Effects of Waste Composition and Loading on the Chemical Durability of a Borosilicate Glass

1981 ◽  
Vol 11 ◽  
Author(s):  
D.E. Clark ◽  
C.A. Maurer ◽  
A.R. Jurgensen ◽  
L. Urwongse
Keyword(s):  
Corrosion Resistance ◽  
Borosilicate Glass ◽  
Chemical Durability ◽  
Deionized Water ◽  
High Alumina ◽  
Savannah River ◽  
Corrosion Performance ◽  
Waste Composition ◽  
Savannah River Plant ◽  
High Level

ABSTRACTThe effects of waste composition and percent loading in a borosilicate glass designed for US defense high level wastes (HLW) have been evaluated. Three types of simulated wastes were investigated; high alumina, high iron and a composite representative of an average waste composition from Savannah River Plant (SRP) waste tanks. Corrosion resistance of the borosilicate glass is significantly enhanced by the presence of any of the three types of wastes. Additionally, corrosion resistance is improved as the % waste loading is increased in the glass. The best corrosion performance was obtained with the high alumina waste in deionized water.

Effects of Alpha, Gamma, and Alpha-Recoil Radiation on Borosilicate Glass Containing Savannah River Plant Defense High-Level Nuclear Waste

1981 ◽  
Vol 6 ◽  
Author(s):  
Ned E. Bibler
Keyword(s):  
Borosilicate Glass ◽  
Deionized Water ◽  
Savannah River ◽  
Γ Radiation ◽  
Leach Rate ◽  
Geologic Storage ◽  
Savannah River Plant ◽  
Pb Ions ◽  
High Level ◽  
Doped Glass

ABSTRACTAt the Savannah River Plant, the reference process for the immobilization of defense high-level waste (DHLW) for geologic storage is vitrification into borosilicate glass. During geologic storage for 106y, the glass would be exposed to ∼3 × 1010 rad of β radiation, ∼1010 rad of γ radiation, and 1018 particles/g glass for both α and α-recoil radiation. This paper discusses tests of the effect of these radiations on the leachability and density of the glass. No effect of the radiations was detected that reduced the effectiveness of the glass for long-term storage of DHLW even at doses corresponding to 106 years storage for the actual glass. For the tests, glass containing simulated DHLW was prepared from frit of the reference composition. Three methods were used to irradiate the glass: external irradiations with beams of ∼200 keV or Pb ions, internal irradiations with Cm–244 doped glass, and external irradiations with Co–60 γ rays. Results with both Xe and Pb ions indicate that a dose of 3 × 1013 ions/cm2 (simulating >106 years storage) does not significantly increase the leachability of the glass in deionized water. Tests with Cm–244 doped glass show no increase in leach rate in deionized water up to a dose of 1.3 × 1018 α and α-recoils/g glass. The density of the Cm–244 doped glass has decreased by 1% at a dose of 1018 particles/g glass. With γ-radiation, the density has changed by <0.05% at a dose of 8.5 × 1010 rad. Results of leach tests in deionized water and brine indicated that this very large dose of γ-radiation increased the leach rate by only 20%. Also, the leach rates are 3 to 4 times lower in brine.

An Experimental Comparison of Alternative Solid Forms for Savannah River High-Level Wastes

1981 ◽  
Vol 6 ◽  
Author(s):  
John A. Stone
Keyword(s):  
Borosilicate Glass ◽  
Silica Glass ◽  
Deionized Water ◽  
Experimental Comparison ◽  
Savannah River ◽  
Waste Forms ◽  
High Silica ◽  
High Level ◽  
Selection Of ◽  
High Aluminum

ABSTRACTSamples of borosilicate glass, high-silica glass, tailored ceramic, and SYNROC, incorporating simulated Savannah River high-level defense waste sludges, were leached by the MCC-1 procedure for times up to 28 days. Cesium, uranium, and cerium leach rates are reported for waste forms containing a composite sludge, at 40°C in deionized water, and at 90°C in deionized water, silicate water, and brine. The ordering of the waste forms from best to worst differs for each element leached, and none of the forms show a clear advantage for all the key radwaste elements. Some cesium leach rates for forms containing high-aluminum or high-iron sludges also are presented. So far, only small effects of sludge type have been observed, with one exception. This study is one of several inputs for selection of an alternative waste form for Savannah River waste.

scholarly journals High-Level Radioactive Insoluble Waste Preparation for Vitrification

