A Laboratory Method to Predict Hydriding Properties of Zirconium Alloys Under Irradiation

Abstract

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 same 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 1.0 M) aqueous lithium hydroxide solutions. The test times ranged from 7 to 35 days, and the specimens were held at constant temperatures within the water reactor coolant temperature range (280 to 315°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. Potential applications of the method include the following: (1) screening to predict relative in-reactor hydriding behavior of zirconium based materials, (2) investigating compositional and fabrication variables that influence hydriding within a given zirconium alloy system, and (3) investigating hydriding mechanisms.