55 results about "Neutron radiation" patented technology

The invention relates to the technical field of underground geotechnical engineering and tunnel construction monitoring, in particular to a shield tunneling rock stratum real-time judgment method. The method comprises the steps that according to signals of neutron radiation characteristics released when different lithologic samples are fractured under a shield tunneling stress path simulated in a laboratory, a knowledge base of the neutron radiation characteristics of a rock stratum is constructed; deploying an on-site multi-source data acquisition system, and acquiring data of neutron radiation characteristics and shield tunneling parameters; jointly forming a fusion feature vector according to the data of the neutron radiation features and the shield tunneling parameters; the method has the advantages that signals of neutron radiation characteristics released in the rock fracture process are directly monitored, the signals are fused with shield tunneling parameters in a multi-source mode, a machine learning model is combined, and the lithology classification probability and the rock mass integrity index are obtained. Real-time and accurate discrimination of lithology and integrity of a front rock stratum is realized, physical characteristics of a rock mass are reflected, and a discrimination effect is continuously optimized.

The invention relates to a method for separating heavy metal cations and radionuclides from aqueous solutions. The method according to the invention can be used to separate dissolved and / or suspended naturally occurring ions or radioactive ions produced by radioactive decay, nuclear transformation reactions such as nuclear fission, or, for example, by activation with neutron radiation, from aqueous solutions.

A radiation dose quantification method and system, the method comprising: detecting, with one or more detectors with respective sensitive volumes, gamma-rays emitted as a result of capture of neutrons by a composition in a subject subjected to an irradiation program, the composition comprising one or more thermal neutron capture agents, the neutrons having been generated by non-elastic collisions between a primary beam of particles and nuclei in the subject, wherein the particles consist of any one or more of protons, deuterons, tritons and heavy ions, the irradiation program comprises at least one period of irradiation with a beam duration that includes beam-on periods and beam-off periods; applying at least one predefined energy window or filter configured to accept only detection events in the one or more detectors resulting from gamma-rays with an energy indicative of selected gamma-rays arising from the capture of thermal neutrons by the one or more thermal neutron capture agents, wherein the thermal neutrons are neutrons with energies below approximately 0.4 eV; applying a timing window configured to reject or ignore detection events in the one or more detectors resulting from at least prompt gamma-rays produced in non-neutron capture events; and determining the radiation dose of neutron radiation received by the subject during the irradiation program from at least the accepted detection events or determining a dose map of the radiation received by the subject from at least the accepted detection events.

The present invention relates to a remotely operated system (200) to deploy and recover materials (25) into and from a nuclear fusion device (100), for the purpose of exposing said materials (25) to neutron radiation, which will cause the generation of radioactive isotopes, comprising at least one loop made of pipes (14) with constant inner diameter, a plurality of capsules (23) with an outer diameter slightly smaller than the inner diameter of the at least one loop (14), wherein the capsules (23) are adapted to contain the material (25) to be irradiated by the neutron radiation, a propulsion system for circulating the capsules (23) through the at least one loop (14), a remotely operated capsule filling system (18) for filling said capsules (23) into the at least one loop (14), and a remotely operated capsule recovery system (19) for recovering said capsules (23) from the at least one loop (14).

The invention relates to an under-voltage protection circuit suitable for a neutron radiation environment. The circuit comprises NMOS (N-channel Metal Oxide Semiconductor) tubes Q1 and Q2, a voltage-regulator tube D1, diodes D2 and D3 and resistors R1-R5, the anode of the voltage-regulator tube D1 is connected with the anode of the diode D2; the cathode of the diode D2 is connected with the first end of the resistor R1 and the first end of the resistor R2 at the same time. The second end of the resistor R2 is connected with the first end of the resistor R5 and the grid electrode of the NMOS tube Q1 at the same time. The drain electrode of the NMOS tube Q1 is connected with the first end of the resistor R3 and the grid electrode of the NMOS tube Q2; the second end of the resistor R5, the first end of the resistor R4, the cathode of the diode D3 and the drain electrode of the NMOS tube Q2 are connected; the cathode of the diode D1, the second end of the resistor R3 and the second end of the resistor R4 are connected to a power supply VCC; the second end of the resistor R1, the source electrode of the NMOS tube Q1 and the source electrode of the NMOS tube Q2 are connected to a reference ground end GND. The device is realized based on a localized VDMOS, and has the characteristics of neutron displacement radiation resistance, simple design, easy realization, low cost and the like.

