This work is focused on the material migration into gaps between tiles and into castellation grooves on plasma-facing components from JET: water-cooled Mk-I divertors and belt limiter blocks. The essential results are summarised by the following: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the the CFC tiles gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner and in belt limiter the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on top surface; (iv) fuel inventory in the castellation of the Be divertor and limiter tiles is strongly associated with co-deposition of carbon. Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.