69 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 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.

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 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 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 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 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.

The invention relates to a method for estimating the number of critical accident fission times of an inner core of a cylindrical uranium solution storage tank. By adopting the method for estimating the number of critical accident fission times of the inner core of the cylinder uranium solution storage tank, according to geometric structure parameters of the uranium solution storage tank, material component parameters of contents of the uranium solution storage tank and technological process parameters. a three-dimensional Monte Carlo neutron transport program is used for modeling and calculating to obtain a reactivity rate [rho]t added into a system when a critical accident occurs, and then the fission frequency during the nuclear critical accident is estimated by combining a parameter change trend obtained by fitting early-stage experimental data. Compared with a simple empirical formula, the estimation value obtained by the method provided by the invention is more reasonable. Meanwhile, only a common three-dimensional Monte Carlo neutron transport program needs to be utilized, a neutron dynamics calculation program special for nuclear critical accident simulation does not need to be developed and used, and estimation of the nuclear critical accident is more convenient.