The Effect of Temperature on the Redox Constraints for the Processing of High-Level Nuclear Waste into a Glass Waste Form

1989 ◽  
Vol 176 ◽  
Author(s):  
Henry D. Schreiber ◽  
Charlotte W. Schreiber ◽  
Margaret W. Riethmiller ◽  
J. Sloan Downey

ABSTRACTThe oxidation-reduction equilibria of selected multivalent elements in an alkali borosilicate glass melt (Savannah River Laboratory frit #131) were measured as a function of the imposed oxygen fugacity over the temperature range from 950°C to 1350°C. Redox constraints on the processing of high-level nuclear waste into the glass melt require that the prevailing oxygen fugacity be about 10−5 to 10−12 Zatm at 950°C, about 10−2 to 10−9 atm at 1150°C, and about 100 to 10−7 atm at 1350°C. Such conditions circumvent foaming under oxidizing situations and metal/sulfide precipitation if the system becomes too reducing. The defined oxygen fugacity ranges correspond to the previously prescribed range of 0.1 to 0.5 for the [Fe2+]/[Fe3+] ratio in the resulting glass, independent of the processing temperature from 950°C to 1350°C.

MRS Bulletin ◽  
1987 ◽  
Vol 12 (5) ◽  
pp. 61-65 ◽  
Author(s):  
M.J. Plodinec

At the Savannah River Plant (SRP), construction of what will be the world's largest solidification facility for nuclear waste has been under way since 1983. Beginning in 1990, the nearly 100 million liters of liquid high-level nuclear waste now stored on the site will be made into a durable borosilicate glass in this Defense Waste Processing Facility (DWPF).In developing a slurry-fed melting process for the DWPF, we made advances in understanding both glass processing and glass durability. This article focuses on what we learned and what further advances are likely to be made.Generally speaking, the goal of any glass technologist is to make a good glass and to make it well. In the glass industry a good product is whatever people will buy. To make it well means, above all, to make the product as economically as possible. Thus, the commercial glass technologist will control the composition of the melter feed material very closely to ensure that only the components necessary for glass performance are included, and in the least expensive form possible. The commercial glass technologist may also tolerate low yields or specify several stages of post-melt processing if it is necessary to produce a product to demanding specifications.To the nuclear waste glass technologist, however, a good product is one which will be stable in geologic environments for millions of years.


1999 ◽  
Vol 556 ◽  
Author(s):  
H. Gan ◽  
A. C. Buechele ◽  
C.-W. Kim ◽  
X. Huang ◽  
R. K. Mohr ◽  
...  

AbstractInconel-690, a Cr-Ni-Fe-based “superalloy,” has become the material of choice for electrodes in joule-heated waste glass melters and is currently employed in the high-level nuclear waste vitrification systems at West Valley and DWPF, as well as in GTS Duratek's privatized M-Area mixed waste vitrification facility at Savannah River. Future applications of joule-heated vitrification technologies will necessitate an assessment of the limits of performance of this material under more demanding conditions than have been studied previously. In this work, Inconel 690 electrodes were tested in several simulated sodium-rich aluminosilicate waste glasses in wide ranges of AC current density, electrical waveform, temperature, and glass composition.


2017 ◽  
Vol 27 (6) ◽  
pp. 1431-1438 ◽  
Author(s):  
De-cheng KONG ◽  
Chao-fang DONG ◽  
Kui XIAO ◽  
Xiao-gang LI

Author(s):  
D. T. Hobbs ◽  
T. B. Peters ◽  
M. C. Duff ◽  
M. J. Barnes ◽  
S. D. Fink ◽  
...  

A significant fraction of the high-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) must be pretreated to remove 137Cs, 90Sr and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. Separation processes planned at the SRS include caustic side solvent extraction for 137Cs and sorption onto monosodium titanate (MST) for 90Sr and alpha-emitters. The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes 238Pu, 239Pu and 240Pu. This paper describes the planned Sr/actinide separation process and summarizes recent tests and demonstrations with simulated and actual tank waste solutions.


2013 ◽  
Vol 3 (1) ◽  
pp. 60-69 ◽  
Author(s):  
Hamid Aït Abderrahim ◽  
Didier De Bruyn ◽  
Gert Van den Eynde ◽  
Sidney Michiels

2021 ◽  
pp. 153423
Author(s):  
José Marcial ◽  
Jaroslav Kloužek ◽  
Miroslava Vernerová ◽  
Pavel Ferkl ◽  
SeungMin Lee ◽  
...  

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