Passive nuclear steam supply system

Abstract

A passive nuclear steam supply system (100), comprising a reactor body (1), the reactor body (1) comprising a reactor pressure vessel (11); and a reactor coolant system (2), the reactor coolant system (2) comprising two coolant loops (21), a pressurizer (22) and a surge line (23), and each coolant loop (21) comprising a steam generator (24), two primary pumps (25), a hot leg main pipe (26) and two cold leg main pipes (27), wherein the hot leg main pipe (26) is provided with a first liquid level pressure tap (261) and a second liquid level pressure tap (262), the first liquid level pressure tap (261) being located at the bottom of the hot leg main pipe (26), and the second liquid level pressure tap (262) approaching the steam generator (262) and being located at the top of the hot leg main pipe (26).

Description

Passive nuclear steam supply system

[0001] CROSS-REFERENCE TO RELATED APPLICATIONS

[0002] This application claims priority to Chinese patent application No. 202311184945.9, filed on September 13, 2023, entitled “Passive Nuclear Steam Supply System,” the entire contents of which are incorporated herein by reference. Technical Field

[0003] The present invention relates to the field of nuclear power technology, and in particular to a passive nuclear steam supply system. Background Art

[0004] In the early stages of the development of third-generation nuclear power models, the European Pressurized Water Reactor (EPR) developed by Electricité de France and the AP1000 developed by Westinghouse Electric Company of the United States were representative.

[0005] The EPR model adopts an active redundancy technology route in its safety concept and system configuration, relying on AC power supply, power equipment (main pump), ventilation and refrigeration, and external water sources to achieve accident mitigation and improve the reliability of performing its intended functions under accidents. It increases the multiplicity of the system and improves the reliability of the system by "adding". Although this method can significantly reduce the probability of random failures, it cannot improve the ability to resist common cause failures, nor can it fundamentally reduce human errors and extend the operator's non-intervention time.

[0006] The AP1000 utilizes a passive, simplified technology approach, relying not on external triggers or power sources but solely on natural convection, gravity, pre-pressurization, evaporation, and condensation to achieve safety. Despite its simplified design, the AP1000 unit, with a rated power of only 1000MW, is still considered a medium-sized unit. This results in a relatively high cost per unit of power, making it uneconomical for construction in areas with high electricity loads.

[0007] While the AP1000 provides a relatively mature nuclear steam supply system, increasing electrical power isn't simply a matter of scaling up system parameters. The number of loops determines the number of steam generators, main pumps, and main pipelines, as well as the number of pressure vessel connections. This number of loops is directly related to their heat transfer capacity. Constrained by factors such as safety, economics, equipment manufacturing, waste disposal, and limitations on passive cooling capacity, the nuclear steam supply system required a redesign.

[0008] Summary of the Invention

[0009] In response to the shortcomings of the existing technology, the present invention provides a passive nuclear steam supply system and nuclear power plant based on the 193 core, which significantly improves the power generation capacity of the passive system, reduces the core damage frequency and the frequency of large-scale radioactive material release in the safety evaluation indicators, and meets the requirements of the third generation of nuclear power.

[0010] A first aspect of the present invention provides a passive nuclear steam supply system, comprising: a reactor body including a reactor pressure vessel; a reactor coolant system including two coolant loops, a pressurizer, and a surge tank, each coolant loop including a steam generator, two main pumps, a hot section main pipeline, and two cold section main pipelines;

[0011] Among them, the two ends of the hot section main pipeline are respectively connected to the reactor pressure vessel and the steam generator, the main pump is connected to the lower end of the steam generator, and the two ends of the cold section main pipeline are respectively connected to the reactor pressure vessel and its corresponding main pump. After the reactor coolant is heated in the reactor pressure vessel, it enters the primary side of the steam generator through the hot section main pipeline, and after heat exchange through the steam generator, it returns to the reactor pressure vessel through the main pump and the cold section main pipeline; one end of the surge pipe is connected to the hot section main pipeline of one of the coolant loops, and the other end of the surge pipe is connected to the bottom of the stabilizer; the hot section main pipeline is provided with a first liquid level pressure port and a second liquid level pressure port. The first liquid level pressure port is located at the bottom of the hot section main pipeline, and the second liquid level pressure port is close to the steam generator and located at the top of the hot section main pipeline.

[0012] Preferably, two groups of pressure relief pipelines are provided on the top of the pressure stabilizer, each group of pressure relief pipelines includes a pressure relief main pipe connected to the pressure stabilizer and three parallel pressure relief branch pipes downstream of the pressure relief main pipe, a spring-loaded safety valve is provided on the pressure relief main pipe, a short pipe with a bursting disc is provided downstream of the spring-loaded safety valve, and automatic pressure relief valves are provided on the three pressure relief branch pipes respectively. The downstream of the three pressure relief branch pipes merge into an exhaust pipeline, a vacuum breaker is provided on the exhaust pipeline, and the pressure relief pipeline is arranged downward from the pressure stabilizer to the automatic pressure relief valve.

[0013] Preferably, the passive nuclear steam supply system also includes a normal waste heat removal system, which includes a normal waste heat removal pump and a normal waste heat removal heat exchanger. One end of the normal waste heat removal system is connected to a hot section main pipeline, and the other end is connected to the reactor pressure vessel.

[0014] Preferably, the passive nuclear steam supply system further comprises a refueling water tank in the containment vessel, wherein the refueling water tank in the containment vessel stores refueling water, and during refueling, the refueling water floods the cylinder of the reactor pressure vessel.

[0015] Preferably, a bubbler is provided in the refueling water tank in the containment vessel, and the end of the exhaust pipeline is connected to the bubbler.

[0016] Preferably, a hot section main pipeline is further connected to a fourth-stage pressure relief pipeline, and a fourth-stage automatic pressure relief valve is provided on the fourth-stage pressure relief pipeline. The fourth-stage automatic pressure relief valve is a burst valve.

[0017] Preferably, the passive nuclear steam supply system also includes dedicated safety facilities, the dedicated safety facilities include a passive core cooling system, the passive core cooling system includes a passive safety injection subsystem, the passive safety injection subsystem includes a core make-up water tank, a safety injection tank and a refueling water tank in the containment, the core make-up water tank, the safety injection tank and the refueling water tank in the containment are connected to the nozzle of the reactor pressure vessel through a common pressure vessel direct injection pipeline.

[0018] Preferably, the first safety water source is stored in the core water supply tank, the second safety water source is stored in the injection tank, and the refueling water in the refueling water tank in the containment is the third safety water source. The order of input after the accident is: first safety water source, second safety water source, and third safety water source.

[0019] Preferably, the passive core cooling system also includes a passive residual heat removal subsystem, which includes a passive residual heat removal heat exchanger located in the refueling water tank inside the containment vessel, the inlet end of the hot side of the passive residual heat removal heat exchanger is connected to a hot section main pipeline, and the outlet end is connected to a steam generator.

