83 results about "Reactor safety" patented technology

To suppress horizontal vibration of a nuclear pressure vessel during an earthquake, and make a dynamic vibration absorber compact to suppress an installation space of the dynamic vibration absorber, in a nuclear power plant.SOLUTION: An earthquake resistance support device for a nuclear power plant is arranged in a reactor containment vessel, and suppresses horizontal vibration of a reactor pressure vessel supported by a pedestal when an earthquake occurs. The earthquake resistance support device comprises a horizontal beam that is arranged between the reactor containment vessel and the reactor pressure vessel, and the pedestal, and extends in the radial direction of the reactor pressure vessel and the pedestal; and a dynamic vibration absorber installed on the horizontal beam. The dynamic vibration absorber comprises a hollow weight extrapolated to the horizontal beam and movable in an axial direction of the horizontal beam, and a coil spring device that elastically supports the hollow weight in the axial direction of the horizontal beam, to allow the hollow weight to vibrate in the axial direction of the horizontal beam.SELECTED DRAWING: Figure 3

The embodiment of the invention relates to the field of processes specially used for manufacturing a reactor or parts thereof, in particular to an assembling and charging testing method of a reactor using bulk nuclear fuel and the reactor comprising the bulk nuclear fuel. According to the method, loading test is carried out on the reactor by using the test facility after assembly, so that the backup reactivity of the reactor core can be determined, safe assembly and safe criticality of the reactor are realized, and the safety of experimenters is ensured; by gradually adding the partitioned side reflecting layers outside the reactor core formed by the bulk nuclear fuel, the reactor gradually reaches the critical value, the operation risk is reduced, and meanwhile, the critical point can be accurately obtained; after the critical loading of the reactor core is determined, the partitioned side reflecting layers are replaced by the complete side reflecting layers with the same value, so that the influence of gaps existing between the partitioned side reflecting layers can be avoided, and the reaction efficiency of nuclear fuel is improved. The reactor assembled according to the method has high neutron economy and low critical risk.

The invention relates to the technical field of lead-bismuth reactors, and particularly discloses a liquid lead-bismuth alloy thermal stripe oscillation experiment device and method. The device comprises an experiment loop, a gas protection loop, a data acquisition device and an integrated control module. According to the experimental loop, a magnetic pump is used for driving lead-bismuth fluid and an air cooler to realize cooling through a cold and hot double-loop structure, lead-bismuth jet flow with temperature difference is formed, and the lead-bismuth jet flow is mixed in a three-inlet experimental section to simulate a hot stripe oscillation phenomenon. The method comprises the steps of gas replacement, preheating, loop filling, temperature and flow rate regulation and control, data acquisition and the like. The device can truly simulate a lead bismuth alloy thermal stripe oscillation process, effectively prevents high-temperature oxidation, has a modular experimental section structure, can research thermal stress responses under different structures, and provides reliable experimental data for reactor safety design.

Device and Method for Maintaining a Pressure Reactor Containment The present invention relates to a maintenance device (10) for a pressure reactor containment, comprising a connecting fitting (120) on a wall (20) of the containment, said connecting fitting (120) extending in a longitudinal direction (X) between an internal base (55) and an external base (60) and comprising: a) an insertion portion (65) intended to be positioned in a receiving recess (90) machined in the wall (20) of the containment, b) an external portion (70) intended to be positioned outside (30) of the containment and comprising a support element (105) intended to be placed opposite the wall (20) of the containment on a periphery (P) of the receiving recess (90), a fluid conduit (75) extending through the connecting fitting (120), a groove (88) of reception of a compression seal (100) being provided in the insertion portion,the maintenance device (10) comprising means for fixing said connecting branch (120) to the wall (20). Figure for the abbreviation: 4,

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 invention relates to the technical field of reactor safety analysis, in particular to a test device for simulating steam-water two-phase jet flow impact under a cold loss accident, which comprises a blow-off container provided with a test cavity, the top of the test cavity is provided with a blow-off nozzle, and the blow-off nozzle is arranged downwards and is used for connecting an upstream stagnation system and a simulation crevasse; the mounting assembly is mounted in the test cavity, is positioned below the spraying nozzle and is used for mounting a test piece; and the trash holding assembly is mounted in the test cavity, is positioned below the mounting assembly, and can collect fragments impacted by the steam-water two-phase jet flow from the test piece. The testing device for simulating steam-water two-phase jet flow impact under the cold loss accident can carry out crevasse impact tests on materials such as heat preservation, plugging, fire prevention, painting and cables, and provides test data support in the aspect of disintegrating slag source items for upstream effect analysis of a pit filter.

