CN114121326A - Supercritical or ultra-supercritical nuclear power generation system - Google Patents

Abstract

The invention provides a supercritical or ultra-supercritical nuclear power generation system, which comprises a reactor device, a steam power conversion device and a supercritical or ultra-supercritical boiler, wherein a first loop medium for carrying heat is arranged in the reactor device, the steam power conversion device is used for converting latent heat of a second loop medium into electric energy, the supercritical or ultra-supercritical boiler comprises a pressure-bearing container and a heat exchange device, the pressure-bearing container is provided with a heat exchange region, a loop medium inlet and a loop medium outlet, the loop medium inlet is communicated with the heat exchange region inlet and the reactor device outlet, and the loop medium outlet is communicated with the heat exchange region outlet and the reactor device inlet; the heat exchange device is positioned in the heat exchange area, the outlet of the heat exchange device is communicated with the inlet of the steam-electric power conversion device, and the second loop medium can exchange heat with the first loop medium in the heat exchange device. The supercritical or ultra-supercritical boiler is used for replacing a steam generator in a nuclear island of a high-temperature gas cooled reactor nuclear power plant, so that the power generation efficiency of the whole power generation system is improved.

Description

Supercritical or ultra-supercritical nuclear power generation system

Technical Field

The invention relates to the technical field of nuclear power generation, in particular to a supercritical or ultra-supercritical nuclear power generation system.

Background

Nuclear power generation technology is one of the main power generation technologies worldwide at present and has long been a key technology developed in various major countries. With the international requirement for the emission of greenhouse gas (CO2) becoming stricter, nuclear energy is one of the main clean energy sources, and the nuclear energy is once again in the process of important development. The high-temperature gas cooled reactor is used as a fourth-generation nuclear power generation technology, has the characteristics of inherent safety, high power generation efficiency and capability of developing process heat application, and becomes a hot spot researched and developed by major economic countries.

The nuclear island part of an existing high temperature gas cooled reactor nuclear power plant mainly comprises three parts, namely a reactor part, a steam generator and a helium fan part, and a hot gas guide pipe and a shell part thereof which connect the two parts. The primary function of the reactor section is to controllably convert nuclear energy into thermal energy of the primary loop cooling medium while ensuring intrinsic safety. The functions of the steam generator and the helium fan are to convert the heat energy of the first loop cooling medium into the heat energy of the second loop medium (water/steam) through heat exchange and generate qualified steam meeting the requirements of the steam turbine, and the helium fan provides the power for the flow of the first loop cooling medium. Because the steam generator is used as the pressure boundary of the first loop, the design, manufacture, inspection, test, acceptance and the like of the whole equipment are strictly carried out according to the requirements of the nuclear safety first-level component.

However, the existing steam generator has great conservation (fatigue-creep analysis of materials is performed by adopting a linear elastic method), only subcritical superheated steam can be generated by the highest heat exchange, and parameters of a heat exchange pipe and the steam are difficult to further improve, so that the further improvement of the power generation efficiency is limited.

Disclosure of Invention

The invention provides a supercritical or ultra-supercritical nuclear power generation system, which is used for further improving the power generation efficiency of a high-temperature gas-cooled reactor power generation system.

The invention provides a supercritical or ultra-supercritical nuclear power generation system, which comprises a reactor device, a steam power conversion device and a supercritical or ultra-supercritical boiler, wherein a first loop medium for carrying heat is arranged in the reactor device, the steam power conversion device is used for converting latent heat of a second loop medium into electric energy, and the supercritical or ultra-supercritical boiler comprises:

the pressure-bearing container is provided with a heat exchange region, a primary loop medium inlet for the first loop medium to enter and a primary loop medium outlet for the first loop medium to discharge, the primary loop medium inlet is communicated with the inlet of the heat exchange region and the outlet of the reactor device, and the primary loop medium outlet is communicated with the outlet of the heat exchange region and the inlet of the reactor device;

and the heat exchange device is positioned in the heat exchange area, the outlet of the heat exchange device is communicated with the inlet of the steam-power conversion device, the inlet of the heat exchange device is communicated with the water supply system of the steam-power conversion device, and a second loop medium in the heat exchange device can exchange heat with the first loop medium.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the system further comprises a hot gas guide pipe and a fan, wherein the hot gas guide pipe comprises:

a high-temperature medium pipeline, wherein a first end of the high-temperature medium pipeline is communicated with an outlet of the reactor device, and a second end of the high-temperature medium pipeline is communicated with a loop medium inlet of the supercritical or ultra-supercritical boiler;

a first end of the low-temperature medium pipeline is communicated with an inlet of the reactor device, and a second end of the low-temperature medium pipeline is communicated with a loop medium outlet of the supercritical or ultra-supercritical boiler;

the fan is arranged on the low-temperature medium pipeline, and the air supply direction of the fan is the direction from the supercritical or ultra-supercritical boiler to the reactor device.

