56results about "Plant pumping arrangements" patented technology

The present application relates to a method and system for filling water in a primary loop after overhaul charging of a nuclear power unit, comprising: performing preparation work before filling water; after the preparation work is completed, performing switch operation test of a to-be-tested actuator and water filling and exhaust work of a flowmeter; after the switch operation test of the to-be-tested actuator is completed, performing performance test of a high-pressure safety injection pump; after the performance test of the high-pressure safety injection pump is completed, performing performance test of a low-pressure safety injection pump; after the performance test of the low-pressure safety injection pump is completed, filling water in the primary loop according to a preset water filling mode; judging whether a real-time water level of a reactor pool reaches a full water level; if yes, performing alarm action test. The present application can solve the problem of long time consumption, reduce workload, significantly improve work efficiency, reduce work risk and effectively save production cost by completing the preparation work in advance and arranging the alarm action test at the end.

The present application relates to nuclear power technology field, provide a kind of steam seal air extraction system, comprising: steam turbine unit, refrigeration subsystem and absorber, steam turbine unit has steam seal cavity in two ends;Absorber is connected with refrigeration subsystem to form circulation loop, absorber is connected with air extraction pipe and exhaust pipe, air extraction pipe is used to suck air and water vapor in steam seal cavity into absorber, lithium bromide solution is provided in absorber, and lithium bromide solution is used to absorb water vapor.The steam seal air extraction system described above can suck air and water vapor leaked in steam seal cavity of steam turbine unit into absorber, and water vapor is absorbed by absorber, realizes refrigeration and air extraction simultaneously, meets the air extraction demand of steam turbine unit, also takes into account refrigeration demand, eliminates the problem of system complex, numerous equipment, high auxiliary energy consumption and large volume, weight caused by air extractor and refrigeration unit independent operation, realizes the simplification of power conversion device overall composition and operation efficiency improvement.

A curvilinear electromagnetic pump is configured to follow a curve, such as by joining together multiple linear pump segments offset at an angle relative to one another. A curvilinear electromagnetic pump can be curved in two dimensions or in three dimensions. A curvilinear electromagnetic pump allows for more efficient placement of components and systems within a reactor vessel, significantly reducing the height of the reactor vessel compared to vertically positioned linear pumps. A curvilinear electromagnetic pump can follow the curvature of the reactor vessel wall and can be positioned generally near the bottom of the reactor vessel.

The present invention belongs to the technical field of nuclear power plant equipment maintenance, and specifically relates to an exhaust device of a nuclear island drain and exhaust system of a nuclear power plant, which is achieved by connecting a set of exhaust devices in parallel on the exhaust pipeline at the top of the nuclear island pressurizer equipment, and by adding two tees and isolation valves, vacuum pumps (vacuum ejectors) and other devices on the original pipeline. The vacuum ejector is powered by compressed air, and the exhaust gas mixed with the gas of a circuit returns to the nuclear island drain and exhaust system. The device of the present invention for the flushing stage of a circuit after the overhaul of a nuclear power plant is equipped with a set of exhaust devices of the nuclear island drain and exhaust system, and a vacuum ejector is added to the exhaust pipeline, and the gas of the circuit is discharged through the vacuum ejector, and the circuit is filled with water. The vacuum ejector is powered by maintenance compressed air, and the exhaust gas mixed with the gas of the circuit returns to the nuclear island drain and exhaust system. When the pressurizer liquid level reaches full water, the pressurizer exhaust valve and the isolation valve of the vacuum ejector are closed.

This invention discloses a high-frequency suction device between heat transfer tubes of an evaporator, comprising a suction belt assembly, which includes a support belt body that can be inserted into the gap between the heat transfer tubes and has a suction channel inside. The support belt body has a suction hole that penetrates the side wall and communicates with the suction channel. A wall-climbing mechanism is also included, with the support belt body mounted on the wall-climbing mechanism. The wall-climbing mechanism allows the suction belt assembly to enter the heat transfer tubes of the evaporator, and then the high-frequency suction mechanism generates negative pressure to suction foreign objects inside the heat transfer tubes through the suction hole. Simultaneously, the high-frequency suction mechanism controls the suction frequency, and the suction belt assembly vibrates as the suction frequency changes. The vibration is transmitted to the foreign object body, changing the static state of the foreign object between the heat transfer tubes, making it easier to be suctioned. The vibration of the suction belt assembly also makes it less likely for foreign objects entering the suction belt assembly to block the channel.

