Abstract

A high-temperature dense composite nuclear fuel material and a method of its production. The high-temperature dense composite nuclear fuel material comprises a radiation inert ceramic matrix in which nuclear fuel particles are distributed. The matrix is made of a powder of a silicon carbide based material. The nuclear fuel particles are particles of an oxygen-free nuclear fuel. The method of producing the high-temperature dense composite nuclear fuel material includes preparation of a mixture of nuclear fuel particles and a powder of the radiation inert ceramic matrix, molding of the mixture by pressing, and sintering of the molded mixture. The molded mixture is sintered by hot pressing. The group of inventions makes it possible to produce the high-temperature dense composite nuclear fuel material which, after annealing in vacuum at 0.63Tmelt for 10 hours, exhibits a loss in mass of not more than 2%.

IPC Classes  ?

• G21C 3/64 - Ceramic dispersion fuel, e.g. cermet

Abstract

A device for decomposing ammonia into a nitrogen-hydrogen mixture. The device comprising a cylindrical combustion chamber, a device for supplying air-ammonia mixture, at least one main channel for its supply, a tangential swirler, a spark plug installed in the combustion chamber, and a channel for supplying auxiliary fuel with increased flammability. A cylindrical body is introduced, inside of which a cylindrical combustion chamber is located coaxially with the formation between them, the entrance of the main channel and the tangential swirler, an additional annular channel for supplying and heating the air-ammonia mixture, wherein the cylindrical side surfaces of the body and the combustion chamber have at least one through hole for installation of the spark plug, a flame sensor, a device for secondary air supply and a flame stabilizer are introduced, wherein the combustion chamber is a cooling chamber, at the outlet of which a catalytic unit is installed.

IPC Classes  ?

• F23C 1/00 - Combustion apparatus specially adapted for combustion of two or more kinds of fuel simultaneously or alternately, at least one kind of fuel being either a fluid fuel or a solid fuel suspended in air
• F23C 7/00 - Combustion apparatus characterised by arrangements for air supply
• F23C 13/08 - Apparatus in which combustion takes place in the presence of catalytic material characterised by the catalytic material

Abstract

A device for measuring bending of extended vertically directed channels, including nuclear reactor fuel channels of the high-power channel-type reactor (HPCR). The device comprises a flexible hollow carrier rod having at least one fibre-optic sensor equipped with a core closed by a sealed tubular casing. The upper part of the core comprises a mounting sleeve with a central cylindrical cavity. The lower part of the core is comprises a central cylindrical rod. Longitudinal sector cuts are made on the rod of the core, in which ferrules with longitudinal through channels are fixed, and a gravity pendulum is placed under the lower end of the rod. Through the through channels of the ferrules, fiber-optic lines are drawn, connecting each fibre-optic sensor with a tunable laser and a photoreceiver connected to a computer. The cavity of the sealed tubular casing of the fibre-optic sensor is filled with an inert gas.

IPC Classes  ?

• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations

Abstract

Method of measuring bending of a nuclear reactor fuel channel. A fiber-optic sensor comprises a gravity pendulum that is at the lower end of the fibre-optic sensor; a flexible hollow carrier rod with the fibre-optic sensor is passed along the central tube of a fuel assembly and detecting a gas gap with a photoreceiver, said gas gap varying during the passage of the fibre-optic sensor as a result of the angular motion of the gravity pendulum away from the axis of the bowed central tube of the fuel assembly. Profilograms of the variations of the gas gap for each fibre-optic line of each fibre-optic sensor are recorded; and the magnitude and direction of bending of the central tube of the fuel assembly from the vertical axis are calculated, according to which the presence and magnitude of bending of the nuclear reactor fuel channel are determined.

IPC Classes  ?

• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations

Abstract

A method for measuring bending of extended vertically directed channels. A fibre-optic sensor having a gravity pendulum fixed at the end of a flexible hollow carrier rod is placed inside the extended vertically directed channel. A light signal is supplied via fibre-optic lines connected to the sensor to record light signals. The flexible hollow carrier rod with the fibre-optic sensor is placed along the channel and detects interference in a gas gap with a photoreceiver, said gas gap varying during the passage of the fibre-optic sensor as a result of the angular motion of the gravity pendulum away from the axis of the bowed channel. Profilograms of the variations of the gas gap for each fibre-optic line are recorded; and the magnitude and direction of bending of the channel from the vertical axis are calculated to simplify measurements of bending of a vertically directed channel while maintaining measurement accuracy.

IPC Classes  ?

• G01B 11/24 - Measuring arrangements characterised by the use of optical techniques for measuring contours or curvatures
• G01B 9/02015 - Interferometers characterised by the beam path configuration

Abstract

A robotic monitoring, detection and fire extinction control system comprising a pumping station, a fire main and at least two robotized fire extinguishing units connected to the fire main comprising a fire monitor with drives for vertical and horizontal aiming and a head with a drive for adjusting the spray angle of the jet. A butterfly valve with a drive is mounted to an inlet of the fire monitor, and a pressure sensor and a portable control console are mounted to an outlet of the fire monitor. The pressure sensor and portable control console are connected to a switching unit at the inlet and, via a network controller, to a control device at the outlet. A fire detection and video surveillance device is connected via a receiving and control device to fire alarms. A foam generating unit is connected to the robotized fire extinguishing units by a foam conduit.

IPC Classes  ?

• A62C 3/02 - Fire prevention, containment or extinguishing specially adapted for particular objects or places for area conflagrations, e.g. forest fires, subterranean fires
• A62C 5/02 - Making of fire-extinguishing materials immediately before use of foam
• A62C 27/00 - Fire-fighting land vehicles
• A62C 31/12 - Nozzles specially adapted for fire-extinguishing for delivering foam or atomized foam
• A62C 35/68 - Details, e.g. of pipes or valve systems
• A62C 37/38 - Control of fire-fighting equipment an actuating signal being generated by a sensor separate from an outlet device by both sensor and actuator, e.g. valve, being in the danger zone

Abstract

Methods for cleaning a NPP steam generator tube bundle. The methods characterized in that a hydrodynamic manipulator connected to a hoist is inserted into a vertical corridor inside a heat exchanger. The methods include taking preliminary measurements of the dose rate inside the steam generator, measuring eddy current signals and analyzing the signals received to assess the condition of the walls of heat exchange tubes and deposit thickness. The methods further include determining the coordinates of a cleaning sector and calculating cleaning time and modes, ejecting a high-pressure jet of water using at least one nozzle configured to rotate around the swiveling axis, and moving the hydrodynamic manipulator according to the coordinates of the cleaning sector with the water jet set at a feed pressure of 1000 to 1500 bar and a flow rate of 100 to 150 l/min and a simultaneous video monitoring of the cleaning zone.

IPC Classes  ?

• F28G 1/16 - Non-rotary, e.g. reciprocated, appliances using jets of fluid for removing debris

Abstract

Systems and arrangements for detecting leans and repairing damaged lining of NPP spent fuel pools filled with water. The arrangements and systems include devices that are not submerged to the spent fuel pool, including welding equipment, and devices submerged to the spent fuel pool, including a submersible repair platform for the working mechanism installation. The working mechanism includes a device for cleaning of weld seams and the surface of the spent fuel pool, and a wire cutting device, comprising a container for collection of cuttings and a pump for particulate removal. The leak sealing device is arranged in the form of a small-scale welding assembly connected to the welding equipment, and the submersible repair platform is equipped with means for its attachment to the surface of the spent fuel pool. The arrangement for repairing the lining of a spent fuel pool includes a submersible repair robot having a working unit.

IPC Classes  ?

• B23K 9/00 - Arc welding or cutting
• B23K 37/02 - Carriages for supporting the welding or cutting element

Abstract

The invention relates to cleaning of tube bundle of NPP steam generator. The device for cleaning heat exchange tubes of NPP steam generator comprising a manipulator, which is installed to be moved and fixed in a vertical corridor between bundles of heat exchange tubes; a nozzle that is installed on the manipulator ad is in the form of a head of injectors connected to a pipeline for supplying high-pressure water jets; and means for remote control and video monitoring, wherein the manipulator is equipped with a body, which is installed on a guiding mounting frame and is connected to a base; a drive for main rotational movement located on the body; a post which is connected to the body and is in the form of a hollow column; a drive for auxiliary movement which is fixed on the post, an elbow which is connected to the post and is in the form of a tube having a gear rack, and an elbow of actuating units and a coordinate head. The technical result is a shorter cleaning time and more reliable removal of deposits.

IPC Classes  ?

• F28G 3/16 - Rotary appliances using jets of fluid for removing debris
• F28G 15/00 - CLEANING OF INTERNAL OR EXTERNAL SURFACES OF HEAT-EXCHANGE OR HEAT-TRANSFER CONDUITS, e.g. WATER TUBES OF BOILERS Details

Abstract

Methods for diagnostics of electrically driven equipment and monitoring of vibrations in rotating equipment at nuclear power plants. The methods include measuring and recording of diagnostic signals, dividing the recordings of a reference signal and the signal from the equipment under inspection into segments, converting each segment of the recordings of the reference and inspection signal into a spectrum which represents the distribution of amplitudes across frequencies, sampling amplitudes of the diagnostic signal spectra from the inspected and fault-free equipment at the frequencies where deviations appear between the inspection and reference signals, calculating absolute amplitude difference in the spectra, ranking of the absolute differences and summarizing the ranks. Comparing summarized ranks with the critical value to establish whether the rotating equipment is in excess of the prescribed vibration values and to establish the presence of fault if the sum of the resultant ranks exceeds the critical value.

IPC Classes  ?

• G01H 1/00 - Measuring vibrations in solids by using direct conduction to the detector
• G01M 13/045 - Acoustic or vibration analysis

Abstract

A method for measuring vibration acceleration in three mutually orthogonal planes via vibration sensors. Vibration accelerations are measured at the monitoring points on an operational and knowingly fault-free diesel generator, and then vibration accelerations are measured at the monitoring points of a diesel generator in the course of its operation with prescribed intervals. Temperature and intensity of an ultrasonic signal are additionally measured at the same monitoring points, and the root mean square values of the ultrasonic signal intensity, temperature and vibration acceleration are determined. The measured vibration acceleration values are used to calculate the root mean square values of vibration velocity and vibratory displacement, the distances between clusters of preceding measurements are determined, and a conclusion on the operating state of the diesel generator is made. The method promotes timely detection of any operating defects through periodic measurements and comparison of calculations.

IPC Classes  ?

• G01M 15/12 - Testing internal-combustion engines by monitoring vibrations

Abstract

Methods for technical diagnostics and inspection of pipelines to detect leaks in nuclear power plants pipelines. Acoustic signals are recorded within a broad ultrasonic range at two points along the pipeline length located on the pipeline downstream and upstream of a valve. The recorded ultrasonic signals are then processed and used to construct two signal spectra. Then the range of 15,000 to 90,000 Hz is isolated, and the highest amplitude value in both signal spectra is selected within this range. Then the amplitudes of the signal spectra within the said frequency range are divided by the highest amplitude value, and the difference between the signal spectra upstream and downstream of the valve is determined using a mathematical formula, thereby determining the degree of the pipeline leak-tightness.

IPC Classes  ?

• F17D 5/06 - Preventing, monitoring, or locating loss using electric or acoustic means
• G01M 3/24 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using infrasonic, sonic, or ultrasonic vibrations

IPC Classes  ?

• F23C 1/00 - Combustion apparatus specially adapted for combustion of two or more kinds of fuel simultaneously or alternately, at least one kind of fuel being either a fluid fuel or a solid fuel suspended in air
• F23C 13/08 - Apparatus in which combustion takes place in the presence of catalytic material characterised by the catalytic material
• F23L 9/00 - Passages or apertures for delivering secondary air for completing combustion of fuel
• F23N 5/10 - Systems for controlling combustion using devices responsive to thermal changes or to thermal expansion of a medium using thermocouples

Abstract

The invention relates to the field of chemical engineering, and more particularly to a device for combusting ammonia, which can be used in heating devices for producing heat and in devices for decomposing ammonia into a nitrogen-hydrogen mixture. A device for combusting ammonia comprises a cylindrical combustion chamber, a device for supplying an air-ammonia mixture, which includes at least one main supply channel having a tangential vortex generator mounted at the end thereof, a spark igniter mounted in the combustion chamber, and a channel for suppling a support fuel that has heightened combustibility. Further provided is a cylindrical housing, inside which the cylindrical combustion chamber is arranged in coaxial alignment and so that an annular additional channel for supplying and heating the air-ammonia mixture is formed between the housing and the combustion chamber by the inlet of the main supply channel and by the tangential vortex generator. The cylindrical side surfaces of the housing and the combustion chamber are provided with at least one through-opening for the spark igniter. Flame sensors, a device for supplying secondary air, and a flame stabilizer are provided, and the combustion chamber is designed to allow cooling and has a catalytic unit mounted at the outlet thereof. The technical result is an improvement in the environmental parameters of the burner, as well as the operating reliability and safety thereof.

