1462 results about "Nuclear reactor" patented technology

The present application relates to the field of nuclear reactor simulation. Disclosed are a fast simulation method and system for multi-physical field coupling and a storage medium. The method comprises: at intervals of a first preset interval time, collecting state information of each physical field of a nuclear reactor in a transient process, so as to obtain physical field snapshots of a plurality of physical fields, the transient process being simulated by a reactor simulation model corresponding to the nuclear reactor; stitching the physical field snapshots of the plurality of physical fields, so as to obtain a fully-concatenated physical field; constructing a matrix on the basis of the fully-concatenated physical field; performing order reduction processing on the matrix, so as to construct a timing model corresponding to the fully-concatenated physical field; and decomposing the timing model corresponding to the fully-concatenated physical field, so as to obtain timing models corresponding to the plurality of physical fields. The fast simulation method for multi-physical field coupling of the present application can reduce the inconsistency between different physical fields while increasing the computation speed.

The invention discloses a coupling iterative algorithm library and a reactor multi-physical field coupling calculation program construction method. The coupling iterative algorithm library comprises an open source library layer used for integrating a plurality of numerical calculation libraries and providing an input / output interface for a reactor multi-physics field coupling framework and each numerical calculation library; the core algorithm layer is used for integrating a plurality of coupling iteration algorithms and a plurality of acceleration algorithms related to each coupling iteration algorithm; the interface layer is used for providing a data interaction interface between a nuclear reactor multi-physics field coupling calculation program and the core algorithm layer; and the iterative algorithm template layer is used for forming an iterative solution process control template function based on the open source library layer, the core algorithm layer and the interface layer, and the iterative solution process control template function is used for constructing a nuclear reactor multi-physics field coupling calculation program based on the process control parameters set in the input file. According to the coupling iterative algorithm library, the development efficiency and the development quality of a nuclear reactor multi-physics field coupling calculation program can be improved.

A nuclear reactor is discussed herein. In some examples, the nuclear reactor may comprise an inner-core region, including liquid fuel, coolant / moderator, and a hydrogen vapor space, an out-of-core region surrounding the inner-core region, the out-of-core region including, an inner reflector adjacent to the inner-core region, the inner reflector including a first irradiation facility, and an outer reflector adjacent to the inner reflector, the outer reflector including a second irradiation facility.

The invention relates to the field of reactor data processing, and particularly discloses a numerical calculation method for solving a neutron transport equation. The method comprises the following steps: 1, describing the geometry of a reactor core of a nuclear reactor by utilizing a construction entity geometric modeling method, and subdividing a curved surface grid along a curved surface structure; 2, directly expanding a neutron flux distribution function in each curved surface grid by using a multivariate polynomial; 3, deriving a neutron transport equation and an integral residual equation of a boundary condition under each curved surface grid; 4, solving geometric integral parameters of each curved surface grid; step 5, establishing an adaptive linear system; and step 6, iteratively calculating and solving a maximum characteristic value and an unknown expansion coefficient.

The present application discloses a method and system for identifying internal surface defects of a nuclear reactor pressure vessel, a computer device, and a computer-readable storage medium. The method comprises: collecting image data of the surface of a pressure vessel; constructing an illumination angle attention mechanism, using an image in a data set to establish a feature map, and producing a spatial attention map on the basis of a spatial relationship of the feature map; and using the illumination angle attention mechanism to improve a YOLOv5 model, and identifying surface defects of the pressure vessel by means of the improved YOLOv5. The present application has the characteristics of high defect identification efficiency, low missed detection rate, etc. The present application can effectively cope with common problems when defect detection is performed under different illumination conditions. This precise defect localization and identification capability is critical for ensuring safe operation of pressure vessels, particularly during routine maintenance and inspection.

