目前許多核子反應器安全分析nuclear reactor safety (NRS) 逐漸採用計算流體力學Computational Fluid Dynamics（CFD）的方法，其最大優點可獲得詳細的局部熱水流資料訊息，因此研究紊流模式對於燃料束組件的影響相當重要。本論文探討發展核電廠燃料組件數值模擬方法與提升CFD計算模組對於核能組件準確性之方法論驗證，進而分析燃料棒束內局部熱水流特性，而其特性對於格架設計以及燃料棒束的熱傳能力與完整性有很大的影響。這種三維局部的熱水流分佈特性，正是傳統安全分析工具無法模擬而CFD可以貢獻其能力之處。
研究針對模擬燃料棒組件，進行設計格架模組，以數值模擬的結果與實驗做比較驗證，進行分析討論燃料束流道熱水流現象，探討格點細化所對於模擬所造成的誤差，全尺寸與簡化尺寸的準確性，以及穩態與暫態的比較分析評估準確性。透過研究所獲得的模式基準來加以處理複雜的燃料組件結構，包括不同紊流模式 (Turbulence model)以及近壁面處理 (Near-wall treatment)方式等研究應用在燃料棒熱水流的誤差分析，其中利用近壁面Nusselt number得到燃料棒橫向與軸向分布的數值模擬比較基礎上。研究確認Realizable k-ε紊流模式對於燃料棒流道熱水流的計算結果提供精準的適用性，包括在使用CFD的計算時間上也較為經濟。在透過局部化的探討，包括流動、紊流和傳熱特性等現象，都能與實驗有相當接近的趨勢。研究將可提供更精確的格架葉片模擬校驗乃至於完整的反應器燃料束模擬與紊流模式靈敏度分析。因此，站在平行驗證或是協助核能管制單位審查分析的參考，對於燃料組件之混和葉片的格架設計其燃料棒束內的熱水流分佈特性之模擬分析，有相當重要的必要性。

Computational fluid dynamics (CFD) is increasingly being used in nuclear reactor safety (NRS) analyses to describe safety-relevant phenomena occurring in the reactor coolant system in greater detail. The majority of this paper is to investigate the CFD modeling and assessment for numerically simulating the thermal–hydraulic characteristics within the fuel rod bundle using CFD methodology. And this characteristic for mixing-vane grids of heat transfer capability and completeness has a great influence. The three-dimensional partial can reasonably reproduce distribution, it cannot simulate for traditional analysis tools and CFD can contribute at this ability.
This paper presents the results of numerical issues such as mesh refinement, wall treatment, are applied to the prediction of turbulent flow. The performance of various turbulence models are evaluated by calculation of the Nusselt number distribution in a fuel bundle. Comparison between numerical and experimental results of lateral and axial distributions for the Nusselt number obtained via turbulence model without near-wall functions is not sufficiently good, while agreement is found the realizable k-ε model with near-wall functions accurately predicts, for locations close to the support grid. As a result of this study, we have been able to determine the most appropriate turbulence models and the best enhanced wall treatment for modeling reactor coolant systems. Therefore, parallel tests or assistance are necessary for the regulator staff to review the license issues for CFD investigating the localized thermal-hydraulic characteristic within mixing-vane grids.

總目錄
中文摘要 I
ABSTRACT II
誌謝 III
總目錄 IV
表目錄 VI
圖目錄 VII
符號說明 X
第一章 緒論 1
1.1 研究背景 1
1.2 研究目的 3
1.3 文獻探討 4
第二章 理論基礎與數值模型 9
2.1 統御方程式 9
2.2 紊流模式 10
2.3 數值模式： 27
2.4 邊界條件 28
第三章 研究方法 32
3.1 研究流程 32
3.2 模型建立 33
3.3 網格設置 34
3.4 網格靈敏度分析 35
3.5 近壁面網格處理 37
3.6 軸向網格處理 39
第四章 燃料組件模式分析評估 40
4.1 燃料組件網格靈敏度分析結果 40
4.2 燃料棒格架標準近壁作用的分析影響 45
4.3 燃料棒格架增強近壁作用的分析影響 51
4.4 燃料棒組件全尺寸的分析影響 55
4.5 燃料棒穩態與暫態結果分析 57
第五章 研究結果與討論 60
5.1 標準格架流場分析結果 60
5.2 SPLIT-VANE2格架分裂葉片格架流場分析 68
5.3 SPLIT-VANE1格架分裂葉片格架流場分析 74
5.4 不同棒束距離影響分析 82
5.5 不同格架葉片影響分析 89
第六章 結論與建議 105
第七章 參考文獻 107

[1] FLUENT Inc., Fluent 12.0 User’s Guide (2009).
