CN113530613A - Control method for conventional island of nuclear power plant after gas loss - Google Patents
Control method for conventional island of nuclear power plant after gas loss Download PDFInfo
- Publication number
- CN113530613A CN113530613A CN202110699007.7A CN202110699007A CN113530613A CN 113530613 A CN113530613 A CN 113530613A CN 202110699007 A CN202110699007 A CN 202110699007A CN 113530613 A CN113530613 A CN 113530613A
- Authority
- CN
- China
- Prior art keywords
- condenser
- power plant
- nuclear power
- conventional island
- controlling
- 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
Images
Classifications
-
- F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
- F01—MACHINES OR ENGINES IN GENERAL; ENGINE PLANTS IN GENERAL; STEAM ENGINES
- F01D—NON-POSITIVE DISPLACEMENT MACHINES OR ENGINES, e.g. STEAM TURBINES
- F01D11/00—Preventing or minimising internal leakage of working-fluid, e.g. between stages
- F01D11/02—Preventing or minimising internal leakage of working-fluid, e.g. between stages by non-contact sealings, e.g. of labyrinth type
- F01D11/04—Preventing or minimising internal leakage of working-fluid, e.g. between stages by non-contact sealings, e.g. of labyrinth type using sealing fluid, e.g. steam
- F01D11/06—Control thereof
-
- F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
- F01—MACHINES OR ENGINES IN GENERAL; ENGINE PLANTS IN GENERAL; STEAM ENGINES
- F01K—STEAM ENGINE PLANTS; STEAM ACCUMULATORS; ENGINE PLANTS NOT OTHERWISE PROVIDED FOR; ENGINES USING SPECIAL WORKING FLUIDS OR CYCLES
- F01K9/00—Plants characterised by condensers arranged or modified to co-operate with the engines
- F01K9/003—Plants characterised by condensers arranged or modified to co-operate with the engines condenser cooling circuits
-
- F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
- F01—MACHINES OR ENGINES IN GENERAL; ENGINE PLANTS IN GENERAL; STEAM ENGINES
- F01K—STEAM ENGINE PLANTS; STEAM ACCUMULATORS; ENGINE PLANTS NOT OTHERWISE PROVIDED FOR; ENGINES USING SPECIAL WORKING FLUIDS OR CYCLES
- F01K9/00—Plants characterised by condensers arranged or modified to co-operate with the engines
- F01K9/006—Vacuum-breakers
-
- F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
- F28—HEAT EXCHANGE IN GENERAL
- F28B—STEAM OR VAPOUR CONDENSERS
- F28B11/00—Controlling arrangements with features specially adapted for condensers
-
- 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
Abstract
The invention discloses a control method for a conventional island of a nuclear power plant after gas loss, which comprises the following steps: 1) after the conventional island loses air, the shaft seal air inlet valve is immediately and manually adjusted on site to maintain the CET shaft seal pressure; 2) after the high-temperature steam entering the condenser is isolated from the drain, the CVI system is shut down, and a vacuum breaker valve is opened; 3) and after the pressure of the condenser is recovered to the atmospheric pressure, the condenser is withdrawn from the CET system. Compared with the prior art, the control method for the conventional island of the nuclear power plant after gas loss can maintain the vacuum of the condenser as much as possible so as to have sufficient time to isolate high-temperature steam entering the condenser from drainage, and simultaneously avoid equipment damage and personnel scald caused by the leakage of shaft seal steam.
Description
Technical Field
The invention relates to the field of debugging of nuclear power plants, in particular to a control method for a conventional island of a nuclear power plant after gas loss.
Background
An EPR (European Pressurized Water Reactor) unit has a great difference from a CPR (China Pressurized Water Reactor) unit in the aspect of power model selection of a valve, a nuclear island area basically adopts an electric valve, and a conventional island area adopts a large amount of pneumatic valves. Aiming at the influences on the system and the unit after the pneumatic valve is deflated and the transient control strategy, the CPR unit is not verified by an integral deflation test, and the EPR unit fills the blank of the part by adopting a COC107 test for the first time.
COC107 (loss of Instrument compressed air test) is used for simulating loss of SAR (Instrument compressed air distribution system) under the condition of normal island hot vacuum, verifying the action condition correctness of each system and equipment and the feasibility of a transient control strategy, providing reference and reference for emergency intervention after the occurrence of air loss transient condition during the normal operation of a subsequent unit, in the prior test scheme, realizing real air loss by closing a main valve on an air inlet main pipe of a normal island, and the system transient control strategy after the air loss of the normal island is shown in figure 1 and is sequentially CVI-CET-SVA-ADG-CEX-CRF-AAD-APA-other systems, wherein CVI (condenser vacuum shaft seal) is a condenser vacuum system, CET (condenser shaft seal) is a steam turbine system, and SVA (vapor distribution system) is an auxiliary steam distribution system, adg (feed water tank and gas stripper system) is a feedwater deaerator system, cex (condensate extraction system) is a condensate extraction system, crf (circulating water system) is a circulating water system, aad (start up and shut down) is a start-up and shut-down feedwater system, and apa (main feed water pump system) is an electric main feedwater pump system.
However, in combination with analysis of influences on aspects such as host equipment, cold source equipment, water supply equipment and the like after gas loss, the above transient control strategy is unreasonable, and safe operation of major equipment is influenced, so that at least the following risks exist for major equipment such as a steam turbine, a condenser, a deaerator, a water supply pump and the like:
1) the shaft seal CET system intervenes too late. After SAR is lost, the CET steam inlet valve of the steam turbine shaft seal system is fully opened after the air loss, and when the adjusting function is unavailable, the shaft seal steam overpressure is leaked. The transient control strategy is characterized in that the CET system is intervened after the CVI vacuum pump is shut down and the vacuum breaking valve is opened, the condenser pressure is increased to atmospheric pressure, and at the moment, the shaft seal steam of the CET system leaks out in an overpressure mode for a long time, so that the risk of damage to equipment such as a turbine bearing and the like and the risk of scalding of personnel exist.
2) Steam entering the condenser and the hydrophobic intervention are too late. After the SAR is lost, a siphon damage valve of a CRF system is deflated and opened, a CRF pump cannot run for a long time and cannot maintain the vacuum of the condenser, and the availability of the condenser is influenced. The high-temperature steam of the ADG system and the high-temperature drain of the VPU system (Main steam system) of the transient control strategy are continuously discharged into the condenser after the condenser is vacuum-damaged, which causes the temperature and the pressure of the condenser to be continuously increased, and risks of damaging the titanium pipe of the condenser and the low-pressure cylinder blasting disc of the steam turbine are existed.
3) And returning the ADG water to a condenser isolation valve and an APA/AAD pump for intervention too late. After losing SAR, lose under the circumstances that CEX valve station governing valve and ADG admission governing valve lost the gas and close, ADG loses into water and admission, and oxygen-eliminating device liquid level and pressure continuously reduce, influence APA AAD pump safe operation and steam generator water supply, and APA AAD pump long-time operation this moment, risk that cavitation erosion and vibration damage improve greatly.
In view of the above, there is a need to provide a conventional post-island outgassing control method for a nuclear power plant, which can overcome the above problems.
Disclosure of Invention
The invention aims to: the control method for the conventional island of the nuclear power plant after gas loss is provided, and the safety operation of major equipment and the safety of personnel of the nuclear power plant are effectively ensured.
In order to achieve the purpose, the invention provides a control method for a conventional island of a nuclear power plant after air loss, which comprises the following steps:
1) after the conventional island loses air, the shaft seal air inlet valve is immediately and manually adjusted on site to maintain the CET shaft seal pressure;
2) after the high-temperature steam entering the condenser is isolated from the drain, the CVI system is shut down, and a vacuum breaker valve is opened;
3) and after the pressure of the condenser is recovered to the atmospheric pressure, the condenser is withdrawn from the CET system.
