本研究希望了解台灣馬鞍山核電廠(核三廠)反應爐頂蓋因溫度梯度造成材料破裂的可能性。同時，利用計算流體力學的方法和STAR-CCM+的軟體可以知道反應器內部以及壁面的溫度分布。計算區域從上內部組件的底部到整個爐頂蓋區域，因此透過此分析可以了解有多少的高溫流體會經由控制棒導管流到頂蓋區域，並決定爐蓋區域內的溫度分布。為了減少計算時間，此次研究採用對稱假設來分析上內部組件和爐頂蓋區域的一半，共利用了六千五百萬個多面體網格，包含三層的邊界層網格。在壁面絕熱條件之下，此研究結果呈現爐頂蓋區域的流體流動之情形以及溫度分布，並得到總共2% 的高溫流體會從上內部組件一直往上升到爐頂蓋區域。 我們引進電廠中使用的有效劣化年(EDY) 和再檢查年 (RIY)並確保這些值在合理範圍時電廠不用增加再檢查的次數。改變冷端進口流量，算出24個噴嘴最少需要0.2%流量才不致使爐頂蓋溫度過高。比較不同的紊流模式所計算的結果，在本篇研究中建議使用RKE 為分析的模式。

This study aims to reveal the possibility of a crack due to the temperature gradient on RPV head in Maanshan Nuclear Power Plant in Taiwan. The temperature distribution inside RPV and on the RPV wall is calculated by Computational Fluid Dynamics (CFD) software STAR-CCM+. The computational domain starts from the bottom part of the upper plenum and continues throughout the entire dome region, so that the high temperature coolant leaking to the dome region through gaps between solids can be included to calculate the leak flow and temperature distribution in the dome region. In the present study, Computational cells are generated for half of the upper plenum region and dome region using symmetry assumption to save computer resources. The total number of cells is ~65 millions including 3 prism layers. The calculations not only provide the detailed information of flow and temperature distribution inside the dome region but also show 2% of high-temperature coolant entering to the dome region by the adiabatic assumption of all solid structures. Total effective degradation years (EDY) and reinspection years (RIY) for investigating the crack probability in Maanshan Nuclear Power Plant are applied to this study. The EDY and RIY factor (<1.0) based on the calculated maximum dome temperature indicate that the previous inspection is conservative. Some scenarios with different magnitude of the flow velocity ejected from head cooling nozzles are calculated to determine the bounding case according to EDY and RIY value. The probability of cracking is small if there is enough coolant (>0.4%) ejected from head cooling nozzles. RKE, SKE, and k-ωSST model are compared in this paper. Results obtained from RKE model are similar for those from SKE model, but slightly different for those form KWSST model.

[1] U.S.NRC, "Davis-Besse Reactor Pressure Vessel Head Degradation: Overview, Lessons Learned, and NRC Actions Based on Lessons Learned", NUREG/BR-0353, Rev. 1, 2008
[2] W.R. Huang, “PWR System Introduction 1 ”, P.5-1~6-37, Taipower Maanshan Nuclear Power Plant, 1983
[3] F. Hofmann, F. Archambeau, C. Chaize, "Computational fluid dynamic analysis of a guide tube in a PWR", Nuclear Engineering and Design, Volume 200, Pages:117-126, 2000
[4] S.C. Chiang, B.S. Pei, F. Peter Tsai, " Pressurized water reactor (PWR) hot-leg streaming Part 1: Computational fluid dynamics (CFD) simulations ", Nuclear Engineering and Design, Volume 241, Pages:1768-1775, 2011
[5] Z. Karoutas, K. Lang, P. Joffre, E. Baglietto, R. Brewster, E. Volpenhein, " Evaluating PWR Fuel Performance Using Vessel CFD Analysis", Proceedings of 2010 LWR Fuel Performance, Orlando, Florida, USA.
[6] M.T. Kao, C.Y. Wu, C.C. Chieng, Y.B. Xu, K. Yuan, M. Dzodzo, M. Conner, S. Beltz, S. Ray, T. Bissett, “CFD Analysis of PWR Core Top and Reactor Vessel Upper Plenum Internals Subdomain Models”, Nuclear Engineering and Design Volume 241, Issue:10, Pages:4181-4193, 2011
[7] C.Y. Wu, Y.M. Ferng, C.C. Chieng, Z.C. Kang, “CFD Analysis for Full Vessel Upper Plenum in Maanshan Nuclear Power Plant”, Nuclear Engineering and Design, Volume 253, Pages 285-293, 2012
[8]Y.B. Xu, M. Conner, K. Yuan, M. B. Dzodzo, Z. Karoutas, S. A. Beltz, S. Ray, T. A. Bissett, C. C. Chieng, M.T. Kao, C. Y. Wu, " Study of impact of the AP1000® reactor vessel upper internals design on fuel performance", Nuclear Engineering and Design, Volume 252, Pages 128-134, 2012
[9] F. C. Zhang, S. G. Tan, X. H. Zheng, J Chen, "CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)", Applied Mechanics and Materials, Volume 444-445, pages:411-415, 2014
[10] K. Honda, H. Ohira, T. Mori, "An Investigation of Thermal-hydraulics Behavior of MONJU Reactor Upper Plenum under 40% Rated Steady state", The 10th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-10) ,NUTHOS10-1114, Okinawa, Japan, December 14-18, 2014
[11] A. Kodama, T. Takata, A. Yamaguchi, " Numerical Study on Structural Integrity of Inner Marrel Caused by Thermal Stratification in Upper Plenum of Monju ", The 10th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-10) ,NUTHOS10-1296, Okinawa, Japan, December 14-18, 2014
[12] Y. S. Tseng, ” Maanshan Nuclear Power Plant RPV Dome of Thermohydraulic Analysis Report”, August 9,2011
[13] J. Uribe, R. Howard, M. Rabbitt, ”Fluid-Solid coupling in advanced gas-cooled reactor thermohydraulics”, 7th International Symposium on Turbulence, Heat and Mass Transfer, Palermo, Sicily, Italy, September 24-27, 2012
[14] Solidworks Corp, Solidworks 2010, www.solidworks.com
[15] CD-adapco, STAR-CCM+ 6.04 UserGuide_6.04.pdf, www.cd-adapco.com
[16] T. H. Shih, W.W. Liou, A. Shabbir, Z. Yang, and J. Zhu, "A New k-εEddy Viscosity Model for High Reynolds Number Turbulent Flows -- Model Development and Validation". NASA TM 106721,1994
[17] W.P. Jones and B.E. Launder, "The Prediction of Laminarization with A Two-equation Model of Turbulence", International Journal of Heat and Mass Transfer, Volume15, pp. 301-314, 1972
[18] F. R. Menter, "Two-Equation Eddy-Viscosity Turbulence Models for Engineering Applications," AIAA Journal, Vol. 32, No. 8, August 1994, pp. 1598-1605.
[19] ASME, 2006. Cases of ASEM boiler and pressure vessel code. ASME N-729-1,March 28, 2006
[20] Atomic Energy Council ,"EOC-15 for Maashan Nuclear Power Plant", NRD-NPP-93-33, 2004
[21]D. Gandy,"Carbon Steel Handbook",Final Report, March 2007
[22] ASME Boiler and Pressure Vessel Code, 2001 Edition, Section II, “Materials,”Part D, “Properties”, American Society of Mechanical Engineers, 2001
[23] S. Ritter and H.P. Seifert, "Stress Corrosion Cracking of Low-Alloy Reactor Pressure Vessel Steels and of a Weld Filler Material under BWR Conditions", CH-5232 Villigen PSI,SWITZERLAND