12 results about "Neutron transport" patented technology

This invention discloses a neutron transport calculation system with decoupled geometric modeling and mesh generation, including a mesh generator and a CSG geometric modeling module. The module is fully embedded into the computational framework of the neutron transport procedure using the feature line method. First, a suitable type of mesh generator is selected based on the geometry of the underlying material cell. The internal structure of the mesh generator is completed by setting the meshing parameters. Subsequently, the mesh generator is set as an attribute of the material cell. The mesh generator's built-in positioning algorithm can quickly determine the fine mesh index of the ray segment endpoints; its intersection algorithm can accurately calculate the intersection distance between the ray and the fine mesh. Thus, all feature line segment information is generated. Finally, source iteration is performed until the convergence condition is met to obtain the key physical parameters. By using this invention, fine mesh generation can be achieved, which is helpful for subsequent reactor parameter acquisition. This invention can be widely applied in the field of nuclear reactor physics numerical calculation technology.

This invention relates to a method for neutron transport calculation or resonance calculation for dispersed fuels. The method is characterized by correcting the matrix chord length probability and average chord length used in related calculations based on the true chord length probability distribution and true average chord length of the matrix, and pre-constructing an interpolation table or fitting formula for solving the true average chord length correction factor using dimensionless parameters. The method for solving the true chord length probability distribution and true average chord length of the matrix includes the following steps: setting the geometric shape of a random medium region according to the fuel type; generating a random distribution of particle positions within the given random medium region according to a given filling rate; randomly selecting particle surfaces to isotropically emit neutrons outwards, or uniformly and isotropically emitting neutrons from the matrix, and calculating the distance to the next particle; statistically analyzing the chord lengths of all neutrons passing through the matrix, thereby obtaining the true chord length probability distribution and true average chord length of the matrix.

This application discloses a method, apparatus, electronic device, and storage medium for calculating fuel sphere burnup, relating to the field of nuclear reactor physics and thermal calculation technology. The method includes: dividing fuel spheres into multiple batches based on their distribution; establishing neutron transport equations for each batch of fuel spheres; calculating the self-shielding factor of each batch of fuel spheres; calculating the neutron flux density of each region of the reactor core using the homogenized cross section of the pebble bed region and the homogenized cross section of the reflector layer region; calculating the power density of each region of the pebble bed based on the neutron flux density of each region of the reactor core; calculating the neutron flux density and power density of each batch of fuel spheres at the current reactor core power level based on the neutron flux density of the pebble bed region and the self-shielding factor of each batch of fuel spheres; and calculating the burnup depth of the corresponding batch of fuel spheres based on the power density of each batch of fuel spheres at the current reactor core power level, thereby improving the technical accuracy of fuel sphere burnup calculation in non-uniformly packed pebble bed reactors.

This description relates to a method (300) for determining, by a computing device, a fuel salt composition of a molten salt nuclear reactor, the method comprising: - the determination of neutron fluxes (ϕ(E)) of several energies and a multiplication factor by a step of solving (302) a neutron transport equation; - the determination, for each of the nuclides, of condensed microscopic cross sections from the neutron fluxes (ϕ(E)); - the determination (303) of the weights of the nuclides on the multiplication factor from the condensed microscopic cross sections; and - the determination (304), from the weights of the nuclides, of a fuel salt inventory including at least one calculation of the evolution of the nuclides, by a step of solving (306) evolution equations, such that the evolution inventory complies with at least one operating condition.

This invention discloses a method, electronic device, and computer storage medium for reactor transport calculation. The method includes invoking multiple parallel processes to calculate neutron transport response matrices for multiple regions within the reactor core. Each region includes several grids allocated to the associated processes, and each grid corresponds to a neutron transport response matrix. A set of basis vectors is set to accelerate the algorithm. These basis vectors include the neutron flux of the grids contained in the multiple regions, a first incident flow at the outer boundary of the core, a second incident flow in a first grid group, and core eigenvalues. The grids in the first grid group are non-adjacent. Multiple processes are invoked to use the accelerated algorithm to perform iterative neutron transport calculations based on the neutron transport response matrices and basis vectors, obtaining target values ​​for the neutron flux and core eigenvalues. This application effectively solves the problem of decreased solution speed while maintaining computational accuracy during neutron transport calculations.

This disclosure provides a method, apparatus, electronic device, and storage medium for homogenizing the cross-sectional parameters of a pebble bed reactor, relating to the field of nuclear reactor technology. It obtains the individual cross-sectional parameters of various fuel spheres in the reactor core to reflect the individual neutron reaction characteristics of different fuel spheres. Then, based on the actual mixing state of the fuel spheres within the core region, it calculates the neutron collision probability between different fuel sphere groups to match the actual neutron transport scenario in the reactor core. Subsequently, based on this collision probability, it establishes and solves the neutron flux density distribution equation to obtain the accurate neutron flux density of each fuel sphere group under different energy groups. Finally, based on the principle of reaction rate conservation, it combines the volume fraction of each fuel sphere group with the corresponding neutron flux density for weighted fusion to generate regional homogenized cross-sectional parameters. Therefore, it can solve the problems of inaccurate regional homogenized cross-sectional parameters and large calculation errors in the core neutron flux and power distribution in existing technologies.

