燃料護套鋯水反應之CFD模式發展與應用研究

2011年三月福島核能事故發生嚴重氫爆，使核子事故嚴重性加劇。其爆炸之氫氣來源即為在冷卻水流失事故發生時，燃料棒表面鋯與水進行鋯水反應作用，產生大量氫氣及熱能。故鋯水反應之相關現象探討，成為近幾年來核能安全分析之重要計算案例。然而，在國際上，利用CFD進行鋯水反應現象分析之文獻參考並不常見，如何建立有系統的計算模式，即為本文章之研究重點。然而，進行鋯水反應現象分析之前，燃料棒周遭流場預測之準確度亦為分析之重點。其因為周遭流場對於氫氣之分布有極大影響，且周遭流場之熱點位置亦與意外事故發生時之起火點息息相關。故本研究之分析流程前期目標為燃料棒周遭流場之模型建立研究，而後加入鋯水反應模式，進行燃料護套鋯水反應現象之熱通量計算及氫氣產生量預測。以期能建立CFD分析冷卻水流失事故(Loss-of-coolane-accident, LOCA)發生，引發鋯水反應時之安全評估審查導則基礎之先期研究。

關鍵字

鋯水反應； CFD核能安全分析；燃料管束；計算流體力學；不準度分析

並列摘要

Traditional safety analysis for nuclear power plants (NPPs) with system codes with more conservative assumptions to ensure the plant safety. However, it would scarify the operation flexibility and efficiency. On the other hand, system codes could not analyze local phenomenon in the NPPs. Therefore, Computational Fluid Dynamics (CFD) is gradually adopted in the nuclear safety analysis. Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis during a loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production that may cause serious accident. Consequently, the development of CFD Model for the zirconium-water reaction of fuel cladding is quite essential. In this paper, the CFD predictions of a temperature rise and hydrogen production due to Zircaloy/steam oxidation were compared with the results of the CFX-10 simulations. From these validation processes, it is shown that the analysis process of Zircaloy/steam reaction was built.

並列關鍵字

Zircaloy/steam reaction ； Rod bundles ； CFD ； Uncertainty analysis ； LOCA

參考文獻

1. International Atomic Energy Agency, Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems, IAEA-TECDOC- 1379, 2003.

3. Nuclear Energy Agency, “Best Practice Guidelines for the use of CFD in nuclear Reactor Safety Applications”, NEA/CSNI/R(2007)5, 2007.

4. The American Society of Mechanical Engineers, “Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer”, ASME V&V 20-2009, 2009.

5. M.V. Holloway, H.L. McClusky, D.E. Beasley, “The Effect of Support Grid Features on Local, Single-Phase Heat Transfer Measurements in Rod Bundles”, Journal of Heat Transfer, Volume 126, 2004, pp.43-53.

6. Chih-Hung Lin , Cheng-Han Yen, Yuh-Ming Ferng, “CFD investigating the flow characteristics in a triangular-pitch rod bundle using Reynolds stress turbulence model” , Annals of Nuclear Energy 65, 2014, pp.357-364

國際替代計量

燃料護套鋯水反應之CFD模式發展與應用研究

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