System and method for verifying passive pulse cooling of nuclear power plant

Abstract

The invention discloses a system and a method for verifying passive pulse cooling of a nuclear power plant, which can be used for verifying a passive pulse cooling mechanism and influence factors under the accident that the whole plant loses all water supply due to power loss superposition. The system comprises a secondary-loop passive injection system, a primary-loop high-temperature and high-pressure water loop, a steam generator experiment piece, a cooling system and a measuring system. In an experiment process, a primary-loop high-temperature high-pressure water loop firstly establishes a water circulation condition, a discharge valve of a steam generator experiment piece is kept to be opened, deaerator feed water of a secondary-loop passive injection system is driven by superheated steam to enter a secondary side of the steam generator experiment piece, and secondary-loop feed water is heated and gasified by primary-loop high-temperature water through a U-shaped pipe; gasification pressurization causes secondary loop water supply stagnation, pressurization stagnation and continuous pressure relief of a discharge valve at the same time so as to enable water supply to circulate, and water supply circulates and stops alternately, so that the passive pulse cooling experiment effect is achieved. The method provides principle-level data for development and transformation of the nuclear power plant, and provides real and effective experimental verification.

Description

TECHNICAL FIELD

[0001] The present application relates to the technical field of nuclear power plant accident protection mechanism verification, and in particular to a system and method for verifying passive pulse cooling of a nuclear power plant. BACKGROUND

[0002] In view of the insufficient ability of the existing second-generation nuclear power plant to cope with the "loss of off-site power superimposed on loss of all feedwater" extreme accident, a leading scholar in the field of nuclear safety disclosed a passive pulse cooling method in patent CN111370153, which divides the steam generator of the nuclear power plant into two power evaporators and one cooling evaporator. First, the cooling evaporator is depressurized, second, without using an additional power source, the power evaporator and the superheated steam of the high-pressure steam pipeline are used in a passive manner to drive the lower-temperature feedwater of the deaerator into the cooling evaporator, and finally, high-temperature and high-pressure water in the cooling loop is injected into the steam generator to relieve the accident danger and reduce the possibility of core meltdown.

[0003] During the injection into the cooling evaporator, the heating and vaporization of the feedwater will increase the secondary side pressure, which will hinder and interrupt the injection of the feedwater. At the same time, the depressurization process of the cooling evaporator continues, and after the secondary side pressure decreases, the feedwater resumes injection, resulting in a cyclic pulse process of "injection-interruption-injection" of the feedwater. The cyclic process increases the time with the decrease of the cooling loop temperature until the end of the pulse effect.

[0004] During the pulse cooling process, there are gas-liquid phase changes and time cycle changes, and the calculation simulation means cannot accurately verify the rationality of this method, so it is of practical significance to develop a principle-level experimental device and method using the prototype parameters of the nuclear power plant. SUMMARY

[0005] In order to solve the problems existing in the prior art, the purpose of the present application is to provide a system and method for verifying passive pulse cooling of a nuclear power plant, which overcomes the lack of principle experimental verification of the passive pulse cooling method.

[0006] In order to achieve the above purpose, the present application adopts the following technical solutions:

[0007] The application provides a system and method for verifying passive pulse cooling of a nuclear power plant, which can be used for verifying passive pulse cooling mechanism and influencing factors under the condition of full-plant loss of off-site power superimposed loss of all feedwater accidents, and the system comprises a secondary loop passive injection system 23, a primary loop high-temperature and high-pressure water circuit 24, a steam generator experimental unit 12, a cooling system 25 and a measuring system. During the experiment, the primary loop high-temperature and high-pressure water circuit 24 first establishes a water circulation condition, the steam generator experimental unit 12 discharge valve 11 is kept open, the deaerator 5 feedwater of the secondary loop passive injection system 23 is driven into the secondary side of the steam generator experimental unit 12 by superheated steam, the secondary loop feedwater is heated and gasified through the U-shaped tube by the primary loop high-temperature water, the gasification pressurization causes the secondary loop feedwater to stagnate, the pressurization stagnation is accompanied by the continuous pressure relief of the discharge valve 11, and then the feedwater flows, so that the alternating flow and stagnation of the feedwater occur, and the passive pulse cooling experiment effect appears.