According to the present invention, there is provided a supercritical or ultra supercritical nuclear power generation system, wherein the hot gas duct further comprises:

the first isolation valve is arranged on the high-temperature medium pipeline and used for controlling the on-off of the high-temperature medium pipeline;

and the second isolation valve is arranged on the low-temperature medium pipeline and is used for controlling the on-off of the low-temperature medium pipeline.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, at least two first isolation valves and at least two second isolation valves are arranged, and the fan is positioned between two adjacent second isolation valves.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the low-temperature medium pipeline is sleeved outside the high-temperature medium pipeline, and a gap is reserved between the low-temperature medium pipeline and the high-temperature medium pipeline so as to allow a low-temperature first loop medium to pass through.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, a plurality of supports are arranged between the low-temperature medium pipeline and the high-temperature medium pipeline and are used for positioning the high-temperature medium pipeline.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the inner wall of the high-temperature medium pipeline and/or the outer wall of the low-temperature medium pipeline is/are provided with the heat insulation layer.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the heat exchange device comprises a plurality of groups of heat exchange tube bundles which are sequentially communicated, the inlet of the first group of heat exchange tube bundles can be communicated with the water supply system of the steam power conversion device, and the last group of heat exchange tube bundles can be communicated with the inlet of the steam power conversion device.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the present invention, the supercritical or ultra-supercritical boiler further comprises:

the inlet of the first header is communicated with a water supply system of the steam-power conversion device, and the inlet of the first group of heat exchange tube bundles is communicated with the outlet of the first header;

an inlet of the second header is communicated with an outlet of the last group of heat exchange tube bundles, and an outlet of the second header is communicated with an inlet of the steam-electric power conversion device;

and the at least one third header is used for connecting outlets and inlets of two adjacent groups of heat exchange tube bundles.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the pressure-bearing container is internally provided with a mixing region of a first loop medium, the mixing region is positioned below the heat exchange region, and the mixing region is communicated with the inlet of the first loop medium.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the pressure-bearing container comprises a pressure-bearing shell and a hearth positioned in the pressure-bearing shell, the heat exchange region is positioned in the hearth, a gap for the first loop medium to pass is arranged between the hearth and the pressure-bearing shell, the gap is communicated with an outlet of the heat exchange region, an inlet of the first loop medium is arranged on the hearth, and an outlet of the first loop medium is arranged on the pressure-bearing shell and is communicated with the gap.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the pressure-bearing container is also provided with a liquid water receiving cavity, and the liquid water receiving cavity is positioned below the mixing area and is communicated with the mixing area.

According to the supercritical or ultra-supercritical nuclear power generation system provided by the invention, the top of the liquid water receiving cavity is provided with the water cooling pipe, and the water cooling pipe is connected with the first header.

According to the present invention there is provided a supercritical or ultra supercritical nuclear power generation system, the reactor apparatus comprising:

the reactor body comprises a reactor pressure vessel and a reactor positioned in the reactor pressure vessel, and the first loop medium is arranged in the reactor pressure vessel;

a containment system including a containment vessel, the containment vessel housed outside the reactor body.

The supercritical or ultra-supercritical nuclear power generation system provided by the invention can realize the combination of a high-temperature gas-cooled reactor technology and a thermal power supercritical or ultra-supercritical technology, not only can utilize the main characteristics of the inherent safety, the high-temperature energy supply and the modular construction of the high-temperature gas-cooled reactor, but also can fully utilize the supercritical or ultra-supercritical thermal power technology, and utilize a supercritical or ultra-supercritical boiler to replace a steam generator in a nuclear island of a high-temperature gas-cooled reactor nuclear power plant, thereby improving the power generation efficiency of the whole power generation system.

Drawings

In order to more clearly illustrate the technical solutions of the present invention or the prior art, the drawings needed to be used in the description of the embodiments or the prior art will be briefly described below, and it is obvious that the drawings in the following description are some embodiments of the present invention, and it is obvious for those skilled in the art to obtain other drawings based on these drawings without creative efforts.

FIG. 1 is a schematic structural diagram of a supercritical or ultra supercritical nuclear power generation system provided by the present invention;

FIG. 2 is a schematic structural view of a reactor apparatus provided by the present invention;

FIG. 3 is a schematic structural view of a hot gas duct provided by the present invention;

FIG. 4 is a schematic vertical cross-sectional view of a supercritical or ultra-supercritical boiler provided by the present invention;

FIG. 5 is a schematic cross-sectional view of a supercritical or ultra-supercritical boiler provided by the present invention;

fig. 6 is a schematic structural view of a steam-electric power conversion device provided by the present invention.

Reference numerals:

1: a reactor device; 2: a hot gas conduit; 3: supercritical or ultra supercritical boilers;

4: a steam-to-power conversion device; 11: a reactor body; 12: a containment vessel;

22: a first isolation valve; 23: a high temperature medium pipe; 24: a protective shell;

25: a second isolation valve; 26: a cryogenic medium conduit; 31: a prestressed concrete shell;

32: a hearth; 33: a heat exchange tube bundle; 34: a first header;

35: a mixing zone; 36: a liquid water receiving cavity; 37: a water-cooled tube;

38: a void; 39: a loop medium inlet; 310: a second header;

311: a third header; 41: a steam turbine; 42: a generator;

43: a condenser; 44: a condensate pump; 45: a low pressure heater;

46: a deaerator; 47: a feed pump; 48: a high pressure heater.

