ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 45

DOE Review Release 01/10/2013

Document Title:

Exotic Radionuclides at the Savannah River 

Site

Document Number: ORAUT-RPRT-0054

Revision: 01

Effective Date: 01/03/2013

Type of Document: Report

Supersedes: Revision 00

Subject Expert(s): Edward D. Scalsky and Donald E. Bihl

Approval: Signature on File Approval Date: 12/18/2012

Edward D. Scalsky, Document Owner

Concurrence: Signature on File Concurrence Date: 12/17/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 12/18/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 12/17/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/03/2013

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 2 of 45

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/20/2012 00 New document initiated in response to questions raised by SC&A 

that not all radionuclides have been identified and considered in the 

analysis of personnel exposures at the Savannah River Site. 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Edward D. Scalsky.

01/03/2013 01 Revision initiated to add new information pertaining to 147Pm in 

Attachment B, 249Bk and 254Fm in Attachment C, and 171Tm in 

Attachment D. Responds to formal NIOSH review comments; For 

information added a sentence at the end of the first paragraph in 

Section 5.0; deleted the last sentence in the plutonium paragraph; 

added information at the end of the uranium paragraph related to the 

types of DU analysis used between 1986 through 1994 and for DU 

and EU after 1994; replaced the two paragraphs discussion in the 

neptunium section with a new paragraph on the separation of 

neptunium from the actinides and uranium and on the analysis of 

neptunium; added additional references to the information in the 

section on 90Sr; clarified the references for the whole body counting 

information and added a discussion on the counting geometry and 

the energy regions of interest and added the references into the 

reference section; revised the title of Table A-1 to reflect the doses 

were due to inhalation from handling the 1.8 x 10-6 g 99Tc; added the 

words “estimates” and “handling” for clarification that estimated 

inhalation doses are a result of handling the 99Tc. There were no 

review comments incorporated as a result of formal internal review. 

Training required: As determined by the Objective Manager. 

Initiated by Donald E. Bihl and Edward D. Scalsky.
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 3 of 45

TABLE OF CONTENTS

SECTION TITLE PAGE

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

1.0 Executive Summary ....................................................................................................................6

2.0 Introduction ................................................................................................................................. 6

3.0 Radionuclide Production.............................................................................................................6

4.0 Savannah River Laboratory Operations....................................................................................16

5.0 Bioassay Programs...................................................................................................................18

5.1 In Vitro Bioassay ...........................................................................................................18

5.2 In Vivo Bioassay............................................................................................................22

6.0 Summary ...................................................................................................................................24

References ...........................................................................................................................................25

ATTACHMENT A, TECHNETIUM-99...................................................................................................34

ATTACHMENT B, PROMETHIUM-147................................................................................................36

ATTACHMENT C, BERKELIUM-249 AND FERMIUM-254 ..................................................................38

ATTACHMENT D, THULIUM-171 ........................................................................................................43

ATTACHMENT E, AMERICIUM-242m................................................................................................. 45

LIST OF TABLES

TABLE TITLE PAGE

3-1 Examples of general samples irradiated during the HFDP.........................................................8

3-2 Samples irradiated during the HFDP for selected months........................................................10

3-3 Radionuclides discussed for possible irradiation ......................................................................13

3-4 Isotopes identified in SRS documents from 1954 through 1971...............................................14

4-1 Positive bioassay results...........................................................................................................17

5-1 Selected isotopes bioassay methods........................................................................................20

A-1 Potential inhalation committed organ doses from handling 1.8 g 99Tc......................................35

E-1 Activity of 242Cm per unit daily intake of 242mAm ........................................................................45
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 4 of 45

ACRONYMS AND ABBREVIATIONS

AHIB alpha-hydroxyisobutyrate

AIMS Automated Inventory Management System

ANL Argonne National Laboratory

BNL Brookhaven National Laboratory

Ci curies

cm centimeter

cpm counts per minute

d day

DNA Delayed Neutron Analysis

DU depleted uranium

EU enriched uranium

g gram

HDEHP di,2-ethylhexylophosphoric acid

HFDP High Flux Demonstration Program

hr hour

in. inch

IPHS Isotopic Power and Heat Sources

kCi kilocurie

keV kiloelectron-volt

kg kilogram

LRL Lawrence Radiation Laboratory

mCi millicurie

MCi megacurie

mg milligram

mm millimeter

MPLB maximum permissible lung burden

mrem millirem

n neutron

nCi nanocurie

ng nanogram

NIOSH National Institute for Occupational Safety and Health

NMMSS Nuclear Material Management and Safeguards System

NTA nuclear track emulsion, type A

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

pCi picocurie

ppm parts per million
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 5 of 45

RC Radiation Control

s second

SC&A Sanford Cohen & Associates

SRDB Ref ID Site Research Database Reference Identification (number

SRL Savannah River Laboratory

SRP Savannah River Plant

SRS Savannah River Site

TIOA tri-isooctlyamine

TRAMEX transuranic metal extraction

UCB University of California, Berkeley

UCL University of California, Livermore

W(t) watt-thermal

WTD Works Technical Department

yr year

μCi microcurie 

μg microgram
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 6 of 45

1.0 EXECUTIVE SUMMARY

The first reactor (R-Reactor) at the Savannah River Plant (SRP) – now know as the Savannah River 

Site (SRS) – started operation on December 28, 1953 and shut down in June 1964. The K-, P-, and 

L-Reactors all started in 1954 and operated until 1988; the L-Reactor was shut down in 1968 and 

restarted in 1985. The C-Reactor operated from 1955 through 1985.

During the years the SRS reactors were operating, there were many nuclides that were irradiated. 

Originally, the irradiations were in support of the nuclear weapons program. However, when military 

requirements slowed down, other missions were sought. Some missions included the production of 

transplutonium elements; development of isotopic power and heat power sources; and development 

of sources for food sterilization, medical uses, radiography, and irradiating samples for national 

laboratories, universities, and other countries in support of their research.

In light of the many radionuclides that resulted from these irradiations, Sanford Cohen & Associates 

(SC&A) questioned if all radionuclides were identified and considered in the analysis of personnel 

exposures (SC&A 2010). According to the SC&A report, the main issues relate to the production, 

irradiation in reactors, and processing (if any) of target materials for the production of special 

radionuclides to clarify the dose reconstruction issues associated with them. SC&A classifies these 

special radionuclides as the “exotic” radionuclides that were produced in the C- and K-Reactors 

during the high flux years of operation (i.e., 1965 to 1970). This report attempts to identify the 

radionuclides, the bioassay methods that were used to evaluate the potential exposures to personnel 

handling and processing the radionuclides and, if possible, the approximate quantities that were 

available at SRS.

2.0 INTRODUCTION

In addition to the materials that were irradiated in the SRS reactors for the weapons program, a great 

deal of material was irradiated for use in commercial, industrial, and medical applications. This report 

discusses the nuclides that were irradiated and processed for various periods in the SRS facilities. 

The bioassay methods that were used to evaluate the potential exposure to these radionuclides are 

also discussed.

Section 3.0 describes production of exotic radionuclides, and Section 4.0 discusses Savannah River 

Laboratory (SRL) operations. Section 5.0 discusses bioassay practices, and Section 6.0 summarizes 

the finding of this report. Attachments A through E provide details on production of 99Tc, 147Pm, 249Bk 

and 254Fm, and 242mAm, respectively.

3.0 RADIONUCLIDE PRODUCTION

The term “exotic radionuclides” was used first in the SC&A report (SC&A 2010). SC&A stated its 

concerns that “exotic radionuclides” that are not explicitly addressed as matrix issues for which 

NIOSH is developing coworker models. Radionuclides excluded from analysis include 232Th, 241Am, 

244Cm, 252Cf, 237Np, fission and activation products,210Po, recycled uranium, natural uranium, 239/240Pu, 

and 241Pu. SC&A included among the “exotic radionuclides” those materials that were irradiated for 

offsite customers during the High Flux Demonstration Program (HFDP). However, for purposes of 

this report any radionuclide with some exceptions (e.g., tritium, thorium) that were produced as a 

result of irradiation in the SRS reactors is considered to be an exotic radionuclide, some of which are 

only briefly mentioned.
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 7 of 45

The SC&A report (SC&A 2010) identified the following radionuclides of concern:

• Alpha emitters: 

– Americium-243

– Protactinium-231

– Plutonium-238

– Plutonium-242

– Uranium-233

• Beta emitters: 

– Cesium-137

– Europium-153

– Iodine-131

– Iridium-192

– Promethium-147

– Technetium-99

– Thulium-170

In the early 1950s, five heavy-water reactors (R, P, L, K, and C) were constructed at SRS. All five 

reactors were operating by 1955. Reactor shutdown began in 1964 with R-Reactor. C-Reactor was 

shut down in 1987; L-Reactor was shut down in 1968, restarted in 1985, and shut down again in 

1988. K- and P-Reactors operated with few interruptions until 1988. A restart of K-Reactor began in 

1991 but was not completed (Till et al. 2001). None of the reactors are currently operational.

Originally, the primary function of the SRS reactors was the production of plutonium and tritium for 

nuclear weapons. However, even in the early 1950s, the reactors were producing radioisotopes for 

nonweapons uses, which continued in the 1960s with the reduction of the military mission at SRS 

(Reed et al. 2002). Depleted uranium (DU), 237Np, and lithium targets were irradiated to produce

239Pu, 238Pu, and tritium, respectively. The targets were irradiated and transferred to the Tritium 

Facilities, H-Area Separations (238Pu), and F-Area Separations (239Pu) for processing. Spent reactor 

fuel was sent to H-Area Separations for processing. The reactors were also used to irradiate targets 

that produced transplutonic elements such as curium and californium.

Between 1955 and 1967 large quantities of special radionuclides were produced in the SRS reactors 

for various campaigns. For example, a series of campaigns produced 66 MCi of 60Co for Oak Ridge 

National Laboratory (ORNL), the U.S. Bureau of Mines, the U.S. Army Food Process Development 

Irradiator, Japan Atomic Energy Research Institution, Brookhaven National Laboratory (BNL), and 

Marine Products Development. SRS also produced 630 kg of 233U from irradiation of thorium for the 

thorium breeder reactor program. In addition, 275 kg of 238Pu from irradiation of 237Np, 600 g of 210Po, 

1,700 Ci of 152Eu, and 370 MCi of 170Tm were produced. Other radionuclides produced during this 

period and in subsequent years included more 210Po, 121Sn, 139Ce, 166Ho, and 41Ca for BNL, the 

University of Illinois, and Denmark (Bebbington 1990).

From 1965 through 1970, C- and K-Reactors were modified to produce a very high flux of neutrons for 

production of unusual (e.g., 170Tm, 171Tm, 46Sc) radionuclide sources. The HFDP began in C-Reactor 

in February 1965 and lasted 1 year. The HFDP in K-Reactor began in August 1969 to produce 252Cf. 

A peak flux of 7 × 1015 n/cm2

-s was achieved, and 2.1 g of 252Cf were produced before termination of 

the production campaign in November 1970. A second Californium Production Campaign was 

planned but never executed because the market for 252Cf never materialized as expected (Boswell 

2000). The target material in the C- and K-Reactors consisted primarily of 242Pu. In addition, three 1-

in.-diameter thimbles contained 243Am, 244Cm, and 150 target materials of 66 elements for nine 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 8 of 45

universities and laboratories (Boswell 2000). Table 3-1 lists the target materials that were used to 

supply irradiated samples to Argonne National Laboratory (ANL), ORNL, the Lawrence Radiation 

Laboratory (LRL) facilities at University of California, Berkeley (UCB) and the University of California, 

Livermore (UCL) (Smith et al. 1965). There were additional targets; for example, one tritiumproducing core (Mark VI) was designed with a removable inner lithium target so fuel could be 

irradiated to a very high exposure. The high exposure depleted 70% of the 235U and thus 

concentrated 236U, which was formed from every sixth neutron capture in 235U. Irradiation of the 236U 

formed 237Np, which was the target material for the production of plutonium (Boswell 2000). In 

addition, the irradiation of the target material 242Pu, after a succession of neutron captures and beta 

decays, successively produced other target materials – 242Pu, 243Pu, 243Am, 244Am, 244Cm, 245Cm, 

246Cm, 247Cm, 248Cm, 249Cm, 249Bk, 250Bk, 250Cf, 251

Cf, and 252Cf (Reed et al. 2002). There were also 

special irradiations during both the HFDP and the Curium I campaign to produce test quantities of 

170Tm, 171Tm, 46Sc, and 204Tl.

