CN111276267A - System and method for quickly cooling primary circuit of high-temperature gas cooled reactor nuclear power station after thermal test - Google Patents
System and method for quickly cooling primary circuit of high-temperature gas cooled reactor nuclear power station after thermal test Download PDFInfo
- Publication number
- CN111276267A CN111276267A CN202010208465.1A CN202010208465A CN111276267A CN 111276267 A CN111276267 A CN 111276267A CN 202010208465 A CN202010208465 A CN 202010208465A CN 111276267 A CN111276267 A CN 111276267A
- Authority
- CN
- China
- Prior art keywords
- steam
- water
- steam generator
- helium
- valve bank
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
Images
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
- G21C15/16—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants comprising means for separating liquid and steam
-
- F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
- F22—STEAM GENERATION
- F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
- F22B1/00—Methods of steam generation characterised by form of heating method
- F22B1/02—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers
- F22B1/16—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers the heat carrier being hot liquid or hot vapour, e.g. waste liquid, waste vapour
- F22B1/162—Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers the heat carrier being hot liquid or hot vapour, e.g. waste liquid, waste vapour in combination with a nuclear installation
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
- G21C15/02—Arrangements or disposition of passages in which heat is transferred to the coolant; Coolant flow control devices
- G21C15/14—Arrangements or disposition of passages in which heat is transferred to the coolant; Coolant flow control devices from headers; from joints in ducts
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Landscapes
- Engineering & Computer Science (AREA)
- Physics & Mathematics (AREA)
- High Energy & Nuclear Physics (AREA)
- General Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- Life Sciences & Earth Sciences (AREA)
- Sustainable Development (AREA)
- Sustainable Energy (AREA)
- Thermal Sciences (AREA)
- Mechanical Engineering (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
The invention discloses a system and a method for rapidly cooling a primary circuit of a high-temperature gas cooled reactor nuclear power station after a thermal test, which 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 bank, a second valve bank, a high-pressure heater, a water feed pump, a fifth valve bank, a gas-water separator, a condenser, a fourth valve bank, a third valve bank and a waste liquid tank.
Description
Technical Field
The invention belongs to the technical field of nuclear power, and relates to a system and a method for quickly cooling a primary circuit of a high-temperature gas cooled reactor nuclear power station after a thermal test.
Background
The primary loop heat test of the high-temperature gas cooled reactor nuclear power station is to heat a primary loop heat test area to 250 ℃ by using the operating power of a helium main fan and maintain a heat balance state, and verify whether the thermal state function of a primary loop system is consistent with the design regulation requirement. The accident dehumidification row in the helium purification system cools a loop after the heat test is finished, and because this accident dehumidification row mainly acts on the moisture of detaching a loop graphite and the carbon brick internals absorption, its effect to a loop cooling is not obvious, and the loop cooling mainly is through the heat dissipation of reactor core to the outside air of pressure vessel cabin, and this natural cooling mode has the cooling time long, and cooling rate is uncontrollable problem.
The primary loop hot test process of the high-temperature gas cooled reactor is different from that of a pressurized water reactor nuclear power station, and the secondary loop system is isolated in the hot test process, so that the reliability of a secondary loop steam system and equipment cannot be verified; meanwhile, the secondary side of the steam generator is not filled with water (nitrogen filling protection), so that steam purging cannot be performed on the secondary side of the steam generator in the hot test process, and according to the past engineering practice experience, the starting process of the unit is severely restricted due to unqualified quality of the two-loop feed water, and the related flushing and verification tests of the two-loop steam system are necessarily completed in the hot test period.
Disclosure of Invention
The invention aims to overcome the defects of the prior art and provides a system and a method for rapidly cooling a primary circuit of a high-temperature gas cooled reactor nuclear power station after a thermal test.
In order to achieve the aim, the rapid cooling system after the loop thermal test of the high-temperature gas cooled reactor nuclear power station 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 bank, a second valve bank, a high-pressure heater, a water feed pump, a fifth valve bank, a gas-water separator, a condenser, a fourth valve bank, a third valve bank and a waste liquid tank;
the reactor pressure vessel is communicated with the primary side of a steam generator, and the outlet of the primary side of the steam generator is communicated with the outlet of the primary side of the steam generator through a helium-water cooler, a gas-water separator and a helium compressor in sequence;
the outlet of the auxiliary steam pipeline is divided into two paths, wherein one path is communicated with the inlet of the first valve bank, the other path is communicated with the inlet of the high-pressure heater through the second valve bank, the outlet of the feed water pump is communicated with the inlet of the high-pressure heater, the outlet of the high-pressure heater is communicated with the inlet of the fifth valve bank, the outlet of the fifth valve bank is communicated with the secondary side inlet of the steam generator through a pipeline and a pipe with the outlet of the first valve bank, the secondary side outlet of the steam generator is communicated with the inlet of the steam-water separator, the top outlet of the steam-water separator is communicated with the inlet of the condenser, the bottom outlet of the steam-water separator is divided into two paths, one path is communicated with the inlet of the condenser through the fourth valve bank, and the other.
The reactor pressure vessel is communicated with the primary side of the steam generator through a hot gas conduit.
The primary side outlet of the steam generator is communicated with the primary side outlet of the steam generator through a first stop valve, a helium-water cooler, a gas-water separator, a helium gas compressor and a second stop valve in sequence.
The method for rapidly cooling the primary circuit of the high-temperature gas cooled reactor nuclear power station after the thermal test comprises the following steps:
1) after a loop hot test is finished, the helium temperature in the reactor pressure vessel and a steam generator is 250 ℃, the first stop valve, the second stop valve, the first valve bank, the second valve bank, the fifth valve bank, the third valve bank and the fourth valve bank are all in a closed state, the helium compressor is stopped, and the auxiliary steam pipeline and the water feeding pump are in a standby state;
2) opening the first stop valve and the second stop valve, starting a helium compressor, and driving the helium in the reactor pressure vessel and the steam generator to flow so as to form a circulating cooling loop;
3) gradually opening the first valve group, filling auxiliary steam output by the auxiliary steam pipeline into the secondary side of the steam generator, simultaneously controlling the filling speed of the steam through the first valve group, ensuring that the secondary side of the steam generator does not generate flow-induced vibration in the filling process, discharging steam-water mixture output by the secondary side of the steam generator into a steam-water separator, and warming the steam generator and the pipeline thereof;
4) the pressure, the temperature and the flow of auxiliary steam are adjusted through a first valve bank, so that the temperature of the auxiliary steam is always lower than the temperature of helium in a primary circuit of the high-temperature gas cooled reactor nuclear power station by 10-20 ℃, the flow of the auxiliary steam is matched with the heat exchange quantity of the helium, the optimal cooling effect is achieved, meanwhile, the cooling rate meets the unit operation limitation requirement, a third valve bank is adjusted to enable liquid in a steam-water separator to be at a normal liquid level, condensed water separated by the steam-water separator is discharged into a waste liquid tank, and steam separated by the steam-water separator enters a condenser;
5) when the temperature of helium in a primary loop of the high-temperature gas cooled reactor nuclear power station is reduced to be lower than 105 ℃, closing the first valve group, gradually opening the second valve group and the fifth valve group, heating water output by a feed pump through a high-pressure heater, then sending the water into the secondary side of a steam generator, controlling the feed water temperature of the secondary side of the steam generator through the second valve group, enabling the feed water temperature of the secondary side of the steam generator to be always lower than the temperature of helium in the primary loop by 10-20 ℃, enabling the feed water flow to be matched with the heat exchange quantity of the helium, adjusting the feed water flow and the pressure of the secondary side of the steam generator through the fifth valve group, and ensuring that the tube side of the steam generator;
when helium in a primary loop of the high-temperature gas cooled reactor nuclear power station is cooled, the secondary side of a steam generator is flushed, and when the quality of the condensate water in the steam-water separator is tested and meets the unit water supply standard, a third valve bank is closed, a fourth valve bank is opened, and the condensate water is recovered to the condenser;
and when the helium temperature of the primary loop of the steam generator is reduced to below 40 ℃, closing the second valve group, stopping the helium compressor, closing the first stop valve and the second stop valve, and finishing the primary loop cooling of the high-temperature gas cooled reactor nuclear power station.
The invention has the following beneficial effects:
the rapid cooling system and the rapid cooling method after the thermal test of the primary loop of the high-temperature gas cooled reactor nuclear power station control the cooling temperature and the flow of auxiliary steam or feed water in real time according to the helium temperature of the primary loop so as to overcome the problem of uncontrollable cooling rate in the natural cooling process, and simultaneously, the secondary side of the steam generator participates in the thermal test to verify the reliability of a secondary loop system and equipment in advance. In addition, when a loop of helium gas is cooled, wash steam generator's secondary side, condensate water chemical examination quality of water satisfies unit water supply standard in catch water, closes the third valves, opens the fourth valves, retrieves condensate water to the condenser in, makes two return circuits water supply quality of water qualified, improves steam generator reliability of operation then.
Drawings
FIG. 1 is a schematic structural diagram of the present invention.
