116results about "Plant parameters regulation" patented technology

The present application relates to a method and system for filling water in a primary loop after overhaul charging of a nuclear power unit, comprising: performing preparation work before filling water; after the preparation work is completed, performing switch operation test of a to-be-tested actuator and water filling and exhaust work of a flowmeter; after the switch operation test of the to-be-tested actuator is completed, performing performance test of a high-pressure safety injection pump; after the performance test of the high-pressure safety injection pump is completed, performing performance test of a low-pressure safety injection pump; after the performance test of the low-pressure safety injection pump is completed, filling water in the primary loop according to a preset water filling mode; judging whether a real-time water level of a reactor pool reaches a full water level; if yes, performing alarm action test. The present application can solve the problem of long time consumption, reduce workload, significantly improve work efficiency, reduce work risk and effectively save production cost by completing the preparation work in advance and arranging the alarm action test at the end.

The invention discloses a power coordination control method of a high-temperature gas cooled reactor nuclear steam supply system, which relates to the technical field of nuclear power, and comprises the following steps of: coupling multiple modules in the high-temperature gas cooled reactor nuclear steam supply system through a common conventional island; the key state parameters of the to-be-adjusted module are used as main control variables, and the other modules are used as auxiliary control variables, so that the mutual influence relationship among the modules can be comprehensively considered; reasoning the control quantities based on a preset fuzzy rule to determine a fuzzy control strategy, so that adverse effects among modules can be counteracted, overall parameters are kept in a design value range, and safe and stable operation of the system is guaranteed; flexible adjustment can be performed according to key state parameters of different modules by using a fuzzy control strategy, the power of the reactor is quickly and accurately controlled, the dynamic response capability of the system is improved, and the requirement that the power of the reactor is quickly adjusted along with the load requirement of the steam turbine generator unit is met.

Techniques for detecting, forecasting, and reducing or avoiding disruptions in fusion devices are described. Data corresponding to operation of a fusion device may be received. Based on the data and one or more physics-based models, an occurrence of an event and a criticality level of the event may be determined. Based on criticality level of the event, whether a disruption of plasm will occur may be determined. An indication identifying the event and whether the disruption of plasma in the fusion device will occur may be output to a computing device associated with a user. In some embodiments, based, at least in part, on the criticality level of the first event, a control system associated with the fusion device may be instructed to perform one or more actions.

The invention discloses a reactor coolant system flow monitoring device and a signal control method, and belongs to the technical field of reactor coolant systems.The reactor coolant system comprises at least three main loops, a first pressure tapping hole located in the outer wall is formed in a main pipeline transition section elbow of each main loop, and a second pressure tapping hole located in the outer wall is formed in the main pipeline transition section elbow of each main loop; the at least four second pressure tapping holes are formed in the inner wall; the flow monitoring device comprises at least four first differential pressure sensors and at least two second differential pressure sensors; the positive pressure ends of the at least four first differential pressure sensors are communicated with the first pressure tapping hole, and the negative pressure ends of the at least four first differential pressure sensors are communicated with the at least four second pressure tapping holes in a one-to-one correspondence mode. The at least two second differential pressure sensors are arranged in parallel, and the two ends of the at least two second differential pressure sensors communicate with an inlet pipeline and an outlet pipeline of the main pump on the main loop correspondingly.

The invention relates to the technical field of compressed air system intelligent control, and discloses an intelligent control and air supply distribution method and system for a nuclear facility compressed air system pipe network. Carrying out numerical simulation on the distributed arrangement and operation conditions of different air compression stations and the flow and pressure characteristics of the compressed air pipe network under the intermittent load of different process equipment; on the basis of data generated by numerical simulation, the pipe network tail end pressure of the compressed air pipe network system is adjusted through an intelligent fuzzy self-tuning PID control method, and the operation state of an air compressor unit of the compressed air pipe network system is optimized on the basis of a flow matching strategy; pipe network operation parameters of the compressed air pipe network system are collected in real time, and the control strategy is fed back and adjusted. The whole compressed air system can operate efficiently and stably, the requirements of all links of nuclear facilities for compressed air are guaranteed, meanwhile, the operation intelligence and reliability of the system are improved, and adverse effects caused by unstable pressure and the like are reduced.

