EPA-670/2-73-053-1
August 1973
Environmental Protection Technology Series
RECOMMENDED METHODS OF
REDUCTION, NEUTRALIZATION, RECOVERY OR
DISPOSAL OF HAZARDOUS WASTE
Volume IX Nuclear
Office of Research and Development
U.S. Environmental Protection Agency
Washington, D.C. 20460
-------
EPA-670/2-73-053-1
August 1973
RECOMMENDED METHODS OF
REDUCTION, NEUTRALIZATION, RECOVERY
OR DISPOSAL OF HAZARDOUS WASTE
Volume IX. National Disposal Site Candidate
Waste Stream Constituent Profile Reports -
Radioactive Materials
By
R. S. Ottinger, J. L. Blumenthal, D. F. Dal Porto,
G. I. Gruber, M. J. Santy, and C. C. Shin
TRW Systems Group
One Space Park
Redondo Beach, California 90278
Contract No. 68-03-0089
Program Element No. 1D2311
Project Officers
Norbert B. Schomaker
Henry Johnson
Solid and Hazardous Waste Research Laboratory
National Environmental Research Center
Cincinnati, Ohio 45268
Prepared for
OFFICE OF RESEARCH AND DEVELOPMENT
U.S. ENVIRONMENTAL PROTECTION AGENCY
WASHINGTON, D.C. 20460
-------
REVIEW NOTICE
The Solid Waste Research Laboratory of the National Environmental
Research Center - Cincinnati, U.S. Environmental Protection Agency has
reviewed this report and approved its publication. Approval does not
signify that the contents necessarily reflect the views and policies of
this Laboratory or of the U.S. Environmental Protection Agency, nor does
mention of trade names of commercial products constitute endorsement or
recommendation for use.
The text of this report is reproduced by the National Environmental
Research Center - Cincinnati in the form received from the Grantee; new
preliminary pages and new page numbers have been supplied.
-------
FOREWORD
Man and his environment must be protected from the adverse
effects of pesticides, radiation, noise and other forms of pollu-
tion, and the unwise management of solid waste. Efforts to protect
the environment require a focus that recognizes the interplay between
the components of our physical environment—air, water, and land.
The National Environmental Research Centers provide this multidisci-
plinary focus through programs engaged in:
• studies on the effects of environmental
contaminants on man and the biosphere, and
• a search for ways to prevent contamination
and to recycle valuable resources.
Under Section 212 of Public Law 91-512, the Resource Recovery
Act of 1970, the U.S. Environmental Protection Agency is charged
with preparing a comprehensive report and plan for the creation of
a system of National Disposal Sites for the storage and disposal of
hazardous wastes. The overall program is being directed jointly by
the Solid and Hazardous Waste Research Laboratory, Office of Research
and Development, National Environmental Research Center, Cincinnati,
and the Office of Solid Waste Management Programs, Office of Hazard-
ous Materials Control. Section 212 mandates, in part, that recom-
mended methods of reduction, neutralization, recovery, or disposal
of the materials be determined. This determination effort has been
completed and prepared into this 16-volume study. The 16 volumes
consist of profile reports summarizing the definition of adequate
waste management and evaluation of waste management practices for
over 500'hazardous materials. In addition to summarizing the defini-
tion and evaluation efforts, these reports also serve to designate a
material as a candidate for a National Disposal Site, if the material
meets criteria based on quantity, degree of hazard, and difficulty of
disposal. Those materials which are hazardous but not designated as
candidates for National Disposal Sites, are then designated as candi-
dates for the industrial or municipal disposal sites.
A. W. Breidenbach, Ph.D., Director
National Environmental Research Center
Cincinnati, Ohio
m
-------
GLOSSARY
Rad - Radiation Absorbed Dose - the absorbed dose of any nuclear
radiation which is accompanied by the liberation of 100 ergs of
energy per gram of absorbing material.
Rem - Roentgen Equivalent Man - a criterion of biological
injury which is defined as:
Dose in rems = dose in rads x (Relative Biological Effectiveness)
= dose in rads x
physical dose of 200 -ky X-rays to produce effect of interest
physical dose of comparison radiation to produce same effect
IV
-------
TABLE OF CONTENTS
VOLUME IX
NATIONAL DISPOSAL SITE CANDIDATE
WASTE STREAM CONSTITUENT PROFILE REPORTS
Radioactive Materials
Page
Carbon-14, Cobalt-60, Iridium-192, Radium-226 . . . . 1
Cesium-134, Cesium-137 (Barium-137m) 17
Hydrogen-3 . . . 41
Iodine-129, Iodine-131, Krypton-85, Xenon-133 53
Plutonium-238, Plutonium-239, Plutonium-240, Plutonium-241,
Americium-241, Americium-243, Curium-242, Curium-244 77
Ruthenium-106 (Rhodium-106), Cerium-144 (Praseodymium-144),
Promethium-147 103
Strontium-90 (Yttrium-90) . 127
Zirconium-95, Niobium-95 147
-------
PROFILE REPORT
Carbon-14, Cobalt-60, Iridium-192, Radium-226
1. GENERAL
Introduction
These four radionuclides are representative of the radioisotopes of
commercial interest which are generally produced, distributed, and used
by the private sector of the economy. With the exception of radium-226,
these radioactive materials are either directly, or indirectly, subject
to federal regulations. In many states individuals can purchase radium
without proof of competence to handle it safely. It is interesting to
note that two-thirds of the companies selling radium devices are located
in non-agreement states and account for 91 percent of the total devices
sold.2296
Radium-226 is a naturally occurring radionuclide first discovered in
1898. It has the longest history of use of any radioactive material
and is also one of the most hazardous radionuclides. The ingestion
of the luminous dial paint prepared from radium-226 was the cause of death
of many of the early dial painters before the hazard was fully understood.
Much of what is known of the biological effects of ionizing radiation on
man is based on the effects of radium ingested by these early watch dial
painters. It is estimated that over 3 million timepieces containing radium
are sold annually and the exposure to the total population is probably greater
than that from all other consumer products containing radioactive material.
A comparison of the estimated annual radiation dose received by the critical
organs from a wrist watch containing 0.15 microcuries of radium-226 with
the International Commission on Radiological Protection (ICRP)0563 annual
limit for that body part is attached (Table 1).
-------
TABLE 1
ESTIMATED ANNUAL RADIATION DOSE FROM A WRIST
WATCH CONTAINING 0.15 MICROCURIES OF RADIUM-226
Organ
Skin of the wrist
Lens of eye
Blood-forming tissue
Gonads
Estimated Annual Dose
millirem
4,800
110
30
10
ICRP Annual Limit
For Body Part °563
millirem
7,500
500
500
500
-------
Radium-226 has a half-life of 1,602 years and decays by the emission
of high energy, 5.68 Mev alpha particles. Radium-222 eventually decays
to lead and its decay chain is attached (Table 2).
Carbon-14, cobalt-60, and iridium-192 are produced in nuclear reactors
by the bombardment of a target material with neutrons. Their principal use
is in the fields of nuclear medicine, radiation detection and control
equipment, and in nuclear power sources. Carbon-14 has a half-life of
5,730 years and emits only a 0.156 Mev beta particle. Cobalt-60 has a
half-life of 5.3 years and emits both beta and gamma particles. It decays
by the emission of a 0.319 Mev beta particle to form stable nickel-60.
Iridium-192 is both a beta and gamma emitter and has a half-life of 74.2 days,
Manufacture
Radium-226 is a daughter of uranium-238 and is found to occur
naturally in the earth's surface. Carbon-14, cobalt-60, and iridium-192
are produced by the bombardment of a target material with neutrons. This
is usually performed in a nuclear reactor. Carbon-14 is produced by the
neutron bombardment of nitrogen-14. Cobalt-60 is produced from cobalt-59
and iridium-192 is produced from iridium-191.
Uses
Radium-226 has been and still is used in a wide variety of products
and applications. In the past 10 years there has been a decrease in
radium usage made possible by the increased availability and acceptance of
use of other less hazardous radioactive materials. Radium-226 is used
in producing timepieces, electron tubes, record player brushes, gauges, fire
detectors, and in various self-luminous products. It is used in medicine
for the treatment of tumors, superficial skin lesions, lymphoid tissue, and
other diseases. It is estimated that 330 curies of radium-226 contained
in 50,000 sources are used in medical applications at 2,300 facilities
which provide approximately 85,000 medical treatments per year.2296
2a
-------
TABLE 2
RADIUM-226 DECAY CHAIN
Nuclide
Ra226
88|a
Jn222
OO 1
T 91Q
84P
82f
831
84'
821
83f
84f
OO'
0
b214
i214
o214
b210
.210
o210
206
b Ub
Name
Radium-226
Radon-222
Polonium-218
Lead-214
Bismuth-214
Polonium-214
Lead-210
Bismuth-210
Polonium-210
Lead-206
Half-Life
1602 years
3.8 days
3.5 minutes
26.8 minutes
19.7 minutes
164 miroseconds
21 years
5 days
138 days
Stable
Major Radiation
Alpha and gamma
Alpha and gamma
Alpha
Beta and gamma
Alpha, beta and gamma
Alpha
Beta and gamma
Beta
Alpha
-------
Carbon-14 is used in medicine to study metabolic diseases and in
radioisotope gauges. Iridium-192 is widely used in the field of medicine
in radiographic units. Cobalt-60 is used in isotopic power devices, in
teletherapy units to treatcancer, and in radiation processing applications.
Presently, there are approximately 1,830 cobalt-60 teletherapy units in
2151
use with each unit containing about 3 thousand curies of cobalt-60.
Sources and Types of Wastes
In almost all applications, these materials are used in the solid form.
They are generally used as a sealed source and the material is retained
within the encapsulating material. When radium-226 is used in self-luminous
compounds,it is retained within the crystalline radium salt; however, there
is some release of radium from luminous compounds. These materials are
generally distributed throughout the country and can be found in hospitals,
commercial facilities, and in households.
Physical and Chemical Properties
The physical and chemical properties of carbon-14, cobalt-60,
iridium-192, and radium-226 are included in the attached worksheets.
Carbon forms a vast number and variety of compounds with hydrogen,
oxygen, nitrogen, and other elements. Cobalt is a brittle, hard material
that resembles iron and nickel. Iridium is a metal of the platinum
family. It is the most corrosion-resistant metal1 known and is the
second heaviest known element. Radium is an alkaline metal that acts
like calcium and barium chemically. It reacts with nitrogen and is
mainly used in the form of salts.
2. RADIATION HAZARD
Radium-226 is one of the most hazardous radioactive materials known.
Radium-226 replaces calcium in the bone structure and is a source of
irradiation to the blood-forming organs. This, along with its long
4
-------
half-life (1,602 years) and high radiation energies, places it in the
highest radiotoxicity group. It also has the longest history of use
of any radioactive material, and most of the standards for the effects of
ionizing radiation on man are based on this material. Carbon-14, cobalt-60,
and iridium-192 are moderately dangerous radioactive materials.
The effects of their radiation exposure are primarily dependent on
the amount of radiation and the portion of the body affected. The effects
of whole-body gamma radiation exposure are: (1) 5 to 25 rads, minimal dose
detectable by chromosome analysis or other specialized analyses, but not by
hemogram; (2) 50 to 75 rads, minimal acute dose readily detectable in a
specific individual (e.g., one who presents himself as a possible exposure
case); (3) 75 to 125 rads, minimal acute dose likely to produce vomiting
in about 10 percent of people so exposed; (4) 150 to 200 rads, acute dose
likely to produce transient disability and clear hematological changes
in a majority of people so exposed; (5) 300 rads, median lethal dose for
Of.CC
single short exposure. These effects are for a single large dose of
radiation or a series of substantial doses in a short interval of time to
the total body. The dose delivered to a particular body organ following
the inhalation of 1 microcurie of each of these radionuclides is attached
(Table 3). For radium-226 the dose delivered to the bone is 300 rem
following the inhalation of 1 microcuries (1.01 micrograms). The dose
delivered to the bone following the injection of 1 microcurie into the
body via a wound is 1,000 rem.
Standards for prolonged exposure over a 50-year period have defined
the single dose limit in terms of the maximum permissible dose accumulated
in a period of 13 weeks. The whole body exposure limit is 3 rem per quarter
for a radiation worker and the accumulated dose limit is 5(N - 18), where
N is the individual's age in years. Limits for the thyroid, bone, and
other organs have also been defined. Values of the total body burden for
each radionuclide required to produce the maximum permissible dose rates
defined above have been compiled. For radium-226 and carbon-14 the
critical organ is the bone and the maximum permissible body burden is 0.1
and 300 microcuries, respectively. For cobalt-60 the critical organ is the
total body and the maximum body burden is 10 microcuries. For iridium-192
-------
TABLE 3
CARBON-14, COBALT-60, IRIDIUM-192, AND RADIUM-226 DOSE
TO A PARTICULAR BODY ORGAN FOLLOWING INHALATION
OF ONE MICROCURIE OF THE NUCLIDE
Isotope
Form
Organ
Dose
rem
Carbon-14
Carbon-14
Insoluble
Soluble
Lung
Bone
0.06
0.002
Cobalt-60
Cobalt-60
Insoluble
Soluble
Lung
Total Body
0.77
0.008
Iridium-192
Iridium-192
Insoluble
Soluble
Lung
Kidney
0.27
0.05
Radium-226
Radium-226
Insoluble
Soluble
Lung
Bone
130
300
-------
the critical organ is the kidney and the maximum body burden is 6 microcuries.
3. OTHER HAZARDS
As an element carbon is not very toxic. In the form
of dust it can cause irritation of the eyes and mucous membranes. The
toxici-ty of cobalt is also low. It can produce dermatitis and is slightly
irritating to the skin. The toxicity of iridium is unknown even though
soluble iridium compounds are said to be toxic. Radium's main hazard
is its radioactivity. The fire and explosive hazard of the above materials
is moderate.
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling, Storage, and Transportation
Since these radionuclides are hazardous to man by inhalation, ingestion,
or direct radiation exposure, great care should be exercised in their
handling. Handling should be set up to prevent excessive exposure to
personnel. Special procedures and adequate radiation shielding are
required in their handling. Radium»226 is an alpha emitter. These alpha
particles have little penetrating power and can be stopped by a sheet of
writing paper. The beta particles emitted by carbon-14, cobalt-60, and
iridium-192 range in energy from 0.16 to 0.67 Mev and cannot penetrate
more than 0.015 to 0.085 inch of water. The gamma radiation emitted by
cobalt, iridium-192, and radium-226 is highly penetrating and high-
density shields, such as lead, are required to stop the radiation. For
cobalt-60 which emits high energy gamma rays (1.33 Mev), approximately
3.5 inches of lead or 20 inches of concrete are required to reduce its
radiation by a factor of 100. To detect and control personnel exposure
to their radiation, all persons working with this material should wear
dosimetry devices which directly indicate the dose. Other commonly used
devices are the film badge and the thermoluminescent dosimeters (TLD).
Specially constructed containers in controlled areas should be used
for storing large quantities of these materials. They should be protected
by both a primary and a secondary containment barrier. Special monitoring
-------
systems and proper warning signs should be located in the general area of
the storage facility. Special precautions are required in the storage
of radium-226 since approximately one gram of radium produces about 0.001
milliliter of radon gas per day. Thus, stored radium should be vented to
prevent the build-up of radon gas.
Radium-226 is classified as a transport group I radionuclide,
cobalt-60 and iridium-192 are classified as transport group III radionuclides,
and carbon-14 as a transport group IV radionuclide by the Department of
Transportation. The rules and regulations governing their transportation
are given in the Code of Federal Regulations (CFR) Title 14—Transportation,
Parts 170 to 190. The radium-226 content is limited to 0.001 curies
for a Type A package and 20 curies for a Type B package defined in 40CFR173.
For cobalt-60 and iridium-192 these limits are increased to 3 and 200 curies.
For carbon-14 the limits are 20 and 200 curies. The release rate of these
materials is limited to zero under the specified accident conditions for
Type A and B packages.2150
Disposal/Reuse
The disposal of these materials is governed by the AEC Manual
Chapter 0524 °559 and 10CFR20.2149 Two sets of standards have been
established for the permissible radiation exposure in unrestricted
areas. One is for the greatest dose received by an individual and the
other for the average dose received by the general population. The
standards for the safe release of these materials to the environment
in an unrestricted area are contained in 10CFR202149 and their release
should not result in concentrations in air and water greater than those
listed in this report (Table 4). These concentrations apply to an
individual and their release may be limited if a suitable sample of the
population is exposed to one-third concentrations in air or water
specified in this report (Table 4).
8
-------
TABLE 4
CARBON-14, COBALT-60, IRIDIUM-192, AND RADIUM-226
MAXIMUM PERMISSIBLE CONCENTRATIONS2149
Radionuclide
Form
Concentration in Air Concentration in Water
(mi crocuri es/mi11i1i ter) (mi crocuri es/mi11i1i ter)
Carbon-14
Carbon-14
Cobalt-60
Cobalt -60
Iridium-192
Iridium-192
Radium-226
Radium-226
Soluble
Submersion*
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
x 10
x 10
1 x 10
3 x 10
4 x 10
9 x 10
3 x 10
2 x 10
-7
-6
-8
-TO
-9
-10
-12
-12
8 x 10
-4
5 x 10
3 x 10
4 x 10
4 x 10
3 x 10
3 x 10
-5
-5
-5
-5
-8
-5
*Submersion means that values given are for submersion in an infinite
cloud of gaseous material.
-------
Although rarely practiced, the disposal by release in a sanitary sewage
system is limited to 0.1 microcuries of radium-226, 10 microcuries of
cobalt-60, 100 microcuries of iridium-192, and 1,000 microcuries of carbon-
14. The disposal by burial at any one location and time is limited to
100 times the above amounts.
The above regulations for the transportation, release, and disposal
of these materials apply only to licensed materials. Radium and other
accelerator-produced radionuclides are exempt from AEC control and
their regulation and control is a state function. In some states the
possession and use of these materials requires licensing and in others
no license is required. Thus, the disposal, transportation, and
use of these materials in some uses is not subject to the above
regulations.
5. EVALUATION OF WASTE MANAGEMENT PRACTICES
Option No.l-Land Burial. Land burial of carbon-14, cobalt-60, and
iridium-192 wastes, in small concentrations, at approved sites that are
acceptable from a geologic standpoint, is an acceptable means of disposal.
4
Their concentration should not be in excess of 10 times the maximum
214Q
permissible concentration for the general population in 10CFR20.
All wastes to be disposed of should be in a solid form and encapsulated
in a suitable container. The burial site design, geology, and hydrology
should be in conformance with the criteria used in selecting and licensing
the present commercial burial sites. ^ This method of disposal is not
considered satisfactory for the disposal of radium-226 because radium-226
has such an extremely long half-life and a high radiotoxicity.
Option No.2-Near-Surface Solid Storage. The storage of solidified
radium-226 wastes in engineered storage facilities offers the best inter-
mediate method for storage of these wastes. The technology for these
facilities has been developed and the wastes will be under surveillance
and control and can be retrieved, should this be required. The wastes
will be stored in stainless-steel lined concrete vaults.
10
-------
Option No.3-Sa1t Deposits. This method offers the best potential
for the disposal of radium-226 wastes since bedded salt deposits are
completely free of circulating ground waters. This method of disposal
has been under study by the Oak Ridge National Laboratory since 1957,
and in November of 1970 a committee of the National Academy of Sciences
recommended that the use of bedded salt for the disposal of radioactive
nyoo
waste is satisfactory. J Recent questions concerning the adequacy of
this method have resulted in the need for further development work before
it can be accepted as an ultimate method of disposal. The critical
problem is the selection of a site that meets the necessary design and
geological criteria.
To summarize, small concentrations of carbon-14, cobalt-60, and
iridium-192 can be disposed of by land burial at approved sites. Large
concentrations of these materials, and radium-226 should be disposed of
in salt deposits following storage if necessary in near-surface
engineered facilities.
6. APPLICABILITY TO NATIONAL DISPOSAL SITE
Carbon-14, cobalt-60, iridium-192, and radium-226 are candidates
for a National Disposal Site due to their large commercial usage and
high health hazard. The recommended process for the disposal of carbon-14,
cobalt-60, and iridium-192 wastes, in small concentrations, is by land
burial. The recommended process for the disposal of radium-226 and large
concentrations of the above radionuclides is disposal in salt beds.
11
-------
7. REFERENCES
0559. AEC Manual, Chapter 0524. Standards for radiation protection
May 12, 1964. lip.
0563. Recommendations of the International Commission on Radiological
Protection, ICRP Publication 2. Report on Committee II on
permissible dose for internal radiation (1959). Peryamon
Press, 1959. 232 p.
0733. Report by the Committee on Radioactive Waste Management. Disposal
of solid radioactive wastes in bedded salt deposits. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p.
0766. Sax, N. I. Dangerous properties of industrial materials. 3 ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
1423. Morton, R. J. Land burial of solid radioactive wastes: study
of commercial operations and facilities. Environmental and
Sanitary Engineering Branch, Atomic Energy Commission,
WASH-1143, 1968. 132 p.
2149. Code of Federal Regulations. Title ID—atomic energy. (Revised
as of Jan. 1, 1971). Washington, U. S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49--transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington, U. S.
Government Printing Office, 1971. 952 p.
2151. Atomic Energy Commission. The nuclear industry, 1971. Washington,
U. S. Government Printing Office, WASH-1174-71, 1971. 193 p.
2296. Pettigrew, G. L., E. W. Rodlnson, and G. D. Schmidt. State and
federal control of health hazards from radioactive materials
other than materials regulated under the Atomic Energy Act
of 1954. U. S. Department of Health, Education, and Welfare,
Rockville, Maryland, June 1971. 101 p.
2666. Recommendations of the National Council on Radiation Protection
and Measurements. Basic radiation protection criteria. NCRP
Report No. 39, Washington, Jan. 1971.
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Carbon-14
Half-life 5730 Years
Type of Decay Negative Beta
Molecular Wt. 14
Density 3.5 gm/cc
Melting Pt. 3550C
C14
1 6L
Boiling Pt. 4827C
Specific Power
Solubility Specific Activity 4.46 curies/gm
Cold Water insoluble _ Hot Water insoluble _
Others: insoluble in acid and alkali _
Decay Chain Radiation Energy Level & Intensities
R—
6C- - -/I14 Beta: 0.155 Mev (100%)
(5730 (Stable) Gamma: None
years)
Shipping Regulations Classified as a transport group IV radionuclide
by the Department or iransportation '
Comments
References: (1) 0766
(2) 2150
13
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name
Cobalt-60
Half-life 5.3 Years
Type of Decay Negative Beta
Molecular Wt. 60 Melting Pt. 1495C
Density 8.9 gm/cc
Structural Formula
27
Co
60
So1ubility__
Cold Water
Others:
Boiling Pt. 2900C
Specific Power 17.4 watts/gm
Specific Activity 1,130 curies/gin
insoluble
Hot Water
Decay Chain
B-
7Co
60
2T
(5.3 years)
soluble in acid
insoluble
,.60
(Stable)
Radiation Energy Level & Intensities
Beta: 0.319 Mev (100%)
Gamma: 1.332 Mev (100%)
1.173 Mev (100%)
Shipping Regulations Classified as a transport group III radionuclide
by the Department of Transportation.
Comments
References: (1) 0766
(2) 2150
14
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name
Iridium-192
Half-life 74.2 days
Type of Decay Negative Beta
Molecular Wt. 192
Density 22.4 gm/cc
Structural Formula
77
Ir
192
Melting Pt. 2443C
Boiling Pt. 4527C
Solubility^
Cold Water
insoluble
Specific Power
Specific Activity 9.17 x 10 curies/qm
Hot Water insoluble
Others: insoluble in aci
Decay Chain
d and alkali, slightly soluble in aqua regi a
Radiation Energy Level & Intensities
192
B-
77lr
(74d)
,0s
192
76'
(Stable)
Beta:
Gamma:
0.240 Mev ( 8%)
0.536 Mev (41%)
0.672 Mev (46%)
0.296 Mev (29%)
0.308 Mev (30%)
0.317 Mev (81%)
0.468 Mev (49%)
0.589 Mev ( 9%)
0.604 Mev ( 9%)
0.612 Mev ( 6%)
Shipping Regulations Classified as a transport group III radionuclide
by the Department of Transportation
Comments
References: (1) 0766
(2) 2150
15
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Radium-226 Structural Formula
Half-life 1602 Years
Ra226
88Ka
Type of Decay Alpha ^
Molecular Wt. 226 Melting Pt. 700C Boiling Pt. 1737C
Density 5 gm/cc Specific Power 1.3 x IP"4 vfatts/am
Solubility Specific Activity 0.99 curies/qm
Cold Hater reacts and evolves H p Hot Water reacts and evolves H^O
Others: reacts with acid
Decay Chain Radiation Energy Level & Intensities
a Alpha: 5.684 Mev (94%)
n,226 ^ 222 Decays 5.447 Mev (6%)
88Ra 86Rn to Lead
Gamma: 0.260 Mev (.007%)
(1602 years) (3.8 days) 0.186 Mev
Shipping Regulations Classified as a transport group I radionuclide by
the Department of Transportation
Comments
References: (1) 0766
(2) 2150
-------
PROFILE REPORT
Cesium-134, Cesium-137 (Barium-137m)
1. GENERAL
Introduction
Cesium-134 and cesium-137 are radioactive isotopes produced by the
fission of uranium and plutonium and will exist in combination with other
radioisotopes in high-level and low-level radioactive waste streams.
Cesium-134 has a fission yield of 8 percent and cesium-137 a fission
yield of 6 percent. Cesium-134 is transformed by beta decay into barium-134
which is a stable element. Cesium-137 has a half-life of 30 years and
emits only beta particles. Cesium-137 is in radiation equilbrium with
its short half-lived daughter (2.6 minutes), barium-137m. Barium-137m,
in turn, is transformed by isometric transition (decay from an excited
metastable state to lower state) into barium-317 which is a stable
element.
From the viewpoint of waste management, cesium-137, because of its
long half-life and high fission yield, is one of the most important
fission products produced in nuclear reactors. The only fission product
of comparable importance is strontium-90. The importance of cesium-137
is realized by the fact that it accounts for 10 percent of the total
fission product activity in nuclear reactor wastes after 1 year, and for
52 percent of the activity after 10 years. Cesium-134 is also of
significance due to its high fission yield (8 percent) and high specific
power (20 watts/gram).
The projected growth in the production of cesium-134 and cesium-137
will parallel that of the civilian nuclear power program. Present
17
-------
projections indicate that nuclear power will account for 30 percent of the
total power production by the year 1980. Annual production figures
from the civilian nuclear power programs for these two isotoped have been
projected to the year 2020 for light-water and fast-breeder reactor fuels
(Tables 1 and 2). The expected total accumulated radioactivity of cesium-
137 from the year 1970 to 2020 in millions of curies is0705:
Radioactivity
Year (megacuries) .
