Topic
Zirconium alloy
About: Zirconium alloy is a research topic. Over the lifetime, 6548 publications have been published within this topic receiving 78954 citations. The topic is also known as: zircaloy.
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TL;DR: A thermodynamic database, Zircobase, was developed for zirconium alloys for use in the nuclear industry as discussed by the authors, and the utility of this database is demonstrated in examples of thermodynamic calculations of the α/β phase transformation temperatures performed on industrial Zr-Nb and Zy4 type alloys.
151 citations
TL;DR: In this paper, an international round-robin experiment was conducted to study the nature of the damage structure in neutron irradiated zirconium and zircaloy-2 using transmission electron microscopy.
150 citations
31 Jul 2008
TL;DR: In this paper, 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).
Abstract: 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 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. 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, fueled-and-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.
149 citations
TL;DR: In this paper, the authors studied the oxidation of chromium-coated zirconium-based alloys under steam at temperatures ranging from 800°C up to 1500°C and for oxidation times ranging from a few minutes up to a few hours.
147 citations
Patent•
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TL;DR: In this paper, a nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconsium alloy tube.
Abstract: A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.
147 citations