Conceptual design of a steam-water power reactor core

被引:0
|
作者
Bessho, Y
Nakayama, T
Yokomi, M
Nakayama, K
Sano, H
Kanazawa, N
机构
[1] Hitachi Ltd, Hitachi Works, Nucl Core Engn & Fuel Design Dept, Nucl Core & Safety Engn Sect, Hitachi, Ibaraki 317, Japan
[2] Japan Atom Power Co, Chiyoda Ku, Tokyo 100, Japan
[3] Power & Ind Syst Res & Dev Div, Hitachi, Ibaraki 316, Japan
[4] Hitachi Engn Co Ltd, Hitachi, Ibaraki 317, Japan
关键词
D O I
10.13182/NT98-A2877
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
The steam-water power reactor cove 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 lar ge void fraction two-phase Slow, 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.
引用
收藏
页码:30 / 43
页数:14
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