CN110148480A - A kind of nuclear power secondary coolant circuit system - Google Patents
A kind of nuclear power secondary coolant circuit system Download PDFInfo
- Publication number
- CN110148480A CN110148480A CN201910451886.4A CN201910451886A CN110148480A CN 110148480 A CN110148480 A CN 110148480A CN 201910451886 A CN201910451886 A CN 201910451886A CN 110148480 A CN110148480 A CN 110148480A
- Authority
- CN
- China
- Prior art keywords
- main
- steam
- steam generator
- main feed
- valve
- 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.)
- Granted
Links
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
- G21C15/18—Emergency cooling arrangements; Removing shut-down heat
-
- 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
- Physics & Mathematics (AREA)
- 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 present invention provides a kind of nuclear power secondary coolant circuit systems, including steam generator system, main steam system, and main feed system, steam generator system includes steam generator, main steam isolation valve and main feed water isolating valve, steam generator is located in containment, main steam isolation valve, main steam system, main feed water isolating valve and main feed system are located at outside containment, first pipe is equipped between steam generator and main steam isolation valve, second pipe is equipped between main steam isolation valve and main steam system, third pipeline is equipped between steam generator and main feed water isolating valve, the 4th pipeline is equipped between main feed water isolating valve and main feed system, first pipe and third pipeline and primary Ioops reactor coolant loop design pressure having the same.In the nuclear power secondary coolant circuit system, when fracture accident occurs in the heat-transfer pipe of steam generator, it is avoided that environment and reveals radioactive substance, moreover it is possible to prevent reactor building superpressure caused by Mass and energy release.
Description
Technical field
The present invention relates to the security technology areas of nuclear power station, more particularly relate to a kind of nuclear power secondary circuit total pressure system.
Background technique
Conventional pressurized water heap secondary coolant circuit system 1 is as shown in Figure 1, the water supply of main feed system 2 is by steam generation in secondary circuit
3 secondary side of device generates steam, and steam is finally delivered to steam turbine 5 through jet chimney 4 and does manual work, on feedwater piping 6 and jet chimney 4
Main steam isolation valve 7 and main feed water isolating valve 8 is respectively set, one group of safety valve 9 is set before main steam isolation valve 7, main steam every
From valve 7 and main feed water isolating valve 8 and its between pipeline design pressure be lower than primary Ioops (reactor coolant loop) design
Pressure, if the thing that the accident (SGTR) ruptured occurs in 3 heat-transfer pipe of steam generator or feedwater piping 6, jet chimney 4 rupture
Therefore the radioactive coolant of primary Ioops leaks to steam generator secondary side by the heat-transfer pipe ruptured, causes steam generator
The pressure rise of secondary side, subsequent main steam isolation valve 7 and main feed water isolating valve 8 are isolated, and main steam safety valve 9 is opened,
The steam of secondary circuit contaminated by radioactive substances inevitably causes radioactive substance to environment through 9 row Xiang great Qi of safety valve
Release.
Therefore, it is necessary to which it is existing to solve the problems, such as to provide a kind of nuclear power secondary coolant circuit system.
Summary of the invention
The purpose of the present invention is to provide a kind of nuclear power secondary coolant circuit systems, under steam generator heat-transfer pipe fracture accident,
It is avoided that environment and reveals radioactive substance.
To achieve the above object, the present invention provides a kind of nuclear power secondary coolant circuit systems, including steam generator system, main steaming
Vapour system and main feed system, the steam generator system include steam generator, main steam isolation valve and master to water segregation
Valve, the steam generator are located in containment, and the main steam isolation valve, main steam system, main feed water isolating valve and master give
Water system is located at outside containment, and first pipe, the main steaming are equipped between the steam generator and the main steam isolation valve
Second pipe is equipped between vapour isolating valve and the main steam system, between the steam generator and the main feed water isolating valve
Equipped with third pipeline, the 4th pipeline, the main feed system are equipped between the main feed water isolating valve and the main feed system
Water supply through the steam generator processing after flow to the main steam system, the first pipe and the third pipeline and one
Circuit reactor coolant loop design pressure having the same.
