Oxides
High-temperature pretreatments that oxidize uranium dioxide to higher oxides have been under development for decades but not deployed in large-scale industrial reprocessing plants. These processes have been developed with different end purposes in mind:
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Volox—voloxidation, a method of removing volatile fission products from fuel to allow abatement and simplify the dissolution cycle (Goode, 1973).
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OREOX—oxidation reduction oxidation process is a development of the Volox process that uses alternate oxidation reduction processes to achieve higher efficiency of fission product removal (Strausberg et al., 1960).
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AIROX—atomics international reduction oxidation process is an OREOX process to allow pulverization of fuel in fuel rods with holes punched in them. This allows fuel and cladding separation before dissolution (Majumdar et al., 1992).
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DUPIC—direct use of PWR fuel in CANDU process is not strictly a headend to reprocessing; it is a pyrochemical method for fuel reprocessing. The concept allows the removal of some fission products prior to refabrication into fuel (Lee et al., 2012).
These processes essentially use the same chemistry, the oxidation of uranium dioxide to higher oxides, Equation 5.10, with the accompanying decrease in density (~ 11-8.4 g cm− 3), thus allowing fuel pulverization. During this process, depending upon the process conditions, various amounts of tritium, carbon-14, iodine-129, kryption-85, and semivolatiles, including caesium and ruthenium, are evolved. The fuel expansion process can also be used to break open cladding that has been punctured (AIROX process), thereby avoiding the need for a shear process, or the process could be carried out upon sheared fuel.
(5.10)
From experiments with unirradiated and irradiated materials, it is known that the presence of elements that cannot be oxidized in the solid state affects the process chemistry. Examples include the presence of plutonium (Rance and Beznosynk, 2005), thorium (Anderson et al., 1954), MOX or alkaline and rare earth elements (SIMFUEL or irradiated fuels) (Kang et al., 2007) with uranium. These reduce the material reactivity as there is a smaller percentage of uranium to provide the driving force for pulverization and release of fission products. The presence of high contents of these dopant elements can stabilize intermediate phases, increase temperature or oxidant conditions to achieve oxidation, and ultimately at high contents prevent oxidation (Thomas et al., 1993; McEachern et al., 1998).
Experiments oxidizing SIMFUEL with varying portions of simulated fission products show that the presence of > 7% fission products (equivalent to ~ 70 GWd tHM− 1) greatly stabilizes the (U,Fp)4O9 phase1 at 400 °C (Cobos et al., 1998). To achieve oxidation, an increase in temperature to 800 °C is needed.
During the oxidation of mixed uranium plutonium oxide, the material oxidizes to a plutonium-rich dioxide phase and a uranium-rich higher oxide phase, Equation 5.11 (Rance and Beznosynk, 2005). During this type of oxidation test, the solubility of plutonium in the M3O8 phase has been observed to be ~ 3% Pu HM− 1, which means that oxidation of most MOX will result in the formation of plutonium-rich oxide particulates. This reduces the proportion of the plutonium dissolved in nitric acid (Cadieux and Stone, 1980). However, at higher plutonium contents oxidation does not occur and experiments up to 900 °C have shown that unirradiated 20-25% Pu HM− 1 MOX does not result in pulverization. Similar segregation is also observed for the oxidation of mixed uranium thorium oxide and, as thorium oxide does not dissolve readily in nitric acid (Anderson et al., 1954; Goode and Stacy, 1979), results in an increase in difficult to dissolve thorium oxide residues.
(5.11)
The oxidation of spent thermal and fast reactor fuels have been studied at the laboratory scale. These experiments have studied the effect of temperature, feed flow rate, and oxygen concentration upon the extent of fission product removal, for example (Goode, 1973). This type of experiment has shown that near-complete tritium removal is possible at moderate temperatures (480 °C) (Goode et al., 1980). Less complete removal of the noble gases and iodine-129 has been achieved (see Table 5.5).
