Hydrogen is absorbed into zirconium alloy–clad fuel elements and structural components as a result of water-side corrosion in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. Increasing hydrogen concentrations in the zirconium alloy affect both its mechanical properties and dimensional stability. It is well known that hydrogen can diffuse relatively rapidly through zirconium base alloys (in relation to the lifetime of the components) in response to solid solution concentration, temperature, and stress gradients. Migration of hydrogen within a component through its life must be considered to reliably predict the performance of the component through life and even post-service life. Many of the physical properties needed to describe this process, such as the solubility of hydrogen in zirconium alloys, diffusion coefficients of hydrogen through zirconium alloys, and the Soret effect within zirconium alloys, already have been published. Many of these measurements have been made outside of the reactor core environment on nonirradiated zirconium alloys.

This paper presents in-reactor, isothermal hydrogen diffusion data that show an effect of the in-reactor radiation environment on the final measured hydrogen concentrations. Samples used are similar to the hydrogen diffusion couples used by several investigators in the past to measure the hydride dissolution solvus as a function of temperature. The experimental results are interpreted on the basis of a physical model for hydrogen trapping that hypothesizes that hydrogen in solid solution in the zirconium lattice comes into thermodynamic equilibria with the hydride phase as well as into a steady-state equilibrium with trapping at irradiation-induced defects within the lattice. In agreement with the model, the irradiation effect on final measured concentrations is seen to be a decreasing function of the irradiation temperature between ∼271°C and 354°C, with any effect essentially eliminated by an irradiation temperature of 354°C.