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995results about "Reactor fuel elements" patented technology
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In-situ global foreign matter inspection grabbing system for lower reactor internals of reactor
The invention discloses an in-situ global foreign matter checking and grabbing system for reactor internals on the lower portion of a reactor. The in-situ global foreign matter checking and grabbing system for the reactor internals on the lower portion of the reactor comprises a base device, an adjusting device and a foreign matter grabbing device. The foreign matter grabbing device is arranged on the base device through the adjusting device. The foreign matter grabbing device comprises a flexible arm assembly, a flexible arm driving assembly, a quick change assembly, at least one clamping jaw used for grabbing foreign matter and a video tracking and checking assembly used for searching the foreign matter and monitoring the grabbing condition. The flexible arm assembly is arranged on the adjusting device; the at least one clamping jaw is detachably arranged on the flexible arm assembly through the quick-change assembly; the video tracking and checking assembly is arranged on the adjusting device; according to the invention, the flexible arm assembly carries various types of clamping jaws to carry out automatic grabbing operation, so that the in-service nuclear power plant reactor internals can accurately, efficiently and safely grab and clean foreign matters in the in-situ interior of a pressure vessel in a whole region.
Owner:YANGJIANG NUCLEAR POWER +1
Device and method for remote installation of heat insulation sleeve relieving cushion block
The invention belongs to the technical field of nuclear power, and particularly relates to a device and method for remote installation of a heat insulation sleeve relieving cushion block. Comprising a gripping apparatus and a screwing tool, the gripping apparatus is used for gripping the cushion block, the bolt and the locking cap to mount and dismount the cushion block, and the screwing tool is used for positioning and mounting the relieving cushion block on the upper cover plate of the control rod guide cylinder and fastening the bolt. The clamping device has the beneficial effects that the air cylinder is adopted to drive the grabhooks to be opened and closed, clamping of the cushion block is achieved, meanwhile, it is guaranteed that the grabhooks are in an open state during air loss through the built-in springs, and the cushion block is prevented from falling off. A grabbing structure that the bolts and the locking caps are lifted along with the cushion blocks is adopted, meanwhile, the anti-falling function is achieved, and integrated operation of clamping, conveying, positioning and screwing of the cushion blocks is achieved.
Owner:CNNC NUCLEAR POWER OPERATION MANAGEMENT CO LTD
High-temperature-irradiation-resistant zirconium alloy and manufacturing method and workpiece thereof
PendingCN120519748AMaterial nanotechnologyOptical rangefindersZirconium alloyPressurised water reactor
The invention discloses a high-temperature-irradiation-resistant zirconium alloy and a manufacturing method and a workpiece thereof, and belongs to the field of zirconium alloys. Wherein the zirconium alloy resistant to high temperature irradiation comprises the following components in percentage by weight: not less than 96% of Zr, 0.61%-1.4% of Sn, 0.41%-1.4% of Nb, 0.1%-0.5% of Fe, 0.11%-0.2% of O and 0.001%-0.3% of Cr; and from 0.015% to 0.03% of Si and / or from 0.001% to 0.06% of Ge; wherein Sn + Nb + O is greater than or equal to 1.5%, Fe / Cr is greater than or equal to 1.5%, Si + Ge is greater than or equal to 0.015%, and (Sn + Nb + O) / (Fe + Cr + Si + Ge) = 1.5-25.0. The alloy has good mechanical properties, corrosion resistance, creep resistance and radiation growth resistance, and is suitable for manufacturing of pressurized water reactor core structural members.
Owner:SHANGHAI NUCLEAR ENGINEERING RESEARCH & DESIGN INSTITUTE CO LTD
Anti-blocking device and anti-blocking method for fuel loading and unloading system of high-temperature gas cooled reactor
PendingCN121768727Aavoid misuseimprove accuracyNuclear energy generationNuclear monitoringReciprocating motionProcess engineering
The embodiment of the invention provides an anti-blocking device and an anti-blocking method for a fuel loading and unloading system of a high-temperature gas cooled reactor, and relates to the field of high-temperature gas cooled reactors. The anti-blocking device of the high-temperature gas cooled reactor fuel loading and unloading system comprises a driving module, a transmission module, an execution module and a control module, the transmission module is connected with the driving module, the execution module is connected with the transmission module, and the transmission module is driven by the driving module to drive the execution module to do linear reciprocating motion. And the control module is connected with the driving module and the execution module at the same time. A cooperative structure of the driving module, the transmission module and the execution module is arranged, a driving signal is converted into linear reciprocating motion of the execution module, mechanical disturbance is applied to a fuel ball stack in a blocked area, the lap joint balance or a bridge junction structure of the fuel ball stack is damaged, and active bridge breaking and jam releasing are achieved. And the jam releasing efficiency and success rate are obviously improved.
Owner:HUANENG NUCLEAR ENERGY TECH RES INST CO LTD
Reactor pressure vessel flange sealing surface defect measuring method and repairing method
The invention provides a reactor pressure vessel flange sealing surface defect measuring method and a reactor pressure vessel flange sealing surface defect repairing method. The measuring method comprises the following steps: S1, measuring and calculating the minimum diameter and the maximum diameter of an annular band formed in a sealing ring indentation range; and S2, by taking the center of the container flange as a circle center, inspecting the defect, falling into the range of the annular belt between the minimum diameter and the maximum diameter, on the container flange, and measuring and recording the size and the orientation of the defect. According to the method, corresponding recommended repair methods are provided according to repair criteria for defects of different scales, the method can be used for guiding implementation of nuclear power plant RPV sealing surface maintenance, the repair workload is effectively reduced, the radiation dose of personnel is reduced, the operation and maintenance efficiency is improved, safe and stable operation of a nuclear power unit is ensured, and the economic benefit of power station operation is improved.
Owner:SHANGHAI NUCLEAR ENGINEERING RESEARCH & DESIGN INSTITUTE CO LTD
Chloride based volatility for the recovery of uranium from nuclear fuel salt
Disclosed herein are methods and systems for the recovery of uranium from used fuel salts using nonaqueous chemistry without electrometallurgy techniques. The used fuel salts are maintained in a molten state in a chlorination chamber, and are then chlorinated for a period of time sufficient to convert at least some UCl3 into a different uranium chloride, thereby generating a uranium-containing gas phase. The uranium-containing gas phase is separated from the resulting residue, and is condensed for collection.
Owner:TERRAPOWER LLC
Radiation-resistant high-entropy alloy magnetron sputtering coating and preparation method thereof
The invention discloses a radiation-resistant high-entropy alloy magnetron sputtering coating and a preparation method thereof, and belongs to the technical field of radiation-resistant coating materials. On the basis of the multi-element synergistic effect and lamination interface design of the high-entropy alloy, FeCrNi and FeNiCrMn are used as target materials, and the coating with the nanometer lamination structure is prepared by accurately regulating and controlling the target current, the gas flow, the deposition time and other process parameters of magnetron sputtering; meanwhile, accurate control over the single-layer film thickness and the interface density is achieved by adjusting the coupling relation between the sputtering power and the deposition rate, and it is ensured that all unit layers are flat and compact, and interface combination is complete. And a multi-heterogeneous interface is formed through the design of a laminated interface, so that a defect capture network is formed, the hardness of a single-layer coating is improved, the hardening rate after irradiation can be lower than that of a traditional single-layer coating, and the problems of crystallization and brittleness failure induced by irradiation are effectively relieved.
Owner:SHANGHAI UNIV
Reactor core high-level detector pulling mechanism
The invention relates to a reactor core high-level detector pulling mechanism which provides effective radiation protection in the pulling process of a detector by introducing a shielding cylinder, so that the radiation dosage possibly received by an operator during operation is remarkably reduced, and the safety of a working environment is improved. Secondly, an alignment connecting device with an alignment camera is further arranged, the position of the lifting and moving assembly can be monitored and adjusted in real time, it is ensured that the lifting and moving assembly is accurately aligned and connected with the detector to be pulled out, and in the process that the shielding cylinder gradually goes deep into the reactor core water pool, the detection accuracy is improved. And position interference between the shielding cylinder and other structural members in the reactor core can be avoided, so that the problem that accurate butt joint is difficult in a traditional method is solved, and the detector pulling-out efficiency and the operation safety are improved.
