CN112542257A - Nuclear power station pressure relief system - Google Patents
Nuclear power station pressure relief system Download PDFInfo
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- CN112542257A CN112542257A CN201910894934.7A CN201910894934A CN112542257A CN 112542257 A CN112542257 A CN 112542257A CN 201910894934 A CN201910894934 A CN 201910894934A CN 112542257 A CN112542257 A CN 112542257A
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- Prior art keywords
- valve
- pressure relief
- pressure
- pipeline
- nuclear power
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21D—NUCLEAR POWER PLANT
- G21D3/00—Control of nuclear power plant
- G21D3/04—Safety arrangements
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C13/00—Pressure vessels; Containment vessels; Containment in general
- G21C13/02—Details
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21D—NUCLEAR POWER PLANT
- G21D1/00—Details of nuclear power plant
- G21D1/006—Details of nuclear power plant primary side of steam generators
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- 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
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- 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
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- Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Business, Economics & Management (AREA)
- Emergency Management (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
The invention provides a nuclear power station pressure relief system, which comprises: the steam generator comprises a pressure vessel, N steam generators, a built-in refueling water tank and a pressure stabilizer. The pressure vessel is connected with the steam generator through a heat pipe section, one end of the pressure stabilizer is connected with the pressure vessel through a fluctuation pipe, and the other end of the pressure stabilizer is communicated with the built-in refueling water tank through a pressure relief valve and a pipeline. According to the embodiment of the invention, high-temperature and high-pressure fluid in an overpressure accident is directly discharged into the internal replacement material water tank for condensation through the pressure stabilizer, the pressure relief valve and the corresponding pipeline, so that the problems of limited pressure bearing capacity and low safety of a pressure relief tank in the pressure relief device of the pressure stabilizer in the prior art are solved.
Description
Technical Field
The invention relates to the technical field of safety control of nuclear power stations, in particular to a pressure relief system of a nuclear power station.
Background
At present, in a pressurized water reactor nuclear power station, a pressure relief system is required to realize rapid condensation and collection of fluid under an accident, radioactive fission products are retained in water, radioactive release to the environment is reduced, and the pressure in a containment vessel is reduced.
The pressure relief system in the prior art generally comprises a pressure stabilizer pressure relief device and a steam generator pressure relief blowdown subsystem, wherein the pressure stabilizer pressure relief device is generally positioned behind a pressure stabilizer discharge pipeline, a pressure relief tank is connected with the pressure stabilizer through a bubbling pipe so as to regulate the pressure in a loop, and the pressure relief tank is a closed container with the lower half part filled with water and the upper half part filled with nitrogen. When the safety valve of the pressure stabilizer or the pressure relief valve in a serious accident is opened, the fluid is discharged from the pressure relief tank through the bubbling pipe through the discharge pipeline and is cooled and condensed after encountering the water in the pressure relief tank. When the absolute pressure in the pressure relief box is larger than a certain value, the safety blasting disc in the pressure relief box is blasted to directly discharge fluid into the containment, so that the atmospheric radioactivity level in the containment is increased, and further the radioactivity release to the external environment is increased. Therefore, in the prior art, the pressure relief device of the voltage stabilizer has the problem of low safety due to the limited pressure bearing capacity of the pressure relief box.
Disclosure of Invention
The embodiment of the invention provides a pressure relief system of a nuclear power station, which aims to solve the problem that a pressure relief device of a voltage stabilizer is low in safety due to the fact that the pressure bearing capacity of a pressure relief box is limited in the prior art.
An embodiment of the present invention provides a pressure relief system for a nuclear power station, including:
including pressure vessel, N steam generator, built-in reloading water tank and stabiliser: the pressure vessel comprises N cold pipe sections and N hot pipe sections;
the N cold pipe sections, the N hot pipe sections, the N steam generators, and N main pumps form a loop having N loops; in a loop, a hot pipe port of the pressure vessel is connected with a primary side inlet of the steam generator through the hot pipe section, a primary side outlet of the steam generator is connected with a main pump inlet through a transition pipe section, and a main pump outlet is connected with a cold pipe port of the pressure vessel through the cold pipe section;
an outlet of a blowdown system at the secondary side of the steam generator is connected with a first end of a first pipeline through a blowdown pipeline and a first switch valve, a second end of the first pipeline is an open end, and the second end of the first pipeline is arranged in the built-in refueling water tank;
one end of the pressure stabilizer is connected with the pressure container through a fluctuation pipe, and the other end of the pressure stabilizer is connected with the first pipeline through a pressure relief valve;
wherein N is a positive integer.
