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Gussev, Maxim N.

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

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Complexity of segregation behavior at localized deformation sites formed while in service in a 316 stainless steel baffle-former bolt

Here, post-irradiation evaluation was performed on a 316 stainless steel baffle former bolt harvested after 40 years of service in a pressurized water reactor. Microstructure analysis revealed the presence of defect-free dislocation channels and strain-induced twins, indicative of loading at a stress level close to yield stress at least once while in service. Primary radiation-induced Ni/Si precipitates were destroyed during channel and twin formation, and secondary, significantly coarser Ni/Si precipitates formed along newly formed Σ3 boundaries during the continued irradiation. Additionally, an elevated phosphorus level was observed inside the strain-induced twin. Complex chemistry inside the strain-induced feature may overlap with dislocation pileups and impact localized corrosion, material long-term performance, and safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Analysis of Deformation and Fracture Mechanisms in the Harvested High-Dose Baffle-Former Bolt with Stress-Corrosion Cracks Formed While in Service

This report presents the results of advanced mechanical testing conducted on miniature tensile specimens excised from an irradiated baffle-former bolt, a commercial pressurized water reactor component. The specimens were extracted from the midsection of the bolt shank, where the estimated damage dose reached 23 displacements per atom. The mechanical testing included the following tests: (1) conventional tensile testing at room temperature, augmented by digital image correlation (DIC) to enable noncontact strain measurements, and (2) in situ tensile testing within a scanning electron microscopy (SEM) instrument equipped with energy-dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD) detectors to evaluate active deformation and fracture mechanisms. Additionally, SEM fractographic analysis revealed alterations in the fracture mechanism from a predominantly ductile fracture in nonirradiated steel to a mixed fracture mode (still dominating ductile and minor cleavage spots) in irradiated specimens derived from the baffle bolts. The findings indicate a complex strain localization behavior for in-service irradiated steel specimens. Beyond conventional necking at the macro scale and defect-free channel formation at the micro scale, DIC results identified the presence of deformation bands approximately 100 μm in width. These bands consist of chains of grains exhibiting elevated local strain levels and may be considered an intermediate or mesoscale level of strain localization. These bands become discernible near the yield stress as localized hot spots, areas of elevated strain, persisting throughout most of the experiment. The formation of mesoscale deformation bands as a strain localization mechanism may exacerbate the defect-free channel formation in irradiated materials, further promoting irradiation-assisted stress corrosion crack initiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

In-service Oxidation and IASCC in High Fluence Baffle-Former Bolts Retrieved from a Westinghouse PWR

Baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts, and life extension of existing pressurized water reactors (PWRs) would only cause more damage to the bolt material. To this end, the Light Water Reactor Sustainability (LWRS) Program Materials Research Pathway (MRP) successfully harvested two high fluence BFBs from a Westinghouse two-loop downflow type PWR in 2016. In the same year, the two BFBs were received at the Westinghouse Churchill facility for inspection and specimen fabrication. The fabrication was completed in 2017 with specimens shipped to Oak Ridge National Laboratory (ORNL) for further testing. The objective of this project is to provide information that is integral to evaluating end of life microstructure and properties as a benchmark of international models developed for predicting radiation-induced swelling, segregation, precipitation, mechanical property degradation, and with the latest findings, susceptibility to irradiation-assisted stress corrosion cracking (IASCC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

How irradiation promotes intergranular stress corrosion crack initiation

Irradiation assisted stress corrosion cracking (IASCC) is a form of intergranular stress corrosion cracking that occurs in irradiated austenitic alloys. It requires an irradiated microstructure along with high temperature water and stress. The process is ubiquitous in that it occurs in a wide range of austenitic alloys and water chemistries, but only when the alloy is irradiated. Despite evidence of this degradation mode that dates back to the 1960s, the mechanism by which it occurs has remained elusive. Here, using high resolution electron backscattering detection to analyze local stress-strain states, high resolution transmission electron microscopy to identify grain boundary phases at crack tips, and decoupling the roles of stress and grain boundary oxidation, we are able to unfold the complexities of the phenomenon to reveal the mechanism by which IASCC occurs. The significance of the findings impacts the mechanical integrity of core components of both current and advanced nuclear reactor designs worldwide.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Deformation and Fracture Behavior of Additively Manufactured 316L Stainless Steel

