A system and method for rapid cooling of a primary circuit of a high-temperature gas-cooled reactor nuclear power plant after hot test

Abstract

The invention discloses a system and method for rapid cooling of a primary circuit of a high-temperature gas-cooled reactor nuclear power plant after a hot test, comprising a reactor pressure vessel, a steam generator, a helium-water cooler, a gas-water separator, a helium compressor, an auxiliary steam pipeline, a first valve group, a second valve group, a high-pressure heater, a feed water pump, a fifth valve group, a gas-water separator, a condenser, a fourth valve group, a third valve group and a waste liquid tank. The system and method can overcome the problem that the hot test rate is uncontrollable during the natural cooling process, shorten the cooling time, and at the same time, the secondary side of the steam generator can participate in the hot test to verify the reliability of the secondary circuit system in advance. At the same time, the hot test process can be used to perform water vapor purge on the secondary side pipeline of the steam generator to ensure that the water quality of the secondary circuit feed water is qualified.

Description

Technical Field

[0001] The invention belongs to the technical field of nuclear power, and relates to a system and method for rapid cooling of a primary circuit after hot test of a high-temperature gas-cooled reactor nuclear power plant. Background Art

[0002] The primary loop hot test of a high temperature gas-cooled reactor nuclear power plant uses the operating power of the main helium fan to heat the primary loop hot test area to 250°C and maintain a thermal equilibrium state to verify whether the thermal function of the primary loop system is consistent with the design requirements. After the hot test is completed, the primary loop is cooled by the accident dehumidification column in the helium purification system. Since the main function of the accident dehumidification column is to remove the moisture adsorbed by the graphite and carbon brick components of the primary loop, its effect on the cooling of the primary loop is not obvious. The primary loop cooling is mainly through the heat dissipation of the core to the air outside the pressure vessel compartment. This natural cooling method has the problems of long cooling time and uncontrollable cooling rate.

[0003] The hot test process of the primary circuit of a high-temperature gas-cooled reactor is different from that of a pressurized water reactor nuclear power plant. During the hot test, the secondary circuit system is isolated, and the reliability of the secondary circuit water steam system and equipment cannot be verified. At the same time, the secondary side of the steam generator is not filled with water (nitrogen protection), and the secondary side of the steam generator cannot be purged with water steam during the hot test. According to previous engineering practice experience, unqualified water quality of the secondary circuit feed water will seriously restrict the unit startup process. It is necessary to complete the relevant flushing and verification tests of the secondary circuit water steam system during the hot test. Summary of the invention

[0004] The purpose of the present invention is to overcome the shortcomings of the above-mentioned prior art and to provide a system and method for rapid cooling of the primary circuit after hot test of a high-temperature gas-cooled reactor nuclear power plant. The system and method can overcome the problem of uncontrollable hot test rate during natural cooling and shorten the cooling time. At the same time, the secondary side of the steam generator can participate in the hot test to verify the reliability of the secondary circuit system in advance. At the same time, the hot test process can be used to perform water vapor purge on the secondary side pipeline of the steam generator to ensure that the water quality of the secondary circuit feed water is qualified.

[0005] To achieve the above-mentioned object, the high-temperature gas-cooled reactor nuclear power plant primary circuit post-hot test rapid cooling system of the present invention comprises a reactor pressure vessel, a steam generator, a helium-water cooler, a gas-water separator, a helium compressor, an auxiliary steam pipeline, a first valve group, a second valve group, a high-pressure heater, a feed water pump, a fifth valve group, a gas-water separator, a condenser, a fourth valve group, a third valve group and a waste liquid tank;

[0006] The reactor pressure vessel is connected to the primary side of the steam generator, and the primary side outlet of the steam generator is connected to the primary side outlet of the steam generator via a helium-water cooler, a gas-water separator and a helium compressor in sequence;

