二重管内強制流動サブクール沸騰限界熱流束の予測

Critical heat flux prediction for subcooled flow boiling in annulus

Liu, W.

Liu, W.

軽水炉の安全性評価のためには、限界熱流束の評価が重要であるが、炉心内強制流動サブクール沸騰条件での限界熱流束の予測手法は確立されていない。本研究では、PWR炉心に対する強制流動サブクール沸騰条件での限界熱流束予測手法確立の一環として、炉心燃料集合体を簡略化した二重管を対象として、強制流動サブクール沸騰限界熱流束の予測手法を検討した。Nouriによる二重管内液相速度分布式をLiquid sublayer dryoutモデルと組み合わせることにより、水及びR113を試験流体とした既存実験データを20%程度で予測できることを確認した。

Subcooled flow boiling is a boiling that begins and develops even though the mean enthalpy of liquid phase is lower than saturation. This forced convective boiling is one of the most efficient ways for the removal of high heat flux. It is widely used in the high heat flux components such as nuclear reactor cores, accelerator targets and fusion reactor components. The thermal outputs of these systems are restricted by Critical Heat Flux(CHF). Because of the importance of the CHF, lots of researches, including both experimental and mechanistic modelling, have been performed. However, the CHF prediction for rod bundles still remains unsolved. As the first step for the CHF prediction in rod bundles, in this paper, we tried to predict the CHF in annulus, which is the most basic flow geometry simplified from a rod bundle. We performed the CHF prediction by using liquid sublayer dryout model, combined with Nouri single phase velocity distribution correlation for annulus. The results show that the CHF in annulus can be predicted in an accuracy of about 20%.