Modular, integrated, and automated proliferation-resistant small method for chemically recycling reactor-derived used nuclear fuel (UNF) to recover transuranic (TRU) element mixture for advanced reactor fuel and to recycle uranium and zirconium

Abstract

To provide an improved process of integrating various different unit operations to a series of connected tanks to realize a role of recycling a nuclear waste.SOLUTION: The present invention includes a decladding tank for separating a fuel pellet from a nuclear fuel rod via oxidation and generating a uranium compound. A fluorination tank is connected to the decladding tank. A fluorinating agent is injected into the fluorination tank and reacts with the uranium compound to convert the uranium compound into UF6. An electrolyte collection tank is connected to the fluorination tank, and plutonium and actinide are taken out.SELECTED DRAWING: Figure 1

Description

[Technical field]

[0001] This patent application is a continuation-in-part (CIP) of U.S. Patent Application No. 17 / 591,167, filed February 2, 2022, in the names of the same inventors and entitled "A MODULAR, INTEGRATED, AUTOMATED, COMPACT, AND PROLIFERATION-HARDENED METHOD TO CHEMICALLY RECYCLE USED NUCLEAR FUEL (UNF) ORIGINATING FROM NUCLEAR REACTORS TO RECOVER A MIXTURE OF TRANSURANIC (TRU) ELEMENTS FOR ADVANCED REACTOR FUEL TO RECYCLE URANIUM AND ZIRCONIUM," which is incorporated herein by reference in its entirety. This patent application claims the benefit of 35 U.S.C. § 119(e).

[0002] Technical Field This application relates generally to nuclear reactors, and more particularly to methods for recycling uranium, plutonium, and possibly minor actinides from spent nuclear fuel (UNF) for disposal of only the actual waste by-products of the nuclear fission process, i.e., the fission products. [Background technology]

[0003] Nearly all commercial thermal reactors are light water reactors (LWRs). LWRs are also thermal reactors, but they use deuterium ( 2It is a type of thermal neutron reactor that uses demineralized water as both coolant and neutron moderator, as opposed to heavy water reactors, which use a form of water containing hydrogen (H). In an LWR, the fast neutrons emerging from fission collide with hydrogen atoms in the water and are slowed down. This increases the probability that the neutrons will interact with the fuel atoms. LWRs use ceramic pellets made of 3-5% enriched uranium oxide as fuel. These pellets are loaded onto a thin-walled cylinder of Zircaloy cladding to stop the release of radioactive fission products (FPs) into the cooling water. Zircaloy is highly corrosion resistant and has a low neutron interaction cross section. The cylindrical tube is sealed with an inert atmosphere of helium gas to stop chemical reactions and increase heat transfer. These "fuel rods" may be arranged in assemblies, typically 17×17 for pressurized water reactors (PWRs) and 10×10 for boiling water reactors (BWRs). The assembly itself may be fabricated from stainless steel and provides structural support using end plates for the control rods and instruments, axially distributed grid spacers, expansion springs, and guide tubes.

[0004] As fuel atoms are split and consumed, the ability of the fuel pellet to sustain a chain reaction decreases. Over time, FPs accumulate and parasitic neutron losses increase. Furthermore, physical damage sustained by the fuel due to heat and radiation also limits its lifespan. Every 18-24 months, one-third of the fuel assemblies can be removed from the core and replaced with new ones. The remainder of the fuel assemblies can be shuffled to ensure consistent fuel burnup. Thus, the average time that a fuel assembly spends in the core is 4.5-6 years. When removed from the core, the spent nuclear fuel (UNF) generates intense heat and radiation from the short-lived FPs, which necessitates cooling and radiation shielding. Thus, the spent fuel assemblies can be stored on-site in pools of demineralized water. After spending about 10 years in this wet storage, the UNFs are packaged in dry casks made of steel and reinforced concrete and stored on concrete pads at the plant site. These casks can later be moved to a geological repository when they become available. This UNF is what is commonly referred to as "nuclear waste."

[0005] UNF is not completely spent fuel. The majority of UNF (>95%) is composed of U-238 with about 0.9% unburned U-235. During its time in the neutron field of the core, some of the U-238 atoms capture neutrons and turn into heavier uranium isotopes called "transuranium" or TRUs. Thus, about 1.0% of UNF may be composed of TRUs, which may be 0.9% unburned plutonium formed from neutron capture in U-238, as well as about 0.1% minor actinides such as neptunium, americium, and curium formed from successive neutron captures. The remainder is a mixture of FPs, many of which are already radioactively stable.

[0006] The radioactive composition of UNF can be as follows: about 1% short-lived (1-<50 year half-life) FPs, about 1% medium-lived (1000-<100,000 years) TRUs, and 95.5% very long-lived (>100 million years) uranium. Fission products are of greatest concern to the environment in the near term and generally require sequestration for hundreds of years. Transuraniums have half-lives of thousands of years and generally require geological sequestration for up to 250,000 years. They are of greatest concern when analyzing means of UNF disposal. However, if recovered, they could potentially serve as fuel. Uranium in UNF is harmless from a radioactive perspective, as its isotopes have half-lives ranging from 2 million to 4 billion years. Each category alone is fairly manageable, but mixed in UNF, they present radiological health risks and require costly monitoring and politically difficult disposal.

