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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

In situ synchrotron X-ray diffraction analysis of ultrasonically assisted microstructure refinement during laser melting process

This work investigates the effect of ultrasonic treatment (UST) on the microstructure of AA4043 alloy during laser melting process. Here, the study utilizes in situ synchrotron X-ray diffraction (SXRD) and postmortem electron backscatter diffraction (EBSD) techniques to examine the ultrasonic grain refinement mechanism. The SXRD diffraction patterns exhibit higher continuity of major diffraction rings in the UST-test pattern, indicating a more refined grain structure. The presence of new diffraction spots in the post-test diffraction pattern with UST suggests the breakdown of epitaxial growth of columnar dendrites during solidification. Thermal profiles determined by analyzing the lattice parameter evolution show a faster cooling rate with UST. The inverse pole figures and aspect ratio distribution from EBSD reveal that the UST samples exhibit a refined, more equiaxed grain structure along the weld centerline. UST sample also develops more fine-sized low angle grain boundaries (LAGBs). These features offer better hot-cracking resistance in aluminum alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of Thermal Aging on Microstructure and Stress Corrosion Cracking Behavior of an Alloy 152 1st Layer Butter Weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40- 80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is critical for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for different durations up to 75,000h (equivalent to 60 years of reactor service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged at 370°C and 450°C to an equivalent of 60 years of service. However, the first weld butter layer has high levels of deformation and is highly susceptible to SCC even in its non-aged condition. Nanohardness testing confirmed the extreme hardening with aging (ΔHV ≅ 100) at two locations within this weldment. In absence of LRO, hardening is suspected to be due to thermally-induced Cr carbide precipitation and coarsening. Testing in a primary water environment of the 1st layer of Alloy 152 weld butter aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. Similarly, the SCC CGR response of the aged weld butter does not appear to show a deterioration in performance, however, the difficult-to-test weldment geometry may affect the test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Effect of thermal aging on microstructure and stress corrosion cracking behavior of Alloy 152 weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged to an equivalent of 60 years of service. Testing in a primary water environment of two heats of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. However, the SCC CGR response of the aged samples appears to show a deterioration in performance, confirming our previous observation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray did not find evidence of LRO in any of the three heats aged to an equivalent of 60 years of service. Testing in a primary water environment of a heat of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. However, the SCC CGR response of the aged sample appears to show a deterioration in performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Pitting behavior of friction stir repair-welded 304L stainless steel in 3.5% NaCl solution at room temperature: role of grain and defect structures

Stainless steel (SS) canisters used for storing spent nuclear waste are prone to chloride-induced stress corrosion cracking. Friction stir welding (FSW), being a low heat input process, can be used for repair welding of these canisters. In this study, an artificial crack was introduced on the 304L SS coupons by an electric discharge machining process. The artificial crack was repair-welded by FSW by maintaining the tool temperature a constant at two levels (725 and 825 °C). Friction stirring significantly reduced the grain size of the stirred zone (SZ) from about 47 to 2–4 μm. The fraction of low-angle grain boundaries increased in the SZ from 2 to 37–43%. Here, on the other hand, the fraction of special grain boundaries (Σ3 and Σ9) that was ~ 50% in the base metal reduced to < 10% in the SZ. During friction stirring, the oxide layer of artificial crack was broken and aligned to form a spiral defect called lazy-S structure. All these microstructural changes affected the corrosion behavior of the FSW specimens when tested in 3.5% NaCl at room temperature using cyclic polarization, chronoamperometry, and electrochemical impedance spectroscopy. The FSW specimens showed lower polarization resistance and lower transpassive potential than those of the base metal specimens. However, the pitting protection potentials of the FSW specimens were higher than that of the base metal. The pitting behavior of the FSW specimens was influenced more by the preferential attack on the lazy-S region than by passive film breakdown. The flat band potentials of the passive film formed on the FSW specimens were more positive than that of base metal. The charge carrier density of passive film formed on the FSW specimens was higher than that of the base metal. The higher fraction of low-angle grain boundaries present in the FSW specimens could supply more number of misorientation dislocations at the metal/film interface which could form anion vacancies by a climb process leading to formation of oxide at these locations without stressing the substrate. Therefore, low-angle boundaries are considered helpful for formation of stress-free passive films that will be highly stable and enhance both pitting and stress corrosion cracking resistance.

