43 results about "Pressurized water reactor" patented technology

This invention discloses a method for producing radioactive isotopes in the core of an advanced high-temperature reactor using a spherical target. This invention overcomes the limitations of using neutron irradiation channels or fuel-related components of nuclear reactors, such as conditioning rods in heavy water reactors and combustible poison rods in pressurized water reactors, to produce radioactive isotopes. It sets a precedent for the mass production of radioactive isotopes using the core of a nuclear reactor, and has significant strategic implications for exploring high-value applications of commercial advanced high-temperature reactors to reduce operating costs and mass-produce radioactive isotopes, including the strategic isotope tritium.

The present application relates to nuclear reactor irradiation technology field, especially provide a kind of coolant bypass control structure for pressurized water reactor test loop, including pressure pipe, the pressure pipe is placed in the inside of heat pipe, spherical lower connector is arranged at the bottom of pressure pipe, the inside of pressure pipe is equipped with shunt pipe, and square round connector is arranged at the bottom of shunt pipe;Loop coolant flows to spherical lower connector through pressure pipe and turns back, and loop coolant is divided into main stream and bypass stream by square round connector at the bottom of shunt pipe;Fuel assembly is installed in the inside of shunt pipe;Lower tube base is arranged at the lower side end of fuel assembly, main stream loop coolant flows through the inside of fuel assembly through the center through-hole of lower tube base, and bypass stream loop coolant flows through the annular gap between lower tube base and shunt pipe from the annular gap between square round connector and lower tube base.The purpose is to retain a small part of bypass stream in fuel assembly irradiation test, and form stable and controllable main stream bypass stream distribution mechanism.

The present application belongs to the technical field of radiation protection and pressurized water reactor source term calculation, and particularly relates to a pressurized water reactor pellet-cladding gap fissile product source term fast calculation method. The method comprises the following steps: step 1: establishing a first-order kinetic balance equation of the fissile product in the fuel rod; step 2: using Chebyshev rational approximation method to solve the accumulation of the nuclides in the decay chain corresponding to different mass numbers in the fuel rod; step 3: calculating the release production ratio of the fissile product nuclides; step 4: judging the release path of the target nuclides from the fuel pellet to the pellet-cladding gap; and step 5: judging the fissile product source term in the pellet-cladding gap according to the release path of the fissile product. The present application has the beneficial effect that the fissile product source term in the fuel rod is calculated in detail, and the fissile product source term in the pellet-cladding gap can be obtained, which is used for the release source term calculation when the fuel cladding is damaged subsequently.

This invention relates to the field of metal additive manufacturing, and more particularly to a pressure vessel steel powder for nuclear power plants and its preparation method. The pressure vessel steel powder comprises the following components: C: ≤0.2wt%; Si: 0.1-0.3wt%; Mn: 1.15-1.60wt%; S: ≤0.005wt%; Cr: ≤0.25wt%; Ni: 0.5-0.8wt%; P: ≤0.008wt%; Mo: 0.43-0.57wt%; V: ≤0.01wt%; Cu: ≤0.08wt%; Al: ≤0.04wt%; Co: ≤0.03wt%; with the balance being iron. This invention achieves precise control of the powder's chemical composition through targeted process design and impurity control strategies, ensuring that all performance indicators fully comply with the requirements of the "Design and Construction Rules for Mechanical Components of Pressurized Water Reactor Nuclear Island" standard.

This invention relates to the field of nuclear power plant maintenance equipment, and more particularly to a pneumatic wrench for rotating the main nut in a pressurized water reactor nuclear power plant. The pneumatic wrench includes: a guide rod connected to the lower end of a first sleeve; a second sleeve mounted on the outer surface of the guide rod's side end, with a third sleeve connected to its lower end; a fixing plate connected to the outer surface of the second sleeve's side end, with a small gear and a transition gear mounted on the upper part of the fixing plate, a pneumatic motor connected to the upper end of the small gear, a large gear ring on one side of the transition gear, a sleeve connected to the lower end of the large gear ring, a bearing installed at the connection between the sleeve and the second sleeve, an inner liner inside the sleeve, a limit block connected to the side end of the inner liner, a groove formed at the connection between the limit block and the large gear ring, a main nut inside the limit block, a main bolt inside the main nut, and a connecting seat connected to the upper end of the main bolt; a connecting plate connected to the outer surface of the first sleeve's side end, with pins installed on the connecting seat and the connecting plate. This invention can disassemble and assemble various types of nuts, and has a wide range of applications.

