國聖電廠目前功率已提升至2943 MWt，且台灣電力公司已有計畫進行中幅功率提升，目標將國聖電廠提升至3030 MWt，為原始功率(2894 MWt)之104.7%。功率提升計畫前須進行一系列暫態分析，確認功率提升後，機組在預期事故下尚能維持燃料棒護套的完整穩定性。本研究使用之RELAP5已整合NSSS與BOP的國聖電廠輸入檔，但是功率仍舊是原定功率2894 MWt。本研究先建構國聖電廠功率提升後RELAP5穩態輸入檔，再進行增壓暫態分析模擬。增壓暫態事件中，假設主蒸氣管路上的三種閥門(主蒸氣隔離閥、汽機斷止閥與汽機控制閥)分別突然關閉，使反應爐壓力劇烈上升，使冷卻水空泡分率下降，中子減速效果增加，反應度上升，使功率大幅上升，直到觸發急停訊號並快速插入控制棒後，鏈式反應減緩而功率下降；本研究的增壓暫態事件中設定安全釋壓閥動作以避免反應爐頂壓力超過ASME規範之限值。以RELAP得到的暫態熱水流資訊和國聖電廠燃料棒幾何性質，做為邊界條件輸入FRAPTRAN程式模擬燃料棒在增壓暫態中的機械性質變化(內外徑、護套溫度、內壓等)，預估燃料棒在功率提升後的增壓暫態事件中，燃料棒護套是否有破損的風險。本研究以SNAP介面中的DAKOTA輸入設定中加入燃料棒製造公差進行不準度分析，評估製造公差的影響。本研究的結果與TRACE與FSAR的結果比較，作為日後國聖電廠功率提升計畫的參考。研究的結果顯示主蒸氣隔離閥關閉事件中有較高的護套溫度、護套環應變與燃料丸熱焓值；而考慮燃料棒製造公差不準度的影響後，燃料棒各項參數仍能滿足引用的法規標準。

Abstract
Kuosheng Nuclear Power Plant, which is the second nuclear power of Taiwan Power Company, has uprated its power from 2894 MWt to 2943 MWt. Taipower has a plan to uprate the power to 3030 MWt (104.7%). In the present study, the uncertainty analyses of fuel rod performance in overpower transients under the stretch power uprate condition are performed using RELAP5 and FRAPTRAN codes. RELAP5 code is used to simulate the thermal hydraulic response of the system in over pressure transients. The results of RELAP5 simulation are used as boundary conditions of FRAPTRAN code calaculations to assess the performance of fuel rods. The target parameters of the assessment are cladding temperature, cladding hoop strain, and fuel enthalpy. The impact of fuel rod manufactures uncertainty on the uncertainty of target parameters are studied statistically. The parameters included in the assessment are cladding outer diameter, cladding roughness, fuel pellet outer diameter, fuel pellet density, fuel pellet roughness, and rod fill pressure. Based on the Wilk’s formula, 207 cases of FRAPTRAN runs are performed to obtain 95th value of the target parameters with 95th confidence. The output of 207 FRAPTRAN runs are also analyzed using parametric method.
The RELAP5 input deck of the plant is initialized to the new steady state conditions under higher power. The steady state conditions are compared with these as predicted by TRACE and as documented in the Final Safety Analysis Report (FSAR). The overpressurization transients considered are main steam line isolation valves closure, turbine stop valves closure and turbine control valves closure. This closure of valves increased dome pressure The pressure of reactor coolant system are controlled by safety relief valves and the dome pressure would not exceed the criteria specified by ASME. The results demonstrate that the responses of target parameters are most serious during the transient of main steam line isolation valves closure event. The results of uncertainty analysis show that the 95/95th values of target parameters are within the regulatory limits for all the transient.

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