

caustic soda to create insoluble oxides or hydroxides and soluble sodium phosphate. The 

materials were then leached with hydrochloric acid to extract various rare earth elements. The 

feed material was sometimes acidic and sometimes basic, depending on pretreatment for a 

particular batch received.

• Sperry Cake: One of several feed materials derived from uranium milling for extraction of 

Protactinium-231 (231Pa). Protactinium-231 extraction and purification was a complex 

process that is poorly described. There is no information to estimate the radioactive 

compounds likely to have been present.

5.1.2

Radionuclides were categorized as either primary or secondary by the authors based on the potential 

extent of use and the amount of bioassay data available. Primary radionuclides of concern are 

discussed in Mound internal dosimetry documentation, History of Mound Bioassay Programs (Meyer 

1992) and Mound Site Radionuclides by Location (King 1995). Primary radionuclides assessed for 

occupational exposure included:

Radionuclides of Concern

• Polonium-210 (210Po)

• Plutonium Isotopes: Plutonium-238 (238Pu), Plutonium-239 (239Pu), Plutonium-240 (240Pu), 

Plutonium-241 (241Pu), Plutonium-242 (242Pu),

• Protactinium-231 (231Pa)

• Tritium (3

H)

These radionuclides were identified as primary because they can deliver relatively high doses to 

human receptors, had widespread use at the site in multiple buildings and facility processes, and had 

sufficient site documentation to evaluate internal dosimetry. Tritium was included due to potential 

absorption via hydrogen exchange across the skin barrier. Primary onsite operations used all of these 

radionuclides in conducting key mission activities. Polonium-210 was referred to as Postum, 

Thorium-230 was referred to as Ionium, and Tritium was referred to as Hot Gas in Mound 

documentation. 

Secondary radionuclides of concern were assessed with limited internal dosimetry data and included 

the following in addition to daughter products from the uranium and thorium decay chains: 

• Americium-241 (241Am)

• Curium-244 (244Cm)
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 9 of 66

• Uranium Isotopes: Uranium-238 (238U), Uranium-235 (235U), Uranium-234 (234U), Uranium233 (233U)

• Thorium Isotopes: Thorium-232 (232Th), Thorium-230 (230Th), Thorium-228 (228Th)

• Actinium-227 (227Ac)

• Radium Isotopes: Radium-228 (228Ra), Radium-226 (226Ra), Radium-224 (224Ra), Radium223 (223Ra)

• Radon Isotopes: Radon-222 (222Rn), Radon-220 (Thoron, 220Rn), 

• Radon-219 (Actinon, 219Rn)

These radionuclides are secondary due to limited use at the site or limited worker exposures. 

Americium-241 should be considered as a separate radionuclide resulting from alpha source 

production, neutron source production, analysis and vaporization of plutonium compounds, 231Pa 

separation, and reactor fuel research programs. Americium-241 was also a minor constituent of 

weapons-grade plutonium materials used at Mound. Short-lived radioactive gases and their 

respective daughter products were present at Mound from activities involving uranium and thorium 

chain extraction and separation activities or the manufacturing of sealed sources. However, worker 

exposures to these gases were minimized by a ventilation system installed in 1980 and were largely 

restricted to the SW building where radium and actinium processing took place. Xenon and krypton 

were present in H Building as a part of the Isotope Separation Program and xenon isotope separation 

activities. However, the International Commission on Radiological Protection (ICRP) stated in 

Publication 30 that external irradiation dominates worker dose to the extent that internal dose should 

be disregarded for most noble gases, including xenon and krypton (ICRP 1979). Radon and thoron 

are noble gases, but exposure could result in significant internal doses. Curium-244 (244Cm) and 

Neptunium-237 (237Np) were present at the site in minor concentrations in special nuclear heat source 

material from the Savannah River Site and present in aqueous wastes derived from plutonium 

recovery operations. Curium-244 and 237Np were also used in reactor fuel research and support to 

the heat source program. Reactor fuel research used 237Np and 244Cm in metallic or oxide form. 

Curium-244 has a relatively high dose conversion factor less than a factor of two below 239Pu and 

232Th and therefore the potential to induce relatively high doses via the inhalation pathway should be 

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

quantities and were of secondary radiological concern. 

Mound engaged in biomedical research with Calcium-45 (45Ca), Iron-59 (59Fe), Cobalt-60 (60Co), Zinc65 (65Zn) and other radionuclides. These were present in relatively minor quantities in comparison to 

the primary and secondary radionuclides and were used in very few buildings for limited periods. 

Most of these radionuclides were present as sealed sources. Worker exposure was less likely, and 

bioassay data was not collected for these radionuclides. Exposure to these radionuclides can 

therefore be considered insignificant.

5.1.3

Bioassay included primarily in vitro urine data and limited fecal sample results. In vitro urine data 

comprises the primary source of information to support dose reconstruction. Nose swipe data were 

considered to be secondary information and were not evaluated. Whole body counting was not 

assessed for tritium, 210Po, 238Pu, 239Pu, and 231Pa because these nuclides are nearly pure alpha 

emitters or have weak gamma emissions. Air-sampling data was also considered secondary 

information contingent upon demonstration that measured concentrations were representative of 

breathing zone concentrations. Various bioassay programs were conducted at Monsanto Chemical 

Internal Dose Control Program
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 10 of 66

Company facilities and at Mound from 1945 to the present using routinely collected urine samples 

from workers (Meyer 1992). The polonium bioassay program described below was also used in Units 

I, III, and IV of the Monsanto Chemical Company; however, this document provides dose 

reconstruction provisions only for the Mound Laboratory. Table 5-1 lists primary bioassay programs 

with reported radionuclide action levels in counts per minute for a urine sample, workers who were 

monitored, and frequency of monitoring. 

Table 5-1. Bioassay programs and action levels. 

Bioassay program Period Action levelsa Monitored workers Frequency

In vitro monitoring

Po-210 (Postum) recovery 1945–1974 10 cpm 50 ml-1 Operations personnel Weekly spots

Weekly 24 hr

Ac-227, Ra-226, Th-228 1951–1960 8.3 cpm 24 hr-1 Research personnel Weekly 24 hr

Th-232 sludge, redrumming1955–1975 2.8 cpm 24 hr-1 Th refinery personnel Monthly 24 hr

Th-230 (Ionium) research 1956–1959 3.5 cpm 24 hr-1 Research personnel Bi-weekly 24 hr

Pa-231 extraction 1956–1960 2.2 cpm 24 hr-1 Research personnel Monthly 24 hr

U-233 research 1958–1960 20 cpm 1,500 

ml-1

10 Research personnel Weekly 24 hr

Tritium (hot gas) production 1957–

present

10 μCi L-1 Recovery personnel Weekly spots 

Weekly 24 hr

Pu-239 neutron source 1957–

present

3.5 cpm 24 hr-1 Processing personnel Weekly spots

Weekly 24 hr

Pu-238 heat source 1960–

present

3.5 cpm 24 hr-1 Processing personnel Weekly spots

Weekly 24 hr

Am-241 & Am-243 1983–

present

0.0 pCi/d b Potentially exposed 

personnel

Daily

Cm-244 1983–

present

0.0 pCi/d b Potentially exposed 

personnel

Daily

In vivo monitoring

Whole body counting 1960–1970 Not available Operations personnel Annual

Chest counting 1970–

present

Variablec Operations personnel Annual

a. AL = action levels; observed counts per minute for 50 ml or 24-hour samples that result in a worker dose of 300 mrem 

wk-1 to the target organ.

b. Smallest reported activity in database.

c. Chest counting MDAs reported in the Mound TBD, Table 5-19 for Am, Pu, Th, and U isotopes. Current day MDAs range 

from 0,1 nCi for Am-241 to 130 nCi for Pu-239.

Routine bioassay monitoring included collection of spot and 24-hr urine samples identified in 

Table 5-1. For example, only those workers conducting research to separate and purify protactinium 

were monitored monthly for 231Pa. Twenty-four hour samples were also requested for 210Po and the 

plutonium isotopes when requested. Monitoring frequencies were surmised from actual performance 

data and personnel sampling records.

Individuals identified as being involved in an internal exposure incident or had urinalysis results in 

excess of the reporting limits listed in Table 5-1 were required to submit additional urine 24-hr 

samples for analysis. Administrative personnel were not monitored and monitoring was discontinued 

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

personnel were monitored annually by whole body or chest counting or if there was a suspected 

exposure incident.

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

target organ dose limit. Results in excess of action levels triggered re-sampling and placement of the 

worker on the “hot roster” to control further exposures if necessary. Reporting limits for americium 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 11 of 66

and curium were 0.0 and any detectable activity was reported and triggered re-sampling. Typical 

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

facility. These controls followed the limits stated in Handbook 52 published in 1953 (NBS 1953).

5.1.4

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 

Gross Alpha Analysis

separation followed by gross alpha counting. Alpha spectrometry capable of identifying specific alpha 

emitting isotopes was not introduced until 1980. Section 5.3.2.4 describes one of these procedures. 

Claimant-favorable assumptions have therefore been made when isotopic data were in question or 

unavailable. Assignment was also based on workplace information.

5.1.5

The decision levels and minimum detectable activities (MDAs) are calculated using the following 

relations for counting of urine (Currie 1968):

Detection Limit and Minimum Detectable Amount Quantification

TREVA

C

D

BACKGROUND

L

3

2

.4 65

+

= (5-1)

where

DL = decision level in dpm L-1

CBACKGROUND = background counts in the region of interest

T = count time

R = recovery

E = average detector efficiency

V = sample volume (usually 1.5 L)

A = the alpha abundance for the radionuclide in question

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

confidence level. 

TREVA

C

MDA

.4 65 BACKGROUND + 3 = (5-2)

The reporting level is different from the MDA; it is the minimally acceptable decision level that the 

bioassay laboratory was required to achieve. MDAs are typically below the reporting levels. Mound 

technical documentation provided no quantitative relationship between the reporting level and MDAs. 

In general, detection limit data for urine bioassay was sparse. 

5.2 BUILDING USE AND SITE PROCESSING ACTIVITIES

Table 5-2 summarizes radionuclide use by building based on information from King (1995). This 

information should be considered when reconstructing occupational doses. The numbers of rooms 

where radionuclides had primary use and were cited as a major dosimetric concern are shown below. 

Many radionuclides were not used or had limited use in buildings. For example, 227Ac was not used in 

HH, PP, SM, and T Buildings. Dose assessments should not be conducted for radionuclides that had 

no demonstrated use in buildings where exposure occurred. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 12 of 66

5.2.1

The Mound Laboratory conducted a wide variety of research and weapons production activities using 

primary and secondary radionuclides. Particle size distributions ranged from submicrometer to large 

metal-clad materials. In addition, Mound used or investigated a complex array of chemical 

compounds, which resulted in a full range of reported lung clearance types. Attachment A, Table A-1 

lists recommended lung clearance class information for radionuclides used in primary building 

activities and processes. A claimant-favorable solubility should be used if a specific solubility type 

Particle Size Distributions

Table 5-2. Radionuclide use by building. 

Nuclide

Buildinga

HH PP R SM SW T WD/WDA

Primary radionuclides

H-3 8/8/13 None 6/15/52 None 19/28/38 24/32/96 1/8/14

Po-210 4/4/13 None 30/29/52 None 4/5/38 49/69/96 4/5/14

Pu-239 None 39/39/39 5/28/52 19/19/19 6/7/38 3/15/96 1/11/14

Pu-238 None 39/39/39 25/28/52 19/1919 7/7/38 4/21/96 10/11/14

Ac-227 None None 24/29/52 None 2/29/38 None 0/6/14

Pa-231 2/2/13 None 5/13/52 None 8/13/38 None 0/2/14

Secondary radionuclides

Cm-244 None 0/7/39 0/6/52 None None None None

Am-241 None 0/7/39 6/13/52 0/5/19 0/1/38 0/6/96 0/10/14

Np-237 None 0/7/39 0/4/52 0/3/19 None None 0/10/14

Uranium 0/2/13 None 6/27/52 0/2/19 0/29/38 0/13/96 0/11/14

U-233 0/1/13 None 2/2/52 None 1/1/38 0/2/96 None

Ra-228 None None None None None None 0/2/14

Ra-226 None None 0/27/52 None 3/9/38 0/3/96 0/3/14

Rn-222 None None 0/27/52 None 3/9/38 0/3/96 0/2/14

Th-232 0/2/13 0/7/13 0/10/52 0/4/19 6/15/38 None 0/5/14

Th-230 2/1/13 None 5/30/52 None 7/11/38 None 0/5/14

Th-228 0/2/13 0/7/13 0/26/52 0/4/19 2/19/38 None 0/5/14

a. nn/nn/nn = number of rooms where a radionuclide was of major concern / total number of rooms 

using a radionuclide / total number of rooms where radioactive materials were used, stored, or 

present. 

cannot be determined. In addition, solubility can be determined based on claim-specific intake or 

bioassay data for a given cancer type and might differ from solubility recommendations stated in table 

A-1. Particle size expressed in activity median aerodynamic diameter (AMAD) and lung clearance 

types were evaluated based on site and process information from Meyer (1992) and King (1995). In 

the absence of specific site data, AMAD particle size and lung clearance types were based on 

recommendations in ICRP Publications 66 and 68 (ICRP 1994, 1995).

Particle sizes at Mound reportedly ranged from submicrometer to 10s of micrometers (King 1995). 

Submicrometer particles were reported to be abundant, and significant quantities of plutonium dioxide 

were vaporized during plasma torch operations (King 1995). Such plutonium particles from plasma 

torch operations would be defined as fumes by industrial hygiene and toxicology terminology (Soule 

1991). In addition, microsphere processing was conducted at Mound with a range of product particle 

sizes from 50 to 250 μm (King 1995). Processes that likely produced submicrometer particle size 

distributions included:

• Plasma torch operations with vaporization of plutonium in the PP building

• Microsphere production using plutonium in the PP and SM buildings

• Polonium heat source manufacturing and research in the R building

• Plutonium metal production in the SM building
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 13 of 66

However, much of the reported particle size distribution information from the site was qualitative in 

nature and could not be verified by evaluation of quantitative particle size study results. In addition, 

it’s not clear whether actual breathing zone particle sizes were based on reported estimates of 

process particle size distributions. 

Due to these limitations, the default particle size distribution AMAD from ICRP Publication 66 (ICRP 

1994) of 5 μm should be used for all processes, as indicated in Table A-1. The default AMAD of 5 μm 

should be used to estimate intake and dose with an associated geometric standard deviation of 2.5 or 

the default value given in IMBA. 

5.2.2

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 low 

background 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 following 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 5-3 lists 

claimant-favorable equilibrium assumptions to infer primary radionuclide activity in urine based on 

detected alpha activity.

Equilibrium Assumptions

Mound did bench-scale reworking of K-65 uranium extraction residue used at Mallinckrodt, Weldon 

Springs, and Fernald as a potential source of radium. There was also an initial attempt to produce 

210Po initiators by reworking radium and uranium tails. This process was quickly abandoned after the 

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

therefore assume equilibrium with 210Pb. 

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 to 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. 

5.3 PRIMARY BIOASSAY DETECTION DATA

This section presents primary bioassay information to support intake and dose reconstruction. 

Primary bioassay data resulted from routinely collected in vitro urine samples followed by various 

analyses for radionuclide content. Analysis data were available for the primary radionuclides. 

Recommendations are provided to facilitate dose reconstruction.

5.3.1

This section discusses specific information for the primary radionuclides, including in vitro bioassay 

data, history of MDAs, detection limits, reporting limits, percent recoveries, and analytical procedures. 

Recommendations are provided to facilitate dose reconstruction.

Primary Radionuclides
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 14 of 66

5.3.1.1 Tritium

There was reported tritium (3

H) recovery, purification, and use for solid tritiated metal compounds in 

1948 in R building and tritium use limited to testing and research activities in T building as early as 

1947 (sic, King 1995)2

Table 5-3. Recommended equilibrium assumptions for secondary radionuclides. 

. Tritium was used extensively at Mound since the start of weapons component 

production in 1954. 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, 

Radionuclide Method of inference Recommendations

Ac-227 Consecutive gross alpha counts of 

extracted radium. Ra-223 with 11.4-d 

half-life detected in urine sample.

Assume Ac-227 and Ra-223 are in 

equilibrium and activity intake rates are 

equal.a 

Th-228 Consecutive gross alpha counts of 

extracted radium. Ra-224 with 3.6-d 

half-life detected in urine sample. 

Assume Th-228 and Ra-224 are in 

equilibrium and activity intake rates are 

equal.a 

Ra-226 Consecutive gross alpha counts of 

extracted radium. Ra-226 with 

1,602-yr half-life detected in urine 

sample.

No assumption necessary.

Th-230 Gross alpha counting of extracted 

thorium in urine starting with sample 

6507-1I on August 1954. Gross 

alpha counting of extracted radium in 

urine beginning with sample 6507-10I 

on September 1954-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 

enriched uranium.

Po-210 Electrodeposition followed by low 

background proportional counting.

Assume equilibrium with Pb-210 to 

determine worker intakes.

Uranium isotopes: 

natural Ud, depleted Ue

Fluorimetric or gross alpha 

determination.

Assume natural uranium ratios for U-238, 

U-234, and U-235 or use IMBA default 

ratios for U-nat and DU.

a IMBA Expert OCAS Edition 3.0.63 (Birchall et al. 2003) does not directly calculate excretion of progeny due to intake of 

parent radionuclides with progeny.

b Thorium-232 is also identified as natural thorium and Thorium-230 as ionium.

c Uranium enriched in 234U and 235U. The degree of enrichment was not reported. These uranium isotopes would not 

be present as isotopically pure materials.

 

2 In fact, the construction of T Building was not complete until December 20, 1948, the first occupation 

of the building was March 15, 1948, and formal acceptance of the building was February 28, 1949 

(SRDB Ref ID 25491: T Building Structural History and Process History Summary Background 

Document).
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 15 of 66

d Uranium with naturally occurring mass and activity ratios.

e Uranium depleted in 234U and 235U.

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 at all buildings 

except PP (Plutonium Processing) and SM (Special Metallurgical). Extensive use of tritium occurred 

in R (Research), SW, and T Buildings. These activities resulted in tritium-contaminated soils under 

SW Building (King 1995). Eight tritium incidents reported by King (1995) occurred in SW Building 

from 1964 to 1986. Tritium exposures ended in 1995 as reported by King (1995). 

In Vitro Bioassay for Tritium

No urinalysis data were available from 1948 to 1956. Tritium urinalysis for workers started in 1957. A 

vibrating Reed Electrometer method of analysis was officially adopted from a Savannah River Plant 

procedure in 1958 and continued into the early 1960s. This method used 20 ml of urine mixed in a 

calcium chloride solution to liberate tritium gas, then analyzed the gas by passing an acetylene carrier 

gas from the urine sample into a 1-L metal ionization chamber. The measured current was therefore 

proportional to the tritium-acetylene mixture. The reported MDA for this method was 1 μCi L-1 of urine 

in Recommended Procedures for Tritium (Meyer 1992). Estimates of accuracy were 1 to 2 μCi L-1 at 

1 μCi L-1, 9.5 to 10.5 μCi L-1 at 10 μCi L-1, and 99 to 101 μCi L-1 at 100 μCi L-1 (Meyer 1992). 

