System and method for taking heat after power failure accident of whole nuclear power plant

Abstract

The invention provides a system and method for carrying heat after a power failure accident of a whole nuclear power plant. The system comprises a pressure container, a voltage stabilizer, a steam generator and a PRS heat exchanger. Wherein the pressure container and the steam generator form water circulation, a first emergency water injection pump is arranged on a path from the steam generator to the pressure container, a generator is arranged on one side of the water injection pump, and the generator is electrically connected with the first emergency water injection pump; a path from the pressure container to the steam generator is communicated with a pressure stabilizer; the air inlet end of the PRS heat exchanger communicates with the top of the steam generator, the air outlet end of the PRS heat exchanger communicates with the middle of the steam generator, and a second emergency water injection pump is arranged on the communicating path of the air outlet end of the PRS heat exchanger and the middle of the steam generator. A safety valve group and a quick pressure relief valve group are arranged at the top of the pressure stabilizer; and an atmosphere discharge valve is arranged at the top of the steam generator. The investment of the nuclear power plant is reduced on the premise that the safety of the nuclear power plant is not reduced, and a theoretical foundation is laid for subsequent practical engineering application.

Description

Technical Field

[0001] The present application belongs to the field of nuclear facility technology, and more specifically, relates to a system and method for providing heat after a power outage in a nuclear power plant, and is particularly suitable for a strategy for providing heat after a power outage in a nuclear power plant without a steam-driven auxiliary feedwater pump. Background Art

[0002] The auxiliary feedwater system is mainly used to supply an appropriate amount of water to the steam generator to discharge the core decay heat and ensure that the steam generator is not overflowed in partial failure modes of the reactor, such as rupture of the main feedwater and main steam pipelines during power operation, and complete failure of the normal feedwater system.

[0003] Among them, the steam-driven auxiliary feedwater pump is mainly used to maintain water injection into the steam generator by pushing the steam in the steam generator when the whole plant is out of power and the power supply cannot be restored for a long time. Steam-driven auxiliary feedwater pumps are equipped in some nuclear power plants, which play a huge role in ensuring the nuclear safety of the reactor.

[0004] However, the steam-driven auxiliary feedwater pump has a complex structure and has long been dependent on imports. The purchase cost of a single unit is more than 10 million yuan higher than that of an electric auxiliary feedwater pump. In addition, the maintenance process of the steam-driven auxiliary feedwater pump is complex and the economic cost is high in daily operation. Therefore, how to cancel the steam-driven auxiliary feedwater pump without reducing the safety of the reactor and find a feasible alternative heat-carrying strategy is an urgent but difficult proposition. Summary of the invention

[0005] The technical problem to be solved by the present invention is that the existing pneumatic auxiliary water supply pump has high purchase cost, and due to its complex structure, its maintenance process is complicated and the economic cost is high in daily operation.

[0006] The first purpose of the present invention is to provide a heat-carrying system suitable for a nuclear power plant power outage accident without a steam-driven auxiliary feedwater pump, comprising a pressure vessel, a pressurizer, a steam generator, and a PRS heat exchanger;

[0007] The pressure vessel and the primary side of the steam generator form a water cycle, and the pressure stabilizer is provided on the path from the pressure vessel to the primary side of a steam generator; the steam end of the PRS heat exchanger is connected to the outlet pipe of the steam generator, and the condensate end of the PRS heat exchanger is connected to the inlet pipe of the steam generator;

[0008] A first emergency water injection pump and a second emergency water injection pump are respectively arranged on the path flowing to the pressure vessel and the secondary side of the steam generator, and the first and second emergency water injection pumps are respectively connected to generators for power supply;

[0009] A safety valve group and a quick pressure relief valve group are arranged on the top of the pressure stabilizer;

[0010] An atmospheric discharge valve group is arranged at the outlet of the steam generator and is communicated with the atmosphere outside the containment vessel.

[0011] As a possible design, the generator is a SBO diesel engine.

[0012] As a possible design, the top connecting pipe on the secondary side of the steam generator is provided with a main steam safety valve group and an atmospheric discharge valve group in sequence, and a main steam isolation valve is provided at the end of the pipe.

[0013] As a possible design, the PRS heat exchanger includes a heat transfer pipe and a water tank, and the heat transfer pipe is immersed in the PRS water tank.

[0014] As a possible design, the pressure stabilizer is connected to the pressure relief box through a pressure stabilizer safety valve group and a quick pressure relief valve group.

[0015] As a possible design, there are multiple valves downstream of the PRS heat exchanger and they are arranged in parallel.

[0016] As a possible design, the number of downstream valves of the PRS heat exchanger is two.

[0017] The second purpose of the present invention is to provide a method for providing heat after a power outage in a nuclear power plant without a steam-driven auxiliary feedwater pump, based on the system for providing heat after a power outage in a nuclear power plant described in any one of the above items;

[0018] The method for removing heat after a power outage in a nuclear power plant includes:

[0019] When the operating condition of the nuclear power plant during a power outage is any one of the modes 1-4 in Table 1, the downstream valve of the PRS heat exchanger automatically opens;

[0020] When the operating condition of the nuclear power plant during a power outage is mode 5 or 6 in Table 1, start the first emergency water injection pump and open the pressurizer safety valve group or the quick pressure relief valve group; or start the second emergency water injection pump and open the atmospheric discharge valve at the same time.

[0021] As a possible design, the injection pressure heads of the first and second emergency water injection pumps are 2.0-2.5 MPa and 1.0-2.0 MPa respectively.

[0022] As a possible design, the injection flow rates of the first and second emergency water injection pumps are 10-12 kg / s.

[0023] The beneficial effects of the present invention are:

[0024] 1. Without reducing the safety of nuclear power plants, the investment in nuclear power plants can be reduced by about 30 million yuan, which improves the economic competitiveness of nuclear power plants and lays a theoretical foundation for subsequent practical engineering applications.

[0025] 2. SBO diesel engines can utilize existing nuclear power plant processes, so no additional investment is required.

[0026] 3. The heat-carrying system disclosed in the present invention, which is applicable to a nuclear power plant without a steam-driven auxiliary feedwater pump and is simple, has high overall reliability, has no influence on the arrangement of other existing systems, and does not affect the design of other reactor systems. BRIEF DESCRIPTION OF THE DRAWINGS

[0027] In order to more clearly illustrate the technical solutions in the embodiments of the present application, the drawings required for use in the embodiments or the description of the prior art will be briefly introduced below. Obviously, the drawings described below are only some embodiments of the present application. For ordinary technicians in this field, other drawings can be obtained based on these drawings without paying creative labor.

[0028] Figure 1 Schematic diagram of the heat transfer system after a power outage in a nuclear power plant in the embodiment.

[0029] Among them, the reference numerals in the figure are:

[0030] 1-pressure vessel, 2-pressure regulator, 3-first emergency water injection pump, 4-steam generator, 5-PRS heat exchanger, 501-heat transfer pipe, 502-water tank, 6-second emergency water injection pump, 7-generator, 8-safety valve group, 9-quick pressure relief valve group, 10-atmospheric discharge valve, 11-buffer tank, 12-main steam safety valve, 13-main steam isolation valve, 14-pressure relief box, 15-downstream valve, 16-refueling water tank, 17-auxiliary water tank. DETAILED DESCRIPTION

[0031] In order to make the technical problems, technical solutions and beneficial effects to be solved by the present application more clearly understood, the present application is further described in detail below in conjunction with the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are only used to explain the present application and are not used to limit the present application.

[0032] It should be noted that when an element is referred to as being "fixed to" or "disposed on" another element, it can be directly on the other element or indirectly on the other element. When an element is referred to as being "connected to" another element, it can be directly connected to the other element or indirectly connected to the other element.

[0033] It should be understood that the terms "length", "width", "up", "down", "front", "back", "left", "right", "vertical", "horizontal", "top", "bottom", "inside", "outside", etc., indicating the orientation or position relationship, are based on the orientation or position relationship shown in the drawings, and are only for the convenience of describing the present application and simplifying the description, and do not indicate or imply that the device or element referred to must have a specific orientation, be constructed and operated in a specific orientation, and therefore should not be understood as a limitation on the present application.

[0034] In addition, the terms "first" and "second" are used for descriptive purposes only and should not be understood as indicating or implying relative importance or implicitly indicating the number of the indicated technical features. Thus, a feature defined as "first" or "second" may explicitly or implicitly include one or more of the features. In the description of this application, the meaning of "plurality" is two or more, unless otherwise clearly and specifically defined.

