Please use this identifier to cite or link to this item:
|Title:||Studies on nuclear fuel evolution during storage and testing of used fuel response to impact loadings|
|Authors:||RONDINELLA Vincenzo; WISS Thierry; PAPAIOANNOU Dimitrios; NASYROW Ramil|
|Citation:||PSAM11-ESREL2012 Proceedings vol. 4 p. 3171-3179|
|Publisher:||Curran Associates Inc.|
|Type:||Articles in periodicals and books|
|Abstract:||Ageing processes occurring during storage of spent nuclear fuel rods are safety-relevant both for storage and for handling/transportation procedures foreseen at the end of the storage period. In particular, it is important to assess all mechanisms that may cause severe degradation of the mechanical integrity of the spent fuel rods. Such mechanisms are strongly affected by the radioactive decay power generated in spent fuel, which determines the temperature level experienced by the rods in the dry storage container. The behaviour of hydrogen present in the metallic cladding of the fuel rods, dissolved or precipitated as hydride, is the key factor affecting the conditions of the cladding. The build-up of alpha-decay damage and helium may become a determining factor in the evolution of the mechanical properties of the spent fuel. At the Institute for Transuranium Elements (ITU), a Joint Research Centre of the European Commission, several methods are applied to obtain experimental data useful to predict the alterations in spent fuel rods during extended storage time (up to a few centuries), and their potential impact on the safety of procedures affecting spent fuel during and after storage. Alterations as a function of time and specific activity are monitored at different levels, from the microstructural defects and the lattice parameter swelling of the fuel up to macroscopic properties such as hardness and thermal conductivity. In order to reproduce cumulative damage effects expected after centuries of storage within acceptable laboratory timescales, accelerated damage build-up conditions are applied e.g. by using unirradiated (U,Pu) oxide with high specific alpha-activity (alpha-doped UO2). The data from these tailor-made samples are compared to actual spent fuel (UO2, MOX) with high burnup and to natural analogues with ages comparable to a geologic era. Microscopy characterization of hydrides in irradiated cladding samples is performed, also in combination with thermal treatments reproducing conditions to be expected during the various stages of handling and storage of spent fuel after discharge from reactor. Hydride analysis is also part of the study concerning the overall response of a fuel rod against impact loadings, e.g. from accidents during handling or transportation. The impact resistance is investigated by performing hammer drop tests on segments (fuel+ cladding) of irradiated fuel rods. These tests are performed using Light Water Reactor rods with UO2 or MOX fuel at different burnup, and simulating different scenarios (e.g. different ages and thermal histories of the rods). The aim is to determine the used fuel rod mechanical resistance after storage, the amount of fuel material released in case of rod fracture (including fine particles), and to assess the potential consequences associated with such release and dispersion.|
|JRC Directorate:||Nuclear Safety and Security|
Files in This Item:
There are no files associated with this item.
Items in repository are protected by copyright, with all rights reserved, unless otherwise indicated.