Table 3-1. Examples of general samples irradiated during the HFDP.

a

Sample Laboratory Sample Laboratory Sample Laboratory

Si-30 ANL Yb-168 ORNL Os UCL

S-33 ORNL Tm-168 UCB, UCL W UCL

Ca-40 ORNL, ANL Er-170 UCB Re UCL

Ca-48 ANL Yb-176 ANL, UCB, UCL Be ANL

Fe-54 ORNL Lu-176 UCB Be3N2 ORNL

Fe-58 ORNL, ANL, UCB Hf-177 UCB Xe UCL

Ni-58 ORNL, ANL Hf-179 ANL, UCL, UCB TiO2 ANL

Ni-62 ORNL Hf-180 ANL, UCL, UCB

Nb-93 ORNL Ta-180 UCB

Nb-96 ANL Ta-181 UCB

Ag-107 UCB W-186 ANL, UCL, UCB Additional isotopesb

Ag-108 UCB Re-187 UCL, UCB Zn-67 ANL

Sn-118 ORNL, UCB, ANL Os-180 UCB Te-124 ANL

Sn-124 ANL Os-182 ANL, UCL, UCB Lu-175 UCL

Ba-130 UCB Bi-209 UCB Be-9 ANL

Sm-144 ORNL, UCB Am-241 ANL, UCL

Nd-146 ORNL Cm-244 ANL, UCL, UCB Sn-120 ANL

Nd-150 ANL Bk-249 ANL

Sm-150 UCB, UCL Cf-252 ANL, UCL, UCB Cd-160 UCL

Sm-154 ORNL Es-254 UCB Am-243 UCL

Dy-156 UCB, UCL LiCl UCB

Dy-158 ANL MgO+Cr-50 UCL

Tb-158 UCB, UCL KPO3 UCL

Gd-160 ANL, UCL, UCB Ar UCL

Er-162 UCB, UCL RbCl ORNL

Dy-164 ANL, UCB Hf ANL

Ho-165 UCB, UCL HFO2 UCL

a. Source: Smith et al. 1965.

b. Additional isotopes from Table 3-2.

Some development work was done mostly with cold Tm2O3 (DuPont 1974, p. 12), and a little work was 

done with 170Tm, 171Tm, 46Sc, and 204Tl (DuPont 1974, pp. 7–8). Scandium-46 is a gamma emitter. 

Thallium-204 is a beta-only emitter. Thulium-170 and -171 have low-yield, low-energy gammas that 

might show up in chest counting if someone was looking closely, but are mostly beta emitters. 

Thulium oxide was characterized as target material for irradiation to form 170Tm, with potential 

applications for short-term missions in space (128-day half-life). As with cobalt, the irradiated product 

is useful without chemical processing. Conditions were established for fabricating wafers, and the 

properties of Tm2O3 were studied to prepare and irradiate it to yield a radioactive product suitable for 

use in heat sources. There was some activity that involved breaching of the source or chemistry on 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 9 of 45

the 170Tm and 171Tm. These were small research and development experiments before actual largescale production of the sources. There was some breakage of 170Tm oxide wafers in the early 

experiments. Metal targets did not have that problem. There was no evidence of ongoing, largescale breaching or chemistry on the 170Tm or 171 Tm sources. Thulium-170 was mentioned as a 

possible heat source early in the program; it emits a gamma ray with about 3% yield. Thulium-171 

only emits a beta in sufficient abundance for bioassay.

SC&A mentioned 147Pm and 192Ir as two radionuclides of concern (SC&A 2010). The National 

Aeronautics and Space Administration’s interest in aged 147Pm as a source of isotopic heat for space 

missions led to a study of the quantity and quality of 147Pm that could be produced annually at SRS. 

The recovery of 147Pm from current high-activity waste was also studied. The major source of 147Pm 

was fission product waste; the quantity available after aging to allow for decay of associated 148Pm 

would have been about 1,700 W(t) per reactor-year of operation. Product quality from current waste, 

as represented by a low ratio of 148Pm to 147Pm, would presumably have been the same as that 

produced at the Hanford Site (about 5 × 10-7). Promethium-148 emits gamma rays that could require 

extra shielding. Irradiation of naturally occurring neodymium would produce only 50% as much 147Pm 

(DuPont 1974). No further mention of 147Pm has been found in this historical reference. DuPont

(1974) mentions that iridium was experimented with but indicates that this experimentation was with 

several inactive iridium capsules. The report makes no mention of 192Ir. 

Table 3-1 shows a list of materials that were irradiated for the U.S. Atomic Energy Commission

laboratories only (i.e., ANL, ORNL, UCB, and UCL). Some of the materials were irradiated for the 

duration of the HFDP (i.e., from February to November 1965). Others were irradiated for single fuel 

cycles of about 1 week (Smith et al. 1965). Table 3-1 does not include isotopes that were irradiated 

for BNL, the University of Illinois, or Denmark, such as 121Sn, 138Ce, and 139Ce, or those irradiated for 

all offsite customers.

Table 3-2 lists some of the specific target elements and compounds that were irradiated for other

customers during the indicated months and years. Many of the samples were repeated in subsequent 

months; for example, March and May 1966 and March 1967 had many repeat samples. There were 

five shipments during January 1966. The 257Fm to LRL and the 2 μg of 252Cf to ANL are listed in the 

table. The other shipments, not listed, included 380 mg of iridium to ORNL, 200 mg of calcium to 

Berkeley, and 1 μg of 253Es to ANL.

The procedure for handling these samples is described by Gray (2006) as follows:

All samples containing the material to be irradiated for off-site customers were in 

welded closed aluminum cans. They came to SRP in that condition. They were placed 

in a basket and the basket lowered into the spent fuel storage basin where the deep 

water provided shielding for handling radioactive components removed from the 

reactor. The research samples were placed in small perforated tubes about 1 inch in 

diameter; maybe five or so samples per a 12" tube. The tubes were then placed in a 

larger (about 1 3/8ths inch diameter) perforated tube about 21 feet long, called a 

thimble. The reactor had many of these thimbles for general use with safety rods, 

instruments, etc. Three of them near the center of the reactor were dedicated to my 

[Gray’s] use for research samples. When it came time for either insertion or removal 

(or both) of a sample, the thimble was removed by remote machine from the reactor 

and placed in the basin. It was moved remotely under water to a location where I could 

have it turned upside down and the tubes unloaded onto a well-lit work table about 12 

feet down. I sorted through the tubes and found the one from which a sample would 

be unloaded (or a new one loaded); all under water. The process was reversed to put 

the whole thing back together and return the loaded thimble to the reactor for the next 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 10 of 45

Table 3-2. Samples irradiated during the HFDP for selected months.

a

August 1965b January 1966c March 1966d May 1966e March 1967f

Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents

ORNL 280 mg natural 

iridium

SRP-SRL 1.4 g Co-59 LRL 11 mg Ag-107 LRL 11 mg Ag-107 LRL 11 mg Ag-107

SRP-SRL 1.4 g Co-59 ANL 40 mg Cm-244 LRL 1.52 g LiCl LRL 1.52 g LiCl LRL 1.52 g LiCl

ANL 40 mg Cm-244 ANL 2 μg Cf-252 LRL 217 mg MgO 

and Cr-50

LRL 217 mg MgO 

and Cr-50

LRL 217 mg MgO 

and Cr-50

ANL 2 μg Cf-252 LRL 240 mg 

Cm-244, 

20 mg Cm-246

LRL 10 mg Fe-58 LRL 10 mg Fe-58 LRL 10 mg Fe-58

LRL 400 mg Cm-244 ANL 77.2 mg Be-9, 

0.6 mg Ca-48, 2 

mg S-30

LRL 2 mg Dy-156 LRL 2 mg Dy-156 LRL 2 mg Dy-156

LRL 450 mg 

Cm-244; 

40 mg Cm-246

ANL 3.95 mg Fe-58, 

1.4 mg Sn-124, 

5.26 mg Hf-180

LRL 9 mg Hf-179 LRL 9 mg Hf-179 LRL 9 mg Hf-179

LRL 60 mg Cm-244; 

25 mg Cm-246

LRL 1.9 mg Cf-252 LRL 10.3 mg Hf-180 LRL 10.3 mg Hf-180 LRL 10.3 mg Hf-180

ANL 240 mg 

Cm-244; 

20 mg Cm-246

LRL 231 mg Gd-160 LRL 58 mg Sm-150 LRL 58 mg Sm-150 LRL 58 mg Sm-150

ANL 77.2 mg Be-9; 

0.6 mg Ca-48; 

2 mg Si-30

48 g Bi-209 GT 200 mg 27% 

Pb-204 metal, 

1 mg PbO 

(Pb-204 

enriched to 

99.7%)

GT 200 mg 27% 

Pb-204 metal, 

1 mg PbO 

(Pb-204 

enriched to 

99.7%)

GT 200 mg 27% 

Pb-204 metal, 

1 mg PbO 

(Pb-204 

enriched to 

99.7%)

LRL 3.95 mg Fe-58; 

1.4 mg Sn-124; 

5.26 mg Hf-180

LRL 48 g Bi-209 BNL 0.5 g C 

(diamond in 

graphite)

LRL 1.5 mg Pu-240, 

0.1 mg Pu-244

LRL 1.5 mg Pu-240, 

0.1 mg Pu-244

ANL 1.9 mg Cf-252 LRL 99 mg Yb-176 ANL 40 mg Cm-244 LRL 48 g Bi-209

(2.5 in.)

LRL 48 g Bi-209

LRL 0.8 g Hf-180 UI 8.3 mg 

Ba(NO3)2, 

12.5 mg CeO2, 

600 mg Sn-118

LRL 1.5 mg Pu-240, 

0.1 mg Pu-244

LRL 48 g Bi-209

(2.54 in.)

LRL 48 g Bi-209

LRL 231 mg Cd-160 LRL 1 g Am-243 LRL 48 g Bi-209 LRL 10.8 mg Hf-177 LRL 10.8 mg Hf-177

LRL 48 g Bi-209 LRL 1 g Am-243 LRL

ANL 

240 mg 

Cm-244; 

20 mg Cm-245

LRL 10 mg Os-192 LRL 10 mg Os-192
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 11 of 45

August 1965b January 1966c March 1966d May 1966e March 1967f

Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents

LRL 48 g Bi-209 LRL 5.7 g Lu-175 LRL 48 g Bi-209 LRL 10.9 mg Er-162 LRL 10.9 mg Er-162

LRL 99 mg Yb-176 ORNL 216 mg Sm2O3 

(enriched to 

94.5% Sm-144) 

and 7.828 mg 

type 302 SS 

monitor

LRL 10.8 mg Hf-177 LRL 9.6 mg W-186 LRL 9.6 mg W-186

SRP 13.47 mg Al ORNL 700 mg gold foil 

and 7.61 mg 

type 302 SS 

monitor

LRL 10 mg Os-192 LRL 9.74 mg Gd-160 LRL 9.74 mg 

Gd-160

UI 8.3 mg 

Ba(NO3)2; 

12.5 mg CeO2; 

600 mg Sn-118

ORNL 418 mg La2O3

and 7.718 mg 

type 302 SS 

monitor

LRL 10.9 mg Er-162 LRL 9.74 mg Re-187 LRL 9.74 mg 

Re-187

ORNL 280 mg natural 

iridium

ANL 1 g Sn-120 LRL 9.6 mg W-186 LRL 7.65 mg Lu-176 LRL 24 mg Sm-144

LRL 1 g Am-243 LRL 40 mg Ba-130 

as BaCO3

LRL 9.74 mg Gd-160 LRL 24 mg Sm-144 LRL 50 mg Sn-118

LRL 1 g Am-243 LRL 40 mg Ba-130 

as BaCO3

LRL 9.74 mg Re-187 LRL 9.6 mg Ag-109 GT 77 g lead

LRL 1 g Am-243 LRL 40 mg Ba-130 

as BaCO3

LRL 24 mg Sm-144 LRL 50 mg Sn-118 SRL 1 g Er2O3

enriched in 

Er-170

LRL 1 mg Hg-203 

(as nitrate)

LRL 1.5 mg Pu-240, 

0.1 mg Pu-244

LRL 9.6 mg Ag-109 ORNL 100 mg Te-128 

(99.16% 

enriched)

ANL 60 mg Te-126 

enriched 

(98.69%)

SRL 90 mg Nd2O3

(enriched Nd146) 

BNL 0.5 g C 

(diamond in 

graphite)

LRL 50 mg Sn-118 GT 77 g Pb Livermore 460 mg 

Cm-244, 

90 mg Cm-246, 

9 mg Cm-248

BNL 0.5 g C 

(diamond in 

graphite)

ORNL 100 mg Te-128 

(99.16% 

enriched)

NC 500 mg SnO2 

enriched 99% 

Sn-120

BNL 0.5 g C 

(diamond in 

graphite)

ORNL 100 mg Te-128

(99.16% 

enriched)

Livermore 300 μg Cf-252

ANL Ca-40, Ca-42, 

Ni-58, Ni-62,

GT 77 g lead ANL Er-162, Yb-176, 

Mo-92
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 12 of 45

August 1965b January 1966c March 1966d May 1966e March 1967f

Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents Sponsor Slug contents

ANL Zn-67, Te-124, 

Pb-204d,g

ANL Fe-54, Yb-168, 

Cr-50, Os-192

ANL Se, Yb-176g ANL Nd-150, Ti-50, 

W-186, Yb-176

ANL Nd-150,

Os-190g

ANL Nb, Ce-136, 

Cf-249, Np-236

ANL U-236, Pa, Nig ANL 90 mg Cm-244, 

30 mg Cm-246

ANL Np-236, Cf-249, 

Th-229g

LRL 5 mg Ln-176 

(70% enriched)

LRL 200 mg calcium ORNL 103.2 mg WO3

(97.2% 

enriched W186)

LRL 0.7 mg platinum 

metal

LRL 1 mg Hg-202 in 

quartz capsule

LRL Fm-257 rabbit 

sample in Al 

capsule (10 μg)

ORNL 8.5 mg type 302 

SS monitor

ORNL 150 μg 236U2O3; 

8.5mg type 302 

SS monitors

a. GT = Georgia Tech; SS = stainless steel; UI = University of Illinois. 

b. DuPont (WTD Aug 1965). (For citations of this form, see the Monthly and Quarterly Reports section at the end of the References list.)

c. DuPont (WTD Jan 1966). 

d. DuPont (WTD Mar 1966). 

e. DuPont (WTD May 1965). 

f. DuPont (WTD Mar 1967). 

g. The sample mass and chemical form of one or more nuclides were unreadable in the referenced document. 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 13 of 45

cycle. Of the 15-18 hours for shut down, I had about 3-5 hours in the middle of it to do 

my work.