IPC Classes  ?

• F23C 1/00 - Combustion apparatus specially adapted for combustion of two or more kinds of fuel simultaneously or alternately, at least one kind of fuel being either a fluid fuel or a solid fuel suspended in air
• F23C 13/08 - Apparatus in which combustion takes place in the presence of catalytic material characterised by the catalytic material
• F23L 9/00 - Passages or apertures for delivering secondary air for completing combustion of fuel
• F23N 5/10 - Systems for controlling combustion using devices responsive to thermal changes or to thermal expansion of a medium using thermocouples

Abstract

meltmelt for 10 hours, exhibits a loss in mass of not more than 2%.

IPC Classes  ?

• G21C 3/64 - Ceramic dispersion fuel, e.g. cermet
• C04B 35/565 - Shaped ceramic products characterised by their compositionCeramic compositionsProcessing powders of inorganic compounds preparatory to the manufacturing of ceramic products based on non-oxides based on carbides based on silicon carbide

Abstract

The group of inventions relates to nuclear fuel material, and more particularly to a high-temperature dense composite nuclear fuel material and a method for producing same. The high-temperature dense composite nuclear fuel material comprises a radiation-inert ceramic matrix in which particles of nuclear fuel are distributed. The matrix is formed from a powder material based on silicon carbide. The particles of nuclear fuel are particles of an oxygen-free nuclear fuel. The method for producing a high-temperature dense composite nuclear fuel material includes preparing a mixture of the nuclear fuel particles and the powder for the formation of a radiation-inert ceramic matrix, compression moulding said mixture and sintering the moulded mixture. The moulded mixture is sintered using the hot pressing method. The group of inventions makes it possible to produce a high-temperature dense composite nuclear fuel material which, after vacuum annealing at 0.63Tmelt for 10 hours, exhibits a loss in mass of not more than 2%.

IPC Classes  ?

• G21C 3/64 - Ceramic dispersion fuel, e.g. cermet

Abstract

The invention relates to the production of carbon products, and more particularly to a technique for treating carbon products during calcination, and can be used in different technical fields for manufacturing electrodes, crucibles and heaters, as well as materials for nuclear engineering such as, for example, uranium-graphite fuel elements. In a method of producing carbon graphite products that includes placing workpieces in a container with a packing and calcining the workpieces in an air atmosphere, carbamide is used as the container packing in an amount of 5-10 wt% of the workpieces. The container is placed in a closed receptacle with a restricted air supply, which is filled with a carbon packing also containing carbamide in an amount of 5-10 wt% of the packing. The aim of the invention and the technical result achieved using the invention is that of simplifying the calcination process and improving the quality of carbon graphite products, particularly products with small dimensions, by preventing oxidation of said products and coking of the packing.

IPC Classes  ?

• C01B 32/21 - After-treatment
• C04B 35/532 - Shaped ceramic products characterised by their compositionCeramic compositionsProcessing powders of inorganic compounds preparatory to the manufacturing of ceramic products based on non-oxides based on carbon, e.g. graphite obtained from carbonaceous particles with or without other non-organic components containing a carbonisable binder

Abstract

The invention relates to the production of carbon products, and more particularly to a technique for treating carbon products during calcination, and can be used in different technical fields for manufacturing electrodes, crucibles and heaters, as well as materials for nuclear engineering such as, for example, uranium-graphite fuel elements. In a method of producing carbon graphite products that includes placing workpieces in a container with a packing and calcining the workpieces in an air atmosphere, carbamide is used as the container packing in an amount of 5-10 wt% of the workpieces. The container is placed in a closed receptacle with a restricted air supply, which is filled with a carbon packing also containing carbamide in an amount of 5-10 wt% of the packing. The aim of the invention and the technical result achieved using the invention is that of simplifying the calcination process and improving the quality of carbon graphite products, particularly products with small dimensions, by preventing oxidation of said products and coking of the packing.

IPC Classes  ?

• C01B 32/21 - After-treatment
• C04B 35/532 - Shaped ceramic products characterised by their compositionCeramic compositionsProcessing powders of inorganic compounds preparatory to the manufacturing of ceramic products based on non-oxides based on carbon, e.g. graphite obtained from carbonaceous particles with or without other non-organic components containing a carbonisable binder

Abstract

The invention relates to measuring technology, and more particularly to equipment for measuring bending of elongate vertically oriented channels, inter alia fuel channels in a high-power channel-type nuclear reactor (RBMK). The technical result is a device that can be manufactured more easily, without detriment to its accuracy in measuring bending of a channel, inter alia a nuclear reactor fuel channel. The device comprises a flexible hollow carrier rod equipped with at least one fibre-optic sensor having a core enclosed in a hermetic tubular casing. The upper part of said core is configured in the form of a positioning sleeve having a central cylindrical cavity. The lower part of the core is configured in the form of a central cylindrical stem. Configured on the central stem of the core are longitudinal sector-shaped cut-outs in which are fastened ferrules with longitudinal through-channels, and disposed below the lower end of the central stem of the core is a gravity pendulum. Passing through the through-channels of the ferrules are fibre-optic lines that connect each fibre-optic sensor to a tunable laser and a photoreceiver that are connected to a computer. The cavity of the hermetic tubular casing of the fibre-optic sensor is filled with an inert gas.

IPC Classes  ?

• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G01B 11/24 - Measuring arrangements characterised by the use of optical techniques for measuring contours or curvatures

Abstract

The invention relates to measuring technology and can be used in the implementation of a method of measuring bending of elongate vertically oriented channels. A fibre-optic sensor fastened to the end of a flexible hollow carrier rod is placed inside a channel. A light signal is fed along fibre-optic lines connected to the sensor, and reflected light signals are recorded. The fibre-optic sensor is equipped with a gravity pendulum. The flexible hollow carrier rod with the fibre-optic sensor is passed along the channel and a photoreceiver and a computer are used to register the shift of an interference pattern of a reflected light signal in a gas gap between the upper end face of the gravity pendulum and the lower end face of fibre-optic lines connected to the photoreceiver and fastened to the sensor, said gas gap varying during the passage of the fibre-optic sensor as a result of the angular motion of the gravity pendulum away from the axis of the bowed channel. On the basis of the registered shifts, profilograms of the variations of the gas gap are recorded for each fibre-optic line, and on the basis of the obtained gas gap profilograms, the magnitude and direction of bending of the channel from the vertical axis are calculated, thereby simplifying the process of measuring bending of a vertically oriented channel while at the same time maintaining measuring accuracy.

IPC Classes  ?

• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations
• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G01B 11/245 - Measuring arrangements characterised by the use of optical techniques for measuring contours or curvatures using a plurality of fixed, simultaneously operating transducers

Abstract

Claimed is a method of measuring bending of a nuclear reactor fuel channel. The method includes equipping a fibre-optic sensor with a gravity pendulum that is suspended for angular motion at the lower end of the fibre-optic sensor; passing a flexible hollow carrier rod with the fibre-optic sensor along the central tube of a fuel assembly and using a photoreceiver and a computer to register the shift of an interference pattern of a reflected light signal in a gas gap between the upper end face of the gravity pendulum and the lower end face of fibre-optic lines connected to the photoreceiver and fastened to the sensor, said gas gap varying during the passage of the fibre-optic sensor as a result of the angular motion of the gravity pendulum away from the axis of the bowed central tube of the fuel assembly; recording, on the basis of the registered shifts of the interference pattern of the reflected light signal, profilograms of the variations of the gas gap for each fibre-optic line of each fibre-optic sensor; and calculating, on the basis of the obtained gas gap profilograms, the magnitude and direction of bending of the central tube of the fuel assembly from the vertical axis, according to which the presence and magnitude of bending of the nuclear reactor fuel channel are determined.

IPC Classes  ?

• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations

Abstract

The invention relates to measuring technology and can be used in the implementation of a method of measuring bending of elongate vertically oriented channels. A fibre-optic sensor fastened to the end of a flexible hollow carrier rod is placed inside a channel. A light signal is fed along fibre-optic lines connected to the sensor, and reflected light signals are recorded. The fibre-optic sensor is equipped with a gravity pendulum. The flexible hollow carrier rod with the fibre-optic sensor is passed along the channel and a photoreceiver and a computer are used to register the shift of an interference pattern of a reflected light signal in a gas gap between the upper end face of the gravity pendulum and the lower end face of fibre-optic lines connected to the photoreceiver and fastened to the sensor, said gas gap varying during the passage of the fibre-optic sensor as a result of the angular motion of the gravity pendulum away from the axis of the bowed channel. On the basis of the registered shifts, profilograms of the variations of the gas gap are recorded for each fibre-optic line, and on the basis of the obtained gas gap profilograms, the magnitude and direction of bending of the channel from the vertical axis are calculated, thereby simplifying the process of measuring bending of a vertically oriented channel while at the same time maintaining measuring accuracy.

IPC Classes  ?

• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations

Abstract

Claimed is a method of measuring bending of a nuclear reactor fuel channel. The method includes equipping a fibre-optic sensor with a gravity pendulum that is suspended for angular motion at the lower end of the fibre-optic sensor; passing a flexible hollow carrier rod with the fibre-optic sensor along the central tube of a fuel assembly and using a photoreceiver and a computer to register the shift of an interference pattern of a reflected light signal in a gas gap between the upper end face of the gravity pendulum and the lower end face of fibre-optic lines connected to the photoreceiver and fastened to the sensor, said gas gap varying during the passage of the fibre-optic sensor as a result of the angular motion of the gravity pendulum away from the axis of the bowed central tube of the fuel assembly; recording, on the basis of the registered shifts of the interference pattern of the reflected light signal, profilograms of the variations of the gas gap for each fibre-optic line of each fibre-optic sensor; and calculating, on the basis of the obtained gas gap profilograms, the magnitude and direction of bending of the central tube of the fuel assembly from the vertical axis, according to which the presence and magnitude of bending of the nuclear reactor fuel channel are determined.

IPC Classes  ?

• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations
• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G01B 11/245 - Measuring arrangements characterised by the use of optical techniques for measuring contours or curvatures using a plurality of fixed, simultaneously operating transducers

Abstract

The invention relates to measuring technology, and more particularly to equipment for measuring bending of elongate vertically oriented channels, inter alia fuel channels in a high-power channel-type nuclear reactor (RBMK). The technical result is a device that can be manufactured more easily, without detriment to its accuracy in measuring bending of a channel, inter alia a nuclear reactor fuel channel. The device comprises a flexible hollow carrier rod equipped with at least one fibre-optic sensor having a core enclosed in a hermetic tubular casing. The upper part of said core is configured in the form of a positioning sleeve having a central cylindrical cavity. The lower part of the core is configured in the form of a central cylindrical stem. Configured on the central stem of the core are longitudinal sector-shaped cut-outs in which are fastened ferrules with longitudinal through-channels, and disposed below the lower end of the central stem of the core is a gravity pendulum. Passing through the through-channels of the ferrules are fibre-optic lines that connect each fibre-optic sensor to a tunable laser and a photoreceiver that are connected to a computer. The cavity of the hermetic tubular casing of the fibre-optic sensor is filled with an inert gas.

IPC Classes  ?

• G01B 11/16 - Measuring arrangements characterised by the use of optical techniques for measuring the deformation in a solid, e.g. optical strain gauge
• G01B 11/24 - Measuring arrangements characterised by the use of optical techniques for measuring contours or curvatures

Abstract

Methods for strengthening soils under bases and foundations of buildings and structures with compaction of the base composed of weak mineral soils by determining the optimal design process parameters of the ground piles over the entire area of the base. The essence of the invention is that the method of compaction of bases composed of weak mineral soils that includes formation of a well, filling each well with the compacting material, and creation of a compacting effect on the compacting material by the hollow tubular working tool to form a ground pile. Preliminary engineering and geological surveys of the base area is performed to determine the values of the modulus of deformation, the Poisson's ratio, the internal friction angle, the specific cohesion, the specific gravity, and the initial void ratio of the weak mineral soil.

IPC Classes  ?

