663results about "Reactor fuel elements" patented technology

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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 relates to the field of nuclear power operation, in particular to a control method for quantitative ammonia potassium saturation of cation resin of a primary loop purification system of a nuclear power plant. The method comprises the following steps: before starting ammonia-potassium saturation, calculating the total amount of ammonia absorbed by resin, the total amount of potassium absorbed by resin and the ideal concentration of a potassium hydroxide solution by utilizing the total exchange amount of nuclear-grade cation resins in a series I and a series II and the ammonia and potassium selectivity coefficients of the nuclear-grade cation resins; the flow of the first series and the flow of the second series are adjusted, the theoretical ammonia adding amount of the first loop under different flows of the second series is calculated, and continuous ammonia adding is conducted on the first loop with the theoretical ammonia adding amount as the target; when the series of diamino potassium is saturated, sampling and analyzing the total alkali metal content and the ammonia content from the outlets of the primary loop and the series II, and calculating and adjusting the adding amount and the frequency of the potassium hydroxide solution to ensure that the total alkali metal concentration of the main loop is in the area A and does not deviate; after the potassium ammonia saturation of the series II is finished, stopping the operation of the series I, calculating the required ammonia adding amount under the normal working condition, timely adjusting the continuous ammonia adding amount of the loop I, quitting the series I, and keeping the operation of the series II. According to the invention, smooth implementation of saturation of series of diamino potassium and safety of primary loop water quality are ensured.

To facilitate separation of fuel debris from a cut-off material and reduce the amount of waste without increasing the amount of cut-off materials.SOLUTION: There is provided a leakage part blocking method for a reactor pressure vessel which blocks a leakage point (opening) at the bottom of the reactor pressure vessel by using a cut-off material. The method uses a camera mounted on a remote device to confirm the shape of the leakage point from the outside of the reactor pressure vessel, when the shape of the leakage point is known, operates the remote device to insert and fix a plug, patch, or mortar being the cut-off material from the outside of the reactor pressure vessel to the leakage point to block the leakage point, when the shape of the leakage point is unknown but the position of the leakage point can be identified, operates the remote device to insert a bag containing fluid to be solidified with accumulation and expansion being the cut-off material from the outside of the reactor pressure vessel to the leakage point, and then, blocks the leakage point with the bag in which the fluid is accumulated and expanded.SELECTED DRAWING: Figure 2

The present application relates to a kind of core high probe extraction mechanism, which provides effective radiation protection during the removal of the probe by introducing a shielding cylinder, significantly reduces the radiation dose that can be received by the operator when working, improves the safety of the working environment.Secondly, the present application also configures a positioning connection device with a positioning camera, which can monitor and adjust the position of the lifting moving assembly in real time, ensure its accurate alignment with the probe to be removed, and avoid positional interference between the shielding cylinder and other structural members in the core during the gradual deepening of the shielding cylinder into the core pool, thereby solving the problem of difficult accurate docking in traditional methods, improving the probe removal efficiency and operation safety.

The application discloses a nuclear power plant control rod guide tube, which comprises a protective cover and a full-range guide assembly installed in the protective cover; the full-range guide assembly comprises at least one full-range continuous guide element, and a full-range continuous guide channel is arranged in the full-range continuous guide element; the full-range continuous guide channel continuously guides at least one control rod in a single control rod assembly in a full range of the guide tube stroke. Compared with the prior art, the nuclear power plant control rod guide tube changes the non-continuous guide segment structure of the interval arranged guide panel into the full-range continuous guide segment formed by the long double-pipe structure, optimizes the control rod stroke protection function, reduces the risk of the control rod bending deformation caused by the transverse impact of the upper chamber fluid on the control rod, and guarantees the integrity of the control rod structure; meanwhile, the wear resistance of the specific position of the control rod guide tube is improved, so that the service life of the control rod guide tube is greatly prolonged, and the later operation cost is reduced.

