40 results about "Fission reactor" patented technology

The invention relates to a new technology for depositing a CrAlSiN gradient coating with high-temperature oxidation corrosion resistance and excellent mechanical properties on a supercritical water-cooled reactor fuel clad surface in a nuclear fission reactor. The CrAlSiN gradient coating is deposited on the surface of a base material by adopting a multi-target reaction magnetron sputtering method, and the technology comprises the following four continuous stages of: 1, preparing a Cr gradient coating; 2, preparing a CrAl gradient coating; 3, preparing a CrAlN gradient coating; and preparing the CrAlSiN gradient coating, and annealing. By reasonably designing the gradient variable microstructure of each element, the anti-oxidation temperature of the coating reaches 950 DEG C, the hardnessof the coating reaches over 37GPa, and the adhesive force of the coating adhered to the base material reaches 34N; the coating also has excellent thermal shock resistance, abrasion resistance and thelike; and the service property of the supercritical water-cooled reactor fuel clad part can be greatly improved, and the service life of the clad part can be greatly prolonged.

The invention provides a production method of a nuclear fuel clad element. The method comprises the following steps: selecting an MAX phase ceramic material, silicon carbide, an MAX phase-based composite ceramic material or a silicon carbide-based composite ceramic material, processing the above ceramic material to prepare a slurry, carrying out vacuum defoaming, making a ceramic film with the thickness of 10[mu]m-10mm on a base band through a curtain coating or draw-off process, winding to make a clad element blank, drying, carrying out rubber discharging, sintering, and carrying out surface treatment to obtain the nuclear fuel clad element. The production method has the advantages of simplicity, easy implementation, low cost, overcoming of the disadvantage of difficult processing of ceramic materials, high production efficiency, short cycle and easy industrialization. When the ceramic material is a Ti3SiC2-baed ceramic material, the ceramic material can resist molten fluorine salt corrosion, and can be used as a fluorine salt fuel clad element material in a nuclear reactor, so practical demands of a thorium-based fourth generation fission reactor nuclear energy system on the structure material are met.

The invention discloses a modeling system which is used in reactor design and used for reactor core component design. According to the modeling system, the geometric specificity of a fission reactor core is analyzed, a great amount of repeatedly constructed geometries are adopted, a CAD engineering model and a Monte Carlo calculation model of the fission reactor core are automatically established by establishing parameters, the Monte Carlo calculation model is used for radiating and transporting input of a Monte Carlo calculation program, and the significant physical quantity of the reactor core can be calculated after the Monte Carlo calculation model is obtained. By constantly evaluating the physical parameters of the calculation result, the modeling system can be used for performing repeated iteration correction on a design scheme of the fission reactor core till a reactor core scheme which makes a user satisfactory is made. The modeling system has the benefits that the tediousness that Monte Carlo calculation programs are written and files are input manually is avoided, the whole modeling process is very visible, the probability of mistakes is greatly reduced, and the design time of the reactor core is shortened.

The invention belongs to the field of cladding designs of a fusion and fission hybrid reactor, and particularly relates to a Th-U self-sustaining circulating full fused salt fuel hybrid reactor system and an operation method thereof. The Th-U self-sustaining circulating full fused salt fuel hybrid reactor system is characterized in that a fast fission breeder reactor provides an initial easily fission fuel required by starting of a thermal fission reactor, a cladding design of the thermal fission reactor utilizes an arrangement strategy of a seed-cladding to improve the entirety neutron economy of the system, and the purposes of a high energy enlargement factor of the system, tritium breeding and thorium-uranium self-sustaining circulating of the system are realized; <233>U is loaded in an energy generating region, so that the energy generating region has a good neutronics property and is mainly used for realizing the purposes of energy amplification of the system, neutron multiplication and most <233>U breeding of the system; superfluous neutrons enter a tritium producing region and are used for tritium breeding and part of <233>U breeding of the system. According to the hybrid reactor, due to self-sustaining circulating of thorium and uranium, a thorium fuel is converted into <233>U and is gradually burn up in an operation process, and the Th-U self-sustaining circulating full fused salt fuel hybrid reactor system can effectively and stably operate for a long time just by gradually adding the thorium fuel and removing the generated fission product.

