19 results about "Fast-neutron reactor" patented technology

The invention belongs to the technical field of research and development of fast neutron reactor jackets and structural materials, and particularly relates to a high-temperature-lead-bismuth-corrosion-resistant high-silicon-content ferrite / martensite steel ladle tube and a preparation method thereof. According to the mass percentage, the composition mainly comprises the following components: 0.16-0.24% of C; 0.4 to 0.8 percent of Mn; cr: 11.6% to 12.4%; 1.2 to 1.8 percent of W; 0.1 to 0.4 percent of Ta; 0.2 to 0.4 percent of V; 1.0%-1.8% of Si and Fe. The preparation method comprises the steps of vacuum induction and vacuum arc remelting. Carrying out homogenization treatment and forging; extruding and annealing; cold rolling and annealing; and performing normalizing and tempering treatment. Through an efficient short-flow process route and reasonable smelting, pipe machining deformation and heat treatment processes, the microscopic structure of the material is remarkably improved, then the mechanical property and high-temperature lead and bismuth corrosion resistance of the finished pipe are improved, and the use requirement of the lead and bismuth cooling fast reactor fuel assembly cladding material under the high-temperature working condition is met.

The invention discloses a preparation method of a high-hydrophobicity liquid lead bismuth alloy nitride coating capable of increasing a high-temperature contact angle, which comprises the following steps of: pretreating a substrate, depositing a FeCrAl coating on the surface of the substrate by adopting magnetron sputtering, and nitriding the FeCrAl coating to obtain a nitride coating; the invention further discloses the high-hydrophobicity liquid lead bismuth alloy nitride coating for improving the high-temperature contact angle. According to the preparation method, the FeCrAl coating is nitrided, so that the surface strength and corrosion resistance of the FeCrAl coating are improved, the formed FeCrAl nitride layer with a compact structure has high lyophobicity on the liquid lead bismuth alloy, the material surface is in a non-wetting state all the time, the liquid lead bismuth alloy can be effectively prevented from permeating into an alloy matrix, and the service life of the material is prolonged. The material is suitable for being used as a cladding material of a lead-cooled fast neutron reactor.

The invention relates to the technical field of nuclear reactor design, in particular to a safety assembly, a lead-cooled fast neutron reactor and a passive negative reactivity feedback method. The safety assembly comprises a first positioning piece, a second positioning piece and a containing piece. The first positioning piece and the second positioning piece are respectively connected with the accommodating piece so as to position the two ends of the accommodating piece in the linear direction, the accommodating piece is provided with an accommodating cavity for accommodating a coolant, and one end of the accommodating cavity in the length direction is communicated with the outside of the accommodating piece. The safety assembly is applied to the lead-cooled fast neutron reactor to serve as a safety barrier on the periphery of the reactor core, passive negative reactivity can be rapidly introduced, the structure is simple, implementation is convenient and fast, the nuclear reactor design specification and related technical requirements can be better met, and safety and reliability are further improved.

A liquid metal cooled nuclear reactor includes a passive decay heat removal system having thermal insulation attached to a wall of a cold source reservoir that holds a phase change material, where the insulation is arranged to automatically fall by gravity from the wall in response to the wall reaching a predetermined temperature. The nuclear reactor may be a fast neutron reactor that incorporates an integral system having a final cold source with a reservoir incorporating an integral exchanger divided into a plurality of parallel tubes between which a phase change material is inserted, the reservoir being surrounded by a thermal insulating layer that can be detached in a passive manner in the event of reaching a predetermined threshold temperature.

Embodiments of the present application relate to the technical field of nuclear reactor design, and more specifically, to a method for determining control rod drop time, which is applied to a fast neutron reactor and includes the following steps: obtaining a shutdown signal from the fast neutron reactor and determining the time corresponding to the shutdown signal as a first time; obtaining a drive signal from a drive device, and determining the time corresponding to the drive signal as a second time when the drive signal indicates that a drive parameter of the drive device is lower than a preset parameter threshold; obtaining a rod position measurement signal from a control rod, and determining the time corresponding to the rod position measurement signal as a third time when the rod position measurement signal indicates that the control rod has reached a rod position corresponding to a shutdown depth requirement; and determining the control rod drop time based on the first time, the second time, and the third time. The method provided in the embodiments of the present application determines the control rod drop time based on the obtained first time, the second time, and the third time, without requiring the participation of an operator or operating personnel, and has high measurement efficiency.

