CROSS-REFERENCE TO RELATED APPLICATION

Pursuant to 35 U.S.C. § 119(a), this application claims the benefit of the earlier filing date and the right of priority to Korean Patent Applications No. 10-2017-0074396, filed on Jun. 13, 2017, the contents of which are incorporated by reference herein in their entirety.

BACKGROUND

1. Technical Field

The present disclosure relates to a reactor cooling method, and more particularly, to power production using heat generated in a core and transferred to a reactor or reactor coolant system during a normal operation, and emergency power production and reactor cooling using heat generated in the core and transferred to the reactor or the reactor coolant system during an accident.

2. Description of the Related Art

Nuclear reactors are divided into loop type reactors (e.g., commercial reactors: Korea) in which major components (steam generator, pressurizer, pump, etc.) are installed outside a reactor vessel and integral reactors (e.g., SMART reactors: Korea) in which the major components are installed inside a reactor vessel.

In addition, Nuclear power plants are divided into active plants and passive plants depending on the implementation of a safety system. An active plant is a reactor using an active component such as a pump operated by electric power of an emergency diesel generator (EDG) or the like to drive a safety system, and a passive plant is a plant using a passive component operated by gravity, gas pressure or the like to drive a safety system.

A passive safety system in a passive plant may maintain the reactor in a safe manner only with a natural force built in the system without an operator action or an AC power source of safety grade such as an emergency diesel generator for more than a period of time (72 hours) required by regulatory requirements in the event of an accident, After 72 hours, using an operator action and a non-safety systems might be allowed to maintain the function of the safety systems and an emergency DC power source (battery).

Unlike a general thermal power plant where heat generation is stopped when fuel supply is stopped, a reactor in a nuclear power plant generates residual heat from a reactor core for a significant period of time by a fission product produced and accumulated during a normal operation even when a fission reaction is stopped in the reactor core. Accordingly, a variety of safety systems for removing the residual heat of the core during an accident are installed in the nuclear power plant.

In case of an active nuclear power plant (Conventional Nuclear Power Plant of Korea), a plurality of emergency diesel generators are provided in preparation for a case of interruption of electric power supply from the inside or outside at the time in an accident, and most active nuclear power plants use a pump to circulate cooling water, and thus a large-capacity emergency AC power source (a diesel generator) is provided due to the high power requirements of those active components. An operator action allowance time for an active nuclear reactor is estimated about 30 minutes.

In order to exclude active components such as a pump that requires a large amount of electricity, a driven force such as gas pressure or gravity is introduced in a passive nuclear reactor (U.S. Westinghouse AP1000, Korean SMART) that has been developed or is being developed to enhance the safety of the nuclear power plant, and thus a large amount of power is not required other than small components such as a valve, which is essentially required for the operation of a passive safety system. However, to enhance the safety in a passive nuclear power plant, an operator action allowance time is drastically extended from 30 minutes to 72 hours or longer, and an emergency active power source (diesel generator) is excluded, and an emergency DC power source (battery) is adopted. And thus the emergency DC power source should be maintained for more than 72 hours. Therefore, the emergency power source capacity required per unit time in a passive nuclear power plant is relatively small compared to an active nuclear power plant, but it is very large in terms of the battery capacity because the emergency power should be supplied for 72 hours or more.

In the other hand, a residual heat removal system (auxiliary feedwater system or passive residual heat removal system) is employed as a system for removing the heat of a reactor coolant system (the sensible heat of the reactor coolant system and the residual heat of the core) using a residual heat removal heat exchanger connected to a primary system or secondary system when an accident occurs in various nuclear power plants including an integral reactor. (AP1000: U.S. Westinghouse, commercial loop type nuclear power plant and SMART reactor: Korea)

Furthermore, a safety injection system is employed as a system for directly injecting cooling water into the reactor coolant system in case of a loss-of-coolant accident to maintain a water level of the reactor core and removing the heat of the reactor coolant system (the sensible heat of the reactor coolant system and the residual heat of the core) using the injected cooling water. (AP1000: U.S. Westinghouse, commercial loop type and SMART reactor: Korea)

Moreover, a reactor containment cooling system or spray system is a system for condensing steam using cooling or spraying to suppress a pressure rise when a pressure inside the reactor containment rises due to an accident such as a loss-of-coolant accident or a steam-line-break accident. Additionally, there are a method of directly spraying cooling water into the reactor containment (commercial loop type reactor: Korea), a method of inducing steam discharged in the reactor containment into a suppression tank (commercial boiling water reactor), a method of using a heat exchanger installed inside or outside the reactor containment (reinforced concrete containment building)) (APR+: Korea), a method of using a surface of the steel containment vessel as a heat exchanger (AP1000: U.S. Westinghouse), or the like.

