CN110148480B - Nuclear power secondary circuit system - Google Patents
Nuclear power secondary circuit system Download PDFInfo
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- CN110148480B CN110148480B CN201910451886.4A CN201910451886A CN110148480B CN 110148480 B CN110148480 B CN 110148480B CN 201910451886 A CN201910451886 A CN 201910451886A CN 110148480 B CN110148480 B CN 110148480B
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- isolation valve
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- 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
- G21C15/18—Emergency cooling arrangements; Removing shut-down heat
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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)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
The invention provides a nuclear power secondary loop system which comprises a steam generator system, a main steam system and a main water supply system, wherein the steam generator system comprises a steam generator, a main steam isolation valve and a main water supply isolation valve, the steam generator is positioned in a containment, the main steam isolation valve, the main steam system, the main water supply isolation valve and the main water supply system are positioned outside the containment, a first pipeline is arranged between the steam generator and the main steam isolation valve, a second pipeline is arranged between the main steam isolation valve and the main steam system, a third pipeline is arranged between the steam generator and the main water supply isolation valve, a fourth pipeline is arranged between the main water supply isolation valve and the main water supply system, and the first pipeline and the third pipeline have the same design pressure as a primary reactor coolant system. In the nuclear power secondary loop system, when a heat transfer pipe of the steam generator breaks, radioactive substances can be prevented from leaking to the environment, and overpressure of a reactor workshop caused by mass-energy release can be prevented.
Description
Technical Field
The invention relates to the technical field of safety of nuclear power stations, in particular to a nuclear power secondary circuit full-voltage system.
Background
In a conventional pressurized water reactor secondary loop system 1, as shown in fig. 1, steam is generated from the feed water of a main feed water system 2 in a secondary loop through a secondary side of a steam generator 3, the steam is finally delivered to a steam turbine 5 to do work through a steam pipeline 4, a main steam isolation valve 7 and a main feed water isolation valve 8 are respectively arranged on a feed water pipeline 6 and the steam pipeline 4, a group of safety valves 9 are arranged in front of the main steam isolation valve 7, the design pressure of the main steam isolation valve 7, the main feed water isolation valve 8 and the pipelines therebetween is lower than that of a primary loop (reactor coolant system), if a rupture accident (SGTR) occurs in a heat transfer pipe of the steam generator 3 or the rupture accidents of the feed water pipeline 6 and the steam pipeline 4 occur, the coolant radioactive in the primary loop leaks to the secondary side of the steam generator through the ruptured heat transfer pipe to cause the pressure increase of the secondary side of the steam generator, and then the main steam isolation valve, the safety valve 9 for main steam is opened, and the steam of the two loops polluted by radioactive substances is discharged to the atmosphere through the safety valve 9, which inevitably causes the release of the radioactive substances to the environment.
Therefore, it is necessary to provide a nuclear secondary loop system to solve the existing problems.
Disclosure of Invention
The invention aims to provide a nuclear power secondary circuit system which can avoid radioactive substances from leaking to the environment under the accident of breakage of a heat transfer pipe of a steam generator.
In order to achieve the purpose, the invention provides a nuclear power secondary loop system, which comprises a steam generator system, a main steam system and a main water supply system, wherein the steam generator system comprises a steam generator, a main steam isolation valve and a main water supply isolation valve, the steam generator is positioned in a containment, the main steam isolation valve, the main steam system, the main water supply isolation valve and the main water supply system are positioned outside the containment, a first pipeline is arranged between the steam generator and the main steam isolation valve, a second pipeline is arranged between the main steam isolation valve and the main steam system, a third pipeline is arranged between the steam generator and the main water supply isolation valve, a fourth pipeline is arranged between the main water supply isolation valve and the main water supply system, and the water supply of the main water supply system flows to the main steam system after being processed by the steam generator, the first and third conduits have the same design pressure as the primary reactor coolant system.