The application discloses a digital fuel physics supervision system and method suitable for a pressurized water reactor, and relates to the field of fuel management of nuclear power plants. The digital fuel physics supervision system comprises a basic information management module, a fuel full life cycle management module, a reactor core supervision module, a physics test management module and a data security and permission control module. The basic information management module stores data such as bases, units and fuel assembly types; the fuel full life cycle management module covers the whole process of purchasing, manufacturing, refueling and spent fuel treatment; the reactor core supervision module monitors parameters such as nuclear power in real time and records abnormalities; the physics test management module controls the whole process of starting a physics test; and the data security and permission control module assigns permissions based on roles. The digital fuel physics supervision system realizes digitalization, standardization and full cycle tracing of fuel supervision, improves management efficiency and safety, and is suitable for fuel physics supervision of pressurized water reactor nuclear power plants.

The present application relates to the technical field of reactors, and particularly relates to a calculation method of corrosion product deposition rate of a pressurized water reactor. The calculation method of corrosion product deposition rate of the pressurized water reactor comprises the following steps: obtaining the concentration distribution and Zeta potential distribution of metal ions related to corrosion products; obtaining the electric field force on the particle state corrosion products based on the Zeta potential distribution; tracking the transport trajectory of the particle state corrosion products by using a discrete phase model, wherein the force on the particles in the transport process comprises the electric field force; calculating the particle mass flow rate according to the discrete phase model; and calculating the deposition rate and erosion rate of the particle state corrosion products on the wall surface according to the particle mass flow rate. The present application can be used to solve the problem that the current calculation model cannot accurately predict the CRUD deposition rate, and can improve the authenticity of CRUD growth prediction and the reliability of reactor safety analysis.

The application provides a fuel and a method for increasing the neutron source intensity of a nuclear reactor, wherein the abundance of O in the fuel is 1% to 50% 18 O is 1% to 50%, and alpha particles and the enriched 18 O can significantly enhance the initial neutron source intensity of a nuclear reactor core in a pressurized water nuclear power plant, and help to improve the start-up performance and reduce the dependence on external neutron sources. In the passive start-up mode without using an external neutron source, the application improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant. 18 O in the fuel, and improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant.

The application provides a small-sized pressurized water reactor, relates to the technical field of nuclear power, and aims at improving the technical problem that the reactor main loop is not simplified and the equipment is not miniaturized due to the existence of a coolant pump in the prior art small-sized pressurized water reactor with forced circulation. The application comprises a pressure vessel, a reactor core, a heat exchanger, a stabilizer and a coolant driving device, wherein the coolant driving device comprises a driving rotor assembly arranged in the inside of the pressure vessel and a driven rotor assembly arranged in the outside of the pressure vessel; the driving rotor assembly comprises a driving motor and a driving rotor connected with the driving motor; the driven rotor assembly comprises a driven rotor and an impeller connected with the driven rotor; and the driving rotor is magnetically connected with the driven rotor. The application adopts the magnetic connection between the driving rotor and the driven rotor, and no corresponding connecting part is arranged on the pressure vessel, so that the safety is ensured while the equipment volume is reduced.

The invention discloses a pressurized water reactor fuel assembly damage position prediction method based on machine learning, and the method comprises the steps: building and training a machine learning model through a decision tree algorithm based on time evolution data of damaged primary loop nuclide concentration obtained by reactor core physical analysis software and evolution simulation calculation software; according to the method, a preprocessed multi-dimensional nuclide concentration stable value feature vector is taken as input, a prediction result of a damage position is taken as output, model overfitting is effectively controlled through pre-pruning and post-pruning strategies, generalization ability is ensured, the damage probability of each position can be provided, and prediction of the damage position of the pressurized water reactor fuel assembly is realized.

The present application relates to a kind of preparation methods of pressurized water reactor MOX fuel powder, comprising the following steps: by mixed actinide salt and urea solid preparation mixed actinide eutectic, mixed actinide eutectic is directly obtained by deamination, denitrogenation / decarbonation, reduction, MOX fuel powder;The mixed actinide salt is hexavalent actinide nitrate and tetravalent actinide oxalate;When the mixed actinide salt is uranyl nitrate hexahydrate solid and thorium oxalate hexahydrate solid, the MOX fuel powder is (U x Th 1‑x )O2;When the mixed actinide salt is uranyl nitrate hexahydrate solid and plutonium oxalate hexahydrate solid, the MOX fuel powder is (U x Pu 1‑x )O2.The method provided by the present application has simple raw materials, can greatly simplify the preparation process of MOX fuel powder composed of UO2 and PuO2 (or ThO2), reduce the use of process reagent, and does not produce any process waste liquid, thereby providing a simple, high production efficiency, waste liquid-free, more economical and green MOX fuel powder preparation method.