1970 5
1980 1,280
1990 6,540
2000 15,600
2020 57,500
Annual production figures for the nuclear weapons program were not
available.
Manufacture
Cesium-137 is one of many fission products produced directly by the
fission of uranium or plutonium. The fission reaction results from the
neutron bombardment of the nucleus. A typical fission reaction is:
1.235 . J rJ37 . D. 96 . 0 1
92U + on •* 55Cs + 37Rb + 2on
Cesium-134 is also produced in the fission reaction. Civilian
nuclear power plants are the major producers and a smaller amount is
produced at AEC facilities. At the present time there are 22 nuclear
power plants in operation with an additional 104 being built or
2151
planned. These isotopes will also be produced in fast-breeder
reactors by the fission of plutonium.
18
-------
TABLE 1
CESIUM-134 CONTENT IN HIGH-LEVEL WASTES PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM
Light-Hater Reactor Fuels
0705
Annual
Production
grams /year
curies/year
watts/year
Annual
Production
grams/year
curies/year
watts/year
Calendar Year
1970 1980 1°S5
.090xl05 5.12xl05 9.22xl05
.12 xlO8 6.66xl08 12.0 xlO8
.123xl06 .86xl05 1.55xl06
Fast-Breeder Reactor
Calendar Year
1970 1980 1985
.079xl05
.10 xlO8
.11 xlO6
1990
10.4xl05
13,6xl08
1.75xl06
Fuels
1990
.48xl05
.62xl08
.65xl06
2000
8.27xl05
10.8 xlO8
1.39xl06
2000
2.04xl05
2.66xl08
2.79xl06
2020
24.3xl05
31.6xl08
4. 07x1 O6
2020
6.13xl05
8.0 xlO8
8.39xl06
-------
TABLE 2
CESIUM-137 CONTENT IN HIGH-LEVEL WASTES PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM
Light-Hater Reactor Fuels
0705
s
Annual
Production
qrams/year
curies/year
watts/year
Annual
Production
grams/year
curies/year
watts/year
Calendar Year
1970 1980 1°85
.064xl06 3.64xl06 6.56xl06
.055xl08 3.16xl08 5.70xl08
.090x105 B.llxlO5 9.21xl05
Fast-Breeder Reactor
Calendar Year
1970 1980 1985
.45xl06
.39xl08
.64xl05
1990
7. 40x1 O6
6.45xl08
10.5xl05
Fuels
1990
2.70xl06
2.34xl08
3.8xl05
2000
5.86xl06
B.llxlO8
8.26xl05
2000
n.sxio6
lOxlO8
16.3xl05
2020
17.3xl06
15.0xl08
23.3xl05
2020
34. 5x1 O6
30x1 O8
49.0xl05
-------
TABLE 3
CESIUM-134 CONTENT IN HIGH-LEVEL FUEL REPROCESSING WASTES
Light-Water Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 30 MW/MTU)
0705
Fission
Products Gnir's. 'Tonne
Cesium- 134 134
Other Fission 34,966
Products
Fission
Products Grans/Tonne
Cesium-134 116
Other Fission 34,984
Products
After 1 year
I'atts/Tnnne
1,830
8,170
(Fuel
After 1 year
llatts/Tonne
220
12,780
Curies/Tonne
175,200
2,045,000
Fast-Breeder
Exposed to 33,
Curies/Tonne
21,200
3,408,800
Grans/Tonne
6
35,096
Reactor Fuels
000 MWD/MTU at
Grams/Tonne
1
34,899
After 10 years
'.'a tts/Tonne
87
943
30 MW/MTU)
After in years
Wa tts/Tonne
11
755
Curies 'Tonne
8,300
308,700
Curies/Tonne
1 ,000
280,000
-------
TABLE 4
CESIUM-137 CONTENT IN HIGH-LEVEL FUEL REPROCESSING WASTES
Light-Water Reactor Fuels
'(Fuel Exposed to 33,000 MWD/MTU at 58 MW/MTU)
0705
Fission After 1 year! After 10 years '
Products Grans/Tonne Uatts/Tonne Curies/Tonne Grams/Tonne l-'atts/Tonne Curies/Tonne
CS137 _ Ba137m
555
203,000
980
451
165,000
fO
Other Fission 32,900
Products
9,445
2,091,000
34,120
579
152,000
Fast Breeder Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 30 MW/MTU)
Fission
Products
Grams/Tonne
After 1 year
Uatts/Tonne
Curies/Tonne Grams/Tonne
After 10 years
Uatts/Tonne Curies/Tonne
Cs137 - Ba137m 1,230
565
207,000
1,000
458
168,000
Other Fission 33,770
Products
13,235
3,123,000
33,900
308
113,000
-------
Physical and Chemical Properties
The physical and chemical properties of cesium-134 and cesium-137
are included in the attached worksheets. Cesium is a member of the
highly electropositive alkeli metal group. In its compounds, cesium
has an oxidation state of -1 and is extremely reactive chemically. It
is a relatively volatile metal, melting at slightly above room temperature
(29..S C) and boiling at 690 C. Most of the salts of cesium are insoluble.
2- RADIATION HAZARD
Cesium-134 and cesium-137 are moderately dangerous radioactive
materials. The effects of their radiation are primarily dependent on
the amount of radiation and the portion of the body affected. The effects
yc.cc.
of whole body or gamma radiation exposure are ^ODD: (1) 5 to 25 rads,
minimum dose detectable by chromosome analysis or other specialized
analyses, but not by hemogram; (2) 50 to 75 rads, minimum acute dose
readily detectable in a specific individual (e.g., one who presents
himself as a possible exposure case); (3) 75 to 125 rads, minimum acute
dose likely to produce vomiting in about 10 percent of people so
exposed; (4) 150 to 200 rads, acute dose likely to produce transient
disability and clear hematological changes in a majority of people so
exposed; (5) 300 rads, median lethal dose for single short exposure.
Standards for prolonged exposure over a fifty-year period have defined
the single dose limit in terms of the maximum permissible dose
accumulated in a period of 13 weeks. The whole body exposure limit is
3 rem per quarter for a radiation worker and the accumulated dose
limit is 5(N - 18), where N is the individual's age in years. Limits
for the thyroid, bone, and other organs have also been defined.
Following the inhalation of 1 microcurie, the cesium-134 dose in
the insoluble form to the lung is 0.56 rem. In the soluble form the
dose delivered to the liver following the inhalation of 1 microcurie of
cesium-134 is 0.09 rem. The dose delivered to the liver following
injection of 1 microcurie into the body via a wound is 0.13 rem. The
cesium-137 dose to the lung following the inhalation of 1 microcurie
23
-------
in the insoluble form is 0.46 rem, and in the soluble form the dose to
the liver is 0.08 rem. The dose to the liver following the injection
of 1 microcurie into the body via a wound is 0.11 rem.
Values of the maximum permissible total body burden of cesium-134
and cesium-137 which are deposited in the total body and produce the
maximum permissible dose rate to a particular body organ have been
compiled. For both isotopes the critical organ is the total body
and the maximum permissible body burden is 20 microcuries for cesium-134
and 30 microcuries for cesium-137. The above body burden rates can be
expressed in terms of the maximum permissible concentration of cesium-134
and cesium-137 in air and water at the continuous exposure rate of
168 hours per week for a period of 50 years. For cesium-134, the maximum
permissible concentration in water is 9x10~ microcuries per milliliter
_Q
and in air 1x10" microcuries per milliliter. For cesium-137, the maximum
-4
permissible concentration in water is 2x10 microcuries per milliliter
-8
and in air 2x10" microcuries per milliliter.
3. OTHER HAZARDS
Besides its radiation hazard, cesium-134 and cesium-137 are only
slightly toxic to the human body if absorbed by inhalation, ingestion,
or through the skin. Their fire hazard is high, since they react
with oxidizing materials and can ignite spontaneously in moist air. Their
explosive hazard is moderate, but they do react with moisture to liberate
hydrogen. Cesium compounds have the same toxicity as cesium unless they
contain a more toxic radical.
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling, Storage, and Transportation
Cesium-134 and cesium-137 are in the greatest abundance in the
mixture of fission products which are separated from the spent fuel
during processing. Since both cesium-134 and cesium-137 are hazardous
to man by inhalation, ingestion, or direct radiation exposure, care
24
-------
Uses
Cesium-137 is used as a radiation source for both industrial and
medical equipment. Cesium-137 is used for radioisotope gauges and in
industrial radiographic units. It is also used in 7 percent of the
worlds total of medical teletherapy units. At present, cesium-137 is
obtained from high-level waste streams in the Richland, Washington,
chemical separations plant and is distributed through the Oak Ridge
National Laboratory, Isotopes Sales Department. In the future, if
its production is warranted by market demands, it could be available in
large quantities from commercial firms operating chemical reprocessing
plants. The AEC revenue from the distribution of cesium-137 through
the fiscal years 1968 to 1971 is: 2151
Year Revenue (Dollars)
1968 83,000
1969 42,000
1970 24,000
1971 59,000
These figures illustrate the relatively small commercial market for
cesium-137 at present. Cesium-134 has very limited commercial use.
Sources and Types of Wastes
Although produced in nuclear power reactors, the primary source
of cesium-134 and cesium-137 is in the high-level, aqueous waste streams
generated at the spent fuel processing plants. The range of chemical
compositions of the various types of waste streams obtained from the
processing step have been tabulated, and in all cases are aqueous
solutions of inorganic nitrate salts. The characteristics of cesium-134
and cesium-137 products in the high-level wastes at two different time
periods have been determined (Tables 3 and 4). These radionuclides are
also found in the secondary waste streams generated at spent fuel pro-
cessing plants. The activity in these wastes is quite low (less than a
tenth of a curie per gal.).
-------
is exercised in their handling. Special procedures and radiation
shielding are utilized in their handling. When separated from the fission
product mixture, both isotopes emit low energy beta particles which
cannot penetrate more than 0.09 in. of water or 0.007 in. of lead. To
stop their highly penetrating gamma rays high-density shields, such as
lead, are required. To reduce their gamma radiation by a factor of 10,
approximately 1.5 to 2 in. of lead or 8 to 9 in. of concrete is required.
To detect and control personnel exposure, all persons working with these
materials should wear dosimetry devices which directly indicate the dose.
Commonly used devices are the film badge and the thermoluminescent
dosimeters (TLD).
Cesium-134 and cesium-137, when separated from the other constituents
of the high-level waste stream, are stored in controlled areas in
specially constructed containers. They are protected by both a primary
and a secondary containment barrier. Special monitoring systems and
proper warning signs are located in the general area of the storage
facility.
Both cesium-134 and cesium-137 are classified as a transport group
III radionuclide by the Department of Transportation, and the rules and
regulations governing their transportation are given in the Code of
Federal Regulations (CFR) Title 49--Transportation, Parts 170 to 190.2150
Their content is limited to 3 curies for a Type A package and 200 curies
for a Type B package defined in 49CFR173. The limits are increased to
20 to 5,000 curies if their physical form meets the requirements of a
special form material. Their release rate is limited to zero under the
specified accident conditions for Type A and B quantities. The
allowable release of radioactivity from packages containing large
quantities of these isotopes is limited to gases and contaminated coolant
containing total radioactivity exceeding neither 0.1 percent of the total
radioactivity of the package nor 10 curies under the hypothetical accident
conditions prescribed in 49CFR173.
-------
The buik of the cesium-134 and cesium-137 arrives at the spent
fuel processing plants 1n the spen;; fuel assemblies which are contained
in Type B packages. The cesium-134 and cesium-137 may be stored as a
liquid at the processing plant for jp to 5 years and are then solidified and
leave the plant in Type B packages.
Disposal/Reuse
After solidification at the f'.:el processing facility, the bulk of
the cesium wastes may be held for UD to 5 years before being shipped to i
Federal repository. This repository has not yet been defined but. work
is in progress, under AEC sponsorship, on engineered surface storage
facilities which will be designed to store solidified high-level waste
for up to 100 years. In the interim, a satisfactory ultimate d-r^posa!
scheme, such as disposal in salt deposits, will be evolved.
The disposalof cesium-134 ana cesium-137 is governed by the AEC
Manual Chapter 0524°559 and 10CFR20. Two sets of standards have
been established for the permissible radiation exposure in unrestricted
areas. One is for the greatest dose received by an individual and the
other for the average dose received by the general population. The
radiation protection standards for these two groups are attached (Table 5).
The standards for the safe disposal of cesium-134 and cesium-137 also define
their maximum concentrations in air and water (Table 6).
The disposal of cesium-134 by release into a sanitary sewage system
is limited to 10 microcuries and the disposal of cesium-137 is limited
to 100 microcuries. The disposal oy burial in the soil at any one
location and time is limited to 1,000 microcuries of cesium-134 and
10,000 microcuries of cesium-137.21-49
-------
TABLE 5
RADIATION PROTECTION STANDARDS FOR INDIVIDUALS AND POPULATION GROUPS
2149
FOP EXTERNAL AND INTERNAL EXPOSURE
Type of Exposure
Dose to Individuals
at Points of Maximum
Probably Exposure
(rem per year)
Average Dose to a
Suitable Population
Sample
(rem per year)
l.'hole bocly, gonads, or
bone marrow
Thyroid or bone
Bone (alternate standards)
0.5
1.5
Body burden at 0.003
micrograms of radium
226 or tts biological
equivalent
0.17
0.5
Body burden of 0.001
micrograms of radium
226 or its biological
eoutvalent
-------
TABLE 6
CESIUM-134 AND CESIUM-137 MAXIMUM PERMISSIBLE CONCENTRATIONS
2149
Isotope
Cs-134
Cs-134
Cs-134
Cs-134
Cs-137
Cs-137
Cs-137
Cs-137
Exposure
Group
Individual
Individual
Population
Population
Individual
Individual
Population
Population
Form
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Concentration in Air %
(mi crocuri es/mi 1 1 i 1 i ter)
IxlO"9
4X10"10
.33xlO"9
1.3xlO-10
2x1 O"9
5X10'10
.67xlO"9
1.7X10-10
Concentration in
(mi crocuri es/mi 11
9x1 O"6
4x1 0~5
3x1 O"6
1.3xlO"5
2xlO"5
4xlO"5
.67xlO"5
1.3xlO"5
Water
ill ter)
-------
5. EVALUATION OF WASTE MANAGEMENT PRACTICES
Recovery
Small quantities of cesium-134 and cesium-137 may be recovered from
the high-level waste streams for industrial purposes. Another potential
advantage to recovery is that two or more radionuclides with similar
characteristics could undergo the same handling, storage, or disposal
processes. Although presently rarely recovered, various techniques,
in various stages of development, are being investigated.
Option No. 1 - Precipitation. Cesium-134 and cesium-137 can be removed
from the high-level reactor wastes by precipitating with nickel
ferrocyanide. Over 99 percent of the cesium can be removed satisfactorily
up to a pH of 10. ' Cesium can also be removed by precipitating
with phosphotungstate or absorbing on alumino-silicate zeolites.
Cesium can be further purified by ion exchange. Cesium is recovered from
the other high-level wastes for commercial purposes or for safety reasons.
Option No. 2 - Scavenging-Precipitation Foam Separation. The removal
of cesium-134 and cesium-137 from low-level radioactive waste water by
scavenging-precipitation foam separation has been studied by Oak Ridge
National Laboratory. The process consists of two steps: (1) precipi-
tating in a sludge-blanket clarification step; and (2) achieving final
decontamination in a foam separation column. The precipitation step
includes the use of Grundite clay for the sorption of cesium. The
Grundite clay increased the cesium decontamination factor (ratio of
initial to final concentration) approximately nine times. The overall
decontamination factor for cesium was eight. At present, further
development is required to reduce operating costs and increase processing
rates.
-------
Option No. 3 - Scavenging-Precipitation Ion Exchange. In this process
final decontamination is obtained by ion exchange columns from the
scavenging-precipitation process. The process includes a provision for
the recycle of the ion exchange waste to the scavenging-precipitation
step. All the removed radionuclides are concentrated in the clarifier
sludge. With most ion exchange resins sodium is removed with cesium.
Cesium can be separated from sodium by phenolic-base cation exchanges
at high pH values. For this process the overall decontamination factor
for cesium varied from 100 to 3,000. The use of ion exchange resins
with a scavenging-precipitation step is a fairly simple and efficient
means of removing cesium-134 and cesium-137 from low-level radioactive
aqueous wastes.
Option No. 4 - Water Recycle. The water recycle process is used for
decontaminating radioactive waste water and recycling the purified
water for reuse. This process has been demonstrated at the pilot plant
scale by Oak Ridge National Laboratory. The steps in the process
include: (1) clarification by the addition of coagulants;
(2) demineralization by cation-anion exchange; and (3) sorption on
granular activated carbon. The majority of the cesium is removed by
the cation-anion exchange. The cesium overall decontamination factor
was 14,000. The method is an improvement to the treat-and-discharge
methods, but further work is required until full-scale production use
is obtained.
Storage/Disposal
Option No. 1 - Land Burial. Land burial of cesium-134 and cesium-134
in low-level wastes, at approved sites that are acceptable
from a geologic and hydrologic standpoint, is an acceptable means of
disposal. The cesium-134 and cesium-137 concentrations should not be
in excess of 10 times their maximum permissible concentrations for the
general population in 10CFR20. All cesium wastes to be disposed
of should be in a solid form and encapsulated in a suitable container.
Liquid wastes should be solidified, preferably using asphalt, in
31
-------
accordance with the methods described in the Radioactive Waste
Solidification report. The burial trenches should be designed not to
intercept the ground water table and constructed with a bottom drain
and sump for water monitoring. The trenches should be covered with
either asphalt or vegetation to limit infiltration of water. The
burial site design, geology, and hydrology should be in conformance
with the criteria used in selecting and licensing the present commericial
1423
burial sites. Since land burial sites meeting the criteria are opera-
ting on a commercial scale currently, they are considered the most satis-
factory method of disposing of dilute concentrations.
Option No.2 - Near-Surface Liquid Storage. Near-surface storage
(using carbon steel and stainless steel tanks encased in concrete and
buried underground) of high-level, reactor-produced aqueous solutions of
cesium and other fission product salts is not considered as a satisfac-
tory means of storage. Aqueous solutions of high-level wastes from
reactor fuelds have been stored in this manner over the past 25 years.
The tanks range in size from 0.33 to 1.3 million gal. and are equipped
with devices for measuring temperatures, liquid levels, and leaks. These
tanks are considered as an interim storage technique due to a general
lack of confidence in their long-term integrity. Therefore, the
near-surface storage of aqueous solutions of cesium and other fission
product salts in steel tanks should only be considered as a near-term
storage technique and not as a permanent storage or disposal technique.
Option No.3 - Near-Surface Solid Storage. The storage of high-
level solidified cesium-134 and cesium-137 wastes and other waste salts
in engineered storage facilities offers the best method for intermediate
storage of these concentrated wastes. The advantages of this method are
that the wastes will be under surveillance and control and that they can
be retrieved, should this be required. The high-level wastes from fuel
processing should be solidified and packaged in a suitable container
(steel). Of the four high-level solidification processes developed,
spray or phosphate glass solidification processes offer better solidified
-------
waste characteristics than the pot calcination or fluidized' bed
i I
calcination processes (see Radioactive Waste Solidification, report).
The wastes will be stored in stainless steel-vaults encased in concrete
and will be either air or water cooled. ! i
Option No. 4 - Salt Deposits. This method offers the greatest
potential for the disposal of high-level wastes since bedded salt
deposits are completely free of circulating ground waters. This method
of disposal has been under study by the Oak Ridge National Laboratory
since 1957, and in November of 1970 a committee of the National Academy
of Sciences recommended that the use of bedded salt for the disposal
of radioactive wastes is satisfactory. ' Recent questions concerning
the adequacy of this method have resulted in the need for further
development work before it can be accepted as a method of ultimate
disposal. The cesium wastes must be solidified and disposed of in the
solid form, encapsulated in a suitable container. The solidified wastes
are buried in rooms carved in the salt deposits approximately 1,000 ft
below the ground. The salt is a good heat transmitter, provides about
the same radioactive shielding as concrete, and can heal its own
fractures by plastic flow. The critical problem is the selection of a
site that meets the necessary design and geological criteria.
Option No. 5 - Bedrock Disposal. The disposal of high-level,
aqueous solutions of cesium-134 and cesium-137 wastes, along with
other radioactive wastes in vaults excavated in crystalline rock over
•
1,500 ft beneath the ground is currently being evaluated by E. I. du
Pont de Nemours, at their Savannah River Plant near Aiken, South
Carolina. 0894,1396 jne wastes would be stored in six tunnels and
once in the tunnels the wastes will seep into the surrounding rock.
Located above the crystalline rock is the Tuscaloosa formation, a good
source of fresh water, which is separated by a layer of clay that would
act as a barrier to the leakage of radioactive wastes. An advisory
committee appointed by the National Academy of Sciences recommended
abandonment of the project. In May 1972, another National Academy of
Sciences panel concluded that bedrock storage provides a reasonable
33
-------
prospect for long-term safe storage, but precise information is needed
to decide if and where underground storage vaults should be built.
Another method has also been proposed for disposing of liquid wastes by
in situ incorporation in molten silicate rock. At the present time
both of these methods are unproven since sufficient engineering data or
exploration has not been completed to verify their suitability. Due
to the long half-lives of certain of the elements in the mixed fission
products and actinides, it is doubtful that data could ever be accumulated
to prove that the geological characteristics of the site over the next
few hundred years are acceptable to ensure the absolute safety of such
a disposal method.
Option No. 6 - Hydraulic Fracturing. The direct disposal of aqueous,
low-level radioactive wastes into shale formations has been investigated
by Oak Ridge National Laboratory. The method consists of mixing
the aqueous wastes with cement and pumping the resulting slurry down a
well out into a nearly horizontal fracture in a thick shale formation.
Additional work is required to demonstrate that this method of disposal
is satisfactory.
To summarize, cesium-134 and cesium-137 can be recovered from the
high-level waste streams for separate disposal or reuse. The acceptable
method of treatment is solidification, followed by storage in a near-
surface engineered storage facility and ultimate disposal in a salt
deposit. Cesium should be recovered from the low-level waste streams
to minimize the amount directly released to the environment. The
recovered cesium should be solidified, preferably using asphalt, and
disposed of at approved sites by land burial.
6. APPLICABILITY TO NATIONAL DISPOSAL SITES
Cesium-134 and cesium-137 are candidates for a National Disposal
Site due to their public health hazard and projected growth with that
of the civilian nuclear power program. From a cost and safety view-
point, it would be desirable to combine the reprocessing plant and the
34
-------
National Disposal Site. Interim storage in near-surface engineered
facilities offers considerable latitude in the site selection. The
site selection for ultimate disposal of these wastes is limited to
particular geological areas in which salt deposits are present.
Temporary, or interim, storage prior to transfer to a National Disposal
Site could be accomplished at many site locations.
The recommended treatment for low-level waste streams is
recovery followed by solidification with asphalt and disposed by land
burial.
35
-------
7. REFERENCES
0553. Etherington, H., ed. Nuclear engineering handbook, New York,
McGraw-Hill Book Company, Inc., 1960. 1,850 p.
0559. AEC Manual, Chapter 0524. Standards for radiation protection
May 12, 1964. 11 p.
0563. Recommendations of the International Commission on Radiological
Protection, ICRP Publication 2. Report on Committee II on
permissible dose for internal radiation (1959). Peryamon
Press, 1959. 232 p.
0703. King, L. J., A. Shimozato, and J. M. Holmes. Pilot plant
studies of the decontamination of low-level process waste
by a scavenging-precipitation foam separation process.
Oak Ridge National Laboratory, ORNL-3803, 1968. 57 p.
0705. Staff of the Oak Ridge National Laboratory. Siting of fuel
reprocessing plants and waste management facilities.
ORNL-4451, 1970. 500 p.
0707. Yee, W. C., F. DeLora, and W. E. Shuckley. Low-level radioactive
waste treatment: the water recycle process. Oak Ridge
National Laboratory, ORNL-4472, 1970. 30 p.
0709. Struxness, R.C. and et. al. Engineering development of hydraulic
fracturing as a method for permanent disposal of radioactive
wastes. Oak Ridge National Laboratory, ORNL-4259, 1968. 261 p.
0714. Touhill, C. J., B. W. Mercer, and A. J. Shuckrow. Treatment of
waste solidification condensates. Battelle Northwest,
B"NWL-723, 1968. 105 p.
0715. Schneider, K. J. Status of technology in the United States for
solidification of highly radioactive liquid wastes. Battelle
Northwest, BNWL-820, 1968. 63 p.
0733. Report by the Committee on Radioactive Waste Management. Disposal
of solid radioactive wastes in bedded salt deposits. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p.
0738. Regan, W. H., ed. Proceedings: the solidification and long-term
storage of highly radioactive wastes. Richland, Washington,
Feb. 14-18, 1966. Atomic Energy Commission, CONF-660208. 877 p,
36
-------
REFERENCES (CONTINUED)
0766. Sax, N. I. Dangerous properties of industrial materials. 3d ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
0894. Girdler, R. M. Storage of liquid radioactive wastes at the
Savannah River plant. Du Pont de Nemours and Company,
Aiken, South Carolina, DPSPU-69-30-9, July 1969. 11 p.
1396. Cooke, J. B. and et. al. Report on the proposal of E. I. Du Pont
de Nemours and Company for the permanent storage of radioactive
separation process wastes in bedrock on the Savannah River,
Du Pont de Nemours (E. I.) and Company, Aiken, South Carolina,
DPST-69-444, May 1970. 57 p.
1423. Morton, R. J. Land burial of solid radioactive wastes: study of
commercial operations and facilities. Environmental and Sanitary
Engineering Branch, Atomic Engergy Commission, WASH-1143, 1968. 132 p,
2146. Cohen, J. J., A. E. Lewis, and R. L. Braun. In situ incorporation
of nuclear waste in deep molten silicate rock. Nuclear Technology,
13: 76-87, Apr., 1972.
2147. Harris, D. and J. Epstein. Properties of selected radioisotopes.
Goddard Space Flight Center, Greenbelt, Maryland, NASA SP-7031,
1968. 89 p.
2148. Atomic Energy Commission. Forecast of growth of nuclear power,
Jan. 1971. Washington, U. S. Government Printing Office,
WASH-1139, 1971. 187 p.
2149. Code of Federal Regulations. Title 10—atomic energy. (Revised
as of Jan. 1, 1971). Washington, U. S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49--transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington, U. S. Government
Printing Office, 1971. 952 p.
