Decommissioning, Immobilisation and Storage
soluTIons for NuClear wasTe InVEntories

AGR, Magnox and Exotic Spent Fuels

DISTINCTIVE is a multi-disciplinary collaboration of 10 universities and 3 key industry partners from across the UK’s civil nuclear sector.

Grain boundary damage mechanisms in strained AGR cladding under irradiation

Grain boundary damage mechanisms in strained AGR cladding under irradiation


Academic Lead – Enrique Jimenez-Melero

Researcher: Chiara Barcellini

University – The University of Manchester


Second-generation Advanced Gas-cooled reactors (AGRs) make use of UO2 pellets contained in austenitic stainless steel SS 20Cr/25Ni/Nb cladding. During the service life inside the reactor, the cladding undergoes significant damage doses of its microstructure (of the order of a few tens of dpa) due to the constant neutron bombardment at temperatures that may vary between 350°C and 700°C. Neutron irradiation will generate additional vacancies and interstitials in the SS structure. Those radiation-induced defects will evolve over time and give rise to extended defects such as dislocation loops or channels, strain localization and radiation induced segregation (RIS). Those nano-micro scale changes in the SS structure will affect its mechanical integrity and its susceptibility to localised corrosion. In order to be able to potentially extend the lifetime of the AGR reactors and also to store the spent fuel cladding in the cooling ponds safely, we need to develop a thorough fundamental understanding of the radiation damaged SS structures, and how those sub-micron structures govern the possible failure mechanisms of the SS claddings.

Current understanding bases the localised corrosion susceptibility of these materials on the RIS phenomenon that depletes chromium from the grain boundaries, while segregating other elements (e.g. Ni or Si) in the vicinity of the grain boundaries. However, our mechanistic understanding of RIS and its link to the failure mechanisms is very limited, and its effect on GB precipitation at the relevant reactor conditions remains largely unknown. Moreover, neutron damage may also lead to preferred dislocation channels. Those preferred channels would cause plastic instabilities and strain localization near the GBs that will affect the structural integrity   and corrosion susceptibility of the cladding. Those localised plastic phenomena are currently being subject to extensive experimental and modelling work in 304 and 316 SS, but have not been considered as a major factor in radiation-induced failure of AGR-type SS so far. To add complexity to this framework, certain parts of the cladding retain significant deformation from the manufacturing route. As a consequence, specific areas in the SS structure may present significant strain localization even before being exposed to neutron bombardment.

Consequently, a systematic study of the GB radiation damage mechanisms in cladding materials at relevant reactor and/or storage conditions is urgently required before extending the time scale of the AGR claddings inside the reactors or in the wet storage ponds. Neutron-irradiated AGR claddings are difficult and expensive to handle and test due to their relatively high levels of activity, requiring specific transport procedures and devoted active labs. Thermally sensitised AGR-type SS materials do not seem to yield equivalent phenomena as those expected in neutron irradiated samples. The alternative proposed in this project is to use systematically ion irradiation to simulate neutron-damaged structures.


To elucidate the principal radiation damage mechanisms operating at the grain boundaries and their local environment, and to link those atomic-scale mechanisms to the structural integrity and potential localised failure phenomena of Nb-stabilised 20Cr/25Ni stainless steel claddings in AGR reactors and storage ponds.

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