93 results about "Neutron transport" patented technology

A method and related system of determining formation density by compensating actual inelastic gamma rays detected from a pulsed-neutron tool for the effects of neutron transport. The method and systems may model response of the tool and use the modeled response as an indication of an amount to compensate detected inelastic gamma rays.

Disclosed is a method for calculating core neutron flux distribution of a small experimental reactor. The method includes steps of 1), determining geometrical and material parameters according to a core structure of the small experimental reactor, and establishing a neutron-transport equation describing movement rules of neutrons in each discrete direction inside a reactor core; 2), preparing corresponding quadrature sets for boundaries unparallel in normal surface vectors and each coordinate axis direction; 3), subjecting the neutron-transport equation to numerical discretion to acquire simultaneous linear algebraic equations of angular neutron flux density by a segment method of approximating unstructured geometry with unstructured grids to establish arbitrary triangular grids; 4), solving the simultaneous linear algebraic equations to acquire the discrete distribution of the angular neutron flux density in the reactor core, and acquiring the discrete distribution of neutron-flux density in the reactor core by the aid of relation of the angular neutron flux density and the neutron-flux density. By the method, the neutron-flux distribution of the small experimental reactor, especially that of an isotope production reactor, a pebble bed reactor and a high-flux reactor in the medical field can be accurately acquired.

A calculation method for fast reactor neutron transportation burn-up coupling analysis comprises the steps of 1, dividing a radial structure of a reactor core into a triangular grid, establishing a triangular prism grid, dividing burn-up regions with axial segmentation of a subassembly as a unit, dividing internal cycle of a reactor into a plurality of burn-up steps, and executing the following steps for each burn-up step; 2, calculating a small group of macroscopic cross-sections of each burn-up region, and carrying out neutron transportation calculation by use of a neutron transportation calculation method based on the triangular prism grid; 3, calculating a burn-up matrix of each burn-up region at the beginning of the burn-up step, and solving a burn-up equation by use of a Chebyshev rational approximation method; 4, carrying out neutron transportation calculation on the reactor core according to a nucleon density vector of each burn-up region at the end of the burn-up step; 5, averaging the burn-up matrixes at the beginning and at the end of the burn-up step to obtain an average burn-up matrix of each burn-up region, and carrying out burn-up calculation on each bur-up region again from the beginning of the burn-up step; and 6, repeating steps 4 and 5 until the nucleon density vectors of each burn-up region at the end of the burn-up step which are obtained through two adjacent calculations are converged.

The present invention discloses a method for obtaining sensitivity coefficients of an effective multiplication factor to a section under different burnups. The method comprises: 1, performing forward burnup calculation: first, using a subgroup method to calculate an effective self-shielding section of each nuclide, second, using a modular characteristic line method to solve a neutron angle flux density and a neutron conjugate angle flux density, and third, using a chebyshev rational approximation method to calculate a nuclear density of each nuclide; 2, performing conjugate burnup calculation: first, using the chebyshev rational approximation method to calculate an initial conjugate nuclear density of each nuclide, and then calculating a conjugate power, second, using the modular characteristic line method to calculate a generalized neutron transport angle flux and a generalized conjugate neutron transport angle flux, and third, calculating a conjugate initial nuclear density of each nuclide of a next step; and 3, calculating sensitivity coefficients of the section of each nuclide to an effective multiplication factor under different burnups. The method provided by the present invention solves a defect of the existing method that sensitivity coefficients of an effective multiplication factor to a nuclear section under different burnups cannot be accurately and effectively calculated.

Systems, methods, and devices are provided to determine an accurate neutron-gamma density (NGD) measurement for a broad range of formations, including low-hydrogen-index or low-porosity formations and formations with heavy elements. For example, such an NGD measurement may be obtained by emitting neutrons into a formation such that some of the neutrons inelastically scatter off elements of the formation and generate inelastic gamma rays. The neutrons and inelastic gamma rays that return to the downhole tool may be detected. Some characteristics of certain formations are believed to affect the fast neutron transport of the formations. Thus, if a formation has one or more of such characteristics, a correction may be applied to the count rate of neutrons, the count rate of inelastic gamma rays, or the neutron transport correction function, upon which the neutron-gamma density (NGD) may be determined.

