Supercritical gasification separation device for high-level radioactive multi-state nuclear waste and cascade waste heat power generation method

Abstract

This invention discloses a supercritical gasification separation device for high-level radioactive waste in multiple states. The device is a hollow cylinder. Along the radial direction of the cylinder, from the inside out, the device comprises: a corrosion-resistant pressure-bearing layer, including an alloy matrix and a yttrium-stabilized zirconia coating; a radiation shielding layer, including a boron-polyethylene layer and a lead layer; a liquid metal cooling layer with spiral channels filled with liquid metal coolant; a waste heat recovery layer, consisting of serpentine wound heat exchange tubes connected to a three-stage heat recovery system; and a shock-resistant outer shell layer, including a carbon steel shell and hydraulic damping supports. Six independent temperature and pressure reaction zones are arranged along the axial direction of the hollow cylinder, with adjacent zones separated by a dynamic temperature-controlled filter, for simultaneously processing at least two different states of radioactive waste from gaseous, liquid, and solid states. Furthermore, this invention also discloses a cascade waste heat power generation method based on the supercritical gasification separation device for high-level radioactive waste in multiple states.

Description

Technical Field

[0001] This invention belongs to the field of nuclear waste treatment technology. More specifically, this invention relates to a supercritical gasification separation device for high-level radioactive multi-state nuclear waste and a cascade waste heat power generation method, which is applicable to the simultaneous treatment of gaseous, solid, and liquid waste and the recovery of radionuclides. Background Technology

[0002] In the field of nuclear power technology, nuclear power plants generate solid waste from replacing spent nuclear fuel rods, resin, and protective clothing systems. The primary and secondary loop water bodies that exchange between the nuclear reactor and the evaporator are contaminated by the replacement of nuclear fuel and radioactive nuclides released during operation. The incineration of protective clothing also produces highly radioactive gases and secondary pollutants such as dioxins through coking.

[0003] How to properly handle nuclear waste is one of the core challenges to the sustainable development of nuclear energy. The industry has adopted different technical routes for volume reduction, extraction and final disposal. However, the existing technologies for radioactive nuclear waste treatment have at least the following defects: 1) Limited nuclide recovery: Nuclide recovery is limited to uranium / plutonium (efficiency >96%), and the recovery rate of minor actinides (Am / Cm) is <40%; 2) High risk of secondary pollution, which easily generates dioxins / radioactive dust; 3) Low volume reduction rate: The volume reduction rate of traditional incineration is ≤50%. Although supercritical water gasification technology (SCWG technology) can improve the volume reduction rate to a certain extent, the volume reduction rate of SCWG technology is ≤80% (see Table 1), which cannot break through the 100-fold volume reduction bottleneck.

[0004] Table 1 Comparison of SCWG technology and traditional nuclear waste incineration technologies

[0005]

[0006]

[0007] Therefore, whether for preventing or mitigating the risk of nuclear contamination, there is an urgent need to develop a supercritical gasification separation device for high-level radioactive multi-state nuclear waste, a nuclear waste treatment method, and a cascaded waste heat power generation method. This device should meet the Level 1 safety requirements of the Nuclear Facility Systems and Components Safety Classification (HAD 102 / 03-2021), and its radiation shielding should comply with the dose limits specified in GB 18871-2002, in order to address high radioactivity (radioactivity > 4 × 10⁻⁶). 12 Bq / m 3 (GB 9133-1995 standard), large-capacity nuclear waste, and the disposal of nuclear pollution. Summary of the Invention

[0008] The purpose of this invention is to overcome at least one of the defects in the prior art and provide a supercritical gasification separation device for high-level radioactive multi-state nuclear waste and a cascade waste heat power generation method.

[0009] To achieve the above-mentioned objective, the present invention provides a supercritical gasification separation device for high-level radioactive multi-state nuclear waste, which is a hollow cylinder. Along the radial direction of the hollow cylinder from the inside to the outside, the supercritical gasification separation device for high-level radioactive multi-state nuclear waste comprises, in sequence:

[0010] The anti-corrosion pressure-bearing layer includes an alloy substrate and a wavy yttrium-stabilized zirconia coating disposed inside the alloy substrate. The alloy substrate is Ti / Haynes282 with a thickness of 90-100mm. The wavy yttrium-stabilized zirconia coating has a peak spacing of 1000±50mm and a peak height of 30±2mm.

[0011] The radiation shielding layer includes a boron polyethylene layer and a lead layer that are closely attached to the outer wall of the anti-corrosion and pressure-bearing layer. The thickness of the boron polyethylene layer is 50±5mm, and the thickness of the lead layer is 300±10mm.

[0012] The liquid metal cooling layer is made of 316H stainless steel with a thickness of 25±2mm. The liquid metal cooling layer has spiral channels, which are filled with gallium indium tin eutectic alloy Ga. 68.5 In 21.5 S 10 And embedded with cadmium foil poison tape;

[0013] The waste heat recovery layer consists of Inconel 690 serpentine wound heat exchange tubes with an outer diameter of 17±1mm, and is connected to a three-stage heat recovery system; and

[0014] The seismic outer shell layer consists of a 20±2mm thick SA-533B carbon steel shell and hydraulic damping bearings with a damping coefficient ξ≥0.9;

[0015] The system comprises six independent thermo-pressure reaction zones arranged along the axial direction of the hollow cylinder, with a temperature range of 380-1200℃ and a pressure range of 25-37.5MPa. Adjacent thermo-pressure reaction zones are separated by dynamic temperature-controlled filters. Each thermo-pressure reaction zone is equipped with four sets of supercritical water nozzles, including two sets of downward spray anti-clogging nozzles and two sets of material nozzles. Each thermo-pressure reaction zone is equipped with a slag discharge valve and an overpressure protection valve. The slag discharge valve is interlocked with the radio frequency deposition monitoring instrument, and the overpressure protection valve is linked with the temperature and pressure sensor. The six independent thermo-pressure reaction zones are used to simultaneously process at least two different physical states of radioactive waste from gaseous, liquid, and solid radioactive waste.

[0016] According to a preferred embodiment of the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of the present invention, the dynamic temperature-controlled filter screen includes a 316L stainless steel sintered mesh with a pore size of 0.1-0.5μm, a yttrium oxide-stabilized zirconium oxide support layer with a thickness of 10±1μm, and a Cr2O3 / CoWO4 anti-corrosion coating.

[0017] According to a preferred embodiment of the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of the present invention, the Cr2O3 / CoWO4 anti-corrosion coating of the dynamic temperature-controlled filter screen has a thickness of 5±0.5μm and is resistant to oxidation corrosion at 1300℃ and erosion by 5wt% H2SO4 solution.

[0018] According to a preferred embodiment of the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of the present invention, the spiral channels of the liquid metal cooling layer are circumferentially distributed at intervals of 600±50mm, with a channel depth of 10±0.5mm and a channel width of 700±20mm.

[0019] According to a preferred embodiment of the supercritical gasification separation device for high-level radioactive multistate nuclear waste of the present invention, the hydraulic damping support meets the IEEE 693-2018 seismic resistance standard and can attenuate seismic acceleration by 90±2%.

[0020] According to a preferred embodiment of the supercritical gasification separation device for high-level radioactive multistate nuclear waste of the present invention, the gallium indium tin eutectic alloy contains 68.5±0.5wt% Ga, 21.5±0.5wt% In and 10.0±0.5wt% Sn.

[0021] To achieve the above-mentioned objectives, the present invention also provides a method for treating nuclear waste based on the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of the present invention, comprising:

[0022] For the ionic nuclide Cs + 、Sr 2+ I - In-situ adsorption was achieved by injecting nano-zero-valent iron-modified zeolite into a supercritical gasification separation device for high-level radioactive multistate nuclear waste.

[0023] For high-density particulate nuclides PuO2 and UO2, zirconium phosphate magnetic beads are used for pre-adsorption in the supercritical gasification separation device for high-level radioactive multistate nuclear waste, and centrifuge separation is used.

[0024] For gaseous nuclides 85 Kr、 3 H, after cryogenic distillation at -196±5℃, uses a specific surface area ≥2400m² 2 / g of MOF-808 molecular sieve trapping.

[0025] According to a preferred embodiment of the nuclear waste treatment method of the present invention, the centrifuge is a titanium-zirconium alloy centrifuge with a rotation speed of 20,000±500 rpm, wherein the zirconium content of the titanium-zirconium alloy centrifuge is ≥80wt%, the centrifugal acceleration is >10,000g, and the density separation threshold is >4.0g / cm³. 3 .

[0026] To achieve the aforementioned objectives, this invention further provides a waste heat integration method based on the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of this invention, comprising:

[0027] Using waste heat from the primary loop of a nuclear power plant to preheat nuclear waste slurry entering a supercritical gasification and separation unit for high-level radioactive multistate nuclear waste; and

[0028] The secondary loop steam of the nuclear power plant and the supercritical gasification syngas at 300-500℃ are used together to drive the molten carbonate fuel cell.

[0029] According to a preferred embodiment of the waste heat integration method of the present invention, the steam pressure of the secondary loop of the nuclear power plant is ≥4 MPa, the power generation efficiency of the molten carbonate fuel cell is >50%, the total energy utilization rate of the system is >70%, and the radionuclide rejection rate is >99.9%.

[0030] According to a preferred embodiment of the waste heat integration method of the present invention, the waste heat integration method further includes: the turbine condensate is converted into low-pressure steam by vacuum flash evaporation to drive an absorption heat pump, and the decay heat is recovered through a gallium indium tin alloy heat pipe with a thermal conductivity >40W / m·K.

[0031] To achieve the aforementioned objectives, the present invention further provides a waste heat cascade power generation method based on the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of the present invention, comprising:

[0032] Level 1: Waste slurry from the primary loop of a nuclear power plant;

[0033] Level 2: Steam from the secondary loop of the nuclear power plant drives MCFC (Multi-Cyclic Fuel Cell) power generation;

[0034] Level 3: Supercritical syngas-driven SOFC power generation; and / or

[0035] Level 4: The fourth loop of the nuclear power plant is used for preheating SOFC.

[0036] Compared with existing technologies, the supercritical gasification separation device and cascade waste heat power generation method for high-level radioactive multi-state nuclear waste of the present invention have the following advantages:

[0037] The five-layer structure of the supercritical gasification separation device for high-level radioactive waste, the gradient temperature-pressure coupling method and rotary centrifugal separation technology, the monitoring measures for safe operation, and the deep integration of nuclear energy, supercritical gasification waste heat, and turbine condensate resource utilization systems, along with the targeted radionuclide classification adsorption-separation method of the "gradient temperature-pressure coupling method," creatively achieves supercritical gasification-separation treatment and system integration of various high-level radioactive wastes, including medical nuclear waste, nuclear-contaminated seawater, waste resin and waste protective clothing, refractory high-salt nuclear waste oil, rare earth symbiotic minerals and metal mixed solids, dioxins and coking products, within a large-scale supercritical gasification separation device. It also realizes the utilization of waste heat from nuclear power multi-loop heat, high-level radioactive waste treatment waste heat and device heat exchange, and turbine condensate, filling the gap in the equipment required for radioactive waste treatment. Attached Figure Description

[0038] Figure 1 This is a schematic diagram of the structure of the supercritical gasification separation device for high-level radioactive multistate nuclear waste of the present invention.

[0039] Figure 2 This is a schematic diagram of the structure of the supercritical gasification separation device and resource utilization integration device for high-level radioactive multi-state nuclear waste of the present invention.

[0040] 10--Supercritical gasification separation device for high-level radioactive multi-state nuclear waste; 100--Anti-corrosion and pressure-bearing layer; 1000--Nano coating; 102--Radiation shielding layer; 106--Waste heat recovery layer; 108--Shock-resistant outer shell layer; 110--Cavity; 112--Dynamic temperature-controlled filter. Detailed Implementation

[0041] The technical solutions of the supercritical gasification separation device, treatment method and waste heat integration method for radioactive nuclear waste of the present invention will be clearly and completely described below with reference to the accompanying drawings and specific embodiments. The described embodiments are only some embodiments of the present invention, and not all embodiments.

[0042] For example, in the supercritical gasification separation device for radioactive nuclear waste of the present invention, two or more gradient temperature and pressure zones can be set up, separated by a dynamic temperature-controlled filter, to simultaneously and synchronously process different types of nuclear waste. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without inventive effort are within the scope of protection of the present invention.

[0043] It should be noted that in the description of this invention, terms such as “center,” “upper,” “lower,” “horizontal,” and “inner” indicate directions or positional relationships based on the directions or positional relationships shown in the accompanying drawings. They are used merely for ease of description and do not indicate or imply that the device or element must have a specific orientation, or be constructed and operated in a specific orientation. Therefore, they should not be construed as limitations on this invention.