A nuclear reactor system for use with an electric power grid is disclosed, the nuclear reactor system including a nuclear reactor, an energy storage system coupled to the nuclear reactor, and control circuitry coupled to the nuclear reactor and the energy storage system, wherein the control circuitry is configured to monitor power demand on the electric power grid, monitor power output generated from the nuclear reactor, detect changes in power demand, cause the energy storage system to temporarily compensate for the changes in power demand, and adjust the power output based on the changes in power demand.

To provide a nuclear power plant and a method for generating power in a nuclear power plant which can improve the load following property of a nuclear power generation plant and can further stabilize the output of a nuclear reactor.SOLUTION: The nuclear power plant having a nuclear reactor includes: an output control unit 70 for controlling the opening degree of an air extraction valve 13 according to a load request; and an energy storage unit for storing renewable energy. The output control unit 70 controls the temperature of water to supply to the nuclear reactor on the basis of the opening degree of the air extraction valve 13 and the amount of stored energy in the energy storage unit.SELECTED DRAWING: Figure 1

To adjust a core flow rate higher than that in a stationary state when a power change operation of a natural circulation reactor is conducted by a control rod, thus improving a power change speed while maintaining a stable condition.SOLUTION: A nuclear power plant includes a large-diameter space chimney at an upper part of a reactor core within a shroud of a reactor pressure vessel. The nuclear power plant includes: a coolant flowmeter configured to measure a coolant flow flowing into the reactor core; a water level gauge configured to measure a water level of a coolant within the reactor pressure vessel; a water-supply flowmeter connected to the reactor pressure vessel and configured to measure a water-supply flow rate; a monitoring device configured to monitor the coolant flow, the water level and the water-supply flow rate; and a controlling device configured to control a control rod and the water-supply flow rate. The controlling device is configured to: calculate, in advance, a fluctuation margin of a core flow rate by a time lag of the water-supply flow rate; change a target value of the water level to be higher than a fluctuation margin of the water level corresponding to the fluctuation margin of the core flow rate so as to change the water-supply flow rate and then change the water level; and thereafter operate and control a position change of the control rod.SELECTED DRAWING: Figure 4

This invention provides a method for controlling the oxygen content in reactor coolant, comprising: determining the relationship between the growth thickness of the oxide layer formed on the surface of the fuel rod cladding within a preset time period under conditions of no coolant corrosion and the oxygen content in the coolant; determining the removal rate of the oxide layer on the surface of the fuel rod cladding after reaction with the coolant; determining a functional relationship between the remaining oxide layer thickness on the cladding surface and the oxygen content of the coolant under varying oxygen content conditions, based on the relationship between the oxide layer growth thickness and the oxygen content in the coolant and the oxide layer removal rate; using the functional relationship between the remaining oxide layer thickness on the cladding surface and the oxygen content of the coolant as the objective function, and determining the optimal oxygen content control strategy of the coolant corresponding to the minimum value of the remaining oxide layer thickness based on a preset optimization algorithm. This invention alleviates the oxidation and corrosion problem of fuel rod cladding materials in metallic coolants by determining the optimal oxygen content control strategy of the coolant.

The application discloses a method for determining feedwater flow of an evaporator of a nuclear power plant and related devices, and relates to the technical field of thermoelectric conversion. The method comprises the following steps: obtaining steam flow at an outlet of the evaporator, secondary loop pressure of the evaporator, feedwater temperature at an inlet of a main feedwater pipe, and nuclear power; using the secondary loop pressure of the evaporator and the feedwater temperature to determine a specific enthalpy coefficient by table lookup, the specific enthalpy coefficient being a ratio of specific enthalpy of steam to specific enthalpy of feedwater; converting the nuclear power into theoretical steam flow, and verifying the steam flow at the outlet of the evaporator by using the theoretical steam flow, if the verification is passed, using the steam flow at the outlet of the evaporator as one of inputs of a feedwater flow soft measurement model; inputting the specific enthalpy coefficient, the steam flow at the outlet of the evaporator and the secondary loop pressure of the evaporator into the feedwater flow soft measurement model to obtain a theoretical feedwater flow. The application realizes high-precision dynamic measurement and calculation of the feedwater flow of the evaporator based on the feedwater flow soft measurement model.

