增材制造碳化硅核燃料包壳元件及其芯块热传导行为模拟研究

Study on Additive Manufacturing and Densification of Silicon Carbide for Nuclear Fuel Elements

本研究采用黏结剂喷射增材制造技术和化学气相渗透(Chemical vapor infiltration,CVI)相结合的方法制备了用于反应堆核燃料元件的碳化硅(Silicon carbide,SiC)样品。重点研究了CVI工艺对增材制造SiC生坯的致密化影响。实验结果表明,利用三氯甲基硅烷(Methyltrichlorosilane,MTS)对增材制造SiC生坯进行CVI处理,可获得较高纯度、较高密度的SiC样品。当CVI温度为1050℃时,MTS在生坯表面沉积,在表面形成坚硬的壳体,形成了外部密度高,内部密度小的不均匀分布;降低温度可以增加MTS的热解时间,延长MTS的迁移距离,使得MTS在样品内部分解。当CVI温度为995℃,MTS在SiC生坯内部分解并生长出SiC晶须,SiC晶须的生长受到气-液-固的过程控制,铁(Fe)作为催化剂,Fe可以和反应体系中的硅在较低的温度下形成低的共融液相,对SiC的生长具有促进作用。经过35 h CVI后,样品密度可以达到2.18 g/cm^(3),样品热导率为3.93 W/mK。在压水堆环境下,对SiC基弥散燃料元件进行了传热数值模拟计算。以CVI处理35 h的SiC样品作为基体,包覆燃料颗粒填充体积百分数为40%时,燃料元件的峰值温度为1200 K,低于无基体UO_(2)燃料芯块。说明CVI致密化后的SiC样品可以降低压水堆的燃料峰值温度,有利于燃料核热释放,可以避免燃料元件内形成局部热点,从而提高元件使用寿命。

in this study,silicon carbide(SiC)ceramics for nuclear reactors were prepared using a combination of binder jetting additive manufacturing technology and chemical vapor infiltration(CVI)process.The effect of CVI on the densification of SiC green bodies by additive manufacturing was studied.The results showed that the SiC samples with higher purity and density could be achieved by utilizing methyltrichlorosilane(MTS)for CVI in additive manufacturing process.At a CVI temperature of 1050℃,MTS deposition on the green body's surface formed a hard shell,resulting in uneven density distribution,with higher density externally and lower density internally.Lowering the temperature extended the thermal decomposition time of MTS,which prolonged the migration distance of MTS,and caused MTS to decompose inside the samples.At a CVI temperature of 995℃,MTS decomposed inside the SiC green body,fostering the growth of SiC whiskers through the gas-liquid-solid process.Iron(Fe)catalysts can form a low eutectic liquid phase with silicon in the reaction system at lower temperatures,promoting the growth of SiC.After 35 hours of CVI,the density of the samples reached 2.18 g/cm^(3),with a thermal conductivity of 3.93 W/mK.Numerical simulations of heat transfer on SiCbased dispersed fuel elements in pressurized water reactor environments were conducted.Utilizing SiC samples treated with CVI for 35 hours as the matrix and filling the cladding fuel particles with 40%volume percentage,the peak temperature of the fuel element was 1200 K.This was lower than that of the UO_(2)fuel pellets without a matrix.These results implied that SiC samples after CVI densification could reduce the peak fuel temperature of pressurized water reactors,facilitating nuclear heat release and preventing the formation of localized hotspots in the fuel element,thus prolonging its service life.