Activation analysis of concrete and graphite in the experimental reactor RUS.

The decommissioning and dismantling of nuclear installations after their service life involves the necessary disassembling, handling and disposing of a large amount of radioactive equipment and structures. In particular, the concrete that has been used as a biological reactor shield and graphite that has been used as a moderator-reflector represent the majority of waste, requiring geological disposal. To reduce this undesirable volume to the minimum and to successfully plan the dismantling and disposal of radioactive materials to storage facilities, the activations of the structures should be accurately evaluated. In the framework of the decommissioning and the dismantling of the experimental reactor of the University of Strasbourg, detailed activation estimates have been conducted to characterise the graphite and the structural materials present in the reactor environment. For this purpose, the chemical compositions of fresh graphite samples and different types of concrete have been determined by activation analysis in the research reactors OSIRIS and ORPHEE of CEA Saclay (France). Then, the activations of graphite, concrete and other materials have been calculated in the whole reactor, as a function of the three main nuclear data libraries, i.e. ENDF, JEF and JENDL. In parallel, the activations of representative graphite and concrete samples have been measured experimentally. The comparison of theoretical predictions with experimental values validates the approach and the methodology used in the present study and tests the consistency and the reliability of the nuclear data used for activation analysis. We believe that a similar approach could also be used for the decommissioning of industrial nuclear reactors.

Non-destructive method of characterisation of radioactive waste containers using gamma spectroscopy and Monte Carlo techniques.

During the decommissioning of the SATURNE accelerator at CEA Saclay (France), a number of concrete containers with radioactive materials of low or very low activity had to be characterised before their final storage. In this paper, a non-destructive approach combining gamma ray spectroscopy and Monte Carlo simulations is used in order to characterise massive concrete blocks containing some radioactive waste. The limits and uncertainties of the proposed method are quantified for the source term activity estimates using 137Cs as a tracer element. A series of activity measurements with a few representative waste containers were performed before and after destruction. It has been found that neither was the distribution of radioactive materials homogeneous nor was its density unique, and this became the major source of systematic errors in this study. Nevertheless, we conclude that by combining gamma ray spectroscopy and full scale Monte Carlo simulations one can estimate the source term activity for some tracer elements such as 134Cs, 137Cs, 60Co, etc. The uncertainty of this estimation should not be bigger than a factor of 2-3.