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Fretting of Fuel Cladding Materials for Pb Cooled Fast Reactors: Long Term Prediction Using Fretting Maps

[+] Author Affiliations
Mattia Del Giacco, Alfons Weisenburger, Georg Müller

Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany

Paper No. ICONE21-16905, pp. V001T02A050; 8 pages
  • 2013 21st International Conference on Nuclear Engineering
  • Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Radiation Protection and Nuclear Technology Applications
  • Chengdu, China, July 29–August 2, 2013
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5578-2
  • Copyright © 2013 by ASME


Fretting is a particular type of wear that is expected to occur in molten lead alloy cooled nuclear reactors due to flow induce vibrations and will mainly affect fuel claddings and heat exchanger tubes. A new facility (FRETHME) designed to investigate this specific type of wear was applied the first time for fretting test in liquid lead alloys at reactor relevant conditions. Numerous fretting tests at severe conditions (accelerated tests) were performed on candidate steels such as the f/m T91 steel, the austenitic 15-15 Ti steel and Al surface alloyed T91 (GESA-T91).

The fretting damage increases with the increasing number of cycles/time and temperature. Fretting interacts with the corrosion mechanisms occurring in liquid Pb alloys (fretting corrosion) and destabilizes the corrosion barriers, favouring e.g. dissolution attacks. Due to the favourable wear and corrosion resistance properties of the surface alloyed layer, GESA-T91 steel showed the best fretting corrosion behaviour up to 550 °C. On the contrary, due to the high Ni content, the 15-15Ti steel is affected by dissolution enhanced fretting; while oxidation enhanced fretting characterizes T91 steel at temperatures higher than 500 °C.

In most of the tests under accelerating conditions, 10% of fuel clad thickness was penetrated after quite short times already. To extrapolate the obtained results to conditions (load and amplitude) that allow long term use of the respective component in Pb cooled reactors, the concept of fretting maps was applied. Fretting maps were constructed using the obtained experimental data especially the fretting wear coefficient, which is a characteristic of a specific fretting regime. The obtained fretting maps were used to determine the tolerable amplitude and load up to which the tested materials can be used in a Pb cooled nuclear reactor.

In addition, dedicated tests suggested that, besides the use of aluminized steels, possible countermeasures to mitigate the fretting impact are the use of pre-oxidized components and Ni-enriched liquid Pb.

Copyright © 2013 by ASME



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