• E02D 27/26 - Compacting soil locally before forming foundationsConstruction of foundation structures by forcing binding substances into gravel fillings
• G01N 33/24 - Earth materials

Abstract

A robotic monitoring, detection and fire extinction control system comprises a pumping station, a fire main and at least two robotized fire extinguishing units connected to the fire main and comprising a nozzle with drives for vertical and horizontal aiming and a head with a drive for adjusting the spray angle of the jet. Mounted at an inlet upstream of the nozzle is a butterfly valve with a drive, and mounted at the outlet of the nozzle, upstream of the head, are a pressure sensor and a portable control console that are connected to a switching unit at the inlet and, via a network controller, to a control device at the outlet. Mounted on the nozzle is a fire detection and video surveillance device that is connected to a signal processing device that is connected to a video monitor device and to a process control system that are connected to a control device that is connected via a fire alarm control device to fire alarms. A foam generating unit is connected to the robotized fire extinguishing units by a foam conduit; additional reserve pipes, connected to the robotized fire extinguishing units, are provided for supplying water and compressed air foam; and an additional controllable butterfly valve, mounted on the foam conduit, is provided and has a drive connected to the switching unit. The head is designed to be capable of supplying water and compressed air foam at a controlled concentration and rate.

IPC Classes  ?

• A62C 37/00 - Control of fire-fighting equipment
• A62C 3/00 - Fire prevention, containment or extinguishing specially adapted for particular objects or places
• A62C 35/58 - Pipe-line systems
• A62C 99/00 - Subject matter not provided for in other groups of this subclass

Abstract

A robotic monitoring, detection and fire extinction control system comprises a pumping station, a fire main and at least two robotized fire extinguishing units connected to the fire main and comprising a nozzle with drives for vertical and horizontal aiming and a head with a drive for adjusting the spray angle of the jet. Mounted at an inlet upstream of the nozzle is a butterfly valve with a drive, and mounted at the outlet of the nozzle, upstream of the head, are a pressure sensor and a portable control console that are connected to a switching unit at the inlet and, via a network controller, to a control device at the outlet. Mounted on the nozzle is a fire detection and video surveillance device that is connected to a signal processing device that is connected to a video monitor device and to a process control system that are connected to a control device that is connected via a fire alarm control device to fire alarms. A foam generating unit is connected to the robotized fire extinguishing units by a foam conduit; additional reserve pipes, connected to the robotized fire extinguishing units, are provided for supplying water and compressed air foam; and an additional controllable butterfly valve, mounted on the foam conduit, is provided and has a drive connected to the switching unit. The head is designed to be capable of supplying water and compressed air foam at a controlled concentration and rate.

IPC Classes  ?

• A62C 3/00 - Fire prevention, containment or extinguishing specially adapted for particular objects or places
• A62C 35/58 - Pipe-line systems
• A62C 37/00 - Control of fire-fighting equipment
• A62C 99/00 - Subject matter not provided for in other groups of this subclass

Abstract

The present invention relates to the field of nuclear power engineering, and more particularly to the field of neutron flux monitoring to enable the monitoring, control and protection of pressure vessel reactors. The invention can be used for correcting errors in the power readings of a nuclear reactor and of a neutron flux monitoring apparatus on the basis of readings from self-powered detectors in an in-core monitoring system. The proposed invention solves the problem of determining with a high degree of accuracy a corrected reactor power value based on the correction of the power of a neutron flux monitoring apparatus according to readings from ex-core neutron flux sensors, taking into account their dependence on the shape of the energy distribution in the core, obtained from in-core detectors in the in-core monitoring system. The technical result of the present invention consists in making it possible to promptly and accurately determine the correction of errors in the power readings of a nuclear reactor in any of its operating modes and thus to provide for safe and reliable operation of the nuclear reactor.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G01T 3/00 - Measuring neutron radiation

Abstract

The invention relates to the field of mechanical engineering and transporting, and more particularly to crane runways for nuclear power plant containment vessels. A crane runway for a nuclear power plant containment vessel comprises cantilever supports uniformly mounted in a circle and fastened at the same height to the inner side of the containment vessel with clearance from one another, and, mounted on top of said cantilever supports, a track in the shape of a circle having the radius of the crane runway, the track being fastened to the cantilever supports by fastening members. Each cantilever support is comprised of an upper horizontal plate and a lower horizontal plate mounted parallel with one another and fastened to one another by vertical longitudinal and transverse ribs. The plates and the transverse ribs are fastened by one end in the reinforced concrete structure of the containment vessel. Both sides of the ends of the plates oriented toward the containment vessel have lateral cutouts, and the upper part of the end face of the cantilever support at the opposite end from the containment vessel has a protruding ledge. The invention makes it possible to increase the robustness of a crane runway for the containment vessel of a nuclear power plant.

IPC Classes  ?

• B66C 7/00 - Runways, tracks, or trackways for trolleys or cranes

Abstract

The invention refers to the area of nuclear power engineering, particularly to auxiliary devices for nuclear power plants, namely to the devices for installation of the outer heat insulation of a nuclear reactor vessel, and can be used at nuclear plants for recovery annealing of the VVER reactor vessel welds. The invention refers to the area of nuclear power engineering, particularly to auxiliary devices for nuclear power plants, namely to the devices for installation of the outer heat insulation of a nuclear reactor vessel, and can be used at nuclear plants for recovery annealing of the VVER reactor vessel welds. The objective to be achieved with the use of the claimed invention is to provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation. The invention refers to the area of nuclear power engineering, particularly to auxiliary devices for nuclear power plants, namely to the devices for installation of the outer heat insulation of a nuclear reactor vessel, and can be used at nuclear plants for recovery annealing of the VVER reactor vessel welds. The objective to be achieved with the use of the claimed invention is to provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation. Reduction of the temperature gradient through the thickness of the nuclear reactor vessel by heat insulation of the external reactor vessel surface, assurance of uniform physical properties for the reactor vessel metal and welds and reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel shall be the technical result of this invention. The invention refers to the area of nuclear power engineering, particularly to auxiliary devices for nuclear power plants, namely to the devices for installation of the outer heat insulation of a nuclear reactor vessel, and can be used at nuclear plants for recovery annealing of the VVER reactor vessel welds. The objective to be achieved with the use of the claimed invention is to provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation. Reduction of the temperature gradient through the thickness of the nuclear reactor vessel by heat insulation of the external reactor vessel surface, assurance of uniform physical properties for the reactor vessel metal and welds and reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel shall be the technical result of this invention. The technical result of the invention is ensured by the fact that the device for installation of the outer heat insulation of a nuclear reactor vessel includes a mobile transfer trolley equipped with the mechanism for its movement, a removable bearing rim located on the mobile transfer trolley with the reactor vessel heat insulation attached to it, at least two lifting devices located on the opposite sides of the reactor vessel at the level of its upper section, in this case the removable bearing rim is connected to the lifting devices with the possibility to lift and lower it. The invention refers to the area of nuclear power engineering, particularly to auxiliary devices for nuclear power plants, namely to the devices for installation of the outer heat insulation of a nuclear reactor vessel, and can be used at nuclear plants for recovery annealing of the VVER reactor vessel welds. The objective to be achieved with the use of the claimed invention is to provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation. Reduction of the temperature gradient through the thickness of the nuclear reactor vessel by heat insulation of the external reactor vessel surface, assurance of uniform physical properties for the reactor vessel metal and welds and reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel shall be the technical result of this invention. The technical result of the invention is ensured by the fact that the device for installation of the outer heat insulation of a nuclear reactor vessel includes a mobile transfer trolley equipped with the mechanism for its movement, a removable bearing rim located on the mobile transfer trolley with the reactor vessel heat insulation attached to it, at least two lifting devices located on the opposite sides of the reactor vessel at the level of its upper section, in this case the removable bearing rim is connected to the lifting devices with the possibility to lift and lower it. Application of the claimed invention will provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation. Heat insulation of the external reactor vessel surface will ensure reduction of the temperature gradient through the thickness of the nuclear reactor vessel, uniformity of the physical properties for its metal and welds as well as reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel.

IPC Classes  ?

• G21C 11/08 - Thermal shieldsThermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield

Abstract

The invention relates to the field of atomic engineering, and more particularly to equipment for detecting leaks in and repairing damage to the inner lining of a water-filled spent fuel pool in a nuclear power station. An arrangement and system for repairing the lining of a spent fuel pool, and more preferably the walls and bottom thereof, are additionally equipped with devices that are not immersible in the spent fuel pool, inter alia, welding equipment, and devices that are immersible in the spent fuel pool, including an immersible repair platform for holding a working mechanism. Said working mechanism is equipped with a device for removing contaminants from weld seams and the surface of the spent fuel pool, and a wire cutting device comprising a receptacle for collecting cuttings and a pump for removing particles. A device for sealing leaks is configured in the form of a small-scale welding assembly coupled to welding equipment, and the immersible repair platform is equipped with means for securing it to the surface of the spent fuel pool. An arrangement for repairing the lining of a spent fuel pool includes an immersible repair robot having a working unit arranged thereon. The invention provides better quality, more reliable repair.

IPC Classes  ?

• G21C 17/07 - Leak testing

Abstract

The invention relates to the field of atomic engineering, and more particularly to equipment for detecting leaks in and repairing damage to the inner lining of a water-filled spent fuel pool in a nuclear power station. An arrangement and system for repairing the lining of a spent fuel pool, and more preferably the walls and bottom thereof, are additionally equipped with devices that are not immersible in the spent fuel pool, inter alia, welding equipment, and devices that are immersible in the spent fuel pool, including an immersible repair platform for holding a working mechanism. Said working mechanism is equipped with a device for removing contaminants from weld seams and the surface of the spent fuel pool, and a wire cutting device comprising a receptacle for collecting cuttings and a pump for removing particles. A device for sealing leaks is configured in the form of a small-scale welding assembly coupled to welding equipment, and the immersible repair platform is equipped with means for securing it to the surface of the spent fuel pool. An arrangement for repairing the lining of a spent fuel pool includes an immersible repair robot having a working unit arranged thereon. The invention provides better quality, more reliable repair.

IPC Classes  ?

• G21C 17/07 - Leak testing

Abstract

The invention relates to cleaning tubes of a tube bundle of a steam generator of a nuclear power plant. A device for cleaning heat exchange tubes of a steam generator of a nuclear power plant comprises: a manipulator which is moveably and fixably mounted in a vertical corridor between bundles of heat exchange tubes; a nozzle that is rotatably mounted on the manipulator and is in the form of a head of injectors connected to a pipeline for supplying high-pressure water jets; and means for remote control and video monitoring, wherein the manipulator is equipped with: a housing which is mounted on a guiding mounting frame and is connected to a base, a drive for main rotational movement which is disposed on the housing, a post which is connected to the housing and is in the form of a hollow column, a drive for auxiliary movement which is fastened on the post, a joint which is connected to the post and is in the form of a tube having a gear rack, and a joint of working elements and a coordinating head. The technical result is a decrease in the cleaning time and reliability of removal of deposits.

IPC Classes  ?

• F28G 3/16 - Rotary appliances using jets of fluid for removing debris
• F28G 15/02 - Supports for cleaning appliances, e.g. frames

Abstract

The invention relates to cleaning tubes of a tube bundle of a steam generator of a nuclear power plant. A device for cleaning heat exchange tubes of a steam generator of a nuclear power plant comprises: a manipulator which is moveably and fixably mounted in a vertical corridor between bundles of heat exchange tubes; a nozzle that is rotatably mounted on the manipulator and is in the form of a head of injectors connected to a pipeline for supplying high-pressure water jets; and means for remote control and video monitoring, wherein the manipulator is equipped with: a housing which is mounted on a guiding mounting frame and is connected to a base, a drive for main rotational movement which is disposed on the housing, a post which is connected to the housing and is in the form of a hollow column, a drive for auxiliary movement which is fastened on the post, a joint which is connected to the post and is in the form of a tube having a gear rack, and a joint of working elements and a coordinating head. The technical result is a decrease in the cleaning time and reliability of removal of deposits.

IPC Classes  ?

• F28G 3/16 - Rotary appliances using jets of fluid for removing debris
• F28G 15/02 - Supports for cleaning appliances, e.g. frames
• B08B 3/02 - Cleaning by the force of jets or sprays

Abstract

The invention relates to technical diagnostics. The claimed method consists in measuring vibration acceleration in three mutually orthogonal planes using vibration sensors mounted at monitoring points on a diesel generator. Vibration accelerations are measured at monitoring points on an operational known good diesel generator, and then vibration accelerations are measured at prescribed intervals at said monitoring points on the diesel generator when the latter is in operation. Temperature and the intensity of an ultrasonic signal are also measured at the aforementioned monitoring points, and the root mean square values of the ultrasonic signal intensity, the temperature and the vibration acceleration are determined. On the basis of the vibration acceleration values measured, the root mean square values of vibration velocity and vibratory displacement are calculated, the distances between clusters of preceding measurements are determined and a conclusion is drawn about the operating state of the diesel generator. The technical result consists in allowing the timely detection of operating defects in a diesel generator by carrying out periodic measurements and comparing the calculation results obtained with one another, thus also providing for the safety of technical equipment in plants.

