337 results about "Pressurized water reactor" patented technology

A pressurized water reactor oxidation corrosion product cross-scale nuclear-thermal-material coupling simulation method comprises the following steps: in the first stage, obtaining a supercooled boiling position through mutual loop iteration simulation of a neutronics model and a thermal hydraulic model; and in the second stage, multi-physics field coupling simulation is carried out through an oxidation corrosion product (CRUD) model, a neutronics model and a thermal hydraulic model, and high-resolution simulation under the pressurized water reactor nuclear-thermal-material coupling condition is obtained. According to the method, a scene in which the thickness range of the oxidation corrosion product is 20-200 microns is considered, multi-physical fields such as pressurized water reactor neutron physics, flow heat transfer and oxidation corrosion product deposition are calculated and analyzed, and the complex physical process can be captured more accurately.

The invention discloses a pressurized water reactor primary loop boron concentration prediction and control method and system, and the method comprises the steps: obtaining an original helium flow sequence of a pressurized water reactor primary loop, and obtaining M boron concentration IMF components IMF1,..., IMFM; the M boron concentration IMF components IMF1,..., IMFM are put into a trained LSTM model for prediction, predicted values Y1, Y2,..., YM of the IMF components are obtained, and error coefficients delta Y1, delta Y2,..., delta YM corresponding to the boron concentration IMF components are calculated according to the predicted values Y1, Y2,..., YM of the IMF components; and inputting error coefficients delta Y1, delta Y2,..., delta YM corresponding to each boron concentration IMF component into the trained DenseNet prediction model to obtain an optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence, based on a DEM-DAC algorithm, controlling the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs1 # of the boron concentration signal sequence, and calculating the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence. The method and the system can solve the technical problems that the boron concentration fluctuation of the primary loop is large and the unit load regulation performance is poor during the reactor variable load operation period.

A pressurized water reactor core oxidation corrosion product deposition multi-physical coupling hardware-in-loop device comprises a corrosion product accelerated deposition system, an element online testing system and an axial power offset digital mirror image system. According to the method, the complex multi-physics field coupling process of hydrochemistry, material corrosion, flow heat transfer, neutron physics and the like involved in oxidation corrosion product deposition can be achieved, off-line data transmission between single physics fields is avoided, the growth rule of a deposition layer is directly obtained, the deposition mechanism of oxidation corrosion products is analyzed, and the deposition efficiency is improved. The defects that an existing device cannot perform entity verification and cannot obtain element concentration and distribution in a corrosion product deposition layer are overcome.

The invention provides a pressurized water reactor core axial power offset phenomenon prediction method and electronic equipment. The prediction method comprises the following steps: firstly, constructing a prediction model of the axial power offset phenomenon of the reactor core; wherein the prediction model is used for representing a model that the axial power offset phenomenon of the reactor core changes along with oxidation corrosion sediment CRUD and fuel burn-up on the surface of a cladding, and the prediction model comprises a neutron physical module, a thermal hydraulic module and a CRUD material module which have a coupling relationship; and then, carrying out coupling calculation on the prediction model within a preset duration, and generating prediction information of the axial power offset phenomenon of the reactor core. According to the prediction mode, the reactor core axial power offset phenomenon changing along with fuel burnup and CRUD deposition in the operation process of the pressurized water reactor is described, and therefore the prediction precision of the reactor core axial power offset phenomenon is improved.

The invention provides a nuclear reactor comprehensive energy system in a data center scene and a control method. The nuclear reactor comprehensive energy system comprises a pressurized water reactor power generation device, an electrolytic cell hydrogen production device, a fuel cell power generation device and an absorption refrigeration device which are communicated with one another, when the power generation of the pressurized water reactor power generation device is sufficient, the pressurized water reactor power generation device is preferentially used for supplying power to the absorption type refrigeration device and supplying surplus power to the electrolytic cell hydrogen production device for hydrogen production; and when the pressurized water reactor power generation device cannot meet the power demand of the absorption type refrigeration device, the fuel cell power generation device supplies power to the absorption type refrigeration device and the electrolytic cell hydrogen production device. The pressurized water reactor is coupled with the wind-light and fuel cell power generation equipment, so that nuclear energy and wind-light resources are fully utilized, and the dual goals of safe and low-carbon efficient power supply are achieved by controlling the full load of the pressurized water reactor and the continuous and safe operation of the electrolytic cell.