A removed research sample, without any opening of the welded-closed aluminum 

cans, was placed directly in a shielded shipping cask. The cask lid was put in place 

and sealed (or locked) down for transport. As the cask was removed from the basin it

was washed down. Then it was placed on a truck for transport, usually to the Atlanta 

airport, for a flight to the outfit that sent it to us originally for irradiation. Many samples, 

because of short half lives of the desired product, were returned in well less than 24 

hours. No samples were opened at SRP [Savannah River Plant]. 

As discussed by Gray (2006), all samples that were to be irradiated for external customers were in 

welded-closed aluminum cans. They came to SRS in that condition. After irradiation they were 

prepared under water for shipment, so the potential for intake was minimal. Similarly, because the 

sources were inspected and loaded into casks under water, the potential for external exposure was 

reduced. As noted above, Table 3-2 lists specific samples that were irradiated for the sponsors listed 

in the odd numbered columns for the indicated month and year. As can be seen in the table, the 

samples were generally small for research purposes at the sponsor’s facilities. However, there are 

five samples in the 5-month period listed in the table that were transferred to SRS-SRL. These 

samples would have been handled using SRS procedures. Based on the limited number of 

radionuclides transferred, it would appear that such irradiations were not routine. 

In addition to the radionuclides and compounds listed in Tables 3-1 and 3-2, a request was made of 

the Isotopes Development Center at SRS to look at the possibility of using irradiation facilities as fully 

as possible for the production of unusual or quite large amounts of isotopes for the potential benefit to 

research and commerce. In response to that request, Rupp (1969) discussed some of the 

characteristics of the radionuclides in Table 3-3, the target materials or compounds that would be 

used to produce some of them, and difficulties that might be encountered. He noted that some of 

these radionuclides were already being produced. As noted above in this section, these radionuclides 

have been identified by SC&A (2010) as being of concern for dose reconstruction. The entire list 

discussed in Rupp (1969) is included in Table 3-3. No documentation has been found indicating that 

program was ever pursued.

Table 3-3. Radionuclides discussed for possible irradiation (Rupp 1969). 

Plutonium-238 Thulium-170 and 171 Calcium-41 Barium-133 Samarium-154

Americium-242m Potassium-40 Iron-55 Lanthanum-137 Samarium-155

Cesium-134 Argon-38 (stable) Nickel-59 Terbium-157 Europium 155

Europium-152 Chlorine-36 Cadmium-109 Holmium-166m

Europium-154 Argon-39 Cadmium-113m Samarium-151

In addition to the information above, a review of monthly reports was conducted to identify the 

radionuclides that were being irradiated or processed at SRS from 1954 through 1971. Table 3-4 lists 

the results of this review (Taulbee 2011). Some of these isotopes were part of the production process 

and many were shipped to off-site customers.

The cells containing an X indicate that the isotope was present several times during the year. Some 

isotopes were present in every month of the year, others only during several months. Those identified 

by an L or M indicate they were present on a limited basis during 1 or 2 months during the year. The 

letters represent the potential for exposure to that radionuclide [i.e., either low (L) or medium (M) 

during that period]. This represents the highest estimated exposure potential for those radionuclides. 

Because the radionuclides indicated by an X appeared many times during the year, an indication of 

their exposure potential could not be identified in the table. Most were in the low to medium range but 

others, such as 241Am, 252Cf, 244Cm, and 237Np, were often listed as a high exposure potential. The 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 14 of 45

Table 3-4. Isotopes identified in SRS documents from 1954 through 1971. 

Isotope 1954 1955 1956 1957 1858 1959 1960 1961 1962 1963 1964 1965 1966 1967 1968 1969 1970 1971

Am-241 X X X X X X

Am-242ma #b Mc

Am-243 X X X X X X X X X X

Ala Ld

Ba-133a L

Bi-209a L

Bk-249 X X X X X X

Bk-250a L M

Ce-144a X X X X X

Cf-252 X X X X X X X X X

Cm-244 X X X X X X X X X X X

Cm-248f f

Co-59 X X

C0-60 X X X X X X X

Cs-137a X X X

Er-170 X X X

Es-253 X X X X

Es-254 X X X X

Es-255 L

Eu-152a L L

Eu-154a L L

Fm-254a M

Flux wire L

Ir-192a L

Nd-148a L

Np-237a X X X X X X X X X X X X X X X

Np-238a M

Np-239 L

Pa-231a M

Pa-233e e

Pm-147a X X X X X X X

Po-210a X X X X X X X

Pu-238a X X X X X X X X

Pu-239a X

Pu-240a X X X X X X # #

Pu-241a X

Pu-242a X X X X X X X X

Ru-106a,e e
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 15 of 45

Isotope 1954 1955 1956 1957 1858 1959 1960 1961 1962 1963 1964 1965 1966 1967 1968 1969 1970 1971

Sb-124a L

Sb-125a L

Sc-46a L L

Si (quartz)a L

Taa L

Tc-99a L

Tl-204a L L

Tm-170a X X X X X

Tm-171a X X X X X

U-233a X X X X X X X X

Y-91a L

Zr-95e e

Zr-95-Nb-95e e

Thoriuma X X X X X X X X X X X X X

a. These isotopes are not included in Table 3-2.

b. # Identified in calculations or plans.

c. M Represents an estimated moderate potential of exposure.

d. L Represents an estimated low potential of exposure. 

e. Not identified in the review of the monthly reports but was identified by the SRL bioassay program (DuPont 1984a. 

f. During separation and purification of transcurium nuclides, 1.7 μg 248Cm was obtained from a filter from which 14 μg 252Cf had previously been obtained. 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 16 of 45

# symbol indicates the isotopes were identified either through existing SRS calculations or in plans. 

There are four radionuclides added to the table that were not identified in the monthly reports but were 

identified in History of the Savannah River Laboratory, Volume IV – General (DuPont 1984a) and 

were observed in the bioassay results. These are 106Ru, 233Pa, 95Zr, and Zr/Nb-95 and are included 

for completeness.

4.0 SAVANNAH RIVER LABORATORY OPERATIONS

The Savannah River Laboratory (SRL) (Building 773-A) and related organizations began organizing in 

July 1953. Over the next couple of years the Engineering Services that were provided in support of 

research and development personnel included operation of all building auxiliaries and waste disposal 

facilities, a glass shop, instrument pool, and a supply room. Close liaison was established with Works 

Engineering Department personnel to coordinate all maintenance, machine shop, and project work for 

the SRL. As radioisotopes were brought into 773-A to start operations in radiochemical laboratories, 

the Radiation Control (RC) Group assisted in establishing necessary procedures and handling 

techniques.

Operations formally accepted occupancy of the cave wing, 773-A, Part II, on December 7, 1953, and 

all construction work was completed in February 1954. Experiments with radioactive materials began 

in the caves in July 1954. In October 1954 the first plutonium produced at the SRP was isolated in 

the caves.

A review of the operations in the SRL was conducted because that was where the products of reactor 

operations were evaluated. While the SRL did not evaluate the radionuclides that were shipped to 

offsite customers, they did handle, evaluate, experiment with, and store “production” radionuclides 

such as 244Cm, 252Cf, 60Co, etc. Reactor incidents are not covered in this report because they would 

primarily involve reactor functionality, such as fuel failures, valve failures, etc., rather than exposure to 

the handlers of the transfers of highly radioactive radioisotopes, which was, of necessity, performed 

under water due to the high exposure potential. There were a considerable number of potentially 

serious incidents in the SRL, as discussed below. Several of the incidents have been investigated 

and have an associated Special Hazards Investigation Report (DuPont 1962–1974), while the nature 

of many of the incidents has not been identified. However, there was an extensive radiation control 

program in terms of identifying potential problems and steps taken to evaluate any internal or external 

exposure as a result of these incidents.

Radiation Control was established in Technical Division facilities so all the survey, monitoring, and 

“containment” services usually assigned to Health Physics could be provided in the Technical Division 

organization, together with a wholly integrated approach and direct responsibility (DuPont 1984a). 

The containment service function required close liaison work with technical personnel in the design 

and construction of the special equipment necessary to contain the radioactive materials completely 

during experimental work. This function was performed by the Apparatus Development Group, a 

subgroup of the RC Group (DuPont 1984a). As radioisotopes were brought into Building 773-A to 

start operations in radiochemical laboratories, the RC Group assisted in establishing necessary 

procedures and handling techniques. Major effort was directed to extensive training of laboratory 

personnel and personnel of assigned groups. The fundamentals of radiation and contamination 

control and aspects of personal protection were stressed. By December 1955, procedures basic to 

the program of personnel and area radiation protection were essentially complete (DuPont 1984a). 

Much of the work in the SRL was carried out in the High-Level Caves, which were designed to 

perform chemical and metallurgical operations on highly radioactive materials. As indicated above, 

operations formally accepted occupancy of the cave wing, 773-A, Part II, on December 7, 1953. In 

October 1954 the first plutonium produced at the SRP was isolated in the caves to provide basic data 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 17 of 45

to substantiate and refine calculations of reactor performance. Operations in the caves continued at 

the maximum obtainable rate from that date.

The following information is a synopsis of the internal exposure history of the SRL from 1953 through 

December 1972, which encompasses the HFDP operational period. The information for these years 

is taken from History of the Savannah River Laboratory, Volume IV – General (DuPont 1984a). 

From July 1953 to the early part of 1960 evaluation of internal doses was based on in vitro analysis. 

In 1960 the SRS whole body counter was completed and provided another means of assessing 

internal doses. From 1953 through 1972 there were thousands of urine samples and whole body 

counts but relatively few documented uptakes in the SRL that resulted in significant doses. There 

were no significant biological uptakes from 1953 through 1959 and none identified in 1963 and 1968. 

Table 4-1 shows the number of positive bioassay results that have been documented in DuPont 

(1984a). Not all positive bioassay results were confirmed to be significant uptakes. 

Table 4-1. Positive bioassay results. 

Year

Number of 

positive results Radionuclide

% 

Maximum permissible 

body burden

1960 5 Not identified <1

1961 3 Not identified <1

1962 51 Not identified <1

1964 1 Pa-233 <1.5

1 Ce-144 <1.5

Pu <1.5

1 Not identified Not identified

1965 1 Am-Cm 30a

1 Not identified <5

1966 1 Zr-95 3.2

Ru-106 1.2

4 Pa-231 <1

1967 1 Ru-106 4.2

Ce-144 1.4

Zr-Nb-95 0.36

1969 11 Cm-244 Not identified

1970 31 Cm-244 Not identified

1 Cm-244 SHI No. 273b

1971 21 Not identified Not identified

1 of the 21 Cm-244 SHI No. 283c

1972 4 Not identified Not identified

2 Cf-252 Not identified

a. No Special Hazards Investigation Report was found for this uptake.

b. Lower (1970).

c. Lower (1971).

Two Special Hazards Investigation Reports (SHI) have been identified for this period. Report No. 273

reported that a general service operator received an uptake of 244Cm when he released airborne 

activity to the decontamination room environment while removing solid radioactive waste from the 

Decontamination Chamber in Building 773-A, Decontamination Facility, Room C-003 (Lower 1970). 

Special Hazards Investigation Report No. 283 reported an Analytical Chemistry Division Chemist 

assimilated some curium while removing samples from storage in Building 773-A, Room B-15 (Lower

1971). 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 18 of 45

5.0 BIOASSAY PROGRAMS

Workers at SRS were monitored for internal deposition of radionuclides from the beginning of 

operations. The methods for evaluation of potential intakes included analysis of the amount of 

radioactivity excreted in the urine or feces (in vitro bioassay) and the measurement of the photon 

radiation emitted from the body by chest counting or whole-body counting (in vivo bioassay). The 

bioassay descriptions in this section are brief. More detailed descriptions can be found in the most 

recent version of the Internal Dosimetry Technical Basis Document (ORAU 2005).