[0007] The outlet of the auxiliary steam pipeline is divided into two routes, one of which is connected to the inlet of the first valve group, and the other is connected to the inlet of the high-pressure heater through the second valve group, the outlet of the feed water pump is connected to the inlet of the high-pressure heater, the outlet of the high-pressure heater is connected to the inlet of the fifth valve group, the outlet of the fifth valve group is connected to the outlet of the first valve group through a pipeline and then connected to the secondary side inlet of the steam generator, the secondary side outlet of the steam generator is connected to the inlet of the steam-water separator, the top outlet of the steam-water separator is connected to the inlet of the condenser, and the bottom outlet of the steam-water separator is divided into two routes, one of which is connected to the inlet of the condenser through the fourth valve group, and the other is connected to the inlet of the waste liquid tank through the third valve group.

[0008] The reactor pressure vessel is connected to the primary side of the steam generator through a hot gas conduit.

[0009] The primary side outlet of the steam generator is connected to the primary side outlet of the steam generator via the first stop valve, the helium-water cooler, the gas-water separator, the helium compressor and the second stop valve in sequence.

[0010] The method for rapid cooling of a primary circuit of a high temperature gas-cooled reactor nuclear power plant after a hot test of the present invention comprises the following steps:

[0011] 1) After the primary loop hot test is completed, the helium temperature in the reactor pressure vessel and steam generator is 250°C, the first stop valve, the second stop valve, the first valve group, the second valve group, the fifth valve group, the third valve group, and the fourth valve group are all in a closed state, the helium compressor is shut down, and the auxiliary steam pipeline and feedwater pump are in a standby state;

[0012] 2) opening the first stop valve and the second stop valve, starting the helium compressor, and driving the flow of helium in the reactor pressure vessel and the steam generator to form a circulating cooling loop;

[0013] 3) gradually opening the first valve group, charging the auxiliary steam output from the auxiliary steam pipeline into the secondary side of the steam generator, and controlling the steam charging speed through the first valve group to ensure that the secondary side of the steam generator does not experience flow-induced vibration during the charging process, and the steam-water mixture output from the secondary side of the steam generator is discharged into the steam-water separator, and the steam generator and its pipelines are warmed up;

[0014] 4) The pressure, temperature and flow rate of the auxiliary steam are adjusted by the first valve group, so that the auxiliary steam temperature is always 10°C-20°C lower than the primary loop helium temperature of the high temperature gas-cooled reactor nuclear power plant, and the flow rate of the auxiliary steam matches the heat exchange of the helium to achieve the best cooling effect. At the same time, the cooling rate meets the unit operation limit requirements. The third valve group is adjusted so that the liquid in the steam-water separator is at a normal liquid level, the condensed water separated by the steam-water separator is discharged into the waste liquid tank, and the steam separated by the steam-water separator enters the condenser;

[0015] 5) When the helium temperature of the primary circuit of the high-temperature gas-cooled reactor nuclear power plant drops below 105°C, the first valve group is closed, and the second valve group and the fifth valve group are gradually opened. The water output by the feedwater pump is heated by the high-pressure heater and then sent to the secondary side of the steam generator. The feedwater temperature on the secondary side of the steam generator is controlled by the second valve group, so that the feedwater temperature on the secondary side of the steam generator is always 10-20°C lower than the helium temperature of the primary circuit, and the feedwater flow rate is matched with the helium heat exchange capacity. The feedwater flow rate and pressure on the secondary side of the steam generator are adjusted by the fifth valve group to ensure that no flow-induced vibration occurs on the pipe side of the steam generator during the charging process, so as to achieve the best cooling effect;

[0016] While cooling the helium in the primary circuit of the high-temperature gas-cooled reactor nuclear power plant, flush the secondary side of the steam generator. When the condensate in the steam-water separator meets the unit feed water standard, close the third valve group, open the fourth valve group, and recycle the condensate to the condenser;

[0017] When the helium temperature in the primary circuit of the steam generator drops below 40°C, close the second valve group, shut down the helium compressor, close the first stop valve and the second stop valve, and the cooling of the primary circuit of the high-temperature gas-cooled reactor nuclear power plant is completed.