This application relates to a fault handling method and apparatus for a nuclear power plant, and relates to the field of nuclear power technology. The method comprises obtaining detection signal sets corresponding to electrical partitions within a pressurized water reactor (PWR), determining target failure modes and temperature correlation values ​​for the PWR based on each detection signal set, and performing fault handling on the PWR based on the target failure modes and temperature correlation values. This method can improve the safety of nuclear power operations.

The invention discloses a primary circuit washing and foreign matter checking method for a pressurized water reactor nuclear power plant. The primary circuit washing and foreign matter checking method for the pressurized water reactor nuclear power plant comprises the following steps: S10, obtaining the size and morphology of a foreign matter; s20, according to the flowing direction of the operation medium of the loop system, the size of the foreign matter and the intercepting area of the passing equipment, the pipeline where the foreign matter possibly flows through and the position of the equipment where intercepting is possible are obtained; s30, calculating the critical flow velocity of the foreign matter in the primary loop medium; s40, quantitatively judging possible paths and closure points of the foreign matters; and S50, determining a flushing path and a flushing flow according to the possible path and the interception point of the foreign matter. According to the primary loop flushing and foreign matter checking method for the pressurized water reactor nuclear power plant, the foreign matter can be effectively checked, and the checking efficiency can be improved.

The utility model discloses an indirect heating type fuel rod simulation piece which comprises a cladding tube and an electric heating element, and the cladding tube is a corrosion product deposition test surface; the electric heating element is arranged in the cladding tube and is used for simulating heat generated by a nuclear fission reaction of the fuel pellet; high-purity helium is filled between the cladding tube and the electric heating element, and sealing of the fuel rod simulation piece is achieved through end plug welding at one end of the cladding tube and circumferential welding between the cladding tube and the electric heating element. According to the utility model, the actual structural characteristics and surface material properties of the fuel element are retained to the greatest extent while the surface thermal load state of the pressurized water reactor fuel element is simulated, so that the authenticity of corrosion product deposition test simulation and the reliability of obtaining test data are improved; and the method can also be used for experimental research on scale-caused accelerated corrosion after deposition of surface corrosion products.

The invention relates to an on-line setting method and system for a spring loading valve of a pressurized water reactor nuclear power plant, which can be used for finishing the adjustment of a spring preloading pressure value under the condition that the valve does not need to be disassembled, so that the time required for maintenance is greatly reduced. And secondly, compared with a traditional method for replacing the whole valve, the method avoids high overall purchase cost of the valve and complex installation engineering, and the maintenance cost is effectively saved. Meanwhile, manpower and material resource investment is reduced, and the overall maintenance cost is further reduced. Furthermore, due to the fact that complex operation such as cutting of connecting pipelines is not needed, potential safety hazards possibly occurring in the maintenance process are remarkably reduced, and therefore the safety of maintenance personnel and stable operation of the system are guaranteed.

The invention discloses a coating zirconium alloy fuel cladding tube corrosion test method, which comprises the following steps: providing a corrosion medium for simulating a pressurized water reactor primary loop water chemical environment, and arranging a coating-free contrast sample, the unit area corrosion rate of a zirconium alloy base material in a corrosion medium and the total mass change of a single-coating target sample in the corrosion medium are obtained by combining a weighing method, and microscopic parameters such as the thickness of a coating corrosion layer, the cross section content of key elements, spatial distribution and the like are measured so as to quantify the coating corrosion consumption; and deducting the base material corrosion mass change from the total mass change of the target sample according to the base material corrosion rate of the comparison sample, separating the coating corrosion mass change, and obtaining a coating corrosion evolution model in combination with the real coating corrosion rate calculated after microscopic parameter verification. The method breaks through the limitation that a traditional weighing method cannot split the independent corrosion behavior of the coating, and provides a basis for corrosion mechanism research, coating corrosion stage division, rate prediction and failure critical time judgment.

The invention discloses a pressurized water reactor primary loop main pump flow prediction and control method and system. The method comprises the steps that an original main pump flow signal of a pressurized water reactor primary loop is obtained; performing multi-scale decomposition on the main pump flow signal by using an HOFEM module of an NL-FCNN to obtain multi-size main pump flow; the multi-scale main pump flow component is input into an ANSA dynamic optimization network structure for feature extraction, and a multi-scale feature vector is obtained; on the basis of the extracted multi-scale feature vector, NL-FCNN is used for predicting a main pump flow signal, and an optimal main pump flow predicted value is obtained; based on a D-ADC algorithm, the optimal main pump flow predicted value is converted into an electric signal through a digital-to-analog conversion circuit, then the electric signal is input into a main pump frequency converter to drive a main pump to operate in a frequency conversion mode, the main pump flow of the pressurized water reactor primary loop is controlled, and the method and system can accurately control the main pump flow of the pressurized water reactor primary loop.