2151. Atomic Energy Commission. The nuclear industry, 1971. Washington,
U. S. Government Printing Office, WASH-1174-71, 1971. 193 p.
37
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Cesium-134 Structural Formula
Half-life 2.05 years
Cs134
55LS
Type of Decay Negative Beta
Molecular Wt. 134 Melting Pt. 28.5C Boiling Pt. 690C
Density 1.88 gm/cc Specific Power 20.4 watts/gm
Solubility Specific Activity 1.220 curies/gin
Cold Mater Reacts Hot Water Reacts v
Others: Soluble in liquid NH-
Decay Chain Radiation Energy Level & Intensities
Beta: 0.662 Mev (71%)
Cs 134 B" Ba134 °'410 Mev ( 1%)
(2.05y) >(stable) 0.089 Mev (28%)
Gamma: 1.365 Mev (3.4%)
1.168 Mev (1.9%)
0.769 Mev (99%)
0.605 Mev (98%)
0.570 Mev (23%)
Shipping Regulations Classified as a transport group III radionuclide
by the Department of Transportation
Comments
References: (1) 0766
(2) 2147
(3) 2150
38
-------
HAZARDOUS WASTES PROPERTIES
. WORKSHEET
Cesium-137
Name
Half-life 30 years
Type of Decay Negative Beta
Molecular Wt. 137
Density 1.88 gm/cc
Structural Formula
55
Cs
137
Melting Pt. 28.5 C
Specific Power
Boiling Pt. 690 C
0.42 watts/gm
Solubility^
Cold Water
Others:
Reacts
Specific Activity
Hot Water Reacts
87 curies/gm
Soluble in liquid NH$
Decay Chain
137m
(2.6m)
Ba
(stable)
Radiation Energy Level & Intensities
Cesium-137
Beta: 1.176 Mev ( 6%)
0.514 Mev (94%)
Barium-137
Gamma: 0.662 Mev (89%)
IT = isometric transition
Shipping Regulations Classified as a transport group III radionuclide
by the Department of Transportation
Comments
References: (1) 0766
(2) 2147
(3) 2150
39
-------
PROFILE REPORT
Hydrogen-3
1. GENERAL
Introduction
Tritium, an isotope of hydrogen,is produced in nuclear reactors in
substantial quantities. It is currently released to the environment in
the form of tritiated coolant water from the reactor and tritium wastes
resulting from processing of the spent nuclear fuels. Although it is one
of the least hazardous radioactive nuclides, the recovery and retention
of tritium may be required in the future due to its long half-life for
radioactive decay, the rapid rate of expansion in the nuclear power and
fuel-reprocessing industries, and its ability to be metabolized in the
form of tritiated water and incorporated into body fluids and tissues.
Tritium is produced in the fission of uranium-235 and plutonium-239
with yields of 0.01 percent and 0.02 percent, respectively. Tritium
decays by the emission of a beta particle and an anti-neutrino to form
stable helium-3. The half-life for this process is approximately 12.3
years. The average energy of the beta emitted by tritium is 5.6 Kev,
which is only about one one-hundredth of the energy emitted by most other
beta emitters.
The projected growth in production of this isotope will parallel that
of the civilian nuclear power program. Present projections indicate that
nuclear power will account for 30 percent of the total power production
by the year 1980- The expected annual production and accumulated
wastes of hydrogen-3 from the civilian nuclear power program are
included (Table 1).
41
-------
fO
TABLE 1
HYDROGEN-3 GENERATED BY THE CIVILIAN NUCLEAR POWER PROGRAM0705
Light-Hater Reactor Fuels
Annual
Production
grams/year
curies/year
watts/year
Annual
Production
grams/year
curies/year
watts/year
Calendar Year
1970 1980 1985
3.74 213 384
. 036x1 O6 2. 06x1 O6 3. 72x1 O6
1.29 73.4 132
Fast-Breeder Reactor
Calendar Year
1970 1980 1985
34.2
.33xl06
11.8
1990
434
4.21xl06
149
Fuels
1990
207
2.01xl06
71.1
2000
344
3. 34x1 O6
119
2000
881
8.55xl06
304
2020
1010
9. 80x1 O6
348
2020
2650
26.8xl06
914
-------
Uses
Tritium is used in luminous devices, tagging, and fusion research.
Sources and Types of Wastes
Tritium may be produced in nuclear reactors by one or more of the
following five methods: f'stoning of uranium, neutron capture reactions
with b-^on and lithium added to the reactor coolant, neutron capture
reactions with boron in conv.ro! rods, activation of deuterium -In water- and
2373
high energy neutron captur? "-eactions with structural materials.
Fission-product tritium is generally contained by the cladding
surrounding each fuel ilenvs'.i. While small amounts of tr1t^um we^e fou"d
in the past to leak to the primary coolant with the stainless-steel clad
elements, the use of zircor.'um cladding has been found to limit the
diffusion of tritium througr. the cladding to very low quantities. The
fission-product tritium generated is contained until the spent fuel is
processed, when it appears as tritiated water in the fuel processing.piant
evaporator condensates. Additional sources of tritium are in the primary
coolant loops of pressurized water reactors and high temperature gas-cooled
reactors. In the pressurized water reactors, boric acid which is dissolved in
the primary coolant for reactivity control causes the production of tritium
by neutron capture reactions. The production of tritium occurs in high
temperature gas cooled reactors by ternary fission and by activation of
helium-3 found in trace amounts in the helium coolant. Although the
primary coolant from these reactors is not routinely discharged to the
waste disposal system, leakage of coolant from pumps and the water utilized
during the refueling operation are discharged to the waste disposal
systems. In the reactor power plant waste disposal systems the water
undergoes several treatment processes before being released to the environment,
but none of these processes is effective in removing tritium. Tritium may
compose between 50 and almost 100 percent of the total amount of radioactive
material discharged as liquid waste from nuclear reactors. The amount of
tritium discharged from reactors in a gaseous form is only about 1 percent
of the total tritium discharge.
43
-------
Physical and Chemical Properties
The physical and chemical properties of tritium are included in the
attached worksheet. It is found as a colorless gas and in the form of
tritiated water. It is undetectable by conventional methods of gross
radioactivity analysis. Special analytical techniques such as liquid
scintillation counting must be used to measure tritium.
2. RADIATION HAZARD
Tritium is a moderately hazardous radioactive material. Because it
is an isotope of hydrogen, it can be metabolized in the form of tritiated
water and incorporated into body fluids and tissues.
The effects of radiation exposure are primarily dependent on the amount
of radiation and the portion of the body affected. A single ingestion of
tritiated water having an activity of 1 microcurie will produce a total
dose to the body tissues of 0.21 rem. Continuous ingestion of water having
a specific activity of 1 microcurie of tritium per mi Hi liter will produce
a dose rate of 170 rem per year to body tissues.
Standards for prolonged exposure over a 50-year period have
defined the single dose limit in terms of the maximum permissible dose
accumulated in a period of 13 weeks. The whole body exposure limit is
3 rem per quarter for a radiation worker and the accumulated dose limit is
5(N-18;, where N is the individual's age in years.
3. OTHER HAZARDS
Tritium has negligible toxicity other than the radiation hazard.
Tritium is flammable in air and mixtures with air are explosive.
-------
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling. Storage, and Transportation
Since tritium is moderately hazardous to man by inhalation or
ingestion, care must be exercised in its handling. Handling must be set
up to prevent excessive exposure to personnel and, in fact, to prevent
any unnecessary exposure.
These materials should be stored in controlled areas in specially
constructed containers. Tritium, as a gas, as luminous paint, or
adsorbed on solid material is classified as a group VII radionuclide by
the Department of Transportation. The rules and regulations governing
its transportation are given in the Code of Federal Regulations (CFR)
Title 49--Transportation, Parts 170 to 190. ° The transport group
VII radionuclide content is limited to 1,000 curies for a Type A package
and 50,000 curies for a Type B package defined in 49CFR173. Their
release rate is limited to zero under the specified accident conditions
for Type A and B quantities.
Disposal/Reuse
The disposal of these materials is governed by the AEC Manual
0559 2149
Chapter 0524 and 10CFR20. Two sets of standards have been
established for the permissible radiation exposure in unrestricted areas.
One is for the greatest dose received by an individual and the other for
the average dose received by the general population. The standards for the
safe release of this material to the environment in an unrestricted area
2149
are contained in 10CFR20 and its release should not exceed the
concentrations in air and water listed in this report (Table 2). These
release rates apply to an individual and its release may be further
limited if a suitable sample of the population is exposed to one-third
the concentrations in air or water specified in this report (Table 2).
45
-------
TABLE 2
2149
TRITIUM ENVIRONMENT RELEASE RATES
Fission Product Form
Hydrogen-3 Soluble
Insoluble
Submersion*
Concentration in Air
(microcuries/milliliter)
2xlO"7
2xlO"7
4x1 O"5
Concentration
(microcuries/mi
3x1 O"3
3xlO"3
* — —
in Water
lliliter)
05
*Submersion means that the values given are for submersion in a semispherical infinite cloud
of airborne material.
-------
Although rarely practiced, the disposal by release in a sanitary seu/age
system is limited to 10,000 microcunas. The disposal by burial at any
one location and time is limited to "00 times the above amount.
5, EVALUATION OF V/AST6 MANAGEMENT PRACTICES
Recovery
Tritium is not currently recovered at the reactor site or t.hvi spent
fuel px'ocessinn plant site. However, the Oak Ridge National Laboratory is
conducting exoerimental work on recovering tritium during fuel reprocessing
for the fast breeder reactor. Voloxidation is the term for the tritium
recovery process being investigated in which the fuel, after being chopped
in the read-end process, is heated tr drive off the tritiated hydrogen
By collecting and condensing the effluent, greater than 99 percent of the
tritium is recovered. It is feasible that such a process could be
incorporated into the reprocessing of fuel from the present, light water
reactors.
Storage/Disposal
The current method of tritium disposal is dilution and dispersion,
both at the reactor site and the spent fuel processing plant site. Most of
this release is in the form of tritiated water in the liquid waste
streams and only a very small percent is released to the air. Tritium
discharge concentrations are usually much less than 1 percent of the
discharge limits. Because of the inherently small radiological hazard
from tritium, the difficulty of removing tritium from the reactor's
tritiated water, and the small percentage of the allowable concentrations
that are currently being released to the environment at the reactor site,
tritium removal is probably not practical in the near future. At the
spent fuel processing plant, however, a process similar to the voloxidation
process should be incorporated in order to recover tritium and dispose
of it by one of the following methods.
47
-------
Option No. 1 - Land Burial. Land burial of tritium wasted, in small
concentrations, at approved sites that are acceptable from a geologic
and hydrologic standpoint, is an acceptable means of disposal. Its
concentrations should not be in excess of 10 times its maximum permissible
2149
concentrations for the general population in 10CFR20. The burial
trenches should be designed not to intercept the ground water table and
constructed with a bottom drain and sump for water monitoring. The
trenches should be covered with either asphalt or vegetation to limit
infiltration of water. The burial site design, geology, and hydrology
should be in conformance with the criteria used in selecting and licensing
1423
the present commercial burial sites. Since land burial has
successfully been practiced on a commercial scale, it should be considered
as the most satisfactory method of disposing of low concentrations of
these wastes.
Option No. 2 - Near Surface Storage. The storage of high concentrations
of these wastes in engineered surface facilities offers the best intermediate
method for storage of these wastes. The technology for these facilities
has been developed and the wastes will be under surveillance and control
and can be retrieved, should this be required. The wastes will be stored
ir, stainless-steel lined concrete vaults which will be either air or water
cooled.
Option No. 3 - Salt Deposits. This method offers the best
potential for the disposal of this radionuclide. This method of disposal
has been under study by the Oak Ridge National Laboratory since 1957,
and in November 1970 a committee of the National Academy of Sciences
recommended that the use of bedded salt for the disposal of radioactive
wastes is satisfactory. Recent questions concerning the adequacy
of this method have resulted in the need for further development work
before it can be accepted as an ultimate method of disposal. The wastes
are buried in rooms carved in the salt deposits approximately 1,000 ft
below the ground. The salt is a good heat transmitter, provides about the
48
-------
same radioactive shielding as concrete, and can heal its own fractures by
plastic flow. The critical problem is the selection of a site that meets
the necessary design and geological criteria.
To summarize, the majority of the tritium generated by the nuclear
power reactors is currently released to the environment in the form of
liquid wastes. Most of this release occurs at the fuel reprocessing plant,
where the tritium could be recovered by voloxidation or a similar process.
The tritium should be encapsulated and disposed of in salt deposits
following storage in engineered storage facilities.
6. APPLICABILITY TO NATIONAL DISPOSAL SITES
Tritium is a candidate for a National Disposal Site due to its
projected growth with that of the civilian nuclear power program. It
should be recovered and encapsulated at the fuel reprocessing plant and
shipped to the National Disposal Site for disposal in salt beds following
storage in engineered storage facilities.
49
-------
7. REFERENCES
0559. AEC Manual, Chapter 0524. Standards for radiation protection
May 12, 1964. 11 p.
0705. Staff of the Oak Ridge National Laboratory. Siting of fuel
reprocessing plants and waste management facilities.
ORNL-4451, 1970. 500 p.
0766. Sax, N. I. Dangerous properties of industrial materials. 3d ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
2148. Atomic Energy Commission. Forecase of growth of nuclear power,
Jan. 1971. Washington, U. S. Government Printing Office,
WASH-1139, 1971. 187 p.
2149. Code of Federal Regulations. Title ID—atomic energy. (Revised
as of Jan. 1, 1971). Washington, U. S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49--transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington,
U. S. Government Printing Office, 1971. 952 p.
2378. Selected materials on environmental effects of producing electric
power. Report of the Joint Committee on Atomic Energy
Congress of the United States. 91st Cong., 1st sess.,
Aug. 1969. Washington, U. S. Government Printing Office. 553 p.
50
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Tritium Structural Formula
Half-life 12-26
Type of Decay Negative Beta
Molecular Wt. 6.05 Melting Pt. Boiling Pt.
Density Specific Power
Solubility Specific Activity 9700 curies/gm
Cold Water _____ Hot Water
Others:
Decay Chain Radiation Energy Level ft Intensities
Beta: 5.6 Kev (average)
18.6 Kev (maximum)
(12.26y) (stable)
Shipping Regulations Classified as a transport group VII radionuclide
by the Department of Transportation
Comments
References: (1) 0766
(2) 2150
(3) 2378
51
-------
PROFILE REPORT
Iodine-129, Iodine-131, Krypton-85, Xenon-133
1. GENERAL
Introduction
Iodine and the noble gases, krypton and xenon, are produced during
the fission of uranium in nuclear reactors. They represent a potential
source of environmental contamination since they are presently released
to the environment during reprocessing of the nuclear fuels. Although
the release rates at present reprocessing plants are low, the recovery
and retention of these gases may be required in the near future due to
the rapid rate of expansion of the nuclear power industry and the
corresponding expansion in the nuclear fuel reprocessing industry.
In addition to the anticipated growth in this industry, economic incentives
exist for reducing fuel decay times prior to reprocessing. Present
reprocessing plants handle 150-day decayed light water reactor fuel. With
the introduction of the fast-breeder reactors it is economically desirable
to reprocess the fuel after 30 days decay. The shortened decay period
will increase the reprocessing plant off-gas handling requirements for
the short-lived isotopes such as iodine-131 (8 days) and xenon-133 (5
days). The iodine-131 activity in 30-day-old reactor fuel is 36,000 times
higher than it is in 150-day-old reactor fuel.
Of the fission-product halogens only the isotopes iodine-129 and
iodine-13:l are physiologically significant after 30 days of post-irradiation
decay. The iodine-131 contents of reactor fuels are approximately 72,000
and 2 curies per metric ton of reactor fuel after decay times of 30 and
150 days respectively. The iodine-129 content is only about 0.03 curies
per metric ton of reactor fuel. Even though the iodine-129 content in
reactor wastes is low, it is significant since it has a half-life of
53
-------
17 million years compared to the iodine-131 half-life of 8 days. Iodine-129
and iodine-131 are also reconcentrated by biological processes in the food
chain leading to man. This reconcentration occurs in the grass-cow-milk
pathway to the thyroids of small children and man.
Krypton and xeiion are both produced in significant quantities in nuclear
reactors. These isotopes are chemically inert and once released they do not
concentrate in body tissues. The various isotopes of krypton and xenon
in reactor fuels are attached (Table 1). The only two isotopes of significance
are krypton-85 and xenon-133. Krypton-85 has a half-life of 10.8 years, a
moderate beta radiation energy, a low gamma radiation level, and a fission
yield of 1.3 percent. Xenon-133 has a half-life of 5.3 days, a moderate
beta and gamma radiation energy, and a fission yield of 6.6 percent.
The projected growth in production of these isotopes will parallel
that of the civilian nuclear power program. Present projections indicate
that nuclear power will account for 30 percent of the total power produc-
?148
tion by the year 1980. The expected annual production of iodine-129
(Table 2) and the expected annual production and accumulated wastes of
krypton-85 (Figure 1) from the civilian nuclear power program are
included. Production figures for iodine-131 and xenon-133 are not
included since due to their short half-life the amount present varies
with the post-irradiation decay time.
Uses
Krypton-85 is used in the preparation of self-illuminating materials
and devices. Its properties also make it useful in leak detection
equipment, thickness gauges, and in gas chromatography. The fission
product xenon, including xenon-133, which is essentially non-radioactive
by the time it is recovered, is used in scintillation devices because
it has a higher proportion of heavy isotopes of xenon than does naturally
occurring xenon. Xenon is also used in the manufacture of light bulbs that
have an extra long life. Iodine-131 is used in medical diagnosis and
therapy and also in agricultural research. Since iodine-131 tends to
54
-------
CA
TABLE 1
THERMAL-NEUTRON FISSION YIELD OF KRYPTON AND XENON FROM U-235 REACTOR FUELS
Fission Products
Krypton-83
Krynton-84
Krypton-85
Krypton-86
Xenon-131
Xenon-132
Xenon-133
Xenon-134
Xenon-135
Xenon-136
Fission Yield
Percent
0.544
1.00
1.293
2.02
2.93
4.38
6.62
8.06
6.30
6.46
Half-Life
Stable
Stable
10.76 years
Stable
Stable
Stable
5.3 days
Stable
9.2 hours
Stable
-------
cn
TABLE 2
IODINE-129 CONTENT IN WASTE PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM0705
Light-Water Reactor Fuels
Annual
Production
grams/year
curies/year
watts/year
Annual
Production
grams/year
curies/year
watts/year
Calendar Year
1970 1980 1985
.01 2x1 O6 ,69xl06 1.24xl06
1.97 112 202
.0085 .0485 .0874
Fast-Breeder Reactor
Calendar Year
1970 1980 1985
.116xl06
19.4
.0083
1990
1.4xl06
229
.099
Fuels
1990
.70xl06
117
.051
2000
l.llxlO6
182
.078
2000
3.0xl06
500
.215
2020
3. 26x1 O6
534
.231
2020
9.0xl06
1500
.649
-------
1970
1980
1990
2000
-2010
Figure 1. Estimated Production of Krypton-85 from the Nuclear Power
Industry.0705
57
-------
concentrate in the thyroid, it is used in the treatment of hyperthyroidism
and in treating congestive heart failure. Iodine-129,due to its long
half-life, has very limited commercial use.
Sources and Types of Wastes
Iodine, krypton, and xenon are produced in nuclear reactors and are
contained within the spent fuel elements. The spent fuel elements are
removed from the reactor and processed for the recovery of usable fissionable
materials. In the spent fuel processing plants the majority of these
radionuclides are released as gases during the dissolution of the fuel
elements in the head-end process. They are also released in the venting
of process vessels and in the treatment of liquid wastes. Iodine-129 and
iodine-131 are released in the off-gas streams in the form of elemental
iodine (I^K hydrogen iodide, or as organic iodide compounds such as methyl
iodide. Krypton and xenon are released in their elemental form and do
not form compounds with other materials since they are chemically inert.
Physical and Chemical Properties
The physical and chemical properties of iodine, krypton, and xenon
are included in the attached worksheets. Iodine is a member of the halogen
group and is a solid at room temperature. Iodine is easily melted and
vaporized and is a very powerful oxidizing agent. Iodine forms compounds
with many elements, but is less active than the other halogens which
displace it from iodide compounds. Krypton and xenon are members of the
rare or inert gas group. These gases are characterized generally by their
zero valence and are considered very inert. Some relatively unstable
compounds of these elements are known, however.
2. RADIATION HAZARD
Iodine, xenon, and krypton are moderately hazardous radioactive
nuclides. Krypton and xenon are chemically inert and once released they
do not concentrate in body tissues. Iodine does reconcentrate by the
grass-cow-milk pathway to the thyroid.
58
-------
The effects of their radiation exposure are primarily dependent on
the amount of radiation and the portion of the body affected. The effects
of whole-body gamma radiation exposure are: (1) 5 to 25 rads, minimal dose
detectable by chromosome analysis or other specialized analyses, but not
by hemogram; (2) 50 to 75 rads, minimal acute dose readily detectable in
a specific individual (e.g., one who presents himself as a possible
exposure case); (3) 75 to 125 rads, minimal acute dose likely to produce
vomiting in about 10 percent of people so exposed; (4) 150 to 200 rads,
acute dose likely to produce transient disability and clear hematological
changes in a majority of people so exposed; (5) 300 rads, median lethal
pccc
dose for single short exposure. The effects are for a single large
dose of radiation or a series of substantial doses in a short interval of
time to the total body.. The dose delivered to a particular body organ
following the inhalation of 1 microcurie of each of these radionuclides
is attached (Table 3). For iodine-131 a dose of 25 rem is delivered to
the thyroid following the inhalation of 15.6 microcuries (0.002 micrograms).
Standards for prolonged exposure over a 50-year period have defined
the single dose limit in terms of the maximum permissible dose accumulated
in a period of 13 weeks. The whole body exposure limit is 3 rem per
quarter for a radiation worker and the accumulated dose limit is 5(N - 18),
where N is the individual's age in years. Limits for the thyroid, bone, and
other organs have also been defined. Values of the total body burden
for each radionuclide required to produce the maximum permissible dose
rates defined above have been compiled. For iodine the critical
organ is the thyroid and the maximum permissible body burden is 2 microcuries
for iodine-131 and 9 microcuries for iodine-129.
3. OTHER HAZARDS
Besides its radiation hazard iodine is moderately toxic whereas
krypton and xenon are only slightly toxic. Iodine contact with the skin
can cause lesions and iodine vapor is intensely irritating to the eyes
and mucous membranes. The recommended maximum allowable concentration
of iodine in the air is 1 mg/cu. meter.
59
-------
TABLE 3
IODINE, KRYPTON, AND XENON DOSE TO A PARTICULAR BODY ORGAN
FOLLOWING INHALATION OF ONE MICROCURIE OF THE NUCLIDE
Isotone
Form
Organ
Dose
rem
Iodine-129
Iodine-129
Iodine-131
Iodine-131
Krypton-85
Xenon-133
Insoluble
Soluble
Insolbule
Soluble
Insoluble
Insoluble
Lung
Thyroid
Lung
Thyroid
Lung
Lunn
0.09
7.0
0.03
1.6
0.26
0.02
60
-------
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling, Storage, and Transportation
Since these radionuclides are hazardous to man by inhalation,
ingestion, or direct radiation exposure, care is exercised in their
handling. Handling is set up to prevent excessive exposure to personnel.
Special procedures and radiation shielding are utilized in their
handling. The beta particles emitted by iodine, xenon, and krypton are
of moderate energy and cannot penetrate more than 0.1 inches of water.
Their gamma radiation is highly penetrating and high-density shields, such
as lead, are required to stop the radiation. To detect and control
personnel exposure to their radiation all persons working with this
material should wear dosimetry devices which directly indicate the dose.
Commonly used devices are the film badge and the thermoluminescent
dosimeters (TLD).
These materials are stored in controlled areas in specially
constructed containers. They are protected by both a primary and a
secondary containment barrier. Special monitoring systems and proper
warning signs should be located in the general area of the storage
facility.
Iodine-129 and iodine-131 are classified as a transport group III
radionuclide and krypton-85 and xenon-133 are classified as a group IV
radionuclide in the uncompressed state, at a pressure less than 14.7 psia,
and as a group III radionuclide at pressures greater than 14.7 psia by
the Department of Transportation. The rules and regulations governing
their transportation are given in the Code of Federal Regulations (CFR)
Title 49--Transportation, Parts 170 to 190. The transport group
III radionuclide content is limited to 3 curies for a Type A package
and 200 curies for a Type B package defined in 49CFR173. Their release
rate is limited to zero under the specified accident conditions for
Type A and B quantities.
61
-------
Disposal/Reuse
The disposal of these materials is governed by the AEC Manual
Chapter 05240559 and 10CFR20.2149 Two sets of standards have been
established for the permissible radiation exposure in unrestricted
areas. One is for the greatest dose received by an individual and the
other for the average dose received by the general population. The
standards for the safe release of these materials to the environment
2149
in an unrestricted area are contained in 10CFR20 and their release
should not cause the concentrations in air and water to exceed the concen-
trations listed in this report (Table 4). These concentrations apply to
an individual and their release may be further limited if a suitable
sample of the population is exposed to one-third the concentrations in
air or water specified in this report (Table 4).
Although rarely practiced, the disposal by release in a sanitary
sewage system is limited to 1 microcurie for iodine-129, 10 microcuries
for iodine-131 and 1,000 microcuries for krypton-85 and xenon-133.
The disposal by burial at any one location and time is limited to 100
times the above amounts.
5. EVALUATION OF WASTE MANAGEMENT PRACTICES
Iodine Recovery
Option No.1 -£austic Scrubbers. Iodine has been removed from gas
streams in many applications by scrubbing with caustic solutions and
reacting with silver nitrate impregnated on ceramic packing. Caustic
scrubbers are effective in removing approximately 90 percent of the iodine
in the off-gas streams and the silver' nitrate towers remove about 99 percent
of the remaining iodine. In general, caustic scrubbers are not effective
in removing organic iodides, such as methyl iodide. Organic iodides can
be effectively removed by using an aqueous scrubbing medium of mercuric
nitrate in nitric acid if sufficient contact time is allowed for the
relatively slow reaction between the absorbed organic iodide and mercury
ion.
62
-------
TABLE 4
IODINE, XRYPTON, AND XENON NATIONAL PERMISSIBLE CONCENTRATIONS
2149
CJ
Radionuclide
Iodine-129
Iodine-129
Iodine-131
Iodine-131
Krypton-85
Xenon*-133
Form
Soluble
Insoluble
Soluble
Insoluble
Submersion*
Submersion*
Concentration in Air
(microcuries/mi 1 1 i 1 i ter )
2X10"11
2xlO"9
IxlO'10
IxlO'8
3x1 O"7
3x1 0"7
Concentration in
(microcuries/mi 11
6x1 O"8
2x1 O"4
3x1 O"7
6x1 O"5
—
---
Water
i liter)
*Submersion means that the values given are for submersion in a semispherical infinite cloud of
airborne material.
