426results about "Reactors manufacture" patented technology

The invention discloses a radiation-resistant high-entropy alloy magnetron sputtering coating and a preparation method thereof, and belongs to the technical field of radiation-resistant coating materials. On the basis of the multi-element synergistic effect and lamination interface design of the high-entropy alloy, FeCrNi and FeNiCrMn are used as target materials, and the coating with the nanometer lamination structure is prepared by accurately regulating and controlling the target current, the gas flow, the deposition time and other process parameters of magnetron sputtering; meanwhile, accurate control over the single-layer film thickness and the interface density is achieved by adjusting the coupling relation between the sputtering power and the deposition rate, and it is ensured that all unit layers are flat and compact, and interface combination is complete. And a multi-heterogeneous interface is formed through the design of a laminated interface, so that a defect capture network is formed, the hardness of a single-layer coating is improved, the hardening rate after irradiation can be lower than that of a traditional single-layer coating, and the problems of crystallization and brittleness failure induced by irradiation are effectively relieved.

The invention provides a reactor core structure capable of controlling an assembly clearance and a manufacturing method thereof. The reactor core structure comprises a reactor core heat management element and a filling material; wherein the base body is provided with a plurality of assembly holes; the reactor core heat management element comprises a fuel rod or a heat pipe; the reactor core heat management element is assembled in the assembly hole of the base body; the filling material is filled in an assembly gap between the reactor core heat management element and the assembly hole; and the filling material is a shape memory alloy. The reactor core structure can improve the safety margin of the heat pipe reactor.

The invention belongs to the technical field of processing of nuclear reactor control rod neutron absorber materials, and particularly relates to a preparation method of a nuclear-grade hafnium cast ingot. The method comprises the following steps: preparing sponge hafnium into an electrode block; performing vacuum electron beam melting on the electrode block for at least two times to obtain a finished cast ingot; and processing the finished cast ingot to prepare the nuclear-grade hafnium cast ingot which can be used for processing the control rod profile for the nuclear reactor assembly. According to the preparation method of the nuclear-grade hafnium cast ingot, the tensile strength, the welding performance and the creep performance of a nuclear-grade hafnium profile can be effectively improved; meanwhile, the technological process is simplified, and the technological economy and the production efficiency are remarkably improved; in addition, sponge hafnium is directly adopted for vacuum electron beam melting, so that the cost of raw materials is greatly reduced. According to the method, the application performance of the nuclear-grade hafnium profile is improved, meanwhile, the process is simplified, the production efficiency is improved, the production cost is reduced, and the economical efficiency of nuclear-grade hafnium profile manufacturing is greatly improved.

The present invention belongs to the technical field of ceramic preparation, and relates to a preparation method and device for a high-purity porous boron carbide nuclear control rod. The present invention uses boron trichloride gas as the boron source, methane as the carbon source, argon as the carrier gas and protective gas, and hydrogen as the catalytic gas and reducing gas to prepare a high-purity porous boron carbide nuclear control rod. The method of increasing the deposition temperature of boron carbide step by step and region by region from the back to the front is adopted, which can not only improve the deposition rate of boron carbide, but also ensure that the physical and mechanical properties of each part of the prepared high-purity porous boron carbide nuclear control rod are consistent. By changing the particle size of the boron carbide powder for preparing the boron carbide blank and coordinating the adjustment of temperature, flow rate, pressure and deposition time, the porosity and pore size of the prepared high-purity porous boron carbide can be regulated and a high total boron and carbon content can be obtained, so as to meet various actual requirements.

Systems, structures, devices, and fabrication processes for ceramic matrix composites suitable for use in a nuclear reactor environment and other applications requiring materials that can withstand high temperatures and / or highly corrosive environments are disclosed. In one aspect, a ceramic composite structure is provided. The structure comprises a chamber including an external shell and a hollow space inside the external shell. The external shell includes an inner composite layer including a first composite structure, a middle composite layer placed outside of the inner composite layer, the middle composite layer including a second composite structure that is different from the first composite structure, and an outer monolithic layer that has a spatially uniform material property and placed outside of the middle composite layer.