Liquid scintillation counting began in 1961, with an approved procedure in 1968. This methodology 

used a 3-ml urine sample placed in an Ultimate Gold cocktail solution (Butler 1961). A quench curve 

was used to correct results for counting efficiency affected by chemical and physical factors that 

inhibit photon detection. In addition, many liquid scintillation detectors used external Barium-133 

(

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-1 of HTO (Mound 1981). Tritium 

results were always reported in units of activity per liter of urine. Table 5-4 lists the reported analytical 

procedures and detection limits for tritium. 

Table 5-4. Reported analytical procedures and detection limits for 

tritium. 

Period Analytical method Detection limit (μCi L-1)

1948–1956 No urinalysis conducted Use 95% dosea

1957–1961 Vibrating reed electrometer 1.0

1962–1995 Liquid scintillation 0.01

a. The upper 95 percentile dose of 340 mrem should be used to reconstruct doses. 

Tritium Dose Calculations

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

used at Mound to current ICRP 67 and ICRP 68 dosimetry. Correction from ICRP 30 to ICRP 77 is 

not needed due to an overall correction factor of 1.0 as shown in equation 5-5 below. 

A dosimetry method developed at Los Alamos National Laboratory (LANL) was used for acute tritium 

dose calculations from 1957 through 1988 (Lawrence 1957). This methodology assumed a maximum 

permissible body burden (MPBB) of 3.7 mCi, a relative biological equivalent of 1.7, an effective halflife of 12 days, and a body water compartment of 43.4 kg. This results in a maximum permissible 

concentration (MPC) of 85 μCi L-1 for a target dose limit of 300 mrem wk-1 under the LANL system of 

dosimetry (Lawrence 1957, DOE 1988). The target dose limit of 300 mrem wk-1 was based on an 

organ dose limit of 15 rem (FRC 1960). The LANL dose relationship used at Mound is shown in 

equation 5-3. 

D(t) = 0.00873 B (5-3)
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 16 of 66

where

B = measured acute body concentration in urine or body water 

(uCi/L)

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

Effective T1/2 = 12 days

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

equation 5-3 for tritium dose calculation and that 6 μCi L-1 corresponds to 52 mrem as entered on the 

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

confirmed basis to correct doses reported by Mound using equation 5-4. 

Current dosimetry for tritium uses a QF = 1.0 compared to an assumed LANL RBE = 1.7, a two 

compartment elimination model, and a target mass of 68.8 kg compared to 43.4 kg used in the LANL 

calculations (ICRP 67; ICRP 68). Equation 5-4 provides a ratio of ICRP 68 dose compared to LANL 

dose, which results in a ratio of 0.34.

=

× ×+

×

× =

∫

∫ ∞ ×−

∞ ×− ×−

0

.0 693 /12

0

.0 693 /10 .0 693 / 40 .0( 97 .0 03 )

68 7.18.

43 4. ( 68 )/

e dt

e e dt

Ratio ICRP LANL

t

t t

0.34 (5-4)

Beginning in 1989, tritium doses were reported in terms of effective dose equivalent under ICRP 30 

methodology. Equation 5-5 determines the ration of ICRP 68 dose compared to ICRP 30, using the 

ICRP 67 target mass of 68.8 kg. The ratio of 1.0 confirms that no dose corrections would be required 

for doses reported from 1989 to 1995.

0.1

.0( 97 .0 03 )

68 8.

63 ( 68 / 30)

0

.0 693 /10

0

.0 693 /10 .0 693 / 40

=

× ×+

×=

∫

∫ ∞ ×−

∞ ×− ×−

e dt

e e dt

Ratio ICRP ICRP

t

t t

(5-5)

Tritium Reporting Limits

Mound reporting limits for tritium impacted 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-1 (Meyer 1992). Reporting limits of 3 uCi/L as discussed by 

James (1997) or 6 μCi L-1 were both reported on the Radioactive Gas Exposure Evaluation Forms as 

actual bioassay results. The 6 uCi/L should be assumed when tritium doses have been reported as 

blanks, otherwise all actual reported bioassay results should be used.

Values of 0.0 were used for non-detection results, which biased the doses below the true estimate of 

the average dose. MDAs in Table 5-4 should therefore be used to estimate intakes at or below 

detection.

Unmonitored Tritium Exposures

This section provides a methodology to reconstruct tritium doses by assuming that doses of tritium 

from 1948 through 1956 were similar to reported doses from 1959 to 1962. Personnel dose data from 

Bigler’s Annual Monitoring Reports from 1959 to 1962 reported numbers of workers by category of 

annual doses as shown in Table 5-5 (Meyer 1992). These data clearly demonstrate that 94% to 98% 

of all workers received doses below one rem from 1959 to 1962. The dose in all years was therefore 

assumed to be the upper end of the 0-1 rem range. The value of 1000 mrem was assumed to be the 

approximate upper 95 percentile and claimant-favorable dose for most workers due to data limitations. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 17 of 66

Table 5-5. Numbers of workers in reported dose ranges by year. 

Dose

rem/year

Number of workers (percent of total)

1959 1960 1961 1962

0–1 117 (94) 249 (96) 286 (98) 303 (98)

1–2 6 (5) 5 (2) 6 (2) 4 (1)

2–3 1 (1) 3 (1) 0 3 (1)

3–4 0 1 0 0

4–5 0 0 0 0

5–10 0 2 (1) 0 0

Max dose (rem) 2.299 8.353 1.693 2.974

Corrected max dose (rem) 0.782 2.840 0.576 1.011

Total # workers 124 260 292 310

All tritium doses reported in Table 5-5 were derived by the LANL dosimetry methodology given in 

equation 5-3. Correction of these doses using equation 5-4 to ICRP68 dosimetry results in an upper 

95 percentile dose of 1000 (0.34) = 340 mrem. This ICRP68 equivalent annual upper 95 percentile 

dose should be assumed for all missing bioassay data from 1947 through 1956.

Reported annual uncorrected and corrected maximum doses in rem are also shown in Table 5-5 and 

ranged from 0.576 to a high of 2.840 corrected (Meyer 1992).

Metal Tritides

Limited information is available on the metal tritides 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 U.S. Department of Energy (DOE) report 

notes that lithium tritide reacts with water vapor to produce mostly HT (DOE 1994). Therefore, the 

dose from intake of lithium tritide is bounded by the assumption that it reacts to produce HTO. 

Insoluble tritides were also used at Mound primarily in SW, R, and T buildings. A lung clearance class 

of S should be assumed for all metal tritides other than lithium. A lung clearance class of S should 

also be assumed for insoluble tritides. 

Tritium Recommendations

The corrected upper 95 percentile dose of 340 mrem should be assumed in assigning annual doses 

to unmonitored radiological workers from 1948 to 1956. When Mound records from 1957 to 1995 only 

include dose without any concentrations, equations 5-4 should be used to correct reported Mound 

doses to current ICRP68 doses. Tritium doses reported as blanks from 1957 to 1995 should be 

reconstructed by assuming a 6 μCi per liter urine concentration reporting limit. Otherwise, all actual 

recorded bioassay results should be used. Doses reported at or below detection or as zeros should 

be corrected using the applicable MDAs given in Table 5-4. Doses would not be assigned to any 

administrative or other non-radiological workers. 

Dose reconstructions should assume that the tritium source term was constant because reliable air 

data was not available from 1947 to 1957 and effluent records started in 1959, which precluded any 

evaluation of the tritium source terms during this period of missing urinalysis. 

Intake of lithium tritides should be estimated by assuming exposure to HTO. Exposure to all other 

tritides should assume class S clearance.

5.3.1.2 Polonium-210

Polonium work transferred to Mound in 1949 (Silverman 1962) from the Monsanto Chemical 

Company facilities (Units III and IV). Polonium-210 was primarily produced by neutron irradiation of 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 18 of 66

naturally occurring Bismuth-209 (209Bi) slugs to produce Bismuth-210 (210Bi) via a (n, γ) reaction 

followed by beta decay to 210Po. The first documented polonium exposure occurred on August 28, 

1944, via a probable skin absorption (Silverman 1962; Spitz 1989a; Runkle 1995). 

Polonium-beryllium (PoBe) neutron source work (210Po) started at Monsanto Chemical Company 

facilities in September 1945. Mound’s first primary mission was to manufacture initiators for nuclear 

weapons. This included the extraction of 210Po from neutron-irradiated bismuth slugs and beryllium 

machining operations. In addition, Mound conducted polonium research and started the 

manufacturing of radioisotope thermal generators using polonium in 1954. Non-weapons polonium 

work at Mound ended in 1972.

Polonium-210 was present in Building HH (Hydrolysis House) from 1949 to 1959 generating sludge 

materials with 0.02 Ci L-1 total alpha activity. Extensive use of 210Po in the R Building involved 

research and electrodeposition studies from 1948 to 1951 and the manufacturing of PoBe neutron 

source from 1956 to 1961. PoBe sources were first manufactured at Mound as neutron source 

initiators to support weapons programs. Polonium was later used to produce 210Po heat source 

generators.

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, storage pool for irradiated bismuth slugs, neutron source processing, 

and analytical laboratories. Aqueous wastes containing Polonium-208 (208Po), Polonium-209 (209Po), 

and 210Po have been present in the WD and WDA Buildings from 1949 to the present. Chemical 

forms included polonium nitrate and polonium chloride. There was no cited 210Po activity in the PP 

and SM Buildings. A single 210Po incident was reported in the SW Building in 1961 and 11 incidents 

are associated with the T Building from 1954 to 1968, where primary polonium separation and 

processing took place.

Polonium-210 Records

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

discussed by Meyer (1992) and Spitz (1989). Weekly Reports of urinalysis data were kept for workers 

grouped by location with entries in cpm per specified volume and 24-hr samples identified. 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 collected on a single day to a volume weighted average 24-hr excretion in cpm. Finally, 

Special and Unscheduled Sample Forms were kept when any additional urine samples were required 

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

duplicate indicated. Bioassay results in activity per day are reported as Bq/d in the PURECON 

database under the heading of BQ_DAY. Sample results have also been reported in cpm and dpm in 

this database. 

In Vitro Bioassay of Polonium-210

Summary data for 210Po are provided in Tables 5-6 and 5-19. Bioassay data for 210Po have been 

placed in the PORECON database by MJW by evaluating individual dosimetry files from original 

logbooks (MJW 2002). This database contains individual sample results from 1944 to 1973, 

encompassing polonium operations at both Monsanto Chemical Company facilities and the Mound 

Laboratory. Possible exposures to 210Po in 1943 are assumed to be insignificant based on limited 

work in a single laboratory and reported administrative activities prior to 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. 00 PC-1 Effective Date: 03/13/2009 Page 19 of 66

Table 5-6. Reported background count rates and derived minimum detectable activities (MDAs) for 

210Po. 

Period

Default sample 

volume (ml)b

Detector 

efficiency 

%

Recovery

efficiency

(%)

Reported

background

(cpm-24 hr)

MDAsa

Spot samples

(dpm-50 ml)

24-hr samples

(dpm-24 hr)

1944c–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

a. Polonium-210 spot and 24-hr sample MDAs were first normalized assuming 85% extraction efficiency and then 

corrected for a 10% chemical recovery from 1944-1963 and 63% from 1964-1973. No aliquot size corrections were 

made for polonium. 

b. Default sample volumes are 50 ml prior to 7-13-64 and 100 ml after 7-13-64. However, sample volumes as reported 

should always be used when available. Default sample volumes should only be used if no volume is reported in primary 

or secondary records.

c. This includes the operational period of the Monsanto Chemical Company facilities at Units III and IV. 

All Mound employees working in polonium operations in T, HH, WD, H, B, 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. Two samples were usually 

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

Friday. Samples were initially treated without HNO3 digestion. Polonium in raw urine samples was 

spontaneously deposited on nickel disks and counted using a low background proportional counter. 

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

and preserved with 50 mg of sulfamic acid before analysis as reported in MLM-460 (Spoerl 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-hr electrodeposition plating time. Electrodeposition was 

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

1950 as 86.1 ±11.9% for 50-ml samples in Spoerl (1950) and 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 non-normality. Spoerl (1950) also reported that samples with 

activities in excess of 200 cpm 50 ml-1 had reduced recoveries from 56% to 64%. However, urine 

sample activity above 200 cpm 50 ml-1 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 was reported 

for this method of analysis in Recommended Bioassay for Polonium (Meyer 1992).

All urine bioassay results were reported as counts per minute in 50 ml of urine and were converted to 

disintegrations per 50 ml using an assumption of 2π geometry and the following relation:

ml

cpm

ml

dpm 2 ×= (5-6)

Chemical Recovery Corrections

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

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

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

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

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

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

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

unmetabolized polonium tracer typically ranged from 75 to 85% (Fellman et al. 1989) as previously 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 20 of 66

observed at Mound (Spoerl 1950). Mound assumed that metabolized polonium is plated with the 

same efficiency as the unmetabolized tracer. Actual chemical recoveries at Mound were therefore 

potentially only 10% rather than the assumed 86%. Reported bioassay urinalysis results would have 

to be corrected by a factor of 10 up until 1964 when complete acid digestion of urine samples was 

required (Meyer 1992). 

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

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

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

210Po dosimetry (Spitz 1989b). A peer review committee was established in 1987 to study polonium 

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

intake assessment (Spitz 1991). 

MJW conducted a subsequent evaluation of the chemical recovery of polonium (MJW 2001). The 

position paper concluded that the 10% recovery observed by Fellman (1989) was not appropriate for 

correction of Mound worker bioassay results due to differing sample storage times following collection 

but prior to analysis. The majority of Mound sample analyses were thought to have occurred within a 

4-hr timeframe post-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 reported by Spoerl

(1950) of 86%. 

However, the data reported by Fellman et al. (1989) collaborated early recoveries reported by Mound 

in 1950 for undigested urine samples and demonstrated similar estimates of urinary 210Po excretion 

for primate and human subjects using undigested urine sample procedures. Differences in sample 

storage times between the original Mound analyses and the Fellman study did not have any 

significant effect on reported recoveries. Fellman et al. (1989) 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. (1989) study is in good 

agreement with a previous Argonne National Laboratory study published in 1971 concerning seven 

individuals 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.” 

A polonium chemical recovery of 10% should therefore be used to correct 210Po data collected before 

1964. Database entries prior to 1964 should be normalized using 85% and then corrected assuming 

10% recovery as shown equation 5-7. This lower recommended recovery has been selected because 

the fundamental assumption of recovery based on deposition of tracer activity has been shown to be 

invalid for metabolized polonium present in raw undigested or unoxidized urine.

After 1964 to 1973, a correction factor of 63% should be utilized based on ashed sample results 

duplicating methods used previously at Mound with a 2-hr spontaneous plating time conducted at 

room temperature (Fellman et al. 1989). Corrections to database entries after 1964 should be 

normalized using 85% and then corrected assuming 63% recovery as shown equation 5-8.

Detection Limits and MDAs

Reductions in MPBBs at Mound resulted in development of improved analytical methods for detection 

of 210Po. The MPBB in μCi and MPC in cpm 50 ml-1 or dpm 50 ml-1 of urine resulting in a target dose 

of 300 mrem wk-1. An MPBB of 2.0 μCi was used in the early years of operations at Mound from 1944 

to 1946. This MPBB was then reduced by a factor of 10 based on recommendations from the 

National Bureau of Standards (NBS 1953, NBS 1959). The U.S. Atomic Energy Commission (AEC) 

rescinded use of the 0.2 μCi MPBB in 1964 and directed Mound to use 0.04 μCi as recommended by 

NCRP (Myers 1992). 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 21 of 66

Polonium-210 detection limits were improved after analytical modifications resulted in a four-fold 

increase in counting sensitivity and a detection limit of 0.2 dpm for a 24-hr sample. Modifications to 

the standard sample preparation procedure in 1964 included use of a larger Lucite plating cell to 

accommodate 100-ml urine samples and electrodeposition onto a single side of a larger 

stainless-steel disk (Sheehan, 1964) If warranted, 24-hr urine samples were collected to determine 

potential dose to target organs. The MD-20736 procedure was vigorously investigated and was the 

primary method of analysis from 1964 to 1973 (Mound 1968).

MDAs were assessed by evaluating reported background counting rates during the early years from 

1944 to 1963 and from 1964 to 1973 when detection limits were improved. Table 5-6 presents MDAs 

calculated using the Currie equation as shown in equation 5-2 (Currie 1968). Background was 

determined to be 0.2 cpm/24-hr urine sample in 1958 in Annual Overviews (Meyer 1992). This 

background counting rate was assumed for the early period of 210Po detection from 1944 to 1963. A 

lower background counting rate of 0.1 cpm was reported in 1964 to 1966, and from 1971 to 1973 in 

Annual Monitoring Reports, Consultations, and Annual Overviews (Meyer 1992). This counting rate

was assumed for the later period from 1964 to 1973. A two-fold reduction in the background counting 

rate is consistent with improved analytical sensitivity resulting from a two-fold increase in sample 

volume used after 1963 (Sheehan 1964). 

MDAs were derived for 50-ml spot samples using a 60 minute count time specified in procedure MD20736 and an assumed 1000 minute count time for 24-hr samples (Mound Laboratory 1968). The 

background value of 0.1 cpm/24-hr sample for the period from 1964 to 1973 results in a derived MDA 

of 0.16 dpm using equation 5-2, a detector efficiency of 50%, and recovery of 63%. Estimated MDAs 

for spot samples were considerably higher due to a much shorter count time of one hour. MDAs 

derived for 1944 to 1963 were higher primarily due to the assumed chemical recovery of 10% prior to 

proper acidification in 1964 and a higher background count rate of 0.2 cpm. MDAs in terms of daily 

excretion are also reported in dpm/d for spot samples.

Daily Excretion Equations

Sample results in activity per day are reported as Bq/d in the PURECON database under the heading 

of BQ_DAY. All Bq/d database values can be corrected by multiplying by 85% and dividing by 10% 

from 1944 to 1963 and 63% from 1964 to 1973 as shown Table 5-19. All Bq/d database values have 

already been corrected for a daily sample size of 1400 ml and counting efficiency of 50%. Urine 

samples identified as 24-hr samples would only require correction for chemical extraction.

Daily excretion in dpm/d for 210Po spot samples using the PORECON database reported in cpm 

should be calculated using equations 5-7 and 5-8 as follows:

Daily Excretion Po 210 =− Database cpm 50% 85 % 10 %×÷×÷ 1400ml /Aliquot ml (1944 − 1963 ) (5-7)

Daily Excretion Po 210 =− Database cpm 50% 85 % 63%×÷×÷ 1400 ml /Aliquot ml (1964 − 1973) (5-8)

Reported Half-Life

The effective half-life was investigated in 1950 and reported by Spoerl et al. (1950) to be 36 d. Dose 

calculations used at Mound assumed this effective half-life from 1950 to the present. Various quality 

factors ranging from 10 in the 1950s to 20 in the 1980s were used to determine dose following 

polonium exposure. ICRP (1979) reported a 50-d biological half-life. However, Sheehan (1951) 

reported a much shorter effective half-life of only 24 d for exposure resulting from a puncture wound 

involving a platinum wire contaminated with 210Po at Mound. 

Mound bioassay data seems to favor a 36-day effective half-time rather than the 50-day half-time 

used by the ICRP in Publication 67. When DRs use the default systemic parameters in IMBA, they 

should be aware that changes to systemic parameters that effectively decrease retention time might 

produce better fits to an individual worker’s urinalysis data. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 22 of 66

Polonium-210 Recommendations 

PURECON database entries in Bq/d should be used when possible. All urinalysis data for 210Po 

collected before 1964 should be normalized assuming 85% extraction efficiency and then corrected 

for a 10% chemical recovery due to plating of metabolized polonium. Polonium-210 data reported 

after 1963 should be normalized assuming 86% extraction efficiency and then corrected for a 63% 

chemical recovery based on an analysis by Fellman et al. (1989). Bioassay results at or below 

detection from 1944 through 1963 should be estimated using derived MDAs in Table 5-6. Daily 

excretion should be determined using equations 5-7 and 5-8. 

