Mechanical integrity studies on spent nuclear fuel rods

The consequences of potential accidents causing spent fuel rod failure may involve fuel particles release and dispersion. This paper presents recent results from spent fuel experimental studies performed at JRC-Karlsruhe addressing handling/transportation and long-term storage issues. An impact test using a hammer drop device in hot cell was performed on a spent fuel segment from a UO2 PWR rod with a burnup of ~67 GWd/tHM. The segment was not defueled and was repressurized to 40 bar before the test. Similarly to what observed in previous impact tests, only the fuel volume directly affected by the rod fracturing was released. In addition to the fuel material released during the impact, neither further particles release nor "flow-out" type of behaviour was observed by further tapping on the fractured segments after the test. Preliminary particle size distribution analysis of the fuel particles deposited on a second stage filter of the testing chamber collecting particles with size ≤8 m indicates a log-normal distribution with main particle size of 2.4 m and standard deviation of 1.1 m. A few sub-micron particles were detected. The detailed analysis of the results, including finer particle fractions, is still ongoing. The final goal of these investigations is to determine criteria and conditions governing the response of spent fuel rods to impact loads and other thermo-mechanical solicitations corresponding to normal and off-normal conditions that may be experienced by the rod during handling, transportation, storage and after extended storage. In addition to impact and other mechanical loading tests, property measurements as a function of accumulated radiation damage are performed on spent fuel and alpha-doped analogues to determine the long term evolution and the potential effects of ageing processes on the mechanical integrity of the spent fuel rod.

RONDINELLA Vincenzo;  NASYROW Ramil;  PAPAIOANNOU Dimitrios;  VLASSOPOULOS Efstathios;  CAPPIA Fabiola;  DIESTE BLANCO Oliver;  WISS Thierry; 

2017-01-16

American Nuclear Society

JRC103277

http://www.ans.org/meetings/m_251,    https://publications.jrc.ec.europa.eu/repository/handle/JRC103277,