HCO3-和SO42-对Cu点蚀行为的影响

在不同浓度配比的HCO₃^-和SO₄^2-混合溶液中,利用循环极化电化学测试方法和SEM,对Cu工作电极的循环极化行为和点蚀表面形貌进行了系统的研究.结果表明,在高电位范围的循环极化实验中,Cu的点蚀行为可分为活性溶解型点蚀和钝化膜破裂型点蚀;随SO₄^2-浓度的升高Cu点蚀的敏感性增大.由于HCO₃^-与SO₄^2-的协同作用,随HCO₂^-浓度升高点蚀敏感性呈先增大后减小的规律.在钝化膜破裂型点蚀中,SO₄^2-提高Cu点蚀的诱发能力;HCO₃^-降低Cu点蚀的诱发能力.2种离子对点蚀自修复能力的影响无明显规律.     

核废料中裂变产生的铂族金属（FPs）的开发、应用和发展

在核废料中存在大量的由核裂变反应产生的铂族金属（FPs：包括Ru，Rh和Pd）。目前全世界的核废料中FPs的总量可以和自然界储量相比拟。并且随时间推移FPs的总量还会增加。未来FPs将是铂族金属的一个重要资源。本文介绍了FPs在核废料中的存在形式以及常见提取方法。详细给出了FPs中各种核素的放射性质，阐述了开发和利用FPs的基本观点、方法和思路。介绍了FPs在核工业和高纯氢气制造上的应用，以及其它潜在的FPs应用领域。     

Stabilizing toxic metal concentrates by using SMITE

Synthetic mineral immobilization technology is an approach for the treatment of heavy-metal concentrates. Although the notion of using synthetic mineral analogs for the stabilization and consolidation of nuclear waste has been discussed for more than 40 years, its application to inorganic hazardous waste, in general, is only now being realized. The advantage of this technology is that high-waste-loaded and high-density waste forms can be fabricated while maintaining excellent chemical durability. These properties translate into considerable savings during transport and disposal.     

Material corrosion issues for nuclear waste disposition in Yucca Mountain

For more than two decades, an extensive scientific effort has been underway to determine whether Yucca Mountain, Nevada, is a suitable site for a deep underground repository for spent nuclear fuel and high-level radioactive waste. Even though the geologic site is stable, additional engineered barriers are planned, including waste packages, drip shields, and tunnel inverts that will be within the emplacement tunnels. Research is under way into the best materials for corrosion prevention in those engineered barriers to ensure their long-term mechanical integrity.     

Corrosion behaviour of carbon steel,titanium and titanium alloy in simulated underground water in Beishan area preselected for nuclear waste repository in China

The groundwater will completely infiltrate to the surface of nuclear waste container after the closure of its deep geological disposal. Therefore, the corrosion behaviour of Q235 carbon steel, titanium and titanium alloy, which were the candidates as the container materials for high-level nuclear waste disposal in simulated groundwater solution of Beishan to be served as the preselected high-level nuclear waste disposal area in China at different temperatures, was studied through electrochemical methods including open circuit potential, electrochemical impedance spectroscopy and potentiodynamic polarisation curve measurements. The results show that the corrosion rate of titanium and titanium alloy is lower than that of carbon steel at all temperatures, and they are more promising as container materials. Another phenomenon is that higher temperatures facilitate the protection performance of corrosion products compared with lower temperatures.     

Online monitoring system for metallic components in underground laboratory

With the increasing application of nuclear energy in power industry, China has accumulated a large amount of high-level radioactive nuclear waste, which are characterized by high toxicity, high radioactivity, and long half-life. Improper disposal will cause great harm to human life and property. China has decided to establish an underground laboratory for geological disposal of high-level radioactive waste (HLW) in Beishan, Gansu Province, to study the scientific issues related to geological disposal of HLW. In this paper, an online monitoring system including two kinds of integrated inductance probe for the corrosion monitoring of metallic materials used in atmospheric or bentonite environment in Beishan underground laboratory was developed, and the reliability and accuracy of the system were studied. The results show that the system can satisfy the requirement of online monitoring of the corrosion rate of metallic materials, ambient temperature, resistivity, and humidity of bentonite in the underground laboratory for geological disposal of HLW in different environments.     

