363results about "Cooling arrangement" patented technology

The embodiment of the invention provides an integrated small reactor passive residual heat removal test heat exchange structure, system and method, and relates to the technical field of nuclear reactor safety system tests. The system aims to solve the problems that in an integrated small reactor passive quay crane test system, split type and pipeline connection are adopted for containment cooling, so that the system is tedious and large in occupied space. And the heat exchange structure is directly connected with the containment simulation device and the cooling water tank through the shared heat exchange part. The test system comprises the test heat exchange structure, and the test method is implemented by adopting the test system. And heat is transferred from the high-temperature containment simulation device to the low-temperature cooling water tank fluid through the heat exchange part. Redundant connecting parts of the common heat exchange part are omitted, and occupied space is reduced. Efficient and rapid heat transfer is achieved, waste heat is discharged in time, safety is guaranteed, and due to the compact structural design, the whole test device is more reasonable in layout, saves space and is convenient to install, debug and maintain.

An installation includes: a plurality of pilings securable to a bed under a surface of a body of water; a base structure disposed atop the plurality of pilings; and a module disposable on the base structure, wherein the module is positioned and securable on the base structure after being floated on the surface of the body of water over the base structure.

The invention relates to the technical field of cooling of particle accelerators and nuclear power equipment, and discloses a cooling device suitable for high-power rays, which comprises an absorber assembly, a cooling water system assembly, a vacuum tube assembly, a shield assembly and a support assembly. The absorber assembly is formed by welding internal graphite and external copper through a brazing process, and air holes penetrating to a graphite layer are formed in the copper and communicated with an air cavity in the graphite to form a vacuum sealing cavity. The cooling water system assembly comprises a water segregator with two layers of water inlet ring cavities and two layers of water outlet ring cavities, and uniform flow distribution is ensured. And the absorber cooling water pipe is embedded into the copper and is connected with the U-shaped pipe to form a cooling loop. The electron beam bombards the graphite to generate heat, and the heat is transferred to the copper and taken away by cooling water in the cooling water pipe embedded in the copper. The shield assembly is arranged around the vacuum tube assembly to prevent radiation leakage. The cooling device realizes effective heat dissipation of 8GeV high-power rays, and has excellent radiation resistance and extremely long fatigue life.

Disclosed in the present invention is a pressure control valve capable of replacement, the pressure control valve comprising a shell, and a throttle valve and a check valve which are mounted in the shell, wherein the shell comprises a first shell portion and a second shell portion detachably connected to a lower end of the first shell portion: the first shell portion is internally provided with a first accommodating cavity penetrating two opposite end surfaces of the first shell portion, the second shell portion is internally provided with a second accommodating cavity penetrating two opposite end surfaces of the second shell portion, and the first accommodating cavity and the second accommodating cavity are connected and in communication with each other: and the throttle valve is limited in the first accommodating cavity, and the check valve is limited in the second accommodating cavity. The pressure control valve of the present invention is used in a nuclear main pump, the separated arrangement of the shell realizes assembly and disassembly and facilitates the replacement of the throttle valve and the check valve inside: and the throttle valve and the check valve can be miniaturized and thus require a small mounting space.

This invention discloses a passive spray system for aerosol removal in nuclear power plants, comprising: a condensate collector, a guide pipe, a spray water tank, a spray pipe, a burst valve, and nozzles. The condensate collector is connected to the spray water tank via the guide pipe and is used to collect condensate within the containment. The spray water tank is connected to the nozzles via the spray pipes. The burst valve is installed on the spray pipes. When the pressure inside the containment reaches the activation trigger pressure of the burst valve, the burst valve opens, and water from the spray water tank flows into the nozzles through the spray pipes, where it is sprayed to remove aerosols from the containment. The beneficial effects of this system are as follows: compared with natural removal methods, this system provides more efficient removal of suspended aerosols within the containment; compared with active spray systems, this system does not require pumps and power supplies, and has higher reliability under accident conditions; this system can collect condensate near the walls, providing a more sustainable water source for the spray system.

