Next Generation Design Codes for a New Technological Platform for Nuclear Power

A preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Tohru Suzuki, Yoshiharu Tobita, Ken-ichi Kawada et al. · Nuclear Engineering and Technology · 2015 · 35 citations · Full text

Engineering, Nuclear Engineering, Nuclear Materials +15

Coupling a CFD code with neutron kinetics and pin thermal models for nuclear reactor safety analyses

Chen Zhao, Xue-Nong Chen, A. Rineiski et al. · Annals of Nuclear Energy · 2015 · 35 citations

Engineering, Computational Engineering, Nuclear Engineering +17

A full-core coupled neutronic/thermal-hydraulic code for the modeling of lead-cooled nuclear fast reactors

R. Bonifetto, Sandra Dulla, Piero Ravetto et al. · Nuclear Engineering and Design · 2013 · 24 citations

Engineering, Nuclear Engineering, Physics +14

Studies of fuel dispersion after pin failure: Analysis of assumed blockage accidents for the MYRRHA–FASTEF critical core

Rui Li, X.-N. Chen, A. Rineiski et al. · Annals of Nuclear Energy · 2015 · 20 citations

Engineering, Nuclear Engineering, Risk Analysis +7

Fuel behavior analysis code FEMAXI-FBR development and validation for core disruptive accident

Tsuyoshi Okawa, Isao Tatewaki, Tomoko Ishizu et al. · Progress in Nuclear Energy · 2015 · 17 citations

Engineering, Automotive Engineering, Nuclear Engineering +8