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Conceptual Design of a Steam-Water Power Reactor Core

Yasunori Bessho, Takashi Nakayama, Michiro Yokomi, Katsuma Nakayama, Hiroki Sano, Nobuhiro Kanazawa

Nuclear Technology

Volume 123 / Number 1 / July 1998 / Pages 30-43


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The steam-water power reactor core concept, originally proposed by several Russian engineers, is expected to improve natural uranium utilization through self-sustaining plutonium by using tight-lattice plutonium fuels and large void fraction two-phase flow, and to realize inherent safety characteristics through large neutron leakage from the core by a flat core configuration.

Results are described for the core conceptual design for specifications meeting a 500-MW(electric) electricity supply for 13 months of continuous operation and 92 GWd/tonne fuel average discharge exposure. The design has core nuclear thermal-hydraulic characteristics that satisfy the specifications and limitations usually applied to boiling water reactors (BWRs), based on analyses by the three-dimensional multineutron-energy group diffusion analysis program CITATION. Further, its safety characteristics satisfy limitations, usually applied to BWRs, by the steam cooling emergency core cooling system and the reflood system, based on analyses of a loss-of-coolant accident, which is thought to be most critical for a core with a small water inventory, by the general transient analysis program TRAC.

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