[2] Yao, S.C., Hochreiter, L.E., Leech, W.J., “Heat-Transfer Augmentation in Rod Bundles Near Grid Spacers” J. Heat Transfer 104, 76 – 81. 1982.
[3] Wu, X., Trupp, A.C., “Experimental study on the unusual turbulence intensity distribution in rod-to-wall gap regions” Exp. Therm. Fluid Sci.6 (4) , 360–370. 1993.
[4] Shih, T.-H., Zhu, J., Lumley, J. L., “Arealizable Reynolds stress algebraic equation model” NASA TM 105993. 1993.
[5] Lim H.T., Jeon K.L., Choi J.H., Park B.S., “Flow characteristics for spacer grid with mixing vane of PWR fuel assembly”, Transactions of the 14th International Conference on Structural Mechanics in Reactor Technology (SMiRT 14), Lyon, France, August 17-22, 1997.
[6] Lee, K.B., Jang, H.G., “A numerical prediction on the turbulent flow in closely spaced bare rod arrays by a nonlinear k−ε model” Nuclear Engineering and Design 172, 351–357. 1997.
[7] Smith III, L.D., Conner, M.E., Liu, B., Dzodzo, M.B., Paramonov, D.V., Beasley, D.E., Langford, H.M., Holloway, M.V., “Benchmarking computational fluid dynamics for application to PWR fuel” In: Proceedings of the 10th International Conference on Nuclear Engineering, Arlington, Virginia, USA, April 14–18. 2002.
[8] Yadigaroglu, G., Anderani, M., Dreier, J., Coddington, P., “Trends and needs in experimentation and numerical simulation for LWR safety” Nuclear Engineering and Design 221, 205–223. 2003.
[9] E. Baglietto, H. Ninokata, “A turbulence model study for simulating flow inside tight lattice rod bundles”Nuclear Engineering and Design 235 (2005) 773–784
[10] Hàzi Gàbor, “On turbulence models for rod bundle flow computations” Annals of Nuclear Energy 32, 755-761. 2005
[11] Gábor Házi, Gusztáv Mayer., “Flow in Rod Bundles ”International Conference Nuclear Energy for New Europe Bled, Slovenia, September 5-8, 2005
[12] E. Baglietto, H. Ninokata, Takeharu Misawa “CFD and DNS methodologies development for fuel bundle simulations” Nuclear Engineering and Design 236 (2006) 1503–1510
[13] Sofiane Benhamadouche and Christelle, Le-Maître,“ Large Eddy Simulation of the flow along four sub-channels downstream a mixing grid in a PWR” The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) Kanazawa City, Ishikawa Prefecture, Japan, September 27-October 2, 2009.
[14] H. Ganjiani and B. Firoozabadi, “Three-Dimensional Simulation of Turbulent Flow in 3-Sub Channels of a VVER-1000 Reactor Transaction” B: Mechanical Engineering Vol. 17, No. 2, pp. 83-92 2010
[15] S. Toth, A. Aszodi, “CFD analysis of flow field in a triangular rod bundle”Nuclear Engineering and Design Volume: 240, Issue: 2, February, 2010, pp. 352-363.
[16] Jun Ho Bae, Joo Hwan Park, “Analytical prediction of turbulent friction factor for a rod bundle”Annals of Nuclear Energy 38 (2011) 348–357
[17] Sun kyu Yang, Heung June Chung, Se Young Chun, and Moon Ki Chung, “Measurements of Turbulent Flow in 5×5 PWR ROD Bundles With Spacer Grids”Journal of Korean Nuclear Society Volume 24, Number 3, September 1992
[18] Karoutas Z, Gu CY, Scholin B. “3-D flow analyses for design of nuclear fuel spacer” In: Proceedings of the seventh international meeting on nuclear reactor thermal-hydraulics, September 1995, New York, United States.
[19] Wang Kee In, Tae Hyun Chun, Dong Seok Oh and Yeon Ho Jung, “CFD Analysis of Turbulent Flow in Rod Bundles for Nuclear Fuel Spacer Design”Transactions of the 15th International Conference on Structural Mechanics in Reactor Technology (SMiRT 15), Seoul, Korea, August 15-20, 1999.
[20] Wang Kee In, Dong Seok Oh, and Tae Hyun Chun, “Flow Analysis for Optimum Design of Mixing Vane in a PWR Fuel Assembly”Journal of Korean Nuclear Society Volume 33, Number 3, PP.327~338, June, 2001.
