Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor

<table>Prediction of SVR with JENDL-3.3 and JENDL-4.0 for several configurations of the BFS facility in IPPE (Russia [<a href="/journals/stni/2012/351809/#B9">9</a>]) (Reproduced from [<a href="/journals/stni/2012/351809/#B8">8</a>]). Vertical bars represent experimental uncertainties. HEU/IEU/LEU: High/Intermediate/Low Enriched Uranium; MOX: Mixed Oxide, that is, uranium–plutonium oxide fuel.</table>

Science and Technology of Nuclear Installations

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Figure 4

Figure 4: Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor 

Figure 4 | Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor 