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

Keyphrases

• Monte Carlo 100%
• Serpent 100%
• DYN3D 100%
• Coding System 100%
• Neutron Physics 100%
• MTR Benchmark 100%
• Full Core 50%
• Burnup Calculation 50%
• Unit-cell Calculations 50%
• OpenMC 50%
• MCNP5 50%
• Research Reactor 25%
• Nodal Diffusion Code 25%
• Power Distribution 25%
• Discontinuity Factor 25%
• Few-group Cross-section 25%
• Lattice Codes 25%
• Eigenvalue Calculation 25%
• Safety Analysis 25%
• Flux Distribution 25%
• Computational Methodology 25%
• Pool-type Reactor 25%
• Generation Units 25%
• Few-group Constants 25%
• Static Calculation 25%
• Group Constant Generation 25%
• K-Eigenvalue 25%
• Plate Fuel Assembly 25%
• IAEA Benchmark 25%
• Serpent Code 25%
• 3D Flux 25%

Engineering

• Burnup 100%
• Research Reactors 50%
• Cross Section 50%
• Power Distribution 50%
• Fuel Assembly 50%
• Reference Solution 50%
• Nuclear Power Plant 50%
• Eigenvalue 50%
• Core Model 50%
• Safety Analysis 50%
• Generation Unit 50%