61 results about "Tritium" patented technology

Main characteristic of the invention is the exploitation of the very large flux of high energy neutrons generated in a nuclear fusion tokamak with beam-driven plasma for the purpose of producing radioisotopes. It specifically relates to the deployment of precursor materials into the core of a nuclear fusion tokamak with beam-driven plasma for the purpose of exposing these materials to the neutron flux generated in the plasma by the process of nuclear fusion of Deuterium (D) and Tritium (T) to generate radioactive isotopes, so-called radioisotopes. The precursor materials contained in metallic capsules are incorporated into the blankets that surround the plasma.

This invention discloses a method for producing radioactive isotopes in the core of an advanced high-temperature reactor using a spherical target. This invention overcomes the limitations of using neutron irradiation channels or fuel-related components of nuclear reactors, such as conditioning rods in heavy water reactors and combustible poison rods in pressurized water reactors, to produce radioactive isotopes. It sets a precedent for the mass production of radioactive isotopes using the core of a nuclear reactor, and has significant strategic implications for exploring high-value applications of commercial advanced high-temperature reactors to reduce operating costs and mass-produce radioactive isotopes, including the strategic isotope tritium.

This invention relates to a vertical oxidation-combustion collection system for sediment or biological radioactive organic carbon-tritium samples and its method of use, comprising: a vertical oxidation-combustion device for oxidizing and combusting the sample and catalyzing it into products to be separated; and a sample collection device comprising: a water vapor trapping module for collecting tritium samples; and a carbon trapping module for collecting... 14 Sample C; A rear-mounted protection module is used to discharge exhaust gas, collect waste liquid, and prevent external air interference with sample collection. This invention, by placing the collection port below the collection container, allows for the collection of tritium samples from combustion products under gravity. 14 Sample C is collected more thoroughly and can also prevent tritium from entering. 14 C-sample collection module, to avoid 14 Sample C enters the post-protection module; more importantly, nitrogen blowing or inert gas blowing can completely remove and collect any residual tritium sample in the collection module. 14 Like C, avoid the occurrence of the memory effect.

The invention relates to the technical field of isotope batteries and discloses a nanopore-containing power generation assembly. The nanopore-containing power generation assembly comprises a transduction layer, the loading layer is arranged on the surface of the transduction layer in advance, and the loading layer is provided with preset nanopores. The technical problem that in the preparation process of the radioactive source in the isotope battery, due to the fact that the hydrogen absorption temperature of a hydrogen storage material is high, large thermal mismatching exists between a transducer device and the hydrogen storage material, growth stress is generated in the phase change process, and consequently the transducer device may have serious structural damage can be solved; the loading layer arranged on the surface of the transduction layer in advance is adopted, and the loading layer is provided with the nanopore structure, so that when the loading layer serves as a basic structure for preparing the isotope power generation assembly, for example, nanopores are filled with radioactive gas such as tritium, the loading layer structure is stable, and the isotope power generation assembly is prepared. The probability that the energy conversion layer structure is damaged due to growth stress generated in the phase change process can be reduced, and then it is guaranteed that the structure of the prepared isotope power generation assembly is stable.

The present application particularly relates to a method for producing helium-3 from heavy water in a heavy water reactor, comprising: step 1, subjecting the heavy water in a moderator system of a heavy water reactor unit to negative pressure rectification to obtain concentrated high-tritium heavy water; step 2, subjecting the concentrated high-tritium heavy water in step 1 to beta decay to produce helium-3 raw gas; and step 3, subjecting the helium-3 raw gas in step 2 to active carbon adsorption, helium-argon separation and hydrogen isotope gas adsorption to obtain helium-3. The method for producing helium-3 from heavy water in a heavy water reactor avoids technical difficulties and high costs of helium-3 / helium-4 isotope separation, and has high technical maturity and good economy.