Wherein, 1 is reactor pressure vessel, 2 is steam generator, 3 is first stop valve, 4 is helium-water cooler, 5 is gas-water separator, 6 is helium compressor, 7 is second stop valve, 8 is auxiliary steam pipeline, 9 is first valves, 10 is second valves, 11 is feed pump, 12 is high pressure heater, 13 is fifth valves, 14 is catch water, 15 is third valves, 16 is fourth valves, 17 is waste liquid tank, 18 is condenser.
Detailed Description
The invention is described in further detail below with reference to the accompanying drawings:
referring to fig. 1, the rapid cooling system after the loop thermal test of the high temperature gas cooled reactor nuclear power plant of the present invention includes 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 bank 9, a second valve bank 10, a high pressure heater 12, a feed water pump 11, a fifth valve bank 13, a gas-water separator 14, a condenser 18, a fourth valve bank 16, a third valve bank 15, and a waste liquid tank 17; the reactor pressure vessel 1 is communicated with the primary side of a steam generator 2, and the outlet of the primary side of the steam generator 2 is communicated with the outlet of the primary side of the steam generator 2 through a helium-water cooler 4, a gas-water separator 5 and a helium gas compressor 6 in sequence; the outlet of the auxiliary steam pipeline 8 is divided into two paths, wherein one path is communicated with the inlet of a first valve group 9, the other path is communicated with the inlet of a high-pressure heater 12 through a second valve group 10, the outlet of a water feed pump 11 is communicated with the inlet of the high-pressure heater 12, the outlet of the high-pressure heater 12 is communicated with the inlet of a fifth valve group 13, the outlet of the fifth valve group 13 is communicated with the outlet of the first valve group 9 through a pipeline and a pipe and then communicated with the secondary side inlet of the steam generator 2, the secondary side outlet of the steam generator 2 is communicated with the inlet of a steam-water separator 14, the top outlet of the steam-water separator 14 is communicated with the inlet of a condenser 18, the bottom outlet of the steam-water separator 14 is divided into two paths, one path is communicated with the inlet of the condenser 18 through a fourth valve group 16, and the other path.
The reactor pressure vessel 1 is communicated with the primary side of the steam generator 2 through a hot gas conduit; the primary side outlet of the steam generator 2 is communicated with 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 gas compressor 6 and a second stop valve 7 in sequence.
The method for rapidly cooling the primary circuit of the high-temperature gas cooled reactor nuclear power station after the thermal test comprises the following steps:
1) after a loop hot test is finished, the helium temperatures in the reactor pressure vessel 1 and the steam generator 2 are 250 ℃, the first stop valve 3, the second stop valve 7, the first valve bank 9, the second valve bank 10, the fifth valve bank 13, the third valve bank 15 and the fourth valve bank 16 are all in a closed state, the helium compressor 6 is stopped, and the auxiliary steam pipeline 8 and the water feed 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 helium in the reactor pressure vessel 1 and the steam generator 2 to flow to form a circulating cooling loop;
3) gradually opening the first valve group 9, filling the auxiliary steam output by the auxiliary steam pipeline 8 into the secondary side of the steam generator 2, and simultaneously controlling the filling speed of the steam through the first valve group 9 to ensure that the secondary side of the steam generator 2 does not generate flow-induced vibration in the filling process, discharging the steam-water mixture output by the secondary side of the steam generator 2 into the steam-water separator 14, and heating the steam generator 2 and the pipeline thereof;
4) the pressure, the temperature and the flow of auxiliary steam are adjusted through a first valve bank 9, so that the temperature of the auxiliary steam is always lower than the temperature of helium in a primary circuit of the high-temperature gas cooled reactor nuclear power station by 10-20 ℃, the flow of the auxiliary steam is matched with the heat exchange quantity of the helium to achieve the optimal cooling effect, meanwhile, the cooling rate meets the unit operation limitation requirement, a third valve bank 15 is adjusted to enable liquid in a steam-water separator 14 to be at a normal liquid level, condensed water separated by the steam-water separator 14 is discharged into a waste liquid tank 17, and steam separated by the steam-water separator 14 enters a condenser 18;
5) when the helium temperature of a primary loop of the high-temperature gas cooled reactor nuclear power station is reduced to be lower than 105 ℃, closing the first valve group 9, gradually opening the second valve group 10 and the fifth valve group 13, heating water output by a water feed pump 11 through a high-pressure heater 12, then sending the water into the secondary side of the steam generator 2, controlling the water supply temperature of the secondary side of the steam generator 2 through the second valve group 10, enabling the water supply temperature of the secondary side of the steam generator 2 to be always lower than the helium temperature of the primary loop by 10-20 ℃, enabling the water supply flow to be matched with the heat exchange quantity of the helium, adjusting the water supply flow and the pressure of the secondary side of the steam generator 2 through the fifth valve group 13, and ensuring that the pipe side of the steam generator 2 does;
when cooling helium in a primary loop of the high-temperature gas cooled reactor nuclear power plant, flushing the secondary side of the steam generator 2, closing the third valve bank 15 and opening the fourth valve bank 16 when the quality of the condensate in the steam-water separator 14 meets the unit water supply standard, and recovering the condensate to the condenser 18;
and when the helium temperature of the primary loop of the steam generator 2 is reduced to below 40 ℃, closing the second valve group 10, stopping the helium compressor 6, closing the first stop valve 3 and the second stop valve 7, and finishing the primary loop cooling of the high-temperature gas-cooled reactor nuclear power station.
