Method to test zirconium alloys in steam and water medium
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
The invention relates to the field of testing of materials, in particular to test the corrosion resistance and vodorodostojkih zirconium alloys developed and used as materials of the elements of the active zone light-water nuclear reactors, in the conditions close to the reactor. The invention can be used to directory testing of zirconium alloys and to study the processes of their interaction with the steam environment in the core of a light water reactor, including the study of the capture of hydrogen ions of various energies and doses in zirconium alloys in steam environments at elevated temperatures and the influence of ion irradiation on the corrosion of zirconium alloys in a given environment.
The known method of corrosion-thermal tests of long unfragmented of Fuel rods, including shells of zirconium alloys (SU 1783383 A1 G01N 17/00, G01N 3/18) in which sections of the test fuel Rod is placed in a quartz chamber, is exposed to heat, water and steam under pressure.
This method is characterized by extremely low productivity tests in connection with the need for sequential testing of each fuel Rod in isolation and impossibility of simultaneous tests of different samples of zirconium alloys. Furthermore, the method features�isoamsa unnecessarily large consumption of expensive materials of Fuel elements.
Closest to the invention is a method of determining the corrosion resistance of zirconium alloys for nuclear reactors (JPH 01250736 A, G01N 17/00, G21C 17/06). The method consists of heating and holding samples of zirconium alloys in steam environments at different temperatures: at temperatures in the range 300-400°C temperature range of the active zone light-water nuclear reactor; at a temperature in the range of 490-530°C for rapid determine the tendency of zirconium alloy to nodular corrosion.
A common drawback of both of analog techniques is that they do not adequately simulate the processes of aggressive environment on zirconium alloys in the reactor core, because it does not include the impact on zirconium alloys of hydrogen ions that occurs in the process of operation of the reactor due to the particles that are created as a result of radiolysis of water.
This disadvantage significantly reduces the use of the produced according to the method, the information for making judgments about the behavior of zirconium alloys studied in the core of a light water reactor in the process of his work.
The technical result of the invention is the approximation of the test conditions samples of zirconium alloys in steam-water environment to the conditions of the active zone light water re�Chora, that can improve the accuracy of the predicted pattern of behavior of zirconium alloys studied in the core of a light water reactor in the process of his work, compiled on the basis of the results of these tests.
The technical result is achieved in that in the proposed method, which includes the extract of samples of zirconium alloys in steam environments in the temperature range of the active zone of a light water reactor according to the invention, in the process keeping the samples in the water vapor environment is created by gas discharge plasma in water vapor, and then irradiated with educated with the positively charged hydrogen ions samples by filing a negative electric potential relative to the plasma.
In the particular case, to ensure uniform over the surface and controlled by the intensity of irradiation of the samples for hydrogen ions, the samples irradiated with positively charged ions of hydrogen plasma glow discharge in water vapor. In diffuse burning glow discharge allows to obtain a uniformly distributed current density of positive ions on the electrode surface under a negative electric potential relative to the plasma.
In the particular case of irradiation of samples positively charged ions of hydrogen plasma clauses� discharge in water vapor for stable diffuse burning glow discharge during irradiation of the samples in the water vapor environment is created by the turbulent flow of steam. Creating a turbulent flow of steam in the region of the discharge leads to an intensification of convective heat transfer, which helps prevent the development of thermal instabilities in the plasma and thus to exclude the possibility of occurrence of spark breakdowns that turns the irradiation process is out of control.
An example of a specific implementation of the method
Fig.1 is a diagram of a device for implementing the proposed method, where 1 - test chamber, 2 - boiler with water, 3 - steam superheater, 4 - capacitor pair, 5 - studied samples of zirconium alloys, 6 - anode 7 - thermocouple temperature sensor.
The method is implemented as follows. Samples 5 E110 alloy is placed in a test chamber 1. In the boiler 2 heat the water to boiling point. Steam at a temperature of 100°C enters test chamber 1. The test chamber 1 steam heated to a temperature of 400°C using steam superheater 3. The temperature of the steam in the test cell 1 is controlled using a thermocouple sensor 7. The vapor pressure in the test cell 1 is close to atmospheric. Between the 5 samples are cathodes and the anode 6 is located at a distance of 0.5 cm from each other, serves a potential difference of 15 kV. The result is a breakdown of the gas gap between them and a mass of plasma glow discharge in pairs in�dy. The 5 samples as cathodes, are under negative potential relative to the plasma and are exposed to the positively charged hydrogen ions. The discharge current of ~7-8 mA, the flux density of ions on samples of ~2×1016cm-2h-1. For the irradiation time of 15 min is dialed dose, equal to 1.7×1019cm-2that corresponds to an estimated radiation dose by rapid proton parts in the core of a light water reactor at fluence neutrons 1022cm-2. Superheated steam passing through the test cell 1 enters the steam condenser 4 - phase pipeline, cooled by water. In the condenser steam 4 steam is cooled, condensed and the resulting water flows back into the boiler, thereby closing the steam-water cycle.
Thus, it follows from the above that the proposed method allows to approximate the conditions of testing of samples of zirconium alloys in steam-water environment to the active area of a light water reactor, because it involves the irradiation of zirconium alloys for hydrogen ions that occurs in the process of operation of the reactor due to the particles that are created as a result of radiolysis of water. Thus, using this method can improve the accuracy of the predicted pattern of behavior of zirconium alloys studied in the core of a light water reactor in PR�process of his work, based on the results of out-of-pile tests.
1. Method of testing of zirconium alloys in steam environments, including exposure of samples of zirconium alloys in steam environments in the temperature range of the active zone of a light water reactor, characterized in that in the process of keeping the samples in the water vapor environment is created by gas discharge plasma in water vapor, and then irradiated with educated with the positively charged hydrogen ions samples by filing a negative electric potential relative to the plasma.
2. A method according to claim 1, characterized in that the samples irradiated with positively charged ions of hydrogen plasma glow discharge in water vapor.
3. A method according to claim 2, characterized in that during the irradiation of samples positively charged ions of hydrogen plasma glow discharge in the water vapor environment is created by the turbulent flow of steam.
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
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
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
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
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
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: 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