In the beginning of operations at SRS, the only method available to evaluate the potential internal 

exposure of personnel was in vitro bioassay, primarily analysis of urine samples. This method was 

typically a gross beta activity analysis although a gross gamma analysis was included from at least 

1959 (Boni undated). Over the years, the bioassay analyses methods changed as new techniques 

were developed.

5.1 IN VITRO BIOASSAY

In vitro bioassay methods changed considerably over time as new techniques and analyses were 

developed and refined. Some of the methods used for analysis of some of the radionuclides are 

briefly discussed below.

Before completion of a whole-body counting facility in 1960, only in vitro bioassay, primarily urinalysis, 

was performed. In vitro urine bioassay for nuclear byproducts performed during this period was 

typically a gross beta activity analysis and/or gross gamma analysis. Since that time, in vitro bioassay 

analysis has become more radionuclide specific, and can be divided into the following categories:

• Plutonium urinalysis

• Uranium urinalysis

• Trivalent actinide urinalysis

• Neptunium urinalysis

• Fission product urinalysis

• Fecal bioassay

Plutonium Urinalysis

The first procedure for plutonium urinalysis at SRS was implemented in 1954 and used until 1959. 

Changes in the processes to prepare the solution for counting were made, but the method for 

counting the alphas tracks on the nuclear track emulsion, Type A (NTA) remained the same until 

1966. In 1964 solid-state surface barrier detectors (gross alpha counting) replaced the NTA 

detectors. In the early 1980s experimentation began with a new technique that used alpha 

spectroscopy, which allowed the detection of the individual radionuclides in the sample (Taylor 2000). 

Uranium Urinalysis

Analysis of uranium by measurement of emitted alpha radiation was referred to as an analysis for 

“enriched uranium.” The analysis of uranium based on its chemical properties was referred to as a 

“depleted uranium” analysis. Enriched uranium (EU) was determined starting in the mid-1950s. 

Analyses for DU were performed with the ORNL fluorophotometric method from 1954 to 1982. This 

method involved using a spectroscope to measure total uranium. The delayed neutron analysis 

(DNA) method was adopted for both EU and DU analyses around 1982 (Taylor 2000). Kinetic 

phosphorimetry analysis was also used for DU during 1986 through 1994. Alpha spectrometry and 

gross alpha counting have been used for both DU and EU since 1994. 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 19 of 45

Trivalent Actinide Urinalysis

The trivalent actinides consist of americium, curium, and californium. There was increased production 

of these materials at SRS in the early 1970s. Analysis was accomplished by extracting the trivalent 

actinides from the plutonium tri-isooctlyamine (TIOA) raffinate with di,2ethylhexylophosphoric acid 

(HDEHP), followed by gross alpha counting. The 1970 method also captured thorium, berkelium, 

einsteinium, and fermium (Butler and Hall 1970). In 1990, ion exchange resin was used to separate 

trivalent actinides from other actinides. The sample was direct-mounted (placed on a planchet with 

rimmed edges and heated to evaporation), and was “gross-alpha” counted by using a wide window on 

the alpha spectrometer. Since 1994, extraction chromatography resin has been used to separate 

trivalent actinides from other actinides. With the implementation of improved software and use of a 

243Am tracer, chemical recovery and counting efficiency could be determined for each sample (Taylor 

2000).

Neptunium Urinalysis

From 1993-1996, extraction chromatography resin was used to separate neptunium from fission 

products and other actinides followed by electrodeposition and gross alpha counting (Taylor 2000). In 

1996, analysis returned to SRS using calcium phosphate precipitation followed by capture on ion 

exchange resin from which plutonium and neptunium were extracted separately followed by alpha 

counting (WSRC 2001 pp. 183-184). Since 2001, TEVA and TRU resins are used to separate 

neptunium from actinides and uranium, and alpha spectrometry has been used (WSRC 2001, pp. 58,

175, 183-184).

Fission Product Urinalysis

Fission products measurements in urine date to at least January 1957, although the method used at 

that time has not been discovered. Since 1959, select fission products in urine were analyzed by beta 

counting on a Geiger-Mueller or proportional counter. The beta count was primarily used to determine 

the presence of 89Sr, 90Sr, and 90Y, which emit no gamma radiation. Because high-energy betas are 

associated with 89Sr, a reliable measurement of its activity was made even though an average of 1 g 

of precipitate was mounted on the planchet. Yttrium-90, which emits a strong beta also yields a 

reliable count, and affords a satisfactory estimate of its parent isobar, 90Sr, since some 90Sr betas are 

not counted due to absorption by the solid. The low-energy beta of most of the induced activities

along with the beta of zirconium-niobium and the rare earths also contributed to the total gross beta 

count. No interference from 40K was observed. The chemistry on the beta sample resulted in 

quantitative recovery of beta-emitting isotopes of iron, cobalt, strontium, niobium, zirconium, 

ruthenium, and cerium. Gross gamma counting was performed using sodium iodide detectors. The 

gamma count was primarily used to determine the presence of induced activity. A gamma count was

used because of the low counting efficiency of the weak beta energies associated with the induced

activities. The gamma from zirconium-niobium and rare earths also contributed to the total gross 

gamma count. Either the beta or the gamma counting statistics are good enough to accurately 

estimate 1% of the maximum permissible amount in the total body of any of the following 

radionuclides (i.e., 90Sr/Y, 89Sr, 95Zr/Nb, 95Nb, 144Ce/Pr, 60Co, 59Fe, 51Cr, 65Zn) with the exception of 

zirconium-niobium. The mixed beta-gamma isotope method provided a simple evaluation of the 

sample regarding the need for an extensive specific analysis, thus eliminating unnecessary and time 

consuming work (Boni undated).

SRS logbooks show that radionuclide-specific concentrations and associated uncertainties were 

quantified for some radionuclides for some of the gamma-counted samples starting in December 1965 

(DuPont 1965–1968). The reported radionuclides were:

• Chromium-51

• Manganese-54

• Cobalt-60

• Zinc-65
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 20 of 45

• Zirconium/Niobium-95

• Ruthenium-106

• Iodine-131

• Cesium-137

• Barium/Lanthanum-140

• Cerium-144.

Strontium Urine Bioassay

In the late 1950s and early 1960s, 89/90Sr was determined by beta counting on automatic and GeigerMueller counters along with zirconium, rare earths, and niobium. From 1969 to 1997, strontium was 

analyzed by a liquid ion exchange method that extracted 90Y, which was counted on a beta 

proportional counter (Taylor et al 1995, p. 79). In 1998, the strontium method was modified by using 

an ion exchange column to extract 90Sr, which was also counted on a proportional counter (Taylor 

2000, p. 5). Since 2001, 90Sr has been separated as part of the use of TRU and TEVA resins for 

plutonium, uranium, actinides, and strontium (WSRC 2001, pp. 58, 168, 169, 172, 173).

Table 5-1 lists the isotopes that were listed as being present, their radiations, bioassay method for 

detection, half-life, and the number of times they were identified in the monthly reports from 1954 to 

1971. An X indicates they were listed numerous times.

Table 5-1. Selected isotopes bioassay methods. 

Radionuclide

Times 

identified Half-lifea

Principal 

radiation

Bioassay method used or 

available for use

Bioassay 

period of 

availability

Am-243 X 7.4E3 yr Alpha, gamma Trivalent actinide urinalysis From mid 

1960s

Whole-body counting of Np-239 From 1960

Chest counting From 1966b

Bk-249 X 320 d Beta None discoveredc

Co-60 X 5.26 yr Beta, gamma Whole-body counting From 1960

Gamma urinalysis IA From 1960

Cs-137 X 30.2 yr Beta, gamma Whole-body counting From 1960

Es-253

X 20.5 d

Alpha Trivalent actinide urinalysisd From mid 

1960s

Es-254 X 276 d Alpha Trivalent actinide urinalysisd From mid 

1960s

Whole-body counting of Bk-250 

progeny

From 1960

Fm-254 X 3.24 h Alpha None discoverede

Pm-147 X 2.63 yr Beta None discoveredf

Pu-242 X 2.9E5 yr Alpha Pu urinalysisg From 1954

Tm-170h X 129 d Beta, gamma Chest countingi From 1966

Tm-171h X 1.92 yr Beta None discoveredj

U-233 X 1.6E5 yr Alpha EU urinalysis From 1959

Bk-250 2 3.22 hr Beta, gamma Whole-body countingk From 1960

Ce-144

2 284 d

Beta, gamma Fission product urinalysis FP From 1950s

Whole-body counting From 1960

Eu-152

2 12.7 y

Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Eu-154

2 8.8 yr

Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Tl-204 2 3.8 yr Beta Fission product urinalysis FP From 1950s

Am242ml

1 152 yr

Beta (Am-242) Trivalent actinide urinalysis From mid 

1960s
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 21 of 45

Radionuclide

Times 

identified Half-lifea

Principal 

radiation

Bioassay method used or 

available for use

Bioassay 

period of 

availability

Ba-133h 1 10.5 yr Gamma Gamma urinalysis From 1960

Cm-248 1 3.4E5 yr Alpha Trivalent actinide urinalysis From mid 

1960s

Es-255 1 38 d Alpha, beta 

(including 

progeny)

Trivalent actinide urinalysisd From mid 

1960s

Np-239 1 2.35 d Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Pa-231 1 3.2E4 yr Alpha, gamma 

(7%)

Gamma urinalysis IA From 1960

Whole-body counting From 1960

Sb-124 1 60.2 yr Beta, gamma Gamma urinalysis IA From 1960

Whole body counting From 1960

Sb-125 1 2.73 yr Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Sc-46 1 83.8 d Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Tc-99 1 2.1E5 yr Beta Nonem

Y-91 1 58.8 d Beta Fission product urinalysis FP From 1950s

Ir-192 Shipment 

only

74.3 d Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

Cd-113mh 0n 13.6 yr Beta Fission product urinalysis FP From 1950s

Cs-134h 0 2.06 yr Beta, gamma Fission product urinalysis FP From 1950s

Whole-body counting From 1960

Eu-155h 0 5.0 yr Beta, gamma Chest counting From 1966

(also whole-body counting of 

associated Eu-154)

From 1960

Ho-166mh 0 1.2E3yr Beta, gamma Gamma urinalysis IA From 1960

Whole-body counting From 1960

I-131 0 8.06 d Beta, gamma Whole-body counting From 1960

Thyroid counting

K-40h 0 1.3E9 yr Beta, gamma Whole body countingo From 1960

Np-238 1 2.12 d Beta, gamma Whole-body counting if soon 

after intake

From 1960

a. Half-lives are taken from Lederer, Hollander, and Perlman (1967) and the Radiological Health Handbook (PHS 1970). 

b. The 40-cm arc whole-body detector had a lung calibration for the 98-keV gamma from Nd-147, which also would have 

detected the 99- 111-keV X-rays from Np-239.

c. See Attachment C.

d. The trivalent actinide method from 1970 forward captured Am, Cm ,Cf, Th, Bk, Es, Fm (Butler and Hall 1970). Alphaemitting isotopes would have been counted.

e. The half-life of Fm-254 is so short that bioassay would not have been feasible. However, the cooling of the target 

before handling in the laboratory resulted in little or no leftover 254Fm. See Attachment C.

f. See Attachment B.

g. Pu-242 was used as an internal tracer starting in about 1981, but would have been detected in Pu urinalysis prior to 

that.

h. Isotopes that were discussed by Rupp (1969) in response to the request to use the SRS irradiation facilities for 

commercial and industrial development.

i. Because Tm is a bone seeker, the 84-kev gamma from the ribs would have been detectable in chest counts, despite the 

low yield (3.3%).

j. See Attachment D.

k. The half-life for Bk-250 is so short that the whole body count would have had to occur shortly after intake; however, the 

DAC (1E-7 μCi/cm3

) is 50 times larger than for Sr-90 so a significant intake would have been readily detectable in the 

workplace.

l. See Attachment E.

m. See Attachment A.

n. Not identified in the referenced monthly reports but were available along with bioassay methods for their detection.

o. An occupational intake would have been detectable only if the count was significantly larger than natural background.
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 22 of 45

Fecal Bioassay

Chemical bioassay of urine and fecal samples has been carried out at SRS since the 1950s to 

determine an individual’s uptake of uranium and actinide elements such as americium, plutonium, and 

californium. Over the years, several techniques have been employed for these types of analyses. 

Methods ranging from autoradiography to high-resolution solid-state alpha spectroscopy have been 

used at one time or another.

Many different methods of sample preconcentration, treatment, and preparation have been evaluated 

and used over the years. Each technique had some particular advantage at the time it was being 

used, either in terms of analysis time, the number of samples that could be processed simultaneously, 

the physical form of the final material to be counted, compatibility with other analyses that could be 

carried out sequentially or in parallel with the main analysis, compliance with more stringent detection 

levels, etc. (WSRC 1990). In contrast to a urine bioassay, a fecal bioassay can (and usually does) 

contain material that was never in the systemic body (i.e., insoluble material cleared from the lungs 

and gastrointestinal tract). Fecal bioassay is thus complementary to urine bioassay. In fact, at SRS, 

fecal bioassay is always used in conjunction with a urine bioassay and never alone (WSRC 2001). 

