基体为含Si铝合金的U-Mo弥散燃料辐照后的最新显微结构分析结果

RERTRU-Mo弥散燃料板正在研制中，准备将其应用于全世界的研究堆中。特别值得关注的问题是基体为含Si铝合金的U-Mo弥散燃料的辐照性能。添加Si是为了提高U-Mo弥散燃料的性能。采用光学金相和扫描电子显微镜，对基体为Al-0．2Si和4043Al合金（～4．8％Si）的U-Mo燃料板在制造条件和辐照条件下的显微组织进行了分析。两种基体的燃料板在制造中都在U-7Mo颗粒周围产生了富Si的反应层；在辐照中观察到这些反应层厚度增大，燃料板的某些区域出现Si贫化。对于4043Al基体的燃料板，只有在非常恶劣的辐照状态下才会在暴露的燃料板区域中观察到这些现象。     

弥散燃料板芯体中U-Mo/Al-Si合金基体反应层性质研究

针对弥散型燃料板采用实验方法分析U-Mo燃料相与Al-Si基体反应层的性质。实验结果表明:反应层主要出现在U-Mo燃料颗粒的内部微裂纹处及燃料颗粒与基体界面处,其形貌和厚度均不规则。U-Mo与Al-Si遵循空位扩散机制,扩散过程主要为Al、Si向U-Mo合金的扩散。在反应层中Al含量基本维持不变,Si含量沿基体-燃料相方向递增,并聚集在U-Mo侧的反应层中。当基体中Si含量达到5%时,可明显抑制扩散反应的进行,从而改进燃料板性能。     

低富集U-Mo弥散型燃料小板RERTR-6辐照试验结果——Si对U-Mo／Al弥散燃料辐照行为影响的初步分析

辐照时Al基体中弥散的U-Mo燃料颗粒表面形成包覆的反应物。在一些辐照试验中,反应物和舢基体交界面处有气孔产生。受辐照条件的影响,气孔可能会长大并彼此连接形成连通的大气孔,严重时形成连续网络结构使燃料板出现不可接受的枕形肿胀。在美国和其他国家的辐照实验中都观察到了这种现象。冶金学和热动力学分析表明,Al基体中加入Si以及U-Mo燃料中添加Zr或Ti都可以提高U-Mo／Al弥散型燃料反应产物的稳定性。本文介绍了添加Si辐照试验的初步结果,即将适量的Si添加到Al基体中能有效降低U-Mo／Al燃料扩散反应和消除气孔形成。     

阿贡国家实验室单片式燃料研制

最初的U-Mo弥散型燃料元件已经最示出铀装载量方面的优越性，可以用它来成功地把部分研究堆燃料转换成低浓铀。另外，辐照实验显示出在高温下U-Mo弥散型燃料元件的芯体与铝基发生了相互反应。解决这些现象的潜在方法是使用燃料合金片代替铝基弥散型燃料。这种单片式燃料提供了比弥散型燃料更低的燃料一基体反应层和更高的铀密度。由于弥散型燃料基体的不足，单片式燃料元件生产需要研究新的制造方法。经过阿贡国家实验室研究人员的努力，已经得到一种可行的单片式燃料板生产方法，本文将介绍这种方法。     

RERTR燃料研制项目的进展

始于20世纪90年代末期的U-Mo合金弥散燃料辐照试验确定了这些燃料令人满意的辐照行为。但美国国内外随后的实验暴露了这种燃料在高温高功率下的缺点。详细的辐照后检验表明，燃料性能问题不是因为U-Mo燃料颗粒性能差，而是由辐照中燃料和Al基体反应形成的反应层的肿胀行为引起的。 极高密度低富集度燃料的继续开发需要一份详尽的计划，包括燃料制造研究、堆外性能鉴定、辐照试验、辐照后检验、燃料性能评估和模型建立。一些潜在的补救措施对锵决公认的燃料性能问题是有效的；补救措施包括燃料和基体化学成分的微小改变。用另外一种材料代替Al基体，或者完全不要基体。所有的这些变动目前都作为六国（阿根廷、加拿大、法国、韩国、俄罗斯和美国）参与的燃料开发合作的一个部分在调查研究当中。本文将回顾目前RERTR-6辐照试验及其支撑实验的试验结果，并且讨论了到2010年底低富集度高密度燃料合格性鉴定的前进途径。     