5.3.1.3 Protactinium-231

Mound conducted limited research to investigate 231Pa as a long-lived alpha emitter for possible use 

as initiators or radiothermal generator source material. ICRP lists 231Pa with a 32,700-yr half-life 

(ICRP 1995). The AEC investigated the chemistry and potential applications of 231Pa. The separation 

and processing of 231Pa from Sperry Cake raffinates was conducted in the HH Building, rooms HH5 

and HH6, from 1955 to 1956. Separation of 231Pa from Sperry Cake was also reported in R Building 

(R148, R149, R167) from 1954 to 1959, and limited laboratory use was reported from 1956 to 1987. 

In addition, minor quantities of 231Pa were present in samples subjected to spectrographic analysis 

conducted in R140 and R142 from 1956 to 1969. Minor concentrations of 231Pa were found in 

WD/WDA Building, which received aqueous wastes from R and SW Buildings. No 231Pa use was 

reported for SM, SW, T, and PP Buildings. 

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.

In Vitro Bioassay of 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-hr samples. A significant volume of 

urine was required to detect an MPC target dose of 300 mrem wk-1 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 

Protactinium-231 from natural sources was discussed by Kirby and Figgins (1959), Kirby (1959), and 

Salutsky et al. (1956). Residues were initially filtered from aqueous raffinate solutions following a 

diethyl ether extraction of uranium. Filtrates were generally dissolved in HCL saturated with NaCl, 

precipitated with a silicate or Ce carrier and NaOH or NH4OH, digested in 9 N HCl or HNO3 solutions, 

and either count the precipitate directly or isolate by anion exchange.

The Thorium Fraction results in cph were converted to cpm, corrected for a 50% counting efficiency, 

and entered directly into the ORAU database as dpm (ORAU 2003e). 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 1959). Analyses were later conducted in 

accordance with procedure MD-20736 (Mound 1968).

Table 5-7 lists reported background count rates given in Section 4, Bigler’s Annual Monitoring Reports 

by time period and MDAs derived using Currie’s equation 5-2 (Meyer 1992). 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 23 of 66

Table 5-7. Protactinium detection limits (dpm).

a

Period

Background

count rate (cpm)

Detection limit

(dpm) 

1954–8/17/55 No bioassay 0.3b

8/17/55–1956 0.3 0.3

1957–1958 0.2 0.02

1959–12/7/59 0.15 0.08

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

b. A detection limit of 0.3 dpm should be used when there are no 

bioassay data available.

Reported bioassay results in cph were generally corrected for a 78% extraction efficiency, 50% 

detector efficiency, and background counting rate. However, no corrections were made for aliquot 

volumes reported on primary records. An evaluation of the 231Pa ORAU database determined that 

some sample volumes were not reported and the modal aliquot was 120 ml. If aliquot or sample 

volumes are not reported in the ORAU database a modal aliquot of 120 ml and a standard man 

sample volume of 1400 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 (Th) ORAU 

database dpm as shown in equation 5-9 using actual sample and aliquot volumes or by assuming a 

standard man sample volume of 1400 ml and a modal aliquot volume of 120 ml (ORAU 2003e):

orAliquot ml

Sample or ml Daily Excretion Pa Th Fraction dpmbaseData 1201400 231 =− × (5-9)

Protactinium Recommendations

Protactinium-231 use did not occur in Buildings PP, SM, and T. Intakes to workers should be 

estimated only for exposures in Buildings HH, R, and SW. Activity data in dpm should only be used 

as taken from the ORAU database for the Thorium Fraction (ORAU 2003e). MDAs in Table 5-7 

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 8/17/1955. Daily excretion should be estimated 

using equation 5-9. 

5.3.1.4 Plutonium-238 and Plutonium-239

This document uses the terms plutonium-238 and plutonium-239 as synonyms for heat source 

plutonium and weapons-grade plutonium. Both heat source plutonium and weapons-grade plutonium 

were actually mixtures of several plutonium isotopes with smaller amounts of 241Am and other 

isotopes. Plutonium-239 was separated from neutron-irradiated uranium metal from 1949 to 1952. In 

addition, plutonium was used to conduct reactor fuel research from 1956 to 1959. The Neutron 

Source Group started production of heat sources in the R Building using 239Pu in 1956. The 

Metallurgical Research Group used plutonium to conduct research on plutonium metal compounds 

from 1956 to 1959. 

In 1959, the Mound facility initiated use of 238Pu. Plutonium metal production was used to 

manufacture heat sources from 1961 to 1965 in support of the SNAP Program in the SM Building. 

Uranium-234 was separated from aged 238Pu from 1970 to 1980.

Plutonium use at Mound was primarily conducted in the R, SM, T, and PP Buildings in support of 

metallurgical and chemical research, heat source production, and microsphere fabrication. Feed 

material received from the Savannah River Plant as plutonium-dioxide powder was enriched in 238Pu 

and processed at Mound to manufacture heat sources. Table 5-8 lists the percentage isotopic 

mixture by weight and activity for this material.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 24 of 66

Table 5-8. Heat source plutonium feed material 

characteristics.a

Nuclide % Mass

% Activity by effective ageb (years)

1.0 yr 3.2 yr 10 yr

Pu-236 0.0001 0.0 0.0 0.0

Pu-238 80.2 95.6 95.9 96.8

Pu-239 15.9 0.1 0.1 0.1

Pu-240 3.0 0.0 0.0 0.1

Pu-241 0.6 4.3 4.0 3.0

Pu-242 0.1 0.0 0.0 0.0

Am-241 0.0 0.0 0.0 0.1

a. Source: King (1995).

b. Effective age is the age when 241Am activity is zero.

However, 238Pu and 239Pu were both used at Mound, including research work confined to the use of 

weapons-grade 239Pu and the manufacture of initiators. Weapons-grade 239Pu used in various 

projects was reported to be over 95% pure with trace amounts of 240Pu and 241Pu (King 1995). 

Table 5-9 lists the percentage mixture by weight and activity for this material.

Table 5-9. Weapons-grade plutonium feed material 

characteristics.a

Nuclide % Mass

% Activity by effective ageb (years)

1.0 yr 3.2 yr 10 yr

Pu-236 0.0001 0.0 0.0 0.0

Pu-238 0.029 1.2 1.3 1.6

Pu-239 93.2 14.0 15.1 19.3

Pu-240 5.71 3.1 3.4 4.3

Pu-241 0.33 81.6 79.7 73.3

Pu-242 0.1 0.0 0.0 0.0

Am-241 0.02 0.1 0.4 1.5

a. Source: HPS (1985).

b. Effective age is the age when 241Am activity is zero.

In general, 238Pu compounds are more soluble than 239Pu due to greater specific activity and therefore 

a greater energetic alpha recoil for 238Pu.

A particle size of 5 μm should be assumed for plutonium and all other alpha emitters in the R Building 

at Mound in the absence of specific information on actual particle size distributions (King 1995). The 

R Building included process laboratories, research laboratories, analytical laboratories, and vault 

storage facilities that supported plutonium work. After 1969, heat source manufacturing and 

plutonium separation were associated with R Building, and 238Pu was the major radionuclide of 

concern. Plutonium compounds were subjected to spectrographic analysis in R140 and R142 with 

possible inhalation of fumes. Various incidents were reported in R Building from 1960 to 1974 

involving exposure to 238Pu. A single plutonium–beryllium exposure incident occurred in 1960.

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 (King 1995). 

Microspheres would be quite stable and their initial particle size distribution would only be affected by 

depositional processes that remove particles above 10 μm from worker breathing zones. Deposition 

would therefore tend to increase the inhalation of submicrometer particles. Reported solubility varied 

widely from type S in room SM60 to types M and F in SM27 and SM28. Use of a plasma torch 

process primarily in SM39 resulted in PuO2 fumes.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 25 of 66

Encapsulation, X-ray inspection, and calibration of plutonium–beryllium neutron sources was 

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

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

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

building was to process PuO2 powder from the Savannah River Plant. In addition, a number of 

processes were developed to manufacture plutonium microspheres, which occurred primarily in the 

PP Building. These processes involved the following:

• Plutonium–molybdenum Cermet process to coat plutonium particles from 43 to 250 μm with 

molybdenum from 1969 to 1973

• Pressed plutonium-oxide process to fabricate plutonium spheres for encapsulation as heat 

sources from 1970 to 1979

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

• Solid solution process from 1979 to 1974

• Solution gel process from 1969 to 1971

• Microsphere process to produce 50- to 250-μm 239Pu-dioxide ceramic spheres from 1967 to 

1970

The primary radiological concern at the PP Building was the possible inhalation of respirable 

plutonium particles. Plutonium microsphere processing in rooms PP162 and PP163 was reported to 

produce a product that had a solubility type S due to plasma processing and high-temperature fusing 

into oxides (King 1995). However, between 1967 and 1970, plutonium processing in rooms PP144, 

PP147, and PP148 was reported to produce class D or W materials (King 1995). Classes D and W 

were assumed to correspond to current ICRP lung clearance types F and M (ICRP 1995). A number 

of plutonium compounds were used in PP Building. Table 5-10 summarizes the number of rooms 

where plutonium compounds were used.

Table 5-10. Plutonium compound usage in the PP 

Building. 

Plutonium compound # Rooms Use period

Plutonium dioxide–PuO2 26 1967–present

Plutonium nitrate–Pu(NO3)4 15 1967–present

Plutonium oxalate–Pu(C2O4)4 13 1967–present

Plutonium hydroxide–Pu(OH)4 14 1967–present

Plutonium fluoride–PuF4 5 1967–1980

Plutonium carbide–PuC 4 1967–1980

Many of the reported plutonium compounds are highly soluble and are ICRP type M. Plutonium 

nitrate is considered type M, and plutonium ethylene diamine tetraacetic acid (EDTA) complexes 

should be considered type F due to rapid assimilation in the body (ICRP 1986). Plutonium-238 

bioassay data was evaluated in 1996 and determined that the likely lung solubility was a mixture of 

50% class W and 50% class Y (Siebert 1996). This mixture was assumed to correspond to ICRP 

Publication 68 solubility classes 50% M and 50% S (ICRP 1995).

In Vitro Bioassay of Plutonium-238/239

Bioassay of the plutonium isotopes at Mound underwent considerable change and improvement with 

the development of specific methodologies to isolate and count 238Pu and 239Pu. In 1952, an attempt 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 26 of 66

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% should be assumed for plutonium bioassay data collected from 1954 to 1960, based 

on the analysis of reduced polonium recoveries discussed in Section 5.3.1.2.

Starting in 1961, urine samples were acidified to a pH of 2 with nitric acid, resulting 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 (Mound 1968). This 

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

quantify detection limits, and alpha spectroscopy (Mound 1981). 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 

and thorium alpha activity. The gross alpha results were reported as 239Pu for workers in R Building 

and 238Pu for SM and PP Building workers until 1980. Table 5-11 summarizes the bioassay history for 

plutonium.

Table 5-11. Summary of historical plutonium analytical procedures. 

Period Reporting levela Analytical method

1954–1960 Kirby and Brodbeck 1954

0.05 dpm 1.5 L-1 

0.5 dpm 1.5 L-1(Corrected 

DL)b

Alkaline earth phosphate coprecipitation; plutonium separation 

with cerium carrier; gross alpha for plutonium; proportional 

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

metabolized Pu.

1961–1967 Kirby and Brodbeck 1954

0.1 dpm 1.5 L-1

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 (Mound 1963)

0.1 dpm 1.5 L-1

Alkaline earth phosphate coprecipitation; plutonium separation 

with cerium carrier; anion exchange; electrodeposition; gross 

alpha reported for plutonium; proportional counting. 

1978–1979 MD-80030 (Mound 1981)

0.1 dpm 1.5 L-1

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 (Mound 1981)

0.07 dpm 1.5 L-1 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-hr 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 will vary.

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

The precision of 239Pu analysis was reported in 1981 as 6.7% at a concentration of 3.2 dpm kg-1 in 

urine with a large counting error of 73% in Section 3, Bioassay Data Quality Assessment/Control 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 27 of 66

(Meyer 1992). The ratio of measured to standard results was 1.06 ± 0.08, which indicates high 

accuracy. 

Table 5-12 summarizes the MDAs for plutonium isotopes in 24-hr bioassay urine samples over time 

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

for accreditation of bioassay laboratories was 0.06 pCi L-1 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-min counts, and a lower recovery of well below 85% was reported for 1,000-min counts (Mound 

1996). Approximately 91% of all samples were counted for 1,000 min. 

Daily excretion of the plutonium isotopes should be estimated using the following equations 5-10 and 

5-11.

Daily Excretion Pu−238,239 = Pu Database dpm 85 % ÷× 10 (% 1957 − 1960 ) (5-10)

Daily Excretion Pu 238,239 =− Pu Database dpm ( −Pr1961 esent ) (5-11)

Table 5-12. Plutonium for reported periods (pCi/sample 

of urine).

a,b

Period Pu-238 MDAs Pu-239/240 MDAs

Before 1957 (0.85) (0.85)

1957 (0.94) (0.94)

1958 (0.6) (0.6)

1959 (0.5) (0.5)

1960 (0.5) (0.5)

1961–1966 0.05 0.05

1967–May 1981 0.11 0.11

Jun–Dec 1981 0.066, 0.027 0.044, 0.020

1982 0.063, 0.023 0.042, 0.017

1983 0.064, 0.024 0.043, 0.018

1984 0.066, 0.026 0.045, 0.019

1985 0.047, 0.017 0.031, 0.012

1986 0.042, 0.015 0.028, 0.011

1987 0.022 0.011

1988 0.017 0.01

Jan–Jul 1989 0.019 0.008

Aug 1989-Dec 1991 0.018 0.008

1992-1995 0.018a 0.009a

a. Plutonium MDAs from before 1957 to 1960 were normalized

assuming 85% extraction efficiency and then corrected for 10% 

chemical recovery due to plating. Based on Mound (1996).

b. All table values based on 1000-minute counts followed by 3000-

minutes counts from 1981–1986.

Plutonium Recommendations

A default AMAD of 5 μm should be used for the plutonium isotopes. However, possible exposure to 

submicrometer materials during plasma torch operations, microsphere processing, and plutonium 

metal production should be considered. Plutonium nitrate and plutonium EDTA complexes should be 

considered soluble type M (ICRP 1986). Exposures during final microsphere production should be 

considered type S. All other plutonium compounds should be considered a mixture of 50% class M 

and 50% class S.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 28 of 66

All 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 prior to sample analysis (Sheehan, Kirby, and Wood 1963; Spoerl 1950). 

MDAs in Table 5-12 should be used to estimate missing or undetected bioassay results unless 

reporting levels in Table 5-11 are greater than the MDA. 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 5-10 and 5-11 as appropriate.

5.3.2

Several hundred isotopes were present in Mound facilities because the compositions of irradiated 

materials are complex. Of these, the radionuclides of secondary interest at Mound are 230Th, 232Th, 

226Ra, 227Ac, 241Am, and 233U through 238U. With the exception of 228Ra and 226Ra, most of these 

isotopes have a low abundance of low-energy gamma emissions. As a group, these isotopes have 

restrictive annual limits of intake, and with the exception of soluble forms of uranium, have slow 

biological clearance by urinary excretion. These factors make it difficult to attain useful in vitro

detection limits by urine bioassay. 

Secondary Radionuclides

Attachment A, Table A-1, lists lung clearance type assumptions for radionuclides used in site process 

activities, and Table 5-3 lists assumptions about the degree of equilibrium. Attachment B, Table B-1, 

lists the room numbers where secondary radionuclides were used.

5.3.2.1 Radon Isotopes

Source terms for radon (222Rn), thoron (220Rn), and actinon (219Rn) were present at Mound due to 

radium and thorium processing. Exposure to these radioactive gases should be included in dose 

reconstruction. 

Source terms for radon (222Rn), thoron (220Rn), and actinon (219Rn) were present at Mound due to 

radium and thorium processing and separation of 231Pa and 227Ac. 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 5-13 shows the results of various radon surveys at Mound.

Table 5-13. Radon isotope survey data by building and period. 

Building

Survey 

dates

Gas concentration (pCi/l)

WL WLM yr-1 (a) Radon Thoron Actinon

SW Tunnel 10/12/79 88,000 28,000 640,000

SW19 before vent 6/79–10/79 67–160

SW19 after vent 3/80–4/80 7.7–13.4 0.03 0.4

12/89 2.6 Not 

detected

12/99 1.0–5.8b

4/00 0.7–7.2

1998–2000 5.8–12

SW21 5/83 17.5–52.8

12/89 34–118 0.3–0.7 0.4–1.3

0.85 

median

4.7–15.3

4/90–6/90 117–161 10.0 median

125

SW22 6/90 125

SW48 12/89 3.2

Old SD 12/89 1.5–2.4
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 29 of 66

Fire station 12/89 1.2–1.4

Paint shop 12/89 0.9–1.2

A 10/88–2/89 0.1–1.1

All other buildings 12/89 0.4–0.9

Outside facilities 5/83 0.24–0.59

Reported background 0.1 to 2.1

Average = 0.5

a. Working level months per year (WLM yr-1) were determined as WL (2000 hr yr-1) (1/ 170 hr WLM-1). 

b. Bolded concentrations were derived from charcoal canister measurements.

The SW19 Building had elevated radon concentrations emanating from radium and thorium 

processing wastes in the “Old Cave.” High concentrations of radon, thoron, and actinon were 

transported from an underground tunnel into SW19 work areas. Measurements of 222Rn near an 

employee’s desk ranged from 67 to 160 pCi L-1 in 1979. On October 12, 1979, inert gas 

concentrations in the tunnel were measured at 88,000 pCi L-1 222Rn, 28,000 pCi L-1 220Rn, and 

640,000 pCi L-1 219Rn. A 222Rn concentration of 88,000 pCi L-1 would correspond to 352 WL, 

assuming an ICRP 50 indoor mean radon daughter equilibrium factor of 0.45 (ICRP 1987). A 

ventilation system was installed in early 1980 to vent SW17 tunnel air to an SW18 exhaust. 

This system reduced 222Rn concentrations in SW19 by a factor of 10 or more to 8.2 pCi L-1 in March 

1980. This result was confirmed with a Passive Environmental Radon Measurement (PERM) 

measurement of 15.4 pCi L-1 and RDT310A continuous measurements from 7.7 pCi L-1 to 13.4 pCi 

L-1. A working level (WL) measurement by the employee’s desk was 0.03 WL or 0.03 (12 mo yr-1)=0.4 

working level month (WLM) compared with an occupational limit of 4 WLM yr-1. 

Radon surveys in Building 21 were conducted with PERMs in 1983 and ranged from 17.5 to 

52.8 pCi L-1 with WLs from 0.40 to 1.3. WLMs exceeded occupational limits of 4 WLM with calculated 

values from 4.8 to 15.6 WLM. Radon measured during the radon study with charcoal canisters 

ranged from 34 to 118 pCi L-1 in 1989. 

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-1 222Rn, except SW and Old SD buildings. In June 2000 a radon study summary report was 

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

reported to be 0.5 pCi L-1 with a range of 0.1 to 2.1 pCi L-1. SW-19 was the only building at Mound 

identified as an area of potential occupational exposure to 222Rn and 220Rn. 