1984 ◽  
Vol 44 ◽  
Author(s):  
B. A. Hamm ◽  
R. E. Eibling ◽  
M. A. Ebra ◽  
T. Motyka ◽  
H. D. Martin
Keyword(s):  
Radioactive Waste ◽  
Borosilicate Glass ◽  
Savannah River ◽  
Soluble Salts ◽  
High Level Radioactive Waste ◽  
Savannah River Plant ◽  
High Level

AbstractAt the Savannah River Plant (SRP), a process has been developed for immobilizing high-level radioactive waste in a borosilicate glass. The waste is currently stored as soluble salts and insoluble solids. Insoluble waste as stored requires further processing before vitrification is possible. The processes required have been developed and demonstrated with actual waste. They include removal of aluminum in some waste, washing soluble salts out of the insoluble waste, and mercury stripping. Each of the processes and the results with actual SRP waste will be discussed. The benefits of each step will also be included.

Mercury Reduction and Removal During High-Level Radioactive Waste Processing and Vitrification

1981 ◽  
Vol 6 ◽  
Author(s):  
Russell E. Eibling ◽  
John R. fowler
Keyword(s):  
Radioactive Waste ◽  
Formic Acid ◽  
Borosilicate Glass ◽  
Processing Temperature ◽  
Savannah River ◽  
High Level Radioactive Waste ◽  
Savannah River Plant ◽  
Reference Process ◽  
Glass Melter ◽  
High Level

ABSTRACTA reference process for immobilizing the high-level radioactive waste in borosilicate glass has been developed at the Savannah River Plant. This waste contains a substantial amount of mercury from separations processing. Because mercury will not remain in borosilicate glass at the processing temperature, mercury must be removed before vitrification or must be handled in the off-gas system. A process has been developed to remove mercury by reduction with formic acid prior to vitrification. Additional benefits of formic acid treatment include improved sludge handling and glass melter redox control.

Up Close: Immobilization of High-Level Nuclear Waste at the Savannah River Plant

MRS Bulletin ◽  
1987 ◽  
Vol 12 (5) ◽  
pp. 61-65 ◽  
Author(s):  
M.J. Plodinec
Keyword(s):  
Nuclear Waste ◽  
Melting Process ◽  
Melt Processing ◽  
Savannah River ◽  
Good Product ◽  
Glass Processing ◽  
Savannah River Plant ◽  
Commercial Glass ◽  
High Level Nuclear Waste ◽  
High Level

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.

Evaluation of Cesium Sujcotitanates as an Alternative Waste Form

1996 ◽  
Vol 465 ◽  
Author(s):  
Yali Su ◽  
M. Lou Balmer ◽  
Bruce C. Bunker
Keyword(s):  
Borosilicate Glass ◽  
Thermal Conversion ◽  
Chemical Durability ◽  
Waste Form ◽  
High Level Waste ◽  
Ion Exchangers ◽  
Heat Treated ◽  
High Level ◽  
Excellent Chemical ◽  
Better Than

ABSTRACTSilicotitanate ion exchangers are potential materials for the removal of radioactive Cs and Sr from tank wastes. In this paper the viability of direct thermal conversion of Cs-loaded silicotitanates to an acceptable high level waste form has been examined. Results show that in aqueous solutions, the Cs leach rates of crystalline silicotitanates (heat treated at 800°C) are 0.04, 0.18, 0.4 g/m2day for Cs loadings of 1, 5, and 20 wt%, respectively. Heating the Cs-loaded (up to 20 wt %) silicotitanates at or above 900 °C for 1 hour further reduces the Cs leach rates to approximately zero (beyond the lppm detection limits). Moreover, Cs volatilization was found to be < 0.8 wt% at temperatures as high as 1000 °C. These results suggest that thermally converted silicotitanate ion exchangers exhibit excellent chemical durability (comparable to or better than borosilicate glass) and thus, have great potential as an alternative waste form.