The invention discloses a reactor thermal hydraulic characteristic rapid prediction method based on dynamic mode decomposition and deep learning, and the method comprises the steps: obtaining thermal hydraulic characteristic data of a reactor under different operation conditions, and selecting steady-state operation data to establish a snapshot; initial dimension reduction is carried out on the snapshot through singular value decomposition, and snapshot representation after dimension reduction is obtained; constructing an order reduction model based on the snapshot representation after dimension reduction; wherein the step of constructing the reduced-order model comprises the steps of performing dynamic mode decomposition on the snapshot after dimension reduction, constructing a parameterized low-dimensional linear operator, and fitting a mapping relation between the operator and an operation parameter through deep learning; verifying the reduced-order model according to a simulation result of the nuclear reactor mechanism model, and obtaining a parameterized reduced-order model when the verification is passed; and inputting the actual operation condition data into the parameterized reduced-order model to obtain a prediction result of the thermal hydraulic characteristics of the reactor. According to the method, the simulation efficiency is remarkably improved by combining the dynamic mode decomposition and the deep learning technology.

The invention discloses a diamond piezoresistive material and pressure sensing application thereof, and particularly relates to a boron-doped diamond-based piezoresistive material and an application method thereof in a sensor, a gradient doping structure and heterogeneous layer design are adopted, and the diamond piezoresistive material with high sensitivity, high linearity, temperature stability, corrosion resistance and radiation resistance is constructed; by optimizing a CVD process and a pattern etching structure, stable piezoresistive response in an environment of-60 DEG C to 250 DEG C is realized, the cycle life exceeds 10 times, and the high-reliability piezoresistive sensor is suitable for high-reliability sensing devices under extreme working conditions such as aerospace engines, deep sea detection equipment and nuclear reactors.

The invention discloses a method for automatically solving the critical heat flux density of a reactor based on a statistical principle, and belongs to the field of safety analysis of nuclear reactors. The problems that an over-fitting problem exists, a data set cannot be verified clearly through a fitting data set, and the development process is complicated when a traditional method is used for predicting independent variable combination items of a relational expression artificially given by CHF relational expression development are solved. The method comprises the following steps: collecting reactor CHF experimental data points; eliminating possible repeated experimental points and cold rod critical points to form a development database; based on a layered random sampling method, extracting data from the development database, and determining a fitting data set and a verification data set; based on the fitting data set, selecting a curve form of a single independent variable or a combination item; the independent variables and the combination items with the collinearity problem are removed; importance ranking analysis of the variables and the combination items is obtained through regression coefficient standardization; forming a final CHF relational expression; and based on the verification data set, verifying the CHF relational expression.

The invention relates to the technical field of safety monitoring of a nuclear reactor, and particularly discloses a multi-region pressure overrun leakage monitoring method for a reactor pressure vessel. The method comprises the steps of multi-region pressure data acquisition, real-time pressure field model construction, dynamic pressure field baseline modeling, pressure overrun anomaly detection, leakage source accurate positioning and evaluation based on spatial gradient analysis, topological structure mapping and multi-sensor data fusion, and graded early warning and response. By the adoption of the technical scheme, real-time and fine monitoring of pressure distribution in the reactor pressure vessel can be achieved, the rapid and accurate judgment and positioning capacity of leakage events is improved, the operation safety and reliability of nuclear power facilities are enhanced, and the reliability of the nuclear power facilities is improved. The defects of independent monitoring, local accurate positioning and complex working condition adaptability of multi-region pressure over-limit leakage in the prior art are overcome.

An exemplary system and method provide nuclear reactors with optimized detector placement. The exemplary system and method include a nuclear reactor model, Markov decision process, and reward function, where reinforcement learning can be used to iteratively generate placements of detectors to candidate positions within the nuclear reactor.

The invention relates to the field of supercritical water reactor thermal analysis, in particular to a positioning grid correction-based multichannel supercritical water reactor thermal analysis method, which comprises the following steps of: acquiring real-time nuclear reactor working condition data under a target supercritical condition; the nuclear reactor working condition data are input into a multi-channel supercritical water reactor thermal model, a global calculation result is obtained, and the multi-channel supercritical water reactor thermal model is a multi-channel model established based on the influence of a positioning grid; and analyzing according to the global calculation result to obtain an evaluation result of the influence of the positioning grillwork on the performance of the reactor core under the supercritical condition. According to the method, local positioning grillwork simulation is not needed, the geometric dependence and the complexity of earlier-stage treatment are reduced, and the efficiency of analyzing the thermal performance of the supercritical water reactor influenced by the positioning grillwork is improved.

