Subject
Waste Management and Disposal,Energy Engineering and Power Technology,Safety, Risk, Reliability and Quality,Nuclear Energy and Engineering
Reference6 articles.
1. MCNP – a General Monte Carlo N-Particle Transport Code (Version C);Briesmeister,1997
2. Safety Analysis Report of the MNSR Reactor;China Institute of Atomic Energy,1993
3. Nuclear Reactor Analysis;Duderstadt,1976
4. Measurement of the fast neutron flux in the MNSR inner irradiation site;Khattab;Applied Radiation and Isotopes,2007
5. Calculations of the thermal and fast neutron fluxes in the Syrian MNSR irradiation sites using the MCNP-4C code;Khattab;Applied Radiation and Isotopes,2009
Cited by
8 articles.
订阅此论文施引文献
订阅此论文施引文献,注册后可以免费订阅5篇论文的施引文献,订阅后可以查看论文全部施引文献