13 results about "Irradiated materials" patented technology

The invention relates to the technical field of nuclear reactor fuel element analysis, in particular to an irradiated zirconium alloy plasticity limit prediction method based on a crystal plasticity finite element, which comprises the following steps: establishing a three-dimensional polycrystalline representative body cell model conforming to the actual grain size of zirconium alloy; based on a crystal plasticity theoretical model, zirconium alloy material parameters are obtained through calculation; boundary conditions and load conditions are determined, finite element calculation is executed in combination with material parameters, and an element stiffness matrix of each increment step is output; assembling the unit stiffness matrix, and obtaining a macroscopic stress-strain matrix of the three-dimensional polycrystal cell model through a homogenization method; and substituting the macroscopic stress-strain matrix into the Rice theoretical criterion, traversing the macroscopic stress-strain matrix of each increment step, and determining the plastic strain meeting the judgment critical condition. According to the method, accurate finite element plastic limit analysis can be carried out on the irradiation material, and the plastic limit strain of the zirconium alloy under the irradiation condition can be accurately judged.

Cutting techniques to split a single workpiece into two portions are disclosed herein. A cutting machine may be configured to rotate the workpiece relative to a cutting wheel. The cutting wheel may be configured to precisely and controllably move towards the center of the work piece, which may result in the workpiece being cleanly cut. The rotation of the workpiece and the movement of the cutting wheel may be controlled by a controller. In some embodiments, a tray may be disposed below the workpiece such that the tray is configured to safely catch a portion of the workpiece that has been cut off. Further, if the workpiece has contents housed within, the tray may safely catch the contents.

The embodiment of the invention relates to the technical field of mechanical property research of reactor irradiation materials, in particular to a method for predicting the slippage position of spiral dislocation of a material, and the method comprises the following steps: determining parameters of a slippage motion relation between irradiation defects generated after the material is irradiated and the spiral dislocation in the material; the spiral dislocation is divided into a plurality of dislocation sections; according to the determined parameters, determining a first probability of occurrence of kink pairs on the spiral dislocations caused by the irradiation defects and a second probability of diffusion of the kink pairs on the spiral dislocations; determining the type of each dislocation segment according to the first probability and the second probability; according to the type of each dislocation section, the slippage position of each dislocation section in the spiral dislocation is determined; and determining the slippage position of the spiral dislocation according to the slippage position of each dislocation section. The method provided by the embodiment of the invention is beneficial to improving the accuracy of description of the spiral dislocation slippage behavior influenced by irradiation, and further is beneficial to researching the mechanical properties of the material influenced by irradiation.

Provided is a radiation decomposition device and a radiation decomposition method that allow efficient decomposition of an irradiated substance while reducing the size of equipment as much as possible. A radiation decomposition device (100) according to the present invention is characterized by including: a radiation source (101) that emits radiation released from a sealed radioactive substance; a reaction tank (103) that contains a to-be-irradiated substance (104); and a heat transfer tank (102) that covers the radiation source (101) and transfers, to the reaction tank (103), heat generated by the decay of the radioactive substance. It is preferable that the radiation source (101) is: a radiation source formed of spent nuclear fuel or vitrified waste containing fission products extracted from spent nuclear fuel; a radiation source containing fission products extracted from spent nuclear fuel; a radiation source containing Co-60; a radiation source containing Cs-137; or a high-level radioactive waste liquid.

The invention provides an irradiation-resistant material as well as a preparation method and application thereof. The radiation-resistant material comprises a base body and a gradient composite coating arranged on the surface of the base body, the base body is an isostatic pressing graphite material, the gradient composite coating comprises a transition layer and a functional layer, and the transition layer is made of beta-SiC whiskers; the material of the functional layer is a SiC-Si3N4 composite layer, and the Si3N4 phase content in the functional layer is gradually increased from the position close to the transition layer to the position close to the outer side face, so that the SiC-Si3N4 gradient composite layer with continuous components is formed. The material has excellent corrosion (fused salt / helium) resistance, neutron irradiation resistance, thermal shock resistance and scouring resistance, is simple in preparation process and low in cost, is particularly suitable for reactor core complex structural parts of fourth-generation nuclear reactors such as fused salt reactors and high-temperature gas cooled reactors, and has a wide industrial application prospect.

The invention discloses a method for measuring temperature in a reactor based on SiC thermal diffusion coefficient change after neutron irradiation. The method comprises the following steps: S1, processing a SiC crystal material into a flaky sample; s2, packaging the SiC crystal sample and the to-be-irradiated material sample into an irradiation sample box, and then putting the irradiation sample box into a reactor for a neutron irradiation experiment; s3, after the neutron irradiation experiment is completed, taking out the irradiation sample box from the reactor, cutting and disintegrating the irradiation sample box, and taking out the irradiated SiC crystal sample; s4, putting the SiC crystal sample into thermal conductivity meter equipment, and measuring thermal diffusion coefficients of the SiC crystal sample at different temperatures; and S5, drawing a curve chart of the thermal diffusion coefficient of the SiC crystal sample after irradiation along with temperature change, wherein the inflection point of the curve is the irradiation temperature of the position of the SiC crystal sample in the reactor. The method has the beneficial effects that the neutron irradiation temperature in the reactor at the position of the SiC crystal can be efficiently and quickly obtained.