Provided is a high-accuracy fast neutron reactor assembly few-group cross section obtaining method. The fine-group cross section of a fast neutron reactor assembly is calculated in the mode that point cross section data and multi-group data are combined, online calculation of an elastic scattering matrix is performed, a neutron transport equation is solved by using the fine-group cross section to obtain fine-group neutron-flux density, energy groups are merged on the basis, and accordingly few-group parameters of the fast reactor assembly are obtained. Due to the fact that the method directly uses the point cross section data, the elastic scattering resonance effect of a medium mass nuclide and the resonance interference effects of all nuclides are accurately considered. The neutron-flux density of a real system is adopted in the energy group merging process, and merging results are more accurate. The overall errors of fast reactor assembly few-group parameters calculated by adopting the method are within 1% compared with a reference value, and the method has higher accuracy.

The invention discloses a method for accurately calculating space-time neutron distribution in a nuclear reactor. The method comprises the following steps of 1, directly solving a steady-state neutron transport equation by using a one-step method to obtain neutron flux density distribution and precursor concentration in a steady state; 2, solving a space-time neutron transport equation within 0.25ms, obtaining an amplitude value of 0.25ms through neutron flux density and a neutron flux density shape function at the moment of 0.25, and taking the amplitude value as an initial value of point reactor calculation; and 3, calculating the space-time neutron transport equation by using the one-step method in a large step length to obtain a predicted solution, solving a point reactor equation in an extremely small step length, and correcting the neutron flux density by using the amplitude value of a point reactor. By improving an existing prediction and correction quasi-static method, on the premise of not remarkably prolonging calculation time, rapid change of neutrons of the nuclear reactor in an extremely short time can be accurately captured, and the defects of the conventional prediction and correction quasi-static method are radically overcome, so that the method provided by the invention can play a real role in high-fidelity space-time neutron dynamics calculation.

The invention discloses a restarting method and system for making a few-group cross section used in a reactor nuclear design. The method comprises the following steps: (a) acquiring the burn-up depth of a fuel assembly; (b) acquiring changes of one or more parameters of the fuel assembly under the condition of the burn-up depth; (c) executing parameter preprocessing directed at the changes of the one or more parameters; and (d) executing neutron transport calculation based on results of the parameter preprocessing so as to obtain few-group cross sections. The system comprises a burn-up depth acquisition unit which is used for acquiring the burn-up depth of the fuel assembly, a fuel parameter acquisition unit which is used for acquiring the changes of the one or more parameters of the fuel assembly under the condition of the burn-up depth, a parameter preprocessing unit which is used for executing parameter preprocessing directed at the changes of the one or more parameters, and a few-group cross section acquisition unit which is used for executing neutron transport calculation based on the results of the parameter preprocessing so as to obtain the few-group cross sections.

The invention discloses a three-dimensional neutron flux numerical simulation method for a pressurized water reactor core. The method comprises the following steps: firstly, dividing a three-dimensional pressurized water reactor core to be simulated into a plurality of layers along the axial direction, and establishing a two-dimensional neutron transport model for each layer based on a characteristic line method; secondly, geometrically dividing a to-be-simulated three-dimensional pressurized water reactor core into a plurality of long strips based on radial lattice cells, and establishing a one-dimensional neutron transport model for each long strip based on a discrete longitudinal marking method; and iterating a residual error model formed by the two-dimensional neutron transport model and the one-dimensional neutron transport model through a JFNK method until convergence to obtain neutron flux distribution of the three-dimensional pressurized water reactor core. Compared with the prior art, the two-dimensional neutron transport model and the one-dimensional neutron transport model are converted into the residual error model to be solved at the same time, the second-order convergence speed is achieved, and the iterative solution stability is good. The method can be used for transport module calculation of a numerical reactor, the numerical reactor transport calculation efficiency is improved, the calculation stability is improved, and the nuclear time cost generated by numerical calculation is saved.

