MODULAR, INTEGRATED, AUTOMATED, COMPACT AND ANTI-PROLIFERATION PROCESS FOR CHEMICALLY RECYCLING A FUEL...

Abstract

A unique integrated system for recycling spent nuclear fuel (SNF) from a reactor includes a stripping tank to separate fuel pellets from nuclear fuel rods by oxidation to produce a uranium compound. A fluorination tank is connected to the stripping tank. A fluorinating agent is injected into the fluorination tank and reacts with the uranium compound to convert the uranium compound into UF6. An electrolytic extraction tank is connected to the fluorination tank to remove plutonium and actinides by electrolytic extraction. (Fig. 8)

Description

Title of the invention: Modular, integrated, automated, compact and anti-proliferation process for chemically recycling spent nuclear fuel (SNF) from nuclear reactors to recover a mixture of transuranium elements (TRU) for advanced reactor fuel, and to recycle uranium and zirconium Technical field

[0001] This patent application is a continuation in part (CIP) of U.S. Patent Application Serial No. 17 / 591,167, filed on February 2, 2022, entitled “MODULAR, INTEGRATED, AUTOMATED, COMPACT, AND ANTIPROLIFERATION METHOD FOR CHEMICALLY RECYCLING SPENT NUCLEAR FUEL (SNU) FROM NUCLEAR REACTORS TO RECOVER A TRANSURANIA (TRU) ELEMENT MIXTURE FOR ADVANCED REACTOR FUEL, AND TO RECYCLE URANIUM AND ZIRCONIUM” in the names of the same inventors. This patent application claims the benefit under 35 USC §119(e).

[0002] The present invention relates generally to nuclear reactors and, more specifically, to a method for recycling uranium, plutonium and possibly minor actinides from spent nuclear fuel (SNF) in order to remove only the actual residual by-products of the nuclear fission process, namely the fission products. CONTEXT OF THE INVENTION

[0003] Almost all commercial thermal neutron reactors are light water reactors (LWRs). LWRs are a type of thermal neutron reactor that uses deionized water as both a coolant and a neutron moderator, as opposed to heavy water reactors, which are also a type of thermal neutron reactor but use a form of water containing deuterium (2H). In an LWR, fast neutrons from fission collide with hydrogen atoms in the water and slow down. This increases the likelihood of the neutron interacting with the fuel atoms. LWRs use ceramic pellets composed of uranium oxide enriched to 3-5% as fuel. These pellets are loaded onto thin-walled cylinders clad in zircaloy to prevent radioactive fission products (FPs) from being released into the cooling water.Zircaloy is very corrosion resistant and has . a low neutron interaction cross section. The cylindrical tubes are sealed with an inert atmosphere of helium gas to prevent chemical reactions and increase heat transfer. These "fuel rods" can be arranged in assemblies, typically 17x17 for a pressurized water reactor (PWR) and 10x10 for a boiling water reactor (BWR). The assemblies themselves can be made of stainless steel and provide structural support using end plates, axially distributed spacer grids, expansion springs, and guide tubes for the control rods and instrumentation.

[0004] As fuel atoms split and are consumed, the fuel pellets' ability to sustain a chain reaction decreases. Over time, FPs accumulate and increase stray neutron losses. In addition, physical damage to the fuel from heat and radiation also limits its lifetime. Every 18 to 24 months, one-third of the fuel assemblies can be removed from the reactor core and replaced with new ones. The remaining fuel assemblies can be reworked to ensure consistent fuel burnup. Thus, the average residence time of a fuel assembly in the core is between 4.5 and 6 years. When removed from the core, spent nuclear fuel (SNF) produces intense heat and radiation from the short-lived FPs, making cooling and radiation shielding necessary.Therefore, spent fuel assemblies can be stored on-site in demineralized water basins. After spending about 10 years in this wet storage, the spent fuel assemblies are packaged in dry steel and reinforced concrete drums and stored on concrete slabs at the power plant sites. These drums can then be moved to a geological repository when one becomes available. This spent fuel assemblies are commonly referred to as "nuclear waste."

[0005] CNU is not completely defueled. The majority of CNU (>95%) is composed of U-238, with about 0.9% unfissioned U-235. While in the neutron field of the reactor core, some U-238 atoms capture neutrons and transform into isotopes heavier than uranium called "transuranic" or TRU. Therefore, about 1.0% of CNU may be composed of TRU, of which 0.9% may be unfissioned plutonium formed by neutron captures in U-238, and about 0.1% may be minor actinides formed by successive neutron captures, such as neptunium, americium, curium, etc. The remainder is a mixture of FPs, much of which has already become radioactively stable.

[0006] The radioactive composition of CNU may be as follows: about 1% short-lived FP (half-life of 1 to < 50 years), about 1% medium-lived TRU (1000 to < 100,000 years), and 95.5% very long-lived uranium (> 100 million years). Fission products may represent the greatest environmental risk in the short term and generally require isolation for a few hundred years. Transuranic elements have half-lives of several thousand years and generally require geological isolation for up to 250,000 years. They are a major concern when analyzing means of CNU disposal. However, they could potentially be used as fuel if recovered. The uranium contained in CNU is harmless from a radioactivity point of view since its isotopes have half-lives ranging from 2 million to 4 billion years.Each category taken in isolation is quite manageable, but as they are mixed in the CNU, they pose a radiological health risk and require expensive monitoring and politically difficult disposal.

[0007] Reprocessing generally means effectively “shutting down” the fuel cycle by recycling uranium, plutonium, and possibly minor actinides and removing only the actual residual by-products of the nuclear fission process, namely the fission products. Current commercial reprocessing methodologies are based on the plutonium and uranium recovery extraction (PUREX) process or its derivatives (https: / / link.springer.com / chapter / 10.1007%2Fl-4020-3598-5_24). PUREX is an aqueous process, meaning it requires dissolving CNU pellets in solutions of various acids of varying strengths. Water, which is an excellent neutron moderator, limits the size of the equipment due to nuclear criticality issues, making larger facilities necessary for sufficient efficiency.The use of such acids and reagents also leads to the production of large quantities of slightly radioactive chemical waste due to radiolysis. In addition, PUREX results in the production of a purified plutonium stream creating proliferation concerns and, as such, PUREX derivatives aim to minimize proliferation risks either by co-extraction of minor actinides or by denaturation of the product. An alternative approach called pyroprocessing has also been developed, but not deployed industrially (please refer to https: / / www.sciencedirect.com / science / article / abs / pii / 0149197096000078 and https: / / www.sciencedirect.com / science / article / pii / B97 80128197257002208). Pyroprocessing is generally a batch process in which solids are transferred from one unit operation to another.

[0008] Therefore, it would be desirable to provide a system and method that overcome the aforementioned limitations. The system and method will utilize a improved process that integrates the various unit operations into a series of connected vessels to accomplish the task of recycling nuclear waste. The system and process will utilize chemistry derived from pyroprocessing, fluorination, and electrowinning, with the process flow providing significant improvements over existing CNU recycling processes. SUMMARY

[0009] According to one embodiment, a unique integrated system for recycling spent nuclear fuel (SNF) from a reactor is disclosed. The system includes a stripping vessel for separating fuel pellets from nuclear fuel rods by oxidation to produce a uranium compound. A fluorination vessel is connected to the stripping vessel, wherein a fluorinating agent injected into the fluorination vessel reacts with the uranium compound to convert the uranium compound to UF6. An electrowinning vessel is connected to the fluorination vessel for removing plutonium and actinides by electrowinning.

[0010] According to one embodiment, a method for recycling spent nuclear fuel (SNF) from a reactor is disclosed. The method for recycling spent nuclear fuel (SNF) from a reactor comprises: separating fuel pellets from nuclear fuel rods by oxidation to produce a uranium compound; converting the uranium compound to UF6 by fluorination; and removing plutonium and actinides by electrowinning. Brief description of the drawings

[0011] The present application is described in more detail with reference to the following drawings. These figures are not intended to limit the scope of this application, but rather illustrate certain attributes thereof. The same reference numbers will be used throughout the drawings to designate the same or similar parts.

[0012] [Fig. 1] is a process diagram of an exemplary embodiment of a method for recycling spent nuclear fuel from light water reactors according to an embodiment of the present invention;

[0013] Figures 2A and 2B are exemplary embodiments of a stripping tank used in the method shown in [Fig.l] according to one embodiment of the present invention;

[0014] [Fig. 3] is an exemplary diagram showing a chemical composition of spent nuclear fuel according to one embodiment of the present invention;

[0015] [Fig.4] is an exemplary table showing the generalized behavior of fission products in fluoride salts according to one embodiment of the present invention;

[0016] [Fig.5A] is an exemplary embodiment of a fluorination tank used in the process shown in [Fig.1] according to one embodiment of the present invention;

[0017] [Fig.5B] is an exemplary embodiment of a fluorination tank used in the process shown in [Fig.1] according to one embodiment of the present invention;

[0018] [Fig.6] is an exemplary diagram showing the relative "stability" of different salt components that can be used for recycling nuclear fuel according to one embodiment of the present invention;

[0019] Figures 7A and 7B are exemplary embodiments of an electrowinning tank used in the method shown in [Fig.l] according to one embodiment of the present invention; and

[0020] [Fig.8] is an exemplary embodiment of a single integrated system for recycling spent nuclear fuel from light water reactors according to an embodiment of the present invention. DESCRIPTION OF THE REQUEST

[0021] The description presented below in connection with the accompanying drawings is intended to be a description of presently preferred embodiments of the disclosure and is not intended to represent the only forms in which the present disclosure may be constructed and / or used. The description sets forth the functions and sequence of steps for constructing and operating the disclosure in connection with the illustrated embodiments. However, it is to be understood that the same or equivalent functions and sequences may be accomplished by different embodiments which are also intended to be encompassed within the spirit and scope of the present disclosure.

[0022] Pyroprocessing can be defined as a process in which materials are subjected to high temperatures (typically above 800 °C) in order to cause a chemical or physical change. Nuclear pyroprocessing can be defined as a high-temperature electrochemical fuel processing technology to recycle spent fuel into metallic fuel for a fast nuclear reactor. Nuclear pyroprocessing uses electrorefining to separate uranium, plutonium + TRU, and FPs using a suitable electrolytic solution (e.g., molten eutectic LiCl-KCl). Corrosive salts and high temperatures cause high stresses on the containment materials. The high radiation fields encountered due to the shorter cooling time before reprocessing make it more difficult to use sensors and probes for continuous monitoring.The process is more suitable for metallic fuels, hence its use in the IFR project. but it has been tested with combustible oxides (https: / / www.sciencedirect.com / scien ce / article / pii / S1002007115001197) by reducing the fuel to metallic form before processing. Although originally studied as a batch process, recent studies have experimented with continuous operation by removing PFs from the electrolytic solution by passing it through zeolite columns (https: / / www.sciencedirect.com / science / article / pii / B9781845695019500104).

