Devices or arrangements for monitoring or testing fuel or fuel elements outside the reactor core, e.g. for burn-up, for contamination (G21C17/06)

G   Physics(394526)
G21C17        onitoring; testing(571)
G21C17/06                     Devices or arrangements for monitoring or testing fuel or fuel elements outside the reactor core, e.g. for burn-up, for contamination (g21c0017080000, g21c0017100000 take precedence;detecting leaking fuel elements during reactor operation g21c0017040000)(152)

Device for detecting defects on the end surface of cylindrical products // 2645436
FIELD: monitoring systems.SUBSTANCE: invention relates to devices for monitoring the surface of cylindrical objects and, in particular, can be used in the nuclear fuel production in monitoring the appearance of the end face of fuel pellets. Device comprises two end-product inspection units sequentially installed on the conveyors, two product flow separation units installed one at a time in front of each end-product inspection unit, as well as two defective product blow-off assemblies installed after each end-product inspection unit. Each flow separation unit comprises means for longitudinal feeding the products to the conveyor one at a time with certain intervals. Each end-product inspection unit comprises an optical sensor for detecting products, n illumination means for the products to be monitored, a means for generating visible spectrum radiation, and a means for recording and transferring images of the product end surfaces to the analytical device. Each defective product blow-off unit comprises an optical sensor for detecting products and a blow-off means to form a directed air flow.EFFECT: technical result consists in automated, operational, high-reliable, careful and human error excluding control of cylindrical objects for availability and character of surface defects, high efficiency of process operation of the control.7 cl, 4 dwg

Device for express control of uranium enrichment in powders // 2645307
FIELD: power industry.SUBSTANCE: device for express-control of uranium enrichment in powders comprises a tank located above the scintillation gamma-radiation detector connected to the unit for controlling and processing measurement results. The device is provided with a unit for protection against the background, which is made in the form of a lead cylinder and placed in a steel frame with the possibility of end-to-end output of the cables to the unit for controlling and processing measurement results. The unit for controlling and processing measurement results is made in the form of a computer with a pulse signal processor.EFFECT: invention allows to provide a fast technique for controlling the enrichment of 235U in powders of uranium oxides at an arbitrary degree of radiation equilibrium disturbance based on the application of a scintillation detector easily adaptable to production conditions.4 cl, 4 dwg
Device for continuous controlling enrichment and content of gadolinium oxide in nuclear fuel moulding powder when filling it in fuel tablet pressing device // 2629371
FIELD: physics.SUBSTANCE: device for controlling enrichment of U235, as well as content of Gd2O3 in the nuclear fuel moulding powder, containing a hopper connected by a filling tube to the tablet pressing device, a detector of its own gamma radiation located in close proximity to the side wall of the filling tube. The device is equipped with an industrial computer containing a spectrometer, which allows obtaining and analyzing the spectral characteristic of the registered gamma radiation of the uranium moulding powder containing gadolinium oxide.EFFECT: operational control of uranium enrichment 235, as well as the content of gadolinium oxide, taking into account the density of the measured sample.3 cl, 3 dwg

ethod for continuous maintenance of stability of measurements of spectrometer channel for controlling uniformity of distribution of fuel in fuel element by gamma-absorption method // 2603351
FIELD: nuclear industry.SUBSTANCE: invention relates to nuclear industry and can be used for controlling uniformity of distribution of fuel in fuel elements by gamma-absorption method using a scintillation spectrometer. Method for continuous maintenance of stability of measurements of a spectrometer channel comprises detecting flux density of gamma-radiation from an external source, passed through limited areas of holders on a standard sample of fuel element and on controlled fuel element during movement thereof along longitudinal axis of standard sample/fuel element. Measured flux density of gamma-radiation is converted by means of spectrometer into sequence of electric pulses and value of pulse counting rate on holders and on fuel column of fuel element in each point of PTS spectrum. Method includes determining correction factor, calculating and recording value of reduced count rate PTSpriv for fuel column of fuel element in each point of spectrum.EFFECT: technical result is providing automatic adjustment of readings of spectrometer channel by taking into account background gamma-radiation.1 cl, 2 dwg, 2 tbl

Apparatus for controlling characteristics of fuel column annular fuel element // 2603017
FIELD: energy.SUBSTANCE: invention relates to nuclear power engineering and can be used in making annular fuel elements of nuclear reactors. Apparatus for controlling characteristics of fuel column of annular fuel elements comprises a row of blocks 1-4 for detection of gamma-radiation of fuel column and blocks 5, 6 for detection of gamma-radiation passing through fuel column. Source 13 of gamma-radiation is fixed on end of bar 12, intended for introduction into cavity of fuel element 9. Mechanism for moving fuel element is configured for translational movement of fuel element 9 along its axis and includes mechanism 8 for gripping and turning fuel element 9 about its axis by 90 degrees. Two blocks 5, 6 for detection of gamma-radiation are located on opposite sides of fuel element displacement axis 9. Control unit is connected to detection blocks and to displacement mechanism of fuel element 9.EFFECT: technical result is possibility, in one pass of annular fuel element, of obtaining all necessary quality characteristics of its production.1 cl, 3 dwg

Fuel element simulator for adjustment of devices measuring boundaries of various media in fuel column // 2594566
FIELD: nuclear energy.SUBSTANCE: invention relates to nuclear power engineering and can be used in production of fuel elements for nuclear reactor fuel assemblies. Fuel element (fuel element) simulator for adjustment of devices used for determining boundaries of different media in fuel column of fuel element is made in the form of a plate, which includes series-arranged sections of various thickness complying with order of arrangement and the number of media in fuel column of fuel element, wherein the thickness of each section is selected from the condition: x=µcpxcp / µimit, where x is a thickness of this plate section; xcp is a thickness of corresponding medium in fuel element; µcp - linear gamma-quanta attenuation coefficient for corresponding medium in fuel element; µimit - linear coefficient of attenuation of gamma-quanta for plate material. Simulator has unlimited service life, is not subjected to wear and does not require special measures for storage, since it does not contain fissile material.EFFECT: technical result consists in the fact that fuel element simulator enables to completely replace standard samples made in the form of real fuel elements.1 cl, 1 dwg

Electrical fluid heating device, method of its production and application for electric simulating nuclear fuel rods // 2587980
FIELD: heating.SUBSTANCE: invention relates to electric heaters, preferably using a electric simulation of nuclear fuel rods to be connected in assembly in power reactors. Device (1) for heating of liquid (Liq.) with increased heat flow includes tubular resistor (2) fed with direct current to heat the liquid due to heat conductivity through embraced by direct mechanical contact electrically insulating and heat conducting intermediate element (6, 22), the complex tubular resistor/intermediate element is surrounded by a cover (7) intended for submersion into a heated liquid, at least part of its length.EFFECT: device provides creation of uniform heat flow, reliable in operation and has long service life.20 cl, 3 dwg

ethod of testing for compatibility of nuclear fuel powder with material of fuel element cladding // 2581846
FIELD: engines.SUBSTANCE: invention relates to methods of determining compatibility of various types of nuclear fuel and structural materials. Method of testing for compatibility of nuclear fuel powder with fuel rod cladding material comprises diffusion annealing of nuclear fuel powder and fuel rod cladding pair. From material of fuel rod cladding is made crucible with a polished inner surface and lid, after which it is moulded into a powder with test fuel and fission products simulators and sealing crucible in an inert gas atmosphere, followed by annealing in temperature range of 600-1,000 °C. Testing is carried out using uranium alloy powders or uranium mononitride with particle size of 10-20 mcm. To produce crucible and cover method uses corrosion-resistant steel or zirconium alloys, and as imitators of chemically active fission products, iodine and/or caesium and/or tellurium.EFFECT: technical result is reliable contact (adhesion) of fuel and structural materials, which increases reliability and information value of diffusion tests.8 cl, 3 dwg

Device for continuous monitoring of density of pressed powder of nuclear fuel during its loading in device for fuel tablets pressing // 2572241
FIELD: power industry.SUBSTANCE: device contains a silo 1 with pressed powder, connected by vertical loading pipe 2 with tablets pressing device 3. Near the pressing device 3 at opposite sides of the pipe 2 the gamma source 4 is installed (at sufficient quantity of pressed powder for registration its own gamma radiation can be used), and detection unit 5, connected with registration unit comprising signal transducer 6, and electronic graphic register 7, connected in series with communication lines 8. The registration unit is made with possibility to communicate signals to control system of the pressing device to regulate pressed powder delivery or for its shutdown.EFFECT: timely monitoring of density decreasing of pressed powder and pressing termination to exclude production of the fuel tablets with wrong geometry and reduced density.2 cl, 3 dwg

Hydrogen igniter and reactor plant having said igniter // 2554115
FIELD: chemistry.SUBSTANCE: invention relates to igniting hydrogen which is part of a gas medium. The igniter consists of a housing having openings for inlet and outlet of the gas medium, and filler in the form of bismuth oxide Bi2O3 and/or lead oxide, placed in the housing. The igniter can be used in a nuclear reactor plant.EFFECT: obtaining a hydrogen igniter which does not contaminate the gas medium, particularly reactor cover gas, with impurities which are hazardous for installation components and/or coolant, for example, lead-bismuth coolant; removing, from the gas medium passing through the igniter, steam formed as a result of igniting hydrogen.14 cl, 1 dwg

Fuel element test method // 2552839
FIELD: heating.SUBSTANCE: method involves determination of helium pressure under cover (9) of a fuel element after its sealing, at which fuel element (1) is kept in a measurement position during the whole test period; local pulse heating of the fuel element is performed in area (4) of compensation volume; time dependence of temperature of cover sections at heating point (10) and in section (12) of the cover, which is remote from the heating point, is recorded during the whole test period. Then, helium pressure and state of the fuel element is evaluated based on it. Prior to local heating throughout the perimeter of the cover part in the compensation volume area a provision is made for elimination of heat transfer. The remote section is chosen on the other side of the compensation volume area; after that, the fuel element is exposed till its temperature is equalised with ambient temperature. Then, ambient temperature below 0°C is created; prior to local heating, the fuel element is kept till its temperature is equalised with new ambient temperature; a heating and measurement cycle is repeated with exclusion of heat transfer along the cover body from the heating point to the remote section.EFFECT: possible testing of a fuel element on one side of a cover.1 dwg

Heat-producing element monitoring method // 2552526
FIELD: power industry.SUBSTANCE: invention relates to control devices of heat-producing elements (fuel elements). The method includes determination of helium pressure under the shell (11) of the heat-producing element after its sealing at which the heat-producing element (1) is sustained at the measurement position, the local pulse heating of the heat-producing element is performed in the field of the compensation volume (8), the time dependence of temperature of shell sections in the place of heating (10) is registered and on the opposite side of the shell, it is used for estimation of helium pressure and the state of the heat-producing element.. Before local heating the heat-producing element is held until equalization of its temperature with ambient temperature, and after completion of monitoring the ambient temperature below 0°C is formed, before the local heating the heat-producing element is held until equalization of its temperature with new ambient temperature, then the cycle heating-measurement is repeated and the obtained time dependences of pressure at different temperatures are compared with calibration dependences for different helium pressure and different levels of the content of air in it.EFFECT: providing additional possibility of non-destructive control of heat-producing elements.1 dwg

ethod to test zirconium alloys in steam and water medium // 2550347
FIELD: testing equipment.SUBSTANCE: in the method in process of exposure of samples of zirconium alloys in the steam and water medium in the temperature range of the light water reactor core they develop a gas discharge plasma in water vapours, afterwards they radiate samples by positively charged hydrogen ions by means of applying of negative electric potential to them relative to the plasma.EFFECT: approximation of testing conditions of samples of zirconium alloys in steam and water medium to conditions of light water reactor core, which makes it possible to increase validity of predicted picture of behaviour of investigated zirconium alloys in light water reactor core in process of its operation made on the basis of results of these tests.3 cl, 1 dwg

Apparatus for automated inspection of surface and volume defects of ceramic nuclear fuel // 2548564
FIELD: physics, atomic power.SUBSTANCE: invention relates to means of inspecting nuclear fuel in the form of cylindrical tablets. The apparatus for automated inspection of surface and volume defects of ceramic nuclear fuel comprises an optical image transformer, optical and thermal image recording channels, illumination sources, a system for inputting pulsed thermal flux into the inspected article and a selector which provides synchronous recording of both optical and thermal images.EFFECT: obtaining reliable results on presence or absence of defects in inspected articles and, as a result, reliable selection of defective and non-defective articles.7 cl, 6 dwg

Ampoule device for in-reactor analysis // 2526328
FIELD: power engineering.SUBSTANCE: device comprises shell with sealing end covers to house at least one capsule with analysed specimens fitted in unsealed thin-wall shell of refractory material. Said capsule is connected with gas lines intended for streaming ventilation of capsule working space. Outlet of every line is plugged for capsule sealing, plugs being composed of sleeves with axial holes filled with fusible material. One of the lines houses thermometer transducers. Note here that sensor of every transducer is fitted inside capsule working space.EFFECT: measurement of temperatures of emissions at nuclear disintegration during experiments, simplified design of capsule seals.4 cl, 1 dwg

Nuclear reactor fuel element simulator // 2523423
FIELD: physics.SUBSTANCE: fuel element simulator has a shell in which there is a column of natural fuel tablets with a centre hole, and an electric heater placed with clearance in the holes of the tablets. The heater is in form of pipe made of heat-resistant material on the outer surface of which is formed a microrelief which varies on the length of the heater and which provides optically variable properties on the length of the surface, which correspond to the simulated temperature profile. A shielding pipe made of heat-resistant material is also placed with clearance on the outside coaxial to the shell, the inner and outer surfaces of said pipe also having a varying microrelief which provides optically variable properties on the length of the heater.EFFECT: high accuracy of simulating the thermal state of fuel elements under investigation by obtaining temperature levels, thermal flux and temperature profiles similar to those in full-scale conditions.7 cl, 2 dwg

Device to measure and correct deviation from parallelism into rods for nuclear fuel // 2507473
FIELD: power engineering.SUBSTANCE: device arranged on a stand (4), comprises a place (31) with a horizontal axis (X) for placement of the above fuel rod; a facility (20) for measurement of deviation from parallelism and a facility (22) for correction of the above deviation. The device comprises a facility (14) of device positioning relative to the fuel rod comprising two parallel supports arranged at the distance from each other, at the same time each of them supports the end of the above fuel rod. The supports are made in the form of two horseshoe-shaped parts (16.1. 16.2), the inner ends of which are designed for resting against the fuel rod, and are distanced from each other at the specified distance to ensure the coverage of the stand support, at which the end rests with the upper plug of the fuel rod, and which has thickness that is substantially equal to the distance between two horseshoe-shaped parts (16.1, 16.2). Also the device comprises a facility (32) to retain a fuel rod made as capable of providing for rotation of the fuel rod around its longitudinal axis, which is arranged between the facility (14) of positioning and facilities of measurement and correction. The facility (32) comprises a lower grip (34) and an upper grip (36), to hold the fuel rod, at the same time the lower grip (34) forms a base for measurement of deviation from parallelism.EFFECT: provision of measurement of deviation from parallelism during correction of the above deviation.12 cl, 15 dwg

Test method of materials in nuclear reactor // 2494480
FIELD: power industry.SUBSTANCE: specimen is made of two coaxially combined tubular elements; one of which is fully or partially located inside the other one; gas pressure is created in a cavity between elements, sealed, arranged in a nuclear reactor and irradiated.EFFECT: increasing informativity and reliability of results of change of properties of reactor materials at irradiation in the reactor at various types of stress-and-strain state.3 cl, 1 dwg

ethod to measure doppler coefficient of reactivity // 2491664
FIELD: power engineering.SUBSTANCE: time-series data by reactivity is produced from time-series data by a neutron bundle by the method of reverse dynamic characteristic in respect to a single-point kinetic equation of the reactor. Time-series data by fuel temperature exposed to previously determined averaging is produced using time-series data by power output of the reactor and pre-determined dynamic model. The component of contribution to feedback by reactivity is determined using time-series data by reactivity and introduced reactivity. The Doppler coefficient of reactivity is determined using the received time-series data by average temperature of a moderator in the reactor, time-series data by fuel temperature exposed to previously determined averaging, isothermic temperature coefficient of reactivity and component of contribution to feedback by reactivity.EFFECT: increased accuracy and simplicity of measurements of the Doppler coefficient and possibility of its usage in case of use of discrete data.8 cl, 7 dwg

Nuclear fuel pellet density monitoring plant // 2458416
FIELD: power industry.SUBSTANCE: nuclear fuel pellet density monitoring plant includes measuring unit including gamma radiation source and detection unit, transfer mechanism for movement of pellets and hold-down device, as well as measuring result control and processing unit intended to control the operation of transfer mechanism for processing of measuring results and rejection of pellets. Transfer mechanism includes the first transfer assembly for movement of column of pellets through measuring assembly with reference to outlet pallet, the second transfer assembly for movement of reference and outlet pallet for columns of pellets in transverse direction, and hold-down device has the possibility of pressing the pellets during movement of column of pellets through the measuring unit.EFFECT: invention allows increasing the monitoring efficiency due to supply to monitoring zone of nuclear fuel pellets in the form of columns and performance of measurement during movement of columns through the monitoring zone.2 cl, 1 dwg

ethod of creep-rupture test of tubular samples in non-instrumentation channel of nuclear reactor // 2451349
FIELD: power engineering.SUBSTANCE: method of creep-rupture test of tubular samples in a non-instrumentation channel of a nuclear reactor includes the following operations. At least one reference tubular sample loaded with inert gas pressure is placed into a heating furnace, maintained at the preset temperature in the heating furnace until destroyed, and time is measured to the moment of its destruction. Two tubular sample accordingly loaded and non-loaded with inert gas pressure are simultaneously placed into an ampoule. The tight ampoule with both types of tubular samples is radiated in a nuclear reactor channel. The radiated tubular samples are placed into a heating furnace and tested until destroyed under pressures and temperatures similar to the ones in the reactor. The time is measured to the moment of destruction of tubular samples of the first and second types in the heating furnace. The time to the moment of tubular sample destruction under conditions of reactor radiation at the preset pressure and temperature is determined using the ratio that takes into account time values measured in process of method realisation.EFFECT: invention makes it possible to increase accuracy of detection of strength characteristics of materials.2 cl