[0020] Preferably, the dedicated safety facilities also include a containment system and a passive containment cooling system. The containment system includes a steel containment, which accommodates a nuclear steam supply system. The passive containment cooling system includes a passive containment cooling water tank, an air flow channel, and a water diversion hopper arranged outside the steel containment. The passive containment cooling water tank stores a fourth safety water source. The passive containment cooling water tank is higher than the steel containment. The water diversion hopper is connected to the passive containment cooling water tank through a pipeline. The water diversion hopper can provide spray water to the steel containment. The air flow channel is located circumferentially on the outside of the steel containment, and air can be introduced into the air flow channel.

[0021] Preferably, the passive containment cooling system further comprises a passive containment cooling auxiliary water tank and a recirculation loop, and the passive containment cooling auxiliary water tank replenishes water to the passive containment cooling water tank through the recirculation loop.

[0022] The nuclear steam supply system of the present invention uses 193 fuel assemblies with an active area height of 14 feet in the core, which are cooled by two coolant loops. This can enable the core rated power to reach 4040MWt, the power plant gross power to reach 1500MWe, and the power plant availability to reach 93%, significantly improving the economy of the model. BRIEF DESCRIPTION OF THE DRAWINGS

[0023] In order to more clearly illustrate the technical solutions of the embodiments of the present application, the following briefly introduces the drawings required for use in the embodiments of the present application.

[0024] FIG1 is a schematic structural diagram of a nuclear steam supply system according to a specific embodiment of the present invention;

[0025] FIG2 is a schematic diagram of a reactor body of a nuclear steam supply system according to a specific embodiment of the present invention;

[0026] FIG3 is a schematic flow diagram of a nuclear steam supply system according to a specific embodiment of the present invention;

[0027] FIG4 is a schematic flow chart of a passive safety injection subsystem according to a specific embodiment of the present invention;

[0028] FIG5 is a schematic flow diagram of a passive core decay heat removal subsystem according to a specific embodiment of the present invention;

[0029] FIG6 is a schematic flow diagram of a passive containment cooling system according to a specific embodiment of the present invention.

[0030] Reference Signs: 100 - Nuclear Steam Supply System; 1 - Reactor Proper; 11 - Reactor Pressure Vessel; 111 - Lower Head; 112 - Cylinder; 113 - Upper Head; 12 - Integrated Top Assembly; 13 - Internals; 131 - Lower Internals; 132 - Upper Internals; 14 - Core Components; 15 - Control Rod Drive Mechanism; 16 - Leakage Guidance Line; 17 - Direct Injection Nozzle; 2 - Reactor Coolant System; 21 - Coolant Loop; 21a - First Loop; 21b - Second Loop; 22 - Pressurizer; 221 - Pressurizer Upper Head; 2211 - Spray Pipe; 2212 - Two Safety Relief Pipes; 222-Pressurizer cylinder; 223-Pressurizer lower head; 224-Sprayer; 23-Surge pipe; 24-Steam generator; 241-Water chamber head; 242-Lower cylinder; 243-Upper cylinder, 244-Steam generator upper head; 245-First nozzle; 246-Second nozzle; 247-Steam outlet; 248-Feedwater inlet; 25-Main pump; 251-Inlet nozzle; 252-Vibration monitoring device; 26-Hot section main pipeline; 261-First liquid level pressure port; 262-Second Second liquid level pressure tap; 27 - cold section main pipeline; 271 - spray branch pipe; 272 - spray valve; 273 - spray main pipe; 274 - needle valve; 28 - pressure relief line; 281 - short pipe; 282 - pressure relief branch pipe; 283 - automatic pressure relief valve; 283a - first-stage automatic pressure relief valve; 283b - second-stage automatic pressure relief valve; 283c - third-stage automatic pressure relief valve; 284 - exhaust line; 285 - safety valve; 286 - vacuum breaker; 29 - fourth-stage pressure relief line; 291 - blast valve; 3-Normal residual heat removal system; 31-Normal residual heat removal pump; 4-Containment refueling water tank; 41-Bubbler; 42-Containment refueling water tank filter; 5-Passive core cooling system; 51-Passive safety injection subsystem; 511-Core make-up water tank; 512-Safety injection tank; 52-Passive core decay heat removal subsystem; 521-Passive residual heat removal heat exchanger; 53-Reactor pressure vessel direct injection line; 54-Recirculation injection line; 6-Containment system; 61-Steel containment; 7-Passive containment cooling system; 71-Passive containment cooling water tank; 72-Air flow channel; 73-Water distribution hopper; 74-Passive containment cooling auxiliary water tank; 75-Recirculation pipeline. DETAILED DESCRIPTION

[0031] In order to better understand the technical solution of the present application, the embodiments of the present application are described in detail below with reference to the accompanying drawings.

[0032] The present invention provides a passive nuclear steam supply system based on the 193 reactor core. It adopts a passive accident mitigation design concept. Under the premise of ensuring the passive cooling capacity after an accident, through the integration of fuel design, reactor core design, thermal hydraulic design, equipment design, and system design, the thermal power of the nuclear steam supply system and the electric power of the nuclear power plant are significantly improved, thereby improving the economy of the nuclear power model. The safety indicators of the reactor are also significantly improved compared with traditional second-generation nuclear power plants and existing third-generation nuclear power plants.

[0033] The passive nuclear steam supply system of the following embodiment is described in detail based on the 193 reactor core.

[0034] FIG1 is a schematic structural diagram of a nuclear steam supply system 100 according to a specific embodiment of the present invention.

[0035] As shown in Figure 1, the passive nuclear steam supply system 100 of this embodiment includes a reactor body 1 and a reactor coolant system 2. Reactor body 1 contains nuclear fuel, which continuously generates heat through a controlled chain fission reaction. Reactor coolant system 2 contains reactor coolant, which circulates to cool the nuclear fuel and remove heat from reactor body 1. The heat removed by the reactor coolant then heats circulating water on the secondary side, generating steam. The reactor coolant used in the nuclear steam supply system 100 of the present invention is light water.

[0036] The reactor body 1 comprises a reactor pressure vessel 11 and an integrated top assembly 12. The reactor pressure vessel 11, specifically cylindrical, provides a high-integrity pressure boundary to contain the reactor coolant, the heat-generating core, and fuel fission products. The integrated top assembly 12 is mounted on the upper end of the reactor pressure vessel 11. During reactor shutdown and refueling, the integrated top assembly 12 is hoisted, transferred, and stored as a whole through integrated operations. This simplifies refueling operations, shortens refueling time, and improves the availability of the nuclear power plant.

[0037] FIG2 is a schematic diagram of a reactor body 1 of a nuclear steam supply system 100 according to a specific embodiment of the present invention.

[0038] As shown in FIG2 , the reactor body 1 is composed of at least a reactor pressure vessel 11 , internal components 13 , core components 14 , a control rod drive mechanism 15 , and an integrated reactor top assembly 12 .

