Engineering topics

Katoh, Yutai

Publications and source records attributed to Katoh, Yutai.

At least 37 records · Page 2

DOE OSTI2022

Stress rupture of SiC/SiC composite tubes under high-temperature steam

SiC-fiber–reinforced SiC matrix composite cladding for light water reactor fuel elements must withstand high-temperature steam oxidation in a loss-of-coolant accident scenario (LOCA). Current composite designs include an outer monolithic SiC layer, in part, to increase steam oxidation resistance. However, it is not clear how such a structure would behave under high-temperature steam in the case when the monolithic layer cracks and carbon interphases and SiC fibers are exposed to the environment. To fill this knowledge gap, stress-rupture tests of prototypic SiC composite cladding at 1000°C under steam and inert environments were conducted. The applied stress was ~120 MPa, which was beyond the initial cracking stress. The failure lifetime under steam was 400–1300 s, while 75% of the composite specimens did not fail after 3 h of total exposure under inert gases. Microstructural observations suggest that steam oxidation activated slow crack growth in the fibers, which led to failure of the composite. The results from this study suggest that stress rupture in steam environments could be a limiting factor of the cladding under reactor LOCA conditions.

36 MATERIALS SCIENCE↗

DOE OSTI2022

A review on additive manufacturing of refractory tungsten and tungsten alloys

Here we review the progress of additive manufacturing effort on refractory metal tungsten and tungsten alloys. These materials are excellent candidates for a variety of high temperature applications but extremely challenging to fabricate via additive manufacturing due to a series of existing issues during the manufacturing. We outline these issues and discuss the current understanding and progress to tackle them. Laser powder-bed-fusion, laser directed-energy-deposition, and electron beam powder-bed-fusion are three common techniques that have been applied to additively manufacture pure tungsten. This overview discusses current observations and understanding on the issues associated with each of these techniques. We identify future research opportunities in additive manufacturing of refractory metals.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Failure evaluation of neutron-irradiated SiC/SiC composites by underwater acoustic emission

Here, understanding the statistical properties of mechanical properties of non-irradiated and neutron-irradiated SiC/SiC composites is essential for component design. This study aims to evaluate the detailed damage accumulation behavior of composites focusing on two fracture parameters: proportional limit stress (PLS) and acoustic emission (AE) onset stress. The developmental underwater AE technique, which is benefit in non-contact in-situ failure monitoring method during mechanical testing and in handle of the irradiated material, was first applied to evaluate damage accumulation behavior. Two types of chemical vapor infiltration SiC/SiC composites were used: one reinforced with Hi-Nicalon Type-S SiC fiber and one reinforced with Tyranno-SA3 SiC fiber in the form of satin-woven 2D fabrics with pyrolytic carbon interface. Neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory reached a fluence of 30 dpa at a temperature of 620–670°C. Four-point flexural tests were conducted to evaluate post-irradiation strength. Weibull statistics did not suggest marked degradation of composite strength. Detailed failure behavior evaluated by AE demonstrated no irradiation-induced change of the AE onset stress (i.e., crack initiation equivalent stress). Failure probability analysis suggests that increasing the reliability of composites (i.e., the Weibull modulus rather than strength itself) is essential to expanding the design margin and benefiting from a probabilistic design approach.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Anisotropic thermal diffusivity and conductivity in SiC/SiC tubes studied by infrared imaging and X-ray computed tomography

Traditional flash diffusivity evaluation of thermal diffusivity/conductivity of composite tubes require machining of specimens. For a thin-wall tube, this method can only be used to obtain through-thickness transport property. A novel method to evaluate anisotropic thermal diffusivity in a composite tube has been developed. Braided SiC/SiC composite tubes were subjected to a xenon flash heating pulse. A high-speed, high-sensitivity infrared camera was used to measure surface temperature changes as a function of time and nondestructively detect subsurface defects/damages, such as macroscopic pores. Standard reference material (Pyroceram 9606) and curved SiC/SiC composite tube specimens were used to validate thermal diffusivity obtained from infrared imaging. Unlike the traditional method, there is no need prepare special specimens, and thermal diffusivity values in three orientations are obtained after a single flash. A finite element analysis model based on x-ray computed tomography scans was developed to simulate the heat transfer. In conclusion, this technique is significant in assessing thermal conductivity and inspecting the health of ceramic tubes during and after service.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Neutron irradiation of tungsten in hydrogen environment at HFIR

