29 results about "Heavy water reactors" patented technology

The invention provides a disassembling and assembling device for a heavy water reactor sealing plug. The disassembling and assembling device comprises a base and a disassembling and assembling mechanism. One end of the base is mounted on the truss manipulator, a rotating shaft is arranged on the base, and one end of the rotating shaft is rotationally matched with the base to drive the dismounting mechanism to grab a target sealing plug or drive the dismounting mechanism to dismount the target sealing plug; the dismounting mechanism comprises an outer cylinder, a positioning cylinder and at least two clamping mechanisms, the outer cylinder is used for axially positioning a target sealing plug, the positioning cylinder and the outer cylinder are coaxially arranged and are arranged on the inner side of the outer cylinder, and the at least two clamping mechanisms are arranged in the peripheral direction of the positioning cylinder; the clamping mechanism comprises a first driving piece and a clamping jaw rotationally matched with the first driving piece, and the clamping jaw is configured to be capable of being driven by the first driving piece to contract towards the central axis of the positioning barrel so as to release the target sealing plug or extend away from the central axis of the positioning barrel and be clamped to a clamping protrusion of an inner cavity of the target sealing plug so as to grab the target sealing plug.

This invention discloses a method for producing radioactive isotopes in the core of an advanced high-temperature reactor using a spherical target. This invention overcomes the limitations of using neutron irradiation channels or fuel-related components of nuclear reactors, such as conditioning rods in heavy water reactors and combustible poison rods in pressurized water reactors, to produce radioactive isotopes. It sets a precedent for the mass production of radioactive isotopes using the core of a nuclear reactor, and has significant strategic implications for exploring high-value applications of commercial advanced high-temperature reactors to reduce operating costs and mass-produce radioactive isotopes, including the strategic isotope tritium.

The application discloses a kind of digital fuel physics supervision method and system suitable for heavy water reactor, it is related to nuclear reactor fuel management technical field, system includes data acquisition module, document scale calculation interface, system management module and central processing module, central processing module configuration basic data, new fuel, FCO, spent fuel, dry storage area management unit.Method is standardized by data acquisition, whole-process digital control, dynamic monitoring and tracing three steps, realize fuel full life cycle closed-loop management.System adopts PDA to cooperate bar code / RFID to realize entity information acquisition and offline synchronization, constructs hierarchical correlation model, with new fuel isolation control, FCO closed-loop approval, spent fuel transfer verification, fine authority control and other functions, can improve management efficiency, reduce human error, strengthen data tracing and operation standardization, guarantee nuclear safety, provide support for management decision, suitable for heavy water reactor fuel physics supervision and whole-process management.

The invention belongs to the technical field of laser welding, and particularly relates to a super-long slender rod type zirconium 4 alloy integrated laser welding device which comprises an assembling platform, a device control table, a protective gas hood, a welding robot provided with a laser welding gun, a robot power source, a rotating tool and a main body mechanism support. And the robot power supply is connected with the device console and a welding robot equipped with a laser welding gun. The device console is connected with the welding robot and the rotating tool. The length of the workpieces is larger than or equal to 12,000 mm, the wall thickness of the workpieces is larger than or equal to 3 mm, and the workpieces are zirconium-4 alloy workpieces with different diameters. The device console is arranged to control the rotating tool to rotate to the welding position and control the welding robot provided with the laser welding gun to conduct rotating motion in the welding process so as to achieve multi-posture welding at different positions; the technical problem that a production channel assembly and a protective sleeve assembly in an existing heavy water reactor online irradiation isotope production device are difficult to assemble and weld is solved.

An apparatus and a method for separating a shielding ball are disclosed. An apparatus for separating a shielding ball comprises: an end shield which is installed in a spinning structure of a heavy water reactor facility and includes a fueling tubesheet and a calandria tubesheet facing each other and multiple shielding balls for filling a space between the calandria tubesheet and the fueling tubesheet; as a device for removing the shielding balls from a lattice tube connected to the calandria tube by being connected to the calandria tubesheet through the fueling tubesheet, a magnetic force applying unit inserted into the lattice tube in a direction of the fueling tubesheet to apply a magnetic force for fixing positions of the multiple shielding balls positioned within a first distance from the calandria tubesheet; a mover for moving the magnetic force applying unit in a direction of the lattice tube; and a storage rack for separately storing, inside the end shield, the shielding balls positioned within the first distance from the calandria tubesheet and shielding balls positioned at a distance greater than the first distance.