[21] Xiang-Zhe CUI and Kwang-Yong KIM. “Three-Dimensional Analysis of Turbulent Heat Transfer and Flow through Mixing Vane in A sub-channel of Nuclear Reactor”Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 40, No. 10 , p. 719-724 (October 2003)
[22] Lee, C.M., Choi, Y.D., “Comparison of thermo-hydraulic performances of large scale vortex flow (LSVF) and small scale vortex flow (SSVF) mixing vanes in 17×17 nuclear rod bundle” Nuclear Engineering and Design 237, 2322–2331. 2007.
[23] Seok Kyu Chang, Sang Ki Moon, Won Pil Baek, Young Don Choi. “Phenomenological investigations on the turbulent flow structures in a rod bundle array with mixing devices”Nuclear Engineering and Design 238 (2008) 600–609
[24] Sang Yong Han, Jeong Sik Seo, Min Su Park, Young Don Choi, “ Measurements of the flow characteristics of the lateral flow in the 6×6 rod bundles with Tandem Arrangement Vanes” Nuclear Engineering and Design 239 (2009) 2728–2736
[25] Mary V. Holloway, Donald E. Beasley, and Michael E. Conner, “ Investigation of swirling flow in rod bundle sub-channels using computation fluid dynamics” Proceedings of ICONE14 International Conference on Nuclear Engineering July 17-20, 2006 Miami, Florida, USA.
[26] J. S. Seo, J. K. Shin, Y. D. Choi, K. R. Kim, “Local heat transfer coefficient measurement in 6×6 Rod bundles with mixing vanes”The 19th International Symposium on Transport Phenomena, 17-20 August, 2008, Reykjavik, ICELAND
[27] M.R. Nematollahi , M. Nazifi. “Enhancement of heat transfer in a typical pressurized water reactor by different mixing vanes on spacer grids”Energy Conversion and Management 49 (2008) 1981–1988.
[28] Yakhot , V., Orszag, S.A., Thangam, S., Gatski, T.B., Speziale, C.G., “Development of turbulence models for shear flows by a double expansion technique”Physics of Fluids A4 (7), 1510–1520. 1992.
[29] Conner, M.E., Baglietto, E., Elmahdi, A.M., 2009. “CFD methodology and validation for single-phase flow in PWR fuel assemblies”Nuclear Engineering and Design, published online 2009.
[30] C.C. Liu and Y.M. Ferng, “Numerically simulating the thermal-hydraulic characteristics within the fuel rod bundle using CFD methodology,” Nuclear Engineering and Design Volume: 240, 2010, pp.3078-3086.
[31] B.H. Yan, H.Y. Gu, Y.H. Yang, L. Yu, “Effect of rolling on the flowing and heat transfer characteristic of turbulent flow in sub-channels” Progress in Nuclear Energy 53 (2011) 59-65
[32] C.C. Liu, Y.M. Ferng , C.K. Shih “CFD evaluation of turbulence models for flow simulation of the fuel rod bundle with a spacer assembly” Applied Thermal Engineering 40 (2012) 389-396
[33] Michael E. Conner, Yassin A. Hassan, Elvis E. Dominguez-Ontiveros “Hydraulic benchmark data for PWR mixing vane grid” Nuclear Engineering and Design Volume: 264, 2013, pp. 97-102.
[34] Rowe D.S., Johnson B.M. and Knudsen J.GA., “ Implications Concerning Rod Bundle Crossflow Mixing based on Measurements of Turbulent Flow Structure” Int. J. Heat Mass Transfer, Vol. 17, 1974, 407-419.
[35] Kim, N.H., Chun, T.H., Lee, S.K., Kim, S.H., 1991. “Application on the law of the wall to predict friction factors for turbulent flow in a rod bundle” In: Proc. First JSME/ASME Joint Int. Conf. Nuclear Engineering, pp. 231–235.
[36] Tsutomu Ikeno, Takeo Kajishima., “ Analysis of dynamical flow structure in a square arrayed rod bundle” Nuclear Engineering and Design Volume: 240, Issue: 2, February, 2010, pp. 305-312.
[37] Jun Ho Bae, Joo Hwan Park, “Analytical prediction of turbulent friction factor for a rod bundle”Annals of Nuclear Energy 38 (2011) 348–357
[38] Holloway, M. V., McClusky, H. L., Beasley, D. E. and Conner, M. E., “ The Effect of Support Grid Features on Local, Single-Phase Heat Transfer Measurements in Rod Bundles” Journal of Heat Transfer 126, 43-53. 2004.
[39] Holloway, M. V., Conover Timothy A., McClusky Heather L., Beasley, Donald E. “The Effect Of Support Grid Design on Azimuthal Variation in Heat Transfer Coefficient for Rod Bundles. ” Journal of Heat Transfer Vol. 127/599. June 2005.
[40] Launder, B.E., Spalding, D.B., 1973. “ The numerical computational of turbulent flows”Comp. Method Appl. Mech. Eng. 3, 269.