Device for pelletising of nuclear fuel and method to manufacture nuclear fuel pellets with application of this device // 2414760
FIELD: power engineering.SUBSTANCE: device to pelletise nuclear fuel comprises press, conveyor (4) for transportation of pellets from press to sintering area, facility (26) of pellets reloading from press to conveyor (4) and facility of inspection of at least one pellet of nuclear fuel at the outlet of press, besides, facility of inspection comprises facility for detection of matrix, where each pellet is made. Method to manufacture pellets of nuclear fuel with application of device, which includes stages, when matrices (10) are filled with powder, powder is pressed, pellets (P) are reloaded to conveyor (4), conveyor (4) is started, pellet (P) is taken, manufactured in certain matrix (10), proper operation of this matrix is inspected by results of inspection of pellets manufactured in it, pellets (P) are transported to sintering area.EFFECT: control of manufactured pellets density, control of pellets without increasing duration of production cycle.24 cl, 4 dwg

Control method of gas pressure in fuel element of nuclear reactor // 2408098
FIELD: power industry.SUBSTANCE: control method of gas pressure in fuel element of nuclear reactor consists in the fact that fuel element is located horizontally, inserted in annular induction heater, heat impulse is generated, which induces convective gas current in fuel element, change of temperature is measured with temperature sensors pressed to the cover and gas pressure is calculated on the basis of temperature change value; at that, shoes and couplings are installed on temperature sensors prior to measurements; sensors are pressed to the cover opposite to each other, one is from above, the other is from below, heat-insulating patches are installed between sensors and difference of temperatures shown with sensors is measured, then heat impulse is supplied and difference of temperatures is measured again in certain time τ1; after that, fuel element is turned together with patches, sensors and induction heater through 180° and after it is turned, temperature difference is measured in certain time τ2, then the second heat impulse is supplied and temperature difference is measured again in time τ1; then fuel element is turned together with patches, temperature sensors and induction heater through 180° back to initial position; then temperature difference is measured again in time τ2; cycle is repeated for several times; after that obtained results are mathematically processed, and as a result gas pressure value is determined inside fuel element.EFFECT: improving measurement accuracy of gas pressure inside fuel element.

Device for reading serial numbers of fuel assemblies // 2400840
FIELD: power industry.SUBSTANCE: device contains the first housing with through holes for passage of fuel assemblies (FA), around which illuminators are equally installed. Mirrors receiving the optical radiation reflected from fragments of side FA surface and installed with various turning angles of images provide uniform transfer of reflected mirror images to the plane of openings. The second housing with openings, which is located at some distance from the first one, is provided with radiation protection. Inside housing there arranged are video cameras consisting of video matrixes and objectives, and mirror labyrinths formed with inlet mirrors and outlet mirrors. Inlet mirrors are oriented towards outlet openings, and outlet mirrors - towards the objectives. External image control and processing unit is taken to clean room and connected to video cameras through cable communication lines. Invention is aimed at increasing radiation protection of video cameras owing to their possibility of being compactly arranged in remote housing.EFFECT: radiation protective material and mirror labyrinths in the second housing provide additional radiation protection of video cameras.5 cl, 4 dwg

Control device of gas pressure in fuel element of nuclear reactor // 2399970
FIELD: power industry.SUBSTANCE: invention refers to control devices of gas pressure in fuel element of reactor. Device containing annular induction heater (inductor), temperature sensors located on one side of the heater at the distance close to fuel element diametre on opposite generatrixes of fuel element cover coaxially perpendicular to fuel element axis; in order to improve accuracy characteristics of pressure measurement there additionally introduced are heat-insulation patches between temperature sensors in thermal contact zone; sensors have metal shoes in the form of rectangular copper plates bent along the radius of surface generatrix of fuel element cover, covered with electrically insulating thermally conductive film, and flexible (for example rubber) couplings; there also introduced is the device of turning the fuel element through 180° relative to its longitudinal axis together with inductor, sensors and heat-insulation patches.EFFECT: improving accuracy measurement characteristics of gas pressure inside fuel element.

ethod of controlling mass ratio of uranium-235 isotope in gaseous uranium hexafluoride // 2396613
FIELD: chemistry.SUBSTANCE: method of controlling mass ratio of uranium-235 isotope in gaseous uranium hexafluoride involves desublimation of gaseous uranium hexafluoride in a measuring chamber by lowering temperature of the base of the chamber, determination of gamma-ray intensity of the uranium-235 isotope in the solid phase and calculation of the mass ratio of the uranium-235 isotope in uranium hexafluoride using the formula: C = α*Iγ/M, where: M is mass of uranium hexafluoride in the measuring chamber determined using a mass flowmeter or a weight measuring system, g; Iγ is gamma-ray intensity of uranium-235 in solid uranium hexafluoride in the measuring chamber, s-1; α is a calibration coefficient.EFFECT: higher efficiency and accuracy of determining mass ratio of uranium-235 in gaseous uranium hexafluoride.7 cl

Device for controlling gas gap of graphite-uranium reactor process channel // 2377672
FIELD: nuclear physics.SUBSTANCE: invention relates to operation of graphite-uranium reactors. The device for controlling the gas gap of the process channel of a graphite-uranium reactor has a calibration zirconium pipe fitted on the channel pipe of the process channel. On the outer surface of the pipe there is a block of graphite rings with fixed gaps, and a vertically movable electromagnetic radiation sensor is placed coaxially inside the pipe. The sensor is made in form of two measuring coils, compensated on the surface a uniform conducting medium, and one exciting coil above which there is a short-circuited winding made from non-magnetic current conducting material. The coils are mounted on a permalloy flat-topped magnetic conductor. The device also has a mechanism for moving the sensor and an electronic signal processing unit which is connected to the sensor and a computer. Measuring coils are accordingly connected to the electronic signal processing unit through an amplitude-phase balancing bridge circuit of the sensor, and the exciting coil is connected the electronic signal processing unit through an exciting current stabiliser.EFFECT: more accurate control when measuring gas gaps due to possible readjustment of the sensor in the control zone.3 dwg

