Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Byun, Thak-Sang; Garrison, Ben; Mcalister, Michael; Chen, Xin; Gussev, Maxim N.; Lach, Timothy G.; Coq, Annabelle Le; Linton, Kory; Joslin, Chase; Carver, Keith; List, F.A.; Dehoff, Ryan R.; Terrani, Kurt A.

doi:10.1016/j.jnucmat.2021.152849

Cited by 52 publications

(24 citation statements)

References 33 publications

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“…Postbuild annealing treatments can modify the as-built microstructure and result in mechanical behavior changes, progressing toward that of the fully annealed WT 316L. Along with the temperature dependence of tensile property parameters, the effects of postbuild annealing treatments on those parameters are discussed in this section [10,11]. Figures 5 and 6 compare the baseline strength and ductility data determined from the load-displacement data.…”

Section: Temperature Dependence Of Strength and Ductilitymentioning

confidence: 99%

“…Recent research efforts confirmed that austenitic SSs are highly suitable for additive manufacturing of complex shaped reactor components [8][9][10][11]. This is likely because the fast cooling that occurs during the AM process prevents the formation of the high-temperature ferrite (i.e., δ-ferrite) phase during cooling, which is metastable and undergoes degradation at high temperatures caused by phase decomposition and segregation.…”

Section: Introductionmentioning

confidence: 99%

“…Neutron or ion irradiation could also significantly affect the behaviors (i.e., shortened creep life and reduced fracture toughness) observed in AM alloys [11,[23][24][25].…”

Section: Introductionmentioning

confidence: 99%

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Mechanical Properties of Additively Manufactured 316L Stainless Steel Before and After Neutron Irradiation (FY21)

Collins¹,

Coq²,

Linton³

et al. 2021

Self Cite

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This report presents the materials property data of additively manufactured (AM) 316L stainless steel (SS) accumulated for the assessment of core materials in the Transformational Challenge Reactor (TCR) program. The TCR manufacturing approach includes using the laser powder bed fusion (LPBF) method for metallic (316L and Inconel 718) components. To assess the mechanical performance of printed components in reactor-relevant conditions and build a property database for the AM materials, mechanical tests and evaluations were performed before and after neutron irradiation. Miniature tensile specimens were irradiated in the High Flux Isotope Reactor to 0.2, 2, 8, and 10 dpa at target temperatures of 300 and 600°C. Postirradiation evaluation for the 0.2 and 2 dpa specimens was performed during FY20 and FY21, and the results are presented and discussed in this document.To obtain the baseline mechanical property data, uniaxial tension testing over a wide temperature range from room temperature (RT) to 600°C was performed for the same materials that were irradiated, including AM 316L SS in the as-built, stress-relieved, and solution-annealed conditions, as well as reference wrought (WT) 316L SS. Regardless of the postbuild heat treatment, the AM 316L SS showed higher strength than the WT 316L SS, but they had similar ductility. A statistical treatment of RT tensile data indicated that variations in the strength and ductility datasets of AM 316L steels were smaller than or similar to those of WT 316L SS. Additionally, 2D mapping of tensile properties for the stress-relieved AM plates also showed clear location dependence of the properties but with limited magnitudes.Postirradiation tensile testing was conducted at RT, 300, 500, and 600°C for selected irradiation conditions. Additional tests were performed near the measured irradiation temperatures of 260 and 390°C for checking the impact of the difference between the targeted and actual irradiation temperatures. Neutron irradiation induced significant changes in the mechanical behavior of the AM SSs, including hardening and softening. Although the as-built 316L SS tested at 300°C after irradiation to 2 dpa at 390°C showed unstable plastic deformation (i.e., necking) immediately after yielding, the overall property changes of the as-printed alloy became much less significant after higher temperature (610 and 690°C) irradiations or when tested at different temperatures. Irradiation-induced ductilization was also observed in the higher strength materials (i.e., in as-built and stress-relieved conditions) after higher temperature (>600°C) irradiations. The strength change after irradiation was generally smaller in the relatively stronger materials (i.e., the as-built and stress-relieved AM SSs) than in the solution-annealed AM and WT SSs. Overall, these relatively lower strength 316L SSs retained higher ductility in the irradiation conditions tested, but the stronger 316L SSs demonstrated a similar level of ductility after higher temperature irradiations. For the AM 316L mate...

