化学气相沉积法制备ZrC涂层的热力学分析

ZrC涂层可能在新一代TRISO包覆颗粒上被用作阻挡裂变产物和承受主要载荷的关键层，是先进高温气冷堆燃料元件研究的一个重要方向。文章利用HSC-CHEMISTRY4.1软件分析化学气相沉积工艺参数对所制备的ZrC涂层的影响。分析结果表明，在载气中加入足够的氢气对制备单一ZrC涂层很有必要。ZrCl₄的转化率随着沉积温度的升高而增加，当温度过高时，其影响不明显；较佳的沉积温度范围为1400~1600℃。随着反应物浓度的增加，获得单一ZrC涂层对应的最低ZrCl₄与CH₄的摩尔分数比增加；反应物摩尔分数的最佳范围可选为：甲烷，1.0%～2.0%；ZrCl₄，为甲烷的1.5倍。     

ZrCl4蒸汽沉积法制备高温气冷堆包覆燃料颗粒ZrC涂层

采用ZrCl₄蒸汽、H₂和C₃H₆作为化学反应体系,以Ar为载气,在流化床沉积炉中制备高温气冷堆包覆燃料颗粒ZrC涂层。对所制备涂层进行了分析表征,结果显示：ZrC涂层剖面均匀光滑,无明显孔洞;与内致密热解炭层的界面清晰,厚度约为35μm;涂层主要成分为Zr和C,Zr/C摩尔比接近化学计量比1∶1,其主要相组成为面心立方的ZrC;晶粒生长无明显的择优取向。     

塑料微球厚有机涂层制备研究

采用低压等离子体化学气相沉积(LPP-CVD)并结合反弹盘方法在靶球表面涂敷厚的CxH1-x涂层.使用反式-2-丁烯为工作气体的沉积速率相对较慢,最高为1 μm/h,而使用苯乙烯时,沉积速率提高到3～4 μm/h.结合反弹盘技术,在塑料微球上涂敷了厚度为50～80 μm的CxH1-x涂层,涂敷薄膜的表面均方根粗糙度小于50 nm.     

高温气冷堆包覆燃料颗粒SiC涂层的制备与表征

采用化学气相沉积法在包覆燃料颗粒上制得SiC涂层.对SiC涂层的功能、沉积设备、沉积工艺及组织结构进行了较系统的描述.研究表明制得的SiC涂层表面光滑、致密、无明显孔洞;与内、外致密热解碳层的界面清晰,无明显扩散渗透现象.对SiC涂层的组分及晶体结构分析表明涂层主要元素为Si和C,且Si/C的摩尔比接近1∶1,反应生成化学计量比的β-SiC.价键结构分析则表明涂层中主要连接键为Si-C共价键.     

锆合金包壳水侧SiC涂层研究

研究了聚碳硅烷（PCS）粉末的高温裂解特性及PCS粉末与锆粉间的化学反应机理,并在900 ℃制备了SiC涂层。研究发现,900 ℃开始,PCS裂解产物由无定形态SiC向结晶态转变。不同温度下,PCS粉末与锆粉的混合物发生一系列化学反应,产物为ZrC、Zr₂Si、Si₃Zr₅,通过调节反应温度,可控制该化学反应的程度,进而实现对涂层成分的调节。采用先驱体转化法（PIP）在锆合金包壳表面制备了SiC涂层,经PCS溶液浸涂-裂解3次循环可得到SiC陶瓷层,厚度为4 μm,涂层成分为SiC,ZrC为过渡层。划痕法测试得到涂层附着力等级为1～2级。     

聚变堆第一壁涂层材料TiC和TiN的残余应力研究

采用Ｘ衍射法测定了聚变堆第一壁涂层材料ＴｉＣ和ＴｉＮ薄膜的残余应力．对涂层材料不同的制备方法（化学气相沉积ＣＶＤ和物理气相沉积ＰＶＤ）、基体材料（Ｍｏ、石墨和３１６ＬＳＳ）、涂层厚度及沉积温度对残余应力的影响进行了研究．结果表明，ＣＶＤ与ＰＶＤ制备的涂层的残余应力均为压应力，且ＣＶＤ较ＰＶＤ产生的残余应力要低，Ｔｉ／Ｍｏ（ＣＶＤ）随涂层厚度（１４μｍ～６０μｍ）的增加，残余应力增加．ＰＶＤ涂层的残余应力主要为本征应力，高达数ＧＰａ．其值随沉积温度（２００℃～６５０℃）的升高而降低．对残余应力产生的原因作了初步讨论     

