Source Trends for Performance Assessment of HLW Glass and Spent Fuel as Waste Forms

Determination of the Forward Dissolution Rate for International Simple Glass in Alkaline Solutions

The International Simple Glass (ISG) is considered as reference benchmark glass and is used in the frame of an international collaboration for the study of the dissolution mechanisms of high-level vitrified nuclear waste.

In this work the forward dissolution rate of the ISG was determined in different alkaline solutions, as a simulation of the disposal conditions foreseen by the Belgian concept for geological disposal of vitrified waste. The determination of the forward dissolution rate was done by measuring the Si released from the glass in solution in tests performed at 30 °C in four different KOH solutions with pH varying from 9 to 14 and in artificial cementitious water at pH 13.5.

The forward dissolution rates determined for the ISG in high pH solutions are in good agreement with the results obtained for a lower pH range.

The rates obtained in this study, moreover, were compared with the rates measured in the same conditions for SON68 glass from a previous work. The values obtained for the two glasses are comparable in artificial cementitious water and in KOH at moderately alkaline pH. At higher pH, ISG glass shows a lower forward dissolution rate with respect to SON68 (0.20 g•m−2•d for ISG and 0.35 g•m−2 d for SON68 at pH 14).

Modeling spent nuclear fuel alteration and radionuclide migration in disposal conditions

A physico-chemical model developed for spent fuel alteration was integrated in a global reactive transport model of a spent fuel disposal system, considering both decaying and stable isotopes, corroded steel canisters, bentonite backfills and a clayey host-rock. Fuel evolution took into account radiolytic-enhanced corrosion and long-term solubility-controlled dissolution as well as instantaneous release fractions. The calculations show that spent-fuel dissolution has no significant alteration effect on the near-field components except an oxidizing plume in the vicinity of the waste packages. The dissolved uranyl species, partly precipitate as schoepite on the fuel pellets, and partly diffuse in the near-field where magnetite and pyrite reduce U(VI) to yield uraninite precipitation. Under disposal conditions, preliminary calculations indicate that steel corrosion may generate sufficient dissolved hydrogen as to react with radiolytic oxidants and inhibit fuel dissolution. The formation of a protective schoepite layer could also reduce the alteration of fuel pellets. Radionuclides migration (Am, Cs, I) in the near-field is discussed in a second stage discriminating between sorption, precipitation and radioactive decay processes. The migration of Cs is translated in terms of cumulative activity profiles useful for integrated performance assessment.