2641 直接接触沸騰水型鉛ビスマス冷却高速炉の概念検討(S57-4 原子炉システムおよびその要素技術(4),S57 原子炉システムおよびその要素技術)

高橋 実; NOVITRIAN; 山田 由美; Abu Khalid Rivai; 内田 正治

doi:10.1299/jsmemecjo.2006.3.0_35

2006.3

Session ID : 2640

DOI https://doi.org/10.1299/jsmemecjo.2006.3.0_35

Conference information

2641 Conceptual Study of Direct Contact Water Boiling Lead-Bismuth Cooled Fast Reactor

Minoru TAKAHASHI, NOVITRIAN, Yumi YAMADA, Abu Khalid Rivai, Shouji UCHIDA

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Keywords: Lead-bismuth, Fast reactor, Safety, Corrosion, Boiling, Direct contact heat transfer, Gas lift pump

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Abstract

Conceptual design and safety analysis were performed for Pb-Bi Cooled Direct Contact Boiling Water Fast Reactor (PBWFR). Steels were surface-treated using sputtering method with Al and SUS304 targets for compatibility with Pb-Bi. The steels were corrosion-resistant in Pb-Bi at 700℃. Pb-Bi-water direct contact boiling two-phase flow in PBWFR was simulated numerically and experimentally. Measured void fraction could be described by modified drift flux model.

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