No abstract

It is shown by means of calculations and experiments that the generally accepted practice of using an unconditional instead of a conditional probability density distribution when estimating the error in monitoring the power in the fuel channels with the highest power density in the cores of a wide class of reactors with standard power monitoring error greater than 1.5% results in a systematic underestimation of the margin up to the limits for safe operation by up to approximately 5 and 10% for regimes with and without real-time optimization of the energy release distributions, respectively. This underestimation decreases to zero monotonically when the average power is used instead of the maximum power.Two types of errors are unavoidably encountered in designing and operating nuclear reactors with high power density -a measurement error and a regulation error, most often smoothing and distribution of the energy release and other distributed parameters of the core [1]. Ordinarily, the margin up to the safe operating limit is calculated assuming that the errors are independent of the measured value and the regulation algorithm, though, as shown in [2], this is not so for large measurement errors when corrections due to modeling of the smoothing and measurement errors must be introduced. It is legitimate to examine separately to problems of determining the distribution of the measurement error: for the fuel channel with the highest power density and no concrete coordinate and for each channel of the reactor with different parameter regulation methods, for example, without smoothing or with minimization of the parameter nonuniformity over the core. A detailed examination of the first problem is presented in [2]. The present article is devoted to solving the second problem and generalizing the results obtained.First, we shall examine the results of the solution of the first problem [2], where the errors in monitoring the coefficient of nonuniformity K r of the radial-azimuthal distribution of energy release are determined in regimes without and with smoothing of the energy release distribution in RBMK-1000. However, they can also be used for other reactors with high power density, which are equipped or not equipped with in-reactor monitoring systems. In the latter case, design neutronphysical calculations are used as a means for monitoring. The regime without smoothing was studied analytically and by the method of statistical modeling. The regime with smoothing was studied only using the latter method. The two-dimensional

An algorithm is proposed for reconstructing the radial-azimuthal distribution of the energy release in an RBMK-1000 core and its program implementation is constructed. Calculations of the dependence of the ratio of the sensitivity of a hafnium detector on the integrated current are presented. It is shown that the proposed algorithm will make it possible to increase the reconstruction accuracy of the power of process channels.A great deal of attention has been devoted recently to increasing the power of power-generating units with RBMK-1000 reactors. In this connection, it is necessary to increase the margins to the limits of safe operation, specifically, by using more accurate algorithms to reconstruct the radial-azimuthal distribution of the energy release in the special mathematical computational program Prizma-M, which takes account of the spectral sensitivity of the in-reactor detectors with hafnium-dioxide emitters [1].The existing algorithm for calculating the power of the process channels neglects the spectral sensitivity of in-reactor detectors, even though the ratio between the cell-averaged flux density of the suprathermal and thermal neutrons changes appreciably from one process channel to another and especially with a transition to the periphery of the core. The deviation of this ratio from the average is +82%, −19%, and the rms deviation in the plateau zone is 9%. Since the contribution of the suprathermal neutrons to the detector signal can be 20-30%, this will lead to large errors in the determination of the power of the process channels.One weakness in the existing algorithms of the SKALA-Mikro information-measurement system is that the Prizma-M program uses an empirical relation for calculating the fuel assembly power with an in-reactor detectorwhere J is the detector current, in μA; K cal is a calibration coefficient, in MW/μA; ξ d (I) is the dependence taking account of the sensitivity of the detector on the increase of its integral current with increasing integral current of the detector I; ξ t.d (E) is the dependence taking account of energy production of the fuel assembly.Relation (1) contains a methodological error, since it neglects the spectral sensitivity of the detector. In addition, when a new type of fuel assembly is put into operation the new dependences ξ t.d (E) must be calculated.The main problems of improving the algorithm are separating the contribution of the thermal and suprathermal neutrons and introducing the dependence of their sensitivity ratio to the suprathermal and thermal neutrons on the burnup and establishing the relation between the thermal neutron flux density at the detector's location and the power of the process channel. Obtaining the dependence of the detector's ratio of the sensitivity by a computational method is a difficult problem, so

A method developed for performing direct measurements of three-dimensional distributions of energy release and energy production in RBMK fuel assemblies is described. The method is based on performing measurements with a gamma-neutron chamber and comparing the neutron and gamma signals. The results of the measurements of the neutron flux density, energy release, and energy production are compared with the values obtained with the Prizma-M program of the Skala-micro information-measurement system. It is confirmed experimentally that the Prizma-M system can be used to monitor the distribution of not only the neutron flux density and energy release of fuel assemblies but also the energy production of off-loaded fuel assemblies.The need to perform periodic measurements of the thermal-neutron flux density distribution, energy release, and energy production in a RBMK-1000 core is due to the need for regular metrological certification, as provided in the operational documentation, of the Skala-micro information-measurement system and calibration of in-reactor detectors. This work is performed by scanning the core along the radius with Dt.6 hafnium detectors and scanning along the height of the core with Kt-19 fission chambers, However, when measurements of the energy release in fuel assemblies are performed the detectors do not provide the desired accuracy and do not permit determining the distribution of energy production in an operating reactor. The latter problem must be solved for complete certification of the Skala-micro system and, ultimately, to increase the safety of not only the reactor but also the storage facilities for spent nuclear fuel. The present work is devoted to solving these problems.Two-section triaxial Kt-18 ionization chambers were used to scan in an operating reactor fuel assemblies with assembly 49, specially provided for this purpose, with dry central tubes. One section of such a chamber is a fission chamber with 90% enrichment 235 U, and the other is a γ chamber. In chambers of this type, all electrodes are made of corrosion-resistant steel, and the insulation is made of compacted magnesium oxide; argon is the working gas [1]. The outer diameter of the chamber in the core is 6 mm, and the sensitive part of this section is 50 mm long; the distance between the centers of the sections is 85 mm. The ratio of the sensitive and total volume of the fission and γ chambers is ~0.047. In the course of the measurements, the chambers were moved with constant velocity along the height of a fuel assembly using a crane from the central hall. The measurements were performed at the No. 1 unit of the Kursk nuclear power plant on December 20, 2004 at 95% nominal power. Twenty five fuel assemblies were scanned. For technical reasons, some of these assemblies were scanned only along half of the core height. For this reason, we shall examine the value of any parameter only for the top half of a fuel

A significant influence on stability of the process of filling the CCM mold with liquid metal is exerted by the structural and technological schemes and designs of used devices, modes and parameters of filling the mold with the melt. All this is due to the features of the devices used and the improvement of their design. The high requirements for such devices have determined the need to create new devices designs to reduce the time spent on preparation for work and maintenance and to improve the quality of resulting metal billets. In scientific literature, including patents, more and more articles and materials are devoted to the development of new and improvement of the existing methods of supplying and stirring liquid metal in CCM and devices for their implementation. Experimental studies of liquid metal flow in CCM are a long, complex and laborious process. Therefore, mathematical modeling by numerical methods is increasingly used for this purpose. The authors have proposed a new technology for pouring liquid metal into a mold and a device for its implementation due to the use of effect of a deep-bottom submersible nozzle rotating in the mold with eccentric outlet holes. The purpose of this work is to simulate by proven numerical method a new process of filling a rectangular CCM mold with liquid steel and stirring it. Based on the developed numerical schemes and algorithms, a calculation program was compiled. The article describes an example of calculating the steel casting into a mold of rectangular cross-section and flow diagrams of liquid metal in it.