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In the UK, the typical route being followed to handle metallic Intermediate Level Waste (ILW) derived from decladding of irradiated nuclear fuel is to primarily deposit it within stainless steel drums/boxes and subsequently encapsulate it into grout. Thus, a monolithic bond between the waste and the encapsulants can be achieved. However, recent inspections have revealed bulges around the circumference in a considerable proportion of the inspected packages. This distortion has been ascribed to the interior metallic corrosion–induced volume expansion, occurring throughout the waste packages storage period. Throughout this period, the volume expansion taking place in the core of the ILW drums has imposed continuously growing levels of strain on the surrounding encapsulants, i.e. grout and steel confinement. As a result, the encapsulants may exhibit considerable damage, in the form of cracks in the grout or/and steel distortion, threatening the integrity of the packages and, consequently, raising safety concerns within the nuclear community. Herein, experimental work performed on ILW drum–simulant packages is described, attempting to evaluate the performance of the packages under an internal corrosion–induced volume expansion.

Commercial graphites are used for a wide range of applications. For example, Gilsocarbon graphite is used within the reactor core of advanced gas-cooled reactors (AGRs, UK) as a moderator. In service, the mechanical properties of the graphite are changed as a result of neutron irradiation induced defects and porosity arising from radiolytic oxidation. In this paper, we discuss measurements undertaken of mechanical properties at the micro-length-scale for virgin and irradiated graphite. These data provide the necessary inputs to an experimentally-informed model that predicts the deformation and fracture properties of Gilsocarbon graphite at the centimetre length-scale, which is commensurate with laboratory test specimen data. The model predictions provide an improved understanding of how the mechanical properties and fracture characteristics of this type of graphite change as a result of exposure to the reactor service environment.

Previous studies have indicated that carbon is deposited on spinel-type oxides containing manganese, iron, nickel and chromium manganese oxides and uranium oxides, especially where rapid electron exchange is possible in mixed valence compounds. In this investigation, characterised magnetite, Fe 3 O 4 , surfaces have been subjected to γ-irradiation under conditions of temperature, pressure and atmosphere similar to those experienced in reactor with the aim of furthering our understanding of the catalytic processes involved in deposit initiation and growth. Although it was not possible to produce large (111) Fe 3 O 4 planes, enhanced carbon deposition was observed on small (111) regions showing the enhanced catalytic effect of these planes for carbon deposition out of the gas phase.

The composition of oxides formed on steel surfaces within power reactors may influence heat transfer efficiency. Previous studies have indicated that carbon is deposited on spinal-type oxides containing manganese, iron, cobalt, nickel and chromium. In this investigation, characterised manganese oxides have been subjected to γ-irradiation under conditions similar to those experienced in reactors in an effort to understand the catalytic processes involved in deposit initiation and growth. Mn3O4 and Mn2O3, under the conditions present in the γ-cell, were reduced to MnO during the time of exposure. Relative carbon deposition rates were observed to follow the trend MnO>Mn3O4≈Mn2O3.