Stainless steel-zirconium waste forms from the treatment of spent nuclear fuel

Stainless steel-zirconium waste-form alloys have been developed for the disposal of metallic wastes recovered from spent nuclear fuel using the electrometallurgical process developed by Argonne National Laboratory. The metal waste comprises the spent-fuel cladding, noble-metal fission products, and other metallic constituents remaining after electrorefining. Two nominal waste-form compositions have been slected: stainless steel-clad fuels and zirconium-8 wt.% stainless steel for Zircaloy-clad fuels. These alloys are very corrosion resistant. Tests performed with these alloys indicate favorable behavior for use high-level nuclear waste forms.     

Entwicklung,Verarbeitung und Einsatz des stickstofflegierten,hochmolybdänhaltigen Stahles X 3 CrNiMoN 17 13 5

Development, processing and use of nitrogen alloyed high molybdenum steel X3CrNiMoN 17 13 5 Addition of 0,15% Ni to the steel X5-CrNiMo 17 13 delays the precipitation of intermetallic phases and chromium carbides. The nitrogen containing steel can be welded up to a plate thickness of 30 mm without precipitation; hot cracking has not been observed either. The weld metal corresponds to the base material with respect to ductility and corrosion resistance. It is resistant to pitting, crevice corrosion and stress corrosion cracking, in particular in mixed acids. Successful applications include textile industry (bleaching equipment, heating equipment), shipbuilding (driving shafts, exhaust gas coolers, waste gas condenser tubing), water desalination equipment, nuclear industry (evaporator for radioactive waste water, processing of radioactive waste in fluoride containing nitric acid), chemical industry (pressure container material, heat exchangers for brackish water).     

Corrosion domain analysis of copper corrosion in aqueous media

Abstract

The technology that is being developed in Sweden for the disposal of high level nuclear waste calls for storage of the waste in copper canisters, which are encapsulated in a bentonite buffer contained in drill holes in the floors of drifts (tunnels) in a granitic rock repository. A controversial issue has arisen during the development of this technology: that copper, when in contact with pure water under anoxic conditions corrodes and hence is not immune as previously believed. This issue is resolved in the present paper by deriving corrosion domain diagrams as a means of presenting the thermodynamics of the system in the clearest form possible, when assessing the immunity and activation of copper.     

Predicting the Lifetimes of Nuclear Waste Containers

As for many aspects of the disposal of nuclear waste, the greatest challenge we have in the study of container materials is the prediction of the long-term performance over periods of tens to hundreds of thousands of years. Various methods have been used for predicting the lifetime of containers for the disposal of high-level waste or spent fuel in deep geological repositories. Both mechanical and corrosion-related failure mechanisms need to be considered, although until recently the interactions of mechanical and corrosion degradation modes have not been considered in detail. Failure from mechanical degradation modes has tended to be treated through suitable container design. In comparison, the inevitable loss of container integrity due to corrosion has been treated by developing specific corrosion models. The most important aspect, however, is to be able to justify the long-term predictions by demonstrating a mechanistic understanding of the various degradation modes.     

Isolating wastes in the electrometallurgical treatment of spent nuclear fuel

The main hazards in spent nuclear fuel are fission products and transuranic radionuclides. An electrometallurgical treatment is designed to isolate these elements by electrorefining and then place them in waste forms suitable for geologic disposal. In the highly reducing chemical environment used for electrometallurgical treatment, fuel cladding and transition-metal fission products remain as metals; these metals are collected and melted to form a highly corrosion-resistant waste form. Other fission-product elements and transuramic elements collect in the molten-salt process fluid and are removed by ion exchange into zeolite, which is further processed to make a durable-composite ceramic waste form.     

Lead as a containment material for nuclear waste

Summarizing, it can be said that lead—even with its tendency toward grain growth, which results in decreased corrosion resistance and a need for structural support—is potentially the ideal material from which to manufacture high-level nuclear waste “coffins” for subterranean burial.     

Cladding and duct materials for advanced nuclear recycle reactors

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.     

