CN111963264A - System and method for non-nuclear steam rush-transfer of sodium-cooled fast reactor steam turbine - Google Patents
System and method for non-nuclear steam rush-transfer of sodium-cooled fast reactor steam turbine Download PDFInfo
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- CN111963264A CN111963264A CN202010974599.4A CN202010974599A CN111963264A CN 111963264 A CN111963264 A CN 111963264A CN 202010974599 A CN202010974599 A CN 202010974599A CN 111963264 A CN111963264 A CN 111963264A
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- 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
- F01K11/00—Plants characterised by the engines being structurally combined with boilers or condensers
- F01K11/02—Plants characterised by the engines being structurally combined with boilers or condensers the engines being turbines
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- 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
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- F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
- F02—COMBUSTION ENGINES; HOT-GAS OR COMBUSTION-PRODUCT ENGINE PLANTS
- F02C—GAS-TURBINE PLANTS; AIR INTAKES FOR JET-PROPULSION PLANTS; CONTROLLING FUEL SUPPLY IN AIR-BREATHING JET-PROPULSION PLANTS
- F02C1/00—Gas-turbine plants characterised by the use of hot gases or unheated pressurised gases, as the working fluid
- F02C1/04—Gas-turbine plants characterised by the use of hot gases or unheated pressurised gases, as the working fluid the working fluid being heated indirectly
- F02C1/05—Gas-turbine plants characterised by the use of hot gases or unheated pressurised gases, as the working fluid the working fluid being heated indirectly characterised by the type or source of heat, e.g. using nuclear or solar energy
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- F—MECHANICAL ENGINEERING; LIGHTING; HEATING; WEAPONS; BLASTING
- F22—STEAM GENERATION
- F22B—METHODS OF STEAM GENERATION; STEAM BOILERS
- F22B33/00—Steam-generation plants, e.g. comprising steam boilers of different types in mutual association
- F22B33/18—Combinations of steam boilers with other apparatus
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Abstract
The invention discloses a system and a method for non-nuclear steam rush-transfer of a sodium-cooled fast reactor steam turbine, which comprises an auxiliary steam header, a deaerator, a high-pressure heater, a steam generator, a superheater, a reactor first/second loop, a steam turbine and a steam turbine shaft seal system, wherein the auxiliary steam header is connected with the deaerator; the outlet of the auxiliary steam header is communicated with a steam inlet of the deaerator and a steam inlet of the high-pressure heater, the water outlet of the deaerator is communicated with the water inlet of the high-pressure heater, the water outlet of the high-pressure heater is communicated with the heat absorption side inlet of the steam generator, the heat absorption side outlet of the steam generator is communicated with the heat absorption side inlet of the superheater, and the heat absorption side outlet of the superheater is communicated with the steam inlet of the steam turbine; the system can effectively solve the problem that a sodium-cooled fast reactor steam turbine cannot carry out non-nuclear steam rush-rotation, and reduce the thermal shock frequency of the first loop/the second loop of the reactor.
Description
Technical Field
The invention belongs to the technical field of nuclear power plant debugging, and relates to a system and a method for non-nuclear steam rush-transfer of a sodium-cooled fast reactor steam turbine.
Background
The technical scheme that the steam turbine of the current nuclear power unit utilizes non-nuclear steam to carry out turbine impulse rotation is generally two: the heat generated by the rotation of a main pump of a loop and the friction of a coolant and the heat generated by an electric heater of a voltage stabilizer of the loop are utilized to improve the temperature of the coolant and the reactor components of the loop for heat storage, a steam generator generates steam, or the steam generated by an auxiliary boiler is utilized to enable a steam turbine to run at a speed increasing and constant speed, and the performances of the steam turbine and auxiliary machines thereof are verified in the period.
For the sodium-cooled fast reactor, the above method has a plurality of disadvantages: firstly, a steam turbine is flushed by utilizing heat stored in a reactor, the reactor needs repeated heat storage for many times, the consumption time period is long, and certain thermal shock is generated on the first reactor or the second reactor by repeated temperature rise and temperature reduction; and secondly, the heat generated by the reactor primary pump/secondary loop and the electric heater and the system heat accumulation are limited, and because the deaerator, the turbine shaft seal and the high-pressure or low-pressure heater need to consume a large amount of steam and are limited by the cooling rate of the primary loop/secondary loop equipment, the available steam amount of the turbine is small on the premise that the steam pressure and the temperature are met, and the running requirement of the turbine cannot be met. Thirdly, the temperature and the pressure of the auxiliary steam are lower than the required values of the turbine impulse rotation, according to calculation, the auxiliary steam can only impulse the turbine to 2000rpm, and the requirement that the turbine is impulse rotated to 3000rpm and rotates at a constant speed cannot be met.
