135 results about "Neutron irradiation" patented technology

The invention relates to a device and a method for measuring concrete volume change caused by long-term neutron irradiation, the device comprises a neutron emitter and a neutron receiver which are arranged in a laboratory in a matching manner, a test concrete test piece is fixed between the neutron emitter and the neutron receiver, and the test concrete test piece measures the temperature through a temperature detector; a contrast concrete test piece is arranged in the test chamber, a collection mold wraps the outer side of the contrast concrete test piece, and a condensation pipe, a water vapor collector and a high-precision electronic scale are arranged. The method comprises the following steps: 1, measuring the overall expansion rate of a concrete test piece through a strain measurer; 2, testing the volume change of the concrete test piece in the environment; 3, calculating the volume change rate of the concrete caused by neutron radiation by combining the thermal expansion coefficient of the concrete and the drying shrinkage of the concrete; and 4, the relationship between the volume expansion rate of the concrete and the neutron irradiation dose is tested, and the relationship between the volume change rate of the shielding concrete of the engineering structure and the time is speculated, so that the safety of the nuclear engineering structure is ensured.

The utility model discloses a low-temperature adsorption pump and a low-temperature adsorption array suitable for a fusion reactor. The low-temperature adsorption pump comprises a heat shielding baffle, a supercritical helium pipeline, a helium low-temperature plate, an adsorption bin, a high heat conduction plate and a gas baffle, the supercritical helium pipeline, the helium low-temperature plate, the adsorption bin, the high-thermal-conductivity plate and the gas baffle are sequentially arranged in the thermal baffle; and the adsorption bin is filled with an adsorbent. The adsorbent contained in the adsorption bin is multiple times that of a low-temperature adsorption plate with the same area, and particularly, the difference is larger when a low-temperature adsorption array is formed. Meanwhile, a single-stage heat shielding baffle can be used, enough high conductance is guaranteed, high pumping speed and large capacity can be achieved, the problem that activated carbon falls off in the operation process of the low-temperature adsorption pump is effectively solved, and the low-temperature adsorption pump can be used in the strong neutron irradiation environment such as a fusion device in the future.

The invention provides a preparation method of a silicon carbide substrate, the silicon carbide substrate and a semiconductor device. The preparation method comprises the following steps: preprocessing a silicon carbide substrate raw material; performing neutron irradiation treatment on the material; and the material is subjected to annealing treatment. The silicon carbide substrate prepared by the preparation method provided by the invention can have relatively high mechanical property, conductivity, uniformity and stability.

The invention discloses an equivalent shear force detection system for a nuclear fusion superconducting coil under a multi-element coupling working condition, and belongs to the field of conductor insulation development of the nuclear fusion superconducting coil. The system comprises a nuclear magnetic resonance scanner, a signal exchanger, a liquid helium cooling system, a superconducting coil immersed curing insulation composite material sample piece, a four-stage magnetic field, a bipolar magnetic field, a neutron irradiation probe, a liftable rotary mechanical arm and an AI large model processing server. Completing the detection of the equivalent shearing force of the superconducting coil insulating material under the working conditions of magnetic coupling, neutron irradiation and the like in the nuclear fusion reaction stage; the insulation material equivalent shear force detection system can detect the endogenous organization structure change of the insulation composite material of the superconducting coil in the nuclear fusion reaction stage, and can avoid the situation that the electric insulation composite material of the superconducting coil is forced to be interrupted in the nuclear fusion reaction due to fatigue damage caused by concentrated equivalent shear force.

The invention relates to a method for evaluating the crack initiation probability of an RPV surface crack under a transient working condition, and belongs to the field of safety analysis of nuclear pressure-bearing equipment. The method comprises the following steps: sampling a random variable and determining a sampled multivariate sample; calculating a stress intensity factor of the crack evaluation point caused by multiple loads under the transient condition; calculating a stress intensity factor of the crack evaluation point after the influence of the crack tip plastic zone is brought into the transient condition; calculating the reference temperature of the crack evaluation point after neutron irradiation embrittlement is considered under the transient condition; constructing a fracture toughness probability distribution model of the crack evaluation points under the transient condition; calculating the crack initiation probability of the crack at the current moment under the transient condition; judging whether the current moment is the last moment of the transient working condition, if so, ending the calculation process of the crack initiation probability under the transient working condition, and otherwise, updating the transient moment; and evaluating the crack initiation probability of the RPV surface cracks under all samples. According to the method, the crack initiation probability of the RPV surface crack under the transient working condition can be efficiently and accurately evaluated.

