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Thermohydraulic studies of reactor facilities with fast reactors are complex experimentally and computationally. Extensive experimental data are obtained on the velocity and temperature profiles, hydrodynamic resistance and heat emission, initial heat section, and interchannel mixing of the coolant in the fuel assemblies. These are used to develop engineering methods of performing thermohydraulic calculations of fuel assemblies as well as computational compute codes. The particulars of the hydrodynamics and heat transfer in intermediate heat exchangers and steam generators of reactor facilities with fast reactors are studied. This has made it possible to validate their thermohydraulic characteristics.Thermohydraulic analysis is a very important avenue for validating the parameters of reactor facilities that pertain to neutron physics, thermomechanics, structural strength, and others. The first-loop tank contains the reactor core and the breeding zone, the pumps and intermediate heat exchangers, the safety-and-control system column, and other units. The tank contains zones of three-dimensional flow of the coolant with different flow intensity, including stagnant zones, zones of stream flow, and by-pass flows. The large heating with relatively low coolant velocity requires taking account of the buoyancy contribution.The development of the head unit BN-1200 is based on maximum use of tested and scientifically validated technical solutions implemented in the BN-350, -600, -800 designs and new technical solutions which increase safety and provide high cost-effectiveness. These include placement of the sodium systems and first-loop equipment in the reactor tank, a flexible layout of the structural elements of the core, an emergency heat-removal system with built-in heat exchangers, vessel or large-modular steam generators. Correspondingly, maximum use will be made of scientific resources in the thermophysical validation of BN-1200.The integrated layout determined the complex of thermohydraulic studies for BN-600 and -800: 1) hydrodynamics of the flows of liquid-metal coolant in the channels, fuel assemblies, intertube space of the intermediate heat exchangers and steam generators, mixing chambers, and distribution collector systems;2) heat emission in tubes, fuel assemblies with liquid metal coolant; 3) interchannel exchange in fuel-element bundles; and 4) hydrodynamics and heat transfer in sodium-sodium heat-exchangers and sodium-water steam generators. Core. A high efficiency of a nuclear power plant with a fast reactor is attained by means of the high temperature of the steam and, naturally, high sodium temperature (up to 550°C as compared with 300°C in VVER) [1]. The high heating of the sodium (ΔT ƒ ) as compared with the low wall-liquid temperature head (ΔT α )

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.