RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
NUCLEAR WASTE MANAGEMENT AND THE NUCLEAR FUEL
CYCLE
Patricia. A. Baisden
National Ignition Facility Programs Directorate, Lawrence Livermore National
Laboratory, Livermore, CA, USA
Gregory R. Choppin
Department of Chemistry and Biochemistry, Florida State University, Tallahassee, FL,
USA
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Keywords: actinide chemistry, advanced fuel cycle, aqueous and non-aqueous
separations technologies, fast reactors, Gen IV nuclear energy systems, nuclear fuel
cycle, nuclear power and spent fuel, nuclear waste, partitioning and transmutation.
Contents
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1. Introduction
2. Classification of radioactive wastes
3. Who is responsible for radioactive wastes?
4. Splitting the atom for energy
5. Status of nuclear power world-wide
6. Nature of HLW as a function of time
7. Fast reactors
8. The nuclear fuel cycle
9. Important characteristics of actinides
10. Separations technologies for the nuclear fuel cycle
11. Advanced fuel cycle concepts and partitioning and transmutation (P&T)
12. Aqueous chemical processing
13. Non-aqueous chemical processing
14. Transmutation devices for the advanced fuel cycle
15. Strategies for implementation of an advanced fuel cycle
16. Generation IV nuclear energy systems
17. Future of P&T
Appendix
Glossary
Bibliography
Biographical sketches
To cite this chapter
Summary
In addition to a rapidly growing demand for more electricity, especially in Asia,
concerns over energy resource availability, climate change, air quality, and energy
security suggest a larger and more important role for nuclear power in the future.
However, it is unlikely the public will accept growth of nuclear power until issues
associated with nuclear waste management, reactor safety, economics, and non-
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
proliferation are addressed by both the nuclear industry and government. In this chapter,
the management of nuclear waste from power generation is described in terms the three
major fuel cycle options that are under consideration by various countries around the
world. The importance of the actinide elements and certain fission products as the long
term waste issues is discussed along with both aqueous and non-aqueous separation
technologies that can be used to partition these from spent fuel for subsequent
transmutation using reactor or accelerator-driven devices in an advanced fuel cycle
scenario. The strategy and challenges associated with the deployment of Generation IV
Energy systems currently under development are also discussed.
1. Introduction
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Radioactive waste, like other wastes, may be composed of materials varying in origin,
chemical composition, and physical state. However, what differentiates radioactive
waste from other waste forms is that it contains components that are unstable due to
radioactive decay. Managing radioactive waste requires different approaches to ensure
the protection of both humans and the environment from the radiation.
In general, three options exist for managing radioactive waste: (1) concentrate and
contain (concentrate and isolate the wastes in an appropriate environment); (2) dilute
and disperse (dilute to regulatory-acceptable levels and then discharge to the
environment); and; (3) delay to decay (allow the radioactive constituents to decay to an
acceptable or background level). The first two options are common to managing nonradioactive waste but the third is unique to radioactive waste. Eventually all radioactive
wastes become benign because they decay to stable elements while non-radioactive,
hazardous waste remains hazardous forever or until their chemical speciation is changed
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By far the largest source of radioactive waste from the civilian sector results from the
generation of power in nuclear reactors. Much smaller quantities of civilian radioactive
waste result from use of radionuclides for scientific research as well as from industrial
sources such as medical isotope production for diagnostic and therapeutic use and from
X-ray and neutron sources. The other significant sources of radioactive waste are due to
defense related activities that support the production and manufacture of nuclear
weapons. This chapter deals primarily with the management of civilian nuclear waste
and focuses on that generated from nuclear power.
2. Classification of Radioactive Wastes
In order to manage nuclear wastes, it is useful to classify or group them into categories
based on the properties of the waste, which can be done in a number of ways. For
example, radioactive waste can be classified by the level of radioactivity present (high,
intermediate, low, or below regulatory concern), by the dominant type of radiation
emitted (alpha, beta, gamma, or X-ray) or, by its half-life (a length of time required for
the material to decay to half of its original value). Also, radioactive wastes can be
classified by their physical characteristics (primarily, solid or liquid, but they can also
exist in the gaseous state). A quantitative way to classify radioactive waste is by
specific activity or activity concentration; i.e., by the activity per quantity of waste
(mass or volume). The heat generated in a sample (that depends on half-life,
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
concentration, and type of radiation) can also be used for classification. Finally, waste
can be classified for security and non-proliferation purposes (e.g., designated as “special
nuclear materials”), for worker safety, and for transportation.
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These various classifications have advantages and disadvantages depending upon how
the information is used. In the United States and many places around the world,
radioactive wastes have traditionally been classified on the basis of their characteristics
and how the waste was produced via a top down, generator-oriented approach. This has
resulted in inconsistencies, overlaps, and omissions which lead to conflicts between the
source-defined classifications and the waste acceptance criteria developed from the
disposal systems. Such generator-oriented waste classifications do not fully capture the
associated hazards and, thus, are insufficient to ensure public health and environmental
safety. Whatever qualitative framework is used, the length of time radioactive waste
must be isolated from the public is determined primarily by its half-life and energy. The
heat generation, concentration, and type of radiation determine the shielding and
handling requirements for the disposal of the waste.
Although the classification of radioactive wastes varies from country to country, three
groupings are generally accepted internationally.
2.1. High-level Waste (HLW)
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HLW generally refers to the radioactive nuclides at high levels from nuclear power
generation, (i.e. reprocessing waste streams or unprocessed spent fuel) or from the
isolation of fissile radionuclides from irradiated materials associated with nuclear
weapons production. When the spent nuclear fuel from reactor operations (civilian or
defense) is chemically processed, the radioactive wastes include nuclides from the
aqueous phase from the first extraction cycle (and other reprocessing waste streams) as
it contains high concentrations of radioactive fission products. As a result, HLW is
highly radioactive, generates a significant amount of heat, and contains long-lived
radionuclides. Typically these aqueous waste streams are treated by the principle of
“concentrate and contain,” as the HLW is normally further processed and solidified into
either a glass (vitrification) or a ceramic matrix waste form. Spent nuclear fuel not
reprocessed is also considered as HLW. Because of the highly radioactive fission
products contained within the spent fuel, it must be stored for “cooling” for many years
before final disposal by isolation from the environment. This final disposal of HLW is
placement within deep geologic formations. Relative to the total volume of waste
produced from commercial power generation, HWL constitutes only a small fraction (a
few percent). However, the vast majority of the radioactivity (> 95%) resides in the
HLW. The only disposal option for this class of waste is burial in a deep geologic
repository.
2.2. Intermediate-level Waste (ILW)
ILW contains lower amounts of radioactivity than HLW but still requires use of special
shielding to assure worker safety. Reactor components, contaminated materials from
reactor decommissioning, sludge from spent fuel cooling and storage areas, and
materials used to clean coolant systems such as resins and filters are generally classified
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
as ILW. The most common management option is “delay to decay” for short-lived solid
waste, but for the long-lived waste, the “concentrate and contain” principle
(solidification for deep geologic disposal) is required. ILW comprises about 7% of the
volume and, roughly, 4% of the radioactivity of all radioactive wastes. The disposal
options for this class of waste are burial in a deep geologic repository for the long-lived
radionuclides and near-surface burial for the short-lived ones.
2.3. Low-level Waste (LLW)
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All civilian and defense-related facilities that use or handle radioactive materials
generate some LLW. These include research laboratories, hospitals using radionuclides
for diagnostic and therapeutic procedures, as well as nuclear power plants. LLW
includes materials that become contaminated by exposure to radiation or by contact
with radioactive materials. Items such as paper, rags, tools, protective clothing, filters
and other lightly contaminated materials that contain small amounts of short-lived
nuclides are usually classified as LLW. By its nature, LLW does not require shielding
during normal handling and transportation and both principles of “delay to decay” and
“dilute and disperse” can be employed for disposal depending on the exact nature of the
waste. Often, it is advantageous to reduce the volume of LLW by compaction or
incineration before disposal. Worldwide it constitutes ~90% of the volume but only
~1% of the radioactivity associated with all radioactive waste. However, wastes
containing small amounts of long-lived radionuclides can be included under the LLW
classification. The disposal options for this class of waste are near-surface burial or no
restrictions depending on level of radioactivity.
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The International Atomic Energy Agency (IAEA) noted that these classifications,
HLW, ILW and LLW, although useful for some purposes, have some important
limitations. Specifically the limitations cited are no “clear linkage to the safety aspects
in radioactive waste managements, especially disposal,” the “current classification
system lacks quantitative boundaries between classes,” and no “recognition of a class of
waste that contains so little radioactive material that it may be exempted from control as
a radioactive waste.” To overcome these limitations, the IAEA proposed a modified
system shown in Figure 1.
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
Figure 1: From the IAEA publication: Classification of Radioactive Waste, A Safety
Guide, Safety Series No 111-G1.1 (1994).
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In this simple but eloquent scheme, the principal waste classes are exempt waste
(explained in the next paragraph), low and intermediate level waste (which may be
subdivided into short and long-lived waste and alpha-bearing waste) and high level
waste. On the y-axis the radioactivity ranges from very low (simple and conventional
disposal) to very high (isolation in a geological repository). As the radioactivity
increases so does the heat content, the need for shielding, and need for better and longer
isolation from the biosphere. On the x-axis, the decay periods range from very short
(seconds) to very long (millions of years) and the associated wastes range in
composition from those containing small to significant quantities of long-lived
radionuclides.
Exempt waste, also identified as “below regulatory concern” or “de minimis”, contains
very low concentrations of radioactivity. The associated radiological hazards are
generally considered to be negligible as they are less than for naturally-occurring
radioactivity.
To date, the classification scheme proposed by the IAEA has not had universal
acceptance although this scheme or slight variations of it are in use in a number of
countries. For example, most countries recognize and accept exempt waste as the fourth
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
principal classification along with HLW, ILW, and LLW.
In the United States waste classifications are defined by the US Nuclear Regulatory
Commission (NRC):
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1. High Level Waste (HLW) – (A) The highly radioactive material resulting from
the reprocessing of spent nuclear fuel, including liquid waste produced directly
in reprocessing and any solid material derived from such liquid waste that
contains fission products in sufficient concentrations; and (B) other highly
radioactive material for which the NRC, consistent with existing law, requires
permanent isolation;
2. Spent Nuclear Fuel (SNF) – Fuel that has been withdrawn from a nuclear reactor
following irradiation, the constituent elements of which have not been separated
by reprocessing; the NRC requires SNF be regulated as HLW;
3. Transuranic (TRU) Waste – Waste contaminated with alpha-emitting transuranic
elements (Z > 92) with half-lives greater than 20 years and in concentrations
higher than 100 nanocuries/gram (3700 Bq g-1);
4. Uranium mining and mill tailings – Tailings or wastes produced by the
extraction or concentration of uranium or thorium from any ore processed
primarily for its source material content (also called by-product material);
5. Low-Level Waste (LLW) – Radioactive material that is not high-level
radioactive waste, spent nuclear fuel, nor by-product material; low level waste is
further subdivided as shown in Table 1;
6. Naturally Occurring and Accelerator produced Radioactive Materials
(NORM/NARM).
LLW Waste Class
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Class A
Class B
Class C
Greater than Class C
Definition
Low levels of radiation and heat, no shielding
required to protect workers or public, rule of thumb
states that it should decay to acceptable levels within
100 years
Higher concentrations of radioactivity than Class A
and requires greater isolation and packaging (and
shielding for operations) than Class A waste.
Requires isolation from the biosphere for 500 years.
Must be buried at least 5m below the surface and
must have an engineered barrier (container and
grouting)
This is the LLW that does not qualify for near-surface
burial. This includes commercial transuranics (TRUs)
that have half-lives > 5 years and activity > 100
nCi/g.
Table 1: Subclasses of low-level waste according to the US Nuclear Regulatory
Commission (US NRC) - U.S. Code of Federal Regulations, Title 10, Part 61 (10 C.F.R.
61).
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
3. Who is Responsible for Radioactive Wastes?
Internationally it is accepted that each country is ethically and legally responsible for
the radioactive wastes generated within its borders. Specifics related to what constitutes
radioactive waste, the parties responsible, as well as the bodies charged with regulating
its use and ultimate disposal, are issues defined by governmental processes.
3.1. Pertinent Legislation in the US Regarding Radioactive Wastes: An Example
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In the United States, legislation has resulted in laws that govern the production, use, and
means of disposal of radioactive materials. In 1946 the McMahon Energy Act created
the Atomic Energy Commission (AEC) with a charter to operate the Manhattan Project
facilities. The Atomic Energy Act of 1954, as amended, defined responsibilities related
to production, use, ownership, liability, and disposal. This legislation also made
possible the creation of the civilian nuclear power program within the US. It was not
until about 1970 that the AEC defined high level waste. HLW was defined as aqueous
waste resulting from solvent extraction cycles associated with reprocessing spent
reactor fuel. Four years later, Congress enacted the Energy Reorganization Act that split
the AEC into two organizations, the Energy Resource and Development Administration
(ERDA) and the Nuclear Regulatory Commission (NRC). ERDA, now the Department
of Energy (DOE), was given the task of developing and promoting nuclear power while
the NRC was tasked with licensing and regulating the emerging nuclear power industry.
In 1980, the Low-Level Radioactive Waste Policy Act and an amendment in 1985,
transferred responsibility for disposal of non-defense related low level waste from the
federal government to the states but kept commercial transuranic (TRU) waste and
certain low-level wastes containing long lived radionuclides within the responsibility of
the federal government. This legislation also mandated that the Environmental
Protection Agency (EPA) establish radiation protection from LLW. Disposal of HLW
and spent fuel was the focus of the Nuclear Waste Policy Act (NWPA) in 1982 which
called for the construction of two permanent repositories on the west and east coasts.
The NWPA was amended in 1987 to name Yucca Mountain as the only site that would
be developed as a repository.
The disposal of HLW remains a major public policy issue in the US as well as abroad.
Public concerns and political pressures have led to the need to demonstrate the
feasibility and safety of HLW disposal and, in some countries, to implement laws that
require operational HLW disposal capability in the next 10–50 years. Many countries,
including the United States, have programs dedicated to the disposal of HLW. However,
all HLW produced to date is being stored and no permanent disposal system has been
approved.
4. Splitting the Atom for Energy
Nuclear power produces energy by fissioning nuclides, most notably 235 U or 239 Pu .
When these nuclides fission, the resulting fission fragments and neutrons have a
combined mass less than the parent nuclide. During the fission process, the difference in
mass is converted to energy and results in the release of about 200 MeV of energy (or
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
3.2×10–11 joules) for every atom that fissions. The partitioning of the energy resulting
from the fission of 235 U is shown in Table 2.
Description
Kinetic energy of fission fragments
Kinetic energy of prompt neutrons
(2.5 neutrons/fission event with a
average energy of ~2 MeV)
prompt γ-rays
Subtotal
Kinetic energy of β-particles
emitted in decay of fission products
Neutrinos
γ-rays
Subtotal
Energy/fission of an atom of U-235
in MeV
5.0
7.0
176.50
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Prompt
Energy
Approx.