Optionally, the pressure relief valve includes at least three safety valves arranged in parallel, where the first threshold values of different safety valves are different; when the safety valve is larger than or equal to the corresponding first threshold, the safety valve is in an opening state; when the safety valve is smaller than the corresponding first threshold, the safety valve is in a closed state.
Optionally, the pressure relief valve further comprises a sub-valve and a controller, wherein the sub-valve is connected in parallel with the safety valve; the controller is electrically connected with the sub-valve and used for controlling the sub-valve to be opened, the sub-valve is in a closed state under the condition that the sub-valve is smaller than a second threshold value, and the second threshold value is smaller than a first threshold value.
Optionally, the pressure relief system of the nuclear power station further includes a second pipeline, one end of the second pipeline is connected to the first pipeline, and the other end of the second pipeline is an open end and is suspended in the air; and a second switch valve is arranged on the second pipeline.
Optionally, the controller is electrically connected with the second switch valve; the controller is used for controlling the second switch valve to be opened after the sub-valve is opened for a first preset time; and after the sub-valve is closed for a second preset time, controlling the second switch valve to be closed.
Optionally, a check valve is disposed on the first pipeline.
Optionally, the number of check valves is at least two.
Optionally, the internal replacement water tank is an accommodating tank with an opening at the top end.
Optionally, a gas containing box is arranged in the built-in refueling water tank, the gas containing box is fixedly connected with the inner wall of the built-in refueling water tank, and a gas inlet is formed in one side, facing the bottom end of the built-in refueling water tank, of the gas containing box; the second end of the first pipeline passes through the air inlet and is positioned in the gas containing box.
Optionally, a spray head is arranged at the second end of the first pipeline, and the spray head is located inside the gas containing box.
According to the embodiment of the invention, one end of the voltage stabilizer is connected with the pressure container through the fluctuation pipe, the other end of the voltage stabilizer is connected with the built-in material changing water tank through the pressure relief valve, fluid in the voltage stabilizer is directly discharged into the internal replacement material water tank through the pressure relief valve when an accident occurs, and high-temperature and high-pressure fluid is rapidly condensed in the internal replacement material water tank, so that a radioactive fission product is retained in water, and therefore, the problems of limited pressure bearing capacity and low safety of the pressure relief tank in the voltage stabilizer pressure relief device in the prior art are solved.
Drawings
In order to more clearly illustrate the technical solutions of the embodiments of the present invention, the drawings required to be used in the description of the embodiments of the present invention will be briefly introduced below, and it is obvious that the drawings in the following description are only some embodiments of the present invention, and it is obvious for those skilled in the art to obtain other drawings based on these drawings without inventive exercise.
FIG. 1 is a schematic view of a nuclear power plant pressure relief system of the present invention.
Detailed Description
The technical solutions in the embodiments of the present invention will be clearly and completely described below with reference to the drawings in the embodiments of the present invention, and it is obvious that the described embodiments are some, not all, embodiments of the present invention. All other embodiments, which can be derived by a person skilled in the art from the embodiments given herein without making any creative effort, shall fall within the protection scope of the present invention.
Unless defined otherwise, technical or scientific terms used herein shall have the ordinary meaning as understood by one of ordinary skill in the art to which this invention belongs. The use of "first," "second," and similar terms in the present application do not denote any order, quantity, or importance, but rather the terms are used to distinguish one element from another. Also, the use of the terms "a" or "an" and the like do not denote a limitation of quantity, but rather denote the presence of at least one. The terms "connected" or "coupled" and the like are not restricted to physical or mechanical connections, but may include electrical connections, whether direct or indirect. "upper", "lower", "left", "right", and the like are used merely to indicate relative positional relationships, and when the absolute position of the object being described is changed, the relative positional relationships are changed accordingly.
