Oxidation kinetics of Zircaloy-4 and Zr–1Nb–1Sn–0.1Fe at temperatures of 700–1200°C

Baek, Jeong‐In; Park, Ki Bum; Jeong, Yong Hwan

doi:10.1016/j.jnucmat.2004.08.007

Cited by 79 publications

(27 citation statements)

References 22 publications

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“…Then all the specimens were pickled in a solution of 5% HF, 45% HNO 3 , and 50% H 2 O and cleaned ultrasonically in an ethanol and acetone solution. The oxidation tests in steam were conducted by using the apparatus already described in a previous paper [22]. The steam supply rate was 0.6 g/cm 2 -min at standard temperature and pressure (STP).…”

Section: Methodsmentioning

confidence: 99%

Breakaway phenomenon of Zr-based alloys during a high-temperature oxidation

Baek

Jeong

2008

Journal of Nuclear Materials

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Section: Methodsmentioning

confidence: 99%

Breakaway phenomenon of Zr-based alloys during a high-temperature oxidation

Baek

Jeong

2008

Journal of Nuclear Materials

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“…Similar trends have been observed by other investigators in steam-oxidized Zircaloy-2 4, 18 and Zircaloy-4. 16,17,19 The relative thickness of the fully converted oxide layer and Otemperature and time. Weight gain measurements in these studies indicated that the total oxygen uptake for samples tested at 1000 O C obeys a parabolic rate law, W 2 = K p t where K p is the temperature dependent rate constant, W denotes oxygen weight gain, and t is time.…”

Section: Resultsmentioning

confidence: 99%

High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

McHugh

Garnier

Rashkeev

et al. 2013

Ceramic Materials for Energy Applications III

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Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support downselection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000 O C. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a -SiC CMC overbraid, and sintered -SiC were tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of Hepressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined. INTRODUCTIONA primary requirement of advanced nuclear fuel cladding designs is to contain the nuclear fuel inside the fuel rods in the event of a loss-of-coolant accident (LOCA) without diminishing heat transfer characteristics or impairing coolant flow during normal operation. This critical function would increase the safety margin of nuclear reactor designs significantly. While Zr-based alloys are currently used as cladding in most nuclear reactor designs, a variety of advanced cladding materials including metals, ceramics and metal/ceramic composites are being considered as replacements. The corrosion behavior of these materials depends on many factors, including the chemical nature of the cladding material, the reaction temperature, and the type and concentration of the oxidizing agent. For example, corrosion of Zr in air follows a different oxidation pathway than it does in steam. In air, the predominant reaction is:

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“…oxygen and becomes the oxide. The diffusion depth of the oxygen is expanded with the increasing temperatures and the thickness of the zirconium oxide layer is increased [16]. When the burn-up of the fuel is increased or a loss of coolant accident (LOCA) occurs, the temperature of a fuel system rises.…”

Section: Determination Of the Chemical Structures For The Oxidized Zimentioning

confidence: 99%

“…The importance of a study on a Zircaloy oxidation at a high temperature should be emphasized in relation with the irradiation performance and safety of fuel pins. The oxidation behavior of zirconium alloys as the cladding materials have been discussed at high temperature [15][16][17][18][19][20] outer surface oxidation of a cladding, but not so many studies have been done on the inner surface oxidation.…”

Section: Introductionmentioning

confidence: 99%