CRITICALITY SAFETY ANALYSIS FOR A FINAL REPOSITORY FOR SPENT NUCLEAR FUEL: COMMONNESS AND DIFFERENCES TO THE CLASSICAL APPROACH

摘要

In a classical criticality safety analysis of a system containing fissile material, one has to perform code validation exercises to determine the bias of the calculation tool(s) with regard to the application system. In comparison to this approach, the criticality analysis of a final geological repository for spent nuclear fuel faces two additional problems: 1) Insufficient knowledge of the future long term behavior of a repository in the post-closure phase, and 2) the validation of nuclear data for the material mixtures specific for the repository site. To face these drawbacks, different actions can be taken. For example by probabilistic, risk informed approaches, the risk of criticality can be quantified, or even excluded. This may be treated by analyzing scenarios e.g. by means of Monte Carlo Sampling, or comparable tools. For the creation of such time dependent geochemical models, numerical tools capable to regard for the relevant evolution of the chemical milieu are to be applied. The paper at hand presents some proposals to implement such advanced approaches into the "classical" criticality safety analysis for a final repository at a generic site. Additionally, it presents some approaches recently followed in Germany, where also guidance for probabilistic means is under discussion in the industrial standard (DIN) currently under development.