Fuel, intended to be irradiated in a fast neutron nuclear reactor, comprising grains of at least one metal selected from U, Pu or Th, the grains being compressed in a volume, the fuel being porous, the fuel being characterized in that it comprises open intergranular pores, forming an interconnected porosity, the volume fraction of porosity preferably being between 3% and 30%, more than 50% or 80% of the porosity being an open intergranular porosity,

The invention belongs to the technical field of fourth-generation advanced fast neutron reactors, and particularly relates to an activity monitoring and coating testing method and device in a Po-210 aerosol migration loop. The device comprises an aerosol migration loop and a Po-210 in-situ monitoring device, and the aerosol migration loop is used for carrying out in-situ monitoring on mCi-magnitude Po-210 released at high temperature in a lead-bismuth alloy sample; the gaseous Po-210 moving in the sealed environment is in contact with the surfaces of four functional coating materials of antifouling, protection, sealing and decontamination and generates electrostatic adsorption, chemical adsorption or particle elastic collision effects, and the Po-210 in-situ monitoring device is used for monitoring the activity concentration level of Po-210 aerosol circularly moving in an aerosol migration loop in the whole test process. According to the invention, the adsorption performance test of the Po-210 aerosol on the surfaces of special coatings with four functions of antifouling, decontamination, sealing and protection and the like and transported can be safely and efficiently completed, and the activity level of the Po-210 aerosol transported in the gas space of a sealing system can be accurately monitored.

A compact, liquid-metal cooled, loop-type, fast neutron reactor. The invention relates to a primary / secondary heat exchanger with identical loops, each consisting of a heat transfer module arranged axisymmetrically around the reactor block. Each module comprises an expansion tank (20), a heat exchanger (21), and an electromagnetic pump (22), which alone forms a heat exchange loop between the liquid sodium primary circuit and the gaseous secondary circuit, such as nitrogen, which is a closed Brayton cycle. Figure for the abstract: Fig. 11

A nuclear power plant with a nuclear building housing at least one molten salt fast neutron reactor and a fission gas management system for the accumulated gases in the reactor pile headspace. The invention relates to a nuclear power plant comprising: a reactor building (10); at least one reactor vessel shaft (9), located inside the reactor building; at least one reactor vessel (2) (1) filled with a molten salt fuel bath and defining within it a gas space above the free level of the bath, called the pile headspace; another vessel (8), arranged around the reactor vessel and inside the reactor vessel shaft, defining an inter-vessel space (E) with the reactor vessel; at least one fission gas management system (gaseous PF) for the gases accumulated in the reactor vessel pile headspace during normal reactor operation. Figure for the abstract: Fig. 2

The invention relates to a fast neutron reactor core with optimized irradiation. The fast neutron reactor core comprises a reflecting layer with a hollow cylinder structure and an active region positioned in the reflecting layer, the active area comprises a coolant located outside, a central irradiation pore channel located in the center, peripheral irradiation pore channels symmetrically arranged in pairs relative to the center, an adjusting rod assembly, a safety rod assembly and a plurality of fuel assemblies arranged in the coolant, and the peripheral irradiation pore channels, the adjusting rod assembly and the safety rod assembly are located on the side length of a hexagon in the center. Under the condition that the thermal power is less than or equal to 210 MW, the maximum neutron flux in the reactor core is greater than or equal to 1.30 * 10 < 16 > cm <-2 > s <-1 >, the average neutron flux of the central irradiation pore channel is greater than or equal to 1.20 * 10 < 16 > cm <-2 > s <-1 >, the volume of the central irradiation pore channel is greater than or equal to 10000 cm < 3 >, and the fuel circulation period is greater than or equal to 100 equivalent full power days. And the neutron flux level and the irradiation capability of the reactor are obviously improved. Meanwhile, reactor core thermal hydraulic analysis is carried out based on the reactor, the highest temperature of a fuel pellet is smaller than or equal to 536 DEG C, the highest temperature of a cladding is smaller than or equal to 505 DEG C, and the highest temperature of a coolant is smaller than or equal to 456 DEG C. All thermal parameters of the hottest assembly meet thermal design criteria.

This invention relates to the field of nuclear reactors, specifically to a lead-based fast neutron reactor exhaust device and a fission gas collection device. The exhaust device includes fuel rods and a one-way exhaust device. The fuel rods include nuclear fuel pellets and cladding tubes. The one-way exhaust device connects the gas chamber to the outside of the cladding tubes in one direction. A metal coolant is provided inside the pressure vessel to submerge the fuel rods and the one-way exhaust device. In use, the fission gas generated by the nuclear fuel pellets passes through the one-way exhaust device and enters the metal coolant in one direction. The fission gas collection device includes a lead-based fast neutron reactor exhaust device, a temporary storage tank, a multi-stage purification device, and a circulation pump. This invention achieves the purpose of one-way exhaust by connecting a one-way exhaust device to the cladding tube of the fuel rods, so that the fission gas inside the fuel rods is discharged into the primary coolant when the pressure exceeds a certain level. The fission gas collection device collects gaseous fission products in the upper chamber of the reactor pressure vessel.

The naturally controlled temperature low-pressure fast neutron reactor maintains reactivity by evaporating and condensing nuclear fuel. The reactor evaporates a portion of the fissionable nuclear fuel when temperature exceeds the boiling temperature of the molten salt fuel. By the evaporation, the reactor controls the core reactivity chain reaction, preventing the temperature from exceeding the operation temperature. The evaporated fuel gas phase is condensed externally or internally. The reactor includes breedable materials like U238 and Thorium to generate additional fissionable fuel. Because of its inherent safety and simplicity, the reactor can be modernized into a small modular reactor and operate at a designed temperature controlled by the nuclear fuel chemistry. The reactor reflector can include ceramic solid particles of used nuclear fuel, U238, and Thorium.