The present invention proposes an automatic loading system and method for new fuel of a pebble bed high temperature reactor. The system includes a loading subsystem, a pipeline monitoring subsystem, a purge subsystem and a lifting subsystem. The loading subsystem includes a loader, a surface quality detection device and a reversing conveying device which are sequentially connected through a ball-dropping pipeline. The outlets of the reversing conveying device are respectively connected to a fuel operating system and a broken ball storage tank. The reversing conveying device determines the conveying direction of the new fuel according to the detection results of the surface quality detection device and the control instructions. The present invention realizes fully automatic single loading of new fuel and improves the efficiency of new fuel loading. The present invention is equipped with a pipeline monitoring subsystem, which can monitor the isolation effect of the fuel operating system and the new fuel loading subsystem in real time. Once radioactive substances leak out, the boundary isolation valve of the new fuel loading subsystem can be closed in real time to prevent the leakage of radioactive substances and protect the safety of operators.

The invention discloses a hydrogenation system and a hydrogenation method for a high-temperature and high-pressure irradiation test loop. The hydrogenation system comprises a hydrogen supply unit, a flow control unit and a feeding unit which are connected in sequence; the flow control unit comprises an automatic adding unit and a manual adding unit which are arranged in parallel; the manual adding unit comprises a return water providing unit and at least two hydrogenation units, and the hydrogenation unit is connected to the feeding unit and the hydrogen supply unit; at least two hydrogenation units are arranged in parallel in at least two rows; each row of hydrogenation units is connected to the return water providing unit; each row of hydrogenation units is provided with a temperature measuring instrument, and a pressure measuring instrument is provided on a pipeline connecting the hydrogenation unit and the feeding unit; isolation valves are provided at both ends of each row of hydrogenation units; the operation is convenient, safe and reliable, the dissolved hydrogen concentration in the loop has a fast response speed, a high control accuracy, and a high hydrogen use efficiency, which meets the hydrogenation requirements in the high-temperature and high-pressure irradiation test loop.

The utility model provides an anti-contamination instrument water filling and air exhausting tool which comprises an equipment body, a lower sealing component, an upper sealing component and a drainage pipeline, a pipeline channel is formed in one end of the equipment body, and the lower sealing component and the upper sealing component are connected to the two ends of the pipeline channel of the equipment body respectively. The drainage pipeline is connected to the upper sealing component and communicated with the pipeline channel of the equipment body, and the pipeline channel of the equipment body is connected with the instrument water filling and exhausting connector through the lower sealing component. According to the utility model, the equipment main body made of a radiation-resistant material and the built-in rotating ratchet wheel are matched with the instrument pollution discharge plug, and are matched with the two built-in sliding sealing rings in the lower sealing part to be connected with the instrument pollution discharge port, so that the radioactive medium is drained, and the pressurized radioactive medium can be effectively prevented from splashing; the safety problem caused by the fact that radioactive media in a pipeline splash or flow out when a traditional water filling and air exhausting tool such as an open spanner is used for instrument water filling and air exhausting is solved.

This application provides a method and apparatus for water injection in the primary loop of a nuclear reactor, belonging to the field of nuclear control. The method includes: obtaining the original enthalpy and residual enthalpy of the reactor core; obtaining the critical time for successful water injection and the critical time for unsuccessful water injection corresponding to the primary loop of the reactor core; determining the critical total flow rate for successful water injection in the primary loop based on the critical time for successful water injection, the original enthalpy and the residual enthalpy, and the energy conservation principle of the reactor core; and determining the critical total flow rate for unsuccessful water injection in the primary loop based on the critical time for unsuccessful water injection, the original enthalpy and the residual enthalpy, and the energy conservation principle of the primary loop; and controlling the total flow rate of water injection in the primary loop based on the critical total flow rate for successful water injection and the critical total flow rate for unsuccessful water injection. This application embodiment can improve the efficiency of primary loop water injection.