A covering for reinforcing a base layer of a nuclear fuel cladding is provided. The covering comprises a first layer configured to cover a first portion of the outer surface of the base layer of the nuclear fuel cladding, a second layer surrounding the first layer and the base layer and a third layer surrounding the second layer. The first layer comprises a fiber based material, the second layer comprises an interfacing material and the third layer comprises Chromium. A reinforced cladding for nuclear fuel and a method for producing a reinforced nuclear fuel cladding are also provided.

The invention discloses a zirconium alloy fuel rod end plug composite welding method and a zirconium alloy fuel rod, and belongs to the field of welding. The zirconium alloy fuel rod end plug composite welding method comprises the following steps that a zirconium alloy end plug and a zirconium alloy fuel rod cladding are spliced together and fixed, and an inert gas environment is provided after vacuumizing; firstly, a zirconium alloy end plug and a zirconium alloy fuel rod cladding are subjected to TIG welding, and meanwhile ultrasonic stirring and electromagnetic stirring are provided; then, TIG welding is conducted on an inflation hole of the zirconium alloy end plug, and ultrasonic stirring and electromagnetic stirring are provided at the same time; and after welding is completed, a welding area is subjected to polishing and quality detection. According to the method, air holes in the zirconium alloy welding area can be effectively eliminated, uniform distribution of structures and components is promoted, and the welding quality and the corrosion resistance of the welding area are improved.

The invention discloses a cold rolling die with a straight rib thin-wall cladding pipe, a production process and the cladding pipe, and relates to the technical field of reactor fuel cladding, the cold rolling die comprises a cold rolling outer die for controlling the forming of the outer contour of a pipeline and a cold rolling core head for controlling the inner diameter; the cold rolling outer die is a die with a hollow area inside, the hollow area comprises a pitch circle section and a fixed rib section, and the fixed rib section is provided with at least one section of rib groove; a pipeline forming die cavity is formed by hollow areas of the two cold rolling outer dies, and a roll gap is formed between the two cold rolling outer dies; when the cold rolling die is used for producing the cladding pipe, the requirement for room-temperature plasticity of cladding materials in the forming process is low, the quality of the surface of the cladding pipe cannot be affected in the forming process, the quality of a formed product is high, and the cold rolling die is simple in structure, low in manufacturing cost and suitable for popularization and application. The thin-wall cladding tube with the straight ribs produced by the cold rolling die is good in surface quality, and the ribs of the cladding tube are good in mechanical property.

A system is described that includes a sputter target and a magnetic element array including multiple sets of magnets arranged to have a Hall-Effect region that extends along a length of the sputter target. The elongated sputtering electrode material tube is interposed between the magnetic array and an object to be deposited with a sputtered material from the sputter target. During a direct current high-power impulse magnetron sputtering operation, the system performs a depositing on a surface of the object by generating and controlling an ion and neutral particle flux by: providing a vacuum apparatus containing a sputter target holder electrode; first generating a high-power pulsed plasma magnetron discharge with a high-current negative direct current (DC) pulse to the sputter a target holder electrode; and second generating a configurable positive voltage kick pulse to the sputter target holder electrode after terminating the negative DC pulse.

The invention relates to a nuclear fuel cladding produced with a substrate (14) which is made of pure zirconium or of a zirconium based alloy and a multilayer protective coating (16) which covers a surface (14B) of the substrate (14), the protective coating (16) comprising a main layer (18) made of pure chromium and one or more additional layers (20), each additional layer (20) being made of pure chromium or from a material made of chromium and, additionally, oxygen and / or nitrogen, with the possible presence of unavoidable impurities.

The invention provides a high-strength winding thin-wall SiC composite cladding tube and a preparation method thereof. A preparation method of a high-strength winding thin-wall SiC composite cladding tube comprises the steps that silicon carbide fibers are wound and laid around a core mold, the winding pre-tightening force is smaller than or equal to 10 N. The winding mode comprises the first mode that one layer of annular winding is conducted firstly, then one layer of spiral winding is conducted, and the number of fiber intersection points in the whole annular direction at a certain position of the outer surface of the core mold is 1. One-layer annular winding is carried out firstly, then two-layer spiral winding is carried out, the number of fiber intersection points in the whole annular direction at a certain position of the outer surface of the core mold is 3, and a prefabricated body is obtained; carrying out a densification deposition process on the preform, wherein the densification deposition process comprises PyC pyrolytic carbon interface deposition, CVI chemical vapor infiltration and CVD chemical vapor deposition; and precise inner circle and outer circle machining is conducted, and the SiC composite cladding pipe is obtained. According to the preparation method disclosed by the invention, through a specific dry winding mode and a densification deposition process, the SiC composite material tube blank is higher in density and better in mechanical property.

