James Stubbins

This research qualitatively reviews literature regarding energy system modeling in Japan specific to the future hydrogen economy, leveraging quantitative model outcomes to establish the potential future deployment of hydrogen in Japan. The analysis focuses on the four key sectors of storage, supplementing the gas grid, power generation, and transportation, detailing the potential range of hydrogen technologies which are expected to penetrate Japanese energy markets up to 2050 and beyond. Alongside key model outcomes, the appropriate policy settings, governance and market mechanisms are described which underpin the potential hydrogen economy future for Japan. We find that transportation, gas grid supplementation, and storage end-uses may emerge in significant quantities due to policies which encourage ambitious implementation targets, investment in technologies and research and development, and the emergence of a future carbon pricing regime. On the other hand, for Japan which will initially be dependent on imported hydrogen, the cost of imports appears critical to the emergence of broad hydrogen usage, particularly in the power generation sector. Further, the consideration of demographics in Japan, recognizing the aging, shrinking population and peoples’ energy use preferences will likely be instrumental in realizing a smooth transition toward a hydrogen economy.

In this study, we employed pressurized creep tubes to investigate the biaxial thermal creep behavior of Inconel 617 (Alloy 617) and Haynes 230 (Alloy 230). Both alloys have been considered to be the primary candidate structural materials for very high temperature reactors (VHTRs) due to their exceptional high-temperature mechanical properties. The current creep experiments were conducted at 900°C for the effective stress range of 15–35 MPa. For both alloys, complete creep strain development with primary, secondary, and tertiary regimes were observed in all studied conditions. The tertiary creep was found to be dominant in the entire creep lives of both alloys. With increasing applied creep stress, the fraction of the secondary creep regime decreases. The nucleation, diffusion, and coarsening of creep voids and carbides on grain boundaries was found to be the main reason for the limited secondary regime, and was also found to be the major cause of creep fracture. The creep curves computed using the adjusted creep equation of the form ε = Aσ cosh−1(1 + rt) + Pσntm agree well with the experimental results for both alloys at the temperatures of 850–950°C.

Paper published with permission.

In this study, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region ~7.5 μm in depth; the peak damage (~40 dpa) was estimated to be at ~7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing of the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials’ radiation damage in a dynamic manner.

Many R&D programs are being conducted worldwide to study and, in some cases, develop materials with the properties required by Gen IV reactor systems. These advanced reactor systems require materials having sufficient dimensional stability and mechanical properties to meet the anticipated near- and long-term service demands under extreme conditions. These conditions are most challenging in environments with high levels of irradiation damage coupled with requirements for corrosion resistance and severe mechanical loading conditions. Based on past experience, austenitic alloys based around 316SS compositions, 15% Cr-15% Ni-Ti, have not been able to meet these demanding conditions at elevated temperatures. Ferritic-martensitic steels and ferritic-martensitic ODS steels have been proposed for these conditions. However, long-term fast neutron irradiation results show that austenitic alloys of the 15%Cr–25%Ni-Ti-Nb class have excellent swelling resistance at temperatures up to 600°C and at doses larger than 150 dpa. A major reason for this excellent behaviour is the thermal-mechanical treatment of the alloy and the presence of both Ti and Nb to doubly stabilize the precipitate structure.

This material could represent an effective option for claddings in SFR and other advanced reactor applications. For this reason, in consideration of 60-y-of-service requirement, predictions of long-term materials behaviour will be carried out taking into account ageing effects (from 500° to 700°C). A complete analysis aiming to characterize the microstructure of the as-received material, cold-worked at 20%, and also after thermal treatment is under way. Ageing tests as well as TEM and SEM analysis of the microstructural evolution as a function of aging condition will be performed. The evolution of some phases in the common nucleation and growth sites will be presented.