The invention discloses an energy resolution cluster dynamics solving method for multi-principal-element alloy irradiation damage. The method comprises the following steps: establishing dynamics ordinary differential equations of various defect clusters in physical time through a reaction mechanism; converting the concentration variable of the cluster in the dynamic ordinary differential equation set from a concentration domain to a logarithm domain to obtain a logarithm domain ordinary differential equation set; performing time marching calculation on the log-domain ordinary differential equation set by adopting a numerical integration method, and updating the defect concentration change rate in the log-domain ordinary differential equation set in the solving process to obtain a defect evolution result; according to the method, the cluster dynamics simulation process can be accelerated, the evolution process of defect clusters can be accurately simulated, the method is widely suitable for defect behavior research of various irradiation materials, and the method has remarkable calculation efficiency and physical accuracy.

The invention discloses a method for measuring temperature in a reactor based on SiC thermal expansion coefficient change after neutron irradiation. The method comprises the following steps: S1, processing a SiC crystal material into a sample with a columnar structure; s2, packaging the SiC crystal sample and the to-be-irradiated material sample into an irradiation sample box, and putting the irradiation sample box into a reactor for a neutron irradiation experiment; s3, after the neutron irradiation experiment is completed, taking out the irradiation sample box from the reactor, cutting and disassembling the irradiation sample box, and then taking out the irradiated SiC crystal sample; s4, putting the SiC crystal sample into thermal expansion measurement equipment, and measuring the change of the elongation of the SiC crystal sample along with the gradual increase of the temperature; and S5, calculating the change rate of the elongation of the SiC crystal sample gradually rising along with the temperature, and drawing a change curve of the instantaneous thermal expansion coefficient of the SiC crystal sample along with the temperature to obtain the neutron irradiation temperature of the position of the SiC crystal sample in the reactor.

The invention relates to the technical field of strength prediction, in particular to a continuous strength prediction method and device for an engineering size irradiation material. The method comprises the following steps: acquiring first strength data obtained by a non-irradiation tensile test of a first tensile sample at different size points under different size spectrum systems; inputting the cross sectional area of the scale distance section in the first tensile sample and the first intensity data into a non-irradiation intensity prediction model to obtain a first parameter set; obtaining second strength data obtained by a tensile test after irradiation of a second tensile sample at different size points; inputting the cross sectional area of the scale distance section in the second tensile sample, the first parameter set and the second intensity data into the post-irradiation intensity prediction model to obtain a second parameter set; and inputting the cross sectional area of the engineering size irradiation material into the fitted post-irradiation intensity prediction model to obtain predicted intensity data of the engineering size irradiation material. According to the technical scheme, the strength of the engineering size irradiation material can be continuously predicted.

This invention discloses a method for testing materials via online irradiation in a heavy water reactor, comprising the following steps: installing a concave base and guide rod; installing the irradiated material and steel wire rope into the groove of the irradiation device; installing the irradiation device onto a designated VFD; conducting sample irradiation; dismantling the irradiation device; cutting the steel wire rope to remove the irradiated sample; cutting the remaining steel wire and storing it as waste; restoring the equipment to its original state on the designated VFD; and decontaminating the irradiation and shearing devices. The advantages of this invention are: it allows the irradiation channel to be opened during high-power operation of the reactor core, enabling the sample to be placed inside and irradiated for a relatively short period (usually several weeks or months) before being retrieved for analysis and testing.

The present application relates to the technical field of strength prediction, in particular to a method and device for continuously predicting the strength of engineering size irradiated materials. The method comprises: obtaining first strength data of first tensile samples of different size points in different size spectra obtained through unirradiated tensile test; inputting the cross-sectional area of the gauge length section in the first tensile sample and the first strength data into an unirradiated strength prediction model to obtain a first parameter set; obtaining second strength data of second tensile samples of different size points obtained through tensile test after irradiation; inputting the cross-sectional area of the gauge length section in the second tensile sample, the first parameter set and the second strength data into an irradiated strength prediction model to obtain a second parameter set; and inputting the cross-sectional area of the engineering size irradiated material into the fitted irradiated strength prediction model to obtain predicted strength data of the engineering size irradiated material. The above technical solution can continuously predict the strength of the engineering size irradiated material.

The embodiment of the invention relates to the field of nuclear reactor device testing, in particular to a method for determining the temperature and time of a high-temperature annealing test of an irradiated material. According to the method, by determining first microdefect information of a material, specific characteristics of microdefects of the material after irradiation can be recognized, the irradiation damage state is quantified, first hardening information is determined accordingly, and correlation between the microdefects of the material and macroscopic performance is established; by determining second microdefect information under different temperature and time conditions, microdefect long-time evolution simulation under different high-temperature annealing conditions can be realized, test trials and errors are reduced, and selection of annealing conditions is optimized; according to the first hardening information and the second hardening information, the annealing temperature and time of material performance recovery can be determined, the number of times of actual annealing tests is further reduced, the test trial and error cost is saved, and the expected test effect is obtained more efficiently.