[0035] At present, when the whole nuclear power plant is out of power and the power supply cannot be restored for a long time, the steam-driven auxiliary feedwater pump is generally used to maintain water injection into the steam generator through the steam in the steam generator. However, the steam-driven auxiliary feedwater pump has a complex composition and has long been dependent on imports. The purchase cost of a single unit is more than 10 million yuan higher than that of an electric auxiliary feedwater pump. In addition, the maintenance process of the steam-driven auxiliary feedwater pump is complex and the economic cost is high in daily operation. Therefore, it is necessary to study an alternative solution to reduce costs while ensuring safety.

[0036] Based on the above problems, Figure 1 As shown, the present invention provides a heat-carrying system suitable for a power outage accident in a nuclear power plant without a steam-driven auxiliary feedwater pump, comprising a pressure vessel 1, a pressurizer 2, a steam generator 4 and a PRS heat exchanger 5; wherein, the pressure vessel 1 and the primary side of the steam generator 4 form a water cycle, a first emergency water injection pump 3 is arranged on the path from the steam generator 4 to the pressure vessel 1, the first emergency water injection pump 3 takes water from a refueling water tank 16, and the first emergency water injection pump 3 is powered by a generator 7; the pressure vessel 1 is connected to the pressurizer 2 on the path from the pressure vessel 1 to one of the steam generators 4; the steam end of the PRS heat exchanger 5 is connected to the outlet pipe of the steam generator 4, and the condensate end of the PRS heat exchanger 5 is connected to the inlet pipe of the steam generator 4, a second emergency water injection pump 6 is arranged on the path from the condensate end of the PRS heat exchanger 5 to the inlet pipe of the steam generator 4, the second emergency water injection pump 6 takes water from an auxiliary feedwater tank 17, and is also powered by the generator 7;

[0037] A safety valve group 8 and a quick pressure relief valve group 9 are provided on the top of the pressure stabilizer 2;

[0038] An atmospheric discharge valve 10 is disposed on the top of the steam generator 4 .

[0039] The operation mode of a nuclear power plant when the entire plant is out of power is shown in Table 1.

[0040] Table 1 Comparison table of operation modes and standard operating conditions

[0041]

[0042] It should be noted that in Table 1, [1] means that decay heat is not included. [2] means that Pn is the rated power. [3] means that all bolts of the reactor pressure vessel top cover are in a fully tightened state. [4] means that one or more bolts of the reactor pressure vessel top cover are not in a fully tightened state.

[0043] In the present invention, when the operation mode of the nuclear power plant is in any mode of mode 1-4, the operation mode of the heat system is: the PRS heat exchanger 5 is started, and the residual heat of the core is exported to the PRS water tank 502 by natural circulation. It should be noted that in actual application, the start-up of the PRS heat exchanger 5 can be designed to be an automatic mode, that is, after the power is cut off, when the intelligent control system determines that the operation mode of the nuclear power plant is any one of 1-4, the intelligent control system immediately opens the electric isolation valve downstream of the PRS heat exchanger 5, and immediately performs heat exchange to avoid the accumulation of heat in the core. It should be noted that the intelligent control system is the original control system of the nuclear power plant and only adds the program for starting the electric isolation valve. This operation belongs to the routine operation of those skilled in the art, that is, no creative work is required.

[0044] When the operation mode of a nuclear power plant is in mode 5 or 6, if a power outage occurs in the entire plant and the power supply cannot be restored for a long time, there are two ways to remove the residual heat from the core.

[0045] The first method is that the operator manually starts the first emergency water injection pump 3, draws water from the internal refueling water tank 16, and injects water into the primary circuit through the pipeline. At the same time, the operator manually opens the safety valve group 8 or the quick pressure relief valve group 9 of the pressurizer 2 to remove the residual heat of the core through the filling and discharging method of the primary circuit.

[0046] The second method is that the operator manually starts the second emergency water injection pump 6, draws water from the auxiliary water supply tank 17, and injects water into the secondary side of the steam generator 4 through the water supply pipeline. The operator manually opens the atmospheric discharge valve 10 on the steam pipeline to remove the residual heat of the core through the secondary circuit filling and discharging method.