国外U3Si2与UMo燃料的发展现状

U3Si2-Al弥散型燃料是一种成功的低浓铀燃料，但在较高温度和较深燃耗运行时，其抗辐照性能急剧下降；UMo-Al弥散型燃料可能使任何高性能研究堆改用低浓铀，可是燃料相与铝基体的广泛反应引起严重的肿胀，期待含硅的铝基体能成功阻止这种反应的发生；单片型UMo合金燃料板具有较好的抗辐照性能，但制造方法尚不成熟。所有这些问题都亟待解决。本文首先简介了研究堆低浓铀燃料的发展简史，分析了U3Si2-Al弥散型燃料的成就与不足，讨论了UMo合金燃料所遇到的问题与需要解决的途径，提出了U3Si2-Al、UMo-Al弥散型燃料和单片型UMo合金燃料板的研究现状。     

中国核动力研究设计院研究堆燃料元件的发展现状

文章介绍中国核动力研究设计院(NPIC)紧跟国际原子能机构(IAEA)确认的防止核扩散、降低研究和试验反应堆用燃料富集度研究计划(RERTR)的进展,在研究堆低浓铀燃料元件开发研究方面进行的一系列工作,描述了NPIC的U3Si2-Al燃料元件研究及生产现状和在新开发的UMo合金燃料研究方面的最新进展.     

U_3Si_2-Al弥散型燃料单板质量微观分析述评

利用光学显微镜、扫描电镜、X射线衍射分析、X射线能谱和图像分析仪等方法、手段,综合分析了U_3Si_2-Al弥散板型燃料板和主要质量问题,即燃料相的微观组织结构、燃料相和Al基体的相容性,燃料相的分析情况、狗骨形态、单板包壳厚度和芯体与包覆材料的结合质量等,以及它们对产品质量的影响。分析结果表明,U-7.7％Si铸态合金中,主要是     

单片式和弥散型U-Mo燃料研究进展及前景

本文介绍了单片式和弥散型U-Mo燃料的研究进展，以便实现用其他替代品来进行高浓铀向低浓铀转换（HEU→LEU）的可能性。在RERTR（研究实验堆降低铀浓缩度计划）-7A辐照试验中，以Zr-4合金为包壳的20％铀富集度U-7Mo单片式燃料板的辐照后初步检验结果表明其性能优良。正在进行中的弥散型燃料的加工性能和性能预测试验具有以下特征：①用锗替代硅与基体铝形成合金；②对U-Mo燃料颗粒涂硅、锗、镍涂层，基体铝与这些元素合金化，从而减少界面反应的动力；③在基体中加入多孔氧化铝，用来吸收裂变气体产物。     

铀钼合金燃料合格性鉴定试验的现状及发展前景

美国、法国、阿根廷和俄罗斯正就铀钼（U-Mo）合金燃料的开发、合格性鉴定和商业应用的许可证问题进行国际合作。进行合格性鉴定试验的样品分别是小板、全尺寸板和组件。美国、法国和阿根廷采用的是轧制工艺，铀密度为6～8g·cm^-3；俄罗斯采用的是挤压工艺，挤压工艺对于提高铀密度比较困难。铀密度为5～6g·c^-3。根据合格性鉴定计划，每个合作国家都根据本国的辐照试验给出了综合性的辐照结果。结果表明，铀钼合金燃料在低燃耗时具有很好的辐照性能，但最近发现，在高温高能耗辐照条件下，铀钼合金弥散型燃料出现了严重的肿胀。即所谓的枕头效应，这主要是由于铀燃料与铝基体发生了过度反应引起的。解决办法有两个，一个是在铝基体材料中添加硅，以减少U～Mo与Al基的反应；另一个办法是采用U-Mo片状合金燃料。     

Microstructural characterization of U-7Mo/Al-Si alloy matrix dispersion fuel plates fabricated at 500 °C

