Mechanical properties of neutron irradiated 316L stainless steel additively manufactured by laser powder bed fusion: Effect of post-manufacturing heat treatments

Milan Konstantinovic, Didier Bardel

    Research outputpeer-review

    Abstract

    This study was conducted on 316L material produced by additive manufacturing, laser powder bed fusion process, with four heat treatment conditions: stress relief, low and high temperature solution annealing and hot isostatic pressing. The materials were neutron irradiated in the Belgian material testing reactor BR2 at 300 0C and with a dose up to about 4 dpa. Charpy impact and tensile test results revealed that all materials gradually harden and loose ductility with increasing dose. At the highest dose, the decrease of absorbed energy is the largest for the high temperature solution annealed sample, of about 60% of its initial value, while the stress relieved material exhibits the smallest reduction, of about 20% of its initial value. Yield stress is observed to be more sensitive to neutron dose than tensile stress. The bigger the initial hardening, the smaller the irradiation hardening. In addition, all 316L additive manufactured materials reach the same hardening level of about 600 MPa at a dose of about 4 dpa. Most probably, initial contribution to hardening, originating from dislocation density differences between material batches, is gradually overtaken by the irradiation induced defect contribution that becomes the most dominant contribution to hardening at 4 dpa and explains the convergence of yield stress values. Depending of the dose and the heat treatment conditions, the yield stress values of all batches lay below the material constitutive model of irradiated solution annealed 316L stainless steel. These results could potentially indicate a beneficial effect of the additive manufacturing process to irradiation hardening resistance of 316L stainless steel.
    Original languageEnglish
    Article number155662
    Number of pages10
    JournalJournal of Nuclear Materials
    Volume607
    DOIs
    StatePublished - Mar 2025

    ASJC Scopus subject areas

    • Nuclear and High Energy Physics
    • General Materials Science
    • Nuclear Energy and Engineering

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