The invention provides a heat exchange medium which comprises solid particles and fluid. The invention further provides a heat exchange system which comprises the heat exchange medium, a first heat exchanger, a mixing device, a separating device, a second heat exchanger and a first conveying device, wherein the mixing device is arranged on the upstream of the first heat exchanger and used for mixing the solid particles of the heat exchange medium and conveying the mixture to the first heat exchanger; the separating device is arranged at the downstream of the first heat exchanger and used for separating the solid particles and the fluid of the heat exchange medium discharged from the first heat exchanger; and the first conveying device is used for conveying the solid particles separated by using the separating device into the mixing device after passing through the second heat exchanger. Furthermore, the invention further provides a nuclear reactor system comprising the heat exchange system. The gas-solid or liquid-solid cooling medium provided by the invention has the advantages of high thermal capacity, low-pressure system, no corrosion, off-line processing and the like. The fission reactor provided by the invention can safely and reliably operate under the high power density or ultrahigh power density.

The invention discloses a mineral stantardless argon-argon dating method. The mineral stantardless argon-argon dating method comprises wrapping a mineral sample into a sample piece through aluminum foil, pasting high-purity nickel pieces on the front side and the rear side of the ample piece respectively and placing the sample piece into an accelerator neutron source for neutron irradiation; detecting the nickel pieces after the irradiation, determining a value of a neutron reaction section sigma and calculating the neutron flux N of the accelerator neutron source; loading the mineral sample after the irradiation to a rare gas measurement system, heating and melting the sample, determining the argon isotope content of the mineral sample through a rare gas mass spectrometer and obtaining the required formula through data calculation; substituting the neutron flux N of the accelerator neutron source, the reaction section sigma and the formula into the formula to calculate the argon-argon age of the mineral sample. According to the mineral stantardless argon-argon dating method, the problem that a J value which is related to the neutron flux of a 235 U fission reactor needs to be corrected through a known age standard sample in the traditional argon-argon dating is solved, the argon-argon dating can be implemented without any geological standard sample, and the accuracy of an argon-argon dating method is improved.

A liquid lithium-cooled fission reactor optimized for producing radioactive materials. The reactor is designed to enhance the availability of rare radioactive materials that have significant value for national defense, industrial research, and medical care. This invention has tangible design attributes that can be tailored to create one or more scarce and valuable radioactive materials. In particular, the reactor design is optimized for low-cost production of large quantities of radioactive tritium needed in national-defense and fusion-breeder programs. There are four core designs applied to this invention, all of which produce tritium and surplus heat that can generate byproduct electricity, thereby reducing the cost of radioactive-material production. Three of the embodiments furnish radioactive fission products, such as molybdenum-99, that can be extracted with high efficiency and rapid processing, thus fulfilling a critical supply and price shortfall in radioisotopes used for medical diagnosis and treatment.

A traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein. In a traveling wave nuclear fission reactor, a nuclear fission reactor fuel assembly comprises a plurality of nuclear fission fuel rods that are exposed to a deflagration wave burnfront that, in turn, travels through the fuel rods. The excess reactivity is controlled by a plurality of movable neutron absorber structures that are selectively inserted into and withdrawn from the fuel assembly in order to control the excess reactivity and thus the location, speed and shape of the burnfront. Controlling location, speed and shape of the burnfront manages neutron fluence seen by fuel assembly structural materials in order to reduce risk of temperature and irradiation damage to the structural materials.

A traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein. In a traveling wave nuclear fission reactor, a nuclear fission reactor fuel assembly comprises a plurality of nuclear fission fuel rods that are exposed to a deflagration wave burnfront that, in turn, travels through the fuel rods. The excess reactivity is controlled by a plurality of movable neutron absorber structures that are selectively inserted into and withdrawn from the fuel assembly in order to control the excess reactivity and thus the location, speed and shape of the burnfront. Controlling location, speed and shape of the burnfront manages neutron fluence seen by fuel assembly structural materials in order to reduce risk of temperature and irradiation damage to the structural materials.

The invention relates to a silicon oxide dispersion strengthening steel and a preparation method thereof, and belongs to the technical field of low-activation and anti-irradiation metal materials. Thesilicon oxide dispersion strengthening steel is composed of a dispersion strengthening phase nano SiO2 particle and a matrix 8-20 Cr martensite / ferrite steel, the characteristic of conversion of crystal structures such as quartz, scale quartz and square quartz of the SiO2 crystal is mainly used, and the oxide / matrix interface structure is adjusted through reasonable thermal treatment, so that themechanical property and the radiation resistance of the steel can be remarkably improved. In addition, the silicon oxide dispersion strengthening steel is used as first wall structure material of a future fusion reactor and a nuclear fuel cladding material of a fourth-generation fission reactor, thus the safety of the reactor system can be improved, and the power generation efficiency and economical efficiency of the reactor system can be improved.

The invention relates to a wide-energy-spectrum neutron fluence on-line monitoring system. The wide-energy-spectrum neutron fluence on-line monitoring system comprises a neutron probe, a signal acquisition device and a data processing device, wherein the neutron probe comprises a cylindrical moderator, a thermal neutron detector, a fast neutron detector and a shielding layer, and can obtain relatively flat neutron fluence response in an energy range of 1 eV to 20 MeV; the signal acquisition and processing part can realize setting of parameters such as high voltage and magnification times of the detector and real-time online display of neutron fluence. The system can remarkably solve the defect that an existing long neutron counter is heavy in size and has the advantages of being light, portable, high in gamma-ray and X-ray resisting capacity and the like, so that the system can be applied to neutron fluence monitoring of strong neutron radiation fields such as accelerators, fission reactors and fusion test devices.

The invention relates to a reactor internal component critical heat flux real-time monitoring method based on a sonic sensor, which is used for real-time monitoring on the event of critical heat flux (CHF) caused when abnormal working condition occurs on the first wall of the fusion reactor, facing to a plasma component, and a fission reactor core fuel assembly; according to the invention, since sound wave technology is adopted to monitor the event of CHF caused by a high heat component, the method has the remarkable advantages of good real-time property, large monitoring range, high precision and the like, overcomes the limitations that detection points are large in number, data acquisition is lagged and only limited parts can be detected when a thermocouple is adopted for monitoring the CHF event, avoids the defects of low surface emissivity, great change, uncertain temperature distribution caused by high parasitic reflection, and the like when infrared testing mode is adopted for monitoring the CHF event, and solves the difficulty in realizing real-time monitoring on the reactor internal component critical heat flux.

The invention relates to a fuel component and a method for manufacturing of a fuel component. The fuel component is adapted to be used in fission reactors. The fuel component comprises a core consisting of a first material, and a layer consisting of a second material. The layer encloses at least partly the core. The first material comprises a fissile substance. The fuel component comprises an intermediate layer between the core and the layer. The intermediate layer has a material gradient that comprises a decrease of the concentration of the first material from the core to the layer and an increase of the concentration of the second material from the core to the layer.

The present invention provides a nuclear fuel cladding pipe, which is made of an aluminum silicon carbon ceramic material or an aluminum silicon carbon ceramic matrix composite material, or the surface coating of the nuclear fuel cladding pipe is made of the aluminum silicon carbon ceramic material or aluminum silicon carbon ceramic matrix composite material, wherein the general structure formulaof the aluminum silicon carbon ceramic material is (Al4C3)n(SiC)m, wherein, n denotes the layer number of Al4C3 in an aluminum silicon carbon unit cell and is a natural number; m denotes the layer number of SiC in the aluminum silicon carbon unit cell and is a natural number. The nuclear fuel cladding pipe can meet the requirements of structural elements in the fourth generation fission reactor nuclear power system on high temperature resistance, corrosion resistance, oxidation resistance, neutron radiation resistance and good neutron radiation stability, and has a good application prospect.