The primary method for fecal bioassay in the early and mid-1980s was gamma spectrometry with a 

phoswich detector. Some of the samples that were analyzed by this method were subsequently dried 

and sent to offsite laboratories for analysis. Since 1994, fecal samples have been analyzed by a 

modified urinalysis procedure (Taylor et al. 1995; Taylor 2000).

5.2 IN VIVO BIOASSAY

History of Whole-Body Counting at SRS (Taylor 2000) 

In vitro bioassay was the only method available for assessing occupational intakes of radionuclides in 

defense facilities like SRS before the mid-1950s. Before the development of large NaI detectors, in

vivo bioassay was performed with ionization chambers or Geiger-Mueller tubes that were rather 

insensitive and not practical for assessing occupational intakes of radionuclides, with the possible 

exception of 131I. In 1957, a whole body counting facility similar to that at ANL using NaI detectors 

was proposed for SRS. Before the completion of the facility in 1960, only in vitro bioassay, primarily 

urinalysis, was performed at SRS. In vitro urine bioassay for fission and activation products during

this period was typically a gross beta or gross gamma analysis.

The original SRS Whole Body Counter in Building 735-A was completed in 1960 and ceased 

operations in September 1995. The original geometry was a single 8- by 4-in. NaI detector with the 

person sitting in a chair. This was referred to as the 40-cm arc geometry. The energy spectrum was 

divided into 15 regions ranging from 60 -100 keV (147Nd) to 1760 -1990 keV (unnamed) (Taylor 2000, 

pp. 5-6, Taylor et al 1995 pp.60-63, Watts et al 1964, Watts 1961-67). Sometime during the early 

1970s, the original chair geometry was replaced with a recumbent (bed) geometry. The single NaI 

detector was replaced with five 4- by 4-in. cylindrical NaI detectors under the bed. The NaI detectors 

were in an arc under the bed to provide constant efficiency for all detectors (Taylor 2000). Starting in 

about 1979, the gamma spectrum was reorganized into 10 regions of interest from 104 keV to 1,534 

keV (as shown on the In Vivo Count Results form). In the mid-1980s, whole body counting using a 

stand-up geometry and large (4” × 4” × 16” or 5” × 3” × 16”) NaI detectors and much shorter counting 

times (2-4 min.) was implemented (Taylor 200, p. 7; Taylor et al 1995, pp. 74-76). 

History of Chest Counting at SRS (Taylor 2000)

The whole-body counter, which was useful for monitoring workers for intakes of gamma-emitting 

fission and activation products, was of no use for detecting or quantifying intakes of many actinides 

such as plutonium and americium. The 40-cm arc whole body counter was calibrated for 147Nd in the 

lung as well as in the whole body so low-energy gamma emitters (around 100 keV) could be 

quantified, although the sensitivity was poor. Efforts began at SRS in the mid-1960s to develop low-
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 23 of 45

energy photon detectors that would permit detection of actinides in the chest. Most efforts involved 

either thin NaI detectors or xenon-filled proportional counters. The xenon proportional counters, with 

their high resolution, showed great promise but were eventually abandoned because of their high

background and low efficiency. Development of the thin NaI detector continued, and in 1966 a 1-mmthick by 5-in.-diameter detector with a 0.001-in. window was first used to count workers involved in an 

incident. Thin NaI detectors were used with varying degrees of success from 1966 until 1971, when 

they were replaced with phoswich detectors. The results of the thin NaI counts were verified to be 

officially recorded in the personnel files of the workers counted during this period.

In the late 1960s, the phoswich detectors were developed for low-energy photon and X-ray 

measurement. Phoswich detectors consist of a “sandwich” of phosphorescent materials (NaI and 

CsI), hence, the name “phoswich.” This detector took advantage of its thin construction and inherent 

anti-coincidence counting capabilities to greatly reduce the background in the low-energy region of the 

spectrum. In 1972, a dual phoswich system composed of two 5-in.-diameter detectors was placed in 

service. Although improvements were made in the analysis of data and calibration over the years, 

this basic system was used for chest counting for nearly 15 years until the adoption of planar 

germanium detectors in the late 1980s.

In the late 1970s, small coaxial germanium detectors were installed in the steel room. These 

detectors were pointed at the sides of the chest in an effort to detect the 186-keV photon from the 235U 

in EU, which was not readily detected by the phoswich or NaI detector. These detectors were also 

used to identify high-energy photon emitters, taking advantage of the superior resolution of the 

germanium detectors. In the mid-1980s, thick phoswich detectors specifically designed for EU were 

purchased. This turnkey system was used for only a few years before the germanium chest counter 

replaced it. In the early 1980s, solid-state planar germanium detectors were introduced at several 

facilities in the complex for chest counting. Routine use of a six-detector germanium chest counter 

began at SRS in about 1987 or 1988. This was a turnkey system purchased from a commercial 

vendor.

The utility of routine chest counting for actinides declined throughout the 1980s and early 1990s. This 

decline can be related to fundamental changes in the occupational internal dose limits. During the 

1970s, the occupational limit for plutonium was expressed in terms of a quantity of plutonium that 

could be present in the chest at any given time – the maximum permissible lung burden (MPLB). The 

MPLB for plutonium was 16 nCi, which at the time was considered to be detectable with the dual

phoswich system. Thus, compliance with the occupational dose limits could be demonstrated with a 

simple chest count. In 1980, the limits changed to annual effective dose equivalent, which caused an 

increase in the complexity of calculations but did not cause insurmountable problems because it is 

closely related to the MPLB. The routine chest count was still a valuable tool for demonstrating 

compliance. In 1992, the limits were changed to committed effective dose equivalent. The effect of

this was to reduce drastically the amount of plutonium in the chest that corresponded to the 

occupational limit. The reduction was so great that chest counters could no longer detect plutonium at 

compliance levels, which rendered routine chest counts useless for determining compliance. 

However, chest counters still play an important role for special chest counts and for routine chest 

counts if a tracer such as 241Am might be present.

Germanium detectors have become the detectors of choice for measurement of actinide radioisotopes 

in the lungs. They are less efficient but much more specific (i.e., better resolution) than NaI or 

phoswich detectors. The SRS germanium detector system became operational in December 1989, 

replacing the phoswich detector system.
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6.0 SUMMARY

Based on the various isotopes that were irradiated in the SRS reactors, it is clear that more than 100 

nuclides have been identified as having been irradiated for the SRL and many offsite customers. As 

indicated in the tables, aside from the “production” radioisotopes, most materials that were irradiated 

were relatively small quantities, which were, according to Gray (2006), sealed in aluminum cans on 

arrival, handled under water, and shipped as soon as possible on removal from the reactor. During 

the early years, urine bioassay (either gross alpha or gross beta analyses) was the only method of 

evaluating intakes of radioactive materials. SRS built a whole-body counter in 1960 for detecting 

fission and activation products. Because the whole-body counter had limited use for the detection of 

actinides, efforts began in the mid-1960s to develop low-energy photon detectors that would permit 

detection of actinides in the chest. Improvements in whole-body counters and chest counters 

provided the means of detecting lower amounts of activities in the body. The RC Group provided 

training and retraining of all personnel and maintained records of internal and external exposures from 

the beginning of operations including exposures related to incidents. 

SRL did handle high activities of actinides, such as isotopes of americium, curium, and californium, for 

which bioassay was established. A program involving 210Po was conducted by a small group of 

personnel for whom bioassay was conducted (Mahathy 2011). Whole-body counting was used to 

monitor intakes of most of the exotic irradiated targets. The SRL did early development work 

involving chemistry on 171Tm and broken wafers (170Tm), which was one-time, small-scale work. 

Other than that, no evidence was found that the SRL worked with exotic radionuclide sources in a 

manner that breached the integrity of the source. Although this was small-scale work, a method for 

bioassay for the 170Tm was available as indicated in Table 5-1. No evidence was found for bioassay 

for 171Tm, 249Bk, 254Fm, 147Pm, and 99Tc. These five radionuclides are discussed in Attachments A 

through E. There were exotic radionuclides created at SRS for which no bioassay was performed 

(mostly beta-only emitters), there was no evidence found that indicated that bioassay was needed. 

This report addresses 249Bk during the HFDP (1965 to 1970) (see Attachment C). There was 

significant production of 249Bk in the reactors in the 1970s, but nothing was found in the records that 

indicated the targets were processed in a way that purified the 249Bk or that pure 249Bk was handled in 

quantities that could have produced significant potential for intakes.
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 25 of 45

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DuPont IPHS Apr–Jun 1968 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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Carolina. [SRDB Ref ID: 104546]

DuPont SRL Nov 1964 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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DuPont SRL Jan 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

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DuPont SRL Feb 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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DuPont SRL Mar 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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DuPont SRL Apr 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

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[SRDB Ref ID: 106719]

DuPont SRL Jun 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

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DuPont SRL Aug 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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[SRDB Ref ID: 106746]

DuPont SRL Sep 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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Carolina. [SRDB Ref ID: 106810]

DuPont SRL Nov 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, November 1965, DP-65-1-11 (extract), Savannah River Laboratory, Aiken, 

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DuPont SRL Dec 1965 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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DuPont SRL Feb 1966 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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DuPont SRL Mar 1966 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

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DuPont SRL Apr 1966 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, April, DP-66-1-4 (extract), Savannah River Laboratory, Aiken, South Carolina. [SRDB 

Ref ID: 106831]

DuPont SRL Jun 1966 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, June 1966, DP-66-1-6 (extract), Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106838]

DuPont SRL Sep 1966 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, September 1966, DP-66-1-9 (extract), Savannah River Laboratory, Aiken, 

South Carolina. [SRDB Ref ID: 106750] 

DuPont SRL Nov 1966 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, November 1966, DP-66-1-11 (extract), Savannah River Laboratory, Aiken, 

South Carolina. [SRDB Ref ID: 106755] 

DuPont SRL Dec 1966 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, December 1966, DP-66-1-12 (extract), Savannah River Laboratory, Aiken, 

South Carolina. [SRDB Ref ID: 106758]

DuPont SRL Feb 1967 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, February 1967, DP-67-1-2 (extract), Savannah River Laboratory, Aiken, 

South Carolina. [SRDB Ref ID: 106765]

DuPont SRL May 1967 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, May 1967, DP-67-1-5 (extract), Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106799]

DuPont SRL Jun 1967 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, June 1967, DP-67-1-6 (extract), Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106800]

DuPont SRL Jul 1967 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, July 1967, DP-67-1-7 (extract), Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106803]

DuPont SRL Aug 1967 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, August 1967, DP-67-1-8, Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106664]
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 32 of 45

DuPont SRL Sep 1967 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, September 1967, DP-67-1-9, Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106665] 

DuPont SRL Dec 1967 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, December 1967, DP-67-1-12 (extract), Savannah River Laboratory, Aiken, 

South Carolina. [SRDB Ref ID: 106671] 

DuPont SRL Mar 1968 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, March 1968, DP-68-1-3 (extract), Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106874]

DuPont SRL Apr 1968 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, April 1968, DP-68-1-4 (extract), Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106681]

DuPont SRL May 1968 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, May 1968, DP-68-1-5 (extract), Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106682]

DuPont SRL Mar 1969 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, March 1969, DP-69-1-3, Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106872]

DuPont SRL Apr 1969 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, April 1969, DP-69-1-4 (extract), Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106807]

DuPont SRL Oct 1969 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, October 1969, DP-69-1-10 (extract), Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106852]

DuPont SRL Dec 1969 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, December 1969, DP-69-1-12, Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106873]

DuPont SRL Apr 1970 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, April 1970, DP-70-1-4, Savannah River Laboratory, Aiken, South Carolina. [SRDB 

Ref ID: 106866]

DuPont SRL Jul 1970 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, July 1970, DP-70-1-7, Savannah River Laboratory, Aiken, South Carolina. [SRDB Ref 

ID: 106863]

DuPont SRL Aug 1970 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, August 1970, DP-70-1-8 (extract), Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106862]

DuPont SRL Sep 1970 (E. I. du Pont de Nemours and Company), Savannah River Laboratory 

Monthly Report, September 1970, DP-70-1-9 (extract), Savannah River Laboratory, Aiken, 

South Carolina. [SRDB Ref ID: 106861]
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 33 of 45

DuPont SRL Oct 1970 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, October 1970, DP-70-1-10 (extract), Savannah River Laboratory, Aiken, South 

Carolina. [SRDB Ref ID: 106859]

DuPont SRL Apr 1971 (E. I. du Pont de Nemours and Company), Savannah River Laboratory Monthly 

Report, April 1971, DP-71-1-4 (extract), Savannah River Laboratory, Aiken, South Carolina. 

