The Recommended Publication for Citing
  1. Jaerim Jang, Wonkyeong Kim, Sanggeol Jeong, Eun Jeong, Jinsu Park, Matthieu Lemaire, Hyunsuk Lee, Yongmin Jo, Peng Zhang, Deokjung Lee, "Validation of UNIST Monte Carlo Code MCS for Criticality Safety Analysis of PWR Spent Fuel Pool and Storage Cask," Annals of Nuclear Energy, 114: 495-509. https://doi.org/10.1016/j.anucene.2017.12.054 (2018)
A Monte Carlo code MCS has being developed at Ulsan National Institute of Science and Technology (UNIST) since 2013. The target of MCS is to solve complex whole core problems like BEAVRS. MCS can treat the 3D whole core geometry with universe and lattice, and the neutron physics with probability-table, free-gas treatment, S(a,b) and Doppler Broadening Rejection Correction.

Monte Carlo Code MCS
 • Language: Fortran 2003
 • Purpose
  - Large Scale Reactor Analysis with accelerated Monte Carlo simuation 
   - University research: MC methodology development, advanced reactor design 
  General 3-D geometry
 • Nuclear Data
   - ENDF-B/VII.0 and ENDF-B/VII.1
   - Continuous energy and multi-group
   - Double indexing method
 • Physics 
- Resonance upscattering (DBRC, FESK)
- Probability  table method
- S(𝜶,𝜷)
- On the fly Doppler broadening
- CTF Coupling
  • Acceleration
- MOC and MC Hybrid solver
- Modified power iteration
- Wielandt method
   - Parallel fission bank
 • Depletion
     - CRAM , MEM, Krylov Subspace

CORE Benchmark

                Normalized fission reaction rate and flux

 Fuel temperature and coolant density 

BEAVRS Cycle 1 Whole Core Depletion with Full Feedback

Flow Chart

BEAVRS TH Calculation with depletion effect