IPC Classes  ?

• G01M 15/04 - Testing internal-combustion engines
• G01M 15/12 - Testing internal-combustion engines by monitoring vibrations

Abstract

The invention relates to a method for cleaning the surfaces of the tubes of a steam generator tube bundle in a nuclear power station. The method for cleaning the heat exchange tubes of steam generators in a nuclear power station consists in introducing a hydrodynamic manipulator connected to a hoist into a vertical passage inside a heat exchanger, taking preliminary measurements of the dose rate inside the steam generator, measuring eddy current signals and analyzing the signals received in order to assess the condition of the walls of the heat exchange tubes and the thickness of the deposits thereon, determining the coordinates of a cleaning sector and calculating the time and cleaning modes, emitting a high-pressure jet of water using at least one nozzle mounted for rotation about an axis of rotation, and moving the hydrodynamic manipulator according to the coordinates of the cleaning sector, wherein the water jet is set at a feed pressure of from 1000 to 1500 bar and a flow rate of from 100 to 150 l/min and the cleaning zone is subject to video monitoring. The technical result is a shorter cleaning time and more reliable removal of deposits.

IPC Classes  ?

• F28G 3/16 - Rotary appliances using jets of fluid for removing debris

Abstract

The invention relates to technical diagnostics. The claimed method consists in measuring vibration acceleration in three mutually orthogonal planes using vibration sensors mounted at monitoring points on a diesel generator. Vibration accelerations are measured at monitoring points on an operational known good diesel generator, and then vibration accelerations are measured at prescribed intervals at said monitoring points on the diesel generator when the latter is in operation. Temperature and the intensity of an ultrasonic signal are also measured at the aforementioned monitoring points, and the root mean square values of the ultrasonic signal intensity, the temperature and the vibration acceleration are determined. On the basis of the vibration acceleration values measured, the root mean square values of vibration velocity and vibratory displacement are calculated, the distances between clusters of preceding measurements are determined and a conclusion is drawn about the operating state of the diesel generator. The technical result consists in allowing the timely detection of operating defects in a diesel generator by carrying out periodic measurements and comparing the calculation results obtained with one another, thus also providing for the safety of technical equipment in plants.

IPC Classes  ?

• G01M 15/04 - Testing internal-combustion engines
• G01M 15/12 - Testing internal-combustion engines by monitoring vibrations

Abstract

The invention relates to the field of technical diagnostics and can be used for investigating the fluid-tightness of pipes and detecting leaks in pipes in nuclear power stations. Acoustic signals are recorded at two points along the length of a pipe, and the received acoustic signals are subsequently processed. The acoustic signals are recorded in a broad ultrasonic range at two points along the length of the pipe, said points being situated on the pipe upstream and downstream of a valve. The ultrasonic signals recorded at the point on the pipe upstream of the valve and at the point downstream of the valve are then processed, and the values obtained are used to construct two signal spectra. Next, a range of from 15000 to 90000 Hz is isolated and, within this range, the highest amplitude value in each of the signal spectra is selected. Then, the amplitudes of the signal spectra in said frequency range are divided by the highest amplitude value and the difference between the signal spectra upstream and downstream of the valve is determined using a mathematical formula, thereby making it possible to determine the degree of fluid-tightness of the pipe in order to assess its fitness for further service, as well as allowing better quality, more efficient detection of leaks in the pipe.

IPC Classes  ?

• G01M 3/24 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using infrasonic, sonic, or ultrasonic vibrations
• F17D 5/02 - Preventing, monitoring, or locating loss

Abstract

The invention relates to the field of technical diagnostics and can be used for investigating the fluid-tightness of pipes and detecting leaks in pipes in nuclear power stations. Acoustic signals are recorded at two points along the length of a pipe, and the received acoustic signals are subsequently processed. The acoustic signals are recorded in a broad ultrasonic range at two points along the length of the pipe, said points being situated on the pipe upstream and downstream of a valve. The ultrasonic signals recorded at the point on the pipe upstream of the valve and at the point downstream of the valve are then processed, and the values obtained are used to construct two signal spectra. Next, a range of from 15000 to 90000 Hz is isolated and, within this range, the highest amplitude value in each of the signal spectra is selected. Then, the amplitudes of the signal spectra in said frequency range are divided by the highest amplitude value and the difference between the signal spectra upstream and downstream of the valve is determined using a mathematical formula, thereby making it possible to determine the degree of fluid-tightness of the pipe in order to assess its fitness for further service, as well as allowing better quality, more efficient detection of leaks in the pipe.

IPC Classes  ?

• F17D 5/02 - Preventing, monitoring, or locating loss
• G01M 3/24 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using infrasonic, sonic, or ultrasonic vibrations

Abstract

The invention relates to methods for diagnosing the technical condition of electrically driven equipment and can be used for monitoring vibrations in rotating equipment in nuclear power stations. The claimed method includes measuring and recording diagnostic signals in a test period while the equipment under inspection is in operation, dividing the recordings of a reference signal and a signal from the equipment under inspection into five segments of not less than two seconds in length, converting each segment of the recordings of the reference and inspection signals into a spectrum that represents the distribution of amplitudes across frequencies, sampling amplitudes of the spectra of the diagnostic signals of the inspected and fault-free equipment at the frequencies at which deviations appear between the inspection and reference signals, calculating the absolute differences of the amplitudes of the spectra of the inspected and fault-free equipment, ranking the absolute differences and summing the resultant ranks. The sum of the resultant ranks is compared with a critical value to establish whether the rotating equipment is in excess of prescribed vibration values, and to establish the presence of a fault if the sum of the resultant ranks exceeds the critical value.

IPC Classes  ?

• G01M 13/045 - Acoustic or vibration analysis

Abstract

The invention relates to methods for diagnosing the technical condition of electrically driven equipment and can be used for monitoring vibrations in rotating equipment in nuclear power stations. The claimed method includes measuring and recording diagnostic signals in a test period while the equipment under inspection is in operation, dividing the recordings of a reference signal and a signal from the equipment under inspection into five segments of not less than two seconds in length, converting each segment of the recordings of the reference and inspection signals into a spectrum that represents the distribution of amplitudes across frequencies, sampling amplitudes of the spectra of the diagnostic signals of the inspected and fault-free equipment at the frequencies at which deviations appear between the inspection and reference signals, calculating the absolute differences of the amplitudes of the spectra of the inspected and fault-free equipment, ranking the absolute differences and summing the resultant ranks. The sum of the resultant ranks is compared with a critical value to establish whether the rotating equipment is in excess of prescribed vibration values, and to establish the presence of a fault if the sum of the resultant ranks exceeds the critical value.

IPC Classes  ?

• G01M 13/045 - Acoustic or vibration analysis

Abstract

The invention relates to a method for cleaning the surfaces of the tubes of a steam generator tube bundle in a nuclear power station. The method for cleaning the heat exchange tubes of steam generators in a nuclear power station consists in introducing a hydrodynamic manipulator connected to a hoist into a vertical passage inside a heat exchanger, taking preliminary measurements of the dose rate inside the steam generator, measuring eddy current signals and analyzing the signals received in order to assess the condition of the walls of the heat exchange tubes and the thickness of the deposits thereon, determining the coordinates of a cleaning sector and calculating the time and cleaning modes, emitting a high-pressure jet of water using at least one nozzle mounted for rotation about an axis of rotation, and moving the hydrodynamic manipulator according to the coordinates of the cleaning sector, wherein the water jet is set at a feed pressure of from 1000 to 1500 bar and a flow rate of from 100 to 150 l/min and the cleaning zone is subject to video monitoring. The technical result is a shorter cleaning time and more reliable removal of deposits.

IPC Classes  ?

• F28G 3/16 - Rotary appliances using jets of fluid for removing debris
• B08B 3/02 - Cleaning by the force of jets or sprays

Abstract

Methods for non-destructive testing of engineering materials. In one aspect, a method can be used to calculate residual longitudinal and annular welding stresses in welded joints and can be used to assess the quality of pipeline welds according to the criterion of the level of residual stresses and to determine the initial parameters for the pipeline strength calculation. In some aspects, the method enables independent calculation of the longitudinal and hoop residual stresses. Thus, the stresses can be calculated in the seam metal of the pipelines welds, where they reach their maximum values. The method can be used to test a pipeline section using an ultrasonic echo method to measure the propagation time for longitudinal waves and transverse waves polarized along and across the pipe axis. The measurement results define the distinguishing features of the stress state of a welded joint for a specific type of pipe by numerical modeling.

IPC Classes  ?

• G01N 29/07 - Analysing solids by measuring propagation velocity or propagation time of acoustic waves

Abstract

A method for treatment of spent ion-exchange resins for disposal includes feeding a mixture of spent ion-exchange resins to the a loading tank, separating the ion-exchange resins, feeding separated ion-exchange resins into the a drying chamber, vacuum drying the ion-exchange resins and subjecting the resins to additional heat treatment in a high-temperature furnace, and unloading the treated ion-exchange resins into a transport container. A device for treatment of spent ion-exchange resins includes a loading tank, a metering device connected to a drying chamber, an inclined feed screw located between the loading tank and the metering device, a vacuum pump, a heated gas filter, a high-temperature furnace equipped with a vacuum drying and gas purification system, and a feeding device located between the drying chamber and the high-temperature furnace. A docking unit is connected to a lower part of the high-temperature furnace.

IPC Classes  ?

• G21F 9/00 - Treating radioactively contaminated materialDecontamination arrangements therefor
• G21F 9/32 - Processing by incineration
• G21F 9/02 - Treating gases
• G21F 9/34 - Disposal of solid waste
• G21F 9/12 - Processing by absorptionProcessing by adsorptionProcessing by ion-exchange

Abstract

Decontamination method for a nuclear reactor design component, comprising treating the nuclear reactor design component with a low-temperature plasma under the flow of reactionless gas. A selected site on the surface of the design component is supplied to the electrode. Plasma discharge is ignited between the surface of the design component chosen as the cathode, and the electrode chosen as the anode. Operating parameters of the discharge effective to sputtering of the cathode are chosen. The cathode is sputtered. The electrode and a gas pipeline diverting inert gas from the discharge zone are cooled to a temperature sufficient for the precipitation of the sputtered atoms on the surface of the electrode and lines. After sputtering the cathode to the specified depth, the electrode is moved to a new area of treatment and the process steps are repeated until the complete treatment of the entire surface of the design component.

IPC Classes  ?

• G21F 9/00 - Treating radioactively contaminated materialDecontamination arrangements therefor
• G21F 9/30 - Processing

Abstract

Steels and alloys for use in the nuclear power engineering industry during the manufacture of equipment, including manufacturing an in-vessel baffle of pressurized water power reactors (VVER) with a life less than 60 years. In one aspect, an austenite chromium-nickel steel is provided with increased resistance to swelling on impact of neutron flows at doses up to 150 dpa. In some aspects, the result is accomplished by adding molybdenum, calcium and rare earth metals (REM) lanthanum and cerium to steel containing carbon, silicon, manganese, chromium, nickel, titanium and iron at specified mass ratios.

IPC Classes  ?

• C22C 38/60 - Ferrous alloys, e.g. steel alloys containing lead, selenium, tellurium or antimony, or more than 0.04% by weight of sulfur
• C22C 38/58 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with more than 1.5% by weight of manganese
• C22C 38/52 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with cobalt
• C22C 38/50 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with titanium or zirconium
• C22C 38/44 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with molybdenum or tungsten
• C22C 38/42 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with copper
• C22C 38/02 - Ferrous alloys, e.g. steel alloys containing silicon
• C22C 38/00 - Ferrous alloys, e.g. steel alloys

Abstract

Apparatus for drying Spent Ion-Exchange Resins (SIER), which can intensify the SIER drying process, reduce power consumption, and accelerate discharge of SIER when the drying process is completed. The apparatus comprises a sealed cylindrical body, and a blow-down choke installed in an upper part of the sealed cylindrical body and a nozzle to feed the spent ion-exchange resins is installed inside the body, and a nozzle to retrieve dried ion-exchange resins is installed in its bottom part and equipped with a locking device. An external heater is provided for the body, and a drive shaft that is installed in alignment inside the body, capable of rotation, and equipped with a stirrer. A lower part of the drive shaft with lower screw winding is installed in alignment inside the nozzle to retrieve dried ion-exchange resins. The nozzle to retrieve dried ion-exchange resins is equipped with a water draining device.