Table 5.5. Effect of voloxidation conditions upon noble gas and iodine-129 removal (Goode, 1973)
T (°C) | [O2] (%) | Gas flow (cfh) | Volatilised (%) |
---|
85Kr-133Xe | 131I |
---|
450 | 75 | 0.2 | 19 | ~ 8 |
750 | 75 | 1.0 | 69 | ~ 99 |
450 | 25 | 0.2 | 73 | ~ 85 |
750 | 75 | 1.0 | 42 | ~ 90 |
450 | 75 | 1.0 | 36 | ~ 33 |
750 | 75 | 0.2 | 76 | ~ 95 |
450 | 25 | 1.0 | 11 | ~ 38 |
750 | 25 | 0.2 | ~ 99 | ~ 92 |
Oxidation of fuel at higher temperatures, for example, greater than 700 °C, increases the amount of the so-called semivolatile elements that are released (Goode, 1973). Examples of fission products that are evolved in significant quantities are caesium-137, ruthenium-106, antimony-125, and niobium-105. Experiments underpinning the DUPIC cycle development show that oxidation at higher temperatures quantitatively increases the release and number of semivolatile elements (Bateman et al., 2006). These experiments also observe that control of the oxidation conditions can suppress the distillation of caesium and other elements by formation of less-volatile oxides. Oxidation of spent fuel and distillation of fission product elements under high-temperature conditions can be compared with studies of fuel distillation under severe reactor failure, such as Kundsen cell mass spectrometry studies (Capone et al., 1996).
The oxidation of irradiated fast reactor 20% Pu HM− 1 MOX has also been demonstrated (Goode, 1973). These experiments have shown that despite the difficulties in the oxidation of unirradiated 20% Pu HM− 1 MOX (Rance and Beznosynk, 2005), the oxidation of irradiated MOX is achievable. Removal efficiencies of noble gases of up to 98% at 750 °C has been demonstrated, which is excellent considering a portion of the noble gases are likely to be trapped in plutonium-rich oxides and not released. Comparing these results with irradiated uranium oxide experiments suggests good removal efficiencies of tritium, carbon-14, and iodine-129 will also be possible. This type of test clearly shows oxidation of 20% Pu HM− 1 MOX is possible.
More recent advanced voloxidation studies have focused on the use of strong oxidants such as nitrogen dioxide and ozone (DelCul, 2010). These oxidants are capable of oxidizing uranium dioxide to trioxide, for example, Equation 5.12, and show promise that near-quantitative removal of tritium, carbon-14, noble gases, and iodine-129 are possible at temperatures lower than air or oxygen.
(5.12)
Carbides and nitrides
The conversion of carbides and nitrides to oxides could be an important step to allow
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Carbides—removal of carbon that prevents the generation of organics during dissolution.
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Nitrides—evolution of nitrogen-15 predissolution for recovery.
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Carbides and nitrides—the removal of fission products for abatement.
These processes can be carried out with a variety of oxidants and temperatures. The conditions necessary tend to be more moderate compared to oxide due to the reactivity of the materials. However, the reactivity also leads to concerns over pyrophoricity of the materials, which means careful control to prevent ignition is a key consideration.
The oxidation of uranium carbide and nitrides with oxygen or air leads to the formation of triuranium octoxide. The accompanying density change is even greater than that for uranium dioxide oxidation, so although these reactions have not been studied in as much detail as the oxidation of uranium dioxide, the uranium dioxide oxidation literature can be used as a starting point. This leads to two conclusions:
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Efficient removal of volatile fission products should be achievable.
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Segregation of uranium and plutonium phases will lead to incomplete plutonium recovery during dissolution in nitric acid. The extreme case for carbide is illustrated in Equation 5.13.
(5.13)
However, there is a significant density change in the conversion of carbides or nitrides to dioxides, 13.6 or 14.3-11.0 g cm− 3, respectively. This density change is sufficient to cause pulverization of pellets. This could be used to as a pretreatment process to convert mixed uranium plutonium carbides or nitrides to oxides, with high plutonium solubility. This type of process has been demonstrated for unirradiated mixed uranium plutonium carbides using carbon dioxide (Murbach and Turner, 1963) and steam (Flanary et al., 1964). These processes proceed with two steps: oxidation to the dioxide and carbon followed by reaction of carbon. Of the two reactants, carbon dioxide is often preferred as water is prone to leave residual carbon and produces hydrogen as a product. However, the use of carbon dioxide or water as an oxidant for irradiated carbons would result in the abatement of very large quantities of carbon dioxide or water during the abatement of carbon-14 and tritium.
The conversion of mixed uranium plutonium nitride to solid-state mixed dioxide may provide a method of removing nitrogen-15 without reducing the plutonium solubility.