Example one
In this embodiment, a nuclear power plant of a domestic high-temperature gas cooled reactor is taken as an example.
1) During the thermal test, helium parameters in a reactor pressure vessel 1 and a steam generator 2 in a primary loop of the high-temperature gas cooled reactor nuclear power station are 7MPa, 250 ℃ and 6.331kg/s, a first stop valve 3, a second stop valve 7, a first valve bank 9, a second valve bank 10, a fifth valve bank 13, a third valve bank 15 and a fourth valve bank 16 are all in a closed state, a helium compressor 6 is stopped, and an auxiliary steam pipeline 8 and a water feed pump 11 are in a standby state; the auxiliary steam is provided with two parameters of steam by an auxiliary electric boiler, namely saturated steam with the pressure of 1.25MPa, the flow of 34t/h and the temperature of 193 ℃ and superheated steam with the pressure of 1.25MPa, the flow of 5.7t/h and the temperature of 310 ℃, the water feed pump 11 is an electric debugging water feed pump, and the water feed flow and the water feed pressure are adjusted by the water feed pump 11;
2) after the thermal test is finished, a loop cooling is started, the first stop valve 3 and the second stop valve 7 are opened, the helium compressor 6 is started, 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 into the steam generator 2 through the helium compressor 6 to form a circulating cooling loop, wherein the operation parameters of the helium-water cooler 4 are as follows: the flow rate is 300kg/h, the inlet/outlet temperature of the shell pass is 7/12 ℃, the inlet/outlet temperature of the tube pass is 250/10 ℃, and the pressure is 8.1 MPa; the operating parameters of the gas-water separator 5 are as follows: the flow rate is 150kg/h, the working temperature is 10 ℃, and the pressure is 7.0 Mpa; the operating parameters of the helium compressor 6 were: the suction/discharge pressure is 7.0/8.1MPa, and the flow is 150 kg/h;
3) slowly opening the first valve group 9, filling auxiliary steam into the secondary side of the steam generator 2, controlling the filling speed of the auxiliary steam through the first valve group 9 to ensure that the secondary side of the steam generator 2 does not generate flow-induced vibration in the filling process, discharging a steam-water mixture output by the secondary side of the steam generator 2 into the steam-water separator 14, and warming the steam generator 2 and pipelines thereof;
4) the pressure, the temperature and the flow of the auxiliary steam are adjusted through the first valve group 9, so that the temperature of the auxiliary steam is always lower than the temperature of helium in a primary circuit by 10-20 ℃, the flow is matched with the heat exchange quantity of the helium, the optimal cooling effect is achieved, and meanwhile, the cooling rate meets the unit operation limitation requirement. Adjusting the third valve bank 15 to enable liquid in the steam-water separator 14 to be at a normal liquid level, discharging condensed water separated by the steam-water separator 14 into a waste liquid tank 17, and discharging steam separated by the steam-water separator 14 into a condenser 18; the auxiliary steam input parameters are shown in table 1 below.
TABLE 1
When the temperature range of helium in the primary loop is 240-250 ℃, adjusting the auxiliary steam pressure to be 1.25MPa, the temperature to be 230 ℃ and the flow to be 15 t/h; when the temperature range of helium in the primary loop is 230-240 ℃, the auxiliary steam pressure is adjusted to be 1.25MPa, the temperature is 220 ℃, and the flow is 20 t/h; and (3) sequentially adding auxiliary steam according to the thermal parameters in the table 1 until the temperature range of helium in the primary loop is reduced to 105-120 ℃, then adjusting the auxiliary steam pressure to be 0.11MPa, the temperature to be 100 ℃ and the flow to be 34 t/h.