The invention belongs to the technical field of heavy water reactor nuclear power plant auxiliary facilities, and particularly relates to a heavy water reactor nuclear power plant auxiliary channel shielding plug dismounting device. The fixed support is hinged to the rotating arm, a rotating bolt bearing is installed in a bolt hole below the fixed support, the rotating arm rotates relative to the fixed support and is provided with a through hole, and the shielding plug connecting rod can penetrate through the through hole of the rotating arm to be connected with the shielding plug. The shielding plug tray is fixedly installed on the rotating arm, a through hole is designed in the middle of the shielding plug tray, and the lifting top block penetrates through the through hole. A threaded hole is designed in the tail end of the shielding plug tray, the centering bolt is connected with the threaded hole, the centering angle of the shielding plug tray is controlled by adjusting the exposed length of the centering bolt, and the shielding plug tray is locked and fixed through the centering locking nut. The shielding plug connecting rod is connected with the shielding plug, and when the shielding plug is horizontally located in the shielding plug tray, the shielding plug connecting rod penetrates through the through hole of the rotating arm. The dismounting and mounting efficiency of the shielding plug is improved, and the safety of personnel and equipment in the dismounting and mounting process is ensured.

The present application particularly relates to a method for producing helium-3 from heavy water in a heavy water reactor, comprising: step 1, subjecting the heavy water in a moderator system of a heavy water reactor unit to negative pressure rectification to obtain concentrated high-tritium heavy water; step 2, subjecting the concentrated high-tritium heavy water in step 1 to beta decay to produce helium-3 raw gas; and step 3, subjecting the helium-3 raw gas in step 2 to active carbon adsorption, helium-argon separation and hydrogen isotope gas adsorption to obtain helium-3. The method for producing helium-3 from heavy water in a heavy water reactor avoids technical difficulties and high costs of helium-3 / helium-4 isotope separation, and has high technical maturity and good economy.

This invention relates to the field of target box transportation technology after heavy water reactor irradiation. Addressing the problem that existing transportation channels cannot effectively reduce radiation exposure to personnel during the transportation of radioactive target boxes, this invention provides a target box transportation device after heavy water reactor irradiation, comprising a transportation pipe, a shielded transfer container, and a conveyor belt. A negative pressure is established within the transportation pipe to draw the target box into a vertical section. A separator is installed on the vertical section of the transportation pipe to allow the target boxes to enter the shielded transfer container one by one within a shielded box. The shielded transfer container is transported via the conveyor belt. Furthermore, this invention also provides a method for transporting target boxes after heavy water reactor irradiation, sequentially loading the target boxes from the vertical section into designated positions within the shielded transfer container in the shielded box, and completing radiation protection treatment of the target boxes within the shielded box. By loading the irradiated target boxes into a shielded working box, compared to containerized transport, the dose rate in the operation and maintenance area can be reduced from 2 mSv / h to 25 μV / h, significantly reducing the dose level for operators.

Thorium-based fuel bundles are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300 / 600 / 900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to 13-19.95% 235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt%. Reprocessing of spent fuel bundles is also avoided.

Disclosed is a method for dismantling a D2O tank of a heavy water reactor installed on an inner wall of a containment building and surrounded by a concrete structure, the method including cutting and opening a front surface of the concrete structure so as to dismantle the D2O tank, descending the D2O Tank by using an dedicated dismantling apparatus, and taking the D2O tank out of the containment building by using a crane, and a dismantling apparatus.