The starting microstructure of a dispersion fuel plate will impact the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of U-Mo dispersion fuel plates, particularly the interaction layers that can form between the fuel particles and the matrix, scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses have been performed on samples from depleted U-7Mo (U-7Mo) dispersion fuel plates with either Al-2 wt.% Si(Al-2Si) or AA4043 alloy matrix. It was observed that in the thick interaction layers, U(Al, Si)3 and U6Mo4Al43 were present, and in the thin interaction layers, (U, Mo) (Al, Si)3, U(Al, Si)4, U3Si3Al2, U3Si5, and possibly USi-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this investigation, the time that a dispersion fuel plate is exposed to a relatively high temperature during fabrication will impact the nature of the interaction layers around the fuel particles. Uniformly thin, Si-rich layers will develop around the U-7Mo particles for shorter exposure times, and thicker, Si-depleted layers will develop for the longer exposure times.     

CARR用U-Mo合金燃料的堆芯物理方案研究

U-Mo合金燃料具有铀密度高、辐照稳定性好和后处理简单等优点,是未来研究堆燃料的理想选择。在保持中国先进研究堆(CARR)主体结构不变的基础上,使用合适的U-Mo合金燃料替换CARR现有燃料,进行堆芯方案初步研究。通过对中子注量率、循环长度等关键参数的对比分析,给出了较优的堆芯物理设计方案。该堆芯物理方案具有更好的设计参数,并可节省大量的燃料经费支出,提高了反应堆运营的经济性。     

Properties of DU-10 wt% Mo alloys subjected to various post-rolling heat treatments

Studies were completed to obtain mechanical properties of depleted uranium-molybdenum (U-Mo) alloys subjected to different post-processing treatments using microhardness, quasi-static tensile tests, and scanning electron microscopy failure analysis. U-Mo alloy foils are currently under investigation for potential fuel conversion of high power research reactors to low enriched uranium fuel. Although mechanical properties take on a secondary effect during irradiation, an understanding of the alloy behavior during fabrication and the effects of irradiation on the integrity of the fuel are essential. In general, the microhardness was insensitive to annealing temperature but decreased with annealing duration. Yield strength, Young's modulus, and ultimate tensile strength were affected in varying manners with both increasing annealing temperature and duration, and subjecting the alloy to rolling. The failure mode was insensitive to annealing conditions, but was significantly controlled by the impurity concentration of the alloy, especially carbon. Values obtained from literature are also provided with reasonable agreement based on extrapolation of annealing duration, even though processing conditions and applications were quite different in some instances.     

Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix

In the framework of the IRIS-TUM irradiation program, several full size, flat dispersion fuel plates containing ground U(Mo) fuel kernels in an aluminum matrix, with and without addition of silicon (2.1 wt.%), have been irradiated in the OSIRIS reactor. The highest irradiated fuel plate (with an Al-Si matrix) reached a local maximum burnup of 88.3% ²³⁵U LEU-equivalent and showed a maximum thickness increase of 323 μm (66%) but remained intact. This paper reports the post irradiation examination results obtained on four IRIS-TUM plates. The evolution of the fission gas behavior in this fuel type from homogeneously dispersed nanobubbles to the eventual formation of large but apparently stable fission gas bubbles at the interface of the interaction layer and the fuel kernel is illustrated. It is also shown that the observed moderate, but positive effect of Si as inhibitor for the U(Mo)-Al interaction is related to the dispersion of this element in the interaction layer, although its concentration is very inhomogeneous and appears to be too low to fully inhibit interaction layer growth.     

Microstructural development in irradiated U-7Mo/6061 Al alloy matrix dispersion fuel

A U-7Mo alloy/6061 Al alloy matrix mini-dispersion fuel plate was irradiated in the Advanced Test Reactor and then examined using optical metallography and scanning electron microscopy to characterize the developed microstructure. Results were compared to the microstructure of the as-fabricated dispersion fuel to identify changes that occurred during irradiation. The layer that formed on the surface of the fuel U-7Mo particles during fuel plate fabrication exhibits stable irradiation performance as a result of the ∼0.88 wt% Si present in the fuel meat matrix. During irradiation, the pre-formed interaction layer changed very little in thickness and composition. The overall irradiation performance of the fuel plate to moderate power and burnup was considered excellent.     