The invention relates to the technical field of nuclear reactor safety, and particularly discloses a reactor waste heat export, recovery and comprehensive utilization system and method based on a heat pipe, and the system comprises a nuclear energy generation system, a secondary loop energy conversion system, a primary loop waste heat discharge system, a secondary loop waste heat discharge system, and a waste heat recovery and comprehensive utilization system. According to the waste heat recycling system, the heat pipe can be used for heat transfer and storage under the peak regulation operation condition, the nuclear heat use efficiency is remarkably improved, the requirements of various types of energy sources such as electricity and heat are met, and passive export of reactor core waste heat can be achieved through multi-natural-circulation coupling cooperative operation under the condition of a whole-field outage accident. By means of the efficient heat transfer characteristic and the modularization characteristic of the heat pipes, independent operation of the heat pipes can be achieved, the problem of single-point failure of the system can be avoided, meanwhile, separation of a reactor coolant and a cooling medium of a waste heat removal system can be achieved, the inherent safety of a nuclear reactor is remarkably improved, and the requirements of multi-purpose scenes are met; and the purpose of fully utilizing resources is achieved.

The invention belongs to the technical field of nuclear reactor physical calculation, and particularly discloses a multi-group steady-state neutron transport coarse mesh node global efficient solving method and system, which effectively fuse and exert the advantages of a coarse mesh node method and a Newton-Kralov method. A general coarse mesh segment transport calculation model suitable for various polyhedral segments is constructed, high-precision and high-efficiency transport scanning calculation is realized under sparse segments, the number of variables is greatly reduced, and the calculation scale of single transport scanning is reduced; a Newton-Kralov subspace global solution framework is established, a global solution variable is selected based on a variable coupling relationship of a coarse network nodal method, a linear equation corresponding to a Newton step length is solved by using a Kralov subspace method, and the Newton step length is used for updating the global solution variable to realize rapid and efficient convergence. On the premise of ensuring the calculation precision, the calculation efficiency is remarkably improved, and the calculation scale is reduced, so that the method is particularly suitable for the problem of large-scale and high-dominance-ratio neutron transportation.

The invention belongs to the technical field of nuclear reactor physical numerical calculation, and particularly relates to a method for improving precision and convergence of solving a diffusion equation based on a PINN, and the method comprises the following steps: S1, building a neural network algorithm model combining a physical information neural network PINN and source iteration; s2, aiming at a multi-region smooth transition neutron diffusion problem with small cross section difference between materials, optimizing the neural network algorithm model by adopting an optimization strategy of a training level; and S3, aiming at the multi-region neutron diffusion problem of large neutron flux gradient at the boundary due to large cross section difference between materials, improving the neural network algorithm model by adopting a domain-divided neural network architecture. According to the method, the problem that the calculation cost is exponentially increased in a high-dimensional or complex geometric scene due to the fact that a traditional finite difference method and a finite element method depend on grid division is avoided.

The application belongs to the technical field of thermal hydraulic analysis in nuclear reactors, and proposes an initial parameter generation method based on NUMAP steady-state calculation, which comprises: obtaining a thermal hydraulic simulation system model to be initialized in NUMAP, and decomposing the model into multiple single-flow single-pressure boundary chain structure sub-models; for each single-flow single-pressure boundary chain structure sub-model, setting the gas mass fraction of each in-line component and the non-condensable gas mass fraction to 0, and setting the flow to a constant value at the flow boundary; setting the temperature of each in-line component to the average of the inlet and outlet temperatures, or linearly filling the temperature of each in-line component according to the length of the component with the inlet and outlet temperatures as the endpoints; obtaining the total friction pressure drop along the line by subtracting the total gravity pressure drop along the line from the total pressure drop along the line, linearly dividing the total friction pressure drop along the line according to the parameters, and sequentially assigning the initial pressure of the in-line component from the inlet boundary. The application can make NUMAP reach steady state as soon as possible.

The invention discloses a stress corrosion crack propagation simulation and prediction method and system based on a phase field method. The method comprises the following steps: constructing a stress corrosion dynamic evolution model based on the phase field method; the stress corrosion dynamic evolution model is a control equation set for coupling a displacement field, a material concentration field and a crack phase field to carry out stress corrosion cracking evolution; establishing a discrete numerical value format for analyzing the stress corrosion dynamic evolution model, and performing simulation program development of the stress corrosion dynamic evolution model based on an ABAQUS platform according to the discrete numerical value format; and acquiring a specific load working condition, and performing stress corrosion crack growth rate simulation prediction on the nuclear reactor primary circuit 690 alloy material under different working conditions based on a developed simulation program to obtain the crack growth rate. According to the invention, the phase-field function is introduced to describe the crack morphology to simulate the stress corrosion crack propagation, and the rule of the stress corrosion crack propagation of the primary loop 690 alloy under different working conditions is accurately simulated.