The invention relates to a tectonic entity geometric modeling method suitable for a multi-heap geometry and characteristic line method, which comprises the following steps of: (1) defining a geometricunit and Boolean operation of the geometric unit, and giving a description mode of geometric input data; (2) designing an interface for reading an input file, abstracting each geometric unit in the input file into an object-oriented data type, and carrying out geometric modeling; and (3) according to the constructed geometric model, realizing searching of the bottommost material element where thepoint is located, and realizing generation of a flat source approximate region. Geometric structures of different reactor types of reactors can be described according to a specified input rule, a geometric model is stored as object-oriented data through an interface, and geometric preprocessing operation can be completed for solving a neutron transport equation by using a characteristic line method subsequently.

The invention relates to a computing method of a neutron flux spectrum in a reactor, particularly to a computing method of an absolute neutron flux spectrum of a refrigerant in a reactor core active section. According to the computing method, a method combining reactor core neutron transport program with Monte Carlo program is used, the reactor core computing program is used for computing the absolute value of average flux of the reactor core, the flux ratio of the refrigerant and the reactor core, and the neutron flux distribution of a customized energy group in the refrigerant are generated through analog computation of the Monte Carlo program, and the absolute neutron flux spectrum distributed in different energy groups of the refrigerant in the reactor core active section is finally computed, so that the computing method of traditionally using the value of average neutron flux to compute related source items of the reactor is improved, and the computing accuracy is improved.

The invention discloses a method for calculating an axial swelling effect of a fast neutron reactor assembly. The method comprises the following steps of 1, separating a transport computation grid from a burnup computation grid, wherein the transport computation grid is kept unchanged in the whole computation, and the burnup computation grid keeps growth of a same proportion along with axial swelling of a fuel assembly; 2, based on a variational nodal method, establishing a weak form of a neutron transport equation solution, wherein a section in a node is a function related to a spatial position, so that appearance of various materials is allowed in a transport computation node; 3, expanding a flux by utilizing a spherical harmonic function and a spatial orthogonal polynomial, and solving a flux expansion moment by adopting a response matrix method so as to obtain flux distribution of each node; and 4, calculating a burnup level of the node according to average power of a burnup node, obtaining a homogenized section of the burnup node according to burnup interpolation, and adjusting the homogenized section in the burnup node according to axial elongation of the assembly to obtain a required homogenized section in the computation node.

The invention belongs to the technical field of nuclear power, and particularly relates to a reconstruction method and device for burn-up characteristics of reactor core lattice cells. The neutron flux of each energy group in each segment divided in the axial direction of the reactor core assembly and the microscopic cross section of each energy group are determined by adopting a neutron diffusionmodel, and microscopic burnup calculation of a lattice cell level fine grid is carried out on the reactor core by adopting a modulation method. And thus, parameters such as the microscopic reaction rate and the important nuclide nucleon density in the lattice cell can be obtained more accurately. Compared with a neutron transport model, the neutron diffusion model is small in calculation amount and high in speed, so that a large number of calculation resources can be saved, efficient calculation can be realized, parameters such as the microcosmic reaction rate and the important nuclide nucleon density in a lattice cell can be obtained more accurately, and the accuracy and efficiency of reactor core burn-up characteristic calculation are greatly improved.

The invention discloses a dual non-uniform space self-shielding effect correction method, a device, equipment and a medium. The method comprises the steps: calculating the number of neutrons escapingfrom the surface of fuel and the number of neutrons entering adjacent fuel lattice cells for collision in a moderator after escaping from the surface of the fuel through a Monte Carlo method, and obtaining the number of neutrons escaping from the surface of the fuel; obtaining a Dankov factor, and then correcting the neutron escape probability through the Dankov factor so as to correct the dual non-uniformity of neutron resonance calculation; according to the fact that the escape probability of neutrons in the randomly distributed medium area before and after homogenization is not changed, enabling the randomly distributed medium area to be equivalent to a homogenization medium; calculating the equivalent cross section of the homogenizing medium based on the macroscopic cross sections of the dispersed particles and the sub-regions of the matrix and the volume shares and the space self-shielding factors corresponding to the sub-regions, so as to correct the dual non-uniformity of neutron transport calculation and realize the correction of the dual non-uniformity space self-shielding effect of the randomly distributed medium fuel element. The neutron calculation accuracy is improved.