A deployment system for a molten salt reactor includes a reactor module, a cooling module, a coolant preparation module, a fuel shipping module, and a fuel preparation module. Each of the reactor module, the cooling module, the coolant preparation module, the fuel shipping module, and the fuel preparation module may be supplied to a deployment site. Using the coolant preparation module, a coolant salt held within the reactor module may be transferred to the cooling module. Using the fuel preparation module, a fuel salt held within the fuel shipping module may be transferred to the reactor module.

The invention belongs to the technical field of nuclear fuel element surface protection, and particularly relates to a multi-layer composite (Cr, Al) Nx / Cr protective coating for a nuclear reactor zirconium alloy cladding and a preparation method of the multi-layer composite (Cr, Al) Nx / Cr protective coating. The coating adopts a multi-layer structure design consisting of a buffer layer, a middle layer and an outermost layer in sequence from inside to outside, wherein the buffer layer is a Cr coating; the middle layer is a nitrogen-saturated (Cr, Al) Nx coating, x is larger than or equal to 1, the middle layer is of a columnar crystal structure, and the hardness of the middle layer is 5-10 GPa; the outermost layer is a non-stoichiometric ratio (Cr, Al) Nx coating, x is larger than or equal to 0.6 and smaller than or equal to 0.9, a nanocrystalline structure is shown, and the hardness is 12-25 GPa. Component regulation, hardness gradient and a microstructure are subjected to collaborative design, so that the coating has excellent wear resistance, high oxidation resistance and structural integrity at the same time under a harsh nuclear working condition, the safety margin of the zirconium alloy cladding under normal service and accident working conditions is greatly improved, and the service life of the zirconium alloy cladding under normal service and accident working conditions is greatly prolonged.

Insulated fuel assembly core with plurality of fuel monoliths, exhaust support plate, exhaust shield assembly, and insulation layer in which the plurality of fuel monoliths are located axially along a longitudinal axis and each of the plurality of fuel monoliths has a shape of an eccentric cylinder and a composition including a fissionable fuel component. Channels in the exhaust support plate are oriented so that propellant gas flow through the exhaust support plate does not impinge the exhaust shield assembly. The insulated fuel assembly core is manufactured by forming a tensioned fuel monolith stack mandrel assembly using mandrels spacers and internal tensioning components and forming an insulation layer on an outer surface of the tensioned fuel monolith stack mandrel assembly by mandrel winding. The insulated fuel assembly core can be incorporated into a fuel assembly of nuclear propulsion fission reactor structure of, for example, a nuclear thermal propulsion engine.

The invention provides a reactor core structure capable of improving an assembly clearance and a manufacturing method thereof. The reactor core structure comprises a reactor core heat management element and a filling material; wherein the base body is provided with a plurality of assembly holes; the reactor core heat management element comprises a fuel rod or a heat pipe; the reactor core heat management element is assembled in the assembly hole of the base body; the filling material is filled in an assembly gap between the reactor core heat management element and the assembly hole; the filling material is a foam porous material, and the pore diameter of the foam porous material is micron-sized. The reactor core structure can improve the safety margin of the heat pipe reactor.

There is provided a method of dismantling a dousing tank of a heavy water reactor structure. The method of dismantling a dousing tank of a heavy water reactor structure includes: selecting a dousing tank of a heavy water reactor structure for dismantlement; dismantling a 1st structure inside the dousing tank; fixing the dousing tank to a dome structure portion provided above the dousing tank through a mounting port; positioning a cutting unit on a 2nd structure including a concrete structure inside the dousing tank; and performing, by the cutting unit, cutting based on a forward or reverse rotation manner on an inner peripheral surface of the dousing tank.