[0008] The secondary loop passive injection system 23 comprises an electric heating boiler 1 for providing superheated steam, a first regulating valve 2-1 arranged on a superheated steam pipeline, a pressure reducing valve 4 and a second regulating valve 2-2, the superheated steam outlet of the electric heating boiler 1 passes through the first regulating valve 2-1 for rough flow adjustment, passes through the pressure reducing valve 4 for adjusting the superheated steam to a specified pressure, and passes through the second regulating valve 2-2 for fine flow adjustment; the superheated steam pipeline is connected to the inlet of a deaerator 5, a one-way check valve 8 and a fourth regulating valve 2-4 are connected to the outlet of the deaerator 5, and the deaerator 5 is internally provided with a heating rod; during the experiment, the superheated steam provided by the electric heating boiler 1 drives the feedwater in the deaerator 5 to enter the steam generator experimental unit 12.

[0009] The primary side U-shaped tube of the steam generator experimental unit 12 circulates the high-temperature and high-pressure water of the primary loop high-temperature and high-pressure water circuit 24, and the secondary side circulates the saturated water of the secondary loop passive injection system 23, and the top pipeline is provided with a discharge valve 11.

[0010] Preferably, the primary high-temperature and high-pressure water circuit 24 comprises a preheating section 18, a first electrically-operated stop valve 13-1 and a bypass passage release valve 14 arranged between the outlet of the preheating section 18 and the inlet of the steam generator test piece 12, a second electrically-operated stop valve 13-2 connecting the outlet of the steam generator test piece 12 and the inlet of the condenser 15, a closed water tank 16, a sixth regulating valve 2-6, a centrifugal pump 17, a fifth regulating valve 2-5 and the preheating section 18 connected in sequence at the outlet of the condenser 15. During the test, the high-temperature and high-pressure water is driven by the centrifugal pump 17, passes through the inlet and outlet of the steam generator test piece 12, is cooled by the condenser 15 after passing through the second electrically-operated stop valve 13-2, reaches the closed water tank 16, passes through the sixth regulating valve 2-6 to the centrifugal pump 17, passes through the fifth regulating valve 2-5 to the preheating section 18, is heated by the preheating section 18, and then reaches the inlet of the steam generator test piece 12 through the first electrically-operated regulating valve 13-1, thereby completing the circuit cycle.

[0011] Preferably, the cooling system 25 comprises the condenser 15, a filter 19, a third electrically-operated stop valve 13-3, a circulating pump 20, a fourth electrically-operated stop valve 13-4, a first open water tank 21-1 and a cooling water tower 22 connected in sequence at the outlet of the condenser 15, and the cooling water tower 22 is connected to the inlet of the condenser 15 through a seventh regulating valve 2-7. During the operation of the condenser 15, the cooling water is driven by the circulating pump 20, enters the inlet of the condenser 15 from the cooling water tower 22 through the seventh regulating valve 2-7, passes through the filter 19, the third electrically-operated stop valve 13-3, the circulating pump, the fourth electrically-operated stop valve 13-4 and the first open water tank 21-1, and finally reaches the cooling water tower 22, thereby completing the circuit cycle.

[0012] Preferably, the measuring system comprises a first pressure sensor 3-1 and a second pressure sensor 3-2 arranged before and after the pressure-reducing valve 4 of the high-pressure steam pipeline, a third pressure sensor 3-3 arranged at the outlet of the deaerator 5, a fourth pressure sensor 3-4 arranged at the upper side of the steam generator test piece 12, a fifth pressure sensor 3-5 arranged at the lower side of the U-shaped tube of the steam generator test piece 12, a sixth pressure sensor 3-6 and a seventh pressure sensor 3-7 arranged at the inlet and outlet of the primary side of the steam generator test piece 12, a first thermocouple 10-1 arranged at the outlet of the deaerator 5, a second thermocouple 10-2 arranged at the lower side of the U-shaped tube of the steam generator test piece 12, a fourth thermocouple 10-4 and a fifth thermocouple 10-5 arranged at the inlet and outlet of the primary side of the steam generator test piece 12, a sixth thermocouple 10-6 and a third thermocouple 10-3 arranged at the inlet and outlet of the preheating section 18, a differential pressure transmitter 6 arranged at the deaerator 5, a first flowmeter 9-1 arranged before the inlet of the secondary side of the steam generator test piece 12, a second flowmeter 9-2 arranged after the discharge valve 11 of the steam generator test piece 12, and a third flowmeter 9-3 arranged before the inlet of the primary side of the steam generator test piece 12.

[0013] The electric heating boiler 1 preferably generates 2.5 MPa superheated steam to meet the actual high-pressure steam parameters of a nuclear power plant, and the water source of the electric heating boiler 1 is provided by the second open water tank 21-2.

