195 results about "Tritium" patented technology

The invention discloses a helium cold fusion cladding experiment module device, which mainly comprises a first wall proliferation agent area and a rear baffle plate, is circumferentially provided with a cover plate, a bottom plate, a left side plate and a right side plate, has higher engineering characteristics similar to a cladding design scheme, is convenient to carry out neutronics measurement, and is high in reliability. Based on the device, the technical level of fusion blanket neutronics verification and measurement technology and blanket tritium breeding performance evaluation in China can be improved, independent tritium production blanket performance data is accumulated, the requirement of ITER-TBM engineering development is met, and a foundation is laid for CFETR helium cooling blanket engineering design.

Disclosed in the present invention is a tritium production module system in a reactor. The tritium production module system has a multi-layer structure distributed in the radial direction of the reactor, and the tritium production module system comprises a tritium production area and a circulation area, wherein the tritium production area comprises a tritium breeding material, the tritium breeding material interacting with neutrons to produce tritium elements; the tritium production module system is in fluid communication with the outside of the tritium production module system via the circulation area; the circulation area comprises a tritium pipeline, the tritium pipeline being used for circulation of a reactor coolant and a tritium-containing medium, and the tritium pipeline extending in the axial direction of the reactor; a plurality of holes are provided in a pipe wall of the tritium pipeline, and the tritium pipeline is in fluid communication with the tritium production area via the holes, thereby improving the ability and efficiency of tritium transfer in the tritium production module system.

The invention provides a normal-temperature fusion target assembly, a laser fusion reactor system and a laser fusion power generation method.The fusion target assembly comprises a pair of coaxial compression cones with opposite cone tops, the compression cones are filled with normal-temperature boron deuteride lithium spherical crown targets, the fusion target assembly further comprises an ignition cone, and the cone tip end of the ignition cone is sleeved with a magnetic field coil; according to the light fusion power generation method, a spherical crown target is driven by compressed laser to generate centripetal implosion, high-density isometric plasma with a steep edge is formed near the sphere center in a collision mode, and tritium production reaction and fusion ignition combustion are generated by the collision plasma under driving of ignition laser. Energy released by high repetition frequency fusion is converted into electric energy which is stably output by a thermodynamic system; the invention further provides a laser nuclear fusion reactor system which is high in repetition frequency, low in cost and low in radioactivity.

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.

The invention discloses a charged particle identification method, a charged particle identification system based on a screen grid ionization chamber, computer equipment and a storage medium, and belongs to the technical field of radiation detection. The charged particle identification method comprises the following steps: acquiring signals generated by charged particles at a cathode and an anode of a detector and generating characteristic quantities; constructing particle identification parameters by using the obtained characteristic quantity; generating a two-dimensional spectrogram according to the identification parameters and observing distribution characteristics of different particle events; and identifying different particles according to the distribution characteristics of different particle events in the two-dimensional spectrogram. Particle energy is associated with an ionization range during identification, the problem of limitation of a traditional particle identification technology in a high event rate and complex particle product environment is solved, especially the capability of processing a tritium nuclear event superposition phenomenon in a resonance energy region is improved, and therefore the problem that due to the fact that the energy resolution is low, the probability of tritium nuclear event superposition is low can be solved. And the defect problem in the prior art that the detection range is limited due to the fact that complex particle types cannot be accurately identified is solved.

The invention discloses a method and a device for calculating the concentration of gaseous tritium discharged by a nuclear facility and an evaluation method and a device. The method comprises the following steps: acquiring meteorological data, environmental parameters and discharge parameters of the nuclear facility; determining an atmospheric dispersion factor of gaseous tritium discharged by the nuclear facility based on the meteorological data, the environmental parameters and the discharge parameters; determining a re-evaporation factor of the discharged gaseous tritium based on the atmospheric dispersion factor, the deposition property of the gaseous tritium, meteorological data and environmental parameters; determining a reevaporation dispersion factor of gaseous tritium based on the meteorological data, the environmental parameters and the reevaporation factor; and adding the atmospheric dispersion factor and the reevaporation dispersion factor to obtain the concentration of the gaseous tritium discharged by the nuclear facility. The method can accurately evaluate the concentration of gaseous tritium in the surrounding environment after the nuclear facility discharges gaseous tritium.

