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Sample records for austenitic alloys irradiated

  1. High post-irradiation ductility thermomechanical treatment for precipitation strengthened austenitic alloys

    DOEpatents

    Laidler, James J.; Borisch, Ronald R.; Korenko, Michael K.

    1982-01-01

    A method for improving the post-irradiation ductility is described which prises a solution heat treatment following which the materials are cold worked. They are included to demonstrate the beneficial effect of this treatment on the swelling resistance and the ductility of these austenitic precipitation hardenable alloys.

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

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Munro, B.

    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

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.; Grossbeck, M.L.

    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.

  6. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA. Revision 1

    SciTech Connect

    Huang, F.H.

    1992-02-01

    Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken.

  7. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA

    SciTech Connect

    Huang, F.H.

    1992-02-01

    Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken.

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

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.; Grossbeck, M.L.

    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.

  9. Austenitic alloy and reactor components made thereof

    DOEpatents

    Bates, John F.; Brager, Howard R.; Korenko, Michael K.

    1986-01-01

    An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

  10. Corrosion of austenitic alloys in aerated brines

    SciTech Connect

    Heidersbach, R.; Shi, A.; Sharp, S.

    1999-11-01

    This report discusses the results of corrosion exposures of three austenitic alloys--3l6L stainless steel, UNS N10276, and UNS N08367. Coupons of these alloys were suspended in a series of brines used for processing in the pharmaceutical industry. The effects of surface finish and welding processes on the corrosion behavior of these alloys were determined. The 316L coupons experienced corrosion in several environments, but the other alloys were unaffected during the one-month exposures of this investigation. Electropolishing the surfaces improved corrosion resistance.

  11. Investigation of joining techniques for advanced austenitic alloys

    SciTech Connect

    Lundin, C.D.; Qiao, C.Y.P.; Kikuchi, Y.; Shi, C.; Gill, T.P.S.

    1991-05-01

    Modified Alloys 316 and 800H, designed for high temperature service, have been developed at Oak Ridge National Laboratory. Assessment of the weldability of the advanced austenitic alloys has been conducted at the University of Tennessee. Four aspects of weldability of the advanced austenitic alloys were included in the investigation.

  12. Improved high temperature creep resistant austenitic alloy

    DOEpatents

    Maziasz, P.J.; Swindeman, R.W.; Goodwin, G.M.

    1988-05-13

    An improved austenitic alloy having in wt% 19-21 Cr, 30-35 Ni, 1.5-2.5 Mn, 2-3 Mo, 0.1-0.4 Si, 0.3-0.5 Ti, 0.1-0.3 Nb, 0.1-0.5 V, 0.001-0.005 P, 0.08-0.12 C, 0.01-0.03 N, 0.005-0.01 B and the balance iron that is further improved by annealing for up to 1 hour at 1150-1200/degree/C and then cold deforming 5-15%. The alloy exhibits dramatically improved creep rupture resistance and ductility at 700/degree/C. 2 figs.

  13. High temperature creep resistant austenitic alloy

    DOEpatents

    Maziasz, Philip J.; Swindeman, Robert W.; Goodwin, Gene M.

    1989-01-01

    An improved austenitic alloy having in wt % 19-21 Cr, 30-35 Ni, 1.5-2.5 Mn, 2-3 Mo, 0.1-0.4 Si, 0.3-0.5 Ti, 0.1-0.3 Nb, 0.1-0.5 V, 0.001-0.005 P, 0.08-0.12 C, 0.01-0.03 N, 0.005-0.01 B and the balance iron that is further improved by annealing for up to 1 hour at 1150.degree.-1200.degree. C. and then cold deforming 5-15 %. The alloy exhibits dramatically improved creep rupture resistance and ductility at 700.degree. C.

  14. Radiation resistant austenitic stainless steel alloys

    DOEpatents

    Maziasz, P.J.; Braski, D.N.; Rowcliffe, A.F.

    1987-02-11

    An austenitic stainless steel alloy, with improved resistance to radiation-induced swelling and helium embrittlement, and improved resistance to thermal creep at high temperatures, consisting essentially of, by weight percent: from 16 to 18% nickel; from 13 to 17% chromium; from 2 to 3% molybdenum; from 1.5 to 2.5% manganese; from 0.01 to 0.5% silicon; from 0.2 to 0.4% titanium; from 0.1 to 0.2% niobium; from 0.1 to 0.6% vanadium; from 0.06 to 0.12% carbon; from 0.01 to 0.03% nitrogen; from 0.03 to 0.08% phosphorus; from 0.005 to 0.01% boron; and the balance iron, and wherein the alloy may be thermomechanically treated to enhance physical and mechanical properties. 4 figs.

  15. Radiation resistant austenitic stainless steel alloys

    DOEpatents

    Maziasz, Philip J.; Braski, David N.; Rowcliffe, Arthur F.

    1989-01-01

    An austenitic stainless steel alloy, with improved resistance to radiation-induced swelling and helium embrittlement, and improved resistance to thermal creep at high temperatures, consisting essentially of, by weight percent: from 16 to 18% nickel; from 13 to 17% chromium; from 2 to 3% molybdenum; from 1.5 to 2.5% manganese; from 0.01 to 0.5% silicon; from 0.2 to 0.4% titanium; from 0.1 to 0.2% niobium; from 0.1 to 0.6% vanadium; from 0.06 to 0.12% carbon; from 0.01% to 0.03% nitrogen; from 0.03 to 0.08% phosphorus; from 0.005 to 0.01% boron; and the balance iron, and wherein the alloy may be thermomechanically treated to enhance physical and mechanical properties.

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

    SciTech Connect

    Hankin, G.L.; Faulkner, R.G.; Hamilton, M.L.; Garner, F.A.

    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.

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

    SciTech Connect

    Not Available

    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)

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

    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.

  1. Mn-Fe base and Mn-Cr-Fe base austenitic alloys

    DOEpatents

    Brager, Howard R.; Garner, Francis A.

    1987-01-01

    Manganese-iron base and manganese-chromium-iron base austenitic alloys designed to have resistance to neutron irradiation induced swelling and low activation have the following compositions (in weight percent): 20 to 40 Mn; up to about 15 Cr; about 0.4 to about 3.0 Si; an austenite stabilizing element selected from C and N, alone or in combination with each other, and in an amount effective to substantially stabilize the austenite phase, but less than about 0.7 C, and less than about 0.3 N; up to about 2.5 V; up to about 0.1 P; up to about 0.01 B; up to about 3.0 Al; up to about 0.5 Ni; up to about 2.0 W; up to about 1.0 Ti; up to about 1.0 Ta; and with the remainder of the alloy being essentially iron.

  2. Mn-Fe base and Mn-Cr-Fe base austenitic alloys

    DOEpatents

    Brager, Howard R.; Garner, Francis A.

    1987-09-01

    Manganese-iron base and manganese-chromium-iron base austenitic alloys designed to have resistance to neutron irradiation induced swelling and low activation have the following compositions (in weight percent): 20 to 40 Mn; up to about 15 Cr; about 0.4 to about 3.0 Si; an austenite stabilizing element selected from C and N, alone or in combination with each other, and in an amount effective to substantially stabilize the austenite phase, but less than about 0.7 C, and less than about 0.3 N; up to about 2.5 V; up to about 0.1 P; up to about 0.01 B; up to about 3.0 Al; up to about 0.5 Ni; up to about 2.0 W; up to about 1.0 Ti; up to about 1.0 Ta; and with the remainder of the alloy being essentially iron.

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

    SciTech Connect

    Grossbeck, M.L.; Gibson, L.T.; 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.

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

    SciTech Connect

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    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

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

    SciTech Connect

    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.

  6. The role of irradiated microstructure in the localized deformation of austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Jiao, Z.; Was, G. S.

    2010-12-01

    Localized deformation has emerged as a potential factor in irradiation-assisted stress corrosion cracking of austenitic stainless steels in LWR environments and the irradiated microstructure may be a critical factor in controlling the degree of localized deformation. Seven austenitic alloys with various compositions were irradiated using 2-3 MeV protons to doses of 1 and 5 dpa at 360 °C. The irradiated microstructure consisting of dislocation loops and voids was characterized using transmission electron microscopy. The degree of localized deformation was characterized using atomic force microscopy on the deformed samples after conducting constant extension rate tension tests to 1% and 3% strain in argon. Localized deformation was found to be dependent on the irradiated microstructure and to correlate with hardening originating from dislocation loops. Dislocation loops enhance the formation of dislocation channels and localize deformation into existing channels. On the contrast, voids mitigate the degree of localized deformation. The degree of localized deformation decreases with SFE with the exception of alloy B. Localized deformation was found to have similar dependence on SFE as loop density suggesting that SFE affects localized deformation by altering irradiated microstructure.

  7. Fracture toughness of irradiated wrought and cast austenitic stainless steels in BWR environment.

    SciTech Connect

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

    2007-01-01

    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. 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). We look at the results of a study of simulated light-water reactor coolants, material chemistry, and irradiation damage and their effects on the susceptibility to stress-corrosion cracking of various commercially available and laboratory-melted stainless steels.

  8. Entropic stabilization of austenite in shape memory alloys

    NASA Astrophysics Data System (ADS)

    Elliott, Ryan S.; Karls, Daniel S.

    2013-12-01

    Martensitic transformations (MTs) are the key phenomena responsible for the remarkable properties of Shape Memory Alloys (SMAs). Recent Density Functional Theory (DFT) electronic structure calculations have revealed that the austenite structure of many SMAs is a saddle-point of the material's potential energy landscape. Correspondingly, the austenite is unstable and thus unobservable at zero temperature. Thus, the observable high temperature austenite structure in many SMAs is entropically stabilized by nonlinear dynamic effects. This paper discusses the phenomenon of entropic stabilization of the austenite phase in SMAs and explicitly demonstrates it using Molecular Dynamics (MD) and a three-dimensional all-atom potential energy model whose equilibria crystal structures correspond to commonly observed SMA phases. A new technique is used to carefully select a model so that it is likely to lead to entropic stabilization of a B2 cubic austenite from a B19 orthorhombic martensite. This is accomplished by using a detailed branch-following and bifurcation (BFB) parametric study of the Morse pair potential binary alloy model. The results of the MD simulation clearly demonstrate the entropic stabilization of the B2 austenite phase at high temperature. Analysis of the dynamics of the B2 austenite phase indicates that its stabilization may be viewed as a result of individual atoms randomly visiting the B19 and αIrV phases (with only occasional visits to the B2 and L10 phases). This occurs without long-range correlations in such a way that each atom's time-average configuration corresponds to the B2 structure.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Chopra, O. K.; Roa, A. S.; Environmental Science Division; U.S. NRC

    2010-12-15

    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.

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

    SciTech Connect

    Chen, Y.; Yang, Yong; Huang, Yina; Allen, T.; Alexandreanu, B.; Natesan, K.

    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.

  12. The role of dislocation channeling in IASCC initiation of neutron irradiated austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale Jennings

    The objective of this study was to understand the role of dislocation channeling in the initiation of irradiation-assisted stress corrosion cracking (IASCC) of neutron irradiated austenitic stainless steel using a novel four-point bend test. Stainless steels used in this study were irradiated in the BOR-60 fast reactor at 320 °C, and included a commercial purity 304L stainless steel irradiated to 5.5, 10.2, and 47.5 dpa, and two high purity stainless steels, Fe-18Cr-12Ni and Fe-18Cr-25Ni, irradiated to ~10 dpa. The four-point bend test produced the same relative IASCC susceptibility as constant extension rate tensile (CERT) experiments performed on the same irradiated alloys in boiling water reactor normal water chemistry. The cracking susceptibility of the CP 304L alloy was high at all irradiation dose levels, enhanced by the presence of MnS inclusions in the alloy microstructure, which dissolve in the NWC environment. Dissolution of the MnS inclusion results in formation of an oxide cap that occludes the inclusion site, creating a crevice condition with a high propensity for crack initiation. Crack initiation at these locations was induced by stress concentration at the intersecting grain boundary, resulting from the intersection of a discontinuous dislocation channels (DC). Stress to initiate an IASCC crack decreased with dose due earlier DC initiation. The HP Fe-18Cr-12Ni alloy had low susceptibility to IASCC, while the high Ni alloy exhibited no cracking susceptibility. The difference in susceptibility among these conditions was attributed to the propensity for DCs to transmit across grain boundaries, which controls stress accumulation at DC -- grain boundary intersections.

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  15. Advanced austenitic alloys for fossil power systems. CRADA final report

    SciTech Connect

    Swindeman, R.W.; Cole, N.C.; Canonico, D.A.; Henry, J.F.

    1998-08-01

    In 1993, a Cooperative Research and Development Agreement (CRADA) was undertaken between Oak Ridge National Laboratory and ABB Combustion Engineering t examine advanced alloys for fossil power systems. Specifically, the use of advanced austenitic stainless steels for superheater/reheater construction in supercritical boilers was examined. The strength of cold-worked austenitic stainless steels was reviewed and compared to the strength and ductility of advanced austenitic stainless steels. The advanced stainless steels were found to retain their strength to very long times at temperatures where cold-worked standard grades of austenitic stainless steels became weak. Further, the steels exhibited better long-time stability than the stabilized 300 series stainless steels in either the annealed or cold worked conditions. Type 304H mill-annealed tubing was provided to ORNL for testing of base metal and butt welds. The tubing was found to fall within range of expected strength for 304H stainless steel. The composite 304/308 stainless steel was found to be stronger than typical for the weldment. Boiler tubing was removed from a commercial boiler for replacement by newer steels, but restraints imposed by the boiler owners did not permit the installation of the advanced steels, so a standard 32 stainless steel was used as a replacement. The T91 removed from the boiler was characterized.

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

    SciTech Connect

    Hankin, G.L.; Faulkner, R.G.; Hamilton, M.L.; Garner, F.A.

    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.

  17. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    SciTech Connect

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-02-24

    We investigated dynamics of deformation localization and dislocation channel formation in situ in a neutron irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Channel formation was observed at 70% of the formal tensile yield stress for both alloys. It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the yield stress, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young modulus) in the channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in soft grains with a high Schmid factor located near stiff grains with high elastic stiffness. Moreover, the spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. Finally, it was shown that in the AISI 304 steel, channels were twin-free in grains oriented close to [001] and [101] of standard unit triangle; [111]-grains and grains oriented close to Schmid factor maximum contained deformation twins.

  18. Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

    DOE PAGES

    Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T.

    2015-02-24

    We investigated dynamics of deformation localization and dislocation channel formation in situ in a neutron irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Channel formation was observed at 70% of the formal tensile yield stress for both alloys. It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the yield stress, channels often formed near the middle of the grain boundary. For amore » single grain, the role of elastic stiffness value (Young modulus) in the channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in soft grains with a high Schmid factor located near stiff grains with high elastic stiffness. Moreover, the spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. Finally, it was shown that in the AISI 304 steel, channels were twin-free in grains oriented close to [001] and [101] of standard unit triangle; [111]-grains and grains oriented close to Schmid factor maximum contained deformation twins.« less

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

    SciTech Connect

    Ashdown, B.G.

    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)

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

    SciTech Connect

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

    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.

  1. Effects of focused ion beam milling on austenite stability in ferrous alloys

    SciTech Connect

    Knipling, K.E.; Rowenhorst, D.J.; Fonda, R.W.; Spanos, G.

    2010-01-15

    The susceptibility of fcc austenite to transform to bcc during focused ion beam milling was studied in three commercial stainless steels. The alloys investigated, in order of increasing austenite stability, were: (i) a model maraging steel, Sandvik 1RK91; (ii) an AISI 304 austenitic stainless steel; and (iii) AL-6XN, a super-austenitic stainless steel. Small trenches were milled across multiple austenite grains in each alloy using a 30 kV Ga{sup +} ion beam at normal incidence to the specimen surface. The ion beam dose was controlled by varying the trench depth and the beam current. The factors influencing the transformation of fcc austenite to bcc (listed in order of decreasing influence) were found to be: (i) alloy composition (i.e., austenite stability), (ii) ion beam dose (or trench depth), and (iii) crystallographic orientation of the austenite grains. The ion beam current had a negligible influence on the FIB-induced transformation of austenite in these alloys.

  2. Modeling of radiation-induced segregation in austenitic Fe-Cr-Ni alloys

    NASA Astrophysics Data System (ADS)

    Allen, Todd Randall

    Radiation-induced segregation (RIS) was studied in Fe-Cr-Ni alloys irradiated with protons to better understand the mechanisms causing changes in grain boundary chemistry and to improve the ability to predict RIS in austenitic Fe-Cr-Ni alloys. Ni-18Cr, Ni-18Cr-9Fe, Ni-18Cr-0.08P, and Fe-20Cr-9Fe were irradiated with 3.2MeV protons at temperatures from 200sp°C to 500sp°C and to doses from 0.1 to 3 dpa. Grain boundary chemistry was measured using both Auger electron spectroscopy (AES) and scanning transmission electron microscopy with energy dispersive x-ray spectroscopy (STEM/EDS). The significant driving mechanism far segregation in Fe-Cr-Ni alloys is shown to be the inverse Kirkendall (IK) mechanism, specifically the coupling between alloying elements and the vacancy flux. The inclusion of interstitial binding effects to RIS models results in poor agreement between model predictions and segregation measurements, severely overpredicting the measured Ni enrichment and Fe depletion. Grain boundary segregation is unique for each bulk alloy composition in that the amount and the rate of segregation differs for alloys irradiated under the same conditions. Kinetic parameters must be known for each alloy to accurately predict segregation, but the kinetic parameters in Fe-Cr-Ni alloys at low temperature are not well studied. Additionally, short range ordering interactions are important in determining the segregation in all Fe-Cr-Ni alloys. Ordering enthalpies must be included in RIS models to correctly describe the segregation process. Therefore, to develop a predictive RIS model, a method for calculating diffusivities from the bulk composition that includes ordering enthalpies was developed. The Perks (IK) model has been modified to account for composition dependent segregation kinetics by calculating the migration energy using pair interaction potentials, ordering enthalpies, and the local concentration. Based on segregation measurements from seven different alloys

  3. Influence of displacement damage on deuterium and helium retention in austenitic and ferritic-martensitic alloys considered for ADS service

    NASA Astrophysics Data System (ADS)

    Voyevodin, V. N.; Karpov, S. A.; Kopanets, I. E.; Ruzhytskyi, V. V.; Tolstolutskaya, G. D.; Garner, F. A.

    2016-01-01

    The behavior of ion-implanted hydrogen (deuterium) and helium in austenitic 18Cr10NiTi stainless steel, EI-852 ferritic steel and ferritic/martensitic steel EP-450 and their interaction with displacement damage were investigated. Energetic argon irradiation was used to produce displacement damage and bubble formation to simulate nuclear power environments. The influence of damage morphology and the features of radiation-induced defects on deuterium and helium trapping in structural alloys was studied using ion implantation, the nuclear reaction D(3He,p)4He, thermal desorption spectrometry and transmission electron microscopy. It was found in the case of helium irradiation that various kinds of helium-radiation defect complexes are formed in the implanted layer that lead to a more complicated spectra of thermal desorption. Additional small changes in the helium spectra after irradiation with argon ions to a dose of ≤25 dpa show that the binding energy of helium with these traps is weakly dependent on the displacement damage. It was established that retention of deuterium in ferritic and ferritic-martensitic alloys is three times less than in austenitic steel at damage of ˜1 dpa. The retention of deuterium in steels is strongly enhanced by presence of radiation damages created by argon ion irradiation, with a shift in the hydrogen release temperature interval of 200 K to higher temperature. At elevated temperatures of irradiation the efficiency of deuterium trapping is reduced by two orders of magnitude.

  4. Crack growth rates of irradiated austenitic stainless steel weld heat affected zone in BWR environments.

    SciTech Connect

    Chopra, O. K.; Alexandreanu, B.; Gruber, E. E.; Daum, R. S.; Shack, W. J.; Energy Technology

    2006-01-31

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of reactor pressure vessels because of their superior fracture toughness. However, exposure to high levels of neutron irradiation for extended periods can exacerbate the corrosion fatigue and stress corrosion cracking (SCC) behavior of these steels by affecting the material microchemistry, material microstructure, and water chemistry. Experimental data are presented on crack growth rates of the heat affected zone (HAZ) in Types 304L and 304 SS weld specimens before and after they were irradiated to a fluence of 5.0 x 10{sup 20} n/cm{sup 2} (E > 1 MeV) ({approx} 0.75 dpa) at {approx}288 C. Crack growth tests were conducted under cycling loading and long hold time trapezoidal loading in simulated boiling water reactor environments on Type 304L SS HAZ of the H5 weld from the Grand Gulf reactor core shroud and on Type 304 SS HAZ of a laboratory-prepared weld. The effects of material composition, irradiation, and water chemistry on growth rates are discussed.

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

    NASA Astrophysics Data System (ADS)

    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.

  6. Laser beam surface melting of high alloy austenitic stainless steel

    SciTech Connect

    Woollin, P.

    1996-12-31

    The welding of high alloy austenitic stainless steels is generally accompanied by a substantial reduction in pitting corrosion resistance relative to the parent, due to microsegregation of Mo and Cr. This prevents the exploitation of the full potential of these steels. Processing to achieve remelting and rapid solidification offers a means of reducing microsegregation levels and improving corrosion resistance. Surface melting of parent UNS S31254 steel by laser beam has been demonstrated as a successful means of producing fine, as-solidified structures with pitting resistance similar to that of the parent, provided that an appropriate minimum beam travel speed is exceeded. The use of N{sub 2} laser trail gas increased the pitting resistance of the surface melted layer. Application of the technique to gas tungsten arc (GTA) melt runs has shown the ability to raise the pitting resistance significantly. Indeed, the use of optimized beam conditions, N{sub 2} trail gas and appropriate surface preparation prior to laser treatment increased the pitting resistance of GTA melt runs to a level approaching that of the parent material.

  7. Copper modified austenitic stainless steel alloys with improved high temperature creep resistance

    DOEpatents

    Swindeman, R.W.; Maziasz, P.J.

    1987-04-28

    An improved austenitic stainless steel that incorporates copper into a base Fe-Ni-Cr alloy having minor alloying substituents of Mo, Mn, Si, T, Nb, V, C, N, P, B which exhibits significant improvement in high temperature creep resistance over previous steels. 3 figs.

  8. The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part I. Ductility and fracture toughness

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

    The radiation swelling effect on the fracture properties of irradiated austenitic steels under static loading has been studied and analyzed from the mechanical and physical viewpoints. Experimental data on the stress-strain curves, fracture strain, fracture toughness and fracture mechanisms have been represented for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various swelling. Some phenomena in mechanical behaviour of irradiated austenitic steels have been revealed and explained as follows: a sharp decrease of fracture toughness with swelling growth; untypical large increase of fracture toughness with decrease of the test temperature; some increase of fracture toughness after preliminary cyclic loading. Role of channel deformation and channel fracture has been clarified in the properties of irradiated austenitic steel and different tendencies to channel deformation have been shown and explained for the same austenitic steel irradiated at different temperatures and neutron doses.

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

    SciTech Connect

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

    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.

  10. Effect of heavy ion irradiation on microstructural evolution in CF8 cast austenitic stainless steel

    DOE PAGES

    Chen, Wei-Ying; Li, Meimei; Kirk, Marquis A.; ...

    2015-08-21

    The microstructural evolution in ferrite and austenitic in cast austenitic stainless steel (CASS) CF8, as received or thermally aged at 400 °C for 10,000 h, was followed under TEM with in situ irradiation of 1 MeV Kr ions at 300 and 350 °C to a fluence of 1.9 × 1015 ions/cm2 (~3 dpa) at the IVEM-Tandem Facility. For the unaged CF8, the irradiation-induced dislocation loops appeared at a much lower dose in the austenite than in the ferrite. At the end dose, the austenite formed a well-developed dislocation network microstructure, while the ferrite exhibited an extended dislocation structure as linemore » segments. Compared to the unaged CF8, the aged specimen appeared to have lower rate of damage accumulation. The rate of microstructural evolution under irradiation in the ferrite was significantly lower in the aged specimen than in the unaged. Finally, we attributed this difference to the different initial microstructures in the unaged and aged specimens, which implies that thermal aging and irradiation are not independent but interconnected damage processes.« less

  11. Effect of heavy ion irradiation on microstructural evolution in CF8 cast austenitic stainless steel

    SciTech Connect

    Chen, Wei-Ying; Li, Meimei; Kirk, Marquis A.; Baldo, Peter M.; Lian, Tiangan

    2015-08-21

    The microstructural evolution in ferrite and austenitic in cast austenitic stainless steel (CASS) CF8, as received or thermally aged at 400 °C for 10,000 h, was followed under TEM with in situ irradiation of 1 MeV Kr ions at 300 and 350 °C to a fluence of 1.9 × 1015 ions/cm2 (~3 dpa) at the IVEM-Tandem Facility. For the unaged CF8, the irradiation-induced dislocation loops appeared at a much lower dose in the austenite than in the ferrite. At the end dose, the austenite formed a well-developed dislocation network microstructure, while the ferrite exhibited an extended dislocation structure as line segments. Compared to the unaged CF8, the aged specimen appeared to have lower rate of damage accumulation. The rate of microstructural evolution under irradiation in the ferrite was significantly lower in the aged specimen than in the unaged. Finally, we attributed this difference to the different initial microstructures in the unaged and aged specimens, which implies that thermal aging and irradiation are not independent but interconnected damage processes.

