72 results about "Fission" patented technology

A system for radioisotope production uses fast-neutron-caused fission of depleted or naturally occurring uranium targets in an irradiation chamber. Fast fission can be enhanced by having neutrons encountering the target undergo scattering or reflection to increase each neutron's probability of causing fission (m, f) reactions in U-238. The U-238 can be deployed as layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material.

The invention belongs to the technical field of nuclear fuel reprocessing and discloses a process for recovering and purifying neptunium from waste liquor discharged from 2AW+2DW in Purex flow. The process comprises the primary neptunium purification process and the secondary neptunium purification process; the process comprises the following steps: by taking 30 volume percent of TBP-kerosene as an extraction agent, extracting Np, U and Pu to the organic phase, and allowing the fission fragments to enter the aqueous phase; adding a reducing agent and an extraction agent into the organic phase, reversely extracting the Np to the aqueous phase; further extracting uranium and plutonium, and performing oxalate precipitation, filtering and calcining on the obtained aqueous phase product to obtain the purified neptunium oxide solid. The process has the advantages that the neptunium recovery rate is over 98 percent, and the uranium and plutonium removal coefficient in neptunium is high.

The invention relates to a novel global critical accident alarm method and system. The method comprises the following steps: judging whether a critical accident occurs or not; when a critical accidentoccurs, judging the position of a critical accident area according to dose rate responses of detectors at different positions; and carrying out integral processing on a measurement value of the related detector to obtain the total fission times under the critical accident condition. The system comprises a detection unit; a critical accident judgment unit used for judging whether a critical accident occurs or not; a position confirmation unit used for judging the position of a critical accident area according to the dose rate response of detectors at different positions when a critical accident occurs; and a total fission frequency calculation unit used for carrying out integral processing on the measurement value of the related detector to obtain the total fission frequency under the critical accident condition. The beneficial effects of the invention are that the method can give an alarm and determine an accident point under the condition of a critical accident, and quickly estimatesthe size of the accident and the total number of fission times after the accident is finished.

The invention relates to a method for specially processing decay heat calculation in a compression process of a fuel consumption database, which comprises the following steps of: calculating a representative nuclear fuel assembly by using a fine fuel consumption database to obtain a set of related data, and restarting fuel consumption calculation under the condition of assuming that a fission product does not react with neutrons to obtain another set of data; and using two sets of data for carrying out decay heat calculation twice, firstly, selecting a few of heavy nuclides with great contribution of heavy nuclear decay heat, secondly, selecting important contribution nuclides of the fission product irradiation effect, adding a few of heavy nuclides with great contribution and the important contribution nuclides of the fission product irradiation effect into the compressed fuel consumption database, and deducting the fission product irradiation effect from decay heat, so as to obtain afission system decay heat release function, further obtaining a fission sub-system decay heat release function based on the fission product decay hypothesis, finally performing twice nonlinear fitting on the fission sub-system decay heat release function to obtain multiple groups of decay heat precursor kernels, and adding the multiple groups of decay heat precursor kernels into the compressed fuel consumption database. The decay heat result calculated by the method is more accurate.

The invention discloses a measuring experimental device for fission gas leakage after fuel element damage and an experimental method thereof. The device comprises a water inlet pipe, a lower-part sealing cover, a sealed can, a pumping chamber, a water outlet pipe, two fixing devices, a fuel rod fixing ring, a lifting console, a connection chamber, a fuel rod, a sealing head and a pumping tube, wherein the water inlet pipe is connected with the lower-part sealing cover; the lower-part sealing cover is connected with the sealed can; the fuel rod is placed in the sealed can and fixed by the fuelrod fixing ring; the water outlet pipe is connected with the sealed can; the two fixing devices are used for fixing the entire sealed tank; the sealed tank is connected with the pumping chamber; the pumping chamber is connected with the lifting console through the connecting chamber; and the lifting console is used for controlling the sealing head to go up and down in the pumping chamber. The device has the advantages that the overall structure is simple and easy to process; and a measurement scheme is provided for the study of fission gas leakage characteristics after the fuel element is damaged. The invention also provides the experimental method of the measuring experimental device.

The invention discloses a system for releasing and measuring fission gas of heavy-water reactor fuel elements. The system comprises a drilling device (1), a calibrating mechanism (2), a collecting device (3) and a vacuum unit (4). The drilling device (1) is used for drilling the heavy-water reactor fuel elements (7) and completely releasing the fission gas, the calibrating mechanism (2) is used for measuring the volumes of cavities inside the heavy-water reactor fuel elements (7), the collecting device (3) is used for pressurizing and collecting the fission gas released by the drilling device (1), and the vacuum unit (4) is used for vacuumizing the drilling device (1), the calibrating mechanism (2) or the collecting device (3). The invention further discloses a process for releasing and measuring the fission gas. The system and the process have the advantages that technologies for releasing and collecting the fission gas of the heavy-water reactor fuel elements are established, and accordingly the fission gas of the heavy-water reactor fuel elements can be released, collected and measured.