[0018] The present invention has the following beneficial effects:

[0019] The system and method for rapid cooling after hot test of the primary circuit of a high-temperature gas-cooled reactor nuclear power plant described in the present invention, in specific operation, controls the cooling temperature and flow of the auxiliary steam or feed water in real time according to the helium temperature of the primary circuit, so as to overcome the problem of uncontrollable cooling rate during natural cooling. At the same time, the secondary side of the steam generator participates in the hot test to verify the reliability of the secondary circuit system and equipment in advance. In addition, while the helium of the primary circuit is cooling, the secondary side of the steam generator is flushed. When the condensate in the steam-water separator meets the water supply standard of the unit, the third valve group is closed, the fourth valve group is opened, and the condensate is recovered to the condenser, so that the water quality of the secondary circuit feed water is qualified, thereby improving the reliability of the steam generator operation. BRIEF DESCRIPTION OF THE DRAWINGS

[0020] Figure 1 It is a structural schematic diagram of the present invention.

[0021] Among them, 1 is the reactor pressure vessel, 2 is the steam generator, 3 is the first stop valve, 4 is the helium-water cooler, 5 is the gas-water separator, 6 is the helium compressor, 7 is the second stop valve, 8 is the auxiliary steam pipeline, 9 is the first valve group, 10 is the second valve group, 11 is the feed water pump, 12 is the high-pressure heater, 13 is the fifth valve group, 14 is the gas-water separator, 15 is the third valve group, 16 is the fourth valve group, 17 is the waste liquid tank, and 18 is the condenser. DETAILED DESCRIPTION

[0022] The present invention is further described in detail below in conjunction with the accompanying drawings:

[0023] refer to Figure 1 The high-temperature gas-cooled reactor nuclear power plant primary circuit post-hot test rapid cooling system of the present invention comprises a reactor pressure vessel 1, a steam generator 2, a helium-water cooler 4, a gas-water separator 5, a helium compressor 6, an auxiliary steam pipeline 8, a first valve group 9, a second valve group 10, a high-pressure heater 12, a feed water pump 11, a fifth valve group 13, a gas-water separator 14, a condenser 18, a fourth valve group 16, a third valve group 15 and a waste liquid tank 17; the reactor pressure vessel 1 is connected to the primary side of the steam generator 2, and the primary side outlet of the steam generator 2 is connected to the primary side outlet of the steam generator 2 via the helium-water cooler 4, the gas-water separator 5 and the helium compressor 6 in sequence; the outlet of the auxiliary steam pipeline 8 is divided into two paths, one of which is connected to the first side outlet of the steam generator 2; The inlet of one valve group 9 is connected, and the other is connected to the inlet of the high-pressure heater 12 through the second valve group 10. The outlet of the feed water pump 11 is connected to the inlet of the high-pressure heater 12. The outlet of the high-pressure heater 12 is connected to the inlet of the fifth valve group 13. The outlet of the fifth valve group 13 is connected to the outlet of the first valve group 9 through a pipeline and then connected to the secondary side inlet of the steam generator 2. The secondary side outlet of the steam generator 2 is connected to the inlet of the steam-water separator 14. The top outlet of the steam-water separator 14 is connected to the inlet of the condenser 18. The bottom outlet of the steam-water separator 14 is divided into two routes, one of which is connected to the inlet of the condenser 18 through the fourth valve group 16, and the other is connected to the inlet of the waste liquid tank 17 through the third valve group 15.

[0024] The reactor pressure vessel 1 is connected to the primary side of the steam generator 2 through a hot gas conduit; the primary side outlet of the steam generator 2 is connected to the primary side outlet of the steam generator 2 through a first stop valve 3, a helium-water cooler 4, a gas-water separator 5, a helium compressor 6 and a second stop valve 7 in sequence.