[0007] Reprocessing generally means effectively "closing" the fuel cycle by recycling uranium, plutonium, and possibly minor actinides, and disposing of only the actual waste by-products of the fission process, i.e., the fission products. Current commercial reprocessing methods are based on the Plutonium and Uranium Recovery by Extraction (PUREX) process or its variants (https: / / link.springer.com / chapter / 10.1007%2F1-4020-3598-5_24). PUREX is an aqueous process, meaning that UNF pellets must be dissolved in solutions of various acids of different strengths. Water, being an excellent neutron moderator, limits the size of the equipment due to nuclear criticality concerns that require larger facilities for sufficient throughput. The use of such acids and reagents also leads to the generation of large amounts of weakly radioactive chemical waste resulting from radiolysis. Furthermore, PUREX leads to the production of purified plutonium streams that raise proliferation concerns, so PUREX variants aim to minimize proliferation risks by either co-extracting minor actinides or denaturing the product. An alternative approach called pyroprocessing has also been developed but has not been industrially deployed (see https: / / www.sciencedirect.com / science / article / abs / pii / 0149197096000078 and https: / / www.sciencedirect.com / science / article / pii / B9780128197257002208). Pyroprocessing is generally a batch process in which solids are transferred from one unit operation to another. Summary of the Invention [Problem to be solved by the invention]

[0008] It would therefore be desirable to provide a system and method that overcomes the above limitations, which would use an improved process that integrates various unit operations into a series of connected vessels to accomplish the task of recycling nuclear waste. The system and method would use chemical reactions that are derivatives of dry reprocessing, fluorination, and electrowinning, and this process flow provides a significant improvement over existing methods for recycling UNF. [Means for solving the problem]

[0009] Summary of the Invention According to one embodiment, a single integrated system for recycling spent nuclear fuel (UNF) emerging from a reactor is disclosed. The system includes a decladding vessel that separates fuel pellets from nuclear fuel rods via oxidation to produce uranium compounds. A fluoridation vessel is coupled to the decladding vessel, where a fluoridation agent injected into the fluoridation vessel reacts with the uranium compounds to convert them to UF6. An electrowinning vessel coupled to the fluoridation vessel removes plutonium and actinides via electrowinning.

[0010] According to one embodiment, a method for recycling spent nuclear fuel (UNF) emerging from a reactor is disclosed, the method comprising separating fuel pellets from nuclear fuel rods via oxidation to produce uranium compounds, converting the uranium compounds to UF6 via fluorination, and removing plutonium and actinides via electrowinning.

[0011] The present application will be described in further detail with reference to the following drawings, which are not intended to limit the scope of the present application, but rather to illustrate certain attributes thereof. The same reference numbers will be used throughout the drawings to refer to the same or similar parts. [Brief description of the drawings]

[0012] [Figure 1]1 is a flow chart of an exemplary embodiment of a method for recycling spent nuclear fuel from a light water reactor, in accordance with an embodiment of the present invention. [Figure 2A] 2 is an exemplary embodiment of a decoating vessel used in the method shown in FIG. 1, according to one embodiment of the present invention. [Figure 2B] 2 is an exemplary embodiment of a decoating vessel used in the method shown in FIG. 1, according to one embodiment of the present invention. [Diagram 3] 1 is an exemplary diagram illustrating the chemical composition of spent nuclear fuel, according to one embodiment of the present invention. [Figure 4] 1 is an exemplary chart illustrating the generalized behavior of fission products in fluoride salts, according to one embodiment of the present invention. [Figure 5A] 2 is an exemplary embodiment of a fluorination vessel used in the method shown in FIG. 1, in accordance with one embodiment of the present invention. [Figure 5B] 2 is an exemplary embodiment of a fluorination vessel used in the method shown in FIG. 1, in accordance with one embodiment of the present invention. [Figure 6] 1 is an exemplary diagram illustrating the relative "stability" of different salt components that may be used in nuclear fuel recycling, according to one embodiment of the present invention. [Figure 7A] 2 is an exemplary embodiment of an electrowinning cell used in the method shown in FIG. 1, in accordance with one embodiment of the present invention. [Figure 7B] 2 is an exemplary embodiment of an electrowinning cell used in the method shown in FIG. 1, in accordance with one embodiment of the present invention. [Figure 8] 1 is an exemplary embodiment of a single integrated system for recycling spent nuclear fuel from a light water reactor, in accordance with an embodiment of the present invention. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0013] The description set forth below in conjunction with the accompanying drawings is intended as a description of the presently preferred embodiments of the disclosure and is not intended to represent the only manner in which the disclosure may be constructed and / or utilized. The description sets forth functions and sequences of steps for constructing and operating the disclosure in conjunction with the illustrated embodiments. However, it should be understood that the same or equivalent functions and sequences may be accomplished by different embodiments that are also intended to be encompassed within the spirit and scope of the disclosure.

