1. Briesmeister, J.F., 2000. MCNP – A General Monte Carlo N-Particle Transport Code, Version 4C, Los Alamos National Laboratory Report LA-13709-M (April 2000).
2. ORIGEN2: a versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials;Croff;Nucl. Technol.,1983
3. DeHart, M.D., Brady, M.C., Parks, C.V., 1996. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results, ORNL-6901. Oak Ridge National Laboratory (June 1996).
4. ALEPH 1.1.2, A Monte Carlo Burnup Code;Haeck,2006
5. Kelly, D.J., 1995. Depletion of a BWR lattice using the RACER continuous energy Monte Carlo code. In: Proceedings of the International Conference on Mathematics and Computations, Reactor Physics and Environmental Analyses, vol. II, Portland, p. 1011.