Owner:GUANGXI FANGCHENGGANG NUCLEAR POWER +1
Nuclear fuel cladding composite material and preparation method thereof
The invention mainly provides a nuclear fuel cladding composite material and a preparation method thereof. The nuclear fuel cladding composite material comprises a zirconium alloy matrix, the niobium-rich diffusion barrier layer is formed on the surface of the zirconium alloy matrix; the Cr coating covers the niobium-rich diffusion barrier layer; and an Al-rich layer formed on the surface of the Cr coating. The preparation method comprises the following steps: firstly, carrying out surface pretreatment on the zirconium alloy matrix; a niobium-rich diffusion barrier layer is formed on the surface of the zirconium alloy matrix through ion implantation; then depositing a Cr coating on the niobium-rich layer; and finally, an Al element is injected into the surface of the Cr coating, and an Al-rich layer is formed so that an Al2O3 protective layer can be formed under the high-temperature working condition. According to the technical scheme, interface mutual diffusion can be effectively inhibited, the stability of the coating is improved, meanwhile, the Cr coating structure is optimized, the oxidation process is delayed, a gradient protection system is constructed, and the oxidation resistance of a high-temperature material is enhanced.
Owner:HARBIN INSTITUTE OF TECHNOLOGY (SHENZHEN) (INSTITUTE OF SCIENCE AND TECHNOLOGY INNOVATION HARBIN INSTITUTE OF TECHNOLOGY SHENZHEN)
Remote communication method, device and equipment for mixing and clarifying tank stirring device
ActiveCN120431708AElectric signal transmission systemsMeasurement devicesFuel reprocessingSignal quality
The invention provides a remote communication method, device and equipment for a mixing and clarifying tank stirring device, belongs to the technical field of nuclear fuel post-processing, and solves the problem of control signal fluctuation caused by electromagnetic interference, cable aging and environment abrupt change in a complex industrial scene. The method comprises the following steps: acquiring signal quality data; preprocessing the signal quality data to obtain preprocessed signal quality data; determining predicted signal attenuation data according to the preprocessed signal quality data; comparing and analyzing the predicted signal attenuation data with a preset threshold value to obtain a control instruction; and controlling the mixing and clarifying tank to communicate according to the control instruction. According to the scheme, the anti-interference performance and the prediction precision can be improved, and the communication failure rate and the energy consumption are reduced.
Owner:THE 404 COMPANY LIMITED CHINA NAT NUCLEAR
Process and device for maintaining upper reactor internals control rod guide cylinder bolt locking cap
The invention belongs to the technical field of maintenance of upper reactor internals of a reactor, and particularly relates to a process and a device for maintaining a bolt locking cap of a control rod guide cylinder of an upper reactor internals. Comprising a bolt torque reinspection tool and a locking cap automatic bending tool. The bolt torque reinspection tool is matched with a torque wrench to realize reinspection of bolt torque; the automatic bending tool for the locking cap is used for controlling the bulging claw to correspond to a groove of a locking bolt head and driving the bulging claw to press downwards to bend the locking cap, so that looseness prevention of the bolt is achieved. The device has the beneficial effects that a bolt torque reinspection tool and an automatic locking cap bending tool are conveyed to a working area through a ring crane, and the field environment is not changed; the bolt torque reinspection tool and the locking cap automatic bending tool are both of a long-rod-shaped structure as a whole, an operator can perform remote operation on an operation platform, and whether the torque of the guide cylinder bolt meets the locking requirement or not can be inspected in cooperation with a torque spanner.
Owner:CNNC FUJIAN FUQING NUCLEAR POWER
Reactor lower reactor internal reactor core supporting plate global foreign matter video inspection grabbing equipment
The invention discloses a reactor lower internals reactor core supporting plate global foreign matter video inspection grabbing device which comprises a rack, a video inspection device and a foreign matter grabbing device, a positioning pin is arranged at the bottom of the rack, and a containing groove used for storing foreign matter is further formed in the rack; the video inspection device is movably installed on the rack in the height direction and comprises a video inspection mechanism used for conducting foreign matter video inspection. The foreign matter grabbing device is movably installed on the rack in the height direction and comprises a foreign matter grabbing mechanism used for grabbing foreign matter. By applying the video inspection and grabbing equipment for the global foreign matters on the reactor core supporting plate of the reactor lower reactor internals, the foreign matters on the reactor core supporting plate can be inspected and grabbed when the reactor internals are in situ, the damage to the components and a pressure vessel body caused by hoisting of the reactor internals can be effectively avoided, the steps of hoisting of the reactor internals and the like are reduced, and the production efficiency is improved. And the overall foreign matter checking and grabbing operation efficiency can be effectively improved.
