P21-06 低減速スペクトル炉稠密格子炉心内限界熱流束

概要

論文の詳細を見る
The purposes of the critical heat flux (CHF) experiments are to investigate the system parameter effects on critical power and to confirm the thermal-hydraulic feasibility of the Reduced-Moderation Water Reactor (RMWR). The test section was a 7-rod bundle with rod diameter of 12.3mm, rod gap of 1.0mm and heated length of 1.8m. Axial power distribution was flat. Exit pressure of the test section was ranged from 1.0 to 8.5MPa, inlet water subcooling from 2 to 50 K, mass velocity from 0 to 2500 kg(m^2s), and transverse heat flux ratio from 0.97 to 1.6. In this manuscript, system parameter effects on critical quality briefly, discussion of the heat up point and applicability of the Arai critical power correlations to the present experiments was described. It was found from comparison between correlations and present database that the Arai correlation underestimated the critical power. It was confirmed from this study that existing design of RMWR core can thermally be feasible.
一般社団法人日本機械学会の論文
2002-06-14