Regular Article

Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

Radioactive Materials Division, National Cooperative for the Disposal of Radioactive Waste (NAGRA), Hardstrasse 73, 5430 Wettingen, Switzerland

^* e-mail: stefano.caruso@nagra.ch

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 history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for UO₂, whereas the latter is reduced to 9% when more accurate irradiation data are used (core-follow flux data instead of life-average flux data).