High-energy synchrotron radiation has proven to be a powerful technique for investigating fundamental deformation processes for various materials, particularly metals and alloys. In this study, high-energy synchrotron X-ray diffraction (XRD) was used to evaluate Alloy 617 and Alloy 230, both of which are top candidate structural materials for the very-high-temperature reactor (VHTR). Uniaxial tensile experiments using in-situ high-energy X-ray exposure showed the substantial advantages of this synchrotron technique. First, the small volume fractions of carbides, e.g., ∼6% of M6C in Alloy 230, which are difficult to observe using laboratory-based X-ray machines or neutron scattering facilities, were successfully examined using high-energy X-ray diffraction. Second, the loading processes of the austenitic matrix and carbides were separately studied by analyzing their respective lattice strain evolutions. In the present study, the focus was placed on Alloy 230. Although the Bragg reflections from the γ matrix behave differently, the lattice strain measured from these reflections responds linearly to external applied stress. In contrast, the lattice strain evolution for carbides is more complicated. During the transition from the elastic to the plastic regime, carbide particles experience a dramatic loading process, and their internal stress rapidly reaches the maximum value that can be withstood. The internal stress for the particles then decreases slowly with increasing applied stress. This indicates a continued particle fracture process during plastic deformations of the γ matrix. The study showed that high-energy synchrotron X-ray radiation, as a nondestructive technique for in-situ measurement, can be applied to ongoing material research for nuclear applications.

The very high temperature gas-cooled reactor (VHTR), with dual capacities of highly efficient electricity generation and thermochemical production of hydrogen, is considered as one of the most promising Gen-IV nuclear systems. The primary candidate materials for construction of the intermediate heat exchanger (IHX) for the VHTR are alloy 617 and alloy 230. To have a better understanding of the degradation process during high temperature long-term service and to provide practical data for the engineering design of the IHX, aging experiments were performed on alloy 617 and alloy 230 at 900°C and 1000°C. Mechanical properties (hardness and tensile strength) and microstructure were analyzed on post-aging samples after different aging periods (up to 3000 h). Both alloys attained increased hardness during the early stages of aging and dramatically soften after extended aging times. Microstructural analysis including transmission electron microscopy, scanning electron microscopy, energy dispersive X-ray spectroscopy, and electron backscatter diffraction was carried out to investigate the microstructure evolution during aging. A carbide particle precipitation, growth, and maturing process was observed for both alloys, which corresponds to the changes of the materials’ mechanical properties. Few changes in grain boundary character distribution and grain size distribution were observed after aging. In addition, high temperature corrosion studies were performed at 900°C and 1000°C for both alloys. Alloy 230 exhibits much better corrosion resistance at elevated temperature compared with alloy 617.

Alloy 617 and Alloy 230 are solid-solution strengthened nickel based superalloys, which have been considered two of the most promising structural materials for the Very-High-Temperature Reactor (VHTR). In order to have a better understanding of the degradation process of the materials in the VHTR, long-term aging experiments have been carried out to investigate the dynamic process of microstructure evolution at 900 and 1000°C for Alloy 617 and Alloy 230. The microstructural evolution process in different aging periods (up to 3000 hours) was analyzed by Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM) and Electron Backscatter Diffraction (EBSD). A diffusion-controlled precipitation and coarsening of carbide particles (mainly M23C6 and M6C) for both alloys was observed. The corresponding characteristics of the precipitates, i.e. type, size and coherence, were analyzed. The coarsening rate of the intergranular precipitates in Alloy 617 was found to be much faster compared to Alloy 230’s. The inhomogeneous precipitation process in the transverse plane of Alloy 617 was observed, which may be attributed to the alignment of the inclusion particles induced by the hot rolling. Hardness and tensile tests were carried out to investigate the aging impacts on materials’ strength. Both alloys obtained increased hardness and strength during early stages of aging and softened after elongated time. The results of mechanical tests were in a good agreement with the microstructure evolution process.

Direct liquid fuel cells, such as the direct methanol fuel cell (DMFC) and borohydride fuel cell, may experience severe electrochemical corrosion problems due to the presence of conducting liquid fuels. This type of fuel cells has catalytic electrodes that are permeable to the liquid fuels. This permeability to liquids or solutions then opens up a possibility to the electrochemical improvement of catalytic electrodes once excessive corrosion has occurred. The in-situ electro-deposition or electroplating of the catalytic diffusion electrode is developed to first improve the performance and secondly mitigate the corrosion problem. The experimental details of such a technique are described in this paper. Direct liquid fuel cells thus treated typically have power density up to 40% higher, compared to fuel cells whose catalysts are prepared the traditional way. It is well known that power density is one of the enabling factors for portable power sources, for which the typical direct liquid fuel cells are designed. Therefore, the in-situ electroplating technique is expected to play an important role in the commercialization of fuel cell technology.

