scholarly journals Direct Conversion of Halogen-Containing Wastes to Borosilicate Glass

1996 ◽  
Vol 465 ◽  
Author(s):  
C. W. Forsberg ◽  
E. C. Beahm ◽  
J. C. Rudolph

ABSTRACTGlass has become a preferred waste form worldwide for radioactive wastes; however, there are limitations. Halogen-containing wastes can not be converted to glass because halogens (chlorides, fluorides, etc.) form poor-quality waste glasses. Furthermore, halides in glass melters often form second phases that create operating problems. A new waste vitrification process, the Glass Material Oxidation and Dissolution System (GMODS), removes these limitations by converting halogen-containing wastes into borosilicate glass and a secondary, clean, sodium-halide stream.

1981 ◽  
Vol 6 ◽  
Author(s):  
A. Briggs ◽  
D.V.C. Jones ◽  
G.B. Cole

ABSTRACTA possible method of treatment for Magnox cladding waste is by dissolution in nitric acid and precipitation of barium sulphate-based floc with which radioactive ions are co-precipitated. The floc could then be immobilised in a matrix material such as cement or bitumen to give the waste form, or alternatively can be converted directly into a waste form by hot pressing.This paper describes the direct conversion of barium sulphate floc, containing simulated radwaste, into a synthetic, ceramic version of the natural mineral barite by a hot-pressing route. By variation of the parameters pressure, temperature and time, optimum conditions for consolidation of the floc to > 90% theoretical density on a laboratory scale are found to be 22.5 MPa, 900°C for 10 minutes. Using a pressure of 15 MPa, at 900°C for 30 min., hot-pressed billets of BaSO4 have been made on a 5 kg scale. In going from the Magnox waste to the hot-pressed barium sulphate a volume reduction factor ∼ 18 is achieved. The principal phases in the product are found to be BaSO4 , MgO and Fe3O4, and the degree of consolidation achieved depends on the MgO content.The leaching behaviour of the hot-pressed materials in 100°C, 3 day Soxhlet tests also depends on the MgO content, and on the consequent level of open porosity. If there is porosity accessible to the leach water, MgO at the internal surfaces is converted to Mg(OH)2, which deposits within the pores, and a weight gain is registered in the Soxhlet test. If, however, there is no open porosity, a weight loss occurs, and leach rates ∼ 4 × 10−7 kg/m2/sec are found. In contrast, pure BaSO4, hot-pressed to similar densities, shows no variation in leaching behaviour over a wide range of o en porosities, and gives Soxhlet leach rates ∼ 8 × 10−8 kg/m2/sec.


1981 ◽  
Vol 6 ◽  
Author(s):  
Clyde J. M. Northrup ◽  
George W. Arnold ◽  
Thomas J. Headley

ABSTRACTThe first observations of physical and chemical changes induced by lead implantation damage and leaching are reported for two proposed U.S. nuclear waste forms (PNL 76–68 borosilicate glass and Sandia titanate ceramics) for commercial wastes. To simulate the effects of recoil nucleii due to alpha decay, the materials were implanted with lead ions at equivalent doses up to approximately 1 × 1019 a decays/cm3 . In the titanate waste form, the zirconolite, perovskite, hollandite, and rutile phases all exhibited a mottled appearance in the transmission electron microscope (TEM) typical of defect clusters in radiation damaged, crystalline solids. One titanate phase containing uranium was found by TEM to be amorphous after implantation at the highest dose. No enhanced leaching (deionized water, room temperature, 24 hours) of the irradiated titanate waste form, including the amorphous phase, was detected by TEM, but Rutherford backscattering (RBS) suggested a loss of cesium and calcium after 21 hours of leaching. The RBS spectra also indicated enhanced leaching from the PNL 76–68 borosilicate glass after implantation with lead ions, in general agreement with the observations of Dran, et al. [6,7] on other irradiated materials. Elastic recoil detection spectroscopy (ERD), used to profile hydrogen after leaching, showed penetration of the hydrogen to several thousand angstroms for both the implanted and unimplanted materials. These basic studies identified techniques to follow the changes that occur on implantation and leaching of complex amorphous and crystalline waste forms. These studies were not designed to produce comparisons between waste forms of gross leach rates.


1986 ◽  
Vol 84 ◽  
Author(s):  
Rodney C. Ewing ◽  
Michael J. Jercinovic

AbstractOne of the unique and scientifically most difficult aspects of nuclear waste isolation is the extrapolation ofshot-term laboratory data (hours to years) to the long time periods (103-105 years) required by regulatory agencies for performance assessment. The direct verification of these extrapolations is not possible, but methods must be developed to demonstrate compliance with government regulations and to satisfy the lay public that there is a demonstrable and reasonable basis for accepting the long-term extrapolations. Natural analogues of both the repository environment (e.g. radionuclide migration at Oklo) and nuclear waste form behavior (e.g. alteration of basaltic glasses and radiation damage in minerals) have been used to demonstrate the long-term behavior of large scale geologic systems and, on a smaller scale, waste form durability. This paper reviews the use of natural analogues to predict the long-term behavior of nuclear waste form glasses. Particular emphasis is placed on the inherent limitations of any conclusions that are based on “proof” by analogy. An example -- corrosion of borosilicate glass -- is discussed in detail with specific attention to the proper and successful use of natural analogues (basaltic glass) in understanding the long-term corrosion behavior of borosilicate glass.