Device and method for checking outward appearance of nuclear reactor fuel rods // 2367039
FIELD: nuclear physics.SUBSTANCE: invention relates to checking outward appearance of nuclear reactor fuel rods at the end of a manufacturing cycle. The device for checking outward appearance of nuclear reactor fuel rods has optical apparatus. These apparatus include at least one camera and are connected to an image reading and processing system. This system can detect presence of geometric defects on each examined fuel rod. The device additionally contains a controlled profilometer. The method of checking the outward appearance of nuclear reactor fuel rods involves two stages. Geometric defects are first detected on each examined fuel rod using optical apparatus. Right after detection of a geometrical defect, its depth is measured using the profilometer.EFFECT: possibility of faster checking rods, since there is possibility of determining presence and depth of defects without scanning the entire surface with a profilometer.18 cl, 6 dwg

ethod and device for determination of density and size of object and their use for inspection of nuclear fuel tablets in production process // 2362140
FIELD: physics.SUBSTANCE: invention relates to the area of non-destructive testing methods. Device for automatic density determination of object (100) includes device (2) for determination of significant size x of specified object (100); device (30) for determination of photon beam intensity (I) weakened due to passing through specified object (100); device (200) for data collection, processing and analysis, transportation means (70, 72, 80, 82, 84, 86, 88) for object (100); the first regulator of (74, 76, 78) object position (100); the second regulator of (90, 92, 94, 96, 98) object position (100). Method of the above device application includes calibration stages of device (2) and device (30) and stages of actual significant size measurement for object (100), which are performed for each object (100) in specified batch of objects.EFFECT: increase of measurement accuracy.33 cl, 15 dwg

Device for measurement of geometric dimensions of nuclear reactor fuel elements // 2338276
FIELD: nuclear power engineering.SUBSTANCE: device for measurement of dimensions of nuclear reactor fuel elements is equipped with linear electromechanical drive with unit of automatic measurement of metering frame displacement value. Drive is fixed on the column. Device is equipped with guides with pneumatic drive for orientation of fuel assembly during loading and balloon cylinder mounted in seat-caliper, and device for generation of beams that are parallel to axis of fuel assembly represents laser units, which are installed on the foundation in boxes filled with sand, and are equipped with pendant compensators for automatic retention of beams in vertical position.EFFECT: obtained for measurement of dimensions of nuclear reactor fuel elements.2 dwg

Process and device of spent nuclear fuel burnup detection // 2328043
FIELD: physics.SUBSTANCE: can be applied in burnup control of spent nuclear fuel (SNF) at the facilities storing or operating with SNF, in order to increase efficiency of SNF technological processing cycle due to the optimal configuration. During fuel burnup check fuel assemblies by gamma-ray spectrometric method burnup check process is combined with fuel assembly canisters unloading from transport. At that, fuel assembly canister is fixed, so that fuel assembly core centre is placed at the detection unit axis. Flux of gamma-ray radiation emitted by the whole fuel assembly core is passed through a collimator. Then, the flux of gamma-ray radiation is passed through dissipating filter, and photon gamma-ray radiation spectrum is measured. Peak of total energy of Cesium-137 radioactive nuclide with energy of 662 keV shows Cesium-137 content in uranium. The device for spent nuclear fuel burnup detection includes bridge crane actuator, gamma-ray dissipating filter, while collimator and detection block protection are united in a single protection monoblock. At that, the hole of the collimator is a penetration in the protection monoblock, and bridge crane actuator holds fuel assembly canister steadily against detection block.EFFECT: fast detection of fuel burnup in fuel assemblies on industrial scale; higher efficiency of spent nuclear fuel processing and simpler construction of measurement plant.2 cl, 4 dwg

ethod of remote measurement of fuel element parameters // 2325718
FIELD: physics.SUBSTANCE: said utility invention relates to the instrumentation and may be used for determining parameters of bodies, mainly for remote determination of parameters of radiated fuel elements. According to the invention, for remote measurement of fuel element parameters, an empty grip is weighed and the sample held by the grip is weighed, in the air. After that, the sample and the grip are immersed in the working fluid and weighed after their immersion in the working fluid. The results are used for calculating the initial density of the working fluid. The fuel element held with the grip in the air is weighed. The fuel element with the grip are immersed in the liquid, to various depths, and weighed after each immersion. After that, the partial volume of the fuel element is calculated; the partial volume being the volume contained between the two successive cross sections of the fuel element coinciding with the surface of the working fluid in the vessel at two successive immersion stages; after that, this volume is used for calculating the average area of the fuel element cross section and the full volume of the fuel element.EFFECT: increased accuracy of fuel element parameter determination.1 dwg

ethod for detecting surface flaws on cylindrical pieces of equipment // 2323492
FIELD: nuclear fuel production.SUBSTANCE: proposed method for detecting surface flaws on cylindrical pieces of equipment includes sequential delivery of piece of equipment under inspection to surface inspection position. Butt-end surfaces of piece of equipment delivered to inspection position are illuminated by radiating flux. Radiation detectors receive radiation reflected from butt-end surfaces. Images received from detectors are treated in analyzing device. Side surface of piece of equipment under inspection is illuminated by radiation flux passed at angle φ to normal to its surface. Image reflected at angle to normal equal to incident angle of radiating flux is received. Butt-end surfaces of piece of equipment under inspection are illuminated by radiating flux passed at angle α to normal to butt-end surface. Radiation reflected from butt-end surfaces at angle to normal equal to incident angle of radiating flux is received. Image boundaries of piece-of-equipment surfaces are determined in image frames by means of analyzing device using boundary tracing method. Same method is used to find surface flaw sections on surface images. Surface flaws are described by geometric figures. Surface areas of these figures are calculated. Type of flaws is determined, and decision is taken on fitness of piece of equipment under inspection basing on logic decision rules.EFFECT: enhanced reliability of on-line inspection of cylindrical pieces of equipment for surface flaws and their type.1 cl 4 dwg