The invention particularly relates to a method for reducing hydrogen and deuterium concentration at a key position of a heavy water reactor pressure pipe, and belongs to the technical field of aging protection of heavy water reactor pressure pipe materials, the pressure pipe is made of zirconium alloy, two ends of the pressure pipe are in expanded connection with end parts, and an expanded connection pressure stress area, an expanded connection tensile stress area and a pipe body area are defined; the method comprises the steps that at least one hydrogen absorption material block is fixedly installed in the area, corresponding to an expanded connection tensile stress area, of the outer surface of the pressure pipe; the hydrogen absorption material block is made of yttrium or yttrium alloy; and the hydrogen absorption material block is used for absorbing hydrogen and deuterium diffused to the expanded connection tensile stress area from the expanded connection pressure stress area of the pressure pipe. According to the method, the hydrogen equivalent concentration of the expanded connection tensile stress area, especially the polishing mark position is obviously reduced, the hydrogen equivalent concentration is prevented from exceeding the safety limit value, and the service life of the pressure pipe is prolonged.

The invention belongs to the technical field of nuclear fuel, and particularly relates to a heavy water reactor fuel rod bundle damage positioning and searching method. Comprising the following steps of: 1, judging reactor entering time according to a change trend of Xe-133 nuclide in a coolant; 2, confirming a reactor core loop where the damaged fuel is located; 3, formulating a damaged fuel positioning system scanning strategy; 4, making and evaluating a damaged fuel reloading scheme; 5, changing and monitoring the damaged fuel; and 6, judging the position of the damaged fuel rod bundle. The method has the beneficial effects that the troubleshooting range of the damaged fuel is further narrowed through coolant nuclide analysis and damaged loop confirmation, delayed neutron scanning is started in combination with a damaged fuel positioning system scanning strategy, and compared with the prior art, the method can shorten the positioning and searching time of the damaged fuel, improve the accuracy of the positioning and searching result of the damaged fuel, and improve the accuracy of the positioning and searching result of the damaged fuel. And the influence of fuel damage is reduced.

Proposed is a cutting apparatus for a fuel basket accommodating spent nuclear fuel of a heavy water reactor, wherein the apparatus includes a base, a vertical support arranged on one side of the base; a turn table arranged to be reciprocated between one side and an opposite side of the base and be horizontally rotatable on the base; a horizontal support arranged to extend from an upper-end of the vertical support in a horizontal direction toward the turn table; a side-cutting means arranged on the vertical support protruding toward the turn table for cutting a side welding bead of the fuel basket mounted on the turn table; and an upper-cutting means arranged on the horizontal support protruding toward the turn table for cutting an upper welding bead of the fuel basket mounted on the turn table.

The application relates to the technical field of fuel rod bundle underwater detection, and discloses a heavy water reactor fuel rod bundle underwater inspection platform and a detection method. The heavy water reactor fuel rod bundle underwater inspection platform comprises a supporting mechanism, a three-dimensional moving mechanism arranged on the supporting mechanism, a rotating positioning mechanism for clamping and driving the fuel rod bundle to rotate, an element outer diameter measuring mechanism for self-adaptive centering and measuring the outer diameter of a fuel element, an element oxide film thickness measuring mechanism for self-adaptive centering and measuring the thickness of an oxide film on the surface of the fuel element, and a calibration platform arranged on the supporting mechanism and used for calibrating the element outer diameter measuring mechanism and the element oxide film thickness measuring mechanism. The three-dimensional moving mechanism drives the element outer diameter measuring mechanism and the oxide film thickness measuring mechanism to move, and the element outer diameter measuring mechanism and the oxide film thickness measuring mechanism are in contact with a measuring calibration rod on the calibration platform to perform a calibration operation and in contact with a fuel element to perform a measuring operation. The underwater accurate measurement of the outer diameter of the fuel element and the thickness of the oxide film can be realized.

The invention belongs to the technical field of nuclear fuel, and particularly relates to a heavy water reactor fuel rod bundle in-reactor damage monitoring method. Comprising the following steps: 1, carrying out online monitoring by using a fission gas monitoring system; 2, chemical sampling analysis of the coolant; and 3, performing data analysis by using a damaged fuel positioning system. The method has the beneficial effects that the method solves the problems that the monitoring means for the damage in the heavy water reactor fuel rod bundle is single, and the time for intervening in searching the damaged fuel is late, the fuel damage phenomenon can be found in the first time, and the work for searching the damaged fuel is started in the first time. The damaged fuel searching work is involved in advance, the searching time of the damaged fuel can be shortened, and the influence of coolant nuclide on personnel and the environment after damage can be reduced.

