Search NASA⌕ Search

Engineering topics

Gussev, Maxim N.

Publications and source records attributed to Gussev, Maxim N..

54 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

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

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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Microstructure and Mechanical Performance of the Friction Stir Welds Made on Neutron-Irradiated Steel with Helium

This report describes new experimental results on the microstructure and mechanical performance of the friction stir welds made on neutron-irradiated steel with helium. The report focuses on helium-related issues, specifically, helium-induced degradation in the welded joint, aiming to repair irradiated components of nuclear power plants (NPPs). The friction stir welds analyzed here were produced at Oak Ridge National Laboratory previously, and limited characterization work was performed, mostly addressing the presence or absence of macroscopic cracks. The present work attempts to perform a more comprehensive study to assess microstructure conditions, grain size, plastic strain gradients, the morphology of the helium-induced damage, and deformation behavior.

42 ENGINEERING↗
DOE OSTI2021

Ultrasonic additive manufacturing of zirconium: Pilot results

Ultrasonic additive manufacturing (UAM) was successfully applied to the Zirconium material system to create a three-dimensional component prototype. The UAM process resulted in grain size refinement and significant localized deformation. Electron backscatter diffraction analysis revealed activated slip systems only applicable at high deformation temperatures, while texture analysis showed a decrease in measurable texture of the UAM build in comparison to the initial Zr-foil. Average tensile strengths in the X (sonotrode travel), Y (vibration), and Z (build) directions averaged 435 MPa, 458 MPa, and 359 MPa, respectively, demonstrating the viability of UAM for Zr-based materials. Furthermore, the delamination of some Z specimens along foil boundaries during loading suggests an interplay between interfacial Ti impurities introduced during welding and the spatial dependence of weld quality for this material system.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Complexity of deformation mechanism in neutron-irradiated 304L austenitic stainless steel at microstructural scale

The deformation mechanisms of neutron-irradiated 304 L stainless steel have been investigated using the in-situ electron backscattered diffraction (EBSD) technique. To decipher the role of irradiation, a detailed analysis was performed on irradiated miniature samples and compared with nonirradiated samples. Furthermore, at three levels of engineering strain, —0.0%, 4.8%, and 14.5%—microstructural features as well as EBSD were recorded at the identical location during in-situ deformation testing. Post-mortem analysis of the EBSD data revealed that the neutron-irradiated samples had higher fractions of the slip bands even at low amounts of deformation. Along with the slip bands, the neutron-irradiated sample showed the twins and martensitic phase formed almost parallel to the slip bands. In addition, the image results also indicated that the formation of twins and phases associated with the slip bands. A detailed mechanistic understanding has been investigated to identify the (111)γ planes orientation with respect to the loading direction associated with the formation of twins and martensite. The newly formed deformation twins and martensite were also found to maintain this relationship when they existed together.

304L austenitic stainless steel↗
DOE OSTI2021

Quantifying adherence of oxide scales on steels exposed to high temperature and pressure steam

Oxide scale exfoliation is a major concern in fossil fuel power generation because it can cause tube blockages and erode valves and steam turbine components downstream. There is still considerable scientific and commercial interest to improve the mechanistic understanding of oxide failures by developing models to predict exfoliation and the extent of tube blockage as a function of operating conditions and component geometries. Tensile testing inside a scanning electron microscope was conducted on ferritic–martensitic and austenitic steel specimens with the steam side (Fe,Cr)-rich oxides grown after exposures for up to 1000 h in steam with ~100 ppb O 2 at 276 bar and 550°C. Multiple oxide layer cracks and delamination events were observed and analyzed in detail during the tests. Results from the testing agreed well with earlier observations that had identified the failure location at the outer–inner oxide layer for all tested materials. Calculated adhesion energies identified the outer–inner oxide interface of alloy 347HFG as the weakest interface.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Influence of neutron irradiation on Al-6061 alloy produced via ultrasonic additive manufacturing

Samples of aluminum alloy 6061 produced via ultrasonic additive manufacturing (UAM) were irradiated in the High Flux Isotope Reactor (HFIR) up to 17.3 dpa at ~70°C while in contact with water using perforated rabbit capsules. The irradiation campaign included as-received (AR) material, specimens subjected to various post-weld heat treatments (PWHTs, including hot isostatic pressing [HIP]), and reference (wrought) alloy samples. Mechanical tensile tests, accompanied by digital image correlation (DIC) analysis, fractography, and metallography, were performed as a part of the post-irradiation evaluation. The X- and Y-specimens (i.e., oriented in the sonotrode moving and vibration directions, respectively) showed pronounced radiation hardening and ductility decrease. Specific serration flow behavior and propagation of deformation bands were observed under various material conditions up to 3.5 dpa but disappeared at 17.3 dpa. In all cases, the fracture mechanism of X- and Y-specimens was ductile; ductile dimples dominated the fracture surface. Irradiated X- and Y-specimens showed good performance, regardless of material conditions (AR or PWHT). The performance of Z-specimens oriented in the build direction was strongly dependent on the PWHT. The AR and aged specimens showed fracture stress decrease with dose, and they experienced fracture under irradiation after 3.5 dpa; specimen cross section analysis revealed specific interface degradation that was likely related to corrosion. Recrystallization significantly improved in-reactor performance. Finally, HIP suppressed interface degradation due to recrystallization and pore removal, which led to good in-reactor performance for Z-specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation

FeCrAl alloys have been extensively investigated over the past decade as a candidate material for accident-tolerant fuel cladding in light water reactors. Here, we have completed a first of its kind study where Al and Cr concentrations are varied systematically under neutron irradiation to elucidate the post irradiation microstructural and mechanical response as a direct result of alloying content. Neutron irradiations were performed on alloys with composition Fe-(10-13)Cr-(5-7)Al in wt.% at temperatures of 214, 357, and 557°C to a dose of ~1.8 dpa (displacements per atom). Dislocation loop sizes and dispersion characteristics did not show strong compositional dependence. However, there were noticeable effects of composition on the dispersion of Cr-rich α'-precipitates, particularly a decrease in number density with increasing Al content or a decrease in Cr content. The present results confirm previous studies indicating the Cr concentration in these precipitates is lower than that expected in binary FeCr alloys and that Al can act as a destabilizing alloying element for the deleterious α'-phase.

36 MATERIALS SCIENCE↗
DOE OSTI2021

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

Fabrication of nuclear reactor components using additive manufacturing (AM) methods is now a practical option since the AM technologies have advanced to allow for building of complex parts with high quality materials. To assess the mechanical performance of printed components in reactor-relevant conditions and to build a property database for the AM 316L stainless steel (SS), mechanical testing and characterization were performed before and after neutron irradiation. In this work, miniature tensile specimens were irradiated at the High Flux Isotope Reactor (HFIR) to 0.2 and 2 displacements per atom (dpa) at 300 and 600°C. The AM 316L SS was tested in the as-built, stress-relieved, and solution-annealed conditions, and the wrought (WT) 316L SS in solution-annealed condition as a reference alloy. The baseline test result showed that the AM 316L SS, regardless of the post-build heat treatment, had higher strength than the WT 316L SS, but similar ductility. Post-irradiation tensile testing was conducted at RT, 300°C, and 500°C for selected irradiation conditions. Neutron irradiation induced significant changes in the mechanical behavior of the AM stainless steels, including both hardening and softening. Although the as-built 316L steel after 300°C irradiation showed necking just after yielding, the overall property changes of the as-printed alloy became less significant after 600°C irradiation. Irradiation-induced ductilization was also observed after the higher temperature irradiation. In general, the strength change was smaller in the relatively stronger as-built and stress-relieved AM SSs than in the solution-annealed AM and WT SSs. These relatively lower strength 316L SSs overall retained higher ductility in the irradiation conditions tested, but the stronger 316L SSs demonstrated a similar level of ductility after the higher temperature (600°C) irradiation. It is a positive assessment for the AM 316L materials that no embrittlement was observed within the test and irradiation conditions of the experiment.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Irradiation-induced amorphization of Fe-Y-based second phase particles in accident-tolerant FeCrAl alloys

Here, second phase intermetallic particles in an advanced accident-tolerant FeCrAl (Fe-13Cr-5Al-2Mo) alloy are formed in the α-Fe matrix during processing. These particles are prominently related to the added Y. Neutron irradiation to ~7 displacements per atom (dpa) with a dose rate of ~8.16 × 10 -7 dpa/s at 282 °C resulted in the amorphization of these precipitates which could degrade the mechanical properties of the FeCrAl alloys. Analytical electron microscopy and diffraction analysis combined with structural freedom analysis have been used to investigate the radiation resistance of the second phase particles. Radiation tolerance is closely linked to the particle Fe-Y content and can be tailored using the structure freedom value.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Characterization of mechanical properties and deformation behavior of highly irradiated 316L stainless steel from target modules at the Spallation Neutron Source using digital image correlation analysis

Tensile specimens with irradiation doses ranging from approximately 0.7 to 7.6 displacements per atom (dpa) were fabricated from 316L samples removed from Spallation Neutron Source (SNS) target vessels after service. Tensile tests were performed in conjunction with digital image correlation (DIC) characterization, which permitted detailed analysis of the strain distribution and deformation behavior of the specimen gauge sections during testing. The results show, while considerable hardening (500–700 MPa increase in yield strength) occurred in the 316L SNS target material with increasing dose, appreciable ductility (19–23% total elongation) remained in the material after irradiation to 7.6 dpa in the mixed neutron/proton environment of the SNS target region. The DIC data showed the strain distribution and deformation behavior of all irradiated specimens were nonuniform during testing. Strain localization in the form of deformation bands or “hotspots” was observed during testing, and in some specimens, the bands propagated as deformation waves along the gauge section. The “traveling wave” deformation behavior affected the amount of elongation each specimen achieved, and the scatter in the elongation data was attributed to the large variation in deformation behavior observed among different specimens. Bands of localized strain, or deformation bands, oriented at specific angles were observed in two-dimensional plots of the strain distribution in the gauge sections of some specimens. Overall, this paper describes the experimental procedures, results, and deformation behavior of tensile specimens from SNS targets after service and illustrates the value of DIC characterization during tensile testing of irradiated materials.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Plane-Strain Tension Testing of ATF FeCrAl Cladding (Progress Report)

