889 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.

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 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.

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 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.

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.

A nuclear fuel cell cladding that includes a zirconium-carbon covetic material. The zirconium-carbon covetic material has a carbon component associated with the surface of zirconium particles. The amount of carbon present in the zirconium-carbon covetic material is in a range of greater than 0.1 wt % to about 25 wt % of the zirconium-carbon covetic material. The carbon component may include carbon nanotubes, carbon nanomaterials, graphene, or graphene nanoplatelets. The carbon component may be uniformly distributed within the zirconium matrix. The zirconium-carbon covetic material may be formed from a zirconium alloy. The zirconium-carbon covetic material may be configured for use in various types of nuclear reactors. The synthesis of the cladding involves a process of plasma-enhanced chemical vapor deposition. The resulting nuclear fuel cell cladding offers improved performance and reliability for nuclear reactor applications.

The embodiment of the invention provides a calibration method and device for an out-of-pile detector of a nuclear reactor, and belongs to the technical field of nuclear power. The method comprises the following steps: acquiring an upper current signal, a lower current signal and a first axial power offset detected by an out-of-pile detector of a nuclear reactor; determining an axial power offset calibration coefficient of the out-of-pile detector and a nuclear power calibration coefficient of the out-of-pile detector according to the upper current signal, the lower current signal and the first axial power offset; respectively correcting second axial power offset and nuclear power currently indicated by the out-of-pile detector through the axial power offset calibration coefficient and the nuclear power calibration coefficient to obtain target nuclear power; and obtaining the axial power deviation indicated by the out-of-pile detector through the target axial power deviation and the target nuclear power. According to the embodiment of the invention, the axial power deviation and the nuclear power are respectively corrected to obtain more accurate nuclear power and axial power deviation, so that the accuracy of the out-of-pile detector indication signal is improved.

The embodiment of the invention discloses a pressure control method and system for a once-through steam generator of a nuclear reactor, and relates to the technical field of nuclear power. Determining power deviation according to the operation power and the secondary circuit load; according to the power deviation, the rated power corresponding to the nuclear reactor and the rated feed water flow corresponding to the once-through steam generator in the nuclear reactor, pressure regulation information is determined; and adjusting the pressure of the once-through steam generator according to the pressure adjusting information. According to the embodiment of the invention, a PID controller is not needed in the process of determining the pressure regulation information, but the pressure regulation information determined based on the power deviation directly acts on the feed water flow regulation valve, so that rapid and accurate OTSG pressure control can be realized.

The invention belongs to the technical field of nuclear fuel element surface protection, and particularly relates to a multi-layer composite (Cr, Al) Nx / Cr protective coating for a nuclear reactor zirconium alloy cladding and a preparation method of the multi-layer composite (Cr, Al) Nx / Cr protective coating. The coating adopts a multi-layer structure design consisting of a buffer layer, a middle layer and an outermost layer in sequence from inside to outside, wherein the buffer layer is a Cr coating; the middle layer is a nitrogen-saturated (Cr, Al) Nx coating, x is larger than or equal to 1, the middle layer is of a columnar crystal structure, and the hardness of the middle layer is 5-10 GPa; the outermost layer is a non-stoichiometric ratio (Cr, Al) Nx coating, x is larger than or equal to 0.6 and smaller than or equal to 0.9, a nanocrystalline structure is shown, and the hardness is 12-25 GPa. Component regulation, hardness gradient and a microstructure are subjected to collaborative design, so that the coating has excellent wear resistance, high oxidation resistance and structural integrity at the same time under a harsh nuclear working condition, the safety margin of the zirconium alloy cladding under normal service and accident working conditions is greatly improved, and the service life of the zirconium alloy cladding under normal service and accident working conditions is greatly prolonged.

The invention provides a shell-and-tube phase change heat storage system for a nuclear reactor containment and a nuclear accident heat energy management method, and belongs to the technical field of nuclear safety engineering and heat energy storage. The problems that an existing shell-and-tube heat reservoir is designed for industrial waste heat recovery and is not suitable for nuclear accidents, and the internal space structure of a containment is complex and limited are solved. The heat storage units are distributed in parallel, and each heat storage unit comprises a shell and a plurality of heat exchange tube bundles arranged in the shell; the heat storage unit has an interchangeable heat storage medium arrangement mode which comprises the first mode that the space in the heat exchange tube bundle is filled with a phase change material, and accident steam flows through a shell side channel between the shell and the heat exchange tube bundle; in the second mode, the shell side space between the shell and the heat exchange tube bundle is filled with the phase change material, and accident steam flows through an in-tube channel of the heat exchange tube bundle; and the accident steam passively flows through the channel which is not filled with the phase change material under the driving of the pressure difference. The method is used in the phase-change heat storage field.