A target used for heavy water reactor production of C-14 isotopes, which comprises: a target tube, which has a target material (11) and / or an absorber material (14), the target material (11) being a nitrogen-containing solid material; end plates (2), which are arranged at two ends of the target tube; a connection rod (3), which is provided with a connection component that is connected to a positioning mechanism used for positioning the target, wherein the connection rod (3) passes through the target tube along the axial direction of the target tube, and the target tube is fixedly connected to the connection rod (3) by means of the end plates(2). The present invention can be adapted to be placed at different locations such as at a guide tube, an inspection passage, or a fuel channel to undergo irradiation and produce C-14 isotopes.

The application relates to a heavy water reactor fuel channel discharge station transverse potentiometer protection device, which comprises a protective cover (05); the protective cover (05) is a semicircular tile structure, is buckled on the upper portion of a transverse potentiometer (04) and covers the portion above the horizontal center of the shell of the transverse potentiometer (04); a shunt strip is arranged on the upper center of the protective cover (05) in an axial direction. The heavy water reactor fuel channel discharge station transverse potentiometer protection device can prevent the leakage of heavy water and hydraulic oil from entering the transverse potentiometer during the operation of a loading and unloading machine, ensures that the transverse potentiometer can normally operate, prolongs the service life of the transverse potentiometer and saves the cost of a company.

The present invention relates to a method for dynamically controlling corrosion in a steam-water circuit of a nuclear power plant, the method comprising: injecting a compound ethanolamine and optionally ammonia, the system comprising: an injection unit located at a feed point and configured for injecting the compound ethanolamine and ammonia in a fluid circulating in the circuit; and a measurement unit configured for measuring in the loop a parameter pH value and at least one parameter selected from the group consisting of an oxidation-reduction potential and a corrosion potential; the system is characterized in that it further comprises a control unit comprising: a first acquisition module connected to the measurement unit, configured for acquiring a measured value of the parameter, and configured for acquiring at least one setpoint value of said parameter; a command module, connected to the injection unit, configured for commanding the injection of the compounds ethanolamine and ammonia through the injection unit according to at least one setpoint value and a measured value of the acquired parameter, characterized in that ethanolamine is used as a reducing agent. The invention also relates to a system for dynamically controlling corrosion in a steam-water circuit of a nuclear power plant.

The application relates to the technical field of fuel rod bundle underwater size measurement, and provides a heavy water reactor fuel rod bundle underwater size measurement system and a measurement method. The heavy water reactor fuel rod bundle underwater size measurement system comprises a support platform, a rotating positioning mechanism, a three-dimensional moving mechanism, a first underwater measurement camera used for collecting overall images of the fuel rod bundle and images of single fuel elements, a camera measurement calibration platform comprising a rod bundle measurement calibration plate and an element measurement calibration plate, a first mirror assembly used for reflecting the images of the first end plate to the first underwater measurement camera, a camera movement module comprising two module guide rails and a camera support capable of sliding along the module guide rails, a second underwater measurement camera used for collecting images of a second end plate, and a second mirror assembly used for reflecting the images of the second end plate to the second underwater measurement camera. The moving platform and the underwater camera can be used for completing fuel rod bundle image collection and splicing, and the detection speed is high and the measurement precision is high.

The present application belongs to the technical field of heavy water reactor nuclear power plant auxiliary facilities, and particularly relates to a heavy water reactor nuclear power plant auxiliary channel shielding plug dismounting and mounting device. A fixed support is hinged to a rotating arm, a rotating bolt bearing is installed in a bolt hole below the fixed support, the rotating arm rotates relative to the fixed support, the rotating arm is designed with a through hole, a shielding plug connecting rod can pass through the through hole of the rotating arm and be connected with the shielding plug; a shielding plug tray is installed and fixed on the rotating arm, a through hole is designed in the middle of the shielding plug tray, and a lifting top block passes through the through hole; a screw hole is designed at the end of the shielding plug tray, a centering bolt is connected with the screw hole, the centering angle of the shielding plug tray is controlled by adjusting the exposed length of the centering bolt, and the centering bolt is locked and fixed by a centering locking nut; the shielding plug connecting rod is connected with the shielding plug, and the shielding plug connecting rod passes through the through hole of the rotating arm when the shielding plug is horizontally located in the shielding plug tray. The present application improves the dismounting and mounting efficiency of the shielding plug and ensures the safety of personnel and equipment during the dismounting and mounting process.

