32 results about "Fission neutron" patented technology

The invention discloses a nuclear reactor core capable of improving neutron flux rate; the core is a circular structure; the center of the core is a Be thermal neutron well region, the circumference of which is a fuel component region, the circumference of which is a Be reflecting layer; the whole core is a circular structure. The maximum thermal neutron flux rate of the central Be thermal neutron well region can be over twice larger than that of the fuel region ; as the thermal neutron well region can be provided with fuel component with high uranium content, the fuel component center can form fast neutron region, therefore, the generated fission neutron has high flux rate.

The present invention includes an apparatus and method for neutron radiation detection. The apparatus comprises combining thin walled, boron-coated straw tubes with a plastic moderator material interspersed around the tubes. The method involves using such an apparatus through application of voltage to a central wire running inside the tubes and collecting electrical pulses generated thereby.

The invention relates to a hybrid reactor cladding for realizing long-term energy amplification by using a fast-thermal coupling mixed energy spectrum, which comprises a first wall, a fast fission region, a non fission neutron multiplication area, a thermal fissile area and a tritium breeding area. The structure of the hybrid reactor cladding is the first wall, the fast fission area, the non fission neutron multiplication area, the thermal fission area and the tritium breeding area in sequence from inside to outside. The cladding adopts the way that helium flows circumferentially and is cooled step by step. Helium enters into the cladding from a side pipeline of the first wall, enters into a fast neutron fission area transversely through the cladding after cooling the first wall, enters into the non fission neutron multiplication area and the thermal fission area from the other side of the cladding after cooling the fast neutron fission area and then enters the side pipeline of the first wall and flows out. Liquid lithium lead in the tritium breeding area enters into the cladding from the upper part of the cladding and flows out from the lower part and brings out heat and tritium relying on the flow. According to the hybrid reactor cladding, the long-term energy amplification, higher power generation efficiency and better safety can be achieved.

A π-conjugated organic material for detecting ionizing radiation, and particularly for detecting low energy fission neutrons. The π-conjugated materials comprise a class of organic materials whose members are intrinsic semiconducting materials. Included in this class are π-conjugated polymers, polyaromatic hydrocarbon molecules, and quinolates. Because of their high resistivities (≧109 ohm·cm), these π-conjugated organic materials exhibit very low leakage currents. A device for detecting and measuring ionizing radiation can be made by applying an electric field to a layer of the π-conjugated polymer material to measure electron / hole pair formation. A layer of the π-conjugated polymer material can be made by conventional polymer fabrication methods and can be cast into sheets capable of covering large areas. These sheets of polymer radiation detector material can be deposited between flexible electrodes and rolled up to form a radiation detector occupying a small volume but having a large surface area. The semiconducting polymer material can be easily fabricated in layers about 10 μm to 100 μm thick. These thin polymer layers and their associated electrodes can be stacked to form unique multi-layer detector arrangements that occupy small volume.

The present invention includes an apparatus and method for neutron radiation detection. The apparatus comprises combining thin walled, boron-coated straw tubes with a plastic moderator material interspersed around the tubes. The method involves using such an apparatus through application of voltage to a central wire running inside the tubes and collecting electrical pulses generated thereby.

The invention relates to a gas scintillation based fission neutron detector. The detector comprises a detector barrel, an entrance window hermetically arranged at the front end face of the detector barrel, an exit window hermetically arranged at the rear end face of the detector barrel, two quartz windows arranged at two lateral surfaces of the detector barrel, at least two target stands fixed in the detector barrel, fission targets suspended on the corresponding target stands, two photoelectric detectors arranged outside the two quartz windows respectively, a charging tool communicated with the detector barrel, and scintillation gas filled in a barrel core; and the scintillation gas adopts inert gas or nitrogen. The gas scintillation based fission neutron detector solves the technical problems of low neutron / gamma resolving power and low neutron sensitivity in the conventional neutron detector, and has the advantages of smooth fast neutron energy response and quick time response.

The invention relates to a fission neutron well logging correction method, which comprises the following steps that: 1, a correction counting rate of a neutron monitor of a tested wellhole is obtained; 2, a water correction factor is obtained; 3, the neutron counting rates near a well wall and along a well axis after the water correction in a tested well are obtained; and 4, the neutron counting rates near the well wall and along the well axis after the epithermal neutron service life correction in the tested well are obtained. The method provided by the invention can be used for directly measuring the uranium content. Compared with drill hole rock core sampling chemical analysis results, the method has the advantages that the deviation does not exceed 10 percent. The uranium ore grade is more objectively reflected; particularly in sections with damaged uranium-radium balance, conventional gamma well logging is combined, and a uranium-radium balance factor of a drill hole ore block can be determined in site; and the rock core sampling quantity is reduced, and the uranium resource investigation cost is reduced. The method can be used for accurately measuring the uranium leaching rate in mining using the leaching technology, avoids the environment pollution and the cost waste due to unquestioning excessive liquid injection in the process of the mining using the leaching technology, and has important practical significance on improving the efficiency of the mining using the leaching technology, reducing the cost and protecting the environment.

The invention discloses a prompt uranium fission neutron logging technique based on the epithermal neutron and thermal neutron ratio. The technique is specifically characterized in that in boreholes, fast neutrons, also called virgin neutrons, are transmitted to formation rock by means of pulsing; the virgin neutrons are successively moderated by in-well medium into epithermal neutrons and thermal neutrons; the thermal neutrons induce 235U fission to radiate prompt uranium fission neutrons, also called secondary neutrons; the secondary neutrons are successively moderated into epithermal neutrons and thermal neutrons, inducing 235U fission again; from any moment t after the virgin neutrons are moderated into the thermal neutrons, the total nE(t) of existing epithermal neutrons in the formation rock is in direct proportion to the total nT(t) of existing thermal neutrons and uranium content qu, namely nE(t)<nT(t) qu. By using a bi-neutron detector and a dual-time detector disposed in the invention to measure time spectra of the epithermal and thermal neutrons and defining an attenuation ratio of the epithermal neutrons to the thermal neutrons as 'E / T', from a moment, when the virgin neutrons are moderated into the thermal neutrons, to any moment, a uranium quantitative real-time algorithm based on saturated seams is constructed.

The invention relates to a critical alarm device and system. A neutron detection unit detects the number of fission neutrons of a critical accident to obtain neutron detection data, and a lower computer compares the neutron detection data with a set alarm threshold and drives an alarm unit to send an alarm signal when the neutron detection data is greater than or equal to the set alarm threshold.Therefore, the situation that a critical alarm cannot be sent in time due to the fact that a detector exceeds the range under a high dose rate is avoided, and the accuracy and safety of the critical alarm are improved.

The invention discloses a fission neutron converter. The fission neutron converter comprises an inner sleeve and an outer sleeve coaxially sleeved with the inner sleeve, wherein an upper grid plate and a lower grid plate are respectively mounted at two end parts between the inner sleeve and the outer sleeve, and cross-shaped fuel rods are mounted between the upper grid plate and the lower grid plate. According to the fission neutron converter disclosed by the invention, the bottleneck problem of material irradiation is solved by adequately utilizing technical conditions of the convention research reactor, the twofold effect can be achieved with half effort, and the current problems that on the one hand, the advanced research reactor cannot be built in one day, and on the other hand, a large amount of in-service research reactors are still operated are solved; by utilizing the cross-shaped fuel rods made of high-fission-density fuels, the fluence rate of local fast neutrons of a reactor core is substantially improved under a condition that the physical and thermotechnical safety of an HFETR (High Flux Engineering Test Reactor) is met, so that the technical potentials of the HFETR effectively are developed, and furthermore, the irradiation cost of materials is saved.

The invention discloses a nuclear fission chain reaction dynamic demonstrator, which comprises a bracket, a glass container, a bottom plate, a goal opening, a swing cylinder, an air pump, an air blowing port, a programmable controller, a splitting neutron ball and an impacting neutron ball , the glass container is arranged on the support, the bottom plate is arranged under the glass container, the atomic nucleus model map arranged together is arranged on the bottom plate, a notch is arranged on the bottom plate corresponding to the atomic nucleus model map, and the split neutron ball is placed in the notch, The blowing port is arranged at the lower side of the glass container, the shaft of the swing cylinder is supported on the bottom side of the bottom plate, the ball opening is located at the top of the glass container, and the swing cylinder is connected with the programmable controller. The demonstrator of the invention can dynamically demonstrate nuclear fission chain reaction, and is convenient for science popularization and education of nuclear fission knowledge.

The invention discloses a fuel assembly available for integral replacement. By fixation of the fuel assembly and a top cover, axial fixation of the fuel assembly is realized, a fuel assembly fixation problem caused by a high-density cooling agent of a liquid-state lead-based reactor is solved, and integral replacement of the fuel assembly is realized. Since fuel elements with fixed upper ends and free lower ends are reinforced under the action of axial flushing force and buoyancy of the cooling agent, stability of the fuel elements in operation of the reactor is improved; owing to a moderator filled in each fixed rod, fission neutron moderation can be accelerated to improve reactivity of a reactor core and reduce the size of the reactor core. The fuel assembly has advantages of high safety, simple structure, small reactor core size and the like.

The invention relates to a solution system plutonium concentration estimation method based on neutron coincidence counting and a monitoring system. The method comprises steps of calculating correctionfactor by adopting a three-dimensional Monte Carlo program; correcting the neutron absorption effect of the point model equation set, correcting the energy spectrum difference between (alpha, n) neutrons and fission neutrons, and correcting the detection efficiency difference between induced fission neutrons and spontaneous fission neutrons, so the neutron coincidence counting calculation resultof the improved point model equation set can accurately reflect the measurement result of neutron coincidence counting of the neutron detection system outside the uranium plutonium solution system, and therefore the plutonium concentration of the uranium plutonium solution system can be accurately predicted and estimated. The method realizes non-destructive monitoring and analysis of the plutoniumconcentration of the uranium plutonium solution system, and is an advanced solution system plutonium concentration estimation method with engineering feasibility.

The invention discloses a 4H-SiC neutron detector applied to prompt fission neutron uranium mine logging. The detector is of a Schottky structure, an n<+> type 4H-SiC crystal serves as a substrate, and a chemical vapor deposition technology is utilized for performing isoepitaxial growth of an n<-> type 4H-SiC epitaxial layer on the front side of the substrate; a magnetron sputtering technology isutilized for respectively depositing multiple layers of metals and protective layers on the back side of the substrate and one side of the epitaxial layer and performing annealing treatment, thereby forming ohmic contact and Schottky contact; and finally, a photoetching technology and the magnetron sputtering technology are utilized for sputtering 6LiF on the Schottky contact to form a neutron conversion layer. In addition to the advantages of conventional semiconductor neutron detectors, the 4H-SiC neutron detector also has the advantages of small volume, easy n / gamma discrimination, high temperature resistance and irradiation resistance. The 4H-SiC neutron detector disclosed by the invention has high detection efficiency and high counting rate and is suitable for the uranium mine loggingenvironment with narrow space, high temperature and intense radiation.

The invention discloses a nuclear reactor core capable of improving neutron flux rate; the core is a circular structure; the center of the core is a Be thermal neutron well region, the circumference of which is a fuel component region, the circumference of which is a Be reflecting layer; the whole core is a circular structure. The maximum thermal neutron flux rate of the central Be thermal neutron well region can be over twice larger than that of the fuel region ; as the thermal neutron well region can be provided with fuel component with high uranium content, the fuel component center can form fast neutron region, therefore, the generated fission neutron has high flux rate.