Compared with prior art, in the nuclear power secondary coolant circuit system of the application, the water supply of main feed system through the 4th pipeline and
Third pipeline to steam generator secondary side generates steam, and steam is delivered to main steam system through first pipe and second pipe
The steam turbine generator of system does work.Wherein, first pipe and third pipeline have identical with primary Ioops (reactor coolant loop)
Design pressure, i.e., steam generator system total pressure design.When there is fracture accident (SGTR) in the heat-transfer pipe of steam generator,
Steam generator secondary side is leaked at the cut that primary Ioops radioactivity coolant passes through steam generator heat-transfer pipe, master gives at this time
Water isolation valve and main steam isolation valve are isolated, and radioactive substance is inclusive in steam generator secondary side, steam generation
Device secondary pressure maximum can reach the design pressure of primary Ioops, since first pipe and the third pipeline are reacted with primary Ioops
Reactor coolant system design pressure having the same, there is no superpressure situations for steam generator secondary side, and radioactive substance is not
It can be discharged into environment.Therefore, it avoids revealing radioactive substance to environment, reduces the waste of secondary circuit desalination deaerated water.When
When first pipe or third pipeline breaking, main feed water isolating valve and main steam isolation valve are isolated, and can prevent Mass and energy release from leading
The reactor building superpressure of cause.Due also to being not provided with safety valve between main steam isolation valve and steam generator, safety can avoid
Accident conditions caused by valve error starting.
Preferably, being equipped with heat-transfer pipe, the heat-transfer pipe and primary Ioops reactor coolant loop in the steam generator
Design pressure having the same.
Preferably, the main steam isolation valve and the main feed water isolating valve have with primary Ioops reactor coolant loop
Identical design pressure.
Preferably, nuclear power secondary coolant circuit system further include be arranged in the third pipeline and be located at containment in main water supply stop
Valve is returned, the reflux of water is prevented.
Preferably, the nuclear power secondary coolant circuit system of the application further includes that the first pipe is arranged in and is located at outside containment
First waste heat guiding system and the second waste heat being arranged between the main feed check (valve) and steam generator export system
System, the first waste heat guiding system and the second waste heat guiding system are used for the Residual heat removal for generating primary Ioops.
Preferably, the steam generator is once through steam generator.
Detailed description of the invention
Fig. 1 is the structural schematic diagram of nuclear power secondary coolant circuit system in the prior art.
Fig. 2 is the structural schematic diagram of nuclear power secondary coolant circuit system of the present invention.
Specific embodiment
The embodiment of the present invention described referring now to the drawings, similar element numbers represent similar element in attached drawing.
Referring to FIG. 1, the nuclear power secondary coolant circuit system 100 of the application, including steam generator system (not shown), main steam
System 30 and main feed system 50, steam generator system include steam generator 10, main steam isolation valve 20 and main water supply every
From valve 40, steam generator 10 is located at K1 in containment, main steam isolation valve 20, main steam system 30, main feed water isolating valve 40
It is located at K2 outside containment with main feed system 50, first pipe 61 is equipped between steam generator 10 and main steam isolation valve 20,
Be equipped with second pipe 62 between main steam isolation valve 20 and main steam system 30, steam generator 10 and main feed water isolating valve 40 it
Between be equipped with third pipeline 63, between main feed water isolating valve 40 and main feed system 50 be equipped with the 4th pipeline 64, main feed system 50
Water supply through steam generator 10 processing after flow to main steam system 30, first pipe 61 and third pipeline 63 are reacted with primary Ioops
Reactor coolant system design pressure having the same.
Further, heat-transfer pipe is equipped in steam generator 10, heat-transfer pipe has phase with primary Ioops reactor coolant loop
Same design pressure.Steam generator 10 is once through steam generator 10 in this implementation.Further, 20 He of main steam isolation valve
Main feed water isolating valve 40 and primary Ioops reactor coolant loop design pressure having the same.That is work as steam generator
Heat transfer tracheal rupture in 10, primary Ioops reactor coolant loop radioactivity coolant pass through the broken of 10 heat-transfer pipe of steam generator
Be leaked to 10 secondary side of steam generator at mouthful, at the same main steam isolation valve 20 and main feed water isolating valve 40 by operation carry out every
From, due to secondary coolant circuit system and primary Ioops design pressure having the same, 10 secondary side of steam generator will not superpressure, radioactivity
Substance will not can be discharged into environment.In addition, when first pipe 61, second pipe 62, third pipeline 63 or the rupture of the 4th pipeline 64
When, by the isolation of main steam isolation valve 20 and main feed water isolating valve 40, it can also prevent reactor building caused by Mass and energy release
Superpressure.
