This hydriding effect is particularly related to the use of zirconium alloy as fuel cladding. Hydrogen is produced due to reaction between water and zirconium in LWR environment. But moisture can also attack from within the fuel pellets if the pellets are not dried properly before insertion into the reactor. Hydrogen can then be absorbed and diffuse in the interior of zirconium cladding. The low-temperature a-zirconium (HCP) phase has a very low solubility of hydrogen, leading to any excess hydrogen-forming zirconium hydride precipitates. This leads to embrittle­ment, delayed hydride cracking, and hydride blistering, all of which diminish the lifetime of fuel rods and cause safety concerns in spent nuclear fuel rod reposito­ries. It has been known that crystallographic texture of the zirconium alloy fuel cladding tube can be tailored to minimize the effect of hydride formation. With a suitable texture, the hydrides form along the hoop direction in place of radial direc­tion, making it less susceptible to fracture. Figure 6.45 shows a microscopic cross section of hydrides oriented along the hoop direction in a zircaloy tube (the usual

texture consists of majority of grains with c-axis close to the radial direction: +/ — 30° from the radial direction toward the hoop).

Stress Corrosion Cracking

In Section 5.3, we have learned about corrosion basics, including stress corrosion cracking. Inside the reactor, these effects manifest themselves in various ways with respect to different components. Here, we introduce an important effect that occurs in the fuel rod known as pellet-cladding interaction (PCI). Figure 6.46a shows a schematic of the PCI effect. PCI is associated with local power ramps at start-up or during operation and occurs due to the effect of fission products like I, Cs, Cd, and so on, resulting in stress corrosion cracking. The crack generally initi­ates in the inner wall of the cladding and then progresses outward. Figure 6.46b shows an actual example of PCI failure in a fuel cladding system in a commercial power reactor. PCI minimization/elimination can be achieved by the following: (i) reduced ramp rates (flux variation or thermal gradient), but it is not an easy solu­tion, (ii) coated (barrier) fuel, that is, surface coated with proper lubricant, and (iii) barrier cladding, that is, inner surface coated with graphite, copper, or pure zirconium to minimize stresses at the ID surface of the cladding; in BWRs, crystal bar Zr and zircaloy-2 are coextruded to form a thin Zr liner on the ID. The latter is discussed in detail in the following.

A modified Zr-lined barrier cladding known as TRICLAD™ has been developed by GE by adding a thin layer of corrosion-resistant zircaloy-2 bonded to the inner surface of the Zr barrier [40, 41]. In addition, the outer surface is made resistant to nodular corrosion by heat treatment that results in small second-phase particles, while major­ity of the parent zircaloy-2 material contained characteristic large SPP size distribu­tion for improved crack growth resistance (toughness) [42, 43]. Figure 6.47 is a schematic of the Triclad for BWR service with the following four layers from ID to OD: (i) an inner layer of corrosion-resistant zircaloy-2 to slow oxidation and hydrogen generation, and to delay local hydride formation in the case of rod perforation, (ii) a Zr barrier for PCI resistance to blunt cracks nucleated at the inner surface, (iii) bulk

Corrosion-Resistant Outer Zircaloy-2 Surface

parent zircaloy-2 tubing with good toughness during irradiation, and (iv) an outer layer of zircaloy-2 processed for high resistance to nodular corrosion. An important feature of the final product is the integrity of the metallurgical bonding between the Zr barriers with both the inner zircaloy-2 liner and the bulk zircaloy-2 tubing.