-------
Option No.2-Activated Charcoal. Charcoal has been extensively
used for the removal of elemental iodine from off-gas streams. Elemental
Iodine is readily trapped from air streams by activated charcoal, even
at relative humidities asprcacMng saturation. Iodine absorption
1395
efficiencies of 99.99 percent have been obtained. However, organic
iodides, such as methyl iodide are not effectively absorbed. To remove
organic iodides the act-'vated-charcoal beds should be preceded by a
catalytic oxidation step to completely oxidize all organic vapors in
the gas stream. The addition of the oxidation step prior to the charcoal
bed is effective for removing organic iodides, but the bed life may be
limited to a few months or ? residence time of a few seconds may be
required which is not practical in a large gas flow system. The charcoal
beds are quite effective fov the removal of trace quantities of elemental
iodine and their use should probably be limited to the final iodine
removal step.
Option No.3-Impregnated Charcoal. Special impregnated charcoals
have been developed for removing methyl iodide. The Impregnated charcoals
presently under consideration are: (1) iodized with one or more
iodine-containing substances, to provide 1-131, 1-127 exchange capability;
or (2) triethylinediamine - impregnated to effect removal by the
utilization of the alkyl halide-organic amlne reaction. Methyl iodine
retention efficiencies greater than 99 percent have been obtained with
both these impregnated charcoals. The removal efficiencies for the
iodine-impregnated charcoal refer only to 1-131. The performance of
these impregnated charcoal beds is drastically reduced by poisoning due
1 ^Rfi
to impurities in the air stream. Iodine-Impregnated charcoal and
triethylinediamine-impregnated charcoal are satisfactory absorbents
for organic iodides, such as methyl iodide, but an upstream filter
system may be required to '•emove impurities in the air.
64
-------
Option No.4-$i"lver Zeolite. Inorganic and organic forms of iodine
are both effectively removed from the off-gas stream by silver zeolite.
The silver zeolite has ideal properties of nonflammability, capability of
iodine absorption up to 500 C, retention of absorbed iodine up to
1,000 C, and rapid reaction times. The zeolite is a synthetic zeolite
absorber (Linde Molecular Sieve 13X) in which the normal sodium zeolite
matrix has been converted to silver zeolite. The silver zeolite retention
efficiency of elemental iodine and methyl iodide is greater than 99.8
2291
percent at temperatures varying from room temperature to 500 C.
The efficiency of the silver zeolite is adversely effected by long-term
exposure to organic contaminants in the air stream. The zeolite life
expectancy can be extended by preceding the zeolite with an oxidation
step. This is economically desirable since the basic cost of silver
zeolite is 3 to 10 times the cost of charcoal. Silver zeolite is the
most efficient absorber of all airborne iodine species and is particularly
applicable to the large gas flows encountered in nuclear fuel
reprocessing plants,
Krypton and Xenon Recovery
Option No.1 Activated Charcoal. Krypton and xenon can be retained by
adsorption beds of activated charcoal. In this process the krypton and
xenon are cooled by means of cold trays and are then passed through a constant
temperature (25 C) charcoal column. These absorption beds are used to
provide holdup time which is long compared to the half-life of the isotopes.
This process is effective for the 5.3 day xenon-133 but not for 10.8 year
krypton-85. For xenon-133 the gas could be held up in a charcoal bed
for a period of 18 days to reduce its activity 10 times or a period of
54 days for a reduction of a 1,000 times.
Option Nq.2-Cryogenic Distillation. A cryogenic distillation process
for the recovery of krypton and xenon from off-gas streams generated during
the processing of spent fuels has been applied at the Idaho
2292
Chemical Processing Plant. In this process the off-gas stream is
first treated prior to entering the cryogenic system by passing the
gas through a catalytic conversion system which removes nitrous oxide and
65
-------
hydrogen. In an off-gas stream with a high hydrogen content a special
catalytic unit and condenser would also be required to reduce the hydrogen
concentration in the off-gas to levels below the explosive limit. The
off-gas is then fed to the cryogenic system which is cooled by liquid
nitrogen. This system consists primarily of two regenerators, a
distillation unit, and a batch still. The gas stream is cooled to
-260 F in the two regenerators prior to entering the distillation column.
In the distillation column, krypton and xenon are condensed and absorbed
by liquid nitrogen which flows through the plates in the column. The
krypton and xenon are dissolved in the liquid nitrogen and are transferred
to a batch still where they are separated and purified in a fractionation
step. This process has been applied in actual plant operation at the
Idaho Chemical Processing Plant and is capable of processing up to
20 scfm of off-gas. The process has the potential for recovering
99 percent of the gases with 4 to 10 percent of nitrogen and oxygen
impurities in the final product. Actual overall krypton-xenon recovery
efficiencies have varied from 30 to 60 percent. The low recovery
efficiencies were due to difficulties in operation of the gas cleanup
step and to leaks within the product bottling system.
Option No.3-Fluorocarbon Solvents. Krypton and xenon can be
effectively removed from contaminated air streams by selective absorption
in fluorocarbon solvents, such as refrigerant-!2 (dichlorodifluoromethane).
This process has been tested on a pilot plant scale at processing rates
up to 20 scfm of gas by Union Carbide at their Oak Ridge, Tennessee,
1397
facility. In this process, the removal of krypton and xenon is
accomplished by contacting the gas stream with a stream of fluorocarbon
solvent (refrigerant-11 or -12) in a packed absorber column at low
temperatures (-177 to -21 F) and high pressure (169 to 437 psia). Besides
krypton and xenon substantial amounts of oxygen, nitrogen, and argon are
dissolved in the solvent. Since krypton and xenon are substantially more
soluble in the solvent than these other gases, the absorber column is
followed by a fractionator system where most of the dissolved nitrogen,
argon, and oxygen is driven off. The solvent is then driven to a stripper
system where the product gas concentrated in krypton and xenon is evolved
66
-------
and the solvent is recycled back to the absorber. Based on the pilot
plant test data, a plant operating with an absorber at -30 F and
425 psia should give a krypton and xenon removal efficiency of over
1397
99.7 percent with refrigerant-12 as the solvent. For application to
fuel reprocessing plants with feed rates in excess of 100 scfm some
scale-up of the pilot plant facility is required. This method offers
the best potential for the effective removal of krypton and xenon from
off-gas streams. It has been successfully operated on a pilot-plant
scale but has not yet been applied in actual plant operations.
Storage/Disposal
Option No.l-Land Burial. Land burial of iodine, krypton, and xenon
wastes, in small concentrations, at approved sites that are acceptable
from a geologic and hydrologic standpoint, is an acceptable means of
disposal. Their concentrations should not be in excess of 10 times
their maximum permissible concentrations for the general population in
2149
10CFR20. The burial trenches should be designed not to intercept
the ground water table and constructed with a bottom drain and sump for
water monitoring. The trenches should be covered with either asphalt or
vegetation to limit infiltration of water. The burial site design,
geology, and hydrology should be in conformance with the criteria used
1423
in selecting and licensing the present commercial burial sites.
Since land burial has successfully been practiced on a commercial scale,
it should be considered as the most satisfactory method of disposing
of low concentrations of these wastes.
Option No.2-Near-Surface Storage. The storage of high concentrations
of these wastes in engineered storage facilities offers the best immediate
method for storage of these wastes. The technology for these facilities
has been developed. The wastes will be under surveillance and control
and can be retrieved, should this be required. The wastes will be stored
in stainless-steel lined concrete vaults which will be either air or water
cooled.
67
-------
Option No.3-Sa1t Deposits. This method offers the best potential for
the disposal of this radionucl ide. This method of disposal has been undsr
study by the Oak Ridge National Laboratory since 1957, and in November DT
1970 a committee of the National Academy of Sciences recommended that the
use of bedded salt for the disposal of radioactive wastes is satisfactory.
Recent questions concerning tne adequacy of this method have resulted in
the need for further development work before it can be accepted as an
utlimate method of disposal. The wastes are buried in rooms carved in the
salt deposits approximately 1,000 ft below the ground. The salt is a good
heat transmitter, provides about the same radioactive shielding as concrete
and can heal its own fractures by plastic flow. The critical problem i«.
the selection of a site that meets the necessary design and geological
criteria.
To summarize, krypton, xenon, and iodine are released in the off -gas
streams during processing of the spent fuels. The iodine can be removed
from the off -gas stream by silver zeolite. Krypton and xenon can be
effectively removed by cryogenic distillation or by selective absorption
in fluorcarbon solvents. Following recovery the wastes can be held to
allow the short-lived isotopes such as iodine-131 and xenon-133 to decay.
cor iodine-131 and xenon-133 their activity is reduced by a factor of
10,000 in 106 days and 71 days, respectively. The long-lived isotopes
krypton-85 (10.8 years) and iodine-129 (17 million years) should be
encapsulated and disposed of in salt deposits following storage in near-
surface engineered facilities. The wastes to be disposed of can be
encapsulated in high-pressure cylinders or solidified. A method for
solidifying the gases xenon and krypton includes dispersion in glasses
or resins and entrapment in molecular sieves or small pressurized steel
bulbs which are in turn encased in epoxy resin.
6. APPLICABILITY TO NATIONAL DISPOSAL SITES
Iodine-129 and krypton-85 are candidates for a National Disposal Site
due to their long half-lives and projected growth with that of the civilian
nuclear power program. Iodine-131 and xenon-133 are not candidates for a
68
-------
National Disposal Site because of their short half-lives.
The recommended treatment for the recovery of iodine-129 and iodine-131
from reprocessing plant off-gas streams is by the use of silver zeolite.
Krypton-85 and zenon-133 can be removed by cryogenic distillation or
absorption in fluorocarbon solvents. Iodine-131 and xenon-133 can be
stored in engineered storage facilities to allow their radioactive decay
and then disposed of in low concentrations in land burial facilities.
Iodine-129 and xenon-133 should be stored in engineered storage facilities
followed by disposal in salt beds.
-------
7, REFERENCES
0559. AEC Manual, Chapter 0524. Standards for radiation protection
May 12, 1964. 11 p.
0563. Recommendations of the International Commission on Radiological
Protection, ICRP Publication 2. Report on Committee II on
permissible dose for internal radiation (1959). Peryamon
Press, 1959. 232 p.
0669. Adams, R. E., R. D. Ackley, and W. E. Browning. Removal of
radioactive methyl iodine from steam-air systems. Oak Ridge
National Laboratory, ORNL-4040, 1967. 26 p.
0705. Staff of the Oak Ridge National Laboratory. Siting of fuel
reprocessing plants and waste management facilities.
ORNL-4451, 1970. 500 p.
0733. Report by the Committee on Radioactive Waste Management. Disposal
of solid radioactive wastes in bedded salt deposits. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p..
0766. Sax, N. I. Dangerous properties of industrial materials. 3d ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
1386. Ackley, Z. Combs, and R. E. Adams. Aging, weathering, and poisoning
of impregnated charcoals used for trapping radioiodine. Oak
Ridge National Laboratory, ORNL-TM-2860, 1970. 21 p.
1395. Milham, R. C. and L. R. Jones. Iodine retention studies.
Du Pont de Nemours (E. I.) and Co., Savannah River Laboratory,
DP-1234, 1970. 16 p.
1397. Stephenson, M. J., J. R. Merriman, D. I. Duthorn. Experimental
investigation of the removal of krypton and xenon from
contaminated gas streams by selective absorption in fluorocarbon
solvents. Union Carbide Corp., Oak Ridge Gaseous Diffusion
Plant, K-1780, 1970. 71 p.
1423. Morton, R. J. Land burial of solid radioactive wastes: study
of commercial operations and facilities. Environmental and
Sanitary Engineering Branch, Atomic Energy Commission,
WASH-1143, 1968. 132 p.
2148. Atomic Energy Commission. Forecast of growth of nuclear power,
Jan. 1971. Washington, U. S. Government Printing Office,
WASH-1139, 1971. 187 p.
70
-------
REFERENCES (CONTINUED)
2149. Code of Federal Regulations. Title 10--atomic energy. (Revised
as of Jan. 1, 1971). Washington, U. S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49--transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington, U. S.
Government Printing Office, 1971. 952 p.
2291. Forbes, S. G., and R. F. Berger. Fission product sampling and
decontamination development program. Phillips Petroleum
Co., Idaho Operations Office, IDO-172581, 1969. 37 p.
2292. Bendixsen, C. L., and G. F. Offutt. Rare gas recovery facility
at the Idaho Chemical Processing Plant. Idaho Nuclear Corp.,
Idaho Falls, Idaho, IN-1221, 1969. 43 p.
2666. Recommendations of the National Council on Radiation Protection
and Measurements. Basic radiation protection criteria, NCPR
Report No. 39, Washington, Jan. 1971.
71
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Iodine-129 Structural Formula
Half-life 1.7xl07
years
53
Type of Decay Negative Beta
129
Molecular Wt. 129 Melting Pt. 114 C Boiling Pt. 184 C
Density Gas: 11.27 grams/liter Specific Power 7.08xlO"8 watts/gm
Solid: 4.93 qm/cc at 20 C . co in-fl :—•—
Solubility " Specific Activity 1.62x10 H curies/gm
Cold Water Slightly soluble Hot Water Slightly soluble
Others: Dissolves readily in chloroform, carbon tetrachloride. & carbon
dlsulflde
Decay Chain Radiation Energy Level & Intensities
I ~7~~ *"Xe129 Beta: 0.150 mev (100%)
<1.7xlOy) (stable)
Shipping Regulations Classified as a
by the Department of Transportation
transport group III radlonuclide
Comments
References: (1) 0766
0
(2) 2150
72
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Iodine-131
Half-life 8 days
Type of Decay Negative Beta
Molecular Wt. 131 Melting Pt. 114 C
Density Gas: 11.27 gm/liter
Solid: 4.93 am/cc at 20
Structural Fnrtnula
53J
T131
Boiling Pt. 1840 C
Specific Power
„ .„. . . , 00_ .
Specific Act! vity 1,230 cunes/gm
Hot Water Slightly soluble __
Solubility
Cold Water Slightly soluble
Others: Dissolves readily in chloroform, carbon tetrachloride,& carbon
Decay Chain
disulflde
Radiation Energy Level & Intensities
131
(8d)
-Xe131
(stable)
Beta:
Gamma\
0.810 Mev
0.608 Mev
0.330 Mev
0.250 Mev
0.723 Mev
0.637 Mev
0.364 Mev
0.284 Mev
0.080 Mev
(0.7%)
(82.7%)
(9.3%)
(2.8%)
1.6%)
6.8%)
82%)
5.4%)
2.6%)
Shipping Regulations Classified as a transport group
the'Department of Transportation'
III radionuclide by
Comments
References: (1) 0766
(2) 2150
73
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Xenon-133
Name
Half-life 5.3 days
Type of Decay Negative Beta
Molecular Wt. 133 Melting Pt.
-11
Y 133
54Xe
2 C
Boiling Pt. -107 C
fie Power
Chemically inert
Solubility
Cold Hater
Others:
Dens i ty
Gas: 5.495 gm/liter at 20 C
Liquid: 3.52 qm/cc at -107C
Solid: 2.70 gm/cc at
Specific Activity
Hot Water
1.86x10 curies/gm
Decay Chain
B
-1
133
Xe
(5.3 days)
-Cs
133
(stable)
Radiation Energy Level & Intensities
Beta: 0.346 Mev (100%)
Gamma: 0.809 Mev (100%)
Shipping Regulations Classified as a transport group VI radionuclide at
pressures less than 14.7 psia and as a group III radionuclide at pressures
greater than 14.7 psia by the Department of Transportation
Comments Xenon is a member of the so-called rare or inert gases
References: (1) 0766
(2) 2150
74
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name
Krypton-85
Half-life 10.8 years
Type of Decay Negative Beta
Molecular Wt. 85 Melting Pt. -157 C
Structural Formula
36
Kr
85
Boiling Pt. -152 C
Densit^Liqauid:¥l6 gS/icT-?53 C Specific Power 0.53 watts/am
" - " Spec1fic Activit* 397 curies/gm
Cold Water .
Others: Chemically inert
Hot Water
Decay Chain
85
,-1
Kr
(10.8y)
.85
(stable)
Radiation Energy Level & Intensities
Beta: 0.690 Mev (99.4%)
0.150 Mev (0.6%)
Gamma: 0.514 Mev (0.4%)
Shipping Regulations Classified as a transport group VI radionuclide at
pressures less than 14.7 psia and as a group III radionuclide at pressures
greater than 14.7 psia by the Department of Transportation
Comments Krypton is a member of the so-called rare or inert gases.
References: (1) °766
(2) 2150
75
-------
PROFILE REPORTS
Plutonium-238, Plutonium-239. Plutonium-240. Plutonium-241
Americium-241, Americium-243, Cun'um-242, Curium-244
1. GENERAL
Introduction
Plutonium, americium, and curium are artificially produced radionuclides
that do not exist in nature except for very small amounts of plutonium.
These elements are characterized by their high radiotoxicity, long
half-life, and ability to fission. Plutonium-239 is the most important
among these elements because of its use in nuclear weapons and the place
it holds as the key material in the development of fast-breeder reactors
for the civilian nuclear power program. Plutonium-239 is readily
fissionable with neutrons and one lb of this material is equivalent
to about ten billion watt-hours of heat energy.
Americium, curium, and plutonium (except for Pu-241) decay by the
emission of high energy, 5 to 6 Mev, alpha particles. Plutonium-241
decays by beta emission to form americium-241. All these elements
eventually decay to lead, and the time required is on the order of millions
of years. The americium-243 decay chain is attached (Table 1) and the
three long-lived isotopes in this chain are: Pu-239, half-life
2.4xl04 years; U-235, half-life 7.1xl08 years; and Pu-231, half-life
A
3.3x10 years.
The projected growth in production of these isotopes will parallel
that of the civilian nuclear power program. The production of
plutonium-239 will increase with the planned introduction of the
fast-breeder reactors in the 1980's. Present projections indicate
that nuclear power will account for 30 percent of the total
power production by the year 1980. The expected total
77
-------
TABLE 1
AMERICIUM-243 DECAY CHAIN
Nuclide
Am243
95 ,
9-30
93NP"-
94P,
I
92U
I
gQTI
'
;239
?TC
-O3
^ QT
1
[
rooi
9, P."'
f997
Ar1""'
89 *
•
90TI
88RJ
86
277
^'
223
i~^\j
219
Name
Americium-243
Neptunium-239
Plutonium-239
Uranium-235
Thorium-231
Protactinium-231
Actinium-227
Thorium-227
Radium-223
Radon-219
Half-Life
7,950 years
2.4 days
24,400 years
0
7.1xlO°years
25 hours
n
3.3x10 years
21 .6 years
18.2 days
11.4 days
4 seconds
Major Radiation
Alpha and gamma
Beta and gamma
Alpha and gamma
Alpha and gamma
Beta and gamma
Alpha and gamma
Beta and gamma
Alpha and gamma
Alpha and gamma
Alpha and gamma
\J\J |
*01 C
84Po
84
Polonium-215
1 .8 minutes
Alpha
?91 1
82P
83B
81T
82P
r"
211
207
*- v/ /
207
\
Lead-211
Bismuth-211
Thallium-207
Lead-207
36 minutes
2 minutes
4.8 minutes
Stable
Beta and gamma
Alpha, beta, and gamma
Beta and gamma
78
-------
accumulated wastes for these isotopes from the civilian nuclear power
program to the year 2020 is attached (Table 2). These data assume that
99.5 percent of plutonium is recovered from the spent fuel elements.
Annual production figures or accumulated totals for these materials
produced as part of the nuclear weapons program were not available.
Manufacture
The transuranium elements are produced in cyclotrons and in thermal
power reactors. The majority of the isotopes are produced in civilian
nuclear power plants. At the present time there are 22 nuclear power
2151
plants in operation with an additional 104 being built or planned.
The present reactor fuel is uranium dioxide. In the future, plutonium
dioxide will be used as the fuel in the fast-breeder reactors.
Plutonium. Plutonium-238 is produced in cyclotrons by deuteron
bombardment of uranium or in nuclear reactors from neutron bombardment
by the reaction:
U238(n,2n)U237-l^NP238~Pu238
Plutonium-238 is also a daughter of Cm-242. Plutonium-239 is produced
in extensive quantities in nuclear reactors from natural uranium by
the reaction:
B"
100 oon ** Mn^-Jy On'1'-'
U238(n, Y)U239 ^P ^Pu
Plutonium-240 and plutonium-241 are produced by multiple* n-capture from
U-238 and Pu-239.
Americium. Americium-241 is a daughter of plutonium-241 and is
formed from plutonium-239 in a nuclear reactor by successive
neutron capture. Americium-243 is produced by multiple n-capture from
U-238 and Pu-239.
79
-------
TABLE 2
ACCUMULATED ACTINIDES IN SPENT FUEL PROCESSING WASTES FROM THE CIVILIAN NUCLEAR POWER PROGRAM1
0705
$
Accumulated Radionuclides
meqacunes
Plutonium-238*
41
Plutonium-239*
Plutonium-240*
Plutonium-241*
Americium-241
Americium-243
Curium-242
Curium-244
1970
0.002
0.00009
0.0001
0.03
0.009
0.0009
0.73
0.13
1980
1.2
0.02
0.04
6.6
2.3
0.23
43.2
29.9
Calendar- Year
1990
8.3
0.24
0.4
47.2
22.7
1.5
185.0
137.0
2000
30.7
1.3
1.9
191
121
5.2
487.0
255.0
2020
166
8.5
11.4
909
763
27.0
1,490
700
Assumes that 0.5 percent of the plutonium in the spent fuel is lost to waste.
-------
Curium. Curium-242 is produced in cyclotrons by the helium-ion
bombardment of Pu-239. It is also produced by multiple n-capture from
U-238 and Pu-238 and by the reaction:
Cerium-244 is produced by multiple n-capture from U-238, Pu-239, and Am-243
Uses
Plutonium-238 is used extensively for radioisotopic power devices
for space electric power, for radioisotopic heaters, and as an energy
source for cardiac pacemakers, heart pumps, and small undersea propulsion
systems. Plutonium-239 is used as a fuel in fast-breeder reactors and
in nuclear weapons. Presently, plutonium-239 is used in experimental
reactors but with the introduction of the fast-breeder reactors in the
civilian nuclear power program the requirements for its use will increase
rapidly. Curium and americium are used in various national laboratories
for research purposes and to produce heavier elements.
Sources and Types of Wastes
The majority of the transuranium elements are produced in nuclear
reactors and are contained within the spent fuel elements. The spent
fuel elements are processed for the recovery of the usable fission
materials (uranium and plutonium) which are refabricated into new fuel
elements. Solvent extraction using nitric acid is the means currently
used for the first-stage removal of the fissionable materials.
Approximately 99 percent of the plutonium is removed during the
processing. The other radionucl ides along with the unrecovered
plutonium are contained in the aqueous effluent from the reprocessing
step. The composition of fuel elements prior to processing have
been tabulated, and the .activity of plutonium, americium, and
curium products present in the waste stream following the processing
step are included (Table 3).
-------
TABLE 3
ACTIVITY OF PLUTONIUM, AMERICIUM, AND CURIUM PRODUCTS PRESENT IN WASTES GENERATED BY
THE PROCESSING OF SPENT REACTOR FUELS0705*
Light Water Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 30 MM/MTU)
00'
Actinide
Products
Pu238, Pu239, Pu240, Pu24!
Am241, Am243
Cm , Cm
Other Actinides
After 1 year
Grams/Tonne
48
144
32
5,436
Watts/Tonne
2.7
6.2
306
0.1
Curies/Tonne
641
189
8,430
30
Grams /Tonne
71
140
1
5,438
After 100 ^ears
Watts /Tonne
2
6
2
0.1
Curies/Tonne
64
179
57
25
Fast-Breeder Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 30 MW/MTU)
Actinide
Products
Pu238, Pu239, Pu240, Pu241
Am241 , Am243
Cm , Cm
Other Actinides
After 1 ^year
Grams/Tonne
428
742
20
4,550
Watts/Tonne
11
54
625
2
Curies/Tonne
3,203
1,620
17,000
277
Grams /Tonne
412
794
1
4,637
After lOCLyears
Watts/Tonne
8
48
3
1
Curies/Tonne
264
1,469
72
165
Assumes that 0.5 percent of the plutonium in the spent fuel is lost to waste.
-------
Physical and Chemical Properties
The physical and chemical properties of plutonium, americium, and
curium are included in the attached worksheets. Plutonium is highly
reactive, and in moist air oxidation proceeds rapidly, especially at
elevated temperatures. It dissolves in concentrated hydrochloric acid,
J.O
hydroiodic acid, or perchloric-acid with the formation of the Pu ion.
Plutonium exhibits six allotropic modifications having various
crystalline structures whose densities vary from 15.92 to 19.84.
Plutonium forms binary compounds with oxygen, carbon, nitrogen, silicon,
and the halides.
Americium is a ductile, malleable metal which is precipitated by the
fluorides, hydroxides, and oxalates. The element exists in three
oxidation states in aqueous solution. The trivalent state is highly
stable and difficult to oxidize. Curium is chemically similar to
gadolenium and is carried on rare-earth precipitates. The only stable
oxidation state that has been definitely identified in aqueous solutions
is +3.
2. RADIATION HAZARD
Plutonium is one of the most dangerous poisons known. The
permissible levels of concentration of plutonium in air and water are
the lowest of any of the radioactive elements. This is a result of the
concentration of plutonium directly in the blood-forming sections of
the bone, rather than the more uniform bone distribution shown by
other heavy elements. Americium and curium are also extremely hazardous
radioactive materials.
The effects of their radiation exposure are primarily dependent on
the amount of radiation and the portion of the body affected. The
dose delivered to a particular body organ following the inhalation of 1
microcurie of each of these radionuclides is attached (Table 4). For
plutonium-239 the dose delivered to the bone is 7,000 rem following the
83
-------
TABLE 4
PLUTONIUM, AMERICIUM, AND CURIUM DOSE TO A PARTICULAR BODY ORGAN FOLLOWING
INHALATION OF ONE MICROCURIE OF THE NUCLIDE
Isotope
Plutonium -2 38
Plutonium-238
Plutonium-239
Plutonium-239
Plutonium-240
Plutonium-240
Plutonium-241
Plutonium-241
Americium-241
Americium-241
Americium-243
Americium-243
Curium-242
Curium-242
Curium-244
Curium-244
Form
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Organ
Lung
Bone
Lung
Bone
Lung
Bone
Lung
Bone
Lung
Bone
Lung
Bone
Lung
Liver
Lung
Bone
Dose
rem
64
700
60
7,000
60
7,000
0.06
130
65
2,200
60
2,000
43
50
67
1,300
84
-------
inhalation of 1 microcurie (16.4 micrograms). The dose delivered to the
bone following the injection of 1 microcurie into the body via a wound
is 30,000 rem.
