James Cole

A lanthana-containing nanostructured ferritic steel (NFS) was processed via mechanical alloying (MA) of Fe-14Cr-1Ti-0.3Mo-0.5La2O3 (wt.%) and consolidated via spark plasma sintering (SPS). In order to study the consolidation behavior via SPS, sintering temperature and dwell time were correlated with microstructure, density, microhardness and shear yield strength of the sintered specimens. A bimodal grain size distribution including both micron-sized and nano-sized grains was observed in the microstructure of specimens sintered at 850, 950 and 1050 oC for 45 min. Significant densification occurred at temperatures greater than 950 oC with a relative density higher than 98%. A variety of nanoparticles, some enriched in Fe and Cr oxides and copious nanoparticles smaller than 10 nm with faceted morphology and enriched in La and Ti oxides were observed. After SPS at 950 oC, the number density of Cr-Ti-La-O enriched nanoclusters with an average radius of 1.5 nm was estimated to be 1.2 ×10^24 m^-3 . The La + Ti : O ratio was close to 1 after SPS at 950 and 1050 C; however, the number density of nanoclusters decreased at 1050 C. With SPS above 950 C, the density improved but the microhardness and shear yield strength decreased due to partial coarsening of the grains and nanoparticles.

Elemental powder mixture of Fe–Cr–Ti–Mo and La2O3 were ball milled for different milling times in a high energy shaker mill. Effects of ball milling time on crystallite size, particle size and hardness were investigated using X-ray diffraction (XRD), scanning electron microscopy (SEM) and microhardness tester. After 10 h of ball milling, the smallest crystallite size and highest hardness were ~24 nm and ~970 HV, respectively. Transmission electron microscopy (TEM) studies have revealed nanoscale features 2–5 nm in diameter present in the milled powder. Local atom probe tomography studies have shown that these nanoscale features were possibly nanoclusters enriched in La, TiO and O.

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are also analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. Notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

Formation of a high density of Mn–Ni–Si nanoscale precipitates in irradiated Cu-free and Cu-bearing reactor pressure vessel steels could lead to severe unexpected embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement prediction models, would emerge only at high fluence. However, the mechanisms and variables that control Mn–Ni–Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni contents were carried out at ~295 °C to high and very high neutron fluences of ~1.3 × 1020 and ~1.1 × 1021 n cm-2. Atom probe tomography shows that significant mole fractions of Mn–Ni–Si-dominated precipitates form in the Cu-bearing steels at ~1.3 × 1020 n cm-2, while they are only beginning to develop in Cu-free steels. However, large mole fractions of these precipitates, far in excess of those found in previous studies, are observed at 1.1 × 1021 n cm-2 at all Cu contents. At the highest fluence, the precipitate mole fractions primarily depend on the alloy Ni, rather than Cu, content. The Mn–Ni–Si precipitates lead to very large increases in measured hardness, corresponding to yield strength elevations of up to almost 700 MPa.

This work investigates fuel-cladding chemical interaction (FCCI) between U-25Pu-14Zr (in wt pct) fuel and pure Fe at elevated temperatures, understanding of which is critical for evaluation of the fuel performance. Phases and microstructure formed in the quaternary uranium-plutonium-zirconium-iron (U-Pu-Zr-Fe) system were characterized using the transmission electron microscopy technique. Phases formed within the FCCI layer were identified using selective area electron diffraction (SAED) analysis as Fe2U (Fd- m), Fe2Zr (Fd-3m), a-U (Cmcm), Fe2Pu (Fd-3m), ß-Pu (C12/m1), and ß-Zr (Im-3m). U.S. Government Work. Not Protected by U.S. Copyright. Manuscript submitted June 23, 2015.