An arylboron compound, its preparation method, applications, and pharmaceutical compositions are disclosed for tumor imaging. The compound of formula (I) exhibits high fluorescence quantum efficiency through intramolecular charge transfer. The compound of formula (I) synthesized in this invention can be linked to different antibodies via Q3 at certain concentrations, allowing for precise targeting of corresponding tumors based on the antibody. This type of boron drug formulation can be used in boron neutron capture therapy to kill tumor cells. It exhibits better killing effects; the compound of formula (I) targets tumor cells through antigen-antibody interaction. After the boron-10-containing drug accumulates at the tumor site, neutron radiation triggers a nuclear reaction releasing alpha particles and... 7 Lithium particles kill cells; the range of these two particles is approximately 10 μm. High-energy-density heavy ions disrupt the DNA double helix structure in tumor cells, causing irreparable cell death. Simultaneously, the 10 μm distance avoids damage to normal cells and tissues.

The utility model relates to a nuclear power station spent fuel transportation container shielding protection device. The device comprises a container body; the neutron shielding layer is arranged on the outer side of the container body in a surrounding mode, a filling cavity is defined by the neutron shielding layer and the outer side wall of the container body, and the filling cavity is filled with neutron shielding materials. By arranging the neutron shielding layer surrounding the outer side of the container body and the neutron shielding material filled in the neutron shielding layer, the neutron radiation shielding problem in the spent fuel transportation and assembly loading process is solved, the neutron radiation dose on the outer surface of the transportation container can be effectively reduced, a stable shielding structure is formed, and safe transportation of spent fuel is guaranteed.

The present application relates to cement-based building material technical field, especially to a kind of carbonized steel slag phosphogypsum light aggregate anti-radiation super high performance concrete and preparation method thereof, the present application uses cement as main cementitious material, adds composite admixture, utilizes steel slag and phosphogypsum to prepare light aggregate with core-shell structure, and cooperates with copper-plated steel fiber, and the anti-radiation super high performance concrete with good fluidity, high mechanical property, low shrinkage, strong neutron radiation resistance is prepared.

The invention discloses a neutron radiation shielding material and a preparation method thereof, the neutron radiation shielding material comprises a neutron radiation protection coating agent, and the neutron radiation protection coating agent comprises the following raw materials: polyvinyl alcohol, N, N-dihydroxyethyl-2-methacrylamide, di (carborane methoxy ester) phthalate, a dispersant, a cross-linking agent and an accelerant. According to the neutron radiation protection coating agent, N, N-dihydroxyethyl-2-methacrylamide is introduced and can be polymerized under the thermal action to form a Gel-PMA and PVA hydrogel system, a PVA-boric acid ester covalent network system is further constructed under the action of boric acid ester, and the neutron radiation protection coating agent is prepared. And a strong bisamide hydrogen bond interpenetrating network is generated under complete dispersion of phthalic acid di (carborane methoxyl ester) in a Gel-PMA polymer system, and the dual effects of mechanical property improvement and efficient neutron shielding effectiveness are synergistically realized.

The invention relates to an accelerator thermal neutron radiation field double-target beam automatic adjusting device and method, and the device is characterized in that a particle transmitting end is used for transmitting an ion beam; the particle receiving end is arranged on the emission axis of the ion beam; the particle receiving end comprises two semicircular beryllium targets and a tantalum target, the projections of the two semicircular beryllium targets form a complete circle on the ion beam emission axis, the tantalum target is arranged at the outermost side line of the two semicircular beryllium targets, and the area of the tantalum target is smaller than one third of the area of the two semicircular beryllium targets; and the xy guider is arranged between the particle transmitting end and the particle receiving end, a deflection electric field is arranged in the xy guider, and the xy guider is used for acquiring the quantity of ions received by the two semicircular beryllium targets and the quantity of ions received by the tantalum targets respectively, adjusting the electric field intensity of the deflection electric field based on the quantity of ions and changing the shooting direction of the ion beam. The technical effects that manual adjustment is not needed, the monitoring efficiency is improved, and meanwhile the experience degree needed by adjustment is reduced are achieved.

The embodiment of the invention provides a neutron source emissivity valuing device and a neutron source emissivity measuring method. The neutron source emissivity valuing device comprises a solution pool, a measuring chamber, a circulating pipeline, a shielding body and a detector. The solution tank is provided with a first accommodating cavity for accommodating an activating solution and a second accommodating cavity for accommodating a neutron radioactive source; the measuring chamber is provided with a third accommodating cavity and a measuring cavity; the shielding body is provided with a fourth accommodating cavity for accommodating the measuring chamber and a communication channel communicated with the outside, and the communication channel is communicated with the measuring cavity; the detector is arranged in the communication channel, and one end of the detector extends into the measuring cavity. According to the neutron source emissivity valuing device, the detector in the measurement chamber can be taken out through the communication channel, so that on one hand, the detector is simpler and more convenient to disassemble and assemble; on the other hand, the detector can be independently packaged and stored, and collision damage of the detector is effectively prevented.

