44 results about "Nuclear power reactor" patented technology

The invention relates to the field of nuclear power reactor core water replenishing tank outlet check valve welding, in particular to a welding alignment tool for a reactor core water replenishing tank outlet check valve and a welding method. The tool comprises a rack, two pipeline fixing frames and a valve body supporting and adjusting frame, the rack provides a fixing foundation for other parts, the valve body supporting and adjusting frame is arranged in the center of the rack, and the two pipeline fixing frames are arranged at the two ends of the rack; the valve body supporting and adjusting frame is used for supporting and adjusting the horizontal position and the vertical height of a check valve body, and the pipeline fixing frame is used for supporting and fixing a check valve butt joint pipeline. The valve body supporting and adjusting frame is used for supporting and adjusting the horizontal position and the vertical height of the valve, the worm gear lifting machine and the guide column are installed in the support to achieve adjustment and locking of any height, the horizontal position adjusting mechanism freely slides through the linear guide rail and the sliding block, and then rapid centering of the CMT outlet check valve and the butt joint pipeline welding seam lip edge is achieved. The labor intensity of personnel can be reduced and the working efficiency is effectively improved.

The invention discloses a xenon oscillation automatic suppression auxiliary early warning prompting method based on a 3AO method, and belongs to the technical field of nuclear power reactor core monitoring. Firstly, various high-precision sensors are arranged at key parts of a nuclear reactor, occurrence signs of xenon oscillation are captured in time, and the system can accurately identify early signals of xenon oscillation by analyzing data of the sensors, so that a basis is provided for formulation of a subsequent control strategy. The core of the control strategy is fusion of a 3AO method and a deep Q network algorithm, the algorithm comprehensively considers the real-time state of the reactor core, and a control decision is optimized by learning historical data, so that the optimal control rod position and boron concentration adjustment scheme is calculated. And then, the system automatically prompts an operator to adjust the control rod and the boron concentration according to a control instruction generated by an intelligent algorithm so as to respond to the change requirement of the nuclear reactor in real time, and an efficient closed-loop control system is formed by feeding back an adjustment effect in real time, so that the accuracy and the reaction speed of a control strategy can be continuously improved.

The invention relates to a nuclear power reactor building outer dome and an installation method thereof.In the nuclear power reactor building outer dome, on one hand, steel molds are prefabricated into steel mold units in a factory in a blocking mode, after the steel mold units are transported to a site, the steel mold units are spliced into a finished steel mold on the site, and then the steel mold is integrally hoisted to the top of a corresponding reactor building; and a scaffold does not need to be erected on the inner dome, so that the construction process is greatly simplified, a large amount of time and manpower resources are saved, and the construction efficiency is improved. And on the other hand, all the steel molds are directly hoisted after off-site assembly is completed, a scaffold for supporting the outer dome does not need to be erected outside the inner dome, the construction period is effectively shortened, and the project progress is accelerated.

The invention mainly relates to the technical field of nuclear power reactors, in particular to a modeling method of a non-uniform power distribution rod bundle CHF test piece and a CHF prediction method. The modeling method of the non-uniform power distribution rod bundle CHF test piece mainly comprises the following ten steps: determining geometric parameters of a fuel rod bundle, dividing a radial channel, dividing an axial node block, determining geometric parameters of a channel, determining heating power distribution, determining a working condition range, selecting a two-phase flow model, determining a turbulence mixing coefficient, determining a flow resistance coefficient and determining a grillwork position. The invention provides a complete and systematic sub-channel modeling method for a non-uniform power distribution rod bundle CHF test piece, and the CHF parameters of the non-uniform power distribution rod bundle can be predicted more accurately.

The invention discloses a nuclear power control system safety risk prediction method based on a Transform model, which is characterized in that a nuclear power control system safety risk prediction model is constructed by adopting a nuclear power control system risk feature selection and small probability feature value generation algorithm, and intelligent safety risk prediction of a nuclear power control system is realized. The method specifically comprises the steps of performing preliminary selection on nuclear power system characteristics, performing a small probability characteristic value generation algorithm, constructing a nuclear power control system intelligent safety risk prediction model based on Transform and the like. Compared with the prior art, the method has the advantages that safe operation of the nuclear power reactor is ensured, various parameters of the reactor are monitored to be in a safe range, actions are automatically taken when abnormal conditions are detected to prevent accidents, and the method is convenient, simple, good in monitoring effect and high in safety. The method has important significance and a good application prospect for safety management of nuclear power.