As described above, various safety systems configured with multiple trains with two or more trains are installed in each system such as a residual heat removal system and a safety injection system for cooling the reactor coolant system (including the reactor vessel) to protect the reactor core at the time of an accident. However, in recent years, there has been a growing demand for safety enhancement of nuclear power plants due to the impact of Fukushima nuclear power plant (boiling water reactor) accident and the like, and thus there is a rising demand for safety facilities against a severe accident such as an external reactor vessel cooling system even in a pressurized water reactor (PWR) with a very low risk of leakage of large amounts of radioactive materials due to employing a very large-internal-volume nuclear reactor containment.

In detail, various safety facilities are provided to relieve an accident in case of the accident. In addition, each of the safety facilities is configured with multiple trains, and the probability that all systems fail simultaneously is very small. However, as a public requirement for the safety of nuclear power plants increases, safety facilities have been enhanced in preparation for a severe accident even with a very low probability of occurrence.

The external reactor vessel cooling system is a system provided to cool the outside of reactor vessel during core meltdown to prevent damage of the reactor vessel, assuming that a serious damage occurs in the core cooling function and a severe accident that the core is melted occurs since various safety facilities do not adequately perform functions due to multiple failure causes at the time of an accident. (AP1000 U.S. Westinghouse)

When the reactor vessel is damaged, a large amount of radioactive material may be discharged into the reactor containment, and a pressure inside the reactor containment may rise due to an large amount of steam generated by corium (melted core)-water reaction and gas formed by the core melt-concrete reaction. The reactor containment serves as a final barrier to prevent radioactive materials from being discharged into an external environment during an accident. When the reactor containment is damaged due to an increase in internal pressure, a large amount of radioactive material may be released to an external environment. Therefore, the external reactor vessel cooling system performs a very important function of suppressing radioactive materials from being discharged into the reactor containment and an increase of the internal pressure during a severe accident to prevent radioactive materials from being discharged into an external environment.

The external reactor vessel cooling system which is adopted in many countries is a system in which cooling water is filled in the reactor cavity located at a lower part of the reactor vessel and the cooling water is introduced into the cooling flow path in a space between the thermal insulation material and the reactor vessel and then steam is discharged to an upper part of the cooling flow path. In addition, a method of injecting a liquid metal at the time of an accident to mitigate the critical heat flux phenomenon, a method of pressurized cooling water to induce single phase heat transfer, a method of modifying a surface of the external reactor vessel to increase the heat transfer efficiency, a method of forming a forced flow, and the like, may be taken into consideration.

On the other hand, Stirling engine was developed by Robert Stirling (1816) as an external combustion engine, which tightly holds gas in a closed cylinder and drives an actuator (a type of piston) and a piston according to the strokes of heating, expansion, cooling and compression to produce power.

The Stirling engine is classified into α type, β type, γ type, and dual-acting type according to the configuration of the cylinders and the pistons, and classified into a mechanically driven Stirling engine (kinematic engine) and a free piston Stirling engine (FPSE), and the like depending on the piston movement.

Stirling engines may also operate with heat having a small temperature difference (e.g., 2° C.) and have a very high theoretical efficiency and low noise and low vibration compared to internal combustion engines.

In addition, the Stirling engine may use various heat sources such as solar heat and geothermal heat, and the use of environmentally friendly heat sources has an advantage of low discharge of pollutants.

In an external reactor vessel cooling system in the related art, since a thermal insulation material has to perform an appropriate thermal insulation function during a normal operation of the nuclear power plant, a flow path is sealed such that the inlet and outlet flow paths formed in the thermal insulating material at the time of an accident must be properly opened in a timely manner, and there is a delay time for filling the reactor cavity, and the heat removal ability may be reduced due to a critical heat flux phenomenon or the like while evaporating cooling water to form a steam layer on the external reactor vessel.

