CFD analysis of the Lower Plenum mixing test of the high temperature test facility benchmark by NRG PALLAS and OSU
To facilitate the deployment of High Temperature Gas-cooled Reactors (HTGRs), often designed as Small Modular Reactors (SMRs), validated modelling and simulation tools are essential. While traditional lumped parameter System Thermal Hydraulic (STH) codes, developed originally for Light Water Reactors, are common, increase in computational power has led to the rise of Computational Fluid Dynamics (CFD) but this requires rigorous validation. The OECD/NEA addresses this by administering a comprehensive benchmark, utilizing high-quality experimental data from the Oregon State University (OSU) High Temperature Testing Facility (HTTF), a 1:4 scaled Integrated Effects Test (IET) of the General Atomics’ (GA)-MHTGR design. This benchmark includes the Lower Plenum (LP) mixing problem, offering both code-to-code and code-to-experiment comparison cases. The current article presents and compares the attempts by NRG PALLAS and OSU towards the code-to-code study, where the boundary conditions try to mimic the full power conditions of the experiment. Both institutions independently use Unsteady-RANS (URANS) solvers as available in the commercial code Simcenter STAR-CCM+ while differing in the modelling approaches. The qualitative and quantitative results demonstrate good overall agreement between their simulations, yet also reveal specific areas of disparity in the predicted temperatures and velocities, providing valuable insights for HTGR V&V efforts.
Authors
- Ben Nakhnikian-Weintraub (ORCID: https://orcid.org/0009-0006-9388-1733)
- Izabela Gutowska
- Ferry Roelofs (ORCID: https://orcid.org/0000-0002-9685-7054)
- Mazhar Hussain (ORCID: https://orcid.org/0009-0003-7759-8602)
Institutions
- Oregon State University (US)
- Nuclear Research and Consultancy Group (NL)
Publication Details
- Journal
- Nuclear Engineering and Technology
- Published
- 2025-11-03
- DOI
- https://doi.org/10.1016/j.net.2025.104008
- Primary Topic
- Nuclear Engineering Thermal-Hydraulics
- Type
- article
- Field-Weighted Citation Impact
- 0.00