The present application relates to a kind of nuclear power reactor plant outer dome and its installation method, in the nuclear power reactor plant outer dome, on the one hand, steel mould is prefabricated into each steel mould unit in factory block, after being transported to site, each steel mould unit is assembled into completed steel mould in site, and steel mould is hoisted to the top of corresponding reactor plant as a whole, without erecting scaffold on inner dome, so that the construction process is greatly simplified, a large amount of time and human resources are saved, and construction efficiency is improved.Another aspect, each steel mould of the present application is directly hoisted after being assembled to completion outside site, without erecting scaffold for supporting outer dome outside inner dome, effectively shortening construction period, and accelerating engineering progress.

The invention belongs to the technical field of nuclear industry equipment maintenance, and particularly relates to a nuclear power reactor bottom simulation experiment device which comprises a pressure vessel wall, a plurality of replaceable flux tube assemblies, a plurality of fixed flux tubes and at least one supporting leg. At least one supporting leg is installed at the edge of the bottom end of the pressure vessel wall and used for supporting the pressure vessel wall. The replaceable flux pipe assemblies and the fixed flux pipes penetrate through the wall of the pressure container and are detachably installed on the wall of the pressure container. The top end of each replaceable flux tube assembly is provided with at least one quick release connector. The quick-release connector is used for being externally connected with machining equipment. The simulation experiment device for the reactor bottom of the nuclear power reactor is arranged for an experiment instead of being operated on a reactor site, so that the technical problems that the neutron flux measurement precision is seriously influenced and potential safety hazards of a unit are generated due to the fact that weld joints of existing neutron flux tubes are prone to deformation and cracking are solved.

The invention discloses a nuclear storage and supply integrated method and system based on a pressurized water reactor and fused salt heat storage. The method comprises the following steps that steam is provided for a steam turbine device through nuclear energy of a reactor device; power is generated through the steam turbine device; heat is provided for the fused salt heat storage device through the steam turbine device, and heat is stored and provided through fused salt in the fused salt heat storage device. A nuclear power'machine '(steam turbine device) side is coupled with the fused salt heat storage device to form a nuclear storage and supply integrated system, that is, flexible connection is added between nuclear power and a user, so that decoupling of a nuclear power reactor device and the steam turbine device is realized, a flexible nuclear energy concept is realized, and the system can be applied to peak regulation, large-scale heating and steam supply scenes; the requirements for nuclear power peak regulation and nuclear energy comprehensive utilization are met through the fused salt heat storage device, the reactor power can not be adjusted frequently, potential safety hazards caused by frequent reactor power adjustment are avoided, and the nuclear power safety is improved.

The present invention provides a bidirectional ultrasonic inspection probe for the inner hole of a nuclear power reactor bolt, comprising a mounting bracket and a probe, wherein the mounting bracket comprises a support frame and a support tube, and the support frame is also elastically connected to a probe bracket, and the upper surface of the probe bracket has symmetrical inclined surfaces on both sides, and the probe is fixed on the inclined surfaces, and the lower surface of the probe bracket is inwardly recessed and has two symmetrically arranged inner concave surfaces pointing to the upper surface of the probe bracket and intersecting to form a tip; an expansion sleeve that can be supported between the outer wall of the support tube and the inner hole of the bolt is fixed on the support tube. The present invention has a reasonable structural design, can effectively improve the stability of the probe when entering the inner hole of the bolt for detection, maintain the concentricity of the probe as a whole and the inner hole of the bolt, and improve the detection quality and detection accuracy.