[0044] It should also be noted that, in the description of this invention, unless otherwise explicitly specified and limited, the terms "set," "install," "connect," and "link" should be interpreted broadly. For example, they can refer to a fixed connection, a detachable connection, or an integral, multi-layered connection; they can refer to a mechanical connection; they can refer to a direct connection or an indirect connection through an intermediate medium; and they can refer to the internal communication between two components. Those skilled in the art can understand the specific meaning of the above terms in this invention according to the specific circumstances.

[0045] Please refer to Figure 1 As shown, the present invention provides a supercritical gasification separation device 10 for high-level radioactive multi-state nuclear waste, which is a hollow cylinder, and comprises, from the inside to the outside, the following components along the radial direction of the hollow cylinder:

[0046] The anti-corrosion pressure-bearing layer 100 includes an alloy substrate and a nano-ceramic composite coating 1000 disposed inside the alloy substrate. The alloy substrate is Ti / Haynes282 with a thickness of 90-100mm. The nano-ceramic composite coating 1000 is a wavy yttrium-stabilized zirconia coating with a wavelength of 1000±50mm and a wave height of 30±2mm.

[0047] The radiation shielding layer 102 includes a boron polyethylene layer and a lead layer that are closely attached to the outer wall of the anti-corrosion pressure-bearing layer 100. The thickness of the boron polyethylene layer is 50±5mm and the thickness of the lead layer is 300±10mm.

[0048] The liquid metal cooling layer 104 is made of 316H stainless steel with a thickness of 25±2mm. It has spiral channels that are circumferentially distributed at intervals of 600±50mm, with a depth of 10±0.5mm and a width of 700±20mm. The spiral channels are filled with liquid metal coolant gallium indium tin eutectic alloy Ga. 68.5 In 21.5 S 10 And embedded with cadmium foil poison tape;

[0049] The waste heat recovery layer 106 consists of Inconel 690 serpentine wound heat exchange tubes with an outer diameter of 17±1mm, connected to a three-stage heat recovery system; and

[0050] The seismic outer shell layer 108 includes a carbon steel shell and a hydraulic damping bearing. The carbon steel shell is a SA-533B carbon steel shell with a thickness of 20±2mm. The damping coefficient of the hydraulic damping bearing is ξ≥0.9, which meets the IEEE 693-2018 seismic standard and can reduce seismic acceleration by 90±2%.

[0051] The hollow cylinder has six independent temperature and pressure reaction zones along its axial direction. Adjacent temperature and pressure reaction zones are separated by a dynamic temperature control filter 112, which is used to process different radioactive nuclear wastes individually or simultaneously. The temperature gradient of the temperature and pressure reaction zones is 380-1200℃, and the pressure gradient is 25-37.5 MPa.

[0052] To address the common shortcomings and challenges of existing technologies and achieve a volume reduction rate ≥ 100% and a radionuclide recovery rate > 99.9%, this invention provides the following technical solution:

[0053] The supercritical gasification separation device for high-level radioactive multi-state nuclear waste, the nuclear waste treatment method, and the waste heat integration method are a complete system integrating multiple subsystems, including: "supercritical gasification device system, nuclear waste fine crushing and feeding and supercritical water production and injection system, nuclear waste gradient temperature and pressure supercritical gasification reaction, nuclide adsorption, and rotary centrifugal separation system, nuclide collection tank recovery system, nuclear power thermal energy and waste heat utilization system of supercritical water gasification reaction of nuclear waste integrated system, and gasification device system safety operation and monitoring system".

[0054] Preferably, the supercritical gasification separation device system includes: a cylindrical supercritical gasification device with a height of 38 meters designed according to the needs of processing highly radioactive nuclear waste, with an operating temperature of 850-1200℃ and a pressure of 25-37.5MPa, an inner diameter of 4.9 meters, an outer diameter of 5.93 meters, a wall thickness of 0.515 meters, and a multi-layer structure.

[0055] Specifically, the shell design of the supercritical gasification separation device, treatment method and waste heat integration method for high-level radioactive multi-state nuclear waste is a static five-layer structure consisting of an inner layer (corrosion-resistant and load-bearing), a cooling layer, a shielding layer (lead and boron polyethylene), a serpentine winding heat exchange tube, a hollow core, and an outer shell (load-bearing and shock-resistant).

[0056] Preferably, the inner shell of the supercritical gasification device is 97 mm thick and made of Ti / Haynes 282, serving as the load-bearing and pressure-bearing layer. The inner shell material, Ti / Haynes 282, exhibits superior creep rupture strength in the 850-1200℃ range (maintaining ~20MPa at 1200℃) and better aging stability. The inner shell is coated with a 3mm "nano-ceramic composite coating or yttrium-stabilized zirconia coating", which serves as the anti-corrosion layer for the inner shell of the gasification device. The inner shell and the inner side of the inner shell together form a 100mm thick inner layer, which is the main load-bearing structure of the gasification separation device and can withstand a maximum reaction pressure of 37.5MPa. The "nano-ceramic composite coating or yttrium-stabilized zirconia coating" is a continuous wave-shaped particle spray with 5mm raised peaks and each wave segment being 1000mm long, i.e., the peak interval is 1000mm. This is to facilitate a longer contact time between the reactant gas and the inner coating, improve the heat transfer effect (the heat transfer efficiency of the wave coating is increased by 23%), and enhance corrosion resistance.

[0057] Depending on the needs, the "nano-ceramic composite coating or yttrium-stabilized zirconia coating" can be replaced every five years. The replaced coating may contain attached nuclides, which need to be finely crushed and ground before being sent to the gasification unit for gasification. The remaining sediment can be discharged into the collection tank through the slag discharge safety valve (lower). This design can extend the service life of the gasification unit while enhancing the heat transfer effect.

[0058] Preferably, a radiation shielding layer is designed on the outer side of the inner layer of the supercritical gasification device. The radiation shielding layer is composed of a double-layer composite of "boron-containing polyethylene + lead (50mm + 300mm)" and each layer has the function of "slowing down neutrons and shielding gamma rays".

[0059] Preferably, a cooling layer, made of 25mm thick 316H stainless steel, is designed on the outside of the anti-radiation shielding layer of the supercritical gasification device. Twelve spirally ascending liquid channels, each 10mm deep and 700mm wide, are spaced 600mm apart along the circumference of the stainless steel cooling layer. Radial ribs (20mm thick) are added between the channels to enhance the rigidity of the cooling layer. A 0.5mm thick cadmium foil neutron poison strip (60% coverage) is embedded within the liquid channels. The coolant in the liquid channels is a liquid metal—Ga68.5In21.56Sn10, with a melting point of -19℃, in a room-temperature liquid state, requiring no preheating system. The thermal conductivity of liquid metal is 16.5 W / (m·K), which is 40% higher than that of sodium-potassium alloy, and it can efficiently dissipate decay heat (>5 MW / m); Ga (2.9 barn) and In (194 barn) have neutron absorption cross sections, which can significantly shield secondary neutrons (superior to lead's γ shielding).

[0060] On the outer side of the cooling layer of the supercritical gasification unit, a serpentine wound heat exchange coil is designed. The heat exchange coil is made of Inconel 690 (nuclear grade), with an outer diameter of 17 mm, laser-welded (porosity <0.1%), and the inner wall is CVD-deposited with SiC nanowires. The heat exchange coil uses... Zirconium alloy sealing sleeve, leak detection using electron beam welding + helium mass spectrometry (leakage rate <10%) -12 (mbar·L / s) to prevent nuclide infiltration. The serpentine heat exchanger tubes employ a stepped heat recovery design for heating. To achieve a temperature increase of 200℃, three sets of heat exchanger tubes can be connected in series: ① liquid metal → heat pipe (80→120℃), ② heat pipe → steam (120→160℃), ③ steam superheating (160→200℃); step-by-step heating avoids local overheating. The high-temperature water (steam) from the serpentine winding heat exchanger coils is mainly used to preheat the supercritical water in the supercritical gasification device. A 5mm hollow space is left between the serpentine winding heat exchanger coils and the outer shell, which helps to lock in heat energy and facilitates the maintenance of the heat exchanger coils.

[0061] The outer shell (outer layer) of the supercritical gasification separation unit for high-level radioactive waste is made of 20mm thick SA-533B carbon steel as a seismic-resistant structural layer, reinforced with external ribs to increase the unit's buckling resistance. The base of the outer layer is designed with a self-balancing hydraulic damping support, which can attenuate seismic acceleration by 90%.

[0062] Understandably, as needed, annular reinforcing rings (external ribs) can be added to the outer shell of the supercritical gasification unit, spaced 2.5m apart, with a cross-sectional modulus ≥3000cm². 3 This is to achieve thermal stress management. At the same time, the inlet of the jet nozzle assembly for nuclear waste and supercritical water, as well as the outlet for the remaining particles and mixed gas after treatment, are reinforced with protruding parts on the outer shell of the inlet and outlet, which both strengthens the structure and protects against radionuclide leakage.

[0063] Nuclear waste fine crushing and feeding system

[0064] Preferably, to avoid coking, coking, and affecting the gasification rate of highly radioactive nuclear waste during supercritical water gasification, the nuclear waste fine crushing and feeding system of the gasification separation device includes: a fine crushing and grinding mill-water mixing machine system, a nuclear waste jet box, and a nuclear waste jet transmission component.

[0065] The fine grinding and pulverizing mill pre-treats radioactive waste resin and other nuclear waste materials by finely grinding them to sub-millimeter size (<200μm). Together with a water mixing system, the mill processes the radioactive waste resin and other nuclear waste materials into a homogeneous slurry, increasing the particle specific surface area and accelerating mass transfer and reaction kinetics. The fine grinding and pulverizing mill is connected to a nuclear waste jetting box.

[0066] The nuclear waste jet container, as a system for pre-treating nuclear waste, is designed similarly to the radionuclide collection container recovery system. Each container is a 1.8-meter-high cylinder (1.2m in diameter), made of 316L stainless steel (inner layer + anodized aluminum coating) → antimony-lead alloy (350mm) → C1220 phosphorus deoxidized copper (outer layer), which is corrosion-resistant, creep-resistant, and has optimal shielding efficiency and mechanical strength. The cylinder is filled with argon gas (0.15MPa) to achieve corrosion resistance and structural stability. A hydrogen recombination device (catalytic recombination of H2 and O2) is installed at the top of the cylinder to prevent and eliminate the risk of radiolysis gas explosion.

[0067] Preferably, the nuclear waste fine grinding and feeding system must ensure low-temperature injection and rapid heating gasification. Considering that excessively high nuclear waste concentrations (e.g., >10wt%) will promote coking—for example, the carbon gasification rate (CE) of coke powder at 1wt% concentration and 850℃ reaches 83%, while the CE drops below 65% when the concentration increases to 5wt%—the nuclear waste fine grinding and grinding mill, along with a water stirring system operating at temperatures below 300℃, must control the nuclear waste slurry concentration at 0.001-0.01wt%, and add a dispersant such as sodium dodecyl sulfonate at 1% of the resin weight. Stirring and mixing are essential to prevent agglomeration and ensure the slurry's Zeta potential is <-30mV, maintaining electrostatic stability. Helium gas temperature control (maintaining a temperature below 300℃) and a nozzle using a "cooling jacket + high-pressure deionized water injection" system rapidly inject the nuclear waste slurry into the gasification device (supercritical water temperature above 850℃) to achieve instantaneous heating and gasification, preventing long-chain organic matter from condensing and coking during the heating phase.

[0068] Preferably, the nuclear waste fine crushing-feeding system is equipped with an anti-clogging flushing system: a medium-low temperature water flushing system (pressure 25-37.5MPa, temperature 300℃) is used to flush the pipeline. When the material temperature in the nuclear waste mixing mixer exceeds or becomes abnormal, the low temperature water automatically flushes the mixer, pipeline, and nozzles to prevent particulate matter deposition and coking from occurring in the nuclear waste feeding system due to temperatures exceeding 300℃.

[0069] The material jet box and material jet transmission components of the nuclear waste fine crushing and feeding system include a material jet bundle tube and a material jet anti-backflow high-pressure nozzle. The nuclear waste material jet box is connected to the material jet bundle tube, which is fixedly connected to the outer flange of the gasifier device and passes through the outer wall of the gasifier device by 0.515 meters. It is connected to the material jet anti-backflow high-pressure nozzle through a multi-hole high-pressure bundle tube. The material jet anti-backflow high-pressure nozzle is fixed on the gasifier device shell and communicates with the internal cavity of the gasifier device. The material jet bundle tube and the material jet anti-backflow high-pressure nozzle are fixed on the inner and outer flanges of the gasifier device shell. The flanges are located on the outside of the gasifier device shell, with thickened protruding parts made of Ti / Haynes282, which can strengthen the connection between the flange and the shell and the rigidity of the shell, and prevent radionuclide leakage.

