40 results about "Isotope" patented technology

The application provides a temperature calculation method for an isotope irradiation hole of a side reflection layer of a pebble bed type high-temperature gas cooled reactor, and the method comprises the following steps: screening and determining an envelope target type which causes the maximum temperature rise of the irradiation hole based on the material nuclear characteristics and the expected heat release rate of each type of isotope target arranged in the irradiation hole; establishing a local refined three-dimensional thermal calculation model based on the parameters of the envelope target; inputting the maximum heat release rate of the envelope target into the three-dimensional thermal calculation model as a volume heat source, running a thermal fluid analysis program to perform steady-state or transient thermal calculation, and obtaining temperature field distribution data of the irradiation hole and the surrounding area; performing post-processing analysis on the temperature field distribution data, extracting the temperature distribution curve of the key path, identifying the highest temperature point and its position; and comparing the highest temperature with the allowable working temperature threshold of the irradiation hole material to evaluate the temperature safety of the irradiation hole.

This invention provides an application of ion emitters based on covalent organic framework materials in thermal ionization mass spectrometry (TIEMS), specifically for improving the ionization efficiency of uranium during TIEMS measurements. Compared to conventional dual-band source modes in TIEMS, and ionization enhancement techniques such as graphite methods, benzene vapor carburizing methods, and graphene oxide methods, this invention significantly improves uranium ionization efficiency, facilitating the accurate measurement of ultra-trace and low-abundance uranium isotopes.

This invention provides an apparatus, system, and method for analyzing altered carbonate carbon isotopes. The apparatus includes: an enrichment component comprising at least one enrichment tube having an enrichment section; and a freezing component comprising a container filled with a refrigerant. The enrichment section is completely immersed in the refrigerant to freeze the target gas generated by sample decomposition within the enrichment section to a target concentration. Based on the technical solution of this invention, analysis of low-content altered carbonate carbon isotopes can be achieved; it can also simultaneously acquire multiple sets of samples required for altered carbonate carbon isotope analysis, saving time for detecting large quantities of altered carbonate carbon isotopes. Furthermore, the operation is simple, reducing analytical costs and providing a reliable technical means for the application of geochemical exploration technology.

The application provides a fuel and a method for increasing the neutron source intensity of a nuclear reactor, wherein the abundance of O in the fuel is 1% to 50% 18 O is 1% to 50%, and alpha particles and the enriched 18 O can significantly enhance the initial neutron source intensity of a nuclear reactor core in a pressurized water nuclear power plant, and help to improve the start-up performance and reduce the dependence on external neutron sources. In the passive start-up mode without using an external neutron source, the application improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant. 18 O in the fuel, and improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant.

An X-ray scanning system includes: a housing configured to be handheld; an X-ray source for scanning a target; a set of first detectors configured to be mechanically coupled to the housing and to detect resulting X-rays received from the target responsive to scanning the target; a second detector configured to be mechanically coupled to the housing; and a set of processors mechanically coupled to the housing and communicatively linked to the first detector and to the second detector so as to receive first and second signals, respectively, therefrom. The processor(s) are configured to: generate an image of the target based on the first signal; distinguish, in the second signal, a contribution of passive nuclear radiation detected at the second detector; and provide an output corresponding to the contribution of passive nuclear radiation. A related system for identifying a radioactive isotope, along with corresponding methods, are provided.

A target support for isotope production includes a target cylinder with an inner flow channel and an end cap. The inner wall of the end of the target cylinder has a radially inner boss. The end cap has a fixed locking foot and a rotating locking foot, which can respectively engage with the inner boss. The fixed locking foot of the end cap is disposed on a cover plate of a fixed part, and the rotating locking foot is disposed on a rotating chuck. The rotating chuck can rotate relative to the cover plate to form a free position or a locked position. In the free position, axial rotation of the end cap ensures that both the fixed locking foot and the rotating locking foot are located within the gap of the inner boss. In the locked position, axial rotation of the end cap ensures that at least some of the fixed locking foot and / or the rotating locking foot are always engaged with the inner boss. This target support for isotope production has a simple structure, high reliability, and can be reused for a long time, reducing isotope production costs and decreasing the generation of radioactive waste.

