Monte Carlo transport and burnup calculation

Annals of Nuclear Energy

Volumn 30, Issue 1, 2003, Pages 127-132

  Li, Deng ^a   Zhongsheng, Xie ^b   Shu, Li ^a  

^a INSTITUTE OF APPLIED PHYSICS AND COMPUTATIONAL MATHEMATICS   (China)

^b XI'AN JIAOTONG UNIVERSITY   (China)

Author keywords

[No Author keywords available]

Indexed keywords

COMPUTER PROGRAMMING; COMPUTER SIMULATION; MONTE CARLO METHODS; NUCLEAR REACTORS;

NEUTRON TRANSPORT;

NEUTRONS;

EID: 0037208731     PISSN: 03064549     EISSN: None     Source Type: Journal    
DOI: 10.1016/S0306-4549(02)00044-0     Document Type: Article

References (5)

1. MCNP - A general Monte Carlo code for N-particle transport code
2. Transfer matrix treatments for multigroup Monte Carlo calculations - The elimination of ray effects
3. A general description of the Lattice code WIMS
4. MCMG: A 3-D Multigroup Monte Carlo code and its benchmarks