46 results about "Neutron activation analysis" patented technology

A method for determining fracture geometry of a subterranean formation from radiation emitted from a fracture in the formation, including measuring gamma-radiation emitted from the fracture; subtracting background radiation from the measured gamma-radiation to obtain a peak-energy measurement; comparing the peak-energy measurement with a gamma-ray transport/spectrometer response model; and determining formation fracture geometry of the fracture in accordance with values associated with the response model.

The present invention relates to a resonant vacuum arc discharge apparatus for producing nuclear fusion. A resonant high-frequency high-voltage alternating current (AC) power supply is used to efficiently power a fusion tube normally containing deuterium, tritium and/or helium-3 vapor. Metals that can hold large amounts of hydrogen isotopes such as palladium and titanium can be used to increase the target density. The nuclear fusion device can be used for energy production, well logging, uranium mining, neutron activation analysis, isotope production or other applications that require a neutron source.

The invention relates to a sample changer for irradiated samples. Aiming at solving the problems of complex structure, high equipment cost and difficulty in mastering a control method of an existing automatic sample changer for neutron activation analysis, the invention provides a multi-channel sample changer. The multi-channel sample changer comprises a bracket, a sample transfer tube, an air cylinder, a driving device and two or more rabbit interface tubes fixed on the same side of the bracket, wherein the sample transfer tube is positioned on the other side of the bracket and realizes the longitudinal movement through the control of the air cylinder and the horizontal movement through the control of the driving device, and under the collaborative control of the air cylinder and the driving device, the sample transfer tube can move to all through holes of the bracket to be communicated with the rabbit interface tubes. The multi-channel sample changer provided by the invention is small in size, simple and compact in structure, low in cost, and easy to maintain, the control method is easy to master, and the automation of sample irradiation, cooling and measurement in a neutron activation analysis process can be realized.

The invention discloses a compact type D-D neutron generator applicable to neutron activation analysis or neutron radiography. The compact type D-D neutron generator comprises a vacuum cavity body, an ion source system, a beam extraction acceleration system, a target system, a high voltage feed system and a vacuum pump system, wherein ion beams generated by the ion source are extracted, accelerated and subjected to a D-D nuclear fusion reaction with a deuterium target on the target system to discharge neutrons; the two ends of the vacuum cavity body are connected with the ion source system, the beam extraction acceleration system, the target system and the high voltage feed system through an ion source positive electrode flange and a second flange respectively; the vacuum cavity body, the ion source positive electrode flange and the high voltage feed system are sealed through an O-shaped ring; and the ion source is a dual-plasma source, and the D gas is pumped in the ion source. The neutron yield can be more than 1*108s(1) magnitude, and the ion source, or target sheet and other system and internal parts can be replaced quite conveniently; and meanwhile, the usage cost is greatly lowered.

The invention relates to a sample element distribution measuring device and a method based on prompt gamma-ray neutron activation analysis. The sample element distribution measuring device based on prompt gamma-ray neutron activation analysis comprises a neutron source system, a sample control system, and a data detection processing system; the neutron source system is used for generating neutrons, and slowing-down is carried out so as to obtain thermal neutrons, collimation is carried out so as to obtain neutron beams, a sample to be detected is subjected to bombardment, and radiation capturereaction with nuclide in the sample is carried out so as to generate characteristic gamma-ray; the characteristic gamma-ray is detected and processed by the data detection processing system, gamma-ray energy and intensity are analyzed so as to obtain the nuclide information of a corresponding measured point; a measuring platform is adopted for rotation and moving of the object to be detected based on preset measuring step length, measuring of different positions is carried out, the corresponding relationship of the sample positions with the nuclide content is obtained, and sample element space distribution information is obtained.

The invention relates to an energy spectrum analysis method for large-size samples. Based on an instantaneous gamma-ray neutron activation analysis technique, gamma energy spectra are obtained by a scintillator detector. The obtained gamma energy spectra are subjected to neutron self-shielding correction treatment according to the element nonlinear response. The gamma energy spectra are analyzed by a least square method, single element response energy spectra are obtained, a single element spectral library is established, and nuclides in the samples can be identified and analyzed by the singleelement spectral library. Aiming at the defect that other spectral library analysis methods require a large number of samples, the single element spectral library which is not affected by backgroundelements is obtained by a few of calibrated samples based on a physical mechanism of the influence of multielement components on PGNAA analysis results. The measured energy spectra can be fitted directly by the single element spectral library, and the fast and accurate analysis of the nuclides in complex samples having multiple components and different reaction cross sections can be achieved.

The invention discloses a high-precision scanning mine detection device and a scanning detection method, and particularly relates to a device for judging existence of landmines and precisely locatingthe landmines through a PGNAA (Prompt Gamma-Ray Neutron Activation Analysis) technology. The high-precision scanning mine detection device is composed of a neutron source item system, a detection system, a shielding protection system and an energy spectrum analysis system, wherein the neutron source item system is used for exciting nitrogen feature gamma rays in a landmine sample, detectors collect the gamma rays and measure the energy spectrum, the shielding protection system of the detectors is made of a boron-containing polyethylene material, the activation interference of the detectors isreduced, and meanwhile, neutron damage of the detectors is lowered. According to the high-precision scanning mine detection device, the signal-to-noise ratio is regarded as a device design evaluationcriterion, the overall scanning mine detection device is designed based on a novel device evaluation optimization method, the signal-to-noise ratio is high, and the measurement level is improved; meanwhile, a novel scanning method is utilized for mine detection, the measurement time is shortened, and the precision of position detection of the landmines is improved at the same time.

The system for measuring chlorine concentration in fly ash cement concrete utilizes a portable neutron generator and a gamma-ray detector for performing prompt gamma neutron activation analysis of chlorine concentration in a fly ash cement concrete specimen. The system includes a portable neutron generator for generating a pulsed neutron beam having a neutron energy of approximately 2.5 MeV and a gamma-ray detector, such as a bismuth germanate (BGO) gamma-ray detector. A moderator having opposed first and second faces is further provided. The first face is positioned adjacent a target plane of the portable neutron generator, and the second face is adapted for positioning adjacent the fly ash cement concrete specimen. A detection axis of the gamma-ray detector is angled at approximately 45° with respect to an axis of symmetry of the fly ash cement concrete specimen.

The invention discloses a crack body density quantitative calculation method based on neutron activation analysis self-shielding correction. The crack body density quantitative calculation method comprises the steps that tracer elements with high thermal neutron capture cross sections are added into a propping agent, capture gamma-ray energy spectra before and after fracturing construction are measured through a pulsed neutron instrument, and a measured value of fracture body density after fracturing is obtained; the capture gamma-ray energy spectra before and after fracturing are analyzed based on an instantaneous gamma neutron activation analysis technology; elements with invariable contents before and after fracturing are extracted and any one element from the elements is elected, and aself-shielding correction factor is calculated by utilizing a characteristic peak counting ratio before and after fracturing of the element; a true value of the self-shielding correction factor is determined by performing self-shielding correction on a measured value of characteristic peak counting of the tracing element; and numerical fitting is carried out on the true value of the characteristic peak count of the tracing element and the fracture body density to determine a fracture body density quantitative calculation formula. According to the crack body density quantitative calculation method based on the neutron activation analysis self-shielding correction, the saturation effect caused by self-shielding is eliminated, the crack response sensitivity is further improved, and the quantitative calculation of the crack body density is realized.

The invention discloses a high-precision integrated explosion protection and detection device, which is an integrated device for detecting and determining existence of an explosive by use of a instantgamma-ray neutron activation analysis technology and with an explosion protection function. The device consists of an explosion protection system, a neutron source item system, a sample detection system, an energy spectrum analysis system, a shielding protection system and a mechanical power system, wherein a neutron source is used for activating a characteristic gamma ray of a detected object, adetector is used for gamma ray collection and energy spectrum measurement, the explosion protection system has a relatively good energy absorption cushioning function, the shielding protection systemis used for implementing dose protection over the environment and a human body and reducing harms of neutrons to the environment and the human body, and the mechanical power system has a convenient measurement function. According to the device, a signal to noise ratio is used as a design and evaluation criterion of the device and the whole device is optimized and designed on the basis of a noveldevice evaluation and optimization method, so that the device is endowed with a relatively high signal to noise ratio and the measurement level of the device is improved; in addition, an explosion protection function and a detection function are organically combined, so that the device is high in safety and convenient to operate.

The invention provides a method for performing dry-process mercury concentration measurement on power plant flue gas by replacing activated carbon with special modified high-sulfur petroleum coke adsorbents. The process of preparing modified high-sulfur petroleum coke comprises the two steps of Na2S solution activation and NH4Br solution modification performed on original petroleum coke. A special adsorption tube contains two sections of special modified high-sulfur petroleum coke adsorbents, namely an adsorption section and a penetration section in the flue gas flowing sequence. During measurement, the flue gas sequentially flows through the dust removal device, the adsorption tube and a water removal device. In the adsorption tube, gas-phase mercury in the flue gas is absorbed by the special modified high-sulfur petroleum coke adsorbents. The content of the mercury in the adsorbents is analyzed by utilizing a neutron activation analysis method or the content of the mercury in a filtrate is analyzed by utilizing a cold vapor atomic absorption spectrometry (CVAAS) or a cold vapor atomic fluorescence spectrometry (CVAFS) after the adsorbents are leached with an acid liquor. The method provided by the invention has the benefit that the expensive activated carbon is replaced with the special modified high-sulfur petroleum coke adsorbents, so that the dry-process flue gas mercury concentration measurement cost is effectively reduced.

