E. V. Usov

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.

Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code

Experiments with impulse gas injection into model coolants, such as water or the Rose alloy, performed at the Novosibirsk Branch of the Nuclear Safety Institute, Russian Academy of Sciences, are described. The test facility and the experimental conditions are presented in details. The dependence of coolant pressure on the injected gas flow and the time of injection was determined. The purpose of these experiments was to verify the physical models of thermohydraulic codes for calculation of the processes that could occur during the rupture of tubes of a steam generator with heavy liquid metal coolant or during fuel rod failure in water-cooled reactors. The experimental results were used for verification of the HYDRA-IBRAE/LM system thermohydraulic code developed at the Nuclear Safety Institute, Russian Academy of Sciences. The models of gas bubble transportation in a vertical channel that are used in the code are described in detail. A two-phase flow pattern diagram and correlations for prediction of friction of bubbles and slugs as they float up in a vertical channel and of two-phase flow friction factor are presented. Based on the results of simulation of these experiments using the HYDRA-IBRAE/LM code, the arithmetic mean error in predicted pressures was calculated, and the predictions were analyzed considering the uncertainty in the input data, geometry of the test facility, and the error of the empirical correlation. The analysis revealed major factors having a considerable effect on the predictions. The recommendations are given on updating of the experimental results and improvement of the models used in the thermohydraulic code.

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification

The article presents information on the validation and verification (V&V) of the first version (V1) of the EUCLID integrated code intended for safety analysis of operating or designed liquid metal (sodium, lead, or lead–bismuth) cooled reactors under normal operation and under anticipated operational occurrences by carrying out interconnected neutronics, thermal–mechanical, and thermal–hydraulic calculations. The list of processes and phenomena that have to be modeled in the integral code for correctly describing the above-mentioned operating conditions is given. Based on this list, the most high-quality experimental data are selected for carrying out the validation. It is shown that, for sodium cooled reactors, a significant number of experiments was carried out around the world on studying individual thermal–hydraulic processes and phenomena, which made it possible to perform validation of the thermal–hydraulic module. The validation of the code—as applied to description of processes that take place in fuel rods with oxide or nitride fuel and gas gap—is carried out against the results of post-pile investigations of fuel rods irradiated in fast sodium cooled research and power-generating reactors. The obtained results opened up the possibility to determine the errors of calculating such fuel rod parameters as release of gaseous fission products from the fuel and sizes of pellet and cladding in a limited range of burnup values. To perform validation of the neutronics module as applied to calculation of such parameters as power density distribution over the core and decay heat release, a sufficient number of experiments and benchmarks were selected. The results obtained from experimental operating conditions of a BN-600 reactor and startup conditions of a BN-800 reactor made it possible to estimate how correctly the integral code performs calculations of interconnected thermal–hydraulic and neutronic processes. Only a limited set of experimental investigations is available for heavy liquid metal cooled reactors. In view of this circumstance, programs for obtaining the lacking data are developed. To estimate the quality with which the experiments are modeled by means of the EUCLID/V1 integrated code, a procedure for evaluating the errors of calculation results is developed. In accordance with this procedure, the error of calculating the parameters playing the main role in the reactor safety assessment is evaluated.

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.

Numerical analysis of experiments with gas injection into liquid metal coolant

Presented paper contains results of a numerical analysis of experiments with gas injection in water and liquid metal which have been performed at the Institute of Thermophysics Russian Academy of Science (IT RAS). Obtained experimental data are very important to predict processes that take place in the BREST-type reactor during the hypothetical accident with damage of the steam generator tubes, and may be used as a benchmark to validate thermo-hydraulic codes. Detailed description of models to simulate transport of gas phase in a vertical liquid column is presented in a current paper. Two-fluid model with closing relation for wall friction and interface friction coefficients was used to simulate processes which take place in a liquid during injection of gaseous phase. It has being shown that proposed models allow obtaining a good agreement between experimental data and calculation results.

Development and Verification Models of Vertical Stratification, Dryout and Slug Boiling of Superheated Sodium for LMFBR Safety Analyses

A new model which describes the dynamics of a vertically stratified flow correctly within the limits of a single-pressure two-fluid model has been developed. The model is based on the modification of finite-differences of convective terms and pressure gradients taking into account a distinct interface.We propose to use the vapor quality as a criterion for the onset of dryout. The choice of the criterion is based on the analysis of experimental and theoretical studies. To determine the boundary vapor quality we used the correlation xcr = 1.26·G0.2, which was found from experimental data fit.A review of articles has shown that for today it is impossible to predict correct superheat value. Therefore the superheat value was determined as a parameter of the model from the experimental data of a particular simulated experiment. Thus a boiling up regime was selected. The model described in this paper allows us to calculate the boiling up of sodium under the superheat conditions as well as problems of the evolution of the vapor volume.The verification of the models was done by using the SOCRAT-BN code [1]. SOCRAT-BN is a coupled code which consists of modules for calculation of damage and melting of a reactor’s core, thermohydraulic processes and neutron physics.The models of vertical stratification, dryout and slug boiling of superheated sodium are described in details in this paper. Also we present the results of verification for the models within analytic tests and experimental data.Copyright

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