The invention relates to a preparation method of a molybdenum-niobium thermocouple wire, according to the method, dispersion strengthening and toughening are adopted, the problem of uniformity of molybdenum-niobium alloy components is solved through a method of combining high-energy ball milling, vacuum gradient sintering and electron beam melting, and the prepared molybdenum-niobium alloy thermocouple wire is high in thermoelectric stability and precision through drawing and surface treatment. The high-temperature-resistant strength and the neutron irradiation-resistant performance are good.

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.

The application belongs to the technical field of gallium nitride substrate preparation, and particularly relates to a preparation method of n-type gallium nitride self-supporting substrate and a heavily doped GaN substrate and application. The preparation method of the n-type gallium nitride self-supporting substrate comprises the following steps: (1) sequentially epitaxially growing a buffer layer and a gallium nitride thick film on an original substrate; (2) peeling off the gallium nitride thick film to obtain a gallium nitride peeled sheet; (3) according to a germanium atom theoretical doping concentration of the gallium nitride peeled sheet, performing neutron irradiation treatment on the gallium nitride peeled sheet to obtain a germanium-doped gallium nitride peeled sheet; and (4) performing annealing treatment on the germanium-doped gallium nitride peeled sheet to obtain an n-type gallium nitride self-supporting substrate. The n-type gallium nitride self-supporting substrate obtained by the preparation method has a carrier concentration of 5x10 19 / cm 3 Without exogenous silicon or germanium dopants, the n-type gallium nitride self-supporting substrate obtained by the preparation method can be doped after the gallium nitride peeled sheet is formed, and the doping amount and doping concentration of the germanium element can be accurately controlled, which is suitable for large-scale production.

The present application relates to the technical field of neutron irradiation, and provides a neutron generating device, comprising a graphite stack, a neutron source, a shielding shell and a plurality of movable plates, wherein the graphite stack forms a radiation cavity and a detection cavity, the neutron source is arranged in the radiation cavity, the neutron beam emitted by the neutron source is moderated by the graphite stack to form thermal neutrons, and a thermal neutron reference radiation field is generated, the shielding shell forms a accommodating space, the graphite stack is arranged in the accommodating space, the shielding shell is provided with a plurality of beam ports connected to the accommodating space, the plurality of movable plates are movably connected to the shielding shell to selectively open or cover the beam ports, wherein the thermal neutron reference radiation field forms an inner calibration area and an outer calibration area, and detector traceability is achieved by placing a detector in the inner calibration area or the outer calibration area.

The invention provides a large phased array device and a communication method. The device comprises a base band unit (BBU); the active antenna unit AAU comprises an AAU master controller and an AAU subarray; the BBU and the AAU master control are integrated on a single PCB (Printed Circuit Board); the BBU is in communication connection with the AAU master control through a PCB (Printed Circuit Board) microstrip line, and a communication interface between the BBU and the AAU master control is a CPRI (Common Public Radio Interface) or an eCPRI (Enhanced Common Public Radio Interface); the AAU master control and the AAU sub-arrays are in communication connection through a radio frequency coaxial cable. The base band unit BBU and the AAU master control are integrated on the single PCB, a base band and radio frequency physical layer fusion framework is achieved, the bandwidth upper limit is greatly increased, and a multi-beam large-bandwidth large-scale array is supported; the neutron irradiation resistance of the coaxial cable is far higher than that of optical fibers, the interface is simplified, and the reliability is multiplied; a low-cost commercial device can be selected as a chip high-speed interface of the AAU subarray.

The invention discloses a preparation method of an extremely-low-activation neutron irradiation micro-nano sample, which comprises the following steps: S1, according to activation characteristics of a selected material, calculating a micro-nano sample volume critical value which can be reduced to a low level without surface radioactivity of the selected material; s2, selecting a low-activation single crystal material as a single crystal substrate; s3, performing pit digging treatment on the single crystal substrate by using a focused ion beam technology to form one or more pits of which the volume is smaller than the volume critical value of the micro-nano sample; and S4, preparing one or more micro-nano samples matched with the pits by using a focused ion beam technology, and welding the micro-nano samples into the pits. The preparation method has the advantages that the activation level is reduced, the efficiency is high, the experiment safety is enhanced, and the application range is wide.

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.

The invention discloses an aluminum alloy for simulating a neutron irradiation transmutation silicon effect and a manufacturing method thereof, and belongs to the field of aluminum alloys. The manufacturing method of the aluminum alloy for simulating the neutron irradiation transmutation silicon effect comprises the following steps: determining the Si element content in the alloy according to the state of the aluminum alloy to be simulated, and preparing a supersaturated solid solution raw material with corresponding components, an aluminum alloy bar is prepared through a powder metallurgy process, and an aluminum alloy structure for simulating the generation of the transmutation silicon effect is obtained through heat treatment. According to the aluminum alloy preparation method, the simulated alloy can be prepared to effectively simulate a precipitated phase structure formed by transmutation reaction after the aluminum alloy is irradiated by neutrons for a long time, and the aluminum alloy preparation method plays an important role in researching the influence of irradiation transmutation products on the performance of the aluminum alloy; the current situation that irradiation performance evaluation of a metal material with a solid transmutation product is extremely dependent on neutron irradiation resources is relieved, the research efficiency is improved, and therefore the design level and the safety performance of an aluminum alloy structural part in a research reactor are improved.

The invention relates to the technical field of automatic parameter optimization, and particularly provides a neutron treatment plan parameter optimization method and a neutron treatment plan system. The neutron treatment plan parameter optimization method provided by the invention comprises geometric automatic optimization, physical automatic optimization, biological automatic optimization and prescription automatic optimization, and formulation of an optimal neutron treatment plan is completed through cascading of the four automatic optimizations. According to the method, the irradiation parameters meeting the prescription requirement can be automatically optimized and output based on numerous parameters such as neutron source distribution, relative positions of a target region and an organ at risk, central positions such as a plan, a neutron irradiation angle and a neutron irradiation distance, and the time for making the neutron irradiation plan for the first time is saved.

The invention belongs to the technical field of gallium oxide semiconductor materials, and particularly relates to an n-type high-carrier-concentration beta-Ga2O3 single crystal material and a preparation method and application thereof. The invention provides an n-type high-carrier-concentration beta-Ga2O3 single-crystal material and a preparation method for the n-type high-carrier-concentration beta-Ga2O3 single-crystal material, aiming at the technical bottleneck that a high-carrier-concentration gallium oxide single-crystal material cannot be prepared by a Czochralski method. The preparation method comprises the following steps: (1) preparing a low-carrier-concentration beta-Ga2O3 single-crystal material by adopting a melt method; (2) calculating the theoretical doping concentration of germanium atoms in the beta-Ga2O3 single crystal material; the method comprises the following steps: carrying out neutron irradiation treatment on a beta-Ga2O3 single crystal material with low carrier concentration to obtain a germanium-doped beta-Ga2O3 single crystal material; and (3) carrying out activation treatment on the germanium-doped beta-Ga2O3 single crystal material, so as to obtain the n-type beta-Ga2O3 single crystal material with high carrier concentration. The problem of spiral growth in the process of preparing the high-electron-concentration gallium oxide single crystal through a pulling method is solved, the process is simple, the carrier concentration of the prepared beta-Ga2O3 single crystal material is adjustable in the range of 1016-5 * 1020 / cm < 3 >, and industrial large-scale production can be achieved.

The invention relates to the technical field of nuclear reactors, and provides a monocrystalline silicon wafer neutron irradiation device which comprises an outer barrel, a plurality of inner barrels stacked in the axial direction are arranged in the outer barrel, a first supporting plate used for bearing the inner barrels is arranged at the bottom of the outer barrel, a tray cover is detachably connected to the top end of the outer barrel, and the tray cover abuts against the top end of the uppermost inner barrel. A plurality of monocrystalline silicon wafers are stacked in the inner barrel, a second supporting plate used for bearing the monocrystalline silicon wafers is arranged at the bottom of the inner barrel, an end ring used for being sleeved outside the top end of the adjacent inner barrel in a matched mode is arranged below the second supporting plate, and a clamping ring abutting against the inner wall of the outer barrel is limited between the end ring and the adjacent inner barrel below; through the grouping loading design of the outer barrels and the stacked inner barrels, monocrystalline silicon wafers are dispersed into the inner barrels and are borne by the corresponding second bearing plates, the monocrystalline silicon wafers cannot be transferred to the lower-layer inner barrels, the situation that the lower monocrystalline silicon wafers are damaged due to the fact that overall stacking is too high, and the damage risk of the monocrystalline silicon wafers is high, and the irradiation yield is insufficient is avoided.

