The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200ºC, ring compression tests were performed to determine post-quench ductility at ≤135ºC. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000ºC. Among other findings, embrittlement was found to be sensitive to fabrication processes -especially surface finish -but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueledand-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur. ForewordFuel rod cladding is the first barrier for retention of fission products, and the structural integrity of the cladding ensures coolable core geometry. In the early 1990s, new data from foreign research programs showed degraded cladding behavior for high-burnup fuel compared with low-burnup fuel in tests designed to simulate postulated accidents. Interim actions were taken, but it became clear that extrapolation from a low-burnup data base needed to be reassessed more carefully for regulatory purposes.One of NRC's central regulations used in plant licensing deals with postulated loss-of-coolant accidents (LOCAs). A portion of that regulation in 10 CFR 50.46(b) specifies criteria that were derived from tests with unirradiated Zircaloy cladding, and these criteria limit the peak cladding temperature and the maximum cladding oxidation during the accident. These two limits are known as embrittlement criteria. Their purpose is to prevent cladding embrittlement during a LOCA, thus ensuring that the general core geometry will be maintained and be coolable.In the mid-1990s, NRC sponsored a cooperative research program at Argonne National Laboratory to reassess these limits for the possible effects of fuel burnup. The program's industry partners included the Electric Power Research Institute, Framatome ANP (now AREVA), Westinghouse, and Global Nuclear Fuel; in general, the industry partners were responsible fo...

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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.

ABSTRACT: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 ringstretch 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 µm (≈700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness <90 µm (≈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.

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