耐事故燃料用Cr涂层锆合金包壳研究进展

2011年日本福岛核事故后,燃料包壳表面涂层技术成为耐事故燃料研发的主要方向之一。国内外对此开展了大量的研究工作。经过10年多的技术探索,Cr涂层包壳从众多涂层方案中脱颖而出,已成为涂层包壳研发主要技术路线。目前国际上Cr涂层包壳技术已完成了制备工艺、性能评价及设计准则等研究工作,进入了由技术研发到工程应用的重要转型阶段。梳理国外的研发经验可为我国的Cr涂层研究提供参考。法国和美国在Cr涂层包壳研发中开展了大量的堆内外试验,在工程应用上取得了实质性的突破。因此,本文系统梳理了到目前为止法国和美国在Cr涂层研发方面主要研究内容、研究方法及其未来规划。     

中子辐照对耐事故燃料FeCrAl合金力学性能的影响研究

FeCrAl合金具有良好的抗高温氧化和力学性能,能够作为燃料包壳材料。为研究FeCrAl合金的辐照力学性能,开展了不同元素成分含量和2×1019 cm?2、8×1019 cm?2 2种中子注量辐照下的FeCrAl合金力学性能试验,并在室温和380℃下测试了FeCrAl合金的拉伸性能,获得了不同Cr和Al含量FeCrAl合金的抗拉强度和屈服强度,并研究了Al含量、Cr/Al含量配比及中子辐照对FeCrAl合金力学性能的影响。研究表明,FeCrAl合金强度随着Al含量增加大致呈增加趋势;经2×1019 cm?2中子辐照后,FeCrAl合金强度有较大提升;再经8×1019 cm?2中子辐照后,FeCrAl合金强度升高不明显。该研究结果为耐事故燃料（ATF）包壳材料的研发选型提供了重要的数据支撑。      

常温C4+辐照对多晶HIP-SiC力学性能的影响

热等静压(HIP)法具有提高材料致密度、抑制晶粒生长、避免晶粒取向等优点,常用于制备多晶6H-SiC材料。为探究HIP法制备的多晶6H-SiC辐照损伤特性,评价HIP工艺下碳化硅材料作为耐事故燃料包壳的可行性,以多晶6H-SiC为研究对象,分析样品辐照前后的性能、结构等变化。为防止其他元素对实验造成影响,实验采用的辐照离子为C⁴⁺,辐照剂量为1.8 dpa和5 dpa,并设置1组未辐照样品作对比。通过运用SEM、纳米压痕、XRD、拉曼光谱等测试分析多晶6H-SiC在离子辐照前后表面特征和性能参数的变化。研究结果表明,样品材料元素成分占比无明显变化,而硬度略有下降,晶格呈现损伤现象,但辐照很快趋于饱和且结构上未发生改变。因此从离子辐照方面分析,整体上HIP制备的多晶6H-SiC抗辐照能力较强,具有作为未来事故容错燃料基体的可能性。     

Zr－4合金的中子辐照生长

对由两厂分别生产的Ｚｒ－４包壳管样品在重水堆内进行中子辐照试验，辐照温度为６１０Ｋ，快中子注量为４．２×１０￣（２４）ｍ￣（－２）（Ｅ＞１．０ＭｅＶ）。试验结果表明，Ｚｒ－４管的辐照生长应变随辐照中子注量增加呈线性增加。两厂生产的Ｚｒ－４包壳管的生长应变可用Ｇ＝Ａ（φｔ）￣ｎ或Ｇ＝Ｂ＋Ｃ（φｔ）表达式描述，两者的差异可能是合金元素和杂质的综合影响所致。     

中子辐照对15MnTi钢力学性能影响

通过冲击试验、单向拉伸试验研究了中子辐照对15MnTi钢拉伸、冲击性能影响。中子辐照温度50℃、累计快中子注量1.5066×10~(18)cm~(-2),获得15MnTi钢母材、热影响区辐照前后冲击功及15MnTi钢室温、300℃下辐照前后拉伸曲线。试验结果表明,中子辐照导致15MnTi钢母材、热影响区韧脆转变温度上升,其中母材增幅较热影响区大,但冲击功上平台变化不大;15MnTi钢母材屈服强度、拉伸强度上升,其中室温屈服强度变化大,高温拉伸性能变化不明显。     