We study the mechanisms by which gas atoms such as helium and hydrogen diffuse and interact with other defects in bcc metals and investigate the effect of these mechanisms on the nucleation of embryonic gas bubbles. Large quantities of helium and hydrogen are produced due to spallation and transmutation in structural materials in fusion and accelerator-driven reactors. The long time evolution of the extrinsic gas atoms and their accumulation at vacancies is studied using a kinetic Monte Carlo algorithm that is parameterized by the migration energies of the point defect entities. First-order reaction kinetics are observed when gas clusters with vacancies. If gas-gas clustering is allowed, mixed-order diffusion limited kinetics are observed. When dissociation of gas from clusters is allowed, gas-vacancy clusters survive to steady state while gas-gas clusters dissolve. We obtain cluster size distributions and reaction rate constants that can be used to quantify microstructural evolution of the irradiated metal.

316L stainless steel has been widely used for structural component applications in irradiation environments. The impact of manufacturing or service-induced defects on material tensile response and fracture behavior has been a longstanding concern. In this study, electron backscattering diffraction and finite element methods have been utilized to examine and model the tensile response of 316L stainless steel with different notch geometries. The tensile response is influenced by twinning at room temperature, while it is completely absent at temperatures higher than 200°C. In addition, irradiation exposure increases the yield point of the material and reduces the ductility. It is found that the larger plastic deformation area and severe tip blunting in ductile, unirradiated material promote uniformly distributed post-yield strain hardening processes. In this case, fracture initiates internally for tensile and C-notched specimen geometries, but from the notch tip for sharp V-notched specimen geometries. When irradiation-induced hardening is considered, similar fracture modes are found, however, strains to fracture are significantly reduced.

We report here that high energy electron irradiation of multiwalled boron nitride nanotubes can be used to form sharp, crystalline, conical tips, or to cut boron nitride nanotubes by controlling the electron beam size. Electron beam cutting is observed when a focused electron beam with a diameter much smaller than the tube diameter is used. The tip formation is observed when a shaped, disklike, electron beam is used to irradiate the tube; the diameter of the beam in this case is similar to the tube diameter. In situ electron microscopy observation shows that the tip formation effect is driven by layer peeling and the collapse of the inner walls of the nanotube. This is very different from the formation of nanoarches observed during cutting. The combination of shaping and cutting can be used to fabricate atomically sharp tips for field emitters, nanoimaging, and manipulations.

Extended abstract of a paper presented at Microscopy and Microanalysis 2006 in Chicago, Illinois, USA, July 30 – August 3, 2005

Microstructural defects are introduced in materials upon irradiation with energetic particles. These defects can cause degradation of mechanical properties and contribute to material failure. Transmuted helium in irradiated stainless steels exerts deleterious effects on material properties. We have performed kinetic Monte Carlo (kMC) simulations of point defect diffusion and clustering in bcc alpha iron. The model includes helium and vacancy diffusion and spontaneous clustering and dissociation of the point defects from the clusters. We employ the kMC simulations to investigate the time evolution of the point defect configuration leading to defect clustering and bubble formation. The concentration of embryonic point defect clusters is determined as a function of the simulation time.

We report on a high-resolution electron diffraction study of the structure of individual multiwalled boron nitride nanotubes (MW-BNNTs). The tube chirality was determined by electron diffraction. Diffraction patterns were recorded from small sections of the nanotubes, ∼125nm long, using the nanoarea electron diffraction technique. Accurate measurements of the MW-BNNT chiral angles and their distribution were made from diffraction patterns. Generally, the tube chiralities within each MW-BNNT are strongly correlated; clustering around a single chirality with a dispersion of a few degrees. Multihelix nanotubes were rarely observed. Statistics based on 67 nanotubes revealed a dispersion of the chiral angles (α) with some preference of tubes in the ranges of 10°⩽α⩽15° and 25°⩽α⩽30°. Since various properties of nanotubes depend on the tube structure (diameter and chirality), the results presented here have general significances to nanotube growth and applications.

