Abdullah S. Alomari
King Abdulaziz City for Science and Technology
Ultimate tensile strengthDynamic strain agingEmbrittlementDeformation (engineering)Strain rateCreepStrain hardening exponentDislocationComposite materialAmplitudeTensile testingMetallurgyMaterials scienceStructural materialStress relaxationDeformation (meteorology)Deformation mechanismDamage toleranceWork hardeningAustenitic stainless steelLow-cycle fatigueAlloyPressurized water reactorHold timeCreep fatigueElongationZirconium alloyCladding (metalworking)MicromechanicsReliability (statistics)Grain boundarySodiumNuclear reactorDuctilityStrain (chemistry)Forensic engineeringStress (mechanics)Activation energyAustenite
12Publications
4H-index
53Citations
Publications 12
Newest
#1Abdullah S. Alomari (KACST: King Abdulaziz City for Science and Technology)
#1Abdullah S. Alomari (KACST: King Abdulaziz City for Science and Technology)
Last. Abdullah S. Alomari (KACST: King Abdulaziz City for Science and Technology)H-Index: 4
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Austenitic stainless steels are widely employed in various applications due to many desirable properties such as excellent mechanical properties and corrosion/oxidation resistance. Serrated yieldin...
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#1Zeinab Y. Alsmadi (NCSU: North Carolina State University)H-Index: 2
#2Abdullah S. Alomari (NCSU: North Carolina State University)H-Index: 4
Last. Korukonda L. Murty (NCSU: North Carolina State University)H-Index: 27
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Abstract To understand high temperature creep-fatigue interaction of the Alloy 709, strain-controlled low-cycle fatigue (LCF) tests were performed at strain ranges varying from 0.3% to 1.2% with fully reversible cycle of triangular waveform at 750 °C. In addition, different hold times of 60, 600, 1800 and 3600 s were introduced at the maximum tensile strain to investigate the effect of creep damage on the fatigue-life at strain range of 1% at 750 °C. The creep-fatigue life and the number of cycl...
10 CitationsSource
#1Zeinab Y. Alsmadi (NCSU: North Carolina State University)H-Index: 2
#2Abdullah S. Alomari (NCSU: North Carolina State University)H-Index: 4
Last. Korukonda L. Murty (NCSU: North Carolina State University)H-Index: 27
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Preliminary investigation on mechanical properties of a Nb-strengthened and nitrogen-stabilized Fe-25wt.%Ni-20Cr (Alloy 709) advanced austenitic stainless steel suggests that it might be a potential candidate for Sodium-Cooled Fast Reactor (SFR), which has higher technology readiness level for deployment. However, the creep-fatigue deformation behaviour is unknown for this alloy. To understand high temperature creep-fatigue interaction of the Alloy 709, strain-controlled low-cycle fatigue (LCF) ...
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#1Zeinab Y. AlsmadiH-Index: 2
Last. K. Murty
view all 4 authors...
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#1Abdullah S. Alomari (NCSU: North Carolina State University)H-Index: 4
#2Nilesh Kumar (UA: University of Alabama)H-Index: 18
Last. Korukonda L. Murty (NCSU: North Carolina State University)H-Index: 27
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Abstract Understanding serrated yielding behavior resulting from dynamic strain aging (DSA) is essential for design and safety considerations. In this work, uniaxial tensile tests were carried out at temperatures ranging from 298 to 1073 K and strain rates 10 −5 – 10 −3 s −1 followed by microstructural examination of Fe-25Ni-20Cr (wt%) austenitic stainless steel (Alloy 709), a candidate structural material for Sodium-cooled Fast Reactors. Serrated yielding was found to occur in this alloy in two...
6 CitationsSource
#1Abdullah S. Alomari (NCSU: North Carolina State University)H-Index: 4
#2Nilesh Kumar (UA: University of Alabama)H-Index: 18
Last. Korukonda L. Murty (NCSU: North Carolina State University)H-Index: 27
view all 3 authors...
Understanding creep properties and microstructural evolution for candidate materials of the next-generation nuclear reactors is essential for design and safety considerations. In this work, creep tests were carried out at temperatures ranging from 973 to 1073 K and stresses 40 to 275 MPa followed by microstructural examinations of a Fe-20Cr-25Ni (mass pct) austenitic stainless steel (Alloy 709), a candidate structural material for the Sodium-cooled Fast Reactors. The apparent stress exponent and...
13 CitationsSource
1 Citations
#1Nilesh Kumar (NCSU: North Carolina State University)H-Index: 18
#2Abdullah S. Alomari (NCSU: North Carolina State University)H-Index: 4
Last. Korukonda L. Murty (NCSU: North Carolina State University)H-Index: 27
view all 3 authors...
Gen-IV nuclear reactors require materials to operate under much harsher conditions necessitating the development of advanced structural materials. Sodium-cooled Fast Reactor (SFR) is a Gen-IV nuclear reactor with a high level of technology readiness. Alloy 709, Fe-25Ni-20Cr (wt%) alloyed with Nb and stabilized with nitrogen, is an advanced austenitic stainless steel having promising set of properties for SFRs. However, the creep-fatigue deformation behavior is unknown for this alloy. This work f...
1 CitationsSource
#1Abdullah S. Alomari (NCSU: North Carolina State University)H-Index: 4
#2Nilesh Kumar (NCSU: North Carolina State University)H-Index: 18
Last. Korukonda L. Murty (NCSU: North Carolina State University)H-Index: 27
view all 3 authors...
Abstract Contrary to the commonly observed embrittlement during dynamic strain aging, we report in this note distinct enhancement in ductility in a Nb-containing and nitrogen stabilized Fe-25(wt%)Ni-20Cr austenitic stainless steel (Alloy 709) at temperatures from 623 K to 873 K at 10−4 s−1 where serrated flow is noted. This observation is rationalized in terms of the influence of strain hardening parameters and strain rate sensitivity on uniform elongation and ductility respectively.
16 CitationsSource
#1Nilesh Kumar (NCSU: North Carolina State University)H-Index: 18
#2Abdullah S. Alomari (NCSU: North Carolina State University)H-Index: 4
Last. Korukonda L. Murty (NCSU: North Carolina State University)H-Index: 27
view all 3 authors...
Abstract Understanding micromechanics of plastic deformation of existing materials is essential for improving their properties further and/or developing advanced materials for much more severe load bearing applications. The objective of the present work was to understand micromechanics of plastic deformation of Zircaloy-4, a zirconium-based alloy used as fuel cladding and channel (in BWRs) material in nuclear reactors. The Zircaloy-4 in recrystallized (at 973 K for 4 h) condition was subjected t...
2 CitationsSource