678results about "Cooling arrangement" patented technology

The invention relates to the technical field of nuclear reactor safety, and particularly discloses a reactor waste heat export, recovery and comprehensive utilization system and method based on a heat pipe, and the system comprises a nuclear energy generation system, a secondary loop energy conversion system, a primary loop waste heat discharge system, a secondary loop waste heat discharge system, and a waste heat recovery and comprehensive utilization system. According to the waste heat recycling system, the heat pipe can be used for heat transfer and storage under the peak regulation operation condition, the nuclear heat use efficiency is remarkably improved, the requirements of various types of energy sources such as electricity and heat are met, and passive export of reactor core waste heat can be achieved through multi-natural-circulation coupling cooperative operation under the condition of a whole-field outage accident. By means of the efficient heat transfer characteristic and the modularization characteristic of the heat pipes, independent operation of the heat pipes can be achieved, the problem of single-point failure of the system can be avoided, meanwhile, separation of a reactor coolant and a cooling medium of a waste heat removal system can be achieved, the inherent safety of a nuclear reactor is remarkably improved, and the requirements of multi-purpose scenes are met; and the purpose of fully utilizing resources is achieved.

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 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 provides a long-term passive cooling system of a compact small reactor, which is characterized in that a reactor and steam generators are arranged in a containment, the steam generators are connected with the reactor, and each steam generator is provided with a secondary side residual heat removal system; the liquid storage chamber is mounted on the outer wall surface of the containment in a surrounding manner; the passive heat exchanger is arranged in the liquid storage chamber and is connected with a secondary side residual heat removal system of the corresponding steam generator, so that steam of the steam generator enters the passive heat exchanger for heat exchange; the plurality of passive heat pipes are mounted on the wall surface of the containment in a penetrating manner and are used for guiding heat in the containment to the liquid storage chamber. The secondary side waste heat removal system is arranged on the secondary side of the steam generator and used for relieving non-loss-of-water accidents, short-term and long-term reactor core cooling is achieved outside the containment through the passive heat exchanger combining water cooling and air cooling, the peak pressure of the containment after the accidents is effectively reduced, long-term passive heat carrying is achieved, and the unit operation safety is improved.

This invention discloses a firecracker-style controlled nuclear fusion and fission synergistic power generation device and method. The power generation device includes a fission-fusion reaction system, a circulating cooling system, and other accessories. The fission-fusion reaction system includes a neutron gun, fuel spheres, a fuel sphere storage and delivery device, and a vacuum reactor. The circulating cooling system includes a circulating fan, a voltage regulator, a steam turbine, a steam generator, delivery pipelines, and a water pump. Other accessories include an energy storage and control system and a vacuum pump. This invention innovatively integrates the start-up mechanism, drawing on the firecracker-style multi-reactor relay concept, combining nuclear fission and nuclear fusion. It utilizes the high temperature and pressure generated by nuclear fission to start nuclear fusion, reducing start-up difficulty and cost. Simultaneously, it efficiently converts and utilizes energy, employing helium for heat transfer. Helium's high thermal conductivity, low density, high specific heat capacity, and chemical inertness allow it to effectively carry away the heat stored in the phase change microspheres, comprehensively collecting and converting nuclear reaction heat energy, thus improving energy conversion efficiency.

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.

The invention provides a shell-and-tube phase change heat storage system for a nuclear reactor containment and a nuclear accident heat energy management method, and belongs to the technical field of nuclear safety engineering and heat energy storage. The problems that an existing shell-and-tube heat reservoir is designed for industrial waste heat recovery and is not suitable for nuclear accidents, and the internal space structure of a containment is complex and limited are solved. The heat storage units are distributed in parallel, and each heat storage unit comprises a shell and a plurality of heat exchange tube bundles arranged in the shell; the heat storage unit has an interchangeable heat storage medium arrangement mode which comprises the first mode that the space in the heat exchange tube bundle is filled with a phase change material, and accident steam flows through a shell side channel between the shell and the heat exchange tube bundle; in the second mode, the shell side space between the shell and the heat exchange tube bundle is filled with the phase change material, and accident steam flows through an in-tube channel of the heat exchange tube bundle; and the accident steam passively flows through the channel which is not filled with the phase change material under the driving of the pressure difference. The method is used in the phase-change heat storage field.