[0047] It should be noted that the primary circuit refers to the water circuit / gas circuit formed by the first emergency water injection pump 3 , the pressure vessel 1 , the primary side of the steam generator 4 and the pressure stabilizer 2 .

[0048] The secondary circuit refers to the water circuit / gas circuit formed by the second emergency water injection pump 6, the pressure vessel 1, the secondary side of the steam generator 4, the atmospheric discharge valve group and the main steam isolation valve 13 group.

[0049] It should be noted that the layout diagram of the nuclear power plant is already available.

[0050] As a possible implementation, Figure 1 As shown, the generator 7 is generally a power generation device commonly used in the art, such as a diesel generator 7, and specifically may be an SBO diesel engine.

[0051] As a possible implementation, Figure 1 As shown, the top outlet of the steam generator 4 is provided with a main steam safety valve group 12 and an atmospheric discharge valve group 10 through a bypass pipeline, and the end of the top outlet pipeline of the steam generator 4 is provided with a main steam isolation valve 13, which is connected to the steam turbine. While achieving rapid steam release, the safety of the release process and the safety of other equipment can also be guaranteed.

[0052] As a possible implementation, Figure 1 As shown, the PRS heat exchanger 5 comprises a heat transfer pipe 501 and a water tank 502, wherein the heat transfer pipe 501 is immersed in the water tank 502. The purpose of heat exchange can be achieved through a simple heat exchange method, and it is worthy of popularization and use.

[0053] As a possible implementation, Figure 1 As shown, a pressure relief box 14 is provided beside the pressure stabilizer 2, and the pressure stabilizer 2 is connected to the pressure relief box 14 through the safety valve group 8 and the quick pressure relief valve group 9, which is conducive to achieving safe, stable and rapid pressure relief.

[0054] As a possible implementation, Figure 1 As shown, there are two downstream valves 15 of the PRS heat exchanger 5 and they are arranged in parallel to avoid failure of one of the valves, which affects the timely heat exchange.

[0055] The present invention also discloses a heat transfer method based on the heat transfer system after a power outage in the entire nuclear power plant, which is specifically as follows:

[0056] When the operating condition of the nuclear power plant during a power outage is any one of modes 1-4 in Table 1, the downstream valve 15 of the PRS heat exchanger 5 is opened.

[0057] When the operating condition of the nuclear power plant during a power outage is mode 5 or 6 in Table 1, the first emergency water injection pump 3 and the safety valve group 8 or the quick pressure relief valve group 9 of the pressurizer 2 are started, and at the same time, the downstream valve 15 of the PRS heat exchanger 5 will not automatically open the second emergency water injection pump or; the second emergency water injection pump 6 is started, the safety valve group 8 and / or the quick pressure relief valve group 9 of the pressurizer 2 are closed, and the atmospheric exhaust valve is opened at the same time.

[0058] It should be noted that the injection pressure heads of the first emergency water injection pump 3 and the second emergency water injection pump 6 are generally 2.0-2.5 MPa and 1.0-2.0 MPa, for example, 1.0 MPa, 1.2 MPa, 1.5 MPa, 1.7 MPa, 2.0 MPa, 2.1 MPa, 2.2 MPa, 2.3 MPa, 2.4 MPa, 2.5 MPa and any value between any two values. The injection flow rate is generally 8-12 kg / s, for example, 8 kg / s, 9 kg / s, 10 kg / s, 11 kg / s, 12 kg / s and any value between any two values.

[0059] 1. The present invention discloses a heat removal system and method suitable for a power outage accident in a nuclear power plant without a steam-driven auxiliary feedwater pump. The system and method realize the removal of decay heat after a power outage accident in a nuclear power plant by appropriately expanding the application scope of the original safety system and appropriately adding some equipment. The system can reduce the investment in the nuclear power plant by about 30 million yuan without reducing the safety of the nuclear power plant, thereby improving the economic competitiveness of the nuclear power plant and laying a theoretical foundation for subsequent practical engineering applications.

[0060] 2. The present invention discloses a heating system suitable for a nuclear power plant power outage without a steam-driven auxiliary feedwater pump. The operating conditions, flow rate and pressure head of the newly added water injection pump have been considered and calculated in detail. The power is within the range that can be provided by the existing equipment, and the power of the existing SBO diesel engine will not be required to be increased.

