1933 results about "Nuclear reactor" patented technology

One variation of a system includes a pressure vessel: including a wall; defining a primary working fluid circuit extending vertically within the wall; defining a secondary working fluid circuit extending vertically within the wall and fluidly isolated from the primary working fluid circuit; and configured to store a nuclear fuel, a primary working fluid, and a secondary working fluid. The wall of the pressure vessel: defines a heat exchanger configured to transfer thermal energy from the primary working fluid flowing through the primary working fluid circuit into the secondary working fluid flowing through the secondary working fluid circuit; and defines a radiation shield configured to attenuate radiation emitted by the nuclear fuel.

An integral molten salt nuclear reactor includes a drain tank section configured to hold a volume of fuel salt. The integral molten salt nuclear reactor further includes a reactor section configured to receive the volume of fuel salt from the drain tank and cause fission reactions that heats the molten salt. The integral molten salt nuclear reactor further includes a heat exchange section configured to receive a flow of the heated fuel salt from the reactor section and remove heat therefrom.

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 relates to the technical field of laser welding, and discloses a laser machining device and method for a nuclear reactor shell, and the method comprises the steps that a shell plate welding area to be welded is subjected to surface modification pretreatment; the laser processing device is used for carrying out initial welding on the pretreated shell plate, in the initial welding process, the temperature sensing assembly is used for monitoring temperature distribution of the welding area in real time, and temperature information of the welding area is obtained; the temperature information is input into a preset temperature-deformation prediction model for prediction of a variable trend and a stress distribution condition so as to obtain a first prediction result; based on the obtained first prediction result, the welding parameters of the laser processing device are dynamically adjusted for the first time; welding deformation and stress concentration are effectively coped with, the welding quality and structural strength of the nuclear reactor shell after laser welding are improved, and the reliability and safety of the nuclear reactor shell during use are effectively improved.

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.

The invention provides a fouling risk calculation method and equipment of a reactor and a storage medium, and relates to the technical field of nuclear reactors. The reactor fouling risk calculation method provided by the invention comprises the following steps: step S110, acquiring input parameters; s120, the mass evaporation rate is calculated based on the input parameters; step S130, calculating the dirt thickness of the reactor core based on the mass evaporation rate; step S140, calculating boron deposition mass based on the dirt thickness of the reactor core; step S150, judging whether a convergence condition is met or not, if the convergence condition is met, entering step S160, and otherwise, returning to step S110; and step S160, outputting a calculation result. According to a risk formation mechanism and important phenomena caused by reactor fouling, the fouling thickness and the boron deposition quality of the reactor core are calculated based on the input parameters, so that risks, including CILC risks and CIPS risks, caused by fouling are evaluated.

The invention relates to the field of theoretical calculation simulation of a nuclear reactor core state, in particular to a simulation method for dynamic reactor core power distribution measurement and a related product, and the method comprises the following steps: obtaining accurate burnup data; establishing a steady-state reactor core theoretical model based on accurate burnup; solving a three-dimensional neutron kinetic equation by taking the steady-state reactor core theoretical model as an initial state and taking a relation curve of reactor core operation power and operation time as input to obtain a dynamic reactor core theoretical model taking time as a function; inputting the power and time relation corresponding to the power and time relation corresponding to the power plant test plan into the dynamic reactor core theoretical model to obtain a database; according to the method, the steady-state theoretical model is constructed by utilizing accurate burnup data, dynamic reactor core simulation is carried out through a transient reactor core neutronics calculation model and the like, a dynamic reactor core model completely consistent with an actual test state can be generated, and actually measured data of dynamic reactor core power distribution can be directly processed and accepted.

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 invention discloses a reactor core nuclear thermal coupling calculation method based on an embedded physical knowledge neural network. The method comprises the following steps: 1, dividing reactor core nodes of a nuclear reactor; 2, initializing neutron physical field parameters and thermal hydraulic parameters of the reactor core; 3, building, training and storing an embedded physical knowledge neural network model; 4, predicting the neutron number density of each node of the reactor core by using the stored model to obtain neutron flux distribution; 5, calculating reactor core power distribution according to neutron flux distribution; 6, calculating and updating the thermal hydraulic parameters of the reactor core at the tn + 1 moment by using a system analysis program; 7, calculating and updating reactivity of each node of the reactor core; and 8, judging whether the appointed calculation time is reached, if so, ending, otherwise, returning to the step 4. The method is based on an embedded physical knowledge neural network method, an accurate and efficient calculation method is provided for solving distribution of the neutron physical field and the thermal field in the reactor core, and the method has important significance on reactor design optimization.

An integrated energy system comprising a power plant including at least one nuclear reactor and electrical power generation system, the at least one nuclear reactor being configured to generate steam, and the electrical power generation system being configured to generate electricity, a desalination system configured to receive at least a portion of the electricity and steam to produce brine, an electrolysis process configured to process the brine into Sodium Hydroxide (NaOH), a Sodium Formate (HCOONa) production process configured to receive the Sodium Hydroxide (NaOH) to produce Sodium Formate (HCOONa), a Hydrogen (H2) extraction reactor configured to receive the Sodium Formate (HCOONa) and produce Hydrogen (H2), and a fuel cell configured to receive the Hydrogen (H2).