IPC Classes  ?

• F26B 9/08 - Machines or apparatus for drying solid materials or objects at rest or with only local agitationDomestic airing cupboards in stationary drums or chambers including agitating devices
• G21F 9/16 - Processing by fixation in stable solid media
• F26B 25/04 - Agitating, stirring, or scraping devices
• G21F 9/28 - Treating solids

Abstract

Device for deactivating radioactive elements comprising a loading-unloading manipulator, a process chamber with a lid with a round turn-around table located inside, a waste disposal container and an exhaust ventilation unit. The process chamber is equipped with a remote handling unit mounted on its outer part, a device for a dust-free abrasive blast cleaning, a waterjet cutting unit, connected through a leak-tight installed hoses in the process chamber walls, an abrasive and slurry collection unit and at least one device for radiation control, located inside the process chamber, the solid and liquid waste separation unit and discharge liquid waste. The process chamber is equipped with a chemical-radiation-shock proof sound-absorbing coating, and the remote control unit is connected with the device for a dust-free abrasive blast cleaning, waterjet cutting unit, abrasive and slurry collection device, exhaust ventilation unit, the process chamber lid, the radiation unit and the waste separation unit.

IPC Classes  ?

• G21F 9/00 - Treating radioactively contaminated materialDecontamination arrangements therefor
• G21F 9/30 - Processing
• B24C 3/32 - Abrasive blasting machines or devicesPlants designed for abrasive blasting of particular work, e.g. the internal surfaces of cylinder blocks
• B24C 9/00 - Appurtenances of abrasive blasting machines or devices, e.g. working chambers, arrangements for handling used abrasive material

Abstract

A method of fragmentation of elements of a nuclear reactor includes placement of elements inside a cask and subsequent cutting, the cask being perforated. Each element is lowered into the cask by a full internal height of the cask using a gripper having clamping jaws. The element is intercepted at an upper edge of the cask, lifted, and positioned using video surveillance and artificial lighting so that a hydraulic cutter is directly under the clamping jaws. The element is cut at a point corresponding to a level of the upper edge of the cask, separating from the element a fragment equal to the internal height of the cask. Then the upper part of the element remaining after cutting is lowered inside the cask by the full internal height of the cask and the cutting of the element into fragments is repeated until the element is fully cut to fragments.

IPC Classes  ?

• G21C 19/36 - Mechanical means only
• G21D 1/00 - Details of nuclear power plant

Abstract

The invention relates to the field of nuclear engineering, and more particularly to auxiliary devices for nuclear power plants, and even more particularly to devices for installing external thermal insulation on a nuclear reactor vessel, and can be used in nuclear power stations for performing recovery annealing operations on weld seams of pressurized water reactor vessels. The claimed invention is directed toward solving the problem of allowing the installation and removal of thermal insulation on the outer surface of a pressurized water reactor vessel from the confined space below a reactor and at elevated levels of ionizing radiation. The technical result of the present invention is that of reducing the temperature gradient across the thickness of a nuclear reactor vessel by thermally insulating the outer surface of the reactor vessel, providing for uniformity of the physical properties of the metal and weld seams of the reactor vessel, and reducing the impact of thermal effects on surrounding structures during the recovery annealing of weld seams and (or) of the main metal of a pressurized water reactor vessel. The technical result is achieved in that a device for installing external thermal insulation on a nuclear reactor vessel comprises a mobile transport trolley equipped with a mechanism for the movement thereof, a detachable carrier rim disposed on the transport trolley and to which thermal insulation for the reactor vessel is attached, and at least two hoists disposed on opposite sides of the reactor vessel at the level of the top of the reactor vessel, the detachable carrier rim being connected to the hoists to allow raising and lowering of the rim. Use of the claimed invention makes it possible to install and remove thermal insulation on the outer surface of a pressurized water reactor vessel from the confined space below a reactor and at elevated levels of ionizing radiation. The thermal insulation of the outer surface of a reactor vessel makes it possible to reduce the temperature gradient across the thickness of the nuclear reactor vessel, provide for uniformity of the physical properties of the metal and weld seams of the reactor vessel, and reduce the impact of thermal effects on surrounding structures during the recovery annealing of weld seams and (or) of the main metal of a pressurized water reactor vessel.

IPC Classes  ?

• F16L 59/12 - Arrangements for supporting insulation from the wall or body insulated, e.g. by means of spacers between pipe and heat-insulating materialArrangements specially adapted for supporting insulated bodies
• G21C 11/08 - Thermal shieldsThermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield

Abstract

An external thermal insulation for a nuclear reactor vessel comprises posts, supporting rings, and thermal insulation rings. The posts are equidistantly mounted on the floor of the space beneath a reactor and are provided with guide grooves. Each post is hingedly mounted on a post base, wherein a hinge connecting a post and a post base is offset from the centre of gravity of the post to allow the post to deviate from and return to a vertical position, and the post base is provided with an adjustable screw-type support and has a supporting platform. A system for installing external thermal insulation for a nuclear reactor vessel comprises hoists and a mobile transport trolley with removable installation equipment. The transport trolley is provided with a mechanism for controlling the removable installation equipment and is further provided with a drive for moving the transport trolley, an end-of-travel device for stopping movement of the transport trolley, a pump unit, and, connected to said pump unit and disposed at the centre of the transport trolley, a hydraulic jack with a rotatable disk fastened to the piston of the jack, and the removable installation equipment includes a device for installing posts, a device for installing hoists, and a device for installing supporting rings and thermal insulation rings.

IPC Classes  ?

• F16L 59/12 - Arrangements for supporting insulation from the wall or body insulated, e.g. by means of spacers between pipe and heat-insulating materialArrangements specially adapted for supporting insulated bodies
• G21C 11/08 - Thermal shieldsThermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield

Abstract

The invention relates to the field of nuclear engineering, and more particularly to auxiliary devices for nuclear power plants, and even more particularly to devices for installing external thermal insulation on a nuclear reactor vessel, and can be used in nuclear power stations for performing recovery annealing operations on weld seams of pressurized water reactor vessels. The claimed invention is directed toward solving the problem of allowing the installation and removal of thermal insulation on the outer surface of a pressurized water reactor vessel from the confined space below a reactor and at elevated levels of ionizing radiation. The technical result of the present invention is that of reducing the temperature gradient across the thickness of a nuclear reactor vessel by thermally insulating the outer surface of the reactor vessel, providing for uniformity of the physical properties of the metal and weld seams of the reactor vessel, and reducing the impact of thermal effects on surrounding structures during the recovery annealing of weld seams and (or) of the main metal of a pressurized water reactor vessel. The technical result is achieved in that a device for installing external thermal insulation on a nuclear reactor vessel comprises a mobile transport trolley equipped with a mechanism for the movement thereof, a detachable carrier rim disposed on the transport trolley and to which thermal insulation for the reactor vessel is attached, and at least two hoists disposed on opposite sides of the reactor vessel at the level of the top of the reactor vessel, the detachable carrier rim being connected to the hoists to allow raising and lowering of the rim. Use of the claimed invention makes it possible to install and remove thermal insulation on the outer surface of a pressurized water reactor vessel from the confined space below a reactor and at elevated levels of ionizing radiation. The thermal insulation of the outer surface of a reactor vessel makes it possible to reduce the temperature gradient across the thickness of the nuclear reactor vessel, provide for uniformity of the physical properties of the metal and weld seams of the reactor vessel, and reduce the impact of thermal effects on surrounding structures during the recovery annealing of weld seams and (or) of the main metal of a pressurized water reactor vessel.

IPC Classes  ?

• G21C 11/08 - Thermal shieldsThermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield
• F16L 59/12 - Arrangements for supporting insulation from the wall or body insulated, e.g. by means of spacers between pipe and heat-insulating materialArrangements specially adapted for supporting insulated bodies

Abstract

An external thermal insulation for a nuclear reactor vessel comprises posts, supporting rings, and thermal insulation rings. The posts are equidistantly mounted on the floor of the space beneath a reactor and are provided with guide grooves. Each post is hingedly mounted on a post base, wherein a hinge connecting a post and a post base is offset from the centre of gravity of the post to allow the post to deviate from and return to a vertical position, and the post base is provided with an adjustable screw-type support and has a supporting platform. A system for installing external thermal insulation for a nuclear reactor vessel comprises hoists and a mobile transport trolley with removable installation equipment. The transport trolley is provided with a mechanism for controlling the removable installation equipment and is further provided with a drive for moving the transport trolley, an end-of-travel device for stopping movement of the transport trolley, a pump unit, and, connected to said pump unit and disposed at the centre of the transport trolley, a hydraulic jack with a rotatable disk fastened to the piston of the jack, and the removable installation equipment includes a device for installing posts, a device for installing hoists, and a device for installing supporting rings and thermal insulation rings.

IPC Classes  ?

• G21C 11/08 - Thermal shieldsThermal linings, i.e. for dissipating heat from gamma radiation which would otherwise heat an outer biological shield
• F16L 59/12 - Arrangements for supporting insulation from the wall or body insulated, e.g. by means of spacers between pipe and heat-insulating materialArrangements specially adapted for supporting insulated bodies

Abstract

Devices for eliminating radioactive contamination of radioactive waste by providing adaptive processing of the decontamination solution for reuse. The plant for electrochemical decontamination of metal radioactive waste includes a pipe equipped with shut-off valves, a radioactive waste processing module that comprises a unit for electrochemical decontamination connected by a ventilation channel to the ventilation module and pipe for decontamination solution supply and discharge equipped with shut-off valves. The plant is equipped with a decontamination solution preparation module connected with a pipe for decontamination solution supply and discharge, at least one pump, while the module for decontamination solution receiving is equipped with devices for cleaning and pH correction of decontamination solution, and the unit for electrochemical decontamination of metal radioactive waste, the module for decontamination solution receiving and the decontamination solution preparation module are equipped with pH measurement elements.

IPC Classes  ?

• G21F 9/00 - Treating radioactively contaminated materialDecontamination arrangements therefor
• G21F 9/34 - Disposal of solid waste

Abstract

The invention can be used for cutting out holes and cylindrical through samples from the housing of a nuclear reactor. The present device comprises a mechanism for rotating and feeding a cutting tool, and a movable platform. A vertical wall of a maintenance cabin is provided with a through-hole for a positioning bushing that has a mounting assembly fastened to an inside flange thereof, said mounting assembly comprising a mounting plate and, hingedly mounted thereto, a pivotable positioning plate. Said plates are provided with a mechanism for fixing the plates in an interlocked position. The movable platform is equipped with height-adjustable pivotable wheeled supports, is mounted inside the maintenance cabin on the floor thereof, is rigidly fastened to a gusset plate of the pivotable positioning plate, and comprises a mechanism for rotating and feeding a cutting tool. Fastened to an outside flange of the positioning bushing is a catching member for catching trepanned samples and collecting shavings. The invention allows operations to be performed easily and safely.

IPC Classes  ?

• B23B 35/00 - Methods for boring or drilling, or for working essentially requiring the use of boring or drilling machinesUse of auxiliary equipment in connection with such methods
• G21C 19/00 - Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel

Abstract

The invention relates to the field of mechanical engineering, namely to process equipment for nuclear power industry. The objective of the present invention is to ensure the convenience and safety of work when performing process operations for cutting trepans at one workplace in a limited space of the repair cabin. The technical result achieved by the proposed invention is to ensure the accuracy of the alignment of the cutting tool and reduce the cutting tool installation time and, as a result, reduce the radiation exposure loads on the service personnel. This technical result is achieved due to the fact that the device for holes and trepans cutting containing a mechanism of rotation and feeding of the cutting tool and a movable platform is, according to the present invention, further provided with a box-shaped repair cabin performing the role of a device body, with athrough hole made in a vertical wall of the repair cabin, the said holeaccommodating a positioning sleeve mounted with an inner flange facing the inside of the repair cabin and an outer flange fixed outside the repair cabin, with a mounting fixture secured on the inner flange of the positioning sleeve, where the said mounting fixture comprises a mounting plate fixed on the inner flange of the positioning sleeve and a rotating positioning plate pivotally installed on the mounting plate with a possibility of rotating around the vertical axis, where the mounting plate and the rotating positioning plate are provided with a mechanism for fixing in the closed position, the movable platform is provided with height-adjustable rotatable wheel supports installed inside on the bottom of the repair cab and rigidly attached to the gusset of the rotating positioning plate, the mechanism of rotation and feeding of the cutting tool is installed on the movable platform with the mounting and rotating positioning plates being in the closed position coaxially with the positioning sleeve, and a catcher is fixed on the outer flange of the positioning sleeve for primary storage of cut trepans and collection of chips. Preferably, the movable platform of the device for holes and trepans cutting is a spatial truss with height-adjustable rotatable wheel supports. The device for holes and trepans cutting may be provided with a fixture for branding trepans beforecutting the same, including a hollow cylindrical body, end covers installed at the ends of the cylindrical body having groove holes made on them along the diameter line, connected to each other in pairs on different end covers by through guide tubes placed inside the cylindrical body, a rod for branding trepans placed in one of the guide tubes, with a removable striker at one end and a digital stamp on the other, and removable plugs installed in the other guide tubes, where the outer surface of the cylindrical body is made according to the running fit for installation inside the positioning sleeve, and the space inside the cylindrical body between the guide tubes is filled with lead shot to a part of the length of the cylindrical body. The device for holes and trepans cutting is preferably provided with a lead safety plug installed in the positioning sleeve from the end facing the inside of the repair cabin. The mechanism of rotation and feeding of the cutting tool of the device for holes and trepans cutting may be implemented in the form of a drilling machine, preferably pneumatic, and the cutting tool is implemented in the form of a core drill. A drilling fixture of the device for holes and trepans cutting may be made in the form of a hollow cylinder provided with end caps installed at its ends, with central holes made in them, interconnected by a hollow tube configured to accommodate an ejector and to supply a cooling fluid to the cutting tool, where one of the end caps has a mounting bore for installing the cutting tool, and the other end cap is configured to be installed and fixed in the chuck of the drilling machine.