5) When the helium temperature of a primary loop is reduced to be lower than 105 ℃, the first valve group 9 is closed, the second valve group 10 and the fifth valve group 13 are opened slowly, feed water is heated by the high-pressure heater 12 and then is sent to the secondary side of the steam generator 2, the feed water temperature is controlled by the second valve group 10, the feed water flow and pressure are controlled by the fifth valve group 13, the phenomenon of flow-induced vibration on the pipe side of the steam generator 2 in the filling process is guaranteed not to occur, the feed water temperature is controlled by the second valve group 10, the feed water temperature is always lower than the helium temperature of the primary loop by 10-20 ℃, matching of flow and helium heat exchange quantity is achieved, the optimal cooling effect is achieved, and feed water input parameters.
TABLE 2
| Temperature range (. degree. C.) of primary circuit | 90～105 | 80～90 | 70～80 | 60～70 |
| Feed water temperature (. degree. C.) | 80 | 70 | 60 | 40 |
| Feed water flow (t/h) | 10 | 15 | 20 | 30 |
| Feed water pressure (MPa) | 2.5 | 2.0 | 1.5 | 1.0 |
Referring to table 2, when the temperature range of the helium in the primary loop is 90-105 ℃, the pressure of the feed water is adjusted to be 2.5MPa, the temperature is 80 ℃, and the flow is 10 t/h; when the temperature range of helium in the primary loop is 80-90 ℃, the pressure of feed water is adjusted to be 2.0MPa, the temperature is 70 ℃, and the flow is 15 t/h; when the temperature range of helium in the primary loop is 70-80 ℃, the pressure of feed water is adjusted to be 1.5MPa, the temperature is 60 ℃, and the flow is 20 t/h; when the temperature range of helium in the primary loop is 60-70 ℃, the pressure of feed water is adjusted to 1.0MPa, the temperature is 40 ℃, and the flow is 30 t/h;
and when the helium in the primary loop is cooled, the secondary side of the steam generator 2 is flushed, when the quality of the condensate in the steam-water separator 14 meets the unit water supply standard, the third valve bank 15 is closed, the fourth valve bank 16 is opened, the condensate is recovered to the condenser 18, and the qualified standard of water supply sampling is shown in the following table 3.
TABLE 3
| Parameter of | Unit of | Data of |
| Conductivity of cation | μs/cm | ≤0.2 |
| Dissolved oxygen | μg/L | ≤3 |
| Chlorine | μg/L | ≤5 |
| Iron | μg/L | ≤10 |
| Copper (Cu) | μg/L | ≤2 |
| Sodium salt | μg/L | ≤2 |
| Silicon dioxide | μg/L | ≤15 |
| Suspended solids | μg/L | ≤10 |
| Organic matter | μg/L | 0 |
| Hardness of | μmol/L | 0 |
| pH | 9.6-9.8 |
And when the temperature of the helium in the primary circuit is reduced to be below 40 ℃, closing the second valve group 10, stopping the helium compressor 6, closing the first stop valve 3 and the second stop valve 7, and completing cooling of the primary circuit.
Claims (4)
1. A rapid cooling system after a loop thermal test of a high-temperature gas cooled reactor nuclear power station is characterized by comprising 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 bank (9), a second valve bank (10), a high-pressure heater (12), a water feed pump (11), a fifth valve bank (13), a gas-water separator (14), a condenser (18), a fourth valve bank (16), a third valve bank (15) and a waste liquid tank (17);
the reactor pressure vessel (1) is communicated with the primary side of the steam generator (2), and the outlet of the primary side of the steam generator (2) is communicated with the outlet of the primary side of the steam generator (2) through 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, wherein one path is communicated with the inlet of a first valve bank (9), the other path is communicated with the inlet of a high-pressure heater (12) through a second valve bank (10), the outlet of a water feed pump (11) is communicated with the inlet of the high-pressure heater (12), the outlet of the high-pressure heater (12) is communicated with the inlet of a fifth valve bank (13), the outlet of the fifth valve bank (13) is communicated with the outlet of the first valve bank (9) through a pipeline and a pipe and then communicated with the secondary side inlet of a steam generator (2), the secondary side outlet of the steam generator (2) is communicated with the inlet of a steam-water separator (14), the top outlet of the steam-water separator (14) is communicated with the inlet of a condenser (18), the bottom outlet of the steam-water separator (14) is divided into two paths, and one path is communicated with the inlet of the condenser (18) through a, the other path is communicated with the inlet of a waste liquid tank (17) through a third valve group (15).
