14 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 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.

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 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 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 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.