The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. | Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis EPJ Nuclear Sci. Technol. 2 4 2016 Nuclear Sciences S. Caruso published by EDP Sciences 2016 amp Technologies DOI epjn e2015-50057-8 Available online at http REGULAR ARTICLE Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis Stefano Caruso Radioactive Materials Division National Cooperative for the Disposal of Radioactive Waste NAGRA Hardstrasse 73 5430 Wettingen Switzerland Received 25 September 2015 Received in final form 4 November 2015 Accepted 24 November 2015 Published online 15 January 2016 Abstract. The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact the safety assessments of geological repositories require the average and maximum in the sense of very conservative inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE TRITON simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation .
