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Research and development activities waste fixation program. Quarterly progress report, October--December 1973

Author
Platt, A. M.
Published
United States : [publisher not identified], 1974.
[Oak Ridge, Tennessee] : [U.S. Atomic Energy Commission], 1974.
Physical Description
microfiche : negative ; 11 x 15 cm

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Summary
Two 40-hr nonradioactive spray solidification runs were completed in which both PU-4b and PW-6 reference waste compositions were successfully converted into borosilicate glass. The throughput capacity of the 13-in. diameter spray calciner while processing PW-4b waste is over 38 1/hr with a 700 deg C calciner wall and 24-30 1/hr with a 550 deg C wall. Mechanical agitation increases the capacity of the inconel meiter to over 4.5 l of melt per hour and greatly improves product homogeneity. Both PW-4b and PW-6 reference waste compositions were successfully concentrated in the nonradioactive, wiped film evaporator (WFE) facility. Either waste can be concentrated to approximately 60 wt% total solids in the WFE. For the PW-6 waste, this represents removal of about 80% of the initial liquid present in the waste. Two ceramic melters were built and operated, each using molybdenum electrodes for heating. While the performance of these first melters indicated a number of design and construction problems which need to be resolved, a ceramic-electrode heated melter appears to be a viable alternative to metallic melters. The PW-4b melt composition 72-68 contains finely divided crystallites when large quantities are melted at 1150 deg C. The crystallites were identified as CeO/sub 2/, a spinel and zircon, with CeO/ sub 2/ the predominant species. Several options for eliminating the crystallites include increasing the melt temperature, agitating the melt to prevent crystal settling, modification of the composition to promote CeO/sub 2/ solubility, and reducing the concentration of CeO/sub 2/ in the glass. Helium gas formed by alpha particle emission from actinide nuclides in high-level waste glasses can produce large internal stresses in the glass. A porosity of 0.2% in typical waste glasses, from uranium oxide fuel, may be sufficient to reduce the internal stresses to inconsequential levels by providing sites for gas accumulation. However, for waste glasses from plutonium recycle fuel, a higher than anticipated porosity of approximately 3% may be required. Mass spectrographic analysls was performed on the gas phase from several canisters of solidified radioactive waste from the previous Waste Solidification Engineering Prototypes (WSEP) program. Analysis of the gas in several phosphate glass canisters revealed CO/sub 2/, present from the decomposition of residual Purex solvent and silicone antifoam agent, and N/sub 2/O from the decomposition of nitrate. Hydrogen was below detection limits. Oxygen depletion of the original air atmosphere was the only change found in limited sampling of spray solidified and pot calcine product canisters. (auth)
Report Numbers
BNWL-1809
Other Subject(s)
Collection
U.S. Atomic Energy Commission depository collection.
Note
DOE contract number: AT(45-1)-1830
NSA number: NSA-29-024091
OSTI Identifier 4338386
Research organization: Battelle Pacific Northwest Labs., Richland, WA (United States).
Research organization: comp.
Funding Information
Sponsored by USDOE.
View MARC record | catkey: 42542129