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Phase stability of laves intermetallics in stainless steel-zirconium alloys.

Description: Phase transformations occurring in a stainless steel-15 wt% zirconium (SS-15Zr) alloy were studied by in situ neutron diffraction. Neutron diffraction patterns as a function of time were obtained on alloys that were held at various elevated temperatures (1084-1275 C). As-cast SS-15Zr alloys contain ferrite, austenite, ZrFe{sub 2}-type Laves polytypes C36 and C15, and small amounts of a Fe{sub 23}Zr{sub 6}-type intermetallic. Annealing at high temperatures resulted in an increase of the Fe{sub 23}Zr{sub 6}, intermetallic content. The C15 Laves polytype is the equilibrium phase for T {le} 1230 C; C36 is the stable polytype at higher temperatures ({approximately}1275 C). Phase changes were slow for temperatures <1100 C.These findings have important implications for use of the SS-15Zr alloy as a nuclear waste form.
Date: April 8, 1999
Creator: Abraham, D. P.
Partner: UNT Libraries Government Documents Department

Corrosion behavior of stainless steel-zirconium alloy waste forms.

Description: Stainless steel-zirconium (SS-Zr) alloys are being considered as waste forms for the disposal of metallic waste generated during the electrometallurgical treatment of spent nuclear fuel. The baseline waste form for spent fuels from the EBR-II reactor is a stainless steel-15 wt.% zirconium (SS-15Zr) alloy. This article briefly reviews the microstructure of various SS-Zr waste form alloys and presents results of immersion corrosion and electrochemical corrosion tests performed on these alloys. The electrochemical tests show that the corrosion behavior of SS-Zr alloys is comparable to those of other alloys being considered for the Yucca Mountain geologic repository. The immersion tests demonstrate that the SS-Zr alloys are resistant to selective leaching of fission product elements and, hence, suitable as candidates for high-level nuclear waste forms.
Date: January 13, 1999
Creator: Abraham, D. P.
Partner: UNT Libraries Government Documents Department

Corrosion testing of stainless steel-zirconium metal waste form.

Description: Stainless steel-zirconium (SS-Zr) alloys are being considered as waste forms for the disposition of metallic waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste forms contain irradiated cladding hulls, components of the alloy fuel, noble metal fission products, and actinide elements. The baseline waste form is a stainless steel-15 wt% zirconium (SS-15Zr) alloy. This article presents microstructure and some of the corrosion studies being conducted on the waste form alloys. Electrochemical corrosion, immersion corrosion, and vapor hydration tests have been performed on various alloy compositions to evaluate corrosion behavior and resistance to selective leaching of simulated fission products. The SS-Zr waste forms are successful at the immobilization and retention of fission products and show potential for acceptance as high-level nuclear waste forms.
Date: December 14, 1998
Creator: Abraham, D. P.
Partner: UNT Libraries Government Documents Department

Metal waste forms from treatment of EBR-II spent fuel.

Description: Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper.
Date: May 18, 1998
Creator: Abraham, D. P.
Partner: UNT Libraries Government Documents Department

TEM characterization of corrosion products formed on a SS-15ZR alloy.

Description: The corrosion products formed on a stainless steel-15Zr (SS-15Zr) alloy have been characterized by transmission electron microscopy (TEM) and energy dispersive x-ray spectroscopy (EDS). Examination of alloy particles that were immersed in 90 C deionized water for two years revealed that different corrosion products were formed on the stainless steel and intermetallic phases. Two corrosion products were identified on an austenite particle: trevorite (NiFe{sub 2}O{sub 4}) in the layer close to the metal and maghemite (Fe{sub 2}O{sub 3}) in the outer layer. The corrosion layer formed on the intermetallic was uniform, adherent, and amorphous. The EDS analysis indicated that the layer was enriched in zirconium when compared with the intermetallic composition. High-resolution TEM images of the intermetallic-corrosion layer interface show an interlocking metal-oxide interface which may explain the relatively strong adherence of the corrosion layer to the intermetallic surface. These results will be used to evaluate corrosion mechanisms and predict long-term corrosion behavior of the alloy waste form.
Date: January 4, 2000
Creator: Luo, J. S. & Abraham, D. P.
Partner: UNT Libraries Government Documents Department

Metal waste forms from the electrometallurgical treatment of spent nuclear fuel

Description: Stainless steel-zirconium alloys are being developed for the disposal of radioactive metal isotopes isolated using an electrometallurgical treatment technique to treat spent nuclear fuel. The nominal waste forms are stainless steel-15 wt% zirconium alloy and zirconium-8 wt% stainless steel alloy. These alloys are generated in yttria crucibles by melting the starting materials at 1,600 C under an argon atmosphere. This paper discusses the microstructures, corrosion and mechanical test results, and thermophysical properties of the metal waste form alloys.
Date: May 1, 1996
Creator: Abraham, D.P.; McDeavitt, S.M. & Park, J.
Partner: UNT Libraries Government Documents Department

Characterization of oxidation products on a ZrFe{sub 2}-type laves intermetallic exposed to 200{degree}C steam.