The invention provides a sampling method suitable for continuous-discrete mixed variable design optimization of an advanced reactor. The sampling method comprises the following steps: (1) screening an effective discrete variable combination; (2) constructing a cumulative probability distribution function of the discrete variable combination; (3) hierarchically designing a space based on a cumulative probability distribution function interval; (4) distributing the number of samples according to the interval width proportion of the cumulative probability distribution function of each layer; (5) Latin hypercube sampling is carried out on the continuous variables in each layer; (6) integrating a global sample set and verifying the quality of the global sample set; and (7) constructing an agent model and evaluating precision. According to the method, the probability characteristic of the discrete variable and the space filling characteristic of the continuous variable are integrated, efficient exploration of the nuclear reactor design space is achieved, and compared with a traditional method, the sample quality and the calculation efficiency are remarkably improved.

The embodiment of the invention relates to the technical field of nuclear reactors, in particular to a gaseous fuel reactor, a reactor and a reactivity adjusting method.The gaseous fuel reactor comprises a reactor body, a containing cavity is formed in the reactor body, the containing cavity is arranged to contain gaseous fuel, and a gaseous fuel injection port is formed in the reactor body; the gaseous fuel injection pipe is arranged at the gaseous fuel injection port, and gaseous fuel is input into the containing cavity through the gaseous fuel injection port and the gaseous fuel injection pipe; the gaseous fuel provides heat and transfers the heat to the heat exchange assembly; the reactivity control assembly is arranged to be capable of changing the volume of the gaseous fuel in the containing cavity. According to the gaseous fuel reactor provided by the embodiment of the invention, the gaseous fuel is adopted as the reactor core fuel of the reactor, and the reactivity of the gaseous fuel reactor is controlled by changing the volume of the gaseous fuel in the accommodating cavity through the reactivity control assembly, so that the redundancy and reliability of reactivity control are improved.

The invention relates to the technical field of nuclear reactor and electrical equipment fault diagnosis, in particular to a nuclear reactor electrical penetration assembly fault mode diagnosis method and system, and provides a new nuclear reactor electrical penetration assembly fault mode diagnosis method. The method comprises the following steps of: designing a self-adaptive threshold value of each dimension of the electrical penetration device and an operation state reliability evaluation weight algorithm according to the characteristics of an electrical operation fault of the electrical penetration device and the characteristics of a sealed operation fault of the electrical penetration device, comprehensively performing operation diagnosis of the new mass reactor electrical penetration device, and comprehensively performing multi-dimensional comprehensive fault characteristic variables including the relation of original fault characteristics and the relation of the new mass reactor electrical penetration device. And meanwhile, the original fault features are utilized more sufficiently, so that the multi-dimensional operation diagnosis result of the electrical penetration device is more accurate, and the actual application requirements of a new mass heap site are fully met.

The invention discloses an intelligent path planning simulation method and system for nuclear reactor decommissioning engineering equipment. The method comprises the following steps: calling scene data and model data from a field database of a nuclear reactor decommissioning engineering; the model data comprises a structure model, a task equipment model and an operation limiting condition of the nuclear reactor; based on the scene data and the model data, assembling parameters of a path planning algorithm; the parameters comprise path safety requirements, path optimization targets and environment constraint conditions; calling a path planning algorithm based on the parameters; executing a path planning algorithm and outputting a result; and simulating an intelligent path of the dynamic decommissioning engineering equipment based on the result, and generating and outputting a UE animation. According to the method, a high-precision three-dimensional simulation model is established by deeply combining a simulation technology and combining decommissioning environment data of a nuclear reactor, a simulation platform supports multi-scene switching, different decommissioning process requirements are met, an optimal path verified through simulation can directly guide actual operation, and a clear navigation instruction is provided for equipment.