[0025] The method for rapid cooling of a primary circuit of a high temperature gas-cooled reactor nuclear power plant after a hot test of the present invention comprises the following steps:

[0026] 1) After the primary loop hot test is completed, the helium temperature in the reactor pressure vessel 1 and the steam generator 2 is 250°C, the first stop valve 3, the second stop valve 7, the first valve group 9, the second valve group 10, the fifth valve group 13, the third valve group 15, and the fourth valve group 16 are all in a closed state, the helium compressor 6 is shut down, and the auxiliary steam pipeline 8 and the feedwater pump 11 are in a standby state;

[0027] 2) Open the first stop valve 3 and the second stop valve 7, start the helium compressor 6, and drive the flow of helium in the reactor pressure vessel 1 and the steam generator 2 to form a circulating cooling loop;

[0028] 3) gradually opening the first valve group 9, charging the auxiliary steam output from the auxiliary steam pipeline 8 into the secondary side of the steam generator 2, and controlling the steam charging speed through the first valve group 9 to ensure that the secondary side of the steam generator 2 does not experience flow-induced vibration during the charging process, and the steam-water mixture output from the secondary side of the steam generator 2 is discharged into the steam-water separator 14, and the steam generator 2 and its pipelines are warmed up;

[0029] 4) The pressure, temperature and flow rate of the auxiliary steam are adjusted by the first valve group 9, so that the auxiliary steam temperature is always 10°C-20°C lower than the primary loop helium temperature of the high temperature gas-cooled reactor nuclear power plant, and the flow rate of the auxiliary steam matches the heat exchange of the helium to achieve the best cooling effect. At the same time, the cooling rate meets the unit operation limit requirements, and the third valve group 15 is adjusted so that the liquid in the steam-water separator 14 is at a normal liquid level, the condensed water separated by the steam-water separator 14 is discharged into the waste liquid tank 17, and the steam separated by the steam-water separator 14 enters the condenser 18;

[0030] 5) When the helium temperature of the primary loop of the high-temperature gas-cooled reactor nuclear power plant drops to below 105°C, the first valve group 9 is closed, and the second valve group 10 and the fifth valve group 13 are gradually opened. The water output by the feedwater pump 11 is heated by the high-pressure heater 12 and then sent to the secondary side of the steam generator 2. The feedwater temperature of the secondary side of the steam generator 2 is controlled by the second valve group 10, so that the feedwater temperature of the secondary side of the steam generator 2 is always 10-20°C lower than the helium temperature of the primary loop, and the feedwater flow rate is matched with the helium heat exchange capacity. The feedwater flow rate and pressure of the secondary side of the steam generator 2 are adjusted by the fifth valve group 13 to ensure that no flow-induced vibration occurs on the pipe side of the steam generator 2 during the charging process, so as to achieve the best cooling effect;

[0031] While cooling the helium gas in the primary loop of the high-temperature gas-cooled reactor nuclear power plant, the secondary side of the steam generator 2 is flushed. When the condensate in the steam-water separator 14 meets the unit feed water standard, the third valve group 15 is closed, the fourth valve group 16 is opened, and the condensate is recovered to the condenser 18;

[0032] When the helium temperature of the primary circuit of the steam generator 2 drops below 40° C., the second valve group 10 is closed, the helium compressor 6 is shut down, the first stop valve 3 and the second stop valve 7 are closed, and the cooling of the primary circuit of the high temperature gas-cooled reactor nuclear power plant is completed.

[0033] Embodiment 1

[0034] This embodiment takes a domestic high temperature gas-cooled reactor nuclear power unit as an example.