Description: The release of radioactive elements from the stainless steel-15 wt% zirconium (SS-15Zr) metal waste form will be governed by the corrosion behavior of ZrFe{sub 2}-type intermetallics phases present in the alloy. In this article, oxidation products that formed on a ZrFe{sub 2}-type intermetallic sample exposed to 200 C steam were characterized by Auger Electron Spectroscopy (AES) and Transmission Electron Microscopy (TEM). The data revealed two oxide layers on the sample surface: an outer crystalline iron-oxide layer and an inner amorphous zirconium-rich layer believed to be zirconium oxide. Thermodynamic considerations indicate that the zirconium-rich layer formed first. The iron-oxide layer appears to have resulted from the diffusion of iron through the zirconium-rich layer to the oxide-vapor interface.
Date: November 20, 2000
Creator: Abraham, D. P.; Dietz, N. & Finnegan, N.
Partner: UNT Libraries Government Documents Department

Stainless steel-zirconium alloy waste forms

Description: An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ``noble`` nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation.
Date: July 1996
Creator: McDeavitt, S. M.; Abraham, D. P.; Keiser, D. D., Jr. & Park, J. Y.
Partner: UNT Libraries Government Documents Department

Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel

Description: Stainless steel-zirconium waste form alloys are being developed for the disposal of metallic wastes recovered from spent nuclear fuel using an electrometallurgical process developed by Argonne National Laboratory. The metal waste form comprises the fuel cladding, noble metal fission products and other metallic constituents. Two nominal waste form compositions are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels. The noble metal fission products are the primary source of radiation and their contribution to the waste form radioactivity has been calculated. The disposition of actinide metals in the waste alloys is also being explored. Simulated waste form alloys were prepared to study the baseline alloy microstructures and the microstructural distribution of noble metals and actinides, and to evaluate corrosion performance.
Date: July 1, 1996
Creator: McDeavitt, S.M.; Abraham, D.P.; Park, J.-Y. & Keiser, D.D. Jr.
Partner: UNT Libraries Government Documents Department

Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

Description: Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion.
Date: October 1, 1996
Creator: McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr. & Park, J.Y.
Partner: UNT Libraries Government Documents Department

Laves intermetallics in stainless steel-zirconium alloys

Description: Laves intermetallics have a significant effect on properties of metal waste forms being developed at Argonne National Laboratory. These waste forms are stainless steel-zirconium alloys that will contain radioactive metal isotopes isolated from spent nuclear fuel by electrometallurgical treatment. The baseline waste form composition for stainless steel-clad fuels is stainless steel-15 wt.% zirconium (SS-15Zr). This article presents results of neutron diffraction measurements, heat-treatment studies and mechanical testing on SS-15Zr alloys. The Laves intermetallics in these alloys, labeled Zr(Fe,Cr,Ni){sub 2+x}, have both C36 and C15 crystal structures. A fraction of these intermetallics transform into (Fe,Cr,Ni){sub 23}Zr{sub 6} during high-temperature annealing; the authors have proposed a mechanism for this transformation. The SS-15Zr alloys show virtually no elongation in uniaxial tension, but exhibit good strength and ductility in compression tests. This article also presents neutron diffraction and microstructural data for a stainless steel-42 wt.% zirconium (SS-42Zr) alloy.
Date: May 1, 1997
Creator: Abraham, D.P.; McDeavitt, S.M. & Richardson, J.W. Jr.
Partner: UNT Libraries Government Documents Department

The effect of actinides on the microstructural development in a metallic high-level nuclear waste form

Description: Waste forms to contain material residual from an electrometallurgical treatment of spent nuclear fuel have been developed by Argonne National Laboratory. One of these waste forms contains waste stainless steel (SS), fission products that are noble to the process (e.g., Tc, Ru, Pd, Rh), Zr, and actinides. The baseline composition of this metallic waste form is SS-15wt.% Zr. The metallurgy of this baseline alloy has been well characterized. On the other hand, the effects of actinides on the alloy microstructure are not well understood. As a result, SS-Zr alloys with added U, Pu, and/or Np have been cast and then characterized, using scanning electron microscopy, transmission electron microscopy, and neutron diffraction, to investigate the microstructural development in SS-Zr alloys that contain actinides. Actinides were found to congregate non-uniformally in a Zr(Fe,Cr,Ni){sub 2+x} phase. Apparently, the actinides were contained in varying amounts in the different polytypes (C14, C15, and C36) of the Zr(Fe,Cr,Ni){sub 2+x} phase. Heat treatment of an actinide-containing SS-15 wt.% Zr alloy showed the observed microstructure to be stable.
Date: October 25, 1999
Creator: Keiser, D. D., Jr.; Sinkler, W.; Abraham, D. P.; Richardson, J. W., Jr. & McDeavitt, S. M.
Partner: UNT Libraries Government Documents Department

Corrosion of structural materials by lead-based reactor coolants.