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

in the underground tunnel. The only actinon measurements were made in Building 21 at 0.3 to 

0.7 pCi L-1, but the holding time before analysis was excessive compared to the half-life of 219Rn in 

seconds. The only thoron measurement in SW19 was below detection. Thoron and actinon 

exposures would not be detectable in excreta by bioassay due to very short half-lives. 

Radon Isotope Recommendations

The recommended WLM values in Table 5-14 should be used to reconstruct doses following radon 

exposures. All WLMs should be assumed to be median values of lognormal distributions with a GSD 

of 3.0.

Table 5-14. Radon-222 WLM assumptions for dose reconstruction. 

Building conditions Period

Radon 

(pCi/L)

Radon 

(WL)

Radon 

(WLM)

SW Tunnel prior to ventilation 1949–1979 88,000 352 4,100

SW-19 Ra-Ac pilot project 1949–1952 160 0.72 8.5

SW-19 Ra-Ac Old Cave production 1952–1960 160 0.72 8.5
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 30 of 66

SW-19 post production, prior to ventilation 1961–1979 160 0.72 8.5

SW-19 post production, post to ventilation 1980–1995 0.7–13.4 0.03 0.4

SW-21 1952–1995 17.5–125 0.85 10.0

SW-22 1966–1995 125 0.56 6.6

SW-48 1980–1995 3.2 0.01 0.2

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

All other buildings 1949–1995 0.9 0.004 0.05

Background–indoors 1949–1995 0.5 0.002 0.03

Exposure to radioactive gases in the tunnel itself should assume concentrations of 88,000 pCi L-1

222Rn, 28,000 pCi L-1 220Rn, and 640,000 pCi L-1 219Rn. A WL of 352 and WLM per year of 4100 

should be used to estimate dose to workers only during documented exposure inside the tunnel. 

To reconstruct missed doses a background radon concentration of 0.5 pCi L-1 should be used, which 

converts to 0.03 WLM indoors assuming 100% occupancy and 0.45 (ICRP 1987).

5.3.2.2 Thorium-230

Attachment B, Table B-1, lists the buildings and rooms where 230Th exposures could have occurred. 

Most of these exposures would have occurred in conjunction with exposures to other thorium isotopes 

or 231Pa. Small research activities involving the extraction of 230Th (ionium) from raffinate materials 

occurred in R, HH, and SW Buildings. The program began in early 1956 and ended in 1958, with the 

last bioassay sample processed on September 9, 1958. This activity was evidently related to an 

interest in 230Th as a tracer and as a target material for the production of the isotopes 231Pa and 

232,233U by neutron irradiation of 230,232Th (Figgins and Kirby 1966). Exposure to 230Th waste materials 

also would have occurred in Building WD/WDA. 

Important feed materials included raffinates and their precipitates such as Cotter Cake, which is also 

known as Airport Cake. Cotter Cake appears to be derived from a raffinate that had been neutralized 

by caustic or lime after ore material was dissolved in nitric or hydrofluoric acid and extracted with 

diethyl ether. 

Thorium-230 Records

Primary 230Th bioassay records consisted of a logbook, and apparently duplicate records in a brown 

notebook. Count data were typically recorded on Form O-318 followed by an “I” or “Io”. Secondary 

230Th results started as weekly reports on March 17, 1958. Weekly reports included Name, Isotope, 

and Result. Prior to 1958, secondary 230Th results were reported on Form O-634 including Name, 

Badge Number, Date, Type of Analysis, Isotope, and Result. However, some secondary documents 

have problems with reporting units. Some results are reported to be cph per 24-hour sample when 

primary records indicate that they are actually cpm per aliquot. The ORAU database should therefore 

be considered a secondary record extracted from primary records.

Bioassay Procedures and Detection Limits

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-hr 

urine specimens. The objective of the urine bioassay sampling was to detect an MPC that 

corresponded to a 300-mrem wk-1 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-hr urine sample corresponded to a dose equivalent 

of 300 mrem wk-1. Based on Meyer (1992, Section 2), three different urine bioassay procedures were 

tried for thorium. A thenoyl triflouro acetone (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 

given in MLM-1003 dated August, 1954 (Kirby and Brodbeck 1954, p. 14). 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 31 of 66

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 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 provided by Potter (2002). 

Excretion curves for acute and chronic exposures can also be readily calculated using the IMBA 

software (Birchall et al. 2003). ICRP 68 assigns radium to absorption type M; however, thorium can 

be either type M or S. 

When reconstructing a particular dose history, careful attention must be paid to the reporting basis 

that Mound used because several different bioassay strategies were attempted for 230Th. If the 

reporting basis for an Ionium bioassay result is ambiguous, assume that 230Th was extracted and 

counted. Excretion rates in dpm/day should be calculated using equation 5-12 and entries from the 

ORAU database that contain information on volume, aliquot size, and dpm results, unless 

contradicted by another bioassay record (ORAU 2003e):

Aliquot ml

mlSample Daily ExcretionTh 230 =− Database dpm × (5-12)

If no aliquot volume is given in the ORAU database 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 detection limits. Where no aliquot size is given in the ORAU database, 

assume the detection limit is in dpm excreted per 24-hr sample (ORAU 2003e).

Table 5-15. Thorium-230 detection limits (dpm). 

Period Detection limit Reference

Prior to 1/1994 0.3 MDA (MJW 2002)a

January 1994–present 0.17 MDA (Barton 1995)

a. Detection limit recommended for all secondary radionuclides (MJW 

2002). Correction for aliquot size should be conducted if available.

5.3.2.3 Natural Thorium

Monitoring for natural thorium began in 1955 to support a thorium refining; preliminary research began 

in the R Building. Exposures to 232Th and daughters 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. 

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 thorium dioxide. Material 

received by Mound was sometimes basic and sometimes acidic.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 32 of 66

Thorium-232 Records

Thorium-232 records are diverse due to programs conducted for many years. Primary 232Th bioassay 

data were entered into a small brown spiral notebook marked “Radium-Thorium” and “RadiumThorium 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 

were removed from 7/6/59 to 1/9/61, 7/13/64 to 11/15/64 and 5/30/65 to 6/6/65. These record books 

apparently do not contain true primary data, but calculated results such as cpm excreted per day. 

Secondary records in weekly reports contained 232Th results as cpm/24-hr samples beginning March 

17, 1958. In August 1959, secondary results were reported on form O-756. The ORAU database is a 

record of secondary 232Th bioassay data extracted from other primary records (ORAU 2003e). Most 

232Th records in the ORAU database have both a thorium and radium entry in dpm (ORAU 2003e). 

The corresponding primary 232Th records from the 1950s carry a sample number that ends in “T”. 

Assume that the “T” result is from the natural thorium fraction and “R” indicates that results are from 

the radium fraction for a given sample. 

Bioassay Procedures and Detection Limits

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

MLM-460 (Spoerl 1950) and MD-20736 (Mound 1968). Meyer (1992, Section 1) indicates that 

2.8 cpm on a 24-hr urine sample corresponded to an intake that would result in a 300-mrem wk-1 dose 

equivalent to the critical organ under the dosimetry system in effect when the sample was collected. 

Meyer (1992), Section 2 recommends correcting bioassay data by converting to counts per minute 

and then dividing by 1.11 to obtain units of picocuries per 24-hr period. 

Where no aliquot volume is given in the ORAU database, assume that the dpm value has been 

corrected for background, detection efficiency, and represents a 24-hr urine excretion. If an aliquot 

volume is given in the ORAU database, then assume that the bioassay result is in units of dpm per 

aliquot. No other corrections are required for urinalysis results in dpm. If urinalysis results are in cpm 

then 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 thorium-232 detection limits by period. Detection limits in Table 5-14 have units of 

dpm excreted per 24-hr when no aliquot size is given in the ORAU database. However, detection limit 

units should be dpm per aliquot, if an aliquot volume was reported in the database. When interpreting 

gross alpha thorium bioassay data for thorium sludge activities, dose reconstructors should use the 

degree of equilibrium assumptions in Table 5-3 and the lung solubility assumptions in Attachment A, 

Table A-1, unless additional information is available. 

Table 5-16. Natural thorium detection limits (dpm). 

Period Detection limit Reference

Prior to 12/1987 0.3 MDA (MJW 2002)a

December 1987–1994 0.02 (Th-232) MDA (Barton 1995)

January 1994–Present 0.08 (Th-228) MDA (Barton 1995)

a. Correction for aliquot size should be conducted if available. Most

database entries do not have reported aliquot volumes. 

Thorium-232 bioassay results in the ORAU database contain a result for radium (Ra) and/or for 

thorium (Th) (ORAU 2003e). If a radium result is provided for a 232Th urine bioassay, the primary 

record sample ID usually ends in “R” and refers to Ra-224. From these results, the daily excretion of 

Ra-224 can be obtained using equation 5-13 as follows:

Daily Excretion Ra 224 =− Ra Database dpm ÷ 90 % (5-13)
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 33 of 66

A primary record ID that ends in “T” corresponds to a natural thorium (Th) result that includes 232Th 

and 228Th. The daily excretion of natural thorium in the ORAU database can be calculated using 

equation 5-14 as follows: 

Daily Excretion Natural Thorium = Th Database dpm ÷ 92 % (5-14)

Equations 5-13 and 5-14 include correction factors of 90% and 92% for radiochemical yield 

respectively, because database results for radium and thorium have not corrected for chemical 

extraction.

Local background thorium urinary excretion rate data is not available. However, ICRP Publication 23 

(ICRP 1975) reports that background excretion rates for natural thorium range from 0.05 to 1.0 

dpm/day, and that Reference Man excretes 0.05 dpm/day. For dose reconstruction, assume that 

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

pCi/day (ICRP 1975).

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

Mound completed a facility in the SW Building in 1952 for processing neutron-irradiated 226Ra. The 

principal activation products of interest were 227Ac and 228Th. Actinium-227 and 228Th decay to 223Ra

and 224Ra, respectively. Work continued on irradiated 226Ra until 1960. Attachment A, Table A-1, and 

Table 5-2 contain additional details of when and where these radionuclides were used. 

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

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

forms O-318. Notebooks are labeled RA#1a and RA#1c. Secondary urinalysis results were written 

on hand drawn forms and placed in the “J. B. Black binder.” Each form had an employee name, 

sample ID, and the 223Ra, 224Ra, 226Ra, and 227Ac results.

In addition, manila folders contained primary bioassay data with counts collected, assumptions made, 

equations used, theoretical curve computations, and a written analysis of the observed data (Meyers 

1992). Much of this information has already been included in the DOE response to claims. 

Secondary data appeared in various forms such as weekly reports, and on forms O-756 and O-634. 

The ORAU database contains available secondary bioassay records for the Radium Irradiation 

project, and excerpts are typically found in the DOE response to claims. 

The ORAU radium-actinium database provides results for 223Ra, 224Ra, 226Ra, and 227Ac in units of cph 

for 24-hour urine samples (ORAU 2003d). No aliquot volumes were given in this database.

Bioassay Procedures and Detection Limits

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 (Mound 1968). 

Radium was extracted from collected 24-hr urine bioassay samples. Intakes of 227Ac and 228Th should 

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

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

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

periods of up to 30 or 35 d. The resultant count data for a single sample was manipulated to form a 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 34 of 66

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. After the 223Ra and 

224Ra excretion rates are determined, it is necessary to estimate the corresponding excretion rates 

that would have occurred from equal intakes of 227Ac and 228Th. 

To convert cph urinalysis data a counting efficiency of 50% and 90% extraction efficiency must be 

applied as documented by Meyer (1992). Assume that all secondary cph records are based on 24-hr 

samples.

MDAs were not evaluated for 223Ra, 224Ra, and 226Ra because observed gross alpha counts were 

partitioned into activities attributable to individual radionuclides using complex equations, 

assumptions, and corrections for ingrowth would be required. However, evaluation of selected 

records indicated that a reasonable gross alpha background count rate is 4 cph. The background 

gross alpha count rate can then be used to derive an MDA of 0.09 dpm for missed worker doses by 

assuming a 50% detector efficiency, 1000 minute count time, and a 90% extraction efficiency in 

equation 5-2. The 0.09 dpm of total alpha activity should be claimant-favorably assigned to 226Ra.

The ORAU radium-actinium database resulted from the radium irradiation project (ORAU 2003d). 

Assume that intakes of Ac-227 and Th-228 were in equilibrium with their radium progeny. Urine 

bioassay results are in units of cph for Ra-223, Ra-224, and Ra-226. Actinium-227 intakes must be 

derived from Ra-223 bioassay result, 228Th intakes must be surmised from Ra-224 bioassay results, 

and Ra-226 daily excretion is calculated directly using equations 5-15, 5-16, and 5-17 as follows: 

cph

dpm Daily Excretion Acof Ra Database cph 602 227 −=− 223 90 % ×÷ (5-15)

cph

dpm Daily Excretion of Th Ra Database cph 602 228 −=− 224 90 % ×÷ (5-16)

cph

dpm Daily Excretion of Ra Ra Database cph 602 226 −=− 226 90 % ×÷ (5-17)

5.3.2.5 Uranium Isotopes

A small-scale research project involving 233U was conducted from 1958 to 1960. Uranium-233 and 

other uranium isotopes were analyzed by chemical separation and gross alpha counting using Mound 

procedure MD-20736 (Mound 1968). It appears that 233U activities stopped in 1960, but this 

procedure continued to be used for other uranium isotopes. 

Uranium Records

Primary records for the uranium isotopes were found in individual claims on forms such as O-318. 

Information from these records appears to be summarized in the ORAU database (ORAU 2003e). 

Results for spot uranium samples are in units of dpm per sample aliquot and have been corrected for 

background, carrier count, extraction efficiency of 80%, and 50% detector efficiency. 

Bioassay Procedures and Detection Limits

Both spot and 24-hr 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-hr sample result. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 35 of 66

Sample results from the ORAU database in cph need to be corrected for aliquot size. The following 

equation should be used to obtain excretion rates in units of dpm for 24-hr samples using equation 5-

18 and from data presented in the ORAU database (ORAU 2003e): 

Aliquot ml

ml Daily Excretion Uranium Isotopes dpmbaseData 1400 = × (5-18)

Site-specific values for background uranium dietary intake at Mound were not available for this 

analysis. Typical intakes of uranium attributable to background should be on the order of 4.9 Bq yr-1

238U, 4.9 Bq yr-1 234U, and 0.2 Bq yr-1 235U (United Nations 1993) and 12 Bq yr-1 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/day in the urine and 2.1 to 2.7 dpm/day in the 

feces. For dose reconstruction, assume that uranium in the diet of monitored workers resulted in the 

excretion of 0.07 dpm/day (urine) and 2.1 dpm/day (feces). 

Table 5-17 lists detection limits by period for the uranium isotopes, including a DL for fluorometric 

results if available. 

Table 5-17. Uranium isotope detection limits (dpm). 

Period Detection limit Reference

Early 1950s–1985 0.4 (U-nat) μg L-1 ORAU 2003ba

Prior to 5/1985 0.3 MDA (MJW 2002)b

May 1985–present 0.07 (U-234)

0.03 (U-235)

0.09 (U-238)

MDA (Barton 1995)

a. Use for fluorometric results if reported. Statistical basis for this MDA 

in urine is unknown.

b. Correction for aliquot size should be made for the DL if available.

Attachment A, Table A-1, lists assumptions for lung clearance properties, and Table 5-3 lists isotopic 

equilibrium assumptions.

5.3.2.6 Americium-241 and Curium-244

Americium-241 is a ubiquitous constituent of heat source plutonium (238Pu) and weapons-grade 

plutonium (239Pu). 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 5-8.

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. 

Americium-241 and Curium-244 Records

Documentation for 241Am and 244Cm is very limited. Primary records for 241Am and 244Cm are 

unknown. Secondary data for 241Am and 244Cm are found in the ORAU database (ORAU 2003e). 

The ORAU database indicates that 244Cm results are related to the plutonium logbooks. All 241Am and 

244Cm urinalysis data are in units of activity (pCi). The MJW Report (MJW 2002) 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 presented in the database.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 36 of 66

Bioassay Procedures and Detection Limits

Assume that all data in the ORAU database are in units of pCi excreted per 24 hours and that no 

additional corrections are required. The DL of 0.3 recommended by MJW (2002) is consistent with 

the DL of 0.2 for the period from the 1950s to 1989. The 0.3 DL should therefore be used as the more 

claimant-favorable estimate of detection from the 1950s to 7/1989. Table 5-18 lists detection limits by 

period for 241Am and 244Cm.

Table 5-18. Americium-241 and curium-244 detection 

limits (dpm excreted per 24 hours). 

Period

Detection limit

(dpm/d) Reference

Early 1950s–1989 0.2 ORAU 2003ba

Prior to 7/1989 0.3 MDA (MJW 2002)b

July 1989–present 0.05 MDA (Barton 1995)

a. Statistical basis is unknown and therefore recommend using 0.3 

dpm/d as reported by MJW (2002).

b. Correction for aliquot size should be made for the DL if available.

Daily excretion should be determined using equation 5-19.

Daily Excretion −241orAm Cm 244 =− pCibaseData (5-19)

5.3.2.7 Americium-241 and Curium-244 Recommendations

Exposure to 241Am in heat source plutonium should be considered negligible. Intakes of 241Am 

resulting from exposure to weapons-grade plutonium should be estimated using the percent activities 

given in Table 5-8 of the TBD as 0.1% at one year, 0.4% at 3.2 years, and 1.5% at 10 years of ingrowth. Assume that the bioassay data provided in the ORAU database were for personnel who were 

working with 241Am and 244Cm source materials (ORAU 2003e).

All data in the ORAU database are in units of pCi excreted per 24 hours and no additional corrections 

are required. The 0.3 DL should be used as the more claimant-favorable estimate of detection from 

the 1950s to 7/1989. 

5.3.3

A summary of corrections for all radionuclides is provided below in Table 5-19. The primary data 

types are listed and those corrections required to estimate intake. In addition, equations are provided 

to calculate daily excretion. Most radionuclides are in units of dpm and do not require detector or 

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

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

aliquot corrections when an aliquot volume is available and MDAs should also be corrected when 

necessary. If no aliquot volume is reported, MDAs should be assumed to be dpm/24-hr sample. 

Summary of Correction Factors and Excretion Equations

Thorium-232 will require aliquot corrections in only a few instances. If 230Th, 232Th, or uranium 

urinalysis data are ever reported in cpm or cph, rather than the standard dpm units, correction factors 

for detector and extraction efficiencies are presented below in Table 5-19. 

Extraction efficiencies should be used to correct data reported in cpm or cph. Protactinium-231, 230Th, 

and uranium data have been corrected for chemical recovery in the ORAU database (ORAU 2003e). 

However, if any data for these radionuclides are reported in cpm or cph, extraction efficiencies of 

78%, 90% and 80% should be used for 231Pa, 230Th, and uranium respectively. PORECON database 

entries for 210Po must first be normalized assuming 86% extraction efficiency and then corrected for a 

chemical recovery 10% prior to 1964 and 63% after 1964. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 37 of 66

5.3.4

This section presents information required to ensure proper evaluation of bioassay data as well as a 

means to assign missed dose. Table 5-20 summarizes radionuclide analytical methods and MDAs by 

time period for urinalysis results. 

Summary of Analytical Methods and Detection Limits

5.4 SECONDARY BIOASSAY DETECTION DATA

This section discusses secondary bioassay data to support estimation of radionuclide intakes and 

dose calculations, including fecal analyses, in vivo chest counting, and air sampling. These sources 

of bioassay information were considered secondary due to limited availability and likely uncertainty. 