The invention provides a method for preparing a uranium dioxide single crystal by using a molten salt method and the obtained uranium dioxide single crystal, aiming at solving the problem that the growth of different crystal faces cannot be controlled when the uranium dioxide single crystal is prepared by using an existing fluxing agent method. The method comprises the following steps: in the presence of a reducing agent and a surface active substance, heating U3O8 powder in molten salt to a heat preservation temperature, preserving heat, and then cooling and growing to obtain the uranium dioxide single crystal, wherein the surface active substance is SiO2 (silicon dioxide), GeO2 (germanium dioxide), V2O5 (vanadium pentoxide) or Na2WO4 (titanium dioxide). According to the method, U3O8 is taken as a raw material, a molten salt fluxing agent is combined, the reducing agent is used for reducing U3O8 and growing the UO2 crystal, and the preferred crystal face of UO2 in the nucleation and growth processes is effectively controlled by selecting a proper surface active substance and changing the surface energy of UO2 in the nucleation and growth processes, so that the UO2 single crystal with the preferred crystal face and an ordered structure is obtained.

The invention discloses a uranium nitride spherical particle preparation method, which comprises: mixing a carbon source and a solution to prepare gel particles; the carbon source comprises diamond powder; the gel particles are subjected to pretreatment and carbon thermal reduction nitridation treatment, and the uranium nitride spherical particles are obtained. According to the preparation method, the C / U molar ratio can be effectively controlled, and the prepared uranium nitride spherical particles have the advantages of being compact in structure, smooth in surface and high in purity.

The invention belongs to the technical field of nuclear fuel preparation, aims to solve the problems of non-uniform particle distribution and high breakage rate in traditional coated particle fuel preparation, and discloses a coated particle dispersion fuel pellet and a preparation method thereof.The method comprises the steps that firstly, silicon carbide slurry is prepared, a silicon carbide base band is prepared through tape casting equipment, and then the silicon carbide base band is prepared; after tape casting, the base band enters a drying area to be partially dried, TRISO particles are arranged on the surface of the base band at set intervals, the silicon carbide base band is freely curled and formed from one end, degreasing treatment and sintering densification are conducted, and the core block is prepared through the method. Through high-precision tape casting and three-dimensional curling technologies, uniform distribution of TRISO particles in CDF is achieved, and the contact and breakage rate of fuel in the preparation process is reduced.

Additive manufacturing compositions include low-absorbing particles or non-absorbing particles that have an absorbance for wavelengths of 300 nm to 700 nm that is equal to or greater than 0 Au and is less 1.0 Au, such as 0.001 Au ≤ absorbance ≤ 0.7 Au. Slurries including such particles and an uranium-containing particle and that are used in additive manufacturing processes have an increased penetration depth for curative radiation. Removal of low-absorbing particles or non-absorbing particles during post-processing of as-manufactured products results in pores that create porosity in the as-manufactured product that provide a volume accommodating fission gases and / or can enhance wicking of certain heat pipe coolant liquids. Low-absorbing particles or non-absorbing particles can be functionalized for improved properties, for example, with fissionable material for improved ceramic yields, with burnable poisons or stabilizers for increased homogeneity, with stabilizers for localized delivery of the stabilizer, or with combinations thereof.

The embodiment of the invention relates to the field of processes specially used for manufacturing a reactor or parts thereof, in particular to an assembling and charging testing method of a reactor using bulk nuclear fuel and the reactor comprising the bulk nuclear fuel. According to the method, loading test is carried out on the reactor by using the test facility after assembly, so that the backup reactivity of the reactor core can be determined, safe assembly and safe criticality of the reactor are realized, and the safety of experimenters is ensured; by gradually adding the partitioned side reflecting layers outside the reactor core formed by the bulk nuclear fuel, the reactor gradually reaches the critical value, the operation risk is reduced, and meanwhile, the critical point can be accurately obtained; after the critical loading of the reactor core is determined, the partitioned side reflecting layers are replaced by the complete side reflecting layers with the same value, so that the influence of gaps existing between the partitioned side reflecting layers can be avoided, and the reaction efficiency of nuclear fuel is improved. The reactor assembled according to the method has high neutron economy and low critical risk.

The present invention relates to a system (1) for extracting UF6 from a UF6 container placed in an autoclave (5), comprising a pump (24) having an inlet port (26) being connected to an upstream supply circuit (10, 20), the upstream supply circuit (10, 20) transporting UF6 from the autoclave to the pump (24); wherein the system further comprises a temperature- controlled subsystem (3) providing a heat supply to the pump (24) and at least a downstream portion of the upstream supply circuit (10, 20), the temperature-controlled subsystem (3) heat supply being adapted to maintain a temperature within the upstream supply circuit and the pump (24) during operation such that said temperature is above the crystallization temperature of UF6.

