現今核能安全議題已成為國際間最被關注的焦點，核電如能繼續被選項，擔任未來供電主流，核安之改良及增進勢不可免。第四代核反應器已透過一段時間之研究與開發，其技術已達成熟。第四代核反應器之一－高溫氣冷式反應器(High Temperature Gas cooled Reactor, HTGR)，其設計目標朝著更深度之被動安全，更高的熱功效率及發電效率，更高之經濟性與更良好的環境友善度，將為未來核能發展之趨勢電廠之一。HTGR核反應器最嚴重之事故為空氣侵入事故，因此在HTGR核反應器發生空氣侵入事故時，分析並了解事故發生後的物理現象及其嚴重性是必要的。
本研究主要以商用流體力學軟體CFD來模擬計算三維暫態模型，分析高溫氣冷式反應器內部爐心之熱流現象。主要研究對象為空氣侵入反應器爐心內部所造成之事故影響，研究中所分析之事故是使用穩態運轉之狀態當作事故發生時之初始條件。此事故之發生是由於連接蒸氣產生器與反應器壓力槽之間之Hot gas duct發生斷裂，造成外界空氣經由破口處利用氣體分子擴散運動進入爐心內部。空氣進入反應器爐心後，由於空氣中之氧分子與高溫石墨產生一系列之化學反應，此反應將會造成燃料元件腐蝕而使燃料元件表面受損，嚴重可能導致放射性物質外洩。因此燃料組件表面之腐蝕嚴重情況就成了一個輻射物質會否外釋之安全性之重要指標。
本研究中，隨著事故發生時間之演變，石墨腐蝕最為嚴重的地方均在爐心燃料區域之下半部，因此未來第四代HTGR反應器之設計上，可以嘗試在腐蝕最為嚴重之位置設置一些安全保障作為，讓氧氣之濃度可以平均散佈在爐心燃料區內部，而不會集中在特定區域。

Nuclear power plant safety is one of the most important part issues of international energy usage nowadays. Generation IV nuclear reactors have been investigated and developed for a period of time, and will be one of the the major electricity supplies for the future global energy requirements efficiently and safely. High Temperature Gas Cooled Reactor (HTGR) is one type of generation IV reactors, which is developed successfully and waited for being used to solve future the energy shortage problem worldwide. Although, there is already highly safe design for HTGR, but deeper understanding and manipulating the hypothetical severe accident is still necessary. Air-ingress accident analysis of GEN IV HTGR is the main purpose of this study.
In this study, CFD (Computational Fluid Dynamics) is used for investigating the thermal-hydraulic phenomena in HTGR while air-ingress is happening there after. At the beginning, normal operating conditions is obtained by steady-state simulation. Transient simulation is used for understanding the thermal-hydraulic phenomena of the HTGR core since the severe accident incipience. The pipe between reactor vessel and steam generator, which is called hot gas duct, broke during the accident started. This pipe broken will cause the outside air diffused into the reactor vessel. Fuel assembly will be corroded by oxygen, due to a series of chemical interaction between oxygen and high temperature graphite. As a result, radioactive fission products may be released into outside environment space. Obviously, the corrosion problem on fuel assembly is the key issue of HTGR safety.
After a long period of interaction between oxygen and fuel assembly, the fuels in the lower region of the core damaged the most. As a consequence, some safety devices are needed to be design and equipped for making air diffuses uniformly in vessel to avoid the non uniformed corrosion situation.

1. Chang H. Oh*, Hyung S. Kang, Eung S. Kim, “Air-ingress analysis: Part2-Computational fluid dynamic models”, Nuclear Engineering and Design 213-225(2011).
2. Nam-il Tak*, Min-Hwan Kim, Won Jae Lee, “Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor”, Annals of Nuclear Energy 1892-1899(2008).
3. Hiroyuki Sato, Richard Johnson, Richard Schultz, “Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR”, Annals of Nuclear Energy 1172-1185(2010).
4. Ms Raji Tripathi, “GT-MHR conceptual design description report”, General atomics-337-02(2002)
5. Takeda, T., Hishida, M., “Studies on molecular diffusion and natural convection in a multicomponent gas system.” Int. J. Heat Mass Transfer 39(3), 527-536(1996).
6. Eung Soo Kim, Hee Cheon No, et al, “Estimation of graphite density and mechanical strength variation of VHTR during air-ingress accident”, Nuclear Engineering and Design 897-847(2008)
7. Chang H. Oh*, Eung S. Kim, Hyung S. Kang, “Natural circulation patterns in the VHTR air-ingress accident and related issues”, Nuclear Engineering and Design(2011).
8. Zhai, T., Kadak, A. and NO, H.C., “LOCA and air ingress accident analysis of a pebble bed reactor”, CANES MIT, MIT-ANP-TR-102(2004).
9. Rehm, W., John, W., Verfondern, K., “Present results and further developments on safety analysis of small and medium sized HTRs for core heat-up accidents.” Nucl. Eng. Des. 109, 281-287(1988).
10. Blanchard. A, “The thermal oxidation of graphite”, IAEA-TECDOC-1154, Appendix 2(2002).
11. Wu Yuanqiang., DiaoXingzhong., Zhou Huizhong., Huang Zhiyong., “Design and tests for the HTR-10 control rod system” , Nuclear Engineering and Design 218 (2002) 147–154.
12. S. Ergun. “Fluid Flow through Packed Columns,” Chem. Eng. Prog., 48, 89-94, (1952).
13. B. E. Launder and D. B. Spalding. “Lectures in mathematical models of turbulence, ” Academic Press, London, England, (1972).
14. M Necati ÖZISIK, Heat transfer abasicapproch, p.383(1985).
15. 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。
16. 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).
17. 趙兆颐與朱瑞安，反應器熱工流體力學，清華大學出版社。
18. Perry. John H, Chemical engineers handbook, 3-285 (1941).
19. EungSoo Kim., HeeCheon No., “Experimental study on the reaction between nucleargraphite IG-110 and carbon dioxide”, Journal of Nuclear Materials 350 (2006) 96–100.
20. Marie-Anne V. Brudieu., “Blind benchmark predictions of the NACOK air ingress tests using the CFD code FLUENT”., Massachusetts Institute of Technology, February-2007.
21. 陳宗廷，「CFD應用於HTGR爐心事故熱流分析」，國立清華大學工程與系統科學研究所，碩士論文，2009
22. 祈振瑋，「利用CFD熱流模式探討空氣進入HTR-10爐心之事故分析」，國立清華大學核子工程與科學研究所，碩士論文，2011