The embodiment of the invention relates to the field of cooling devices of pressure vessels, in particular to an oxygen control device, a lead-bismuth reactor and a method for adjusting the oxygen content of lead-bismuth alloy in the reactor. According to the oxygen control device provided by the embodiment of the invention, the oxygen content data of the lead-bismuth alloy in the container is measured and obtained in real time by arranging the measuring piece and the control assembly, and then the oxygen content level in the container is adjusted in real time by utilizing the deoxygenation mass exchanger and the oxygen release mass exchanger; therefore, the oxygen concentration in the lead-bismuth alloy is accurately maintained in a preset range, so that stable operation of a cooling system is ensured; the closed space is formed among the container, the oxygen removal mass exchanger and the oxygen release mass exchanger, so that the oxygen content adjustment process of the lead-bismuth alloy is realized in a closed environment, impurities are prevented from being mixed in, and the quality of the lead-bismuth alloy is ensured. According to the lead-bismuth reactor provided with the oxygen control device, the content of oxygen in the lead-bismuth alloy can be actively adjusted, and long-term stable operation of the reactor is achieved.

The application discloses a kind of liquid metal cooling pool type reactor multifunctional passive residual heat removal systems based on heat pipe, belongs to reactor system component design technical field, the system is by high-temperature heat pipe, main container wall pit, gas storage tank, storage tank valve, gas heater, gas output pipe, gas input pipe, air extractor, safety container, main container, reactor pit, cold air descending passage, cylindrical thermal insulation layer, hot air ascending passage, connecting pipeline, exhaust chimney, heat radiation plate and in-pile fixed facility composition.The application can be in the condition of reactor total plant power failure, main heat exchanger water interruption and earthquake and other accident conditions, start passive accident residual heat removal system, and the residual heat of reactor core is removed to final heat sink by high-temperature heat pipe.Reactor is in low-power operation or shutdown condition, and auxiliary heating is carried out to in-pile liquid metal, to ensure that coolant is in molten state and promote flow.The system structure is simple, and has independence, passivity, inherent safety.

This invention belongs to the field of online water filling technology after nuclear power plant heat exchanger maintenance, specifically relating to an online water filling system for nuclear power plant heat exchangers after maintenance. The invention includes a No. 1 heat exchanger and a No. 2 heat exchanger, which are connected in parallel via pipelines. The inlet ends of the parallel pipelines form two parallel loops: a flow loop and a bypass loop. It also includes a No. 1 cooling water pump and a No. 2 cooling water pump, which are connected in parallel. One end of the parallel loop of the No. 1 and No. 2 cooling water pumps is connected to the outlet end of the parallel pipeline of the No. 1 and No. 2 heat exchangers, and the other end is connected to the bypass loop via a user connection to the cooling water system. This invention can reduce the risk of human error during online water filling of nuclear power plant heat exchangers after maintenance, and improve the safety and reliability of the heat exchanger filling process.

A reactor containment vessel vent system continuously releases steam generated in a reactor containment vessel to the atmosphere even when a power supply is lost. In the reactor containment vessel vent system, the noble gas filter that allows steam to pass through but does not allow radioactive noble gases to pass through among vent gas discharged from the reactor containment vessel is provided at a most downstream portion of the vent line. An immediate upstream portion of the noble gas filter and the reactor containment vessel are connected to each other by the return pipe via the intermediate vessel. When the radioactive noble gases having pressure equal to or higher than predetermined pressure stay in the immediate upstream portion of the noble gas filter, the staying radioactive noble gases flows into the intermediate vessel by the relief valve. Thus, steam permeability is maintained and the system continuously releases the steam.

A nuclear power plant comprising at least one nuclear reactor having a steam generator, a steam cycle including the steam generator, a first turbine fluidly coupled to the steam generator, a first condenser fluidly coupled to the first turbine, and a feedwater heater fluidly coupled to the first condenser and the steam generator, an intermediate thermal loop including the first condenser, and an evaporator, and an organic cycle including the evaporator, a second turbine fluidly coupled to the evaporator, a second condenser fluidly coupled to the second turbine, and a heat exchanger fluidly coupled to the second condenser and the evaporator.