The role of hydrometallurgy in treating nuclear waste

This paper describes three case histories in which Eichrom's Diphonix^TM resin was used successfully in pilot-scale equipment for the treatment of radioactive waste streams. Two of the cases describe the use of the resin for reducing actinides, primarily uranium, in an analytical waste stream and in a fuel processing effluent. The third case history reports the use of Diphonix resin to control radioactive zinc and cobalt in a nuclear power plant effluent.     

Modeling the hydrogen-induced cracking of titanium alloys in nuclear waste repository environments

This article considers hydrogen-induced cracking of titanium grade 7 and other relevant titanium alloys in the current waste package design for the environmental conditions anticipated within the proposed Yucca Mountain nuclear waste repository in Nevada. In particular, corrosion processes possible in the aqueous environments expected within this site are considered, including key corrosion processes that could occur and the expected corrosion performance of these alloys. It can be concluded that, based on the conservative modeling approaches adopted, hydrogen-induced cracking of titanium alloys will not occur under nuclear waste repository conditions since there will not be sufficient hydrogen in the alloy even after 10,000 years of emplacement. For more information, contact Fred Hua, Bechtel SAIC Company, LLC, 1180 Town Center Drive, Las Vegas, NV 89144; e-mail Fred_Hua@ymp.gov.     

The effect of radiation on nuclear waste forms

Radioactive decay within solidified nuclear wastes leads to physical and chemical changes that can alter the rate at which radionuclides are released to the biosphere. Under current policy, the geologic repository for such wastes is relied upon to meet federal release-rate regulations. If these regulatory and acceptance criteria cannot be achieved or are tightened, it may become necessary to rely more heavily on the physical and chemical durability of the waste form itself. Although radiation effects are known to degrade nuclear waste forms, the mechanisms and extent of the degradation are neither well understood nor under current investigation.     

高放废物包装容器材料腐蚀研究进展

高放废物安全处置的各项研究工作中,废物包装容器的选材及在处置环境下的腐蚀性问题是重点内容之一。各国针对高放废物的地质处置库概念设计中,对拟选高放废物容器包装材料的腐蚀速率、腐蚀产物、使用寿命及影响因素进行了实验测试和分析。国外关于高放废物包装容器的研究经验对我国开展相关工作具有积极的借鉴意义。     

硼酸缓冲溶液中pH值和Cl~-浓度对Cu腐蚀行为的影响

在硼酸缓冲溶液中,采用动电位极化、电化学阻抗谱(EIS)和半导体电容分析方法分别研究了Cu电极的极化行为及其表面人工Cu_2O钝化膜的化学稳定性.结果表明,低pH值、高Cl~-浓度均造成Cu_2O钝化膜的破坏和溶解.高Cl~-浓度时,Cu_2O钝化膜的半导体性质由p型转变为n型,使Cl~-更容易进入钝化膜与Cu~+络合,并破坏钝化膜从而加速腐蚀.高pH值、低Cl~-浓度有利于Cu_2O钝化膜稳定.     

Ceramic-composite waste forms from the electrometallurgical treatment of spent nuclear fuel

Argonne National Laboratory is developing a method to treat spent nuclear fuel in a molten-salt electrorefiner. Glass-bonded zeolite and sodalite are both being developed as ceramic waste forms. The ceramic waste form will contain the fission product (e.g., rare earth, alkali and alkaline-earth metals, halogens, and chalcogens) and transuranic radionuclides that accumulate in the electrorefiner salt. Zeolite A can fully incorporate both the salt and the radionuclides into its crystal structure. Salt-loaded zeolite A is mixed with glass frit; the blend undergoes hot isostatic pressing to produce a monolithic leach-resistant waste form. Alternatively, the salt-loaded zeolite may be converted to sodalite simply by heat treating first, then adding the glass and hot pressing.     

Radiation effects in a model ceramic for nuclear waste disposal

The safe immobilization of nuclear waste in geological repositories is one of the major scientific challenges facing humanity today. Crystalline ceramics hold the promise of locking up actinides from nuclear fuel and excess weapons plutonium in their structure thereby isolating them from the environment. This paper presents the atomistic details of radiation damage in a model ceramic, zircon.