Standards for prolonged exposure over a fifty-year period have defined
the single dose limit in terms of the maximum permissible dose accumulated
in a period of 13 weeks. The whole body exposure limit is 3 rem per
quarter and the accumulated dose limit is 5(N - 18), where N is the
individual 's age in years. Limits for the thyroid, bone, and other organs
have also been defined. Values of the total body burden for each
radionuclide required to produce the maximum permissible dose rates
defined above have been compiled. For plutonium the critical organ
is the bone and the maximum permissible body burden is 0.04 microcuries
for Pu-238, Pu-239, and Pu-240 and 0.9 microcuries for Pu-241. For
americium-241 and americium-243 the critical organ is the bone and the
maximum permissible body burden is 0.05 microcuries. For curium-242
the critical organ is the liver and the maximum body burden is. 0.05
microcuries. For curium-244 the critical organ is the bone and the
maximum body burden is 0.1 microcuries.
3. OTHER HAZARDS
Besides its radiation hazard, plutonium is highly toxic. It is also
highly reactive in moist air. The toxicity of curium and americium are
unknown. Plutonium, americium, and curium also present a potential
hazard due to nuclear fission. Plutonium-239 represents the greatest
hazard since it readily undergoes fission with thermal neutrons, as well as
with those of higher energy and is capable of maintaining a self-sustaining
reaction.
85
-------
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling, Storage, and Transportation
These radionuclides are all alpha emitters except for plutonium-241.
The alpha particles have little penetrating power and can be stopped by
a sheet of writing paper. They are extremely hazardous if inhaled. Semi-
remote handling techniques are required to protect personnel from
radiation injury and adequate radiation shielding is required to avoid
excessive radiation exposure. These materials are safely handled in
glove boxes. The beta particles emitted by plutonium-241 are of low
energy (0.021 Mev) and cannot penetrate more than 0.002 in. of water.
These radionuclides are stored in controlled areas in
specially constructed containers. Extreme precautions are taken
to avoid the release of any material. Plutonium is stored at low
temperatures and in dry air to avoid corrosion. The storage container
is often protected by both a primary and a secondary containment
barrier. Precautions are also taken to prevent the formation of a
critical mass. Plutonium in a liquid solution is more likely to become
critical than solid plutonium. The container design and arrangement must
preclude any possibility of obtaining nuclear criticality. Special
monitoring systems and proper warning signs are located in the general
area of the storage facility.
Plutonium, americium, and curium are classified as a transport
group I radionuclide by the Department of Transportation, and the rules
and regulations governing their transportation are given in the Code
of Federal Regulations (CFR) Title 49—Transportation, Parts 170 to
190. Their content is limited to 0.001 curies for a Type A
package and 20 curies for a Type B package defined in 49CFR173. The
limits are increased to 20 and 500 curies if their physical form meets
the requirements of a special form material. Their release rate is
limited to zero under the specified accident conditions for Type A
and B quantities. ° Plutonium-239 is designated as a ffssile
material by the Department of Transportation and must meet the controls
required to provide nuclear criticality safety.
86
-------
Disposal/Reuse
The disposal of these materials is governed by the AEC Manual
Chapter 0524 and 10CFR20.2149 Two sets of standards have been
established for the permissible radiation exposure in unrestricted
areas. One is for the greatest dose received by an individual and the
other for the average dose received by the general population. The
standards for the safe release of these materials to the environment
in an unrestricted area are contained in lOCFRZO^9 and their
concentrations in air and water should not exceed the values listed
in this report (Table 5). These concentrations apply to an individual
and their release may be limited if a suitable sample of the population
is exposed to one-third the concentrations in air or water specified
in this report (Table 5).
Although not practiced, the disposal by release in a sanitary
sal t
2149
sewage system is limited to 0.1 microcurie. The disposal by burial
at any one location and time is limited to 10 microcuries.
5. EVALUATION OF WASTE MANAGEMENT PRACTICES
Recovery
Option No. 1 ~ Solvent Extraction. Solvent extraction processes have
successfully been used for well over a decade as the standard method
for the recovery of up to 99.8 percent of the plutonium and uranium
in spent reactor fuel elements. The basis for separation in this process
is the differences in solubility of these materials in the organic and
aqueous phase. Nearly all major fuel reprocessing facilities use the
Purex solvent extraction process. These facilities include those
operated under contract to the Atomic Energy Commission (AEC) and
Nuclear Fuel Services facility at West Valley, New York, which is
privately owned and operated. In the Purex process, the spent fuel
elements are dissolved in nitric acid and an organic compound, tributyl
phosphate (TBP) in an inert hydrocarbon, is added. When this organic
mixture is added, the TBP extracts both the uranium and the plutonium
87
-------
TABLE 5
PLUTONIUM, AMERICIUM, AND CURIUM MAXIMUM PERMISSIBLE CONCENTRATIONS2149*
Radionuclide
Plutonium- 238
Plutonium-238
Plutonium-239
Plutonium-239
Plutonium-240
Plutonium- 240
Plutonium-241
Plutonium-241
Americium-241
Americium-241
Americium-243
Americium-243
Curium-242
Curi um-242
Curium-244
Curium-244
Form Concentration in Air Concentration in Water
(mi crocuries/mi 11 i 1 1 ter) (mi crocuries/mi 1 1 i 1 i ter)
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
7x1 O-'4
IxlO'12
6x1 O"14
IxlO-12
6x1 O"14
IxlO"12
3xlO-12
IxlO"9
2xlO"13
4x1 O"12
2xlO"13
4x1 O"12
4xlO"12
6xlO-12
3x1 O"13
3xlO"12
5x1 O"6
3xlO"5
5x1 O"6
3x1 O"5
5x1 O"6
3x1 O"5
2x1 O"4
IxlO"3
4x1 O"6
2xlO"5
4x1 O"6
3xlO"5
2x1 O"5
3xlO"5
7x1 O"6
3x1 O"5
area.
Maximum permissible concentrations, apply to an individual in an unrestricted
88
-------
into the organic phase, while the fission products and other wastes
remain in the aqueous phase. Plutonium and uranium are separated by
reducing the plutonium to the trivalent form, usually with ferrous
sulfamate, and contacting with nitric acid. In the aquafluor process,
to be used at General Electric's Morris, Illinois plant, the uranium
and plutonium separation will be accomplished by ion exchange. The
ion exchange separation step will limit the plant's capacity since they
must be designed to prevent the accumulation of a critical mass.
Storage/Disposal
u
Option No. 1 - Land Burial. Disposal of these high-level radionuclides,
in small concentrations, by direct burial in unlined trenches is not
considered a satisfactory means of disposal. These types of wastes are
presently disposed of in this manner at approved sites. This method of
disposal does not provide for any type of secondary containment barrier
to prevent their release by leaching with the local ground water. Since
these materials have an extremely long half-life and a high radiotoxicity,
this method of disposal is not satisfactory.
Option No. 2 - Near-Surface Liquid Storage. Near-surface storage of
these radionuclides in stainless steel tanks encased in concrete and
buried underground is not considered as a satisfactory means of disposal.
Aqueous solutions of spent fuel reprocessing wastes have been stored in
this manner over the past 25 years. These tanks are considered as an
interim storage technique due to a general lack of confidence in their
long-term integrity. Therefore, the near-surface storage of these
wastes in steel tanks should only by considered as an interim storage
technique and not as a permanent storage or disposal technique.
Option No. 3 - Near-Surface Solid Storage. The storage of solidified
wastes in engineered surface facilities offers the best immediate method
for storage of these wastes. The necessary technology for these facilities
has been developed. The wastes will be under constant surveillance and
control and can be retrieved, should this be required. The wastes
89
-------
should be solidified and encapsulated in a suitable container (steel).
Of the four high-level solidification processes developed for reactor-
produced wastes, spray or phosphate glass solidification offer the
better solidified waste characteristics than the pot calcination or
fluidized bed calcination processes (see Radioactive Waste Solidifi-
cation report). The wastes will be stored in stainless steel-lined
concrete vaults which will be either air or water cooled.
Option No. 4 - Salt Deposits. This method offers the best potential
for the disposal of solidified, high-level wastes since bedded salt
deposits are completely free of circulating ground waters. This method
of disposal has been under study by the Oak Ridge National Laboratory
since 1957, and in November of 1970 a committee of the National Academy
of Sciences recommended that the use of bedded salt for the disposal of
radioactive waste is satisfactory. Recent questions concerning the
adequacy of this method have resulted in the need for further development
work before it can be accepted as an ultimate method of disposal. The
wastes must be solidified and disposed of in the solid form encapsulated
in a suitable container. The solidified wastes are buried in rooms
carved in the salt deposits approximately 1,000 ft below the ground.
The salt is a good heat transmitter, provides about the same radioactive
shielding as concrete, and can heal its own fractures by plastic flow.
The critical problem is the selection of a site that meets the necessary
design and geological criteria for the mixture of fission products and
actinide wastes.
Option No. 5 - Bedrock Disposal. The disposal of high-level liquid
wastes along with other spent fuel processing wastes in vaults excavated
in crystalline rock over 1,500 ft beneath the ground is currently being
evaluated by E. I. Du Pont de Nemours, at their Savannah River Plant
near Aiken, South Carolina. ' An advisory committee appointed by
the National Academy of Sciences recommended abandonment of the project.
In May 1972, another National Academy of Sciences panel concluded that
bedrock storage provides a reasonable prospect for long-term safe storage,
but precise information is needed to decide if and where underground
90
-------
storage vaults should be built. Another method has also been proposed
for disposing of liquid wastes by in situ incorporation in molten
silicate rock. At the present time both of these methods are
unproven since sufficient engineering data or exploration has not been
completed to verify their suitability. Due to the long half-lives and
high biological hazard of the actinides, it is doubtful that data could
ever be accumulated to prove that the geological characteristics of the
site over the next few thousand years are acceptable to ensure the
absolute safety of such a disposal method.
To summarize, plutonium, americium, and curium are hazardous,
long-lived isotopes for which extreme precautions are required in
their disposal. Plutonium is separated from the spent fuel processing
wastes by solvent extraction for reuse. Recovery factors as high as
99.8 percent can be obtained. The americium, curium, and remaining
plutonium are contained with the final reprocessing wastes. The
acceptable methods of treatment are spray or phosphate glass solidifi-
cation and disposal in either salt deposits or near-surface engineered
facilities. Plutonium, americium, and curium wastes from research
laboratories and other sources should also be solidified, encapsulated,
and disposed of in the same manner.
6. APPLICABILITY TO NATIONAL DISPOSAL SITE
Plutonium, americium, and curium are candidates for a National
Disposal Site due to their long half-lives, health hazard, and projected
growth with that of the civilian nuclear power program. From a cost
and safety viewpoint, it would be desirable to combine the reprocessing
plant and the National Disposal Site. Interim storage in near-surface,
engineered storage facilities offers considerable latitude in the site
selection. The site selection for ultimate disposal of these wastes is
limited to particular geological areas in which salt deposits are present.
91
-------
7. REFERENCES
0559. AEC Manual, Chapter 0524. Standards for radiation protection
May 12, 1964. 11 p.
0563. Recommendations of the International Commission on Radiological
Protection, ICRP Publication 2. Report on Committee II on
permissible dose for internal radiation (1959). Peryamon
Press, 1959. 232 p.
0705. Staff of the Oak Ridge National Laboratory. Siting of fuel
reprocessing plants and waste management facilities.
ORNL-4451, 1970. 500 p.
0733. Report by the Committee on Radioactive Waste Management. Disposal
of solid radioactive wastes in bedded salt deposits. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p.
0766. Sax, N. I. Dangerous properties of industrial materials. 3d ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
0894. Girdler, R. M. Storage of liquid radioactive .wastes at the
Savannah River Plant. Du Pont de Nemours and Company,
Aiken, South Carolina, DPSPU-69-30-9, July 1969. 11 p.
1396. Cooke, J. B. and et. al. Report on the proposal of
E. I. Du Pont de Nemours and Company for the permanent
storage of radioactive separation process wastes in bedrock
on the Savannah River, Du Pont de Nemours (E. I.) and Company,
Aiken, South Carolina, DPST-69-444, May 1970. 57 p.
2146. Cohen, J. J., A. E. Lewis, and R. L. Braun. In situ incorporation
of nuclear waste in deep molten silicate rock. Nuclear Technology,
13: 76-87, Apr. 1972.
2147. Harris, D., and J. Epstein. Properties of selected radioisotopes.
Goddard Space Flight Center, Greenbelt, Maryland,
NASA SP-7031, 1968. 89 p.
2148. Atomic Energy Commission. Forecast of growth of nuclear power,
Jan. 1971. Washington, U.S. Government Printing Office,
WASH-1139, 1971. 187 p.
-------
REFERENCES (Continued)
2149. Code of Federal Regulations. Title 10--atomic energy. (Revised
as of Jan. 1, 1971). Washington, U.S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49—transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington, U.S.
Government Printing Office, 1971. 952 p.
2151. Atomic Energy Commission. The nuclear industry, 1971. Washington,
U.S. Government Printing Office, WASH-1174-71, 1971. 193 p.
93
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Plutonium-238 Structural Formula
Half-life 86.4 years
Type of Decay Alpha and Spontaneous Fission
PU238
94PU
Molecular Wt. 238 _ Melting Pt. 639 C Boiling Pt. 3235 C
Density 19.84 gm/cc Specific Power 0.56 watts/gm. _
Solubility Specific Activity 17.4 curries/gm
Cold Hater _ Hot Water _
Others: Soluble HC1 ; insoluble in HN03 and concentrated H^SO,
Decay Chain Radiation Energy Level & Intensities
.234
94 ^92 Alpha: 5.499 Mev (72%)
5 5'456 Mev
(86. 4y) (2.5X10y)
Gamma: 0.0435 Mev (0.04%)
Shipping Regulations Classified as a transport group I radionuclide
by the Department of Transportation
Comments
References: (1) 0766
(2) 2147
(3) 2150
94
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Plutonium-239 Structural
Half-life 24,400 years
Type of Decay Alpha and Spontaneous Fission
Pu239
94PU
Molecular Wt. 239 _ Melting Pt. 639 C Boiling Pt. 3235 C
Density 19.84 gm/cc Specific Power 0.0019 watts /am
Solubility Specific Activity 0.061 curries/am
Cold Water __ _ Hot Water _
Others: Soluble HC1 . insoluble in HNOj and concentrated H^S04
Decay Chain Radiation Energy Level & Intensities
Pu239 ° U235
94 "*92U Alpha: 5.157 Mev (74%)
(2.4X10',) (7.,X10
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Plutonium-240
Half-life 6.580 years
Type of Decay Alpha and Spontaneous Fission
Molecular Wt. 240 Melting Pt. 639 C
Density 19.84 gm/cc
Structural Formula
94
Pu
240
Boiling Pt. 3235 C
Solubility^
Cold Water
Others:
Specific Power 0.0069 watts/qm
Specific Activity
Hot Water
0.227 curries/gm
Decay Chain
Pu?40
94PU
(6580y)
soluble HC1; insoluble in HNO^ and concentrated HoSO,
236
92U
(2.4X107y)
Radiation Energy Level & Intensities
Alpha: 5.168 Mev (76%)
5.123 Mev (29%)
Gamma: 0.045 Mev (24%)
Shipping Regulations Classified as a transport group I radionuclide
by the Department of Transportation
Comments
References: (1) 0766
(2) 2147
(3) 2150
96
-------
241
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Plutonium-241 Structural
Half-life 13.2 years
Type of Decay Negative Beta
Molecular Wt. 241 Melting Pt. 639 C Boiling Pt. 3235 C
Density 19.84 gm/cc Specific Power 0.0048 watts/gm
Solubility Specific Activity 112 curries/gm
Cold Water Hot Water ^
Others: Soluble HC1; insoluble in HN03 and concentrated H^SO.
Decay Chain Radiation Energy Level ?< Intensities
Pu241 B" .241
94ru - 95nm Beta: 0.021 Mev (100%)
03.2y) (458y) Also emits alpha particle 0.0023%
of the time.
Shipping Regulations Classified as a transport group I radionuclide
by the Department or iransportation.'
Comments
References: (1) Q766
(il\ 2147
(3) 2150
97
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Americium-241 Structural Formula
Half-life 458 years
Type of Decay Alpha and Spontaneous Fission
Molecular Wt. 241 _ Melting Pt. 994 C Boiling Pt. 2607 C
Density 11.7 gm/cc _ Specific Power 0.11 watts /qm _
Solubility Specific Activity 3.24 curries /am
Cold Water t _ . _ Hot Water _ k
Others: Soluble in dilute acid _ _
Decay Chain Radiation Energy Level & Intensities
Am241 a N 237
95Hm -- - 93Mp Alpha: 5.486 Mev (86%)
6 5'443 Mev
Gamma: 0.060 Mev (36%)
0.026 Mev (3%)
Shipping Regulations Classified as a transport group I radionuclide
by the Department of Transportation.
Comments
References: (1) 0766
(2) 2150
98
-------
243
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Americium-243 Structural
Half-life 7.950 years
Type of Decay Aljha
Molecular Wt. 243 Melting Pt. 994 C Boiling Pt. 2607 C
Density 11.7 gm/cc Specific Power 0.0062 watts/gm
Solubility Specific Activity 0.185 curries/gm
Cold Water • Hot Water
Others: Soluble in dilute acid
Decay Chain Radiation Energy Level & Intensities
Am243 a Nn239 B- p 239 Americium-243
95 »• 93"" +• Q4KU
y* Alpha: 5.276 Mev (88%)
5.234 Mev (11%)
5.181 Mev (1.1%)
m f t • fSl I M • ««r • l.r W I IV. * ^ \J\JIQ f
(7,950y) (2.4d) (2.4X10>) 5.234 Mev (11%)
(1.
Gamma: 0.075 Mev (50%)
0.044 Mev (4%)
Neptum'um-239
Beta: 0.437 Mev (Max)
Shipping Regulations Classified as a transport group I radionuclide
by the Department or iransportationT"
Comments
References: (1) 0766
(2) 2150
99
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Curium-242
Name
Half-life 163 days
Type of Decay Alpha and Spontaneous Fission
Molecular Wt. 242 Melting Pt. 1340 C
Density 7 gm/cc
Structural Fnrmnla
96
Cm'
.242
Boiling Pt.
So1ubi1ity_
Cold Water
Others:
96Cm'
(163d)
Decay Chain
242 a
Pu238
94PU
(86 years)
Specific Power 122 watts/gm
Specific Activity
Hot Water
3,320 curries/gm
Radiation Energy Level & Intensities
Alpha: 6.115 Mev (74%)
6.071 Mev (26%)
Gamma: 0.158 Mev (.003%)
0.102 Mev (.004%)
0.044 Mev (.041%)
Shipping Regulations Classified as a transport group I radionuclide
by the Department ot iransportatlon.
Comments
References: (1) 0766
(2) 2147
(3) 2150
100
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Curium-244 Structural Formula
Half-life 17.6 years
96
Type of Decay Alpha and Spontaneous Fission
Molecular Wt. 244 _ Melting Pt. 1340 C Boiling Pt. _
Density 7 gm/cc _ Specific Power 2.7 watts/gm _
Solubility Specific Activity 83 curries /gm
Cold Water ' _ Hot Water _
Others: _
Decay Chain Radiation Energy Level & Intensities
244 a 240
(17. 6y) (6580y) 5'766 Mev
Gamma: 0.150 Mev (.0013%)
0.100 Mev (.0015%)
0.043 Mev (.0210%)
Shipping Regulations Classified as a transport group I radionuclide by
the Department of Transportation.
Comments
References: (1)0766
(2)2147
(3)2156
101
-------
PROFILE REPORT
Ruthenium-106 (Rhodium-106), Cerium-144 (Praseodymium-!44).
Promethium-147
1. GENERAL
Introduction
Ruthenium-106, cerium-144, and promethium-147 are radioactive
isotopes that are produced in nuclear reactors by the fission of
uranium. Ruthenium-106 has a half-life of 368 days and emits only
beta particles. It exists in radiation equilibrium with its short
half-lived daughter, rhodium-106. Rhodium-106 has a half-life of
30 seconds and emits both gamma rays and high energy (3.55 Mev)
beta particles. Cerium-144 is both a beta and gamma emitter and has
a half-life of 284 days. It decays by beta emission to form
praseodymium-144 which has a half-life of 17.3 minutes. Promethium-147
has a half-life of 2.62 years and decays by the emission of a 0.224
beta particle to form stable samarium-147.
These three isotopes, Ru-106, Ce-144, and Pm-147 have moderately
long half-lives (284 days to 2.6 years) and account for a majority of
the total fission product activity and heat content in the spent fuel
processing. After one year of radioactive decay, they are responsible
for 65 percent of the total heat content and for 70 percent of the
total activity in the high-level waste.
Since these isotopes are principally produced in nuclear reactors,
their projected growth will parallel that of the civilian nuclear
power program. Present projections indicate that nuclear power will
account for 30 percent of the total power production by the year
1980. Annual production figures from the civilian and nuclear
power program for these three radionuclides to the year 2020
103
-------
for light-water and fast-breeder reactor fuels are attached (Tables 1 to 3),
These figures illustrate the tremendous growth anticipated in their
production.
Manufacture
Presently, these three radionuclides are produced in thermal
reactors by the fission of uranium. The ffssion yield for these three
radionuclides is: cerium-144, 5.6 percent; promethium-147, 2.6 percent;
and ruthenium-106, 0.38 percent. Civilian nuclear power plants are the
major producers and a smaller amount is produced at AEC facilities. At
the present time there are 22 nuclear power plants in operation with
21 51
an additional 104 being built or planned. They will also be
produced in fast-breeder reactors by the fission of plutonium.
Uses
Promethi.um-147 is used in radioisotope power generators and as a
beta source in the radioactive gauge field. It is also used in the
preparation of self-illuminating materials and devices for signs and
signals. At the present time promethium-147 is separated from the
reactor-produced waste streams and distributed by the AEC through the
Oak Ridge National Laboratory, Isotope Sales Department. The AEC
pi r I
revenue from its distribution for the years 1968 to 1971 is :
Year Revenue (Dollars)
1968 56,000
1969 45,000
1970 16,000
1971 30,000
In the future when its production is warranted by market demands,
it would be available in large quantities from commercial firms
operating chemical reprocessing plants. The commercial use for
ruthenium-106 and cerium-144 is very limited.
104
-------
TABLE 1
RUTHEMIUM-106 CONTENT IN HIGH-LEVEL WASTE PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM'
Light-Hater Reactor Fuels
,0705
Annual
Production 1970
5
grams/year .071x10
L— i * Q
O curies/year .024x10
W watts/year .141xl04
Annual
Production 1970
grams/year
curies/year
watts/year
Calendar Year
1980 1985
4.06xl05 7.30xl05
1.36xl09 2.45xl09
8.05xl04 14.5xl04
Fast-Breeder Reactor
Calendar Year
1980 1985
1.37xl05
0.46xl09
2. 73x1 O4
1990
8,26xl05
2. 77x1 O9
16.4xl04
Fuels
1990
8.26xl05
2.77xl09
16.5xl04
2000
6.55xl05
Z.19xl09
13.0xl04
2000
35. 3x1 O5
ll.SxlO9
70. 2x1 O4
2020
19.2xl05
6. 45x1 O9
38.2xl04
2020
106X105
35. 6x1 O9
211xl04
-------
TABLE 2
CERIUM-144 CONTENT IN HIGH-LEVEL WASTE PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM'
Light-Hater Reactor Fuels
0705
M
Annual
Production
grams/year
curies/year
watts/year
1970
.15xl05
.046xl09
.041xl06
1980
8.26xl05
2. 64x1 O9
2.31xl06
Calendar Year
1°85
14.9xl05
4.75xl09
4.17xl06
1 990
16.8xl05
5,38xl09
4.72xl06'
2000
13.3xl05
4.26xl09
3.74xl06
2020
39.2xl05
12.5xl09
ll.OxlO6
Fast-Breeder Reactor Fuels
Annual
Production 1970
grams/year
curies/year
watts/year
Calendar
1980 1985
1.43xl05
.46xl09
.40xl06
Year
1990
8.6xl05
2.75xl09
2.41xl06
2000
36.7xl05
ll.SxlO9
10.3xl06
2020
noxio5
35. 3x1 O9
30.9xl06
-------
TABLE 3
PROMETHIUM-147 CONTENT IN HIGH-LEVEL WASTE PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM0705
Light-Hater Reactor Fuels
8
si
Annual
Production
grams/year
curies/year
watts/year
1970
. 058x1 O5
.054xl08
.028xl05
1980
3. 32x1 05
3. 08x1 O8
1.58xl05
Calendar Year
1985 1990
5.97xl05 6.75xl05
5.54xl08 6.27xl08
2.85xl05 3.23xl05
2000
5.35xl05
4.97xl08
2. 56x1 O5
2020
15.7xl05
14.6xl08
7.52xl05
Fast-Breeder Reactor Fuels
Annual
Production 1970
grams/year
curies/year
watts/year
Calendar Year
1 980 1 985 1 990
1.36xl05 8.17xl05
1.26xl08 7.59xl08
.65xl05 3.91xl05
2000
34.9xl05
32. 4x1 O8
16.7xl05
2020
105xl05
97.4xl08
50.2xl05
-------
Sources and Types of Wastes
Ruthenium-106, cerium-144, and promethium-147 are only a few of the
many fission products produced in nuclear reactors and are contained within
the spent fuel elements. The spent fuel elements are removed from the
reactor and processed for the recovery of usable fissionable materials.
These radionuclides and the other high-level wastes are contained in the
aqueous effluent from the spent-fuel processing step. The range of
chemical compositions of these waste streams have been tabulated.
These waste streams are primarily aqueous solutions of inorganic nitrate
salts £nd any differences in their composition occurs mainly in the
amounts and types of salts added during the processing step. The amount
and activity of Ru-106, Ce-144, and Pm-147 present in the high-level
waste streams after 1 year and 10 years is attached (Table 4).
These radionuclides can also be found in the secondary waste streams
generated at the spent fuel processing plant. The volume of these waste
streams can be quite large but the radionuclide activity in these streams
is quite low. Ruthenium-106 which usually oxidizes and volatilizes
during the fuel reprocessing step is also found in the off-gas streams.
Physical and Chemical Properties
The physical and chemical properties of ruthenium-106, cerium-144,
and promethium-147 are included in,the attached worksheets. Ruthenium-106
exhibits all possible positive valences; consequently, its chemistry is
exceedingly complex. In the lower oxidation states it is basic in
nature, while in the higher oxidation states it tends to be acidic. In
alkaline solutions it may be oxidized to the ruthenite, ruthenate,
perruthenate, or the tetroxide, depending upon the strength of the
oxidizing agent. The tetroxide is volatile, boiling at around 100 C.