The present invention relates to an electromagnetic induction pump, comprising: - a tubular internal sensor including a plurality of internal combs and an internal coil positioned between the teeth of the internal combs, the interior of the internal sensor forming a cavity; - an internal tube positioned around the internal sensor; - a pumping channel allowing circulation of the fluid to be pumped; - an external tube, the channel being located between the internal tube and the external tube; - a tubular external sensor positioned around the external tube and including a plurality of external combs and an external coil positioned between the teeth of the external combs; - the electromagnetic pump further comprising a moving device for moving the internal combs and capable of changing the radial clearance between the internal combs and the internal tube, and including a first portion positioned inside the internal sensor and connected to the internal combs to enable radial movement of the internal combs, and a second portion extending at least partially outside the internal sensor and connected to the first portion in a manner that controls the first portion.

The application provides a hole plate reaming device for non-active system injection flow channel test, which is capable of installing a sleeve along a test pipeline in a radial direction, moving in an axial direction by cooperating with an axial guide rail, and rotating and reciprocating synchronously with a reaming cutter driven by a main shaft cutter bar, so as to complete online reaming operation without disassembling the throttle hole plate, effectively save the complicated process of multiple disassembly, processing and reassembly in traditional debugging, and greatly shorten the non-active system injection flow channel test and debugging period.

The present application relates to a kind of nuclear power plant surge line nozzle heat shield sleeve structure and system, comprising: support ring, fluid distribution cover, heat shield sleeve, nuclear power plant surge line nozzle and cladding layer.The nuclear power plant surge line nozzle heat shield sleeve structure and system described above when actually used, through the one end of support ring and the cladding layer of surge line nozzle connection, the other end and fluid distribution cover connection, the one end of heat shield sleeve is inserted into support ring and is connected with support ring, to form gap between the other end of heat shield sleeve and the safety end of surge line nozzle, radioactive particles generated in the annular space between heat shield sleeve and surge line nozzle during reactor operation can be discharged from the gap, thereby avoiding the difficulty and cost of in-service inspection caused by the fact that the safety end of heat shield sleeve surge line nozzle forms a sealed space, so that the risk of high radiation faced by inspection and maintenance personnel is reduced.

The present application relates to nuclear reactor irradiation technology field, especially provide a kind of coolant bypass control structure for pressurized water reactor test loop, including pressure pipe, the pressure pipe is placed in the inside of heat pipe, spherical lower connector is arranged at the bottom of pressure pipe, the inside of pressure pipe is equipped with shunt pipe, and square round connector is arranged at the bottom of shunt pipe;Loop coolant flows to spherical lower connector through pressure pipe and turns back, and loop coolant is divided into main stream and bypass stream by square round connector at the bottom of shunt pipe;Fuel assembly is installed in the inside of shunt pipe;Lower tube base is arranged at the lower side end of fuel assembly, main stream loop coolant flows through the inside of fuel assembly through the center through-hole of lower tube base, and bypass stream loop coolant flows through the annular gap between lower tube base and shunt pipe from the annular gap between square round connector and lower tube base.The purpose is to retain a small part of bypass stream in fuel assembly irradiation test, and form stable and controllable main stream bypass stream distribution mechanism.

A fuel assembly (400) for a nuclear fission reactor system (100). The fuel assembly (400) may extend along a longitudinal axis (420). The fuel assembly (400) may comprise a pressure tube (402) and a single sleeve member (430) provided radially inward of a radially inner surface (432) of the pressure tube (402). The sleeve member (430) may define a chamber (434) which extends along a length of the sleeve member (430). The chamber (434) may comprise, in series, a first sub-chamber (436), a second sub-chamber (438) and a third sub-chamber (440). The first sub-chamber (436) may define a chamber flow inlet (442) and the third sub-chamber (440) defines chamber flow outlet (444). There may be provided a fuel compact unit (500) located in the second sub-chamber (438) within the sleeve member (430) such that a clearance gap (522) is maintained between a radially inner surface (450) of the sleeve member (430) and the fuel compact unit (500) to define a sleeve member flow passage (452).

A method of operating a nuclear reactor system includes: operating a first nuclear reactor to produce steam, the first nuclear reactor positioned in a borehole that extends from a terranean surface through one or more subterranean formations; shutting down the first nuclear reactor; placing a second nuclear reactor into the borehole; and operating the second nuclear reactor to produce steam. A method for controlling a nuclear reactor system includes boron injection and removal, and adjusting a pressure in at least one pressurizer vessel of a pair of pressurizer vessels. A first pressurizer vessel is in fluid communication with primary coolant at a first depth of a reactor vessel. A second pressurizer vessel is in fluid communication with the primary coolant at a second depth of the reactor vessel. The reactor vessel is positioned in a borehole that extends from a terranean surface through one or more subterranean formations.

