馬鞍山電廠TRACE模式發展與設計基準喪失冷卻水事故分析之應用

Development of Maanshan TRACE Model and Application of Design Basis LOCA Analysis

楊融華(Jung-Hua Yang)

指導教授 ： 施純寬 ； 王仲容

國立清華大學/原子科學院/工程與系統科學系/博士(2013年)

摘要

馬鞍山電廠為三迴路壓水式反應器電廠，由西屋公司設計製造，廠區內有兩座運轉機組，兩座機組設計100%熱功率為2775MW。喪失冷卻水事故(Loss of Coolant Accident, LOCA)為輕水式反應器設計基準事故中其中一項重要的安全分析，雖然電廠發生破口喪失冷卻水事故的機率極低，但安全系統設計上仍然要確保當喪失冷卻水事故發生時電廠有足夠的冷卻能力以防止爐心燃料熔毀。因此探討台灣馬鞍山電廠小功率提昇後若發生冷卻水流失事故，電廠是否能維持法規中所要求的燃料護套溫度不能超過2200℉之規定，對於經濟貢獻具有相當的價值性。本研究首先利用分離效應測試(Separate Effects Tests)，進行一些重要的熱水流現象的驗證以及分析探討，如臨界流(Critical Flow)、逆向流(Countercurrent Flow)、緊急爐心注水旁通(ECC Bypass)等，接著建立馬鞍山電廠整廠TRACE分析模式，並利用電廠起動測試數據以及FSAR報告數據來驗證馬鞍山電廠整廠TRACE分析模式的準確性。此外，核研所為了研究西屋三迴路壓水式反應器(特別是針對馬鞍山電廠)的熱流現象而建造了IIST實驗設施，因此在本研究中也建立了IIST TRACE分析模式並利用IIST的實驗數據來證明其模式模擬的準確性。IIST破口實驗分析的經驗也回饋於馬鞍山電廠發生喪失冷卻水事故的安全分析上。
在經過前述的評估與驗證後，本研究利用馬鞍山電廠TRACE模式進行喪失冷卻水事故之分析，分析案例包含大破口事故以及小破口事故。 安全分析中，針對電廠許多重要熱水力參數(例如爐心水位、破口流率、系統壓力、和燃料護套尖峰溫度等)做計算與討論。破口喪失冷卻水事故分析結果顯示馬鞍山電廠TRACE分析模式不僅能預測電廠熱流現象與行為，更能提供評估燃料護套尖峰溫度較大的餘裕。

關鍵字

馬鞍山電廠 ； TRACE ； 喪失冷卻水事故

並列摘要

The Maanshan nuclear power plant operated by Taiwan Power Company is a Westinghouse PWR. The rated core thermal power is 2775 MW. The Loss of coolant accident (LOCA) is one of the design basis accidents (DBAs) in light water reactors. Although a LOCA is considered very unlikely to happen, the safety systems must be designed to secure adequate cooling of the reactor core during the accident to prevent the meltdown of the core. By the separate effects tests, several important phenomena of thermal-hydraulic, such as critical flow, CCFL, and ECC bypass, are discussed and verified. Then each separate models are integrated to the Maanshan TRACE model. The Maanshan TRACE model is also verified by referring to the start-up test data and FSAR data. Furthermore, IIST facility had been established for safety studies of the Westinghouse three- loops PWR, especially for Maanshan nuclear power plant. The IIST TRACE model also has been established and verified its accuracy by referring ot the test data of IIST facility LOCA experiments. The experiences of IIST LOCA analysis are used in the Maanshan LOCA analysis.
After verifications, this study uses the Maanshan TRACE model to analysis the two types of LOCA, LBLOCA and SBLOCA. Several important thermal-hydraulic parameters, such as core water level, break flow rate, system pressure, and peak cladding temperature, are calculated and discussed in this study. LOCA analytical results indicate that the Maanshan TRACE model predicts not only the behaviors of important plant parameters in consistent trends with the Maashan FSAR data, but also provides a greater margin for the PCT evaluation.

並列關鍵字

Maanshan ； TRACE ； LOCA

參考文獻

3. Ferng, Y. M. and Lee, C. H., “Numerical simulation of atural circulation experiments conducted at IIST facility”, Nucl. Eng. Des., Vol. 31, 83, 1994.

4. Ferng, Y. M. and Lee, C. H., “A comparison of the RELAP5/MOD3 code with IIST natural circulation experiments”, Nucl. Technol., Vol. 111, 34, 1995.

7. Lee, C. H., Liu, T. J., Way., Y. S., and Hsia, D. Y., “Investigation of mid-loop operation with loss of RHR at INER Integral System Test (IIST) facility”, Nucl. Eng. Des., Vol. 163, 349, 1996.

8. Lee, C. H., Huang, I. M., Chang, C. J., Liu, T. J., and Ferng, Y. M., “Using an IIST SBLOCA experiment to assess RELAP5/MOD3.2”, Nucl. Technol., Vol. 126, 48, 1999.

10. C. H. Lee, T. J. Liu, Y. S. Way, andD. Y.Hsia, “Investigation of mid-loop operation with loss of RHR at INER integral system test (IIST) facility,” Nuclear Engineering and Design, vol. 163, no. 3, pp. 349–358, 1996.