Compared with the prior art, in the nuclear power secondary loop system, the feedwater of the main feedwater system is conveyed to the secondary side of the steam generator through the fourth pipeline and the third pipeline to generate steam, and the steam is conveyed to the steam turbine generator of the main steam system to do work through the first pipeline and the second pipeline. Wherein the first and third tubes have the same design pressure as the primary loop (reactor coolant system), i.e. a full pressure design of the steam generator system. When a rupture accident (SGTR) occurs to a heat transfer pipe of the steam generator, a primary radioactive coolant leaks to the secondary side of the steam generator through a rupture of the heat transfer pipe of the steam generator, at the moment, a main water supply isolation valve and a main steam isolation valve are isolated, radioactive substances are contained on the secondary side of the steam generator, the secondary side pressure of the steam generator can maximally reach the design pressure of a primary loop, and as the first pipeline and the third pipeline have the same design pressure with a primary reactor coolant system, the secondary side of the steam generator cannot have an overpressure condition, and the radioactive substances cannot be released into the environment. Therefore, radioactive substances are prevented from leaking to the environment, and waste of the two-loop desalting deoxygenated water is reduced. When the first pipeline or the third pipeline is broken, the main water supply isolation valve and the main steam isolation valve are isolated, and overpressure of a reactor factory caused by mass-energy release can be prevented. And because a safety valve is not arranged between the main steam isolation valve and the steam generator, the accident condition caused by the false start of the safety valve can be avoided.
Preferably, the steam generator is provided with heat transfer tubes therein, the heat transfer tubes having the same design pressure as the primary reactor coolant system.
Preferably, the primary steam isolation valve and the primary feedwater isolation valve are at the same design pressure as the primary reactor coolant system.
Preferably, the nuclear power secondary circuit system further comprises a main water supply check valve arranged in the third pipeline and located in the containment, and backflow of water is prevented.
Preferably, the nuclear power secondary loop system further comprises a first waste heat deriving system arranged outside the first pipeline and located outside the containment and a second waste heat deriving system arranged between the main water supply check valve and the steam generator, and the first waste heat deriving system and the second waste heat deriving system are used for discharging waste heat generated by a loop.
Preferably, the steam generator is a once-through steam generator.
Drawings
FIG. 1 is a schematic structural diagram of a nuclear power secondary circuit system in the prior art.
FIG. 2 is a schematic structural diagram of a nuclear power secondary circuit system according to the present invention.
Detailed Description
Embodiments of the present invention will now be described with reference to the drawings, wherein like element numerals represent like elements.
Referring to fig. 2, the nuclear power secondary circuit system 100 of the present application includes a steam generator system (not shown), a main steam system 30 and a main water supply system 50, the steam generator system includes a steam generator 10, a main steam isolation valve 20 and a main water supply isolation valve 40, the steam generator 10 is located in an in-containment K1, the main steam isolation valve 20, the main steam system 30, the main water supply isolation valve 40 and the main water supply system 50 are located outside a containment K2, a first pipeline 61 is arranged between the steam generator 10 and the main steam isolation valve 20, a second pipeline 62 is arranged between the main steam isolation valve 20 and the main steam system 30, a third pipeline 63 is arranged between the steam generator 10 and the main water supply isolation valve 40, a fourth pipeline 64 is arranged between the main water supply isolation valve 40 and the main water supply system 50, the water supply of the main water supply system 50 flows to the main steam system 30 after being processed by the steam generator 10, and the first pipeline 61 and the third pipeline 63 have the same design pressure as a primary reactor coolant system.
Further, heat transfer tubes are provided within the steam generator 10, which have the same design pressure as the primary reactor coolant system. The steam generator 10 in this embodiment is a once-through steam generator 10. Still further, the main steam isolation valve 20 and the main feedwater isolation valve 40 have the same design pressure as the primary reactor coolant system. That is, when the heat transfer tubes inside the steam generator 10 are broken, the radioactive coolant of the primary loop reactor coolant system leaks to the secondary side of the steam generator 10 through the broken parts of the heat transfer tubes of the steam generator 10, and the main steam isolation valve 20 and the main water supply isolation valve 40 are operated to be isolated, because the two loop systems and the primary loop have the same design pressure, the secondary side of the steam generator 10 cannot generate overpressure, and radioactive substances cannot be released into the environment. Furthermore, when the first, second, third or fourth piping 61, 62, 63, 64 is broken, the release of mass-produced overpressure in the reactor building is also prevented by the isolation of the main steam isolation valve 20 and the main feedwater isolation valve 40.