Dry storage canister↗
DOE OSTI2024

Mechanical Failure Risk Management for In-Service CSP Nitrate Hot Tanks

The hot thermal energy storage tanks in many operating (Gen2) concentrating solar power plants are designed to contain molten nitrate solar salt at 565 Degrees Celsius and are constructed from 347H stainless steel (SS). From previous work, it is known that 347H SS weldments are potentially susceptible to stress-relaxation cracking (SRC), particularly in the heat-affected zone above 540 Degrees Celsius when no post-weld heat-treatment (PWHT) is performed. However, residual stresses can be reduced when weldments receive PWHT, while at the same time a crack-resistant microstructure is being generated. The overarching goal of this project was to optimize cost-effective, thickness-dependent, localized PWHT procedures and to investigate for feasible non-destructive evaluation (NDE) at commercial scale for hot tanks in Gen2 CSP using 347H SS.

14 SOLAR ENERGY↗
DOE OSTI2022

Microstructure and creep strength of simulated intercritical heat-affected zone of grade 91 steel

The intercritical heat affected zone (ICHAZ) has been reported as one of the most Type IV cracking susceptible regions in 9Cr creep-resistant steel weldments. However, creep degradation mechanisms within the ICHAZ itself need further clarifications. In this work, two ICHAZ specimens of Grade 91 steel, low-temperature ICHAZ (LT-ICHAZ) and high-temperature ICHAZ (HT-ICHAZ), were simulated using the Gleeble thermomechanical system by exposing to two peak temperatures (860 °C and 900 °C) between $A_{C1}$ and $A_{C3}$. Dramatically different creep strengths of two simulated ICHAZs were observed and studied. We report dilation curve analysis indicates a high fraction of newly transformed martensite formed in the HT-ICHAZ, which results in a much higher hardness (360 HV0.5) of the HT-ICHAZ than 266 HV0.5 of LT-ICHAZ. Precipitates, especially $M_{23}C_6$ carbides, were not fully dissolved in both ICHAZ specimens. After a typical postweld heat treatment (760 °C-2 hours), the faster recovery of low-carbon martensite and reduced precipitation strengthening due to $M_{23}C_6$ carbides coarsening in the HT-ICHAZ led to a significant hardness reduction. These microstructural degradations in the HT-ICHAZ made its creep lifetime about 36 times shorter than that of the LT-ICHAZ tested at 650 °C. The remaining tempered martensite from base metal in the LT-ICHAZ was the primary contributor to maintain its high creep resistance.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Understanding failure in austenitic steels: Key considerations for molten salt storage in CSP applications

Owing to their excellent properties, austenitic stainless steels are extensively used in boilers, furnaces, molten salt tanks, and other applications that are subjected to extreme mechanical loads and high-temperature conditions. Their high corrosion and creep resistance make them suitable for high-temperature operating environments. Additionally, good fatigue resistance and favorable mechanical and visual properties are essential. However, the premature failure of several components at elevated temperatures has been previously reported. Although the failure analysis of components in service is complex, processes such as cold work and welding have been identified as contributing factors to the performance degradation of these steels. This study aims to analyze the various documented failure modes and mechanisms in austenitic steels, including creep, cracking, stress relaxation cracking, and fatigue, to better understand the multi-objective design requirements for these alloys as structural materials for high-temperature molten salt tanks in Concentrating Solar Power (CSP) plants. Stabilized austenitic grades, such as AISI 347H, demonstrate superior resistance to creep and corrosion-related degradation when compared to non-stabilized grades like AISI 316L at temperatures relevant to concentrated solar power (CSP) applications. In contrast, nickel-based alloys offer enhanced corrosion resistance, albeit at a higher cost. This review underscores that creep, stress relaxation cracking, and thermo-mechanical fatigue are the predominant long-term failure risks in CSP hot tanks.

