Conference Agenda
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Session Overview |
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Tech. Session 12-7. MMR - IV & GCR - III
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9:00am - 9:25am
ID: 1814 / Tech. Session 12-7: 1 Full_Paper_Track 6. Advanced Reactor Thermal Hydraulics and Safety Keywords: HTTF, Lower Plenum, HTGR, STAR-CCM+, URANS Grid Independence Test on the Lower Plenum Mixing Test of the High Temperature Test Facility Benchmark NRG PALLAS, Netherlands, The To facilitate the deployment of High Temperature Gas-cooled Reactors (HTGRs), modeling and simulation tools that have been validated for such systems are required. The most common methods for HTGR systems analysis are lumped parameter System Thermal Hydraulic (STH) codes that were originally developed and validated for Light Water Reactors (LWRs). The Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD/NEA) is currently administering a benchmark that provides a set of Verification and Validation (V&V) problems and exercises using high quality experimental data from the Oregon State University’s (OSU’s) High Temperature Testing Facility (HTTF), a 1:4 scaled Integrated Effects Test (IET) of the General Atomics’ (GA) MHTGR design. The OECD/NEA benchmark consists of three separate problems to be analyzed, one of which is the Lower Plenum (LP) mixing exercise. This problem can be tackled in two cases: a code-to-code comparison study with fixed boundary conditions mimicking the full power conditions of the experiment and a code-to-experiment comparison study with best estimate boundary conditions, the former being the focus of current efforts. Previous articles have respectively showcased the time independence study on a coarse mesh and the results of a medium resolution mesh. The current article presents the grid-independence study using three mesh sizes. All cases use Unsteady-RANS (URANS) solvers employing the Realizable K-Epsilon turbulence model as available in the commercial code Simcenter STAR-CCM+. The results show that although not all considered points converge, the obtained Grid Convergence Index (GCI) is quite low. 9:25am - 9:50am
ID: 1332 / Tech. Session 12-7: 2 Full_Paper_Track 6. Advanced Reactor Thermal Hydraulics and Safety Keywords: SFR, MOOSE, Subchannel, SCM, SAM Multiscale Thermal-hydraulic Analysis of the MARVEL Micro-reactor Using Coupled MOOSE Subchannel (SCM) and SAM INL, United States of America MARVEL is a natural-convection-cooled sodium-potassium microreactor that is anticipated to generate 85 kilowatts of thermal energy. It will operate within Idaho National Laboratory’s Transient Reactor Test Facility and is being developed by the DOE Microreactor Program. MARVEL will be used to test microreactor applications, generate operational data, and pave the path for commercial demonstrations. A thermal-hydraulic computational model of this facility is a valuable tool to study important transients and calculate the safety limits of the micro-reactor design. For this purpose, the authors propose to use a multiscale coupled simulation: SCM for modeling the reactor core and SAM for the reactor’s primary cooling system. SCM is MOOSE physics module for subchannel analysis, which was designed to model single-phase flows through liquid-metal cooled, wire-wrapped fuel pin sub-assemblies, ordered in a triangular lattice. The SCM code was modified to be able to model MARVEL’s unique geometry. SAM is a systems analysis module based on the MOOSE framework. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for various advanced reactor types. The coupling between the two SCM and SAM for MARVEL modeling is done implementing a domain over-lapping approach. The resulting coupled simulation can model transients such as reactor startup/shutdown and provide an intermediate fidelity picture of the temperature field and other variables, in the core. Results for the steady-state simulations are presented in the article as well as flow blockage transient. 9:50am - 10:15am
ID: 1261 / Tech. Session 12-7: 3 Full_Paper_Track 6. Advanced Reactor Thermal Hydraulics and Safety Keywords: Heat Pipe-Cooled Micro Modular Reactor, Supercritical CO2 (sCO2), Air, Brayton cycle Evaluation of Air and sCO2 Brayton Cycle for Heat Pipe-Cooled Micro Modular Reactor University of Stuttgart, Germany Micro Modular Reactors (MMRs) present a promising solution for decentralized power generation, particularly in remote areas. Among the various designs under investigation, Heat Pipe-Cooled MMRs (HP-MMRs) have gathered significant interest. As power conversion unit (PCU), two distinct cycles are being investigated: an open-air recuperated Brayton cycle and a supercritical CO2 (sCO2) recuperated Brayton cycle. The goal of this research is to provide a useful comparison at a system level between these two power conversion strategies, offering insights that could inform future design choices for HP-MMRs in off-grid applications. The thermodynamic modelling and optimization of the two cycles, employed as PCUs for a 5 MWth Heat Pipe-Cooled MMR, are investigated employing the system code ATHLET. The analysis focuses on the design of key system components, such as the Heat Pipe Heat Exchanger (HPHX), the recuperator, the ultimate heat sink (for the sCO2 case), and the turbomachinery. Preliminary findings suggest that the air cycle offers operational flexibility, can leverage the maturity of existing technologies from the power and aerospace industries, and does not require an ultimate heat sink. In contrast, the sCO2 cycle demonstrates advantages in terms of more compact turbomachinery and higher thermal efficiency. Additionally, the study explores potential control strategies and their feasibility for part-load operations, with the aim of enhancing system adaptability under variable load conditions. | ||