2. The system for rapidly cooling the primary circuit of the nuclear power plant after the thermal test of the high-temperature gas-cooled reactor according to claim 1, wherein the reactor pressure vessel (1) is communicated with the primary side of the steam generator (2) through a hot gas conduit.
3. The system for rapidly cooling the primary loop of the high-temperature gas-cooled reactor nuclear power plant after the thermal test according to claim 1, wherein a primary side outlet of the steam generator (2) is communicated with a 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.
4. A method for rapidly cooling a primary circuit of a high-temperature gas-cooled reactor nuclear power plant after a thermal test is characterized in that the method is based on the system for rapidly cooling the primary circuit of the high-temperature gas-cooled reactor nuclear power plant after the thermal test of claim 3 and comprises the following steps:
1) after a loop hot test is finished, the helium temperature in the reactor pressure vessel (1) and the steam generator (2) is 250 ℃, the first stop valve (3), the second stop valve (7), the first valve bank (9), the second valve bank (10), the fifth valve bank (13), the third valve bank (15) and the fourth valve bank (16) are all in a closed state, the helium compressor (6) is stopped, and the auxiliary steam pipeline (8) and the water feed 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 helium in the reactor pressure vessel (1) and the steam generator (2) to flow to form a circulating cooling loop;
3) gradually opening a first valve bank (9), filling auxiliary steam output by an auxiliary steam pipeline (8) into the secondary side of the steam generator (2), controlling the filling speed of the steam through the first valve bank (9) to ensure that the secondary side of the steam generator (2) does not generate flow-induced vibration in the filling process, discharging steam-water mixture output by the secondary side of the steam generator (2) into a steam-water separator (14), and heating the steam generator (2) and the pipeline thereof;
4) the pressure, the temperature and the flow of auxiliary steam are adjusted through a first valve bank (9), so that the temperature of the auxiliary steam is always lower than the temperature of helium in a primary loop of the high-temperature gas cooled reactor nuclear power station by 10-20 ℃, the flow of the auxiliary steam is matched with the heat exchange quantity of the helium to achieve the optimal cooling effect, meanwhile, the cooling rate meets the unit operation limitation requirement, a third valve bank (15) is adjusted to enable liquid in a steam-water separator (14) to be at a normal liquid level, condensed water separated by the steam-water separator (14) is discharged into a waste liquid tank (17), and steam separated by the steam-water separator (14) enters a condenser (18);
5) when the helium temperature of a primary loop of the high-temperature gas cooled reactor nuclear power station is reduced to be lower than 105 ℃, closing the first valve bank (9), gradually opening the second valve bank (10) and the fifth valve bank (13), heating water output by a feed pump (11) through a high-pressure heater (12), then sending the water into the secondary side of the steam generator (2), controlling the feed water temperature of the secondary side of the steam generator (2) through the second valve bank (10), enabling the feed water temperature of the secondary side of the steam generator (2) to be always lower than the helium temperature of the primary loop by 10-20 ℃, enabling the feed water flow to be matched with the helium heat exchange quantity, adjusting the feed water flow and pressure of the secondary side of the steam generator (2) through the fifth valve bank (13), and ensuring that the tube side of the steam generator (2) does not generate flow-induced vibration phenomenon in;
when helium in a primary loop of the high-temperature gas cooled reactor nuclear power plant is cooled, the secondary side of the steam generator (2) is flushed, when the quality of the condensate water in the steam-water separator (14) is tested and meets the unit water supply standard, the third valve bank (15) is closed, the fourth valve bank (16) is opened, and the condensate water is recovered to the condenser (18);
and when the helium temperature of the primary loop of the steam generator (2) is reduced to below 40 ℃, closing the second valve group (10), stopping the helium compressor (6), closing the first stop valve (3) and the second stop valve (7), and finishing the cooling of the primary loop of the high-temperature gas-cooled reactor nuclear power station.