Owner:YANGJIANG NUCLEAR POWER +1
Fiber reinforced multi-layered wear and corrosion coatings of zirconium alloy nuclear fuel cladding
A covering for reinforcing a base layer of a nuclear fuel cladding is provided. The covering comprises a first layer configured to cover a first portion of the outer surface of the base layer of the nuclear fuel cladding, a second layer surrounding the first layer and the base layer and a third layer surrounding the second layer. The first layer comprises a fiber based material, the second layer comprises an interfacing material and the third layer comprises Chromium. A reinforced cladding for nuclear fuel and a method for producing a reinforced nuclear fuel cladding are also provided.
Owner:WESTINGHOUSE ELECTRIC CORP
Cutting and jam releasing device for fuel ball conveying pipeline of high-temperature gas cooled reactor
PendingCN121964217AGuarantee continuous and stable operationEasy to unlockNuclear energy generationReactor fuel elementsThermodynamicsProcess engineering
The embodiment of the invention provides a high-temperature gas cooled reactor fuel ball conveying pipeline cutting and jam releasing device, and relates to the technical field of high-temperature gas cooled reactors. The high-temperature gas cooled reactor fuel ball conveying pipeline cutting jam releasing device comprises a cutting function block, the cutting function block is provided with a first end and a second end which are opposite, a first mounting hole penetrating through the first end and the second end is formed in the cutting function block, and the first end is used for abutting against a blocked fuel ball; the cutting assembly is arranged in the first mounting hole and connected with the cutting function block, and the cutting assembly is used for cutting the fuel ball abutting against the first end; and the driving assembly is detachably arranged at the second end, and the driving assembly is used for driving the cutting function block to move. The problem that the blocked fuel balls in the conveying pipeline are inconvenient to release at present is solved, and continuous and stable operation of the high-temperature gas cooled reactor is guaranteed.
Owner:HUANENG NUCLEAR ENERGY TECH RES INST CO LTD
Nuclear power station pool foreign matter collecting and transferring device
The nuclear power station pool foreign matter collecting and transferring device comprises a shielding container and a collecting basket, the shielding container comprises a first containing cavity and is provided with shielding materials capable of shielding rays, and the collecting basket comprises a second containing cavity and can be arranged in the first containing cavity; and drainage structures are arranged on the collecting basket and the shielding container. According to the nuclear power station pool foreign matter collecting and transferring device, after the foreign matter is collected, the foreign matter does not need to be transferred underwater again, the collecting basket can be placed in the shielding container or the foreign matter is directly placed in the shielding container, and the shielding container can shield the foreign matter with high radiation dose; radiation of foreign matters to human bodies and the environment can be reduced, the operation risk of personnel is reduced, and use is simpler and more convenient; the shielding container and the collecting basket can be used in cooperation and can also be used independently, and on-site flexible use is facilitated; the arrangement of the drainage structure facilitates the discharge of radioactive residual water in the foreign matters, so as to reduce the foreign matter collection and transfer weight.
Owner:LIAONING HONGYANHE NUCLEAR POWER
Ball path dredging device for high-temperature gas cooled reactor fuel loading and unloading system
The embodiment of the invention provides a ball path dredging device for a high-temperature gas cooled reactor fuel loading and unloading system, and relates to the technical field of high-temperature gas cooled reactor fuel dredging. The ball path dredging device for the high-temperature gas cooled reactor fuel loading and unloading system comprises a box body, a coiling machine, a gas cylinder set, a tool head set, a filtering piece and a negative pressure piece, the box body is provided with a first installation cavity, a second installation cavity, a third installation cavity and a fourth installation cavity, the coiling machine is installed in the first installation cavity, and the gas cylinder set is installed in the second installation cavity; the tool head set is installed in the second installation cavity, the filtering piece is installed in the third installation cavity, and the negative pressure piece is installed in the fourth installation cavity. Key components such as a coiling machine, a gas cylinder group, a tool head group, a filtering piece and a negative pressure piece are intensively arranged in different mounting cavities in the same box body to form an integrated equipment framework. All the functional units are rapidly connected through preset gas paths, circuits and signal paths, and the operation process is remarkably simplified.
Owner:HUANENG NUCLEAR ENERGY TECH RES INST CO LTD
Process for production of actinium-225
InactiveUS20250273355A1Conversion outside reactor/acceleratorsNuclear energy generationRadio isotopesProcess engineering
Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.
