ONIX: An open-source depletion code

Inter-Code Comparison of Computational VERA Depletion Benchmark Using OpenMC, OpenMC-ONIX and DRAGON

“…Figures 9-10 show a comparison of multi-group scalar neutron flux among OpenMC, ONIX, and DRAGON with the two different energy structures. OpenMC and ONIX calculated neutron flux is obtained with pre-defined multi-group (47-groups) energy structure as stated in the benchmark problem [9] whereas DRAGON computed neutron flux obtained using SHEM-361 energy group structure. The results reveal that neutron flux in the thermal energy region is well agreed upon in both models obtained using OpenMC, ONIX, and DRAGON codes.…”

“…Fuel burnup or depletion analysis plays a vital role on both research and power nuclear reactor since reactor safety, cost-effective utilization and core lifetime are depending on fuel burnup/depletion process during operation. The objective of this study is to validate the open source codes OpenMC, ONIX-OpenMC [9] and DRAGON against the PWR fuel assembly based on VERA depletion/burnup benchmark problems for the first time. The purpose of burnup calculation is to determine the characteristics of the core that vary with time and improve the safety and performance of the core.…”

“…Burnup equations that describe the time rate of change of isotopic composition form a system of first-order linear differential equations. Chebyshev Rational Approximation Method (CRAM) is the currently most widely used method due to its stability for solving the equations with many isotopes and large time steps [9].…”

“…The VERA depletion benchmark problems [10][11] were proposed by the Consortium for Advanced Simulation of Light Water Reactor (CASL) from Oak Ridge National Laboratory (ORNL), with detailed guidelines including the burnup chain, fission Q-value for a depletion analysis. The depletion calculations of the VERA depletion benchmark suite were performed using Monte Carlo codes OpenMC, ONIX [9] and deterministic code DRAGON. OpenMC calculation result is considered as a standard solution due to explicit geometry modeling capability in 2D and 3D, usage of continuous energy cross-sections and with only a limited number of approximations for physics interaction and six dimension phase of a neutron such as the location, energy, and direction.…”

“…OpenMC calculation result is considered as a standard solution due to explicit geometry modeling capability in 2D and 3D, usage of continuous energy cross-sections and with only a limited number of approximations for physics interaction and six dimension phase of a neutron such as the location, energy, and direction. To perform this study with OpenMC code, ENDF/B VII.1 data library was utilized to generate data set in HDF5 format for OpenMC at 565 K, 600 K, 900 K, and 1200 K temperatures utilizing NJOY2016, while DRAGLIB [9] format for DRAGON was processed using PyNJOY 2016 to obtain the cross section at specific temperatures. In this paper, 2D results of VERA depletion problems are reported.…”