The invention relates to the technical field of nuclear power reactors, and discloses a nuclear power unit secondary circuit support system and a once-through steam generator nuclear power system. According to the nuclear power unit secondary loop support system, a plurality of groups of steam generator output pipelines are communicated with a main steam pipeline in parallel; the start-stop reactor main pipe is communicated with the output pipeline of the steam generator through a start-stop branch, a steam-water separator is mounted on the start-stop reactor main pipe in parallel, a steam supply branch is communicated between the start-stop reactor main pipe and the main steam pipeline, the steam supply branch is communicated with a steam exhaust pipeline, and two groups of steam exhaust outlets are formed in the tail end of the steam exhaust pipeline in parallel; and the drainage pipeline is communicated with the liquid phase outlet of the steam-water separator, and two groups of drainage outlets are arranged on the drainage pipeline in parallel. Start-stop steam exhaust of multiple groups of steam generators is treated through the shared start-stop reactor main pipe and the steam-water separator, graded recycling of working media and heat energy in the start-stop process is achieved, vapor-phase media are recycled, and waste of heat energy resources and water resources in the initial stage of start is reduced.

The utility model relates to a hoisting tool for an outer dome of a nuclear power reactor, which is provided with a first hoisting groove and a second hoisting groove which are symmetrical to each other, so that a hoisting rope is uniformly stressed in a hoisting process, and potential safety hazards caused by overlarge stress of a single point are avoided. And due to the arrangement of the lifting rope assembly and the short lifting assembly, stress in the lifting process is more reasonable, shaking and deformation of the steel mold in the lifting process are reduced, and the lifting safety is improved. Furthermore, the lifting hook is arranged on a large crawler crane, the nuclear power reactor outer dome lifting tool can be correspondingly mounted on the large crawler crane, so that the outer dome steel mold is integrally lifted, the problem that the lifting load of a traditional tower crane is limited is effectively solved, and large-span and large-load lifting of the outer dome steel mold is realized.

The utility model relates to the technical field of thermal power transformation, in particular to a fused salt-based generator set for transforming thermal power into nuclear power. The combined generator set comprises a nuclear island steam generator, a steam fused salt heat exchanger, a fused salt assembly, a heat exchange assembly, a water supply pipeline, a steam turbine and a generator, and the steam fused salt heat exchanger is connected with the nuclear island steam generator and the fused salt assembly and used for storing steam heat of the nuclear island steam generator into fused salt of the fused salt assembly. The heat exchange assembly is connected with the fused salt assembly and the water supply pipeline, the steam turbine is connected with the heat exchange assembly through a steam pipeline, the steam turbine is connected with the generator, and parameters of steam supplied to the steam turbine by the steam pipeline are controlled by controlling the temperature and flow of fused salt in the fused salt assembly and the pressure and flow of feed water in the water supply pipeline. Therefore, thermodynamic system matching of the nuclear power reactor and the thermal power generating unit is realized. According to the technical scheme, the thermodynamic system matching problem of an original thermal power generating unit and a nuclear power reactor can be solved.

The invention mainly relates to the technical field of nuclear power reactors, in particular to a method for determining a low-flow CHF test working condition of a fuel assembly. The method comprises the steps that aiming at the working condition that the flow range is 20-300 kg / (m < 2 >. S), a plurality of first CHF values are calculated through a plurality of first CHF relational expressions; the first CHF relational expression with the minimum deviation between the first CHF value and the calculation result of the WLOP relational expression based on the sub-channel program is selected as the CHF relational expression of the low-flow working condition; a plurality of low-flow CHF test working conditions are constructed according to the CHF relational expression of the low-flow working conditions and the flow range of 20-300 kg / (m < 2 >. S); and screening the plurality of low-flow CHF test working conditions according to a working condition effectiveness screening criterion to obtain a final CHF test working condition.