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

    NASA Astrophysics Data System (ADS)

    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.

  13. Precipitation hardening austenitic superalloys

    DOEpatents

    Korenko, Michael K.

    1985-01-01

    Precipitation hardening, austenitic type superalloys are described. These alloys contain 0.5 to 1.5 weight percent silicon in combination with about 0.05 to 0.5 weight percent of a post irradiation ductility enhancing agent selected from the group of hafnium, yttrium, lanthanum and scandium, alone or in combination with each other. In addition, when hafnium or yttrium are selected, reductions in irradiation induced swelling have been noted.

  14. Effect of Alloying Element Partition in Pearlite on the Growth of Austenite in High-Carbon Low Alloy Steel

    NASA Astrophysics Data System (ADS)

    Yang, Z. N.; Xia, Y.; Enomoto, M.; Zhang, C.; Yang, Z. G.

    2016-03-01

    The growth of austenite from pearlite in high-carbon low alloy steel occurs with and without alloy element redistribution depending on the amount of superheating above the eutectoid temperature. The transition temperature of austenite growth (denoted PNTT) is calculated as a function of pearlite transformation temperature and subsequent holding time, which affect the degree of partitioning in pearlite, using experimental partition coefficients k θ/ α of Mn, Cr, Co, Si, and Ni reported in the literature. PNTT is the highest in Cr-containing alloys which have the largest k θ/ α in pearlite. Post-transformation aging, usually accompanied by cementite spheroidization, leads to a marked increase of PNTT in Mn and Cr alloys. PNTT of Ni alloy does not depend on pearlite transformation temperature because practically the formation of partitioned pearlite is severely limited in this alloy for kinetic reasons. Above PNTT, austenite growth occurs fast initially, but slows down in the order of ten seconds when the ferrite disappears, and the remaining small carbide particles dissolve very slowly under the control of alloy element diffusion.

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

    SciTech Connect

    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

  16. Austenitic stainless steel alloys having improved resistance to fast neutron-induced swelling

    DOEpatents

    Bloom, Everett E.; Stiegler, James O.; Rowcliffe, Arthur F.; Leitnaker, James M.

    1979-01-01

    The present invention is based on the discovery that radiation-induced voids which occur during fast neutron irradiation can be controlled by small but effective additions of titanium and silicon. The void-suppressing effect of these metals in combination is demonstrated and particularly apparent in austenitic stainless steels.

  17. Austenitic stainless steel alloys having improved resistance to fast neutron-induced swelling

    DOEpatents

    Bloom, Everett E.; Stiegler, James O.; Rowcliffe, Arthur F.; Leitnaker, James M.

    1977-03-08

    The present invention is based on the discovery that radiation-induced voids which occur during fast neutron irradiation can be controlled by small but effective additions of titanium and silicon. The void-suppressing effect of these metals in combination is demonstrated and particularly apparent in austenitic stainless steels.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Stubbins, James; Heuser, Brent; Robertson, Ian; Sehitoglu, Huseyin; Sofronis, Petros; Gewirth, Andrew

    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

  20. Evolution of microstructure after irradiation creep in several austenitic steels irradiated up to 120 dpa at 320 °C

    NASA Astrophysics Data System (ADS)

    Renault-Laborne, A.; Garnier, J.; Malaplate, J.; Gavoille, P.; Sefta, F.; Tanguy, B.

    2016-07-01

    Irradiation creep was investigated in different austenitic steels. Pressurized tubes with stresses of 127-220 MPa were irradiated in BOR-60 at 320 °C to 120 dpa. Creep behavior was dependent on both chemical composition and metallurgical state of steels. Different steels irradiated with and without stress were examined by TEM. Without stress, the irradiation produced high densities of dislocation lines and Frank loops and, depending on the type of steels, precipitates. Stress induced an increase of the precipitate mean size and density and, for some grades, an increase of the mean loop size and a decrease of their density. An anisotropy of Frank loop density or size induced by stress was not observed systematically. Dislocation line microstructure seems not to be different between the stressed and unstressed specimens. No cavities were detectable in these specimens. By comparing with the data from this work, the main irradiation creep models are discussed.

  1. The irradiation effects on zirconium alloys

    NASA Astrophysics Data System (ADS)

    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.

  2. Machining and Phase Transformation Response of Room-Temperature Austenitic NiTi Shape Memory Alloy

    NASA Astrophysics Data System (ADS)

    Kaynak, Yusuf

    2014-09-01

    This experimental work reports the results of a study addressing tool wear, surface topography, and x-ray diffraction analysis for the finish cutting process of room-temperature austenitic NiTi alloy. Turning operation of NiTi alloy was conducted under dry, minimum quantity lubrication (MQL) and cryogenic cooling conditions at various cutting speeds. Findings revealed that cryogenic machining substantially reduced tool wear and improved surface topography and quality of the finished parts in comparison with the other two approaches. Phase transformation on the surface of work material was not observed after dry and MQL machining, but B19' martensite phase was found on the surface of cryogenically machined samples.

  3. STUDY OF GRAIN BOUNDARY CHARACTER ALONG INTERGRANULAR STRESS CORROSION CRACK PATHS IN AUSTENITIC ALLOYS

    SciTech Connect

    Guertsman, Valery Y.; Bruemmer, Stephen M.

    2001-05-25

    Samples of austenitic stainless alloys were examined by means of scanning and transmission electron microscopy. Misorientations were measured by electron backscattered diffraction. Grain boundary distributions were analyzed with special emphasis on the grain boundary character along intergranular stress-corrosion cracks and at crack arrest points. It was established that only coherent twin S3 boundaries could be considered as "special" ones with regard to crack resistance. However, it is possible that twin interactions with random grain boundaries may inhibit crack propagation. The results suggest that other factors besides geometrical ones play an important role in the intergranular stress-corrosion cracking of commercial alloys.

  4. Irradiation creep of dispersion strengthened copper alloy

    SciTech Connect

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  6. Deformation and cracking of irradiated austenitic stainless steels

    SciTech Connect

    Carter, R.D.; Atzmon, M.; Was, G.S.

    1995-12-31

    Samples of proton-irradiated 304L stainless steel were deformed by constant extension rate tensile tests at strain rates of 3 {times} 10{sup {minus}7} s{sup {minus}1} and 3 {times} 10{sup {minus}8} s{sup {minus}1} to strains of up to 10% at 288--350 C in argon. Minor cracking was observed in and around spinel inclusions in the material, however no intergranular cracking of the type observed in water environments was found. Thus intergranular cracking cannot occur by a radiation hardening mechanism alone. The microstructures that resulted from irradiation and deformation were characterized using electron microscopy. Surface slip band formation is observed on one or two {l_brace}111{r_brace} slip systems in each grain. The slip bands correspond to dislocation channels in the material as identified by transmission electron microscopy. The channels form by activation of grain-boundary dislocation sources, with the emitted dislocations sweeping through the grain interior to the opposing rain boundaries. During this process, the dislocations remove the radiation-produced defects. Slip band and dislocation channel densities increase with increasing strain in the samples. These results are used to interpret stress corrosion cracking behavior in this material.

  7. Cracking behavior of thermally aged and irradiated CF-8 cast austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Chen, Y.; Alexandreanu, B.; Chen, W.-Y.; Natesan, K.; Li, Z.; Yang, Y.; Rao, A. S.

    2015-11-01

    To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8.

  8. High strength nickel-chromium-iron austenitic alloy

    DOEpatents

    Gibson, Robert C.; Korenko, Michael K.

    1980-01-01

    A solid solution strengthened Ni-Cr-Fe alloy capable of retaining its strength at high temperatures and consisting essentially of 42 to 48% nickel, 11 to 13% chromium, 2.6 to 3.4% niobium, 0.2 to 1.2% silicon, 0.5 to 1.5% vanadium, 2.6 to 3.4% molybdenum, 0.1 to 0.3% aluminum, 0.1 to 0.3% titanium, 0.02 to 0.05% carbon, 0.002 to 0.015% boron, up to 0.06 zirconium, and the balance iron. After solution annealing at 1038.degree. C. for one hour, the alloy, when heated to a temperature of 650.degree. C., has a 2% yield strength of 307 MPa, an ultimate tensile strength of 513 MPa and a rupture strength of as high as 400 MPa after 100 hours.

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

    NASA Astrophysics Data System (ADS)

    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.

  10. The Austenitizing Behavior of a Low Alloy Steel

    DTIC Science & Technology

    1980-05-01

    to differences in melting and solidification between the two types of steels ; ESR exhibiting a more uniform dendritic grain structure and possibly...AD-AoVtS^ MCHNICAi; LIBRARY AD TECHNICAL REPORT ARLCB-TR-80013 TIE AUSIENITIZING BEHAVIOR OF A LDW ALLDY STEEL P, A. Thornton May 1980 US...Behavior of a Low Alloy Steel 5. TYPE OF REPORT 4 PERIOD COVERED 6. PERFORMING ORG. REPORT NUMBER 7. AUTHORf*; 8. CONTRACT OR GRANT NUMBERf*) Peter A

  11. Ultrafine-Grained Structure of Fe-Ni-C Austenitic Alloy Formed by Phase Hardening.

    PubMed

    Danilchenko, Vitalij

    2016-12-01

    The X-ray and magnetometry methods were used to study α-γ transformation mechanisms on heating quenched Fe-22.7 wt.% Ni-0.58 wt.% С alloy. Variation of heating rate within 0.03-80 K/min allowed one to switch from diffusive to non-diffusive mechanism of the α-γ transformation. Heating up primary austenitic single crystal specimen at a rate of less than 1.0-0.5 K/min has led to formation of aggregate of grains with different orientation and chemical composition in the reverted austenite. Significant fraction of these grains was determined to have sizes within nanoscale range.

  12. Carburization of austenitic alloys by gaseous impurities in helium

    SciTech Connect

    Lai, G.Y.; Johnson, W.R.

    1980-03-01

    The carburization behavior of Alloy 800H, Inconel Alloy 617 and Hastelloy Alloy X in helium containing various amounts of H/sub 2/, CO, CH/sub 4/, H/sub 2/O and CO/sub 2/ was studied. Corrosion tests were conducted in a temperature range from 649 to 1000/sup 0/C (1200 to 1832/sup 0/F) for exposure time up to 10,000 h. Four different helium environments, identified as A, B, C, and D, were investigated. Concentrations of gaseous impurities were 1500 ..mu..atm H/sub 2/, 450 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and 50 ..mu..atm H/sub 2/O for Environment A; 200 ..mu..atm H/sub 2/, 100 ..mu..atm CO, 20 ..mu..atm CH/sub 4/, 50 ..mu..atm H/sub 2/O and 5 ..mu..atm CO/sub 2/ for Environment B; 500 ..mu..atm H/sub 2/, 50 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and < 0.5 ..mu..atm H/sub 2/O for Environment C; and 500 ..mu..atm H/sub 2/, 50 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and 1.5 ..mu..atm H/sub 2/O for Environment D. Environments A and B were characteristic of high-oxygen potential, while C and D were characteristic of low-oxygen potential. The results showed that the carburization kinetics in low-oxygen potential environments (C and D) were significantly higher, approximately an order of magnitude higher at high temperatures, than those in high-oxygen potential environments (A and B) for all three alloys. Thermodynamic analyses indicated no significant differences in the thermodynamic carburization potential between low- and high-oxygen potential environments. It is thus believed that the enhanced carburization kinetics observed in the low-oxygen potential environments were related to kinetic effects. A qualitatively mechanistic model was proposed to explain the enhanced kinetics. The present results further suggest that controlling the oxygen potential of the service environment can be an effective means of reducing carburization of alloys.

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

    SciTech Connect

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

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  15. Irradiation creep of vanadium-base alloys

    SciTech Connect

    Tsai, H.; Billone, M.C.; Strain, R.V.; Smith, D.L.; Matsui, H.

    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.

  16. Influence of Hold Time on Creep-Fatigue Behavior of an Advanced Austenitic Alloy

    SciTech Connect

    Mark Carroll; Laura Carroll

    2011-09-01

    An advanced austenitic alloy, HT-UPS (high temperature-ultrafine precipitate strengthened), is a candidate material for the structural components of fast reactors and energy-conversion systems. HT-UPS provides improved creep resistance through a composition based on 316 stainless steel (SS) with additions of Ti and Nb to form nano-scale MC precipitates in the austenitic matrix. The low cycle fatigue and creep-fatigue behavior of a HT-UPS alloy has been investigated at 650 C, 1.0% total strain, and an R ratio of -1 with hold times as long as 9000 sec at peak tensile strain. The cyclic deformation response of HT-UPS is compared to that of 316 SS. The cycles to failure are similar, despite differences in peak stress profiles and the deformed microstructures. Cracking in both alloys is transgranular (initiation and propagation) in the case of continuous cycle fatigue, while the primary cracks also propagate transgranularly during creep-fatigue cycling. Internal grain boundary damage as a result of the tensile hold is present in the form of fine cracks for hold times of 3600 sec and longer and substantially more internal cracks are visible in 316 SS than HT-UPS. The dislocation substructures observed in the deformed material are different. An equiaxed cellular structure is observed in 316 SS, whereas tangles of dislocations are present at the nanoscale MC precipitates in HT-UPS and no cellular substructure is observed.

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

    SciTech Connect

    Hashimoto, N.; Robertson, J.P.; Grossbeck, M.L.; Rowcliffe, A.F.; Wakai, E.

    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.

  18. Irradiation testing of 316L(N)-IG austenitic stainless steel for ITER

    NASA Astrophysics Data System (ADS)

    van Osch, E. V.; Horsten, M. G.; de Vries, M. I.

    1998-10-01

    In the frame work of the European Fusion Technology Programme and the International Thermonuclear Experimental Reactor (ITER), ECN is investigating the irradiation behaviour of the structural materials for ITER. The main structural material for ITER is austenitic stainless steel Type 316L(N)-IG. The operating temperatures of (parts of) the components are envisaged to range between 350 and 700 K. A significant part of the dose-temperature domain of irradiation conditions relevant for ITER has already been explored, there is, however, very little data at about 600 K. Available data tend to indicate a maximum in the degradation of the mechanical properties after irradiation at this temperature, e.g. a minimum in ductility and a maximum of hardening. Therefore an irradiation program for plate material 316L(N)-IG, its Electron Beam (EB) weld and Tungsten Inert Gas (TIG) weld metal, and also including Hot Isostatically Pressed (HIP) 316L(N) powder and solid-solid joints, was set up in 1995. Irradiations have been carried out in the High Flux Reactor (HFR) in Petten at a temperature of 600 K, at dose levels from 1 to 10 dpa. The paper presents the currently available post-irradiation test results. Next to tensile and fracture toughness data on plate, EB and TIG welds, first results of powder HIP material are included.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  1. Orientation relationships between M2C carbide and the austenite matrix in an Fe-Mn-AI-Mo-C alloy

    NASA Astrophysics Data System (ADS)

    Peng, Shang-Wen; Chou, Chang-Pin

    1993-05-01

    M2C carbides were observed to precipitate within the austenite matrix of an Fe-24.6Mn-6.6Al-3. IMo-1.0C alloy after quenching from 1200 °C and aging at 750 °C. By means of transmission electron microscopy (TEM) and diffraction techniques, the orientation relationships between M2C (p) and the austenite (γ) matrix were determined to be: (0001)p//(111)γ, (11- bar 20)p// (1 bar 10)γ, ( bar 1100)p//(11 bar 2)γ. M2C carbide has been reported by many researchers to precipitate from the ferrite matrix or along austenite/ferrite boundaries in alloy steels containing Mo. However, little information concerning the formation of M2C in the austenite matrix has been provided. This investigation presents the first evidence for the existence of M2C carbide wholly within the austenite matrix and its relationship to the austenite. The energy-dispersive spectrometry (EDS) analyses were performed on M2C carbides, and the results indicate that the solubility of the M2C carbide for foreign atoms other than Mo is very limited.

  2. Investigation of the effect of cyclic laser heating for creating dispersed structures in the austenitic-martensitic alloys based on Fe-Cr-Ni system

    NASA Astrophysics Data System (ADS)

    Andreev, A. O.; Mironov, V. D.; Petrovskii, V. N.; Orlov, A. V.; Libman, M. A.

    2016-09-01

    The effect of cyclic laser heating on the formation of the austenite structure in the austenitic-martensitic alloys based on Fe-Cr-Ni system is investigated. It is shown that under the influence of ultra-fast laser heating on the martensite, which was formed during plastic deformation, the reverse martensitic transformation occurs, and austenite with high strength characteristics is formed. Repeated and multiple laser heating effectively grinds areas of austenite to a size close to the large nanoparticles. There is an additional increase in the strength characteristics of austenite as a result of this fragmentation.

  3. Manufacture of Alumina-Forming Austenitic Steel Alloys by Conventional Casting and Hot-Working Methods

    SciTech Connect

    Brady, M.P.; Yamamoto, Y.; Magee, J.H.

    2009-03-10

    Oak Ridge National Laboratory (ORNL) and Carpenter Technology Corporation (CarTech) 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 scale up of developmental ORNL alumina-forming austenitic (AFA) stainless steels by conventional casting and rolling techniques. CarTech successfully vacuum melted 301b heats of four AFA alloy compositions in the range of Fe-(20-25)Ni-(12-14)Cr-(3-4)Al-(l-2.5)Nb wt.% base. Conventional hot/cold rolling was used to produce 0.5-inch thick plate and 0.1-inch thick sheet product. ORNL subsequently successfully rolled the 0.1-inch sheet to 4 mil thick foil. Long-term oxidation studies of the plate form material were initiated at 650, 700, and 800 C in air with 10 volume percent water vapor. Preliminary results indicated that the alloys exhibit comparable (good) oxidation resistance to ORNL laboratory scale AFA alloy arc casting previously evaluated. The sheet and foil material will be used in ongoing evaluation efforts for oxidation and creep resistance under related CRADAs with two gas turbine engine manufacturers. This work will be directed to evaluation of AFA alloys for use in gas turbine recuperators to permit higher-temperature operating conditions for improved efficiencies and reduced environmental emissions. AFA alloy properties to date have been obtained from small laboratory scale arc-castings made at ORNL. The goal of the ORNL-CarTech CRADA was to establish the viability for producing plate, sheet and foil of the AFA alloys by conventional casting and hot working approaches as a first step towards scale up and commercialization of the AFA alloys. The AFA alloy produced under this effort will then be evaluated in related CRADAs with two gas turbine engine manufacturers for gas turbine recuperator applications.

  4. Fatigue and Creep-Fatigue Deformation of an Ultra-Fine Precipitate Strengthened Advanced Austenitic Alloy

    SciTech Connect

    M.C. Carroll; L.J. Carroll

    2012-10-01

    An advanced austenitic alloy, HT-UPS (high-temperature ultrafine-precipitation-strengthened), has been identified as an ideal candidate material for the structural components of fast reactors and energy-conversion systems. HT-UPS alloys demonstrate improved creep resistance relative to 316 stainless steel (SS) through additions of Ti and Nb, which precipitate to form a widespread dispersion of stable nanoscale metallic carbide (MC) particles in the austenitic matrix. The low-cycle fatigue and creep-fatigue behavior of an HT-UPS alloy have been investigated at 650 °C and a 1.0% total strain, with an R-ratio of -1 and hold times at peak tensile strain as long as 150 min. The cyclic deformation response of HT-UPS is directly compared to that of standard 316 SS. The measured values for total cycles to failure are similar, despite differences in peak stress profiles and in qualitative observations of the deformed microstructures. Crack propagation is primarily transgranular in fatigue and creep-fatigue of both alloys at the investigated conditions. Internal grain boundary damage in the form of fine cracks resulting from the tensile hold is present for hold times of 60 min and longer, and substantially more internal cracks are quantifiable in 316 SS than in HT-UPS. The dislocation substructures observed in the deformed material differ significantly; an equiaxed cellular structure is observed in 316 SS, whereas in HT-UPS the microstructure takes the form of widespread and relatively homogenous tangles of dislocations pinned by the nanoscale MC precipitates. The significant effect of the fine distribution of precipitates on observed fatigue and creep-fatigue response is described in three distinct behavioral regions as it evolves with continued cycling.

  5. IASCC susceptibility of irradiated austenitic stainless steel under very low dissolved oxygen

    SciTech Connect

    Kodama, Mitsuhiro; Katsura, Ryoei; Morisawa, Junichiro; Nishimura, Seiji; Suzuki, Shunichi; Takamori, Kenro; Shima, Seishi; Kato, Takahiko

    1995-12-31

    Slow strain rate tests of Type 304 stainless steel (SS) irradiated to 1.3 {times} 10{sup 26} n/m{sup 2} (E>1MeV) were conducted in high-temperature water and argon gas environment to discuss irradiation-assisted stress corrosion cracking (IASCC) mechanism with respect to the dissolved oxygen (DO) effect. IASCC susceptibility of Type 304 SS decreased with decreasing DO. However, IASCC was not mitigated completely in the hydrogen injected water. And IG fracture was not observed in the case of argon gas environment. These results indicated that the high-temperature aqueous environment was an indispensable condition for the occurrence of IASCC. Moreover, lowering DO(<1ppb) did not necessarily eliminate IASCC susceptibility when austenitic stainless steel was irradiated to high neutron fluence. By considering H{sub 2}O{sub 2} formed by {gamma}-ray irradiation, IASCC at very low DO could not be explained by an active path corrosion model. At high DO, IASCG would be affected by the active path corrosion of radiation-induced chromium depletion. However, at very low DO, the possibility that IASCC would be affected by other mechanisms such as hydrogen embrittlement was suggested.

  6. Development of 1100 °C Capable Alumina-Forming Austenitic Alloys

    DOE PAGES

    Brady, M. P.; Muralidharan, G.; Yamamoto, Y.; ...

    2016-11-18

    Recently dalumina-forming austenitic (AFA) alloys based on ~12–32 weight % (wt%) Ni have been developed and offer an attractive combination of oxidation resistance and creep resistance at relatively low alloy cost. But, they exhibit a transition to internal oxidation and nitridation of Al above ~750–950 °C depending on composition and exposure environment. In order to identify AFA compositions capable of higher-temperature operation for applications such as ethylene cracking, the oxidation behavior of a series of developmental, as-cast nominal Fe–(25–45)Ni–(10–25)Cr–(4–5)Al–1Si–0.15Hf–0.07Y–0.01B wt% base alloys with and without Nb, Ti, and C additions was evaluated at 1100 °C in air with 10% watermore » vapor. Furthermore, we observed protective alumina scale formation at levels of 35Ni, 25Cr, and 4Al with additions of Nb and C, indicating promise for 1100°C capable cast AFA alloys.« less

  7. Austenitic stainless steels and high strength copper alloys for fusion components

    NASA Astrophysics Data System (ADS)

    Rowcliffe, A. F.; Zinkle, S. J.; Stubbins, J. F.; Edwards, D. J.; Alexander, D. J.

    1998-10-01

    An austenitic stainless steel (316LN), an oxide-dispersion-strengthened copper alloy (GlidCop Al25), and a precipitation-hardened copper alloy (Cu-Cr-Zr) are the primary structural materials for the ITER first wall/blanket and divertor systems. While there is a long experience of operating 316LN stainless steel in nuclear environments, there is no prior experience with the copper alloys in neutron environments. The ITER first wall (FW) consists of a stainless steel shield with a copper alloy heat sink bonded by hot isostatic pressing (HIP). The introduction of bi-layer structural material represents a new materials engineering challenge; the behavior of the bi-layer is determined by the properties of the individual components and by the nature of the bond interface. The development of the radiation damage microstructure in both classes of materials is summarized and the effects of radiation on deformation and fracture behavior are considered. The initial data on the mechanical testing of bi-layers indicate that the effectiveness of GlidCop Al25 as a FW heat sink material is compromised by its strongly anisotropic fracture toughness and poor resistance to crack growth in a direction parallel to the bi-layer interface.

  8. Development of 1100 °C Capable Alumina-Forming Austenitic Alloys

    SciTech Connect

    Brady, M. P.; Muralidharan, G.; Yamamoto, Y.; Pint, B. A.

    2016-11-18

    Recently dalumina-forming austenitic (AFA) alloys based on ~12–32 weight % (wt%) Ni have been developed and offer an attractive combination of oxidation resistance and creep resistance at relatively low alloy cost. But, they exhibit a transition to internal oxidation and nitridation of Al above ~750–950 °C depending on composition and exposure environment. In order to identify AFA compositions capable of higher-temperature operation for applications such as ethylene cracking, the oxidation behavior of a series of developmental, as-cast nominal Fe–(25–45)Ni–(10–25)Cr–(4–5)Al–1Si–0.15Hf–0.07Y–0.01B wt% base alloys with and without Nb, Ti, and C additions was evaluated at 1100 °C in air with 10% water vapor. Furthermore, we observed protective alumina scale formation at levels of 35Ni, 25Cr, and 4Al with additions of Nb and C, indicating promise for 1100°C capable cast AFA alloys.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Pawel, J.E.; Rowcliffe, A.F.; Alexander, D.J.; Grossbeck, M.L.; 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.