[0023] Exemplary embodiments of the system and method disclose an improved process that integrates the various unit operations in a series of connected vessels to accomplish the task of recycling nuclear waste. The chemistry is derived from pyroprocessing, fluorination, and electrowinning, and the process flow provides significant improvements over existing CNU recycling processes.

[0024] Exemplary embodiments of the system and method are intended to optimize the recycling of CNU from LWRs, although the applicability may be extended to metal salts, molten salts, and other fuel forms. The unit operations are adaptations / modifications of chemical processes currently under development. The chemistry consists primarily of redox reactions that are propagated by the use of reagents or by the application of an appropriate electrical potential. The main component of CNU—uranium—can be removed in a single step by conversion to a gas, thereby improving decontamination factors. After uranium removal, the “ash” containing the TRU fluorides as well as the short-lived fission products that form stable fluorides are converted to a molten salt form that is distinct from PUREX where the fission product stream is separated in the early stages.An electrowinning process uses an electric potential to cause the deposition of actinide fluorides (only TRUs at this stage) present in the molten salt phase in metallic form onto a suitable cathode. This is followed by an increase in the electric potential to deposit lanthanide species onto a separate cathode. The reactive fission product species continue to remain in the molten salt and can be separated by a combination of ion exchange, fractional crystallization, and distillation.

[0025] With reference to the figures, a system 1 and method for optimizing the recycling of CNU from LWRs may be disclosed. The system 1 may integrate various sub-operations into a continuous process that allows for the transfer of nuclear material from one operation to another with minimal operator intervention. As can be seen, the first step in this process is to efficiently separate the fuel pellets from the zircaloy cladding. The fuel oxide is not the most efficient heat conductor. When operating at full power, the The temperature at the center of the pellet is above 1200 °C, while the outer surface of the pellet is close to 400 °C, while the temperature at the cladding surface is around 300 °C. In addition, gaseous PFs may accumulate during operation, leading to swelling and cracking of the cladding. In the case of severe swelling, the pellet surface presses against the inner walls of the cladding and may physically bond to it or sometimes even crack the cladding. These phenomena are collectively known as pellet-cladding interaction. Therefore, "stripping" may be a necessary starting point step. Stripping can take different forms.

[0026] According to one embodiment, the decladding process oxidizes UO2 under a dry oxygen atmosphere at approximately atmospheric pressure and at a temperature in the range of about 400 to 500°C to U3O8, as shown in the equation below. Upon reaction with oxygen, the uranium oxide is pulverized into a tri-uranium octoxide powder, the volume of the pellets increases, which allows for efficient separation from the cladding. The general process is as follows:

[0027] UO2 + O2^U3O8

[0028] According to another embodiment, a mixture of oxygen and ozone (O3) can be used, which reduces the temperature of the oxidation reaction to < 400 °C and gives a more pulverized product in the form of UO3.

[0029] In recycling metallic fuel, a starting point process separates the cladding, if present, from the metal, including any reactive binding material (e.g., sodium), and then subjects the metallic fuel to the above oxidation reaction to produce pulverized U3O8. The remaining operations do not change in the case of metallic fuel.

[0030] It has been observed that pulverized fuel has a minimum particle size and a maximum particle size such that contamination by fine powdery residues is minimal. At these high temperatures, gaseous fission products and other volatile fission products, such as krypton, xenon, iodine, bromine, tritium, and radon decay product, which have accumulated at the grain boundaries can be released. Studies show that drilling holes in the cladding prior to oxidation can increase the reaction rate. With hole drilling and shearing, volatile FPs in the fuel rod expansion chamber (i.e., the space in the clad rods for gas accumulation) can be released. Therefore, it is contemplated that these two unit operations can be performed in a single space, regardless of the number of units. parallel subsequent operations, which would minimize the volume of gas requiring purification.

[0031] In the method used herein, after some physical unbundling, i.e., removal of structural components and drilling of holes to allow for efficient reaction, the CNU assemblies may be loaded into a removable basket mechanism 10 (hereinafter mechanism 10), as seen in Figures 2A-2B. The removable basket mechanism 10 may have a plurality of slots 12 formed within the mechanism 10. Each slot 12 may be of a suitable size to secure the fuel rods 14 vertically. The vessel 16 may be provided with a removable enclosure 23 secured to an upper region of the vessel 16. The removable enclosure 23 may be used to provide isolation from the internal atmosphere and may include an outlet 20 for circulating the atmosphere and collecting any off-gases from the process.

[0032] The mechanism 10, loaded with the spent fuel rods 14, may then be positioned within a stripping vessel 16 (hereinafter vessel 16). The vessel 16 may be composed of metal alloys capable of withstanding high temperatures without corrosion, such as various stainless steels, Inconels, and the like. The vessel 16 may have sufficient thickness to withstand the desired temperatures / pressures used during the stripping process.

[0033] The interior of the bowl 16 may be formed of a heat-resistant ceramic material 16A. According to one embodiment, the heat-resistant ceramic material 16A may be capable of resisting oxidation at temperatures up to about 1500°C. According to one embodiment, the same ceramic material may be used to coat the mechanism 10 and the surfaces of the slots 12 formed in the mechanism 10.

[0034] Inlets 18 and outlets 20 may be formed in the vessel 16 and in communication with the interior of the vessel 16. The inlets 18 and outlets 20 may be formed to allow oxygen to enter the vessel 16 and various gaseous fission products to exit the vessel 16. Heaters 22 may be coupled to the vessel 16. The heaters 22 may be used to maintain the temperature of the vessel 16.

