2 results
Immobilization of rocky flats graphite fines residues
- T. S. Rudisill, J. C. Marra, D. K. Peeler
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- Journal:
- MRS Online Proceedings Library Archive / Volume 556 / 1999
- Published online by Cambridge University Press:
- 10 February 2011, 231
- Print publication:
- 1999
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- Article
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The Savannah River Technology Center (SRTC) is developing an immobilization process for graphite fines residues generated during nuclear materials production activities at the Rocky Flats Environmental Technology Site (Rocky Flats). The continued storage of this material has been identified as an item of concern. The residue was generated during the cleaning of graphite casting molds and potentially contains reactive plutonium metal. The average residue composition is 73 wt% graphite, 15 wt% calcium fluoride (CaF2), and 12 wt% plutonium oxide (PuO2 ). Approximately 950 kg of this material are currently stored at Rocky Flats.
The strategy of the immobilization process is to microencapsulate the residue by mixing with a sodium borosilicate (NBS) glass frit and heating at nominally 700°C. The resulting waste form would be sent to the Waste Isolation Pilot Plant (WIPP) for disposal. Since the PuO2 concentration in the residue averages 12 wt%, the immobilization process was required to meet the intent of safeguards termination criteria by limiting plutonium recoverability based on a test developed by Rocky Flats. The test required a plutonium recovery of less than 4 g/kg of waste form when a sample was leached using a nitric acid/CaF2 dissolution flowsheet.
Immobilization experiments were performed using simulated graphite fines with cerium oxide (CeO2) as a surrogate for PuO2 and with actual graphite fines residues. Small-scale surrogate experiments demonstrated that a 4:1 frit to residue ratio was adequate to prevent recovery of greater than 4 g/kg of cerium from simulated waste forms. Additional experiments investigated the impact of varying concentrations of CaF2 and the temperature/heating time cycle on the cerium recovery. Optimal processing conditions developed during these experiments were subsequently demonstrated at full-scale with surrogate materials and on a smaller scale using actual graphite fines.
In general, the recovery of cerium from the full-scale waste forms was higher than for smaller scale experiments. The presence of CaF2 also caused a dramatic increase in cerium recovery not seen in the small-scale experiments. However, the results from experiments with actual graphite fines were encouraging. A 4:1 frit to residue ratio, a temperature of 700°C, and a 2 hr heating time produced waste forms with plutonium recoveries of 4±1 g/kg. With an increase in the frit to residue ratio, waste forms fabricated at this scale should meet the Rocky Flats product specification. The scale-up of the waste form fabrication process to nominally 3 kg is expected to require a 5:1 to 6:1 frit to residue ratio and maintaining the waste form centerline temperature at 700°C for 2 hr.
Americium/Curium Extraction from a Lanthanide Borosilicate Glass
- T. S. Rudisill, J. M. Pareizs, W. G. Ramsey
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- Journal:
- MRS Online Proceedings Library Archive / Volume 465 / 1996
- Published online by Cambridge University Press:
- 03 September 2012, 111
- Print publication:
- 1996
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A solution containing kilogram quantities of highly radioactive isotopes of amerícium and curium (Am/Cm) and lanthanide fission products is currently stored in a process tank at the Department of Energy's Savannah River Site (SRS). This tank and its vital support systems are old, subject to deterioration, and prone to possible leakage. For this reason, a program has been initiated to stabilize this material as a lanthanide borosilicate (LBS) glass.1 The Am/Cm has commercial value and is desired for use by the heavy isotope programs at the Oak Ridge National Laboratory (ORNL).
A recovery flowsheet was demonstrated using a curium-containing glass to extract the Am/Cm from the glass matrix. The procedure involved grinding the glass to less than 200 mesh and dissolving in concentrated nitric acid at 110°C. Under these conditions, the dissolution was essentially 100% after 2 hours except for the insoluble silicon. Using a nonradioactive surrogate, the expected glass dissolution rate during Am/Cm recovery was bracketed by using both static and agitated conditions. The measured rates, 0.0082 and 0.040 g/hrcm2, were used to develop a predictive model for the time required to dissolve a spherical glass particle in terms of the glass density, particle size, and measured rate. The calculated dissolution time was in agreement with the experimental observation that the curium glass dissolution was complete in less than 2 hr.