本研究目的為進行六角柱型高溫氣冷式反應器全爐心的驗證計算與燃料組件計算。所探討的反應器以日本的高溫工學試驗研究爐(High Temperature Test Reactor)為對象。六角柱型高溫氣冷式反應器包含了兩個特性，第一個特性是燃料區域由微小的TRISO粒子所組成並散佈於石墨中，具雙層異質特性，影響燃料晶格計算處理方式。第二個特性是在反應器內是用石墨作為緩速材料，所以中子在反應器爐心所走的路徑相對較長，影響燃料組件計算的截面生成方式。
全爐心驗證計算使用蒙地卡羅方法。使用連續點能量截面計算求得的爐心有效增殖因數(keff)與實驗值相比，有約20~40mk的高估。採用多群截面比使用連續點能量截面計算keff值低5mk。
燃料組件計算使用決定論法的遷移計算以產生少群截面。結果顯示相較於輕水式反應器，高溫氣冷式反應器燃料組件計算需要產生比較多能群的少群截面組。決定論法程式進行燃料組件計算產生的26群截面與238/361群截面的遷移計算的kinf相當接近，相差不到1 mk。使用決定論法程式採supercell模型與採無限複製模型相比，產生的少群截面會對燃料組件kinf有最大約13~17mk的影響，對燃料組件內的中子通率分布影響有最大約4.5%的影響。

The purpose of this study is to carry out core benchmark calculation and fuel assembly calculation for prismatic-type high temperature gas-cooled reactor. The prismatic-type high temperature gas-cooled reactor chosen is the High Temperature Test Reactor (HTTR) of Japan. High temperature gas-cooled reactor has two features. (1) The fuel region is made of tiny TRISO particles dispersed in graphite matrix, therefore, double heterogeneity effect should be considered in the cell-level calculation. (2) Graphite is used as moderator, which makes the migration length relatively large comparing to the assembly dimension. This affects the proper way of assembly calculation for production of the few group cross sections.
Core benchmark calculations are performed by using Monte Carlo transport code. The multiplication factor keff obtained by using continuous-energy cross section is 20~40mk larger than the experimental data. When multi-group calculation is used, keff is about 5mk smaller than that calculated by using continuous-energy cross sections.
Deterministic transport codes are used for fuel assembly calculation. The results showed that, more energy group is needed for few group cross sections generation for high temperature gas-cooled reactor than for light water reactor. The 26-group constants can almost reproduce the 238/361 group calculation results (< 1mk in kinf ). The use of supercell model shows that incorrect environment while producing fuel assembly cross sections will result in 13~17mk difference in kinf, and 4.5% difference in the neutron flux distribution in the fuel assembly.

目錄
摘要 I
Abstract II
誌謝 IV
目錄 V
表目錄 IX
圖目錄 XIV
第一章 緒論 1
1.1 高溫氣冷式反應器介紹 1
1.2 高溫氣冷六角柱型反應器 2
1.3 高溫氣冷球床式反應器 4
1.4 研究動機 5
第二章 計算工具介紹 7
第三章 HTTR爐心模型建立與驗證計算 21
3.1 HTTR爐心介紹 21
3.2 HTTR爐心模型建立 26
3.3 HTTR驗證問題計算 32
第四章 HTTR爐心燃耗計算 42
4.1 溫度效應計算 42
4.2 燃耗計算 44
第五章 HTTR燃料組件計算 51
5.1 HTTR燃料組件遷移計算 53
5.2 燃料組件計算差異分析 66
5.3 室溫燃料組件計算 74
5.4 Supercell模型室溫燃料組件計算 83
5.5 高溫燃料組件計算 102
5.6 Supercell模型高溫燃料組件計算 107
第六章 結論與未來建議工作 119
參考文獻 122

[1] http://www.gen-4.org/
[2] Javier Ortensi, “Prismatic Core Neutronics Design and Fuel Cycle”, Training Course on HTGR Technology, China, 2012.
[3] J.D. Bes, N. Fujimoto, B.H. Dolphin, L. Snoj, A. Zukeran, Evaluation of the Start-Up Core Physics Tests at Japan’s High Temperature Engineering Test Reactor (Fully-Loaded Core), INL/EXT-08-14767 Rev. 2, Idaho National Laboratory (INL), 2010.
[4] Kunitomi, K., et al., Evaluation of High Temperature Gas-cooled Reactor Performance: Benchmark Analysis Related to Initial Testing of the HTTR and HTR-10, IAEA-TECDOC-1382, International Atomic Energy Agency, Vienna, Austria, 2003.
[5] Shinzo Saito, Toshiyuki Tanaka, Yukio Sudo, et al., Design of high temperature engineering test reactor (HTTR), Report of the Japan Atomic Energy Research Institute, JAERI 1332, 1994.
[6] A. Yamamoto, T. Endo, "Overview of Core Simulation Methodologies for Light Water Reactor Analysis, International Electronic Journal of Nuclear Safety and Simulation, 2, pp.12-21 (2011).
[7] Javier Ortensi, “Supercell Depletion Studies for Prismatic High Temperature Reactors”, HTR 2012, 2012.
[8] Y. Wang, et al., “Improved Neutronics Treatment of Burnable Poisons for the Prismatic HTR”, HTR 2012, 2012.
[21] Mark D. DeHart, “Advancements in Generalized-Geometry Discrete Ordinates Transport for Lattice Physics Calculations”, Proc. of PHYSOR-2006, Vancouver, British Columbia, Canada, Sept. 10–14, 2006.
[22] Zhaopeng Zhong, et al. "Implementation of Two-level Coarse-mesh Finite Difference Acceleration in an Arbitrary Geometry, Two-dimensional Discrete Ordinates Transport Method." Nuclear science and engineering, 158.3 (2008): 289-298.
[25] Michael L. Fensin, John S. Hendricks, Samim Anghaie, “Improved Reaction Rate Tracking and Fission Product Yield Determinations for the Monte Carlo-Linked Depletion Capability in MCNPX”, Nuclear technology 164.1 (2008): 3-12.
[27] A. Hébert, “Towards DRAGON Version4”, paper presented at the Topical Meeting on Advances in Nuclear Analysis and Simulation, September 10 - 14, Vancouver, Canada (2006).
[28] Vincent Descotes, Javier Ortensi, Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor, INL/EXT-10-20677, Idaho National Laboratory (INL), 2011.
[29] Javier Ortensi, Joshua J. Cogliati, et al., Deterministic Modeling of the High Temperature Test Reactor, INL/EXT-10-18969, Idaho National Laboratory (INL), 2010.
[30] Michael A. Pope, Javier Ortensi, Abderrafi M. Ougouag, “Investigation of Supercells for Preparation of Homogenized Cross Sections for Prismatic Deep Burn VHTR Calculations”, HTR 2010, 2010.
[31] T. A. Taiwo, et al., Evaluation of High Temperature Gas-Cooled Reactor Physics Experiments as VHTR Benchmark Problems, ANL-GenIV-059, Argonne National Laboratory, 2005.
[32] T.A. Taiwo, T.K. Kim, Evaluation of the DRAGON code for VHTR design analysis, ANL-GENIV-060. Argonne National Laboratory, 2006.
[33] A. Hébert , “A Collision Probability Analysis of the Double-Heterogeneity Problem”, Nuclear science and engineering: 115, 177-184 (1993).
[35] http://httr.jaea.go.jp
[36] D. E. Ames II, “Configuration Adjustment Potential of the Very High Temperature Reactor Prismatic Cores with Advanced Actinide Fuels”, Master Thesis, Texas A&M University (2006).
[38] Ilas, Dan, Jess Gehin, "HTTR Fuel Block Simulations with SCALE", PHYSOR2010, 2010.
[39] J.Y. Wang, M.H. Chiang, R.J. Sheu, Y-W. H. Liu, “HTTR Criticality Calculations
with SCALE6: Studies of Various Geometric and Unit-Cell Options in Modeling”, PHYSOR2012 – Advances in Reactor Physics – Linking Research, Industry, and Education, Knoxville, Tennessee, USA, April 15-20, 2012.
[40] N. Nojiri, S. Shimakawa, N. Fujimoto, and M. Goto, “Characteristic Test of Initial HTTR Core” , Nucl. Eng. Des., 233: 283-290 (2004).
[41] M. Goto, S. Shimakawa, Y. Nakao, “Impact of Revised Thermal Neutron Capture Cross Section of Carbon Stored in JENDL-4.0 on HTTR Criticality Calculation”, J. Nucl. Sci. Tech., 48, p.965, 2011.
[42] J.D. Bess, N. Fujimoto, J.W. Sterbentz, L. Snoj, A. Zukeran, Evaluation of Zero-Power, Elevated-Temperature Measurements at Japan's High Temperature Engineering Test Reactor, INL/EXT-10-19627, Idaho National Laboratory (INL), 2011.
[43] M.H. Chiang, J.Y. Wang, R.J. Sheu, Y-W. H. Liu, “Evaluation of the HTTR Criticality and Burnup Calculations with Continuous-energy and Multigroup Cross Sections”, HTR 2012, 2012.
[44] http://www.polymtl.ca/merlin/libraries.htm
[45] N. Fujimoto, N. Nojiri, “Burnup Characteristics of Burnable Poison and Core Characteristics of HTTR”, JAEA-Technology-2005-008, Japan Atomic Energy Agency, January (2006).
[46] “HTR Design Criteria”, http://www.janleenkloosterman.nl/vhtr_course_200811.php
[47] “IAEA Activities on Gas-Cooled Reactor Technology Development”, Training Course on HTGR Technology, China, 2012
[48] M. D. DeHart, M. L. Williams, S. M. Bowman, “Parallel Computing in SCALE”, PHYSOR 2010 – Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburgh, Pennsylvania, USA, May 9-14, 2010