The method for determining heat generation in the fuel element during development in the hinge channel of a nuclear reactor
The invention relates to nuclear power, namely, the development of fuel elements, their experimental development in nuclear reactors, in particular high-temperature thermionic fuel elements when creating electricity generating channels thermionic reactor Converter. Constant thermal power of the reactor at the desired time t measure the pressure of the gaseous fission product P in the ventilation system of a fuel rod, which used a vented fuel element, measure the temperature of the shell of a fuel rod Taboutafter which measure the amount of fuel material m, released from the fuel, determine the maximum temperature in the fuel core of a fuel rod T0and evaluation of heat qvin Fe on the proposed expressions. Technical result - increase the accuracy of determination of heat generation in the fuel rod during the experimental testing in a nuclear reactor. 4 Il.The efficiency of the fuel rods is studied in the loops of research reactors, loop channels (PC). The test loop is a separate channel in the reactor, equipped with Autonomous cooling circuit with its own pumps, heat exchangers, poetry of the coolant at the inlet and outlet, the coolant flow rate, pressure, power supply of fuel cladding tightness, condition of the coolant, sometimes measure the temperature of the fuel rods.Control of heat in the fuel rods, in particular in thermionic fuel elements, is of paramount importance for loop testing, in particular to determine such fundamental parameters as efficiency, as well as for proper and safe organization of tests .There are ways to determine the heat release in the fuel rods, in particular thermionic fuel elements of power-generating elements (AGE) forming a power generation Assembly (EHS), when experimental development EHS in the hinge channel of a nuclear reactor.The method of determining thermal capacity of the fuel rods EHS the recalculation method  is to use a for loop test methods relative translation. Its essence lies in the fact that if one tested PC relatively reliably known heat capacity of the fuel rods of the EHS, the deviation designs EHS and PC environments and next PC from known can be taken into account by introducing some correction factors that can be known a priori, estimated, calculated or determined experimentally. Prim, including changing to a different fuel enrichment; changing the position of the PC according to the height of the reactor core; introduction to PC-absorbing neutrons screens, etc. However, the relative translation does not have high accuracy, but sometimes it is convenient to estimate the expected heat dissipation at small differences of the new PC and EHS.Thermal method of determining thermal capacity of thermionic fuel elements EHS  is the measurement values of the flow G and the difference between the enthalpies of the fluid between the points corresponding to the edges of the active part of the EHS. In this case, the heat power Q discharged by the pumped fluid, is determined by the formulawhere C is the specific heat of the fluid at temperature T=(T1+T2)/2;T=T2-T1- heated coolant.In calculated by (1) the value of Q is included radiative dissipation Qpstructural materials PC. Therefore, thermal power generated in the fuel cores thermionic fuel elements EHSwhere WEHSuseful electric power Assembly.Thermal control method, mainly owing to uncertainties in the values of radiation heat generation structural materials PC.Closest to the invention to the technical essence is a method of determining thermal capacity of the fuel elements under development in the hinge channel of a nuclear reactor, including the assessment of heat generation in the fuel rod, which is known thermophysical properties of materials, described in . This so-called radial thermal conductivity, namely, that in the system of heat PC set differential thermocouple. Block differential thermocouples previously in laboratory conditions tarinoita depending on the heat flux. The calibration table is transferred and loop experiment. Here, just as in thermal method, you must take into account radiative heat Qpall materials between the interelectrode gap EHS and a differential thermocouple. Thermal power in a fuel rod is determined by the expression (2). Setting the differential thermocouple opposite each AGE, you can determine the approximate distribution of heat on the height of the EHS. This method gives a significant error because of the difficulty of accounting for heat losses in the axial direction in the additional conditions (distributed in height) radiation heat generation in the PC materials, which is not ocherovatelna in materials EHS and PC this improves the accuracy in determining the heat release in the fuel rod during the experimental testing in a nuclear reactor.This object is achieved by the proposed method to determine the heat release in the fuel element during development in the hinge channel of a nuclear reactor, including the assessment of heat generation in the fuel rod, which is known thermophysical properties of materials as a fuel rod used vented fuel elements, and constant thermal power of the reactor at the desired time t measure the pressure of the gaseous fission product P in the ventilation system of a