常压下ATF锆合金包壳Cr涂层表面饱和池式沸腾气泡行为实验研究

铬（Cr）涂层锆合金包壳是最有前途的耐事故燃料（ATF）的新型包覆材料之一,对其表面的气泡动力学进行研究有助于评估是否具有更好的传热性能。在常压下的Cr涂层锆合金包壳池式沸腾实验装置中对不同工艺方法下制备的Cr涂层锆合金包壳进行实验,研究了粗糙度等表面状态对气泡产生、长大以及脱离等气泡行为的影响。结果表明,气泡接触角与Cr涂层表面粗糙度有关,粗糙度越大,表面气泡接触角越小;不同涂层工艺下制备的4种Cr涂层锆合金包壳样件表面的气泡脱离直径范围为1.256～1.446 mm,气泡脱离频率范围为29.99～50.97 Hz;气泡脱离直径与粗糙度呈负相关,脱离频率与粗糙度呈正相关;气泡脱离直径预测模型与实验数据之间的偏差为±6%,脱离频率预测模型与实验数据之间的偏差为±3%。     

Cr涂层锆合金包壳模拟LOCA试验研究

2011年日本福岛核事故暴露传统锆合金燃料包壳在失水事故(LOCA)工况下的安全性问题。为了探究新型Cr涂层锆合金包壳在LOCA工况下的性能,本研究针对物理气相沉积(PVD)工艺涂覆的12～15μm厚度Cr涂层Zr-1Nb合金包壳管,开展模拟LOCA工况下的高温蒸汽氧化-淬火试验,氧化温度为1200℃和1300℃,单面氧化时间为10 min和20 min,淬火温度约800℃,之后对淬火后试样进行环压测试。结果发现,在研究条件下,Cr涂层未出现剥落,涂层完整;Cr涂层锆合金包壳外表面形成较为致密Cr₂O₃层,抑制O原子扩散至锆合金基体,阻止锆合金基体被氧化为ZrO₂层和α-Zr(O)层,环压测试发现淬火后包壳保持良好塑性。研究表明,在本测试工况下Cr涂层锆合金包壳相比传统锆合金包壳具有更强的抗LOCA事故能力。     

中子对氟聚合物力学性能的影响

对中子辐照前后氟聚合物 F2311和 F2314的静态力学性能、动态力学性能以及分子量进行了考察。结果表明,经注量为 1.5×1013/ cm2的中子辐照后,F2311和 F2314的静态力学性能稍有增强,分子量变化不大;F2311的动态力学性能基本不变,F2314的储存模量和损耗模量却有所降低,经注量为 2.5×1013/ cm2的中子辐照后,F2311的静态力学性能和 F2314的拉伸性能明显增强,F2314的压缩性能反而降低;F2311的动态力学性能基本不变,F2314的储存模量和损耗模量却明显减小。     

Cr涂层锆包壳池式沸腾传热实验研究

铬（Cr）涂层锆合金包壳被认为是最有前途的耐事故燃料（ATF）包壳材料之一,这种材料的表面状态对传热性能的影响程度将极大地影响着涂层锆包壳的工艺优化方向。本文在常压下的Cr涂层锆合金包壳池式沸腾实验装置中对不同工艺方法下制备的Cr涂层锆合金包壳进行实验,研究了粗糙度等表面状态对传热的影响规律及其机制。结果表明,表面粗糙度的提高能降低汽化核心产生的条件,在相同壁面过热度下可显著强化传热。在本文研究参数范围内,随着传热表面粗糙度的提高,临界热流密度（CHF）相应地呈上升趋势,增加表面粗糙度能有效提高CHF值。在此基础上,本文还建立了粗糙度对传热系数影响的预测关系式。     

Effect of Neutron Irradiation on Mechanical Properties of Niobium and Nb-Y Alloy

In a high-temperature gas-cooled reactor core, which consists of prismatic graphite fuel elements, leakage flows of coolant gas occur through gaps between blocks. Since the effects of these leakage flows on the total flow distribution are significant, their flow features must be clarified. In this paper, the leakage flows (crossflow through the interface gap between contacting fuel elements and the permeation flow through the fuel elements) in the normally stacked fuel elements were studied. In the basic experiments, leakage flow rates were measured using small-scale graphite blocks to determine the equivalent interface gap width and the permeability. The experiments using the full-scale fuel element were also carried out and the results agreed well with those of the basic experiments. Furthermore, a simple flow model was devised to predict the leakage flow in the fuel element.     

Effect of Neutron Irradiation on Mechanical Properties of Al-Li Alloys

An increase in yield stress at room temperature was observed in Al-0.6W/0 Li alloy irradiated to thermal neutron doses of 2.9 × 10¹⁹ to 7.2 × 10¹⁹ cm^?2. The hardening of as-irradiated specimens is accompanied with yield point followed by jerky yield-elongation in the stress-strain curve. The radiation hardening could not be annealed out by heating for 30 min at temperatures up to 350°C, whereas the yield-elongation disappeared gradually with increasing heating temperature in the l mm diam. specimens; with the 2 mm diam. specimens the yield-elongation still remained even after post-irradiation heating for 30 min at 350°C. Strengthening accompanied by jerky yield-elongation is considered to be due to He atom clusters precipitated along the dislocation. The hardening observed in the specimens heat-treated after irradiation at temperatures above 250°C is caused by randomly distributed gas bubbles.

In heavily cold-worked Al-0.6%W/o Li specimens, recovery of work hardening occurred during neutron irradiation to 4.2 × 10¹⁹ cm^?2. Hardening due to gas bubbles was also observed in the cold-worked specimens. In Al-2.7W/0 Li alloy, an increase in yield stress took place in the specimens irradiated to 4.2 × 10¹⁹ cm^?2 and heated for 30 min at temperatures of 155° to 260°C. The hardening is thought to be due to re-precipitation of β-phase resolved during the neutron irradiation.     

Effect of Neutron Irradiation on Deformation Behavior of Zirconium

The effect of neutron irradiation on the tensile deformation behavior of zirconium was examined at room temperature at various strain rates ranging of 2.2×10^?4~2.2× 10^?2 sec^?1. The microstructure of the deformed specimens was observed by transmission electron microscopy. It was established that neutron irradiation diminishes the uniform elongation and the strain hardening rate, and hastens the onset of plastic instability. These phenomena are attributed to inhomogeneous deformation in the dislocation channels in the irradiated and deformed zirconium.

From the relation between strain rate and tensile properties (yield stress, ultimate tensile stress, uniform elongation and strain hardening rate), it was established that in unirradiated zirconium deformation is controlled by slip at strain rates below 6×10^?3 sec^?1, while above this threshold, twinning as well as slip contribute to deformation.

Neutron irradiation markedly inhibits deformation twinning in zirconium at room temperature. At 77 K, on the other hand, deformation by twinning is more prominent in irradiated specimens. The mechanism of twinning inhibition due to neutron irradiation is discussed.     

Short-Time Mechanical Properties of EP823-SH Ferrite-Martensite Steel After High-Dose Neutron Irradiation

Atomic Energy - Irradiation of experimental fuel assemblies with uranium-plutonium nitride fuel and shells made of ferritic- martensitic steel EP823-Sh is conducted in order to ensure normal...     

新型锆合金包壳材料商业反应堆辐照考验方案设计

对于一种新型锆合金包壳材料,在商业反应堆中开展服役条件下的辐照考验是其研发必不可少的关键环节。相比于国际核电发达国家在锆合金包壳材料研发中积累的丰富商业堆辐照考验,国内自主锆合金仅开展了有限的商业堆先导辐照考验,且考验的燃耗水平偏低。本文将通过对国外锆合金辐照考验经验的总结及自身实践经验,给出锆合金包壳商业反应堆辐照考验时在方案设计上的一般方法和堆芯安全评估、风险应对上的基本考虑,为国内锆合金商业堆辐照考验研究提供参考。