Cerium-144 is a member of the rare-earth group. Cerium has a
valence of +3 but can also assume a valence of +4. Tetravalent cerium
is a powerful oxidizing agent. The hydroxides and nitrates of cerium
are slightly soluble in water. The carbonates, oxalates, and phosphates
108
-------
TABLE 4
RUTHENIUM-106, CERIUM-144, AND PROMETHIUM-147 CONTENT IN HIGH-LEVEL FUEL REPROCESSING WASTE0705
Light-Water Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 30 MW/MTU)
Fission"
Products Graris /Tonne
M
O
CD
Ru10§ Rh106
Ce144 pr!44
PJ47
Other Fission
Products
81
143
92
34,784
After 1 year
tlatts/Tnnne
2,656
3,800
44
3,500
Curies/Tonne
546,000
912,000
85,000
679,000
Grans/Tonne
0.16
0.05
8
35,002
After 10 years
l-'atts /Tonne
5
1.5
4
1,019
Curies/Tonne
1,100
300
8,000
307,600
Fast Breeder Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 58 MW/MTU)
Fission
Products Grams/Tonne
Ru™ Rh106
Ce144 pr!44
Pn,147
Other Fission
Products
204
177
300
34,219
After 1 year
Uatts/Tonne
6,660
4,700
150
2,290
Curies/Tonne
1,370,000
1,128,000
279,000
653,000
After in years
Grams/Tonne
0.41
0.06
28
34,872
Watts/Tonne
14
2
13
737
Curies/Tonne
2,760
370
26,000
251 ,870
-------
are not soluble. Promethium-147 is also a member of the rare-earth group
and its chemistry is similar to that of cerium.
2. RADIATION HAZARD
Ruthenium-106 and cerium-144 are dangerous radioactive materials.
Promethium-147 is a moderately dangerous radioactive material. The
effects of their radiation exposure are primarily dependent on the
amount of radiation and the portion of the body affected. The effects
of acute whole-body gamma radiation exposure are: (1) 5 to 25 rads, minimal
dose detectable by chromosome analysis or other specialized analyses, but
not by hemogram; (2) 50 to 75 rads, minimal acute dose readily detectable
in a specific individual (e.g., one who presents himself as a possible
exposure case); (3) 75 to 125 rads, minimal acute dose likely to produce
vomiting in about 10 percent of people so exposed; (4) 150 to 200 rads,
acute dose likely to produce transient disability and clear hematological
changes in a majority of people so exposed; (5) 300 rads, median lethal
pccc
dose for single short exposure. These effects are for a single large
dose of radiation or a series of substantial doses in a short interval of
time to the total body. Standards for prolonged exposure over a fifty-
year period have defined the single dose limit in terms of maximum
permissible dose accumulated in a period of 13 weeks. The whole body
exposure limit is 3 rem per quarter for a radiation worker and the
accumulated dose limit is 5(N - 18), where N is the individual's age in
years. Limits for the thyroid, bone, and other organs have also been
defined.0563
Values of the total body burden for each radionuclide required to
produce the maximum permissible dose rates defined above have been
compiled. For ruthenium-106 the critical organ is the kidney and
the maximum permissible body burden is 3 microcuries. For cerium-144
and promethium-147 the critical organ is the bone and the maximum
permissible body burden is 5 microcuries and 60 microcuries, respectively.
110
-------
Their radiological toxicity can also be expressed in terms of the
dose delivered to a particular body organ following the inhalation of
1 microcurie. For each of these radionuclides the single inhalation
of 1 microcurie will produce the following doses:
Isotope Form Organ Dose
Ru-106 Insoluble Lung 1.2 rem
Ru-106 Soluble Kidney 0.04 rem
Ce-144 Insoluble Lung 1.0 rem
Ce-144 Soluble Bone 1.1 rem
Pm-147 Insoluble Lung 0.07 rem
Pm-147 Soluble Bone 0.2 rem
3. OTHER HAZARDS
Cerium-144 resembles aluminum in its pharmacological action as well
as in its chemical properties. Besides its radiation hazard, cerium-144
is not toxic to the skin or mucous membranes by repeated exposure and is
only slightly toxic if absorbed into the body by inhalation or injection.
Its fire and explosive hazard is moderate. It ignites spontaneously
in air at temperatures of 150 C to 180 C. The toxicity of cerium compounds
is the same as cerium, except when the anion has a toxicity of its own.
The details of ruthenium-!06 toxicity are unknown, but it is believed
to be toxic. It is dangerous when heated to decomposition since
it emits toxic fumes of ruthenium oxide. Data on the toxicity of
promethium-147 were not available.
Ill
-------
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling, Storage, and Transportation
Since all three radionuclides are hazardous to man by inhalation,
ingestion, or direct radiation exposure, care is exercised in their
handling. Special procedures and radiation shielding are utilized in
their handling. The beta particles emitted by Ru-106, Ce-144, and Pm-147
are of low energy and cannot penetrate more than 0.02 in. of water.
The daughter products of Ru-106 and Ce-144 emit high energy beta
particles which can penetrate up to 0.8 in. of water. The principal
shielding problem associated with their daughter's beta radiation is the
bremsstrahlung effect which results from the presence of highly penetrating
rays resulting from the deflection of the beta particles by the nuclei
in the shielding medium. Since the bremsstrahlung effect increases with
atomic number, shielding materials of high atomic weight are not
satisfactory. Ru-106, Ce-144, and their daughters also emit gamma
radiation. The gamma radiation is highly penetrating and high-density
shields, such as lead, are required to stop the radiation. To detect
and control personnel exposure to their radiation all persons working
with this material should wear dosimetry devices which directly
indicate the dose. Commonly used devices are the film badge
and the thermoluminescent dosimeters (TLD).
These materials are stored in controlled reservations in
specially constructed containers which are protected by both a primary
and a secondary containment barrier. Special monitoring systems and
proper warning signs are located in the general area of the storage
facility.
Ruthenium-106 and cerium-144 are classified as a transport group
III radionuclide and promethium-147 as a transport group IV radionuclide
by the Department of Transportation. The rules and regulations
governing their transportation are given in the Code of Federal
Regulations (CFR) Title 49—Transportation, Parts 170 to 190.2150
112
-------
The Ru-106 and Ce-144 content is limited to 3 curies for a Type A
package and 200 curies for a Type B package defined in 49CFR173. The
Pm-147 content is limited to 20 curies for a Type A package and
200 curies for a Type B package. These limits for each radionuclide
are increased to 20 and 5,000 curies if their physical form meets the
requirements of a special form material. Their release rate is limited
to zero under the specified accident conditions for Type A and B
quantities. The allowable release of radioactivity from packages
containing large quantities of these materials is limited to gases and
contaminated coolant containing total radioactivity exceeding neither
0.1 percent of the total radioactivity of the package nor 10 curies
under the hypothetical accident conditions prescribed in 49CRF173.
Disposal/Reuse
The disposal of these materials is governed by the AEC Manual
0550 ?14Q
Chapter 0524U3M and 10CFR20. Two sets of standards have been
established for the permissible radiation exposure in unrestricted
areas. One is for the greatest dose received by an individual and
the other for the average dose received by the general population.
The radiation protection standards for these two groups are attached
(Table 5). These standards also define their maximum concentrations in
air and water for exposure to either an individual or to the general
population. For an individual, the safe release of these materials to
the environment in an unrestricted area should not exceed the
concentrations in microcuries per milliliter (yc/ml) listed below :
Concentration Concentration
Isotope Form in Air in Hater
Ru-106 Soluble 3xlOr9 yc/ml IxlO"5 yc/ml
_7n -5
Ru-106 Insoluble 2x10 yc/ml 1x10 D yc/ml
Ce-144 Soluble 3xlO"10 yc/ml IxlO"5 yc/ml
Ce-144 Insoluble 2xlO"10 yC/ml IxlO"5 yc/ml
Pm-147 Soluble 2xlO"9 uc/ml 2xlO~4 yc/ml
Pm-147 . Insoluble 3xlO"9 yc/ml 2xlO"4 yc/ml
113
-------
TABLE 5
RADIATION PROTECTION STANDARDS FOR INDIVIDUALS AND POPULATION GROUPS
2149
FOP. EXTERNAL AND INTERNAL EXPOSURE
Tyoe of Exposure
Dose to Individuals
at Points of Maximum
Probably Exposure
(rem per year)
Average Dose to a
Suitable Population
Sample
(rem per year)
'..'hole bociy, gonads, or
bone marrov.1
Thyroid or bone
Bone (alternate standards)
0.5
1.5
Body burden at 0.003
micrograms of radium
226 or fts biological
equivalent
0.17
0.5
Body burden of 0.001
micrograms of radium
226 or its biological
eoutvalent
-------
The concentrations for the safe release of these materials to the general
population are one-third the above values.
The disposal by release into a sanitary sewage system is limited
to 10 microcuries for Ru-106 and Ce-144 and to 100 microcuries for
Pm-147. The disposal by burial in the soil at any one location and
time is limited to 1,000 microcuries for Ru-106 and Ce-144 and to
2149
10,000 microcuries for Pm-147.
5. EVALUATION OF WASTE MANAGEMENT PRACTICES
Recovery
Option No. 1 - Volatilization-Absorption. Ruthenium-106 removal from
solutions can be accomplished by oxidizing the ruthenium to its tetroxide
state and volatilizing it by simultaneous heating and gas sparging. In the
tetroxide state, ruthenium bolls at 100 C. Ruthenium can be removed
from uranium nitrate solutions by sparging with air containing ozone.
Ruthenium decontamination factors (ratio of initial to final concentration)
of about 100 were obtained during an 8-hour sparge with 1 percent ozone in
air. Ruthenium can also be oxidized to the volatile form by hot
concentrated nitric acid. The volatized ruthenium is then recovered by
absorption. Two such absorbents are pyrolusite and activated carbon.
Decontamination factors with pyrolusite alone are less than 100.0714
Activated carbon used subsequent to pyrolusite absorption increases
the decontamination factor 10 times.
Option No. 2 - Scavenging-Precipitation Foam Separation. The removal
of radionuclides from low-level radioactive waste water by scavenging-
precipitation form separation has been studied by Oak Ridge National
Laboratory. The process consists of two steps: (1) precipitating
in a sludge-blanket clarification step, and (2) achieving final
decontamination in a foam-separation column. Processing rates of 300 gal.
per hour (gph) for scavenging-precipitation and 120 gph for foam separation
115
-------
have been achieved. Cerium-144 decontamination factors of greater than
20 have been obtained. Ruthenium-106 decontamination factors of only
four were obtained. The low decontamination factor is expected since
ruthenium is one of the most difficult radionuclides to remove by
scavenging.
Option No. 3 - Scavenging-Precipitation Ion Exchange. In this process
final decontamination is obtained by ion exchange columns from the
scavenging-precipation process. The process includes a provision
for the recycle of the ion exchange waste to the scavenging-precipitation
step. All the removed radionuclides are concentrated in the clarifier
sludge. For cerium-144 and promethium-147 the overall decontamination
factors varied from 20 to 700 and for ruthenium-106 the overall
decontamination factor varied from 1.5 to 8. For cerium-144 and
promethium-147, the use of ion exchange resins with a scavenging-precipitation
step is a fairly simple and efficient means of removing these radionuclides
from low-level, radioactive aqueous wastes. For ruthenium-106 wastes
this method is not satisfactory since ruthenium in the liquid form is
not very susceptible to ion exchange.
Option No. 4 - Water Recycle. The water recycle process is used for
decontaminating radioactive waste water and recycling the purified
water for reuse. This process has been demonstrated at the pilot plant
scale by Oak Ridge National Laboratory. 07° The steps in the
process include: (1) clarification by controlled addition of coagulants,
(2) demineralization by cation-am'on exchange, and (3) sorption on
granular activated carbon. For optimum removal of the ruthenium-106
during the coagulation-clarification step the pH of the waste must
range between 7 and 8 for proper alum floe formation. High decontamination
factors for ruthenium-106 can only be obtained during the final step by
sorption on activated carbon. The overall decontamination factor for
ruthenium-106 was 1,230 and for cerium-144 it was 800. The method is
an improvement to the treat-and-discharge methods, but further work is
required until full-scale production use is obtained.
116
-------
Disposal
Option No. 1 - Land Burial. Land burial of Ru-106, Ce-144 and Pm-147
wastes, in small concentrations, at approved sites that are acceptable
from a geologic and hydro!ogic standpoint, is an acceptable means of
disposal. Their concentrations should not be in excess of 10 times
the maximum permissible concentration for the general population in
10CFR20.2149 All wastes to be disposed of should be in a solid form
and encapsulated in a suitable container. Liquid wastes should be
solidified, preferably using asphalt, in accordance with the methods
described in the radioactive waste solidification report. The burial
trenches should be designed not to intercept the ground water table
and constructed with a bottom drain and sump for water monitoring.
The trenches should be covered with either asphalt or vegetation to
limit infiltration of water. The burial site design, geology, and
hydrology should be in conformance with the criteria used in selecting
and licensing the present commercial burial sites. Since land
burial is successfully practiced on a commercial scale and since
these isotopes have moderate half-lives, it should be considered as
the most satisfactory method of disposing of all dilute concentrations
of these wastes. '
Option No. 2 - Near-Surface Liquid Storage. Near-surface ;
storage of aqueous wastes in stainless steel tanks encased in
concrete and buried underground is not considered as a satisfactory
means of disposal. Aqueous wastes have been stored in this
manner over the past 25 years. The tanks range in size from 0.33 to
1.3 million gal. and are generally equipped with devices for
measuring temperatures, liquid levels, and leaks. At the present time
these tanks are considered as an interim storage technique due to a
general lack of confidence in their long-term integrity. Since
present regulations require the solidification of all reactor wastes
2149
within 5 years following reprocessing,- • the near-surface storage
of these wastes.in steel tanks should only be considered as a near-term
storage technique and not as a permanent storage or disposal technique.
117
-------
Option No. 3 - Near-Surface Solid Storage. The storage of high-level
solidified Ru-106, Ce-144, and Pm-147 wastes in engineered surface
facilities offers the best immediate method for storage of these
wastes. The technology for these facilities has been developed. The
wastes will be under surveillance and control and can be retrieved,
should this be required. The aqueous wastes should be solidified and
packaged in a suitable container (steel). Of the four high-level
solidification processes developed, spray and phosphate glass solidifi-
cation offer the best solidified waste characteristics (see Radioactive
Waste Solidification report). The wastes will be stored in stainless
steel-lined concrete vaults which will be either air or water cooled.
The periodic replacement of waste containers probably will not be
required since their activity is reduced by a factor of 1,000 in 10
years for Ru-106 wastes, in 21 years for Ce-144 wastes, and in 26
years for Pm-147 wastes. The replacement of the waste containers
might be required if other long-lived isotopes are also present.
Option No. 4 - Salt Deposits. This method offers the best
potential for the disposal of wastes from fuel reprocessing since
bedded salt deposits are completely free of circulating ground
waters. This method of disposal has been under study by .the Oak Ridge
National Laboratory since 1957, and in November 1970 a committee of
the National Academy of Sciences recommended that the use of bedded
salt for the disposal of radioactive wastes is satisfactory.
Recent questions concerning the adequacy of this method have resulted
in the need for further development work before it can be accepted as
an ultimate method of disposal. The reactor-produced wastes must be
solidified and disposed of in the solid form encapsulated in a
suitable container. The preferred solidification process is spray
solidification. The solidified wastes are then buried in rooms carved
in the salt deposits approximately 1,000 ft below the ground. The
salt is a good heat transmitter, provides about the same radioactive
shielding as concrete, and can heal its own fractures by plastic
flow. Salt deposits should be considered for the disposal of these
wastes at approved sites that are acceptable from a design and geo-
logical standpoint.
118
-------
Option No. 5 - Bedrock Disposal. The disposal of liquid wastes along
with other spent fuel processing wastes in vaults excavated in crystalline
rock over 1,500 ft beneath the ground is currently being evaluated by
E. I. du Pont de Nemours, at their Savannah River Plant near Aiken,
South Carolina. ' The wastes would be stored in six tunnels
and once in the tunnels the wastes will seep into the surrounding
rock. Located above the crystalline rock is the Tuscaloosa formation,
a good source of freshwater, which is separated by a layer of clay that
would act as a barrier to the leakage of radioactive wastes. An
advisory committee appointed by the National Academy of Sciences
recommended abandonment of the project. In May 1972, another National
Academy of Sciences panel concluded that bedrock storage provides a
reasonable prospect for long-term safe storage but precise information
is needed to decide if and where underground storage vaults should be
built. Another method has also been proposed for disposing of liquid
wastes by in situ incorporation in molten silicate rock. At the
present time both of these methods are unproved since sufficient
engineering data or exploration has not been completed to verify their
suitability.
Option No. 6 - Hydraulic Fracturing. The direct disposal of aqueous
low-level radioactive wastes into shale formations has been investigated
by Oak Ridge National Laboratory.0 The method consists of mixing
the aqueous wastes with cement and pumping the resulting slurry down a
well and out into a nearly horizontal fracture in a thick shale formation.
Additional work is required to demonstrate that this method of
disposal is satisfactory.
To summarize, ruthenium-106, cerium-144, and promethium-147 are
radioactive isotopes that account for a majority of the initial
activity present in spent fuel processing wastes! The acceptable method
of treatment is solidification and interim storage in near-surface
engineered facilities followed by permanent disposal in salt deposits.
Cerium-144 and promethium-147 can be recovered from low-level aqueous
waste streams by scavenging-precipitation ion exchange. Ruthenium-106
119
-------
can be recovered by volatization and absorption on pyrolusite and
activated carbon. Generally, during the reprocessing of uranium and
Plutonium it is desirable to prevent the volatization of ruthenium.
This can be accomplished during the leaching operation by adding a
reducing agent such as sodium nitrite or nitrogen dioxide to suppress
the ruthenium volatization.
6. APPLICABILITY TO NATIONAL DISPOSAL SITES
Ruthenium-106, cerium-144, and promethium-147 are candidates for a
National Disposal Site. From a cost and safety viewpoint, it would be
desirable to combine the reprocessing plant and the National Disposal
Site. The recommended treatment for the high-level wastes is solidifica-
tion by either the spray or phosphate glass solidification processes,
interim storage in near-surface, engineered facilities, and ultimate
disposal in salt deposits.
120
-------
7. REFERENCES
0559. AEC Manual, Chapter 0524. Standards for radiation protection,
May 12, 1964. 11 p.
0563. Recommendations of the International Commission on Radiological
Protection, ICRP Publication 2. Report on Committee II on
permissible dose for internal radiation (1959). Peryamon Press,
1959. 232 p.
0703. King, L. J., A. Shimozato, and J. M. Holmes. Pilot plant studies
of the decontamination of low-level process waste by a
scavenging-precipitation foam separation process. Oak Ridge
National Laboratory, ORNL-3803, 1968. 57 p.
0705. Staff of the Oak Ridge National Laboratory. Siting of fuel
reprocessing plants and waste management facilities.
ORNL-4451, 1970. 500 p.
0707. Yee, W. C., F. DeLora, and W. E. Shuckley. Low-level radioactive
waste treatment: the water recycle process. Oak Ridge
National Laboratory, ORNL-4472, 1970. 30 p.
0709. Struxnes, R. C. and et. al. Engineering development of hydraulic
fracturing as a method for permanent disposal of radioactive
wastes. Oak Ridge National Laboratory, ORNL-4259, 1968. 261 p.
0714. Touhill, C. J., B. W. Mercer, and A. J. Shuckcrow. Treatment of
waste solidification condensates. Battelle-Northwest,
BNWL-723, 1968. 102 p.
0715. Schneider, K. J. Status of technology in the United States for
solidification of highly radioactive liquid wastes. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p.
0733. Report by the Committee on Radioactive Waste Management. Disposal
of solid radioactive wastes in bedded salt deposits. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p.
0766. Sax, N. I. Dangerous properties of industrial materials. 3d ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
121
-------
REFERENCES (CONTINUED)
0894. Girdler, R. M. Storage of liquid radioactive wastes at the Savannah
River Plant. Du Pont de Nemours and Company, Aiken, South
Carolina, DPSPU-69-30-9, July 1969. 11 p.
1396. Cooke, J. B. and et. al. Report on the proposal of E. I. Du Pont
de Nemours and Company for the permanent storage of radioactive
separation process wastes in bedrock .on the Savannah River,
Du Pont de Nemours (E. I.) and Company, Aiken, South Carolina,
DPST-69-444, May 1970. 57 p.
1423. Morton, R. J. Land burial of solid radioactive wastes: study of
commercial operations and facilities. Environmental and
Sanitary Engineering Branch, Atomic Energy Commission, WASH-1143,
1968. 132 p.
2146. Cohen, J. J., A. E. Lewis, and R. L. Braun. In situ incorporation
of nuclear waste in deep molten silicate rock. Nuclear Technology,
13: 76-87, Apr. 1972.
2147. Harris, D. and J. Epstein. Properties of selected radioisotopes.
Goddard Space Flight Center, Greenbelt, Maryland, NASA SP-7031,
1968. 89 p.
2148. Atomic Energy Commission. Forecast of growth of nuclear power,
Jan. 1971. Washington, U. S. Government Printing Office,
WASH-1139, 1971. 187 p.
2149. Code of Federal Regulations. Title lO—atomic engergy. (Revised
as of Jan. 1, 1971). Washington, U. S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49--transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington, U. S.
Government Printing Office, 1971. 952 p.
2151. Atomic Energy Commission. The nuclear industry, 1971. Washington,
U. S. Government Printing Office, WASH-1174-71, 1971. 193 p.
2666. Recommendations of the National Council on Radiation Protection
and Measurements. Basic radiation protection criteria. NCRP
Report No. 39, NCRP Publications, Washington, 1971.
122
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
'Name
Ruthenium-106
Half-life 368 days
Type of Decay
Molecular Wt.
Density 12.30 gm/cm3
Negative Beta
106 Melting Pt.
Structural Formula
44
Ru
106
2250 C
Boiling Pt. 3900 C
Specific Power 33.1 watts/gm
Solubility
Cold Water
Insoluble
Specific Activity 3.360 curies/gm
Hot Water Insoluble
Others: Insoluble in aqua regia, acid & alcohol; soluble in fused alkali
Decay Chain
106
106
Ru Rh
(368d) -* (30s)
Pd106
(stable)
Radiation Energy Level & Intensities
Ruthenium-106
Beta: 0.0392 Mev (max)
Gamma: None
Rhodium-106
Beta:
Gamma :
55 Mev (90%)
10 Mev ( 3%)
40 Mev ( 5%)
0.512 Mev (21%)
0.622 Mev ( 1%)
1.050 Mev (1.5%)
Shipping Regulations Classified as a transport group III radionuclide
by'the Department of Transportation '
Comments
References: (1) 0766
(2) 2150
123
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Cerium-144 Structural Formula
Half-life 284 days
Type of Decay Negative Beta
.144
Molecular Wt. 144 Melting Pt. 795 C Boiling Pt. 3468 C
Density 6.78 gm/cc Specific Power 25.6 watts/gm
So1ubi1ity Specific Activity 3.180 curies/gm
Cold Water Slightly reactive Hot Water Reacts
Others: Soluble in dilute mineral acid; i
Decay Chain
Ce144 5 Pr144 °I Nd144
(284d) (17.3m) (stable)
nsoluble in
alkali
Radiation Energy Level & Intensities
Cerium- 144
Beta:
Gamma:
0.309
0.175
0.134
0.080
Mev
Mev
Mev
Mev
(76%)
(24%)
(11%)
( 2%)
Praseodymi um-1 44
.
Beta:
Gamma:
2.996
2.186
1.487
0.695
Mev
Mev
Mev
Mev
(max)
(0.7%)
(0.3%)
(1.5%)
Shipping Regulations Classified as
by the Department of Transportation
Classified as a transport group III radionuclide
Comments
References: (1) 0766
(2) 2147
(3) 2150
124
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
61Pm
147
Name Promethium-147 Structural Formula
Half-life 2.62 years
Type of Decay Negative beta
Molecular Wt. 147 Melting Pt. 1035 C Boiling Pt. 2730 C
Density 7.3 gm/cm Specific Power 0.33 watts/gm
Solubility Specific Activity 929 curies/gm
Cold Water ; Hot Water ^
Others:
Decay Chain Radiation Energy Level & Intensities
Pm147 - Sm147
(2.62y) (stable)
Pm147 -* Sm147 Beta: °'224 Mev
Gamma: None
Shipping Regulations Classified as a transport group IV radionuclide by
the Department of TraH5portat16H'
Comments ^ery limited chemical and physical data. Is a member of the
rare-earth group.\
References: (1) 2147
(2) 2150
125
-------
PROFILE REPORT
Strontium-90 (Yttrium-90)
1. GENERAL
Introduction
Strontium-90 is a radioactive isotope principally produced by the
'fission of uranium and plutonium. It has a half-life of 28 years and
emits a 0.546 Mev beta particle. Strontfum-90 produces a short
half-lived daughter, yttrium-90. It has a half-life of 64 hours and
emits a 2.27 Mev beta particle. Yttrium-90, in turn, produces
zirconium-90 which is a stable element.
Strontium-90 has attracted great interest as a public health hazard
since it is the most biologically significant of the radioactive fission
products produced in either nuclear weapon tests or nuclear reactors.
Its biological significance is derived from several factors: (1) it has
a large fission yield (5.9%); (2) it has a long effective half-life; and
(3) it tends to deposit and concentrate in the bone tissue (due to the
fact that strontium is chemically similar to calcium). In the wastes
generated at nuclear power plants strontium-90 and its daughter yttrium-90
are responsible for approximately 7 percent and 38 percent of the total
fission product activity in the wastes after 1 year and 10 years,
respectively*
The projected growth in the production of strontium-90 will parallel
that of the civilian nuclear power program. Present projections indicate
that nuclear power will account for 30 percent of the total power
production by the year 1980. Projected annual production figures
from the civilian nuclear power program for stront1um-90 to the year 2020
127
-------
for light-water and fast-breeder reactor fuels are attached (Table 1). The
expected total accumulated radioactivity for strontium-90 from the year 1970
to 2020 in millions of curies is:
Radioactivity
Year (Megacuries)
1970 4
1980 962
1990 4,640
2000 9,550
2020 29,400
These figures along with the annual production figures illustrate the
tremendous growth anticipated in the production of this isotope. Annual
production figures or accumulated totals for strontium-90 produced as
part of the nuclear weapons program were not available.
Manufacture
Strontium-90 is produced by the fission of the heavy elements, i.e.,
those elements heavier than lead. The fission reaction results from
neutron bombardment or spontaneous decomposition of the nucleus. A
typical fission reaction is:
Presently, strontium-90 is produced in thermal power reactors .by the
fission of uranium. Civilian nuclear power plants are the major
producers and a smaller amount is produced at AEC facilities. At the
present time there are 22 nuclear power plants in operation with an
additional 104 being built or planned.2151 In the future,,strontium-90
will also be produced in fast-breeder reactors by the fission of
Plutonium.
