25results 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.

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 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.

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 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.

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.

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.

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.

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.

The invention provides a multi-state sensing self-adaptive water filling method for a primary loop of a reactor, and relates to the technical field of reactor cooling, and the method comprises the following steps: collecting water flow related characteristic data in the primary loop; calculating an overall change value of the water flow; collecting data of temperature and water level in the reactor, and comparing the collected data with the initial temperature and water level of the reactor before the time T; comparing the total change value of the water flow velocity, the water flow rate, the water flow temperature and the water flow pressure with the total change coefficient of the reactor; through continuous contrastive analysis, how the temperature and the water level of the reactor are influenced by changes of various data of water flow in the loop can be quantified. The disturbance source is judged through a dynamic comparison mechanism of the total change value of the water flow and the total change value of the reactor, and meanwhile, the system can dynamically switch the control modes according to the comparison result and can adaptively adjust parameters, so that the water filling adjustment is quicker and more accurate.

The invention relates to the technical field of thermal hydraulic control of a nuclear power plant, and discloses a starting water supply system applied to a modular high-temperature gas cooled reactor nuclear power plant, which comprises a starting water supply deaerator, a starting water supply pump, a water supply pipe network assembly and a distributed control system. The water supply pipe network is physically connected to the downstream of the main feed pump outlet isolation valve. The control system adjusts the deaerator to heat steam and replenish water based on the energy balance principle, so that the feed water temperature tracks the dynamic set value in real time; thermal shock protection logic is integrated, water inlet temperature difference and temperature change rate are limited, and thermal stress damage of equipment is prevented; undisturbed switching logic based on flow conservation and cross amplitude limiting is adopted, and the total flow is maintained to be constant through double-valve cooperative adjustment when switching from starting water supply to main water supply. The problem of feed water temperature matching under asymmetric operation of a multi-module unit is solved, stable switching of water supply modes is achieved, and the standby water supply function under the working condition that main water supply is lost is achieved.

This invention discloses a coupled operation system and method for molten salt reactor power generation, energy storage, and heating, including a reactor primary loop, a molten salt secondary loop, a high- and low-temperature molten salt energy storage and release loop, a molten salt reactor power generation loop, and a heating loop. The reactor primary loop is connected to the molten salt secondary loop, and the molten salt secondary loop is connected to the high- and low-temperature molten salt energy storage and release loop, the molten salt reactor power generation loop, and the heating loop. This system and method can realize the coupled operation of molten salt reactor heating and power generation, achieving energy cascade utilization, while ensuring the stability of system operation.

The invention provides a method and a device for controlling a water feeding pump of a nuclear reactor, and relates to the field of nuclear reactor control, in-pipe pressure of a water feeding main pipe and in-pipe pressure of a steam main pipe are obtained at preset time intervals within a preset time length, and the water feeding pump of the nuclear reactor is controlled according to a deviation value between a first pressure difference determined by the two pressure values and a preset target pressure difference. If the deviation value is continuously larger than the first preset threshold value, the rotating speed of the water feeding pump is increased; and if the deviation value is continuously smaller than a second preset threshold value, the rotating speed of the water feeding pump is reduced, and therefore pressure sudden change of the steam main pipe and the water feeding main pipe is responded and adjusted. Thus, in the scene that the water level deviates from the safety interval, the secondary loop feed water flow can be adjusted according to the deviation value, rapid recovery and stabilization of the water level in the steam generator are achieved, and the adverse effect of violent fluctuation of the water level on safe operation of a nuclear reactor is avoided.

The invention relates to the technical field of high-temperature gas cooled reactors, and discloses a high-temperature gas cooled reactor primary loop atmosphere switching device and a high-temperature gas cooled reactor primary loop atmosphere switching method. The pressure relief system is connected with the pressure reduction adjusting system; the vacuumizing system is detachably connected with the pressure relief system; the pressurization adjusting system is connected with a primary loop of the reactor; and the depressurization adjusting system, the pressure relief system, the vacuumizing system and the pressurization adjusting system jointly realize atmosphere switching, pressurization and depressurization operations of the primary loop of the reactor. According to the invention, a temporary interface does not need to be additionally arranged on the primary loop of the existing reactor, and the boundary of the primary loop of the reactor does not need to be opened during atmosphere switching, so that suffocation risks and radiation risks caused by leakage of micro-positive pressure helium can be avoided, and the sealing uncertainty caused by repeated disassembly and assembly of the boundary of the primary loop is reduced; the operation convenience and the operation efficiency are improved, and rework and industrial potential safety hazards caused by leakage are reduced.

