72 results about "Gamma detection" patented technology

Dual modality detection devices and methods are provided for detecting nuclear material, the devices include a neutron detector including multiple neutron detection modules; and a gamma detector including multiple gamma detection modules, where the multiple neutron detection modules and the multiple gamma detection modules are integrated together in a single unit to detect simultaneously both gamma rays and neutrons.

The invention discloses a neutron-gamma ray detection device which belongs to the detection technology field. The detector adopts a conventional plastic scintillator or an inorganic crystal to detect gamma rays in the environment and adopts a layer of neutron absorber on the surface of the plastic scintillator or the inorganic crystal to detect neutrons; the detector can be separately used as a neutron detector or be separately used as the gamma detector or be used as the detector for simultaneously detecting the neutrons and the gamma. The neutron-gamma ray detection device can simultaneously detect and monitor the neutrons and/or gamma (X rays) in the environment and the gamma detector can be used as a moderator of the neutron detector, thereby greatly improving the neutron detection efficiency. Working substances of the whole detector are all solids, thereby facilitating the production, the installation, the transportation and the long-term outdoor works.

The present disclosure provides systems and methods for crystal growth of cadmium zinc tellurium (CZT) and cadmium tellurium (CdTe) crystals with an inverted growth reactor chamber. The inverted growth reactor chamber enables growth of single, large, high purity CZT and CdTe crystals that can be used, for example, in X-ray and gamma detection, substrates for infrared detectors, or the like. The inverted growth reactor chamber enables reductions in the presence of Te inclusions, which are recognized as an important limiting factor in using CZT or CdTe as radiation detectors. The inverted growth reactor chamber can be utilized with existing crystal growth techniques such as the Bridgman crystal growth mechanism and the like. In an exemplary embodiment, the inverted growth reactor chamber is a U-shaped ampoule.

The present invention discloses a multiple-nuclide gamma imaging system and method and belongs to the field of medical imaging. The multiple-nuclide gamma imaging system comprises a multiple-nuclide tracer injection module, a multi-nuclide scintillation crystal detector module, a configurable circuit module and an image reconstruction and imaging module. The multiple-nuclide gamma imaging method comprises the following steps: S1, starting the imaging system, setting acquisition time of the imaging system, and injecting 11C, 13N, 15O and 18F labeled compounds into an imaging object; S2, obtaining a decay event emission gamma pulse data set by positioning a gamma detector; S3, outputting an electrical signal (scintillation pulse) by a silicon photomultiplier; S4, conducting time discrimination, energy discrimination, data acquisition and coincidence processing by a detector; and S5, displaying images in a tomographic manner by a computer. The disclosed multiple-nuclide gamma imaging system and method can obtain more comprehensive data than the prior art, comprehensively analyze multi-molecule events and related influences, reduce requirements on the total gamma photon counting, reduce irradiation risks on organisms and improve a signal-to-noise ratio of reconstructed images.

The invention discloses a method for measuring neutron, gamma energy spectrum and dosage in a nuclear power station. The method comprises the following steps: 1 ) the neutron and gamma rays are inputinto a neutron detection system to output a signal A, and the gamma rays are input into a gamma detection system to output a signal B; 2) the signal A and the signal B are input into a digital signalprocessing circuit to be subjected to operation processing to obtain the neutron, gamma energy spectrum and dosage, wherein the digital signal processing circuit takes a field-programmable gate arraychip as a core chip. Compared with the prior art, the method for measuring neutron, gamma energy spectrum and dosage in the nuclear power station disclosed by the invention has the advantages that thesimulations measurement of the neutron, gamma energy spectrum and dosage can be realized based on the field-programmable gate array technology. In addition, the invention further discloses a system for measuring the neutron, gamma energy spectrum and dosage in nuclear power station.