The invention provides a system for online monitoring of the integrity of a lead-bismuth reactor fuel element can. The system comprises a delayed fission neutron detection unit which is used for monitoring and acquiring data of a delayed fission neutron fluence rate in real time, a characteristic fission nuclide detection unit which is used for monitoring and acquiring data of characteristic fission nuclide radioactivity and a back-end data centralized processing unit, wherein the back-end data centralized processing unit has the functions of automatic alarm and output of an evaluation conclusion report, and is used for receiving the data of the delayed fission neutron fluence rate and the data of the characteristic fission nuclide radioactivity, comparing the data with preset corresponding thresholds, and giving a comparison conclusion. Through the online monitoring system, automatic and real-time evaluation of the integrity of the lead-bismuth reactor fuel is achieved, and the data is monitored and acquired through the delayed fission neutron detection unit and characteristic fission nuclide detection unit simultaneously, and is verified through the data centralized processing unit, so that the reliability of on-line integrity evaluation of the lead-bismuth fuel is improved greatly, and the system can also be used for guiding shutdown leakage tests.

The invention relates to a method for calibrating and correcting a fission neutron logging instrument. The neutron count rate by the well wall and along the well axis after water correction; 4. Obtain the neutron count rate by the well wall and along the well axis after the epithermal neutron life correction in the measured well; the present invention can directly measure the uranium content, and Compared with the chemical analysis results of hole core sampling, the deviation does not exceed 10%; the present invention reflects the grade of uranium ore more objectively, especially in the section where the balance of uranium and radium is destroyed, combined with conventional gamma logging, the drilling section can be determined on site The uranium-radium balance coefficient reduces the number of core samples and reduces the cost of uranium resource exploration; the invention can accurately measure the uranium leaching rate in in-situ leaching mining, and avoid environmental pollution and Waste of cost has important practical significance for improving the efficiency of in-situ leaching mining and smelting, reducing costs and protecting the environment.

The invention discloses a dynamic demonstration instrument for nuclear fission chain reaction, which comprises a bracket, a glass container, a baseplate, a ball inlet, a rocking cylinder, an air pump, a blowing mouth, a programmable controller, a fission neutron ball and an incident neutron ball. The glass container is arranged on the bracket; the baseplate is arranged under the glass container; a together-aligned atomic nucleus model is arranged on the baseplate; a recession is positioned on the baseplate, which is corresponding to the atomic nucleus model; the fission neutron ball is arranged inside the recession; the blowing mouth is arranged at the lower part at the lateral side of the glass container; the shaft of the rocking cylinder bears one side of the baseplate at the bottom of the baseplate; the ball inlet is situated upon the glass container; and the rocking cylinder is connected with the programmable controller. The demonstration instrument of the invention dynamically demonstrates the nuclear fission chain reaction, thus facilitating the popularization of science and education for the knowledge about nuclear fission.

The invention discloses a neutron transport solving method for a reactor with uniformly distributed materials. According to the method, machine learning and a Monte Carlo method are combined. Firstly, a series of reactors with uniformly distributed materials need to be constructed, and each reactor contains different material components or densities. Solving the reactors by applying a Monte Carlo method, and counting the scattering and fission probabilities of neutrons and the characteristic distribution of the neutrons after simulation is finished; and training to obtain a full-connection neural network model which takes neutron energy and material composition as input and takes characteristic distribution of scattering and fission neutrons as output by utilizing the probabilities and characteristic distribution data. And finally, by using the obtained full-connection neural network model, taking neutrons and materials generated by fission as input to obtain fission neutrons of the next generation, and carrying out iterative calculation until convergence. Compared with an existing Monte Carlo method, the method has the advantages that a large number of sampling processes are not needed any more, the calculation speed is higher, and neutron transport calculation can be rapidly completed for a reactor with materials evenly distributed.

A fission source extrapolation method for accelerating the Monte Carlo criticality calculation. By setting an initial fission source for the Monte Carlo criticality calculation, neutron transport simulation can be performed according to the initial fission source, and the spatial distribution of the fission source can be counted; Then, based on the setting that the fission source will gradually converge to the real solution, the fission source is linearly extrapolated. Finally, the entire calculation model is divided into multiple fission regions, and the source extrapolation factor is used to adjust the number of fission neutrons in each region. The invention can remarkably improve the convergence speed of the fission source of the Mongolian card calculation of the reactor, and can be used as an acceleration module of the Mongolian card program, so that it can be used by a variety of Mongolian card programs, and the efficiency of the critical calculation of the Mongolian card is significantly improved.

The invention relates to an out-of-pile detector calibration method, which comprises the following steps of: S1, acquiring detector response factors at all positions in a pile by simulating a process of transporting neutrons from the inside of the pile to an out-of-pile detector; s2, a theoretical xenon oscillation process is simulated and constructed, and fission neutron generation rates of all positions in the reactor in the xenon oscillation process are obtained; s3, according to the detector response factors and the fission neutron generation rate, response current distribution of all the out-of-pile detectors in the xenon oscillation process is obtained; and S4, calculating the calibration coefficient by using the current distribution. According to the invention, while the implementation of a field xenon oscillation test is cancelled, a set of complete calibration coefficient with enough precision can still be provided, the disturbance to a reactor core is reduced, and the control difficulty of the reactor (especially at the end of the service life) is reduced, so that the pressure of operators is also reduced.

The invention relates to a critical alarm device and system. A neutron detection unit detects the number of critical accident fission neutrons to obtain neutron detection data. The lower computer compares the neutron detection data with the set alarm threshold. When equal to the set alarm threshold, the drive alarm unit sends out an alarm signal. Based on this, the critical alarm cannot be issued in time to avoid the overrange of the detector at a high dose rate, and improve the accuracy and safety of the critical alarm.

The invention discloses a slurry density correction method based on uranium fission instantaneous neutron logging data. A numerical logging model is established through Monte Carlo (MCNP) simulation;according to the change relation of transient fission neutron uranium mine logging thermal neutrons and epithermal neutron densities along with the stratum environment, related factors influencing thelogging neutron time spectrum are studied; the influence of different mud densities under the same borehole diameter on E / T (the ratio of epithermal neutrons to thermal neutrons) in transient fissionneutron uranium mine logging is discussed, a correction function influenced by the stratum environment is given through an MCNP simulation data fitting method, and the influence of different mud densities on uranium content measurement can be effectively reduced.

The invention belongs to the technical field of nuclear physics. It discloses a systemic calculation method for the change of fission yield with neutron energy. The method is to fit two sets of parameters to determine the Gaussian function at two points, and the parameters at other excitation energies are determined by their respective functional relationships, so that The mass distribution of the original fission fragments is obtained; then according to the prompt neutron emission number v(A, Ec) of each fragment during the composite nuclear fission of the excitation energy, the chain yield of each mass chain after fission induced by different neutron energies is obtained. The calculation results reasonably explain the experimental results of the yield variation with the incident neutron energy, and a more accurate and reliable yield-energy curve is obtained.

The present invention relates to a gas scintillation based fission neutron detector. The detector comprises a detector barrel, an entrance window hermetically arranged at the front end face of the detector barrel, an exit window hermetically arranged at the rear end face of the detector barrel, two quartz windows arranged at two lateral surfaces of the detector barrel, at least two target stands fixed in the detector barrel, fission targets suspended on the corresponding target stands, two photoelectric detectors arranged outside the two quartz windows respectively, a charging tool communicated with the detector barrel, and scintillation gas filled in a barrel core; and the scintillation gas adopts inert gas or nitrogen. The gas scintillation based fission neutron detector solves the technical problems of low neutron / gamma resolving power and low neutron sensitivity in the conventional neutron detector, and has the advantages of smooth fast neutron energy response and quick time response.