Study of Melted Synroc Doped with Simulated High-Level Waste

1996 ◽  
Vol 465 ◽  
Author(s):  
I. A. Sobolev ◽  
S. V. Stefanovsky ◽  
S. V. Ioudintsev ◽  
B. S. Nikonov ◽  
B. I. Omelianenko ◽  
...  
Keyword(s):  
Bulk Composition ◽  
Barometric Pressure ◽  
High Level Waste ◽  
Savannah River ◽  
Savannah River Plant ◽  
High Level ◽  
A Minor ◽  
Simulated High Level Waste ◽  
Additional Phase

ABSTRACTPreparation and characterization of inductively-melted Synroc containing 20 wt% simulated plant “Mayak” reprocessing waste were performed. The sample bulk composition was as follows, (in wt.%): 55.4 TiO2; 15.8 ZrO2; 7.5 CaO; 7.4 BaO; 4.3 Al2O3 2.0 MnO; 1.8 SiO2; 0.7 Na2O; 1.9 K2O, 0.5 Ce2O3; 1.0 UO2; 0.9 NiO; 0.6 Cr2O3, and 0.2 FeO. The sample was produced by melting in air at 1550–1600 °C under barometric pressure. It is composed of a few crystalline phases and a minor glass phase. Most of the phases (hollandite, zirconolite, perovskite and rutile) are similar to the analogous phases found in the other Synroc formulations. An additional phase with average composition, wt.%: 59.8 TiO2; 15.6 CaO; 7.0 UO2; 5.6 ZrO2; 4.7 MnO; 4.1 Ce2O3, and 1.8 Al2O3 was found. Some elements (Ba, Si, Ni, K, Na, Fe) were present in the phase in negligible quantities. Its formula (Ca2.65U0.3Ce0.2)(Ti7.3Mn0.6Zr0.4Al0.3)O20.0 is rather close to a rare mineral uhligite - Ca3(Ti,Zr,Al)9O20. Another possible counterpart of the phase is murataite-like mineral previously described in tailored ceramic designed for Savannah River Plant wastes fixation. This phase as well as zirconolite are the major host for U in the sample Preliminary data on the material leachability in water at 350 °C and 50 MPa have been obtained Uranium contents in the solution were about 1 ppb and close to the uranium dioxide solubility in deionized water under the same P-T conditions.

Titanate Ceramic Matrices for Alumina-Rich Wastes

2003 ◽  
Vol 807 ◽  
Author(s):  
Bruce D. Begg ◽  
Eric R. Vance ◽  
Huijun Li ◽  
Terry McLeod ◽  
Nicholas Scales ◽  
...  
Keyword(s):  
Savannah River ◽  
Fuel Cladding ◽  
Process Flexibility ◽  
Magnesia Spinel ◽  
Savannah River Plant ◽  
Waste Loading ◽  
High Level ◽  
Sodium Incorporation ◽  
Titanate Ceramic ◽  
Ceramic Matrices

ABSTRACTIn the early 1980s a synroc variant, SYNROC-D, was developed for immobilisation of high-level defence waste stored at the Savannah River Plant, USA. A key phase in the immobilisation matrix was spinel, used to immobilise the large proportion of iron and alumina in the waste. Here we examine the feasibility of this approach for other alumina-rich wastes, not necessarily containing iron, derived from the dissolution of aluminium fuel cladding. The advantages of using a magnesia spinel, as opposed to hercynite (FeAl2O4), as the primary alumina-bearing phase are discussed in terms of an increase in waste loading and process flexibility. Two options for sodium incorporation, glass and the titanate phase freudenbergite, are considered.

Influence of Calcium Addition on the Chemical Durability of the Model Alumino-Silicate Glasses in Aqueous Solutions

2012 ◽  
Vol 512-515 ◽  
pp. 1009-1014
Author(s):  
Dagmar Galusková ◽  
Miroslav Hnatko ◽  
Jozef Kraxner ◽  
Dušan Galusek ◽  
Pavol Šajgalík
Keyword(s):  
Corrosion Resistance ◽  
Grain Boundary ◽  
Aqueous Solutions ◽  
Chemical Durability ◽  
Deionized Water ◽  
Silicate Glasses ◽  
Hydrothermal Conditions ◽  
Aqueous Environments ◽  
Alumino Silicate ◽  
Static Conditions

The corrosion resistance of liquid phase sintered (LPS) alumina ceramics in aqueous environments strongly depends on composition and chemistry of grain boundary glass formed during sintering. The chemical durability of model alumino-silicate glasses with various contents of CaO in aqueous solutions was therefore evaluated. Prepared glasses were corroded under hydrothermal conditions in deionized water under static conditions. The examination of surface morphology of corroded specimens after the contact with deionized water, together with the analysis of corrosion solution provided information on mechanism of dissolution of grain boundary glasses in LPS aluminas and confirmed that dissolution process is hindered due to saturation of solution with respect to leached elements. The initial dissolution rates for studied glasses were determined. The results are applicable for optimization and enhancement of corrosion resistance of LPS alumina under hydrothermal conditions.

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