[0035] 1) During the hot test, the helium parameters in the reactor pressure vessel 1 and the steam generator 2 in the primary circuit of the high temperature gas-cooled reactor nuclear power plant are 7MPa, 250℃, and 6.331kg / s. The first stop valve 3, the second stop valve 7, the first valve group 9, the second valve group 10, the fifth valve group 13, the third valve group 15, and the fourth valve group 16 are all in a closed state, the helium compressor 6 is shut down, and the auxiliary steam pipeline 8 and the feed water pump 11 are in a standby state; the auxiliary steam is provided by the auxiliary electric boiler with two parameters of steam, namely, saturated steam with a pressure of 1.25MPa, a flow rate of 34t / h, and a temperature of 193℃ and superheated steam with a pressure of 1.25MPa, a flow rate of 5.7t / h, and a temperature of 310℃. The feed water pump 11 is an electrically debugged feed water pump, and the feed water flow rate and feed water pressure are adjusted by the feed water pump 11;

[0036] 2) After the hot test, the cooling of the primary circuit begins, the first stop valve 3 and the second stop valve 7 are opened, the helium compressor 6 is started, and the helium enters the helium-water cooler 4 and the gas-water separator 5 through the primary side of the steam generator 2, and then is sent to the steam generator 2 through the helium compressor 6 to form a circulating cooling circuit, wherein the operating parameters of the helium-water cooler 4 are: flow rate of 300kg / h, shell side inlet / outlet temperature of 7 / 12°C, tube side inlet / outlet temperature of 250 / 10°C, and pressure of 8.1MPa; the operating parameters of the gas-water separator 5 are: flow rate of 150kg / h, working temperature of 10°C, and pressure of 7.0Mpa; the operating parameters of the helium compressor 6 are: suction / exhaust pressure of 7.0 / 8.1MPa, and flow rate of 150kg / h;

[0037] 3) Slowly open the first valve group 9 to charge the auxiliary steam into the secondary side of the steam generator 2. The charging speed of the auxiliary steam is controlled by the first valve group 9 to ensure that the secondary side of the steam generator 2 does not experience flow-induced vibration during the charging process. The steam-water mixture output from the secondary side of the steam generator 2 is discharged into the steam-water separator 14 to warm up the steam generator 2 and its pipelines.

[0038] 4) The pressure, temperature and flow rate of the auxiliary steam are adjusted through the first valve group 9, so that the temperature of the auxiliary steam is always 10-20°C lower than the temperature of the primary helium, so that the flow rate matches the heat exchange of the helium to achieve the best cooling effect, and the cooling rate meets the unit operation limit requirements. The third valve group 15 is adjusted to make the liquid in the steam-water separator 14 at a normal liquid level, the condensed water separated by the steam-water separator 14 is discharged into the waste liquid tank 17, and the steam separated by the steam-water separator 14 is discharged into the condenser 18; the auxiliary steam input parameters are shown in Table 1 below.

[0039] Table 1

[0040]

[0041]

[0042] When the temperature range of helium in the first circuit is 240-250℃, adjust the auxiliary steam pressure to 1.25MPa, the temperature to 230℃ and the flow rate to 15t / h; when the temperature range of helium in the first circuit is 230-240℃, adjust the auxiliary steam pressure to 1.25MPa, the temperature to 220℃ and the flow rate to 20t / h; according to the thermal parameters in Table 1, add auxiliary steam in sequence until the temperature range of helium in the first circuit drops to 105-120℃, and then adjust the auxiliary steam pressure to 0.11MPa, the temperature to 100℃ and the flow rate to 34t / h.

[0043] 5) When the helium temperature of the primary circuit drops to below 105°C, close the first valve group 9, slowly open the second valve group 10 and the fifth valve group 13, and feed water is heated by the high-pressure heater 12 and then sent to the secondary side of the steam generator 2. The feed water temperature is controlled by the second valve group 10, and the feed water flow and pressure are controlled by the fifth valve group 13 to ensure that no flow-induced vibration occurs on the pipe side of the steam generator 2 during the charging process. The feed water temperature is controlled by the second valve group 10 so that the feed water temperature is always 10-20°C lower than the helium temperature of the primary circuit, so as to match the flow rate with the helium heat exchange capacity and achieve the best cooling effect. The feed water input parameters are shown in Table 2 below.