Detailed Description

In order to make the objects, technical solutions and advantages of the present invention clearer, the technical solutions of the present invention will be clearly and completely described below with reference to the accompanying drawings, and it is obvious that the described embodiments are some, but not all embodiments of the present invention. All other embodiments, which can be derived by a person skilled in the art from the embodiments given herein without making any creative effort, shall fall within the protection scope of the present invention.

The supercritical or ultra supercritical nuclear power generation system of the present invention will be described below with reference to fig. 1 to 6.

As shown in fig. 1, the supercritical or ultra-supercritical nuclear power generation system provided by the present invention includes a reactor device 1, a steam-power conversion device 4, and a supercritical or ultra-supercritical boiler 3. Wherein, a first loop medium is arranged in the reactor device 1 and is used for leading out the heat energy generated in the reactor device 1; the steam power conversion device 4 is used for converting latent heat of the second loop medium into electric energy to realize power generation; the supercritical or ultra-supercritical boiler 3 is used to exchange heat between the first loop medium and the second loop medium to transfer thermal energy in the first loop medium to the second loop medium.

And, the supercritical or ultra-supercritical boiler 3 includes a pressure-bearing vessel and a heat exchange device. Specifically, the pressure-bearing vessel is provided with a heat exchange region, a primary loop medium inlet 39 and a primary loop medium outlet, and the heat exchange device is located in the heat exchange region. A loop medium inlet 39 is communicated with the inlet of the heat exchange area and the outlet of the reactor device 1 so as to guide the first loop medium in the reactor device 1 into the heat exchange area; when the first loop medium passes through the heat exchange device, the second loop medium in the heat exchange device can exchange heat with the first loop medium, and the outlet of the heat exchange device is communicated with the inlet of the steam power conversion device 4, so that the second loop medium after heat exchange can be guided into the steam power conversion device 4, and the steam power conversion device 4 converts the latent heat of the second loop medium into electric energy. The outlet of the primary loop medium of the supercritical or ultra-supercritical boiler 3 is communicated with the outlet of the heat exchange area and the inlet of the reactor device 1 so as to lead out the primary loop medium after heat exchange with the secondary loop medium and lead the primary loop medium after heat exchange to flow back to the reactor device 1.

It should be noted that the first loop medium may be an inert gas, specifically, helium; the second loop medium may be water or steam.

By the arrangement, the combination of the high-temperature gas-cooled reactor technology and the thermal power supercritical or ultra-supercritical technology can be realized, the main characteristics of inherent safety, high-temperature energy supply and modular construction of the high-temperature gas-cooled reactor can be utilized, the supercritical or ultra-supercritical thermal power technology can be fully utilized, the supercritical or ultra-supercritical boiler 3 is utilized to replace a steam generator in a nuclear island of a high-temperature gas-cooled reactor nuclear power plant, and the power generation efficiency of the whole power generation system is improved.

In an optional embodiment of the present invention, the supercritical or ultra-supercritical nuclear power generation system further includes a hot gas conduit 2, and the hot gas conduit 2 includes a high temperature medium pipe 23 and a low temperature medium pipe 26, wherein a first end of the high temperature medium pipe 23 is configured to communicate with the outlet of the reactor apparatus 1, a second end of the high temperature medium pipe 23 is configured to communicate with a primary loop medium inlet 39 of the supercritical or ultra-supercritical boiler 3, a first end of the low temperature medium pipe 26 is configured to communicate with the inlet of the reactor apparatus 1, and a second end of the low temperature medium pipe 26 is configured to communicate with a primary loop medium outlet of the supercritical or ultra-supercritical boiler 3. In this way, the high-temperature first loop medium in the reactor apparatus 1 can be introduced into the supercritical or ultra-supercritical boiler 3 through the high-temperature medium pipe 23, and the low-temperature first loop medium in the supercritical or ultra-supercritical boiler 3 can be returned into the reactor apparatus 1 through the low-temperature medium pipe 26.

The supercritical or ultra-supercritical nuclear power generation system further comprises a fan, wherein the fan is arranged on the low-temperature medium pipeline 26, and the air supply direction of the fan can be the direction from the supercritical or ultra-supercritical boiler 3 to the reactor device 1, namely the fan can supply air to the reactor device 1, so that power can be provided for the flow of the first loop medium, and the first loop medium in the supercritical or ultra-supercritical boiler 3 flows back to the reactor device 1. Here, the fan is a helium fan.