[0014] The superheated steam at the outlet of the electric heating boiler 1 is preferably adjusted in flow rate by the first regulating valve 2-1, adjusted to a specified pressure by the pressure reducing valve 4, and adjusted in flow rate by the second regulating valve 2-2.

[0015] The deaerator 5 is preferably a pressure-bearing heated water tank with an effective volume of 1 m 3 , which can withstand a pressure of 2.5 MPa, and is provided with an electric heating rod. Before the feed water is injected, the electric heating rod can heat the feed water in the deaerator 5 to a saturated temperature of 0.93 MPa, so that the feed water reaches a saturated temperature of 0.93 MPa. The water source of the deaerator 5 is provided by the second open water tank 21-2 through the third regulating valve 2-3.

[0016] The deaerator 5 preferably sets the saturated water temperature at 0.93 MPa, 10℃ lower than the saturated water temperature, 20℃ lower than the saturated water temperature, and 30℃ lower than the saturated water temperature, respectively, to investigate the influence of the feed water temperature on the pulse cooling.

[0017] The deaerator 5 preferably adopts two driving modes of air-side pressurization and water-side pressurization. The air-side pressurization means that the superheated steam is injected from the upper end of the deaerator, and the superheated steam is directly introduced into the saturated steam at the upper end of the deaerator 5. The water-side pressurization means that the superheated steam is injected from the lower end of the deaerator, and the superheated steam is directly introduced into the saturated water at the lower end of the deaerator 5.

[0018] The one-way check valve 8 preferably ensures that when the pressure in the steam generator experimental device 12 rises, the steam will not flow back to the deaerator.

[0019] The opening degree adjustment of the fourth regulating valve 2-4 can preferably simulate the flow fluctuation of the main feed water pipeline and the bypass channel.

[0020] The first thermocouple 10-1 and the third pressure sensor 3-3 are preferably arranged on the feed water pipeline connecting the deaerator 5 and the steam generator experimental device 12. The first thermocouple 10-1 and the third pressure sensor 3-3 can be arranged at an equal distance of 3-4 along the feed water pipeline to investigate the phase change that may occur in the feed water during transportation and analyze the possibility of water hammer phenomenon.

[0021] The steam generator experimental device 12 preferably adopts 15 U-shaped tubes with less number to represent the heat exchange capacity after modeling. The U-shaped tubes are made of stainless steel or inconel material. Stainless steel can save economic cost, and inconel is used as the actual material of a nuclear power plant, which has a large cost. Inconel is used to additionally investigate the structural strength of the steam generator in the extreme pulse cooling process.

[0022] Preferably, to avoid the second thermocouple 10-2 affecting the phase change flow field inside the steam generator test piece 12, the second thermocouple 10-2 is arranged at the position where the U-tube of the steam generator test piece 12 contacts the tube plate.

[0023] Preferably, the discharge valve 11 simulates the atmospheric bypass discharge valve of a nuclear power plant, and further two safety valves are added to investigate the influence of the additional safety valve number on the pulse cooling.

[0024] Preferably, the primary high-temperature high-pressure water circuit 24 meets the actual primary circuit parameters and fluctuation requirements 15.5 MPa, 330℃ of a nuclear power plant.

[0025] Preferably, the preheating section 18 precisely adjusts the water temperature entering the primary side inlet of the steam generator test piece 12, which can simulate the change of the actual decay power of a nuclear power plant with the primary circuit temperature after pulse cooling.

[0026] Preferably, the measurement system records the water temperature and pressure at the inlet and outlet of the primary side of the steam generator test piece 12, the water temperature at the inlet and outlet of the secondary side of the steam generator test piece 12, the temperature at the bottom of the U-tube of the steam generator test piece 12, the pressure at the upper side and the lower side, the flow rate of the feedwater entering the secondary side of the steam generator test piece 12, and the steam flow rate of the feedwater leaving the secondary side of the steam generator test piece 12.

[0027] Preferably, the thermocouple adopts an industrial I-grade precision T-type armored thermocouple with a temperature measurement accuracy of 0.5℃ and a temperature measurement range of -200-350℃.

[0028] Preferably, the first flow meter 9-1 of the secondary passive injection system 23 and the third flow meter 9-3 of the primary high-temperature high-pressure water circuit 24 adopt a Coriolis force mass flow meter to measure the water flow rate, and the second flow meter 9-2 of the steam generator test piece 12 adopts a thermal gas mass flow meter to measure the steam flow rate.

