壓水式反應器壓熱震之CFD技術建立與應用

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。

關鍵字

壓熱震；壓水式反應器；計算流體力學； PIRT ； V&V ； MSLB ； SO-1 ； SBLOCA

並列摘要

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.

並列關鍵字

PTS ； PWR ； CFD ； PIRT ； Verification and Validation ； SO-1 ； MSLB ； SBLOCA

參考文獻

3. U.S. NRC 10 CFR 50 Appendix G,”Fracture Toughness Requirements”, January 2011.

5. U.S. NRC 10 CFR 50.61,”Fracture Toughness Requirements”, January 2011.

11. ASME, “EN-N-641 Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements Section XI Division 1 ,” 2004.

12. B.R. Bass, C.E. Pugh, J. Sievers, and H. Schulz, “Overview of the International Comparative Assessment Study of Pressurized Thermal-Shock in Reactor Pressure Vessels,” Int. J. Pressure Vessels and piping, ol 78, p.p. 197-211, 2001.

13. I. Jeong, C. Jang, J.H. Park, S.Y. Yull, T.E. Jin, H.G. Yuem, and S.G. Jeong, “Lessons learned from the plant-specific pressure thermal shock intergrity analysis on an embrittled reactor pressure vessel,” Int. J. Pressure Vessel and Piping, Vol. 78, p.p. 99-109, 2001.

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壓水式反應器壓熱震之CFD技術建立與應用

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