Advances in steam generator design have been made in recent years. The demands of gas-cooled nuclear power plants mean high-temperature operating conditions and space limitations. The feasibility of the high-temperature gascooled reactor (HTGR) concept and the 235U-Th233U fuel cycle was demonstrated by 6 yr of operation of the 40-MW(e) Peach Bottom prototype HTGR power plant. Two steam generators located outside the pressure vessel were used to exchange the heat from the primary coolant (helium) to the secondary coolant (water). A prestressed concrete reactor vessel (PCRV) was used in the design of the 330-MW(e) Fort St. Vrain power demonstration plant. Use of the PCRV made the integration of all the nuclear steam supply system components practical. The primary coolant inventory was reduced and external piping and steam generator pressure shells were eliminated. A once-through-type steam generator system was selected. Materials selected for use in the pressure parts exceeded American Society of Mechanical Engineers Code requirements. The next step in the development of HTGR technology is the large commercial HTGR plant, which has once-through-type steam generators with a nominal capacity of 500 MW(th). Materials used in the main steam section range from 2¼ Cr—1 Mo to Ni-Fe-Cr (Alloy 800). High carbon levels were used to increase the creep strength of the materials. Gas cooling for fast breeder reactors is being studied by designing a 300-MW(e) demonstration plant. The steam generators are similar to the design of the Fort St. Vrain and large commercial plants. Tubes made of Alloy 800 are used.