The application relates to a neutron absorption material and application thereof to a control rod for a nuclear reactor core. The neutron absorption material provided by the application is prepared by mixing ytterbium elements and holmium elements in a certain proportion, can effectively play the high neutron capture cross-section characteristics of the ytterbium elements and the holmium elements, effectively absorb thermal neutrons, and can ensure good chemical and thermal stability, so that the neutron absorption material can maintain its structure and performance stability for a long time in a high irradiation environment, effectively avoids the conversion of the neutron absorption material into a weak neutron absorption material after absorbing neutrons, and further controls the neutron absorption characteristics of the neutron absorption material. Further, the neutron absorption material provided by the application can still ensure the neutron absorption characteristics thereof and maintain the reactivity value under high burnup depth, so as to prolong the service life of the control rod for the nuclear reactor core.

The invention relates to the technical field of nuclear thermal coupling, in particular to a general multi-physical coupling method, device and equipment for a nuclear reactor and a computer storage medium. According to the method, Monte Carlo transport software, a computing environment of a thermotechnical solver and a preCICE communication interface are initialized through a predefined configuration file, Monte Carlo transport calculation is directly completed according to an initial material in first iteration, body power density data are obtained, and the body power density data are transmitted to the thermotechnical solver through preCICE to serve as a heat source to be mapped to a finite element grid. Further obtaining an updated temperature and density field and feeding back the updated temperature and density field to Monte Carlo transportation software; and controlling an iteration process through a convergence criterion until the requirement is met. According to the method, full-process automation and memory-level data exchange in the multi-physical coupling process are achieved, the calculation efficiency and precision are remarkably improved, manual intervention and file read-write bottlenecks are eliminated, and meanwhile, the strong adaptability and good expansibility of the method to different geometric models and physical solvers are guaranteed by relying on the universal interface design and modular architecture of preCICE.

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

The invention provides a rotating target for generating monoenergetic neutrons, which comprises a target chamber, a target sheet, a sealing flange and a vacuum system, the target sheet is composed of a TiD / TiT target material and a substrate, a deviation distance of 5-10mm exists between the center of the target material and the center of the substrate, the substrate is connected with a motor outside the target chamber through a fixed connecting rod, and the target chamber is driven by the motor to rotate at a speed of 3-5min / r. The target material is designed to be the eccentric target rotating slowly, so that the utilization rate of the target material can be improved, and the phenomena that the temperature of a single point is too high, a coating layer penetrates or falls off, enriched gas escapes and neutron yield fluctuation is caused due to the fact that beam bombards one position of the target material for a long time can be avoided. The energy width of neutrons generated by the rotating target is lower than 5%, the rotating target has good monochromaticity and low energy loss, the design precision requirement of the fourth-generation nuclear reactor can be met, and a possible guarantee is provided for commercialized popularization and application of the fourth-generation nuclear reactor.

A nuclear reactor core includes a plurality of fuel elements and moderator blocks to form a nuclear reactor core inner portion and a nuclear reactor core outer portion. The nuclear reactor core inner portion includes an inner moderator matrix formed of a plurality of inner holes that include a plurality of inner fuel openings with one or more fuel elements disposed therein. The plurality of inner holes further include a plurality of inner coolant passages to flow a coolant. The nuclear reactor core outer portion includes an outer moderator matrix formed of a plurality of outer holes that include a plurality of outer fuel openings with one or more fuel elements disposed therein. The plurality of outer holes further include a plurality of outer coolant passages to flow the coolant. The outer coolant passage diameter exceeds the inner coolant passage diameter.

