Dennis D. Keiser

Interdiffusion behavior in UPuZr fuel versus stainless steel couples

Abstract Irradiation of metallic Uue5f8Zr or Uue5f8Puue5f8-Zr nuclear fuel alloys in a reactor results in fuel swelling that can lead to interactions between the fuel and cladding steel. These interactions are complicated, involving many variables and producing complex microstructures. This paper contributes to the understanding of fuel-cladding interaction by reporting results from diffusion couples annealed at 650°C for 100 h between a Uue5f8Puue5f8Zr alloy and stainless steels with and without Ni. SEM-EDX analyses of these couples were utilized to determine the composition of phases throughout the diffusion zone and to generate composition profiles. Overall, a diffusion couple between a Uue5f8Puue5f8Zr fuel alloy and a cladding steel that does not contain Ni develops the largest diffusion zone of any of the fuel-cladding steel diffusion couples; a slightly smaller diffusion zone develops for a couple between a Uue5f8Puue5f8Zr alloy and a cladding steel containing Ni. By comparing these results with those from earlier binary fuel studies, it is concluded that the presence of Pit increases interdiffusion of the various components in fuel-cladding steel diffusion couples.

Influence of technetium on the microstructure of a stainless steel-zirconium alloy

Stainless steel–zirconium alloys are being developed for the disposal of metallic waste generated during the electrometallurgical treatment of spent Experimental Breeder Reactor (EBR-II) fuel. The metallic waste contains the fission product technetium, which must be incorporated into a stable waste form matrix to prevent its release into the environment. The baseline waste form for metallic waste from EBR-II fuels is a stainless steel–15 wt% zirconium (SS–15Zr) alloy. The microstructure of SS–15Zr alloys containing 2 wt% technetium was characterized using a combination of microscopy, spectroscopy, diffraction, and chemical analysis techniques. Peaks corresponding to the iron solid solutions ferrite and austenite, ZrFe2-type Laves polytypes C36 and C15, and an Fe23Zr6-type intermetallic were identified in diffraction patterns of the alloy. Discrete technetium-rich phases were not observed either in diffraction patterns or in the microstructure; the element partitioned into various phases of the SS–15Zr alloy. Technetium favors ferrite and austenite over the Zr-based intermetallics. The lattice parameters of the Zr-based intermetallics are smaller than those in an alloy without technetium, which appears to substitute at the zirconium sites of the intermetallic lattice.

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.

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.

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

Abstract Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U–Zr alloy fuel elements irradiated in the Experimental Breeder Reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining.

Diffusion in the Ce-Nd system

To date, information on interdiffusion in intralanthanide systems is limited. Interdiffusion studies have been reported for the Pr-Nd, La-Pr, and La-Ce systems. Yet, no data is available for the Ce-Nd system. This paper reports the results of Ce-Nd diffusion couples annealed at 425 C, 550 C, and 650 C. Both intrinsic and interdiffusion data have been calculated, as well as activation energy information as a function of composition. Comments will be made on how well the composition of the boundaries of the [({gamma}Ce) + ({beta}Ce, {alpha}Nd)] two-phase region in the Ce-Nd phase diagram agree with diffusion couple results.

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

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.

Analysis of interdiffusion in multi-phase systems

Interdiffusion in multicomponent systems is typically studied through diffusion couple experiments in which two alloys are abutted and annealed at an elevated temperature for a fixed time. For diffusion zones with two-phase or multi-phase layers, however, the determination of fluxes and interdiffusion coefficients for a component is not straightforward. There is no single concentration profile for a component. Instead, there are distinct profiles for that component in each phase. In this paper the authors present two techniques to overcome the difficulties of analyzing multi-phase diffusion. For two-phase regions, quantitative metallography can be used to evaluate average compositions along the diffusion zone. For layers with two or more simultaneous phases, SEM/EDX linescan average composition measurements can be taken along with spot readings. The two techniques are compared in a study of a six-component diffusion couple experiment between a U-Pu-Zr alloy and a Fe-Ni-Cr stainless steel.

Phase identification in a U–Zr/Ni–Cr diffusion couple using synchrotron radiation

The diffusion zone between a U-23 at.{percent} Zr alloy and a Ni-16.4 at.{percent} Cr alloy exhibited nine distinct phase layers, many of which were mixtures of two phases. Four single-phase regions were less than 10 {mu}m wide. To identify these phases by diffractometry, a synchrotron x-ray beam was collimated by a 50 {mu}m by 1 mm slit. This beam was translated across the sample to obtain diffraction patterns throughout the diffusion zone. In this way, only a few phases were simultaneously within the beam, easing identification of the phases. Strains in the lattice due to solid solution were also observed. These microdiffraction techniques are applicable to a wide range of material systems. {copyright} {ital 1996 Materials Research Society.}