128
-------
TABLE 1
STRONTIUM-90 CONTENT IN HIGH-LEVEL WASTE PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM1
,0705
Light-Water Reactor Fuels
Annual
Production
grams/year
curies/year
watts/year
^HMHM^M^MH^HM^^MMM^H^M^B^
Annual
Production
grams/year
caries/year
watts/year
Calendar Year
1970 1980 1985 1990
2.83xl04 161xl04 291xl04 328xl04
4.0xl06 228x1 O6 41 Oxl O6 463x1 O6
.52xl04 29.6xl04 53.3xlC4 60.30xl04
Fast-Breeder Reactor Fuel
Calendar Year
1985 1990
llxlO4 66xl04
15.5xl06 93.4xl06
2.0xl04 12.2xl04
2000
260x1 O4
367x1 O6
47.8xl04
s
2000
282x1 O4
399x1 O6
52x1 O4
2020
764x1 O4
lOSOxlO6
140.5xl04
2020
846x1 O4
1 200x1 O6
156xl04
-------
Uses
Strontium-90 is a long-lived high-energy beta emitter. It is
used in military and space applications as an energy source for
small auxiliary power units. It is also used in the radioactive
gauge field. Beta gauges activated by strontium-90 are extensively
used in industry to measure and control the thickness or density of thin
materials (paper, textiles, etc.) or coatings. In the field of nuclear
medicine strontium-90 is used as a source of beta radiation for the
treatment of skin diseases and eye disorders and also as a medical blood
irradiator. At the present time the strontium-90 used in the above
applications is obtained from the waste streams in the Richland, Washington,
chemical separations plant and is distributed through the Oak Ridge
National Laboratory, Isotopes Sales Department. In the future, when
its production is warranted by market demands, it could be available in
large quantities from commercial firms operating chemical processing
plants. The AEC revenue from the distribution of strontium-90 through
the fiscal years 1965 to 1971 is2151: -
Year Revenue (Dollars)
1965 3,000
1966 65,000
1967 59,000
1968 188,000
1969 169,000
1970 72,000
1971 248,000
These figures illustrate the relatively small commercial market for
the use of strontium-90 at the present.
Sources and Types of Strontium-90 Wastes
Although produced in nuclear power reactors the primary source of
strontium-90 is in the high-level aqueous waste streams generated
at the spent fuel processing plants. The range of chemical compositions
130
-------
of the various types of waste streams obtained from the processing step
have been tabulated, and in all cases are primarily aqueous solutions
of inorganic nitrate salts. Strontium-90 is normally found in the waste
stream in the form of a nitrate or oxide. The characteristics of strontium-
90 products in the high-level wastes at two different time periods is
attached (Table 2).
Strontium-90 is also found in the low-level waste streams generated
at spent fuel processing facilities. The strontium-90 activity in
these waste streams is from a few hundredths to several tenths of a curie
per gallon of waste. Strontium-90 is also found in the wastes resulting
from research laboratories and medical and industrial applications.
Generally the strontium-90 waste from these sources is found in the
solid form.
Physical and Chemical Properties
The physical and chemical properties of strontium-90 and its short
half-lived daughter yttrium-90 are included in the attached worksheet.
Strontium-90 is a member of the alkaline-earth metal group. It is a
highly metallic and highly electropositive element with a valence of
+2. It dissolves readily in acids and will burn when heated in air,
oxygen, or carbon dioxide. At low temperatures, oxidation is slow
owing to the formation of a protective oxide film. Strontium-90 reacts
readily with water, releasing hydrogen and forming metal and hydroxyl
ions. The carbonates and sulfates of strontium are very insoluble.
The fluorides and the oxalate of strontium are also insoluble. The
sulfides, however, are soluble in water.
2. RADIATION HAZARD
Strontium-90 is a very dangerous radioactive material. It tends to
concentrate in the bone and irradiate the adjacent soft tissue. The
symptoms of strontium-90 radiation exposure are nausea and fatigue
followed by vomiting and diarrhea. The possible types of injury include:
131
-------
TABLE. 2
STRONTIUM-90 CONTENT IN HIGH-LEVEL FUEL REPROCESSING WASTESU/UO
Light-Water Reactor Fuels
'(Fuel Exposed to 33,000 MWD/MTU at 30 MW/MTU)
Fission After 1 _year After 10 years
h* Products Grapis/Tonne Watts./Tonne Curies/Tonne Grams/Tonne '.4atts /Tonne Curies/Tonne
W . ""^"™"™~""^™""
JO Sr90 _ y90 53a 53Q 151,000 427 425 120,900
Other Fission
Products 34,566 9,470 2,071,000 34,673 605 196,100
Fast^Breeder Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 58 MW/MTU)
Fission After 1 year • . After 10 years
Products Grams/Tonne Watts/Tonne Curies/Tonne Grams/Tonne I'atts/Tonne Curies/Tonne
Sr90 - Y90 300 298 85,000 241 239 68,000
Other Fission
Products 34,600 13,§Q2 3,345,000 34,658 52? 213,000
-------
leukopenia, anemia, leukemia, cataracts, and increases in the average
rate of genetic mutation. The effects of strontium-90 radiation exposure
are primarily dependent on the amount of radiation and the portion of
the body affected. Standards for prolonged exposure over a 50-year period
have defined the single dose limit in terms of the maximum permissible
dose accumulated in a period of 13 weeks. The whole body exposure limit is
3 rem per quarter for a radiation worker and the accumulated dose limit is
5(N - 18), where N is the individual's age in years. Limits for the
thyroid, bone, and other organs have also been defined (Table 3}.0563
Following the inhalation of 1 microcurie of strontium-90 in the
insoluble form the critical organ is the bone and the dose delivered to
the .bone following the inhalation of 1 microcurie of strontium-90 is
56 rem. The dose delivered to the bone following injection of 1 microcurie
into the body via a wound is 90 rem. Values of the maximum permissible
total body burden of strontium-90 which are deposited in the total body
and produces the maximum permissible dose rate to a particular body
organ have been compiled. The maximum permissible body burden for
the bone is 2.0 microcuries and for the total body 20 microcuries. The
above body burden rates can be expressed in terms of the maximum
permissible concentrations of strontium-90 in air and water. These
concentrations can be expressed in terms of a daily intake of air and
water at the continuous exposure rate of 168 hours per week for a period
of 50 years. For strontium-90 the critical organ is the bone and the
maximum permissible concentration in water is lxlO~ microcuries per
milliliter and in air 1x10 microcuries per milliliter.
3. OTHER HAZARDS
Strontium-90 resembles calcium in its metabolism and behavior.
Strontium-90, besides its radiation hazard, is moderately toxic to the
skin or mucous membranes following a single exposure. Its fire and
explosive hazard is moderate, in the form of dust, when exposed to flame
or chemical reaction. 6 It reacts with water to evolve hydrogen.
Strontium nitrate or peroxide are powerful oxiding materials. In
133
-------
TABLE 3
RADIATION PROTECTION STANDARDS FOR INDIVIDUALS AND POPULATION GROUPS
FOP EXTERNAL AMD INTERNAL EXPOSURE 2149
Type of Exposure Dose to Individuals Average Dose to a
at Points of Maximum Suitable Population
Probable Exposure Sample
(rem per year) (rem per year)
Whole body, gonads, or 0.5 0.17
bone marrov;
Thyroid or bone 1.5 0.5
Bone (alternate standards) Body burden at 0.003 Body burden of 0.001
micrograms of radium micrograms of radium
226 or tts biological 226 or its biological
equivalent eoutvalent
-------
contact with easily oxidizable substances they react rapidly to cause
ignition or explosion by friction or on contact with a small amount
.of water.
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling, Storage, and Transportation
Since strontium-90 is hazardous to man by inhalation, ingestion, or
direct radiation exposure, care is exercised in its handling. Special
procedures and radiation shielding are required in the handling of
strontium-90. The beta particles emitted by strontium-90 generally
cannot penetrate more than one centimeter of metal or glass. The
principal problem associated with the beta radiation is the bremsstrahlung
effect which results from the presence of highly penetrating rays
resulting from the deflection of beta particles by the nuclei in the
shielding medium. Since the bremsstrahlung effect increases with atomic
number, shielding materials of high atomic weight are not satisfactory.
Lead and high-density concrete are commonly used shielding materials for
strontium-90 radiation. To detect and control personnel exposure to
strontium-90 radiation all persons working with this material should
wear dosimetry devices which directly indicate the dose. Commonly used
devices are the film badge and the thermoluminescent dosimeters (TLD).
Strontium-90 is stored in controlled reservations in specially
constructed containers. The strontium-90 is protected by both a
primary and a secondary containment barrier. Special monitoring systems
and proper warning signs are located in the general area of the storage
facility.
Strontium-90 is classified as a transport group II radionuclide by
the Department of Transportation, and the>rules and regulations governing
their transportation are given in the Code of Federal Regulations (CFR)
Title 49—Transportation, Parts 170 to 190. 215° The strontium-90
content is limited to 0.05 curies for a Type A package and 20 curies for
135
-------
a Type B package defined in 49CFR173. The limits are increased to 20
and 500 curies if the physical form of the strontium-90 meets the
requirements of a special form material. The strontium-90 release rate
is limited to zero under the specified accident conditions for Type A
and B quantities. The allowable release of radioactivity from
packages containing large quantities of strontium-90 is limited to
gases and contaminated coolant containing total radioactivity exceeding
neither 0.1 percent of the total radioactivity of the package nor 0.5
curies under the hypothetical accident conditions prescribed in 49CFR173.
Disposal/Reuse
The disposal of strontium-90 is governed by the AEC Manual
O'iRQ 2149
Chapter 0524UODy and 10CFR20. Two sets of standards have been
established for the permissible radiation exposure in unrestricted
areas. One is for the greatest dose received by an individual and the
other for the average dose received by the general population. The
radiation protection standards for these two groups is attached. The
standards for strontium-90 also define its maximum concentrations in
air and water in microcuries per milliliter (yc/ml) listed below'
2149.
Exposure
Group
Individual
Individual
Population
Population
Strontium-90
Form
Soluble
Insoluble
Soluble
Insoluble
Concentration
in Air
3x1O"11 yc/ml
2x1O"10 yc/ml
IxlO"11
.67xlO"10 yc/ml
Concentration
in Water
3xlO"7 yc/ml
4x10~5 yc/ml
IxlO"7 yc/ml
1.3xlO"5 yc/ml
Although rarely utilized, the disposal of strontium-90 by release
into a sanitary sewage system is limited to 1 microcurie or a concen-
-5
tration of 1x10 microcuries per milliliter of water in the soluble form
_3
or 1x10 microcuries per milliliter of water in the insoluble form. The
disposal of strontium-90 by burial in the soil at any one location and
time is limited to 100 microcuries of strontium-90.
136
-------
5. EVALUATION OF WASTE MANAGEMENT PRACTICES
Recovery
Option No. 1 - Precipitation. Strontium-90 can be removed from high-
level processing wastes following the spent fuel processing step by
precipitating as a carbonate or fluoride or absorbing on alumino-
silicate zeolites. The standard procedure for the recovery of
strontium-90 is to precipitate as a carbonate and then add titanium
dioxide and heat to drive off the C02> forming strontium titanite. The
strontium-90 can be recovered from the other high-level wastes for
commercial purposes, or to provide isolation of the strontium-90 for
heat transfer and safety reasons (see Radioactive Waste Solidification
Report).
Option No. 2 - Scavenging-Precipitation Foam Separation. The removal
of strontium-90 from radioactive waste water by scavenging-precipitation
foam separation has been studied by Oak Ridge National Laboratory.0703
The process consists of two steps: (1) precipitating in a sludge
blanket clarification step, and (2) achieving final decontamination in
a foam-separation column. Processing rates of 300 gal. per hour (gph)
for scavenging precipitation and 120 gph for foam separation have been
achieved. Strontium-90 is precipitated by the addition of calcium
carbonate or calcium phosphate. Strontium-90 decontamination factors
(ratio of initial to final concentration) of 1050 have been obtained.0703
At the present time, further development is required to reduce
operating costs and increase processing rates.
Option No. 3 - Scavenging-Precipitation Ion Exchange. In this process
final decontamination is obtained by ion exchange columns from the
scavenging-precipitation process. In the phenolic ion exchange column
calcium and magnesium are removed with the strontium. The process
includes a provision for the recycle of the ion exchange waste to the
scavenging-precipitation step. All the removed radionuclides are
concentrated in the clarifier sludge. The overall decontamination
137
-------
factors for strontium-90 varied from 1,200 to 12,000.°705 The use of
ion exchange resins with a scavenging-precipitation step is a fairly
simple and efficient means of removing strontium-90 from low-level
radioactive aqueous wastes.
Option No. 4 - Water Recycle. The water recycle process is used for
decontaminating radioactive waste water and recycling the purified
water for reuse. This process has been demonstrated at the pilot
plant scale by Oak Ridge National Laboratory. 7 The steps in the
process include: (1) clarification by controlled addition of
coagulants; (2) demineralization by cation-anion exchange; and
(3) sorption on granular activated carbon. The majority of the
strontium-90 is removed by the cation-anion exchange. The strontium-90
overall decontamination factor was 5,700. The method is an improvement
to the treat-and-discharge methods, but further work is required until
full-scale production use is obtained.
Storage/Disposal
Option No. 1 - Land Burial. Land burial of strontium-90 wastes, in small
concentrations, at approved sites that are acceptable from a geologic
and hydrologic standpoint, is an acceptable means of disposal.
Strontium-90 concentration should not be in excess of 104 times the
maximum permissible concentration for the general population in
pi nn r r
10CFR20. All strontium-90 wastes to be disposed of should be
in a solid form and packaged in a suitable container. Liquid
strontium-90 wastes should be solidified, preferably using asphalt,
in accordance with the methods described in the radioactive waste
solidification report. The burial trenches should be designed not
to intercept the ground water table and constructed with a bottom
drain and sump for water monitoring. The trenches should be covered
with either asphalt or vegetation to limit infiltration of water.
The burial site design, geology, and hydrology should be in conformance
with the criteria used in selecting and licensing the present
commercial burial sites.1423 Since land burial has successfully been
practiced on a commercial scale, it should be considered as the most
satisfactory method of disposing of all dilute concentrations of
strontium-90 wastes.
138
-------
Option No. 2 - Near-Surface Liquid Storage. Near-surface storage of
reactor-produced aqueous solutions of strontium-90 and other fission
products salts which are stored in carbon steel or stainless steel tanks
encased in concrete and buried underground is not considered as a
satisfactory means of disposal. Aqueous solutions of high-level wastes
have been stored in this manner over the past 25 years. The tanks range
in size from 0.33 to 1.3 million gal. and are generally equipped with
devices for measuring temperatures, liquid levels, and leaks. These
tanks are considered as an interim storage technique due to a general
lack of confidence in their long-term integrity. Since present
regulations require the solidification of all reactor wastes within
5 years following reprocessing, the near-surface storage of aqueous
solutions of strontium-90 and other fission product salts in steel tanks
should only be considered as a near-term storage technique and not as a
permanent storage or disposal technique.
Option No. 3 - Near-Surface Solid Storage. The storage of solidified,
high-level wastes containing strontium-90 and other waste salts in
engineered storage facilities offers the best intermediate method for
storage of these wastes. The technology for these facilities has been
developed. The wastes will be under surveillance and control and
can be retrieved, should this be required. The high-level wastes from
fuel reprocessing should be solidified and packaged in a suitable
container (steel). Of the four high-level solidification processes
developed, spray or phosphate glass solidification offer the best
solidified waste characteristics (see Radioactive Waste Solidification
report). The wastes will be stored in stainless steel-lined concrete
vaults which will be either air or water cooled.
Option No. 4 - Salt Deposits. This method offers the best potential
for the disposal of wastes from fuel reprocessing since bedded salt
deposits are completely free of circulating ground waters. This method
of disposal has been under study by the Oak Ridge National Laboratory
since 1957, and in November of 1970 a committee of the National Academy
of Sciences recommended that the use of bedded salt for the disposal of
139
-------
radioactive wastes is satisfactory. Recent questions concerning
the adequacy of this method have resulted in the need for further
development work before it can be accepted as an ultimate method of
disposal. The wastes containing strontium-90 must be solidified and
disposed of in the solid form and packaged in a suitable container. The
solidified wastes are then buried in rooms carved in the salt deposits
approximately 1,000 ft below the ground. The salt is a good heat
transmitter, provides about the same radioactive shielding as concrete,
and can heal its own fractures by plastic flow. The critical problem
is the selection of a site that meets the necessary design and
geological criteria for the mixture of fission product and actinide wastes.
Option No. 5 - Bedrock Disposal. The disposal of low-level aqueous
solutions of strontium-90 wastes along with other spent fuel reprocessing
wastes in vaults excavated in crystalline rock over 1,500 ft beneath
the ground is currently being evaluated by E. I. du Pont de Nemours, at
their Savannah River Plant near Aiken, South Carolina.0894'1396 The wastes
would be stored in six tunnels and once in the tunnels the wastes will seep
into the surrounding rock. Located above the crystalline rock is the
Tuscaloosa formation, a good source of freshwater, which is separated
by a layer of clay that would act as a barrier to the leakage of
radioactive wastes. An advisory committee appointed by the National
Academy of Sciences recommended abandonment of the project. In
May 1972, another National Academy of Sciences panel concluded that
bedrock storage provides a reasonable prospect for long-term safe
storage, but precise information is needed to decide if and where
underground storage vaults should be built. Another method has also
been proposed for disposing of liquid wastes by in situ incorporation
in molten silicate rock. At the present time both of these methods
are unproven since sufficient engineering data or exploration has not
been completed to verify their suitability. Due to the long half-lives
of certain of the elements of the mixed fission products and actinides
and high biological hazard, it is.doubtful that data could ever be
accumulated to prove that the geological characteristics of the site over
the next few hundred years are acceptable to ensure the absolute safety
of such a disposal method.
140
-------
Option No.6 - Hydraulic Fracturing. The direct disposal of aqueous,
low-level radioactive wastes into shale formations has been
investigated by Oak Ridge National Laboratory.0 The method
consists of mixing the aqueous wastes with cement and pumping the
resulting slurry down a well and out into a nearly horizontal fracture in
a thick shale formation. Since strontium-90 is leached from cement and
no additional encapsulation is provided and since it has long half-life
this method of disposal is not recommended.
To summarize, strontium-90 can be recovered from the high-level
waste streams from the spent fuel processing facilities for separate
disposal or reuse. The separation of strontium-90 from these waste
streams will probably be required with the introduction of the fast-
breeder reactor due to its high heat content (see Radioactive Waste
Solidification report). The acceptable method of treatment is spray
or phosphate glass solidification, followed by storage in near-surface
engineered facilities. Finally, these wastes should be disposed of in
salt deposits. Strontium-90 should be recovered from the low-level
waste streams to minimize the amount directly released to the environ-
ment. An adequate method of recovery is scavenging-precipitation ion
exchange. The recovered strontium-90 can then be solidified, preferably
using asphalt, and disposed of at approved sites by land burial.
6. APPLICABILITY TO NATIONAL DISPOSAL SITES
Strontium-90 is a candidate for a National Disposal Site due to its
health hazard and its projected growth with that of the civilian nuclear
power program. From a cost and safety viewpoint, it would be desirable
to combine the reprocessing plant and the National Disposal Site. Interim
storage in near-surface engineered facilities offers considerable latitude
in the storage site selection. The site selection for ultimate disposal
of these wastes is limited to particular geological areas in which salt
deposits are present.
141
-------
The recommended treatment for strontium-90 in low-level waste
streams is recovery by scavenging-precipitation ion exchange followed
by solidification with asphalt and disposal by land burial. For
the high-level, strontium-90 wastes, the recommended processes are
recovery by either spray or phosphate glass solidification, interim
storage of the solidified waste in near-surface engineered storage
facilities, and disposal in salt beds.
142
-------
7. REFERENCES
0559. AEC Manual, Chapter 0524. Standards for radiation protection
May 12, 1964. 11 p.
0563. Recommendations of the International Commission on Radiological
Protection, ICRP Publication 2. Report on Committee II on
permissible dose for internal radiation (1959). Pergamon
Press, 1959. 232 p.
0686. LaRiviere, J. R. Packaging and storing radioactive wastes.
Chemical Engineering Progress. 66: 42-44, Feb. 1970.
0703. King, L. J., A. Shimozato, and J. M. Holmes. Pilot plant studies
of the decontamination of low-level process waste by a
scavenging-precipitation foam separation process. Oak Ridge
National Laboratory, ORNL-3803, 1968. 57 p.
0705. Staff of the Oak Ridge National Laboratory. Siting of fuel
reprocessing plants and waste management facilities.
ORNL-4451, 1970. 500 p.
0707. Yee, W. C., F. DeLora, and W. E. Shuckley. Low-level radioactive
waste treatment: the water recycle process. Oak Ridge
National Laboratory, ORNL-4472, 1970. 30 p.
0709. Struxnes, R. C. and et. al. Engineering development of hydraulic
fracturing as a method for permanent disposal of radioactive
wastes. Oak Ridge National Laboratory, ORNL-4259, 1968. 261 p.
0715. Schneider, K. J. Status of technology in the United States for
solidification of highly radioactive liquid wastes. Battelle
Northwest, BNWL-820, 1968. 63 p.
0733. Report by the Committee on Radioactive Waste Management. Disposal
of solid radioactive wastes in bedded salt deposits. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p.
0738. Regan, W. H., ed. Proceedings: the solidification and long-term
storage of highly radioactive wastes. Richland, Washington,
Feb. 14-18, 1966. Atomic Energy Commission, CONF-660208. 877 p.
0766. Sax, N. I. Dangerous properties of industrial materials. 3d ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
143
-------
REFERENCES (CONTINUED)
0894. Girdler, R. M. Storage of liquid radioactive wastes at the
Savannah River Plant. Du Pont de Nemours and Company,
Aiken, South Carolina, DPSPU-69-30-9, July 1969. 11 p.
1396. Cooke, J. B. and et. al. Report on the proposal of
E. I. Du Pont de Nemours and Company for the permanent
storage of radioactive separation process wastes in bedrock
on the Savannah River, Du Pont de Nemours (E. I.) and Company,
Aiken, South Carolina, DPST-69-444, May 1970. 57 p.
1423. Morton, R. J. Land burial of solid radioactive wastes: study
of commercial operations and facilities. Environmental and
Sanitary Engineering Branch, Atomic Energy Commission,
WASH-1143, 1968. 132 p.
2146. Cohen, J.- J., A. E. Lewis, and R. L. Braun. In situ incorporation
of nuclear waste in deep molten silicate rock. Nuclear Technology,
13: 76-87, Apr. 1972.
2147. Harris, D. and J. Epstein. Properties of selected radioisotopes.
Goddard Space Flight Center, Greenbelt, Maryland,
NASA SP-7031, 1968. 89 p.
2148. Atomic Engergy Commission. Forecast of growth of nuclear power,
Jan. 1971. Washington, U. S. Government Printing Office,
WASH-1139, 1971. 187 p.
2149. Code of Federal Regulations. Title lO—atomic energy. (Revised
as of Jan. 1, 1971). Washington, U. S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49--transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington, U. S.
Government Printing Office, 1971. 952 p.
2151. Atomic Energy Commission. The nuclear industry, 1971. Washington,
U. S. Government Printing Office, WASH-1174-71, 1971. 193 p.
144
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name Strontlum-90 Structurfl1
Half-life 28 years 90
7ft
Type of Decay Negative Beta J0
Molecular Wt. 90 _ Melting Pt. 769C Boiling Pt. 1384C
Density 2.6 gm/cc _ Specific Power 0.98 watts /qm _
Solubility Specific Activity 142 curies/gm
Cold Water Reacts _ Hot Water Reacts _ ^
Others: Soluble in acid, alcohol, and liquid NHj _
Decay Chain Radiation Energy Level & Intensities
Beta: 0.549 Mev (100%)
(28y) - - (64h)
Sr90 B Y90 Gamma: None
Shipping Regulations Classified as a transport group II radionuclide
by the Department of Transportation
Comments
References: (1) 0766
(2) 2147
(3) 2150
145
-------
HAZARDOUS WASTES PROPERTIES
HORKSHEET
Name
Yttrium-90
Half-life 64 hours
Type of Decay Negative Beta
Molecular Wt. 90 Melting Pt.
Density 4.34
Structural Formula
,90
39
1495C
Boiling Pt. 2927C
Specific Power
Solubility^
Cold Water Slightly reactive
Others: Very soluble in dilute acid; soluble in hot KOH
Specific Activity 5.3x10 curies/qm
Hot Water Reacts
Decay Chain
,90
(64h)
Z.r90
(Stable)
Radiation Energy Level & Intensities
Beta: 2.27 Mev(max)
Gamma: None
Shipping Regulations Classified as a transport group IV
radionuclide by the Department of Transportation
Comments
References: (1) 0766
(2) 2150
146
-------
PROFILE REPORT
Zirconium-95. Niobium-95
1. GENERAL
Zirconium-95 is a radioactive isotope principally produced in nuclear
reactors by the fission of uranium and plutonium. Zirconium-95 has a
half-life of 65 days and emits both beta particles and gamma rays.
Zirconium-95 produces, by beta decay, niobium-95. Niobium-95 has a
half-life of 35 days, emits a 0.160 Mev beta particle and a 0.765 Mev
gamma ray. Niobium-95 is transformed by beta decay into molybdenum-95
which is a stable element.
Zirconium-95 and niobium-95 are both short-lived isotopes and account
for 25 percent of the total activity in spent fuel processing wastes after
90 days. Their activity decreases to less than 1 percent of the total after
1 year. They also present considerable difficulty in the separation
of uranium and plutonium from these elements since they both form radioactive
colloids in solution. They also tend to be absorbed on surfaces
such as container walls.
The projected growth in zirconium-95 and niobium-95 will parallel that
of the civilian nuclear power program. Present projections indicate that
nuclear power will account for 30 percent of the total power production
by the year I960.2 Projected annual production figures from the
civilian nuclear power program for these two isotopes to the year 2020
for light-water and fast-breeder reactor fuels are attached (Tables 1 and
2). These figures illustrate the tremendous growth anticipated in their
production.