The invention provides a reactor internal capable of realizing integral movement of a reactor core, and relates to the technical field of research, development and design of reactor body structures, the reactor internal comprises a hanging basket barrel, a reactor core shroud and a reactor core; the reactor core is arranged in the reactor core shroud; the reactor core shroud is arranged in the hanging basket barrel body, the reactor core shroud is movably connected with the hanging basket barrel body, the reactor core shroud slides up and down in the hanging basket barrel body, and the bottom of the hanging basket barrel body supports the reactor core shroud. According to the invention, the material changing and maintenance efficiency can be effectively improved, the downtime is shortened, the safety is enhanced, the personnel radiation exposure risk is reduced, the inspection and maintenance are convenient, and the service life of equipment is prolonged. The system adapts to advanced reactor design and conforms to the development trend of modularization and high efficiency, and the performance, safety and economical efficiency of the nuclear reactor are remarkably improved through the design of the system.

The application is suitable for the field of nuclear reactor technology and discloses a filtering device and filtering method for a lead-bismuth cooled pool type reactor, which comprises a fixing assembly, a branch pipeline and a filtering assembly; the fixing assembly is arranged on the side of the main channel, and the fixing assembly comprises the branch pipeline in which the filtering assembly is detachably installed; the branch pipeline is connected with the side of the main channel through the fixing assembly to form the branch pipeline, and the filtering assembly is used for filtering and purifying in the branch pipeline without providing filtering power by itself; the filtering device as a whole can be arranged in the pool type reactor to ensure the integrity of the boundary of the reactor body; the filtering assembly has a simple structure and is convenient to install and replace, and can be operated in the reactor for a long time, so that the filtering device has a long service life and the replacement period of the filtering device is reduced.

The application discloses a kind of structural grid methods and systems of dividing triangular arrangement rod bundle assembly with wire winding, it is related to nuclear reactor core thermal-hydraulic analysis and calculation field, by using structured hexahedral grid to divide the flow channel of triangular arrangement rod bundle with wire winding, grid quality is higher, and calculation efficiency is more optimal;When establishing the structured grid of the flow cross section of triangular arrangement rod bundle assembly with wire winding, we fully consider the geometric characteristics of wire winding, thereby improve the geometric reduction degree of grid model;According to the calculation requirement, the number and distribution of grid nodes can be flexibly adjusted, and the axial length is shortened accordingly, so as to not affect the calculation precision, reduce the number of grids, save the calculation resource, and ensure the calculation quality.

The invention relates to the technical field of nuclear reactor fault detection, in particular to a method and system for extracting weak collision and abrasion noise of loose parts in a reactor based on signal processing and machine learning, and the method comprises the steps: obtaining a reactor vibration signal, and decomposing the vibration signal into a plurality of sub-signals through empirical mode decomposition; performing Hilbert transform processing on each sub-signal, extracting an instantaneous frequency and an instantaneous amplitude of each sub-signal, and constructing a time-frequency two-dimensional matrix of each sub-signal; superposing the time-frequency two-dimensional matrixes of the sub-signals along the dimensions of the sub-signals to generate a three-dimensional instantaneous time-frequency matrix; and inputting the three-dimensional instantaneous time-frequency matrix into a pre-trained convolutional neural network model, inhibiting the operation noise of the reactor, and extracting a weak collision and abrasion noise signal after noise reduction. The method aims at extracting weak collision and abrasion noise signals of the loose part in the reactor and achieving collision and abrasion fault recognition of the loose part in the reactor.

The present application relates to the technical field of nuclear reactor thermal hydraulic simulation analysis, and provides a thermal stratification phenomenon simulation method and apparatus, a device, a medium, and a program product. The method comprises: on the basis of an interactive model computation result in a pre-configured program interface, controlling a one-dimensional system program and a three-dimensional computational fluid mechanics program to perform coupling computation, wherein the program interface is configured between the one-dimensional system program and the three-dimensional computational fluid mechanics program, and the model computation result comprises results obtained by performing simulation computation on a liquid metal reactor by the one-dimensional system program and the three-dimensional computational fluid mechanics program; and obtaining a real-time computation result of coupling computation of the one-dimensional system program and the three-dimensional computational fluid mechanics program, wherein the real-time computation result comprises a simulation result of a thermal stratification phenomenon of the liquid metal reactor in a transient operation process. The present application can reduce the computing resource demand and the time cost and improve the efficiency while ensuring that the thermal stratification phenomenon of the liquid metal reactor is accurately simulated.