A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs, each of the functionally independent modules comprising a digital module or a combination digital and analog module, an analog module electrically coupled to one or more of the functionally independent modules, and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

To provide a concrete structure capable of preventing foreign matter from entering a steam exhaust pipe during concrete placing as well as after completion, and reactor containment vessel using the concrete structure.SOLUTION: A concrete structure comprises a concrete skeleton 2, a steel plate (an inner steel plate) 3 stretched on a surface of the concrete skeleton 2 via studs 8, and a steam exhaust pipe 10 buried inside the concrete skeleton 2 for exhausting steam generated from the concrete skeleton 2 when the concrete skeleton 2 is exposed to high temperatures. The steam exhaust pipe 10 includes one end disposed at a boundary part between the concrete skeleton 2 and the steel plate 3, and the other end disposed outside the concrete skeleton 2, and the steam exhaust pipe 10 includes a steam introduction mechanism 11 at least at one end for introducing steam into the steam exhaust pipe 10.SELECTED DRAWING: Figure 1

The invention provides a method, system, equipment and medium for identifying the number and size of reactor fuel cladding damage, and the method comprises the steps: S1, obtaining the unit crevasse release amount reference data of each nuclide based on transient release modeling experiments of different groove sizes; s2, obtaining the total release amount of nuclides based on the obtained unit crevasse release amount reference data of each nuclide; and S3, constructing a matrix reversibility constraint, and realizing the unique solution inversion of the damage parameter. The method can improve the accuracy and efficiency of fuel cladding damage detection, provides reliable method support for safe operation of a reactor, and can be widely applied to monitoring and evaluation of the safety of the fuel assembly of the pressurized water reactor nuclear power station.

The invention discloses a natural circulation system based on in-pipe condensation and an operation method, and the system comprises a cooling water tank which is arranged outside a reactor containment and is filled with cooling water. And the heat exchanger is arranged in the cooling water tank. The inlet end of the air inlet pipeline is connected inside the containment, and the outlet end is connected with the heat exchanger. The inlet end of the drainage pipeline is connected with the heat exchanger, and the outlet end extends into the in-shell water tank. The inlet end of the exhaust pipeline is communicated with the heat exchanger, and the outlet end extends into the suppression pool. And the suppression pool is internally filled with water, is arranged outside the containment, has a pressure difference with the containment, and is communicated with the containment through a low-pressure-difference communication device and a pipeline. And the in-shell water tank is arranged in the safety shell. The inlet end of the gravity injection pipeline is communicated with the in-shell water tank, and the outlet end is connected with the reactor. The problem that an existing pipe external condensation type natural circulation system is not suitable for the low level difference condition is solved, the operation stability of the natural circulation system under the condition is improved, and the starting time of the natural circulation system is shortened.

The invention discloses an advanced safety injection system and a pressurized water reactor and belongs to the field of nuclear power. 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 divided by a conical surface wall with the axis vertically extending, the ACC cavity surrounds the outer side of the CMT cavity, the sectional area of the ACC cavity is gradually reduced from top to bottom, and an upper through hole and a lower through hole are formed in the conical surface wall so that the CMT cavity can be communicated with the ACC cavity. Pressure gas is stored at the top of the ACC cavity; a top inlet communicated with a reactor primary loop cooling system is formed in the top end of the CMT cavity, and an injection pipeline for injecting a coolant into the reactor core is arranged at the bottom end of the CMT cavity. The advanced safety injection system is simple in structure, the CMT and ACC functions can be achieved through the single water tank under different accident working conditions, the manufacturing and maintenance cost is effectively reduced, and the reactor safety is improved.

