Develop Method for Hydriding Fuel Cladding and Characterize Influence of Hydriding on Mechanical Behavior, 20-R8269
Inclusive Dates: 11/01/11 – Current
Background — Zirconium-based cladding material exposed to coolant water during nuclear reactor operations could absorb hydrogen ranging from less than 100 to up to 600 ppm, depending on temperature, fuel burnup, and material type. During extended dry storage, cladding plays an important role in safely handling, storing, and transferring spent nuclear fuel. As the cladding cools during extended storage, the hydrogen inside the cladding may precipitate as hydrides; furthermore, both existing and newly formed hydrides may reorient. Depending on size, distribution, and orientation, these hydrides may lead to premature fracture as a result of hydride embrittlement or delayed hydride cracking. Because the United States is actively considering extended dry storage as an alternative approach to managing the spent nuclear fuel and has an increase amount of high burnup fuel as a result of changes in plant-operating conditions, there is a strong need for relevant data. The objectives of this project are to develop methods and identify parameters controlling hydride formation at various hydrogen concentrations, identify conditions when hydrides reorient under stress, and characterize the influence of hydrides and their orientation on mechanical properties.
Approach — The primary objectives of this project are being achieved through the following four tasks:
- Develop method and parameters to prepare specimens with various hydrogen concentration levels.
- Identify conditions when hydrides reorient under stress.
- Conduct mechanical tests to characterize the influence of hydriding on cladding mechanical behavior.
- Provide required periodic reports and develop journal and conference papers.
Accomplishments — Significant progress has been made in the following areas.
Methods of hydriding Zircaloy-2: Four methods have been used to hydride the material: (i) electrochemical method—cathodic charging followed by diffusion annealing; (ii) hydrogen charging in a tubular reactor with continuous flow of a mixture of hydrogen-Argon gas; (iii) hydriding in pure hydrogen in a pressurized vessel at 300°C for 12 hours; and (iv) hydriding in supercritical water at 350°C in a pressurized vessel. Figure 1 shows the cross section of one specimen charged with hydrogen, where the hydrides were highlighted after etching.
Hydride reorientation: A hydride reorientation test was conducted on a hydrogen-charged tensile specimen using an in situ loading stage inside a scanning electron microscope (SEM). Before loading, reference photographs of the hydrides were obtained at several magnifications ranging from 500X to 3,000X. The hydride microstructure viewed at 3,000X magnification shows that the hydrides, which appear as needle-shaped features, are parallel to the axial direction of the tensile specimen and the loading direction.
Mechanical properties of the hydrided Zircaloy-2: Fracture testing was conducted on hydride reoriented three-point bend specimens at 200°C in the SEM. Direct observations indicated that the reoriented hydrides, which ranged from ~1 to 22 µm in lengths, were more prone to fracture at larger sizes (>10 µm) compared to smaller sizes (<0.5 µm). The reoriented hydrides reduced fracture resistance through a void nucleation, growth, and coalescence process at the crack tip, as shown in Figure 2. The resulting crack resistance curves for Zircaloy-2, with reoriented hydrides, decreased from 38 MPa(m)1/2 to 21 MPa(m)1/2, with increasing hydrogen contents from 51 wt. ppm to 1,265 wt. ppm, as shown in Figure 3.