[0014] Pyrochemical reprocessing may be defined as a process in which materials are subjected to high temperatures (typically above 800°C) to effect chemical or physical changes. Nuclear pyrochemical reprocessing may be defined as a high-temperature electrochemical fuel processing technique to recycle spent fuel into metallic fuel for nuclear fast reactors. Nuclear pyrochemical reprocessing utilizes electrorefining to separate uranium, plutonium + TRUs, and FPs by using a suitable electrolyte solution (e.g., molten LiCl-KCl eutectic). The corrosive salts and high temperatures result in significant stresses on containment materials. The high radiation fields encountered due to the shorter cooling time before reprocessing make the use of sensors and probes for continuous monitoring more difficult. This process is more suitable for metallic fuels and is therefore used in the IFR project, but has been tested with oxide fuels by reducing the fuel to metallic form before processing (https: / / www.sciencedirect.com / science / article / pii / S1002007115001197). Although originally studied as a batch process, recent work has experimented with continuous operation by removing FPs from the electrolyte solution by passing it through a zeolite column (https: / / www.sciencedirect.com / science / article / pii / B9781845695019500104).

[0015] Exemplary system and method embodiments disclose an improved process that integrates various unit operations into a series of connected tanks to fulfill the role of recycling nuclear waste. The chemistries are derivatives of dry reprocessing, fluorination, and electrowinning, and the process flows provide significant improvements over existing methods for recycling UNF.

[0016] Although the exemplary system and method embodiments aim to optimize the recycling of UNF emerging from LWRs, applicability can extend to metals, molten salts, and other fuel forms. The unit operations are adaptations / modifications of the chemical processes under development. The chemical reactions consist mainly of redox reactions driven by the use of reagents or application of suitable potentials. The bulk component of UNF, uranium, can be removed in a single step by conversion to gas allowing improved decontamination factors. After removing the uranium, the "ash" containing the TRU fluorides along with short-lived fission products that form stable fluorides, is in molten salt form, which is distinct from PUREX, where the fission product stream is separated in a first step. The electrowinning process uses a potential to activate the deposition of actinide fluorides (at this stage only TRU) present in the molten salt phase as metals on a suitable cathode. Following this, the potential is increased to deposit the lanthanide species on a separate cathode. The reactive fission product species remain in the molten salt and can be separated through a combination of ion exchange, fractional crystallization, and distillation.

[0017] With reference to the figure, a system 1 and method can be disclosed that aims to optimize the recycling of UNF emerging from LWRs. The system 1 can integrate various sub-operations into a continuous process that allows the nuclear material to be transferred 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. Oxide fuel is not the most efficient conductor. At full power operation, the temperature of the pellet center is over 1200°C, while the outer surface of the pellet is close to 400°C, and the temperature of the cladding surface is about 300°C. Furthermore, FP gases can accumulate during operation, leading to cladding bulges and cracks. When there is significant bulging, the pellet surface presses against the inner wall of the cladding and can physically bond with it or sometimes even cause the cladding to break. These phenomena are collectively known as pellet-cladding interactions. Thus, "de-cladding" can be a necessary pre-processing step. De-cladding can take a variety of different forms.

[0018] According to one embodiment, the decladding method oxidizes UO2 to U3O8 in a dry oxygen atmosphere at slightly atmospheric pressure and at a temperature in the range of about 400-500°C, as seen in the following equation. Upon reaction with oxygen, the uranium oxide is pulverized to triuranium octoxide powder, increasing the pellet volume, which provides for efficient separation from the cladding. The general process is as follows: UO2+O2→U3O8

[0019] 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 results in a more finely divided product in the form of UO3.

[0020] When recycling metallic fuel, the pretreatment process separates the cladding from the metal, including any reactive binding material (e.g., sodium), if present, and then subjects the metallic fuel to the oxidation reaction described above to produce finely divided U3O8. The remaining operations remain the same in the case of metallic fuel.

[0021] It has been observed that the pulverized fuel has minimum and maximum particle sizes so that contamination from the fine residue is minimal. At these high temperatures, gaseous and other volatile fission products can be released, such as krypton, xenon, iodine, bromine, tritium, and the decay product radon that accumulates at the grain boundaries. Studies have shown that puncturing the cladding before oxidation can increase the rate of the reaction. The puncturing and shearing can release volatile FPs in the fuel rod plenum (i.e., the space within the cladding rod for gas accumulation). It is therefore envisioned that either of these unit operations can be carried out in a single space, regardless of the number of parallel units of subsequent operations, thereby minimizing the gas volume that requires purification.

[0022] In the process used herein, after some physical disassembly, i.e., removal of structural components, and drilling to allow efficient reaction, the UNF assembly may be loaded into a removable basket mechanism 10 (hereafter mechanism 10) as seen in FIGS. 2A-2B. The removable basket mechanism 10 may have a number of slots 12 formed therein. Each slot 12 may be of an appropriate size to secure a fuel rod 14 in an upright fashion. The vessel 16 may be provided with a removable housing 23 attached to an upper region of the vessel 16. The removable housing 23 may be used to provide internal atmospheric isolation and may include an outlet 20 for circulating atmospheric air and collecting any off-gases from the process.

[0023] The mechanism 10, loaded with spent fuel rods 14, may then be positioned inside a decladding vessel 16 (hereinafter vessel 16). The vessel 16 may be constructed of metal alloys capable of withstanding high temperatures without corroding, such as various stainless steels, Inconel, etc. The vessel 16 may be formed with a thickness sufficient to withstand the desired temperatures / pressures used during the decladding process.