Energy
(MeV)
164.5
Delayed
Energy
Total
Table 2: Partitioning of energy in the fission of
6.5
10.5
6.5
23.5
200.0
235
U
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A great advantage of nuclear power is its ability to extract a tremendous amount of
energy from a small volume of fuel. In comparison to fission, the burning of fossil fuels
involves only the breaking of chemical bonds, and as a result, only 4 eV or 6.5×10–19
joules/molecule of CO 2 are released in the combustion of a carbon atom. Another
advantage is related to the inherent characteristics of the fuel itself. Aside from its
potential use in nuclear weapons, uranium ore has little use other than for the
production of electricity through nuclear power. In contrast, fossil fuels can potentially
be used for a wide variety of applications from the production of pharmaceuticals to
various important synthetic materials production and to transportation. A third inherent
advantage of nuclear power is that under normal operations, energy is produced with
minimal environmental impact in terms of air and water pollution compared with
burning of fossil fuel, particularly coal. Fossil fuels release large quantities of CO 2 and
sulfur and nitrogen oxides (acid rain) to the atmosphere and substantial amounts of fly
ash as solid waste. International concern is increasing over the “Greenhouse Gas” threat
from fossil fuel. Energy created by splitting the atom does not produce greenhouse
gases.
5. Status of Nuclear Power World-wide
At the present time, nuclear energy provides over 17% of the world’s total electricity as
base load power. As of 2006, 441 commercial nuclear reactors operating in 31 countries
produce this electricity with a total generating capacity of over 380 GWe (gigawatts of
electrical energy). Of the existing commercial power reactors shown in Table 3, the
overwhelming majority (> 99%) are thermal reactors, which means that fission is
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
induced by low energy, thermal neutrons (~0.025 eV). Of the thermal reactors, greater
than 80% are light water reactors (LWRs) There are two primary designs or types of
LWRs, the pressurized water (PWR) or boiling water (BWR) reactors. In a LWR,
normal hydrogen in water is used as the moderating material that reduces the energy of
the neutrons released in the fission from approximately 2–3 MeV to the thermal regime
(~0.025 eV). If the neutrons are not so moderated (e.g., in a fast reactor), more energetic
neutrons (10s to 100s of keV) can also induce fission. The probability of fission for a
given nucleus depends on the energy of the neutron inducing the fission. Some nuclides
fission more easily with thermal neutrons while others require more energetic neutrons
to induce fission.
Country
Number GWe
Fuel
Coolant
Moderator
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Reactor
Type
Pressurized
Water
Reactor
(PWR)
Boiling
Water
Reactor
(BWR)
Gas-cooled
Reactor
(Magnox &
AGR)
Pressurized
Heavy
Water
Reactor
“CANDU”
(PHWR)
Light Water
Graphite
Reactor
(RBMK)
Fast Neutron
Reactor
(FBR)
US,
France,
Japan,
Russia
268
249
enriched
UO2
water
water
US,
Japan,
Sweden
94
85
enriched
UO2
water
water
CO2
graphite
23
12
Canada
40
22
natural UO2
heavy
water
heavy
water
12
12
enriched
UO2
water
graphite
4
1
PuO2 and
UO2
liquid
sodium
none
441
381
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UK
natural U
(metal),
enriched
UO2
Russia
Japan,
France,
Russia
Total
Table 3: Types of commercial nuclear reactors.(GWe=capacity in thousands of
megawatts)Courtesy of the World Nuclear Association (http://www.worldnuclear.org/info/inf32.html)
5.1. Commercial Nuclear Power Generation
The power produced by current commercial reactors ranges from about 1 to 4.5 GWth
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
(gigawatts thermal). These reactors have average efficiencies of about 33% and produce
between 500 to 1500 MWe (megawatts of electrical energy). The core of a typical large
PWR reactor consists of between 80 to 100 tons of enriched uranium (~3.5% 235 U )
divided into several hundred fuel assemblies. Each fuel assembly is made up of 200-300
fuel rods roughly 3.5 m long. Each fuel rod is composed of ceramic uranium dioxide
( UO2 ) pellets typically ~1 cm in diameter and ~1.5 cm in length.
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In the operation cycle of a nuclear reactor, several competing processes determine the
final radionuclide inventory in the spent fuel. These processes are (1) neutron induced
fission which produces energy, 2-3 neutrons on the average, and two fission products
for each atom that fissions; (2) neutron capture which produces an isotope of the same
element but of one mass unit higher; and (3) radioactive decay. In a nuclear reactor,
there are two primary modes of radioactive decay, β − and alpha decay. For the process
of β − decay, a nuclide of mass number A and atomic number Z is transformed into
another nuclide of the same mass number A but with atomic number Z + 1 . Alpha
decay results in the formation of a helium nucleus plus a nuclide four mass units and
two Z units lower than the original nuclide.
The specific amounts of fission products and transuranium nuclides produced in a
reactor depend on the mode of operation, i.e., on the energy (low, thermal, or high, fast
energy) and the flux of the neutrons (number of neutrons per cm2 per second) inducing
the reactions in the reactor fuel. In general, the components in spent fuel include:
• fission products that range in half-life from a few days or less to hundreds of
thousands to millions of years;
• uranium (element 92), primarily 235 U and 238 U along with 236 U ( T1/ 2 = 2.4×107
N
•
a) from neutron capture of 235 U ;
isotopes of the actinide elements heavier than uranium including Np, Pu, Am
and Cm.
U
At the present time, the majority of spent fuel being stored around the world has been
used in LWRs. The breakdown of LWR spent fuel is shown in Table 4.
Percentage
95.6
3
0.3
0.1
0.9
0.1
Component
Uranium
Stable or shortlived fission
products
Disposition
Recycled or disposed of as low-level,
Class C waste
Decays in a short time resulting in no
significant disposal problem
Cesium and strontium are the
primary near-term heat sources in
HLW: decays in a few centuries
Primarily iodine and technetium: can
Long-lived fission
be stored or removed and
products
transmuted*
Plutonium
Can be separated and burned as fuel
Long-lived minor Primarily Np, Am, and Cm: can be
Major fission
products
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
actinides
separated and fissioned using fast
neutrons
Table 4: Typical percentages of components making up spent nuclear fuel from LWRs
along with suggested approaches for disposal.* Transmutation is discussed in Section
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Some of the isotopes of the transuranium elements ( Z > 92 ) which are formed during
reactor operation can undergo fission with release of energy. For example, the loading
and discharge inventories associated with a 1 GWe PWR running at a burn rate of ~33
GWd/ton (gigawatt-days of operation/metric ton) is shown in Table 4. From the Table,
674 kg of the original 954 kg of 235 U loaded are consumed in the production of 111 kg
of 236 U and 563 kg of fission products. Since the total amount of fission products
produced was 946 kg of which 563 kg of fission products were from the fission of 235 U ,
~383 kg of fission products must have been produced from the fission of heavier
nuclides. Most of the additional fission products may be due to the fission of fissile
239
Pu isotope. Since the probability of fission of 238 U with thermal neutrons in a PWR
is very low, the loss of 238 U is due to the neutron capture reaction that produces 239 U
which decays by beta emission ( T1/ 2 = 23.5 months) to produce 239 Np which, again,
beta decays ( T1/ 2 = 2 d) to produce
Pu ( T1/ 2 = 24 000 a). Using this logic, the net loss
of U corresponds to the production of approximately 674 kg of 239 Pu . At the end of
the reactor cycle, however, only 286 kg of total Pu and higher actinides remain leaving
approximately 387 kg of “missing” 239 Pu . The mass balance between initial and
discharge inventories in the table is inaccurate by (1) the mass equivalence of the
energy produced (~1 kg) and (2), the mass associated with the neutrons escaping the
reactor vessel and those captured in the moderator (~2 kg), the ~387 kg of “missing”
239
Pu fissions is consistent with the excess fission product estimate of ~383 kg. Thus,
during the normal operation of a uranium-fueled PWR, roughly 40% of the energy
comes from the fission of 239 Pu produced by neutron capture reactions during the
reactor cycle.
U
N
238
239
Nuclides
U-235
U-236
U-238
U Total
Pu-239
Pu Total
Z > 94
Sr-90
Cs-137
Long-lived FP
FP Total
Total Mass
Initial Load (kg) Discharge Mass (kg)
954
280
111
26,328
25,655
27,282
26,047
56
266
20
13
30
63
946
27,282
27,279
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
Z > 94 (transplutonium elements)
FP = Fission Products
Table 5: Initial and Discharge Radionuclide Inventory in a nominal 1000 MWe Reactor
Courtesy: Elsevier from NIM A463, 428-467 (2001) H. Niefenecker, et. al.Source: P.
Schapira, in R. Turlay (Ed.) Les Dechets Nucleaires les Editions de Physique, 1997
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Table 6 provides a more detailed breakdown of the “isotopics” for spent fuel at
discharge for a typical 1000 MWe PWR. In the previous discussion, we attributed all
the “missing fission fragment mass” to the fission of 239 Pu . In reality, however, during
the course of operation of the reactor, some of the 239 Pu captured a neutron to form
240
Pu ( T1/ 2 = 6600 a). Because of its long half-life and low thermal fission cross section
(actinide isotopes with odd mass numbers are more fissile than even mass isotopes).
240
Pu most likely captures a neutron and forms 241 Pu which is a fissile nuclide.
U
N
Actinide Isotope
234
U
235
U
236
U
237
U
238
U
237
Np
239
Np
236
Pu
238
Pu
239
Pu
240
Pu
241
Pu
242
Pu
241
Am
242m
Am
243
Am
242
Cm
243
Cm
244
Cm
245
Cm
246
Cm
Half-life (years)
2.5 E+5
7.1 E+8
2.4 E+7
1.8 E-2
4.5 E+9
2.1 E+6
6.4 E-3
2.6 E+0
8.6 E+1
2.4 E+4
6.6 E+3
1.3 E+1
3.8 E+5
4.6 E+2
1.5 E+2
7.9 E+3
4.0 E-1
3.2 E+1
1.8 E+2
9.3 E+3
5.5 E+3
Kg/year in Spent Fuel*
3.1 E+0
2.1 E+2
1.1 E+2
9.1 E-7
2.6 E+4
2.0 E+1
2.0 E-6
2.5 E-4
6.0 E+0
1.4 E+2
5.9 E+1
2.8 E+1
9.7 E+0
1.3 E+0
1.2 E-2
2.5 E+0
1.3 E-1
2.0 E-3
9.1 E-1
5.5 E-2
6.2 E-3
* 1000 MWe LWR, discharge plus 150 d
Table 6: Major and minor actinide content of spent LWR fuel. Courtesy: Benedict et al,
Nuclear Chemical Engineering, p. 369, 2nd Ed., McGraw Hill, NY (1981). Source: (NAS
1996) Nuclear Wastes, Technologies for Separations and Transmutation, National
Academy Press, p. 25
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Because the fission process is asymmetric, the fission products (FPs) tend to distribute
themselves around mass 83-105 (light FP) and mass 129–149 (heavy FP). Some of the
more common light FPs are 85 Kr, 90 Sr, 95 Zr, 99 Tc 8 and the corresponding heavy FPs
are 129 I, 137 Cs, 144 Ce, and 147 Nd .
To maintain efficient performance of a nominally 1000 MWe LWR, about one-third of
the spent fuel is removed from the reactor core and replaced with fresh fuel each year.
However, over time, the concentration of fission products and heavy elements in a fuel
bundle increases to the point where it is no longer practical to continue to use the fuel
and after 2-3 years, the entire fuel core is replaced with a fresh one.
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6. Nature of HLW as a Function of Time
In the short-term, fission products with half-lives of 30-50 years dominate HLW and
constitute the main radiological factor impacting the design of a repository. This is due
to the intense gamma radiation and the resulting decay heat associated with the FPs.
The heat load initially comes primarily from the decay of 137 Cs ( T1/ 2 = 30 years) and
90
Sr ( T1/ 2 = 28 years). However, after 300 to 500 years the majority of such short-lived
fission products are gone and the dominant hazards in HLW become the actinides and
long-lived fission products (LLFP = primarily 99 Tc and 129 I , as well as 79 Se , 93 Zr , and
135
Ce , see Table 7).
Half-life (a)
2.13 × 105
1.57 × 107
3.0 × 106
1.5 × 106
1.0 × 105
6.5 × 104
Thermal Fission (TBq/tU)
0.48400
0.00120
0.00980
0.06800
0.02900
0.01510
U
N
Fission Product
99
Tc
129
I
135
Cs
93
Zr
126
Sn
79
Se
Table 7: Some long-lived fission products from a thermal reactor.a a After 10 years of
cooling a 1 t U as 3.2% enriched UO2 fuel with 33 MWd/kg U burn-up at an average
flux of 3.24×1018 n m-2 s-1 in a typical PWR.
Figure 2 shows the relative levels of the dominant fission products and selected
actinides in terms of the parameter, ingestion radiotoxicity, as a function of decay time
after discharge from the reactor. From the plot it is clear that 90 Sr and 137 Cs dominate
the hazard level of HLW at early times and then after about 30-50 years, Pu isotopes
dominate.
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Figure 2: Relative ingestion toxicity of representative fission products and some
actinides in spent fuel as a function of time after discharge from a PWR. Courtesy: Oak
Ridge National Laboratory, managed by UT-Battelle, LLC, for the U.S. Dept. of Energy
Source: (NAS 1996) Nuclear Wastes, Technologies for Separations and Transmutation,
National Academy Press, p. 24.
7. Fast Reactors
Fast reactor technology was developed during the time that thermal reactor technology
was being used for commercial power production. Since less than 1% of the energy
content of mined uranium is realized in burning 235 U in conventional LWRs, the initial
motivation for fast reactors was to breed 239 Pu from 238 U to maximize uranium fuel
utilization. In order to exploit a higher fraction of the energy content of uranium, Pu
could be produced in fast “breeder” reactors (FBRs) by placing a “blanket of uranium”
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around the core. In the blanket, 238 U would capture fast neutrons and produce more Pu
than consumed in the reactor. In fact, about 70% of all 238 U placed in a fast reactor can
be converted to 239 Pu . Since the natural abundance of 238 U is 99.3%, this conversion
results in the production of a vast supply of fissile fuel for future centuries.