As a preferred embodiment of the control method after the gas loss of the conventional island of the nuclear power plant, the step 3) further comprises shutting down the SVA system after exiting the CET system.
As a preferred embodiment of the control method for the conventional island after gas loss in the nuclear power plant, during the step 1) of manually adjusting the shaft seal steam inlet valve and maintaining the pressure of the CET shaft seal, high-temperature steam entering a condenser needs to be isolated as soon as possible from drainage.
As a preferred embodiment of the control method for the conventional island of the nuclear power plant after gas loss, the isolation of the high-temperature steam and the drainage which enter the condenser comprises the steps of cutting the exhaust steam of the ADG from the condenser to the atmosphere, isolating the return water from the oxygen remover of the ADG to the condenser, and avoiding the high-temperature steam of the oxygen remover from entering the condenser in a large amount.
As a preferred embodiment of the control method for the conventional island after gas loss in the nuclear power plant, the isolating the high-temperature steam and the drain water entering the condenser further comprises closing a VPU system to the drain water of the condenser.
As a preferred embodiment of the control method for the conventional island after gas loss in the nuclear power plant, the step 3) is followed by a step 4), after all heat sources entering the condenser are isolated, the CRF pump is stopped, and the CRF pump is prevented from operating for a long time under the condition that the siphon break valve is opened.
As a preferred embodiment of the control method for the conventional island after gas loss of the nuclear power plant, the method further comprises the step 5) of stopping the APA and the AAD pumps after isolating the return water from the ADG deaerator to the condenser.
As a preferred embodiment of the control method after the gas loss of the conventional island of the nuclear power plant, in the step 5), after the operation of the APA and the AAD pump is stopped, the CEX pump is gradually removed from operation.
As a preferred embodiment of the control method after the gas loss of the conventional island of the nuclear power plant, the step 5) further comprises a step 6) of stopping the ADG.
As a preferred embodiment of the control method for the conventional island after gas loss of the nuclear power plant, the step 5) and the step 6) are performed after the step 2) and in parallel with the step 3) and the step 4).
As a preferred embodiment of the control method for the conventional island after gas loss in the nuclear power plant, the method further comprises the step 7), after the step 4) and the step 6) are completed, other systems are checked and operated, and the other systems comprise an ABP system, an AHP system, an SEK system, an SIR system and a GSS system.
Compared with the prior art, the control method for the conventional island of the nuclear power plant after gas loss at least has the following advantages:
1) according to the invention, after the conventional island is exhausted, the shaft seal steam inlet valve is immediately and manually adjusted on site to maintain the pressure of the CET shaft seal, then the CVI system is stopped after the high-temperature steam entering the condenser is isolated from the drain water, the vacuum breaking valve is opened, and the CET system is withdrawn after the pressure of the condenser is recovered to the atmospheric pressure, so that the vacuum of the condenser can be maintained as far as possible to isolate the high-temperature steam and the drain water entering the condenser for a sufficient time, and simultaneously, the damage to equipment and the scald of personnel caused by the leakage of the shaft seal steam are avoided;
2) according to the invention, during the period of manually adjusting the shaft seal steam inlet valve and maintaining the pressure of the CET shaft seal, the exhaust steam of the ADG is cut to the atmosphere from the condenser, the return water from the ADG deaerator to the condenser is isolated, the high-temperature steam of the deaerator is prevented from entering the condenser in large quantity, and the exhaust from the VPU system to the condenser is closed, so that the high-temperature steam and the drainage entering the condenser are isolated as soon as possible, and the condenser and low-pressure cylinder equipment can be prevented from being damaged; after all heat sources entering the condenser are isolated, the CRF pump is stopped, so that the CRF pump is prevented from running for a long time under the working condition that the siphon break valve is opened;
3) according to the invention, after the conventional island is degassed, the return water from the ADG deaerator to the condenser is isolated in time, the APA and AAD pumps are stopped, and the CEX pump is gradually withdrawn from operation, so that the damage caused by cavitation erosion and over-high vibration of the APA/AAD pump due to continuous reduction of the water level and pressure of the deaerator is avoided.