In recent years, it has been observed that Inconel X-750 spacers in CANDU reactors exhibits lower ductility with reduced load carrying capacity following irradiation in a reactor environment. The fracture behaviour of ex-service material was also found to be entirely intergranular at high doses. The thermalized flux spectrum in a CANDU reactor leads to transmutation of 58Ni to 59Ni. The 59Ni itself has unusually high thermal neutron reaction cross-sections of the type: (n, c), (n, p), and (n, a). The latter two reactions,in particular,contribute to a significant enhancement of the atomic displacements in addition to creating high concentrations of hydrogen and helium within the material. Microstructural examinations by transmission electron microscopy (TEM) have confirmed the presence of helium bubbles in the matrix and aligned along grain boundaries and matrix–precipitate interfaces. Helium bubble size and density are found to be highly dependent on the irradiation temperature and material microstructure; the bubbles are larger within grain boundary precipitates. TEM specimens extracted from fracture surfaces and crack tips provide information that is consistent with crack propagation along grain boundaries due to the presence of He bubbles.

A lanthana-containing nanostructured ferritic steel (NFS) was processed via mechanical alloying (MA) of Fe-14Cr-1Ti-0.3Mo-0.5La2O3 (wt.%) and consolidated via spark plasma sintering (SPS). In order to study the consolidation behavior via SPS, sintering temperature and dwell time were correlated with microstructure, density, microhardness and shear yield strength of the sintered specimens. A bimodal grain size distribution including both micron-sized and nano-sized grains was observed in the microstructure of specimens sintered at 850, 950 and1050 °C for 45 min. Significant densification occurred at temperatures greater than 950 °C with a relative density higher than 98%. A variety of nanoparticles, some enriched in Fe and Cr oxides and copious nanoparticles smaller than 10 nm with faceted morphology and enriched in La and Ti oxides were observed. After SPS at 950 °C, the number density of Cr–Ti–La–O-enriched nanoclusters with an average radius of 1.5 nm was estimated to be 1.2 × 1024 m−3. The La + Ti:O ratio was close to 1 after SPS at 950 and 1050 °C; however, the number density of nanoclusters decreased at 1050 °C. With SPS above 950 °C, the density improved but the microhardness and shear yield strength decreased due to partial coarsening of the grains and nanoparticles.

A novel nanostructured ferritic steel powder with the nominal composition Fe–14Cr–1Ti–0.3Mo–0.5La2O3 (wt.%) was developed via high energy ball milling. La2O3 was added to this alloy instead of the traditionally used Y2O3. The effects of varying the ball milling parameters, such as milling time, steel ball size and ball to powder ratio, on the mechanical properties and microstructural characteristics of the as-milled powder were investigated. Nanocrystallites of a body-centered cubic ferritic solid solution matrix with a mean size of approximately 20 nm were observed by transmission electron microscopy. Nanoscale characterization of the as-milled powder by local electrode atom probe tomography revealed the formation of Cr–Ti–La–O-enriched nanoclusters during mechanical alloying. The Cr:Ti:La:O ratio is considered “non-stoichiometric”. The average size (radius) of the nanoclusters was about 1 nm, with number density of 3.7x10^24 m^-3. The mechanism for formation of nanoclusters in the as-milled powder is discussed. La2O3 appears to be a promising alternative rare earth oxide for future nanostructured ferritic steels.

Thermally stable nanofeatures with high number density are expected to impart excellent high temperature strength and irradiation stability in nanostructured ferritic steels (NFSs) which have potential applications in advanced nuclear reactors. A lanthana-bearing NFS (14LMT) developed via mechanical alloying and spark plasma sintering was used in this study. The sintered samples were irradiated by Fe2+ ions to 10, 50 and 100 dpa at 30 °C and 500 °C. Microstructural and mechanical characteristics of the irradiated samples were studied using different microscopy techniques and nanoindentation, respectively. Overall morphology and number density of the nanofeatures remained unchanged after irradiation. Average radius of nanofeatures in the irradiated sample (100 dpa at 500 °C) was slightly reduced. A notable level of irradiation hardening and enhanced dislocation activity occurred after ion irradiation except at 30 °C and ⩾50 dpa. Other microstructural features like grain boundaries and high density of dislocations also provided defect sinks to assist in defect removal.

Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) have been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.37 dpa. Atom probe tomography revealed manganese and silicon-enriched clusters in both UFG and CG steel after neutron irradiation. Mechanical properties were characterized using microhardness and tensile tests, and irradiation of UFG carbon steel revealed minute radiation effects in contrast to the distinct radiation hardening and reduction of ductility in its CG counterpart. After irradiation, micro hardness indicated increases of around 9% for UFG versus 62% for CG steel. Similarly, tensile strength revealed increases of 8% and 94% respectively for UFG and CG steels while corresponding decreases in ductility were 56% versus 82%. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation while no significant change was observed in UFG steel, revealing better radiation tolerance. Quantitative correlations between experimental results and modeling were demonstrated based on irradiation induced precipitate strengthening and dislocation forest hardening mechanisms.

Nano-structured ferritic steels (NFSs) with 12-14 wt% Cr have attracted widespread interest for potential high temperature structural and fuel cladding applications in advanced nuclear reactors. They have excellent high temperature mechanical properties and high resistance to radiation-induced damage. The properties of the NFSs depend on the composition that mainly consists of Cr, Ti, W or Mo, and Y2O3 as alloying constituents. In this study, a novel nano-structured ferritic steel (Fe-14Cr-1Ti-0.3Mo-0.5La2O3, wt%) termed as 14LMT was developed via high energy ball milling and spark plasma sintering. Vickers microhardness values were measured. Microstructural studies of the developed NFSs were performed by EBSD and TEM, which revealed a bimodal grain size distribution. A significant number density of nano-precipitates was observed in the microstructure. The diameter of the precipitates varied between 2-70 nm and the morphology from the spherical to faceted shape. The Cr-La-Ti-O-enriched nano-clusters were identified by APT studies.

Elemental powder mixture of Fe–14Cr–1Ti–0.3Mo–0.5La2O3(wt%) composition is mechanicallyalloyed for different milling durations (5, 10 and 20 h) and subsequently consolidated via spark plasmasintering under vacuum at 950?C for 7 min. The effects of milling time on the densi?cation behaviorand density/microhardness are studied. The sintering activation energy is found to be close to that ofgrain boundary diffusion. The bimodal grain structure created in the milled and sintered material isfound to be a result of milling and not of sintering alone. The oxide particle diameter varies between2 and 70 nm. Faceted precipitates smaller than 10 nm in diameter are found to be mostly La–Ti–Cr-enriched complex oxides that restrict further recrystallization and related phenomena

Exposure to high temperatures and irradiation results in interaction and interdiffusion between fuel and cladding constituents that can lead to formation of undesirable brittle or low-melting point phases. A diffusion couple study has been conducted to understand fuel-cladding interaction occurring between Ue22Pue4Zr (in wt%) fuel and pure Fe at elevated temperatures. The phases formed within fuel claddingchemical interaction (FCCI) layer have been characterized in the transmission electron microscope (TEM). The phases formed within FCCI layer have been identified as Fe2U (Fd-3m), FeU6 (I4/mcm), Fe2Zr (Fd-3m), FeZr2 (I4/mcm), Fe2Pu (Fd-3m), UZr2 (P6/mmm), b-Zr (Im-3m), and ZrO2 (Fm-3m).

Phases and microstructure in as-cast, annealed at 850 °C, and subsequently cooled U–23Pu–9Zr fuel were characterized using scanning and transmission electron microscopy techniques. SEM examination shows formation of three phases in the alloy that were identified in TEM using selective area diffraction pattern analysis: α-Zr globular and elongated δ-UZr2 inclusions and a thick oxide layer formed on top of β-Pu phase, which has been initially assumed to be ζ-(U, Pu). However, further examination of the cross-sectional TEM specimens identified the matrix phases as δ-UZr2, β-Pu, and (U, Zr)ht. Two types of inclusions were observed in the immediate vicinity of the specimen surface and they were consistent with α-Zr and ζ-(U, Pu).