This invention discloses a method for accurately tracing geological hydrogen migration, belonging to the field of geological hydrogen exploration and migration tracing technology. The invention involves preparing stable hydrogen isotopes and radioactive tritium tracers with specific geochemical fingerprints and injecting them into target geological units; deploying a three-dimensional monitoring network including sensor arrays and automatic samplers to collect hydrogen concentration data and gas samples in real time; laboratory analysis to determine hydrogen isotope composition and rare gas concentrations, using an N2-He-Ar ternary diagram to identify gas sources and eliminate interference; integrating data and performing numerical simulations using migration control equations coupled with isotope fractionation effects, optimizing parameters through inversion algorithms; and generating a three-dimensional visualization map of hydrogen migration paths and a quantitative report based on the inversion results. This invention achieves a breakthrough in geological hydrogen migration from qualitative detection to quantitative tracking, significantly improving the reliability and accuracy of tracing, enabling real-time dynamic capture of the migration process, and providing precise decision-making basis for hydrogen exploration.

The invention relates to the technical field of nuclear facilities, and discloses a deuterium-tritium fusion device containing zoning method, device, equipment and medium, and the method comprises the steps: preliminarily dividing each tritium-related room into a plurality of tritium-related regions with different tritium containing grades based on a preset tritium-related concentration limit value; identifying a transition room between two tritium-related areas with adjacent spatial positions and adjacent tritium containing levels, and generating a plurality of different candidate zoning schemes by adjusting the tritium-related area to which the transition room belongs; for any candidate zoning scheme, determining system balance time of the candidate zoning scheme based on the space volume of each tritium-related region, the annual average tritium release rate and the preset net exhaust air rate; and identifying the shortest system balance time in each system balance time, and determining the candidate partitioning scheme corresponding to the shortest system balance time as a target partitioning scheme. The method has the beneficial effect that the dynamic response performance of the negative pressure ventilation system is optimized.

In order to realize a fuel breeding unit that does not require the introduction of tritium recovery piping, a fuel breeding unit (1) converts kinetic energy of neutrons generated by nuclear fusion in plasma into thermal energy and generates tritium using neutrons. The fuel breeding unit (1) discharges the generated tritium directly into the reactor.

The invention discloses a tritium monitor for realizing alpha, beta and gamma background deduction and a tritium activity concentration detection method.The tritium monitor comprises a detector I, a detector II and a signal processing unit, and the detector I is used for detecting alpha particles, beta particles and gamma particles; wherein the beta particles comprise tritium decay beta particles and high-energy beta particles with energy higher than that of the tritium decay beta particles, and the detector II is used for detecting alpha particles, gamma particles and high-energy beta particles and outputting signals II corresponding to counting rates of the particles. The signal processing unit is electrically connected with the detector I and the detector II, and the signal processing unit is configured to be used for receiving the signal I and the signal II, calculating a background equivalent current jointly generated by the alpha particles, the gamma particles and the high-energy beta particles based on the signal II and a preset calibration coefficient, and calculating the total ionization current based on the difference value of the total ionization current and the background equivalent current. And obtaining a net tritium current, and determining the tritium activity concentration based on the net tritium current.

Electrolysis induces cold fusion, producing helium-4. [Solution] A mixture of heavy water D and tritium T is placed in a container and electrolyzed using nickel and palladium electrodes to produce helium. The produced helium is collected in a container.

The present application relates to a kind of tritium separation method of tritium-containing wastewater and its reactor system, application, the tritium separation method of tritium-containing wastewater of the present application is ultraviolet irradiation promotes separation method, comprising: step 1) tritium-containing wastewater is introduced into the reaction containment of reactor system and is contacted with adsorption material to form liquid-solid two-phase contact system, step 2) after ultraviolet light is irradiated to liquid-solid two-phase contact system, make tritium enrich in adsorption material;Step 3) after the liquid-solid two-phase contact system of irradiation is completed, solid-liquid separation is obtained to low concentration tritium treatment water and tritium-rich adsorption material.The method of the present application is irradiated by ultraviolet light, accelerates hydrogen / tritium isotope exchange and adsorption process, can obtain higher tritium removal rate and adsorption amount under the same processing time.The present application belongs to energy-saving and environmental protection industry, especially in wastewater environmental control and nuclear pollution wastewater treatment in the environmental protection application of the recovery treatment of nuclear radioactive pollution elements plays a key role.