The invention discloses a floating grid-based nuclear-thermal-mechanical multi-physical coupling calculation method. The method comprises the steps of establishing a fuel rod geometric model, carryingout a nuclear-thermal-force multi-physical iterative calculation process, carrying out plane heat transfer calculation, plane mechanics calculation and plane neutron transport calculation, and carrying out axial one-dimensional neutron diffusion calculation. According to the invention, neutronics, thermal hydraulic and mechanical problems are solved based on deformable floating grid coupling, andbehaviors of heat conduction, mechanical deformation, neutron transport, nuclide burnup and the like of a fuel rod in an extreme environment in a reactor are simulated; and meanwhile, multiple physical states such as temperature, stress, deformation, in-rod pressure, neutron group flux, power, burnup and isotope density of the fuel rod are obtained.

The invention discloses a method for simulating the three-dimensional neutron flux of a pressurized water reactor core based on axial expansion. The method comprises the following steps: firstly, dividing the three-dimensional pressurized water reactor core to be simulated into a plurality of layers along the axial direction, expanding the three-dimensional neutron angular flux for each layer, andapproximately simulating the axial change of the neutron flux by adopting a linear model in the axial direction; establishing a two-dimensional neutron transport model based on a characteristic linemethod in the radial direction according to the three-dimensional neutron flux after axial linear expansion; obtaining the relationship between the zero-order neutron angular flux and the first-orderneutron angular flux according to boundary conditions so as to simplify the two-dimensional neutron transport model established by the characteristic line method; and finally, solving the simplified two-dimensional neutron transport model along each layer from bottom to top in the axial direction to obtain neutron flux distribution of the three-dimensional pressurized water reactor core of each layer. Compared with the prior art, neutron leakage items generated by transverse integration in the two-dimensional neutron transport model and the one-dimensional neutron transport model are avoided,the solving stability is high, the method can be used for transport module calculation of a numerical reactor, and the stability of numerical simulation is improved.

The invention relates to a load balancing parallel method of a three-dimensional neutron transport characteristic line method. The method is based on geometric symmetry. The method comprises steps ofcalculating the number of angles needing to be considered; calculating the number and length of periodic tracks or reflective tracks at each angle; calculating the load of each process according to the length of the periodic or reflective track, the number of the processes and the number of the tracks, performing ray tracing on each process, and iteratively solving a characteristic equation according to the geometric grid information, completing the material data and the line segment information after ray tracing. According to the method, the periodic trajectory and the reflective trajectory are completely allocated to the threads, parallel tasks can be efficiently allocated, load balance is ensured, and point-to-point communication between the processes is eliminated.

The invention provides a neutron transport acceleration method, electronic equipment and a computer readable storage medium, and the method comprises the steps: building a multi-group gCMFD equation equivalent to an initial neutron transport equation based on an initial parameter set of the neutron transport equation; performing iterative acceleration processing on the neutron transport equation through the multi-group gCMFD equation in a spatial dimension until iteration meets a first convergence condition, and establishing an equivalent single-group gCMFD mode equation; and performing iteration acceleration processing on the multi-group gCMFD equation through the single-group gCMFD equation in an energy dimension until iteration meets a second convergence condition. According to the technical scheme, the equivalence between the multi-group gCMFD equation and the neutron transport equation and the equivalence between the single-group gCMFD equation and the multi-group gCMFD equation and the stability of acceleration calculation are strictly guaranteed, neutron transport calculation can be accelerated more quickly, and the overall calculation efficiency is improved.