1. X-5 Monte Carlo Team, MCNP Version 5 - A General Monte Carlo N-Particle Transport Code, Report LA-UR-03-1987, Los Alamos National Laboratory (2003).
2. Méplan O. et al., MURE, MCNP Utility for Reactor Evolution: couples Monte Carlo transport with fuel burnup and thermal-hydraulics calculations, Data Bank Computer Program Services, OECD Nuclear Energy Agency (2009).
3. Basile D. et al., COBRA-EN, an upgraded version of the COBRA-3C/MIT subchannel code for thermal-hydraulics, Data Bank Computer Program Services, OECD Nuclear Energy Agency (2001).
4. Capellan N. et al., “3D coupling of Monte Carlo neutronics and thermal-hydraulics calculations as a simulation tool for innovative reactors”, Proc. Int. Conf. global 2009.
5. Brown F.B., “Review of Monte Carlo Criticality Calculations Convergence, Bias, Statistics”, Proc. Int. Conf. M&C 2009.