The present invention provides thorium-based fuels for pressurized heavy water reactors (e.g., fuel pellets, fuel pins, and / or fuel bundles), as well as methods for manufacturing and using the same. [Solution] The fuel composition of the bundle consists of 60 wt% or more thorium, with the remaining fuel being low-enriched uranium (LEU) of 235U with an enrichment level of 13-19.95%. The use of such a thorium-based fuel bundle provides a non-brederate spent fuel bundle having (1) 100% of the nominal power throughout the entire lifecycle of the core, (2) high burnup, and (3) a total isotope uranium concentration of less than 12 wt%. Reprocessing of spent fuel can also be avoided.

The present application relates to a kind of methods for controlling the concentration of deuterium gas in heavy water reactor under severe accident, by arranging different types of deuterium gas elimination device in different compartment region of heavy water reactor containment, so that the deuterium gas concentration in different compartment region of containment is reduced, and the goal of eliminating the risk of deuterium gas in heavy water reactor containment under severe accident is achieved. The deuterium gas concentration in heavy water reactor containment under severe accident is controlled by using ignition device and catalytic composite device jointly. For the compartment region where the main heat transport system break, the discharge pipe container burst disc and the fluid release valve are located, the ignition device is arranged to rapidly reduce the deuterium gas concentration in the compartment; for other compartment region, catalytic composite device is arranged to eliminate the deuterium gas in other compartment region. The deuterium gas elimination technology of the present application can effectively control the deuterium gas concentration in heavy water reactor containment, thereby realizing the elimination of the risk of deuterium gas in heavy water reactor containment.

The invention belongs to the technical field of heavy water reactor nuclear power plant fuel operation system maintenance, and particularly relates to a heavy water process controller battery replacement method, which comprises the following steps of: 1, data backup: performing data backup on a process controller of which the battery is to be replaced; step 2, program backup: performing program backup on the process controller of which the battery is to be replaced; step 3, battery replacement: disassembling the controller mainboard and replacing the battery under the condition of keeping power supply; and 4, data checking: checking data and programs of the process controller after battery replacement. The method is high in reliability, high in operability and small in introduction risk, the reliability, the working efficiency and the equipment reliability of the method are greatly improved through field long-term actual application, and the power station management requirement is met.

The invention provides a dismantling device for a heavy water reactor sealing structure, which comprises a base and a dismantling mechanism, and one end of the base is provided with a mounting interface connected with a truss manipulator; the dismantling mechanism comprises a sleeve, a first driving part, a shearing part and a clamping assembly, the shearing part is clamped on the inner side of the sleeve, and one end of the first driving part is connected with the shearing part and used for driving the shearing part to move in the axial direction of the sleeve; the clamping assembly comprises a second driving piece, a supporting column, a driving rod and a limiting block. The supporting column is arranged on the inner side of the sleeve, and an annular cavity for containing a target pipe fitting is formed between the supporting column and the shearing piece. The supporting column is provided with a containing hole channel in the axial direction, the driving rod is arranged in the mounting hole, the supporting column is provided with at least one limiting hole in the circumferential direction, and the limiting block is elastically clamped in the limiting hole; one end of the second driving piece is connected with the driving rod to drive the driving rod to move in the axial direction of the mounting hole and can drive the limiting block to at least partially stretch out of the limiting hole to abut against the pipe wall of the target pipe so that the shearing piece can cut off the connector assembly on the target pipe.

The invention belongs to the technical field of heavy water reactor nuclear power plant fuel operation system maintenance, and particularly relates to a control cabinet for a separator of a loading and unloading machine. Comprising a control cabinet and a cabinet door hinged to the control cabinet, a power supply mechanism is arranged at one end of the control cabinet, a partition plate is arranged in the middle of the inner wall of the control cabinet, a first control cavity and a second control cavity are formed in the control cabinet through the partition plate, and an adjustable power source and a central controller are arranged in the first control cavity; a vacuumizing device and an air supply device are arranged in the second control cavity, an air cooler is arranged at the top of the control cabinet, a damping protection mechanism is arranged between the air cooler and the control cabinet, the output end of the air cooler is communicated with an air conveying hose, and one end of the air conveying hose is communicated with a cold air distribution mechanism used for controlling energy consumption. The cooling device has the beneficial effects that the control cavity with the high temperature can be cooled more quickly, meanwhile, waste of cold air can be reduced, and energy consumption is reduced.

