Irradiation stability and thermomechanical properties of 3D-printed SiC

Terrani, Kurt A.; Wang, Hsin; Coq, Annabelle Le; Linton, Kory; Petrie, Christian M.; Koyanagi, Takaaki; Byun, Thak Sang

doi:10.1016/j.jnucmat.2021.152980

Cited by 13 publications

(8 citation statements)

References 30 publications

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“…The Transformational Challenge Reactor (TCR) program has fully adopted the LPBF process and binder jetting CVI methods and is using them to build most of its reactor components [8,9,26]. The extensive knowledge of conventionally manufactured 316L SS combined with the well-established AM processing route makes the 316L alloy an ideal core structural material for TCR.…”

Section: Introductionmentioning

confidence: 99%

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%

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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show abstract

“…A novel methodology that combines these processes was recently developed to produce carbide ceramics. This new methodology has been leveraged to produce an inert SiC fuel matrix or structure for the TCR core [22,[26][27][28]. Specimens used for this testing and characterization task were produced using the newly established manufacturing process.…”

Section: Materials and Specimens 21 A Combined Process Of Binderjet Printing And CVImentioning

confidence: 99%

“…An infrared detector (InSb: λ = 3-5 µm) was used to record the top surface temperature transient through a sapphire window. The half-rise time (t 1/2 ) was determined by software, and thermal diffusivity (α) was calculated using Parker's equation [28], assuming no heat loss:…”

Section: Thermal Diffusivitymentioning

confidence: 99%

Mechanical and Thermophysical Properties of 3D-Printed SiC before and after Neutron Irradiation – FY21

Parish¹,

Wang²,

Trofimov³

et al. 2021

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“…• Demonstrating an agile design process to leverage AM and rapidly converge on an optimized, advanced nuclear microreactor design [9][10][11][12][13][14][15] • Advancing new reactor materials such as an yttrium hydride moderator [16][17][18][19][20][21][22], AM 316 stainless steel (316SS) [23], AM silicon carbide (SiC) [24,25], and the novel integration of uranium nitride tristructural-isotropic fuel [26] densely packed in an AM SiC matrix [27] • Developing the digital platform necessary to certify and qualify AM materials for nuclear applications [28][29][30] • Integrating and embedding spatially distributed sensors within AM materials for nuclear applications [31][32][33][34] • Progressing toward semi-autonomous reactor operation [35,36] • Evaluating and understanding radiation effects on AM SiC [37,38], 316SS [39], and integral TCR fuel compacts [27,40] In fiscal year (FY) 2021, the TCR program priorities shifted away from a nuclear reactor demonstration, but the focus on advancing ceramic AM for nuclear applications and qualifying AM components remained. Eventually, the TCR program was merged into the AMMT program and focused on the broader adoption of AM for nuclear applications compliant with American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance (NQA-1) standards.…”

Section: Introductionmentioning

confidence: 99%