A mobile fusion device (MFD) comprises three main sections: the MFD reactor (467), MFD support sub-systems (468), and the MFD support and fuel injection sub-systems (469), which are indicated by broken lines in Figure 1 The MFD reactor (467) is where the Deuterium and Tritium fuel is heated to high enough temperatures and densities and confinement times that fusion occurs, releasing neutrons that are absorbed in and heats up the 17Li-83Pb breeder blanket where more Tritium fuel is bred and where heat is transferred to Helium coolant. The MFD support sub-systems (468) houses auxiliary power units that generate electricity to power the electrical sub-systems of the MFD, and houses the MFD control computer, and where the Helium coolant is cryogenically cooled and purified, and where the 17Li-83Pb breeder blanket fluid is purified by separating and storing Deuterium fuel and Tritium fuel and Helium coolant, and where a Helium heat exchanger is located. The MFD support and fuel injection sub-systems (469) houses Deuterium and Tritium neutral beam injection plasma heating systems and a Deuterium and Tritium fuel gas puffer system and a Deuterium and Tritium fuel storage and pellet manufacturing system, and where the 17Li-83Pb breeder blanket fluid is purified by separating and storing Deuterium fuel and Tritium fuel and Helium coolant, and where a Helium heat exchanger is located, and where the Helium coolant is cryogenically cooled and purified, and houses auxiliary power units that generate electricity to power the electrical sub-systems of the MFD. The MFD is housed within a multi-layered pressurised outer containment pod MLP, which has a multi-layered pressurised rear access hatch MLH. The MFD is mounted on a sliding raft SR for ease of access.

The invention discloses an online monitoring system for tritium in water based on a PSF array, and relates to the field of nuclear energy. The on-line monitoring system for tritium in water comprises a fluid control system, a PSF detector module and an electronics system, the fluid control system automatically controls to open electromagnetic valves of a sample inlet and a sample outlet, and a peristaltic pump is started, so that a water sample to be detected is pumped into a detection chamber of the PSF detector module through the sample inlet; after the detection chamber is filled with the to-be-detected water sample, the electromagnetic valve and the peristaltic pump are automatically controlled to be closed; the PSF detector module captures precipitation energy generated by tritium decay beta particles in a water sample to be detected through a PSF array, converts the precipitation energy into photon signals, and converts the photon signals into electric signals through a detection window and a double photomultiplier; and the electronics system performs signal preprocessing on the received electric signal, and obtains corresponding tritium content monitoring data to be detected based on the preprocessed electric signal. The on-line monitoring system for tritium in water can ensure that the finally obtained measurement result is more accurate.

The invention belongs to the technical field of tritium removal pump sets, and discloses a molecular pump set for evacuating a hydrogen isotope-containing super-large cavity, which comprises a stainless steel closed cavity and a super-large vacuum cavity, and is characterized in that a first tritium removal pump set and a second tritium removal pump set are respectively arranged in the stainless steel closed cavity through a fixing frame; the first tritium removal pump set and the second tritium removal pump set each comprise four tritium removal pumps, and the first tritium removal pump set and the second tritium removal pump set are integrally arranged in parallel and both communicate with the ultra-large vacuum cavity. According to the invention, the eight tritium removal pumps are mounted in the stainless steel chamber, and a connection mode of'parallel connection + series connection 'is creatively adopted, so that the requirement of continuous evacuation of the large chamber is met. In the initial stage, the eight tritium removal pumps operate in parallel so as to improve the overall pumping speed, the latter four tritium removal pumps serve as backing pumps, and the other four tritium removal pumps serve as main pumping pumps so as to improve the compression ratio of pumped tritium, increase the pressure of an exhaust port and improve the pumping efficiency of tritium.

