MCNPProject
Simulate PWR,CANDU and VVER reactor (Assembly Model) using MCNP-v6.0.0 https://mcnp.lanl.gov/
MCNP Code
Monte Carlo N-Particle Transport Code (MCNP) is a software package for simulating nuclear processes. It is developed by Los Alamos National Laboratory since at least 1957 with several further major improvements. It is distributed within the United States by the Radiation Safety Information Computational Center in Oak Ridge, TN and internationally by the Nuclear Energy Agency in Paris, France. It is used primarily for the simulation of nuclear processes, such as fission, but has the capability to simulate particle interactions involving neutrons, photons, and electrons among other particles. "Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, detector design and analysis, nuclear oil well logging, accelerator target design, fission and fusion reactor design, decontamination and decommissioning."
This Project
In this project, three different types of reactor had been simulated with three different Uo2 enrichment.Pressurized water reactor (PWR AP1000), CANDU reactor(ARC100) and water-cooled water-moderated energy reactor(VVER-1000) had been chosen to be simulated due to the difference in geometry of assembly and fuel rod composition.
Geometry layout of AP1000 PWR assembly using MCNP-v6.0.0.
Geometry layout of CANDU ACR700 assembly using MCNP-v6.0.0.
Geometry layout of VVER-1000 assembly using MCNP-v6.0.0.