Disclosure of Invention
The invention aims to overcome the defects of the prior art and provides a system and a method for non-nuclear steam transfer of a sodium-cooled fast reactor steam turbine, which can effectively solve the problem that the sodium-cooled fast reactor steam turbine cannot carry out non-nuclear steam transfer and reduce the thermal shock times of a reactor one/two loop.
In order to achieve the aim, the system for non-nuclear steam rush-transfer of the sodium-cooled fast reactor steam turbine comprises an auxiliary steam header, a deaerator, a high-pressure heater, a steam generator, a superheater, a reactor first/second loop, a steam turbine and a steam turbine shaft seal system;
the outlet of the auxiliary steam header is communicated with a steam inlet of the deaerator and a steam inlet of the high-pressure heater, the water outlet of the deaerator is communicated with the water inlet of the high-pressure heater, the water outlet of the high-pressure heater is communicated with the heat absorption side inlet of the steam generator, the heat absorption side outlet of the steam generator is communicated with the heat absorption side inlet of the superheater, and the heat absorption side outlet of the superheater is communicated with the steam inlet of the steam turbine; the outlet of the first loop/the second loop of the reactor is communicated with the inlet of the first loop/the second loop of the reactor after sequentially passing through the heat release side of the steam generator and the heat release side of the superheater.
The steam-water separator also comprises an auxiliary boiler, wherein a steam outlet of the auxiliary boiler is communicated with an inlet of the auxiliary steam header.
The water outlet of the deaerator is communicated with the water inlet of the high-pressure heater through a main water feed pump.
The water outlet of the high-pressure heater is communicated with the heat absorption side inlet of the steam generator through a water supply control valve group.
The auxiliary steam header is also communicated with a steam turbine shaft seal system.
A method for non-nuclear steam rush-transfer of a sodium-cooled fast reactor steam turbine comprises the following steps:
the superheated steam output by the auxiliary steam header is divided into three paths, wherein one path of the superheated steam enters the deaerator, the other path of the superheated steam enters the high-pressure heater, and the other path of the superheated steam is supplied to the steam turbine shaft seal system. The method comprises the following steps that feed water is deaerated and preheated by a deaerator, then enters a heater for preheating, then sequentially enters a heat absorption side of a steam generator and a heat absorption side of a superheater for absorbing heat to form superheated steam, and finally enters a steam turbine for applying work;
working media output by the first loop/the second loop of the reactor sequentially enter the heat release side of the steam generator and the heat release side of the superheater for releasing heat, and then enter the first loop/the second loop of the reactor.
The invention has the following beneficial effects:
the system and the method for the non-nuclear steam rush-rotation of the sodium-cooled fast reactor steam turbine have the advantages that when the system and the method are operated specifically, energy is stored by the first loop/the second loop of the reactor and the electric heater, meanwhile, superheated steam provided by the auxiliary steam header is used for supplying water for preheating and supplying the water to the steam turbine shaft seal system, and the cooling rate of the first loop and the second loop of the reactor is reduced, so that the steam turbine can be continuously and stably at a rated rotating speed for a long time during the non-nuclear rush-rotation period, meanwhile, the first loop/the second loop of the reactor does not need to store heat for many times, the cooling rate is far smaller than a design value, the continuity and the stability of the non-nuclear steam rush-rotation process of the steam turbine are improved, the non-nuclear rush-rotation efficiency is improved, the thermal shock times of the first.
Drawings
FIG. 1 is a schematic structural diagram of the present invention.
Wherein, 1 is auxiliary boiler, 2 is auxiliary steam header, 3 is oxygen-eliminating device, 4 is main feed water pump, 5 is high pressure heater, 6 is water supply control valve group, 7 is steam generator, 8 is superheater, 9 is reactor one/two loop, 10 is steam turbine, 11 is steam turbine shaft seal system.