[0039] The reactor pressure vessel 11 consists of a lower head 111, a barrel 112, and an upper head 113. Both the lower head 111 and the upper head 113 are hemispherical and integrally forged. The upper head 113 and the barrel 112 are connected by flanges. Both the barrel 112 and the upper head 113 have flanges. After the barrel 112 and the upper head 113 are connected, the flanges are sealed by a sealing ring. Two sets of sealing rings are provided, one inside and one outside, concentrically arranged. In this embodiment, the inner diameter of the reactor pressure vessel barrel 112 is 4430 mm.

[0040] The core structure 14 includes a fuel assembly and fuel-related components that cooperate with the fuel assembly.

[0041] The fuel assemblies generate heat through nuclear reactions, providing excellent heat transfer conditions for dissipating the heat. They are supported by the core components 13, maintaining stability during the nuclear reaction. The reactor pressure vessel 11 of this embodiment contains 193 fuel assemblies, arranged in a 15x15 pattern within the core component 14. The height of the fuel assembly active area is 426.72 cm, or 14 feet. The average linear power density of the fuel assemblies is 181.0 W / cm.

[0042] Each fuel assembly consists of 264 fuel rods and a fuel assembly skeleton arranged in a 17x17 square. The fuel assembly skeleton comprises an upper and lower tube base, eight intermediate grids, four intermediate stirring grids, a top grid, a bottom grid, a guard grid, 24 control rod guide tubes, and an instrument tube. The fuel rods are constructed from ZIRLO cladding tubes and Zr-4 end plugs encapsulating uranium dioxide ceramic pellets. The pellets are cylindrical and made from low-enriched uranium dioxide powder that is cold-pressed and sintered to the desired density. The fuel rods utilize a monolithic fuel-burnable poison structure, with the fuel pellets coated with boride.

[0043] Fuel-related assemblies include control rod assemblies, burnable poison assemblies, neutron source assemblies, and drag plug assemblies. These assemblies are used to control the reactivity of the fuel assembly. The fuel assemblies and fuel-related assemblies constitute the reactor core, which has an equivalent diameter of 337.10 cm in the active region and a core height-to-diameter ratio of 1.27.

[0044] The reactor body 1 of this embodiment can be loaded with not only uranium dioxide fuel assemblies but also mixed uranium-plutonium oxide fuel, so that the core has the MOX loading capability.

[0045] The nuclear steam supply system 100 of the present invention uses 193 fuel assemblies with an active area height of 14 feet in the core, which is cooled by two coolant loops. The rated power of the core can reach 4040MWt, the gross power of the power plant can reach 1500MWe, and the power plant availability can reach 93%, which significantly improves the economy of the model.

[0046] The nuclear steam supply system 100 of the present invention has a core design margin of not less than 15%, a core damage frequency of less than 4.02×10-7 / reactor-year, and a probability of large-scale radioactive material release of less than 5.07×10 -8 / reactor year, safety is further improved.

[0047] The in-core components 13 are located within the reactor pressure vessel 11 and consist of a lower in-core component 131, an upper in-core component 132, and a guide plate. The in-core components 13 support and position the fuel assembly and related components, guide and distribute the coolant flow, shield the reactor pressure vessel 11, and support and guide the in-core measurement components, thereby ensuring the safe and reliable operation of the reactor. Compared to existing nuclear power plant models, the in-core components 13 of this embodiment eliminate the neutron shielding plate, thereby reducing the risk of loose components within the reactor. The lower chamber structure eliminates the flow distribution skirt and adopts a removable flow equalizing plate, thereby optimizing flow distribution performance. In this embodiment, the availability of the in-core components 13 is no less than 93%.

[0048] The integrated top assembly 12 is installed on the upper head 113 of the reactor pressure vessel 11. During reactor shutdown and refueling, the top assembly 12 is hoisted, transferred, and stored in its entirety through integrated operation with the upper head 113. This simplifies refueling operations, shortens refueling time, and improves the availability of the nuclear power plant. The integrated top assembly 12 primarily consists of a shroud assembly, a control rod drive mechanism, a ventilation and cooling system, a cantilevered crane assembly, a control rod drive mechanism seismic support structure, a top cover hoist, and a pressure vessel top cover venting system support.

[0049] The control rod drive mechanism 15 is installed on the integrated reactor top assembly 12 and extends through the upper head 113. The control rod drive mechanism 15 is connected to the control rod assembly and controls the insertion and advancement of the control rod assembly, thereby controlling the average core temperature and enabling safe reactor startup, reactor power regulation, and reactor shutdown. In this embodiment, the control rod drive mechanism 15 is a magnetic lift type.

[0050] The fuel assemblies utilize a high-burnup, low-leakage core loading method for both the initial and refueling cycles. This loading method improves neutron economy and extends the operating life of the reactor pressure vessel 11. During operation, the refueling cycle is 18-24 months, with an 18-month refueling cycle alternating with a 16- or 20-month refueling cycle to accommodate the varying refueling schedules of power plants. Under various refueling strategies, the average fuel element burnup after unloading is no less than 50,000 MW·d / tU, and can reach a maximum of 62,000 MW·d / tU.

[0051] The control rod assembly meets the reactivity control requirements in baseload, load following, initial startup, shutdown restart, and rapid power reduction modes, enabling daily load following operation using a mechanical compensation strategy. This strategy significantly reduces the number of boron adjustments in baseload mode and the amount of boron-containing wastewater in load following mode, enabling unadjusted boron load following above a certain power level for most of the cycle life.

[0052] Two leakage monitoring lines 16 are installed between the shell flange and the upper head flange of the reactor pressure vessel 11: one between the inner and outer sealing rings, and the other outside the outer sealing ring. These two lines share a common main pipe, on which a high-temperature alarm temperature detector is installed to monitor the reactor pressure vessel flange seal for leakage. When the monitored temperature exceeds the ambient temperature, it indicates a seal leak and triggers a high-temperature alarm.

[0053] FIG3 is a flow diagram of a nuclear steam supply system 100 according to a specific embodiment of the present invention.

[0054] As shown in FIG1 and FIG3 , the reactor coolant system 2 is connected to the reactor pressure vessel 11 , and the decay heat generated by the core is taken away by driving the coolant to circulate between the reactor coolant system 2 and the reactor pressure vessel 11 .

[0055] The reactor coolant system 2 includes two coolant loops 21, a pressurizer 22, and a surge pipe 23. Each coolant loop 21 includes a steam generator 24, two main pumps 25, a hot section main pipeline 26, and two cold section main pipelines 27.

[0056] As shown in Figure 3, the two coolant loops 21 are respectively a first loop 21a and a second loop 21b. The first loop 21a includes a steam generator 24a, two main pumps 25a, a hot section main pipeline 26a, and two cold section main pipelines 27a; the second loop 21b includes a steam generator 24b, two main pumps 25b, a hot section main pipeline 26b, and two cold section main pipelines 27b.