Air-sampling data must be demonstrated to be representative of breathing zone concentrations to be 

used effectively in dose reconstruction.

Table 5-19. Summary of correction factors and excretion equations. 

Nuclide Period

Primary

data type

Aliquot

correction

Detector

efficiencya

Extraction

efficiencyb

Po-210c

1944–1963 cpm or Bq/d no no 10

1964–1973 cpm or Bq/d no no 63

Daily Excretion Po 210 =− Po Database Bq / d 86 % ÷× 10 (% 1944 − 1963 )

Daily Excretion Po 210 =− Po Database Bq / d 86 % ÷× 63 (% 1964 − 1973)

Pu-238

Pu-239

1957–1960 dpm no no 10

1961–1966 dpm no no no

1967–5/81 dpm no no no

6/81–1995 dpm no no no

Daily Excretion Pu = Pu Database dpm 85 % ÷× 10 (% 1957 − 1960 )

Daily Excretion = PuPu Database dpm ( −Pr1961 esent )

Pa-231

8/17/55–1956 dpm yes no no

1957–1958 dpm yes no no

1959–12/759 dpm yes no no

Aliquot ml

Sample ml Daily Excretion Pa 231 =− Thorium eData dpmbas ×

Th-230

Prior 1/94 dpm yes no no 1/94–1995

Aliquot ml

mlSample Daily ExcretionTh 230 =− Database dpm ×

Th-Natural

Prior 12/87

12/87–1994 dpm yes no 92

1/94–1995

Daily Excretion Ra 224 =− Ra Database dpm ÷ 90 %

Daily Excretion Natural Thorium = Th Database dpm ÷ 92 %

Ra-226

Ac-227

Th-228

1952–1960 cph no 50 90

cph

dpm Daily Excretion of Ra Ra Database cph 60

2 226 −=− 226 90 % ×÷

cph

dpm Daily Excretion Acof Ra Database cph 602 227 −=− 223 90 % ×÷

cph

dpm Daily Excretion of Th Ra Database cph 602 228 −=− 224 90 % ×÷
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 38 of 66

Nuclide Period

Primary

data type

Aliquot

correction

Detector

efficiencya

Extraction

efficiencyb

Uranium

1950s–1985

Prior 1985 dpm yes no no

5/85–1995

Aliquot ml

ml Daily Excretion Uranium Isotopes dpmbaseData 1400 = ×

Am-241

Cm-244

1950–1989

Prior 7/89 pCi no no no

7/89–1995

Daily Excretion −241orAm Cm 244 =− pCibaseData

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

b. Assume that intake of radium-223 and radium-224 were in equilibrium with actinium-227 and thorium228 respectively. 

c. Polonium-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. 00 PC-1 Effective Date: 03/13/2009 Page 39 of 66

Table 5-20. Analytical methods and detection limits by time period. 

Nuclide Method/description Period MDA

H-3 

(hot gas)

No urinalysis conducted 1948–1956 Use 95% Dose 340 mrem

Vibrating reed electrometer 1957–1961 1.0 μCi L-1

Liquid scintillation 1962–1995 0.01 μCi L-1

Po-210

(Postum) 

Spontaneous electrodeposition from raw urine, low 

background proportional counting.

1944–1963 Spot samples

6.0a dpm

168 dpm/d

24-hr samples

1.3a 

dpm-24 hr

Spontaneous electrodeposition from acidified urine, 

low background proportional counting.

1964–1973 Spot samples

0.72a dpm

20 dpm/d

24-hr samples

0.15a

 

dpm-24 hr

Pa-231 Radium extraction and activity inferred from 227Th 

or 223Ra dpm assuming secular equilibrium.

8/17/55–1956 0.3 dpm

1957–1958 0.02 dpm

1959–12/7/59 0.08 dpm

Pu-238, 

Pu-239

Alkaline earth phosphate coprecipitation, Pu 

separation with cerium carrier gross alpha 

proportional counting. 

Before 1957 (0.85) pCi/sampleb

1957 (0.94) pCi/sampleb

1958 (0.6) pCi/sampleb

1959 (0.5) pCi/sampleb

1960 (0.5) pCi/sampleb

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

Above with anion exchange, electrodeposition, 

gross alpha proportional counting.

1967–May 1981 0.11 pCi/sample

Above with alpha spectroscopy for Pu-238 and 

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

detection limits quantified.

Jun–Dec 1981 0.066, 0.044 pCi/sample

1982 0.063, 0.042 pCi/sample

1983 0.064, 0.043 pCi/sample

1984 0.066, 0.045 pCi/sample

1985 0.047, 0.031 pCi/sample

1986 0.042, 0.028 pCi/sample

1987 0.022, 0.011 pCi/sample

1988 0.017, 0.01 pCi/sample 

Jan–Jul 1989 0.019, 0.008 pCi/sample

Aug 1989–Dec 1991 0.018, 0.008 pCi/sample

1992–present 0.018, 0.009 pCi/sample

Th-230

(Ionium)

Radium separation, intakes and excreted Ra-226 

and Th-230 assumed to be in equilibrium.

Prior 1/94 0.3 dpm

Th-230 extraction, alpha spectroscopy. 1/94–1995 0.17 dpm

Th-natural Thorium separation, proportional counting. Prior 12/87 0.3 dpm

12/87–1994 0.02 (Th-232) dpm

Thorium separation, alpha spectroscopy. 1/94–1995 0.08 (Th-228) dpm

Uranium LiF fusion, fluoroscopy. 1950s–1985 0.4 ug/l

Uranium separation, proportional counting. Prior 1985 0.3 dpm

Uranium separation, alpha spectroscopy. 5/85–1995 0.07 (U-234) dpm

0.03 (U-235) dpm

0.09 (U-238) dpm

Am-241, 

Cm-244

Americium or curium separation, proportional 

counting.

1950–1989 0.2 pCi

Prior 7/89 0.3 pCi

Americium or curium separation, alpha 

spectroscopy.

7/89–1995 0.05 pCi

a. Polonium-210 spot and 24-hr sample MDAs were first normalized assuming 86% extraction efficiency and then corrected for a 10%

chemical recovery from 1944-1963 and 63% from 1964-1973. No aliquot size corrections were made for polonium.

b. Plutonium MDAs from before 1957 to 1960 were normalized assuming 85% extraction efficiency and then corrected for 10% chemical 

recovery due to plating.

5.4.1

Average percent recoveries for plutonium, uranium, thorium, and americium for fecal samples were 

determined from an analysis of fecal sample data collected from July to October 1989 ORAU files 

(ORAU 2003c). Table 5-21 summarizes these chemical recoveries as determined using Plutonium242 (242Pu), Uranium-232 (232U), Thorium-229 (229Th), and Americium-243 (243Am) as tracers. Specific 

MDAs for these fecal analyses could not be quantified without reported background sample counts. 

However, the MDAs appeared to range from 0.01 to 0.04 pCi/sample for 238Pu and from 0.001 to 

Fecal Sample Analysis
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 40 of 66

0.008 for 239Pu, assuming that reported error terms associated with the sample activities were 1 

sigma.

Table 5-21. Radionuclide chemical recoveries for 

collected fecal samples. 

Nuclide Samples Average Minimum Maximum

Pu-242 27 0.77 0.32 1.02

U-232 9 0.69 0.13 1.05

Th-229 2 0.37 0.74 1.03

Am-243 2 0.77 0.71 0.82

5.4.2

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

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

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

Whole Body Counting 

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

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

5-in.-diameter thallium-activated Phoswich scintillation detectors were employed to detect 

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

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

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

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

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

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

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

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

hard-copy record. 

Reported 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 (File Records-2 1990). Personnel records from 1994 reported 

MDAs associated with chest counting of selected individuals and 171 counts (File Records-3 1994). 

Table 5-22 summarizes an evaluation of these reported MDAs by radionuclide and current MDAs. 

Table 5-22. Reported 1994 chest counting MDAs (nCi).

a

Radionuclide

MDA range

(nCi)

MDA mean ± SD

(nCi)

Current MDAsb

(nCi)

Am-241 < 0.92 0.3 ±0.1 0.10

Pu-240 --c -- 47d

Pu-239 9–4910 568 ±842 130d

Pu-238 0.6–1870 214 ±318 58

Th-234 0–4.35 1.8 ±0.8 --

Th-232 -- -- 31

Th-228 -- -- 3.2

U-238 -- -- 1.1

U-235 -- -- 0.10

U-234 -- -- 30

a. Chest counting MDAs assumed a 1.9-cm chest wall thickness.

b. Current MDAs were taken from ORAU (2003).

c. = none.

d. For bioassay samples it would not be possible to differentiate between 239Pu 

and 240Pu due to identical energy peaks at 13.6 keV. Separate MDAs were 

reported by ORAU (2003).
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 41 of 66

MDAs are 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 1962). In vivo chest counting was therefore 

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

bioassay urinalysis. In addition, MDA data was quite limited and unavailable prior to 1990. Dose 

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

Chest counting data should only be used to assess intake when counting was conducted as a followup for a known worker intake.

5.4.3

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

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

sampling equipment, procedures, and calibration efforts. NRC Regulatory Guide 8.25 requires 

stringent acceptance criteria to demonstrate that such air sampling is representative of breathing zone 

air (NRC 1992). In addition, area air sampling cannot be used effectively in some workplace 

scenarios and can greatly underestimate actual worker inhalation intakes. Obtaining air sample data 

would probably entail a time-consuming search of archived site records.

Personnel Air Sampling Data 

Personnel breathing zone 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 breathing zone concentrations. In addition, the DOE claims data do not 

generally provide air concentration data for nuclides other than tritium.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 PC-1 Effective Date: 03/13/2009 Page 42 of 66

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Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 47 of 66

ATTACHMENT A

LUNG SOLUBILITY RECOMMENDATIONS FOR SITE PROCESS ACTIVITIES

Page 1 of 3

This attachment lists lung solubility site specific recommendations for Mound process activities. A 

claimant-favorable solubility should be used if a specific solubility type cannot be determined. In 

addition, solubility can be determined based on claim-specific intake or bioassay data and therefore 

cannot be definitely stated in the TBD. Table A-1 summarizes this information by building. Table A-1 

information was largely qualitative in nature and derived almost entirely on King (1995). 

Table A-1. Recommended lung clearance types for site process activities.

a

Building Site process activities

Lung 

clearance 

HH Building Hydrolysis of irradiated bismuth slugs. Gaseous diffusion and isotope separation 

activities. Th-230 extraction operations. Pa-231 separation from Sperry Cake.

Po-210 Separation Program: Hydrolysis of irradiated bismuth and aluminum chloride 

solutions to produce gelatinous bismuth-oxychloride sludge. Sludge shipped off site 

4/1954. 

Po-210: F or M

Pa-231 Separation Program: Extraction of Pa-231 from Sperry Cake. Extraction of 

Th-230 and Pa-231 from ore residues.

Pa-231: M 

Th-230: M 

Ionium Program: Extraction of Th-230 from Airport Cake raffinate during. Raffinate 

from uranium processing. Expected carbonate and sulfate compounds. 

Th-230: M 

Isotope Separation Program: Gaseous thermal using H-3 waste gas. Xenon isotope 

separation.

(ICRP 1979)

PP Building Primarily processing of Pu-238 dioxide from Savannah River Site feed materials. 

Various plutonium compounds used and soluble compounds reported. Various lung 

clearance types and particle sizes reported. Processing in rooms PP-144, PP-147 

and PP-148 and PP-162 and PP-163. 

Sol Gel processing with soluble EDTA complexes: No heat, vaporization, crushing, or 

grinding involved in process. 

Pu isotopes: M 

Microsphere pre-processing: Majority of particles reported as submicrometer from 

sieving and sintering operation. Soluble Pu compounds reported.

Pu isotopes: 

Oxides M 

Soluble F 

Microsphere Final Production: High temperature plasma processing, Pu condensation 

aerosols from plasma torch operations.

Pu isotopes: S

PMC Process: Microsphere production with crushing and sieving. Particle range of 

product reported as 43-250 μm. Mo used to coat Pu microspheres. Pu machining 

used.

Pu isotopes: S

PPO Process: Microsphere production with crushing and sieving. Hot press used to 

encapsulate.

Pu isotopes: S

Solid Solution Process: Th nitrate added to a 9:1 Pu/Th mixture. Crushing, sieving, 

and hot pressing used in process.

Pu isotopes: M

Th isotopes: M 

R Building Research and Development laboratories. Use of primarily Po-210, Pu-238/239, H-3, 

Ac-227, Th-230, Ra-226/228, Pa-231, and U-234 in neutron source and microsphere 

production, reactor fuel studies, Ionium Program, isotope separation programs, and 

limited H-3 recovery.

Neutron Source Program: Po-210 source processing. Electroplate deposition and 

precipitate processes. High vapor pressures. Submicrometer particles reported.

Po-210: F or M

Microsphere Processing: Pu-239 microspheres with 43-250 μm sizes from PP 

Building were encapsulated. 

Pu isotopes: S

Radium-Actinium Separation: Ac-227 separated from irradiated radium-bromide-226 

and K65 ore residues in SW-19 Old Cave.

Actinium Processing: Precipitates in SW-19 processed. Soluble Ac reported in 

precipitates.

Ra isotopes: M 

Ac-227: F or M
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 48 of 66

ATTACHMENT A

LUNG SOLUBILITY RECOMMENDATIONS FOR SITE PROCESS ACTIVITIES

Page 2 of 3

Building Site process activities

Lung 

clearance 

R Building 

(cont’d.)

Protactinium Separation: Pa-231 separation from Sperry Cake. Pa-231: M

U-234 Separation: Oxalate precipitation followed by ion exchange to separate U-234 

from aged Pu-238. Shipped to Oak Ridge National Laboratory.

U isotopes: M

Pu isotopes: M 

Pu Reactor Studies: Fuel studies using molten materials and mixtures. Pu isotopes: M

Ionium Program: Separation of Th-230 from Sperry Cake raffinate. Depleted U and 

Th isotopes reported for feed material compared to Cotter Concentrate. Possible 

Th-228, Th-232 present.

Th isotopes: M 

Metal Tritide Program: Production and studies of metal tritides including uranium 

tritides.

Metal tritide: S

U tritides: S

H-3 recovery operations: Recovery of H-3 from R Building operations. (ICRP 1998)

SM Building Received powered Pu dioxide and liquid Pu nitrate from Savannah River Site as feed 

materials. Microsphere and Plutonium metal production. U-234 separation. Heat 

source projects. 

Pu Metal Production: Liquid Pu nitrate and powdered Pu dioxide from Savannah 

River converted to Pu metal for SNAP. Some Pu compounds reported as soluble. 

Submicrometer to 20 m reported for various processing.

Pu isotopes: 

Oxides S All 

others M

Microsphere Production: High temperature plasma processing of submicrometer 

particles, Pu aerosols and fumes from plasma torch operations. Majority of particles 

reported as submicrometer from sieving and sintering operation.

Pu isotopes: S

Aqueous Waste Recovery: Pu isotopes recovered from high-risk wastewater. 

Wastewater equipment removed. Soluble Pu compounds reported.

Pu isotopes: M

U-234 Separation Program: Separation and recovery of U-234 from aged Pu-238 

sources using tributyl solvent extraction and ion exchange. Shipped to Oak Ridge 

National Laboratory.

U-234: M

SW Building Research and Development laboratories. Neutron source program support. Thorium 

refinery program. H-3 recovery and research. Pa-separation and purification. 

Radium and Actinium separation. Th-229 and U-234 separation activities.

Metal Tritide Program: Production and studies of metal tritides including uranium 

tritides.

Tritium Program: Thermal diffusion and enrichment, research, waste classification, 

and effluent capture. Remaining diffusion columns.

Metal tritide: S

U tritides: S 

(ICRP 1998)

U-234 Separation Program: Separation and recovery of U-234 from aged Pu-238 

sources using tri-butyl solvent extraction and ion exchange. Shipped to Oak Ridge 

National Laboratory.

U isotopes: M 

Pu isotopes: M 

Thorium Refinery Program: Extraction of thorium and radium from Brazilian monazite 

sludge (Monex). Oxalate and sulfate compounds reported. Some feed material could 

have been oxides or hydroxides due to digestion at high temperatures.

Th isotopes: M

Ra isotopes: M 

Pa-231 Purification: Processing of Cotter Concentrate residues to recover Th-230 and 

Pa-231. Use of organic phosphates. Th-230 had 95% of total activity. Reported 

soluble compounds.

Th isotopes: M

Ra isotopes: M 

Th-230: M 

Pa-231: M

Radium and thorium from activities involving production and separation of Ac-227. Th isotopes: M

Ra isotopes: M

Radium-Actinium Separation: Ac-227 separated from irradiated radium-bromide-226 

and K-65 ore in SW-19 Old Cave.

Actinium Processing: Precipitates from SW-19 processed in New Cave. Highly 

soluble Ac compounds reported. Ac-227 production activities conducted without 

radium.

Ra isotopes: M 

Ac-227: F or M

Thorium-229 Separation: Recovery of Th-229 from aged U-238 and U-233 using a 

chloride complexation process.

Th isotopes: M

U isotopes: M

Neutron Source Program: Po-209 and Po-210 recovery. Source manufacturing of 

Pu-238/Be and Pu-239/Be sources. Soluble Po compounds reported.

Po-210: F or M

Pu isotopes: S 

Am-241: S
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ATTACHMENT A

LUNG SOLUBILITY RECOMMENDATIONS FOR SITE PROCESS ACTIVITIES

Page 3 of 3

Building Site process activities

Lung 

clearance 

T Building Research and development laboratories. Extensive H-3 facilities and operations. 

Po-210 separation and fuels research. Neutron source program. Source storage 

operations.

Tritium Program: Thermal diffusion facility, H-3 and HTO recycle and enrichment 

system, H-3 effluent reduction facility, H3 isotope separation system, H-3 engineering 

and development laboratory. Gas preparation systems for large quantities of H-3 

(T-48, T-49, T-50). Cryogenic distillation columns with U capture beds. Uranium 

tritides, tritiated water, and tritium gas all present.

(ICRP 1998) 

Metal tritide: S 

U tritides: S 

Po-210 Separation Program: Dissolution of Al cans with irradiated Bismuth slugs. 

Processing and separation of Po-210. Storage of irradiated Bi slugs in T236, 237 

pool. Soluble Po compounds reported.

Po-210: F or M

Polonium Fuels Research: Po metal coupled with rare earths in metal matrix. Particle 

size of < 44 m reported in T260.

Po isotopes: S

Source Storage and Calibration: Storage of sealed orphan sources at T300A. 

Calibration of U Mill Tailings surveillance monitors in T-229. 

Ra-226+D: M

Neutron Source Program: Source manufacturing with Po-210 and Pu isotopes 

coupled with Be. Source testing and calibration. Soluble Po compounds reported.

Po-210: F or M 

Pu isotopes: S

WD/WDA

Buildingb

Treatment of liquid radioactive wastewater. Treatment systems for alpha and beta 

emitters. Polonium treatment process. Cotter Concentrate acid waste treatment. H-3 

wastewater treatment. Alpha emitter treatment facility. Effluent monitoring. 

Monex Sludge Storage: Storage of Th sludge from Cotter Concentrate processing. 

No processing in WD Building.

Th isotopes: M

Ra isotopes: M

Effluent Monitoring: Monitoring of wastewater effluents and treatment processes. 

Th-230 from Cotter processing. Plutonium decontamination activities contributed. 

Overflow contamination event with Po-210.