With continued reference to FIG. 2, the nuclear power secondary circuit system 100 further includes a main water check valve 70 disposed in the third conduit 63 and located in the containment K1 to prevent backflow of water.
Referring to fig. 2, the nuclear power secondary loop system 100 further includes a first waste heat removal system 80 disposed in the first pipe 61 and located outside the containment K2, and a second waste heat removal system 90 disposed between the primary water supply check valve 70 and the steam generator 10, where the first waste heat removal system 80 and the second waste heat removal system 90 are configured to discharge waste heat generated by a primary loop, that is, to take away the primary loop reactor waste heat.
Compared with the prior art, in the nuclear power secondary loop system 100, the feed water of the main water supply system 50 is conveyed to the secondary side of the steam generator 10 through the fourth pipeline 64 and the third pipeline 63 to generate steam, and the steam is conveyed to the steam turbine generator of the main steam system 30 through the first pipeline 61 and the second pipeline 62 to do work. Wherein the first and third conduits 61, 63 have the same design pressure as the primary circuit (reactor coolant system). When a rupture accident (SGTR) occurs in the heat transfer tubes of the steam generator 10, a primary radioactive coolant leaks to the secondary side of the steam generator 10 through the breaches of the heat transfer tubes of the steam generator 10, at this time, the main feed water isolation valve 40 and the main steam isolation valve 20 isolate the radioactive substance contained in the secondary side of the steam generator 10, the secondary pressure of the steam generator 10 can maximally reach the design pressure of the primary loop, and because the first pipeline 61 and the third pipeline 63 have the same design pressure as the primary loop, the secondary side of the steam generator 10 cannot have an overpressure condition, and the radioactive substance cannot be released into the environment. Therefore, radioactive substances are prevented from leaking to the environment, and waste of the two-loop desalting deoxygenated water is reduced. When the first pipe 61 or the third pipe 63 is broken, the main feed water isolation valve 40 and the main steam isolation valve 20 are isolated, and overpressure of the reactor building caused by mass-energy release can be prevented. And because a safety valve is not arranged between the main steam isolation valve 20 and the steam generator 10, the accident condition caused by the error starting of the safety valve can be avoided.
It should be noted that the above-mentioned embodiments illustrate rather than limit the scope of the invention, and that those skilled in the art will be able to modify the invention in its various equivalent forms after reading the present invention and to fall within the scope of the invention as defined in the appended claims.
Claims (3)
1. A nuclear power secondary loop system is characterized by comprising a steam generator system, a main steam system and a main water supply system, wherein the steam generator system comprises a steam generator, a main steam isolation valve, a main water supply isolation valve, a first waste heat deriving system and a second waste heat deriving system, the steam generator is positioned in a containment, the main steam isolation valve, the main steam system, the main water supply isolation valve and the main water supply system are positioned outside the containment, a first pipeline is arranged between the steam generator and the main steam isolation valve, a second pipeline is arranged between the main steam isolation valve and the main steam system, a third pipeline is arranged between the steam generator and the main water supply isolation valve, a fourth pipeline is arranged between the main water supply isolation valve and the main water supply system, and water supplied by the main water supply system flows to the main steam system after being processed by the steam generator, the first and third conduits have the same design pressure as the primary reactor coolant system,
the steam generator is internally provided with a heat transfer pipe which has the same design pressure with a primary loop reactor coolant system,
the main steam isolation valve and the main feed water isolation valve have the same design pressure as a primary reactor coolant system,
the first waste heat deriving system is arranged outside the first pipeline and located outside the containment, the second waste heat deriving system is arranged between the main water supply check valve and the steam generator, and the first waste heat deriving system and the second waste heat deriving system are used for discharging waste heat generated by a primary loop reactor coolant system.