A heat pipe heat exchanger device for use with cooling water in nuclear power plants is provided. [Solution] The heat pipe heat exchanger device includes a treatment device, an anti-clogging assembly, a drive assembly, and an ash cleaning assembly. The anti-clogging assembly includes a filter screen installed inside a housing included in the treatment device, a groove on the outer wall of the filter screen, a partition plate installed inside the housing, and an ash drop groove opening through the outer wall of the partition plate. The ash cleaning assembly includes a storage chamber installed inside the housing, a guide rail on the inner wall of the storage chamber, an ash collection box movably installed inside the storage chamber, and a guide groove on the outer wall of the ash collection box. When the motor is started, the moving column moves up and down to push the filter screen and shake off impurities on the filter screen. As the mounting plate moves, the cleaning brush completes the cleaning work, and as the mounting plate moves, a moving door opens to make it easier for operators to remove dust.

This invention discloses an in-core component with highly uniform core outlet flow rate and its design method, relating to the field of nuclear reactor design technology. It solves the problem of poor uniformity in coolant flow distribution at the core outlet in existing reactors, balances flow resistance at various points at the core outlet, and reduces lateral flow between different channels. The specific solution is as follows: it includes a core top plate and a support column assembly mounted above the core top plate. The core top plate is provided with several water flow holes and mounting holes. The water flow holes consist of circular and square water flow holes. The circular water flow holes are divided into first and second circular water flow holes with different functions. The area ratio of the square water flow holes to the circular water flow holes is 1.1:1 to 1.3:1.

The invention discloses an assembly method of a heat pipe type reactor and the heat pipe type reactor, and belongs to the field of nuclear power. The assembly method of the heat pipe type reactor comprises the following steps: closing one end of each mounting hole of the grid base body, vertically arranging the grid base body to enable the closed end to be located at the bottom, injecting filling slurry containing heat-conducting powder and a dispersing agent into the mounting holes, then inserting pipe fittings into the mounting holes until the filling slurry overflows, and finally assembling the grid base body into the mounting holes. The gap between the pipe fitting and the mounting hole is uniformly filled with the filling slurry in a forced flowing mode; and heating to cure and shape the filling slurry. The method can effectively eliminate the assembly gap between the grid substrate and the pipe fitting of the heat pipe type reactor, and improves the heat transfer efficiency and the safety margin of the reactor.

A nuclear reactor core includes a plurality of fuel elements and moderator blocks to form a nuclear reactor core inner portion and a nuclear reactor core outer portion. The nuclear reactor core inner portion includes an inner moderator matrix formed of a plurality of inner holes that include a plurality of inner fuel openings with one or more fuel elements disposed therein. The plurality of inner holes further include a plurality of inner coolant passages to flow a coolant. The nuclear reactor core outer portion includes an outer moderator matrix formed of a plurality of outer holes that include a plurality of outer fuel openings with one or more fuel elements disposed therein. The plurality of outer holes further include a plurality of outer coolant passages to flow the coolant. The outer coolant passage diameter exceeds the inner coolant passage diameter.

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.

To provide a heat pipe heat exchanger device to be used for cooling water of a nuclear power plant.SOLUTION: A heat pipe heat exchanger device includes a processing instrument, an anti-blocking assembly, a driving assembly, and a dust removing assembly. The anti-blocking assembly comprises: a filter screen arranged in a housing included the processing instrument; a recess groove formed in the outer wall of the filter screen; a partition plate arranged in the housing; and a dust falling groove formed in the outer wall of the partition plate in a penetrating mode. The dust removing assembly comprises: a containing room arranged in the housing; a guide rail arranged on the inner wall of the containing room; a dust collecting box movably arranged in the containing room; and a guide groove formed in the outer wall of the dust collecting box. A motor is started to move a movable column up and down to push the filter screen, thereby shaking off impurities on the filter screen. The mounting plate moves to enable the cleaning brush to complete cleaning work, and in the moving process of the mounting plate, the movable door is opened so that workers can conveniently remove dust.SELECTED DRAWING: Figure 1

The application discloses a kind of high temperature gas cooled reactor module arrangement structures, including reactor plant, and nuclear auxiliary plant, electrical plant and main control room for reactor plant;Two independent, identical high temperature gas cooled reactor nuclear steam supply system modules are arranged in reactor plant.In the present scheme, the high temperature gas cooled reactor module containing two high temperature gas cooled reactor nuclear steam supply system modules is regarded as a standardized product, and can be assembled into a type unit according to market demand, under the premise that each high temperature gas cooled reactor nuclear steam supply system module has independent, identical inherent safety characteristics, so that multiple standardized modules can be independently operated and maintained, and the demand for long-term, continuous and stable supply of high temperature and high pressure industrial steam in various petrochemical parks is met.Furthermore, multiple high temperature gas cooled reactor modules can be flexibly combined to adapt to different load capacity, site conditions and construction scale.

