S. I. Lezhnin

System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment

The system of equations from a two-fluid model is widely used in modeling thermohydraulic processes during accidents in nuclear reactors. The model includes conservation equations governing the balance of mass, momentum, and energy in each phase of the coolant. The features of heat and mass transfer, as well as of mechanical interaction between phases or with the channel wall, are described by a system of closing relations. Properly verified foreign and Russian codes with a comprehensive system of closing relations are available to predict processes in water coolant. As to the sodium coolant, only a few open publications on this subject are known. A complete system of closing relations used in the HYDRA-IBRAE/LM/V1 thermohydraulic code for calculation of sodium boiling in channels of power equipment is presented. The selection of these relations is corroborated on the basis of results of analysis of available publications with an account taken of the processes occurring in liquid sodium. A comparison with approaches outlined in foreign publications is presented. Particular attention has been given to the calculation of the sodium two-phase flow boiling. The flow regime map and a procedure for the calculation of interfacial friction and heat transfer in a sodium flow with account taken of high conductivity of sodium are described in sufficient detail. Correlations are presented for calculation of heat transfer for a single-phase sodium flow, sodium flow boiling, and sodium flow boiling crisis. A method is proposed for prediction of flow boiling crisis initiation.

Development of a unified library of sodium properties

A new approach for describing the properties of sodium vapor is described. An equation for the state of sodium vapor is obtained as a dependence s(p, h), which can be used for engineering calculations of power installations with a sodium coolant.

Pressure Waves in Two-Phase Bubble/Slug Flows and Interphase Process

This paper considers the experimental results and physical effects on the pressure waves dynamics of vapour-liquid two-phase medium of bubble and slug structure. The general mechanisms of wave formation, behaviour and instability of two-phase flow under pressure waves, basic peculiarities of the interphase heat transfer is obtained. The role of destruction and collapse of bubbles and slugs, phase transition (condensation and evaporation), structural transition on wave dynamics is studied also. Experiments were carried out at the 52 and 8 mm ID tubes with Freon-11 and water as a test liquid. High-speed filming, laser optical methods, together with pressure history recording used in experiments. Theoretical models are suggested for wave evolution in media with nonuniform void fraction, which describe amplification of reflected waves and pressure pulses generation.

Experimental Simulation of Hydrodynamics and Heat Transfer in Bubble and Slug Flow Regimes in a Heavy Liquid Metal

For the confirmation of the claimed design properties of a reactor plant with a heavy liquid-metal coolant, computational and theoretical studies should be performed in order to justify its safety. As one of the basic scenarios of an accident, the leakage of water into the liquid metal is considered in the case of steam generator tube decompression. The most important in the analysis of such a kind of accidents are questions concerning the motion and heat exchange of steam bubbles in the steam generator and the probability of blocking the flow area owing to freezing coolant, since the temperature of boiling feedwater in the steam generator can become lower than the melting point of lead. In this paper, we present main approaches and relationships used for the simulation of the motion of gas bubbles and heat transfer between bubbles and liquid metal flow. A brief description of the HYDRA-IBRAE/LM computational code that can be used to analyze emergency situations in a liquid metal-cooled reactor facility is also presented. It should be noted that the existing experimental data on the motion and heat transfer of gas bubbles in a heavy liquid metal are insufficient. For this reason, in order to verify the HYDRA-IBRAE/LM code models, experiments have been performed on the cooling of liquid lead by argon and on the motion of gas bubbles in the Rose’s alloy. In particular, a change in the temperature of the coolant over time has been studied, and the void fraction of gas at different flow rates of gas has been measured. A detailed description of the experiments and a comparison of the results of the calculations with the experimental data are presented. The analysis of uncertainties made it possible to reveal the main factors that exert the greatest effect on the results of calculations. The numerical analysis has shown that the models incorporated into the HYDRA-IBRAE/LM code allow one to describe to a sufficient degree of confidence the process of cooling liquid lead melt when argon bubbles pass through it, simulating the flow of water into the liquid metal in the course of steam generator tube rupture.

Model of vapor slug growth in the channels of power engineering equipment with sodium coolant

The problem of vapor volume growth in superheated liquid is very important both in practical and in theoretical aspects as the behavior of the solitary bubble governs the general patterns and the proceeding character of the boiling process. On the basis of the analysis of experimental and theoretical works, we create the physical and the numerical model of vapor volume growth in superheated sodium in power engineering equipment channels. On that basis, we implement the program module, making it possible to calculate the sodium boiling process under superheating. We perform a numerical analysis of the dependence of the vapor volume growth rate on the superheating value, the liquid film thickness, and the coefficient of the heat transfer with the wall. The numerical modeling results are compared to experimental data on the boiling of the sodium in the column and with the data on the sodium boiling under a decrease of the mass flow rate in the contour.

Recommendations on adopting the values and correlations for calculating the thermophysical and kinetic properties of liquid lead

Recent years have seen an essentially increased interest in studying the properties of liquid lead, which is primarily connected with the possibility of using it as coolant in nuclear power installations, first of all, in reactors based on fission of heavy nuclei by fast neutrons. The article presents an analysis of published data on the thermophysical and kinetic properties of lead in liquid state, the results of which served as a basis for selecting and recommending correlations to be used in carrying out scientific and engineering calculations. A general assessment of the state of experimental investigations into the thermophysical properties of liquid lead is presented. The presented value of lead solidification temperature is the maximally reliable one. The data on the boiling temperature, melting and vaporization enthalpies, and saturated vapor pressure have been determined with satisfactory accuracy. The published data on the liquid lead heat capacity differ considerably from each other; therefore, the recommended values should be experimentally checked and determined more exactly. The available experimental data on surface tension density, volumetric expansion coefficient, sound velocity, viscosity, and thermal conductivity do not cover the entire range of liquid phase existence temperatures. The temperature region above 1200 K and the crystal-liquid phase transition region are the least studied ones. Additional investigations of these properties in the above-mentioned temperature intervals are necessary. The question about the influence of impurities on the thermophysical properties of lead still remains to be answered and requires experimental investigations.

Implementation of the library of properties of sodium vapor on the basis of the s (p, h) formulation in the thermohydraulic module of the SOKRAT-BN integral code

The paper presents the results of incorporating into the SOKRAT-BN code a new library of properties of sodium vapor on the basis of the s (p, h) formulation: the entropy of sodium vapor as a function of pressure and enthalpy. A double increase in the computational speed and a good agreement with the results of the previous version of the library of properties of sodium vapor are attained.

Coupled Code SOCRAT-BN Development for Safety Analysis of Sodium-Cooled Fast Reactors

SOCRAT-BN is a software package to simulate design and severe accidents of sodium-cooled fast reactors. The package consists of modules for calculating damage to the reactor’s core, thermohydraulic processes and neutron physics.The thermohydraulic module has been developed to calculate one- and two-phase flows in channels with different geometry and bundles. The module is based on a two-fluid model for equal pressures of phases.In this paper we present an explanation of the deciding constitutive models for equations used in the system. Validation of the module was performed on the experimental data for one- and two-fluid flows in complex geometry channels and on calculation of running a first loop of the reactor BN-600 in nominal mode.Copyright