Description: Advanced nuclear reactor design has, in recent years, focused increasingly on the use of heavy-liquid-metal coolants, such as lead and lead-bismuth eutectic. Similarly, programs on accelerator-based transmutation systems have also considered the use of such coolants. Russian experience with heavy-metal coolants for nuclear reactors has lent credence to the validity of this approach. Of significant concern is the compatibility of structural materials with these coolants. We have used a thermal convection-based test method to allow exposure of candidate materials to molten lead and lead-bismuth flowing under a temperature gradient. The gradient was deemed essential in evaluating the behavior of the test materials in that should preferential dissolution of components of the test material occur we would expect dissolution in the hotter regions and deposition in the colder regions, thus promoting material transport. Results from the interactions of a Si-rich mild steel alloy, AISI S5, and a ferritic-martensitic stainless steel, HT-9, with the molten lead-bismuth are presented.
Date: November 16, 2000
Creator: Abraham, D. P.; Leibowitz, L.; Maroni, V. A.; McDeavitt, S. M. & Raraz, A. G.
Partner: UNT Libraries Government Documents Department

Electrochemical corrosion testing of metal waste forms

Description: Electrochemical corrosion tests have been conducted on simulated stainless steel-zirconium (SS-Zr) metal waste form (MWF) samples. The uniform aqueous corrosion behavior of the samples in various test solutions was measured by the polarization resistance technique. The data show that the MWF corrosion rates are very low in groundwaters representative of the proposed Yucca Mountain repository. Galvanic corrosion measurements were also conducted on MWF samples that were coupled to an alloy that has been proposed for the inner lining of the high-level nuclear waste container. The experiments show that the steady-state galvanic corrosion currents are small. Galvanic corrosion will, hence, not be an important mechanism of radionuclide release from the MWF alloys.
Date: December 14, 1999
Creator: Abraham, D. P.; Peterson, J. J.; Katyal, H. K.; Keiser, D. D. & Hilton, B. A.
Partner: UNT Libraries Government Documents Department

Modeling corrosion and constituent release from a metal waste form.

Description: Several ANL ongoing experimental programs have measured metal waste form (MWF) corrosion and constituent release. Analysis of this data has initiated development of a consistent and quantitative phenomenology of uniform aqueous MWF corrosion. The effort so far has produced a preliminary fission product and actinide release model based on measured corrosion rates and calibrated by immersion test data for a 90 C J-13 and concentrated J-13 solution environment over 1-2 year exposure times. Ongoing immersion tests of irradiated and unirradiated MWF samples using more aggressive test conditions and improved tracking of actinides will serve to further validate, modify, and expand the application base of the preliminary model-including effects of other corrosion mechanisms. Sample examination using both mechanical and spectrographic techniques will better define both the nature and durability of the protective barrier layer. It is particularly important to assess whether the observations made with J-13 solution at 900 C persist under more aggressive conditions. For example, all the multiplicative factors in Table 1 implicitly assume the presence of protective barriers. Under sufficiently aggressive test conditions, such protective barriers may very well be altered or even eliminated.
Date: December 4, 2000
Creator: Bauer, T. H.; Fink, J. K.; Abraham, D. P.; Johnson, I.; Johnson, S. G. & Wigeland, R. A.
Partner: UNT Libraries Government Documents Department

Diagnostic examination of Generation 2 lithium-ion cells and assessment ofperformance degradation mechanisms.

Description: The Advanced Technology Development (ATD) Program is a multilaboratory effort to assist industrial developers of high-power lithium-ion batteries overcome the barriers of cost, calendar life, abuse tolerance, and low-temperature performance so that this technology may be rendered practical for use in hybrid electric vehicles (HEVs). Included in the ATD Program is a comprehensive diagnostics effort conducted by researchers at Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), and Lawrence Berkeley National Laboratory (LBNL). The goals of this effort are to identify and characterize processes that limit lithium-ion battery performance and calendar life, and ultimately to describe the specific mechanisms that cause performance degradation. This report is a compilation of the diagnostics effort conducted since spring 2001 to characterize Generation 2 ATD cells and cell components. The report is divided into a main body and appendices. Information on the diagnostic approach, details from individual diagnostic techniques, and details on the phenomenological model used to link the diagnostic data to the loss of 18650-cell electrochemical performance are included in the appendices. The main body of the report includes an overview of the 18650-cell test data, summarizes diagnostic data and modeling information contained in the appendices, and provides an assessment of the various mechanisms that have been postulated to explain performance degradation of the 18650 cells during accelerated aging. This report is intended to serve as a ready reference on ATD Generation 2 18650-cell performance and provide information on the tools for diagnostic examination and relevance of the acquired data. A comprehensive account of our experimental procedures and resulting data may be obtained by consulting the various references listed in the text. We hope that this report will serve as a roadmap for the diagnostic analyses of other lithium-ion technologies being evaluated for HEV applications. It is our hope that the information contained ...
Date: July 15, 2005
Creator: Abraham, D. P.; Dees, D. W.; Knuth, J.; Reynolds, E.; Gerald, R.; Hyung,Y.-E. et al.
Partner: UNT Libraries Government Documents Department