國聖電廠透過小幅度功率提升，目前已經功率提升至2943 MWt，接下來台灣電力公司計畫進行中幅度功率提升，將功率提升至3030 MWt，為原定運轉功率之105%。功率提升計畫前須進行一系列暫態分析，確認功率提升後，機組遇上嚴重事故尚能維持其穩定性。本研究使用美國核管會(United States Nuclear Regulatory Commission, U.S. NRC)最先進熱水流安全分析程式TRACE，進行國聖電廠功率提升後，增壓暫態反應之模擬分析。在此研究之增壓暫態反應中，假設主蒸汽管路上三種閥門(包含主蒸氣隔離閥、汽機斷止閥與汽機控制閥)突然關閉，使得反應爐頂壓力上升，造成空泡分率下降，中子減速效果增加，反應度突然上升，因此功率大幅上升，直到急停訊號釋出並插入控制棒後，核分裂反應逐漸受到控制、功率下降；而安全釋壓閥的作動亦保護反應爐頂壓力不超過ASME規範之限值。此外，為彌補熱水流模擬程式於暫態時對於燃料棒分析之限制，本研究尚結合了FRAPTRAN程式，將TRACE所得到之暫態熱水流性質作為邊界條件，搭配暫態功率變化，並蒐集國聖電廠燃料棒幾何性質資料，進行燃料棒暫態性質分析；並且搭配DAKOTA不準度分析程式，考慮燃料棒的製造公差所造成的影響，得到燃料棒於增壓暫態時的性質變化及其不準度區間，藉由觀察此不準度區間，能更有信心地進行燃料棒的損壞判斷。本研究成功建立了一套熱水流與燃料性質模擬分析之流程，並成功利用SNAP介面結合DAKOTA與FRAPTRAN程式進行燃料棒暫態不準度分析，且自行編寫出一個FRAPTRAN不準度分析數據擷取程式，方便未來研究者快速地擷取與讀取數據。

After the measurement uncertainty recapture power uprates, Kuosheng Nuclear Power Plant has uprated the power to 2943 MWt. Recently, Taiwan Power Company is concerned in stretch power uprated plan and uprates the power to 3030 MWt, which is 105% of the designed power. Before the stretch power uprates, several transient analysis should be done for ensuring that the power plant could maintain stability in high power operating conditions. The advanced thermal hydraulic analysis code TRACE which is conducted by U.S. NRC was applied for overpressurization transient including main steam line isolation valves closure, turbine stop valves closure and turbine control valves closure. This closure of valves increased the dome pressure; as a result, the void fraction inside the reactor core decreased. Declination of the void fraction will increase the reactivity feedback; hence, the power increased rapidly until the reactor scram. Further, with safety relief valves open, the dome pressure would not exceed the criteria regulated by ASME. In addition, to cover the insufficiency of thermal hydraulic code, fuel rod transient analysis code FRAPTRAN was applied. To perform fuel rod transient analysis, thermal information from TRACE code would be entered into FRAPTRAN with power history. In addition, DAKOTA code was applied to concern the influence of fuel rod manufacturing tolerance. With uncertainty bands from DAKOTA analysis, the criteria could be judged with more confidence. This research successfully established a procedure of thermal hydraulic and fuel rod property analysis. Further, with SNAP interface, the FRAPTRAN and DAKOTA were combined successfully to perform the uncertainty analysis. Also, a FRAPTRAN uncertainty data extracting program, which can capture several interesting parameters at the same time, was developed in this research.

摘要
Abstract
致謝
目錄
表目錄
圖目錄
壹、 緒論
壹、一. 前言
壹、二. 研究目的與方法
壹、三. 程式簡介與分析流程
貳、 文獻回顧
貳、一. 增壓暫態反應
貳、二. 不準度分析模式
貳、二-1. University of Pisa method：
貳、二-2. AEA Technology method：
貳、二-3. Probabilistic method：
參、 國聖電廠之TRACE與FRAPTRAN模式建立
參、一. TRACE模式建立
參、二. FRAPTRAN模式建立
參、三. 不準度分析模式建立
肆、 分析結果與討論
肆、一. 主蒸氣隔離閥關閉 (Main Steam Line Isolation Valves Closure)
肆、二. 汽機斷止閥關閉(Turbine Stop Valves Closure)
肆、三. 汽機控制閥關閉(Turbine Control Valves Closure)
肆、四. 暫態比較與討論
伍、 結論與建議
參考文獻
附錄A 國聖電廠TRACE模式設定
附錄B 國聖電廠FRAPTRAN模式設定
附錄C 國聖電廠DAKOTA模式設定

1. Taiwan Power Company, Final Safety Analysis Report for Kuosheng Nuclear Power Station Units 1&2 (FSAR). 2001, Taiwan Power Company: Republic of China (Taiwan).
2. Lin, K.Y., Verification of the Kuosheng BWR/6 TRACE Model with Load Rejection Startup Test, in ASME verification and validation 2012. 2012: Las Vegas, NV.
3. TRACE V5.0 user’s manual, U.S.N.R. Commission, Editor. 2010, U. S. Nuclear Regulatory Commission: Washington, DC.
4. K.J. Geelhood, W.G.L., C.E. Beyer, FRAPTRAN 1.4: Integral Assessment. 2011, Pacific Northwest National Laboratory.
5. K.J. Geelhood, W.G.L., C.E. Beyer, FRAPCON-3.4: Integral Assessment. 2011, Pacific Northwest National Laboratory.
6. Brian M. Adams et al, Dakota, A Multilevel Parallel Object-Oriented Framework for Design Optimization, Parameter Estimation, Uncertainty Quantification, and Sensitivity Analysis: Version 6.0 User's Manual. 2014, Albuquerque, New Mexico, USA: Sandia National Laboratories.
7. Applied Programming Technology, I., Symbolic Nuclear Analysis Package (SNAP) User's Manual. 2007: Applied Programming Technology, Inc.
8. THE U. S. NUCLEAR REGULATORY COMMISSION'S STRATEGY FOR REVISING THE RIA ACCEPTANCE CRITERIA. 2012 Top Fuel Reactor Fuel Performance Meeting. 2012.
9. Nuclear Fuel Safety Criteria Technical Review Second Edition. 2012, Issy-les-Moulineaux, France: ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT NUCLEAR ENERGY AGENCY.
10. Briggs, L.L., Uncertainty Quantification Approaches for Advanced Reactor Analyses. 2008, Nuclear Engineering Division, Argonne National Laboratory. p. 32.
11. Seung Wook Lee et al, ANALYSIS OF UNCERTAINTY QUANTIFICATION METHOD BY COMPARING MONTE-CARLO METHOD AND WILKS’ FORMULA. NUCLEAR ENGINEERING AND TECHNOLOGY, 2014. 46(4).
12. Wilks, S.S., Determination of sample sizes for setting tolerance limits. Annals of Mathematical Statistics, 1941. 12: p. 91-96.
13. Guba, A., M. Makai, and L. Pal, Statistical aspects of best estimate method - I. Reliability Engineering & System Safety, 2003. 80(3): p. 217-232.
14. K.J. Geelhood, W.G.L., C.E. Beyer, D.J. Senor, M.E. Cunningham, D.D. Lanning, H.E. Adkins, Predictive Bias and Sensitivity in NRC Fuel Performance Codes. 2009, Pacific Northwest National Laboratory: Richland, WA.