Methods of controlling the reactivity of a molten salt fission reactor. The molten salt fission reactor comprises a core and a coolant tank (101), the core comprising fuel tubes (103) containing a molten salt fissile fuel, and the coolant tank containing a molten salt coolant (102), wherein the fuel tubes are immersed in the coolant tank. The methods comprise dissolving a neutron absorbing compound in the molten salt coolant, the neutron absorbing compound comprising a halogen and a neutron absorbing element. The first method further comprises reducing the neutron absorbing compound to a salt of the halogen and an insoluble substance comprising the neutron absorbing element, the halogen being fluorine or chlorine, wherein the insoluble substance is not volatile at a temperature of the coolant during operation of the reactor. In the second method the one or more neutron absorbing compounds are chosen such that reduction of the neutron absorbing capacity of the one or more neutron absorbing compounds due to absorption of neutrons compensates for a fall in reactivity of the core in order to control fission rates in the core. Apparatus for implementing the methods are also provided.

A target irradiation system for irradiating a radioisotope target in a vessel penetration of a fission reactor, including a target elevator assembly including a body portion defining a central bore and an open bottom end, a center tube that is disposed within the central bore of the body portion, a target basket that is slidably receivable within the center tube, and a winch that is connected to the target basket by a cable, wherein the target basket is configured to receive the radioisotope target therein and be lowered into the vessel penetration of the reactor when irradiating the radioisotope target.

The invention belongs to a nuclear power design technique, and particularly relates to a loading scheme of an advanced fusion-fission subcritical energy reactor core. The reactor core structure comprises a plurality of fuel assembly modules which are arranged in a hoop along an annular plasma fusion area, wherein each fuel assembly module comprises a plurality of fuel assemblies arranged in a polar direction of the plasma fusion area; a first heat-resistant and radiation-resistant wall is arranged at one side of each fuel assembly towards the plasma fusion area; a tritium-producing coating is arranged at the other side of each fuel assembly opposite to each first wall; an outer shielding layer is arranged outside each tritium-producing coating. By adopting the advanced fusion-fission subcritical energy reactor core, the design requirements of the energy magnification factor and the tritium breeding ratio can be simultaneously met; the continuous growth of the magnification factor and the tritium breeding ratio can be maintained within a long period of time; the breeding performance of the reactor core is good.

The invention relates to a fission reactor criticality calculation system based on a Monte Carlo method. The fission reactor criticality calculation system comprises a grid numerical analysis module, a shannon entropy calculation module and a shannon entropy evaluation module. The system is implemented through the following steps of: firstly, resolving grid partition information filled in an input file by a user, then calculating shannon entropy of each generation according to a shannon entropy calculation formula, performing numerical correction on a shannon entropy curve, quantitatively giving algebra required by system source convergence, and automatically regulating non-active generation algebra in the system, wherein the shannon entropy evaluation module fits a shannon entropy curve according to the least square method, judges the algebra required by convergence by use of the corrected shannon entropy according to convergence condition, and then sets the non-active generation algebra of the system. According to the fission reactor criticality calculation system based on the Monte Carlo method, the restriction that the non-active generation algebra of a traditional fission reactor criticality calculation system is manually set by a user is changed, and the efficiency of calculating the fission reactor criticality by use of the Monte Carlo method is improved.

