Criteria for Fracture Initiation at Hydrides in Zirconium-2.5 Pct Niobium Alloy

Simpson, L. A.

doi:10.1007/bf02644181

Cited by 63 publications

(22 citation statements)

References 20 publications

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“…It is well known that elevating the level of the stress triaxiality ratio decreases the fracture strain of recrystallized Zircaloy sheet containing uniform distributions of hydrides [8,9]. Conducting a laboratory test to assess the safety of Zr alloy cladding during an RIA presents difficulties.…”

Section: Introductionmentioning

confidence: 99%

“…[6][7][8][9][10][11][12]). However, when compared to the ductility behavior in these previous laboratory studies, an RIA event involving high burnup fuel presents two important new conditions that affect cladding response: (a) non-uniformly distributed hydrides in the form of a layer or rim and (b) failure under a multi-axial tension stress state.…”

Section: Introductionmentioning

confidence: 99%

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On the Embrittlement of Zircaloy-4 Under RIA-Relevant Conditions

Daum

Majumdar

et al. 2002

Zirconium in the Nuclear Industry: Thirteenth International Symposium

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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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Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

On the Embrittlement of Zircaloy-4 Under RIA-Relevant Conditions

Daum

Majumdar

et al. 2002

Zirconium in the Nuclear Industry: Thirteenth International Symposium

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“…None of these studies were specifically focused on the fracture properties of ''radial'' hydrides that would be of interest for input into models of DHC initiation and growth, which is the main emphasis of this chapter. The following summarizes key findings from these latter types of studies that were initiated by Simpson [45].…”

Section: Early Workmentioning

confidence: 96%

“…A program with this in mind was, therefore, initiated by Simpson [45]. This author used Acoustic Emission (AE) to detect cracking in hydrides during continuous tensile loading of smooth and notched tensile specimens containing a uniform distribution of hydrides.…”

Section: Early Workmentioning

confidence: 99%

Fracture Strength of Embedded Hydride Precipitates in Zirconium and its Alloys

Püls

2012

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components

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“…[1][2][3][4][5][6][7] It has been found, at room temperature, In this series of articles, Part I is concerned with the that ZIRCALOYs undergo a ductile-to-brittle transition general influence of hydride on the mechanical properties when the hydrogen content (hydride volume fraction) in the of ZIRCALOYs at two temperature (20 ЊC and 300 ЊC) and specimen is higher than some critical value. [1][2][3][4][5][6] Final heat for hydrogen content up to 4500 ppm. A novel ring tensile treatment of the material can shift this transition considertest will be used to obtain a gauge length in which the hoop ably.…”

Section: Introductionmentioning

confidence: 99%

Hydride embrittlement and irradiation effects on the hoop mechanical properties of pressurized water reactor (PWR) and boiling-water reactor (BWR) ZIRCALOY cladding tubes: Part I. Hydride embrittlement in stress-relieved, annealed, and recrystallized ZIRCALOYs at 20 °C and 300 °C

Arsène¹,

Bai

Bompard

2003

Metall Mater Trans A

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This series of articles on hydride embrittlement effect in ZIRCALOYs consists of three parts. Part I is concerned with the general influence of hydride on the mechanical properties of ZIRCALOYs at two temperatures (20 ЊC and 300 ЊC) and for hydrogen content up to 4500 ppm. The damage and fracture mechanisms were investigated by scanning electron microscopy (SEM) fracture surfaces and profiles observations. Special attention was paid to better understanding the ductile-brittle transition as a function of the hydrogen content at two temperatures. The macroscopical behavior of the hydrided ZIRCALOYs was then correlated to the damage and fracture mechanisms. In Part II, the morphology and crystallographic structure of hydrides in two ZIRCALOYs is examined at different magnifications and by different techniques (SEM, transmission electron microscopy (TEM), and X-ray diffraction) as a function of their pre-heat treatment and their hydrogen content. Part III, on the other hand, discusses the mechanical properties of the hydrides under the influence of several parameters based mainly on the SEM in-situ observations and the mechanical modeling.

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Criteria for Fracture Initiation at Hydrides in Zirconium-2.5 Pct Niobium Alloy

Cited by 63 publications

References 20 publications

On the Embrittlement of Zircaloy-4 Under RIA-Relevant Conditions

On the Embrittlement of Zircaloy-4 Under RIA-Relevant Conditions

Fracture Strength of Embedded Hydride Precipitates in Zirconium and its Alloys

Hydride embrittlement and irradiation effects on the hoop mechanical properties of pressurized water reactor (PWR) and boiling-water reactor (BWR) ZIRCALOY cladding tubes: Part I. Hydride embrittlement in stress-relieved, annealed, and recrystallized ZIRCALOYs at 20 °C and 300 °C

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