Main objective of this work is to develop, by systematic variation of the chemical composition, and TMP, 14% Cr nano-structured ferritic alloys with significantly improved high-temperature properties compared to currently available ODS alloys. Application of state-of-the-art characterization tools shall lead to an integrated understanding of structure-property correlation and the formation mechanism of nanoparticles.
Written in a user-friendly, conversational style, the fourth edition of this groundbreaking text helps pre-service and in-service mathematics teachers build the comfort and confidence they need to begin talking to children about fractions and ratios, distilling complex ideas and translating research into usable ideas for the classroom. For two decades, Teaching Fractions and Ratios for Understanding has pushed readers beyond the limits of their current understanding of fractions and rational numbers, challenging them to refine and explain their thinking without falling back on rules and procedures they have relied on throughout their lives. All of the material offered in the book has been used with students, and is presented so that readers can see the brilliance of their insights as well as the issues that challenge their understanding. Each chapter includes children’s strategies and samples of student work for teacher analysis, as well as activities for practicing each thinking strategy, designed to be solved without rules or algorithms, using reasoning alone. The fourth edition of this popular text has been updated throughout and includes new examples of student work, updated artwork, and more. As with previous editions, an equally valuable component of this text is the companion book MORE! Teaching Fractions and Ratios for Understanding (2012), a supplement that is not merely an answer key but a resource that provides the scaffolding for the groundbreaking approach to fraction and ratio instruction explored here. MORE! includes in-depth discussions of selected problems in the main text, supplementary activities, Praxis preparation questions, more student work, and templates for key manipulatives.
Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and grain boundaries in the ion-irradiated alloys. More significant segregation was observed in the neutron irradiated alloys. For the more concentrated alloys, irradiation did not affect existing Cr segregation to grain boundaries and segregation to dislocation loops was not observed perhaps due to a change in the dislocation loop structure with increasing Cr concentration. Precipitation of [alpha]' was observed in the neutron irradiated alloys containing over 9 at.% Cr. However ion irradiation appears to suppress the precipitation process. Initial low dose ion irradiation experiments strongly suggest a cascade recoil effect. The systematic analysis of grain boundary orientation on Cr segregation was significantly challenged by carbon contamination during ion irradiation or by existing carbon and therefore carbide formation at grain boundaries (sensitization). The combination of the proposed systematic experimental approach with atomistic modeling of diffusion processes directly addresses the programmatic need for developing and benchmarking predictive models for material degradation taking into account atomistic kinetics parameters.
An oxide dispersion strengthened (ODS) ferritic steel with high temperature strength has been developed in line with low activation criteria for application in fusion power systems. The composition Fe-13.5Cr-2W-0.5Ti-0.25Y2O3 was chosen to provide a minimum chromium content to insure fully delta-ferrite stability. High temperature strength has been demonstrated by measuring creep response of the ODS alloy in uniaxial tension at 650 and 900 C in an inert atmosphere chamber. Results of tests at 900 C demonstrate that this alloy has creep properties similar to other alloys of similar design and can be considered for use in high temperature fusion power system designs. The alloy selection process, materials production, microstructural evaluation and creep testing are described.
Uniaxial tension creep response is reported for an oxide dispersion strengthened (ODS) steel, Fe-13.5Cr-2W-0.5Ti-0.25 Y2O3 (in weight percent) manufactured using the mechanical alloying process. Acceptable creep response is obtained at 900°C.
