Key Topic 2: Understanding the Wastes

Overview

This research area deals with understanding the behaviour of various wastes in geological disposal facilities (repositories).  The waste types include spent uranium oxide (UO2) and mixed oxide (MOX) fuels, vitrified high level waste (HLW) and long-lived intermediate level radioactive wastes (ILW).  The various waste types represent the potential source terms for release of radionuclides after waste disposal canisters, or containers, are breached.

Studies in this area aim to better quantify the processes controlling radionuclide, chemical and gas release from waste forms, in order to improve the quality of models used in safety assessments and to adequately define the types and magnitudes of uncertainties associated with various processes.  It is expected that, because experimental facilities and specialised equipment for work with highly radioactive materials are available in only a few countries, benefits for other countries could become available through use of these facilities under international co-operation arrangements.

Objective

The purpose of this research is to understand safety-relevant processes, in particular the contribution of the waste form to radionuclide retention in the repository.  It is important to define the total inventory of various radionuclides and their time-dependent release in mathematical models in order to assess radionuclide release from a repository.

The 2011 IGD-TP Strategic Research Agenda identified two research topics of high importance and urgency, a further two of medium importance and urgency, and a fifth of low importance:

  • Topic 1 (High). Improved data for the rapid release fraction for spent uranium oxide fuel and improved understanding of its dissolution behaviour.  While significant improvements in understanding of the performance of spent fuel under repository conditions have taken place over the past 10–15 years, the gradual increase in discharge burn-up means that the present spent fuel behaviour database should be extended.  Most of the published data on rapid release fractions of various long-lived radionuclides and matrix dissolution relates to fuel with burn-up values below 45 GWd/tU.  Over the next ten years the average burn-up is expected to exceed this value for many reactors, reaching average values of about 60 GWd/tU.  Therefore, licence applications related to disposal of such fuel may be inadequately supported unless further work is performed.
  • Topic 2 (High). Improved data and understanding of the release of radionuclides and chemical species from various long-lived ILW.  This includes detailed characterisation methods, issues related to adequate inventory determination, chemical form, speciation on release, and transport in the near-field and in the far-field.
  • Topic 3 (Medium). Improved data and understanding of the behaviour of spent MOX fuel – the rapid release fractions and dissolution behaviour under repository conditions should be further studied to determine how fuel structure influences the dissolution.  Quantities of such fuel are presently small and their disposal is likely to be deferred to allow cooling because of their high heat generation.
  • Topic 4 (Medium). Further development of burn-up credit methodology and its application for fuels with higher enrichment that allow higher burn-up.  A number of studies related to post-emplacement criticality evaluation have been performed, with results showing that, with the application of burn-up credit, canisters of spent fuel in a repository will remain sub-critical provided a moderate burn-up is reached.  These studies need to be extended to high burn-up fuel, which has a higher 235U enrichment.  This is a broader issue than for high burn-up fuel, as a consistent methodology is needed to evaluate fuel over the entire burn-up range, including MOX fuel.
  • Topic 5 (Low). Improved data and understanding of the performance of vitrified high level waste.  Although this has been extensively studied and considerable data and information is available, further work is required, albeit with a lower importance reflecting its relatively lower importance in safety assessments.