FeCrAl包壳燃料棒辐照行为研究

铁素体FeCrAl不锈钢具有成为耐事故燃料包壳材料的潜在价值。通过FeCrAl包壳燃料棒堆内性能的初步分析,评估FeCrAl包壳的堆内性能,并对FeCrAl包壳后续的研发及应用提出建议。使用FUPAC程序对FeCrAl包壳燃料棒的堆内稳态辐照行为进行了初步研究。分析结果表明,FeCrAl包壳燃料棒的温度、内压、应力应变均低于设计限值。     

辐照蠕变对锆合金包壳管吸氢所致多场耦合行为的影响

本文考虑辐照效应,改进了锆合金包壳管内部的氢原子扩散-氢化物析出-热-力耦合行为的微分控制方程。根据多物理场等效积分弱形式和所建立的耦合计算方法,在FEPG软件平台编制文件,生成多场耦合计算的有限元程序,并对程序进行了验证。计算分析了辐照蠕变对锆合金包壳管堆内吸氢所致多场耦合行为演化的影响,结果表明:辐照蠕变导致包壳管内产生应力松弛,促使Mises应力显著降低,同时导致静水应力由负值转变为正值,进而影响氢原子的扩散;与不考虑辐照蠕变的结果进行对比,发现辐照蠕变会增大燃料芯块与包壳管局部接触区域的负的静水应力的绝对值及向外的静水应力梯度,导致接触区域内的氢原子浓度减小,接触区域周围的氢原子浓度增大。     

N36锆合金包壳辐照生长经验模型研究

利用N36锆合金包壳燃料棒堆内辐照考验的部分池边检查数据,计算了4个典型辐照生长经验模型对N36锆合金包壳的适用参数。计算结果表明,在典型辐照生长经验模型中,双曲正切经验模型最适合描述N36锆合金包壳辐照生长行为。在双曲正切经验模型基础上,建立了N36锆合金包壳辐照生长最佳估算模型和包络模型。通过添加工程因子,建立了不同加工工艺的N36锆合金包壳辐照生长经验模型。利用池边检查剩余数据对N36锆合金包壳辐照生长经验模型进行了验证,模型与数据吻合较好。     

不锈钢包壳管蠕变坍塌临界时间计算

压水堆燃料相关组件棒在堆内使用寿期中．不得发生蠕变坍塌以保证包壳结构完整性。这些棒通常使用不锈钢包壳。运用假想夹层分析理论,导出了一个较为简便的不锈钢包壳在通常工作环境下蠕变坍塌临界时间的计算公式。     

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。     

船用堆燃料棒包壳疲劳寿命分析

船用堆瞬态变工况下燃料棒包壳温度和冷却剂压力波动较大,引起包壳的疲劳损伤,因此包壳疲劳寿命分析至关重要。本文利用ANSYS软件模拟船用堆瞬态变工况下燃料棒的热机械行为,结合锆包壳疲劳寿命设计曲线,考察包壳温度、冷却剂压力、燃料棒内压以及辐照对船用堆燃料棒包壳疲劳寿命的影响。计算结果表明,瞬态变工况使得包壳疲劳寿命有很大降低;包壳温度变化与冷却剂压力变化相比,前者对包壳疲劳寿命的影响小;辐照会降低包壳疲劳寿命。在不影响核动力船舶机动性的前提下,可采取一些必要的措施来降低包壳的疲劳损伤。     

燃料棒包壳氧化层溶解失效准则研究

为确定严重事故条件下燃料棒包壳温度达到金属锆的熔点后包壳氧化层的失效时间、再定位熔融物的成分以及氧化层失效对堆芯熔化进程的影响,本文基于熔融锆同时溶解UO_2和ZrO_2动力学模型及燃料棒包壳水侧氧化层的受力分析建立了氧化层在熔融锆中溶解失效的准则。以FPT-0实验结果验证后发现该失效准则可以较准确地预测包壳氧化层的溶解失效。为增加该准则在严重事故计算程序中的适用性,在燃料棒设计结构一定的条件下,进一步将该准则量化为温度的函数,分析表明包壳氧化程度和燃料棒温度上升速率是影响包壳氧化层失效温度的主要因素。利用该失效准则可以同时获得包壳氧化层失效后再定位的熔融物的质量及成分含量。     

碳化硅复合材料包壳燃料棒在LOCA事故中的特性研究

碳化硅复合材料具有热膨胀系数低、中子吸收截面低、抵抗热冲击、高温下耐腐蚀、强度高等特点,是反应堆耐事故燃料的候选材料之一。为分析碳化硅复合材料包壳燃料棒在失水事故下的综合性能,利用公开文献实验数据,拟合碳化硅复合材料物性公式,用FRAPTRAN对比分析与Zr4包壳燃料棒在同种工况下的性能差异。本文整理的公式和使用的方法可初步用于分析碳化硅复合材料燃料棒的综合性能。计算结果表明:在失水事故中,碳化硅复合材料的燃料棒失效时间长,平均温度低,可有效延缓事故进程。     