In an alternative embodiment, the hot gas conduit 2 further comprises a first isolation valve 22 and a second isolation valve 25, the first isolation valve 22 is arranged on the high temperature medium pipeline 23 and is used for controlling the on-off of the high temperature medium pipeline 23; the second isolation valve 25 is arranged on the low-temperature medium pipeline 26 and is used for controlling the on-off of the low-temperature medium pipeline 26. In this way, the connection between the supercritical or ultra-supercritical boiler 3 and the reactor device 1 can be disconnected through the first isolation valve 22 and the second isolation valve 25, when the first loop medium leaks from the supercritical or ultra-supercritical boiler 3, the first loop medium in the reactor device 1 can be prevented from entering the supercritical or ultra-supercritical boiler 3 in time, and the first loop medium can be prevented from being released radially by closing the first isolation valve 22 and the second isolation valve 25, so that the safety of the power generation system is improved. By providing the first isolation valve 22 and the second isolation valve 25, the safety level of the equipment or parts far from the reactor apparatus 1 can be reduced, thereby reducing the difficulty in designing and manufacturing the equipment to achieve the design and manufacturing specifications capable of using the thermoelectric supercritical or ultra-supercritical boiler 3.

In the present embodiment, each of the first and second isolation valves 22 and 25 may be provided with at least two, and it is arranged such that when one of the isolation valves is damaged, the high temperature medium pipe 23 or the low temperature medium pipe 26 can be shut off by closing the other isolation valve.

Also, the blower may be located between two adjacent second isolation valves 25, thus facilitating the servicing of the helium blower. When the fan needs to be overhauled, the two second isolation valves 25 are closed firstly, the low-temperature medium pipeline 26 is cut off, the first loop medium in the supercritical or ultra-supercritical boiler 3 is not discharged any more, and the first loop medium in the reactor device 1 cannot run out through the outlet.

In one embodiment, the low temperature medium pipe 26 and the high temperature medium pipe 23 are provided independently of each other, and the primary medium inlet 39 and the primary medium outlet of the supercritical or ultra-supercritical boiler 3 are provided independently of each other, and the outlet and the inlet of the reactor device 1 are provided independently of each other.

In another embodiment, the low-temperature medium pipe 26 is sleeved outside the high-temperature medium pipe 23, and a gap is left between the low-temperature medium pipe 26 and the high-temperature medium pipe 23 for the low-temperature first loop medium to pass through, so that the first loop medium after heat exchange can flow back into the reactor device 1.

A plurality of supports are provided between the low temperature medium pipe 26 and the high temperature medium pipe 23. The supports are distributed in the axial direction of the hot gas duct for positioning the central hot medium duct 23.

Specifically, one end of the support is fixedly connected with the inner wall of the low-temperature medium pipeline 26, and the other end of the support abuts against the high-temperature medium pipeline 23, so that radial support is realized. And, in same support cross-section, support and include three supporting structure at least to realize the location of central high temperature medium pipeline 23, supporting structure includes the base that is equipped with compression spring and the branch that first end is connected with compression spring, base and the inner wall fixed connection of low temperature medium pipeline 26, and the second end of branch supports on the outer wall of high temperature medium pipeline 23. This enables the support structure to constantly support the high-temperature medium pipe 23. Here, a plurality of support structures are distributed around the outside of the hot medium duct 23, i.e. on the same support cross section, with at least three struts arranged, so that the hot medium duct 23 is fixed in the center of the cold medium duct 26.

The first isolation valve 22 and the second isolation valve 25 can be electric isolation valves so as to realize automatic control, thereby ensuring the safety of personnel.

In some embodiments, the inner wall of the high temperature medium pipe 23 and/or the outer wall of the low temperature medium pipe 26 is provided with an insulation layer. Specifically, the inner wall of the high temperature medium pipe 23 may be provided with an insulation layer to withstand the temperature difference between the high temperature medium and the low temperature medium, and the outer wall of the low temperature medium pipe 26 may also be provided with an insulation layer to reduce the heat loss of the low temperature medium pipe 26. Here, the insulating layer may be made of aluminum silicate fiber.

Further, a thermal expansion absorption structure may be provided on the insulation layer of the high temperature medium pipe 23 to absorb a thermal expansion difference between the high temperature medium pipe 23 and the insulation layer fixing structure. Here, the thermal expansion absorbing structure may be a socket structure, and is not particularly limited.

Here, a bellows may be provided at an end of the high temperature medium pipe 23 for absorbing thermal expansion generated in the high temperature medium pipe 23, and a bellows may be provided at an appropriate position of the high temperature medium pipe 23 for absorbing thermal expansion generated in the high temperature medium pipe 23.

In this embodiment, as shown in fig. 3, the hot gas duct 2 further includes a protective shell 24, the high temperature medium pipeline 23, the low temperature medium pipeline 26 and the fan are all located in the protective shell 24, and the protective shell 24 may be a concrete shell, so as to avoid helium leakage caused by breakage of the high temperature medium pipeline 23 or the low temperature medium pipeline 26. And an access door for an access man is provided on the protective case 24.

In an alternative embodiment of the present invention, the heat exchange device may comprise a plurality of sets of heat exchange tube bundles 33 which are sequentially communicated, wherein an inlet of the first set of heat exchange tube bundles 33 can be communicated with a water supply system of the steam-electric power conversion device, and the last set of heat exchange tube bundles 33 can be communicated with an inlet of the steam-electric power conversion device, so as to realize circulation of the second loop medium.

Also, the supercritical or ultra supercritical boiler 3 further comprises a first header 34, a second header 310 and at least one third header 311.