Embodiments of the present application relate to the technical field of nuclear reactors, and disclose a method and system for predicting the state of a reactor core, a device, and a medium. The method comprises: acquiring first reactor core state measurement parameters of a target reactor core in a first time step, the first time step being a time step obtained by dividing a combustion process of the target reactor core, and the first reactor core state measurement parameters comprising state measurement parameters of the target reactor core in a plurality of physical fields; when the first reactor core state measurement parameters are inputted into a data fusion model, by means of the data fusion model, performing fusion processing on the state measurement parameters of the plurality of physical fields, and obtaining a target data matrix; and when the target data matrix is inputted into a state prediction model, by means of the state prediction model, performing prediction to obtain first reactor core state information, the first reactor core state information being used for representing a reactor core state of the target reactor core in a second time step, and the second time step being later than the first time step.

The invention belongs to the field of graphite materials, and particularly relates to nuclear-grade natural graphite, preparation of the nuclear-grade natural graphite and application of the nuclear-grade natural graphite to preparation of high-temperature gas cooled reactor fuel spherulis.The preparation method comprises the steps that a natural graphite raw material is subjected to first-stage heat preservation treatment at the temperature T1 in advance, and a first-stage material is obtained; grinding and grading the first-section material to obtain graphite powder A and graphite powder B; then shaping and grading the graphite powder A to obtain graphite powder C; mixing the graphite powder B and the graphite powder C to obtain graphite powder D; the graphite powder D is subjected to liquid phase treatment in acid liquor in advance, and graphite powder E is obtained; the graphite powder E is subjected to two-stage gas-solid treatment with gas control and pressure control, and the nuclear-grade natural graphite is prepared. According to the method, the purity of the graphite can be obviously optimized, and in addition, the loss of strength in the purification process can be avoided, so that the graphite which is more than 5N, ultralow in nuclear reaction degradation impurities, excellent in strength and capable of meeting the application requirements of a nuclear reaction device can be obtained.

The invention belongs to the technical field of processing of nuclear reactor control rod neutron absorber materials, and particularly relates to a preparation method of a nuclear-grade hafnium cast ingot. The method comprises the following steps: preparing sponge hafnium into an electrode block; performing vacuum electron beam melting on the electrode block for at least two times to obtain a finished cast ingot; and processing the finished cast ingot to prepare the nuclear-grade hafnium cast ingot which can be used for processing the control rod profile for the nuclear reactor assembly. According to the preparation method of the nuclear-grade hafnium cast ingot, the tensile strength, the welding performance and the creep performance of a nuclear-grade hafnium profile can be effectively improved; meanwhile, the technological process is simplified, and the technological economy and the production efficiency are remarkably improved; in addition, sponge hafnium is directly adopted for vacuum electron beam melting, so that the cost of raw materials is greatly reduced. According to the method, the application performance of the nuclear-grade hafnium profile is improved, meanwhile, the process is simplified, the production efficiency is improved, the production cost is reduced, and the economical efficiency of nuclear-grade hafnium profile manufacturing is greatly improved.

The embodiment of the invention provides a method and equipment for measuring phase interface parameters between nuclear reactor fuel assemblies, and belongs to the technical field of thermal hydraulic analysis of nuclear power plants. The method comprises the steps that cavitation bubbles are generated in the flow channel of the special-shaped structure; images of cavitation bubbles in the flow channel of the special-shaped structure in multiple directions are collected; performing three-dimensional reconstruction according to the images of the cavitation bubble in the multiple directions to obtain the volume of the cavitation bubble; the void fraction is determined according to the volume of the void and the total volume of the fluid domain in the special-shaped structure flow channel, and the phase interface parameters comprise the volume of the void and the void fraction. According to the embodiment of the invention, the measurement precision of phase interface parameters can be improved.

The invention discloses a method for improving the quality of a laser cladding refractory high-entropy alloy coating, and belongs to the technical field of alloy coatings. The refractory high-entropy alloy coating is composed of WMoTaNb and hydride of Ti or V, the refractory high-entropy alloy coating is a single BCC phase, and the refractory high-entropy alloy coating is prepared through coaxial powder feeding laser cladding. The laser cladding refractory high-entropy alloy coating prepared through an introduction method (solid solution strengthening and in-situ deoxidation) has excellent wear resistance and high-temperature oxidation resistance, the quality and performance of the refractory high-entropy alloy coating are improved through additional powder, and the microhardness of the section of the refractory high-entropy alloy coating reaches up to 674.52 HV0. 2. The coating can be applied to the fields of aerospace, nuclear reactors and the like.