[0024] The interior of vessel 16 may be formed of a high temperature ceramic material 16A. According to one embodiment, high temperature 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 surfaces of mechanism 10 and slots 12 formed therein.

[0025] An inlet 18 and an outlet 20 may be formed within and in communication with the interior of the vessel 16. The inlet 18 and outlet 20 may be formed to allow oxygen to enter the vessel 16 and various fission product gases to exit the vessel 16. A heater 22 may be coupled to the vessel 16. The heater 22 may be used to maintain the temperature of the vessel 16.

[0026] By positioning the fuel rods 14 inside the vessel 16, the vessel can be heated to over 400-500°C in a dry excess oxygen atmosphere to convert the uranium oxide to U3O8. Reaction with oxygen pulverizes the pellets into U3O8 powder, which the vessel 16 can collect in the lower portion of the vessel 16. Similarly, reaction with an O2 / O3 mixture at <400°C converts UO2 to UO3. The reaction can also simultaneously rupture the cladding of the fuel rods 14.

[0027] The finely divided U3O8 / UO3 powder formed during the decladding process is collected in the lower part of the vessel 16 for further processing, while the hollow zircaloy cladding tubes of the fuel rods 14 are recovered and can be purified through electrotransport, thereby recovering high purity hafnium-free zirconium.

[0028] According to one embodiment, the vessel 16 may serve a dual purpose in that the vessel 16 may be used for the recovery of zirconium after oxidation. Following the decladding process, the outer shell of the zircaloy cladding tube of the fuel rod 14 may be left behind. Molten salt with poor solubility for uranium oxide may be introduced into the vessel 16 by the mechanism 10 containing the fuel rod 14 and positioned between the cathode-plate and the anode+plate. A small amount of zirconium halide may be added to the salt to initiate electrotransport. Upon application of a suitable potential to the cathode-plate and the anode+plate, the zirconium from the fuel rod 14 is electrotransported / electrorefined through dissolution into 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. This process liberates U3O8 / UO3 particles that may be collected in the vessel 16 and can be transported to downstream processes. It should be noted that the above steps can be modified such that the vessel 16 is first filled with molten salt, followed by the de-cladding process.

[0029] The following reactions may occur: Anode: Zr → Zr n+ +ne - Cathode: Zr n+ +ne - =Zr

[0030] The cadmium cathode or other suitable cathode carrying its load of zirconium can be discharged at the bottom of the vessel 16 via the outlet 25. Any salts emerging therefrom can be easily separated. Meanwhile, upon cooling, the zirconium can be separated from the cadmium via phase separation through fractional crystallization. Similarly, in the case of solid cathodes, the deposit can be harvested by using mechanical means such as scraping or by applying a thermal shock that takes advantage of the difference in the thermal expansion coefficients of the cathode material and the deposited metal.

[0031] According to another embodiment, decoating can be performed by electrotransport of zirconium, after which the uranium oxide pellets can be collected at the bottom of vessel 16, and then, after draining the molten salt, the UO2 can be roasted under a dry oxygen atmosphere to form U3O8 powder, or under an O2 / O3 atmosphere to form UO3.

[0032] UNF has a nominal composition that can be seen in Figure 3. In conventional pyroprocessing, uranium oxide can be reduced to metals prior to an electrorefining process to recover and separate uranium and plutonium. However, pyroprocessing suffers from poor separation factor for uranium as well as the need to transfer materials between vessels.

[0033] For process efficiency, uranium should be recovered with minimal contamination. Some systems propose using uranium valence chemistry to collect it as volatile UF6. This can be accomplished by contacting it with a gaseous or liquid fluorinating agent. Many of the primary fission products form fluorides (as shown in Figure 4) and dissolve in the molten salt in this process.

[0034] For the separation of nuclear materials from UNF, volatile hexafluorides of U, Pu, and Np can be removed from complex matrices such as the pulverized U3O8 / UO3 / UO2 matrix containing FPs and TRUs, as presented in this proposed process flow. Efficient separation of the resulting volatile fluorides from each other can be achieved by their volatility, by modifying the conditions regarding their physical properties such as boiling and sublimation points. For these purposes, several strong fluorinating agents have been investigated, such as HF, F2, ClF3, BrF5, etc. Large-scale fluorination in such facilities relies on the storage, handling, and transportation of such fluorinating agents, which presents significant risks to human health and the environment. Such agents are corrosive, so their accidental release in UNF recycling facilities further complicates the accident prevention requirements of the design basis.

[0035] Alternatively, nitrogen trifluoride (NF3) can serve as an efficient fluorinating agent. NF3 is used in the electronics industry to remove silicon and silicon compounds during the fabrication of semiconductor devices and in cleaning microelectronic devices. Unlike the above-mentioned fluorinating agents, NF3 is not corrosive and does not react significantly with moisture or common acids and bases at room temperature. NF3 is also thermally stable to relatively high temperatures and is not sensitive to shock, so it can be safely and economically transported, stored, and handled for large-scale processes. Most importantly, NF3 possesses unique temperature-based selective reactivity with various components of UNF. This feature can be exploited to separate volatile fluoride-forming fission products from uranium and to efficiently separate uranium from other actinides with minimal cross-contamination.