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.

A molten salt nuclear reactor includes a vessel and a control blade assembly disposed within the vessel. The control blade assembly includes a drive rod and a control blade. The drive rod further includes an oblique surface wherein vertical motion of the drive rod and its oblique surface causes horizontal motion of the control blade into and out of a reactor section of the molten salt reactor.

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 provides a repairing method and a repairing system for sealing surface defects of a nuclear reactor pressure vessel, and belongs to the technical field of nuclear power maintenance. The repairing method comprises the following steps: performing defect detection on the sealing surface of the pressure vessel; performing polishing repair or welding repair by adopting a corresponding repair method according to the distribution mode, type and size of the defects, wherein the welding repair process and process parameters are determined according to the thickness of the residual anticorrosive coating at the bottom of the welding groove after the defects are removed, the finite element simulation welding temperature field and the welding effect evaluation on the test piece; and after welding, the welding repair part is ground, so that the welding repair area meets the requirements of flatness and roughness. The repairing efficiency is greatly improved, the repairing effect is guaranteed, the personnel radiation time is shortened, and the service life of the sealing surface of the pressure vessel is prolonged.

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 embodiment of the invention relates to the technical field of nuclear reactors, in particular to a cock assembly for a reactor. The reactor comprises a reactor container and a reactor core assembly arranged in the reactor container, the reactor container comprises a container opening, and the reactor further comprises a refueling machine. The cock assembly comprises a first cock, a second cock and a third cock. The first cock is rotatably arranged at a container opening of a reactor container of a reactor, the first cock is provided with an eccentric first through hole, and the size of the first through hole is larger than the radius of the first cock. The second cock is arranged in the first through hole of the first cock and can rotate relative to the first cock, and the second cock is provided with an eccentric second through hole. The third cock is arranged in the second through hole of the second cock and can rotate relative to the second cock, and the third cock is provided with an eccentric third through hole. And the refueling machine is arranged in a third through hole of the third cock. According to the cock assembly, the moving range of the refueling machine can be increased, and the size of the first cock is reduced.

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.

A loading system for a molten salt reactor system includes a slug loading assembly. The slug loading assembly includes an inert chamber having an entry port therein that is configured to receive a solid slug. The loading system further includes a chute having a pipe run extending from the entry port and elevationally below the slug loading assembly. The loading system further includes a terminal sieve fluidically coupled with the pipe run opposite the entry port and configured to receive the solid slug via the pipe run. The terminal sieve is positionable with a flow of a liquid molten salt for dissolution of the solid slug therein.

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 invention discloses a calculation method for a nuclear reactor neutron transport equation, and belongs to the field of nuclear reactor core design and safety. According to the method, the problem that the convergence speed of an inner iteration process of a transport equation and a low-order equation in a traditional PI-CMFD method is low is solved. The method specifically comprises the following steps: step 1, reading a coarse mesh model, a fine mesh model, material information and boundary conditions of a nuclear reactor; step 2, establishing a numerical discrete model of a steady-state multi-group transport equation according to the fine grid model read in the step 1; 3, after boundary constraint and material constraint are applied to the established numerical discrete model of the steady-state multi-group transport equation, a discrete equation set is constructed by transforming a test function in the model; and 4, solving the constructed discrete equation set based on the grid mapping relation between the coarse grid model and the fine grid model by adopting a PI method and a CMFD method to obtain a neutron angular flux value and a characteristic value of the neutron transport equation. The method can be applied to calculation of the neutron transport equation of the nuclear reactor.

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 discloses a preparation method for smelting a high-rhenium-molybdenum alloy pipe. The method comprises the following steps: 1, uniformly mixing Mo powder and Re powder, performing cold isostatic pressing, and sintering to obtain a high-rhenium-molybdenum alloy sintered blank; 2, carrying out vacuum electron beam smelting for many times to obtain a smelted high-rhenium molybdenum alloy cast ingot; 3, performing high-temperature extrusion combination and forging to obtain a smelted high-rhenium molybdenum alloy bar; 4, performing deep hole drilling, honing and machining treatment to obtain a smelted high-rhenium molybdenum alloy pipe blank; and 5, performing multi-pass high-temperature rolling, intermediate heat treatment, surface treatment and finished product heat treatment to obtain the smelted high-rhenium molybdenum alloy pipe. According to the method, repeated vacuum electron beam smelting is adopted, and repeated pressure machining such as high-temperature extrusion, forging and high-temperature rolling is combined, so that the purity and component uniformity of the smelted high-rhenium molybdenum alloy pipe are improved, the structure is uniform, the mechanical property is excellent, and higher dimensional precision and surface quality are obtained; the material is suitable for the fields of cladding pipes, heat pipes, medical implants and the like in nuclear reactors.

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.