核能之使用主要是建構於核能保安、核能安全與核不擴散做為和平使用核能的基礎原則。近年來由於電腦技術進步飛快，除了電廠安全分析進步外，電廠各組件之結構也成為了另一項分析之要點。本研究即利用計算流體力學(Computational Fluid Dynamics, CFD)針對壓水式反應器壓力槽(Reactor Pressure Vessel, RPV)進行壓熱震(Pressurized thermal shock, PTS)之模擬。本研究根據過去針對德國ROCOM之實驗數據進行分析技術校驗之研究經驗作為基礎，以確保後續在核三廠反應器壓力槽模型建立上具有一定之可行性。另外也根據美國機械工程師學會(American Society of Mechanical Engineers, ASME)所提出之驗證及確認(Verification and Validation, V&V)步驟進行不準度分析，使模型之網格不準度合乎國際規範之標準。在暫態事故下，壓熱震對壓水式反應器於事故或暫態期間的完整性有顯著影響之現象，當壓熱震導致反應器壓力槽破損，將導致更嚴重的冷卻水流失與後續救援的問題。本研究根據資料整理與比對，利用清楚簡單之方式建立PIRT表。最後，再經由PIRT表所評分出之MSLB、SO-1、SBLOCA作為反應器壓力槽熱流分析之條件，並且根據所收集的事故邊界條件，進行了事故狀態下反應器壓力槽模型之溫度變化分析，其結果顯示，降流區內部中子屏蔽板與冷熱端環路的幾何關係，將導致部分壓力槽內壁具有較高的降溫速率。此一成果除了證實本研究之成果達到本研究精進我國結構可靠度分析技術之研究目的外，未來更可為我國核電機組之結構可靠度安全評估貢獻一已之力。

Safety analysis has been the most important issue for the design of nuclear power plant (NPP) recently. Computational Fluid Dynamics (CFD) methodology can be applied in predicting the detailed knowledge of thermal-hydraulic phenomena in the applications of nuclear safety. In the present work, PWR pressure vessel model has been developed with CFD to predict the exhaustive of thermal-hydraulic behaviour in the downcomer of PWR. In the meanwhile, the framework composed of PIRT, validation and mesh uncertainty has been made to discuss the effects under different mesh designs followed from ASME V&V. Based on the analysis of PIRT, SO-1, MSLB, and SBLOCA were set as transient events for simulation. The results showed that unexpected variations on temperature, shear stress and velocity can be found near the vessel wall during a short period. The positions in downcomer for the inlet of cold leg and the neutron shielding plate may affect the core flow distribution and enhance cooling ability for core in some parts of vessel. This study presented the procedure of uncertainty qualification, including PIRT, mesh uncertainty and the methodology of CFD model, for PWR has been developed. The result can be regarded as the damage risk in RPV for comparison with the safety regulation requirements on vessel as well as a reference for the operation of PWR plants in Taiwan.

中文摘要 i
Abstract ii
致謝 iii
目錄 iv
表目錄 vi
圖目錄 vii
第一章 緒論 1
1.1 研究動機與背景 1
1.2 分析技術發展工具 2
第二章 文獻回顧 3
第三章 理論模式與數值分析 7
3.1 基礎方程式 7
3.2 紊流模式 8
3.3 數值分析設定 13
第四章 反應器壓力槽熱水流分析模式與靈敏度測試 16
4.1 模型建立與網格測試 16
4.2 網格不準度測試 23
4.3 紊流模式影響評估 38
第五章 分析案例選定與邊界條件建立 46
5.1 壓熱震事件概述 46
5.2 邊界條件建立 52
第六章 分析結果討論 58
6.1 正常運轉狀態熱流分析 58
6.2 緊迫性壓熱鎮事故熱流分析 60
6.3 預期性壓熱震事故熱流分析 69
6.4 相關運轉程序書精進之可行性評估 75
第七章 結論與未來工作 77
參考資料 79
附錄A MSLB事故UDF程式碼 82
附錄B SO-1事故UDF程式碼 87
附錄C SBLOCA事故UDF程式碼 88

 

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