fuel rod, measure the temperature of the shell of a fuel rod Taboutafter which measure the amount of fuel material m, released from the fuel, determine the maximum temperature T0in the fuel core of a fuel rod from the relationand evaluation of heat qvin Fe is performed according to the expressionwhereR is the total resistance of the ventilation system, 1/m;A and b are coefficients depending on the type of the fuel material;r - radius of the inner membrane of a fuel rod, m;- coefficient of thermal conductivity of the fuel material, W/(m·deg);While [m2·with3·deg1/2/kg2]; [Deg]; t [p]; qv[W/m3]; m [kg]; P [PA]; T0[K]; Tabout[K].In Fig.1 and 2 schematically presents the main structural variants common types of vented fuel rods, which can be implemented this way. In Fig.3 schematically shows a nuclear reactor, where the composition of the loopback channel is being TVEL. In Fig.4 is a graph illustrating the inventive method.In Fig.1 and 2 denote: 1 - Fe, 2 - shell, 3 - fuel material (TM), 4 - ventilation system, 5 - tube, 6 - capillary tip 7 - in temperature sensor, 8 - condensate TM 9 - substrate, 10 - camera for condensate TM. In Fig.1 the ventilation system 4 consists of a Central axisymmetric tube 5 with the capillary tip 6. In Fig.2 ventilation system 4 is made in the form of the Central channel, piercing TM for the entire length of a fuel rod. In Fig.3 marked: 11 - nuclear reactor, 12 - reflector control, 13-active area, a 14 - loop channel, 15 - electricity generating Assembly (EHS), 16 - pressure sensor, 17 - tank-tank gaseous fission products (GPA).The method is implemented as follows.TVEL 1 in the composition of the electricity generating Assembly 15 is placed in a loopback channel 14 provided with Noahs 15 is placed in the cell active area 13 of a nuclear reactor 11. The reactor 11 are outputted by the planned thermal power and maintain it constant over the interest of time t. During operation of the reactor 11 at a constant level of heat capacity in a vented fuel element 1 EHS 15 is a division of nuclear fuel in TM 3 with the formation of gaseous fission products (GPA) coming through the ventilation system 4 outside of the fuel rod 1, and then the reactor 11 in the reservoir tank 17. Simultaneously with the GPA through the ventilation system 4 of the fuel rod 1 out and molecules TM 3, diffusing into the gas-vapor medium consisting of GPA and TM.To prevent clogging of channels of the output gap, a pair of TM released from the fuel rod 1, are separated from the GPA in the chamber 10 by condensation on the substrate 9. In any moment of time t measured by GPA pressure P in the ventilation system 4 fuel rod 1, for example using a pressure sensor 16, and the temperature of the shell 2 Taboutthe fuel rod 1, for example using a temperature sensor 7. Then measure the amount of fuel material m, vented out of the fuel rod 1 and condensed on the substrate 9 in the chamber 10. To measure the amount of condensate 8 can, for example, directly using the method of neutron radiography of irradiated loop channels, as is done in experienee studied object beam of neutrons with the subsequent registration of radiographic detector of the distribution of neutron flux, passed through the object in the plane perpendicular to the direction of scanning. The resulting distribution characterizes the degree of attenuation of the neutron flux in the beam sections of the object in the direction of scanning. In this case, the PC 14 at the time of registration is extracted from the cells of the reactor 11 and is placed in a neutron diffraction unit (not shown), and then again PC 14 return to the cell of the reactor 11 to continue experimental studies. Neutron radiography record the volume V of the condensate 8 and, when the density TMdefine m=V·. Knowing the total resistanceR ventilation system 4, determine the maximum temperature0in the fuel core of a fuel rod 1 from the relation (3)m=t/(A·P·R)·(T1/20/exp(B/T0)),and evaluation of heat qvin Fe is performed according to the expression (4)qv=4·r-2·(T0-Tabout)/(·ln-+1)).In the derivation of the relation (3) we use the phenomenon of diffusion of TM in the one-dimensional case in the two-component system described by the first law is TM inside it or this condensation is negligible and does not affect the performance of the ventilation system.In this case, the first law Fika can be written in the form:where m is the number of TM released from vented TVEL;t - time;D is the diffusion coefficient of molecules TM in the gas mixture GPA and molecules TM;molecular weight TM;nothe concentration of TM at the outlet of the ventilation system of a fuel rod, corresponding to the vapour pressure of TM when the temperature of the substrate where it condenses TM;n0- the maximum concentration of molecules TM in Fe;R is the total resistance of the ventilation system.While we believe that noand n0do not depend on t, which corresponds to the operating condition of the reactor at a constant heat capacity.In case of performance of the ventilation system in the form of Central axisymmetric tube with a capillary tip (Fig.1) the total resistance of the ventilation systemR=R1+R2+R3,where R1, R2, R3resistance provided by the capillary tip, tube and chamber condensate TM, respectively.Due to the smallness summand R3compared to R1and R2they can be neglected.