[0057] Reactor coolant flows through the primary side of the steam generator 24. The two ends of the hot-section main pipeline 26 are connected to the reactor pressure vessel 11 and the steam generator 24, respectively. Two main pumps 25 are connected to the lower ends of the corresponding steam generators 24. The two ends of the cold-section main pipeline 27 are connected to the reactor pressure vessel 11 and the corresponding main pumps 25, respectively. After being heated in the reactor pressure vessel 11, the reactor coolant enters the primary side of the steam generator 24 through the hot-section main pipeline 26. After heat exchange in the steam generator 24, it returns to the reactor pressure vessel 11 through the main pump 25 and the cold-section main pipeline 27, completing the circulation. The main pump 25 provides the driving force for the circulation of the reactor coolant. After absorbing heat from the primary side, the secondary side of the steam generator 24 generates steam, which is used to drive the steam turbine to produce work.

[0058] As shown in Figure 1, the steam generator 24 is a steam generating device that integrates a vertical shell, U-shaped tubes, and a steam-water separator. The cylinder of the steam generator 24 includes a water chamber head 241, a lower cylinder 242, an upper cylinder 243, and a steam generator upper head 244, which are connected in sequence. Inside the cylinder of the steam generator 24, there are tube sheets, a water chamber baffle, a tube bundle, and a steam-water separator (not shown). The tube sheet is located inside the steam generator near the connection between the water chamber head 241 and the lower cylinder 242, dividing the internal space of the steam generator 24 into upper and lower parts. The tube bundle is a U-shaped tube whose ports pass through the tube sheet. The water chamber baffle is located between the water chamber head 241 and the tube sheet, dividing the space of the water chamber head 241 into left and right parts, and separating the two ports of each U-shaped tube into two compartments of the water chamber head 241.

[0059] One side compartment of the water chamber head 241 houses a first nozzle 245 connected to the hot section main pipeline 26, while the other side compartment houses two second nozzles 246 connected to the main pump 25. This allows the reactor coolant flowing from the hot section main pipeline 26 to enter one side of the water chamber head 241, then pass through the U-shaped tube bundle and enter the other side of the water chamber head 241. It then passes through the main pump 25 and the cold section main pipeline 27 before returning. The steam-water separator, located on the secondary side of the tube bundle and comprising a primary separator and dryer, separates the steam generated on the secondary side. A steam outlet 247 is located at the top of the steam generator upper head 244, and a water inlet 248 is located on the side of the upper cylinder 243, enabling secondary steam-water circulation.

[0060] The steam generator 24 further includes auxiliary components (not shown) such as tube support plates and anti-vibration bars to maintain the stability of the tube bundle within the steam generator 24 .

[0061] In this embodiment, the heat transfer area of ​​a single steam generator 24 is 14666.5m 2 The allowable pipe blocking rate is 10%, the primary side operating gauge pressure is 15.4 MPa, and the thermal design flow rate of the primary side reactor coolant is 41206 m3 / h, thereby achieving a heat transfer capacity of 2029MW for a single steam generator 24.

[0062] The main pumps 25 are used to pump the reactor coolant, circulating it between the core, the coolant loop 21, and the steam generators 24. Each steam generator 24 has two main pumps 25 connected to its water chamber header 241. The main pumps' inlet nozzles 251 are directly connected to the steam generator's second nozzle 246. The two main pumps 25 on each steam generator 24 operate in the same direction. The main pumps 25 are vertical and can be either canned motor or wet-wound motor. Under optimal flow estimates, the flow rate of a single main pump is 21,642 m³ / s. 3 / h.

[0063] The motor and all rotating components of a canned motor pump are contained within a pressure vessel, which consists of the pump casing, heat shield ring, stator top cover, stator flange and casing assembly, stator bottom cover, as well as piping, flanges, and the external heat exchanger tube side. The canned motor pump's drive motor is a vertical, water-cooled, squirrel-cage induction motor with a canned rotor and stator, and a flywheel is located on the motor side. The motor and all rotating components of a wet-wound motor pump are contained within a pressure vessel, which consists of the pump casing, heat shield, main flange bolts and nuts, motor casing, bearing bracket, thrust bearing housing, piping, flanges, and the external heat exchanger tube side. The stator and rotor are in direct contact with the reactor coolant. The wet-wound pump's motor is a vertical, water-cooled, squirrel-cage wet-wound induction motor with a flywheel located on the motor side.

[0064] The main pump 25 is suspended below the water chamber head 241 of the steam generator 24. During operation, it generates a certain amount of vibration. Therefore, each main pump 25 is equipped with a vibration monitoring device 252. Specifically, the vibration monitoring device 252 includes five speed sensors and one phase sensor. The five speed sensors are used to measure radial vibration in the X direction (two), radial vibration in the Y direction (two), and axial vibration in the Z direction (one). The phase sensor is used to output the rotational frequency and is used in conjunction with the speed sensor. When the speed sensor detects different vibration values, it triggers different alarms, thereby ensuring the safe operation of the main pump 25.

[0065] Relative to the reactor pressure vessel 11, both coolant loops 21 employ a two-inlet, one-outlet configuration. That is, the reactor coolant enters the reactor pressure vessel 11 through two cold-segment main pipes 27 and exits the reactor pressure vessel 11 through a single hot-segment main pipe 26. The two coolant loops 21 are symmetrically arranged relative to the reactor pressure vessel 11. In each coolant loop 21, the hot-segment main pipe 26 is located between the cold-segment main pipes 27. The center elevation of the cold-segment main pipe 27 is 445 mm higher than that of the hot-segment main pipe 26. This allows the reactor coolant in the cold-segment main pipe 27 to naturally sink due to density differences when entering the reactor pressure vessel 11, facilitating mixing of the reactor coolant. The hot-segment main pipe 26 has an outer diameter of 1088 mm and a wall thickness of 94 mm. Its gauge pressure during normal operation is 15.7 MPa, and its operating temperature is 324°C. The cold-segment main pipe 27 has an outer diameter of 795 mm and a wall thickness of 72.5 mm. Its gauge pressure during normal operation is 16.2 MPa, and its operating temperature is 291°C. At the optimal estimated flow rate, the flow velocity in both the hot-segment main pipe 26 and the cold-segment main pipe 27 is reduced, alleviating the problem of velocity-accelerated corrosion in the pipes during long-term operation.

[0066] As shown in Figures 1 and 3, a surge pipe 23 and a pressurizer 22 are connected to the hot section main pipe 26 of one of the two coolant loops 21. One end of the surge pipe 23 is connected to the hot section main pipe 26 of one of the coolant loops 21, and the other end of the surge pipe 23 is connected to the bottom of the pressurizer 22. For example, the surge pipe 23 and the pressurizer 22 are connected to the hot section main pipe 26a of the first loop 21a.

[0067] The pressurizer 22 is the equilibrium point where the liquid and steam in the reactor coolant system 2 maintain a saturated state. It controls the pressure of the reactor coolant system 2 during steady-state operation and transiently in the power plant. Under accident conditions, it cooperates with other facilities to control the pressure change of the reactor coolant system 2 within the allowable range. Under major accident conditions, it provides overpressure or low pressure protection to prevent damage to the core or other equipment.