The invention discloses a pulse neutron generation device for cutting steady-state neutron radiation and a pulse time measurement method. Comprising a steady-state neutron radiation source device, a shielding structure, a supporting frame, a collimator provided with a conical beam limiting diaphragm, a cutting sliding block made of neutron absorbing materials and internally provided with a penetrating beam limiting hole, a sliding block falling stabilizing device, a sliding block height adjusting device, a sliding block releasing device and a detector located on the emergent side. The cutting sliding block freely falls down along a vertical track limited by the falling stabilizing device after being released at a preset height, time cutting is carried out on a steady-state neutron beam passing through the collimator, a reference point is arranged on the supporting frame according to the pulse time measuring method, and a photoelectric probe A, a photoelectric probe B and a timer are arranged near a falling path; in the falling process of the cutting sliding block, the two photoelectric probes are sequentially shielded to trigger the timer to start / stop timing, and the time interval recorded by the timer is determined as the actual pulse width of pulsed neutron radiation.

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 belongs to the technical field of neutron radiation protection, and discloses a flexible neutron radiation protection material and a preparation method thereof. Comprising a high-density polyethylene / modified boron nitride composite fabric, and the surface of the composite fabric is coated with a polymer / boron powder composite coating. The preparation method comprises the following steps: (1) preparing modified boron nitride; (2) preparing a polyethylene / boron nitride blend; (3) spinning polyethylene / boron nitride fibers; (4) weaving the fabric; (5) preparing a coating dispersion liquid; and (6) fabric coating treatment. The shielding efficiency of the product prepared by the invention at a 5.3-wavelength neutron absorption position can reach 50.9-78%, and the protection requirement of neutron radiation is met; the surface coating of the composite fabric obtained after batch production is good in quality, and the composite fabric has good popularization significance.

The application discloses a bat moth lepidoptera cordyceps strain and application thereof in preparation of anti-neutron radiation medicines. Samsoniella hepiali The strain is the bat moth lepidoptera cordyceps strain P969 (Pleurotus spodoleucus), which is preserved in the China Center for Type Culture Collection and has a preservation number of CCTCC NO: M 20251172 and a preservation date of May 26, 2025. Mycelium extracts of the strain can protect biological tissues against neutron radiation and repair organ damage caused by neutron radiation, so that the strain has a wide development space and a good development and application prospect.

The invention relates to the technical field of neutron radiation measurement, discloses a neutron nondestructive testing wide-spectrum moderation device for a laser neutron source and a detection method, and adopts a three-in-one design idea of material function compounding, space structure nesting and energy step matching. An integrated structure formed by concentrically nesting an inner layer, a middle layer and an outer layer in sequence from inside to outside is adopted; the inner layer adopts a high-Z inelastic moderator, the middle layer adopts a low-absorption elastic moderator layer, and the outer layer adopts a thermalization and mechanical constraint shell; through physical combination and space structure optimization of a high atomic number material and a low atomic number material, step-by-step, collaborative and stepped efficient moderation of the neutron spectrum in the laser white light is realized. An integrated moderation structure formed by sequentially and concentrically nesting a high atomic number inelastic moderation layer, a low-absorption elastic moderation layer and a thermalization restraint layer from inside to outside is adopted, and step-by-step, collaborative and stepped efficient moderation and regulation and control of the wide energy spectrum of the laser white light neutron source are achieved.

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 invention relates to a steel plate concrete shielding workshop and reinforced concrete wall arrangement and connection node, and relates to the technical field of concrete shielding workshops, the steel plate concrete shielding workshop and reinforced concrete wall arrangement and connection node comprises a steel plate concrete assembly, a reinforced concrete wall assembly and a composite connection assembly, the steel plate concrete assembly is fixedly arranged on the outer side surface of the reinforced concrete wall assembly; the steel plate concrete assembly comprises an outer-layer steel plate and inner-layer radiation shielding concrete. The radiation shielding performance is excellent, 15%-20% of barite aggregate and borax are doped into the inner-layer radiation shielding concrete, the barite aggregate can effectively block penetration of gamma rays and the like, the borax can specifically absorb neutron radiation, a dual-radiation protection barrier is formed, and the core requirement of a shielding plant for radiation isolation is met. In the composite connecting assembly, the radiation shielding plug filled in the shear-resistant sleeve is tightly attached to the steel plate concrete assembly and the reinforced concrete wall assembly, and a radiation leakage gap at a connecting node can be filled.

The invention relates to the technical field of space positioning, in particular to a neutron radiation source positioning method based on a sensor array. The method comprises the following steps: acquiring array geometric data and scene structure data; background sub-field measurement is carried out according to the array geometric data and the scene structure data to obtain background sub-field data; performing target monitoring according to the array geometric data and the background sub-field data to obtain target monitoring data; performing neutron feature map construction according to the target monitoring data to obtain neutron feature map data; and performing neutron source position inversion according to the neutron feature map data to obtain neutron source position data. According to the invention, the method can achieve the positioning of the neutron radiation source in a structural scene with shielding, reflection and multi-path propagation, and effectively improves the recognition capability of weak signal interference and the robustness of space matching. Through a source position network model combining array geometry, structure visibility and time sequence characteristics, radiation source positioning reliability in a semi-closed environment is enhanced.

The application discloses a preparation method of SnBi / B4C radiation shielding composite material. The material selects SnBi alloy with high atomic number, large density, strong processability and corrosion resistance as a matrix, B4C particles with light weight, high strength, good stability and large neutron reaction cross section as reinforcing fillers, and is formed through a powder metallurgy process, so that the composite material with high strength and the ability of shielding thermal neutrons, gamma rays, X rays and alpha particles is obtained. Compared with traditional lead gamma ray shielding materials with single function, the SnBi / B4C radiation shielding composite material solves the problem that neutrons cannot be shielded, and the selected raw materials are non-toxic, so that the composite material is expected to replace lead and be used for radiation shielding in the medical field.

The invention relates to the technical field of signal processing of neutron detectors, in particular to a neutron pulse signal collecting and counting device and method. According to the technical scheme, the neutron detector comprises a front signal source, a low-noise signal amplifier, an analog-to-digital conversion module and a signal collecting and processing unit, and the front signal source adopts a microstructure semiconductor neutron detector and is used for receiving neutron radiation and generating charge pulse signals; and the low-noise signal amplifier is connected with the front signal source and is used for amplifying the charge pulse signal. A set of complete neutron signal acquisition and processing system is constructed by integrating a microstructure semiconductor neutron detector, a three-stage low-noise amplification circuit, a high-speed analog-to-digital conversion module based on double-ADC alternate sampling and a self-adaptive digital processing algorithm, so that high sensitivity, low-noise amplification and high-speed sampling are realized, and the detection accuracy is improved. The method has excellent pulse identification precision and real-time energy spectrum reconstruction capability.

The invention discloses a hepialus hepiali strain and application thereof in preparation of anti-neutron radiation drugs, the strain is hepialus hepiali strain P969 (Samsoniella hepiali) and is preserved in the China Center for Type Culture Collection, the preservation number is CCTCC NO: M 20251172, the preservation date is 2025-05-26, hypha extract of the strain can protect biological tissues against neutron radiation, and the strain can be used for preparing anti-neutron radiation drugs. And the method has a wide development space, and has a very good development and application prospect.

The application relates to a mix proportion design method of a high-temperature environment neutron shielding serpentine concrete, which comprises the following steps: S1, calculating a neutron shielding attenuation coefficient of the serpentine concrete; S2, calculating the relationship between the neutron shielding attenuation coefficient of the serpentine concrete and combined water content through experiments; S3, calculating the relationship between cementing materials and aggregate mixing amounts through the combined water content of the concrete; S4, calculating the relationship between the cementing materials and the aggregate through the compressive strength requirement of the concrete; and S5, obtaining a mix proportion calculation formula of the serpentine concrete in a high-temperature environment. Through the thickness requirement of the serpentine concrete in structural design, the neutron shielding requirement and the like, and in combination with relevant physical parameters, the mix proportion design method of the neutron radiation shielding serpentine concrete in a high-temperature environment is provided, and the method has important significance for the development of nuclear engineering.

A neutron flux detector comprising a source material and a solid state radiation detector is disclosed. The source material is configured to produce gamma photons greater than or equal to 6.8 MeV during neutron capture. The solid state radiation detector comprises a Schottky diode and an emitter layer comprising a Compton and photoelectron source material. The emitter layer is configured to receive the gamma photons produced by the source material. The emitter layer is spaced apart from the Schottky diode a distance such that a gap is defined between the emitter layer and the Schottky diode. The distance is selected such that only electrons produced by the emitter layer as a result of the emitter layer absorbing the gamma photons will contribute to a measured output signal of the Schottky diode.