Owner:TERRAPOWER ISOTOPES LLC
Circulating device of spherical element and pebble bed high-temperature gas cooled reactor
The invention discloses a circulating device of a spherical element and a pebble bed high-temperature gas cooled reactor, the circulating device of the spherical element comprises a shell provided with a ball receiving pipe used for receiving the spherical element and a ball outlet pipe used for conveying the spherical element, the axial direction of the ball receiving pipe is consistent with the falling direction of the spherical element, and the axial direction of the ball receiving pipe is consistent with the falling direction of the spherical element. An included angle is formed between the ball outlet direction of the spherical element and the axial direction of the ball receiving pipe, so that the spherical element can be conveyed back to the reactor core along the ball outlet pipe after circulating in the shell; a rotor is coaxially and rotatably arranged in the shell, the rotor is provided with a bearing assembly for bearing the spherical element, and the spherical element can synchronously rotate along with the rotor; the shell is internally provided with the stator fixed relative to the shell, the stator can be matched with the bearing assembly to convey the spherical element to the ball outlet pipe, in this way, after entering the bearing assembly, the spherical element rotates along with the rotor until the stator is matched with the bearing assembly to convey the spherical element to the ball outlet pipe, and circulating conveying of the spherical element is achieved; and the circular conveying stability of the spherical elements is improved.
Owner:CHINERGY CO LTD
Nuclear fuel cladding and method for producing such cladding
The invention relates to a nuclear fuel cladding produced with a substrate (14) which is made of pure zirconium or of a zirconium based alloy and a multilayer protective coating (16) which covers a surface (14B) of the substrate (14), the protective coating (16) comprising a main layer (18) made of pure chromium and one or more additional layers (20), each additional layer (20) being made of pure chromium or from a material made of chromium and, additionally, oxygen and / or nitrogen, with the possible presence of unavoidable impurities.
Owner:FRAMATOME SA
Lifting and transfer device for radioactive filter element
PCT designated stage expiredWO2025108045A1Nuclear energy generationReactor fuel elementsClassical mechanicsTransfer mechanism
The present application relates to a lifting and transfer device for a radioactive filter element, the lifting and transfer device comprising a grasping mechanism, a displacement mechanism and a transfer mechanism. The grasping mechanism comprises a swinging assembly, a first rotating assembly and a claw, the first rotating assembly being arranged at one end of the swinging assembly, the claw being arranged at the end of the first rotating assembly away from the swinging assembly, and the claw being configured to grasp a filter element handle. The displacement mechanism comprises a lifting component, which drives the claw to rise or lower, and on which the grasping mechanism is arranged. The displacement mechanism is arranged on the transfer mechanism, and the transfer mechanism is configured to transfer the filter element. That is, by means of the rising and lowering of the lifting assembly, the swinging of the swinging assembly and the 360° rotation of the rotating assembly, filter elements at different heights and different inclination angles and filter handles in different orientations can be grasped.
Owner:CHINA NUCLEAR POWER TECH RES INST CO LTD
Spherical fuel element lifting device and lifting method
PendingCN120636877ANuclear energy generationNuclear monitoringStructural engineeringMechanical engineering
The invention discloses a spherical fuel element lifting device and method.The spherical fuel element lifting device comprises a lifting shell, a lifting body, a detection assembly and a control assembly, and the lifting shell is provided with a spherical fuel element inlet, a spherical fuel element outlet and a spherical fuel element lifting channel; the spherical fuel element inlet and the spherical fuel element outlet are both communicated with the spherical fuel element lifting channel, and a plurality of electromagnetic coils are arranged on the lifting shell in the vertical direction of the lifting shell; the lifting body is movably arranged in a lifting channel of the spherical fuel element; the detection assembly is arranged on the lifting shell and can detect the position of the spherical fuel element; all the electromagnetic coils and the detection assembly are connected with the control assembly. According to the lifting device for the spherical fuel element, the lifting body is lifted by sequentially electrifying the electromagnetic coils, so that the spherical fuel element is driven to be lifted, the lifting efficiency of the spherical fuel element is improved, and collision between the spherical fuel element and the lifting shell can be reduced.
Owner:CHINERGY CO LTD
Process for the repeated irradiation of a radium target
ActiveUS20250266182A1Conversion outside reactor/acceleratorsNuclear energy generationRadio isotopesEngineering
Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.