With continued reference to FIG. 1, nuclear power secondary coolant circuit system 100 further includes setting in third pipeline 63 and is located at K1 in containment
Main feed check (valve) 70, prevent the reflux of water.
With continued reference to FIG. 1, the nuclear power secondary coolant circuit system 100 of the application further includes setting in first pipe 61 and is located at peace
The first waste heat guiding system 80 of the outer K2 of full shell and be arranged between main feed check (valve) 70 and steam generator 10 more than second
The waste heat that hot guiding system 90, the first waste heat guiding system 80 and the second waste heat guiding system 90 are used to generate primary Ioops is arranged
Out, primary Ioops reactor waste can be taken out of.
Compared with prior art, in the nuclear power secondary coolant circuit system 100 of the application, the water supply of main feed system 50 is through the 4th pipe
Road 64 and third pipeline 63 are delivered to 10 secondary side of steam generator and generate steam, and steam is through first pipe 61 and second pipe 62
It is delivered to the steam turbine generator acting of main steam system 30.Wherein, first pipe 61 and third pipeline 63 and primary Ioops (reactor
Coolant system) design pressure having the same.When there is fracture accident (SGTR) in the heat-transfer pipe of steam generator 10, one time
Road radioactivity coolant is leaked to 10 secondary side of steam generator at the cut by 10 heat-transfer pipe of steam generator, and master gives at this time
Water isolation valve 40 and main steam isolation valve 20 are isolated, and radioactive substance is inclusive in 10 secondary side of steam generator, are steamed
10 secondary pressure maximum of vapour generator can reach the design pressure of primary Ioops, due to first pipe 61 and the third pipeline 63
With primary Ioops design pressure having the same, for 10 secondary side of steam generator there is no superpressure situation, radioactive substance will not
It can be discharged into environment.Therefore, it avoids revealing radioactive substance to environment, reduces the waste of secondary circuit desalination deaerated water.When
When one pipeline 61 or third pipeline 63 rupture, main feed water isolating valve 40 and main steam isolation valve 20 are isolated, and mass-energy can be prevented
Reactor building superpressure caused by discharging.Due also to it is not provided with safety valve between main steam isolation valve 20 and steam generator 10,
It can avoid accident conditions caused by safety valve error starting.
It should be pointed out that the above specific embodiment is only illustrative of the invention and is not intended to limit the scope of the invention,
After having read the present invention, those skilled in the art each fall within power appended by the application to the modification of various equivalent forms of the invention
Benefit requires the range limited.
Claims (6)
1. a kind of nuclear power secondary coolant circuit system, which is characterized in that including steam generator system, main steam system and master to water system
System, the steam generator system includes steam generator, main steam isolation valve, main feed water isolating valve, the steam generator
In containment, the main steam isolation valve, main steam system, main feed water isolating valve and main feed system are located at containment
Outside, first pipe, the main steam isolation valve and the master are equipped between the steam generator and the main steam isolation valve
It is equipped with second pipe between vapour system, third pipeline, institute are equipped between the steam generator and the main feed water isolating valve
It states and is equipped with the 4th pipeline between main feed water isolating valve and the main feed system, the water supply of the main feed system is through the steam
The main steam system, the first pipe and the third pipeline and primary Ioops reactor coolant are flowed to after generator processing
System design pressure having the same.
2. nuclear power secondary coolant circuit system according to claim 1, which is characterized in that be equipped with heat transfer in the steam generator
Pipe, the heat-transfer pipe and primary Ioops reactor coolant loop design pressure having the same.
3. nuclear power secondary coolant circuit system according to claim 1, which is characterized in that the main steam isolation valve and the master give
Water isolation valve and primary Ioops reactor coolant loop design pressure having the same.
4. nuclear power secondary coolant circuit system according to claim 1, which is characterized in that further include setting in the third pipeline and
Main feed check (valve) in containment prevents the reflux of water.
5. nuclear power secondary coolant circuit system according to claim 1, which is characterized in that further include setting in the first pipe and
The first waste heat guiding system outside containment and it is arranged between the main feed check (valve) and the steam generator
Second waste heat guiding system, the first waste heat guiding system and the second waste heat guiding system are used for primary Ioops reactor
The Residual heat removal that coolant system generates.