To reduce a material quantity while maintaining the reliability of a fast reactor.SOLUTION: A pump motor control device 100 comprises: an inverter 101 that converts a voltage and a frequency of AC power supplied from a power supply P; bypass lines 102A, 102B, and 102C for directly supplying AC power supplied from the power supply P to a pump motor 7; and a switching control unit 108 that switches between outputting converted power that is the AC power after being converted by the inverter 101 to the pump motor 7 and outputting bypass-line power that is the AC power supplied from the power supply P via the bypass lines 102A, 102B, and 102C to the pump motor 7.SELECTED DRAWING: Figure 2

A canned rotodynamic flow machine (1) configured for operating with a working fluid such as molten salt of a molten salt nuclear reactor. The stator windings are formed by one or more electrically conductive solid bars (12).

The invention provides an intelligent water filling and draining and volume control system for a primary loop of a nuclear reactor, which comprises a main pipeline, a compensation tank is mounted on the upper side of the main pipeline, a screw rod is rotatably matched with the inner wall of the compensation tank, a threaded sleeve is in threaded fit with the lower end of the screw rod, and a piston head is fixedly connected to the lower end of the threaded sleeve. And a driving mechanism is arranged on the upper side of the compensation tank. A lead screw driven by a servo motor and a threaded sleeve transmission structure are arranged, coordinated regulation and control are matched with a water filling and draining regulating valve, and a pawl-ratchet wheel locking mechanism and a double-unlocking design are combined, so that the volume of a loop is rapidly and accurately compensated, the pressure fluctuation problem caused by sudden volume change is reduced, mechanical locking takes effect automatically during power failure to prevent cooling liquid from flowing backwards, and the service life of the loop is prolonged. The manual unlocking mechanism and the lead screw are designed in a separated mode to guarantee emergency operation reliability, the sealing guide structure and the isolated locking component are installed in a matched mode, the operation safety and stability of the system are improved, and the requirements of the nuclear reactor for the severe working condition are met.

The utility model belongs to the technical field of nuclear industry, and particularly relates to a nuclear main pump motor and a liquid level detection device for lubricating oil of a bearing chamber of the nuclear main pump motor, the liquid level detection device comprises a shell and a liquid level detection assembly, the shell is provided with a containing cavity, and the inner side wall, facing the containing cavity, of the shell is provided with a shielding part; the liquid level monitoring module of the liquid level detection assembly is installed in the containing cavity, and the shielding part surrounds the containing cavity, so that the shielding part can effectively prevent external nuclear radiation from entering the containing cavity, the influence of the nuclear radiation on electronic components such as the liquid level monitoring module is reduced, and normal work of the liquid level detection device is ensured; according to the liquid level detection device, the liquid level measurement precision of the liquid level monitoring module and the accuracy of liquid level information signal transmission are improved, meanwhile, the risk that the structure of the liquid level monitoring module is damaged by nuclear radiation can be reduced through the arrangement of the shielding part, the maintenance frequency of the liquid level monitoring module is reduced, the service life of the liquid level monitoring module is prolonged, and therefore the practicability of the liquid level detection device is overall improved.

The disclosure relates to a device and a method for determining data representing the risk of a pumping station clogging event at an industrial site, which includes the following steps: obtaining (S01) at least one time series Q of current environmental conditions (DRCEQ); determining (S02), within historical data in the form of at least one past time series X of environmental conditions (DRCEX), a correspondence between a variable in time series Q and a variable in said at least one time series X; when at least a partial correspondence is determined, obtaining (S03), using a time series Y of past pumping station clogging events, the risk data for the occurrence of a pumping station clogging event. Fig. 1

This invention discloses a method for reducing effluent emissions from the conventional island of a nuclear power plant. The conventional island of a nuclear power plant is equipped with a feedwater chemical sampling system, a steam conversion system, and a condenser circulation system. This method reduces the overall effluent emissions from the conventional island of the nuclear power plant by shutting down some pipelines of the feedwater chemical sampling system, reducing the effluent emissions from the steam conversion system, and reducing the backwashing frequency of the condenser circulation system. This reduces ammonia nitrogen emissions, alleviates environmental monitoring pressure, and improves environmental friendliness. At the same time, the downstream wastewater discharge is reduced, and the upstream water production can be further reduced, thus reducing water production costs and workload and improving economic efficiency.

A system for producing electricity having a primary coolant loop having, in operable and fluid coupling, a vertical nuclear reactor, a vertical once-through steam generator, at least one external hot leg conduit, and at least one external hot leg conduit. The at least one external hot leg conduit fluidly is coupled with the vertical nuclear reactor and to the vertical once-through steam generator to deliver a primary coolant from the vertical nuclear reactor to the vertical once-through steam generator. The at least one external cold leg conduit is fluidly coupled with the vertical nuclear reactor to the vertical once-through steam generator to return the primary coolant from the vertical once-through steam generator back to the vertical nuclear reactor. At least one booster pump is operably coupled to the primary coolant loop to force flow of the primary coolant through the primary coolant loop during normal operation of the system.