Priority Applications (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202010208465.1A CN111276267A (en) | 2020-03-23 | 2020-03-23 | System and method for quickly cooling primary circuit of high-temperature gas cooled reactor nuclear power station after thermal test |
| PCT/CN2021/078750 WO2021190258A1 (en) | 2020-03-23 | 2021-03-02 | Rapid cooling system and method for high-temperature gas cooled reactor nuclear power station primary loop after thermal test |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202010208465.1A CN111276267A (en) | 2020-03-23 | 2020-03-23 | System and method for quickly cooling primary circuit of high-temperature gas cooled reactor nuclear power station after thermal test |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CN111276267A true CN111276267A (en) | 2020-06-12 |
Family
ID=71003860
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN202010208465.1A Pending CN111276267A (en) | 2020-03-23 | 2020-03-23 | System and method for quickly cooling primary circuit of high-temperature gas cooled reactor nuclear power station after thermal test |
Country Status (2)
| Country | Link |
|---|---|
| CN (1) | CN111276267A (en) |
| WO (1) | WO2021190258A1 (en) |
Cited By (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN109767852A (en) * | 2019-02-22 | 2019-05-17 | 西安热工研究院有限公司 | A kind of secondary circuit security system and its working method for reactor emergency shut-down |
| CN112435765A (en) * | 2020-11-23 | 2021-03-02 | 华能山东石岛湾核电有限公司 | High-temperature gas cooled reactor steam generator small-flow cooling system and control method |
| WO2021190258A1 (en) * | 2020-03-23 | 2021-09-30 | 西安热工研究院有限公司 | Rapid cooling system and method for high-temperature gas cooled reactor nuclear power station primary loop after thermal test |
| CN113819453A (en) * | 2021-10-28 | 2021-12-21 | 华能山东石岛湾核电有限公司 | Device and method for increasing feed water temperature in starting stage of high-temperature gas cooled reactor |
| CN114017759A (en) * | 2021-11-02 | 2022-02-08 | 华能山东石岛湾核电有限公司 | Cooling system of high-temperature gas cooled reactor nuclear power station |
| CN114165778A (en) * | 2021-11-04 | 2022-03-11 | 华能核能技术研究院有限公司 | High-temperature gas cooled reactor secondary loop system and method for improving main water supply operation temperature |
| CN114220574A (en) * | 2021-11-03 | 2022-03-22 | 华能核能技术研究院有限公司 | System and method for quickly cooling steam generator of high-temperature gas cooled reactor |
| CN114220579A (en) * | 2021-10-29 | 2022-03-22 | 华能核能技术研究院有限公司 | Boosting system and method for pressure test before service of primary loop of high-temperature gas cooled reactor |
Families Citing this family (6)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN114038587B (en) * | 2021-11-10 | 2023-11-24 | 中国核动力研究设计院 | Start-stop system and method suitable for modularized small pressurized water reactor nuclear power plant |
| CN114459261B (en) * | 2022-03-11 | 2024-03-19 | 浙江嘉诚动能科技股份有限公司 | Multi-module reactor nuclear power unit evaporator water supply system |
| CN115083646B (en) * | 2022-06-23 | 2023-06-27 | 华能核能技术研究院有限公司 | Method for rapidly cooling steam generator after emergency shutdown of high-temperature gas cooled reactor |
| CN115719655B (en) * | 2022-11-21 | 2024-05-24 | 华能山东石岛湾核电有限公司 | Device and method for high-temperature gas cooled reactor main helium fan baffle test |
| CN116525166A (en) * | 2023-04-28 | 2023-08-01 | 华能山东石岛湾核电有限公司 | High temperature gas cooled reactor primary loop coolant multi-form 3 H、 14 C activity concentration calculation method and device |
| CN117476254B (en) * | 2023-09-27 | 2024-06-04 | 华能核能技术研究院有限公司 | Method and system for stopping modular high-temperature gas cooled reactor unit |
Family Cites Families (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2015068563A1 (en) * | 2013-11-07 | 2015-05-14 | 日立Ｇｅニュークリア・エナジー株式会社 | Cooling system for nuclear reactor suppression pool |
| CN106887265B (en) * | 2017-03-14 | 2018-05-15 | 国核电力规划设计研究院有限公司 | The start and stop shut-down system of one bulb bed modular high temperature gas cooled reactor |
| CN108278590B (en) * | 2018-03-14 | 2023-08-01 | 西安热工研究院有限公司 | System and method for shutdown cooling of high-temperature gas cooled reactor nuclear power plant |
| CN111276267A (en) * | 2020-03-23 | 2020-06-12 | 西安热工研究院有限公司 | System and method for quickly cooling primary circuit of high-temperature gas cooled reactor nuclear power station after thermal test |
| CN211788194U (en) * | 2020-03-23 | 2020-10-27 | 西安热工研究院有限公司 | High temperature gas cooled reactor nuclear power station primary circuit rapid cooling system after heat test |
-
2020
- 2020-03-23 CN CN202010208465.1A patent/CN111276267A/en active Pending
-
2021
