Monte Carlo and nodal neutron physics calculations of the IAEA MTR benchmark using Serpent/DYN3D code system

• The adequacy of applying Serpent and DYN3D codes to an MTR reactor core is assessed.
• The Serpent code capabilities as a lattice code for MTR plate-type fuel are evaluated.
• The DYN3D code capabilities in modeling full 3D MTR cores are evaluated and compared.
• Serpent/DYN3D static and burnup calculations are thoroughly compared with other codes.
• The Serpent/DYN3D results are consistent with MCNP5 and OpenMC results for MTR core.

As part of recent efforts to utilize NPPs computational methodologies to safety analysis of research reactors, the Serpent and DYN3D codes were extensively compared with a variety of static and burnup calculations as defined in the IAEA benchmark for 10 MW MTR pool-type reactor. These calculations include unit cell calculations and few group constants generation, unit cell and full core k-eigenvalue and burnup calculations, and full core 3D flux and power distributions. The Serpent code capabilities as a lattice code for MTR plate-type fuel assemblies were evaluated and compared with EPRI-CELL and WIMS-D4 results and reference solutions for full 3D core models were compared with MCNP5 and OpenMC results. The DYN3D nodal diffusion code capabilities in modeling full 3D MTR cores were also evaluated using few group cross sections and assembly discontinuity factors obtained by Serpent unit cell calculations. The DYN3D results were compared with Serpent, MCNP5 and OpenMC.