The invention discloses a uranium nitride-silicon carbide composite pellet and a preparation method thereof, the composite pellet comprises a dispersed phase and a continuous phase, and the dispersed phase comprises uranium nitride; the continuous phase comprises silicon carbide, and the continuous phase is dispersed in the dispersed phase. The composite pellet disclosed by the invention has relatively high density and uranium density, relatively low fission product release rate, relatively good water and oxygen corrosion resistance at high temperature and relatively high safety and economical efficiency, the preparation method is simple, and the composite pellet has a wide application prospect in the field of nuclear fuel and can be applied to various advanced pressurized water reactors, fast neutron reactors and space reactors.

The invention discloses a uranium nitride-silicon carbide-carbon composite pellet and a preparation method thereof. The composite pellet comprises a dispersed phase, and the dispersed phase comprises uranium nitride; the continuous phase comprises silicon carbide and a carbon elementary substance, and the continuous phase is dispersed in the dispersed phase. The composite pellet has high density, heat conductivity and uranium density, low fission product release rate and high safety and economical efficiency, the preparation method is simple, and the composite pellet has wide application prospects in the field of nuclear fuel. The reactor can be applied to various advanced pressurized water reactors, fast neutron reactors and space reactors.

A nuclear power plant with a nuclear building housing at least one molten salt fast neutron reactor and a fission gas management system for the accumulated gases in the reactor pile headspace. The invention relates to a nuclear power plant comprising: a reactor building (10); at least one reactor vessel shaft (9), located inside the reactor building; at least one reactor vessel (2) (1) filled with a molten salt fuel bath and defining within it a gas space above the free level of the bath, called the pile headspace; another vessel (8), arranged around the reactor vessel and inside the reactor vessel shaft, defining an inter-vessel space (E) with the reactor vessel; at least one fission gas management system (gaseous PF) for the gases accumulated in the reactor vessel pile headspace during normal reactor operation. Figure for the abstract: Fig. 2

This invention belongs to the field of nuclear reactor core design, specifically relating to a method for constructing an equivalent model of the non-fuel region of a fast neutron reactor with complex geometry. First, the core of the three-dimensional fast neutron reactor to be simulated is geometrically divided into several layers along the axial direction. For each layer, a fine-group embedding self-screen resonance model is established. Then, the material of all fuel regions in the core is homogenized and used as average fuel. This average fuel is constructed outside the cylindrical model of the non-fuel region, ensuring that the non-fuel nuclides calculated by the cylindrical model are consistent. n The escape cross section is conserved with the escape cross section calculated by resonance, and an equivalent model of the outermost fuel radius is iteratively searched. Its advantages are: this method can quickly and online obtain the equivalent model of each non-fuel zone in the reactor core; it offers excellent geometric flexibility for complex reactor cores, fully considers environmental effects, and has higher computational accuracy; it can be used for rapid simulation calculations of numerical reactors and engineering applications.

The invention consists essentially of a fuel assembly (1) of hexagonal or parallelepiped cross-section, without a hexagonal duct, preferably intended for low-power-density sodium-cooled fast neutron reactors (SFR).

This invention relates to an integral fast neutron reactor including safety devices to mitigate core meltdown accidents. The main objective of this invention is to provide an integral fast neutron nuclear reactor comprising: a main tank including a core; an integrated main loop including a main pump, an intermediate heat exchanger, and a heat exchanger for discharging excess power; and a convex bastion dividing the volume of heat transfer fluid into a heat collector and a cold collector. The core includes safety devices consisting of at least one basic safety device specifically designed to mitigate core meltdown accidents. The at least one basic safety device includes a delivery pipe and a hollow through-tube positioned to pass through the core grid and a strong backplate to the bottom of the tank. The delivery pipe includes one or more injection nozzles and a reduced-thickness region to increase the puncture velocity in the event of a core meltdown accident. Equipment for recovering molten fuel is located at the bottom of the tank.

Provided is a light water reactor uranium fuel assembly capable of reducing the heat generation amount of both Am-241 and Cm-244, which are causes of glassification, by a light water reactor alone without using a fast neutron reactor, and capable of reducing the amount of glassification. A light water reactor uranium fuel assembly for a nuclear fuel cycle in which an americium isotope is extracted at the time of reprocessing of used fuel and the extracted americium isotope is added to the nuclear fuel cycle, the weight ratio W (wt% unit) of the added americium 241 with respect to the weight of the fuel heavy metal being W < -0.006e with respect to the average uranium 235 enrichment e (wt% unit) of the fuel assembly 2 +0.12e-0.43 (enrichment 5 wt% or more), W < -0.000356e+0.00357 (enrichment 4.2 wt% or more and less than 5.0 wt%).