Low stress creep behavior of Zircaloy-2 vis-a-vis zirconium

Journal Article · Fri Feb 14 23:00:00 EST 1992 · Scripta Metallurgica; (United States)

OSTI ID:5633816

Prasad, N ^[1]; Malakondaiah, G ^[2]; Rao, P R ^[3]

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1. Dept. of Metallurgical Engineering, Banaras Hindu Univ., Varanasi-221005 (IN)
2. Defense Metallurgical Research Lab., Hyderabad-500258 (IN)
3. Dept. of Science and Technology, Government of India, New Delhi-110016 (IN)

Zircaloy-2 is a popular zirconium base nuclear fuel cladding material having tin as a major alloying additions along with Fe, Ni and Cr, each around 0.1 wt %. Low stress creep behavior of a-zirconium and zircaloy-2 has been studied employing helical spring specimens. Detailed experimental results are reported and analyzed elsewhere. Here we present data on a-Zr and zircaloy-2 obtained under identical test conditions of grain size and test temperature with the aim of bringing out the effect of alloying additions to zirconium through the results obtained on the low stress creep behavior of zircaloy-2. The high purity zirconium contained, by ppm, 110 Sn, 350 Fe, 37 Ni, 150 Cr and 880 O while zircaloy-2 contained, by wt %, 1.3 Sn, 0.16 Fe, 0.06 Cr, 0.04 Ni and 0.11 O. Helical spring specimens of wire radius 500{mu}m and spring radius 0.012m were employed in both the cases. Spring specimens were annealed under vacuum (Zr at 923K/1.8ks and zircaloy-2 at 1113K/2.7ks) while on mandrel to develop a matching grain size in both the materials. The mean linear intercept grain size was found to be 16.8{mu}m in alpha zirconium and 16.6{mu}m in zircaloy-2. Close coiled helices were creep tested at 873K using the set-up and procedure described elsewhere. The maximum applied stress was 5.3MN/m{sup 2} in the case of zirconium while the same in the case of zircaloy-2 was 3.9MN/m{sup 2}. Deflection versus time curves for tests conducted at 873K for both the materials are shown in this paper.

• OSTI ID: 5633816
• Journal Information: Scripta Metallurgica; (United States), Journal Name: Scripta Metallurgica; (United States) Vol. 26:4; ISSN 0036-9748; ISSN SCRMB
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
220504 -- Nuclear Reactor Technology-- Environmental Aspects-- Design Basis & Hypothetical Accidents-- (1992-)
36 MATERIALS SCIENCE
360202 -- Ceramics
Cermets
& Refractories-- Structure & Phase Studies
360203* -- Ceramics
Cermets
& Refractories-- Mechanical Properties
ALLOY-ZR98SN-2
ALLOYS
CHEMICAL COMPOSITION
CHEMICAL REACTIONS
CHROMIUM ADDITIONS
CHROMIUM ALLOYS
CLADDING
CORROSION
CORROSION RESISTANT ALLOYS
CREEP
CRYSTAL STRUCTURE
DATA
DEPOSITION
ELECTRON MICROSCOPY
ELEMENTS
ENVIRONMENTAL EFFECTS
EXPERIMENTAL DATA
FABRICATION
GRAIN BOUNDARIES
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HEAT TREATMENTS
INFORMATION
IRON
IRON ADDITIONS
IRON ALLOYS
MATERIALS
MATERIALS TESTING
MATERIALS WORKING
MECHANICAL PROPERTIES
METALS
MICROSCOPY
MICROSTRUCTURE
NICKEL
NICKEL ADDITIONS
NICKEL ALLOYS
NUMERICAL DATA
PRESSURE TUBES
REACTOR COMPONENTS
STRAINS
STRESS CORROSION
SURFACE COATING
TESTING
THERMOMECHANICAL TREATMENTS
TIN ALLOYS
TRANSITION ELEMENTS
TRANSMISSION ELECTRON MICROSCOPY
TUBES
ZIRCALOY
ZIRCALOY 2
ZIRCONIUM ALLOY
ZIRCONIUM BASE ALLOYS