Drawings
The following describes in detail the control method after the conventional island of the nuclear power plant loses air and the beneficial effects thereof with reference to the accompanying drawings and the specific implementation.
Fig. 1 illustrates a system transient control strategy after a conventional island is de-aerated.
Fig. 2 is a schematic diagram of a conventional island gas loss related system of a nuclear power plant.
Fig. 3 is a flow diagram of a control method of a nuclear power plant after a conventional island is degassed.
Detailed Description
In order to make the objects, technical solutions and advantageous technical effects of the present invention clearer, the present invention is further described in detail below with reference to the accompanying drawings and the detailed description. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention.
Referring to fig. 2 and 3, in order to solve the problems existing in the conventional system transient control strategy after air loss of the island, the inventor analyzes each risk point by combing the state and influence of each system after air loss, and optimizes and improves the control strategy from the perspective of ensuring the safe withdrawal of system equipment by taking the main risk points as traction.
The method specifically comprises the following points:
1) post-deflation control strategy optimization for host device
After the conventional island loses air, the pressure adjusting function of the CET shaft seal is unavailable, but the turbine lubricating oil and the turning gear system still normally operate, and the CET system can be continuously put into use. In order to maintain the vacuum of the condenser as far as possible so as to have sufficient time to isolate high-temperature steam and drain water entering the condenser and simultaneously avoid equipment damage and personnel scald caused by the leakage of shaft seal steam, a shaft seal steam inlet valve is manually controlled to maintain CET shaft seal pressure immediately, then after the high-temperature steam entering the condenser is isolated from the drain water, the CVI system is stopped, a vacuum destruction valve is opened, and after the pressure of the condenser is recovered to the atmospheric pressure, the CET system is withdrawn.
2) Control strategy optimization after air loss of cold source equipment
During manual adjustment of the CET shaft seal pressure, since the CRF cannot be operated for a long time to maintain the condenser vacuum, in order to prevent damage to the condenser and the low-pressure cylinder equipment, high-temperature steam and drain entering the condenser need to be isolated as soon as possible. According to the influence degree to the condenser, preferentially cut ADG's steam exhaust to the atmosphere, keep apart the return water of oxygen-eliminating device to the condenser by the condenser, avoid oxygen-eliminating device high temperature steam to get into the condenser in a large number.
And after all heat sources entering the condenser are isolated, stopping running the CRF pump, and preventing the CRF pump from running for a long time under the working condition that the siphon break valve is opened.
3) Post-air-loss control strategy optimization for water supply equipment
After the conventional island loses gas, the CEX valve station regulating valve and the deaerator main and auxiliary steam inlet regulating valve lose gas and are closed, and the deaerator cannot maintain the water level and the pressure. On the other hand, after the APA pump is started by the air loss signal, the APA pump and the AAD pump simultaneously operate under a low-flow working condition. In order to prevent the APA/AAD pump from cavitation erosion and damage caused by overhigh vibration due to the reduction of the water level and pressure of the deaerator, the return water from the ADG to the condenser is isolated in time, the APA and AAD pumps are stopped, and the CEX pump is gradually stopped from running.
Through the above analysis, the conventional island air loss control method shown in fig. 3 is obtained.
The control method for the conventional island of the nuclear power plant after gas loss comprises the following steps:
and step S1, immediately and manually adjusting the shaft seal steam inlet valve on site after the conventional island loses air, and maintaining the CET shaft seal pressure.
During the period of manually adjusting a shaft seal steam inlet valve and maintaining the pressure of a CET shaft seal, high-temperature steam and drainage which enter a condenser need to be isolated as soon as possible: include that the exhaust steam with ADG is cut to the atmosphere by the condenser, keep apart the return water of ADG oxygen-eliminating device to condenser, avoid the oxygen-eliminating device high temperature steam to get into the condenser in a large number, still including the drainage of closing VPU system to condenser.
And step S2, after the high-temperature steam entering the condenser is isolated from the drainage, the CVI system is shut down, and the vacuum breaking valve is opened.
And step S3, after the pressure of the condenser is recovered to the atmospheric pressure, the condenser is withdrawn from the CET system. This step also includes shutting down the SVA system after exiting the CET system.
And step S4, after all heat sources entering the condenser are isolated, stopping the CRF pump, and preventing the CRF pump from operating for a long time under the working condition that the siphon break valve is opened.
Step S5, after the ADG deaerator is isolated to the return water of the condenser, the APA and the AAD pump are stopped, and the CEX pump is gradually stopped from running; this step may be performed after step S2, in parallel with step S3.
Step S6, after step S5 is completed, ADG is shut down.
In step S7, after steps S4 and S6 are completed, other systems including an ABP (Low pressure fed water heater) system, an AHP (High pressure fed water heater) system, an SEK (Conventional island waste liquid collection) system, an SIR (Chemical reagents injection system) system and a GSS (steam-water separator-reheater) system are checked and operated.
Compared with the prior art, the control method for the conventional island of the nuclear power plant after gas loss at least has the following advantages:
1) according to the invention, after the conventional island is exhausted, the shaft seal steam inlet valve is immediately and manually adjusted on site to maintain the pressure of the CET shaft seal, then the CVI system is stopped after the high-temperature steam entering the condenser is isolated from the drain water, the vacuum breaking valve is opened, and the CET system is withdrawn after the pressure of the condenser is recovered to the atmospheric pressure, so that the vacuum of the condenser can be maintained as far as possible to isolate the high-temperature steam and the drain water entering the condenser for a sufficient time, and simultaneously, the damage to equipment and the scald of personnel caused by the leakage of the shaft seal steam are avoided;
2) according to the invention, during the period of manually adjusting the shaft seal steam inlet valve and maintaining the pressure of the CET shaft seal, the exhaust steam of the ADG is cut to the atmosphere from the condenser, the return water from the ADG deaerator to the condenser is isolated, the high-temperature steam of the deaerator is prevented from entering the condenser in large quantity, and the exhaust from the VPU system to the condenser is closed, so that the high-temperature steam and the drainage entering the condenser are isolated as soon as possible, and the condenser and low-pressure cylinder equipment can be prevented from being damaged; after all heat sources entering the condenser are isolated, the CRF pump is stopped, so that the CRF pump is prevented from running for a long time under the working condition that the siphon break valve is opened;
3) according to the invention, after the conventional island is degassed, the return water from the ADG deaerator to the condenser is isolated in time, the APA and AAD pumps are stopped, and the CEX pump is gradually withdrawn from operation, so that the damage caused by cavitation erosion and over-high vibration of the APA/AAD pump due to continuous reduction of the water level and pressure of the deaerator is avoided.
Through practical verification, the state of the COC107 test unit adopting the control method after the conventional island of the nuclear power plant is out of gas is safe and controllable, no safety quality event occurs, the construction period is reduced to a certain extent compared with a plan, the safety quality is improved, and human errors are reduced.
In conclusion, the control method for the conventional island of the nuclear power plant after the air loss eliminates risk factors in the system transient control strategy after the conventional island loses air, can ensure the safety of conventional island systems and equipment such as a steam turbine, a condenser, an APA/AAD pump and the like, and effectively shortens the control time after the conventional island loses air.
Appropriate changes and modifications to the embodiments described above will become apparent to those skilled in the art from the disclosure and teachings of the foregoing description. Therefore, the present invention is not limited to the specific embodiments disclosed and described above, and some modifications and variations of the present invention should fall within the scope of the claims of the present invention. Furthermore, although specific terms are employed herein, they are used in a generic and descriptive sense only and not for purposes of limitation.