Po-210: F or M

Pu isotopes: S 

Th isotopes: S

Cotter Waste Treatment: Treatment of acidic wastewater from processing of Cotter 

Concentrate.

Th isotopes: S

Tritium Waste Treatment: Influent holding tanks in WDA -9. (ICRP 1998)

Polonium Treatment: Filtration using 1 m in WD-1. Drumming of sludge, shipment 

offsite. Soluble Po compounds reported.

Po-210: F or M

Plutonium Waste Treatment: Waste treatment facility. High- level wastewater from 

SM and PP contained alpha emitters. Acidic solutions of Pu isotopes, but primarily 

Pu-238. High risk drumming station.

Pu isotopes: S

Alpha Waste Treatment: High-level non-plutonium alpha emitters in wastewater from 

R and SW. Primarily Po-210. Tritium and Pu-238 secondary contaminants.

Po-210: F or M

Ac-227: M 

Pa-231: M

a. Source: King (1995).

b. WD/WDA received wastewater from other buildings, so specific information on form is unavailable. Use generic assumptions for 

nuclides when no specific information on process is available.

c. Assume tritium (H-3) is tritiated water vapor (HTO) unless specified otherwise. Assume 3% organically bound 3H with T1/2 = 40 d 

(ICRP 1998).

d. External irradiation from xenon and krypton dominates worker dose, so internal dose should be disregarded as stated in ICRP 30 

(ICRP 1979).
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ATTACHMENT B 

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 1 of 12

Table B-1 summarizes locations and dates of use for secondary radionuclides; the information is 

based on MD-22153, Mound Site Radionuclides by Location (King 1995). Secondary radionuclides 

are listed alphabetically by building for all rooms.

Radionuclides listed in Table B-1 should be considered of major concern only where the Comment 

column states that a primary or major concern exists. It should be noted that other primary or 

secondary radionuclides might have been present in a building and room along with each radionuclide 

listed in Table B-1. For example, 227Ac was listed as used in room R-140, but 226Ra and 231Pa were 

the primary radionuclides of concern. Therefore, 227Ac was not a primary radionuclide of concern.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 51 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 2 of 12

Table B-1. Secondary radionuclides by location and period. 

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

R R crawl space Ac-227 Table 3 1948 Present Crawl space above R building.

R R-113, 114, 115 Ac-227 Table 3 1951 1955 Ac-227 as nitrate, oxide, and oxalate, reportedly all soluble. Ac-227 major hazard followed by Ra-226 

and Th-228.

R R-118 Ac-227 Table 3 1955 1955 Soluble actinium as nitrate, oxalate, and oxide. Storage Area. Ac-227, Ra-226, Th-228 major hazards in 

1955. Th-230 major hazard in 1956-1958.

R R-118, 119, 

121, 123

Ac-227 Table 3 1955 1955 Soluble actinium as nitrate, oxalate, and oxide. Storage Area. Ac-227, Ra-226, Th-228 major hazards in 

1955. Th-230 the major hazard in 1956.

R R-120 Ac-227 Table 3 1951 1953 Actinium-227 processing. Actinium as nitrate, oxalate and oxide, chloride, fluoride. Also rooms 118, 

119, 121, 123. 1951-1953: Ac-227 primary hazard with Ra-226 and Th-228 secondary concerns. Th230 primary concern from 1956-1958.

R R-129, 130 Ac-227 Table 3 1951 1953 Actinium-227 processing. Actinium oxide, nitrate, and oxalate. Ac-227 major hazard followed by Ra-226 

and Th-228.

R R-133 Ac-227 Table 3 1951 1953 Actinium as nitrate, oxide, oxalate, fluoride, iodide. Th-230 major hazard 1956-1958.

R R-137 Ac-227 Table 3 1951 1953 Reduction of actinium. Ac as oxide, nitrate, oxalate, fluoride, iodide. Ac-227, Ra-226, and Th-228 major 

hazards from 1951-1953.

R R-140 Ac-227 Table 3 1950 1969 Analytical lab for R-120 work. Pa-231 and Ra-226 primary hazards. Am-241 secondary significance in 

late 1960s.

R R-142 Ac-227 Table 3 1949 1953 Radium-actinium separation. Ac oxalate, fluoride, radium bromide, nitrate, and carbonate. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-145, 147 Ac-227 Table 3 1949 1954 Radium-actinium separation. Radium carbonate and nitrate. Ac-227 major hazard followed by Ra-226 

and Th-228. Pa-231 separation from raffinate. Actinium oxalate, oxide. Pa-231 major hazard from 

1955-1956.

R R-148, 149 Ac-227 Table 3 1949 1954 Reduction and separation of actinium. Radium carbonate. Actinium nitrate, oxalate, oxide. Uranium 

nitrate. Ac-227 major hazard followed distantly by Ra-226 and Th-228. Separation of Pa-231 from 

residues. U-234 separation from Pu-238 from 1954-1958 & 1970-1980.

R R-152 Ac-227 Table 3 1951 1953 Actinium oxide, nitrate, oxalate, fluoride, iodide.

R R-160 Ac-227 Table 3 1951 1953 Actinium separation. Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228.

R R-161 Ac-227 Table 3 1951 1953 Purification of Actinium, Ac-227 as oxide, nitrate, oxalate. Ac-227 major hazard followed by Ra-226 and 

Th-228.

R R-169 Ac-227 Table 3 1960 1980 Low level counting room. U-238 separation research from 1980-present.

R R-171, 172 Ac-227 Table 3 1965 1991 Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228. U-235, 238 primary 

concerns: 1961-1963. Am-241 also present 1965-1991.

R R-171, 172 Ac-227 Table 3 1949 1954 Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228. U-235, 238 primary 

concern: 1961-1963.

R R-198 Ac-227 Table 3 1956 1982 Numerous isotopes. Decontamination lab. Ac-227 hazard was secondary to Pu-238. Am-241 present 

1956-1982.

R R-Corr-6 Ac-227 Table 3 1948 1984 Access corridor for hot side of R-Building. Ac-227 followed by Ra-226 and Th-228.

SW SW-128, 129, 

130, 134

Ac-227 Table 5 1970 1979 Support rooms. No processing. Th-230 was major radiological hazard.

SW SW-132 Ac-227 Table 5 1970 1979 Recovery of Pa-231 and Th-230 from Cotter concentrates. Actinium as oxides or nitrates.

SW SW-136, 137 Ac-227 Table 5 1964 1979 Provide access port to hot cell in SW-140 hot cell. Actinium as nitrate and oxide. Major concerns Th228, Ra-226, Ac-227, Pa-231, and U-234, U-235, U-238.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 52 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 3 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

SW SW-140 Ac-227 Table 5 1974 1979 Hot cell facility. Actinium processing. Recovery of Pa-231 and Th-230 from Cotter concentrates. Ac227, Pa-231, Th-230 major radionuclides.

SW SW-140 Ac-227 Table 5 1974 1979 Hot cell facility. Actinium processing. Recovery of Pa-231 and Th-230 from Cotter concentrates. Ac227, Pa-231, Th-230 major radionuclides.

SW SW-142 Ac-227 Table 5 1950 1961 Radium as nitrate, carbonate, bromide. Analytical laboratory. Actinium as fluoride and oxalate. U as 

deuteride and tritide. Major nuclides: Ra-226, Ac-227, Th-228. Metal tritides. 

SW SW-149, 149a Ac-227 Table 5 1965 1968 All SW Building Nuclides. Decontamination lab. Tritium was major isotope of concern.

SW SW-19 Ac-227 Table 5 1951 1953 Ac-227 separated from irradiated Ra-226. Radium as carbonate, nitrate, bromide. Ac-227 major hazard 

followed by Ra-226 and Th-228. Ac as metal, oxalate, fluoride.

SW SW-22 Ac-227 Table 5 1966 1975 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction, processing U ore. U-232, 233, Th-228, Th-230 present.

SW SW-22 Ac-227 Table 5 1970 1979 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction, processing U ore. U-232, 233, Th-228, Th-230 present. 

WD, WDA WD-1, 8, 101, 

104

Ac-227 Table 7 1949 1990 Liquid radioactive waste processing. “Low level waste” from SM building. Am-241 present 1949-1990.

WD, WDA WD-1, 8, 101,

104

Ac-227 Table 7 1974 1979 Liquid radioactive waste processing. “Low level waste” from SM building. Am-241 present 1949-1990.

WD, WDA WD Penthouse Ac-227 Table 7 1949 Present Filter bank for WD and WDA. All radionuclides are a hazard. Am-241 present 1949-present.

R R-111 Ac-227 & 

progeny

Table 3 1951 1965 Soluble actinium. Ac-227 major hazard followed by Ra-226 and Th-228.

R R-116 Am-241 Table 3 1956 1962 Production of alpha sources by plating various metal substrates. Am-241 was major concern, Pu-239 

secondary.

R R-120 R-118, 

119, 121, 123

Am-241 Table 3 1956 1962 Production of alpha sources by plating various metal substrates. Am-241 was major concern, Pu-239 

secondary.

R R-140 Am-241 Table 3 Late 

1960s

Present Analysis of Pu compounds via chemical analysis and emission spectroscopy. Vaporization of 

compounds by arcing. Pu-238 major hazard. Am-241 and Ac-227 of equal secondary concern.

R R-145, 147 Am-241 Table 3 1956 1959 Chemistry studies of reactor fuels. Pu compounds studied using thermal processes. Pu-239 major 

hazard. Am-241 a distant secondary concern.

R R-155 Am-241 Table 3 1956 1969 Reactor fuels program. Materials characterized by electron microprobe. U-238 major concern, then Pu239.

R R-159 Am-241 Table 3 1956 1969 Reactor Fuel Group experimentation with U and Pu compounds. Physical properties measured for metal 

alloys. U-238 major concern and Pu-239 secondary.

R R-167 Am-241 Table 3 1956 1987 Isotopic separation of Pa-231, Pa-233, characterization and analysis of Ce-141, Ce-144, Am-241, Pa231. Pa-231 major concern with all others a close second.

R R-171, 172 Am-241 Table 3 1965 1991 Vaporization of oxide solutions by arcing. Pu-238 major concern with all others a distant second.

R R-198 Am-241 Table 3 1956 1982 Numerous isotopes. Decontamination lab. Pu-238 major concern. Ac-227 secondary.

T T-5E Am-241 Table 5 1955 Present Nondestructive testing using X-ray techniques and radiography. Tritium major concern, Pu-238 

secondary, and Pu-239 tertiary concern.

T T-19, 19A Am-241 Table 6 1985 Present Nondestructive X-ray testing. Tritium major concern, Pu-238 secondary, and Pu-239 tertiary concern.

T T-208A Am-241 Table 6 1947 Present Tritium major concern, Pu-238 secondary, and Pu-239 tertiary concern.

PP PP-146, 150, 

151, 152, 153, 

154, 156

Am-241 Table 2 1967 1980 Anion exchange recovered Pu via DOWEX-1 resin. Pu dissolved in nitric acid, hydrofluoric acid, 

hydroxylamine, sodium carbonate. Other radionuclides present, but of minor concern. 

SM SM-34 Am-241 Table 4 1963 1967 Treatment of solid and liquid wastes. Pu primary hazard, other radionuclides minor.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 53 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 4 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

SM SM-38 Am-241 Table 4 1963 1967 Treatment of solid and liquid wastes. Pu primary hazard, other radionuclides minor.

SM SM-59 Am-241 Table 4 1963 1967 Pu waste recovery. Pu isotopes primary radiological concern, including Pu dioxide, sulfate, nitrate, 

fluoride.

T T-310 Am-241 Table 6 1969 1983 Calorimetric measurements and encapsulation of other radionuclides. Pu-238 major concern, H-3 

secondary, and Pu-239 tertiary concern.

T T-319 Am-241 Table 6 1949 Present Calorimetric measurements and encapsulation of other radionuclides. Pu-238 major concern, H-3 

secondary, and Pu-239 tertiary concern. 

WD WD-1 Am-241 Table 7 1949 1990 Materials in aqueous waste. Po-210 major concern, with H-3 and Pu-238 secondary. Pa-231 a distant 

third.

WD WD-1 Am-241 Table 7 1961 1974 Low level acidic and basic aqueous wastes Pu wastes. Pu-238 major concern with Pu-239 secondary.

WD WD-8 Am-241 Table 7 1949 1990 Aqueous waste operations and decontamination of Pu facility. Po-210 major concern, tritium and Pu-238 

secondary. Ac-227 and Pa-231 third.

WD WD-8 Am-241 Table 7 1961 1974 Pu-238 was major concern with Pu-239 secondary. Am-241 present 1949-1990.

WD WD-9 Am-241 Table 7 1973 Present Tritium major concern with all other radionuclides secondary.

WD WD-10 Am-241 Table 7 1961 1974 Aqueous plutonium solutions. Pu-238 major concern with Pu-239 secondary. Room was highly 

contaminated.

WD WD-101 Am-241 Table 7 1949 1990 Aqueous waste operations and decontamination of Pu facility. Po-210 major concern, tritium and Pu-238 

secondary. Ac-227 and Pa-231 third.

WD WD-101 Am-241 Table 7 1961 1974 Low-level aqueous plutonium wastes. Pu-238 major concern, then Pu-239.

WD WD-104 Am-241 Table 7 1949 1990 Aqueous waste operations and decontamination of Pu facility. Po-210 major concern, tritium and Pu-238 

secondary. Ac-227 and Pa-231 third.

WD WD-104 Am-241 Table 7 1961 1974 Low-level aqueous plutonium wastes. Pu-238 major concern, then Pu-239.

WD WD-107 and 

basement

Am-241 Table 7. 1976 1981 Ultra filtration and reverse osmosis. All radionuclides in trace amounts and of minor concern.

WD WD-108 Am-241 Table 7 1949 1981 Highly contaminated stack with filter.

WD WDA-112 Am-241 Table 7 1966 1979 Techniques developed to separate Pu from wastewater. Pu-238 primary concern with Pu-239 second.

WD WD-Penthouse Am-241 Table 7 1949 Present Filter bank for the building. Any radionuclide in the building potentially present on filters.

SW SW-219 Am-241 Table 5 1962 1965 Pu neutron source production. Pu-238 was major concern, followed by Pu-239, then Am-241.

T T-277, 279 Am-241 Table 6 1949 1967 Polonium neutron source program, which used Be, Bo, Li, etc. 

HH HH-5 Pa-231 Table 1 1955 1958 Separation of Pa-231 from raffinates from 1955-1956. Decommissioned by 1958. Th-230 extraction 

from Airport Cake. Pa-231 and Th-230 major nuclides.

HH HH-6 Pa-231 Table 1 1955 1956 Pilot plant for separation of Pa-231 from Raffinates. Separation of Th-230 from Airport Cake raffinate. 

Th-230 and Pa-231 predominant nuclides.

R R-117 Pa-231 Table 3 1948 1990 Storage vault. Pa-231, Th-230, Pu, and Po. Pa-231 was major nuclide before 1955.

R R-118 Pa-231 Table 3 1956 1958 Soluble actinium as nitrate, oxalate and oxide. Storage Area. Ac-227, Ra-226, Th-228 major hazard in 

1955. Th-230 major hazard from 1956-1958.

R R-118, 119, 

121, 123

Pa-231 Table 3 1956 1958 Soluble actinium as nitrate, oxalate and oxide. Storage Area. Ac-227, Ra-226, Th-228 major hazard in 

1955. Th-230 major hazard from 1956-1958. Am-241 present 1956-1962.

R R-120 Pa-231 Table 3 1956 1958 Actinium as nitrate, oxalate and oxide, chloride, fluoride. Also rooms 118, 119, 121, 123. 1951-1953: 

Ac-227 primary hazard with Ra-226 and Th-228 secondary concerns. Th-230 primary concern from 

1956-1958. Am-241 present 1956-1962.

R R-133 Pa-231 Table 3 1956 1958 Actinium as nitrate, oxide, oxalate, fluoride, iodide. Th-230 major hazard from 1956-1958.

R R-140 Pa-231 Table 3 1950 1969 Analytical lab for R-120 work. Pa-231 and Ra-226 primary hazards. Am-241 of secondary significance 

in late 1960s.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 54 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 5 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

R R-142 Pa-231 Table 3 1956 1969 Ac-oxalate, fluoride, radium bromide, nitrate, and carbonate. Radium-actinium separation. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-167 Pa-231 Table 3 1960 1962 Many isotopes. Separation of Pu from irradiated U from 1949-1952. Extraction of Pa-231 from residues 

from 1954-1955. U-233 major nuclide in 1958. Am-241 secondary 1956-1987.

R R-169 Pa-231 Table 3 1960 1980 Low level counting room. 1980-present: U-238 separation research.

SW SW-128, 129, 

130, 134

Pa-231 Table 5 1970 1979 Support rooms. No processing. Th-230 was major radiological hazard.

SW SW-132 Pa-231 Table 5 1970 1979 Actinium as oxide or nitrate. Recovery of Pa-231 and Th-230 from Cotter concentrate.

SW SW-136, 137 Pa-231 Table 5 1964 1979 Actinium as nitrate and oxide. Provide access port to hot cell in SW-140. Th-228, Ra-226, Ac-227, Pa231, U-234, U-235, and U-238 major nuclides.

SW SW-140 Pa-231 Table 5 1974 1979 Hot cell facility. Actinium processing of Pa-231 and Th-230 from Cotter concentrate. Ac-227, Pa-231, 

Th-230 major nuclides.

SW SW-22 Pa-231 Table 5 1970 1979 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction, processing U ore. U-232, 233, Th-228, Th-230 present.

WD, WDA WD-1, 8, 101, 

104

Pa-231 Table 7 1974 1979 Liquid radioactive waste processing. "Low level waste" from SM building. All SM Buildings were a 

concern.

WD, WDA WD Penthouse Pa-231 Table 7 1949 Present Filter bank for WD and WDA. All radionuclides were a hazard. Am-241 present 1949-present.

R R-167 Pa-231, 233 Table 3 1956 1987 Many nuclides present. Separation of Pu from irradiated U from 1949-1952. Extraction of Pa-231 from 

residues from 1954-1955. U-233 major nuclide in 1958. Am-241 secondary 1956-1987.

SW SW-19 Ra & progeny Table 5 1973 Present Radium as carbonate, nitrate, bromide. Ac-227 separated from irradiated Ra-226. Ac-227 major hazard 

followed by Ra-226 and Th-228. Actinium as metal, oxalate, fluoride.

R R-111 Ra-223, 224, 226 Table 3 1951 1965 Soluble actinium. Ac-227 major hazard followed by Ra-226 and Th-228.

R R-118 Ra-223, 224, 226 Table 3 1955 1955 Soluble actinium as nitrate, oxalate and oxide. Storage Area. Ac-227, Ra-226, Th-228 major hazard in 

1955. Th-230 major hazard from 1956-1958.

R R-118, 119, 

121, 123

Ra-223, 224, 226 Table 3 1955 1955 Soluble actinium as nitrate, oxalate and oxide. Storage Area. Ac-227, Ra-226, Th-228 major hazard in 

1955. Th-230 major hazard from 1956-1958.

R R-129, 130 Ra-223, 224, 226 Table 3 1951 1953 Actinium oxide, nitrate, and oxalate. Actinium -227 processing. Ac-227 major hazard followed by Ra226 and Th-228.

R R-140 Ra-223, 224, 226 Table 3 1950 1969 Analytical lab for R-120 work. Pa-231 and Ra-226 primary hazards. Am-241 of secondary significance 

in late 1960s.