[0036] The Gibbs free energies for the main reactions of NF3 with uranium metal and various uranium and plutonium compounds are presented below. The reactions are at 300°C, however there should be little variation at the temperatures envisaged for the process.

number

[0037] The free energies clearly indicate that reactions with metals are most favorable, however such reactions may sometimes proceed with the formation of a UF4 intermediate. Reactions with triuranium octoxide also proceed spontaneously, but may lead to the formation of intermediate acid fluorides. The bulk reaction will be affected by other factors such as temperature and how well the reactants are mixed, which may require optimization. Similarly, NF3 will convert PuO2 to PuF4, but note that the formation of hexafluoride is not favored.

[0038] When NF3 gas mixed with an inert medium such as helium or argon is permeated through the pulverized UNF mixture, triuranium octoxide volatilizes at the process temperature and readily forms UF6, which separates from the bulk material bubbles out of solution. At low temperatures, fission products that form volatile fluorides react with NF3 and are removed from the bulk material. While uranium volatilization occurs at higher temperatures, as will be discussed, unlike F2 gas, NF3 is not an effective agent to produce PuF6 and maintain volatility. The reaction is unfavorable and any PuF6 formed is readily reduced to PuF4 by UF4 in the medium salts, which acts as a reducing agent. Neptunium, present in UNF primarily as Np-237, may volatilize in this process, but only in a well-defined manner and at higher temperatures than uranium. Np-237 can be collected and is the feedstock for the production of valuable Pu-238, used in powering deep space satellites and rovers.

[0039] Starting with a UO3 reactant, a reducing agent such as hydrogen gas / ammonia can be used to reduce the UO3 to UO2.

[0040] If UO3 is reduced to UO2, the product can be contacted with a fluorinating agent such as HF to convert UO2 to UF4 before being contacted with a fluorinating agent strong enough to produce volatile fluorides.

[0041] Regardless of the starting uranium compound in this reaction, i.e., U3O8, UO3, UO2, UF4UO2F2, etc., the reaction of the uranium compound with NF3 to form volatile UF6 occurs at a well-defined temperature that allows for the separation of uranium from fission products and other transuraniums.

[0042] The outgoing gas will require treatment to separate the uranium from any remaining volatilized FP fluorides. This separation can be accomplished through selective adsorption and desorption of the trace gas components by passing over beds of NaF, MgF2, or alumina. NpF6 can be absorbed by a bed of MgO. One process developed in the past utilizes a bed of NaF pellets at 400°C, which irreversibly captures Nb, Ru, and Zr fluorides from the gas stream. The second adsorption bed is NaF at 100°C, which reversibly captures U and Mo fluorides. 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 be passed over a third bed of MgF2 at 150°C, which captures MoF6, thus obtaining a purified stream of UF6. Similarly, a rectification sublimation process can be applied whereby the gas stream is passed through columns held at different temperatures where individual fluoride species or groups of fluoride species condense while other components remain volatile and pass through.

[0043] If the fluoride molten salt fuel form is 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 fully favorable. The remainder of the operation is the same as for the other fuel types.

[0044] By first removing the uranium, the enriched fission products and minor actinides block the diffusion of plutonium. The uranium thus recovered has a very high decontamination factor and can be used directly as feed for enrichment, converted into breeding material for fast spectrum reactors, reconstituted with the recovered TRU mixture to fuel advanced reactors, or simply returned to Earth in the form of oxide.

[0045] Volatilization of uranium results in very high decontamination factors (>10) between uranium and other actinides and fission product contaminants. 5), which allows for the recovery of neptunium in UNF (composed almost entirely of Np-237), which can be cryo-distilled from the stream. Np-237 is the feed material in high neutron flux reactors to produce Pu-238, which is used in radioisotope thermoelectric generators (RTGs) for deep space probes.

[0046] Referring to FIG. 5A, a vessel 30 for the fluorination of U3O8 may be shown. The vessel 30 may be formed of a first tank 32. The first tank may be formed of a suitable alloy that is compatible with the high temperature corrosive salt environment and whose chemical components participate little in redox chemical reactions under the operating conditions of the proposed process. Various grades of high nickel alloys such as Inconel, as well as alloys of nickel and chromium with refractory metals such as Monel, are suitable candidates, as are carbon composites of lanthanide borides.

[0047] The first tank 32 may feature a double wall (i.e. a tank within a tank structure). The outer wall 34 of the first tank may act as a catch basin in case of leakage. The space 37 between the outer wall 34 and the inner wall 36 may be used to cool the first tank 32. According to one embodiment, the space between the outer wall 34 and the inner wall 36 may be cooled by circulating a coolant or the outer surface of the inner wall 36 may be constructed with a network of heat pipes providing passive cooling. According to one embodiment, a coolant such as NaK alloy may be used, which may also be the coolant in the heat pipes if that design is used. This serves two purposes: 1) the inner surface of the inner wall 36 is capable of constantly maintaining a layer of "frozen" salt that provides a barrier against chemical corrosion as the fluorination reaction takes place; and 2) Passive cooling is provided for the highly exothermic reaction that forms volatile fluorides.

