研究生: |
吳炘融 Wu, Shin-Rong. |
---|---|
論文名稱: |
球床型高溫氣冷式反應器不停爐裝卸燃料之臨界及燃耗分析 Criticality and burnup analysis of online refueling process in pebble-bed high temperature gas-cooled reactor |
指導教授: |
梁正宏
Liang, Jenq-Horng |
口試委員: |
陳健湘
Chen, Chien-Hsiang 宋大崙 Sung, Ta-Lun 趙得勝 Chao, Te-Sheng |
學位類別: |
碩士 Master |
系所名稱: |
原子科學院 - 核子工程與科學研究所 Nuclear Engineering and Science |
論文出版年: | 2017 |
畢業學年度: | 106 |
語文別: | 中文 |
論文頁數: | 62 |
中文關鍵詞: | 高溫氣冷式反應器 、球床型反應器 、燃耗計算 、燃料裝卸模型 |
外文關鍵詞: | HTR-10, pebble-bed reactor, burnup analysis |
相關次數: | 點閱:1 下載:0 |
分享至: |
查詢本校圖書館目錄 查詢臺灣博碩士論文知識加值系統 勘誤回報 |
本論文之目的為針對高溫氣冷式反應器HTR-10,建立一可模擬不停爐的連續裝卸燃料之軸、徑向燃料裝卸的模型。HTR-10為北京清華研究院所主導設計及建造的第四代反應器,係世界第三座模組式球床式反應器。球床式反應器的特點在於不停爐裝卸燃料 (online-refueling),能大幅減少停爐時間並有效使用燃料,獲得高燃耗的用過燃料。傳統的輕水式反應器中,其爐心幾何不會隨著時間改變,故可用靜態模擬的方式探討其爐心特性。然而,球床式反應器之燃料營運採用不停爐裝卸燃料,導致無法直接以靜態方式模擬其運轉情形。也因此,建立可近似動態模擬的模型是非常重要的。本論文使用MCNPX以及ENDF/B-VII截面資料庫來進行相關的計算工作,爐心溫度假設為900 K,並為均勻分佈。
不停爐裝卸燃料之模擬上因幾何複雜,計算時間過長,因此本論文為了進行較快速且不失準確度的計算,在模型幾何上進行了部分簡化,並從文獻與模擬結果證實簡化只略為影響計算。不停爐裝卸燃料程序上,分為兩種不同的模型:環狀殼層燃料裝卸模型與混合式燃料裝卸模型。環狀殼層燃料裝卸模型,其針對裝卸燃料時徑向燃耗分佈與爐心特性進行分析,為補充前人研究上只探討軸向之燃耗分佈的不足。混合式燃料裝卸模型則試圖以流道的概念模擬軸向與徑向之裝卸燃料,使結果更貼近實際運轉。結果顯示,環狀殼層燃料裝卸模型可在較短的時間上得出具代表性的徑向燃料特性。而混合式燃料裝卸模型則在燃耗表現上相當優異,平均卸載燃耗與文獻中的80 GWd/tHM符合,各流道間的燃耗分佈也十分接近文獻中5次通過的結果,可證明此混合式燃料裝卸模型較以往模型更加貼近實際運轉狀況。
This study aims to propose an accurate online refueling model for HTR-10 to simulate the burnup features in both axial and radial direction. HTR-10 is constructed and designed by Institute of Nuclear and New Energy Technology (INET), which is located in Tsinghua university, Beijing, China. HTR-10 is a pebble bed reactor (PBR) and also belongs to the type of high temperature gas-cooled reactors (HTGRs), which is one of the most promising candidates in GenIV reactors. One of the features in PBR is the online-refueling, which can reduce outage time, enhance the efficiency in fuel use, and increase the burnup value for discharged fuel. However, this feature would make the calculation more difficult. Therefore, a simplified model is highly demanded to dynamically simulate the online-refueling process. All the calculations were performed using MCNPX and ENDF/B-VII neutron cross-section data library. In addition, the temperature was set to be uniform at 900 K for all elements in the core.
For rapid and precise calculations, some simplifications of geometry were assumed in this study. A series of verification tests were conducted to figure out the deviations of these simplifications, which showed acceptable results. Two different fuel movement models were proposed and established in this study, namely, shell-by-shell and layer-mix-shell fuel movement models. The purpose of the former one is to analyze the burnup characteristics in radial direction, which is a supplement for the previous fuel movement model that only considered the axial burnup characteristics. The latter one considers the fuel movement in both axial and radial directions based on the concept of fuel channels. The results revealed that the shell-by-shell fuel movement model can present the radial burnup characteristics in relative short time. Furthermore, the layer-mix-shell fuel movement model showed that the average discharged burnup is approximately 80 GWd/tHM, making a good agreement with the designed burnup value of spent fuels in HTR-10 as published in the literatures.
[1] Statistical Review of World Energy 2016, bp Global, 2016.
[2] A Technology Roadmap for Generation IV Nuclear Energy Systems, Generation IV International Forum, 2002.
[3] 吴宗鑫,张作义,先进核能系统和高温气冷堆,第一版,清华大学出版社,北京,西元2004年。
[4] R.D. Burnette, N.L. Baldwin, Primary coolant chemistry of the Peach Bottom and Fort St. Vrain high-temperature gas-cooled reactors, General Atomic Company, San Diego, California, USA, 1980.
[5] K. Kunitomi, X. Yan, T. Nishihara, N. Sakaba, T. Mouri, “JAEA’S VHTR for hydrogen and electricity cogeneration: GTHTR300C”, Nuclear Engineering and Technology, Vol. 39, No. 1, pp. 9-20, 2007.
[6] 百度科技HTR-PM頁面,http://baike.baidu.com。
[7] K. Kunitomi, Y. Sun, S. Ball, H.L. Brey, M. Methnani, Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to Initial Testing of the HTTR and HTR-10, IAEA-TEDOC-1382, International Atomic Energy Agency, 2007.
[8] 吳尚謙,「高溫氣冷式反應器HTR-10之模擬計算與燃耗特性分析」,國立清華大學核子工程與科學研究所,碩士論文,西元2013年。
[9] Y. Yang, Z. Luo, X. Jing, Z. Wu, “Fuel Management of the HTR-10 including the Equilibrium state and Running-in Phase”, Nuclear Engineering and Design, Vol. 218, pp. 33-41, 2002.
[10] W.K. Terry, S.S. Kim, L.M. Montierth, J.J. Cogliati, A.M. Ougouag, “Evaluation of HTR-10 Reactor as a Benchmark for Physics Code QA”, PHYSOR-2006, Vol. 39, No. 4, ANS Topical Meeting on Reactor Physics, 2006.
[11] 王孟仁,「高溫氣冷式研究用反應器HTR-10之爐心特性分析與計算」,國立清華大學核子工程與科學研究所,碩士論文,西元2011年。
[12] X-5 Monte Carlo Team, MCNP-A General Monte Carlo N-Particle Transport Code. Version 5 Volume I: Overview and Theory, LA-UR-03-1987, Los Alamos National Laboratory, 2003.
[13] J.S. Hendricks, G.W. McKinney, M.L. Fensin, M.R. James, J.W. Durkee, J.P. Finch, D.B. Pelowitz, L.S. Waters, M.W. Johnson, MCNPX 2.6.0 Extensions, LA-UR-08-2216, Los Alamos National Laboratory, 2008.
[14] T. Goorley, M. James, T. Booth, F. Brown, J. Bull, L.J. Cox, J. Durkee, J. Elson, M. Fensin, R.A. Forster, J. Hendricks, H.G. Hughes,1 R. Johns, B. Kiedrowski, R. Martz, S. Mashnik, G. Mckinney, D. Pelowitz, R. Prael, J. Sweezy, L. Waters, T. Wilcox, T. Zukaitis, Initial MCNP6 Release Overview – MCNP6 version 1.0, LA-UR-13-22934, Los Alamos National Laboratory, 2013.
[15] M.J. Wang, J.J. Peir, R.J. Sheu, J.H. Liang, “Effects of geometry homogenization on the HTR-10 criticality calculations”, Nuclear Engineering and Design, Vol. 271, pp. 356-360, 2014.
[16] 张竞宇,李富,魏春琳,孙玉良,「基于蒙特卡罗方法对VSOP模型的keff验证」,原子能科学技术,第47卷第3期,第350-354頁,西元2013年。
[17] J. Rosales, A. Muñoz, C. García, L. García, C. Brayner, J. Pérez, A. Abánades, “Computational Model for the Neutronic Simulation of Pebble Bed Reactor’s Core Using MCNPX”, International Journal of Nuclear Energy, Vol. 2014, pp. 1-12, 2014.
[18] 朱贵凤,邹杨,李明海,严睿,彭红花,徐洪杰,「球床高温堆平衡态燃耗计算程序的开发」,原子能科学技术,第49卷第5期,第890-896頁,西元2015年。
[19] J.R. Lamarsh, A.J. Baratta, Introduction to Nuclear Engineering, 3rd edition, Prentice Hall, 2011.
[20] 经荥清,张旭,罗经宇,「HTR-10-MW高温气冷实验堆换料方式的研究」,核科学与工程,第13卷第2期,第119-125頁,西元1993年。
[21] 吴宗鑫,经荥清,「HTR-10的燃料管理」,清华大学学报:自然科学版,第41卷第4期,第120-123頁,西元2001年。
[22] H. Jeong, S.H. Chang, “Estimation of the Fission Products, Actinides and Tritium of HTR-10”, Nuclear Engineering and Technology, Vol. 41, pp. 729-738, 2009.
[23] D.R. Vondy, J.A. Lane, A.T. Gresky, “Production of Np237 and Pu238 in Thermal Power Reactors,” I & EC Process Design and Development, Vol. 3, No. 4, pp. 293-296, 1964.