The application discloses a reactor safety system, a shell member of which comprises a first shell and a second shell; a first injection pipeline is arranged in the first shell and the second shell and is communicated with a reactor core of a reactor; a first injection subsystem of a passive injection system comprises a first injection pool arranged outside the second shell, the first injection pool is higher than the reactor, and the first injection pool is communicated with the first injection pipeline; the first injection pool is configured to make the stored coolant in the first injection pool under the action of gravity to be injected into the reactor core through the first injection pipeline under a first working condition; a passive heat removal system comprises a first heat removal subsystem, a heat exchange mechanism of the first heat removal subsystem comprises an evaporator and a condenser communicated with the evaporator, the evaporator is arranged in the inside of the first shell, the condenser is arranged outside the second shell, and the evaporator is provided with a cooling medium. The first injection subsystem and the first heat removal subsystem are matched, so that the accident can be effectively controlled.

According to an embodiment, a nuclear plant has: a plurality of inner shell radial rebars 27 that are arranged at least partially in a base slab 26 of a double shell containment vessel 20, arranged radially having a center line of the base slab as a center of radiation angle, and penetrate an inner shell base part 21a; a plurality of outer shell radial rebars 28 that are arranged radially having the center line of the base slab as a center of radiation angle, and penetrate the outer shell base part 22a; and a plurality of common radial rebars 29 that penetrate both the inner shell base part 21a and the outer shell base part 22a, being a part of the inner shell radial rebars 27 and also a part of the outer shell radial rebars 28, wherein the number of the outer shell radial rebars 28 is larger than the number of the inner shell radial rebars 27; and the maximum number of the common radial rebars 29 is equal to the greatest common divisor of the number of the inner shell radial rebars 27 and the number of the outer shell radial rebars 28.

A liquid metal cooled nuclear reactor includes a passive decay heat removal system having thermal insulation attached to a wall of a cold source reservoir that holds a phase change material, where the insulation is arranged to automatically fall by gravity from the wall in response to the wall reaching a predetermined temperature. The nuclear reactor may be a fast neutron reactor that incorporates an integral system having a final cold source with a reservoir incorporating an integral exchanger divided into a plurality of parallel tubes between which a phase change material is inserted, the reservoir being surrounded by a thermal insulating layer that can be detached in a passive manner in the event of reaching a predetermined threshold temperature.

The utility model provides a residual heat removal and auxiliary heating system of a reactor. The residual heat removal and auxiliary heating system comprises a heat exchanger, a pipeline group, a valve group, a cooler, a pump and an electric heater, and the heat exchanger is filled with a heat transfer medium. The pipeline group comprises a liquid outlet pipeline, a middle pipeline and a liquid return pipeline; a liquid outlet of the heat exchanger, the liquid outlet pipeline, the middle pipeline, the liquid return pipeline and a liquid inlet of the heat exchanger are connected in sequence; wherein the liquid outlet pipeline is provided with a first branch pipeline and a second branch pipeline which are arranged in parallel, and the liquid return pipeline is provided with a third branch pipeline and a fourth branch pipeline which are arranged in parallel. The cooler is arranged on the first branch pipeline. The pump is arranged on the middle pipeline. The electric heater is arranged on the third branch pipeline. According to the waste heat removal and auxiliary heating system, two different functional requirements of waste heat removal and auxiliary heating can be integrated into one system through the shared pipeline and the shared equipment, and the waste heat removal and auxiliary heating system has the advantages of being simple in system design, low in construction difficulty, low in equipment acquisition cost and the like.