  11. Effects of water chemistry on intergranular cracking of irradiated austenitic stainless steels

    SciTech Connect

    Chung, H.M.; Ruther, W.E.; Sanecki, J.E.; Hins, A.; Kassner, T.F.

    1995-12-31

    To determine the effects of water chemistry on the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) in austenitic stainless steels, constant-extension-rate tests were conducted in simulated BWR environments on several heats of high- and commercial-purity (HP and CP) Type 304 SS specimens from BWR components irradiated to fluences up to 2.4 {times} 10{sup 21} n cm{sup {minus}2} (E > 1 MeV). Effects of dissolved oxygen (DO) and electrochemical potential (ECP) in 289 C water were investigated. Dependence of susceptibility to intergranular stress corrosion cracking (IGSCC) on DO was somewhat different for the two materials. Susceptibility of the HP heats, less influenced by DO and ECP, was higher than that of CP material for all DO and fluence levels. Percent IGSCC in the CP material was negligible for DO < 0.01 ppm or ECP <{minus}140 mV SHE. Results of analysis by Auger electron spectroscopy indicated that the HP neutron absorber tubes were characterized by relatively lower concentrations of Cr, Ni, and Li and relatively higher concentrations of F and N on grain boundaries than those of the CP materials. It is suggested that a synergism between irradiation-induced grain-boundary Cr depletion and fabrication-related fluorine contamination plays an important role in the stress corrosion cracking behavior of the HP neutron absorber tubes.

  12. Alloy development for irradiation performance in fusion reactors. Annual report, September 1979-September 1980

    SciTech Connect

    Harling, O K; Grant, N J

    1980-12-01

    This report summarizes the research and development work performed during the second year of an M.I.T. project directed toward the development of improved structural alloys for the fusion reactor first wall application. Several new alloys have been produced by rapid solidification. Emphasis in alloy design and production has been placed on producing austenitic Type 316SS with fine dispersions of TiC and Al/sub 2/O/sub 3/ particles. Results of mechanical and microstructural tests are presented. A number of neutron irradiations have been initiated on samples fabricated from alloys produced in this project. A dual beam, heavy ion and helium ion, irradiation was completed using several alloys and a range of temperatures, damage rates and total doses. Modeling of irradiation phenomena has been continued with emphasis in the last year upon understanding the effect of recoil resolution on relatively stable second phase particles. Work continued to fully characterize the microstructure of several ZrB/sub 2/ doped stainless steels.

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

    NASA Astrophysics Data System (ADS)

    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.

  14. Re-weldability tests of irradiated austenitic stainless steel by a TIG welding method

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Kalinin, George

    2000-12-01

    Austenitic stainless steel (SS) is widely used for the in-vessel and ex-vessel components of fusion reactors. In particular, SS316L(N)-IG (IG-ITER Grade) is used for the vacuum vessel (VV), pipe lines, blanket modules, branch pipe lines connecting the module coolant system with the manifold and for the other structures of ITER. One of the most important requirements for the VV and the water cooling branch pipelines is the possibility to repair different defects by welding. Those components which may require re-welding should be studied carefully. The SS re-weldability issue has a large impact on the design of in-vessel components, in particular, the design and efficiency of radiation shielding by the modules. Moreover, re-welded components should operate for the lifetime of the reactor. This paper deals with the study of re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. Tungsten inert gas (TIG) welding was used for re-welding of the SS.

  15. HREM study on the ledge structures, transient lattices and dislocation structures at the austenite-martensite and austenite-bainite interfaces in Fe-based alloys

    NASA Astrophysics Data System (ADS)

    Kajiwara, S.

    2003-10-01

    High-resolution electron microscopy (HREM) has been performed to know the atomic arrangement of the austenite-martensite interface and the austenite-bainite interface in Fe-based alloys. The alloys studied are Fe-23.0Ni-3.8Mn, Fe-8.8Cr-l.lC, Fe-30.5Ni-lOCo-3Ti (mass %) for martensitic transformation and Fe-2Si-1.4C (mass %) for bainitic transformation. These alloys have various transformation characteristics depending on the alloy; for martensitic transformation, athermal and isothermal kinetics, the Kurdjumow-Sachs (K-S) and Nishiyama (N) orientation relationships, reversible and irreversible movement of the interface, and for bainitic transformation, upper bainite and lower bainite. All the interfaces observed had to be limited to 112 (macroscopically 225) or very close to 112 because of the geometrical condition that the atom rows of <110>f, b and <100>b must be observed parallel to the interface, i.e., the edge-on orientation. The austenite-martensite interface is (121)f with the K-S orientation relationship of (lll)f//(011)b and [ bar{1}01] f//[ bar{1}bar{1}1] b, and the interface is basically composed of the terrace of (lll)f and the ledge of (010)f, which have the average ratio of 2:1 for the number of atom rows of [ bar{1}01] //[ bar{1}bar{1}1] b on these planes. This interface always accompanies the transient lattice region with the thickness of 0.4-1.0 nm, where the lattice changes continuously from fcc to bcc (or bct). No extra-half plane is observed at the (121)f interface over a large distance of 100-200 lattice planes. The interface for both the upper and lower bainites is close to (112)f with the N orientation relationship of (lll)f/(011)b and [ bar{1}bar{1}0] f//[ bar{1}00] b'. Contrary to the interface for martensite, this interface for bainite has many extra-half planes except when the interface is close to (112)f. The interface is basically made up of the terrace of (lll)f/(011)b and the ledge of (0bar{1}l)b'//(bar{1}bar{1}2)f, and the

  16. Performance of Alumina-Forming Austenitic Steels, Fe-base and Ni-base alloys exposed to metal dusting environments

    SciTech Connect

    Vande Put Ep Rouaix, Aurelie; Unocic, Kinga A; Pint, Bruce A; Brady, Michael P

    2011-01-01

    A series of conventional Fe- and Ni- base, chromia- and alumina- forming alloys, and a newly developed creep-resistant, alumina-forming austenitic steel were developed and its performance relative to conventional Fe- and Ni-based chromia-forming alloys was evaluated in metal dusting environments with a range of water vapor contents. Five 500h experiments have been performed at 650 C with different water vapor contents and total pressures. Without water vapor, the Ni-base alloys showed greater resistance to metal dusting than the Fe-base alloys, including AFA. However, with 10-28% water vapor, more protective behavior was observed with the higher-alloyed materials and only small mass changes were observed. Longer exposure times are in progress to further differentiate performance.

  17. A comparative evaluation of welding consumables for dissimilar weids between 316LN austenitic stainless steel and Alloy 800

    NASA Astrophysics Data System (ADS)

    Sireesha, M.; Albert, Shaju K.; Shankar, V.; Sundaresan, S.

    2000-03-01

    Transition joints in power plants between ferritic steels and austenitic stainless steels suffer from a mismatch in coefficients of thermal expansion (CTE) and the migration of carbon during service from the ferritic to the austenitic steel. To overcome these, nickel-based consumables are commonly used. The use of a trimetallic combination with an insert piece of intermediate CTE provides for a more effective lowering of thermal stresses. The current work envisages a trimetallic joint involving modified 9Cr-1Mo steel and 316LN austenitic stainless steel as the base materials and Alloy 800 as the intermediate piece. Of the two joints involved, this paper describes the choice of welding consumables for the joint between Alloy 800 and 316LN. Four consumables were examined: 316, 16-8-2, Inconel 82 and Inconel 182. The comparative evaluation was based on hot cracking tests and estimation of mechanical properties and coefficient of thermal expansion. While 16-8-2 exhibited highest resistance to solidification cracking, the Inconel filler materials also showed adequate resistance; additionally, the latter were superior from the mechanical property and coefficient of thermal expansion view-points. It is therefore concluded that for the joint between Alloy 800 and 316LN the Inconel filler materials offer the best compromise.

  18. Tomographic atom probe characterization of the microstructure of a cold worked 316 austenitic stainless steel after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Etienne, A.; Radiguet, B.; Pareige, P.; Massoud, J.-P.; Pokor, C.

    2008-11-01

    For the first time, chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after 17 years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa (7.81 × 10 25 neutrons.m -2, E > 1 MeV). The samples were analysed with a laser assisted tomographic atom probe. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A high number density (6 × 10 23 m -3) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. Mo and P segregations at the interfaces of these clusters were also observed. Finally, Si enriched atmospheres were seen.

  19. Ion-irradiation-assisted tuning of phase transformations and physical properties in single crystalline Fe7Pd3 ferromagnetic shape memory alloy thin films

    NASA Astrophysics Data System (ADS)

    Arabi-Hashemi, A.; Witte, R.; Lotnyk, A.; Brand, R. A.; Setzer, A.; Esquinazi, P.; Hahn, H.; Averback, R. S.; Mayr, S. G.

    2015-05-01

    Control of multi-martensite phase transformations and physical properties constitute greatly unresolved challenges in Fe7Pd3-based ferromagnetic shape memory alloys. Single crystalline Fe7Pd3 thin films reveal an austenite to martensite phase transformation, continuously ranging from the face-centered cubic (fcc) to the face-centered tetragonal (fct) and body-centered cubic (bcc) phases upon irradiation with 1.8 MeV Kr+ ions. Within the present contribution, we explore this scenario within a comprehensive experimental study: employing atomic force microscopy (AFM) and high resolution transmission electron microscopy (HR-TEM), we first clarify the crystallography of the ion-irradiation-induced austenite \\Rightarrow martensite and inter-martensite transitions, explore the multi-variant martensite structures with c-a twinning and unravel a very gradual transition between variants at twin boundaries. Accompanying magnetic properties, addressed locally and globally, are characterized by an increasing saturation magnetization from fcc to bcc, while coercivity and remanence are demonstrated to be governed by magnetocrystalline anisotropy and ion-irradiation-induced defect density, respectively. Based on reversibility of ion-irradiation-induced materials changes due to annealing treatment and a conversion electron Mößbauer spectroscopy (CEMS) study to address changes in order, a quantitative defect-based physical picture of ion-irradiation-induced austenite ⇔ martensite transformation in Fe7Pd3 is developed. The presented concepts thus pave the way for ion-irradiation-assisted optimization strategies for tailored functional alloys.

  20. Slag remelt purification of irradiated vanadium alloys

    SciTech Connect

    Carmack, W.J.; Smolik, G.R.; McCarthy, K.A.; Gorman, P.K.

    1995-07-01

    This paper describes theoretical and scoping experimental efforts to investigate the decontamination potential of a slag remelting process for decontaminating irradiated vanadium alloys. Theoretical calculations, using a commercial thermochemical computer code HSC Chemistry, determined the potential slag compositions and slag-vanadium alloy ratios. The experiment determined the removal characteristics of four surrogate transmutation isotopes (Ca, Y - to simulate Sc, Mn, and Ar) from a V-5Ti-5Cr alloy with calcium fluoride slag. An electroslag remelt furnace was used in the experiment to melt and react the constituents. The process achieved about a 90 percent removal of calcium and over 99 percent removal of yttrium. Analyses indicate that about 40 percent of the manganese may have been removed. Argon analyses indicates that 99.3% of the argon was released from the vanadium alloy in the first melt increasing to 99.7% during the second melt. Powder metallurgy techniques were used to incorporate surrogate transmutation products in the vanadium. A powder mixture was prepared with the following composition: 90 wt % vanadium, 4.7 wt % titanium, 4.7 wt % chromium, 0.35 wt % manganese, 0.35 wt % CaO, and 0.35 wt % Y{sub 2}O{sub 3}. This mixture was packed into 2.54 cm diameter stainless steel tubes. Argon was introduced into the powder mixture by evacuating and backfilling the stainless steel containers to a pressure of 20 kPa (0.2 atm). The tubes were hot isostatically pressed at 207 MPa (2000 atm) and 1473 K to consolidate the metal. An electroslag remelt furnace (crucible dimensions: 5.1 cm diameter by 15.2 cm length) was used to process the vanadium electrodes. Chemical analyses were performed on samples extracted from the slags and ingots. Ingot analyses results are shown below. Values are shown in percent removal of the four targeted elements of the initial compositions.

  1. Evaluation of Alumina-Forming Austenitic Stainless Steel Alloys in Microturbines

    SciTech Connect

    Brady, M.P.; Matthews, W.J.

    2010-09-15

    Oak Ridge National Laboratory (ORNL) and Capstone Turbine Corporation (CTC) 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 microturbine recuperator components. ORNL delivered test coupons of three different AFA compositions to CTC. The coupons were exposed in steady-state elevated turbine exit temperature (TET) engine testing, with coupons removed for analysis after accumulating ~1,500, 3,000, 4,500, and 6,000 hours of operation. Companion test coupons were also exposed in oxidation testing at ORNL at 700-800°C in air with 10% H2O. Post test assessment of the coupons was performed at ORNL by light microscopy and electron probe microanalysis. The higher Al and Nb containing AFA alloys exhibited excellent resistance to oxidation/corrosion, and thus show good promise for recuperator applications.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    Enomoto, M.; Wan, X. L.

    2017-04-01

    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.

  4. Electrochemical and microstructural characterization of an austenitic stainless steel irradiated by heavy ions at 515°C

    NASA Astrophysics Data System (ADS)

    Bell, G. E. C.; Inazumi, T.; Kenik, E. A.; Kondo, T.

    1992-04-01

    The electrochemical and microstructural behavior of a solution-annealed, heavy-ion-irradiated, austenitic stainless steel. designated LS1A, have been investigated at 515°C after doses of 1.10 and 30 displacements per atom (dpa). Changes in electrochemical properties due to radiation-induced segregation in thin radiation-affected layers of the material were detected by the electrochemical potentiokinetic reactivation (EPR) technique using TEM disk specimens. At all doses, the Flade potential and reactivation charge were greater than those measured for thermally-aged control specimens. Grain face etching, similar to that found on EPR-tested neutron irradiated austenitic stainless steels, was observed on all specimens after testing. Duplicate heavy ion irradiated specimens were also examined by high resolution analytical electron microscopy (AEM). The 1 dpa specimen showed only a high density of small faulted dislocations (~ 10 nm), and no grain boundary precipitation or grain boundary segregation was detected. AEM confirmed chromium and molybdenum depletion at grain boundaries as measured by EPR for the 10 and 30 dpa specimens.

  5. BWRVIP-140NP: BWR Vessel and Internals Project Fracture Toughness and Crack Growth Program on Irradiated Austenitic Stainless Steel

    SciTech Connect

    Gilman, J

    2005-03-15

    To prepare for this project, EPRI and BWRVIP conducted a workshop at Ponte Vedra Beach, Florida during February 19-21, 2003 (EPRI report 1007822). Attendees were invited to exchange relevant information on the effects of irradiation on austenitic materials in light water reactors and to produce recommendations for further work. EPRI reviewed the data, recommendations, and conclusions derived from the workshop and developed prioritized test matrices defining new data needs. Proposals were solicited, and selected proposals are the basis for the program described in this report. Results The planned test matrix for fracture toughness testing includes 21 tests on 5 materials.

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

    NASA Astrophysics Data System (ADS)

    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.

  7. Irradiation-assisted stress corrosion cracking in HTH Alloy X-750 and Alloy 625

    SciTech Connect

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-12-31

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water to determine the irradiation-assisted stress corrosion cracking (IASCC) behavior of HTH Alloy X-750 and direct-aged Alloy 625. New data confirm previous results showing that high irradiation levels reduce SCC resistance in Alloy X-750. Heat-to-heat variability correlates with boron content, with low boron heats showing improved IASCC properties. Alloy 625 is resistant to IASCC, as no cracking was observed in any Alloy 625 specimens. Microstructural, microchemical and deformation studies were performed to characterize the mechanisms responsible for IASCC in Alloy X-750 and the lack of an effect in Alloy 625. The mechanisms under investigation are: boron transmutation effects, radiation-induced changes in microstructure and deformation characteristics, and radiation-induced segregation. Irradiation of Alloy X-750 caused significant strengthening and ductility loss that was associated with the formation of cavities and dislocation loops. High irradiation levels did not cause significant segregation of alloying or trace elements in Alloy X-750. Irradiation of Alloy 625 resulted in the formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to the loops and precipitates was apparently offset by a partial dissolution of {gamma}{double_prime} precipitates, as Alloy 625 showed no irradiation-induced strengthening or ductility loss. In the nonirradiated condition, an IASCC susceptible HTH heat containing 28 ppm B showed grain boundary segregation of boron, whereas a nonsusceptible HTH heat containing 2 ppm B and Alloy 625 with 20 ppm B did not show significant boron segregation. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in Alloy X-750, and the absence of these two effects results in the superior IASCC resistance displayed by Alloy 625.

  8. Deformation Microstructure and Deformation-Induced Martensite in Austenitic Fe-Cr-Ni Alloys Depending on Stacking Fault Energy

    NASA Astrophysics Data System (ADS)

    Tian, Ye; Gorbatov, Oleg I.; Borgenstam, Annika; Ruban, Andrei V.; Hedström, Peter

    2017-01-01

    The deformation microstructure of austenitic Fe-18Cr-(10-12)Ni (wt pct) alloys with low stacking fault energies, estimated by first-principles calculations, was investigated after cold rolling. The ɛ-martensite was found to play a key role in the nucleation of α'-martensite, and at low SFE, ɛ formation is frequent and facilitates nucleation of α' at individual shear bands, whereas shear band intersections become the dominant nucleation sites for α' when SFE increases and mechanical twinning becomes frequent.

  9. Ab initio investigation of the surface properties of austenitic Fe-Ni-Cr alloys in aqueous environments

    NASA Astrophysics Data System (ADS)

    Rák, Zs.; Brenner, D. W.

    2017-04-01

    The surface energetics of two austenitic stainless steel alloys (Type 304 and 316) and three Ni-based alloys (Alloy 600, 690, and 800) are investigated using theoretical methods within the density functional theory. The relative stability of the low index surfaces display the same trend for all alloys; the most closely packed orientation and the most stable is the (111), followed by the (100) and the (110) surfaces. Calculations on the (111) surfaces using various surface chemical and magnetic configurations reveal that Ni has the tendency to segregate toward the surface and Cr has the tendency to segregate toward the bulk. The magnetic frustration present on the (111) surfaces plays an important role in the observed segregation tendencies of Ni and Cr. The stability of the (111) surfaces in contact with aqueous solution are evaluated as a function of temperature, pH, and concentration of aqueous species. The results indicate that the surface stability of the alloys decrease with temperature and pH, and increase slightly with concentration. Under conditions characteristic to an operating pressurized water reactor, the Ni-based alloy series appears to be of better quality than the stainless steel series with respect to corrosion resistance and release of aqueous species when in contact with aqueous solutions.

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

    SciTech Connect

    Muralidharan, Govindarajan; Yamamoto, Yukinori; Brady, Michael P; Pint, Bruce A; Pankiw, Roman; Voke, Don

    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.

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

    NASA Astrophysics Data System (ADS)

    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

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

    PubMed

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

    2016-12-01

    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 × 10(14) to 2.7 × 10(18) D/cm(2). 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 × 10(17) D/cm(2)); II-the nonlinear region of medium implantation doses (1 × 10(17) to 8 × 10(17) D/cm(2)); III-the linear region of high implantation doses (8 × 10(17) to 2.7 × 10(18) D/cm(2)). 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

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

    NASA Astrophysics Data System (ADS)

    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.

  14. Irradiation creep and creep rupture of titanium-modified austenitic stainless steels and their dependence on cold work level

    SciTech Connect

    Garner, F.A.; Hamilton, M.L. ); Eiholzer, C.R. ); Toloczko, M.B. ); Kumar, A.S. )

    1991-11-01

    A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% from the conventional 20% level was detrimental to its performance, especially for applications above 550{degrees}c. The 20% cold work level is preferable to the 10% level, in terms of both in-reactor creep rupture response and initial strength.

  15. Influence of cold work level on the irradiation creep and creep rupture of titanium-modified austenitic stainless steels

    SciTech Connect

    Garner, F.A.; Hamilton, M.L. ); Eiholzer, C.R. ); Toloczko, M.B. ); Kumar, A.S. )

    1992-06-01

    A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% form the conventional 20% level was detrimental to its performance, especially for applications above 550{degrees}C. The 20% cold work level is preferable to the 10% level, in terms of both in- reactor creep rapture response and initial strength.

  16. An analysis of the kinetics, morphology, and mechanism of austenite formation during thermal processing of iron alloys

    NASA Astrophysics Data System (ADS)

    Schmidt, Eric

    The solid state phenomenon of austenite precipitation from ferrite occurs at some point during the thermal processing of nearly all steels. Austenitization in pure iron is expected to be controlled by processes which occur at the migrating austenite/ferrite interfaces. An analytic expression which accounts for these processes has been proposed which generally follows the transition state theory for thermally activated processes. The velocity of an interface controlled by this mechanism should be very fast (for pure iron, a velocity of 100s of mum/s in a temperature range from about 915°C to 940°C has been measured), will be linear with temperature, and is not time dependant. This model for interface-reaction controlled migrating interfaces has been found to be consistent with observations in pure iron, and in interstitial free steels. The morphology of austenite precipitates during the interface reaction controlled transformation suggests that this phase transformation is a massive transformation with incoherent interfaces and no partitioning of solute atoms. The mobility of interface reaction-controlled transformation boundaries reported in the present and previous investigations have been discussed in further detail. The morphology of austenite precipitates, with regard to the appearance of the migrating interfaces and the initial location of carbon in the microstructure, have been found to be consistent with the massive transformation in pure iron. This can he shown in binary iron-carbon alloy and in a set of carbon steels which contain various amounts of e.g. manganese, chromium, and nickel. The mobility of partitionless, massive transformation interfaces has been found generally to range over 6 orders of magnitude, and is a few to several orders of magnitude larger in pure iron than in Fe-C or Fe-C-X steels. If the transformation can be made to occur in the single phase austenite region for an alloy, the interface mobility may increase significantly at long

  17. Irradiation performance of FFTF drivers using the D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1994-12-31

    Five test assemblies similar in design to the Fast Flux Test Facility driver fuel assembly , but employing the alloy D9 in place of stainless steel 316 for duct, cladding, and wire wrap compnents were irradiated to demonstrate the improved performance of the new design. Results of post-irradiation examinations are discussed.

