In Situ and Ex Situ Observations of the Influence of Twin Boundaries on Heavy Ion Irradiation Damage Effects in 316L Austenitic Stainless Steels

  • G. Meric de BellefonEmail author
  • J. C. van Duysen
  • K. Sridharan
Conference paper
Part of the The Minerals, Metals & Materials Series book series (MMMS)


Austenitic stainless steels are widely used in Light Water Reactors (LWR) in annealed or cold-worked conditions. Their in-service temperature ranges between 270 and 370 °C, and in some regions of the reactor they can receive radiation damage levels up to about 4 dpa/year. Radiation can induce defects such as micro-compositional segregation and formation of microstructural defects such as dislocation loops, stacking fault tetrahedra, precipitates, and voids, which may undermine corrosion and mechanical properties. One consequence of these damage effects is irradiation-assisted stress corrosion cracking (IASCC), which is one of the most important degradation mechanisms in austenitic stainless steels in LWR.

It is well-known that interfaces in materials can act as defect sinks for radiation induced defects and thus can reduce neutron-induced damage (or slow down damage accumulation with dose) [1]. In addition to grain boundaries, polycrystalline austenitic stainless steels...


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Copyright information

© The Minerals, Metals & Materials Society 2019

Authors and Affiliations

  • G. Meric de Bellefon
    • 1
    Email author
  • J. C. van Duysen
    • 2
    • 3
    • 4
  • K. Sridharan
    • 1
  1. 1.Department of Nuclear EngineeringUniversity of WisconsinMadisonUSA
  2. 2.Unité Matériaux et Transformation (UMET), CNRSVilleneuve-d’AscqFrance
  3. 3.Department of Nuclear EngineeringUniversity of TennesseeKnoxvilleUSA
  4. 4.EDF—Centre de Recherche des RenardieresMoret sur LoingFrance

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