Disclosed are a 177-core pressurized water reactor control rod bundle grouping arrangement method and a 177-core pressurized water reactor control rod in-bundle arrangement method. The control rod in-bundle arrangement method comprises respectively arranging four strong absorber rods at the position of coordinates (4, 4) in a grid array and positions of rotating same 90°, 180° and 270° around a gauge pipe grid cell. The control rod bundle grouping arrangement method comprises arranging control rod bundles which are divided into T rod groups, S rod groups and AO rod groups in a reactor core of 177 fuel assemblies. While the safety of reactor cores is ensured, the capability of adjusting and controlling control rods can be enhanced, so as to satisfy the control requirements for various load changes. Given a specified quantitative ratio of strong absorbers to weak absorbers, a stronger control capability and less disturbance to local power distribution are allowed.

The application discloses a method for hot stretching of a bolt of a reactor coolant shielding motor main pump of a non-active advanced pressurized water reactor, and the method is suitable for installation and tightening of large full-thread bolts and nuts, can be installed by using the thermal expansion and cold shrinkage characteristics, and is suitable for equipment operating at high temperature; the heating rod of the method can be installed in the bolt, so that the space requirement of the mechanical equipment required for installation equipment is reduced, and the method can be used in a narrow space; the required elongation is divided into two times of hot stretching and one time of balance stretching, the first time of hot stretching can eliminate the influence of the compressible amount between the pump shell and the main pump, the second time of hot stretching is carried out on the bolts according to the stretching law, and the last balance stretching is carried out by the stretching law to stretch the bolts to the position, so that all the bolts can obtain the same stretching amount.

The application discloses a nuclear material accounting method, system and device for a damaged fuel assembly, and relates to the technical field of nuclear material accounting management of nuclear power plants. The method comprises the following steps: acquiring fissile product specific activity measurement data in a pressurized water reactor primary coolant; establishing a fissile product release and migration dynamics model; under the condition that the fissile product release reaches a steady state, calculating the specific activity of the fissile product diffused and released into the primary coolant and the specific activity of the primary loop fissile product released by recoil, and obtaining a total specific activity; converting the total specific activity into a normalized fission rate; drawing a diffusion release curve and a recoil release curve; extrapolating the recoil release curve to a preset decay constant threshold to obtain a fission rate corresponding to complete recoil release; and realizing nuclear material accounting of the damaged fuel assembly by calculating the fission rate of the free heavy uranium nuclide in the primary loop. The method makes up for the blind area that the damaged fuel assembly cannot implement nuclear material accounting, and provides a reference for monitoring the damage degree of the fuel assembly of the power plant.

This invention provides a passive pressurized water reactor (PWR) nuclear island main building and nuclear power plant. The main building comprises a reactor building, a nuclear auxiliary building, and a nuclear island electrical building. The reactor building provides containment space for the nuclear fission reaction; the nuclear auxiliary building houses radioactive materials and non-radioactive materials preferably placed close to radioactive areas; the nuclear island electrical building houses non-radioactive materials. In a top view, the main building has a rectangular outline. The nuclear auxiliary building and the nuclear island electrical building surround the reactor building, with the reactor building located in the central region of the rectangle. The nuclear auxiliary building and the nuclear island electrical building are located on either side of the reactor building, and their outer outlines together form a rectangle. This passive PWR nuclear island main building and nuclear power plant improves the compatibility of the building layout structure.

The present application relates to a pressurized water reactor fuel cladding breakage online monitoring method and a monitoring system thereof, the method comprising the following steps: step S1: obtaining a pre-established energy calibration relationship, which is a corresponding relationship between gamma ray energy and channel number; step S2: obtaining pre-calibrated efficiency curve parameters, which are obtained through active efficiency calibration based on a simulated tubular standard radioactive source, the material, outer diameter and inner diameter of the simulated tubular standard radioactive source being consistent with those of the measured pipeline; step S3: obtaining gamma energy spectrum data through a back-to-back coincidence measurement technology, and automatically deconvoluting the gamma energy spectrum data according to the energy calibration relationship and the efficiency curve parameters to calculate the type and activity concentration of fissile nuclides in the measured pipeline; and step S4: determining a corresponding preset threshold value according to the type of fissile nuclides, and comparing the activity concentration with the preset threshold value to obtain a monitoring result of pressurized water reactor fuel cladding breakage.