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There are several major differences in the design and operation of fast reactors
compared with that of thermal reactors. In a fast reactor, no moderator is used and
neutrons induce fission at or near the energies at which they are released in the fission
reaction (~2 MeV). Although fast reactors do not require a moderator, they do require a
coolant. Since water is such an efficient medium for reducing the energy of neutrons, it
cannot be used as a coolant; therefore, liquid metals such as sodium, lead, and a lead–
bismuth alloy or gases, such as helium and carbon–dioxide have been used. Neutrons
travel at such high speeds in fast reactors that the probability of their escaping to the
surroundings is much higher. As a result of this “neutron leakage,” fewer neutrons
would be available to react with other fissile nuclides to sustain the chain reaction in
fast reactors. To overcome this problem, fast reactor fuel contains a higher ratio of
fissile-to-fertile nuclides. Common fast reactor fuel is either highly enriched 235 U or a
mixture of 239 Pu and 238 U . As Pu has a rather low fast neutron absorption cross-section,
a loading of Pu of at least 20–30% is required with 70–80% 238 U .
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In the 1980s, an expected shortage of uranium did not materialize and uranium
remained abundant and cheap. Consequently, the need for fast reactors to breed
plutonium was neither compelling nor economical. This, along with concerns related to
the potential diversion of plutonium to weapons, decreased interest in fast reactors.
Prototype demonstration fast reactors such as the BN-600 (600 MWe) in Russia (1980)
and the Superphenix (1240 MWe) in France (1985) had started to come on line and the
US had initiated the Integral Fast Reactor Program. This program was built on the
successful operation of Argonne National Laboratory’s experimental breeder reactor,
EBR-II, a liquid metal (Na) cooled fast reactor, and its associated pyrochemical process
that enabled the practical recycling of a metallic alloy fuel. However, in 1977, President
Carter halted reprocessing in the US, and in 1994, the US terminated its fast reactor
program. Likewise, the Superphenix reactor in France was shut down in 1999. The
Japanese currently have the most active FBR program in the world with both the Joyo
(100 MWth) and the Monju (280 MWe) reactors in operation.
In the future, many countries are expected to increase their electrical generating
capacity by increased reliance on nuclear power. In addition, many countries are
exploring ways to reduce the amount of nuclear waste requiring disposal in geologic
repositories by exploiting fast reactors to fission the minor actinides. Fast reactor
technology is discussed further in later sections describing partitioning and
transmutation of waste.
8. The Nuclear Fuel Cycle
The nuclear fuel cycle, shown schematically in Figure 3, consists of the processes used
to produce electricity from nuclear reactions. The cycle begins with mining the uranium
and ends with disposal of the wastes generated. Such wastes include the spent fuel itself
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and/or the material from reprocessing to recover the unused fissile material.
Figure 3: Schematic of the nuclear fuel cycle. Courtesy of the World Nuclear
Association (http://www.world-nuclear.org/info/inf03.html)
N
The fuel cycle is often described in terms of the front-end activities, beginning with
mining and milling up through and including burning the uranium fuel in a nuclear
reactor and the back-end activities in which the vast majority of the waste is isolated
and the unburned uranium and/or plutonium recovered.
U
As the term “nuclear fuel cycle” suggests, the standard scenario envisioned by the
nuclear power industry was to reprocess the spent fuel to recover the unburnt uranium
and plutonium and recycle it to generate additional energy. In addition, by recovering
uranium and returning it to the fuel cycle, the demand for new uranium and the need for
enrichment services are reduced by about 30% compared to “once-through” fuel
management. A main advantage of reprocessing in waste management is the large
reduction in volume (factor of ~100) of the materials to be disposed of as HLW. Some
countries, most notably the US and Canada, at the present time are not reprocessing
fuel, and thus do not close the fuel cycle. France, Russia, the United Kingdom, India
and some other nations are currently reprocessing spent fuel and recycling plutonium.
Japan is expected to join this group with an active reprocessing program.
8.1. Options in the Fuel Cycle that Impact Waste Management
The three major back-end fuel cycle options currently under consideration by various
countries are:
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•
•
•
a once through fuel option with direct disposal of spent fuel as HLW;
reprocessing fuel cycle (RFC) with mixed oxide (MOX) recycle of U and Pu in
light water or fast breeder reactors and disposal of HLW;
an advanced fuel cycle (AFC), an extension of RFC in which the wastes are
partitioned and some are transmuted (P&T) to reduce the long-term
radiotoxicity.
These options have different advantages and disadvantages and pose different
challenges that vary in complexity to the management of nuclear wastes.
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8.1.1. Once-Through Fuel Option
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In the “once through” option, shown in Figure 4, fuel is burned in a reactor to make
energy.
Figure 4: Schematic of the once-through fuel option
When energy can no longer be efficiently produced, the spent fuel is removed from the
reactor. Because the spent fuel is both thermally hot and intensely radioactive due to the
decay of the short-lived fission products, it is normally stored in special, water-filled
cooling ponds at or near the reactor. The water serves to absorb and dissipate heat and is
also used as a radiation barrier. Since water can moderate the neutron energy, the
cooling ponds are designed to physically separate the fuel to prevent criticality. The fuel
is allowed to cool for periods varying from 7-10 years to over 20 years depending on
the capacity at the reactor site to store spent fuel after which it is removed and placed in
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casks under a blanket of inert gas to be stored in shielded, dry storage areas for further
cooling. The once-through fuel cycle places the largest demand on repository capacities
and after 50-100 years poses one of the biggest proliferation concerns since decay of the
fission products over time could allow access and retrieval of plutonium. The advantage
of the once-through fuel cycle is that, in the near-term, potentially dangerous and
expensive handling and processing of spent fuel is avoided. However, in the long term,
large, expensive geologic repositories are needed that must be secured over very long
times to prevent removal of plutonium. Currently, the US, Sweden, Spain, Canada, and
S. Korea are using the once-through fuel option.
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8.1.2. The Reprocessing Fuel Cycle (RFC)
Figure 5: Schematic of a simple RFC.
RFC, Figure 5, was the option envisioned by the nuclear power industry from the
beginning because it offers the following advantages:
•
•
•
Uranium and plutonium would be recovered and recycled to make more energy.
The volume of material to be disposed of as HLW would be significantly
reduced since the uranium and plutonium comprise about 97% of the total mass
of spent fuel.
Recovered uranium could be re-enriched, and returned to the LWR fuel cycle,
reducing the demand for mining and milling new uranium and the need for
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enrichment services by about 30%. The option to reprocess fuel effectively
closes the fuel cycle and has the advantages noted above. However, the
development of new separation processes that are efficient and can be carried
out under a high-radiation environment is required.
Recycling Plutonium as Mixed–Oxide Fuel (MOX)
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Reprocessed plutonium can be used as fuel in modified LWRs. By adding more control
rods to manage the reactivity over the reactor volume, LWRs can be configured to
operate using fuel consisting of both uranium and plutonium–oxide or what is termed
mixed–oxide (MOX) fuel. Over the past 10 years, reactors in France, Germany,
Switzerland and Belgium have been licensed to use MOX fuel to generate electricity.
Japan plans to have one-third of its 53 reactors using MOX fuel by 2010.
MOX fuel is made by mixing plutonium with depleted uranium (a by-product of the
uranium enrichment process) or natural uranium. Two sources of plutonium are
available. The first is reprocessed plutonium from commercial spent fuel and the second
is surplus weapons-grade plutonium from the nuclear stockpiles of the nuclear weapons
countries. Such material, reactor-grade and weapons-grade plutonium, differ in the
content of 239 Pu and the minor plutonium isotopes, especially 240 Pu . Commercial
reactor-grade plutonium is nominally about 55–60% 239 Pu and has a relatively high
amount ~20–25% of 240 Pu whereas weapons-grade plutonium contains greater than
95% 239 Pu and a minor amount of 240 Pu .
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MOX fuel fabrication facilities are currently operating in France, Belgium, the United
Kingdom, India, and Japan and collectively produce about 460 tons per year. For the
fabrication of MOX fuel, the “age” of the plutonium (i.e., the time since the plutonium
was last separated from americium) is very important. Am-241, which results from the
decay of short-lived 241 Pu isotope ( T1/ 2 = 14.4 a), builds up at a rate of about 2 parts per
U
million per month or about 0.5% per year. Since 241 Am emits a 60 keV
X-ray, additional shielding and remote handling is required during MOX fabrication if
the 241 Am content becomes greater than 5–6%. As a result, MOX is usually made with
recently reprocessed plutonium.
As part of a long-term strategy for reducing the inventory of weapons-grade plutonium,
both the US and Russia, plan to construct their MOX fabrication facilities to allow
burning this excess plutonium in LWRs. Fabrication plants using weapons-grade
plutonium would have to incorporate additional controls to monitor and prevent
diversion of plutonium for weapons use.
In practice, MOX fuel can be made with fairly high (> 50%) loadings of plutonium but
to burn such a fuel would require specially designed reactors. Loadings of the order of
5% (for weapons-grade Pu) to 7% (for LWR reprocessed Pu) are more common for use
in modified LWRs. The characteristics of MOX fuel containing 5–7% plutonium are
quite similar to those of the more conventional (4–5% enriched) uranium–oxide fuel. As
a result, MOX fuel can be substituted for part of a reactor’s fuel loading without any
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significant changes in its operating conditions and performance. However, reactors
burning MOX typically limit the MOX loading to about one-third of the core fuel
assemblies.
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Since most plutonium isotopes have small thermal neutron fission cross sections, the
minor actinides, by neutron capture, build up in the fuel over time and recycle of
plutonium in LWRs has serious limitations. However, burning plutonium in MOX fuel
using conventional LWRs does have two advantages. (1) it produces useable energy
and, (2) it reduces the inventory, preventing the accumulation of separated plutonium
resulting from reprocessing commercial power reactor spent fuel. If weapons-grade
plutonium is burned as MOX fuel, there are additional advantages over and above
reducing the inventory of such plutonium. The first is that, while in the reactor, the
weapons-grade plutonium in the fuel would be protected against diversion by the
intense radioactive field of the fission products. This is similar to the argument used by
those who are against any reprocessing of fuel and claim that by not reprocessing,
plutonium left in spent fuel is protected against proliferation by the fission products in
the fuel, or what is called the “spent fuel standard.” Second, while in the reactor, the
isotopic composition of the weapons-grade Pu would be degraded and, eventually,
resemble reactor-grade plutonium (ca. 65% 239 Pu + 241 Pu and high 240 Pu content).
Burning plutonium in MOX fuel reduces the overall radiotoxicity compared with
untreated LWR spent fuel but the resulting minor actinides and fission products would
require disposition in a geologic repository. Within the commercial nuclear power
community, recycling of plutonium in MOX fuel in LWRs is viewed as a mature
industry.
Uranium Recycling
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The uranium recovered from the reprocessing of commercial spent fuel usually contains
about 1% 235 U , and, because the PUREX process does not remove 100% of the fission
products and other actinides, reprocessed uranium is slightly more radioactive than
natural uranium. The increased radioactivity is primarily due to the presence of other α emitting contaminants. [According to the International Atomic Energy Agency (IAEA
1977), the level of fission products in reprocessed uranium cannot exceed 19 MBq/kg of
U. Similarly, reprocessed U can contain no more than 250 kBq/kg U of other nonuranium alpha activity.] During irradiation, the level of 234 U increases over that in
natural uranium. In addition, 232 U , not present in natural uranium, is produced by the
decay of 2.85 a 236 Pu isotope. Although the total accumulation of 232 U is small, it
decays with a 72 a half-life by α-emission through a chain of other α - and γ -ray
emitters including the 1.9 a 228 Th . The decay chain finally culminates in 208 Tl , a high
energy γ -ray emitter. The long-lived 236 U ( T1/ 2 = 2.4×107 a) is also produced by the
(n, γ) reaction on 235 U but adds little to the α -activity. The (n, γ) reaction on
236
U produces the 6.75 d 237U isotope that decays by beta-emission to 237Np. This decay
occurs during storage prior to reprocessing and does not impact the radioactivity
associated with reprocessed uranium. In addition to these uranium isotopes, small
amounts of plutonium and neptunium remain in the uranium fraction after reprocessing
which increases the α -activity of reprocessed uranium.
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For reprocessed uranium to be used again in a reactor, it must be enriched to about
3.5%. Since new uranium is in plentiful supply, it has not been necessary to re-enrich
uranium recovered from reprocessing to meet the requirements for new fuel. If and
when it becomes necessary to enrich reprocessed uranium, the potential occupational
hazard associated with UF6 conversion, enrichment, and fuel fabrication processes due
to the small amounts of residual fission products in the uranium have to be addressed.
8.1.3. Advanced Fuel Cycle (AFC)
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The potential advantages of the AFC are similar in many respects to those described
above for the RFC. However, in the AFC, the HLW problem is reduced by removing
and destroying the minor actinides and the long-lived fission products through a process
called partitioning and transmutation (P&T). Partitioning involves carrying out a
complex series of advanced aqueous chemical and/or non-aqueous metallurgical
operations to selectively remove or partition the minor actinides (MAs) and long-lived
fission products (LLFPs). Transmutation involves the transformation of the nuclear
structure of an atom to form a different atom that is either stable or has a shorter halflife and thus is less of a concern than the original atom. P&T is discussed in greater
detail in a later section.
9. Important Characteristics of Actinides
To understand the science of actinide separation chemistry, it is necessary to be familiar
with certain characteristic properties of actinide elements. The more important of these
are discussed subsequently and more complete discussions can be found in the related
EOLSS Chapter by N. M. Edelstein and L. Morss (see: Chemistry of the Actinide
Elements), as well as in reviews and books cited in the reference list in the Appendix.
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The actinides of Z < 96 can exist in oxidation states of III to VI in solution, some in 2
or 3 such states in equilibrium. The changes III → IV and V → VI are rapid, involving
only electron exchange. However, the changes III/IV → V/VI are slower as chemical
bonds must be broken or made in addition to electron exchange since the An(V) and
An(VI) species exist as linear dioxo cations. In all oxidation states, the actinide cations
are hard acids and show little interaction with soft base donors such as S, P, etc. in
aqueous media. Also, as expected for hard-hard interactions, the bonding to donors is
primarily ionic.
Am(III) and Am(IV) cations have coordination numbers (CN) of 6 to 12 while the
dioxo actinyls have CN = 4–6. These ranges in CN reflect the ionic bonding which
results in the coordination number and symmetry being dominated by steric and net
charge effects. The complexation strength normally follows the pattern:
An 4+ >AnO 2 2+ >An 3+ >AnO 2 +
This has been attributed to an effective charge on the An in AnO 2 2+ of 3.3 ± 0.1 and in
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AnO 2 + of 2.2 ± 0.
In designing specific ligands, it is important to be aware of the following properties:
An-N bonds are longer-lived than An-O bonds;
An-O bonds promote and may be necessary to the formation of An-N bonds;
five-membered chelate rings are the most stable;
prearranged structures are more stable;
if too rigid, prearranged structures may be too slow to form;
bulky steric groups can slow dissociation kinetics;
redox, steric effects, high CN, and weak bond covalency (in extractant ligands)
can be exploited to improve specificity in separations.
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•
•
•
•
•
•
10. Separations Technologies for the Nuclear Fuel Cycle
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Separation processing of irradiated nuclear samples has been a major emphasis in
nuclear technology since the first separation experiments by W. Crookes and H.