The invention discloses a yttrium hydride-based high-temperature-resistant neutron moderation shielding material containing gadolinium oxide / samarium oxide / europium oxide and a preparation method and application thereof, and belongs to the technical field of nuclear radiation shielding. The problems that existing pure yttrium hydride is weak in neutron absorbing and shielding capacity, prone to cracking and pulverization and the like are solved. The density of the shielding material is 4.36-4.62 g / cm < 3 >, and the mass proportions of the substances in the total amount of YH3, Sm2O3, Eu2O3 and Gd2O3 are as follows: YH3 is more than or equal to 99% and less than 100%, Gd2O3 is more than or equal to 0% and less than or equal to 1%, Sm2O3 is more than or equal to 0% and less than or equal to 1%, Eu2O3 is more than or equal to 0% and less than or equal to 1%, Gd2O3 + Sm2O3 + Eu2O3 is more than 0% and less than or equal to 1%, and Gd2O3 + Sm2O3 + Eu2O3 is more than 0% and less than or equal to 1%. The material can effectively absorb full-spectrum neutrons, has high density, low crack rate and low defect rate, is resistant to high temperature and excellent in shielding performance, adopts hydrogenation, nano-ball milling and pressing sintering preparation processes, and is suitable for a micro nuclear power reactor at the temperature of 900 DEG C or below.

The invention discloses a reactor multi-source heterogeneous data fusion method and device based on PBS modeling, and relates to the field of nuclear power reactor data management. First data transmitted by a target data source is acquired through a target plug-in unit; the method comprises the following steps: performing PBS modeling on a reactor in advance to obtain a preset product structure tree, and determining a target object entity to which first data belongs, so as to associate the first data with the target object entity to obtain an association relationship; and storing the association relationship to a first relational database under a data organization layer, and storing the first data to a target database under a source data layer of a data storage module according to a data type corresponding to the first data. Through the arrangement of the plug-in unit, communication with data sources following various communication protocols and / or communication formats can be realized, acquisition of multi-source heterogeneous data is realized, a preset product structure tree is established to realize data fusion organization, data fusion storage is realized by using the data storage module, and fusion and unified management of the multi-source heterogeneous data are realized.

A hexagonal silicon carbide ceramic core structure optimization method, the method is suitable for the hexagonal core structure involved in space nuclear power reactor, accelerator driven subcritical nuclear energy system, modular small reactor and other advanced reactors. The structure optimization method comprises the following steps: 1. Establishing a geometric model: setting the geometric size of the hexagonal silicon carbide ceramic core, setting the size, number and position of the fuel element or the coolant channel; 2. Adding a multi-physics field calculation module: according to the service condition, adding the irradiation field, temperature field, flow field and solid stress-strain field of the hexagonal silicon carbide ceramic core; 3. Setting the calculation domain boundary: setting the neutron flux density and irradiation time in the core, the coolant inlet temperature, the core volume heat source input, the coolant inlet velocity, and the material properties of silicon carbide; 4. Setting and parallel solving parameters: setting the number of parallel jobs and the number of cores occupied by a single job; 5. Carrying out the structure optimization design of the hexagonal silicon carbide ceramic core.

The invention discloses a device and a method for measuring a three-dimensional velocity field of a rod cluster channel, which are applied to the field of nuclear power reactor core thermal hydraulic power and are used for accurately measuring the three-dimensional velocity field of a coolant in the rod cluster channel. The lower rectifying chamber, the experimental square cavity and the upper rectifying chamber are sequentially communicated from bottom to top to form a fluid channel; a plurality of hollow-core fuel rods which are arranged in an array are arranged in the experimental square cavity, and the hollow-core fuel rods are fixed through a positioning grillwork and a positioning partition plate. The UDV ultrasonic probe system comprises a transverse UDV probe array, an axial UDV probe array and an embedded measuring unit, wherein the lateral surface of the transverse UDV probe array is arranged in a layered manner and corresponds to the rod gap; the axial UDV probe array is arranged at the top of the transverse UDV probe array; the axis of the axial UDV probe array is vertical to the cross section; the embedded measurement unit comprises an embedded UDV probe and a positioning device, the positioning device adjusts the position of the probe through an axial adjusting device, a circumferential adjusting knob and the like, and measurement coverage and accuracy are guaranteed.

A laser welding tool for the lower guide tube of a nuclear power reactor internal component includes a welding tool for the guide tube assembly and a welding tool for the double-tube in the guide tube assembly. The welding tool for the guide tube assembly includes a welding trough frame, the trough frame trough body facing one side, on which a plurality of side baffles are spaced apart, a transverse top block is provided on the other side of the welding trough frame, and a longitudinal top block is provided on the top of the welding trough frame. Both the transverse top block and the longitudinal top block are inserted into the welding trough frame and used to tighten the double-tube in the welding trough frame. The welding tool for the double-tube includes a lower support frame and an upper support frame. The lower support frame and the upper support frame clamp the guide tube assembly. The lower support frame and the upper support frame are fixed at both ends by plates and form two welding operation areas on both sides for welding operations on both sides of the guide tube assembly. The present invention adopts the above structure to effectively improve the welding quality and efficiency of the lower guide tube of the nuclear power reactor internal component.

The invention relates to the technical field of nuclear power, and provides a reactor passive starting device and a reactor passive starting method in order to reduce or eliminate a blind area range during passive starting of a nuclear power reactor. The device comprises a detector arranged in an out-of-reactor detection duct and a neutron channel box arranged in a descending section coolant layer between a pressure vessel and a hanging basket of a reactor, wherein the neutron channel box is close to the out-of-reactor detector; when the neutron channel box is a hollow channel, the passive starting method of the reactor comprises the following steps: in the starting process of the reactor, keeping that the neutron channel box is filled with high-pressure gas, keeping that a coolant does not circulate in the neutron channel box, and keeping that the coolant still circulates in other areas between the hanging basket and the pressure vessel; after the critical start of the reactor is completed, injecting a coolant into the neutron channel box; according to the method, the critical monitoring blind area in the critical starting process of the first-cycle reactor is eliminated, and blind-area-free passive starting of the first-cycle reactor is realized.

The utility model discloses a high-strength fastener for nuclear power reactor internals, which comprises an upper fastener, the lower end of the upper fastener is provided with threaded cylinders at the inner sides of the left end and the right end, and during installation, the threaded cylinders at the lower end of the upper fastener are inserted into the inner position of a total passing round hole of the lower fastener. The gasket is inserted from the lower end position of the stainless steel bolt, so that the threaded cylinder and the stainless steel bolt are connected to the inner side position in a sleeving manner, the nut B can be connected to the lower end position of the threaded cylinder and rotate along the outer wall position of the threaded cylinder, and finally the nut A is connected to the lower end position of the stainless steel bolt; when the bolt is loosened, the nut A is rotated to be tightly attached to the gasket, at the moment, the nut A wraps the nut B at the inner side position through the groove in the inner position, through the connection mode, in the vibration environment, the higher anti-loosening capacity is provided, and the bolt can be effectively prevented from loosening.

The present invention discloses a truncated antimony-beryllium neutron source rod and a reactor core for a large-scale power reactor, comprising a neutron source rod cladding, a pressing tube and a neutron source core block. The neutron source core block is arranged inside the neutron source rod cladding; the top end of the neutron source core block is connected to one end of the neutron source rod cladding through the pressing tube, and the bottom end of the neutron source core block is connected to the other end of the neutron source rod cladding through the pressing tube; the pressing tube is used to limit the axial displacement of the neutron source core block along the neutron source rod cladding; the central height of the neutron source core block is consistent with the central height of the sensitive area of the out-of-reactor detector, and the relationship between the total length of the neutron source core block and the total length of the sensitive area of the out-of-reactor detector is: h2 - h1 ≤ 50 cm, where h2 is the total length of the neutron source core block and h1 is the total length of the sensitive area of the out-of-reactor detector. The present invention is applicable to the secondary neutron source assembly of a large-scale nuclear power reactor, can reduce the emission amount of the tritium source term in the nuclear reactor environment, and helps to extend the service life and reliability of the neutron source rod.

