incorporate thorium fuel elements. This reactor could be operated either on a once-through fuel cycle or with reprocessing to recover the 233U and thorium for recycle. A once-through cycle would require, however, considerably more uranium mining and enrichment (See Table 5–4 for comparative uranium consumption.)

The high-temperature gas-cooled reactor loads fuel elements into graphite blocks that serve as the reactor’s moderator. The coolant is high-temperature helium. A 330-megawatt (electric) (MWe) commercial plant is operating near Fort Saint Vrain, Colorado. HTGR’s appear capable of operating at both higher thermal efficiency and higher conversion ratios than light water reactors, but their expanded use depends on successful development of economical reprocessing for graphite-based fuel. If the high-temperature gas used to cool the graphite core could be used to drive a gas turbine directly, the thermal efficiency of this reactor could be further improved and the reactor’s operation would be freed of requirements for water. Moreover, HTGR’s, or a version of the related German pebble-bed reactor (whose fuel is contained inside balls of graphite), could be used to supply process heat as well as electricity.

In the molten-salt reactors, solid fuel assemblies are replaced by uranium fluoride and thorium fluoride dissolved in a molten fluoride-salt mixture. The salt is circulated to a heat exchanger external to the core, The molten-salt breeder reactor adds chemical kidneys in the fuel’s external circulation system for continuous reprocessing of the fuel/coolant. This feature reduces the total inventory of fissile material committed to the power plant and its fuel cycle, as compared to other breeder reactors. Volatile fission products are also continuously removed, a step that permits true breeding in the Th-U fuel cycle with a thermal-neutron spectrum. A small pilot molten-salt reactor (10 megawatt (thermal) (MWt)) was operated at Oak Ridge National Laboratories. The development of this concept is far behind that of other advanced converters and breeders. Success cannot be guaranteed because of formidable materials problems, but the advantages that might be realized in this type of system are considerable.


Several factors must be considered in evaluating advanced converters: ease of development, economic prospects, and compatibility with other policy objectives of the nuclear program. On this latter point, since policy objectives change, the important criterion is that a reactor type perform well under different, or variable, policy limitations.

Two advanced converter systems stand out as being clearly favored by these criteria: the prismatic HTGR (named for the shape of its fuel

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