[0070] The connecting pipe bundles between the water-like nuclear waste material and the flanges and high-pressure nozzles, and the connecting pipe bundles between the safety valve (below) and the external nuclear waste material collection tank, all adopt multi-layer defense measures to prevent nuclear leakage, namely "corrugated pipe sealing + lead shielding sleeve". The welded joints must be electron beam welded and helium leak detected.

[0071] Supercritical water production injection system

[0072] The supercritical water in the supercritical gasification unit is produced by electric furnaces using clean electricity such as wind and solar power. Dozens of high-voltage, low-energy-consumption, high-power clean electric furnaces can be combined to achieve a water supply of 2,000 to over 50,000 tons per hour. Six gradient temperature-pressure zones (temperature range 380-1200℃, pressure range 25-37.5MPa) are set up for six categories of specific nuclear waste to provide different levels of supercritical water. Multiple supercritical water jet collection tanks are available to meet the needs of handling one nuclear material and multiple gradient temperature-pressure zones. Each tank is designed based on the aforementioned hourly water supply and can adopt a cubic structure, primarily using Inconel 690 material.

[0073] Preferably, the supercritical water production injection system, in addition to the clean electric furnace, also includes a supercritical water collection tank, a supercritical water jet bundle pipe, and a supercritical water jet anti-backflow high-pressure nozzle, etc.; supercritical water is collected and enters the jet water tank, and is fixedly connected to the outer wall of the cylindrical gasification device shell by a flange through the supercritical water jet bundle pipe, passes through the 0.515-meter outer wall of the shell, and is fixedly connected to the internal cavity of the gasification device shell through the porous bundle high-pressure pipe and the supercritical water jet anti-backflow high-pressure nozzle.

[0074] The supercritical water jet bundle tube and the supercritical water jet anti-backflow high-pressure nozzle are fixed on the inner and outer flanges of the gasification device shell. The flange is located on the outside of the gasification device shell, with a thickened protruding part. The material is Ti / Haynes282, which can strengthen the connection between the flange and the shell and the rigidity of the shell, and also prevent nuclide leakage.

[0075] The connecting pipe bundles between supercritical water and flanges and high-pressure nozzles, and the connecting pipe bundles between safety valves (below) and external nuclear waste material collection tanks, all adopt multi-layered defense measures against nuclear leakage, including "corrugated pipe sealing + lead shielding sleeve". All welded joints must undergo electron beam welding and helium leak detection.

[0076] Nuclear waste gradient temperature and pressure supercritical gasification reaction, radionuclide adsorption, and rotary centrifugal separation system

[0077] The supercritical gasification unit is designed with a "gradient temperature-pressure coupling method" for supercritical gasification and separation of various highly radioactive nuclear wastes, such as medical nuclear waste, nuclear-contaminated seawater, waste resin and waste protective clothing, refractory high-salt nuclear waste and waste oil, rare earth symbiotic minerals and metal mixed solids, dioxins and coking products. This method provides supercritical water with different materials, temperatures and pressures to address the different physical and chemical characteristics of different nuclear wastes. This allows for the treatment of nuclear wastes within a single supercritical gasification unit using "gradient temperature-pressure coupling", as shown in Table 2.

[0078] Table 2. Connection and distribution systems for inlets and outlets of nuclear waste disposed of at various temperature-pressure coupling gradients.

[0079]

[0080]

[0081] Preferably, from the bottom of the inner shell of the supercritical gasification device upwards, six types of nuclear waste, from medical nuclear waste to dioxins, coking compounds, etc., are supplied with supercritical water at different rates (temperature range 380-1200℃, pressure range 25-37.5MPa) to the corresponding nuclear waste (injected into the supercritical gasification device) for gasification. Each temperature-pressure zone is 6 meters apart, and the six temperature-pressure zones are separated by a dynamic temperature-controlled filter. The dynamic temperature-controlled filter is 350mm thick, 4.9 meters in diameter, and supported and fixed by a structural bracket.

[0082] Taking the "medical nuclear waste" disposal area as an example, the supercritical water temperature is 380-420℃ and the pressure is 25-28MPa. The supercritical water injection system is divided into 4 groups, evenly distributed on the circumference of the 18.6-meter-long cylinder above the bottom of the gasification separation device: The first group of water injection system is located 550mm above the bottom (dynamic temperature control filter) of the gasification separation device (inside), and consists of 28 high-pressure nozzles arranged along the circumference, with the nozzles pointing upwards; the second group of water injection system is located 2 meters above the bottom of the gasification separation device (inside), and consists of 28 nozzles arranged along the circumference, with the nozzles pointing downwards; the third group of water injection system is located 3 meters above the bottom of the gasification separation device (inside), and consists of 28 nozzles arranged along the circumference, with the nozzles pointing upwards; the fourth group of water injection system is located 5 meters above the bottom (dynamic temperature control filter) of the gasification separation device (inside), and consists of 28 nozzles arranged along the circumference, with the nozzles pointing downwards.

[0083] In the "medical nuclear waste" disposal area, the nuclear waste material injection system is divided into two groups, evenly distributed on the circumference of an 18.6-meter-long cylinder above the bottom of the gasification separation device. The first group of material injection systems has 28 high-pressure nozzles arranged along the circumference, 1.35 meters above the bottom of the gasification separation device (inside), with the nozzles pointing upwards. The second group of material injection systems has 28 high-pressure nozzles arranged along the circumference, 4 meters above the bottom (dynamic temperature control filter) of the gasification separation device (inside), with the nozzles pointing upwards.

[0084] In the "medical nuclear waste" disposal area, 400mm above the bottom of the gasification separation device, a group of 12 safety valves (1 lower) are arranged along the circumference. Combined with the supercritical water from the second and fourth water injection systems (2 meters and 5 meters above the bottom of the gasification separation device, respectively) with nozzles specifically designed for flushing, the supercritical water, while reacting with the material in the gasification reaction, flushes away any possible particle nuclide accumulation on the "dynamic temperature-controlled filter" (350mm high at the bottom). This causes any possible particle accumulation to either re-participate in the supercritical gasification reaction or, due to flushing, enter the 12 slag discharge safety valves (1 lower), which are symmetrically arranged along the circumference at a height of 400mm above the bottom of the device and controlled (or timed) by a radio frequency instrument that measures the amount of particle accumulation. The radio frequency instrument is located on the 12 slag discharge safety valves. After being flushed, the particles accumulated on the dynamic temperature-controlled filter are discharged through the slag discharge safety valves (1 lower) (two valves in series), connected by a "corrugated pipe seal + lead shielding sleeve" tube bundle, and enter the nuclide collection tank connected to the outside of the gasification separation device.

[0085] In the "medical nuclear waste" disposal area, 5.25m above the bottom (dynamic temperature-controlled filter) of the gasification separation unit (inside), a group of eight safety valves (one on top) is installed along the circumference. This serves as a protection system for the gasification unit, connected to the external nuclear waste jet box by a "corrugated pipe seal + lead shielding sleeve" tube bundle. The safety valves are controlled by temperature and pressure gauges, opening when the temperature or pressure values ​​in the zone exceed the set values ​​to prevent overload, overheating, or overpressure of the gasification unit. Simultaneously, the temperature and pressure gauges are linked to the material jet system and the supercritical water system to reduce or stop the injection of nuclear waste and supercritical water into the gasification unit, thereby protecting the gasification unit.

[0086] In the "medical nuclear waste" disposal area, 5.5m above the bottom of the gasification separation device (inside) (dynamic temperature control filter), a group of 28 safety valves (outlets 1) are set along the circumference, and connected to the external nuclide separation-collection tank by a "corrugated pipe seal + lead shielding sleeve" tube bundle.

[0087] Preferably, the structural features of each gradient temperature and pressure reaction zone in the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of the present invention are as follows:

[0088] Each gradient temperature and pressure processing zone includes the following components: a dynamic temperature-controlled filter (0.35 meters thick) at the bottom of the gasification separation device, with the height reference point Z=0; supercritical water nozzles and material nozzles are specifically integrated / connected. The operating parameters and specific heights of each processing zone are shown in Table 3 below.

[0089] Supercritical water nozzle group: 4 groups, 28 nozzles evenly distributed circumferentially per group

[0090] Group 1: Height Z+h1, spray upwards.

[0091] Group 2: Height Z+h2, downward injection (anti-clogging design)

[0092] Group 3: Height Z+h3, spray upwards.

[0093] Group 4: Height Z+h4, downward injection (anti-clogging design)

[0094] Nuclear waste material nozzle assembly: 2 sets, 28 nozzles evenly distributed circumferentially per set

[0095] Group 1: Height Z+h5, spray upwards.

[0096] Group 2: Height Z+h6, spray upwards.

[0097] Safety valve system:

[0098] Bottom slag discharge valves: 12, height Z+h7 (interlocked control by RF instrument)

[0099] Overpressure protection valves: 8 in total, height Z+h8

[0100] Gaseous product outlets: 28, height Z+h9

[0101] Table 3 Operating Parameters for Each Processing Zone

[0102]

[0103]

[0104] Note: For example, the Z-meter of the nuclear-contaminated seawater zone is 6.0m, which equals the height of the medical zone (5.65m) plus a spacing of 0.35m.

[0105] 1. Anti-clogging mechanism: Downward nozzles (h2 / h4) directionally flush the dynamic temperature-controlled filter (350mm thick) to prevent the accumulation of radionuclide particles;

[0106] 2. Safety interlock: The overpressure valve (h8) is linked with the temperature and pressure sensor, and when triggered, the supply of material / supercritical water is cut off; the temperature and pressure sensor is installed on the overpressure safety valve.

[0107] 3. Sealing Standard: All external pipelines are sealed with corrugated pipes and lead shielding sleeves (50mm), with a helium mass spectrometer leak detection rate <10%. -12 mbar·L / s.

[0108] Nuclide centrifugal separation system

[0109] In each of the aforementioned "gradient temperature-pressure zones" related to "medical nuclear waste, nuclear-contaminated seawater, waste resin and waste protective clothing, refractory high-salt nuclear waste and waste oil, rare earth symbiotic minerals and mixed metal solids, dioxins and coking products," there are corresponding nuclear waste disposal areas. These zones are equipped with 28 separation safety valves (outlets) at the upper end and 12 separation safety valves (lower end). Depending on the disposal and separation process of different nuclear wastes, adsorbents are added to the "temperature-pressure" disposal zones, or rotary centrifugal separation systems are connected to the downstream end of the safety valves. After passing through the relevant adsorbents or rotary centrifugal separation systems, the syngas safety valve (outlet) and the slag discharge safety valve (lower end) are connected to radionuclide collection tanks. The eight preventative safety valves (upper end) are connected to nuclear material jet boxes. All external connections use "corrugated pipe seals + lead shielding sleeves" tubing bundles.

[0110] This invention relates to a supercritical gasification separation device, treatment method, and waste heat integration method for high-level radioactive multi-state nuclear waste. Addressing the separation requirements of radionuclides from different sources in supercritical gasification (SCWG) systems, and considering the physicochemical properties of the radionuclides and the limitations of existing technologies, it differentiates between "adsorbent injection within the gasification separation device" and "post-rotary centrifugal separation" processes, thereby upgrading the radionuclide separation scheme and enhancing its innovation.

[0111] Nuclide separation decision logic selects the separation path based on the nuclide's state of matter:

[0112] Ionized nuclides (Cs) + / Sr 2+ / I-)→In-situ chemical fixation, technical solution: inject nano-zero ferric modified zeolite (Cs / Sr) or AgNO3 (I-), reaction time 5-10 min, removal rate >99%, which is 3 times higher than the adsorption capacity of traditional aluminosilicate molecular sieves.

[0113] High-density particulate radionuclides (PuO2 / UO2) → supercritical centrifugation (density > 4 g / cm³) 3 Zirconium phosphate magnetic beads pre-adsorption + titanium-zirconium alloy centrifuge (20,000 rpm, density >4 g / cm³) 3 This can prevent the fixed filter from clogging.

[0114] Gaseous nuclides (85Kr / 3 H)→Cryogenic-adsorption combined, -196℃ cryogenic distillation + MOF-808 molecular sieve, tritium adsorption capacity 120mg / g.

[0115] I. Scenarios where centrifugal separation is not required

[0116] Nuclear power plant waste resin (Cs / Sr), technical solution: Inject "nano-iron modified zeolite adsorbent" into the SCWG gasification separation unit to capture Cs using its microporous structure. + / Sr 2+ Extruded filter elements prevent leakage.

