Process of boric acid mass transfer during accidents accompanied with rupture of circulation pipelines in VVER reactors of new generation equipped with passive safety systems are examined. Results of calculation of variation of boric acid concentration in VVER-TOI reactor in case of accident development process are presented. Positive effects of boric acid droplet entrainment on the processes of acid accumulation and crystallization in the reactor core are demonstrated. The obtained results allow formulating the conclusion on the possibility of these processes in the reactor core which may lead to the disruption of heat removal from fuel pins. Review of available published reference data on physical properties of boric acid solutions (density, viscosity, thermal conductivity) is given. It is established that available information is of too general nature and fails to cover the whole range of parameters (acid temperature, pressure and concentration) typical for potential emergency situation on NPP equipped with VVER reactor. Necessity of experimental study of processes of droplet entrainment under parameters typical for VVER emergency operation conditions, as well as investigation of thermal physics properties of boric acid within wide range of acid concentration values is required.

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The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with VVER-1200 reactor) reactor core in the event of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed based on computational simulation of the related processes in the reactor and containment. The computational simulation has been performed with regard to the detrimental effect of non-condensable gases on steam generator (SG) condensation power. Nitrogen arriving at the circuit with the actuation of hydroaccumulators of the 1st stage and products of water radiolysis are the main sources of non-condensable gases in the primary circuit. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the course of emptying of the 2nd stage hydroaccumulators system (HA-2) the gas-steam mixture spontaneously flows out from SG cold headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculations carried out by different integral thermal hydraulic codes revealed that this volume flow rate of gas-steam mixture from SG to the HA-2 system would suffice to eliminate the “poisoning” of SG piping and to maintain necessary condensation power. In support of the calculation results, the experiments were carried out at the HA2M-SG test facility constructed at IPPE. The test facility incorporates a VVER steam generator model of volumetric-power scale of 1:46. Steam to the HA2M-SG test facility is supplied fed from the IPPE heat power plant. Gas addition to steam coming to the SG model is added from high pressure gas cylinders. Nitrogen and helium are used in the experiments for simulating hydrogen. The report presents the basic results of experimental investigations aimed at the evaluation of SG condensation power under the inflow of gas-steam mix with different gases concentration to the tube bundle, both under the simulation of gas-steam mixture outflow from SG cold header to the HA-2 system and without outflow. As a result of the research performed at the HA2M-SG test facility, it has been substantiated experimentally that in the event of an emergency leak steam generators have condensation power sufficient for effective heat removal from the reactor provided by PHR system.

The strategy for the development of nuclear power presupposes the construction of thermal water-cooled reactors and fast reactors [1][2][3]. At present, water moderated and cooled power reactors are the foundation of our country's nuclear power. The organizations of the State Corporation Rosatom have developed designs for new-generation reactor setups -AES-2006 and VVER-TOI, which are the result of evolutionary development and improvement of operating water moderated and cooled power reactors, which proved their reliability in thousands of reactor-years of accident-free operation. VVER are characterized by elevated power, the best economic and operational performance as well as enhanced safety compared with currently operating reactors. The thermal power is increased by, for example, using fuel assemblies with a new design incorporating heat-exchange intensifiers. The safety of the new designs of NPP with VVER is enhanced by means of the principle of technological diversity, which consists in combining active and passive safety systems. The passive systems ensure shutdown and prolonged removal of residual heat in the presence of a sealed loop or a depressurized loop and do not require operator intervention or an external long-time power source.To secure the possibility of increasing the power and validating the serviceability of new passive safety systems for VVER, a large-scale program of thermophysical and experimental studies has been completed.Increasing the Power Density and Efficiency of the VVER Core. The new NPP designs with VVER presuppose an improved design of the core as a whole and the fuel assemblies in particular. One promising avenue for increasing the capacity of power-generating units and the efficiency of the fuel cycle is to use fuel assemblies with improved thermohydraulic characteristics, which is accomplished by using mixing or intensifying lattices to increase heat transfer. Acting on the coolant flow these setups decrease the nonuniformity of the heating of the coolant (enthalpy) in the transverse section of a fuel assembly. They also make it possible to increase the turbulence properties of the flow, which causes moisture to settle on the walls of the fuel elements and promotes the removal of a large quantity of heat.In order to validate the heat-engineering reliability of a VVER core with fuel assemblies with mixing lattices, it is necessary to determine the effect of their design on the hydrodynamics and mass-transfer processes in the core. Important problems are the choice of the arrangement of the fuel assemblies and optimization of the design of the mixing lattices, which must possess the optimal combination of parameters such as the hydraulic resistance, the intensity of mixing of the coolant and a safety margin to crisis of boiling. The main method for studying the hydrodynamics and mass transfer in reactor fuel assemblies is the experimental investigation of fuel-assembly models on thermohydraulic stands: SVD-2, STF, and the TRASSER setup [4]. The main characteristics of the tripl...

Results obtained from a research work on experimentally substantiating the serviceability of the additional system for passively flooding the core of a VVER reactor from the second stage hydro accumula tors are presented.