To provide a floating nuclear power generation system capable of controlling the position of a floating body.SOLUTION: The floating nuclear power generation system includes a reactor, a turbine generator driven by steam generated by heat of the reactor, a floating body moored on the sea on which the reactor and the turbine generator are disposed, and a circulating water pump for supplying cooling water to a condenser for condensing steam having passed through a turbine of the turbine generator, the circulating water pump including: a first circulating water pump responsible for supplying cooling water discharged from the floating body in a first direction when the floating body is viewed from above; and a second circulating water pump responsible for supplying cooling water discharged from the floating body in a second direction opposite to the first direction.SELECTED DRAWING: Figure 3

The invention discloses a temporary ventilation system arranged in a steel containment vessel, which is suitable for the steel containment vessel and comprises an upper exhaust system, a lower exhaust system, a first transverse and vertical main pipe assembly and a second transverse and vertical main pipe assembly, and the upper exhaust system and the lower exhaust system are both arranged in the steel containment vessel. The upper exhaust system is arranged at the upper end of the lower exhaust system, one end of the first transverse and vertical main pipe assembly extends into the steel containment and is connected with the upper exhaust system, and one end of the second transverse and vertical main pipe assembly extends into the steel containment and is connected with the second transverse and vertical pipe assembly. Through the application of the invention, the temporary ventilation system arranged in the steel containment vessel is provided, and the temporary ventilation system can effectively collect flue gas generated at different heights, so that the sedimentation and residual rate of the flue gas in the steel containment vessel can be reduced; in addition, the probability that the temporary ventilation item occupies the space of the formal item is reduced, more sufficient space is provided for construction of the formal item, and the space utilization rate of the steel containment is increased.

The invention discloses a nuclear power station composite passive safety system which comprises a first thermoelectric conversion subsystem, a second thermoelectric conversion subsystem and a third thermoelectric conversion subsystem, the first thermoelectric conversion subsystem is arranged at the upper part of a containment, the second thermoelectric conversion subsystem is arranged at the lower part of the containment, and the third thermoelectric conversion subsystem is arranged at the lower part of the containment. The third thermoelectric conversion subsystem is arranged in a fuel plant outside the containment; the first thermoelectric conversion subsystem, the second thermoelectric conversion subsystem and the third thermoelectric conversion subsystem each comprise a hot end, a cold end and a thermoelectric conversion device, the hot ends and the cold ends are arranged at different positions respectively, temperature differences exist between the hot ends and the cold ends, and the thermoelectric conversion devices can convert the temperature differences into electric energy. Therefore, electric energy is provided for each device of the nuclear power station cooling system.

The invention belongs to the technical field of VVER reactors, and discloses a power distribution control method for VVER reactor bias loop operation, which comprises axial power distribution control and radial power distribution control, can control axial and radial power distribution during VVER reactor bias loop operation, is applied to a certain VVER reactor bias loop operation condition in China for several times, and can control the power distribution of the VVER reactor bias loop operation condition. The reactor core power distribution Kv factors are all controlled within the limit value of the technical specification, so that the protection action caused by over-limit of local parameters of the reactor core is avoided, and the safety of the reactor core is effectively guaranteed when the reactor is in bias operation.

The utility model relates to the field of nuclear power plants, and discloses a nuclear power plant wastewater collection system which can optimize the design of the nuclear power plant wastewater collection system and reduce system equipment configuration and arrangement space so as to save project investment and project cost. The system comprises a conventional island waste liquid collecting system and a waste oil and non-radioactive water discharging system. The conventional island waste liquid collecting system comprises a single first water collecting pit, the first water collecting pit is configured to contain radioactive oil-free waste water, and the rear end of the water collecting pit is connected to the conventional island liquid effluent discharging system through a pipeline. The conventional island liquid effluent discharge system is configured to perform radioactive substance attenuation treatment on radioactive oil-free wastewater. The waste oil and non-radioactive water discharging system comprises a single second water collecting pit, the front end of the second water collecting pit is connected to a plurality of water feeding pump sets, a sealing oil system and a lubricating oil storage room through pipelines, the second water collecting pit is configured to contain non-radioactive oily wastewater, and the rear end of the second water collecting pit is connected to a nuclear island dirty oil treatment system through a pipeline; and the nuclear island sump oil treatment system is configured to treat oily wastewater without radioactivity.

The invention relates to the technical field of nuclear power, and discloses a modular nuclear energy small reactor efficient combined heat and power generation heating system which is characterized in that a modular nuclear energy small reactor supplies steam to a steam turbine high-pressure cylinder through a once-through steam generator, and exhaust steam enters a reheater after passing through a steam-water separator. An outlet of the reheater forms three parallel branches, the first branch is a heating steam extraction pipeline, the second branch is a low-pressure cylinder cooling steam pipeline, and the third branch is a low-pressure cylinder steam inlet pipeline. And the heat supply network heater and the secondary network water-water heat exchanger are subjected to medium isolation, and generated drain water is sent back to the deaerator through a heat supply network drain pump and is merged into the heat regeneration system. The control device adjusts the opening degree of the heating steam extraction adjusting valve and the rotating speed of the first-level network heat supply network circulating water pump and limits the upper opening degree limit of the heating steam extraction adjusting valve and the lower opening degree limit of the low-pressure cylinder cooling steam adjusting valve. High-proportion steam extraction of the reheat steam is achieved, the water supply temperature of the user side and the pressure difference of the two networks are stabilized, cold source losses are reduced, and the disturbance resistance of heat supply operation is enhanced.