Becquerel in the late 1890s. These initial separations, including those that led to the
discovery of Po and Ra in 1898, used precipitation as the chemical isolation and
purification method for the radionuclides. Precipitation remained the predominant
separation technique in radiochemistry even in the Manhattan atomic bomb project of
the 1940s, which achieved radiochemical separations of plutonium from uranium and
the decay and fission products. Initially, the separation used was the Bismuth Phosphate
Process with BiPO4 as the carrier for the insoluble phosphates of Pu(III) and Pu(IV) to
obtain separation from the soluble UO 2 2+ phosphate complexes.
Figure 6: Fuel reprocessing chemistry
The BiPO4 method was replaced by solvent extraction processes which used
continuous, remote controlled operations for isolation of uranium and plutonium from
the fission products and from each other. The initial solvent extraction system, REDOX,
used methylisobutyl ketone (MIBK or kexone) as the extracting solvent. However the
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REDOX process had the problem of hexone decomposition in the nitric acid solution
from the dissolution of the spent fuel. It was displaced by the PUREX (Plutonium
Uranium Redox Extraction) which continues to be the major system for spent nuclear
fuel processing. Figure 6 summarizes these separation methods.
10.1. PUREX Process
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The major system used internationally for processing spent nuclear fuel has been
PUREX. The PUREX process is shown in Figure 7.
Figure 7: Schematic of the PUREX process
The PUREX process uses tributylphosphate, TBP, dissolved in kerosene to separate
UO 2 2+ and Pu 4+ from fission products. The PUREX process depends on the relative
ease of redox of plutonium which allows recovery of purified plutonium from much
greater concentrations of uranium. In PUREX, both solvent and aqueous reagent
streams can be recycled for reuse, reducing the amount of process waste. The plutonium
and uranium are separated with decontamination factors greater than 107 from fission
products. The amount of plutonium and uranium which remain with the fission products
in the aqueous waste streams is usually tenths to a few percent.
Depending on the composition of the spent fuel, additional separation steps may have to
be added to the PUREX process to obtain satisfactory separation of certain fission
products in the wastes. For example, it is possible to selectively remove long-lived
fission products (e.g., 137 Cs ) from the waste streams using ion-exchange processes. An
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example of an additional solvent extraction system is the SREX process, which uses
complexation to a crown ether to remove Sr-90 from the PUREX waste streams.
10.2. DIAMEX Process
The DIAMEX process is the result of a research program of the European Union with
the goal of improved actinide element separation from fission products in acidic nuclear
waste solutions. Bifunctional diamides have advantages for actinide extraction in such
waste solutions such as their incinerability and the fact that their radiolytic and
hydrolytic degradation products do not hamper the process.
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The diamide N,N’-dimethyldioctylhexyloxyethylmalonamide,
(CH 3C8 H17 NCO) 2 CHC2 H 4 OC6 H13 (DMDOHEMA) was found to have the necessary
properties for use. It has disadvantages, such as the formation of degradation products
which are more soluble in the organic phase due to the replacement of the CH 3C4 H 9 N
group by CH 3C8 H17 N . However, these are balanced by its improved separation factors
from fission products such as Mo(VI) and Ru as well as from contaminates such as
Fe(III). An important factor is that the recovery of An(III) and An(IV) and
AnO 2 2+ species is quantitative. The stripping of Pu(IV) and U(VI) from the DIAMEX
extractant solution is satisfactory but the extraction and stripping from DIAMEX of Pd,
Tc and Np is not satisfactory and is being studied for improvement.
10.3. TRUEX Process
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This solvent extraction process is designed to separate transuranic elements from the
acidic high-level waste solutions of PUREX. The extractant, octyl(pheny)-N,Ndibutylcarbamoyl-methylphosphine oxide (CMPO), is dissolved in an alkane solvent.
CMPO can extract transuranic elements from acidic solutions almost quantitatively and
selectively. TRUEX treatment of the waste streams from PUREX separation of
plutonium and uranium reduces the concentrations of residual U and Pu in the wastes by
factors of 100 to 1000 but does not separate tri- and tetravalent actinides from the
lanthanide fission products.
A problem with the use of TRUEX is the need for aggressive stripping conditions to
recover the An(III) and An(IV) species from the organic phase due to the strength of
their complexation by CMPO.
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Figure 8: The TRUEX Process
10.4. TRAMEX Process
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Liquid cation exchangers, such as trialkylamines and tetraalkylammonium salts
dissolved in organic solvents, provide high selectivity in the separation of tri- and
tetravalent actinides from lanthanides and most other fission products. This is the basis
of the TRAMEX process. The separation is usually accomplished using aqueous
solutions with high chloride (e.g., 10 M LiCl) concentrations and requires pretreatment
to convert the aqueous nitrate solution of the wastes from the PUREX process to the
chloride feed solution. As the strong corrosive nature of the concentrated chloride
solution requires special processing equipment, it would be an advantage to develop
separations from high nitrate salt solutions. A partitioning/transmutation system
(discussed below) requires the separation of actinides as the first step for which the
TRAMEX process could be used.
10.5. TALSPEAK Process
The high cross sections for neutron absorption of some lanthanides require their
removal from recycled spent fuel as well as from the target material for transmutation of
actinides. The TALSPEAK (Trivalent Actinide Lanthanide Separation by Phosphorus
Extractants and Aqueous Komplexes) process separates lanthanides from trivalent
actinides in acidic solutions of pH 2.5–3.0. In the “normal” version of the TALSPEAK
process proposed for use in the partitioning/transmutation technologies, the trivalent
lanthanides are extracted by di(2-ethylhexyl) phosphoric acid (HDEHP) from an
aqueous solution of lactic and diethylenetriaminepentaacetic (DTPA) acids while the
trivalent actinides remain in the aqueous phase. In the “reversed” version, the Am(III)
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and Ln(III) cations are extracted into an organic phase containing HDEHP in the first
step, followed by stripping of the trivalent actinides into an aqueous phase containing
lactic and DTPA acids. Subsequently, the trivalent lanthanides in the organic phase can
be recovered by stripping with 6 M nitric acid.
A disadvantage of the TALSPEAK process is the formation of radiolytic degradation
products of the organic complexing agents and pH buffering agents which could affect
control of the process.
10.6. Stereospecific Extractants
Sr and 137 Cs are the major sources of radioactivity and heat in nuclear wastes and their
removal prior to further separation greatly simplifies subsequent waste handling and
storage. A Sr(II) extraction/recovery process, SREX, uses a solution of a crown ether,
4,4’(5’)-di-t-butylcyclohexano-18-crown-6 (DtBuCH18C6), and TBP in Isopar-L.
SREX can remove strontium from the waste streams of the PUREX process. A
modification of the process, called CSEX-SREX, provides the removal of both Cs and
Sr by extraction with a dibenzo-18-crown-6 derivative of proprietary composition. Both
processes achieve very high decontamination factors for Sr and Cs.
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While many of the extractants discussed have strong complexation properties, their
selectivity in the separation of some radionuclides is insufficient. Crown ethers,
cryptands and similar macrocyclic ligands can provide the desired high stereoselectivity
for specific nuclides that match their cavity size and shape. Good separation factors
have been reported between trivalent lanthanides and trivalent actinides for solvent
extraction systems based on complexants with “soft” donor groups (e.g., nitrogen and
sulfur). These complexants are based on amide functional groups or on sulfur based β –
diketone extractants. However, redox techniques can separate the actinides, including
thorium and uranium, more simply and, usually, more efficiently. Research on soft
donor ligand systems for separation of the transplutonium actinides from the
lanthanides continues and such a process could be used if a separating ligand with very
high radiation stability and low aqueous solubility is found. The efficiency of
regeneration of reagents for reuse is a problem if the specificity is due to a strong
rigidity and, hence, inertness of the ligand structure.
Table 8 lists a number of aqueous-based separations used or under development to
support reprocessing spent nuclear fuels and advanced fuel cycle concepts.
Ligand Class
Phosphoric acid based DIDPA,
HDEHP, TBP
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Separation Process
DIDPA- di-isodecylphosphoric acid – uses TBP as
a modifier to avoid third phase formation
HDEHP – di-2-ethyl-hexyl-phosphoric acid
TALSPEAK – uses HDEHP with TBP as a
modifier to avoid third phase formation to extract
Ln from an aqueous solution containing lactic or
glycolic acid and diethylenetriaminiopentaacetic
acid (DPTA). Ac stay in aqueous phase
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CMPO-carbamoylmethylphosphine oxide
Alkyl-phosphine oxides
Azines
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Diamides
TRUEX – TRans Uranium EXtraction
n-octyl-phenyl-di-isobutyl-carbamoylmethylphosphine-oxide (CMPO),strong complexing
makes stripping with conventional dilute solutions
difficult, requires more powerful stripping agents
that increase quantity of secondary waste
TRPO - Trialkyl phosphine oxide, must neutralize
feed to allow extraction and then raise the acidity
to strip the Ln and Ac, Zr, Mo, Ru and Tc interfere
with separation
DIAMEX – DIAMide EXtraction – uses
malonamide extractants, extraction/ scrubbing/
stripping cycle similar to TRUEX but no solid
secondary waste expected because extractant is
fully destroyed by incineration.
TPTZ – Tripyridyltriazine used with dinonylnapthalenesulfonic acid (HDNNS) to
selectively extract An
Picolinamides – N and O chelating agents that can
selectively extract Ac over Ln but requires high
nitrate and low proton concentration
CYANEX – 301 mixture consisting of mainly
bis(2.4.4. –trimethylpentyl) dithiophosphinic acid,
requires relatively high pH (3.5-4.0), often is used
synergistically with TBP
SESAME – electrochemical method used to
oxidize Am to either the +6 or +5 state in nitrate
media but because of difficulty keeping Am
oxidized in aqueous media, use of a complexant to
reduce the apparent redox potential is needed. The
cage-like heteropolyion, potassium
phosphotungstate is used. Oxidized species is then
separated by extraction.
SREX – Sr Extraction – uses dicyclohexano 18crown-6 ether to selectively extract Sr from acidic
liquid waste
Pyridine and Amides
Thio-phosphinic acid based
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Electrochemical redox with
heteropolyions
Crown ethers
Table 8: Some of the aqueous-based separation methods used or under development for
reprocessing spent nuclear fuel.
10.7. Non-aqueous Processes
Non-aqueous processes have been developed to various levels for separation of
radionuclides. These non-aqueous techniques use differences in such properties as
volatility and redox behavior of the radioactive elements in molten salt solvents and
have been employed successfully in uranium isotope separations, in electrorefining of
plutonium, and in the production of metallic fuel for advanced nuclear reactors. Some
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non-aqueous processes being evaluated as separation technologies for treatment of
nuclear wastes are reviewed.
10.7.1. Volatility Processes
Uranium hexafluoride has been used for uranium isotopic enrichment in the gaseous
diffusion process. Plutonium and neptunium also form volatile hexafluoro compounds.
The radiolytic decomposition of PuF6 to PuF4 and F2 in the volatilization process is a
disadvantage of the method, since PuF4 is less volatile and deposits on the surfaces of
the equipment.
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Some β -diketone ligands form relatively volatile compounds with trivalent lanthanides,
indicating the volatility can be expected for trivalent actinides although uranium and
plutonium tetravalent and hexavalent β -diketones have much lower volatilities. For β diketonates such as “fod” (6,6,7,7,8,8,8-hepta-fluoro-2,2-dimethyl-3.5-octanedione),
separation factors greater than or equal to 104 for uranium and plutonium from Am can
be expected. Fod has the desirable properties of stability to air and water at
temperatures ≥ 100°C and can be recovered by extraction with an organic solvent.
Processing of large volumes of aqueous waste by extraction with β -diketones followed
by volatilization of the β -diketonate complexes may have engineering problems to
overcome before use of these volatile compounds could be considered as a viable
separation option for the large volumes of defense wastes.
10.7.2. Molten Salt Processes
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Two basic types of molten salt systems have been used in separations: (1) a mixed
fluoride salt which was used both as the coolant and as the fuel and blanket system in
the Molten Salt Reactor Experiment (MSRE) and (2) the chloride eutectic salts which
have been used as an ionic solvent in pyrochemical spent fuel reprocessing systems
proposed for use with highly irradiated metallic reactor fuels.
An important advantage of the molten inorganic salts as well as of molten metals for
spent fuel processing is their inherent resistance to radiation damage. For example,
pyrochemical systems allow processing reactor fuels after minimum “cooling” time
even through they have very high concentrations of fission products, high decay-heat
output and intense radiation emission. In fact, the decay heat can be an advantage since
it can be used to maintain the pyrochemical fluid at the process temperatures of 500 to
800°C.
The Salt Transport Process is a sophisticated pyrochemical method for recovering U
and Pu from residues of fast reactor fuels. In this process, plutonium and uranium are
separated from fission product residues and other spent reagents. The basis of the
separation is the transport of actinide metals between two molten alloys of magnesium,
one containing copper and the other containing zinc. The alloys are chemically
connected by an ionic molten salt solvent which allows movement of ions between the
alloys. Equilibrium is established rapidly at the operating temperature (800°C) and
plutonium can be separated from uranium by a factor of about 10 and the fission
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products are isolated from the actinide elements. The noble element fission products are
retained in the copper-magnesium alloy, and the more reactive products (Cs, Sr, and the
rare earth elements) remain as ionic species in the transport salt phase. No external
electrical potential is needed in this process as the separation is accomplished by
oxidation of the plutonium by the chloride solvent salt at the copper-magnesium
interface, and reduction of the plutonium chloride by the magnesium alloy at the
magnesium-zinc interface.
10.7.3. Electrochemical Separations using Non-Aqueous Processes
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The electrochemical technique known as electrorefining was developed to purify
plutonium metal alloys to be used in the LAMPRE reactor project (Los Alamos Molten
Plutonium Reactor Experiment). LAMPRE never operated but the separations technique
was adapted to purification of plutonium and uranium metals for both weapons and
breeder reactor fuels. Electrorefining is classified as a pyrochemical process since it
uses a molten ionic salt as a transfer medium of pure Pu/U metal to the cathode. The
process somewhat resembles the Salt Transport Process described above, but the driving
potential is provided electrically and can be controlled very precisely to provide a
cathode product that is extremely pure. Spent driver fuel and blanket fuel used to breed
plutonium from the Experimental Breeder Reactor (EBR II) has been successfully
reprocessed by the electrometallurgical process developed by Argonne National
Laboratory (ANL). EBR-II driver fuel consists of highly enriched uranium metal
alloyed with about 10% Zr and clad in stainless steel. The blanket fuel (depleted
uranium metal) also uses stainless steel cladding. Briefly, the electrometallurgical
treatment (shown schematically in Figure 9) involves chopping the spent fuel into small
pieces and placing it in an anode basket inside an electrorefiner vessel containing a
LiCl/KCl salt mixture heated to about 500°C. Prior to electrolysis, an oxidant such as
CdCl2 or UCl3 is added to the salt. Upon passage of a constant electrolysis current
between the anode basket and a steel cathode, elements that form very stable chloride
compounds (e.g., actinides, alkali, alkaline earth and rare earth elements) dissolve into
the salt leaving the less stable chloride compounds (e.g., Cd, Fe, Nb, Mo, noble metals)
with the cladding hulls in the anode basket. The electrolysis is carried out under
controlled current conditions such that principally U3+ is reduced to the metal at the
cathode. After a period of time the cathode is removed from the electrorefiner and the U
metal is recovered after the adhering salt is removed by volatilization in a vacuum
furnace. The uranium is then cast into an ingot in a high-temperature furnace and stored.
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Figure 9: Flow diagram of the non-aqueous electrometallurgical process developed and
used to reprocess EBR II spent fuel.
Courtesy: (NAS 2002) Electrometallurgical Techniques for DOE Spent Fuel – Final
Report, National Academy Press, p. 21, Fig. 2.2.
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ANL also demonstrated a variation to this flowsheet that uses a liquid Cd cathode in
addition to the steel cathode, as shown in Figure 10. The Cd cathode permitted the
separation of Pu, the transuranium elements, and some rare earth fission products from
the salt. Without the Cd cathode, these elements remained along with most of the fission
products in the salt. The salt was subsequently processed by adding zeolite, heating to
adsorb the salt into the zeolite, adding glass and then hot isostatically press the mixture
to form a glass bonded sodalite ceramic waste form. The cladding hulls, noble metal
fission products, and any unoxidized material that remained in the anode basket were
subjected to removal of adherent salt by volatilization and converted to a metallic waste
form by adding more Zr and cast into an ingot in a furnace. By changing the head end
treatment steps, other types of fuel can be accommodated in this non-aqueous process.
For example, an initial Li/Li2O step can be added to convert oxide fuel to the metallic
form. To be compatible with the electrometallurgical treatment just described, the
electrorefiner requires the spent fuel be in a metallic form.
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Figure 10: Modified flow diagram of the non-aqueous electrometallurgical process that
used a liquid Cd electrode to separate the TRU elements from the majority of the fission
products.
Courtesy: (NAS 2002) Electrometallurgical Techniques for DOE Spent Fuel – Final
Report, National Academy Press, p. 21, Fig. 2.3.
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This electrometallurgical process with some modifications and additional development
could be used in transmutation schemes involving other fuel types such as metal–oxides
and nitrides and in situations where cooling periods as short as 3–5 months are required.
ANL has already developed an electrochemical head end process to convert the oxide
fuel to the metallic form by reducing the oxide in a LiCl/Li2O molten salt bath. The
resulting metal fuel could then be electrorefined using the process described for EBR-II
fuel.
11. Advanced Fuel Cycle Concepts and Partitioning and Transmutation (P&T)
The Advanced Fuel Cycle (AFC) is an option that not only closes the fuel cycle but also
allows significant reduction of the inventory of long-lived radionuclides in the waste.
This option is viewed favorably because (1) unused uranium and plutonium in the spent
fuel is recovered for future use to generate energy (economic issue); (2) the associated
radiological hazard is greatly reduced because long-lived actinide isotopes are removed
(public health and safety issue; (3) the volume, heat load, containment times, and
repository requirements are significantly decreased (cost issue). The AFC requires
coupling and integration of all of the processes and steps in the fuel cycle from fuel
fabrication through disposal. It differs from the RFC in that, in addition to the removal
of uranium and plutonium, the minor actinides (neptunium, americium, and curium, and
small amounts of the heavier actinides) and some of the long-lived fission products
(principally, technetium, and iodine) are partitioned from the waste stream and
transmuted prior to vitrification and final disposal in the HLW. Partitioning involves a
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complex series of aqueous chemical and/or non-aqueous metallurgical operations to
selectively remove or partition the minor actinides (MAs) and long-lived fission
products (LLFPs). Transmutation involves the transformation of a minor actinide or
long-lived fission product into another nuclide that is either stable or has a shorter halflife and thus, is less of a problem than the original radionuclide. Transmutation involves
using nuclear reactions and processes such as neutron capture followed by radioactive
decay, (n, 2n) reactions with energetic neutrons, or neutron induced fission (thermal or
fast) to achieve the desired transformation. Transmutation, sometimes referred to as
“incineration,” can be accomplished through a combination of LWRs, fast reactors
(configured either as breeders or as minor actinide or transuranic [TRU] “burners”), and
subcritical, accelerator-based transmutation devices. The AFC option is the least
developed of the fuel cycles although many of the technologies that are needed are well
understood and, in some cases, quite mature. Because of concerns about greenhouse
gases and global warming, many countries are moving toward increasing the amount of
energy produced by nuclear reactors and closing the fuel cycle with various adaptations
of AFC.
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Figure 11 shows the relative ingestion toxicity of HLW from spent fuel as a function of
time after discharge from the reactor by various scenarios of partitioning and
transmutation (P&T). If no attempt is made to remove the MA, it takes on the order of
105 years for the spent fuel to decay to the level of 1 ton of natural uranium. However, if
99.9% or a ten-fold decrease in the MA content is achieved by P&T, it takes only about
500-700 years to reach the same level. Although not shown on the graph, a partitioning
scheme that includes only the partitioning of plutonium for recycle and not the minor
actinides for transmutation, reduces the long-term radiotoxicity of the waste by only
about a factor of five. Therefore, plutonium and the minor actinides have to be
considered together for any transmutation scheme to be beneficial. It should be further
recognized that any fuel cycle strategy involving partition and transmutation that does
not include breeding of plutonium addresses only long-term waste issues but does
nothing to improve the resource utilization of uranium. If reducing the MA content 100fold requires isolation of HLW for only 500-700 years to reach the level of natural
uranium, containers designed to enclose the HLW would fulfill this safety requirement.
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Figure 11: Relative ingestion toxicity of representative fission products and some
actinides in spent fuel as a function of time after discharge from a PWR. Note reduction
in toxicity resulting from P&T under various scenarios of MA and LLFP removal.
11.1. Transmutation of Minor Actinides
For the purpose of transmutation, the minor actinides of concern in spent fuel are
neptunium, americium and curium although small amounts of the higher actinides,
berkelium, californium, einsteinium, and fermium can be made under the right
conditions. The reaction pathways leading to the production of the transuranium
elements are given in Figure 12.
For the minor actinides, the transmutation process consists in the capture of one or more
neutrons until a more fissionable isotope is formed. For the actinides, the most
important transmutation reaction is fission because it results in the removal of the
isotope from the minor actinide inventory and replaces it with two typically shorter
lived, less toxic fission fragments. With more energetic neutrons, the (n, 2n) reaction is
also useful because this reaction transforms fertile actinides with low fission
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probabilities to actinides with higher fission probabilities. Neutron capture reactions
that produce less fissionable isotopes add to the inventory of minor actinides.
Figure 12: Neutron capture pathways of importance for the production of actinides.
Courtesy: (NAS 1996) Nuclear Wastes, Technologies for Separations and
Transmutation, National Academy Press, p. 22, Fig. 2-1.
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Because thermal neutrons are not as effective for inducing fission reactions as more
energetic neutrons, conventional LWRs alone cannot be used to transmute minor
actinides. Thermal reactors tend to preferentially produce minor actinides which
(especially, non-fissile even A isotopes of plutonium) increase in amount as a function
of time. Fast reactors or fast neutron spectrum devices more effectively destroy the
minor actinides because the probability of fission is considerably higher with fast
neutrons compared to thermal neutrons for both even and odd A isotopes of plutonium
fission. Thus, fast neutron devices are much preferred for the recycle of plutonium and
the ultimate complete destruction of all of the minor actinides. The transmutation yield
in reactors using either thermal or fast neutrons is not the only factor that must be taken
into account when developing transmutation scenarios for the minor actinides. Of equal
or greater importance is the impact of loading of MAs on the stability of the reactor
during normal operations. For example, the addition of minor actinides can result in
transients or local spikes of the neutron flux within the reactor volume effecting the
overall reactivity and stability of the reactor core.
Transmutation of 237 Np (with half-life of about 2 million years) is possible in both
thermal and fast reactors. For reactor transmutation, neptunium can be incorporated in
either LWR– UO2 or MOX fuel and transmuted homogeneously with UO2 or
heterogeneously as a mixture of UO2 and PuO 2 . Neither approach presents any
significant problems with respect to reactor operation at small loadings (a few percent).
Fuel fabrication involving neptunium however is a bit more difficult because 237Np
decays to 233Pa which emits a number of energetic γ-rays, requiring additional shielding.
As a result, the alpha activity and gamma radiation levels are significantly higher when
neptunium is added to the mix during conventional UO2 fuel fabrication. For MOX fuel
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fabrication, the impact is much less. In this case, the presence of neptunium is
inconsequential with respect to alpha activity but the gamma radiation level is
significantly increased. This situation changes, however, upon multiple recycling of
MOX fuel. Neutron capture by 237Np leads to an increase of the short-lived
238
Pu isotope ( T1/ 2 = 88 y) which decays by alpha emission with a high decay energy.
As a result the alpha activity increases significantly as well as does the heat load during
processing. The (α, n) reaction of light elements also becomes important as a result of
the high alpha energy associated with the decay of 238 Pu . Neutrons emitted from (α, n)
reactions complicate not only fuel fabrication but also the transportation and storage of
the fuel.
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Figure 12 shows the transmutation pathways for the production and destruction of Am
isotopes. For americium, there are primarily two isotopes of concern in developing any
transmutation scheme. They are 241 Am ( T1/ 2 = 460 a) and the longer-lived 243 Am ( T1/ 2 =
7900 a) produced by the beta decay of 241 Pu and 243 Pu respectively. The 152 a
242m
Am isotope, formed by the (n,γ) reaction of 241 Am is also of importance because its
decay leads to 242 Cm . This nuclide, with its very short 163 d half-life, is the most
intense α-emitter in spent fuel. Successive neutron captures by 243 Am , followed by beta
decay, leads to the production of curium with masses of 244, 245, and 246.
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The ability to transmute 241 Am is particularly advantageous because it is the parent of
long-lived 237 Np . Thus, if 241 Am can be removed from the HLW stream and destroyed
by fission, a major source of 237 Np can be eliminated. However, transmutation of
241
Am is complicated due to its emission of a 60 keV γ -ray. Because of this gammaray, fuel fabrication involving significant quantities of americium requires shielded
facilities with equipment modified with remote operation. For these reasons, blending
americium to form MOX fuel (homogeneous recycle) is difficult and the preferred
method of transmutation is by using specially prepared targets (heterogeneous recycle).
Because curium isotopes have both shorter half-lives and higher spontaneous fission
branches compared with neptunium and americium, their transmutation in reactors
poses another set of problems. In spent fuel, 242 Cm ( T1/ 2 = 163 d) is, by mass, about 100
times less abundant than the 18 a 244 Cm isotope but its specific activity is about six
times higher. Both the high specific activity and the neutron intensity associated with
curium makes handling operations and fuel fabrication much more difficult than even
americium. Any transmutation scheme involving curium is complicated by the fact that
curium decays primarily by α-emission to produce isotopes of plutonium. Specifically,
242
Cm α-decays to 238 Pu , 243 Cm to 239 Pu , 244 Cm to 240 Pu and 245 Cm to 241 Pu , etc. For
the case of transmuting the most abundant curium isotope by mass in spent fuel, 244 Cm ,
the in-growth of 240 Pu causes problems because of its low thermal fission cross section.
As a result, transmuting 244 Cm in a LWR is not practical. However, this is a case where
using the “delay and decay” can be advantageous. For example, if 244 Cm is stored for
about 100 years, it will decay to ~2% of its original value and the 240 Pu that has grown
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in could be separated and transmuted in a fast reactor along with normal plutonium
recycling operations.
11.2. Transmutation of the Long-lived Fission Products
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In addition to removing the MA, there are a few long-lived fission products that, if they
could be partitioned and transmuted as well, would further reduce the radiotoxicity of
HLW by about another factor of ten. (See Figure 11.) The most important fission
products associated with long-term risks are technetium (existing only as 99 Tc ), iodine,
and possibly cesium. The transmutation process for fission products is basically the
capture of a neutron (or several neutrons) followed by beta decay until a less toxic
(shorter-lived) or stable nuclide is formed. For technetium and iodine, the transmutation
can be accomplished in thermal reactors through the following pathways:
Tc (T1/2 = 2.13×105 a)
99Tc + n → 100 Tc
⎯⎯
→
•
(T1/2 = 16 s)
129
Ru + n →
101
(stable)
Ru + n → 102 Ru
(stable)
(stable)
I (T1/2 = 1.6×107 a)
129
I+n →
130
I
⎯⎯
→
•
(T1/2 = 12.4 h)
127
100
130
Xe + n → 131 Xe + n → 132 Xe
(stable)
(stable)
(stable)
I (stable)
127
→
I+n
128
I
⎯⎯
→
•
Xe + n
(stable)
→ 129 Xe + n → 130 Xe
(stable)
(stable)
N
(T1/2 = 25 min)
128
U
For both technetium and iodine, capture of a single neutron followed by β − decay
results in the production of stable isotopes of ruthenium and xenon respectively.
Further, as the irradiation proceeds additional neutrons can be captured but these
products are also stable. In practice, the transmutation of technetium and iodine in
conventional thermal reactors is difficult for several reasons. First, because these
isotopes are major fission products, they are produced in high yield and, over time,
result in large inventories that must be separated from the HLW and transmuted.
Second, the thermal neutron capture cross sections are not particularly large so much
higher neutron fluxes are required to reduce significantly the inventory in a reasonable
amount of time. Third, the available neutron flux from conventional LWRs is too small
(1013 n cm-2 s-1) to achieve the required level of transmutation in a reasonable length of
time.
For cesium, the long-lived fission product of concern is 135 Cs . However 135 Cs
constitutes only about 15% of isotopes of cesium that are produced in fission. The
relatively short-lived 137 Cs ( T1/ 2 = 30 a) and stable 133 Cs are present in amounts similar
or greater than
135
Cs. Cs-134 ( T1/ 2 = 2 a) is also present at about the percent level.
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Although 135 Cs can be converted to stable 136Ba by neutron capture followed by beta
decay, this is not practical because the highly radioactive 137 Cs isotope would make
handling the material extremely difficult. In addition, neutron capture by the other
cesium isotopes, 133 Cs and 134 Cs , would produce more 135 Cs than would be
transmuted. The only way around these problems would be to remove 135 Cs by isotope
separation prior to transmutation. Since isotope separation is a very costly process,
transmuting 135 Cs may no be cost effective. Similar situations exist for transmuting the
long-lived isotopes of zirconium ( 93 Zr ), selenium ( 79 Se ), and tin ( 126 Sn ).
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Although some of the long-lived fission products have thermal neutron capture cross
sections suitable for transmutation in reasonable irradiation times, others do not and
would require reactors with very high neutron fluxes to significantly reduce inventories.
For such reduction, dedicated reactors with large thermal fluxes and/or dedicated
accelerator-driven systems (ADS) with thermal neutron fluxes of the order of 1016 n cm2 -1
s are being considered. Fast neutron capture cross sections for most long-lived fission
products are several orders of magnitude smaller than those for thermal neutrons,
making transmutation by fast reactors very difficult. Because of the relatively low
transmutation rates for long-lived fission products, transmutation of these radionuclides
is expected to be very expensive unless new, cost-effective technologies are developed.
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In general, transmutation of long-lived fission products, although theoretically possible,
is very difficult and would require significant development to be successful. Fission
products under transmutation conditions are only consumers of neutrons and, unlike the
actinides, do not produce additional neutrons useful for further transmutation.
Consequently, the product of the capture cross section and the neutron flux must be
higher than the decay constant of the isotope for it to be a realistic candidate for
transmutation. Without prior isotopic separation, the transmutation of only a few longlived fission products, namely iodine and technetium and, possibly 107 Pd , are feasible.
Thus, the best way to minimize the long-term radiological impact of most long-lived
fission products is to separate and immobilize them in specially designed, stable waste
forms.
11.3. Partitioning Schemes for the Minor Actinides and Long-lived Fission
Products
In order to realize the AFC and significantly reduce the inventories of minor actinides
and possibly, of some long-lived fission products, advances are required in the chemical
processes used to separate or partition these elements from spent fuel. A number of
aqueous processes have been studied that are either variations or extensions of the
conventional PUREX process. In addition to separating uranium and plutonium for
recycle as fuel, aqueous processes could be used for the initial, first cycle separation of
selected long-lived fission products, the minor actinides, and possible cesium and
strontium from spent fuel. Because aqueous processes undergo radiolysis in intense
radiation fields from “fresh” spent fuel, non-aqueous processes are required, particularly
in multiple recycle situations where minimizing the time between fuel discharge and
fuel reprocessing is critical. In addition to developing new chemical separation
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processes, a significant amount of development is needed on fabrication of targets of
minor actinides and long-lived fission products compatible with transmutation devices.
The reprocessing of conventional LWR–MOX fuel for recycling the transuranium
elements (TRU) and burning them in fast reactors or in dedicated transmutation devices
is conceptually possible in present PUREX reprocessing plants. The same PUREX
process used for the removal of transuranium elements from HLW from the
reprocessing of once through LWR– UO2 fuel from conventional nuclear power plants
can be used for MOX. However, reprocessing of MOX fuel is more difficult because
the radiotoxicity is about eight times greater for MOX compared with once through
spent LWR– UO2 fuel. Subsequent reprocessing of recycled MOX fuel would require
dedicated facilities with more shielding and remote handling capabilities and new, nonaqueous processes.
12. Aqueous Chemical Processing
12.1. Improved PUREX Process – Removal of Np, I, and Tc
Neptunium can be effectively separated by a relatively simple modification to the
conventional PUREX process. Neptunium can be oxidized to the VI state by nitrous
acid and extracted into the organic phase with U and Pu in the first cycle and, later,
separated during the second purification cycle of uranium. Neptunium can be
transmuted by blending with MOX fuel (homogeneous recycle) or fabricated into
special targets (heterogeneous recycle) for irradiation. Unlike neptunium, the other
minor actinides, americium and curium cannot be separated by modifying the PUREX
process. Processes such as the TRUEX process, described previously, must be designed
specifically to remove americium and curium (trivalent actinides) from HLW.
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Technetium and iodine can also be separated by reasonable modifications to the
traditional PUREX process. During dissolution, 10–20% of the technetium remains as
an insoluble residue with the noble metals Ru, Rh, and Pd with the cladding hulls. The
technetium that does dissolve in the nitric acid is Tc(VII) as TcO4– which can be
extracted by TBP. Iodine is released as a gas in the dissolution step and is separated by
trapping it from the off-gas. A flow diagram showing the removal of Np, I, and Tc from
such a modified PUREX process is shown in Figure 13.
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Figure 13: The PUREX process modified for the separation of Np, I, and Tc.
12.2. UREX and UREX+ Processes
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A variation of the PUREX process that has been proposed as a possible alternative
partitioning scheme for the transmutation of wastes is the UREX process. In this
aqueous based process, shown schematically in Figure 14, uranium and technetium are
co-extracted from a nitric acid solution containing the dissolved spent commercial LWR
fuel with plutonium in the HLW stream with the other minor actinides and fission
products. UREX, like the PUREX process, uses tri-butyl phosphate (TBP) as the
extractant. However, in the UREX process, uranium is extracted at a high separation
factor in one extraction cycle. The distinguishing feature of the UREX process is the
addition of a reductant/complexant, acetohydroxamic acid (AHA), to the dissolver
solution. The addition of AHA results in the suppression of plutonium extraction and in
the retention of certain fission products (e.g., Mo, Zr, and Ru) that would otherwise
contaminate the organic phase. Technetium is back-extracted into a high acidity
aqueous phase, yielding a relatively pure technetium stream. Uranium can then be
stripped from the organic phase. The UREX liquid waste stream or raffinate contains all
of the transuranic elements and all of the fission products except technetium, iodine,
xenon and krypton. The separation requirements for the UREX process development
include recoveries of uranium and technetium at greater than 99.9% and 95%
respectively. (Note: Iodine is released as a gas in the dissolution step and is separated
by trapping it from the off-gas line.)
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Figure 14: Flow diagram for the UREX process.
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The UREX+ process is a series of five solvent-extraction flowsheets that perform the
following operations: (1) recovery of Tc and U (UREX), (2) recovery of Cs and Sr
(CCD-PEG), (3) recovery of Pu and Np (NPEX), (4) recovery of Am, Cm, and rareearth fission products (TRUEX), and; (5) separation of Am and Cm from the rare earth
fission products (Cyanex 301). The goals of the separation process development for
UREX+ include:
•
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Uranium
o
o
o
o
Recovery ≥ 90%
Purity to allow disposal as low-level waste
Fission products to meet regulatory limits and TRU < 100 nCi/g (3700 Bq g-1)
If U is to be recycled, additional separations would be required to increase
the purity to the acceptable standard
• Technetium
o Recovery ≥ 95%
o Purity only important for use as transmutation target
• Cesium/Strontium
o Recovery ≥ 97%
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o Contribution to the heat load in the repository equal to that of all other fission
products
o Purity to allow ultimate disposal as low-level waste after decay storage
• Plutonium/Neptunium
o Recovery ≥ 99% to allow 100-fold reduction of heat load to repository
o Purity to meet mixed-oxide (MOX) fuel specifications
• Americium/Curium
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o Recovery ≥ 99.5% to allow 100-fold reduction of heat load to the repository
o Purity to meet fast-spectrum reactor lanthanide specification of < 20 mg/g
Uranium plus TRU
• Raffinates from the UREX+ process
o TRUEX, all soluble fission products but Cs, Sr, Tc, I, and REEs
o Cyanex 301, RE elements
o Purity requirements are basis for recoveries in other streams
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The flow diagrams for steps 2 and 3, and steps 4 and 5, in the UREX+ process are
shown in Figures 15 and 16 respectively.
Figure 15: The flow diagram for Step 2 (recovery of Cs and Sr using CDC-PEG, a
mixture of chlorinated cobalt dicarbollide (CCD) for cesium extraction and
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polyethylene glycol (PEG) for strontium extraction diluted by phenyltrifluoromethyl
sulfone) and Step 3 (for recovery of a pure Pu/Np product) using the NPEX process.
Figure 16: The flow diagram for Step 4 (recovery of Am, Cm, and rare-earth fission
products using TRUEX) and Step 5 (separation of Am and Cm from the rare earths
using Cyanex 301).
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If UREX+ can be successfully used, it has the advantage that iodine and technetium are
captured for safe disposal, uranium is recovered as U 3O8 for recycle or disposal as LLW
Class C waste, Np and Pu are recovered together for blending into MOX fuel for
thermal reactors, Am and Cu are separated together and can be transmuted in a fast
reactor, Cs and Sr are removed from the waste stream going to a repository, greatly
reducing the heat load in the waste form, and, finally, the remaining fission products are
stored in the repository.
13. Non-Aqueous Chemical Processing
One of the most challenging R&D areas impacting waste management for the advanced
fuel cycle is the development of non-aqueous processes. Because of the higher burn-up
needed for transmutation as well as the need for multiple recycling with minimum
cooling time between fuel discharge and reprocessing, non-aqueous processes offer
important advantages. Non-aqueous processes are much more radiation resistant than
aqueous processes and can be used to reprocess spent fuel after considerably shorter
cooling periods. Other advantages include the potential to produce much smaller
quantities of secondary waste and waste forms suitable for long-term disposal, the ease
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of recycling of reagents used in separations, and finally, the compactness of equipment
which reduces the physical size of the plant needed. However, the separation factors are
usually smaller, requiring multiple stages to achieve relatively high levels of
decontamination. Also, the processes must be conducted under highly controlled
atmospheres (e.g., dry argon) to avoid hydrolysis and precipitation reactions. Nonaqueous systems typically have a rather limited throughput as they are usually operated
in batch rather than continuous processes. However, this is not a severe limitation in a
transmutation scenario as after the initial processing of the spent fuel, the volume of
material to be processed in the transmutation cycle is only a fraction of the initial fuel
volume.
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Currently, non-aqueous separation schemes are under investigation in several countries
including the US, Russia, and Japan. However, the most advanced non-aqueous
processing scheme that has been demonstrated on an engineering scale is the
electrometallurgical process developed by Argonne National Laboratory (ANL). This
process, developed for metallic fuels, is described in detail in the earlier section entitled
Electrochemical Separations using Non-aqueous Processes. However, by changing the
head end treatment steps, the ANL electrometallurgical process can be modified to
accommodate other fuel types such as oxides and nitrides. For example, an initial
Li/Li 2 O reduction step can be used to convert oxide fuel used in LWRs to the metallic
form. Other such modifications could be made to allow this process to be used in
situations where cooling periods as short as 3–5 months are required.
14. Transmutation Devices for the Advanced Fuel Cycle
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For the AFC to be most effective in reducing the inventory of minor actinides at a
reasonable rate, dedicated devices that produce a “fast” neutron spectrum are required.
Such devices include the advanced liquid metal reactors (ALMR), a fast reactor
configured to operate as an actinide “incinerator” rather than as a “breeder” and
accelerator-driven systems (ADS). Fast reactor technology, discussed earlier in this
chapter, is relatively mature whereas the development of ADS is not. Accelerator-based
waste transmutation programs are being developed in France, Japan, USA, and by
CERN. An ADS facility shown in Figure 17 consists of (1) a high power proton
accelerator; (2) a subcritical target that produces a high neutron flux upon bombardment
with high-energy protons; (3) a blanket system that utilizes this intense source-driven
neutron flux to fission actinides and transmute fission products; and (4) a chemical
processing plant to support the ADS operations.
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Figure 17: Schematic of and Accelerator-Driven System (ADS) for transmutation of
waste.
Courtesy: World Nuclear Association (http://world.nucelar.org/sym/1999/venneri.html)
The Uranium Institute, Twenty Fourth Annual International Symposium 1999,
“Disposition of Nucelar Wastes Using Subcritical Accelerator-Driven Systems” by F.
Venneri, M. Williamson, N. Li, M. Houts, R. Morley, D. Beller, W. Sailor, and G.
Lawrence
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The central part of the system is the subcritical target where an intense (> 100 mA),
high-energy (1-2 GeV) proton beam impinges on a high Z target to produce neutrons in
a process called spallation. The effectiveness of an ADS target is usually given by the
number of neutrons produced per incident proton and is linearly proportional to both the
atomic number and density of the target. For this reason targets such as molten lead, or
lead/bismuth eutectic are being proposed as the spallation source. Accelerator-driven
systems producing 40 or more neutrons per incident proton are also being proposed.
The waste material is loaded into the blanket where it is “incinerated” by capture of
neutrons causing it either to fission or to be converted to short-lived or stable isotopes.
This process, therefore, reduces the radiotoxic inventory of the waste material. The heat
produced from the transmuter, Figure 18, (the combination of the target and blanket)
can then be used to generate power, ~10% of which can be used to run the accelerator
and the remaining 90% can be put on the power grid. Unlike fast reactors that operate
on the criticality principle (at least one of the neutrons emitted during a fission event
goes on to cause another atom to fission) to sustain nuclear reaction, ADS systems are
subcritical. When the proton beam is off in an ADS, no neutrons are created and no
nuclear reaction occurs.
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Figure 18: Schematic of a transmuter for an ADS.
Courtesy: World Nuclear Association (http://world.nucelar.org/sym/1999/venneri.html)
The Uranium Institute, Twenty Fourth Annual International Symposium 1999,
“Disposition of Nucelar Wastes Using Subcritical Accelerator-Driven Systems” by F.
Venneri, M. Williamson, N. Li, M. Houts, R. Morley, D. Beller, W. Sailor, and G.
Lawrence
An often overlooked but key component of an ADS is the chemical processing plant
needed to support the operations. Within the plant, separation processes are used to
partition the spent fuel to obtain the waste material that is introduced into the ADS and
also to support multiple recycle, fuel fabrication, and finally, the production of the final
waste forms acceptable for disposal in a geologic repository. Figure 19 shows a flow
diagram for chemical processing plant designed to support transmutation with an ADS.
All of these activities are crucial to the success of transmutation using ADS and
considerable R&D is needed to make this a reality.
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Figure 19: Flow diagram for a chemical processing plant for an ADS.
Courtesy: World Nuclear Association (http://world.nucelar.org/sym/1999/venneri.html)
The Uranium Institute, Twenty Fourth Annual International Symposium 1999,
“Disposition of Nucelar Wastes Using Subcritical Accelerator-Driven Systems” by F.
Venneri, M. Williamson, N. Li, M. Houts, R. Morley, D. Beller, W. Sailor, and G.
Lawrence
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15. Strategies for Implementation of an Advanced Fuel Cycle
In general, the different strategies for implementation of AFC concepts depend on
tradeoffs between the effectiveness of the partitioning steps (chemical yields and
acceptable losses of the desired elements), the degree of transmutation that can be
achieved by various types of devices (burnup rates, neutron spectra and fluxes
available), and the potential for proliferation of the associated fissile material. However,
the two main approaches differ in how the plutonium is managed. In the first, plutonium
is separated and used to create energy and the minor actinides, neptunium and above,
are treated and managed as waste products. In the second, the transuranium elements,
plutonium and the minor actinides are managed as a single unit.
Where plutonium is separated from LWR spent fuel, the utilization of plutonium as an
energy source and the management of the minor actinides as a waste product is a
relatively simple process and the fuel cycle is closed in an evolutionary way. This is
shown schematically in Figure 20. Initially, plutonium is recycled in a light water
reactor as a mixed oxide fuel, LWR–MOX, and later, in fast reactors as FR-MOX fuel.
In this scenario, the fuel cycle is optimized for the production of energy from
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plutonium, especially if the fast reactor is a breeder reactor, but does not reduce the
burden of the MAs and LLFPs on the repository.
Figure 20: Schematic of an advanced fuel cycle scenario where Pu is separated and
recycled to optimize the production of energy from plutonium.
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The second approach which does not separate plutonium and the minor actinides during
the processing of LWR fuel has as its principal driver reduction of the risk of
proliferation of plutonium. To accomplish the transmutation of this waste stream, a
“TRU burner” is required that operates in a closed cycle where the wastes are recycled
multiple times. Further, to be effective, non-aqueous chemical processing is necessary
to minimize the time between irradiations and to shorten the overall recycle time. The
main difference between the two approaches is that, in the former, plutonium would be
removed by conventional wet chemical processing and only the minor actinides are
reprocessed using non-aqueous methods. The options for burning TRU include both
thermal and fast critical reactors and thermal and special, fast-spectrum subcritical
accelerator-driven systems (ADS). Figure 21 is a schematic of one of many possible
P&T schemes that involves both aqueous and non-aqueous partitioning that has as an
ultimate goal, removal of Pu, the MAs, and selected LLFPs from the waste going to a
repository.
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Figure 21: One of several Partitioning and Transmutation (P&T) strategies being
considered to substantially reduce minor actinides produced by conventional LWR
power reactors.
16. Generation IV Nuclear Energy Systems
Nuclear energy is expected to grow considerably over the next 50 years for several
reasons:
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• the demand for energy exceeds the supply even though significant improvements
in energy efficiency and conservation have been made. The demand is increasing
exponentially in Asia;
• energy resource availability is a growing concern;
• decline in air quality and climate change due to global warming caused by
greenhouse gases produced by the burning of fossil fuels for energy continue;
• the need for countries for energy security requires a sustainable nuclear energy
component that is acceptable to the public.
For nuclear energy to become a larger part of an overall energy strategy, other issues
such as reactor safety, economics, and non-proliferation must be considered. These
issues, as well as the waste issue, are being addressed in an initiative of the US by
Department of Energy’s (DOE) Office of Nuclear Energy called the “Generation IV
International Forum for Nuclear Energy Cooperation (GIF).” Charter nations, in
addition to the US, include Argentina, Brazil, Canada, France, Japan, Republic of
Korea, and the United Kingdom. Figure 22 shows the evolution of nuclear reactor
technology starting with the early prototype reactors in the 1950s through the proposed
Generation IV energy systems that are under design. The Generation IV nuclear
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systems are to be ready for deployment by 2030 with standardized designs to expedite
licensing, reduce capital costs and allow rapid construction with simpler and more
robust designs that incorporate passive or inherently safe features such as no active
controls or controls that rely on gravity or natural convective cooling; higher
availability and longer operating life, and higher, more efficient fuel burnup with
greater minimization of waste.
Figure 22: Time line showing the evolution of nuclear energy systems.
Courtesy: US Dept. of Energy, Office of Nuclear Energy
(http://nuclear.energy.gov/genIV/neGenIV1.html)
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A roadmap has been developed that defines the next generation nuclear energy systems
to achieve the required safety performance, waste reduction, and proliferation
resistance, demanded by the public at an economically competitive cost. The six most
promising energy systems that have been identified for further development include:
Thermal-spectrum systems
• Very-high temperature reactor (VHTR)
• Supercritical-water-cooled reactor (SCWR)
Fast-spectrum systems
• Gas-cooled fast reactor (GFR)
• Lead-cooled fast reactor (LFR)
• Sodium-cooled fast reactor (SFR)
Liquid-fuel system
• Molten salt reactor (MSR)
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In the US, the Generation IV program is being coupled to the Nuclear Hydrogen
Initiative to develop hydrogen production technologies compatible with nuclear energy
systems of the future.
16.1. Advanced Fuel Cycle Development to Support Generation IV Energy
Systems
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For an energy strategy involving significant growth in nuclear energy to be successful,
advances must be made in all areas of the fuel cycle. For the deployment of Generation
IV reactors, fuel cycle technologies must be developed that include new fuels, cladding,
and fuel fabrication, improved separation and fuel fabrication processes that support
P&T, better and more robust waste forms and disposal technologies. These technologies
must be:
• economical, having a financial risk comparable to other energy sources,
• inherently reliable, safe, and resistant to proliferation,
• significantly reduce the toxicity of the waste destined for geologic storage by
removal of the transuranic elements, and,
• reclaim the valuable energy remaining in spent nuclear fuel.
16.2. The Advanced Fuel Cycle Initiative (AFCI)
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The US is pursuing a national energy policy which, over time, will significantly
increase reliance on nuclear energy. The US nuclear energy enterprise will transition
from the current once-through fuel option to a fuel cycle based on sustained actinide
recycle that (1) reduces repository burdens for storing nuclear waste, (2) employs
advanced treatment technologies that are more resistant to proliferation, and, (3) more
effectively utilizes nuclear fuel resources. The US through its Advanced Fuel Cycle
Initiative (AFCI) program is developing separation processes that do not produce pure
plutonium. This phased approach, shown in Table 9, includes extending the life of the
current fleet of primarily LWRs while introducing successive generations of more
advanced light water reactors and ultra-high burnup fuels. This approach also includes
the introduction of Generation IV thermal reactors, more proliferation-resistant
technologies for Pu recycle, and then Generation IV fast reactors requiring new fuel
systems and even more advanced separation technologies for destruction of actinides
and long-lived fission products via transmutation.
Implementation:
approx. time
period
Now
Fuel Cycle Phase
2010-2015
Once through fuel
option
2020-2025
Limited closed
fuel cycle
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Disposition Actions
Spent fuel stored at reactor sites
Geologic repositories open
High burn up fuels implemented
Limited fuel recycling
Recycling of weapons useable
material begins
Cs/Sr removal: allowed to decay
RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
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2030-2035
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2035-2040
Geologic disposal of:
U as low level waste
TRU and remaining FPs as HLW
Selected TRU recovery for recycle in
thermal reactors
Thermal recycle
Cs/Sr removal: allowed to decay
of TRU
Geologic disposal of:
U as low level waste
Remaining MAs and FPs as HLW
U recovery for reuse or storage
Cs/Sr removal: allowed to decay
Dedicated burners TRU recovery and recycle in
for TRU
dedicated fast “burner” reactors
Geologic disposal of:
Remaining FPs and HLW
U recovery for reuse or storage
Cs/Sr removal: allowed to decay
Gen IV closed fuel TRU recovery and recycle in Gen IV
cycle
reactors
Geologic disposal of:
Remaining FPs and HLW
Spent fuel continually recycled
Cs/Sr removal: allowed to decay
Transitional
Geologic disposal of:
Recycle
Only remaining FP disposed as
HLW
Recovered U routinely used for fuel
Cs/Sr removal: allowed to decay
Sustainable
Geologic disposal of:
Recycle
Only remaining FP disposed as
HLW
2040-2050
2050
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N
~ 2100
Table 9: Phased strategy for implementation of Advanced Nuclear Fuel Cycles in the
U.S.
Times are only approximate and durations of the phases or the extent of their overlap
are subject to evolution of nuclear future.
Courtesy: US Dept. of Energy, Office of Nuclear Energy, Advanced Fuel Cycle
Program Plan,
May 2005, p.5-3. (http://www.ne.doe.gov/AFCI/neAFCI.html)
Chemical separations technology development for advanced fuel cycle applications
using the AFCI Program as an example are focused on separations processes that
facilitate removal of those constituents of spent fuel that contribute most to the heat load
and waste volume imposed on the disposal of high-level waste in the repository.
Processes requirements include:
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• removal of over 90% of the uranium in sufficiently pure form that it can be
disposed as low-level waste or re-enriched for recycle to LWRs;
• removal of over 99% of the cesium and strontium present in spent fuel, thereby
greatly reducing the short-term heat load;
• separation of the transuranic elements (plutonium, neptunium, americium and
curium) for storage or recycle to LWRs or future advanced reactors for
fissioning, thereby greatly reducing the long-term heat load.
16.3. Areas of Scientific Concerns in the AFCI
Dual-Tier
System
Long-Term
Options
Advanced
processes
Advanced light
water reactor, very
high temperature
reactor
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Single-Tier
System
Front-end fuels
separations
UREX
UREX
Thermal reactor
for recycle
LWR, ALWR
Fuel for thermal
reactors
Mixed oxide
(MOX) With
UREX
Inert matrix fuel
(IMF)
Non-aqueous
pyroprocessin
g or advanced
aqueous
Non-aqueous
Advanced
pyroprocessing or
processes
advanced aqueous
Fast reactor fuel
Metal or MOX
metal or MOX
U
N
Separations for
fast reactor fuel
Fast reactor
Sodium-cooled Sodium-cooled
Nitrides, carbides,
composites
Lead-cooled or
gas cooled fast
reactors;
Advanced systems
(fusion-fission;
accelerator driven
system, molten
salt reactors)
Table 10: Proposed research areas for the advanced fuel cycle program.
Courtesy: US Dept. of Energy, Office of Nuclear Energy, Path to Sustainable Nuclear
Energy, Basic and Applied Research Opportunities for Advanced Fuel Cycles, Sept.
2005, p. 8.
(http://www.science.doe.gov/bes/reports/list.html)
Science areas important in the development of advanced fuel cycles are materials,
separations, modeling and simulation and proliferation resistance of the fuel cycle.
The major challenges in the material science area are maximization of the efficiency in
use of nuclear fuels, minimization of the effects and amounts of wastes that are
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disposed, and reduction of the risks of proliferation of potential weapons materials from
the fuel cycles. The main scientific challenges in materials development for AFCI are in
these areas:
• data on relevant thermodynamic and thermophysical properties which require
theoretical systems to predict such data from alloys, chemical complexes,
colloids, etc., at relevant temperatures of reactor operation, of conditions in
waste storage facilities, and in operations in separations plants;
• solid/solid and solid/liquid interface interactions in reactor operation, and in
separations and waste disposal;
• radiation effects in all these systems.
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The separation systems require much fundamental research to achieve an adequate
knowledge of the important factors to ensure success in their use. These include:
• a valid understanding of the ionic and molecular behavior of the multicomponent fluids;
• design, synthesis and study of the materials (i.e., complexants, etc.) for highly
selective reagents;
• sufficient understanding of the effects of radiation and interfacial phenomena in
the separation systems.
Critical needs in modeling and simulation are methods to deal with the chemistry and
physics by different models. These must involve widely varying temporal and spatial
scales, and methods to deal with uncertainties including propagation of errors in the
data and in models. The scientific challenges facing modeling and simulation for AFCI
can be placed in the following areas:
U
N
• improve nuclear data covariance matrices and determine precise actinide cross
sections,
• develop modeling for materials in extreme environments such as high radiation
fields and elevated temperatures,
• develop models for advanced separations of spent fuels which minimize waste,
• model the entire fuel cycle with uncertainties in the data and the models.
Nuclear fuel management systems in present use have certain proliferation risks. The
growth, internationally, of spent nuclear fuel and of separated plutonium is of growing
global concern in this area.
A definite and workable program is needed in global management of nuclear materials.
A number of factors can be cited as targets for improvement which are several “top
level” goals and require application of new science and technology.
• strengthening safeguards technology by improving means of reducing the risk of
diversion by advanced radiation monitoring detection systems, specific sensors
to monitor chemical process conditions and chemical effluents,
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• decreasing the purity of nuclear materials in fuels and bulk materials (isotopic
and chemical composition).
17. Future of P&T
Japan, France and the US are the most active pursuers of partitioning and transmutation
technologies. In addition, several international groups (NEA, IAEA, and the EC) are
engaged in such studies. Each organization regularly publishes their progress, status and
future activities. Many of these reports are available on the respective websites.
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Appendix: Additional Bibliography
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TACHIMORI, S., NAKAMURA, H., 1982b,Extraction of some elements by a mixture of DIDPA-TBP
and its application to actinide partitioning process, J. Nucl. Sci. Technol., 19(4), pp 326-33
The League of Women Voters Education Fund, The Nuclear Waste Primer 1993 Revised Edition, Lyons
& Burford, New York (1993): pp 25-26.
United States Code, title 42, section 10101 (42 U.S.C. § 10101)
U.S. Code of Federal Regulations, Title 10, Part 60.2 (10 C.F.R. 60.2)
United States Code, title 42, section 2014 (42 U.S.C. § 2014)
United States Code, title 42, section 2021 (42 U.S.C. § 2021)
VANDEGRIFT, G. F., REGALBUTO, M.C., AASE, S.B., ARAFAT, H.A., BAKEL, A.J., BOWERS,
D.L., BYRNES, J.P., CLARK, M.A., EMERY, J.W., FALKENBERG, J.R., GELIS, A.V.,
HAFENRICHTER, L.D., LEONARD, R.A., PEREIRA, C., QUIGLEY, K.J., TSAI, Y., VANDER
POL, M.H., and LAIDLER, J.J., 2004, Lab-Scale Demonstration of the UREX+ Process, Waste
Management ’04 Conference, Tucson, AZ
VITORGE, P.,1984, Lanthanides and Trivalent Actinides Complexation by Tripyridyl, Triazine,
Applications to Liquid-Liquid Extraction, CEA-R-5270
WANNER, H., Forest, I. Editors, 1992, Chemical Thermodynamics of Uranium, North Holland,
Amsterdam.
WEAVER, B., F.A. KAPPELMANN, 1964, Talspeak: A New Method of Separating Americium and
Curium from Lanthanides by Extraction from an Aqueous Solution of Aminopolyacetic Acid
Complex with a Monoacidic Phosphate or Phosphonate, ORNL-3559, Oak Ridge, Tenn, Oak
Ridge National Laboratory.
WEIGL, M., GEIST, A.,M GOMPPER, K., KIM, J.I., 2001, Kinetics of Lanthanide/Actinide Coextraction with N,N’-dimethyl, N,N’-dibutyltetradecylmalonic diamide (DMDBTDMA). Solvent
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YAKOVLEV, G.N., GORBENKO-GERMANOV, D.S., 1956, Coprecipitation of Americium (V) with
Double Carbonates Uranium (VI) or Platinum (vi) with Potassium, Proc. 1st United Nations
International Conference on Peaceful Uses of Atomic Energy, Vol. 7, Vienna, International
Atomic Energy Agency.
YOUNG, W. A., 1994, Hazard-Based Classification of Nuclear Waste – A Wiser Arrangement, Nuclear
Guardianship Forum, #3.
ZHU, Y., SONG, C., XU, J., YANG, D., LUI, B., CHEN, J., 1989, The Removal of Actinides from High
Level Radioactive Waste by TRPO Extraction. The Extraction of Americium and Some
Lanthanides, Chinese J. Nucl. Sci. Eng., Vol . 9, pp 141-150.
ZHU, Y., 1995, The Separation of Americium from Light Lanthanides by Cyanex-301 Extraction,
Radiochim. Acta, 68, pp 95-98.
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Glossary
A:
ADS:
AFC:
AFCI:
AHA:
ALMR:
ANL:
Bq:
BRC:
Mass number, sum of the number of protons and neutrons in the
nucleus.
Accelerator-driven system.
Advanced Fuel Cycle.
Advanced Fuel Cycle Initiative.
Acetohydroxamic acid, complexing agent used in the UREX
process to suppress the extraction of plutonium.
Advanced Liquid Metal Reactor.
Argonne National Laboratory.
Abbreviation for Becquerel, the SI derived unit for radioactivity
defined as the activity of a quantity of radioactive material in
which one atom decays per second (s-1).
Below regulatory concern or de minimis refers to material
containing very low concentrations of radioactivity such that the
associated radiological hazards are generally considered to be
©Encyclopedia of Life Support Systems (EOLSS)
RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
Breeder reactor:
BWR:
CERN:
E
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SC
M
O
PL
-E
E
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AP
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TE
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CSEX:
Cooling time:
Decommissioning
of a nuclear
reactor:
DIAMEX:
EBR-II:
EC:
eV:
EW:
Fissile nuclear fuel:
FBRs:
negligible since they are less than for naturally-occurring
radioactivity.
A nuclear reactor that produces more fissile nuclear fuel that it
consumes.
Boiling water reactor.
Organisation Européenne pour la Recherche Nucléaire
(European Organization for Nuclear Research) - World’s
largest particle physics laboratory located near Geneva on the
border between France and Switzerland.
Chemical process designed to selectively extract cesium.
The amount of time spent nuclear fuel is allowed to decay in
pools of water before handling for the purpose of chemical
reprocessing or dry storage.
Involves removing the spent fuel (the fuel that has been in the
reactor vessel), dismantling any systems or components
containing activation products (such as the reactor vessel and
primary loop), and cleaning up or dismantling contaminated
materials from the facility.
Extraction process for improved separation of actinide from
acidic solutions using bifunctional diamides.
Experimental Breeder Reactor – II.
European Community.
Electron volt.
Exempt waste – waste that is identified as below regulatory
concern or de minimis.
Fuel that can be easily made to undergo the nuclear fission
reaction.
Fast Breeder Reactors – reactors that breed new fuel using fast
neutrons.
Fast reactor mixed oxide fuel.
Gas-cooled fast reactor.
Generation IV International Forum for Nuclear Energy
Cooperation.
Gases in the atmosphere that absorb light radiated from the
earth’s surface and lead to global warming.
Gigawatts of electrical energy.
High-level waste.
Term used interchangeably with transmutation.
International Atomic Energy Agency.
Intermediate level waste.
Distribution of isotopes such as in spent nuclear fuel.
Kiloelectron volts.
Los Alamos Molten Plutonium Reactor Experiment.
Los Alamos National Laboratory.
Lead-cooled fast reactor.
Long-lived fission products.
Low-level waste.
U
N
FR-MOX:
GFR:
GIF:
Greenhouse Gas:
GWe:
HLW:
Incineration:
IAEA:
ILW:
Isotopics:
keV:
LAMPRE:
LANL:
LFR:
LLFP:
LLW:
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
LWR:
LWR:MOX:
MA:
MeV:
MOX:
MSR:
MWth:
MWe:
NEA:
NORM/NARM:
E
SA
SC
M
O
PL
-E
E
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NPEX:
Light water reactors.
Light water reactor mixed oxide fuel.
Minor actinides, i.e. actinides other than U and Pu.
Million electron volts.
Mixed oxide fuel.
Molten salt reactor.
Megawatts of thermal energy.
Megawatts of electrical energy.
Nuclear Energy Agency (France).
Naturally Occurring and Accelerator produced Radioactive
Material.
Neptunium extraction – chemical process designed to
selectively extract neptunium.
A complex series of aqueous chemical and/or non-aqueous
metallurgical operations used to selectively remove or partition
the minor actinide and long-lived fission products from highlevel waste.
Partitioning and Transmutation.
Plutonium Redox Extraction – chemical process designed to
selectively extract plutonium by changing its oxidation state.
Pressurized water reactor.
Pertaining to chemical reactions carried out at high temperatures
such as in a molten salt.
A measure of the toxicity or the harmful potential of a quantity
radioactive material assess in terms of dose received.
In the PUREX process, the raffinate is the solution containing
the dissolved spent nuclear fuel that remains after the extraction
of plutonium and uranium.
Rare earth elements also known as the lanthanides.
Reprocessing Fuel Cycle.
Supercritical-water-cooled reactor.
Sodium-cooled fast reactor.
Spent Nuclear Fuel.
Process by which a heavy nucleus emits a large number of
nucleons (primarily neutrons and protons) as a result of being hit
by a high-energy (several 100s of MeV or greater) proton.
Chemical process designed to selectively extract strontium.
A reactor in which a self-sustaining nuclear reaction can not be
maintained; thus, when the source of neutrons is removed the
reactor shuts down.
Trivalent Actinide Lanthanide Separation by Phosphorus
Extractants and Aqueous Komplexes.
The transformation of a minor actinide or long-lived fission
product into another nuclide that is either stable or has a shorter
half-life and thus, is less of a problem than the original
radionuclide.
Process for separating tri- and tetravalent actinides from
Partitioning:
P&T:
PUREX:
PWR:
Pyrochemical:
Radiotoxicity:
Raffinate:
U
N
REEs:
RFC:
SCWR:
SFR:
SNF:
Spallation:
SREX:
Subcritical reactor:
TALSPEAK:
Transmutation:
TRAMEX:
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RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
TRU:
TRU burner:
TRUEX:
E
SA
SC
M
O
PL
-E
E
O
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UREX:
UREX+:
US NRC:
USDOE:
VHTR:
Vitrification:
lanthanides and most other fission products using liquid cation
exchangers, such as trialkylamines and tetraalkylammonium
salts dissolved in organic solvents.
Transuranium elements.
A reactor or accelerator-driven device designed to efficiently
reduce the inventory of transuranium elements by fissioning.
Transuranium element extraction – separation scheme to recover
the trivalent actinides and trivalent rare-earth (lanthanide)
fission products.
Uranium extraction.
Uranium extraction plus addition separation steps.
United State Nuclear Regulatory Commission.
United State Department of Energy.
Very-high temperature reactor.
Process by which nuclear waste is incorporated into a noncrystalline or glass-like matrix.
Bibliography
Interactive Chart of the Nuclides from the National Nuclear Data Center at the Brookhaven National
Laboratory. Website address: nndc.bnl.gov/chart [This site provides quick and easy access to nuclear data
for known isotopes such a half-life, decay modes, etc].
International Atomic Energy Agency (IAEA). Web site address: www.iaea.org [This United Nations site
is a comprehensive source of detailed, high quality international nuclear and radiation related
information. Search this site for articles and reports on a wide variety of nuclear related topics such as
nuclear waste management, transportation of nuclear materials, nuclear power world-wide, issues related
to nuclear non-proliferation, etc].
Nuclear Energy Institute (NEI). Web site address: www.nei.org. [This site provides a general overview of
nuclear energy information in an easily understood format].
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Nuclear Waste Management Organization of Canada (NWMOC). Web site address: nwmo.ca
[Information on the strategy being used by Canada to handle radioactive waste from nuclear power
generation].
Nuclear Instruments and Methods (A) Journal. Plendl, H.S., Editor, 1994, Special Issue on Nuclear
Transmutation Methods and Technologies for Disposition of Long-lived Radioactive Materials, ,
Vol.414, 1-126 and Plendl, H.S., Editor, 2001 Special Issue on Accelerator Driven Systems, Vol. 463,
425-666. [These two dedicate issues of this journal contain a number of excellent articles containing
technical information on methods and technologies for disposition of long-lived nuclear waste].
Organization for Economic Co-operation and Development, Nuclear Energy Agency (OECD/NEA). Web
site address: nea.fr. [This site is a valuable resource for technical information covering research and
development in the area of nuclear waste disposal. See especially proceedings from annual International
Information Exchange Meetings on Actinide and Fission Product Partitioning and Transmutation].
U.S. Department of Energy (DOE). Web site address: www.eia.doe.gov [This very large site provides
comprehensive data on energy use throughout the U.S. with links to numerous sites for specific energy
information. Also see DOE Office of Civilian Radioactive Waste Management at web site
ocrwm.doe.gov/; DOE, Basic Energy Sciences Workshop, 2005, The Path to Sustainable Nuclear Energy
Basic and Applied Research Opportunities for Advanced Fuel Cycles, September 2005, available from
http://www.er.doe.gov/bes/reports/files/PSNE_rpt.pdf; DOE Office of Nuclear Energy, Science and
Technology, Office of Advanced Nuclear Research. Generation IV Nuclear Energy Systems, Ten-year
Program
Plan,
Fiscal
Year
2005,
March
2005,
available
from
http://nuclear.energy.gov/genIV/neGenIV1.html].
©Encyclopedia of Life Support Systems (EOLSS)
RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
Uranium Information Center (UIC). Web site: uic.com [This site, maintained by the Australian Uranium
Association, contains a wide variety of briefing materials and educational resources on nuclear energy
topics].
US Nuclear Regulatory Commission (NRC). Web site address: nrc.gov/waste [This site provides
information about types of regulated waste and activities, responsibilities and related information].
Wastelink and Radwaste.org. Website address: radwaste.org/ [Site are maintained by Herne Data Systems
Ltd includes more than 10,000 links to radwaste, nuclear, radiation and environmental related companies,
research centers, regulatory agencies, government organizations, non-governmental organizations, etc.].
Wikipedia Encyclopedia. Website address: en.wikipedia.org [A wide variety of nuclear topics can be
accessed by searching this site by topic].
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World Nuclear Association. Web site address: world-nuclear.org. [This site provides comprehensive and
factual information covering a wide variety of nuclear-related topics from around the world. The
information is easily understood and updated frequently].
Biographical Sketches
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Patricia Ann Baisden received her BS degree (with honors) in 1971 and her Ph.D. in chemistry in 1975
under the direction of Prof. Gregory R. Choppin at the Florida State University. After completing her
graduate degree, Dr. Baisden accepted a postdoctoral appointment with Nobel Laureate, Glenn T.
Seaborg at the Lawrence Berkeley Laboratory. Since 1978, she has worked for the Lawrence Livermore
National Laboratory in California holding a number of technical and management positions including
Deputy Director of the Glenn T. Seaborg Institute for Transactinium Science and Division Leader,
Materials Program Leader, Chief Scientist, and Deputy Associate Director in the Chemistry & Materials
Science Directorate. Her expertise is in nuclear and radiochemistry; actinide chemistry, and lasers and
optical materials and over her career she has authored or co-authored over 50 journal papers. Dr. Baisden
has also held a number of professional positions including serving as Chairman of the American
Chemical Society’s Division of the Nuclear Chemistry and Technology, as a member of the Department
of Energy (DOE)/National Science Foundation (NSF) Nuclear Science Advisory Committee, and as an
editor of Radiochimica Acta, the International Journal for the Chemical Aspects of Nuclear Science and
Technology. Additionally, she has been appointed to a number of National Academy of
Sciences/National Resource Council study committees including the Committee on Environmental.
Technologies for Decontamination and Decommissioning, Committee on Electrometallurgical
Techniques for DOE Spent Fuel and Excess Weapons Plutonium Disposition; Committee on the
Assessment of the National Institutes of Standards and Technology (NIST) Physics Laboratory, and the
Chemical Sciences Roundtable. Dr. Baisden is also credited with starting in 1983 the "Summer School in
Nuclear Chemistry," an Undergraduate Fellowship Program for DOE. This program, which will be held
for its 25th consecutive year in 2007, has to date, introduced over 550 undergraduate students to the field
of nuclear and radiochemistry. Currently Dr. Baisden holds the position of Assistant Associate Director
in the National Ignition Facility (NIF) Programs Directorate at LLNL and is on special assignment to the
Office of Inertial Confinement Fusion with the DOE/National Nuclear Security Administration in
Washington, DC.
Gregory R. Choppin received a Maxima Cum Laude Bachelor of Science degree for Chemistry from
Loyola University, New Orleans, Louisiana, in 1949 and a Ph.D. from the University of Texas, Austin,
in1953. He then began postdoctoral research with Glenn T. Seaborg in the new element research group at
the Lawrence Radiation Laboratory, University of California-Berkeley. In 1955, he accepted a permanent
position in the Lawrence Laboratory and was one of the four co-discovers of element 101, Mendelevium.
These experiments were the first on single atoms as only one atom of Mendelevium was made per hour
of irradiation in the cyclotron. In 1956, he moved to Florida State University as an Assistant Professor of
Chemistry. At FSU, he initiated a research program in f-element chemistry, studying both lanthanide (4f)
and actinide (5f) behavior. He spent sabbatical years at the Belgian Nuclear Center (Mol) in 1962-63 and
at the German Nuclear Center (Karlsruhe) in 1979-80. He has spent shorter periods for research in
nuclear laboratories in France, Portugal, Sweden, and Japan. His research has resulted in over 400 journal
publications and 8 books. A major focus of his research has been the thermodynamics and kinetics of f-
©Encyclopedia of Life Support Systems (EOLSS)
RADIOCHEMISTRY AND NUCLEAR CHEMISTRY – Nuclear Waste Management and the Nuclear Fuel Cycle - Patricia. A.
Baisden, Gregory R. Choppin
element complexation reactions. Since 1975, he has focused attention increasingly on the environmental
behavior of the actinide elements. Recognition of Professor Choppin’s research is reflected in a number
of awards among which are American Chemical Society Award in Nuclear Chemistry; Presidential
Citation Award, American Nuclear Society; Chemical Pioneer Award, American Institute of Chemistry;
Award in Chemical Education, Manufacturing Chemists Association; Award in Rare Earth Chemistry,
Rare Earth Society, Japan; Becquerel Medal in Radiochemistry, British Royal Society of Chemistry;
Honorary Doctor of Science, Loyola University, New Orleans; Honorary Doctor of Technology,
Chalmers University, Sweden; Alexander von Humboldt Prize, Germany; Hevesy Medal for Nuclear
Research, Hungary. In 1969, he was given the honor of becoming an R.O. Lawton Distinguished
Professor at Florida State University. In 2001, Professor Choppin became a Professor Emeritus.
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To cite this chapter
Patricia. A. Baisden and Gregory R. Choppin,(2007),NUCLEAR WASTE MANAGEMENT AND
THE NUCLEAR FUEL CYCLE, in Radiochemistry and Nuclear Chemistry, [Ed. Sándor Nagy], in
Encyclopedia of Life Support Systems (EOLSS), Developed under the Auspices of the UNESCO, Eolss
Publishers, Oxford ,UK, [http://www.eolss.net]
©Encyclopedia of Life Support Systems (EOLSS)