The application provides a sealed weld cutting device and method, which comprises a base, a rotary driving assembly arranged on the base, a cutting mechanism, a connecting assembly, a positioning adjusting assembly for adjusting the cutting position, and a control assembly, the rotary driving assembly is drivingly connected with the cutting mechanism through the connecting assembly, the positioning adjusting assembly comprises an axial positioning assembly and a radial positioning assembly, the axial positioning assembly is arranged on the base and connected with the connecting assembly, the radial positioning assembly is arranged in the cutting mechanism, and the axial positioning assembly and the radial positioning assembly are both in communication connection with the control assembly.In the application, the base, the rotary driving assembly, the cutting mechanism, the connecting assembly and the positioning adjusting assembly are arranged, the cutting position is aligned with the sealed weld through the positioning adjusting assembly, and the remote control is realized through the control assembly, so that the problems of narrow space, high radiation and unreachable personnel in the nuclear reactor are solved, and the remote control and the accurate alignment of the cutting position are realized.

The embodiment of the invention provides a nuclear reactor melt physical property parameter determination method and device, and relates to the technical field of nuclear reactors. The method comprises the following steps: receiving a physical property parameter acquisition request, wherein the physical property parameter acquisition request comprises the type of a target substance in a nuclear reactor melt; in response to the physical property parameter acquisition request, acquiring a base class pointer corresponding to the target substance according to the type of the target substance; based on the base class pointer, a physical property calculation relation related to the target substance is called, and the physical property calculation relation comprises a mapping relation or a function calculation relation; according to the physical property calculation relation, physical property parameters of the target substance are determined, and the physical property parameters comprise at least one of density, specific heat capacity, emissivity and heat conductivity. According to the embodiment of the invention, the reliability of physical property parameters can be improved.

A nuclear power plant comprising at least one nuclear reactor having a steam generator, a steam cycle including the steam generator, a first turbine fluidly coupled to the steam generator, a first condenser fluidly coupled to the first turbine, and a feedwater heater fluidly coupled to the first condenser and the steam generator, an intermediate thermal loop including the first condenser, and an evaporator, and an organic cycle including the evaporator, a second turbine fluidly coupled to the evaporator, a second condenser fluidly coupled to the second turbine, and a heat exchanger fluidly coupled to the second condenser and the evaporator.

An underwater crawling device for a nuclear reactor comprises: a drive control mechanism; the crawling mechanism is connected with the driving control mechanism, the crawling mechanism comprises a machine body and a plurality of crawling legs, each crawling leg comprises a translation component and a lifting component, the translation component is connected with the machine body, the lifting component is connected with the translation component, and the machine body is connected with the machine body. The driving control mechanism controls the translation part and the lifting part to move in sequence according to a set sequence; the plurality of crawling legs are matched with one another to drive the machine body to crawl and move to a set position; wherein the lifting component is provided with a positioning guide pin corresponding to a flow hole of a lower grid plate of a reactor core in a reactor pressure vessel, and the lifting component is inserted into the flow hole through the positioning guide pin to realize crawling movement; a machine body guide pin matched with the flow hole is arranged on the bottom face of the machine body, and the machine body guide pin is inserted into the flow hole to achieve positioning and bear horizontal component force.

A simulation virtual debugging method and system are applied to debugging of a nuclear reactor refueling machine control system, and the method comprises the following steps: generating an operation sequence according to a received refueling demand, the operation sequence comprising a plurality of equipment instructions; simulating the equipment instruction through a pre-constructed refueling machine simulation model and a preset buoyancy algorithm rule to obtain simulation result information; and debugging the nuclear reactor refueling machine control system according to the simulation result information. According to the scheme, the debugging work of the nuclear reactor refueling machine control system can be started without waiting for mechanical equipment to arrive at a factory, the project production time can be shortened, and the debugging accuracy, reliability and debugging efficiency can be ensured.

This invention discloses a method, system, and medium for data mapping between a solid geometric model and an unstructured mesh, relating to the field of nuclear reactor core coupling programming technology. By converting the CSG geometric entity into a coupling intermediate mesh (unstructured mesh), and using the coupling intermediate mesh as an intermediary between the CSG geometric entity and other professional unstructured meshes, high-precision coupled data transmission and coupled calculation between the CSG geometric entity and other professional unstructured meshes can be achieved through mesh mapping and other methods, based on the geometric correspondence between the CSG geometric entity and the coupling intermediate mesh.