A method and a device for nuclear fusion with a deuterium alloy reactor are provided. The device includes: a high-energy neutron generator unit for generating a high-energy neutron current to bombard densely confined static deuterons in the deuterium alloy, producing high-energy deuterium nuclei; a deuterium alloy reactor unit, where deuterium gas forms a deuterium alloy upon entering the reactor, when bombarded by an external high-energy neutron current, high-energy deuterium nuclei are generated, the high-energy deuterium nuclei collide with tightly confined static deuterium nuclei in the deuterium alloy reactor, overcoming a Coulomb barrier and undergoing a nuclear fusion reaction; the high-energy deuterium nuclei further collide violently with the high-energy tritium nuclei produced by the nuclear fusion reaction to continue the nuclear fusion reaction; and an energy conversion unit configured to convert energy generated by the nuclear fusion reaction into usable energy.

The invention relates to the technical field of nuclear radiation monitoring, in particular to an automatic gaseous tritium sampling and measuring method which comprises the following steps: a feature space mapping step: acquiring ionization current and environmental data, and extracting multi-dimensional signal features; a memory entropy observation step: inputting the features into a dynamic memory entropy observer model, and resolving a current memory entropy value representing the nuclide adsorption degree; an intelligent decision-making step: generating an optimal sampling control strategy by using a reinforcement learning decision-making model based on the memory entropy; a dynamic regulation and control and output step: controlling the valve to be switched between a continuous sampling mode and a pulse cleaning mode in response to the strategy, and outputting a measurement result; by introducing the memory entropy state variable, the sensor is converted into an active gamer, and the effective service life of the sensor is prolonged to the maximum extent.

This application provides a steel material for reactors and a method for preparing the same. The steel material for reactors includes a steel substrate and a protective coating. The protective coating includes an in-situ oxide layer on the surface of the steel substrate and an alumina layer on the in-situ oxide layer. The in-situ oxide layer includes α-Fe₂O₃, and the alumina layer includes α-Al₂O₃. In the steel material for reactors, the α-Fe₂O₃ in the in-situ oxide layer serves as a growth template for the alumina layer. Through the mutual matching and induction effect of crystal structures, the protective coating forms a dense protective barrier on the surface of the steel substrate, effectively reducing the permeation of hydrogen and its isotopes deuterium and tritium.

The application discloses a neutron tube deuterium-tritium reservoir, a nickel tube is arranged in the nickel cylinder body, through holes are arranged on the nickel tube, a hydrogen absorption coating is arranged on the inner and outer surfaces of the nickel tube and the side surface of the nickel cylinder body, ceramic tubes are arranged between the nickel cylinder body and the nickel tube and in the nickel tube, sealing openings are arranged at the bottom of the nickel cylinder body and the nickel tube, insulating sealing ceramic glue is arranged at the sealing openings, and the two ends of the ceramic tubes pass through the insulating sealing ceramic glue at the bottom of the nickel tube and the bottom of the nickel cylinder body respectively; the internal double-layer structure greatly increases the storage capacity of the hydrogen absorption material for deuterium-tritium gas, the double-layer spiral ceramic tubes are heated to cooperate with the double-layer structure, the hydrogen absorption material heating surface of the device can be increased, the size of the heating current is controlled to realize accurate control of the deuterium-tritium gas release amount, the closed reservoir made of alloy metal nickel avoids exposing the hydrogen storage material, has the function of hydrogen permeation, and guarantees the deuterium-tritium release.

The application discloses a one-dimensional hydrogen isotope transport analysis method and system, and the method comprises the following steps: S1, a complete multi-trap tritium transport parameter system is constructed by inputting multiple sets of trap characteristic parameters and distinguishing the differential influence of different types of traps; S2, the retardation effects of different types of traps are superimposed by adopting a multi-trap interaction mechanism to represent the interaction of different traps in competing for tritium atoms and releasing tritium atoms cooperatively; S3, the breakthrough time is calculated based on the effective diffusion coefficient and the material thickness; S4, the hydrogen isotope permeation flux is calculated by adopting an infinite series method based on the breakthrough time; and S5, the permeation flux is associated and matched with the corresponding time to generate corresponding relationship data of the permeation flux and the time. The application reduces the simulation error of the hydrogen isotope retention amount, and can accurately analyze the permeation amount and the retention amount of tritium under diffusion limitation.