This invention discloses a gas purification device for neutron-activated argon isotope dating based on a D-D accelerator, belonging to the field of rare gas isotope mass spectrometry analysis and geological dating technology. Addressing the limitations of traditional gas purification systems, which suffer from large internal volumes and high static release rates, making them unsuitable for the efficient purification and enrichment of trace argon isotopes generated by the D-D accelerator neutron activation method, this invention employs an integrated 316L stainless steel substrate, internally integrating a main gas flow channel and multiple branch channels. The substrate houses a sample gas injection module, a primary purification module, a secondary purification module, a tertiary purification module, a vacuum preparation and monitoring module, a standard gas injection module, and a gas enrichment module. Each module is connected to the main gas flow channel via branch channels, achieving multi-stage purification of the mixed gas and low-temperature enrichment of trace argon isotopes. This device is used for gas purification in D-D accelerator neutron-activated argon isotope dating, meeting the requirements for efficient purification and enrichment of extremely small amounts of argon gas.

This invention belongs to the field of nuclear radiation detection and nanomaterials technology, specifically relating to a submicron-resolution boron-10 isotope-enriched boron nitride neutron scintillator. Addressing the technical bottleneck of severe light scattering in existing physically hybrid scintillators, this invention innovatively employs anodized aluminum nanopore templates as a confined reaction vessel and optical waveguide structure. A precursor solution enriched with boron-10 isotopes is filled into the template pores, and a hexagonal boron nitride scintillator with a nanoarray structure and boron-10 isotope enrichment is generated through in-situ reaction at high temperature. This structure strictly confines the generation and propagation of scintillating light within independent nanopillars, fundamentally suppressing transverse light scattering by utilizing the waveguide effect of the pore walls. Thus, while ensuring the macroscopic thickness of the scintillator and high neutron absorption efficiency, it achieves micron- to submicron-level spatial resolution.

A direct fusion heat exchanger technology allows for the efficient transfer of fusion plasma heat energy to a working fluid for aerospace propulsion and other applications. Aspects of this disclosure also enable the generation of high neutron fluxes for applications like medical isotope production, materials testing, and neutron radiography. The use of a thin-film beam window array allows charged particles from a fusion plasma or accelerator to pass through and directly interact and heat the working fluid or fusion target. The thin-film beam window array offers advantages in efficiency, compactness, and safety compared to traditional fusion reactor designs. The technology also addresses the limitations of existing neutron sources by providing a simpler, cheaper, and more efficient method for generating high neutron fluxes.

The utility model relates to radioactive isotope marking transportation device with prevent leaking structure, including transportation base, one end of transportation base is connected with the device shell, the top movable joint of device shell is provided with device top, device top bottom is provided with sealing strip, the inboard fixed connection of device shell has the isolation chamber, the inboard of isolation chamber is provided with prevent leaking container, the outside of prevent leaking container is provided with limit stop, limit stop bottom is provided with transmission assembly. The novel through the rotation knob, knob cooperation transmission gear A and transmission ring and and transmission gear B drive the limit stop of isolation chamber top hold prevent leaking container, prevent in the transportation process prevent leaking container to pour, and when device top is closed, the sealing strip of device top bottom surface cooperates with sealed baffle, thereby improved the sealing of device, ensure device inside and outside environment complete isolation, cause the radioactive material leakage of prevent leaking container inner wall.

The present application relates to a kind of based on EA-IRMS measurement sample δ 13 C method, comprising the following steps: dilution is carried out to the sample to be measured to obtain sample aqueous solution, and a predetermined volume of sample aqueous solution is placed in a vessel to carry out drying treatment;The δ 13 C value of the sample after pretreatment is measured by using element analysis-isotope mass spectrometer instrument, and a plurality of δ 13 C measurement value is obtained after calibration by standard curve;The plurality of δ 13 C measurement value is drift corrected based on the instrument drift function determined in advance, and a plurality of corrected sample δ 13 C value is obtained, and the final sample δ 13 C value is obtained after averaging. The present application can ensure that the mass deviation of the sample is smaller by using dissolution technology to prepare the sample, so as to improve the measurement accuracy of the final sample δ 13 C value;The present application can further determine the minimum sample peak value and drift correct the EA-IRMS measurement, to improve the result accuracy.