Described herein is a process for recovering uranium tetrafluoride (UF4) from hexafluoride uranium (UF6), the process comprising dissolving UF6 directly in an ionic liquid and recovering UF4, which can be processed to obtain UO2(s) or metallic uranium.

A method for depositing divalent metal compounds on the surface of a nuclear power plant component, the component being a nickel-based or austenitic stainless steel alloy includes: providing within the component an aqueous treatment solution containing at least one soluble metal-containing compound such as a zinc salt and at least one source of oxygen; allowing the treatment solution to remain in the component until the compound is deposited on the wetted surface of the component; and, removing the aqueous solution after exposure. The treatment may be applied more than once, using more than one divalent metal compound, and the surface may further be exposed to a solution containing a noble metal species and a reducing agent. The treatment temperature is preferably below 100°C.

A nuclear reactor is designed to couple the load path of the control elements with the reactor core, thus reducing the opportunity for differential movement between the control elements and the reactor core. A cartridge core barrel can be fabricated in a manufacturing facility to include the reactor core, control element supports, and control element drive system. The cartridge core barrel can be mounted to a reactor vessel head. Thus, any movement, such as through seismic forces, transmits an equal direction and magnitude to the control elements and the reactor core. This arrangement reduces the opportunity for differential movement between the control elements and the reactor core.

The present invention provides a method for manufacturing a multilayer nuclear fuel cladding tube, which includes the following steps: providing a preliminary cladding tube, in which an inner tube with a rod-shaped insert is inserted is arranged in an outer tube; reducing the diameter of the preliminary cladding tube by applying pressure from the outside to the inside of the preliminary cladding tube; and removing the insert from the inner tube by providing a force in the direction in which the insert extends, wherein the inner tube and the outer tube can be formed of metals different from each other.

The invention discloses a large boron carbide ceramic ring manufacturing method and belongs to the field of nuclear power. The method comprises the following steps: providing first powder, second powder and third powder of which the average particle sizes are sequentially increased as boron carbide raw material powder, uniformly mixing, and sintering for 1-5 hours at the temperature of 1800-2000 DEG C and the pressure intensity of 20-40 MPa to obtain an annular blank; and grinding and polishing the plane of the annular blank, and performing ultrasonic-assisted vibration grinding and polishing on the outer arc surface to obtain a finished product of the large boron carbide ceramic ring. According to the method, the production efficiency and quality of the large boron carbide ceramic ring can be improved, and the edge breakage or cracking risk is reduced.

In many embodiments, a plurality of neutron moderators may be arranged to define a plurality of flow channels within a reactor vessel, wherein each neutron moderator of the plurality of neutron moderators includes at least one surface face, and each neutron moderator is not machined. In some embodiments, the flow channel may be defined by the at least one surface face of each neutron moderator of the plurality of neutron moderators. In at least one embodiment, the plurality of neutron moderators may include at least three neutron moderators. In several embodiments, the plurality of neutron moderators may include a first pair of neutron moderators and a second pair of neutron moderators. In one or more embodiments, the first pair of neutron moderators and the second pair of neutron moderators may be arranged to form a lattice configuration in a reactor vessel.

The invention belongs to the nuclear industry and can be used for the production of fuel pellets from uranium-molybdenum metal powder enriched with uranium-235 to 7%, which can be used as fuel elements for nuclear reactors. The sintering of the pellets is carried out in an inert argon atmosphere at a temperature in the range of 1100°C to 1155°C, as the initial powder, a particle size of 160 μm is used, with a molybdenum content of 9.0 to 10.5 wt.%. The powder is preheated at a temperature of 500°C for 10-20 hours (in an argon medium), and then cold-pressed into pellets in a mold at a pressure of up to 950 MPa. According to another method, when preparing uranium-molybdenum pellets using a binder (plasticizer), the pellets are heated in argon from 300°C to 450°C for a duration of 2-4 hours before sintering, in order to remove the binder. The invention allows to increase the uranium content in the fuel, to reduce the heat accumulated in the core of the nuclear reactor, and to reduce the energy release in the event of a violation of the normal operating conditions of the nuclear reactor, which will improve its safety and emergency stability.