A hybrid power generation system which includes a nuclear facility comprising a nuclear reactor and an exclusion zone. A thermal energy storage, a nuclear steam supply system located, and a solar energy collection system are all located within the exclusion zone. The thermal energy storage vessel contains a thermal mass composition operable to store thermal energy. The nuclear steam supply system has a nuclear reactor and a working fluid loop. The working fluid loop is configured to circulate a working fluid from a steam generator through the thermal energy storage vessel to absorb thermal energy and heat the working fluid for introduction to an electricity generating system. The solar energy collection system includes a heat transfer loop heated via a solar collector. The heat transfer loop is configured to circulate a heated heat transfer fluid to add thermal energy to the thermal mass composition in the thermal energy storage vessel.

The present invention relates to a system and a method for automatically setting a main steam condensate drain valve closing setpoint of a nuclear power plant, and, specifically, to a system and a method for automatically setting a main steam condensate drain valve closing setpoint of a nuclear power plant, which apply different setpoints to respective condensate drain valves since the pressure of main steam and a condenser of a nuclear power plant varies according to operating conditions such as turbine output, and the discharge speeds of the condensate vary according to the different installation positions of condensate drain valves.

The application relates to the technical field of periodic testing, and discloses a safety injection logic test optimization method, device, equipment and medium for a nuclear power station. The method comprises the following steps: in the process of performing a safety injection logic test on the nuclear power station, if it is detected that a fault scene simulation of the nuclear power station is ended, a small-flow electric valve on a small-flow pipeline of an RCV pump of the nuclear power station is controlled to be started, and after the small-flow electric valve is started, an adjustable isolation valve is controlled to be adjusted from a closed state to a fully open state within a set opening time range, so that the adjustable isolation valve is used for slowly injecting reactor coolant into a primary pump No. 1 sealing leakage flow pipeline of the nuclear power station through the small-flow pipeline of the RCV pump, wherein the adjustable isolation valve is installed on the small-flow pipeline of the RCV pump and located upstream of the small-flow electric valve. The method can improve the safety of nuclear power station equipment operation.

The invention relates to the technical field of reactor power flow control, in particular to an automatic rapid power and flow reduction control method for a sodium-cooled fast reactor, which comprises the following steps: step 1, when the reactor power is greater than 80% of rated power, setting a target set value of the reactor power; 2, a reactor power adjusting system is adjusted to be in a manual mode, and a first loop flow control system and a second loop flow control system are adjusted to be in an automatic mode; 3, the rod control and rod position indication system controls the first compensation rod to execute descending action; 4, when the reactor power is smaller than or equal to 80% of rated power, the rod control and rod position indicating system controls the first compensation rod to stop descending, and the reactor power adjusting system adjusts the reactor power to a power target set value. According to the invention, the effect of rapidly reducing the nuclear power and the flow of the primary loop and the secondary loop to 75% of the rated power level is realized through a non-rod-falling type automatic rapid power and flow reduction technical means.

The invention is suitable for the technical field of system control, and provides a control method for the rotating speed of a main water supply pump of a nuclear power plant and a steam generator system.The method comprises the steps that when it is detected that a main water supply valve breaks down, the first rotating speed of the main water supply pump is determined according to the first output power of the steam generator system at the current moment; correcting the first rotating speed according to the numerical value of the process parameter of the steam generator system to obtain a corrected second rotating speed; and the rotating speed of the main water supply pump is adjusted to be the second rotating speed. According to the method, under the condition that the main water supply valve of the steam generator breaks down, the rotating speed of the main water supply pump can be timely and accurately adjusted, and the loss of the nuclear power station can be reduced.

The invention discloses a pressurized water reactor nuclear power unit power adjusting system and method, and belongs to the technical field of pressurized water reactor nuclear reactors, the system comprises a pressurized water reactor nuclear steam supply system and a bypass system arranged on a steam generator secondary side steam pipeline of the pressurized water reactor nuclear steam supply system; wherein the bypass system comprises a steam bypass subsystem, a composite material heat storage subsystem and an air power generation subsystem; the steam bypass subsystem is used for introducing excess steam in the pressurized water nuclear reactor nuclear steam supply system into the composite material heat storage subsystem for heat energy storage, and the air power generation subsystem absorbs the heat energy stored in the composite material heat storage subsystem through compressed air to drive a turbine to generate power. Under the condition that the operation power of the nuclear reactor is not changed, the steam heat energy of the bypass steam turbine is stored in a mode of sensible heat energy of the composite material, the problem of mismatching between power supply and power demand is solved, and power adjustment of the pressurized water reactor nuclear power unit is achieved.