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.

The invention provides a slope fission early warning method and system. According to the invention, early warning is carried out on the slope fission according to the principle that the landslide is separated from the slope body before happening, so that the technical defect that sufficient time cannot be left for personnel and material transfer in the existing image monitoring method is overcome,and the advancement and accuracy of landslide early warning are improved. According to the invention, before fission judgment, the monitoring area is determined based on the InSAR image; a fission monitoring area is delineated; an InSAR image at the (i + 3) th moment is obtained through real-time monitoring; fission monitoring and early warning are carried out on the monitoring points of the delineated fission monitoring area, the monitoring range is reduced, the calculated amount is reduced, fission monitoring is carried out on the basis of an isomorphic curve division mode by utilizing theprinciple that the change trends of the deformation quantities of the fission area are different, and the scientificity and precision of fission monitoring are guaranteed.

The invention relates to a thermal history simulation method, system and equipment based on fission track length distribution on a vertical section, and the method comprises the steps: obtaining N samples at different heights on the vertical section, and analyzing and processing the N samples to obtain samples meeting requirements, generating the final thermal history simulation result of the to-be-simulated geology through the thermal history curve of each sample meeting the requirement, and the thermal history simulation result being high in accuracy. The thermal history simulation method based on fission track length distribution on the vertical section is not limited by geological types in geological thermal history simulation, is widely used, and solves the technical problems that thermal history simulation obtained by adopting data obtained by a fission track method in the prior art is limited in energy exploration and the simulation data is inaccurate.

This invention relates to a method for fault-tolerant, reliable stream processing, where a data stream is processed by a processing means for yielding a result data stream, where before the data stream is processed, the data stream is decomposed (fission) into at least two fission sub-streams, such that each fission sub-stream carries a partial information of said data stream, the data stream is processed by processing the at least two fission sub-streams independently (P.a, P.b) yielding at least two fission result sub-streams, where each fission result sub-stream carries a partial information of the result data stream, and where the at least two fission result sub-streams are composable(fusion) to the result data stream, such that in case of unavailability of a part of the fission sub-streams the missing information could be interpolated. The invention also relates to an apparatus, a system and a computer software product therfor.

An accelerator-driven sub-critical reactor providing: 1) a significantly more efficient Thorium-cycle system, in one configuration, 2) a more energy-productive nuclear waste reduction system, 3) accelerator driven systems for other fertile-fission candidate elements, and 4) which may be applied to fusion systems (substituting the fission unit in the proposed system class and category) in a way that may lower the break-even point for such systems and thus make the advent of practical fusion sooner than otherwise possible. 5) In addition and importantly, an optical-power processing and distribution is also enabled by the proposed, providing both optical power as base power for telecom, process energy for industrial uses, and lighting and other wavelengths for consumer and general business use.

The invention provides a material source analysis method based on measurement of a Dpar value. The method comprises the following steps: measuring fission track Dpar values of samples in a deposition area and a source area; then grouping the obtained Dpar values by adopting a multi-dimensional scaling analysis method to obtain grouped data, and constructing a sedimentary area-source area coupling model by taking the Dpar values as parameters and matching the grouped data with the same Dpar value characteristics of the sedimentary area and the source area, so that the object source area and the rock type thereof can be accurately delineated, the sediment source can be determined, and the object source analysis is more accurate and credible. According to the material source analysis method, the material source analysis result can be accurately obtained only by measuring the Dpar value of the sample, thermal neutron irradiation does not need to be carried out on the sample in the Dpar value measuring process, and the time consumption for obtaining the Dpar value is short; the defect that in the traditional material source analysis process of measuring the fission track age, an analysis sample needs to be sent to an atomic nuclear reactor for thermal neutron irradiation or the irradiation time is long, and radioactivity exists is effectively overcome.

Methods and systems for non-intrusively detecting the existence of fissile materials in a container via the measurement of energetic prompt neutrons are disclosed. The methods and systems use the unique nature of the prompt neutron energy spectrum from neutron-induced fission arising from the emission of neutrons from almost fully accelerated fragments to unambiguously identify fissile material. These signals from neutron-induced fission are unique and allow the detection of any material in the actinide region of the nuclear periodic table.