The present application belongs to the technical field of nuclear reactor core, and particularly relates to a self-sufficient detector arrangement method suitable for a modular small-sized pressurized water reactor. The method comprises the following steps: determining the arrangement space of the self-sufficient detector in the assembly; establishing the influence function value of the self-sufficient detector at each different position; calculating the three-dimensional core calculation modeling by adopting the whole reactor neutron transport calculation; randomly establishing a three-dimensional arrangement scheme set of the self-sufficient detector, and obtaining the normalized theoretical current of the assembly where no measuring point is located; calculating the root mean square error of the relative deviation of the normalized theoretical current; setting the root mean square error acceptance limit value, and selecting the self-sufficient detector arrangement scheme with the least number of detectors and the root mean square error lower than the limit value as the optimal scheme. The beneficial effect lies in that the reasonable arrangement of the self-sufficient detector is realized in the limited space, so as to obtain the detector arrangement scheme satisfying the power distribution precision of the online monitoring.

Method for determining the fuel salt composition of a molten salt reactor. This description relates to a method (300) for determining, by a computing device, the fuel salt composition of a molten salt nuclear reactor, the method comprising: - the determination of neutron fluxes (ϕ(E)) of several energies and a multiplication factor by a step of solving (302) a neutron transport equation; - the determination, for each of the nuclides, of condensed microscopic cross sections from the neutron fluxes (ϕ(E)); - the determination (303) of the weight of the nuclides on the multiplication factor from the condensed microscopic cross sections;and- the determination (304), from the weights of the nuclides, of a fuel salt supply inventory including at least one calculation of the evolution of the nuclides, by a step of solving (306) evolution equations, so that the evolution inventory respects at least one operating condition. Figure for the abstract: Fig. 3;

The present application belongs to the technical field of plate fuel temperature model, and particularly relates to a plate fuel effective temperature model construction method based on high-fidelity multi-physical field data mapping, which comprises the steps of establishing a fine three-dimensional geometric model, grid division and multi-physical field domain decomposition, setting thermal hydraulic boundary conditions, multi-physical field coupling calculation, and constructing an effective temperature model based on the interpolation idea. Based on the Cardinal framework of the MOOSE platform, a customized coupling is established between the continuous energy neutron transport solver and the solid heat conduction module, the moderator feedback is isolated by adopting the frozen fluid attribute strategy, a high-resolution intra-plate temperature field is generated for the plate fuel grid, and the Doppler reactivity is accurately evaluated. The interpolation type effective temperature model introduced in the present application quantifies the spatial self-shielding effect by introducing a dimensionless parameter γ, and taking the U3Si2-Al dispersed fuel in the standard MFR range as an example, the multi-scale verification from a single plate to a five-plate subassembly proves that the model has good universality.

A method for calculating multi-group data of nuclear fuel considering resonance upscattering effects includes: 1. For nuclides in the nuclear fuel with resonant scattering cross-sections, a nuclear data processing program generates a resonance upscattering correction factor table for their total cross-section, fission cross-section, and absorption cross-section; 2. A fuel assembly calculation program performs resonance calculations to obtain the average multi-group cross-section of the nuclides, and interpolates the average multi-group cross-section in a traditional multi-group cross-section table that does not consider resonance upscattering to obtain the corresponding dilution cross-section; 3. Using the dilution cross-section, interpolation in the resonance upscattering correction factor table yields a correction factor, which is used to correct the average multi-group cross-section, resulting in a nuclear fuel multi-group cross-section considering resonance upscattering effects. This method solves the problem that fuel assembly calculation programs using ultrafine group methods for resonance calculations cannot consider resonance upscattering effects, providing accurate multi-group data for neutron transport calculations and improving the calculation accuracy of nuclear reactor safety parameters such as the effective multiplication coefficient and fuel temperature coefficient.

This invention discloses a method for constructing a neutron transport order reduction model based on component-level domain decomposition coupling. It employs a core architecture of "divide and conquer, numerical flux coupling." In the offline stage, "operational condition vectors" are extracted from fuel components as subdomains, and their dimensionality is reduced through sampling and independent solving. This constructs a dedicated neutron angular flux basis function library categorized by component type, energy group, and orientation angle. In the online stage, the entire reactor core is treated as a component network. Low-dimensional expansion is performed within each subdomain using basis functions. Based on the upwind numerical flux mechanism of discontinuous Galerkin (DG), robust coupling with physical conservation between components is achieved. Finally, a global block sparse linear equation system is assembled and solved to reconstruct the full-core angular flux distribution. This significantly reduces the offline cost to an acceptable range, ensuring high-precision reconstruction and numerical stability of the method for local non-uniformity. While maintaining high fidelity accuracy, it achieves a computational efficiency improvement of more than three orders of magnitude.