6. nuclear power secondary coolant circuit system according to claim 1, which is characterized in that the steam generator is direct-flow steam hair
Raw device.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201910451886.4A CN110148480B (en) | 2019-05-28 | 2019-05-28 | Nuclear power secondary circuit system |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201910451886.4A CN110148480B (en) | 2019-05-28 | 2019-05-28 | Nuclear power secondary circuit system |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN110148480A true CN110148480A (en) | 2019-08-20 |
| CN110148480B CN110148480B (en) | 2021-01-12 |
Family
ID=67593351
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN201910451886.4A Active CN110148480B (en) | 2019-05-28 | 2019-05-28 | Nuclear power secondary circuit system |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN110148480B (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN111540487A (en) * | 2020-04-30 | 2020-08-14 | 中国核动力研究设计院 | Cooling treatment method for reactor after steam generator heat transfer pipe failure accident |
Citations (11)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| 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 |
| 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 |
| CN104321825A (en) * | 2012-06-13 | 2015-01-28 | 西屋电气有限责任公司 | 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 |
| 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 |
| 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 |
| 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 |
-
2019
- 2019-05-28 CN CN201910451886.4A patent/CN110148480B/en active Active
Patent Citations (11)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| 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 |
| 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 |
| CN104321825A (en) * | 2012-06-13 | 2015-01-28 | 西屋电气有限责任公司 | Pressurized water reactor compact steam generator |
| 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 |
| 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 |
| 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 |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN111540487A (en) * | 2020-04-30 | 2020-08-14 | 中国核动力研究设计院 | Cooling treatment method for reactor after steam generator heat transfer pipe failure accident |
Also Published As
| Publication number | Publication date |
|---|---|
| CN110148480B (en) | 2021-01-12 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| Bae et al. | Safety evaluation of the inherent and passive safety features of the smart design | |
| CN104361913A (en) | Secondary side passive waste heat removal system | |
| US20210202121A1 (en) | Flow Mixing T-Unit of Reactor Volume Control System | |
| EP3937188A1 (en) | Depressurisation valve | |
| WO2014090106A1 (en) | High-pressure safe injection system for nuclear power stations | |
| CN103871509A (en) | Pressurized water reactor nuclear power plant reactor coolant system | |
| Chang et al. | SMART behavior under over-pressurizing accident conditions | |
| CN105070329A (en) | Nuclear power station secondary side passive residual heat removal system | |
| CN203366766U (en) | Secondary side discharge system for alleviating vapor generator's heat-transfer pipe cracking accidents | |
| CN110148480A (en) | A kind of nuclear power secondary coolant circuit system | |
| CN203338775U (en) | Nuclear power plant steam generator overflow prevention structure | |
| CN110911024A (en) | Sustained release device for steam generator heat transfer tube rupture accident | |
| JP2019520562A (en) | Upgrading the power output of previously deployed nuclear power plants | |
| CN204242602U (en) | Secondary side passive residual heat guiding system | |
| Kim et al. | Safety analysis of SMART | |
| Uspuras et al. | Specific features of the RBMK-1500 reactor and BDBA management | |
| US3488254A (en) | Containment arrangement for water cooled nuclear reactor installation | |
| Kitou et al. | Development of Inherently Safe Technologies for Large Scale BWRs:(1) Plant System | |
| Ullah et al. | Steam Generator Tube Rupture Accident at a NPP and Exploration of Mitigation Strategies for Its Consequence | |
| Zhiqiang et al. | Research on severe accident mitigation strategy of injecting into the OTSG of ACP100 | |
| Kang et al. | Concept Design Evaluation of Passive Containment Pressure and Radioactivity Suppression System with Cooling Tank for SMR | |
| Lim et al. | Comparative Analysis of Existing and Modified Parameters in MAAP-ISAAC 4.03 Code for MSGTR Accident | |
| Lim et al. | Steam Generator Tube Integrity Analysis of a Total Loss of All Heat Sinks Accident for Wolsong NPP Unit 1 | |
| Alblouwy et al. | Evaluation of Multiple Steam Generator Tubes Rupture for SMART | |
| Li et al. | Study of severe accident induced by SBLOCA for a marine nuclear power plant |
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 | ||
| GR01 | Patent grant | ||
| GR01 | Patent grant |