  18. Characterization of the Carbon and Retained Austenite Distributions in Martensitic Medium Carbon, Low Alloy, Steel

    SciTech Connect

    Sherman, D. H.; Cross, Steven M; Kim, Sangho; Grandjean, F.; Long, G. J.; Miller, Michael K

    2007-01-01

    The retained austenite content and carbon distribution in martensite were determined as a function of cooling rate and temper temperature in steel that contained 1.31 at. pct C, 3.2 at. pct Si, and 3.2 at. pct non-iron metallic elements. Mossbauer spectroscopy, transmission electron microscopy (TEM), transmission synchrotron X-ray diffraction (XRD), and atom probe tomography were used for the microstructural analyses. The retained austenite content was an inverse, linear function of cooling rate between 25 and 560 K/s. The elevated Si content of 3.2 at. pct did not shift the start of austenite decomposition to higher tempering temperatures relative to SAE 4130 steel. The minimum tempering temperature for complete austenite decomposition was significantly higher (>650 C) than for SAE 4130 steel ({approx}300 C). The tempering temperatures for the precipitation of transition carbides and cementite were significantly higher (>400 C) than for carbon steels (100 C to 200 C and 200 C to 350 C), respectively. Approximately 90 pct of the carbon atoms were trapped in Cottrell atmospheres in the vicinity of the dislocation cores in dislocation tangles in the martensite matrix after cooling at 560 K/s and aging at 22 C. The 3.2 at. pct Si content increased the upper temperature limit for stable carbon clusters to above 215 C. Significant autotempering occurred during cooling at 25 K/s. The proportion of total carbon that segregated to the interlath austenite films decreased from 34 to 8 pct as the cooling rate increased from 25 to 560 K/s. Developing a model for the transfer of carbon from martensite to austenite during quenching should provide a means for calculating the retained austenite. The maximum carbon content in the austenite films was 6 to 7 at. pct, both in specimens cooled at 560 K/s and at 25 K/s. Approximately 6 to 7 at. pct carbon was sufficient to arrest the transformation of austenite to martensite. The chemical potential of carbon is the same in martensite

  19. Dissolution and oxidation behaviour of various austenitic steels and Ni rich alloys in lead-bismuth eutectic at 520 °C

    NASA Astrophysics Data System (ADS)

    Roy, Marion; Martinelli, Laure; Ginestar, Kevin; Favergeon, Jérôme; Moulin, Gérard

    2016-01-01

    Ten austenitic steels and Ni rich alloys were tested in static lead-bismuth eutectic (LBE) at 520 °C in order to obtain a selection of austenitic steels having promising corrosion behaviour in LBE. A test of 1850 h was carried out with a dissolved oxygen concentration between 10-9 and 5 10-4 g kg-1. The combination of thermodynamic of the studied system and literature results leads to the determination of an expression of the dissolved oxygen content in LBE as a function of temperature: RT(K)ln[O](wt%) = -57584/T(K) -55.876T(K) + 254546 (R is the gas constant in J mol-1 K-1). This relation can be considered as a threshold of oxygen content above which only oxidation is observed on the AISI 316L and AISI 304L austenitic alloys in static LBE between 400 °C and 600 °C. The oxygen content during the test leads to both dissolution and oxidation of the samples during the first 190 h and leads to pure oxidation for the rest of the test. Results of mixed oxidation and dissolution test showed that only four types of corrosion behaviour were observed: usual austenitic steels and Ni rich alloys behaviour including the reference alloy 17Cr-12Ni-2.5Mo (AISI 316LN), the 20Cr-31Ni alloy one, the Si containing alloy one and the Al containing alloy one. According to the proposed criteria of oxidation and dissolution kinetics, silicon rich alloys and aluminum rich alloy presented a promising corrosion behaviour.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  2. Isolation of the role of radiation-induced segregation in irradiation-assisted stress corrosion cracking of proton-irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Busby, Jeremy Todd

    2001-11-01

    The role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) was studied in order to better understand the underlying mechanisms of IASCC. High-purity 304L (HP-304L), commercial purity 304 (CP-304) and commercial purity 316 (CP-316) stainless steel alloys were irradiated with 3.2 MeV protons at 400°C (HP-304L) and 360°C (CP-304 and CP-316) to doses ranging from 0.1 and 5.0 dpa. Grain boundary chemistry was measured using scanning transmission electron microscopy with energy-dispersive spectroscopy (STEM/EDS) in both unirradiated and irradiated samples. Unirradiated and irradiated samples of the two commercial purity alloys were also strained to failure in an aqueous environment representative of boiling water reactor cores. The cracking susceptibility and RIS in the proton-irradiated CP-304 is very similar to that from the neutron-irradiated samples. The CP-316 alloy did not crack. Radiation-induced segregation, cracking susceptibility, and dislocation loop microstructure developed at the same rate as a function of dose in the CP-304 alloy. To isolate the effects of RIS in IASCC, post-irradiation annealing was utilized. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure show that dislocation loops are removed preferentially over RIS due to the density of vacancies required and kinetic considerations. Experimental anneals were conducted on HP-304L samples irradiated to 1.0 dpa and CP-304 samples irradiated to 1.0 and 2.5 dpa. Post-irradiation anneals were performed at temperatures ranging from 400°C to 650°C for times between 45 minutes and 5 hours. At all temperatures, the hardness and dislocation densities decreased with increasing annealing time much faster than RIS did. Annealing at 600°C for 90 minutes removed virtually all dislocation microstructure while leaving RIS intact. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing

  3. Manufacture of Alumina-Forming Austenitic Stainless Steel Alloys by Conventional Casting and Hot-Working Methods

    SciTech Connect

    Brady, M.P.; Yamamoto, Y.; Magee, J.H.

    2009-03-23

    Oak Ridge National Laboratory (ORNL) and Carpenter Technology Corporation (CarTech) 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 scale up of developmental ORNL alumina-forming austenitic (AFA) stainless steels by conventional casting and rolling techniques. CarTech successfully vacuum melted 30lb heats of four AFA alloy compositions in the range of Fe-(20-25)Ni-(12-14)Cr-(3-4)Al-(1-2.5)Nb wt.% base. Conventional hot/cold rolling was used to produce 0.5-inch thick plate and 0.1-inch thick sheet product. ORNL subsequently successfully rolled the 0.1-inch sheet to 4 mil thick foil. Long-term oxidation studies of the plate form material were initiated at 650, 700, and 800 C in air with 10 volume percent water vapor. Preliminary results indicated that the alloys exhibit comparable (good) oxidation resistance to ORNL laboratory scale AFA alloy arc casting previously evaluated. The sheet and foil material will be used in ongoing evaluation efforts for oxidation and creep resistance under related CRADAs with two gas turbine engine manufacturers. This work will be directed to evaluation of AFA alloys for use in gas turbine recuperators to permit higher-temperature operating conditions for improved efficiencies and reduced environmental emissions.

  4. Swelling of several commercial alloys following high fluence neutron irradiation

    NASA Astrophysics Data System (ADS)

    Powell, R. W.; Peterson, D. T.; Zimmerschied, M. K.; Bates, J. F.

    Swelling values have been determined for a set of commercial alloys irradiated to a peak fluence of 1.8 × 10 23 n/cm 2 (E >0.1 MeV) over the temperature range of 400 to 650°C. The alloys studied fall into three classes: the ferritic alloys AISI 430F, AISI 416, EM-12, H-11 and 2 {1}/{4}Cr-1Mo; the superalloys Inconel 718 and Inconel X-750; and the refractory alloys TZM and Nb-1Zr. All of these alloys display swelling resistance far superior to cold worked AISI 316. Of the three alloy classes examined the swelling resistance of the ferritics is the least sensitive to composition.

  5. Microstructural stability and mechanical behavior of FeNiMnCr high entropy alloy under ion irradiation

    SciTech Connect

    Leonard, Keith J.; Bei, Hongbin; Zinkle, Steven J.; Kiran Kumar, N. A. P.; Li, C.

    2016-05-13

    In recent years, high entropy alloys (HEAs) have attracted significant attention due to their excellent mechanical properties and good corrosion resistance, making them potential candidates for high temperature fission and fusion structural applications. However there is very little known about their radiation resistance, particularly at elevated temperatures relevant for energy applications. In the present study, a single phase (face centered cubic) concentrated solid solution alloy of composition 27%Fe-28%Ni-27%Mn-18%Cr was irradiated with 3 or 5.8 MeV Ni ions at temperatures ranging from room temperature to 700 °C and midrange doses from 0.03 to 10 displacements per atom (dpa). Transmission electron microscopy (TEM), scanning transmission electron microscopy with energy dispersive x-ray spectrometry (STEM/EDS) and X-ray diffraction (XRD) were used to characterize the radiation defects and microstructural changes. Irradiation at higher temperatures showed evidence of relatively sluggish solute diffusion with limited solute depletion or enrichment at grain boundaries. The main microstructural feature at all temperatures was high-density small dislocation loops. Voids were not observed at any irradiation condition. Nano-indentation tests on specimens irradiated at room temperature showed a rapid increase in hardness ~35% and ~80% higher than the unirradiated value at 0.03 and 0.3 dpa midrange doses, respectively. The irradiation-induced hardening was less pronounced for 500 °C irradiations (<20% increase after 3 dpa). Overall, the examined HEA material exhibits superior radiation resistance compared to conventional single phase Fe-Cr-Ni austenitic alloys such as stainless steels. Furthermore, the present study provides insight on the fundamental irradiation behavior of a single phase HEA material over a broad range of irradiation temperatures.

  6. Microstructural stability and mechanical behavior of FeNiMnCr high entropy alloy under ion irradiation

    DOE PAGES

    Leonard, Keith J.; Bei, Hongbin; Zinkle, Steven J.; ...

    2016-05-13

    In recent years, high entropy alloys (HEAs) have attracted significant attention due to their excellent mechanical properties and good corrosion resistance, making them potential candidates for high temperature fission and fusion structural applications. However there is very little known about their radiation resistance, particularly at elevated temperatures relevant for energy applications. In the present study, a single phase (face centered cubic) concentrated solid solution alloy of composition 27%Fe-28%Ni-27%Mn-18%Cr was irradiated with 3 or 5.8 MeV Ni ions at temperatures ranging from room temperature to 700 °C and midrange doses from 0.03 to 10 displacements per atom (dpa). Transmission electron microscopymore » (TEM), scanning transmission electron microscopy with energy dispersive x-ray spectrometry (STEM/EDS) and X-ray diffraction (XRD) were used to characterize the radiation defects and microstructural changes. Irradiation at higher temperatures showed evidence of relatively sluggish solute diffusion with limited solute depletion or enrichment at grain boundaries. The main microstructural feature at all temperatures was high-density small dislocation loops. Voids were not observed at any irradiation condition. Nano-indentation tests on specimens irradiated at room temperature showed a rapid increase in hardness ~35% and ~80% higher than the unirradiated value at 0.03 and 0.3 dpa midrange doses, respectively. The irradiation-induced hardening was less pronounced for 500 °C irradiations (<20% increase after 3 dpa). Overall, the examined HEA material exhibits superior radiation resistance compared to conventional single phase Fe-Cr-Ni austenitic alloys such as stainless steels. Furthermore, the present study provides insight on the fundamental irradiation behavior of a single phase HEA material over a broad range of irradiation temperatures.« less

  7. A scanning Hall probe imaging study of the field induced martensite-austenite phase transition in Ni50Mn34In16 alloy.

    PubMed

    Sharma, V K; Moore, J D; Chattopadhyay, M K; Morrison, Kelly; Cohen, L F; Roy, S B

    2010-01-13

    The martensite to austenite phase transition in the off-stoichiometric Heusler alloy Ni(50)Mn(34)In(16) can be induced both by temperature change and by application of a magnetic field. We have used scanning Hall probe imaging to study the magnetic field induced martensite-austenite phase transition. The study provides clear visual evidence of the coexistence of the martensite and austenite phases across this field induced transition in both increasing and decreasing magnetic fields. Clear evidence of thermomagnetic history effects associated with the martensite-austenite phase transition is also obtained. Quantitative analysis of the magnetic field dependence of the volume fraction of the austenite phase in Ni(50)Mn(34)In(16) shows evidence of a nucleation and growth mechanism across the field induced martensite-austenite phase transition. The local M-H loops constructed from the Hall images indicate the presence of a landscape of the critical magnetic field (for the field induced transition) distributed over the sample volume and thus confirm the disorder influenced nature of this first-order magnetic phase transition.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Ermi, A.M.; Gelles, D.S.

    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.

  10. Effects of self-irradiation in plutonium alloys

    DOE PAGES

    Chung, B. W.; Lema, K. E.; Allen, P. G.

    2015-09-16

    In this paper, we present updated results of self-irradiation effects on 238Pu-enriched 239Pu alloys measured by immersion density, dilatometry, and tensile tests. We obtained the self-irradiation equivalent time of nearly 200 years, nearly 100 years longer than in our previous papers. At this extended aging, we find the rate of decrease in density has slowed significantly, stabilizing around 15.73 g/cc, without signs of void swelling. The volume expansion measured at 35°C also shows apparent saturation at less than 0.25%. Quasi-static tensile measurement still show gradual increase in the strength of plutonium alloys with age.

  11. Effect of Alloying Additions on Phase Equilibria and Creep Resistance of Alumina-Forming Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Yamamoto, Y.; Santella, M. L.; Brady, M. P.; Bei, H.; Maziasz, P. J.

    2009-08-01

    The high-temperature creep properties of a series of alumina-forming austenitic (AFA) stainless steels based on Fe-20Ni-(12-14)Cr-(2.5-4)Al-(0.2-3.3)Nb-0.1C (weight percent) were studied. Computational thermodynamics were used to aid in the interpretation of data on microstructural stability, phase equilibria, and creep resistance. Phases of MC (M: mainly Nb), M23C6 (M: mainly Cr), B2 [ β-(Ni,Fe)Al], and Laves [Fe2(Mo,Nb)] were observed after creep-rupture testing at 750 °C and 170 MPa; this was generally consistent with the thermodynamic calculations. The creep resistance increased with increasing Nb additions up to 1 wt pct in the 2.5 and 3 Al wt pct alloy series, due to the stabilization of nanoscale MC particles relative to M23C6. Additions of Nb greater than 1 wt pct decreased creep resistance in the alloy series due to stabilization of the Laves phase and increased amounts of undissolved, coarse MC, which effectively reduced the precipitation of nanoscale MC particles. The additions of Al also increased the creep resistance moderately due to the increase in the volume fraction of B2 phase precipitates. Calculations suggested that optimum creep resistance would be achieved at approximately 1.5 wt pct Nb in the 4 wt pct Al alloy series.

  12. Effect of alloying additions on phase equilibria and creep resistance of alumina-forming austenitic stainless steels

    SciTech Connect

    Yamamoto, Yukinori; Santella, Michael L; Brady, Michael P; Bei, Hongbin; Maziasz, Philip J

    2009-01-01

    The high-temperature creep properties of a series of alumina-forming austenitic (AFA) stainless steels based on Fe-20Ni-(12-14)Cr-(2.5-4)Al-(0.2-3.3)Nb-0.1C (weight percent) were studied. Computational thermodynamics were used to aid in the interpretation of data on microstructural stability, phase equilibria, and creep resistance. Phases of MC (M: mainly Nb), M{sub 23}C{sub 6} (M: mainly Cr), B2 [{beta}-(Ni,Fe)Al], and Laves [Fe{sub 2}(Mo,Nb)] were observed after creep-rupture testing at 750 C and 170 MPa; this was generally consistent with the thermodynamic calculations. The creep resistance increased with increasing Nb additions up to 1 wt pct in the 2.5 and 3 Al wt pct alloy series, due to the stabilization of nanoscale MC particles relative to M{sub 23}C{sub 6}. Additions of Nb greater than 1 wt pct decreased creep resistance in the alloy series due to stabilization of the Laves phase and increased amounts of undissolved, coarse MC, which effectively reduced the precipitation of nanoscale MC particles. The additions of Al also increased the creep resistance moderately due to the increase in the volume fraction of B2 phase precipitates. Calculations suggested that optimum creep resistance would be achieved at approximately 1.5 wt pct Nb in the 4 wt pct Al alloy series.

  13. Impact of Mn on the solution enthalpy of hydrogen in austenitic Fe-Mn alloys: a first-principles study.

    PubMed

    von Appen, Jörg; Dronskowski, Richard; Chakrabarty, Aurab; Hickel, Tilmann; Spatschek, Robert; Neugebauer, Jörg

    2014-12-05

    Hydrogen interstitials in austenitic Fe-Mn alloys were studied using density-functional theory to gain insights into the mechanisms of hydrogen embrittlement in high-strength Mn steels. The investigations reveal that H atoms at octahedral interstitial sites prefer a local environment containing Mn atoms rather than Fe atoms. This phenomenon is closely examined combining total energy calculations and crystal orbital Hamilton population analysis. Contributions from various electronic phenomena such as elastic, chemical, and magnetic effects are characterized. The primary reason for the environmental preference is a volumetric effect, which causes a linear dependence on the number of nearest-neighbour Mn atoms. A secondary electronic/magnetic effect explains the deviations from this linearity.

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

    NASA Astrophysics Data System (ADS)

    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.

  15. Irradiation damage in multicomponent equimolar alloys and high entropy alloys (HEAs).

    PubMed

    Nagase, Takeshi; Rack, Philip D; Egami, Takeshi

    2014-11-01

    To maintain sustainable energy supply and improve the safety and efficiency of nuclear reactors, development of new and advanced nuclear materials with superior resistance to irradiation damage is necessary. Recently, a new generation of structural materials, termed as multicomponent equimolar alloys and/or high entropy alloys (HEAs), are being developed. These alloys consist of multicomponent elements for maximizing the compositional entropy, which stabilizes the solid solution phase. In this paper, preliminary studies on the irradiation damage in equimolar alloys and HEAs by High Voltage Electron Microscopy (HVEM) are reported [1-4]. (1) ZrHfNb equimolar alloys [1, 2]A multicomponent ZrHfNb alloy was prepared by a co-sputtering process using elemental Zr, Hf, and Nb targets using an AJA International ATC 2000-V system. A single-phase bcc solid solution was obtained in the ZrHfNb alloy with an approximately equiatomic ratio of its constituent elements. The irradiation-induced structural change in the ZrHfNb equimolar alloys with the bcc solid solution structure was investigated by HVEM using the Hitachi H-3000 installed at Osaka University. The polycrystalline bcc phase shows high phase stability against irradiation damage at 298 K; the bcc solid solution phase, whose grain size was about 20 nm, remained as a main constituent phase even after the severe irradiation damage that reached 10 dpa. (2) CoCrCuFeNi HEAs [3]A single-phase fcc solid solution was obtained in a CoCrCuFeNi alloy. The microstructure of the alloy depended on the preparation technique: a nanocrystalline CoCrCuFeNi alloy with an approximately equiatomic ratio of its constituent elements was obtained by a co-sputtering process with multi-targets, while polycrystalline structures were formed when the arc-melting method was used. Both nanocrystalline and polycrystalline structures showed high phase stability against fast electron irradiation at temperatures ranging from 298 K to 973 K; a fcc

  16. Structural transformations in austenitic stainless steel induced by deuterium implantation: irradiation at 100 K

    NASA Astrophysics Data System (ADS)

    Morozov, Oleksandr; Zhurba, Volodymyr; Neklyudov, Ivan; Mats, Oleksandr; Rud, Aleksandr; Chernyak, Nikolay; Progolaieva, Viktoria

    2015-03-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic stainless steel 18Cr10NiTi preimplanted at 100 K with deuterium ions in the dose range from 3 × 1015 to 5 × 1018 D/cm2. The kinetics of structural transformation development in the implantation steel layer was traced from deuterium thermodesorption spectra as a function of implanted deuterium concentration. At saturation of austenitic stainless steel 18Cr10NiTi with deuterium by means of ion implantation, structural-phase changes take place, depending on the dose of implanted deuterium. The maximum attainable concentration of deuterium in steel is C = 1 (at.D/at.met. = 1/1). The increase in the implanted dose of deuterium is accompanied by the increase in the retained deuterium content, and as soon as the deuterium concentration attains C ≈ 0.5 the process of shear martensitic structural transformation in steel takes place. It includes the formation of bands, body-centered cubic (bcc) crystal structure, and the ferromagnetic phase. Upon reaching the deuterium concentration C > 0.5, the presence of these molecules causes shear martensitic structural transformations in the steel, which include the formation of characteristic bands, bcc crystal structure, and the ferromagnetic phase. At C ≥ 0.5, two hydride phases are formed in the steel, the decay temperatures of which are 240 and 275 K. The hydride phases are formed in the bcc structure resulting from the martensitic structural transformation in steel.

  17. Study of irradiation creep of vanadium alloys

    SciTech Connect

    Tsai, H.; Strain, R.V.; Smith, D.L.

    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. Development of Advanced Corrosion-Resistant Fe-Cr-Ni Austenitic Stainless Steel Alloy with Improved High-Temperature Strength and Creep-Resistance

    SciTech Connect

    Maziasz, P.J.; Swindeman, R.W.

    2001-06-15

    In February of 1999, a Cooperative Research and Development Agreement (CRADA) was undertaken between Oak Ridge National Laboratory (ORNL) and Special Metals Corporation - Huntington Alloys (formerly INCO Alloys International, Inc.) to develop a modified wrought austenitic stainless alloy with considerably more strength and corrosion resistance than alloy 800H or 800HT, but with otherwise similar engineering and application characteristics. Alloy 800H and related alloys have extensive use in coal flue gas environments, as well as for tubing or structural components in chemical and petrochemical applications. The main concept of the project was make small, deliberate elemental microalloying additions to this Fe-based alloy to produce, with proper processing, fine stable carbide dispersions for enhanced high temperature creep-strength and rupture resistance, with similar or better oxidation/corrosion resistance. The project began with alloy 803, a Fe-25Cr-35NiTi,Nb alloy recently developed by INCO, as the base alloy for modification. Smaller commercial developmental alloy heats were produced by Special Metal. At the end of the project, three rounds of alloy development had produced a modified 803 alloy with significantly better creep resistance above 815 C (1500 C) than standard alloy 803 in the solution-annealed (SA) condition. The new upgraded 803 alloy also had the potential for a processing boost in that creep resistance for certain kinds of manufactured components that was not found in the standard alloy. The upgraded 803 alloy showed similar or slightly better oxidation and corrosion resistance relative to standard 803. Creep strength and oxidation/corrosion resistance of the upgraded 803 alloy were significantly better than found in alloy 800 H, as originally intended. The CRADA was terminated in February 2003. A contributing factor was Special Metals Corporation being in Chapter 11 Bankruptcy. Additional testing, further commercial scale-up, and any potential

  19. Elucidating the Effect of Alloying Elements on the Behavior of Austenitic Stainless Steels at Elevated Temperatures

    NASA Astrophysics Data System (ADS)

    Naghizadeh, Meysam; Mirzadeh, Hamed

    2016-12-01

    The effect of carbon and molybdenum on elevated temperature behavior of austenitic stainless steels was studied. It was revealed that carbon does not alter the overall grain coarsening behavior but molybdenum significantly retards the growth of grains toward higher temperatures and slower kinetics and effectively increases the grain growth activation energy due to an interaction energy between Mo and grain boundaries. These observations were based on especial activation energy plots, which facilitate the interpretation of results.

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

    SciTech Connect

    Vorobjev, A.N.; Porollo, S.I.; Konobeev, Yu.V.

    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.

  1. Structural transformations in austenitic stainless steel induced by deuterium implantation: irradiation at 100 K.

    PubMed

    Morozov, Oleksandr; Zhurba, Volodymyr; Neklyudov, Ivan; Mats, Oleksandr; Rud, Aleksandr; Chernyak, Nikolay; Progolaieva, Viktoria

    2015-01-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic stainless steel 18Cr10NiTi preimplanted at 100 K with deuterium ions in the dose range from 3 × 10(15) to 5 × 10(18) D/cm(2). The kinetics of structural transformation development in the implantation steel layer was traced from deuterium thermodesorption spectra as a function of implanted deuterium concentration. At saturation of austenitic stainless steel 18Cr10NiTi with deuterium by means of ion implantation, structural-phase changes take place, depending on the dose of implanted deuterium. The maximum attainable concentration of deuterium in steel is C = 1 (at.D/at.met. = 1/1). The increase in the implanted dose of deuterium is accompanied by the increase in the retained deuterium content, and as soon as the deuterium concentration attains C ≈ 0.5 the process of shear martensitic structural transformation in steel takes place. It includes the formation of bands, body-centered cubic (bcc) crystal structure, and the ferromagnetic phase. Upon reaching the deuterium concentration C > 0.5, the presence of these molecules causes shear martensitic structural transformations in the steel, which include the formation of characteristic bands, bcc crystal structure, and the ferromagnetic phase. At C ≥ 0.5, two hydride phases are formed in the steel, the decay temperatures of which are 240 and 275 K. The hydride phases are formed in the bcc structure resulting from the martensitic structural transformation in steel.

  2. Swelling suppression in phosphorous-modified Fe-Cr-Ni alloys during neutron irradiation

    SciTech Connect

    Lee, E.H.; Packan, N.H.

    1988-01-01

    Phosphorous-containing austenitic alloys in the solution annealed condition were irradiated at 745--760/degree/K. The alloys were variations on Fe--13Cr--15Ni--0.05P with respective additions of 0.8 Si, 0.2 Ti, or 0.8 Si /plus/ 0.2 Ti; also included were low (0.01) and zero P compositions (all values in wt. %). The reference ternary and the two phosphorous-only variations contained little precipitation and numerous voids and swelled rapidly, while the three variants containing P with Si and/or Ti showed little or no void formation and profuse phosphide precipitation. Results indicate that phosphorous in solution alone does not have a major influence on void swelling, whereas fine-scale phosphide precipitation is quite effective at eliminating void formation. The principal mechanism restricting swelling is the effect of the dense precipitate microstructure. These precipitates foster profuse cavity nucleation which in turn dilutes the helium atoms (and more time) in order for individual cavities to surpass their critical size and number of gas atoms necessary for subsequent growth as voids. This mechanism for swelling suppression was not found to be particularly sensitive to moderate variations in either the dislocation or cavity densities; the mechanism is strongest at elevated temperature where the critical quantities are large and is less effective at lower temperatures where the critical quantities are small. 19 refs., 10 figs., 3 tabs.

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

    NASA Astrophysics Data System (ADS)

    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.

  4. Irradiation-assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625

    SciTech Connect

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-07-01

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water. New data confirms previous results that high irradiation levels reduce SCC resistance in Alloy X-750. Low boron heats show improved IASCC (irradiation-assisted stress corrosion cracking). Alloy 625 is resistant to IASCC. Microstructural, microchemical, and deformation studies were carried out. Irradiation of X-750 caused significant strengthening and ductility loss associated with formation of cavities and dislocation loops. High irradiation did not cause segregation in X-750. Irradiation of 625 resulted in formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to loops and precipitates was apparently offset in 625 by partial dissolution of {gamma} precipitates. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in X-750, and the absence of these two effects results in superior IASCC resistance in 625.

  5. Ultra high vacuum fracture and transfer device for AES analysis of irradiated austenitic stainless steel

    SciTech Connect

    Urie, M.W.; Panayotou, N.F.; Robinson, J.E.

    1980-01-01

    An ultrahigh vacuum fracture and transfer device for analysis of irradiated and non-irradiated SS 316 fuel cladding is described. Mechanical property tests used to study the behavior of cladding during reactor transient over-power conditions are reported. The stress vs temperature curves show minimal differences between unirradiated cladding and unfueled cladding. The fueled cladding fails at a lower temperature. All fueled specimens failed in an intergranular mode. (FS)

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

    SciTech Connect

    Swindeman, R.W.; Ren, W.

