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Sample records for austenitic alloys irradiated

  1. Flux and composition dependence of irradiation creep of austenitic alloys irradiated in PFR at ˜420°C

    Science.gov (United States)

    Toloczko, M. B.; Garner, F. A.; Standring, J.; Munro, B.; Adaway, S.

    1998-10-01

    Swelling and irradiation creep of five austenitic stainless steel alloys irradiated at ˜420°C in the Prototypic Fast Reactor (PFR) were examined. The specimens were in the form of pressurized creep tubes, constructed in the USA and irradiated in PFR in a joint USA/UK experiment. The alloy compositions varied greatly, with the greatest elemental variation in the nickel content, which ranged from 15% to 40% over the five alloys. For each alloy, at least two identical sets of tubes were constructed. Each tube-set was irradiated at a different neutron flux level. Swelling was observed to vary with both alloy composition and flux. Irradiation creep was examined from the perspective of the overlineB= ɛ¯˙/ overlineσ=B 0+D Ṡ creep model. The values of both creep coefficients, B0 and D, were typical for austenitic stainless steels and were found to be insensitive to flux over the range of fluxes in this experiment. However, the creep coefficients may be mildly sensitive to alloy composition.

  2. The compositional dependence of irradiation creep of austenitic alloys irradiated in PFR at 420{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Munro, B. [AEA Technology, Dounreay (United Kingdom)] [and others

    1997-04-01

    Irradiation creep data are expensive and often difficult to obtain, especially when compared to swelling data. This requires that maximum use be made of available data sources in order to elucidate the parametric dependencies of irradiation creep for application to new alloys and to new environments such as those of proposed fusion environments. One previously untapped source of creep data is that of a joint U.S./U.K. experiment conducted in the Prototype Fast Reactor (PFR) in Dounreay, Scotland. In this experiment, five austenitic steels were irradiated in a variety of starting conditions. In particular, these steels spanned a large range (15-40%) of nickel contents, and contained strong variations in Mo, Ti, Al, and Nb. Some alloys were solution-strengthened and some were precipitation-strengthened. Several were cold-worked. These previously unanalyzed data show that at 420{degrees}C all austenitic steels have a creep compliance that is roughly independent of the composition of the steel at 2{+-}1 x 10{sup {minus}6}MPa{sup {minus}1} dpa{sup {minus}1}. The variation within this range may arise from the inability to completely separate the non-creep strains arising from precipitation reactions and the stress-enhancement of swelling. Each of these can be very sensitive to the composition and starting treatment of a steel.

  3. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States)] [and others

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

  4. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    Science.gov (United States)

    Chernov, I. I.; Kalashnikov, A. N.; Kalin, B. A.; Binyukova, S. Yu

    2003-12-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe-C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion irradiation up to a fluence of 5 × 10 20 m -2 at the temperature of 920 K. It was shown that helium-ion irradiation at high temperature resulted in formation of bubbles with a greater size and a smaller density in Fe and ferritic-martensitic steels than those in nickel and austenitic steels. Large gaseous bubbles in ferritic component are uniformly distributed in grains body in Fe-C alloys as well as in ferritic-martensitic steels. The bubbles with a higher density and a smaller size than those in ferritic component are formed in martensitic grains of steels and Fe-C alloys with a high carbon content ( NC>0.01 wt%), which leads to a small level of swelling of martensite in comparison with that of ferrite. In addition, the bubbles in martensitic grains have a tendency to ordered distribution.

  5. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Chernov, I.I. E-mail: chernov@phm.mephi.ru; Kalashnikov, A.N.; Kalin, B.A.; Binyukova, S.Yu

    2003-12-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe-C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion irradiation up to a fluence of 5 x 10{sup 20} m{sup -2} at the temperature of 920 K. It was shown that helium-ion irradiation at high temperature resulted in formation of bubbles with a greater size and a smaller density in Fe and ferritic-martensitic steels than those in nickel and austenitic steels. Large gaseous bubbles in ferritic component are uniformly distributed in grains body in Fe-C alloys as well as in ferritic-martensitic steels. The bubbles with a higher density and a smaller size than those in ferritic component are formed in martensitic grains of steels and Fe-C alloys with a high carbon content (N{sub C}>0.01 wt%), which leads to a small level of swelling of martensite in comparison with that of ferrite. In addition, the bubbles in martensitic grains have a tendency to ordered distribution.

  6. Relationship between localized strain and irradiation assisted stress corrosion cracking in an austenitic alloy

    Energy Technology Data Exchange (ETDEWEB)

    McMurtrey, M.D., E-mail: mdmcm@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Was, G.S. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Patrick, L.; Farkas, D. [Department of Materials Science and Engineering, Virginia Tech, Blacksburg, VA 24061 (United States)

    2011-04-25

    Research highlights: {yields} Austenitic steel is more susceptible to intergranular corrosion after irradiation. {yields} Simulation and experiment used to study cracking in irradiated austentic steel. {yields} Cracking occurs at random high angle boundaries normal to the tensile stress. {yields} Cracking at boundaries with high normal stress and inability to accommodate strain. {yields} Boundary type, angle, and Taylor and Schmid factors affect strain accommodation. - Abstract: Irradiation assisted stress corrosion cracking may be linked to the local slip behavior near grain boundaries that exhibit high susceptibility to cracking. Fe-13Cr-15Ni austenitic steel was irradiated with 2 MeV protons at 360 deg. C to 5 dpa and strained in 288 deg. C simulated BWR conditions. Clusters of grains from the experiment were created in an atomistic simulation and then virtually strained using molecular dynamic simulation techniques. Cracking and grain orientation data were characterized in both the experiment and the simulation. Random high angle boundaries with high surface trace angles with respect to the tensile direction were found to be the most susceptible to cracking. Grain boundary cracking susceptibility was also found to correlate strongly with slip continuity, indicating that the strain accommodation at the boundary is related to cracking resistance. Higher cracking susceptibility was also found at grain boundaries adjacent to grains with low Schmid factor or high Taylor factor. The basic trends reported here are supported by both the experiments and the simulations.

  7. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  8. The independence of irradiation creep in austenitic alloys of displacement rate and helium to dpa ratio

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States); Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    The majority of high fluence data on the void swelling and irradiation creep of austenitic steels were generated at relatively high displacement rates and relatively low helium/dpa levels that are not characteristic of the conditions anticipated in ITER and other anticipated fusion environments. After reanalyzing the available data, this paper shows that irradiation creep is not directly sensitive to either the helium/dpa ratio or the displacement rate, other than through their possible influence on void swelling, since one component of the irradiation creep rate varies with no correlation to the instantaneous swelling rate. Until recently, however, the non-swelling-related creep component was also thought to exhibit its own strong dependence on displacement rate, increasing at lower fluxes. This perception originally arose from the work of Lewthwaite and Mosedale at temperatures in the 270-350{degrees}C range. More recently this perception was thought to extend to higher irradiation temperatures. It now appears, however, that this interpretation is incorrect, and in fact the steady-state value of the non-swelling component of irradiation creep is actually insensitive to displacement rate. The perceived flux dependence appears to arise from a failure to properly interpret the impact of the transient regime of irradiation creep.

  9. Swelling, mechanical properties and structure of austenitic high-nickel alloy irradiated in a fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K.; Neustroev, V.S.; Povstyanko, A.V.; Bulanova, T.M.; Ostrovsky, Z.E. [Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation); Kuznetzov, A.A.; Kursevitch, I.P.; Nikolaev, V.A. [Central Research Inst. of Structural Materials, St. Petersburg (Russian Federation)

    1996-12-31

    Specimens from fuel assembly wrappers and control rod cases of 0.07C-15Cr35Ni-3Mo-B-Zr-Y alloy were investigated after irradiation in the BOR-60 and BN-600 reactors up to a maximum damage dose of about 110 dpa at temperatures between 340 to 550 C. Maximum swelling occurs at 420 to 430 C with 6% at a dose of about 80 dpa and 13.5% at a dose of 108 dpa. The maximum change in strength properties and corresponding decrease of ductility occur at an irradiation temperature of 385 C. With increasing testing temperature the ductility increases from 3--6% up to 22% at 450 C for the material irradiated at 60--80 dpa. Frank dislocation loops with an average diameter of 25 nm voids having diameter between 20 to 23 nm and semicoherent needle-like precipitates with an average length of 21 nm, oriented to several directions, were observed in the alloy structure irradiated to 60 dpa at 430 C. Analysis of the experimental data was performed and the service life time of the investigated alloy was estimated for use as a material for control rod cases of the BOR-60 reactor.

  10. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    NARCIS (Netherlands)

    Chernov, [No Value; Kalashnikov, AN; Kahn, BA; Binyukova, SY

    2003-01-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion it. radiation up to a fluence of 5 x 10(20) m(-2) at the temperature of 920 K. It

  11. The dependence of irradiation creep in austenitic alloys on displacement rate and helium to dpa ratio

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Toloczko, M.B. [Washington State Univ., WA (United States); Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Before the parametric dependencies of irradiation creep can be confidently determined, analysis of creep data requires that the various creep and non-creep strains be separated, as well as separating the transient, steady-state, and swelling-driven components of creep. When such separation is attained, it appears that the steady-state creep compliance, B{sub o}, is not a function of displacement rate, as has been previously assumed. It also appears that the formation and growth of helium bubbles under high helium generation conditions can lead to a significant enhancement of the irradiation creep coefficient. This is a transient influence that disappears as void swelling begins to dominate the total strain, but this transient can increase the apparent creep compliance by 100--200% at relatively low ({le}20) dpa levels.

  12. Shear punch testing of {sup 59}Ni isotopically-doped model austenitic alloys after irradiation in FFTF at different He/dpa ratios

    Energy Technology Data Exchange (ETDEWEB)

    Hankin, G.L.; Faulkner, R.G. [Loughborough Univ., Leicestershire (United Kingdom). I.P.T.M.E.; Hamilton, M.L.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    A series of three model alloys, Fe-15Cr-25Ni, Fe-15Cr-25Ni-0.04P and Fe-15Cr45Ni were irradiated side-by-side in FFTF-MOTA in both the annealed and the cold worked condition in each of two variants, one using naturally occurring isotopic mixtures, and another doped with {sup 59}Ni to generate relatively high helium-to-dpa ratios. Previous papers in this series have addressed the influence of helium on radiation-induced evolution of microstructure, dimensional stability and mechanical properties, the latter using miniature-tensile specimens. In the final paper of this experimental series, three sets of irradiations conducted at different temperatures and displacement rates were examined by shear punch testing of standard microscopy disks. The results were used to determine the influence of helium generation rate, alloy starting condition, irradiation temperature and total neutron exposure. The results were also compared with the miniature tensile data obtained earlier. In general, all alloys approached saturation levels of strength and ductility that were relatively independent of He/dpa ratio and starting condition, but were sensitive to the irradiation temperature and total exposure. Some small influence of helium/dpa ratio on the shear strength is visible in the two series that ran at {approximately}490 C, but is not evident at 365 C.

  13. MODULATED STRUCTURES AND ORDERING STRUCTURES IN ALLOYING AUSTENITIC MANGANESE STEEL

    Institute of Scientific and Technical Information of China (English)

    L. He; Z.H. Jin; J.D. Lu

    2001-01-01

    The microstructure of Fe-10Mn-2Cr-1.5C alloy has been investigated with transmission electron microscopy and X-ray diffractometer. The superlattice diffraction spots and satellite reflection pattrens have been observed in the present alloy, which means the appearence of the ordering structure and modulated structure in the alloy. It is also proved by X-ray diffraction analysis that the austenite in the alloy is more stable than that in traditional austenitic manganese steel. On the basis of this investigation,it is suggested that the C-Mn ordering clusters exist in austenitic manganese steel and the chromium can strengthen this effect by linking the weaker C-Mn couples together,which may play an important role in work hardening of austenitic manganese steel.

  14. On the formation of stacking fault tetrahedra in irradiated austenitic stainless steels – A literature review

    Energy Technology Data Exchange (ETDEWEB)

    Schibli, Raluca, E-mail: raluca.stoenescu@gmail.com; Schäublin, Robin

    2013-11-15

    Irradiated austenitic stainless steels, because of their low stacking fault energy and high shear modulus, should exhibit a high ratio of stacking fault tetrahedra relative to the overall population of radiation induced nanometric defects. Experimental observations of stacking fault tetrahedra by transmission electron microscopy in commercial-purity stainless steels are however scarce, while they abundantly occur in high-purity or model austenitic alloys irradiated at both low and high temperatures, but not at around 673 K. In commercial alloys, the little evidence of stacking fault tetrahedra does not follow such a trend. These contradictions are reviewed and discussed. Reviewing the three possible formation mechanisms identified in the literature, namely the Silcox and Hirsch Frank loop dissociation, the void collapse and the stacking fault tetrahedra growth, it seems that the later dominates under irradiation.

  15. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-04-01

    Progress is reported in eight sections: analysis and evaluation studies, test matrices and test methods development, Path A Alloy Development (austenitic stainless steels), Path C Alloy Development (Ti and V alloys), Path D Alloy Development (Fe alloys), Path E Alloy Development (ferritic steels), irradiation experiments and materials inventory, and materials compatibility and hydrogen permeation studies. (DLC)

  16. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.

    1987-12-01

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.

  17. First-principles study of helium, carbon, and nitrogen in austenite, dilute austenitic iron alloys, and nickel

    Science.gov (United States)

    Hepburn, D. J.; Ferguson, D.; Gardner, S.; Ackland, G. J.

    2013-07-01

    An extensive set of first-principles density functional theory calculations have been performed to study the behavior of He, C, and N solutes in austenite, dilute Fe-Cr-Ni austenitic alloys, and Ni in order to investigate their influence on the microstructural evolution of austenitic steel alloys under irradiation. The results show that austenite behaves much like other face-centered cubic metals and like Ni in particular. Strong similarities were also observed between austenite and ferrite. We find that interstitial He is most stable in the tetrahedral site and migrates with a low barrier energy of between 0.1 and 0.2 eV. It binds strongly into clusters as well as overcoordinated lattice defects and forms highly stable He-vacancy (VmHen) clusters. Interstitial He clusters of sufficient size were shown to be unstable to self-interstitial emission and VHen cluster formation. The binding of additional He and V to existing VmHen clusters increases with cluster size, leading to unbounded growth and He bubble formation. Clusters with n/m around 1.3 were found to be most stable with a dissociation energy of 2.8 eV for He and V release. Substitutional He migrates via the dissociative mechanism in a thermal vacancy population but can migrate via the vacancy mechanism in irradiated environments as a stable V2He complex. Both C and N are most stable octahedrally and exhibit migration energies in the range from 1.3 to 1.6 eV. Interactions between pairs of these solutes are either repulsive or negligible. A vacancy can stably bind up to two C or N atoms with binding energies per solute atom up to 0.4 eV for C and up to 0.6 eV for N. Calculations in Ni, however, show that this may not result in vacancy trapping as VC and VN complexes can migrate cooperatively with barrier energies comparable to the isolated vacancy. This should also lead to enhanced C and N mobility in irradiated materials and may result in solute segregation to defect sinks. Binding to larger vacancy clusters

  18. Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Shack, W. J.

    2008-01-21

    In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1 MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.

  19. Formation of austenite in peritectic Fe-C-X alloy

    Energy Technology Data Exchange (ETDEWEB)

    Kalinushkin, E.P.; Sitalo, J. [State Metall. Acad. of Ukraine, Dnepropetrovsk (Ukraine); Fras, E.; Kapturkiewicz, W.; Burbelko, A.A. [Akademia Gorniczo-Hutnicza, Cracow (Poland)

    2000-07-01

    The mechanism for the formation of peritectic austenite in ferrous alloys was examined. The basic role, played by the mechanisms, is well known in technical literature; like diffusion transport through the solid phase which forms an envelope of austenite (peritectic transformation) and a mechanism of transport through channels of liquid in the envelope of austenite (peritectic reaction). Our calculations show that the peritectic transformation prevails at the initial stage of the grain growth, but afterwards the leading role is taken over by the transport through the channels of liquid. Images of the microstructure support the calculations and transport mechanism. (orig.)

  20. Irradiation creep in austenitic and ferritic steels irradiated in a tailored neutron spectrum to induce fusion reactor levels of helium

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, M.L.; Gibson, L.T. [Oak Ridge National Laboratory, TN (United States); Jitsukawa, S.

    1996-04-01

    Six austenitic stainless steels and two ferritic alloys were irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor where an atomic displacement level of 7.4 dpa was achieved and was then transferred to the High Flux Isotope Reactor for the remainder of the irradiation to a total displacement level of 19 dpa. Temperatures of 60 and 330{degree}C are reported on. At 330{degree}C irradiation creep was found to be linear in stress and fluence with rates in the range of 1.7 - 5.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. Annealed and cold-worked materials exhibited similar creep rates. There is some indication that austenitic alloys with TiC or TiO precipitates had a slightly higher irradiation creep rate than those without. The ferritic alloys HT-9 and Fe-16Cr had irradiatoin creep rates about 0.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. No meaningful data could be obtained from the tubes irradiated at 60{degree}C because of damage to the tubes.

  1. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Chopra, O. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Gruber, Eugene E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shack, William J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2010-06-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC

  2. Effects of titanium additions to austenitic ternary alloys on microstructural evolution and void swelling

    Energy Technology Data Exchange (ETDEWEB)

    Okita, T; Wolfer, W G; Garner, F A; Sekimura, N

    2003-12-01

    Ternary austenitic model alloys were modified with 0.25 wt.% titanium and irradiated in FFTF reactor at dose rates ranging over more than two orders in magnitude. While lowering of dose rate strongly increases swelling by shortening the incubation dose, the steady state swelling rate is not affected by dose rate. Although titanium addition strongly alters the void microstructure, swelling at {approx} 420 C does not change with titanium additions, but the sensitivity to dose rate is preserved.

  3. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

    2015-01-15

    Highlights: • Dislocation loops were the prominent defect, but neutron irradiation caused higher loop density. • Grain boundaries had similar amounts of radiation-induced segregation. • The increment in hardness and yield stress due to irradiation were very similar. • Relative IASCC susceptibility was nearly identical. • The effect of dislocation channel step height on IASCC was similar. - Abstract: The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni–Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed

  4. A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

    Science.gov (United States)

    Chopra, O. K.; Rao, A. S.

    2011-02-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.

  5. Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR - 60 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, Yong [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, Yina [Argonne National Lab. (ANL), Argonne, IL (United States); Allen, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Alexandreanu, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-11-01

    As nuclear power plants age and neutron fluence increases, detrimental effects resulting from radiation damage have become an increasingly important issue for the operational safety and structural integrity of core internal components. In this study, irradiated specimens of reactor core internal components were characterized by transmission electron microscopy. The specimens had been irradiated to 5.5-45 dpa in the BOR-60 reactor at a dose rate close to 10-6 dpa/s and temperature of about 320°C. No voids were observed in the austenitic stainless steels and nickel alloys at all doses. Despite the possibility that fine voids below the TEM resolution limit may be present, it was clear that void swelling was insignificant in all examined alloys up to 45 dpa. Irradiated microstructures of the studied alloys were dominated by a high density of Frank loops. The mean size and density of the Frank loops varied from one material to another, but saturated with increasing dose above ~10 dpa. While no irradiation-induced precipitations were present below 24.5 dpa, fine precipitates were evident in several alloys at 45 dpa.

  6. Application of advanced austenitic alloys to fossil power system components

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.

    1996-06-01

    Most power and recovery boilers operating in the US produce steam at temperatures below 565{degrees}C (1050{degrees}F) and pressures below 24 MPa (3500 psi). For these operating conditions, carbon steels and low alloy steels may be used for the construction of most of the boiler components. Austenitic stainless steels often are used for superheater/reheater tubing when these components are expected to experience temperatures above 565{degrees}C (1050{degrees}F) or when the environment is too corrosive for low alloys steels. The austenitic stainless steels typically used are the 304H, 321H, and 347H grades. New ferritic steels such as T91 and T92 are now being introduced to replace austenitic: stainless steels in aging fossil power plants. Generally, these high-strength ferritic steels are more expensive to fabricate than austenitic stainless steels because the ferritic steels have more stringent heat treating requirements. Now, annealing requirements are being considered for the stabilized grades of austenitic stainless steels when they receive more than 5% cold work, and these requirements would increase significantly the cost of fabrication of boiler components where bending strains often exceed 15%. It has been shown, however, that advanced stainless steels developed at ORNL greatly benefit from cold work, and these steels could provide an alternative to either conventional stainless steels or high-strength ferritic steels. The purpose of the activities reported here is to examine the potential of advanced stainless steels for construction of tubular components in power boilers. The work is being carried out with collaboration of a commercial boiler manufacturer.

  7. Sodium corrosion behavior of austenitic alloys and selective dissolution of chromium and nickel

    Science.gov (United States)

    Suzuki, T.; Mutoh, I.; Yagi, T.; Ikenaga, Y.

    1986-06-01

    The corrosion behavior of six austenitic alloys and reference Type 316 stainless steel (SS) has been examined in a flowing sodium environment at 700°C for up to about 4000 h. The alloys with a range of nickel content between ~ 15 and 43 wt% were designed and manufactured with an expectation of improved swelling resistance during fast neutron irradiation, compared to reference Type 316 SS. The corrosion loss of the alloys at zero downstream position and the concentrations of chromium, nickel and iron in the surface region were determined as a function of corrosion time. The selective dissolution of nickel and chromium played an important role in sodium corrosion of the alloys. During the initial period, accelerated corrosion took place and selective dissolution of chromium and nickel proceeded at a rapid rate. During the subsequent period, the overall corrosion rate and depletion of chromium and nickel decreased with increasing time until the corrosion rate and the surface concentrations of chromium, nickel and iron, which depended on composition of the alloys, reached the steady-state after about 2000 h. Also, the corrosion rate increased with increasing original nickel content of the alloys. Microstructural examination revealed surface attack of the alloys with higher nickel contents, in particular for the two precipitation strengthened Fe-Ni alloys. The alloys showed a trend of increasing carbon and nitrogen contents.

  8. The development of a tensile-shear punch correlation for yield properties of model austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hankin, G.L.; Faulkner, R.G. [Loughborough Univ. (United Kingdom); Hamilton, M.L.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-08-01

    The effective shear yield and maximum strengths of a set of neutron-irradiated, isotopically tailored austentic alloys were evaluated using the shear punch test. The dependence on composition and neutron dose showed the same trends as were observed in the corresponding miniature tensile specimen study conducted earlier. A single tensile-shear punch correlation was developed for the three alloys in which the maximum shear stress or Tresca criterion was successfully applied to predict the slope. The correlation will predict the tensile yield strength of the three different austenitic alloys tested to within {+-}53 MPa. The accuracy of the correlation improves with increasing material strength, to within {+-} MPa for predicting tensile yield strengths in the range of 400-800 MPa.

  9. Influence of the austenitic stainless steel microstructure on the void swelling under ion irradiation

    Directory of Open Access Journals (Sweden)

    Rouxel Baptiste

    2016-01-01

    Full Text Available To understand the role of different metallurgical parameters on the void formation mechanisms, various austenitic stainless steels were elaborated and irradiated with heavy ions. Two alloys, in several metallurgical conditions (15Cr/15Ni–Ti and 15Cr/25Ni–Ti, were irradiated in the JANNUS-Saclay facility at 600 °C with 2 MeV Fe2+ ions up to 150 dpa. Resulting microstructures were observed by Transmission Electron Microscopy (TEM. Different effects on void swelling are highlighted. Only the pre-aged samples, which were consequently solute and especially titanium depleted, show cavities. The nickel-enriched matrix shows more voids with a smaller size. Finally, the presence of nano-precipitates combined with a dense dislocation network decreases strongly the number of cavities.

  10. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (comp.)

    1980-04-01

    Progress is reported concerning preparation of a materials handbook for fusion, creep-fatigue of first-wall structural materials, test results on miniature compact tension fracture toughness specimens, austenitic stainless steels, Fe-Ni-Cr alloys, iron-base alloys with long-range crystal structure, ferritic steels, irradiation experiments, corrosion testing, and hydrogen permeation studies. (FS)

  11. Irradiation creep of various ferritic alloys irradiated {approximately}400 C in the PFR and FFTF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B. [Washington State Univ., WA (United States); Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Co., WA (United States)

    1998-03-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400 C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400 C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 {times} 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  12. Bainitic stabilization of austenite in low alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Brandt, M.L.; Olson, G.B.

    1992-01-01

    Stabilization of retained austenite via bainitic transformation was studied in a triple-phase, ferrite/bainite/austenite steel 0.26C1.52Si-1.2Mn. Volume fraction and stability of retained austenite are varied by isothermal transformation time at 752F following intercritical annealing at 1418F. Austenite stability is measured using the Bolling-Richman technique. Austenite content is measured by and austenite carbon content is estimated from lattice parameters. Strength and ductility measured in both uniaxial and plane-strain tension are correlated with austenite amount and stability. While austenite content peaks at 3 minutes transformation time, stability continues to increase out to 5 minutes associated with a saturation of austenite carbon content and continued refinement of austenite particle size. Despite the reduced austenite content of 8 percent, the higher stability provided by the 5 minutes treatment gives superior mechanical properties.

  13. Bainitic stabilization of austenite in low alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Brandt, M.L.; Olson, G.B.

    1992-12-31

    Stabilization of retained austenite via bainitic transformation was studied in a triple-phase, ferrite/bainite/austenite steel 0.26C1.52Si-1.2Mn. Volume fraction and stability of retained austenite are varied by isothermal transformation time at 752F following intercritical annealing at 1418F. Austenite stability is measured using the Bolling-Richman technique. Austenite content is measured by and austenite carbon content is estimated from lattice parameters. Strength and ductility measured in both uniaxial and plane-strain tension are correlated with austenite amount and stability. While austenite content peaks at 3 minutes transformation time, stability continues to increase out to 5 minutes associated with a saturation of austenite carbon content and continued refinement of austenite particle size. Despite the reduced austenite content of 8 percent, the higher stability provided by the 5 minutes treatment gives superior mechanical properties.

  14. Effect of Plastic Deformation on Magnetic Properties of Fe-40%Ni-2%Mn Austenitic Alloy

    Institute of Scientific and Technical Information of China (English)

    Selva Büyükakkas; H Aktas; S Akturk

    2007-01-01

    The effects of plastic deformation on the magnetic properties of austenite structure in an Fe-40%Ni-2%Mn alloy is investigated by using Mssbauer spectroscopy and Differential Scanning Calorimetry (DSC) techniques The morphology of the alloy has been obtained by using Scanning Electron Microscopy (SEM). The magnetic behaviour of austenite state is ferromagnetic. After plastic deformation, a mixed magnetic structure including both paramagnetic and ferromagnetic states has been obtained at the room temperature. The volume fraction changes, the effective hyperfine fields of the ferromagnetic austenite phase and isomery shift values have also been determined by Mssbauer spectroscopy. The Curie point (TC) and the Neel temperature (TN) have been investigated by means of DSC system for non-deformed and deformed Fe-Ni-Mn alloy. The plastic deformation of the alloy reduces the TN and enhances the paramagnetic character of austenitic Fe-Ni-Mn alloy.

  15. Precipitation hardening in Fe--Ni base austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chang, K.M.

    1979-05-01

    The precipitation of metastable Ni/sub 3/X phases in the austenitic Fe--Ni-base alloys has been investigated by using various combinations of hardening elements, including Ti, Ta, Al, and Nb. The theoretical background on the formation of transition precipitates has been summarized based on: atomic size, compressibility, and electron/atom ratio. A model is proposed from an analysis of static concentration waves ordering the fcc lattice. Ordered structure of metastable precipitates will change from the triangularly ordered ..gamma..', to the rectangularly ordered ..gamma..'', as the atomic ratio (Ti + Al)/(Ta + Nb) decreases. The concurrent precipitation of ..gamma..' and ..gamma..'' occurs at 750/sup 0/C when the ratio is between 1.5 and 1.9. Aging behavior was studied over the temperature range of 500/sup 0/C to 900/sup 0/C. Typical hardness curves show a substantial hardening effect due to precipitation. A combination of strength and fracture toughness can be developed by employing double aging techniques. The growth of these coherent intermediate precipitates follows the power law with the aging time t : t/sup 1/3/ for the spherical ..gamma..' particles; and t/sup 1/2/ for the disc-shaped ..gamma..''. The equilibrium ..beta.. phase is observed to be able to nucleate on the surface of imbedded carbides. The addition of 5 wt % Cr to the age-hardened alloys provides a non-magnetic austenite which is stable against the formation of mechanically induced martensite.Cr addition retards aging kinetics of the precipitation reactions, and suppresses intergranular embrittlement caused by the high temperature solution anneal. The aging kinetics are also found to be influenced by solution annealing treatments.

  16. Influence of Simulated Outside-Reactor Irradiation on Anticorrosion Property of Austenitic Stainless Steel

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The influence of γ-ray irradiation on the properties of inside-reactor stainless steel structures was studied by simulating the working condition of pressurized water reactor (PWR) first circuit and the outside-reactor γ-ray irradiation. The result shows that the simulated outside-reactor irradiation (irradiation dose 4.4 × 104 Gy) has no influence on anticorrosion properties of solutionized SUS304 austenitic stainless steel, including intergranular corrosion (IC) and stress corrosion cracking (SCC). Anticorrosion properties (IC, SCC) of sensitized SUS304 austenitic stainless steel are reduced by simulated outside-reactor irradiation. The longer the sensitizedtime is, the more obvious the influence is.

  17. Static Recrystallization Behavior of Hot Deformed Austenite for Micro-Alloyed Steel

    Institute of Scientific and Technical Information of China (English)

    Jie HUANG; Zhou XU; Xin XING

    2003-01-01

    Static recrystallization behavior of austenite for micro-alloyed steel during hot rolling was studied and the influence (τ-ε diagram) of holding time and deformation at different deformations and isothermal temperatures on microstructuralstate of austen

  18. Development of radiation damage during in-situ Kr++ irradiation of Fesbnd Nisbnd Cr model austenitic steels

    Science.gov (United States)

    Desormeaux, M.; Rouxel, B.; Motta, A. T.; Kirk, M.; Bisor, C.; de Carlan, Y.; Legris, A.

    2016-07-01

    In situ irradiations of 15Cr/15Nisbnd Ti and 15Cr/25Nisbnd Ti model austenitic steels were performed at the Intermediate Voltage Electron Microscope (IVEM)-Tandem user Facility (Argonne National Laboratory) at 600 °C using 1 MeV Kr++. The experiment was designed in the framework of cladding development for the GEN IV Sodium Fast Reactors (SFR). It is an extension of previous high dose irradiations on those model alloys at JANNuS-Saclay facility in France, aimed at investigating swelling mechanisms and microstructure evolution of these alloys under irradiation [1]. These studies showed a strong influence of Ni in decreasing swelling. In situ irradiations were used to continuously follow the microstructure evolution during irradiation using both diffraction contrast imaging and recording of diffraction patterns. Defect analysis, including defect size, density and nature, was performed to characterize the evolving microstructure and the swelling. Comparison of 15Cr/15Nisbnd Ti and 15Cr/25Nisbnd Ti irradiated microstructure has lent insight into the effect of nickel content in the development of radiation damage caused by heavy ion irradiation. The results are quantified and discussed in this paper.

  19. Study on comprehensive properties of duplex austenitic surfacing alloys for impacting abrasion

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    In this paper, comprehensive property crack resistance, work hardening and abrasion resistance of a series of double-phases austenitic alloys(FAW) has been studied by means of SEM, TEM and type MD-10 impacting wear test machine. FAW alloys are of middle chromium and low manganese, including Fe-Cr-Mo-C alloy,Fe-Cr-Mn-C alloy and Fe-Cr-Mn-Ni-C alloy, that are designed for working in condition of impacting abrasion resistance hardfacing.Study results show that the work hardening mechanism of FAW alloys are mainly deformation high dislocation density and dynamic carbide aging, the form of wearing is plastic chisel cutting. Adjusting the amount of carbon, nickel, manganese and other elements in austenitic phase area, the FAW alloy could fit different engineering conditions of high impacting, high temperature and so on.

  20. Compatibility of Austenitic Steel With Molten Lead-Bismuth-Tin Alloy

    Institute of Scientific and Technical Information of China (English)

    ZHANG Rui-qian; LI Yan; WANG Xiao-min

    2011-01-01

    The compatibility of the austenitic AISI 304 steel with Pb-Bi-Sn alloy was analyzed. The AISI 304 steels were immersed in stagnant molten Pb-33.3Bi-33. 3Sn alloy at 400, 500 and 600℃ for different exposure times (100-2 000 h) respectively. XRay diffractio

  1. Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation

    Science.gov (United States)

    Lu, Z.; Faulkner, R. G.; Morgan, T. S.

    2008-12-01

    High Cr ferritic/martensitic steels are leading candidates for structural components of future fusion reactors and new generation fission reactors due to their excellent swelling resistance and thermal properties. A commercial grade 12%CrMoVNb ferritic/martensitic stainless steel in the form of parent plate and off-normal weld materials was fast neutron irradiated up to 33 dpa (1.1 × 10 -6 dpa/s) at 400 °C and 28 dpa (1.7 × 10 -6 dpa/s) at 465 °C, respectively. TEM investigation shows that the fully martensitic weld metal transformed to a duplex austenite/ferrite structure due to high fluence neutron irradiation, the austenite was heavily voided (˜15 vol.%) and the ferrite was relatively void-free; whilst no austenite phases were detected in plate steel. Thermodynamic and phase equilibria software MTDATA has been employed for the first time to investigate neutron irradiation-induced phase transformations. The neutron irradiation effect is introduced by adding additional Gibbs free energy into the system. This additional energy is produced by high energy neutron irradiation and can be estimated from the increased dislocation loop density caused by irradiation. Modelling results show that neutron irradiation reduces the ferrite/austenite transformation temperature, especially for high Ni weld metal. The calculated results exhibit good agreement with experimental observation.

  2. Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Z. [IPTME, Loughborough University, Loughborough LE11 3U (United Kingdom)], E-mail: zheng.lu@lboro.ac.uk; Faulkner, R.G.; Morgan, T.S. [IPTME, Loughborough University, Loughborough LE11 3U (United Kingdom)

    2008-12-01

    High Cr ferritic/martensitic steels are leading candidates for structural components of future fusion reactors and new generation fission reactors due to their excellent swelling resistance and thermal properties. A commercial grade 12%CrMoVNb ferritic/martensitic stainless steel in the form of parent plate and off-normal weld materials was fast neutron irradiated up to 33 dpa (1.1 x 10{sup -6} dpa/s) at 400 deg. C and 28 dpa (1.7 x 10{sup -6} dpa/s) at 465 deg. C, respectively. TEM investigation shows that the fully martensitic weld metal transformed to a duplex austenite/ferrite structure due to high fluence neutron irradiation, the austenite was heavily voided ({approx}15 vol.%) and the ferrite was relatively void-free; whilst no austenite phases were detected in plate steel. Thermodynamic and phase equilibria software MTDATA has been employed for the first time to investigate neutron irradiation-induced phase transformations. The neutron irradiation effect is introduced by adding additional Gibbs free energy into the system. This additional energy is produced by high energy neutron irradiation and can be estimated from the increased dislocation loop density caused by irradiation. Modelling results show that neutron irradiation reduces the ferrite/austenite transformation temperature, especially for high Ni weld metal. The calculated results exhibit good agreement with experimental observation.

  3. Evaluation of Ion Irradiation Behavior of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin Sung; Kim, Min Chul; Hong, Jun Hwa; Han, Chang Hee; Chang, Young Mun; Bae, Chang Soo; Bae, Yoon Young; Chang, Moon Hee

    2006-08-15

    FM steel (Grade 92) and ODS alloy(MA956) specimens were ion irradiated with 122 MeV Ne ions. Irradiation temperatures were about 450 and 550 .deg. C and the peak dose was 1, 5, and 10 dpa. Cross-sectional TEM samples were prepared by the electrolytic Ni-plating after pre-treatment of the irradiated specimens. Irradiation cavities in FM steel and ODS alloy specimens were not much different in size; about 20 nm in diameter in both specimens irradiated at around 450 .deg. C. However, the size distribution of cavities in FM steel specimens was broader than that in ODS alloy specimen, indicating that the cavity growth probably via coalescence). It was noticeable that the location and the preferential growth of the cavities in FM steel specimens: cavities on the PAGB (prior austenite grain boundary) was significantly larger than those within the grains. This could be an important issue for the mechanical properties, especially high temperature creep, fracture toughness, and so on. The dependency of the dose threshold and swelling on the ratio of the inert gas concentration/dpa was analysed for the various irradiation source, including He, Ne, Fe/He, and fast neutron, and the empirical correlation was established.

  4. Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Company, Richland, WA (United States)

    1997-04-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  5. Modeling of Austenite Grain Growth During Austenitization in a Low Alloy Steel

    Science.gov (United States)

    Dong, Dingqian; Chen, Fei; Cui, Zhenshan

    2016-01-01

    The main purpose of this work is to develop a pragmatic model to predict austenite grain growth in a nuclear reactor pressure vessel steel. Austenite grain growth kinetics has been investigated under different heating conditions, involving heating temperature, holding time, as well as heating rate. Based on the experimental results, the mathematical model was established by regression analysis. The model predictions present a good agreement with the experimental data. Meanwhile, grain boundary precipitates and pinning effects on grain growth were studied by transmission electron microscopy. It is found that with the increasing of the temperature, the second-phase particles tend to be dissolved and the pinning effects become smaller, which results in a rapid growth of certain large grains with favorable orientation. The results from this study provide the basis for the establishment of large-sized ingot heating specification for SA508-III steel.

  6. Effect of alloying elements on solidification of primary austenite in Ni-Mn-Cu cast iron

    Directory of Open Access Journals (Sweden)

    A. Janus

    2011-04-01

    Full Text Available Within the research, determined were direction and intensity of alloying elements influence on solidification way (directional orvolumetric of primary austenite dendrites in hypoeutectic austenitic cast iron Ni-Mn-Cu. 50 cast shafts dia. 20 mm were analysed.Chemical composition of the alloy was as follows: 1.7 to 3.3 % C, 1.4 to 3.1 % Si, 2.8 to 9.9 % Ni, 0.4 to 7.7 % Mn, 0 to 4.6 % Cu, 0.14 to0.16 % P and 0.03 to 0.04 % S. The discriminant analysis revealed that carbon influences solidification of primary austenite dendrites most intensively. It clearly increases the tendency to volumetric solidification. Influence of the other elements is much weaker. This means that the solidification way of primary austenite dendrites in hypoeutectic austenitic cast iron Ni-Mn-Cu does not differ from that in an unalloyed cast iron.

  7. The Effect of Post-Bond Heat Treatment on Tensile Property of Diffusion Bonded Austenitic Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sunghoon; Kim, Sung Kwan; Jang, Changheui [KAIST, Daejeon (Korea, Republic of); Sah, Injin [KAERI, Daejeon (Korea, Republic of)

    2015-12-15

    Diffusion bonding is the key manufacturing process for the micro-channel type heat exchangers. In this study, austenitic alloys such as Alloy 800HT, Alloy 690, and Alloy 600, were diffusion bonded at various temperatures and the tensile properties were measured up to 650 ℃. Tensile ductility of diffusion bonded Alloy 800HT was significantly lower than that of base metal at all test temperatures. While, for Alloy 690 and Alloy 600, tensile ductility of diffusion bonded specimens was comparable to that of base metals up to 500 ℃, above which the ductility became lower. The poor ductility of diffusion bonded specimen could have caused by the incomplete grain boundary migration and precipitates along the bond-line. Application of post-bond heat treatment (PBHT) improved the ductility close to that of base metals up to 550 ℃. Changes in tensile properties were discussed in view of the microstructure in the diffusionbonded area.

  8. Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Shack, W. J.; Energy Technology

    2006-01-31

    This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At 3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent intergranular stress corrosion cracking determined at {approx}3 dpa is a good measure of IASCC susceptibility. At {approx}1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At {approx}3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to {approx}3 dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility of Types 304 and 316 steels that contain >0.003 wt.% S increased drastically. This indicates that a sulfur related critical phenomenon plays an important role in IASCC. A sulfur content of <0.002 wt.% is the primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L steel and their high-purity counterparts, a sulfur content of <0.002 wt.% alone is not a sufficient condition to ensure good resistance to IASCC. This is in distinct contrast to

  9. Swelling analysis of austenitic stainless steels by means of ion irradiation and kinetic modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kohyama, Akira [Kyoto Univ., Institute of Advanced Energy, Uji, Kyoto (Japan); Donomae, Takako [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-03-01

    The influences of irradiation environment on the swelling behavior of austenitic stainless steel has been studied, to aid understanding the origin of the difference in swelling response of PNC316 stainless steel in fuel-pin environment and in materials irradiation capsules, in terms of irradiation conditions, damage mechanism and material conditions. This work focused on the theoretical investigation of the influence of temperature variation on microstructural development of austenitic stainless steels during irradiation, using a kinetic rate theory model. A modeling and calculation on non-steady irradiation effects were first carried out. A fully dynamic model of point defect evolution and extended defect development, which accounts for cascade damage, was developed and successfully applied to simulate the interstitial loop evolution in low temperature regimes. The influence of cascade interstitial clustering on dislocation loop formation has also been assessed. The establishment of a basis for general assessment of non-steady irradiation effects in austenitic stainless steels was advanced. The developed model was applied to evaluate the influences of temperature variation in formerly carried out CMIR and FFTF/MFA-1 FBR irradiation experiments. The results suggested the gradual approach of microstructural features to equilibrium states in all the temperature variation conditions and no sign of anomalous behavior was noted. On the other hand, there is the influence of temperature variation on microstructural development under the neutron irradiation, like CMIR. So there are some possibilities of the work of mechanism which is not taken care on this model, for example the effect of the precipitate behavior which is sensitive to irradiation temperature. (author)

  10. Migration and accumulation at dislocations of transmutation helium in austenitic steels upon neutron irradiation

    Science.gov (United States)

    Kozlov, A. V.; Portnykh, I. A.

    2016-04-01

    The model of the migration and accumulation at dislocations of transmutation helium and the formation of helium-vacancy pore nuclei in austenitic steels upon neutron irradiation has been proposed. As illustrations of its application, the dependences of the characteristics of pore nuclei on the temperature of neutron irradiation have been calculated. The results of the calculations have been compared with the experimental data in the literature on measuring the characteristics of radiation-induced porosity that arises upon the irradiation of shells of fuel elements of a 16Cr-19Ni-2Mo-2Mn-Si-Ti-Nb-V-B steel in a fast BN600 neutron reactor at different temperatures.

  11. Thermal stability of the cellular structure of an austenitic alloy after selective laser melting

    Science.gov (United States)

    Bazaleeva, K. O.; Tsvetkova, E. V.; Balakirev, E. V.; Yadroitsev, I. A.; Smurov, I. Yu.

    2016-05-01

    The thermal stability of the cellular structure of an austenitic Fe-17% Cr-12% Ni-2% Mo-1% Mn-0.7% Si-0.02% C alloy produced by selective laser melting in the temperature range 20-1200°C is investigated. Metallographic analysis, transmission electron microscopy, and scanning electron microscopy show that structural changes in the alloy begin at 600-700°C and are fully completed at ~1150°C. Differential scanning calorimetry of the alloy with a cellular structure reveals three exothermic processes occurring upon annealing within the temperature ranges 450-650, 800-1000, and 1050-1200°C.

  12. Development of Austenitic ODS Strengthened Alloys for Very High Temperature Applications

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Robertson, Ian [Kyushu Univ. (Japan); Sehitoglu, Huseyin [Univ. of Illinois, Urbana-Champaign, IL (United States); Sofronis, Petros [Kyushu Univ. (Japan); Gewirth, Andrew [Kyushu Univ. (Japan)

    2015-04-22

    This “Blue Sky” project was directed at exploring the opportunities that would be gained by developing Oxide Dispersion Strengthened (ODS) alloys based on the Fe-Cr-Ni austenitic alloy system. A great deal of research effort has been directed toward ferritic and ferritic/martensitic ODS alloys which has resulted in reasonable advances in alloy properties. Similar gains should be possible with austenitic alloy which would also take advantage of other superior properties of that alloy system. The research effort was aimed at the developing an in-depth understanding of the microstructural-level strengthening effects of ODS particles in austentic alloys. This was accomplished on a variety of alloy compositions with the main focus on 304SS and 316SS compositions. A further goal was to develop an understanding other the role of ODS particles on crack propagation and creep performance. Since these later two properties require bulk alloy material which was not available, this work was carried out on promising austentic alloy systems which could later be enhanced with ODS strengthening. The research relied on a large variety of micro-analytical techniques, many of which were available through various scientific user facilities. Access to these facilities throughout the course of this work was instrumental in gathering complimentary data from various analysis techniques to form a well-rounded picture of the processes which control austenitic ODS alloy performance. Micromechanical testing of the austenitic ODS alloys confirmed their highly superior mechanical properties at elevated temperature from the enhanced strengthening effects. The study analyzed the microstructural mechanisms that provide this enhanced high temperature performance. The findings confirm that the smallest size ODS particles provide the most potent strengthening component. Larger particles and other thermally- driven precipitate structures were less effective contributors and, in some cases, limited

  13. Formation of austenite and dissolution of carbides in Fe-8.2Cr-C alloys

    Energy Technology Data Exchange (ETDEWEB)

    Shtansky, D.; Nakai, K.; Ohmori, Y. [Ehime Univ., Matsuyama (Japan). Faculty of Engineering

    1999-01-01

    The mechanism of austenite formation and the kinetics of carbide dissolution have been studied in Fe-8.2Cr-0.2C and Fe-8.2Cr-0.96 C (numbers indicate mass%) alloys with ferrite lamellar carbide and ferrite spheroidized carbide initial structures. The morphology of austenite formation in the range of 850 to 900 C has been examined in detail by transmission electron microscopy. The mechanisms of austenite nucleation and growth have been distinguished as they depend on the composition, starting microstructure and austenitizing temperature. The effects of both austenitising temperature in the range of 850 to 1150 C and a time on the M{sub 23}C{sub 6} and M{sub 7}C{sub 3} carbide evolution have also been investigated. Different morphologies of transformation products have been observed. The orientation relationships between ferrite, austenite and carbides were determined. The observed results can be explained by assuming local equilibrium at the moving interfaces during the reactions. (orig.) 47 refs.

  14. Evaluation of High Temperature Corrosion Resistance of Finned Tubes Made of Austenitic Steel And Nickel Alloys

    Directory of Open Access Journals (Sweden)

    Turowska A.

    2016-06-01

    Full Text Available The purpose of the paper was to evaluate the resistance to high temperature corrosion of laser welded joints of finned tubes made of austenitic steel (304,304H and nickel alloys (Inconel 600, Inconel 625. The scope of the paper covered the performance of corrosion resistance tests in the atmosphere of simulated exhaust gases of the following chemical composition: 0.2% HCl, 0.08% SO2, 9.0% O2 and N2 in the temperature of 800°C for 1000 hours. One found out that both tubes made of austenitic steel and those made of nickel alloy displayed good resistance to corrosion and could be applied in the energy industry.

  15. The irradiation effects on zirconium alloys

    Science.gov (United States)

    Negut, Gh.; Ancuta, M.; Radu, V.; Ionescu, S.; Stefan, V.; Uta, O.; Prisecaru, I.; Danila, N.

    2007-05-01

    Pressure tube samples were irradiated under helium atmosphere in the TRIGA Steady State Research and Material Test Reactor of the Romanian Institute for Nuclear Research (INR). These samples are made of the Zr-2.5%Nb alloy used as structural material for the CANDU Romanian power reactors. After irradiation, mechanical tests were performed in the Post Irradiation Examination Laboratory (PIEL) to study the influence of irradiation on zirconium alloys mechanical behaviour. The tensile test results were used for structural integrity assessment. Results of the tests are presented. The paper presents, also, pressure tube structural integrity assessment.

  16. Magnetic analysis of martensitic and austenitic phases in metamagnetic NiMn(In, Sn) alloys

    Energy Technology Data Exchange (ETDEWEB)

    Lázpita, P., E-mail: patricia.lazpita@ehu.es [University of Basque Country (UPV/EHU), Leioa (Spain); Escolar, J. [University of Basque Country (UPV/EHU), Leioa (Spain); Chernenko, V.A. [University of Basque Country (UPV/EHU), Leioa (Spain); BCMaterials, Parque Tecnológico de Bizkaia, Ed. 500, Derio 48160 (Spain); Ikerbasque, Basque Foundation for Science, Bilbao 48013 (Spain); Barandiarán, J.M. [University of Basque Country (UPV/EHU), Leioa (Spain); BCMaterials, Parque Tecnológico de Bizkaia, Ed. 500, Derio 48160 (Spain)

    2015-09-25

    Highlights: • NiMnIn austenite and martensite have similar Ising-type critical exponents. • NiMnIn critical exponents rule out disordered states as spin-glass in martensite. • In NiMnIn alloys, magnetism arises mainly from moments localized at Mn atoms. • NiCoMnSn critical exponents are close to the ones from tricritical mean field model. • NiCoMnSn complex magnetic state results from three different magnetic atoms. - Abstract: Two different metamagnetic shape memory alloys of nominal composition Ni{sub 50}Mn{sub 36}In{sub 14} and Ni{sub 42}Co{sub 8}Mn{sub 39}Sn{sub 11} have been studied by means of modified Arrott plots to give insight into the magnetic states of both the austenitic and martensitic phases. For Ni{sub 50}Mn{sub 36}In{sub 14} alloy, the same critical exponents (β = 0.32 and γ = 2.0) are obtained in austenite and martensite. They suggest that localized moments at Mn atoms are responsible for the magnetism of both phases according to the Ising model. The martensite, however, displays a rather complex behavior because β continuously changes with temperature. In Ni{sub 43}Co{sub 6.5}Mn{sub 39}Sn{sub 11.5}, critical exponents in the austenite are β = 0.27 and γ = 1.0. They are close to the tricritical mean field model, but no reliable fits were obtained in the martensite. The results are discussed in terms of microscopically different magnetic states in two alloys reflecting a complex interplay between the ferromagnetic and antiferromagnetic contributions.

  17. Modeling precipitation thermodynamics and kinetics in type 316 austenitic stainless steels with varying composition as an initial step toward predicting phase stability during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Jae-Hyeok, E-mail: jhshim@kist.re.kr [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); High Temperature Energy Materials Research Center, Korea Institute of Science and Technology, Seoul 136-791 (Korea, Republic of); Povoden-Karadeniz, Erwin [Christian Doppler Laboratory for Early Stages of Precipitation, Vienna University of Technology, A-1040 Vienna (Austria); Kozeschnik, Ernst [Institute of Materials Science and Technology, Vienna University of Technology, A-1040 Vienna (Austria); Wirth, Brian D. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States)

    2015-07-15

    Highlights: • We model the precipitation kinetics in irradiated 316 austenitic stainless steels. • Radiation-induced phases are predicted to form at over 10 dpa segregation conditions. • The Si content is the most critical for the formation of radiation-induced phases. - Abstract: The long-term evolution of precipitates in type 316 austenitic stainless steels at 400 °C has been simulated using a numerical model based on classical nucleation theory and the thermodynamic extremum principle. Particular attention has been paid to the precipitation of radiation-induced phases such as γ′ and G phases. In addition to the original compositions, the compositions for radiation-induced segregation at a dose level of 5, 10 or 20 dpa have been used in the simulation. In a 316 austenitic stainless steel, γ′ appears as the main precipitate with a small amount of G phase forming at 10 and 20 dpa. On the other hand, G phase becomes relatively dominant over γ′ at the same dose levels in a Ti-stabilized 316 austenitic stainless steel, which tends to suppress the formation of γ′. Among the segregated alloying elements, the concentration of Si seems to be the most critical for the formation of radiation-induced phases. An increase in dislocation density as well as increased diffusivity of Mn and Si significantly enhances the precipitation kinetics of the radiation-induced phases within this model.

  18. Irradiation creep of dispersion strengthened copper alloy

    Energy Technology Data Exchange (ETDEWEB)

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

    1997-04-01

    Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

  19. Nickel-based alloy/austenitic stainless steel dissimilar weld properties prediction on asymmetric distribution of laser energy

    Science.gov (United States)

    Zhou, Siyu; Ma, Guangyi; Chai, Dongsheng; Niu, Fangyong; Dong, Jinfei; Wu, Dongjiang; Zou, Helin

    2016-07-01

    A properties prediction method of Nickel-based alloy (C-276)/austenitic stainless steel (304) dissimilar weld was proposed and validated based on the asymmetric distribution of laser energy. Via the dilution level DC-276 (the ratio of the melted C-276 alloy), the relations between the weld properties and the energy offset ratio EC-276 (the ratio of the irradiated energy on the C-276 alloy) were built, and the effects of EC-276 on the microstructure, mechanical properties and corrosion resistance of dissimilar welds were analyzed. The element distribution Cweld and EC-276 accorded with the lever rule due to the strong convention of the molten pool. Based on the lever rule, it could be predicted that the microstructure mostly consists of γ phase in each weld, the δ-ferrite phase formation was inhibited and the intermetallic phase (P, μ) formation was promoted with the increase of EC-276. The ultimate tensile strength σb of the weld joint could be predicted by the monotonically increasing cubic polynomial model stemming from the strengthening of elements Mo and W. The corrosion potential U, corrosion current density I in the active region and EC-276 also met the cubic polynomial equations, and the corrosion resistance of the dissimilar weld was enhanced with the increasing EC-276, mainly because the element Mo could help form a steady passive film which will resist the Cl- ingress.

  20. High Nitrogen Austenitic Stainless Steels Manufactured by Nitrogen Gas Alloying and Adding Nitrided Ferroalloys

    Institute of Scientific and Technical Information of China (English)

    LI Hua-bing; JIANG Zhou-hua; SHEN Ming-hui; YOU Xiang-mi

    2007-01-01

    A simple and feasible method for the production of high nitrogen austenitic stainless steels involves nitrogen gas alloying and adding nitrided ferroalloys under normal atmospheric conditions. Alloying by nitrogen gas bubbling in Fe-Cr-Mn-Mo series alloys was carried out in MoSi2 resistance furnace and air induction furnace under normal atmospheric conditions. The results showed that nitrogen alloying could be accelerated by increasing nitrogen gas flow rate, prolonging residence time of bubbles, increasing gas/molten steel interfaces, and decreasing the sulphur and oxygen contents in molten steel. Nitrogen content of 0.69% in 18Cr18Mn was obtained using air induction furnace by bubbling of nitrogen gas from porous plug. In addition, the nickel-free, high nitrogen austenitic stainless steels with sound and compact macrostructure had been produced in the laboratory using vacuum induction furnace and electroslag remelting furnace under nitrogen atmosphere by the addition of nitrided alloy with the maximum nitrogen content of 0.81 %. Pores were observed in the ingots obtained by melting and casting in vacuum induction furnace with the addition of nitrided ferroalloys and under nitrogen atmosphere. After electroslag remelting of the cast ingots, they were all sound and were free of pores. The yield of nitrogen increased with the decrease of melting rate in the ESR process. Due to electroslag remelting under nitrogen atmosphere and the consequential addition of aluminum as deoxidizer to the slag, the loss of manganese decreased obviously. There existed mainly irregular Al2O3 inclusions and MnS inclusions in ESR ingots, and the size of most of the inclusions was less than 5 μm. After homogenization of the hot rolled plate at 1 150 ℃× 1 h followed by water quenching, the microstructure consisted of homogeneous austenite.

  1. Effect of Multi-Step Tempering on Retained Austenite and Mechanical Properties of Low Alloy Steel

    Institute of Scientific and Technical Information of China (English)

    Hamid Reza Bakhsheshi-Rad; Ahmad Monshi; Hossain Monajatizadeh; Mohd Hasbullah Idris; Mohammed Rafiq Abdul Kadir; Hassan Jafari

    2011-01-01

    The effect of multi-step tempering on retained austenite content and mechanical properties of low alloy steel used in the forged cold back-up roll was investigated.Microstructural evolutions were characterized by optical microscope,X-ray diffraction,scanning electron microscope and Feritscope,while the mechanical properties were determined by hardness and tensile tests.The results revealed that the content of retained austenite decreased by about 2% after multi-step tempering.However,the content of retained austenite increased from 3.6% to 5.1% by increasing multi-step tempering temperature.The hardness and tensile strength increased as the austenitization temperature changed from 800 to 920 ℃,while above 920 ℃,hardness and tensile strength decreased.In addition,the maximum values of hardness,ultimate and yield strength were obtained via triple tempering at 520 ℃,while beyond 520 ℃,the hardness,ultimate and yield strength decreased sharply.

  2. Effects of Nitrogen Content and Austenitization Temperature on Precipitation in Niobium Micro-alloyed Steels

    Institute of Scientific and Technical Information of China (English)

    Lei CAO; Zhong-min YANG; Ying CHEN; Hui-min WANG; Xiao-li ZHAO

    2015-01-01

    The influences of nitrogen content and austenitization temperature on Nb(C,N)precipitation in niobium micro-alloyed steels were studied by different methods:optical microscopy,tensile tests,scanning electron mi-croscopy,transmission electron microscopy,physicochemical phase analysis,and small-angle X-ray scattering. The results show that the strength of the steel with high nitrogen content is slightly higher than that of the steel with low nitrogen content.The increase in the nitrogen content does not result in the increase in the amount of Nb(C,N) precipitates,which mainly depends on the niobium content in the steel.The mass fraction of small-sized Nb(C,N) precipitates (1-10 nm)in the steel with high nitrogen content is less than that in the steel with low nitrogen con-tent.After austenitized at 1 150 ℃,a number of large cuboidal and needle-shaped particles are detected in the steel with high nitrogen content,whereas they dissolve after austenitized at 1 200 ℃ and the Nb(C,N)precipitates become finer in both steels.Furthermore,the results also show that part of the nitrogen in steel involves the formation of al-loyed cementite.

  3. Correlation between locally deformed structure and oxide film properties in austenitic stainless steel irradiated with neutrons

    Science.gov (United States)

    Chimi, Yasuhiro; Kitsunai, Yuji; Kasahara, Shigeki; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-07-01

    To elucidate the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water for neutron-irradiated austenitic stainless steels (SSs), the locally deformed structures, the oxide films formed on the deformed areas, and their correlation were investigated. Tensile specimens made of irradiated 316L SSs were strained 0.1%-2% at room temperature or at 563 K, and the surface structures and crystal misorientation among grains were evaluated. The strained specimens were immersed in high-temperature water, and the microstructures of the oxide films on the locally deformed areas were observed. The appearance of visible step structures on the specimens' surface depended on the neutron dose and the applied strain. The surface oxides were observed to be prone to increase in thickness around grain boundaries (GBs) with increasing neutron dose and increasing local strain at the GBs. No penetrative oxidation was observed along GBs or along surface steps.

  4. Effect of austenitizing conditions on the impact properties of an alloyed austempered ductile iron of initially ferritic matrix structure

    Energy Technology Data Exchange (ETDEWEB)

    Delia, M.; Alaalam, M.; Grech, M. [Univ. of Malta (Malta). Dept. of Metallurgy and Materials

    1998-04-01

    The effect of austenitizing conditions on the microstructure and impact properties of an austempered ductile iron (ADI) containing 1.6% Cu and 1.6% Ni as the main alloying elements was investigated. Impact tests were carried out on samples of initially ferritic matrix structure and which had been first austenitized at 850, 900, 950, and 1,000 C for 15 to 360 min and austempered at 360 C for 180 min. Results showed that the austenitizing temperature, T{sub {gamma}}, and time, t{sub {gamma}} have a significant effect on the impact properties of the alloy. This has been attributed to the influence of these variables on the carbon kinetics. Microstructures of samples austenitized at 950 and 1,000 C contain no pro-eutectoid ferrite. The impact properties of the former structures are independent of t{sub {gamma}}, while those solution treated at 1,000 C are generally low and show wide variation over the range of soaking time investigated. For fully ausferritic structures, impact properties fall with an increase in T{sub {gamma}}. This is particularly evident at 1,000 C. As the T{sub {gamma}} increases, the amount of carbon dissolved in the original austenite increases. This slows down the rate of austenite transformation and results in coarser structures with lower mechanical properties. Optimum impact properties are obtained following austenitizing between 900 and 950 C for 120 to 180 min.

  5. The evolution of mechanical property change in irradiated austenitic stainless steels

    Science.gov (United States)

    Lucas, G. E.

    1993-11-01

    The evolution of mechanical properties in austenitic stainless steels during irradiation is reviewed. Changes in strength, ductility and fracture toughness are strongly related to the evolution of the damage microstructure and microstructurally-based models for strengthening reasonably correlate the data. Irradiation-induced defects promote work softening and flow localization which in turn leads to significant reductions in ductility and fracture toughness beyond about 10 dpa. The effects of irradiation on fatigue appear to be modest except at high temperature where helium embrittlement becomes important. The swelling-independent component of irradiation creep strain increases linearly with dose and is relatively insensitive to material variables and irradiation temperature, except at low temperatures where accelerated creep may occur as a result of low vacancy mobility. Creep rupture life is a strong function of helium content, but is less sensitive to metallurgical conditions. Irradiation-induced stress corrosion cracking appears to be related to the evolution of radiation-induced segregation/depletion at grain boundaries, and hence may not be significant at low irradiation temperatures.

  6. Tensile properties of a titanium modified austenitic stainless steel and the weld joints after neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, K.; Ioka, I.; Jitsukawa, S.; Hamada, A.; Hishinuma, A. [and others

    1996-10-01

    Tensile specimens of a titanium modified austenitic stainless steel and its weldments fabricated with Tungsten Inert Gas (TIG) and Electron Beam (EB) welding techniques were irradiated to a peak dose of 19 dpa and a peak helium level of 250 appm in the temperature range between 200 and 400{degrees}C in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) and the High Flux Isotope Reactor (HFIR). The He/dpa ratio of about 13 appm/dpa is similar to the typical helium/dpa ratio of a fusion reactor environment. The tensile tests were carried out at the irradiation temperature in vacuum. The irradiation caused an increase in yield stress to levels between 670 and 800 MPa depending on the irradiation temperature. Total elongation was reduced to less than 10%, however the specimens failed in a ductile manner. The results were compared with those of the specimens irradiated using irradiation capsules producing larger amount of He. Although the He/dpa ratio affected the microstructural change, the impact on the post irradiation tensile behavior was rather small for not only base metal specimens but also for the weld joint and the weld metal specimens.

  7. Irradiation creep of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Billone, M.C.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1998-03-01

    A study of irradiation creep in vanadium-base alloys is underway with experiments in the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) in the United States. Test specimens are thin-wall sealed tubes with internal pressure loading. The results from the initial ATR irradiation at low temperature (200--300 C) to a neutron damage level of 4.7 dpa show creep rates ranging from {approx}0 to 1.2 {times} 10{sup {minus}5}/dpa/MPa for a 500-kg heat of V-4Cr-4Ti alloy. These rates were generally lower than reported from a previous experiment in BR-10. Because both the attained neutron damage levels and the creep strains were low in the present study, however, these creep rates should be regarded as only preliminary. Substantially more testing is required before a data base on irradiation creep of vanadium alloys can be developed and used with confidence.

  8. High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.B.; Solly, B.

    1966-12-15

    The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties.

  9. Impact of the nanostructuration on the corrosion resistance and hardness of irradiated 316 austenitic stainless steels

    Science.gov (United States)

    Hug, E.; Prasath Babu, R.; Monnet, I.; Etienne, A.; Moisy, F.; Pralong, V.; Enikeev, N.; Abramova, M.; Sauvage, X.; Radiguet, B.

    2017-01-01

    The influence of grain size and irradiation defects on the mechanical behavior and the corrosion resistance of a 316 stainless steel have been investigated. Nanostructured samples were obtained by severe plastic deformation using high pressure torsion. Both coarse grain and nanostructured samples were irradiated with 10 MeV 56Fe5+ ions. Microstructures were characterized using transmission electron microscopy and atom probe tomography. Surface mechanical properties were evaluated thanks to hardness measurements and the corrosion resistance was studied in chloride environment. Nanostructuration by high pressure torsion followed by annealing leads to enrichment in chromium at grain boundaries. However, irradiation of nanostructured samples implies a chromium depletion of the same order than depicted in coarse grain specimens but without metallurgical damage like segregated dislocation loops or clusters. Potentiodynamic polarization tests highlight a definitive deterioration of the corrosion resistance of coarse grain steel with irradiation. Downsizing the grain to a few hundred of nanometers enhances the corrosion resistance of irradiated samples, despite the fact that the hardness of nanocrystalline austenitic steel is only weakly affected by irradiation. These new experimental results are discussed in the basis of couplings between mechanical and electrical properties of the passivated layer thanks to impedance spectroscopy measurements, hardness properties of the surfaces and local microstructure evolutions.

  10. Damage structure of austenitic stainless steel 316LN irradiated at low temperature in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, N.; Robertson, J.P.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States); Wakai, E. [Japan Atomic Energy Research Inst. (Japan)

    1998-03-01

    TEM disk specimens of austenitic stainless steel 316LN irradiated to damage levels of about 3 dpa at irradiation temperatures of either about 90 C or 250 C have been investigated by using transmission electron microscopy. The irradiation at 90 C and 250 C induced a dislocation loop density of 3.5 {times} 10{sup 22} m{sup {minus}3} and 6.5 {times} 10{sup 22} m{sup {minus}3}, a black dot density of 2.2 {times} 10{sup 23} m{sup {minus}3} and 1.6 {times} 10{sup 23} m{sup {minus}3}, respectively, in the steels, and a high density (<1 {times} 10{sup 22} m{sup {minus}3}) of precipitates in matrix. Cavities could be observed in the specimens after the irradiation. It is suggested that the dislocation loops, the black dots, and the precipitates cause irradiation hardening, an increase in the yield strength and a decrease in the uniform elongation, in the 316LN steel irradiated at low temperature.

  11. Effect of austenitizing conditions on the impact properties of an alloyed austempered ductile iron of initially ferritic matrix structure

    Science.gov (United States)

    Delia, M.; Alaalam, M.; Grech, M.

    1998-04-01

    The effect of austenitizing conditions on the microstructure and impact properties of an austempered ductile iron (ADI) containing 1.6% Cu and 1.6% Ni as the main alloying elements was investigated. Impact tests were carried out on samples of initially ferritic matrix structure and which had been first austenitized at 850,900, 950, and 1000°C for 15 to 360 min and austempered at 360°C for 180 min. Results showed that the austenitizing temperature, Tγ, and time, tγ, have a significant effect on the impact properties of the alloy. This has been attributed to the influence of these variables on the carbon kinetics. The impact energy is generally high after short tγ, and it falls with further soaking. In samples austenitized at 850 and 900°C, these trends correspond to the gradual disappearance of the pro-eutectoid ferrite and the attainment of fully developed ausferritic structures. In initially ferritic structures, the carbon diffusion distances involved during austenitization are large compared to those in pearlitic structures. This explains the relatively long soaking periods required to attain fully ausferritic structures, which in spite of the lower impact energy values, have a better combination of mechanical properties. Microstructures of samples austenitized at 950 and 1000°C contain no pro-eutectoid ferrite. The impact properties of the former structures are independent of tγ, while those solution treated at 1000°C are generally low and show wide variation over the range of soaking time investigated. For fully ausferritic structures, impact properties fall with an increase in Tγ. This is particularly evident at 1000°C. As the Tγ increases, the amount of carbon dissolved in the original austenite increases. This slows down the rate of austenite transformation and results in coarser structures with lower mechanical properties. Optimum impact properties are obtained following austenitizing between 900 and 950°C for 120 to 180 min.

  12. Hydrogen-plasticity in the austenitic alloys; Interactions hydrogene-plasticite dans les alliages austenitiques

    Energy Technology Data Exchange (ETDEWEB)

    De lafosse, D. [Ecole Nationale Superieure des Mines, Lab. PECM-UMR CNRS 5146, 42 - Saint-Etienne (France)

    2007-07-01

    This presentation deals with the hydrogen effects under stresses corrosion, in austenitic alloys. The objective is to validate and characterize experimentally the potential and the limits of an approach based on an elastic theory of crystal defects. The first part is devoted to the macroscopic characterization of dynamic hydrogen-dislocations interactions by aging tests. then the hydrogen influence on the plasticity is evaluated, using analytical classic models of the elastic theory of dislocations. The hydrogen influence on the flow stress of bcc materials is analyzed experimentally with model materials. (A.L.B.)

  13. Mechanical characteristics and swelling of austenitic Fe-Cr-Mn steels irradiated in the SM-2 and BOR-60 reactors

    Science.gov (United States)

    Shamardin, V. K.; Bulanova, T. M.; Neustroev, V. S.; Ivanov, L. I.; Djomina, E. V.; Platov, Yu. M.

    1991-03-01

    Three types of austenitic Fe-Cr-Mn stainless steels were irradiated simultaneously with Fe-Cr-Ni austenitic steel at temperatures from 400 to 800°C in the mixed spectrum of the high flux SM-2 reactor to 10 dpa and 700 appm of He and in the BOR-60 reactor to 60 dpa without He generation. The paper presents the swelling and mechanical properties of steels irradiated in the BOR-60 and SM-2 as a function of the concentration of transmuted He and the value of atomic displacement.

  14. Improvement of steam oxidation resistance of martensitic and austenitic alloys by Al-containing coatings

    Energy Technology Data Exchange (ETDEWEB)

    Knoedler, Reinhard; Straub, Stefan [Alstom Power Systems GmbH, Mannheim (Germany)

    2010-07-01

    An increase of steam power plant efficiency is necessary to reduce the emissions and to reduce fuel consumption. To obtain this goal, the steam temperature must be increased considerably. Present alloys, however, show oxide scale growth and spallation at elevated temperatures. These shortcomings can be avoided by applying coatings to martensitic and austenitic steels. Therefore, diffusion coatings on martensitic 9 - 11 % - Cr steels and 79 % - Cr austenitic steels were applied and exposed to flowing steam for operating times up to 15.000 h at 650 C. The coating process was optimized with respect to surface preparation, heat treatment and other process parameters. Metallographic analysis was performed after the oxidation tests by light optical (OM) and scanning electron microscopy (SEM). With energy dispersive X-ray analysis (EDX) in SEM the distribution of the elements was determined in order to assess the diffusion velocity of different coating constituents. This allows an estimation of the coating lifetime. The best coating showed that only a few {mu}m of oxide scales have formed as compared to several 100 {mu}m on the uncoated steel (under the same test conditions). Thus, these types of coatings can be a promising solution for preventing oxidation of martensitic and austenitic steels. (orig.)

  15. Microstructural evolution in austenitic stainless steel irradiated with triple-beam

    Energy Technology Data Exchange (ETDEWEB)

    Hamada, Shozo; Miwa, Yukio; Yamaki, Daiju [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Zhang Yichuan

    1997-03-01

    An austenitic stainless steel was simultaneously irradiated with nickel, helium and hydrogen ions at the temperature range of 573-673 K. The damage level and injected concentration of He and H ions in the triple-beam irradiated region are 57 dpa, 19000 and 18000 at.ppm, respectively. Following to irradiation, the cross sectional observation normal to the incident surface of the specimen was carried out with a transmission electron microscope. Two bands parallel to the incident surface were observed in the irradiated specimen, which consist of dislocation loops and lines of high number density. These locate in the range of the depth of 0.4 to 1.3 {mu}m and 1.8 to 2.4 {mu}m from the incident surface, respectively. The region between two bands, which corresponds to the triple beam irradiated region, shows very low number density of dislocations than that in each band. Observation with higher magnification of this region shows that fine cavities with high number density uniformly distribute in the matrix. (author)

  16. Metallurgical Source of Cryogenic Intergranular Fracture of Fe-38Mn Austenitic Alloy

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    SEM and Field emitting TEM-EDAX were used to investigate the fracture surface of series impact specimens and the grain boundary chemistries of VIM (vacuum-induction-melted) Fe-38Mn austenitic alloy before and after ESR (electroslag remelting,). The quantity and the size of inclusions were also examined. The results show that the VIM Fe-38Mn aust enitinic alloy water-quenched from 1 100 ℃ undergoes an obvious ductile-to-brittle transition, and the impact work at ambient temperature is 242 J, the corresponding fracture surface exhibits adimple character. However, the impact work at 77 K of VIM alloy is only 25 J and the fracture mode is IGF (intergranular f racture). After ESR, the impact work at ambient temperature is 320 J and the fra cture surface exhibits a character of "volcano lava" (meaning excellent toughn ess); The impact work at 77 K is up to 300 J and the fracture mode is microvoid coalescence mixed with quasi-cleavage. The segregation of Mn is not found in all specimens, but the segregation of S is observed, and the S segregation is decreased after ESR. The examined results of inclusions show that ESR reduces the quantity and improves the morphology of inclusions. From the above results it can be seen that the cryogenic IGF of VIM Fe-38Mn austenitic alloy is related to the S segregation at grain boundary. After ESR the decrease in the quantity and size of inclusion results in the increase of the impact work at ambient temperature, while the restriction of IGF is related to the decrease in the total level, and hence in the grain boundary segregation of S.

  17. Effect of Grain Size on Void Formation during High-Energy Electron Irradiation of Austenitic Stainless Steel

    DEFF Research Database (Denmark)

    Singh, Bachu Narain

    1974-01-01

    Thin foils of an ‘ experimental ’ austenitic stainless steel, with and without dispersions of aluminium oxide particles, are irradiated with 1 MeV electrons in a High Voltage Electron Microscope at 600°C. Evidence of grain size dependent void nucleation, void concentration, and void volume swelling...

  18. Effect of austenitization heat treatment on the magnetic properties of Fe-40wt% Ni-2wt% Mn alloy

    Institute of Scientific and Technical Information of China (English)

    S. Buyukakkas; H. Aktas; S. Akturk

    2007-01-01

    The effect of austenitization heat treatment on magnetic properties was examined by means of M(o)ssbauer spectroscopy on an Fe-40wt%Ni-2wt%Mn alloy. The morphology of the alloy was obtained by using scanning electron microscopy (SEM) under different heat treatment conditions. The magnetic behavior of the non heat-treated alloy is ferromagnetic. A mixed magnetic structure including both paramagnetic and ferromagnetic states was obtained at 800℃ after 6 and 12 h heat treatments. In addition, the magnetic structure of the heat-treated alloy at 1150℃ for 12 h was ferromagnetic. With the volume fraction changing, the effective hyperfine field of the ferromagnetic austenite phase and isomery shift values were also determined by M(o)ssbauer spectroscopy.

  19. Sub-zero austenite to martensite transformation in a Fe-Ni-0.6wt.%C alloy

    DEFF Research Database (Denmark)

    Villa, Matteo; Pantleon, Karen; Somers, Marcel A. J.

    2011-01-01

    Martensitic transformation in a model Fe-Ni-0.6wt%C alloy was investigated at sub-zero Celsius temperature. The influence of the thermal path in determining the conditions leading to the formation of martensite was studied. In the investigation, samples were austenitized and quenched, whereafter...

  20. Irradiation-induced changes of martensitic transformation temperatures in a TiNiNb shape memory alloy

    Energy Technology Data Exchange (ETDEWEB)

    Mo, H.Q. [Department of Applied Physics, University of Electronic Science and Technology of China, Chengdu 610054 (China); Department of Material Forming and Controlling Engineering, Sichuan University, Chengdu 610065 (China); Zu, X.T. [Department of Applied Physics, University of Electronic Science and Technology of China, Chengdu 610054 (China)]. E-mail: xiaotaozu@yahoo.com; Huo, Y. [Department of Mechanics, Fudan University, Shanghai 200433 (China)

    2005-04-15

    Effects of electron irradiations on the transition behavior of 1123 K annealed Ti{sub 44}Ni{sub 47}Nb{sub 9} shape memory alloy specimens were studied. The transformation temperatures and the latent heat of phase transformation were measured by differential scanning calorimeter (DSC). The microstructure changes were determined by XRD and TEM. The 1.7 MeV electron irradiation increases the martensitic transformation start temperature, finish temperature, austenite transformation start, finish temperature by {approx}20 K. The XRD and TEM observation showed that the volume fraction of {beta}-Nb precipitate increased after electron irradiation, which contributed to the observed changes of the transformation temperatures.

  1. Effects of low temperature neutron irradiation on deformation behavior of austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, J.E.; Rowcliffe, A.F.; Alexander, D.J.; Grossbeck, M.L. [Oak Ridge National Laboratory, TN (United States); Shiba, K.

    1996-04-01

    An austenitic stainless steel, designated 316LN-IG, has been chosen for the first wall/shield (FW/S) structure for the International Thermonuclear Experimental Reactor (ITER). The proposed operational temperature range for the structure (100 to 250{degree}C) is below the temperature regimes for void swelling (400-600{degree}C) and for helium embrittlement (500-700{degree}C). However, the proposed neutron dose is such that large changes in yield strength, deformation mode, and strain hardening capacity could be encountered which could significantly affect fracture properties. Definition of the irradiation regimes in which this phenomenon occurs is essential to the establishment of design rules to protect against various modes of failure.

  2. Composite model of microstructural evolution in austenitic stainless steel under fast neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.; Odette, G.R.

    1986-01-01

    A rate-theory-based model has been developed which includes the simultaneous evolution of the dislocation and cavity components of the microstructure of irradiated austenitic stainless steels. Previous work has generally focused on developing models for void swelling while neglecting the time dependence of the dislocation structure. These models have broadened our understanding of the physical processes that give rise to swelling, e.g., the role of helium and void formation from critically-sized bubbles. That work has also demonstrated some predictive capability by successful calibration to fit the results of fast reactor swelling data. However, considerable uncertainty about the values of key parameters in these models limits their usefulness as predictive tools. Hence the use of such models to extrapolate fission reactor swelling data to fusion reactor conditions is compromised.

  3. The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part II. Fatigue crack growth rate

    Science.gov (United States)

    Margolin, B.; Minkin, A.; Smirnov, V.; Sorokin, A.; Shvetsova, V.; Potapova, V.

    2016-11-01

    The experimental data on the fatigue crack growth rate (FCGR) have been obtained for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various radiation swelling. The performed study of the fracture mechanisms for cracked specimens under cyclic loading has explained why radiation swelling affects weakly FCGR unlike its effect on fracture toughness. Mechanical modeling of fatigue crack growth has been carried out and the dependencies for prediction of FCGR in irradiated austenitic steel with and with no swelling are proposed and verified with the obtained experimental results. As input data for these dependencies, FCGR for unirradiated steel and the tensile mechanical properties for unirradiated and irradiated steels are used.

  4. Irradiation Behavior in High Entropy Alloys

    Institute of Scientific and Technical Information of China (English)

    Song-qin XIA; Zhen WANG; Teng-fei YANG; Yong ZHANG

    2015-01-01

    As an increasing demand of advanced nuclear fission reactors and fusion facilities, the key requirements for the materials used in advanced nuclear systems should encompass superior high temperature property, good behavior in corrosive environment, and high irradiation resistance, etc. Recently, it was found that some selected high entropy alloys (HEAs) possess excellent mechanical properties at high temperature, high corrosion resistance, and no grain coarsening and self-healing abil-ity under irradiation, especially, the exceptional structural stability and lower irradiation-induced volume swelling, compared with other conventional materials. Thus, HEAs have been considered as the potential nuclear materials used for future ifssion or fusion reactors, which are designed to operate at higher temperatures and higher radiation doses up to several hundreds of displacement per atom (dpa). An insight into the irradiation behavior of HEAs was given, including fundamental researches to investigate the irradiation-induced phase crystal structure change and volume swelling in HEAs. In summary, a brief overview of the irradiation behavior in HEAs was made and the irradiation-induced structural change in HEAs may be relatively insensi-tive because of their special structures.

  5. Carburization of austenitic and ferritic alloys in hydrocarbon environments at high temperature

    Directory of Open Access Journals (Sweden)

    Serna, A.

    2003-12-01

    Full Text Available The technical and industrial aspects of high temperature corrosion of materials exposed to a variety of aggressive environments have significant importance. These environments include combustion product gases and hydrocarbon gases with low oxygen potentials and high carbon potentials. In the refinery and petrochemical industries, austenitic and ferritic alloys are usually used for tubes in fired furnaces. The temperature range for exposure of austenitic alloys is 800-1100 °C, and for ferritic alloys 500-700 °C, with carbon activities ac > 1 in many cases. In both applications, the carburization process involves carbon (coke deposition on the inner diameter, carbon absorption at the metal surface, diffusion of carbon inside the alloy, and precipitation and transformation of carbides to a depth increasing with service. The overall kinetics of the internal carburization are approximately parabolic, controlled by carbon diffusion and carbide precipitation. Ferritic alloys exhibit gross but uniform carburization while non-uniform intragranular and grain-boundary carburization is observed in austenitic alloys.

    La corrosión a alta temperatura, tal como la carburación de materiales expuestos a una amplia variedad de ambientes agresivos, tiene especial importancia desde el punto de vista técnico e industrial. Estos ambientes incluyen productos de combustión, gases e hidrocarburos con bajo potencial de oxígeno y alto potencial de carbono. En las industrias de refinación y petroquímica, las aleaciones austeníticas y ferríticas se utilizan en tuberías de hornos. El rango de temperatura de exposición para aleaciones austeníticas está entre 800-1.100°C y para aleaciones ferríticas está entre 500-700°C, con actividades de carbono ac>1 en algunos casos. En tuberías con ambas aleaciones, el proceso de carburación incluye deposición de carbón (coque en el diámetro interno, absorción de carbono en la superficie

  6. Project on New Domestic Zirconium Alloy Fuel Assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Pei-sheng; ZHANG; Ai-min

    2012-01-01

    <正>The objectives of the project is to conduct irradiation at research reactor for small fuel assembly with domestic new zirconium alloy, and then to carry out post irradiation examination, and finally to acquire

  7. Alloy development for irradiation performance: program strategy

    Energy Technology Data Exchange (ETDEWEB)

    Bloom, E. E.; Stiegler, J. O.; Wiffen, F. W.; Dalder, E. N.C.; Reuther, T. C.; Gold, R. E.; Holmes, J. J.; Kummer, D. L.; Nolfi, F. V.

    1978-01-01

    The objective of the Alloy Development for Irradiation Performance Program is the development of structural materials for use in the first wall and blanket region of fusion reactors. The goal of the program is a material that will survive an exposure of 40 MWyr/m/sup 2/ at a temperature which will allow use of a liquid-H/sub 2/O heat transport system. Although the ultimate aim of the program is development of materials for commercial reactors by the end of this century, activities are organized to provide materials data for the relatively low performance interim machines that will precede commercial reactors.

  8. Effect of alloying elements on branching of primary austenite dendrites in Ni-Mn-Cu cast iron

    Directory of Open Access Journals (Sweden)

    A. Janus

    2011-04-01

    Full Text Available Within the research, determined were direction and intensity of influence of individual alloying elements on branching degree of primary austenite dendrites in austenitic cast iron Ni-Mn-Cu. 30 cast shafts dia. 20 mm were analysed. Chemical composition of the alloywas as follows: 2.0 to 3.3 % C, 1.4 to 3.1 % Si, 2.8 to 9.5 % Ni, 0.4 to 7.7 % Mn, 0 to 4.6 % Cu, 0.14 to 0.16 % P and 0.03 to 0.04 % S.Analysis was performed separately for the dendrites solidifying in directional and volumetric way. The average distance "x" between the2nd order arms was accepted as the criterion of branching degree. It was found that influence of C, Si, Ni, Mn and Cu on the parameter "x"is statistically significant. Intensity of carbon influence is decidedly higher than that of other elements, and the influence is more intensive in the directionally solidifying dendrites. However, in the case of the alloyed cast iron Ni-Mn-Cu, combined influence of the alloying elements on solidification course of primary austenite can be significant.

  9. Fracture behavior of neutron-irradiated high-manganese austenitic steels

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Narui, M.; Kayano, H.

    1991-03-01

    The instrumented Charpy impact test was applied to study the fracture behavior of high-manganese austenitic steels before and after neutron irradiations. Quarter-size specimens of a commercial high-manganese steel (18% Mn-5% Ni-16% Cr), three reference steels (21% Mn-1% Ni-9% Cr, 20% Mn-1% Ni-11% Cr, 15% Mn-1% Ni-13% Cr) and two model steels (17% Mn-4.5% Si-6.5% Cr, 22% Mn-4.5% Si-6.5% Cr-0.2% N) were used for the impact tests at temperatures between 77 and 523 K. The load-deflection curves showed typical features corresponding to characteristics of the fracture properties. The temperature dependences of fracture energy and failure deflection obtained from the curves clearly demonstrate only small effects up to 2 × 10 23 n/m 2 ( E > 0.1 MeV) and brittleness at room temperature in 17% Mn-Si-Cr steel at 1.6 × 10 25 n/m 2 ( E > 0.1 MeV), while ductility still remains in 22%Mn-Si-Cr steel.

  10. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    Science.gov (United States)

    Subbotin, A. V.; Panyukov, S. V.

    2016-08-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  11. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: the case of Fe-Cu model alloys

    CERN Document Server

    Subbotina, A V

    2016-01-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. We show that the obtained results are in a good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  12. Determination of the orientation relationship between austenite and incommensurate 7M modulated martensite in Ni-Mn-Ga alloys

    Energy Technology Data Exchange (ETDEWEB)

    Li, Z.B. [Key Laboratory for Anisotropy and Texture of Materials (Ministry of Education), Northeastern University, Shenyang 110819 (China); Zhang, Y.D. [Laboratoire d' Etude des Microstructures et de Mecanique des Materiaux (LEM3), CNRS UMR 7239, Universite Paul Verlaine - Metz, 57045 Metz (France); Esling, C., E-mail: claude.esling@univ-metz.fr [Laboratoire d' Etude des Microstructures et de Mecanique des Materiaux (LEM3), CNRS UMR 7239, Universite Paul Verlaine - Metz, 57045 Metz (France); Zhao, X. [Key Laboratory for Anisotropy and Texture of Materials (Ministry of Education), Northeastern University, Shenyang 110819 (China); Zuo, L., E-mail: lzuo@mail.neu.edu.cn [Key Laboratory for Anisotropy and Texture of Materials (Ministry of Education), Northeastern University, Shenyang 110819 (China)

    2011-04-15

    For Ni-Mn-Ga ferromagnetic shape memory alloys, a large magnetic-field-induced strain could be reached through the reorientation of martensitic variants in the martensite state. Owing to the collective and displacive nature of the austenite to martensite transformation, a certain orientation relationship (OR) between the parent and the product phase is required to minimize the transformation strain and the strain energy generated, which brings about self-accommodating groups of martensitic variants with specific orientation correlations. In this work, the microstructural and crystallographic characteristics of martensitic variants in a polycrystalline Ni{sub 50}Mn{sub 30}Ga{sub 20} alloy were investigated by electron backscatter diffraction analysis. With accurate orientation measurement on inherited martensitic variants, the local orientations of parent austenite grains were predicted using four classical OR for the martensitic transformation. Furthermore, a specific OR, namely the Pitsch relation with (1 0 1){sub A}//(1 2-bar 10-bar){sub 7M} and [1 0 1-bar]{sub A}//[10-bar 10-bar 1]{sub 7M}, was unambiguously determined by considering the magnitude of discontinuity between the lattices of the product and parent phases and the structural modulation of the incommensurate 7M modulated martensite. The present procedure to determine the OR, without recourse to the presence of retained austenite, is in general applicable to a variety of materials with modulated superstructure for insight into their martensitic transformation processes.

  13. Evaluation of Tensile Property of Austenitic Alloys Exposed to High-Temperature S-CO{sub 2} Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyunmyung; Lee, Ho Jung; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-12-15

    Super-critical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been considered to replace the current steam Rankine cycle in Sodium-cooled Fast Reactor (SFR) in order to improve the inherent safety and thermal efficiency. Several austenitic alloys are considered as the structural materials for high temperature S-CO{sub 2} environment. Microstructural change after long-term exposure to high temperature S-CO{sub 2} environment could affect to the mechanical properties. In this study, candidate materials (austenitic stainless steels and Alloy 800HT) were exposed to S-CO{sub 2} to assess oxidation resistance and the change in tensile properties. Loss of ductility was observed for some austenitic stainless steels even after 250 h exposure. The contribution of S-CO{sub 2} environment on such changes was analyzed based on the characterization of the surface oxide and carburization of the materials in which 316H and 800H showed different oxidation behaviors.

  14. In Situ Observation of Austenite Growth During Continuous Heating in Very-Low-Carbon Fe-Mn and Ni Alloys

    Science.gov (United States)

    Enomoto, M.; Wan, X. L.

    2017-02-01

    The growth of austenite during continuous heating was observed in situ under a confocal scanning laser microscope in Fe-Mn and Ni alloys containing less than 0.01 mass pct C. The advancements of the α/γ boundary were measured in the temperature range of ca. 40 K, which encompassed the Ae3 line of the alloys. Below Ae3, the growth rates were of the same order of magnitude as those predicted from the carbon diffusion-controlled negligible partition local equilibrium in the (α + γ) two-phase region, whereas those observed near and above the Ae3 were ca. two orders of magnitude greater. The α/γ boundary mobilities evaluated therefrom were somewhat smaller than those obtained previously in massive ferrite transformation during continuous cooling in the same alloys, albeit the experimental scatter was large and fell near the mobilities proposed in the literature. The α/γ boundary migrated probably with a carbon diffusion spike ahead of the boundary and the solute drag of the carbon or alloy element is unlikely to be operative during the growth of austenite.

  15. Corrosion behaviour of austenitic stainless steel, nickel-base alloy and its weldments in aqueous LiBr solutions

    Energy Technology Data Exchange (ETDEWEB)

    Blasco-Tamarit, E.; Igual-Munoz, A.; Garcia Anton, J.; Garcia-Garcia, D. [Departamento de Ingenieria Quimica y Nuclear. E.T.S.I.Industriales, Universidad Politecnica de Valencia, P.O. Box 22012 E-46071 Valencia (Spain)

    2004-07-01

    With the advances in materials production new alloys have been developed, such as High- Alloy Austenitic Stainless Steels and Nickel-base alloys, with high corrosion resistance. These new alloys are finding applications in Lithium Bromide absorption refrigeration systems, because LiBr is a corrosive medium which can cause serious corrosion problems, in spite of its favourable properties as absorbent. The objective of the present work was to study the corrosion resistance of a highly alloyed austenitic stainless steel (UNS N08031) used as base metal, a Nickel-base alloy (UNS N06059) used as its corresponding filler metal, and the weld metal obtained by the Gas Tungsten Arc Welding (GTAW) procedure. The materials have been tested in different LiBr solutions (400 g/l, 700 g/l, 850 g/l and a commercial 850 g/l LiBr heavy brine containing Lithium Chromate as corrosion inhibitor), at 25 deg. C. Open Circuit Potential tests and potentiodynamic anodic polarization curves have been carried out to obtain information about the general electrochemical behaviour of the materials. The polarization curves of all the alloys tested were typical of passivable materials. Pitting corrosion susceptibility has been evaluated by means of cyclic potentiodynamic curves, which provide parameters to analyse re-passivation properties. The galvanic corrosion generated by the electrical contact between the welded and the base material has been estimated from the polarization diagrams according to the Mixed Potential Method. Samples have been etched to study the microstructure by Scanning Electron Microscopy (SEM). The results demonstrate that the pitting resistance of all these materials increases as the LiBr concentration decreases. In general, the presence of chromate tended to shift the pitting potential to more positive values than those obtained in the 850 g/l LiBr solution. (authors)

  16. Analysis of phase transformation from austenite to martensite in NiTi alloy strips under uniaxial tension

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Phase transformation from austenite to martensite in NiTi alloy strips under the uniaxial tension has been observed in experiments and numerically simulated as a localized deformation. This work presents an analysis using the theory of phase transformation. The jump of deformation gradient across the interface between two phases and the Maxwell relation are considered. Governing equations for the phase transformation are derived. The analysis is reduced to finding the minimum value of the loading at which the governing equations have a unique, real and physically acceptable solution. The equations are solved numerically and it is verified that the unique solution exists definitely.The Maxwell stress, the stresses and strains inside both austenite and martensite phases,and the transformation-front orientation angle are determined to be in reasonably good agreement with experimental observations.

  17. In situ TEM study of G-phase precipitates under heavy ion irradiation in CF8 cast austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Wei-Ying [Argonne National Laboratory, Argonne, IL 60439 (United States); University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Li, Meimei; Zhang, Xuan; Kirk, Marquis A.; Baldo, Peter M. [Argonne National Laboratory, Argonne, IL 60439 (United States); Lian, Tiangan [Electric Power Research Institute, Palo Alto, CA 94304 (United States)

    2015-09-15

    Thermally-aged cast austenitic stainless steels (CASS) CF8 was irradiated with 1 MeV Kr ions at 300, 350 and 400 °C to 1.88 × 10{sup 19} ions/m{sup 2} (∼3 dpa) at the IVEM-Tandem Facility at the Argonne National Laboratory. Before irradiation, the distribution of G-phase precipitates in the ferrite showed spatial variations, and both their size and density were affected by the ferrite–austenite phase boundary and presence of M{sub 23}C{sub 6} carbides. Under 300 °C irradiation, in situ TEM observation showed G-phase precipitates were relatively unchanged in the vicinity of the phase boundary M{sub 23}C{sub 6} carbides, while the density of G-phase precipitates increased with increasing dose within the ferrite matrix. Coarsening of G-phase precipitates was observed in the vicinity of phase boundary M{sub 23}C{sub 6} carbides at 350 °C and 400 °C.

  18. The formation of radiation-induced segregation at twin bands in ion-irradiated austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hyung-Ha; Lee, Gyeong-Geun; Kwon, Junhyun; Hwang, Seong Sik [Nuclear Materials Division, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Shin, Chansun, E-mail: c.shin@mju.ac.kr [Department of Materials Science and Engineering, Myongji University, 116 Myongji-ro, Cheoin-gu, Youngin, Gyeonggi-do 449-728 (Korea, Republic of)

    2014-11-15

    Radiation-induced segregation (RIS) at twins was investigated using transmission electron microscopy (TEM) for ion-irradiated austenitic stainless steel. Significant RIS was found to occur at twin boundaries. TEM analysis indicates that interfacial dislocations at partially coherent twin boundaries are potential sites for strong RIS phenomenon. The RIS causes the formation of thin bands having a higher Ni and lower Cr concentration in twin bands with a width less than 15 nm. In wider twin bands, strong RIS occurs only at the outer twin boundaries, but not inside the band. The possible mechanism for the formation of the RIS thin band is discussed.

  19. Microstructure of Au-ion irradiated 316L and FeNiCr austenitic stainless steels

    Science.gov (United States)

    Jublot-Leclerc, S.; Li, X.; Legras, L.; Lescoat, M.-L.; Fortuna, F.; Gentils, A.

    2016-11-01

    Thin foils of 316L were irradiated in situ in a Transmission Electron Microscope with 4 MeV Au ions at 450 °C and 550 °C. Similar irradiations were performed at 450 °C in FeNiCr. The void and dislocation microstructure of 316L is found to depend strongly on temperature. At 450 °C, a dense network of dislocation lines is observed in situ to grow from black dot defects by absorption of other black dots and interstitial clusters whilst no Frank loops are detected. At 550 °C, no such network is observed but large Frank loops and perfect loops whose sudden appearance is concomitant with a strong increase in void density as a result of a strong coupling between voids and dislocations. Moreover, differences in both alloys microstructure show the major role played by the minor constituents of 316L, increasing the stacking fault formation energy, and possibly leading to significant differences in swelling behaviour.

  20. Microstructural Features of Austenite Formation in C35 and C45 alloys

    OpenAIRE

    2007-01-01

    The microstructural evolution during continuous heating experiments has been studied for two C-Mn steels with carbon contents in the range 0.35 to 0.45 wt pct using optical microscopy, scanning electron microscopy (SEM), and electron probe microanalysis (EPMA). It is shown that the formation of the austenitic phase is possible in pearlite as well as in ferrite regions. Thus, a considerable overlap in time of ferrite-to-austenite and pearlite-to-austenite transformations is likely to occur. An...

  1. Stacking faults and microstrains in strain-hardened surface of nitrogen-alloyed austenitic steel

    Science.gov (United States)

    Narkevich, N.; Syrtanov, M.; Mironov, Yu.; Surikova, N.

    2016-11-01

    X-ray diffractometry has been applied to examine the effect of ultrasonic forging and frictional treatment on structural parameters and oriented microstrains responsible for the generation of residual microstresses in austenitic steel Fe-17Cr-19Mn-0.52N. The maximum stacking fault density α = 0.067 is observed in the steel surface layer of thickness 5 µm after frictional treatment. A decrease in the austenite lattice parameter after deformation treatment is associated with the change in the sign (direction) of residual stresses. Surface deformation treatment induces compression of the austenite lattice along the normal to the surface.

  2. Subgrain and dislocation structure changes in hot-deformed high-temperature Fe-Ni austenitic alloy

    Energy Technology Data Exchange (ETDEWEB)

    Ducki, K.J.; Rodak, K.; Hetmanczyk, M.; Kuc, D

    2003-08-28

    The influence of plastic deformation on the substructure of a high-temperature austenitic Fe-Ni alloy has been presented. Hot-torsion tests were executed at constant strain rates of 0.1 and 1.0 s{sup -1}, at testing temperatures in the range 900-1150 deg. C. The examination of the microstructure was carried out, using transmission electron microscopy. Direct measurements on the micrographs allowed the calculation of structural parameters: the average subgrain area, and the mean dislocation density. A detailed investigation has shown that the microstructure is inhomogeneous, consisting of dense dislocation walls, subgrains and recrystallized regions.

  3. Development of Cast Alumina-forming Austenitic Stainless Steel Alloys for use in High Temperature Process Environments

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Yamamoto, Yukinori [ORNL; Brady, Michael P [ORNL; Pint, Bruce A [ORNL; Pankiw, Roman [Duraloy Technologies Inc; Voke, Don [Duraloy Technologies Inc

    2015-01-01

    There is significant interest in the development of alumina-forming, creep resistant alloys for use in various industrial process environments. It is expected that these alloys can be fabricated into components for use in these environments through centrifugal casting and welding. Based on the successful earlier studies on the development of wrought versions of Alumina-Forming Austenitic (AFA) alloys, new alloy compositions have been developed for cast products. These alloys achieve good high-temperature oxidation resistance due to the formation of protective Al2O3 scales while multiple second-phase precipitation strengthening contributes to excellent creep resistance. This work will summarize the results on the development and properties of a centrifugally cast AFA alloy. This paper highlights the strength, oxidation resistance in air and water vapor containing environments, and creep properties in the as-cast condition over the temperature range of 750°C to 900°C in a centrifugally cast heat. Preliminary results for a laboratory cast AFA composition with good oxidation resistance at 1100°C are also presented.

  4. Structural Transformations in Austenitic Stainless Steel Induced by Deuterium Implantation: Irradiation at 295 K

    Science.gov (United States)

    Morozov, Oleksandr; Zhurba, Volodymir; Neklyudov, Ivan; Mats, Oleksandr; Progolaieva, Viktoria; Boshko, Valerian

    2016-02-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic steel 18Cr10NiTi pre-implanted at 295 K with deuterium ions in the dose range from 8 × 1014 to 2.7 × 1018 D/cm2. The kinetics of structural transformation development in the steel layer was traced from deuterium thermodesorption spectra as a function of deuterium concentration. Three characteristic regions with different low rates of deuterium amount desorption as the implantation dose increases were revealed: I—the linear region of low implantation doses (up to 1 × 1017 D/cm2); II—the nonlinear region of medium implantation doses (1 × 1017 to 8 × 1017 D/cm2); III—the linear region of high implantation doses (8 × 1017 to 2.7 × 1018 D/cm2). During the process of deuterium ion irradiation, the coefficient of deuterium retention in steel varies in discrete steps. Each of the discrete regions of deuterium retention coefficient variation corresponds to different implanted-matter states formed during deuterium ion implantation. The low-dose region is characterized by formation of deuterium-vacancy complexes and solid-solution phase state of deuterium in the steel. The total concentration of the accumulated deuterium in this region varies between 2.5 and 3 at.%. The medium-dose region is characterized by the radiation-induced action on the steel in the presence of deuterium with the resulting formation of the energy-stable nanosized crystalline structure of steel, having a developed network of intercrystalline boundaries. The basis for this developed network of intercrystalline boundaries is provided by the amorphous state, which manifests itself in the thermodesorption spectra as a widely temperature-scale extended region of deuterium desorption (structure formation with a varying activation energy). The total concentration of the accumulated deuterium in the region of medium implantation doses makes 7 to 8 at.%. The resulting structure shows stability against the action of

  5. Analysis Of The Austenite Grain Growth In Low-Alloy Boron Steel With High Resistance To Abrasive Wear

    Directory of Open Access Journals (Sweden)

    Białobrzeska B.

    2015-09-01

    Full Text Available Today low-alloy steels with boron achieve high resistance to abrasive wear and high strength. These features are obtained by using advanced technology of manufacturing. This makes boron steels increasingly popular and their application more diverse. Application of these steels can extend the lifetime of very expensive machine construction in many industries such as mining, the automotive, and agriculture industries. An interesting subgroup of these materials is steel with boron intended for heat treatment. These steels are supplied by the manufacturer after cold or hot rolling so that it is possible for them to be heat treated in a suitable manner by the purchaser for its specific application. Very important factor that determines the mechanical properties of final product is austenite grain growth occurring during hot working process such us quenching or hot rolling. Investigation of the effect of heating temperature and holding time on the austenite grain size is necessary to understand the growth behavior under different conditions. This article presents the result of investigation of austenite grain growth in selected low-allow boron steel with high resistance to abrasive wear and attempts to describe the influence of chemical composition on this process.

  6. Study of Fe-12Cr-20Mn-W-C austenitic steels irradiated in the SM-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K.; Bulanova, T.M.; Neustroyev, V.S.; Ostrovsky, Z.E.; Kosenkov, V.M. (V.I. Lenin Research Inst. of Atomic Reactors, Dimitrovgrad (Russia)); Ivanov, L.I.; Djomina, E.V. (A.A. Baikov Inst. of Metallurgy, Academy of Science, Moscow (Russia))

    1992-09-01

    A comparison has been made between the mechanical properties and swelling of austenitic stainless steels EP-838 (Fe-Cr-Mn) and 316SS (Fe-Cr-Ni) irradiated in the mixed-neutron spectrum of the SM-2 reactor in the temperture range 400-800deg C (every 100deg C) to 16 dpa dose with 1000 and 3000 appm helium generation, correspondingly, determined by nickel content. EP-838 exhibited less susceptibility to void swelling and radiation hardening. Fe-12Cr-20Mn-W-0.1C steel without nickel irradiated at 100deg C to 21 dpa exhibited significant radiation hardening accompanied by [alpha]-phase formation in the steel structure. (orig.).

  7. Study of Fe-12Cr-20Mn-W-C austenitic steels irradiated in the SM-2 reactor

    Science.gov (United States)

    Shamardin, V. K.; Bulanova, T. M.; Neustroyev, V. S.; Ostrovsky, Z. E.; Kosenkov, V. M.; Ivanov, L. I.; Djomina, E. V.

    1992-09-01

    A comparison has been made between the mechanical properties and swelling of austenitic stainless steels EP-838 (Fe-Cr-Mn) and 316SS (Fe-Cr-Ni) irradiated in the mixed-neutron spectrum of the SM-2 reactor in the temperature range 400-800°C (every 100°C) to 16 dpa dose with 1000 and 3000 appm helium generation correspondingly, determined by nickel content. EP-838 exhibited less susceptibility to void swelling and radiation hardening. Fe-12Cr-20Mn-W-0.1C steel without nickel irradiated at 100°C to 21 dpa exhibited significant radiation hardening accompanied by α-phase formation in the steel structure.

  8. Hydrogen embrittlement and hydrogen induced stress corrosion cracking of high alloyed austenitic materials; Wasserstoffversproedung und wasserstoffinduzierte Spannungsrisskorrosion hochlegierter austenitischer Werkstoffe

    Energy Technology Data Exchange (ETDEWEB)

    Mummert, K.; Uhlemann, M.; Engelmann, H.J. [Institut fuer Festkoerper- und Werkstofforschung Dresden e.V. (Germany)

    1998-11-01

    The susceptiblity of high alloyed austenitic steels and nickel base alloys to hydrogen-induced cracking is particularly determined by 1. the distribution of hydrogen in the material, and 2. the microstructural deformation behaviour, which last process is determined by the effects of hydrogen with respect to the formation of dislocations and the stacking fault energy. The hydrogen has an influence on the process of slip localization in slip bands, which in turn affects the microstructural deformation behaviour. Slip localization increases with growing Ni contents of the alloys and clearly reduces the ductility of the Ni-base alloy. Although there is a local hydrogen source involved in stress corrosion cracking, emanating from the corrosion process at the cathode, crack growth is observed only in those cases when the hydrogen concentration in a small zone ahead of the crack tip reaches a critical value with respect to the stress conditions. Probability of onset of this process gets lower with growing Ni content of the alloy, due to increasing diffusion velocity of the hydrogen in the austenitic lattice. This is why particularly austenitic steels with low Ni contents are susceptible to transcrystalline stress corrosion cracking. In this case, the microstructural deformation process at the crack tip is also influenced by analogous processes, as could be observed in hydrogen-loaded specimens. (orig./CB) [Deutsch] Die Empfindlichkeit von hochlegierten austentischen Staehlen und Nickelbasislegierungen gegen wasserstoffinduziertes Risswachstum wird im wesentlichen bestimmt durch 1. die Verteilung von Wasserstoff im Werkstoff und 2. das mikrostrukturelle Verformungsverhalten. Das mikrostrukturelle Deformationsverhalten ist wiederum durch den Einfluss von Wasserstoff auf die Versetzungsbildung und die Stapelfehlerenergie charakterisiert. Das mikrostrukturelle Verformungsverhalten wird durch wasserstoffbeeinflusste Gleitlokalisierung in Gleitbaendern bestimmt. Diese nimmt mit

  9. Phase diffusionless γ↔α transformations and their effect on physical, mechanical and corrosion properties of austenitic stainless steels irradiated with neutrons and charged particles

    Science.gov (United States)

    Maksimkin, O. P.

    2016-04-01

    The work presents relationships of γ→α' and α'→γ-transformations in reactor 12Cr18Ni10Ti and 08Cr16Ni11Mo3 austenitic stainless steels induced by cold work, irradiation and/or temperature. Energy and mechanical parameters of nucleation and development of deformation-induced martensitic α'-phase in the non-irradiated and irradiated steels are given. The mechanisms of localized static deformation were investigated and its effect on martensitic γ→α' transformation is determined. It has been shown that irradiation of 12Cr18Ni10Ti steel with heavy Kr ions (1.56MeV/nucleon, fluence of 1·1015 cm-2) results in formation of α'-martensite in near-surface layer of the sample. Results of systematic research on reversed α'→γ-transformation in austenitic metastable stainless steels irradiated with slow (VVR-K) and fast (BN-350) neutrons are presented. The effect of annealing on strength and magnetic characteristics was determined. It was found that at the temperature of 400 °C in the irradiated with neutrons samples (59 dpa) an increase of ferromagnetic α'-phase and microhardness was observed. The obtained results could be used during assessment of operational characteristics of highly irradiated austenitic steels during transportation and storage of Fuel Assemblies for fast nuclear reactors.

  10. Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

    Science.gov (United States)

    Kasahara, Shigeki; Kitsunai, Yuji; Chimi, Yasuhiro; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-11-01

    This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. One of the temperature profiles was that the specimens experienced neutron irradiation in both reactors, under which the irradiation temperature transiently increased to 290 °C from room temperature with increasing reactor power during reactor startup periods. Another was that the specimens were pre-heated to about 150 °C prior to the irradiation to suppress the transient temperature increase. Tensile tests at 290 °C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Difference of the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. Although influence of neutron irradiation involving transient temperature increase to 290 °C from room temperature on the yield strength and the Vickers hardness is buried in the trend curves of existing data, the influence was also found certainly in increment of in yield strength, existence of modest yield drop, and loss of strain hardening capacity and ductility. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, seemed to have important implications regarding the interpretation of not irradiation hardening, but deformation of the austenitic stainless steel.

  11. Swelling and microstructure of austenitic stainless steel ChS-68 CW after high dose neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Porollo, S.I.; Konobeev, Yu.V. [State Scientific Center of Russian Federation - Institute of Physics and Power Engineering (IPPE), Obninsk, Kaluga Region (Russian Federation); Garner, F.A., E-mail: frank.garner@dslextreme.co [Radiation Effects Consulting, 2003 Howell Avenue, Richland, WA 99354 (United States)

    2009-08-15

    Austenitic stainless steel ChS-68 serving as fuel pin cladding was irradiated in the 20% cold-worked condition in the BN-600 fast reactor in the range 56-84 dpa. This steel was developed to replace EI-847 which was limited by its insufficient resistance to void swelling. Comparison of swelling between EI-847 and ChS-68 under similar irradiation conditions showed improvement of the latter steel by an extended transient regime of an additional approx10 dpa. Concurrent with swelling was the development of a variety of phases. In the temperature range 430-460 deg. S where the temperature peak of swelling was located, the principal type of phase generated during irradiation was G-phase, with volume fraction increasing linearly with dose to approx0.5% at 84 dpa. While the onset of swelling is concurrent with formation of G-phase, the action of G-phase cannot be confidently ascribed to significant removal from solution of swelling-suppressive elements such as silicon. A plausible mechanism for the higher resistance to void swelling of ChS-68 as compared with EI-847 may be related to an observed higher stability of faulted dislocation loops in ChS-68 that impedes the formation of a glissile dislocation network. The higher level of boron in ChS-68 is thought to be one contributor that might play this role.

  12. Cluster dynamics modeling of the effect of high dose irradiation and helium on the microstructure of austenitic stainless steels

    Science.gov (United States)

    Brimbal, Daniel; Fournier, Lionel; Barbu, Alain

    2016-01-01

    A mean field cluster dynamics model has been developed in order to study the effect of high dose irradiation and helium on the microstructural evolution of metals. In this model, self-interstitial clusters, stacking-fault tetrahedra and helium-vacancy clusters are taken into account, in a configuration well adapted to austenitic stainless steels. For small helium-vacancy cluster sizes, the densities of each small cluster are calculated. However, for large sizes, only the mean number of helium atoms per cluster size is calculated. This aspect allows us to calculate the evolution of the microstructural features up to high irradiation doses in a few minutes. It is shown that the presence of stacking-fault tetrahedra notably reduces cavity sizes below 400 °C, but they have little influence on the microstructure above this temperature. The binding energies of vacancies to cavities are calculated using a new method essentially based on ab initio data. It is shown that helium has little effect on the cavity microstructure at 300 °C. However, at higher temperatures, even small helium production rates such as those typical of sodium-fast-reactors induce a notable increase in cavity density compared to an irradiation without helium.

  13. Preliminary report on the irradiation conditions of the HFIR JP-23 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ermi, A.M. [Westinghouse Hanford Company, Richland, WA (United States); Gelles, D.S. [Pacific Northwest Laboratory, Richland, WA (United States)

    1995-04-01

    The objective of this effort was to irradiate a series of alloys over the temperature range 300 to 600{degrees}C to approximately 10 dpa in the High Flux Isotope Reactor (HFIR). The alloys covered a wide range of materials and treatments. The Japanese specimen matrix consisted of ferritic steels, vanadium alloys, copper alloys, molybdenum alloys, and titanium-aluminum compounds. The US specimen matrix consisted of vanadium alloys, 316 stainless steels, and isotopically tailored ferritic and austenitic alloys.

  14. Radiation Damages in Aluminum Alloy SAV-1 under Neutron Irradiation

    Science.gov (United States)

    Salikhbaev, Umar; Akhmedzhanov, Farkhad; Alikulov, Sherali; Baytelesov, Sapar; Boltabaev, Azizbek

    2016-05-01

    The aim of this work was to study the effect of neutron irradiation on the kinetics of radiation damages in the SAV-1 alloy, which belongs to the group of aluminum alloys of the ternary system Al-Mg-Si. For fast-neutron irradiation by different doses up to fluence 1019 cm-2 the SAV-1 samples were placed in one of the vertical channels of the research WWR type reactor (Tashkent). The temperature dependence of the electrical resistance of the alloy samples was investigated in the range 290 - 490 K by the four-compensation method with an error about 0.1%. The experimental results were shown that at all the temperatures the dependence of the SAV-1 alloy resistivity on neutron fluence was nonlinear. With increasing neutron fluence the deviation from linearity and the growth rate of resistivity with temperature becomes more appreciable. The observed dependences are explained by means of martensitic transformations and the radiation damages in the studied alloy under neutron irradiation. The mechanisms of radiation modification of the SAV-1 alloy structure are discussed.

  15. Static Softening in a Ni-30Fe Austenitic Model Alloy After Hot Deformation: Microstructure and Texture Evolution

    Science.gov (United States)

    Beladi, Hossein; Cizek, Pavel; Taylor, Adam S.; Rohrer, Gregory S.; Hodgson, Peter D.

    2017-02-01

    In the current study, the microstructure and texture characteristics of a model Ni-30Fe austenitic alloy were investigated during hot deformation and subsequent isothermal holding. The deformation led to the formation of self-screening arrays of microbands within a majority of grains. The microbands characteristics underwent rather modest changes during the post-deformation annealing, which suggests that limited dislocation annihilation occurs within the corresponding dislocation walls. The fraction of statically recrystallized (SRX) grains progressively increased with the holding time and closely matched the softening fraction measured from the offset flow stress approach. The corresponding texture was weak and preserved its character with the holding time. There was no pronounced temperature effect on the grain boundary character distribution after the completion of SRX. The Σ3 and Σ9 coincidence site lattice boundaries were characterized as (111) pure twist and (1-14) symmetric tilt types, respectively. Nonetheless, the recrystallization temperature slightly affected the grain boundary network.

  16. Recovery process of neutron-irradiated vanadium alloys in post-irradiation annealing treatment

    Energy Technology Data Exchange (ETDEWEB)

    Fukumoto, K., E-mail: fukumoto@u-fukui.ac.jp [Research Institute for Nuclear Engineering, University of Fukui, Tsuruga, Fukui 914-0055 (Japan); Iwasaki, M. [Research Institute for Nuclear Engineering, University of Fukui, Tsuruga, Fukui 914-0055 (Japan); Xu, Q. [KUR, Kyoto University, Kumatori, Osaka (Japan)

    2013-11-15

    Experiments to determine the influence of post-irradiation annealing on the mechanical properties and microstructures of neutron-irradiated V–4Cr–4Ti alloys were conducted. Two groups of specimens (as-irradiated specimens and specimens which underwent the post-irradiation annealing treatment) were subjected to tensile tests at room temperature and 773 K. Post-irradiation annealing experiments carried out over periods of up to 50 h were used to restore strength and ductility. As annealing time was extended, ductility was recovered up to 5% at 50 h anneal; however irradiation hardening was not recovered completely. Microstructural changes due to post-irradiation annealing corresponded to the amount that yield stress increased in tensile behavior in the irradiated specimen. The recovery in ductility was likely caused by the dissolution of interstitial impurities from defect clusters and dislocation cores produced by neutron irradiation during post-irradiation anneal treatment. A 3% elongation recovery in V–4Cr–4Ti alloys was achieved by annealing at 773 K for 20 h in a vacuum for neutron-irradiated samples at low temperature.

  17. Study of irradiation creep of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    Thin-wall tubing was produced from the 832665 (500 kg) heat of V-4 wt.% Cr-4 wt.% Ti to study its irradiation creep behavior. The specimens, in the form of pressurized capsules, were irradiated in Advanced Test Reactor and High Flux Isotope Reactor experiments (ATR-A1 and HFIR RB-12J, respectively). The ATR-A1 irradiation has been completed and specimens from it will soon be available for postirradiation examination. The RB-12J irradiation is not yet complete.

  18. A Hybrid Low Temperature Surface Alloying Process for Austenitic Stainless Steels

    Institute of Scientific and Technical Information of China (English)

    Y. Sun

    2004-01-01

    This paper describes a novel, hybrid process developed to engineer the surfaces of austenitic stainless steels at temperatures below 450℃ for the improvement in wear and corrosion resistance. The process is carried out in the plasma of a glow discharge containing both nitrogen and carbon reactive species, and facilitates the incorporation of both nitrogen and carbon into the austenite surface to form a dual-layer structure comprising a nitrogen-rich layer on top of a carbon-rich layer.Both layers can be precipitation-free at sufficiently low processing temperatures, and contain nitrogen and carbon respectively in supersaturated fcc austenite solid solutions. The resultant hybrid structure offers several advantages over the conventional low temperature nitriding and the newly developed carburizing processes in terms of mechanical and chemical properties, including higher surface hardness, a hardness gradient from the surface towards the layer-core interface, uniform layer thickness, and much enhanced corrosion resistance. This paper discusses the main features of this hybrid process and the various structural and properties characteristics of the resultant engineered surfaces.

  19. Irradiation creep and stress-enhanced swelling of Fe-16Cr-15Ni-Nb austenitic stainless steel in BN-350

    Energy Technology Data Exchange (ETDEWEB)

    Vorobjev, A.N.; Porollo, S.I.; Konobeev, Yu.V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

    1997-04-01

    Irradiation creep and void swelling will be important damage processes for stainless steels when subjected to fusion neutron irradiation at elevated temperatures. The absence of an irradiation device with fusion-relevant neutron spectra requires that data on these processes be collected in surrogate devices such as fast reactors. This paper presents the response of an annealed austenitic steel when exposed to 60 dpa at 480{degrees}C and to 20 dpa at 520{degrees}C. This material was irradiated as thin-walled argon-pressurized tubes in the BN-350 reactor located in Kazakhstan. These tubes were irradiated at hoop stresses ranging from 0 to 200 MPa. After irradiation both destructive and non-destructive examination was conducted.

  20. Change in the properties of Fe-Cr-Ni and Fe-Cr-Mn austenitic steels under mixed and fast neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Bulanova, T.M. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Golovanov, V.N. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Neustroyev, V.S. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Povstyanko, A.V. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Ostrovsky, Z.E. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors

    1996-10-01

    Detailed investigations are performed on mechanical properties, swelling and structure of different types of Fe-Cr-Ni and Fe-Cr-Mn austenitic stainless steels irradiated in the SM-2 high-flux research reactor and BOR-60 fast reactor. Steel irradiation temperatures are ranging from 100 up to 800 C and the maximum achieved level of damage doses is 60 dpa for Fe-Cr-Mn steel (with 4-5% of Ni) and 30 dpa for steels of the C-12Cr-20Mn-W-T type. Presented are dose dependencies of swelling and mechanical properties of Fe-Cr-Ni and Fe-Cr-Mn steels. It is shown that at temperatures below 530 C the investigated Fe-Cr-Mn steel systems are less susceptible to swelling as compared to Fe-Cr-Ni ones. Fe-Cr-Mn steels showed a lower value of irradiation embrittlement after irradiation in the mixed spectrum at temperatures from 100 up to 400 C and much higher embrittlement after irradiation from 350 up to 400 C in the fast spectrum in comparison with Fe-Cr-Ni steels. Higher hardening rate of Fe-Cr-Mn steels after their irradiation in BOR-60 is attributed to the presence of dislocation loops and defects of high density in the structure. The structural change features in Fe-Cr-Mn steels under irradiation are considered taking into account austenite stabilization in the initial state. (orig.).

  1. Change in the properties of FeCrNi and FeCrMn austenitic steels under mixed and fast neutron irradiation

    Science.gov (United States)

    Shamardin, V. K.; Bulanova, T. M.; Golovanov, V. N.; Neustroyev, V. S.; Povstyanko, A. V.; Ostrovsky, Z. E.

    1996-10-01

    Detailed investigations are performed on mechanical properties, swelling and structure of different types of FeCrNi and FeCrMn austenitic stainless steels irradiated in the SM-2 high-flux research reactor and BOR-60 fast reactor. Steel irradiation temperatures are ranging from 100 up to 800°C and the maximum achieved level of damage doses is 60 dpa for FeCrMn steel (with 4-5% of Ni) and 30 dpa for steels of the C12Cr20MnWT type. Presented are dose dependencies of swelling and mechanical properties of FeCrNi and FeCrMn steels. It is shown that at temperatures below 530°C the investigated FeCrMn steel systems are less susceptible to swelling as compared to FeCrNi ones. FeCrMn steels showed a lower value of irradiation embrittlement after irradiation in the mixed spectrum at temperatures from 100 up to 400°C and much higher embrittlement after irradiation from 350 up to 400°C in the fast spectrum in comparison with FeCrNi steels. Higher hardening rate of FeCrMn steels after their irradiation in BOR-60 is attributed to the presence of dislocation loops and defects of high density in the structure. The structural change features in FeCrMn steels under irradiation are considered taking into account austenite stabilization in the initial state.

  2. A new alloy design concept for austenitic stainless steel with tungsten modification for bipolar plate application in PEMFC

    Science.gov (United States)

    Kim, Kwang Min; Kim, Kyoo Young

    The feasibility of a new alloy design concept utilizing the principle of 'tungsten bronze effect' is critically evaluated for the development of metallic bipolar plates for proton exchange membrane fuel cell (PEMFC). An austenitic stainless steel (ASS) is modified with W and La to improve the stability of the passive film in an acidic environment as well as to reduce the contact resistance by the tungsten bronze effect. The experimental ASS containing W and La was evaluated in a simulated PEMFC environment of H 3PO 4 and H 2SO 4 solutions at 80 °C, and the electrical property was evaluated by performing a contact resistance test. The test results show that the ASS modified with W and La has good passive film stability for corrosion resistance and low contact resistance. The X-ray photoelectron spectroscopy (XPS) analysis clearly suggests the possibility of the tungsten bronze effect from the change in valency state of W 6+ to W 5+ in the passive film formed on the modified ASS. The feasibility of a new alloy design concept utilizing the 'tungsten bronze effect' is well demonstrated; however, more study is highly required for the development of metallic bipolar plates of PEMFC.

  3. Investigation of austenitic alloys for advanced heat recovery and hot gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    Alloys for design and construction of structural components needed to contain process streams and provide internal structures in advanced heat recovery and hot gas cleanup systems were examined. Emphasis was placed on high-strength, corrosion-resistant alloys for service at temperatures above 1000 {degrees}F (540{degrees}C). Data were collected that related to fabrication, joining, corrosion protection, and failure criteria. Alloys systems include modified type 310 and 20Cr-25Ni-Nb steels and sulfidation-resistance alloys HR120 and HR160. Types of testing include creep, stress-rupture, creep crack growth, fatigue, and post-exposure short-time tensile. Because of the interest in relatively inexpensive alloys for high temperature service, a modified type 310 stainless steel was developed with a target strength of twice that for standard type 310 stainless steel.

  4. Investigation of austenitic alloys for advanced heat recovery and hot gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W. [Oak Ridge National Lab., TN (United States)

    1996-08-01

    Materials properties were collected for the design and construction of structural components for use in advanced heat recovery and hot gas cleanup systems. Alloys systems included 9Cr-1Mo-V steel, modified 316 stainless steel, modified type 310 stainless steel, modified 20Cr-25Ni-Nb stainless steel, modified alloy 800, and two sulfidation resistant alloys: HR160 and HR120. Experimental work was undertaken to expand the databases for potentially useful alloys. Types of testing included creep, stress-rupture, creep-crack growth, fatigue, and post-exposure short-time tensile tests. Because of the interest in relatively inexpensive alloys for service at 700{degrees}C and higher, research emphasis was placed on a modified type 310 stainless steel and a modified 20Cr-25Ni-Nb stainless steel. Both steels were found to have useful strength to 925{degrees}C with good weldability and ductility.

  5. Anomalous transport properties of N i2M n1 -xC rxGa Heusler alloys at the martensite-austenite phase transition

    Science.gov (United States)

    Khan, Mahmud; Brock, Jeffrey; Sugerman, Ian

    2016-02-01

    The martensite-austenite phase transition in a series of N i2M n1 -xC rxGa Heusler alloys has been investigated by x-ray diffraction, dc magnetization, and electrical resistivity measurements. With increasing Cr concentration, the martensitic phase transformation shifts to higher temperature while the ferromagnetic transition shifts to lower temperature. For x 0.5 , the transition occurs in a paramagnetic state. The Cr doping results in a reconstruction of the electronic structure, particularly, near the Fermi level, which is indicated in the resistivity data where a systematic jumplike anomaly is observed in the vicinity of the martensite-austenite phase transformation. With increasing Cr concentration, the magnitude of the jump in resistivity changes dramatically from less than 1 % to nearly 18 % The results are discussed considering the fundamental interactions in Heusler alloys.

  6. Swelling and tensile properties of neutron-irradiated vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.; Smith, D.L.

    1990-07-01

    Vanadium-base alloys are candidates for use as structural material in magnetic fusion reactors. In comparison to other candidate structural materials (e.g., Type 316 stainless and HT-9 ferritic steels), vanadium-base alloys such as V-15Cr-5Ti and V-20Ti have intrinsically lower long-term neutron activation, neutron irradiation after-heat, biological hazard potential, and neutron-induced helium and hydrogen transmutation rates. Moreover, vanadium-base alloys can withstand a higher surface-heat, flux than steels because of their lower thermal stress factor. In addition to having these favorable neutronic and physical properties, a candidate alloy for use as structural material in a fusion reactor must have dimensional stability, i.e., swelling resistance, and resistance to embrittlement during the reactor lifetime at a level of structural strength commensurate with the reactor operating temperature and structural loads. In this paper, we present experimental results on the swelling and tensile properties of several vanadium-base alloys after irradiation at 420, 520, and 600{degree}C to neutron fluences ranging from 0.3 to 1.9 {times} 10{sup 27} neutrons/m{sup 2} (17 to 114 atom displacements per atom (dpa)).

  7. Investigation of austenitic alloys for advanced heat recovery and hot-gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W. [Oak Ridge National Lab., TN (United States)

    1997-12-01

    Materials properties were collected for the design and construction of structural components for use in advanced heat recovery and hot gas cleanup systems. Alloys systems included 9Cr-1Mo-V steel, modified 316 stainless steel, modified type 310 stainless steel, modified 20Cr-25Ni-Nb stainless steel, and modified alloy 800. Experimental work was undertaken to expand the databases for potentially useful alloys. Types of testing included creep, stress-rupture, creep-crack growth, fatigue, and post-exposure short-time tensile tests. Because of the interest in relatively inexpensive alloys for service at 700 C and higher, research emphasis was placed on a modified type 310 stainless steel and a modified 20Cr-25Ni-Nb stainless steel. Both steels were found to have useful strength to 925 C with good weldability and ductility.

  8. Microstructural and Stress Corrosion Cracking Characteristics of Austenitic Stainless Steels Containing Silicon

    Science.gov (United States)

    Andresen, Peter L.; Chou, Peter H.; Morra, Martin M.; Lawrence Nelson, J.; Rebak, Raul B.

    2009-12-01

    Austenitic stainless steels (SSs) core internal components in nuclear light water reactors (LWRs) are susceptible to irradiation-assisted stress corrosion cracking (IASCC). One of the effects of irradiation is the hardening of the SS and a change in the dislocation distribution in the alloy. Irradiation may also alter the local chemistry of the austenitic alloys; for example, silicon may segregate and chromium may deplete at the grain boundaries. The segregation or depletion phenomena at near-grain boundaries may enhance the susceptibility of these alloys to environmentally assisted cracking (EAC). The objective of the present work was to perform laboratory tests in order to better understand the role of Si in the microstructure, properties, electrochemical behavior, and susceptibility to EAC of austenitic SSs. Type 304 SS can dissolve up to 2 pct Si in the bulk while maintaining a single austenite microstructure. Stainless steels containing 12 pct Cr can dissolve up to 5 pct bulk Si while maintaining an austenite structure. The crack growth rate (CGR) results are not conclusive about the effect of the bulk concentration of Si on the EAC behavior of SSs.

  9. Preliminary Analysis of the General Performance and Mechanical Behavior of Irradiated FeCrAl Base Alloys and Weldments

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-30

    The iron-based, iron-chromium-aluminum (FeCrAl) alloys are promising, robust materials for deployment in current and future nuclear power plants. This class of alloys demonstrates excellent performance in a range of environments and conditions, including high-temperature steam (>1000°C). Furthermore, these alloys have the potential to have prolonged survival under loss-of-coolant accident (LOCA) conditions compared to the more traditional cladding materials that are either Zr-based alloys or austenitic steels. However, one of the issues associated with FeCrAl alloys is cracking during welding. The present project investigates the possibility of mitigating welding-induced cracking via alloying and precise structure control of the weldments; in the frame work of the project, several advanced alloys were developed and are being investigated prior to and after neutron irradiation to provide insight into the radiation tolerance and mechanical performance of the weldments. The present report provides preliminary results on the post-irradiation characterization and mechanical tests performed during United States Fiscal Year (FY) 2016. Chapter 1 provides a general introduction, and Chapter 2 describes the alloy compositions, welding procedure, specimen geometry and manufacturing parameters. Also, a brief discussion of the irradiation at the High Flux Isotope Reactor (HFIR) is provided. Chapter 3 is devoted to the analysis of mechanical tests performed at the hot cell facility; tensile curves and mechanical properties are discussed in detail focusing on the irradiation temperature. Limited fractography results are also presented and analyzed. The discussion highlights the limitations of the testing within a hot cell. Chapter 4 underlines the advantages of in-situ testing and discusses the preliminary results obtained with newly developed miniature specimens. Specimens were moved to the Low Activation Materials Development and Analysis (LAMDA) laboratory and prepared for

  10. Cast, heat-resistant austenitic stainless steels having reduced alloying element content

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [Knoxville, TN; Sikka, Vinod Kumar [Oak Ridge, TN; Maziasz, Philip J [Oak Ridge, TN; Pankiw, Roman I [Greensburg, PA

    2010-07-06

    A cast, austenitic steel composed essentially of, expressed in weight percent of the total composition, about 0.4 to about 0.7 C, about 20 to about 30 Cr, about 20 to about 30 Ni, about 0.5 to about 1 Mn, about 0.6 to about 2 Si, about 0.05 to about 1 Nb, about 0.05 to about 1 W, about 0.05 to about 1.0 Mo, balance Fe, the steel being essentially free of Ti and Co, the steel characterized by at least one microstructural component selected from the group consisting of MC, M.sub.23C.sub.6, and M(C, N).

  11. Cast, heat-resistant austenitic stainless steels having reduced alloying element content

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [Knoxville, TN; Sikka, Vinod Kumar [Oak Ridge, TN; Maziasz, Philip J [Oak Ridge, TN; Pankiw, Roman I [Greensburg, PA

    2011-08-23

    A cast, austenitic steel composed essentially of, expressed in weight percent of the total composition, about 0.4 to about 0.7 C, about 20 to about 30 Cr, about 20 to about 30 Ni, about 0.5 to about 1 Mn, about 0.6 to about 2 Si, about 0.05 to about 1 Nb, about 0.05 to about 1 W, about 0.05 to about 1.0 Mo, balance Fe, the steel being essentially free of Ti and Co, the steel characterized by at least one microstructural component selected from the group consisting of MC, M.sub.23C.sub.6, and M(C, N).

  12. Extreme embrittlement of austenitic stainless steel irradiated to 75--81 dpa at 335--360 C

    Energy Technology Data Exchange (ETDEWEB)

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approximately}270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the swelling regime to displacement rate and the large, unprecedented levels of swelling reached at 335--360 C at these high neutron fluences. The failure mechanism appears to be identical to that observed at similar swelling levels in other austenitic steels irradiated in US fast reactors at 400--425 C, whereby stress-concentration between voids and nickel segregation at void surfaces predisposes the steel to an epsilon martensite transformation followed by formation of alpha martensite at crack tips. The very slow strain rate inherent in such creep tests and the relatively high helium levels may also contribute to the failure.

  13. Extreme embrittlement of austenitic stainless steel irradiated to 75-81 dpa at 335-360{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

    1997-04-01

    It is generally accepted that void swelling of austenitic steels ceases below some temperature in the range 340-360{degrees}C, and exhibits relatively low swelling rates up to {approximately}400{degrees}C. This perception may not be correct at all irradiation conditions, however, since it was largely developed from data obtained at relatively high displacement rates in fast reactors whose inlet temperatures were in the range 360-370{degrees}C. There is an expectation, however, that the swelling regime can shift to lower temperatures at low displacement rates via the well-known {open_quotes}temperature shift{close_quotes} phenomenon. It is also known that the swelling rates at the lower end of the swelling regime increase continuously at a sluggish rate, never approaching the terminal 1%/dpa level within the duration of previous experiments. This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0-200 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approx}270{degrees}C. Tubes in the annealed condition reached 75 dpa at 335{degrees}C, and another set in the 20% cold-worked condition reached 81 dpa at 360{degrees}C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes.

  14. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part III: Phase stability during heavy ion irradiation

    Science.gov (United States)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    The phase stability in Fe-15Ni-13Cr alloys was investigated as a function of minor alloying additions after 4 MeV Ni ion irradiation at 948 K. The results showed that the stability of precipitate phases was dictated mainly by the defects produced by radiation damage and preferential segregation of Si and Ni at defects. In addition, radiation enhanced diffusion and cascade induced dissolution and mixing allowed kinetically sluggish phases to form rapidly under irradiation. These radiation effects caused an enhancement, retardation, or modification of thermal phases, and formation of new phases. The relative stability of precipitate phases varied sensitively with alloy composition. The roles of each alloying element on phase stability and the impact of radiation on the mechanisms of phase evolution were systematically studied and documented. The knowledge obtained from this work provides guidelines for designing alloys that lead to develop desired precipitate microstructures under irradiation.

  15. A Comparison of the Corrosion Resistance of Iron-Based Amorphous Metals and Austenitic Alloys in Synthetic Brines at Elevated Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J C

    2008-11-25

    Several hard, corrosion-resistant and neutron-absorbing iron-based amorphous alloys have now been developed that can be applied as thermal spray coatings. These new alloys include relatively high concentrations of Cr, Mo, and W for enhanced corrosion resistance, and substantial B to enable both glass formation and neutron absorption. The corrosion resistances of these novel alloys have been compared to that of several austenitic alloys in a broad range of synthetic brines, with and without nitrate inhibitor, at elevated temperature. Linear polarization and electrochemical impedance spectroscopy have been used for in situ measurement of corrosion rates for prolonged periods of time, while scanning electron microscopy (SEM) and energy dispersive analysis of X-rays (EDAX) have been used for ex situ characterization of samples at the end of tests. The application of these new coatings for the protection of spent nuclear fuel storage systems, equipment in nuclear service, steel-reinforced concrete will be discussed.

  16. Neutron-Induced Microstructural Evolution of Fe-15Cr-16Ni Alloys at ~400 C During Neutron Irradiation in the FFTF Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.; Wolfer, W. G.; Isobe, Yoshihiro

    2001-06-30

    An experiment conducted at ~400 degrees C on simple model austenitic alloys (Fe-15Cr-16Ni and Fe-15Cr-16Ni-0.25Ti, both with and without 500 appm boron) irradiated in the FFTF fast reactor at seven different dpa rates clearly shows that lowering of the atomic displacement rate leads to a pronounced reduction in the transient regime of void swelling. While the steady state swelling rate (~1%/dpa) of these alloys is unaffected by changes in the dpa rate, the transient regime of swelling can vary from <1 to ~60 dpa when the dpa rate varies over more than two orders of magnitude. This range of dpa rates covers the full span of fusion, PWR and fast reactor rates. The origin of the flux sensitivity of swelling arises first in the evolution of the Frank dislocation loop population, its unfaulting, and the subsequent evolution of the dislocation network. There also appears to be some flux sensitivity to the void nucleation process. Most interestingly, the addition of titanium suppresses the void nucleation process somewhat, but does not alter the duration of the transient regime of swelling or its sensitivity to dpa rate. Side-by-side irradiation of boron-modified model alloys in this same experiment shows that higher helium generation rates homogenize the swelling somewhat, but do not significantly change its magnitude or flux sensitivity. The results of this study support the prediction that austenitic alloys irradiated at PWR-relevant displacement rates will most likely swell more than when irradiated at higher rates characteristic of fast reactors. Thus, the use of swelling data accumulated in fast reactors may possibly lead to an under-prediction of swelling in lower-flux PWRs and fusion devices.

  17. Fatigue strain-life behavior of carbon and low-alloy steels, austenitic stainless steels, and Alloy 600 in LWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Keisler, J.; Chopra, O.K.; Shack, W.J. [Argonne National Lab., IL (United States)

    1995-08-01

    The existing fatigue strain vs. life (S-N) data, foreign and domestic, for carbon and low-alloy steels, austenitic stainless steels, and Alloy 600 used in the construction of nuclear power plant components have been compiled and categorized according to material, loading, and environmental conditions. Statistical models have been developed for estimating the effects of the various service conditions on the fatigue life of these materials. The results of a rigorous statistical analysis have been used to estimate the probability of initiating a fatigue crack. Data in the literature were reviewed to evaluate the effects of size, geometry, and surface finish of a component on its fatigue life. The fatigue S-N curves for components have been determined by adjusting the probability distribution curves for smooth test specimens for the effect of mean stress and applying design margins to account for the uncertainties due to component size/geometry and surface finish. The significance of the effect of environment on the current Code design curve and on the proposed interim design curves published in NUREG/CR-5999 is discussed. Estimations of the probability of fatigue cracking in sample components from BWRs and PWRs are presented.

  18. Microstructural examination of irradiated vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Chung, H.M. [Argonne National Lab., IL (United States)

    1997-04-01

    Microstructural examination results are reported for a V-5Cr-5Ti unirradiated control specimens of heat BL-63 following annealing at 1050{degrees}C, and V-4Cr-4Ti heat BL-47 irradiated in three conditions from the DHCE experiment: at 425{degrees}C to 31 dpa and 0.39 appm He/dpa, at 600{degrees}C to 18 dpa and 0.54 appm He/dpa and at 600{degrees}C to 18 dpa and 4.17 appm He/dpa.

  19. CRADA NFE-08-01456 Evaluation of Alumina-Forming Austenitic Stainless Steel Alloys in Industrial Gas Turbines

    Energy Technology Data Exchange (ETDEWEB)

    Brady, Michael P [ORNL; Pint, Bruce A [ORNL; Unocic, Kinga A [ORNL; Yamamoto, Yukinori [ORNL; Kumar, Deepak [ORNL; Lipschutz, Mark D. [Solar Turbines, Inc.

    2011-09-01

    Oak Ridge National Laboratory (ORNL) and Solar Turbines Incorporated (Solar) participated in an in-kind cost share cooperative research and development agreement (CRADA) effort under the auspices of the Energy Efficiency and Renewable Energy (EERE) Technology Maturation Program to explore the feasibility for use of developmental ORNL alumina-forming austenitic (AFA) stainless steels as a material of construction for industrial gas turbine recuperator components. ORNL manufactured lab scale foil of three different AFA alloy compositions and delivered them to Solar for creep properties evaluation. One AFA composition was selected for a commercial trial foil batch. Both lab scale and the commercial trial scale foils were evaluated for oxidation and creep behavior. The AFA foil exhibited a promising combination of properties and is of interest for future scale up activities for turbine recuperators. Some issues were identified in the processing parameters used for the first trial commercial batch. This understanding will be used to guide process optimization of future AFA foil material production.

  20. Effect of alloy grain size on the high-temperature oxidation behavior of the austenitic steel TP 347

    Directory of Open Access Journals (Sweden)

    Vicente Braz Trindade

    2005-12-01

    Full Text Available Generally, oxide scales formed on high Cr steels are multi-layered and the kinetics are strongly influenced by the alloy grain boundaries. In the present study, the oxidation behaviour of an austenite steel TP347 with different grain sizes was studied to identify the role of grain-boundaries in the oxidation process. Heat treatment in an inert gas atmosphere at 1050 °C was applied to modify the grain size of the steel TP347. The mass gain during subsequent oxidation was measured using a microbalance with a resolution of 10-5 g. The scale morphology was examined using SEM in combination with energy-dispersive X-ray spectroscopy (EDS. Oxidation of TP347 with a grain size of 4 µm at 750 °C in air follows a parabolic rate law. For a larger grain size (65 µm, complex kinetics is observed with a fast initial oxidation followed by several different parabolic oxidation stages. SEM examinations indicated that the scale formed on specimens with smaller grain size was predominantly Cr2O3, with some FeCr2O4 at localized sites. For specimens with larger grain size the main oxide is iron oxide. It can be concluded that protective Cr2O3 formation is promoted by a high density of fast grain-boundary diffusion paths which is the case for fine-grained materials.

  1. Preliminary Analysis of the General Performance and Mechanical Behavior of Irradiated FeCrAl Base Alloys and Weldments

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N [ORNL; Field, Kevin G [ORNL; Yamamoto, Yukinori [ORNL

    2016-09-01

    The iron-based, iron-chromium-aluminum (FeCrAl) alloys are promising, robust materials for deployment in current and future nuclear power plants. This class of alloys demonstrates excellent performance in a range of environments and conditions, including high-temperature steam (>1000 C). Furthermore, these alloys have the potential to survive greater durations under lost-of-coolant incident (LOCA) conditions compared to the more traditional cladding materials that are Zr-based or austenitic steels. However, one of the issues associated with FeCrAl alloys is cracking during welding. The present project investigates the possibility to mitigate welding-induced cracking via alloying and precise structure control of the weldments; in the frame work of the project, several advanced alloys were developed and are being investigated prior to and after neutron irradiation to provide insight into the radiation tolerance and mechanical performance of the weldments. The present report provides preliminary, working results on the post-irradiation characterization and mechanical tests performed during United States Fiscal Year (FY) 2016. Chapter 1 provides a general introduction, and Chapter 2 describes the alloy compositions, welding procedure, specimen geometry and manufacturing. Also, a brief discussion of the irradiation at the High-Flux Isotope Reactor (HFIR) is provided. Chapter 3 is devoted to the analysis of mechanical tests performed at the hot cell facility; tensile curves and mechanical properties are discussed in details focusing on the irradiation temperature role. Limited fractography results are also given and analyzed. The discussion highlights the limitations of the testing at the hot cell. Chapter 4 underlines the advantages of in-situ testing and discusses the preliminary results obtained with newly developed miniature specimens. Specimens were moved to the Low Activation Materials Development Laboratory (LAMDA) and prepared for mechanical tests. Follow-on SEM

  2. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Huang, J.S.; Guinan, M.W.; Hahn, P.A.

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

  3. Investigation of austenitic alloys for advanced heat recovery and hot gas cleanup systems

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W.

    1996-06-01

    The objective of the research is to provide databases and design criteria to assist in the selection of optimum alloys for construction of components needed to contain process streams in advanced heat recovery and hot-gas cleanup systems. Typical components include: steam line piping and superheater tubing for low emission boilers (600 to 700{degrees}C), heat exchanger tubing for advanced steam cycles and topping cycle systems (650 to 800{degrees}C), foil materials for recuperators, on advanced turbine systems (700 to 750{degrees}C), and tubesheets for barrier filters, liners for piping, cyclones, and blowback system tubing for hot-gas cleanup systems (850 to 1000{degrees}C). The materials being examined fall into several classes, depending on which of the advanced heat recovery concepts is of concern. These classes include martensitic steels for service to 650{degrees}C, lean stainless steels and modified 25Cr-30Ni steels for service to 700{degrees}C, modified 25Cr-20Ni steels for service to 900{degrees}C, and high Ni-Cr-Fe or Ni-Cr-Co-Fe alloys for service to 1000{degrees}C.

  4. Irradiation creep of various ferritic alloys irradiated at ˜400°C in the PFR and FFTF reactors

    Science.gov (United States)

    Toloczko, M. B.; Garner, F. A.; Eiholzer, C. R.

    1998-10-01

    Irradiation creep of three ferritic alloys at ˜400 ∘C has been studied. Specimens were in the form of pressurized tubes. In a joint US/UK creep study, two identical sets of creep specimens constructed from one heat of HT9 were irradiated in fast reactors, one in the Prototypic Fast Reactor (PFR) and the other in the Fast Flux Test Facility (FFTF). The specimens in PFR were irradiated to a dose of ˜50 dpa, whereas the specimens in FFTF were irradiated to a dose of 165 dpa. The observed swelling and creep behavior were very different in the two reactors. Creep specimens constructed from D57, a developmental alloy ferritic alloy, were also irradiated in PFR to a dose of ˜50 dpa. Creep behavior typical of previous studies on ferritic alloys was observed. Finally, creep specimens constructed from MA957, a Y 2O 3 dispersion-hardened ferritic alloy, were irradiated in FFTF to a dose of ˜110 dpa. This alloy exhibited a large amount of densification, and the creep behavior was different than observed in more conventional ferritic or ferritic-martensitic alloys.

  5. Microstructure and dimensional changes of neutron-irradiated zirconium alloys

    Science.gov (United States)

    Pedraza, A. J.; Fainstein-Pedraza, D.

    1982-08-01

    Experimental observations concerning the neutron-irradiation-induced defect structure in zirconium-based alloys are analyzed within the framework of an irradiation growth theory developed in the past years. The competition of those defects and the microstructure present in the material prior to irradiation as point defect sinks is studied as a function of irradiation temperature and dose. Owing to the different growth behavior of recrystallized and of cold-worked specimens at reactor temperatures, the cellular microstructure of the latter is considered in detail. In view of the highly anisotropic dislocation system in these materials, cell boundaries are reasoned to form with essentially edge components in the walls parallel to the c-axis, while the boundary segments normal to that axis should be of screw type. Since the latter would induce no dimensional change if the cell boundary absorbs defects by dislocation climb, it is argued — on the basis of growth data — that it must behave as a point defect sink/source of a different nature than that of free dislocations. The possibility of dislocation segments with a ( c + a)-type Burgers vector in the cell boundary or rapid point defect diffusion along it are also discussed. The existing growth model is then enlarged in order to account quantitatively for the dimensional changes of cold-worked materials, and its results are compared with available experimental data.

  6. Influence of composition, heat treatment and neutron irradiation on the electrical conductivity of copper alloys

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, B.N.

    1998-01-01

    The electrical conductivity of three different types of copper alloys, viz. CuNiBe, CuCrZr and Cu-Al(2)O(3) as well as of pure copper are reported. The alloys have undergone different pre-irradiation heat treatments and have been fission-neutron irradiated up to 0.3 dpa. In some cases post...

  7. NON-EQUILIBRIUM SOLUTE SEGREGATION TO AUSTENITIC GRAIN BOUNDARY IN FERRUM-NICKLE ALLOY

    Institute of Scientific and Technical Information of China (English)

    P. Wu; D.Y. Yu; X.L. He

    2001-01-01

    The development of non-equilibrium segregation of boron at grain boundaries in Fe-40%Ni alloy during continuous cooling process was experimentally observed with boronParticle Tracking Autoradiography (PTA) and Transmission Electron Microscopy(TEM). The samples with 10ppm boron were cooled at 2℃/s to 1040, 980, 920,860, 780 and 640℃ respectively after pre-heat treatment of 1150℃ for 15min witha Gleeble-1500 heat simulating machine, then water quenched to room temperature.The width of segregation layer and boron depletion zone, rich factor and other pc-rameters were measured by a special image analysis system. The experimental resultsof PTA show that the grain boundary segregation of boron during cooling process is adynamic process and the development of the non-equilibrium segregation experiencesthree stages: first increases rapidly from 1150 to 1040℃, then gently from 1040 to860℃, and rapidly again from 860℃ to 640℃. The width of boron depletion zoneincreases from about 11μm at 1040℃ to 26μm at 640℃. TEM observation showsthat boron precipitates exist at grain boundaries when the samples are cooled to below860℃. The experimental phenomena are briefly discussed.

  8. Hydrogen effects in nitrogen-alloyed austenitic steels; Wirkung von Wasserstoff in stickstofflegierten austenitischen Staehlen

    Energy Technology Data Exchange (ETDEWEB)

    Uhlemann, M.; Mummert, K. [Institut fuer Festkoerper- und Werkstofforschung Dresden e.V. (Germany); Shehata, M.F. [National Research Centre, Cairo (Egypt)

    1998-12-31

    Hydrogen increases the yield strength of nitrogen-alloyed steels, but on the other hand adversely affects properties such as tensile strength and elongation to fracture. The effect is enhanced with increasing nitrogen and hydrogen contents. Under the effect of hydrogen addition, the discontinuous stress-strain characteristic and the distinct elongation limit of hydrogen-free, nitrogen containing steels is no longer observed in the material. This change of mechanical properties is attributed to an interatomic interaction of nitrogen and hydrogen in the lattice, which is shown for instance by such effects as reduction of hydrogen velocity, high solubility, and a particularly strong lattice expansion. The nature of this interaction of nitrogen and hydrogen in the fcc lattice remains to be identified. (orig./CB) [Deutsch] Wasserstoff fuehrt in stickstofflegierten Staehlen zu einer Erhoehung der Streckgrenze, aber gleichzeitig zu einer Abnahme der Zugfestigkeit und Bruchdehnung. Dieser Effekt verstaerkt sich mit zunehmenden Stickstoff- und Wasserstoffgehalten. Ein diskontinuierlicher Spannungs-Dehnungsverlauf mit einer ausgepraegten Streckgrenze in wasserstofffreien hochstickstoffhaltigen Staehlen wird nach Wasserstoffeinfluss nicht mehr beobachtet. Die Aenderung der mechanischen Eigenschaften, wird auf eine interatomare Wechselwirkung von Stickstoff und Wasserstoff im Gitter zurueckgefuehrt, die sich u.a. in geringer Wasserstoffdiffusionsgeschwindigkeit, hoher Loeslichkeit und vor allem in extremer Gitteraufweitung aeussert. Insgesamt ist die Natur der Wechselwirkung zwischen Stickstoff und Wasserstoff im kfz Gitter noch nicht aufgeklaert. (orig.)

  9. Correlation between shear punch and tensile data for neutron-irradiated aluminum alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, M.L.; Edwards, D.J. [Pacific Northwest Laboratory, Richland, WA (United States); Toloczko, M.B. [Univ. of California, Santa Barbara, CA (United States)] [and others

    1995-04-01

    This work was performed to determine whether shear punch and tensile data obtained on neutron irradiated aluminum alloys exhibited the same type of relationship as had been seen in other work and to assess the validity of extrapolating the results to proton-irradiated alloys. This work was also meant to be the first of a series of similar test matrices designed to determine whether the shear punch/tensile relationship varied or was the same for different alloy classes.

  10. Case reviews on the effect of microstructure on the corrosion behavior of austenitic alloys for processing and storage of nuclear waste

    Science.gov (United States)

    Kain, V.; Sengupta, P.; de, P. K.; Banerjee, S.

    2005-05-01

    This article describes the corrosion behavior of special austenitic alloys for waste management applications. The special stainless steels have controlled levels of alloying and impurity elements and inclusion levels. It is shown that “active” inclusions and segregation of chromium along flow lines accelerated IGC of nonsensitized stainless steels. Concentration of Cr+6 ions in the grooves of dissolved inclusions increased the potential to the transpassive region of the material, leading to accelerated attack. It is shown that a combination of cold working and controlled solution annealing resulted in a microstructure that resisted corrosion even after a sensitization heat treatment. This imparted extra resistance to corrosion by increasing the fraction of “random” grain boundaries above a threshold value. Randomization of grain boundaries made the stainless steels resistant to sensitization, IGC, and intergranular stress corrosion cracking (IGSCC) in even hot chloride environments. The increased corrosion resistance has been attributed to connectivity of random grain boundaries. The reaction mechanism between the molten glass and the material for process pot, alloy 690, during the vitrification process has been shown to result in depletion of chromium from the reacting surfaces. A comparison is drawn between the electrochemical behavior of alloys 33 and 22 in 1 M HCl at 65 °C. It is shown that a secondary phase formed during welding of alloy 33 impaired corrosion properties in the HCl environment.

  11. Thermally Nitrided Stainless Steels for Polymer Electrolyte Membrane Fuel Cell Bipolar Plates: Part 1 Model Ni-50Cr and Austenitic 349TM alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Heli [National Renewable Energy Laboratory (NREL); Brady, Michael P [ORNL; Turner, John [National Renewable Energy Laboratory (NREL)

    2004-01-01

    Thermal nitridation of a model Ni-50Cr alloy at 1100 C for 2 h in pure nitrogen resulted in the formation of a continuous, protective CrN/Cr{sub 2}N surface layer with a low interfacial contact resistance. Application of similar nitridation parameters to an austenitic stainless steel, 349{sup TM}, however, resulted in a discontinuous mixture of discrete CrN, Cr{sub 2}N and (Cr,Fe){sub 2}N{sub 1-x} (x = 0--0.5) phase surface particles overlying an exposed {gamma} austenite-based matrix, rather than a continuous nitride surface layer. The interfacial contact resistance of the 349{sup TM} was reduced significantly by the nitridation treatment. However, in the simulated PEMFC environments (1 M H{sub 2}SO{sub 4} + 2 ppm F{sup -} solutions at 70 C sparged with either hydrogen or air), very high corrosion currents were observed under both anodic and cathodic conditions. This poor behavior was linked to the lack of continuity of the Cr-rich nitride surface formed on 349{sup TM} Issues regarding achieving continuous, protective Cr-nitride surface layers on stainless steel alloys are discussed.

  12. Effects of irradiation on chromium's behavior in ferritic/martensitic FeCr alloy

    Institute of Scientific and Technical Information of China (English)

    Xinfu HE; Wen YANG; Zhehao QU; Sheng FAN

    2009-01-01

    The effects of irradiation on chromium performance under different temperatures in Fe-20at%Cr were modeled by modified Marlowe code. Chromium precipitation was observed in FeCr alloy after irradiation; interstitial Chromium atoms are the preferred formation of mixed FeCr dumbbells in the direction ofand; interstitial chromium atoms congregated on {111} and {110} plane. The results are compared with experiment observations and are useful to understanding the irradiation performances of FeCr alloy.

  13. Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

    Energy Technology Data Exchange (ETDEWEB)

    J. Gan; D. Keiser; D. Wachs; B. Miller; T. Allen; M. Kirk; J. Rest

    2009-11-01

    Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up to ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.

  14. Proton irradiation damage of an annealed Alloy 718 beam window

    Science.gov (United States)

    Bach, H. T.; Anderoglu, O.; Saleh, T. A.; Romero, T. J.; Kelsey, C. T.; Olivas, E. R.; Sencer, B. H.; Dickerson, P. O.; Connors, M. A.; John, K. D.; Maloy, S. A.

    2015-04-01

    Mechanical testing and microstructural analysis was performed on an Alloy 718 window that was in use at the Los Alamos Neutron Science Center (LANSCE) Isotope Production Facility (IPF) for approximately 5 years. It was replaced as part of the IPF preventive maintenance program. The window was transported to the Wing 9 hot cells at the Chemical and Metallurgical Research (CMR) LANL facility, visually inspected and 3-mm diameter samples were trepanned from the window for mechanical testing and microstructural analysis. Shear punch testing and optical metallography was performed at the CMR hot cells. The 1-mm diameter shear punch disks were cut into smaller samples to further reduce radiation exposure dose rate using Focus Ion Beam (FIB) and microstructure changes were analyzed using a Transmission Electron Microscopy (TEM). Irradiation doses were determined to be ∼0.2-0.7 dpa (edge) to 11.3 dpa (peak of beam intensity) using autoradiography and MCNPX calculations. The corresponding irradiation temperatures were calculated to be ∼34-120 °C with short excursion to be ∼47-220 °C using ANSYS. Mechanical properties and microstructure analysis results with respect to calculated dpa and temperatures show that significant work hardening occurs but useful ductility still remains. The hardening in the lowest dose region (∼0.2-0.7 dpa) was the highest and attributed to the formation of γ″ precipitates and irradiation defect clusters/bubbles whereas the hardening in the highest dose region (∼11.3 dpa) was lower and attributed mainly to irradiation defect clusters and some thermal annealing.

  15. Void and precipitate structure in ion- and electron-irradiated ferritic alloys

    Science.gov (United States)

    Ohnuki, Soumei; Takahashi, Heishichiro; Takeyama, Taro

    1984-05-01

    Void formation and precipitation were investigated in Fe10Cr and Fe13Cr base alloys by 200 keV C + ion and 1 MeV electron irradiation. The ferritic alloys exhibited significant resistance to void swelling. In FeCr and FeCr-Si alloys, ion-irradiation produced the precipitates of M 23C 6 type. In the FeCrTi alloy, Ti-rich precipitates were formed with high number density on {100} plane. During electron-irradiation Fe-10Cr alloy, complex dislocation loops were produced with high number density, of which Burgers vector was mostly . EDX analysis showed slightly enrichment of chromium on dislocation loops. These results suggested that the stability of type dislocation structure at high dose is an important factor on good swelling resistance in the ferritic alloys, moreover, titanium addition will intensify the stability of the doslocations through the fine precipitation on dislocations.

  16. Charpy impact test results for low activation ferritic alloys irradiated to 30 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Laboratory, Richland, WA (United States)

    1996-04-01

    Miniature specimens of six low activation ferritic alloys have been impact field tested following irradiation at 370{degrees}C to 30 dpa. Comparison of the results with those of control specimens and specimens irradiated to 10 dpa indicates that degradation in the impact behavior appears to have saturated by {approx}10 dpa in at least four of these alloys. The 7.5Cr-2W alloy referred to as GA3X appears most promising for further consideration as a candidate structural material in fusion reactor applications, although the 9Cr-1V alloy may also warrant further investigation.

  17. Kr ion irradiation study of the depleted-uranium alloys

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

    2010-12-01

    Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  18. Phase Transformations in Austenitic 0Cr18Ni10Ti Steel Irradiated with High-Energy Heavy Ions

    CERN Document Server

    Hofmann, A; Semina, V K

    2000-01-01

    Radiation-induced segregation and phase transformations in 0Cr18Ni10Ti steel irradiated with high-energy heavy Ar^{+6} ions at 625^o up to 1 dpa (from 0.01 to 1 dpa) have been studied. It was found that ion irradiation accelerates carbide precipitation and EDX-analysis showed irradiation-induced segregation near grain boundaries.

  19. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors; Caracterisation microstructurale et modelisation du durcissement des aciers austenitiques irradies des structures internes des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C

    2003-07-01

    The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

  20. Evaluation of High-Temperature Tensile Property of Diffusion Bond of Austenitic Alloys for S-CO{sub 2} Cycle Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sunghoon; Sah, Injin; Jang, Chanheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-12-15

    To improve the inherent safety of the sodium-cooled fast reactor (SFR), the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle is being considered as an alternative power conversion system to steam the Rankine cycle. In the S-CO{sub 2} system, a PCHE (printed circuit heat exchanger) is being considered. In this type of heat exchangers, diffusion bonding is used for joining the thin plates. In this study, the diffusion bonding characteristics of various austenitic alloys were evaluated. The tensile properties were measured at temperatures starting from the room temperature up to 650℃. For the 316H and 347H types of stainless steel, the tensile ductility was well maintained up to 550℃. However, the Incoloy 800HT showed lower strength and ductility at all temperatures. The microstructure near the bond line was examined to understand the reason for the loss of ductility at high temperatures.

  1. Surface modification of Zr-based bulk amorphous alloys by using ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, M., E-mail: miqbalchishti@yahoo.com [Physics Division, Pakistan Institute of Nuclear Science and Technology (PINSTECH), P.O. Nilore, Islamabad, 45650 (Pakistan); Qayyum, A.; Akhter, J.I. [Physics Division, Pakistan Institute of Nuclear Science and Technology (PINSTECH), P.O. Nilore, Islamabad, 45650 (Pakistan)

    2011-02-10

    Research highlights: > Ion irradiations of two multicomponent bulk amorphous alloys have been done. Ion irradiation produced crystalline phases in the amorphous matrix due to which mechanical properties enhanced. Considerable increase in hardness and elastic modulus was observed. The results are verified, authentic and confirmed. - Abstract: Surfaces of the [Zr{sub 0.65}Cu{sub 0.18}Ni{sub 0.09}Al{sub 0.08}]{sub 98}M{sub 2} (M = Er and Gd) bulk amorphous alloys were modified by irradiation with energetic singly charged argon (Ar{sup +}) ions. Samples of both the alloys were irradiated with 2.17 x 10{sup 17} argon ions of 10 keV energy. As cast and ion irradiated samples were characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). Mechanical properties like Vicker's hardness, nanohardness, elastic modulus and elastic recovery were measured. Considerable increase in elastic modulus and hardness was observed because of ion irradiation in these alloys. The ion irradiated samples of the [Zr{sub 0.65}Cu{sub 0.18}Ni{sub 0.09}Al{sub 0.08}]{sub 98}Er{sub 2} alloy showed better properties as compared to [Zr{sub 0.65}Cu{sub 0.18}Ni{sub 0.09}Al{sub 0.08}]{sub 98}Gd{sub 2} alloy. CuZr{sub 2} phase was detected in ion irradiated alloys by XRD and confirmed by EDS. The range of Ar{sup +} ions was found to be approximately 9.3 {+-} 5.4 nm in both alloys.

  2. Effect of fission neutron irradiation on the tensile and electrical properties of copper and copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Fabritsiev, S.A. [D.V. Efremov Institute, St. Petersburg (Russian Federation); Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1995-04-01

    The objective of this study is to evaluate the properties of several copper alloys following fission reactor irradiation at ITER-relevant temperatures of 80 to 200{degrees}C. This study provides some of the data needed for the ITER research and development Task T213. These low temperature irradiations caused significant radiation hardening and a dramatic decrease in the work hardening ability of copper and copper alloys. The uniform elongation was higher at 200{degree}C compared to 100{degree}C, but still remained below 1% for most of the copper alloys.

  3. Ion irradiation induced disappearance of dislocations in a nickel-based alloy

    Energy Technology Data Exchange (ETDEWEB)

    Chen, H.C.; Li, D.H.; Lui, R.D.; Huang, H.F.; Li, J.J.; Lei, G.H.; Huang, Q.; Bao, L.M.; Yan, L., E-mail: yanlong@sinap.ac.cn; Zhou, X.T., E-mail: zhouxingtai@sinap.ac.cn; Zhu, Z.Y.

    2016-06-15

    Under Xe ion irradiation, the microstructural evolution of a nickel based alloy, Hastelloy N (US N10003), was studied. The intrinsic dislocations are decorated with irradiation induced interstitial loops and/or clusters. Moreover, the intrinsic dislocations density reduces as the irradiation damage increases. The disappearance of the intrinsic dislocations is ascribed to the dislocations climb to the free surface by the absorption of interstitials under the ion irradiation. Moreover, the in situ annealing experiment reveals that the small interstitial loops and/or clusters induced by the ion irradiation are stable below 600 °C.

  4. Embrittlement behaviour of different international low activation alloys after neutron irradiation

    Science.gov (United States)

    Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2001-05-01

    The embrittlement behaviour of ferritic/martensitic steels after irradiation in the Petten high flux reactor (HFR) was investigated by instrumented Charpy-V tests with subsize specimens. The main objective, apart from studying effects of particularly low doses, was a comparison of low activation alloys (LAA) from various countries with different Cr contents and different types and concentrations of minor alloying elements and impurities. In the present report, the results of another three materials (OPTIMAR, OPTIFER-IV, GA3X) obtained within the second phase of the MANITU programme (0.8 dpa, at 250-450°C) were analysed and assessed in comparison to the results of the first irradiation up to 0.8 dpa. The evaluation clearly showed a reduced embrittlement problem for the advanced reduced-activation alloys. Of the examined alloys, the GA3X steel shows the very best embrittlement behaviour after neutron irradiation.

  5. Evolution of precipitate in nickel-base alloy 718 irradiated with argon ions at elevated temperature

    Science.gov (United States)

    Jin, Shuoxue; Luo, Fengfeng; Ma, Shuli; Chen, Jihong; Li, Tiecheng; Tang, Rui; Guo, Liping

    2013-07-01

    Alloy 718 is a nickel-base superalloy whose strength derives from γ'(Ni3(Al,Ti)) and γ″(Ni3Nb) precipitates. The evolution of the precipitates in alloy 718 irradiated with argon ions at elevated temperature were examined via transmission electron microscopy. Selected-area electron diffraction indicated superlattice spots disappeared after argon ion irradiation, which showing that the ordered structure of the γ' and γ″ precipitates became disordered. The size of the precipitates became smaller with the irradiation dose increasing at 290 °C.

  6. Irradiation effects in oxide dispersion strengthened (ODS) Ni-base alloys for Gen. IV nuclear reactors

    Science.gov (United States)

    Oono, Naoko; Ukai, Shigeharu; Kondo, Sosuke; Hashitomi, Okinobu; Kimura, Akihiko

    2015-10-01

    Oxide particle dispersion strengthened (ODS) Ni-base alloys are irradiated by using simulation technique (Fe/He dual-ion irradiation) to investigate the reliability to Gen. IV high-temperature reactors. The fine oxide particles with less than 10 nm in average size and approximately 8.0 × 1022 m-3 in number density remained after 101 dpa irradiation. The tiny helium bubbles were inside grains, not at grain-boundaries; it is advantageous effect of oxide particles which trap the helium atoms at the particle-matrix interface. Ni-base ODS alloys demonstrated their great ability to overcome He embrittlement.

  7. Cast heat-resistant austenitic steel with improved temperature creep properties and balanced alloying element additions and methodology for development of the same

    Energy Technology Data Exchange (ETDEWEB)

    Pankiw, Roman I; Muralidharan, Govindrarajan; Sikka, Vinod Kumar; Maziasz, Philip J

    2012-11-27

    The present invention addresses the need for new austenitic steel compositions with higher creep strength and higher upper temperatures. The new austenitic steel compositions retain desirable phases, such as austenite, M.sub.23C.sub.6, and MC in its microstructure to higher temperatures. The present invention also discloses a methodology for the development of new austenitic steel compositions with higher creep strength and higher upper temperatures.

  8. Neutron diffraction analysis of Cr-Ni-Mo-Ti austenitic steel after cold plastic deformation and fast neutrons irradiation

    Science.gov (United States)

    Voronin, V. I.; Valiev, E. Z.; Berger, I. F.; Goschitskii, B. N.; Proskurnina, N. V.; Sagaradze, V. V.; Kataeva, N. F.

    2015-04-01

    A quantitative assessment is presented of the dislocation density and relative fractions of edge and screw dislocations in reactor-steel samples 16Cr-15Ni-3Mo-1Ti subjected to preliminary cold deformation by rolling and subsequent fast neutron irradiation using neutron diffraction analysis. The Williamson-Hall modified method was used for calculations. It is shown that the fast neutron irradiation leads to a decrease in the density of dislocations that appeared after samples deformation. The applicability of neutron diffraction analysis to the examination of dislocation structure of deformed and irradiated materials is shown.

  9. Phase Stability under Irradiation of Precipitates and Solid Solutions in Model ALloys and in ODS Alloys Relevant for Gen IV

    Energy Technology Data Exchange (ETDEWEB)

    Arthur T. Motta; Robert C. Birtcher

    2007-10-17

    The overall objective of this program is to investigate the irradiation-altered phase stability of oxide precipitates in ODS steels and of model alloy solid solutions of associated systems. This information can be used to determine whether the favorable mechanical propertiies of these steels are maintained under irradiation, thus addressing one of the main materials research issues for this class of steels as identified by the GenIV working groups. The research program will also create fundamental understanding of the irradiation precipitation/dissolution problem by studying a "model" system in which the variables can be controlled and their effects understood individually.

  10. The Influence of Austenite Grain Size on the Mechanical Properties of Low-Alloy Steel with Boron

    Directory of Open Access Journals (Sweden)

    Beata Białobrzeska

    2017-01-01

    Full Text Available This study forms part of the current research on modern steel groups with higher resistance to abrasive wear. In order to reduce the intensity of wear processes, and also to minimize their impact, the immediate priority seems to be a search for a correlation between the chemical composition and structure of these materials and their properties. In this paper, the correlation between prior austenite grain size, martensite packets and the mechanical properties were researched. The growth of austenite grains is an important factor in the analysis of the microstructure, as the grain size has an effect on the kinetics of phase transformation. The microstructure, however, is closely related to the mechanical properties of the material such as yield strength, tensile strength, elongation and impact strength, as well as morphology of occurred fracture. During the study, the mechanical properties were tested and a tendency to brittle fracture was analysed. The studies show big differences of the analysed parameters depending on the applied heat treatment, which should provide guidance to users to specific applications of this type of steel.

  11. Numerical Modeling of Growth Kinetics of Pro-eutectoid Ferrite Transformed from Austenite in Fe-C-Σ X Alloys

    Institute of Scientific and Technical Information of China (English)

    Zhenyu LIU; Guodong WANG; Toshio Narita

    2005-01-01

    In the present paper, a numerical modeling was developed to simulate the growth kinetics of ferrite transformed from austenite in Fe-C-∑X (X denotes substitution elements, such as Mn, Ni, Cr etc.) steels by solving the diffusion equation using finite difference method (FDM). Coupled with the kinetic modeling, thermodynamic calculations were carried out to determine the γ/α phase equilibrium conditions using a para-equilibrium (PE) model. The dissipation of free energy for γ→α phase transformation due to the so-called solute drag effect (SDE) was taken into account in the thermodynamic modeling. With this modeling, simulations on the growth kinetics of ferrite in the steels containing austenite-stabilizing and ferrite-stabilizing elements (such as Ni, Mn and Si, Cr, respectively) were performed, which indicates that it deviates from the parabolic growth rate law after the initial stage of transformation. The results were compared with the experimental values given by Bradley and Aaronson, showing that this model has a reasonably good accuracy to predict the growth kinetics of ferrite.

  12. Effect of the carbide phase on the tribological properties of high-manganese antiferromagnetic austenitic steels alloyed with vanadium and molybdenum

    Science.gov (United States)

    Korshunov, L. G.; Kositsina, I. I.; Sagaradze, V. V.; Chernenko, N. L.

    2011-07-01

    Effect of special carbides (VC, M 6C, Mo2C) on the wear resistance and friction coefficient of austenitic stable ( M s below -196°C) antiferromagnetic ( T N = 40-60°C) steels 80G20F2, 80G20M2, and 80G20F2M2 has been studied. The structure and the effective strength (microhardness H surf, shear resistance τ) of the surface layer of these steels have been studied using optical and electron microscopy. It has been shown that the presence of coarse particles of primary special carbides in the steels 80G20F2, 80G20M2, and 80G20F2M2 quenched from 1150°C decreases the effective strength and the resistance to adhesive and abrasive wear of these materials. This is caused by the negative effect of carbide particles on the toughness of steels and by a decrease in the carbon content in austenite due to a partial binding of carbon into the above-mentioned carbides. The aging of quenched steels under conditions providing the maximum hardness (650°C for 10 h) exerts a substantial positive effect on the parameters of the effective strength ( H surf, τ) of the surface layer and, correspondingly, on the resistance of steels to various types of wear (abrasive, adhesive, and caused by the boundary friction). The maximum positive effect of aging on the wear resistance is observed upon adhesive wear of the steels under consideration. Upon friction with enhanced sliding velocities (to 4 m/s) under conditions of intense (to 500-600°C) friction-induced heating, the 80G20F2, 80G20M2, and, especially, 80G20F2M2 steels subjected to quenching and aging substantially exceed the 110G13 (Hadfield) steel in their tribological properties. This is due to the presence in these steels of a favorable combination of high effective strength and friction heat resistance of the surface layer, which result from the presence of a large amount of special carbides in these steels and from a high degree of alloying of the matrix of these steels by vanadium and molybdenum. In the process of friction

  13. Observations of defect structure evolution in proton and Ni ion irradiated Ni-Cr binary alloys

    Science.gov (United States)

    Briggs, Samuel A.; Barr, Christopher M.; Pakarinen, Janne; Mamivand, Mahmood; Hattar, Khalid; Morgan, Dane D.; Taheri, Mitra; Sridharan, Kumar

    2016-10-01

    Two binary Ni-Cr model alloys with 5 wt% Cr and 18 wt% Cr were irradiated using 2 MeV protons at 400 and 500 °C and 20 MeV Ni4+ ions at 500 °C to investigate microstructural evolution as a function of composition, irradiation temperature, and irradiating ion species. Transmission electron microscopy (TEM) was applied to study irradiation-induced void and faulted Frank loops microstructures. Irradiations at 500 °C were shown to generate decreased densities of larger defects, likely due to increased barriers to defect nucleation as compared to 400 °C irradiations. Heavy ion irradiation resulted in a larger density of smaller voids when compared to proton irradiations, indicating in-cascade clustering of point defects. Cluster dynamics simulations were in good agreement with the experimental findings, suggesting that increases in Cr content lead to an increase in interstitial binding energy, leading to higher densities of smaller dislocation loops in the Ni-18Cr alloy as compared to the Ni-5Cr alloy.

  14. Subtask 12F1: Effect of neutron irradiation on swelling of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Loomis, B.A.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of this work is to determine the effects of neutron irradiation on the density change, void distribution, and microstructural evolution of vanadium-base alloys. Swelling behavior and microstructural evolution of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were investigated after irradiation at 420-600{degrees}C up to 114 dpa. The alloys exhibited swelling maxima between 30 and 80 dpa and swelling decreased on irradiation to higher dpa. This is in contrast to the monotonically increasing swelling of binary alloys that contain Fe, Ni, Cr, Mo, W, and Si. Precipitation of dense Ti{sub 5}Si{sub 3} promotes good resistance to swelling of the Ti-containing alloys, and it was concluded that Ti of >3 wt.% and 400-1000 wppm Si are necessary to effectively suppress swelling. Swelling was minimal in V-4Cr-4Ti, identified as the most promising alloy based on good mechanical properties and superior resistance to irradiation embrittlement. 18 refs., 6 figs., 1 tab.

  15. Subtask 12F3: Effects of neutron irradiation on tensile properties of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of this work is to determine the effects of neutron irradiation on the tensile properties of candidate vanadium-base alloys. Vanadium-base alloys of the V-Cr-Ti system are attractive candidates for use as structural materials in fusion reactors. The current focus of the U.S. program of research on these alloys is on the V-(4-6)Cr-(3-6)Ti-(0.05-0.1)Si (in wt.%) alloys. In this paper, we present experimental results on the effects of neutron irradiation on tensile properties of selected candidate alloys after irradiation at 400{degrees}C-600{degrees}C in lithium in fast fission reactors to displacement damages of up to {approx}120 displacement per atom (dpa). Effects of irradiation temperature and dose on yield and ultimate tensile strengths and uniform and total elongations are given for tensile test temperatures of 25{degrees}C, 420{degrees}C, 500{degrees}, and 600{degrees}C. Effects of neutron damage on tensile properties of the U.S. reference alloy V-4Cr-4Ti are examined in detail. 7 refs., 10 figs., 1 tab.

  16. Improvement of the Corrosion Resistance of High Alloyed Austenitic Cr-Ni-Mo Stainless Steels by Solution Nitriding

    Institute of Scientific and Technical Information of China (English)

    Christine Eckstein; Heinz- Joachim Spies; Jochen Albrecht

    2004-01-01

    Characteristic features of austenitic steel grades combine a good corrosion resistance with a low hardness, wear resistance and scratch resistance. An interesting possibility for improving the wear behaviour of these steels without loss of their corrosion resistance lies in enriching the near surface region with nitrogen. The process of a solution nitriding allows the rise of the solution of nitrogen in the solid phase. On this state nitrogen increases the corrosion resistance and the tribilogical load-bearing capacity. The aim of the study was, to investigate the improvement of the pitting corrosion behaviour by solution nitriding. A special topic was to observe the effect of nitrogen by different molybdenum content. So austenitic stainless steels (18% Cr, 12% Ni, Mo gradation between 0.06 to 3.6%) had been solution nitrided. The samples could be prepared with various surface content of nitrogen from 0.04 to 0.45% with a step-by-step grinding. The susceptibility against pitting corrosion of these samples had been tested by determination of the stable pitting potential in 0.5M and 1M NaCl at 25℃. For the investigated steel composition and the used corrosion system there is no influence of molybdenum on the effectiveness of nitrogen. The influence of nitrogen to all of the determined parameters can be corrosion tests. Additionally surface investigations with an acid elektolyte (0,1M HCl + 0,4M NaCI) were performed. In this case the passivation effective nitrogen content increases markedly with rising molybdenum concentration of the steel.Obviously an interaction of Mo and N is connected with a strongly acid electrolyte.

  17. Microstructural evolution of austenitic stainless steels irradiated to 17 dpa in spectrally tailored experiment of the ORR and HFIR at 400{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Wakai, E.; Hashimoto, N.; Gibson, L.T. [Oak Ridge National Lab., TN (United States)] [and others

    1997-08-01

    The microstructural evolution of austenitic JPCA aged and solution annealed JPCA, 316R, C, K, and HP steels irradiated at 400{degrees}C in spectrally tailored experiments of the ORR and HFIR has been investigated. The helium generation rates were about 12-16 appm He/dpa on the average up to 17.3 dpa. The number densities and average diameters of dislocation loops in the steels have ranges of 3.3 x 10{sup 21} m{sup -3} and 15.2-26.3 nm, respectively, except for HP steel for which they are 1.1 x 10{sup 23} m{sup -3} and 8.0 nm. Precipitates are formed in all steels except for HP steel, and the number densities and average diameters have ranges of 5.2 x 10{sup 20} - 7.7 x 10{sup 21} m{sup -3} and 3.4- 19.3 nm, respectively. In the 216R, C, and K steels, the precipitates are also formed at grain boundaries, and the mean sizes of these are about 110, 50, and 50 nm, respectively. The number densities of cavities are about 1 x 10{sup 22} m{sup -3} in all the steels. The swelling is low in the steels which form the precipitates.

  18. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  19. Precipitation behavior of AlxCoCrFeNi high entropy alloys under ion irradiation

    Science.gov (United States)

    Yang, Tengfei; Xia, Songqin; Liu, Shi; Wang, Chenxu; Liu, Shaoshuai; Fang, Yuan; Zhang, Yong; Xue, Jianming; Yan, Sha; Wang, Yugang

    2016-01-01

    Materials performance is central to the satisfactory operation of current and future nuclear energy systems due to the severe irradiation environment in reactors. Searching for structural materials with excellent irradiation tolerance is crucial for developing the next generation nuclear reactors. Here, we report the irradiation responses of a novel multi-component alloy system, high entropy alloy (HEA) AlxCoCrFeNi (x = 0.1, 0.75 and 1.5), focusing on their precipitation behavior. It is found that the single phase system, Al0.1CoCrFeNi, exhibits a great phase stability against ion irradiation. No precipitate is observed even at the highest fluence. In contrast, numerous coherent precipitates are present in both multi-phase HEAs. Based on the irradiation-induced/enhanced precipitation theory, the excellent structural stability against precipitation of Al0.1CoCrFeNi is attributed to the high configurational entropy and low atomic diffusion, which reduces the thermodynamic driving force and kinetically restrains the formation of precipitate, respectively. For the multiphase HEAs, the phase separations and formation of ordered phases reduce the system configurational entropy, resulting in the similar precipitation behavior with corresponding binary or ternary conventional alloys. This study demonstrates the structural stability of single-phase HEAs under irradiation and provides important implications for searching for HEAs with higher irradiation tolerance. PMID:27562023

  20. Precipitation behavior of AlxCoCrFeNi high entropy alloys under ion irradiation

    Science.gov (United States)

    Yang, Tengfei; Xia, Songqin; Liu, Shi; Wang, Chenxu; Liu, Shaoshuai; Fang, Yuan; Zhang, Yong; Xue, Jianming; Yan, Sha; Wang, Yugang

    2016-08-01

    Materials performance is central to the satisfactory operation of current and future nuclear energy systems due to the severe irradiation environment in reactors. Searching for structural materials with excellent irradiation tolerance is crucial for developing the next generation nuclear reactors. Here, we report the irradiation responses of a novel multi-component alloy system, high entropy alloy (HEA) AlxCoCrFeNi (x = 0.1, 0.75 and 1.5), focusing on their precipitation behavior. It is found that the single phase system, Al0.1CoCrFeNi, exhibits a great phase stability against ion irradiation. No precipitate is observed even at the highest fluence. In contrast, numerous coherent precipitates are present in both multi-phase HEAs. Based on the irradiation-induced/enhanced precipitation theory, the excellent structural stability against precipitation of Al0.1CoCrFeNi is attributed to the high configurational entropy and low atomic diffusion, which reduces the thermodynamic driving force and kinetically restrains the formation of precipitate, respectively. For the multiphase HEAs, the phase separations and formation of ordered phases reduce the system configurational entropy, resulting in the similar precipitation behavior with corresponding binary or ternary conventional alloys. This study demonstrates the structural stability of single-phase HEAs under irradiation and provides important implications for searching for HEAs with higher irradiation tolerance.

  1. Mechanical and microstructural properties of neutron irradiated Fe-Cr-C alloys

    Energy Technology Data Exchange (ETDEWEB)

    Konstantinovic, M.J.; Renterghem, W. van; Matijasevic, M.; Minov, B.; Lambrecht, M.; Chiapetto, M.; Malerba, L. [Studiecentrum voor Kernenergie/Centre d' Etude de l' Energie Nucleaire (SCK-CEN), Mol (Belgium); Toyama, T. [Institute for Materials Research, Tohoku University, Sendai (Japan)

    2016-11-15

    Defect properties of neutron irradiated Fe-Cr-C alloys and their influence on the mechanical behavior are studied by combining mechanical tests, microstructural examination, and the results of models. It is found that the initial microstructure of these alloys, determined by the Cr and C concentrations, as well as by the thermal treatment, can account for different defect formation and distribution after neutron irradiation. On the basis of these results, a correlation between defect properties and macroscopic mechanical behavior is proposed. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  2. The strong influence of displacement rate on void swelling in variants of Fe-16Cr-15Ni-3Mo austenitic stainless steel irradiated in BN-350 and BOR-60

    Energy Technology Data Exchange (ETDEWEB)

    Budylkin, N.I.; Bulanova, T.M.; Mironova, E.G.; Mitrofanova, N.M.; Porollo, S.I.; Chernov, V.M.; Shamardin, V.K.; Garner, F.A. E-mail: frank.garner@pnl.gov

    2004-08-01

    Recent irradiation experiments conducted on a variety of austenitic stainless steels have shown that void swelling appears to be increased when the dpa rate is decreased, primarily by a shortening of the transient regime of swelling. This paper presents results derived from nominally similar irradiations conducted on six Russian steels, all laboratory heat variants of Fe-16Cr-15Ni-3Mo-Nb-B, with each irradiated in two fast reactors, BOR-60 and BN-350. The BN-350 irradiation proceeded at a dpa rate three times higher than that conducted in BOR-60. In all six steels, a significantly higher swelling level was attained in BOR-60, agreeing with the results of earlier studies.

  3. Effects of irradiation to 4 dpa at 390 C on the fracture toughness of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gruber, E.E.; Galvin, T.M.; Chopra, O.K. [Argonne National Lab., IL (United States)

    1998-09-01

    Fracture toughness J-R curve tests were conducted at room temperature on disk-shaped compact-tension DC(T) specimens of three vanadium alloys having a nominal composition of V-4Cr-4Ti. The alloys in the nonirradiated condition showed high fracture toughness; J{sub IC} could not be determined but is expected to be above 600 kJ/m{sup 2}. The alloys showed very poor fracture toughness after irradiation to 4 dpa at 390 C, e.g., J{sub IC} values of {approx}10 kJ/m{sup 2} or lower.

  4. Various categories of defects after surface alloying induced by high current pulsed electron beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Dian [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Tang, Guangze, E-mail: oaktang@hit.edu.cn [School of Material Science & Engineering, Harbin Institute of Technology, Harbin 150001 (China); Ma, Xinxin [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Gu, Le [School of Mechatronics Engineering, Harbin Institute of Technology, Harbin 150001 (China); Sun, Mingren [School of Material Science & Engineering, Harbin Institute of Technology, Harbin 150001 (China); Wang, Liqin [School of Mechatronics Engineering, Harbin Institute of Technology, Harbin 150001 (China)

    2015-10-01

    Highlights: • Four kinds of defects are found during surface alloying by high current electron beam. • Exploring the mechanism how these defects appear after irradiation. • Increasing pulsing cycles will help to get good surface quality. • Choosing proper energy density will increase surface quality. - Abstract: High current pulsed electron beam (HCPEB) is an attractive advanced materials processing method which could highly increase the mechanical properties and corrosion resistance. However, how to eliminate different kinds of defects during irradiation by HCPEB especially in condition of adding new elements is a challenging task. In the present research, the titanium and TaNb-TiW composite films was deposited on the carburizing steel (SAE9310 steel) by DC magnetron sputtering before irradiation. The process of surface alloying was induced by HCPEB with pulse duration of 2.5 μs and energy density ranging from 3 to 9 J/cm{sup 2}. Investigation of the microstructure indicated that there were several forms of defects after irradiation, such as surface unwetting, surface eruption, micro-cracks and layering. How the defects formed was explained by the results of electron microscopy and energy dispersive spectroscopy. The results also revealed that proper energy density (∼6 J/cm{sup 2}) and multi-number of irradiation (≥50 times) contributed to high quality of alloyed layers after irradiation.

  5. Processing of Refractory Metal Alloys for JOYO Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    RF Luther; ME Petrichek

    2006-02-21

    This is a summary of the refractory metal processing experienced by candidate Prometheus materiats as they were fabricated into specimens destined for testing within the JOYO test reactor, ex-reactor testing at Oak Ridge National Laboratory (ORNL), or testing within the NRPCT. The processing is described for each alloy from the point of inception to the point where processing was terminated due to the cancellation of Naval Reactor's involvement in the Prometheus Project. The alloys included three tantalum-base alloys (T-111, Ta-10W, and ASTAR-811C), a niobium-base alloy, (FS-85), and two molybdenum-rhenium alloys, one containing 44.5 w/o rhenium, and the other 47.5 w/o rhenium. Each of these alloys was either a primary candidate or back-up candidate for cladding and structural applications within the space reactor. Their production was intended to serve as a forerunner for large scale production ingots that were to be procured from commercial refractory metal vendors such as Wah Chang.

  6. The effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 100 C

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M. [Risoe National Lab., Roskilde (Denmark)

    1998-03-01

    This report describes the final irradiation experiment in a series of screening experiments aimed at investigating the effects of bonding and bakeout thermal cycles on irradiated copper alloys. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment. The post-irradiation tests at 100 C revealed the greatest loss of ductility occurred in the CuCrZr alloys, irrespective of the pre-irradiation heat treatment, with the uniform elongation dropping to levels of less than 1.5%. The yield and ultimate strengths for all of the individual heat treated samples increased substantially after irradiation. The same trend was observed for the CuNiBe alloys, which overall exhibited a factor of 3 higher uniform elongation after irradiation with almost double the strength. In both alloys irradiation-induced precipitation lead to a large increase in the strength of the solution annealed specimens with a noticeable decrease in uniform elongation. The Al25 alloy also experienced an increase in the overall strength of the alloy after irradiation, accompanied by approximately a 50% decrease in the uniform and total elongation. The additional bakeout treatments given to the CuCrZr and CuNiBe before irradiation served to increase the strength, but in terms of the ductility no improvement or degradation resulted from the additional thermal exposure. The results of this experiment confirm that the al25 possesses the most resistant microstructure to thermal and irradiation-induced changes, while the competing effects of ballistic dissolution and reprecipitation lead to important changes in the two precipitation strengthened alloys. This study and others have repeatedly shown that these materials can only be used if the very low uniform elongation (1% or less) can be accounted for in the design since pre-irradiation thermal processing cannot mitigate the irradiation embrittlement.

  7. Development of a robust modeling tool for radiation-induced segregation in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Allen, Todd R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Busby, Jeremy T [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels in Light Water Reactor (LWR) components has been linked to changes in grain boundary composition due to irradiation induced segregation (RIS). This work developed a robust RIS modeling tool to account for thermodynamics and kinetics of the atom and defect transportation under combined thermal and radiation conditions. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. Both cross and non-cross phenomenological diffusion coefficients in the flux equations were considered and correlated to tracer diffusion coefficients through Manning’s relation. The preferential atomvacancy coupling was described by the mobility model, whereas the preferential atom-interstitial coupling was described by the interstitial binding model. The composition dependence of the thermodynamic factor was modeled using the CALPHAD approach. Detailed analysis on the diffusion fluxes near and at grain boundaries of irradiated austenitic stainless steels suggested the dominant diffusion mechanism for chromium and iron is via vacancy, while that for nickel can swing from the vacancy to the interstitial dominant mechanism. The diffusion flux in the vicinity of a grain boundary was found to be greatly influenced by the composition gradient formed from the transient state, leading to the oscillatory behavior of alloy compositions in this region. This work confirms that both vacancy and interstitial diffusion, and segregation itself, have important roles in determining the microchemistry of Fe, Cr, and Ni at irradiated grain boundaries in austenitic stainless steels.

  8. Effects of energetic ion irradiation on the magnetism of Fe–Ni Invar alloy

    Energy Technology Data Exchange (ETDEWEB)

    Matsushita, M., E-mail: matsushita@eng.ehime-u.ac.jp [Graduate School of Science and Engineering, Ehime University, 3-Bunkyocho, Matsuyama (Japan); Akamatsu, S. [Graduate School of Science and Engineering, Ehime University, 3-Bunkyocho, Matsuyama (Japan); Matsushima, Y. [Graduate School of Natural Science and Technology, Okayama University, Tsushima-naka Kitaku, Okayama (Japan); Iwase, A. [Department of Materials Science, Osaka Prefecture University, Gakuencho, Sakai (Japan)

    2013-11-01

    Highlights: •16-MeV Au{sup 3+} ions were irradiated to Fe{sub 66}Ni{sub 34} alloy. •Magnetic properties of Fe{sub 66}Ni{sub 34} were changed by the irradiation. •T{sub c} of a part of sample increases due to the irradiation. •FCC structure is stable before and after irradiation. -- Abstract: The magnetic properties of Fe–Ni Invar alloys are significantly affected by ion irradiation. Au{sup 3+} with the energy of 16 MeV irradiation effects on the magnetism of Fe{sub 66}Ni{sub 34} have been reported in this paper. Considering from the temperature variations of AC susceptibility of irradiated Fe{sub 66}Ni{sub 34}, Curie temperature of a part of sample increase with increasing incident ion fluence, and the magnetization of irradiated Fe{sub 66}Ni{sub 34} is also increase. The FCC structure of Fe{sub 66}Ni{sub 34} is not changed by ion irradiation; however peaks become broader with increasing ion fluence. It means that lattice fluctuations are generated owing to ion irradiation. However it cannot be considered that lattice fluctuations observed X-ray diffraction measurements are enough to increase the Curie temperature observed in AC susceptibility measurements. Then, we suggest as the considerable origin of increasing T{sub C}, atomic mixing effects owing to the ion irradiation. It might change the chemical ordering reported in the diffused scattering, such as Fe–Fe coupling.

  9. Study of austenitic stainless steel welded with low alloy steel filler metal. [tensile and impact strength tests

    Science.gov (United States)

    Burns, F. A.; Dyke, R. A., Jr.