A method and system for calibrating nuclear instrumentation of a reactor is disclosed. The system comprises a reactor configured to operate at a low power level and an external heater configured to heat the reactor to its operating temperature and maintain the reactor in a critical state. The reactor maintains its temperature at a steady state by repeating steps of reducing the heater's power and increasing the reactor's power by changing one or more reactivity settings, wherein the reduced heater's power is compensated by using the increased power generated by the reactor itself. The nuclear instrumentation is configured to measure the reactor's output power and compare it with the calculated reactor's power. Based on the discrepancy between the calculated power and the measured power exceeds a threshold; and the nuclear instrumentation is appropriately calibrated by adjusting one or more settings.

To obtain an operation plan having high operability including width and a period of adjustment force.SOLUTION: A nuclear power generation plant 100 includes a power generation system 101 and an output control device 102. The output control device includes: an evaluation criterion input device 28 for inputting a market price prediction value of wholesale power and a consumption adjustment; an adjustment force setting device 26 for setting an adjustment force supply amount for adjusting consumption from information on a price prediction value and a reactor core / plant state including a margin for a thermal restriction of at least a reactor core and a control rod position; an operation plan evaluation device 25 for generating an operation plan from an adjustment force supply amount and evaluation propriety of the plan; a dynamic characteristic evaluation system 24 for evaluating a reactor core / plant state on the basis of an operation plan; and an operation plan output device 27 for outputting an operation plan obtained by a repeated evaluation of the adjustment force setting device and the operation plan evaluation device.SELECTED DRAWING: Figure 1

The application provides a unit control method, device and equipment suitable for a third generation pressurized water reactor, and the method comprises the following steps: if it is detected that a single-column AHP system fails to be isolated, the feedwater temperature of a secondary loop feedwater system is monitored in real time; if the difference of the feedwater temperature in a preset unit time is greater than a preset feedwater deviation value, the real-time operation power of the unit is obtained; whether the real-time operation power meets a preset unit power adjustment condition is judged according to a preset operation control strategy; if yes, the operation power of the unit is adjusted according to the unit power adjustment value in the operation control strategy. The embodiment of the application can adjust the operation power of the unit according to the unit power adjustment value in the operation control strategy, effectively improves the availability of the unit operation after the single-column AHP system is isolated, and ensures the safety and reliability of the unit operation.

The invention provides a method for reducing tritium emission of a pressurized water reactor unit, which comprises the following steps of: discharging a primary loop coolant into a refueling water tank in a containment vessel for temporary storage for t time, reducing tritium concentration in the primary loop coolant by utilizing natural decay, and then discharging the primary loop coolant so as to reduce the tritium emission of the pressurized water reactor unit, the t time is a fuel circulation period of the pressurized water reactor unit. The method is reasonable in design and easy to operate, the primary loop coolant is temporarily stored through the refueling water tank in the existing containment, the tritium concentration in the primary loop coolant is reduced through tritium natural decay, tritium emission of the pressurized water reactor unit can be reduced, additional devices and equipment are not needed, and the tritium treatment cost of the pressurized water reactor unit cannot be increased.

The present disclosure provides a nuclear reactor reactivity compensation method and device, electronic equipment and storage medium, relates to the technical field of nuclear reactors, and reflects the physical model of the core structure and the arrangement of the irradiation target in the nuclear reactor by constructing the physical model, inputs the neutron physical parameters related to the core and the target, and carries out numerical simulation based on the neutron transport theory, determines the reactivity change by comparing the effective multiplication factor under the reference state and the target different loading state, and generates a compensation operation strategy combined with the inherent reactivity control parameters of the nuclear reactor, so that the problem that the reactivity evaluation is inaccurate due to the lack of a systematic quantitative evaluation method and the failure to comprehensively consider the key influencing factors of the target in the prior art can be solved, and the reactivity regulation range of the reactor may be exceeded, and the physical safety of the core operation is threatened, so that the technical effect of realizing the precise quantitative evaluation of the reactivity change introduced by the irradiation target is achieved.

A method and system for monitoring nuclear power plant axial power deviation, and a terminal device. The method for monitoring nuclear power plant axial power deviation comprises: acquiring real-time monitoring data (S10); wherein the real-time monitoring data is from a plurality of nuclear power plant monitoring systems, and the plurality of nuclear power plant monitoring systems comprise at least two of an operation monitoring system, a control rod control and position system, a nuclear sampling system and a reactor core measurement system; according to the real-time monitoring data, determining important display parameters and preset display parameters (S20); wherein the preset display parameters comprise a real-time axial power deviation, a continuous parameter, an axial power deviation influence parameter and an auxiliary function parameter; and outputting the preset display parameters and the important display parameters (S30).