[0044] Table 2

[0045] Primary circuit temperature range (℃) 90～105 80～90 70～80 60～70 Feed water temperature (℃) 80 70 60 40 Water flow rate (t / h) 10 15 20 30 Water supply pressure (MPa) 2.5 2.0 1.5 1.0

[0046] Referring to Table 2, when the primary helium temperature range is 90-105°C, the feed water pressure is adjusted to 2.5MPa, the temperature is 80°C, and the flow rate is 10t / h; when the primary helium temperature range is 80-90°C, the feed water pressure is adjusted to 2.0MPa, the temperature is 70°C, and the flow rate is 15t / h; when the primary helium temperature range is 70-80°C, the feed water pressure is adjusted to 1.5MPa, the temperature is 60°C, and the flow rate is 20t / h; when the primary helium temperature range is 60-70°C, the feed water pressure is adjusted to 1.0MPa, the temperature is 40°C, and the flow rate is 30t / h;

[0047] While the primary helium is being cooled, the secondary side of the steam generator 2 is flushed. When the condensate in the steam-water separator 14 meets the unit feed water standard, the third valve group 15 is closed, the fourth valve group 16 is opened, and the condensate is recovered to the condenser 18. The qualified standards for feed water sampling are shown in Table 3 below.

[0048] Table 3

[0049] Parameters unit data Cation conductivity μs / cm ≤0.2 Dissolved oxygen μg / L ≤3 chlorine μg / L ≤5 iron μg / L ≤10 copper μg / L ≤2 sodium μg / L ≤2 Silicon dioxide μg / L ≤15 Suspended Solids μg / L ≤10 Organic matter μg / L 0 hardness μmol / L 0 pH 9.6-9.8

[0050] When the helium temperature of the primary circuit drops below 40° C., the second valve group 10 is closed, the helium compressor 6 is shut down, the first stop valve 3 and the second stop valve 7 are closed, and the primary circuit cooling is completed.

Claims

1. A method for rapid cooling of a primary circuit of a high temperature gas-cooled reactor nuclear power plant after a hot test, characterized in that: Based on a primary loop hot test rapid cooling system for a high temperature gas-cooled reactor nuclear power plant, the primary loop hot test rapid cooling system for a high temperature gas-cooled reactor nuclear power plant comprises a reactor pressure vessel (1), a steam generator (2), a helium-water cooler (4), a gas-water separator (5), a helium compressor (6), an auxiliary steam pipeline (8), a first valve group (9), a second valve group (10), a high pressure heater (12), a feed water pump (11), a fifth valve group (13), a gas-water separator (14), a condenser (18), a fourth valve group (16), a third valve group (15) and a waste liquid tank (17); The reactor pressure vessel (1) is connected to the primary side of the steam generator (2), and the primary side outlet of the steam generator (2) is connected to the primary side outlet of the steam generator (2) via a helium-water cooler (4), a gas-water separator (5) and a helium compressor (6) in sequence; The outlet of the auxiliary steam pipeline (8) is divided into two paths, one of which is connected to the inlet of the first valve group (9), and the other is connected to the inlet of the high-pressure heater (12) through the second valve group (10). The outlet of the feed water pump (11) is connected to the inlet of the high-pressure heater (12), and the outlet of the high-pressure heater (12) is connected to the inlet of the fifth valve group (13). The outlet of the fifth valve group (13) and the outlet of the first valve group (9) are connected to the secondary side inlet of the steam generator (2) through a pipeline. The secondary side outlet of the steam generator (2) is connected to the inlet of the steam-water separator (14). The top outlet of the steam-water separator (14) is connected to the inlet of the condenser (18). The bottom outlet of the steam-water separator (14) is divided into two paths, one of which is connected to the inlet of the condenser (18) through the fourth valve group (16), and the other is connected to the inlet of the waste liquid tank (17) through the third valve group (15); The primary side outlet of the steam generator (2) is connected to the primary side outlet of the steam generator (2) via the first stop valve (3), the helium-water cooler (4), the gas-water separator (5), the helium compressor (6) and the second stop valve (7) in sequence; The following steps are involved: 1) After the primary circuit hot test is completed, the helium temperature in the reactor pressure vessel (1) and the steam generator (2) is 250° C., the first stop valve (3), the second stop valve (7), the first valve group (9), the second valve group (10), the fifth valve group (13), the third valve group (15), and the fourth valve group (16) are all in a closed state, the helium compressor (6) is shut down, and the auxiliary steam pipeline (8) and the feed water pump (11) are in a standby state; 2) opening the first stop valve (3) and the second stop valve (7), starting the helium compressor (6), and driving the flow of helium in the reactor pressure vessel (1) and the steam generator (2) to form a circulating cooling loop; 3) gradually opening the first valve group (9) to charge the auxiliary steam outputted from the auxiliary steam pipeline (8) into the secondary side of the steam generator (2), and at the same time controlling the charging speed of the steam through the first valve group (9) to ensure that the secondary side of the steam generator (2) does not experience flow-induced vibration during the charging process, and discharging the steam-water mixture outputted from the secondary side of the steam generator (2) into the steam-water separator (14), thereby warming the steam generator (2) and its pipelines; 4) adjusting the pressure, temperature and flow rate of the auxiliary steam by means of the first valve group (9) so that the auxiliary steam temperature is always 10°C-20°C lower than the helium temperature of the primary circuit of the high temperature gas-cooled reactor nuclear power plant, and the flow rate of the auxiliary steam matches the heat exchange capacity of the helium to achieve the best cooling effect, while the cooling rate meets the unit operation restriction requirements, adjusting the third valve group (15) so that the liquid in the steam-water separator (14) is at a normal liquid level, the condensed water separated by the steam-water separator (14) is discharged into the waste liquid tank (17), and the steam separated by the steam-water separator (14) enters the condenser (18); 5) When the helium temperature of the primary circuit of the high temperature gas-cooled reactor nuclear power plant drops to below 105° C., the first valve group (9) is closed, the second valve group (10) and the fifth valve group (13) are gradually opened, the water output by the feed water pump (11) is heated by the high pressure heater (12) and then sent to the secondary side of the steam generator (2), the feed water temperature of the secondary side of the steam generator (2) is controlled by the second valve group (10), so that the feed water temperature of the secondary side of the steam generator (2) is always lower than the helium temperature of the primary circuit by 10-20° C., and the feed water flow rate is matched with the helium heat exchange capacity, the feed water flow rate and pressure of the secondary side of the steam generator (2) are adjusted by the fifth valve group (13), so that the flow-induced vibration phenomenon does not occur on the pipe side of the steam generator (2) during the charging process, so as to achieve the best cooling effect; While cooling the primary helium loop of the high temperature gas-cooled reactor nuclear power plant, the secondary side of the steam generator (2) is flushed, and when the condensate in the steam-water separator (14) meets the unit feed water standard after testing, the third valve group (15) is closed, the fourth valve group (16) is opened, and the condensate is recovered to the condenser (18); When the helium temperature in the primary circuit of the steam generator (2) drops below 40° C., the second valve group (10) is closed, the helium compressor (6) is shut down, the first stop valve (3) and the second stop valve (7) are closed, and the primary circuit of the high temperature gas-cooled reactor nuclear power plant is cooled.

2. The method for rapid cooling of a primary circuit of a high temperature gas-cooled reactor nuclear power plant after hot test according to claim 1, characterized in that: The reactor pressure vessel (1) and the primary side of the steam generator (2) are connected via a hot gas conduit.

Images