Chemical durability is not a single material property that can be uniquely measured. Instead it is the response to a host of coupled material and environmental processes whose rates are estimated by a combination of theory, experiment, and modeling. High-level nuclear waste (HLW) glass is perhaps the most studied of any material yet there remain significant technical gaps regarding their chemical durability. The phenomena affecting the long-term performance of HLW glasses in their disposal environment include surface reactions, transport properties to and from the reacting glass surface, and ion exchange between the solid glass and the surrounding solution and alteration products. The rates of these processes are strongly influenced and are coupled through the solution chemistry, which is in turn influenced by the reacting glass and also by reaction with the near-field materials and precipitation of alteration products. Therefore, those processes must be understood sufficiently well to estimate or bound the performance of HLW glass in its disposal environment over geologic time-scales. This article summarizes the current state of understanding of surface reactions, transport properties, and ion exchange along with the near-field materials and alteration products influences on solution chemistry and glass reaction rates. Also summarized are the remaining technicalmore » gaps along with recommended approaches to fill those technical gaps.« less

Current Understanding and Remaining Challenges in Modeling Long‐Term Degradation of Borosilicate Nuclear Waste Glasses

The kinetics of matrix dissolution and alkali-exchange for a series of sodium aluminosilicate glass compositions was determined at constant temperature and solution pH(D) under conditions of silica-saturation. Steady state release rate for sodium was 10–50 times faster than the rate of matrix dissolution, demonstrating that alkali exchange is an important long-term reaction mechanism that must be considered when modeling systems near saturation with respect to dissolved glass components. Sodium release rates were 30% slower in D 2 O compared to rates in H 2 O; but matrix dissolution rates were unaffected. These results are consistent with rupture of the OH bond as the rate-limiting reaction in Na + –H + exchange whereas matrix dissolution is controlled by OH − or H 2 O catalyzed hydrolysis of SiOSi and SiOAl bonds. Changes in Na exchange rate with increasing Al 2 O 3 content could not be reconciled with changes in the number of non-bridging oxygen (NBO) sites in the glass alone. A simple model was used to estimate a structural energy barrier for alkali ion exchange using NaO bond length and co-ordination as measured by Na K-edge X-ray absorption spectroscopy, and binding energy shifts for SiONa sites measured by X-ray photoelectron spectroscopy (XPS). The energy barrier was calculated to increase from 34 kJ mol −1 for Na 2 O·2SiO 2 glass to 49 kJ mol −1 for a glass containing 15 mol% Al 2 O 3 , consistent with stronger bonding of Na on NBO sites and increasing mechanical stiffness of the glass network with increasing Al content. The calculated ion-exchange enthalpies were then used to calculate Na ion-exchange rates as a function of glass composition. Agreement between the calculated and measured Na ion exchange rates was excellent.

The structure of Na2O–Al2O3–SiO2 glass: impact on sodium ion exchange in H2O and D2O

Water diffusion in the simulated French nuclear waste glass SON 68 contacting silica rich solutions: Experimental and modeling

Several series of SrTiO3–aluminoborate and SrTiO3–aluminoborosilicate glasses were prepared, which exhibited low corrosion rates in water at 90 °C. The additions of up to 9 mol% P2O5 to these glasses were found to significantly improve the chemical durability. The weight loss rate of a glass with the nominal composition (mol%) 15SrO–15TiO2–23.3Al2O3–46.7B2O3 decreased by over two orders of magnitude with the additions of up to 9 mol% P2O5. In contrast, P2O5-additions increased the dissolution rate of the silicate analog of the initial borate glass, 15SrO–15TiO2–23Al2O3–47SiO2. The positive effect of small P2O5 additions on the chemical durability of the aluminoborate glasses may be attributed to the replacement of BO3 units by BO4 units in the glass network, as indicated by infrared and Raman spectra.

Aluminoborate and aluminoborosilicate glasses with high chemical durability and the effect of P2O5 additions on the properties

Alkali ion exchange in γ-irradiated glasses

The leaching behavior of borate waste glass SL-1

Temperature has strong effects on the leaching behavior of the 90-19/U simulated high-level waste glass form. The Arrhenius equation is used to determine the change in the glass corrosion mechanism...

Temperature Effects on the Leaching Behavior of a High-Level Waste Glass Form

A comparison of HLW-glass and PWR-borate waste glass

Borate waste is the main liquid waste generated by nuclear power plants (NPPs). Vitrification is conceptually attractive because of the potential durability of the final product, the flexibility of the process in treating a wide variety of waste streams, and the economy of large volume reduction. The vitrification of borate waste from NPPs, including the glass formulation and product characterization, is examined. The Minimum Additive Waste Stabilization (MAWS) concept was utilized to design the glass formulation. The glass formulation named SL-1, which can incorporate 45 wt% of waste oxides, was selected. The SL-1 glass has good chemical stability, the melting temperature is 1000°C, and the viscosity of molten glass is ∼5.0 Pa.S at 1000°C. The borosilicate glass form could satisfactorily solidify borate waste with high volume reduction.

Vitrification of Borate Waste Generated by Nuclear Power Plants

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The leaching behavior of borate waste glass SL-1

Temperature Effects on the Leaching Behavior of a High-Level Waste Glass Form

A comparison of HLW-glass and PWR-borate waste glass

Vitrification of Borate Waste Generated by Nuclear Power Plants