Sean M. McDeavitt

Evaluation of stainless steel–zirconium alloys as high-level nuclear waste forms

Abstract Stainless steel–zirconium (SS–Zr) alloys have been developed for the consolidation and disposal of waste stainless steel, zirconium, and noble metal fission products such as Nb, Mo, Tc, Ru, Pd, and Ag recovered from spent nuclear fuel assemblies. These remnant waste metals are left behind following electrometallurgical treatment, a molten salt-based process being demonstrated by Argonne National Laboratory. Two SS–Zr compositions have been selected as baseline waste form alloys: (a) stainless steel–15 wt% zirconium (SS–15Zr) for stainless steel-clad fuels and (b) zirconium–8 wt% stainless steel (Zr–8SS) for Zircaloy-clad fuels. Simulated waste form alloys were prepared and tested to characterize the metallurgy of SS–15Zr and Zr–8SS and to evaluate their physical properties and corrosion resistance. Both SS–15Zr and Zr–8SS have multi-phase microstructures, are mechanically strong, and have thermophysical properties comparable to other metals. They also exhibit high resistance to corrosion in simulated groundwater as determined by immersion, electrochemical, and vapor hydration tests. Taken together, the microstructure, physical property, and corrosion resistance data indicate that SS–15Zr and Zr–8SS are viable materials as high-level waste forms.

Laves intermetallics in stainless steel-zirconium alloys

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.

Microstructure and phase identification in type 304 stainless Steel-Zirconium alloys

Stainless steel-zirconium alloys have been developed at Argonne National Laboratory to contain radioactive metal isotopes isolated from spent nuclear fuel. This article discusses the various phases that are formed in as-cast alloys of type 304 stainless steel and zirconium that contain up to 92 wt pct Zr. Microstructural characterization was performed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS), and crystal structure information was obtained by X-ray diffraction. Type 304SS-Zr alloys with 5 and 10 wt pct Zr have a three-phase microstructure—austenite, ferrite, and the Laves intermetallic, Zr(Fe,Cr,Ni)2+x. whereas alloys with 15, 20, and 30 wt pct Zr contain only two phases—ferrite and Zr(Fe,Cr,Ni)2+x. Alloys with 45 to 67 wt pct Zr contain a mixture of Zr(Fe,Cr,Ni)2+x and Zr2(Ni,Fe), whereas alloys with 83 and 92 wt pct Zr contain three phases—α-Zr, Zr2(Ni,Fe), and Zr(Fe,Cr,Ni)2+x. Fe3Zr-type and Zr3Fe-type phases were not observed in the type 304SS-Zr alloys. The changes in alloy microstructure with zirconium content have been correlated to the Fe-Zr binary phase diagram.

Chemical partitioning technologies for an ATW system

Abstract A roadmap for the development of the technology of an Accelerator Transmutation of Waste (ATW) system was recently submitted to the U.S. Congress by the U.S. Department of Energy. One element of this roadmap was a development plan for the separations technologies that would be required to support an ATW system operating with a sustained feed of 1,450 tonnes of commercial light water reactor spent fuel per year. A Technical Working Group was constituted to identify appropriate separations processes and prepare a plan for their development. The baseline process selected combines aqueous and pyrochemical processes to enable efficient separation of uranium, technetium, iodine, and the transuranic elements from LWR spent fuel in the head-end step. For the recycle of unburned transuranics and newly-generated technetium and iodine from irradiated ATW transmuter assemblies, which were given to be metallic in form, a second and quite different pyrochemical process was identified. The diversity of processing methods was chosen for both technical and economic factors; aqueous methods are deemed to be better suited to large tonnages of commercial oxide spent fuel, while it is considered that pyrochemical processes can be exploited effectively in smaller-scale operations, particularly when the application is to metallic fuels or targets. A six-year technology evaluation and development program is foreseen, by the end of which an informed decision can be made on proceeding with demonstration of the ATW system.

High temperature interaction behavior at liquid metal-ceramic interfaces.

Liquid metal/ceramic interaction experiments were undertaken at elevated temperatures with the purpose of developing reusable crucibles for melting reactive metals. The metals used in this work included zirconium (Zr), Zr-8 wt.% stainless steel, and stainless steel containing 15 wt.% Zr. The ceramic substrates include yttria, Zr carbide, and hafnium (Hf) carbide. The metal-ceramic samples were placed on top of a tungsten (W) dish. These experiments were conducted with the temperature increasing at a controlled rate until reaching set points above 2000 °C; the systems were held at the peak temperature for about five min and then cooled. The atmosphere in the furnace was argon (Ar). An outside video recording system was used to monitor the changes on heating up and cooling down. All samples underwent a post-test metallurgical examination. Pure Zr was found to react with yttria, resulting in oxygen (O) evolution at the liquid metal-ceramic interface. In addition, dissolved O was observed in the as-cooled Zr metal. Yttrium (Y) was also present in the Zr metal, but it had segregated to the grain boundaries on cooling. Despite the normal expectations for reactive wetting, no transition interface was developed, but the Zr metal was tightly bound to yttria ceramic. Similar reactions occurred between the yttria and the Zr-stainless steel alloys. Two other ceramic samples were Zr carbide and Hf carbide; both carbide substrates were wetted readily by the molten Zr, which flowed easily to the sides of the substrates. The molten Zr caused a very limited dissolution of the Zr carbide, and it reacted more strongly with the Hf carbide. These reactive wetting results are relevant to the design of interfaces and the development of reactive filler metals for the fabrication of high temperature components through metal-ceramic joining. Parameters that have a marked impact on this interface reaction include the thermodynamic stability of the substrate, the properties of the modified interface, the temperature-dependent solubility limits of the liquids and solid phases, and the high-temperature stoichiometry of the ceramic.

