The Advanced Fuel Cycle Initiative (AFCI) is a Department of Energy (DOE)
program, that has been investigating technologies to improve fuel cycle sustainability
and proliferation resistance. One of the program's goals is to reduce the amount of
radioactive waste requiring repository disposal.
Cesium and strontium are two primary heat sources during the first 300 years
of spent nuclear fuel's decay, specifically isotopes Cs-137 and Sr-90. Removal of
these isotopes from spent nuclear fuel will reduce the activity of the bulk spent
fuel, reducing the heat given off by the waste. Once the cesium and strontium are
separated from the bulk of the spent nuclear fuel, the isotopes must be immobilized.
This study is focused on a method to immobilize a cesium- and strontium-bearing
radioactive liquid waste stream. While there are various schemes to remove these
isotopes from spent fuel, this study has focused on a nitric acid based liquid waste.
The waste liquid was mixed with the bentonite, dried then sintered. To be effective
sintering temperatures from 1100 to 1200 degrees C were required, and waste concentrations
must be at least 25 wt%. The product is a leach resistant ceramic solid with the
waste elements embedded within alumino-silicates and a silicon rich phase. The
cesium is primarily incorporated into pollucite and the strontium into a monoclinic
feldspar.
The simulated waste was prepared from nitrate salts of stable ions. These ions
were limited to cesium, strontium, barium and rubidium. Barium and rubidium will
be co-extracted during separation due to similar chemical properties to cesium and
strontium. The waste liquid was added to the bentonite clay incrementally with
drying steps between each addition. The dry powder was pressed and then sintered
at various temperatures. The maximum loading tested is 32 wt. percent waste,
which refers to 13.9 wt. percent cesium, 12.2 wt. percent barium, 4.1 wt. percent
strontium, and 2.0 wt. percent rubidium. Lower loadings of waste were also tested.
The final solid product was a hard dense ceramic with a density that varied from
2.12 g/cm3 for a 19% waste loading with a 1200 degrees C sintering temperature to 3.03
g/cm3 with a 29% waste loading and sintered at 1100 degrees C.
Differential Scanning Calorimetry and Thermal Gravimetric Analysis (DSC-TGA)
of the loaded bentonite displayed mass loss steps which were consistent with water
losses in pure bentonite. Water losses were complete after dehydroxylation at ~650 degrees C.
No mass losses were evident beyond the dehydroxylation. The ceramic melts at temperatures
greater than 1300 degrees C.
Light flash analysis found heat capacities of the ceramic to be comparable to
those of strontium and barium feldspars as well as pollucite. Thermal conductivity
improved with higher sintering temperatures, attributed to lower porosity. Porosity
was minimized in 1200 degrees C sinterings. Ceramics with waste loadings less than 25
wt% displayed slump, the lowest waste loading, 15 wt% bloated at a 1200 degrees C sintering.
Waste loading above 25 wt% produced smooth uniform ceramics when sintered
>1100 degrees C.
Sintered bentonite may provide a simple alternative to vitrification and other
engineered radioactive waste-forms.
| Identifer | oai:union.ndltd.org:tamu.edu/oai:repository.tamu.edu:1969.1/ETD-TAMU-2009-12-7548 |
| Date | 2009 December 1900 |
| Creators | Ortega, Luis H. |
| Contributors | McDeavitt, Sean M. |
| Source Sets | Texas A and M University |
| Language | English |
| Detected Language | English |
| Type | Book, Thesis, Electronic Dissertation, text |
| Format | application/pdf |
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