The invention belongs to the field of manufacturing of nuclear fuel elements for nuclear power stations, and particularly relates to a method for adjusting the rod position of a nuclear fuel assembly. Comprising the following conditions: 1, the condition that an upper end plug of a fuel rod penetrates through the eleventh layer of grillwork of the fuel assembly; 2, the lower end plug of the fuel rod does not penetrate through the first layer of grillwork of the fuel assembly; 3, the lower end plug of the fuel rod is close to the end face of the lower tube seat and cannot be adjusted by a pull claw. The nuclear fuel assembly rod position adjusting tool and method have the beneficial effects that the nuclear fuel assembly rod position adjusting tool and method are applied to manufacturing of pressurized water reactor nuclear fuel assemblies. According to the method, the position of the fuel rod in the rod pulling process can be effectively adjusted, and the technical requirements of nuclear fuel assembly products are met. The adjusted rod position of the nuclear fuel assembly meets the expectation.

The utility model relates to a shielding device and recovery system for a nuclear reactor detector assembly, and the shielding device comprises a shielding pipe body which is provided with a pipe cavity, and the pipe cavity is used for accommodating a high-level radioactive detector; the driving assembly is arranged outside the shielding pipe body; the pulling assembly is arranged in the tube cavity, one end of the pulling assembly is connected with the driving assembly through a traction strip, and the other end of the pulling assembly is used for being connected with a high-level radiometric detector; the pulling assembly is driven by the driving assembly to move in the axial direction of the shielding tube body so that the high-level detector can be contained in the tube cavity or the high-level detector can be discharged out of the tube cavity. The structure design is ingenious, safe transfer of the high-level radioactive detector is achieved, radiation can be effectively shielded through the shielding pipe body, and operators and the surrounding environment are protected.

Systems and methods use inserted dividers with control rod drives (CRDs) to flush CRDs from inside a reactor. Dividers separate flush zones from the reactor. Pumps, injectors, gravity, or any other drive, potentially in combination with draining pumps, vacuums, or other suctioning device, flow the fluid through the divider into the zone and back out of the same for flushing out radiation and debris or bringing about any other desired condition. A handling pole may extend several meters to allow operators to reach any CRD and place the divider and flushing components about the same. Flow paths may extend with the pole to deliver and remove the flush fluid to divider ports. Additional cameras, locators, detectors, batteries, etc. may be used to inspect and clean. The systems may clean CRDs whenever accessible, such as after a control blade or rod has been removed during a maintenance outage.

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.

Described herein is a process for recovering uranium tetrafluoride (UF4) from hexafluoride uranium (UF6), the process comprising dissolving UF6 directly in an ionic liquid and recovering UF4, which can be processed to obtain UO2(s) or metallic uranium.

The invention provides a storage rack capable of separating reactor internals, which comprises a support column, one end of the support column is connected with the bottom of a refueling pool, and the other end of the support column extends towards the pool surface of the refueling pool; wherein the supporting column is used for supporting a reactor core shroud and a reactor core lower plate which are positioned in the hanging basket barrel when the hanging basket barrel axially slides along the supporting column, so that the distance from the reactor core shroud and the reactor core lower plate to the surface of the refueling pool is shortened. According to the storage rack capable of realizing separation of the reactor internals, when materials are changed, the hanging basket barrel body, the reactor core shroud and the reactor core lower plate can be lifted away together by utilizing a loading and unloading machine and then lifted to the supporting columns, the reactor core shroud and the reactor core lower plate are jacked up by the tops of the supporting columns and cannot fall off, and the hanging basket barrel body falls off due to gravity, so that the reactor core shroud and the reactor core lower plate are lifted up and then are separated from each other. The fuel rods can be conveniently grabbed by the loading and unloading machine, and the difficulty of deep reactor core reloading is solved.

