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Probably the most important range of materials to consider for the blanket material in the tokamak design for fusion reactors such as ITER and DEMO is the high alloy Fe9Cr oxide dispersion strengthened (ODS) ferritic steels. These steels possess exceptional thermal conductivity and low thermal expansion while being strongly resistant to void swelling. Their main drawback is the high ductile-to-brittle transition temperature (DBTT), particularly in the ODS versions of the material. This paper describes attempts that are being made to reduce this DBTT in as yet unirradiated materials by a novel heat treatment procedure. The principle behind this approach is that low DBTT in the unirradiated materials will lead to relatively low DBTT even in He-containing material that has been irradiated with fusion blanket-type irradiations. New batches of high alloy Fe9Cr ODS (EUROFER) ferritic steel have been produced by a powder metallurgical route, and relatively homogeneous material has been produced by a hot isostatic pressing procedure. Mini-Charpy test specimens were made from materials that had been subjected to a matrix of heat treatments designed to show up variations in solution treatment (ST) temperature, cooling rate from the ST temperature and tempering treatment. The initial DBTT was in the range 150-200 degrees C. Extremely interesting results have been obtained. DBTT downward shifts of up to 200 degrees C have been observed by using a high 1300 degrees C ST temperature and a low cooling rate. The paper goes on to describe the microstructure of this material, and discusses the possible microstructural factors needed to produce these very high DBTT downward shifts. Low dissolved carbon and higher proportions of low-angle grain boundaries seem to provide the key to the understanding of the alloy behaviour.