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The thermal and economic performance indices of a present-day low-capacity reactor facility with a boiling-water vessel reactor are presented. The operation of a reactor with natural coolant circulation is highly reliable in different operating regimes of a nuclear heat and power plant.

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Measurements of the propagation velocity of small perturbations of the neutron flux along the height of the core of a VK-50 boiling water vessel reactor are presented. The method of measurement is based on the evaluation of the delay time of a signal between the axially positioned sensitive elements of a moving double direct-charge detector. The measurements are compared with calculations for the parameters of two-phase coolant in the measurement channels. It is shown that the agreement between the measurements and the calculations is best for the velocity of an interphase surface.Correlation methods of measuring coolant flow, which are based on evaluating the delay time of the signal between the axially positioned sensors, are widely used in in-core systems monitoring boiling-water vessel reactors. One of the most widely used schemes for such measurements is to determine the mutual probabilistic characteristics of neutron flux fluctuations using stationary local neutron intensity sensors based on, for example, direct-charge detectors [1,2].Another variant is to use mobile paired detectors separated by a relatively small distance. Such sensors make it possible to determine the dependence of the propagation velocity of a neutron flux perturbation over the height of the core from the local flow structure of the steam-water mixture. The results of height measurements with in-core monitoring of the energy release in VK-50 are presented in the present article [3].Review of Foreign Studies. The cross-correlation function of the currents of sensors distributed over height was first used to determine the parameters of the steam-water flow in the core of boiling-water vessel reactors at the beginning of the 1970s, when publications appeared concerning measurements performed on the Lingen (West Germany) [4], Garigliano (Italy) [5], Fukushima-1 (Japan) [6], and other reactors. These studies were based on a methodology presented in [7] and developed in subsequent publications, for example, [8,9].A two-component model of neutron-flux noise was examined in [7]. The local component of the noise was attributed to the generation and motion of steam bubbles, while the global component was due to the general effect of the coolant density on the reactivity of the system. Hence it is natural to suppose that the velocity determined by means of the cross-correlation function of the currents is the velocity of the steam phase, as done in the very early studies [10].