The present invention relates to a simple and novel process to produce a beryllium containing lithium-based triphasic composite material for application as tritium breeder material in nuclear fusion reactors. The composite material comprises three different lithium-based breeder materials, out of which one phase contains beryllium in its structure that may serve as a neutron multiplier. The process does not yield any liquid or solid waste during production of the composite material and therefore is a safe and environment friendly process.

The invention relates to a method for measuring the sampling efficiency of a bubbling tritium sampler, which is characterized in that gas generated by a water vapor generator is introduced into the tritium sampler, the sampling flow in a gas circuit is controlled by a flow meter, and after sampling is finished, the sampling efficiency is the ratio of the increased mass of the tritium sampler to the reduced mass of the water vapor generator. According to the method, the non-radioactive method that water vapor generated by the water vapor generator serves as a sampling gas source to replace a tritiated water vapor gas source is achieved, meanwhile, a weighing method is adopted for measuring and calculating the sampling efficiency of the bubbling method tritium sampler and the sampling device, the problem that tritiated water vapor activity concentration tracing is difficult is effectively solved, and the sampling efficiency of the sampling device is improved. Pollution to a tritium sampler and a sampler and risks of tritiated water vapor leakage possibly caused in the efficiency measurement process by using a tritiated water vapor source are avoided. The invention provides a convenient, quick, pollution-free and safe measuring method for measuring the sampling efficiency of the bubbling tritium sampler and the sampler. The radioactive source does not need to be operated.

The invention relates to the technical field of engineering logging, and discloses a low-power-consumption radio frequency ion source logging neutron tube and a working method, deuterium gas is ionized into deuterium ions through an ion source, the deuterium ions are accelerated under the action of a strong electric field to obtain energy, and the high-energy deuterium ions bombard a deuterium / tritium target to generate nuclear reaction to generate neutrons. The power requirement of the radio frequency source can be greatly reduced. Compared with an inductive coupling type radio frequency ion source neutron tube, the capacitive coupling type neutron tube has the advantages that D1 + 1 in plasma is increased to 50%, the required power is 25%-30% of that of the inductive coupling type neutron tube, the load of a neutron generator is greatly reduced, the D-D deuterium neutron tube is further applied to engineering practice, and powerful technical support is provided for green and environment-friendly well logging.

To solve the problem that muon is supplemented to a generated substance after nuclear fusion, not to a nuclear fusion fuel substance, and muon catalyzed fusion reaction is unlikely to proceed because in a well-known muon-catalyzed fusion, muon is adhered to / captured / trapped by helium to be a generated substance after nuclear fusion, not to a fusion fuel substance such as hydrogen / heavy hydrogen D / tritium T, and catalyzed fusion reaction stops; and to propose a system which is unlikely to generate a neutron.SOLUTION: When nuclear fusion of an atomic nucleus occurs in muon catalyzed fusion, electric charges of atomic nucleus of a substance generated by nuclear fusion is used for a nuclear fusion reaction system which has a feature smaller than electric charges of an atomic nucleus of an atom to become nuclear fusion fuel. To be specific, a muon catalyzed fusion system using borohydride (diborane or the like) including proton as well as boron for nuclear fusion fuel is proposed.SELECTED DRAWING: Figure 8

