The results of the first fatigue crack growth rate (FCGR) tests of irradiated pressure vessel steels in a simulated reactor coolant environment are presented. These results are compared with data on the same and similar unirradiated steels fatigue-tested in high-temperature air, and in high-temperature pressurized reactor-grade water. In the aqueous environment, irradiated condition test results show a slight increase in FCGR over those for unirradiated materials, while both classes of data show a substantial increase above FCGR data for the unirradiated condition in a high-temperature air environment. For unirradiated FCGR data, this increase has been attributed to a hydrogen-assistance model, and this is also suggested to be the case for the irradiated results. Possible interactions of material and environment, irradiation, damage, and hydrogen embrittlement effects are discussed.