The invention belongs to the technical field of semiconductor materials, and particularly relates to a preparation method of an n-type phosphorus-doped silicon carbide crystal. According to the preparation method of the n-type phosphorus-doped silicon carbide crystal, phosphorus atoms are obtained through reaction and conversion of silicon atoms in the silicon carbide crystal and neutron capture nucleuses, and the preparation method comprises the following steps that the initial electrical property of the silicon carbide crystal to be doped is measured; calculating the target doping concentration of phosphorus atoms according to the target resistivity, and judging whether the silicon carbide crystal is subjected to phosphorus atom doping or not; carrying out neutron irradiation treatment on the silicon carbide crystal to be doped by adopting a neutron source to convert silicon atoms into phosphorus atoms so as to realize doping of the phosphorus atoms; and activating the doped phosphorus atoms to obtain the n-type phosphorus-doped silicon carbide crystal. The method has the beneficial effects that endogenous silicon in the silicon carbide is accurately regulated and controlled to be converted into phosphorus through the neutron irradiation condition, so that the method for doping phosphorus in the silicon carbide is realized; the process is simple, the cost is low, and the crystal form of the silicon carbide crystal is not changed.

This invention discloses a cerium-doped InP-based single-photon avalanche detector for space communication and its fabrication method. The detector's epitaxial structure, from bottom to top, includes a substrate, an N-type buffer layer, an absorption layer, a band transition layer, an N-type charge layer, an intrinsic multiplication layer, an intrinsic InP layer, and a cerium-doped P-type contact layer. The core innovation lies in the dual radiation-resistant design of the Ce-doped P-type contact layer and the Ce-doped modified stacked passivation film. The fabrication method involves MOCVD epitaxial growth, spin-coating source diffusion to prepare the P-type contact layer, ICP etching and wet repair, magnetron sputtering doping, and electrode fabrication. This invention utilizes Ce's unique electronic structure to suppress radiation-induced defects and combines this with an SACM structure to optimize the electric field distribution, achieving low dark count rate, high detection sensitivity, and excellent neutron radiation resistance. The process is compatible with existing semiconductor production lines and is suitable for wafer-level mass production, meeting the high-reliability single-photon detection requirements of aerospace missions such as space laser communication and quantum key distribution.

The invention provides an irradiation-resistant material as well as a preparation method and application thereof. The radiation-resistant material comprises a base body and a gradient composite coating arranged on the surface of the base body, the base body is an isostatic pressing graphite material, the gradient composite coating comprises a transition layer and a functional layer, and the transition layer is made of beta-SiC whiskers; the material of the functional layer is a SiC-Si3N4 composite layer, and the Si3N4 phase content in the functional layer is gradually increased from the position close to the transition layer to the position close to the outer side face, so that the SiC-Si3N4 gradient composite layer with continuous components is formed. The material has excellent corrosion (fused salt / helium) resistance, neutron irradiation resistance, thermal shock resistance and scouring resistance, is simple in preparation process and low in cost, is particularly suitable for reactor core complex structural parts of fourth-generation nuclear reactors such as fused salt reactors and high-temperature gas cooled reactors, and has a wide industrial application prospect.

The application discloses a kind of for plant seed neutron irradiation 241 Am- 9 Be source dose distribution irradiation device mainly includes cylindrical shield, transport plate, 241 Am- 9 Be neutron source, spherical shielding layer, rubber stick and sample irradiation area;The neutron source will be accompanied by a large number of 59.5keV, 4.438MeV gamma ray production while spontaneous fission, by wrapping spherical lead layer around neutron source shielding gamma ray, so that the main plant seed radiation effect comes from neutron radiation, while the geometric center of neutron source is the spherical center, set up sample irradiation area, ensure the uniformity of plant seed absorbed dose in each area, and directly given without going through tedious measurement and calculation, with the characteristics of standardization dose distribution, can meet the scientific research needs of most legume plant seed neutron irradiation, and can be applied to other kinds of plant seed irradiation and neutron radiation agricultural production and mutation breeding.

Systems and methods for sorting metal scrap such as but not limited to aluminum scrap includes a transport system for transporting the metal scrap to a prompt gamma neutron activation analysis device and / or with a prompt gamma neutron activation analysis device integrated into the transport system. A method of sorting metal scrap may include receiving metal scrap of an aluminum alloy and determining composition information of the metal scrap based on gamma radiation emitted by the metal scrap as a result of neutron irradiation. The method includes assigning the metal scrap to a class of a plurality of predetermined classes of the aluminum alloy based on the determined composition information and sorting the metal scrap based on the assigned class.