Tensile deformation and fracture behavior of an additively manufactured (AM) 316L stainless steel (SS) in the as-built, stress-relieved, and solution-annealed conditions was investigated using in situ tensile testing in a scanning electron microscope with an electron back scattering diffraction (SEM–EBSD) detector. Analyses were performed to discuss the characteristic deformation and fracture process of the fine-grained AM 316L SS with and without relaxation heat treatment. The as-built 316L showed the highest strength, and both post-build heat treatments lowered the strength of the alloy. Regardless of the post-build processing, the AM 316L SS showed overall higher strength but slightly lower ductility when compared to the wrought (WT) 316L SS. Analysis of EBSD data indicated that the characteristic microstructural features from AM, such as the complex and fine grain morphology, dislocation network, pores, and silicon-rich oxides, evolved and exerted various roles during the tensile deformation and fracture processes. Further, it was obvious that the interaction of dislocation slips with oxide particles and cavities (or pores) resulted in an accelerated cracking in AM 316L SS. Overall, however, their influence on mechanical behavior was limited, as the genuinely high ductility of 316L SS could help avoid any premature or brittle fracture.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Comprehensive Characterization of Helium-Induced Degradation of the Friction Stir Weld on Neutron-Irradiated 304L Stainless Steel

The report describes new experimental results on the mechanical performance of the friction stir welds made on neutron-irradiated 304L stainless steel with helium. The report focuses on helium-related issues (e.g., the helium-induced degradation in the welded joint), aiming at the need to repair irradiated components of nuclear power plants. The friction stir welds analyzed here were previously produced at the US Department of Energy’s Oak Ridge National Laboratory, and initial characterization work was performed, mostly addressing the microstructure and if macroscopic cracks and helium bubbles present. The present work attempts to perform a more comprehensive study to assess mechanical performance (i.e., microhardness distribution, tensile properties, tensile deformation behaviors, and fractography analysis).

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructure and in-service degradation of baffle former bolts – in-core components of light water reactors

This report presents a microstructure evaluation of the baffle former bolt, a commercial pressurized water reactor component. Using SEM, EDS, and EBSD analysis the material was identified as AISI 316 steel with an annealed austenite microstructure. Findings include abnormal grain growth in a fraction of the grains and the absence of retained ferrite. Notable observations include an overlapping of pre- and post-irradiation deformation, including defect-free channel formation as an active deformation mechanism in the near-surface layer of BFB. Specimen surfaces, exposed to high-temperature, high-pressure water, exhibited signs of in-service corrosion degradation. EBSD and EDS analyses highlighted intergranular corrosion, possible grain boundary oxidation with depths less than 3 µm, and, unexpectedly, short cracks filled with Cr-rich oxides measuring around 5-6 µm. The presence of defect-free channels from EBSD data suggests episodes of high mechanical stresses during service. The specimens provide a potentially unique insight into the in-service degradation of NPP components, strain localization, and crack initiation processes. An intact corrosion layer reveals a complex component's in-service history. Further research is underway for a detailed analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Microstructural Characterizations of Two High Fluence Baffle-Former Bolts Retrieved from a Westinghouse Two-loop Downflow Type PWR

As one of the pressurized water reactor (PWR) internal components, baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts. In support of evaluating long-term operational performance of materials used in core internal components, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Office of Nuclear Energy, Light Water Reactor Sustainability (LWRS) Program, Materials Research Pathway (MRP), has harvested two high fluence BFBs from a commercial Westinghouse two-loop downflow type PWR. The two bolts of interest, i.e. bolts # 4412 and 4416, were withdrawn from service in 2011 as part of a preventative replacement plan. No identification of cracking or potential damage was found for these bolts during their removal in 2011. However, the bolts required a lower torque for removal from the baffle structure than the original torque specified during installation. Irradiation damage levels in the bolts range from 15 to 41 displacements per atom. The goal of this project is to perform detailed microstructural and mechanical property characterization of BFBs following in-service exposures. The information from these bolts will be integral to the LWRS program initiatives in evaluating end of life microstructure and properties. Furthermore, valuable data will be obtained that can be incorporated into model predictions of long term irradiation behavior and compared to results obtained from experimental reactor conditions

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Mechanical properties of Zircaloy cladding tubes and contributions to M.E.T.A. mechanical property database

