The early characterization of irradiation effects in stainless steels at the Experimental Breeder Reactor-II

The new Global Nuclear Energy Partnership (GNEP) program is revitalizing interest in fast spectrum reactors. With this revitalization comes the need to engineer new high-performance reactors using improved materials, especially materials that can perform adequately at higher temperatures than has previously been required. The previous fast reactor development programs in the United States developed a large knowledge base and a reasonable understanding of the irradiation effects on certain classes of materials, especially stainless steels. With the need for new, high-performance materials today, this established knowledge provides advantages in alloy design.     

Environmental degradation of materials during wet storage of spent nuclear fuels

Wet storage is the predominant mode of storage of spent nuclear fuels. Due to legislation and other constraints, many countries do not reprocess spent fuels and have to store these for extended periods in spent fuel storage pools (SFSPs). Although the water chemistry of the pool is benign, certain factors such as stagnancy of water, crevices, and galvanic contacts between various materials of the fuel clad and the lining of the pools can result in unexpected localized corrosion. In this study, the susceptibility to localized corrosion of aluminum-1S (Al-1S), Zircaloy-2, and type 304 stainless steel (SS) has been assessed using accelerated tests with crevice bent beam (CBB) assemblies. The pool water constituents have been analyzed and electrochemical potentials (ECPs) measured in water samples drawn from different locations of the pool. The ECP has also been measured in situ, in the pools. It has been demonstrated that under conditions of crevice and galvanic contact, aluminum clad fuels from research reactors are prone to localized corrosion even in the benign environments of a SFSP. The ECP experiments indicate the importance of surface condition of the material and irradiation on degradation of various materials due to corrosion.     

Heavy Liquid Metal Corrosion of Structural Materials in Advanced Nuclear Systems

Interest in advanced nuclear concepts using liquid metal coolant has increased in the past few years. Liquid metal coolants have been proposed for the next generation of small-sized nuclear reactors, which offer exceptional safety and reliability, sustainability, nonproliferation, and economic competitiveness. Heavy liquid metal coolants are investigated for advanced fast reactors that operate at high temperatures, reaching high efficiencies. Lead and lead-bismuth eutectic (LBE) coolants are also proposed as coolants and targets of accelerator driven systems. High temperature, corrosive environment, high fast neutron flux, high fluence, and radiation damage, among other physical phenomena, challenge the integrity of materials in these advanced systems. Excellent compatibility with the liquid coolant is recognized as a key factor in the selection of structural materials for advanced concepts. In this article, we review materials requirements for heavy metal cooled systems with emphasis on lead and LBE materials corrosion properties. We describe experimental corrosion tests currently ongoing at the Los Alamos National Laboratory (LANL) Development of Lead Alloy Technical Applications (DELTA) loop. DELTA is a facility designed to study the long-term corrosive effects of LBE on structural materials under relevant conditions of chemistry, flow, and temperature. The research studies will provide data of corrosion rates and corrosion mechanisms in selected steel exposed to high velocity (above 2 m/s) in flowing LBE at 500°C. Fundamental research studies will help support conceptual design efforts and further the development of heavy liquid metals technology.     

The Nuclear Renaissance: A challenge for the materials community

The “Nuclear Renaissance” is at hand and the materials community is integral to its success. Increased reliability of current nuclear components and the construction of the next generation of nuclear reactors depend on advancements in materials knowledge. This article introduces some of the ways materials will enable the future of nuclear power. Todd M. Osman is the technical director at TMS Dr. Osman can be reached at tosman@tms.org.     

乏燃料水池转运舱氦质谱检漏技术的应用

乏燃料水池转运舱是核电站乏燃料水池的组成部分,用于存放燃料组件倒料设备和水。监测乏燃料水池转运舱的水的泄漏率,通过气体和水的泄漏率转换,分析和计算出氦气检漏的泄漏率范围,并建立检测区域模型,优化检测范围和流程,通过范围查找和精确查找的结合,快速和准确地找到泄漏位置,以为乏燃料水池的查漏工作提供经验。     

Government nanotechnology funding: An international outlook

This article provides a global overview of the emerging fields of nanoscale science, engineering, and technology. Nanotechnology spending in different countries andits importance inindustrial revolution are highlighted. As the 21st century unfolds, nanotechnology’s impact on the health, wealth, and security of the world’s people is expected to be at least as significant as the combined influences in this century of antibiotics, integrated circuits, and other advanced materials.     

