Calculations of a steady state of the OECD/NRC PWR MOX/UO2 transient benchmark with DYN3D

Steady states and transients in thermal reactors with standard core loadings are simulated with the nodal codes based on the three-dimensional diffusion equation for two energy groups of neutrons. The assembly wise cross sections are generated with cell codes solving the transport equation for many energy groups in two-dimensions assuming reflecting boundary conditions. Loading cores with higher content of MOX fuel and the increase of the fuel cycle length is challenging for the standard methods. Different neutron spectra of UO2 and MOX fuel, inhomogeneities inside the fuel as-semblies arising during the exposure in the reactor lead to discrepancies which needs an improvement of the applied methods. The core model DYN3D which is ap-plied for three-dimensional analyses of thermal reactor cores with quadratic or hex-agonal fuel assemblies is based on standard methods. The first step of the code enhancement consists in the extension to multi energy groups. The OECD/NEA and U.S. NRC PWR MOX/UO2 Core Transient Benchmark which is a well defined problem with a complete set of data is used as a means for verification of the modifi-cations of DYN3D. The core chosen for the simulation is based on four-loop West-inghouse PWR power plant similar to the reactor used for Plutonium disposition in the U.S. Steady-state calculations with the non-consideration and consideration of the assembly discontinuity factors (ADF), with a different number of energy groups, and with different number of nodes per assembly are compared with a reference so-lution generated by the DeCart. A Two-group solutions is also compared with a published PARCS diffusion solution.