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

Stainless steel-zirconium alloy waste forms

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

Corrosion of structural materials by lead-based reactor coolants.

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

Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase

Synthesis and casting of a lithium-bismuth compound for an ion-replacement electrorefiner.

The intermetallic compound Li{sub 3}Bi played an integral part in the demonstration of an ion replacement electrorefining method developed at Argonne National Laboratory. The Li{sub 3}Bi compound was generated in a tilt-pour casting furnace using high-purity lithium and bismuth metals as the initial charge. At first, small-scale ({approximately}20 g) experiments were conducted to determine the materials synthesis parameters. In the end, four larger-scale castings (500 g to 1250 g) were completed in a tantalum crucible. The metals were heated slowly to melt the charge, and the formation reaction proceeded vigorously above the melting point of bismuth ({approximately}270 C). For the large-scale melts, the furnace power was temporarily turned off at this point. After several minutes, the tantalum crucible stopped glowing, and the furnace power was turned on. The temperature was then increased to {approximately}1200 C to melt and homogenize the compound, and liquid Li{sub 3}Bi was cast into cold stainless steel molds. Approximately 3.7 kg of Li{sub 3}Bi was generated by this method

Actinide-containing metal disposition alloys

Argonne National Laboratory is currently developing an electro-metallurgical process for treating a wide array of spent nuclear fuels. As part of this process, two waste streams will be consolidated into waste forms; one will be a mineral and the other a metal alloy. The metal waste form is an alloy that contains cladding hulls, ``noble`` metal fission products, and Zr from alloy fuels. The nominal composition of the metal waste form alloys are stainless steel-15 wt.% Zr (SS-15Zr) for stainless steel clad fuel and Zircaloy-8 wt.% stainless steel (Zr-8SS) for Zircaloy clad fuel, with both alloys also containing up to 4 wt.% noble metal fission products. This paper investigates using the two nominal metal alloy compositions described above as a possible Pu and TRU disposition form

Fundamental studies of ceramic/metal interfacial reactions at elevated temperatures.

This work characterizes the interfaces resulting from exposing oxide and non-oxide ceramic substrates to zirconium metal and stainless steel-zirconium containing alloys. The ceramic/metal systems together were preheated at about 600 C and then the temperatures were increased to the test maximum temperature, which exceeded 1800 C, in an atmosphere of high purity argon. Metal samples were placed onto ceramic substrates, and the system was heated to elevated temperatures past the melting point of the metallic specimen. After a short stay at the peak temperature, the system was cooled to room temperature and examined. The chemical changes across the interface and other microstructural developments were analyzed with energy dispersive spectroscopy (EDS). This paper reports on the condition of the interfaces in the different systems studied and describes possible mechanisms influencing the microstructure

Interaction phenomena at reactive metal/ceramic interfaces.

The objective of this study was to understand the interface chemical reactions between stable ceramics and reactive liquid metals, and developing microstructure. Experiments were conducted at elevated temperatures where small metal samples of Zr and Zr-alloy were placed on top of selected oxide and non-oxide ceramic substrates (Y{sub 2}O{sub 3}, ZrN, ZrC, and HfC). The sample stage was heated in high-purity argon to about 2000 C, held in most cases for five minutes at the peak temperature, and then cooled to room temperature at {approximately}20 c/min. An external video camera was used to monitor the in-situ wetting and interface reactions. Post-test examinations of the systems were conducted by scanning electron microscopy and energy dispersive spectroscopy. It was determined that the Zr and the Zr-alloy are very active in the wetting of stable ceramics at elevated temperatures. In addition, in some systems, such as Zr/ZrN, a reactive transition phase formed between the ceramic and the metal. In other systems, such as Zr/Y{sub 2}O{sub 3}, Zr/ZrC and Zr/HfC, no reaction products formed, but a continuous and strong joint developed under these circumstances also