Detailed Description
The invention is described in further detail below with reference to the accompanying drawings:
referring to fig. 1, the system for non-nuclear steam rush-transfer of the sodium-cooled fast reactor steam turbine of the invention comprises an auxiliary steam header 2, a deaerator 3, a high-pressure heater 5, a steam generator 7, a superheater 8, a steam turbine 10 and a reactor first/second loop 9; an outlet of the auxiliary steam header 2 is communicated with a steam inlet of the deaerator 3 and a steam inlet of the high-pressure heater 5, a water outlet of the deaerator 3 is communicated with a water inlet of the high-pressure heater 5, a water outlet of the high-pressure heater 5 is communicated with a heat absorption side inlet of the steam generator 7, a heat absorption side outlet of the steam generator 7 is communicated with a heat absorption side inlet of the superheater 8, and a heat absorption side outlet of the superheater 8 is communicated with a steam inlet of the steam turbine 10; the outlet of the first loop/second loop 9 of the reactor is communicated with the inlet of the first loop/second loop 9 of the reactor after passing through the heat release side of the steam generator 7 and the heat release side of the superheater 8 in sequence.
The invention also comprises an auxiliary boiler 1, wherein a steam outlet of the auxiliary boiler 1 is communicated with an inlet of the auxiliary steam header 2.
The water outlet of the deaerator 3 is communicated with the water inlet of the high-pressure heater 5 through the main water feed pump 4. The water outlet of the high-pressure heater 5 is communicated with the heat absorption side inlet of the steam generator 7 through a water supply control valve group 6. The auxiliary steam header 2 is also communicated with a steam turbine shaft seal system 11.
The method for non-nuclear steam rush-transfer of the sodium-cooled fast reactor steam turbine comprises the following steps:
the superheated steam output by the auxiliary steam header 2 is divided into three paths, wherein one path of the superheated steam enters the deaerator 3, the other path of the superheated steam enters the high-pressure heater 5, and the other path of the superheated steam is supplied to the steam turbine shaft seal system 11. The feed water is deoxidized and preheated by the deaerator 3, then enters the heater for preheating, then sequentially enters the heat absorption side of the steam generator 7 and the heat absorption side of the superheater 8 for absorbing heat to form superheated steam, and finally enters the steam turbine 10 for doing work;
working media output by the reactor I/II loop 9 enter the heat release side of the steam generator 7 and the heat release side of the superheater 8 in sequence to release heat, and then enter the reactor I/II loop 9.
Example one
The specific working process of this embodiment is as follows:
the method comprises the following steps that an auxiliary boiler 1 generates superheated steam with the pressure of 2.1MPa, the temperature of 280 ℃ and the highest flow rate of 115t/h, the superheated steam enters an auxiliary steam header 2, the steam is stabilized to 1.8MPa and 270-274 ℃ in the auxiliary steam header 2, the steam output by the auxiliary steam header 2 enters a deaerator 3, and feed water in the deaerator 3 is heated to 210 ℃; meanwhile, the steam output by the auxiliary steam header 2 is also supplied to a steam turbine shaft seal system; the water supply pressure of the main water supply pump 4 is adjusted by controlling the rotating speed of the main water supply pump 4, and the water supply flow is controlled by the water supply control valve group 6, so that the inlet pressure of the water side of the steam generator 7 is 3.55MPa, and the flow is 73.56 t/h. Steam output by the auxiliary steam header 2 enters the high-pressure heater 5, and the temperature of feed water in the high-pressure heater 5 is heated from 210 ℃ to 240 ℃;
the highest temperature of the output working medium of the reactor one/two loop 9 is 358 ℃, the feedwater absorbs heat in the steam generator 7 and the superheater 8, the feedwater at 240 ℃ is vaporized into superheated steam at 300 ℃, the required heat absorption capacity is 39.702MW, wherein the main pump and the electric heater of the reactor one/two loop 9 provide 17.3MW, and the reactor one/two loop 9 provides 22.402MW of heat through temperature reduction.
Calculated, in the process of providing 22.402MW heat by utilizing the heat storage and temperature reduction of the reactor I/II circuit 9, the temperature reduction rate of the reactor I/II circuit 9 is 17.43 ℃/h which is far less than the safety limit value of 30 ℃/h required by design, the temperature reduction amplitude of the reactor I/II circuit 9 can be 50 ℃, namely 172min steam can be continuously provided, the process of flushing the steam turbine 10 to 3000rpm takes 95min, and theoretically, the steam turbine 10 can continuously operate at 3000rpm rated speed for 77 min.