The invention relates to the field of nuclear reactors, in particular to a lower chamber structure suitable for a pressurized water reactor. The lower chamber structure is characterized in that a reactor core supporting plate is welded at the lower end of the hanging basket barrel, and a water flowing through hole is formed in the reactor core supporting plate; a plurality of radial supporting key grooves are uniformly formed in the lower surface of the reactor core supporting plate in the circumferential direction; a plurality of radial supporting keys are arranged on the RPV lower sealing head, the positions of the radial supporting key grooves correspond to the positions of the radial supporting keys, and the radial supporting keys extend into the corresponding radial supporting key grooves in the reactor core supporting plate; supporting tables are arranged on the two sides of the radial supporting key; a flow distribution cover is hung on the supporting table; drain holes are formed in the surface of the flow distribution cover; and a vortex suppression assembly is arranged in the flow distribution cover and is connected with the reactor core supporting plate. Through the integral flow distribution cover structure, the flow distribution structure is simplified; the reactor core supporting structure is simplified through the reactor core supporting plate, and meanwhile, the secondary supporting structure is simplified through the radial supporting keys on the RPV. And generation of unstable eddy current is inhibited through the eddy current inhibition assembly.

The invention relates to the technical field of high-temperature gas cooled reactor cooling, in particular to a cooling circulation device, a high-temperature gas cooled reactor comprehensive cooling system and a using method. A cooling circulation device comprises a liquid inlet circulation pipeline; a liquid outlet circulation pipeline; the circulating pump set is arranged between the liquid inlet circulating pipeline and the liquid outlet circulating pipeline, one end of the liquid inlet circulating pipeline is suitable for being communicated with the first liquid inlet main pipeline, one end of the liquid outlet circulating pipeline is suitable for being communicated with the first liquid inlet main pipeline, and cooling liquid flows in the first liquid inlet main pipeline; a first valve is arranged on the first liquid inlet main pipeline between the liquid inlet circulating pipeline and the liquid outlet circulating pipeline, and the communicating position of the liquid inlet circulating pipeline and the first liquid inlet main pipeline is located on the upstream of the communicating position of the liquid outlet circulating pipeline and the first liquid inlet main pipeline. The invention provides a cooling circulation device, a high-temperature gas cooled reactor comprehensive cooling system and a use method, and aims to solve the problems that power pumps and pipelines need to be additionally arranged when each system of a reactor is in an active state, the arrangement is complicated, and the cost is increased.

A buoyancy driven fluid system coupled to a reactor system configured to achieve free convection operable to cool the reactor system is disclosed. The buoyancy driven fluid system of the present disclosure generates natural circulation by designing the reactor system to have a large vertical offset between the heat exchanger and the reactor core thereby generating a large buoyancy force between a thermal center of the reactor core and a thermal center of the heat exchanger. By ensuring that the sum pressure drop of the components connected to the primary fluid loop is no greater than the buoyancy force of the system, the fluid may circulate throughout the reactor system without the aid of pumps or other forced flow mechanism. The reactor system may be designed within certain size constraints to maintain a compact form while still providing free convection.

Holding device, in particular for attaching a high-pressure cooler (A) to a reactor coolant pump (B), with a first mounting assembly (1) for attachment to an upper area of ​​a cylinder of the high-pressure cooler (A), a second mounting assembly (10) for attachment to a lower area of ​​the cylinder of the high-pressure cooler (A), a third mounting assembly (4) for attaching the first mounting assembly (1) to a motor housing of the reactor coolant pump (B) and for attaching it to the mounting assembly (1), and a fourth mounting assembly (7) for attachment to the motor housing of the reactor coolant pump (A) and for attachment to the second mounting assembly (10) and to the third mounting assembly (4).

This invention provides a method for controlling the oxygen content in reactor coolant, comprising: determining the relationship between the growth thickness of the oxide layer formed on the surface of the fuel rod cladding within a preset time period under conditions of no coolant corrosion and the oxygen content in the coolant; determining the removal rate of the oxide layer on the surface of the fuel rod cladding after reaction with the coolant; determining a functional relationship between the remaining oxide layer thickness on the cladding surface and the oxygen content of the coolant under varying oxygen content conditions, based on the relationship between the oxide layer growth thickness and the oxygen content in the coolant and the oxide layer removal rate; using the functional relationship between the remaining oxide layer thickness on the cladding surface and the oxygen content of the coolant as the objective function, and determining the optimal oxygen content control strategy of the coolant corresponding to the minimum value of the remaining oxide layer thickness based on a preset optimization algorithm. This invention alleviates the oxidation and corrosion problem of fuel rod cladding materials in metallic coolants by determining the optimal oxygen content control strategy of the coolant.

Safety container cooling system (10) comprising a sealed safety container (12) of a nuclear facility, - with a heat exchanger for heat transfer between a medium to be cooled and a cooling medium, - with first means to extract the medium to be cooled from the safety vessel (12) in a first cooling circuit and supply it to the heat exchanger and, after it has flowed through the heat exchanger, return it to the safety vessel (12), wherein the first means comprise a first pumping device (38) to pump the medium, now cooled by the heat exchanger, back into the safety vessel (12), - with second means to supply the recoolant from a coolant reservoir to the heat exchanger and to return it to the coolant reservoir after it has passed through the heat exchanger, wherein the second means comprise a second pumping device located outside the safety vessel (12) to set the recoolant in motion, wherein the second means comprise a turbine (40) which is driven by the flow of the cooling medium, wherein the first pump device (38) is coupled to the turbine (40) in such a way that it is driven by it, characterized by the fact that the first pumping device (38) and the turbine (40) are arranged outside the safety container (12), that the heat exchanger is a condenser (30) arranged outside the safety container (12), that the first means have a withdrawal device (24) in the safety container (12) for withdrawing the medium to be cooled from the safety container (12), which has a steam component and an air component, and at least the steam component is directed to the condenser (30), that the steam component is condensed to water by the condenser (30), that the water is pumped back into the safety container (12) as a cooled medium by the first pumping device (38).

The heat conduction containment is provided with a concrete shell (1) and a bottom plate (2) and is characterized by further comprising a first steel shell (3), a second steel shell (4), a plurality of annular partition plates (5) and an exhaust water conveying pipe (6), the concrete shell (1) is arranged on the outermost side, the first steel shell (3) is arranged on the inner edge face of the concrete shell (1) in an attached mode, the second steel shell (4) is arranged in the first steel shell (3), and the annular partition plates (5) are arranged on the outer edge face of the second steel shell (4). A heat exchange cavity is formed between the second steel shell (4) and the first steel shell (3), the annular partition plate (5) is arranged between the first steel shell (3) and the second steel shell (4) and divides the heat exchange cavity into a plurality of water storage layers, through holes (51) are formed in the annular partition plate (5), so that the water storage layers are in fluid communication, and the exhaust water delivery pipe (6) penetrates through the heat exchange cavity and the exterior of the containment.

The application provides a bubble behavior test device and method for simulating a liquid metal cooled reactor, and belongs to the technical field of nuclear engineering. The device aims to solve the problems of high operating temperature, large power consumption and poor safety of a traditional experimental loop using lead-bismuth alloy as a liquid phase medium, and the shortcoming that a traditional contact type measurement method easily causes disturbance to the flow characteristics of fluid. The bubble behavior test device for simulating the liquid metal cooled reactor comprises a flow channel body, a rod bundle matrix and liquid gallium-indium-tin alloy metal arranged in the flow channel body, and a plurality of gas check valves arranged on each rod of the rod bundle matrix; an inert gas output assembly is provided with a plurality of gas output ports and is connected with the gas injection ports on each rod of the rod bundle matrix; an ultrasonic probe array is arranged on the side of the rod bundle matrix and in the central rod; and a fluid circulation assembly is connected with the flow channel body at the inlet end and the outlet end and is used for driving the fluid in the flow channel body to move from top to bottom and separating the gas from the flow channel body. The device is mainly used for simulating the bubble behavior of the liquid metal cooled reactor.

A nuclear power plant has a nuclear reactor including a reactor pressure vessel which houses plural fuel rods containing fissile material. The nuclear power plant further has means for submerging the reactor pressure vessel in water and thereby water-cooling the reactor pressure vessel in the event of an emergency requiring cooling of the nuclear reactor. The nuclear power plant further has a primary core catcher outwardly of the reactor pressure vessel, the primary core catcher being formed of a material suitable for retaining molten corium in the event corium escapes the reactor pressure vessel. The nuclear power plant further has secondary core catcher outwardly of the primary core catcher, the secondary core catcher lining a tank which is water-filled in normal use of the plant to submerge and thereby water-cool the primary core catcher. The secondary core catcher is also is formed of a material suitable for retaining molten corium in the event corium escapes the primary core catcher.