CN104538068B - A kind of method that steam generator spill-over is prevented under shape pipe breaking accident operating mode of conducting heat - Google Patents
A kind of method that steam generator spill-over is prevented under shape pipe breaking accident operating mode of conducting heat Download PDFInfo
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- CN104538068B CN104538068B CN201310306912.7A CN201310306912A CN104538068B CN 104538068 B CN104538068 B CN 104538068B CN 201310306912 A CN201310306912 A CN 201310306912A CN 104538068 B CN104538068 B CN 104538068B
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C17/00—Monitoring; Testing ; Maintaining
- G21C17/017—Inspection or maintenance of pipe-lines or tubes in nuclear installations
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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)
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
A kind of method the present invention relates to prevent steam generator spill-over under shape pipe breaking accident operating mode of conducting heat, including：After step one, nuclear power station occur steam generator tube rupture accident, emergency shut-down is triggered；Step 2, judge whether to occur heat transfer shape pipe breaking accident accident and determine damaged steam generator；And isolate the steam pipe line of loop where damaged steam generator；Step 3, operator adjust the water level of damaged steam generator to steam generator high water level high manually by the exhaust system on steam generator blowdown pipeline；The airborne release system that control simultaneously does not occur heat transfer shape pipe breaking accident accident loop is lowered the temperature to primary Ioops, is depressured, to balance the pressure of damaged primary side of steam generator, secondary side.The inventive method achieves the spill-over for preventing breakage SG secondary sides under SGTR accident conditionses, burst size of the radioactive substance to environment is greatly reduced.
Description
Technical field
The present invention relates to nuclear power plant accident mitigation technique field, and in particular to prevented under one kind heat transfer shape pipe breaking accident operating mode
The method of steam generator spill-over.
Background technology
Steam generator may occur during nuclear power station runs（SG）Heat transfer shape pipe breaking accident（SGTR）.The accident will
Cause main coolant loop that the secondary side of SG is flowed into by cut, so as to cause damaged SG secondary sides water level to rise, ultimately result in
Damaged SG secondary sides full water.Damaged SG secondary sides full water will cause with active water directly by the airborne release system of secondary circuit
System enters environment, so as to be polluted to environment.
Active service nuclear power station is triggered when there is SGTR accidents by voltage-stablizer pressure low signal or the low low signal of voltage-stablizer pressure
Reactor shutdown and engineered safeguards features put into, and it is logical that the corresponding accident treatment code of Nuclear Power Plant Operators application afterwards carries out isolation
Toward the operation such as breakage SG auxiliary feedwaters, stoppage in transit engineered safeguards features, the final main coolant loop that terminates is leaked to secondary side, and will
Reactor brings corresponding treatment state into.But, accident sunykatuib analysis result shows, using the alleviation method of this accident, breaks
SG spill-overs are damaged, burst size increased adverse consequences of the radioactive substance to environment is caused.
The content of the invention
The technical problem to be solved of the invention be to provide it is a kind of can pressurized-water reactor nuclear power plant steam generator conduct heat
During shape pipe breaking accident operating mode, prevent damaged steam generator spill-over and avoid radioactive water from entering environment by airborne release system
Method.
In order to solve the above-mentioned technical problem, the technical scheme is that, one kind heat transfer shape pipe breaking accident operating mode under prevent
The method of steam generator spill-over, comprises the following steps：
Action before step one, Operator actions
After there is steam generator tube rupture accident in nuclear power station, the different capacity level residing for foundation nuclear power station, by
Voltage-stablizer pressure low signal or the steam generator water level signal that meets low with voltage-stablizer water level high trigger emergency shut-down；
The action for performing at once is needed after step 2, Operator actions
After Operator actions, the radioactivity alarm signal according to radiation monitoring system judges that whether heat-transfer pipe occurs is broken
Split accident accident and determine damaged steam generator；And isolate the steam pipe line of loop where damaged steam generator, will be broken
The discharge definite value for damaging the airborne release system of loop where steam generator brings up to 8.2MPa；
Post-actions after step 3, Operator actions
Heat transfer shape pipe breaking accident occurs and after Operator actions, and operator is by the discharge on steam generator blowdown pipeline
System adjusts the water level of damaged steam generator to steam generator high water level high manually；There is no heat transfer tracheal rupture in control simultaneously
The airborne release system of accident accident loop is lowered the temperature to primary Ioops, is depressured, to balance damaged primary side of steam generator, two
The pressure of secondary side, brings reactor into state that cut flow is zero, final to realize preventing damaged steam generator spill-over.
