Test method of materials in nuclear reactor
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
The invention relates to the field of reactor materials and can be used for reactor tests of structural materials for nuclear reactors.
There is a method of testing materials for long-term strength in the channel of a nuclear reactor: the patent for the invention №93013102 (1995.07.20, authors: Popov V.V. Potapov J.V.). The method applies to research the strength characteristics of materials, in particular tests of thin-walled tubular specimens loaded by internal pressure, and can be used to test fuel element cladding of nuclear reactor irradiation conditions. The method improves the accuracy of determining the time to failure of the sample in terms of exposure in reinstrumentation channel of the reactor. The inventive test piece, made in the form of thin-walled tubes, load gas under pressure, is placed in the insulated tube is installed in the reactor in such a way that part of the ampoule with the sample is placed in an active area, and some outside the active area, and is exposed to failure of the sample, and the time to failure of the sample is determined by comparison of the relative deformations of the non-irradiated part of the ampoule and the ampoule analogue tested outside of the reactor at identical parameters of loading. This method allows you to gain strength characteristics the specifications specifications of the investigated material through known properties of another material. The disadvantage of this method is the implementation of the stress-strain state of the test material under the action of stresses only one type of tensile stresses. This disadvantage is due to structural properties of the device for implementing this method.
The closest analogue, coinciding with the claimed invention by the greatest number of essential features, a method described in the literature: "Stress state dependence of in-reactor creep and swelling. Part 2: Experimental Results". M.M. Hall, Jr., J.E. Flinn // Journal of nuclear materials, 2010. V.396. 119-129. In this work the method of testing samples of steel 304 on in-reactor creep and swelling. The constructional schema devices used for sample loading and setting various types of stress-strain state of the material. This method of testing allows you to implement uniaxial tension, uniaxial compression, biaxial stretching, biaxial compression and circumferential stretching. However, the disadvantage of this method of testing is the inability to realize simultaneously the impact on the material as tensile and compressive stresses. This disadvantage is caused by the design of the devices developed to implement the described method of testing materials.
The aim of the invention is the higher is their informativeness and reliability of the results modify the properties of reactor materials under irradiation in the reactor under different types of stress-strain state.
This objective is achieved in that in the method of testing materials in a nuclear reactor, comprising loading the sample gas under pressure and placing it in a nuclear reactor, produce a sample of two coaxially aligned tubular elements, one of which is fully or partially inside the other, creating a gas pressure in the cavity between the elements, seal, place in a nuclear reactor and irradiated.
The minimum clearance between the elements is not less than 0.1 mm, it is Technologically difficult to provide a smaller gap, potential distortions that can lead to inaccurate results.
The wall thickness of the tubular elements does not exceed the size of the gap between the elements - the condition of tonkostennoy tubular sample.
Sealing compound tubular sample perform with end fittings, welding system, them to both ends of the inner and outer elements by way of argon-arc welding under pressure.
The internal volume of the composite sample between the inner and outer element is filled with an inert gas (argon, helium) under pressure, ensuring the required level of stresses in tubular structural elements.
The use of a composite sample consisting of two coaxially aligned tubular elements, internal, under esteem compressive stresses and external, under the action of tensile stresses, allows at the same temperature and dose under reactor irradiation simultaneously implement two completely different types of stress-strain state experienced a structural material.
After conducting reactor tests are really measuring the geometric dimensions of the tubular structural elements to determine the magnitude of their deformation.
New the essential feature of the proposed method is the production of such sample, which created the opportunity to implement in a single sample simultaneously two fundamentally different types of stress-strain state experienced a structural material under the action of tensile and compressive stresses at the same temperature and dose under reactor irradiation.
In the patent and technical literature, no information about the use of the same test method and devices with this significant feature, which allows to conclude that the claimed solution is not known from the prior art and novel and not obvious from the prior art, i.e. involves an inventive step.
The proposed method provides a technical effect, and may be implemented using known what technology means. Therefore, it has industrial applicability.
The inventive method is technically implemented using the fabricated device is shown on the accompanying drawing, where:
1 - the inner tubular element;
2 - external tubular element;
3 is a bottom annular tube;
4 - upper annular tube;
5 - technological stub.
The invention is illustrated by the following example. For implementing the inventive method test material in a nuclear reactor BOR-60 were manufactured composite tubular samples. Material used pipe from steel 18CR10NITI in austeniana condition with a diameter of 10.2 mm with wall thickness of 0.5 mm Pipe for the inner part of the samples was tied up at the rolling machine from the source pipe 10,2×0.5 mm in size 6,03×0,55. After tightening the pipe held joint heat treatment of pipes of large and small diameters: the austenization at a temperature of 1050°C for 30 minutes with cooling with the furnace. Thus, achieved similar structural state of the metal pipes.
Made 8 samples, two for each pressure level - 0, 25, 70 and 100 ATM. The samples were filled with argon and germetizirovany using argon-arc welding under conditions of excessive pressure with subsequent testing for tightness. All the fabricated samples were hermeti is low.
The samples were placed in a reactor and tested at temperatures (420-450°C to damaging doses of 15 and 30 sleep. After conducting reactor tests were conducted measuring the geometric dimensions of the tubular structural elements to determine the magnitude of their deformation.
1. The method of testing materials in a nuclear reactor, in which the test sample is used for gas under pressure and place it in a nuclear reactor, characterized in that the manufactured sample of two coaxially aligned tubular elements, one of which is fully or partially inside the other, creating a gas pressure in the cavity between the elements, seal, place in a nuclear reactor and irradiated.
2. The method according to claim 1, characterized in that the minimum gap between the elements is not less than 0.1 mm
3. The method according to claim 1, characterized in that the thickness of the walls of the tubular elements does not exceed the magnitude of the gap between the elements.
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
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
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.
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
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.
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
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.
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.
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.
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
FIELD: operating uranium-graphite reactors.
SUBSTANCE: proposed method for serviceability check of process-channel gas gap in graphite stacking of RBMK-1000 reactor core includes measurement of diameters of inner holes in graphite ring block and process-channel tube, exposure of zirconium tube joined with graphite rings to electromagnetic radiation, reception of differential response signal from each graphite ring and from zirconium tube, integration of signal obtained, generation of electromagnetic field components from channel and from graphite rings, separation of useful signal, and evaluation of gap by difference in amplitudes of signals arriving from internal and external graphite rings, radiation amplitude being 3 - 5 V at frequency of 2 - 7 kHz. Device implementing this method has calibrated zirconium tube installed on process channel tube and provided with axially disposed vertically moving differential vector-difference electromagnetic radiation sensor incorporating its moving mechanism, as well as electronic signal-processing unit commutated with sensor and computer; sensor has two measuring and one field coils wound on U-shaped ferrite magnetic circuit; measuring coils of sensor are differentially connected and compensated on surface of homogeneous conducting medium such as air.
EFFECT: ability of metering gas gap in any fuel cell of reactor without removing process channel.
2 cl, 9 dwg
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
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
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
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
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
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
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.
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
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