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Section: Temperature Dependence Of Strength and Ductilitymentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Mechanical Properties of Additively Manufactured 316L Stainless Steel Before and After Neutron Irradiation (FY21)

Collins¹,

Coq²,

Linton³

et al. 2021

Self Cite

View full text Add to dashboard Cite

show abstract

“…The AM aluminum alloy 6061 specimens that were subjected to post-weld heat treatment and hot isostatic pressing were irradiated in the high flux isotope reactor up to 17.3 dpa at ∼ 70 °C while in contact with water using perforated rabbit capsules [56]. The AM 316L stainless steel in the as-built, stress-relieved and solution-annealed conditions were irradiated at the high flux isotope reactor to 0.2 and 2 displacements per atom at 300 °C and 600 °C [57]. These irradiation experiments revealed that:…”

Section: Irradiation Damage Behaviorsmentioning

confidence: 99%

A Review of the Latest Developments in the Field of Additive Manufacturing Techniques for Nuclear Reactors

Zhang

et al. 2022

Crystals

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This review paper provides insights the into current developments in additive manufacturing (AM) techniques. The comprehensive presentations about AM methods, material properties (i.e., irradiation damage, as-built defects, residual stresses and fatigue fracture), experiments, numerical simulations and standards are discussed as well as their advantages and shortages for the application in the field of nuclear reactor. Meanwhile, some recommendations that need to be focused on are presented to advance the development and application of AM techniques in nuclear reactors. The knowledge included in this paper can serve as a baseline to tailor the limitations, utilize the superiorities and promote the wide feasibilities of the AM techniques for wide application in the field of nuclear reactors.

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“…316L austenitic stainless steel is widely used in petroleum, chemical, pharmaceutical, and nuclear power fields because of its high anti‐creep strength, good corrosion resistance, and excellent sensitization resistance 19 . Most of 316L equipment serve under harsh working conditions such as high temperature and high pressure and fluctuating load 20 . In engineering practice, the maximum service temperature of 316L steel has reached 650°C, which is close to its ultimate service temperature 21 .…”

Section: Introductionmentioning

confidence: 99%

Experimental and theoretical research on creep‐fatigue behaviors of 316L steel with and without 650°C thermal aging

Mao

Yang

Zhu

et al. 2022

Fatigue Fract Eng Mat Struct

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In this paper, tensile test and creep‐fatigue test at 650°C were carried out for 316L steel with and without prior thermal aging. The effect of aging on creep‐fatigue interaction of 316L steel is analyzed in detail. Microscopic observations are performed on fatigue and creep‐fatigue specimens to study the microstructure characteristics affected by aging, creep, and fatigue. Actually, three strain amplitude loadings of ±0.3%, ±0.4%, and ±0.5% are adopted in creep‐fatigue tests. The creep‐fatigue behaviors of unaged and aged 316L steel are investigated in depth for varying strain amplitude. The creep‐fatigue constitutive model is constructed by using Chaboche mixed softening and hardening model combined with strain hardening creep. Life prediction models are compared and used to do prediction. The applicability of different models is recommended. The accuracy of the models is analyzed by comparing with the test data. The agreement is achieved between theoretical and experimental predictions.

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Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Cited by 52 publications

References 33 publications

Mechanical Properties of Additively Manufactured 316L Stainless Steel Before and After Neutron Irradiation (FY21)

Mechanical Properties of Additively Manufactured 316L Stainless Steel Before and After Neutron Irradiation (FY21)

A Review of the Latest Developments in the Field of Additive Manufacturing Techniques for Nuclear Reactors

Experimental and theoretical research on creep‐fatigue behaviors of 316L steel with and without 650°C thermal aging

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