To support a multi-laboratory Methodology, Evaluation, Testing, and Analysis (M.E.T.A.) cladding properties database, Oak Ridge National Laboratory’s (ORNL’s) cladding mechanical test geometries were manufactured from several nuclear-relevant cladding alloys and subsequently tested. These geometries were developed as mechanical test specimens to evaluate the properties of tube materials that may be used for irradiation testing at ORNL’s High Flux Isotope Reactor. They may also be used as test articles to be harvested—via in-cell machining—from commercially irradiated fuel rods and later tested. This report explores the differences among axial, hoop, and SSJ tensile geometries with partially recrystallized Zircaloy-2 to test ORNL correlation-based methods on a plate material that approximates, to the greatest extent possible, the characteristics of nuclear industry tubing. Furthermore, several tests were conducted with ORNL’s Zircaloy-4 tube inventory to (1) develop material properties as a standard for future tests, (2) determine the effect of the US Department of Energy’s Advanced Fuels Campaign coating processes on tube mechanical properties, and (3) evaluate the effect of specimen machining methods on the mechanical properties of tube geometries.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Mechanical responses of 316L stainless steel printed by wire arc additive manufacturing with different thermal histories

In this paper, 316L stainless steel walls were printed using single-bead pass and multiple-bead pass wire arc additive manufacturing (WAAM), which represent the internal basket wall and external wall of a spent nuclear fuel (SNF) canister; respectively. Scanning electron microscopy (SEM) microstructure characterization and tensile tests at room temperature (21 °C) and elevated temperature (250 °C) were performed on specimens extracted along the length and height directions of the printed walls. Results showed that grain morphology, grain angles, and retained ferrite are different between the two printed walls, which received different heat input in WAAM. The single-bead pass printed wall exhibited higher yield strength than the multiple-bead pass printed wall because of more retained ferrite and larger portion of random-high angle boundaries (RHAB) grains.

Tang, Wei↗
DOE OSTI2023

Analysis of Deformation and Fracture Mechanisms in the Harvested Low-Dose Baffle Former Bolt via Advanced Mechanical Tests

This report details the production of tensile specimens from in-service irradiated baffle former bolts (BFBs) and presents the results of a pilot in-situ SEM/EBSD test conducted at Oak Ridge National Laboratory’s Low Activation Materials Development and Analysis (LAMDA) facility. The SEM/EDS results confirmed that the material is AISI 316L steel, not AISI 347, and that no Nb addition was detected in several analyzed locations. EBSD analysis of the microstructure revealed in-grain misorientation gradients indicative of cold work, likely caused by material processing, and no retained ferrite was identified. During straining, dislocation channel formation was the primary deformation mechanism, with twinning also observed in favorably oriented grains at local strains as small as ~1.5%. No strain-induced phase instability was observed, likely due to the fact that the Ni and Cr content reached the upper limits for the 316L steel specification. Multiple microfracture events were observed during the tensile test. Fine strain-induced pores formed at the channel-grain boundary intersection points, while coarser microcracks were associated with non-metallic inclusions. Data analysis is ongoing, and additional tests are planned in the near future to provide more statistics and details.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Two-step deformation-induced martensitic transformation in additively manufactured High-Si stainless steel

Deformation-induced phase transformation in the additively manufactured (AM) Fe–18Cr–8Ni-5.5Si (in wt.%) alloy was analyzed using the in-situ electron backscattered diffraction (EBSD) record during uniaxial tensile testing in a scanning electron microscope (SEM). Analysis of EBSD data was focused to investigate phase transformation behavior during deformation with consideration of mechanical interactions among phases and grains in the ultrafine AM microstructure. The initial EBSD phase map illustrated that the as-built microstructure was a duplex structure consisting of major austenite (γ) and minor ferrite phases, some of which showed non-traditional orientation relationship since the rapid solidification in the AM process has produced non-equilibrium state in phase formation. In the deformed as-built AM alloy, the two-step martensitic transformation (γ→ε→α´) was identified as the main deformation-induced transformation mechanism. Martensitic transformation steps were found to be associated with glides of Shockley partial dislocations on closed packed (111) planes under the influence of atomic disturbance in the γ and ε phases, respectively. Using the closed-packed plane pole figures and unit cell orientations, their evolutions and mechanisms accompanying with the transformation were analyzed. Finally, the results indicate that one variant of ε-martensite forms from their parent γ phase, while two variants of α´ form from their parent ε phase.