Neutron irradiation of tungsten with and without the presence of hydrogen is needed to understand the influence of hydrogen on microstructure development under fusion reactor conditions. However, there is a risk of ignition if air ingress occurs during seal welding of irradiation capsules in a pressurized hydrogen environment. Therefore, an irradiation capsule was designed that contains several disks of vanadium hydrides at a 30% hydrogen-to-metal atomic ratio. During irradiation, the hydrogen is released from the hydrides as the capsule temperature increases, so the irradiation capsule environment is mostly hydrogen when the capsule reaches its 400 °C design temperature. In this study, we describe the design and operating characteristics of this first-of-a-kind irradiation capsule.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Thermo-Mechanical Distortion of Tungsten-Coated Steel During High Heat Flux Testing Using Plasma Arc Lamps

An experimental setup and a test section were designed and fabricated for high heat flux testing (HHFT) of neutron-irradiated specimens using water-wall plasma arc lamps. Because of the radiological considerations and limitations of reactor irradiation, the size of the test articles was limited to disks less than 10 mm in diameter. The specimen was clamped onto an actively cooled block, and clamping allowed the insertion of several thermocouples on the back surface of the specimen through a copper (Cu) block. Five vacuum plasma sprayed tungsten (W)–coated F82H steel specimens were subjected to HHFT. Surface profilometry measurements, which were conducted after HHFT, revealed central bowing of the top W surface. This type of residual distortion occurred for all of the specimens, and the larger the specimens were, the larger was the distortion.In an attempt to understand specimen distortion and address the science questions related to the testing of subsize specimens during HHFT, a simplified thermo-mechanical model was developed. By using a measured temperature in the Cu as an isothermal boundary condition, the model eliminated the need for coupling cooling fluid flow models with stress models, greatly simplifying the analysis. The main variable in the proposed model is hC, i.e., the thermal contact conductance between the F82H and the Cu washer. Inelastic properties, including hardening properties, were considered for F82H steel and Cu. Numerical simulation results demonstrated a buildup of residual deformation during HHFT and a very complex state of stress and deformation during typical heat flux (HF) cycling. Additionally, hoop stress evolution during a high heat flux cycle reveals that F82H at an interface with W would be mainly in compression during HF application and experienced a transition to a tension state during cooldown. Also, specimen distortion evolves during each HF cycle, as the specimen bows downward during HF application and upward during the cooldown period between HF cycles. The final specimen distortion, i.e., upward bowing of the specimen center, was qualitatively predicted for hC values of 4000 to 5000 W/(m 2 ·K). This hC range of values, for which bulging is obtained, is at the lower spectrum of the range of values for hC, consistent with the low thermal contact conductance expected from the unpolished F82H surface.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Additive manufacturing of ceramic materials for energy applications: Road map and opportunities

Among engineering materials, ceramics are indispensable in energy applications such as batteries, capacitors, solar cells, smart glass, fuel cells and electrolyzers, nuclear power plants, thermoelectrics, thermoionics, carbon capture and storage, control of harmful emission from combustion engines, piezoelectrics, turbines and heat exchangers, among others. Advances in additive manufacturing (AM) offer new opportunities to fabricate these devices in geometries unachievable previously and may provide higher efficiencies and performance, all at lower costs. This article reviews the state of the art in ceramic materials for various energy applications. The focus of the review is on material selections, processing, and opportunities for AM technologies in energy related ceramic materials manufacturing. The aim of the article is to provide a roadmap for stakeholders such as industry, academia and funding agencies on research and development in additive manufacturing of ceramic materials toward more efficient, cost-effective, and reliable energy systems.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Neutron irradiation-enhanced grain growth in tungsten and tungsten alloys

Herein, to understand the microstructural stability of candidate plasma-facing materials under fusion-relevant environments, neutron irradiation of W and W-3%Re alloys with and without K and La dopants was performed in the mixed-spectrum High Flux Isotope Reactor at nominal temperatures of ~850 °C and ~1100 °C to calculated doses between 0.42 and 0.47 dpa. To the best of our knowledge, this study presents the first experimental evidence of radiation-enhanced recrystallization in W and undoped W–Re alloys at ~850 °C, conditions where thermal annealing does not cause any grain growth in a similar timescale. Potassium- or lanthanum-doped tungsten alloys showed more resistance to radiation-enhanced grain growth. We explain the acceleration of grain growth by analyzing the self-diffusion constant under atomic displacement environments. The microstructural observations of the studied W variants suggest that La doping is more effective than K doping for mitigating recrystallization. This study also found that radiation-enhanced recrystallization is an important consideration when designing and applying W to plasma-facing components in future nuclear fusion reactors.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Irradiation hardening and ductility loss of Eurofer97 steel variants after neutron irradiation to ITER-TBM relevant conditions

Ten Eurofer97 steel variants, produced by non-standard fabrication-processing routes and modified alloying chemistries, were studied by neutron irradiations in the high flux isotope reactor. The irradiations were performed to ITER-TBM relevant conditions of ~255–350 °C, 2.94–3.24 dpa. We quantified the irradiation-induced degradation of the steels using mechanical property tests. All the steels suffered from irradiation hardening, where a significant increase in Vickers microhardness and yield stress (σ YS ) occurred, accompanied with severe loss of tensile elongation. The extent of hardening was material dependent. For T irr = 300±30 °C, most steels showed σ YS increase in the range of ~30% to as high as ~66%, except for a low temperature tempered steel with σYS increase below 15%. Despite large losses in elongation, most failures were ductile. Significant post-necking ductility was retained with reduction in area (RA) between 65–75%, but <50% for low temperature tempered steels. The ultimate tensile stress to yield stress (σ UTS /σ YS ) ratios decreased significantly after irradiation, highlighting irradiation-induced strain hardening capacity reduction. No major effect of irradiation on the plastic instability stress (σ PIS ) and true fracture stress of the steels was observed. By comparing the tensile stresses in true stress units and with literature, the results suggest that RAFM steel designing should target materials with a large separation between σ PIS and σ YS , to ensure the materials can maintain large work hardening and uniform deformation capability after irradiation. The tensile data of the steels additionally revealed a compelling evidence of an inverse trend between the change in RA and increase in σ YS of the neutron irradiated Eurofer97 type steels.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Thermal diffusivity and thermal conductivity of SiC composite tubes: the effects of microstructure and irradiation

Cladding thermal conductivity is an important physical property in assessing the performance of silicon carbide (SiC)-cladded fuels for nuclear reactors. However, there is a significant lack of reliable data, particularly for irradiated materials, because the geometry complicates the measurement. This study investigates the thermal diffusivity of coupons with a curvature, machined from SiC fiber–reinforced SiC matrix composite tubes, with and without neutron irradiation under light water reactor–relevant temperature and dose conditions. The tested materials included full composite and duplex SiC composite tubes. The measurements were conducted using a modern flash diffusivity apparatus. The analyzed area on the specimen during diffusivity testing was reduced for improved measurement accuracy due to sample curvature. Post-irradiation measurements showed that the effects of neutron irradiation on thermal conductivity (e.g. thermal defect resistivity) are different between SiC composite plates versus tubes. The difference was explained by higher matrix density of the tube than the plate. This study provides reliable thermal properties of prototypic SiC composite tubes useful for fuel performance modeling of SiC-based cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Fiber/matrix debonding evaluation of SiCf/SiC composites using micropillar compression technique