The invention provides a reactor core. The reactor core comprises: a core grid that includes a plurality of unit grids connected with each other; a fuel assembly, which is arranged in the unit grids and provides fuel for fission reactions; a control part, which is arranged in the unit grids and is used for controlling the reactivity of the fission reactions; a reflection assembly, which is arranged in the unit grids and is used for reflecting neutrons generated by fission reactions so as to increase the neutron fluence rate of the reactor core; and production assemblies, which are disposed inthe cell grid, receive neutron irradiation and generate radioisotopes. The core is configured to include at least two production assemblies for generating different radioisotopes. The reactor core hasthe characteristics of simple structure, multiple types of produced isotopes and high efficiency.

The invention relates to a fission ionization chamber, in particular to a sampling fission ionization chamber and a fission total number measuring method based on the same. The problems that when an existing ionization chamber is used for measuring the fission yield, due to the fact that the ionization chamber has self-shielding and is poor in neutron energy spectrum consistency, large measurementuncertainty is introduced, gas fission products and short-life fission products cannot be measured, and the influence on the health of measuring personnel is large are solved. The ionization chamberis characterized by comprising a cathode support, a cathode and an anode support, wherein the cathode support and the cathode are coaxially arranged in sequence from bottom to top, and the anode support is arranged in an inner cavity of the cathode and is coaxial with the cathode. An annular sealing boss and an isolation boss are arranged on the lower bottom surface of the anode support. The sealing boss is positioned on the outer side of the isolation boss. The lower end surface of the sealing boss is hermetically and fixedly connected with the bottom inner surface of the cathode. A gap is formed between the lowermost end of the isolation boss and the inner surface of the cathode bottom. The sampling anode and the annular anode are attached to the lower bottom surface of the anode support, and the sampling anode is located in the isolation boss. The annular anode is located between the isolation boss and the sealing boss.

The invention discloses a pressurized water reactor neutron transport method based on deep learning-interface flow coupling, and the method comprises the steps: simulating a neutron migration process after the capture, fission or scattering of neutrons and materials in a local simple geometric problem through a Monte Carlo method; a simulation result of a Monte Carlo method and a deep learning method are utilized to learn neutron incidence-emergence distribution rules of different materials and geometric units, and a connection relation between simple geometric units and a neutron transport calculation process are established based on interface flow conservation. According to the method, the traditional process that in a deterministic theory method, resonance calculation must be carried out firstly to obtain a multi-group cross section, and then complex geometric multi-group transport calculation is carried out is subverted, and a novel neutron transport calculation method integrating a Monte Carlo method, deep learning and the deterministic theory method is formed.

The invention belongs to the technical field of nuclear reactor fuel elements, and particularly relates to a dual-cladding fuel element with enhanced moderating power, which comprises an upper end plug, air cavity springs A, inner cladding, an air cavity spring B, outer cladding, a plurality of core blocks A, a plurality of core blocks B and a lower end plug, wherein the plurality of core blocks Aare stacked at the lower end in the inner cladding; an inner cladding fission gas cavity is formed above the upper parts of the core blocks A; one end of the inner cladding fission gas cavity is separated from the core blocks A through the gas cavity spring A, and the other end of the inner cladding fission gas cavity is also provided with a gas cavity spring A; the outer wall of the inner cladding is sleeved with the outer cladding; the plurality of core blocks B are stacked between the outer wall of the inner cladding and the inner wall of the outer cladding from bottom to top; the air cavity spring B is arranged above the upper parts of the core blocks B; the upper end plug is in positioning connection with the tops of the inner cladding and the outer cladding; the lower end plug is welded with the bottoms of the inner cladding and the outer cladding; the materials of the core blocks A and the core blocks B are combined in the form of a fission material-moderating material or a moderating material-fission material.

The invention relates to a method and device for separating uranium particles. The method includes adopting the fission track etching method to obtain the fission track of uranium on the fission track etching film, and then using a laser beam to focus on the fission track of uranium to evaporate the uranium particles, and carry the evaporated uranium particles Into the nitric acid absorption solution. The uranium particle separation device of the present invention includes a sample moving platform, a sample chamber is arranged above the sample moving platform, a laser for generating laser beams is located above the sample chamber, one side of the sample chamber is connected with an air filter, and the other side is connected with an absorption tube The air inlet is connected, the gas outlet of the absorption pipe is connected with the atmospheric sampler, and the nitric acid absorption liquid is filled in the absorption pipe. Through the uranium particle separation method and device of the invention, the rapid separation of uranium particles can be realized, and the operation is simple and convenient.