[0061] 3. The present invention discloses a heating system suitable for a nuclear power plant without a steam-driven auxiliary feedwater pump after a power outage. The system has a simple layout, fewer devices, and a higher overall system reliability. It has no effect on the layout of other existing systems and does not affect the design of other reactor systems.

[0062] The specific implementation methods described above further illustrate the objectives, technical solutions and beneficial effects of the present invention in detail. It should be understood that the above description is only a specific implementation method of the present invention and is not intended to limit the scope of protection of the present invention. Any modifications, equivalent substitutions, improvements, etc. made within the spirit and principles of the present invention should be included in the scope of protection of the present invention.

Claims

1. A heat-carrying system suitable for a nuclear power plant without a steam-driven auxiliary feedwater pump after a power outage, characterized in that: Including pressure vessels, pressurizers, steam generators, and PRS heat exchangers; The pressure vessel and the primary side of the steam generator form a water cycle, and the pressure stabilizer is provided on the path from the pressure vessel to the primary side of one of the steam generators; the steam end of the PRS heat exchanger is connected to the outlet pipe of the steam generator, and the condensate end of the PRS heat exchanger is connected to the inlet pipe of the steam generator; A first emergency water injection pump and a second emergency water injection pump are respectively arranged on the path flowing to the pressure vessel and the secondary side of the steam generator, and the first and second emergency water injection pumps are respectively connected to generators for power supply; A safety valve group and a quick pressure relief valve group are provided on the top of the pressure stabilizer to release pressure; An atmospheric discharge valve group is arranged at the outlet of the steam generator and is communicated with the atmosphere outside the containment vessel.

2. According to claim 1, a heat-generating system suitable for a nuclear power plant power outage accident without a steam-driven auxiliary feedwater pump, characterized in that: The generator is a SBO diesel engine.

3. The heat-generating system for a nuclear power plant after a power outage without a steam-driven auxiliary feedwater pump according to claim 1, characterized in that: The top connecting pipeline on the secondary side of the steam generator is provided with a main steam safety valve group and an atmospheric discharge valve group in sequence, and a main steam isolation valve is provided at the end of the pipeline.

4. The heat-generating system for a nuclear power plant after a power outage without a steam-driven auxiliary feedwater pump according to claim 1, characterized in that: The PRS heat exchanger includes a heat transfer pipe and a water tank. The heat transfer pipe is immersed in the PRS water tank.

5. The heat-generating system for a nuclear power plant after a power outage without a steam-driven auxiliary feedwater pump according to claim 1, characterized in that: The pressure stabilizer is connected to the pressure relief box through the pressure stabilizer safety valve group and the quick pressure relief valve group to relieve pressure.

6. The heat-generating system for a nuclear power plant after a power outage without a steam-driven auxiliary feedwater pump according to claim 1, characterized in that: The number of downstream valves of the PRS heat exchanger is multiple and they are arranged in parallel.

7. A heat-generating system suitable for a nuclear power plant power outage after a power outage without a steam-driven auxiliary feedwater pump according to claim 6, characterized in that: The number of downstream valves of the PRS heat exchanger is two.

8. A method for removing heat after a power outage in a nuclear power plant without a steam-driven auxiliary feedwater pump, characterized in that: A system for carrying heat after a power outage in a nuclear power plant according to any one of claims 1 to 7; The method for removing heat after a power outage in a nuclear power plant includes: When the operating condition of the nuclear power plant during a plant-wide blackout is any one of the modes 1-4 in Table 1, the downstream valve of the PRS heat exchanger automatically opens; When the operating condition of the nuclear power plant during a power outage is mode 5 or 6 in Table 1, start the first emergency water injection pump and open the pressurizer safety valve group or the quick pressure relief valve group; or start the second emergency water injection pump and open the atmospheric discharge valve at the same time.

9. The method for removing heat after a power outage in a nuclear power plant without a steam-driven auxiliary feedwater pump according to claim 7 is characterized in that: The injection pressure heads of the first and second emergency water injection pumps are 2.0-2.5 MPa and 1.0-2.0 MPa respectively.

10. The method for removing heat after a power outage in a nuclear power plant without a steam-driven auxiliary feedwater pump according to claim 8, characterized in that: The injection flow rates of the first and second emergency water injection pumps are 10-12 kg / s.

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