V. F. Strizhov

A numerical study on natural convection of a heat-generating fluid in rectangular enclosures

Abstract Unsteady natural convection of a heat-generating fluid in the enclosures of a rectangular section with isothermal or adiabatic rigid walls is investigated numerically in the present work. Using the high-performance finite-difference scheme in the 2D stream function-vorticity formulation, developed by the authors, the peculiarities of convective heat transfer are studied in a wide range of thermal and geometric parameters for the laminar regime of fluid motion. Steady-state as well as oscillating solutions obtained in this work are compared with available numerical and experimental results of other researchers.

Challenges left in the area of in-vessel melt retention

The in-vessel melt retention becomes an important safety objective for the present or future middle power nuclear plants, so care has to be taken in the evaluation of the various phenomena related to ensuring the feasibility of this objective. Since the prediction of the relevant phenomena has to be performed for the prototypical accident conditions, the applicability of the measured data or of the correlations derived from these measurements have to be established and the uncertainties determined. In this context, most uncertainties are introduced by the non-prototypicalities in the experiments. The paper describes the major findings from the OECD RASPLAV project and discusses the remaining challenges left in the area of in-vessel molten corium coolability.

A semiquantitative theory of convective heat transfer in a heat-generating fluid

Abstract A semiquantitative theory of heat transfer in a heat-generating fluid within a closed volume is developed. The analysis is based on relationships derived from the condition of energy balance and on modern physical concepts of heat transfer processes in an energy-neutral fluid. Four main regimes and one asymptotic regime of heat transfer are singled out, which differ in the exponents of power in the expression relating the Nusselt number, Nu, and the modified Rayleigh number, RaI. In the asymptotic limit, with Ra I 1 32 ⪢ 1 , heat transfer to the upper horizontal part of the boundary and the curved part of the boundary facing downward obeys dependences Nu up ∼ Ra I 7 32 and Nu dn ∼ Ra I 1 4 , respectively. In the range of values of RaI which is of interest with regard to the safety problem of the nuclear power engineering the established theoretical correlations are in good agreement with the experimental data.

Development of a diffusion-inertia model for calculating bubble turbulent flows: Isothermal polydispersed flow in a vertical pipe

This work is dedicated to simulating polydispersed bubble turbulent flows. The work is aimed at development of an approach to simulating polydispersed bubble flows on the basis of the combination of a diffusion-inertia model with the delta-approximation method. The coagulation and fragmentation models of bubbles have been checked. The comparison of the simulation result with the experimental data has shown that the elaborated approach makes it possible to simulate the bubble polydispersed flows in wide range of gas content values.

A Model for Calculating Composition and Density of the Core Melt in the Water-Moderated Water-Cooled Reactor in Case of Severe Accident

Thermochemical behavior of the core melt in the VVER-type reactor at severe accident is discussed. Experimental information gained made it possible to construct a thermodynamic model of the O-U-Zr-Fe system. The model describes the immiscibility of the oxide and metal phases of the core melt and makes it possible to estimate their densities. Parameters of the model were obtained by comparison with diagrams of binary and ternary subsystems, as well as with results of the MASCA experiment. Additional verification calculations show good agreement of the theory with experiment. In particular, parametric calculations showed that there is a wide enough region of the core melt parameters at which densities of the oxide and metal phases are comparable in values, but this fact does not rule out the probability of stratification with upper position of the oxide layer. In its simplified formulation the model has been incorporated into the HEFEST code. Results are reported for calculations of heat transfer in the stratified core melt in the reactor vessel for two different cases of relative location of the melt layers.

Best Estimate Methodology for Modeling Bubble Flows

In this paper the numerical–experimental methodology of studying bubble flows is considered. The effective numerical CFD algorithms for 3D calculation of two-phase flows with explicitly chosen of interface boundary and calculation of surface tension forces are offered. The new high-precise experimental technique permitting to control and to describe 3D effects of bubble flows as average in time and space, so local instant is described. The main attention of this paper is focused on validation and verification of numerical model (before submitted at NURETH11 conference) as with usage of the tests, accessible from the literature and outcomes of experiments obtained in IT SB RAS (Novosibirsk).Copyright

Severe accident codes status and future development

New demands for acceptance of nuclear power require full deterministic evidence of nuclear power plants (NPP) safety. From this point of view, the role of precise deterministic analysis of NPP safety plays a very important role for both the existing and future generation of NPPs. Considering the current status of existing severe accident codes, one may conclude that their capabilities are quite limited and not sufficient to prove NPP safety. This conclusion is based on the experience of usage of these codes, analysis of models and experimental database supporting codes and used for their validation. At the same time, the modern level of development of computer techniques and numeric methods allows the use of equations based on first principles rather than correlation. The transition to physical modeling appears to be more effective in the cases of designing and validation of codes using both separate effect and integral tests, and allows predictive power of codes to be increased and the range of uncertainties to be reduced. Moreover, physical modeling allows critical points of models and codes to be understood, and permits the planning of integral tests to resolve severe accident and accident management issues.

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.

Nonlinear Algebraic Reynolds Stress Model for Two-Phase Turbulent Flows Laden with Small Heavy Particles in Circular Tube

The purpose of the study is to present an explicit algebraic Reynolds stress (nonlinear turbulent viscosity) model combined with modified k – e turbulence model taking into account particles effect on turbulence for calculating the main turbulent characteristics of two-phase flows. For calculating particles distribution in space we used diffusion-inertia model (DIM). The turbulence attenuating in the presence of particles is clearly observed, investigated and compared with the experimental data. The developed model adequately described turbulence anisotropy and the influence of particles inertia and concentration on the turbulence intensity.

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.