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Irradiation Induced Damage of Fe-10%Cr Under Uniaxial Pressure

[+] Author Affiliations
Mohammad Abu-Shams, Ishraq Shabib

Central Michigan University, Mt. Pleasant, MI

Paper No. IMECE2015-51686, pp. V009T12A073; 7 pages
  • ASME 2015 International Mechanical Engineering Congress and Exposition
  • Volume 9: Mechanics of Solids, Structures and Fluids
  • Houston, Texas, USA, November 13–19, 2015
  • Conference Sponsors: ASME
  • ISBN: 978-0-7918-5752-6
  • Copyright © 2015 by ASME


Structural materials of next generation nuclear reactors are expected to experience severe operating conditions including intense heat, high irradiation doses, thermal and mechanical stresses, and corrosive environments, which would potentially degrade material properties and impose severe threat to structural integrity. For example, high irradiation doses cause the evolution of displacement cascades, consisting of point defects, which lead to void swelling, irradiation creep, irradiation assisted stress corrosion cracking, and embrittlement. Over the last several decades, extensive computational researches have been conducted to study displacement cascades and generate defect statistics over a wide range of irradiation doses and temperatures for pure materials, primarily Fe. However, very limited data can be found to determine cascade evolution and defect statistics of Fe-alloys under pressure. In this work, large-scale molecular dynamics simulations have been performed to study displacement cascade and generate defect statistics of Fe-10%Cr alloy under uniaxial pressure. The selection of the material is based on the fact that Fe and Cr are the two major alloying elements of Ferritic-martensitic steels, which have shown promise to be a candidate material for future generation reactors due to high temperature stability and reduced swelling under irradiation. The simulated material is built from a single crystal Fe model of [130], [310], and [001] orientation, and randomly substituting Fe atoms by Cr. Empirical EAM potential has been used to define interatomic interactions. Irradiation simulations are performed for doses between 2–15keV, and pressure ranges between −10,000 bars to +10,000 bars applied along the x-direction. Simulation temperature is kept at a minimum, e.g. 15K, to minimize thermal influences. Displacement cascades are generated by imparting kinetic energy to a lattice atom (i.e. primary-knock-on-atom, PKA) along an arbitrary crystallographic direction (i.e. the diagonal direction of the simulation cell). Point defects are identified using the Wigner-Seitz method. Upon collision, the PKA atom displaces the surrounding atoms from their perfect lattice cites and causes a rapid increase in defect numbers. As the imposed energy is dissipated through the crystal, the displaced atoms recombine with the vacancies and the defect numbers gradually decrease and become stable. The cascade structure shows the presence of the vacancies at the core of the cascades surrounded by the interstitials. The number of defects increases almost linearly with increasing the irradiation dose for any pressure. The effect of pressure is found to be more profound within the intermediate pressure range, e.g. between −100 to +1000 bar, within which the number of point defects continually decreases as the pressure changes from tension to compression. The trend is found to be consistent for the whole PKA energy range. Point defects are also found to form defect clusters. The common neighbor analyses haves been performed to determine the structure of the clustered defects. It has been revealed that the defect clusters are of cubic diamond type. Additional analyses are currently under progress to evaluate the influence of pressure on cascade volume, point defect composition, and cluster composition.

Copyright © 2015 by ASME



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