The invention relates to the technical field of reactor safety analysis, and provides a pressurized water reactor pressure vessel lower chamber flow distribution analysis method. The method comprises the following steps: establishing a fluid domain three-dimensional geometric model; performing unstructured grid division on the fluid domain three-dimensional geometric model to generate a three-dimensional calculation grid; selecting a turbulence model; setting inlet boundary conditions and physical property parameters of the coolant; the reactor core area is equivalent to a porous medium area, and flow resistance parameters of the porous medium area are set; based on the three-dimensional calculation grid, adopting a selected turbulence model, loading flow resistance parameters of a porous medium area and inlet boundary conditions and physical property parameters of a coolant, and carrying out three-dimensional flow simulation to obtain coolant flow distribution at the section of a reactor core inlet; and evaluating the flow distribution uniformity of the reactor core inlet according to a preset flow distribution evaluation criterion. The method is used for simulating three-dimensional flow of a coolant in a lower cavity of a reactor pressure vessel, acquiring a flow field structure and flow distribution data of the coolant, and performing quantitative evaluation.

The invention relates to the technical field of reactors, in particular to a method for calculating the deposition rate of a pressurized water reactor corrosion product. The method for calculating the deposition rate of the corrosion product of the pressurized water reactor comprises the following steps: acquiring concentration distribution and Zeta potential distribution of metal ions related to the corrosion product; acquiring an electric field force applied to the particulate corrosion product based on the Zeta potential distribution; a discrete phase model is adopted to track the transportation track of the particle-state corrosion product, and the force borne by particles in the transportation process comprises the electric field force; according to the discrete phase model, calculating to obtain a particle mass flow rate; and according to the particle mass flow rate, the deposition rate and the erosion rate of the particle-state corrosion product on the wall surface are calculated. The method can be used for relieving the problem that a current calculation model cannot accurately predict the CRUD deposition rate, and the authenticity of CRUD growth prediction and the credibility of reactor safety analysis can be improved.

An integrated PWR nuclear reactor that requires no refueling and is cooled by natural convection. It is housed within a reactor containment unit which is transportable to a placement site and placed within an underground building that allows heat transfer from the reactor to the surrounding soil. Tire reactor containment unit encloses an integrated PWR nuclear reactor, a pressure and chemical control system, a safety system and a turbo generator. After its operational life, the reactor's water and air are replaced with inert gases, circulating by natural convection to remove decay heat to the surrounding soil. This enables the integrated PWR reactor, initially designed to generate electricity, to act as a dry storage facility for uranium fuel elements after the end of its operational life.

The invention belongs to the field of nuclear materials, and particularly relates to a nuclear fuel and a preparation method thereof and a reactor, the nuclear fuel comprises a nitride matrix and an MAX phase material distributed in the nitride matrix; the chemical formula of the MAX phase material is Mn + 1AXn, M is selected from at least one of Ti and Cr, A is selected from at least one of Si and Al, X is selected from C, and n is 1 or 2. According to the nuclear fuel disclosed by the invention, the MAX phase material is introduced into the nitride matrix, and a corrosion-resistant oxide layer is generated in situ by utilizing the MAX phase material to block water permeation, so that the nuclear fuel has the properties of high thermal conductivity, relatively strong water corrosion resistance, steam oxidation resistance and radiation resistance, and can be applied to a light water reactor; the application risk of the existing uranium nitride fuel in the light-water reactor is reduced, and the safety and economy of the reactor are improved.

This invention discloses a method and apparatus for simulating heat pipe reactor loss-of-heat-sink accidents, including a heat pipe reactor simulation device, a DC power supply, and a secondary loop auxiliary system. The DC power supply is independently located outside the heat pipe reactor simulation device and electrically connected to it. The secondary loop auxiliary system is connected to the piping of the heat pipe reactor simulation device. This solution, based on "scaled modeling + precise modeling + safe substitution," constructs a heat pipe reactor accident simulation system that combines realism, safety, and economy. It avoids nuclear risks while accurately reproducing key scenarios of coolant loss accidents, solving the pain points of real reactor testing. It provides an efficient technical means for the safe design, optimization, and accident response of heat pipe reactors, and is of great significance for improving the safe operation level of nuclear reactors.