    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.

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

    SciTech Connect

    Swindeman, R.W.; Ren, W.

    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.

  8. Irradiation assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625

    SciTech Connect

    Mills, W.J.; Lebo, M.R.; Bajaj, R.; Kearns, J.J.; Hoffman, R.C.; Korinko, J.J.

    1994-06-01

    In-reactor testing of bolt-loaded precracked compact tension specimens was performed in 360{degree}C water to determine effect of irradiation on the SCC behavior of HTH Alloy X-750 and direct aged Alloy 625. Out-of-flux and autoclave control specimens provided baseline data. Primary test variables were stress intensity factor, fluence, chemistry, processing history, prestrain. Results for the first series of experiments were presented at a previous conference. Data from two more recent experiments are compared with previous results; they confirm that high irradiation levels significantly reduce SCC resistance in HTH Alloy X-750. Heat-to-heat differences in IASCC were related to differences in boron content, with low boron heats showing improved SCC resistance. The in-reactor SCC performance of Alloy 625 was superior to that for Alloy X-750, as no cracking was observed in any Alloy 625 specimens even though they were tested at very high K{sub 1} and fluence levels. A preliminary SCC usage model developed for Alloy X-750 indicates that in-reactor creep processes, which relax stresses but also increase crack tip strain rates, and radiolysis effects accelerate SCC. Hence, in-reactor SCC damage under high flux conditions may be more severe than that associated with postirradiation tests. In addition, preliminary mechanism studies were performed to determine the cause of IASCC In Alloy X-750.

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

    SciTech Connect

    Swindeman, R.W.

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Gussev, Maxim N.; Field, Kevin G.; Briggs, Samuel A.; Yamamoto, Yukinori

    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

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

    DOEpatents

    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).

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

    DOEpatents

    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).

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

    SciTech Connect

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V.; Garner, F.A.

    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.

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

    SciTech Connect

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V.

    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.

  16. Development of Advanced Corrosion-Resistant Fe-Cr-Ni Austenitic Stainless Steel Alloy with Improved High Temperature Strenth and Creep-Resistance

    SciTech Connect

    Maziasz, PJ

    2004-09-30

    In February of 1999, a Cooperative Research and Development Agreement (CRADA) was undertaken between Oak Ridge National Laboratory (ORNL) and Special Metals Corporation-Huntington Alloys (formerly INCO Alloys International, Inc.) to develop a modified wrought austenitic stainless alloy with considerably more strength and corrosion resistance than alloy 800H or 800HT, but with otherwise similar engineering and application characteristics. Alloy 800H and related alloys have extensive use in coal flue gas environments, as well as for tubing or structural components in chemical and petrochemical applications. The main concept of the project was make small, deliberate elemental microalloying additions to this Fe-based alloy to produce, with proper processing, fine stable carbide dispersions for enhanced high temperature creep-strength and rupture resistance, with similar or better oxidation/corrosion resistance. The project began with alloy 803, a Fe-25Cr-35NiTi,Nb alloy recently developed by INCO, as the base alloy for modification. Smaller commercial developmental alloy heats were produced by Special Metals. At the end of the project, three rounds of alloy development had produced a modified 803 alloy with significantly better creep resistance above 815EC (1500EC) than standard alloy 803 in the solution-annealed (SA) condition. The new upgraded 803 alloy also had the potential for a processing boost in that creep resistance for certain kinds of manufactured components that was not found in the standard alloy. The upgraded 803 alloy showed similar or slightly better oxidation and corrosion resistance relative to standard 803. Creep strength and oxidation/corrosion resistance of the upgraded 803 alloy were significantly better than found in alloy 800H, as originally intended. The CRADA was terminated in February 2003. A contributing factor was Special Metals Corporation being in Chapter 11 Bankruptcy. Additional testing, further commercial scale-up, and any potential

  17. Effects of self-irradiation in plutonium alloys

    SciTech Connect

    Chung, B. W.; Lema, K. E.; Allen, P. G.

    2015-09-16

    In this paper, we present updated results of self-irradiation effects on 238Pu-enriched 239Pu alloys measured by immersion density, dilatometry, and tensile tests. We obtained the self-irradiation equivalent time of nearly 200 years, nearly 100 years longer than in our previous papers. At this extended aging, we find the rate of decrease in density has slowed significantly, stabilizing around 15.73 g/cc, without signs of void swelling. The volume expansion measured at 35°C also shows apparent saturation at less than 0.25%. Quasi-static tensile measurement still show gradual increase in the strength of plutonium alloys with age.

  18. Void-precipitate association during neutron irradiation of austenitic stainless steel

    SciTech Connect

    Pedraza, D.F.; Maziasz, P.J.

    1986-01-01

    Microstructural data has recently become available on a single heat of 316 stainless steel irradiated in EBR-II and HFIR, over a wide range of irradiation temperature (55 to 750/sup 0/C), dose (7 to 75 dpa), and helium generation rate (0.5 to 55 at. ppM He/dpa). Extensive information on precipitate compositions and characteristics are included. The data reveal several important relationships between the development of voids and precipitation. Precipitate associated voids dominate the swelling of (DO heat) 316 at 500 to 650 C from 8.4 to 36 dpa in EBR-II. Cold work (CW) or helium preinjection delay void formation in EBR-II. Higher helium generation in HFIR also delays void formation at 500 to 640/sup 0/C in SA 316 and CW DO heat 316. The delay persists in CW 316 at least to 61 dpa in HFIR, but abundant matrix and precipitate-associated voids form in SA after 47 dpa. In another heat of CW 316 (N-lot) irradiated in HFIR matrix and precipitate voids form readily after 22 to 44 dpa at 500 to 600/sup 0/C.

  19. Irradiation testing of high density uranium alloy dispersion fuels

    SciTech Connect

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.

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

    SciTech Connect

    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.

  1. Microstructural development of diffusion-brazed austenitic stainless steel to magnesium alloy using a nickel interlayer

    SciTech Connect

    Elthalabawy, Waled M.; Khan, Tahir I.

    2010-07-15

    The differences in physical and metallurgical properties of stainless steels and magnesium alloys make them difficult to join using conventional fusion welding processes. Therefore, the diffusion brazing of 316L steel to magnesium alloy (AZ31) was performed using a double stage bonding process. To join these dissimilar alloys, the solid-state diffusion bonding of 316L steel to a Ni interlayer was carried out at 900 deg. C followed by diffusion brazing to AZ31 at 510 deg. C. Metallographic and compositional analyses show that a metallurgical bond was achieved with a shear strength of 54 MPa. However, during the diffusion brazing stage B{sub 2} intermetallic compounds form within the joint and these intermetallics are pushed ahead of the solid/liquid interface during isothermal solidification of the joint. These intermetallics had a detrimental effect on joint strengths when the joint was held at the diffusion brazing temperature for longer than 20 min.

  2. Microstructural examination of irradiated vanadium alloys

    SciTech Connect

    Gelles, D.S.; Chung, H.M.

    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.

  3. Effect of silicon on stability of austenite during isothermal annealing of low-alloy steel with medium carbon content in the transition region between pearlitic and bainitic transformation

    NASA Astrophysics Data System (ADS)

    Jeníček, Š.; Vorel, I.; Káňa, J.; Ibrahim, K.; Kotěšovec, V.

    2017-02-01

    In a vast majority of steels, a prerequisite to successful heat treatment is the phase transformation of initial austenite to the desired type of microstructure which may consist of ferrite, pearlite, bainite, martensite or their combinations. Diffusion plays an important role in this phase transformation. Together with enthalpy and entropy, two thermodynamic quantities, diffusion represents the decisive mechanism for the formation of the particular phase. The basis of diffusion is the thermally-activated movement of ions of alloying and residual elements. It is generally known that austenite becomes more stable during isothermal treatment in the transitional region between pearlitic and bainitic transformation. This is due to thermodynamic processes which arise from the chemical composition of the steel. The transformation of austenite to pearlite or bainite is generally accompanied by formation of cementite. The latter can be suppressed by adding silicon to the steel because this element does not dissolve in cementite, and therefore prevents its formation. The strength of this effect of silicon depends mainly on the temperature of isothermal treatment. If a steel with a sufficient silicon content is annealed at a temperature, at which silicon cannot migrate by diffusion, cementite cannot form and austenite becomes stable for hours.

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

    SciTech Connect

    Brady, Michael P; Pint, Bruce A; Unocic, Kinga A; Yamamoto, Yukinori; Kumar, Deepak; Lipschutz, Mark D.

    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.

  5. Low-cycle fatigue of two austenitic alloys in hydrogen gas and air at elevated temperatures

    NASA Technical Reports Server (NTRS)

    Jaske, C. E.; Rice, R. C.

    1976-01-01

    The low-cycle fatigue resistance of type 347 stainless steel and Hastelloy Alloy X was evaluated in constant-amplitude, strain-controlled fatigue tests conducted under continuous negative strain cycling at a constant strain rate of 0.001 per sec and at total axial strain ranges of 1.5, 3.0, and 5.0 percent in both hydrogen gas and laboratory air environments in the temperature range 538-871 C. Elevated-temperature, compressive-strain hold-time experiments were also conducted. In hydrogen, the cyclic stress-strain behavior of both materials at 538 C was characterized by appreciable cyclic hardening at all strain ranges. At 871 C neither material hardened significantly; in fact, at 5% strain range 347 steel showed continuous cyclic softening until failure. The fatigue resistance of 347 steel was slightly higher than that of Alloy X at all temperatures and strain ranges. Ten-minute compressive hold time experiments at 760 and 871 C resulted in increased fatigue lives for 347 steel and decreased fatigue lives for Alloy X. Both alloys showed slightly lower fatigue resistance in air than in hydrogen. Some fractographic and metallographic results are also given.

  6. Correlation between mechanical properties and retained austenite characteristics in a low-carbon medium manganese alloyed steel plate

    SciTech Connect

    Chen, Jun; Lv, Mengyang; Tang, Shuai; Liu, Zhenyu; Wang, Guodong

    2015-08-15

    The effects of retained austenite characteristics on tensile properties and low-temperature impact toughness have been investigated by means of transmission electron microscopy and X-ray diffraction. It was found that only part of austenite phase formed during heat treating was left at room temperature. Moreover, the film-like retained austenite is displayed between bcc-martensite laths after heat treating at 600 °C, while the block-form retained austenite with thin hcp-martensite laths is observed after heat treating at 650 °C. It has been demonstrated that the film-like retained austenite possesses relatively high thermal and mechanical stability, and it can greatly improve low-temperature impact toughness, but its contribution to strain hardening capacity is limited. However, the block-form retained austenite can greatly enhance ultimate tensile strength and strain hardening capacity, but its contribution to low-temperature impact toughness is poor. - Highlights: • Correlation between retained austenite and impact toughness was elucidated. • The impact toughness is related to mechanical stability of retained austenite. • The effect of retained austenite on tensile and impact properties is inconsistent.

  7. Effect of cryogenic irradiation on NERVA structural alloys

    NASA Technical Reports Server (NTRS)

    Dixon, C. E.; Davidson, M. J.; Funk, C. W.

    1972-01-01

    Several alloys (Hastelloy X, AISI 347, A-286 bolts, Inconel 718, Al 7039-T63 and Ti-5Al-2.5Sn ELI) were irradiated in liquid nitrogen (140 R) to neutron fluences between 10 to the 17th power and 10 to the 19th power nvt (E greater than 1.0 Mev). After irradiation, tensile properties were obtained in liquid nitrogen without permitting any warmup except for some specimens which were annealed at 540 R. The usual trend of radiation damage typical for materials irradiated at and above room temperature was observed, such as an increase in strength and decrease in ductility. However, the damage at 140 R was greater because this temperature prevented the annealing of radiation-induced defects which occurs above 140 R.

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

    NASA Astrophysics Data System (ADS)

    Shim, Jae-Hyeok; Povoden-Karadeniz, Erwin; Kozeschnik, Ernst; Wirth, Brian D.

    2015-07-01

    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.

  9. Influence of laser irradiation on change properties of bulk amorphous Zr-Pd metallic alloys

    NASA Astrophysics Data System (ADS)

    Fedorov, V. A.; Yakovlev, A. V.; Pluzhnikova, T. N.; Shlikova, A. A.; Berezner, A. D.

    2017-01-01

    We study the morphological features of laser irradiation zones formed on the surface of the bulk metallic glasses. We use the nanoindentation method for estimation alloys properties caused by impulse heating during irradiation.

  10. Characterization of Nonmetallic Inclusions in High-Manganese and Aluminum-Alloyed Austenitic Steels

    NASA Astrophysics Data System (ADS)

    Park, Joo Hyun; Kim, Dong-Jin; Min, Dong Joon

    2012-07-01

    The effects of Al and Mn contents on the size, composition, and three-dimensional morphologies of inclusions formed in Fe- xMn- yAl ( x = 10 and 20 mass pct, y = 1, 3, and 6 mass pct) steels were investigated to enhance our understanding of the inclusion formation behavior in high Mn-Al-alloyed steels. By assuming that the alumina is a dominant oxide compound, the volume fraction of inclusions estimated from the chemical analysis, i.e., insoluble Al, in the Fe-Mn-3Al steels was larger than the inclusion volume fractions in the Fe-Mn-1Al and Fe-Mn-6Al steels. A similar tendency was found in the analysis of inclusions from a potentiostatic electrolytic extraction method. This finding could be explained from the terminal velocities of the compounds, which was affected by the thermophysical properties of Fe-Mn-Al steels. The inclusions formed in the Fe-Mn-Al-alloyed steels are classified into seven types according to chemistry and morphology: (1) single Al2O3 particle, (2) single AlN or AlON particle, (3) MnAl2O4 single galaxite spinel particle, (4) Al2O3(-Al(O)N) agglomerate, (5) single Mn(S,Se) particle, (6) oxide core with Mn(S,Se) skin (wrap), and (7) Mn(S,Se) core with Al2O3(-Al(O)N) aggregate (or bump). The Mn(S,Se) compounds were formed by the contamination of the steels by Se from the electrolytic Mn. Therefore, the raw materials (Mn) should be used carefully in the melting and casting processes of Fe-Mn-Al-alloyed steels.

  11. Irradiation-Induced Thermal Effects in Alloyed Metal Fuel of Fast Reactors

    NASA Astrophysics Data System (ADS)

    Kryukov, F. N.; Nikitin, O. N.; Kuzmin, S. V.; Belyaeva, A. V.; Gilmutdinov, I. F.; Grin, P. I.; Zhemkov, I. Yu

    2017-01-01

    The paper presents the results of studying alloyed metal fuel after irradiation in a fast reactor. Determined is the mechanism of fuel irradiation swelling, mechanical interaction between fuel and cladding, and distribution of fission products. Experience gained in fuel properties and behavior under irradiation as well as in irradiation-induced thermal effects occurred in alloyed metal fuel provides for a fuel pin design to have a burnup not less than 20% h. a.

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

    SciTech Connect

    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.

  13. Carbon--silicon coating alloys for improved irradiation stability

    DOEpatents

    Bokros, J.C.

    1973-10-01

    For ceramic nuclear fuel particles, a fission product-retaining carbon-- silicon alloy coating is described that exhibits low shrinkage after exposure to fast neutron fluences of 1.4 to 4.8 x 10/sup 21/ n/cm/sup 2/ (E = 0.18 MeV) at irradiation temperatures from 950 to 1250 deg C. Isotropic pyrolytic carbon containing from 18 to 34 wt% silicon is co-deposited from a gaseous mixiure of propane, helium, and silane at a temperature of 1350 to 1450 deg C. (Official Gazette)

  14. TEM Examination of Advanced Alloys Irradiated in ATR

    SciTech Connect

    Jian Gan, PhD

    2007-09-01

    Successful development of materials is critical to the deployment of advanced nuclear power systems. Irradiation studies of candidate materials play a vital role for better understanding materials performance under various irradiation environments of advanced system designs. In many cases, new classes of materials have to be investigated to meet the requirements of these advanced systems. For applications in the temperature range of 500 800ºC which is relevant to the fast neutron spectrum burner reactors for the Global Nuclear Energy Partnership (GNEP) program, oxide dispersion strengthened (ODS) and ferritic martensitic steels (e.g., MA957 and others) are candidates for advanced cladding materials. In the low temperature regions of the core (<600ºC), alloy 800H, HCM12A (also called T 122) and HT 9 have been considered.

  15. Irradiation-induced patterning in dilute Cu-Fe alloys

    NASA Astrophysics Data System (ADS)

    Stumphy, B.; Chee, S. W.; Vo, N. Q.; Averback, R. S.; Bellon, P.; Ghafari, M.

    2014-10-01

    Compositional patterning in dilute Cu1-xFex (x ≈ 12%) induced by 1.8 MeV Kr+ irradiation was studied as a function of temperature using atom probe tomography. Irradiation near room temperature led to homogenization of the sample, whereas irradiation at 300 °C and above led to precipitation and macroscopic coarsening. Between these two temperatures the irradiated alloys formed steady state patterns of composition where precipitates grew to a fixed size. The size in this regime increased somewhat with temperature. It was also observed that the steady state concentrations of Fe in Cu matrix and Cu in the Fe precipitates both greatly exceeded their equilibrium solubilities, with the degree of supersaturation in each phase decreasing with increasing temperature. In the macroscopic coarsening regime, the Fe-rich precipitates showed indications of a “cherry-pit” structure, with Cu precipitates forming within the Fe precipitates. In the patterning regime, interfaces between Fe-rich precipitates and the Cu-rich matrix were irregular and diffuse.

  16. Austenite Grain Structures in Ti- and Nb-Containing High-Strength Low-Alloy Steel During Slab Reheating

    NASA Astrophysics Data System (ADS)

    Roy, S.; Chakrabarti, D.; Dey, G. K.

    2013-02-01

    Austenite-grain growth was investigated in a couple of microalloyed steels, one containing Ti and the other containing Nb, Ti, and V, using different reheating temperatures between 1273 K and 1523 K (1000 °C and 1250 °C). Nature and distribution of microalloy precipitates were quantitatively analyzed before and after reheating. Interdendritic segregation (or microsegregation) during casting can result in an inhomogeneous distribution of microalloy precipitates in the as-cast slabs, which can create austenite grain size variation (even grain size bimodality) after reheating. Ti addition reduced the grain size variation; however, it could not eliminate the grain size bimodality in Nb-containing steel, due to the differential pinning effect of Nb precipitates. A model was proposed for the prediction of austenite grain size variation in reheated steel by combining different models on microsegregation during solidification, thermodynamic stability, and dissolution of microalloy precipitates and austenite grain growth during reheating.

  17. Kinetic evaluation of intergranular fracture in austenitic stainless steels

    SciTech Connect

    Simonen, E.P.; Bruemmer, S.M.

    1995-12-31

    A second, higher-dose threshold exists for irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels in non-oxidizing environments. The data supporting this concept have stimulated interest in the mechanical aspects of intergranular (IG) fracture. Cracking in a non-oxidizing environment suggests that mechanically-induced IG fracture may play an important role in the IASCC mechanism under these conditions. Radiation alters deformation processes in austenitic alloys and may influence the fracture mode during either in-situ or post-irradiation straining. Radiation effects that must be considered include radiation strengthening, radiation creep and radiation-induced flow localization. The present evaluation relates these radiation-induced phenomena to IG fracture relevant to IASCC. The evaluation indicates that radiation strengthening retards matrix deformation and allows intergranular fracture to occur at higher stresses and lower temperatures than expected for unirradiated stainless steel.

  18. Complex nanoprecipitate structures induced by irradiation in immiscible alloy systems

    NASA Astrophysics Data System (ADS)

    Shu, Shipeng; Bellon, P.; Averback, R. S.

    2013-04-01

    We investigate the fundamentals of compositional patterning induced by energetic particle irradiation in model A-B substitutional binary alloys using kinetic Monte Carlo simulations. The study focuses on a type of nanostructure that was recently observed in dilute Cu-Fe and Cu-V alloys, where precipitates form within precipitates, a morphology that we term “cherry-pit” structures. The simulations show that the domain of stability of these cherry-pit structures depends on the thermodynamic and kinetic asymmetry between the A and B elements. In particular, both lower solubilities and diffusivities of A in B compared to those of B in A favor the stabilization of these cherry-pit structures for A-rich average compositions. The simulation results are rationalized by extending the analytic model introduced by Frost and Russell for irradiation-induced compositional patterning so as to include the possible formation of pits within precipitates. The simulations indicate also that the pits are dynamical structures that undergo nearly periodic cycles of nucleation, growth, and absorption by the matrix.

  19. Bactericidal activity of copper and niobium-alloyed austenitic stainless steel.

    PubMed

    Baena, M I; Márquez, M C; Matres, V; Botella, J; Ventosa, A

    2006-12-01

    Biofouling and microbiologically influenced corrosion are processes of material deterioration that originate from the attachment of microorganisms as quickly as the material is immersed in a nonsterile environment. Stainless steels, despite their wide use in different industries and as appliances and implant materials, do not possess inherent antimicrobial properties. Changes in hygiene legislation and increased public awareness of product quality makes it necessary to devise control methods that inhibit biofilm formation or to act at an early stage of the biofouling process and provide the release of antimicrobial compounds on a sustainable basis and at effective level. These antibacterial stainless steels may find a wide range of applications in fields, such as kitchen appliances, medical equipment, home electronics, and tools and hardware. The purpose of this study was to obtain antibacterial stainless steel and thus mitigate the microbial colonization and bacterial infection. Copper is known as an antibacterial agent; in contrast, niobium has been demonstrated to improve the antimicrobial effect of copper by stimulating the formation of precipitated copper particles and its distribution in the matrix of the stainless steel. Thus, we obtained slides of 3.8% copper and 0.1% niobium alloyed stainless steel; subjected them to three different heat treatment protocols (550 degrees C, 700 degrees C, and 800 degrees C for 100, 200, 300, and 400 hours); and determined their antimicrobial activities by using different initial bacterial cell densities and suspending solutions to apply the bacteria to the stainless steels. The bacterial strain used in these experiments was Escherichia coli CCM 4517. The best antimicrobial effects were observed in the slides of stainless steel treated at 700 degrees C and 800 degrees C using an initial cell density of approximately 10(5) cells ml(-1) and phosphate-buffered saline as the solution in which the bacteria came into contact with

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

    SciTech Connect

    Hamilton, M.L.; Edwards, D.J.; Toloczko, M.B.

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  2. Effects of alloying additions and austenitizing treatments on secondary hardening and fracture behavior for martensitic steels containing both Mo and W

    NASA Astrophysics Data System (ADS)

    Lee, K. B.; Kwon, H.; Kwon, H.; Yang, H. R.

    2001-07-01

    The effects of alloying additions and austenitizing treatments on secondary hardening and fracture behavior of martensitic steels containing both Mo and W were investigated. The secondary hardening response and properties of these steels are dependent on the composition and distribution of the carbides formed during aging (tempering) of the martensite, as modified by alloying additions and austenitizing treatments. The precipitates responsible for secondary hardening are M2C carbides formed during the dissolution of the cementite (M3C). The Mo-W steel showed moderately strong secondary hardening and delayed overaging due to the combined effects of Mo and W. The addition of Cr removed secondary hardening by the stabilization of cementite, which inhibited the formation of M2C carbides. The elements Co and Ni, particularly in combination, strongly increased secondary hardening. Additions of Ni promoted the dissolution of cementite and provided carbon for the formation of M2C carbide, while Co increased the nucleation rate of M2C carbide. Fracture behavior is interpreted in terms of the presence of impurities and coarse cementite at the grain boundaries and the variation in matrix strength associated with the formation of M2C carbides. For the Mo-W-Cr-Co-Ni steel, the double-austenitizing at the relatively low temperatures of 899 to 816 °C accelerated the aging kinetics because the ratio of Cr/(Mo + W) increased in the matrix due to the presence of undissolved carbides containing considerably larger concentrations of (Mo + W). The undissolved carbides reduced the impact toughness for aging temperatures up to 510 °C, prior to the large decrease in hardness that occurred on aging at higher temperatures.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    DOE PAGES

    Bach, H. T.; Anderoglu, O.; Saleh, T. 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 cutmore » 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.« less

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

    SciTech Connect

    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.

  6. New insights to the promoted bainitic transformation in prior deformed austenite in a Fe-C-Mn-Si alloy

    NASA Astrophysics Data System (ADS)

    Hu, Hai-jiang; Xu, Guang; Zhou, Ming-xing; Yuan, Qing

    2017-02-01

    The varying trends of the amount and rate of bainitic transformation with strains at low temperature were investigated through metallography, X-ray diffraction and dilatometry. The results show that deformation at 573 K promotes bainitic transformation, whereas the promotion degree on bainite transformation by ausforming is nonlinear with strains. The amount of bainite in deformed austenite first increases and then decreases with the increase of strains. There exists a maximum value of the promotion effect corresponding to a critical small strain at a low temperature. Bainitic transformation rate can be increased by ausforming at low temperature, whereas a large strain weakens the acceleration effect. The amount of bainite in deformed materials is synthetically depended on the effect of enhanced nucleation and repressed growth. In addition, the volume fraction of retained austenite is not completely consistent with carbon content, indicating that ausforming plays a important role in determining the amount of austenite.