The invention belongs to the technical field of data processing, and relates to a neutron generator operation state monitoring method and system based on digital twinning. The method comprises the following steps: acquiring real-time operation electrical parameters to construct a virtual twinborn body; inverting the effective deuterium and tritium atom concentration in the target film in real time by using a physical difference equation containing a particle injection item and a thermal diffusion loss item; a beam spot drift correction factor is determined by calculating the difference between the real-time impedance of the neutron tube and the nominal design impedance, the beam spot area of the virtual twin is dynamically corrected, and the problem of prediction model failure caused by electrode ablation is solved; and finally, on the basis of the corrected state parameters, using a multi-parameter product model to predict the neutron yield and perform health degree diagnosis. Transparent monitoring of the internal state of the neutron tube is achieved, the problem of model distortion caused by beam spot drift is effectively solved, and the prediction precision of the whole life cycle is improved.

The invention relates to the technical field of nuclear radiation monitoring, in particular to a remote tritium on-line monitoring system, which comprises a data acquisition module for synchronously acquiring tail end tritium concentration, pipeline on-way environmental parameters and environmental background radiation data; the coupling model building module is used for building a gas-solid interface adsorption kinetics and fluid transportation coupling model and determining a dynamic transmission transfer function; the source item inversion reconstruction module is used for executing deconvolution operation and reversely deducing the real tritium concentration of a source end; the state evaluation and threshold adjustment module is used for calculating pipeline adsorption saturation in real time and dynamically adjusting an alarm threshold; the early warning generation module is used for generating an early leakage early warning instruction; the operation and maintenance prediction module is used for identifying pipeline maintenance requirements based on the adsorption performance decline trend; according to the invention, the defects of signal lag and peak smoothing are overcome, and second-level response to sudden pulse leakage is realized.

The invention relates to the technical field of deuterium-tritium fusion fuel measurement, in particular to a laser spectrum and mass spectrum combined fuel retention in-situ measurement method. Aiming at the problems of incapability of real-time in-situ monitoring, limited measurement precision, incapability of providing fuel retention space distribution characteristics and the like, the invention provides the following technical scheme: step 1, realizing external trigger time sequence synchronous control of an instrument; 2, acquiring a mass spectrum signal and determining a calibration coefficient; step 3, acquiring the distance between the endoscope and the position to be measured and feeding back the distance; 4, focusing the laser to a to-be-measured position; 5, synchronously collecting spectrum and mass spectrum signals; step 6, converting the mass spectrum signal into the total number of fuel particles in unit area on the wall surface; step 7, combining the spectrum and the mass spectrum to obtain depth distribution of fuel particle number at the position to be measured; 8, remote scanning of different areas of the inner wall of the Tokamak is achieved. Quantitative information of local fuel particles is provided, and overall distribution characteristics are obtained.

This invention relates to the field of environmental radiation measurement technology, and particularly to a system and method for sampling tritium concentration in trace atmosphere samples. The system includes: a sampling device for converting the gas to be measured into water tritide, and then measuring the tritium concentration in the trace atmosphere using a measuring device passing through the sample; the sampling device is configured to: evacuate the buffer using a vacuum pump to a preset vacuum level before introducing the gas to be measured into the circulating catalytic loop; in the circulating catalytic loop, the gas to be measured circulates multiple times through the catalytic furnace and the bubbler, achieving the conversion and collection of gaseous tritium into water tritide. This invention, by setting up a circulating catalytic loop, allows the gas to be measured to circulate repeatedly between the catalytic furnace and the collection bottle, allowing unreacted gas to repeatedly enter the catalytic furnace, ensuring that trace amounts of gaseous tritium are fully oxidized into water tritide, significantly improving the conversion efficiency.