The invention relates to the field of nuclear power plants, in particular to a coolant circulation loop pressure control method and device, electronic equipment and a medium. After a nuclear power unit start-stop instruction is obtained, a real-time data flow of a cooling loop is collected in real time; once the data flow meets the preset boundary condition, a backpressure regulating valve at the downstream of a high-pressure reducing valve is locked at a first opening degree, then high-pressure downstream isolating valves are sequentially opened, the high-pressure reducing valve is gradually opened according to a two-stage flow target, the high-pressure side flow is firstly regulated according to a preset numerical value, and then the operation mode of the backpressure regulating valve is switched; after the flow of the high-pressure side is stable, the high-pressure reducing valve continues to be finely adjusted, the low-pressure discharging total flow is reduced, so that the load is stably transferred from the low-pressure station to the high-pressure station, then the high-pressure reducing valve is automatically put into use, the low-pressure reducing valve is closed, switching success is confirmed according to the real-time flow, manual intervention is not needed in the whole process, human errors are remarkably reduced, and the working efficiency is improved. And the operation safety and reliability are improved.

The application discloses a nuclear power KRTN16 channel in-situ processor hardware integrated system and a hot switching method, which comprises a total power supply, a main power supply, a backup power supply, a normal channel switch, a backup channel switch, a switching execution unit, a switching control unit, a mainboard card and a backup board card. The total power supply is connected with the input end of the normal channel switch and the input end of the backup channel switch. The main power supply is connected with the output end of the normal channel switch and the input end of the mainboard card. The control signal output end of the mainboard card and the control signal output end of the backup board card are connected with the input end of the switching control unit. The output end of the switching control unit is connected with the input end of the switching execution unit. The output end of the switching execution unit is connected with the execution signal input end of the normal channel switch and the execution signal input end of the backup channel switch. The output end of the backup channel switch is connected with the input end of the backup board card through the backup power supply. The application can realize hot switching quickly.

The invention discloses a high-temperature gas cooled reactor electric power control method and system based on model predictive control, and belongs to the technical field of nuclear reactor control. The method comprises the following steps: constructing a prediction model based on system dynamic characteristics of a high-temperature gas cooled reactor to quantitatively predict a future electric power output trend; at each sampling moment, on the basis of a current actual measurement state and a prediction model, an optimal control input sequence is calculated in a rolling manner by solving an optimization problem with constraints on line; and applying the control quantity at the current moment to the reactor, and performing feedback correction by using an error between actual output and a predicted value to realize closed-loop control. Through combination of prediction, rolling optimization and feedback correction, complex dynamic characteristics of nonlinearity, strong coupling and parameter time varying of the high-temperature gas cooled reactor are effectively processed, overshoot is remarkably reduced, adjustment time is shortened, tracking precision is improved, control fluctuation is reduced, and the method is suitable for large-scale popularization and application. And an advanced and reliable technical scheme is provided for safe, stable and efficient power regulation of the high-temperature gas cooled reactor.

This invention relates to the field of intelligent control technology for compressed air systems, and discloses an intelligent control and gas supply distribution method and system for the compressed air system pipeline network of a nuclear facility. The method includes: constructing a simulation model of the compressed air pipeline network system, and numerically simulating the flow and pressure characteristics of the compressed air pipeline network under different air compressor station distributions, operating conditions, and intermittent loads of different process equipment; based on the data generated by the numerical simulation, using an intelligent fuzzy self-tuning PID control method to adjust the pressure at the end of the compressed air pipeline network system, and optimizing the operating status of the air compressor units in the compressed air pipeline network system based on a flow matching strategy; and real-time acquisition of the pipeline network operating parameters and feedback to adjust the control strategy. Overall, this method can achieve efficient and stable operation of the compressed air system, ensuring the compressed air needs of all aspects of the nuclear facility, while improving the intelligence and reliability of system operation and reducing the adverse effects caused by pressure instability and other problems.

