In new Russian NPP with VVER reactor in the event of LOCA, provision is made for the use of passive heat removal system for necessary core cooling. In the case of leakage in the primary circuit this system assures the transition of steam generators to operation in the mode of condensation of the primary circuit steam coming to steam generator piping from the reactor. As a result, the condensate from the steam generators arrives to the core providing its additional cooling. The steam generator off-design condensation mode has the following features: undeveloped nucleate boiling on the horizontal tubes heated by condensing steam; natural circulation processes in the both steam generator circuits; low heat fluxes and temperature differences. The experimental study of undeveloped nucleate boiling on the single horizontal tube heated by condensing steam has been carried out in the Institute for Physics and Power Engineering. The experiments have been carried out on the GROT test facility. The heart of the test facility is the single VVER steam generator tube (length l = 10.2 m, outer diameter D = 16 mm, wall thickness d = 1.5 mm). The tube is fabricated of the original stainless steel 08Cr18Ni10Ti. The length and geometry of the test tube corresponded to that of real steam generator. The test facility was equipped with thermocouples enabling the temperatures of primary and secondary facility circuits to be controlled. The experiments were carried out at three heating steam pressures Ps1: 0.21, 0.35, 0.55 MPa. The main task of the research was to study the pressure effect on the process of undeveloped nucleate boiling on the single horizontal tube. On the base of the results of these experiments the empirical correlations for prediction of heat transfer coefficient and heat flux were obtained. The generalizing empirical correlations obtained can be used for the substantiation of work heat-exchanging equipment of NPP with VVER reactor in the condensation mode, and also can be applied to the verification of computer codes.

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The article presents the results of experimental studies of the thermophysical (density, viscosity) and physicochemical (degree of acidity - pH) properties of water solutions of boric acid. A review of the available literature data on the effect of the properties of boric acid solutions on heat removal from the reactor core is presented. It has been established that the available information is very general and does not cover the entire range of parameters (temperature, pressure, acid concentration) characteristic of a possible emergency at a nuclear power plant with a VVER reactor. Methods and facilities for conducting experimental studies are described. The results of experimental studies are presented. The density of aqueous solutions of boric acid with a concentration of 2.5-450 g/kg H2O at a temperature of 25-130 °C was determined. The dependence of the investigated characteristic on temperature and concentration was also obtained. The results of an experimental study of the kinematic viscosity of water solutions of boric acid with a concentration of 2.5-200 g/kg H2O at a temperature of 25-90 °C were obtained. The total error in measuring the viscosity of aqueous solutions of boric acid did not exceed 2 %. The pH values of water solutions of boric acid in the temperature range 25-50 and concentrations of 2.5-450 g/kg H2O were determined. The dependence for calculating the degree of acidity of boric acid is obtained. Experimental data on the thermophysical and physicochemical properties of water solutions of boric acid can be used to refine the results of calculations of emergency heat removal processes in a reactor facility, carried out using both one-dimensional calculation programs and three-dimensional CFD codes.

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