    1979-01-01

    The tensile and impact strength properties of 316L stainless steel plate welded with low alloy steel filler metal were determined. Tests were conducted at room temperature and -100 F on standard test specimens machined from as-welded panels of various chemical compositions. No significant differences were found as the result of variations in percentage chemical composition on the impact and tensile test results. The weldments containing lower chromium and nickel as the result of dilution of parent metal from the use of the low alloy steel filler metal corroded more severely in a marine environment. The use of a protective finish, i.e., a nitrile-based paint containing aluminum powder, prevented the corrosive attack.

  10. EFFECTS OF IRRADIATION ON THERMAL CONDUCTIVITY OF ALLOY 690 AT LOW NEUTRON FLUENCE

    Directory of Open Access Journals (Sweden)

    WOO SEOG RYU

    2013-04-01

    Full Text Available Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7−28 × 1019n/cm2 (E>0.1MeV at 250°C, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at 250°C did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1% at fluences of (0.7∼28 × 1019n/cm2 (E>0.1MeV.

  11. Ion irradiation testing and characterization of FeCrAl candidate alloys

    Energy Technology Data Exchange (ETDEWEB)

    Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wang, Yongqiang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-10-29

    The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commercially available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.

  12. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-15

    Highlights: • Thermal properties of irradiated U–Mo alloy monolithic fuel samples were measured. • Density, thermal diffusivity, and thermal conductivity are influenced by increasing burnup. • U–Mo chemistry and specific heat capacity was not as sensitive to increasing burnup. • Thermal conductivity decreased approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} at 200 °C. • An empirical model developed previously agrees well with the experimental measurements. - Abstract: A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U–Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U–Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U–Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U–Mo alloy decreased approximately 30% for a fission density of 3.30 × 10{sup 21} fissions cm{sup −3} and approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  13. Project Progress of New Domestic Zirconium Alloy Fuel Sub-assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Ai-min; ZHANG; Pei-sheng; LIU; Jia-zheng; LIU; Wei

    2015-01-01

    At present,the project of new domestic zirconium alloy fuel sub-assembly irradiation is ongoing according to schedule.This paper presents progress of the project such as fuel sub-assembly detailed design,manufacturing process and fuel transportation method.1 Fuel sub-assembly detailed designing

  14. Properties of unirradiated and irradiated Ti-6Al-4V alloy for ITER flexible connectors

    Energy Technology Data Exchange (ETDEWEB)

    Rodchenkov, B.S., E-mail: rodchen@nikiet.ru [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, Moscow 101000 (Russian Federation); Evseev, M.V. [Institute of Reactor Materials, Zarechnyi, Sverdlovsk Region 624051 (Russian Federation); Strebkov, Yu.S. [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, Moscow 101000 (Russian Federation); Sinelnikov, L.P.; Shushlebin, V.V. [Institute of Reactor Materials, Zarechnyi, Sverdlovsk Region 624051 (Russian Federation)

    2011-10-01

    The high strength ({alpha} + {beta}) Ti-6Al-4V alloy was selected as the material for flexible attachments of the shield blanket modules in the ITER reactor. The different technologies used for manufacturing this alloy are: forging, stamping or pressing. The microstructures resulting from these processing methods can vary significantly and as a consequence the properties, including irradiation behavior, also vary. There are limited data available on the irradiation behavior of these materials. Specimens cut in the longitudinal and transversal directions of forged and stamped material were studied, with some of the specimens hydrogen charged to {approx}400 ppm H{sub 2}. In the unirradiated condition the forged alloy had slightly more ductility than the stamped alloy. The strength properties of both were practically the same. Neutron irradiation of these materials in the IVV-2M reactor reached doses of {approx}0.2 and 0.3 dpa at temperatures 240-260 deg. C. Irradiation resulted in substantial hardening and significant decrease of the fracture toughness of specimens from both materials.

  15. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    Science.gov (United States)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  16. Tensile properties of vanadium alloys irradiated at 390{degrees}C in EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Tsai, H.C.; Nowicki, L.J. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    Vanadium alloys were irradiated in Li-bonded stainless steel capsules to {approx}390{degrees}C in the EBR-II X-530 experiment. This report presents results of postirradiation tests of tensile properties of two large-scale (100 and 500 kg) heats of V-4Cr-Ti and laboratory (15-30 kg) heats of boron-doped V-4Cr-4Ti, V-8Cr-6Ti, V-5Ti, and V-3Ti-1Si alloys. Tensile specimens, divided into two groups, were irradiated in two different capsules under nominally similar conditions. The 500-kg heat (No. 832665) and the 100-kg heat (VX-8) of V-4Cr-4Ti irradiated in one of the subcapsules exhibited complete loss of work-hardening capability, which was manifested by very low uniform plastic strain. In contrast, the 100-kg heat of V-4Cr-4Ti irradiated in another subcapsule exhibited good tensile properties (uniform plastic strain 2.8-4.0%). A laboratory heat of V-3Ti-1Si irradiated in the latter subcapsule also exhibited good tensile properties. These results indicate that work-hardening capability at low irradiation temperatures varies significantly from heat to heat and is influenced by nominally small differences in irradiation conditions.

  17. Nonequilibrium self-organization in alloys under irradiation leading to the formation of nano composites

    CERN Document Server

    Enrique, R A; Averback, R S; Bellon, P

    2003-01-01

    Alloys under irradiation are continuously driven away from equilibrium: Every time an external particle interacts with the atoms in the solid, a perturbation very localized in space and time is produced. Under this external forcing, phase and microstructural evolution depends ultimately on the dynamical interaction between the external perturbation and the internal recovery kinetics of the alloy. We consider the nonequilibrium steady state of an immiscible binary alloy subject to mixing by heavy-ion irradiation. It has been found that the range of the forced atomic relocations taking place during collision cascades plays an important role on the final microstructure: when this range is large enough, it can lead to the spontaneous formation of compositional patterns at the nanometer scale. These results were rationalized in the framework of a continuum model solved by deriving a nonequilibrium thermodynamic potential. Here we derive the nonequilibrium structure factor by including the role of fluctuations. In ...

  18. Initial tensile test results from J316 stainless steel irradiated in the HFIR spectrally tailored experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1995-04-01

    The objective of this work is to determine the effects of neutron irradiation on the mechanical properties of austenitic stainless steel alloys. In this experiment, the spectrum has been tailored to reduce the thermal neutron flux and achieve a He/dpa level near that expected in a fusion reactor.

  19. Database on Performance of Neutron Irradiated FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Univ. of Wisconsin, Madison, WI (United States); Littrell, Ken [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    The present report summarizes and discusses the database on radiation tolerance for Generation I, Generation II, and commercial FeCrAl alloys. This database has been built upon mechanical testing and microstructural characterization on selected alloys irradiated within the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to doses of 13.8 dpa at temperatures ranging from 200°C to 550°C. The structure and performance of these irradiated alloys were characterized using advanced microstructural characterization techniques and mechanical testing. The primary objective of developing this database is to enhance the rapid development of a mechanistic understanding on the radiation tolerance of FeCrAl alloys, thereby enabling informed decisions on the optimization of composition and microstructure of FeCrAl alloys for application as an accident tolerant fuel (ATF) cladding. This report is structured to provide a brief summary of critical results related to the database on radiation tolerance of FeCrAl alloys.

  20. U.S. Contribution 1994 Summary Report Task T12: Compatibility and irradiation testing of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L. [comp.

    1995-03-01

    Vanadium alloys exhibit important advantages as a candidate structural material for fusion first wall/blanket applications. These advantages include fabricability, favorable safety and environmental features, high temperature and high wall load capability, and long lifetime under irradiation. Vanadium alloys with (3-5)% chromium and (3-5)% titanium appear to offer the best combination of properties for first wall/blanket applications. A V-4Cr-4Ti alloy is recommended as the reference composition for the ITER application. This report provides a summary of the R&D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. Separate abstracts have been prepared for individual reports from this publication.

  1. Metastable phases in Zr-Excel alloy and their stability under heavy ion (Kr2+) irradiation

    Science.gov (United States)

    Yu, Hongbing; Zhang, Ken; Yao, Zhongwen; Kirk, Mark A.; Long, Fei; Daymond, Mark R.

    2016-02-01

    Zr-Excel alloy (Zr-3.5Sn-0.8Nb-0.8Mo, wt.%) has been proposed as a candidate material of pressure tubes in the CANDU-SCWR design. It is a dual-phase alloy containing primary hcp α-Zr and metastable bcc β-Zr. Metastable hexagonal ω-Zr phase could form in β-Zr as a result of aging during the processing of the tube. A synchrotron X-ray study was employed to study the lattice properties of the metastable phases in as-received Zr-Excel pressure tube material. In situ heavy ion (1 MeV Kr2+) irradiations were carried out at 200 °C and 450 °C to emulate the stability of the metastable phase under a reactor environment. Quantitative Chemi-STEM EDS analysis was conducted on both un-irradiated and irradiated samples to investigate alloying element redistribution induced by heavy ion irradiation. It was found that no decomposition of β-Zr was observed under irradiation at both 200 °C and 450 °C. However, ω-Zr particles experienced shape changes and shrinkage associated with enrichment of Fe at the β/ω interface during 200 °C irradiation but not at 450 °C. There is a noticeable increase in the level of Fe in the α matrix after irradiation at both 200 °C and 450 °C. The concentrations of Nb, Mo and Fe are increased in the ω phase but decreased in the β phase at 200 °C. The stability of metastable phases under heavy ion irradiation associated with elemental redistribution is discussed.

  2. Ablation experiment and threshold calculation of titanium alloy irradiated by ultra-fast pulse laser

    Directory of Open Access Journals (Sweden)

    Buxiang Zheng

    2014-02-01

    Full Text Available The interaction between an ultra-fast pulse laser and a material's surface has become a research hotspot in recent years. Micromachining of titanium alloy with an ultra-fast pulse laser is a very important research direction, and it has very important theoretical significance and application value in investigating the ablation threshold of titanium alloy irradiated by ultra-fast pulse lasers. Irradiated by a picosecond pulse laser with wavelengths of 1064 nm and 532 nm, the surface morphology and feature sizes, including ablation crater width (i.e. diameter, ablation depth, ablation area, ablation volume, single pulse ablation rate, and so forth, of the titanium alloy were studied, and their ablation distributions were obtained. The experimental results show that titanium alloy irradiated by a picosecond pulse infrared laser with a 1064 nm wavelength has better ablation morphology than that of the green picosecond pulse laser with a 532 nm wavelength. The feature sizes are approximately linearly dependent on the laser pulse energy density at low energy density and the monotonic increase in laser pulse energy density. With the increase in energy density, the ablation feature sizes are increased. The rate of increase in the feature sizes slows down gradually once the energy density reaches a certain value, and gradually saturated trends occur at a relatively high energy density. Based on the linear relation between the laser pulse energy density and the crater area of the titanium alloy surface, and the Gaussian distribution of the laser intensity on the cross section, the ablation threshold of titanium alloy irradiated by an ultra-fast pulse laser was calculated to be about 0.109 J/cm2.

  3. Heavy ion irradiation induced dislocation loops in AREVA's M5 Registered-Sign alloy

    Energy Technology Data Exchange (ETDEWEB)

    Hengstler-Eger, R.M., E-mail: Rosmarie.Hengstler-Eger@areva.com [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Baldo, P. [Argonne National Laboratory, Materials Science Division, 9700 South Cass Avenue, 60439 Argonne IL (United States); Beck, L. [Maier-Leibnitz-Laboratorium (MLL), Am Coulombwall 6, 85748 Garching (Germany); Dorner, J.; Ertl, K. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Hoffmann, P.B. [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Hugenschmidt, C. [Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II), Lichtenbergstr. 1, 85747 Garching (Germany); Kirk, M.A. [Argonne National Laboratory, Materials Science Division, 9700 South Cass Avenue, 60439 Argonne IL (United States); Petry, W.; Pikart, P. [Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II), Lichtenbergstr. 1, 85747 Garching (Germany); Rempel, A. [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)

    2012-04-15

    Pressurized water reactor (PWR) Zr-based alloy structural materials show creep and growth under neutron irradiation as a consequence of the irradiation induced microstructural changes in the alloy. A better scientific understanding of these microstructural processes can improve simulation programs for structural component deformation and simplify the development of advanced deformation resistant alloys. As in-pile irradiation leads to high material activation and requires long irradiation times, the objective of this work was to study whether ion irradiation is an applicable method to simulate typical PWR neutron damage in Zr-based alloys, with AREVA's M5 Registered-Sign alloy as reference material. The irradiated specimens were studied by electron backscatter diffraction (EBSD), positron Doppler broadening spectroscopy (DBS) and in situ transmission electron microscopy (TEM) at different dose levels and temperatures. The irradiation induced microstructure consisted of - and -type dislocation loops with their characteristics corresponding to typical neutron damage in Zr-based alloys; it can thus be concluded that heavy ion irradiation under the chosen conditions is an excellent method to simulate PWR neutron damage.

  4. Impact property of low-activation vanadium alloy after laser welding and heavy neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Nagasaka, Takuya, E-mail: nagasaka@nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu (Japan); The Graduate University for Advanced Studies, Toki, Gifu (Japan); Muroga, Takeo [National Institute for Fusion Science, Toki, Gifu (Japan); The Graduate University for Advanced Studies, Toki, Gifu (Japan); Watanabe, Hideo [Research Institute for Applied Mechanics, Kyushu University, Kasuga (Japan); Miyazawa, Takeshi [The Graduate University for Advanced Studies, Toki, Gifu (Japan); Yamazaki, Masanori [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki (Japan); Shinozaki, Kenji [Department of Mechanical System Engineering, Graduate School of Engineering, Hiroshima University, Higashi Hiroshima (Japan)

    2013-11-15

    Weld specimens of the reference low activation vanadium alloy, NIFS-HEAT-2, were irradiated up to a neutron fluence of 1.5 × 10{sup 25} n m{sup −2} (E > 0.1 MeV) (1.2 dpa) at 670 K and 1.3 × 10{sup 26} n m{sup −2} (5.3 dpa) at 720 K in the JOYO reactor in Japan. The base metal exhibited superior irradiation resistance with the ductile-to-brittle transition temperature (DBTT) much lower than room temperature (RT) for both irradiation conditions. The weld metal kept the DBTT below RT after the 1.2 dpa irradiation; however, it showed enhanced irradiation embrittlement with much higher DBTT than RT after the 5.3 dpa irradiation. The high DBTT for the weld metal was effectively recovered by a post-irradiation annealing at 873 K for 1 h. Mechanisms of the irradiation embrittlement and its recovery are discussed, based on characterization of the radiation defects and irradiation-induced precipitation.

  5. Summary of the U.S. specimen matrix for the HFIR 13J varying temperature irradiation capsule

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    The US specimen matrix for the collaborative DOE/Monbusho HFIR 13J varying temperature irradiation capsule contains two ceramics and 29 different metals, including vanadium alloys, ferritic/martensitic steels, pure iron, austenitic stainless steels, nickel alloys, and copper alloys. This experiment is designed to provide fundamental information on the effects of brief low-temperature excursions on the tensile properties and microstructural evolution of a wide range of materials irradiated at nominal temperatures of 350 and 500 C to a dose of {approximately}5 dpa. A total of 340 miniature sheet tensile specimens and 274 TEM disks are included in the US-supplied matrix for the irradiation capsule.

  6. Nanoparticles alloying in liquids: Laser-ablation-generated Ag or Pd nanoparticles and laser irradiation-induced AgPd nanoparticle alloying

    Science.gov (United States)

    Semaltianos, N. G.; Chassagnon, R.; Moutarlier, V.; Blondeau-Patissier, V.; Assoul, M.; Monteil, G.

    2017-04-01

    Laser irradiation of a mixture of single-element micro/nanomaterials may lead to their alloying and fabrication of multi-element structures. In addition to the laser induced alloying of particulates in the form of micro/nanopowders in ambient atmosphere (which forms the basis of the field of additive manufacturing technology), another interesting problem is the laser-induced alloying of a mixture of single-element nanoparticles in liquids since this process may lead to the direct fabrication of alloyed-nanoparticle colloidal solutions. In this work, bare-surface ligand-free Ag and Pd nanoparticles in solution were prepared by laser ablation of the corresponding bulk target materials, separately in water. The two solutions were mixed and the mixed solution was laser irradiated for different time durations in order to investigate the laser-induced nanoparticles alloying in liquid. Nanoparticles alloying and the formation of AgPd alloyed nanoparticles takes place with a decrease of the intensity of the surface-plasmon resonance peak of the Ag nanoparticles (at ∼405 nm) with the irradiation time while the low wavelength interband absorption peaks of either Ag or Pd nanoparticles remain unaffected by the irradiation for a time duration even as long as 30 min. The nanoalloys have lattice constants with values between those of the pure metals, which indicates that they consist of Ag and Pd in an approximately 1:1 ratio similar to the atomic composition of the starting mixed-nanoparticle solution. Formation of nanoparticle networks consisting of bimetallic alloyed nanoparticles and nanoparticles that remain as single elements (even after the end of the irradiation), joining together, are also formed. The binding energies of the 3d core electrons of both Ag and Pd nanoparticles shift to lower energies with the irradiation time, which is also a typical characteristic of AgPd alloyed nanoparticles. The mechanisms of nanoparticles alloying and network formation are also

  7. Nanoparticles alloying in liquids: Laser-ablation-generated Ag or Pd nanoparticles and laser irradiation-induced AgPd nanoparticle alloying.

    Science.gov (United States)

    Semaltianos, N G; Chassagnon, R; Moutarlier, V; Blondeau-Patissier, V; Assoul, M; Monteil, G

    2017-04-18

    Laser irradiation of a mixture of single-element micro/nanomaterials may lead to their alloying and fabrication of multi-element structures. In addition to the laser induced alloying of particulates in the form of micro/nanopowders in ambient atmosphere (which forms the basis of the field of additive manufacturing technology), another interesting problem is the laser-induced alloying of a mixture of single-element nanoparticles in liquids since this process may lead to the direct fabrication of alloyed-nanoparticle colloidal solutions. In this work, bare-surface ligand-free Ag and Pd nanoparticles in solution were prepared by laser ablation of the corresponding bulk target materials, separately in water. The two solutions were mixed and the mixed solution was laser irradiated for different time durations in order to investigate the laser-induced nanoparticles alloying in liquid. Nanoparticles alloying and the formation of AgPd alloyed nanoparticles takes place with a decrease of the intensity of the surface-plasmon resonance peak of the Ag nanoparticles (at ∼405 nm) with the irradiation time while the low wavelength interband absorption peaks of either Ag or Pd nanoparticles remain unaffected by the irradiation for a time duration even as long as 30 min. The nanoalloys have lattice constants with values between those of the pure metals, which indicates that they consist of Ag and Pd in an approximately 1:1 ratio similar to the atomic composition of the starting mixed-nanoparticle solution. Formation of nanoparticle networks consisting of bimetallic alloyed nanoparticles and nanoparticles that remain as single elements (even after the end of the irradiation), joining together, are also formed. The binding energies of the 3d core electrons of both Ag and Pd nanoparticles shift to lower energies with the irradiation time, which is also a typical characteristic of AgPd alloyed nanoparticles. The mechanisms of nanoparticles alloying and network formation are also

  8. Tensile properties of vanadium alloys irradiated at 200{degrees}C in the HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Nowicki, L.; Smith, D.L. [Argonne National Lab., IL (United States)

    1997-08-01

    Vanadium alloys were irradiated in a helium environment to {approx}10 dpa at {approx}200{degrees}C in the High Flux Isotope Reactor (HFIR). This report presents results of postirradiation tests of tensile properties of laboratory heats of V-(1-18)Ti, V-4Cr-4Ti, V-8Cr-6Ti, V-9Cr-5Ti, V-3Ti-1Si, and V-3Ti-0.1C alloys. Because of significant loss of work-hardening capability, all alloys except V-18Ti exhibited a very low uniform plastic strain <1%. For V-Ti. The mechanism of the loss of work-hardening capability in the other alloys is not understood.

  9. Positron annihilation study of the hardening behavior in Al-Cu based alloy by electron and heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hori, Fuminobu; Kobayashi, Ippei; Iwase, Akihiro [Department of Materials Science, Osaka Prefecture University, 1-1 Gakuen-cho, Sakai, Osaka 599-8531 (Japan); Saito, Yuichi; Ishikawa, Norito; Oshima, Takeshi, E-mail: horif@mtr.osakafu-u.ac.j [JAEA Tokai, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2010-04-01

    Al-Cu based alloy, which is generally called duralumin (JIS2017), was irradiated with 10 MeV Iodine ions, 200 MeV Xenon ions and 3 MeV electrons at room temperature respectively. The micro Vicker's hardness and positron annihilation coincidence Doppler broadening (CDB) measurements have been performed before and after irradiation. Only in the case of ion irradiation, the Vicker's hardness increases with increasing ion dose. Nevertheless, there was no difference in the profile CDB spectrum for before and after irradiation. On the other hand, we found that the micro hardness of this alloy, which was Xe ion irradiated and subsequently annealed at 423 K, is greater than that of age hardened alloy without irradiation. CDB ratio curve of the age hardened Duralumin is clearly different in the electron momentum range around 0.015-0.025 mc from that of the ion irradiated alloy. The results of three-dimensional atom probe (3DAP) also show that a lot of small clusters were found after ion irradiation but large precipitations have found in annealed Duralumin. These results reveal that a number of small clusters formed in this alloy after ion irradiation, and they should strongly affects the micro hardness.

  10. Effects of Grit Blasting and Annealing on the High-Temperature Oxidation Behavior of Austenitic and Ferritic Fe-Cr Alloys

    Science.gov (United States)

    Proy, M.; Utrilla, M. V.; Otero, E.; Bouchaud, B.; Pedraza, F.

    2014-08-01

    Grit blasting (corundum) of an austenitic AISI 304 stainless steel (18Cr-8Ni) and of a low-alloy SA213 T22 ferritic steel (2.25Cr-1Mo) followed by annealing in argon resulted in enhanced outward diffusion of Cr, Mn, and Fe. Whereas 3 bar of blasting pressure allowed to grow more Cr2O3 and Mn x Cr3- x O4 spinel-rich scales, higher pressures gave rise to Fe2O3-enriched layers and were therefore disregarded. The effect of annealing pre-oxidation treatment on the isothermal oxidation resistance was subsequently evaluated for 48 h for both steels and the results were compared with their polished counterparts. The change of oxidation kinetics of the pre-oxidized 18Cr-8Ni samples at 850 °C was ascribed to the growth of a duplex Cr2O3/Mn x Cr3- x O4 scale that remained adherent to the substrate. Such a positive effect was less marked when considering the oxidation kinetics of the 2.25Cr-1Mo steel but a more compact and thinner Fe x Cr3- x O4 subscale grew at 650 °C compared to that of the polished samples. It appeared that the beneficial effect is very sensitive to the experimental blasting conditions. The input of Raman micro-spectroscopy was shown to be of ground importance in the precise identification of multiple oxide phases grown under the different conditions investigated in this study.

  11. Helium effects on irradiation dmage in V alloys

    Energy Technology Data Exchange (ETDEWEB)

    Doraiswamy, N.; Alexander, D. [Argonne National Lab., IL (United States)

    1996-10-01

    Preliminary investigations were performed on V-4Cr-4Ti samples to observe the effects of He on the irradiation induced microstructural changes by subjecting 3 mm electropolished V-4Cr-4Ti TEM disks, with and without prior He implantation, to 200 keV He irradiation at room temperature and monitoring, in-situ, the microstructural evolution as a function of total dose with an intermediate voltage electron microscope directly connected to an ion implanter. A high density of black dot defects were formed at very low doses in both He pre-implanted and unimplanted samples.

  12. Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

    Science.gov (United States)

    Idrees, Y.; Yao, Z.; Cui, J.; Shek, G. K.; Daymond, M. R.

    2016-11-01

    Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters. Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen.

  13. Tensile properties of vanadium alloys irradiated at <430{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

    1997-08-01

    Recent attention to vanadium alloys has focused on significant susceptibility to loss of work-hardening capability in irradiation experiments at <430{degrees}C. An evaluation of this phenomenon was conducted on V-Ti, V-Cr-Ti, and V-Ti-Si alloys irradiated in several conventional and helium-charging irradiation experiments in the FFTF-MOTA, HFIR, and EBR-II. Work hardening capability and uniform tensile elongation appear to vary strongly from alloy and heat to heat. A strong heat-to-heat variation has been observed in V-4Cr-4Ti alloys tested, i.e., a 500-kg heat (No. 832665), a 100-kg heat (VX-8), and a 30-kg heat (BL-47). The significant differences in susceptibility to loss of work-hardening capability from one heat to another are estimated to correspond to a difference of {approx}100{degrees}C or more in minimum allowable operating temperature (e.g., 450 versus 350{degrees}C).

  14. A novel way to estimate the nanoindentation hardness of only-irradiated layer and its application to ion irradiated Fe-12Cr alloy

    Science.gov (United States)

    Kim, Hoon-Seop; Lee, Dong-Hyun; Seok, Moo-Young; Zhao, Yakai; Kim, Woo-Jin; Kwon, Dongil; Jin, Hyung-Ha; Kwon, Junhyun; Jang, Jae-il

    2017-04-01

    While nanoindentation is a very useful tool to examine the mechanical properties of ion irradiated materials, there are some issues that should be considered in evaluating the properties of irradiated layer. In this study, in order to properly extract the hardness of only-irradiated layer from nanoindentation data, a new procedure is suggested in consideration of the geometry of indentation-induced plastic zone. By applying the procedure to an ion irradiated Fe-12Cr alloy, the reasonable results were obtained, validating its usefulness in the investigation of practical effect of irradiation on the mechanical behavior of future nuclear materials.

  15. Impact properties of vanadium-base alloys irradiated at < 430 C

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

    1998-03-01

    Recent attention to vanadium-base alloys has focused on the effect of low-temperature (<430 C) neutron irradiation on the mechanical properties, especially the phenomena of loss of work-hardening capability under tensile loading and loss of dynamic toughness manifested by low impact energy and high ductile-brittle-transition temperature (DBTT). This paper summarizes results of an investigation of the low-temperature impact properties of V-5Ti, V-4Cr-4Ti, and V-3Ti-Si that were irradiated in several fission reactor experiments, i.e., FFTF-MOTA, EBR-II X-530, and ATR-A1. Irradiation performance of one production-scale and one laboratory heat of V-4C-4Ti and one laboratory heat of V-3Ti-Si was the focus of the investigation. Even among the same lass of alloy, strong heat-to-heat variation was observed in low-temperature impact properties. A laboratory heat of V-4Cr-4Ti and V-3Ti-1Si exhibited good impact properties whereas a 500-kg heat of V-4Cr-4Ti exhibited unacceptably high DBTT. The strong heat-to-heat variation in impact properties of V-4Cr-4Ti indicates that fabrication procedures and minor impurities play important roles in the low-temperature irradiation performance of the alloys.

  16. Microstructure analysis of magnesium alloy melted by laser irradiation

    Science.gov (United States)

    Liu, S. Y.; Hu, J. D.; Yang, Y.; Guo, Z. X.; Wang, H. Y.

    2005-12-01

    The effects of laser surface melting (LSM) on microstructure of magnesium alloy containing Al8.57%, Zn 0.68%, Mn0.15%, Ce0.52% were investigated. In the present work, a pulsed Nd:YAG laser was used to melt and rapidly solidify the surface of the magnesium alloy with the objective of changing microstructure and improving the corrosion resistance. The results indicate that laser-melted layer contains the finer dendrites and behaviors good resistance corrosion compared with the untreated layer. Furthermore, the absorption coefficient of the magnesium alloy has been estimated according to the numeral simulation of the thermal conditions. The formation process of fine microstructure in melted layers was investigated based on the experimental observation and the theoretical analysis. Some simulation results such as the re-solidification velocities are obtained. The phase constitutions of the melted layers determined by X-ray diffraction were β-Mg 17Al 12 and α-Mg as well as some phases unidentified.

  17. Irradiation Embritlement in Alloy HT-­9

    Energy Technology Data Exchange (ETDEWEB)

    Serrano De Caro, Magdalena [Los Alamos National Laboratory

    2012-08-27

    HT-9 steel is a candidate structural and cladding material for high temperature lead-bismuth cooled fast reactors. In typical advanced fast reactor designs fuel elements will be irradiated for an extended period of time, reaching up to 5-7 years. Significant displacement damage accumulation in the steel is expected (> 200 dpa) when exposed to dpa-rates of 20-30 dpa{sub Fe}/y and high fast flux (E > 0.1 MeV) {approx}4 x 10{sup 15} n/cm{sup 2}s. Core temperatures could reach 400-560 C, with coolant temperatures at the inlet as low as 250 C, depending on the reactor design. Mechanical behavior in the presence of an intense fast flux and high dose is a concern. In particular, low temperature operation could be limited by irradiation embrittlement. Creep and corrosion effects in liquid metal coolants could set a limit to the upper operating temperature. In this report, we focus on the low temperature operating window limit and describe HT-9 embrittlement experimental findings reported in the literature that could provide supporting information to facilitate the consideration of a Code Case on irradiation effects for this class of steels in fast reactor environments. HT-9 has an extensive database available on irradiation performance, which makes it the best choice as a possible near-term candidate for clad, and ducts in future fast reactors. Still, as it is shown in this report, embrittlement data for very low irradiation temperatures (< 200 C) and very high radiation exposure (> 150 dpa) is scarce. Experimental findings indicate a saturation of DBTT shifts as a function of dose, which could allow for long lifetime cladding operation. However, a strong increase in DBTT shift with decreasing irradiation temperature could compromise operation at low service temperatures. Development of a deep understanding of the physics involved in the radiation damage mechanisms, together with multiscale computer simulation models of irradiation embrittlement will provide the basis to

  18. Alloy development for irradiation performance. Quarterly progress report for period ending March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (comp.)

    1980-06-01

    This report is organized along topical lines in parallel to a Program Plan of the same title so that activities and accomplishments may be followed readily relative to that Program Plan. Thus, the work of a given laboratory may appear throughout the report. Chapters 1, 2, 8, and 9 review activities on analysis and evaluation, test methods development, status of irradiation experiments, and corrosion testing and hydrogen permeation studies, respectively. These activities relate to each of the alloy development paths. Chapters 3, 4, 5, 6, and 7 present the ongoing work on each alloy development path. The Table of Contents is annotated for the convenience of the reader.

  19. Effect of strain on ferrite transformation from super-cooled austenite in Fe-0. 5%C alloy. Fe-0. 5%C gokin no karei osutenaito/feraito hentai ni oyobosu kako no eikyo

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, K.; Ito, Y.; Narita, T. (Hokkaido Univ., Sapporo (Japan). Faculty of Engineering)

    1993-08-01

    During the cooling of a steel, when austenite is applied by strain, the temperature of ferrite transformation would increase accompanied with decrease of its given temperature and increase of strain. In this study, the isothermal transformation behaviour from austenite to ferrite applied by strain in the super-cooled state was investigated, effect of strain on size of ferrite particles and increase of volume rate during transformation were explained by using the velocity theory. That is, concerning to the alloy of two-elemental system Fe-0.51%C cooled at 0.3[degree]C/s and applied by strain at 710[degree]C, at which austenite was super-cooled by 55[degree]C, its isothermal transformation behaviour was investigated. As a result, the following conclusions were obtained. Time required for the transformation remarkably decreased and the size of ferrite particles became ultra-fine subjected to strain. The nucleation rate of ferrite particles remarkably increased with increasing strain. 14 refs., 11 figs., 1 tab.

  20. Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

  1. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Gomes, I.; Chung, H.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-04-01

    The ATR-A1 irradiation experiment in the Advanced Test Reactor (ATR) was a collaborative U.S./Japan effort to study at low temperatures the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation was completed on May 5, 1996, as planned, after achieving an estimated neutron damage of 4.7 dpa in vanadium. The capsule has since been kept in the ATR water canal for the required radioactivity cool-down. Planning is underway for disassembly of the capsule and test specimen retrieval.

  2. Phase stability and microstructures of high entropy alloys ion irradiated to high doses

    Science.gov (United States)

    Xia, Songqin; Gao, Michael C.; Yang, Tengfei; Liaw, Peter K.; Zhang, Yong

    2016-11-01

    The microstructures of AlxCoCrFeNi (x = 0.1, 0.75 and 1.5 in molar ratio) high entropy alloys (HEAs) irradiated at room temperature with 3 MeV Au ions at the highest fluence of 105, 91, and 81 displacement per atom, respectively, were studied. Transmission electron microscopy (TEM) and high-resolution TEM (HRTEM) analyses show that the initial microstructures and phase composition of all three alloys are retained after ion irradiation and no phase decomposition is observed. Furthermore, it is demonstrated that the disordered face-centered cubic (FCC) and disordered body-centered cubic (BCC) phases show much less defect cluster formation and structural damage than the NiAl-type ordered B2 phase. This effect is explained by higher entropy of mixing, higher defect formation/migration energies, substantially lower thermal conductivity, and higher atomic level stress in the disordered phases.

  3. Density decrease in vanadium-base alloys irradiated in the dynamic helium charging experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Galvin, T.M.; Smith, D.L. [Argonne National Laboratory, Chicago, IL (United States)

    1996-04-01

    Combined effects of dynamically charged helium and neutron damage on density decrease (swelling) of V-4Cr-4Ti, V-5Ti, V-3Ti-1Si, and V-8Cr-6Ti alloys have been determined after irradiation to 18-31 dpa at 425-600{degrees}C in the Dynamic helium Charging Experiment (DHCE). To ensure better accuracy in density measurement, broken pieces of tensile specimens {approx} 10 times heavier than a transmission electron microscopy (TEM) disk were used. Density increases of the four alloys irradiated in the DHCE were <0.5%. This small change seems to be consistent with the negligible number density of microcavities characterized by TEM. Most of the dynamically produced helium atoms seem to have been trapped in the grain matrix without significant cavity nucleation or growth.

  4. Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

    Science.gov (United States)

    Sun, Yong; Huo, He-yong; Wu, Yang; Li, Jiangbo; Zhou, Wei; Guo, Hai-bing; Li, Hang; Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin

    2016-11-01

    It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 μm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

  5. Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-04-01

    Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}C to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.

  6. Thermal-hydraulic Analysis of New Zirconium Alloys Assembly Irradiated in CARR

    Institute of Scientific and Technical Information of China (English)

    YIN; Hao; ZHAO; Shou-zhi; LIU; Xing-min

    2013-01-01

    This article is mainly about the thermal-hydraulic analysis of the new zirconium alloys assembly on irradiation test of China Advanced Research Reactor(CARR),so as to provide security assessment throughout the design.CFD software was used for three-dimensional simulation.Firstly,the geometric model,mesh,specified boundary condition types and region types were constructed.Then importing the

  7. Irradiation behavior of Ti 4Al 2V (ΠT-3B) alloy for ITER blanket modules flexible attachment

    Science.gov (United States)

    Rodchenkov, B. S.; Kozlov, A. V.; Kuznetsov, Yu. G.; Kalinin, G. M.; Strebkov, Yu. S.

    2007-08-01

    Titanium alloys are recommended as a material to manufacture flexible attachments of the shield blanket modules in the ITER reactor owing to their advantageous combination of properties, i.e., high resistance to impact loading, strength, density and low thermal expansion coefficient. An additional factor for selecting Ti alloys is their fast induced radioactivity decay. The (α + β)-Ti alloys have higher strength than (α)-Ti alloys but are less developed. The data base on the irradiation behavior of these materials is limited. Neutron irradiation of (α)-Ti-4Al-2V (ΠT-3B) alloy has been performed in the framework of the ITER R&D programme. Specimens from a forging of Ti-4Al-2V alloy were irradiated in the IVV-2M reactor to doses of (0.32-0.43) dpa at temperatures of (240-260) °C. This paper describes the results of tensile, low cycle fatigue and fracture toughness tests of alloy in the unirradiated and neutron irradiated conditions. The results obtained are compared with those of the (α + β)-Ti-6Al-4V alloy.

  8. Neutron irradiation effects on the microstructural development of tungsten and tungsten alloys

    Science.gov (United States)

    Hasegawa, Akira; Fukuda, Makoto; Yabuuchi, Kiyohiro; Nogami, Shuhei

    2016-04-01

    Data on the microstructural development of tungsten (W) and tungsten rhenium (Re) alloys were obtained after neutron irradiation at 400-800 °C in the Japan Materials Testing Reactor (JMTR), the experimental fast test reactor Joyo, and the High Flux Isotope Reactor (HFIR) for irradiation damage levels in the range of 0.09-1.54 displacement per atom (dpa). Microstructural observations showed that a small amount of Re (3-5%) in W-Re alloys is effective in suppressing void formation. In W-Re alloys with Re concentrations greater than 10%, acicular precipitates are the primary structural defects. In the HFIR-irradiated specimen, in which a large amount of Re was expected to be produced by the nuclear transmutation of W to Re because of the reactor's high thermal neutron flux, voids were not observed even in pure W. The synergistic effects of displacement damage and solid transmutation elements on microstructural development are discussed, and the microstructural development of tungsten materials utilized in fusion reactors is predicted.

  9. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    Science.gov (United States)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  10. TEM in situ micropillar compression tests of ion irradiated oxide dispersion strengthened alloy

    Science.gov (United States)

    Yano, K. H.; Swenson, M. J.; Wu, Y.; Wharry, J. P.

    2017-01-01

    The growing role of charged particle irradiation in the evaluation of nuclear reactor candidate materials requires the development of novel methods to assess mechanical properties in near-surface irradiation damage layers just a few micrometers thick. In situ transmission electron microscopic (TEM) mechanical testing is one such promising method. In this work, microcompression pillars are fabricated from a Fe2+ ion irradiated bulk specimen of a model Fe-9%Cr oxide dispersion strengthened (ODS) alloy. Yield strengths measured directly from TEM in situ compression tests are within expected values, and are consistent with predictions based on the irradiated microstructure. Measured elastic modulus values, once adjusted for the amount of deformation and deflection in the base material, are also within the expected range. A pillar size effect is only observed in samples with minimum dimension ≤100 nm due to the low inter-obstacle spacing in the as received and irradiated material. TEM in situ micropillar compression tests hold great promise for quantitatively determining mechanical properties of shallow ion-irradiated layers.

  11. Effects of ultraviolet irradiation on bonding strength between Co-Cr alloy and citric acid-crosslinked gelatin matrix.

    Science.gov (United States)

    Inoue, Motoki; Sasaki, Makoto; Katada, Yasuyuki; Taguchi, Tetsushi

    2014-02-01

    Novel techniques for creating a strong bond between polymeric matrices and biometals are required. We immobilized polymeric matrices on the surface of biometal for drug-eluting stents through covalent bond. We performed to improve the bonding strength between a cobalt-chromium alloy and a citric acid-crosslinked gelatin matrix by ultraviolet irradiation on the surface of cobalt-chromium alloy. The ultraviolet irradiation effectively generated hydroxyl groups on the surface of the alloy. The bonding strength between the gelatin matrix and the alloy before ultraviolet irradiation was 0.38 ± 0.02 MPa, whereas it increased to 0.48 ± 0.02 MPa after ultraviolet irradiation. Surface analysis showed that the citric acid derivatives occurred on the surface of the cobalt-chromium alloy through ester bond. Therefore, ester bond formation between the citric acid derivatives active esters and the hydroxyl groups on the cobalt-chromium alloy contributed to the enhanced bonding strength. Ultraviolet irradiation and subsequent immobilization of a gelatin matrix using citric acid derivatives is thus an effective way to functionalize biometal surfaces.

  12. Properties of V-(8-9)Cr-(5-6)Ti alloys irradiated in the dynamic helium charging experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Nowicki, L.; Smith, D.L. [Argonne National Lab., IL (United States)

    1996-10-01

    In the Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in vanadium alloy specimens by the decay of tritium during irradiation to 18-31 dpa at 425-600{degrees}C in lithium-filled capsules in the Fast Flux Test Facility. This report presents results of postirradiation tests of tensile properties and density change in V-8Cr-6Ti and V-9Cr-5Ti. Compared to tensile properties of the alloys irradiated in the non-DHCE (helium generation negligible), the effect of helium on tensile strength and ductility of V-8Cr-6Ti and V-9Cr-5Ti was insignificant after irradiation and testing at 420, 500, and 600{degrees}C. Both alloys retained a total elongation of >11 % at these temperatures. Density change was <0.48% for both alloys.

  13. Texture and microstructure of the austenite in multiphased steel sheets

    Energy Technology Data Exchange (ETDEWEB)

    Regle, H. [ARCELOR Group, CMC-IRSID, Maizieres-les-Metz (France); Maruyama, N.; Yoshinaga, N. [Nippon Steel Corp. - Technical Development Bureau, Futtsu (Japan)

    2004-07-01

    In this paper we present results obtained in collaboration between NSC and Arcelor on the austenite of a multiphased steel and on a 70%Ni-30Fe alloy. The work concerns the formation of the crystallographic textures during the recrystallisation of austenite, since these textures have a strong influence, after the phase transformation, on the forming properties of the sheets. The microstructure and the textures of the austenite and the FeNi alloy were measured with X-Ray diffraction techniques and with EBSD on a high resolution FEG-SEM. (orig.)

  14. Embrittlement behaviour of low-activation alloys with reduced boron content after neutron irradiation

    Science.gov (United States)

    Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2003-09-01

    Ferritic/martensitic steels for fusion applications have been irradiated up to 2.4 dpa in the Petten high flux reactor (HFR); their embrittlement behaviour was investigated by instrumented Charpy-V tests with subsize specimens. The aim of this mid-dose range programme was a comparison of low-activation alloys subjected to different heat treatments and with reduced B contents (down to 2 wt ppm). In the present report, the results of different OPTIFER alloys (Ia, II, IV, V, VI), as obtained in Phases IA and IB of the HFR-irradiation programme (2.4 dpa, at 250-450 °C), are analysed and assessed in comparison to the results of the former MANITU programme. The evaluation clearly shows the eliminated embrittlement problem for the advanced European reduced-activation alloys in comparison to international reference steels. This improvement can be clearly correlated to the reduction of the boron content. Furthermore, the influence of different heat treatments on the impact properties is presented.

  15. Modelling the effect of texture and dislocation structure on irradiation creep of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Christodoulou, N.; Causey, A.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Woo, C.H.; Tome, C.N. [Atomic Energy of Canada Limited, Whiteshell Labs., Pinawa, Manitoba (Canada); Klassen, R.J.; Holt, R.A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1994-07-01

    The effect of texture and dislocation structure on irradiation creep of Zircaloy-2 (irradiated at about 340 K) and Zr-2.5 wt% Nb alloys (irradiated at about 558 K) is studied by means of a self-consistent model. The model relates the creep behaviour of polycrystals to that of single crystals by taking into account the crystallographic texture, dislocation density, grain shape and the intergranular stresses generated due to the crystallographic anisotropy. Three independent creep compliances of the polycrystal obtained from creep tests on a reference material are used to derive the single crystal creep compliances. These are used to calculate the polycrystalline compliances for the remaining materials. At low irradiation temperatures the predicted polycrystalline compliances agree well with the measured values. The observed behaviour can be described by a climb-assisted glide mechanism in which the creep strain is accommodated mainly by prismatic slip with smaller contributions from basal and pyramidal slip systems. At higher irradiation temperatures, the self-consistent approach can also describe well the creep behaviour of Zr-2.5 wt% Nb samples. (author)

  16. Effects of tensile stress on Cu clustering in irradiated Fe–Cu alloy

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, K., E-mail: fujiik@inss.co.jp [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Fukuya, K. [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Kasada, R.; Kimura, A. [Institute of Advanced Energy, Kyoto University, Uji 611-0011 (Japan); Ohkubo, T. [National Institute for Materials Science, Tsukuba 305-0047 (Japan)

    2015-03-15

    Effects of tensile stress on Cu clustering were explained using atom probe tomography (APT) results of Fe–0.6 wt.%Cu alloy specimens irradiated with 6.4 MeV Fe ions while applying a tensile stress of 60 MPa at room temperature (less than 50 °C) and 290 °C. The hardening under the tensile-stressed irradiation was smaller than that under the stress-free irradiation at both room temperature and 290 °C. APT results showed that well-defined Cu clusters were formed in all specimens even under the room temperature irradiation. The Cu clusters under the tensile-stressed condition were smaller and had higher densities than those under the stress-free condition. The lower Cu content in clusters and more diffuse Cu clustering were obtained for the specimens irradiated under the tensile-stressed condition. The hardening efficiency of Cu clusters was correlated with the Cu content in clusters and the coherency of interface between a cluster and the matrix. Application of tensile stress would control hardening by changing the nature of Cu clusters.

  17. Tensile and impact properties of vanadium-base alloys irradiated at low temperatures in the ATR-A1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Nowicki, L.J.; Billone, M.C.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

    1998-03-01

    Subsize tensile and Charpy specimens made from several V-(4-5)Cr-(4-5)Ti alloys were irradiated in the ATR-A1 experiment to study the effects of low-temperature irradiation on mechanical properties. These specimens were contained in lithium-bonded subcapsules and irradiated at temperatures between {approx}200 and 300 C. Peak neutron damage was {approx}4.7 dpa. Postirradiation testing of these specimens has begun. Preliminary results from a limited number of specimens indicate a significant loss of work-hardening capability and dynamic toughness due to the irradiation. These results are consistent with data from previous low-temperature neutron irradiation experiments on these alloys.

  18. Simulation of defect evolution in electron-irradiated dilute FeCr alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz, Christophe J., E-mail: christophe.ortiz@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico - CIEMAT, 28040 Madrid (Spain); Terentyev, Dmitry, E-mail: dterenty@sckcen.be [Institute of Nuclear Materials Science, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Olsson, Paer, E-mail: par.olsson@edf.fr [Department of Materials and Mechanics of Components, EDF R and D, F-77250 Moret-sur-Loing (France); Vila, Rafael, E-mail: rafael.vila@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico - CIEMAT, 28040 Madrid (Spain); Malerba, Lorenzo, E-mail: lmalerba@sckcen.be [Institute of Nuclear Materials Science, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2011-10-01

    A rate theory model based on ab initio data was used to predict defect evolution in electron-irradiated dilute FeCr alloys during isochronal annealing. A good correlation was found between the prediction of the model and existing isochronal resistivity recovery measurements. In agreement with experimental results, our model predicts a shift of stage I{sub E} towards lower temperature with increasing Cr concentration. According to our model, stage II is found to be not only due to the recombination of I{sub 2} clusters with vacancies but also due to the annihilation of ICr and I{sub 2}Cr complexes at vacancies.

  19. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

    Energy Technology Data Exchange (ETDEWEB)

    Onimus, F

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  20. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Gomes, I.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The ATR-A1 irradiation experiment was a collaborative U.S./Japan effort to study at low temperature the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation started in the Advanced Test Reactor (ATR) on November 30, 1995, and ended as planned on May 5, 1996. Total exposure was 132.9 effective full power days (EFPDs) and estimated neutron damage in the vanadium was 4.7 dpa. The vehicle has been discharged from the ATR core and is scheduled to be disassembled in the next reporting period.

  1. Evolution Law of Helium Bubbles in Hastelloy N Alloy on Post-Irradiation Annealing Conditions

    Directory of Open Access Journals (Sweden)

    Jie Gao

    2016-10-01

    Full Text Available This work reports on the evolution law of helium bubbles in Hastelloy N alloy on post-irradiation annealing conditions. After helium ion irradiation at room temperature and subsequent annealing at 600 °C (1 h, the transmission electron microscopy (TEM micrograph indicates the presence of helium bubbles with size of 2 nm in the depth range of 0–300 nm. As for the sample further annealed at 850 °C (5 h, on one hand, a “Denuded Zone” (0–38 nm with rare helium bubbles forms due to the decreased helium concentration. On the other hand, the “Ripening Zone” (38–108 nm and “Coalescence Zone” (108–350 nm with huge differences in size and separation of helium bubbles, caused by different coarsening rates, are observed. The mechanisms of “Ostwald ripening” and “migration and coalescence”, experimentally proved in this work, may explain these observations.

  2. Influence of Surface Roughness on Morphology of Aluminum Alloy After Pulsed-Laser Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sung Ho; Kim, Chung Seok; Jhang, Kyung Young [Hanyang University, Seoul (Korea, Republic of); Shin, Wan Soon [Agency for Defense Development, Daejeon (Korea, Republic of)

    2011-09-15

    The objective of this study is to investigate the influence of surface roughness on the morphology of aluminum 6061- T6 alloy after irradiation with a Nd:YAG pulsed laser. The test specimen was prepared by a polishing process using a diamond paste (1 {mu}m) and emery polishing papers (100, 220, 600, 2400) to obtain different initial surface roughness. After irradiation with ten pulsed-laser shots, the surface morphology was examined by using scanning electron microscopy (SEM), optical microscopy (OM), and atomic force microscopy (AFM). The diameter of the melted zone increased with the surface roughness because the multiple reflections and absorption of the laser beam occurred on the surface because of the surface roughness, so that the absorptance of the laser beam changed. This result was verified using the relative absorptance calculated from the diameter of the melted zone with the surface roughness and the diameter increased with the average surface roughness.

  3. INFLUENCE OF ABNORMAL AUSTENITE GRAIN GRAIN GROWTH IN QUENCHED ABNT 5135 STEEL

    Directory of Open Access Journals (Sweden)

    Camila de Brito Ferreira

    2015-03-01

    Full Text Available Grain size in the steels is a relevant aspect in quenching and tempering heat treatments. It is known that high austenitizing temperature and long time provide an increase in austenitic grain sizes. Likewise, after hardening of low alloy steel, the microstructure consists of martensite and a volume fraction of retained austenite. This paper evaluates the influence of austenite grain size on the volume fraction of retained austenite measured by metallographic analyses and X-ray diffraction. The Mi and Mf temperatures were calculated using an empirical equation and experimentally determined by differential thermal analysis. The mechanical behavior of the steel was evaluated by Vickers microhardness testing. Differently from other results published in the literature that steel hardenability increases with the austenite grain size, it was observed that the increase in austenite grain promotes greater volume fraction of retained austenite after water quenching.

  4. Tensile properties of V-(4-15)Cr-5Ti alloys irradiated at 400{degrees}C in the HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Nowicki, L.; Smith, D.L. [Argonne National Lab., IL (United States)

    1996-10-01

    V-(4-15)Cr-5Ti alloys were irradiated in a helium environment to {approx}10 dpa at {approx}400{degrees}C in the High Flux Isotope Reactor (HFIR). This report presents results of postirradiation tests of tensile properties of V-4Cr-4Ti, V-8Cr-6Ti, V-10Cr-5Ti, and V-15Cr-5Ti. Despite concerns on the effects of transmutation of vanadium to Cr and impurity pickup from the helium environment, all of the alloys exhibited ductile tensile behavior. However, the alloys exhibited ductilities somewhat lower than those of the specimens irradiated to a similar dose and at a similar temperature in an Li environment in fast reactors. Uniform plastic strain in the V-Cr-(4-5)Ti alloys decreased monotonically with increasing Cr content.

  5. Subtask 12F4: Effects of neutron irradiation on the impact properties and fracture behavior of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Loomis, B.A.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    Up-to-date results on the effects of neutron irradiation on the impact properties and fracture behavior of V, V-Ti, V-Cr-Ti and V-Ti-Si alloys are presented in this paper, with an emphasis on the behavior of the U.S. reference alloys V-4Cr-4Ti containing 500-1000 wppm Si. Database on impact energy and cluctile-brittle transition temperature (DBTT) has been established from Charpy impact tests of one-third-size specimens irradiated at 420{degrees}C-600{degrees}C up to {approx}50 dpa in lithium environment in fast fission reactors. To supplement the Charpy impact tests fracture behavior was also characterized by quantitative SEM fractography on miniature tensile and disk specimens that were irradiated to similar conditions and fractured at -196{degrees}C to 200{degrees}C by multiple bending. For similar irradiation conditions irradiation-induced increase in DBTT was influenced most significantly by Cr content, indicating that irradiation-induced clustering of Cr atoms takes place in high-Cr (Cr {ge} 7 wt.%) alloys. When combined contents of Cr and Ti were {le}10 wt.%, effects of neutron irradiation on impact properties and fracture behavior were negligible. For example, from the Charpy-impact and multiple-bend tests there was no indication of irradiation-induced embrittlement for V-5Ti, V-3Ti-1Si and the U.S. reference alloy V-4Cr-4Ti after irradiation to {approx}34 dpa at 420{degrees}C to 600{degrees}C, and only ductile fracture was observed for temperatures as low as -196{degrees}C. 14 refs., 8 figs., 1 tab.

  6. Development of Semi-Stochastic Algorithm for Optimizing Alloy Composition of High-Temperature Austenitic Stainless Steels (H-Series) for Desired Mechanical and Corrosion Properties.

    Energy Technology Data Exchange (ETDEWEB)

    Dulikravich, George S.; Sikka, Vinod K.; Muralidharan, G.

    2006-06-01

    The goal of this project was to adapt and use an advanced semi-stochastic algorithm for constrained multiobjective optimization and combine it with experimental testing and verification to determine optimum concentrations of alloying elements in heat-resistant and corrosion-resistant H-series stainless steel alloys that will simultaneously maximize a number of alloy's mechanical and corrosion properties.

  7. 沉淀强化奥氏体合金的氢致断裂行为%Study on Behaviors of Hydrogen-induced Fracture of Precipitation Strengthened Austenitic Alloy

    Institute of Scientific and Technical Information of China (English)

    李忠文; 赵明久; 戎利建

    2012-01-01

    Correlation of hydrogen embritt/ement sensitivity and fracture behavior was investigated by means of tensile tests and fractographic examination in a precipitation strengthened austenitic alloy with and without hydrogen. Additionally, slip bands during tensile deformation were observed in order to determine the effect of hydrogen on localized plasticity and microcrack nucleation. The results show that the fracture mode of the precipitation strengthened austenitic alloy exhibits dramatic transition from dimple fracture in uncharged specimens to the mixed mode with dimple, intergranular and slip band fracture in charged specimens. As for the reason, it can be related to not only the slip planarity and localization induced by hydrogen but also the dislocation pile-up and hydrogen accumulation formed at the sites of grain boundaries, twin boundaries and intersecting slip bands.%研究了未充氢和热充氢沉淀强化奥氏体合金的拉伸断裂行为,分析了其氢脆敏感性与拉伸断裂行为间的联系,研究了氢对合金局部塑性变形及微裂纹形核的影响。结果表明:氢使沉淀强化合金由单一的韧窝断裂转变为韧窝断裂、沿晶断裂和滑移带开裂的混合断裂方式。其原因是:一方面,氢促进位错平面化滑移趋势、加剧局部塑性变形;另一方面,滑移带被晶界、孪晶界以及不同取向的滑移带所阻碍,引起了位错塞积和氢聚集。

  8. Effect of alloying and heat treatment on the structure and tribological properties of nitrogen-bearing stainless austenitic steels under abrasive and adhesive wear

    Science.gov (United States)

    Korshunov, L. G.; Goikhenberg, Yu. N.; Chernenko, N. K.

    2007-05-01

    The effect of nitrogen, silicon, and aging modes on the structure, resistance to abrasive and adhesive wear, friction factor, and mechanical properties of nitrogen-bearing (0.27-0.83% N) chromium-manganese austenitic steels is studied. It is shown that it is possible to ensure a favorable combination of mechanical and tribological properties in such steels by choosing the appropriate chemical composition and aging mode.

  9. Phase decomposition of AuFe alloy nanoparticles embedded in silica matrix under swift heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Pannu, Compesh, E-mail: Compesh@gmail.com [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi (India); Bala, Manju; Singh, U.B. [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi (India); Srivastava, S.K. [Department of Physics and Astronomy, Indian Institute of Technology Kharagpur, Kharagpur (India); Kabiraj, D. [Inter University Accelerator Centre, Aruna Asaf Ali Marg, New Delhi (India); Avasthi, D.K. [Amity University, Noida 201313, Uttar Pradesh (India)

    2016-07-15

    AuFe alloy nanoparticles embedded in silica matrix are synthesized using atom beam sputtering technique and subsequently irradiated with 100 MeV Au ions at various fluences ranging from 1 × 10{sup 13} to 6 × 10{sup 13} ions/cm{sup 2}. The X-ray diffraction, absorption spectroscopy, X-ray photo electron spectroscopy and transmission electron microscopy results show that swift heavy ion irradiation leads to decomposition of AuFe alloy nanoparticles from surface region and subsequent reprecipitation of Au and Fe nanoparticles occur. The process of phase decomposition and reprecipitation of individual element nanoparticles is explained on the basis of inelastic thermal spike model.

  10. Void structure and density change of vanadium-base alloys irradiated in the dynamic helium charging experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Nowicki, L.; Gazda, J. [Argonne National Lab., Chicago, IL (United States)] [and others

    1995-04-01

    The objective of this work is to determine void structure, distribution, and density changes of several promising vanadium-base alloys irradiated in the Dynamic Helium Charging Experiment (DHCE). Combined effects of dynamically charged helium and neutron damage on density change, void distribution, and microstructural evolution of V-4Cr-4Ti alloy have been determined after irradiation to 18-31 dpa at 425-600{degree}C in the DHCE, and the results compared with those from a non-DHCE in which helium generation was negligible.

  11. Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

    Science.gov (United States)

    Field, Kevin G.; Briggs, Samuel A.; Hu, Xunxiang; Yamamoto, Yukinori; Howard, Richard H.; Sridharan, Kumar

    2017-01-01

    FeCrAl alloys are an attractive class of materials for nuclear power applications because of their increased environmental compatibility compared with more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300 and 400 °C have shown post-irradiation microstructures containing dislocation loops and a Cr-rich α‧ phase. Although these initial studies established the post-irradiation microstructures, there was little to no focus on understanding the influence of pre-irradiation microstructures on this response. In this study, a well-annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 displacements per atom (dpa) at 382 °C and then the effect of random high-angle grain boundaries on the spatial distribution and size of a dislocation loops, a/2 dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with a/2 dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and a dislocation loops exhibiting an increased size in the vicinity of the grain boundary. These results suggest the importance of the pre-irradiation microstructure and, specifically, defect sink density spacing to the radiation tolerance of FeCrAl alloys.

  12. Characterization of ion-irradiated ODS Fe–Cr alloys by doppler broadening spectroscopy using a positron beam

    Energy Technology Data Exchange (ETDEWEB)

    Parente, P.; Leguey, T. [Departamento de Física and IAAB, Universidad Carlos III de Madrid, 28911 Leganés (Spain); Castro, V. de, E-mail: vanessa.decastro@uc3m.es [Departamento de Física and IAAB, Universidad Carlos III de Madrid, 28911 Leganés (Spain); Gigl, T.; Reiner, M.; Hugenschmidt, C. [FRM II and Physics Department, Technische Universität München, 85747 Garching (Germany); Pareja, R. [Departamento de Física and IAAB, Universidad Carlos III de Madrid, 28911 Leganés (Spain)

    2015-09-15

    The damage profile of oxide dispersion strengthened steels after single-, or simultaneous triple-ion irradiation at different conditions has been characterized using a low energy positron beam in order to provide information on microstructural changes induced by irradiation. Doppler broadening and coincident Doppler broadening measurements of the positron annihilation line have been performed on different Fe–Cr–(W,Ti) alloys reinforced with Y{sub 2}O{sub 3}, to identify the nature and stability of irradiation-induced open-volume defects and their possible association with the oxide nanoparticles. It was found that irradiation induced vacancy clusters are associated with Cr atoms. The results are of highest interest for modeling the damage induced by 14 MeV neutrons in reduced activation Fe–Cr alloys relevant for fusion devices.

  13. High Mn austenitic stainless steel

    Science.gov (United States)

    Yamamoto, Yukinori [Oak Ridge, TN; Santella, Michael L [Knoxville, TN; Brady, Michael P [Oak Ridge, TN; Maziasz, Philip J [Oak Ridge, TN; Liu, Chain-tsuan [Knoxville, TN

    2010-07-13

    An austenitic stainless steel alloy includes, in weight percent: >4 to 15 Mn; 8 to 15 Ni; 14 to 16 Cr; 2.4 to 3 Al; 0.4 to 1 total of at least one of Nb and Ta; 0.05 to 0.2 C; 0.01 to 0.02 B; no more than 0.3 of combined Ti+V; up to 3 Mo; up to 3 Co; up to 1W; up to 3 Cu; up to 1 Si; up to 0.05 P; up to 1 total of at least one of Y, La, Ce, Hf, and Zr; less than 0.05 N; and base Fe, wherein the weight percent Fe is greater than the weight percent Ni, and wherein the alloy forms an external continuous scale including alumina, nanometer scale sized particles distributed throughout the microstructure, the particles including at least one of NbC and TaC, and a stable essentially single phase FCC austenitic matrix microstructure that is essentially delta-ferrite-free and essentially BCC-phase-free.

  14. Gas accumulation at grain boundaries during 800 MeV proton irradiation of aluminium and aluminium-alloys

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Horsewell, Andy; Sommer, W. F.;

    1986-01-01

    Samples of pure aluminium (99.9999%) and commercial Al-2.7%Mg(AlMg3) and Al-1.1%Mg-0.5%Si(Al6061) alloys were irradiated with 800 MeV protons at the Los Alamos Meson Physics Facility (LAMPF) at a temperature between 40-100°C to a maximum dose of 0.2 dpa. Transmission electron microscopy (TEM......) showed a complete absence of voids or bubbles in the grain interiors of the aluminium and the aluminium-alloys. Bubbles were clearly visible by TEM at grain boundaries in pure Al and the AlMg3 alloy; but bubbles were not visible in the Al6061 alloy. The bubble density in the AlMg3 alloy was considerably...

  15. High strength, tough alloy steel

    Science.gov (United States)

    Thomas, Gareth; Rao, Bangaru V. N.

    1979-01-01

    A high strength, tough alloy steel is formed by heating the steel to a temperature in the austenite range (1000.degree.-1100.degree. C.) to form a homogeneous austenite phase and then cooling the steel to form a microstructure of uniformly dispersed dislocated martensite separated by continuous thin boundary films of stabilized retained austenite. The steel includes 0.2-0.35 weight % carbon, at least 1% and preferably 3-4.5% chromium, and at least one other substitutional alloying element, preferably manganese or nickel. The austenite film is stable to subsequent heat treatment as by tempering (below 300.degree. C.) and reforms to a stable film after austenite grain refinement.

  16. Microstructural changes of Y-doped V-4Cr-4Ti alloys after ion and neutron irradiation

    Directory of Open Access Journals (Sweden)

    H. Watanabe

    2016-12-01

    Full Text Available High-purity Y-doped V-4Cr-4Ti alloys (0.1–0.2wt. % Y, manufactured by the National Institute for Fusion Science (NIFS, were used for this study. Heavy-ion and fission-neutron irradiation was carried out at temperatures 673–873K. During the ion irradiation at 873K, the microstructure was controlled by the formation of Ti(C,O,N precipitates lying on the (100 plane. Y addition effectively suppressed the growth of Ti(C,O,N precipitates, especially at lower dose irradiation to up to 4 dpa. However, at higher dose levels (12.0 dpa, the number density was almost at the same levels irrespective of the presence of Y. After neutron irradiation at 873K, fine titanium oxides were also observed in all V alloys. However, smaller oxide sizes were observed in the Y-doped samples under the same irradiation conditions. The detailed analysis of EDS showed that the center of the Ti(C,O,N precipitates was mainly enriched by nitrogen. The results showed that the contribution of not only oxygen atoms picked up from the irradiation environment but also nitrogen atoms is essential to understand the microstructural evolution of V-4Cr-4Ti-Y alloys.