The invention discloses a method for calculating a multi-group slow luminescence photon yield by using a fission yield and decay data. The method comprises the following steps: firstly, calculating a multi-group decay photon yield of each nuclide based on a decay photon energy spectrum given in a decay sub-library of an evaluation nuclear database; constructing decay information of each nuclide according to the decay mode and the branching ratio data; secondly, fission yield data under different neutron incident energies are selected according to different requirements; then constructing decay chain information of all fission products until each fission product decays to a stable nuclide; according to fission yield data, adding decay photons generated in the process of decaying fission products to stable nuclides in different energy ranges, so as to obtain the multi-group slow luminescence photon yield of the fission nuclides; and finally, in order to ensure the accuracy of heat release calculation of the slow luminophores, correcting the yield of the multi-group slow luminophores obtained by calculation by using the average fission slow luminophore energy given in the evaluation kernel database. The method is high in calculation precision and calculation efficiency.

The invention discloses a multi-source heterogeneous data fusion method based on the Internet of Things, and the method comprises the steps that a DTS collects the data of a plurality of IoT devices,wherein the data comprises the type identifiers of different IoT devices; the DTS performs data combination on the IoT device data, and defines the data combination as a source data set; the DTS performs primary fission on part of IoT data in a source data set based on the type identifiers of the IoTs and the encryption levels corresponding to the type identifiers, and inserts the fission IoT datainto the source data set to form a primary fission data set; the DTS performs secondary fission on part of IoT data in the primary fission data set, and inserts the fission IoT data into the primaryfission data set according to a random insertion algorithm to form a secondary fission data set; the DTS performs tertiary fission on part of IoT data in the secondary fission data set, encrypts the fission IoT data according to an asymmetric encryption algorithm, and inserts the fission IoT data into the secondary fission data set according to a random insertion algorithm to form a tertiary fission data set; and the DTS sends the tertiary fission data set to the core layer.

The invention relates to a method for eliminating neutrons from fission, fusion or aneutronic nuclear reactions in a reactor (100), in particular in a laser-driven nuclear fusion reactor (100) which operates with hydrogen and the boron-11 isotope, in which method at least some moderated neutrons are made to undergo a nuclear reaction with tin (11). As a result of the nuclear reactions with tin, the neutrons convert the tin nuclei into stable nuclei having a higher atomic weight resulting from neutron capture. The invention also relates to a reactor (100) which is designed for energy conversion by means of fission, fusion or aneutronic nuclear reactions and for generating electric energy, wherein the reactor contains a neutron elimination device (50) which contains tin and is arranged such that at least some moderated neutrons are made to undergo a nuclear reaction with the tin.

The invention provides a method and a system for producing Mo-99 by using a liquid molten salt reactor. The method comprises the following steps: providing a liquid molten salt reactor, wherein a plurality of graphite moderation assemblies containing channels are arranged in the reactor core, the channels of the graphite moderation assemblies are filled with molten salt composed of low-enriched uranium and base salt, and Mo-99 is generated by fission in the liquid molten salt reactor. When the liquid molten salt reactor runs, an on-line solid-liquid separation method is adopted to separate indissolvable solid fission products on line, then a cooling method is adopted to reduce the radioactive activity of the indissolvable solid fission products, and finally a chemical separation method is adopted to separate and recover Mo-99 from the indissolvable solid fission products, so that the preparation of Mo-99 is realized. According to the invention, the production efficiency is improved, operation is convenient and fast, the economic cost is low, the fuel demand quantity is low, the radioactive shielding requirement is low, and the current Mo-99 supply demand problem can be effectively solved.

The invention discloses an organic-matter-rich mud shale formation fluid pressure coefficient prediction method and device, a storage medium and computer equipment, and the method comprises the steps: based on apatite fission track age and length distribution data, through fission track length simulation, determining the uplift time and the uplift denudation process, and according to ancient temperature scale Ro data, through ancient temperature scale inversion, calculating the fission track length distribution data; the method comprises the following steps: determining the uplift time, determining the denudation amount, on the basis of the uplift time, the denudation amount and denudation amount recovery, obtaining the maximum vertical effective pressure endured by the shale in the geological historical period through burial history recovery, and after the maximum vertical effective pressure and the current vertical effective pressure are determined, determining the OCR value of the shale. According to the method, the formation fluid pressure coefficient of the shale rich in organic matters is predicted by utilizing the OCR value according to the mathematical model, so that a key parameter index is provided for geological evaluation and engineering evaluation of exploration.

