The dispersion fuel concept offers a method for minimizing irradiation- induced property changes in nuclear fuel systems by localizing fission damage into regulated domains within the system. The system is thereby provided with a structural framework, or matrix, whose properties are relatively unaffected by the dynamic and impurity effects of fission fragments. Dispersion-type fuel systems whose irradiation behavior has been, or is being, investigated in the United States are listed. The extent of property retention in a dispersion fuel alloy is dependent on the materials selected for the fuel-bearing and matrix phases and on their microstructural arrangement in the alloy. The consequences of varying the materials and fuel-phase particle size and volume fraction are discussed. From experimental observations it is shown that, by proper control of these parameters within the bounds of existing technology, the dispersion concept does provide improvement in property retention over alternate alloy design concepts. (auth)