PACVD的金属陶瓷的制备,特征和磨损性能

用脉冲ｄｃ辉光放电通过等离子辅助化学气相沉积工业用金属陶瓷刀片镀上ＴｉＮｘ，ＴｉＣｘＮｙ和Ｎｘ。在７７３－９７３Ｋ的沉积温度范围内，研究了涂层参数对沉积率，涂层成分，涂层－基底分界面，涂层结构及微硬度的影响。采用最佳过程参数，获得了具有低氧和氯杂质，高微硬度和良好粘附强度的涂层。     

塑料微球厚有机涂层制备研究

采用低压等离子体化学气相沉积（LPP－CVD）并结合反弹盘方法在靶球表面涂敷厚的CxH1-xI涂层。使用反式－2－丁烯为工作气体的沉积速率相对较慢,最高为1μm/h,而使用苯乙烯时,沉积速率提高到3－4μm/h。结合反弹盘技术,在塑料微球上涂敷了厚度为50－80μm的CxH1-x涂层,涂敷薄膜的表面均方根粗糙度小于50nm。     

射频磁控溅射法制备氧化铝涂层绝缘性能及吸氢特性

氧化铝具有优良的绝缘和阻氚性能,是ITER候选功能材料之一。本工作采用射频磁控溅射法在中国低活化马氏体（CLAM）钢基底上制备了氧化铝涂层。分别采用掠入射X射线衍射、Raman激光光谱和原子力显微镜对氧化铝涂层的结构和表面形貌进行了表征;测量了氧化铝涂层体电阻率;研究了氧化铝涂层样品的吸氢特性。结果表明：氧氩比为0.1和0.5下制备的氧化铝涂层为非晶结构,氧氩比为0.4下制备的涂层中出现了结晶程度较差的氧化铝δ相结构;氧氩比为0.1和0.4下制备的涂层粗糙度和粒径均小于氧氩比为0.5下制备的涂层;不同氧氩比下制备的氧化铝涂层体电阻率均超过2.7×10¹⁴Ω•cm,氧氩比为0.4下制备的涂层电阻率最高,达到2.1×10¹⁵Ω•cm;氧氩比为0.5下制备的涂层样品具有最低的吸氢量。氧氩比对涂层的电绝缘特性和吸氢特性有显著影响。     

锆合金包壳水侧SiC涂层研究

研究了聚碳硅烷(PCS)粉末的高温裂解特性及PCS粉末与锆粉间的化学反应机理,并在900℃制备了SiC涂层。研究发现,900℃开始,PCS裂解产物由无定形态SiC向结晶态转变。不同温度下,PCS粉末与锆粉的混合物发生一系列化学反应,产物为ZrC、Zr_2Si、Si_3Zr_5,通过调节反应温度,可控制该化学反应的程度,进而实现对涂层成分的调节。采用先驱体转化法(PIP)在锆合金包壳表面制备了SiC涂层,经PCS溶液浸涂-裂解3次循环可得到SiC陶瓷层,厚度为4μm,涂层成分为SiC,ZrC为过渡层。划痕法测试得到涂层附着力等级为1～2级。     

Performance modeling of Deep Burn TRISO fuel using ZrC as a load-bearing layer and an oxygen getter

The effects of design choices for the TRISO particle fuel were explored in order to determine their contribution to attaining high-burnup in Deep Burn modular helium reactor fuels containing transuranics from light water reactor spent fuel. The new design features were: (1) ZrC coating substituted for the SiC, allowing the fuel to survive higher accident temperatures; (2) pyrocarbon/SiC “alloy” substituted for the inner pyrocarbon coating to reduce layer failure and (3) pyrocarbon seal coat and thin ZrC oxygen getter coating on the kernel to eliminate CO. Fuel performance was evaluated using General Atomics Company’s PISA code. The only acceptable design has a 200-μm kernel diameter coupled with at least 150-μm thick, 50% porosity buffer, a 15-μm ZrC getter over a 10-μm pyrocarbon seal coat on the kernel, an alloy inner pyrocarbon, and ZrC substituted for SiC. The code predicted that during a 1600 °C postulated accident at 70% FIMA, the ZrC failure probability is <10^?4.     

Fabrication of uniform ZrC coating layer for the coated fuel particle of the very high temperature reactor

Fuel for the very high temperature reactor is required to be used under severer irradiation conditions and higher operational reactor temperatures than those of present high temperature gas cooled reactors. Japan Atomic Energy Agency has developed zirconium carbide (ZrC)-coated fuel particles previously in laboratory scale which are expected to maintain their integrity at higher temperatures and burnup conditions than conventional silicon carbide-coated fuel particles. As one of the important R&D items, ZrC coating process development has been started in the year 2004 to determine the coating conditions to fabricate uniform structure of ZrC layers by using a new large-scale coater up to 0.2 kg batch. It was thought that excess carbon formed in the ZrC layer under the oscillation of coating temperature would cause non-uniformity of the ZrC layer. Finally, uniform ZrC coating layer has been fabricated successfully by adjusting the time constant of the coater and keeping the coating temperature at around 1400 °C.     

