A Zirconium Cermet for Transuranic Isotope Storage and Burning
- Fuel Cycle and Materials Laboratory, Texas A and M University, College Station, TX 77843 (United States)
Transuranic (TRU) isotopes comprise {approx} 1% of the spent nuclear fuel (SNF) inventory from commercial light water reactors and supply the majority of the late-term (>300 years) heat load produced by SNF. These isotopes also have significant potential fission energy content that could be recovered in a fast reactor or accelerator driven system. Estimates by Wigeland, et al. suggest that removal of 99.9% of TRU isotopes and 97% of cesium, strontium, and europium from SNF could increase the linear drift load (i.e. capacity) of the proposed geologic repository at Yucca Mountain, NV by an order of magnitude. In other words, the waste packing capacity in Yucca Mt. would be increased and the recovered TRU could be used in a fast reactor system for energy production. One of the goals of the Advanced Fuel Cycle Initiative (AFCI) of the U.S. Department of Energy is to develop advanced nuclear technologies to recycle spent nuclear fuel without separating pure plutonium. The UREX+ family of solvent extraction and ion exchange processes can be manipulated to isolate TRU isotopes from uranium and other short-lived isotopes while leaving the Pu intimately mixed with other minor actinides (this corresponds with the UREX+1a process). The product TRU nitrate solution can be converted into oxide microspheres via direct denitration or sol-gel processing, leaving the TRU product in a stable oxide form for further processing. In addition, nuclear-grade zirconium metal can be recovered by recycling Zircaloy from the spent fuel cladding after the fuel has been chopped and dissolved away. The Zircaloy cladding hulls may be processed via a hydride/de-hydride process to produce high purity zirconium powder.These powder products can be combined via powder metallurgy techniques to produce a stable, safe storage form for TRU waste with the possibility of converting a long term waste liability into an advanced energy resource. Previous work by McDeavitt et al. at Argonne National Laboratory analyzed the production of a (Th,U)O-2-Zr cermet fuel using a cold-drawing process. This was a multi-pass process wherein a series of reductions were performed on each individual rod with a high temperature anneal between successive reductions to relieve strain hardening. This process allowed excellent control over the finished product matrix density allowing the user to pre-select the desired density by manipulating the reduction scheme. The final cold-drawn cermet contained a large fraction of damaged oxide particles, though the cause of this damage was not investigated due to project time constraints. It is believed that the high stress profile developed during cold deformation was the primary cause of this damage. The Fuel Cycle and Materials Laboratory (FCML) at Texas A and M University is currently analyzing a hot extrusion process to fabricate the proposed cermet and provide comparison data with the previous work. Initial studies have been completed and a system redesign is near completion. This paper presents the initial findings of the initial work and describes the new system setup. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 20979596
- Resource Relation:
- Conference: Advanced nuclear fuel cycles and systems (GLOBAL 2007), Boise - Idaho (United States), 9-13 Sep 2007; Other Information: Country of input: France; 13 refs; Related Information: In: Proceedings of GLOBAL 2007 conference on advanced nuclear fuel cycles and systems, 1873 pages.
- Country of Publication:
- United States
- Language:
- English
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