Owner:TERRAPOWER ISOTOPES LLC
Blowing and sucking jam releasing device for high-temperature gas cooled reactor fuel ball conveying pipeline
PendingCN121812220ANuclear energy generationReactor fuel elementsThermodynamicsMechanical engineering
Owner:HUANENG NUCLEAR ENERGY TECH RES INST CO LTD
Gradient CrN / Cr composite coating structure for nuclear fuel cladding and preparation method
The invention mainly provides a gradient CrN / Cr composite coating structure for a nuclear fuel cladding and a preparation method. The preparation method comprises the following steps: providing a zirconium alloy substrate; forming a ZrN diffusion barrier layer on the surface of the zirconium alloy substrate; a Cr gradient transition layer is formed on the ZrN layer; and CrN / Cr multilayer structures are alternately deposited on the gradient transition layer. According to the technical scheme, Zr-Cr mutual diffusion can be effectively inhibited, generation of a brittle ZrCr2 phase at a substrate / coating interface is avoided, meanwhile, the CrN transition layer further strengthens the diffusion barrier effect through chemical stability, and the structural integrity of the coating in a high-temperature environment is improved; the design of periodically introducing the CrN layer is adopted, and a new interface is introduced to break continuous growth of thick columnar crystals of the Cr layer, so that a compact structure is formed; and the wear resistance and the cracking resistance of the coating are improved through the high hardness characteristic of CrN. Meanwhile, the high-power pulse magnetron sputtering technology is combined with automatic control, so that the thickness and the structure of the coating can be accurately regulated and controlled, and the process repeatability is ensured.
Owner:HARBIN INSTITUTE OF TECHNOLOGY (SHENZHEN) (INSTITUTE OF SCIENCE AND TECHNOLOGY INNOVATION HARBIN INSTITUTE OF TECHNOLOGY SHENZHEN)
High-temperature gas cooled reactor fuel loading and unloading system, dredging method and high-temperature gas cooled reactor
The invention relates to the field of nuclear reactors, and discloses a high-temperature gas cooled reactor fuel loading and unloading system, a dredging method and a high-temperature gas cooled reactor. The high-temperature gas cooled reactor fuel loading and unloading system comprises a reactor core falling ball returning pipe section, and the reactor core falling ball returning pipe section conveys circulating fuel elements back to a reactor core. The blockage dredging device is installed on a reactor core falling ball returning pipe section and comprises a telescopic assembly, the telescopic assembly is arranged at the end, away from a reactor core, of the reactor core falling ball returning pipe section and comprises a fixed end and a driving end capable of stretching out and drawing back relative to the fixed end, the fixed end is fixedly arranged, and the driving end faces the reactor core falling ball returning pipe section. The telescopic path of the driving end is arranged in the axial direction of the reactor core falling ball pipe section, and the driving end can enter and exit from the reactor core falling ball pipe section and drive the spherical element. The blocked spherical element is driven through the telescopic assembly, compared with a traditional pneumatic dredging mode, the blocked spherical element can be treated more directly and effectively, the dredging efficiency is greatly improved, dredging time is shortened, and the reliability is also remarkably improved.
Owner:HUANENG SHANDONG SHIDAOBAY NUCLEAR POWER CO LTD
Impact jam releasing device for fuel ball conveying pipeline of high-temperature gas cooled reactor
PendingCN121812221AEasy to dredgeGuarantee continuous and stable operationNuclear energy generationReactor fuel elementsProcess engineeringImpact
The embodiment of the invention provides an impact jam releasing device for a fuel ball conveying pipeline of a high-temperature gas cooled reactor, and relates to the field of high-temperature gas cooled reactors. The impact jam releasing device comprises an impact block, wherein the impact block is provided with a first end and a second end which are opposite in the axial direction; the moving part is connected with the first end of the impact block and used for driving the impact block to move along the ball path pipeline; the impact head is movably arranged in the impact block; and the impact driving piece is arranged in the impact block, is connected with the impact head and is used for driving the impact head to move in the axial direction of the impact block so as to drive the impact head to extend out of the second end or retract into the impact block. The impact jam releasing device provided by the invention can solve the problem that the blocked fuel balls in the ball path pipeline are inconvenient to dredge at present, and ensures the continuous and stable operation of the high-temperature gas cooled reactor.