The application is suitable for the field of nuclear reactor technology and discloses a filtering device and filtering method for a lead-bismuth cooled pool type reactor, which comprises a fixing assembly, a branch pipeline and a filtering assembly; the fixing assembly is arranged on the side of the main channel, and the fixing assembly comprises the branch pipeline in which the filtering assembly is detachably installed; the branch pipeline is connected with the side of the main channel through the fixing assembly to form the branch pipeline, and the filtering assembly is used for filtering and purifying in the branch pipeline without providing filtering power by itself; the filtering device as a whole can be arranged in the pool type reactor to ensure the integrity of the boundary of the reactor body; the filtering assembly has a simple structure and is convenient to install and replace, and can be operated in the reactor for a long time, so that the filtering device has a long service life and the replacement period of the filtering device is reduced.

The application provides a deep learning method for predicting thermal hydraulic characteristics of a sodium-cooled fast reactor heat exchanger, and belongs to the technical field of a sodium-cooled fast reactor passive residual heat removal system; the method comprises the following steps: a first DNN is used to approximate low-fidelity flow and temperature distribution; a second DNN is used to build a nonlinear relationship between low-fidelity and high-fidelity data; and a physical information partial differential equation and a boundary condition are used as constraints to inject a third DNN, so as to optimize a prediction result; by combining the advantages of a fluid heat transfer partial differential equation and data driving, the method can realize high-fidelity prediction of flow velocity and temperature in a sodium-cooled fast reactor PCHE by using only low-fidelity data, solves the problem that physical information in a heat exchanger is difficult to measure, and has important significance for improving the inherent safety of a sodium-cooled fast reactor and the safety of a reactor core.

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.

In a sodium-cooled fast reactor, a breached fuel pin releases fission and activation products into the primary sodium coolant and into the cover gas. A portion of the primary sodium coolant is diverted to a distillation system where its constituents can be separated based on volatility. Condensing these constituents enables precise measurement of their concentration without delays caused by waiting for the decay of radioactive sodium background scatter, enhancing the detection and quantification of the dilute constituents. Additionally, the method continuously determines the initial concentration of constituents in the feed. Quantification is achieved using detectors, facilitating the continuous assessment of dilute constituent concentrations in the sodium coolant feed stream.

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.

A nuclear reactor is designed to couple the load path of the control elements with the reactor core, thus reducing the opportunity for differential movement between the control elements and the reactor core. A cartridge core barrel can be fabricated in a manufacturing facility to include the reactor core, control element supports, and control element drive system. The cartridge core barrel can be mounted to a reactor vessel head. Thus, any movement, such as through seismic forces, transmits an equal direction and magnitude to the control elements and the reactor core. This arrangement reduces the opportunity for differential movement between the control elements and the reactor core.

The application discloses a kind of efficient supercritical carbon dioxide high temperature gas cooled reactor systems and its operating method, the system includes pre-cooler, main compressor, re-compressor, high temperature turbine, low temperature regenerator, high temperature regenerator, high temperature helium-supercritical carbon dioxide heat exchanger, low temperature helium-supercritical carbon dioxide heat exchanger and reactor.The application proposes the power generation system of two-stage split supercritical carbon dioxide cycle comprising high-temperature split re-compression cycle loop and low-temperature re-split reheat expansion cycle loop, by arranging high-temperature helium-supercritical carbon dioxide heat exchanger and low-temperature helium-supercritical carbon dioxide heat exchanger in power generation system, ensure the good matching of helium loop heat release temperature window and supercritical carbon dioxide power cycle loop heat absorption temperature window, realize the efficient cascade utilization of high-temperature gas cooled reactor helium loop large-span temperature interval heat.

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.