Design Study on BN-600 Hybrid Core (II) -Evaluation of Fuel Integrity and Core Neutronic Characteristics by Japanese Analysis Methods-

Sugino, Kazuteru; Uto, Nariaki  ; Naganuma, Masayuki  ; Mizuno, Tomoyasu

A program of disposal of Russian surplus weapon-grade plutonium by containing the plutonium in vibropacked MOX fuel subassemblies and burning them in the BN-600 hybrid reactor core has been progressed. The relevant design works on the BN-600 hybrid core have been carried out under the contract between Japan Nuclear Cycle Development Institute (JNC) and OKB Mechanical Engineering (OKBM), Russian public enterprise. JNC obtained a series of design technical reports. Japanese analysis methods were adopted to evaluate fuel integrity in the design basis transients and neutronic characteristics of the BN-600 hybrid core, based on the design technical data described in the obtained reports. The evaluation results of the key performances, such as maximum cladding and fuel temperatures, coolant (sodium) void reactivity, reactivity coefficient, were found to satisfy the design criteria and/or target provided by Russia, and meet the Russian rule. The results of this study showed that the core and fuel specifications determined by Russia can be considered reasonable and proper from the viewpoint of safety and neutronic designs, and that the Japanese analysis methods are expected to contribute to increasing reliability of the Russian design works.