The present application relates to a kind of medical isotope production base reactor planning limit area determination method, with the center of single reactor pressure vessel in multiple reactor as origin, the outer boundary radius of multiple reactor planning limit area is determined;Within the outer boundary radius of planning limit area, the accident analysis is carried out to multiple reactor, and the source term reference value of each reactor is adjusted based on inherent safety correction coefficient;Using linear superposition mode, the source term of multiple reactor is superposed, and the superposition source term is obtained;According to superposition source term, the corresponding atmospheric dispersion state of multiple reactor is determined, and multiple-dimensional consequence evaluation index is obtained;Based on multiple-dimensional consequence evaluation index, the outer boundary radius of multiple reactor planning limit area is iteratively adjusted until multiple-dimensional consequence evaluation index meets preset threshold, and the boundary of planning limit area is determined.The present application realizes scientific quantification of radiation influence under normal operation and accident condition, and solves the problem that traditional method has poor adaptability and insufficient evaluation accuracy.

Embodiments of the present disclosure provide an image-guided radiation therapy system and a method thereof. The system includes a first detection assembly, at least one imaging source, and a radiation source. The first detection assembly may be configured to determine a first image of a target region of an object by receiving a photon signal generated by a radioactive isotope within the object. The at least one imaging source may be configured to determine a second image by emitting an imaging beam to the target region of the object. The radiation source may be configured to emit a therapeutic beam to the target region of the object. The first detection assembly and the at least one imaging source may be mounted on a first rotatable part, the radiation source may be mounted on a second rotatable part, and the first rotatable part and the second rotatable part may be coaxially and coplanarly arranged. The first rotatable part may be configured to be rotatable independently relative to the second rotatable part.

This invention provides a calibration device and method for an isotope level gauge. The calibration device includes a ground rail with a pylon mounted on it. A lifting calibration component is mounted on the cantilever of the pylon. The lifting calibration component is used to hook and suspend a calibration block and control its vertical movement. By collecting the measurement values ​​of the level gauge sensor when the calibration block is in different positions, the level gauge is calibrated. The lifting calibration component hooks and suspends the calibration block through a rigid connection, avoiding swaying during vertical movement. Whether in manual or automatic mode, the operator can complete all operations through the control panel without approaching the crystallizer, thus avoiding radiation damage to the operator and significantly improving work safety. Furthermore, the ground rail is parallel to the arrangement direction of the crystallizer, and a shielding wall is provided between the ground rail and the crystallizer to shield the radiation from the level gauge's radiation source, further reducing the safety risks to personnel working behind it.

The present disclosure provides an isotope irradiation safety analysis method and device, electronic equipment and storage medium, relating to the technical field of nuclear power generation, by comprehensively considering the irradiation requirements of target isotopes, side reflection layer hole configuration parameters, and material properties and structural parameters of irradiation targets, two types of calculation models corresponding to unloaded and loaded irradiation target states are constructed to obtain baseline and perturbed core power distribution, and then the degree of core power distortion is quantitatively characterized, and based on the comparison results, the irradiation target design parameters and / or hole configuration parameters are adjusted to iteratively optimize the irradiation scheme, so that the problem that the prior art lacks a systematic analysis method of comprehensive key factors, can only qualitatively describe or simply estimate the core power distortion, and cannot meet the reactor safety evaluation requirements, can be solved, and the technical effects of accurately quantifying the degree of core power distortion and ensuring that the isotope irradiation scheme meets the safety requirements of the reactor are achieved.

This invention relates to the field of isotope source technology and discloses an isotope source carrier, an isotope source, and a method for preparing the isotope source. The isotope source carrier includes a preset curvature shape memory layer and a loading layer. The preset curvature shape memory layer has a first shape state at a first preset temperature and a second shape state at a second preset temperature. The preset curvature shape memory layer has a first surface recessed in the thickness direction of the preset curvature shape memory layer in the first shape state. The loading layer is disposed on the preset curvature shape memory layer, and the shape of the loading layer is adapted to the shape of the preset curvature shape memory layer. This isotope source carrier can overcome the technical problem of how to ensure the flatness of the isotope source. During the loading of isotopes, the loading layer will generate a certain growth stress. The preset curvature shape memory layer can offset the growth stress of the loading layer by setting a specific curvature radius, thereby obtaining an isotope source of a predetermined shape.

The application discloses a kind of strong current proton cyclotron production 225 Ac- 22 Na- 68 Method for producing Ge isotopes, comprising the following steps: selecting a plurality of target nuclides capable of fully utilizing the accelerator energy interval; each nuclide produces an energy interval corresponding to the production of two target nuclides, and the reaction cross-section peak region of one of the target nuclides corresponds to the reaction cross-section valley region of the other target nuclide; selecting target materials and shells for producing different target nuclides in different nuclide production energy intervals, in the same nuclide production energy region, the reaction cross-section peak region of the target nuclide produced by the target material corresponds to the reaction cross-section valley region of the target nuclide produced by the shell; or in the same nuclide production energy region, the reaction cross-section peak region of the target nuclide produced by the shell corresponds to the reaction cross-section valley region of the target nuclide produced by the target material; The present application converts the original radioactive waste into valuable products, reduces the production and processing cost of radioactive waste, and realizes the recycling of resources.

A target support for isotope production includes a target cylinder, two end covers and two locking rings, the end cover includes a through hole, a plurality of insertion plates which can be inserted and clamped in the target cylinder are arranged around the through hole, and the wall surface on one side of the insertion plate is integrated with the wall surface of the through hole, and the surface of the insertion plate is provided with a clamping part protruding outward; the target cylinder has a clamping structure matched with the clamping part; the locking ring is in interference fit with the through hole and the insertion plate, and can be locked by pressing the surface of the insertion plate from the inside after the clamping part and the clamping structure are clamped. The target support for isotope production can be recycled after the completion of isotope production in one batch, effectively reducing the material cost and reducing the generation of radioactive waste.

A target used for heavy water reactor production of C-14 isotopes, which comprises: a target tube, which has a target material (11) and / or an absorber material (14), the target material (11) being a nitrogen-containing solid material; end plates (2), which are arranged at two ends of the target tube; a connection rod (3), which is provided with a connection component that is connected to a positioning mechanism used for positioning the target, wherein the connection rod (3) passes through the target tube along the axial direction of the target tube, and the target tube is fixedly connected to the connection rod (3) by means of the end plates(2). The present invention can be adapted to be placed at different locations such as at a guide tube, an inspection passage, or a fuel channel to undergo irradiation and produce C-14 isotopes.

This application provides a modeling method, a modeling apparatus, an electronic device, and a storage medium; the modeling method includes: constructing a fourth model based on a first model, a second model, and a third model; the fourth model is used to predict the lifetime of an isotope power source; wherein, the first model characterizes the effect of device temperature difference decay on the lifetime of the isotope power source; the second model characterizes the effect of device material irradiation damage decay on the lifetime of the isotope power source; and the third model characterizes the effect of device material thermal aging on the lifetime of the isotope power source.

The present disclosure provides a target quantity optimization method and device for irradiation, electronic equipment and storage medium, relating to the technical field of nuclear power generation, by constructing a physical model of target radioisotope production according to the neutron characteristics of the target nuclear reaction and the neutron field parameters of the reactor, then adjusting the target quantity to obtain isotope yields under different quantities and combining market value parameters to generate correlation data between target quantity and irradiation economic benefits, and then using the correlation data to train a prediction model for the mapping relationship between target quantity and irradiation economic benefits, and finally determining the target target quantity that maximizes the benefits according to the model output, so that the problem of not establishing a quantitative mapping relationship between target quantity and irradiation economic benefits and relying only on experience to determine the target quantity in the prior art can be solved, achieving the technical effects of accurately determining the optimal target quantity, maximizing the economic benefits of radioisotope irradiation, and reducing the waste of target material and the generation of radioactive waste.

