[en] A corrosion and hydriding test series on zirconium alloys in the Engineering Test Reactor G-7 loop demonstrated relatively large lot-to-lot and alloy-to-alloy differences in hydriding rates under irradiation. Similar differences were also found among irradiated Zircaloy-2 pressure tubes fabricated by three suppliers for the Hanford Site N Reactor. This substantial in-reactor hydriding data base and access to archive materials from these alloys permitted an investigation of methods to reproduce the in-reactor hydriding orders-of-merit by an out-of-reactor method. The out-of-reactor method selected for investigation consisted of autoclaving alloys in relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solutions. The test times ranged from 7 to 35 d, and the samples were held at constant temperatures within the water reactor coolant temperature range (280/degree/C to 315/degree/C). The in-reactor hydriding behavior for several lots of Zircaloy-2, one lot of Zircaloy-4, and one lot of Zr--2.5Nb was reproduced in the lithium hydroxide tests. The hydriding rates were compared on the basis of the ratio of hydrogen weight gain to oxide weight gain. 13 refs., 4 figs., 6 tabs