摘要
本研究以三維暫態可壓縮計算流體模型為本，研究高溫氣冷式反應器(HTGR)爐心內部之熱流現象。此模型主要求解動量方程式、能量方程式、k-ε紊流方程式以及化學物種傳遞方程式。燃料球部份以多孔性介質方式探討並假設冷卻劑為理想氣體分析。
在此選用HTR-10做為模擬對象，此反應器設計於北京清華大學核能工藝協會，是以石墨作為緩和劑、氦氣作為冷卻劑並且它的爐心為燃料球堆積而成。初步先計算穩態反應器爐心內部熱流參數，計算結果獲得速度、壓力、溫度等等…。接著研究發生預期暫態事故或事件時爐心內部之特性。
事故分析分為二種。其一為水蒸汽經由蒸汽產生器管路發生破裂出釋放，水蒸汽經由主循環迴路進入反應器爐心；另一種事故為連接反應器與蒸汽產生器之Hot gas duct發生斷裂，空氣經由自然對流與分子擴散進入反應器爐心。研究結果顯示出。蒸汽產生器管路發生破裂事故，水蒸汽進入爐心與燃料球外層的石墨產生之化學反應所導致的氫氣體積分率不足以構成發生氫氣爆炸危險。連接反應器與蒸汽產生器之Hot gas duct斷裂事故，可行性測試結果，CFD有足夠能力可以進行分析。

關鍵字：高溫氣冷式反應器、暫態CFD模型、分子擴散、自然對流

ABSTRACT
A transient three-dimensional (3-D) compressible CFD model is developed in this paper to investigate the thermal-hydraulic characteristics within the core of a high-temperature gas-cooled reactor (HTGR). This transient model essentially includes the Navier-Stoke’s equation, the energy equation, k-ε two-equation turbulent model and the chemical species transport equations. The fuel elements are reasonably treated as the porous media and the gas coolant is assumed to be the ideal gas.
The HTR-10 is selected in the present simulations, which was designed by Institute of Nuclear and New Energy Technology of Tsinghua University in China and is a graphite-moderated and helium-cooled one with the pebble bed core of spherical fuel elements. A calculation is preliminarily performed to obtain the steady-state distributions of thermal-hydraulic parameters within the reactor core, including the velocity, pressure, and temperature, etc. Then, under the postulated event or accident, the transient characteristics in the core can be investigated.
There are two kinds of accident analysis. One is the water vapor will rupture through the steam generation tube, then the water vapor will enter the reactor core through the main circulation loop. Another accident happened where the hot gas duct between reactor and steam generator double ended. The air will enter the reactor core by the natural convection and molecular diffusion. From the investigating result, it was obvious that the steam generation ruptured, and the water vapor entered the reactor core and occurred chemical reaction with graphite, the hydrogen volume fraction does not reach the dangerous hydrogen detonation level. Besides, the test result shows that CFD has the capability to analysis the accident that the hot gas duct between reactor and steam generator double ended.

Keywords: HTGR, Transient CFD Model, Molecular diffusion, Natural convection

參考文獻
1. DOE, 2000. Discussion on Goals for Generation IV Nuclear Power Systems—from a workshop held on May 1– 3, Bethesda, Maryland, USA.
2. L Lohnert, G.H., 1990. “Technical design features and essential safety-related properties of the HTR-module”.Nuclear Engineering and Design 121, 259–275. (1990)
3. Zuying Gao, Chunyun Wang, Baoyan Li, Transient Analysis of Water Ingress Into The HTR-10 High Temperature Gas Cooled Test Reactor, Proceedings of the IAEA Technical Committee meeting, Beijing, China, Oct. 25-27 (1993)
4. Zuying GAO, Baoyan LI, Chunyun WANG, Zhiqiang JJANG . “Transient analysis of air ingress from broken pipe into the HTR-10 reactor pressure vessel.”.Institute of Nuclear Energy Technology, Tsinghua University, Beijing, China
5. Gao Zuying and Shi Lei. “Thermal hydraulic transient analysis of the HTR-10. Nuclear Engineering and Design 218 (2002) 65–80
6. Jing Xingqing and Yang Yongwei,“physical designs and calculations for 10MW high temperature gas cooled reactor –test module (HTR-10)., ” Institute of Nuclear Energy Technology, Tsinghua University, Beijing 100084, P. R. China (2004).
7. Zuying Gao and Lei Shi ,“Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core, ”Nuclear Engineering and Design 218, 51–64 (2002)
8. T. H. Huang, “Integrated system CFD modeling of the flow distribution within a pebble bed modular reactor.”, ICONE16 May 11-15, 2008, Orlando, Florida, USA
9. Walter Schmitz and Albert Koster, “Air ingress and corrosion potential for PBMR direct cycle.”, HTR2008, September 28-October 1, 2008, Washington, DC USA
10. Ugur Emre SIKIK, “Simulations of air and water ingress transients for the pebble bed modular reactor (PBMR) by means of the TINTE code.”, HTR2008, September 28-October 1, 2008, Washington, DC USA
11. Andrew C. Kadak and Tieliang Zhai, “Air ingress benchmarking with computational fluid dynamics analysis”, Nuclear Engineering and Design 236 (2006) 587–602
12. M.S. Yao and R.P. Wang et al., “The helium purification system of the HTR-10”, Nuclear Engineering and Design 218 (2002) 163–167
13. ANSYS Inc., 2006. Fluent 6.3.26 User’s Guide.
14. S. Ergun. “Fluid Flow through Packed Columns,” Chem. Eng. Prog., 48, 89-94, (1952).
15. B. E. Launder and D. B. “Spalding. Lectures in Mathematical Models of Turbulence, ” Academic Press, London, England, (1972).
16. M Necati ÖZISIK, Heat Transfer A BASIC APPROCH, p.383(1985).
17. 劉建忠，「被動式衰變熱移除系統於第四代高溫氣冷式與液態鈉核子反應器之熱流分析」，國立高雄應用科技大學機械與精密工程研究所，碩士論文，2008。
18. Jung-Jae Lee, Goon-Cherl Park, Kwang-Yong Kim, Won-Jae Lee, “Numerical treatment of pebble contact in the flow and heat transfer analysis of a pebble bed reactor core, ” Nuclear Engineering and Design 237 , 2183–2196 (2007).
19. Perry. John H, Chemical engineers handbook, 3-285 (1941)
20. 趙兆颐與朱瑞安，反應器熱工流體力學，清華大學出版社。