Abstract

Bentonites are commonly proposed for use in the geological disposal of high heat generating radioactive wastes. Repository designs include bentonite as a buffer to occupy the void space around the waste canisters because of the favourable properties it can exhibit that enhance the isolation and containment functionality of the repository. Many repository concepts introduce small voids within the bentonite because the buffer is incorporated as individual bricks stacked around the waste canisters. These voids must be closed to prevent the persistence of high permeability pathways for fluids. As bentonite hydrates, it expands and can exert a considerable swelling pressure on the surrounding host rock. The bentonite also expands to fill the engineering cavities inherently present in the repository, but the non-uniform development of total stress and pore pressure could cause persistent material heterogeneities to occur. This is likely to be exacerbated by the thermal gradients existing between the hot waste and the temperature of the surrounding host rock in the early stages of repository post-closure. Whilst this is an area of ongoing research, the final extent of bentonite homogenisation within the repository and for how long property variations persist, is not well understood.In this study, four tests were conducted on pre-compacted, sodium-activated MX80 bentonite samples placed next to a water-filled engineering void, to examine the effect of elevated temperature on the development of swelling and swelling pressure as a function of sample size. The sample lengths were chosen to give small bentonite-to-void ratios, and to represent extremities of behaviour, such that an acceptable upper limit of void size might be established. The results demonstrated that even under extreme bentonite-to-void ratios, the bentonite was able to swell and completely fill the void space, exerting a small but measurable swelling pressure. Under the conditions of this study, the results have shown larger end-of-test swelling pressures and higher final dry densities along their entire length than shown in equivalent tests conducted at ambient temperature. In addition, the elevated temperature tests showed a rapid initial increase and then decrease in swelling pressure at the start of testing, approaching an asymptote in swelling pressure more quickly, whilst uptaking less water than in the ambient temperature case. This implied that heating the bentonite reduced the test duration by about 60%, which is most likely explained by a reduction in the viscosity of the test permeant at higher temperatures.

Highlights

  • Radioactive waste is generated as a by-product of the production of energy from nuclear fuels, and the safe and long-term disposal of this waste is an inevitable requirement

  • The swelling pressure recorded in this radial load cell (R2), which was located at the bottom of the vessel where the sample was positioned at the start of the test, rose to a peak value of 930 kPa very early in the test

  • How­ ever, for the data available and akin to the Dueck et al (2019) data, this study shows slightly higher average swelling pressures when compared to the MX80 model of Borgesson et al (1995) and slightly lower average swelling pressures than those expected by the model of Akesson et al (2010), which was derived from the experimental results of Karnland et al (2006)

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Summary

Introduction

Radioactive waste is generated as a by-product of the production of energy from nuclear fuels, and the safe and long-term disposal of this waste is an inevitable requirement. In the UK, the Nu­ clear Decommissioning Authority is responsible for clearing up 17 nu­ clear legacy sites from the post-war era (NDA, 2016), and safely disposing of High (HLW), Intermediate (ILW) and Low Level (LLW) Wastes in a repository will need to take place. The high heat generating wastes (HHGW) (e.g. HLW and spent fuels (SF)) will be placed in a canister with a clay buffer material occupying the void space around the waste (Sellin and Leupin, 2013). Low heat generating wastes (LHGW) (e.g. ILW and LLW) on the other hand will be disposed of in a separate part of the geological disposal facility (GDF) and likely be surrounded by a cementitious backfill. For HHGWs, this means that the buffer should have a very low permeability whilst being able to withstand the high temperatures produced by decaying radioactive waste without degrading. Degrada­ tion of the EBS could occur mechanically and/or chemically and would likely lead to an increased permeability, reducing its sealing perfor­ mance (Nakayama et al, 2004; Karnland et al, 2007; Cuisinier et al, 2008; Herbert et al, 2008; Fernandez et al, 2009; Ye et al, 2014)

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