The invention discloses a pressurized water reactor steam generator liquid level control method and system, and belongs to the technical field of nuclear power station control. The method comprises the steps that a liquid level set value and a measured value of the steam generator are obtained, and deviation is calculated; inputting the set value, the measured value, the deviation and the bias constant into a preset BP neural network; the network dynamically outputs three coefficients of proportion, integral and differential through forward propagation to serve as real-time parameters of a PID (Proportion Integration Differentiation) controller; the PID controller calculates and outputs a water supply flow demand value, and a water supply valve is adjusted after conversion; and calculating a root-mean-square error of liquid level control as a loss function, and updating the weight of the neural network on line by using the new liquid level deviation through a back propagation algorithm to realize adaptive optimization of PID parameters. According to the invention, the self-learning ability of the BP neural network is combined with the reliability of PID control, the control quality and stability of the liquid level of the steam generator under various power levels are significantly improved, the system can be deployed in a power plant DCS through modular packaging, and the engineering practicability is high.

The invention belongs to the field of manufacturing of nuclear fuel elements for nuclear power stations, and particularly relates to a method for adjusting the rod position of a nuclear fuel assembly. Comprising the following conditions: 1, the condition that an upper end plug of a fuel rod penetrates through the eleventh layer of grillwork of the fuel assembly; 2, the lower end plug of the fuel rod does not penetrate through the first layer of grillwork of the fuel assembly; 3, the lower end plug of the fuel rod is close to the end face of the lower tube seat and cannot be adjusted by a pull claw. The nuclear fuel assembly rod position adjusting tool and method have the beneficial effects that the nuclear fuel assembly rod position adjusting tool and method are applied to manufacturing of pressurized water reactor nuclear fuel assemblies. According to the method, the position of the fuel rod in the rod pulling process can be effectively adjusted, and the technical requirements of nuclear fuel assembly products are met. The adjusted rod position of the nuclear fuel assembly meets the expectation.

The invention provides a rapid power reduction method and system based on an adjusting rod group and a pressurized water reactor, and relates to the technical field of nuclear reactors. According to the method and the device, the target rod group for rapid power reduction is selected from the grey rod group and the at least part of the black rod group in the adjusting rod group according to the target sequence, and the target sequence is the rod falling sequence determined based on the insertion sequence of the grey rod group and the at least part of the black rod group, so that rapid power reduction based on the adjusting rod group is realized; and the disturbance on the reactor core is small after the rod is dropped, and the influence on the shutdown capacity of a unit and the power distribution of the reactor core is reduced.

A kind of pressurized water reactor fuel assembly bending calculation method, by pressurized water reactor assembly calculation program, the fuel assembly of different water gap size is calculated, obtains the few-group constant of each type fuel assembly different water gap size;The few-group constant of each type fuel assembly is functioned, and the few-group constant library of fuel assembly of different water gap size is obtained by merging;Based on the measured data of pressurized water reactor fuel assembly bending, obtain the three-dimensional water gap distribution data of reactor core;Based on the three-dimensional water gap distribution data of reactor core, update each square block of reactor core to rectangular block, obtain the few-group constant of rectangular block by linear interpolation to water gap;Based on the conservation of interfacial neutron flux density and the conservation of interfacial neutron flux density of rectangular block, the staggered grid of rectangular block caused by the difference of water gap size is considered in reactor core calculation;The calculation of key physical quantity of reactor core is completed by pressurized water reactor core calculation program.The present application can provide accurate calculation method for pressurized water reactor fuel assembly bending, and fully guarantee the safety of reactor operation.

The utility model belongs to the technical field of fuel elements for nuclear reactors, and discloses a guide pipe, a guide pipe assembly and a pressurized water reactor fuel assembly. The guide pipe comprises a pipe body and two water hole assemblies, the pipe body is oppositely provided with a first end and a second end in the axial direction of the pipe body, the two water hole assemblies are arranged at the position, close to the first end, of the pipe body and the position, close to the second end, of the pipe body respectively, and the two water hole assemblies are symmetrically arranged in the radial direction of the center of the pipe body; the water hole assemblies are arranged at the two ends of the pipe body in the axial direction of the pipe body and used for discharging water flow during rod falling, that is, the pipe body has a mistake-proof function, so that when the guide pipe is installed, whether the guide pipe and the end plug are installed reversely or not does not need to be considered, the problems that the guide pipe assembly, a framework and a pressurized water reactor fuel assembly are scrapped and the like are avoided, and the production efficiency is improved. The installation reliability of the guide pipe is improved, and the safety of the pressurized water reactor fuel assembly loaded with the guide pipe is further improved.