R R-142 Ra-223, 224, 226 Table 3 1949 1953 Ac-oxalate, fluoride, radium bromide, nitrate, carbonate. Radium-actinium separation. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-169 Ra-223, 224, 226 Table 3 1960 1980 Low level counting room. 1980-present: U-238 separation research.

R R-171, 172 Ra-223, 224, 226 Table 3 1949 1954 Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228. U-235, U-238 

primary concern from 1961-1963.

SW SW-128, 129, 

130, 134

Ra-223, 224, 226 Table 5 1970 1979 Support rooms. No processing. Th-230 was major radiological hazard.

SW SW-132 Ra-223, 224, 226 Table 5 1970 1979 Actinium as oxide or nitrate. Recovery of Pa-231 and Th-230 from Cotter concentrates.

SW SW-136, 137 Ra-223, 224, 226 Table 5 1964 1979 Actinium as nitrate and oxide. Provide access port to hot cell in SW-140. Th-228, Ra-226, Ac-227, Pa231, U-234, U-235, and U-238 all present.

SW SW-140 Ra-223, 224, 226 Table 5 1964 1968 Hot cell facility. Actinium processing. Recovery of Pa-231 and Th-230 from Cotter concentrates. Ac227, Pa-231, Th-230 major nuclides.

WD, WDA WD-1, 8, 101, 

104

Ra-223, 224, 226 Table 7 1974 1979 Liquid radioactive waste processing. "Low level waste" from SM building. All SM Buildings were a 

concern. Am-241 present 1949-1990.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 55 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 6 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

WD, WDA WD-9 Ra-223, 224, 226 Table 7 1973 Present Uranium as deuterides and tritides. Pumps for influent tritiated wastewater. Tritium was primary 

radiological hazard. Am-241 present 1973-present.

R R-113, 114, 115 Ra-223, 224, 

226, 

Table 3 1951 1955 Ac-227 as nitrate, oxide and oxalate, reportedly all soluble. Ac-227 major hazard followed by Ra-226 

and Th-228.

WD, WDA WD-Penthouse Ra-223, 224, 

226, 228

Table 7 1949 Present Filter bank for WD and WDA. All radionuclides are a hazard. Am-241 present 1949-present.

WD, WDA WD-1, 8, 101, 

104

Ra-223, 226 Table 7 1949 1990 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.

SW SW-22 Ra-224, 225 Table 5 1966 1975 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction, processing U ore. U-232, 233, Th-228, Th-230 all 

present.

SW SW-22 Ra-224, 226, 228 Table 5 1970 1979 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction, processing U ore. U-232, 233, Th-228, Th-230 all 

present.

SW SW-11, 12, 13, 

14, 15, 16

Ra-224, 228 Table 5 1955 1955 Numerous nuclides. Th-232 extraction from monazite sand. Th-232, Th-228, Ra-224 major hazard. Pa231 purification process. Pa-231 major concern, U-238 secondary hazard. 

WD, WDA WD-1, 8, 101, 

104

Ra-224, 228 Table 7 1955 1955 Liquid radioactive waste processing. "Low level waste" from SM building. All SM Buildings were a 

concern. Am-241 present 1949-1990.

R R crawl space Ra-226 Table 3 1948 Present Crawl space above R building.

R R-120, 118, 

119, 121, 123.

Ra-226 Table 3 1951 1953 Actinium as nitrate, oxalate and oxide, chloride, fluoride. Ac-227 primary hazard with Ra-226 and Th-228 

secondary concerns from 1951-1953. Th-230 primary concern from 1956-1958.

R R-133 Ra-226 Table 3 1951 1953 Actinium as nitrate, oxide, oxalate, fluoride, iodide. Th-230 major hazard from 1956-1958.

R R-137 Ra-226 Table 3 1951 1953 Ac as oxide, nitrate, oxalate, fluoride, iodide. Reduction of actinium. Ac-227, Ra-226, Th-228 major 

hazards from 1951-1953.

R R-142 Ra-226 Table 3 1949 1953 Radium-actinium separation. Ac-oxalate, fluoride, radium bromide, nitrate, carbonate. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-145, 147 Ra-226 Table 3 1949 1954 Radium-actinium separation. Radium carbonate and nitrate. Actinium oxalate, oxide. Ac-227 major 

hazard followed by Ra-226 and Th-228. Pa-231 separation from raffinate. Pa-231 major hazard from 

1955-1956.

R R-148, 149 Ra-226 Table 3 1949 1954 Radium-actinium separation. Reduction of actinium. Radium carbonate. Actinium nitrate, oxalate, 

oxide. Ac-227 major hazard followed distantly by Ra-226 and Th-228. Separation of Pa-231 from 

residues. U-234 separation from Pu-238 from 1954-1958 & 1970-1980. Uranium nitrate.

R R-152 Ra-226 Table 3 1951 1953 Actinium oxide, nitrate, oxalate, fluoride, iodide.

R R-160 Ra-226 Table 3 1951 1953 Actinium separation. Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228.

R R-161 Ra-226 Table 3 1951 1953 Purification of Actinium. Ac-227 as oxide, nitrate, oxalate. Ac-227 major hazard followed by Ra-226 and 

Th-228.

R R-167 Ra-226 Table 3 1956 1987 Many nuclides. Separation of Pu from irradiated U from 1949-1952. Extraction of Pa-231 from residues 

from 1954-1955. U-233 major nuclide in 1958. Am-241 secondary 1956-1987.

R R-171, 172 Ra-226 Table 3 1965 1991 Actinium separation. Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228. 

U-235, U-238 primary concern from 1961-1963. Am-241 also present 1965-1991.

SW SW-142 Ra-226 Table 5 1950 1961 Analytical laboratory. Radium as nitrate, carbonate, bromide. Metal tritides. Actinium as fluoride and 

oxalate. U as deuteride and tritide. Major nuclides Ra-226, Ac-227, Th-228. 

SW SW-19 Ra-226 Table 5 1951 1953 Ac-227 separated from irradiated Ra-226. Radium as carbonate, nitrate, bromide. Actinium as metal, 

oxalate, fluoride. Ac-227 major hazard followed by Ra-226 and Th-228. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 56 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 7 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

T T-19, 19A Ra-226 Table 6 1985 Present Encapsulated sources. X-ray and non-destructive testing. Am-241 present 1985-present.

T T-208A Ra-226 Table 6 1947 Present Encapsulated sources. X-ray and non-destructive testing. Am-241 also present.

SW SW-142 Rn & progeny Table 5 1950 1961 Analytical laboratory. Radium as nitrate, carbonate, bromide. Metal tritides. Actinium as fluoride and 

oxalate. U as deuteride and tritide. Major nuclides Ra-226, Ac-227, Th-228.

SW SW-19 Rn & progeny Table 5 1951 1953 Ac-227 separated from irradiated Ra-226. Radium as carbonate, nitrate, bromide. Actinium as metal, 

oxalate, fluoride. Ac-227 major hazard followed by Ra-226 and Th-228. 

T T-229 Rn & progeny Table 6 1978 1988 Neutron source calibration lab. Other radiological instrument calibration.

WD, WDA WD-9 Rn & progeny Table 7 1973 Present Uranium as deuterides and tritides. Pumps for influent tritiated wastewater. Tritium primary radiological 

hazard. Am-241 present 1973-present.

R R-142 Rn, various 

isotopes

Table 3 1949 1953 Radium-actinium separation. Ac-oxalate, fluoride. Radium bromide, nitrate, carbonate. Ac-227 major 

hazard followed by Ra-226 and Th-228.

WD, WDA WD-9 Rn & progeny Table 7 1973 Present Pumps for influent tritiated wastewater. Uranium as deuterides and tritides. Tritium was primary 

radiological hazard. Am-241 present 1973-present.

HH HH-6 Th Table 1 1955 1956 Pilot plant for separation of Pa-231 from Raffinates. Separation of Th-230 from Airport Cake raffinate. 

Th-230 and Pa-231 were predominant nuclides.

SW SW-22 Th-228 & 

progeny

Table 5 1966 1975 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction, processing U ore. U-232, U-233, Th-228, Th-230 all 

present.

SW SW-22 Th-228, 229 Table 5 1966 1975 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction, processing U ore. U-232, U-233, Th-228, Th-230 all 

present.

R R crawl space Th-228, 229, 230 Table 3 1948 Present Crawl space above R building.

R R-111 Th-228, 229, 230 Table 3 1951 1965 Soluble actinium. Ac-227 major hazard followed by Ra-226 and Th-228.

R R-120, Also 

118, 119, 121, 

123.

Th-228, 229, 230 Table 3 1951 1953 Actinium as nitrate, oxalate and oxide, chloride, fluoride. 1951-1953: Ac-227 primary hazard with Ra226 and Th-228 secondary concerns. Th-230 primary concern from 1956-1958. Am-241 present 1956-

1962.

R R-129, 130 Th-228, 229, 230 Table 3 1951 1953 Actinium -227 processing. Actinium oxide, nitrate, and oxalate. Ac-227 major hazard followed by Ra226 and Th-228.

R R-133 Th-228, 229, 230 Table 3 1951 1953 Actinium as nitrate, oxide, oxalate, fluoride, iodide. 1956-1958: Th-230 major hazard.

R R-137 Th-228, 229, 230 Table 3 1951 1953 Reduction of actinium. Ac as oxide, nitrate, oxalate, fluoride, iodide. Ac-227, Ra-226, Th-228 major 

hazards from 1951-1953.

R R-142 Th-228, 229, 230 Table 3 1949 1953 Radium-actinium separation. Ac-oxalate, fluoride, radium bromide, nitrate, carbonate. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-145, 147 Th-228, 229, 230 Table 3 1949 1954 Radium-actinium separation. Radium carbonates and nitrates Actinium oxalate, oxide. Ac-227 major 

hazard followed by Ra-226 and Th-228. Pa-231 separation from raffinate. Pa-231 major hazard from 

1955-1956.

R R-148, 149 Th-228, 229, 230 Table 3 1949 1954 Reduction of actinium. Separation work. Radium nitrate and carbonate. Actinium nitrate, oxalate, oxide. 

Uranium nitrate. Ac-227 major hazard followed distantly by Ra-226 and Th-228. Separation Pa-231 

from residues. U-234 separation from Pu-238 from 1954-1958 & 1970-1980.

R R-152 Th-228, 229, 230 Table 3 1951 1953 Actinium oxide, nitrate, oxalate, fluoride, iodide.

R R-160 Th-228, 229, 230 Table 3 1951 1953 Actinium separation. Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228.

R R-161 Th-228, 229, 230 Table 3 1951 1953 Purification of Actinium. Ac-227 as oxide, nitrate, oxalate. Ac-227 major hazard followed by Ra-226 and 

Th-228.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 57 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 8 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

R R-171, 172 Th-228, 229, 230 Table 3 1949 1954 Actinium separation. Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228. 

U-235, 238 primary concern from 1961-1963.

SW SW-142 Th-228, 229, 230 Table 5 1950 1961 Analytical laboratory. Radium as nitrate, carbonate, bromide. Metal tritides. Actinium as fluoride and 

oxalate. U as deuteride and tritide. Major nuclides: Ra-226, Ac-227, Th-228. 

SW SW-19 Th-228, 229, 230 Table 5 1951 1953 Ac-227 separated from irradiated Ra-226. Radium as carbonate, nitrate, bromide. Actinium as metal, 

oxalate, fluoride. Ac-227 major hazard followed by Ra-226 and Th-228. 

WD, WDA 8, 101, 104 Th-228, 229, 230 Table 7 1949 1990 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.

R R-118 Th-228, 229, 

230, 232

Table 3 1955 1955 Storage Area. Soluble actinium as nitrate, oxalate and oxide. Ac-227, Ra-226, Th-228 major hazards in 

1955. Th-230 major hazard from 1956-1958.

R R-140 Th-228, 230, 232 Table 3 1950 1969 Analytical lab for R-120 work. Pa-231 and Ra-226 primary hazards. Am-241 of secondary significance 

in late 1960s.

SW SW-11, 12, 13, 

14, 15, 16

Th-228, 232 Table 5 1955 1955 Numerous nuclides. Th-232 extraction from monazite sand. Th-232, Th-228, Ra-224 major hazards. 

Pa-231 purification process. U-238 secondary hazard. 

WD, WDA WD-1, 8, 101, 

104

Th-228, 232 Table 7 1955 1955 Liquid radioactive waste processing. “Low level waste" from SM building. Am-241 present 1949-1990.

R R-167 Th-229, 230, 232 Table 3 1958 1958 Many isotopes. Separation of Pu from irradiated U (1949-1952), Extraction of Pa-231 from residues from 

1954-1955. U-233 major nuclide in 1958. Am-241 secondary 1956-1987.

HH HH-5 Th-230 Table 1 1956 1958 Separation of Pa-231 from 1955-1956. Decommissioned by 1958. Th-230 extraction from Airport Cake 

raffinate. Pa-231 and Th-230 major nuclides.

R R-117 Th-230 Table 3 1948 1990 Storage vault. Pa-231, Th-230, Pu, and Po. Pa-231 was major nuclide before 1955.

R R-142 Th-230 Table 3 1956 1969 Radium-actinium separation. Ac-oxalate, fluoride, radium bromide, nitrate, carbonate. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-148, 149 Th-230 Table 3 1970 1980 Reduction of actinium. Separation work. Radium carbonate. Actinium nitrate, oxalate, oxide. Uranium 

nitrate. Ac-227 major hazard followed distantly by Ra-226 and Th-228. Separation of Pa-231 from 

residues. U-234 separation from Pu-238 from 1954-1958 & 1970-1980.

R R-142 Th-230 & 

progeny

Table 3 1956 1969 Radium-actinium separation. Ac-oxalate, fluoride, radium bromide, nitrate, carbonate. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-118 Th-230, 232 Table 3 1956 1958 Storage Area. Soluble actinium as nitrate, oxalate and oxide. Ac-227, Ra-226, Th-228 major hazard in 

1955. Th-230 major hazard from 1956-1958. Am-241 present 1956-1962.

R R-120, Also 

118, 119, 121, 

123.

Th-230, 232 Table 3 1956 1958 Actinium as nitrate, oxalate and oxide, chloride, fluoride. Ac-227 primary hazard with Ra-226 and Th-228 

secondary concerns from 1951-1953. Th-230 primary concern from 1956-1958. Am-241 present 1956-

1962.

R R-133 Th-230, 232 Table 3 1956 1958 Actinium as nitrate, oxide, oxalate, fluoride, iodide. Th-230 major hazard from 1956-1958.

R R-167 Th-230, 232 Table 3 1956 1987 Many isotopes. Separation of Pu from irradiated U from 1949-1952. Extraction of Pa-231 from residues 

from 1954-1955. U-233 major nuclide in 1958. Am-241 secondary 1956-1987.

R R-169 Th-230, 232 Table 3 1960 1980 Low level counting room. 1980-present: U-238 separation research.

SW SW-128, 129, 

130, 134

Th-230, 232 Table 5 1970 1979 Support rooms. No processing. Th-230 was major radiological hazard.

SW SW-132 Th-230, 232 Table 5 1970 1979 Actinium as oxide or nitrate. Recovery of Pa-231 and Th-230 from Cotter concentrates.

SW SW-136, 137 Th-230, 232 Table 5 1964 1979 Provide access port to hot cell in SW-140. Actinium as nitrate and oxide. Nuclides of concern Th-228, 

Ra-226, Ac-227, Pa-231, U-234, U-235, and U-238.

SW SW-140 Th-230, 232 Table 5 1974 1979 Hot cell facility. Actinium processing and recovery of Pa-231 and Th-230 from Cotter concentrates. 

Major nuclides were Ac-227, Pa-231, and Th-230. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 58 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 9 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

SW SW-22 Th-230, 232 Table 5 1970 1979 New Cave. Th-229 separated from U-233. Actinium nitrate and oxide. Other Th and Pa projects 

included Cotter Concentrate extraction and U ore processing. U-232, U-233, Th-228, Th-230 present.

WD, WDA WD-1, 8, 101, 

104

Th-230, 232 Table 7 1974 1979 Liquid radioactive waste processing. "Low level waste" from SM building. 

WD, WDA WD Penthouse Th-230, 232 Table 7 1949 Present Filter bank for WD and WDA. All radionuclides are a hazard. Am-241 present 1949-present.

R R-118, 119, 

121, 123

Th-230, 232 & 

progeny

Table 3 1956 1958 Storage Area. Soluble actinium as nitrate, oxalate and oxide. Ac-227, Ra-226, Th-228 major hazards in 

1955. Th-230 major hazard in 1956. Am-241 present 1956-1962.

WD, WDA WD-1, 8, 101, 

104

Th-230, 232 Table 7 1974 1979 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.

38 & PP 147, 148 Th-232 Table 2 1967 1980 Precipitation, crushing, hot-pressing, sintering operations. Thorium used as analogue for Pu. Th-nitrate, 

oxide, oxalate, hydroxide and fluoride. 

38 & PP Corr-141 Th-232 Table 2 1967 Present Pu Processing building, Th nitrate, oxalate, and hydroxides.

38 & PP Corr-142 Th-232 Table 2 1967 Present Pu Processing bldg. Th nitrate, oxalate, and hydroxides present.

38 & PP PP-143 (A1) Th-232 Table 2 1967 1980 Pu Processing Building. Precipitation, crushing, hot-pressing, sintering. Thorium used as analogue for 

Pu. Th-nitrate, oxalate, and hydroxides. 

38 & PP PP-144 (A2) Th-232 Table 2 1967 1980 Precipitation, crushing, hot-pressing, sintering. Thorium used as analogue for Pu. Th nitrate, oxalate, 

hydroxides and insoluble forms. 

38 & PP PP-Corr-140 Th-232 Table 2 1967 Present Pu building, also Th nitrate, oxalate, and hydroxides.

HH HH-5 Th-232 Table 1 1956 1958 Separation of Pa-231 from 1955-1956. Decommissioned by 1958. Th-230 extraction from Airport Cake 

raffinate. Pa-231 and Th-230 major nuclides.

R R-129, 130 Th-232 Table 3 1965 1978 Actinium-227 processing. Actinium oxide, nitrate, and oxalate. Ac-227 major hazard followed by Ra-226 

and Th-228.

SM SM-21 Th-232 Table 4 1961 1967 Analytical Laboratory. Physical measurements. Th used as analogue to Pu. EDTA and citrates as 

complexing agents. Storage area for radioactive materials. 

SM SM-61, 62 Th-232 Table 4 1961 1967 Analytical labs. Studies of Pu. Th-232 used as an analogue for Pu. Complexing agents EDTA and 

citrates used with Pu.

R R-118 Th-232 & 

progeny

Table 3 1956 1958 Storage Area. Soluble actinium as nitrate, oxalate and oxide. Ac-227, Ra-226, Th-228 major hazards in 

1955. Th-230 major hazard from 1956-1958. Am-241 present 1956-1962.

SW SW-11, 12, 13, 

14, 15, 16

Th-232 & 

progeny

Table 5 1955 1955 Numerous isotopes. Th-232 extraction from monazite sand. Th-232, Th-228, Ra-224 major hazards. 

Pa-231 purification process. U-238 secondary hazard. 

SM SM-1 Th-232, 228 Table 4 1962 1967 Pu facility. Treatment of low risk waste.

HH HH-6 U Table 1 1955 1956 Pilot plant for separation of Pa-231 from Raffinates. Separation of Th-230 from Airport Cake raffinate. 

Th-230 and Pa-231 were predominant nuclides.

