CFD燃料組件分析模式誤差之研究

Investigating the simulation errors for CFD models in the fuel bundle

葉承翰(Cheng-Han Yeh)

指導教授 ： 馮玉明

國立清華大學/原子科學院/工程與系統科學系/碩士(2014年)

摘要

安全分析一直是核電廠設計、運轉與安全不可或缺的重要工具。傳統式的核電廠安全分析，端賴系統分析程式，並運用保守度或安全餘裕補足模式之不足以確保核電廠之安全。近年來在電腦運算與儲存能力突飛猛進的助益下，核能安全分析界逐漸地應用計算流體力學(Computational Fluid Dynamics, CFD)程式進行相關之分析。國內既有電廠運轉安全相關改善案或者廠家引進新型燃料之相關安全分析上，也採用CFD程式進行全部或部份分析的案例，並進行分析法制化或申請執照之工作。CFD程式應用於核電廠安全分析，最重要的是理論模式與格點模式的選取以及不準度的評估。然而，核電廠複雜的幾何配置與熱水流現象已超過CFD程式內建一般理論模式的適用範圍。因此，管制單位對於利用泛用型CFD程式進行核電廠安全分析案例之審查，除了如審查相關分析結果與程式驗證報告外，獨立執行CFD分析以做為交叉驗證是必須的。
本研究主要以三維穩態模型為計算基礎，其計算流程包括連續方程式、動量方程式、能量方程式以及紊流方程式。針對不同的燃料管束流進行熱水流分析，並探討不同格點所造成之離散化誤差，量化結果之不準度，作為日後建立CFD分析燃料管束流幾何配置之模式與評估審查導則之先期研究。另外進行紊流模式所造成的不準度進行文獻蒐集及先期分析，作為日後接續計畫之起始研究。

關鍵字

燃料管束 ； 計算流體力學 ； 格點誤差 ； 模式誤差 ； 不準度分析

並列摘要

Safety analysis is one of essential tools for the design, operation and safety of nuclear power plants (NPPs). Traditional safety analysis for NPPs depends on system codes with more conservative assumptions or margins to ensure the plant safety, which would scarify the operation flexibility and efficiency. With advantages of dramatic progress in computer power, Computational Fluid Dynamics (CFD) is gradually adopted in the nuclear safety analysis. In addition, Taipower has licensed some safety analysis cases using or partially using the CFD. The most important things for the CFD simulations are the establishment of mesh and models as well as the error estimation. The thermal-hydraulic phenomena related to the reactor safety are so complicated that the models adopted in the commercial CFD may not reasonably capture these characteristics. Independent simulations and cross-check are necessary for the regulator staff to review the CFD issues. Therefore, it is crucial for the regulatory staff to investigate the CFD modeling and assessment. Focusing on modeling the characteristics of fuel bundle flow, this project is to investigate the CFD methodology, influence of different mesh models and its uncertainty. These simulation results can assist the regulator staff in providing the basis of review guidelines for this issue.

並列關鍵字

Rod bundles ； CFD ； Mesh Error ； Model Error ； Uncertainty analysis

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. 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. L.F. Richardson, “The Approximate Arithmetical Solution by Finite Differences of Physical Problems Involving Differential Equations, with an Application to the Stresses in a Masonry Dam”, Philosophical Transactions of the Royal Society of London. Series A, Volume 210, 1911, pp.307-357.

7. P.J. Roache, “Quantification of Uncertainty in Computational Fluid Dynamics”, Annual Review of Fluid Mechanics, Volume 29, 1997, pp.123-160.