[0068] As shown in Figure 1, the pressurizer 22 is a cylindrical pressure vessel consisting of an upper head 221, a barrel 222, and a lower head 223. The upper head 221 is a hemispherical shell equipped with a spray tube 2211 and two safety pressure relief tubes 2212. A sprayer 224 is also located within the head (see Figure 3). The lower head 223 is a hemispherical shell with a central connection to the surge tube 23. Electric heaters (not shown) are installed around the surge tube 23, each of which is individually replaceable.

[0069] As shown in Figure 3, each of the two cold-segment main pipes 27a of the first loop 21a is connected to a spray branch pipe 271, each equipped with a spray valve 272. The spray branch pipes 271 merge into a spray main pipe 273, which is then connected to the spray pipe 2211 of the pressurizer upper head 221, forming the pressurizer spray subsystem. When the main pump 25 is running, the pressure differential generated at the connection between the cold-segment main pipe 27a and the spray branch pipe 271, and at the connection between the hot-segment main pipe 26a and the surge pipe 23, acts as a driving force to propel the spray flow from the cold-segment main pipe 26a through the sprayer 224 and into the pressurizer upper head 221, spraying the supercooled atomized water into the pressurizer 22 to suppress pressure increases. A needle valve 274 is installed in parallel with each of the two spray branches 271 and the spray valve 272. When the spray valve 272 is closed, the needle valve 274 allows a continuous small bypass flow to the pressurizer spray pipe 2211 and the surge pipe 23. This prevents overcooling of the spray line and the surge pipe 23 when spray is not needed to control pressure during normal operation. It also reduces thermal stress and thermal shock when the spray valve 272 is opened or transiently causes coolant to surge in and out of the surge pipe 23, helping to maintain uniform water chemistry and temperature in the pressurizer 22. A water seal (not shown) is installed on the spray main pipe 273 to prevent steam from accumulating and returning to the spray valve 272.

[0070] As shown in Figure 1, two pressure taps are provided on the hot section main pipe 26 for connecting to a liquid level transmitter to monitor the liquid level of the reactor coolant system during the semi-filling period after shutdown. These two pressure taps include a first liquid level pressure tap 261 and a second liquid level pressure tap 262. The first liquid level pressure tap 261 is located at the bottom of the hot section main pipe 26, while the second liquid level pressure tap 262 is located at the top of the hot section main pipe 26, as close as possible to the steam generator 24.

[0071] A first liquid level pressure port 261 and a second liquid level pressure port 262 are provided on the hot section main pipe 26 so that the reactor coolant level can be continuously monitored during half-pipe operation, thereby improving the reliability of monitoring.

[0072] Two safety relief pipes 2212 on the pressurizer upper head 221 are connected to the pressure relief line 28. Each of these lines is equipped with a safety valve 285, providing overpressure protection for the reactor coolant system 2. Safety valve 285 is a spring-loaded valve that opens to relieve pressure when the pressure within the pressurizer 22 exceeds its set value. A short pipe 281 with a bursting disc is installed downstream of the outlet of safety valve 285, providing a sealed collection space for condensate leaking from the valve. Compared to existing nuclear power plants, the safety valve inlet pipeline in this embodiment does not have a water seal, thus preventing pressure setpoint drift. The pressure relief line 28 is equipped with three parallel pressure relief branches 282, each equipped with an automatic pressure relief valve 283: a first-stage automatic pressure relief valve 283a, a second-stage automatic pressure relief valve 283b, and a third-stage automatic pressure relief valve 283c. The downstream junction of the three automatic pressure relief valves 283 connects to an exhaust line 284, which is equipped with a vacuum breaker 286 to ensure that the discharge pipe pressure remains at atmospheric pressure after the valve releases. This prevents steam from condensing in the pipe and creating a vacuum, which could lead to water siphoning. The pressure relief line 28 runs downward from the pressure stabilizer 22 to the automatic pressure relief valves 283 on the pressure relief line 28, forming a cold water seal before the automatic pressure relief valves 283, minimizing steam leakage during normal power operation.

[0073] As shown in Figure 1, a surge tube 23 connects the hot section main pipeline 26 and the pressurizer 22. When the coolant density within the reactor coolant system 2 changes, the expanding or contracting reactor coolant flows into or out of the pressurizer 22 through the surge tube 23. An electric heater heats the subcooled liquid entering the pressurizer 22, vaporizing a portion of the water in the pressurizer 22 to control the pressure drop. The surge tube 23 has a nearly spiral structure, providing sufficient flexibility to adapt to changes in reactor coolant temperature. The surge tube 23 has a continuous downward slope from the pressurizer 22 nozzle to the hot section main pipeline 26, minimizing thermal stratification within the surge tube 23 and preventing the accumulation of wastewater in the pipeline.

[0074] In normal transient operation, in order to reduce the high thermal stress caused by the rapid change of fluid temperature, a thermal sleeve (not shown) is provided at the nozzle of the surge pipe 23 at the bottom of the stabilizer 22 and the nozzle of the stabilizer spray pipe 2211 to avoid excessive thermal stress.

[0075] A vent line is provided on the top of the reactor pressure vessel 11 , and a stack top vent valve is provided on the vent line. Reactor coolant is discharged through the stack top vent valve to prevent a large amount of coolant from being injected into the reactor pressure vessel 11 and causing the pressurizer 22 to overflow.

[0076] In this embodiment, the inner diameter of the stabilizer 22 is 2696 mm, the height is about 14574 mm, and the minimum internal volume is 70.8 m3 The rated total power of the electric heater is 1950kW and the designed spray flow rate is 205m 3 / h, the set pressure of the safety valve 285 is 17.2 MPa. Including the volume of the pressurizer 22 and the surge tube 23 in the cold state, the total volume of the reactor coolant system 2 is 390m3 when equipped with a wet winding pump. 3 , when equipped with a shielded pump, it is 386m 3 .

[0077] The nuclear steam supply system 100 has 193 14-foot fuel assemblies and is cooled by two cooling loops 21. The core power exceeds 4000MWt and the rated power can reach 4040MWt. The core design margin is greater than or equal to 15%, which significantly improves the core power. The two-loop reactor coolant system 2 also reduces the number of equipment and the space occupied by the equipment, thereby significantly improving the economy of the model.

[0078] FIG4 is a schematic diagram of a safety system flow chart of a specific embodiment of the present invention.

[0079] As shown in FIG4 , to cool the reactor core during shutdown, the nuclear steam supply system 100 of this embodiment is equipped with a normal residual heat removal system 3. This system 3 includes a normal residual heat removal pump 31, a normal residual heat removal heat exchanger (not shown), and related piping and valve components. The inlet of the normal residual heat removal system 3 is connected to the hot section main pipeline 26, and the outlet is connected to the reactor pressure vessel 11. During shutdown, reactor coolant is transported by the normal residual heat removal pump 31 to the normal residual heat removal heat exchanger for cooling. The cooled reactor coolant then returns to the reactor pressure vessel 11. The normal residual heat removal system 3 is an active system that cools the reactor core only during shutdown.