This paper analyzes experimental fatigue response for selected unirradiated and irradiated reactor materials: AISI 316 stainless steel, ferritic/martensitic MANET and HT9 steels. Available tensile test results on the same or similar materials are used to predict changes in fatigue response using the Universal Slopes method. The predictions are compared with the experimental data to assess the potential for using tensile data to predict reactor component fatigue response. It was found that the effect of irradiation on fatigue life was less severe than on tensile properties. However, tensile properties are useful for qualitative predictions of fatigue response.

In this work, the potential for obtaining reliable fatigue crack growth data from miniature notched bar specimens in three-point bend loading is examined. These specimens, at 2.0 mm wide by 0.8 mm thick by 7.9 mm long, are substantially smaller than any existing fatigue crack growth specimens. This work describes the development of subsize specimen fatigue crack growth rate test techniques using these miniature bar specimens. Fatigue crack growth data are reported for three alloys, 304L SS, Alloy 718 and Mod. 9Cr-1Mo steel. The data obtained using the subsize specimens compare favorably to crack growth rate data obtained on ASTM standard size specimens. The fracture surface microstructures of the subsize and standard size specimens are compared to show the similarities of the crack growth mechanisms in the two specimen configurations. Since three different materials were tested, it is possible to draw conclusions about the limitations of using these extremely small specimens in terms of material yield and ultimate strength and other standard tensile properties.

The effect of S segregation to grain boundaries on the intergranular embrittlement of Ni has been studied at room temperature using Auger electron spectroscopy and slow strain rate tensile tests. The grain-boundary S concentration was varied by time-controlled annealing of dilute Ni–S alloy specimens at 625 °C. The ductile-to-brittle transition in Ni, as determined from percent integranular fracture and reduction-in-area measurements, occurred over a narrow range of S concentrations centered on 15.5±3.4 at. % S. This critical S concentration for 50% intergranular fracture of polycrystalline Ni is similar to the 14.2±3.3 at. % S required to induce 50% amorphization of single-crystal Ni by S+-ion implantation. This suggests that segregation-induced intergranular fracture, like implantation-induced amorphization, may be a disorder-induced polymorphous melting process. In agreement with experimental observations, the polymorphous melting curve for the Ni–S solid solution on the phase diagram drops rapidly to zero as the alloy composition approaches ∼18 at. % S. The critical grain-boundary concentration for intergranular fracture, while slightly less, is within experimental error of the concentration predicted for polymorphous melting as well as that measured for ion-implantation-induced amorphization.

Carbon dioxide (CO2) corrosion products formed on low-carbon steel test coupons exposed to multiphase flows of CO2-oil-salt water mixtures were analyzed. The effects of flow, temperature, pressure, and fluid composition on formation of these corrosion surfaces also were examined. Type 1018 steel (UNS G10180) coupons were exposed in a high-pressure horizontal flow loop. Examination of the exposed coupons was performed using scanning electron microscopy (SEM) with energy-dispersive x-ray (EDX) analysis, Auger electron spectroscopy (AES), and x-ray photoelectron spectroscopy (XPS). Three common features were discovered among the coupons: (1) films with an iron corrosion product, (2) bare steel surfaces (with no film) containing iron carbide (Fe3C) structures in a ferrite matrix, and (3) crystalline features found to be iron carbonate (FeCO3) or salts from the seawater. The effects of flow, temperature, pressure, and fluid composition on formation of these surfaces were studied by comparing coupons exposed to similar conditions. Little change in morphology of the corrosion surface was found as the temperature and CO2 partial pressure increased. More apparent changes, however, appeared with declining salt water percentages in the fluid. The most evident changes in corrosion morphology emerged with variations in flow.

