The purpose, structure, and basic characteristics of an acoustic system for monitoring coolant leaks in the first loop of a VVER reactor are presented. The principle of operation of the diagnostics algorithm is described. The system has been checked on a special stand. The system satisfies all sensitivity and temporal requirements. In 2005, SAKT was put into experimental operation in the No. 3 unit of the Kalinin nuclear power plant.One of the most dangerous accidents at a nuclear power plant with VVER is a guillotine break in first-and secondloop pipelines with bilateral coolant outflow. Expensive and labor-intensive protective measures and technical means must be used to prevent the consequences of such damage, which results in beating and knocking of pipelines together and substantial reactive forces acting on the equipment connected with the pipelines. The modern approach to solving this problem is based on the concept "leak before rupture" [1], according to which when a non-through defect forms complete failure of the pipeline will not occur during operation and the defects will not transform into a complete through rupture of the pipe as long as the length of the through defect does not exceed a critical value. In this case, early detection of a leak will make it possible to stop the reactor safely and perform repairs or replace the pipeline. Adhering to this concept makes it possible to do away with examining guillotine rupture and the associated consequences and also eliminate expensive protective facilities (support, screens, tie rods, and so forth).Systems for detecting leaks in the first and second groups are provided as technical support for the designs of reactor facilities with VVER. To increase detection reliability, as a rule, it is assumed that at least three systems operating on different physical principles will be used. Acoustic systems are at the top of the list. The principal of operation of such systems is well known, and by now there are several developments, mainly foreign developments, which have found application in nuclear power plants with VVER.The present paper describes the first domestic multi-channel automated acoustic system for monitoring leaks, called
Conclusions are drawn, on the basis of experimental data obtained during experimental operation of the acoustic system for monitoring the leaks in the No. 3 unit of the Kalinin nuclear power plant, concerning the source of background noise which competes with the noise due to the efflux of coolant through leaks in the first loop. Two main sources are identified. The first one at pressure less than 8 MPa in the loop is due to pulsations of bubbles of undissolved gas in the coolant. The second source, which produces correlated noise in all circuits of the loop, is due to noise originating in the deaerator of the purge-makeup system. The high degree of correlation of the intensity of background noises is important for filtering such noise.Acoustic noise generated during liquid outflow under high pressure carries information about leaks in the circulation loops of a VVÉR reactor. Acoustic methods of diagnostics which are based on this effect a have two important advantages: they are fast-acting and, in principle, permit estimating the size and coordinates of a leak.The acoustic method can be implemented in two ways -contact, when the information indicator is the intensity (spectral composition) of stress waves generated on the surface of pipelines and equipment vessels by the outflowing steam-liquid medium, and purely acoustic, associated with the detection of low-frequency acoustic pressure waves in air near the equipment being monitored. For several reasons, the contact acoustic method is widely used in nuclear power plants with VVÉR reactors.A salient feature of a reactor system as a source of acoustic noise is the complexity of the processes occurring in a system with many connections which are difficult to take into account -mechanical, hydrodynamic, vibration, and impact processes, those associated with the generation and bubbling of steam, and so forth. All this engenders multiple acoustic sources which act in different frequency ranges that depend on the operating regime of the system, the composition of the equipment, and other factors. Naturally, there is no hope of detecting coolant leaks without studying the characteristics of such noise and finding their basic sources.Publications devoted to the diagnostics of leaks in the loops of nuclear power plants have not devoted the proper attention to background acoustic noise. Consequently, the present paper will be of interest to specialists working in the field of acoustic leak detection.
The subject of this paper involves monitoring the temperature of the heat transfer agent above the core and fluctuations in the temperature of the sodium, affecting the strength of the constructions under steady-state reactor operating conditions. Sources of temperature differences may be local superheating with deformations of the fuel element lattices, blocking of parts of the cross sections of the fuel assemblies, and other such factors. Some information about this is presented in [1][2][3].The experimental assembly consisted of seven models of fuel assemblies (Fig. 1). Nineteen fuel element simulators separated by wire coils were mounted in each model. Electrical heating of the simulators was provided only in the central model. The thermocouple probe (Fig. 2) consisted of 30 chromel-alumel thermocouples in stainless-steel capillaries of diameter 0.8 mm. The probe could be placed along the height from the middle of the caps (h = 0) to 200 mm. The placement of the thermocouples is shown in Fig. la. There were also thermocouples in the caps of the fuel assemblies in the cross section of the lower edges of their windows, at the outlet from the model and at the inlet to the "cold" fuel assemblies.There were two independent heat-transfer-agent feeds: into the central model and simultaneously into the six lateral models. We considered the case of superheating of the heat transfer agent in the central model, accomplished either by feeding the heat transfer agent at a higher temperature or by heating it with the electric heaters of the fuel element simulators. The basic parameters for the two series of experiments are given in Table 1. When feeding hotter heat-transfer agent, the possibility of generating additional thermal noise from the electric heaters is eliminated.The average temperature was measured by means of a scanning system. The temperature pulsations were measured and statistically treated using a system including an amplifier circuit, a magnetograph, and a computer. We entered 8100 numbers for each statistical treatment.The temperature distribution over the radius of the model, averaged over the thermocouple readings, is shown in Fig. 3. From this figure it follows that the "hot spot" from the central fuel assembly spreads out with an increase in height. There is some anomaly for low velocity of the "hot" flow (see Fig. 3a, b, c): at some distances from the end of the cap of the fuel assembly, a temperature minimum appears. The effect is possibly connected with cold sodium from adjacent cells entering this region.The flow of heat with an increase in distance from the caps of the models occurs differently for the central sodium flow and the flows from the lateral windows (at the radius rl) (Fig. 4). At a height of 200 mm, the temperature difference between the central flow (t 0) and the cold sodium tr4 (at the radius r4) is only 20% of the initial temperature difference (t h -tc) (the indices h, c mean hot, cold at the inlet to the model). Within the cap, the temperature of this flow decreases by 10%:tO-tr...
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