The invention belongs to a radiation dose measurement technique, and particularly relates to a personal neutron dosimeter with energy compensation. The personal neutron dosimeter comprises a casing, a gamma detection element and a solid track detector, wherein the casing is made of a boron-containing polyethylene material; an air cavity is formed in the casing made of the boron-containing polyethylene material; and the solid track detector is arranged in the air cavity. The personal neutron dosimeter ensures that the energy compensation effect is realized by using two conversion bodies which are the air layer and the boron-containing material.

The invention discloses highly stable small type Gamma detection system and gain stabilization method based on temperature correction. In the system, the temperature of a detector is measured by usinga digital temperature sensor and a thermal insulation layer to improve the accuracy of temperature measurement at different environmental temperature rates, especially at the temperature abrupt change; adjusting a FIR filter and forming parameters in real time according to the monitored temperature change, and reducing the influence of the noise characteristic change caused by temperature on theenergy resolution; a gain stabilization module is realized directly in the FPGA by using the pulse amplitude real-time compensation technology. The measures such as external high-voltage regulation oramplifier gain adjustment and the like are not needed. The use of a digital-to-analog converter is avoided. The system complexity is effectively reduced. The digital signal processing technology simultaneously realizes the energy spectrum measurement and the gain stabilization system. The system noise suppression parameter is optimized by utilizing the temperature feedback method. For miniaturized Gammaradioactive monitoring device which is required to operate in a rapidly temperature changing environment, the system has good application prospect.

The invention discloses a double monoclonal-antibodies sandwich ELISA method for detecting ChIFN-gamma, wherein the specific monoclonal antibodies (mAbs) against ChIFN-gamma are used to build the double monoclonal-antibodies sandwich ELISA method for detecting ChIFN-gamma. The double monoclonal-antibodies sandwich ELISA method for detecting ChIFN-gamma, with mAb3E5 as capture antibody and biotin-labeled mAb 3E3 as detection antibody, can quickly detect the trace ChIFN-gamma (125-500pg/ml). The method is rapid, useful, sensitive, and effective, can be used to detect the ChIFN-gamma and recombinant ChIFN-gamma in biological sample, and provides a sensitive detection tool for detecting and evaluating the cell immune response of poultry, and is useful for the research on the function of the ChIFN-gamma in various immune response mechanisms of poultry, and efficiently makes up for the prior ChIFN-gamma detection.

The invention discloses a display screen detection method and device, and belongs to the technical field of display. The method comprises that a light display parameter of a display screen to be detected is adjusted via an automatic gamma technology; a frame compensation parameter of the display screen is determined via a gamma detection technology, and a frame compensation function of the displayscreen is switched off in the automatic gamma technology and gamma detection technology; when the display screen needs frame compensation, the frame compensation function of the display screen is switched on, and the frame compensation parameter is burned in a specific position of the display screen; and the display screen is turned on so that the display screen obtains the frame compensation parameter in the specific position. Thus, the problems of low efficiency and high cost due to the fact that display screens which need or do not need compensation are detected independently in the related art are solved, the detection efficiency is improved and the detection cost is reduced. The method and system are used for detection of display screens.

The invention discloses a novel while-drilling azimuth gamma forward modeling method based on integral calculation. According to the method, while-drilling azimuth gamma forward modeling can be rapidly and accurately carried out; According to the forward modeling method, the radioactive parameter difference and the stratum attenuation coefficient difference of different stratums are considered, the influence of a shielding layer in the orientation-while-drilling gamma logging instrument on the gamma ray range received by the gamma detector is studied, and only gamma rays in the detection rangeare counted. And according to the stratum structure condition set by the user, carrying out forward modeling to obtain a prior model of the azimuth gamma data while drilling, and guiding geosteeringdrilling. The azimuth gamma image response characteristics of the drill bit when meeting different stratums can be simulated according to the azimuth gamma forward modeling algorithm, and a user can accurately and timely obtain the underground condition according to the image characteristics.