Referring to fig. 1, an embodiment of the present invention provides a pressure relief system for a nuclear power plant, including: pressure vessel 10, N steam generators 20, built-in refueling water tank 30 and pressurizer 40: the pressure vessel 10 comprises N cold pipe sections 51 and N hot pipe sections 52; the N cold pipe sections 51, the N hot pipe sections 52, the N steam generators 20, and the N main pumps form a loop having N loops; in a loop, the hot pipe orifice 12 of the pressure vessel 10 is connected with the primary side inlet 21 of the steam generator 20 through the hot pipe section 52, the primary side outlet 23 of the steam generator 20 is connected with the primary pump inlet through the transition pipe section 54, and the primary pump outlet is connected with the cold pipe orifice 11 of the pressure vessel 10 through the cold pipe section 51; a blowdown system outlet 22 at the secondary side of the steam generator 20 is connected with a first end of a first pipeline 53 through a blowdown pipeline and a first switch valve 61, a second end of the first pipeline 53 is an open end, and the second end of the first pipeline 53 is arranged in the built-in refueling water tank 30; one end of the pressure stabilizer 40 is connected with the pressure vessel 10 through a surge pipe, and the other end of the pressure stabilizer 40 is connected with the first pipeline 53 through a pressure relief valve 62; wherein N is a positive integer.
The structure of the pressure vessel 10 may be set according to actual requirements, and in the embodiment of the present invention, the N cold nozzles 11 are uniformly distributed on the same circumference of the pressure vessel 10; the N thermal nozzles 12 are uniformly distributed on the same circumference of the pressure vessel 10.
Furthermore, in order to improve the safety of the system and prevent the liquid in the pipeline from leaking, the cold pipe section 51 and the cold pipe orifice 11 are connected in a sealing manner. Accordingly, the hot pipe nozzle 12, the primary side inlet 21 of the steam generator 20, the secondary side blowdown system outlet 22 of the steam generator 20, and the primary side outlet 23 of the steam generator 20 are all connected to corresponding pipes in a sealing manner.
Specifically, as shown in fig. 1, an end a of the cold pipe section 51 is connected to an outlet of a main pump in a primary circuit of the nuclear power plant; the B-end of the transition section 54 is connected to the inlet of the main pump in a primary circuit of the nuclear power plant.
Further, the shape of the pressure stabilizer 40 may be set according to actual needs, and in the embodiment of the present invention, in order to better control the pressure in the pipeline and the pressure container 10, the pressure stabilizer 40 may be a cylindrical container with openings at both ends for accommodating a part of the high-temperature and high-pressure fluid in the pipeline.
When the nuclear power plant is in normal operation, the pressure relief system of the nuclear power plant is in a usable state, and when an overpressure accident occurs, the high-temperature and high-pressure fluid in the pressurizer 40 enters the large-capacity built-in refueling water tank 30 through the pressure relief valve 62 and the first pipeline 53 to be cooled and condensed, so that the radioactive fission product is retained in water.
Wherein, because the high-temperature high-pressure fluid may have a certain radioactivity, in order to ensure that the water quality is maintained in a good state, the built-in refueling water tank 30 may be provided with a corresponding deionization device; in order to accelerate the condensation speed of the high-temperature and high-pressure fluid, a cooling device can be arranged in the built-in refueling water tank 30.
In the embodiment of the invention, one end of the pressure stabilizer 40 is connected with the pressure container 10 through a pipeline, and the other end of the pressure stabilizer is connected with the built-in refueling water tank 30 through the pressure relief valve 62, so that the fluid in the pressure stabilizer 40 is directly discharged into the built-in refueling water tank 30 through the pressure relief valve 62 when an accident occurs. The high-temperature high-pressure fluid is rapidly condensed in the built-in refueling water tank 30, so that the radioactive fission product is retained in water, and the problems of limited pressure bearing capacity and low safety of a pressure relief tank in the pressure relief device of the voltage stabilizer in the prior art are solved.
It should be noted that the first switch valve 61 may be connected to a controller, and when a Steam Generator Tube Rupture (SGTR) accident occurs during operation of the nuclear power plant and the water level in the damaged steam generator 20 is too high, the first switch valve 61 may be controlled to open, and the secondary side fluid of the damaged steam generator 20 may directly enter the built-in refueling water tank 30 through the first pipe 53; when the water level of the damaged steam generator 20 is reduced to a certain degree, the first switch valve 61 can be controlled to be closed, so that blowdown and pressure relief of the steam generator 20 are realized, the problem that fluid in the damaged steam generator 20 enters other intact steam generators 20 through a transfer pipeline in the prior art is avoided, the radioactivity of the intact steam generators 20 is increased, the pollution problem is caused, the blowdown structural design of the steam generator 20 is simplified, the arrangement difficulty of the system is reduced, and the reliability of blowdown and pressure relief functions under the accident is improved under the condition of low cost.