In situ fracture studies have been carried out on thin films of the NiTi intermetallic compound under plane stress, tensile loading conditions in the high-voltage electron microscope. Local stress-induced amorphization of regions directly in front of moving crack tips has been observed. The upper cutoff temperature, TC–Amax, for the stress-induced crystalline-to-amorphous transformation was found to be 600 K, identical to that for heavy ion-induced amorphization of NiTi and for ion-beam mixing-induced amorphization of Ni and Ti multilayer specimens. 600 K is also both the lower cutoff temperature, TA–Cmin, for radiation-induced crystallization of initially-unrelaxed amorphous NiTi and the lowest isothermal annealing temperature, TXmin, at which stress-induced amorphous NiTi crystallizes. Since TXmin should be TK, the ideal glass transition temperature, the discovery that TC–Amax=TA–Cmin=TXmin=TK implies that disorder-driven crystalline-to-amorphous transformations result in the formation of the ideal glass, i.e., the glassy state that has the same entropy as that of the defect-free crystal. As the glassy state with the lowest free energy, its formation can be understood as the most energetically-favored, kinetically-constrained response of crystalline alloys driven far from equilibrium.

As part of a study to determine the potential for using aluminum alloys as structural components in accelerators used to produce tritium, tensile specimens and transmission electron microscopy (TEM) disks of two aluminum alloys in two tempers were irradiated with neutrons from spallation reactions in the Los Alamos Spallation Radiation Effects Facility (LASREF) at the Los Alamos Meson Physics Facility (LAMPF) at 90 to 120°C to a fluence of ~4×1020 n/cm2. Tensile and shear punch tests were performed on the irradiated alloys as well as on a similar unirradiated control matrix with a larger number of specimens, and a correlation between shear and tensile properties was developed for the unirradiated alloys and the irradiated alloys. Comparison of the correlations for the unirradiated and irradiated alloys showed that small differences in the correlations existed, however for purposes of estimation of uniaxial tensile properties, either correlation was found to be adequate.

The Accelerator Production of Tritium (APT) and the Accelerator Transmutation of Waste (ATW) programs require structural materials which retain good mechanical properties when exposed in a spallation neutron irradiation environment. One group of materials likely to withstand the environment anticipated for these systems is the aluminum alloy series. To characterize this class of materials in a prototypical irradiation environment, Al5052 (Al-2.7Mg) and Al6061 (Al-1.1Mg-0.5Si) in hardened and annealed conditions were irradiated to a fluence of 4.2×1020 neutrons/cm2 at ~100°C in a spallation neutron source. Following irradiation, tensile tests and post-test examinations were performed to determine the influence of irradiation and test temperature on mechanical properties and fracture mode. It was found that, the properties of these two aluminum alloys were not significantly affected by the irradiation exposure conditions examined here. Thus these materials may be acceptable as structural materials for APT and ATW applications. This conclusion is based on limited mechanical properties testing, supported by other information in the literature on the performance of these materials in other irradiation environments.

Pure nickel was irradiated in EBR-II without active temperature control to two exposures levels (14 and 31 dpa) at target temperatures of 425, 500, and 600°C in three starting conditions: annealed, 30% cold worked, 30% cold worked and aged at 650°C for 10 hr. The same material conditions were irradiated with active temperature control in FFTF to 33–36 dpa at 430, 520, and 600°C. Post-irradiation density measurements show that in the annealed condition swelling tends to saturate at levels strongly dependent on irradiation temperature. In EBR-II at 600°C. saturation occurs before 12 dpa but swelling is still increasing slowly at 425 and 500°C. The tendency toward saturation of swelling is enhanced by cold working. In addition, the temperature dependence of the saturation level is strongly decreased by cold-working. The maximum swelling under most irradiation conditions is on the order of 8%. In the isothermal FFTF irradiations, the effectiveness of cold working in promotion of swelling is much less pronounced at the higher irradiation temperatures, particular at 600°C. These results are consistent with most evidence from other neutron and charged particle irradiation studies on pure nickel. In comparing the results of these and previous studies, it appears that the saturation is linked to a progressive collapse of the dislocation network when voids begin to dominate the microstructure.