ethod for distant nondestructive inspection of irradiated nuclear fuel // 2315378
FIELD: nondestructive inspection of fissionable materials in irradiated nuclear fuel of nuclear reactor fuel assemblies.SUBSTANCE: proposed method includes additional measurements integrated gamma-radiation in fuel assembly of geometry close to that used in recording inherent neutron radiation. Then ratio of neutron count rate (NCR) found from fuel assembly inherent neutron radiation detectors to exposure dose rate (EDR)obtained from integrated gamma-radiation detectors (NCR/EDR) is evaluated, this ratio is used to identify fuel assembly; amount of fissionable material in fuel assembly is determined by inherent neutron radiation.EFFECT: enhanced reliability of irradiated nuclear fuel inspection with respect to inherent neutron radiation, enlarged functional capabilities due to pre-identification of fuel assembly.1 cl, 1 dwg

ethod for checking nuclear fuel rod // 2303302
FIELD: nuclear power engineering; checking nuclear reactor fuel elements.SUBSTANCE: proposed method involves following procedures. Welded joint between plug and fuel rod can incorporating spring is checked for condition by means of electromagnetic induction detector. Excess-energy formed welded joint loosens spring metal structure which can reduce electromagnetic coupling and level of detector-recorded signal. Criterion of fuel rod quality estimate is found by comparing peak values of signal and those of signal on straight-line section of curve.EFFECT: enhanced quality control level of welded joint between plug and can charged with fuel pellets.1 cl, 2 dwg

Nuclear reactor pelletized fuel production process // 2303301
FIELD: nuclear technology; production of pelletized nuclear, mainly uranium-gadolinium, fuel for power reactors.SUBSTANCE: proposed process for producing pelletized fuel for nuclear reactors includes following operations: mixing of uranium and gadolinium oxides, preparation of molding powder, formation of molded parts followed by their sintering. N equal-size samples are taken during each production step. Each sample is placed on support disposed in immediate proximity of magnet surface. Magnet is connected to force-metering device. Readings of the latter are used to evaluate magnetic susceptibility of fuel sample. Chosen sample height is twice as great as that of magnet thickness.EFFECT: enhanced product quality due to enhanced precision of main process control and measurement characteristics, facilitated procedure, simplified design of process hardware.2 cl, 3 dwg

Apparatus for measuring fuel assembly dimensions // 2302675
FIELD: nuclear power engineering; manufacturing and checking fuel assemblies mainly for water-cooled and water-moderated power reactors.SUBSTANCE: proposed apparatus for measuring fuel assembly dimensions has column vertically mounted on base, turnbuckles, seat-gage to receive fuel assembly bottom nozzle, movable measuring frame with two diametrically opposite video cameras which is provided with inductive sensors, contactless sensors responding to measuring frame positioning at points of measurement, their quantity being equal to sum of measuring points for top nozzle, spacer grids, and bottom nozzle, and device generating beams parallel to fuel assembly axis. Sensors are brought to fuel assembly under measurement by means of air cylinders. Device generating beams parallel to fuel assembly axis has gas laser and set of mirrors, Beam generating device is disposed on vibration-damping base. The latter is isolated from column-supporting base as well as from turnbuckles and seat-gage. One of mirrors is semitransparent and is fixed in position at certain angle to vibration-damping base.EFFECT: enhanced quality and reliability of fuel assembly dimensions measurement results.2 cl, 2 dwg

Fuel element simulator (alternatives) // 2273063
FIELD: nuclear power engineering; production and application of nuclear reactor fuel assemblies at nuclear power stations.SUBSTANCE: proposed simulator designed for investigating impact of axial tensile stresses in fuel element claddings and their elongation on formation of fuel assembly as a whole has its core installed in can as far as it will go until stopped by ends of plugs and is built integral with the latter to form single thermomechanical system; cylindrical rods of core are made of material whose coefficient of linear thermal expansion is not lower than that of fuel element cladding. As an alternative, at least two of core cylindrical rods are made of material whose coefficient of linear thermal expansion is not lower than that of nuclear fuel. As still another alternative, cylindrical rods of core are made of material whose coefficient of linear thermal expansion on length equal to fuel pile length equals that of nuclear fuel and on remaining length cylindrical rods of core are made of material whose coefficient of linear thermal expansion equals that of cladding material.EFFECT: reduced cost and enhanced precision of experiment.3 cl, 2 dwg

ethod for evaluating service life of pressurized-tube reactor graphite stack // 2266576
FIELD: nuclear power engineering; evaluating service life of pressurized-tube reactor and its graphite stack.SUBSTANCE: proposed method for evaluating service life of pressurized-tube reactor graphite stack includes step-by-step selective accelerated irradiation of graphite blocks, evaluation of limiting value of fluence as soon as graphite has acquired ultimate strength, and its comparison with fluence of graphite blocks of other reactor subchannels. Graphite blocks are subjected to step-by-step irradiation on running reactor in subchannels with power generation amounting to 90 - 100% of current maximum value attained for reactor; mean power level higher than average for reactor by 20 - 30%, but not higher than maximal admissible level, is maintained in them. Upon termination of each irradiation step strength of graphite blocks of selected subchannels is measured, service life of reactor graphite stack is evaluated as soon as they have reached permissible ultimate strength through margin of fluence by difference between maximum value of graphite block fluence in chosen subchannels measured for graphite brought to ultimate strength and fluence of graphite blocks in remaining subchannels of reactor.EFFECT: enhanced reliability of evaluating graphite stack life and its extension ensuring desired safety in reactor operation.7 cl