< The TM, as shown in Fig.2, in the first approximation can be consideredIn a first approximation, the diffusion coefficient D of molecules TM to nonequilibrium stationary gas mixture of molecules TM and GPA (mostly molecules Heh ) is calculated by the formula where u is the average velocity of thermal motion of molecules TM;* the average free path length of the molecules TM.The velocity u, we define the expression given in , and* - from the expression given in , taking into account the ratio P=nkT0from  and considering that the GPA consist mainly of Heh, as follows from where k is Boltzmann's constant;T0temperature;d, dXethe diameters of the molecules TM and Eh, respectively;,Heh- molecular weight molecules TM and Eh, respectively; P is the pressure in GPA.Knowing the density TMto determine d from the relation d=1,122·(/)1/3, and dXefrom .Given the exponential dependence of the vapor pressure Pthe>/p>where a* and b are coefficients depending on the type TM.Where the expression for the maximum concentration of TM in Fe can be written asGiven that the temperature of the substrate on which condensation occurs TM at the exit of the fuel rod in the camera (see Fig.1, 2), much less than the maximum temperature TM in the fuel elements and taking into account the exponential dependence of the concentration of TM temperature (12)Given the above, substituted in (5) expressions (8) and (12) taking into account(9), (10), (13), we obtain the relation (3)m=t/(A·P·R)·(T1/20/exp(B/T0)),where the factor a depends on the type TM and is determined from the expressionTo determine the maximum temperature T0the fuel material in a vented fuel element of (3) can, for example, an iterative method or graphically.The expression (4) to determine the heat release qvin Fe, we get a differential equation of thermal conductivity, using a special case solution for a hollow cylinder with heat, cooled from the outer surface , which is typical for our case,where qv
ClaimsThe method for determining heat generation in the fuel element during development in the hinge channel of a nuclear reactor, including the assessment of heat generation in the fuel rod, which is known thermophysical properties of materials, characterized in that the quality of fuel used vented fuel elements, and constant thermal power of the reactor in interewt temperature membrane fuel rod Taboutafter which measure the amount of fuel material m, released from the fuel, determine the maximum temperature T0in the fuel core of a fuel rod from the relationm=t/(A·P·R)·(T1/20/exp(a/T0)),and evaluation of heat qvin Fe is performed according to the expressionqv=4·r-2·(T0-Tabout)/(·ln-+1)),whereR is the total resistance of the ventilation system, 1/m;A and b are coefficients depending on the type of the fuel material;r - radius of the inner membrane of a fuel rod, m;- coefficient of thermal conductivity of the fuel material, W/(m·deg);the relative volume fraction of porosity of the fuel material in the fuel rod, Rel. units;While [m2·with3·deg1/2/kg2]; [Deg]; t [p]; qv[W/m3]; m [kg]; P [PA]; T0[K]; Tabout[To].
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