Wherein, the inlet of the first header 34 is communicated with the water supply system of the steam-electric power conversion device, the inlet of the first group of heat exchange tube bundles 33 is communicated with the outlet of the first header 34, the inlet of the second header 310 is communicated with the outlet of the last group of heat exchange tube bundles 33, and the outlet of the second header 310 is communicated with the inlet of the steam-electric power conversion device.

And the outlets and inlets of two adjacent groups of heat exchange tube bundles 33 are communicated through a third header 311 to realize the collection and the flow division of the second loop medium. That is, the outlet of the previous group of heat exchange tube bundles and the inlet of the next group of heat exchange tube bundles are communicated through the third header 311, so that the second loop medium in the previous group of heat exchange tube bundles is collected through the third header 311, and the collected second loop medium is distributed to the next group of heat exchange tube bundles through the third header 311.

In this way, the second loop medium (water) supplied by the water supply system of the steam-electric power conversion device is collected in the first header 34, and then the second loop medium is supplied to the first group of heat exchange tube bundles 33, and the second loop medium in the plurality of groups of heat exchange tube bundles 33 is collected and divided sequentially by the plurality of third headers 311 until the second loop medium in the last group of heat exchange tube bundles 33 is collected in the second header 310 and flows into the steam-electric power conversion device through the second header 310.

Every group heat exchange tube bank all can include the heat exchange tube of many settings side by side, and the heat exchange tube can be U type pipe to adjacent two sets of heat exchange tube bank 33's U type mouth orientation can be opposite. It should be noted that the heat exchange power and heat exchange parameters of each group of heat exchange tube bundle 33 can be designed and arranged according to actual requirements.

Further, a gap for the first loop medium to pass through is left between the adjacent heat exchange tube bundles 33.

In an alternative embodiment, a mixing zone 35 is provided within the pressure containing vessel for homogenizing the first loop medium in the mixing zone 35 to homogenize the flow rate of the first loop medium flowing into the heat transfer zone.

In an optional embodiment, the pressure vessel includes a pressure-bearing shell and a furnace 32, the furnace 32 is located in the pressure-bearing shell, a gap 38 is arranged between the furnace 32 and the pressure-bearing shell, the heat exchange region is located in the furnace 32, and the gap 38 is communicated with an outlet of the heat exchange region, so that the first loop medium after heat exchange can enter the gap 38 from the outlet of the heat exchange region. Here, a primary-loop medium inlet 39 is provided on the furnace 32, a primary-loop medium outlet is provided on the pressure-bearing housing, and the primary-loop medium outlet is communicated with the gap 38, so that the heat-exchanged primary-loop medium flows out of the primary-loop medium outlet.

It should be noted that the top end of the furnace 32 is not in contact with the top end of the pressure-bearing shell, and a certain distance is left between the top end of the furnace and the top end of the pressure-bearing shell, so that the outlet of the heat exchange region is not closed by the top end of the pressure-bearing shell, and the gap 38 is communicated with the outlet of the heat exchange region.

Moreover, the furnace 32 can be configured as a membrane wall, which separates the interior and exterior of the furnace 32 into two different zones to reduce the temperature of the outer zone of the furnace 32.

In this embodiment, the pressure-bearing housing can bear the pressure of the first loop medium. The pressure-bearing shell comprises a cylindrical barrel wall, a top cover for sealing the top end of the barrel wall and a bottom plate for sealing the bottom end of the barrel wall. The cylinder wall, the top cover and the bottom plate can be of an integrated structure so as to improve the sealing performance of the pressure-bearing shell.

Here, the pressure-bearing housing may be a prestressed concrete housing 31, a metal housing, or another type of housing.

In an alternative embodiment of the present invention, as shown in fig. 4, the pressure-bearing container is further provided with a liquid water receiving chamber 36, the liquid water receiving chamber 36 is located below the mixing zone, and the liquid water receiving chamber 36 is communicated with the mixing zone. Thus, when a tube breakage accident occurs in the heat exchange tube bundle 33, the liquid water (i.e. the second loop medium) in the heat exchange tube bundle 33 falls into the liquid water receiving cavity 36 by gravity, and is received in a centralized manner, so as to prevent the liquid water from entering the reactor device 1.

And, there is a water cooling pipe 37 on the top of the liquid water receiving cavity 36, and the water cooling pipe 37 is communicated with the first header 34, so that the arrangement can make the low-temperature second loop medium pass through the water cooling pipe 37, thereby reducing the temperature in the liquid water receiving cavity 36 and further reducing the evaporation capacity of the liquid water.

It should be noted that the liquid water receiving chamber 36 is located in a stagnation region where the inert gas flows, so as to prevent the first loop medium from bringing the liquid water into the reactor apparatus 1. Because the first loop medium is gas and moves upwards after entering the mixing area, a stagnation area is formed in the area below the mixing area, the convection influence of the gas on the surface of the liquid water in the liquid water containing cavity 36 is reduced, and the evaporation capacity of the liquid water is reduced.

In this embodiment, be provided with the access door that supplies the maintainer to get into on the pressure-bearing casing.