2. The nuclear power secondary circuit system of claim 1 further comprising a main water feed check valve disposed in the third conduit and within the containment vessel to prevent backflow of water.
3. The nuclear power secondary loop system of claim 1 wherein the steam generator is a once-through steam generator.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
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| CN201910451886.4A CN110148480B (en) | 2019-05-28 | 2019-05-28 | Nuclear power secondary circuit system |
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| CN201910451886.4A CN110148480B (en) | 2019-05-28 | 2019-05-28 | Nuclear power secondary circuit system |
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| CN110148480A CN110148480A (en) | 2019-08-20 |
| CN110148480B true CN110148480B (en) | 2021-01-12 |
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Families Citing this family (2)
| Publication number | Priority date | Publication date | Assignee | Title |
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| CN111540487B (en) * | 2020-04-30 | 2022-03-01 | 中国核动力研究设计院 | Cooling treatment method for reactor after steam generator heat transfer pipe failure accident |
| CN115083630A (en) * | 2022-06-02 | 2022-09-20 | 中广核研究院有限公司 | Secondary side waste heat discharge system of small pressurized water reactor |
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| RU2137034C1 (en) * | 1993-12-15 | 1999-09-10 | Вискумни устав ядрович электрарни а.с. | Device for increasing opening pressure of steam generator safety valves |
| JP2010112773A (en) * | 2008-11-05 | 2010-05-20 | Hitachi-Ge Nuclear Energy Ltd | Nuclear power plant |
| CN102759098A (en) * | 2012-05-09 | 2012-10-31 | 杨子路 | Non-kinetic energy water supplying system |
| CN202855318U (en) * | 2012-09-04 | 2013-04-03 | 中科华核电技术研究院有限公司 | Passive starting cooling system for nuclear power plant and nuclear power plant system |
| CN104520941A (en) * | 2012-04-17 | 2015-04-15 | 巴布科克和威尔科克斯M能量股份有限公司 | Auxiliary condenser system for decay heat removal in a nuclear reactor system |
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| Publication number | Priority date | Publication date | Assignee | Title |
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| US5309487A (en) * | 1992-06-24 | 1994-05-03 | Westinghouse Electric Corp. | Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems |
| US9206978B2 (en) * | 2012-06-13 | 2015-12-08 | Westinghouse Electric Company Llc | Pressurized water reactor compact steam generator |
| CN204229849U (en) * | 2014-11-18 | 2015-03-25 | 上海核工程研究设计院 | The non-active emergency feedwater supply system of a kind of nuclear power station |
| CN105070329A (en) * | 2015-08-31 | 2015-11-18 | 上海核工程研究设计院 | Nuclear power station secondary side passive residual heat removal system |
| CN107464590A (en) * | 2017-08-23 | 2017-12-12 | 中国船舶重工集团公司第七〇九研究所 | Marine PWR Passive residual heat removal system |
| 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 |
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Patent Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| RU2137034C1 (en) * | 1993-12-15 | 1999-09-10 | Вискумни устав ядрович электрарни а.с. | Device for increasing opening pressure of steam generator safety valves |
| JP2010112773A (en) * | 2008-11-05 | 2010-05-20 | Hitachi-Ge Nuclear Energy Ltd | Nuclear power plant |
| CN104520941A (en) * | 2012-04-17 | 2015-04-15 | 巴布科克和威尔科克斯M能量股份有限公司 | Auxiliary condenser system for decay heat removal in a nuclear reactor system |
| CN102759098A (en) * | 2012-05-09 | 2012-10-31 | 杨子路 | Non-kinetic energy water supplying system |
| CN202855318U (en) * | 2012-09-04 | 2013-04-03 | 中科华核电技术研究院有限公司 | Passive starting cooling system for nuclear power plant and nuclear power plant system |
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