23 results about "Nuclear power reactor" patented technology

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 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 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 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.

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.

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 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 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 application discloses an Am-Be neutron source assembly and a reactor core suitable for a large nuclear power reactor, the neutron source assembly comprising a star-shaped frame and a plurality of neutron source rods hung on the star-shaped frame; the neutron source rods are loaded with Am-Be materials to provide a neutron source for physical start of the reactor core; the number of the neutron source rods depends on the attenuation caused by the water gap thickness of the lower section of the pressure vessel. The neutron source rod comprises a neutron source rod cladding, a pressing tube and Am-Be material blocks, and the Am-Be material blocks are arranged in the neutron source rod cladding. The neutron source rod further comprises Sb-Be material blocks, and the Am-Be material blocks and the Sb-Be material blocks are alternately arranged along the axial direction of the neutron source rod cladding. The application saves the high procurement cost of Cf-252 sources, avoids the problem that there is no neutron source available due to long-term shutdown of the power plant caused by various accidents, and no longer uses a separate secondary neutron source of the reactor core.

This invention discloses a refueling system and method for a small nuclear power reactor. The system, located within the reactor building, includes a reactor pool, a spent fuel pool, and a fuel transfer channel. The spent fuel pool comprises a spent fuel storage area and other storage areas separated by walls lower than the spent fuel pool itself. The fuel transfer channel is a passage enclosed by walls and extends from the bottom of the spent fuel pool to the reactor pool. During refueling, the water in the reactor pool and spent fuel pool is at the refueling level, which is higher than the walls. The reactor pool is connected to the spent fuel pool via the fuel transfer channel, and the spent fuel storage area is connected to the other storage areas. After refueling, the water in the reactor pool and fuel transfer channel is drained, and the spent fuel pool is drained to the storage level, which is lower than the walls. The reactor pool and spent fuel pool, as well as the spent fuel storage area and other storage areas, are now isolated. This invention improves the reliability of the refueling system.

The application discloses a zirconium alloy cladding for nuclear power reaction accident fault-tolerant fuel and a preparation method thereof. The zirconium alloy cladding for nuclear power reaction accident fault-tolerant fuel comprises a zirconium alloy base material, and a zirconium oxide bottom film layer and an aluminum-zirconium mixed film layer are sequentially arranged on the surface of the zirconium alloy base material. The zirconium alloy cladding for nuclear power reaction accident fault-tolerant fuel has excellent interface bonding between layers, uniform thickness of each layer, compact structure, and finally has good corrosion resistance, wear resistance and high-temperature water vapor corrosion resistance.

The invention discloses a coupling virtual power plant peak regulation method and system based on heating and steam supply of a small nuclear power reactor, and belongs to the technical field of nuclear energy peak regulation. The coupling virtual power plant peak regulation method comprises the steps that three types of core influence factors of the small nuclear power reactor side, the virtual power plant side and multi-energy cooperation are disassembled; constructing a four-dimensional index system covering peak regulation efficiency, peak regulation capability, system stability and economy; the cooperative peak regulation performance coefficient is constructed by quantitatively calculating index values, the performance of a coupling system and the performance of a single peak regulation system are compared, and operation parameters are optimized for short board indexes. The method realizes nuclear energy and virtual power plant multi-energy cooperation, improves the peak regulation capability, provides a scientific decision basis through quantitative evaluation, has the advantages of economy and environmental protection, can adapt to various differentiated scenes, and is significant in application value.

The utility model relates to the technical field of nuclear power reactor underwater measurement, and provides an underwater three-dimensional measurement device, the underwater three-dimensional measurement device comprises a shell and a rotation measurement mechanism, the shell is used for fixedly sleeving a control rod guide cylinder, so that a cushion block at the top end of the control rod guide cylinder is arranged in the shell; the rotation measuring mechanism comprises a rotation driving assembly and a three-dimensional scanning assembly which are arranged in the shell, the rotation driving assembly is rotationally connected with the shell, the three-dimensional scanning assembly is fixedly connected to the rotation driving assembly and corresponds to the cushion block, and the rotation driving assembly is used for driving the three-dimensional scanning assembly to rotate around the cushion block in the circumferential direction so as to collect images of the cushion block. The device can accurately measure the wear condition of the cushion block, and has the characteristics of simple overall structure installation, stability, reliability and the like.

The invention discloses a zirconium hydride-based moderated shielding material containing gadolinium oxide / samarium oxide / europium oxide and a preparation method and application thereof, and belongs to the technical field of radiation shielding. The problems of hydrogen loss, cracking, poor thermal stability and the like of the existing zirconium hydride are solved. The density of the material is 5.58-5.71 g / cm < 3 >, the hydrogen content is 1.6 wt%-1.83 wt%, the mass proportions of all the substances in the total amount of ZrH2, Gd2O3, Sm2O3 and Eu2O3 are as follows: 99% < = ZrH2 < 100%, 0% < = Gd2O3 < = 1%, 0% < = Sm2O3 < = 1%, 0% < = Eu2O3 < = 1%, 0% < Gd2O3 + Sm2O3 + Eu2O3 < = 1%, the hydrogen desorption initial temperature is greater than or equal to 557 DEG C, multi-element synergistic absorption is utilized, the shielding efficiency for 1MeV fast neutrons is greater than or equal to 97.5%, the thermal neutron ratio is reduced by greater than or equal to 80%, and the material is suitable for small nuclear power reactors below 600 DEG

A zirconium hydride-based moderation shielding material containing gadolinium oxide / samarium oxide / europium oxide, its preparation method, and its application belong to the field of radiation shielding technology. It solves the problems of hydrogen loss, cracking, and poor thermal stability inherent in existing zirconium hydride materials. The material of this invention has a density of 5.58–5.71 g / cm³. 3 The hydrogen content is 1.6wt%~1.83wt%. The mass proportions of each substance in the total amount of ZrH2, Gd2O3, Sm2O3 and Eu2O3 are: 99%≤ZrH2<100%, 0%≤Gd2O3≤1%, 0%≤Sm2O3≤1%, 0%≤Eu2O3≤1%, 0%<Gd2O3+Sm2O3+Eu2O3≤1%. The hydrogen release initiation temperature is ≥557℃. Utilizing multi-element synergistic absorption, the shielding efficiency for 1MeV fast neutrons is ≥97.5%, and the thermal neutron proportion is reduced by ≥80%. It is suitable for small nuclear power reactors below 600℃.

The embodiment of the invention provides a nuclear instrumentation system alarm method and device, electronic equipment and a storage medium, and belongs to the technical field of nuclear power. The method comprises the following steps: calculating based on initial source intensity of a preset neutron source and a nuclear power reactor core charging sub-step sequence to obtain a sub-step sequence theoretical counting rate of the nuclear power reactor core charging sub-step sequence; screening the sub-step sequence theoretical counting rates to obtain the minimum theoretical counting rate of the preset neutron source measured by the source range detector; simulating the original counting rate distribution data based on the minimum theoretical counting rate; performing loss probability calculation based on the original counting rate loss alarm threshold and the original counting rate distribution data to obtain a target counting rate loss probability; performing counting rate optimization on the nuclear instrumentation system based on the target counting rate loss probability; based on the current reactor core charging step sequence and the current reactor core charging moment, a current counting rate is obtained through calculation; and alarming the current counting rate based on the optimized nuclear instrumentation system. According to the embodiment of the invention, the accuracy of nuclear instrument system alarm can be improved.

The application relates to a supporting system for a half-pool type loop nuclear power reactor, which comprises a supporting ring, a plurality of ring cavities formed in the longitudinal direction on the supporting ring, a plurality of vertical supporting structures arranged at the peripheral positions of the ring cavities, each vertical supporting structure comprising a plurality of vertical supporting structures, any vertical supporting structure comprising a support arranged at the peripheral position of the corresponding ring cavity of the supporting ring and a lug used for fixing a corresponding device in the half-pool type loop nuclear power reactor, the lug being mounted on the support to arrange the corresponding device in the corresponding ring cavity, and a plurality of positioning structures distributed below the supporting ring to horizontally fix the supporting ring on a concrete foundation. The supporting system can realize the supporting function of a main container, a main heat exchanger and a main pump and ensure the levelness of the main container, the main heat exchanger and the main pump.