ethod for detection of failed fuel elements // 2262757
FIELD: nuclear engineering.SUBSTANCE: the proposed invention is related to the ultrasonic methods of testing for detection of failed fuel elements and may be used when checking the lead-proofness of the fuel elements at spent fuel assemblies, being in water. The proposed methods includes excitation of ultrasonic waves in the fuel element can, detection of the useful reflected and twice reflected signals and determination of the nonleak-proofness when using the amplitude of the useful reflected signal. As a preliminary the ultrasonic waves are excited in the can of an air-tight fuel element or in its simulator by means at a transducer. In this case the wave frequency increases from 0.25 MHz. The value of the frequency of, at which the twice reflected signal is not detected, is determined. Then the ultrasonic waves at the frequency of are excited in the can of a tested fuel element and the conclusion is made on its nonleak-proofness.EFFECT: increased sensitivity and reliability at detection of failed fuel elements.3 cl, 10 dwg

ethod and device for inspecting and grading fuel elements // 2261489
FIELD: nuclear engineering; manufacture of fuel elements for nuclear reactor fuel assemblies.SUBSTANCE: fuel element is detected by sensor on measurement position. In response to sensor signal it is clamped and held in position for local pulsed heating in compensating volume region. Temperature variations are recorded while fuel element is turned in beginning and end of its heating, and during turn intervals. Helium pressure and pressure measurement ranges are given in description of invention. Pressure measurement assembly of inspection device is built of four functional units incorporating pneumatic distributor, primary transducer, induction heating generator, as well as treatment and control module built around industrial computer.EFFECT: enhanced quality and operating reliability of fuel elements due to detection of leaky ones and their isolation.3 cl, 3 dwg

ethod of controlling mass fraction of uranium-235 isotope in gas phase of uranium hexafluoride and measuring system for implementation of the method // 2256963
FIELD: analytical methods in nuclear engineering.SUBSTANCE: invention relates to analysis of fissile materials by radiation techniques and intended for on-line control of uranium hexafluoride concentration in gas streams of isotope-separation uranium processes. Control method comprises measuring, within selected time interval, intensity of gamma-emission of uranium-235, temperature, and uranium hexafluoride gas phase pressure in measuring chamber. Averaged data are processed to create uranium hexafluoride canal in measuring chamber. Thereafter, measurements are performed within a time interval composed of a series of time gaps and average values are then computed for above-indicated parameters for each time gap and measurement data for the total time interval are computed as averaged values of average values in time gaps. Intensity of gamma-emission of uranium-235, temperature, and pressure, when computing current value of mass fraction of uranium-235 isotope, are determined from averaged measurement data obtained in identical time intervals at variation in current time by a value equal to value of time gap of the time interval. Computed value of mass fraction of uranium-235 isotope is attached to current time within the time interval of measurement. Method is implemented with the aid of measuring system, which contains: measuring chamber provided with inlet and outlet connecting pipes, detection unit, and temperature and pressure sensors, connected to uranium hexafluoride gas collector over inlet connecting pipe; controller with electric pulse counters and gamma specter analyzer; signal adapters; internal information bus; and information collection, management, and processing unit. Controller is supplemented by at least three discriminators and one timer, discriminator being connected to gamma-emission detector output whereas output of each discriminator is connected to input of individual electric pulse counter, whose second input is coupled with timer output. Adapter timer output is connected to internal information bus over information exchange line. Information collection, management, and processing unit is bound to local controlling computer network over external interface network.EFFECT: enabled quick response in case of emergency deviations of uranium hexafluoride stream concentration, reduced plant configuration rearrangement at variation in concentration of starting and commercial uranium hexafluoride, and eliminated production of substandard product.24 cl, 5 dwg

ethod for evaluating burnout margin in nuclear power units // 2256962
FIELD: power engineering; evaluating burnout margin in nuclear power units.SUBSTANCE: proposed method intended for use in VVER or RBMK, or other similar reactor units includes setting of desired operating parameters at inlet of fuel assembly, power supply to fuel assembly, variation of fuel assembly power, measurement of wall temperature of fuel element (or simulator thereof), detection of burnout moment by comparing wall temperatures at different power values of fuel assembly, evaluation of burnout margin by comparing critical heat flux and heat fluxes at rated parameters of fuel assembly, burnout being recognized by first wall temperature increase disproportional relative to power variation. Power is supplied to separate groups of fuel elements and/or separate fuel elements (or simulators thereof); this power supplied to separate groups of fuel elements and/or to separate fuel elements is varied to ensure conditions at fuel element outlet equal to those preset , where G is water flow through fuel element, kg/s; iout, iin is coolant enthalpy at fuel element outlet and inlet, respectively, kJ/kg; Nδi is power released at balanced fuel elements (or simulators thereof) where burnout is not detected, kW; n is number of balanced fuel elements; Nbrn.i is power released at fuel elements (or element) where burnout is detected; m is number of fuel elements where burnout is detected, m ≥ 1; d is fuel element diameter, mm.EFFECT: enhanced precision of evaluating burnout margin for nuclear power plant channels.1 cl, 2 dwg

ethod of controlling nuclear fuel for thermal resistance // 2256961
FIELD: nuclear fuel technology.SUBSTANCE: invention relates to production of pelleted fuel and consists in controlling nuclear fuel for thermal resistance involving preparation for selecting pellets from nuclear fuel lot for measuring diameter, which preparation consists in dedusting. Selected pellets are placed in temperature-stabilized box together with measuring instrument. Diameter of each pellet is them measured and measurement data are entered into computer. Thereafter, pellets are charged into heat treatment vessel, wherein pellets are heated in vacuum at residual pressure not exceeding 7·10-2 Pa at heating velocity not higher than 10°C/min to 100-160°C and held at this temperature at most 2 h, whereupon heating is continued under the same conditions to 1470-1530°C and this temperature is maintained for a period of time not exceeding 4 h, after which hydrogen is fed with flow rate 2-6 L/min. Humidity of gas mix is measured in the heat treatment outlet. If humidity of gas mixture in the heat treatment outlet exceeds 800 ppm, hydrogen feeding is stopped and material is subjected to additional vacuum degassing at residual pressure below 7·10-2 Pa and held at 1470-1530°C in vacuum for further 4 h. Hydrogen feeding is the repeated at 2-6 L/min. If humidity of gas mixture in the heat treatment outlet is below 800 ppm, preceding temperature is maintained not longer than 2 h and raised to 1625-1675°C at velocity 40-60°C/h and then to 1700-1750°C at velocity 15-45°C/h. When outlet humidity of mixture is 500-750 ppm, hydrogen feeding is lowered to 1 L/min. Temperature 1700-1750°C is maintained during 24±2 h, after which pellets are cooled to 1470-1530ºC at velocity not higher than 10°C/min. Hydrogen is replaced with argon and cooling is continued to temperature not higher than 40°C, which temperature is further maintained. Outside diameter of each pellet from the selection is measured to find average diameter of pellets before and after heat treatment in order to calculate residual sintering ability. When this parameter equals 0.0-0.4%, total lot of pellets is used in fuel elements and in case of exceeding or negative residual sintering ability the total lot of pellets is rejected.EFFECT: improved pellet quality control.2 dwg