A liquid metal cooled nuclear reactor includes a reactor pressure vessel having a core and a multi-stage annular linear induction pump (ALIP) configured to circulate liquid metal coolant through the core, the multi-stage ALIP having a plurality of sets of induction coils at least partially defining separate respective stages of the multi-stage ALIP, the plurality of sets of induction coils configured to be electrically connected to separate respective multi-phase power sources such that the stages of the multi-stage ALIP are independently controlled from one another to adjustably control the flow of liquid metal coolant through the core based on independent control of the multiple multi-phase power sources.

The present application relates to a kind of nuclear power plant airborne radioactive containment ventilation device and group control method, process function piece includes fresh air treatment function piece, air supply and exhaust and negative pressure regulation function piece and purification function piece, fresh air treatment function piece, air supply and exhaust and negative pressure regulation function piece and purification function piece are sequentially connected by pipeline, control piece includes global group control module, fresh air treatment group control module, air supply and exhaust and negative pressure regulation group control module, purification group control module.The nuclear power plant airborne radioactive containment ventilation device and group control method of the present application can make operator only need to send a small amount of simple control instruction by global group control module, realize the function control of each component, realize the automatic operation regulation of ventilation system under different conditions, system one-key start-stop function, can greatly improve the automation level of ventilation system, reduce operator operation load, reduce human error risk, simplify ventilation system control design.

A pressurized water reactor includes a primary reactor coolant circuit flown through by a primary reactor coolant during operation, and a chemical and volume control system for the primary reactor coolant. The chemical and volume control system includes, along the direction of flow of the primary reactor coolant, a letdown line, a high-pressure charging pump with a given discharge pressure, and a charging line leading to the primary reactor coolant circuit. The chemical and volume control system further includes a hydrogenation system with a hydrogen supply and a hydrogen feeding line. In order to achieve efficient and fast hydrogen injection into the primary reactor coolant, a high-pressure feeding pump is arranged in the feeding line to provide a gas pressure higher than the discharge pressure of the charging pump. The feeding line discharges into the charging line.

The invention relates to the technical field of high-temperature gas cooled reactors, and discloses a primary loop micro-positive pressure establishing system and a micro-positive pressure establishing method.The primary loop micro-positive pressure establishing system is characterized in that the first end of a helium purification subsystem is connected with a primary loop, and the second end of the helium purification subsystem is connected with a helium supply and storage subsystem through a first suction structure; one end of the fuel loading and unloading subsystem is connected with a loop; one end of the helium-assisted exhaust subsystem is connected with the other end of the fuel loading and unloading subsystem through a second suction structure; the other end of the helium auxiliary exhaust subsystem is connected with the third end of the helium purification subsystem through a third suction structure; primary loop helium is sucked to the helium supply and storage subsystem through the first suction structure, the second suction structure and the third suction structure, and the primary loop is in a micro-positive pressure state. The primary loop is in the micro-positive pressure state, the maintenance requirement is met, it is guaranteed that helium is controllable, working conditions are created for maintenance, and accidental irradiation events caused by uncontrollable leakage of helium of the primary loop are prevented.

The present application discloses a floating nuclear power generation system in which the position of the float can be controlled. [Solution] This floating nuclear power generation system includes a nuclear reactor, a turbine generator driven by steam generated by the heat of the reactor, a float on which the reactor and turbine generator are located and which is moored at sea, and a circulating water pump that supplies cooling water to a condenser which condenses the steam that has passed through the turbine of the turbine generator, wherein the circulating water pump has a first circulating water pump that is responsible for transporting cooling water discharged from the float in a first direction when the float is viewed from above, and a second circulating water pump that is responsible for transporting cooling water discharged from the float in a second direction opposite to the first direction when the float is viewed from above.