[0029] Preferably, the working temperature of the centrifugal pump 17 is 60℃, and the water from the outlet of the primary side of the steam generator test piece 12 needs to be cooled by a condenser to meet the working temperature of the centrifugal pump 17.

[0030] A method for verifying passive pulse cooling of nuclear power plants, according to the working condition requirements, adjusting the fourth regulating valve 2-4, the discharge valve 11; after the flow of the primary loop high temperature and high pressure water circuit 24 is stable, adjust the opening degree of each valve in the primary loop high temperature and high pressure water circuit 24, and reach the water flow and pressure conditions required by the experimental working condition; open the preheating section 18, stepwise and slowly increase the power of the preheating section power supply, heat the primary side water of the steam generator experimental piece 12 to the predetermined steam generator experimental piece 12 inlet temperature; start the built-in heating rod in the deaerator 5, gradually increase the heating power, and the feedwater temperature reaches the predetermined temperature; start the electric heating boiler 1, gradually increase the heating power, and generate stable superheated steam; record the temperature, pressure and flow data of the measurement system as a steady state control; close the built-in heating rod in the deaerator 5, open the one-way check valve 8 at the deaerator outlet, the first regulating valve 2-1 at the outlet of the electric heating boiler 1 and the pressure reducing valve 4 in turn, record and observe the measurement system data until the deaerator storage water is completely injected into the steam generator experimental piece 12; in turn close the first regulating valve 2-1 at the outlet of the electric heating boiler 1, the pressure reducing valve 4 and the one-way check valve 8 at the deaerator outlet, complete a group of working condition experiments; repeat the operation in the steps to carry out the next working condition experiment; during the experiment, the cooling system 25 cools the steam generator experimental piece 12 primary side outlet water; after the experiment, close the primary side preheating section 18 and the electric heating boiler 1, and then close the primary side centrifugal pump 17 to unload the pressure of the primary loop.

[0031] Preferably, a method for verifying passive pulse cooling of nuclear power plants is applied to the steam generator experimental piece 12 secondary side without water, and there may be some water in the steam generator secondary side under the actual accident working condition of the nuclear power plant. The one-way check valve 8 and the fourth regulating valve 2-4 are opened during the experimental preparation stage, and then the one-way check valve 8 and the fourth regulating valve 2-4 are closed after the deaerator 5 provides some water, and the water volume of the deaerator 5 is supplemented, and then the experiment is carried out.

[0032] Compared with the prior art, the present application has at least the following advantages:

[0033] 1) The electric heating boiler 1 is used instead of the power steam generator used in the passive pulse cooling of the nuclear power plant, the deaerator 5 uses a pressure-bearing heated water tank instead of the prototype secondary loop deaerator, and the opening degree of the fourth regulating valve 2-4 is adjusted to simulate the flow fluctuation of the main feedwater pipe and the bypass channel, thereby saving experimental cost and reducing experimental complexity.

[0034] 2) The preheating section 18 is used to heat the primary loop high temperature and high pressure water circuit 24, and the centrifugal pump 17 adjusts the flow and pressure, so that the inlet temperature, flow and pressure of the steam generator experimental piece 12 can be accurately adjusted, and the working condition after the actual accident of the nuclear power plant can be effectively simulated.

[0035] 3) The volume of the deaerator 5, the heat transfer area of the heat transfer pipe of the steam generator experimental piece 12 and the volume of the steam space of the secondary side of the steam generator experimental piece 12 are modeled in the same proportion, which can effectively reduce the equipment scale and save the experimental cost under the premise of ensuring the effectiveness of the principle level data.

[0036] 4) The experimental working condition of the present application is redundant, and the experimental parameters can be adjusted in a large range to meet the verification needs of the actual working condition of the nuclear power plant, and the calculation simulation multi-condition change results can be further verified to broaden the calculation and theoretical research range.

[0037] 5) The present application provides two driving modes of air-side pressurization and water-side pressurization for the deaerator 5, which provides a reference for the modification of different nuclear reactor plants.

[0038] 6) The present application can be used as a principle experimental device to directly verify the feasibility of the passive pulse cooling method in the nuclear power plant, and provide data verification for the modification and operation of the existing second-generation nuclear power plant. BRIEF DESCRIPTION OF DRAWINGS

[0039] Figure 1 The present application is a first example of a system and method for verifying the passive pulse cooling of a nuclear power plant.

[0040] Figure 2 The present application is a second example of a system and method for verifying the passive pulse cooling of a nuclear power plant.

