AbstractThe critical heat flux (CHF) of subcooled flow boiling for water was experimentally investigated as a fundamental study on high-heat-flux cooling of fusion reactor components. A circular tube with 3 mm inside diameter was heated directly by Joule heating. The experiments were conducted in the pressure range 0.3–1.0 MPa under uniform heating conditions. It was verified that the modified Tong correlation, whivh was proposed by the authors, predicts the accurate CHF of lower than about 20 MW/m2 in the pressure range from 0.1 to 13.8 MPa and was also useful in the prediction of the high CHF from 20 to 130 MW/m2 at low and medium pressures, which might correspond to the cooling conditions of fusion reactor components. The high-CHF data were compared with the Gunther, the Gambill-Greene and the Weisman-Pei correlation and the effects of the experimental parameters on the CHF-pressure relation were discussed. The flow conditions of the cooling water at which a CHF from 10 to 40 MW/m2 might be achieved, were examined using the modified Tong correlation.
限界熱流束（CHF）のサブクール流動沸騰に関する研究を、核融合炉構成要素の高熱流束冷却の基礎研究として実施した。内径3mmの円管をジュール加熱により直接加熱した。実験は圧力範囲0.3〜1.0MPa、均一加熱条件下で実施した。著者らが提案した修正Tong相関式が、圧力範囲0.1〜13.8MPaにおいて約20MW/m²以下の限界熱流束を正確に予測できることを検証し、また核融合炉構成要素の冷却条件に相当し得る低圧・中圧域における20〜130MW/m²の高限界熱流束の予測にも有用であることを確認した。高限界熱流束のデータをGunther、Gambill-Greene、Weisman-Peiの各相関式と比較し、実験パラメータが限界熱流束と圧力の関係に及ぼす影響について考察した。修正Tong相関式を用いて、10〜40MW/m²の限界熱流束が達成され得る冷却水の流動条件を検討した。