[0035] With the fuel rods 14 placed inside the vessel 16, the vessel can be heated to between 400 and more than 500 °C in a dry atmosphere containing excess oxygen to convert the uranium oxide to U3O8. The reaction with the oxygen pulverizes the pellets into U3O8 powder and the vessel 16 can collect this powder at the bottom of the vessel 16. Similarly, a reaction with an O2 / O3 mixture at < 400 °C converts UO2 to U03. The reaction can also simultaneously rupture the cladding of the fuel rods 14

[0036] The pulverized U3O8 / UO3 powder formed during the decladding process can be collected at the bottom of the vessel 16 for further processing, while the hollow zircaloy cladding of the fuel rods 14 can be recovered and refined by electrotransport, whereby high-purity hafnium-free zirconium can be recovered.

[0037] According to one embodiment, the vessel 16 may have a dual purpose, wherein the vessel 16 may be used for the recovery of zirconium after oxidation. After the decladding process, the zircaloy cladding shells of the fuel rods 14 may be left behind. A molten salt having poor solubility for uranium oxides may be introduced into the vessel 16 with the mechanism 10 containing the fuel rods 14 and positioned between cathode - and anode + plates. A small amount of zirconium halide may be added to the salt to initiate electrotransport.Upon application of a suitable electrical potential to the cathode- and anode- plates, zirconium from the fuel rods 14 is electrotransported / electrorefined by dissolution in the molten salt while the zirconium halide is reduced to metal on a suitable cathode, such as a liquid cadmium cathode or a solid stainless steel cathode. The process releases pieces of U3O8 / UO3 which can be collected in the vessel 16 and can be transferred to the downstream process. It should be noted that the above steps can be modified when the vessel 16 is first filled with the molten salt, followed by the stripping process.

[0038] The following reactions may occur:

[0039] Anode: Zr Zr"+ + ne

[0040] Cathode: Zrn+ + ne = Zr

[0041] The cadmium cathode or other suitable cathode carrying its zirconium load may be discharged to the bottom of the tank 16 via an outlet 25. Any salt finding its way to the outlet can easily be separated. Furthermore, upon cooling, the zirconium may be separated from the cadmium by phase separation by fractional crystallization. Similarly, in the case of a solid cathode, the deposit may be recovered using mechanical means, such as scraping, or by applying a thermal shock which utilizes the differences between the coefficients of thermal expansion of the cathode material and the deposited metal.

[0042] According to another embodiment, the stripping can be carried out by electrotransport of the zirconium, after which the uranium oxide pellets can be collected at the bottom of the tank 16 and then, after evacuation of the molten salt, the UCL can be roasted. under a dry oxygen atmosphere to form U3O8 powder, or under an O2 / O3 atmosphere to form UO3.

[0043] CNU has a nominal composition as seen in [Fig.3]. In conventional pyroprocessing, uranium oxide can be reduced to metal prior to an electrorefining process to recover and separate uranium and plutonium. However, its uranium separation factors are poor and it also requires material transfer between vessels.

[0044] For process efficiency, uranium must be recovered with minimal contamination. Several systems propose to use the valence chemistry of uranium to collect it as volatile UF6. This can be achieved by contact with a gaseous or liquid fluorinating agent. Many important fission products form fluorides (as shown in [Fig. 4]) and are dissolved in the molten salt during the process.

[0045] To separate nuclear materials from CNU, volatile hexafluorides of U, Pu, and Np can be removed from a complex matrix, such as the pulverized U3O8 / UO3 / UO 2 matrix containing PFs and TRUs as present in this proposed process stream. Efficient separation of the obtained volatile fluorides from each other can be achieved due to their volatility by changing the conditions regarding their physical properties, such as boiling and sublimation points. Several strong fluorinating agents have been investigated for this purpose - HF, F2, C1F3, BrF5, etc. Since large-scale fluorination in such a facility could depend on the storage, handling, and transportation of these fluorinating agents, there is a significant risk to human health and the environment.Since these agents are corrosive, their accidental release at a CNU recycling facility will further complicate theoretical technical incident prevention requirements.

[0046] Alternatively, nitrogen trifluoride (NF3) can serve as an effective fluorinating agent. NF3 has been used in the electronics industry to remove silicon and silicon compounds during the fabrication of semiconductor devices and during the cleaning of microelectronic devices. Unlike the fluorinating agents described above, NF3 is not corrosive and does not react significantly with moisture or common acids and bases at room temperature. NF3 is also thermally stable at relatively high temperatures and insensitive to shock, so that it can be transported, stored, and handled safely and economically for a large-scale process. Most importantly, NF3 has a unique temperature-based selective reactivity with the various components of CNU. This characteristic can be exploited to enable the separation of fission products that form volatile fluorides from uranium, as well as for the efficient separation of uranium from other actinides with minimal cross-contamination.

[0047] The Gibbs free energy of the principal reactions of metallic uranium and various uranium and plutonium compounds with NF3 are presented below. The reactions take place at 300 °C, however there should be little variation at the temperatures envisaged for the process.

[0048] U + ZNF^g) ^UF^gJ+N^g))=-1884.8

[0049] UO2 + 2NF3(g) —UF„(g) +N2(g) + O2(g) AGm = -901.1

[0050] U3O8 + 6NF3(g) -^UF6(g)+3N2(g)+4O2(g) AGrxn = -823.2

[0051] UF4 + NF3(g) ^UF6(g)+N2(g) AG^ -204.9

[0052] 3 PuO2 + 4 NF3(g) PuF4 + 2 N2(g)+3 O2(g) AGrxn = -880

[0053] 3 PuF4 + 2 NF3(g) PuF6(g)+N2(g) AGrxn = +54.1

[0054] The free energies clearly show that the reaction with the metal is the most favorable, however such a reaction can sometimes proceed with the formation of a UF4 intermediate. The reaction with tri-uranium octoxide will also proceed spontaneously, but may lead to the formation of intermediate oxyfluorides. The overall reaction will be sensitive to temperature and other factors, such as the quality of the mixing of the reactants, which will require optimization. Similarly, it should be noted that although NF3 converts PuO2 to PuF4, the formation of a hexafluoride is not favorable.

[0055] When gaseous NF3 mixed with an inert medium, such as helium or argon, permeates through the pulverized CNU mixture, triuranium octoxide readily forms UF6 which volatilizes at process temperatures and bubbles out of solution from the bulk material. At lower temperatures, fission products forming volatile fluorides react with NF3 and are removed from the bulk material. Volatilization of uranium occurs at a higher temperature, whereas, as mentioned, unlike gaseous F2, NF3 is not an effective agent for producing PuF6 and retaining volatility. The reaction is unfavorable and the formed PuF6 is readily reduced to PuF4, with the UF4 in the salt-like medium acting as a reducing agent.Neptunium, which exists in the CNU primarily as Np-237, can partially volatilize during this process, but at a separate and higher temperature than uranium. Np-237 can be collected, which is a raw material for producing the valuable Pu-238 used to power deep-space satellites and rovers.

[0056] Starting with a UO3 reactant, a reducing agent such as hydrogen gas / ammonia can be used to reduce the UO3 to UO2.

[0057] If the UO3 is reduced to UO2, then this product can be contacted with a fluorinating agent, such as HF, to convert the UO2 to UF4, before being contacted with a fluorinating agent strong enough to produce volatile fluorides.

[0058] Regardless of the starting uranium compound in this reaction, i.e., U3O8, UO3, UO2, UF4, UO2F2, etc., the reaction of the uranium compound with NF3 to form volatile UF6 occurs at a distinct temperature that allows the separation of uranium from the fission products and other transuranics.

[0059] The emanating gases will require treatment to separate the uranium from any remaining volatilized FP fluorides. This separation can be achieved by selective adsorption and desorption of the minor gaseous components by passing them over beds of NaF, MgF2, or alumina. The NpF6 can be absorbed by MgO beds. A process developed in the past used a bed of NaF pellets at 400 °C that irreversibly captured the Nb, Ru, and Zr fluorides from the gas stream. The second sorption bed is NaF at 100 °C, which captures the U and Mo fluorides, albeit reversibly. The remaining volatile FPs continue to pass through and are captured as waste. When the second fluoride bed is heated to > 150 °C, the U and Mo fluorides desorb and can pass onto a third MgF2 bed at 150 °C which captures the MoF6 and thus a purified UF6 stream can be obtained.Similarly, a rectification and sublimation process can be applied, whereby the gas stream passes through columns maintained at different temperatures where individual or group fluoride species condense while the other components remain volatile and pass through.

[0060] If a fuel in the form of molten fluoride salt is used as the feed to the process, it is introduced into the process at this stage and subjected to the fluorination operation. The free energy of formation of UF6 from UF4 means that the process is still sufficiently favorable. The remaining operations are the same as for other types of fuel.

[0061] By first removing the uranium, the concentrated fission products and minor actinides block the proliferation of plutonium. The recovered uranium has a very high decontamination factor and could be used directly as an enrichment feedstock, converted into breeder material for fast-spectrum reactors, reconstituted with the recovered TRU mixture to fuel advanced reactors, or simply returned to Earth in oxide form.

[0062] Volatilization of uranium allows very high decontamination factors (> 105) to be obtained between uranium and other actinides and fission products. contaminants. It allows the recovery of neptunium in the CNU (which is composed almost entirely of Np-237) which can be cryo-distilled from the flux. Np-237 is a raw material in high-flux reactors to produce Pu-238 which is used in radioisotope thermoelectric generators (RTGs) for deep space probes.

[0063] Referring to [Fig. 5A], a vessel 30 for the fluorination of U3C)8 may be shown. The vessel 30 may be formed from a first reservoir 32. The first reservoir may be formed from a suitable alloy that is compatible with the high temperature corrosive salt environment and whose chemical components do not participate appreciably in the redox chemistry under the operating conditions of the proposed process. Various grades of high nickel alloys, such as Inconel, and alloys of nickel and chromium with refractory metals, such as Monel, are suitable candidates, as are lanthanide boride carbon composites.

[0064] The first tank 32 may have double walls (i.e., a tank within a tank structure). An outer wall 34 of the first tank may serve as a catch basin in the event of a leak. The space 37 between the outer wall 34 and an inner wall 36 may be used to cool the first tank 32. In one embodiment, the space between the outer wall 34 and the inner wall 36 may be cooled by circulating a coolant, or the outer face of the inner wall 36 may be constructed with a network of heat pipes that provide passive cooling. In one embodiment, a coolant, such as a NaK alloy, may be used, which could also be the coolant within the heat pipes if this design is used. This serves two purposes:

[0065] 1) the inner face of the inner wall 36 can always maintain a layer of salt " frozen” which constitutes a barrier against chemical corrosion when the fluorination reactions are carried out, and

[0066] 2) passive cooling is provided for highly exothermic reactions which form volatile fluorides.