[SRDB Ref ID: 106698]

DuPont WTD May 1965 (E. I. du Pont de Nemours and Company), Progress Report, May 1965, 

Works Technical Department, DPSP-65-1-5 (extract), Savannah River Plant, Aiken, South 

Carolina. [SRDB Ref ID: 68105]

DuPont WTD Aug 1965 (E. I. du Pont de Nemours and Company), Progress Report, August 1965, 

Works Technical Department, DPSP-65-01-08 DV, Savannah River Plant, Aiken, South 

Carolina, September 17. [SRDB Ref ID: 11875]

DuPont WTD Jan 1966 (E. I. du Pont de Nemours and Company), Progress Report, January 1966, 

Works Technical Department, DPSP-65-1-1 DV, Savannah River Plant, Aiken, South Carolina, 

February 18. [SRDB Ref ID: 11879]

DuPont WTD Mar 1966 (E. I. du Pont de Nemours and Company), Progress Report, March 1966, 

Works Technical Department, DPSP-66-1-3 Extract, Savannah River Plant, Aiken, South 

Carolina, April 20. [SRDB Ref ID: 68109]

DuPont WTD Mar 1967 (E.I. du Pont de Nemours and Company), Progress Report, March 1967, 

Works Technical Department, DPSP 67-1-3 DV, Savannah River Plant, Aiken, South Carolina, 

April 20. [SRDB Ref ID: 12455]

DuPont WTD Feb 1971 (E. I. du Pont de Nemours and Company), Works Technical Department, 

Report for February 1971, DP-71-1-2 (extract), Aiken, South Carolina. [SRDB Ref ID: 68020]
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 34 of 45

In the latter half of 1964 and into 1965 work was performed in the SRL on extraction of 99Tc from highlevel waste (DuPont 1984b, pp. 94, 118; DuPont SRL Sep 1964,

1 p. 44). The feed material would 

have had plenty of fission products, and monitoring by fission product urinalyses and whole body 

counting would have been appropriate. In June 1964, 1.1 g of 99Tc was isolated and work toward 

extracting another 1.8 g was conducted in September (DuPont SRL Sep 1964, p. 44). The purity of 

the product was not mentioned. No mention of technetium extraction was made for subsequent years 

in the historical summary document (DuPont 1984b), and no mention was made in the series of 

quarterly reports of the Isotopic Power and Heat Sources (IPHS) Project starting in July 1966 (for 

instance, DuPont IPHS Jul–Sep 1966, DuPont IPHS Jan–Mar 1967, and others). [Note: the October 

to December 1966 report has not been captured.] In addition, no mention of work with 99Tc was made 

in any of the SRL RC Group monthly reports in 1966 (whereas there is frequent mention of work 

specifically with curium, thorium, 238Pu, 237Np and occasional mention of work with thulium and 

polonium in 1966 or other years). 

The work with 99Tc extraction appears to be limited from about midyear 1964 to about midyear 1965, 

when it appears work on this radionuclide was abandoned.

Because the specific activity of 99Tc is small (0.017 Ci/g), the potential for consequential intakes was

smaller than for the other heat source radionuclides and was very small in terms of potential for 

inducing cancer. Table A-1 shows estimates of potential chronic intakes and committed organ doses 

from handling 1.8 g of 99Tc using the common Brodsky’s “magic” number (10-6) to relate activity in 

process to potential for intakes (Brodsky 1980).

 

1 For citations of this form, see the Monthly and Quarterly Reports section at the end of the reference list.

ATTACHMENT A

TECHNETIUM-99

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Table A-1. Potential inhalation committed organ 

doses (mrem) from handling 1.8 g 99Tc.

a

Target organs

Organ committed doses

Chronic intake (365 d), 

83.7 pCi/d

Type F 

inhalation

Type M 

inhalation

Adrenals 3.86E-03 3.47E-03

Urinary bladder 1.60E-02 1.44E-02

Brain 3.86E-03 3.47E-03

Breast 3.86E-03 3.47E-03

Gall bladder 3.86E-03 3.47E-03

Heart wall 3.86E-03 3.47E-03

Kidneys 3.86E-03 3.47E-03

Liver 5.14E-03 4.63E-03

Muscle 3.86E-03 3.47E-03

Ovaries 3.86E-03 3.47E-03

Pancreas 3.86E-03 3.47E-03

Testes 3.86E-03 3.47E-03

Thyroid 1.01E-01 9.07E-02

Red bone marrow 3.86E-03 3.47E-03

Bone surface 3.86E-03 3.47E-03

Stomach 2.07E-01 1.89E-01

Small intestine 5.18E-03 6.24E-03

Upper large intestine 5.59E-02 5.99E-02

Lower large intestine 1.56E-01 1.69E-01

Skin 3.86E-03 3.47E-03

Spleen 3.86E-03 3.47E-03

Thymus 3.86E-03 3.47E-03

Uterus 3.86E-03 3.47E-03

Extrathoracic airways 1.92E-01 9.33E-01

Lung 6.10E-03 2.69E+00

Colon 9.94E-02 1.07E-01

Esophagus 3.86E-03 3.47E-03

Gonads 3.86E-03 3.47E-03

a. Maximum doses highlighted.

ATTACHMENT A

TECHNETIUM-99

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Promethium-147 was among the unusual nuclides that were produced in research quantities and 

evaluated for a variety of uses from 1965 to 1970. Promethium-147 has a half-life of 2.6 years and is 

a beta emitter with an average energy of 62 keV. The low energy allows for ease of handling and 

does not require significant shielding, which, in turn, reduces the weight required of other 

radioisotopes for similar applications.

As early as 1958 there was an interest in using the fission product 147Pm as a fuel for Systems for 

Nuclear Auxiliary Power (SNAP) batteries. As a result, a process for the recovery of 147Pm was 

started. At that time a method was also developed for the radiochemical determination of 147Pm, 

along with 144Ce and 91Y, which were the major contaminants of promethium (DuPont 1984b, p. 23). 

As can be seen in Table 3-4 147Pm was identified several times in 1959 and 1960. However, it was 

not until 1965 that studies began on the use of various radionuclides as heat source materials, 

including 147Pm. 

Because the National Aeronautics and Space Administration (NASA) had an interest in aged 147Pm as 

a source of isotopic heat for space missions, a study of the quantity and quality of 147Pm that could be 

produced annually at SRP was undertaken. The major source of 147Pm for its recovery would be 

fission product waste. The quantity available after aging to allow for decay of associated 148Pm would 

be about 1,700 W(t) per reactor-year of operation. Product quality from current waste in 1965, as 

represented by a low ratio of 148Pm/147Pm, would presumably be the same as that produced at 

Hanford (about 5 × 10-7). Promethium-148 emits gamma rays that could require extra shielding, while 

147Pm is a beta emitter as indicated above. In addition, a scoping study was made to determine the 

quality of 147Pm in fission product sludge from SRP operation. The measured isotopic ratio of 148Pm to 

147Pm was found to be similar to the isotope ratio in Hanford fission products (0.4 ppm) (DuPont 1974, 

p. 7). 

A process to recover 147Pm from plant high-activity waste was also developed in the laboratory in 

1965 for use as a power source. During the extraction process, the 147Pm had to be separated from 

144Ce since that is a major source of heat and its removal would allow easier storage of the 147Pm. 

After separation the 147Pm was then stored until most of the 148Pm impurity decayed and was then 

purified by ion exchange (DuPont 1984b, p.119). Starting in February 1965, an estimate was made of 

the annual production at SRP (in the fission product fraction) of 147Pm, and the recovery of 147Pm from 

current high-activity waste was studied (DuPont SRL Feb 1965, pp. 37–40; DuPont SRL Aug 1965, 

pp.43, 45). The February study was conceptual; work at the tracer level started in August. The early 

work had considerable 144Ce, so intake would have been monitorable by whole body counting 

(DuPont SRL Dec 1965, pp. 29–32). The isotopic purity of aged 147Pm (from the SRP waste tanks) 

was also measured (DuPont 1974, pp.5–6; DuPont SRL Mar 1965, p. 24). Laboratory-scale testing 

and improvements to the chemical separation procedure continued in 1966, still with issues with 144Ce 

in the 147Pm fraction (DuPont SRL Feb 1966, p. 20; DuPont SRL Sep 1966, pp. 23–24). A plant test 

to demonstrate the initial recovery steps was begun in November 1966 and demonstrated the 

recovery of a 147Pm-La mixture from plant waste (DuPont 1984b, p. 142; DuPont SRL Nov 1966, p. 

33). Work continued in the laboratory on a method to remove the cerium (DuPont SRL Nov 1966, p. 

34). The first test with essentially pure 147Pm occurred in June 1967 involving about 500 Ci (DuPont 

SRL Jun 1967, pp. 26–27). By late 1967 another test of the process to recover 147Pm from highactivity waste was successful in producing about 500 g from a lanthanide mixture. The first steps of 

the process were conducted in F Area to produce the Pm-La mixture, which underwent multiple 

extractions (back-extractions) to remove 80% to 90% of the cerium. The resulting material was to be

stored 2 years for decay of 148Pm. Laboratory tests then demonstrated that pressurized ion exchange 

columns could be used for final purification of 147Pm. The initial separation steps also captured 238Pu 

ATTACHMENT B

PROMETHIUM-147

Page 1 of 2 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 37 of 45

in a separate fraction (DuPont SRL Sep 1967, pp. 32–33; DuPont SRL Dec 1967, p. 37; DuPont

1984b, p. 163). Although the 148Pm contaminant in the Pm-La mixture is a strong gamma emitter and 

was of concern for the desired use of the 147Pm, the activity ratio was about 0.002 148Pm/147Pm, so 

whole body counting would not have been a sensitive bioassay for this mixture; however, even with 

80% to 90% removal, there was still considerable 144Ce (a little more than 1,700 Ci/g 147Pm or about 

2/1 activity ratio 144Ce/147Pm) (DuPont SRL Apr 1968, p. 17). Whole body counting would have been 

an adequate bioassay for the Pm-La mixture in the F Area.

For another application, suitable isotopic purities were calculated for 238Pu, 171Tm, and 147Pm as power 

sources for pumps to assist or replace the heart. A heart pump would be about 5 in. in diameter and 

installed in the abdomen. It would include a steam boiler and heat exchanger with blood as coolant

(DuPont IPHS Oct–Dec 1967, p. 45). Plutonium-238 appeared to be the most promising power 

source for such pumps because it would be available in adequate supply and could be aged to 

remove 236Pu to sustain long cycles of operation with reasonable weight of shielding. Thulium-171 

would require potentially the least shielding, but was expected to be more expensive than 238Pu in the 

necessary, highly purified form and was also handicapped by much shorter half-life (1.9 vs. 87 years). 

Promethium-147, readily available as a fission product, had similar disadvantages of being difficult

and more expensive to produce the necessary isotopic purification required for minimum shielding, 

and it also has a shorter half-life (2.6 years). In 1968, 147Pm was also considered as a heat source for 

thermoelectric and thermionic power generation. Studies of compounds of promethium were 

conducted to provide information on the optimum fuel form selection (DuPont 1984b, p.182). A 

January to March 1968 plant test of the process in H Area recovered about 2 kg of 147Pm as a 

lanthanide mixture. In laboratory tests in April 1968 pressurized ion exchange columns were used to 

isolate approximately 1 g of 147Pm from a portion of the 500-g lanthanide mixture that had been 

created the previous December (DuPont 1984b, p. 182; DuPont SRL Apr 1968, pp. 15–17). The feed 

for these tests had 1,700 Ci of 144Ce per g 147Pm, but the final product had less than 1 mCi 144Ce per g 

147Pm and similarly small amounts of 106Ru, 95Zr/Nb, and 154Eu (DuPont SRL Apr 1968, p. 17). No 

further work with pure 147Pm was found in the SRL monthly reports for the remainder of 1968 or for 

1969 to 1971.

While there was considerable development work with 147Pm, there is no record that indicates it was 

ultimately applied to the above-mentioned applications. The Pm-La separations of 500 g and 2 kg in 

different operations in the F and H Areas represented multicuries of 147Pm activity, but there was 

sufficient 144Ce contamination for adequate bioassay monitoring by whole body counting. The smaller 

amounts of pure 147Pm that were used for methods testing in June 1967 (about 500 Ci) and April 1968 

(about 1,000 Ci) were too pure to use whole body counting for bioassay. These tests were conducted 

in the SRL. 

In 1969 work on the development of 147Pm as a heat source was concluded and no further work was 

planned (DuPont 1984b, pp. 193, 200–201). In 1971 about 460 g 147Pm and 100 g 144Ce, stored for 

several years, were discarded to make room for 244Cm (DuPont 1971, p. 8). 

ATTACHMENT B

PROMETHIUM-147

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Berkelium-249 and 254Fm were two of many radionuclides created as part of the Transplutonium 

Project, evolving out of the work to create 252Cf and to produce 3 kg of 244Cm by 1967 (DuPont SRL 

Sep 1964, p. 25). Berkelium-249 resulted from decay of 249Cm, which was produced from multiple 

(n,γ) reactions starting from 244Cm. Berkelium-249 emits a weak beta particle (average energy 

33 keV) and no photons. Its progeny is 249Cf, which is an alpha emitter that would have been counted 

in the trivalent actinide bioassay procedure, but the long half-life (351 years) precludes ingrowth of 

sufficient activity to be useful for bioassay monitoring.