The invention provides a heat exchange medium which comprises solid particles and fluid. The invention further provides a heat exchange system which comprises the heat exchange medium, a first heat exchanger, a mixing device, a separating device, a second heat exchanger and a first conveying device, wherein the mixing device is arranged on the upstream of the first heat exchanger and used for mixing the solid particles of the heat exchange medium and conveying the mixture to the first heat exchanger; the separating device is arranged at the downstream of the first heat exchanger and used for separating the solid particles and the fluid of the heat exchange medium discharged from the first heat exchanger; and the first conveying device is used for conveying the solid particles separated by using the separating device into the mixing device after passing through the second heat exchanger. Furthermore, the invention further provides a nuclear reactor system comprising the heat exchange system. The gas-solid or liquid-solid cooling medium provided by the invention has the advantages of high thermal capacity, low-pressure system, no corrosion, off-line processing and the like. The fission reactor provided by the invention can safely and reliably operate under the high power density or ultrahigh power density.

The invention discloses a modeling system which is used in reactor design and used for reactor core component design. According to the modeling system, the geometric specificity of a fission reactor core is analyzed, a great amount of repeatedly constructed geometries are adopted, a CAD engineering model and a Monte Carlo calculation model of the fission reactor core are automatically established by establishing parameters, the Monte Carlo calculation model is used for radiating and transporting input of a Monte Carlo calculation program, and the significant physical quantity of the reactor core can be calculated after the Monte Carlo calculation model is obtained. By constantly evaluating the physical parameters of the calculation result, the modeling system can be used for performing repeated iteration correction on a design scheme of the fission reactor core till a reactor core scheme which makes a user satisfactory is made. The modeling system has the benefits that the tediousness that Monte Carlo calculation programs are written and files are input manually is avoided, the whole modeling process is very visible, the probability of mistakes is greatly reduced, and the design time of the reactor core is shortened.

The invention discloses a method for controlling the reactivity of a molten salt fission reactor. The molten salt fission reactor comprises a core including fuel tubes (103) containing molten salt fission fuel and coolant tanks (101) containing molten salt coolant (102), wherein the fuel tubes are submerged in the coolant in the slot. The method includes the steps of dissolving a neutron absorbing compound including a halogen and a neutron absorbing element in a molten salt coolant. The first method also includes reducing a neutron absorbing compound to a salt of said halogen and an insoluble species comprising said neutron absorbing element, said halogen being fluorine or chlorine, wherein said insoluble species undergoes fission in said molten salt Non-volatile at the temperature of the molten salt coolant during operation of the reactor. In a second approach, the one or more neutron-absorbing compounds are selected such that the one or more neutron-absorbing compounds compensate for a decrease in neutron-absorbing capacity due to absorbing neutrons in the core The resistance is reduced to control the fission rate of the core. The invention also provides a device for implementing the method.

The invention belongs to the field of cladding designs of a fusion and fission hybrid reactor, and particularly relates to a Th-U self-sustaining circulating full fused salt fuel hybrid reactor system and an operation method thereof. The Th-U self-sustaining circulating full fused salt fuel hybrid reactor system is characterized in that a fast fission breeder reactor provides an initial easily fission fuel required by starting of a thermal fission reactor, a cladding design of the thermal fission reactor utilizes an arrangement strategy of a seed-cladding to improve the entirety neutron economy of the system, and the purposes of a high energy enlargement factor of the system, tritium breeding and thorium-uranium self-sustaining circulating of the system are realized; <233>U is loaded in an energy generating region, so that the energy generating region has a good neutronics property and is mainly used for realizing the purposes of energy amplification of the system, neutron multiplication and most <233>U breeding of the system; superfluous neutrons enter a tritium producing region and are used for tritium breeding and part of <233>U breeding of the system. According to the hybrid reactor, due to self-sustaining circulating of thorium and uranium, a thorium fuel is converted into <233>U and is gradually burn up in an operation process, and the Th-U self-sustaining circulating full fused salt fuel hybrid reactor system can effectively and stably operate for a long time just by gradually adding the thorium fuel and removing the generated fission product.