[0067] The first reservoir 32 may have a plurality of inlets 38. The inlets 38 may be in communication with the interior of the first reservoir 32. The inlets 38 may be used to insert different agents into the interior of the first reservoir 32 during the fluorination process. In the present illustrated embodiment, the first reservoir 32 may have a first inlet 38A, a second inlet 38B, a third inlet 38C, and a fourth inlet 38D. In the present embodiment, the first inlet 38A may be formed in a lower area of ​​the first reservoir 32. The first inlet 38A may be used to insert a fluorinating agent into the first reservoir 32. The second inlet 38B may be formed in an upper region of the first reservoir 32. The second inlet 38B may be used to insert the U3O8 suspension into the first reservoir 32. The third inlet 38C may be formed in a lower side region of the first reservoir 32. The third inlet 38C may be used to circulate a coolant into the space 37 formed in the first reservoir 32. The fourth inlet 38D may be formed in a side region of the first reservoir 32. The fourth inlet 38D may be used to circulate a molten salt inside the first reservoir 32.

[0068] The first reservoir 32 may have a plurality of outlets 40. The outlets 40 may be in communication with the interior of the first reservoir 32. The outlets 40 may be used to remove different materials from the interior of the first reservoir 32 during the fluorination process. In the present illustrated embodiment, the first reservoir 32 may have a first outlet 40A, a second outlet 40B, a third outlet 40C, and a fourth outlet 40D. In the present embodiment, the first outlet 40A may be formed in a lower region of the first reservoir 32. The first outlet 40A may be used to remove molten salt inserted into the first reservoir 32. The second outlet 40B may be formed in an upper region of the first reservoir 32. The second outlet 40B may be used to remove UF6 from the first reservoir 32.The third outlet 40C may be formed in an upper side area of ​​the first tank 32. The third outlet 40C may be used to remove coolant flowing through the space 37 formed by the inner wall 36 and the outer wall 34 in the first tank 32. The fourth outlet 40D may be formed in a side area of ​​the first tank 32. The fourth outlet 40D may be used to remove molten salt inside the first tank 32.

[0069] The vessel 30 may have a second reservoir 39 connected to the first reservoir 32. The second reservoir 39 may be used to circulate molten salt through the first reservoir 32. The second reservoir 39 may have an inlet 42A formed on a lower region of the second reservoir 39. The inlet 42A may be connected to the fourth outlet 40D of the first reservoir 32. The second reservoir 39 may have an outlet 44 formed on an upper region of the second reservoir 39. The outlet 44 may be connected to the fourth inlet 38D of the first reservoir 32. The molten salt in the second reservoir 36 may exit through the outlet 44 and flow into the fourth inlet 38D of the first reservoir 32. The molten salt may exit the first reservoir 32 via the fourth outlet 40D and return to the second reservoir 39 via the inlet 42A. The second reservoir 39 may have another inlet 42B formed on a side area thereof.Inlet 42B may be used to inject a replacement salt into the second reservoir 39.

[0070] Heaters 46 may be coupled to the second reservoir 39. The heaters 46 may be used to transfer thermal energy to the second reservoir 39 to regulate and maintain the liquidus of the molten salt in the second reservoir 39 and thus the conditions in the first reservoir 32.

[0071] The tank 30 may have a pump 46. The pump 46 may be connected between the first tank 32 and the second tank 39. The pump 46 may be used to circulate the molten salt between the first tank 32 and the second tank 39.

[0072] Referring to [Fig.5B], another embodiment of fluorination may be disclosed. In this embodiment, the fluorination is carried out without using a molten salt medium. When carried out without using a molten salt medium, the uranium compound (U3O8 or UO3 or UO2 or UF4 or UO2F2 hereinafter (uranium compound)) from the decladding step may be introduced into a vessel 30'. The vessel 30' may be a fluidized bed reactor 70 (hereinafter bed reactor 70). The bed reactor 70 may be made of the same material as that of the other fluorination vessel 30 described above.

[0073] The bed reactor 70 may have a structure similar to that of the first tank 32. The bed reactor 70 may have double walls (i.e., a vessel within a vessel structure). An outer wall 72 may serve as a catch pan in the event of a leak. The space 74 between the outer wall 72 and an inner wall 76 may be used to cool the bed reactor 70. In one embodiment, the space between the outer wall 72 and the inner wall 76 may be cooled by circulating a coolant or the outer face of the inner wall 76 may be constructed with a network of heat pipes that provide passive cooling. In one embodiment, a coolant, such as a NaK alloy, may be used, which could also be the coolant within the heat pipes if this design is used.

[0074] The bed reactor 70 may have a distribution plate 78. The distribution plate 78 may be positioned proximate a lower area within the bed reactor 70. The distribution plate 78 may extend through the interior of the bed reactor 70, to form a bed within the bed reactor 70. According to one embodiment, the distribution plate 78 may be formed of nickel.

[0075] The bed reactor 70 may have a plurality of inlets 80. A first inlet 80A may be formed near the top of the reactor vessel 70. The first inlet 80A may be where the pulverized uranium compound may be introduced into the bed reactor 70. A second inlet 80B may be formed in a bottom of the bed reactor 70. The second inlet 80B may be where the gaseous fluorinating agent may be introduced into the bed reactor 70.