IPC Classes  ?

• B23B 41/00 - Boring or drilling machines or devices specially adapted for particular workAccessories specially adapted therefor
• B23B 47/00 - Constructional features of components specially designed for boring or drilling machinesAccessories therefor

Abstract

The invention relates to nuclear engineering, and more particu-larly to the drying of spent ion-exchange resins. An apparatus for drying spent ion-exchange resins comprises: a hermetically sealed cylindrical housing hav-ing in the top part thereof a bleed nozzle and a pipe for feeding spent ion-ex-change resins into the housing, and having in its bottom part a pipe for remov-ing dried spent ion-exchange resins which is provided with a closing device; an external heater for heating the housing; and also a drive shaft coaxially and ro-tatably mounted inside the housing and equipped with an agitator. Said agitator is configured in the form of an anchor impeller, the blades of which are shaped to be conformal with the inside surface of the bottom and side parts of the hous-ing, and codirectional upper and lower screw flights arranged on the drive shaft above and below the point of attachment of said anchor impeller. The lower part of the drive shaft with the lower screw flight is arranged coaxially inside the pipe for removing dried spent ion-exchange resins. The pipe for removing dried spent ion-exchange resins is provided with a means for draining water. The invention makes it possible to reduce the time and energy intensiveness of the process of drying spent ion-exchange resins.

IPC Classes  ?

• G21F 9/28 - Treating solids

Abstract

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IPC Classes  ?

• C22C 38/58 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with more than 1.5% by weight of manganese
• C22C 38/52 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with cobalt
• C22C 38/50 - Ferrous alloys, e.g. steel alloys containing chromium with nickel with titanium or zirconium

Abstract

The invention relates to nuclear engineering, and more particularly to the drying of spent ion-exchange resins. An apparatus for drying spent ion-exchange resins comprises: a hermetically sealed cylindrical housing having in the top part thereof a bleed nozzle and a pipe for feeding spent ion-exchange resins into the housing, and having in its bottom part a pipe for removing dried spent ion-exchange resins which is provided with a closing device; an external heater for heating the housing; and also a drive shaft coaxially and rotatably mounted inside the housing and equipped with an agitator. Said agitator is configured in the form of an anchor impeller, the blades of which are shaped to be conformal with the inside surface of the bottom and side parts of the housing, and codirectional upper and lower screw flights arranged on the drive shaft above and below the point of attachment of said anchor impeller. The lower part of the drive shaft with the lower screw flight is arranged coaxially inside the pipe for removing dried spent ion-exchange resins. The pipe for removing dried spent ion-exchange resins is provided with a means for draining water. The invention makes it possible to reduce the time and energy intensiveness of the process of drying spent ion-exchange resins.

IPC Classes  ?

• G21F 9/28 - Treating solids

Abstract

The invention relates to metallurgy of alloyed steels and alloys that are intended for use in nuclear power engineering in the production of main equipment for nuclear power plants, namely for the manufacture of an internal baffle for pressurized water reactors (PWR) with a service life of at least 60 years. The technical result of the present invention is the creation of an austenitic nickel-chromium steel that has increased swelling resistance when exposed to neutron fluxes at doses of up to 150 dpa. The technical result is achieved in that molybdenum, calcium and the rare-earth metals (REM) lanthanum and cerium are additionally introduced into the composition of a known steel containing carbon, silicon, manganese, chromium, nickel, titanium, and iron with the ratio of elements as follows: 0.06-0.10 wt% carbon, 0.40-0.60 wt% silicon, 1.50-2.00 wt% manganese, 15.0-16.0 wt% chromium, 24.00-26.00 wt% nickel, 0.70-1.40 wt% molybdenum, (5 * carbon content + 0.10) - 0.80 wt% titanium, 0.001-0.003 wt% calcium, 0.001-0.005 wt% lanthanum and cerium, < 0.035 wt% phosphorus, < 0.008 wt% sulfur, < 0.020 wt% nitrogen, < 0.025 wt% cobalt, < 0.3 wt% copper, < 0.001 wt% tin, < 0.001 wt% antimony, < 0.001 wt% arsenic, < 0.001 wt% bismuth, < 0.001 wt% lead, and the remainder being iron. Moreover, in order to most fully achieve the technical result, the following ratios should simultaneously be used: The chromium equivalent, calculated according to the formula CCr eq=CCr+CMo+1.5*CSi+0.5*CTi, should not exceed the value Ceq/Cr<18.0, where CCr is the chromium content, wt%; CMo is the molybdenum content, wt%; CSi is the silicon content, wt%; and CTi is the titanium content, wt%. The nickel equivalent, calculated according to the formula CNi eq=CNi+30*CC+0.5*CMn+30*CN, should not be below the value of Ceq/Ni<27?5, where CNi is the nickel content, wt%; CC is the carbon content, wt%; CMn is the manganese content, wt%; and CN is the nitrogen content, wt%. The proposed austenitic nickel-chromium steel with a nickel content of 25% makes it possible to achieve the indicated technical result (increased resistance to swelling while maintaining the required mechanical properties).

Abstract

The group of inventions relates to non-destructive testing of engineering materials. It can be used for the calculation of residual longitudinal and annular welding stresses in welded joints and can be used to assess the quality of the pipelines welds according to the criterion of the level of residual stresses and to determine the initial parameters for the pipeline strength calculation. The technical result of the claimed group of inventions is to provide the possibility of independent calculation of the longitudinal and hoop residual stresses. Thus, the stresses are calculated in the seam metal of the pipelines welds, where they reach their maximum values. The technical result of the invention according to the first embodiment is provided in that in the method of the calculation of residual stresses in the weld metal of pipelines welds, consisting in that on the pipeline section under tests the propagation time for longitudinal waves and transverse waves polarized along and across the pipe axis are measured with the ultrasonic echo method, and the measurement results define the distinguishing features of the stress state of a welded joint according to the claimed solution for a specific type of pipe by numerical modeling to pre-define the position of the cross-sections balancing, which balancing the hoop stresses in the base metal reaches the minimum value, and the balancing coefficient value, equal to the ratio of the maximum membrane residual tensile hoop stresses in the seam metal to the value of the minimum residual compressive membrane hoop stresses in the base metal. Further, prior to the welded joints completion, measurements of the initial values of the propagation time of the longitudinal wave and transverse waves polarized along and across the pipe axis are performed in the balancing sections are performed; after the welded joints completion at the same measurement points, measurements of the performance values of the propagation time of the same types of waves are performed, according to the results of measurements applying the acoustoelasticity equations for a biaxial stress state:where tol, t02, t03 are the initial and ti, t2, t3 are the performance values of the propagation times of the transverse waves polarized along and across the generating line of a tube and longitudinal wave, respectively, K1 and K2 are the coefficients of the acoustic-elastic linkage, for each section of measurements, the values of membrane longitudinal and hoop stresses are calculated, as well as bending moments, using which, based on the principle of balancing membrane stresses and taking into account the balancing coefficient, the maximum values of residual longitudinal and hoop local stresses in the seam metal are calculated. The technical result of the invention according to the second embodiment is provided in that in the method of the calculation of residual stresses in the weld metal of pipelines welds made of the acoustically isotropic metal, consisting in that on the pipeline section under tests the propagation time for the bulk waves are measured with the ultrasonic echo method, and the measurement results define the distinguishing features of the stress state of a welded joint according to the claimed solution for a specific type of pipe by numerical modeling to pre-define the position of the cross-sections balancing, which balancing the hoop stresses in the base metal reaches the minimum value, and the balancing coefficient value, equal to the ratio of the maximum membrane tensile hoop stresses in the seam metal to the value of the minimum compressive membrane hoop stresses in the base metal and also the value of the intrinsic acoustic anisotropy of the base metal. After the welded joint completion, in the balancing sections, measurements of the performance values of the propagation time of transverse waves polarized along and across the pipe axis are performed, according to the results of measurements using the acoustoelasticity equation for the difference between the longitudinal and hoop membrane stresses the specified difference for each measuring point is calculated:where, ao is the intrinsic acoustic anisotropy of the base metal, D is the elastic-acoustic coupling coefficient, the separation of longitudinal and hoop membrane stresses is performed using the results of defining the longitudinal stresses in the additional section, then for each measurement section, the values of membrane longitudinal and hoop stresses are calculated, as well as bending moments, using which, based on the principle of balancing membrane stresses and taking into account the balancing coefficient, the maximum values of residual longitudinal and hoop local stresses in the seam metal are calculated.21

IPC Classes  ?

• G01N 29/07 - Analysing solids by measuring propagation velocity or propagation time of acoustic waves

Abstract

On a region of interest of a pipeline, an ultrasonic pulse-echo technique is used to measure the propagation time of a longitudinal wave and of transverse waves polarized along and across the axis of the pipe, and the measuring results are used to determine the characteristics of the stress state of a welded joint. Using computational modelling, the position of equilibrium sections, in which equilibrating circumferential stresses in the main metal exhibit minimum values, and the value of an equilibrium coefficient that is equal to the ratio of the maximum residual circumferential membrane tensile stresses in the metal of the seam to the value of the minimum residual circumferential membrane compressive stresses in the main metal are pre-calculated. Prior to formation of a welded joint, initial values of the propagation time of a longitudinal wave and of transverse waves polarized along and across the axis of the pipe are measured in the equilibrium sections. After formation of the welded joint, working values of the propagation time of the same type of waves are measured at the same points. On the basis of the measuring results, the values of the longitudinal and circumferential membrane stresses and also of the bending moments are calculated for each section and are used, together with the equilibrium coefficient, to calculate the maximum values of the residual longitudinal and circumferential local stresses in the metal of the seam.

IPC Classes  ?

• G01N 29/07 - Analysing solids by measuring propagation velocity or propagation time of acoustic waves

Abstract

The invention relates to nuclear engineering. A method for decontaminating a structural element of a nuclear reactor involves treating a structural element of a nuclear reactor with low-temperature plasma while supplying a stream of a chemically inert gas. An electrode is applied to a selected area on a surface of the structural element, a plasma discharge is ignited between the surface of the structural element functioning as a cathode and an electrode functioning as an anode, and working parameters of the discharge are selected which are effective for sputtering the surface of the cathode. The cathode is sputtered. The electrode and a gas pipeline for removing the chemically inert gas from the discharge zone are cooled to a temperature sufficient for the sputtered atoms to be deposited on the surfaces of the electrode and the pipeline. Once the surface of the cathode has been sputtered to a given depth, the electrode is moved to a newly selected treatment area and the operations of the method are repeated until the entire surface of the structural element to be decontaminated has been fully treated. The invention makes it possible to reduce significantly the radioactivity of the power plant structures treated and to reduce the amount of radioactive waste produced.

IPC Classes  ?