粉末冶金法制备多孔U-10%Mo合金及其微观组织结构分析

U-Mo合金是近期广受关注的金属型核燃料材料之一,它具有良好的抗辐照肿胀能力。分析认为,若能在U-Mo合金中加入一定的孔隙,可起到容纳裂变气体以进一步提高其抗辐照肿胀性能的作用。本文利用冷等静压 真空固相烧结的粉末冶金方法制备低密度U-10%Mo合金材料,探索了烧结工艺对产品密度的影响规律。实验得到了一系列不同孔隙度的U-Mo合金材料,利用金相显微镜（OM）、扫描电子显微镜（SEM）对其微观结构进行了表征。结果证明,样品在1 100 ℃下烧结时密度随烧结时间的延长而提高,因此可通过改变烧结时间控制其孔隙率。     

中国先进研究堆标准燃料组件堆外水力稳定性试验

中国先进研究堆（CARR）标准燃料组件由滚压在两块侧板上的21块燃料板组成。堆外水力试验的目的是考验在水力冲刷条件下燃料组件的结构稳定性。试验件是按照正式产品制造工艺制造的贫铀组件，试验平均流速为12m/s，是满功率运行流速的120％。先后试验了2个组件，第1个组件试验60d，是满功率运行时间的120％，试验后观察到固定下定位梳的销钉松动，下定位梳严重磨损了燃料板;工艺改进后制造的第2个组件试验120d，是满功率运行时间的240％，试验表明，第2个组件结构完整。试验中对组件结构稳定性和燃料板腐蚀性能，诸如组件的压差、燃料板振动、包壳表面腐蚀深度等进行了研究。     

Transmission electron microscopy characterization of irradiated U–7Mo/Al–2Si dispersion fuel

The plate-type dispersion fuels, with the atomized U(Mo) fuel particles dispersed in the Al or Al alloy matrix, are being developed for use in research and test reactors worldwide. It is found that the irradiation performance of a plate-type dispersion fuel depends on the radiation stability of the various phases in a fuel plate. Transmission electron microscopy was performed on a sample (peak fuel mid-plane temperature ～109 °C and fission density ～4.5 × 10²⁷ f m^?3) taken from an irradiated U–7Mo dispersion fuel plate with Al–2Si alloy matrix to investigate the role of Si addition in the matrix on the radiation stability of the phase(s) in the U–7Mo fuel/matrix interaction layer. A similar interaction layer that forms in irradiated U–7Mo dispersion fuels with pure Al matrix has been found to exhibit poor irradiation stability, likely as a result of poor fission gas retention. The interaction layer for both U–7Mo/Al–2Si and U–7Mo/Al fuels is observed to be amorphous. However, unlike the latter, the amorphous layer for the former was found to effectively retain fission gases in areas with high Si concentration. When the Si concentration becomes relatively low, the fission gas bubbles agglomerate into fewer large pores. Within the U–7Mo fuel particles, a bubble superlattice ordered as fcc structure and oriented parallel to the bcc metal lattice was observed where the average bubble size and the superlattice constant are 3.5 nm and 11.5 nm, respectively. The estimated fission gas inventory in the bubble superlattice correlates well with the fission density in the fuel.     

Development of very-high-density low-enriched-uranium fuels

Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm⁻³ U₃Si₂–Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8–9 g cm⁻³ in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work.     

Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

This paper describes experiences and present status of research and development works for the high temperature gas-cooled reactor (HTGR) fuel in Japan. Recently, Very High Temperature Reactor (VHTR) is evaluated highly worldwide, and is a principal candidate for the Generation IV reactor systems. In Japan, HTGR fuel fabrication technologies have been developed through the High Temperature Engineering Test Reactor (HTTR) project in Japan Atomic Energy Agency since 1960’s. In total about 2 tons of uranium of the HTTR fuel has been fabricated successfully and its excellent quality has been confirmed through the long-term high temperature operation. Based on the HTTR fuel technologies, SiC TRISO fuel has been newly developed for burnup extension targeted VHTR. For ZrC-TRISO coated fuel as an advanced fuel designs, R&Ds for fabrication and inspection have been carried out in JAEA. The irradiation with the Japanese uniform stoichiometric ZrC coating has been completed in the cooperation with Oak Ridge National Laboratory of the United States.