[0076] The uranium compound introduced through the first inlet 80A may be positioned on the distribution plate 78. When the gaseous fluorinating agent enters the bed reactor 70 through the second inlet 80B, the distribution plate 78 may bubble the gaseous fluorinating agent through the formed bed of the uranium compound and may cause the bed of the uranium compound to expand due to fluidization. The fluorinating agent may react with the uranium compound and convert the uranium compound to UF6. As before, different reactions occur at different temperatures, resulting in the separation of the uranium as UF6, while the transuranics and non-volatile fission products remain behind.

[0077] The exothermic reaction may be cooled by a suitable coolant circulating in the space 74 between the outer wall 72 and an inner wall 76. In the present embodiment, the bed reactor 70 may have a third inlet 80C formed in an upper side region. The third inlet 80C may be used to insert the coolant into the space 74 between the outer wall 72 and an inner wall 76. A first outlet 82A may be formed in a lower side region of the bed reactor 70 to remove the coolant circulating in the space 74.

[0078] A second outlet 82B may be formed in an upper region of the bed reactor 70 to allow the gaseous UF6 to exit from the top of the bed reactor 70. The gaseous UF6 may exit through a bayonet filter 84 connected to the second outlet 82B. The bayonet filter 84 may be made of a suitable material, such as sintered nickel. The bayonet filter 84 may prevent any solid particles larger than a certain size entrained in the gas mixture from exiting the bed reactor 70. A third outlet 82C is provided near the bottom of the bed reactor 70. The third outlet 82B may be used for products remaining after fluorination and volatilization of the UF6 to be removed from the bed reactor 70.

[0079] [Fig.6] may represent a diagram of various fluorides of interest for CNU recycling. The eutectic mixture LiF-CaF2 may serve as a very stable mixture having an acceptable eutectic melting point of 767 °C. Dissolving uranium and other actinide fluorides may serve to further reduce the eutectic point. Alternatively, LiF-NaF or LiF-KF which have lower eutectic points could also be used depending on technical needs.

[0080] Electrowinning is one of the oldest known electrolytic processes and is widely used in modern metal recovery, mining, refining and even wastewater treatment. It has long been proposed as a means of extracting actinides. In the process As presented, the plutonium and minor actinides remaining in the saline mixture after uranium volatilization are well suited for coextraction on a solid or liquid electrode via electrowinning. The extraction can be controlled by regulating the applied potential and the process can be scaled up while ensuring that the recovered plutonium is never pure.

[0081] Since there is a large reduction potential gap between the actinides and lanthanides in a molten salt solution such as that present herein, the actinides present in the salt can be coextracted. As the electrical potential applied between the electrodes increases, lanthanide PF species with more negative reduction potentials can be extracted for disposal. In the proposed method, the post-fluorination molten salt mixture flows into an electrowinning vessel 50, as seen in FIGS. 7A and 7B. The electrowinning vessel 50 may have an inlet 66 formed therein. The inlet 66 may be used to allow the molten salt mixture to flow into the electrowinning vessel 50. Heaters 68 may be attached around the perimeter of the electrowinning vessel 50.The heaters 68 may be used to maintain the liquidus conditions of the molten salt mixture. In the case of a combination of high eutectic point salts, the cathode may have an electrically connected pan underneath to collect the deposited species in liquid form. The continuity of the cathode surface with the same reduction potential ensures prevention of redissolution of the reduced metals in the fluoride medium.

[0082] The electrowinning vessel 50 may consist of a central anode 52 and a plurality of cathodes 54 disposed circumferentially around the electrowinning vessel 50. The vessel 50 may consist of circular or semi-circular support rails 51 made of a conductive material, such as graphite, to accommodate the plurality of cathodes 54 and provide electrical conduction. The support rails 51 may be placed on top of the vessel 50 and electrically insulated from the body of the vessel 50. In one embodiment, the anode 52 may be a porous graphite anode. The cylindrical cathodes 54 may be mounted on motors 56 that rotate the cathodes 54 to provide continuous relative motion between the molten salt and the cathode surface and ensure that there is uniform deposition of the electrolytically produced actinides. Examples of reactions may include:

[0083] D. electrolysis PuF4 ---^“Pu4+ + 4F

[0084] Anode: 4F 2F2 + 4e

[0085] Cathode: Pu4*+4e" = Pu

[0086] A separation jacket 58 may be positioned around the anode 52 to allow capture of the reactant fluorine gas formed on the anode 52. The separation jacket 58 creates an isolated gas space for collection of the fluorine gas formed on the porous anode 52. A noble gas atmosphere must be maintained within the electrowinning cell 50. The electrowinning cell 50 may have a noble gas inlet 60 formed therein. In the present embodiment, the noble gas inlet 60 may be formed in a lower region of the electrowinning cell 50. The noble gas inlet 60 may allow a noble gas to bubble and continuously circulate through the mixture to provide a carrier gas that dilutes the formed fluorine gas. The lower bottom 62 of the electrowinning cell 50 may be sloped. The inclined lower bottom 62 may allow the removal of insoluble PFs which may accumulate after treatment.A slurry outlet 64 may be formed on a lower side region of the electrowinning tank 50, adjacent a lowest region of the inclined lower bottom 62. The slurry outlet 64 may be used to allow slurry to be discharged and removed from the electrowinning tank 50.

[0087] The plutonium thus extracted can be considered "anti-proliferation" given the presence of other actinides (i.e. neptunium, americium and curium) and contamination by fission products (quantities > 500 ppm of rare earths). This flow makes it impossible to use the plutonium for clandestine purposes without extensive processing.