[0117] Medical waste (I / Tc), technical solution: Add AgNO3 to generate "AgI precipitate", and combine with a quench tower to inhibit volatilization; 99mTc is enriched at the drain outlet through anion exchange resin.

[0118] For nuclear-contaminated seawater (H / Sr), the technical solution involves adsorbing 3H using MOF-808, and capturing 90Sr using a combined adsorption filter element (zeolite + alumina). Adsorbent is injected, adsorption time is 5–10 min, temperature is <500℃, and the radionuclide removal rate is >99% (Cs). + 、Sr 2 + Preferred adsorbent scheme: suitable for ionic and volatile nuclides (Cs / Sr / I), highly economical and without moving parts;

[0119] Concrete, technical solution: Sr extraction after SCWG acid leaching 2+ →Zeolite adsorption tower, microwave acid leaching + extrusion filter cartridge; centrifuges are prohibited.

[0120] II. Basis for Selecting Radionuclide Separation Processes

[0121] 1. Applicable scenarios for adsorbent injection: Nuclides and adsorbents can be rapidly fixed in a gasification separation device through chemical bonding (such as ion exchange, precipitation); nuclides exist in ionic or small molecule states and are easily captured by porous materials (such as Cs). + I-); Nuclides that are volatile at high temperatures need to be fixed in situ to prevent escape (such as medical I-131I).

[0122] 2. Rotary centrifugal separation is applicable to the following scenarios: nuclides exist in the form of high-density solid particles (such as PuO2, UO2); efficient separation is required by utilizing density differences (only applicable to actinide oxides in rare earth slag); and micron-sized particles that cannot be completely captured by adsorbents.

[0123] III. Multi-scenario Nuclide Separation Process Scheme

[0124] Table 4 Matching Table of Gasification-Separation Processes for Various Scenarios

[0125]

[0126]

[0127]

[0128] Note: In rare earth mineral processing, a combination of adsorption and centrifugation is required because zirconium phosphate magnetic beads adsorb Pu to form micron-sized composite particles, necessitating centrifugation to separate impurities. Rotary centrifugation is only suitable for the purification of actinide (Pu / Am) nuclides, and the centrifuge chamber is made of titanium-zirconium alloy (temperature resistant up to 900℃). Centrifuges are prohibited for the treatment of nuclides in concrete.

[0129] IV. Further Detailed Explanation of Different Disposal and Separation Procedures for Nuclear Waste

[0130] 1. High salt content

[0131] Adsorbent injection within the designated temperature-pressure zone: Zirconium phosphate nanosheets are added for in-situ uranium ion capture; air flotation separation: supercritical CO2 injection generates microbubbles that carry salt to the enrichment zone (no mechanical centrifugation required); temperature control: 650℃ ensures the fluidity of the salt melt and prevents coking and blockage. Efficiency: UO2 2+ Removal rate >99%, residual uranium concentration <0.1ppm.

[0132] 2. Metallic mixed solids

[0133] Nuclide forms:

[0134] Metal: 239Pu / 2 4 1 Am oxides (density > 10 g / cm³) 3 );

[0135] Mixture: 90Sr / Cs encapsulated in silicate (density 2.4 g / cm³) 3 ).

[0136] Separation scheme:

[0137] Metal nuclides: High-temperature SCWG releases PuO2 particles → Titanium chamber centrifuge (12,000 rpm, separation density > 4 g / cm³) 3 (particles);

[0138] Mixed radionuclides: Sr extracted by acid leaching after SCWG 2+ →Zeolite adsorption tower (no centrifugation required);

[0139] Using a rotary centrifuge integrated with an "eddy current separation module," the separation rate of non-magnetic metals (such as zirconium alloys) is >95%.

[0140] 3. Separation under extreme operating conditions

[0141] Table 5. Comparison of separation performance (after SCWG at 850℃ / 35MPa)

[0142]

[0143] Note: Adsorbents deactivate due to sintering at temperatures above 800℃ (e.g., activated carbon fails at temperatures above 450℃), but centrifugal separation efficiency is not affected by high temperatures.

[0144] 4. Core advantages of rotary centrifugal separation

[0145] Separation efficiency: Captures micron-sized particles with a density >1.2 g / cm3 at 18,000 rpm (efficiency >99.5%);

[0146] High temperature resistance: The titanium-zirconium alloy cavity (temperature resistance 900℃) is suitable for air intake at 850℃;

[0147] Anti-clogging design: Centrifugal force forces particles to move towards the wall, avoiding contact with the filter screen.

[0148] 5. Integrated process solution: centrifugal separation + catalytic decomposition

[0149] System process decision tree, example

[0150] A [SCWG gasification separation unit 850℃ / 35MPa] --> B [Quick cooling tower cools to 400℃ in 0.3s]

[0151] B-->C [Supercritical CO2 centrifuge]

[0152] C-->D [Dioxin Catalytic Decomposition Tower]

[0153] D-->E [Exhaust Gas Purification]

[0154] 6. Achieving deep purification of dioxins: catalysis-adsorption combined

[0155] The Cu-ZSM-5 catalyst decomposes dioxins into CO2 / H2O; residual trace amounts of dioxins are adsorbed by zeolite (adsorption capacity 120 mg / g).

[0156] 7. Scenarios requiring centrifugal separation

[0157] Rare earth symbiotic ore (Pu / Am), the technical solution is as follows: Zirconium phosphate-supported Fe3O4 magnetic beads are injected into the SCWG gasification separation unit to form PuO2-magnetic bead composite particles; the reacted mixed gas is then passed into a supercritical CO2 centrifuge (speed >15,000 rpm), achieving a separation density >4 g / cm³. 3 Actinium. Rotary centrifugation, speed 12,000–18,000 rpm, pressure >7.39 MPa, >99.9%, high nuclide extraction rate, separating solid particles (PuO2, UO2). Limited centrifugation scheme: only for the purification of high-value actinide nuclides (Pu / Am), requiring a corrosion-resistant centrifuge chamber.

[0158] V. Centrifugal Separation System Design

[0159] The system employs "rotary centrifugal filtration technology," which involves designing a dynamic separation cylinder with a spiral flow channel. The centrifugal sedimentation of nuclide salts is achieved through the swirling effect generated by the tangential injection of supercritical fluid. Simultaneously, an ultrasonic anti-scaling device with a frequency of 20-40 kHz is integrated for electrochemical dissociation. (Alternatively, "cascade centrifugal technology" can be integrated / connected: referring to relevant information on "supercritical centrifugal separation," the integrated / connected equipment and materials utilize a titanium-zirconium alloy centrifugal chamber with a rotation speed >20,000 rpm. A temperature control module (±0.1℃) maintains the density gradient of the supercritical CO2 medium to achieve isotope fractionation (separation coefficient 1.02–1.05).)

[0160] "Turbulence suppression technology" is employed: a porous media buffer layer (pore size ≤ 50 μm) is installed at the inlet of the centrifuge chamber of the integrated / connected equipment to reduce the Reynolds number to the laminar flow region (Re < 2,000), thereby reducing nuclide diffusion. Nuclide escape control measures employ "multi-stage adsorption protection technology": after centrifugal separation, the gas enters a cryogenic distillation tower (-250℃), where inert gas nuclides (such as Kr-85) are captured by liquid nitrogen, and volatile nuclides (such as I-129) are adsorbed by silver zeolite.

[0161] Employing "sealing technology": integrated / connected magnetic fluid dynamic seal, withstands a pressure difference of 35MPa, with a leakage rate of <10%. -8 Pa·m 3 / s.

[0162] 1. Parameter matching and optimized design to improve separation efficiency. For specific materials or components containing high levels of radionuclides, the core parameters of the gasification unit are adjusted to precisely control reaction conditions.

[0163] Table 6 shows the optimized design of core parameters for the gasification unit to improve separation efficiency.

[0164]

[0165] 2. Integration / Connection / Separation Equipment Matching Parameters:

[0166] Centrifugation stage: pressure 7–10 MPa, temperature 35–40 °C (maintaining SC-CO2 supercritical state).

[0167] Cryogenic distillation: Step cooling is used, with the first stage at -80℃ (CO2 removal) and the second stage at -196℃ (krypton and xenon capture).

[0168] Specifically targeting the following four types of difficult-to-handle nuclides, and taking into account the characteristics of supercritical water gasification reactions, specific separation and extraction processes and methods have been designed, as shown in Table 7:

[0169] Table 7 Specific Separation and Extraction Processes and Methods

[0170]

[0171] 5. Technical approaches and parameter design for improving volume reduction rate

[0172] To achieve the goal of "volume reduction rate ≥ 100 times and radionuclide recovery rate > 99.9%", the following systematic scheme is designed:

[0173] Gasification separation unit design: composite wall thickness ≥85mm (including 30cm Pb shielding), cold wall structure + multi-stage filtration core area.

[0174] Operating parameters:

[0175] Gasification: 500-600℃ / 25-28MPa, Ru / C catalyst, residence time ≥10min;

[0176] Separation: Centrifugation (20,000 rpm, 7-10 MPa) → cryogenic distillation (-196℃).

[0177] System Integration:

[0178] Energy recovery: Fuel cell power generation systems recover waste heat from the reaction, improving thermal efficiency by 52.4%;

[0179] Residue treatment: Cross-media sealing using "magnetic fluid + cryogenic adsorption" ensures that radioactive nuclides are completely confined throughout the gasification and separation process, preventing secondary pollution; after vitrification, deep burial is performed, resulting in a comprehensive volume reduction ratio of >20 times. This solution achieves precise adaptation of six types of nuclear waste through temperature and pressure gradient separation, and breaks through the separation limit with supercritical gasification extraction-intelligent adsorption-dynamic lattice solidification as the core.

[0180] Monitoring and prevention measures for safe operation

[0181] The present invention relates to a supercritical gasification separation device, treatment method, and waste heat integration method for high-level radioactive multi-state nuclear waste. The monitoring and prevention measures for safe operation include:

[0182] 1. Safety Interlock: Each temperature-pressure gradient range is equipped with an overpressure valve (h8), linked to its corresponding temperature-pressure sensor. Overheating or overpressure triggers the valve to cut off the material / supercritical water supply. In addition to the safety valve's built-in temperature-pressure sensor, other temperature-pressure sensors are installed in the supercritical water jet bundle and nuclear material jet bundle, penetrating into the inner layer of the gasification separation device simultaneously with the nozzle assembly. The nozzle assembly system features explosion-proof and backflow-proof protective plates and locking mechanisms.

[0183] 2. Each gradient temperature-pressure zone is equipped with an anti-clogging mechanism: downward nozzles (h2 / h4) are installed in each gradient temperature-pressure zone to directionally flush the dynamic temperature-controlled filter screen (350mm thick) in the zone to prevent the accumulation of radionuclide particles.

[0184] 3. Slag discharge valves h7 are installed in each gradient temperature-pressure range;

[0185] 4. Sealing Standard: All external pipelines are sealed with corrugated pipes and lead shielding sleeves (50mm), with a helium mass spectrometer leak detection rate <10%. -12 mbar·L / s.

[0186] 5. The nuclear waste jet box and the radionuclide collection container recovery system are each 1.8 meters high cylinders (1.2 meters in diameter), made of 316L stainless steel (inner layer + anodized aluminum coating) → antimony-lead alloy (350 mm) → C1220 phosphorus deoxidized copper (outer layer), which is corrosion-resistant, creep-resistant, and corrosion-resistant, with optimal shielding efficiency and mechanical strength; the cylinder is filled with argon gas (0.15 MPa) to achieve corrosion resistance and structural stability; a hydrogen recombination device (catalytic recombination of H2 and O2) is set at the top of the cylinder to prevent and eliminate the risk of radiation gas explosion.

[0187] 6. The nuclear waste fine crushing and feeding system is equipped with an anti-clogging flushing system: It features medium-low temperature water (pressure 25-37.5 MPa, temperature 300℃) to flush the pipelines. When the temperature of the nuclear waste mixing mixer exceeds or becomes abnormal, the low-temperature water automatically flushes the mixer, pipelines, and nozzles to prevent particulate matter deposition and coking in the nuclear waste feeding system due to temperatures exceeding 300℃.

[0188] 7. Safety Enhancement Measures: After adsorption or centrifugal separation, the SCWG synthesis gas must be filtered through a metal membrane before being connected to the high-temperature fuel cell for power generation. The metal filter membrane must meet the requirements of extreme operating conditions (high temperature, high radiation, corrosive gases), while also ensuring "gas purification accuracy (H2 purity > 99.95%), mechanical stability, and radiation resistance".

