The ductile-brittle transition of a zirconium alloy due to hydrogen
Journal Article
·
· Scripta Metallurgica et Materialia; (United States)
- National Tsing Hua Univ., Hsinchu (Taiwan, Province of China)
Zircaloy-4 is generally used as fuel element cladding and an in-core structure component in light water reactors. Its mechanical properties are degraded by the presence of hydrides which are primarily formed by the absorption of excess hydrogen from corrosion reactions. Lin et al. and Bai et al. studied the mechanical properties of hydrided Zircaloy-4 alloys independently by tensile testing the smooth specimens. Both of them observed a room temperature ductile-brittle transition on the reduction of area, when the hydrogen content in the specimen is higher than some critical value. Their definition of 'ductile-brittle transition' is the abrupt change of the value of reduction of area, which is somewhat different from the conventional definition determined from impact tests. Since the effect of hydrogen on reduction of area is more distinct than on elongation, the reduction of area has been used as a parameter to exhibit hydrogen effect. In this paper the authors followed the definition of Lin and Bai. The observations of ductile-brittle transitions mentioned above were mostly performed using smooth tensile specimens in standard tensile tests. In the present study, notch tensile tests were performed on Zircaloy-4 alloys at various temperatures up to 300[degree]C in hydrided conditions to investigate the change of ductile-brittle transition due to notch effect. Additionally, no previous study has been done on Zircaloy-4 alloy to determine whether the ductile-brittle transition is existed in a hydrogen gas environment. To determine this phenomenon, notched, uncharged Zircaloy-4 specimens were tensile tested in a hydrogen gas environment at various temperatures up to 200[degree]C. They found that the ductile-brittle transition exists not only for the hydrided Zircaloy-4 but also for Zircaloy-4 tested in the hydrogen gas environment.
- OSTI ID:
- 6442189
- Journal Information:
- Scripta Metallurgica et Materialia; (United States), Journal Name: Scripta Metallurgica et Materialia; (United States) Vol. 28:12; ISSN 0956-716X; ISSN SCRMEX
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
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Boiling Water Cooled
210200 -- Power Reactors
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36 MATERIALS SCIENCE
360102 -- Metals & Alloys-- Structure & Phase Studies
ALLOY-ZR98SN-4
ALLOYS
CHROMIUM ADDITIONS
CHROMIUM ALLOYS
CORROSION RESISTANT ALLOYS
DATA
DUCTILE-BRITTLE TRANSITIONS
ELEMENTS
EMBRITTLEMENT
EXPERIMENTAL DATA
FUEL CANS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HYDROGEN
HYDROGEN EMBRITTLEMENT
INFORMATION
IRON ADDITIONS
IRON ALLOYS
MATERIALS
MECHANICAL PROPERTIES
NONMETALS
NUMERICAL DATA
REACTOR MATERIALS
REACTORS
TEMPERATURE RANGE
TEMPERATURE RANGE 0273-0400 K
TEMPERATURE RANGE 0400-1000 K
TIN ALLOYS
WATER COOLED REACTORS
ZIRCALOY
ZIRCALOY 4
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
36 MATERIALS SCIENCE
360102 -- Metals & Alloys-- Structure & Phase Studies
ALLOY-ZR98SN-4
ALLOYS
CHROMIUM ADDITIONS
CHROMIUM ALLOYS
CORROSION RESISTANT ALLOYS
DATA
DUCTILE-BRITTLE TRANSITIONS
ELEMENTS
EMBRITTLEMENT
EXPERIMENTAL DATA
FUEL CANS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HYDROGEN
HYDROGEN EMBRITTLEMENT
INFORMATION
IRON ADDITIONS
IRON ALLOYS
MATERIALS
MECHANICAL PROPERTIES
NONMETALS
NUMERICAL DATA
REACTOR MATERIALS
REACTORS
TEMPERATURE RANGE
TEMPERATURE RANGE 0273-0400 K
TEMPERATURE RANGE 0400-1000 K
TIN ALLOYS
WATER COOLED REACTORS
ZIRCALOY
ZIRCALOY 4
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS