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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
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 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
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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
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 non-radioactive waste but the third is unique to non-radioactive waste. Eventually all non-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
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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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.
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)
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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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)
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.
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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Figure 1: From the IAEA publication: Classification of Radioactive Waste, A Safety Guide, Safety Series No 111-G1.1 (1994).
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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principal classification along with HLW, ILW, and LLW.
In the United States waste classifications are defined by the US Nuclear Regulatory Commission (NRC):
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 Definition
Class A
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
Class B
Higher concentrations of radioactivity than Class A and requires greater isolation and packaging (and shielding for operations) than Class A waste.
Class C
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)
Greater than Class C
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.
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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 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 or .
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
235
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3.2×10–11 joules) for every atom that fissions. The partitioning of the energy resulting from the fission of 235
is shown in Table 2. U Description Approx. Energy (MeV)
Kinetic energy of fission fragments 164.5
Kinetic energy of prompt neutrons (2.5 neutrons/fission event with a average energy of ~2 MeV)
5.0 Prompt
Energy
prompt γ-rays 7.0
Subtotal 176.50
Kinetic energy of β-particles
emitted in decay of fission products 6.5
Neutrinos 10.5 Delayed
Energy
γ-rays 6.5
Subtotal 23.5
Total Energy/fission of an atom of U-235
in MeV 200.0
Table 2: Partitioning of energy in the fission of 235U
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 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 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.
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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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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.
Reactor
Type Country Number GWe Fuel Coolant Moderator
Pressurized Water Reactor (PWR) US, France, Japan, Russia 268 249 enriched UO2 water water Boiling Water Reactor (BWR) US, Japan, Sweden 94 85 enriched UO2 water water Gas-cooled Reactor (Magnox & AGR) UK 23 12 natural U (metal), enriched UO2 CO2 graphite Pressurized Heavy Water Reactor “CANDU” (PHWR)
Canada 40 22 natural UO2 heavy
water heavy water Light Water Graphite Reactor (RBMK) Russia 12 12 enriched UO2 water graphite Fast Neutron Reactor (FBR) Japan, France, Russia 4 1 PuO2 and UO2 liquid sodium none Total 441 381
Table 3: Types of commercial nuclear reactors.(GWe=capacity in thousands of megawatts)Courtesy of the World Nuclear Association
(http://www.world-nuclear.org/info/inf32.html)
5.1. Commercial Nuclear Power Generation
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(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% )
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 ( ) pellets typically ~1 cm in diameter and ~1.5 cm in length.
235
U
2
UO
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 and along with ( = 2.4×107
a) from neutron capture of ;
235
U 238U 236U T1/ 2
235
U
• isotopes of the actinide elements heavier than uranium including Np, Pu, Am
and Cm.
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 Component Disposition
95.6 Uranium Recycled or disposed of as low-level,
Class C waste
3
Stable or short-lived fission products
Decays in a short time resulting in no significant disposal problem
0.3 Major fission
products
Cesium and strontium are the primary near-term heat sources in HLW: decays in a few centuries
0.1 Long-lived fission
products
Primarily iodine and technetium: can be stored or removed and
transmuted*
0.9 Plutonium Can be separated and burned as fuel
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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
11
Some of the isotopes of the transuranium elements ( ) 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
loaded are consumed in the production of 111 kg
of 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 , ~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 isotope. Since the probability of fission of with thermal neutrons in a PWR is very low, the loss of is due to the neutron capture reaction that produces
which decays by beta emission ( = 23.5 months) to produce which, again,
beta decays ( = 2 d) to produce ( = 24 000 a). Using this logic, the net loss
of corresponds to the production of approximately 674 kg of . At the end of
the reactor cycle, however, only 286 kg of total Pu and higher actinides remain leaving
approximately 387 kg of “missing” . 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” 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 produced by neutron capture reactions during the
reactor cycle. 92 Z > U 236 U 235 U 239 Pu 238 U 238 U 239U 1/ 2 T 239Np 1/ 2 T 239Pu T1/ 2 238 U 239Pu 239 Pu 239 Pu 239 Pu
Nuclides Initial Load (kg) Discharge Mass (kg)
U-235 954 280 U-236 111 U-238 26,328 25,655 U Total 27,282 26,047 Pu-239 56 Pu Total 266 Z > 94 20 Sr-90 13 Cs-137 30 Long-lived FP 63 FP Total 946 Total Mass 27,282 27,279
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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
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 . In reality, however, during
the course of operation of the reactor, some of the captured a neutron to form
( = 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).
most likely captures a neutron and forms which is a fissile nuclide.
239 Pu 239 Pu 240 Pu T1/ 2 240 Pu 241Pu
Actinide Isotope Half-life (years) Kg/year in Spent Fuel*
234 U 2.5 E+5 3.1 E+0 235 U 7.1 E+8 2.1 E+2 236 U 2.4 E+7 1.1 E+2 237 U 1.8 E-2 9.1 E-7 238 U 4.5 E+9 2.6 E+4 237 Np 2.1 E+6 2.0 E+1 239 Np 6.4 E-3 2.0 E-6 236 Pu 2.6 E+0 2.5 E-4 238 Pu 8.6 E+1 6.0 E+0 239 Pu 2.4 E+4 1.4 E+2 240 Pu 6.6 E+3 5.9 E+1 241 Pu 1.3 E+1 2.8 E+1 242 Pu 3.8 E+5 9.7 E+0 241 Am 4.6 E+2 1.3 E+0 242m Am 1.5 E+2 1.2 E-2 243 Am 7.9 E+3 2.5 E+0 242 Cm 4.0 E-1 1.3 E-1 243 Cm 3.2 E+1 2.0 E-3 244 Cm 1.8 E+2 9.1 E-1 245 Cm 9.3 E+3 5.5 E-2 246 Cm 5.5 E+3 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 8 and the corresponding heavy FPs
are 129 .