[0048] The first tank 32 may have multiple inlets 38. The inlets 38 may be in communication with the interior of the first tank 32. The inlets 38 may be used to insert different agents into the first tank 32 during the fluorination process. In the present embodiment shown, the first tank 32 may have a first inlet 38A, a second inlet 38B, a third inlet 38C, and a fourth inlet 38D. In this embodiment, the first inlet 38A may be formed in a lower region of the first tank 32. The first inlet 38A may be used to insert a fluorination agent into the first tank 32. The second inlet 38B may be formed in an upper region of the first tank 32. The second inlet 38B may be used to insert a U3O8 slurry into the first tank 32. The third inlet 38C may be formed in a lower side region of the first tank 32. The third inlet 38C may be used to circulate a coolant within the space 37 formed in the first tank 32. The fourth inlet 38D may be formed in a side region of the first tank 32. The fourth inlet 38D may be used to circulate a molten salt within the first tank 32.

[0049] The first tank 32 may have multiple outlets 40. The outlets 40 may be in communication with the interior of the first tank 32. The outlets 40 may be used to remove different materials from within the first tank 32 during the fluorination process. In the embodiment shown, the first tank 32 may have a first outlet 40A, a second outlet 40B, a third outlet 40C, and a fourth outlet 40D. In this embodiment, the first outlet 40A may be formed in a lower region of the first tank 32. The first outlet 40A may be used to remove the molten salt inserted in the first tank 32. The second outlet 40B may be formed in an upper region of the first tank 32. The second outlet 40B may be used to remove UF6 from the first tank 32. The third outlet 40C may be formed in an upper side region of the first tank 32. The third outlet 40C may be used to remove the coolant circulating in 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 region of the first tank 32. The fourth outlet 40D may be used to remove the molten salt in the first tank 32.

[0050] The vessel 30 may have a second tank 39 coupled to the first tank 32. The second tank 39 may be used to circulate the molten salt through the first tank 32. The second tank 39 may have an inlet 42A formed on a lower region of the second tank 39. The inlet 42A may be coupled to a fourth outlet 40D of the first tank 32. The second tank 39 may have an outlet 44 formed on an upper region of the second tank 39. The outlet 44 may be coupled to a fourth inlet 38D of the first tank 32. The molten salt in the second tank 36 may exit the outlet 44 and flow into the fourth inlet 38D of the first tank 32. The molten salt may exit the first tank 32 via the fourth outlet 40D and return to the second tank 39 via the inlet 42A. The second tank 39 may have another inlet 42B formed on a side region thereof. Inlet 42B can be used to inject replacement salt into second tank 39.

[0051] A heater 46 may be coupled to the second tank 39. The heater 46 may be used to transfer thermal energy to the second tank 39 to control and maintain the liquid phase of the molten salt in the second tank 39 and thus the conditions in the first tank 32.

[0052] The vessel 30 may have a pump 46. The pump 46 may be coupled 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.

[0053] Referring to FIG. 5B, another fluorination embodiment may be disclosed. In this embodiment, the fluorination is performed without the use of a molten salt medium. When performed without the use of a molten salt medium, the uranium compounds (U3O8 or UO3 or UO2 or UF4 or UO2F2, hereafter (uranium compounds)) from the decladding step may be introduced into vessel 30'. Vessel 30' may be a fluidized bed furnace 70 (hereafter bed furnace 70). Bed furnace 70 may be made of the same materials as the other fluorination vessels 30 disclosed above.

[0054] The bed furnace 70 may have a similar structure to the first tank 32. The bed furnace 70 may feature a double wall (i.e., a bath within a bath structure). The outer wall 72 may act as a pan in case of leakage. The space 74 between the outer wall 72 and the inner wall 76 may be used to cool the bed furnace 70. According to one embodiment, the space between the outer wall 72 and the inner wall 76 may be cooled by circulating a coolant or the outer surface of the inner wall 76 may be constructed with a network of heat pipes that provide passive cooling. According to one embodiment, a coolant such as NaK alloy may be used and may also be the coolant in the heat pipes whose design should be used.

[0055] The bed furnace 70 may have a distributor plate 78. The distributor plate 78 may be positioned near a lower region of the interior of the bed furnace 70. The distributor plate 78 may extend across the interior of the bed furnace 70 forming a bed within the interior of the bed furnace 70. According to one embodiment, the distributor plate 78 may be formed of nickel.

[0056] The bed furnace 70 may have multiple inlets 80. A first inlet 80A may be formed near the top of the furnace vessel 70. The first inlet 80A may be where a finely divided uranium compound may be introduced into the bed furnace 70. A second inlet 80B may be formed at the bottom of the bed furnace 70. The second inlet 80B may be where a gaseous fluorinating agent may be introduced into the bed furnace 70.

[0057] The uranium compounds introduced through the first inlet 80A can be positioned on the distributor plate 78. When the gaseous fluorinating agent enters the bed furnace 70 through the second inlet 80B, the distributor plate 78 can bubble the gaseous fluorinating agent through the bed formed of the uranium compounds, causing the bed of the uranium compounds to expand as a result of fluidization. The fluorinating agent can react with the uranium compounds and convert them to UF6. As before, different reactions occur at different temperatures, resulting in the separation of uranium as UF6, while transuranium and non-volatile fission products remain behind.

[0058] The exothermic reaction may be cooled by a suitable coolant circulated in the space 74 between the outer wall 72 and the inner wall 76. In this embodiment, the bed furnace 70 may have a third inlet 80C formed in an upper side region. The third inlet 80C may be used to insert a coolant into the space 74 between the outer wall 72 and the inner wall 76. A first outlet 82A may be formed in a lower side region of the bed furnace 70 to remove the coolant circulating in the space 74.

[0059] A second outlet 82B may be formed in the upper region of the bed furnace 70 to allow the UF6 gas to exit from the top of the bed furnace 70. The UF6 gas may exit through a plug filter 84 coupled to the second outlet 82B. The plug filter 84 may be made of a suitable material such as sintered nickel. The plug filter 84 may prevent any solid particles entrained in the gaseous mixture that are larger than a certain size from exiting the bed furnace 70. A third outlet 82C is provided near the bottom of the bed furnace 70. The third outlet 82B may be used to remove the product remaining after fluorination and volatilization of UF6 from the bed furnace 70.

[0060] FIG. 6 may show a diagram for various fluorides relevant to UNF recycling. A eutectic mixture of LiF-CaF2 may serve as a very stable mixture with an acceptable eutectic melting point of 767°C. Dissolution of uranium and other actinide fluorides may serve to further reduce the eutectic point. Alternatively, LiF-NaF or LiF-KF with lower eutectic points may also be used depending on the engineering needs.

[0061] Electrowinning is one of the oldest known electrolytic processes and is widely used in modern metal recovery, mining, refining, and even wastewater treatment. Electrowinning has long been proposed as a means to extract actinides. In the process presented, plutonium and minor actinides remaining in the salt mixture after volatilization of uranium are quite suitable to be co-extracted on solid or liquid electrodes via electrowinning. Extraction can be controlled by controlling the applied potential, making the process suitable for scale-up while ensuring that the recovered plutonium is never pure.

[0062] Considering the large gap in reduction potential between actinides and lanthanides in molten salt solutions such as those presented herein, the actinides in the salts can be co-extracted. As the potential applied between the electrodes is increased, the lanthanide FP species with more negative reduction potentials can be extracted for disposal. In the proposed process, the molten salt mixture after fluorination is discharged into an electrowinning cell 50, as seen in Figures 7A-7B. The electrowinning cell 50 can have an inlet 66 formed therein. The inlet 66 can be used to allow the molten salt mixture to be discharged into the electrowinning cell 50. A heater 68 can be coupled to the periphery of the electrowinning cell 50. The heater 68 can be used to maintain liquid phase conditions of the molten salt mixture. In the case of salt combinations with high eutectic points, the cathode can have an electrically connected flat dish underneath to collect the deposited species in liquid form. The continuity of the cathode surfaces with the same reduction potential ensures the prevention of re-dissolution of the reduced metals into the fluoride medium.

[0063] The electrowinning cell 50 may consist of a central anode 52 and multiple cathodes 54 arranged circumferentially around the electrowinning cell 50. The cell 50 may consist of a circular or semicircular support rail 51 made of a conductive material, such as graphite, to accommodate the multiple cathodes 54 and provide electrical conduction. The support rail 51 may be positioned at the top of the cell 50 and electrically insulated from the body of the cell 50. According to one embodiment, the anode 52 may be a porous graphite anode. The cylindrical cathode 54 may be mounted on a motor 56 that rotates the cathode 54 to provide continuous relative motion between the molten salt and the cathode surface to ensure that there is a uniform deposition of the electrowinning actinides. An example reaction may be as follows:

number

[0064] A split shroud 58 may be positioned around the anode 52 to allow for capture of reactive fluorine gas formed on the anode 52. The split shroud 58 creates an isolated gas space for collection of fluorine gas formed on the porous anode 52. Within the electrowinning cell 50, a noble gas atmosphere must be maintained. The electrowinning cell 50 may have a noble gas inlet 60 formed therein. In this 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 for noble gas to be continuously bubbled and circulated through the mixture to provide a carrier gas to dilute the formed fluorine gas. The lower bed 62 of the electrowinning cell 50 may be sloped. The sloped lower bed 62 may allow for removal of insoluble FPs that may accumulate after processing. A sludge outlet 64 may be formed on a lower side region of the electrowinning cell 50 adjacent to the lowest region of the sloped lower bed 62. The sludge outlet 64 may be used to allow for draining and removal of sludge from the electrowinning cell 50.

[0065] Plutonium extracted in this way can be considered "proliferation-hardened" given the presence of other actinides (i.e., neptunium, americium, and curium) and contamination by fission products (rare earths in amounts >500 ppm). This process makes it impossible to use the plutonium for clandestine purposes without extensive processing.

[0066] The molten salt bath is then passed through a column of zeolite to adsorb any trace fission product impurities present and purify the salt sufficiently for reuse in the process. When sufficient amounts of reactive FPs (mainly cesium and strontium) have accumulated in the salt, they can be subjected to fractional crystallization to produce a mostly dispersed phase of reactive FP fluorides. This salt phase can then be extracted for disposal. Additionally, distillation can be used to purify the salt mixture leaving behind the reactive FP fluoride species.

[0067] System 1 integrates the various described sub-operations into a continuous process that allows the transfer of nuclear material from one operation to another with minimal operator intervention. This allows for significantly improved material control and metering, improved instrument positioning, enhanced control, while at the same time reducing the amount of unmetered material. The molten salt used in System 1 is not susceptible to radiation damage and does not lose its chemical properties. The molten salt can therefore be continuously reused in the process.