The invention provides a heavy water reactor water supply pipe dismantling method and device, and the dismantling method comprises the steps: 1, carrying out the partition division of a fuel channel according to the connection layout of a water supply pipe and the fuel channel in a heavy water reactor plant, the plant item distribution and the transfer space condition; 2, aiming at the water supply pipes connected at different subareas of the fuel channel, determining the number and specific positions of shear points of the water supply pipes in each subarea in combination with interference limitation, broken pipe transfer size and bearing requirements of non-removable items in a plant; and step 3, shearing the water supply pipe at the shearing point by using a shearing tool, and synchronously taking down a broken pipe formed by shearing by using a clamping tool and placing the broken pipe in a collecting box.

The present disclosure belongs to the technical field of nuclear power, and particularly relates to a heavy water reactor nuclear power plant atmospheric release valve grinding device and method. The heavy water reactor nuclear power plant atmospheric release valve sealing surface grinding device provided by the present disclosure effectively clamps the part to be ground through a positioning module, designs and manufactures a grinding mold matched with the shape and size of the part to be ground, ensures that the grinding is in place, and guarantees the maintenance quality; the rotation driving and up-down driving of the grinding mold are realized through a grinding module, so that the grinding intensity can be adjusted at any time during the grinding of the part, the grinding surface can not only be completely attached to the part to be ground, but also can provide effective grinding pressure during the grinding process; thus, safe and efficient grinding treatment of the heavy water reactor nuclear power plant atmospheric release valve sealing surface is realized.

The invention belongs to the technical field of heavy water reactor nuclear power plant auxiliary facilities, and particularly relates to a heavy water reactor nuclear power plant auxiliary facility sleeve dismounting device. Two turning cover lugs are arranged on one side of the upper part of the main frame, and a turning cover is mounted on the turning cover lugs; the linear sliding tables are installed on short square steel transversely arranged in the middle of the main frame in parallel, each linear sliding table is provided with two sliding table sliding blocks, and the sliding table sliding blocks are connected with the bottom of the lifting bottom plate so that the lifting bottom plate can slide on the linear sliding tables. A transverse moving bearing seat cushion block is welded to rectangular steel in the middle of a main frame, a transverse moving bearing seat is fixed to the transverse moving bearing seat cushion block, a transverse moving ruler is welded to the upper surface of the transverse moving bearing seat, one end of a transverse moving lead screw penetrates through a bearing in the transverse moving bearing seat to be connected with a transverse moving hand wheel, and the other end of the transverse moving lead screw penetrates through a transverse moving mounting hole of a lifting bottom plate to be connected with a transverse moving nut in a matched mode. The transverse moving nut is fixedly installed on the lifting bottom plate, and when the transverse moving hand wheel is rotated, the transverse moving nut drives the lifting bottom plate to move front and back. The sleeve dismounting efficiency is improved, and the safety of personnel and equipment is ensured.

The invention belongs to the technical field of heavy water reactor spent fuel pools, aims to solve the problem of water quality of spent pools, and discloses an underwater dust suction device and method for a heavy water reactor nuclear power plant spent fuel pool. The device comprises a driving motor, a driving screw rod, a dirt suction head, a water inlet pipe, a bottom plate, a filter, a pump front pipe, a water outlet pipe and a water suction pump; the filter is connected with the dirt suction head through a water inlet pipe, the dirt suction head is connected with the long push rod, the filter is connected with the water suction pump through a pump front pipe, and the water suction pump is connected with a water outlet pipe; according to the method, underwater dust collection of the spent fuel pool is implemented by using the device. Particles suspended in the middle of the pool can be absorbed, the manual cleaning difficulty is reduced, and dust particles in the pool can be absorbed and concentrated, so that the water quality is improved, and the critical safety is ensured.