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Session Chair: Changho Lee, Argonne National Laboratory-ANL, United States of America
Application of the Coupled Code ATHLET-ANSYS CFX for the Simulation of the Flow Mixing Inside the ROCOM Test Facility
Angel Papukchiev, Zhi Yang
Gesellschaft- für Anlagen- und Reaktorsicherheit (GRS) gGmbH
System thermal-hydraulics (STH) codes have been developed and applied in the last 40 years for the design and the assessment of nuclear power reactors. These tools are extensively validated against experiments and provide reliable results. Nevertheless, since STH are based on 1D lump parameter approach, their capabilities to model transients with 3D flow mixing phenomena such as boron dilution, main steam line break, etc. are very limited. Therefore, to close this gap, STH are coupled today with modern computational fluid dynamics (CFD) programs. In this multiscale coupling approach the regions in the computational domain, where complex 3D flow effects are present, are simulated with the 3D CFD code, while the rest of it is calculated with the 1D STH program in a very efficient way. The development and validation of such numerical tools received recently strong attention in numerous international projects like NURISP, THINS, SESAME. It was found that an extensive validation of the new multiscale programs is necessary in order to deliver reliable and accurate best-estimate solution. GRS coupled its STH code ATHLET with the CFD programs ANSYS CFX and OpenFOAM. Within the validation process the ROCOM PKLIIIT1.1 experiment was simulated. The ROCOM facility represents a scaled German Konvoi reactor and was used to perform investigations on boron dilution, pressurized thermal shock and other reactor safety related issues. In the selected PKLIIIT1.1 experiment ethanol is injected in the water-filled primary ROCOM circuit at quasi-stationary conditions leading to complex flow mixing phenomena in the whole primary circuit. In this work the simulated transient is analyzed and the ATHLET-ANSYS CFX multiscale results are compared with experimental data. Moreover, important modeling aspects and some lessons learned are discussed.
Preliminary Multiphysics Simulation of Heat Pipe Cooled Micro Nuclear Reactors Using PROTEUS / FLUENT / ANLHTP
Changho Lee1, Yeon Sang Jung1, San Lee2, Hyoung Kyu Cho2, Alex Levinsky3
1Argonne National Laboratory; 2Seoul National University; 3Westinghouse Electric Company
In recent years, concerns in compact power generation are re-emerging due to the growing demand of affordable and sustainable energy resources even in remote locations, military bases, etc. where electricity supply is limited. Multiphysics simulations of a heat pipe cooled micro reactor were performed using the MOC solver of PROTEUS for neutronics analysis, FLUENT for thermal analysis, and ANLHTP for heat pipe performance analysis. The coupling system of the three codes was developed using the Python-based external drivers, which coordinate the overall workflow including data exchange (power, temperature, and heat transfer rate) required for the coupling and also control the individual calculation steps of the three codes such as convergence check and boundary conditions. In this study, FLUENT was used instead of ANSYS-mechanical because of the limited code availability, therefore no effect of structural change was accounted for. 3D test cases composed of 6 fuel rods and 7 heat pipes in the hexagonal lattice configuration were generated based on the MegaPower geometry and composition. Steady-state and transient simulations were performed using PROTEUS/FLUENT/ANLHTP. For transient simulations, one (a heat pipe at the center or outside) out of 7 heat pipes in the test cases was made fail to find temperature and power changes with time. ANLHTP was updated to analyze a slow transient using the steady-state model, assuming that the vapor core has relatively negligible thermal inertia and quick response to the change of the wick-vapor interface temperature. The components of a heat pipe (heat pipe container, wick and liquid in the wick) that have relatively large thermal inertia were modeled using FLUENT. The transient simulation results from the coupled system showed that the power and temperature changes due to a heat pipe failure were qualitatively reasonable. Details will be presented in the full paper.
Alternative Moderators for Modular High Temperature Gas-Cooled Reactors
Edward Duchnowski1, Nicholas Robert Brown1, Lance Snead2, Jason Trelewicz2
1University of Tennessee, Knoxville; 2State University of New York at Stony Brook
The modular High Temperature Gas-Cooled Reactor (mHTGR) is a thermal nuclear reactor that typically consists of a helium cooled, graphite moderated system. Graphite-moderated HTGR systems are large, with low power density, and are therefore costly to build and operate. The Advanced Research Projects Agency-Energy (APRA-E) under the Modeling-Enhanced Innovations Trailblazing Nuclear Energy Reinvigoration (MEITNER) Program aims to identify alternatives to graphite moderators that will enable transformational design improvements for mHTGRs. These alternative moderator concepts focus on beryllium and hydrogen for their increased moderating power relative to graphite. The alternative moderators have been assessed from a reactor physics stand point and parameters including reactor cycle length, neutron flux spectra, few-group moderator cross section, critical size, and discharge burnup were evaluated in this paper. These preliminary results show comparable performance for unoptimized cases and improved performance for optimized lattice pitch cases. The proposed alternative moderators have the potential to minimize the core size of mHTGRs without compromising on performance or safety.