Austenitic Stainless Steels for Fast Reactors -Irradiation Experiments, Property Evaluation and Microstructural Studies

Karthik, V.; Murugan, S.; Parameswaran, P.; Venkiteswaran, C.N.; Gopal, K.A.; Muralidharan, N.G.; Saroja, S.; Kasiviswanathan, K.V.

doi:10.1016/j.egypro.2011.06.033

Cited by 29 publications

(4 citation statements)

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“…Therefore, the core materials for any high-performance reactor will require excellent high-temperature mechanical properties, high-radiation resistance, and high-corrosion resistance, in addition to the feasibility of manufacturing processes. The 300 series austenitic stainless steels (SSs) are widely used in the reactor core and coolant system components of current nuclear power plants and have been among the key candidate structural materials for advanced future reactors, including sodium-cooled fast reactors and fusion energy systems [2][3][4][5][6]. The austenitic alloys have been consistently used for nuclear applications because they provide a good combination of strength, ductility, toughness, and oxidation-corrosion resistance over an exceptionally wide temperature range [7].…”

Section: Introductionmentioning

confidence: 99%

“…The TCR core structural materials are exposed to the He gas coolant with an outlet temperature of ~500°C and to the high-flux neutron irradiation on the order of tens of displacements per atom. Because the materials degradation caused by the coolant corrosion, void swelling, and He embrittlement [2][3][4][5][6] will be insignificant in the TCR environment, some of the high-temperature mechanical properties-including deformation and fracture, creep, and creep-fatigue properties-must be evaluated for the assessment of the core materials [21].…”

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

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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: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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

Collins¹,

Coq²,

Linton³

et al. 2021

View full text Add to dashboard Cite

show abstract

“…But these alloys are susceptible to irradiation hardening and embrittlement due to the formation of voids, dislocation loops, and stacking fault tetrahedra [ 3 , 4 , 5 , 6 , 7 , 8 , 9 , 10 ]. These alloys also exhibit poor creep resistance, especially at elevated temperatures as would be required in advanced Generation IV nuclear reactors such as supercritical water-cooled reactors (SCWR) and sodium-cooled fast reactors (SFR) [ 11 , 12 , 13 , 14 ]. Consequently, 316 SS has been modified with a higher Ni/Cr ratio and higher Si and Ti concentrations for enhanced radiation tolerance [ 15 , 16 , 17 , 18 , 19 ] and high-temperature creep strength [ 20 ].…”

Section: Introductionmentioning

confidence: 99%

Nanocluster Evolution in D9 Austenitic Steel under Neutron and Proton Irradiation

Mullurkara,

Bejawada,

Sen

et al. 2023

Materials

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Austenitic stainless steel D9 is a candidate for Generation IV nuclear reactor structural materials due to its enhanced irradiation tolerance and high-temperature creep strength compared to conventional 300-series stainless steels. But, like other austenitic steels, D9 is susceptible to irradiation-induced clustering of Ni and Si, the mechanism for which is not well understood. This study utilizes atom probe tomography (APT) to characterize the chemistry and morphology of Ni–Si nanoclusters in D9 following neutron or proton irradiation to doses ranging from 5–9 displacements per atom (dpa) and temperatures ranging from 430–683 °C. Nanoclusters form only after neutron irradiation and exhibit classical coarsening with increasing dose and temperature. The nanoclusters have Ni3Si stoichiometry in a Ni core–Si shell structure. This core–shell structure provides insight into a potentially unique nucleation and growth mechanism—nanocluster cores may nucleate through local, spinodal-like compositional fluctuations in Ni, with subsequent growth driven by rapid Si diffusion. This study underscores how APT can shed light on an unusual irradiation-induced nanocluster nucleation mechanism active in the ubiquitous class of austenitic stainless steels.

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“…Karthik et al [ 25 ] carried out the systematic microstructural studies of the austenitic stainless steel irradiation performance. The following salient features were analyzed: (a) behavior of SS316 at different fluence levels and (b) irradiation experiments of classical and modified D9 versions.…”

Section: Introductionmentioning

confidence: 99%

Stability Improvement of Flexible Shallow Shells Using Neutron Radiation

et al. 2020

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Microelectromechanical systems (MEMS) are increasingly playing a significant role in the aviation industry and space exploration. Moreover, there is a need to study the neutron radiation effect on the MEMS structural members and the MEMS devices reliability in general. Experiments with MEMS structural members showed changes in their operation after exposure to neutron radiation. In this study, the neutron irradiation effect on the flexible MEMS resonators’ stability in the form of shallow rectangular shells is investigated. The theory of flexible rectangular shallow shells under the influence of both neutron irradiation and temperature field is developed. It consists of three components. First, the theory of flexible rectangular shallow shells under neutron radiation in temperature field was considered based on the Kirchhoff hypothesis and energetic Hamilton principle. Second, the theory of plasticity relaxation and cyclic loading were taken into account. Third, the Birger method of variable parameters was employed. The derived mathematical model was solved using both the finite difference method and the Bubnov–Galerkin method of higher approximations. It was established based on a few numeric examples that the irradiation direction of the MEMS structural members significantly affects the magnitude and shape of the plastic deformations’ distribution, as well as the forces magnitude in the shell middle surface, although qualitatively with the same deflection the diagrams of the main investigated functions were similar.

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Austenitic Stainless Steels for Fast Reactors -Irradiation Experiments, Property Evaluation and Microstructural Studies

Cited by 29 publications

References 1 publication

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)

Nanocluster Evolution in D9 Austenitic Steel under Neutron and Proton Irradiation

Stability Improvement of Flexible Shallow Shells Using Neutron Radiation

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