14 SOLAR ENERGY↗
DOE OSTI2025

The Effect of Thermal Aging on the Performance of High-Cr Ni-based Alloys in Primary Water Environment

Nickel-based Alloy 690, and the associated weld Alloys 52 and 152, are used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels. They are chosen for their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate beyond their first or second license renewal (e.g., 40 to 60 years or beyond 60 years). Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants. Moreover, Ni-Cr alloys are being considered for high-level radioactive waste disposal containers that will be required to contain the waste for at least 10,000 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Effect of post-weld heat treatment on microstructure and mechanical properties of rotary inertia friction welded dissimilar 422-4140 martensitic steel piston joints

The study investigated how post‐weld heat treatment (PWHT) temperature affects the microstructure and localized deformation/fracture during bend testing of rotary inertia friction welds (RIFW) between AISI 422 stainless steel and AISI 4140 steel. RIFW produced a fully martensitic interface with approximately 550 HV hardness in both the thermo-mechanically affected (TMAZ) and heat‐affected zones (HAZ). Due to differences in temper resistance, the 4140 TMAZ/HAZ softened progressively under PWHT temperatures from 525 °C to 700 °C, while the 422 TMAZ unexpectedly maintained about 550 HV up to 600 °C before significantly softening at temperatures ≥625 °C. This asymmetric softening generated steep hardness gradients across the interface at temperatures ≤600 °C. Furthermore, carbon migration across the interface was minimal up to 600 °C, moderate at 625 °C, and by 700 °C produced a carbide‐rich eutectoid layer in the 422 TMAZ alongside a carbon‐depleted soft ferrite layer in the 4140 TMAZ. Strain during bending was PWHT‐dependent, concentrating on the 4140 side; in as-welded joints, the high hardness led to deformation and crack initiation in the base metals, whereas in PWHT samples, cracking initiated in the softened 4140 TMAZ near the interface. The intermediate PWHT temperature of 625 °C offered the best balance of limited carbon diffusion across the interface, relatively low peak weld hardness and minimized hardness gradients across the interface, more homogenous deformation, and good bend test performance.