[0080] The reactor pressure vessel 11 is located within the reactor cavity. During refueling, the cylinder 112, lower head 111, and internals 14 of the reactor pressure vessel 11 are submerged. To facilitate refueling, the nuclear steam supply system 100 of this embodiment is provided with an in-containment refueling water tank 4. Refueling water is stored in the in-containment refueling water tank 4. During refueling, the refueling water in the in-containment refueling water tank 4 is transported to the reactor cavity, submerging the cylinder 112 of the reactor pressure vessel 11. After refueling is completed, the refueling water is transported back to the in-containment refueling water tank 4.

[0081] A bubbler 41 is installed in the refueling water tank 4 within the containment vessel. An exhaust line 284, connected to the first-stage automatic pressure relief valve 283a, the second-stage automatic pressure relief valve 283b, and the third-stage automatic pressure relief valve 283c, discharges water and steam into the refueling water tank 4 within the containment vessel through the bubbler 41. By opening these automatic pressure relief valves 283, the steam and water within the pressurizer 22 are discharged, gradually reducing the pressure of the reactor coolant system 2. A vacuum breaker 286 on the pressure relief line 28 prevents water from being siphoned from the bubbler 41 within the refueling water tank 4 within the containment vessel.

[0082] A containment refueling water tank filter 42 is provided in the containment refueling water tank 4. There are three containment refueling water tank filters 42. Each refueling water tank injection line corresponds to one containment refueling water tank filter 42. The containment refueling water tank filters 42 are connected by pipes.

[0083] A fourth-stage pressure relief line 29 is also connected to the hot section main pipeline 26a of the first loop 21a. This line is equipped with a burst valve 291, with its terminal end facing the atmosphere. Following an accident, burst valve 291 opens, reducing the pressure in the reactor coolant system 2. Burst valve 291 is a propellant-actuated valve, designed for single-action operation. Normally closed, it receives an activation signal, triggering the ignition of the propellant, opening the valve and providing pressure relief to the fourth-stage pressure relief line 29.

[0084] During normal shutdown, the decay heat of the core is removed through the active normal residual heat removal system 3. Under accident conditions, dedicated safety facilities are used to limit, control, mitigate and terminate the accident, maintain the safety of the core, and keep the public's radioactive radiation level within the specified limits. The dedicated safety facilities use passive methods to achieve their functions and do not rely on active components to prevent and mitigate accidents; specific passive measures include gravity, natural circulation, energy storage devices, check valves, and non-circulating power valves. The dedicated safety facilities specifically include the passive core cooling system 5, the containment system 6 (see Figure 6), the passive containment cooling system 7 (see Figure 6), and the hydrogen control system in the containment (not shown).

[0085] The passive core cooling system 5 is used to discharge the core decay heat after an accident, reduce the fuel cladding temperature, and reduce the release of radioactive substances into the environment. It can be specifically divided into a passive safety injection subsystem 51 and a passive core decay heat discharge subsystem 52.

[0086] The passive safety injection subsystem 51 is composed of at least two core water supply tanks 511, two safety injection tanks 512, a refueling water tank 4 in the containment, and related pipes and valves. The core water supply tanks 511, the safety injection tanks 512, the refueling water tank 4 in the containment, and the containment recirculation injection line 54 are injected into the direct injection nozzle 17 of the reactor pressure vessel through two shared reactor pressure vessel direct injection lines 53 (see Figure 1).

[0087] After a hypothetical loss of coolant (LOCA) accident, the pressure in the reactor coolant system 2 decreases, triggering a reactor shutdown and safety injection. A dedicated safety device 10 triggers a signal to shut down the main pump 25 and open the isolation valve at the outlet of the core makeup water tank 511, activating high-pressure injection. Once the pressure drops to a certain level, the safety injection tank 512 begins injection. After the valve on the fourth-stage pressure relief line 29 activates, the injection burst valve 291 in the containment refueling water tank 4 activates. Finally, the check valve on the injection line of the containment refueling water tank 4 opens, providing gravity injection. After a severe accident, the electric valve and burst valve on the recirculation injection line 54 open, allowing the water in the containment refueling water tank 4 to flood the exterior of the reactor pressure vessel 11, cooling the molten core within.

[0088] The core water supply tank 511 provides the first safety water source. It is a cylindrical vertical container with a total volume of 85m 3 The injection tank 512 provides a second safe water source. It is a spherical container with a total volume of 78.6m 3 , the water volume is 66.4m 3 The refueling water in the refueling water tank 4 in the containment is the third safety water source, and its minimum water volume is 2686.9m 3 The order of safe water sources after an accident is: first safe water source, second safe water source, and third safe water source.

[0089] FIG5 is a schematic flow diagram of a passive core decay heat removal subsystem 52 according to a specific embodiment of the present invention.

[0090] The passive core decay heat removal subsystem 52 includes a passive residual heat removal heat exchanger 521 and associated valves and piping. The passive residual heat removal heat exchanger 521 is located within the refueling water tank 4 within the containment vessel. Its inlet pipeline is connected to a hot-section main pipeline 26, and its outlet pipeline is connected to the cold-section chamber of the steam generator 24. Following an accident, the passive residual heat removal heat exchanger 521 automatically activates. After the main pump 25 stops operating, the passive residual heat removal heat exchanger 521 operates under natural circulation conditions to remove core decay heat. The passive residual heat removal heat exchanger 521 is designed to transfer 75.9 MW of heat.

[0091] The present invention adopts a passive dedicated safety facility, simplifies the system configuration, and can alleviate the accident without human operation 72 hours after the accident, thereby reducing the probability of human error.

[0092] FIG6 is a schematic flow diagram of a passive containment cooling system according to a specific embodiment of the present invention.

[0093] As shown in FIG6 , the dedicated safety facilities also include a containment system 6 and a passive containment cooling system 7 .

[0094] The containment system 6 consists of a steel containment 61 and mechanical penetrations, valves, and pipes on the steel containment 61. The nuclear steam supply system 100, except for the normal residual heat removal pump 31, the normal residual heat removal heat exchanger 32, and some pipes and valves of the normal residual heat removal system 3, is located inside the steel containment 61. A closed space is formed inside the steel containment 61 to ensure integrity under accident conditions, thereby providing a barrier to prevent the leakage of radioactive products. When the reactor coolant pipe or the secondary side pipe of the containment system 6 ruptures at both ends, the peak pressure inside the steel containment 6 will not exceed the design pressure value. The steel containment 61 includes a head and a cylinder. The inner diameter of the cylinder is 43 mm, the thickness of the head is 43 mm, the wall thickness of the cylinder is mostly 52 mm, the thickness of the bottom circle of the cylinder is 55 mm, and the free volume inside the steel containment 61 is 7.5×10 4 m 3 .