The step one also includes that the low low signal triggering safety injection system of voltage-stablizer pressure injects water into steam generator.
The steam generator water level is high to isolate the auxiliary of damaged steam generator with the low low signal triggering of voltage-stablizer water level
Help feedwater.
It refers to that the narrow range water level of steam generator reaches the narrow range of steam generator that the steam generator water level is high
75%。
It refers to that voltage-stablizer water level is reduced to the 14% of voltage-stablizer water level range that the voltage-stablizer water level is low.
Beneficial effects of the present invention：
The inventive method achieves the spill-over for preventing breakage SG secondary sides under SGTR accident conditionses, greatly reduce and put
Burst size from penetrating property material to environment.
Specific embodiment
The present invention is described further with reference to embodiments.
The method that steam generator spill-over is prevented under a kind of shape pipe breaking accident operating mode of conducting heat of the present invention, comprises the following steps：
Action before step one, Operator actions
After there is steam generator tube rupture accident in nuclear power station, the different capacity level residing for foundation nuclear power station, by
Voltage-stablizer pressure low signal or the steam generator water level signal that meets low with voltage-stablizer water level high trigger emergency shut-down；
Inject to make up the loss of cooling agent by the low low signal triggering peace note of voltage-stablizer pressure；
Steam generator water level is high to isolate damaged steam generator with the low low signal triggering of voltage-stablizer water level（Occur high
The steam generator of water level）Auxiliary feedwater, with slow down damaged steam generator secondary side water loading amount increase；
It refers to that the narrow range water level of steam generator reaches setting valve high, i.e. steam hair that the steam generator water level is high
The 75% of the narrow range of raw device；
It refers to that voltage-stablizer water level is reduced to low setting valve, i.e. voltage-stablizer water level range that the voltage-stablizer water level is low
14%；
The action for performing at once is needed after step 2, Operator actions
After Operator actions, whether the radioactivity alarm signal according to radiation monitoring system judges steam generation occurs
Device TR accidents and the damaged steam generator of determination；
Isolate the steam pipe line of loop where damaged steam generator at once, and by loop where damaged steam generator
The discharge definite value of airborne release system brings up to 8.2MPa；To reduce the airborne release system of damaged steam generator loop as far as possible
Unite and discharged to the steam of environment；The discharge definite value for improving the airborne release system of loop where damaged steam generator can also subtract
Pressure difference between small damaged primary side of steam generator and secondary side, is reduced primary Ioops cooling agent and is sent out to damaged steam by cut
The release of raw device secondary side.
Post-actions after step 3, Operator actions
Steam generator tube rupture accident occurs and after Operator actions, to prevent damaged steam generator spill-over,
Operator is sent out by the water level that the exhaust system on steam generator blowdown pipeline adjusts damaged steam generator manually to steam
Raw device high water level high；
The airborne release system by there is no steam generator tube rupture accident loop is carried out to primary Ioops simultaneously
Cooling, step-down, to balance the pressure of damaged primary side of steam generator, secondary side, and then terminate main coolant loop to secondary
Side leaks, and brings reactor into state that cut flow is zero, final to realize preventing damaged steam generator spill-over, reduces broken
Damage burst size of the steam generator to environment.
Claims (4)
1. a kind of method that steam generator spill-over is prevented under shape pipe breaking accident operating mode of conducting heat, it is characterised in that：Including following step
Suddenly：
Action before step one, Operator actions
After nuclear power station occurs steam generator tube rupture accident, according to the different capacity level residing for nuclear power station, by voltage stabilizing
Device pressure low signal or the steam generator water level signal that meets low with voltage-stablizer water level high trigger emergency shut-down；
The action for performing at once is needed after step 2, Operator actions
After Operator actions, whether the radioactivity alarm signal according to radiation monitoring system judges tracheal rupture thing of conducting heat
Therefore and the damaged steam generator of determination；And isolate the steam pipe line of loop where damaged steam generator, by damaged steam hair
The discharge definite value of the airborne release system of loop brings up to 8.2MPa where raw device；
Post-actions after step 3, Operator actions
Heat transfer shape pipe breaking accident occurs and after Operator actions, and operator is by the exhaust system on steam generator blowdown pipeline
The water level of the damaged steam generator of regulation is to steam generator high water level high manually；There is no heat transfer shape pipe breaking accident in control simultaneously
The airborne release system of loop is lowered the temperature to primary Ioops, is depressured, to balance the pressure of damaged primary side of steam generator, secondary side
Power, brings reactor into state that cut flow is zero, final to realize preventing damaged steam generator spill-over；
It refers to that the narrow range water level of steam generator reaches the 75% of the narrow range of steam generator that the steam generator water level is high.