The present invention discloses a method for improving the reverse breakdown voltage of a gallium oxide diode based on neutron radiation, which mainly solves the problem of the low breakdown voltage of the existing gallium oxide diode. The implementation scheme is as follows: Select a neutron source device, and set the neutron energy of the device to 0.1 MeV - 20 MeV and the dose to 1×10<supgt;13< / supgt> - 1×10<supgt;16< / supgt> n / cm<supgt;2< / supgt> according to the tolerance of gallium oxide material to neutron radiation; Horizontally and evenly attach multiple Ga<subgt;2< / subgt>O<subgt;3 diodes to the PCB board, and then vertically place the PCB board with the devices attached at the pore position of the neutron source radiation device; After checking that the setting parameters of the neutron source device are correct, turn on the power to irradiate the attached devices with neutrons until the neutron dose emitted by the neutron source reaches the preset value, then turn off the power and take out the Ga<subgt;2O<subgt;3 diodes. The present invention effectively improves the reverse breakdown voltage of the Ga<subgt;2O<subgt;3 diode and can be applied to high-voltage and high-power power electronic devices.

The invention provides a particle-reinforced copper-based composite shielding material as well as a preparation method and application thereof. The particle-reinforced copper-based composite shielding material is prepared from a copper matrix and an additive, the additive comprises gadolinium hexaboride and / or europium hexaboride; the mass of the additive is 0.05 wt%-5wt% of the mass of the particle reinforced copper-based composite shielding material. According to the invention, the additive with specific mass percentage is used, and the shielding capability of boron, gadolinium and / or europium on neutron irradiation is utilized, so that the irradiation damage of the copper-based material is avoided; moreover, the gadolinium hexaboride and / or europium hexaboride are / is stable in property and have a particle reinforcing effect, so that the mechanical property of the copper matrix is further improved.

The invention discloses a method for measuring temperature in a reactor based on SiC thermal diffusion coefficient change after neutron irradiation. The method comprises the following steps: S1, processing a SiC crystal material into a flaky sample; s2, packaging the SiC crystal sample and the to-be-irradiated material sample into an irradiation sample box, and then putting the irradiation sample box into a reactor for a neutron irradiation experiment; s3, after the neutron irradiation experiment is completed, taking out the irradiation sample box from the reactor, cutting and disintegrating the irradiation sample box, and taking out the irradiated SiC crystal sample; s4, putting the SiC crystal sample into thermal conductivity meter equipment, and measuring thermal diffusion coefficients of the SiC crystal sample at different temperatures; and S5, drawing a curve chart of the thermal diffusion coefficient of the SiC crystal sample after irradiation along with temperature change, wherein the inflection point of the curve is the irradiation temperature of the position of the SiC crystal sample in the reactor. The method has the beneficial effects that the neutron irradiation temperature in the reactor at the position of the SiC crystal can be efficiently and quickly obtained.

The invention relates to the technical field of functional materials, in particular to a preparation method and application of a fast neutron scintillation screen. The preparation method of the fast neutron scintillation screen comprises the following steps: (1) preparing a perovskite fast neutron scintillator material; and (2) preparing a scintillation screen by using a hydrogen-rich substrate packaging luminescent material, wherein S1, the hydrogen-rich substrate and a scintillator material are prepared: the scintillator material in the step (1) is dissolved in a solvent for ultrasonic treatment; meanwhile, taking a proper amount of hydrogen-rich matrix, heating and dissolving the hydrogen-rich matrix into a liquid state at high temperature, adding the solution subjected to ultrasonic treatment into the dissolved hydrogen-rich matrix, and stirring until the solvent is completely volatilized to obtain a mixed solution in which the scintillator material and the hydrogen-rich matrix are completely mixed; s2, the fast neutron scintillation screen is prepared, specifically, the mixed solution in the S1 is cast into a mold to be prepared into a thin film, the fast neutron scintillation screen can freely fall off after a moment, and the fast neutron scintillation screen is obtained. The three types of fast neutron scintillation screens prepared by the method have good luminous efficiency and spatial resolution under the irradiation of fast neutrons, and show huge potential in fast neutron imaging.