In addition, there is also a research on cooling the external reactor using a liquid metal, but the liquid metal method has difficulties in the maintenance of the liquid metal. In addition, the method of cooling the external reactor using a pressurization method has difficulties in the application of a natural circulation flow, and the method of modifying the reactor vessel surface has difficulties in the fabrication and maintenance of the surface, and the forced flow method has a disadvantage in that it must be supplied with electric power.

On the other hand, the large-capacity steam turbine method has a large size of the facility, thus increasing the cost when the strengthened seismic design is applied thereto. Therefore, there is a limitation in being designed to produce electric power during a normal operation and during an accident of the nuclear power plant.

In addition, since an external reactor vessel cooling system in the related art is operated by an operator action at the time of an accident, various instruments and components for monitoring the accident are required for the operation, and a probability that a system in a standby mode fails to operate at the time of an accident is higher than a probability that a system being operated is stopped to operate at the time of an accident.

Accordingly, the present disclosure presents a reactor cooling and power generation system in which a large-scale turbine power generation facility in the related art is maintained almost same design, and a small-scale power generation facility including the Sterling engine is additionally installed to receive heat generated and discharged from the core during a normal operation or during an accident of the nuclear power plant.

SUMMARY

An object of the present disclosure is to provide a reactor cooling and power generation system having system reliability in which safety class or seismic design are easily applicable, and reactor cooling is carried out while continuously operating during a normal operation as well as during an accident to produce emergency power.

Another object of the present disclosure is to propose a reactor cooling and power generation system having enhanced safety in which residual heat of a certain scale or more is removed during a normal operation as well as during an accident.

Still another object of the present disclosure is to propose a nuclear power plant having economic efficiency and safety due to the downsizing and reliability enhancement of an emergency power system of the nuclear power plant.

A reactor cooling and power generation system according to the present disclosure may include a reactor vessel, a heat exchange section formed to receive heat generated from a core inside the reactor vessel through a fluid, and an electric power production section comprising a Sterling engine formed to produce electric energy using the energy of the fluid whose temperature has increased while receiving the heat of the reactor, wherein the system is formed to circulate the fluid that has received heat from the core through the electric power production section, and operate even during a normal operation and during an accident of the nuclear power plant to produce electric power.

According to an embodiment, the electric power produced during the normal operation of the nuclear power plant may be supplied to an internal and external electric power system and an emergency battery.

According to an embodiment, the electric energy charged in the emergency battery may be formed to supply an emergency electric power as an emergency power source during an accident. Furthermore, the electric power produced during an accident of the nuclear power plant may be formed to be supplied to an emergency power source of the nuclear power plant.

According to an embodiment, the emergency power source may be formed to be supply an electric power source for the operation of a nuclear safety system or valve manipulating for the operation of the nuclear safety system or monitoring the nuclear safety system or operation of the reactor cooling and power generation system during an accident of the nuclear power plant.

According to an embodiment, a seismic design of seismic category I, II or III may be applied thereto, and a safety grade of safety class 1, 2 or 3 may be applied thereto.

According to an embodiment, the system may further include a first discharge section connected to the heat exchange section, wherein the first discharge section is formed to allow at least a part of the fluid excessively supplied to the electric power production section to bypass the electric power production section.

According to an embodiment, the heat exchange section may be formed to surround at least a part of the reactor vessel, and may have a shape capable of cooling an outer wall of the reactor vessel formed to receive heat discharged from the reactor vessel that has received heat generated from the core. Furthermore, at least a part of the shape of the heat exchange section having a shape of cooling the outer wall of the reactor vessel may include a cylindrical shape, a hemispherical shape, a double vessel shape, or a mixed shape thereof.

According to an embodiment, the system may be connected to an in-containment refueling water storage tank (IRWST) to supply refueling water to the heat exchange section having a shape capable of cooling the outer wall of the reactor vessel. Furthermore, a second discharge section may be provided in a heat exchange section having a shape capable of cooling the outer wall of the reactor vessel, and the second discharge section may be formed to discharge the refueling water supplied from the in-containment refueling water storage tank (IRWST).

According to an embodiment, a coating member may further be formed on the heat exchange section having a shape capable of cooling the outer wall of the reactor vessel to prevent the corrosion of the reactor vessel. A surface of the coating member may be chemically treated to increase a surface area thereof. Furthermore, a heat transfer member may further be formed to efficiently transfer heat discharged from the reactor vessel. A surface of the heat transfer member may be chemically treated to increase a surface area thereof.