[0041] 1 is an electric boiler, 2 is a regulating valve, 3 is a pressure sensor, 4 is a pressure reducing valve, 5 is a deaerator, 6 is a differential pressure transmitter, 7 is a safety valve, 8 is a one-way check valve, 9 is a flowmeter, 10 is a thermocouple, 11 is a discharge valve, 12 is a steam generator experimental piece, 13 is an electric stop valve, 14 is a release valve, 15 is a condenser, 16 is a closed water tank, 17 is a high-pressure centrifugal pump, 18 is a preheating section, 19 is a filter, 20 is a circulating pump, 21 is a starting water tank, 22 is a cooling tower, 23 is a two-loop passive injection system, 24 is a one-loop high-temperature high-pressure water circuit, and 25 is a cooling system. DETAILED DESCRIPTION

[0042] The technical solutions in the embodiments of the present application are described in detail in combination with the example drawings of the present application. Obviously, the described embodiments are only a part of the embodiments of the present application, not all the embodiments. Based on the embodiments of the present application, all other embodiments obtained by those skilled in the art without creative labor are within the protection scope of the present application.

[0043] The present application provides a system and method for verifying the passive pulse cooling of a nuclear power plant, Figure 1A system diagram for verifying passive pulse cooling of a nuclear power plant using water-side pressurization according to the present application, superheated steam is injected from the lower end of the deaerator, and the superheated steam is directly introduced into the saturated water in the lower end of the deaerator 5.

[0044] The system of the present example includes a secondary circuit passive injection system 23, a primary circuit high-temperature high-pressure water circuit 24, a steam generator test piece 12, a cooling system 25, and a measurement system; the secondary circuit passive injection system 23 includes an electrically heated boiler 1, a first regulating valve 2-1, a pressure reducing valve 4, a second regulating valve 2-2, a third regulating valve 2-3, a deaerator 5, a one-way check valve 8, and a fourth regulating valve 2-4, and a second open water tank 21-2.

[0045] During the experiment, 0.93 MPa saturated feedwater in the deaerator 5 is driven into the steam generator test piece 12 by 2 MPa superheated steam provided by the electrically heated boiler 1.

[0046] The primary circuit high-temperature high-pressure water circuit 24 includes a preheating section 18, a first electrically driven stop valve 13-1, a release valve 14, a second electrically driven stop valve 13-2, a condenser 15, a closed water tank 16, a sixth regulating valve 2-6, a centrifugal pump 17, and a fifth regulating valve 2-5.

[0047] During the experiment, the centrifugal pump 17 provides the driving force, and 15.5 MPa, 330°C high-temperature high-pressure water passes through the inlet and outlet of the steam generator test piece 12, is cooled to 50°C by the condenser 15 after passing through the second electrically driven stop valve 13-2, reaches the closed water tank 16, reaches the centrifugal pump 17 after passing through the sixth regulating valve 2-6, reaches the preheating section 18 after passing through the fifth regulating valve 2-5, and finally reaches the inlet of the steam generator test piece 12 after being warmed up by the preheating section 18, completing the loop cycle.

[0048] The cooling system 25 includes the condenser 15, a filter 19, a third electrically driven stop valve 13-3, a circulating pump 20, a fourth electrically driven stop valve 13-4, a first open water tank 21-1, a cooling water tower 22, and a seventh regulating valve 2-7.

[0049] During the experiment, the condenser 15 is working, and cold water from the cooling water tower 22 enters the inlet of the condenser 15 through the seventh regulating valve 2-7, passes through the filter 19, the third electrically driven stop valve 13-3, the circulating pump 20, the fifth electrically driven stop valve 13-5, and the first open water tank 21-1, and finally reaches the cooling water tower 22, completing the loop cycle. The water from the outlet of the primary side of the steam generator test piece 12 is cooled by the condenser 15 to meet the working temperature of the centrifugal pump 17.

[0050] The measuring system comprises a first pressure sensor 3-1 and a second pressure sensor 3-2 arranged before and after a pressure reducing valve 4 on a high-pressure steam pipeline, a third pressure sensor 3-3 arranged at an outlet of a deaerator 5, a fourth pressure sensor 3-4 arranged at an upper side of a steam generator test piece 12, a fifth pressure sensor 3-5 arranged at a lower side of a U-shaped tube of the steam generator test piece 12, a sixth pressure sensor 3-6 and a seventh pressure sensor 3-7 arranged at an inlet and an outlet of a primary side of the steam generator test piece 12, a first thermocouple 10-1 arranged at an outlet of the deaerator, a second thermocouple 10-2 arranged at a lower side of the U-shaped tube of the steam generator test piece 12, a fourth thermocouple 10-4 and a fifth thermocouple 10-5 arranged at an inlet and an outlet of a primary side of the steam generator test piece 12, a sixth thermocouple 10-6 and a third thermocouple 10-3 arranged at an inlet and an outlet of a preheating section 18, a differential pressure transmitter 6 arranged at the deaerator 5, a first flowmeter 9-1 arranged before an inlet of a secondary side of the steam generator test piece 12, a second flowmeter 9-2 arranged after a drain valve 11 of the steam generator test piece 12, and a third flowmeter 9-3 arranged before an inlet of a primary side of the steam generator test piece 12.