[0095] The refueling water tank 4 within the containment vessel is located within the steel containment vessel 61, and the passive residual heat removal heat exchanger 521 is placed within the refueling water tank 4 within the containment vessel. After an accident, when the passive residual heat removal heat exchanger 521 is in operation, coolant flowing through the core enters the primary side of the passive residual heat removal heat exchanger 521, transferring heat to the secondary side. The secondary side directly contacts the water within the refueling water tank 4 within the containment vessel. The water within the refueling water tank 4 within the containment vessel evaporates upon heating, and the water vapor rises to the inner wall of the steel containment vessel 61 above, where it liquefies upon cooling and transfers heat back to the steel containment vessel 61. A condensation reflux device is provided on the steel containment vessel 61 to return the condensate to the refueling water tank 4 within the containment vessel, achieving long-term circulating cooling within the steel containment vessel 61.

[0096] The passive containment cooling system 7 transfers heat from the air within the steel containment 61 to the external environment, limiting and reducing the temperature and pressure within the containment following a loss of coolant accident or a rupture of the main steam or main feedwater pipeline within the steel containment 61. This maintains the air pressure within the steel containment 61 within design limits, ensuring the integrity of the steel containment 61. The passive containment cooling system 7 includes at least one passive containment cooling water tank 71, a cooling water delivery channel 74 for draining water to the containment, an air flow channel 72 circumferentially located outside the steel containment 61, and a cofferdam and diversion hopper 73 for distributing water within the steel containment 61. The air flow channel 72 is circumferentially located outside the steel containment 61 and allows air to be introduced into the air flow channel 72.

[0097] The passive containment cooling water tank 71 is positioned higher than the steel containment 61, and the cofferdam and water diversion bucket 73 are located above the steel containment 61. The passive containment cooling water tank 71 is connected to the water diversion bucket 73 through a pipe, and the water diversion bucket 73 can provide spray water to the steel containment 61. The water in the passive containment cooling water tank 71 is sprayed onto the outer wall of the steel containment 61 through the water diversion bucket 73 under the action of gravity, absorbs the heat transferred from the steel containment 61, and is heated to become steam, which is carried out with the air in the air flow channel to achieve cooling of the steel containment 61.

[0098] The passive containment cooling water tank 71 stores the fourth safety water source, with a rated available volume of 3700m 3 Within 72 hours after a design basis accident, the passive containment cooling system 7 passively removes heat from the containment without requiring external intervention. In another embodiment, the passive containment cooling system 7 further comprises a passive containment cooling auxiliary water tank 74 and a recirculation line 75. After 72 hours, the recirculation line transfers water from the passive containment cooling auxiliary water tank 74 to the passive containment cooling water tank 71 for continued cooling.

[0099] The hydrogen control system in the containment (not shown) includes a hydrogen monitoring subsystem, a hydrogen recombination subsystem and a hydrogen ignition subsystem. The hydrogen monitoring subsystem includes three hydrogen sensors located in the dome area of ​​the steel containment for monitoring the hydrogen concentration; the hydrogen recombination subsystem includes a passive autocatalytic hydrogen recombiner for handling the slowly generated hydrogen expected under the design basis loss of coolant accident; the hydrogen ignition subsystem includes a hydrogen igniter, which is activated by the operator before the hydrogen concentration reaches the flammable limit.

[0100] With the nuclear steam supply system 100 equipped with the above-mentioned dedicated safety facilities, the power plant availability rate reached 93%, and the core damage frequency was less than 4.02×10 -7 / reactor-year, the frequency of large-scale radioactive material release is less than 5.07×10 -8 / Stack year, the economy and safety of the model are further improved.

[0101] The nuclear steam supply system 100 of this embodiment also includes instrumentation and a control system (not shown). These instruments collect physical parameters such as temperature, pressure, flow rate, liquid level, and neutron flux rate from the system, and after signal processing, mathematical calculations, and logical analysis, issue control commands to the equipment. The instrumentation includes process measurement, external nuclear measurement, and internal measurement instruments. The control system includes a power plant control system, a reactor protection and safety monitoring system, and a diverse drive system.

[0102] Process measurement instruments are used to measure most thermal parameters such as temperature, pressure, liquid level, flow, etc. They are set according to the needs of parameter monitoring. The collected data is transmitted to the power plant control system, reactor protection and safety monitoring system or diversified drive system. The measurement channels adopt multi-channel redundant settings.

[0103] External nuclear instrumentation measures core power information, ensuring continuous monitoring of reactor power. It provides neutron flux rate signals to the reactor protection and safety monitoring system. High neutron flux rates and high rates of change trigger a reactor shutdown through the system, providing startup and overpower protection.

[0104] The in-core measurement system monitors the neutron flux rate at various locations in the core and the core outlet temperature. All neutron flux rate signals are sent to the core online monitoring system to form a three-dimensional distribution of core power and support core operation optimization.

[0105] Power plant control systems are used to maintain normal plant operation and prevent reactors from deviating into unsafe conditions. During power operation, they regulate the parameters of the reactor and reactor coolant system by controlling the status of equipment such as reactor control rods, valves, pumps, and heaters to maintain constant thermal output. These controls keep equipment and systems operating under normal conditions, and plant parameters stable or fluctuating smoothly within specified limits, preventing the need for an emergency shutdown.

[0106] The reactor protection and safety monitoring system provides essential protection and oversight for reactor control during normal operation, enabling safety functions to be triggered and the plant to maintain a safe shutdown during abnormal, design basis event, and accident conditions. When plant parameters reach their shutdown settings, the reactor protection and safety monitoring system automatically triggers an emergency reactor shutdown, actuating dedicated safety features to mitigate the consequences of design basis accidents.

[0107] The diversified control system serves as a backup for the reactor protection and safety monitoring systems and is physically separate from them. It does not share the same measuring instruments, processing components, indicators, drives, or power supplies, nor do they have interconnected circuits. The diversified drive system mitigates the consequences of anticipated transients resulting from a failed emergency shutdown, reduces the probability of core damage caused by hypothetical common failures of the protection and monitoring systems in probabilistic safety assessments, and prevents containment failure due to overpressure after a core melt.

[0108] The present invention also provides a nuclear power plant that utilizes the aforementioned passive nuclear steam supply system 100 for the 193 core. The system also includes a steel plate concrete containment vessel, a steam turbine, and a generator. The steel plate concrete containment vessel is located outside the steel containment vessel 61, forming a double-layer containment structure. The steam turbine's air inlet and exhaust are connected to the steam outlet and water inlet of the steam generator 24, respectively. The steam turbine and generator are coaxially connected. Steam generated on the secondary side of the steam generator 24 enters the steam turbine, driving it to produce work. The steam turbine then drives the generator's rotor, generating electricity.

[0109] The steam turbine is a half-speed, single-shaft, four-cylinder, six-exhaust, condensing type equipped with two steam-water separator reheaters. The steam flow at the main steam inlet reaches 7,733 t / h, the steam absolute pressure reaches 5.78 MPa, and the steam temperature reaches 273.2°C. The unit has a rated speed of 1,500 r / min and a rated output of 1,534 MW, resulting in a total gross power output of 1,500 MWe.