In an alternative embodiment of the present invention, as shown in fig. 2, the reactor apparatus 1 includes a reactor body 11 and a containment system, the reactor body 11 includes a reactor pressure vessel and a reactor, the reactor is disposed in the reactor pressure vessel, and a first loop medium is disposed in the reactor pressure vessel so as to lead out heat energy generated by the reactor. The containment system includes a containment vessel 12, the containment vessel 12 being housed outside the reactor body 11 to protect the public from nuclear reactor accidents.

Specifically, the core of the reactor employs spherical fuel elements of triple-coated ceramic-type particles, and a graphite core support structure is provided outside the spherical fuel elements. And the residual heat carried out by the reactor core can be completely led out by a passive mechanism. The material of the reactor pressure vessel is metal. And a gas guide pipe for leading out the first loop medium and an air inlet pipe for leading in the first loop medium are arranged on the reactor pressure vessel, an outlet and an inlet which are connected with the gas guide pipe and the air inlet pipe are arranged on the containment vessel 12, the outlet is used for being connected with a high-temperature medium pipeline 23, and the inlet is used for being connected with a low-temperature medium pipeline 26.

It should be noted that, the containment system may further include a negative pressure ventilation system, a vent hole communicated with the negative pressure ventilation system and a chimney for exhausting air after being processed are arranged on the containment 12, when the reactor pressure vessel leaks, the gas in the containment 12 is extracted through the negative pressure ventilation system, and after being filtered, the radioactive gas (the first loop medium) is exhausted to the atmosphere through the chimney, so that the influence of radioactivity on the environment is reduced, and the safety of the reactor device 1 is improved. A water-cooled wall is provided on the inner wall of the containment vessel 12.

Here, the containment system does not bear high pressure, does not have the functional requirements of spray cooling, combustible gas control and the like.

Further, thermal expansion absorption structures may be disposed on the inner walls of the reactor pressure vessel and the containment vessel 12 to absorb thermal expansion of the reactor pressure vessel and the containment vessel 12 in a hot state.

In an optional embodiment of the present invention, the steam-power conversion device 4 may be similar to a thermal power supercritical or ultra-supercritical steam-power conversion system, and may utilize high-temperature and high-pressure steam to drive the steam turbine 41 to do work, and generate electricity through a generator, and simultaneously supply deionized water to the supercritical or ultra-supercritical boiler 3 according to the parameter requirements of the system.

Specifically, as shown in fig. 6, the steam-electric power conversion device 4 may include a steam turbine 41, a generator 42, a condenser 43, a condensate pump 44, a low-pressure heater 45, a deaerator 46, a water feed pump 47, and a high-pressure heater 48, wherein the steam turbine 41 includes a high-pressure cylinder, an intermediate-pressure cylinder, and a low-pressure cylinder connected in sequence, a steam inlet of the high-pressure cylinder is communicated with an outlet of the heat exchange device, and the steam turbine 41 is connected to the generator through a shaft. The outlet of the low-pressure cylinder, the condenser 43, the condensate pump 44, the low-pressure heater 45, the deaerator 46, the water feed pump 47, the high-pressure heater 48 and the first header are sequentially connected, so that condensed water obtained by condensing low-temperature steam is conveyed to the first header to supply water to the heat exchange device, and recycling of water (second loop medium) is achieved.

Here, the water supply system of the steam power conversion device 4 may be a water supply pump.

The invention provides a supercritical or ultra-supercritical nuclear power generation system which comprises a reactor device 1, a hot gas guide pipe 2, a supercritical or ultra-supercritical boiler 3 and a steam power conversion device 4.

The reactor device 1 comprises a reactor body 11 and a containment system arranged at the periphery of the reactor body 11, and the reactor device 1 is used for stably providing a high-temperature and high-pressure first loop medium for the whole power generation system by utilizing the heat generated by the reactor. The reactor device 1 adopts a modularized high-temperature gas-cooled reactor to fully utilize the intrinsic safety of the high-temperature gas-cooled reactor, the modularized construction and the main characteristics of providing a high-temperature heat source. The containment system serves as a safety barrier to improve the safety of the reactor apparatus 1. The reactor in the reactor body 11 can be a pebble bed type module high-temperature gas cooled reactor, the reactor core structure of the reactor is the same as that of a 20 million module type high-temperature gas cooled reactor demonstration power station, the thermal power of a single reactor is about 250MW, helium is used as a coolant, the highest temperature of the helium can reach 750 ℃ or higher, and a high-temperature gas medium can be stably provided for supercritical and ultra-supercritical boilers. In addition, the reactor can also adopt a prismatic high-temperature gas cooled reactor.