• G21F 9/28 - Treating solids

Abstract

The invention relates to nuclear engineering. A method for decontaminating a structural element of a nuclear reactor involves treating a structural element of a nuclear reactor with low-temperature plasma while supplying a stream of a chemically inert gas. An electrode is applied to a selected area on a surface of the structural element, a plasma discharge is ignited between the surface of the structural element functioning as a cathode and an electrode functioning as an anode, and working parameters of the discharge are selected which are effective for sputtering the surface of the cathode. The cathode is sputtered. The electrode and a gas pipeline for removing the chemically inert gas from the discharge zone are cooled to a temperature sufficient for the sputtered atoms to be deposited on the surfaces of the electrode and the pipeline. Once the surface of the cathode has been sputtered to a given depth, the electrode is moved to a newly selected treatment area and the operations of the method are repeated until the entire surface of the structural element to be decontaminated has been fully treated. The invention makes it possible to reduce significantly the radioactivity of the power plant structures treated and to reduce the amount of radioactive waste produced.

IPC Classes  ?

• G21F 9/28 - Treating solids

Abstract

A mobile compressed foam firefighting unit comprising a mixing chamber connected at the outlet to the foam feeder, and the following systems connected to the mixing chamber inlet: a water supply system comprising a water pump and a water pump drive, a foam concentrate supply system comprising a foam pump and a foam pump drive, and an air supply system comprising an air compressor and an air pump drive. The system includes a drive motor, and drives of the air compressor and the foam pump comprise variable hydraulic transmissions kinetically connected to the drive motor, and the system is equipped with a water flow meter, a throttle valve with an electric drive and a check valve, and an electronic control unit of the throttle valve installed in the water supply pipe between the water pump and the mixing chamber.

IPC Classes  ?

• A62C 5/02 - Making of fire-extinguishing materials immediately before use of foam
• A62C 27/00 - Fire-fighting land vehicles
• A62C 37/36 - Control of fire-fighting equipment an actuating signal being generated by a sensor separate from an outlet device
• B60L 1/00 - Supplying electric power to auxiliary equipment of electrically-propelled vehicles

Abstract

The invention relates to devices for ridding radioactive waste of radioactive contamination. A device for deactivating radioactive elements comprises a loading and unloading manipulator, a working chamber with a roof, a rotary table arranged inside said chamber, a container for collecting waste and an exhaust ventilation device. The working chamber is equipped with a remote control unit, a device for dust-free abrasive blast-cleaning, an abrasive water jet cutter, which is connected to a device for collecting abrasive and cuttings by means of hoses hermetically mounted in the walls of the working chamber, a radiation monitoring device, which is arranged inside the working chamber, a device for separating liquid and solid waste, and a liquid waste drain, an abrasive water jet cutter, a device for collecting abrasive and cuttings, an exhaust ventilation system, a roof of the working chamber, a radiation monitoring device and a device for separating liquid and solid waste. The working chamber is equipped with an impact-resistant, soundproof cover which is resistant to chemicals and radiation. The remote control unit is connected to the device for dust-free abrasive blast-cleaning. The invention makes it possible to reduce the amount of waste and also to minimize hazards and radiation exposure.

IPC Classes  ?

• G21F 9/28 - Treating solids
• G21F 9/30 - Processing

Abstract

The invention relates to devices for ridding radioactive waste of radioactive contamination. A device for deactivating radioactive elements comprises a loading and unloading manipulator, a working chamber with a roof, a rotary table arranged inside said chamber, a container for collecting waste and an exhaust ventilation device. The working chamber is equipped with a remote control unit, a device for dust-free abrasive blast-cleaning, an abrasive water jet cutter, which is connected to a device for collecting abrasive and cuttings by means of hoses hermetically mounted in the walls of the working chamber, a radiation monitoring device, which is arranged inside the working chamber, a device for separating liquid and solid waste, and a liquid waste drain, an abrasive water jet cutter, a device for collecting abrasive and cuttings, an exhaust ventilation system, a roof of the working chamber, a radiation monitoring device and a device for separating liquid and solid waste. The working chamber is equipped with an impact-resistant, soundproof cover which is resistant to chemicals and radiation. The remote control unit is connected to the device for dust-free abrasive blast-cleaning. The invention makes it possible to reduce the amount of waste and also to minimize hazards and radiation exposure.

IPC Classes  ?

• G21F 9/28 - Treating solids
• G21F 9/30 - Processing

Abstract

There is disclosed herein methods of treatment of spent ion-exchange resins for disposal, and devices for implementation of the same. The disclosed methods comprise feeding a mixture of spent ion-exchange resins with transport water to a loading tank, separating ion-exchange resins from transport water by settling the mixture thereof and draining the transport water from the loading tank. The subsequent metered feed of ion-exchange resins separated from the transport water into the drying chamber, vacuum drying with simultaneous mixing of the ion-exchange resins is in the drying chamber at a temperature not exceeding 90 C and unloading the treated ion-exchange resin into a transport container.

IPC Classes  ?

• B01J 49/00 - Regeneration or reactivation of ion-exchangersApparatus therefor

Abstract

The invention relates to the complex processing of a nuclear power plant's NPP's liquid, boron-containing waste with a complex composition, being generated during the operation of NPPs, including of ones VVER-type, and can be used to isolate boric and nitric acids and hydroxides of sodium and potassium for their reuse in the NPP process cycle. The invention allows to obtain crystalline boric acid and highly concentrated solutions of nitric acid and hydroxides of sodium and potassium, suitable for reuse in the NPP process cycle and for general industrial use. Conducting electrodialysis at low values of current and voltage provides a reduction of the method's energy intensity. The involvement of all major components of waste mother liquors into the processing reduces the amount of stored and disposed hazardous waste.

IPC Classes  ?

• B01D 61/44 - Ion-selective electrodialysis
• C01B 35/10 - Compounds containing boron and oxygen
• C01D 1/20 - Preparation by reacting oxides or hydroxides with alkali metal salts
• G21F 9/12 - Processing by absorptionProcessing by adsorptionProcessing by ion-exchange

Abstract

A method of processing spent ion-exchange resins for burial includes delivering a mixture of spent ion-exchange resins with transport water to a loading tank, separating the ion-exchange resins from the transport water by means of settling of the mixture and draining of the transport water from the loading tank, subsequent metered delivery of the ion-exchange resins separated from the transport water to a drying chamber, vacuum drying with simultaneous mixing of the ion-exchange resins in the drying chamber at a temperature of no more than 90°C and unloading the spent ion-exchange resin to the transport container. The ion-exchange resins after completion of vacuum drying in the drying chamber undergo additional heat treatment in a high-temperature furnace at a temperature of 250-300°C with simultaneous mixing and vacuum drying. The spent ion-exchange resin is unloaded into the transport container after heat treatment in the high-temperature furnace.

IPC Classes  ?

• G21F 9/16 - Processing by fixation in stable solid media

Abstract

The invention relates to the field of nuclear technology. A method for dividing long elements of a nuclear reactor into fragments comprises placing the long elements inside a container and subsequently cutting said long elements. A long element is lowered into the container over the entire height thereof. The long element is cut level with the upper edge of the container, with a fragment equal to the height of the container being separated from said long element. The upper part of the long element that remains after cutting is again lowered inside the container over the entire height thereof and cutting of the long element into fragments is repeated until the long element is completely divided. There is also a device for dividing long elements of a nuclear reactor into fragments. The invention makes it possible to reduce the laboriousness of and timeframes involved in carrying out work to divide long elements, and also to minimize the exposure of service personnel radiation.

IPC Classes  ?

• B26D 5/00 - Arrangements for operating and controlling machines or devices for cutting, cutting-out, stamping-out, punching, perforating, or severing by means other than cutting
• G21F 9/28 - Treating solids

Abstract

The invention relates to the field of nuclear technology. A method for dividing long elements of a nuclear reactor into fragments comprises placing the long elements inside a container and subsequently cutting said long elements. A long element is lowered into the container over the entire height thereof. The long element is cut level with the upper edge of the container, with a fragment equal to the height of the container being separated from said long element. The upper part of the long element that remains after cutting is again lowered inside the container over the entire height thereof and cutting of the long element into fragments is repeated until the long element is completely divided. There is also a device for dividing long elements of a nuclear reactor into fragments. The invention makes it possible to reduce the laboriousness of and timeframes involved in carrying out work to divide long elements, and also to minimize the exposure of service personnel radiation.

IPC Classes  ?

• G21F 9/28 - Treating solids
• B26D 5/00 - Arrangements for operating and controlling machines or devices for cutting, cutting-out, stamping-out, punching, perforating, or severing by means other than cutting

Abstract

The invention relates to devices for ridding radioactive waste of radioactive contamination. An assembly for the electrochemical decontamination of metallic radioactive waste includes a pipeline equipped with a shutoff valve, and a metallic radioactive waste treatment module comprising a unit for the electrochemical decontamination of metallic radioactive waste which is connected by a ventilation channel to a ventilation module and by a decontaminating solution supply and discharge pipeline equipped with a shutoff valve to a module for receiving decontaminating solution. The installation is equipped with a module for preparing decontaminating solution connected by the decontaminating solution supply and discharge pipeline, which is equipped with at least one pump, to the unit for the electrochemical decontamination of metallic radioactive waste and to the module for receiving decontaminating solution. The module for receiving decontaminating solution is provided with purification and pH correction devices for the decontaminating solution, and the unit for the electrochemical decontamination of metallic radioactive waste, the module for receiving decontaminating solution and the module for preparing decontaminating solution are equipped with elements for measuring the pH level. The invention makes it possible to provide adaptive reprocessing of decontaminating solution for repeat use.

IPC Classes  ?

• G21F 9/34 - Disposal of solid waste

Abstract

The invention relates to devices for ridding radioactive waste of radioactive contamination. An assembly for the electrochemical decontamination of metallic radioactive waste includes a pipeline equipped with a shutoff valve, and a metallic radioactive waste treatment module comprising a unit for the electrochemical decontamination of metallic radioactive waste which is connected by a ventilation channel to a ventilation module and by a decontaminating solution supply and discharge pipeline equipped with a shutoff valve to a module for receiving decontaminating solution. The installation is equipped with a module for preparing decontaminating solution connected by the decontaminating solution supply and discharge pipeline, which is equipped with at least one pump, to the unit for the electrochemical decontamination of metallic radioactive waste and to the module for receiving decontaminating solution. The module for receiving decontaminating solution is provided with purification and pH correction devices for the decontaminating solution, and the unit for the electrochemical decontamination of metallic radioactive waste, the module for receiving decontaminating solution and the module for preparing decontaminating solution are equipped with elements for measuring the pH level. The invention makes it possible to provide adaptive reprocessing of decontaminating solution for repeat use.

IPC Classes  ?

• G21F 9/34 - Disposal of solid waste

Abstract

The invention relates to compact mobile robotic fire-extinguishing systems for monitoring emergency situations and carrying out disaster response operations in a hazard zone under particularly dangerous conditions and/or in inaccessible locations. A compact mobile robotic fire-extinguishing system allows fire extinguishants to be delivered at a rate of from 15 to 20 l/s. The present compact mobile robotic fire-extinguishing system comprises a tracked chassis, a body mounted on said chassis, a control unit and a drive part with a rechargeable battery, which are mounted inside the body, a fire extinguishant delivery element mounted on the body such as to be remotely controllable, a video surveillance system, and also a connecting pipe, the outlet of which is connected to the inlet of the fire extinguishant delivery element. The connecting pipe is mounted inside the body with its central axis situated in the vertical plane of symmetry of the body; the axis of the inlet of the connecting pipe is situated lower than the axis of the outlet of the fire extinguishant delivery element by a distance of at least half of the height of the compact mobile robotic fire-extinguishing system, and the outlet of the connecting pipe is situated in the centre of the top surface of the body.

IPC Classes  ?

• A62C 27/00 - Fire-fighting land vehicles
• A62C 31/03 - Nozzles specially adapted for fire-extinguishing adjustable, e.g. from spray to jet or vice-versa
• A62C 37/00 - Control of fire-fighting equipment

Abstract

The invention relates to nuclear power engineering. A unit for reprocessing radioactive waste comprises a waste loading assembly, a plasma shaft furnace, a device for feeding air into the furnace and a pyrogas combustion chamber, an evaporative heat exchanger, a gas purification system comprising a bag filter, a heat-exchanger device and a scrubber, pumps and containers for reagents and reprocessing products, and fittings. The unit is equipped with at least one control module and a monitoring module of a slag drainage assembly, said monitoring module being electrically connected to said control module, a module for monitoring the state of the internal environment, a module for the state of equipment, and at least one gas-analyser module, a plasma shaft furnace, a cell for receiving slag melt, and electrical equipment of the device for feeding air into the furnace and into the pyrogas combustion chamber. The control module is electrically connected to the electrical equipment of the waste loading assembly. The monitoring module of the slag drainage assembly is electrically connected to the electrical equipment of the slag drainage assembly. The invention makes it possible to provide an adaptive operating mode of the unit in which radioactive waste of varying radioactive levels is reprocessed in automatic mode.