The utility model discloses a nuclear power plant pipeline deposition source item simulation body, this nuclear power plant pipeline deposition source item simulation body includes closure, radiation source suction accessory and pipeline, radiation source suction accessory adsorbs multiple radionuclide of gamma ray energy 59.5keV-1332.5keV to be closed in closure, the closure is pasted in the inner surface of pipeline, the utility model discloses can simulate the deposition source item of pipeline inner surface through the closure of radiation source suction accessory, the closure of radiation source suction accessory is pasted in the inner surface of pipeline, can be used to simulate the condition of the consistent deposition source item of pressurized water reactor primary loop pipeline in the pipeline, is used to confirm the active efficiency scale of deposition source item measurement, greatly reduces the risk of radioactive contamination in the active efficiency scale process.

The application discloses an advanced safety injection system and a pressurized water reactor, and belongs to the nuclear power field. The advanced safety injection system comprises an integrated pressure water tank, the integrated pressure water tank comprises a CMT cavity and an ACC cavity which are separated by a conical wall extending vertically along an axis, the ACC cavity is arranged outside the CMT cavity, the cross-sectional area of the ACC cavity gradually decreases from top to bottom, and upper and lower through holes are arranged on the conical wall to communicate the CMT cavity and the ACC cavity. The ACC cavity is provided with pressure gas at the top; the CMT cavity is provided with a top inlet communicated with a reactor primary cooling system at the top end and is provided with an injection pipeline for injecting coolant into a reactor core at the bottom end. The advanced safety injection system has simple structure, can play the CMT and ACC functions under different accident conditions by using a single water tank, effectively reduces the manufacturing and maintenance costs, and improves the safety of the reactor.

This invention belongs to the technical field of nuclear reactor systems, specifically relating to an automatic depressurization system for a pressurized water reactor. It includes four stages of depressurization valves. The first three stages are connected to the pressurizer steam chamber and share a connecting pipe with two of the three pressurizer safety valves. The discharge from the first three stages is collected and then enters the depressurization water pool inside the containment via a sprayer, and is subsequently discharged into the containment through the depressurization water pool. The automatic depressurization system proposed in this invention has a simpler configuration, reduces the number of devices by 20%, has higher reliability, simplifies operation, and reduces the power supply load. The fourth stage uses an electric valve instead of the explosion valve in units such as the AP1000, meeting the start-up and operation requirements of the passive safety injection system and improving the safety of the nuclear power unit.

This invention specifically relates to a method for measuring the quadrant power tilt ratio of a pressurized water reactor core, comprising the following steps: S101, based on the power distribution test data of the pressurized water reactor at various power levels, calculating the temperature difference ΔT between the thermocouple thermometer temperature at the core outlet and the core inlet temperature at various power levels of the pressurized water reactor. i With each fuel assembly containing a thermocouple thermometer, the power p i The correspondence f i And the power of all fuel assemblies in each quadrant j of the reactor core and P Qi The reactor core contains thermocouple thermometers in each quadrant j, and the fuel assembly power and P... Tj Ratio K j S102. Based on the power distribution data under the operating power of the pressurized water reactor, combined with f i and K j The power tilt ratio (QPTR) of the core in each quadrant is calculated under the operating power of the pressurized water reactor. This invention uses f... i Calculate the power p of each fuel assembly containing a thermocouple thermometer. i Reduce the calculation bias of the core quadrant power tilt ratio; consider K j This reduces the calculation bias of the core quadrant power tilt ratio.

The application discloses a method and device for predicting a peak value of a fissile product of a pressurized water reactor nuclear power plant. The method comprises: obtaining a first nuclide concentration of each fissile product in a primary coolant and an operating parameter at a current time in an operating interval of the pressurized water reactor nuclear power plant; determining a fuel damage parameter corresponding to the first nuclide concentration and the operating parameter based on a preset fuel cladding rupture diagnosis algorithm; determining a second nuclide concentration prediction value of each fissile product in the primary coolant at the end of the current operating interval based on a first prediction model, the operating parameter and the fuel damage parameter; and determining a third nuclide concentration prediction peak value of each fissile product in the primary coolant in a shutdown interval based on a second prediction model, the second nuclide concentration prediction value, the operating parameter and the fuel damage parameter. The method can realize the prediction of the nuclide concentration peak value of the fissile product in the shutdown condition, and then the shutdown strategy can be formulated or modified according to the predicted concentration peak value, so as to guarantee the safety and reliability of the shutdown.