A split-flap integral fully synchronous bolt twisting and stretching mechanism belongs to the technical field of twisting and stretching of bolts for pressure vessels of nuclear power reactors. The scheme is as follows: it includes a bolt adapter, a C-type bracket, a B-type bracket, an A-type bracket, a plurality of bolt storage racks, bolts, a robot for twisting bolts, a split-flap fully synchronous bolt stretching machine for bolt stretching, bolt elongation measurement and main nut twisting, and a control console. A C-type bracket, or a B-type bracket, or an A-type bracket is respectively arranged on each bolt storage rack. Each C-type bracket, B-type bracket, and A-type bracket are provided with a plurality of bolts. A bolt adapter is respectively connected to each bolt. The control console is connected to the twisting split-flap fully synchronous bolt stretching machine. Beneficial effect: The present invention adopts a split-flap mechanism to install an integral fully synchronous bolt twisting and stretching mechanism. Because of its small size and light weight, it is more convenient and quick to install, saves time and effort, and is particularly suitable for occasions where the installation space is limited.

An underground nuclear power reactor has a hollow blast tunnel extending from one end of a containment member and containing a nuclear reactor, heat exchanger, generator, etc. The nuclear reactor, etc. is on a movable support member. The blast tunnel defines a blast chamber having spaced apart multiple debris deflector members therein. The blast chamber has an upper wall with a top opening formed therein that is selectively closed by a top portion. If the reactor needs to be serviced or replaced, the top portion is opened to allow the reactor to pass therethrough into the blast chamber and out through the top opening. If the reactor explodes, the explosion thereof drives debris therethrough a blast door into the blast chamber where the deflector members reduce the blast force as the debris passes through the blast chamber. A simplified cooling system is provided.

The invention belongs to the technical field of nuclear reaction safety, and particularly relates to a nuclear power reactor high-current glass-metal sealing electrical penetration assembly, which comprises a metal shell with two through ends and a metal pin arranged in the metal shell in a penetrating manner, and the metal pin is in sealing connection with the metal shell through sealing glass; the two ends of the metal pin are respectively provided with a cavity, internal threads for connection are arranged in the cavities, and the cavities at the two ends of the metal pin are not communicated. One end of the metal pin is in threaded connection with an adapter, and the adapter can be in tight conductive connection with the copper conductor. Through the arrangement of the adapter, the direct contact between the copper conductor and the metal pin is avoided, and the phenomenon that the connection part of the copper conductor and the Kovar pin is burnt out due to overlarge heat productivity in a rated short-circuit heat capacity test is avoided.

The invention discloses a control rod sequence exchange method, device and equipment for a nuclear power reactor core, a medium and a product. The method comprises the following steps: determining a target control rod group meeting a current sequence exchange period according to the current sequence exchange period corresponding to at least one control rod group of a target nuclear power reactor core; if it is determined that the target control rod group meets the preset control rod sequence exchange condition, determining a current insertion position of the target control rod group, and determining a target insertion position according to the current insertion position; determining a sequence exchange execution strategy of the target control rod group according to the current insertion position and the target insertion position of the target control rod group; and controlling the control rods in the target control rod group to perform sequence exchange according to the sequence exchange execution strategy. According to the scheme of the embodiment, the control rods can be subjected to sequence exchange according to the sequence exchange period of the control rod group in the nuclear power reactor core, so that the influence of burnup shadow on the nuclear power reactor core is reduced, and the operation efficiency and safety of the nuclear power reactor core are improved.

The invention relates to the technical field of nuclear power reactor core outlet temperature detection, in particular to a reactor core outlet temperature measuring structure, which comprises an upper supporting column, a lower supporting column and a temperature sensor, the end plug is connected to the bottom of the upper supporting column, a circulation part is arranged on the end plug in the radial direction, and the end plug is communicated with the reactor core outlet through the circulation part; the detector assembly penetrates through the upper supporting column and the end plug and is provided with a thermocouple, and the thermocouple is used for detecting the outlet temperature of the reactor core; the detector assembly is inserted into the fuel assembly instrument tube; wherein the high-temperature coolant at the outlet of the reactor core enters the end plug through the circulating part and is mixed with the coolant from the instrument pipe of the fuel assembly. Through the arrangement, sufficient mixing can be ensured, so that the temperature deviation between the thermocouple measurement temperature and the reactor core outlet coolant temperature is reduced, and the thermocouple detection temperature reading can reflect the reactor core outlet coolant temperature more accurately.