  17. Effect of irradiation temperature on crystallization of {alpha}-Fe induced by He irradiations in Fe{sub 80}B{sub 20} amorphous alloy

    Energy Technology Data Exchange (ETDEWEB)

    San-noo, Toshimasa; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi [Musashi Inst. of Tech., Tokyo (Japan); Hayashi, Nobuyuki; Sakamoto, Isao

    1997-03-01

    Since amorphous alloys are generally highly resistant to irradiation and their critical radiation dose is an order of magnitude higher for Fe-B amorphous alloy than Mo-methods, these alloys are expected to become applicable as for fusion reactor materials. The authors investigated {alpha}-Fe crystallization in an amorphous alloy, Fe{sub 80}B{sub 20} using internal conversion electron Moessbauer spectroscopy. The amount of {alpha}-Fe component was found to increase by raising the He-irradiation dose. The target part was modified to enable He ion radiation at a lower temperature (below 400 K) by cooling with Peltier element. Fe{sub 80}B{sub 20} amorphous alloy was cooled to keep the temperature at 300 K and exposed to 40 keV He ion at 1-3 x 10{sup 8} ions/cm{sup 2}. The amount of {alpha}-Fe crystal in each sample was determined. The crystal formation was not observed for He ion radiation below 2 x 10{sup 18} ions/cm{sup 2}, but that at 3 x 10{sup 8} ions/ cm{sup 2} produced a new phase ({delta} +0.40 mm/sec, {Delta} = 0.89 mm/sec). The decrease in the radiation temperature from 430 to 300 K resulted to extremely repress the production of {alpha}-Fe crystal, suggesting that the crystallization induced by He-radiation cascade is highly depending on the radiation temperature. (M.N.)

  18. Irradiation-induced precipitation and mechanical properties of vanadium alloys at <430 C

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Gazda, J.; Smith, D.L. [Argonne National Lab., IL (United States)

    1998-09-01

    Recent attention to V-base alloys has focused on the effect of low-temperature (<430 C) irradiation on tensile and impact properties of V-4Cr-4Ti. In previous studies, dislocation channeling, which causes flow localization and severe loss of work-hardening capability, has been attributed to dense, irradiation-induced precipitation of very fine particles. However, efforts to identify the precipitates were unsuccessful until now. In this study, analysis by transmission electron microscopy (TEM) was conducted on unalloyed V, V-5Ti, V-3Ti-1Si, and V-4Cr-4Ti specimens that were irradiated at <430 C in conventional and dynamic helium charging experiments. By means of dark-field imaging and selected-area-diffraction analysis, the characteristic precipitates were identified to be (V,Ti{sub 1{minus}x})(C,O,N). In V-3Ti-1Si, precipitation of (V,Ti{sub 1{minus}x})(C,O,N) was negligible at <430 C, and as a result, dislocation channeling did not occur and work-hardening capability was high.

  19. Stable atomic structure of NiTi austenite

    Energy Technology Data Exchange (ETDEWEB)

    Zarkevich, Nikolai A [Ames Laboratory; Johnson, Duane D [Ames Laboratory

    2014-08-01

    Nitinol (NiTi), the most widely used shape-memory alloy, exhibits an austenite phase that has yet to be identified. The usually assumed austenitic structure is cubic B2, which has imaginary phonon modes, hence it is unstable. We suggest a stable austenitic structure that “on average” has B2 symmetry (observed by x-ray and neutron diffraction), but it exhibits finite atomic displacements from the ideal B2 sites. The proposed structure has a phonon spectrum that agrees with that from neutron scattering, has diffraction spectra in agreement with x-ray diffraction, and has an energy relative to the ground state that agrees with calorimetry data.

  20. Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kohnert, Aaron A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dasgupta, Dwaipayan [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-23

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensure that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.

  1. Effect of neutron irradiation and postradiation annealing on the microstructure and properties of an Al-Mg-Si alloy

    Science.gov (United States)

    Maksimkin, O. P.; Tsai, K. V.; Rofman, O. V.; Sil'nyagina, N. S.

    2016-09-01

    The effect of long-term neutron irradiation and postradiation thermal-induced aging on the microstructure and mechanical properties of an aluminum-based reactor Al-Mg-Si alloy grade SAV-1 has been studied. The material under study is the shell of an automatic fine-control rod used to control the reactivity of the core of a VVR-K research reactor. Successive 1-h annealings of specimens of the SAV-1 alloy irradiated to doses of 0.001 and 5 dpa in the temperature range of 100-550°C have been carried out. The evolution of the fine structure of the material and changes in its mechanical characteristics have been studied. The phenomenon of the acceleration of the aging of the SAV-1 alloy under the effect of a high neutron fluence at an irradiation temperature of 80°C has been observed, which involves the formation of numerous lineage (stitch) Guinier-Preston zones in the alloy. It has been shown that the strength characteristics of the SAV-1 alloy depend significantly on the degree of its radiation- and thermal-induced aging.

  2. Final report on characterization of physical and mechanical properties of copper and copper alloys before and after irradiation

    DEFF Research Database (Denmark)

    Singh, B.N.; Tähtinen, S.

    2002-01-01

    The present report summarizes and highlights the main results of the work carried out during the last 5-6 years on effects of neutron irradiation on physical and mechanical properties of copper and copper alloys. The work was an European contribution toITER Research and Development programme...... amount of further effort is needed to find a realistic and optimum solution....

  3. A generalized Ising model for studying alloy evolution under irradiation and its use in kinetic Monte Carlo simulations

    Science.gov (United States)

    Huang, Chen-Hsi; Marian, Jaime

    2016-10-01

    We derive an Ising Hamiltonian for kinetic simulations involving interstitial and vacancy defects in binary alloys. Our model, which we term ‘ABVI’, incorporates solute transport by both interstitial defects and vacancies into a mathematically-consistent framework, and thus represents a generalization to the widely-used ABV model for alloy evolution simulations. The Hamiltonian captures the three possible interstitial configurations in a binary alloy: A-A, A-B, and B-B, which makes it particularly useful for irradiation damage simulations. All the constants of the Hamiltonian are expressed in terms of bond energies that can be computed using first-principles calculations. We implement our ABVI model in kinetic Monte Carlo simulations and perform a verification exercise by comparing our results to published irradiation damage simulations in simple binary systems with Frenkel pair defect production and several microstructural scenarios, with matching agreement found.

  4. Formation of Pt-Zn Alloy Nanoparticles by Electron-Beam Irradiation of Wurtzite ZnO in the TEM

    Science.gov (United States)

    Lee, Sung Bo; Park, Jucheol; van Aken, Peter A.

    2016-07-01

    As is well documented, platinum nanoparticles, promising for catalysts for fuel cells, exhibit better catalytic activities, when alloyed with Zn. Pre-existing syntheses of Pt-Zn alloy catalysts are composed of a number of complex steps. In this study, we have demonstrated that nanoparticles of Pt-Zn alloys are simply generated by electron-beam irradiation in a transmission electron microscope of a wurtzite ZnO single-crystal specimen. The initial ZnO specimen is considered to have been contaminated by Pt during specimen preparation by focused ion beam milling. The formation of the nanoparticle is explained within the framework of ionization damage (radiolysis) by electron-beam irradiation and accompanying electrostatic charging.

  5. Thermal substructure of hot deformed austenite substructure

    Energy Technology Data Exchange (ETDEWEB)

    Bernshtejn, M.L.; Kaputkina, L.M.; Nikishov, N.A. (Moskovskij Inst. Stali i Splavov (USSR))

    1982-01-01

    Effect of hot working different regimes on formation of austenite structure and substructure of the 60N20 and 60Kh5G6 steels and kinetics of softening processes at postdeformation isothermal (at deformation temperature) heating, is investigated. It is shown, that variation of hot working regimes permits to obtain a wide range of structural and substructural austenite conditions. Rate decrease and temperature increase promotes obtaining after hot working and conservation under cooling conditions of developed polygonized substructure. Similar polygonized isotropic substructure with a rather low density of dislocations inside of subgrains promotes decelerating of initial stages of recrystallization development under conditions of hot working and regulated post-deformation heatings. Alloying by carbide-forming elements (chromium and manganese) delays development of recrystallization (in comparison with alloying with nickel), even if the steel is in the condition of single-phase solid solution.

  6. Pitting corrosion resistant austenite stainless steel

    Science.gov (United States)

    van Rooyen, D.; Bandy, R.

    A pitting corrosion resistant austenite stainless steel comprises 17 to 28 wt. % chromium, 15 to 26 wt. % nickel, 5 to 8 wt. % molybdenum, and 0.3 to 0.5 wt. % nitrogen, the balance being iron, unavoidable impurities, minor additions made in the normal course of melting and casting alloys of this type, and may optionally include up to 10 wt. % of manganese, up to 5 wt. % of silicon, and up to 0.08 wt. % of carbon.

  7. Mechanical characteristics and swelling of austenitic Fe-Cr-Mn steels irradiated in the SM-2 and BOR-60 reactors. [0. 4C-12Cr-19Mn-2Ni-Mo-N; 0. 4C-12Cr-14Mn-5Ni-Mo-2Al-B; 0. 4C-17Cr-17Mn-Cu-Mo-Nb-N; Fe-Cr-Ni steel: 0. 8C-16Cr-15Ni-3Mo-Nb; 316; 304

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K.; Bulanova, T.M.; Neustroev, V.S. (Lenin (V.I.) Research Inst. of Atomic Reactors, Dimitrovgrad (USSR)); Ivanov, L.I.; Djomina, E.V.; Platov, Yu.M. (AN SSSR, Moscow (USSR). A.A. Baikov Inst. of Metallurgy)

    Three types of austenitic Fe-Cr-Mn stainless steels were irradiated simultaneously with Fe-Cr-Ni austenitic steel at temperatures from 400 to 800deg C in the mixed spectrum of the high flux SM-2 reactor to 10 dpa and 700 appm of He and in the BOR-60 reactor to 60 dpa without He generation. The paper presents the swelling and mechanical properties of steels irradiated in the BOR-60 and SM-2 as a function of the concentration of transmuted He and the value of atomic displacement. (orig.).

  8. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Science.gov (United States)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  9. Recent results on the neutron irradiation of ITER candidate copper alloys irradiated in DR-3 at 250{degrees}C to 0.3 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M.

    1997-04-01

    Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment with additional specimens re-aged and given a reactor bakeout treatment at 350{degrees}C for 100 h. CuAl-25 was also heat treated to simulate the effects of a bonding thermal cycle on the material. A number of heat treated specimens were neutron irradiated at 250{degrees}C to a dose level of {approximately}0.3 dpa in the DR-3 reactor as Riso. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250{degrees}C showed a severe loss of ductility in the case of the CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens exhibited a reasonable amount of uniform elongation, with CuCrZr possessing a lower strength.

  10. Structural and chemical evolution in neutron irradiated and helium-injected ferritic ODS PM2000 alloy

    Science.gov (United States)

    Jung, Hee Joon; Edwards, Dan J.; Kurtz, Richard J.; Yamamoto, Takuya; Wu, Yuan; Odette, G. Robert

    2017-02-01

    An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron 58Ni(nth,γ) 59Ni(nth,α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al2YO3 oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

  11. Structural and chemical evolution in neutron irradiated and helium-injected ferritic ODS PM2000 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hee Joon; Edwards, Dan J.; Kurtz, Richard J.; Yamamoto, Takuya; Wu, Yuan; Odette, G. Robert

    2017-02-01

    An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron 59Ni(nth, 59Ni(nth,α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al2YO3 oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

  12. {sup 197}Au irradiation study of phase-change memory cell with GeSbTe alloy

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Liangcai; Song, Zhitang; Lian, Jie; Rao, Feng; Liu, Bo; Song, Sannian; Liu, Weili; Feng, Songlin [State Key Laboratory of Functional Materials for Informatics, Laboratory of Nanotechnology, Shanghai Institute of Micro-system and Information Technology, Chinese Academy of Sciences, Shanghai 200050 (China); Zhou, Xilin; Liu, Xuyan [State Key Laboratory of Functional Materials for Informatics, Laboratory of Nanotechnology, Shanghai Institute of Micro-system and Information Technology, Chinese Academy of Sciences, Shanghai 200050 (China); Graduate School of the Chinese Academy of Sciences, Beijing 100080 (China)

    2010-10-15

    A {sup 197}Au ion source was used to irradiate a Ge{sub 2}Sb{sub 2}Te{sub 5}-alloy-based phase-change memory (PCM) cell to study the ion-irradiation effect on the properties of the cell. The PCM devices with the tungsten (W) heating electrode of 260 nm diameter were fabricated by 0.18 {mu}m complementary metal-oxide-semiconductor (CMOS) technology. Four different doses (10{sup 10}, 10{sup 11}, 10{sup 12}, and 5 x 10{sup 12} ions/cm{sup 2}, respectively) were applied to irradiate the PCM cell. The samples before and after irradiation were characterized by current-voltage and resistance measurements at room temperature. It is found that the cell properties (resistance value of the amorphous and crystalline states, threshold voltage, and current for phase transition, etc.) have hardly changed, even for the sample irradiated up to 10{sup 12} ions/cm{sup 2} dose, and the cell still has good set-reset operation ability (above 10{sup 5} cycles). Furthermore, the resistance ratio remains at 1000 even after 10{sup 5} cycles of the set-reset operation. The results show the PCM cell with Ge{sub 2}Sb{sub 2}Te{sub 5} alloy has a strong ion-irradiation tolerance. (Abstract Copyright [2010], Wiley Periodicals, Inc.)

  13. Microstructure of V-4Cr-4Ti alloy after low-temperature irradiation by ions and neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Gazda, J.; Meshii, M. [Northwestern Univ. (United States); Chung, H.M. [Argonne National Lab., IL (United States)

    1998-03-01

    Mechanical properties of V-4Cr-4Ti alloy were investigated after low-temperature (<420 C) irradiation. The effects of fast neutrons at 390 C were investigated by irradiation to {approx}4 dpa in the X530 experiment in the EBR-II reactor; these tests were complemented by irradiation with single (4.5-MeV Ni{sup ++}) and dual ion beams (350-keV He{sup +} simultaneously with 4.5-MeV Ni{sup ++}). TEM observations showed the formation of a high density of point-defect clusters and dislocation loops (<30 nm diameter) distributed uniformly in the specimens. Mechanical-property testing showed embrittlement of the alloy. TEM investigations of deformed microstructures were used to determine the causes of embrittlement and yielded observation of dislocation channels propagating through the undeformed matrix. Channels are the sole slip paths and cause early onset of necking and loss of work-hardening in this alloy. Based on a review of the available literature, suggestions are made for further research of slip localization in V-base alloys.

  14. Effect of Temperature on the Deformation Behavior of B2 Austenite in a Polycrystalline Ni49.9Ti50.1 (at.Percent) Shape Memory Alloy

    Science.gov (United States)

    Garg, A.; Benafan, O.; Noebe, R. D.; Padula, S. A., II; Clausen, B.; Vogel, S.; Vaidyanathan, R.

    2013-01-01

    Superelasticity in austenitic B2-NiTi is of great technical interest and has been studied in the past by several researchers [1]. However, investigation of temperature dependent deformation in B2-NiTi is equally important since competing mechanisms of stress-induced martensite (SIM), retained martensite, plastic and deformation twinning can lead to unusual mechanical behaviors. Identification of the role of various mechanisms contributing to the overall deformation response of B2-NiTi is imperative to understanding and maturing SMA-enabled technologies. Thus, the objective of this work was to study the deformation of polycrystalline Ni49.9Ti50.1 (at. %) above A(sub f) (105 C) in the B2 state at temperatures between 165-440 C, and generate a B2 deformation map showing active deformation mechanisms in different temperature-stress regimes.

  15. Enhancement in anomalous Hall resistivity of Co/Pd multilayer and CoPd alloy by Ga+ ion irradiation

    KAUST Repository

    Guo, Zaibing

    2014-02-01

    In this paper, we report the effect of Ga+ ion irradiation on anomalous Hall effect (AHE) and longitudinal resistivity (ρxx) in [Co(3 Å)/Pd(5 Å)]80 multilayer and Co 42Pd58 alloy. 4- and 2-fold increases in anomalous Hall resistivity (ρAH) in the Co/Pd multilayer and CoPd alloy have been observed after irradiations at doses of 2.4 × 1015 and 3.3×10 15 ions/cm2, respectively. Skew scattering and side jump contributions to AHE have been analyzed based on the scaling relationship ρAH = aρxx + bρ2xx. For the Co/Pd multilayer, AHE is mainly affected by ion irradiation-induced interface diffusion and defects. For the CoPd alloy, the increase in doses above 1.5 × 1015 ions/cm2 induces a sign change in skew scattering, followed by the skew scattering contribution to AHE overwhelming the side jump contribution, this phenomenon should be attributed to irradiation-induced defects and modifications in chemical ordering. © Copyright EPLA, 2014.

  16. Enhancement in anomalous Hall resistivity of Co/Pd multilayer and CoPd alloy by Ga+ ion irradiation

    Science.gov (United States)

    Guo, Z. B.; Mi, W. B.; Li, J. Q.; Cheng, Y. C.; Zhang, X. X.

    2014-02-01

    In this paper, we report the effect of Ga+ ion irradiation on anomalous Hall effect (AHE) and longitudinal resistivity (\\rho_{\\textit{xx}}) in [Co(3 Å)/Pd(5 Å)]80 multilayer and Co42Pd58 alloy. 4- and 2-fold increases in anomalous Hall resistivity (\\rho_{\\textit{AH}}) in the Co/Pd multilayer and CoPd alloy have been observed after irradiations at doses of 2.4\\times 10^{15} and 3.3\\times 10^{15}\\ \\text{ions/cm}^{2} , respectively. Skew scattering and side jump contributions to AHE have been analyzed based on the scaling relationship \\rho_{\\textit{AH}}=a\\rho_{\\textit{xx}}+b\\rho_{\\textit{xx}}^{2} . For the Co/Pd multilayer, AHE is mainly affected by ion irradiation-induced interface diffusion and defects. For the CoPd alloy, the increase in doses above 1.5\\times 10^{15}\\ \\text{ions/cm}^{2} induces a sign change in skew scattering, followed by the skew scattering contribution to AHE overwhelming the side jump contribution, this phenomenon should be attributed to irradiation-induced defects and modifications in chemical ordering.

  17. Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

    Energy Technology Data Exchange (ETDEWEB)

    D D. Keiser, Jr.; A. B. Robinson; D. E. Janney; J. F. Jue

    2008-03-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. As part of this development, reactor experiments are being conducted in the Advanced Test Reactor to determine the irradiation performance of different dispersion fuels that contain U-Mo alloys with different Mo contents and Al alloy matrices with different Si contents. Of particular interest is the performance of the dispersion fuels depending on the Si content of the Al alloy matrix, since the addition of Si is being looked to for improving the performance of these dispersion fuels. This paper will describe the results of recent microstructural examinations that have been performed using optical metallography and scanning electron microscopy on as-fabricated and as-irradiated dispersion fuels with different amounts of Si added to the Al matrix. Differences in the microstructural development during irradiation as a function of the Si content in the Al matrix will be discussed, and comments will be made about the development and stability of the fuel/matrix interaction layers that are commonly present in irradiated dispersion fuels.

  18. Radiation effects on microstructure and hardness of a titanium aluminide alloy irradiated by helium ions at room and elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Tao, E-mail: tao@ansto.gov.au [Institute of Materials Engineering, Australian Nuclear Science & Technology Organisation, Locked Bag 2001, Kirrawee DC, Sydney, NSW 2232 (Australia); Zhu, Hanliang [Institute of Materials Engineering, Australian Nuclear Science & Technology Organisation, Locked Bag 2001, Kirrawee DC, Sydney, NSW 2232 (Australia); Ionescu, Mihail [Institute for Environment Research, Australian Nuclear Science & Technology Organisation, Locked Bag 2001, Kirrawee DC, Sydney, NSW 2232 (Australia); Dayal, Pranesh; Davis, Joel; Carr, David; Harrison, Robert; Edwards, Lyndon [Institute of Materials Engineering, Australian Nuclear Science & Technology Organisation, Locked Bag 2001, Kirrawee DC, Sydney, NSW 2232 (Australia)

    2015-04-15

    A 45XD TiAl alloy possessing a lamellar microstructure was irradiated using 5 MeV helium ions to a fluence of 5 × 10{sup 21} ion m{sup −2} (5000 appm) with a dose of about 1 dpa (displacements per atom). A uniform helium ion stopping damage region about 17 μm deep from the target surface was achieved by applying an energy degrading wheel. Radiation damage defects including helium-vacancy clusters and small helium bubbles were found in the microstructure of the samples irradiated at room temperature. With increasing irradiation temperature to 300 °C and 500 °C helium bubbles were clearly observed in both the α{sub 2} and γ phases of the irradiated microstructure. By means of nanoindentation significant irradiation hardening was measured. For the samples irradiated at room temperature the hardness increased from 5.6 GPa to 8.5 GPa and the irradiation-hardening effect reduced to approximately 8.0 GPa for the samples irradiated at 300 °C and 500 °C.

  19. Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Jr, D D; Robinson, A B; Janney, D E; Jue, J F

    2008-03-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with either Al-0.2Si or 4043 Al (~4.8% Si) alloy matrix in the as-fabricated and/or as-irradiated condition using optical metallography and/or scanning electron microscopy. Fuel plates with either matrix can have Si-rich layers around the U-7Mo particles after fabrication, and during irradiation these layers were observed to grow in thickness and to become Si-deficient in some areas of the fuel plates. For the fuel plates with 4043 Al, this was observed in fuel plate areas that were exposed to very aggressive irradiation conditions.

  20. Bactericidal and antimicrobial effects of pure titanium and titanium alloy treated with short-term, low-energy UV irradiation

    Science.gov (United States)

    Narita, K.; Ono, A.; Wada, K.; Tanaka, T.; Kumagai, G.; Yamauchi, R.; Nakane, A.; Ishibashi, Y.

    2017-01-01

    Objectives The surface of pure titanium (Ti) shows decreased histocompatibility over time; this phenomenon is known as biological ageing. UV irradiation enables the reversal of biological ageing through photofunctionalisation, a physicochemical alteration of the titanium surface. Ti implants are sterilised by UV irradiation in dental surgery. However, orthopaedic biomaterials are usually composed of the alloy Ti6Al4V, for which the antibacterial effects of UV irradiation are unconfirmed. Here we evaluated the bactericidal and antimicrobial effects of treating Ti and Ti6Al4V with UV irradiation of a lower and briefer dose than previously reported, for applications in implant surgery. Materials and Methods Ti and Ti6Al4V disks were prepared. To evaluate the bactericidal effect of UV irradiation, Staphylococcus aureus 834 suspension was seeded onto the disks, which were then exposed to UV light for 15 minutes at a dose of 9 J/cm2. To evaluate the antimicrobial activity of UV irradiation, bacterial suspensions were seeded onto the disks 0, 0.5, one, six, 24 and 48 hours, and three and seven days after UV irradiation as described above. In both experiments, the bacteria were then harvested, cultured, and the number of colonies were counted. Results No colonies were observed when UV irradiation was performed after the bacteria were added to the disks. When the bacteria were seeded after UV irradiation, the amount of surviving bacteria on the Ti and Ti6Al4V disks decreased at 0 hours and then gradually increased. However, the antimicrobial activity was maintained for seven days after UV irradiation. Conclusion Antimicrobial activity was induced for seven days after UV irradiation on both types of disk. Irradiated Ti6Al4V and Ti had similar antimicrobial properties. Cite this article: T. Itabashi, K. Narita, A. Ono, K. Wada, T. Tanaka, G. Kumagai, R. Yamauchi, A. Nakane, Y. Ishibashi. Bactericidal and antimicrobial effects of pure titanium and titanium alloy treated with

  1. An austenitic steel for fuel cladding tubes and core components of LMFBR`s with high ductility after neutron irradiation; Ein austenitischer Stahl fuer Huellrohre und Kernkomponenten natriumgekuehlter Brueter mit hoher Duktilitaet nach Neutronenbestrahlung

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, L.; Kempe, H.

    1994-06-01

    Two heats of an austenitic stainless steel with different priority concerning the resistance against Helium-embrittlement (B801) and void-swelling (F218) had been developed and tested as a material for fuel rod claddings and core components of liquid metal fast breeder reactors. The two steels show a ductility five times higher than the reference steel 1.4970 in tensile - and creep-rupture-tests after irradiation in reactors with fast and mixed neutron flux respectively. Just so the swelling resistance had been confirmed up to 40 dpa. Checked claddings of the heat F218 in the dimensions 6x0.38 mm, 6.55x0.45 mm and 7.6x0.5 mm are available for pin- and bundle irradiation experiments. (orig.) [Deutsch] Im Rahmen der Entwicklung austenitischer Staehle als Werkstoffe fuer Huellrohre und Kernkomponenten Schneller Natriumgekuehlter Brutreaktoren wurden zwei Chargen mit unterschiedlicher Prioritaet fuer ihre Widerstandsfaehigkeit gegen Heliumversproedung (B801) und Porenschwellen (F218) konzipiert und untersucht. Beide Staehle zeigten nach Bestrahlung in Reaktoren mit schnellem bzw. gemischtem Neutronenfluss sowohl im Warmzugversuch als auch im Zeitstandversuch eine Duktilitaet, die um den Faktor 5 hoeher liegt als die des Referenzstahles 1.4970. Fuer beide Staehle konnte die Schwellresistenz bis 40 dpa Neutronenbestrahlung nachgewiesen werden. Fuer Brennstab- und Buendelbestrahlungsexperimente stehen gepruefte Huellrohre der Charge F218 mit den Abmessungen 6x0.38 mm, 6.55x0.45 mm und 7.6x0.5 mm zur Verfuegung. (orig.)

  2. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States); Tsai, H.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  3. Radiation-induced segregation and corrosion behavior on Σ3 coincidence site lattice and random grain boundaries in proton-irradiated type-316L austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Sakaguchi, N., E-mail: sakaguchi@eng.hokudai.ac.jp [Center for Advanced Research of Energy and Materials, Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Hokkaido (Japan); Endo, M.; Watanabe, S. [Center for Advanced Research of Energy and Materials, Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Hokkaido (Japan); Kinoshita, H. [Fukushima National College of Technology, Iwaki 970-8034, Fukushima (Japan); Yamashita, S. [Fuels and Materials Department, O-arai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Kokawa, H. [Graduate School of Engineering, Tohoku University, Sendai 980-8579 (Japan)

    2013-03-15

    The behavior of radiation-induced segregation (RIS) and intergranular corrosion at random grain boundaries and Σ3 coincidence site lattice (CSL) boundaries in proton-irradiated 316L stainless steel was examined. The frequency of the CSL boundaries was enhanced up to 86.6% by grain boundary engineering treatment prior to irradiation. Significant nickel enrichment and chromium depletion were induced at the random grain boundary owing to the RIS. At faceted Σ3 CSL boundaries, chromium depletion occurred at the asymmetrical boundary facet plane whereas no RIS was observed at the coherent twin boundary. After the electrochemical etching test, an intergranular corrosion groove was found along the random grain boundaries because of the low chromium concentration (∼12%) at the boundaries. At the faceted Σ3 CSL boundaries, the discontinuous groove along the asymmetric facet plane was completely disrupted by the non-corrosive coherent twin boundary.

  4. Anisotropic Radiation-Induced Segregation in 316L Austenitic Stainless Steel with Grain Boundary Character

    Energy Technology Data Exchange (ETDEWEB)

    Christopher M. Barr; Gregory A. Vetterick; Kinga A. Unocic; Khalid Hattar; Xian-Ming Bai; Mitra L. Taheri

    2014-04-01

    Radiation-induced segregation (RIS) and subsequent depletion of chromium along grain boundaries has been shown to be an important factor in irradiation-assisted stress corrosion cracking in austenitic face-centered cubic (fcc)-based alloys used for nuclear energy systems. A full understanding of RIS requires examination of the effect of the grain boundary character on the segregation process. Understanding how specific grain boundary structures respond under irradiation would assist in developing or designing alloys that are more efficient at removing point defects, or reducing the overall rate of deleterious Cr segregation. This study shows that solute segregation is dependent not only on grain boundary misorientation, but also on the grain boundary plane, as highlighted by markedly different segregation behavior for the __3 incoherent and coherent grain boundaries. The link between RIS and atomistic modeling is also explored through molecular dynamic simulations of the interaction of vacancies at different grain boundary structures through defect energetics in a simple model system. A key insight from the coupled experimental RIS measurements and corresponding defect–grain boundary modeling is that grain boundary–vacancy formation energy may have a critical threshold value related to the major alloying elements’ solute segregation.

  5. Effects of bonding bakeout thermal cycles on pre- and post irradiation microstructures, physical, and mechanical properties of copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J. [Pacific Northwest National Laboratory, Richland, WA (United States)

    1996-10-01

    At present, dispersion strengthened (DS) copper is being considered as the primary candidate material for the ITER first wall and divertor components. Recently, it was agreed among the ITER parties that a backup alloy should be selected from the two well known precipitation hardened copper alloys, CuCrZr and CuNiBe. It was therefore decided to carry out screening experiments to simulate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties, and electrical resistivity of CuCrZr and CuNiBe alloys. On the basis of the results of these experiments, one of the two alloys will be selected as a backup material. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime ageing, and bonding thermal cycle followed by reageing and the reactor bakeout treatment at 623K for 100 hours. Tensile specimens of the DS copper were also given the heat treatment corresponding to the bonding thermal cycle. A number of these heat treated specimens of CuCrZr, CuNiBe, and DS copper were neutron irradiated at 523K to a dose level of {approx}0.3 dpa (NRT) in the DR-3 reactor at Riso. Both unirradiated and irradiated specimens with the various heat treatments were tensile tested at 532K. The dislocation, precipitate and void microstructures and electrical resistivity of these specimens were also determined. Results of these investigations will be reported and discussed in terms of thermal and irradiation stability of precipitates and irradiation-induced precipitation and recovery of dislocation microstructure. Results show that the bonding and bakeout thermal cycles are not likely to have any serious deleterious effects on the performance of these alloys. The CuNiBe alloys were found to be susceptible to radiation-induced embrittlement, however, the exact mechanism is not yet known. It is thought that radiation-induced precipitation and segregation of the beryllium may be responsible.

  6. The dependence of helium generation rate on nickel content of Fe-Cr-Ni alloys irradiated at high dpa levels in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Oliver, B.M.; Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-04-01

    With a few exceptions in the literature, it is generally accepted that it is nickel in Fe-Cr-Ni alloys that produces most of the transmutant helium and that the helium generation rate should scale linearly with the nickel content. Surprisingly, this assumption is based only on irradiations of pure nickel and has never been tested in an alloy series. There have also been no extensive tests of the predictions for helium production in alloys in various fast reactors spectra.

  7. Void swelling in high purity FeCrNi and FeCrNiTi alloys irradiated in JOYO

    Science.gov (United States)

    Muroga, T.; Araki, K.; Miyamoto, Y.; Yoshida, N.

    1988-07-01

    Microstructures have been observed in Fe-13Cr-14Ni and Fe-13Cr-14Ni-0.12Ti alloys irradiated in JOYO (Japanese Fast Experimental Reactor) at 400, 500 and 600 °C to the fluence of 0.079, 0.81 and 6.2 × 10 25n/ m2 ( E > 0.1 MeV). In the Fe-13Cr-14Ni alloy, voids are observed in all cases. The dose dependence of swelling seems to obey the kinetics of linear increase with or without initial short transient. On the other hand, remarkable swelling suppression effects are observed in the Fe-13Cr-14Ni-0.12Ti alloy. The detailed microstructural observation suggests the titanium addition effects suppress the void nucleation in the matrix by gettering impurities and obstructing dislocation climb by precipitate decoration on dislocation lines.

  8. Carbon Concentration of Austenite

    Directory of Open Access Journals (Sweden)

    Z. Ławrynowicz

    2007-07-01

    Full Text Available The investigation was carried out to examine the influence of temperature and times of austempering process on the maximum extend towhich the bainite reaction can proceed and the carbon content in retained austenite. It should be noted that a small percentage change in theaustenite carbon content can have a significant effect on the subsequent austempering reaction changing the volume fraction of the phasespresent and hence, the resulting mechanical properties. Specimens were prepared from an unalloyed ductile cast iron, austenitised at 950oCfor 60 minutes and austempered by the conventional single-step austempering process at four temperatures between BS and MS, eg., 250,300, 350 and 400oC. The samples were austempered at these temperatures for 15, 30, 60, 120 and 240 minutes and finally quenched toambient temperature. Volume fractions of retained austenite and carbon concentration in the residual austenite have been observed byusing X-ray diffraction. Additionally, carbon concentration in the residual austenite was calculated using volume fraction data of austeniteand a model developed by Bhadeshia based on the McLellan and Dunn quasi-chemical thermodynamic model. The comparison ofexperimental data with the T0, T0' and Ae3' phase boundaries suggests the likely mechanism of bainite reaction in cast iron is displacive rather than diffusional. The carbon concentration in retained austenite demonstrates that at the end of bainite reaction the microstructure must consist of not only ausferrite but additionally precipitated carbides.

  9. Microchemical effects in irradiated Fe–Cr alloys as revealed by atomistic simulation

    Energy Technology Data Exchange (ETDEWEB)

    Malerba, L., E-mail: lmalerba@sckcen.be [Structural Materials Modelling and Microstructure Unit, SMA/NMS, Studiecentrum voor Kernenergie, Centre d’Etudes de l’Energie Nucléaire (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Bonny, G.; Terentyev, D. [Structural Materials Modelling and Microstructure Unit, SMA/NMS, Studiecentrum voor Kernenergie, Centre d’Etudes de l’Energie Nucléaire (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Zhurkin, E.E. [Experimental Nuclear Physics Department, K-89, Faculty of Physics and Mechanics, Saint-Petersburg State Polytechnical University, 29 Polytekhnicheskaya Str., 195251 St. Petersburg (Russian Federation); Hou, M. [Physique des Solides Irradiés et des Nanostructures CP234, Faculté des Sciences, Université Libre de Bruxelles, Bd du Triomphe, B-1050 Bruxelles (Belgium); Vörtler, K.; Nordlund, K. [Association EURATOM-Tekes, Department of Physics, P.O. Box 43, FI-00014, University of Helsinki (Finland)

    2013-11-15

    Neutron irradiation produces evolving nanostructural defects in materials, that affect their macroscopic properties. Defect production and evolution is expected to be influenced by the chemical composition of the material. In turn, the accumulation of defects in the material results in microchemical changes, which may induce further changes in macroscopic properties. In this work we review the results of recent atomic-level simulations conducted in Fe–Cr alloys, as model materials for high-Cr ferritic–martensitic steels, to address the following questions: 1. Is the primary damage produced in displacement cascades influenced by the Cr content? If so, how? 2. Does Cr change the stability of radiation-produced defects? 3. Is the diffusivity of cascade-produced defects changed by Cr content? 4. How do Cr atoms redistribute under irradiation inside the material under the action of thermodynamic driving forces and radiation-defect fluxes? It is found that the presence of Cr does not influence the type of damage created by displacement cascades, as compared to pure Fe, while cascades do contribute to redistributing Cr, in the same direction as thermodynamic driving forces. The presence of Cr does change the stability of point-defects: the effect is weak in the case of vacancies, stronger in the case of self-interstitials. In the latter case, Cr increases the stability of self-interstitial clusters, especially those so small to be invisible to the electron microscope. Cr reduces also significantly the diffusivity of self-interstitials and their clusters, in a way that depends in a non-monotonic way on Cr content, as well as on cluster size and temperature; however, the effect is negligible on the mobility of self-interstitial clusters large enough to become visible dislocation loops. Finally, Cr-rich precipitate formation is favoured in the tensile region of edge dislocations, while it appears not to be influenced by screw dislocations; prismatic dislocation loops

  10. Synthesis of per-fluorinated polymer-alloy based on PTFE by high temperature EB-irradiation

    Science.gov (United States)

    Oshima, Akihiro; Mutou, Fumihiro; Hyuga, Toshiyuki; Asano, Saneto; Ichizuri, Shogo; Li, Jingye; Miura, Takaharu; Washio, Masakazu

    2005-07-01

    In this study, synthesis of per-fluorinated polymer-alloy based on polytetrafluoroethylene (PTFE) has been demonstrated by high temperature irradiation techniques. The per-fluorinated polymer-blend thin films originated from polymer dispersion (PTFE, PTFE/PFA polymer-blend: FA and PTFE/FEP polymer-blend: FE) have been fabricated by the wire-bar coating equipment. The obtained films (thickness: 5-15 μm) were irradiated by EB at 335 °C ± 5 °C in nitrogen gas atmosphere. Characterization of irradiated polymer-blends has been performed by 19F solid-state NMR spectroscopy, thermal analysis and so on. By DSC analysis, the heat of crystallization (ΔHc) of both irradiated polymer-blends were decreased with increase in absorbed dose. Moreover, the melting and crystallization temperatures of both materials shift to lower temperatures, compared with crosslinked PTFE. The obtained materials showed the lower crystallinity. By 19F solid-state NMR spectroscopy, the new signals appeared at around -160 ppm and at -188 ppm. The signals are assigned to the fluorine signals of CF groups, which represent crosslinking sites with Y-type (>CF-) and Y‧-type (>Cdbnd CF-) in the polymer-blend chains. Thus, it is confirmed that the polymer-alloys with good performance based on PTFE are synthesized through the radiation crosslinking reaction between PTFE and PFA or FEP molecules.

  11. EL2-related defects in neutron irradiated GaAs/sub 1//sub -x/P/sub x/ alloys

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, E.; Garcia, F.; Jimenez, B.; Calleja, E.; Gomez, A.; Alcober, V.

    1985-10-15

    The generation of EL2-related defects in GaAsP alloys by fast neutron irradiation has been studied through deep level transient spectroscopy and photocapacitance techniques. After irradiation p-n junctions were not annealed at high temperatures. In the composition range x>0.4, fast neutrons generate a broad center at E/sub c/-0.7 eV that it is suggested to belong to the EL2 family. The presence of photocapacitance quenching effects has been taken as a preliminary fingerprint to make the above assignment. From computer analysis of the nonexponential transient capacitance waveforms, evidence that neutron irradiation creates a family of midgap levels, EL2-related, is found.

  12. Effects of irradiation induced Cu clustering on Vickers hardness and electrical resistivity of Fe–Cu model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tobita, Tohru, E-mail: tobita.tohru@jaea.go.jp [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai, Naka-gun, Ibaraki-prefecture 319-1195 (Japan); Nakagawa, Shou [Department of Materials Science, Osaka Prefecture University, Sakai-shi, Osaka 599-8531 (Japan); Takeuchi, Tomoaki; Suzuki, Masahide [Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency, Narita, Oarai, Higashiibaraki-gun, Ibaraki-prefecture 311-1393 (Japan); Ishikawa, Norito [Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Tokai, Naka-gun, Ibaraki-prefecture 319-1195 (Japan); Chimi, Yasuhiro [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai, Naka-gun, Ibaraki-prefecture 319-1195 (Japan); Saitoh, Yuichi [Department of Advanced Radiation Technology, Japan Atomic Energy Agency, Watanuki, Takasaki-shi, Gunma-prefecture 370-1292 (Japan); Soneda, Naoki; Nishida, Kenji; Ishino, Siori [Central Research Institute of Electric Power Industry, Komae-shi, Tokyo 201-8511 (Japan); Iwase, Akihiro [Department of Materials Science, Osaka Prefecture University, Sakai-shi, Osaka 599-8531 (Japan)

    2014-09-15

    Three kinds of Fe-based model alloys, Fe–0.018 atomic percent (at.%) Cu, Fe–0.53at.%Cu, and Fe–1.06at.%Cu were irradiated with 2 MeV electrons up to the dose of 2 × 10{sup −5} dpa at 250 °C. After the irradiation, the increase in Vickers hardness and the decrease in electrical resistivity were observed. The increase in hardness by electron irradiation is proportional to the product of the Cu contents and the square root of the electron dose. The decrease in electrical resistivity is proportional to the product of the square of Cu contents and the electron dose. Cu clustering in the materials with electron irradiation and thermal aging was observed by means of the Atom Probe Tomography (APT). The change in Vickers hardness and electrical resistivity is well correlated with micro-structure evolution related to the Cu clustering process. The irradiation hardening was proportional to the square root of volume fraction of the Cu clusters from early stage of irradiation.

  13. Study of the interactions between irradiation and chemical order effects in ternary alloys Ni-Cr-Fe; Etude des interactions entre effets d`irradiation et effets d`ordre chimique dans les alliages ternaires Ni-Cr-Fe

    Energy Technology Data Exchange (ETDEWEB)

    Frely, E

    1997-12-31

    Because of its resistance to corrosion even under stress, the alloy 69 (nickel-based alloy with a chemically disordered F.c.c. structure) is a promising material for application in some of the inner parts of nuclear reactor. However, the eventual formation of an ordered NI{sub 2}Cr superstructure under irradiation or thermal ageing might diminish its performances. We have studied the binary model alloy Ni-Cr33at.% as well as the ternary alloys Ni-Cr3at.%-Fe5cat.% and Ni-Cr32at.%-Fe10at.%, the last one having a chemical composition similar to that of the industrial alloy. After irradiation experiments with 2.5 MeV electrons in the 300-500 deg C temperature range, all the model alloys show the Ni{sub 2}Cr superstructure. The samples irradiated at fluences between 2 and 8. 10 d.p.a. have been characterized by X-ray and neutron diffraction. The superlattice reflexions and the ordered domains have been observed by electron microscopy. The critical temperature of the order-disorder transformation, measured under 1 MeV electron irradiation, decreases linearly with iron content. The evolution of the chemical corder has been traced by means of in situ resistivity measurements. We have used the pair exchange based Dienes model of ordering kinetics for studying the long range order S (S between 0.5 and 0.8 after irradiation). The iron seems to remain in disorder in the sublattices. The similarity of the results under thermal ageing and under irradiation shows that the main effect of the electronic irradiation is to accelerate ordering. Under both treatments increasing the iron content or the dislocation density reduce the ordering kinetics. However, this effect is not sufficient to explain the lack of order in alloy 690 after a fluence of 1 d.p.a. (author). 95 refs.

  14. The Kinetics of Dislocation Loop Formation in Ferritic Alloys Through the Aggregation of Irradiation Induced Defects

    Science.gov (United States)

    Kohnert, Aaron Anthony

    The mechanical properties of materials are often degraded over time by exposure to irradiation environments, a phenomenon that has hindered the development of multiple nuclear reactor design concepts. Such property changes are the result of microstructural changes induced by the collision of high energy particles with the atoms in a material. The lattice defects generated in these recoil events migrate and interact to form extended damage structures. This study has used theoretical models based on the mean field chemical reaction rate theory to analyze the aggregation of isolated lattice defects into larger microstructural features that are responsible for long term property changes, focusing on the development of black dot damage in ferritic iron based alloys. The purpose of such endeavors is two-fold. Primarily, such models explain and quantify the processes through which these microstructures form. Additionally, models provide insight into the behavior and properties of the point defects and defect clusters which drive general microstructural evolution processes. The modeling effort presented in this work has focused on physical fidelity, drawing from a variety of sources of information to characterize the unobservable defect generation and agglomeration processes that give rise to the observable features reported in experimental data. As such, the models are based not solely on isolated point defect creation, as is the case with many older rate theory approaches, but instead on realistic estimates of the defect cluster population produced in high energy cascade damage events. Experimental assessments of the microstructural changes evident in transmission electron microscopy studies provide a means to measure the efficacy of the kinetic models. Using common assumptions of the mobility of defect clusters generated in cascade damage conditions, an unphysically high density of damage features develops at the temperatures of interest with a temperature dependence

  15. Effects of solute atoms on evolution of vacancy defects in electron-irradiated Fe-Cr-based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Druzhkov, A.P., E-mail: druzhkov@imp.uran.r [Institute of Metal Physics, Ural Branch RAS, 18 Kovalevskaya St., 620041 Ekaterinburg (Russian Federation); Nikolaev, A.L. [Institute of Metal Physics, Ural Branch RAS, 18 Kovalevskaya St., 620041 Ekaterinburg (Russian Federation)

    2011-01-15

    The evolution of vacancy-type defects in Fe-Cr alloys (13-16 at.% Cr) undoped and doped with C, N, Au, or Sb and in conventional ferritic-martensitic steel ({approx}13% Cr) has been investigated using positron annihilation spectroscopy under electron irradiation at room temperature and subsequent stepwise annealing. Small vacancy clusters are formed in the undoped Fe-16Cr alloy, which anneal out between 320 and 550 K. It is shown that oversized substitutional solute atoms (Sb, Au) in the Fe-Cr alloy interact with vacancies and form complexes, which are stable up to 600 and 420 K, respectively. It is found that the accumulation of vacancy defects considerably increases in the alloys and the steel with an enhanced content of interstitial impurities. It is shown that this effect is related to the formation of vacancy-carbon complexes. It is known that chromium in iron decreases the diffusion mobility of carbon. Therefore, the structure of vacancy-carbon complexes and the kinetics of their annealing in Fe-Cr alloys differ from those in the Fe-C system.

  16. Nanostructure evolution under irradiation in FeMnNi alloys: A “grey alloy” object kinetic Monte Carlo model

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France)

    2015-07-15

    This work extends the object kinetic Monte Carlo model for neutron irradiation-induced nanostructure evolution in Fe–C binary alloys developed in [1], introducing the effects of substitutional solutes like Mn and Ni. The objective is to develop a model able to describe the nanostructural evolution of both vacancy and self-interstitial atom (SIA) defect cluster populations in Fe(C)MnNi neutron-irradiated model alloys at the operational temperature of light water reactors (∼300 °C), by simulating specific reference irradiation experiments. To do this, the effects of the substitutional solutes of interest are introduced, under simplifying assumptions, using a “grey alloy” scheme. Mn and Ni solute atoms are not explicitly introduced in the model, which therefore cannot describe their redistribution under irradiation, but their effect is introduced by modifying the parameters that govern the mobility of both SIA and vacancy clusters. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proved to be key to explain the experimentally observed disappearance of detectable defect clusters with increasing solute content. Solute concentration is explicitly taken into account in the model as a variable determining the slowing down of self-interstitial clusters; small vacancy clusters, on the other hand, are assumed to be significantly slowed down by the presence of solutes, while for clusters bigger than 10 vacancies their complete immobility is postulated. The model, which is fully based on physical considerations and only uses a few parameters for calibration, is found to be capable of reproducing the experimental trends in terms of density and size distribution of the irradiation-induced defect populations with dose, as compared to the reference experiment, thereby providing insight into the physical mechanisms that influence the nanostructural evolution undergone by this material during irradiation.

  17. The effect of oxygen on void stability in ion-irradiated steel

    Science.gov (United States)

    Seitzman, Larry E.; Dodd, R. Arthur; Kulcinski, Gerald L.

    1990-07-01

    The effect of oxygen on void stability in an Fe-17Ni-13Cr austenitic ternary alloy has been investigated using 15 MeV nickel-ion irradiation at elevated temperatures and preimplantation of 6 MeV oxygen at room temperature. The nickel irradiation was performed over a temperature range of 550 °C to 650 °C. Utilizing transverse specimen preparation techniques, the irradiated steel was examined by transmission electron microscopy (TEM). As little as 10 appm preimplanted oxygen caused a significant increase in the void number density when the steel was irradiated at 550 °C. A near-surface void-denuded zone occurs in the irradiated steel, while a region depleted of visible voids also occurs in the steel injected with 300 appm oxygen or greater and irradiated at 550 °C.

  18. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part 1: Effects of minor alloying elements on precipitate phases in melt products and implication in alloy fabrication

    Science.gov (United States)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    In an effort to develop alloys for fission and fusion reactor applications, 28Fe-15Ni-13Cr base alloys were fabricated by adding various combinations of the minor alloying elements, Mo, Ti, C, Si, P, Nb, and B. The results showed that a significant fraction of undesirable residual oxygen was removed as oxides when Ti, C, and Si were added. Accordingly, the concentrations of the latter three essential alloying elements were reduced also. Among these elements, Ti was the strongest oxide former, but the largest oxygen removal (over 80%) was observed when carbon was added alone without Ti, since gaseous CO boiled off during melting. This paper recommends an alloy melting procedure to mitigate solute losses while reducing the undesirable residual oxygen. In this work, 14 different types of precipitate phases were identified. Compositions of precipitate phases and their crystallographic data are documented. Finally, stability of precipitate phases was examined in view of Gibbs free energy of formation.

  19. Development of Cast Alumina-Forming Austenitic Stainless Steels

    Science.gov (United States)

    Muralidharan, G.; Yamamoto, Y.; Brady, M. P.; Walker, L. R.; Meyer, H. M., III; Leonard, D. N.

    2016-09-01

    Cast Fe-Ni-Cr chromia-forming austenitic stainless steels with Ni levels up to 45 wt.% are used at high temperatures in a wide range of industrial applications that demand microstructural stability, corrosion resistance, and creep strength. Although alumina scales offer better corrosion protection at these temperatures, designing cast austenitic alloys that form a stable alumina scale and achieve creep strength comparable to existing cast chromia-forming alloys is challenging. This work outlines the development of cast Fe-Ni-Cr-Al austenitic stainless steels containing about 25 wt.% Ni with good creep strength and the ability to form a protective alumina scale for use at temperatures up to 800-850°C in H2O-, S-, and C-containing environments. Creep properties of the best alloy were comparable to that of HK-type cast chromia-forming alloys along with improved oxidation resistance typical of alumina-forming alloys. Challenges in the design of cast alloys and a potential path to increasing the temperature capability are discussed.

  20. Development of Cast Alumina-Forming Austenitic Stainless Steels

    Science.gov (United States)

    Muralidharan, G.; Yamamoto, Y.; Brady, M. P.; Walker, L. R.; Meyer, H. M., III; Leonard, D. N.

    2016-11-01

    Cast Fe-Ni-Cr chromia-forming austenitic stainless steels with Ni levels up to 45 wt.% are used at high temperatures in a wide range of industrial applications that demand microstructural stability, corrosion resistance, and creep strength. Although alumina scales offer better corrosion protection at these temperatures, designing cast austenitic alloys that form a stable alumina scale and achieve creep strength comparable to existing cast chromia-forming alloys is challenging. This work outlines the development of cast Fe-Ni-Cr-Al austenitic stainless steels containing about 25 wt.% Ni with good creep strength and the ability to form a protective alumina scale for use at temperatures up to 800-850°C in H2O-, S-, and C-containing environments. Creep properties of the best alloy were comparable to that of HK-type cast chromia-forming alloys along with improved oxidation resistance typical of alumina-forming alloys. Challenges in the design of cast alloys and a potential path to increasing the temperature capability are discussed.

  1. A study on the influence of trace elements (C, S, B, Al, N) on the hot ductility of the high purity austenitic alloy Fe-Ni 36% (INVAR)

    Energy Technology Data Exchange (ETDEWEB)

    Simonetta-Perrot, M.T.

    1994-11-01

    In order to study the damage mechanisms leading to the ductility decrease of the Invar alloy at 600 C, a high-purity Fe-Ni 36% sample has been doped with trace elements with the purpose of identifying the role of sulfur, sulfur with Al N or B N precipitates and sulfur with boron, on the ductility, the failure modes, the intergranular damage and the plastic deformation mechanisms prior to failure. A new AES segregation quantification method has been used to study the kinetics and thermodynamics of intergranular and surface segregations and determine the relation between sulfur segregation and grain joint fragility. refs., figs., tabs.

  2. On the crystal structure of Cr2N precipitates in high-nitrogen austenitic stainless steel. II. Order-disorder transition of Cr2N during electron irradiation.

    Science.gov (United States)

    Lee, Tae Ho; Kim, Sung Joon; Takaki, Setsuo

    2006-04-01

    The crystal structure and order-disorder transition of Cr2N were investigated utilizing transmission electron microscopy (TEM). Based on the analyses of selected-area diffraction (SAD) patterns, the crystal structure of the ordered Cr2N superstructure was confirmed to be trigonal (P31m), characterized by three sets of superlattice reflections (001), ((11/33)0) and ((11/33)1). During electron irradiation, the superlattice reflections gradually disappeared in the regular sequence (001), ((11/33)0) and ((11/33)1), indicating that the order-disorder phase transition of Cr2N occurred. The convergent-beam electron diffraction (CBED) observation revealed that the space group of disordered Cr2N is P6(3)/mmc, which corresponds to an h.c.p. (hexagonal close packed) sublattice of metal atoms with a random distribution of N atoms in six octahedral interstices. The redistribution model of N atoms through the order-disorder transition is discussed based on the characteristics and disappearing sequence of superlattice reflections.

  3. Impact of Ion Irradiation upon Structure and Magnetic Properties of NANOPERM-Type Amorphous and Nanocrystalline Alloys

    Directory of Open Access Journals (Sweden)

    Marcel Miglierini

    2015-01-01

    Full Text Available Structural modifications and their impact upon magnetic properties are studied in amorphous and nanocrystalline NANOPERM-type 57Fe75Mo8Cu1B16 alloy. They are introduced by irradiation with 130 keV N+ ions to the total fluencies of up to 2.5 × 1017 ions/cm2 under different cooling conditions. Increased temperature during the irradiation triggers formation of nanocrystallites of bcc-Fe in those subsurface regions that are affected by bombarding ions. No crystallization occurs when good thermal contact between the irradiated sample and a sample holder is assured. Instead, structural rearrangement which favours development of magnetically active regions was determined by the local probe methods of Mössbauer spectrometry. Dipole magnetic interactions dominate in subsurface regions on that side of the ribbons which was exposed to ion irradiation. Nevertheless, structural modifications demonstrate themselves also via macroscopic magnetic parameters such as temperature dependence of magnetization, Curie temperature, and hysteresis loops. Impact of only the temperature itself to the observed effects is assessed by the help of samples that were subjected just to heat treatment, that is, without ion irradiation.

  4. Defects in hyperpure Fe-based alloys created by 3 MeV e{sup -}-irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Li, X.H.; Moser, P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Recherche Fondamentale sur la Matiere Condensee; Akamatsu, M.; Van Duysen, C. [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    Information about vacancy defects created in RPV (Reactor Pressure Vessels) steels after neutron irradiations are obtained via a simulation: the RPV steels are simulated by a series of high purity Fe-based alloys; the neutron irradiation is simulated by a 3 MeV electron irradiation; vacancy defects characteristics are obtained by positron lifetime techniques. Irradiations are made at 150 or 288 deg C, with a dose of 4*10{sup 19} e-/cm{sup 2}, and followed by isochronal annealing in the range 20-500 deg C. The observed vacancy defects are single trapped vacancies and small vacancy clusters, the size of which being lower than 10 empty atomic volumes (vacancy clusters containing more than 50 empty atomic volumes were never found). A large recovery step is observed between 200 and 400 deg C, after 150 deg C irradiation and attributed to vacancy-impurity detrapping, and also, vacancy cluster evaporation. The influence of C, Cu and Mo are presented. These results are in agreement with a model supposing, in pure Fe, single vacancy migration at -50 deg C and vacancy-impurity detrapping at 200 deg C. (authors). 4 figs., 15 refs.

  5. Effects ofγ-irradiation and Deformation Temperature on Tensile Properties of Pb-2 mass% Sb Alloy

    Institute of Scientific and Technical Information of China (English)

    Gh MOHAMMED; SEI-GAMAL

    2016-01-01

    Effects ofγ-irradiation and deformation temperature (T)on the tensile properties of Pb-2 mass% Sb alloys were studied.The samples were annealed at 458 K for 2 h in air,then water quenched after they wereγ-irradiated (the different doses were 0·5,1·0,1·5,and 2·0 MGy).The tensile properties were performed using stress-strain measurements at a constant strain rate (1·2×10-3 s-1 )and at different T (303-393 K).It was found that at con-stant dose,the fracture stress (σF )decreases while the fracture strain (εF )increases as T increases.At particular T,σF increases whileεF decreases with increasing dose.The strain-hardening exponent (n),which is the slope of the relation between ln(σ)and ln(ε)of the parabolic part of the stress-strain curve,was determined and its values in-crease as T increases and decrease as the dose increases.The value of the activation energy increases as the dose in-creases from 0·07 eV for un-irradiated sample to 0·1 eV for the 2 MGy-irradiated sample.These values are in ac-cordance with that needed for dislocation movement and ordering process.An interpretation of the results was given, based on the creation of point and line defects due toγ-irradiation,and that results in a distribution of beta phase (Sb-phase),leading to a difficulty in the movement of dislocations,so there is an increase in alloy hardness.

  6. Austenite nucleation and growth observed on the level of individual grains by three-dimensional X-ray diffraction microscopy

    OpenAIRE

    Savran, V.I.; Offerman, S. E.; Sietsma, J.

    2010-01-01

    Austenite nucleation and growth is studied during continuous heating using three-dimensional X-ray diffraction (3-D XRD) microscopy at the European Synchrotron Radiation Facility (ESRF) (Grenoble, France). Unique in-situ observations of austenite nucleation and growth kinetics were made for two commercial medium-carbon low-alloy steels (0.21 and 0.35 wt pct carbon with an initial microstructure of ferrite and pearlite). The measured austenite volume fraction as a function of temperature shows...

  7. The influence of silicon and aluminum on austenite deformation behavior during fatigue and tensile loading

    Science.gov (United States)

    Lehnhoff, Gregory R.

    Advanced high strength steels (AHSS) for automobile light-weighting utilize Si and Al alloying to retain austenite in the microstructure during thermal partitioning treatments. This research project utilized fully austenitic steels with varied Si and Al compositions to understand the effect of these elements on austenite deformation response, including deformation induced martensite formation and deformation twinning. Specific focus was directed at understanding austenite deformation response during fatigue loading. Independent alloying additions of 2.5 wt pct Si and Al were made to a base steel composition of 15 Ni - 11 Cr - 1 Mn - 0.03 C (wt pct). Weak beam dark field transmission electron microscopy (TEM) imaging of dissociated dislocations was implemented to experimentally determine the influences of Si and Al on austenite stacking fault energy (SFE). The 2.5 wt pct Si alloying addition decreased the SFE by 6.4 mJ/m2, while the 2.5 wt pct Al alloying increased the SFE by 12 mJ/m2. Fully reversed, total strain controlled, low cycle fatigue (LCF) tests indicated that all four alloys underwent primary cyclic hardening and stabilization. Secondary cyclic strain hardening was correlated to BCC martensite formation using Feritscope magnetic fraction measurements of LCF specimens; the formation of 1 pct martensite led to 7 MPa of secondary hardening. TEM showed that martensite predominantly formed as parallel, irregular bands through strain induced nucleation on austenite shear bands. The austenite shear bands consisted of austenite {111} planes with concentrated dislocations, stacking faults, and/or HCP epsilon-martensite. Aluminum alloying promoted martensite formation during LCF, while Si suppressed martensite. Therefore, the strain induced nucleation process was not suppressed by the increased SFE associated with Al alloying. Tensile testing indicated that Si alloying promoted deformation twinning by lowering the SFE. Similarly to LCF loading, Al promoted

  8. Synergic effects of ion irradiations (La, Ce) and alkaline pretreatment (KOH) on hydriding kinetic property of a Mm–Ni based alloy

    Energy Technology Data Exchange (ETDEWEB)

    Abe, H., E-mail: abe.hiroshi10@jaea.go.jp [Quantum Beam Science Directorate, Japan Atomic Energy Agency, 1233 Watanuki, Takasaki, Gunma 370-1292 (Japan); Aone, S.; Morimoto, R.; Uchida, H. [Course of Applied Science, Graduate School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan)

    2013-12-15

    Highlights: •Ion irradiations by various ions onto a Mm–Ni based hydrogen storage alloy. •The generation of vacancy type defects by ion irradiations in the surface of a Mm–Ni based alloy. •The enhancement of the initial rate by the ion irradiations of the Mm surface. •The enhancement of the initial rate by an alkaline treatment of the Mm surface. -- Abstract: The ion beam irradiation is known to produce a high density of vacancy type defects in the surface region of a metal and found to be an effective method as a surface modification in order to enhance the hydriding rate of a metal. In this study, we examined synergic effects of both surface modifications of ion irradiations and alkaline treatment on the initial hydriding rate of a Mm–Ni based alloy. In this study, the irradiations by lanthanum (La) and cerium (Ce) ions combined with an alkaline KOH pretreatment were found much more effective in the enhancement of the initial hydriding rate compared with irradiations with other ions. This study reports the synergic effects of the surface modifications by the both the surface irradiations with rare earth ions of La and Ce, and an alkaline surface treatment on the hydriding kinetics.

  9. Oxidation resistant high creep strength austenitic stainless steel

    Science.gov (United States)

    Brady, Michael P.; Pint, Bruce A.; Liu, Chain-Tsuan; Maziasz, Philip J.; Yamamoto, Yukinori; Lu, Zhao P.

    2010-06-29

    An austenitic stainless steel displaying high temperature oxidation and creep resistance has a composition that includes in weight percent 15 to 21 Ni, 10 to 15 Cr, 2 to 3.5 Al, 0.1 to 1 Nb, and 0.05 to 0.15 C, and that is free of or has very low levels of N, Ti and V. The alloy forms an external continuous alumina protective scale to provide a high oxidation resistance at temperatures of 700 to 800.degree. C. and forms NbC nanocarbides and a stable essentially single phase fcc austenitic matrix microstructure to give high strength and high creep resistance at these temperatures.

  10. Laser etching of austenitic stainless steels for micro-structural evaluation

    Science.gov (United States)

    Baghra, Chetan; Kumar, Aniruddha; Sathe, D. B.; Bhatt, R. B.; Behere, P. G.; Afzal, Mohd

    2015-06-01

    Etching is a key step in metallography to reveal microstructure of polished specimen under an optical microscope. A conventional technique for producing micro-structural contrast is chemical etching. As an alternate, laser etching is investigated since it does not involve use of corrosive reagents and it can be carried out without any physical contact with sample. Laser induced etching technique will be beneficial especially in nuclear industry where materials, being radioactive in nature, are handled inside a glove box. In this paper, experimental results of pulsed Nd-YAG laser based etching of few austenitic stainless steels such as SS 304, SS 316 LN and SS alloy D9 which are chosen as structural material for fabrication of various components of upcoming Prototype Fast Breeder Reactor (PFBR) at Kalpakkam India were reported. Laser etching was done by irradiating samples using nanosecond pulsed Nd-YAG laser beam which was transported into glass paneled glove box using optics. Experiments were carried out to understand effect of laser beam parameters such as wavelength, fluence, pulse repetition rate and number of exposures required for etching of austenitic stainless steel samples. Laser etching of PFBR fuel tube and plug welded joint was also carried to evaluate base metal grain size, depth of fusion at welded joint and heat affected zone in the base metal. Experimental results demonstrated that pulsed Nd-YAG laser etching is a fast and effortless technique which can be effectively employed for non-contact remote etching of austenitic stainless steels for micro-structural evaluation.

  11. Effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 350 degrees C

    DEFF Research Database (Denmark)

    Singh, B.N.; Edwards, D.J.; Eldrup, Morten Mostgaard

    2001-01-01

    specimens were given heat treatments corresponding to solution anneal, prime-aging, bonding thermal treatment followed by re-aging and the reactor bakeout treatment. A number of specimens were irradiated at 350 degreesC to a dose level of similar or equal to0.3 dpa in the DR-3 reactor at Riso. Both......Screening experiments were carried out to determine the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) copper alloys. Tensile...

  12. Detection of helium in irradiated Fe9Cr alloys by coincidence Doppler broadening of slow positron annihilation

    Science.gov (United States)

    Cao, Xingzhong; Zhu, Te; Jin, Shuoxue; Kuang, Peng; Zhang, Peng; Lu, Eryang; Gong, Yihao; Guo, Liping; Wang, Baoyi

    2017-03-01

    An element analysis method, coincidence Doppler broadening spectroscopy of slow positron annihilation, was employed to detect helium in ion-irradiated Fe9Cr alloys. Spectra with higher peak to background ratio were recorded using a two-HPGe detector coincidence measuring system. It means that information in the high-momentum area of the spectra can be used to identify helium in metals. This identification is not entirely dependent on the helium concentration in the specimens, but is related to the structure and microscopic arrangement of atoms surrounding the positron annihilation site. The results of Doppler broadening spectroscopy and transmission electron microscopy show that vacancies and dislocations were formed in ion-irradiated specimens. Thermal helium desorption spectrometry was performed to obtain the types of He traps.