The application discloses a pressurized water reactor heat storage energy peak shaving system, and belongs to the field of reactor engineering and energy storage technology. The system comprises a pressurized water reactor secondary loop, a carbon dioxide energy storage loop and a carbon dioxide power generation loop. At the electricity trough of the power grid, the pressurized water reactor secondary loop drives a steam turbine with excess steam, the turbine is coaxially connected with a compressor of the energy storage loop, so that the excess electric energy is converted into heat energy and stored in a high-temperature molten salt tank, and domestic hot water is supplied at the same time. At the electricity peak of the power grid, the steam turbine in the pressurized water reactor secondary loop generates power at full load, and the power generation loop generates power by using the stored molten salt heat energy, and provides low-temperature heat energy required for the next cycle of the energy storage loop. In the application, molten salt and water are combined as heat storage medium, and through multi-loop coupling and key equipment optimization, efficient, compact and flexible heat storage energy and peak shaving are realized, the application is suitable for energy storage reconstruction of nuclear power plants, and has the advantages of high energy density, small land occupation, low investment cost and environmental friendliness.

The invention relates to the field of nuclear safety, in particular to an online monitoring and online protection calculation method and system suitable for a large pressurized water reactor core. The method comprises the following steps: selecting a typical operation state parameter combination according to the range of key operation state parameters of a reactor core; receiving real-time measurement data of the operation state parameters, calculating a plurality of sets of intermediate coefficients based on the reactor core state at the current moment t0 according to the selected operation state parameter combination, forming an intermediate coefficient library containing the plurality of sets of intermediate coefficients, and then carrying out rapid protection calculation on the formed intermediate coefficient library; meanwhile, a set of intermediate coefficients is calculated based on the reactor core state at the current moment t1 according to the current running state parameters measured in real time, and after calculation is completed, rapid monitoring calculation is conducted through the set of intermediate parameters. The system comprises a measuring system, an upper layer unit and a lower layer unit, the upper layer unit comprises a fine tracking calculation module and a rapid monitoring calculation module; and the lower layer unit comprises a rapid protection calculation module. According to the method, accurate and efficient calculation of the key safety parameters of the online protection system is realized, and the unit operation flexibility and economy of the pressurized water reactor core nuclear power plant are improved.

The present disclosure discloses a method and device for reducing temperature and pressure of a loop, electronic equipment and a storage medium. According to the present application, the deviation information is dynamically determined based on the real-time temperature and pressure parameters of the reactor primary loop, the fuzzy controller domain parameters are adaptively adjusted, the precise association of dynamic parameters and control rules is established, the valve control instruction is automatically generated through fuzzy reasoning to adjust the opening degree, and the subjectivity and hysteresis of manual experience judgment and manual adjustment are avoided. Therefore, the technical problem that the response lag and control deviation accumulation are caused by manual adjustment and the adaptive association of dynamic parameters and control rules is not established in the existing pressurized water reactor primary loop control method, which further affects the operation safety and economy of the nuclear power plant, can be solved, the response speed and control accuracy of the temperature and pressure reduction process are improved, the control deviation is reduced, the stability of the transition condition from hot shutdown to cold shutdown is ensured, and the technical effects of strengthening the operation safety and economy of the nuclear power plant are achieved.

A method treats an internal surface of a metal part made from a nickel-based or cobalt-based alloy of a primary circuit during an operating cycle of a PWR reactor. The method includes estimating a quantity of nickel discharged into the primary circuit by the metal parts of the primary circuit in contact with water of the primary circuit, during an operating cycle of the reactor; calculating a quantity of magnetite necessary for fixing the whole of the quantity of nickel estimated; forming a treated water by dissolving the quantity of magnetite calculated in the water of the primary circuit when the water has a temperature of between 80° C. and 180° C.; and putting this treated water in circulation in the primary circuit so as to form, on the internal surface of at least one metal part, a layer of a protective oxide to fix the quantity of nickel estimated.