  7. Heavy ion irradiation induced dislocation loops in AREVA's M5® alloy

    NASA Astrophysics Data System (ADS)

    Hengstler-Eger, R. M.; Baldo, P.; Beck, L.; Dorner, J.; Ertl, K.; Hoffmann, P. B.; Hugenschmidt, C.; Kirk, M. A.; Petry, W.; Pikart, P.; Rempel, A.

    2012-04-01

    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® 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.

  8. Effects of neutron irradiation on deformation behavior of nickel-base fastener alloys

    SciTech Connect

    Bajaj, R.; Mills, W.J.; Kammenzind, B.F.; Burke, M.G.

    1999-07-01

    This paper presents the effects of neutron irradiation on the fracture behavior and deformation microstructure of high-strength nickel-base alloy fastener materials, Alloy X-750 and Alloy 625. Alloy X-750 in the HTH condition, and Alloy 625 in the direct aged condition were irradiated to a fluence of 2.4x10{sup 20} n/cm{sup 2} at 264 C in the Advanced Test Reactor. Deformation structures at low strains were examined. It was previously shown that Alloy X-750 undergoes hardening, a significant degradation in ductility and an increase in intergranular fracture. In contrast, Alloy 625 had shown softening with a concomitant increase in ductility and transgranular failure after irradiation. The deformation microstructures of the two alloys were also different. Alloy X-750 deformed by a planar slip mechanism with fine microcracks forming at the intersections of slip bands with grain boundaries. Alloy 625 showed much more homogeneous deformation with fine, closely spaced slip bands and an absence of microcracks. The mechanism(s) of irradiation assisted stress corrosion cracking (IASCC) are discussed.

  9. Effect of irradiation on the stress corrosion cracking behavior of Alloy X-750 and Alloy 625

    SciTech Connect

    Mills, W.J.; Lebo, M.R.; Kearns, J.J.; Hoffman, R.C.; Korinko, J.J.; Luther, R.F.; Sykes, G.B.

    1993-10-01

    In-reactor testing of bolt-loaded precracked and as-notched compact tension specimens was performed in 360{degrees}C water to determine effect of irradiation on SCC of Condition HTH and Condition BH Alloy X-750 and age-hardened Alloy 625. Variables were stress intensity factor (K{sub I}) level, fluence, grade of HTH material, prestraining and material chemistry. Effects of irradiation on high temperature SCC and the rapid cracking that occurs during cooldown below 150{degrees}C were characterized. Significant degradation in the in-reactor SCC resistance of HTH material was observed at initial K{sub I} levels above 30 MPa{radical}m and fluences greater than 10{sup 19} n/cm{sup 2} (E > 1 MeV). A small degradation in SCC resistance of HTH material was observed at low fluences (<10{sup 16} n/cm{sup 2}). As-notched specimens displayed less degradation in SCC resistance than precracked specimens. Prestraining greatly improved in-flux and out-of-flux SCC resistance of HTH material, as little or no SCC was observed in precracked specimens prestrained 20 to 30%, whereas extensive cracking was observed in nonprestrained specimens. Condition HTH heats with low boron (10 ppM or less) had improved in-reactor SCC resistance compared to heats with high and intermediate boron (>20 ppM). Age-hardened Alloy 625 exhibited superior in-reactor SCC behavior compared to HTH material as no crack extension occurred in any of the precracked Alloy 625 specimens tested at initial K{sub I} levels up to 80 MPa{radical}m.

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

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    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.

  11. Evaluation of critical resolved shear strength and deformation mode in proton-irradiated austenitic stainless steel using micro-compression tests

    NASA Astrophysics Data System (ADS)

    Jin, Hyung-Ha; Ko, Eunsol; Kwon, Junhyun; Hwang, Seong Sik; Shin, Chansun

    2016-03-01

    Micro-compression tests were applied to evaluate the changes in the strength and deformation mode of proton-irradiated commercial austenitic stainless steel. Proton irradiation generated small dots at low dose levels and Frank loops at high dose levels. The increase in critical resolved shear stresses (CRSS) was measured from micro-compression of pillars and the Schmid factor calculated from the measured loading direction. The magnitudes of the CRSS increase were in good agreement with the values calculated from the barrier hardening model using the measured size and density of radiation defects. The deformation mode changed upon increasing the irradiation dose level. At a low radiation dose level, work hardening and smooth flow behavior were observed. Increasing the dose level resulted in the flow behavior changing to a distinct heterogeneous flow, yielding a few large strain bursts in the stress-strain curves. The change in the deformation mode was related to the formation and propagation of defect-free slip bands. The effect of the orientation of the pillar or loading direction on the strengths is discussed.

  12. Disassembly of irradiated lithium-bonded capsules containing vanadium alloy specimens

    SciTech Connect

    Tsai, H.; Strain, R.V.

    1996-04-01

    Capsules containing vanadium alloy specimens from irradiation experiments in FFTF and EBR-II are being processed to remove the lithium bond and retrieve the specimens for testing. The work has progressed smoothly.

  13. HIGH TEMPERATURE BRAZING ALLOY FOR JOINT Fe-Cr-Al MATERIALS AND AUSTENITIC AND FERRITIC STAINLESS STEELS

    DOEpatents

    Cost, R.C.

    1958-07-15

    A new high temperature brazing alloy is described that is particularly suitable for brazing iron-chromiumaluminum alloys. It consists of approximately 20% Cr, 6% Al, 10% Si, and from 1.5 to 5% phosphorus, the balance being iron.

  14. Evaluation of hardening behaviors in ion-irradiated Fe-9Cr and Fe-20Cr alloys by nanoindentation technique

    NASA Astrophysics Data System (ADS)

    Li, Shilei; Wang, Yanli; Dai, Xianyuan; Liu, Fang; Li, Jinyu; Wang, Xitao

    2016-09-01

    The ion irradiation hardening behaviors of Fe-9 wt% Cr and Fe-20 wt% Cr model alloys were investigated by nanoindentation technique. The specimens were irradiated with 3 MeV Fe11+ ions at room temperature up to 1 and 5 dpa for Fe-9Cr alloy and 1 and 2.5 for Fe-20Cr alloy. The ratio of average hardness in the same depth of irradiated and unirradiated (Hirr. av/Hunirr. av) was used to determine the critical indentation depth hcrit to eliminate the softer substrate effect. The Nix-Gao model was used to explain the indentation size effect. Irradiation hardening is clearly observed in both Fe-9Cr alloy and Fe-20Cr alloy after ion irradiation. The differences of ISE and irradiation hardening behaviors between Fe-9Cr and Fe-20Cr alloys are considered to be due to their different microstructures and microstructural evolution under ion irradiation.

  15. Simulation of the elastic deformation of laser-welded joints of an austenitic corrosion-resistant steel and a titanium alloy with an intermediate copper insert

    NASA Astrophysics Data System (ADS)

    Pugacheva, N. B.; Myasnikova, M. V.; Michurov, N. S.

    2016-02-01

    The macro- and microstructures and the distribution of elements and of the values of the microhardness and contact modulus of elasticity along the height and width of the weld metal and heat-affected zone of austenitic corrosion-resistant 12Kh18N10T steel (Russian analog of AISI 321) and titanium alloy VT1-0 (Grade 2) with an intermediate copper insert have been studied after laser welding under different conditions. The structural inhomogeneity of the joint obtained according to one of the regimes selected has been shown: the material of the welded joint represents a supersaturated solid solution of Fe, Ni, Cr, and Ti in the crystal lattice of copper with a uniformly distributed particles of intermetallic compounds Ti(Fe,Cr) and TiCu3. At the boundaries with steel and with the titanium alloy, diffusion zones with thicknesses of 0.1-0.2 mm are formed that represent supersaturated solid solutions based on iron and titanium. The strength of such a joint was 474 MPa, which corresponds to the level of strength of the titanium alloy. A numerical simulation of the mechanical behavior of welded joints upon the elastic tension-compression has been performed taking into account their structural state, which makes it possible to determine the amplitude values of the deformations of the material of the weld.

  16. Fine structure characterization of martensite/austenite constituent in low-carbon low-alloy steel by transmission electron forward scatter diffraction.

    PubMed

    Li, C W; Han, L Z; Luo, X M; Liu, Q D; Gu, J F

    2016-11-01

    Transmission electron forward scatter diffraction and other characterization techniques were used to investigate the fine structure and the variant relationship of the martensite/austenite (M/A) constituent of the granular bainite in low-carbon low-alloy steel. The results demonstrated that the M/A constituents were distributed in clusters throughout the bainitic ferrite. Lath martensite was the main component of the M/A constituent, where the relationship between the martensite variants was consistent with the Nishiyama-Wassermann orientation relationship and only three variants were found in the M/A constituent, suggesting that the variants had formed in the M/A constituent according to a specific mechanism. Furthermore, the Σ3 boundaries in the M/A constituent were much longer than their counterparts in the bainitic ferrite region. The results indicate that transmission electron forward scatter diffraction is an effective method of crystallographic analysis for nanolaths in M/A constituents.

  17. Effect of yttrium on martensite-austenite phase transformation temperatures and high temperature oxidation kinetics of Ti-Ni-Hf high-temperature shape memory alloys

    NASA Astrophysics Data System (ADS)

    Kim, Jeoung Han; Kim, Kyong Min; Yeom, Jong Taek; Young, Sung

    2016-03-01

    The effect of yttrium (< 5.5 at%) on the martensite-austenite phase transformation temperatures, microstructural evolution, and hot workability of Ti-Ni-Hf high-temperature shape memory alloys is investigated. For these purposes, differential scanning calorimetry, hot compression, and thermo-gravimetric tests are conducted. The phase transformation temperatures are not noticeably influenced by the addition of yttrium up to 4.5 at%. Furthermore, the hot workability is not significantly affected by the yttrium addition up to 1.0 at%. However, when the amount of yttrium addition exceeds 1.0 at%, the hot workability deteriorates significantly. In contrast, remarkable improvement in the high temperature oxidation resistance due to the yttrium addition is demonstrated. The total thickness of the oxide layers is substantially thinner in the Y-added specimen. In particular, the thickness of (Ti,Hf) oxide layer is reduced from 200 µm to 120 µm by the addition of 0.3 at% Y.

  18. Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

    DOE PAGES

    Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; ...

    2015-07-14

    The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition.more » Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.« less

  19. Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

    SciTech Connect

    Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; Yamamoto, Yukinori; Snead, Lance Lewis

    2015-07-14

    The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

  20. Damage accumulation in ion-irradiated Ni-based concentrated solid-solution alloys

    DOE PAGES

    Ullah, Mohammad W.; Aidhy, Dilpuneet S.; Zhang, Yanwen; ...

    2016-01-01

    We investigate Irradiation-induced damage accumulation in Ni0.8Fe0.2 and Ni0.8Cr0.2 alloys by using molecular dynamics simulations to assess possible enhanced radiation-resistance in these face-centered cubic (fcc), single-phase, concentrated solid-solution alloys, as compared with pure fcc Ni.

  1. Critical behavior and exponent parameters of the austenitic phase in Ni50- x Pr x Mn37Sn13 alloys with x = 1

    NASA Astrophysics Data System (ADS)

    Phan, T. L.; Manh, T. V.; Ho, T. A.; Yu, S. C.; Dan, N. H.; Yen, N. H.; Thanh, T. D.

    2014-06-01

    We fabricated Huesler alloy ingots of Ni50- x Pr x Mn37Sn13 with x = 0 - 5 by using an arcmelting method. Crystalline-structural analyses revealed the coexistence of austenitic and martensitic phases in the samples with x = 0 and 1, in which the volume fraction of the austenitic phase for x = 1 was higher than that for x = 0. With higher Pr concentrations, x > 1, Pr- and Ni3Sn-related secondary phases, which reduced the magnetic order of the alloys, were formed. Thus, only the sample with x = 1 was more suitable for studying the critical behavior. Based on Landau's phase-transition theory and Banerjee's criteria, we found that this sample undergoes a second-order magnetic phase transition (SOMT) at a temperature around the Curie temperature T C ≈ 299 K. Using the modified Arrott plots, asymptotic relations, and a universal scaling law, we determined the values of the critical exponents β = 0.501 ± 0.009 and γ = 1.045 ± 0.006. These values are very close to those expected for the mean-field theory with β = 0.5 and γ = 1, proving the existence of long-range ferromagnetic (FM) order in the sample with x = 1. Particularly, around at temperature T C , the magnetic-entropy change reaches the maximum value (∆ S max ). Its magnetic-field dependences can be described by using a power law |∆ S max | ∝ H n , where n = 0.687 is close to the value 0.677 calculated from the theoretical relation n = 1 + ( β - 1)/( β + γ). We believe that the doping of a suitable Pr amount in Ni50- x Pr x Mn37Sn13 ( x ≈ 1) promotes the formation of the austenitic phase and results in long-range FM order. However, the persistence of the martensitic phase and secondary phases favors short-range FM order and thus decreases the FM order in Ni50- x Pr x Mn37Sn13.

  2. Effects of alloying elements on the formation of < c >-component loops in Zr alloy Excel under heavy ion irradiation.

    SciTech Connect

    Idrees, Yasir; Francis, Elisabeth M.; Yao, Zhongwen; Korinek, Andreas; Kirk, Marquis A.; Sattari, Mohammad; Preuss, Michael; Daymond, M. R.

    2015-05-14

    We report here the microstructural changes occurring in the zirconium alloy Excel (Zr-3.5 wt% Sn-0.8Nb-0.8Mo-0.2Fe) during heavy ion irradiation. In situ irradiation experiments were conducted at reactor operating temperatures on two Zr Excel alloy microstructures with different states of alloying elements, with the states achieved by different solution heat treatments. In the first case, the alloying elements were mostly concentrated in the beta (beta) phase, whereas, in the second case, large Zr-3(Mo,Nb,Fe)(4) secondary phase precipitates (SPPs) were grown in the alpha (alpha) phase by long term aging. The heavy ion induced damage and resultant compositional changes were examined using transmission electron microscopy (TEM) in combination with scanning transmission electron microscope (STEM)-energy dispersive x-ray spectroscopy (EDS) mapping. Significant differences were seen in microstructural evolution between the two different microstructures that were irradiated under similar conditions. Nucleation and growth of < c >-component loops and their dependence on the alloying elements are a major focus of the current investigation. It was observed that the < c >-component loops nucleate readily at 100, 300, and 400 degrees C after a threshold incubation dose (TID), which varies with irradiation temperature and the state of alloying elements. It was found that the TID for the formation of < c >-component loops increases with decrease in irradiation temperature. Alloying elements that are present in the form of SPPs increase the TID compared to when they are in the beta phase solid solution. Dose and temperature dependence of loop size and density are presented. Radiation induced redistribution and clustering of alloying elements (Sn, Mo, and Fe) have been observed and related to the formation of < c >-component loops. It has been shown that at the higher temperature tests, irradiation induced dissolution of precipitates occurs whereas irradiation induced

  3. Evaluation of irradiation hardening of ion-irradiated V-4Cr-4Ti and V-4Cr-4Ti-0.15Y alloys by nanoindentation techniques

    NASA Astrophysics Data System (ADS)

    Miyazawa, Takeshi; Nagasaka, Takuya; Kasada, Ryuta; Hishinuma, Yoshimitsu; Muroga, Takeo; Watanabe, Hideo; Yamamoto, Takuya; Nogami, Shuhei; Hatakeyama, Masahiko

    2014-12-01

    Irradiation hardening behavior of V-4Cr-4Ti and V-4Cr-4Ti-0.15Y alloys after Cu-ion beam irradiation were investigated with a combination between nanoindentation techniques and finite element method (FEM) analysis. The ion-irradiation experiments were conducted at 473 K with 2.4 MeV Cu2+ ions up to 7.6 dpa. For the unirradiated materials, the increase in nanoindentation hardness with decreasing indentation depth, so-called indentation size effect (ISE), was clearly observed. After irradiation, irradiation hardening in the measured depth was identified. Hardening behavior of bulk-equivalent hardness for V-4Cr-4Ti-0.15Y alloy was similar to that for V-4Cr-4Ti alloy. Y addition has little effect on irradiation hardening at 473 K. Adding the concept of geometrically necessary dislocations (GNDs) to constitutive equation of V-4Cr-4Ti alloy, the ISE was simulated. A constant value of α = 0.5 was derived as an optimal value to simulate nanoindentation test for ion-irradiated V-4Cr-4Ti alloy. Adding the term of irradiation hardening Δσirrad. to constitutive equation with α = 0.5, FEM analyses for irradiated surface of V-4Cr-4Ti alloy were carried out. The analytic data of FEM analyses based on neutron-irradiation hardening equivalent to 3.0 dpa agreed with the experimental data to 0.76 dpa. The comparison indicates that irradiation hardening by heavy ion-irradiation is larger than that by neutron-irradiation at the same displacement damage level. Possible mechanisms for extra hardening by heavy ion-irradiation are the processes that the injected Cu ions could effectively produce irradiation defects such as interstitials compared with neutrons, and that higher damage rate of ion-irradiation enhanced nucleation of irradiation defects and hence increased the number density of the defects compared with neutron-irradiation.

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

    SciTech Connect

    Fabritsiev, S.A.; Zinkle, S.J.; Rowcliffe, A.F.

    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.

  5. Modelling of the effect of dislocation channel on intergranular microcrack nucleation in pre-irradiated austenitic stainless steels during low strain rate tensile loading

    NASA Astrophysics Data System (ADS)

    Evrard, Pierre; Sauzay, Maxime

    2010-10-01

    In the present article, the effect of dislocation channel on intergranular microcrack nucleation during the tensile deformation of pre-irradiated austenitic stainless steels is studied. Because several slip planes are activated within the dislocation channel, the simple dislocation pile-up model seems not well suited to predict grain boundary stress field. Finite element computations, using crystal plasticity laws and meshes including a channel of finite thickness, are also performed in order to study the effect of some microstructural characteristics on grain boundary stress field. Numerical results show that: the thickness and the length of the dislocation channel influence strongly the grain boundary normal stress field. The grain boundary orientation with respect the stress axis does not affect so much the grain boundary normal stresses close to the dislocation channel. On the contrary far away the dislocation channel, the grain boundary stress field depends on the grain boundary orientation. Based on these numerical results, an analytical model is proposed to predict grain boundary stress fields. It is valuable for large ranges of dislocation channel thickness, length as well as applied stress. Then, a macroscopic microcrack nucleation criterion is deduced based on the elastic-brittle Griffith model. The proposed criterion predicts correctly the influence of grain boundary characteristics (low-angle boundaries (LABs), non-coincident site lattice (non-CSL) high-angle boundaries (HABs), special grain boundaries (GBs)) on intergranular microcrack nucleation and the macroscopic tensile stress required for grain boundary microcrack nucleation for pre-irradiated austenitic stainless steels deformed in argon environment. The criterion based on a dislocation pile-up model (Smith and Barnby) underestimates strongly the nucleation stress. These results confirm that pile-up models are not well suited to predict microcrack nucleation stress in the case of dislocation

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

    NASA Astrophysics Data System (ADS)

    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.; Zhou, X. T.; Zhu, Z. Y.

    2016-06-01

    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.

  7. Helium generation rates in isotopically tailored Fe-Cr-Ni alloys irradiated in FFTF/MOTA

    SciTech Connect

    Greenwood, L.R.; Garner, F.A.; Oliver, B.M.

    1991-11-01

    Three Fe-Cr-Ni alloys have been doped with 0.4% {sup 59}Ni for side-by-side irradiations of doped and undoped materials in order to determine the effects of fusion-relevant levels of helium production on microstructural development and mechanical properties. The alloys were irradiated in three successive cycles of the Materials Open Test Assembly (MOTA) located in the Fast Flux Test Facility (FFTF). Following irradiation, helium levels were measured by isotope dilution mass spectrometry. The highest level of helium achieved in doped alloys was 172 appm at 9.1 dpa for a helium(appm)-to-dpa ratio of 18.9. The overall pattern of predicted helium generation rates in doped and undoped alloys is in good agreement with the helium measurements.

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

    NASA Astrophysics Data System (ADS)

    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.

  9. Magnetic patterning using ion irradiation for highly ordered CoPt alloys with perpendicular anisotropy

    SciTech Connect

    Abes, M.; Venuat, J.; Muller, D.; Carvalho, A.; Schmerber, G.; Beaurepaire, E.; Dinia, A.; Pierron-Bohnes, V.

    2004-12-15

    We used a combination of ion irradiation and e-beam lithography to magnetically pattern an ordered CoPt alloy with strong perpendicular magnetic anisotropy. Ion irradiation disorders the alloy and strongly reduces the magnetic anisotropy. Magnetic force microscopy showed a regular array of 1 {mu}m{sup 2} square dots with perpendicular anisotropy separated by 1 {mu}m large ranges with in-plane anisotropy. This is further confirmed by magnetic measurements, which showed that arrays protected by a 200 nm Pt layer present the same coercive field and the same perpendicular anisotropy as before irradiation. This is promising for applications in magnetic recording technologies.

  10. Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

    NASA Astrophysics Data System (ADS)

    Jin, K.; Bei, H.; Zhang, Y.

    2016-04-01

    In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 1013 to 5 × 1015 ions cm-2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. With continuously increasing the ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.

  11. Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

    SciTech Connect

    Jin, Ke; Zhang, Yanwen; Bei, Hongbin

    2016-01-01

    In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 1013 to 5 × 1015 ions cm–2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. With continuously increasing the ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Here, under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.

  12. Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

    DOE PAGES

    Jin, Ke; Zhang, Yanwen; Bei, Hongbin

    2016-01-01

    In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 1013 to 5 × 1015 ions cm–2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. With continuously increasing themore » ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Here, under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.« less

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

    DOEpatents

    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.

  14. Microstructural evolution in nickel alloy C-276 after Ar-ion irradiation at elevated temperature

    SciTech Connect

    Jin, Shuoxue; He, Xinfu; Li, Tiecheng; Ma, Shuli; Tang, Rui; Guo, Liping

    2012-10-15

    In present work, the irradiation damage in nickel-base alloy C-276 irradiated with Ar-ions was studied. Specimens of C-276 alloy were subjected to an irradiation of Ar-ions (with 120 keV) to dose levels of 6 and 10 dpa at 300 and 550 Degree-Sign C, respectively. The size distributions and densities of dislocation loops caused by irradiation were investigated with transmission electron microscopy. Irradiation hardening due to the formation of the loops was calculated using the dispersed barrier-hardening model, showing that irradiation hardening was greatest at 300 Degree-Sign C/6 dpa. The microstructure evolution induced by Ar-ion irradiation (0-10 dpa) in nickel-base alloy C-276 has been studied using a multi-scale modeling code Radieff constructed based on rate theory, and the size of dislocation loops simulated by Radieff was in good agreement with the experiment. - Highlights: Black-Right-Pointing-Pointer High density of dislocation loops appeared after Ar ions irradiation. Black-Right-Pointing-Pointer Irradiation hardening due to the formation of loops was calculated by the DBH model. Black-Right-Pointing-Pointer Size of loops simulated by Radieff was in good agreement with the experiment.

  15. ATR-A1 irradiation experiment on vanadium alloys and low activation steels

    SciTech Connect

    Tasi, H.; Strain, R.V.; Gomes, I.; Hins, A.G.; Smith, D.L.

    1996-04-01

    To study the mechanical properties of vanadium alloys under neutron irradiation at low temperatures, an experiment was designed and constructed for irradiation in the Advanced Test Reactor (ATR). The experiment contained Charpy, tensile, compact tension, TEM, and creep specimens of vanadium alloys. It also contained limited low-activation ferritic steel specimens as part of the collaborative agreement with Monbusho of Japan. The design irradiation temperatures for the vanadium alloy specimens in the experiment are {approx}200 and 300{degrees}C, achieved with passive gap-gap sizing and fill gas blending. To mitigate vanadium-to-chromium transmutation from the thermal neutron flux, the test specimens are contained inside gadolinium flux filters. All specimens are lithium-bonded. The irradiation started in Cycle 108A (December 3, 1995) and is expected to have a duration of three ATR cycles and a peak influence of 4.4 dpa.

  16. Hydrogen Release from Irradiated Vanadium Alloy V-4Cr-4Ti

    SciTech Connect

    Klepikov, A. Kh.; Romanenko, O. G.; Chikhray, E. V.; Tazhibaeva, I. L.; Shestakov, V. P.; Longhurst, Glen Reed

    1999-09-01

    The present work is an attempt to obtain data concerning the influence of neutron and ? irradiation upon hydrogen retention in V-4Cr-4Ti vanadium alloy. The experiments on in-pile loading of vanadium alloy specimens at the neutron flux density 1014 n/cm2s, hydrogen pressure of 80 Pa, and temperatures of 563, 613, and 773 K were carried out using the IVG.1M reactor of the Kazakhstan National Nuclear Center. A preliminary set of loading/degassing experiments with non-irradiated material has been carried out to obtain data on hydrogen interaction with vanadium alloy. The, data presented in this work are related both to non-irradiated and irradiated samples.

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

    SciTech Connect

    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.

  18. Neutron-irradiated model alloys and pressure-vessel steels studied using positron spectroscopy

    NASA Astrophysics Data System (ADS)

    Cumblidge, Stephen Eric

    We have used positron-annihilation-lifetime spectroscopies to examine microstructural evolution of pressure vessel steels and model alloys that have systematically varied amounts of copper, nickel, and phosphorus during neutron irradiation and post-irradiation annealing. The objective of this work was to characterize the neutron-irradiation induced microstructural features that cause the embrittlement of nuclear reactor pressure-vessel steel. We used positron annihilation lifetime spectroscopy and Doppler-broadening spectroscopy to examine the model alloys and pressure-vessel steels before and after irradiation and after post-irradiation annealing. We followed the changes in the mechanical properties of the materials using Rockwell 15N hardness measurements. The results show that in both the model alloys and pressure-vessel steels neutron irradiation causes the formation of vacancy-type defect clusters and a fine distribution of copper- and nickel-enriched metallic precipitates. The vacancy clusters are small in size and were present in all samples, and disappear upon annealing at 450°C. The metallic precipitates are present only in the model alloy samples with either high Cu or a combination of medium Cu and high Ni, and they remain in the microstructure after annealing up to 550°C, starting to anneal possibly at 600°C. The neutron-irradiated pressure vessel steels behave similarly to the high Cu samples, indicating that neutron irradiation induced precipitation occurs in these alloys as well. This work provides independent evidence for the irradiation-induced metallic precipitates seen by other techniques, gives evidence for the exact nature of the matrix damage, and is significant to understanding the in-service degradation of pressure vessel materials.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    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.

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

    SciTech Connect

    Loomis, B.A.; Chung, H.M.; Smith, D.L.

    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.

  2. A comparative study of the in vitro corrosion behavior and cytotoxicity of a superferritic stainless steel, a Ti-13Nb-13Zr alloy, and an austenitic stainless steel in Hank's solution.

    PubMed

    Assis, S L; Rogero, S O; Antunes, R A; Padilha, A F; Costa, I

    2005-04-01

    In this study, the in vitro corrosion resistance of a superferritic stainless steel in naturally aerated Hank's solution at 37 degrees C has been determined to evaluate the steel for use as a biomaterial. The potentiodynamic polarization method and electrochemical impedance spectroscopy (EIS) were used to determine the corrosion resistance. The polarization results showed very low current densities at the corrosion potential and electrochemical behavior typical of passive metals. At potentials above 0.75 V (SCE), and up to that of the oxygen evolution reaction, the superferritic steel exhibited transpassive behavior followed by secondary passivation. The superferritic stainless steel exhibited high pitting resistance in Hank's solution. This steel did not reveal pits even after polarization to 3000 mV (SCE). The EIS results indicated high impedance values at low frequencies, supporting the results obtained from the polarization measurements. The results obtained for the superferritic steel have been compared with those of the Ti-13Nb-13Zr alloy and an austenitic stainless steel, as Ti alloys are well known for their high corrosion resistance and biocompatibility, and the austenitic stainless steel is widely used as an implant material. The cytotoxicity tests indicated that the superferritic steel, the austenitic steel, and the Ti-13Nb-13Zr alloy were not toxic. Based on corrosion resistance and cytotoxicity results, the superferritic stainless steel can be considered as a potential biomaterial.

  3. Effect of preliminary irradiation on the bond strength between a veneering composite and alloy.

    PubMed

    Matsumoto, Yoshifumi; Furuchi, Mika; Oshima, Akiko; Tanoue, Naomi; Koizumi, Hiroyasu; Matsumura, Hideo

    2010-01-01

    The shear bond strength of a veneering composite (Solidex) and silver-palladium-copper-gold alloy (Castwell M.C.12) was evaluated for different duration times and irradiance for preliminary photo-polymerization. A veneering composite was applied onto a cast disk. Preliminary photo irradiation was performed using different duration times or irradiance. After final polymerization, the bond strength and the spectral distribution of each curing unit were determined. Shear bond strength was significantly higher for 90 s (12.4 MPa), than that for 0 s (8.3 MPa). With regard to the effect of irradiance, that from Solidilite (11.4 MPa) was significantly higher than that from Sublite S at 3 cm (8.7 MPa). The irradiance of Hyper LII and Sublite S at 3 cm was higher than Sublite S at 15 cm or Solidilite unit. Long time irradiation and low intensity is effective for preliminary irradiation in order to enhance the bond strength.

  4. Dose dependence of mechanical properties in tantalum and tantalum alloys after low temperature irradiation

    SciTech Connect

    Byun, Thak Sang

    2008-01-01

    The dose dependence of mechanical properties was investigated for tantalum and tantalum alloys after low temperature irradiation. Miniature tensile specimens of three pure tantalum metals, ISIS Ta, Aesar Ta1, Aesar Ta2, and one tantalum alloy, Ta-1W, were irradiated by neutrons in the High Flux Isotope Reactor (HFIR) at ORNL to doses ranging from 0.00004 to 0.14 displacements per atom (dpa) in the temperature range 60 C 100 oC. Also, two tantalum-tungsten alloys, Ta-1W and Ta-10W, were irradiated by protons and spallation neutrons in the LANSCE facility at LANL to doses ranging from 0.7 to 7.5 dpa and from 0.7 to 25.2 dpa, respectively, in the temperature range 50 C 160 oC. Tensile tests were performed at room temperature and at 250oC at nominal strain rates of about 10-3 s-1. All neutron-irradiated materials underwent progressive irradiation hardening and loss of ductility with increasing dose. The ISIS Ta experienced embrittlement at 0.14 dpa, while the other metals retained significant necking ductility. Such a premature embrittlement in ISIS Ta is believed to be because of high initial oxygen concentrations picked up during a pre-irradiation anneal. The Ta-1W and Ta-10W specimens irradiated in spallation condition experienced prompt necking at yield since irradiation doses for those specimens were high ( 0.7 dpa). At the highest dose, 25.2 dpa, the Ta-10W alloy specimen broke with little necking strain. Among the test materials, the Ta-1W alloy displayed the best combination of strength and ductility. The plastic instability stress and true fracture stress were nearly independent of dose. Increasing test temperature decreased strength and delayed the onset of necking at yield.

  5. TEM observations and finite element modelling of channel deformation in pre-irradiated austenitic stainless steels - Interactions with free surfaces and grain boundaries

    NASA Astrophysics Data System (ADS)

    Sauzay, Maxime; Bavard, Karine; Karlsen, Wade

    2010-11-01

    Transmission electron microscopy (TEM) observations show that dislocation channel deformation occurs in pre-irradiated austenitic stainless steels, even at low stress levels (˜175 MPa, 290 °C) in low neutron dose (˜0.16 dpa, 185 °C) material. The TEM observations are utilized to design finite element (FE) meshes that include one or two "soft" channels (i.e. low critical resolved shear stress (CRSS)) of particular aspect ratio (length divided by thickness) embedded at the free surface of a "hard" matrix (i.e. high CRSS). The CRSS are adjusted using experimental data and physically based models from the literature. For doses leading to hardening saturation, the computed surface slips are as high as 100% for an applied stress close to the yield stress, when the observed channel aspect ratio is used. Surface slips are much higher than the grain boundary slips because of matrix constraint effect. The matrix CRSS and the channel aspect ratio are the most influential model parameters. Predictions based on an analytical formula are compared with surface slips computed by the FE method. Predicted slips, either in surface or bulk channels, agree reasonably well with either atomic force microscopy measures reported in the literature or measures based on our TEM observations. Finally, it is shown that the induced surface slip and grain boundary stress concentrations strongly enhance the kinetics of the damage mechanisms possibly involved in IASCC.

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

    SciTech Connect

    Wakai, E.; Hashimoto, N.; Gibson, L.T.

    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.

  7. Cr precipitation in neutron irradiated industrial purity Fe-Cr model alloys

    NASA Astrophysics Data System (ADS)

    Kuksenko, V.; Pareige, C.; Pareige, P.

    2013-01-01

    The microstructure of four neutron irradiated Fe-Cr model alloys of industrial purity (Fe-2.5%Cr, Fe-5%Cr, Fe-9%Cr and Fe-12%Cr) has been characterized by atom probe tomography (APT). Irradiation has been performed at 300 °C up to 0.6 dpa in MTR reactor. APT investigations confirmed the enhanced precipitation of α' clusters as these clusters have only been observed in supersaturated model alloys. In addition a nonexpected family of clusters has been revealed due to irradiation induced segregation of impurities: NiSiPCr-enriched clusters. They might be associated to defect clusters invisible by transmission electron microscopy (TEM). A quantitative description of these objects is presented in this paper and results are compared with TEM and SANS data of the literature obtained on the same model alloy.

  8. Effects of irradiation temperature and dose rate on the mechanical properties of self-ion implanted Fe and Fe-Cr alloys

    NASA Astrophysics Data System (ADS)

    Hardie, Christopher D.; Williams, Ceri A.; Xu, Shuo; Roberts, Steve G.

    2013-08-01

    type defects; and defect migration in a material depends strongly on the presence of impurity atoms [3]. In iron the migration energy for a vacancy is 0.67 eV and that of an interstitial 0.34 eV; carbon forms strongly bound complexes with vacancies and a vacancy-carbon complex has migration energy of 1.08 eV [3]. Depending on temperature, this may result in unequal fluxes of mobile interstitials and vacancies, known as a production bias [4], and thus influence the relative fractions of the various reaction paths described above.Variations in the fraction of reaction paths with dose rate have been inferred from the swelling and creep behaviour of several materials [2,5,6]. The majority of research regarding the dose rate dependence of radiation damage was conducted by using fission reactors in the 1980s, which focused on the swelling and creep rates of austenitic stainless steels and their variation with neutron flux. For example, swelling of 316 stainless steel cladding and the creep rate of numerous steels under irradiation has been shown to decrease with increasing dose rate [7,8]. This decrease in swelling and creep is believed to be due to a reduction in the density of active point defects, resulting from heightened rates of defect clustering or a higher fraction of recombination [6]. Muroga et al. [9] compared the saturated dislocation loop densities in Fe-Cr-Ni austenitic alloys after irradiation with high flux electron, fast neutron and fusion D-T neutron sources, showing a considerable increase in saturated dislocation loop density in the irradiated alloys as the dose rate increased from ˜10-9 dpa/s (D-T neutron source) to ˜10-4 dpa/s (electrons). In a subsequent investigation, Fe-15Cr-16Ni irradiated with 4 MeV nickel ions at a dose rate of 10-4 dpa/s exhibited even higher loop densities than those from high energy electron irradiation at a comparable dose rate [10]. This difference may be attributed to heightened rates of defect recombination, resulting

  9. Thermal properties of U-Mo alloys irradiated to moderate burnup and power

    NASA Astrophysics Data System (ADS)

    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 3.30 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.52 × 1021 fissions cm-3 from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

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

    SciTech Connect

    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.

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

    PubMed Central

    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

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

    PubMed

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

    2016-08-26

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  14. Damage accumulation in ion-irradiated Ni-based concentrated solid-solution alloys

    SciTech Connect

    Ullah, Mohammad W.; Aidhy, Dilpuneet S.; Zhang, Yanwen; Weber, William J.

    2016-01-01

    We investigate Irradiation-induced damage accumulation in Ni0.8Fe0.2 and Ni0.8Cr0.2 alloys by using molecular dynamics simulations to assess possible enhanced radiation-resistance in these face-centered cubic (fcc), single-phase, concentrated solid-solution alloys, as compared with pure fcc Ni.

  15. Charpy impact properties of low activation alloys for fusion applications after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Rieth, M.; Dafferner, B.; Röhrig, H. D.

    1996-10-01

    The MANITU irradiation and fracture-toughness testing program although initially foreseen to clarify the early dose-saturation of ΔDBTT for commercial ferritic steels has been extended to include the medium temperature (≥ 250°C) irradiation hardening behaviour of promising low-activation alloys. The results after a first 0.8 dpa irradiation clearly show a much better behaviour of the new alloys in any respect (e.g. DBTT after irradiation always below +50°C for subsize specimens, for the ORNL steel even below -20°C). The complexity of temperature dependency is probably caused by the transition range in dose accumulation, and should therefore not be 'over-interpreted'.

  16. 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

    NASA Astrophysics Data System (ADS)

    Budylkin, N. I.; Bulanova, T. M.; Mironova, E. G.; Mitrofanova, N. M.; Porollo, S. I.; Chernov, V. M.; Shamardin, V. K.; Garner, F. A.

    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.

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

    SciTech Connect

    Gruber, E.E.; Galvin, T.M.; Chopra, O.K.

    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.

  18. Effects of compositional complexity on the ion-irradiation induced swelling and hardening in Ni-containing equiatomic alloys

    SciTech Connect

    Jin, K.; Lu, C.; Wang, L. M.; Qu, J.; Weber, W. J.; Zhang, Y.; Bei, H.

    2016-04-14

    The impact of compositional complexity on the ion-irradiation induced swelling and hardening is studied in Ni and six Ni-containing equiatomic alloys with face-centered cubic structure. The irradiation resistance at the temperature of 500 °C is improved by controlling the number and, especially, the type of alloying elements. Alloying with Fe and Mn has a stronger influence on swelling reduction than does alloying with Co and Cr. Lastly, the quinary alloy NiCoFeCrMn, with known excellent mechanical properties, has shown 40 times higher swelling tolerance than nickel.

  19. Impact of irradiation on the tensile and fatigue properties of two titanium alloys

    NASA Astrophysics Data System (ADS)

    Marmy, P.; Leguey, T.

    2001-07-01

    The attachment of the first wall modules of the ITER FEAT fusion reactor is designed using flexible connectors made from titanium alloys. An assessment of the tensile and fatigue performance of two candidate alloys, a classical two phase Ti6Al4V alloy and a monophase α alloy Ti5Al2.5Sn, has been carried out using 590 MeV protons for the simulation of the fusion neutrons. The dose deposited was up to 0.3 dpa and the irradiation temperature was between 40°C and 350°C. The unirradiated tensile performances of both alloys are roughly identical. The radiation hardening is much stronger in the α+β alloy compared with the α alloy, and the ductility is correspondingly strongly reduced. A very fine precipitation observed by TEM in the primary and secondary α grains of the dual phase alloy seems to be the cause of the intense radiation hardening observed. Two different regimes have been observed in the behaviour of the cyclic stresses. At a high imposed strain, the softening is small in the Ti6Al4V and larger in the Ti5Al2.5Sn. At a low imposed strain, and for both alloys, cyclic softening occurs up to about 800 cycles, but then a transition occurs, after which a regime of cyclic hardening appears. This cyclic hardening disappears after irradiation. In both materials, and for all test conditions, the compressive stress of the hysteresis loop was found to be larger than the tensile stress. The stress asymmetry seems to be triggered by the plastic deformation. The fatigue resistance of the Ti5Al2.5Sn alloy is slightly better than that of the Ti6Al4V alloy. The irradiation did not significantly affect the fatigue performance of both alloys, except for high imposed strains, where a life reduction was observed in the case of the Ti6Al4V alloy. SEM micrographs showed that the fractures were transgranular and pseudo-brittle.

  20. Processing of Refractory Metal Alloys for JOYO Irradiations

    SciTech Connect

    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.

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

    SciTech Connect

    Edwards, D.J.; Singh, B.N.; Toft, P.; Eldrup, M.

    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.

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

    SciTech Connect

    Yang, Ying; Field, Kevin G; Allen, Todd R.; Busby, Jeremy T

    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.

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

    NASA Technical Reports Server (NTRS)

    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.

  4. Development of a high strength, hydrogen-resistant austenitic alloy. [Fe-36 Ni-3 Ti-3 Ta-1. 3 Mo

    SciTech Connect

    Chang, K.M.; Klahn, D.H.; Morris, J.W. Jr.

    1980-08-01

    Research toward high-strength, high toughness nonmagnetic steels for use in the retaining rings of large electrical generators led to the development of a Ta-modified iron-based superalloy (Fe-36 Ni-3 Ti-3 Ta-0.5 Al-1.3 Mo-0.3 V-0.01 B) which combines high strength with good toughness after suitable aging. The alloy did, however, show some degradation in fatigue resistance in gaseous hydrogen. This sensitivity was associated with a deformation-induced martensitic transformation near the fracture surface. The addition of a small amount of chromium to the alloy suppressed the martensite transformation and led to a marked improvement in hydrogen resistance.

  5. The influence of microstructure on blistering and bubble formation by He ion irradiation in Al alloys

    NASA Astrophysics Data System (ADS)

    Soria, S. R.; Tolley, A.; Sánchez, E. A.

    2015-12-01

    The influence of microstructure and composition on the effects of ion irradiation in Al alloys was studied combining Atomic Force Microscopy, Scanning Electron Microscopy and Transmission Electron Microscopy. For this purpose, irradiation experiments with 20 keV He+ ions at room temperature were carried out in Al, an Al-4Cu (wt%) supersaturated solid solution, and an Al-5.6Cu-0.5Si-0.5Ge (wt.%) alloy with a very high density of precipitates, and the results were compared. In Al and Al-4Cu, He bubbles were found with an average size in between 1 nm and 2 nm that was independent of fluence. The critical fluence for bubble formation was higher in Al-4Cu than in Al. He bubbles were also observed below the critical fluence after post irradiation annealing in Al-4Cu. The incoherent interfaces between the equilibrium θ phase and the Al matrix were found to be favorable sites for the formation of He bubbles. Instead, no bubbles were observed in the precipitate rich Al-5.6Cu-0.5Si-0.5Ge alloy. In all alloys, blistering was observed, leading to surface erosion by exfoliation. The blistering effects were more severe in the Al-5.6Cu-0.5Si-0.5Ge alloy, and they were enhanced by increasing the fluence rate.

  6. Ion irradiation testing and characterization of FeCrAl candidate alloys

    SciTech Connect

    Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew; Wang, Yongqiang

    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.

  7. Alloying effect of Ni and Cr on irradiated microstructural evolution of type 304 stainless steels

    NASA Astrophysics Data System (ADS)

    Tan, L.; Busby, J. T.

    2013-11-01

    Life extension of the existing nuclear power plants imposes significant challenges to core structural materials that suffer increased fluences. This paper presents the microstructural evolution of a type 304 stainless steel and its variants alloyed with extra Ni and Cr under neutron irradiation at ˜320 °C for up to 10.2 dpa. Similar to the reported data of type 304 variants, a large amount of Frank loops, ultrafine G-phase/M23C6 particles, and limited amount of cavities were observed in the irradiated samples. The irradiation promoted the growth of pre-existing M23C6 at grain boundaries and resulted in some phase transformation to CrC in the alloy with both extra Ni and Cr. A new type of ultrafine precipitates, possibly (Ti,Cr)N, was observed in all the samples, and its amount was increased by the irradiation. Additionally, α-ferrite was observed in the type 304 steel but not in the Ni or Ni + Cr alloyed variants. The effect of Ni and Cr alloying on the microstructural evolution is discussed.

  8. Irradiation effect of swift heavy ion for Zr50Cu40Al10 bulk glassy alloy

    NASA Astrophysics Data System (ADS)

    Onodera, Naoto; Ishii, Akito; Ishii, Kouji; Iwase, Akihiro; Yokoyama, Yoshihiko; Saitoh, Yuichi; Ishikawa, Norito; Yabuuchi, Atsushi; Hori, Fuminobu

    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, Zr50Cu40Al10 bulk glassy alloys were irradiated with heavy ions of 10 MeV I at room temperature. The maximum fluence was 3 × 1014 ions/cm2. 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.

  9. Mechanical properties and microstructural change of W-Y2O3 alloy under helium irradiation

    NASA Astrophysics Data System (ADS)

    Tan, Xiaoyue; Luo, Laima; Chen, Hongyu; Zhu, Xiaoyong; Zan, Xiang; Luo, Guangnan; Chen, Junling; Li, Ping; Cheng, Jigui; Liu, Dongping; Wu, Yucheng

    2015-07-01

    A wet-chemical method combined with spark plasma sintering was used to prepare a W-Y2O3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance.

  10. Mechanical properties and microstructural change of W-Y2O3 alloy under helium irradiation.

    PubMed

    Tan, Xiaoyue; Luo, Laima; Chen, Hongyu; Zhu, Xiaoyong; Zan, Xiang; Luo, Guangnan; Chen, Junling; Li, Ping; Cheng, Jigui; Liu, Dongping; Wu, Yucheng

    2015-07-31

    A wet-chemical method combined with spark plasma sintering was used to prepare a W-Y2O3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance.

  11. Mechanical properties and microstructural change of W–Y2O3 alloy under helium irradiation

    PubMed Central

    Tan, Xiaoyue; Luo, Laima; Chen, Hongyu; Zhu, Xiaoyong; Zan, Xiang; Luo, Guangnan; Chen, Junling; Li, Ping; Cheng, Jigui; Liu, Dongping; Wu, Yucheng

    2015-01-01

    A wet-chemical method combined with spark plasma sintering was used to prepare a W–Y2O3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance. PMID:26227480

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

    SciTech Connect

    Chung, H.M.; Tsai, H.C.; Nowicki, L.J.

    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.

  13. Irradiation-enhanced α' precipitation in model FeCrAl alloys

    SciTech Connect

    Edmondson, Philip D.; Briggs, Samuel A.; Yamamoto, Yukinori; Howard, Richard H.; Sridharan, Kumar; Terrani, Kurt A.; Field, Kevin G.

    2016-02-17

    We have irradiated the model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) with a neutron at ~ 320 to damage levels of ~ 7 displacements per atom (dpa) to investigate the compositional influence on the formation of irradiation-induced Cr-rich α' precipitates using atom probe tomography. In all alloys, significant number densities of these precipitates were observed. Cluster compositions were investigated and it was found that the average cluster Cr content ranged between 51.1 and 62.5 at.% dependent on initial compositions. Furthermore, this is significantly lower than the Cr-content of α' in binary FeCr alloys. As a result, significant partitioning of the Al from the α' precipitates was also observed.

  14. Irradiation-enhanced α' precipitation in model FeCrAl alloys

    DOE PAGES

    Edmondson, Philip D.; Briggs, Samuel A.; Yamamoto, Yukinori; ...

    2016-02-17

    We have irradiated the model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) with a neutron at ~ 320 to damage levels of ~ 7 displacements per atom (dpa) to investigate the compositional influence on the formation of irradiation-induced Cr-rich α' precipitates using atom probe tomography. In all alloys, significant number densities of these precipitates were observed. Cluster compositions were investigated and it was found that the average cluster Cr content ranged between 51.1 and 62.5 at.% dependent on initial compositions. Furthermore, this is significantly lower than the Cr-content of α' in binary FeCr alloys. As a result, significant partitioning ofmore » the Al from the α' precipitates was also observed.« less

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

    SciTech Connect

    Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    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.

  16. Database on Performance of Neutron Irradiated FeCrAl Alloys

    SciTech Connect

    Field, Kevin G.; Briggs, Samuel A.; Littrell, Ken; Parish, Chad M.; Yamamoto, Yukinori

    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.

  17. Effects of solute elements on irradiation hardening and microstructural evolution in low alloy steels

    NASA Astrophysics Data System (ADS)

    Fujii, Katsuhiko; Ohkubo, Tadakatsu; Fukuya, Koji

    2011-10-01

    The effects of the elements Mn, Ni, Si and Cu on irradiation hardening and microstructural evolution in low alloy steels were investigated in ion irradiation experiments using five kinds of alloys prepared by removing Mn, Ni and Si from, and adding 0.05 wt.%Cu to, the base alloy (Fe-1.5Mn-0.5Ni-0.25Si). The alloy without Mn showed less hardening and the alloys without Ni or Si showed more hardening. The addition of Cu had hardly any influence on hardening. These facts indicated that Mn enhanced hardening and that Ni and Si had some synergetic effects. The formation of solute clusters was not confirmed by atom probe (AP) analysis, whereas small dislocation loops were identified by TEM observation. The difference in hardening between the alloys with and without Mn was qualitatively consistent with loop formation. However, microstructural components that were not detected by the AP and TEM were assumed to explain the hardening level quantitatively.

  18. Microstructural changes in dilute Ni-Be alloys during HVEM sub-threshold irradiations

    SciTech Connect

    Regnier, P.G.; Lam, N.Q.

    1984-10-01

    The microstructural sensitivity of Ni-0.7 at.% Be alloys to HVEM sub-threshold irradiations was investigated. Several aspects were examined: (1) dose dependence of the microstructure change during irradiation with 350-keV electrons at 350/sup 0/C; (2) energy dependence of the incubation dose for the first appearance of black dots in the alloy films under irradiation at 350/sup 0/C along various crystallographic directions; and (3) temperature dependence of the microstructural evolution during 350-keV electron irradiation. It was found that sub-threshold irradiations were capable of inducing nonequilibrium solute segregation. Below approx. 400/sup 0/C, segregation-induced homogeneous precipitation of the ..gamma..-phase occurred in the alloy matrix, whereas at higher temperatures, only heterogeneous precipitation was observed at defect sinks. From the information about the energy dependence of the incubation dose for precipitation, the displacement threshold energy for Ni and point-defect production rate by secondary Be-Ni collisions were estimated.