147
-------
00
TABLE 1
ZIRCONIUM-95 CONTENT IN HIGH-LEVEL WASTES PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM'
Light-Hater Reactor Fuels
,0705
Annual
Production
grams/year
curies/year
watts/year
Annual
Production
grams/year
curies/year
watts/year
Calendar Year
1970 1980 1°85
.0128xl04 7.31X104 13.2xl04
.272xl08 15.5xl08 27.9xl08
,142xl06 S.llxlO6 14.6xl06
Fast-Breeder Reactor
Calendar Year
1970 1980 1985
3.54xl04
7. 50x1 O8
3.93xl06
199,0
14.9xl04
31.6xl08
16.5xl06
Fuels
1990
21.3xl04
45.2xl08
23.7xl06
n
n.sxio4
25. Oxl O8
13.1xl06
2000
"gi.ixio4
193xl08
lOlxlO6
20?n
39. 7x1 O4
73. 6x1 O8
38.5xl06
2020
274x1 O4
580x1 O8
304x1 O6
-------
C£
TABLE 2
NIOBIUM-95 CONTENT IN HIGH-LEVEL WASTE PRODUCED BY THE CIVILIAN NUCLEAR POWER PROGRAM
Light-Water Reactor Fuels
,0705
Annual
Production
grams /year
curies/year
watts/year
Annual
Production
grams/year
curies/year
watts/year
Calendar Year
1970 1980 1985
.012xl04 6.54xl04 11.8xl04
.45xl08 25. 7x1 O8 46. 3x1 O8
.22xl06 12.3xl06 22.2xl06
Fast-Breeder Reactor
Calendar Year
1970 1980 1985
2. 42x1 O4
9.50xl08
4.56xl06
1990
13.3xl04
52. 4x1 O8
25.1xl06
Fuels
1990
14.6xl04
57.2xl08
27.5xl06
2000
10.6xl04
41. 5x1 O8
19.9xl06
2000
62. 2x1 O4
244x1 O8
118xl06
2020 •
Sl.lxlO4
122xl08
58.6xl06
2020
187xl04
734x1 O8
353x1 O6
-------
The primary sources of zirconium-95 and niobium~95 wastes are the
spent fuel processing plants. These isotopes and the other resultant
high-level radioactive wastes are contained in the aqueous effluent from
the spent fuel processing step. The range of chemical compositions of
the various types of waste streams obtained from the spent fuel processing
step have been tabulated. The activity of zirconium-95 and niobium-95
in the spent reactor wastes at two different times is°attached (Tables
3 and 4). Due to their short half-lives, their activity after 10 years
is reduced to zero. They are also found in very small quantities in the
low-level radioactive waste streams generated at spent fuel processing
facilities.
The physical and chemical properties of zirconium-95 and niobium-95
are included in the attached worksheets. Zirconium-95 is a member of the
group IV series of metals, all of which have a characteristic +4
oxidation state. Zirconium-95 is not readily soluble in nitric acid but
dissolves at substantial rates in both hydrofluoric and moderately
concentrated sulfuric acid. Niobium-95 exists in the +5 oxidation state
in a majority of its compounds. Niobium-95 is a passive metal and does
not dissolve in most acids when pure, but is attacked by these acids if
impurities are present. Both isotopes tend to form radioactive colloids
in solution which do not act as true solutes but tend to be absorbed
on surfaces.
2. RADIATION HAZARD
Zirconium-95 and niobium-95 both are moderately dangerous
radioactive materials. Zirconium-95 is classified as a slightly
more hazardous material than niobium-95. The effects of their
radiation are primarily dependent on the amount of radiation and
the portion of the body affected. The effects of acute whole-body
gamma radiation exposure are: (1) 5 to 25 rads, minimal dose detectable
by chromosome analysis or other specialized analyses, but not by hemogram;
(2) 50 to 75 rads, minimal acute dose readily detectable in a specific
individual (e.g., one who presents himself as a possible exposure case);
150
-------
TABLE 3
ZIRCONIUM-95 CONTENT IN HIGH-LEVEL FUEL REPROCESSING WASTES
Light-Water Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at 30 MW/MTU)
0705
Fission
Products Grans/Tonne
Zr95 1.3
Other Fission 35,099
Products
Fission
Products Grams/Tonne
Zr95 630
Other Fission 34,897
Products
After 1 year
Watts/Tonne
146
9,854
(Fuel
After 1 year
Ha tts /Tonne
310
13,490
Curies/Tonne
28,000
2,194,000
Grans/Tonne
0
35,100
Fa?t-Breeder Reactor Fuels
Exposed to 33,000 MWD/MTU at
Curies/Tonne
59,000
3,371,000
Grams/Tonne
0
34,900
After lOj^ears
I'atts/Tonne Curies/Tonne
0 0
1030 317,000
58 MW/MTU)
After 10 years
Watts/Tonne Curies/Tonne
0 0
776 281,000 "
-------
tn
to
TABLE 4
NIOBIUM-95 CONTENT IN HIGH-LEVEL FUEL REPROCESSING WASTES
0705
Light-Water Reactor Fuels
jFuel Exposed to 33.fcQOO-'M1KD/MTU at 30 -MW/MTO) •
Fission After 1
year
Products Grans/Tonne tlatts/Tnnne Curies/Tonne Grans/Tonne
OK
NbyD 2 285
Other Fission 35,098 9,715
Products
Fission After
59,000 0
2,163,000 35,100
Fast-Breeder Reactor Fuels
(Fuel Exposed to 33,000 MWD/MTU at
year
Products. Grams/Tonne Uatts/Tonne Curies/Tonne Grams/Tonne
95
Nb 3 600 125,000 0
Other Fission 34,897 -13,200 . 3,305,000. 34,900
Products -
After 10 j/ears
'..'-at ts/Tonne
0
1030
58 MW/MTU)
After 10 years
l.'atts/Tonne
0
776
Curies-'Tonne
0
317,000
Curies/Tonne
0
281 ,000
-------
(3) 75 to 125 rads, minimal acute dose likely to produce vomiting 1n
about 10 percent of people so exposed; (4) 150 to 200 rads, acute dose
likely to produce transient disability and clear hematological changes
in a majority of people so exposed; (5) 300 rads, median lethal dose
Pfififi
for single short exposure. These effects are for a single large dose
of radiation or a series of substantial doses in a short interval of
time to the total body. Standards for prolonged exposure over a 50-
year period have defined the single dose limit in terms of the maximum
permissible dose accumulated in a period of 13 weeks. The whole body
exposure limit is 3 rem per quarter for a radiation worker and the
accumulated dose limit is 5(N - 18), where N is the individual's age in
years. Limits for the thyroid, bone, and other organs have been
defined (Table 5).0563
The zirconium-95 dose in the insoluble form to the lung is 0.22 rem
following the inhalation of 1 microcurie. In the soluble form the dose
delivered to the bone following the inhalation of 1 microcurie of
zirconium-95 is 0.06 rem. The dose delivered to the bone following
injection of 1 microcurie into the body via a wound is 0.22 rem. For
niobium-95 the above doses are reduced by a factor of three to four.
Values of the maximum permissible total body burden of zirconium-95
and niobium-95 which are deposited in the total' body and produce the
maximum permissible, dose rate to a particular body organ have been
compiled. °563 For both isotopes the critical organ is the bone and '
the maximum permissible body burden is 20 microcuries for zirconium-95
and 40 microcuries for niobium-9,5.
3. OTHER HAZARDS
Besides its radiation hazard, zirconium-95 is not significantly
poisonous, and so far as is known, the inherent toxicity of zirconium
compounds is low.0776 The toxicity of niobium is unknown.0776
Zirconium fire and explosive hazard is high, especially in the form of
dust when exposed to chemical reaction with air or oxidizing agents.
The fire and explosive hazard of niobium is low.
153
-------
TABLE 5
RADIATION PROTECTION STANDARDS FOR INDIVIDUALS AND POPULATION GROUPS
214Q
FOP. EXTERNAL AND INTERNAL EXPOSURE
Tyne of Exposure'Dose to IndividualsAverage Dose to a
at Points of Maximumn Suitable Population
Probably Exposure Sample
^ . (rem per year) frem per year)
CA
Whole body, gonads, or 0.5 0.17
bone marrov/
Thyroid or bone 1.5 0.5
Bone (alternate standards) Body burden at 0.003 Body burden of 0.001
micrograms of radium micrograms of radium
226 or tts biological 226 or its biological
equivalent eoutvalent
-------
4. DEFINITION OF ADEQUATE WASTE MANAGEMENT
Handling, Storage, and Transportation
i
Since zirconium-95 and m'obium-95 are hazardous to man by inhalation,
ingestion, or direct radiation exposure, care is exercised in their
handling. Special procedures and radiation shielding are utilized in
their handling. The beta particles emitted by them generally cannot
penetrate more than 0.1 in. of water or 0.04 in. of glass. Their gamma
rays are highly penetrating. Lead and concrete are commonly-used
shielding materials. To detect and control personnel exposure, all
persons working with these materials should wear dosimetry devices which
directly indicate the dose. Commonly used devices are the film
badge and the thermoluminescent dosimeters (TLD).
Zirconium-95 and niobium-95 are stored in controlled areas in
specially-constructed containers. They are protected by both a primary
and a secondary containment barrier. Special monitoring systems and
proper warning signs are located in the general area of the storage
facility.
Zirconium-95 is classified as a transport group III and niobium-95
as a transport group IV radionuclide by the Department of Transportation,
and the rules and regulations governing their transportation are given
in the Code of Federal Regulations (CFR) Title 49--Transportation,
Parts 170 to 190.2150 The zirconium-95 content is limited to 3 curies
for a Type A package and 200 curies for a Type B package defined in
49CFR173. For niobium-95 the limits are 20 curies for a Type A package
and 200 curies for a Type B package. The limits are increased to 20
and 5,000 curies if their physical form meets the requirements of a
special form material. Their release rate is limited to zero under
the specified accident conditions for Type A and B quantities.2150
The allowable release of radioactivity from packages containing large
quantities of these isotopes is limited to gases and contaminated coolant
155
-------
containing total radioactivity exceeding neither 0.1 percent of the
total radioactivity of the package nor 10 curies under the hypothetical
accident conditions prescribed in 49CFR173.
Disposal/Reuse
The disposal of zirconium-95 and niobium-95 is governed by the AEC
0559 2149
Manual Chapter 0524 and 10CFR20. Two sets of standards have
been established for the permissible radiation exposure in unrestricted
areas. One is for the greatest dose received by an individual and the other
for the average dose received by the general population. The radiation
protection standards for these two groups is attached (Table 5). The
standards for the safe disposal of zirconium-95 and niobium-95 also define
their maximum concentrations in air and water. Their concentrations in
an unrestricted area should not exceed the concentrations in microcuries
per milliliter included in this report (Table 6).
Although rarely practiced, the disposal of zirconium-95 and
niobium-95 into a sanitary sewage system is limited to 100 microcuries.
Their disposal by burial in the soil at any one location and time is
limited to 10,000 microcuries.
5. EVALUATION OF WASTE MANAGEMENT PRACTICES
Recovery
Option No.l - Scavenging-Precipitation Foam Separation. The removal
of zirconium-95 and niobium-95 from low-level radioactive waste water by
scavenging-precipitation foam separation has been studied by Oak Ridge
National Laboratory. The process consists of two steps:
(1) precipitating in a sludge-blanket clarification step; and
(2) achieving final decontamination in a foam-separation column.
Processing rates of 300 gal. per hour (gph) for scavenging
precipitation and 120 gph for foam separation have been achieved.
Decontamination factors (ratio of initial to final concentration)
156
-------
Table 6
?14Q
ZIRCONIUM-95 AND NIOBIUM-95 MAXIMUM PERMISSIBLE CONCENTRATIONS*'H*
Isotope
Zirconium-95
Zirconium-95
Zirconium-95
Zirconium-95
Niobium-95
Niobium-95
Niobium-95
Niobium-95
Exposure
Group
Individual
Individual
Population
Population
Individual
Individual
Population
Population
Form
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Soluble
Insoluble
Concentration
in Air
microcuries/mi Hi liter
4xlO"9
IxlO"9
1.3xlO"9
.33xlO"9
2xlO"8
3xlO"9
.67xlO"8
IxlO"9
Concentration
in Water
mi crocuries/mi 11 i 1 i ter
6xlO"5
6xlO"5
2xlO"5
2xlO"5
IxlO"4
IxlO"4
.33xlO"4
.33xlO"4
-------
greater than 50 have been obtained for both isotopes. At the
present time, further development is required to reduce operating
costs and increase processing rates.
Option No.2 - Scavenging-Precipitation Ion Exchange. In this process
final decontamination is obtained by ion exchange columns from the
scavenging-precipitation process. The process includes a provision
for the recycle of the ion exchange waste to the scavenging-precipitation
step. All the removed radionuclides are concentrated in the clarifier
sludge. The overall decontamination factors for these isotopes varied
from 10 to 150. The use of ion exchange resins with a
scavenging-precipitation step is a fairly simple and efficient means
of removing these isotopes from low-level radioactive aqueous wastes.
Option No.3 - Water Recycle. The water recycle process is used for
decontaminating radioactive waste water and recycling the purified
water for reuse. The process has been demonstrated at the pilot plant
scale by Oak Ridge National Laboratory. The steps in the process
include: (1) clarification by the addition of coagulants;
(2) demineralization by cation-anion exchange; and (3) sorption on granular
activated carbon. Overall decontamination factors of 350 were obtained.
This is the minimum value since the concentrations of zirconium-95 and
niobium-95 were reduced to the analytical limits of detection (background).
This method is an improvement to the treat-and-discharge methods, but
further work is required until full-scale production use is obtained.
Storage/Disposal
Option No. 1 - Land Burial. Land burial of low-level zirconium-95
and niobium-95 wastes in small concentrations at approved sites
meeting the geologic and hydro!ogic criteria, is an acceptable
means of disposal. Their concentrations should not be in excess of 10**
times their maximum permissible concentrations for the general population
in 10CFR20.2149 All wastes to be disposed of should be in a solid form
and encased in a suitable container. Liquid wastes should be
158
-------
solidified, preferably using asphalt, in accordance with the methods
described in the Radioactive Waste Solidification report. The burial
trenches should be designed not to intercept the ground water table
and constructed with a bottom drain and sump for water monitoring. The
trenches should be covered with either asphalt or vegetation to limit
infiltration of water. The burial site design, geology, and hydrology
should be in conformance with the criteria used in selecting and
1423
licensing the present commercial burial sites. Since land burial
has successfully been practiced on a commercial scale and both zirconium-95
and niobium-95 have short half-lives, it should be considered as the most
satisfactory method of disposing of all dilute concentrations of these
wastes.
Option No. 2 ~ Near-Surface Liquid Storage. Near-surface storage of
reactor-produced aqueous solutions containing zirconium-95 and niobium-95
high-level waste in carbon steel and stainless steel tanks encased in
concrete and buried underground is a satisfactory means of short-term
storage. Aqueous solutions of high-level waste have been stored in this
manner over the past 25 years. The tanks range in size from 0.33 to 1.3
million gal. and are generally equipped with devices for measuring
temperatures, liquid levels, and leaks. These tanks are considered as an
interim storage technique due to a general lack of confidence in their
long-term integrity. Since present regulations require the solidi-
2149
fication of all reactor wastes within 5 years following reprocessing,
the near-surface storage of aqueous solutions of zirconium-95 and
niobium-95 and other fission product salts in steel tanks should only
be considered as a near-term storage technique and not as a permanent
storage or disposal technique.
Option No. 3 - Near-Surface Solid Storage. The storage of high-level
solidified zirconium-95 and niobium-95 wastes and other waste salts in
engineered storage facilities offers the best intermediate method for
storage of these wastes. The technology for these facilities has been
developed. The wastes will be under surveillance and control and can be
retrieved, should this be required. The high-level wastes from spent fuel
159
-------
processing should be solidified and packaged in a suitable container. Of the
four high-level solidification processes developed, spray or phosphate glass
solidification processes offer the better solidified waste characteristics
than the pot calcination or the fluidized bed calcination processes (see
Radioactive Waste Solidification report). The wastes will be stored in
stainless steel-lined concrete vaults which will be air or water cooled.
The periodic replacement of the waste containers probably will not be
required since the activity of zircom'um-95 is reduced by a factor of
1,00 in 432 days and niobium-95 activity is reduced by a factor of 100 in
232 days. The replacement of the containers may be required if other
long-lived isotopes are present.
Option No. 4 - Salt Deposits. This methqd offers the best potential
for the disposal of Bastes from fuel reprocessing since bedded salt •
deposits are completely free of circulating ground waters. This method
of disposal has been under study by the Oak Ridge National Laboratory
since 1957, and in November of 1970 a committee of the National Academy
of Sciences recommended that the use of bedded salt for the disposal of
0733
radioactive wastes is satisfactory. Recent questions concerning the
adequacy of this method have resulted in the need for further development
work before it can be accepted as an ultimate method of disposal. The
wastes must be solidified and packaged in a suitable container. The
solidified wastes are buried in rooms carved in the salt deposits,
approximately 1,000 ft below the ground. The salt is a good heat
transmitter, provides about the same radioactive shielding as concrete,
and can heal its own fractures by plastic flow. The critical problem
is the selection of a site that meets the necessary design and geological
criteria for the mixture of fission product and actinide wastes.
Option No. 5 - Bedrock Disposal. The disposal of high-level aqueous
solutions containing zirconium-95 and niobium-95, along with other
wastes from spent fuel reprocessing wastes in vaults excavated in
crystalline rock over 1,500 ft beneath the ground is currently being
evaluated by E. I. du Pont de Nemours, at their Savannah River Plant
near Aiken, South Carolina. ' The wastes would be stored in
160
-------
six tunnels and once in the tunnels the wastes will seep into the
surrounding rock. Located above the crystalline rock is the Tuscaloosa
formation, a good source of freshwater, which is separated by a layer of
clay that would act as a barrier to the leakage of radioactive wastes.
An advisory committee appointed by the National Academy of Sciences
recommended abandonment of the project. In May 1972, another National
Academy of Sciences panel concluded that bedrock storage provides a
reasonable prospect for long-term safe storage, but precise information
is needed to decide if and where underground storage vaults should be
built. Another method has also been proposed for disposing of liquid
wastes by in situ incorporation in molten silicate rock. At the
present time both of these methods are unproven since sufficient
engineering data or exploration has not been completed to verify their
suitability.
Option No. 6 - Hydraulic Fracturing. The direct disposal of aqueous
low-level radioactive wastes into shale formations has been
investigated by Oak Ridge National Laboratory.0709 The method consists
of mixing the aqueous wastes with cement and pumping the resulting
slurry down a well out into a nearly horizontal fracture in a thick shale
formation. Since zirconium-95 and niobium-95 are leached from cement
and no additional encapsulation is provided, this method of disposal
requires additional work to determine its suitability.
To summarize, both zirconium-95 and niobium-95 are short-lived
isotopes with a high initial activity but whose activity decreases by a
factor of 1,000 within 2 years. These isotopes will probably be
contained with the long-lived fuel reprocessing wastes since their
recovery is not required due to their short half-life and limited
commercial use. The acceptable method of treatment is solidification
followed by storage in near-surface engineered facilities and disposal
in salt deposits.
161
-------
6. APPLICABILITY TO NATIONAL DISPOSAL SITES
Zirconium-95 and niobium-95 are both candidates for a National
Disposal Site due to their projected growth with that of the civilian
nuclear power program. From a cost and safety viewpoint, it would be
desirable to combine the reprocessing plant and the National Disposal
Site. Interim storage in near-surfaqe engineered facilities offers
considerable latitude in the site selection. The site selection for
ultimate disposal of these wastes is limited to particular geological
areas in which salt deposits are present.
The recommended treatment for high-level aqueous wastes containing
zirconium-95 and niobium-95 is solidification using either spray or
phosphate glass solidification processes. If required for economic
reasons, recovery prior to solidification could be effected by
scavenging-precipitation ion exchange. The solidified wastes should
then be held in fn-t.erlrn- storage in engineered storage facilities, and
finally disposed of in salt deposits.
162
-------
7. REFERENCES
0559. AEC Manual, Chapter 0524. Standards for radiation protection,
May 12, 1964. 11 p.
0563. Recommendations of the International Commission on Radiological
Protection, ICRP Publication 2. Report on Committee II on
permissible dose for internal radiation (1959). Peryamon Press,
1959. 232 p.
0703. King, L. J., A. Shimozato, and J. M. Holmes. Pilot plant studies
of the decontamination of low-level process waste by a
scavenging-precipitation foam separation process. Oak Ridge
National Laboratory, ORNL-3803, 1968. 57 p.
0705. Staff of the Oak Ridge National Laboratory. Siting of fuel
reprocessing plants and waste management facilities.
ORNL-4451, 1970. 500 p.
0707. Yee, W. C., F. DeLora, and W. E. Shuckley. Low-level radioactive
waste treatment: the water recycle process. Oak Ridge
National Laboratory, ORNL-4472, 1970. 30 p.
0709. Struxness,R, C. and et.al. Engineering development of hydraulic
fracturing as a method for permanent disposal of radioactive
wastes. Oak Ridge National Laboratory, ORNL-4259, 1968. 261 p.
0715. Schneider, K. J. Status of technology in the United States for
solidification of highly radioactive liquid wastes. Battelle
Northwest, BNWL-820, 1968. 63 p.
0733. Report by the Committee on Radioactive Waste Management. Disposal
of solid radioactive wastes in bedded salt deposits. National
Academy of Sciences—National Research Council, Washington,
1970. 28 p.
0766. Sax, N. I. Dangerous properties of industrial materials. 3d ed.,
New York, Reinhold Publishing Corp., 1968. 1,251 p.
0894. Girdler, R. M. Storage of liquid radioactive wastes at the
Savannah River plant. Du Pont de Nemours and Company,
Aiken, South Carolina, DPSPU-69-30-9, July 1969. 11 p.
1396. Cooke, J. B. and et. al. Report on the proposal of E. I. Du Pont
de Nemours and Company for the permanent storage of radioactive
separation process wastes in bedrock on the Savannah River,
Du Pont de Nemours (E. I.) and Company, Aiken, South Carolina,
DPST-69-444, May 1970. 57 p.
163
-------
REFERENCES (CONTINUED)
1423. Morton, R. J. Land burial of solid radioactive wastes: study of
commercial operations and facilities. Environmental and Sanitary
Engineering Branch, Atomic Energy Commission, WASH-1143, 1968. 132 p.
2146. Cohen, J. J., A. E. Lewis, and R. L. Braun. In situ incorporation of
nuclear waste in deep molten silicate rock. Nuclear Technology,
13: 76-87, Apr., 1972.
2148. Atomic Energy Commission. Forecast of growth of nuclear power,
Jan. 1971. Washington, U. S. Government Printing Office
WASH-1139, 1971. 187 p.
2149. Code of Federal Regulations. Title lO—atomic energy. (Revised
as of Jan. 1, 1971). Washington, U. S. Government Printing
Office, 1971. 429 p.
2150. Code of Federal Regulations. Title 49--transportation, parts 1
to 199. (Revised as of Jan. 1, 1971). Washington, U. S.
Government Printing Office, 1971. 952 p.
2666. Recommendations of the National Council on Radiation Protection
and Measurements. Basic radiation protection criteria,
Jan. 15, 1971, Washington, NCRP Publications.
164
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Zirconium-95
Name
Half-life 65 days
Type of Decay Negative Beta
Molecular Wt. 95
Density
Fnrmula
40
Zr
95
6.49 gm/cc
Melting Pt. 1852C
Specific Power
Boiling Pt.
Ill watts/gm
3578C
Solubility^
Cold Water
Others: Soluble in HF and aqua regia, slightly soluble in acid
Insoluble
Specific Activity 2,12xlOH curies/gm
Hot Water Insoluble
Decay Chain
(90h)
95m
Radiation Energy Level ft Intensities
Beta: 0.99 Mev (2%)
0.396 Mev (55%) :
0.360 Mev (43%)
Gamma: 0.724 Mev (49%)
0.756 Mev (49%)
Shipping Regulations Classified as a transport groyp III radlonuqHde
by the Department of Transportation
Comments
References:
0776
2150
165
-------
HAZARDOUS WASTES PROPERTIES
WORKSHEET
Name N1ob1um-95
Structural Formula
Half-life 35 days
41
Nb
95
Type of Decay Negative Beta
Molecular Wt. 95 Melting Pt. 2468C Boiling Pt. 4927C
Density 8.57 gm/cc
Specific Power 190 watts/gin
Insoluble
Specific Activity 3.93x10^ curies/gm
Hot Water Insoluble _v
Solubility
Cold Water
Others: Soluble in fused alkali; insoluble in HC1. HNCU and aoua regia
""• . WM^^^B«M^"™M*^^^^^»"^««*«^^^™^«^B«IMB«^^^^^»^^^BM^«B™M«WM«^B^^*^^»^*^h«M»^^»^^^^^^M«**
Decay Chain . Radiation Energy Level & Intensities
Nb
(35d)
95 B"
95
->Mo
(stable)
Beta: 0.160 Mev (99%)
0.924 Mev (1%)
Gamma: 0.765 Mev (100%)
Shipping Regulations Classified as a transport group IV
radionuclide bv the Department of Transportation
Comments
References: (1) 0776
(2) 2150
166
-------
BIBLIOGRAPHIC DATA
SHEET
I
Report No.
EPA-670/2-73-053-J
3. Recipient's Accession .No.
4. Tide and Subtitle Recommended Methods of Reduction, Neutralization,
Recovery, or Disposal of Hazardous Waste. Volume 'IX, National
Disposal Site Candidate Waste Stream Constituent Profile
Reports - Radioactive-Materials
5. Report Date
Issuing date - Aug. 197;
6.
7. Author(s) R. s. Ottinger, J. L. Blumenthal,
G. I. Gruber, M. J. Santy. and C. C. Shih
D.-F.
Dal
I
Porto,
8- Performing Organization Rept.
N°" 21485-6013-RU-QQ
9. Performing Organization Name and Address
TRW Systems Group, One Space Park
Redondo Beach, California 90278
10. Project/Task/Work Unit No.
11. Contract/Grant No.
68-03-0089
12. Sponsoring Organization Name and Address
National Environmental Research Center
Office of Research and Development
U.S. Environmental Protection Agency
Cincinnati, Ohio 45268
13. Type of Report & Period
Covered
Final
14.
15. Supplementary Notes
Volume IX of 16 volumes.
16. Abstracts
This volume contains summary information and evaluation of waste management methods in
the form of Profile Reports for radioactive materials. These Profile Reports were pre-
pared for either a particular hazardous waste stream constituent or a group of related
constituents. Each Profile Report contains a discussion of the general characteristics
of the waste stream constituents, their toxicology and other associated hazards, the
definition of adequate management for the waste material, an evaluation of the current
waste management practices with regard to their adequacy, and recommendation as to the
most appropriate processing methods available and whether the waste material should be
considered as a candidate for National Disposal, Industrial Disposal, or Municipal
Disposal.
17. Key Words and Document Analysis. 17a. Descriptors
Radioactive Materials
National Disposal Site Candidate
Hazardous Wastes
Carbon-14
Cobalt-60
Iridium-192
Radium-226
Cesium-134
Cesium-137
Hydrogen-3
17b. Identifiers/Open-Ended Terms
Iodine-129
Iodine-131
Krypton-85
Xenon-133
Plutonium-238, 239, 240, 241
Amercium-241, 243
Curium-242, 244
Ruthenium-106
Cerium-144
Prometh i urn-147
Strontium-90
Zirconium-95
Niobium-95
17c. COSATI Field/Group Qgp.
35 .
gg
18. Availability Statement
Release to public.
- 167 -
19.. Security Class (This
Report)
UNC I..ASSIFIED
20. Security Class (This
Page
UNCLASSIFIED
21. No. of Pages
173
22. Price
------- |