CAP1400熔融物堆内滞留试验验证研究
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摘要
通过反应堆压力容器外部冷却(ERVC)实现熔融物堆内滞留(IVR)技术是核电厂严重事故缓解的重要措施之一。针对CAP1400 IVR措施实施,开展了提高临界热通量关键因素(FIRM)试验研究,本论文详细介绍了验证试验的台架设计、主要技术参数和试验结果。本试验结果对于研究反应堆压力容器IVR-ERVC条件下的外部临界热流密度(CHF)特性具有重要学术意义,并对提高反应堆压力容器的安全性具有重要工程应用价值。
In-vessel retention(IVR) of molten corium through external reactor vessel cooling(ERVC) is a key severe accident management strategy in nuclear power plants. With respect to the implementation of the CAP1400 IVR strategy, one verification experiment about heat transfer characteristic of key factors of improving ERVC-CHF "was carried out. This paper discussed the experimental design, primary test parameters and the results of the facility in detail. The experimental results had important academic significance for studying external critical heat flux(CHF) characteristics of the IVR-ERVC conditions, as well as engineering application value for improving the safety of reactor pressure vessels.
In-vessel retention(IVR) of molten corium through external reactor vessel cooling(ERVC) is a key severe accident management strategy in nuclear power plants. With respect to the implementation of the CAP1400 IVR strategy, one verification experiment about heat transfer characteristic of key factors of improving ERVC-CHF "was carried out. This paper discussed the experimental design, primary test parameters and the results of the facility in detail. The experimental results had important academic significance for studying external critical heat flux(CHF) characteristics of the IVR-ERVC conditions, as well as engineering application value for improving the safety of reactor pressure vessels.
引文
[1] Tuomisto H, Thoefanous T G. A Consistent Approach toSevere Accident Management[J]. Nuclear Engineering and Design, 1994, 148:171,
[2] Theofanous T G, Liu C, Additon S, et al. In-vessel coolability and retention of a core melt[R]. Argonne National Lab., IL(United States); California Univ., Santa Barbara, CA(United States). Center for Risk Studies and Safety, 1996, 1.
[3] Kymalainen Q, Tuomisto H, Theofanous TG. In-Vessel Retention of Corium at the Loviisa Plant[J]. Nuclear Engineering and Design, 1997, 169:109-130.
[4] Theofanous T G, Syri S L. The coolability limits of a reactor pressure vessel lower head[J].Nuclear engineering and design, 1997, 169:59-76.
[5] Theofanous T G, Angelini S. Natural convection for invessel retention at prototypie Rayleigh numbers[J].Nuclear Engineering and Design, 2000, 200:1-9.
[6] Dinh T N, Tu J P, Salmassi T, et al. Limits of Coolability in the AP1000-Related ULPU-2400 Configuration V Facility[C]. The 10th International Topical Meeting On Nuclear Reactor Thermal Hydraulics, NURETH-10,Seoul, Korea, 2003.
[7] Rouge S SULTAN test facility for large-scale vessel coolability in natural convection at low pressure[J]. Nuclear Engineering and Design, 1997, 169:185-195.
[8] Cheung F B, Haddad K H. A hydrodynamic critical heat flux model for saturated pool boiling on a downward facing curved heating surface[J]. International Journal of Heat and Mass Transfer, 1997, 40:1291-1302.
[9] Cheung F B, Yang J, Dizon MB, et al. Scaling of downward facing boiling and steam venting in a hemispherical heated channel[J]. International Journal of Transport Phenomena, 2004, 6:81-96.
[10]陆维,胡腾,赵宇峰等.真实表面材料及其老化效应对反应堆压力容器ERVC-CHF影响的试验研究[J].原子能科学技术,2016,50(10):1782-1786.
[11]杨胜,胡腾,陆维,等.针对IVR-ERVC策略的朝下曲面化学水沸腾CHF特性试验[J].核动力工程,2016(6):23-27.
[12] Chang H J, Hu T, Lu Wei, et al. Experimental study on CHF using a full scale 2-D curved test section with additives and SA508 heater for IVR-ERVC strategy[J].Experimental Thermal and Fluid Science, 2017,84:1-9.
[13] Lee J, Jeong Y H, Chang S H. CHF enhancement in flow boiling system with TSP and boric acid solutions under atmospheric pressure[J]. Nuclear Engineering and Design, 2010, 240:3594-3600.
[2] Theofanous T G, Liu C, Additon S, et al. In-vessel coolability and retention of a core melt[R]. Argonne National Lab., IL(United States); California Univ., Santa Barbara, CA(United States). Center for Risk Studies and Safety, 1996, 1.
[3] Kymalainen Q, Tuomisto H, Theofanous TG. In-Vessel Retention of Corium at the Loviisa Plant[J]. Nuclear Engineering and Design, 1997, 169:109-130.
[4] Theofanous T G, Syri S L. The coolability limits of a reactor pressure vessel lower head[J].Nuclear engineering and design, 1997, 169:59-76.
[5] Theofanous T G, Angelini S. Natural convection for invessel retention at prototypie Rayleigh numbers[J].Nuclear Engineering and Design, 2000, 200:1-9.
[6] Dinh T N, Tu J P, Salmassi T, et al. Limits of Coolability in the AP1000-Related ULPU-2400 Configuration V Facility[C]. The 10th International Topical Meeting On Nuclear Reactor Thermal Hydraulics, NURETH-10,Seoul, Korea, 2003.
[7] Rouge S SULTAN test facility for large-scale vessel coolability in natural convection at low pressure[J]. Nuclear Engineering and Design, 1997, 169:185-195.
[8] Cheung F B, Haddad K H. A hydrodynamic critical heat flux model for saturated pool boiling on a downward facing curved heating surface[J]. International Journal of Heat and Mass Transfer, 1997, 40:1291-1302.
[9] Cheung F B, Yang J, Dizon MB, et al. Scaling of downward facing boiling and steam venting in a hemispherical heated channel[J]. International Journal of Transport Phenomena, 2004, 6:81-96.
[10]陆维,胡腾,赵宇峰等.真实表面材料及其老化效应对反应堆压力容器ERVC-CHF影响的试验研究[J].原子能科学技术,2016,50(10):1782-1786.
[11]杨胜,胡腾,陆维,等.针对IVR-ERVC策略的朝下曲面化学水沸腾CHF特性试验[J].核动力工程,2016(6):23-27.
[12] Chang H J, Hu T, Lu Wei, et al. Experimental study on CHF using a full scale 2-D curved test section with additives and SA508 heater for IVR-ERVC strategy[J].Experimental Thermal and Fluid Science, 2017,84:1-9.
[13] Lee J, Jeong Y H, Chang S H. CHF enhancement in flow boiling system with TSP and boric acid solutions under atmospheric pressure[J]. Nuclear Engineering and Design, 2010, 240:3594-3600.