Inspection and sorting-out line for fuel elements // 2256248
FIELD: atomic industry.SUBSTANCE: proposed line is provided with computer-aided system for contactless control of flaw depth and profile on surface of fuel element can and on end parts including sorting-out device that functions to reject faulty fuel elements. This line is characterized in high capacity and reduced labor consumption.EFFECT: enlarged functional capabilities, improved quality of fuel elements.1 cl, 2 dwg

Pulse-current simulator of nuclear reactor kinetics // 2256221
FIELD: analog computer engineering; verifying nuclear reactor reactivity meters (reactimeters).SUBSTANCE: proposed simulator has m threshold devices, m threshold selector switches, m series-connected decade amplifiers, m electronic commutators, n - m - 1 series-connected decade frequency dividers, first group of m parallel-connected frequency selector switches, second group of n - m frequency selector switches, and group of n - m parallel-connected mode selector switches. Integrated inputs of threshold selector switches are connected to output of high-voltage amplifier and output of each threshold selector switch, to input of respective threshold device; output of each threshold device is connected to control input of respective electronic commutator; inputs of electronic commutators are connected to outputs of decade amplifiers and outputs are integrated with output of group of mode selector switches and with input of voltage-to-frequency converter; output of inverting amplifier is connected to input of first decade amplifier and to that of group of mode selector switches; input of first group of frequency selector switches is connected to output of voltage-to-frequency converter and to input of first decade frequency divider and output, to integrated outputs of first group of frequency selector switches and to input of division-chamber pulse shaper input; each of inputs of second group of frequency selector switches is connected to input of respective decade frequency divider except for last one of this group of switches whose input is connected to output of last decade frequency divider; threshold selector switches and frequency selector switches of first group, as well as m current selector switches have common operating mechanism; mode selector and frequency selector switches of second group have common operating mechanism with remaining n - m current selector switches. Such design makes it possible to realize Coulomb law relationship at all current ranges of simulator for current and frequency channels.EFFECT: ability of verifying pulse-current input reactimeters by input signals adequate to signals coming from actual neutron detector.2 cl, 1 dwg

Installation for determination of hydrogen in uranium dioxide fuel pellets // 2253915
FIELD: the invention refers to analytical chemistry particular to determination of general hydrogen in uranium dioxide pellets.SUBSTANCE: the installation has an electrode furnace with feeding assembly , an afterburner, a reaction tube with calcium carbide, an absorption vessel with Ilovay's reagent for absorption of acetylene, a supply unit. The afterburner of hydrogen oxidizes hydrogen to water which together with the water exuding from pellets starts reaction with carbide calcium. In result of this equivalent amount of acetylene is produced. The acetylene passing through the absorption vessel generates with Ilovay's reagent copper acietilenid which gives red color to absorption solution. According to intensity of color of absorption solution the contents of general hydrogen are determined.EFFECT: simplifies construction of the installation, increases sensitivity and precision of determination of the contents of hydrogen in uranium dioxide pellets.2 cl, 1 dwg

Hydrogen analyzer for uranium dioxide fuel pellets // 2253157
FIELD: analyzing metals for oxygen, nitrogen, and hydrogen content including analyses of uranium dioxide for total hydrogen content.SUBSTANCE: proposed analyzer depending for its operation on high-temperature heating of analyzed specimens has high-temperature furnace for heating uranium dioxide pellets and molybdenum evaporator; molybdenum evaporator is provided with water-cooled lead-in wire, and molybdenum deflecting screen is inserted between molybdenum evaporator and furnace housing.EFFECT: simplified design of electrode furnace, reduced power requirement.1 cl, 1 dwg

ethod for identifying spent fuel assembly // 2249265
FIELD: identifying o spent fuel assemblies with no or lost identifying characteristics for their next storage and recovery.SUBSTANCE: identifying element is made in the form of circular clip made of metal snap ring or of two metal semi-rings of which one bears identification code in the form of intervals between longitudinal through slits. Clip is put on fuel assembly directly under bracing bushing and clip-constituting semi-rings are locked in position relative to the latter without protruding beyond its outline. For the purpose use is made of mechanical device of robot-manipulator type. Identification code is read out by means of mechanical feeler gage and sensor that responds to feeler gage displacement as it engages slits. Identifying elements are installed under each bracing bushing.EFFECT: ability of identifying fragments of spent fuel assembly broken into separate parts before recovery.10 cl, 4 dwg

Nuclear fuel granule simulator // 2248053
FIELD: nuclear power engineering.SUBSTANCE: proposed invention may be found useful for optimizing manufacturing process of dispersion-type fuel elements using granules of uranium, its alloys and compositions as nuclear fuel and also for hydraulic and other tests of models or simulators of dispersion-type fuel elements of any configuration and shape. Simulators of nuclear fuel granules of uranium and its alloys are made of quick-cutting steel alloys of following composition, mass percent: carbon, 0.73 to 1.12; manganese and silicon, maximum 0.50; chromium, 3.80 to 4.40; tungsten, 2.50 to 18.50; vanadium, 1.00 to 3.00; cobalt, maximum 0.50; molybdenum, 0 to 5.30; nickel, maximum 0.40; sulfur, maximum 0.025-0.035; phosphor, maximum 0.030; iron, the rest.EFFECT: enhanced productivity, economic efficiency, and safety of fuel element process analyses and optimization dispensing with special shielding means.1 cl, 3 dwg