This invention discloses a high-boron-loading manganese-based boron drug for BNCT diagnosis and treatment, its preparation, and its application, belonging to the fields of biomedicine and nuclear medicine. The boron drug structure is shown in general formula 7, with manganese porphyrin as the core skeleton. A benzene ring side chain at the meso position of the porphyrin ring is connected to a boron-10-containing catechol borate ester group via an amide bond. All boron atoms are boron-10 isotopes. The preparation method includes: first constructing the manganese porphyrin skeleton through condensation, cyclization, and metal coordination reactions; then independently preparing boron source units; and finally obtaining the target product through amide bond coupling. This boron drug possesses high boron-10 loading, efficient blood-brain barrier penetration, and excellent MRI imaging performance, enabling real-time and precise monitoring of boron drug distribution in vivo. The quantitative correlation between imaging signals and boron concentration guides the timing of neutron irradiation, integrating MRI tracing and BNCT treatment, effectively solving the problems of low boron loading, inability to penetrate the blood-brain barrier, and difficulty in real-time monitoring of in vivo distribution in existing boron drugs.

The application discloses a numerical simulation method for predicting the bearing capacity of an irradiated concrete structure, and comprises the following steps: a neutron flux distribution model in a reactor is established by using a Monte Carlo method, and a three-dimensional neutron radiation field inside a concrete shielding structure is calculated; a random distribution aggregate-mortar micromechanics model is generated based on a Fuller grading algorithm, and an interference discrimination method is used to ensure that the aggregate space is not overlapped; a mapping relationship between neutron flux data and elastic modulus degradation is established, and neutron irradiation expansion and thermal expansion effects are coupled through an equivalent thermal expansion theory; node coordinate parameters are dynamically converted in finite element analysis, a neutron flux threshold is matched according to a preset space condition, and a corresponding elastic modulus degradation coefficient is returned; and comparison data of stress fields and displacement fields of the structure before and after irradiation are output. The application can accurately match the degradation requirements of complex geometric shapes through joint judgment of polar coordinates and three-dimensional space.

The present application relates to a kind of long-term neutron irradiation caused concrete volume change determination device and method, the device includes: the neutron emitter and neutron receiver of matching setting in laboratory interior, test concrete test piece is fixed between neutron emitter and neutron receiver, test concrete test piece is measured temperature by temperature detector;Test chamber is equipped with control concrete test piece, control concrete test piece outside is wrapped up collection mould, setting condenser tube, water vapor collector and high-precision electronic scale.The method: one, the overall expansion rate of concrete test piece is determined by strain measuring device;Two, the self volume change of test concrete test piece in environment;Three, the volume change rate of concrete caused by neutron radiation is calculated in combination with the thermal expansion coefficient of concrete and the drying shrinkage of concrete;Four, the volume change rate of engineering structure shielding concrete with time is inferred by the relationship between the volume expansion rate of experimental concrete and neutron irradiation dose, to ensure the safety of nuclear engineering structure.

The invention discloses an imaging device based on a hexagonal boron nitride neutron imaging plate and a preparation method thereof, and the device comprises a neutron source, a collimator, a neutron imaging plate, a first reflector and an imaging device, and also comprises an excitation source which is used for emitting laser; the neutron imaging plate is an h-BN imaging plate and comprises a substrate and an h-BN film covering the substrate. The imaging device is a CCD (Charge Coupled Device) fluorescence imaging device; thermal neutrons emitted by the neutron source irradiate a to-be-detected sample, part of the thermal neutrons penetrate through the to-be-detected sample and then irradiate the h-BN imaging plate, thermal neutron irradiation damage defects occur, photoluminescence is generated under excitation of laser emitted by the laser source, and the photoluminescence is reflected to the CCD fluorescence imaging device through the first reflector to complete imaging of the to-be-detected sample. According to the invention, neutron current detection can be realized according to the boron vacancy defect of h-10BN and by utilizing the photoluminescence characteristic of h-10BN, the gamma sensitivity is low, the device process is simple, and the neutron measurement cost is reduced.

The present invention discloses a nano-boron delivery agent and its preparation method and application. The nano-boron delivery agent h-10BN-PG nanoparticles provided by the present invention have the characteristics of being easy to prepare, non-toxic, good biocompatibility, high 10B content, good dispersibility and stability in physiological solutions, etc. The h-10BN-PG nanoparticles can inhibit protein adsorption and non-specific uptake by macrophages, and have good "stealth" characteristics, thus avoiding rapid clearance by the MPS, and their blood circulation time is up to 8.5 hours. The h-10BN-PG nanoparticles can accumulate efficiently in CT26 tumor tissues through the EPR effect and can be further endocytosed by tumor cells; the concentration of 10B in tumors can reach 8.8% ID / g or 102.1 ppm. BNCT mediated by h-10BN-PG nanoparticles can completely eradicate subcutaneous CT26 tumors with only one injection of nanoparticles and one neutron irradiation.