The present invention simplifies a system for fuel circulation. A nuclear fusion reactor system (10) comprises: a vacuum container (20); an exhaust device (31) that is connected to the vacuum container (20) and discharges gas inside the vacuum container (20); a blanket (50) that is disposed inside the vacuum container (20) and surrounds a space in which plasma is formed; a tritium breeding material that is supplied to the blanket (50); and a fuel storage device (35) that is installed outside the vacuum container (20). The tritium breeding material is a liquid metal containing lithium and tin, and releases, from the tritium breeding material in the vacuum container (20), tritium generated by the tritium breeding material. The exhaust device (31) discharges the tritium released from the tritium breeding material to outside the vacuum container (20) toward the fuel storage device (35).

The invention discloses a method for calculating the effective tritium breeding ratio of a fusion reactor tritium self-sustaining breeding blanket. The method comprises the steps that a physical target and an engineering target of fusion reactor tritium self-sustaining are determined; establishing a cladding model, and carrying out neutronics calculation to obtain neutron thermal deposition and distribution results of the cladding model; calculating to obtain a temperature field and a distribution result based on the cladding model, the neutron thermal deposition and the distribution result; according to the obtained temperature field and the distribution result, obtaining an effective tritium release temperature zone of the cladding model, counting the area or volume of the effective tritium release temperature zone, and calculating the effective tritium release amount of the cladding model; calculating the total effective tritium release amount of the fusion reactor according to the effective tritium release amount of the cladding model; the effective tritium breeding ratio is calculated and compared with the theoretical total tritium breeding ratio, and the fusion reactor tritium self-sustaining capacity is judged. According to the invention, actual releasable tritium is quantified for the first time, and a real and reliable basis is provided for cladding design; the method can be directly used for cladding design optimization of major fusion devices such as ITER, CFETR and the like.

The application discloses a laser inertial fusion cladding and a design method thereof. The cladding is divided into a tungsten armor, a first wall, a tritium breeding self-cooling area and a back plate from near to far according to the distance from the fusion reaction center. The design method firstly establishes a simplified radial partition three-dimensional model of the cladding, then performs neutron transport calculation on the model to obtain the tritium breeding rate and the nuclear heat deposition spatial distribution of the cladding, combines material characteristics to perform thermal calculation and obtain the temperature distribution in the cladding, comprehensively considers the tritium breeding rate and the temperature distribution, adopts an optimization algorithm to determine the reasonable interval of the radial size of each functional area of the cladding, then performs fine modeling, and optimizes the design in the interval until the standard is met. The fusion cladding and the design method thereof can replace the complete cladding model in the traditional cladding engineering design with the spherical radial partition model on the basis of ensuring the accuracy, avoid the establishment of a complex model in the multi-physical field coupling calculation, optimize the design process of the cladding, and design the cladding suitable for the laser inertial confinement fusion.

This invention discloses a method for achieving high hydrogen / deuterium / tritium permeation resistance in oxide coatings. By applying an electric current to an oxide coating (such as zirconium oxide, chromium oxide, titanium oxide, aluminum oxide, etc.) at a certain temperature, the lattice thermodynamic equilibrium is disrupted by the electric field, resulting in a high concentration of lattice point defects generated in situ within the coating, significantly enhancing its hydrogen / deuterium / tritium permeation resistance. Experimental results show that these lattice point defects can form deep-level traps, capturing hydrogen isotope atoms and effectively inhibiting the diffusion and permeation of hydrogen and its isotopes within the coating, thus achieving high hydrogen / deuterium / tritium permeation resistance. This process is simple, efficient, low-cost, and widely applicable, enabling the efficient preparation of oxide coatings with high hydrogen / deuterium / tritium permeation resistance. It has significant application prospects in the fields of tritium permeation protection in fusion reactor structural materials, hydrogen embrittlement protection of metals, and hydrogen permeation resistance in materials used in hydrogen-containing and hydrogen isotope-containing environments.

This invention relates to the field of nuclear radiation monitoring technology, specifically to an automated gaseous tritium sampling and measurement method, comprising: a feature space mapping step: acquiring ionization current and environmental data, and extracting multidimensional signal features; a memory entropy observation step: inputting the features into a dynamic memory entropy observer model, and calculating the current memory entropy value characterizing the degree of nuclide adsorption; an intelligent decision-making step: generating an optimal sampling control strategy based on the memory entropy value using a reinforcement learning decision model; and a dynamic control and output step: switching the valve between continuous sampling and pulse cleaning modes in response to the strategy, and outputting the measurement results. This invention, by introducing a memory entropy state variable, transforms the sensor into an active player, maximizing the effective service life of the sensor.