[0088] The molten salt bath can then be passed through zeolite columns to adsorb minor fission product impurities present to sufficiently purify the salt for reuse in the process. When sufficient amounts of reactive PFs (primarily cesium and strontium) accumulate in the salt, it can be subjected to fractional crystallization which creates a phase in which most of the reactive PF fluorides are dispersed. This salt phase can then be extracted for disposal. In addition, distillation could be used to purify the salt mixture leaving the reactive PF fluoride species.

[0089] System 1 can integrate the various sub-operations described into a continuous process that allows for the transfer of nuclear material from one operation to another with minimal operator intervention. This allows for significantly improved material control and accounting, better instrument placement, improved controls, while reducing the amount of unaccounted materials. The molten salt used in System 1 is immune to radiation damage and does not lose its chemical properties. Therefore, the molten salt can be continuously reused in the process.

[0090] As can be seen in one embodiment shown in [Fig. 8], the system 1 combines, sizes and optimizes the stripping vessel 16, the vessel 30 for fluorination of U3O8 and the electrowinning vessel 50 for throughput after taking into account nuclear criticality safety. Embodiments of the system 1 may also use more than one unit of a sub-operation in parallel, either in combination or independently connected in series to the successive sub-operation. Several of these integrated systems 1 modules can then be built within a larger processing facility. As processing requirements increase, more systems 1 modules can be added to the facility to improve cost-effectiveness.

[0091] The foregoing description illustrates particular embodiments of the application, but is not intended to be a limitation on its practice. The following claims, including all equivalents thereof, are intended to define the scope of the application.

Claims

Claims

1. A unique integrated system for recycling spent nuclear fuel (SNF) from a reactor comprising: a stripping vessel for separating fuel pellets from nuclear fuel rods by oxidation to produce a uranium compound; a fluorination vessel connected to the stripping vessel, wherein a fluorination agent injected into the fluorination vessel reacts with the uranium compound to convert the uranium compound to UF6; and an electrowinning vessel connected to the fluorination vessel for removing plutonium and actinides by electrowinning.

2. The system of claim 1, wherein the fluorination vessel comprises: a vessel having a double wall, wherein a space formed between the double wall is used to cool the vessel; a distribution plate formed inside the vessel containing the uranium compound injected into the fluorination vessel; and a fluorination agent inlet formed in the vessel injecting the fluorination agent into the vessel.

3. The system of claim 2, wherein the distribution plate bubbles the fluorinating agent through the uranium compound on the distribution plate to cause the uranium compound to expand, the fluorinating agent reacting with the uranium compound to convert the uranium compound to UF6.

4. The system of claim 2, wherein a coolant circulates through the space formed between the double wall to cool the tank.

5. The system of claim 2, comprising a gas outlet formed in an upper region of the vessel allowing gaseous UF6 to exit through the top of the vessel.

6. A system according to claim 5, comprising a bayonet filter connected to the gas outlet.

7. A system according to claim 5, comprising a bayonet filter connected to the gas outlet, wherein the bayonet filter is made of sintered nickel.

8. A system according to claim 2, comprising a product outlet formed in a lower area of ​​the tank for removing products remaining after fluorination and volatilization of the uranium compound from the vessel.

9. The system of claim 1, wherein the uranium compound is one of U3O8> UO3, UO2, UF4 or UO2F2 or combinations thereof.

10. A method of recycling spent nuclear fuel (SNF) from a reactor comprising: separating fuel pellets from nuclear fuel rods by oxidation to produce a uranium compound; converting the uranium compound to UF6 by fluorination; and removing plutonium and actinides by electrowinning.

11. The method of claim 10, wherein converting the uranium compound to UF6 by fluorination comprises passing a fluorinating agent through the uranium compound to cause the uranium compound to expand, the fluorinating agent reacting with the uranium compound to convert the uranium compound to UF6.

12. The method of claim 10, wherein converting the uranium compound to UF6 by fluorination comprises: forming a distribution plate within a fluorination vessel; placing the uranium compound injected into the fluorination vessel on the distribution plate; and injecting the fluorinating agent into the fluorination vessel.

13. The method of claim 12, comprising bubbling the fluorinating agent through the uranium compound on the distribution plate to cause the uranium compound to expand, the fluorinating agent reacting with the uranium compound to convert the uranium compound to UF6.

14. The method of claim 10, wherein the fluorinating agent is one of gaseous F2, HF, C1F3, BrF5, or NF3 mixed with an inert gas or combinations thereof.

15. The method of claim 10, wherein the uranium compound is one of U3O8> UO3, UO2, UF4 or UO2F2 or combinations thereof.

16. A method of recycling spent nuclear fuel (SNF) from a reactor comprising: separating the fuel pellets from the nuclear fuel rods to produce a uranium compound converting the uranium compound to UF6 by fluorination, wherein converting the uranium compound to UF6 by fluorination comprises: forming a distribution plate within a fluorination vessel; placing the uranium compound injected into the fluorination vessel on the distribution plate; injecting the fluorinating agent into the fluorination vessel; and removing the plutonium and actinides by electrowinning.

17. The method of claim 16, comprising bubbling the fluorinating agent through the uranium compound on the distribution plate to cause the uranium compound to expand, the fluorinating agent reacting with the uranium compound to convert the uranium compound to UF6.

18. The method of claim 16, wherein the fluorinating agent is one of gaseous F2, HF, C1F3, BrF5, or NF3 mixed with an inert gas or combinations thereof.

19. The method of claim 16, wherein the uranium compound is one of U3O8> UO3, UO2, UF4 or UO2F2 or combinations thereof.

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