The embodiment of the invention provides a radioactive inert gas measurement device and method, equipment and a storage medium. According to the device, signal detection equipment detects a beta-gammacoincidence pulse electric signal obtained through the action of a sample and a detection crystal, and a beta detection crystal and a gamma detection crystal in a detector are arranged in a combinedmanner; the beta-gamma coincidence pulse electric signal is converted into a digital form from an analog form by signal acquisition and processing equipment; a computing device analyzes the amplitudeof the beta pulse signal in the beta-gamma coincidence pulse electric signal to obtain a first amplitude and the amplitude of the gamma pulse signal to obtain a second amplitude; and the activity concentration of the inert gas nuclide in the sample is calculated according to the pulse count of the digital beta-gamma coincidence pulse electric signal in the amplitude range formed by the first amplitude and the second amplitude. According to the scheme, the accuracy and reliability of measurement can be improved, radioactive inert gas nuclides can be distinguished, the concentration value of activity can be accurately and continuously measured, and the size, weight and cost of the detection device can be reduced.

The invention provides a nuclide identification method. The nuclide identification method comprises the steps of constructing a radionuclide library and an energy spectrum thereof by utilizing Monte Carlo simulation; constructing an energy response model of a gamma detector by utilizing the actual measurement energy spectrum; utilizing the energy response model to realize matching of a simulationenergy spectrum and an actual measurement energy spectrum; generating a gamma-ray energy spectrum with a spectral line drift effect by utilizing the matched nuclide gamma-ray energy spectrum; carryingout dimension reduction on the energy spectrum data of the gamma-ray energy spectrum with the spectral line drift effect; constructing a neural network, training the neural network, and realizing optimization of node parameters of the neural network; and identifying the gamma-ray energy spectrum of the unknown nuclide detected by the gamma detector by using the neural network. According to the nuclide identification method, the time consumption of the neural network training process and the calculation overhead of the nuclide identification algorithm are reduced; the nuclide identification speed is improved; and the nuclide identification type and identification precision are improved.

A handheld device for the location and identification of a radiation source, including: a radiation transparent housing; a radiation locator device disposed within the radiation transparent housing operable for determining the location of the radiation source, wherein the radiation locator device includes a plurality of gamma detection crystals arranged in a geometric pattern and separated by a gamma shielding material, a plurality of detectors coupled to the plurality of gamma detection crystals, and a processor module coupled to the plurality of detectors; one or more of a neutron detection crystal and a gamma spectroscopy crystal disposed within the radiation transparent housing adjacent to the radiation locator device; and one or more detectors coupled to the one or more of the neutron detection crystal and the gamma spectroscopy crystal and the processor module; wherein the one or more of the neutron detection crystal and the gamma spectroscopy crystal, the one or more detectors, and the processor module are collectively operable for identifying the radiation source.

The invention discloses a fast nuclide recognition method based on feature transformation and a neural network. The method comprises the steps that a gamma detector is used to measure a natural background spectrum and a radionuclide spectrum; a filtering algorithm is used to smooth the natural background spectrum to acquire a standard background spectrum; the filtering algorithm is used to smooth the radionuclide spectrum to acquire a standard radionuclide spectrum; the standard background spectrum is subtracted from the standard radionuclide spectrum to acquire a net quantization spectrum; eigentransformation is carried out on the net quantization spectrum; a certain number of transform coefficients are sequentially extracted as energy spectrum feature vectors and are normalized; and the normalized feature vectors are input into the neural network for nuclide recognition. The nuclide recognition method provided by the invention is not affected by the measurement time, the detection distance and the nuclide activity, has the advantages of fast response speed, natural background radiation, small noise interference and the like, and can be used for rapid nuclide recognition of a portable radiation detector.

The invention discloses a position reading-out device of gamma rays and a position reading-out system of the gamma rays. The position reading-out device of the gamma rays comprises a signal processing module, a discrimination module and a switch module, wherein the signal processing module is used for processing a current signal output by an SSPM (Security Service Provider Module) gamma ray detector and outputting an amplitude discrimination signal and a gating output signal; the discrimination module is used for discriminating the amplitude discrimination signal according to a pre-set signal and outputting a first control signal and a second control signal; the switch module is used for closing a corresponding switch according to the first control signal and the second control signal to output the gating output signal. By the aid of the position reading-out device of the gamma rays, an SSPM gamma ray detector array unit with the current signal which is higher than the pre-set signal is gated so that the influences on the resolution ratio of the position of the detector by dark noises are greatly reduced, and an application range of the SSPM detector in gamma detection is extremely expanded.