SM SM-34 U Table 4 1963 1971 U as oxide, fluoride and sulfate. Pu major hazard. Am-241 present 1963-1967.

SM SM-58 U Table 4 1963 1967 Pu facility waste recovery. Pu- isotopes were primary radiological concern. U oxide, fluoride, nitrate, 

sulfate.

SW SW-11, 12, 13, 

14, 15, 16

U Table 5 1968 Present Numerous nuclides. Th-232 extraction from monazite sand. Th-232, Th-228, Ra-224 major hazards. 

Pa-231 purification process. U-238 secondary hazard. 

SW SW-11, 12, 13, 

14, 15, 16

U Table 5 1959 1978 Numerous nuclides. Th-232 extraction from monazite sand. Th-232, Th-228, Ra-224 major hazards. 

Pa-231 purification process. U-238 secondary hazard.

SW SW-9 U Table 5 1989 Present U deuteride, tritide, and hydrides. Tritium was primary radiological hazard.

SW SW-22 U-232, 233 Table 5 1966 1975 New Cave. Th-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa projects 

including Cotter Concentrate extraction, processing U ore. U-232, U-233, Th-228, Th-230 all present.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 59 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 10 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

HH HH-5 U-233 Table 1 1955 1958 Separation of Pa-231 from 1955-1956. Decommissioned by 1958. Th-230 extraction from Airport Cake 

raffinate. Pa-231 and Th-230 major nuclides.

R R-167 U-233 Table 3 1956 1987 Many isotopes. Separation of Pu from irradiated U (1949-1952), Extraction of Pa-231 from residues from 

1954-1955. U-233 major nuclide in 1958. Am-241 secondary 1956-1987.

R R-167 U-233 Table 3 1958 1958 Numerous nuclides. Separation of Pu from irradiated U from 1949-1952. Extraction of Pa-231 from 

residues from 1954-1955. U-233 major nuclide in 1958. Am-241 secondary 1956-1987.

T T-310 U-233, 235 Table 6 1969 1983 Calorimetry. Pu isotopes and tritium were primary hazard. Am-241 present 1969-1983.

T T-319 U-233, 235 Table 6 1949 Present Calorimetry. Pu isotopes and tritium were primary hazard. Am-241 present 1949-present.

R R-162 U-233, 235, 238 Table 3 1958 1969 Analytical lab. Pu and fission products major concern 1948-1951. U-233, U-235, U-238 major concerns 

1958-1959.

HH HH-5 U-234 Table 1 1955 1958 Separation of Pa-231 from 1955-1956. Decommissioned by 1958. Th-230 extraction from Airport Cake 

raffinate. Pa-231 and Th-230 major nuclides.

R R-148, 149 U-234 Table 3 1970 1980 Reduction of actinium. Separation work. Radium carbonate. Actinium nitrate, oxalate, oxide. Uranium 

nitrate. Ac-227 major hazard followed distantly by Ra-226 and Th-228. Separation of Pa-231 from 

residues. U-234 separation from Pu-238 from 1954-1958 & 1970-1980.

WD, WDA WD-1, 8, 101, 

104

U-234 Table 7 1961 1974 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.

WD, WDA WD-10 U-234 Table 7 1961 1974 High activity influent storage tanks in room. Highly contaminated area. Pu-238 was primary concern. 

Am-241 present 1961-1974.

WD, WDA WD-108 U-234 Table 7 1949 1981 Room having stack with filter. Highly contaminated area. Uranium and transuranics present. Am-241 

from 1949-1981.

WD, WDA WDA-110 U-234 Table 7 1966 1970 Alpha emitter wastewater treatment. Pu isotopes were major radiological hazard.

WD, WDA WDA-112 U-234 Table 7 1966 1979 Extraction of Pu from wastewater. Pu-238 was primary radiological hazard. Am-241 present 1966-1979.

R R-118 U-234, 235, 238 Table 3 1956 1958 Storage Area. Soluble actinium as nitrate, oxalate and oxide. Ac-227, Ra-226, Th-228 major hazards in 

1955. Th-230 major hazard from 1956-1958.

R R-118, 119, 

121, 123

U-234, 235, 238 Table 3 1956 1958 Storage Area. Soluble actinium as nitrate, oxalate and oxide. Ac-227, Ra-226, Th-228 major hazards in 

1955. Th-230 major hazard in 1956. Am-241 present 1956-1962.

R R-120, also 

rooms 118, 119, 

121, 123.

U-234, 235, 238 Table 3 1956 1958 Actinium as nitrate, oxalate and oxide, chloride, fluoride. 1951-1953: Ac-227 primary hazard with Ra226 and Th-228 secondary concerns. Th-230 primary concern from 1956-1958. Am-241 present 1956-

1962.

R R-133 U-234, 235, 238 Table 3 1956 1958 Actinium as nitrate, oxide, oxalate, fluoride, iodide. Th-230 major hazard from 1956-1958.

R R-140 U-234, 235, 238 Table 3 1950 1969 Analytical lab for R-120 work. Pa-231 and Ra-226 primary hazards. Am-241 of secondary significance 

in late 1960s.

R R-142 U-234, 235, 238 Table 3 1956 1969 Radium-actinium separation. Ac-oxalate, fluoride, radium bromide, nitrate, carbonate. Ac-227 major 

hazard followed by Ra-226 and Th-228.

R R-145, 147 U-234, 235, 238 Table 3 1968 1975 Radium-actinium separation. Radium carbonate and nitrate. Ac-227 major hazard followed by Ra-226 

and Th-228. Pa-231 separation from raffinate. Actinium oxalate, oxide. Pa-231 major hazard 1955-

1956. 

R R-167 U-234, 235, 238 Table 3 1949 1952 Numerous nuclides. Separation of Pu from irradiated U from 1949-1952. Extraction of Pa-231 from 

residues from 1954-1955. U-233 major nuclide in 1958.

R R-169 U-234, 235, 238 Table 3 1960 1980 Low level counting room. 1980-present: U-238 separation research.

R R-198 U-234, 235, 238 Table 3 1956 1982 Numerous isotopes. Decontamination lab. Ac-227 hazard was secondary hazard to Pu-238. Am-241 

present 1956-1982.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 60 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 11 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

SW SW-128, 129, 

130, 134

U-234, 235, 238 Table 5 1970 1979 Support rooms. No processing. Th-230 was major radiological hazard.

SW SW-132 U-234, 235, 238 Table 5 1970 1979 Recovery of Pa-231 and Th-230 from Cotter concentrates. Actinium as oxide or nitrate.

SW SW-136, 137 U-234, 235, 238 Table 5 1964 1979 Provide access port to hot cell in SW-140. Actinium as nitrate and oxide. Concerns were Th-228, Ra226, Ac-227, Pa-231, U-234, U-235, and U-238.

SW SW-140 U-234, 235, 238 Table 5 1974 1979 Hot cell facility. Actinium processing and recovery of Pa-231 and Th-230 from Cotter concentrates. Ac227, Pa-231, Th-230 major nuclides of concern.

SW SW-149, 149a U-234, 235, 238 Table 5 1965 1968 All SW Building Nuclides. Decontamination lab. Tritium was major isotope of concern.

SW SW-22 U-234, 235, 238 Table 5 1970 1979 New Cave. Thorium-229 separated from U-233. Actinium as nitrate and oxide. Other Th and Pa 

projects including Cotter Concentrate extraction and U ore processing. U-232, 233, Th-228, Th-230.

WD, WDA WD-1, 8, 101, 

104

U-234, 235, 238 Table 7 1974 1979 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.

WD, WDA WD-1, 8, 101, 

104

U-234, 235, 238 Table 7 1949 1990 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.

WD, WDA WD-1, 8, 101, 

104

U-234, 235, 238 Table 7 1974 1979 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.

WD, WDA WD-Penthouse U-234, 235, 238 Table 7 1949 Present Filter bank for WD and WDA. All radionuclides are a hazard. Am-241 present 1949-present.

HH HH-5 U-235 Table 1 1955 1958 Separation of Pa-231 from 1955-1956. Decommissioned by 1958. Th-230 extraction from Airport Cake 

raffinate. Pa-231 and Th-230 are major nuclides.

T T-236 U-235, U-238, U239Table 6 1985 Present Dock. Tritium was primary hazard from 1985-present.

R R-169 U-235, 238 Table 3 1980 Present Low level counting room. U-238 separation research from 1980-present.

R R-171, 172 U-235, 238 Table 3 1961 1963 Ac as nitrate, oxalate, oxide. Ac-227 major hazard followed by Ra-226 and Th-228. U-235, 238 primary 

concern from 1961-1963.

WD, WDA WD-107 U-235, 238 Table 7 1976 1981 Ultra-filtration and reverse osmosis. Th-230 was primary isotope of concern in trace amounts only. Am241 present 1976-1981.

WD, WDA WD-9 U-235, 238, 239 Table 7 1973 Present Pumps for influent tritiated wastewater. Uranium as deuterides and tritides. Tritium was primary 

radiological hazard. Am-241 present 1973-present.

R R-155 U-235, U-238 Table 3 1956 1969 Microprobe Lab, 1956-1969. U-238 major radionuclide.

R R-159 U-235, U-238 Table 3 1956 1969 Numerous other isotopes. U-238 major radionuclide

HH HH-5 U-238 Table 1 1955 1958 Separation of Pa-231 from 1955-1956. Decommissioned by 1958. Th-230 extraction from Airport Cake 

raffinate. Pa-231 and Th-230 are major nuclides.

R R-105, 106 U-238 Table 3 1965 1978 Uranium deuteride and uranium tritide. Tritium gas was the major radiological hazard followed by 

uranium.

R R-108 U-238 Table 3 1948 1965 Uranium-deuteride and tritide. Tritium gas was the major radiological hazard followed by uranium.

R R-112 U-238 Table 3 1973 1993 U-Nat deuteride and tritide. Tritium hazard >> uranium hazard.

R R-116 U-238 Table 3 1960 1965 Numerous isotopes, U-238 as deuteride and tritide. Tritium hazard >> uranium hazard.

R R-128 U-238 Table 3 1987 Present Uranium as deuteride and tritide. Tritium hazard >> Uranium hazard. Am-241 used 1956-1962.

R R-134 U-238 Table 3 1986 Present U deuteride and tritide. Tritium primary radiological hazard. Low radionuclide inventory.

R R-168 U-238 Table 3 1958 1958 Physical properties of U-238 salts and oxide including fluorides. Major hazard was U-238.

R R-170 U-238 Table 3 1960 1964 U-238 mass spectroscopy.

SW SW-11, 12, 13, 

14, 15, 16

U-238 Table 5 1955 1955 Numerous nuclides. Th-232 extraction from monazite sand. Th-232, Th-228, Ra-224 major hazards. 

Pa-231 purification process. U-238 secondary. 
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 61 of 66

ATTACHMENT B

SECONDARY RADIONUCLIDES BY LOCATION AND PERIOD

Page 12 of 12

Building Rooms Isotopes

MD-22153 

reference

Start

year

End

year Comments

SW SW-11, 12, 13, 

14, 15, 16

U-238 Table 5 1968 Present Numerous nuclides. Th-232 extraction from monazite sand. Th-232, Th-228, Ra-224 major hazards. 

Pa-231 purification process. U-238 secondary.

SW SW-142 U-238 Table 5 1969 1985 Analytical laboratory. Radium as nitrate, carbonate, bromide. Metal tritides. Actinium as fluoride and 

oxalate. U as deuteride and tritide. Major nuclides Ra-226, Ac-227, Th-228. 

SW SW-146, 147 U-238 Table 5 1967 Present Uranium as deuteride, tritide. Tritium was radiological concern from 1967-1985.

SW SW-150, 152 U-238 Table 5 1968 Present Uranium as deuterides, tritides. Tritium was primary radiological hazard.

SW SW-208 U-238 Table 5 1965 Present Uranium as deuterides and tritides. Tritium was primary radiological hazard.

SW SW-210 U-238 Table 5 1968 1969 Uranium as deuterides and tritides. Tritium was primary radiological hazard.

SW SW-219 U-238 Table 5 1985 Present Uranium as deuterides and tritides. Tritium was primary radiological hazard.

SW SW-238, 240 U-238 Table 5 1968 Present Uranium-238 and tritium. Tritium was major hazard.

SW SW-8 U-238 Table 5 1960 1985 Primarily deuterium and tritium research. Tritium was primary radiological hazard.

T T-266, 267 U-238 Table 6 1988 Present Tritium engineering & development. Tritium was primary radiological hazard. Uranium as deuterides 

and tritides from 1988-present.

T T-274 U-238 Table 6 1986 Present Uranium as deuterides and tritides. Tritium was primary radiological hazard.

T T-37 U-238 Table 6 1985 Present Decontamination area. Provided access to other parts of building. Tritium was primary radiological 

hazard. Uranium as deuterides and tritides.

T T-48, 49, 50 U-238 Table 6 1983 Present Glovebox labs and access corridor. Tritium was primary hazard. Uranium as deuterides and tritides.

T T-59 U-238 Table 6 1985 Present Tritium thermal diffusion facility. Tritium was primary hazard. Uranium as deuterides and tritides.

T T-61, 61a U-238 Table 6 1973 Present Uranium as deuteride and tritide. Tritium recovery.

WD, WDA WD-1, 8, 101, 

104

U-238 Table 7 1955 1955 Liquid radioactive waste processing. "Low level waste" from SM building. Am-241 present 1949-1990.
Document No. ORAUT-TKBS-0016-5 Revision No. 00 Effective Date: 09/09/2004 Page 62 of 66

ATTACHMENT C 

CODING INSTRUCTIONS FOR URINALYSIS RECORDS

The following coding instructions are summarized in Table 5-9 for urinalysis records reported on 

Urinalysis Cards, Weekly Reports, and 24-hour reports at Mound.

Table C-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 Disintegrations / minute

DM-24 Disintegrations / minute / 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. 00 PC-1 Effective Date: 03/13/2009 Page 63 of 66

ATTACHMENT D 

UNCERTAINTY ASSESSMENT

Page 1 of 4

TABLE OF CONTENTS

Section

D.0 UNCERTAINTY ASSESSMENT ............................................................................................

Page

63

D.1 CONTAMINATION OF SAMPLES .........................................................................................63

D.2 ASSIGNMENT OF GROSS ALPHA RESULTS AS PLUTONIUM .......................................... 63

D.3 MISSING BIOASSAY RECORDS ..........................................................................................63

D.4 ASSUMPTION OF EQUILIBRIUM .........................................................................................64

D.5 PARTICLE SIZE DISTRIBUTIONS ........................................................................................64

D.6 SOLUBILITY AND LUNG CLEARANCE TYPE ......................................................................64

D.7 RADIONUCLIDE USE BY LOCATION...................................................................................65

D.8 ANALYTICAL PROCEDURES ...............................................................................................65

D.0

Uncertainties in estimating worker dose commitments from bioassay data result from uncertainty in 

monitoring measurements and assessment of intake (ICRP 1998). Uncertainties for the bioassay 

measurements were not stated in the records. 

UNCERTAINTY ASSESSMENT

D.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.

D.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 allocated improperly among the organs. 

D.3 MISSING BIOASSAY RECORDS

The record of urinalysis data is generally complete for the Mound site. Some data are available for 

the Monsanto Chemical Company facilities operated prior to the beginning of operations at the Mound 

laboratory as well. Reportable worker exposures in the early years could also have been undetected 

due to low recovery caused by plating on sample containers before sample analysis. Intakes for 

missing bioassay results should be estimated using the MDAs reported in Table 5-5 for spot and 24-hr 

samples. 

Tritium bioassay data were not reported from 1948 to 1956. The dose reconstructor was advised to 

use a value of 1000 mrem corrected to 340 mrem ICRP68 dosimetry for unmonitored workers. This 

value was assumed to be the approximate upper 95 percentile and considered a claimant-favorable 

dose for all workers. Uncertainty of extrapolating doses from 1959–1962 to the 1947–1956 time

period are considered minor and compensated by use of the upper 95 percentile dose estimate. 

Doses should not be assigned to administrative or other non-radiological workers. 
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D.4 ASSUMPTION OF EQUILIBRIUM

Dose uncertainty related 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 where acute intakes of unspecified feed materials such as 

raffinate, tailings, and various production residues occurred. For instance, raffinate produced by 

sulfuric acid leaching of ore would have a relatively low concentration of radium compared to 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 that workers were exposed to 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) 238U chain and possible cases of disequilibria are not 

involved, and (2) both 223Ra and 224Ra have short half-lives in days and therefore quickly achieve 

equilibrium with their respective parent radionuclides. 

D.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 involved. 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 1994, 1995). Alpha emitters with larger 5-μm particles and lung 

clearance types S or M tend to have smaller inhalation effective dose coefficients than do smaller 

1-μm particles. Use of a smaller AMAD of 1 μm increases worker dose following inhalation of type M 

239Pu by a factor of 1.5. 

Of the secondary radionuclides, the difference in dose conversion factors for 1-μm versus 5-μm 

particles with clearance types S or M was greatest for 231Pa by a factor of 2. The opposite trend was 

seen for type F alpha emitters, where 5-μm particles of uranium or 227Ac have inhalation effective 

dose coefficients that are 20% larger than corresponding dose coefficients for 1-μm particles. 

Therefore, uncertainty in dose coefficients related to particle size distribution is dependent on lung 

clearance type and is comparatively small based on data provided 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.

D.6 SOLUBILITY AND LUNG CLEARANCE TYPE

Uncertainties in ICRP Publication 68 dose coefficients due to lung clearance kinetics are much 

greater than uncertainties related to particle size (ICRP 1995). Solubility of a radionuclide is one of 

the most important parameters affecting internal dose (NIOSH 2002). Lung clearance types should 

therefore be assigned on a case-by-case basis if there is sufficient bioassay data for radionuclides to 

surmise lung clearance behavior for acute exposures. Assigning a radionuclide to type S when it is 

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. 
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D.7 RADIONUCLIDE USE BY LOCATION

The uncertainty associated with radionuclide use by location was almost entirely documented by King 

(1995) and was considered minor compared with other sources of uncertainty. Radionuclide use by 

location has been sufficiently documented for primary radionuclides (Table 5-1). There is little 

uncertainty associated with documentation for tritium, 210Po, and the plutonium isotopes and their 

processes by building. Protactinium-231 had limited use and, therefore, there is 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 to those for natural uranium. Isotopic radium and thorium data were also limited. Almost 

no information was available for 244Cm and 237Np other than as a contaminant in the Savannah River 

Plant source material. 

Information on the location and use of radioactive gases was primarily only available for 222Rn (radon) 

from 1979 to the present. Very limited information was available for 220Rn (thoron) and 219Rn 

(actinon), and the measurements were questionable due to excessive decay times before analysis. 

However, measurements conducted in the underground tunnel in 1979 confirmed the presence of 

222Rn at 88,000 pCi L-1, 220Rn at 28,000 pCi L-1, and 219Rn at 640,000 pCi L-1. 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 4.0 s and direct deposition of actinon daughters 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 compensated for this uncertainty.

D.8 ANALYTICAL PROCEDURES

Uncertainties associated with analytical procedures used throughout the history of bioassay 

measurements at Mound are significant. Early analytical procedures resulted in large errors and 

MDAs relative to required regulatory dose limits. Reductions in regulated body burdens for 210Po 

required modifications to improve analytical sensitivities. Positive bioassay detections would thus 

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.
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Additional errors occurred in the early years as a result of the inability to accurately assign gross 

alpha activity determined with low background proportional counting to specific alpha-emitting 

radionuclides. Gross alpha activity was thus 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 detection limits.