1987 ◽  
Vol 112 ◽  
Author(s):  
P. L. Chambré ◽  
C. H. Kang ◽  
W. W.-L. Lee ◽  
T. H. Pigford

AbstractThe dissolution rate of waste solids in a geologic repository is a complex function of waste form geometry, chemical reaction rate, exterior flow field, and chemical environment. We present here an analysis to determine the steady-state mass transfer rate, over the entire range of flow conditions relevant to geologic disposal of nuclear waste. The equations for steady-state mass transfer with a chemical-reaction-rate boundary condition are solved by three different mathematical techniques which supplement each other. This theory is illustrated with laboratory leach data for borosilicate-glass and a spherical spent-fuel waste form under typical repository conditions. For borosilicate glass waste in the temperature range of 57°C to 250°C, dissolution rate in a repository is determined for a wide range of chemical reaction rates and for Peclet numbers from zero to well over 100, far beyond any Peclet values expected in a repository. Spent-fuel dissolution in a repository is also investigated, based on the limited leach data now available.


2009 ◽  
Vol 1193 ◽  
Author(s):  
Barbara F. Dunnett ◽  
Nick R. Gribble ◽  
Andrew D. Riley ◽  
Carl J. Steele

AbstractSellafield Ltd operates a Waste Vitrification Plant (WVP) to immobilise the arisings from the reprocessing of spent nuclear fuel. Washout of solids from the base of waste storage tanks in preparation for decommissioning is likely to produce feeds enriched in molybdenum to the WVP. Vitrification of such feeds in the borosilicate glass formulation currently used by the WVP for vitrification of reprocessing waste has been investigated to determine the maximum achievable loading of MoO3.The vitrification of molybdenum in the absence and presence of reprocessing waste was studied. A number of glasses were manufactured in the laboratory containing various waste loadings. The resultant glasses were examined both visually and under the scanning electron microscope for the presence of any phase separation. Additional aluminium was added to the glasses manufactured in the absence of reprocessing waste to improve the durability of the glass. In borosilicate glass containing 3.5 wt% Al2O3 the onset of a molybdenum phase separation was observed in glasses containing 2.6 wt% MoO3. In the presence of Magnox reprocessing waste, phase separation was observed when the product contained >3.8 wt% MoO3. Soxhlet durability testing of a selection of the glasses manufactured was carried out. The Soxhlet durability of glasses in the absence of phase separation was good.


2022 ◽  
Vol 578 ◽  
pp. 121352
Author(s):  
Ryuhei Motokawa ◽  
Koji Kaneko ◽  
Yojiro Oba ◽  
Takayuki Nagai ◽  
Yoshihiro Okamoto ◽  
...  

1981 ◽  
Vol 6 ◽  
Author(s):  
G. Bandyopadhyay

ABSTRACTSeveral simulated interim waste forms have been investigated in the laboratory to study their suitability for application in handling and transportation of high-level radioactive wastes to terminal processing sites. In the fused-salt/sludge option, the neutralized supernatant liquid and the precipitated sludge are treated simultaneously to form fused-salt cakes. Silicate-based options, in which sodium silicate or sodium silicate and Ca(OH)2 act as binders for the sludge, require prior separation of the sludge and the soluble radioactive constituents from the supernatant before the waste form can be prepared. The results from tests on simulated fused-salt waste forms indicated that the process simplicity of this option is partially offset by the high water solubility and hygroscopicity of the product, which would necessitate special precautions during transportation and storage. The most promising silicate-based option is the ambienttemperature silicate sludge process, in which the sludge is mixed with sodium silicate [and sometimes with Ca(OH)2] and subsequently exposed to a contrelled-humidity environment at room temperature to form a chemical bond. Solid material containing 75 wt % synthetic calcined sludge, prepared by this process, has sufficient physical, chemical, and mechanical stability for use as an interim waste form.


1999 ◽  
Vol 556 ◽  
Author(s):  
M. Nyman ◽  
T. M. Nenoff ◽  
Y. Su ◽  
M. L. Balmer ◽  
A. Navrotsky ◽  
...  

AbstractThe radioactivity of the Hanford site waste tanks is primarily from 137Cs and 90Sr, of which can both be selectively removed from solution using a crystalline silicotitanate (CST) ion exchanger. We are currently seeking waste forms alternative to borosilicate glass for Cs-CSTs. In order to obtain a fundamental basis for the development of an alternative waste form, we are investigating synthesis and characterization of CST component phases, namely Cs-Si-Ti-O phases. Two novel Cs-Ti-Si-O phases (one porous, one condensed) have been hydrothermally synthesized, characterized and evaluated as waste form candidates based on chemical and thermal stability, leachability, and ion exchange capabilities.


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