Specifically, the pressure relief valve 62 may include at least three safety valves 621 arranged in parallel, where the first threshold values of different safety valves 621 are different; when the safety valve 621 is greater than or equal to a corresponding first threshold, the safety valve 621 is in an open state; when the safety valve 621 is smaller than the corresponding first threshold, the safety valve 621 is in a closed state.
The first threshold of the safety valve 621 is the pressure in the pipeline connected with the safety valve 621, when an overpressure accident of the DBC-3 or DBC-4 primary circuit occurs, the safety valve 621 opens to release pressure when the pressure reaches the first threshold, high-temperature and high-pressure fluid enters the built-in refueling water tank 30 through the first pipeline 53 to be cooled and condensed, and when the pressure falls below the first threshold, the safety valve 621 closes.
Further, the pressure relief valve 62 further includes a sub-valve 622 connected in parallel with the safety valve 621, and a controller for controlling the opening of the sub-valve 622, wherein when the pressure is smaller than a second threshold, the sub-valve 622 is closed, and the second threshold is smaller than the first threshold.
The pressure relief system of the nuclear power station further comprises a second pipeline 55, wherein one end of the second pipeline 55 is connected with the first pipeline 53, and the other end of the second pipeline 55 is an open end and is arranged in a suspended manner; a second switch valve 63 is disposed on the second pipeline 55, the controller is electrically connected to the second switch valve 63, and when the sub-valve 622 is opened for a first preset time period, the second switch valve 63 is controlled to be opened; after the sub-valve 622 is closed for a second preset time period, the second on-off valve 63 is controlled to be closed.
The connecting mode between the pipelines can be welding, and the connecting mode between the pipelines and the valve is sealing connection.
The controller for controlling the opening and closing of the first on-off valve 61 is different from the controller for controlling the safety valve 621 and the sub-valve 622.
When a serious accident occurs in the nuclear power plant and the pressure of a primary circuit needs to be rapidly reduced, the sub-valve 622 is controlled to be opened by a controller, after a first preset time, the second switch valve 63 is controlled to be opened, high-temperature and high-pressure fluid enters the second pipeline 55 through the first pipeline 53 and is finally discharged into the atmosphere in the containment vessel, and the pressure inside the pressure vessel 10 is rapidly reduced; when the pressure drops below the second threshold, the sub-valve 622 is closed, and the second on-off valve 63 is closed after a second predetermined time.
Further, in order to prevent the high-temperature and high-pressure fluid flowing through the first pipe line 53 from flowing backward, the first pipe line 53 is provided with check valves 64, and the number of the check valves 64 is at least two in order to prevent the check function from being disabled due to the failure of the check valves 64.
Specifically, a joint between the safety valve 621 and the first pipe line 53 is a first joint, and a joint between the first pipe line 53 and the second pipe line 55 is a second joint. The second connection is between the first connection and the check valve 64.
The structure of the built-in refueling water tank 30 can be set according to actual needs, in the embodiment of the present invention, the built-in refueling water tank 30 is a containing tank with an open top end, wherein a gas containing tank 31 is arranged in the built-in refueling water tank 30, and the gas containing tank 31 and the outer wall of the built-in refueling water tank 30 can be fixedly connected through a connecting rod.
Wherein, an air inlet is provided at one side of the air accommodating tank 31 facing the bottom end of the built-in refueling water tank 30, a second end of the first pipeline 53 passes through the air inlet and is positioned in the air accommodating tank 31, a spray head 531 is provided at the second end of the first pipeline 53, and the spray head 531 is positioned in the air accommodating tank 31.
The gas storage tank 31 is normally entirely submerged below the liquid level of the built-in refueling water tank 30 and filled with water, and when high-temperature and high-pressure fluid is discharged, the gas storage tank 31 stores therein water-insoluble gas.
The above description is only for the specific embodiments of the present invention, but the scope of the present invention is not limited thereto, and any person skilled in the art can easily conceive of the changes or substitutions within the technical scope of the present invention, and the changes or substitutions should be covered within the scope of the present invention. Therefore, the protection scope of the present invention shall be subject to the protection scope of the claims.
