Multigroup Scattering Matrix Generation Method Using Weight-to-Flux Ratio Based on a Continuous Energy Monte Carlo Technique
Tóm tắt
Từ khóa
Tài liệu tham khảo
Ando Y., 2003, Proc. 7th Int. Conf. on Nuclear Criticality Safety (ICNC2003), 494
Briesmeister J. F., 1993, MCNP—A General Monte Carlo N-Particle Transport Code, Version 4A
Fowler T. B., 1969, Nuclear Reactor Core Analysis Code: CITATION
Alcouffe R. E., 1995, DANTSYS: A Diffusion Accelerated Neutral Particle Transport Code System, 10.2172/212580
Okumura K., 2007, SRAC2006: A Comprehensive Neutronics Calculation Code System
Yamamoto T., 2006, Proc. 2005 Symp. on Nuclear Data, 7