A reactor pressure vessel support cooling device

Abstract

The application relates to a reactor pressure vessel support cooling device and belongs to the technical field of nuclear power, which comprises a support body, a support chamber is arranged in the support body, the support chamber is fixedly connected with a reactor pressure vessel body through a plurality of concrete support frames, a cold storage tank is arranged at the top end of the support body, a cooling liquid output pipeline is arranged at the bottom end of the cold storage tank, a control valve is arranged on the cooling liquid output pipeline, a heat conduction pipeline is arranged in each concrete support frame, the inlet end of the heat conduction pipeline is communicated with the cooling liquid output pipeline, the outlet end of the heat conduction pipeline is connected with a cooling liquid discharge pipeline, and the free end of the cooling liquid discharge pipeline is connected with a monitoring water pool. The application provides a cooling measure for the reactor pressure vessel support of a high-temperature gas cooled reactor under an ultra-design reference accident, effectively prevents the over-temperature failure of the concrete at the reactor pressure vessel support, and guarantees the structural integrity of the reactor pressure vessel.

Description

Technical Field

[0001] This invention relates to a reactor pressure vessel support cooling device, belonging to the field of nuclear power technology. Background Technology

[0002] Modular high-temperature gas-cooled reactor technology is a fourth-generation nuclear reactor technology. It employs a structure in which two nuclear steam supply system (NSSS) modules are connected to a turbine generator unit for operation and power generation. During operation, one or two NSSS modules supply steam to the turbine unit and perform work, with the turbine driving the generator unit to generate electricity.

[0003] The reactor pressure vessel and steam generator shell of a high-temperature gas-cooled reactor (HTGR) are supported by a metal structure within a reinforced concrete structure of the primary circuit compartment. A load-bearing support is located at the bottom of the reactor pressure vessel to ensure its structural integrity under various operating conditions. However, the design temperatures of the working chamber, reactor pressure vessel, hot gas duct shell, and steam generator shell are all significantly higher than the allowable temperature of the concrete. To protect the concrete structure, the design must avoid transferring unacceptable heat from the supporting structure to the concrete structure. Therefore, the supports of the HTGR reactor pressure vessel must be cooled by a cooling system to ensure that the concrete temperature meets design requirements.

[0004] The existing high-temperature gas-cooled reactor pressure vessel support cooling system is a passive natural circulation system, with the air cooler located indoors (reactor hall). Under beyond-design-baseline accident conditions, the system's mitigation capabilities are limited by the following factors: (1) Ambient temperature: The reactor pressure vessel support cooling system of the high-temperature gas-cooled reactor is located indoors (reactor hall), and its heat carrying capacity is affected by the temperature of the reactor hall. Under accident conditions, there is a risk of complete loss of plant power, shutdown of the reactor hall ventilation and air conditioning system, rise in ambient temperature of the reactor hall, and decrease in heat carrying capacity of the reactor pressure vessel support cooling system; there is also a risk of overheating of the concrete at the reactor pressure vessel support.

[0005] (2) Under the above-design-base accident conditions, when the temperature of the reactor pressure vessel support exceeds 100°C, the cooling water boils and vaporizes at atmospheric pressure, and there is a risk of interruption of the natural circulation of the support cooling system.

[0006] (3) Under the above-design-base accident conditions, the reference value of heat dissipation of the reactor pressure vessel support of the high-temperature gas-cooled reactor far exceeds the design heat carrying capacity of the reactor pressure vessel support cooling system.

[0007] In summary, under conditions exceeding design basis accidents, the concrete at the support of the high-temperature gas-cooled reactor pressure vessel is at risk of overheating failure, which could affect the structural integrity of the reactor pressure vessel. Therefore, improvements are urgently needed. Summary of the Invention

[0008] To overcome the shortcomings of the prior art, this invention designs a reactor pressure vessel support cooling device, which provides cooling measures for the reactor pressure vessel support of a high-temperature gas-cooled reactor under an over-design-base accident, effectively preventing the concrete at the reactor pressure vessel support from overheating and ensuring the structural integrity of the reactor pressure vessel.

