高速炉燃料集合体熱流動解析手法の開発と適用

Development of thermal hydraulic analysis program for wire-wrapped fuel pin bundle of fast reactor

大島 宏之; 今井 康友*

Ohshima, Hiroyuki; Imai, Yasutomo*

経済性向上のため高性能化(高燃焼度化,高線出力化)を目指すナトリウム冷却高速炉燃料の設計や安全評価に資するため、複数の解析コードで構成される燃料集合体変形・熱流動解析システムの開発を進めている。本研究では、この解析システムの構成コードの1つである局所詳細熱流動解析コードについて、圧力損失など流動解析精度を改善するため、高Re数型/低Re数型モデルを組合せた乱流モデルを開発するとともに、基本検証や実機スケール燃料集合体試験解析により、その高い予測精度と計算効率を確認した。

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions including fuel deformation. In this study, we focused on SPIRAL, which is one component code of the numerical simulation system and plays the role to simulate detailed local flow and temperature fields in a wire-wrapped fuel pin bundle, and incorporated a new hybrid turbulence model to improve the pressure drop predictability. SPIRAL with the new turbulence model was applied to numerical simulations of flows in a parallel channel and several types of fuel assemblies for code validation. Pressure loss coefficients estimated by the simulations are in good agreement with theoretical and measured data from laminar to turbulent regions.