The invention belongs to the technical field of treatment, and discloses a proton application system for automatic and rapid detection of boron concentration. The proton application system for the automatic and rapid detection of the boron concentration comprises a low energy linear accelerator system used for generating proton beams and accelerating energy of the proton beams and an automatic boron-containing sample boron concentration rapid detection system which is used for automatic and rapid boron concentration quantitative measurement of batch boron samples and comprises a detection transmission section, a neutron irradiation device, an automatic boron-containing sample inlet device, a prompt gamma detection and analysis device, a beam current gathering device and a detection and control system. The proton application system can not only realize automatic and rapid detection of the boron concentration in the boron-containing samples, but also connect a neutron capture treatment system and/or a radionuclide production system at the same time, the device function is high, and the integration level is high. The invention further provides a boron-containing sample boron concentration rapid detection method. The efficiency of boron concentration detection for a large number of boron-containing samples is effectively improved and the cost of boron concentration detection is greatly reduced.

The invention discloses a controlled source integrated nuclear logging instrument and a logging method. The controlled source integrated nuclear logging instrument comprises a neutron generator, a near-thermal neutron detector, an ultra-thermal neutron detector, a near-gamma detector, a far-gamma detector and a far-thermal neutron detector which are sequentially arranged from one end to the otherend, wherein the near-thermal neutron detector and the ultra-thermal neutron detector are arranged side by side, and the far-gamma detector and the far-thermal neutron detector are arranged side by side; and the distance between the near-thermal neutron detector and the neutron generator as well as the distance between the ultra-thermal neutron detector and the neutron generator is 30+/-3 cm, thedistance between the near-gamma detector and the neutron generator is 45+/-3 cm, and the distance between the far-gamma detector and the neutron generator as well as the distance between the far-thermal neutron detector and the neutron generator are 75+/-3 cm. The yield of all elements in a formation, the formation porosity and the formation density value can be calculated simultaneously, multipledownhole measurement is avoided, and the measurement efficiency is greatly improved.

A method for setting the working voltage of a silicon photomultiplier in a gamma detection system includes the following steps: placing the gamma detection system in a constant-temperature constant-humidity environment; selecting a gamma radiation source; applying working voltage V to the silicon photomultiplier; placing the selected gamma radiation source in front of the gamma detection system at a distance of L, and using a multichannel collection system of the gamma detection system to collect outputs of the silicon photomultiplier for m times; calculating a signal to noise ratio Y<k> when the working voltage is V; and searching for the maximum element Y<max> in a signal to noise ratio array Y, the corresponding V<max> being the optical working voltage of the silicon photomultiplier in the gamma detection system. Through the method provided by the invention, the optical working voltage can be found out from a breakdown voltage and overvoltage range provided by a manufacturer, thereby providing optimal detectivity for the gamma detection system.

The invention discloses a cabinet body provided with a gamma detector and a neutron detector, comprising an upper frame, a lower frame, a left frame, a right frame, an upper bracket, a lower bracket, a vertical bracket, at least three left longitudinal brackets and at least three right longitudinal brackets, wherein the frames have a preset thickness and are mutually connected to form a cuboid structure; the upper bracket and the lower bracket are arranged on the upper frame and the lower frame along the thickness direction; the vertical bracket can be installed between the upper frame and the lower frame; the left longitudinal brackets and the right longitudinal brackets are installed on the left frame and the right frame with an equal distance along the thickness direction; and the vertical bracket, the left longitudinal brackets and the right longitudinal brackets are respectively and relatively provided with a plurality of first installation parts used for installing the gamma detector and a plurality of second installation parts used for installing the neutron detector. The invention also provides a radioactive substance detection system of the cabinet body.