Claims (11)
1. A control method for a conventional island of a nuclear power plant after gas loss is characterized by comprising the following steps:
1) after the conventional island loses air, the shaft seal air inlet valve is immediately and manually adjusted on site to maintain the CET shaft seal pressure;
2) after the high-temperature steam entering the condenser is isolated from the drain, the CVI system is shut down, and a vacuum breaker valve is opened;
3) and after the pressure of the condenser is recovered to the atmospheric pressure, the condenser is withdrawn from the CET system.
2. The method for controlling the conventional island of the nuclear power plant after the gas loss according to claim 1, is characterized in that: the step 3) further comprises shutting down the SVA system after exiting the CET system.
3. The method for controlling the conventional island of the nuclear power plant after the gas loss according to claim 1 or 2, is characterized in that: in the step 1), a shaft seal steam inlet valve is manually adjusted, and high-temperature steam entering a condenser and drainage are isolated as soon as possible during the period of maintaining the pressure of the CET shaft seal.
4. The method for controlling the conventional island of the nuclear power plant after the air loss according to claim 3, is characterized in that: keep apart the high temperature steam that gets into the condenser and hydrophobic include that the steam extraction with ADG is cut to the atmosphere by the condenser, keeps apart the return water of ADG oxygen-eliminating device to condenser, avoids oxygen-eliminating device high temperature steam to get into the condenser in a large number.
5. The method for controlling the conventional island of the nuclear power plant after the air loss according to claim 4, is characterized in that: and the step of isolating the high-temperature steam and the drain entering the condenser further comprises the step of closing a VPU system to the drain of the condenser.
6. The method for controlling the conventional island of the nuclear power plant after the air loss according to claim 5, is characterized in that: and 4) after the step 3), stopping the CRF pump after isolating all heat sources entering the condenser, and preventing the CRF pump from operating for a long time under the working condition that the siphon break valve is opened.
7. The method for controlling the conventional island of the nuclear power plant after the air loss according to claim 6, is characterized in that: and 5) after the ADG deaerator is isolated from the return water of the condenser, the APA and the AAD pump are stopped.
8. The method for controlling the conventional island of the nuclear power plant after the air loss according to claim 7, is characterized in that: and 5) after the APA and the AAD pumps are stopped, gradually stopping the CEX pump from running.
9. The method for controlling the conventional island of the nuclear power plant after the air loss according to claim 8, is characterized in that: step 6) is also included after step 5), the ADG is shut down.
10. The method for controlling the conventional island of the nuclear power plant after the air loss according to claim 9, is characterized in that: the step 5) and the step 6) are performed after the step 2) and in parallel with the step 3) and the step 4).
11. The method for controlling the conventional island of the nuclear power plant after the gas loss according to claim 10, wherein: and 7), after the steps 4) and 6) are completed, checking and operating other systems, wherein the other systems comprise an ABP system, an AHP system, an SEK system, an SIR system and a GSS system.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202110699007.7A CN113530613B (en) | 2021-06-23 | 2021-06-23 | Control method after gas loss of conventional island of nuclear power plant |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202110699007.7A CN113530613B (en) | 2021-06-23 | 2021-06-23 | Control method after gas loss of conventional island of nuclear power plant |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN113530613A true CN113530613A (en) | 2021-10-22 |
| CN113530613B CN113530613B (en) | 2023-07-14 |
Family
ID=78096527