This invention belongs to the field of reactor instrumentation and control, specifically relating to a hybrid steam emission control system and method applied to a small pressurized water reactor. It includes a data acquisition module, a data processing module, and an output module. The data acquisition module controls the system input signals including main steam header steam pressure, main steam header pressure setpoint, nuclear power, turbine load, and final power setpoint. The data processing module includes a regulating control unit and a rapid opening control unit. The output module, under the premise that the system allows the steam emission valve to operate, fully opens the steam emission valve when the rapid opening control command is "1"; and actuates the steam emission valve to the opening degree required by the regulating control command when the rapid opening control command is "0". The beneficial effect is that by real-time monitoring of the power difference and steam pressure deviation between the primary and secondary loops, the power difference control channel and pressure difference control channel are triggered when transient conditions such as turbine tripping or load shedding occur.

The application discloses a kind of reactor core impurity deposition prediction methods, steps are as follows:1, establish coolant flow heat exchange model and fuel rod heat conduction analysis model;2, establish corrosion product deposition amount model;3, establish deposit thermal resistance model and is coupled with fuel rod heat conduction analysis model;4, establish deposit multi-physics field coupling solution method, obtain the reactor core coolant flow field and temperature field under the influence of impurity deposit at a time;5, carry out the calculation of multiple deposition time results.This method simplifies the geometry structure in the reactor core of pressurized water reactor, considers the influence of deposit on the internal physical quantity of reactor core by establishing the deposit model of reactor core cladding corrosion product, and solves multiple physical models, realizes the accurate prediction of multi-physics field under the condition of full-scale, long-period deposition state of reactor core, and has the advantages of fast calculation speed and less resource consumption.The calculation result can provide an efficient and accurate computational fluid dynamics method for long-term prediction of pressurized water reactor full reactor core and full life cycle.

The dual isolation valve system for a fuel transfer tube in a pressurized light water reactor nuclear power plant, according to the present invention, is characterized in that: in order to withdraw nuclear fuel from a fuel building and reload nuclear fuel into a reactor within a reactor building through a fuel transfer tube, a motor-operated gate isolation valve and a manual isolation valve are installed at the front side of the fuel transfer tube, relative to the fuel building side, in order to fill and drain a reactor refueling water storage tank; the manual isolation valve on the fuel building side is maintained in an open state; a plurality of disc-fastening holes are formed along the outer circumference of the disc of the motor-operated gate isolation valve; a disc protrusion is inserted into the inside of the fuel transfer tube; a disc step circumference is larger than the outer diameter of the fuel transfer tube so as to seal the fuel transfer tube; and the disc of the motor-operated gate isolation valve moves downward due to the driving of a motor (M), causing the motor-operated gate isolation valve to block the fuel transfer tube and maintain the tube in a closed state, the disc being an isolation valve disc that functions as a blind flange.

The application discloses a method for analyzing migration and deposition of activated corrosion products of a pressurized water reactor unit main pipeline, and comprises the following steps: selecting a particle phase deposition-erosion model on a CFD platform; self-defining a dissolution-precipitation mechanism model; determining calculation working conditions and boundary conditions of the CFD platform; wherein the calculation working conditions comprise: transient operation working conditions of the unit under different three-dimensional flow fields, temperature fields and pressure fields in a unit overhaul stage; performing fluid grid division on a pre-built main pipeline model; after the grid is imported into the CFD platform and the particle phase deposition-erosion model and the dissolution-precipitation mechanism model are set and coupled, the deposition rate of the activated corrosion products of the main pipeline under the transient operation working conditions of the unit under different three-dimensional flow fields, temperature fields and pressure fields in the unit overhaul stage is solved to be more accurate and effective, which is of great significance to safe and economic operation of a nuclear power unit.

This invention belongs to the field of pressurized water reactor technology, specifically relating to a deflector plate and reactor top structure. The reactor top structure includes a pressure vessel top cover, a ventilation structure, a cylindrical cofferdam assembly, a seismic support plate, and a deflector plate. The cofferdam assembly is fixedly installed on the upper end of the pressure vessel top cover. A ventilation structure is provided on the outer wall of the cofferdam assembly. The deflector plate is installed on the reactor control rod drive mechanism inside the cofferdam assembly. A flange is provided on the top of the cofferdam assembly for installing the seismic support plate. The seismic support plate is flange-shaped and is bolted to the flange on the top of the cofferdam assembly, fixing it above the cofferdam assembly. This invention solves the problems of forced convection cooling in existing pressurized water reactor tops, such as the need for electricity, noise generation, poor economic efficiency, impact on nuclear safety, difficulty in installation, poor support, and poor heat dissipation.