Coupled OpenMC/CTF to VERA Core Physics Benchmark Problem 6

Abstract

Abstract The multiple-physics modeling has been demonstrated to be a high-fidelity and effective method for the analysis of reactor core physics and thermal-hydraulics. OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. CTF is a subchannel thermal-hydraulics code designed for Light Water Reactor analysis. In this work, OpenMC and CTF are coupled for the analysis of light water reactor fuel assembly. OpenMC provides axial and radial fuel pin normalized power distribution to CTF, and CTF gives the fuel temperatures and coolant properties to the neutronics simulation of OpenMC. The windowed multipole temperature method is used in OpenMC to match the accurate temperature distribution of fuel rods and coolants obtained from CTF. In this study, the OpenMC/CTF is applied and validated using the Virtual Environment for Reactor Applications (VERA) core physics benchmark problem 6, from the Consortium for Advanced Simulation of Light Water Reactors. The problem involves a three-dimensional fuel assembly in Hot Full Power conditions. The Keff eigenvalue, pin power distribution, fuel temperatures, and coolant properties are obtained and compared with VERA’s reference results (MPACT/CTF). Our converged results showed good consistency with the reference solution, eigenvalue differences agreed within 183 pcm, axially-integrated normalized radial fission distribution difference agreed within +1.4%/−1.9%, local volume-averaged fuel pin temperatures agreed within +26.3°C, and local subchannel exit coolant temperatures agreed within +3.6°C/−0.4°C. These preliminary solutions prove that OpenMC coupled to CTF method has shown high-fidelity results in the three-dimensional fuel assembly neutronics and thermal-hydraulics coupled problems.