[0051] The measuring system records the steam pressure and the feedwater temperature of the secondary loop passive injection system 23, judges whether the experimental conditions are met, and analyzes the phase change caused by the pressure drop change of the feedwater flow through the deaerator 5. The measuring system records the water temperature and the pressure of the high-temperature and high-pressure water loop 24 flowing through the inlet and the outlet of the primary side of the steam generator test piece 12, judges whether the experimental conditions are met, and analyzes the cooling effect of the pulse cooling method on the steam generator. The measuring system records the pressure of the upper and lower sides of the steam generator test piece 12 and the wall temperature of the U-shaped tube, and judges whether the feedwater “flow-interruption-flow” pulse phenomenon occurs by combining the pressure fluctuation and the first flowmeter 9-1 of the secondary loop passive injection system 23.

[0052] The method for verifying the passive pulse cooling of a nuclear power plant first adjusts the fourth regulating valve 2-4 and the discharge valve 11 to be fully open; waits for the flow of the primary loop high-temperature high-pressure water circuit 24 to be stable, adjusts the opening degree of each valve in the primary loop high-temperature high-pressure water circuit 24, and the circuit reaches 15.5 MPa; the preheating section 18 is started, the power of the preheating section is slowly increased in steps, and the primary side water of the steam generator test piece 12 is heated to 330 DEG C; the built-in heating rod in the deaerator 5 is started, the heating power is gradually increased, and the water temperature reaches 230 DEG C, which is the saturated water temperature of 0.93 MPa; the electric heating boiler 1 is started, the heating power is gradually increased, and stable 2 MPa superheated steam is generated; the temperature, pressure and flow data of the measurement system are recorded as a steady state control; the built-in heating rod in the deaerator 5 is closed, the one-way check valve 8 at the outlet of the deaerator, the first regulating valve 2-1 at the outlet of the electric heating boiler 1 and the pressure reducing valve 4 are opened in turn, the measurement system data are recorded and observed, and the deaerator storage water is completely injected into the steam generator test piece 12; the first regulating valve 2-1 at the outlet of the electric heating boiler 1, the pressure reducing valve 4 and the one-way check valve 8 at the outlet of the deaerator are closed in turn, and a group of working condition experiments are completed; the operation in the steps is repeated, and the next working condition experiment is carried out; during the experiment, the cooling system 25 cools the primary side outlet water of the steam generator test piece 12; after the experiment, the primary side preheating section 18 and the electric heating boiler 1 are closed, and the primary side centrifugal pump 17 is closed, and the primary loop is depressurized.

[0053] Figure 2 A system schematic diagram for verifying the passive pulse cooling of a nuclear power plant using gas side pressurization, the superheated steam is injected from the upper end of the deaerator, and the superheated steam is directly introduced into the saturated steam at the upper end of the deaerator 5. The other structures and working methods of the gas side pressurization system are the same as those of the water side pressurization system.

[0054] The present application meets the parameter conditions of the prototype working condition of the nuclear power plant, measures the pressure and temperature change of all time nodes of the establishment and disappearance of the pulse cooling, and can verify the rationality of the passive pulse cooling method, and provides principle level data for the development and reconstruction of the nuclear power plant, and provides real and effective experimental verification.

[0055] The above content is only used to illustrate the present application, and cannot be regarded as the limitation of the specific embodiments of the present application, and any modification made on the basis of the technical scheme falls within the protection scope of the claims of the present application.