2. according to the method that steam generator spill-over is prevented under the heat transfer shape pipe breaking accident operating mode described in claim 1, its feature
It is：The step one also includes that the low low signal triggering safety injection system of voltage-stablizer pressure injects water into steam generator.
3. according to the method that steam generator spill-over is prevented under the heat transfer shape pipe breaking accident operating mode described in claim 1, its feature
It is：Steam generator water level low signal triggering low with voltage-stablizer water level high isolate the auxiliary of breakage steam generator to
Water.
4. according to the method that steam generator spill-over is prevented under the heat transfer shape pipe breaking accident operating mode described in claim 1, its feature
It is：It refers to that voltage-stablizer water level is reduced to the 14% of voltage-stablizer water level range that the voltage-stablizer water level is low.
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| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201310306912.7A CN104538068B (en) | 2013-07-22 | 2013-07-22 | A kind of method that steam generator spill-over is prevented under shape pipe breaking accident operating mode of conducting heat |
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| CN201310306912.7A CN104538068B (en) | 2013-07-22 | 2013-07-22 | A kind of method that steam generator spill-over is prevented under shape pipe breaking accident operating mode of conducting heat |
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| CN104538068A CN104538068A (en) | 2015-04-22 |
| CN104538068B true CN104538068B (en) | 2017-06-20 |
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| CN107068214B (en) * | 2017-05-09 | 2024-03-26 | 中广核研究院有限公司 | Nuclear power plant steam atmosphere discharge device and two-loop pressure release method |
| CN107195343B (en) * | 2017-05-23 | 2023-04-18 | 中广核研究院有限公司 | Detection control device and method for breakage and leakage of heat transfer pipe in nuclear reactor |
| CN109859866B (en) * | 2019-03-06 | 2022-02-22 | 中国核动力研究设计院 | Method for relieving accident consequence of main steam pipeline rupture |
| CN109994230A (en) * | 2019-04-12 | 2019-07-09 | 西安热工研究院有限公司 | A kind of passive dump of nuclear power station steam generator and cooling system and method |
| CN110415848B (en) * | 2019-08-05 | 2020-11-24 | 中国核动力研究设计院 | Protection system for reducing superimposed SWCCF accidents in response to heat extraction |
| CN110718312B (en) * | 2019-09-18 | 2023-08-04 | 上海电力大学 | System and method for terminating safety injection under heat transfer pipe fracture accident |
| CN110689973B (en) * | 2019-09-18 | 2023-04-28 | 上海电力大学 | Nuclear power station primary loop depressurization control method under heat transfer pipe rupture accident |
| CN110718313B (en) * | 2019-09-18 | 2023-08-04 | 上海电力大学 | Nuclear power station primary loop cooling control method under heat transfer pipe rupture accident |
| CN110675966B (en) * | 2019-09-18 | 2023-07-04 | 上海电力大学 | System and method for isolating steam generator under heat transfer pipe rupture accident |
| CN110911024A (en) * | 2019-11-21 | 2020-03-24 | 中国核动力研究设计院 | Sustained release device for steam generator heat transfer tube rupture accident |
| 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 |
| CN111681794B (en) * | 2020-06-19 | 2022-02-22 | 中国核动力研究设计院 | Full-range SGTR accident handling method and system for pressurized water reactor nuclear power plant |
| CN112366012B (en) * | 2020-10-23 | 2022-05-03 | 岭东核电有限公司 | Water level early warning method and device for steam generator, terminal equipment and storage medium |
| CN113421663B (en) * | 2021-06-18 | 2022-04-15 | 中国核动力研究设计院 | Natural circulation cooling method suitable for pressurized water reactor nuclear power plant |
| CN115331858A (en) * | 2022-08-16 | 2022-11-11 | 中国核动力研究设计院 | Method for processing SGTR (steam generator and turbine control unit) accident of pressurized water reactor nuclear power plant and control system |
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| CN103050161A (en) * | 2012-12-11 | 2013-04-17 | 中国核电工程有限公司 | Method for automatically isolating auxiliary water supply pipeline |
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| CN102446566B (en) * | 2009-01-05 | 2014-06-04 | 中国广东核电集团有限公司 | Steam generator water-level control method and device and nuclear power plant of nuclear power generating unit |
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| Publication number | Priority date | Publication date | Assignee | Title |
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| CN103050161A (en) * | 2012-12-11 | 2013-04-17 | 中国核电工程有限公司 | Method for automatically isolating auxiliary water supply pipeline |
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