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Automatic power control is one of the basic problems in the operation of nuclear power reactors. Its successful solution depends on the availability of accurate and reliable information of the power and energy release distribution in the reactor core. In principle, a nuclear reactor can be controlled by measuring its power by heat-engineering methods, the 7 radiation of the fission products, and the produced neutron flux.The principal drawbacks of the heat-engineering method is its slow response and limited measuring range, which are respectively determined by the time required for thermal equilibrium to be established during transition periods and by small temperature differences of the coolant at low reactor powers. Because of these difficulties, the heat-engineering method is rarely used for automatic control, but is widely employed as a reference method for calibration of other reactor power control systems and for providing an auxiliary signal for the reactor control system [1, 2].Power measurement by means of 7 detectors is limited by the slow response of such detectors due to the kinetics of 7 quanta emission. Emelyanov et al. [3] suggested a method of correcting the response of 7 detectors according to which the time dependence of a T-quanta flux is represented by a sum of exponents with appropriate coefficients. Unfortunately, both the exponents and coefficients depend on the reactor type and on the location and structure of detectors, and must be found separately for each specific case; moreover, even a small change in the detector location necessitates recalculation of these quantities which always involves much computation and experimental work. Because of these reasons the suggested method of improving the response of T detectors is rarely used in practice.The neutron method of power monitoring and control [1, 2] is most widely used because it offers the additional advantage that the parameter directly acted upon by control devices is the neutron flux density. The present use of neutron-current ionization chambers located outside the reactor core as input elements for automatic control systems provides no unambiguous relation between the reactor power and detector readings because of the possible "shading" of the detectors by control elements and because of considerable neutron-field gradients in the reactor. Actually, the operation of detectors located outside the core is based on leakage neutrons. In case of a large core, and in particular when the coefficient of reactivity is positive, there is a vital need in core control which is impossible to accomplish with detectors located outside the core.This makes it very desirable to design automatic reactor control systems based on intracore neutron detectors. Such a method makes it possible to combine the intracore monitoring sysfem of energy release distribution and the automatic reactor control system (which would reduce the total number of neutron detectors and make the system less expensive), to obtain objective and more accurate data on the tota...

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The construction of new types of fuel elements requires that reactor tests be carried out on both the fuel components and on experimental half-scale fuel elements. The channel being described is intended for verifying, under the conditions of the reactor of the world's first nuclear power station, new technical solutions for the fuel elements of the Beloyarsk nuclear power station and the Bilibinsk nuclear heat and electric power plant. The positive experience accumulated on the basis of tests of unique experimental fuel elements has enabled an experimental channel to be constructed which, when loaded with fissile material, is identical with the regular-fuel-element channel and can be used in place of it.The channel is designed for a regular cell of the reactor of the world's first nuclear power station and has the same subcoupling dimensions. In the construction of the channel, certain improvements are taken into account which were adopted in the channels of the Beloyarsk nuclear power station and the Bilibinsk nuclear heat and electric power station, and directed at increasing their efficiency.The experimental channel {Fig. 1) consists of two fuel elements, descendingand ascending tubes, installed in metal and graphite sleeves, which form a cylinder with diameter 64 mm for the graphite sleeve and 63 mm for the steel components. The length of the channel is 6.6 m.

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