Claims

1. A system for verifying passive pulse cooling of a nuclear power plant, characterized by: The application relates to a mechanism and influencing factors of a passive pulse cooling under a whole-plant power failure superimposed with a total loss of feedwater accident in a verifiable nuclear power plant, which comprises a secondary circuit passive injection system (23), a primary circuit high-temperature high-pressure water loop (24), a steam generator experimental piece (12), a cooling system (25) and a measuring system; the secondary circuit passive injection system (23) is used for injecting saturated water into the secondary side of the steam generator experimental piece (12); the primary circuit high-temperature high-pressure water loop (24) is used for providing high-temperature high-pressure water for the primary side U-shaped tube of the steam generator experimental piece (12), and a preheating section (18) is arranged in the primary circuit high-temperature high-pressure water loop (24); the cooling system (25) is used for cooling the steam generator experimental piece (12); and the measuring system is used for measuring the pressure, temperature and flow of the secondary circuit passive injection system (23), the primary circuit high-temperature high-pressure water loop (24) and the steam generator experimental piece (12). The secondary circuit passive injection system (23) comprises an electric heating boiler (1) for providing superheated steam, a first regulating valve (2-1) arranged on a superheated steam pipeline, a pressure reducing valve (4) and a second regulating valve (2-2), superheated steam at the outlet of the electric heating boiler (1) is adjusted in flow by the first regulating valve (2-1), the superheated steam is adjusted to a specified pressure by the pressure reducing valve (4), and the flow of the superheated steam is adjusted by the second regulating valve (2-2); the superheated steam pipeline is connected to the inlet of a deaerator (5), a one-way check valve (8) and a fourth regulating valve (2-4) are connected to the outlet of the deaerator (5), and a heating rod is arranged in the deaerator (5); during the experiment, superheated steam provided by the electric heating boiler (1) drives feedwater in the deaerator (5) to enter the steam generator experimental piece (12). The primary side U-shaped tube of the steam generator experimental piece (12) flows through high-temperature high-pressure water of the primary circuit high-temperature high-pressure water loop (24), and the secondary side flows through saturated water of the secondary circuit passive injection system (23); a discharge valve (11) is arranged at the top end of the U-shaped tube; secondary circuit feedwater is heated and gasified by the high-temperature water of the primary circuit, the pressurization of the gasification causes the secondary circuit feedwater to stagnate, the pressurization stagnation and the continuous pressure relief of the discharge valve (11) cause the feedwater to flow, and the alternating flow and stagnation of the feedwater causes the passive pulse cooling experiment effect.

2. A system for verifying passive pulse cooling of a nuclear power plant according to claim 1, characterized in that: The one loop high temperature and high pressure water circuit (24) comprises a preheating section (18), a first electric stop valve (13-1) and a bypass passage release valve (14) arranged between the outlet of the preheating section (18) and the inlet of the steam generator test piece (12), a second electric stop valve (13-2) connected between the outlet of the steam generator test piece (12) and the inlet of the condenser (15), a closed water tank (16), a sixth regulating valve (2-6), a centrifugal pump (17), a fifth regulating valve (2-5) and the preheating section (18) connected in sequence at the outlet of the condenser (15). During the test, the high temperature and high pressure water is driven by the centrifugal pump (17), passes through the inlet and outlet of the steam generator test piece (12), is cooled by the condenser (15) after passing through the second electric stop valve (13-12), reaches the closed water tank (16), passes through the sixth regulating valve (2-6) to the centrifugal pump (17), passes through the fifth regulating valve (2-5) to the preheating section (18), is heated by the preheating section (18), and then reaches the inlet of the steam generator test piece (12) through the first electric regulating valve (13-1), to complete the loop circulation.

3. A system for verifying passive pulse cooling of a nuclear power plant according to claim 1, characterized in that: The cooling system (25) comprises the condenser (15), a filter (19), a third electric stop valve (13-3), a circulating pump (20), a fourth electric stop valve (13-4), a first open water tank (21-1) and a cooling water tower (22) connected in sequence at the outlet of the condenser (15), and the cooling water tower (22) is connected to the inlet of the condenser (15) through a seventh regulating valve (2-7). When the condenser (15) works, the circulating pump (20) provides driving force, and the cold water from the cooling water tower (22) enters the inlet of the condenser (15) through the seventh regulating valve (2-7), passes through the filter (19), the third electric stop valve (13-3), the circulating pump, the fourth electric stop valve (13-4) and the first open water tank (21-1), and finally reaches the cooling water tower (22), to complete the loop circulation.