The present invention discloses an integrated gamma spectrum measurement system suitable for neutron activation analysis, which includes a detector, a header multichannel pulse amplitude analyzer, aninterface and a shell. The detector includes a scintillator and a photomultiplier; and gamma ray enters the detector and is finally converted into an electric pulse signal. The header multichannel pulse amplitude analyzer is packaged by the metal shell, which includes a voltage divider circuit, an adjustable high voltage power circuit and a multichannel spectrum acquisition circuit. The voltage divider circuit and the adjustable high voltage power circuit provide the bias voltage for each dynode of the photomultiplier. The multichannel spectrum acquisition circuit realizes the conversion of the ray pulse signal to the spectrum. An interface circuit adopts a PoE interface, which can not only supply power for the system but also realize the function of network port communication. The shell plays a role of carrying, electromagnetic shielding and radiation shielding; and all components and circuits are enclosed in the detector shell. The invention provides the gamma spectrum measurement system with small volume, low cost, convenient application, a high counting pass rate and high energy resolution.

The invention discloses a method for gradually approximating a scale gamma energy spectrum high-energy region based on peak shape fitting. According to the method, a standard gamma source and gamma rays generated by neutron activation characteristic substances are gradually adopted for calibration; a simple energy spectrum is adopted to scale a low-energy section, then a low-energy section scale result is used to estimate a high-energy section peak position, a peak shape fitting method is adopted to determine a peak center, an energy scale range is continuously expanded, an interested high-energy section is gradually approached, and finally a channel address interval corresponding to the interested high-energy section is obtained. Compared with a traditional method that professionals recognize manual scales after multiple samples are activated at a time and complex gamma energy spectra are obtained, the method provides a set of self-adaptive processes and algorithms, automatic scale isachieved, the requirement for operators is reduced, and an application scene of neutron activation analysis is greatly expanded.

The invention discloses an activated foil clamping structure and activated foil taking-out device for neutron activation analysis. The device comprises a piece of an activated foil clamping structurecomposed of high-purity aluminum (purity of larger than 99.999%) and five pieces of aluminum-nickel or aluminum-cobalt alloy activated foil embedded on the high-purity aluminum, and a device for taking out the aluminum-nickel or aluminum-cobalt alloy activated foil. According to the activated foil clamping structure of the invention, five groups of activated foils are fixed in an inlaying manner,so that the risk of accidental falling of the activated foils can be effectively reduced, the function of obtaining a plurality of data points by single measurement can be realized, the laying efficiency is improved, and the activated foil clamping structure is suitable for reactors running at low power and high power; according to the activated foil taking-out device of the invention, the position of the activated foil can be positioned, the activated foil is taken out in a long-distance mechanical pressure applying mode, and the activated foil taking-out device is provided with the lead glass transparent cover, so that the gamma-ray radiation dosage borne by experiment operators can be effectively reduced; and the activated foil clamping structure and activated foil taking-out device forneutron activation analysis are suitable for measurement of neutron flux at a designated position in a reactor.

The invention discloses a high-flux neutron channel. Under same conditions, more neutrons can be led out of an outlet of the neutron channel, an existing neutron channel is changed from cylinder-shaped to truncated cone-shaped, and a polyethylene slowing-down layer and a graphite reflection layer are added around the neutron channel, so that the neutron flux of the outlet of the neutron channel is increased by over 80%. Graphite is added as the reflection layer firstly, so that the neutron flux is increased by more than 50%; then, the neutron channel is changed from cylinder-shaped to truncated cone-shaped, so that the neutron flux is increased by about 60%; the polyethylene slowing-down layer is added between the neutron channel and the graphite reflection layer, so that neutrons are slowed down firstly and then the neutron flux is increased by more than 80% finally; therefore, 252 Cf neutron sources which are same in activity can lead out more neutrons of the neutron channel in the practical application aspects of neutron activation analysis and irradiation and the like, so that the economic benefits of the neutron sources are improved, the operation time of operators is shortened, and the radiation of the neutron sources to the human body and the environment can be alleviated.

The invention relates to a device for performing flue gas mercury concentration measurement by utilizing modified fly ash. The device comprises a dust removal device, an adsorption tube, a temperature control device, a tail gas treatment device, a flow control device and an atmosphere control device. During measurement, flue gas sequentially flows through dust removal device, the adsorption tube and the tail gas treatment device. In the adsorption tube, gas-phase mercury in the flue gas is absorbed by a special modified fly ash adsorbent. The atmosphere of an adsorption environment is controlled by the atmosphere control device in the process that the flue gas passes through the adsorption tube. The temperature control is performed by the temperature control device in the process that the flue gas passes through the dust removal device and the adsorption tube. The content of the mercury in the adsorbent is analyzed by utilizing a neutron activation analysis method or the content of the mercury in a filtrate is analyzed by utilizing a cold atomic absorption spectrometry or a cold atomic fluorescence spectrometry after the adsorbent is leached with acid liquor. The device provided by the invention has the benefit that the adsorption atmosphere is controlled by adopting the atmosphere control device, so that the adsorption efficiency of the cheap adsorbent is improved, and while the cost is reduced, the flue gas mercury concentration measurement accuracy is ensured.