The invention provides an ultra-pure thorium fluoride production process which comprises the following steps: preparing a nanoscale ThF4.xNH4F precursor through a gas-liquid-solid three-phase reaction, atomizing an ammonium fluoride solution into microdroplets by adopting ultrasonic atomization, and carrying out synergistic reaction with gaseous ammonium fluoride and a liquid-phase thorium nitrate solution; a microwave-plasma synergistic fluorination reaction is adopted, and the precursor is heated in a high-purity inert atmosphere to generate crude thorium fluoride; according to the ultra-pure thorium fluoride production process provided by the invention, the purity of thorium fluoride is remarkably improved through the technical means of preparation of a nanoscale precursor through a gas-liquid-solid three-phase reaction, microwave-plasma synergistic fluorination, multi-stage sublimation combined with graphene-based nano trapping membrane purification, intelligent process control, by-product closed-loop recovery and the like; the impurity content is low, and the requirements of nuclear reactor fuel substrates and high-end optical coatings on ultra-high purity are met. The method has the advantages of high yield, high process stability and small purity fluctuation among batches, and the production efficiency and the product consistency are obviously improved.

Systems, structures, devices, and fabrication processes for ceramic matrix composites suitable for use in a nuclear reactor environment and other applications requiring materials that can withstand high temperatures and / or highly corrosive environments are disclosed. In one aspect, a ceramic composite structure is provided. The structure comprises a chamber including an external shell and a hollow space inside the external shell. The external shell includes an inner composite layer including a first composite structure, a middle composite layer placed outside of the inner composite layer, the middle composite layer including a second composite structure that is different from the first composite structure, and an outer monolithic layer that has a spatially uniform material property and placed outside of the middle composite layer.

A rapid inference method for a neutron flux distribution change of a reactor core of a nuclear reactor. In an offline training stage, a transient neutron transport equation is solved by means of a full-order solver, so as to obtain spatial distribution data of neutron flux and a delayed neutron precursor density changing over time, a reduced-order basis and a reduced-order coefficient data set of a training data set are obtained by means of proper orthogonal decomposition, and a reduced-order model is obtained by means of least squares fitting into the reduced-order coefficient data set; and in an online prediction stage, after the reduced-order model is solved on the basis of parameters of a predicted working state, the reduced-order basis is applied to a solved reduced-order coefficient, so as to reconstruct a physical field, which includes a neutron flux distribution of a reactor core, a delayed neutron precursor density distribution, etc. changing over time. In the present invention, data of a high-fidelity full-order model is used to construct a rapid prediction model, and a high-resolution change in the neutron flux of a reactor core over time is finely and efficiently described in a neutronics simulation, which plays an important role in reactor safety analysis, reactor data assimilation and inverse problem solving.

An installation includes: a plurality of pilings securable to a bed under a surface of a body of water; a base structure disposed atop the plurality of pilings; and a module disposable on the base structure, wherein the module is positioned and securable on the base structure after being floated on the surface of the body of water over the base structure.

According to the KAN nuclear reactor LOCA accident deduction and diagnosis method based on Bayesian optimization, the problems of parameter deduction and fault diagnosis in the LOCA accident working condition process can be solved, and KAN nuclear reactor LOCA working condition accident deduction and diagnosis based on Bayesian optimization are based on multi-pipeline sensor measurement signals; according to the method, modeling is carried out by using a two-fluid six-equation theory and a B-PIKAN network is combined to carry out pipeline core physical quantity prediction and progressive transfer learning. The whole process starts from acquisition of measurement signals of a multi-pipeline sensor, and through steps of physical two-fluid six-equation modeling, B-PIKAN model training, output and progressive transfer learning, deduction change values of core physical parameters of other pipelines along with time are finally obtained. The method is suitable for a complex LOCA pipeline environment, and the fault judgment accuracy is improved; according to the method, generalization and robustness are enhanced, and the modeling cost is reduced; according to the method, the reliability and robustness of the model are improved, and the interpretability is enhanced.