A conceptual design of a helium-cooled blanket that satisfies the boundary conditions specified by the Next European Torus team is presented. The first wall is made of austenitic stainless steel with poloidally running helium cooling tubes and a 10-mm-thick steel erosion layer. The breeding material is lithium orthosilicate (Li4SiO4) with 60% 6Li enrichment and it is contained in the form of 2-mm pebbles in a bed together with 2-mm beryllium particles. Zirconium hydride is used in the back of the blanket to improve the tritium breeding. The main helium cooling system and the helium purge system for the tritium extraction are both at 80 bar, but they are completely separate for tritium-control reasons. An oxidizing atmosphere in the helium purge system ensures that the tritium losses from the plant are <10 Ci/day. The tritium inventory in the blanket is mainly due to tritium adsorption on the surface of the ceramic material. It is <1000 g, provided that the specific surface of the ceramic material is <0.25 m2/g. The rather leaky structure provided by the poloidally running breeder tubes is the main reason for the rather modest tritium breeding ratio. Improvement of the breeding ratio could be obtained by using a high melting point multiplier (beryllium or Zr5Pb3) in the first-wall region. This would also have the advantage of increasing the inlet helium temperature in the blanket region. The helium temperature resulting from the present design would allow a plant efficiency comparable to that of a pressurized water reactor. A higher plant efficiency would require the use of a more advanced structural material than austenitic stainless steel.