The application discloses a fusion reactor solid-state blanket design method based on topology optimization, and is characterized in that based on the topology structure characteristics of the fusion reactor solid-state blanket arranged in layers along the radial direction, the coupling mechanism of the multi-physical fields of neutronics, thermal-hydraulics and structural mechanics is researched, a coupling model between various parameters of the multi-physical fields is established through theoretical analysis and a multivariate nonlinear regression method; under the constraint conditions of material temperature and structural mechanical properties, a constrained topology optimization model for maximizing the tritium breeding ratio is constructed; the constrained topology optimization model is converted into an unconstrained topology optimization model for maximizing the tritium breeding ratio by using a Lagrange function and a KKT condition, the unconstrained topology optimization model is solved by using a gradient descent algorithm, and a blanket radial topology structure capable of meeting the design criteria is obtained. A blanket three-dimensional model is established, and the performance of neutronics, thermal-hydraulics and structural mechanics in the whole reactor is analyzed, the tritium production, heat removal and structural safety performance of the blanket are evaluated, and it is verified that the topology optimization model can efficiently obtain a blanket design scheme with optimal comprehensive performance of multiple fields.

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.

A controlled fusion process is provided that can produce a sustained series of fusion reactions: a process that (i) uses a substantially higher reactant density of the deuterium and tritium gases by converging cationic reactants into the higher reaction density at a target cathode rather than relying on random collisions, the converging producing a substantially higher rate of fusion and energy production; (ii) uses a substantially lower input of energy to initiate the fusion; (iii) can be cycled at a substantially higher cycle frequency; (iv) has a practical heat exchange method; (v) is substantially less costly to manufacture, operate, and maintain; and, (vi) has a substantially improved reaction efficiency as a result of not mixing reactants with products.

The application provides an automatic sampling and measuring device for radioactive aerosol and gaseous tritium in radioactive gaseous effluent. The measuring device comprises a radioactive aerosol measuring module and a gaseous tritium measuring module integrated in the measuring device, and the radioactive aerosol measuring module and the gaseous tritium measuring module are communicated with a gaseous effluent pipeline through sampling pipelines. The measuring device of the application can greatly save labor, realize unattended and long-time online monitoring, flexibly adjust the sampling and monitoring frequency, and improve the monitoring efficiency. In addition, in the case of a nuclear accident, the emergency monitoring mode can be quickly started to quickly respond to the accident monitoring requirements, so as to realize continuous, online and automatic monitoring of the concentration of radioactive aerosol and gaseous tritium in the effluent of a nuclear facility.

The present invention discloses a system and method for measuring the deuterium-tritium fuel ratio in plasma, comprising an electron cyclotron radiometer, a microwave reflectometer, a Mirnov probe, electron cyclotron radiation imaging, a first collector, a central processing unit, an interferometer, a dynamic Stark effect spectrometer, a magnetic coil, a charge exchange spectrometer, a second collector, a high-performance computer, and a TAE frequency real-time processor. The deuterium-tritium fuel ratio can be obtained by accurately measuring the frequency of the toroidal Alfvén eigenmode in the burning plasma and accurately calculating the TAE frequency in the deuterium-deuterium plasma under the same conditions. Compared to the existing technology, even if the burning plasma is in an uneven distribution state, it can be measured as a whole, and the deuterium-tritium fuel ratio can be calculated based on the measured data, thereby providing a safe, accurate, and rapid method for measuring the deuterium-tritium fuel ratio for the heating effect of fusion product alpha particles, the high and low confinement mode conversion power thresholds, and the confinement time.

The present invention relates to the field of nuclear radiation detection technology, and provides a tritium measurement device with radon discrimination capabilities. The device comprises a gas flow ionization chamber, a current amplification circuit, a pulse amplification circuit, and a signal processing circuit. The collector of the gas flow ionization chamber is connected to the current amplification circuit, which amplifies and measures the current signal in the gas flow ionization chamber. The measured signal includes the current signals of tritium beta particles and radon alpha particles. The high-voltage electrode of the gas flow ionization chamber is connected to the pulse amplification circuit after passing through a DC-blocking capacitor, which measures the pulse signal generated by radon alpha particles. The device structure of the present invention is convenient to implement, comprising only a single detection device. It can measure and discriminate radon alpha particles entering the gas flow tritium ionization chamber in real time, accurately compensate for tritium measurement results, and prevent further increase in the total mean square error of the measurement results after compensation.

