Filonov V. Prediction of the regimes with deteriorated heat transfer in perspective IV generation reactors with supercritical coolant parameters

The dissertation work is devoted to the development of special procedures and tools for deterioration heat transfer modes assessment in the core of advanced generation IV reactors based on experimental data. Foreword represents the substantiated relevance of the special procedures development. The general characteristics of the work are given. Its purpose, main objectives, object and subject of research are also formulated. At first chapter a critical review of the current state of research on heat transfer at supercritical coolant parameters is given. The main physical aspects of heat transfer deterioration (HTD), as well as the complexity of structural flow evaluation are considered. The aim and objectives of the study are formulated based on the literature review. The second chapter is devoted to the adaptation of the transfer matrix method (TMM) for the analysis of nonlinear thermo-hydraulic processes for the coolant at supercritical coolant parameters (SCWP). The proposed method actually removes the restrictions on the type of correlations for the Euler and Nusselt numbers and has improved stability, and can be interpreted as the basis of modern codes of thermal hydraulics. The third chapter is assigned to extend the capabilities of the 1-D thermal hydraulics methods, which are described in detail in the second section by introducing differential functions for determining the processes of dissipation intensity and heat transfer. There is a logical transition from the governing equations of the 1-D approach to a 2-D axisymmetric formulation in the form of a "narrow channel" model. An alternative form for the functional dependence of tangential stresses is proposed, which allows increasing the stability of the method. Due to the application of the concepts of "base" and "correcting value", which were introduced in the second section, it was possible to construct a solution for the obtained system in the sum of a series, the decomposition coefficients of which are determined by an effective numerical procedure. The presented validation results show that the proposed approach allows predicting the features of the flow structure under the HTD, with a significant reduction in computational resources compared to CFD. Fourth chapter represents the problem of adaptation of existing universal or specialized tools of thermal-hydraulic analysis for nonlinear problems of heat transfer at SCWP with HTD. A simple method of adaptation of the two-zone wall temperature Kader function based on existing probe studies for carbon dioxide is proposed. The problem of implementation in universal CFD computational fluid dynamics packages is discussed, which is based on the method of choosing the wall zone reference coordinate for determining the dynamic velocity and dimensionless temperature. The example of ANSYS CFX shows one of the ways to create a special user procedure, which has an improved tendency to predict the axial temperature profile of the deteriorated heat transfer. In this section, the calibration and validation of the obtained results based on experimental studies for vertical tubes and rod assemblies of fuel element simulators is carried out. The peculiarities of the proposed implementation are also discussed, and recommendations for the application and further improvement of engineering approaches for predicting the HTD at SCWP are formed. Fifth chapter allocated to the development of special tools for prediction of heat transfer modes with supercritical parameters under nuclear heating conditions. For this purpose, the coupling of the thermohydraulic part described in Sections 2 and 3 was performed by combining the pressure field and setting the integral flow characteristics with the neutron-physical problem. In order to optimize the coupled calculations, parametric profiles of energy release were formed using MCNP4C and connected with the WIMS5b cell code. In this section, the peculiarities of the heat transfer modes penetration under nuclear heating conditions are considered, as well as the influence of the DHT form on the criticality of the system. Sixth chapter represents the results of steady-state assessment of the advanced ECC-SMART reactor using the approaches described in Chapters 2-5. The equivalent thermal-hydraulic scheme for preliminary estimation of the turbulent Prandtl number and energy release in the fuel assembly are build. The issues of implementation of transfer coefficients and energy release in the near-wall region based on a special near-wall function (Chapter 4) and conjugate code estimates (Chapter 5) are considered. The applied methods made it possible to reduce the discretization of the full CFD model of the advanced reactor by tens of times, where the reactor core flow part is performed quite accurately.