Materials Challenges in Nuclear Energy

Advanced Oxidation Resistant Iron-Based Alloys for LWR Fuel Cladding

Hydrogen in zirconium alloys: A review

When zirconium alloys are used in water-cooled reactors, they are subjected to waterside corrosion. The emphasis of this chapter is on uniform oxidation and hydrogen embrittlement. The twin processes of oxidation and hydriding are described comprehensively in terms of their underlying mechanisms.

Corrosion of Zirconium Alloys

Ductile-to-Brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

Prestorage drying operations of high-burnup fuel may make Zircaloy-4 (Zry-4) fuel cladding more susceptible to failure, especially during fuel handling, transport, and post-storage retrieval. In particular, hydride precipitates may reorient from the circumferential to the radial direction of the cladding during drying operations if a threshold level of hoop stress at or above a corresponding threshold temperature is exceeded. This study indicates that the threshold stress is approximately 75–80 MPa for both nonirradiated and high-burnup stress-relieved Zry-4 fuel cladding cooled from 400°C and, under ring compression at both room temperature and 150°C, that radial-hydride precipitation embrittles Zry-4. Specifically, the plastic tensile hoop strain needed to initiate unstable crack propagation along radial hydrides decreases dramatically from >8% to lt;1% as radial-hydride fraction increases. Lower hydrogen contents (lr;300wppm) appear to be more susceptible to radial-hydride embrittlement compared to hig...

https://www.tandfonline.com/doi/pdf/10.1080/18811248.2006.9711195

Radial-hydride embrittlement of high-burnup zircaloy-4 fuel cladding

The fracture behavior of unirradiated Zircaloy-4 containing either solid hydride blisters or hydrided rims has been examined for the contrasting conditions of equal-biaxial and plane-strain tensile deformation at three temperatures (25°, 300°, and 375°C). Cold-worked and stress-relieved Zircaloy-4 sheet containing hydride blisters shows nearly identical failure strains in equal-biaxial and plane-strain tensile deformation for a wide range of blister or rim depths. In all cases, failure strains decrease rapidly with increasing hydride blister or rim thickness, especially in the ≤100 µm range. Test temperature has a significant effect on ductility with failure strains at 300° and 375°C being much greater than at room temperature. The results indicate that the ductility of material containing hydride rims/blisters greater than ≈ 30–40 µm deep is limited by crack growth, which occurs in a mode I manner at 25°C but in a mixed mode I/II manner at ≥300°C (and at higher failure strain levels).

Failure of Hydrided Zircaloy-4 Under Equal-Biaxial and Plane-Strain Tensile Deformation

Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop) direction of the cladding. Also, longitudinal tube specimens were also designed to determine the constitutive properties in the axial direction. Finite-element analysis (FEA) and experimental parameters and results were closely coupled to optimize an accurate determination of the stress-strain response and to induce fracture behavior representative of accident conditions. To determine the constitutive properties, a procedure was utilized to transform measured values of load and displacement to a stress-strain response under complex loading states. Additionally, methods have been developed to measure true plastic strains in the gauge section and the initiation of failure using real-time data analysis software. Strain rates and heating conditions have been selected based on their relevance to the mechanical response and temperatures of the cladding during the accidents.

/pdf/mechanical-property-testing-of-irradiated-zircaloy-cladding-57hg8jcp32.pdf

Mechanical property testing of irradiated Zircaloy cladding under reactor transient conditions

The extended use of Zircaloy cladding in light water reactors degrades its mechanical properties by a combination of irradiation embrittlement, coolant-side oxidation, hydrogen pickup, and hydride formation. The hydrides are usually concentrated in the form of a dense layer or rim near the cooler outer surface of the cladding. Utilizing plane-strain ring-stretch tests to approximate the loading path in a reactivity-initiated accident (RIA) transient, we examined the influence of a hydride rim on the fracture behavior of unirradiated Zircaloy-4 cladding at room temperature and 300 C. Failure is sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness >100 {micro}m ({approx}700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness <90 {micro}m ({approx}600 wppm) remain ductile. The mechanism of failure is identified as strain-induced crack initiation within the hydride rim and failure within the uncracked ligament due to either a shear instability or damage-induced fracture. We also report some preliminary results of the uniaxial tensile behavior of low-Sn Zircaloy-4 cladding tubes in a cold-worked, stress-relieved condition in the transverse (hoop) direction at strain rates of 0.001/s and 0.2/s and temperatures of 26-400 C.

/pdf/on-the-embrittlement-of-zircaloy-4-under-ria-relevant-25rnk48gon.pdf

On the embrittlement of zircaloy-4 under RIA-relevant conditions

Sufficient mechanical ductility of high-burnup Zircaloy-4 fuel cladding is important to prevent large-opening ruptures and significant fuel dispersal during postulated in-reactor and spent-fuel processing accidents. The effect of irradiation, oxidation, and hydriding at high fuel burnup may degrade cladding ductility to the extent that such large ruptures are possible under severe loadings. To understand this susceptibility to failure, this study focused on mechanical testing coupled with detailed finite-element modeling and analyses. Under ring-compression-type loading at room temperature, tensile cracks form within the corrosion-induced oxide layer under elastic loading. The oxide crack then propagates into the cladding wall under additional loading with little to no measurable plastic strain, as confirmed by both experiment and analyses of plastic hoop strain in the ring. For cladding with the oxide removed prior to testing at ≤1 %/s, cracking of the underlying hydride rim comprised of circumferentially oriented hydrides occurs at low plastic hoop strain (≤3 %), whereas the finite-element analysis suggests that the base alloy with a relatively small amount of hydrides appears to fail at higher strain (>8 %). At even higher strain rates (≈400 %/s), cracking within the hydride rim occurs at near-zero ductility, but the base alloy continues to remain highly ductile. These room-temperature results indicate that the hydride rim is sensitive to strain rate, whereas the base alloy is relatively not. With the precipitation of ≈100 % radially oriented hydrides, the cladding exhibits near-zero ductility at room temperature and ≈0.1 %/s. This study suggests that the ring-compression test coupled with finite-element modeling and analysis may be used to estimate crack-initiation strains in irradiated cladding materials with susceptible microstructures and under various deformation rates.

Robert S. Daum

Papers

Radial-hydride embrittlement of high-burnup zircaloy-4 fuel cladding

Failure of Hydrided Zircaloy-4 Under Equal-Biaxial and Plane-Strain Tensile Deformation

Mechanical property testing of irradiated Zircaloy cladding under reactor transient conditions

On the embrittlement of zircaloy-4 under RIA-relevant conditions

Experimental and Analytical Investigation of the Mechanical Behavior of High-Burnup Zircaloy-4 Fuel Cladding