85 90 95 99
Kr, Sr, Zr, Tc
137 144 147
I, Cs, Ce, and 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.
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 13
( = 30 years) and
( = 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 and , as well as 79 , , and
, see Table 7). 7 Cs T1/ 2 90 Sr T1/ 2 99 Tc 129I Se 93Zr 135 Ce
Fission Product Half-life (a) Thermal Fission (TBq/tU)
99 Tc 2.13 × 105 0.48400 129 I 1.57 × 107 0.00120 135 Cs 3.0 × 106 0.00980 93 Zr 1.5 × 106 0.06800 126 Sn 1.0 × 105 0.02900 79 Se 6.5 × 104 0.01510
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 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.
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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 and 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 in conventional LWRs, the initial
motivation for fast reactors was to breed from 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”
235
U
239
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around the core. In the blanket, would capture fast neutrons and produce more Pu
than consumed in the reactor. In fact, about 70% of all placed in a fast reactor can
be converted to . Since the natural abundance of is 99.3%, this conversion
results in the production of a vast supply of fissile fuel for future centuries.
238 U 238 U 239 Pu 238U
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 or a
mixture of and . 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% .
235 U 239 Pu 238U 238 U
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)
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.
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.
8.1.1. Once-Through Fuel Option
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.
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,
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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)
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 and the minor plutonium isotopes, especially . Commercial
reactor-grade plutonium is nominally about 55–60% and has a relatively high
amount ~20–25% of whereas weapons-grade plutonium contains greater than
95% and a minor amount of .
239 Pu 240Pu 239 Pu 240 Pu 239 Pu 240Pu
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 isotope ( = 14.4 a), builds up at a rate of about 2 parts per
million per month or about 0.5% per year. Since emits a 60 keV
X-ray, additional shielding and remote handling is required during MOX fabrication if
the content becomes greater than 5–6%. As a result, MOX is usually made with
recently reprocessed plutonium.
241 Pu T1/ 2 241 Am 241 Am
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.
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% and high 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.
239 241
Pu + Pu 240Pu
Uranium Recycling
The uranium recovered from the reprocessing of commercial spent fuel usually contains
about 1% , 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 -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
non-uranium alpha activity.] During irradiation, the level of increases over that in
natural uranium. In addition, , not present in natural uranium, is produced by the
decay of 2.85 a isotope. Although the total accumulation of is small, it
decays with a 72 a half-life by α-emission through a chain of other α- and -ray
emitters including the 1.9 a . The decay chain finally culminates in 20 , a high
energy -ray emitter. The long-lived ( = 2.4×107 a) is also produced by the
reaction on but adds little to the -activity. The (n reaction on
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.
235 U α 234 U 232 U 236 Pu 232U γ 228 Th 8Tl γ 236 U T1/ 2 (n, γ) 235U α , γ) 236 U
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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)
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 half-life 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.
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 I are rapid, involving
only electron exchange. However, the changes 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.
II→IV and V→VI
III/IV→V/VI
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:
4+ 2+ 3+ +
2 2
An >AnO >An >AnO
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+ 2
AnO 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.
10. Separations Technologies for the Nuclear Fuel Cycle
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 2
2
UO + 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
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
and 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.
2 2
UO + Pu4+
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., 137Cs) 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.
The diamide N,N’-dimethyldioctylhexyloxyethylmalonamide,
3 8 17 2 2 4 6 13
(CH C H NCO) CHC H OC H (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
group by . 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
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.
3 4 9 CH C H N 3 8 17 CH C H N 2+ 2 AnO 10.3. TRUEX Process
This solvent extraction process is designed to separate transuranic elements from the acidic high-level waste solutions of PUREX. The extractant, octyl(pheny)-N,N-dibutylcarbamoyl-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
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
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.
90
Srand 13
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.
7
Cs
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 Separation Process
DIDPA- di-isodecylphosphoric acid – uses TBP as a modifier to avoid third phase formation
HDEHP – di-2-ethyl-hexyl-phosphoric acid Phosphoric acid based DIDPA,
HDEHP, TBP
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-carbamoylmethyl-phosphine oxide
TRUEX – TRans Uranium EXtraction n-octyl-phenyl-di-isobutyl-carbamoylmethyl-phosphine-oxide (CMPO),strong complexing makes stripping with conventional dilute solutions difficult, requires more powerful stripping agents that increase quantity of secondary waste
Alkyl-phosphine oxides
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
Diamides
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.
Azines TPTZ – Tripyridyltriazine used with
di-nonylnapthalenesulfonic acid (HDNNS) to selectively extract An
Pyridine and Amides Picolinamides – N and O chelating agents that can
selectively extract Ac over Ln but requires high nitrate and low proton concentration
Thio-phosphinic acid based
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
Electrochemical redox with heteropolyions
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.
Crown ethers
SREX – Sr Extraction – uses dicyclohexano 18-crown-6 ether to selectively extract Sr from acidic liquid waste
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.
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
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
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.
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.