Reduced-activation ferritic steels are considered as the primary candidate materials for structural applications within nuclear fusion power plants. It is known that by mechanically alloying ferritic steel powder with Y (usually in the form ofY203) then consolidating the material by hot isostatic pressing, a nanoscale dispersion of oxygen rich nanoclusters as small as ~2nm is introduced into the microstructure. This vastly improves high temperature strength and creep resistance, and the nanoclusters also act as trapping sites for helium and point defects produced under irradiation. In this thesis, the evolution of the oxide nanoclusters in a Fe-14Cr-2W-0.3Ti & 0.3Y203 ODS alloy was investigated primarily using atom probe tomography. The microstructure was characterized at various points during processing to give an insight into the factors influencing the formation of the nanoclusters. It was found that the nanoclusters nucleated during the mechanical alloying stage, then followed near classical nucleation and growth mechanisms keeping the same composition of ~8%Y, ~12%Ti,~2S%O and ~4S%Cr throughout. The formation and evolution of S-lSnm grain boundary oxides was also observed, and these were shown to form first as Cr203 particles that subsequently transform into a Y-Ti-O based oxide on further processing. The influence of mechanical alloying with 0.Swt.%Fe2Y rather than 0.3wt.%Y203 was also investigated, and this showed that there was no difference in the final microstructure produced provided the level of Ti in the starting powder was tightly controlled. Without sufficient Ti, the nanoclusters were Y-O based and ~6nm diameter. Both the Y-O and Y-Ti-O nanoclusters were moderately stable on annealing at 1200°C for up to 100 hours, with only minimal coarsening observed. Ti was found not to influence the coarsening rate of the nanoclusters significantly. The stability of the oxide nanoclusters under irradiation was investigated by using Fe2+ ion irradiation to simulate displacement cascade damage in the ODS-Eurofer material (the official European candidate material for testing in the ITER fusion test reactor). Doses up to ~6 dpa at 400°C were used, and there was no significant change to the nanocluster distribution. However segregation of Mn to dislocations was observed after irradiation. These results indicate that ODS steels are good candidate structural materials, as the microstructure is stable at high temperature and under irradiation. The starting powders, and processing parameters need to be tightly controlled in order to produce the optimal material for use in service.
Tempered martensite ferritic steels are used for critical components in fossil fired power plants that operate in the creep range. The materials contain a high density of dislocations and precipitates form on all types of internal interfaces, the majority of which represent subgrain boundaries. Most previous studies suffer from either only relating to short term creep experiments or from being incomplete in not considering all relevant elements of the microstructure. No systematic effort was made to investigate the evolution of microstructures under conditions of long term creep. In the present study the evolution of the microstructure of a 12% Cr tempered martensite ferritic steel was investigated under conditions of long term aging and creep. Transmission electron microscopy (TEM) and electron back scattered diffraction (EBSD) techniques were used to characterize materials from interrupted creep tests (0.5%, 1%, 1.6% and rupture at 11.9%; creep conditions: 550°C, 120 MPa, rupture time: 139 971 h). It is shown that subgrains coarsen, that the close correlation between carbides and subgrain boundaries loosens during long term creep, and that the frequency of small angle boundaries increases. In addition, the evolution of dislocation densities during long term aging and creep was studied using high angle annular dark field (HAADF) scanning transmission electron microscopy (STEM). During aging the dislocation density remains constant, while during long term creep the dislocation density continuously decreases. All these elementary deformation processes have already been discussed in short term creep studies. The present study shows that they also govern long term creep, however, during long term creep, precipitation and coarsening reactions occur which are not observed during short term creep. Cr rich M23C6, VX carbides and Laves phase were identified as the major precipitates in the microstructure of the 12% Chromium tempered martensite ferritic steel. Their chemical compositions, sizes, volume fractions and number densities were evaluated in all interrupted specimens. M23C6 particles coarsen and establish their equilibrium concentration after 51072 hours. VX particles are stable. The Laves phase particles do not reach thermodynamic equilibrium as they form and grow during long term creep. This is due to Silicon which is found in the Laves phase particles and which diffuses slowly in the steel matrix.
This paper deals with the irradiation behavior of the oxide dispersion strengthened (ODS) ferritic alloy DT2203Y05, a 13% Cr ferritic alloy strengthened by a fine dispersion of yttrium and titanium oxides. This alloy was irradiated up to 81 dpa in Phénix as fuel pin cladding. The profilometry measurements confirm its high swelling resistance. Few voids, mainly associated with oxides, are observed at low irradiation temperatures, but this alloy is severely embrittled by irradiation. A few cracks are observed in the lower 2/3 of the fissile column, and the longitudinal tensile tests show hardening and severe ductility loss induced by irradiation along the whole fuel column.