The invention is applicable to the field of nuclear power stations, and provides a pressure test system for a reactor containment, which comprises an internal transmission device, a regulating valve and exhaust equipment, the internal transmission device is arranged in the reactor containment, is connected with the regulating valve through a first pipeline and is used for transmitting gas in the reactor containment to the regulating valve; the adjusting valve is arranged outside the reactor containment, connected with the exhaust equipment through a second pipeline and used for guiding the gas to be transmitted to the exhaust equipment through the second pipeline by adjusting the opening degree of the valve to the set opening degree, and the designed flow range of the gas transmitted between the first pipeline and the second pipeline is 0-32,400 m < 3 > / h; and the exhaust equipment is used for exhausting gas. According to the pressure test system, the valve opening degree of the regulating valve is regulated to the set opening degree to guide gas to be quickly transmitted to the exhaust equipment, so that the pressure relief rate of the gas in the containment is greatly improved, the duration required by a CTT test is shortened, and the operation efficiency of a unit is improved.

The invention discloses a nuclear reactor containment vessel thermal hydraulic state prediction method based on machine learning and system program coupling, which comprises the following steps: 1, constructing a machine learning off-line model, including establishing a nuclear reactor typical thermal hydraulic phenomenon experimental data set, preprocessing data and training an automatic machine learning model; 2, initializing a coupling interface module, which mainly comprises initializing a Python environment, loading a dynamic link library of a system program, constructing a variable mapping scheme and initializing a memory; 3, coupling calculation: calling the dynamic link library to obtain the thermal-hydraulic state parameters at the previous moment, calling the machine learning model to calculate key physical model parameters based on the thermal-hydraulic state parameters, and feeding back the key physical model parameters to the dynamic link library to calculate the thermal-hydraulic state parameters at the current moment; 4, updating the current time; and 5, repeating the step 3 until the set time is reached. The machine learning-system program coupling framework provided by the invention can quickly and accurately predict the thermal hydraulic state of the nuclear reactor containment.

The invention relates to the field of reactor core physical calculation, in particular to a reactor core physical optimization calculation method under a nuclear power unit bias operation condition, and the method comprises the following steps: 1, constructing a geometric model and a reactor core grid model of a reactor in fluid dynamics software; 2, establishing a porous medium model in fluid dynamics software; 3, setting boundary conditions in the hydrodynamic software and carrying out thermal hydraulic calculation; 4, constructing a thermotechnical and physical coupling interface, taking a thermotechnical calculation result of the hydrodynamic software as input, and carrying out physical calculation in physical calculation software; 5, taking the three-dimensional power distribution of the physical calculation software as fluid dynamics software input, and carrying out thermal hydraulic calculation to obtain an optimized fuel assembly upper power peak factor and an optimized fuel assembly volume power peak factor; and 6, repeating the steps 3-5 until the power peak factor converges. According to the method, the physical parameters of the reactor core under the bias ring operation condition are calculated, and the control capability of safe operation of the reactor is further improved.

The invention relates to the technical field of reactors, in particular to a corrosion product deposition growth prediction method, a storage medium and electronic equipment. The corrosion product deposition growth prediction method comprises the following steps: establishing a three-dimensional model, and carrying out mesh generation on the three-dimensional model to obtain a mesh model; obtaining the thickness increment of the corrosion product; converting the thickness increment of the corrosion product on the grid unit surface into a node movement amount by using a node movement algorithm, and driving grid nodes to displace in each time step length according to the calculated node movement amount based on a dynamic grid technology; obtaining an updated grid model; the deposition rate and the erosion rate of the corrosion product are solved on the updated grid model, and the thickness increment of the updated corrosion product is obtained. The application of the invention can improve the safety of the reactor, and has important values for protecting the integrity of the fuel rod, preventing the leakage of radioactive substances and ensuring the safe operation of the reactor.