[0009] To achieve the above objectives, the present invention adopts the following technical solution: Technical Solution 1 A reactor pressure vessel support and cooling device includes a support body with a support chamber inside the support body. The support chamber is fixedly connected to the reactor pressure vessel body through multiple concrete support frames. A cold storage tank is installed at the top of the support body, and a coolant output pipe is installed at the bottom of the cold storage tank. A control valve is installed on the coolant output pipe. Each concrete support frame is equipped with a heat conduction pipe. The inlet end of the heat conduction pipe is connected to the coolant output pipe, and the outlet end of the heat conduction pipe is connected to the coolant discharge pipe. The free end of the coolant discharge pipe is connected to a monitoring water tank.

[0010] Furthermore, the heat-conducting pipe includes a spiral pipe, an inlet pipe, and an outlet pipe. The two ends of the spiral pipe are integrally connected to the inlet pipe and the outlet pipe, respectively. The free end of the inlet pipe is connected to the coolant output pipe, and the free end of the outlet pipe is connected to the coolant discharge pipe.

[0011] Furthermore, the spiral pipe is divided into an outer spiral pipe and an inner spiral pipe. The outer spiral pipe is arranged on the side away from the reactor pressure vessel body, and the inner spiral pipe is arranged on the side close to the reactor pressure vessel body. The pitch of the outer spiral pipe is greater than the pitch of the inner spiral pipe.

[0012] Furthermore, the outer spiral pipe is arranged in one half of the pipe laying area inside the concrete support frame, and the inner spiral pipe is arranged in the other half of the pipe laying area.

[0013] Furthermore, the pitch of the outer helical tube is 10cm, and the pitch of the inner helical tube is 5cm.

[0014] Furthermore, the inlet pipe is located below the outlet pipe, and the inner end of the spiral pipe is connected to the inlet pipe, while the outer end of the spiral pipe is connected to the outlet pipe.

[0015] Furthermore, at least four concrete support frames are provided.

[0016] Furthermore, the safety level of the concrete support frame is safety level 3, the seismic resistance category is seismic resistance class I, and the quality assurance level is QA2.

[0017] Furthermore, the monitoring pool is positioned at a height lower than the concrete support frame.

[0018] Technical Solution Two A cooling method for a reactor pressure vessel support cooling device, based on the above-mentioned technical solution one, includes the following steps: Store coolant in the cold water tank and connect the coolant output pipe, each heat conduction pipe and the coolant discharge pipe. In the event of an accident exceeding the design baseline, when the temperature of the concrete support frame exceeds the set safety threshold, or when the cooling system of the normal reactor pressure vessel malfunctions, the staff shall adjust the opening of the control valve to control the flow rate of the coolant. Under the influence of gravity, the coolant in the cold water storage tank automatically flows into each heat conduction pipe through the coolant output pipe to absorb the heat transferred from the concrete support frame. After absorbing heat, the coolant flows out from the heat-conducting pipe and is discharged into a monitoring water tank located at a low position, thereby maintaining the temperature at the concrete support frame below the set safety threshold and preventing the concrete from overheating and failing.

[0019] Compared with the prior art, the present invention has the following features and beneficial effects: 1. This invention arranges heat conduction pipes inside the concrete support frame, achieving precise cooling of the load-bearing structure. This fundamentally prevents the support structure from losing its load-bearing capacity due to overheating, ensuring the structural integrity of the reactor pressure vessel. At the same time, the cold storage tank is installed at the top of the support body, and the cooling water is driven entirely by gravity. No external power source or active equipment is required, and it is not affected by the ambient temperature and operating conditions of the reactor hall. It can still be reliably put into operation under extreme conditions of over-design basis accidents where the plant power supply is lost.

[0020] 2. The heat-conducting pipe of this invention adopts a differentiated pipe layout with dense inner and sparse outer sections. The heat flux density is highest on the side closer to the reactor pressure vessel. The densely coiled inner spiral tubes can provide a larger heat exchange area per unit volume, effectively intercepting concentrated heat with greater threats. Moreover, the coolant with the lowest temperature first contacts the inner end area where the heat is most concentrated, forming a counter-current heat exchange effect, which significantly increases the average heat exchange temperature difference and improves the heat exchange efficiency per unit length of the coil. The heat on the side farther from the reactor pressure vessel has been significantly attenuated after being absorbed by the inner end. The sparser outer spiral tubes reduce the amount of coil material and coolant flow resistance while meeting the heat exchange requirements, achieving a balance between heat exchange efficiency and material economy. Furthermore, by placing the inlet pipe below the outlet pipe, the cooling water flows from bottom to top through the spiral pipe and can naturally fill the entire coil by gravity, effectively avoiding the accumulation of gas in the pipe and forming air resistance, ensuring the smooth flow of the cooling channel and the stability of heat exchange under accident conditions.

[0021] 3. This invention pre-embeds the heat-conducting pipes inside the concrete support frame, and the flow channel space is fixed and shaped by the concrete. It will not be squeezed and deformed due to the high temperature thermal expansion of the reactor pressure vessel body, thus eliminating the risk of flow channel obstruction under high temperature in the existing external cooling method. Moreover, the staff can actively adjust the cooling water flow rate through the control valve to match the heat carrying capacity with the actual heat dissipation at the concrete support frame, effectively solving the problem of insufficient heat carrying capacity of the existing system. Attached Figure Description

[0022] Figure 1 This is a schematic diagram of the connection structure of the present invention; Figure 2 This is a schematic diagram of the arrangement structure of the heat conduction pipe of the present invention.

[0023] The attached diagrams are labeled as follows: 100, cold storage tank; 200, reactor pressure vessel body; 201, support chamber; 300, steam generator; 400, top wall; 500, coolant output pipe; 600, control valve; 700, support body; 800, concrete support frame; 801, spiral pipe; 802, inlet pipe; 803, outlet pipe; 900, coolant discharge pipe; 1000, monitoring pool. Detailed Implementation

[0024] The present invention will now be described in more detail with reference to the embodiments.

[0025] Example 1 Please see Figure 1 and Figure 2The reactor pressure vessel support cooling device of this embodiment includes a support body 700, and a support chamber 201 is provided inside the support body 700. In this embodiment, two support chambers 201 are arranged side by side. The reactor pressure vessel body 200 is installed in the left support chamber 201, and the steam generator 300 is installed in the right support chamber 201. The reactor pressure vessel body 200 and the steam generator 300 together constitute a nuclear steam supply system.

[0026] Specifically, the interior of the support chamber 201 is fixedly connected to the reactor pressure vessel body 200 via four concrete support frames 800.

[0027] In this embodiment, a cold storage tank 100 is installed at the top of the top wall 400 of the support body 700, and a coolant output pipe 500 is installed at the bottom of the cold storage tank 100. A control valve 600 is provided on the coolant output pipe 500. The control valve 600 is a manual regulating valve. The operator can control the output flow of the coolant by adjusting the opening of the manual regulating valve and match the heat inside the concrete support frame 800 according to the temperature of the corresponding temperature measuring point.

[0028] Each concrete support frame 800 is equipped with a heat conduction pipe.

[0029] The inlet end of the heat conduction pipe is connected to the coolant outlet pipe 500, the outlet end of the heat conduction pipe is connected to the coolant discharge pipe 900, and the free end of the coolant discharge pipe 900 is connected to the monitoring water tank 1000.

[0030] In this embodiment, the coolant is cooling water.

[0031] Under normal operating conditions, the cold storage tank 100 stores a sufficient amount of cooling water, and the control valve 600 remains closed. When an accident exceeding the design baseline occurs, or when the normal reactor pressure vessel cooling system malfunctions, if the temperature at the concrete support frame 800 exceeds the set safety threshold (the set temperature safety threshold in this embodiment is 70°C), the control valve 600 will be manually opened by on-site personnel.

[0032] The cold storage tank 100 is installed at the top of the support body 700. It can take advantage of the height difference between the cold storage tank 100 and the concrete support frame 800. The cooling water in the cold storage tank 100 flows automatically into the heat conduction pipe inside each concrete support frame 800 under the action of gravity through the coolant output pipe 500. When it flows through the heat conduction pipe, it absorbs the heat transferred from the concrete support frame 800. The cooled water after absorbing heat is discharged into the monitoring water pool 1000 located at a lower position through the coolant discharge pipe 900.

[0033] Through gravity flow, heat is continuously carried away from the concrete support frame 800, keeping its temperature below 70°C to prevent it from failing due to overheating. This ensures the structural integrity of the reactor pressure vessel body 200 and mitigates the consequences of over-design basis accidents.

[0034] As can be seen from the above description, the cold storage tank 100 is located at the top of the support body 700. The cooling water is driven by gravity and does not require an external power source. It is not affected by the ambient temperature and operating conditions of the reactor hall and can still be reliably put into operation under accident conditions of loss of plant power.

[0035] Meanwhile, the cooling water flow can be actively adjusted by the control valve 600 to match the heat carrying capacity with the heat dissipation at the concrete support frame 800, effectively solving the problem of insufficient heat carrying capacity of the existing system. Moreover, this device relies entirely on gravity when working, without the need for external power, and has a high degree of passive reliability.

[0036] This device, serving as a redundant backup for the normal support cooling system, can effectively cool the concrete support frame 800 under extreme conditions such as over-design-base accidents, preventing the concrete support frame 800 from overheating and failing. Furthermore, the heat-conducting pipes are embedded inside the concrete support frame 800, and their flow channels are fixed in place by the concrete, preventing compression deformation due to the high-temperature thermal expansion of the reactor pressure vessel body 200. This ensures the stability and smoothness of the cooling channels under accident conditions, eliminating the risk of channel obstruction at high temperatures. Moreover, the cooling water removes the heat absorbed by the concrete support frame 800, directly reflecting the core temperature safety index of the concrete, resulting in a more reliable cooling effect.

[0037] Further, please refer to Figure 2 The heat conduction pipe includes a spiral pipe 801, an inlet pipe 802, and an outlet pipe 803. The two ends of the spiral pipe 801 are integrally connected to the inlet pipe 802 and the outlet pipe 803, respectively. The free end of the inlet pipe 802 is connected to the coolant output pipe 500, and the free end of the outlet pipe 803 is connected to the coolant discharge pipe 900.

[0038] The spiral pipe 801 is divided into an outer spiral pipe and an inner spiral pipe. The outer spiral pipe is arranged on the side away from the reactor pressure vessel body 200, and the inner spiral pipe is arranged on the side close to the reactor pressure vessel body 200. The pitch of the outer spiral pipe is greater than that of the inner spiral pipe. In this embodiment, the pitch of the outer spiral pipe is 10cm and the pitch of the inner spiral pipe is 5cm.

[0039] When the control valve 600 is opened under the condition of failure exceeding the design benchmark, the cooling water in the cold storage tank 100 flows into the heat conduction pipes inside each concrete support frame 800 by gravity through the coolant output pipe 500.

[0040] Cooling water first enters the inner spiral tube with a relatively close pitch through the inlet pipe 802. This side is close to the reactor pressure vessel body 200 and is the area where heat is most concentrated. The densely coiled inner spiral tube can arrange more tube length in a limited space, thereby obtaining a larger heat exchange area in this area to efficiently intercept concentrated heat.

[0041] The cooling water then flows through the outer spiral tube with a wider pitch. This side is 200 meters away from the reactor pressure vessel body, where the heat has already been reduced. The wider pitch reduces the amount of coil used while ensuring sufficient heat exchange, thus lowering costs and flow resistance.