- 2021-03-02 WO PCT/CN2021/078750 patent/WO2021190258A1/en active Application Filing
Cited By (10)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN109767852A (en) * | 2019-02-22 | 2019-05-17 | 西安热工研究院有限公司 | A kind of secondary circuit security system and its working method for reactor emergency shut-down |
| CN109767852B (en) * | 2019-02-22 | 2024-06-04 | 西安热工研究院有限公司 | Two-loop safety system for reactor emergency shutdown and working method thereof |
| WO2021190258A1 (en) * | 2020-03-23 | 2021-09-30 | 西安热工研究院有限公司 | Rapid cooling system and method for high-temperature gas cooled reactor nuclear power station primary loop after thermal test |
| CN112435765A (en) * | 2020-11-23 | 2021-03-02 | 华能山东石岛湾核电有限公司 | High-temperature gas cooled reactor steam generator small-flow cooling system and control method |
| CN113819453A (en) * | 2021-10-28 | 2021-12-21 | 华能山东石岛湾核电有限公司 | Device and method for increasing feed water temperature in starting stage of high-temperature gas cooled reactor |
| CN114220579A (en) * | 2021-10-29 | 2022-03-22 | 华能核能技术研究院有限公司 | Boosting system and method for pressure test before service of primary loop of high-temperature gas cooled reactor |
| CN114017759A (en) * | 2021-11-02 | 2022-02-08 | 华能山东石岛湾核电有限公司 | Cooling system of high-temperature gas cooled reactor nuclear power station |
| CN114220574A (en) * | 2021-11-03 | 2022-03-22 | 华能核能技术研究院有限公司 | System and method for quickly cooling steam generator of high-temperature gas cooled reactor |
| CN114220574B (en) * | 2021-11-03 | 2023-01-13 | 华能核能技术研究院有限公司 | System and method for quickly cooling steam generator of high-temperature gas cooled reactor |
| CN114165778A (en) * | 2021-11-04 | 2022-03-11 | 华能核能技术研究院有限公司 | High-temperature gas cooled reactor secondary loop system and method for improving main water supply operation temperature |
Also Published As
| Publication number | Publication date |
|---|---|
| WO2021190258A1 (en) | 2021-09-30 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN111276267A (en) | System and method for quickly cooling primary circuit of high-temperature gas cooled reactor nuclear power station after thermal test | |
| CN211788194U (en) | High temperature gas cooled reactor nuclear power station primary circuit rapid cooling system after heat test | |
| CN111189770A (en) | Supercritical carbon dioxide doped oxygen corrosion test device and method | |
| CN105945014A (en) | Steam blowing system and method for main steam pipeline of nuclear power plant | |
| JPH03124902A (en) | Combined cycle power plant and operating method therefor | |
| JP2013148438A (en) | Emergency cooling system, and nuclear reactor facility | |
| JP2007008756A (en) | Plant for generating hydrogen and electric power together | |
| CN207661753U (en) | A kind of coke dry quenching boiler waterworks | |
| CN216521613U (en) | Water supply heating system | |
| CN101955777A (en) | Method for keeping stable coke dry quenching gas circulation system after pipe explosion of heat pipe exchanger and device thereof | |
| JP2020067240A (en) | Chemical cleaning method of boiler | |
| CN212253791U (en) | Continuous sewage discharge heat recovery system of coke-oven plant coke dry quenching boiler | |
| CN107091471A (en) | Coke pressure gasification process coal gas heat energy recovering method | |
| CN107062176A (en) | Overheat of steaming Integral vertical thin tubesheet waste heat recovery plant | |
| CN207247959U (en) | Overheat of steaming Integral vertical thin tubesheet waste heat recovery plant | |
| CN112944321A (en) | System for increasing boiler feed water temperature during blow tube and method of use thereof | |
| CN112610939A (en) | Recovery method and recovery system for low-level waste heat of sand acid | |
| CN106287655A (en) | A kind of steam heat recovery technique | |
| CN215982494U (en) | Two return circuits rinse-system of high temperature gas cooled reactor | |
| CN215927489U (en) | Non-nuclear steam flushing system of high-temperature gas cooled reactor | |
| CN212392005U (en) | System for sodium-cooled fast reactor waste heat is derived | |
| KR102541651B1 (en) | Apparatus and method for producing high-pressure steam using waste heat of fuel cell stack | |
| CN215988119U (en) | Steam generator cooling system used after emergency shutdown of high-temperature gas cooled reactor | |
| CN113091042A (en) | Two return circuits rinse-system of high temperature gas cooled reactor | |
| CN213456595U (en) | Supercritical carbon dioxide doped oxygen corrosion test device |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| PB01 | Publication | ||
| PB01 | Publication | ||
| SE01 | Entry into force of request for substantive examination | ||
| SE01 | Entry into force of request for substantive examination |