The utility model discloses a reactor containment and residual heat removal system, the reactor containment comprises: an inner shell, an outer shell, and a water curtain forming assembly for removing residual heat in the inner shell, the outer shell is sleeved on the periphery of the inner shell, the water curtain forming assembly surrounds and is fixedly installed on the outer wall of the inner shell, and the water curtain forming assembly is fixedly installed on the outer wall of the inner shell. A storage cavity for containing cooling water is formed between the inner shell and the water curtain forming assembly. The water curtain forming assembly comprises a partition piece and a drainage piece, the drainage piece or the partition piece is fixedly installed on the inner shell, the partition piece is connected with the drainage piece, the top end of the partition piece is arranged to be horizontal, and the bottom end of the drainage piece makes contact with the outer wall of the inner shell. The manufacturing cost is saved, damage is not prone to occurring in the cooling process, the safety in the using process is guaranteed, the cooling effect is improved, the temperature and pressure reduction speed is high, heat conduction is more uniform, the heat dissipation capacity is better, the utilization rate of cooling water is increased, and the use amount of the cooling water is reduced.

The invention discloses a hot hybrid simulation method for a T-shaped pipe of a pressurized water reactor core safety injection system, and the method comprises the steps: building a simplified three-dimensional geometric model comprising a safety injection T-shaped pipe through CFD software, dividing grids through a hybrid grid technology, carrying out the local encryption of a core region, and generating a boundary layer grid, the method comprises the following steps: selecting a VOF multiphase flow model, a k-epsilon turbulence model and a Lee phase change mass transfer model for coupling simulation, setting corresponding boundary conditions and initial parameters, and finally, completing solution method setting through a Coupled algorithm discrete format and an interpolation method, performing adaptive time step strategy solution, accurately capturing key thermal hydraulic parameters such as a temperature field, a velocity field, flow pattern evolution and the like, and obtaining the thermal hydraulic parameters. And a mixing condensation process of a safety injection coolant and a main loop high-temperature coolant or steam under an accident condition is simulated. The method not only reduces the time consumption and cost of traditional experiments, but also improves the prediction precision of key phenomena in reactor safety analysis, provides reliable technical support for emergency cooling strategy optimization of a pressurized water reactor nuclear power plant under a water loss accident or a steam pipeline fracture accident, and has important engineering application value.

The application discloses a nuclear reactor core sub-channel analysis method based on a two-fluid model, comprising the following steps: 1, establishing a fluid domain calculation grid according to the structure of a nuclear reactor core or an assembly; 2, establishing a control equation set and a fuel rod heat conduction model for continuous gas-liquid phases and discrete droplet phases respectively; 3, considering flow heat exchange, resistance pressure drop and turbulent flow mixing and other physical phenomena in the core, implanting corresponding constitutive models; 4, discretizing the control equation based on a second-order discrete format to obtain a discrete equation set; 5, dividing parallel tasks and parallel regions through an MPI parallel strategy to form a large coefficient matrix of a pressure equation; and 6, solving the discrete equation set and the large coefficient matrix through a pressure coupling semi-implicit algorithm and a stable double-conjugate gradient method. Through the method, key physical parameter distribution of a pressurized water reactor can be more accurately and reasonably calculated, thereby providing technical guidance for reactor safety analysis and engineering design.

The invention relates to the field of ultrasonic temperature field tomography, in particular to a dynamic ultrasonic tomography measurement method for a nuclear reactor primary loop coolant temperature field. The method comprises the following steps: A, discretizing and collecting dynamic temperature field information; and B, determining a transfer path and corresponding flight time of a sound wave at a non-self-measurement moment by using spatial correlation. The invention discloses a dynamic ultrasonic chromatography measurement method for a nuclear reactor primary loop coolant temperature field. The dynamic change of the temperature field is fitted by using the strong dependency of the temperature field on the space and time relationship and considering the dynamic evolution information on the time and space dimensions, the temperature field at each moment in the time frame is reconstructed, and the dynamic characteristics of the reactor primary loop rapid change temperature field are accurately captured. The timeliness and accuracy of coolant three-dimensional dynamic temperature field measurement are improved, high-precision rapid calculation of coolant temperature parameters is further realized, and more powerful guarantee is provided for safe and reliable operation of a reactor.