核燃料芯块裂变气体释放行为模拟研究

基于ABAQUS对UO_2核燃料芯块裂变气体行为做了研究,对两个经典模型:ANS-5.4单阶段模型和两阶段Forsberg–Massih释放模型作了详细描述,重点研究了这两个模型在ABAQUS中的实现及模型与ABAQUS结合时的技术难点。将研究模拟结果与该模型在相应成熟软件中的结果作比较,吻合度很好,验证了基于ABAQUS二次开发方式模拟裂变气体行为方法的可行性。     

锆合金辐照蠕变和生长的宏介观模型研究进展

锆合金作为核反应堆堆芯的主要结构材料之一,在服役过程中会发生辐照蠕变和生长行为,严重影响其使用可靠性。预测锆合金的辐照蠕变和生长是保障反应堆安全运行的关键。本文聚焦于两类锆合金构件,包括压水堆用锆合金包壳管及重水堆用Zr-2.5Nb压力管,分别从宏介观尺度详细综述了其辐照变形预测模型。针对适用于包壳管的宏观经验模型及介观力学模型,分别描述了两类模型的特征,重点介绍了介观力学模型的研究现状及最新进展,并对其未来的发展方向提出建议;针对适用于压力管的宏观经验模型,包括加拿大CANDU压力管辐照变形计算方程(简称C6方程)及秦山重水堆压力管辐照变形计算方程(简称秦山方程),分别描述了两类方程的具体形式,并介绍了C6方程国产化的进展。     

Mechanical Properties of Neutron Irradiated Fuel Cladding Tubes

There was few post irradiation examination data on the mechanical properties of domestic fuel cladding tubes used for light water reactors, then those data obtained abroad have been often used in the fuel design or fuel performance codes. Although, many reports discussed the deformation mechanism of the tube, almost all the data were not obtained from irradiated specimens but unirradiated ones. In recent years, systematic post irradiation examinations on domestic fuel elements used in Japanese light water reactors and the related studies were performed.

This report first summarizes briefly the crystallographic texture which characterizes the properties of Zircaloy fuel cladding tubes, followed by an explanation of basic properties such as elasticity, plasticity, creep and fatigue. Finally, the up-to-date results are introduced.     

反应堆用SiC陶瓷基复合包壳材料研究进展

核燃料元件的包壳材料是反应堆安全的重要屏障。随着核动力反应堆向高燃耗、长燃料循环寿命、高安全性趋势的发展,传统Zr合金包壳材料因其铀燃耗极限(62 MW·d/kg)、高温腐蚀、氢脆、蠕变、辐照生长、芯/壳反应等缺陷,已不能满足未来第四代核能系统燃料元件对包壳材料的苛刻要求。Si C因其更小的中子吸收截面、低衰变热、高熔点及优异的辐照尺寸稳定性等优点,以Si C为基体的陶瓷基复合材料成为新一代包壳材料研究的热点。结合Si C的晶体结构、热物理特性,对其在第四代核反应堆包壳材料中的设计思路、中子辐照效应、热一力性能、与UO,的化学反应等进行了概述,对SiC基复合材料在未来核能领域的应用前景进行了展望。     

Deformation Behavior of Zircaloy-4 Cladding under Cyclic Pressurization

The deformation behavior of Zircaloy-4 cladding was investigated under cyclic pressurization. A cyclic pressurization test for a cladding tube was carried out and a stress-life diagram of the cladding tube was constructed. The result showed that the O'Donnell and Langer relationship can be fitted to the fatigue behavior of the Zircaloy-4 cladding under cyclic pressurization with the frequency of 1 Hz by assuming the fatigue limit of 344.7 MPa at 350°C. In terms of the effect of a loading frequency, the stress-life diagram deviated from the O'Donnell and Langer relationship when the applied frequency decreased to 0.1 Hz. Based on the change of the diametral differences and fractographic observations, it was found that a combined creep-fatigue interaction rather than pure fatigue had an influence on the failure of the Zircaloy-4 cladding under cyclic pressurization at 350°C.     

Development of the Ultra-microhardness Technique for Post Irradiation Examination of Fuel Cladding Tubes

Hardness measurements are potentially valuable for a quantitative discussion of embrittlement in the inner portions of fuel cladding tubes. The size of the indentation, however, is not negligible compared to the measuring region, even when a micro Vickers hardness tester is employed. This limits the measuring technique, and very little has been studied about degradation phenomena in the inner portion of the tubes.

A hardness measurement system, equipped with a depth-sensing indentation instrument, and the necessary post irradiation examination technique for specimens with high radioactivity were successfully developed and the following observations were obtained from the system's application example. The diffusion coefficient of oxygen obtained from the hardness of an unirradiated zirconium lined cladding with simulated oxidation in the fuel rod showed good agreement with literature data. The calculated diffusion coefficient from hardness in the inner portion of irradiated Zircaloy-2 fuel rods was almost the same value as that of unirradiated zirconium, which implied that neither neutron irradiation nor fission fragment bombardment enhanced the oxygen diffusion in the inner portion of cladding tube.