4140 low-alloy steel↗
DOE OSTI2025

Slow Strain Rate Testing of A537 Tank Wall Material

At Savannah River Site (SRS), High-Level Waste is stored in below-grade carbon steel tanks. This waste in part consists of sludge, salt cake, and/or supernate. Preparation of this waste for future processing involves dissolution of the salt cake layer. The salt dissolution process can create conditions that leave the carbon steel tanks susceptible to localized corrosion. The salt to be dissolved contains high concentrations of nitrate, that once released, create an environment that may be conducive to pitting corrosion and/or stress corrosion cracking (SCC) of carbon steel. The salt dissolution process also liberates interstitial liquid trapped between the salt crystals. This liquid is initially high in nitrite and hydroxide concentration. High pH and greater ratios of nitrite to nitrate act as inhibitors to minimize corrosion of carbon steel in high nitrate environments. However, as dissolution proceeds, the concentration of nitrate will increase, while the hydroxide and nitrite concentration of the interstitial liquid will deplete and become insufficient to prevent the onset of corrosion attack. Tank blending and the addition of inhibitors are used to ensure adequate concentrations of hydroxide and nitrite. However, this is not desirable during salt dissolution as it can reduce process efficiency and increase the amount of waste that needs processing. This testing program was designed to examine the risk of SCC associated with utilizing the pitting factor (PF) and nitrite/nitrate (NO 2 - /NO 3 - ) ratio limits for handling dissolved salt solutions at an elevated temperature in the carbon steel waste tanks. The previously identified limits are a PF of 1.2 and an NO 2 - /NO 3 - ratio of 0.15. The results indicate that as long as the NO 2 - /NO 3 - ratio exceeds 0.1 and the PF is above approximately 0.8, there is a discernible safety margin between the open circuit potential (OCP) and the critical cracking potential (CCP) observed during applied potential testing. However, this margin, defined by the difference between the OCP and CCP, is relatively narrow, ranging from 0.1 to 0.25 volts. This small margin raises concerns about potential shifts in OCP during waste retrieval operations, which could inadvertently increase the risk of SCC if the OCP approaches or exceeds the CCP. These results confirm that dissolved salt solutions provide a potent chemistry that, under certain conditions, makes carbon steel susceptible to SCC. The next question to consider is the influence these results have on decisions for storage and retrieval of waste from the tanks. For Type III/IIIA waste tanks, the risk of SCC remains very low. First, and most importantly, the post-weld stress relief of the tanks has reduced the residual stress near the welds. Thus, without the stress component, SCC risk is minimized. The material of construction (A537 Carbon steel) for the Type III/IIIA tanks is superior to the steel in its resistance to SCC than the steel that was utilized for the Type I, II, and IV tanks (A285 carbon steel). From a chemistry control standpoint for a Type III/IIIA tank directly involved with handling dissolved salt solutions, the PF and NO 2 - /NO 3 - ratio limits may be utilized wherein chemistry control provides an extra layer of defense against SCC. Chemistry control for a Type III/IIIA tank minimizes the risk for a tank that may receive the dissolved salt solution, particularly if that tank is a Type I, II, or IV waste tank. On the other hand, if the dissolved salt solution is handled by a Type I, II, or IV waste tank the risk of SCC is real. The potent chemistry, absence of stress relief, and inferior material result in a condition that is conducive to cracking. Efforts should be made to either avoid transferring waste that may not meet the PF and NO 2 - /NO 3 - ratio criteria to one of these tanks or if it is unavoidable, take measures to minimize the consequences of a leak. As shown by these tests, even if the PF and NO 2 - /NO 3 - ratio criteria are met, there is a risk that the tank potential may be disturbed in the positive direction and the risk of SCC increase.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Estimation of pipe failure frequencies in the absence of operational experience data: A pilot study

Probabilistic failure metrics such as leak frequency and rupture frequency are commonly used to characterize piping reliability. The methodologies for calculating the failure metrics rely on a complex set of input parameters. Operating experience data and experimental data play an important role in informing the different input parameters. The paper describes results and conclusions of a coordinated research project to benchmark three different reliability models using a four-step procedure: reference case definition of relevance to advanced reactor designs, input parameter calibration, validation of results, and application of different methodologies upon completion of the calibration and validation steps. The reference case is a weld consisting of nickel-base alloy 152/52 and located within a primary pressure boundary of an advanced reactor. This alloy is a class of structural materials known to be highly resistant to stress corrosion cracking. Synergies between the different methods are noted and the importance of a multi-disciplinary approach to input parameter development is underscored. A key conclusion is that the three methods are equally suitable for estimating failure frequencies. In any specific application, a selection of the most practical or effective computational tool can be considered. The comparison of alternative models confirms and helps to gain confidence in the computed failure frequency estimates. The study was part of a coordinated research project organized by the International Atomic Energy Agency.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Experimental evaluation of localized creep deformation in grade 91 steel weldments