According to an embodiment, the heat exchange section may be provided inside the reactor vessel, and may have a shape capable of cooling an inside of the reactor vessel formed to receive heat discharged from a reactor coolant system inside the reactor vessel that has received heat generated from the core.

According to an embodiment, the system may be connected to an in-containment refueling water storage tank (IRWST) to supply refueling water to the heat exchange section having a shape capable of cooling an inside of the reactor vessel. Furthermore, a second discharge section may be provided in a heat exchange section having a shape capable of cooling the inside of the reactor vessel, and the second discharge section may be formed to discharge the frefueling water supplied from the in-containment refueling water storage tank (IRWST).

According to an embodiment, the system may further include an evaporator section connected to the heat exchange section, wherein the evaporator section is formed to exchange heat with an inner fluid of the heat exchange section and an inner fluid of the electric power production section, and comprises a first circulation section formed to circulate through the heat exchange section and the evaporator section; and a second circulation section formed to circulate through the evaporator section and the electric power production section.

According to an embodiment, at least one of the first circulation section and the second circulation section may be formed to circulate by a single-phase fluid.

According to an embodiment, the heat exchange section may further include a core catcher, and the core catcher may be formed to receive and cool a melted core when the core is melted in the reactor vessel.

According to an embodiment, the Stirling engine may include a power generation section including a cylinder having a reciprocator and a piston configured to generate motive power by heat received through the fluid that has received heat, and a power transmission section, and an electricity generation section configured to convert mechanical energy generated by the power generation section into electrical energy.

According to an embodiment, the Stirling engine may include a high temperature section and a low temperature section respectively filled with working gas, and formed with spaces partitioned from each other inside a cylinder, and working gas filled in the high temperature section and the low temperature section may be formed to communicate with each other, and formed to move a reciprocator and a piston according to the communication of the working gas.

According to an embodiment, the Stirling engine may further include a regenerative heat exchange section, and the regenerative heat exchange section may transfer and store heat stored in the working gas in the regenerative heat exchange section when the working gas moves from the high temperature section to the low temperature section, and transfer the heat stored in the regenerative heat exchange section to the working gas when the working gas returns from the low temperature section to the high temperature section.

According to an embodiment, a fan or a pump may be provided in the low temperature section, and the fan or the pump may be formed to supply a cooling fluid to the low temperature section to exchange heat with the working gas of the low temperature section. The cooling fluid may include air, pure water, seawater, or a mixture thereof.

According to an embodiment, the system may further a condensate storage section at a lower portion of the electric power production section to collect condensate generated by condensing the fluid heat-exchanged in the electric power production section.

According to an embodiment, condensate in the condensate storage section may be supplied to the heat exchange section by gravity or the power of the pump.

A nuclear power plant according to the present disclosure may include a reactor vessel, a heat exchange section formed to receive heat generated from a core inside the reactor vessel through a fluid, and an electric power production section comprising a Sterling engine formed to produce electric energy using the energy of the fluid whose temperature has increased while receiving the heat of the reactor, wherein the system is formed to circulate the fluid that has received heat from the core in the heat exchange section through the electric power production section, and formed to operate even during a normal operation and during an accident of the nuclear power plant to produce electric power.

The reactor cooling and power generation system according to the present disclosure is formed to drive an electric power production section including a Sterling engine formed to produce electric energy using the energy of a fluid in a small scale facility. A heat exchange section and an electric power production section of the present disclosure may continuously operate not only during normal operation but also during an accident to cool residual heat and generate emergency power, thereby improving system reliability. A heat exchange section for facilitating the application of safety class and seismic design with a small scale facility, and continuously operating during an accident as well as during a normal operation and performing reactor cooling by the application of safety class or seismic design may be included therein, thereby improving the reliability of the nuclear power plant.

The reactor cooling and power generation system according to the present disclosure may be designed to remove residual heat of a certain scale or more generated from the core of the reactor, and continuously operated not only during a normal operation but also during an accident to reduce a probability of actuation failure at the time of an accident, thereby improving the safety of the nuclear power plant.

The nuclear power plant according to the present disclosure may improve the economic efficiency of the nuclear power plant through the downsizing of an emergency power system through the reactor cooling and power generation system.