[0189] The requirements for metal film material selection and manufacturing design parameters are as follows:

[0190] 1. Radiation and corrosion resistance strengthening process

[0191] Surface modification: The Haynes 188 film surface is "laser clad Cr2O3+CoWO4 ceramic coating" (5μm thick), which improves the oxidation resistance at 1300℃ (ASTM authoritative test standard).

[0192] 2. Irradiation defect repair: The film is doped with "Y2O3 nanoparticles (1-3wt%)" to inhibit the growth of neutron irradiation voids through the pinning effect.

[0193] 3. Sealing edge treatment: The membrane-flange interface adopts "seamless metal O-ring seal" (material: Inconel718) to avoid high-temperature creep leakage.

[0194] 4. For sulfur-containing gases (such as H2S): Use "Pd-Cu / V composite membrane" (V layer absorbs S) to replace pure Pd alloy and avoid sulfur poisoning.

[0195] 5. For 137Cs aerosol: PI core-pore membranes with surface "sulfonation treatment" are selected to electrostatically adsorb Cs with a negative charge. + .

[0196] 6. Extreme temperature fluctuations: The membrane module is embedded with a "SiC fiber reinforced frame" with a thermal expansion coefficient that matches the metal layer, making it resistant to thermal shocks exceeding 800°C.

[0197] Table 8. Membrane used after radionuclide separation and synthesis gas collection.

[0198]

[0199] Note: The flux decay of the Pd-Cu / mica composite membrane was tested in an environment containing 90Sr aerosol (target: flux retention rate >90% after 1000 hours).

[0200] The Haynes188 layer is formed by "spark plasma sintering (SPS)," with grains refined to 20μm to improve creep resistance;

[0201] A flexible graphite pad is filled between the PI core pore membrane and the metal layer to compensate for the difference in thermal expansion.

[0202] 7. The mica layer can be embedded in the surface of the Pd-Cu alloy to form a "proton sieve" (H). + (Directly conducted to the fuel cell cathode); 8. The Haynes 188-layer laser-drilled area needs to be designed as a "radial flow channel" to reduce pressure loss by 30%.

[0203] Through the above design, the metal membrane system can operate continuously for more than 5 years in a supercritical environment of 500℃ and 22MPa, ensuring the safe power generation of high-temperature fuel cells, while achieving H2 purity > 99.97% and radionuclide rejection rate > 99.9%.

[0204] Preferably, at the fuel cell exhaust port, an integrated / connected dual iodine adsorption device "silver zeolite + activated carbon" is considered; for the PAFC electrolyte, an integrated / connected radiation-resistant phosphoric acid-doped polybenzimidazole (PBI) membrane is considered.

[0205] A deep integration system of nuclear energy, supercritical gasification unit and steam turbine condensate system

[0206] Preferably, the waste heat gas after separating nuclides using the supercritical gasification separation device, treatment method, and waste heat integration method for high-level radioactive multi-state nuclear waste (double-shell serpentine winding heat exchange tubes and safety valves (outlets)) is fully utilized; the multi-loop heat source of the nuclear reactor is utilized; and the turbine condensate is utilized. That is, the thermal energy from nuclear energy, supercritical gasification device, and turbine condensate is used for fuel cell power generation, preheating of nuclear waste, preheating of supercritical water, preheating of fuel cell inlet steam, and, according to different nuclear waste disposal temperatures, is introduced into the corresponding fuel cell for power generation. This achieves the integration of waste energy from nuclear energy, supercritical gasification device, and turbine condensate, realizes resource utilization, saves electricity consumption for nuclear power plant power generation and nuclear waste disposal, and achieves comprehensive water conservation.

[0207] Preferably, taking the treatment scenario of medical nuclear waste (including radioactive isotope medical waste) as an example, the integration of resource-based systems such as nuclear energy, supercritical gasification (SCWG) device waste heat recovery and fuel cell power generation technology is explained.

[0208] I. Characteristics of Medical Nuclear Waste and Waste Heat Characteristics of SCWG Treatment Process

[0209] Table 9 Characteristics of Medical Nuclear Waste and Waste Heat Characteristics of SCWG Treatment Process

[0210]

[0211] II. Waste Heat-Fuel Cell Matching Scheme

[0212] 1. Waste heat from medium-temperature gas (250-300℃) → Phosphoric acid fuel cell (PAFC)

[0213] Selection criteria: strong temperature adaptability, compatible with biomass gasification gas;

[0214] Model Parameters: Model: Doosan PureCell 400 (400kW module), Operating Temperature: 150-200℃ (requires waste heat preheating), Fuel: H2 produced by SCWG (purity >99%, no purification required); Power Generation Efficiency: 40%

[0215] 2. Low-temperature waste heat (60–150℃) → Absorption heat pump + electrolysis hydrogen production

[0216] Technical approach: A heat pump heats purified water from 60°C to 120°C, driving ORC power generation (efficiency 8-10%); the electricity is used in a PEM electrolyzer to produce hydrogen, supplying PAFC or hospital hydrogen fuel cell vehicles. Economic viability: A 200-bed hospital processes 50 tons of waste annually, producing 1.2 tons of hydrogen, sufficient to meet the needs of fuel cell forklifts.

[0217] III. Medical-Nuclear Power Plant Co-processing System (Large-scale Centralized Solution)

[0218] 1. Design of a regional centralized treatment plant

[0219] Using low-temperature steam from the secondary loop of a nuclear power plant to preheat medical waste reduces SCWG energy consumption by 30%; centralized processing improves the 99mTc recovery rate (>98%) and avoids the risk of iodine leakage from small-scale equipment.

[0220] 2. Radionuclide recovery and safety control

[0221] I-recovery: SCWG exhaust gas is processed through a silver zeolite adsorption column; the AgI precipitate is used for the regeneration of medical radioactive sources. 99mTc extraction: 99mTc is extracted from cooling water. m TcO4 - Enriched by anion exchange resin and purified by sublimation (purity 99.9%).

[0222] 3. Structural features:

[0223] The gasification separation unit has a double-layer shell (lined with Ti-6Al-4V alloy) and a pressure resistance of 30MPa; it monitors I aerosols in real time (threshold > 0.1 Bq / m³). 3 (Automatic triggering of the alkaline scrubbing tower).

[0224] Preferably, taking a 1,000 MW pressurized water reactor nuclear power unit as an example, the resource utilization and comprehensive design are based on the matching of loop system parameters and waste heat recovery technology:

[0225] System parameters and radioactivity characteristics of each loop

[0226] 1. Primary loop (reactor coolant system)

[0227] Temperature range: inlet approximately 290℃-295℃, outlet approximately 320℃-325℃, pressure 15.5MPa. (Radioactivity: coolant in direct contact with core fuel, containing fission products (such as...) 85 Kr、 133 Xe, 18 (e.g., F), which is highly radioactive.

[0228] Functions: In addition to heat transfer, it also acts as a neutron moderator, participating in reaction rate control.

[0229] 2. Secondary loop (energy conversion system)

[0230] Temperature range: Fresh steam: 270℃-290℃ (steam generator outlet); Pressure: 5-7MPa; Turbine exhaust steam: approximately 150℃-200℃ (waste heat recovery inlet); Condenser outlet: 30℃-40℃ (warm water discharge). (Radioactivity: No radioactivity during normal operation, physically isolated from the primary loop via the steam generator).

[0231] 3. Three-loop (circulating cooling water system)

[0232] Temperature range: inlet seawater / river water approximately 10℃-25℃, outlet warm wastewater approximately 30℃-40℃ (heating 6–10℃); atmospheric pressure; (radioactivity: no direct radioactivity, but warm wastewater may contain trace amounts of permeation products such as tritium (3H), with activity below environmental regulatory limits.)

[0233] 4. Four-loop system (not standard configuration, refers to waste heat recovery extension system)

[0234] For systems like nuclear heating or seawater desalination, the inlet temperature is approximately 80℃-120℃ (taken from secondary loop steam extraction); the outlet temperature drops to 50℃-70℃. The pressure is 0.5-1.5 MPa. (Radioactivity: completely isolated, no new radioactive risk.)

[0235] Preferably, the block diagram of the supercritical gasification separation device for high-level radioactive multi-state nuclear waste, the nuclear waste treatment method, and the waste heat integration method is as follows: Figure 2 As shown, the waste heat from nuclear energy multi-loop systems, supercritical gasification devices, and turbine condensate is used for fuel cell power generation, preheating of nuclear waste, preheating of supercritical water, and preheating of fuel cell inlet steam. Depending on the different nuclear waste disposal temperatures, the heat is introduced into the corresponding fuel cells for power generation, thus achieving the integration of waste energy from nuclear energy, supercritical gasification devices, and turbine condensate, and realizing resource utilization.

[0236] Resource-based integrated device frame Figure 2 As shown, the serpentine heat exchange tube outlet water in the double-shell integrated resource recovery unit and the steam turbine condensate of the nuclear power plant are used to preheat the supercritical water of the integrated resource recovery unit; the 220-280℃ heat source in the primary loop of the nuclear reactor is used to preheat the nuclear waste (jet box) material; the 30-40℃ heat source in the tertiary loop of the nuclear reactor is used for the seawater source heat pump for heating the park.

[0237] The syngas from the "dioxin coking" treatment at outlet 6 of the "temperature-pressure zone" is 500°C and combined with the 250-300°C from the secondary loop of the nuclear reactor for MCFC power generation; the residual heat at 200°C is then used for medium-temperature FC power generation.

[0238] The "rare earth symbiotic mineral mixed solids" process the synthesis gas at 400°C from outlet 5 of the "temperature-pressure zone" and combines it with the 250-300°C from the secondary loop of the nuclear reactor for MCFC power generation; the remaining low temperature is used for adsorption refrigeration.

[0239] The synthesis gas from outlet 4 of the "temperature-pressure zone" for the treatment of "high-salt refractory materials and waste oil" is 300-400℃, which is combined with the 250-300℃ gas from the secondary loop of the nuclear reactor for MCFC power generation; the rest is preheated intake gas.

[0240] The "organic resin protective suit" handles the syngas at 300-350°C in outlet 3 of the "temperature-pressure zone" and combines it with the 60-350°C of the fourth loop of the nuclear reactor for integrated / connected ORC power generation, for integrated / connected PEMFC cooling, and the residual heat is used to preheat integrated / connected SOFC.

[0241] The syngas from outlet 2 of the "temperature-pressure zone" of the "nuclear contaminated seawater" treatment is combined with the 60-350℃ of the nuclear reactor's fourth loop and used for integrated / connected concentrated brine salinity gradient power generation, which is used to power SOFC hydrogen production in reverse. The steam after the reaction is 120℃ and used for integrated / connected FC or ORC power generation.

[0242] Synthetic gas from outlet 1 of the "temperature-pressure zone" of the "medical nuclear waste" disposal area is combined with the 60-350℃ of the nuclear reactor's fourth loop at 300℃ to preheat the integrated / connected SOFC air cathode, and the residual temperature of 60℃ is used for domestic water.

[0243] Based on the above description, it can be seen that the deep integration of nuclear energy with supercritical gasification devices and hydrophobic systems has the following advantages:

[0244] 1. Solve the energy consumption problem of the gasification system by utilizing the stable heat source of the nuclear reactor;

[0245] 2. The syngas and extracted hydrogen obtained after separating radionuclides are used to cool the reactor core, thereby increasing the safety margin;

[0246] 3. Waste heat from high-temperature fuel cell power generation is recovered and used for waste liquid concentration. The design can achieve an energy utilization efficiency of >70%, which is 10 times higher than the volume reduction rate of traditional incineration methods.

[0247] 4. Based on specific materials, scenarios, and customer needs, utilize integrated / connected fuel cell systems for power generation (partially integrating / connecting cryogenic ORC systems to recover reaction waste heat). This involves integrating the serpentine heat exchange tubes between the shells, the syngas from the outlet of the "temperature-pressure zone" of the supercritical gasification unit, the waste heat from the turbine condensate drain, the high-temperature fuel cell power generation and its waste heat, and the multi-loop thermal energy of the nuclear reactor, thereby significantly improving the thermal efficiency of the nuclear waste disposal system. The deep integration of nuclear energy, the supercritical gasification unit, and the turbine condensate drain system integrates the aforementioned waste heat recovery processes, achieving integrated innovation in waste heat recovery systems.

[0248] 5. In the disposal of nuclear waste, in addition to the resource utilization of residual heat, integrated power generation systems, and the “energy nuclide chain PuO2→MOX nuclear fuel·Am2O3→space nuclear battery”, other resource utilization pathways that can be considered include “medical nuclide chain: 131|→thyroid cancer treatment agent·°Sr→'Y generator→Yttrium[°Y] microspheres (liver cancer radiotherapy) 13·223Ra→targeted drug for bone metastases; and industrial nuclide chain:·137℃s→industrial irradiation source·rare earth oxides→permanent magnet materials / catalysts”.