Actinide distribution in a stainless steel-15 wt% zirconium high-level nuclear waste form

Abstract Actinide-bearing waste forms are being produced from metallic remnants resulting from the electrometallurgical extraction of uranium from EBR-II spent fuel. The baseline metal waste form (MWF) is a stainless steel–15 wt% zirconium (SS–15Zr) alloy that may contain up to 10 wt% actinides, mostly in the form of uranium. This article presents the results of scanning electron microscopy (SEM), transmission electron microscopy (TEM), and neutron diffraction on SS–15Zr alloys containing uranium, plutonium, and neptunium. Neutron diffraction results showed that the addition of uranium to SS–15Zr does not result in the formation of discrete uranium-rich phases. The lattice parameters of the ZrFe 2 -type intermetallics are larger in uranium-containing SS–15Zr alloys and are consistent with the substitution of uranium at zirconium sites of the ZrFe 2 lattice. SEM studies showed that actinides are present only in the ZrFe 2 -type intermetallics; moreover, both actinide-rich and actinide-deficient areas are found within the Laves compound. TEM showed that the simultaneous presence of multiple Laves polytypes, each with a different preference for the uranium atom, results in the uranium concentration gradients observed within the Laves intermetallics.

Formation of the Fe23Zr6 phase in an Fe-Zr alloy

Abstract This study confirms the existence of Fe 23 Zr 6 in an Fe-9.8 at.% Zr alloy. This phase forms from the α-Fe + Fe 2 Zr eutectic during high temperature annealing. Previously, Fe 23 Zr 6 formation was explained by simple diffusion of iron atoms into the Laves intermetallic; however, this mechanism does not explain the presence of α-Zr within the Fe 2 Zr 6 phase. We have proposed a two-step mechanism for Fe 23 Zr 6 formation: Fe 2 Zr decomposes into Fe 23 Zr 6 + α-Zr, and α-Zr reacts with α-Fe to form more Fe 23 Zr 6 . The slow kinetics of Fe 23 Zr 6 formation is due to its dependence on atomic diffusion through the intermetallic phases.

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.

Thoria-based cermet nuclear fuel : cermet fabrication and behavior estimates.

Cermet nuclear fuels have been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) develop enabling technologies required for the economic application of this new fuel form. Critical elements in the demonstration of this new fuel form include developing low-cost fabrication methods and characterizing the cermet properties and important performance parameters. A powder-in-tube drawing and heat treatment process is being evaluated as an alternative to hot extrusion. In this method, zirconium metal and ceramic microspheres are mixed, poured into a Zircaloy shell, and compacted into simulated fuel pins. Important processing variables being evaluated include the amount of compaction required to achieve a desired matrix density and the inter-drawing thermal treatment temperature required to achieve adequate matrix fusion and grain growth.

Microscopy and neutron diffraction study of a zirconium-8 wt% stainless steel alloy

The electrometallurgical treatment of zirconium-based and Zircaloy-clad spent nuclear fuels will yield a metal waste form. The baseline composition for the waste form is zirconium-8 wt% stainless steel (Zr-8SS). The microstructure of the Zr-8SS alloy has been studied by scanning electron microscopy, energy dispersive spectroscopy, and neutron diffraction. The phases present in the as-cast alloy include Zr(α), Zr3(Fe,Ni), Zr2(Fe,Ni), Zr2(Fe,Cr), and Zr(Fe,Cr)2; a solidification sequence has been proposed to explain the formation and morphology of these phases. Alloy phase stability has been studied by thermal aging at 780°C for periods up to 30 days. The phase changes that occur during thermal aging include an increase in Zr3(Fe,Ni) and a decrease in Zr2(Fe,Ni) content; reaction mechanisms have been proposed to explain these changes. The lattice parameters of alloy phases have been determined by neutron diffraction and found to be in agreement with those previously reported for similar phases. This study of alloy microstructures is the first step towards understanding the actinide and fission product distribution and predicting the corrosion behavior of the Zr-8SS metal waste form.