The invention belongs to the technical field of nuclear power reactor pressure vessel tools, and provides an embedded metal sealing ring dismantling device and method.The dismantling device comprises a tray, a plurality of stress pieces and an ejector rod; the multiple stress pieces are arranged at equal intervals in the circumferential direction of the tray. The side, facing the groove of the CRDM pipe base, of each stress piece is provided with an extending part, and each extending part is provided with a barb part protruding towards one side in the groove. When the embedded metal sealing ring is dismounted, the barb part of each stress piece corresponds to the part, protruding out of the groove, of the embedded metal sealing ring; the ejector rod is rotatably arranged in the center of the tray in a penetrating mode, and the tray can be driven to move in the axial direction by rotating the ejector rod. According to the method, the difficulty in dismantling the conical sealing ring embedded in the CRDM tube socket can be solved, the dismantling quality of the conical sealing ring embedded in the CRDM tube socket of the EPR type nuclear power reactor pressure vessel top cover can be guaranteed, and technical support is provided for safe operation of nuclear power.

The invention belongs to the field of nuclear power reactor safety design in a marine environment, and particularly relates to a method and system for simulating three-phase contact line behaviors in a local deformation runner, and the system comprises a model building module which is used for building a multi-phase flow model coupled with multi-field force; the boundary processing module is used for processing local deformation wall surface and high-flow-speed outlet boundary conditions; and the simulation platform module is used for performing dynamic simulation and post-processing analysis and outputting key physical parameters. By constructing a special simulation platform, dynamic visualization and quantitative analysis of the whole process of formation and expansion of the dry spots in the narrow rectangular flow channel are realized for the first time, and a powerful scientific tool support is provided for revealing the internal mechanism of boiling heat transfer crisis from the mesoscale.

This invention discloses a calibration device for a reactor pressure vessel inspection robot, comprising a nozzle area calibration component and a cylinder area calibration component that work together to perform remote ultrasonic calibration. This reactor pressure vessel inspection robot calibration device enables remote in-situ ultrasonic calibration, replacing the traditional equipment effluent calibration mode. It avoids the cumbersome process of equipment effluent, calibration, and re-entry, significantly shortening the critical path duration of nuclear power reactor pressure vessel overhauls and improving overall overhaul efficiency. The nozzle area calibration component is fixed to the inspection robot's support legs via a connecting backplate and can be hoisted into and out of the water along with the robot. This eliminates the need for separate scheduling of hoisting resources for the nozzle area test block and eliminates the need for personnel to perform long-pole remote operations on the reactor pool bridge platform, significantly reducing the time spent on hoisting and re-entry of hoisting and bridge resources. The auxiliary installation components of the cylinder area calibration component can adapt to multiple operating conditions, expanding the device's applicability.

The invention relates to a laser welding process for an upper guide cylinder of a nuclear power reactor internal and an auxiliary tool thereof. The tool comprises an adjusting support, an upper base, a lower base and end covers, the top of the adjusting support is movably connected with the lower end face of the lower base, the upper base and the lower base are symmetrically arranged along the axis, and the two ends are fixedly connected through the end covers. And a plurality of adjusting blocks are arranged on the upper base and the lower base and are used for adjusting the welding seam levelness of the nuclear power reactor internals. The welding process comprises the steps that the height of the adjusting support is adjusted to enable the lower base to be horizontal, nuclear power reactor internals are placed, the upper base is buckled, the end cover is connected and pre-tightened, the position of the upper base is adjusted to enable the upper base to be horizontal with the lower base, the adjusting block is adjusted to enable a weld joint to be horizontal, and finally laser welding is conducted. In the welding process, spot welding and fixing are conducted through laser welding, and then laser penetration welding is conducted; through the design of the adjusting bracket and the adjusting block, the levelness and the welding quality of a welding seam are ensured, the welding efficiency is improved, the cost is reduced, the operation process is simplified, and the high-precision welding device is particularly suitable for the high-precision welding requirement of nuclear power reactor internals.