The application relates to a heavy water reactor nuclear power plant single fuel channel dewatering and water filling operation method, which comprises the following steps: step 1, emptying the nuclear fuel of the fuel channel; step 2, installing a fuel channel cutoff tool to realize cutoff; step 3, implementing ice plug isolation to carry out fuel channel dewatering operation; step 4, installing a temporary shielding plug to establish a maintenance condition; step 5, fuel channel water filling operation; and step 6, nuclear fuel reloading of the fuel channel. The application establishes necessary conditions for single fuel channel fault treatment, and after the fault treatment is completed, the fuel channel is water filled, the nuclear fuel is reloaded and the like, so that the standard configuration of the single fuel channel is restored; thus, the maintenance can be safely and smoothly carried out, the operation risk is reduced, the work efficiency is improved, and the personnel radiation safety and the reactor safety are ensured.

The embodiment of the invention provides a critical heat flux density prediction method of an offshore floating nuclear power station and a related device, and relates to the technical field of reactor safety analysis. The method comprises the steps that multi-dimensional operation parameters of the offshore floating nuclear power station are obtained, wherein the multi-dimensional operation parameters comprise thermal parameters, ocean condition parameters, working medium type parameters, working condition parameters and heat exchanger morphological parameters; inputting the multi-dimensional operation parameters into a pre-trained critical heat flux density prediction model; and outputting a corresponding critical heat flux prediction value through the critical heat flux prediction model. Specific to the critical heat flux density characteristics of the marine floating nuclear power station, the method comprehensively considers the coupling influence of multiple factors such as thermal parameters, marine condition parameters, working medium type parameters, working condition parameters and heat exchanger morphological parameters by pre-training a critical heat flux density prediction model; accurate prediction of the critical heat flux density of the offshore floating nuclear power station is realized, and data support is provided for safe operation of the offshore floating nuclear power station.

A method for analyzing the bubbling behavior of high-fuel-consumption fuel under a transient high-temperature condition relates to the technical field of reactor safety analysis, and comprises the following steps: establishing a three-dimensional model; subdividing the three-dimensional model to obtain a computational domain, and performing physical property assignment on each unit grid of the computational domain; an irradiation test process is simulated, and the fission rate of fuel of each unit grid is calculated; within the irradiation time, the fuel consumption is calculated and updated, and irradiation swelling and creep behaviors of the fuel core are calculated; calculating the total amount of fission gas released by the fuel core body; adding a virtual bubbling core in the three-dimensional model; based on the volume expansion increment of the fuel core body and the true stress of the cladding, simulating and calculating the swelling behavior of the fuel core body, the creep behavior of the cladding and the plastic deformation process of the cladding in the gradual heating process of the fuel by setting external heating boundary conditions; the method is used for solving the problem that a traditional method cannot accurately analyze bubbling changes in a fuel assembly.

This invention relates to the field of reactor technology, specifically to a hybrid energy spectrum supercritical water reactor. The hybrid energy spectrum supercritical water reactor includes a shell, thermal fuel assemblies, and fast fuel assemblies. The thermal region is vertically arranged with alternating thermal fission and regeneration zones; the fast region is vertically arranged with alternating fast fission and regeneration zones. The thermal fuel assemblies are housed in the thermal region, and the fast fuel assemblies are housed in the fast region, all with the same fuel arrangement. In this hybrid energy spectrum supercritical water reactor, no separate moderator channel is required to obtain the required neutron energy spectrum, simplifying the core structure while ensuring safe and efficient reactor operation.

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.

The embodiment of the application relates to the field of reactor fuel elements and assembly thereof, in particular to a method for loading a reactor fuel assembly and a method for loading a lead-bismuth reactor. The method for loading the fuel assembly comprises the following steps: S10, reactor body installation is completed, a neutron source assembly and a simulation assembly are loaded into predetermined positions, a fuel assembly is transported to a loading site, a reactivity control system, nuclear measuring instruments and loading special tools are debugged, and experimental personnel are prepared; S20, a control rod is inserted into a reactor core, a reading of the nuclear measuring instruments is taken and recorded; S30, the loading special tools are operated, the simulation assembly in the predetermined position is taken out and placed into the fuel assembly; S40, the step S30 is repeated until loading of a predetermined number of fuel assemblies is completed, a reading of the nuclear measuring instruments is taken and recorded; and S50, the steps S30-S40 are repeated to complete loading of all fuel assemblies. The method for loading the fuel assembly of the reactor provided by the embodiment of the application is beneficial to ensuring that the reactor is safely loaded with the fuel assembly.