The invention relates to a preparation method of a high-temperature-resistant strong-irradiation thermocouple wire for nuclear engineering, which adopts cold isostatic pressing and vacuum vertical sintering to improve the high-temperature strength of the thermocouple wire, and solves the surface quality problem of the thermocouple wire through precision rotary swaging and high-precision wire drawing. The thermocouple wire has the advantages of being good in high-temperature strength, neutron irradiation resistance and mechanical property, high in potential stability, free of drifting and high in temperature measurement precision.

The utility model provides a reactor structure for realizing a starting-free neutron source in subsequent circulation, which comprises a reactor body and a signal amplification device, and the signal amplification device is of a cavity shell structure and is arranged between a fuel assembly in the reactor and a neutron detector outside the reactor. And the neutron detector is located on a neutron radiation path between the fuel assembly and the neutron detector. According to the utility model, by utilizing the arranged signal amplification device, the thickness of a shielding water layer between the spent fuel assembly and the reactor external source range neutron detector is reduced during the subsequent circulating reactor core loading period, and the number of neutrons reaching the detector is increased, so that the reactor core can be loaded with the spent fuel assembly which does not reach the minimum burnup. The detector can obtain effective counting in the loading process without depending on a neutron source, so that the problem that the neutron fluence rate level in a reactor core is difficult to accurately measure by the out-of-core neutron detector under the condition that the signal-to-noise ratio is large enough is solved.

1. Name of the product in this design: Portable Neutron Personal Dosimeter. 2. Purpose of this design: For detecting neutron radiation levels. 3. The key design feature of this product is its shape. 4. The image or photograph that best illustrates the design's key points: a 3D model.

The present application belongs to the technical field of shielding radiation of photoelectric sensor, and specifically relates to a method for preprocessing a star sensor damaged by neutron radiation in nuclear radiation, which comprises the following steps: step 1, analyzing a photoelectric image damaged by neutron radiation to obtain defect distribution response characteristics; step 2, according to the relationship between the defect response characteristics and the gray scale, preliminarily processing the defects in the star sensor image; and step 3, reprocessing the defects with large deviation after compensation and correction, wherein the correctness of the defects after correction is judged through correlation due to the high correlation of the star sensor image. The method provided by the present application improves the neutron radiation resistance of equipment, improves the detection image, maintains the measurement accuracy, prolongs the service life of the equipment, increases the service life of the photoelectric measuring equipment by more than 20 times, and maintains the high accuracy of the equipment.

The invention relates to the technical field of capacitors, in particular to a film capacitor structure for a nuclear fusion device, which comprises a composite shell mechanism, an elliptic cylinder core mechanism is packaged in the composite shell mechanism, terminal shielding mechanisms are symmetrically arranged at two ends of the composite shell mechanism, and a safety protection mechanism is arranged outside the composite shell mechanism. The composite shell mechanism comprises an ellipsoid shell, and the ellipsoid shell is composed of two hemispherical shells which are symmetrical to each other. The composite shell mechanism is arranged, two titanium alloy hemispherical shells are welded to form a long ellipsoid shell, firm packaging with excellent pressure-bearing and vibration-resistant characteristics can be provided for the internal elliptic cylinder core mechanism, in addition, the lead-boron glass composite shielding layer on the inner wall of the mounting cavity can efficiently absorb high-flux neutron radiation, and the shielding effect is good. And an organic dielectric film in the elliptic cylinder core mechanism can be effectively prevented from being degraded, so that the long-term structural integrity and performance stability of the capacitor are ensured.

The utility model discloses a helmet for radiotherapy, which relates to the technical field of anti-radiation instruments and comprises a circular helmet, a cylindrical neck sleeve is fixedly mounted at the bottom of the circular helmet, sweat-absorbent antibacterial components are fixedly mounted on the inner sides of the circular helmet and the cylindrical neck sleeve, and anti-radiation components are fixedly mounted on the outer sides of the circular helmet and the cylindrical neck sleeve. The inner wall of the cylindrical neck sleeve is provided with a plurality of air holes which are circumferentially distributed, the inner walls of the air holes are fixedly provided with air pipes, and the bottom of the cylindrical neck sleeve is provided with a fixing mechanism. According to the utility model, through the arrangement of the fixing mechanism, the sweat-absorbing antibacterial assembly and the anti-radiation assembly, the circular helmet and the cylindrical neck sleeve are prevented from moving when being worn; the anti-bacterial layer can avoid cross infection when the round helmet and the cylindrical neck sleeve are worn, the first radiation protection layer and the second radiation protection layer can effectively resist high-energy radiation such as rays, gamma rays and neutron radiation of the sweat-absorbent anti-bacterial assembly, and radiation protection can be provided for the head and the neck of a patient during radiotherapy.