  13. Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking

    Energy Technology Data Exchange (ETDEWEB)

    Gary S. Was

    2009-03-31

    The objective of this project is to determine whether deformation mode is a primary factor in the mechanism of irradiation assisted intergranular stress corrosion cracking of austenitic alloys in light watert reactor core components. Deformation mode will be controlled by both the stacking fault energy of the alloy and the degree of irradiation. In order to establish that localized deformation is a major factor in IASCC, the stacking fault energies of the alloys selected for study must be measured. Second, it is completely unknown how dose and SFE trade-off in terms of promoting localized deformation. Finally, it must be established that it is the localized deformation, and not some other factor that drives IASCC.

  14. Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

    Science.gov (United States)

    Arbuzov, V. L.; Goshchitskii, B. N.; Sagaradze, V. V.; Danilov, S. E.; Kar'kin, A. E.

    2010-10-01

    The processes of accumulation and annealing of radiation defects at low-temperature (77 K) electron and neutron irradiation and their effect on the physicomechanical properties of Fe-Cr alloys and oxide dispersion strengthened (ODS) steel have been studied. It has been shown that the behavior of radiation defects in ODS steel and Fe-Cr alloys is qualitatively similar. Above 250 K, radiation-induced processes of the solid solution decomposition become conspicuous. These processes are much less pronounced in ODS steel because of specific features of its microstructure. Processes related to the overlapping of displacement cascades under neutron irradiation have been considered. It has been shown that, in this case, it is the increase in the size of vacancy clusters, rather than the growth of their concentration, that is prevailing. Possible mechanisms of the radiation hardening of the ODS steel and the Fe-13Cr alloy upon irradiation and subsequent annealing have been discussed.

  15. Influence of an external magnetic field on damage by self-ion irradiation in Fe90Cr10 alloy

    Directory of Open Access Journals (Sweden)

    Fernando José Sánchez

    2016-12-01

    Full Text Available The effect of an external magnetic field (B=0.5 T on Fe90Cr10 specimens during Fe ion irradiation, has been investigated by means of Conversion Electron Mössbauer Spectroscopy (CEMS. The analysis has revealed significant differences in the average hyperfine magnetic field (=0.3 T between non-irradiated and irradiated samples as well as between irradiations made with B (w/ B and without B (w/o B. It is considered that these variations can be due to changes in the local environment around the probe nuclei (57Fe; where vacancies and Cr distribution play a role. The results indicate that the Cr distribution in the neighbourhood of the iron atoms could be changed by the application of an external field. This would imply that an external magnetic field may be an important parameter to take into account in predictive models for Cr behaviour in Fe–Cr alloys, and especially in fusion conditions where intense magnetic fields are required for plasma confinement.

  16. Revised ANL-reported tensile data for unirradiated and irradiated (FFTF, HFIR) V-Ti and V-Cr-Ti alloys

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States)

    1998-03-01

    The tensile data for all unirradiated and irradiated vanadium alloys samples tested at Argonne National Laboratory (ANL) have been critically reviewed and, when necessary, revised. The review and revision are based on reanalyzing the original load-displacement strip chart recordings by a methodology consistent with current ASTM standards. For unirradiated alloys (162 samples), the revised values differ from the previous values as follows: {minus}11{+-}19 MPa ({minus}4{+-}6%) for yield strength (YS), {minus}3{+-}15 MPa ({minus}1{+-}3%) for ultimate tensile strength (UTS), {minus}5{+-}2% strain for uniform elongation (UE), and {minus}4{+-}2% strain for total elongation (TE). Of these changes, the decrease in {minus}1{+-}6 MPa (0{+-}1%) for UTS, {minus}5{+-}2% for UE, and {minus}4{+-}2% for TE. Of these changes, the decrease in UE values for alloys irradiated and tested at 400--435 C is the most significant. This decrease results from the proper subtraction of nongauge-length deformation from measured crosshead deformation. In previous analysis of the tensile curves, the nongauge-length deformation was not correctly determined and subtracted from the crosshead displacement. The previously reported and revised tensile values for unirradiated alloys (20--700 C) are tabulated in Appendix A. The revised tensile values for the FFTF-irradiated (400--600 C) and HFIR-irradiated (400 C) alloys are tabulated in Appendix B, along with the neutron damage and helium levels. Appendix C compares the revised values to the previously reported values for irradiated alloys. Appendix D contains previous and revised values for the tensile properties of unirradiated V-5Cr-5Ti (BL-63) alloy exposed to oxygen.

  17. Study of clustering point defects under irradiation in dilute iron alloys; Etude de la formation sous irradiation des amas de defauts ponctuels dans les alliages ferritiques faiblement allies

    Energy Technology Data Exchange (ETDEWEB)

    Duong-Hardouin Duparc, T.H.A

    1998-12-31

    In low copper steels for nuclear reactor pressure vessel, point defect clustering plays an important role in hardening. These clusters are very small and invisible by transmission electron microscopy. In order to study the hardening component which results from the clustering of freely migrating point defects, we irradiated in a high voltage electron microscope Fe, the FeCu{sub 0.13%}, FeP{sub 0.015%} and FeN{sub 33ppm} alloys and the complex FeMn{sub 1.5%}Ni{sub 0.8%}Cu{sub 0.13%}P{sub 0.01%} alloy the composition of which is close to the matrix of pressure vessel steel. We studied the nucleation of dislocation loops and their growth velocity. The observations and the analyses have shown that in the complex model alloy, the interstitial dislocation loops are smaller and their density is more important than for the others alloys. The diffusion coefficients of interstitials and vacancies are obtained with the help of a simplified model. The densities of dislocation loops and their growth velocities obtained experimentally are reproduced by means of a cluster dynamics model we have developed. This is achieved self-consistently by using as a first trial the approximated coefficients obtained with the simplified model. The results of calculations have shown that the binding energy of di-interstitials must be very important in the binary iron alloys and only 0.95 eV in iron. Copper, nitrogen and phosphorus stabilize di-interstitials in iron. Finally the distribution of interstitial loops at 290 deg C and at 2.10{sup -9} dpa/s is calculated with the diffusion coefficient of point defects adjusted in FeCu. A distribution of small loops appears which gives an increase of hardening estimated to 10 Hv instead of 33 Hv experimentally observed. This low value can be improved by assuming in agreement with molecular dynamics simulations that a little fraction of di-interstitials is created at 2.5 MeV. (author) 111 refs.

  18. Quantitative Microanalysis with high Spatial Resolution: Application of FEG-DTEM XEDS Microanalysis to the Characterization of Complex Microstructures in Irradiated Low Alloy Steet

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.B., Watanabe, M. and Burke, M.G.

    2001-11-14

    To assist in the characterization of microstructural changes associated with irradiation damage in low alloy steels, the technique of quantitative x-ray mapping using a field emission gun scanning transmission electron microscope (FEG-STEM) equipped with an x-ray energy Dispersive spectrometer (XEDS) has been employed. Quantitative XEDS microanalyses of the matrix and grain boundaries of irradiated specimens have been compared with previous quantitative analyses obtained using 3D-Atom Probe Field-Ion Microscopy (3D-APFIM). In addition, the FEG-STEM XEDS maps obtained from the irradiated steel have revealed the presence of 2 to 3 nm Ni-enriched 'precipitates' in the matrix, which had previously been detected using 3D-APFIM. These quantitative FEG-STEM XEDS results represent the first direct and independent microchemical corroboration of the 3D-APFIM results showing ultra-fine irradiation-induced hardening features in low alloy steel.

  19. Effects of heavy-ion irradiation on the grain boundary chemistry of an oxide-dispersion strengthened Fe-12 wt.% Cr alloy

    Science.gov (United States)

    Marquis, Emmanuelle A.; Lozano-Perez, Sergio; Castro, Vanessa de

    2011-10-01

    Understanding the behaviour of oxide-dispersion strengthened (ODS) ferritic martensitic steels under irradiation is of prime importance in the design of future fusion reactors. Although changes in grain boundary chemistry during irradiation can significantly affect fracture strength, little is known on the behaviour of grain boundaries in ODS steels. Here, the effect of heavy-ion implantation at 500 °C on grain boundary chemistry in a model ODS Fe-12 wt.% Cr alloy was investigated using atom-probe tomography (APT) and analytical scanning-transmission electron microscopy ((S)TEM) techniques. While chromium and carbon segregation at grain boundaries is found in annealed alloys before irradiation, the three-dimensional APT reconstructions and TEM observations after irradiation reveal a complex distribution of Cr segregation and depletion at grain boundaries of varying character.

  20. Phase stability in thermally-aged CASS CF8 under heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Meimei, E-mail: mli@anl.gov [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Miller, Michael K. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Chen, Wei-Ying [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2015-07-15

    Highlights: • Thermally-aged CF8 was irradiated with 1 MeV Kr ions at 400 °C. • Atom probe tomography revealed a strong dose dependence of G-phase precipitates. • Phase separation of α and α′ in ferrite was reduced after irradiation. - Abstract: The stability of the microstructure of a cast austenitic stainless steel (CASS), before and after heavy ion irradiation, was investigated by atom probe tomography (APT). A CF8 ferrite–austenite duplex alloy was thermally aged at 400 °C for 10,000 h. After this treatment, APT revealed nanometer-sized G-phase precipitates and Fe-rich α and Cr-enriched α′ phase separated regions in the ferrite. The thermally-aged CF8 specimen was irradiated with 1 MeV Kr ions to a fluence of 1.88 × 10{sup 19} ions/m{sup 2} at 400 °C. After irradiation, APT analysis revealed a strong spatial/dose dependence of the G-phase precipitates and the α–α′ spinodal decomposition in the ferrite. For the G-phase precipitates, the number density increased and the mean size decreased with increasing dose, and the particle size distribution changed considerably under irradiation. The inverse coarsening process can be described by recoil resolution. The amplitude of the α–α′ spinodal decomposition in the ferrite was apparently reduced after heavy ion irradiation.

  1. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  2. Effects of 600 MeV proton irradiation on nucleation and growth of precipitates and helium bubbles in a high-purity Al-Mg-Si alloy

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Leffers, Torben; Victoria, M.;

    1986-01-01

    Solution treated specimens of a high-purity Al-0.75%Mg-0.42%Si alloy were irradiated with 600 MeV protons at 150 and 240°C to a dose level of 0.47 and 0.55 dpa, respectively. Mg2Si-type precipitates formed during irradiation at 150 and 240°C; at 240°C, however, a large number of precipitates seem...

  3. Irradiation effect of nano-bubble dispersion strengthened (N-BDS) alloy

    Energy Technology Data Exchange (ETDEWEB)

    Oono, Naoko, E-mail: n-oono@eng.hokudai.ac.jp [Faculty of Engineering, Hokkaido University, Kita 13, Nishi 8, Kita-ku, Sapporo, Hokkaido 060-8628 (Japan); Kawano, Ryohei [Faculty of Engineering, Hokkaido University, Kita 13, Nishi 8, Kita-ku, Sapporo, Hokkaido 060-8628 (Japan); Shi, Shi, E-mail: shishiamy@gmail.com [Faculty of Engineering, Hokkaido University, Kita 13, Nishi 8, Kita-ku, Sapporo, Hokkaido 060-8628 (Japan); Ukai, Shigeharu, E-mail: s-ukai@eng.hokudai.ac.jp [Faculty of Engineering, Hokkaido University, Kita 13, Nishi 8, Kita-ku, Sapporo, Hokkaido 060-8628 (Japan); Hayashi, Shigenari, E-mail: hayashi@eng.hokudai.ac.jp [Faculty of Engineering, Hokkaido University, Kita 13, Nishi 8, Kita-ku, Sapporo, Hokkaido 060-8628 (Japan); Kondo, Sosuke, E-mail: kondo@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hashitomi, Okinobu, E-mail: o-hashitomi@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kimura, Akihiko, E-mail: kimura@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2013-11-15

    Nano-bubble dispersion strengthened (N-BDS) Fe was made from Fe and polymethylmethacrylate (PMMA) powder and irradiated by 6.4 MeV Fe{sup 3+} ions to investigate the cavity strengthening and the bubble to void evolution. The bubbles accelerated the irradiation-induced cavity growth. The hardness of the N-BDS Fe was 500 MPa higher than that of unalloyed Fe and the hardness increased by irradiation, while that of unalloyed Fe did not increase. Cavity is probably the origin of the irradiation hardening of N-BDS Fe.

  4. 一种奥氏体铁镍基合金中的锯齿流变现象%SERRATED FLOW IN A FeNi-BASED AUSTENITIC ALLOY

    Institute of Scientific and Technical Information of China (English)

    赵帅; 李秀艳; 戎利建

    2011-01-01

    Tensile tests on a FeNi-base austenitic alloy, with different amount of twin bound aries, were conducted at different temperatures and three strain rates, respectively. The results show that serrated flow occurs at temperatures from 300 to 700 ℃. This serrated flow exhibits bulge-like serrations at temperatures from 300 to 600 ℃ and stress-loss serrations at 700 ℃, which manifests the nature of thermal activation, I.e. Higher temperatures boost serrations and higher strain rates de press them. Investigations on samples deformed at room temperature (no serrated flow) and 400 ℃ (prominent serrated flow) indicate that twin boundaries are strong enough to block slip deformation at 400 ℃. As a result, stress accumulates on twin boundaries and bulge-like serrations appear on the tensile curves. Effect of twin boundary amount on the morphologies of serrations testified this mecha nism.%采用3种应变速率在不同温度下对一种奥氏体铁镍基合金进行拉伸实验,结果发现在300-700℃范围内发生锯齿流变现象;锯齿流变现象在300-600℃范围内表现为凸起式锯齿,而在700℃时表现为下凹式锯齿;此锯齿流变具有热激活性,即较高的拉伸温度使锯齿易于发生,较大的拉伸速率则抑制其发生.对比分析室温(不发生锯齿流变)及400℃(显著发生锯齿流变)拉伸试样的微观组织,发现在400℃拉伸后的孪晶界上存在明显的位错塞积,说明孪晶界可以有效地阻碍滑移变形,引起孪晶界附.近应力积累,从而导致了凸起式锯齿流变的发生.孪晶界数量对锯齿波形的影响证实了此观点.

  5. Effect of Primary Factor on Cavitation Resistance of Some Austenitic Metals

    Institute of Scientific and Technical Information of China (English)

    WANG Zai-you; ZHU Jin-hua

    2003-01-01

    The cavitation resistance of six kinds of austenitic metals was investigated using a rotating disc rig. The research results show that cavitation resistance of the austenitic metals is obviously raised due to cavitation-induced martensite and greatly influenced by mechanism of martensitic transformation. The cavitation resistance of two stress-induced martensite Fe-Mn-Si-Cr shape memory alloys is much better than that of three strain-induced martensite austenitic stainless steels. The Fe-Mn-Si-Cr shape memory alloys possess excellent cavitation resistance mainly because of their excellent elasticity in local small-zone. The first principal factor for cavitation resistance of metastable austenitic metals is unloaded rebounding depth, and the second one is energy dissipation resulted from cavitation-induced martensite.

  6. The Prediction of Long-Term Thermal Aging in Cast Austenitic Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang; Yang, Ying; Lach, Timothy G.

    2017-02-15

    Cast austenitic stainless steel (CASS) materials are extensively used for many massive primary coolant system components of light water reactors (LWRs) including coolant piping, valve bodies, pump casings, and piping elbows. Many of these components are operated in complex and persistently damaging environments of elevated temperature, high pressure, corrosive environment, and sometimes radiation for long periods of time. Since a large number of CASS components are installed in every nuclear power plant and replacing such massive components is prohibitively expensive, any significant degradation in mechanical properties that affects structural integrity, cracking resistance in particular, of CASS components will raise a serious concern on the performance of entire power plant. The CASS materials for nuclear components are highly corrosion-resistant Fe-Cr-Ni alloys with 300 series stainless steel compositions and mostly austenite (γ)–ferrite (δ) duplex structures, which result from the casting processes consisting of alloy melting and pouring or injecting liquid metal into a static or spinning mold. Although the commonly used static and centrifugal casting processes enable the fabrication of massive components with proper resistance to environmental attacks, the alloying and microstructural conditions are not highly controllable in actual fabrication, especially in the casting processes of massive components. In the corrosion-resistant Fe-Cr-Ni alloy system, the minor phase (i.e., the δ-ferrite phase) is inevitably formed during the casting process, and is in a non-equilibrium state subject to detrimental changes during exposure to elevated temperature and/or radiation. In general, relatively few critical degradation modes are expected within the current design lifetime of 40 years, given that the CASS components have been processed properly. It has been well known, however, that both the thermal aging and the neutron irradiation can cause degradation of static

  7. Phase-field Model for Interstitial Loop Growth Kinetics and Thermodynamic and Kinetic Models of Irradiated Fe-Cr Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yulan; Hu, Shenyang Y.; Sun, Xin; Khaleel, Mohammad A.

    2011-06-15

    strength of interstitial loop for interstitials. In part II, we present a generic phase field model and discuss the thermodynamic and kinetic properties in phase-field models including the reaction kinetics of radiation defects and local free energy of irradiated materials. In particular, a two-sublattice thermodynamic model is suggested to describe the local free energy of alloys with irradiated defects. Fe-Cr alloy is taken as an example to explain the required thermodynamic and kinetic properties for quantitative phase-field modeling. Finally the great challenges in phase-field modeling will be discussed.

  8. Defect evolution in a Nisbnd Mosbnd Crsbnd Fe alloy subjected to high-dose Kr ion irradiation at elevated temperature

    Science.gov (United States)

    de los Reyes, Massey; Voskoboinikov, Roman; Kirk, Marquis A.; Huang, Hefei; Lumpkin, Greg; Bhattacharyya, Dhriti

    2016-06-01

    A candidate Nisbnd Mosbnd Crsbnd Fe alloy (GH3535) for application as a structural material in a molten salt nuclear reactor was irradiated with 1 MeV Kr2+ ions (723 K, max dose of 100 dpa) at the IVEM-Tandem facility. The evolution of defects like dislocation loops and vacancy- and self-interstitial clusters was examined in-situ. For obtaining a deeper insight into the true nature of these defects, the irradiated sample was further analysed under a TEM post-facto. The results show that there is a range of different types of defects formed under irradiation. Interaction of radiation defects with each other and with pre-existing defects, e.g., linear dislocations, leads to the formation of complex microstructures. Molecular dynamics simulations used to obtain a greater understanding of these defect transformations showed that the interaction between linear dislocations and radiation induced dislocation loops could form faulted structures that explain the fringed contrast of these defects observed in TEM.

  9. Effet d'un enrichissement en nickel sur la stabilite mecanique de l'austenite de reversion lorsque soumise a de la fatigue oligocyclique

    Science.gov (United States)

    Godin, Stephane

    The effect of nickel enrichment on the mechanical stability of the reversed austenite contained in martensitic stainless steels 13%Cr-4%Ni and 13%Cr-6%Ni was investigated. The main objective of the study was to observe their microstructure and to compare the dynamic behaviour of the reversed austenite. Tempers made at different temperatures showed that the 6% Ni alloy began to form more austenite and at a lower temperature. SEM and TEM analysis were used to see the austenite and measure its chemical composition. It has been observed that it was richer in Ni than the surrounding martensite. This enrichment increased with tempering temperature and caused an impoverishment of the surrounding martensite. The study also showed that the chemical composition of the austenite formed at the peak (maximum) of both alloys was similar. For a same tempering, this suggests Ni can help to form more austenite but this austenite is not necessarily richer in Ni. The analysis also showed that the austenite was predominantly lamellar and located at the interface and/or inside the martensite laths. Low cycle fatigue tests have shown that the austenite of the 6% Ni alloy was the most mechanically stable even if its Ni content was lower than the 4% Ni alloy austenite. This behaviour was explained by a thinner and narrower morphology of this phase. For a different content of Ni and different quantity of austenite, the most mechanically stable one was in the 4% Ni alloy. It turned out that its reversed austenite was thinner and its surrounding martensite was a bit harder than the 6% Ni alloy austenite. The effect of Ni enrichment of an alloy would be beneficial regarding the mechanical stability if a suitable tempering is made. This tempering must form a thin lamellar austenite in a sufficiently hard martensite. More Ni in the austenite would not necessarily raise the mechanical stability. It could contribute but it seems that it is not be the main factor governing the mechanical stability

  10. Experimental Determination of the Primary Solidification Phase dependency on the solidification velocity for 17 different austenitic stainless steel compositions

    DEFF Research Database (Denmark)

    Laursen, Birthe Nørgaard; Olsen, Flemming Ove; Yardy, John;

    1997-01-01

    to the austenite phase.Most stainless steels are weldable by conventional welding techniques. However, during laser weldng the solidification velocities can be very much higher than by conventional welding techniques. By increasing the solidification velocity to a critical value known as the transition velocity......, the primary solidification phase is found to change from ferrite to austenite.A novel laser remelting technique has been modified to enable the transition velocity for laser welded austenitic stainless steels to be deermined experimentally and on the basis of results from 17 different alloy compositions...... an equation for the calculation of the transition velocity from alloy composition is proposed....

  11. Nano-scale chemical evolution in a proton-and neutron-irradiated Zr alloy

    Science.gov (United States)

    Harte, Allan; Topping, M.; Frankel, P.; Jädernäs, D.; Romero, J.; Hallstadius, L.; Darby, E. C.; Preuss, M.

    2017-04-01

    Proton-and neutron-irradiated Zircaloy-2 are compared in terms of the nano-scale chemical evolution within second phase particles (SPPs) Zr(Fe,Cr)2 and Zr2(Fe,Ni). This is accomplished through ultra-high spatial resolution scanning transmission electron microscopy and the use of energy-dispersive X-ray spectroscopic methods. Fe-depletion is observed from both SPP types after irradiation with both irradiative species, but is heterogeneous in the case of Zr(Fe,Cr)2, predominantly from the edge region, and homogeneously in the case of Zr2(Fe,Ni). Further, there is evidence of a delay in the dissolution of the Zr2(Fe,Ni) SPP with respect to the Zr(Fe,Cr)2. As such, SPP dissolution results in matrix supersaturation with solute under both irradiative species and proton irradiation is considered well suited to emulate the effects of neutron irradiation in this context. The mechanisms of solute redistribution processes from SPPs and the consequences for irradiation-induced growth phenomena are discussed.

  12. Thermodynamic stability of austenitic Ni-Mn-Cu cast iron

    Directory of Open Access Journals (Sweden)

    A. Janus

    2014-07-01

    Full Text Available The performed research was aimed at determining thermodynamic stability of structures of Ni-Mn-Cu cast iron castings. Examined were 35 alloys. The castings were tempered at 900 °C for 2 hours. Two cooling speeds were used: furnace-cooling and water-cooling. In the alloys with the nickel equivalent value less than 20,0 %, partial transition of austenite to martensite took place. The austenite decomposition ratio and the related growth of hardness was higher for smaller nickel equivalent value and was clearly larger in annealed castings than in hardened ones. Obtaining thermodynamically stable structure of castings requires larger than 20,0 % value of the nickel equivalent.

  13. Irradiation effect of swift heavy ion for Zr{sub 50}Cu{sub 40}Al{sub 10} bulk glassy alloy

    Energy Technology Data Exchange (ETDEWEB)

    Onodera, Naoto; Ishii, Akito; Ishii, Kouji; Iwase, Akihiro [Department of Materials Science, Osaka Prefecture University, 1-1, Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Yokoyama, Yoshihiko [Institute for Materials Research, Tohoku University, 2-1-1, Katahira, Aoba-ku, Sendai 980-8577 (Japan); Saitoh, Yuichi [Japan Atomic Energy Agency (JAEA), Takasaki Advanced Radiation Research Institute, 1233, Watanuki-machi, Takasaki, Gunma 370-1292 (Japan); Ishikawa, Norito [Japan Atomic Energy Agency (JAEA), Tokai Research and Development Center, Naka-ku, Ibaraki 319-1195 (Japan); Yabuuchi, Atsushi [Research Organization for the 21st Century, Osaka Prefecture University, 1-1, Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Hori, Fuminobu, E-mail: horif@mtr.osakafu-u.ac.jp [Department of Materials Science, Osaka Prefecture University, 1-1, Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan)

    2013-11-01

    It has been reported that heavy ion irradiation causes softening in some cases of Zr-based bulk metallic glass alloys. However, the fundamental mechanisms of such softening have not been clarified yet. In this study, Zr{sub 50}Cu{sub 40}Al{sub 10} bulk glassy alloys were irradiated with heavy ions of 10 MeV I at room temperature. The maximum fluence was 3 × 10{sup 14} ions/cm{sup 2}. The positron annihilation measurements have performed before and after irradiation to investigate changes in free volume. We discuss the relationship between the energy loss and local open volume change after 10 MeV I irradiation compared with those obtained for 200 MeV Xe and 5 MeV Al. The energy loss analysis in ion irradiation for the positron lifetime has revealed that the decreasing trend of positron lifetime is well expressed as a function of total electronic energy deposition rather than total elastic energy deposition. It means that the positron lifetime change by the irradiation has a relationship with the inelastic collisions with electrons during heavy ion irradiation.

  14. Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Marquis, Emmanuelle [Univ. of Michigan, Ann Arbor, MI (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-03-28

    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

  15. Combined nano-SIMS/AFM/EBSD analysis and atom probe tomography, of carbon distribution in austenite/ε-martensite high-Mn steels.

    Science.gov (United States)

    Seol, Jae-Bok; Lee, B-H; Choi, P; Lee, S-G; Park, C-G

    2013-09-01

    We introduce a new experimental approach for the identification of the atomistic position of interstitial carbon in a high-Mn binary alloy consisting of austenite and ε-martensite. Using combined nano-beam secondary ion mass spectroscopy, atomic force microscopy and electron backscatter diffraction analyses, we clearly observe carbon partitioning to austenite. Nano-beam secondary ion mass spectroscopy and atom probe tomography studies also reveal carbon trapping at crystal imperfections as identified by transmission electron microscopy. Three main trapping sites can be distinguished: phase boundaries between austenite and ε-martensite, stacking faults in austenite, and prior austenite grain boundaries. Our findings suggest that segregation and/or partitioning of carbon can contribute to the austenite-to-martensite transformation of the investigated alloy.

  16. Effect of electron irradiation on texturing in electrodeposited nanocrystalline alloy Fe-78%Ni

    Energy Technology Data Exchange (ETDEWEB)

    Bugaychuk, S.M. [G.V. Kurdyumov Institute for Metal Physics, N.A.S. of Ukraine, 36 Acad. Vernadsky avenue, UA-03680 Kyiv-142 (Ukraine)], E-mail: sb@imp.kiev.ua; Nadutov, V.M. [G.V. Kurdyumov Institute for Metal Physics, N.A.S. of Ukraine, 36 Acad. Vernadsky avenue, UA-03680 Kyiv-142 (Ukraine); Karpets, M.V. [I.M. Frantsevich Institute for Problems of Materials Science, N.A.S. of Ukraine, 3 Krzhyzhanovsky Street, UA-03680 Kyiv-142 (Ukraine); Troschenkov, Yu.M. [Institute for Magnetism, N.A.S. of Ukraine, 36-b Acad. Vernadsky avenue, UA-03680 Kyiv-142 (Ukraine)

    2007-12-15

    The effects of annealing and 5 MeV electron irradiation (performed at the same temperature of 150 {sup o}C for 100 h) on texturing in nanocrystalline Permalloy Fe-78%Ni were analyzed. In as-deposited L1{sub 2}-type ordered fcc Ni{sub 3}Fe intermetallic compound, both the texturing caused by annealing and the lowering of saturation magnetization are suppressed by irradiation, whereas atomic distribution remains unchanged.

  17. Characteristics of the aluminum alloy plasma produced by a 1064 nm Nd:YAG laser with different irradiances

    Indian Academy of Sciences (India)

    W F Luo; X X Zhao; Q B Sun; C X Gao; J Tang; H J Wang; W Zhao

    2010-06-01

    The plasma generated by 1064 nm Nd:YAG laser irradiation of aluminum alloy in air at atmospheric pressure was studied spectroscopically. The electron density inferred by measuring the Stark-broadened line profile of Si(I) 288.16 nm decreases with increasing distance from the target surface. The electron temperature was determined using the Boltzmann plot method with nine strong neutral aluminum lines. Due to the thermal conduction towards the solid target and radiative cooling of the plasma as well as conversion of thermal energy into kinetic energy, the electron temperature decreases both at the plasma edge and close to the target surface. Electron density and electron temperature were also studied as functions of laser power density. At the same time, the validity of the assumption of local thermodynamic equilibrium and the effect of self-absorption were discussed in light of the results obtained.

  18. Synthesis of carbon-supported PtRh random alloy nanoparticles using electron beam irradiation reduction method

    Science.gov (United States)

    Matsuura, Yoshiyuki; Seino, Satoshi; Okazaki, Tomohisa; Akita, Tomoki; Nakagawa, Takashi; Yamamoto, Takao A.

    2016-05-01

    Bimetallic nanoparticle catalysts of PtRh supported on carbon were synthesized using an electron beam irradiation reduction method. The PtRh nanoparticle catalysts were composed of particles 2-3 nm in size, which were well dispersed on the surface of the carbon support nanoparticles. Analyses of X-ray diffraction and scanning transmission electron microscopy-energy-dispersive X-ray spectroscopy revealed that the PtRh nanoparticles have a randomly alloyed structure. The lattice constant of the PtRh nanoparticles showed good correlation with Vegard's law. These results are explained by the radiochemical formation process of the PtRh nanoparticles. Catalytic activities of PtRh/C nanoparticles for ethanol oxidation reaction were found to be higher than those obtained with Pt/C.

  19. Development of ODS ferritic-martensitic steels for application to high temperature and irradiation environment; Developpement d'une nouvelle nuance martensitique ODS pour utilisation sous rayonnement a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Lambard, V

    2000-07-01

    Iron oxide dispersion strengthened alloys are candidate for nuclear fuel cladding. Therefore, it is crucial to control their microstructure in order to optimise their mechanical properties at temperatures up to 700 deg C. The industrial candidates, ODS ferritic alloys, present an anisotropic microstructure which induces a weakening of mechanical properties in transversal direction as well as the precipitation of brittle phases under thermal aging and irradiation. For this purpose, we tried to develop a material with isotropic properties. We studied several 9Cr-1Mo ferritic/martensitic alloys, strengthened or not by oxide dispersion. The mechanical alloying was performed by attribution and powders were consolidated by hot extrusion. In this work, different metallurgical characterisation techniques and modelling were used to optimise a new martensitic ODS alloy. Microstructural and chemical characterization of matrix has been done. The effect of austenitizing and isochronal tempering treatments on microstructure and hardness has been studied. Oxide distribution, size and chemical composition have been studied before and after high temperature thermal treatment. The study of phase transformation upon heating has permitted the extrapolation to the equilibrium temperature formation of austenite. Phase transformation diagrams upon cooling have been determined and the transformation kinetics have been linked to austenite grain size by a simple relation. Fine grain size is unfavourable for the targeted application, so a particular thermal treatment inducing a coarser grain structure has been developed. Finally, tensile properties have been determined for the different microstructures. (author)

  20. Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Busby, Jeremy T [ORNL; Gussev, Maxim N [ORNL

    2011-04-01

    Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today s nuclear power reactor fleet and affects critical structural components within the reactor core. The effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be

  1. Influence of irradiation parameters on damage accumulation in metals and alloys

    DEFF Research Database (Denmark)

    Singh, B.N.; Zinkle, S.J.

    1994-01-01

    It is well known that a fraction of defects produced during irradiation accumulate in crystalline solids in the form of clusters of self-interstitial atoms (SIAs) and vacancies, loops, tetrahedra, dislocation segments and cavities. The irradiation parameters such as recoil energy, damage rate...... energy plays a significant role in determining the magnitude of the damage accumulation. Unfortunately, the available experimental results are not sufficient to allow an unambiguous identification of the mechanisms involved in the evolution of the damage accumulation. Some suggestions are made...

  2. Microstructural studies of advanced austenitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Todd, J. A.; Ren, Jyh-Ching [University of Southern California, Los Angeles, CA (USA). Dept. of Materials Science

    1989-11-15

    This report presents the first complete microstructural and analytical electron microscopy study of Alloy AX5, one of a series of advanced austenitic steels developed by Maziasz and co-workers at Oak Ridge National Laboratory, for their potential application as reheater and superheater materials in power plants that will reach the end of their design lives in the 1990's. The advanced steels are modified with carbide forming elements such as titanium, niobium and vanadium. When combined with optimized thermo-mechanical treatments, the advanced steels exhibit significantly improved creep rupture properties compared to commercially available 316 stainless steels, 17--14 Cu--Mo and 800 H steels. The importance of microstructure in controlling these improvements has been demonstrated for selected alloys, using stress relaxation testing as an accelerated test method. The microstructural features responsible for the improved creep strengths have been identified by studying the thermal aging kinetics of one of the 16Ni--14Cr advanced steels, Alloy AX5, in both the solution annealed and the solution annealed plus cold worked conditions. Time-temperature-precipitation diagrams have been developed for the temperature range 600 C to 900 C and for times from 1 h to 3000 h. 226 refs., 88 figs., 10 tabs.

  3. Development of Computational Tools for Predicting Thermal- and Radiation-Induced Solute Segregation at Grain Boundaries in Fe-based Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-30

    Radiation-induced segregation (RIS) has been frequently reported in structural materials such as austenitic, ferritic, and ferritic-martensitic stainless steels (SS) that have been widely used in light water reactors (LWRs). RIS has been linked to secondary degradation effects in SS including irradiation-induced stress corrosion cracking (IASCC). Earlier studies on thermal segregation in Fe-based alloys found that metalloids elements such as P, S, Si, Ge, Sn, etc., embrittle the materials when enrichment was observed at grain boundaries (GBs). RIS of Fe-Cr-Ni-based austenitic steels has been modeled in the U.S. 2015 fiscal year (FY2015), which identified the pre-enrichment due to thermal segregation can have an important role on the subsequent RIS. The goal of this work is to develop thermal segregation models for alloying elements in steels for future integration with RIS modeling.

  4. Development of Computational Tools for Predicting Thermal- and Radiation-Induced Solute Segregation at Grain Boundaries in Fe-based Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-30

    Radiation induced segregation (RIS) has been frequently reported in structural materials such as austenitic, ferritic, and ferritic-martensitic stainless steels (SS) that have been widely used in light water reactors (LWRs). RIS has been linked to secondary degradation effects in SS including irradiation induced stress corrosion cracking (IASCC). Earlier studies on thermal segregation in Fe based alloys found that metalloids elements such as P, S, Si, Ge, Sn etc. embrittle the materials when enrichment was observed at grain boundaries (GBs). RIS of Fe-Cr- Ni based austenitic steels has been modeled in the U.S. 2015 fiscal year (FY2015), which identified the pre-enrichment due to thermal segregation can have an important role on the subsequent RIS. The goal of this work is to develop thermal segregation models for alloying elements in steels for future integration with RIS modeling.

  5. Vanadium alloy irradiation experiment X530 in EBR-II{sup *}

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Hins, A.G. [Argonne National Laboratory, Chicago, IL (United States)] [and others

    1995-04-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994.

  6. Subtask 12H1: Vanadium alloy irradiation experiment X530 in EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Hins, A.G.; Chung, H.M.; Nowicki, L.J.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of the X530 experiment in EBR-II was to obtain early irradiation performance data, particularly the fracture properties, on the new 500-kg production heat of V-4Cr-4Ti material before the scheduled reactor shutdown at the end of September 1994. To obtain early irradiation performance data on the new 500-kg production heat of the V-4Cr-4Ti material before the scheduled EBR-II shutdown, an experiment, X530, was expeditiously designed and assembled. Charpy, compact tension, tensile and TEM specimens with different thermal mechanical treatments (TMTs), were enclosed in two capsules and irradiated in the last run of EBR-II, Run 170, from August 9 through September 27. For comparison, specimens from some of the previous heats were also included in the test. The accrued exposure was 35 effective full power days, yielding a peak damage of {approx}4 dpa in the specimens. The irradiation is now complete and the vehicle is awaiting to be discharged from EBR-II for postirradiation disassembly. 4 figs., 2 tabs.

  7. Wear behavior of austenite containing plate steels

    Science.gov (United States)

    Hensley, Christina E.

    As a follow up to Wolfram's Master of Science thesis, samples from the prior work were further investigated. Samples from four steel alloys were selected for investigation, namely AR400F, 9260, Hadfield, and 301 Stainless steels. AR400F is martensitic while the Hadfield and 301 stainless steels are austenitic. The 9260 exhibited a variety of hardness levels and retained austenite contents, achieved by heat treatments, including quench and tempering (Q&T) and quench and partitioning (Q&P). Samples worn by three wear tests, namely Dry Sand/Rubber Wheel (DSRW), impeller tumbler impact abrasion, and Bond abrasion, were examined by optical profilometry. The wear behaviors observed in topography maps were compared to the same in scanning electron microscopy micrographs and both were used to characterize the wear surfaces. Optical profilometry showed that the scratching abrasion present on the wear surface transitioned to gouging abrasion as impact conditions increased (i.e. from DSRW to impeller to Bond abrasion). Optical profilometry roughness measurements were also compared to sample hardness as well as normalized volume loss (NVL) results for each of the three wear tests. The steels displayed a relationship between roughness measurements and observed wear rates for all three categories of wear testing. Nanoindentation was used to investigate local hardness changes adjacent to the wear surface. DSRW samples generally did not exhibit significant work hardening. The austenitic materials exhibited significant hardening under the high impact conditions of the Bond abrasion wear test. Hardening in the Q&P materials was less pronounced. The Q&T microstructures also demonstrated some hardening. Scratch testing was performed on samples at three different loads, as a more systematic approach to determining the scratching abrasion behavior. Wear rates and scratch hardness were calculated from scratch testing results. Certain similarities between wear behavior in scratch testing

  8. Effect of the Content of Retained Austenite and Grain Size on the Fatigue Bending Strength of Steels Carburized in a Low-Pressure Atmosphere

    Science.gov (United States)

    Kula, P.; Dybowski, K.; Lipa, S.; Januszewicz, B.; Pietrasik, R.; Atraszkiewicz, R.; Wołowiec, E.

    2014-11-01

    The effect of the content of retained austenite and of the initial austenite grain size on high-cycle fatigue of two low-alloy steels 16MnCr5 and 17CrNi6-6 after carburizing in a low-pressure atmosphere (acetylene, ethylene and hydrogen) and subsequent high-pressure gas quenching is investigated.

  9. Austenite Grain Growth and Precipitate Evolution in a Carburizing Steel with Combined Niobium and Molybdenum Additions

    Science.gov (United States)

    Enloe, Charles M.; Findley, Kip O.; Speer, John G.

    2015-11-01

    Austenite grain growth and microalloy precipitate size and composition evolution during thermal processing were investigated in a carburizing steel containing various additions of niobium and molybdenum. Molybdenum delayed the onset of abnormal austenite grain growth and reduced the coarsening of niobium-rich precipitates during isothermal soaking at 1323 K, 1373 K, and 1423 K (1050 °C, 1100 °C, and 1150 °C). Possible mechanisms for the retardation of niobium-rich precipitate coarsening in austenite due to molybdenum are considered. The amount of Nb in solution and in precipitates at 1373 K (1100 °C) did not vary over the holding times evaluated. In contrast, the amount of molybdenum in (Nb,Mo)C precipitates decreased with time, due to rejection of Mo into austenite and/or dissolution of fine Mo-rich precipitates. In hot-rolled alloys, soaking in the austenite regime resulted in coarsening of the niobium-rich precipitates at a rate that exceeded that predicted by the Lifshitz-Slyozov-Wagner relation for volume-diffusion-controlled coarsening. This behavior is attributed to an initial bimodal precipitate size distribution in hot-rolled alloys that results in accelerated coarsening rates during soaking. Modification of the initial precipitate size distribution by thermal processing significantly lowered precipitate coarsening rates during soaking and delayed the associated onset of abnormal austenite grain growth.

  10. Phase stability in thermally-aged CASS CF8 under heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Meimei; Miller, M; Chen, Wei-Ying

    2015-07-01

    The stability of the microstructure of a cast austenitic stainless steel (CASS), before and after heavy ion irradiation, was investigated by atom probe tomography (APT). A CF8 ferrite-austenite duplex alloy was thermally aged at 400 degrees C for 10,000 h. After this treatment, APT revealed nanometer-sized G-phase precipitates and Fe-rich alpha and Cr-enriched alpha' phase separated regions in the ferrite. The thermally-aged CF8 specimen was irradiated with 1 MeV Kr ions to a fluence of 1.88 x 10(19) ions/m(2) at 400 degrees C. After irradiation, APT analysis revealed a strong spatial/dose dependence of the G-phase precipitates and the alpha-alpha' spinodal decomposition in the ferrite. For the G-phase precipitates, the number density increased and the mean size decreased with increasing dose, and the particle size distribution changed considerably under irradiation. The inverse coarsening process can be described by recoil resolution. The amplitude of the alpha-alpha' spinodal decomposition in the ferrite was apparently reduced after heavy ion irradiation. (C) 2015 Elsevier B.V. All rights reserved

  11. Austenite grain growth simulation considering the solute-drag effect and pinning effect

    Science.gov (United States)

    Fujiyama, Naoto; Nishibata, Toshinobu; Seki, Akira; Hirata, Hiroyuki; Kojima, Kazuhiro; Ogawa, Kazuhiro

    2017-01-01

    Abstract The pinning effect is useful for restraining austenite grain growth in low alloy steel and improving heat affected zone toughness in welded joints. We propose a new calculation model for predicting austenite grain growth behavior. The model is mainly comprised of two theories: the solute-drag effect and the pinning effect of TiN precipitates. The calculation of the solute-drag effect is based on the hypothesis that the width of each austenite grain boundary is constant and that the element content maintains equilibrium segregation at the austenite grain boundaries. We used Hillert’s law under the assumption that the austenite grain boundary phase is a liquid so that we could estimate the equilibrium solute concentration at the austenite grain boundaries. The equilibrium solute concentration was calculated using the Thermo-Calc software. Pinning effect was estimated by Nishizawa’s equation. The calculated austenite grain growth at 1473–1673 K showed excellent correspondence with the experimental results. PMID:28179962

  12. Effect of Austenitizing Heat Treatment on the Microstructure and Hardness of Martensitic Stainless Steel AISI 420

    Science.gov (United States)

    Barlow, L. D.; Du Toit, M.

    2012-07-01

    The effect of austenitizing on the microstructure and hardness of two martensitic stainless steels was examined with the aim of supplying heat-treatment guidelines to the user that will ensure a martensitic structure with minimal retained austenite, evenly dispersed carbides and a hardness of between 610 and 740 HV (Vickers hardness) after quenching and tempering. The steels examined during the course of this examination conform in composition to medium-carbon AISI 420 martensitic stainless steel, except for the addition of 0.13% vanadium and 0.62% molybdenum to one of the alloys. Steel samples were austenitized at temperatures between 1000 and 1200 °C, followed by oil quenching. The as-quenched microstructures were found to range from almost fully martensitic structures to martensite with up to 35% retained austenite after quenching, with varying amounts of carbides. Optical and scanning electron microscopy was used to characterize the microstructures, and X-ray diffraction was employed to identify the carbide present in the as-quenched structures and to quantify the retained austenite contents. Hardness tests were performed to determine the effect of heat treatment on mechanical properties. As-quenched hardness values ranged from 700 to 270 HV, depending on the amount of retained austenite. Thermodynamic predictions (using the CALPHAD™ model) were employed to explain these microstructures based on the solubility of the carbide particles at various austenitizing temperatures.

  13. Ordering Intermetallic Alloys by Ion Irradiation: A Way to Tailor Magnetic Media

    Science.gov (United States)

    Bernas, H.; Attané, J.-Ph.; Heinig, K.-H.; Halley, D.; Ravelosona, D.; Marty, A.; Auric, P.; Chappert, C.; Samson, Y.

    2003-08-01

    We show how, combining He ion irradiation and thermal mobility below 600K, the transformation from chemical disorder to order in thin films of an intermetallic ferromagnet (FePd) may be triggered and controlled. Kinetic Monte Carlo simulations show that the initial directional short range order determines the transformation. Magnetic ordering perpendicular to the film plane was achieved, promoting the initially weak magnetic anisotropy to the highest values known for FePd films. Applications to ultrahigh density magnetic recording are suggested.

  14. Comparing the possibilities of austenite content determination in austempered ductile iron

    Directory of Open Access Journals (Sweden)

    D. Myszka

    2011-07-01

    Full Text Available The article presents various methods for assessment of the austenite volume fraction in Austempered Ductile Iron (ADI. Tests were carried out on two types of ADI, i.e. unalloyed and alloyed with the addition of 0.72%Cu and 0.27%Mo, heat treated under different conditions of isothermal transformation to obtain different austenite volume fractions. The test material was then subjected to metallographic examinations, X-ray diffraction (XRD analysis, an analysis using the author's genuine programme of artificial neural networks, image analysis and magnetic measurements. The results were compared with each other indicating the possibility of a quantitative measurement of austenite and other phases present in cast iron. It was found that different methods of measurement are not fully consistent with each other but show similar results of the austenite content.

  15. Role of quaternary additions on dislocated martensite, retain austenite and mechanical properties of Fe/Cr/C structural steels

    Energy Technology Data Exchange (ETDEWEB)

    Rao, B.V.N.

    1978-02-01

    The influence of quaternary alloy additions of Mn and Ni to Fe/Cr/C steels which have been designed to provide superior mechanical properties has been investigated. Transmission electron microscopy and x-ray analysis revealed increasing amounts of retained austenite with Mn up to 2 w/o and with 5 w/o Ni additions after quenching from 1100/sup 0/C. This is accompanied by a corresponding improvement in toughness properties of the quaternary alloys. In addition, the generally attractive combinations of strength and toughness in these quaternary alloys is attributed to the production of dislocated lath martensite from a homogeneous austenite phase free from undissolved alloy carbides. Grain-refining resulted in a further increase in the amount of retained austenite.

  16. Results of irradiating aluminum and homogeneous alloy YMn2 by 23 MeV γ-quanta in a molecular deuterium atmosphere at 2 kbar pressure

    Science.gov (United States)

    Didyk, A. Yu.; Wisniewski, R.

    2014-03-01

    Specimens of a number of metal were placed successively along the length in a deuterium high-pressure chamber of the "finger type" (DHPC-FT). The specimens were: two aluminum rods, a copper rod, two YMn2 alloy specimens, and stainless steel. The molecular deuterium pressure in the DHPC-FT chamber was 2 kbar. The specimens were irradiated by braking γ-quanta with boundary energy 23 MeV. After irradiation, all specimens were investigated on scanning electron microscopes (SEM) with electron probe X-ray microelement analysis (XMA). Considerable changes in the structure of the surfaces and elemental composition of the measured aluminum, destruction of the homogeneous YMn2 alloy specimen, and the "formation of monocrystalline specimens" of the YMn2 type and structures resembling manganese-based "crystals" were observed. A phenomenological explanation of the observed phenomena and effects based on nuclear reactions is proposed with consideration of certain new approaches, which are examined.

  17. Effects of composition and helium injection on dislocation loop development in pure FeNiCr alloys under Ni ion irradiation

    Science.gov (United States)

    Kimoto, Takayoshi

    1993-08-01

    Pure Fe35Ni7Cr, Fe45Ni7Cr, Fe40Ni13Cr and Fe45Ni15Cr alloys were irradiated by 4MeV Ni 2+ ions at 948 K to doses of about 0.05, 0.3 and 1.0 dpa without helium injection or with simultaneous helium injection. With increasing Ni content and decreasing Cr content, the diameter of radiation-induced dislocation loops increased, and the dose at which the dislocation loops disappeared to develop into dislocation networks decreased. The diameter of dislocation loops induced by Ni 2+ ions irradiation with simultaneous helium injection was larger than that without helium injection for the Fe35Ni7Cr and Fe45Ni7Cr alloys.

  18. Alloy

    Science.gov (United States)

    Cabeza, Sandra; Garcés, Gerardo; Pérez, Pablo; Adeva, Paloma

    2014-07-01

    The Mg98.5Gd1Zn0.5 alloy produced by a powder metallurgy route was studied and compared with the same alloy produced by extrusion of ingots. Atomized powders were cold compacted and extruded at 623 K and 673 K (350 °C and 400 °C). The microstructure of extruded materials was characterized by α-Mg grains, and Mg3Gd and 14H-LPSO particles located at grain boundaries. Grain size decreased from 6.8 μm in the extruded ingot, down to 1.6 μm for powders extruded at 623 K (350 °C). Grain refinement resulted in an increase in mechanical properties at room and high temperatures. Moreover, at high temperatures the PM alloy showed superplasticity at high strain rates, with elongations to failure up to 700 pct.

  19. Shape memory alloy thaw sensors

    Science.gov (United States)

    Shahinpoor, Mohsen; Martinez, David R.

    1998-01-01

    A sensor permanently indicates that it has been exposed to temperatures exceeding a critical temperature for a predetermined time period. An element of the sensor made from shape memory alloy changes shape when exposed, even temporarily, to temperatures above the Austenitic temperature of the shape memory alloy. The shape change of the SMA element causes the sensor to change between two readily distinguishable states.

  20. Protective coating of austenitic steel using robotized GMAW temper-bead technique; Rechargement d'inox austenitique en MAG temperbead robotise

    Energy Technology Data Exchange (ETDEWEB)

    Carpreau, J.M. [Electricite de France (EDF/R and D), Recherche et Developpement, 92 - Chatou (France); Dainelli, P. [Institut de Soudure, 57 - Yutz (France)

    2009-07-15

    This paper summarises experimental results obtained in a study of GMAW temper-bead on low alloyed steel with austenitic consumables. Temper-bead on low alloyed steel with austenitic consumables is mainly used for repairing operations of heavy components such as vessel reactor of nuclear power plants. Experimental work aims at showing the performance of GMAW compared to GTAW and the consequences of GMAW temper-bead on 2OMND5 heat affected zones. (authors)

  1. An On-Heating Dilation Conversional Model for Austenite Formation in Hypoeutectoid Steels

    Science.gov (United States)

    Lee, Seok-Jae; Clarke, Kester D.; van Tyne, Chester J.

    2010-09-01

    Dilatometry is often used to study solid-state phase transformations. While most steel transformation studies focus on the decomposition of austenite, this article presents an on-heating dilation conversional model to determine phase fraction based on measured volume changes during the formation of austenite in ferrite-pearlite hypoeutectoid steels. The effect of alloying elements on the transformation strain is incorporated into the model. Comparison of the conversional model predictions to measured transformation temperature ( A c3) shows excellent agreement. The pearlite decomposition finish temperature ( A pf ) predicted by the conversional model more closely matches experimental results when compared to standard lever rule calculations. Results show that including the effects of substitutional alloying elements (in addition to carbon) improves phase fraction predictions. The conversional model can be used to quantitatively predict intercritical austenite fraction with application to modeling, induction heating, intercritical annealing, and more complex heat treatments for hypoeutectoid steels.

  2. Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement en fluage des alliages de zircomium irradies

    Energy Technology Data Exchange (ETDEWEB)

    Ribis, J

    2007-12-15

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  3. Austenite formation in C-Mn steel

    OpenAIRE

    Savran, V.I.

    2009-01-01

    The production process of almost all modern steels involves austenitization formation of the austenite phase upon continuous heating. Many of the microstructural features and properties that are obtained upon subsequent cooling are to a large extend determined by the evolution of the microstructure and chemical inhomogeneities during austenitization. In spite of its importance, austenitization so far has received much less attention than the transformations on cooling; however, the interest i...

  4. A study of austenitization of SG iron

    Indian Academy of Sciences (India)

    Uma Batra; Pankaj Tandon; Kulbir Kaur

    2000-10-01

    Austenitization process of three SG irons with varying compositions and as cast matrix microstructure has been studied at three austenitization temperatures of 850, 900 and 950C for different time periods. Microstructure, hardness and X-ray diffraction have been used to reveal the nature of dependence of the process on austenitization temperature, time and as cast structure. The optimum austenitization time is maximum for ferritic and minimum for pearlitic matrix.

  5. The effects of ion irradiation on the micromechanical fracture strength and hardness of a self-passivating tungsten alloy

    Science.gov (United States)

    Lessmann, Moritz T.; Sudić, Ivan; Fazinić, Stjepko; Tadić, Tonči; Calvo, Aida; Hardie, Christopher D.; Porton, Michael; García-Rosales, Carmen; Mummery, Paul M.

    2017-04-01

    An ultra-fine grained self-passivating tungsten alloy (W88-Cr10-Ti2 in wt.%) has been implanted with iodine ions to average doses of 0.7 and 7 dpa, as well as with helium ions to an average concentration of 650 appm. Pile-up corrected Berkovich nanoindentation reveals significant irradiation hardening, with a maximum hardening of 1.9 GPa (17.5%) observed. The brittle fracture strength of the material in all implantation conditions was measured through un-notched cantilever bending at the microscopic scale. All cantilever beams failed catastrophically in an intergranular fashion. A statistically confirmed small decrease in strength is observed after low dose implantation (-6%), whilst the high dose implantation results in a significant increase in fracture strength (+9%), further increased by additional helium implantation (+16%). The use of iodine ions as the implantation ion type is justified through a comparison of the hardening behaviour of pure tungsten under tungsten and iodine implantation.

  6. Delta ferrite-containing austenitic stainless steel resistant to the formation of undesirable phases upon aging

    Science.gov (United States)

    Leitnaker, James M.

    1981-01-01

    Austenitic stainless steel alloys containing delta ferrite, such as are used as weld deposits, are protected against the transformation of delta ferrite to sigma phase during aging by the presence of carbon plus nitrogen in a weight percent 0.015-0.030 times the volume percent ferrite present in the alloy. The formation of chi phase upon aging is controlled by controlling the Mo content.

  7. All-proportional solid-solution Rh–Pd–Pt alloy nanoparticles by femtosecond laser irradiation of aqueous solution with surfactant

    Energy Technology Data Exchange (ETDEWEB)

    Sarker, Md. Samiul Islam, E-mail: samiul-phy@ru.ac.bd; Nakamura, Takahiro; Sato, Shunichi [Tohoku University, Institute of Multidisciplinary Research for Advanced Materials (Japan)

    2015-06-15

    Formation of Rh–Pd–Pt solid-solution alloy nanoparticles (NPs) by femtosecond laser irradiation of aqueous solution in the presence of polyvinylpyrrolidone (PVP) or citrate as a stabilizer was studied. It was found that the addition of surfactant (PVP or citrate) significantly contributed to reduce the mean size of the particles to 3 nm for PVP and 10 nm for citrate, which was much smaller than that of the particles fabricated without any surfactants (20 nm), and improved the dispersion state as well as the colloidal stability. The solid-solution formation of the Rh–Pd–Pt alloy NPs was confirmed by the XRD results that the diffraction pattern was a single peak, which was found between the positions corresponding to each pure Rh, Pd, and Pt NPs. Moreover, all the elements were homogeneously distributed in every particle by STEM-EDS elemental mapping, strongly indicating the formation of homogeneous solid-solution alloy. Although the Rh–Pd–Pt alloy NPs fabricated with PVP was found to be Pt rich by EDS observation, the composition of NPs fabricated with citrate almost exactly preserved the feeding ratio of ions in the mixed solution. To our best knowledge, these results demonstrated for the first time, the formation of all-proportional solid-solution Rh–Pd–Pt alloy NPs with well size control.

  8. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  9. Study of biocompatibility of medical grade high nitrogen nickel-free austenitic stainless steel in vitro.

    Science.gov (United States)

    Li, Menghua; Yin, Tieying; Wang, Yazhou; Du, Feifei; Zou, Xingzheng; Gregersen, Hans; Wang, Guixue

    2014-10-01

    Adverse effects of nickel ions being released into the living organism have resulted in development of high nitrogen nickel-free austenitic stainless steels for medical applications. Nitrogen not only replaces nickel for austenitic structure stability but also improves steel properties. The cell cytocompatibility, blood compatibility and cell response of high nitrogen nickel-free austenitic stainless steel were studied in vitro. The mechanical properties and microstructure of this stainless steel were compared to the currently used 316L stainless steel. It was shown that the new steel material had comparable basic mechanical properties to 316L stainless steel and preserved the single austenite organization. The cell toxicity test showed no significant toxic side effects for MC3T3-E1 cells compared to nitinol alloy. Cell adhesion testing showed that the number of MC3T3-E1 cells was more than that on nitinol alloy and the cells grew in good condition. The hemolysis rate was lower than the national standard of 5% without influence on platelets. The total intracellular protein content and ALP activity and quantification of mineralization showed good cell response. We conclude that the high nitrogen nickel-free austenitic stainless steel is a promising new biomedical material for coronary stent development.

  10. A review on nickel-free nitrogen containing austenitic stainless steels for biomedical applications.

    Science.gov (United States)

    Talha, Mohd; Behera, C K; Sinha, O P

    2013-10-01

    The field of biomaterials has become a vital area, as these materials can enhance the quality and longevity of human life. Metallic materials are often used as biomaterials to replace structural components of the human body. Stainless steels, cobalt-chromium alloys, commercially pure titanium and its alloys are typical metallic biomaterials that are being used for implant devices. Stainless steels have been widely used as biomaterials because of their very low cost as compared to other metallic materials, good mechanical and corrosion resistant properties and adequate biocompatibility. However, the adverse effects of nickel ions being released into the human body have promoted the development of "nickel-free nitrogen containing austenitic stainless steels" for medical applications. Nitrogen not only replaces nickel for austenitic structure stability but also much improves steel properties. Here we review the harmful effects associated with nickel and emphatically the advantages of nitrogen in stainless steel, as well as the development of nickel-free nitrogen containing stainless steels for medical applications. By combining the benefits of stable austenitic structure, high strength, better corrosion and wear resistance and superior biocompatibility in comparison to the currently used austenitic stainless steel (e.g. 316L), the newly developed nickel-free high nitrogen austenitic stainless steel is a reliable substitute for the conventionally used medical stainless steels.

  11. Sensitization, intergranular attack, stress corrosion cracking, and irradiation effects on the corrosion of iron--chromium--nickel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wu, P.C.S.

    1978-04-01

    A literature review is presented on the sensitization, intergranular attack, and stress corrosion cracking of austenitic stainless steels with emphasis on dilute solutions at temperatures below the boiling point of water. An attempt is made to list the possible sources of contaminants during manufacture, shipping, construction and all phases of operation of the sodium containing components. The susceptibility of the different materials to stress corrosion cracking in the various contaminants is discussed and suggestions to prevent serious problems are made. (GHT)

  12. Synergistic Computational and Microstructural Design of Next- Generation High-Temperature Austenitic Stainless Steels

    Energy Technology Data Exchange (ETDEWEB)

    Karaman, Ibrahim [Texas A& M Engineering Experiment Station, College Station, TX (United States); Arroyave, Raymundo [Texas A& M Engineering Experiment Station, College Station, TX (United States)

    2015-07-31

    The purpose of this project was to: 1) study deformation twinning, its evolution, thermal stability, and the contribution on mechanical response of the new advanced stainless steels, especially at elevated temperatures; 2) study alumina-scale formation on the surface, as an alternative for conventional chromium oxide, that shows better oxidation resistance, through alloy design; and 3) design new generation of high temperature stainless steels that form alumina scale and have thermally stable nano-twins. The work involved few baseline alloys for investigating the twin formation under tensile loading, thermal stability of these twins, and the role of deformation twins on the mechanical response of the alloys. These baseline alloys included Hadfield Steel (Fe-13Mn-1C), 316, 316L and 316N stainless steels. Another baseline alloy was studied for alumina-scale formation investigations. Hadfield steel showed twinning but undesired second phases formed at higher temperatures. 316N stainless steel did not show signs of deformation twinning. Conventional 316 stainless steel demonstrated extensive deformation twinning at room temperature. Investigations on this alloy, both in single crystalline and polycrystalline forms, showed that deformation twins evolve in a hierarchical manner, consisting of micron–sized bundles of nano-twins. The width of nano-twins stays almost constant as the extent of strain increases, but the width and number of the bundles increase with increasing strain. A systematic thermomechanical cycling study showed that the twins were stable at temperatures as high as 900°C, after the dislocations are annealed out. Using such cycles, volume fraction of the thermally stable deformation twins were increased up to 40% in 316 stainless steel. Using computational thermodynamics and kinetics calculations, we designed two generations of advanced austenitic stainless steels. In the first generation, Alloy 1, which had been proposed as an alumina

  13. Microstructure Change of Ag-In-Cd Alloy after Irradiation in Reactor%Ag-In-Cd合金辐照后的微观组织变化

    Institute of Scientific and Technical Information of China (English)

    龙冲生; 肖红星; 高雯; 陈洪生

    2015-01-01

    Ag-In-Cd control rods are widely used in PWR nuclear power plants.The irradiation swelling behavior of Ag-In-Cd alloy is very important to the safety assess-ment of control rod during its operation.In this work,to simulate the change of micro-structure and density of Ag-In-Cd alloy after irradiation in reactor,a series of simulation alloys were prepared,and the effect of composition on the microstructure and density was investigated.A formula to calculate the alloy density with different compositions was obtained by fitting the experimental values.It is found that simulation alloy will consist of two phases,an fcc phase and an hcp phase,when the content of Ag was 77.5% (mass fraction).In the fcc phase,Ag content will be higher than its average content,and there is a little amount of Sn.In the hcp phase,Ag content will be below its average content,and Sn content will be relatively high.After irradiation,Ag-In-Cd alloy will be single hcp phase when Ag content is between 5 5% and 6 1%.%Ag-In-Cd合金在核电站压水堆控制棒中广泛使用,其辐照肿胀行为是评价Ag-In-Cd控制棒使用寿命的关键因素。本文通过制备不同成分的模拟合金,来模拟Ag-In-Cd合金在堆内辐照后的成分变化,分析合金的密度及微观组织特点。结果发现,当Ag含量低至77.5%(质量分数)时,合金会分解为fcc和hcp两相,fcc相中贫Sn高Ag,hcp相中富Sn低Ag。当Ag含量在55%~61%之间时,合金以hcp单相存在。由实测的密度拟合出了合金密度随成分变化的关系式。此结果对于理解和掌握Ag-In-Cd合金的辐照肿胀行为有重要意义。

  14. Formation and evolution of MnNi clusters in neutron irradiated dilute Fe alloys modelled by a first principle-based AKMC method

    Energy Technology Data Exchange (ETDEWEB)

    Ngayam-Happy, R. [EDF-R and D, Departement Materiaux et Mecanique des Composants (MMC), Les Renardieres, F-77818 Moret sur Loing Cedex (France); Unite Materiaux et Transformations (UMET), UMR CNRS 8207, Universite de Lille 1, ENSCL, F-59655 Villeneuve d' Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France); Becquart, C.S., E-mail: charlotte.becquart@univ-lille1.fr [Unite Materiaux et Transformations (UMET), UMR CNRS 8207, Universite de Lille 1, ENSCL, F-59655 Villeneuve d' Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France); Domain, C. [EDF-R and D, Departement Materiaux et Mecanique des Composants (MMC), Les Renardieres, F-77818 Moret sur Loing Cedex (France); Unite Materiaux et Transformations (UMET), UMR CNRS 8207, Universite de Lille 1, ENSCL, F-59655 Villeneuve d' Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France)

    2012-07-15

    An atomistic Monte Carlo model parameterised on electronic structure calculations data has been used to study the formation and evolution under irradiation of solute clusters in Fe-MnNi ternary and Fe-CuMnNi quaternary alloys. Two populations of solute rich clusters have been observed, which can be discriminated by whether or not the solute atoms are associated with self-interstitial clusters. Mn-Ni-rich clusters are observed at a very early stage of the irradiation in both modelled alloys, whereas the quaternary alloys contain also Cu-containing clusters. Mn-Ni-rich clusters nucleate very early via a self-interstitial-driven mechanism, earlier than Cu-rich clusters; the latter, however, which are likely to form via a vacancy-driven mechanism, grow in number much faster than the former, helped by the thermodynamic driving force to Cu precipitation in Fe, thereby becoming dominant in the low dose regime. The kinetics of the number density increase of the two populations is thus significantly different. Finally the main conclusion suggested by this work is that the so-called late blooming phases might as well be neither late, nor phases.