The invention discloses a concentration method and device for a low-specific-activity <99m>Tc solution. The method comprises the following steps: adsorbing low-specific-activity < 99 > Mo (containing a large amount of < 98 > Mo and < 100 > Mo) feed liquid by using an acidic alumina column, generating enough < 99m > Tc after < 99 > Mo is decayed for a certain time, leaching < 99m > Tc by using normal saline, and concentrating the low-specific-activity < 99m >Tc normal saline solution by using an anion resin column and an AgCl column to obtain a high-specific-activity < 99m >Tc solution. According to the method, the popularization and application of the non-fission < 99 > Mo production method can be promoted, the problem of application of the low-specific-activity < 99m >Tc solution in nuclear medicine clinic is solved, and the service life of a fission < 99 > Mo-< 99m >Tc generator can also be prolonged.

The invention discloses a UO2 / Zr nuclear fuel plate fission mechanical analysis finite element model. The model is composed of a Zr alloy matrix, two UO2 core particles and fission air holes wrappingUO2 cores, wherein the UO2 / Zr nuclear fuel plate fission model is of a 1 / 4 circular fan-shaped thin plate structure; a Zr alloy cladding is arranged outside the fuel plate, a UO2 core body is arrangedinside the fuel plate, and meanwhile small particles mixed with UO2 and Zr, air holes and the like are included. Pores, UO2 and Zr particles are assumed to be spherical; and in order to apply a surface load to the LS-DYNA, a sheet three-dimensional body is established by taking a sphere center as a central point as a geometric model, and only one layer of unit is divided in the thickness direction, so that the finite element calculation amount is greatly reduced, and the finite element solving speed is increased.

The invention relates to the technical field of fission type nuclear reactors, and especially relates to an irradiation target for producing a molybdenum-99 isotope in a heavy water reactor. The irradiation target comprises a fuel rod bundle; the fuel rod bundle comprises a plurality of fuel elements and end plates welded to the two ends of the fuel elements; and each fuel element comprises a cladding, a uranium-containing core arranged in the cladding and end plugs welded to the two ends of the cladding, the uranium-containing core in at least one fuel element is a rich uranium core provided with rich uranium fuel, and the <235>U enrichment degree of the rich uranium fuel ranges from 6.0 wt% to 20.0 wt%. Compared with the prior art, the method disclosed by the invention has the advantages that the characteristic that the heavy water reactor is refueled without stopping the reactor is fully utilized, the <99>Mo with short half-life period can be continuously produced by utilizing the existing reactor, a new irradiation facility does not need to be specially constructed, and the <99>Mo produced by using the enriched uranium is high in efficiency and good in quality, namely high in specific activity; and when the irradiation target is used for producing <99> Mo, the influence on power generation of a nuclear power plant can be reduced to the maximum extent.

The invention relates to the technical field of nuclear reactor cores, and particularly discloses a method for accelerating solving of a generalized conjugate neutron transport equation. The method comprises the following steps: respectively solving a neutron transport equation and a conjugate neutron transport equation by utilizing a characteristic line method, and respectively obtaining corresponding neutron flux distribution; calculating to obtain a source item of the generalized conjugate equation according to a response which specifically needs to be solved, and constructing a generalizedconjugate neutron transport equation; constructing a fixed source solver to solve a generalized conjugate neutron transport equation; and solving to obtain a generalized conjugate neutron transport equation. According to the method, the solving accuracy of the generalized conjugate neutron transport equation is ensured through characteristic line scanning, and the high efficiency of the generalized conjugate neutron transport equation is ensured through coarse net finite difference acceleration; when the method is used for solving the sensitivity coefficient of physical parameters such as thereaction rate ratio and the average fission power of the reactor to nuclear data, the calculation time can be remarkably shortened, and the efficiency is improved.

The invention discloses a three-energy-segment gamma imaging system based on energy window modulation and a system design method, and the imaging system comprises a radiation source, a coding hole, a gamma imaging module I, a gamma imaging module II and a gamma imaging module III which are sequentially arranged in series. The computer control equipment is connected with the gamma imaging module I, the gamma imaging module II and the gamma imaging module III; geometric centers of array scintillators in the radiation source, the coding hole, the gamma imaging module I, the gamma imaging module II and the gamma imaging module III are all located on the axis of the ray channel; the invention further discloses a design method of the imaging system. By means of the imaging system, image diagnosis of the three-energy-band gamma rays in the fission source area can be achieved. Spatial distribution information of gamma rays of the fission source region with a multi-energy-spectrum structure can be obtained through single measurement of the imaging system, and the diagnosis efficiency of gamma images of the fission source region is improved.