Fermium-254 resulted from multiple (n, γ) reactions starting from 252Cf with the final two steps being 

an (n,γ) reaction of the 253Es and beta decay of 254Es (DuPont 1971, p. 38). Fermium-254 (half-life 

3.2 hours) was an intermediate step in the pathway to produce 257Fm (half-life 84 days). It is an alpha

emitter, but the short half-life precludes direct bioassay monitoring. Its progeny 250Cf is also an alpha

emitter, but the difference in half-lives (3.2 hours versus 13.1 years) means the 250Cf activity would be 

insignificant. The transplutonium historical document states that “only 257Fm can be recovered in 

significant amounts because, even though it is at the end of the reaction chain, its 84-day half-life is 

much longer than those of the other fermium isotopes” (DuPont 1971, p. 39). 

Berkelium-249

The SRL monthly report for September 1964, in a section on the use of the transuranic metal 

extraction (TRAMEX) process for isolation of californium, stated, “Continuing analytical development 

study will include determination of Bk and Es, in addition to Cf, in the dissolver solution from the 

second Transplutonium campaign” (DuPont SRL Sep 1964, p. 34). There was no mention of 

berkelium in the October and November monthly reports for 1964 (DuPont SRL Oct 1964; DuPont 

SRL Nov 1964). “A series of laboratory tests was made in support of the second Transplutonium 

separations campaign, which was started in the H Area in October.” These tests focused on the 

separations chemistry of plutonium, americium, and curium (DuPont SRL Nov 1964, p.38). In 

January 1965 a 0.03-μg sample of 252Cf was extracted from curium and extraction of a very small 

amount of berkelium and einsteinium from the californium sample was begun (DuPont SRL Jan 1965, 

pp. 7, 14-18). This work continued through March; the activity of the extracted 249Bk was mentioned 

as about 2,600 cpm. The counters were not calibrated for 249Bk so the exact activity was not given, 

but it was probably 1 to 10 nCi (DuPont SRL Mar 1965, pp. 14–16). No mention is made in 

subsequent SRL monthly reports until March 1966, although the January 1966 monthly report states 

that a transplutonium campaign was started in the reactor in February 1965 and was due to be 

removed in February 1966 (DuPont SRL Jan 1966, p. 5). The historical summary document of the 

Separation Technology Section indicated that in 1965 analyses were developed for 249Bk and 253Es as 

part of the Transplutonium Project (DuPont 1984b, pp. 111–112). The March 1966 SRL monthly 

report shows that some 249Bk from LRL had been combined with a small amount produced at SRS 

(presumably from the 0.03-μg 252Cf sample) to produce a thin target for cross-section studies. The 

target was placed in the reactor in February 1966. It contained 485 ng (790 μCi) (DuPont SRL Mar

1966, pp. 14–16). In February 1967, mass spectrometry was performed on a sample from an 

irradiated 238Pu target (P-5) to determine the yields of isotopes of californium, einsteinium, and 

berkelium. The entire P-5 element had 88 μg of 249Bk but, of course, the analyzed sample had much 

less. The sample also had considerable alpha activity from the isotopes of californium and 

einsteinium (DuPont SRL Feb 1967, pp. 21–22). In May 1967 a similar analysis was performed on 

two samples from the P-6 element showing similar isotopic abundances of californium, einsteinium, 

and berkelium (DuPont SRL May 1967, p. 18). Another cross-section study was performed in August 

1967 using two 249Bk targets containing 175 μg (285 mCi) and 105 μg (171 mCi) of 249Bk, respectively. 

Chemistry was involved in the target preparation (DuPont SRL Aug 1967, pp. 39–42).

ATTACHMENT C

BERKELIUM-249 AND FERMIUM-254

Page 1 of 5 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 39 of 45

The chapter in the 50-year symposium document that discusses creation of curium and californium 

states, “In 1968-69, a high-pressure elution development cation exchange chromatographic process 

was developed. ... Milligram quantities of 252Cf, submilligram quantities of 249Bk, and subnanogram 

amounts of 253Es and 254Fm were separated from ... 244Cm and 243Am ...” (Harbour et al. 2000, 

p. 172). There is no mention of work with berkelium, einsteinium, or fermium in the SRL monthly 

reports from September 1967 through February 1968. In March 1968 the yield of berkelium in the 

curium campaign of 1967 was determined by analyzing a small sample from the stored solution. “The 

analytical method was also used to trace berkelium in the Tramex process.” These were tests in 

preparation for extraction of larger amounts of berkelium from the TRAMEX process (DuPont SRL 

Mar 1968, pp. 9–10). Additional tests (scanning coulometry and mass spectrometry) were run on 

berkelium a year later, but information was not available on the amounts handled in these tests 

(DuPont SRL Mar 1969, p. 5; DuPont SRL Apr 1969, pp. 6–8.) Use of the pressurized cation 

exchange chromatographic process first shows in the October and December 1969 monthly reports. 

However, 250Bk was used as the tracer in these tests. The 250Bk was made from irradiation of a 249Bk 

target. Improvements in counting 249Bk using liquid scintillation counters were also being made at this 

time (DuPont SRL Oct 1969, p. 5; Du Pont SRL Dec 1969, pp. 7–8). 

Improvements in the chemical separation techniques were being tested in April 1970 “for isolating 

249Bk, 254Es, and 257Fm produced with the 252Cf in the present high flux campaign ....” That same 

monthly report also mentions that the 250Bk tracer (mentioned earlier) was created from irradiation of a 

10-μg (16 mCi) 249Bk target (DuPont SRL Apr 1970, pp. 18–19; DuPont 1971, pp. 37–40). The 

chemical separation techniques were apparently intended to be used in the Multipurpose Processing 

Facility, under construction in 1970 in the F Area hot canyon (DuPont 1984b, pp. 209–210). In this 

period SRS was also irradiating targets that were shipped to ORNL for separation of the berkelium, 

fermium, and einsteinium (DuPont 1971, pp. 39–40), and then, ironically, ordering the berkelium back 

from ORNL (DuPont SRL Jul 1970, p. 6; DuPont SRL Aug 1970, p. 6). The shipment from ORNL 

involved 250 μg (407 mCi) of 249Bk (DuPont SRL Aug 1970, p. 6). More separations were performed 

in September in what were still referred to as tests. “Batch 9 containing 1.25 mg of 252Cf from the 

244Cm campaign was purified by pressurized ion exchange chromatography with alphahydroxyisobutyrate (AHIB). This test confirms feasibility of separating Bk, Es, and Fm in the present 

252Cf campaign.” The amount of 249Bk separated was not given, but it would have been much less 

than the 1.25 mg of californium (DuPont SRL Sep 1970, pp. 6–7). 

The Separation Technology Section history indicates that work on isolating berkelium, einsteinium, 

and fermium continued into 1971 (DuPont 1984b, p. 225). The April SRL monthly report confirms two 

additional tests on separation of 249Bk from 252Cf and 154Eu (apparently a stubborn contaminant)

referred to as “small-scale” tests versus the “large-scale” tests in September 1970 (DuPont SRL Apr

1971, pp. 7–8). The Transplutonium Project history document, drafted in February 1971, indicates the 

“Californium I campaign is expected to yield 51 milligrams of 249Bk, 6 micrograms of 254Es, and 10 

picograms of 257Fm ( DuPont 1971, p. 40). Note that no mention was made of 254Fm. There are 

conflicting reports on the time of the Californium I campaign. The February 1971 document refers to it 

in the future, and other documents show californium work underway in early 1971 (DuPont WTD Feb

1971, pp. 5–6, 8, 13). However, the reactor history states that the Californium I campaign occurred 

from August 1969 to November 1970 (DuPont 1984c, p. 158). The difference might be the time the 

targets were in the reactor versus the time the targets were being analyzed after a cooling period. 

The later article on this work (from about 2000) refers to submilligram quantities of 249Bk rather than 

10s of milligrams (Harbour et al. 2000, p. 172), so it appears that the predicted amounts in the 

Californium I campaign were either not created or not recovered in the amounts expected. A 

presentation on the history of berkelium and californium through the end of 1974 stated that the 

ATTACHMENT C

BERKELIUM-249 AND FERMIUM-254

Page 2 of 5 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 40 of 45

material from the Californium I campaign was sent to ORNL for separation and purification of the 

berkelium in the Transplutonium Processing Plant (Crandall 1975, p. 11).

There is no indication that the elements higher than californium were actually isolated in the 

Multipurpose Processing Facility when it was operating. SRL monthly reports in 1972 and 1973 

indicate that production was focused on 238Pu heat sources and 252Cf neutron sources. The 1975 

historical presentation on berkelium and californium stated that “Slide 15 illustrates one of the latest 

flow sheets, that for the completed, but as yet unused, Multipurpose Processing Facility (MPPF) at 

Savannah River” (Crandall 1975, p. 11). Presumably “unused” referred to berkelium rather than

californium. The 1982 risk analysis report for the facility indicated that the plant was capable of 

extraction of curies of berkelium from the plant feed, but it did not state that that had actually occurred 

(Stoddard 1982).

A waste characterization report in 1994 indicated that 249Bk was still being produced in the reactors in 

the 1970s. Specifically, Figure 4 of that document shows 2.59 × 105

, 1.04 × 104, 1.16 × 104, and 

95 μg of 249Bk in the “receipt/production” column for fiscal years 1976, 1977, 1978, and 1980, 

respectively (Gibbs 1994, pp. 8, 19). The series of annual reports showing customers and shipments 

of various isotopes made in DOE laboratories for 1972 through 1980 showed that all 249Bk was made 

at ORNL and that SRS only received berkelium in 1972 (Simmons 1973, p. 14; Simmons and Mandell 

ca. 1975, p. 12; Simmons and Gano ca. 1976, p. 13; Simmons 1977, p. 13; Simmons 1978, p. 15; 

Burlison and Laidler 1979, p. 15; Burlison 1980, p. 11; Burlison 1981, p. 12). Therefore, the amounts 

in the Gibbs document must have been produced and used internally at SRS, which is consistent with 

the statement in the Gibbs document that the 249Bk was “tracked in the Accountability records and 

obtained from the reactor production codes.” This is verified by the SRS Automated Inventory 

Management System (AIMS) monthly reports and the Nuclear Materials Management and Safeguards 

System (NMMSS) (DOE 2012a,b) files. For instance, the October 1975 AIMS report listed 65 “pieces” 

(presumably targets) in K-Reactor amounting to about 2.95 × 105 μg, which is nearly twice the 

inventory in the Gibbs document for the end of June 1975 (DuPont 1975–1979, pp. 2–5). The 

difference, according to the monthly NMMSS reports, was production of 8.5 × 104

, 1.8 × 104, and 1.7

× 104 μg in July, August, and September 1975, respectively. This production rate continued through 

December 1975 and occurred again at about the same rate during March through October 1976 (DOE 

2012a). The NMMSS report lists this production as an SRS-to-SRS transaction of “in fuel element 

and target fabrication process,” consistent with irradiation of targets in one of the reactors. The AIMS 

report for October 1976 listed the same 65 pieces in K-Reactor with an inventory increase to 4.08 × 

105 μg and 24 pieces in P-Reactor with a small inventory of 277 μg, which is consistent with the 

inventory in the Gibbs document for the same time (DuPont 1975–1979, pp. 22–24). The October 

1977 AIMS report shows the 65 pieces still in K-Reactor but with much less activity, about 

3.10 × 105 μg, which agrees well with the Gibbs report for calendar year 1977. This amount implies 

that there was some continued production of the 249Bk in the reactor for some of the year, followed by 

radioactive decay. So the targets were still at K-Reactor but not being irradiated at that time. There 

were also 23 pieces in P-Reactor amounting to about 2,170 μg, implying that the targets in P-Reactor 

had continued to be irradiated during the year (DuPont 1975–1979, pp. 35-37). The NMMSS reports 

show continued production through January 1978, although the amounts are much smaller, generally 

in the hundreds of micrograms per month (as opposed to tens of thousands of micrograms per month 

in the 1975 to 1976 period). This production was listed as “in reactor, product” (DOE 2012a), and is 

consistent with the smaller targets in P-Reactor. In February and March 1978 there was a production 

of only 3 μg each month, also listed as “in reactor, product,” then there was no more production of this 

type. The last production occurred in November 1979, consisting of 95 μg listed as “aluminum 

process, irradiated,” which is consistent with the Gibbs report (DOE 2012b).

