本研究承接粒子遷移模擬研究室先前的研究成果並完備了熱流分析方法論以建立中子與熱流分析耦合計算的平台，作為爐心設計的工具。研究對象為日本JAEA的高溫氣冷式反應器設計：HTTR及其後續之改良設計 GTHTR300。
耦合計算的中子部分採用蒙地卡羅程式；熱流分析採用解析方法的簡化解析模型計算及數值模型，兩種熱流模型與中子物理耦合計算的結果相互比較及驗證。
GTHTR300採均勻濃縮度燃料設計。疊代計算在數次之後便收斂。軸向功率分布峯值偏下方。兩種熱流方法所得的溫度分佈相當接近。與JAEA報告中所提供的溫度分佈相比，雖在爐心中段差異較大，但整體仍具有相當的一致性。
HTTR採非均勻濃縮度燃料設計。疊代計算也在數次之後便達收斂。軸向功率分布為峯值偏上方。其徑向功率比較平坦。疊代過程功率分布變動的行為與GTHTR300相近。兩種熱流方法所求得之溫度分佈有數十度之差異，其原因可能為燃料護套表面採用之熱傳係數不同的緣故。相較於參考文獻，簡化熱傳模型燃料溫度差異較小，外圍石墨塊之溫度和文獻差異約180K。同為石墨的燃料護套與文獻溫度相當接近，差異大約在數十度內。
未來在高溫氣冷式反應器自主設計研究上，HTTR由於其爐心燃料設計的複雜性會造成分析上的諸多困擾，較不理想。往後爐心設計可以參考GTHTR300或INL-HTGR的設計方式去發展。兩者在燃料設計上相對簡單，INL-HTGR將氦氣通道與燃料棒獨立(圓形流道)，在熱流分析上會相對容易。另外，HTTR的徑向採不均勻的鈾濃縮度使功率分布比GTHTR300平坦，亦是未來在設計上值得考量之處。

The purpose of this study is to develop a scheme of neutronic and thermal-hydraulic coupling for high temperature gas-cooled reactor core analysis. This is done by gathering and improving previous works in PTSL, and by completing the methodology in thermal-hydraulic calculation. The coupling studies are tested on the JAEA designed HTTR and GTHTR300 core calculation..
Monte Carlo method is used for the neutronic criticality calculation for the eigenvalue and power distribution. Two methods were used in the thermal-hydraulic calculation for solving the temperature distribution, an analytical method using simplified thermal-hydraulic model and a numerical method using computational fluid dynamic model.
The iteration between the neutronic and thermal hydraulic calculation converged in a few steps for both GTHTR300 and HTTR core calculations. The k_eff of is ~ 1.080, and 1.032 respectively. The converged axial power distribution is bottom peak for GTHTR300 and top peak for HTTR core.
For GTHTR300 core, the temperature distribution calculated by the analytical method is close to the one obtained by using the CFD model. Both results are closed to the reference results of JAEA, except near the middle of the core.
For HTTR core, result from calculation using the CFD model is lower than the result of using the simplified model. This may be due to the use of different heat transfer coefficient at the cladding/coolant surface. The calculated fuel temperature of HTTR is close to the reference result. However, difference of~ 180K was observed in the moderator temperature.
For future R&D of HTGR core design, complicated fuel design such as the HTTR core is not desireable. One should aim on simplier designs such as GTHTR300 and HTGR by INL. Non-uniform fuel enrichment will be considered if the power distribution does not meet the design criteria.

Chapter 1 Introduction 1
Chapter 2 Computational Tools and Methodologies 8
Chapter 3 Neutronic and Thermal-hydraulic Coupling Study on GTHTR300 Analysis 22
Chapter 4 Neutronic and Thermal-hydraulic Coupling Study on HTTR Analysis 47
Chapter 5 Conclusion and Future Works 67
References 70

[1] Javier Ortensi et al., “Deterministic Modeling of High Temperature Test Reactor”, Idaho National Laboratory, 2010.
[2] K. Kunitomi, S. Katanishi, S. Takada, X. Yan, and N. Tsuji, “Reactor Core Design of Gas Turbine High Temperature Reactor 300”, Nuclear Engineering and Design, 230, 349-366, 2004.
[3] H. Sato, “HTGR Plant Design”, Training Course on High Temperature Gas-cooled Reactor, Serpong, Indonesia, Oct. 19-23, 2015.
[4] T.W. Lin, “The Establishment and Verification of Neutron Cross Section Processing Procedure for High Temperature Gas Cooled Reactor Core Calculation”, Master thesis, National Tsing Hua University, 2014
[5] M.H. Chiang, ”Prismatic High Temperature Gas Cooled Reactor Benchmark Calculation and Fuel Assembly Calculation”, Master thesis, National Tsing-Hua University, 2014.
[6] 田揚仟，「由六角柱型高溫氣冷式反應器爐心功率分佈計算看燃料組件設計」，國立清華大學核子工程與科學研究所，碩士論文，2013。
[7] C. C. Tsai, “Establishment of Coupling Model between Core Calculations and Thermal-Hydraulic Calculations for Prismatic-Type High Temperature Gas-Cooled Reactor”, Master thesis, National Tsing-Hua University, 2016.
[8] N. Fujimoto and N. Nojiri, “Burnup Characteristic of Burnable Poison and Core Characteristic of HTTR”, JAEA-Technology 2005-008, Japan Atomic Energy Agency, Jan. 2006.
[9] K.Y. Cho, “Oxidation Behavior of Nuclear Graphite(IG110) with Surface Roughness”, Journal of the Korean Ceramic Society, Vol. 43, No. 10, pp. 613-618, 2006.
[10] Folsom CP, Effective Thermal Conductivity of Tri-Isotropic (TRISO) Fuel Compacts, All Graduate Thesis and Dissertations, Paper 1448, Utah State University, 2012.
[11] O. Hirofumi, “Re-evaluation of Maximum Fuel Temperature of the HTTR at Normal Operation”, IAEA Technical Meeting on Re-evaluation of Maximum Operating Temperatures and Accident Conditions for High Temperature Reactor Fuel and Structural Materials, Vienna, Austria, 10-12 July, 2012.
[12] N. Nojiri, S. Shimakawa, K. Takamatu, Y. Ishii, S. Kouno, S. Kobayashi,
T. Kawamoto, and T.Iyoku, “Power Distributions in the High Temperature Engineering Test Reactor (HTTR) by Measuring Gross Gamma Ray from the Fuel Assembles”, JAERI-Tech 2003-086, Japan Atomic Energy Research Institute, Nov. 2003.