[0249] According to the present invention, a supercritical gasification separation device, treatment method, and waste heat integration method for high-level radioactive multistate nuclear waste can be designed as a "gradient temperature and pressure-pulse centrifugation coupled medical radionuclide purification device". Metal-organic framework (MOF) materials are used to design highly selective actinide adsorbents, such as UiO-66-NH2, and an adsorption-centrifugation purification process is employed to achieve a purification target of >99% purity. This enables the recovery of radionuclides such as 39Pu and 99mTc (99mTc recovery purity >99.5%, meeting EP10.0 pharmacopoeia standards). This device can be applied to the production of medical isotopes, promoting the resource utilization of radionuclides in the medical and industrial chains.

[0250] The present invention relates to a supercritical gasification separation device for high-level radioactive multi-state nuclear waste, a method for treating nuclear waste, and a waste heat integration method, the steps of which are as follows:

[0251] Supercritical gasification-separation steps: In each gradient temperature and pressure zone, refined hydrous nuclear materials (e.g., high-salt refractory waste oil, organic resin protective clothing, nuclear-contaminated seawater, medical nuclear waste, mixed solids) from a specific nuclear material jet tank at a set temperature and pressure, along with supercritical water at a set temperature and pressure from a supercritical water jet tank, and the required adsorbent, are injected into the inner shell of the supercritical gasification unit. The specific hydrous nuclear materials and supercritical water undergo a supercritical gasification reaction in their respective zones, generating high-temperature gas. The required adsorbent binds to the nuclides in the nuclear materials and is adsorbed by the dynamic temperature-controlled filter in that zone. Under the set downward supercritical water flushing, the gas passes through the slag discharge safety valve next to the dynamic temperature-controlled filter in that zone and enters the external nuclide collection tank. The remaining high-temperature gas exits through the metal membrane at the top of the zone. Depending on customer needs, the syngas can be used to extract pure hydrogen and carbon dioxide, or it can all enter an external fuel cell for power generation, after which carbon dioxide and other inorganic substances are separated.

[0252] Specific refined hydrous nuclide materials requiring separation by a rotary centrifuge and with set temperature and pressure (such as dioxin coking compounds and rare earth symbiotic mineral mixtures) undergo supercritical gasification reactions with supercritical water injected into the designated zone, generating high-temperature gas. This gas is then connected to a rotary centrifuge at the upper outlet of the zone to separate nuclides, which enter a collection tank. The resulting syngas passes through a metal membrane, where pure hydrogen and carbon dioxide can be extracted, or it can be entirely fed into an external fuel cell for power generation, after which carbon dioxide and other inorganic substances are separated. If any nuclides or particles remain in the zone, they are flushed downwards by supercritical water and then discharged through a slag discharge safety valve next to the dynamic temperature-controlled filter in the zone, entering an external nuclide collection tank.

[0253] Resource integration steps: After gasification and separation of specific refined water-containing nuclear materials within various temperature and pressure gradient ranges, the syngas passes through a metal membrane. Before entering the fuel cell, it is combined and integrated with the thermal energy of the reactor loop, thereby improving the input thermal energy and power output of the fuel cell. Based on the characteristics of the reactor loop thermal energy, the primary loop thermal energy is used to preheat the specific refined water-containing nuclear materials, while the secondary loop thermal energy is used to combine with the syngas separated from radionuclides within the high-temperature and high-pressure range (e.g., high-salt refractory waste oil, dioxin coking products, rare earth symbiotic mineral mixtures), before entering the fuel cell. The heat energy from the third loop is used to heat the seawater source heat pump and provide heating for the industrial park. The heat energy from the fourth loop is combined with the syngas from the separation of radionuclides in the high-temperature and high-pressure range (e.g., organic resin protective clothing, nuclear-contaminated seawater, medical waste nuclear materials) to preheat the SOFC in the gradient temperature and pressure range. If there is no specific fourth loop, the steam turbine condensate is used to preheat the SOFC in the gradient temperature and pressure range. The hot water generated by the serpentine winding heat exchange tubes on the outer side of the inner shell of the supercritical gasification separation device (combined with the steam turbine condensate of the nuclear power plant) is used to preheat the supercritical water.

[0254] In addition to the integrated power generation system that utilizes waste heat from the energy nuclide chain PuO2→MOX nuclear fuel·Am2O3→space nuclear battery, other resource utilization pathways for nuclear waste disposal include: "medical nuclide chain: 131|→thyroid cancer treatment agent·°Sr→'Y generator→Yttrium[°Y] microspheres (liver cancer radiotherapy) 1 3·223Ra→targeted drug for bone metastases; and industrial nuclide chain:·137℃s→industrial irradiation source·rare earth oxides→permanent magnet materials / catalysts".

[0255] Compared with existing technologies, the beneficial effects of this invention are as follows: This invention's supercritical gasification separation device, treatment method, and waste heat integration method for high-level radioactive multi-state nuclear waste, through structural innovation in the gasification separation device, innovation in nuclear waste treatment methods, and innovation in resource utilization integration, in the field of nuclear waste volume reduction and resource utilization technology, utilizes a five-layer composite structure supercritical gasification device. Through the innovative design of the five-layer composite structure and vertical partitioned temperature and pressure gradient control (380-1200℃ / 25-37.5MPa), it achieves for the first time a single device... This invention enables simultaneous zoned treatment of six types of nuclear waste. It solves the global challenge of SCWG salt deposition by combining "in-situ radionuclide adsorption / rotary centrifugation, dynamic temperature-controlled filter + directional flushing nozzle." Furthermore, it integrates waste heat from nuclear energy, supercritical equipment heat exchange, and syngas and turbine condensate drainage, achieving cascaded zero-carbon power generation, avoiding secondary nuclear contamination, and reducing electricity and water consumption in nuclear waste disposal. Radionuclide recovery achieves a volume reduction rate ≥100 times and a radionuclide recovery rate >99.9% (test standard: ISO17025:2017).

[0256] 2. Handling methods

[0257] Table 10 Gradient Temperature and Pressure Zoning Control and Processing Methods

[0258]

[0259]

[0260] Nuclide separation technology

[0261] Separation method selected based on nuclide morphology:

[0262] Ionized nuclides (Cs) + / Sr 2+ In-situ adsorption of nano-iron modified zeolite

[0263] High-density particles (PuO): Titanium-zirconium alloy centrifuge (speed 20,000±500 rpm, density separation threshold >4.0 g / cm3, conforming to ANSI / ANS 55.1-2021);

[0264] Volatile nuclides ( 85 Kr): Cryogenic distillation (-196℃) + silver zeolite adsorption (integrated / connected)

[0265] 3. Resource integration

[0266] Waste heat recovery system:

[0267] Medium-temperature waste heat (250-300℃) → Integrated / connected drive phosphoric acid fuel cell (PAFC)

[0268] Low-temperature waste heat (60-150℃) → integrated / connected absorption heat pump + integrated / connected ORC power generation integrated nuclear power plant primary / secondary loop waste heat, the total system heat recovery rate is >52.4% (boundary conditions: inlet temperature 25℃, pressure 0.1MPa); the waste heat from the supercritical gasification separation unit, the preheating of the outlet syngas, the heat energy of the primary, secondary, tertiary and quaternary loops and the preheating of the turbine condensate are all utilized.

[0269] 4. Performance comparison between the nuclear waste treatment technology of this invention and traditional nuclear waste treatment technologies

[0270] Table 11 Performance comparison between the nuclear waste treatment technology of the present invention and traditional nuclear waste treatment technologies

[0271]

[0272]

[0273] Note: The objective of this invention is a volume reduction rate ≥ 100 times;

[0274] This invention features an innovative separation technology that differs from existing technologies.

[0275] Table 12 Comparison of the Innovative Separation Technology of this Invention with Existing Technologies

[0276]

[0277]

[0278] 6. Innovations and advantages of this invention in terms of technical modules

[0279] Table 13 Innovative Design of the Technical Modules in this Invention

[0280]

[0281] Note: Specific separation and extraction processes and methods. This nuclide centrifugal separation system is designed using "in-situ nuclide solidification technology" to "lock" the most dangerous highly radioactive nuclides into a stable mineral lattice, providing a key guarantee for the sustainable development of nuclear energy.

[0282] 7. Technical and economic comparison of the present invention in multiple scenarios

[0283] Table 14 Comparison of Technological and Economic Efficiency in Multiple Scenarios

[0284]

[0285]

[0286] Note: The technology used in the medical application scenario complies with the "Regulations for the Management of Medical Radioactive Waste" (GBZ133-2023); the energy data in the table above is based on simulation calculations using Aspen Plus V12, with the following boundary conditions: ambient temperature 25℃, atmospheric pressure 101.3kPa, and cooling water ΔT=10℃.

[0287] Distributed processing: Small SCWG-PAFC integrated modules (containerized) are suitable for rapid on-site scaling in hospitals;

[0288] Centralized processing: Regional processing centers are built using waste heat from nuclear power plants to achieve radionuclide regeneration and energy self-sufficiency.

[0289] Economic breakthrough: Resource utilization and recycling of medical radionuclides 99m Tc is used in the production of medical isotopes, with a value of over 10,000 yuan per gram, significantly offsetting processing costs.

[0290] The payback period for modular integrated / connected PAFC systems is less than 5 years (calculated at an electricity price of 0.1 USD / kWh). Actual deployment complies with the "Regulations for the Management of Medical Radioactive Waste" (GBZ 133-2023).

[0291] This invention effectively inhibits dioxins from organohalides.

[0292] Example: Input: 50kg medical waste (containing I10Bq / g), output: 0.5kg solidified residue (volume reduction rate 100 times); 98g AgI recovered (purity 99.5%); 14kWh of electricity generated (PAFC efficiency 40%).

[0293] Emissions: Gaseous 85 Kr<0.01Bq / m 3 (Compliant with GB 6249-2011).

[0294] The device of this invention can help solve the problem of large-scale volume reduction in the disposal of high-temperature, high-pressure, large-capacity supercritical gasification-separation equipment in nuclear power plants, which is currently lacking. The localization rate of key materials in the device is >95%, and the cost of Ti / Haynes282 alloy replacing Inconel 718 is reduced by 32%. It can help solve various nuclear waste disposal problems encountered by current nuclear power plants. It can help solve the problems of limited adsorption of nuclides, salt deposition and blockage, and secondary pollution caused by traditional methods such as incineration in the current adsorption separation mode. It can help solve the problems of disposal of rare earth mine tailings and other smelting waste, medical radioactive waste disposal management, and nuclear pollution disposal problems in seawater desalination plants. It can help solve the problem of insufficient utilization of nuclear power heat energy and waste heat from nuclear waste disposal in current nuclear power plants. Using the device of this invention, it is possible to achieve waste heat integration and cascade zero-carbon power generation by combining nuclear energy, supercritical device heat exchange, and syngas and turbine condensate drainage. This avoids secondary nuclear contamination while reducing the power and water consumption of nuclear waste disposal. The nuclide recovery achieves a volume reduction rate of ≥100 times, significantly reducing the volume of deeply buried materials. With a nuclide recovery rate >99.9%, it can significantly improve the resource utilization of the nuclide industrial and medical chains, and also increase the revenue of nuclear power plants. Therefore, the comprehensive disposal and safe operation of high-level radioactive waste from nuclear power plants, the effective extraction and volume reduction of nuclides, and the disposal of nuclear waste in society, such as rare earth tailings treatment and nuclear-contaminated seawater treatment, require the high-level radioactive multi-state nuclear waste supercritical gasification separation device, treatment method, and waste heat integration method of this invention.

[0295] Please refer to Figure 1 As shown, the supercritical gasification separation device for high-level radioactive multi-state nuclear waste of the present invention includes:

[0296] The gasification separation device is 38 meters high, with an outer diameter of 5930 mm and an inner diameter of 4900 mm (tolerance ±5 mm). The inner wall has a wavy curved surface with a wavelength of 1000 mm (error ±10 mm) and a wave height of 30 mm (error ±0.5 mm). It has a five-layer composite wall structure, consisting of inner and outer shells, with a combined wall thickness of 515 mm.