  15. The effect of irradiation on tensile properties and fracture toughness of CuCrZr and CuCrNiSi alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G.M., E-mail: gmk@nikiet.ru [OJSC ' NIKIET' , P.O.B. 788, Moscow 101000 (Russian Federation); Artyugin, A.S.; Yvseev, M.V.; Shushlebin, V.V.; Sinelnikov, L.P. [OJSC ' IRM' , Zarechnyi, 624250 Sverdlovsk Region (Russian Federation); Strebkov, Yu.S. [OJSC ' NIKIET' , P.O.B. 788, Moscow 101000 (Russian Federation)

    2011-10-01

    This paper deals with the effect of irradiation on tensile properties and fracture toughness of CuCrZr and CuCrNiSi alloys, considered for use in some in-vessel components of ITER, where a combination of high strength and heat conduction is essential. The heat treatments were: -CuCrZr, quenching in water after annealing at 950 {sup o}S, cold worked 40-45%, and aged at 475-500 {sup o}S for 3 h. -CuCrNiSi, quenching in water after annealing at 980 {sup o}S and aged for 4 h at 460 {sup o}S. Specimens were irradiated in the IVV-2 reactor at {approx}200 {sup o}S and with irradiation damage of 0.15 and 0.27 dpa. Post-irradiation tests were carried out to assess the tensile properties and fracture toughness of the materials. The tests results show that CuCrNiSi has better strength and retains higher ductility after irradiation, but has somewhat lower crack resistance than CuCrZr.

  16. Solidification cracking in austenitic stainless steel welds

    Indian Academy of Sciences (India)

    V Shankar; T P S Gill; S L Mannan; S Sundaresan

    2003-06-01

    Solidification cracking is a significant problem during the welding of austenitic stainless steels, particularly in fully austenitic and stabilized compositions. Hot cracking in stainless steel welds is caused by low-melting eutectics containing impurities such as S, P and alloy elements such as Ti, Nb. The WRC-92 diagram can be used as a general guide to maintain a desirable solidification mode during welding. Nitrogen has complex effects on weld-metal microstructure and cracking. In stabilized stainless steels, Ti and Nb react with S, N and C to form low-melting eutectics. Nitrogen picked up during welding significantly enhances cracking, which is reduced by minimizing the ratio of Ti or Nb to that of C and N present. The metallurgical propensity to solidification cracking is determined by elemental segregation, which manifests itself as a brittleness temperature range or BTR, that can be determined using the varestraint test. Total crack length (TCL), used extensively in hot cracking assessment, exhibits greater variability due to extraneous factors as compared to BTR. In austenitic stainless steels, segregation plays an overwhelming role in determining cracking susceptibility.

  17. Work Hardening Behavior and Stability of Retained Austenite for Quenched and Partitioned Steels

    Institute of Scientific and Technical Information of China (English)

    Cun-yu WANG; Ying CHANG; Jie YANG; Wen-quan CAO; Han DONG; Yi-de WANG

    2016-01-01

    Both microstructure and mechanical properties of low alloy steels treated by quenching and partitioning (Q&P)process were examined.The mixed microstructure of martensite and large-fractioned retained austenite (about 27�3%)was characterized and analyzed,excellent combinations of total elongation of 1 9% and tensile strength of 1 835 MPa were obtained,and three-stage work hardening behavior was demonstrated during tensile test.The en-hanced mechanical properties and work hardening behavior were explained based on the transformation-induced plas-ticity effect of large-fractioned austenite.

  18. General and Localized corrosion of Austenitic and Borated Stainless Steels in Simulated Concentrated Ground Waters

    Energy Technology Data Exchange (ETDEWEB)

    D. Fix; J. Estill; L. Wong; R. Rebak

    2004-05-28

    Boron containing stainless steels are used in the nuclear industry for applications such as spent fuel storage, control rods and shielding. It was of interest to compare the corrosion resistance of three borated stainless steels with standard austenitic alloy materials such as type 304 and 316 stainless steels. Tests were conducted in three simulated concentrated ground waters at 90 C. Results show that the borated stainless were less resistant to corrosion than the witness austenitic materials. An acidic concentrated ground water was more aggressive than an alkaline concentrated ground water.

  19. General and Localized Corrosion of Austenitic And Borated Stainless Steels in Simulated Concentrated Ground Waters

    Energy Technology Data Exchange (ETDEWEB)

    Estill, J C; Rebak, R B; Fix, D V; Wong, L L

    2004-03-11

    Boron containing stainless steels are used in the nuclear industry for applications such as spent fuel storage, control rods and shielding. It was of interest to compare the corrosion resistance of three borated stainless steels with standard austenitic alloy materials such as type 304 and 316 stainless steels. Tests were conducted in three simulated concentrated ground waters at 90 C. Results show that the borated stainless were less resistant to corrosion than the witness austenitic materials. An acidic concentrated ground water was more aggressive than an alkaline concentrated ground water.

  20. Thermodynamic Calculation Study on Effect of Manganese on Stability of Austenite in High Nitrogen Stainless Steels

    Science.gov (United States)

    Wang, Qingchuan; Zhang, Bingchun; Yang, Ke

    2016-07-01

    A series of high nitrogen steels were studied by using thermodynamic calculations to investigate the effect of manganese on the stability of austenite. Surprisingly, it was found that the austenite stabilizing ability of manganese was strongly weakened by chromium, but it was strengthened by molybdenum. In addition, with an increase of manganese content, the ferrite stabilizing ability of chromium significantly increased, but that of molybdenum decreased. Therefore, strong interactions exist between manganese and the other alloying elements, which should be the main reason for the difference among different constituent diagrams.

  1. Material Characterization of Fatigue Specimens made from Meta-stable Austenitic Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Niffenegger, M.; Grosse, M.; Kalkhof, D.; Leber, H. [Paul Scherrer Institut Villigen (Switzerland); Vincent, A.; Pasco, L.; Morin, M. [Insa de Lyon (France)

    2003-07-01

    The main objective of the EU-project CRETE (Contract No.: FIS5-1999-00280) was to assess the capability and the reliability of innovative NDT-inspection techniques for the detection of material degradation, induced by thermal fatigue and neutron irradiation, of metastable austenitic and ferritic low-alloy steel. Several project partners tested aged or irradiated samples, using various techniques (acoustic, magnetic and thermoelectric). However, these indirect methods require a careful interpretation of the measured signal in terms of micro-structural evolutions due to ageing of the material. Therefore the material had to be characterized in its undamaged, as well as in its damaged state. The present report summarises only the material characterization of the fatigue specimens. It is issued simultaneously as an PSI Bericht and the CRETE work package 3 (WP3) report. Each partner according to their own specifications purchased three materials under investigation, namely AISI 347, AISI 321 and AISI 304L. After sending the material to PSI, all fatigue specimens were manufactured by the same Swiss company. Each partner was responsible for his fatigue tests which are documented in the report WP1, written by FANP. In order to characterize the material in its unfatigued as well as in its fatigued state and to consider microstructural changes related to fatigue damage the methods listed below were employed either by PSI or by INSA de Lyon: (1) Inductive Coupled Plasma Emission Photometry (ICP-OES) was applied to determine the chemical composition, (2) Scanning electron microscopy (SEM) for observing cracks, slip bands between grain and twin boundaries, - Ferromaster for measuring the magnetic permeability, (3) Physical Properties Measuring System (PPMS) for measuring magnetization characteristics, (4) Neutron- and advanced X-ray diffraction methods for the quantitative determination of martensite, - Transmission electron microscopy (TEM) for the observation of crystalline

  2. Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Alexandreanu, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-11-01

    Crack growth rate and fracture toughness J-R curve tests were performed on CF-3 and CF-8 cast austenite stainless steels (CASS) with 13-14% of ferrite. The tests were conducted at ~320°C in either high-purity water with low dissolved oxygen or in simulated PWR water. The cyclic crack growth rates of CF-8 were higher than that of CF-3, and the differences between the aged and unaged specimens were small. No elevated SCC susceptibility was observed among these samples, and the SCC CGRs of these materials were comparable to those of CASS alloys with >23% ferrite. The fracture toughness values of unirradiated CF-3 were similar between unaged and aged specimens, and neutron irradiation decreased the fracture toughness significantly. The fracture toughness of CF-8 was reduced after thermal aging, and declined further after irradiation. It appears that while lowering ferrite content may help reduce the tendency of thermal aging embrittlement, it is not very effective to mitigate irradiation-induced embrittlement. Under a combined condition of thermal aging and irradiation, neutron irradiation plays a dominant role in causing embrittlement in CASS alloys.

  3. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part II: Effects of minor elements on precipitate phase stability during thermal aging

    Science.gov (United States)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    The precipitate phase stability in Fe-15Ni-13Cr base austenitic alloys was investigated as a function of minor alloying additions after thermally aging at 600°C and 675°C for times ranging from 24 h to one year. Seven major precipitate phases were found in aged specimens, including M 23C 6, Laves, Eta (η), TiO, NbC, MC, and M 2P. The types and amounts of precipitate phases varied with alloying element additions, aging temperature, and aging time. By analyzing the composition of each individual particle, it was possible to determine the essential constituent elements for each phase. From this information, a strategy to promote or suppress certain precipitate phases was developed. Among the seven phases, the most desirable precipitate phases were considered to be MC and M 2P, because these particles form on a fine scale with a high number density and, therefore, can serve as effective gas atom trap sites under irradiation.

  4. Unusual response of the binary V-2Si alloy to neutron irradiation in FFTF at 430-600{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Ohnuki, S.; Konoshita, H.; Takahaski, H. [Hokkaido Univ., Sapparo (Japan); Garner, F.A. [Pacific Northwest National Laboratory, Richland, WA (United States)

    1996-04-01

    When V-2Si was irradiated in FFTF at 430, 500 and 600C to doses as high as 80 dpa, a very unusual swelling response was observed in which the swelling appeared to saturate rather quickly at {approx}35% at 430 and 540C, but approached this swelling same level much more slowly at 600C. The possible causes of this phenomenon are discussed as well as the implications of these findings on the swelling behavior of other high swelling vanadium binary alloys.

  5. Influence of irradiation with energy-rich particles on the hardness of the Fe-Cr alloy; Einfluss der Bestrahlung mit energiereichen Teilchen auf die Haerte von Fe-Cr-Legierungen

    Energy Technology Data Exchange (ETDEWEB)

    Heintze, Cornelia

    2013-01-14

    Ferritic/martensitic and oxide dispersion strengthened ferritic/martensitic steels are candidate structural materials for components exposed to high neutron fluxes in future nuclear applications like fusion and generation IV fission reactors. The ductilebrittle transition and its shift to higher temperatures which is predominantly caused by irradiation hardening are main concerns for these materials. In the present work, the irradiation behaviour of binary Fe-Cr model alloys, which represent a simplified model for ferritic/martensitic steels, is studied. To this end irradiation with iron ions is used in order to simulate the neutron-induced damage. Due to the limited penetration depth characterization methods suitable for thin layers have to be applied. In the present case, nanohardness testing and transmission electron microscopy (TEM) are employed. The results, including the irradiation-induced hardness change of the layer as a function of chromium content, fluence and irradiation temperature and, for selected cases, quantitative TEM analyses, were exploited to identify irradiation-induced dislocation loops as one source of irradiation hardening. Additional results of small-angle neutron scattering experiments on neutron-irradiated specimens of the same alloys show that nm-scaled α'-phase precipitates also significantly contribute to the irradiation-induced hardness increase. An Orowan model is used to estimate the obstacle strengths posed to dislocation glide by these lattice defects. The topic is stepwise extended to more complex situations with respect to the irradiation conditions and the materials. Considering simultaneous and sequential irradiations with iron- and helium-ions it is shown that the effect of helium on irradiation hardening depends on the chronological order of the irradiations and that the simultaneous introduction of helium in fusion-relevant concentrations amplifies irradiation hardening based on a synergistic effect. There is no

  6. Ion irradiation induced structural modifications of Fe{sub 81}Mo{sub 8}Cu{sub 1}B{sub 10} NANOPERM-type alloy

    Energy Technology Data Exchange (ETDEWEB)

    Miglierini, Marcel [Institute of Nuclear and Physical Engineering, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology, Bratislava (Slovakia); Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, Prague (Czech Republic); Hasiak, Mariusz [Department of Mechanics and Materials Science, Wroclaw University of Technology (Poland)

    2016-05-15

    Structural modifications and their impact upon magnetic properties are studied in amorphous NANOPERM-type {sup 57}Fe{sub 81}Mo{sub 10}Cu{sub 1}B{sub 10} metallic glass exposed to irradiation with 130 keV N{sup +} ions to the total fluencies of up to 2.5 x 10{sup 17} ions/cm{sup 2}. Using surface sensitive technique of Moessbauer spectrometry, traces of crystalline phases are found already in the as-quenched state after the sample production. On the air side of the ribbons, bcc-Fe dominates whereas on the opposite wheel side, also a presence of Fe{sub 3}O{sub 4} is unveiled. The amount of surface crystallization is higher on the wheel side of the ribbons. After ion irradiation, mostly the air side is affected because it was facing the incident ions. Gradual formation of iron nitrides is observed with increasing ion fluence. Though the radiation damage exhibits itself only at this side of the ribbons, its influence upon bulk magnetic properties is clearly identified by the help of magnetic measurements. Hysteresis loops exhibit changes in their shape as well as coercive field. Along with the formation of magnetic crystalline phases (bcc-Fe and nitrides), they are caused by structural rearrangement which takes place also inside the amorphous residual phase. Structural modifications are confirmed via evolution of hyperfine magnetic fields with ion fluence. Structural modification of the {sup 57}Fe{sub 81}Mo{sub 10}Cu{sub 1}B{sub 10} alloy caused by ion irradiation as demonstrated by microstructural (Moessbauer spectrometry (a,b)) and macroscopic (hysteresis loops (c,d)) measurements. As-quenched (a,c) and 2.5 x 10{sup 17} N{sup +}/cm{sup 2} irradiated (b,d) alloys are compared. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. Deuterium retention in W and W-Re alloy irradiated with high energy Fe and W ions: Effects of irradiation temperature

    Directory of Open Access Journals (Sweden)

    Y. Hatano

    2016-12-01

    Full Text Available Neutron irradiation to W induces defects acting as traps against hydrogen isotopes and transmutation elements such as Re and Os. To investigate synergetic effects on radiation-induced defects and Re, deuterium (D retention in W and W–5% Re samples were examined after irradiation with 6.4MeV Fe ions at 523–1273K followed by exposure to D2 gas at 673K. The value of D retention in W–5% Re was lower than that in W by orders of magnitude after the irradiation at high temperatures (≥1073K, while no significant effects of Re addition was observed after irradiation at 523K. Irradiation with 20MeV W ions at room temperature followed by exposure to D plasma at 443–743K also resulted in small difference in D retention between W and W–5% Re samples. The results of positron lifetime measurements showed that the reduced D retention by Re observed after high temperature irradiation was due to suppression of formation of vacancy-type defects (monovacancies and vacancy clusters by Re.

  8. Influence of the austenite-martensite transformation in the dimensional stability of a new tool steel alloyed with niobium (0.08% wt.) and vanadium (0.12% wt.); Influencia de la transformacion austenita-martensita en la estabilidad dimensional de un nuevo acero para herramientas aleado con niobio (0,08%) y vanadio (0,12%)

    Energy Technology Data Exchange (ETDEWEB)

    Conejero Ortega, G.; Candela Vazquez, N.; Pichel Martinez, M.; Barea del Cerro, R.; Carsi Cebrian, M.

    2014-07-01

    Austenite-martensite transformation influence on the dimensional stability of a new experimental tool steel alloyed with niobium (0.08% wt.) and vanadium (0.12% wt.) has been studied. The dimensional stability of this new steel was compared with the dimensional stability of commercial steel, after and before two thermal treatments, T1 (860 degree centigrade) and T2 (900 degree centigrade). The thermal treatments consisted on heating and cooling, at 1 atmosphere of pressure, in N{sub 2} atmosphere furnace, following by heating in a conventional furnace at 180 degree centigrade during 1 hour. Initially, the experimental steel composition and Ac{sub 1} and Ac{sub 3} transformation temperatures were determined by glow-discharge luminescence (GDL) and dilatometric tests, respectively, in order to select the austenization temperatures of T1 and T2 treatments. After hardness measurement, the microstructure of both steels was characterized by X-Ray Diffraction (XRD) and optical metallography, before and after of T1 and T2 thermal treatments. Finally, longitudinal and angular dimensional stability analyses were realized for both commercial and experimental steels. After a contrastive hypothesis analysis, the results showed that the longitudinal relative variation of the experimental steel calculated was around 0.2% and the angular relative variation was not significant. (Author)

  9. Unraveling the Effect of Thermomechanical Treatment on the Dissolution of Delta Ferrite in Austenitic Stainless Steels

    Science.gov (United States)

    Rezayat, Mohammad; Mirzadeh, Hamed; Namdar, Masih; Parsa, Mohammad Habibi

    2016-02-01

    Considering the detrimental effects of delta ferrite stringers in austenitic stainless steels and the industrial considerations regarding energy consumption, investigating, and optimizing the kinetics of delta ferrite removal is of vital importance. In the current study, a model alloy prone to the formation of austenite/delta ferrite dual phase microstructure was subjected to thermomechanical treatment using the wedge rolling test aiming to dissolve delta ferrite. The effect of introducing lattice defects and occurrence of dynamic recrystallization (DRX) were investigated. It was revealed that pipe diffusion is responsible for delta ferrite removal during thermomechanical process, whereas when the DRX is dominant, the kinetics of delta ferrite dissolution tends toward that of the static homogenization treatment for delta ferrite removal that is based on the lattice diffusion of Cr and Ni in austenite. It was concluded that the optimum condition for dissolution of delta ferrite can be defined by the highest rolling temperature and strain in which DRX is not pronounced.

  10. Review of environmental effects on fatigue crack growth of austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Shack, W.J.; Kassner, T.F. [Argonne National Lab., IL (United States)

    1994-05-01

    Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry.

  11. Interatomic potential to study plasticity in stainless steels: the FeNiCr model alloy

    Science.gov (United States)

    Bonny, G.; Terentyev, D.; Pasianot, R. C.; Poncé, S.; Bakaev, A.

    2011-12-01

    Austenitic stainless steels are commonly used materials for in-core components of nuclear light water reactors. In service, such components are exposed to harsh conditions: intense neutron irradiation, mechanical and thermal stresses, and aggressive corrosion environment which all contribute to the components' degradation. For a better understanding of the prevailing mechanisms responsible for the materials degradation, large-scale atomistic simulations are desirable. In this framework we developed an embedded atom method type interatomic potential for the ternary FeNiCr system to model movement of dislocations and their interaction with radiation defects. Special attention has been drawn to the Fe-10Ni-20Cr alloy, whose properties were ensured to be close to those of 316L austenitic stainless steel. In particular, the stacking fault energy and elastic constants are well reproduced. The fcc phase for the Fe-10Ni-20Cr random alloy was proven to be stable in the temperature range 0-900 K and under shear strain up to 5%. For the same alloy the stable glide of screw dislocations and stability of Frank loops was confirmed.

  12. Low-Temperature Nitriding of Deformed Austenitic Stainless Steels with Various Nitrogen Contents Obtained by Prior High-Temperature Solution Nitriding

    DEFF Research Database (Denmark)

    Bottoli, Federico; Winther, Grethe; Christiansen, Thomas Lundin;

    2016-01-01

    investigated. Both hardness and yield stress increase and the alloys remain ductile. In addition, strain-induced transformation of austenite to martensite is suppressed, which is beneficial for subsequent low-temperature nitriding of the surface of deformed alloys. The combination of high- and low...

  13. Irradiation damage

    Energy Technology Data Exchange (ETDEWEB)

    Howe, L.M

    2000-07-01

    There is considerable interest in irradiation effects in intermetallic compounds from both the applied and fundamental aspects. Initially, this interest was associated mainly with nuclear reactor programs but it now extends to the fields of ion-beam modification of metals, behaviour of amorphous materials, ion-beam processing of electronic materials, and ion-beam simulations of various kinds. The field of irradiation damage in intermetallic compounds is rapidly expanding, and no attempt will be made in this chapter to cover all of the various aspects. Instead, attention will be focused on some specific areas and, hopefully, through these, some insight will be given into the physical processes involved, the present state of our knowledge, and the challenge of obtaining more comprehensive understanding in the future. The specific areas that will be covered are: point defects in intermetallic compounds; irradiation-enhanced ordering and irradiation-induced disordering of ordered alloys; irradiation-induced amorphization.

  14. Effects of laser irradiation on aluminum alloy tank containing water%1053nm脉冲激光对铝合金/水结构辐照效应

    Institute of Scientific and Technical Information of China (English)

    焦路光; 赵国民; 袁立国

    2011-01-01

    Experimental measurement and numerical simulation are employed to study the effects of laser irradiation on aluminum alloy tank containing water. Firstly, by analyzing the surface pictures and measuring the temperature histories, the effects of water on the ablation of aluminum alloy plate irradiated by I 053 nm pulsed laser are investigated. The results show that heat conduction and convection play a very important role in the ablation of the aluminum alloy plate. When the tank is empty, the aluminum alloy plate is melt through after eight laser pulses irradiation. And when the tank is filled with water, the aluminum alloy plate has not been melt through after ten laser pulses irradiation. Then, a numerical model is presented to calculate the temperature distribution of the aluminum alloy tank, and the dimension of the region that has melted is obtained. The difference between the numerical result and the experimental result is not apparent when the tank is empty, but when the tank is filled with water the difference is a little large. The reason is that the natural convection and boiling of water are not considered in the numerical model.%使用1053 nm脉冲激光分别辐照铝合金单板和铝合金/水结构,通过表面形貌观察、温度场分析、熔穿时间测量等手段,分析了水的存在对铝合金壳体烧蚀的影响.运用有限元软件ANSYS,建立了脉冲激光辐照下单板及结构温度变化的数值模型,计算了铝合金表面熔凝区域的尺寸,并与实验结果进行了对比.结果表明:在相同的实验条件下,辐照8个激光脉冲时,铝合金单板即被熔穿,而辐照10个脉冲后铝合金/水结构仍未发生熔穿,且结构中铝合金表面的熔凝区域要小于单板情形中的熔凝区域,这表明水的存在对延缓铝合金板的烧蚀有较大的作用.对于单板情形,计算结果与实验结果符合较好,而对于铝合金/水结构情形,数值模拟放大了铝合金壳体的温升,这主

  15. Examination of Spheroidal Graphite Growth and Austenite Solidification in Ductile Iron

    Science.gov (United States)

    Qing, Jingjing; Richards, Von L.; Van Aken, David C.

    2016-09-01

    Microstructures of a ductile iron alloy at different solidification stages were captured in quenching experiments. Etched microstructures showed that spheroidal graphite particles and austenite dendrites nucleated independently to a significant extent. Growth of the austenite dendrite engulfed the spheroidal graphite particles after first contacting the nodule and then by forming an austenite shell around the spheroidal graphite particle. Statistical analysis of the graphite size distribution was used to determine the nodule diameter when the austenite shell was completed. In addition, multiple graphite nucleation events were discerned from the graphite particle distributions. Majority of graphite growth occurred when the graphite was in contact with the austenite. Circumferential growth of curved graphene layers appeared as faceted growth fronts sweeping around the entire surface of a spheroidal graphite particle which was at the early growth stage. Mismatches between competing graphene growth fronts created gaps, which divided the spheroidal graphite particle into radially oriented conical substructures. Graphene layers continued growing in each conical substructure to further extend the size of the spheroidal graphite particle.

  16. Investigation of austenitizing temperature on wear behavior of austempered gray iron (AGI)

    Science.gov (United States)

    Sarkar, T.; Sutradhara, G.

    2016-09-01

    This study is about finding the effect of austenitizing temperature on microstructure and wear behavior of copper alloyed austempered gray iron (AGI), and then comparing it with an as- cast (solidified) state. Tensile and wear tests specimens are prepared from as-cast gray iron material, and austenitized at different temperatures and then austempered at a fixed austempering temperature. Resulting microstructures are characterized through optical microscopy, scanning electron microscope (SEM) and X-Ray diffraction. Wear test is carried out using a block-on-roller multi-tribotester with sliding speed of 1.86 m/sec. In this investigation, wear behavior of all these austempered materials are determined and co-related with the micro structure. Hence the wear surface under scanning electron microscope showed that wear occurred mainly due to adhesion and delamination under dry sliding condition. The test results indicate that the austenitizing temperature has remarkable effect on resultant micro structure and wear behavior of austempered materials. Wear behavior is also found to be dependent on the hardness, tensile strength, austenite content and carbon content in austenite. It is shown that coarse ausferrite micro structure exhibited higher wear depth than fine ausferrite microstructure.

  17. A comparative study of ternary Al-Sn-Cu immiscible alloys prepared by conventional casting and casting under high-intensity ultrasonic irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Kotadia, H.R., E-mail: hiren.kotadia@kcl.ac.uk [Department of Physics, King' s College London, Strand, London WC2R 2LS (United Kingdom); Das, A. [Materials Research Centre, College of Engineering, Swansea University, Singleton Park, Swansea, SA2 8PP (United Kingdom); Doernberg, E.; Schmid-Fetzer, R. [Clausthal University of Technology, Institute of Metallurgy, Robert-Koch-Str. 42, D-38678 Clausthal-Zellerfeld (Germany)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer Systematic investigation on the solidification microstructure of ternary Al-Sn-Cu immiscible system aided by computational thermodynamics calculations. Black-Right-Pointing-Pointer Comparative study of conventional casting and casting under high-intensity ultrasonic irradiation. Black-Right-Pointing-Pointer Demonstrated the high effectiveness of ultrasound exposure during solidification. Black-Right-Pointing-Pointer Effect of cavitation on nucleation and the relative effects of cavitation and acoustic streaming on the dispersion of Sn-rich liquid phases have been explained from the experimental observation. Cavitation was found to promote fragmentation and dispersion of Sn-rich liquid leading to homogeneous dispersion of refined Sn phase. Microstructural modification was found to be contributed by cavitation and associated shockwave generation while bulk fluid flow under acoustic streaming was found to be less effective on the microstructure evolution. Black-Right-Pointing-Pointer Globular and highly refined {alpha}-Al formed near the radiator through enhanced heterogeneous nucleation in contrast to dendritic {alpha}-Al observed in conventional solidification. - Abstract: A comparative study on the microstructure of four ternary Al-Sn-Cu immiscible alloys, guided by the recent thermodynamic assessment of the system, was carried out with specific focus on the soft Sn particulate distribution in hard Al-rich matrix in the presence and absence of ultrasonic irradiation during solidification. The results clearly demonstrate high effectiveness of ultrasonication in promoting significantly refined and homogeneously dispersed microstructure, probably aided by enhanced nucleation and droplet fragmentation under cavitation. While conventional solidification produced highly segregated Sn phase at the centre and bottom of Sn-rich alloy ingots, ultrasonic treatment produced effective dispersion irrespective of the alloy constitution in

  18. 奥氏体不锈钢和镍基合金在550℃/25MPa超临界水中的应力腐蚀开裂敏感性%Stress Corrosion Cracking Susceptibility of Austenitic Stainless Steels and Nickel-based Alloy in Supercritical Water at 550℃/25 MPa

    Institute of Scientific and Technical Information of China (English)

    李力; 张乐福; 唐睿

    2012-01-01

    The stress corrosion cracking(SCC) susceptibility of austenitic stainless steels 316Ti, HR3C, TP347 and nickel-based alloy 718 in supercritical water(SCW)at 550℃/25 MPa was studied. Slow strain rate tests (SSRT) were used to obtain tile stress-strain curves. The results show that the yield strength and tensile strength of 718 were much higher than those of the other three austenitic stainless steels while the elongation of 718 was significantly lower. Scanning electron microscopy observations of fracture surfaces reveal, that the failure mode of 316Ti and TP347 was transgranular ductile fracture, the failure mode of HR3C was both intergranular and transgranular ductile fracture and the failure mode of 718 was almost intergranular stress corrosion cracking (IGSCC).%研究了奥氏体不锈钢316Ti、HR3C、TP347和镍基合金718在550℃/25MPa超临界水中的应力腐蚀开裂(SCC)敏感性。通过慢应变速率拉伸试验得到相应的应力-应变曲线。结果表明,在本次试验工况下三种奥氏体不锈钢的屈服强度、抗拉强度和延伸率都非常接近,但镍基合金718的强度远高出其他材料,同时延伸率也大幅降低。扫描电镜对试样侧面以及断口形貌的观察分析发现:316Ti和TP347的失效模式均为穿晶韧性断裂;HR3C则表现为沿晶和穿晶的混合型韧性断裂;718的失效模式则几乎全是沿晶的脆性断裂。

  19. Effect of bainite transformation and retained austenite on mechanical properties of austempered spheroidal graphite cast steel

    Science.gov (United States)

    Takahashi, Toshio; Abe, Toshihiko; Tada, Shuji

    1996-06-01

    Austempered ductile iron (ADI) has excellent mechanical properties, but its Young's modulus is low. Austempered spheroidal graphite cast steel (AGS) has been developed in order to obtain a new material with superior mechanical properties to ADI. Its carbon content (approximately 1.0 pct) is almost one-third that of a standard ADI; thus, the volume of graphite is also less. Young's modulus of AGS is 195 to 200 GPa and is comparable to that of steel. Austempered spheroidal graphite cast steel has an approximately 200 MPa higher tensile strength than ADI and twice the Charpy absorbed energy of ADI. The impact properties and the elongation are enhanced with increasing volume fraction of carbon-enriched retained austenite. At the austempering temperature of 650 K, the volume fraction of austenite is approximately 40 pct for 120 minutes in the 2.4 pct Si alloy, although it decreases rapidly in the 1.4 pct Si alloy. The X-ray diffraction analysis shows that appropriate quantity of silicon retards the decomposition of the carbon-enriched retained austenite. For austempering at 570 K, the amount of the carbon-enriched austenite decreases and the ferrite is supersaturated with carbon, resulting in high tensile strength but low toughness.

  20. Effect of bainite transformation and retained austenite on mechanical properties of austempered spheroidal graphite cast steel

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Toshio; Abe, Toshihiko; Tada, Shuji [Tohoku National Industrial Research Inst., Sendai (Japan). Materials Engineering Div.

    1996-06-01

    Austempered ductile iron (ADI) has excellent mechanical properties, but its Young`s modulus is low. Austempered spheroidal graphite cast steel (AGS) has been developed in order to obtain a new material with superior mechanical properties to ADI. Its carbon content (approximately 1.0 pct) is almost one-third that of a standard ADI; thus, the volume of graphite is also less. Young`s modulus of AGS is 195 to 200 GPa and is comparable to that of steel. Austempered spheroidal graphite cast steel has an approximately 200 MPa higher tensile strength than ADI and twice the Charpy absorbed energy of ADI. The impact properties and the elongation are enhanced with increasing volume fraction of carbon-enriched retained austenite. At the austempering temperature of 650 K, the volume fraction of austenite is approximately 40 pct for 120 minutes in the 2.4 pct Si alloy, although it decreases rapidly in the 1.4 pct Si alloy. The X-ray diffraction analysis shows that appropriate quantity of silicon retards the decomposition of the carbon-enriched retained austenite. For austempering at 570 K, the amount of the carbon-enriched austenite decreases and the ferrite is supersaturated with carbon, resulting in high tensile strength but low toughness.

  1. Expanded austenite; crystallography and residual stress

    DEFF Research Database (Denmark)

    Christiansen, Thomas; Hummelshøj, Thomas Strabo; Somers, Marcel A. J.

    2009-01-01

    The identity of expanded austenite as developing during low temperature nitriding and/or carburizing of austenitic stainless steel has been under debate since the very first observation of this phase. In the present article recent results obtained with i) homogeneous samples of various uniform...

  2. Austenite formation during intercritical annealing

    OpenAIRE

    A. Lis; J. Lis

    2008-01-01

    Purpose: of this paper is the effect of the soft annealing of initial microstructure of the 6Mn16 steel on the kinetics of the austenite formation during next intercritical annealing.Design/methodology/approach: Analytical TEM point analysis with EDAX system attached to Philips CM20 was used to evaluate the concentration of Mn, Ni and Cr in the microstructure constituents of the multiphase steel and mainly Bainite- Martensite islands.Findings: The increase in soft annealing time from 1-60 hou...

  3. Effects of neutron irradiation on microstructure and deformation behaviour of mono- and polycrystalline molybdenum and its alloys

    DEFF Research Database (Denmark)

    Singh, B.N.; Evans, J.H.; Horsewell, A.;

    1998-01-01

    at similar to 320 K to displacement doses in the range 5.4 x 10(-4) to 1.6 x 10(-1) dpa (NRT) in the DR-3 reactor at Riso National Laboratory. For comparison, polycrystalline specimens of Mo-5% Re and TZM were also irradiated together with the monocrystalline specimens. Both unirradiated and irradiated...

  4. Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600{degrees}C. Each of the four heats exhibited a different stress-free swelling behavior at 400{degrees}C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower ({approx}50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400{degrees}C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600{degrees}C.

  5. Method of making high strength, tough alloy steel

    Science.gov (United States)

    Thomas, Gareth; Rao, Bangaru V. N.

    1979-01-01

    A high strength, tough alloy steel, particularly suitable for the mining industry, is formed by heating the steel to a temperature in the austenite range (1000.degree.-1100.degree. C.) to form a homogeneous austenite phase and then cooling the steel to form a microstructure of uniformly dispersed dislocated martensite separated by continuous thin boundary films of stabilized retained austenite. The steel includes 0.2-0.35 weight % carbon, at least 1% and preferably 3-4.5% chromium, and at least one other subsitutional alloying element, preferably manganese or nickel. The austenite film is stable to subsequent heat treatment as by tempering (below 300.degree. C.) and reforms to a stable film after austenite grain refinement.

  6. Influence of retained austenite on the grain size of austenite after reaustenitization of steels for heavy forgings

    Energy Technology Data Exchange (ETDEWEB)

    Peters, H.J.; Tacke, G.; Hougardy, H.P.

    1989-01-01

    In this investigation the grain size of austenite reaustenitization of different microstructures containing different volume fractions of retained austenite was determined. The austenite grain size after austenitization of martensite and lower bainite was coarse for heating rates lower than a minimum value, which is dependent on the chemical composition. In this case, the austenite forms by rapid growth of retained austenite in the initial microstructure. At heating rates higher than the critical value, formation of austenite starts at the ferrite-carbide phase boundaries giving a fine austenite grain. The formation of austenite from microstructures free of retained austenite, such as pearlite, always occurred by nucleation on the ferrite-carbide interphase resulting in fine austenite grains. (orig.).

  7. Mechanical Properties of Austenitic Stainless Steel Made by Additive Manufacturing.

    Science.gov (United States)

    Luecke, William E; Slotwinski, John A

    2014-01-01

    Using uniaxial tensile and hardness testing, we evaluated the variability and anisotropy of the mechanical properties of an austenitic stainless steel, UNS S17400, manufactured by an additive process, selective laser melting. Like wrought materials, the mechanical properties depend on the orientation introduced by the processing. The recommended stress-relief heat treatment increases the tensile strength, reduces the yield strength, and decreases the extent of the discontinuous yielding. The mechanical properties, assessed by hardness, are very uniform across the build plate, but the stress-relief heat treatment introduced a small non-uniformity that had no correlation to position on the build plate. Analysis of the mechanical property behavior resulted in four conclusions. (1) The within-build and build-to-build tensile properties of the UNS S17400 stainless steel are less repeatable than mature engineering structural alloys, but similar to other structural alloys made by additive manufacturing. (2) The anisotropy of the mechanical properties of the UNS S17400 material of this study is larger than that of mature structural alloys, but is similar to other structural alloys made by additive manufacturing. (3) The tensile mechanical properties of the UNS S17400 material fabricated by selective laser melting are very different from those of wrought, heat-treated 17-4PH stainless steel. (4) The large discontinuous yielding strain in all tests resulted from the formation and propagation of Lüders bands.

  8. Report on Status of Shipment of High Fluence Austenitic Steel Samples for Characterization and Stress Corrosion Crack Testing

    Energy Technology Data Exchange (ETDEWEB)

    Clark, Scarlett R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Leonard, Keith J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructural and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The

  9. Phase transformations and microstructure development in low alloy steel welds

    Energy Technology Data Exchange (ETDEWEB)

    Babu, S.S.; David, S.A.; Vitek, J.M. [and others

    1995-07-01

    Microstructure development in low alloy steel welds depends on various phase transformations that are a function of weld heating and cooling. The phase changes include non-metallic oxide inclusion formation in the liquid state, weld pool solidification, and solid state transformations. In this paper the mechanism of inclusion formation during low alloy steel welding is considered and the model predictions are compared with published results. The effect of inclusions on the austenite to ferrite transformation kinetics is measured and the mechanisms of transformation are discussed. The austenite gain development is related to the driving force for transformation of {delta} ferrite to austenite.

  10. Corrosion properties of S-phase layers formed on medical grade austenitic stainless steel.

    Science.gov (United States)

    Buhagiar, Joseph; Dong, Hanshan

    2012-02-01

    The corrosion properties of S-phase surface layers formed in AISI 316LVM (ASTM F138) and High-N (ASTM F1586) medical grade austenitic stainless steels by plasma surface alloying with nitrogen (at 430°C), carbon (at 500°C) and both carbon and nitrogen (at 430°C) has been investigated. The corrosion behaviour of the S-phase layers in Ringer's solutions was evaluated using potentiodynamic and immersion corrosion tests. The corrosion damage was evaluated using microscopy, hardness testing, inductive coupled plasma mass spectroscopy and X-ray diffraction. The experimental results have demonstrated that low-temperature nitriding, carburising and carbonitriding can improve the localised corrosion resistance of both industrial and medical grade austenitic stainless steels as long as the threshold sensitisation temperature is not reached. Carburising at 500°C has proved to be the best hardening treatment with the least effect on the corrosion resistance of the parent alloy.

  11. Intermetallic Strengthened Alumina-Forming Austenitic Steels for Energy Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Bin [Dartmouth College, Hanover, NH (United States); Baker, Ian [Dartmouth College, Hanover, NH (United States)

    2016-03-31

    In order to achieve energy conversion efficiencies of >50 % for steam turbines/boilers in power generation systems, the materials required must be strong, corrosion-resistant at high temperatures (>700°C), and economically viable. Austenitic steels strengthened with Laves phase and L12 precipitates, and alloyed with aluminum to improve oxidation resistance, are potential candidate materials for these applications. The creep resistance of these alloys is significantly improved through intermetallic strengthening (Laves-Fe2Nb + L12-Ni3Al precipitates) without harmful effects on oxidation resistance. Microstructural and microchemical analyses of the recently developed alumina-forming austenitic (AFA) steels (Fe-14Cr-32Ni-3Nb-3Al-2Ti-based) indicated they are strengthened by Ni3Al(Ti) L12, NiAl B2, Fe2Nb Laves phase and MC carbide precipitates. Different thermomechanical treatments (TMTs) were performed on these stainless steels in an attempt to further improve their mechanical properties. The thermo-mechanical processing produced nanocrystalline grains in AFA alloys and dramatically increased their yield strength at room temperature. Unfortunately, the TMTs didn’t increase the yield strengths of AFA alloys at ≥700ºC. At these temperatures, dislocation climb is the dominant mechanism for deformation of TMT alloys according to strain rate jump tests. After the characterization of aged AFA alloys, we found that the largest strengthening effect from L12 precipitates can be obtained by aging for less than 24 h. The coarsening behavior of the L12 precipitates was not influenced by carbon and boron additions. Failure analysis and post-mortem TEM analysis were performed to study the creep failure mechanisms of these AFA steels after creep tests. Though the Laves and B2-NiAl phase precipitated along the boundaries can improve the creep properties, cracks were

  12. Effect of austenite grain size in Fe-Mn alloys on {epsilon} martensitic transformation and their mechanical properties; Fe-Mn gokin no {epsilon} marutensaito hentai oyobi kikaiteki seishitsu ni oyobosu kessho ryukei no eikyo

    Energy Technology Data Exchange (ETDEWEB)

    Nakatsu, H. [Kyushu Univ., Fukuoka (Japan). Graduate School; Takaki, S. [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1996-02-20

    The Fe-Mn two-components alloy samples varied with Mn content of 12 to 31 mass% were prepared by charging {gamma}-crystalline grain size using its recrystallization, and were surveyed on effects of the {gamma}-crystalline grain size on athermal {epsilon}-martensitic ({epsilon})-transformation and machining- induced {epsilon}-transformation. As a result of examining the relationship between the {gamma}-crystalline grain size or the {epsilon}-transformation and their mechanical properties, conclusion shown as follows is obtained. The athermal {epsilon} was formed at the alloy containing more than 10 mass% of Mn, maximum {epsilon} was shown at the composition containing about 17 mass% of Mn and the {epsilon} was almost not formed at the steel containing more than 27 mass% of Mn. When crushing the {gamma}-crystalline grain to fine powder, the {epsilon} martensitic transformation beginning temperature tended to reduce somewhat and production amount of the {epsilon} decreased extremely. On the steel containing Mn ranged 15 to 31 mass%, the fine powdering affected scarcely its durability but improved its elongation and its tensile strength. 26 refs., 11 figs., 1 tab.

  13. Investigations on avoidance of hot cracks during laser welding of austenitic Cr-Ni steels and nickel-based alloys using temperature field tailoring. Final report; Untersuchungen zur Vermeidung von Heissrissen beim Laserstrahlschweissen von austenitischen Cr-Ni-Staehlen und Nickelbasislegierungen mittels Temperaturfeld-Tailoring. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-08

    The aim of the project was to transfer the developed method of laser beam welding of heat treated machining steels of temperature field tailoring on hot crack endangered austenitic Cr-Ni steels and nickel-based alloys. With this method, transient thermal stresses adjacent to the weld are produced by an travelling induction heating so that the hot cracking is prevented during welding. As test materials the austenitic Cr-Ni steel with sulfur additive 1.4305, the Cr-Ni steels 1.4404 and 1.4435 and the nickel-based alloy Udimet 720 were selected. As a result of the research it was shown that a hot crack-free laser welding in the investigated materials using at least three different welding and material-technical approaches is possible. [German] Das Ziel des Forschungsvorhabens bestand darin, das fuer das Laserstrahlschweissen verguetbarer Automatenstaehle entwickelte Verfahren des Temperaturfeld-Tailorings auf heissrissgefaehrdete austenitische Cr-Ni-Staehle und Nickelbasislegierungen zu uebertragen. Mit diesem Verfahren werden waehrend des Schweissens transiente thermische Spannungen neben der Schweissnaht durch eine mitlaufende induktive Erwaermung so erzeugt, dass die Heissrissbildung verhindert wird. Als Versuchswerkstoffe wurden der austenitische Cr-Ni-Stahl mit Schwefelzusatz 1.4305, die Cr-Ni-Staehle 1.4404 und 1.4435 sowie die Nickelbasislegierung Udimet 720 ausgewaehlt. Im Ergebnis des Forschungsvorhabens konnte gezeigt werden, dass ein heissrissfreies Laserstrahlschweissen bei den untersuchten Werkstoffen unter Nutzung von mindestens drei verschiedenen schweiss- und werkstofftechnischen Ansaetzen moeglich ist: Erstens koennen mit einem Temperaturfeld-Tailoring bei im Stumpfstoss zu verschweissenden Blechen aus austenitischen Staehlen bis mindestens 6 mm Dicke senkrecht zur Naht und parallel zur Blechoberflaeche wirkende transiente Druckspannungen erzeugt werden, die der Bildung von Mittelrippenrissen oder dazu parallel liegenden Heissrissen entgegenwirken

  14. Relation of martensite-retained austenite and its effect on microstructure and mechanical properties of the quenched and partitioned steels

    Institute of Scientific and Technical Information of China (English)

    WANG CunYu; CHANG Ying; LI XiaoDong; ZHAO KunMin; DONG Han

    2016-01-01

    A two-step quenching and partitioning (Q&P) treatment was applied to low-carbon alloy steels.The relation of initial martensite-retained austenite-fresh martensite and its effect on microstructure and mechanical properties were investigated by experiments.The results reveal that the volume fraction of retained austenite can reach the peak value of 17%,and the corresponding volume fractions of initial martensite and fresh martensite are 40% and 43%,respectively,when the tested steel is treated by initial quenching at 330℃,partitioning at 500℃ for 60s and final quenching to room temperature.Moreover,the micromorphologies of austenite and martensite become finer with the increasing of initial martensite fraction.The elongation is the highest when the volume fractions of initial martensite and retained austenite are 70% and 11%,respectively,meanwhile,the yield strength increases and tensile strength decreases gradually with the increase of initial martensite fraction,which proves that the mechanical properties including elongation,yield strength and tensile strength are based on the comprehensive effect of the retained austenite fraction,the finer microstructure and austenite stability.

  15. Low-Temperature Nitriding of Deformed Austenitic Stainless Steels with Various Nitrogen Contents Obtained by Prior High-Temperature Solution Nitriding

    Science.gov (United States)

    Bottoli, Federico; Winther, Grethe; Christiansen, Thomas L.; Dahl, Kristian Vinter; Somers, Marcel A. J.

    2016-08-01

    In the past decades, high nitrogen steels (HNS) have been regarded as substitutes for conventional austenitic stainless steels because of their superior mechanical and corrosion properties. However, the main limitation to their wider application is their expensive production process. As an alternative, high-temperature solution nitriding has been applied to produce HNS from three commercially available stainless steel grades (AISI 304L, AISI 316, and EN 1.4369). The nitrogen content in each steel alloy is varied and its influence on the mechanical properties and the stability of the austenite investigated. Both hardness and yield stress increase and the alloys remain ductile. In addition, strain-induced transformation of austenite to martensite is suppressed, which is beneficial for subsequent low-temperature nitriding of the surface of deformed alloys. The combination of high- and low-temperature nitriding results in improved properties of both bulk and surface.

  16. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels; Etude a la sonde atomique de l`evolution microstructurale sous irradiation d`alliages ferritiques Fe-Cu et d`aciers de cuve REP

    Energy Technology Data Exchange (ETDEWEB)

    Pareige, P.

    1996-04-01

    Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends.

  17. Tensile properties of vanadium-base alloys irradiated in the Fusion-1 low-temperature experiment in the BOR-60 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gazda, J.; Nowicki, L.J.; Billone, M.C.; Smith, D.L. [Argonne National Lab., IL (United States)

    1998-09-01

    The irradiation has been completed and the test specimens have been retrieved from the lithium-bonded capsule at the Research Institute of Atomic Reactors (RIAR) in Russia. During this reporting period, the Argonne National Laboratory (ANL) tensile specimens were received from RIAR and initial testing and examination of these specimens at ANL has been completed. The results, corroborating previous findings showed a significant loss of work hardening capability in the materials. There appears to be no significant difference in behavior among the various heats of vanadium-base alloys in the V-(4-5)Cr-(4-5)Ti composition range. The variations in the preirradiation annealing conditions also produced no notable differences.

  18. Simulation of nanostructural evolution under irradiation in Fe-9%CrC alloys: An object kinetic Monte Carlo study of the effect of temperature and dose-rate

    Directory of Open Access Journals (Sweden)

    M. Chiapetto

    2016-12-01

    Full Text Available This work explores the effects of both temperature and dose-rate on the nanostructural evolution under irradiation of the Fe-9%CrC alloy, model material for high-Cr ferritic/martensitic steels. Starting from an object kinetic Monte Carlo model validated at 563K, we investigate here the accumulation of radiation damage as a function of temperature and dose-rate, attempting to highlight its connection with low-temperature radiation-induced hardening. The results show that the defect cluster mobility becomes high enough to partially counteract the material hardening process only above ∼290°C, while high fluxes are responsible for higher densities of defects, so that an increase of the hardening process with increasing dose-rates may be expected.

  19. Duct and cladding alloy

    Science.gov (United States)

    Korenko, Michael K.

    1983-01-01

    An austenitic alloy having good thermal stability and resistance to sodium corrosion at 700.degree. C. consists essentially of 35-45% nickel 7.5-14% chromium 0.8-3.2% molybdenum 0.3-1.0% silicon 0.2-1.0% manganese 0-0.1% zirconium 2.0-3.5% titanium 1.0-2.0% aluminum 0.02-0.1% carbon 0-0.01% boron and the balance iron.

  20. Impact Properties of Copper-Alloyed and Nickel-Copper Alloyed ADI

    Science.gov (United States)

    Batra, Uma; Ray, Subrata; Prabhakar, S. R.

    2007-08-01

    The influence of austenitization and austempering parameters on the impact properties of copper-alloyed and nickel-copper-alloyed austempered ductile irons (ADIs) has been studied. The austenitization temperature of 850 and 900 °C have been used in the present study for which austempering time periods of 120 and 60 min were optimized in an earlier work. The austempering process was carried out for 60 min for three austempering temperatures of 270, 330, and 380 °C to study the effect of austempering temperature. The influence of the austempering time on impact properties has been studied for austempering temperature of 330 °C for time periods of 30-150 min. The variation in impact strength with the austenitization and austempering parameters has been correlated to the morphology, size and amount of austenite and bainitic ferrite in the austempered structure. The fracture surface of ADI failed under impact has been studied using SEM.

  1. Mathematical Model of the Processoof Pearlite Austenitization

    Directory of Open Access Journals (Sweden)

    Olejarczyk-Wożeńska I.

    2014-10-01

    Full Text Available The paper presents a mathematical model of the pearlite - austenite transformation. The description of this process uses the diffusion mechanism which takes place between the plates of ferrite and cementite (pearlite as well as austenite. The process of austenite growth was described by means of a system of differential equations solved with the use of the finite difference method. The developed model was implemented in the environment of Delphi 4. The proprietary program allows for the calculation of the rate and time of the transformation at an assumed temperature as well as to determine the TTT diagram for the assigned temperature range.

  2. Study on austenitic nitrocarburizing without compound layer

    Energy Technology Data Exchange (ETDEWEB)

    Yang, X. [Univ. of Petroleum, Dongying, Shandong (China); Kong, C.; Qiao, Y. [Shandong Polytechnic Univ., Jinan, Shandong (China)

    1995-12-31

    This paper presents an advanced austenitic nitrocarburizing process. Medium-carbon steel was used in austenitic nitrocarburizing with methanol/ammonia atmospheres. A particular hardened case without compound layer was obtained at 680 C processing temperature and a moderate nitrogen potential level and for steel 45 nitrocarburized, there is a fine-grain region beneath the austenite case. The forming and developing mechanism of the fine-grain region was analyzed and the microhardness profiles of the layer before and after ageing were determined. Having the advantages of shorter processing time and a superior hardened case, this treatment is expected to supersede the conventional ferritic nitrocarburizing process in many wear resistance applications.

  3. Mössbauer investigation of austenite formation together with Cr depletion in aged turbine blade steels

    Science.gov (United States)

    Kuzmann, E.; Jaen, J.; Vértes, A.; Csöme, L.; Tibiássy, B.; Káldor, M.

    1990-07-01

    Mössbauer spectroscopy and hardness measurements were used to study annealing effect on turbine blade steels. Hyperfine field distribution method was applied to follow the changes in the concentration of alloying elements being in the martensite after various heat treatments. Our results imply that upon annealing at a given temperature (400-640°C), formation of austenite takes place (similarly as found earlier [1] in some cases), simultaneously with a significant depletion (up to 4%) of Cr (and other alloying elements) in the martensite.

  4. Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, Frank [Case Western Reserve Univ., Cleveland, OH (United States)

    2016-10-13

    We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance.

  5. Efeito da temperatura e tempo de austenitização nas transformações de fase da liga 13Cr2Ni0,1C Effect of temperature and time austenitizing in phase transformations alloy 13Cr2Ni0,1C

    Directory of Open Access Journals (Sweden)

    Neide Aparecida Mariano

    2010-03-01

    Full Text Available Novas classes de aços inoxidáveis martensíticos, com baixo teor de carbono, têm sido desenvolvidas, para atender, principalmente, às necessidades do segmento da indústria do petróleo. Contudo o seu uso tem sido restrito pelo fato de ser um desenvolvimento recente e muitas de suas propriedades ainda são motivos de investigação. Nesse trabalho, foram determinados os valores das temperaturas inicial e final da transformação austenítica e as temperaturas de início e fim da formação martensítica, para a liga 13Cr2Ni0,1C, através de ensaios dilatométricos com resfriamento contínuo. Com base nesses resultados, foram obtidas as condições otimizadas dos tratamentos térmicos de têmpera e revenido. A caracterização microestrutural das ligas na condição de bruta fusão foi realizada por microscopia ótica observando-se uma matriz martensítica com a presença de ferrita delta.New classes of martensitic stainless steels, with low carbon levels, have been developed aiming to meet the needs of the petroleum industry segment. However, their use has been restricted due to the fact it is a recent development and many of its properties are still under investigation. This work determines the values of initial and final temperatures for the austenitic transformation and the initial and final temperatures of martensitic formation for alloy 13Cr2Ni0,1C, by means of dilatometric tests under continuous cooling. Based on these results the optimized conditions for quench and temper heat treatments were obtained. The microstructural characterization of the alloys under coarse fusion condition was carried out by optical microscopy and the presence of delta-ferrite in the martensitic matrix was observed.

  6. Formation of microcraters and hierarchically-organized surface structures in TiNi shape memory alloy irradiated with a low-energy, high-current electron beam

    Energy Technology Data Exchange (ETDEWEB)

    Meisner, L. L., E-mail: llm@ispms.tsc.ru; Meisner, S. N., E-mail: myp@ispms.tsc.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation); National Research Tomsk State University, Tomsk, 634050 (Russian Federation); Markov, A. B., E-mail: a.markov@hq.tsc.ru; Ozur, G. E., E-mail: vrotshtein@yahoo.com; Yakovlev, E. V., E-mail: msn@ispms.tsc.ru [Institute of High Current Electronics SB RAS, Tomsk, 634055 (Russian Federation); Rotshtein, V. P., E-mail: yakovev@lve.hcei.tsc.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation); Tomsk State Pedagogical University, Tomsk, 634050 (Russian Federation); Gudimova, E. Yu., E-mail: ozur@lve.hcei.tsc.ru [Institute of Strength Physics and Materials Science SB RAS, Tomsk, 634055 (Russian Federation)

    2015-10-27

    The regularities of surface cratering in TiNi alloy irradiated with a low-energy, high-current electron beam (LEHCEB) in dependence on energy density and number of pulses are studied. LEHCEB processing of TiNi samples was carried out using RITM-SP facility. Energy density E{sub s} was varied from 1 to 5 J/cm{sup 2}, pulse duration was 2.5–3.0 μs, the number of pulses n = 1–128. The dominant role of non-metallic inclusions [mainly, TiC(O)] in the nucleation of microcraters was found. It was revealed that at small number of pulses (n = 2), an increase in energy density leads both to increasing average diameter and density of microcraters. An increase in the number of pulses leads to a monotonic decrease in density of microcraters, and, therefore, that of the proportion of the area occupied by microcraters, as well as a decrease in the surface roughness. The multiple LEHCEB melting of TiNi alloy in crater-free modes enables to form quasi-periodical, hierarchically-organized microsized surface structures.

  7. Optimization of Melt Treatment for Austenitic Steel Grain Refinement

    Science.gov (United States)

    Lekakh, Simon N.; Ge, Jun; Richards, Von; O'Malley, Ron; TerBush, Jessica R.

    2017-02-01

    Refinement of the as-cast grain structure of austenitic steels requires the presence of active solid nuclei during solidification. These nuclei can be formed in situ in the liquid alloy by promoting reactions between transition metals (Ti, Zr, Nb, and Hf) and metalloid elements (C, S, O, and N) dissolved in the melt. Using thermodynamic simulations, experiments were designed to evaluate the effectiveness of a predicted sequence of reactions targeted to form precipitates that could act as active nuclei for grain refinement in austenitic steel castings. Melt additions performed to promote the sequential precipitation of titanium nitride (TiN) onto previously formed spinel (Al2MgO4) inclusions in the melt resulted in a significant refinement of the as-cast grain structure in heavy section Cr-Ni-Mo stainless steel castings. A refined as-cast structure consisting of an inner fine-equiaxed grain structure and outer columnar dendrite zone structure of limited length was achieved in experimental castings. The sequential of precipitation of TiN onto Al2MgO4 was confirmed using automated SEM/EDX and TEM analyses.

  8. Ion-irradiation effects on dissimilar friction stir welded joints between ODS alloy and ferritic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, C.-L., E-mail: chunliang@mail.ndhu.edu.tw [Department of Materials Science and Engineering, National Dong-Hwa University, Hualien 97401, Taiwan (China); Richter, A. [Department of Engineering, Technical University of Applied Sciences Wildau, Bahnhofstrasse 1, 15745 Wildau (Germany); Kögler, R. [Institute of Ion Beam Physics and Materials Research, Helmholtz Center Dresden-Rossendorf (HZDR), Bautzner Landstraße 400, 01328 Dresden (Germany); Griepentrog, M.; Reinstädt, P. [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 87, 12205 Berlin (Germany)

    2014-12-05

    Highlights: • FSSW has successfully been used in the welding of dissimilar materials. • The irradiation causes different degrees of hardening in the welding zones. • The formation of He bubbles at precipitates was found in the dissimilar joints. • The hardening effect is due to formation of He-filled vacancies. - Abstract: Friction stir spot welding (FSSW) is an advanced technique for the joining of materials to prevent agglomeration of fine oxide particles, grain coarsening, and stress corrosion cracking etc. In this study, the dissimilar FSSW joint of stainless steel 430/ODS was irradiated with a Fe{sup +}/He{sup +} dual ion beam. Irradiation damage can cause deterioration in the mechanical properties especially in the welding zones. The joint quality therefore plays a decisive role in the life expectancy of nuclear reactors. The effect of irradiation on different zones in the joint (the thermo-mechanically affected zone, the heat affected zone and the base material) was investigated by TEM and nanoindentation. Irradiation causes a hardness increase in all welding zones with a characteristic hardness maximum. The relative hardness increase and the related microstructure are discussed. The formation of He bubbles at chromium carbide precipitates and the homogeneous distribution of He filled vacancies in the mixture region of the 430/ODS FSSW joints was observed.

  9. Microstructure evolution and hardness change in ordered Ni{sub 3}V intermetallic alloy by energetic ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, A.; Kaneno, Y. [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Semboshi, S. [Kansai-Center, Institute for Materials Research, Tohoku University, Sakai, Osaka 599-8531 (Japan); Yoshizaki, H. [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Saitoh, Y. [Takasaki Advanced Radiation Research Institute, Japan Atomic Energy Agency Takasaki, Gunma 370-1292 (Japan); Okamoto, Y. [Quantum Beam Science Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Iwase, A., E-mail: iwase@mtr.osakafu-u.ac.jp [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan)

    2014-11-01

    Ni{sub 3}V bulk intermetallic compounds with ordered D0{sub 22} structure were irradiated with 16 MeV Au ions at room temperature. The irradiation induced phase transformation was examined by means of the transmission electron microscope (TEM), the extended X-ray absorption fine structure measurement (EXAFS) and the X-ray diffraction (XRD). We also measured the Vickers hardness for unirradiated and irradiated specimens. The TEM observation shows that by the Au irradiation, the lamellar microstructures and the super lattice spot in diffraction pattern for the unirradiated specimen disappeared. This TEM result as well as the result of XRD and EXAFS measurements means that the intrinsic D0{sub 22} structure of Ni{sub 3}V changes into the A1 (fcc) structure which is the lattice structure just below the melting point in the thermal equilibrium phase diagram. The lattice structure change from D0{sub 22} to A1 (fcc) accompanies a remarkable decrease in Vickers microhardness. The change in crystal structure was discussed in terms of the thermal spike and the sequential atomic displacements induced by the energetic heavy ion irradiation.

  10. Thermo-mechanical processing of austenitic steel to mitigate surface related degradation

    Science.gov (United States)

    Idell, Yaakov Jonathan

    Thermo-mechanical processing plays an important role in materials property optimization through microstructure modification, required by demanding modern materials applications. Due to the critical role of austenitic stainless steels, such as 316L, as structural components in harsh environments, e.g. in nuclear power plants, improved degradation resistance is desirable. A novel two-dimensional plane strain machining process has shown promise achieving significant grain size refinement through severe plastic deformation (SPD) and imparting large strains in the surface and subsurface regions of the substrate in various metals and alloys. The deformation process creates a heavily deformed 20 -- 30 micron thick nanocrystalline surface layer with increased hardness and minimal martensite formation. Post-deformation processing annealing treatments have been applied to assess stability of the refined scale microstructures and the potential for obtaining grain boundary engineered microstructures with increased fraction of low-energy grain boundaries and altered grain boundary network structure. Varying the deformation and heat treatment process parameters, allows for development of a full understanding of the nanocrystalline layer and cross-section of the surface substrate created. Micro-characterization was performed using hardness measurements, magnetometry, x-ray diffraction, scanning and transmission electron microscopy to assess property and microstructural changes. This study provides a fundamental understanding of two-dimensional plane strain machining as a thermo-mechanical processing technique, which may in the future deliver capabilities for creating grain boundary engineered surface modified components, typified by a combination of grain refinement with improved grain boundary network interconnectivity attributes suitable for use in harsh environments, such as those in commercial nuclear power plants where improved resistance to irradiation stress corrosion

  11. Etude de la migration des interstitiels dans des austenites Fe, Cr (18), Ni (14) pures et industrielles par irradiation dans un microscope a tres haute tension: Role du carbone et du titane

    Science.gov (United States)

    Housseau, N.; Pelissier, J.

    1983-12-01

    Nous avons étudié le rôle des impurtés (C ou Ti) dans la condensation et la migration des défauts interstitiels. Les échantillons étudiés sont des aciers austénitiques: (a) acier de synthèse de haute pureté (Cr 18, Ni 14, Fe) avec ou sans carbone; (b) acier industriel avec C (800 ppm) ou Ti (0,45%). Les échantillons ont été irradiés dans un microscope à très haute tension aux doses allant de 10 -4 jusqu'à 10 -1 dpa aux températures de 300°C à 400°C. Dans de telles conditions les défauts observés sont des boucles interstitielles. L'étude de la variation de l'épaisseur de la zone dénudée près du bord de la lame mince en fonction de la température nous a permis d'évaluer l'énergie de migration effective de l'interstitiel dans ces alliages. Dans l'austénite de synthèse carburée ou non sa valeur est de 0.8 eV. Dans l'acier industriel au titane carburé ou non on obtient 2.0 eV. Nous n'avons pas observé d'effet lié au carbone. L'examen de la densité de boucles à saturation dans les divers aciers suggère une forte énergie de liaison interstitiel-titane. Cette énergie de liaison, si l'ont admet que le titane est la seule impureté agissante du système, peut être estimée à 1.2 eV.

  12. Scale-bridging analysis on deformation behavior of high-nitrogen austenitic steels.

    Science.gov (United States)

    Lee, Tae-Ho; Ha, Heon-Young; Hwang, Byoungchul; Kim, Sung-Joon; Shin, Eunjoo; Lee, Jong Wook

    2013-08-01

    Scale-bridging analysis on deformation behavior of high-nitrogen austenitic Fe-18Cr-10Mn-(0.39 and 0.69)N steels was performed by neutron diffraction, electron backscattered diffraction (EBSD), and transmission electron microscopy (TEM). Two important modes of deformation were identified depending on the nitrogen content: deformation twinning in the 0.69 N alloy and strain-induced martensitic transformation in the 0.39 N alloy. The phase fraction and deformation faulting probabilities were evaluated based on analyses of peak shift and asymmetry of neutron diffraction profiles. Semi in situ EBSD measurement was performed to investigate the orientation dependence of deformation microstructure and it showed that the variants of ε martensite as well as twin showed strong orientation dependence with respect to tensile axis. TEM observation showed that deformation twin with a {111} mathematical left angle bracket 112 mathematical right angle bracket crystallographic component was predominant in the 0.69 N alloy whereas two types of strain-induced martensites (ε and α' martensites) were observed in the 0.39 N alloy. It can be concluded that scale-bridging analysis using neutron diffraction, EBSD, and TEM can yield a comprehensive understanding of the deformation mechanism of nitrogen-alloyed austenitic steels.