The invention provides a double-station material automatic distribution and automatic fission system. The system comprises two parallel tracks. The tracks are divided into fission zones, material distribution zones and discharging zones. A set of material trolleys with the two ends connected with winches are arranged on each track. A double-station material distribution system is arranged in the material distribution zones and conveys materials to the material trolleys after receiving the materials. Fission reactors are arranged in the fission zones and used for conducting high-temperature and high-pressure fission treatment on the materials, entering the fission reactors, in the material trolleys. Discharging winches and material receiving conveyors are arranged in the discharging zones. The discharging winches turnover carriages of the material trolleys laterally and pour the materials subjected to fission treatment into the material receiving conveyors, and the materials are conveyed to the downstream part. The double-station material automatic distribution and automatic fission system adopts a series of automatic devices in a combined mode, and the automation degree is high; the fission reactors are of a horizontal structure, mechanisms such as pressure gages and safety valves are interlocked with a whole safety operation mechanism, and the safety of the system is improved; and the conveying and treatment speeds of organic garbage are increased.

The Deuterium Reactor is a fusion reactor whose design is based upon a non-singular electrostatic required by the quantization of electric charge. This potential allows for a significant reduction in the fusion barrier of deuterium nuclei when these nuclei are held in close proximity, as within a crystal, and preconditioned using a magnetic field. This manner of fusion barrier reduction produces direct fusion of two deuterium nuclei into a helium nucleus without attendant hazardous radiation of classical fusion reactors. The energy released in the deuterium reactor may be used in different ways for different applications and its use will result in a significant reduction in fossil fuel use, a significant reduction in radioactive waste by replacing fission reactors, and a significant impact upon the world economy.

The invention relates to a multiscale fissioning core visualization method based on an automatic merging hierarchical structure tree. The visualization method is used in fission reactor design to check the correctness of the design of a reactor core which has been constructed or is being constructed. Based on the hierarchical feature of a fission reactor, a user input model is analyzed, grouping of a hiberarchy relation structure is established intelligently, a corresponding rendering unit is generated based on grouping, an overall rendering tree structure is established, and quick visualization of an area in which a user is interested is achieved; visualization speed is increased greatly, real-time visualization during modeling of a fine reactor core model is realized, and then fissioning core modeling efficiency is improved remarkably.

A traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein. In a traveling wave nuclear fission reactor, a nuclear fission reactor fuel assembly comprises a plurality of nuclear fission fuel rods that are exposed to a deflagration wave burnfront that, in turn, travels through the fuel rods. The excess reactivity is controlled by a plurality of movable neutron absorber structures that are selectively inserted into and withdrawn from the fuel assembly in order to control the excess reactivity and thus the location, speed and shape of the burnfront. Controlling location, speed and shape of the burnfront manages neutron fluence seen by fuel assembly structural materials in order to reduce risk of temperature and irradiation damage to the structural materials.

The invention discloses a high-strength high-toughness vanadium alloy which comprises the following components in percentage by weight: 3.2-6.0 wt% of Cr (chromium), 3.2-6.0 wt% of Ti (titanium), and 0.1-4.0 wt% of Y (yttrium) and Zr (zirconium). The preparation method comprises the following steps: after adding the Y element and Zr element on the basis of a V-Cr-Ti three-component alloy system, carrying out smelting, homogenizing treatment, coating, hot deformation, annealing, cold deformation, recrystallization annealing and the like to synthesize the V-Cr-Ti-(Y+Zr) alloy. The vanadium alloy gives consideration to the strength and toughness at both room temperature and high temperature, and can be used as a potential structural material for fusion reactors and fission reactors. The preparation method can be used for large-scale production of the vanadium alloy.

The present invention discloses a method of operating a nuclear fission reactor. The reactor includes a reactor core including an array of fuel assemblies and a coolant tank containing a coolant. Each fuel assembly extends generally vertically and includes one or more fuel tubes containing fissionable fuel. The fuel lines are immersed in coolant. The method includes: monitoring and / or modeling the fuel concentration and / or fission rate of each of the fuel assemblies; In this case, the fuel assemblies are moved horizontally within the array to control the rate of fission in the reactor core. The invention also discloses a nuclear reactor implementing the method and a fuel assembly used in the method.