The invention relates to the technical field of nuclear power plant engineering, and discloses a WANO chemical index calculation method and system for a pressurized water reactor nuclear power plant, and the method comprises the steps: obtaining the chemical parameters and operation state of a nuclear power plant; determining WANO chemical index calculation logic, and calculating WANO chemical indexes according to the chemical parameters; and triggering an abnormal alarm according to the WANO chemical index. The method is suitable for different reactor types and working conditions, has wide applicability, solves the problem of false alarm or missing alarm in a fixed limit value mode, provides accurate operation guidance for operators, reduces the operation risk, improves the operation and maintenance efficiency, fills the blank of insufficient combination of dynamic adjustment and working conditions in the prior art, and improves the operation and maintenance efficiency. And the method has remarkable innovativeness and practical engineering value.

The invention provides a research method for solving a sub-channel equation based on a physical information neural network, and belongs to the technical field of artificial intelligence. The method comprises the following steps: firstly, determining a to-be-calculated physical problem, collecting related data, screening the data, and carrying out data preprocessing; determining a control equation, boundary conditions and initial conditions of the system according to the system needing to be calculated; selecting a proper adaptation technology based on the data features and the equation features; designing a neural network architecture for solving, and determining input and output of the network and a network structure; according to the selected adaptation technology and problem features, defining adjustment variables in corresponding forms; designing a composite loss function exclusive to the model; training and optimizing the model to ensure that the precision of the model reaches the highest; and outputting a result and completing multi-scene verification solution. According to the method, in a pressurized water reactor sub-channel single-phase liquid coolant one-dimensional transient scene, a flow field can be solved with high precision and key parameters can be synchronously inverted based on a small amount of data, and meanwhile, the method can adapt to sub-channels with different geometrical characteristics and carry out friction parameter inversion in a targeted manner.

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 invention discloses a method and a system for predicting the breakage rate of a pressurized water reactor nuclear fuel rod, which are particularly suitable for a pressurized water reactor AFA3G nuclear fuel rod and are used for predicting the breakage rate of the pressurized water reactor AFA3G nuclear fuel rod by establishing a data-driven nuclear fuel rod fretting wear model and combining random uncertainty analysis. The method specifically comprises the following steps: firstly, constructing a flow-induced vibration-fretting wear coupling analysis finite element model of the nuclear fuel rod, and calculating the wear depth of the nuclear fuel rod along with time change; fitting a nuclear fuel rod wear depth curve according to the operation condition of the nuclear fuel rod by combining the geometric dimension and the material attribute of the nuclear fuel rod, and establishing a data-driven nuclear fuel rod wear mathematical model; further, establishing an uncertainty analysis model of the nuclear fuel rod, and calculating the failure probability of the nuclear fuel rod by using Monte Carlo simulation; according to the method, finite element analysis and uncertainty analysis are combined, so that the damage rate of the pressurized water reactor AFA3G nuclear fuel rod is predicted.

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 invention discloses a neutron source assembly which comprises a neutron source rod body, a lower end plug, a composite upper end plug and a guide cylinder, and the neutron source rod body comprises a neutron source rod end plug, a neutron source rod cladding tube, an upper reflecting layer, a lower reflecting layer, a source box and a matched end plug. The neutron source assembly adopts the rod-in-rod design, the possibility of abrasion of the neutron source rod body in the transportation process and the like can be reduced, the neutron source leakage risk is reduced, the counterweight design of the composite upper end plug can counteract the buoyancy of a coolant in a reactor, and it is ensured that the neutron source assembly smoothly falls to the bottom and keeps stable; the neutron source rod body can be used in a pressurized water reactor and can also be used in a liquid heavy metal reactor. Meanwhile, the sealing packaging structure of the neutron source rod body can reduce the risk of neutron source leakage caused by abrasion in the transportation and in-reactor operation process, the arrangement of the guide cylinder provides guidance for the assembly of the neutron source rod body, and the probability of collision damage of the neutron source rod body in the assembly process is reduced.

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.

The present invention discloses a method, system and medium for adjusting the power setting value of a reactor control system; relates to the technical field of reactor power control; performs online fault detection on an actuator of the reactor control system; when a fault in the actuator is detected, determines a fault symptom variable and an operating power setting value; pre-processes the fault symptom variable to obtain an input variable, inputs the input variable into a constructed fuzzy tuning model to calculate an operating power compensation value; adjusts the power setting value of the reactor control system by combining the operating power setting value and the operating power compensation value; for a multi-variable, strongly coupled nonlinear small pressurized water reactor system, when a feedwater valve fault and a control rod drive mechanism jam occur, a fuzzy tuning model with strong uncertainty processing capability is used to establish a relationship between the operating power compensation value and the fault symptom variable, automatically adjusts the reactor power setting value, and realizes fault-tolerant control of the reactor control system.

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 invention relates to the field of nuclear reactors, in particular to a lower chamber structure suitable for a pressurized water reactor. The lower chamber structure is characterized in that a reactor core supporting plate is welded at the lower end of the hanging basket barrel, and a water flowing through hole is formed in the reactor core supporting plate; a plurality of radial supporting key grooves are uniformly formed in the lower surface of the reactor core supporting plate in the circumferential direction; a plurality of radial supporting keys are arranged on the RPV lower sealing head, the positions of the radial supporting key grooves correspond to the positions of the radial supporting keys, and the radial supporting keys extend into the corresponding radial supporting key grooves in the reactor core supporting plate; supporting tables are arranged on the two sides of the radial supporting key; a flow distribution cover is hung on the supporting table; drain holes are formed in the surface of the flow distribution cover; and a vortex suppression assembly is arranged in the flow distribution cover and is connected with the reactor core supporting plate. Through the integral flow distribution cover structure, the flow distribution structure is simplified; the reactor core supporting structure is simplified through the reactor core supporting plate, and meanwhile, the secondary supporting structure is simplified through the radial supporting keys on the RPV. And generation of unstable eddy current is inhibited through the eddy current inhibition assembly.

Method for renovating a nuclear power plant initially comprising at least one light water nuclear reactor (LWR), in particular a pressurized water reactor (PWR) or boiling water reactor (BWR), replaced by at least one integrated modular nuclear reactor (SMR). The invention essentially consists of a method for retrofitting a nuclear power plant which consists of removing and removing all the components of the primary circuit with the exception of the LWR reactor vessel, which is emptied of all material and neutralized, then replacing in place of a part of these components with sub-assemblies each consisting of an integrated SMR reactor and a mixed concrete / metal structure which acts both as a reactor vessel for the SMR reactor, advantageously filled with water, as an anchor for the SMR inside the reactor building and advantageously as a contribution to the third containment barrier, while minimally modifying the infrastructure of the reactor building. Figure for abstract: Fig. 14D

The present invention discloses a heat export system and method for a pressurized water reactor nuclear power plant. The heat export system for the pressurized water reactor nuclear power plant includes a cooling pool, a water storage tank, a feed water pipeline, and a steam pipeline; the water storage tank is arranged in the cooling pool; the feed water pipeline is connected between the outlet of the water storage tank and the auxiliary feed water pipeline of the steam generator; the steam pipeline is connected between the inlet of the water storage tank and the discharge pipeline of the steam generator; the steam generator is communicated with the outlet of the water storage tank through the auxiliary feed water pipeline and the feed water pipeline; the steam generator is communicated with the inlet of the water storage tank through the discharge pipeline and the steam pipeline, and releases the internally generated steam and sends it into the water storage tank. The present invention is used to export the heat of the steam generator as an auxiliary means when other heat export means fail under severe accidents in the nuclear power plant, and can effectively cool the secondary side of the steam generator and complete the heat export of the reactor core to ensure the integrity of the reactor core.

The present application relates to the field of nuclear power unit accident analysis, and particularly relates to a coupling method of VVER unit system program and neutron dynamics program. The method is as follows: a reactor system program is compiled into a dynamic link library by adding an interface function in a three-dimensional neutron dynamics program of a reactor core, and the dynamic link library is accessed and called by the reactor system program in a plug-in mode; a neutron physics model of a specific loading scheme of a VVER type reactor is established; a simulation model of an accident analysis system of the VVER type pressurized water reactor is established; control logic of each main system and device of the primary loop and the secondary loop of the VVER type reactor is established; control logic and conservative condition settings for specific accidents are established; coupling calculation of the reactor system program and the three-dimensional neutron dynamics of the reactor core is performed; and key safety parameters of the VVER type reactor in specific accidents are obtained based on the results of the coupling calculation. The present application can simulate the changes of physical thermal parameters and the action responses of main systems in the primary loop and the secondary loop of the reactor after accidents, and is convenient for efficient and reliable accident analysis of the VVER type nuclear power unit.