Third-generation silicon carbide (SiC) composites reinforced by SiC fibers (Hi-Nicalon S [HNS] and Tyranno SA3 [SA3]) are attractive for use in next generation reactors owing to their high strength and chemical inertness at high temperatures, as well as enhanced radiation tolerance under neutron irradiation environments. To optimize composite performance, the interfacial mechanical properties of chemical vapor–infiltrated (CVI) SiC f /SiC composites are investigated in this effort by using a slant interface micropillar compression testing procedure. The micropillar test specimens, containing an inclined pyrolytic carbon (PyC) interphase, are prepared using a focused ion beam. The novel microcompression testing successfully quantifies the debond shear strength and internal friction coefficient of micropillar test samples by using the Mohr-Coulomb formulation. Furthermore, according to four types of SiCf/SiC composite microcompression test results, interfacial properties and debond mechanisms are significantly affected by the PyC layer thickness, the local bonding mechanism of the PyC interphase on the SiC fiber surface, and the surface roughness of fibers. Regardless of PyC thicknesses, SA3-reinforced CVI SiC f /SiC composites are found to have much higher debond shear strengths than HNS-reinforced SiC f /SiC CVI composites. By using this micropillar compression technique alongside analytical methods, we uncover new understandings of PyC interface properties. Additionally, the micropillar test results obtained are correlated with macroscopic mechanical properties of neutron-irradiated CVI SiCf/SiC composites.

42 ENGINEERING↗

DOE OSTI2021

Technological aspects in blanket design: Effects of micro-alloying and thermo-mechanical treatments of EUROFER97 type steels after neutron irradiation

Presently available data on neutron irradiation damage raise doubts on the feasibility of using EUROFER97 steel for a water-cooled starter blanket in a DEMO reactor, since the ductile-to-brittle transition temperature (DBTT) increases significantly for irradiation temperatures below 350°C. The additional DBTT shift caused by H and He transmutation can only be estimated based on very few results with isotopically tailored EUROFER97 steel. Conservative calculations show that the DBTT of EUROFER97 steel could exceed the operating temperature in water-cooled starter blankets within a relatively short time period. This paper presents results from a EUROfusion funded irradiation campaign that was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The paper compares ten newly developed reduced activation ferritic-martensitic (RAFM) steels irradiated to a nominal dose of 2.5 dpa at 300°C. The post-irradiation experiments using Small Specimen Test Technology included hardness, tensile, and fracture mechanics tests combined with fractography and microstructure analysis are presented. Results show that micro-alloying EUROFER97-type steels influenced the mechanical properties but a dominating impact on irradiation damage resistance could not be identified. In contrast, specific thermo-mechanical treatments lead to better DBTT behavior. Discussion about irradiation response to heat treatment conditions is also given. Here, despite requiring data also at high dpa values, the results indicate that with these modified materials an increased lifetime and potentially also an increased operating temperature window can be achieved compared to EUROFER97.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Hydrothermal Corrosion of First-Generation Dual-Purpose Coatings on Silicon Carbide for Accident-Tolerant Fuel Cladding

In this study, four different commercial coatings, TiN, Cr, CrN, and Cr/CrN (multilayer) were deposited by physical vapor deposition (PVD) on high resistivity chemical vapor deposited (CVD) SiC for the dual purpose of corrosion mitigation and fission gas hermeticity. The coatings were exposed for up to 2600 hours in flowing high temperature (288-350°C), high purity, liquid water with dissolved hydrogen (0.15-3ppm) or dissolved oxygen (1-2ppm). The Cr/CrN coatings did not adhere well and spalled quickly during exposure. In oxygenated conditions, none of the coatings were protective . CrN coatings spalled preferentially at defect sites, but reacted slowly enough in the absence of defects to potentially be chemically protective. TiN coatings fully oxidized within a few hundred hours and spalled. In hydrogen, all the coatings were chemically stable. Coating failures were identified and attributed to localized spallation, potentially related to coating defects. Coating improvements are recommended to reduce spallation and improve coating protection for future generations of dual-purpose coatings.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Thermal diffusivity of irradiated tungsten and tungsten-rhenium alloys