Owner:HUANENG NUCLEAR ENERGY TECH RES INST CO LTD
Refueling water tank hydrogen removal system and method
The invention provides a hydrogen removal system and method for a refueling water tank, the system comprises a pressure stabilizer, the top end of the pressure stabilizer is connected with the refueling water tank through a first connecting pipeline, one end of the first connecting pipeline is inserted below the liquid level of the refueling water tank, and a pressure release valve is arranged on the first connecting pipeline; the first connecting pipeline is connected to a downstream treatment system through a second connecting pipeline at the upstream of the pressure release valve, and an exhaust valve is arranged on the second connecting pipeline; the refueling water tank is further provided with a nitrogen filling pipeline, the nitrogen filling pipeline comprises a main pipeline and a branch pipeline, one end of the main pipeline is connected to the downstream treatment system, one end of the branch pipeline is connected with the main pipeline, and the other end of the branch pipeline is inserted into the gas space of the refueling water tank; wherein a vacuum ejector is arranged on the main pipeline, an exhaust pipeline is further arranged on the refueling water tank, one end of the exhaust pipeline is communicated with the gas space of the refueling water tank, and the other end of the exhaust pipeline is connected to the vacuum ejector.
Owner:SHANGHAI NUCLEAR ENGINEERING RESEARCH & DESIGN INSTITUTE CO LTD
Heat dissipation adjusting device of high-voltage transformer substation
ActiveCN120784756ATransformers/inductances coolingTransformers/inductances casingsCold airTransformer
The invention relates to the technical field of high-voltage transformer substations, and discloses a heat dissipation adjusting device of a high-voltage transformer substation, which comprises a high-voltage transformer substation assembly, the high-voltage transformer substation assembly further comprises a high-voltage transformer substation shell, and the side surface of the high-voltage transformer substation shell is fixedly connected with an oil outlet pipe; the side face of the high-voltage transformer substation shell is fixedly provided with an external ventilation opening, the inner wall of the high-voltage transformer substation shell is fixedly connected with a fan assembly, the fan assembly comprises a motor, a chain, a fan and a gear, and the external ventilation opening can generate a large amount of heat during operation of high-voltage transformer substation equipment and discharges the heat through air convection. And the fan assembly fan actively sucks cold air, exhausts hot air, strengthens air convection around the equipment and rapidly takes away heat, so that the phenomena that insulation aging is accelerated, the conductivity is reduced and even faults are caused due to the fact that the temperature of the equipment is too high are avoided.
Owner:JIANGSU LONGYUAN JINGHUI ELECTRIC TECH CO LTD
Intelligent composite cladding pipe with real-time monitoring and early warning regulation and control functions
The invention relates to an intelligent composite cladding tube with real-time monitoring and early warning regulation and control functions, which comprises a compact metal layer and a SiCf / SiC composite material layer which are used as an inner layer or an outer layer of the composite cladding tube, and a flexible transition layer positioned between the compact metal layer and the SiCf / SiC composite material layer, the sensing element is a silicon dioxide optical fiber or a molybdenum-rhenium alloy thermocouple and is used for monitoring the temperature change of the reactor core, so that the flexible transition layer has a temperature sensing function. According to the structure, the phenomena of local overheating and stress concentration can be effectively inhibited, so that the operation safety and the long-term service stability of a reactor system are remarkably improved.
Owner:NORTHWESTERN POLYTECHNICAL UNIV
Leakage treatment device for low-radioactivity liquid effluent of drainage pipeline of nuclear power plant
The invention discloses a nuclear power plant drainage pipeline low-radioactivity liquid effluent leakage treatment device which comprises an anti-seepage mechanism, a water collecting mechanism, a leakage monitoring mechanism and a drainage mechanism. According to the leakage treatment device for the low-radioactivity liquid effluent of the nuclear power plant drainage pipeline, the liquid effluent in the foundation pit range can be prevented from permeating into the surrounding environment by arranging the anti-seepage mechanism, and the liquid effluent flowing out of the nuclear power plant drainage pipeline can be collected by arranging the water collecting mechanism; according to the low-radioactivity liquid effluent leakage treatment device for the drainage pipeline of the nuclear power plant, the condition of leaked liquid effluent can be monitored by arranging the leakage monitoring mechanism, the liquid effluent in the water collecting mechanism can be discharged by arranging the drainage mechanism, and the leakage of the liquid effluent can be effectively controlled by the low-radioactivity liquid effluent leakage treatment device for the drainage pipeline of the nuclear power plant; the purposes of automatic seepage prevention, automatic collection and convenient maintenance are achieved, and the influence risk on the environment is reduced.
Owner:CHINA NUCLEAR POWER ENGINEERING COMPANY LTD +1