The application provides an inorganic scintillation crystal performance test system and method based on an adjustable pinhole collimator, and solves the problem that existing inorganic scintillation crystal performance test systems are mostly limited to the research of specific crystals and radioactive sources and are not universal in radiation detection macroscopic applications. The system comprises a base, a radioactive source, a radioactive source limiting unit, a collimation assembly, a collimation adjusting unit, a detector position adjusting unit and a rear-end test unit. Through the use of the system core structure, the collimation assembly composed of multiple pinhole collimators, the radioactive source radiation is accurately injected into the scintillation crystal at a certain angle, reducing the interference caused by scattering and environmental background, and combining standard scintillation crystals and standard radioactive sources of the same kind, the activity and energy consistency of the radioactive source is calibrated with high precision and sensitivity, and then flexible matching of various types of radioactive sources is realized, and the objective conditions of the numerous radiation isotopes in the application environment are restored as much as possible.

This invention belongs to the field of neutron scintillation screen technology, specifically relating to a meter-scale ultra-large boron-10 isotope-enriched boron nitride neutron scintillation screen. The invention mixes boron-10 isotope-enriched, luminescent boron nitride, a binder, and a solvent to form a homogeneous slurry; then, the slurry is coated onto a substrate using a blade coating or spray coating method to form a wet film; finally, after curing, a meter-scale neutron scintillation screen is obtained. This invention achieves integrated neutron absorber and luminescent center at the atomic scale, fundamentally solving the problem of non-radiative recombination loss caused by energy interface transport in traditional physical hybrid scintillators, while overcoming the limitations of chemical vapor deposition methods. 10 The fabricated scintillation screen faces challenges such as difficulty in enriching boron, substrate dependence, difficulty in large-area fabrication, and high toxicity of raw materials. It exhibits high neutron detection efficiency, high light yield, and excellent optical uniformity, providing an advanced material solution for the large-scale application of thermal neutron detection.

An anti-proliferation technique is disclosed to reduce the likelihood of nuclear proliferation due to the use fissionable fuel salts. The technique includes doping the fuel salt with one or more elements (referred to herein as activation dopants) that, upon exposure to neutrons such as would occur in the fuel salt when a reactor is in operation, undergo a nuclear reaction to, directly or indirectly, form highly active “protecting isotopes” (of the same element as the activation dopant or a different element). A sufficient mass of activation dopants is used so that the Figure of Merit (FOM) of the fuel salt is decreased to below 1.0 within some target number of days of fission. This allows the FOM of the fuel salt to be controlled so that the fuel becomes too dangerous to handle before to the creation of a significant amount of weaponizable isotopes.

There is provided a method of analysing isotopes of an element present in a compound of interest from a mass spectrum obtained by a Fourier transform mass spectrometer. The method comprises generating

A system (100) having an analysis device (1) that analyzes a component contained in a gas (9) is provided. The analysis device has a filter (25) that selectively passes the component contained in the gas, a detector (26) that detects the component that has passed through the filter, and a first control device (36) configured so that the detector (26) detects a low-abundance isotope 21Ne that is a main component, i.e., neon, contained in the gas and does not detect a high-abundance isotope 20Ne.

This application discloses a method for verifying the neutron absorption performance of boron stainless steel based on criticality analysis, comprising: obtaining the chemical composition, key isotope abundance, and material density of boron stainless steel with the highest and lowest boron content, respectively; calculating the minimum thickness of boron stainless steel to meet the critical safety acceptance criteria through criticality analysis; processing the boron stainless steel according to the minimum thickness to obtain comparative samples; measuring the neutron absorbance data of each comparative sample using a neutron transmission device; obtaining the acceptance limit based on the neutron absorbance data; measuring the neutron absorbance of the sample to be verified using a neutron transmission device; if the measurement result is greater than or equal to the acceptance limit, it is qualified; otherwise, it is unqualified. This application realizes a standardized acceptance process with measured neutron absorbance data as the sole criterion, which can avoid the use of high-precision standard plates, significantly improve detection efficiency and judgment accuracy, and ensure the safety and reliability of nuclear materials applications.