The high-temperature medium pipeline 23 of the hot gas guide pipe 2 is used for guiding a high-temperature first loop medium (the temperature of high-temperature helium is higher than 600 ℃) in the reactor device 1 into the supercritical or ultra-supercritical boiler 3, and the low-temperature medium pipeline 26 of the hot gas guide pipe 2 is used for guiding a low-temperature first loop medium (the temperature of low-temperature helium is not higher than 350 ℃) after heat exchange in the supercritical or ultra-supercritical boiler 3 into the reactor device 1. The hot gas conduit 2 can adopt a concentric pipe mode, namely a high-temperature medium pipeline and a low-temperature medium pipeline are combined together, wherein the high-temperature medium pipeline is positioned in the center and only bears the pressure difference between a high-temperature first loop medium and a low-temperature first loop medium, the wall thickness of the high-temperature medium pipeline is thinner, and a heat insulation layer needs to be laid on the inner side of the high-temperature medium pipeline 23 so as to bear the temperature difference between the high-temperature first loop medium and the low-temperature first loop medium. The low-temperature medium pipeline is positioned on the outer side and bears the pressure of the low-temperature first loop medium, the pipe wall is thick, and in order to reduce the heat loss of the low-temperature medium pipeline, a heat insulation layer structure can be arranged on the outer side of the low-temperature medium pipeline. In addition, the high temperature medium pipe and the low temperature medium pipe may be independently provided.

The hot gas duct 2 further comprises a first isolation valve 22 for controlling the on/off of the high temperature medium duct and a second isolation valve 25 for controlling the on/off of the low temperature medium duct so as to reduce the safety level of the equipment or components remote from the reactor device 1. The helium fan is arranged between the two second isolation valves 25 of the cryogenic medium pipeline, and when the two second isolation valves 25 are in an isolated state, the helium fan can be repaired or replaced. Furthermore, the helium fan can also be arranged in the lowest-safety part of the cryogenic medium line and as far away from the core as possible.

The supercritical or ultra-supercritical boiler 3 includes a pressure-bearing shell capable of bearing a certain pressure and a heat exchange tube bundle 33 for performing heat exchange between helium gas and water/steam, and since the pressure of helium gas on the pressure-bearing shell side is higher, a prestressed concrete shell 31 can be adopted to bear the pressure of helium gas. In order to ensure the reliability of the prestressed concrete shell 31, the prestressed concrete shell 31 may take the shape of a cylinder plus a top cover. In order to ensure that the temperature of the prestressed concrete shell 31 does not exceed the temperature, a membrane water wall is provided on the inner side of the prestressed concrete shell 31. Meanwhile, a liquid water receiving cavity 36 is arranged below the heat exchange area of the supercritical or ultra-supercritical boiler 3 for receiving liquid water in the heat exchange tube bundle 33 when the heat exchange tube bundle 33 is damaged.

The steam power conversion device 4 is used for converting the latent heat of the high-temperature and high-pressure steam generated by the supercritical and ultra-supercritical boiler into electric energy and providing supercooled water meeting the requirements for the supercritical or ultra-supercritical heat exchange device. The supercritical or ultra-supercritical nuclear power generation system skillfully combines a high-temperature gas cooled reactor technology and a thermal power supercritical or ultra-supercritical power generation technology, can obviously improve the parameters (including the pressure and the temperature of steam) of a steam medium, enables the steam parameters to reach the supercritical or ultra-supercritical quality, further improves the power generation efficiency (the power generation efficiency can be improved by about 4 percent) of the whole power generation system, further reduces the investment cost of unit power generation (per kilowatt), and improves the market competitiveness of the whole nuclear power generation technology. Meanwhile, by reasonably arranging the isolation valve, the safety level of the original power generation system can be reduced, so that the design and manufacturing specifications of the thermal power supercritical ultra-supercritical boiler can be met. Therefore, the design difficulty of the components is reduced, and the mature experiences of China in the aspects of material selection, design, construction, operation, maintenance and the like of the thermal power supercritical or ultra-supercritical boiler 3 can be comprehensively used for reference. So as to realize multiple goals of simplifying design, facilitating maintenance and reducing cost.

Because the high-temperature gas cooled reactor has the characteristic of modularization, different module numbers can be flexibly selected to be matched with the supercritical or ultra-supercritical boiler 3 and the steam turbine 41 with different heat exchange powers, so that different requirements of different sites and users can be met. In addition, the power generation system can also realize the replacement of the existing thermal power plant, and replace the original supercritical or ultra-supercritical boiler 3 on the premise of keeping the power generation equipment such as the steam turbine 41 and the like unchanged so as to realize the sustainable utilization of the thermal power plant site.

The above-described embodiments of the apparatus are merely illustrative, and some or all of the modules may be selected according to actual needs to achieve the purpose of the solution of the present embodiment. One of ordinary skill in the art can understand and implement it without inventive effort.

Finally, it should be noted that: the above examples are only intended to illustrate the technical solution of the present invention, but not to limit it; although the present invention has been described in detail with reference to the foregoing embodiments, it will be understood by those of ordinary skill in the art that: the technical solutions described in the foregoing embodiments may still be modified, or some technical features may be equivalently replaced; and such modifications or substitutions do not depart from the spirit and scope of the corresponding technical solutions of the embodiments of the present invention.