Magnetic properties may be a useful means for monitoring the evolution of irradiation-induced damage in materials which are inherently ferromagnetic. Such materials include pure iron and nickel which are commonly used in a variety of irradiation applications for fluence or spectrum analysis. One such application is for fluence monitoring in Charpy coupon surveillance packets. This study examines the magnetic response of both iron and nickel wires which have been irradiated to doses up to 3.6×1019 cm−2 (E>lMeV) in such surveillance packets at reactor ambient irradiation temperatures of approximately 290°C. The results show that large changes in magnetic response are found for magnetic properties which are sensitive to materials microstructure. The changes are predominantly to decrease the magnetic remanence and coercivity with dose. These changes saturate at the highest doses examined in this study. Though direct microstructural examination is needed to confirm the hypothesis, the results are consistent with the development of a large population of defect structures (vacancy and interstitial clusters) which evolve with irradiation dose.

High temperature crack growth behavior is investigated over a wide range of R-ratios, frequencies, and temperatures in Alloy 800H. It is found that high R-ratio, low frequency, or high temperature can enhance creep damage and thus induce an intergranular crack growth mode. At low frequencies, the nonlinear fracture mechanics parameter, C*, is found to correlate time-dependent fatigue crack growth rate well if the applied mean stress is used in calculating C*. On the other hand, the Paris crack growth law using Keff is proven to be an adequate expression to use when fatigue (time-independent) damage dominates. These conclusions correlate well with damage mechanisms observed from sample fracture surfaces.

A novel technique to encapsulate metallic mechanical properties test specimens in a controlled atmosphere has been developed that eliminates the need for cumbersome retort systems. The procedure consists of brazing stainless steel flexible bellows tubing on a test specimen in such a manner that the tubing acts as a miniature retort around the gage section of the specimen. While convenient, this method demands a special correlation relating strain in the gage section to displacement between the specimen shoulders in order to conduct displacement-controlled fatigue tests. Results of such tests using vacuum encapsulated specimens were generally excellent and agreed reasonably well with data from related studies, supporting the validity of both the encapsulation technique and the subsequent correlation.

Three high-purity Fe-Ni-Cr alloys have been irradiated with heavy ions and with electrons at 450, 550, and 650°C to doses up to 60 dpa. The heavy ion irradiations were carried out with Ni6+ ions, and specimens were irradiated both with and without helium preinjection. The electron irradiations were carried out in a high-voltage electron microscope (HVEM) on specimens without helium preinjection. The alloy compositions were Fe-15 Ni-10 Cr, Fe-15 Ni-16 Cr, and Fe-15 Ni-20 Cr, all with low carbon levels. The latter two of these compositions are austenitic at room temperature, whereas the first can include martensite on rapid cooling. According to the computer calculated phase diagrams of Watkins et al, all of these alloy compositions would be metastable at the lowest irradiation temperature in the present study, but the austenitic phase structure should be stable in all of the alloys at the highest irradiation temperature.

Voids were formed in all of these alloys during irradiation at 650°C to 30 dpa, and swelling levels at that dose were independent of the presence or absence of preinjected helium. The more interesting irradiation behavior consisted of irradiation induced phase transformations, predominantly during irradiation at 450°C. The Fe-15 Ni-16 Cr alloy precipitated a fine distribution of M23C6 after 30 dpa at 450°C under heavy ion irradiation. Similar precipitates were not found in either of the two other alloys under similar irradiation conditions. The formation of these precipitates in a low carbon alloy was surprising. The precipitates were nucleated uniformly and were not associated with dislocation loops or other microstructural features.

The more striking irradiation induced phase transformation was the evolution of austenitic regions in the martensitic laths in the Fe-15 Ni-10 Cr alloy during irradiation at 450°C. This martensitic reversion reaction seems to involve a mechanism associated with dislocation loop formation. The martensitic reversion reaction was observed for materials irradiated both by HVEM and by heavy ions. The austenitic precipitates were stable and grew with irradiation at 450°C. They were found to nucleate in all orientations consistent with the Nishiyama-Wassermann relation in the body centered cubic matrix. This paper develops an explanation of this behavior based on dislocation loop morphology and relative phase stability.