39 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.

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 the technical field of fusion reactors, and particularly relates to a steam generation system suitable for a fusion reactor nuclear power station. The system comprises a hot steam generator, a saturated steam generator, a heat exchanger, a pressure tank, a low-temperature fluid pump, a high-temperature fluid pump, a steam-liquid separator and a plurality of switch valves.Compared with the common steam generator, the system of the invention has the following characteristics that a first loop contact interface, a second loop contact interface and a first loop device arenot additionally arranged, and a first loop is simple, so that the problems that the construction cost is increased and the safety is reduced due to the fact that the design of the first loop is complicated are effectively solved. According to the invention, the steam generated by the steam generation system is basically the same as the steam generated by the conventional fission nuclear power station steam generator; and in a fusion reactor power station, except the steam generation system, a secondary circuit and other corresponding auxiliary systems can adopt the currently mature fission reactor secondary circuit design and construction process and operation experience.

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 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 invention discloses a method for calculating the velocity and argument cosine of heavy nuclei in the process of epithermal neutron scattering. In fission reactors, the maximum elastic scattering cross section method based on the energy range of heavy nuclei in the superheated region is used to simulate the neutron elastic scattering process; when the neutrons in the superheated region with energies from 0.4 eV to 210 eV scatter resonantly and elastically with the heavy nuclei in the nuclear reactor, at first, the velocity and argument cosine of heavy nuclei are sampled with a certain probability, then the velocity and argument cosine are accepted with a certain probability, or the velocity and argument cosine are discarded with a certain probability, so that the final velocity and argument cosine of heavy nuclei are obtained. The invention optimizes and improves the sampling of the velocity and the argument cosine of the heavy nucleus in the resonance elastic scattering process ofthe traditional neutron and the heavy nucleus, and at the same time, the elastic scattering cross section of heavy nuclei scattered by neutrons in the energy range of superheated region is also takeninto account, which enables the model to calculate the elastic scattering process of the neutrons and heavy nuclei more accurately.