Spatially resolved measurement of localized creep deformation in heterogeneous creep resistant steel weldments is crucial but challenging for lifetime assessments of critical steam components in power plants. Here, experimental approaches were established to quantitatively evaluate commonly observed localized creep deformation in multi-pass Grade 91 steel weldments. An in-situ digital image correlation (DIC) system was utilized with a creep testing frame to monitor and measure both full-field strain and localized strain accumulation across the weldments during long-term creep testing at elevated temperatures. The in situ DIC method measured not only the creep deformation behavior of the weld metal, heat affected zone (HAZ), and base metal, but also creep strain evolution for each sub-region within the HAZ itself, including coarse-grained HAZ, fine-grained HAZ (FGHAZ), and intercritical HAZ (ICHAZ). The DIC results revealed that local creep strain in the ICHAZ reached up to 90% strain before final rupture, whereas nominal creep strain measured by the standard extensometer of the tested cross-weld specimen was below 10%, indicative of Type IV cracking of the Grade 91 weld. Microstructural analyses revealed that the faster creep degradation/deformation in the HAZ was caused mainly by accelerated matrix grain recrystallization/growth and a reduced pinning effect from the segregated and coarsened precipitates in the FGHAZ and ICHAZ. The ultimate creep rupture occurred in the ICHAZ owing to its lowest creep resistance induced by the largest recrystallized grain size, the lowest fraction of coincidence site lattice, and the lowest local strain energies/dislocation densities.

36 MATERIALS SCIENCE↗
DOE OSTI2023

The Effects of Silicon and Niobium Concentration on the Solidification Behavior and Microstructure of Cast Monel Alloys

Cast Monel alloys are used in applications requiring a combination of good mechanical properties and excellent resistance to corrosion. Despite prevalent industrial use, relatively few studies have been conducted to investigate the relationships between composition, solidification behavior, and microstructure. Given that these alloys are used in the cast and welded conditions, these factors have a significant influence over the material properties. Here, in this work, microstructural characterization, electron probe microanalysis, X-ray diffraction, and differential scanning calorimetry were used to study how changes in Si and Nb concentrations affected the solidification path and microstructure of Monel alloys. It was found that increasing Nb concentration stabilized higher amounts of MC carbides and suppressed graphite formation during solidification. It was also found that the high nominal concentration and segregation of Si to the liquid led to the formation of Ni 31 Si 12 and other silicides via terminal eutectic reactions at the end of solidification. A pseudo-binary solidification diagram was constructed using experimental data and was applied to predict the mass fraction of solidified eutectic as a function of composition. The modeled microstructures were found to be in good agreement with experimentally measured phase fractions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Determination of the Solid-State Resistance-Weldability of Additively Manufactured 304L Stainless Steel

• A scoping study was undertaken at Savannah River National Laboratory (SRNL) to determine the solid&#x2;state resistance-weldability of additively manufactured 304L stainless steel. • Additive manufacturing of pressure-containing boundaries is of interest to numerous industries. • The approach investigated in this study was a low energy, solid state spot weld (pinch weld). • This general approach has been routinely used for conventionally prepared tubing and is now being expanded to additively manufactured components. • This work aimed to understand the impact of AM on the bond quality and weld geometry of pinch welds.

Rogers, Jeremy K. [Savannah River National Laborat↗
DOE OSTI2025

Suppression of irradiation hardening in tungsten-coated ferritic steel for fusion reactor blanket applications

W-coated reduced activation ferritic steels have been developed for use as plasma facing components in fusion reactor blankets, offering excellent sputtering resistance and structural strength. Previous high-temperature coating methods, such as diffusion bonding and brazing, caused interfacial deterioration due to thermal stress from mismatched thermal expansion between W and reduced activation ferritic steel. To address this, underwater explosive welding was introduced as a high-velocity cold process that joins dissimilar materials while maintaining a strong, thin interface without the thermal issues associated with traditional methods. In this study, the effects of neutron irradiation on the hardness and microstructure in W-coated F82H reduced activation ferritic steel (W/F82H) joined by underwater explosive welding are investigated. Following neutron irradiation at 290 °C, irradiation hardening is suppressed in W, F82H, and their interface within the W/F82H material. Furthermore, microstructural observations indicate that the recovery of work hardening and relaxation of elastic strain introduced during coating significantly contribute to the suppression of irradiation hardening in W/F82H. In conclusion, W/F82H exhibits significantly suppressed irradiation hardening compared with those in stand-alone materials. This suppression is explained by residual stress from thermal expansion mismatch and the unique microstructure at the interface. These results provide valuable insights for the development of more durable materials in nuclear fusion applications.

36 MATERIALS SCIENCE↗