The invention relates to the technical field of reactor waste heat removal, in particular to a heat dissipation system and method for an ocean floating nuclear energy platform reactor, and the heat dissipation system comprises a loop subsystem, a natural circulation loop, a steam power generation subsystem and an ocean hot trap; the first loop subsystem comprises a reactor core, a pressure container and a coolant pipeline; the natural circulation loop comprises an intermediate heat exchanger, a seawater circulation pipeline and a waste heat discharge main pipeline, and the intermediate heat exchanger is connected with the loop subsystem through a coolant pipeline and used for discharging waste heat to the outside through natural circulation; the ocean heat trap takes external seawater as a final heat trap and is connected with a seawater circulating pipeline, a residual heat discharge main pipeline and a steam power generation subsystem; through the dual design of vapor-liquid separation and pressurization backflow prevention, the core condition for generating a condensation water hammer is avoided from the source, pipeline burst is effectively prevented, and the operation safety of the reactor heat dissipation system of the marine floating nuclear energy platform is remarkably improved.

The invention discloses a water system for a nuclear power station and a building method. The water system for the nuclear power station comprises a first pipeline, a second pipeline, a third pipeline and a reservoir. The first pipeline is used for communicating the nuclear power station with a natural water area, and the altitude of the natural water area is H1; the reservoir is built on a mountain, the altitude of the reservoir is H2, and H2-H1 is larger than or equal to 100 m; one end of the second pipeline is communicated with the reservoir, and the other end is communicated with a natural water area; one end of the third pipeline is communicated with the reservoir, and the other end is communicated with the nuclear power station; the water pumping device is arranged on the second pipeline and used for pumping water in the natural water area to the reservoir. The water body is cooled by utilizing the high-altitude reservoir, and an indoor reservoir does not need to be built, so that the building cost of the reservoir is reduced. Therefore, the reservoir can be constructed to be large enough to meet the water consumption of all workshops of the nuclear power station, so that the influence of high temperature on power generation of the nuclear power station is reduced, and the power generation efficiency of the nuclear power station is improved.

The present application relates to the field of nuclear power, in particular to a boron flushing water discharge method and system of an emergency boron injection pump applied to a VVER unit. The method is as follows: the boron flushing water discharge system is set to an initial state; during normal unit power operation, when boron flushing of the emergency boron injection pump is performed, the boron flushing water electric valve is opened, the three-way valve is operated, and the discharge direction is changed from the initial state to the boron-containing wastewater pit, while the flushing water amount is observed to see whether it meets the requirements; if the flushing water amount meets the requirements, the flushing water is adjusted to discharge to the non-boron-containing wastewater pit; if the flushing water amount does not meet the requirements, the water supply flow is adjusted according to the instructions, and after the flushing water amount meets the requirements, the flushing water is adjusted to discharge to the non-boron-containing wastewater pit. The system is connected to the boron flushing water discharge pipeline after the three-way valve, and the boron flushing discharge pipeline is connected to the non-boron-containing wastewater pit. The present application reduces wastewater discharge, increases boron recovery, prevents primary loop water pollution, can also assist in judging valve leakage, and reduces personnel workload.

The invention discloses an indirect air cooling system for a nuclear power plant, comprising a condenser, a bypass processing device, a circulating water pump, an indirect cooling tower and an air cooling radiator; the air cooling radiator is arranged at the bottom of the indirect cooling tower, the exhaust port of an external steam turbine is connected to the shell side inlet of the condenser, the shell side outlet of the condenser is connected to the external condensate pump, the water outlet of the air cooling radiator is connected to the tube side inlet of the condenser, and the tube side outlet of the condenser is connected to the water inlet of the air cooling radiator via the circulating water pump. The system solves the problems of large water consumption and corrosion of open circulating water in nuclear power plants.

The invention discloses a method for detecting an auxiliary water supply system of a nuclear power station unit, and the method comprises the steps: building a temporary pipeline between an output pipeline and a circulation pipeline, and enabling liquid flowing out of a water tank to pass through the output pipeline, the temporary pipeline and the circulation pipeline when the auxiliary water supply system is subjected to an emergency water supply pump test and an emergency diesel generator set performance test; and finally, the water returns to the water tank and is not actually injected into the secondary side of the steam generator. According to the scheme, real water injection to the steam generator can be avoided only by laying one temporary pipeline, the method has the advantages of being simple in construction, short in temporary distance, simple in debugging and temporary management, controllable in water leakage risk and the like, meanwhile, the key path construction period can be effectively saved, a large amount of manpower and construction period investment can be saved, chemical adding operation can be avoided, and the cost is reduced. The detection risk is effectively controlled; whether the auxiliary water supply system meets the design requirement or not can be verified in advance under the condition that the secondary side of the steam generator does not have the water injection condition, problems are found and handled in time, and the hot test is prevented from being affected.