4. A system for verifying passive pulse cooling of a nuclear power plant according to claim 1, characterized in that: The measuring system comprises a first pressure sensor (3-1) and a second pressure sensor (3-2) arranged before and after a pressure reducing valve (4) on a high-pressure steam pipeline, a third pressure sensor (3-3) at an outlet of a deaerator (5), a fourth pressure sensor (3-4) at an upper side of a steam generator test piece (12), a fifth pressure sensor (3-5) at a lower side of a U-shaped tube of the steam generator test piece (12), a sixth pressure sensor (3-6) and a seventh pressure sensor (3-7) at an inlet and an outlet of a primary side of the steam generator test piece (12), a first thermocouple (10-1) arranged at the outlet of the deaerator (5), a second thermocouple (10-2) arranged at the lower side of the U-shaped tube of the steam generator test piece (12), a fourth thermocouple (10-4) and a fifth thermocouple (10-5) arranged at the inlet and the outlet of the primary side of the steam generator test piece (12), a sixth thermocouple (10-6) and a third thermocouple (10-3) arranged at an inlet and an outlet of a preheating section (18), a differential pressure transmitter (6) arranged at the deaerator (5), a first flowmeter (9-1) arranged in front of a secondary side inlet of the steam generator test piece (12), a second flowmeter (9-2) arranged behind a drain valve (11) of the steam generator test piece (12), and a third flowmeter (9-3) arranged in front of a primary side inlet of the steam generator test piece (12).

5. A system for verifying passive pulse cooling of a nuclear power plant according to claim 1, characterized in that: The deaerator (5) adopts two driving modes of air-side pressurization and water-side pressurization, the air-side pressurization being that superheated steam is injected from an upper end of the deaerator, and the superheated steam is directly introduced into saturated steam at the upper end of the deaerator (5), and the water-side pressurization being that superheated steam is injected from a lower end of the deaerator, and the superheated steam is directly introduced into saturated water at the lower end of the deaerator (5).

6. A system for verifying passive pulse cooling of a nuclear power plant according to claim 1, characterized in that: The opening degree of the fourth regulating valve (2-4) is adjusted to simulate flow fluctuation of an analog main feedwater pipeline and a bypass channel.

7. A system for verifying passive pulse cooling of a nuclear power plant according to claim 1, characterized in that: The volume of the deaerator (5), the heat transfer area of a heat transfer tube of the steam generator test piece (12), and the volume of a steam space of a secondary side of the steam generator test piece (12) are modeled in the same proportion.

8. A system for verifying passive pulse cooling of a nuclear power plant according to claim 1, characterized in that: The water source of the electric heating boiler (1) is provided by a second open water tank (21-2), and the water source of the deaerator (5) is provided by the second open water tank (21-2) through the third regulating valve (2-3).

9. A method for verifying the operation of a system for passive pulse cooling of a nuclear power plant according to any one of claims 1 to 8, characterized in that: According to the working condition requirements, adjust the fourth regulating valve (2-4) and the discharge valve (11); after the flow of the high-temperature and high-pressure water circuit (24) is stable, adjust the opening of each valve in the high-temperature and high-pressure water circuit (24) to achieve the required water flow and pressure conditions of the experimental working condition; open the preheating section (18), slowly increase the power of the preheating section in steps to heat the primary side water of the steam generator experimental device (12) to the predetermined inlet temperature of the steam generator experimental device (12); start the built-in heating rod of the deaerator (5) and gradually increase the heating power until the temperature of the feedwater reaches the predetermined temperature; start the electric heating boiler (1) and gradually increase the heating power to generate stable superheated steam; record the temperature, pressure and flow data of the measurement system as a steady-state control; close the built-in heating rod of the deaerator (5), open the one-way check valve (8) at the outlet of the deaerator, the first regulating valve (2-1) at the outlet of the electric heating boiler (1) and the pressure reducing valve (4) in sequence, record and observe the measurement system data until the deaerator storage water is completely injected into the steam generator experimental device (12); close the first regulating valve (2-1) at the outlet of the electric heating boiler (1), the pressure reducing valve (4) and the one-way check valve (8) at the outlet of the deaerator in sequence to complete a group of working condition experiments; repeat the operation to perform the next working condition experiment; during the experiment, the cooling system (25) cools the outlet water of the primary side of the steam generator experimental device (12); after the experiment, close the primary side preheating section (18) and the electric heating boiler (1), and then close the primary side centrifugal pump (17) to depressurize the primary loop.

10. The method of working according to claim 9, characterized in that: For the steam generator experimental device (12) secondary side without water state, under the actual accident working condition of the nuclear power plant, there is partial water in the secondary side of the steam generator. During the experimental preparation stage, open the one-way check valve (8) and the fourth regulating valve (2-4), provide partial water through the deaerator (5), close the one-way check valve (8) and the fourth regulating valve (2-4), supplement the water volume of the deaerator (5), and then carry out the experiment.

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