A method for solving neutron transport for reactors with uniform distribution of materials that combines machine learning and Monte Carlo methods. The first step is to construct a series of reactors with a uniform distribution of materials, each containing a different material composition or density. These reactors were then solved using the Monte Carlo method, and the neutron scattering and fission probabilities and the characteristic distribution of neutrons were calculated after the simulation. Using these probability and feature distribution data, a fully-connected neural network model is trained that takes neutron energy and material composition as input and the feature distribution of scattered and fission neutrons as output. Finally, using the obtained fully connected neural network model, the neutrons and materials produced by fission are used as inputs to obtain the next generation of fission neutrons, and the calculation is iterative until convergence. Compared with the existing Monte Carlo method, the method no longer needs to perform a large number of sampling processes, and the calculation speed is faster, and the neutron transport calculation can be quickly completed for a reactor with uniform distribution of materials.

The invention discloses a fuel assembly available for integral replacement. By fixation of the fuel assembly and a top cover, axial fixation of the fuel assembly is realized, a fuel assembly fixation problem caused by a high-density cooling agent of a liquid-state lead-based reactor is solved, and integral replacement of the fuel assembly is realized. Since fuel elements with fixed upper ends and free lower ends are reinforced under the action of axial flushing force and buoyancy of the cooling agent, stability of the fuel elements in operation of the reactor is improved; owing to a moderator filled in each fixed rod, fission neutron moderation can be accelerated to improve reactivity of a reactor core and reduce the size of the reactor core. The fuel assembly has advantages of high safety, simple structure, small reactor core size and the like.

Methods and systems that utilize centrifugally tensioned metastable fluid detector (CTMFD) sensors and an external probing source to detect the presence of fissile and fissionable materials, including but not limited to special nuclear materials (SNMs), in containers. Such a method includes subjecting a container to probing with a fission-inducing radiation species that induces fission in an fissile or fissionable material, detecting fission neutrons emitted from the fissile or fissionable material with CTMFD sensors that each contain a detection fluid in which the fission neutrons induce cavitation in a centrifugally tensioned portion of the detection fluid. A threshold energy neutron analysis mode is then utilized to reject the radiation species and detect a fraction of the fission neutrons that have energies above a predetermined energy threshold determined by centrifugally-induced tension in the centrifugally tensioned portion of the detection fluid within each of the CTMFD sensors.

Methods and systems that utilize centrifugally tensioned metastable fluid detector (CTMFD) sensors and an external probing source to detect the presence of fissile and fissionable materials, including but not limited to special nuclear materials (SNMs), in containers. Such a method includes subjecting a container to probing with a fission-inducing radiation species that induces fission in an fissile or fissionable material, detecting fission neutrons emitted from the fissile or fissionable material with CTMFD sensors that each contain a detection fluid in which the fission neutrons induce cavitation in a centrifugally tensioned portion of the detection fluid. A threshold energy neutron analysis mode is then utilized to reject the radiation species and detect a fraction of the fission neutrons that have energies above a predetermined energy threshold determined by centrifugally-induced tension in the centrifugally tensioned portion of the detection fluid within each of the CTMFD sensors.

The invention discloses a uranium fission prompt neutron logging technology based on the ratio of epithermal neutrons to thermal neutrons. Specifically, it means that fast neutrons, also known as primary neutrons, are emitted to formation rocks in a pulsed manner in a borehole. Primary neutrons are successively moderated into epithermal neutrons and thermal neutrons by the medium in the well. Thermal neutrons induce 235U fission and emit uranium fission prompt neutrons, also known as secondary neutrons. The secondary neutrons are successively moderated into epithermal neutrons and thermal neutrons, and induce 235U fission again. At any time t after the primary neutrons are moderated into thermal neutrons, the total amount of epithermal neutrons nE(t) in the formation rock is proportional to the total amount of existing thermal neutrons nT(t) and the uranium content qu, that is, nE(t)∝nT(t)·qu. Using the dual neutron detector and dual time spectrometer disclosed in the invention, the time spectrum of epithermal neutrons and thermal neutrons is measured, and the epithermal neutrons from the time when primary neutrons are slowed down to thermal neutrons to any time are defined , The ratio of thermal neutron attenuation is the "E / T" ratio, and a real-time uranium quantification algorithm based on saturated ore bed is constructed.