36 MATERIALS SCIENCE↗
DOE OSTI2022

316L Stainless Steel Wire Arc Additive Manufacturing and Characterization for Potential SNF Canister Production

In this paper, 316L stainless steel (316L SS) straight walls, curved walls, and vessels were printed by wire arc additive manufacturing (WAAM) using the 316L SS welding wire, to demonstrate the feasibility of spent nuclear fuel (SNF) canisters using this advanced manufacturing technique. Helium leak test with leak tight criteria of 1 × 10-7 ref-cc/s were performed on printed vessels with various wall thickness, and testing pressures were from 0.345 MPa (50 psi) to 1.724 MPa (250 psi). Printed wall microstructures were characterized using optical microscopy (OM) and scanning electron microscopy (SEM). Tensile specimens were machined out from printed walls along the length, height, and thickness directions, respectively, and tested at different temperature conditions, room temperature, 80 °C, 150 °C and 250 °C, respectively. Results showed t hat, all five WAAM-printed canisters with different thickness passed the helium leaktight criteria of radioactive materials containers by the ANSI N14.5; Dispersed micrometer-level size inclusions/voids were observed at printed wall cross sections, dendritic microstructures were found with grain size variation through the welding torch shifting direction, and very small amount of δ-ferrite and σ-phase were also observed in those cross sections; The printed walls were ductile, all tensile specimens broke with large plastic deformation/elongation and without brittle failure, and printed specimens yield strengths and tensile strengths under room temperature and 80 °C conditions are higher than minimum values of 316L SS for pressure vessels and for general applications at room temperature required by the ASTM standard, respectively.

Tang, Wei↗
DOE OSTI2022

Origin, parameters, and underlying deformation mechanisms of propagating deformation bands in irradiated 316L stainless steel

Lüders-type propagating deformation bands were observed in specimens of irradiated 316L stainless steel samples removed from Spallation Neutron Source target vessels after service. Mechanical testing with digital image correlation (DIC) and in-situ tensile testing with scanning electron microscopy electron backscatter diffraction showed that the observed Lüders-type behavior was not related to the known transformation-induced plasticity or twinning-induced plasticity behavior. Instead, the phenomenon occurs at small local strain values before a significant amount of martensite or deformation twins appear in the microstructure. Microstructural analysis and in-situ mechanical test results suggest Lüders-type band formation and propagation were related to the appearance and evolution of defect-free channels—analogous to slip bands. A modified Swift equation with a Ludwigson-like component was offered to rationalize the phenomenon and model the strain-softening processes at small strain values. Finally, the results indicate complex microstructural processes and deformation mechanisms were active at small strain values and underline the benefits of advanced mechanical test approaches such as DIC.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Quantifying deformation during Zry-4 burst testing: a comparison of BISON and a combined in-situ digital image correlation and infrared thermography method

The development and verification of nuclear fuel-cladding performance codes require extensive testing to establish empirical correlations, necessitating more accelerated testing methods that can replace large validation studies. Here, in the present work, a technique was developed and implemented to experimentally quantify the relationship between temperature, cladding internal pressure, and strain in-situ during a burst test, with the specific aim of generating data for code validation and model refinement. Digital image correlation was used to measure hoop, axial, and radial strain, and infrared thermography to quantify axial temperature gradients. Several key experimental modifications to traditional LOCA burst testing were necessary, but are shown to effectively have little impact on burst temperatures. The time/temperature dependent pressure and time/spatially dependent thermal gradient data were used with BISON to simulate cladding burst and strain rates. Although the measured DIC strains reach appreciable amounts at slightly lower temperatures than BISON simulated strains, there is good agreement between the measured and simulated strains in terms of magnitude and rates leading up to burst. However, in the few seconds prior to and during burst, the BISON simulated strains are significantly lower than the measured strains, indicating the potential for model improvement. The combined experiment and simulation technique may be applied to any new cladding concept to accelerate fuel qualification.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Micromechanical aspects of deformation and failure of advanced Iron-Chromium-Aluminum alloys

Here this work investigates the micromechanical deformation and failure mechanisms of advanced FeCrAl alloys developed for use in nuclear reactors as cladding material. Three different FeCrAl alloys (82-X)Fe-13Cr-5Al-X (X=Nb, TiC, 0 for base alloy) were investigated both in their as-received and welded states. In-situ neutron diffraction with simultaneous digital image correlation was used to determine micromechanical deformation mechanisms not only as a function of elemental composition but also as a condition of state (as-received vs. welded). Ex-situ X-ray computed tomography was used on as-deformed samples to help determine the failure mechanisms and void initiation strains.

36 MATERIALS SCIENCE↗