The invention discloses a neutron-gamma mixed radiation screening method based on a radiation response signal. The neutron-gamma mixed radiation screening method depends on a nano-coating CMOS sensor neutron measurement system. The neutron measurement system of the nano-coating CMOS sensor comprises a detector, the detector comprises a CMOS sensor; a LiF thin film is attached to the light-sensitive surface of the CMOS sensor; the method comprises the following steps that a part of neutrons enable the CMOS sensor to generate a radiation response signal A; the other part of neutrons and 6Li are subjected to nuclear reaction to generate alpha rays and tritium ions, so that the CMOS sensor respectively generates a radiation response signal B and a radiation response signal C; the gamma ray enables the CMOS sensor to generate a radiation response signal D; and according to the difference of the four radiation response signals in morphology and / or gray values, discriminating the neutron-gamma mixed radiation field. The detector formed by combining the LiF film and the CMOS sensor is used for discriminating a neutron-gamma mixed radiation field and can be used for monitoring nuclear leakage of a reactor.

The invention discloses an alloy for hydrogen isotope gas storage and a preparation method thereof. The chemical formula of the hydrogen isotope gas storage alloy is ZrMn2-x, wherein x is more than or equal to 0.25 and less than or equal to 0.5. Compared with the prior art, the alloy has the following beneficial effects: 1, the alloy prepared by the invention has lower hydrogen absorption / desorption isotope equilibrium pressure, and when the alloy is used for tritium gas storage, the quantity of tritium remaining in a gas phase in the storage process can be greatly reduced, so that helium generated by decay of tritium remaining in the gas phase is reduced to the greatest extent; and 2, the prepared alloy does not contain catalytic active elements, and generation of tritiated methane gas in the tritium release process of the tritium storage alloy is avoided, so that the chemical purity of tritium gas is improved. And 3, the main element of the alloy is the low-cost Mn element, so that the cost of the hydrogen isotope gas storage alloy can be greatly reduced.

The invention discloses a method for effectively reducing adsorption and accumulation of tritium on the wall of an ionization chamber. A non-metal hydrophobic coating is deposited on the surface of the wall of the ionization chamber. At present, the ionization chamber wall is mainly made of a metal material, tritium atoms enter metal lattices to accumulate, and therefore the tritium memory effect is caused, a layer of dense, uniform and firmly-combined hydrophobic coating is deposited on the chamber wall, and adsorption and accumulation of tritium on the chamber wall can be effectively reduced.

The invention discloses a method for producing radioactive isotopes in a reactor core of an advanced high-temperature reactor by adopting a spherical target piece. According to the invention, the limitation that the radioactive isotope can only be produced by using a neutron irradiation channel of a nuclear reactor or fuel-related components such as an adjusting rod of a heavy water reactor and a burnable poison rod of a pressurized water reactor is broken through, and a precedent of mass production of the radioactive isotope by using the reactor core of the nuclear reactor is created; the method has important strategic significance for exploring high-value application scenes of commercial advanced high-temperature reactors to reduce operation cost and producing radioactive isotopes including strategic isotope tritium in batches.

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.

A metal-organic framework coating may be employed in a molten salt reactor system to capture gaseous fission products produced as a consequence of operation. The metal-organic framework coating may include one or more metal-organic frameworks configured to capture certain fission products based on the need, for example tritium fluoride, tritium gas, and hydrogen fluoride. The one or more metal-organic frameworks may be immobilized onto a substrate and formed into a coating covering the substrate. In many cases, the substrate is the stainless-steel surfaces of reactor enclosures and vessels. In other cases, the substrate may be suspended into a headspace of a vessel of a salt-bearing component, or may be included in an off-gas processing system of a reactor system.

The application discloses a fusion reactor liquid blanket tritium breeding rate online regulation system simulation method and device, wherein the simulation device comprises a fusion blanket neutron transport module, a blanket breeding agent Monte Carlo-activation coupling burnup simulation module, a blanket coolant feeding module, a blanket tritium extraction module, a blanket coolant purification treatment module, a blanket tritium breeding rate calculation module and a blanket tritium breeding agent circulation system; the simulation device realizes the circulation of the blanket coolant nuclide based on the mass conservation equation and the burnup equation. The application is based on the Boltzmann neutron transport equation, the neutron flux of the blanket coolant is simulated by a Monte Carlo method, the time and space variation law of the chemical composition of the system material is obtained by solving the burnup equation, the replenishment demand of the blanket coolant system is calculated, and the influence of the replenishment composition on the blanket tritium breeding rate is simulated, so that the application can be widely applied to the fusion reactor liquid lithium lead blanket.

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 invention discloses a thermonuclear fusion reaction combustion product trapping method and system, and belongs to the field of nuclear fusion. The invention discloses a thermonuclear fusion reaction combustion product trapping method. The method comprises the following steps: S1, cooling a combustion product; s2, carrying out primary photolysis treatment on the cooled combustion product; s3, carrying out primary separation treatment on the combustion product subjected to the primary photolysis treatment; s4, carrying out first collection treatment on a first product obtained by the first separation treatment; s5, performing secondary photolysis treatment on the residual combustion product after the primary separation treatment; s6, carrying out second-time separation treatment on the residual combustion product after the second-time photolysis treatment; s7, collecting a second product obtained by the second separation treatment; and S8, residual combustion products obtained after the second-time separation treatment are collected and treated. According to the invention, mixed products with insufficient combustion of deuterium and tritium in the thermonuclear fusion reaction can be efficiently classified and collected.

The invention provides an automatic sampling and measuring method for gaseous tritium, which comprises the following steps: S1, sampling of tritium-containing gas: dissolving the tritium-containing gas in sampling liquid through a tritium-containing gas sampling unit to prepare tritium-containing liquid; s2, tritium-containing liquid sample preparation and sample feeding: tritium-containing liquid and scintillation liquid are proportionally injected into a measuring bottle through a tritium-containing liquid sample preparation unit, and then the moving position of the measuring bottle is controlled through a sample feeding operation unit; and S3, automatic measurement: measuring the flicker spectrum of the measurement bottle placed in an automatic measurement unit through the automatic measurement unit, and calculating the activity of tritium after measurement. According to the invention, automatic sampling, automatic sample preparation, automatic sample feeding and automatic measurement of tritium in air can be realized, automatic online monitoring of gaseous tritium is realized, and improvement of monitoring efficiency is facilitated; unattended operation and long-time full-automatic sampling and measurement of tritium in the atmospheric environment and gaseous effluent can be realized, and on-line monitoring of low-concentration tritium in air is realized; the system can be quickly started when a nuclear radiation accident occurs, and the response speed of nuclear emergency monitoring is improved.

The present invention relates to a high counting rate alpha particle detection method for accompanying particle imaging of a fusion neutron generator, which can effectively improve the measurement counting rate and timing accuracy of the accompanying particle imaging system of a fusion neutron source. The method comprises: by setting a barrier layer of a semiconductor detector, the thickness of a detection sensitive area, and thinning the thickness of an electrode and a substrate, the energy deposition of interference particles in a deuterium-tritium neutron source in the sensitive area of ​​the detector is reduced, and the proportion of energy deposition of accompanying alpha particles in the sensitive area of ​​the detector is increased, so that the amplitude of the current pulse signal output by the detector is proportional to the particle deposition energy; a current sensitive amplifier is used to amplify and reduce noise on the pulse signal output by the detector, and a narrow width voltage signal proportional to the current signal of the detector is output; a leading edge timing discriminator is used to time the pulse leading edge of the amplified voltage signal, and a reasonable trigger threshold is set to eliminate interference particle signals, so that only accompanying alpha particle timing signals are obtained.