Claims (14)

1. A supercritical or ultra supercritical nuclear power generation system, comprising a reactor device, a steam-electric power conversion device and a supercritical or ultra supercritical boiler, wherein the reactor device is internally provided with a first loop medium for heat carrying, the steam-electric power conversion device is used for converting the latent heat of a second loop medium into electric energy, and the supercritical or ultra supercritical boiler comprises:

the pressure-bearing container is provided with a heat exchange region, a primary loop medium inlet for the first loop medium to enter and a primary loop medium outlet for the first loop medium to discharge, the primary loop medium inlet is communicated with the inlet of the heat exchange region and the outlet of the reactor device, and the primary loop medium outlet is communicated with the outlet of the heat exchange region and the inlet of the reactor device;

and the heat exchange device is positioned in the heat exchange area, the outlet of the heat exchange device is communicated with the inlet of the steam-power conversion device, the inlet of the heat exchange device is communicated with a water supply system of the steam-power conversion device, and a second loop medium in the heat exchange device can exchange heat with the first loop medium.

2. The supercritical or ultra supercritical nuclear power generation system according to claim 1 further comprising a hot gas duct and a fan, the hot gas duct comprising:

a high-temperature medium pipeline, wherein a first end of the high-temperature medium pipeline is communicated with an outlet of the reactor device, and a second end of the high-temperature medium pipeline is communicated with a loop medium inlet of the supercritical or ultra-supercritical boiler;

a first end of the low-temperature medium pipeline is communicated with an inlet of the reactor device, and a second end of the low-temperature medium pipeline is communicated with a loop medium outlet of the supercritical or ultra-supercritical boiler;

the fan is arranged on the low-temperature medium pipeline, and the air supply direction of the fan is the direction from the supercritical or ultra-supercritical boiler to the reactor device.

3. The supercritical or ultra supercritical nuclear power generation system according to claim 2 wherein said hot gas duct further comprises:

the first isolation valve is arranged on the high-temperature medium pipeline and used for controlling the on-off of the high-temperature medium pipeline;

and the second isolation valve is arranged on the low-temperature medium pipeline and is used for controlling the on-off of the low-temperature medium pipeline.

4. The supercritical or ultra supercritical nuclear power generation system according to claim 3 wherein there are at least two of said first and second isolation valves and said wind turbine is located between two adjacent second isolation valves.

5. The supercritical or ultra supercritical nuclear power generation system of claim 2 wherein the low temperature medium conduit is sleeved outside the high temperature medium conduit with a gap between the low temperature medium conduit and the high temperature medium conduit for passage of the low temperature first loop medium.

6. The supercritical or ultra supercritical nuclear power generation system according to claim 5 wherein a plurality of supports are provided between the low temperature medium conduit and the high temperature medium conduit for positioning the high temperature medium conduit.

7. The supercritical or ultra supercritical nuclear power generation system according to claim 2 wherein the inner wall of the high temperature medium pipe and/or the outer wall of the low temperature medium pipe is provided with an insulation layer.

8. The supercritical or ultra supercritical nuclear power generation system according to claim 1 wherein the heat exchange device comprises a plurality of sets of sequentially connected heat exchange tube bundles, a first set of the heat exchange tube bundles having inlets capable of communicating with a water supply system of the steam-to-power conversion device and a last set of the heat exchange tube bundles capable of communicating with an inlet of the steam-to-power conversion device.

9. The supercritical or ultra supercritical nuclear power generation system according to claim 8 wherein said supercritical or ultra supercritical boiler further comprises:

the inlet of the first header is communicated with a water supply system of the steam-power conversion device, and the inlet of the first group of heat exchange tube bundles is communicated with the outlet of the first header;

an inlet of the second header is communicated with an outlet of the last group of heat exchange tube bundles, and an outlet of the second header is communicated with an inlet of the steam-electric power conversion device;

and the at least one third header is used for connecting outlets and inlets of two adjacent groups of heat exchange tube bundles.

10. The supercritical or ultra supercritical nuclear power generation system according to claim 9 wherein a mixing zone of the primary loop medium is provided within the pressure containing vessel, the mixing zone being located below the heat exchange zone and the mixing zone being in communication with the primary loop medium inlet.

11. The supercritical or ultra supercritical nuclear power generation system according to claim 1 wherein the pressure-bearing vessel comprises a pressure-bearing housing and a furnace located within the pressure-bearing housing, the heat exchange region is located within the furnace, and a gap through which the first loop medium passes is provided between the furnace and the pressure-bearing housing, the gap being in communication with an outlet of the heat exchange region, the loop medium inlet being provided on the furnace, the loop medium outlet being provided on the pressure-bearing housing and in communication with the gap.

12. The supercritical or ultra supercritical nuclear power generation system according to claim 10 wherein the pressure containing vessel is further provided with a liquid water receiving cavity located below and in communication with the mixing zone.

13. The supercritical or ultra supercritical nuclear power generation system according to claim 12 wherein a water cooled tube is provided at the top of the liquid water receiving cavity and is connected to the first header.

14. The supercritical or ultra supercritical nuclear power generation system according to claim 1 wherein said reactor plant comprises:

the reactor body comprises a reactor pressure vessel and a reactor positioned in the reactor pressure vessel, and the first loop medium is arranged in the reactor pressure vessel;

a containment system including a containment vessel, the containment vessel housed outside the reactor body.

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