[0042] After absorbing heat, the cooling water flows through outlet pipe 803 into coolant discharge pipe 900, and finally into monitoring water tank 1000 located at a lower level.

[0043] By using a differentiated pipe layout with denser inner sections and sparser outer sections, the entire spiral pipe 801 achieves matching with the temperature field distribution characteristics inside the concrete support frame 800. It absorbs water densely on the side close to the heat source and sparsely fills the gaps on the side far from the heat source, ensuring that the temperature of each area of ​​the concrete support frame 800 is uniform and maintained below 70°C.

[0044] As can be seen from the above description, the spiral pipe 801 adopts a differentiated pipe layout with a dense inner spiral pipe pitch of 5cm and a sparse outer spiral pipe pitch of 10cm. The inner spiral pipe is close to the reactor pressure vessel body 200, where the heat flux density is the highest and the temperature gradient is the largest. The densely coiled inner spiral pipe can provide a larger heat exchange area per unit volume and efficiently intercept the most threatening concentrated heat. The outer spiral pipe is far away from the reactor pressure vessel body 200, where the heat has been significantly attenuated after being absorbed by the inner end. The sparser pitch reduces the amount of coil material and coolant flow resistance while meeting the heat exchange requirements, thus achieving a balance between heat exchange efficiency and economy.

[0045] Furthermore, half of the pipe-laying area inside the concrete support frame 800 is equipped with an outer spiral pipe, and the other half of the pipe-laying area is equipped with an inner spiral pipe. In other embodiments, at least half of the pipe-laying area is equipped with an inner spiral pipe.

[0046] Furthermore, the inlet pipe 802 is located below the outlet pipe 803, and the inner end of the spiral pipe 801 is connected to the inlet pipe 802, while the outer end of the spiral pipe 801 is connected to the outlet pipe 803.

[0047] As can be seen from the above description, when the control valve 600 is opened under the condition of exceeding the design benchmark, the cooling water in the cold storage tank 100 flows into the heat conduction pipe inside each concrete support frame 800 by gravity through the coolant output pipe 500. Since the inlet pipe 802 is located below the outlet pipe 803, the cooling water first enters the inner end of the spiral pipe 801 through the inlet pipe 802. This is the area where the heat is most concentrated. The cooling water comes into contact with the highest temperature area first with the lowest temperature, forming a counter-current heat exchange effect and maximizing the heat exchange temperature difference.

[0048] Subsequently, the cooling water flows spirally from the inner end to the outer end along the spiral pipe 801. After passing through the outer spiral pipe with a relatively wide pitch, it flows out from the outlet pipe 803 located above, merges into the coolant discharge pipe 900, and finally flows into the monitoring water pool 1000 at a lower position.

[0049] The arrangement of the inlet pipe 802 at the bottom and the outlet pipe 803 at the top allows the cooling water to flow from bottom to top in the spiral pipe 801. It can naturally fill the entire spiral pipe 801 by gravity, avoiding the accumulation of air in the pipe and forming air resistance. This ensures smooth flow and continuously removes heat from the inside of the concrete support frame 800, keeping the concrete temperature below 70°C.

[0050] In this embodiment, the safety level of the concrete support frame 800 is safety level 3, the seismic resistance category is seismic resistance class I, the quality assurance level is QA2, and the design complies with the relevant provisions of ASME Code, Volume III, ND.

[0051] Example 2 A cooling method for a reactor pressure vessel support cooling device, based on the above-described embodiment one, includes the following steps: Coolant is stored in the cold storage tank 100, and the coolant output pipe 500, each heat conduction pipe and the coolant discharge pipe 900 are connected.

[0052] The system has multiple battery-powered temperature measuring points for auxiliary monitoring. Under conditions exceeding design baseline, when the temperature of the concrete support frame 800 exceeds 70°C, or when the cooling system of the normal reactor pressure vessel malfunctions, the staff can adjust the opening of the control valve 600 to control the flow rate of the coolant.