Cesium solubility,diffusion and permeation in zirconium carbide

To compare the relative effectiveness of ZrC vis-a-vis SiC as a fission product barrier in fuel structures for high temperature gas cooled reactor (HTGR) applications, a series of cesium infusion experiments on various ZrC powders, and ZrC coated graphite structures was performed to study the cesium solubility, diffusivity, and permeability of this coating material. The ZrC powder results yield a solubility of Cs in ZrC, S(ppm wt) = (1.7 × 10^?6) exp[229 kJ/RT], over the temperature range 1485–1896 K. The diffusion coefficient of Cs in ZrC is 10^?18–10^?16 m²/s over a similar temperature interval. The activation energy of diffusion is estimated to be ≈ 50 kJ/mole.The results of experiments in which both SiC and ZrC coated graphite samples were exposed to cesium are more difficult to interpret. The results support the conclusions of the ZrC powder experiments that ZrC is comparable to SiC as a diffusion barrier to cesium.     

Microstructure and Mechanical Properties of Nano-Size Zirconium Carbide Dispersion Strengthened Tungsten Alloys Fabricated by Spark Plasma Sintering Method

W-(0.2,0.5,1.0)wt%ZrC alloys with a relative density above 97.5%were fabricated through the spark plasma sintering(SPS) method.The grain size of W-1.0wt%ZrC is about2.7 μm,smaller than that of pure W and W-(0.2,0.5)wt%ZrC.The results indicated that the W-ZrC alloys exhibit higher hardness at room temperature,higher tensile strength at high temperature,and a lower ductile to brittle transition temperature(DBTT) than pure W.The tensile strength and total elongation of W-0.5wt%ZrC alloy at 700 ℃ is 535 MPa and 24.8%,which are respectively 59%and 114%higher than those of pure W(337 MPa,11.6%).The DBTT of W-(0.2,0.5,1.0)wt%ZrC materials is in the range of 500 ℃-600 ℃,which is about 100 ℃ lower than that of pure W.Based on microstructure analysis,the improved mechanical properties of the W-ZrC alloys were suggested to originate from the enhanced grain boundary cohesion by ZrC capturing the impurity oxygen in tungsten and nano-size ZrC dispersion strengthening.     

Microstructure and mechanical properties of proton irradiated zirconium carbide

Zirconium carbide is a candidate ceramic being considered for metal-carbide-base composite-type fuels, as well as for an alternative coating material for TRISO particle fuels. Ensuring adequate mechanical properties and dimensional stability in response to radiation is a key part in developing a practical ZrC-base fuel. The existing available radiation response data for ZrC is limited and insufficient. In the present study, ZrC was irradiated with a 2.6 MeV proton beam at 800 °C to doses of 0.7 and 1.5 dpa. Following radiation, the radiation induced damage microstructure is comprised of a high density of nanometer-sized Frank loops, but no irradiation induced amorphization, voids, or precipitates were observed. A slight lattice expansion was found in the irradiated ZrC, in good agreement with the reported results from neutron irradiation. The changes in microhardness and fracture toughness properties induced in the irradiated samples were measured using indentation techniques. The hardness and the fracture toughness both increase with increasing radiation dose.     

A new method for coating microspheres with zirconium carbide and zirconium carbide-carbon graded coats

A new method for the chemical vapor deposition of ZrC and C-ZrC alloys has been developed. This process has been applied to the fabrication of coated particle nuclear fuels of the type used in the large High Temperature Gas-Cooled Reactor. A powder feeder is used to supply ZrCl₄ to the fluidized bed coating furnace where it undergoes reaction with a hydrocarbon to form ZrC. Quantitative metering of the ZrCl₄ makes it possible to control the deposition of the ZrC and the codeposition of the C-ZrC alloys on the fuel particles. Examples of both types of coats made using the described technique are discussed and illustrated.     

A study of the reactions between ZrC and some metal oxides

High temperature chemical reactions of ZrC powder or ZrC coating on alumina spheres with CeO₂, UO₂ and SrO were studied. CeO₂ reacts with ZrC at 1473–2073 K to give Ce₂O₃ and Ce₂Zr₂O₇. Virtually no reactions were observed in UO₂-ZrC mixtures at 1673–1973 K. SrO reacts with ZrC above 1273 K to form SrZrO₃, Sr vapor and, presumably, CO. The SrZrO₃ formation on ZrC coated spheres contained in a bed of SrO powder was rapid at 1673 K. All particles suffered serious damages on the ZrC coatings by the reaction with SrO. Though the Sr concentration in the ZrC was below the detection limit, Sr was found to be distributed within the alumina kernel.