Mineralogy and Geochemistry of a Bentonite Pellets Column Heated for 10 years

A common design of a high-level radioactive waste (HLW) disposal system consists of the waste canisters emplaced in tunnels or shafts, with the space between the heat-emitting canisters and the surrounding rock filled with a bentonite-based material. Understanding the behavior of this barrier, in particular the effect of prolonged heating on its properties, is important to assess the barrier's long-term performance. The objective of the present study was to add to this understanding and to supply experimental data about the state of bentonite hydrated and heated for a long period of time. To that end, a 50 cm long column of Wyoming-type bentonite pellets was heated at its base at 140 & DEG;C (simulating the waste canister) while a synthetic sodium-chloride-rich groundwater including sulfate, calcium, and magnesium was supplied through its upper surface for almost 10 y. At the end of the experiment the upper half of the column was saturated, but in the lower half the water content decreased sharply toward the heater, where it was close to 0%. No relevant mineralogical changes occurred, and the main component of the bentonite continued to be a predominantly sodium montmorillonite with no structural changes with respect to the initial one. In the area where the temperatures were >60 & DEG;C and the material very dry, however, the smectite was not able to completely develop the 1-layer hydrate after 48 h of stabilization at RH 55%, although its expandability was fully recovered when the smectite was hydrated with liquid water. The ions solubilized as a result of the water-content increase were transported toward the heater and concentrated in two distinct areas: sodium and calcium chlorides closer to the heater than calcium and sodium sulfates. At the heater contact, the bentonite microstructure was of dense packets with carbon and Na-S-coated cavities. Precipitation of calcite and calcium sulfates and possibly dissolution of silica minerals also took place close to the heater.

Automated summary

The objective of this study was to investigate the behavior of bentonite as a barrier material in a high-level radioactive waste disposal system. Experimental results showed that the main component of bentonite remained as sodium montmorillonite without any structural changes. In a dry and high temperature condition, the expandability of smectite was not fully developed, but it recovered when hydrated with liquid water. Solubilized ions were transported towards the heater and concentrated in two distinct areas. The microstructure of bentonite near the heater exhibited dense packets with carbon and Na-S-coated cavities, and precipitation of calcite and calcium sulfates also occurred.