IPC Classes  ?

• G21F 9/32 - Processing by incineration

Abstract

The invention relates to emergency protection of nuclear power plants, particularly to technologies for mitigation of consequences or fire prevention and prevention of explosive gas accumulation, which ensure hydrogen explosion safety in premises of the containment dome (hereinafter - CNT) at nuclear power plants (hereinafter - NPP) with water-cooled power reactor (hereinafter - VVER). Moreover, the offered invention can be used at other facilities having the risk of development of potentially hazardous emergency processes, related to the emission of a large amount of light combustible gas and its localization in closed premises of a facility. The technical result attained by the declared invention consists in the reduction of the risk of gas medium inflammation in premises of NPP CNT, as well as in the provision of self-damping of weak burning waves, reduction of their intensity in case of gas medium inflammation in premises of NPP CNT and in reduction of dynamic loads on walls of premises of NPP CNT. The specified technical result is attained due to the fact that in the method for ensuring hydrogen explosion safety at nuclear power plants, comprising ventilation of premises of the nuclear reactor premises and hydrogen recombination in premises of the nuclear reactor premises by its catalytic oxidation, in accordance with the declared solution, a reflector (2, 3) is placed on the way of potentially emergency propagation of a pressure hydrogen-containing steam-gas jet, apertures are made in the walls between premises of the nuclear reactor containment dome with a size equal to minimum 35% of the surface area of the said walls, while excess heat is withdrawn in areas of potential localization of hydrogen-containing steam-gas mixture burning sources.

IPC Classes  ?

• G21C 9/06 - Means for preventing accumulation of explosives gases, e.g. recombiners
• G21C 13/02 - Pressure vesselsContainment vesselsContainment in general Details
• G21C 19/317 - Recombination devices for radiolytic dissociation products

Abstract

ABSTRACTThere is disclosed herein a radioactive waste recycling plant including a waste feed unit, a plasma shaft-type furnace with a melter in a hearth of the furnace and a slug discharge unit connected with a receiving tank for molten slug; an air supply unit; an evaporative heat exchanger; a gas purification unit; a heat-exchanger and a scrubber; pumps and tanks for agents and recycled products; fittings; one control module; an interior environment control module; an equipment status control module; and, a gas analytical module; the control module is electrically connected to a waste feed unit, a plasma shaft-type furnace, a receiving tank and the air supply unit which delivers the air to the furnace and to the combustion chamber.14Date Recue/Date Received 2020-09-22

IPC Classes  ?

• G21F 9/32 - Processing by incineration

Abstract

The invention relates to emergency protection of nuclear power plants, particularly to technologies for mitigation of consequences or fire prevention and prevention of explosive gas accumulation, which ensure hydrogen explosion safety in premises of the containment dome (hereinafter - CNT) at nuclear power plants (hereinafter - NPP) with water-cooled power reactor (hereinafter - VVER). Moreover, the offered invention can be used at other facilities having the risk of development of potentially hazardous emergency processes, related to the emission of a large amount of light combustible gas and its localization in closed premises of a facility. The technical result attained by the declared invention consists in the reduction of the risk of gas medium inflammation in premises of NPP CNT, as well as in the provision of self-damping of weak burning waves, reduction of their intensity in case of gas medium inflammation in premises of NPP CNT and in reduction of dynamic loads on walls of premises of NPP CNT. The specified technical result is attained due to the fact that in the method for ensuring hydrogen explosion safety at nuclear power plants, comprising ventilation of premises of the nuclear reactor premises and hydrogen recombination in premises of the nuclear reactor premises by its catalytic oxidation, in accordance with the declared solution, a reflector is placed on the way of potentially emergency propagation of a pressure hydrogen-containing steam-gas jet, apertures are made in the walls between premises of the nuclear reactor containment dome with a size equal to minimum 35% of the surface area of the said walls, while excess heat is withdrawn in areas of potential localization of hydrogen-containing steam-gas mixture burning sources.

IPC Classes  ?

• G21D 3/04 - Safety arrangements

Abstract

The invention relates to construction, and more particularly to stabilizing soils for bases and foundations of buildings and structures. A method for compacting bases laid using weak mineral soils includes drilling holes, feeding a packing material into each hole, and acting on the packing material with a working tool in order to form a compaction pile. First, a geotechnical survey is carried out with respect to the area of a base, and desired parameters are determined. A target modulus of deformation of the soil layer to be compacted is set. Comparative calculations of the results obtained are carried out. The placement interval of the compaction piles is set as equal to three diameters of a hollow tubular working tool, and the value of the actual mean effective modulus of deformation of the base is determined. The latter is compared with the target modulus of deformation of the mineral soil. Additional geotechnical surveys are carried out with respect to the area of the base. The actual mean effective modulus of deformation of the compacted base is calculated and compared with the target value. If the actual mean effective modulus of deformation of the base does not correspond to the target value, additional compaction piles are installed between the existing piles. The technical result is an increase in soil compacting efficiency, and a reduction in material and labour costs.

IPC Classes  ?

• E02D 27/26 - Compacting soil locally before forming foundationsConstruction of foundation structures by forcing binding substances into gravel fillings

Abstract

There is disclosed herein methods for strengthening soils under bases and foundations of buildings and structures, including electric power industry facilities. The methods include compaction of weak mineral soils that includes formation of a well, filling each well with the compacting material, and creation of a compacting effect on the compacting material by the hollow tubular working tool to form a ground pile, proposes to perform preliminary geological engineering surveys of the base area and determine the values of the modulus of deformation.

IPC Classes  ?

• E02D 3/12 - Consolidating by placing solidifying or pore-filling substances in the soil
• E02D 27/26 - Compacting soil locally before forming foundationsConstruction of foundation structures by forcing binding substances into gravel fillings

Abstract

The inventions relate to fire-fighting ground vehicles. A mobile fire extinguishing apparatus with pressurized foam generation comprises: a mixing chamber connected at the outlet to a device for supplying foam to the heart of a fire; and, connected by pipes to an inlet of the mixing chamber, a water supply system comprising a water supply pump with a drive, a foam-forming concentrate supply system comprising a foam pump with a drive, and also an air supply system comprising an air compressor with a drive. The apparatus is provided with a driving motor. The drives of the compressor and of the foam pump are in the form of an adjustable hydraulic transmission of the air compressor drive and an adjustable hydraulic transmission of the foam pump drive, which are kinematically connected to the driving motor. The apparatus is also provided with a water meter, a gate valve having an electric drive, and a non-return valve, which are mounted on a water supply pipe between the water supply pump and the mixing chamber, and is further provided with an electronic control unit for the gate valve. An input of the gate valve control unit is electrically connected to an output of the water meter, and an output of the gate valve control unit is electrically connected to an input of the electric drive of the gate valve. This makes it possible to produce foam having the necessary delivery pressure and density under any pump operating modes.

IPC Classes  ?

• A62C 27/00 - Fire-fighting land vehicles
• A62C 5/02 - Making of fire-extinguishing materials immediately before use of foam

Abstract

The inventions relate to fire-fighting ground vehicles. A mobile lire extinguishing apparatus with pressurized foam generation comprises: a mixing chamber connected at the outlet to a device for supplying foam to the heart of a fire; and, connected by pipes to an inlet of the mixing chamber, a water supply system comprising a water supply pump with a drive, a foam-forming concentrate supply system comprising a foam pump with a drive, and also an air supply system comprising an air compressor with a drive. The apparatus is provided with a driving motor. The drives of the compressor and of the foam pump are in the form of an adjustable hydraulic transmission of the air compressor drive and an adjustable hydraulic transmission of the foam pump drive, which are kinematically connected to the driving motor. The apparatus is also provided with a water meter, a gate valve having an electric drive, and a non-return valve, which are mounted on a water supply pipe between the water supply pump and the mixing chamber, and is further provided with an electronic control unit for the gate valve. An input of the gate valve control unit is electrically connected to an output of the water meter, and an output of the gate valve control unit is electrically connected to an input of the electric drive of the gate valve. This makes it possible to produce foam having the necessary delivery pressure and density under any pump operating modes.

IPC Classes  ?

• A62C 5/02 - Making of fire-extinguishing materials immediately before use of foam
• A62C 27/00 - Fire-fighting land vehicles

Abstract

The invention relates to the comprehensive treatment of complex boron-containing liquid waste from a nuclear power plant. A method for treating liquid waste from a nuclear power plant with boron control includes introducing calcium nitrate into a borate solution with precipitation of calcium borate and the removal of same from the mother liquor, performing electrodialysis using an electrodialysis cell having a cation exchange membrane and an anion exchange membrane, and obtaining boric acid and a solution of the hydroxides of sodium and potassium. The initial borate solution contains salts of sodium and potassium in the form of their nitrates and sulphates. Calcium nitrate is introduced into the initial borate solution, producing a joint precipitate of borate and calcium sulphate. Boric acid is obtained by treating the joint precipitate of borate and calcium sulphate with a solution of nitric acid with separation of a calcium sulphate precipitate from a solution of calcium borate. The calcium borate solution is treated with nitric acid, forming a precipitate of boric acid and a solution of calcium nitrate. The boric acid precipitate is separated and dried. The mother liquor is subjected to electrodialysis to produce solutions of nitric acid and of the hydroxides of sodium and potassium.

IPC Classes  ?

• G21F 9/06 - Processing
• G21F 9/20 - Disposal of liquid waste

Abstract

The invention relates to the comprehensive treatment of complex boron-containing liquid waste from a nuclear power plant. A method for treating liquid waste from a nuclear power plant with boron control includes introducing calcium nitrate into a borate solution with precipitation of calcium borate and the removal of same from the mother liquor, performing electrodialysis using an electrodialysis cell having a cation exchange membrane and an anion exchange membrane, and obtaining boric acid and a solution of the hydroxides of sodium and potassium. The initial borate solution contains salts of sodium and potassium in the form of their nitrates and sulphates. Calcium nitrate is introduced into the initial borate solution, producing a joint precipitate of borate and calcium sulphate. Boric acid is obtained by treating the joint precipitate of borate and calcium sulphate with a solution of nitric acid with separation of a calcium sulphate precipitate from a solution of calcium borate. The calcium borate solution is treated with nitric acid, forming a precipitate of boric acid and a solution of calcium nitrate. The boric acid precipitate is separated and dried. The mother liquor is subjected to electrodialysis to produce solutions of nitric acid and of the hydroxides of sodium and potassium.

IPC Classes  ?

• G21F 9/06 - Processing
• G21F 9/20 - Disposal of liquid waste

Abstract

The invention relates to the field of nuclear engineering and can be used as part of the fuel-handling equipment of a nuclear reactor. The technical problem addressed by the present invention is that of providing linearity of movement for a cable when an angle of inclination of an inclined shaft in an inclined elevator is increased. The solution to this problem allows a reduction in the dimensions of openings in the walls of a building of a reactor compartment and a spent fuel pool, through which a trolley passes. The present inclined elevator of a nuclear reactor comprises a trolley (1) which has a sheath (2) for a spent fuel assembly and which travels along a track (3) in an inclined shaft (4) with the aid of a cable (8). A guide pulley provided with an asymmetric groove for the cable to pass through is mounted at the entrance to the inclined shaft (4), and is mounted relative to the track at an angle β that satisfies the condition tg β= t/πD, where: β is the angle of inclination of the guide pulley relative to the track; t is the pitch of the channels for the cable on a drum; and D is the diameter of the drum, the axis of symmetry of the groove in the guide pulley lying along the motion trajectory of the cable.

IPC Classes  ?

• G21C 19/32 - Apparatus for removing radioactive objects or materials from the reactor discharge area, e.g. to a storage placeApparatus for handling radioactive objects or materials within a storage place or removing them therefrom
• G21F 9/34 - Disposal of solid waste

Abstract

The invention relates to the field of nuclear engineering and can be used as part of the fuel-handling equipment of a nuclear reactor. The technical problem consists in producing a track for an inclined elevator which, by virtue of having a shorter length, allows a reduction in the time that a trolley carrying spent fuel assemblies spends in a gaseous atmosphere. The solution to this problem provides for increased safety when conveying spent fuel assemblies along an inclined elevator of a nuclear reactor. The essence of the invention is that the track of the inclined elevator of a nuclear reactor comprises alternate rectilinear and curvilinear sections, wherein the initial section and the end section are rectilinear and are arranged at the same angle of inclination α relative to a horizontal plane.

IPC Classes  ?

• E01B 23/02 - Tracks for light railways, e.g. for field, colliery, or mine use
• G21F 9/34 - Disposal of solid waste