  13. Assessment of the integrity of ferritic-austenitic dissimilar weld joints of different grades of Cr-Mo ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Laha, K.; Chandravathi, K.S.; Parameswaran, P.; Goyal, Sunil; Mathew, M.D. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Metallurgy and Materials Group

    2010-07-01

    Integrity of the 2.25 Cr-1Mo / Alloy 800, 9Cr-1Mo / Alloy 800 and 9Cr-1Mo-VNb / Alloy 800 ferritic-austenitic dissimilar joints, fusion welded employing Inconel 182 electrode, has been assessed under creep conditions at 823 K. The dissimilar weld joints displayed lower creep rupture strength than their respective ferritic steel base metals. The strength reduction was more for 2.25Cr-1Mo steel joint and least for 9Cr-1Mo steel joint. The failure location in the joints was found to shift from the ferritic steel base metal to the intercritical region of heat-affected zone (HAZ) in ferritic steel (type IV cracking) with decrease in stress. At still lower stresses the failure occurred at the ferritic / austenitic weld interface. Localized creep deformation and cavitation in the soft intercritical HAZ induced type IV failure whereas creep cavitation at the weld interface particles induced ferritic / austenitic interface cracking due to high creep strength mismatch across it. Micromechanisms of type IV failure and interface cracking in the ferritic / austenitic joints and different susceptibility to failure for different grades of ferritic steels are discussed based on microstructural investigation, mechanical testing and finite element analysis. (Note from indexer: paper contains many typographical errors.)

  14. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    Science.gov (United States)

    Brémier, S.; Inagaki, K.; Capriotti, L.; Poeml, P.; Ogata, T.; Ohta, H.; Rondinella, V. V.

    2016-11-01

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15-20 μm and migration of cladding elements to the fuel.

  15. Application of ultrasound irradiation on sol-gel technique for corrosion protection of Al65Cu20Fe15 alloy powder

    Science.gov (United States)

    Liang, Bo; Zhang, Baoyan; Wang, Guodong; Li, Di; Zhang, Xiaoming

    2013-11-01

    Al65Cu20Fe15 alloy powder was firstly encapsulated by the conventional sol-gel technique utilizing tetraethoxysilane (TEOS) as the precursor in order to improve its corrosion resistance. The optimization was based on nine well-planned orthogonal experiments (L9 (34)). Four main factors in the encapsulation process (i.e. reaction temperature, ethylenediamine concentration, TEOS concentration and feeding method) were investigated. According to the visual analyses of the result, the optimum condition was obtained. Based on the optimal condition in the conventional sol-gel technique, the encapsulation process was then conducted under ultrasonic irradiation. The effects of ultrasound amplitude and irradiation time on the encapsulation process were also studied. FTIR, XRD, SEM, DLS and EDS were also used to characterize the resulting sample. Finally, the corrosion inhibition efficiency of encapsulated powder attained 99.3% in the acidic condition of pH 1, and the average grain size (d50) of the encapsulated powder was just 4.8% larger than that of the raw powder, implying that there was a thin silica film on the surface of powder.

  16. Shape-Memory-Alloy Actuator For Flight Controls

    Science.gov (United States)

    Barret, Chris

    1995-01-01

    Report proposes use of shape-memory-alloy actuators, instead of hydraulic actuators, for aerodynamic flight-control surfaces. Actuator made of shape-memory alloy converts thermal energy into mechanical work by changing shape as it makes transitions between martensitic and austenitic crystalline phase states of alloy. Because both hot exhaust gases and cryogenic propellant liquids available aboard launch rockets, shape-memory-alloy actuators exceptionally suited for use aboard such rockets.

  17. Modeling the austenite decomposition into ferrite and bainite

    Science.gov (United States)

    Fazeli, Fateh

    2005-12-01

    during the industrial treatments. The thermodynamic boundary conditions for the kinetic model were assessed with respect to paraequilibrium. The potential interaction between the alloying atoms and the moving ferrite-austenite interface, referred to as solute drag effect, was accounted for rigorously in the model. To quantify the solute drag pressure the Purdy-Brechet approach was modified prior to its implementation into the model. (Abstract shortened by UMI.)

  18. Effects of the shape of the foil corners on the irradiation performance of U10Mo alloy based monolithic mini-plates

    Energy Technology Data Exchange (ETDEWEB)

    Ozaltun, Hakan [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory

    2015-06-01

    Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These fuel elements are comprised of a high density, low enrichment, U-Mo alloy based fuel foil, sandwiched between Zirconium liners and encapsulated in Aluminum cladding. The use of a high density fuel in a foil form presents a number of fabrication and operational concerns, such as: foil centering, flatness of the foil, fuel thickness variation, geometrical tilting, foil corner shape etc. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance have been evaluated. As a part of these series of sensitivity studies, the shape of the foil corners were studied. To understand the effects of the corner shapes of the foil on thermo-mechanical performance of the plates, a behavioral model was developed for a selected plate from RERTR-12 experiments (Plate L1P785). Both fabrication and irradiation processes were simulated. Once the thermo-mechanical behavior the plate is understood for the nominal case, the simulations were repeated for two additional corner shapes to observe the changes in temperature, displacement and stress-strain fields. The results from the fabrication simulations indicated that the foil corners do not alter the post-fabrication stress-strain magnitudes. Furthermore, the irradiation simulations revealed that post-fabrication stresses of the foil would be relieved very quickly in operation. While, foils with chamfered and filleted corners yielded stresses with comparable magnitudes, they are slightly lower in magnitudes, and provided a more favorable mechanical response compared with the foil with sharp corners.

  19. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects; Mecanismes de fragilisation sous irradiation aux neutrons d'alliages modeles ferritiques et d'un acier de cuve: amas de defauts

    Energy Technology Data Exchange (ETDEWEB)

    Meslin-Chiffon, E

    2007-11-15

    The embrittlement of reactor pressure vessel (RPV) under irradiation is partly due to the formation of point defects (PD) and solute clusters. The aim of this work was to gain more insight into the formation mechanisms of solute clusters in low copper ([Cu] = 0.1 wt%) FeCu and FeCuMnNi model alloys, in a copper free FeMnNi model alloy and in a low copper French RPV steel (16MND5). These materials were neutron-irradiated around 300 C in a test reactor. Solute clusters were characterized by tomographic atom probe whereas PD clusters were simulated with a rate theory numerical code calibrated under cascade damage conditions using transmission electron microscopy analysis. The confrontation between experiments and simulation reveals that a heterogeneous irradiation-induced solute precipitation/segregation probably occurs on PD clusters. (author)

  20. Model Predictive Control of the Exit Part Temperature for an Austenitization Furnace

    Directory of Open Access Journals (Sweden)

    Hari S. Ganesh

    2016-12-01

    Full Text Available Quench hardening is the process of strengthening and hardening ferrous metals and alloys by heating the material to a specific temperature to form austenite (austenitization, followed by rapid cooling (quenching in water, brine or oil to introduce a hardened phase called martensite. The material is then often tempered to increase toughness, as it may decrease from the quench hardening process. The austenitization process is highly energy-intensive and many of the industrial austenitization furnaces were built and equipped prior to the advent of advanced control strategies and thus use large, sub-optimal amounts of energy. The model computes the energy usage of the furnace and the part temperature profile as a function of time and position within the furnace under temperature feedback control. In this paper, the aforementioned model is used to simulate the furnace for a batch of forty parts under heuristic temperature set points suggested by the operators of the plant. A model predictive control (MPC system is then developed and deployed to control the the part temperature at the furnace exit thereby preventing the parts from overheating. An energy efficiency gain of 5.3 % was obtained under model predictive control compared to operation under heuristic temperature set points tracked by a regulatory control layer.

  1. The sub-zero Celsius treatment of precipitation hardenable semi-austenitic stainless steel

    DEFF Research Database (Denmark)

    Villa, Matteo; Hansen, Mikkel Fougt; Somers, Marcel A. J.

    2015-01-01

    A precipitation hardenable semi-austenitic stainless steel AISI 632 grade was austenitized according to industrial specifications and thereafter subjected to isothermal treatment at sub-zero Celsius temperatures. During treatment, austenite transformed to martensite. The isothermal austenite...

  2. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  3. Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

    Energy Technology Data Exchange (ETDEWEB)

    J. H. Jackson; S. P. Teysseyre

    2012-02-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials of interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.

  4. Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

    Energy Technology Data Exchange (ETDEWEB)

    J. H. Jackson; S. P. Teysseyre

    2012-10-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials of interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.

  5. Mechanical response of proton beam irradiated nitinol

    Energy Technology Data Exchange (ETDEWEB)

    Afzal, Naveed [Centre for Advanced Studies in Physics, GC University, Lahore (Pakistan); Ghauri, I.M., E-mail: ijaz.phys@gmail.co [Centre for Advanced Studies in Physics, GC University, Lahore (Pakistan); Mubarik, F.E.; Amin, F. [Centre for Advanced Studies in Physics, GC University, Lahore (Pakistan)

    2011-01-01

    The present investigation deals with the study of mechanical behavior of proton beam irradiated nitinol at room temperature. The specimens in austenitic phase were irradiated over periods of 15, 30, 45 and 60 min at room temperature using 2 MeV proton beam obtained from Pelletron accelerator. The stress-strain curves of both unirradiated and irradiated specimens were obtained using a universal testing machine at room temperature. The results of the experiment show that an intermediate rhombohedral (R) phase has been introduced between austenite and martensite phase, which resulted in the suppression of direct transformation from austenite to martensite (A-M). Stresses required to start R-phase ({sigma}{sub RS}) and martensitic phase ({sigma}{sub MS}) were observed to decrease with increase in exposure time. The hardness tests of samples before and after irradiation were also carried out using Vickers hardness tester. The comparison reveals that the hardness is higher in irradiated specimens than that of the unirradiated one. The increase in hardness is quite sharp in specimens irradiated for 15 min, which then increases linearly as the exposure time is increased up to 60 min. The generation of R-phase, variations in the transformation stresses {sigma}{sub RS} and {sigma}{sub MS} and increase in hardness of irradiated nitinol may be attributed to lattice disorder and associated changes in crystal structure induced by proton beam irradiation.

  6. The effect of heat treatment on the gouging abrasion resistance of alloy white cast irons

    Science.gov (United States)

    Are, I. R. S.; Arnold, B. K.

    1995-02-01

    A series of heat treatments was employed to vary the microstructure of four commercially important alloy white cast irons, the wear resistance of which was then assessed by the ASTM jaw-crusher gouging abrasion test. Compared with the as-cast condition, standard austenitizing treatments produced a substantial increase in hardness, a marked decrease in the retained aus-tenite content in the matrix, and, in general, a significant improvement in gouging abrasion resistance. The gouging abrasion resistance tended to decline with increasing austenitizing tem-perature, although the changes in hardness and retained austenite content varied, depending on alloy composition. Subcritical heat treatment at 500 ° following hardening reduced the retained austenite content to values less than 10 pct, and in three of the alloys it caused a significant fall in both hardness and gouging abrasion resistance. The net result of the heat treatments was the development of optimal gouging abrasion resistance at intermediate levels of retained aus-tenite. The differing responses of the alloys to both high-temperature austenitizing treatments and to subcritical heat treatments at 500 ° were related to the effects of the differing carbon and alloying-element concentrations on changes in the M s temperature and secondary carbide precipitation.

  7. Cyclic deformation behaviour of austenitic steels at ambient and elevated temperatures

    Indian Academy of Sciences (India)

    Th Nebel; D Eifler

    2003-02-01

    The aim of the present investigation is to characterise cyclic deformation behaviour and plasticity-induced martensite formation of metastable austenitic stainless steels at ambient and elevated temperatures, taking into account the influence of the alloying elements titanium and niobium. Titanium and niobium are ferrite-stabilising elements which influence the ferrite crystallisation. Furthermore, They form carbides and/or carbonitrides and thus limit the austenite-stabilising effect of carbon and nitrogen. Several specimen batches of titanium and niobium alloyed austenite and of a pure Cr-Ni-steel for comparison were tested under stress and total strain control at a frequency of 5 Hz and triangular load-time waveforms. Stress-strain-hysteresis and temperature measurements were used at ambient temperature to characterise cyclic deformation behaviour. Plasticity-induced martensite content was detected with non-destructive magnetic measuring techniques. The experiments yield characteristic cyclic deformation curves and corresponding magnetic signals according to the actual fatigue state and the amount of martensite. Fatigue behaviour of X6CrNiTi1810 (AISI 321), X10CrNiCb189 (AISI 348) and X5CrNi1810 (AISI 304) is characterised by cyclic hardening and softening effects which are strongly influenced by specific loading conditions. Martensite formation varies with the composition, loading conditions, temperature and number of cycles.

  8. High Nb, Ta, and Al creep- and oxidation-resistant austenitic stainless steel

    Science.gov (United States)

    Brady, Michael P [Oak Ridge, TN; Santella, Michael L [Knoxville, TN; Yamamoto, Yukinori [Oak Ridge, TN; Liu, Chain-tsuan [Oak Ridge, TN

    2010-07-13

    An austenitic stainless steel HTUPS alloy includes, in weight percent: 15 to 30 Ni; 10 to 15 Cr; 2 to 5 Al; 0.6 to 5 total of at least one of Nb and Ta; no more than 0.3 of combined Ti+V; up to 3 Mo; up to 3 Co; up to 1 W; up to 0.5 Cu; up to 4 Mn; up to 1 Si; 0.05 to 0.15 C; up to 0.15 B; up to 0.05 P; up to 1 total of at least one of Y, La, Ce, Hf, and Zr; less than 0.05 N; and base Fe, wherein the weight percent Fe is greater than the weight percent Ni wherein said alloy forms an external continuous scale comprising alumina, nanometer scale sized particles distributed throughout the microstructure, said particles comprising at least one composition selected from the group consisting of NbC and TaC, and a stable essentially single phase fcc austenitic matrix microstructure, said austenitic matrix being essentially delta-ferrite-free and essentially BCC-phase-free.

  9. A phase-field model for incoherent martensitic transformations including plastic accommodation processes in the austenite

    Science.gov (United States)

    Kundin, J.; Raabe, D.; Emmerich, H.

    2011-10-01

    If alloys undergo an incoherent martensitic transformation, then plastic accommodation and relaxation accompany the transformation. To capture these mechanisms we develop an improved 3D microelastic-plastic phase-field model. It is based on the classical concepts of phase-field modeling of microelastic problems (Chen, L.Q., Wang Y., Khachaturyan, A.G., 1992. Philos. Mag. Lett. 65, 15-23). In addition to these it takes into account the incoherent formation of accommodation dislocations in the austenitic matrix, as well as their inheritance into the martensitic plates based on the crystallography of the martensitic transformation. We apply this new phase-field approach to the butterfly-type martensitic transformation in a Fe-30 wt%Ni alloy in direct comparison to recent experimental data (Sato, H., Zaefferer, S., 2009. Acta Mater. 57, 1931-1937). It is shown that the therein proposed mechanisms of plastic accommodation during the transformation can indeed explain the experimentally observed morphology of the martensitic plates as well as the orientation between martensitic plates and the austenitic matrix. The developed phase-field model constitutes a general simulations approach for different kinds of phase transformation phenomena that inherently include dislocation based accommodation processes. The approach does not only predict the final equilibrium topology, misfit, size, crystallography, and aspect ratio of martensite-austenite ensembles resulting from a transformation, but it also resolves the associated dislocation dynamics and the distribution, and the size of the crystals itself.

  10. INFLUENCE OF IMPULSE MAGNETIC FIELD ON GRAPHITE MORPHOLOGY OF HIGH-ALLOY IRON

    Directory of Open Access Journals (Sweden)

    A. G. Anisovich

    2011-01-01

    Full Text Available The results of researches of change of microstructure of heavily alloyed austenitic cast-iron ChN1507 subjected to magnetoimpulse processing are given. It is established that microhardness rises on all section of the sample.

  11. Synchronized metal-ion irradiation as a way to control growth of transition-metal nitride alloy films during hybrid HIPIMS/DCMS co-sputtering

    Science.gov (United States)

    Greczynski, Grzegorz

    2016-09-01

    High-power pulsed magnetron sputtering (HIPIMS) is particularly attractive for growth of transition metal (TM) nitride alloys for two reasons: (i) the high ionization degree of the sputtered metal flux, and (ii) the time separation of metal- and gas-ion fluxes incident at the substrate. The former implies that ion fluxes originating from elemental targets operated in HIPIMS are distinctly different from those that are obtained during dc magnetron sputtering (DCMS), which helps to separate the effects of HIPIMS and DCMS metal-ion fluxes on film properties. The latter feature allows one to minimize compressive stress due to gas-ion irradiation, by synchronizing the pulsed substrate bias with the metal-rich-plasma portion of the HIPIMS pulse. Here, we use pseudobinary TM nitride model systems TiAlN, TiSiN, TiTaN, and TiAlTaN to carry out experiments in a hybrid configuration with one target powered by HIPIMS, the other operated in DCMS mode. This allows us to probe the roles of intense and metal-ion fluxes (n = 1 , 2) from HIPIMS-powered targets on film growth kinetics, microstructure, and physical properties over a wide range of M1M2N alloy compositions. TiAlN and TiSiN mechanical properties are shown to be determined by the average metal-ion momentum transfer per deposited atom. Irradiation with lighter metal-ions (M1 =Al+ or Si+ during M1-HIPIMS/Ti-DCMS) yields fully-dense single-phase cubic Ti1-x (M1)x N films. In contrast, with higher-mass film constituent ions such as Ti+, easily exceeds the threshold for precipitation of second phase w-AlN or Si3N4. Based on the above results, a new PVD approach is proposed which relies on the hybrid concept to grow dense, hard, and stress-free thin films with no external heating. The primary targets, Ti and/or Al, operate in DCMS mode providing a continuous flux of sputter-ejected metal atoms to sustain a high deposition rate, while a high-mass target metal, Ta, is driven by HIPIMS to serve as a pulsed source of energetic

  12. Conversion of stacking fault tetrahedra to voids in electron irradiated Fe-Cr-Ni

    Science.gov (United States)

    Kojima, S.; Sano, Y.; Yoshiie, T.; Yoshida, N.; Kiritani, M.

    1986-11-01

    Electron irradiations of the austenitic Fe-13Cr-14Ni alloy were performed with a high voltage electron microscope at temperatures between room temperature and 650 K. Formation of stacking fault tetrahedra, voids and dislocation loops was observed as vacancy clusters. At the lower temperatures, the dominant vacancy clusters were tetrahedra and at the higher temperatures, voids were dominant. In the temperature range at which both tetrahedra and voids were coexistent, conversion of tetrahedra to voids were observed. These results are interpreted as the preferable nucleation of voids at the site of tetrahedra. Local effects of dilatation field at the corner of tetrahedra and the segregation of solute atoms are considered to enhance the nucleation. Clustered defects which are considered to be stacking fault tetrahedra that are formed with D-T fusion neutrons in SUS 316 stainless steel are suggested as the preferable site for void nucleation.

  13. Effect of Treatment Time on the Microstructure of Austenitic Stainless Steel During Low-Temperature Liquid Nitrocarburizing

    Science.gov (United States)

    Wang, Jun; Lin, Yuanhua; Zhang, Qiang; Zeng, Dezhi; Fan, Hongyuan

    2014-09-01

    The effect of treatment time on the microstructure of AISI 304 austenitic stainless steel during liquid nitrocarburizing (LNC) at 703 K (430 °C) was investigated using X-ray diffraction (XRD), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). Experimental results revealed that the modified layer was covered with the alloy surface and the modified layer depth increased extensively from 2 to 33.4 μm with increasing treatment time. SEM and XRD showed that when the 304 stainless steel sample was subjected to LNC at 703 K (430 °C) for less than 4 hours, the main phase of the modified layer was expanded austenite. When the treatment time was prolonged to 8 hours, the abundant expanded austenite was formed and it partially transformed into CrN and ferrite subsequently. With the increased treatment time, more and more CrN precipitate transformed in the overwhelming majority zone in the form of a typical dendritic structure in the nearby outer part treated for 40 hours. Still there was a single-phase layer of the expanded austenite between the CrN part and the inner substrate. TEM showed the expanded austenite decomposition into the CrN and ferrite after longtime treatment even at low temperature.

  14. Synergistic Computational and Microstructural Design of Next- Generation High-Temperature Austenitic Stainless Steels

    Energy Technology Data Exchange (ETDEWEB)

    Karaman, Ibrahim [Texas A& M Engineering Experiment Station, College Station, TX (United States); Arroyave, Raymundo [Texas A& M Engineering Experiment Station, College Station, TX (United States)

    2015-07-31

    The purpose of this project was to: 1) study deformation twinning, its evolution, thermal stability, and the contribution on mechanical response of the new advanced stainless steels, especially at elevated temperatures; 2) study alumina-scale formation on the surface, as an alternative for conventional chromium oxide, that shows better oxidation resistance, through alloy design; and 3) design new generation of high temperature stainless steels that form alumina scale and have thermally stable nano-twins. The work involved few baseline alloys for investigating the twin formation under tensile loading, thermal stability of these twins, and the role of deformation twins on the mechanical response of the alloys. These baseline alloys included Hadfield Steel (Fe-13Mn-1C), 316, 316L and 316N stainless steels. Another baseline alloy was studied for alumina-scale formation investigations. Hadfield steel showed twinning but undesired second phases formed at higher temperatures. 316N stainless steel did not show signs of deformation twinning. Conventional 316 stainless steel demonstrated extensive deformation twinning at room temperature. Investigations on this alloy, both in single crystalline and polycrystalline forms, showed that deformation twins evolve in a hierarchical manner, consisting of micron–sized bundles of nano-twins. The width of nano-twins stays almost constant as the extent of strain increases, but the width and number of the bundles increase with increasing strain. A systematic thermomechanical cycling study showed that the twins were stable at temperatures as high as 900°C, after the dislocations are annealed out. Using such cycles, volume fraction of the thermally stable deformation twins were increased up to 40% in 316 stainless steel. Using computational thermodynamics and kinetics calculations, we designed two generations of advanced austenitic stainless steels. In the first generation, Alloy 1, which had been proposed as an alumina

  15. Influence of Martensite Fraction on the Stabilization of Austenite in Austenitic-Martensitic Stainless Steels

    Science.gov (United States)

    Huang, Qiuliang; De Cooman, Bruno C.; Biermann, Horst; Mola, Javad

    2016-05-01

    The influence of martensite fraction ( f α') on the stabilization of austenite was studied by quench interruption below M s temperature of an Fe-13Cr-0.31C (mass pct) stainless steel. The interval between the quench interruption temperature and the secondary martensite start temperature, denoted as θ, was used to quantify the extent of austenite stabilization. In experiments with and without a reheating step subsequent to quench interruption, the variation of θ with f α' showed a transition after transformation of almost half of the austenite. This trend was observed regardless of the solution annealing temperature which influenced the martensite start temperature. The transition in θ was ascribed to a change in the type of martensite nucleation sites from austenite grain and twin boundaries at low f α' to the faults near austenite-martensite (A-M) boundaries at high f α'. At low temperatures, the local carbon enrichment of such boundaries was responsible for the enhanced stabilization at high f α'. At high temperatures, relevant to the quenching and partitioning processing, on the other hand, the pronounced stabilization at high f α' was attributed to the uniform partitioning of the carbon stored at A-M boundaries into the austenite. Reduction in the fault density of austenite served as an auxiliary stabilization mechanism at high temperatures.

  16. Tensile properties of Inconel 718 after low temperature neutron irradiation

    Science.gov (United States)

    Byun, T. S.; Farrell, K.

    2003-05-01

    Tensile properties of Inconel 718 (IN718) have been investigated after neutron irradiation to 0.0006-1.2 dpa at 60-100 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The alloy was exposed in solution-annealed (SA) and precipitation-hardened (PH) conditions. Before irradiation, the yield strength of PH IN718 was about 1170 MPa, which was 3.7 times higher than that of SA IN718. In the SA condition, an almost threefold increase in yield strength was found at 1.2 dpa, but the alloy retained a positive strain-hardening capability and a uniform ductility of more than 20%. Comparisons showed that the strain-hardening behavior of the SA IN718 is similar to that of a SA 316LN austenitic stainless steel. In the PH condition, the IN718 displayed no radiation-induced hardening in yield strength and significant softening in ultimate tensile strength. The strain-hardening capability of the PH IN718 decreased with dose as the radiation-induced dissolution of precipitates occurred, which resulted in the onset of plastic instability at strains less than 1% after irradiation to 0.16 or 1.2 dpa. An analysis on plastic instability indicated that the loss of uniform ductility in PH IN718 was largely due to the reduction in strain-hardening rate, while in SA IN718 and SA 316LN stainless steel it resulted primarily from the increase of yield stress.

  17. Characterization of Low Temperature Ferrite/Austenite Transformations in the Heat Affected Zone of 2205 Duplex Stainless Steel Arc Welds

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, T A; Elmer, J W; Babu, S S; Vitek, J M

    2003-08-20

    Spatially Resolved X-Ray Diffraction (SRXRD) has been used to identify a previously unobserved low temperature ferrite ({delta})/austenite({gamma}) phase transformation in the heat affected zone (HAZ) of 2205 Duplex Stainless Steel (DSS) welds. In this ''ferrite dip'' transformation, the ferrite transforms to austenite during heating to peak temperatures on the order of 750 C, and re-transforms to ferrite during cooling, resulting in a ferrite volume fraction equivalent to that in the base metal. Time Resolved X-Ray Diffraction (TRXRD) and laser dilatometry measurements during Gleeble{reg_sign} thermal simulations are performed in order to verify the existence of this low temperature phase transformation. Thermodynamic and kinetic models for phase transformations, including both local-equilibrium and para-equilibrium diffusion controlled growth, show that diffusion of substitutional alloying elements does not provide a reasonable explanation for the experimental observations. On the other hand, the diffusion of interstitial alloying elements may be rapid enough to explain this behavior. Based on both the experimental and modeling results, two mechanisms for the ''ferrite dip'' transformation, including the formation and decomposition of secondary austenite and an athermal martensitic-type transformation of ferrite to austenite, are considered.

  18. Internal hydrogen-induced subcritical crack growth in austenitic stainless steels

    Science.gov (United States)

    Huang, J. H.; Altstetter, C. J.

    1991-11-01

    The effects of small amounts of dissolved hydrogen on crack propagation were determined for two austenitic stainless steel alloys, AISI 301 and 310S. In order to have a uniform distribution of hydrogen in the alloys, they were cathodically charged at high temperature in a molten salt electrolyte. Sustained load tests were performed on fatigue precracked specimens in air at 0 ‡C, 25 ‡C, and 50 ‡C with hydrogen contents up to 41 wt ppm. The electrical potential drop method with optical calibration was used to continuously monitor the crack position. Log crack velocity vs stress intensity curves had definite thresholds for subcritical crack growth (SCG), but stage II was not always clearly delineated. In the unstable austenitic steel, AISI 301, the threshold stress intensity decreased with increasing hydrogen content or increasing temperature, but beyond about 10 wt ppm, it became insensitive to hydrogen concentration. At higher concentrations, stage II became less distinct. In the stable stainless steel, subcritical crack growth was observed only for a specimen containing 41 wt ppm hydrogen. Fractographic features were correlated with stress intensity, hydrogen content, and temperature. The fracture mode changed with temperature and hydrogen content. For unstable austenitic steel, low temperature and high hydrogen content favored intergranular fracture while microvoid coalescence dominated at a low hydrogen content. The interpretation of these phenomena is based on the tendency for stress-induced phase transformation, the different hydrogen diffusivity and solubility in ferrite and austenite, and outgassing from the crack tip. After comparing the embrittlement due to internal hydrogen with that in external hydrogen, it is concluded that the critical hydrogen distribution for the onset of subcritical crack growth is reached at a location that is very near the crack tip.

  19. Transformation in Austenitic Stainless Steel Sheet under Different Loading Directions

    NARCIS (Netherlands)

    Boogaard, van den A.H.; Krauer, J.; Hora, P.

    2011-01-01

    The stress-strain relation for austenitic stainless steels is based on 2 main contributions: work hardening and a phase transformation from austenite to martensite. The transformation is highly temperature dependent. In most models for phase transformation from austenite to martensite, the stress tr

  20. Microstructural observations of HFIR-irratiated austenitic stainless steels including welds from JP9-16

    Energy Technology Data Exchange (ETDEWEB)

    Sawai, T.; Shiba, K.; Hishinuma, A.

    1996-04-01

    Austenitic stainless steels, including specimens taken from various electron beam (EB) welds, have been irradiated in HFIR Phase II capsules, JP9-16. Fifteen specimens irradiated at 300, 400, and 500{degrees}C up to 17 dpa are so far examined by a transmission electron microscope (TEM). In 300{degrees}C irradiation, cavities were smaller than 2nm and different specimens showed little difference in cavity microstructure. At 400{degrees}C, cavity size was larger, but still very small (<8 nm). At 500{degrees}C, cavity size reached 30 nm in weld metal specimens of JPCA, while cold worked JPCA contained a small (<5 nm) cavities. Inhomogeneous microstructural evolution was clearly observed in weld-metal specimens irradiated at 500{degrees}C.

  1. Effect of high-intensity ultrasonic irradiation on the modification of solidification microstructure in a Si-rich hypoeutectic Al-Si alloy

    Energy Technology Data Exchange (ETDEWEB)

    Das, A., E-mail: A.Das@swansea.ac.uk [Materials Research Centre, School of Engineering, Swansea University, Singleton Park, Swansea, SA2 8PP (United Kingdom); Kotadia, H.R. [Brunel Centre for Advanced Solidification Technology, Brunel University, Uxbridge, UB8 3PH (United Kingdom)

    2011-02-15

    Effect of high-intensity ultrasound irradiation in modifying complex solidification microstructure is explored in a high Si containing Al-Si alloy and the origin of microstructural changes explained on the basis of nucleation and growth behaviour. Complete suppression of dendritic growth and dramatic refinement to globular morphology were observed for primary {alpha}-Al grains. Strong supportive evidence is presented towards enhanced and prolonged heterogeneous nucleation triggered by cavitation induced increase in the equilibrium melting point and effective dissipation of latent heat at the solidification front. Morphological evolution of eutectic Si and intermetallic particles is found to be dominated by coarsening and spherodisation from strong fluid flow in areas of intense cavitation near the ultrasonic radiator. Outside the region of direct energy transfer, Si particle morphology appears to be controlled predominantly by the imposed cooling conditions. Extremely fine and short Si-platelets observed in the intergranular spaces near the radiator are explained on the basis of probable rapid cooling of final liquid pockets of small volume and large surface area, rather than refinement through ultrasound.

  2. Modeling of austenite to ferrite transformation

    Indian Academy of Sciences (India)

    Mohsen Kazeminezhad

    2012-06-01

    In this research, an algorithm based on the -state Potts model is presented for modeling the austenite to ferrite transformation. In the algorithm, it is possible to exactly track boundary migration of the phase formed during transformation. In the algorithm, effects of changes in chemical free energy, strain free energy and interfacial energies of austenite–austenite, ferrite–ferrite and austenite–ferrite during transformation are considered. From the algorithm, the kinetics of transformation and mean ferrite grain size for different cooling rates are calculated. It is found that there is a good agreement between the calculated and experimental results.

  3. Cleavage fracture and irradiation embrittlement of fusion reactor alloys: mechanisms, multiscale models, toughness measurements and implications to structural integrity assessment

    Science.gov (United States)

    Odette, G. R.; Yamamoto, T.; Rathbun, H. J.; He, M. Y.; Hribernik, M. L.; Rensman, J. W.

    2003-12-01

    We describe the highly efficient master curves-shifts (MC-Δ T) method to measure and apply cleavage fracture toughness, KJc ( T), data and show that it is applicable to 9Cr martensitic steels. A reference temperature, T0, indexes the invariant MC shape on an absolute temperature scale. Then, T0 shifts (Δ T) are used to account for various effects of size and geometry, loading rate and irradiation embrittlement (Δ Ti). The paper outlines a multiscale model, relating atomic to structural scale fracture processes, that underpins the MC-Δ T method. At the atomic scale, we propose that the intrinsic microarrest toughness, Kμ( T), of the body-centered cubic ferrite lattice dictates an invariant shape of the macroscopic KJc ( T) curve. KJc ( T) can be modeled in terms of the true stress-strain ( σ- ɛ) constitutive law, σ ( T, ɛ), combined with a temperature-dependent critical local stress, σ*( T) and stressed volume, V*. The local fracture properties, σ*( T)- V*, are governed by coarse-scale brittle trigger particles and Kμ( T). Irradiation (and high strain rate) induced increases in the yield stress, Δ σy, lead to Δ Ti, with typical Δ Ti/Δ σy≈0.6±0.15 °C/MPa. However, Δ Ti associated with decreases in σ* and V* can result from a number of potential non-hardening embrittlement (NHE) mechanisms, including a large amount of He on grain boundaries. Estimates based on available data suggest that this occurs at >500-700 appm bulk He. Hardening and NHE are synergistic, and can lead to very large Δ Ti. NHE is signaled by large (>1 °C/MPa), or even negative, values of Δ Ti/Δ σy (for Δ σy1 and Δc/ Δy≫1. Indeed, in some circumstances, the benefits of irradiation due to increases in Pc may more than offset the liabilities of the decreases in Δc.

  4. Ion-irradiation-assisted phase selection in single crystalline Fe7Pd3 ferromagnetic shape memory alloy thin films: from fcc to bcc along the Nishiyama-Wassermann path.

    Science.gov (United States)

    Arabi-Hashemi, A; Mayr, S G

    2012-11-09

    When processing Fe-Pd ferromagnetic shape memory thin films, selection of the desired phases and their transformation temperatures constitutes one of the largest challenges from an application point of view. In the present contribution we demonstrate that irradiation with 1.8 MeV Kr(+) ions is the method of choice to achieve this goal: Single crystalline Fe(7)Pd(3) thin films that are grown with molecular beam epitaxy on MgO (001) substrates and subsequently irradiated with ions reveal a phase transformation along the whole phase transformation path ranging from fcc austenite to bcc martensite. While for 10(14) ions/cm(2) a fcc-fct phase transformation is observed, increasing the fluence to 5 × 10(14) ions/cm(2) and 5 × 10(15) ions/cm(2) leads to a phase transformation to the bcc phase. Pole figure measurements reveal an orientation relationship for the fcc-bcc phase transformation according to Nishiyama and Wassermann.

  5. Investigations on avoidance of hot cracks during laser welding of austenitic Cr-Ni steels and nickel-based alloys using temperature field tailoring. Final report; Untersuchungen zur Vermeidung von Heissrissen beim Laserstrahlschweissen von austenitischen Cr-Ni-Staehlen und Nickelbasislegierungen mittels Temperaturfeld-Tailoring. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-08

    The aim of the project was to transfer the developed method of laser beam welding of heat treated machining steels of temperature field tailoring on hot crack endangered austenitic Cr-Ni steels and nickel-based alloys. With this method, transient thermal stresses adjacent to the weld are produced by an travelling induction heating so that the hot cracking is prevented during welding. As test materials the austenitic Cr-Ni steel with sulfur additive 1.4305, the Cr-Ni steels 1.4404 and 1.4435 and the nickel-based alloy Udimet 720 were selected. As a result of the research it was shown that a hot crack-free laser welding in the investigated materials using at least three different welding and material-technical approaches is possible. [German] Das Ziel des Forschungsvorhabens bestand darin, das fuer das Laserstrahlschweissen verguetbarer Automatenstaehle entwickelte Verfahren des Temperaturfeld-Tailorings auf heissrissgefaehrdete austenitische Cr-Ni-Staehle und Nickelbasislegierungen zu uebertragen. Mit diesem Verfahren werden waehrend des Schweissens transiente thermische Spannungen neben der Schweissnaht durch eine mitlaufende induktive Erwaermung so erzeugt, dass die Heissrissbildung verhindert wird. Als Versuchswerkstoffe wurden der austenitische Cr-Ni-Stahl mit Schwefelzusatz 1.4305, die Cr-Ni-Staehle 1.4404 und 1.4435 sowie die Nickelbasislegierung Udimet 720 ausgewaehlt. Im Ergebnis des Forschungsvorhabens konnte gezeigt werden, dass ein heissrissfreies Laserstrahlschweissen bei den untersuchten Werkstoffen unter Nutzung von mindestens drei verschiedenen schweiss- und werkstofftechnischen Ansaetzen moeglich ist: Erstens koennen mit einem Temperaturfeld-Tailoring bei im Stumpfstoss zu verschweissenden Blechen aus austenitischen Staehlen bis mindestens 6 mm Dicke senkrecht zur Naht und parallel zur Blechoberflaeche wirkende transiente Druckspannungen erzeugt werden, die der Bildung von Mittelrippenrissen oder dazu parallel liegenden Heissrissen entgegenwirken

  6. Development of a Nano-Satellite Micro-Coupling Mechanism with Characterization of a Shape Memory Alloy Interference Joint

    Science.gov (United States)

    2010-12-01

    3 Endothermic for martensite to austenite transformation. Exothermic for austenite to martensite transformation. 4 Note that the...original work by definition . Unique contributions to the related art of coupling devices and shape memory alloys are as follows: 1) A zero impact

  7. Post irradiated microstructural characterization of Zr–1Nb alloy by X-ray diffraction technique and positron annihilation spectroscopy

    Indian Academy of Sciences (India)

    P S Chowdhury; P Mukherjee; N Gayathri; M Bhattacharya; A Chatterjee; P Barat; P M G Nambissan

    2011-06-01

    Zr–1Nb samples were irradiated with 116 MeV O5+ ions at different doses ranging from 5 × 1017 to 8 × 1018 O5+/m2. X-ray diffraction line profile analysis was performed to characterize the microstructural parameters of these samples. Average domain size, microstrain and dislocation density were estimated as a function of dose. An anomaly was observed in the values of these parameters at a dose of 2 × 1018 O5+/m2. Positron annihilation spectroscopy was used to determine the existence and nature of vacancy clusters in the samples. Isochronal annealing was carried out for a sample to study the evolution of defect clusters.

  8. MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

    Energy Technology Data Exchange (ETDEWEB)

    Farrell, K.

    2003-09-26

    Metals that can sustain plastic deformation homogeneously throughout their bulk tend to be tough and malleable. Often, however, if a metal has been hardened it will no longer deform uniformly. Instead, the deformation occurs in narrow bands on a microscopic scale wherein stresses and strains become concentrated in localized zones. This strain localization degrades the mechanical properties of the metal by causing premature plastic instability failure or by inducing the formation of cracks. Irradiation with neutrons hardens a metal and makes it more prone to deformation by strain localization. Although this has been known since the earliest days of radiation damage studies, a full measure of the connection between neutron irradiation hardening and strain localization is wanting, particularly in commercial alloys used in the construction of nuclear reactors. Therefore, the goal of this project is to systematically map the extent of involvement of strain localization processes in plastic deformation of three reactor alloys that have been neutron irradiated. The deformation processes are to be identified and related to changes in the tensile properties of the alloys as functions of neutron fluence (dose) and degree of plastic strain. The intent is to define the role of strain localization in radiation embrittlement phenomena. The three test materials are a tempered bainitic A533B steel, representing reactor pressure vessel steel, an annealed 316 stainless steel and annealed Zircaloy-4 representing reactor internal components. These three alloys cover the range of crystal structures usually encountered in structural alloys, i.e. body-centered cubic (bcc), face-centered cubic (fcc), and close-packed hexagonal (cph), respectively. The experiments were conducted in three Phases, corresponding to the three years duration of the project. Phases 1 and 2 addressed irradiations and tensile tests made at near-ambient temperatures, and covered a wide range of neutron fluences

  9. New observations on formation of thermally induced martensite in Fe–30%Ni–1%Pd alloy

    Indian Academy of Sciences (India)

    Gokcen Yildiz; Yasin Gokturk Yildiz; Saffet Nezir

    2013-02-01

    Kinetical, morphological, crystallographical and thermal characteristics of thermally induced martensite in an Fe–30%Ni–1%Pd alloy has been studied by scanning electron microscopy (SEM), transmission electron microscopy (TEM), differential scanning calorimetry (DSC) and X-ray diffraction method. Kinetics of transformation was found to be as athermal. SEM and TEM observations and X-ray method revealed ' () martensite formation in the austenite phase of alloy by thermal effect. The crystallographic orientation relationship between austenite and ' () martensite was found to be having Kurdjumov–Sachs (K–S) type relationship. In addition, the lattice parameters of austenite and martensite phases were calculated from X-ray diffraction patterns.

  10. Alumina-Forming Austenitics: A New Approach to Thermal and Degradation Resistant Stainless Steels for Industrial Use

    Energy Technology Data Exchange (ETDEWEB)

    David A Helmick; John H Magee; Michael P Brady

    2012-05-31

    A series of developmental AFA alloys was selected for study based on: 25 Ni wt.% (alloys A-F), 20 wt% Ni (alloys G-H), and 12 Ni wt.% (alloys I-L). An emphasis in this work was placed on the lower alloy content direction for AFA alloys to reduce alloy raw material cost, rather than more highly alloyed and costly AFA alloys for higher temperature performance. Alloys A-D explored the effects of Al (3-4 wt.%) and C (0.05-0.2 wt.%) in the Fe-25Ni-14Cr-2Mn-2Mo-1W-1Nb wt.% base range; alloys E and F explored the effects of removing costly Mo and W additions in a Fe-25Ni-14Cr-4Al-2.5Nb-2Mn-0.2C base, alloys G and H examined Nb (1-2.5wt.%) and removal of Mo, W in a Fe-20Ni-14Cr-3Al-2Mn-0.2 C wt.% base; and alloys I-L examined effects of C (0.1-0.2 wt.%) and Mn (5-10 wt.%) on a low cost Fe-14Cr-12Ni-3Cu-2.5Al wt.% base (no Mo, W additions). Creep testing resulted in elemental trends that included the beneficial effect of higher carbon and lower niobium in 20-25%Ni AFA alloys and, the beneficial of lower Mn in 12%Ni AFA alloys. Corrosion tests in steam and sulfidation-oxidation environments showed, in general, these alloys were capable of a ten-fold improvement in performance when compared to conventional austenitic stainless steels. Also, corrosion test results in metal-dusting environments were promising and, warrant further investigation.

  11. Void denuded zone formation for Fe–15Cr–15Ni steel and PNC316 stainless steel under neutron and electron irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Sekio, Yoshihiro, E-mail: sekio.yoshihiro@jaea.go.jp [Oarai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Yamashita, Shinichiro [Oarai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Sakaguchi, Norihito [Center for Advanced Research of Energy Technology, Hokkaido University, Hokkaido 060-0808 (Japan); Takahashi, Heishichiro [Oarai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Center for Advanced Research of Energy Technology, Hokkaido University, Hokkaido 060-0808 (Japan)

    2015-03-15

    Highlights: • Austenitic stainless steel developed to improve void swelling was used. • Void denuded zone formed near grain boundary can be affected by vacancy mobility. • Vacancy migration energy was estimated from void denuded zone width in the steel. - Abstract: Irradiation-induced void denuded zone (VDZ) formation near grain boundaries was studied to clarify the effects of minor alloying elements on vacancy diffusivity during irradiation in practical PNC316 stainless steel developed for nuclear reactor core materials. The test materials were Fe–15Cr–15Ni steel without additives and PNC316 stainless steel; the latter contains minor alloying elements to improve the void swelling resistance. These steels were neutron-irradiated in the experimental fast reactor JOYO at temperatures from 749 K to 775 K and fast neutron doses of 18–103 dpa, and electron irradiation was also carried out using 1 MeV high voltage electron microscopy at temperatures of 723 K and 773 K and doses up to 14.4 dpa. VDZ formation was analyzed by TEM microstructural observation after irradiation by considering radiation-induced segregation near the grain boundaries. VDZs were formed near random grain boundaries with higher misfit angles in both Fe–15Cr–15Ni and PNC316 steels. The VDZ widths in the PNC316 stainless steel were narrower than those for the Fe–15Cr–15Ni steel for all neutron and electron irradiations. The VDZ width analysis implied that the vacancy diffusivity was reduced in PNC316 stainless steel as a result of interaction of vacancies with minor alloying elements.

  12. Microstructure in the Weld Metal of Austenitic-Pearlitic Dissimilar Steels and Diffusion of Element in the Fusion Zone

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Microstructure and alloy element distribution in the welded joint between austenitic stainless steel(1Cr18Ni9Ti)and pearlitic heat-resistant steel (1Cr5Mo)were researched by means of light microscopy, scanning electron microscopy(SEM)and electron probe microanalysis(EPMA).Microstructure, divisions of the fusion zone and elemental diffusion distributions in the welded joints were investigated. Furthermore, solidification microstructure and δ-ferrite distribution in the weld metal of these steels are also discussed.

  13. The effects of alloying elements Al and In on Ni-Mn-Ga shape memory alloys, from first principles.

    Science.gov (United States)

    Chen, Jie; Li, Yan; Shang, Jia-Xiang; Xu, Hui-Bin

    2009-01-28

    The electronic structures and formation energies of the Ni(9)Mn(4)Ga(3-x)Al(x) and Ni(9)Mn(4)Ga(3-x)In(x) alloys have been investigated using the first-principles pseudopotential plane-wave method based on density functional theory. The results show that both the austenite and martensite phases of Ni(9)Mn(4)Ga(3) alloy are stabilized by Al alloying, while they become unstable with In alloying. According to the partial density of states and structural energy analysis, different effects of Al and In alloying on the phase stability are mainly attributed to their chemical effects. The formation energy difference between the austenite and martensite phases decreases with Al or In alloying, correlating with the experimentally reported changes in martensitic transformation temperature. The shape factor plays an important role in the decrease of the formation energy difference.

  14. Fabrication technological development of the oxide dispersion strengthened alloy MA957 for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    ML Hamilton; DS Gelles; RJ Lobsinger; GD Johnson; WF Brown; MM Paxton; RJ Puigh; CR Eiholzer; C Martinez; MA Blotter

    2000-03-27

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material, in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report. The alloy is a ferritic stainless steel developed by International Nickel Company specifically for structural reactor applications. It is strengthened by a very fine, uniformly distributed yttria dispersoid. Its fabrication involves a mechanical alloying process and subsequent extrusion, which ultimately results in a highly elongated grain structure. While the presence of the dispersoid produces a material with excellent strength, the body centered cubic structure inherent to the material coupled with the high aspect ratio that results from processing operations produces some difficulties with ductility. The alloy is very sensitive to variations in a number of processing parameters, and if the high strength is once lost during fabrication, it cannot be recovered. The microstructural evolution of the alloy under irradiation falls into two regimes. Below about 550 C, dislocation development, {alpha}{prime} precipitation and void evolution in the matrix are observed, while above about 550 C damage appears to be restricted to cavity formation within oxide particles. The thermal expansion of the alloy is very similar to that of HT9 up to the temperature where HT9 undergoes a phase transition to austenitic. Pulse magnetic welding of end caps onto MA957 tubing can be accomplished in a manner similar to that in which it is performed on HT9, although the welding parameters appear to be very sensitive to variations in the tubing that result from small changes in fabrication conditions. The tensile and stress rupture behavior of the alloy are acceptable in the unirradiated condition, being comparable to HT9 below about 700 C and exceeding those of HT9

  15. Expanded austenite, crystallography and residual stress

    DEFF Research Database (Denmark)

    Christiansen, Thomas; Hummelshøj, Thomas Strabo; Somers, Marcel A. J.

    2010-01-01

    compositions and (b) unravelling of the contributions of stress-depth and composition-depth profiles in expanded austenite layers are summarised and discussed. It is shown through simulation of line profiles that the combined effects of composition gradients, stress gradients and stacking fault gradients can...

  16. A characteristic of austenitic ductile iron

    Directory of Open Access Journals (Sweden)

    A. Tabor

    2007-04-01

    Full Text Available The article shows the results of investigations of the mechanical properties conducted on austenitic ductile iron with an addi-tion of 23-24% Ni. The examined mechanical properties included: tensile strength (Rm, proof stress (Rp0,2, elongation (A5 and reduction of area (Z at reduced and low temperatures.

  17. Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-09-29

    A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early-generation powder-metallurgy (PM) oxide dispersion-strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonable matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C, but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, To, in alloys irradiated to 7 dpa and higher.

  18. Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-09-29

    A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early generation powder-metallurgy (PM) oxide dispersion strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonable matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, To, in alloys irradiated to 7 dpa and higher.

  19. Environmentally Assisted Cracking of Nickel Alloys - A Review

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, R

    2004-07-12

    Nickel can dissolve a large amount of alloying elements while still maintaining its austenitic structure. That is, nickel based alloys can be tailored for specific applications. The family of nickel alloys is large, from high temperature alloys (HTA) to corrosion resistant alloys (CRA). In general, CRA are less susceptible to environmentally assisted cracking (EAC) than stainless steels. The environments where nickel alloys suffer EAC are limited and generally avoidable by design. These environments include wet hydrofluoric acid and hot concentrated alkalis. Not all nickel alloys are equally susceptible to cracking in these environments. For example, commercially pure nickel is less susceptible to EAC in hot concentrated alkalis than nickel alloyed with chromium (Cr) and molybdenum (Mo). The susceptibility of nickel alloys to EAC is discussed by family of alloys.

  20. Influence of plastic strain localization on the stress corrosion cracking of austenitic stainless steels; Influence de la localisation de la deformation plastique sur la CSC d'aciers austenitiques inoxydables

    Energy Technology Data Exchange (ETDEWEB)

    Cisse, S.; Tanguy, B. [CEA Saclay, DEN, SEMI, 91 - Gif-sur-Yvette (France); Andrieu, E.; Laffont, L.; Lafont, M.Ch. [Universite de Toulouse. CIRIMAT, UPS/INPT/CNRS, 31 - Toulous (France)

    2010-03-15

    The authors present a research study of the role of strain localization on the irradiation-assisted stress corrosion cracking (IASCC) of vessel steel in PWR-type (pressurized water reactor) environment. They study the interaction between plasticity and intergranular corrosion and/or oxidation mechanisms in austenitic stainless steels with respect to sublayer microstructure transformations. The study is performed on three austenitic stainless grades which have not been sensitized by any specific thermal treatment: the A286 structurally hardened steel, and the 304L and 316L austenitic stainless steels

  1. Expanded austenite in nitrided layers deposited on austenitic and super austenitic stainless steel grades; Analise da austenita expandida em camadas nitretadas em acos inoxidaveis austeniticos e superaustenitico

    Energy Technology Data Exchange (ETDEWEB)

    Casteletti, L.C.; Fernandes, F.A.P.; Heck, S.C. [Universidade de Sao Paulo (EESC/USP), Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia de Materais, Aeronautica e Automobilistica; Oliveira, A.M. [Instituto de Educacao, Ciencia e Tecnologia do Maranhao (IFMA), Sao Luis, MA (Brazil); Gallego, J., E-mail: gallego@dem.feis.unesp.b [UNESP, Ilha Solteira, SP (Brazil). Dept. Engenharia Mecanica

    2010-07-01

    In this work nitrided layers deposited on austenitic and super austenitic stainless steels were analyzed through optical microscopy and X-rays diffraction analysis (XRD). It was observed that the formation of N supersaturated phase, called expanded austenite, has promoted significant increment of hardness (> 1000HV). XRD results have indicated the anomalous displacement of the diffracted peaks, in comparison with the normal austenite. This behavior, combined with peaks broadening, it was analyzed in different nitriding temperatures which results showed good agreement with the literature. (author)

  2. Joining silicon carbide to austenitic stainless steel through diffusion welding; Stellingen behorende bij het proefschrift

    Energy Technology Data Exchange (ETDEWEB)

    Krugers, Jan-Paul

    1993-01-19

    In this thesis, the results are presented of a study dealing with joining silicon carbide to austenitic stainless steel AIS316 by means of diffusion welding. Welding experiments were carried out without and with the use of a metallic intermediate, like copper, nickel and copper-nickel alloys at various conditions of process temperature, process time, mechanical pressure and interlayer thickness. Most experiments were carried out in high vacuum. For reasons of comparison, however, some experiments were also carried out in a gas shielded environment of 95 vol.% Ar and 5 vol.% H2.

  3. Austenitic steel corrosion in IGCC environment. Characterisation by photon and nuclear microprobes

    Science.gov (United States)

    Dillmann, Philippe; Weulersse, Katia; Regad, Belkacem; Moulin, Gérard; Barrett, Ray; Bonnin-Mosbah, Michelle; Lequien, Stéphane; Berger, Pascal

    2001-07-01

    An austenitic steel sample was treated simulating particular working conditions of an integrated gasification combined cycle (IGCC) power plant. Several classical characterisation techniques were used to investigate the oxide scales. In addition, micro-particle-induced X-ray emission (PIXE) and Rutherford backscattering spectrometry (RBS) analyses were performed and permit us to identify several phases constitutive of the oxide. Moreover, micro-X-ray absorption near edge structure (XANES) experiments allow us to determine the valence of the vanadium incorporated in the scale in the form of microscopic islets. The comparison of all these results leads to the proposal of a corrosion mechanism for this alloy.

  4. The Formation of Multipoles during the High-Temperature Creep of Austenitic Stainless Steels

    DEFF Research Database (Denmark)

    Howell, J.; Nielsson, O.; Horsewell, Andy

    1981-01-01

    It is shown that multipole dislocation configurations can arise during power-law creep of certain austenitic stainless steels. These multipoles have been analysed in some detail for two particular steels (Alloy 800 and a modified AISI 316L) and it is suggested that they arise either during...... instantaneous loading or during the primary creep stage. Trace analysis has shown that the multipoles are confined to {1 1 1} planes during primary creep but are not necessarily confined to these planes during steady-state creep unless they are pinned by interstitials....

  5. Micromagnetic and Mössbauer spectroscopic investigation of strain-induced martensite in austenitic stainless steel

    Science.gov (United States)

    Mészáros, L.; Kéldor, M.; Hidasi, B.; Vértes, A.; Czakó-Nagy, I.

    1996-08-01

    Strain-induced martensite in 18/8 austenitic stainless steel was studied. Magnetic measurements and Mössbauer spectroscopic investigations were performed to characterize the amount of α’-martensite due to room-temperature plastic tensile loading. The effects of cold work and annealing heat treatment were explored using magnetic Barkhausen noise, saturation polarization, coercive force, hardness, and conversion electron Mössbauer spectra measurements. The results of the magnetic measurements were compared to results obtained by Mössbauer spectroscopy. The suggested Barkhausen noise measurement technique proved to be a useful quantitative and nondestructive method for determining the ferromagnetic phase ratio of the studied alloy.

  6. Feasibility of surface-coated friction stir welding tools to join AISI 304 grade austenitic stainless steel

    Institute of Scientific and Technical Information of China (English)

    A.K. LAKSHMINARAYANAN; C.S.RAMACHANDRAN; V.BALASUBRAMANIAN

    2014-01-01

    An attempt is made to develop the tools that are capable enough to withstand the shear, impact and thermal forces that occur during friction stir welding of stainless steels. The atmospheric plasma spray and plasma transferred arc hardfacing processes are employed to deposit refractory ceramic based composite coatings on the Inconel 738 alloy. Five different combinations of self-fluxing alloy powder and 60% ceramic rein-forcement particulate mixtures are used for coating. The best friction stir welding tool selected based on tool wear analysis is used to fabricate the austenitic stainless steel joints.

  7. Effect of Structural Heterogeneity on In Situ Deformation of Dissimilar Weld Between Ferritic and Austenitic Steel

    Science.gov (United States)

    Ghosh, M.; Santosh, R.; Das, S. K.; Das, G.; Mahato, B.; Korody, J.; Kumar, S.; Singh, P. K.

    2015-08-01

    Low-alloy steel and 304LN austenitic stainless steel were welded using two types of buttering material, namely 309L stainless steel and IN 182. Weld metals were 308L stainless steel and IN 182, respectively, for two different joints. Cross-sectional microstructure of welded assemblies was investigated. Microhardness profile was determined perpendicular to fusion boundary. In situ tensile test was performed in scanning electron microscope keeping low-alloy steel-buttering material interface at the center of gage length. Adjacent to fusion boundary, low-alloy steel exhibited carbon-depleted region and coarsening of matrix grains. Between coarse grain and base material structure, low-alloy steel contained fine grain ferrite-pearlite aggregate. Adjacent to fusion boundary, buttering material consisted of Type-I and Type-II boundaries. Within buttering material close to fusion boundary, thin cluster of martensite was formed. Fusion boundary between buttering material-weld metal and weld metal-304LN stainless steel revealed unmixed zone. All joints failed within buttering material during in situ tensile testing. The fracture location was different for various joints with respect to fusion boundary, depending on variation in local microstructure. Highest bond strength with adequate ductility was obtained for the joint produced with 309L stainless steel-buttering material. High strength of this weld might be attributed to better extent of solid solution strengthening by alloying elements, diffused from low-alloy steel to buttering material.

  8. A simplified LBB evaluation procedure for austenitic and ferritic steel piping

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, R.M.; Wichman, K.R.

    1997-04-01

    The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

  9. Review of environmental effects on fatigue crack growth of austenitic stainless steels.

    Energy Technology Data Exchange (ETDEWEB)

    Shack, W. J.; Kassner, T. F.; Energy Technology

    1994-07-11

    Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry. The corrosion-fatigue data and curves in water were compared with the air line in Section XI of the ASME Code.

  10. Application of response surface methodology to maximize tensile strength and minimize interface hardness of friction welded dissimilar joints of austenitic stainless steel and copper alloy%响应面方法在奥氏体不锈钢与铜合金异种材料摩擦焊接头的抗拉强度最大化和界面硬度最小化中的应用

    Institute of Scientific and Technical Information of China (English)

    G.VAIRAMANI; T.SENTHIL KUMAR; S.MALARVIZHI; V.BALASUBRAMANIAN

    2013-01-01

    在奥氏体不锈钢与铜合金异种材料摩擦焊接过程中,采用响应面方法优化摩擦焊接工艺参数,以获得抗拉强度最大和界面硬度最小的焊接接头。采用三因素、五水平中心复合正交矩阵来确定实验条件。得到20个焊接接头,测定了焊接接头的抗拉强度和界面硬度。采用方差分析(ANOVA)方法来确定起显著作用的、主要的及相互作用的参数,使用回归分析得到经验关系模型。用设计专家软件构造响应图和等高线图来优化摩擦焊接工艺参数。用得到的经验关系模型可以有效地预测焊接接头的抗拉强度和界面硬度,其置信水平达95%。从形成的等高线图可以得到所需的摩擦焊接的最佳条件。%An attempt was made to optimize friction welding parameters to attain a minimum hardness at the interface and a maximum tensile strength of the dissimilar joints of AISI 304 austenitic stainless steel (ASS) and copper (Cu) alloy using response surface methodology (RSM). Three-factor, five-level central composite design matrix was used to specify experimental conditions. Twenty joints were fabricated using ASS and Cu alloy. Tensile strength and interface hardness were measured experimentally. Analysis of variance (ANOVA) method was used to find out significant main and interaction parameters and empirical relationships were developed using regression analysis. The friction welding parameters were optimized by constructing response graphs and contour plots using design expert software. The developed empirical relationships can be effectively used to predict tensile strength and interface hardness of friction welded ASS−Cu joints at 95%confidence level. The developed contour plots can be used to attain required level of optimum conditions to join ASS−Cu alloy by friction welding process.

  11. New Stainless Steel Alloys for Low Temperature Surface Hardening?

    DEFF Research Database (Denmark)

    Christiansen, Thomas Lundin; Dahl, Kristian Vinter; Somers, Marcel A. J.

    2015-01-01

    The present contribution showcases the possibility for developing new surface hardenable stainless steels containing strong nitride/carbide forming elements (SNCFE). Nitriding of the commercial alloys, austenitic A286, and ferritic AISI 409 illustrates the beneficial effect of having SNCFE present...... in the stainless steel alloys. The presented computational approach for alloy design enables “screening” of hundreds of thousands hypothetical alloy systems by use of Thermo-Calc. Promising compositions for new stainless steel alloys can be selected based on imposed criteria, i.e. facilitating easy selection...

  12. Carbon-content dependent effect of magnetic field on austenitic decomposition of steels

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Xiaoxue, E-mail: zhangxiaoxue1213@gmail.com [Key Laboratory for Anisotropy and Texture of Materials (MOE), Northeastern University, Shenyang 110004 (China); Laboratoire d' Etude des Microstructures et de Mecanique des Materiaux (LEM3), CNRS UMR 7239, University of Metz, 57045 Metz (France); Wang Shoujing, E-mail: wsj210725@yahoo.com.cn [Key Laboratory for Anisotropy and Texture of Materials (MOE), Northeastern University, Shenyang 110004 (China); Zhang Yudong, E-mail: yudong.zhang@univ-metz.fr [Laboratoire d' Etude des Microstructures et de Mecanique des Materiaux (LEM3), CNRS UMR 7239, University of Metz, 57045 Metz (France); Esling, Claude, E-mail: claude.esling@univ-metz.fr [Laboratoire d' Etude des Microstructures et de Mecanique des Materiaux (LEM3), CNRS UMR 7239, University of Metz, 57045 Metz (France); Zhao Xiang, E-mail: zhaox@mail.neu.edu.cn [Key Laboratory for Anisotropy and Texture of Materials (MOE), Northeastern University, Shenyang 110004 (China); Zuo Liang, E-mail: lzuo@mail.neu.edu.cn [Key Laboratory for Anisotropy and Texture of Materials (MOE), Northeastern University, Shenyang 110004 (China)

    2012-04-15

    The transformed microstructures of the high-purity Fe-0.12C alloy and Fe-0.36C alloy heat treated without and with a 12 T magnetic field have been investigated to explore the carbon-content dependent field effect on austenitic decomposition in steels. Results show that, the field-induced transformed morphology characteristics in different alloys differ from each other. In the Fe-0.12C alloy, the pearlite colonies are elongated along the field direction, and shaped by the chained and elongated proeutectoid ferrite grains in the field direction. However, in the Fe-0.36C alloy, the field mainly reduces the amount of Widmaenstatten ferrite and elongates the formed proeutectoid ferrite grains in the field direction. No clear field direction alignment is obtained. The magnetic field also demonstrates carbon-content dependent effect on the texture of the formed ferrite. It clearly enhances the Left-Pointing-Angle-Bracket 001 Right-Pointing-Angle-Bracket fiber of the ferrite in the transverse field direction in the Fe-0.36C alloy. This field effect is related to the crystal lattice distortion induced by carbon solution and this impact becomes stronger with the increase of the carbon content. For the Fe-0.12C alloy, this field effect is greatly reduced due to the reduced carbon oversaturation in ferrite and elevated formation temperature. The orientation relationships (ORs) between the pearlitic ferrite and the pearlitic cementite in both alloys are less affected by the magnetic field. No obvious changes in the either type of the appearing ORs and their number of occurrences are detected. - Highlights: Black-Right-Pointing-Pointer The carbon-content dependent field effect on austenitic decomposition is studied. Black-Right-Pointing-Pointer The field-induced morphology features vary with the carbon content. Black-Right-Pointing-Pointer The field effect on ferrite texture is more pronounced in high carbon content alloy. Black-Right-Pointing-Pointer Magnetic field hardly

  13. Effect of sub-zero cooling on microstructure and mechanical properties of a low alloyed austempered ductile iron

    Institute of S