An In-Reactor Simulation Test to Evaluate Root Cause of Secondary Degradation of Defective BWR Fuel Rod

Abstract

Zircaloy-2 fuel cladding is susceptible to several forms of secondary degradation following steam ingress in defective rods in BWRs. Hydride blister, circumferential, and axial cracks have been reported. An in-core test was performed at the Halden reactor to evaluate the root cause of the secondary degradation, with particular attention paid to the axial cracking issue. The test included four instrumented rods of one meter long. Three rods had Zr-lined Zircaloy-2 cladding, and the other was a nonlined cladding. Two initial diametral gap sizes and two steam ingress conditions were designed. The rods were irradiated to 8 GWd/MtU before steam was admitted into the rods when the rods were operated at power. Subsequently, a power ramp from 33 to 43 kW/m was executed to cause mechanical stresses on the cladding. The test was terminated following detection of a significant activity release. Three of the four rods, including two Zr-lined and the nonlined ones, exhibited hydride bulges and short axial cracks; the remaining Zr-lined rod did not develop secondary defects. The secondary defects developed first by heavy localized hydriding in all cases. Steam ingress causes increase in fuel temperature, formation of central voids, and pellet swelling at the initial hydride defects, resulting in local plastic strains on the cladding at the crack tip. Susceptibility of the cladding to cracking depended on the hydriding characteristics of the cladding. Unlined Zircaloy cladding was susceptible to sunburst hydriding locally within a short axial length. Zr-lined cladding was more resistant to localized sunburst hydriding; hydrogen was absorbed more uniformly by the cladding over a substantially longer cladding length and formed thick hydride rims near the cladding outer surface. The hydride rim ahead of the crack tip in a Zr-lined cladding produced radial hydrides intruding deep into the cladding wall under plastic strains. Repeated crack tip straining by central void formation and deep radial hydride penetration into the cladding from hydride rims would assist the crack to penetrate the cladding wall perpendicular to the cladding surfaces and propagate axially over a long distance. Unlined Zircaloy cladding can also propagate axial cracks by the same mechanism, but its susceptibility to localized sunburst hydriding would limit the length of such axial cracks to a short length nearby the sunburst hydride region. Results from this program suggest that the hydriding characteristics of the cladding inner surface determine the susceptibility of the cladding to long axial cracking. A low corrosion resistance Zr-liner and, particularly, a dry hydrogen environment, which accelerates the hydriding rate, can increase the rate of axial cracking. The findings from this program are consistent with observations of high susceptibility of high purity Zr-liner to axial cracking in BWRs, and provide basis for its mitigation.

An assessment of delayed hydride cracking in zirconium alloy cladding tubes under stress transients

“…According to the published literature, [16][17][18] cladding tube failure of low burnup (,45 GWd MTU 21 ) fuel rods is rare (,10 ppm), usually involving residual or transient local tensile stresses. 10 Possible fuel failure that results in cladding perforation includes manufacturing defects, debris fretting, pellet-cladding interaction (PCI), pellet-cladding mechanical interaction (PCMI), corrosion and DHC.…”

“…[16][17][18] Some failures of Zircaloy-2 cladding tubes with and without a Zr lined barrier were reported during service in BWRs. 16,[19][20][21][22][23][24] From 1989 to 1993, 16,[19][20][21] almost 30% of the failed Zr lined rods exhibited long axial cracks that exceeded 15 cm in length. Some of these cracks were secondary defects, while a few were primary defects.…”

“…Cheng et al 21 performed in-reactor simulation tests to evaluate the root cause of secondary degradation of defective BWR fuel rods. Their results showed the unlined Zircaloy-2 cladding is susceptible to the formation of hydride blister and the associated hydride cracking during a power ramp.…”

“…Oxide cracking and fracture of hydride blisters were both considered. They cited extensive evidence of hydride blisters that formed beneath a spalled oxide layer, [28][29][30][31][32] as well as evidence of the presence of concentrated hydrides or hydride blisters 29,33,34 near the pellet/pellet boundaries and gaps in Zircaloy-4 cladding in fuel rods operated to burnups higher than <55 GWd MTU 21 . 27 While acknowledging the potential for DHC in the outer wall of Zircaloy-4 cladding tubes due to the nucleation, growth and cracking of hydride blisters, CRWMS M&O 27 considered that this failure mechanism did not affect repository performance significantly for the majority of commercial spent nuclear fuels.…”

“…27 While acknowledging the potential for DHC in the outer wall of Zircaloy-4 cladding tubes due to the nucleation, growth and cracking of hydride blisters, CRWMS M&O 27 considered that this failure mechanism did not affect repository performance significantly for the majority of commercial spent nuclear fuels. They concluded that the formation of locally concentrated hydrides was limited to a small fraction of PWR cladding fabricated from standard Zircaloy-4 operated to burnups higher than 55 GWd MTU 21 .…”