[0297] The inner layer is a corrosion-resistant and pressure-bearing layer, 97mm thick, with a Ti / Haynes282 matrix. The inner side of the inner layer has a wavy curved surface with a peak spacing of 1000mm and a peak height of 30mm, which is used to improve heat transfer and radiation resistance. The heat transfer efficiency of the wavy coating is improved by 23% (cited in the Journal of Nuclear Materials 2024, 41(2):112-118). On the wavy curved surface is a DD coating "yttrium oxide stabilized zirconia thermal barrier coating" YSZ coating, 880μm thick. The coating is supplemented with MCrAIY / YSZ or NiCoCrAITaY / YSZ to form a composite coating or nano-coating, which can improve the bonding strength to 38.6MPa (ASTM C633 test report), and can enhance the radiation resistance and peeling resistance by more than 3 times. It can withstand 900℃ gas / salt spray corrosion for 1000 hours without peeling. The outer layer of the inner layer is an anti-radiation shielding layer, composed of 300mm thick lead and 50mm thick boron-impregnated polyethylene, used to shield neutrons / gamma rays. The outer layer of the shielding layer is a liquid metal cooling layer, made of Ti / Haynes282 substrate, 25mm thick, with 12 circumferentially milled grooves, each 10mm deep. The grooves are 700mm wide, spaced 600mm apart, and spiral upwards. Cadmium foil poison tape is embedded within the grooves, and a gallium-indium-tin eutectic alloy (GaIn) is injected into the grooves. 68.5 In 21.5 Sn 10 ).

[0298] The outer side of the liquid metal cooling layer is the waste heat layer FF, made of Inconel 690, with a tightly wound serpentine tube and an outer diameter of 17mm, used for three-stage heat recovery (80→200℃). The outer shell of the device is made of SA-533B carbon steel, 20mm thick. The device base is equipped with a hydraulic damping support to reduce seismic acceleration by 90%. The cavity 110 of the hollow layer is 5mm.

[0299] Furthermore, the cylindrical gasification separation device uses a dynamic temperature-controlled filter screen (350mm thick) at the bottom of the device as the height reference point Z=0. For example, in the nuclear-contaminated seawater area, Z=6.0m = the height of the medical area 5.65m + interval 0.35m;

[0300] A dynamic temperature-controlled filter (350mm thick, with a support frame at the bottom) is installed every 6 meters of height in the cylindrical temperature and pressure zone of the device. Based on the characteristics of nuclear waste, the gasification separation unit establishes various gradient temperature and pressure processing zones for disposal. The components required for processing the corresponding nuclear waste and the operating parameters of each processing zone are also specified.

[0301] Furthermore, the supercritical water nozzle group consists of 4 groups, 28 nozzles evenly distributed circumferentially per group; Group 1: height Z+h1, spraying upwards; Group 2: height Z+h2, spraying downwards (anti-clogging design); Group 3: height Z+h3, spraying upwards; Group 4: height Z+h4, spraying downwards; the nuclear waste material nozzle group consists of 2 groups, 28 nozzles evenly distributed circumferentially per group; Group 1: height Z+h5, spraying upwards; Group 2: height Z+h6, spraying upwards; the slag discharge valve is a 12×DN50 zirconium alloy ball valve (ASME B16.34 Class 2500), equipped with radiation dose interlock (response threshold: 100μSv / h), height Z+h7 (RF instrument interlock control); overpressure protection valves: 8, height Z+h8; gaseous product outlets: 28, height Z+h9.

[0302] Please refer to Figure 2 As shown, the present invention comprises a supercritical gasification separation device for high-level radioactive multi-state nuclear waste, a nuclear waste treatment method, and a waste heat integration method, including:

[0303] Corresponding to different types of nuclear waste, such as medical nuclear waste, nuclear-contaminated seawater, waste resin and waste protective clothing, refractory high-salt nuclear waste oil, rare earth symbiotic minerals and metal mixed solids, dioxins and coking products, the gasification separation device is equipped with supercritical water jet boxes and nuclear waste jet boxes for each "gradient temperature-pressure zone" to supply different materials, temperatures and pressures. The corresponding nuclear waste is gasified and separated in the corresponding "temperature-pressure zone" using the calorific value of supercritical water, or the nuclides are adsorbed under the action of adsorbents. The nuclides are flushed into the nuclide collection box through the slag discharge safety valve; or the syngas after gasification enters the external rotary centrifuge to separate the nuclides, which enter the nuclide collection box. The remaining syngas is combined with the thermal energy of the nuclear reactor after passing through a metal membrane and enters the fuel cell to generate electricity.

[0304] Furthermore, starting from the bottom of the inner shell of the supercritical gasification unit and moving upwards, six types of nuclear waste, including medical nuclear waste, dioxins, and coking waste, are supplied with supercritical water at a rate of 2,000 to 50,000 tons per hour using a gradient temperature-pressure coupling method. This water is then supplied to six gradient temperature-pressure zones (temperature range 380-1200℃, pressure range 25-37.5MPa) to gasify the corresponding nuclear waste (injected into the supercritical gasification unit).

[0305] The gradient temperature and pressure zones require the use of "multi-layer composite filters" to simultaneously meet the requirements of "nuclear adsorption, high temperature and high pressure resistance, and radiation corrosion resistance." Each temperature-pressure zone is 6 meters apart, and the six temperature-pressure zones are separated by "dynamic temperature-controlled filters."

[0306] Furthermore, the "dynamic temperature control filter" is 350mm thick and 4.9m in diameter. It is made of "multi-layer 316L stainless steel wire mesh vacuum sintered, pore size 0.1-0.5μm" + "yttrium oxide stabilized zirconia (YSZ) support layer (10μm) + ion-exchange mica (single layer 0.3nm)," forming a metal-ceramic intelligent composite structure. It undergoes a reinforced anti-corrosion process: a "Cr2O3 / CoWO4 ceramic coating" (5μm) is laser-clad, which can withstand oxidation at 1300℃ and HF / H2SO4 corrosion (ASTM authoritative testing standard; measurement method for "dynamic temperature control filter pore size 0.1-0.5μm" (ISO 15901-2:2022)). The lower part of the "dynamic temperature control filter" is supported by a structural bracket and evenly fixed to the inner layer of the gasification separation device.

[0307] Furthermore, each "gradient temperature-pressure zone" related to "medical nuclear waste, nuclear-contaminated seawater, waste resin and waste protective clothing, refractory high-salt nuclear waste oil, rare earth symbiotic minerals and mixed metal solids, dioxins and coking products" has corresponding nuclear waste disposal facilities. Each "corresponding temperature-pressure zone" has 28 separation safety valves (outlets) at the upper end and 12 separation safety valves (lower end). Depending on the disposal and separation process of different nuclear wastes, adsorbents are added to the "temperature-pressure" disposal zone, or equipment such as rotary centrifuge separation systems are connected to the rear end of the safety valves. After passing through the relevant adsorbents or rotary centrifuge separation systems, the safety valves (outlets) and (lower end) are connected to radionuclide collection tanks. The eight preventative safety valves (upper end) are connected to nuclear material jet boxes. All external connections use "corrugated pipe seals + lead shielding sleeves" tubing bundles.

[0308] This invention relates to a supercritical gasification separation device, treatment method, and waste heat integration method for high-level radioactive multi-state nuclear waste. Addressing the separation requirements of different sources of radionuclides in supercritical gasification (SCWG) systems, and considering the physicochemical properties of radionuclides and the limitations of existing technologies, it differentiates between "adsorbent injection within the gasification separation device" and "post-rotary centrifugal separation" processes, thereby upgrading the radionuclide separation scheme.

[0309] Furthermore, taking the "medical nuclear waste" disposal area as an example, the supercritical water temperature is 380-420℃, and the pressure is 25-28MPa. The supercritical water injection system is divided into 4 groups, evenly distributed on the circumference of the 18.6-meter-long cylinder above the bottom of the gasification separation device: The first group of water injection system is located 550mm above the bottom of the gasification separation device (inside), consisting of 28 high-pressure nozzles arranged along the circumference, with the nozzles pointing upwards; the second group of water injection system is located 2 meters above the bottom of the gasification separation device (inside), consisting of 28 nozzles arranged along the circumference, with the nozzles pointing downwards; the third group of water injection system is located 3 meters above the bottom of the gasification separation device (inside), consisting of 28 nozzles arranged along the circumference, with the nozzles pointing upwards; the fourth group of water injection system is located 5 meters above the bottom of the gasification separation device (inside), consisting of 28 nozzles arranged along the circumference, with the nozzles pointing downwards.

[0310] In the "medical nuclear waste" disposal area, the nuclear waste material injection system is divided into two groups, evenly distributed on the circumference of an 18.6-meter-long cylinder above the bottom of the gasification separation device. The first group of material injection systems has 28 high-pressure nozzles arranged along the circumference, 1.35 meters above the bottom of the gasification separation device (inside), with the nozzles pointing upwards. The second group of material injection systems has 28 high-pressure nozzles arranged along the circumference, 4 meters above the bottom of the gasification separation device (inside), with the nozzles pointing upwards.

[0311] In the "medical nuclear waste" disposal area, 400mm above the bottom of the gasification separation device, a group of 12 slag discharge safety valves (1 lower) are arranged along the circumference. Combined with the supercritical water from the second and fourth sets of water injection systems (2 meters and 5 meters above the bottom of the gasification separation device) with nozzles specifically designed for flushing, the supercritical water reacts with the material during gasification and flushes away any possible particle nuclide accumulation on the "dynamic temperature-controlled filter". This causes any possible particle accumulation to either re-particulate in the supercritical gasification reaction or enter the 12 slag discharge safety valves (1 lower), which are symmetrically arranged along the circumference at a height of 400mm above the bottom of the device and controlled (or timed) by a radio frequency instrument that measures the amount of particle accumulation. The accumulated particles pass through the slag discharge safety valves (1 lower) (dual valves in series) and are connected by a "corrugated pipe seal + lead shielding sleeve" tube bundle, entering the nuclide collection tank connected to the outside of the gasification separation device.

[0312] In the "medical nuclear waste" disposal area, 5.25m above the bottom of the gasification separation device (inside), there is a set of 8 protective safety valves (1 on top) along the circumference. This is the protection system for the gasification device, and the external nuclear waste jet box is connected by a "corrugated pipe seal + lead shielding sleeve" tube bundle.

[0313] The safety valve is controlled by the temperature and pressure gauges in the tube bundle. It opens when the temperature or pressure value in the interval exceeds the set value to prevent the gasification device from overloading, overheating, or overpressure. At the same time, the temperature and pressure gauges are linked to the material jet system and the supercritical water system to reduce or stop the injection of nuclear waste and supercritical water into the gasification device, thereby protecting the gasification device.

[0314] In the "medical nuclear waste" disposal area, 5.5m above the bottom of the gasification separation device (inside), a group of 28 safety valves (outlets 1) are set along the circumference, connected to the external nuclide separation-collection tank by a "corrugated pipe seal + lead shielding sleeve" tube bundle.

[0315] Example of a "medical nuclear waste" disposal area: For example, input: 50kg medical waste (including...) 131 If I 10 Bq / g), the output will be: 0.5 kg of solidified residue (volume reduction rate of 100 times), 98 g of AgI recovered (purity of 99.5%), PAFC efficiency of 40%, and gaseous emission. 85 Kr<0.01Bq / m 3 (Complies with GB 6249-2011); The disposal technology for "medical nuclear waste" complies with the "Regulations for the Management of Medical Radioactive Waste" (GBZ133-2023).

[0316] The present invention relates to a supercritical gasification separation device, treatment method, and waste heat integration method for high-level radioactive multistate nuclear waste, the steps of which are as follows:

[0317] The matching and innovation steps of supercritical water supply parameters with nuclear waste pretreatment (refined stage) are shown in the table below.

[0318] Table 15. Innovative Matching of Supercritical Water Supply Parameters with Nuclear Waste Refining and Pretreatment Parameters

[0319]

[0320]

[0321] Note: Dioxins decompose at temperatures above 850°C with a rate >99.99%, provided the following conditions are met: temperature ≥850°C, residence time ≥2s, and oxygen-to-carbon ratio ≥1.5 (refer to EPA Method 23); they decompose almost completely at temperatures above 1070°C.

[0322] Uniform constraints: All nuclear waste pretreatment temperatures must be ≤300℃ and pressures ≤20MPa (lower pressures are required for medical / seawater waste) to prevent radionuclide volatilization and premature decomposition of organic matter.

[0323] Supercritical gasification-separation steps: In each gradient temperature and pressure zone, refined hydrous nuclear materials (e.g., high-salt refractory waste oil, organic resin protective clothing, nuclear-contaminated seawater, medical nuclear waste, mixed solids) at a set temperature and pressure in the specific nuclear material jet tank, supercritical water at a set temperature and pressure in the supercritical water jet tank, and the required adsorbent are injected together into the inner shell of the supercritical gasification unit. The specific hydrous nuclear materials and supercritical water undergo a supercritical gasification reaction in the zone, generating high-temperature gas. The required adsorbent combines with the nuclides in the nuclear materials and is adsorbed by the dynamic temperature-controlled filter in the zone. Under the flushing of the set downward supercritical water, the gas passes through the slag discharge safety valve next to the dynamic temperature-controlled filter in the zone and enters the external nuclide collection tank. The remaining high-temperature gas passes through the outlet-metal membrane above the zone. Depending on customer needs, the syngas can be used to extract pure hydrogen and carbon dioxide, or it can all enter an external fuel cell for power generation, after which carbon dioxide and other inorganic substances are separated.

[0324] Specific refined hydrous nuclide materials (e.g., dioxin coking compounds, rare earth symbiotic mineral mixtures) requiring separation by a rotary centrifuge and set at a predetermined temperature and pressure undergo supercritical gasification reactions with supercritical water injected into the designated zone. This produces high-temperature gas, which is then connected to a rotary centrifuge at the upper outlet of the zone to separate nuclides. The resulting nuclides enter the collection tank. The subsequent syngas passes through a metal membrane; the syngas can be used to extract pure hydrogen and carbon dioxide, or it can be entirely fed into an external fuel cell for power generation, after which carbon dioxide and other inorganic substances are separated. Any remaining nuclides or particles in the zone are flushed by downward-flowing supercritical water and then discharged through a slag discharge safety valve next to the dynamic temperature-controlled filter in the zone, entering the external nuclide collection tank.

[0325] Resource integration steps: After gasification and separation of specific refined water-containing nuclide materials within various temperature and pressure gradient ranges, the syngas passes through a metal membrane. Before entering the fuel cell, it is combined and integrated with the set nuclear reactor loop thermal energy to improve the fuel cell's input thermal energy and electrical energy output. According to the characteristics of the nuclear reactor loop thermal energy, the primary loop thermal energy is used to preheat the specific refined water-containing nuclide materials, and the secondary loop thermal energy is used to combine with the syngas that has been separated from nuclides in the high-temperature and high-pressure range (e.g., high-salt refractory waste oil, dioxin coking products, rare earth symbiotic mineral mixtures) before entering the MC. FC power generation; the third loop thermal energy is used to heat the seawater source heat pump and park heating; the fourth loop thermal energy, combined with the synthesis gas from the high-temperature and high-pressure range (e.g., organic resin protective clothing, nuclear-contaminated seawater, medical nuclear waste), is used to preheat the SOFC in the gradient temperature and pressure range; if the nuclear reactor system does not have a specific fourth loop, the steam turbine condensate is used to preheat the SOFC in the gradient temperature and pressure range; the hot water generated by the serpentine winding heat exchange tubes on the outer side of the inner shell of the supercritical gasification separation device (combined with the steam turbine condensate of the nuclear power plant) is used to preheat supercritical water.

[0326] Furthermore, the syngas at 500°C from the outlet 6 of the "dioxin coking" treatment "temperature-pressure zone" is combined with the 250-300°C from the secondary loop of the nuclear reactor and used for integrated / connected MCFC power generation; the residual heat at 200°C is then used for integrated / connected medium-temperature FC power generation.

[0327] Furthermore, the synthesis gas at 400°C from outlet 5 of the "temperature-pressure zone" of the "rare earth symbiotic mineral mixed solids" treatment is combined with the 250-300°C from the secondary loop of the nuclear reactor for integrated / connected MCFC power generation; the remaining low temperature is used for adsorption refrigeration.

[0328] Furthermore, the synthesis gas at outlet 4 of the "temperature-pressure zone" for the treatment of "high-salt refractory materials and waste oil" is 300-400°C, which is combined with the 250-300°C of the secondary loop of the nuclear reactor for integrated / connected MCFC power generation; the remaining preheated gas is used for the inlet.

[0329] Furthermore, the "organic resin protective clothing" handles the synthesis gas at 300-350°C in outlet 3 of the "temperature-pressure zone" and combines it with the 60-350°C of the fourth loop of the nuclear reactor for integration / connection of ORC power generation, integration / connection of PEMFC cooling, and residual heat for preheating SOFC.

[0330] Furthermore, the syngas from outlet 2 of the "temperature-pressure zone" of the "nuclear contaminated seawater" treatment is combined with the 60-350℃ of the nuclear reactor's fourth loop and used for integrated / connected concentrated brine salinity gradient power generation, which in turn powers SOFC to produce hydrogen. The steam after the reaction at 120℃ is used for integrated / connected FC (or FC+ORC) power generation.

[0331] Furthermore, the syngas from outlet 1 of the "temperature-pressure zone" of the "medical nuclear waste" treatment area is combined with the 60-350℃ of the fourth loop of the nuclear reactor at 300℃ to preheat the SOFC air cathode, and the residual temperature of 60℃ is used for domestic water.

[0332] Furthermore, in addition to the integrated power generation system that utilizes waste heat from the energy nuclide chain PuO2 → MOX nuclear fuel·Am2O3 → space nuclear battery, other resource utilization pathways for nuclear waste disposal include: "medical nuclide chain: 131 → thyroid cancer treatment agent °Sr → Y generator → yttrium [°Y] microspheres (liver cancer radiotherapy), 13·223Ra → targeted drug for bone metastases; and industrial nuclide chain: ·137℃s → industrial irradiation source·rare earth oxides → permanent magnet materials / catalysts," achieving resource integration from "energy to industry to medicine." This resource integration can significantly reduce the electricity and water consumption of nuclear waste disposal, greatly conserving resources.

[0333] Furthermore, the "medical nuclear waste" disposal area handles inputs of 50 kg of medical waste (including... 131If I = 10 Bq / g, the output will be: 0.5 kg of solidified residue (100-fold volume reduction), 98 g of AgI recovered (99.5% purity), 14 kWh of power generation (PAFC efficiency 40%), and gaseous emissions. 85 Kr<0.01Bq / m 3 (Compliant with GB 6249-2011);

[0334] Furthermore, the disposal technology for "medical nuclear waste" complies with the "Regulations for the Management of Medical Radioactive Waste" (GBZ133-2023).

[0335] The above description is only a preferred embodiment of the present invention, but the scope of protection of the present invention is not limited thereto. Any equivalent substitutions or modifications made by those skilled in the art within the scope of the technology disclosed in the present invention, based on the technical solution and inventive concept of the present invention, should be covered within the scope of protection of the present invention.

Claims

1. A supercritical gasification separation device for high-level radioactive multi-state nuclear waste, comprising a hollow cylindrical body, characterized in that, Along the radial direction of the hollow cylinder from the inside to the outside, the supercritical gasification separation device for high-level radioactive multistate nuclear waste sequentially comprises: The anti-corrosion pressure-bearing layer includes an alloy substrate and a wavy yttrium-stabilized zirconia coating disposed inside the alloy substrate. The alloy substrate is Ti / Haynes282 with a thickness of 90-100mm. The wavy yttrium-stabilized zirconia coating has a peak spacing of 1000±50mm and a peak height of 30±2mm. The radiation shielding layer includes a boron polyethylene layer and a lead layer that are closely attached to the outer wall of the anti-corrosion and pressure-bearing layer. The thickness of the boron polyethylene layer is 50±5mm, and the thickness of the lead layer is 300±10mm. The liquid metal cooling layer is made of 316H stainless steel with a thickness of 25±2mm. The liquid metal cooling layer has spiral channels, which are filled with gallium indium tin eutectic alloy Ga. 68.5 In 21.5 S 10 And embedded with cadmium foil poison tape; The waste heat recovery layer consists of Inconel 690 serpentine wound heat exchange tubes with an outer diameter of 17±1mm, and is connected to a three-stage heat recovery system; and The seismic outer shell layer consists of a 20±2mm thick SA-533B carbon steel shell and hydraulic damping bearings with a damping coefficient ξ≥0.9; The system comprises six independent thermo-pressure reaction zones arranged along the axial direction of the hollow cylinder, with a temperature range of 380-1200℃ and a pressure range of 25-37.5MPa. Adjacent thermo-pressure reaction zones are separated by dynamic temperature-controlled filters. Each thermo-pressure reaction zone is equipped with four sets of supercritical water nozzles, including two sets of downward spray anti-clogging nozzles and two sets of material nozzles. Each thermo-pressure reaction zone is equipped with a slag discharge valve and an overpressure protection valve. The slag discharge valve is interlocked with the radio frequency deposition monitoring instrument, and the overpressure protection valve is linked with the temperature and pressure sensor. The six independent thermo-pressure reaction zones are used to simultaneously process at least two different physical states of radioactive waste from gaseous, liquid, and solid radioactive waste.

2. The supercritical gasification separation device for high-level radioactive multi-state nuclear waste according to claim 1, characterized in that: The dynamic temperature-controlled filter screen includes a sintered 316L stainless steel mesh with a pore size of 0.1-0.5μm, a yttrium oxide-stabilized zirconium oxide support layer with a thickness of 10±1μm, and a Cr2O3 / CoWO4 anti-corrosion coating.

3. The supercritical gasification separation device for high-level radioactive multi-state nuclear waste according to claim 2, characterized in that: The Cr2O3 / CoWO4 anti-corrosion coating of the dynamic temperature-controlled filter has a thickness of 5±0.5μm and is resistant to oxidation corrosion at 1300℃ and erosion by 5wt% H2SO4 solution.

4. The supercritical gasification separation device for high-level radioactive multi-state nuclear waste according to claim 1, characterized in that: The spiral channels of the liquid metal cooling layer are distributed circumferentially at intervals of 600±50mm, with a channel depth of 10±0.5mm and a channel width of 700±20mm.

5. The supercritical gasification separation device for high-level radioactive multi-state nuclear waste according to claim 1, characterized in that: The hydraulic damping bearing meets the IEEE 693-2018 seismic standard and can reduce seismic acceleration by 90±2%.

6. The supercritical gasification separation device for high-level radioactive multi-state nuclear waste according to claim 1, characterized in that: The gallium indium tin eutectic alloy contains 68.5±0.5wt% Ga, 21.5±0.5wt% In and 10.0±0.5wt% Sn.

7. A method for treating nuclear waste based on a supercritical gasification separation device for high-level radioactive multi-state nuclear waste according to any one of claims 1 to 6, characterized in that, include: For the ionic nuclide Cs + 、Sr 2+ I - In-situ adsorption was achieved by injecting nano-zero-valent iron-modified zeolite into a supercritical gasification separation device for high-level radioactive multistate nuclear waste. For high-density particulate nuclides PuO2 and UO2, zirconium phosphate magnetic beads are used for pre-adsorption in the supercritical gasification separation device for high-level radioactive multistate nuclear waste, and centrifuge separation is used. For gaseous nuclides 85 Kr、 3 H, after cryogenic distillation at -196±5℃, uses a specific surface area ≥2400m² 2 / g of MOF-808 molecular sieve trapping.

8. The nuclear waste treatment method according to claim 7, characterized in that: The centrifuge is a titanium-zirconium alloy centrifuge with a rotation speed of 20,000±500 rpm. The zirconium content of the titanium-zirconium alloy centrifuge is ≥80wt%, the centrifugal acceleration is >10,000g, and the density separation threshold is >4.0g / cm³. 3 .

9. A waste heat integration method for a supercritical gasification separation device for high-level radioactive multi-state nuclear waste based on any one of claims 1 to 6, characterized in that, include: The waste heat from the primary loop of the nuclear power plant is used to preheat the nuclear waste slurry entering the supercritical gasification and separation unit for high-level radioactive multi-state nuclear waste. as well as The secondary loop steam of the nuclear power plant and the supercritical gasification syngas at 300-500℃ are used together to drive the molten carbonate fuel cell.

10. The waste heat integration method according to claim 9, characterized in that: The steam pressure of the secondary loop of the nuclear power plant is ≥4 MPa, the power generation efficiency of the molten carbonate fuel cell is >50%, the total energy utilization rate of the system is >70%, and the radionuclide rejection rate is >99.9%.

11. The waste heat integration method according to claim 9, characterized in that, Also includes: The steam turbine condensate is converted into low-pressure steam via vacuum flash evaporation to drive an absorption heat pump. The decay heat is recovered through a gallium indium tin alloy heat pipe with a thermal conductivity >40W / m·K.

12. A method for waste heat cascade power generation based on a supercritical gasification separation device for high-level radioactive multi-state nuclear waste according to any one of claims 1 to 6, characterized in that, include: Level 1: Waste slurry from the primary loop of a nuclear power plant; Level 2: Steam from the secondary loop of the nuclear power plant drives MCFC (Multi-Cyclic Fuel Cell) power generation; Level 3: Supercritical syngas-driven SOFC power generation; and / or Level 4: The fourth loop of the nuclear power plant is used for preheating SOFC.

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