It allows for proper control of the indoor atmosphere during emergencies. [Solution] A nitrogen concentration management device for managing the nitrogen concentration of a controlled space in which a nitrogen storage unit is located in which nitrogen is stored at a pressure higher than atmospheric pressure, comprising: a concentration measuring unit for measuring the oxygen concentration or nitrogen concentration of the space; an oxygen supply unit capable of supplying oxygen to the space; and a control device for controlling the supply of oxygen from the oxygen supply unit to the space based on the measurement results of the concentration measuring unit, wherein the oxygen supply unit includes a compressed gas storage unit filled with oxygen-containing gas at a pressure higher than that of the space; a piping unit which is a path for supplying oxygen from the compressed gas storage unit to the space; and a switching unit located in the piping unit which switches whether or not to supply oxygen to the space under the control of the control device.

The embodiment of the application provides a kind of regulating rod control method and device, electronic equipment, nuclear power plant system and medium, belong to nuclear power plant safety technical field.The method comprises: in response to the working condition information of nuclear power plant system is load rejection condition, the target power of reactor is acquired;According to target power and initial correction factor, determine the secondary loop load demand signal;According to the first initial position of power regulating rod and secondary loop load demand signal, control power regulating rod moves;According to the second initial position of temperature regulating rod, adjust initial correction factor to obtain target correction factor;According to target correction factor, target power, the measured temperature of primary loop subsystem and the measured power of reactor, control temperature regulating rod moves to intermediate position;According to intermediate position and preset safety position, desalted water is injected into primary loop subsystem until temperature regulating rod reaches safety position.The embodiment of the application quickly improves the safety of nuclear power plant under load rejection condition.

A water level control system of a pressue regulator of a pressurized water reactor comprises a heat exchanger, a one-loop system, a temperature measurement module, a water level setting value generation module, a pressure regulator, a water level measurement apparatus, a rotational speed control module and an upper filling pump. The heat exchanger is communicated with the one-loopp system by means of an upper filling channel (l) and a lower discharge channel (2), respectively. The temperature measurement module respectively measures the temperature of a cold pipe section (3) and a hot pipe section (4) of the one-loop system and obtains the average temperature of the one-loop system. The temperature measurement module is in communication connection with the water level setting value generation module, and the water level setting value generation module, according to the average temperature, generates a water level setting value of the pressure regulator, The water level measurement apparatus measures the real-time water level value of the pressure regulator, The rotational speed control module, according to the water level setting value and the real-time water level value, generates a control command for controlling the rotational speed of the upper filling pump (5). The upper filling pump (5) is arranged in the upper filing channel, so as to supply water to the one-loop system, and the upper filling pump (5) controls the water supply speed according to the control command, such that adjustment precision can be ensured while frequent actions of an actuating mechanism are avoided.

The invention relates to the field of nuclear energy comprehensive utilization, in particular to a sodium-cooled fast reactor power generation-energy storage-hydrogen production comprehensive energy utilization system. The system comprises a sodium-cooled fast reactor main loop system, a first-stage fused salt heat storage loop, a second-stage fused salt heat storage and heat supply loop and a Cu-Cl circulating hydrogen production system. The main loop of the sodium-cooled fast reactor supplies heat to the first-stage molten salt tank through the sodium-molten salt heat exchanger to form a high-temperature molten salt heat storage layer; primary fused salt transfers heat to a secondary fused salt tank through a fused salt-fused salt heat exchanger, part of the secondary fused salt directly supplies water to the hydrolysis reaction system, part of the secondary fused salt is heated by an electric heater and then supplies heat to the hydrolysis reaction system, and high-temperature fused salt after hydrolysis and pyrolysis is shunted by a flow regulating valve and supplied to the hydrogen production reaction system, so that graded heat supply and gradient utilization of energy are realized. HCl gas generated by the hydrolysis reaction system is absorbed by the absorption tower to release heat and then provides a low-temperature heat source for the oxidation and crystallization system, so that reaction waste heat recovery is realized; high-temperature waste heat output by the hydrolysis and pyrolysis reaction system is used for supplying heat to the hydrogen production reaction system, so that gradient utilization of energy is formed. CuCl is oxidized by the oxidation system to generate a CuCl2 solution, CuCl2 crystals are generated by the crystallization system through evaporation, cooling and solid-liquid separation, and closed circulation of materials is achieved. The system forms a nuclear energy-fused salt energy storage-chemical hydrogen production multi-stage coupling system, time-sharing storage, steady-state heat supply and efficient hydrogen production of nuclear energy can be achieved, the heat energy utilization rate and the hydrogen energy output efficiency of a nuclear power device are improved, and the system has the advantages of being high in safety and energy utilization rate and excellent in economical efficiency.