The present application relates to the technical fields of research reactor material irradiation, and specifically relates to a kind of with autonomous rotating function's pressure hole irradiation device, including remote terminal and sequentially connected driving section, protection tube and test section, driving section is set in the pressure hole mouth outside, including transmission shaft, power transmission component, servo motor and controller, transmission shaft is connected with protection tube, servo motor is installed in the side direction of transmission shaft, controller is electrically connected with servo motor, remote terminal is connected with controller, power transmission component includes transmission shaft gear and the motor gear meshed with transmission shaft gear, transmission shaft gear is coaxially keyed with transmission shaft, motor gear is coaxially keyed with servo motor output shaft;Transmission shaft and protection tube are provided with the gas outlet pipe and several gas inlet pipes communicated with test section.In the present application, the mode of remote control of servo motor is used, the irradiation device of the test piece can be realized autonomous rotation and test piece temperature control function, and the accuracy and reliability of test result are improved.

This invention provides a target loading device for producing short-half-life isotopes in a heavy water reactor, comprising a channel, an electric cylinder, and a target drum assembly. One end of the channel leads to the reactor core. The target drum assembly sequentially pushes the loaded targets to the target outlet. The main shaft of the electric cylinder is coaxially arranged with the target outlet, driving the main shaft to move towards the target outlet to push the target at the outlet into the channel. The target enters the reactor core through the channel. This invention enables the loading of short-half-life radioactive isotope targets for several days or tens of days during heavy water reactor power operation, facilitating installation and maintenance. Different types and quantities of isotope targets can be loaded into the core production channel during heavy water reactor power operation.

The invention discloses a fuel assembly assembling method and belongs to the field of nuclear power. The method comprises the following steps: immersing the zirconium alloy pipe fitting into a shellac resin coating agent dissolved by ethanol, then taking out and drying to form a protective paint film on the surface of the zirconium alloy pipe fitting; and after the fuel assembly is assembled by pulling the rod, the fuel assembly is immersed in a release agent to remove the protective paint film. The method can effectively reduce the risk of damage to the surface coating of the zirconium alloy pipe fitting in the assembling process, and improves the service life and reliability of the fuel assembly.

A cutting device and method for cutting a plurality of parallel fingers of at least one core component in a plane orthogonal to the longitudinal axis of the fingers, the fingers being adapted to be introduced into the fuel assembly of a nuclear reactor core. The cutting device includes: a first bearing plate including a plurality of abutments respectively for a finger to be cut. The cutting device also includes a rotatable cutting plate including a corresponding cutting blade for each abutment in the first bearing plate, the rotatable cutting plate being rotatable about a rotating axis between a starting position and a final position in a circumferential direction, the first bearing plate and the cutting plate being parallel to each other, the cutting blades including a plurality of groups of cutting blades. For each group of cutting blades the angle between the cutting blades and the corresponding abutments is different to an angle between the cutting blades and the corresponding abutments of any other group of cutting blades.

The present application relates to a kind of modular molten salt loading and storage system, including vertical support structure rack, normal temperature electrical equipment group and high temperature molten salt equipment group are physically separated by isolation fence and integrated in it;High temperature molten salt equipment group includes molten salt storage tank and molten salt pipeline connected with molten salt storage tank;Molten salt pipeline includes molten salt loading pipe, molten salt loading pipe has upward inclination relative to horizontal plane, to realize the self-weight backflow of molten salt;The middle section of molten salt loading pipe is connected with molten salt isolation valve, and molten salt isolation valve and normal temperature electrical equipment group are cooperated to realize the on-off control of molten salt loading pipe.The present application effectively avoids the thermal influence of high temperature environment on normal temperature electrical equipment by isolation fence, and the integrated design of overall modularization integrates the core equipment of molten salt loading and storage into one, discards the traditional dispersed equipment arrangement and installation mode, greatly simplifies the field construction process, reduces installation period, also makes the debugging, maintenance and overhaul of system more convenient.