Performance of advanced accident-tolerant fuel (ATF) claddings must be assessed under all relevant scenarios, including normal and off-normal operating conditions, as well as during a reactivity-initiated accident (RIA). During RIA, the fuel pellet may expand and cause significant loads on cladding, a condition known as pellet cladding mechanical interaction (PCMI), which may even lead to cladding failure. This report describes the fiscal year (FY) 2020 research activities and results from the development of separate-effects testing for ATF cladding candidate iron-chromium-aluminum (FeCrAl) tube specimens under loading conditions expected during RIA/PCMI events. To assess material performance under PCMI, a high strain rate burst test, often termed the modified burst test (MBT), is commonly employed. However, execution of such a test in a hot cell requires significant research. To overcome the complexities of performing MBTs in a hot-cell environment on irradiated samples, a plane-strain tension (PST) test was proposed by several authors as an alternative approach for PCMI testing. A new fixture was developed so that PST specimens could be tested using a more accommodating design for hot-cell manipulators. This report presents the PST fixture design, evaluates the possibility of modifying the specimen geometry, and discusses experimental results with unirradiated Al 6061 and FeCrAl tubes at various lengths. Furthermore, this report discusses the relevance and compatibility of the PST results to RIA scenarios and compares the PST and MBT results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mechanical Properties and Deformation Behavior of Additively Manufactured 316L Stainless Steel (FY2020)

The Transformational Challenge Reactor (TCR) program plans to build most of the TCR core components through additive manufacturing (AM) processes. These processes include laser powder bed fusion (LPBF) for the metallic (316L) components and the newly developed combined process of binderjet printing and chemical vapor infiltration (CVI) for the SiC fuel matrix. Mechanical testing and characterization tasks have been carried out since the beginning of the TCR program to (1) build a property database for the AM materials that will be used in TCR core and (2) to assess the materials’ performance in TCR-relevant conditions. This document reports the outcome of the testing and characterization efforts for the fiscal year with a focus on the mechanical performance data of AM 316L stainless steel (SS). Baseline tensile testing over a wide temperature range of room temperature–600 °C was completed for the AM 316L alloy in as-built, stress-relieved, and solution-annealed conditions. The as-built 316L showed the highest strength, and the alloy after the post-build treatments showed reduced strengths in the low-strain range. However, the strength differences among the AM materials became insignificant in the later part of deformation. Furthermore, regardless of post-build processing, the AM 316L SS showed higher strength and comparable ductility when compared with wrought 316L SS. Thermal creep testing and microstructural evolution during creep deformation were also performed under selected conditions. It was found that the AM 316L steel showed the best creep resistance in the stress-relieved condition. In-situ tensile tests were performed using scanning electron microscopy and 1-ID beamline at the Advanced Photon Source to elucidate the deformation and fracture behavior of AM 316L and the evolution of crystalline stress, dislocations, and pore distribution. Using these in-situ testing data, an in-depth analysis of the roles of microstructural features in deformation and fracture processes is presented herein. The final section of the document introduces ongoing and future activities for materials testing and characterization, including irradiation effect studies and ball punch testing on AM 316L and AM IN718 alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

In situ measurement and modelling of the growth and length scale of twins in α -uranium

Twinning is an important deformation mechanism in crystalline materials. Twins nucleate, then grow, forming a discrete pattern of deformation bands with a certain thickness. The mechanism for twin growth is not completely understood. We present a simple phase field model which captures the physics of twin growth and reproduces the details observed during in situ electron backscatter diffraction experiments on α-uranium. The interaction between residual dislocations at twin interfaces and mobile dislocations in untwinned regions increases the stress needed for twin growth. This is described by a nonlocal term in the proposed constitutive equations: the nucleation stress of a twin increases proportionally to the twin phase field in a neighborhood. Competition between slip and twinning favors the nucleation of a new twin rather than the growth of a preexisting twin. The phase field model is coupled with a crystal plasticity finite element solver that includes the plastic deformation due to twinning. Simulations are able to reproduce the number of twins and their thicknesses as a function of the strain observed during in situ electron backscatter diffraction experiments. The stress-strain curve is also reproduced. This comparison allows the values of the nucleation stress and the interaction strength between residual and mobile dislocations to be found. It also gives an estimation of the density of residual dislocations that is consistent with the observed twin thickness. Finally, this model can be applied to understand microstructural effects in materials using twinning as a strengthening mechanism.

36 MATERIALS SCIENCE↗