[0110] The steel-concrete containment vessel is located outside the steel containment vessel 61, with a personnel passageway between them. A passive containment cooling water tank 71 is located on top of the steel-concrete containment vessel to enable gravity-fed spraying. Air inlets are provided around the circumference of the steel-concrete containment vessel near the top to introduce air into the air flow path of the passive containment cooling system 7. The steel-concrete containment vessel is 1.1 meters thick and is capable of withstanding impacts from large commercial aircraft.

[0111] The nuclear power plant of this embodiment is also provided with a main control room. The main control room is provided with a digital instrumentation and control platform, which integrates the power plant control system, protection and safety monitoring system, diversified drive system and other instrumentation and control systems to realize the monitoring and control of the nuclear power plant. An emergency habitability system is provided in the main control room to provide breathable air for the main control room personnel and limit the temperature rise in the area, thereby ensuring the habitability of the main control room after an accident. The emergency habitability system is composed of an emergency air storage tank and related pipelines, valves and instruments, including an air supply storage tank module and a main air supply pipeline and a spare air supply pipeline leading to the main control room. The air supply storage tank can hold enough compressed air to 111m 3 / h to supply air to the main control room for at least 72 hours.

[0112] The nuclear power plant of this embodiment is also equipped with a radioactive liquid waste treatment system. The radioactive liquid waste treatment system is equipped with an ion exchange sequence and a chemical flocculation process, thereby further reducing the concentration of radioactive liquid waste discharge, so that the concentration of radioactive effluent at coastal sites does not exceed 1000 Bq / L, and the concentration of radioactive effluent at inland sites does not exceed 100 Bq / L, thereby greatly reducing the discharge of radioactive liquid waste.

[0113] The foregoing description is merely a preferred embodiment of the present application and is not intended to limit the present application. Persons skilled in the art will readily appreciate that various modifications and variations are possible. Any modifications, equivalent substitutions, or improvements made within the spirit and principles of the present application shall be included within the scope of protection of the present application.

Claims

1. A passive nuclear steam supply system, characterized in that: include: The reactor body, which includes a reactor pressure vessel; A reactor coolant system, comprising two coolant loops, a pressurizer and a surge pipe, each of the coolant loops comprising a steam generator, two main pumps, a hot section main pipeline, and two cold section main pipelines; in, The two ends of the hot section main pipeline are respectively connected to the reactor pressure vessel and the steam generator, the main pump is connected to the lower end of the steam generator, and the two ends of the cold section main pipeline are respectively connected to the reactor pressure vessel and its corresponding main pump. After the reactor coolant is heated in the reactor pressure vessel, it enters the primary side of the steam generator through the hot section main pipeline, and after heat exchange through the steam generator, it returns to the reactor pressure vessel through the main pump and the cold section main pipeline; One end of the surge pipe is connected to the hot section main pipeline of one of the coolant loops, and the other end of the surge pipe is connected to the bottom of the pressurizer; The hot section main pipeline is provided with a first liquid level pressure port and a second liquid level pressure port, wherein the first liquid level pressure port is located at the bottom of the hot section main pipeline, and the second liquid level pressure port is close to the steam generator and located at the top of the hot section main pipeline.

2. The passive nuclear steam supply system according to claim 1, characterized in that: Two groups of pressure relief pipelines are provided on the top of the pressure stabilizer, and each group of the pressure relief pipelines includes a pressure relief main pipe connected to the pressure stabilizer and three parallel pressure relief branch pipes downstream of the pressure relief main pipe. A spring-loaded safety valve is provided on the pressure relief main pipe, and a short pipe with a bursting disc is provided downstream of the spring-loaded safety valve. Automatic pressure relief valves are provided on the three pressure relief branch pipes respectively, and the downstream confluences of the three pressure relief branch pipes are connected to an exhaust pipeline, which is provided with a vacuum breaker, and the pressure relief pipeline is arranged downward from the pressure stabilizer to the automatic pressure relief valve.

3. The passive nuclear steam supply system according to claim 1, characterized in that: It also includes a normal residual heat removal system, one end of which is connected to one of the hot section main pipelines, and the other end is connected to the reactor pressure vessel.

4. The passive nuclear steam supply system according to claim 2, characterized in that: The containment vessel also includes a refueling water tank, which stores refueling water. During refueling, the refueling water floods the cylinder of the reactor pressure vessel.

5. The passive nuclear steam supply system according to claim 4, characterized in that: A bubbler is provided in the refueling water tank in the containment, and the end of the exhaust pipeline is connected to the bubbler.

6. The passive nuclear steam supply system according to claim 2, characterized in that: A fourth-stage pressure relief pipeline is also connected to one of the hot section main pipelines. A fourth-stage automatic pressure relief valve is arranged on the fourth-stage pressure relief pipeline. The fourth-stage automatic pressure relief valve is a burst valve.

7. The passive nuclear steam supply system according to claim 4, characterized in that: It also includes dedicated safety facilities, which include a passive core cooling system, which includes a passive core cooling system, which includes a passive safety injection subsystem, which includes a core water makeup tank, a safety injection tank and a refueling water tank in the containment vessel, and which are connected to the nozzle of the reactor pressure vessel through a common pressure vessel direct injection line.

8. The passive nuclear steam supply system according to claim 7, characterized in that: The core water replenishment tank stores the first safety water source, the injection tank stores the second safety water source, and the refueling water in the refueling water tank in the containment is the third safety water source. The order of input after the accident is: the first safety water source, the second safety water source, and the third safety water source.

9. The passive nuclear steam supply system according to claim 7, characterized in that: The passive core cooling system also includes a passive residual heat removal subsystem, which includes a passive residual heat removal heat exchanger located in the refueling water tank in the containment vessel, the inlet end of the hot side of the passive residual heat removal heat exchanger is connected to one of the hot section main pipelines, and the outlet end is connected to one of the steam generators.

10. The passive nuclear steam supply system according to claim 7, characterized in that: The dedicated safety facilities also include a containment system and a passive containment cooling system. The containment system comprises a steel containment shell, wherein the steel containment shell contains the nuclear steam supply system. The passive containment cooling system comprises a passive containment cooling water tank, an air flow channel, and a water diversion hopper which are arranged outside the steel containment. The passive containment cooling water tank stores a fourth safety water source. The passive containment cooling water tank is higher than the steel containment. The water diversion hopper is connected to the passive containment cooling water tank through a pipeline. The water diversion hopper can provide spray water to the steel containment. The air flow channel is located in the circumferential direction of the outer side of the steel containment, and air can be introduced into the air flow channel.

11. The passive nuclear steam supply system according to claim 10, characterized in that: The passive containment cooling system further comprises a passive containment cooling auxiliary water tank and a recirculation loop, and the passive containment cooling auxiliary water tank replenishes water to the passive containment cooling water tank through the recirculation loop.

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