Oxidation kinetics and related phenomena of zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions

Leistikow, S.; Schanz, G.

doi:10.1016/0029-5493(87)90286-x

Cited by 149 publications

(56 citation statements)

References 2 publications

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“…Belt polishing removes surface impurities in the range of 0.1-0.3 µm and may result in some additional cold working of the near-surface metal. In terms of hydrogen content or hydrogen pickup, the HBR-type Zry-4 (1980s) has a much higher breakaway-oxidation time (≈3800 s) than the time (≈1800 s) determined by Leistikow and Schanz [19,20] for 1970s Zry-4. The breakaway-oxidation time (≈5000 s) for modern AREVA 15×15 Zry-4 tested by ANL is comparable to the breakaway-oxidation time (≈5400 s) determined by Mardon et al [21] for modern AREVA 17×17 Zry-4.…”

Section: Breakaway Oxidation Time For 15×15 Zry-4 Samples Oxidized Atmentioning

confidence: 97%

“…Conversely, irregularities in surface geometry can result in alternating regions of compressive and tensile stresses. According to Leistikow and Schanz [19,20], such irregularities along the metal-oxide surface are a precursor to tetragonal-to-monoclinic transformation and breakaway oxidation.…”

Section: Effects Of Surface Conditions On Cladding Performancementioning

confidence: 99%

“…Breakaway oxidation has been investigated by Leistikow and Schanz [19,20] for Zry-4 and Mardon et al [21] for Zry-4 and M5. Although the breakaway oxidation time is generally defined as the time corresponding to an increase in oxidation rate (i.e., weight gain rate), it is the associated hydrogen pickup that causes embrittlement.…”

Section: Breakaway Oxidation Time For 15×15 Zry-4 Samples Oxidized Atmentioning

confidence: 99%

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Cladding embrittlement during postulated loss-of-coolant accidents.

Billone¹,

Yan²,

Burtseva³

et al. 2008

137

105

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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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Section: Breakaway Oxidation Time For 15×15 Zry-4 Samples Oxidized Atmentioning

confidence: 97%

Section: Effects Of Surface Conditions On Cladding Performancementioning

confidence: 99%

Section: Breakaway Oxidation Time For 15×15 Zry-4 Samples Oxidized Atmentioning

confidence: 99%

See 1 more Smart Citation

Cladding embrittlement during postulated loss-of-coolant accidents.

Billone¹,

Yan²,

Burtseva³

et al. 2008

137

105

View full text Add to dashboard Cite

show abstract

“…The zirconium alloy plus steam reaction has been widely studied under LOCA scenarios. [4,[20][21][22][23][24][25] Zircaloy oxidizes in the presence of steam to form zirconia and hydrogen following an exothermic reaction:…”

Section: B Reaction Of Cladding Materials With Steammentioning

confidence: 99%

Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

Rebak

2015

Metallurgical and Materials Transactions E

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As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO 2 -zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [e.g., 561 K (288°C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288°C).

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“…Breakaway can occur several times, leading to corrosion kinetics showing a cyclic behaviour [1,2]. Post-transition phenomena are also encountered at temperatures expected in accidental situations, either in steam [3,4] or oxygen and air atmospheres [5,6]. The oxidation resistance of a zircaloy, in usual operation conditions as well as in accidental situations, is controlled by the resistance to breakaway of the zirconia scale.…”

Section: Introductionmentioning

confidence: 99%

Influence of Water Vapour on the Zircaloy-4 Oxidation at 700 °C: Stress Determination by In Situ X-ray Diffraction

Rolland¹,

Buscail²,

Issartel³

2017

Oxid Met

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Abstract. Zircaloy-4 has been oxidised in dry and wet air (15 vol.% H 2 O) at 700 °C. At high temperature and at room temperature X-Ray diffraction shows that the oxide scale is composed of monoclinic zirconia (m-ZrO 2 ) and quadratic zirconia q-ZrO 2 . In dry and wet air, growth stress is determined in situ at 700 °C on the m-ZrO 2 . In dry air, results show that the compressive growth stress in the oxide oscillates between -1300 ±100 MPa and -600 ±100 MPa. In wet air, compressive growth stress varies from -1000 ±100 MPa to -400 ±100 MPa during the 20 first hour oxidation. Then, the stress level remains constant (-400 ±100 MPa). In dry air stress oscillations can be explained by successive growth stress increases in the growing scale and stress relaxation by perpendicular cracks propagation. The growth stress relaxation process in wet air is related to a higher zirconia grain size.

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Oxidation kinetics and related phenomena of zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions

Cited by 149 publications

References 2 publications

Cladding embrittlement during postulated loss-of-coolant accidents.

Cladding embrittlement during postulated loss-of-coolant accidents.

Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

Influence of Water Vapour on the Zircaloy-4 Oxidation at 700 °C: Stress Determination by In Situ X-ray Diffraction

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