The Japan-US PHENIX project irradiated tungsten materials in the RB-19J capsule experiment in the High Flux Isotope Reactor (HFIR). A gadolinium (Gd) shielding was used to absorb the thermal neutrons and reduce rhenium and osmium generation in tungsten. Pure tungsten and K-doped W-3% Re samples were irradiated at 532 – 662 °C to dose of 0.21-0.46 dpa, with the grain orientation perpendicular or parallel to the disk surface. Thermal diffusivity measurements were performed from 100 °C to 500 °C. Furthermore, additional measurements followed after annealing up to 900 °C. Irradiated pure tungsten specimens showed similar thermal diffusivity results compared with an unirradiated W-1% Re specimen in another study. The transmutation amount of Re was calculated to be about 0.52% for those specimens that showed good agreement with this study. Specimens irradiated in this study to different doses presented almost the same thermal diffusivity. Annealing up to 800 °C resulted in no recovery of thermal diffusivity. These results show that the contribution of crystalline defects to degradation of thermal diffusivity is quite limited. In addition, the thermal diffusivity of the irradiated specimens was getting close to that of the unirradiated specimens at elevated temperature.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Tensile properties of powder-metallurgical-processed tungsten alloys after neutron irradiation near recrystallization temperatures

The tensile properties of powder-metallurgical-processed Pure W, K-doped W, W-3%Re, and K-doped W-3%Re were examined after neutron irradiation up to 0.7 dpa at 910–1020 °C with a thermal neutron shield in the High Flux Isotope Reactor (HFIR). After irradiation, recrystallized Pure W (R) exhibited a brittle fracture mode, while recrystallized K-doped W-3%Re (R) exhibited a ductile fracture mode at 500 °C. K-doped W-3%Re (R) has fine grains, and hence, contains a considerable number of grain boundaries that act as sinks for irradiation defects. Solid solute Re in the W matrix could improve not only the mechanical properties of W, but also its resistance to neutron irradiation. At 500 °C, the ductility of K-doped W-3%Re after irradiation was significantly higher than that of Pure W. The irradiation at ~1000 °C did not induce hardening of stress-relieved (SR) W materials, but SR W materials tended to exhibit a decrease in the ultimate tensile strength (UTS) and an increase in total elongation (TE). Here, the softening due to the recovery and recrystallization of SR W materials and the hardening due to the formation of irradiation defect clusters were balanced during irradiation at ~1000 °C, and ductility was exhibited without an increase in strength.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Perspectives on the FESAC transformative enabling capabilities: Priorities, plans, and Status

In early 2017, the Fusion Energy Sciences Advisory Committee (FESAC), an advisory committee to the United States Department of Energy, was charged with identifying transformative enabling capabilities (TECs) “that could promote efficient advance toward fusion energy, building on burning plasma science and technology.” A subcommittee with broad expertise was formed and sought feedback from scientific experts, including experts from outside the fusion community. Three workshops were conducted, and a report was approved by FESAC in 2018 that identified four of the “most promising” TECs: advanced algorithms, high-critical-temperature superconductors, advanced materials and manufacturing, and novel technologies for tritium fuel cycle control. In addition, one “promising” TEC was identified: fast flowing liquid metal plasma-facing components. This paper will give details on the promising TECs and an overview on considerations of these TECs in the United States since the publication of the report.