Claims (10)
1. The utility model provides a nuclear power station release system which characterized in that, includes pressure vessel, N steam generator, built-in reloading water tank and stabiliser: the pressure vessel comprises N cold pipe sections and N hot pipe sections;
the N cold pipe sections, the N hot pipe sections, the N steam generators, and N main pumps form a loop having N loops; in a loop, a hot pipe port of the pressure vessel is connected with a primary side inlet of the steam generator through the hot pipe section, a primary side outlet of the steam generator is connected with a main pump inlet through a transition pipe section, and a main pump outlet is connected with a cold pipe port of the pressure vessel through the cold pipe section;
an outlet of a blowdown system at the secondary side of the steam generator is connected with a first end of a first pipeline through a blowdown pipeline and a first switch valve, a second end of the first pipeline is an open end, and the second end of the first pipeline is arranged in the built-in refueling water tank;
one end of the pressure stabilizer is connected with the pressure container through a fluctuation pipe, and the other end of the pressure stabilizer is connected with the first pipeline through a pressure relief valve;
wherein N is a positive integer.
2. The nuclear power plant pressure relief system of claim 1, wherein the pressure relief valve comprises at least three safety valves arranged in parallel, wherein the first threshold values of different safety valves are different; when the safety valve is larger than or equal to the corresponding first threshold, the safety valve is in an opening state; when the safety valve is smaller than the corresponding first threshold, the safety valve is in a closed state.
3. The nuclear power plant pressure relief system of claim 2, wherein the pressure relief valve further comprises a sub-valve in parallel with the safety valve and a controller;
the controller is electrically connected with the sub-valve and used for controlling the sub-valve to be opened, the sub-valve is in a closed state under the condition that the sub-valve is smaller than a second threshold value, and the second threshold value is smaller than a first threshold value.
4. The nuclear power plant pressure relief system according to claim 3, further comprising a second pipeline, one end of which is connected to the first pipeline and the other end of which is open and suspended; and a second switch valve is arranged on the second pipeline.
5. The nuclear power plant pressure relief system according to claim 4, wherein the controller is electrically connected to the second switching valve;
the controller is used for controlling the second switch valve to be opened after the sub-valve is opened for a first preset time; and after the sub-valve is closed for a second preset time, controlling the second switch valve to be closed.
6. The nuclear power plant pressure relief system of claim 1, wherein the first pipeline has a check valve disposed thereon.
7. The nuclear power plant pressure relief system of claim 6, wherein the number of check valves is at least two.
8. The nuclear power plant pressure relief system according to claim 1, wherein the inner replacement waterbox is an open-topped containment box.
9. The nuclear power plant pressure relief system according to claim 8, wherein a gas receiving tank is arranged in the built-in refueling water tank, the gas receiving tank is fixedly connected with the inner wall of the built-in refueling water tank, and a gas inlet is formed in one side of the gas receiving tank, which faces the bottom end of the built-in refueling water tank; the second end of the first pipeline passes through the air inlet and is positioned in the gas containing box.
10. The nuclear power plant pressure relief system according to claim 9, wherein the second end of the first conduit is provided with a spray head located inside the gas containment tank.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201910894934.7A CN112542257A (en) | 2019-09-20 | 2019-09-20 | Nuclear power station pressure relief system |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201910894934.7A CN112542257A (en) | 2019-09-20 | 2019-09-20 | Nuclear power station pressure relief system |
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| Publication Number | Publication Date |
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| CN112542257A true CN112542257A (en) | 2021-03-23 |
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| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN201910894934.7A Pending CN112542257A (en) | 2019-09-20 | 2019-09-20 | Nuclear power station pressure relief system |
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| CN (1) | CN112542257A (en) |
Cited By (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN114999693A (en) * | 2022-06-01 | 2022-09-02 | 中国核动力研究设计院 | Pressure relief protection system for preventing non-condensable gas from entering reactor core of compressed gas pressure stabilizing reactor |
| CN115430534A (en) * | 2022-07-28 | 2022-12-06 | 中国核电工程有限公司 | Rotary injection device for accident pressure relief discharge |
-
2019
- 2019-09-20 CN CN201910894934.7A patent/CN112542257A/en active Pending
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN114999693A (en) * | 2022-06-01 | 2022-09-02 | 中国核动力研究设计院 | Pressure relief protection system for preventing non-condensable gas from entering reactor core of compressed gas pressure stabilizing reactor |
| CN114999693B (en) * | 2022-06-01 | 2024-05-28 | 中国核动力研究设计院 | Pressure relief protection system for preventing noncondensable gas from entering reactor core of compressed gas stabilized pressure reactor |
| CN115430534A (en) * | 2022-07-28 | 2022-12-06 | 中国核电工程有限公司 | Rotary injection device for accident pressure relief discharge |
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