ATTACHMENT C

BERKELIUM-249 AND FERMIUM-254

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None of the reports are completely clear as to what happened to the targets after irradiation 

(production). There were no shipments off the site from 1974 through 1992. All of the transactions 

are listed as from and to 703-42A with the same shipper and receiver code except for two shipments 

in 1979 listed as removals to “underground waste storage – liquid” (DOE 2012a). A little of the 

material is listed as being removed due to “fission and transmutation,” which would have been 

occurring while the targets were still in the reactors. Most of the inventory is just shown as decaying 

away with only 7 μg left by the end of 1992 (DOE 2012b). The decay bookkeeping category does not 

preclude having the targets chemically manipulated and the 249Bk purified. In the transaction reports, 

there is mention of production of a small amount of oxides (totaling 15 μg) in 1974 and a few listings 

of production of “flats, tubes, and extruded metal” in amounts varying from <1 to 44 μg in late 1975 

through June 1976 (DOE 2012a). This might indicate some manipulation of the targets, although 

without chemical purification 249Bk would not have been the predominant radionuclide. The category 

“aluminum process, irradiated” first shows in December 1978 with a small inventory of 95 μg that grew 

to 1,378 μg by March 1979. The inventory in this category was then reduced consistent only with 

radiological decay through 1989 when it decayed below 1 μg (DOE 2012b). So the dates of the first 

entries of “aluminum process, irradiated” align somewhat with the dates of transfer of some material to 

underground waste storage. This might indicate that some of the targets were chemically 

manipulated in late 1978 and 1979, producing some waste. 

The NMMSS report covering 1993 to present (DOE 2012c) shows the continuation of the decay of 

the 7 μg of material shown in the 1974 through 1992 report, down to 4 μg by August 1993. Then in 

October 1993 the inventory jumps to 58 μg listed as burial material. This inventory stays on the books 

undecayed through June 2006. The transaction report for July 2006 shows all 58 μg being removed 

without any description or movement so presumably this was to show that the buried material had 

decayed away (DOE 2012d). Throughout this period there is no movement of material off the site or 

between material accountability areas.

From an intake perspective, it is important to note concerning general contamination in gloveboxes, 

caves, accessible surfaces, and air contamination, that the overwhelming constituents would have 

been 244Cm, 252Cf, and 243Am, which were being handled in large amounts. For instance, disposal of 

244Cm contaminated waste to the burial grounds in 1969 totaled 12,185 Ci (DuPont RC Dec 1969, p. 

8). The RC Group monthly reports show that handling of 252Cf in the hundreds of curies range in the

SRL was common during this period. Possible exposure to pure submilligram quantities of 249Bk (1.6 

Ci/mg) or the microgram/picogram quantities of einsteinium and fermium would have occurred only to 

those handling and packaging the final products of the chemical process, which would have been 

handled with extreme care because of the rarity of the products. For instance, there is no mention of 

contamination by berkelium, einsteinium, or fermium in the April 1970 RC Group report (DuPont RC 

Apr 1970) (when the highest activity tests were conducted), or in any of that series of reports in 1969 

or 1971.

Possible exposure to pure 249Bk appears to have occurred at nanocurie levels in 1965, sub millicurie

levels in 1966, and tens to hundreds of millicurie levels in 1967 and intermittently thereafter through 

1971. There was continued production of 249Bk through irradiation of targets in the reactors from 1974 

to 1979. No conclusive evidence has been found that 249Bk was extracted from those targets, but the 

transfer of 58 μg to underground storage, apparently as a liquid, seems to indicate some chemical 

processing. Most of the about 1 × 105

 μg of material made in the 1970s seems to have just decayed 

away with some being transferred to waste storage. No facilities other than 703-42A are mentioned in 

the transaction reports. General contamination in the SRL workspace would have been dominated by 

other radionuclides, principally 244Cm, 252Cf and other isotopes of californium, 243Am, and 154Eu; the 

ATTACHMENT C

BERKELIUM-249 AND FERMIUM-254

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Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 42 of 45

risk of intake from the small amounts of pure 249Bk seem low. It seems unlikely that there was 

exposure to purified 249Bk after 1972; the 1972 date is based on receipt of a shipment from ORNL in 

1972.

Fermium-254

Both 254Fm and 257Fm are alpha emitters so, if present in the body, would have been excreted in urine 

and in theory would have been measured by the trivalent actinide urinalysis. However, because of its 

3.2-hour half-life, essentially none of the excreted 254Fm would have been counted considering the 

time gap between excretion and counting of the sample. However, the same half-life consideration 

applies to time involved in separating and extracting the fermium from the californium and einsteinium. 

For instance, the 1971 history indicates that a 15-month cooling period between the time the targets 

were removed from the reactor until the start of chemical processing was normal (DuPont 1971, p. 

40). If hardly any cooling time was employed, then no 254Fm would be left in the target (the parent 

254Es also has a half-life in hours). Although the article from 2000 states “subnanogram amounts of ... 254Fm were separated” (Harbour et al. 2000, p. 172), that article does not mention 257Fm, which was 

the actual goal of the irradiation of einsteinium. It is possible that this was a typographic error and the 

sentence should have said 257Fm. The purpose of creating the higher order elements was for 

research on their physics and chemical natures, which would not have been possible on a very small 

amount of something with a 3.2-hour half-life. 

Tracer quantities of 254Fm were created twice in 1970 (April and September) for the special purpose of 

testing the procedure for separating fermium and einsteinium from californium. The activity of 254Fm 

was about 1 μCi in the April test and about 20 μCi in the September test (DuPont SRL Apr 1970, 

pp. 21–23; DuPont SRL Sep 1970, pp. 6–7; DuPont SRL Oct 1970, pp. 9–10). These activities of 

such a short half-life radionuclide material are not of dosimetric concern. It is concluded that no 

significant exposure occurred to 254Fm as a pure radionuclide.

ATTACHMENT C

BERKELIUM-249 AND FERMIUM-254

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Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 43 of 45

Interest in 171Tm (and 170Tm) as potential heat sources started in 1964. In the September monthly 

report of the SRL, under New Products – Special Test Irradiations, “quantities of 46Sc, 170Tm, 171Tm, 

and 204Tl are being made in the current 244Cm-producing charge” (DuPont SRL Sep 1964, p. 43), a 

test that was authorized in July (Allen 1964). The October and November monthly reports do not 

mention any actual handling of the material (DuPont SRL Oct 1964; DuPont SRL Nov 1964). An 

historical summary document also mentions the above irradiation in 1964 and states that an ion 

exchange process for separating the thulium from the erbium target had been developed and would 

be used on the irradiated material (DuPont 1984a, p. 93). 

The first chemistry to extract 170Tm and 171Tm from the irradiated erbium targets occurred in 1965 

using the aforementioned technique. This was at the laboratory development phase (DuPont 1984a, 

p. 118; DuPont SRL Feb 1965, pp. 43–44; DuPont SRL Apr 1965, p. 41). During this period the 

extracted material consisted of a combination of 170Tm and 171Tm because the target material was 

natural erbium. Because 170Tm is a strong beta emitter, intakes of this mixed isotopic material would 

have been detected by both air sampling and the fission product bioassay. Experimentation with 

enriching erbium to enhance the abundance of 170Er, thereby producing more 171Tm, began in the 

second half of 1965 (DuPont SRL Jun 1965, pp. 14–22; DuPont SRL Jul 1965, p 38; DuPont SRL Sep 

1965, p. 56; DuPont SRL Nov 1965, p. 31). Much of that work was on unirradiated erbium. 

Experimentation and improvement of this technique continued in 1966 (DuPont 1984a, p. 141). 

Experiments in January resulted in separation of 2.7 mg (2.9 Ci) of 171Tm (DuPont SRL Jan 1966, 

pp. 21–22). This material was used as a tracer in subsequent tests toward improving separation 

techniques (DuPont SRL Apr 1966, pp. 12–15). The RC Group monthly report for May 1966 indicates 

that a study on the radiation characteristics from and shielding design needed for 170,171Tm sources 

was underway (DuPont RC May 1966, p. 8). In June a different approach involving irradiation of 

169Tm to produce 171Tm was started, which was estimated to produce about 500,000 Ci of 170+171Tm in 

each of two targets. In addition, irradiation of two 1-g samples of erbium oxide enriched in 170Er 

began (DuPont SRL Jun 1966, pp. 9–15). The irradiated 169Tm targets were later shown to contain 

mostly 170Tm (DuPont SRL Dec 1966, p. 9). The RC Group December 1966 monthly report indicated 

that proposed operations for about 100 Ci of 171Tm in gloveboxes B-126 to B-130 were approved and 

that the associated beta and low-energy gamma activity was to be easily shielded. The report added 

that the aluminum cladding was to be removed in the High Level Caves (DuPont RC Dec 1967, p.10). 

Analyses of the irradiated enriched erbium oxide samples were conducted in May and June of 1967; 

these showed that nearly pure 171Tm was produced (DuPont SRL May 1967, p. 9; DuPont SRL Jun

1967, p. 8; DuPont IPHS Apr–Jun 1967, p. 21; DuPont 1984a, p. 163). Two hundred Ci of 171Tm had 

been produced by this method at that time. In the following quarter about 60 Ci of this material was 

purified from the residual erbium using cation exchange chromatography (DuPont IPHS Jul–Sep 

1967, p. 22; DuPont SRL Jul 1967 p.14). No work with 171Tm was mentioned in subsequent SRL

monthly reports until May 1968; it was also not mentioned in the fourth quarter (calendar year) 1967 

IPHS report. The latter report indicated that 238Pu would be a better heat source than either 171Tm or 

147Pm because of their short half-lives, difficulty in separation and purification, and cost (DuPont IPHS 

Oct–Dec 1967, p. 45). The first quarter 1968 IPHS report only discusses work with 244Cm and 170Tm; 

external radiation measurements were made on a 3-kCi 170Tm source that also contained 1.7 kCi of 

171Tm, but no chemistry was done (DuPont IPHS Jan–Mar 1968, p. 29; Sanders et al. 1969). In May 

1968 another separation of 171Tm from enriched erbium was made with improved chemistry methods 

(30 mg 171Tm, 57 Ci) (DuPont IPHS Apr–Jun 1968, pp. 45–46; DuPont SRL May 1968, pp. 12–13). 

Both of these references also stated that this concluded the work with 171Tm. That the work was 

“deferred” was reiterated in a July 1969 letter from E. L. Albenesius, Research Manager in the SRL

(Albenesius 1969). No further work with potential for intakes of pure 171Tm was found.

ATTACHMENT D

THULIUM-171

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Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 44 of 45

No mention of the work with 171Tm was found in the series of monthly reports by the RC Group of the 

SRL in 1968 or 1969, which merely indicates that there was no significant loss of control.

Bebbington, in his history of DuPont years at Savannah River, concluded “among the unusual 

radionuclides produced in research quantities and evaluated as heat sources and for a number of 

other specialized uses were: curium-244, promethium-147, thulium-170 and -171, scandium-46, 

thallium-204, and polonium-210. ... None of these nuclides, though, found significant applications” 

(Bebbington 1990, p. 200).

The period of work with potential for intakes of essentially pure 171Tm appears to be limited to 1966 

through 1968. The work took place in the SRL. 

ATTACHMENT D

THULIUM-171

Page 2 of 2 
Document No. ORAUT-RPRT-0054 Revision No. 01 Effective Date: 01/03/2013 Page 45 of 45

Americium-242m is a beta emitter with a half-life of 152 years. It has an alpha decay branch, but the 

yield is only 0.5%. Because it does not emit significant alpha radiation, it would not be counted in the 

trivalent actinide bioassay method. The principal decay branch of 242mAm (99.5% yield) beta-decays 

to 242Am, which beta-decays to 242Cm with an 82.7% yield and 16-hour half-life. So this latter 

transition is nearly instantaneous, and equilibrium between 242mAm and 242Am is rapidly established. 

Curium-242 is an alpha emitter with a 163-day half-life that would have been counted by the trivalent 

actinide bioassay analysis. However, it takes quite a few years for equilibrium to be established 

between 242mAm and 242Cm, and equilibrium is never 100% because of the branching ratios. 

The question is whether the buildup of 242Cm is rapid enough and sufficient such that the trivalent 

actinide bioassay would have been an appropriate bioassay for 242mAm exposure. Because 

americium and curium behave the same in the body, body biokinetics do not change the ratio of 242Cm

and 242mAm as a function of time. Assuming workers were chronically exposed to initially pure 242mAm, 

the rate at which 242Cm activity would build up in the body is shown in Table E-1, assuming a unit 

activity intake per day of 242mAm.

Table E-1. Activity of 242Cm per 

unit daily intake of 242mAm. 

Time after start of 

chronic intake

Cm-242 

activity

1 d 0.0018

5 d 0.0087

10 d 0.017

50 d 0.082

90 d 0.140

180 d 0.249

1 yr 0.407

1.5 yr 0.504

2 yr 0.571

3 yr 0.648

The derivations and a plot of the buildup as a function of time are shown in La Bone (2011). It is 

unlikely that the 242mAm would have been pure in the workplace environment; that is, airborne or 

surface contamination would have been a mix of 242mAm and 242Cm, so the actual ratios would have 

been larger than those in Table E-1, up to an equilibrium of 0.83. The activity values in Table E-1 

show that an annual or biennial trivalent actinide bioassay would have had sufficient 242Cm to monitor 

for chronic intakes of 242mAm.

ATTACHMENT E

Page 11 of 1 

ATTACHMENT E

AMERICIUM-242M