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Mechanical failure of fresh nuclear grade iron–chromium–aluminum (FeCrAl) cladding under simulated hot zero power reactivity initiated accident conditions
Journal of Nuclear Materials ( IF 3.1 ) Pub Date : 2020-07-02 , DOI: 10.1016/j.jnucmat.2020.152352
Nicholas R. Brown , Benton E. Garrison , Richard R. Lowden , M. Nedim Cinbiz , Kory D. Linton

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens’ outer surfaces. Experiments were conducted at approximately 275 °C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8–3.4%, corresponding to quasi-static energy depositions of approximately 110–260 calories per gram UO2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the iron-chromium-aluminum (FeCrAl) cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.



中文翻译:

在模拟热零功率反应性引发的事故条件下,新核级铁-铬-铝(FeCrAl)熔覆层的机械故障

反应性引发的事故(RIA)是在轻水反应堆中由反应性的迅速增加而引发的假定事故,反应性的迅速增加导致裂变率和燃料温度的增加。RIA期间燃料系统故障的一种潜在模式是由于燃料颗粒的快速热膨胀而导致的颗粒包覆机械相互作用(PCMI)。通过使用液压改进的爆破测试系统对包壳试管样品快速加压,使样品在双轴加载路径下破裂,从而进行了模拟PCMI实验。使用远心镜头和高速摄像系统跟踪样品的变形和破裂。使用数字图像相关性(DIC)在标本外表面上绘制的斑点图案上计算外表面应变。实验是在大约275°C,热零功率RIA条件的代表。DIC计算出的失效箍应变约为1.8–3.4%,相当于准静态能量沉积约110–260卡路里/克UO2假设最初是颗粒-包层接触。这些条件与美国核监管委员会(NRC)草案法规指南1327中提议的未辐照锆基熔覆层的能量沉积限制(每克UO 2提议的150卡路里)非常相似。在数据中观察到非常小的应变速率依赖性,失效箍应变的大小随应变率的增加而略有降低。这种观察到的依赖性可能是相关的,因为由于中子光谱较硬且中子产生时间较短,因此铁铬铝(FeCrAl)熔覆层的应变率将比锆基熔覆层的应变率高约20%。

更新日期:2020-07-02
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