Lumsdaine, Arnold↗

DOE OSTI2020

Electric current–assisted direct joining of silicon carbide

Conventional direct joining technologies are difficult to use with silicon carbide (SiC) materials, especially for fiber composite forms of SiC, because of the harsh conditions required. Therefore, to reduce the temperature and/or process time required for the direct joining process, an electric current–assisted joining (ECAJ) method was studied. Joining of low–resistivity grade, nitrogen doped β-SiC was demonstrated at a relatively low nominal temperature of 1750 °C with a 10 min hold by enhancing the passage of current through the material. The joining mechanism is discussed in terms of localized overheating and accelerated self-diffusion at the interface. In the case of joining at 2160 °C for 1 min, rapid crystal growth of textured SiC was found at the interface. This study indicates that rapid ECAJ-based direct joining is a practical and appropriate method for joining SiC-based materials.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Microstructure Experiments-Enabled MARMOT Simulations of SiC/SiC-based Accident Tolerant Nuclear Fuel System

We have undertaken an experimental-computational project that addresses a few key technology gaps associated with the use of SiC/SiC composites for light water reactor fuel cladding. Two principal endeavors in this project are to assess the irradiation-induced microstructural changes and swelling in SiC/SiC composites, and to characterize and model the porous oxide surface layer (when exposed to steam) along SiC recession that is detrimental to the clad integrity under accident conditions. The project tasks are: (i) ion irradiation and characterization (University of Tennessee, Knoxville), (ii) mechanical tests and analysis (University of South Carolina), (iii) steam exposure tests (Oak Ridge National Laboratory), (iv) electron microscopy and spectroscopic characterization (NC State University), (v) x-ray microscopy and reconstruction (University of South Carolina and NC State University), (vi) phase field modeling and simulations (NC State University and Idaho National Laboratory). The effects of 10 MeV Au ion irradiation at 350°C on the microstructure evolution in SiC/SiC composites are investigated at doses up to 400 displacements per atom (dpa) at the University of Tennessee, Knoxville. Atomic force microscopy and optical profilometry reveal irradiation induced axial and radial shrinkage of the fibers for doses greater than 10 dpa. Based on detailed electron microscopy characterization, the primary cause of the fiber shrinkage is attributed to irradiation-induced loss of carbon packets. Additionally, the multilayer PyC interface is observed to portray high resistance to irradiation damage. The mechanical response and failure mechanisms of un-irradiated samples is also assessed through loading tests and X-ray imaging at the University of South Carolina. Steam exposure tests are performed at the Oak Ridge National Laboratory in a facility that represents a reactor pressure vessel under a loss-of-coolant-accident scenario. The samples that are analyzed methodically in this report are tested for 32/31 hours at 1200°C with a velocity of 0.25 cm/s for the pressures: 0.1 MPa, 0.45 MPa, 0.92 MPa and 1.38 MPa. Scanning/transmission electron microscopy analysis conducted at the NC State University (NCSU) shows that the oxide layer thickness increases with the steam pressure. While the oxide layer is crystalline (α- cristobalite) for the pressures 0.45 MPa, 0.92 MPa and 1.38 MPa, the layer is amorphous at 0.1 MPa. Results from Raman spectroscopy have also confirmed the formation of α-cristobalite phase of SiO₂. The abrupt increase in the integrated Raman intensity ratio between steam pressures 0.1 and 0.45 MPa suggests the onset of accelerated crystallization. Non-destructive three dimensional X-ray microscopy/tomography (XCT) and computational image processing techniques are employed by the University of South Carolina to probe the porous oxidation features on SiC samples at varying pressures. Interestingly, most pores are observed to be located away from the surface as well as the oxide-SiC interface. The average oxide layer thickness assessed from the XCT analysis is seen to be in excellent agreement with the values determined through scanning electron microscopy analysis for the highest pressures where the oxide layer is relatively more uniform. A phase-field model developed by the NC State University and Idaho National Laboratory for simulating oxidation of SiC by steam captures the paralinear kinetics of SiC oxidation with a high degree of fidelity. Results from quasi-one-dimensional and two dimensional simulations show that the pores with oxidizing species lead to a higher volatilization rate. These simulations indicate that the enhanced apparent volatilization rates at higher pressures observed in experiments can be rationalized by the increased volatilization from the pores. The project team has also successfully developed the capability to import images from experiments for realistic evolution of the oxidizing microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