Open AccessComparative Study on Fuel Assembly of Modular Gas-cooled Fast Reactor using MCNP and OpenMC Code
Author(s) -
Helen Raflis,
Muhammad Ilham,
Zaki Su’ud,
Abdul Waris,
Dwi Irwanto
Publication year - 2021
Publication title -
journal of physics. conference series
Language(s) - English
Resource type - Journals
SCImago Journal Rank - 0.21
H-Index - 85
eISSN - 1742-6596
pISSN - 1742-6588
DOI - 10.1088/1742-6596/1772/1/012031
Subject(s) - monte carlo method , nuclear engineering , neutron transport , fissile material , criticality , computer science , neutron , physics , nuclear physics , engineering , mathematics , statistics
The design study of GFR concepts comprises neutronic analysis of fuel pin and fuel assembly. The Monte Carlo method has advantages in three-dimensional (3D) geometry modeling but requires a high computation time. In this research, the comparative study of Gas-cooled Fast Reactor (GFR) using the Monte Carlo code. The GFR feasibility design study will be carried out with natural uranium with plutonium as fuel cycle inputs. The Monte Carlo method simulates GFR model at full-scale and heterogeneous three-dimensional (3D) using Evaluated Nuclear Data File (ENDF/B-VIII.b5) nuclear data. The code of Monte Carlo methods will be used in this research are the Monte Carlo N - Particle (MCNP) and OpenMC. The comparison of the GFR fuel assembly calculation simulation results is made between the MCNP and OpenMC code. The equilibrium cycle configuration is used as the basis model for the comparisons. The comparison of MCNP and OpenMC code gives a good agreement in criticality calculation of GFR that achieves delta k inf less than 1%. The (U-Pu)N fuel is a good candidate to be chosen in GFR research that gives k inf more than 1.1 in fissile contain 10%Pu and has the highest thermal conductivity. The Zircaloy-4 is the best candidate for material cladding in GFR design that provides the highest k inf .