This invention relates to a steam-water separation reheat system and venting method for nuclear power plants. The steam-water separation reheat system connects a condensate drain system, a condensing unit, and a high-pressure heater, and includes: an extraction reheater, a first venting line, and a second venting line. The extraction reheater is connected to the high-pressure heater via the first venting line and to the condensing unit via the second venting line. The extraction reheater is also connected to the condensate drain system, controlling the condensate delivery to the condensate drain system. Under a first operating condition, steam is delivered to the steam side of the high-pressure heater via the first venting line. Under a second operating condition, a portion of the steam is delivered to the steam side of the high-pressure heater via the first venting line, and the remaining steam is delivered to the condensing unit via the second venting line. Implementing the technical solution of this invention, by adding a second venting line and combining it with the first venting line according to actual conditions, allows for flexible responses to different operating conditions and emergency situations, improving the reliability of the entire steam-water separation reheat system and optimizing venting efficiency.

The high-frequency suction device between the evaporator heat transfer pipes comprises a suction belt assembly, the suction belt assembly comprises a supporting belt body which can be arranged in gaps of the heat transfer pipes in an inserted mode, a suction channel is formed in the supporting belt body, and the supporting belt body is provided with a suction hole which penetrates through the side wall and is communicated with the suction channel; the supporting belt body is mounted on the wall climbing mechanism; the wall climbing mechanism can suck the belt assembly into the evaporator heat transfer pipe, and then the high-frequency suction mechanism generates negative pressure to suck foreign matter in the evaporator heat transfer pipe through the suction holes; meanwhile, the high-frequency suction mechanism is used for controlling the suction frequency, the suction belt assembly is driven to vibrate through the change of the suction frequency, the vibration is transmitted to the foreign matter body to change the static state of the foreign matter between the heat transfer pipes, so that the foreign matter is easily sucked, and the vibration of the suction belt assembly enables the foreign matter entering the suction belt assembly not to easily block a hole channel.

This disclosure provides a modular high-temperature gas-cooled reactor (HTGR) secondary loop start-up and shutdown system. The system is configured with a corresponding number of deaerators based on the number of reactors. Each reactor is supplied with water by its own corresponding deaerator, and the start-up and shutdown processes of each reactor are independent of each other. Each deaerator controls its feedwater temperature according to the reactor's stage, reducing mutual interference between reactors, simplifying system operation, and minimizing temporary measures and operating procedures during startup. This system makes the modular HTGR secondary loop system more independent and rational, reduces coupling between reactors, enhances operational adaptability, aligns with the development trend of HTGRs, and provides a more rational and streamlined modular HTGR secondary loop system, making its start-up and operation more flexible and reliable, and better conforming to the modular development trend of HTGRs.

The invention relates to a device special for drainage and exhaust of a nuclear power station and a using method of the device. The device comprises a modularized drainage tank, a three-way connecting fastener, a four-way connecting fastener, a hoop assembly and a drainage hose. A #-shaped flow guide baffle, a stainless steel filter screen, a box body, a drainage pipe frame and a drainage pipeline are integrated in the modular drainage tank. The drain pipe frame is connected with the three-way connecting fastener and the four-way connecting fastener through drain hoses, and a distributed radioactive waste liquid collecting node network is constructed and formed. Radioactive waste liquid at multiple drainage and exhaust points is collected to the modular drainage tank through the network, is guided and buffered by the #-shaped guide baffle and filtered by the stainless steel filter screen, and is discharged to a specified collection point through the drainage pipeline. Compared with the prior art, through modular design and standardized interfaces, mounting and dismounting processes are simplified, working efficiency is improved, centralized diversion and treatment of radioactive waste liquid at multiple drainage and exhaust points are realized, and labor intensity and safety risk of drainage and exhaust maintenance personnel are reduced.

A steam generator for a nuclear power plant is provided. The steam generator has an upright, elongated body and a tension member that extends from a proximal end, which is mounted at each mounting position on the elongated body, to a distal end, which is configured to be mounted at each load-bearing mounting position within the nuclear power plant. Each proximal end is lower than its respective distal end. When in use, the tension member receives a tensile load from the steam generator and transmits substantially all of the weight of the steam generator to the load-bearing mounting positions within the nuclear power plant.