A method of making a radioisotopic power source, including receiving a predetermined amount of a plurality of Metal-Organic Framework (MOF) particles within a reactor vessel, degassing the received predetermined amount of the MOF. The degassing includes placing the predetermined amount of the MOF under vacuum conditions, heating the received predetermined amount of the MOF above a first predetermined temperature for a first predetermined time period, and sealing the heated MOF. The method also includes cooling the heated and sealed predetermined amount of the MOF to a second predetermined temperature, while maintaining a pressure of the reactor vessel to a first predetermined pressure value for a second predetermined time period, receiving a predetermined amount of a plurality of beta emitter particles at a gaseous state and mixing the predetermined amount of beta emitter particles with the cooled predetermined amount of the MOF.

The invention discloses a pebble bed dust migration filtering characteristic simulation method and system, a terminal and a medium, and relates to the field of fusion reactor nuclear fuel circulation. The simulation method comprises the steps of cladding multiplication pebble bed particle accumulation simulation, pebble bed structure data transmission, tritium extraction gas purging simulation in a cladding multiplication pebble bed, flow field data transmission, radioactive dust particle movement migration process simulation in the cladding multiplication pebble bed and radioactive dust particle capture deposition process simulation in the cladding multiplication pebble bed. And finally, obtaining data through result post-processing, and carrying out analog simulation analysis on the blowing migration filtering characteristics of the radioactive dust in the pebble bed of the fusion reactor cladding multiplication tritium production area under different multiplication cladding types, different blowing working conditions and different dust characteristics. According to the method, detailed dust migration filtering characteristic data can be obtained, and the purpose of improving the fineness and accuracy of radioactive dust migration filtering behavior simulation is achieved.

The present disclosure provides a nuclear fusion method and a blanket structure capable of simplifying a structure of a blanket.SOLUTION: The nuclear fusion method includes introducing helium-3 (3He) gas into a nuclear fusion blanket from the outside to produce tritium gas.SELECTED DRAWING: Figure 6

The invention provides a system and a method for removing radioactive tritium from air. The system comprises a saturation and heating unit used for heating air and enabling the air to contain saturated water vapor; the catalytic exchange washing unit is used for transferring radioactive tritium contained in the air from the saturation and heating unit into liquid-phase water; the condensing unit is used for condensing water vapor in the gas from the catalytic exchange washing unit; wherein the catalytic exchange washing unit comprises a reaction column and a natural water supply unit; the reaction column comprises a filler and a catalyst; wherein when the handling capacity of air corresponds to an air tower velocity less than or equal to 0.15 m / s, the filler and the catalyst are filled in bulk by selecting hydrophilic filler and catalyst; when the air handling capacity corresponds to the air tower speed larger than or equal to 0.15 m / s, the packing and the catalyst are selectively structured, and the packing and the catalyst are filled in a layered mode.

The application belongs to the technical field of electric insulation for high pressure of deuterium-tritium radiation in controllable nuclear fusion, and discloses a potential isolation mechanism and a welding method thereof. In the mechanism, the insulation support is sleeved outside the ceramic sleeve; there are two metal bellows, one end of each of the two metal bellows is welded and fixed to two ends of the ceramic sleeve correspondingly; there are two support sections, one end of one of the two support sections is welded to the other end of one of the metal bellows and the other end is fixed to the vacuum chamber, and one end of the other of the two support sections is welded to the other end of the other of the metal bellows and the other end is fixed to the end cover; some adjusting bolts are connected between one of the support sections and the insulation support, and the rest of the adjusting bolts are connected between the other of the support sections and the insulation support. The application can solve the problems of radiation compatibility and potential isolation caused by the cable leading out of the internal high-voltage electrical equipment during deuterium-tritium fusion, and the size can be increased to meet the use requirements of large scientific devices.