熔盐电堆的材料腐蚀

熔盐反应堆是一种以熔融盐为冷却剂和核燃料的反应堆。作为6种第四代核电站概念堆之一,熔盐堆正日益受到人们的关注。材料的腐蚀问题是熔盐堆发展过程中面临的一个技术挑战。为此,相关研究机构开展了大量研究,并取得了积极进展。本文对材料在熔融氟化物中的腐蚀机制、腐蚀行为以及抗蚀材料发展现状等方面的研究进行了综述。     

Radiation-Resistant Alloys of the Nickel-Chromium System

Properties of some alloys of the nickel-chromium system are studied with the aim of determining the possibility of their use as structural materials for nuclear reactors. It is shown that at some compositions such alloys form a structure ensuring high process and service properties under irradiation in boiling water reactors and pressurized water reactors. Commercial production of these alloys has begun. __________ Translated from Metallovedenie i Termicheskaya Obrabotka Metallov, No. 7, pp. 78 – 82, July, 2005.     

Pressurized water reactor fuel crud and corrosion modeling

Pressurized water reactors circulate high-temperature water that slowly corrodes Inconel and stainless steel system surfaces, and the nickel/iron based corrosion products deposit in regions of the fuel where sub-cooled nucleate boiling occurs. The deposited corrosion products, called ‘crud’, can have an adverse impact on fuel performance. Boron can concentrate within the crud in the boiling regions of the fuel leading to a phenomenon known as axial offset anomaly (AOA). In rare cases, fuel clad integrity can be compromised because of crud-induced localized corrosion (CILC) of the zirconium-based alloy. Westinghouse and the Electric Power Research Institute have committed to understanding the crud transport process and develop a risk assessment software tool called boron-induced offset anomaly (BOA) to avoid AOA and CILC. This paper reviews the history of the BOA model development and new efforts to develop a micro-scale model called MAMBA for use in the Consortium for Advanced Light Water Reactor Simulation (CASL) program.     

Advances in Metallic Nuclear Fuel

Metallic nuclear fuels have generated renewed interest for advanced liquid metal reactors (LMRs) due to their physical properties, ease of fabrication, irradiation behavior, and simple reprocessing. Irradiation performance for both steady-state and transient operations is excellent. Ongoing irradiation tests in Argonne-West’s Idaho-based Experimental Breeder Reactor II (EBR-II) have surpassed 100,000 MWd/T burnup and are on their way to a lifetime burnup of 150,000 MWd/T or greater. Metallic fuel also has a unique neutronic characteristic that enables benign reactor responses to loss-of-flow without scram and loss-of-heat-sink without scram accident conditions. This inherent safety potential of metallic fuel was demonstrated in EBR-II just one year ago. Safety tests performed in the reactor have also demonstrated that there is ample margin to fuel element cladding failure under transient overpower conditions. These metallic fuel attributes are key ingredients of the integral fast reactor (IFR) concept being developed at Argonne National Laboratory.     

Progress in the pyrochemical processing of spent nuclear fuels

Pyrochemical processes are being developed to recover actinides from spent fuels from light-water reactors and integral fast reactors. The transuranic elements from light-water reactors will be introduced into the integral fast reactor fuel cycle. To meet the requirements of that fuel cycle, transuranic elements are recovered as oxide free metal containing some fission products. This article discusses pyrochemical processes for recovering actinides from light-water reactor and integral fast reactor fuels and for treating the high-level wastes from these processes. The development status of these processes and the plans to demonstrate them using facilities at Experimental Breeder Reactor II are also described.     

Indirect Versus Direct Heating of Sheet Materials: Superplastic Forming and Diffusion Bonding Using Lasers

Many from within manufacturing industry consider superplastic forming (SPF) to be ‘high tech’, but it is often criticized as too complicated, expensive, slow and, in general, an unstable process when compared to other methods of manipulating sheet materials. Perhaps, the fundamental cause of this negative perception of SPF, and also of diffusion bonding (DB), is the fact that the current process of SPF/DB relies on indirect sources of heating to produce the conditions necessary for the material to be formed. Thus, heat is usually derived from the electrically heated platens of hydraulic presses, to a lesser extent from within furnaces and, sometimes, from heaters imbedded in ceramic moulds. Recent evaluations of these isothermal methods suggest they are slow, thermally inefficient and inappropriate for the process. In contrast, direct heating of only the material to be formed by modern, electrically efficient, lasers could transform SPF/DB into the first choice of designers in aerospace, automotive, marine, medical, architecture and leisure industries. Furthermore, ‘variable temperature’ direct heating which, in theory, is possible with a laser beam(s) may provide a means to control material thickness distribution, a goal of enormous importance as fuel efficient, lightweight structures for transportation systems are universally sought. This paper compares, and contrasts, the two systems and suggests how a change to laser heating might be achieved.     

Will the titanium golf club be tomorrow’s mashy niblick?

In the sporting goods industry, the application of game-improving advanced materials has resulted in titanium golfclubs attaining near-mythical status. These improvements have not come cheaply, however, and titanium clubs are losing ground to creative designs that employ combinations of lower cost materials. As a result, unless the cost of titanium can be lowered, the metal will see its share of the market become significantly reduced by the judicious use of lower-cost materials. For more information, contact F.H. Froes, Institute for Materials and Advanced Processes, University of Idaho, Moscow, Idaho 83844; (208) 885-7989; fax (208) 88504009; e-mail imap@uidaho.edu. Author’s Note: Do not expect good club design and advanced materials to cure all golfing woes. They can help, but they are not a panacea for a bad swing and/or lack of strength. For these problems, see your local, friendly Professional Golfers’ Association professional.     

Advanced high-velocity oxygen-fuel coating and candidate materials for protecting LP steam turbine blades against droplet erosion

This paper describes the water droplet erosion characteristics of advanced high-velocity oxygen-fuel (HVOF) coating and the materials used for steam turbine blading. For droplet erosion study, round samples as per ASTM G73-98 (“Standard Practice for Liquid Impingement Erosion Testing,” Vol 03.02, Annual Book of ASTM Standards, ASTM, 2004) were selected. The materials commonly used for steam turbine blading are X20Cr13, X10CrNiMoV122, and Ti6Al4V. During incubation as well as in the long run, advanced HVOF coating has performed much better than all these materials. This is due to enhanced particle kinetic energy caused by optimum flow of oxygen and fuel injection by modifying the fuel injector. Droplet erosion test results of these materials and advanced HVOF coating along with their properties and damage mechanism are reported and discussed in this paper.     

面向等离子体壁材料的腐蚀行为与涂层防护综述

面向等离子体材料(Plasma Facing Materials,PFMs)可保护磁约束核聚变装置部件,使此部件不受芯部边缘等离子体的影响,但等离子体与壁相互作用(Plasma-Wall Interactions,PWI)所引起的高温腐蚀、辐射损伤和燃料滞留等问题已然成为先进核聚变装置的发展瓶颈.目前,低Z材料(C、B...     

Biaxial creep behavior of textured zircaloy tubing

Zirconium alloys are commonly used in light-water reactors as thin-walled tubing to clad highly radioactive fuel. The tubes experience varied stresses at high temperatures while being exposed to high-neutron radiation, resulting in thermal creep and radiation growth and creep. However, the dimensional stability of these materials is important to preventing leakage of fission gases and contamination of the coolant water. Predicting the dimensional changes of the thin-walled tubes is further complicated by the anisotropic nature of the hexagonal close-packed metals. This article summarizes the procedures used in the texture analyses and crystal plasticity in developing model equations to predict the dimensional changes of Zircaloy fuel cladding, both out-of-pile and in-reactor. These methodologies can be extended to the life prediction of these important structures in nuclear reactors.     

Influence of an O2 Type Oxidizing Environment on SiCf/SiC Composites: Properties/Microstructure Relationship

Ceramic matrix composites have been identified as a potential material of core structure for the fourth generation of fission nuclear reactors. Regarding their excellent mechanical behavior in very harsh conditions (high temperature and high irradiation flux), the CVI–SiC_f/SiC composites with pyrocarbon interlayer are of prime interest for the fuel cladding in the gas-cooled fast reactor. Although the working atmosphere is helium in these advanced reactors, the presence of oxidizing impurities could have a significant role on the mechanical behavior of materials subjected to long-term exposures. Within this framework, this study was intended to investigate the influence of oxidation on the SiC_f/SiC composites mechanical properties. Different pre-damage states were intentionally introduced by mechanical tensile tests on plate specimens before performing an oxidation treatment of 1,000 h at 1,000 °C under helium with 10 ppm of O₂. The degradation of the composite was determined from the mechanical behavior of post-exposure specimens. Results were correlated both with microstructural observations of the damage and with characterizations of the generated oxides at the surface of the composites. The most severe decline of mechanical properties occurs for the higher predamaged loadings. Indeed in this case, the silica formed during the oxidation of SiC is not in sufficient quantities to fill the cracks.     

Mo-Re合金研究进展

“Re效应”使得Mo的理化、热电、力学、加工焊接性能得到全面改善。Mo-Re合金由于具有良好的综合性能，被广泛应用于先进核能、航空航天、电子工业、生物医用等领域。尤其是优异的抗辐照性能、与核燃料及碱金属冷却剂的相容性、中子特性等核物理性能，使其成为核反应堆堆芯结构材料的首选。本文从晶体结构、组织性能、制备加工及应用4个方面系统综述了Mo-Re合金的研究现状，并对其发展前景进行了展望。