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN202110699007.7A Active CN113530613B (en) | 2021-06-23 | 2021-06-23 | Control method after gas loss of conventional island of nuclear power plant |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN113530613B (en) |
Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH06281367A (en) * | 1993-03-26 | 1994-10-07 | Toshiba Corp | Vacuum keeping system for condenser |
| EP2801703A1 (en) * | 2013-05-08 | 2014-11-12 | Siemens Aktiengesellschaft | Steam turbine plant with sealing steam line |
| CN104791022A (en) * | 2015-02-15 | 2015-07-22 | 华北电力科学研究院有限责任公司 | Fuel gas thermal power plant gland seal and vacuum system and start-stop control method thereof |
| CN105781631A (en) * | 2016-05-10 | 2016-07-20 | 中国能源建设集团山西省电力勘测设计院有限公司 | Starting system and starting method for steam turbine running of coal-fired power plant |
| CN110542325A (en) * | 2019-09-26 | 2019-12-06 | 岭澳核电有限公司 | Condenser vacuum system of nuclear power station |
-
2021
- 2021-06-23 CN CN202110699007.7A patent/CN113530613B/en active Active
Patent Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH06281367A (en) * | 1993-03-26 | 1994-10-07 | Toshiba Corp | Vacuum keeping system for condenser |
| EP2801703A1 (en) * | 2013-05-08 | 2014-11-12 | Siemens Aktiengesellschaft | Steam turbine plant with sealing steam line |
| CN104791022A (en) * | 2015-02-15 | 2015-07-22 | 华北电力科学研究院有限责任公司 | Fuel gas thermal power plant gland seal and vacuum system and start-stop control method thereof |
| CN105781631A (en) * | 2016-05-10 | 2016-07-20 | 中国能源建设集团山西省电力勘测设计院有限公司 | Starting system and starting method for steam turbine running of coal-fired power plant |
| CN110542325A (en) * | 2019-09-26 | 2019-12-06 | 岭澳核电有限公司 | Condenser vacuum system of nuclear power station |
Also Published As
| Publication number | Publication date |
|---|---|
| CN113530613B (en) | 2023-07-14 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| US3953966A (en) | Combined cycle electric power plant having a control system which enables dry steam generator operation during gas turbine operation | |
| KR100315342B1 (en) | Water Supply System of Power Plant | |
| JP3800384B2 (en) | Combined power generation equipment | |
| CN110131003B (en) | System and method for starting and stopping two loops of high-temperature gas cooled reactor nuclear power unit | |
| CN113530613B (en) | Control method after gas loss of conventional island of nuclear power plant | |
| CN110425011B (en) | Optimal control method for shaft seal and door rod steam leakage system of power station steam turbine unit | |
| CN112761738B (en) | Novel method for judging tightness of steam extraction check valve of steam turbine | |
| CN107084011A (en) | A kind of nuclear steam turbine method for rapid cooling | |
| CN114112200B (en) | Method for adjusting pipeline system for adjusting depressurization rate in boiler water overpressure experiment | |
| JPH0498198A (en) | Core cooling facility for nuclear power plant | |
| CN111911252B (en) | Supercritical heat supply unit and cold starting method thereof | |
| KR20170098222A (en) | Method for managing stoppage of a pressurised-water nuclear reactor | |
| CN113972016B (en) | Method, device, equipment and medium for coping with water loss accident outside containment of nuclear power plant | |
| CN114811564A (en) | Intelligent nitrogen shutdown protection process for thermal power generating unit | |
| CN111271603B (en) | Nuclear power station valve diaphragm maintenance method | |
| CN104989467B (en) | Steam turbine shutdown cooling control method | |
| JP4155495B2 (en) | Cooling equipment for reactor isolation | |
| CN217584390U (en) | High temperature gas cooled reactor steam recovery system | |
| CN206190335U (en) | Steam turbine emergency trip system | |
| CN217235273U (en) | Device for adding compressed air storage tank to main chassis vehicle of nuclear power station steam turbine | |
| CN106948886B (en) | Steam turbine method for rapid cooling | |
| CN217206587U (en) | Three-extraction temperature control system of straight condensing steam turbine | |
| CN215214150U (en) | Pressurized water reactor loop two-stage series injection type vacuumizing static exhaust device | |
| Hu et al. | Analysis and mitigation strategy on the decreasing opening of thermal pressurizer pilot operated safety relief valves | |
| CN113566186B (en) | Boiler drum water level adjusting method after damaged stage pressure measuring points |
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 |