[0068] As seen in one embodiment shown in FIG. 8, system 1 combines, scales and optimizes the decladding tank 16, the tank for fluorination of U3O8 30 and the electrowinning tank 50 for throughput after considering nuclear criticality safety. An embodiment of system 1 may also utilize two or more units of sub-operations in parallel, either combined into a series of sub-operations or independently connected in series into a series of sub-operations. Several such modules of integrated system 1 can then be built into a larger processing plant. As processing requirements increase, more modules of system 1 can be added to the plant to improve economics.

[0069] The above description illustrates certain embodiments of the present application, but is not meant to be a limitation on its practice. The following claims, including all equivalents thereof, are intended to define the scope of the present application.

Claims

1. A single integrated system for recycling spent nuclear fuel (UNF) emerging from a reactor, comprising: a decladding vessel for separating fuel pellets from nuclear fuel rods via oxidation to produce uranium compounds; a fluorination vessel coupled to the decoating vessel, wherein a fluorination agent injected into the fluorination vessel reacts with the uranium compounds to convert the uranium compounds into UF ６ a fluorination tank for converting an electrowinning cell coupled to said fluorination cell for removing plutonium and actinides via electrowinning.

2. The fluorination tank comprises: A vessel having a double wall, the space formed between the double walls being used to cool the vessel; a distributor plate formed within the vessel for holding the uranium compounds injected into the fluorination vessel; 2. The system of claim 1, further comprising: a fluorinating agent inlet formed in said vessel for injecting said fluorinating agent into said vessel.

3. The distributor plate bubbles the fluorinating agent through the uranium compound on the distributor plate to expand the uranium compound, and the fluorinating agent reacts with the uranium compound to convert the uranium compound into UF. ６ The system of claim 2 .

4. The system of claim 2 , wherein a coolant is circulated through the space formed between the double walls to cool the vessel.

5. UF ６ 3. The system of claim 2 comprising a gas outlet formed in an upper region of the vessel that allows gas to exit the vessel from a top portion.

6. The system of claim 2 comprising an inset filter coupled to the gas outlet.

7. The system of claim 2 , comprising an inset filter coupled to the gas outlet, the inset filter being made from sintered nickel.

8. 3. The system of claim 2 including a product outlet formed in a lower region of said vessel for removing products remaining after fluorination and volatilization of said uranium compounds from said vessel.

9. The uranium compound is U ３ O ８ , U.O. ３ , U.O. ２ , U.F. ４ , or UO ２ F ２ 10. The system of claim 1, wherein the first and second inputs are one of:

10. 1. A method for recycling spent nuclear fuel (UNF) emerging from a reactor, comprising the steps of: Separating fuel pellets from nuclear fuel rods via oxidation to produce uranium compounds; The uranium compound is converted to UF via fluorination. ６ and converting it to and removing plutonium and actinides via electrowinning.

11. The uranium compound is converted to UF via fluorination. ６ includes passing a fluorinating agent through the uranium compound to expand the uranium compound, the fluorinating agent reacting with the uranium compound to convert the uranium compound to UF. ６ The method of claim 10, wherein

12. The uranium compound is converted to UF via fluorination. ６ can be converted to forming a distributor plate within the fluorination vessel; positioning the uranium compounds injected into the fluorination vessel on the distributor plate; and injecting said fluorination agent into said fluorination tank.

13. bubbling the fluorinating agent through the uranium compound on the distributor plate to expand the uranium compound, wherein the fluorinating agent reacts with the uranium compound to convert the uranium compound into UF ６ The method of claim 12, wherein

14. The fluorinating agent is F mixed with an inert gas. ２ , HF, ClF ３ , BrF ５ , or NF ３ 11. The method of claim 10, wherein the gas is one of, or a combination thereof.

15. The uranium compound is U ３ O ８ , U.O. ３ , U.O. ２ , U.F. ４ , or UO ２ F ２ 11. The method of claim 10, wherein the first and second electrodes are one of:

16. 1. A method for recycling spent nuclear fuel (UNF) emerging from a reactor, comprising the steps of: Separating fuel pellets from nuclear fuel rods to produce uranium compounds; The uranium compound is converted to UF via fluorination. ６ converting said uranium compound to UF via fluorination. ６ can be converted to forming a distributor plate within the fluorination vessel; positioning the uranium compounds injected into the fluorination vessel on the distributor plate; injecting a fluorination agent into said fluorination vessel; and removing plutonium and actinides via electrowinning.

17. bubbling the fluorinating agent through the uranium compound on the distributor plate to expand the uranium compound, wherein the fluorinating agent reacts with the uranium compound to convert the uranium compound into UF ６ The method of claim 16, wherein

18. The fluorinating agent is F mixed with an inert gas. ２ , HF, ClF ３ , BrF ５ , or NF ３ 20. The method of claim 16, wherein the gas is one of, or a combination thereof.

19. The uranium compound is U ３ O ８ , U.O. ３ , U.O. ２ , U.F. ４ , or UO ２ F ２ 17. The method of claim 16, wherein the first and second electrodes are one of:

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