Claims (6)
1. A system for non-nuclear steam rush-transfer of a sodium-cooled fast reactor steam turbine is characterized by comprising an auxiliary steam header (2), a deaerator (3), a high-pressure heater (5), a steam generator (7), a superheater (8), a reactor one/two loop (9), a steam turbine (10) and a steam turbine shaft seal system (11);
an outlet of the auxiliary steam header (2) is communicated with a steam inlet of the deaerator (3) and a steam inlet of the high-pressure heater (5), a water outlet of the deaerator (3) is communicated with a water inlet of the high-pressure heater (5), a water outlet of the high-pressure heater (5) is communicated with a heat absorption side inlet of the steam generator (7), a heat absorption side outlet of the steam generator (7) is communicated with a heat absorption side inlet of the superheater (8), and a heat absorption side outlet of the superheater (8) is communicated with a steam inlet of the steam turbine (10); the outlet of the reactor first/second loop (9) is communicated with the inlet of the reactor first/second loop (9) after sequentially passing through the heat release side of the steam generator (7) and the heat release side of the superheater (8).
2. The system for the non-nuclear steam surge of the sodium-cooled fast reactor steam turbine according to claim 1, characterized by further comprising an auxiliary boiler (1), wherein a steam outlet of the auxiliary boiler (1) is communicated with an inlet of the auxiliary steam header (2).
3. The system for the non-nuclear steam surge of the sodium-cooled fast reactor steam turbine according to claim 1, characterized in that a water outlet of the deaerator (3) is communicated with a water inlet of the high-pressure heater (5) through a main water feed pump (4).
4. The system for the non-nuclear steam surge of the sodium-cooled fast reactor steam turbine according to claim 1, characterized in that the water outlet of the high-pressure heater (5) is communicated with the heat absorption side inlet of the steam generator (7) through a water supply control valve group (6).
5. The system for non-nuclear steam surge of the sodium-cooled fast reactor steam turbine according to claim 1, characterized in that the auxiliary steam header (2) is also communicated with a steam turbine shaft seal system (11).
6. A method for non-nuclear steam rush-transfer of a sodium-cooled fast reactor steam turbine is characterized by comprising the following steps:
the superheated steam output by the auxiliary steam header (2) is divided into three paths, wherein one path of the superheated steam enters the deaerator (3), the other path of the superheated steam enters the high-pressure heater (5), and the other path of the superheated steam is supplied to the steam turbine shaft seal system 11. Feed water is deoxidized and preheated by a deaerator (3), then enters a heater for preheating, then sequentially enters a heat absorption side of a steam generator (7) and a heat absorption side of a superheater (8) for absorbing heat to form superheated steam, and finally enters a steam turbine (10) for acting;
working media output by the first loop/the second loop (9) of the reactor enter the heat release side of the steam generator (7) and the heat release side of the superheater (8) in sequence to release heat, and then enter the first loop/the second loop (9) of the reactor.
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| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202010974599.4A CN111963264A (en) | 2020-09-16 | 2020-09-16 | System and method for non-nuclear steam rush-transfer of sodium-cooled fast reactor steam turbine |
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| CN202010974599.4A CN111963264A (en) | 2020-09-16 | 2020-09-16 | System and method for non-nuclear steam rush-transfer of sodium-cooled fast reactor steam turbine |
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Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN113819453A (en) * | 2021-10-28 | 2021-12-21 | 华能山东石岛湾核电有限公司 | Device and method for increasing feed water temperature in starting stage of high-temperature gas cooled reactor |
| CN114165778A (en) * | 2021-11-04 | 2022-03-11 | 华能核能技术研究院有限公司 | High-temperature gas cooled reactor secondary loop system and method for improving main water supply operation temperature |
| CN115083646A (en) * | 2022-06-23 | 2022-09-20 | 华能核能技术研究院有限公司 | Method for quickly cooling steam generator after emergency shutdown of high-temperature gas cooled reactor |
-
2020
- 2020-09-16 CN CN202010974599.4A patent/CN111963264A/en active Pending
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN113819453A (en) * | 2021-10-28 | 2021-12-21 | 华能山东石岛湾核电有限公司 | Device and method for increasing feed water temperature in starting stage of high-temperature gas cooled reactor |
| CN114165778A (en) * | 2021-11-04 | 2022-03-11 | 华能核能技术研究院有限公司 | High-temperature gas cooled reactor secondary loop system and method for improving main water supply operation temperature |
| CN115083646A (en) * | 2022-06-23 | 2022-09-20 | 华能核能技术研究院有限公司 | Method for quickly cooling steam generator after emergency shutdown of high-temperature gas cooled reactor |
| CN115083646B (en) * | 2022-06-23 | 2023-06-27 | 华能核能技术研究院有限公司 | Method for rapidly cooling steam generator after emergency shutdown of high-temperature gas cooled reactor |
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