[0053] Under the influence of gravity, the coolant in the cold storage tank 100 automatically flows into each heat conduction pipe through the coolant output pipe 500, absorbing the heat transferred from the concrete support frame 800.

[0054] After absorbing heat, the coolant flows out from the heat conduction pipe and is discharged into the monitoring water tank 1000 located at a low position, thereby maintaining the temperature of the concrete support frame 800 below 70°C and preventing the concrete from overheating and failing.

[0055] As can be seen from the above description, this method relies entirely on the height difference between the cold storage tank 100 and the concrete support frame 800 for operation. It requires no external power supply, is unaffected by the ambient temperature and operating conditions of the reactor hall, and can still be reliably put into operation under extreme conditions of power outage. Furthermore, the opening of the control valve 600 can be flexibly adjusted to match the real-time heat dissipation, solving the problem of insufficient heat carrying capacity of the existing system. As a redundant backup measure for the normal support cooling system, this method can effectively prevent the concrete support frame 800 from overheating and failing, ensure the structural integrity of the reactor pressure vessel body 200, and significantly mitigate the safety consequences of over-design basis accidents.

[0056] The working principle of this invention is as follows: During normal operation, the cold storage tank 100 stores a sufficient amount of cooling water, and the control valve 600 remains closed.

[0057] In the event of an accident exceeding design baseline or a failure of the normal reactor pressure vessel cooling system, if the corresponding sensor detects that the temperature at point 800 of the concrete support exceeds 70°C, the control valve 600 shall be manually opened by on-site personnel.

[0058] The cold storage tank 100 is installed at the top of the support body 700. Utilizing the height difference between the cold storage tank 100 and the concrete support frame 800, gravity drives the cooling water inside the cold storage tank 100 to automatically flow into the heat conduction pipes inside each concrete support frame 800 via the coolant output pipe 500 under the influence of gravity. The heat conduction pipes include a spiral pipe 801, an inlet pipe 802, and an outlet pipe 803, with the inlet pipe 802 located below the outlet pipe 803. The inner end of the spiral pipe 801 is connected to the inlet pipe 802. The outer end is connected to the outlet pipe 803, so the cooling water flows from bottom to top through the spiral pipe 801 and can naturally fill the entire pipe, avoiding air resistance. At the same time, the cooling water first enters the inner spiral pipe with a denser pitch through the inlet pipe 802, which efficiently intercepts the area with the most concentrated heat. Then it flows through the outer spiral pipe with a sparser pitch, which reduces flow resistance while ensuring sufficient heat exchange. After absorbing heat, the cooling water flows through the outlet pipe 803 into the coolant discharge pipe 900 and finally into the monitoring water pool 1000 located at a low position.

[0059] Throughout the cooling process, staff can adjust the opening of control valve 600 in real time based on the temperature feedback at the concrete support frame 800 to control the cooling water flow rate, so that the heat carrying capacity of the device matches the actual heat dissipation at the concrete support frame 800. Through continuous gravity-driven circulation, the temperature of the concrete support frame 800 is maintained below 70°C to prevent it from failing due to overheating, thereby ensuring the structural integrity of the reactor pressure vessel body 200 and mitigating the consequences of over-design basis accidents.

[0060] In the description of this invention, it should be noted that the terms "inner", "outer", "upper", etc., indicate the orientation or positional relationship based on the orientation or positional relationship shown in the accompanying drawings. They are only for the convenience of describing this invention and simplifying the description, and do not indicate or imply that the device or element referred to must have a specific orientation, or be constructed and operated in a specific orientation. Therefore, they should not be construed as limiting this invention.

[0061] In the description of this invention, it should be noted that, unless otherwise explicitly specified and limited, the term "connection" should be interpreted broadly. For example, it can be a fixed connection, a detachable connection, or an integral connection; it can be a mechanical connection or an electrical connection; it can be a direct connection or an indirect connection through an intermediate medium; it can be a connection within two components. Those skilled in the art can understand the specific meaning of the above terms in this invention according to the specific circumstances.

[0062] Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. All other embodiments obtained by those skilled in the art based on the embodiments of the present invention without inventive effort are within the scope of protection of the present invention.

Claims

1. A reactor pressure vessel support cooling device, characterized in that: It includes a support body (700), and a support chamber (201) is provided inside the support body (700). The support chamber (201) is fixedly connected to the reactor pressure vessel body (200) through multiple concrete support frames (800). A cold storage tank (100) is installed at the top of the support body (700), and a coolant output pipe (500) is installed at the bottom of the cold storage tank (100). A control valve (600) is installed on the coolant output pipe (500). A heat conduction pipe is installed inside each of the concrete support frames (800). The inlet end of the heat conduction pipe is connected to the coolant output pipe (500), and the outlet end of the heat conduction pipe is connected to the coolant discharge pipe (900). The free end of the coolant discharge pipe (900) is connected to the monitoring water tank (1000).

2. The reactor pressure vessel support cooling device according to claim 1, characterized in that: The heat-conducting pipe includes a spiral pipe (801), an inlet pipe (802), and an outlet pipe (803). The two ends of the spiral pipe (801) are integrally connected to the inlet pipe (802) and the outlet pipe (803), respectively. The free end of the inlet pipe (802) is connected to the coolant output pipe (500), and the free end of the outlet pipe (803) is connected to the coolant discharge pipe (900).

3. A reactor pressure vessel support cooling device according to claim 2, characterized in that: The spiral pipe (801) is divided into an outer spiral pipe and an inner spiral pipe. The outer spiral pipe is arranged on the side away from the reactor pressure vessel body (200), and the inner spiral pipe is arranged on the side close to the reactor pressure vessel body (200). The pitch of the outer spiral pipe is greater than the pitch of the inner spiral pipe.

4. A reactor pressure vessel support cooling device according to claim 3, characterized in that: The concrete support frame (800) has an outer spiral pipe arranged in one half of the pipe laying area, and an inner spiral pipe arranged in the other half of the pipe laying area.

5. A reactor pressure vessel support cooling device according to claim 3, characterized in that: The pitch of the outer spiral tube is 10cm, and the pitch of the inner spiral tube is 5cm.

6. A reactor pressure vessel support cooling device according to claim 2, characterized in that: The inlet pipe (802) is located below the outlet pipe (803), and the inner end of the spiral pipe (801) is connected to the inlet pipe (802), while the outer end of the spiral pipe (801) is connected to the outlet pipe (803).

7. A reactor pressure vessel support cooling device according to claim 1, characterized in that: The concrete support frame (800) is provided in at least four parts.

8. A reactor pressure vessel support cooling device according to claim 1, characterized in that: The concrete support frame (800) has a safety level of 3, a seismic resistance category of seismic class I, and a quality assurance level of QA2.

9. A reactor pressure vessel support cooling device according to claim 1, characterized in that: The monitoring pool (1000) is located at a height lower than the concrete support frame (800).

10. A cooling method based on the reactor pressure vessel support cooling device according to any one of claims 1-9, characterized in that: Includes the following steps: Store coolant in a cold water storage tank (100) and connect the coolant output pipe (500), each heat conduction pipe and the coolant discharge pipe (900). In the event of an accident exceeding the design baseline, when the temperature of the concrete support frame exceeds the set safety threshold, or when the cooling system of the normal reactor pressure vessel malfunctions, the operator shall adjust the opening of the control valve (600) to control the flow rate of the coolant. The coolant in the cold water tank (100) flows automatically into each heat conduction pipe through the coolant output pipe (500) under the action of gravity, absorbing the heat transferred from the concrete support frame (800). After absorbing heat, the coolant flows out from the heat-conducting pipe and is discharged into the monitoring water tank (1000) located at a low position, thereby maintaining the temperature at the concrete support frame (800) below the set safety threshold and preventing the concrete from overheating and failing.

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