用過核子燃料最終處置場之母岩裂縫數值模擬、廢料罐體之峰值溫度分析及反應爐壓力容器受熱衝擊之影響

NCUIR > Engineering College > Graduate Institute of Civil Engineering > Electronic Thesis & Dissertation > Item 987654321/91758

Please use this identifier to cite or link to this item: http://ir.lib.ncu.edu.tw/handle/987654321/91758

Title:	用過核子燃料最終處置場之母岩裂縫數值模擬、廢料罐體之峰值溫度分析及反應爐壓力容器受熱衝擊之影響
Authors:	譚宇翔;Tan, Yu-Xiang
Contributors:	土木工程學系
Keywords:	用過核子燃料;深層地質處置場;裂縫之影響分析;廢料罐體之熱傳分析;Spent Nuclear Fuel;Deep Geological Repository;Fault zones;radiative heat transfer
Date:	2023-07-12
Issue Date:	2024-09-19 14:12:40 (UTC+8)
Publisher:	國立中央大學
Abstract:	用過核子燃料具有長期高放射性及高衰變熱熱能，須有效隔離至生物圈外，國際間經評估最安全且有效的處置措施為深層地質處置(deep geologic disposal)。本文以深層地質處置概念為基礎並且利用有限元素ABAQUS模擬處置場母岩裂縫與廢料罐體之熱傳導分析。本研究首先參考國際計畫DECOVALEX文獻所提供的模型幾何、材料參數、初始及邊界條件，並利用四分之一代表體積單元概念建立具同等效力模型，求得緩衝材料峰值溫度及再飽和時間並與文獻平行驗證，且近一步針對母岩裂縫之幾何進行數值影響分析。此外，第五章KBS-3V型處置法中國際間具有多種廢料罐體之型式，本文透過芬蘭POSIVA文獻研究多種燃料棒與廢料罐體於處置過程下的峰值溫度，並且進行邊界條件與發射率之影響的參數分析。最後本研究第六章參考核能工程與設計之文獻進行壓力容器受熱衝擊數值模擬分析，探討容器受熱衝擊下情形，並且與文獻結果進行平行驗證分析。;The use of nuclear fuel results in long-term high-level radiation and decay heat, which must be effectively isolated from the biosphere. The internationally recognized safest and most effective disposal method is deep geological disposal. This study is based on the concept of deep geological disposal and utilizes finite element ABAQUS to simulate the thermal conduction analysis of the host rock fracture zone and waste container body in the disposal site. The research refers to the geometric and material parameters, initial and boundary conditions provided in the DECOVALEX literature of international plans. Additionally, a quarter representative volume element concept is used to establish an equivalent model, which calculates the peak temperature and resaturation time of the buffer material and verifies it in parallel with the literature. Furthermore, numerical impact analysis is carried out on the geometry of the host rock fracture. In addition, there are multiple types of waste container bodies in the KBS-3V disposal method. This paper refers to the Finnish POSIVA literature to study the peak temperature of fuel rods and waste container bodies during the disposal process, and carries out parameter analysis on the influence of boundary conditions. The final chapter of this study conducted a numerical simulation analysis of pressure vessel thermal shock, referring to literature on nuclear engineering and design. The objective was to explore the effects of thermal shock on the vessel and validate the findings by comparing them with existing research results.
Appears in Collections:	[Graduate Institute of Civil Engineering] Electronic Thesis & Dissertation

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