  19. Microstructure and Mechanical Properties of n-irradiated Fe-Cr Model Alloys

    SciTech Connect

    Matijasevic, Milena; Al Mazouzi, Abderrahim

    2008-07-01

    High chromium ( 9-12 wt %) ferritic/martensitic steels are candidate structural materials for future fusion reactors and other advanced systems such as accelerator driven systems (ADS). Their use for these applications requires a careful assessment of their mechanical stability under high energy neutron irradiation and in aggressive environments. In particular, the Cr concentration has been shown to be a key parameter to be optimized in order to guarantee the best corrosion and swelling resistance, together with the least embrittlement. In this work, the characterization of the neutron irradiated Fe-Cr model alloys with different Cr % with respect to microstructure and mechanical tests will be presented. The behavior of Fe-Cr alloys have been studied using tensile tests at different temperature range ( from -160 deg. C to 300 deg. C). Irradiation-induced microstructure changes have been studied by TEM for two different irradiation doses at 300 deg. C. The density and the size distribution of the defects induced have been determined. The tensile test results indicate that Cr content affects the hardening behavior of Fe-Cr binary alloys. Hardening mechanisms are discussed in terms of Orowan type of approach by correlating TEM data to the measured irradiation hardening. (authors)

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

    SciTech Connect

    Smith, D.L.

    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

    NASA Astrophysics Data System (ADS)

    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

    SciTech Connect

    Zheng, Buxiang; Jiang, Gedong; Wang, Wenjun Wang, Kedian; Mei, Xuesong

    2014-03-15

    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/cm{sup 2}.

  3. Irradiation damage behavior of low alloy steel wrought and weld materials

    SciTech Connect

    Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

    1993-10-01

    A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ``superclean`` composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature ({Delta}TT4{sub 41J} or {Delta}TT{sub 47J}) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest {Delta}TT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials.

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

    SciTech Connect

    Zinkle, S.J.

    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.

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

    PubMed

    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

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

    NASA Astrophysics Data System (ADS)

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

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Smith, D.L.

    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.

  8. Helium behavior in vanadium-based alloys irradiated in the dynamic helium charging experiments

    SciTech Connect

    Fukumoto, K.; Matsui, H.; Chung, H.M.; Gazda, J.; Smith, D.L.

    1996-12-31

    Helium effect of neutron irradiated vanadium alloys, containing titanium, has been studied using Dynamic Helium Charging Experiment (DHCE) in FFTF. Cavity formation was observed only in pure vanadium irradiated at 430 to 600 C and in V-5Ti irradiated at 600 C. No apparent cavity formation was obtained in V-3Ti-1Si and V-4Cr-4Ti. The precipitation of titanium oxide in V-5Ti, V-3Ti-1Si and V-4Cr-4Ti occurred in all irradiation conditions in this study and the precipitates of Ti{sub 5}Si{sub 3} only appeared in V-3Ti-1Si irradiated at 600 C up to 15 dpa with helium generation rate of 4 appmHe/dpa. It is suggested that titanium oxide plays an important role for suppression of cavity formation and swelling from early stage of irradiation. Detail characterization of precipitates and He effect for neutron damages in vanadium alloys are discussed here.

  9. The response of dispersion-strengthened copper alloys to high fluence neutron irradiation at 415 degree C

    SciTech Connect

    Edwards, D.J.; Newkirk, J.W. ); Garner, F.A.; Hamilton, M.L. ); Nadkarny, A.; Samal, P. )

    1992-06-01

    Various oxide-dispersion-strengthened copper alloys have been irradiated to 150 dpa at 415{degree}C in the Fast Flux Test Facility (FFTF). The Al{sub 2}0{sub 3} - strengthened GlidCop{trademark} alloys, followed closely by a HfO{sub 2} - strengthened alloy, displayed the best swelling resistance, electrical conductivity, and tensile properties. The conductivity of the HfO{sub 2} - strengthened alloy reached a plateau at the higher levels of irradiation, instead of exhibiting the steady decrease in conductivity observed in the other alloys. A high initial oxygen content resulted in significantly higher swelling for a series of castable oxide-dispersion-strengthened alloys, while a Cr{sub 2}0{sub 3} - strengthened alloy showed poor resistance to radiation.

  10. The response of dispersion-strengthened copper alloys to high fluence neutron irradiation at 415{degree}C

    SciTech Connect

    Edwards, D.J.; Newkirk, J.W.; Garner, F.A.; Hamilton, M.L.; Nadkarny, A.; Samal, P.

    1992-06-01

    Various oxide-dispersion-strengthened copper alloys have been irradiated to 150 dpa at 415{degree}C in the Fast Flux Test Facility (FFTF). The Al{sub 2}0{sub 3} - strengthened GlidCop{trademark} alloys, followed closely by a HfO{sub 2} - strengthened alloy, displayed the best swelling resistance, electrical conductivity, and tensile properties. The conductivity of the HfO{sub 2} - strengthened alloy reached a plateau at the higher levels of irradiation, instead of exhibiting the steady decrease in conductivity observed in the other alloys. A high initial oxygen content resulted in significantly higher swelling for a series of castable oxide-dispersion-strengthened alloys, while a Cr{sub 2}0{sub 3} - strengthened alloy showed poor resistance to radiation.

  11. Helium effects on irradiation dmage in V alloys

    SciTech Connect

    Doraiswamy, N.; Alexander, D.

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

    SciTech Connect

    Chung, H.M.; Smith, D.L.

    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).

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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. Swelling and tensile properties of EBR-II-irradiated tantalum alloys for space reactor applications

    SciTech Connect

    Grossbeck, M.L.; Wiffen, F.W.

    1985-01-01

    The tantalum alloys T-111, ASTAR-811C, Ta-10 W, and unalloyed tantalum were examined following EBR-II irradiation to a fluence of 1.7 x 10/sup 26/ neutrons/m/sup 2/ (E > 0.1 MeV) at temperatures from 650 to 950 K. Swelling was found to be negligible for all alloys; only tantalum was found to exhibit swelling, 0.36%. Tensile testing revealed that irradiated T-111 and Ta-10 W are susceptible to plastic instability, but ASTAR-811C and tantalum were not. The tensile properties of ASTAR-811C appeared adequate for current SP-100 space nuclear reactor designs. Irradiated, oxygen-doped T-111 exhibited no plastic deformation, and the abrupt failure was intergranular in nature. The absence of plastic instability in ASTAR-811C is encouraging for alloys containing carbide precipitates. These fine precipitates might prevent dislocation channeling, which leads to plastic instability in many bcc metals after irradiation. 10 refs., 13 figs., 8 tabs.

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

    SciTech Connect

    Chung, H.M.; Smith, D.L.

    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.

  17. Nonswelling behavior of HT9 alloy irradiated to high exposure

    SciTech Connect

    Pitner, A.L.; Hecht, S.L.; Trenchard, R.G.

    1993-10-01

    In-reactor monitoring of assembly axial growths in the Fast Flux Test Facility (FFTF) has shown the ferritic/martensitic alloy HT9 to be essentially swelling free out to a fast neutron fluence of at least 37 {times} 10{sup 22} n/cm{sup 2}. This superior performance directly contributes to the ability to achieve high fuel burnup levels necessary for the ultimate viability of an economical Liquid Metal Reactor (LMR) fuel system.

  18. Irradiation By Neutrons And Annealing of SiGe Alloys

    NASA Technical Reports Server (NTRS)

    Vandersande, Jan W.; Mccormack, Joseph; Zoltan, Andrew

    1992-01-01

    Heat treatment restores thermoelectric performance having deteriorated under irradiation by neutrons. Discovery suggests SiGe materials used in radioisotope thermoelectric generators and other applications up to fluences of 5.4 X 10(to the 19th power)cm(to the negative 2nd power) and operating at temperatures of 600 to 1,000 degrees C.

  19. Response of nanostructured ferritic alloys to high-dose heavy ion irradiation

    SciTech Connect

    Parish, Chad M.; White, Ryan M.; LeBeau, James M.; Miller, Michael K.

    2014-02-01

    A latest-generation aberration-corrected scanning/transmission electron microscope (STEM) is used to study heavy-ion-irradiated nanostructured ferritic alloys (NFAs). Results are presented for STEM X-ray mapping of NFA 14YWT irradiated with 10 MeV Pt to 16 or 160 dpa at -100°C and 750°C, as well as pre-irradiation reference material. Irradiation at -100°C results in ballistic destruction of the beneficial microstructural features present in the pre-irradiated reference material, such as Ti-Y-O nanoclusters (NCs) and grain boundary (GB) segregation. Irradiation at 750°C retains these beneficial features, but indicates some coarsening of the NCs, diffusion of Al to the NCs, and a reduction of the Cr-W GB segregation (or solute excess) content. Ion irradiation combined with the latest-generation STEM hardware allows for rapid screening of fusion candidate materials and improved understanding of irradiation-induced microstructural changes in NFAs.

  20. Irradiation Embritlement in Alloy HT-­9

    SciTech Connect

    Serrano De Caro, Magdalena

    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

  1. Radiation-induced instability of MnS precipitates and its possible consequences on irradiation-induced stress corrosion cracking of austenitic stainless steels

    SciTech Connect

    Chung, H.M.; Sanecki, J.E.; Garner, F.A.

    1996-12-01

    Irradiation-assisted stress corrosion cracking (IASCC) is a significant materials issue for the light water reactor (LWR) industry and may also pose a problem for fusion power reactors that will use water as coolant. A new metallurgical process is proposed that involves the radiation-induced release into solution of minor impurity elements not usually thought to participate in IASCC. MnS-type precipitates, which contain most of the sulfur in stainless steels, are thought to be unstable under irradiation. First, Mn transmutes strongly to Fe in thermalized neutron spectra. Second, cascade-induced disordering and the inverse Kirkendall effect operating at the incoherent interfaces of MnS precipitates are thought to act as a pump to export Mn from the precipitate into the alloy matrix. Both of these processes will most likely allow sulfur, which is known to exert a deleterious influence on intergranular cracking, to re-enter the matrix. To test this hypothesis, compositions of MnS-type precipitates contained in several unirradiated and irradiated heats of Type 304, 316, and 348 stainless steels (SSs) were analyzed by Auger electron spectroscopy. Evidence is presented that shows a progressive compositional modification of MnS precipitates as exposure to neutrons increases in boiling water reactors. As the fluence increases, the Mn level in MnS decreases, whereas the Fe level increases. The S level also decreases relative to the combined level of Mn and Fe. MnS precipitates were also found to be a reservoir of other deleterious impurities such as F and O which could be also released due to radiation-induced instability of the precipitates.

  2. Alloy development for irradiation performance. Quarterly progress report for period ending September 30, 1980

    SciTech Connect

    Not Available

    1980-12-01

    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.

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

    SciTech Connect

    Ashdown, B.G.

    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.

  4. Alloy development for irradiation performance. Quarterly progress report for period ending June 30, 1980

    SciTech Connect

    Ashdown, B.G.

    1980-10-01

    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.

  5. Swelling and structure of vanadium-base alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1994-08-01

    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 C in the Dynamic Helium Charging Experiment (DHCE), and the results were compared with those from a non-DHCE in which helium generation and negligible. For specimens irradiated to {approx}18-31 dpa at 500--600 with a helium generation rate of 0.4--4.2 appm He/dpa, only a few helium bubbles were observed at the interface of grain matrices and some of the Ti(O,N,C) precipitates, and no microvoids or helium bubbles were observed either in grain matrices or near grain boundaries. Under these conditions, dynamically produced helium atoms seem to be trapped in the grain matrix without significant bubble nucleation or growth, and in accordance with this, density changes from DHCE and non-DHCE (negligible helium generation) were similar for comparable fluence and irradiation temperature. Only for specimens irradiated to {approx}31 dpa at 425 C, when helium was generated at a rage of 0.4--0.8 appm helium/dpa, were diffuse helium bubbles observed in limited regions of grain matrices and near {approx}15% of the grain boundaries in densities significantly lower than those in the extensive coalescences of helium bubbles typical of other alloys irradiated in tritium-trick experiments. Density changes of specimens irradiated at 425 C in the DHCE were significantly higher than those from non-DHCE irradiation. Microstructural evolution in V-4Cr-4Ti was similar for DHCE and non-DHCE except for helium bubble number density and distribution. As in non-DHCE, the irradiation-induced precipitation of ultrafine Ti{sub 5}Si{sub 3} was observed for DHCE at >500 C but not at 425 C.

  6. Atom probe study of irradiation-enhanced α' precipitation in neutron-irradiated Fe–Cr model alloys

    SciTech Connect

    Chen, Wei -Ying; Miao, Yinbin; Wu, Yaqiao; Tomchik, Carolyn A.; Mo, Kun; Gan, Jian; Okuniewski, Maria A.; Maloy, Stuart A.; Stubbins, James F.

    2015-07-01

    Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on a' phase formation in Fe-Cr model alloys (10-16 at.%) irradiated at 300 and 450°C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α' precipitates with an average radius of 1.0-1.3 nm were observed. The precipitate density varied significantly from 1.1x10²³ to 2.7x10²⁴ 1/m³, depending on Cr concentrations and irradiation temperatures. The volume fraction of α' phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe-10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of a' precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.

  7. Effects of neutron irradiation and hydrogen on ductile-brittle transition temperatures of V-Cr-Ti alloys

    SciTech Connect

    Loomis, B.A.; Chung, H.M.; Nowicki, L.J.; Smith, D.L.

    1993-08-01

    The effects of neutron irradiation and hydrogen on the ductile- brittle transition temperatures (DBTTs) of unalloyed vanadium and V-Cr-Ti alloys were determined from Charpy-impact tests on 1/3 ASTM standard size specimens and from impact tests on 3-mm diameter discs. The tests were conducted on specimens containing <30 appm hydrogen and 600-1200 appm hydrogen and on specimens after neutron irradiation to 28-46 dpa at 420, 520, and 600C. The DBTTs were minimum (< {minus}220{degree}C) for V-(105)Ti alloys under for V-4-Cr-4Ti alloy with <30 appm hydrogen. The effect of 600-1200 appm hydrogen in the specimens was to raise the DBTTs by 100--150{degree}C. The DBTTs were minimum (< {minus}220{degree}C) for V-(1-5)Ti alloys and V-4-Cr-4Ti alloys after neutron irradiation.

  8. Correlating radiation exposure with embrittlement: Comparative studies of electron- and neutron-irradiated pressure vessel alloys

    SciTech Connect

    Alexander, D. E.; Rehn, L. E.; Odette, G. R.; Lucas, G. E.; Klingensmith, D.; Gragg, D.

    1999-12-22

    Comparative experiments using high energy (10 MeV) electrons and test reactor neutrons have been undertaken to understand the role that primary damage state has on hardening (embrittlement) induced by irradiation at 300 C. Electrons produce displacement damage primarily by low energy atomic recoils, while fast neutrons produce displacements from considerably higher energy recoils. Comparison of changes resulting from neutron irradiation, in which nascent point defect clusters can form in dense cascades, with electron irradiation, where cascade formation is minimized, can provide insight into the role that the in-cascade point defect clusters have on the mechanisms of embrittlement. Tensile property changes induced by 10 MeV electrons or test reactor neutron irradiations of unalloyed iron and an Fe-O.9 wt.% Cu-1.0 wt.% Mn alloy were examined in the damage range of 9.0 x 10{sup {minus}5} dpa to 1.5 x 10{sup {minus}2} dpa. The results show the ternary alloy experienced substantially greater embrittlement in both the electron and neutron irradiate samples relative to unalloyed iron. Despite their disparate nature of defect production similar embrittlement trends with increasing radiation damage were observed for electrons and neutrons in both the ternary and unalloyed iron.

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

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    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.

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

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Chung, H.; Smith, D.L.

    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.

  11. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    DOE PAGES

    Parish, Chad M.; Unocic, Kinga A.; Tan, Lizhen; ...

    2016-10-24

    Here we irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ~50 dpa, ~15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ~8 nm, ~1021 m-3 (CNA), and of ~3 nm, 1023 m-3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces inmore » all alloys survived ~50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Finally, among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.« less

  12. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    NASA Astrophysics Data System (ADS)

    Parish, C. M.; Unocic, K. A.; Tan, L.; Zinkle, S. J.; Kondo, S.; Snead, L. L.; Hoelzer, D. T.; Katoh, Y.

    2017-01-01

    We irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ∼50 dpa, ∼15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ∼8 nm, ∼1021 m-3 (CNA), and of ∼3 nm, 1023 m-3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces in all alloys survived ∼50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.

  13. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    SciTech Connect

    Parish, Chad M.; Unocic, Kinga A.; Tan, Lizhen; Zinkle, S. J.; Kondo, Sosuke; Snead, Lance Lewis; Hoelzer, David T.; Katoh, Yutai

    2016-10-24

    Here we irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ~50 dpa, ~15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ~8 nm, ~1021 m-3 (CNA), and of ~3 nm, 1023 m-3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces in all alloys survived ~50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Finally, among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Chung, H.M.; Galvin, T.M.; Smith, D.L.

    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.

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

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    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.

  18. Microstructure and mechanical properties of austenitic stainless steel 12X18H9T after neutron irradiation in the pressure vessel of BR-10 fast reactor at very low dose rates

    SciTech Connect

    Porollo, S. I.; Dvoriashin, Alexander M.; Konobeev, Yury V.; Ivanov, A. A.; Shulepin, S. V.; Garner, Francis A.

    2006-12-01

    Results are presented for void swelling, microstructure andmechanical properties of Russian 12X18H9T (0.12C-18Cr-9Ni-Ti) austenitic stainless steel irradiated as a pressure vessel structure material of the BR-10 fast reactor at ~350C to only 0.64 dpa, produced by many years of exposure at the very low displacement rate of only 1.9x10-9 dpa/s. In agreement with a number of other recent studies it appears that lower dpa rates have a pronounced effect on the microstructure and resultant mechanical properties. In general, loweer dpa rates lead to the onset of swelling at much lower doses compared to comparable irradiations conducted at higher dpa rates.

  19. Monte Carlo simulations of copper clustering in Fe-Cu alloys under irradiation

    NASA Astrophysics Data System (ADS)

    Kwon, J.; Kwon, S. C.; Hong, J. H.

    2004-10-01

    We present the computational approach for studying the microstructures of Cu clusters in Fe-Cu alloys by combining the molecular dynamics (MD) simulation and Monte Carlo methods. The MD simulation is used to characterize the primary damage resulting from the displacement cascade in Fe. Then, using the Metropolis Monte Carlo methods, the microstructure of the Cu clusters is predicted under the assumption that the system will evolve towards the equilibrium state. The formation of the Cu clusters is apparent for Fe-Cu alloys of a higher Cu content (1.0 w/o), whereas the degree of Cu clustering is not significant for the lower Cu content (0.1 w/o) alloys. The atomic configuration of the Cu-vacancy complex under irradiation, produced by this simulation, is in a fair agreement with the experiments. The simulation is expected to provide important information on the Cu-cluster morphology, which is useful for experimental data analysis.

  20. Effects of self-irradiated damage on physical properties of stabilized Pu alloys

    NASA Astrophysics Data System (ADS)

    Freibert, F.; Martinez, B.; Baiardo, J. P.; Olivas, J.; Ronquillo, R.

    2000-07-01

    Our team is currently conducting experiments in the areas of thermal, physical and magnetic properties of Pu239 alloys doped with small quantities of Pu238 in an effort to further our understanding of alterations in electronic structure and self-irradiated damage in these alloys. The combination of data from these measurements will provide the following information: elastic properties and material compressibility, relative lattice defect concentration, microstructure alterations and phase homogeneity, phase transition onset temperature, intermediate phase stability, and transformation type. This series of measurements will provide a unique before and after picture of aging in these stabilized alloys, therefore answering important questions concerning these materials and providing valuable comparisons between newly cast materials and site-returned materials.

  1. Alloy development for first wall materials used in water-cooling type fusion reactors

    NASA Astrophysics Data System (ADS)

    Kiuchi, K.; Ishiyama, T.; Hishinuma, A.

    1991-03-01

    Austenitic stainless steels with high resistance to IASCC were developed for the first wall used in a water cooling type fusion reactor. New alloys with ultra low carbon content were designed to improve all-round properties relevant to the reliability below 450°C, by enhancing the austenite phase stability and purifying the austenite matrix. For this purpose, these were manufactured by means of controlling minor elements, adjusting principal elements and applying SAR thermomechanical treatment. The major characteristics of these alloys were compared with that of Type 316 and JPCA. These alloys showed a good swelling resistance to electron irradiation and high cracking resistance to high heat fluxes of hydrogen beam. The ductility loss and decrease of tensile strength at the objective temperature were also minimized.

  2. Mapping Flow Localization Processes in Deformation of Irradiated Reactor Structural Alloys

    SciTech Connect

    Farrell, K.

    2002-07-18

    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.

  3. Localized deformation and IASCC initiation in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Jiao, Z.; Was, G. S.

    2008-12-01

    Localized deformation may play a key role in the underlying mechanism of irradiation assisted stress corrosion cracking (IASCC) in light water reactor core components. In this study, four austenitic alloys, 18Cr8Ni, 15Cr12Ni, 13Cr15Ni and 21Cr32Ni, with different stacking fault energies were irradiated to 1 and 5 dpa at 360 °C using 3.2 MeV protons. Interrupted constant extension rate tensile (CERT) tests were conducted in a simulated BWR environment to determine IASCC susceptibility. In order to characterize the localized deformation in slip channels and grain boundaries, parallel CERT experiments were also performed in an argon atmosphere. Results show that the IASCC susceptibility of the tested alloys increases with increasing irradiation dose and decreasing stacking fault energy. IASCC tends to initiate at locations where slip channels intersect grain boundaries. Localized deformation in the form of grain boundary sliding due to the interaction of slip channels and grain boundaries is likely the primary cause of the observed cracking initiation.

  4. Irradiation performance of Fast Flux Test Facility drivers using D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1995-11-01

    In comparison with the Fast Flux Test Facility Type 316 stainless steel driver design, six test assemblies employing D9 alloy in place of stainless steel for duct, cladding, and wire wrap material were irradiated to demonstrate the improved performance and lifetime capability of an advanced D9 alloy driver design. A single pinhole-type breach occurred in one of the high-exposure tests after a peak fuel burnup of 155 MWd/kg metal (M) and peak fast neutron fluence of 25 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV). Postirradiation examinations were performed on four of the test assemblies and measured results were compared with analytical evaluations. A revised swelling correlation for D9 alloy was developed to provide improved agreement between calculated and measured cladding deformation results. A fuel pin lifetime design criterion of 5% calculated hoop strain was derived from these results. Alternatively, fuel pin lifetimes were developed for two irradiation parameters using statistical failure analyses. For a 99.99% reliability, the analyses indicated a peak fast-fluence lifetime of 21.0 {times} 10{sup 22} n/cm{sup 2}, or a peak fuel burnup >120 MWd/kg M. In comparison with the Fast Flux Test Facility reference driver design, the extended lifetime capability of D9 alloy would reduce fuel supply requirements for the liquid-metal reactor by a third.

  5. Ar irradiated Cr rich Ni alloy studied using positron annihilation spectroscopy

    NASA Astrophysics Data System (ADS)

    Saini, Sanjay; Menon, Ranjini; Sharma, S. K.; Srivastava, A. P.; Mukherjee, S.; Nabhiraj, P. Y.; Pujari, P. K.; Srivastava, D.; Dey, G. K.

    2016-10-01

    The present study focuses on understanding the effect of Ar ion irradiation at room temperature on Cr rich Ni-Cr alloy. The alloy is irradiated with Ar9+ ions (energy 315 keV) for total dose varying from 9.3 × 1014 to 2.3 × 1016 ion/cm2. The changes in the microstructure of the irradiated samples have been characterized by depth dependent Doppler broadening of annihilation radiation (DBAR) measurements using a slow positron beam facility. The variation in S-E profiles as a function of total dose corroborated with S-W curves indicates that the type of defects is also varied with the increase in total dose. The S-E profiles have been fitted using variable energy positron fit (VEPFIT) program considering a three layer structure for the irradiated samples. Estimated displacement damage profile as a function of increasing dose has been analyzed and a possible mechanism has been attributed to explain the observations made from S-parameter variation.

  6. Irradiation creep of VTiCr alloy in BR-10 reactor core instrumented experiments

    NASA Astrophysics Data System (ADS)

    Troyanov, V. M.; Bulkanov, M. G.; Kruglov, A. S.; Krjuchkov, E. A.; Nikulin, M. P.; Pevchykh, J. M.; Rusanov, A. E.; Smirnoff, A. A.; Votinov, S. N.

    1996-10-01

    A thin wall tubular-type speciment of 4%Ti-4%Cr vanadium alloy was tested for creep under irradiation in BR-10 reactor at 713-723 K and at 8.6 × 10 18 n/m 2s fast neutron flux. A fluence at the end of the experiment have reached 5.8 × 10 25 n/m 2. Specimen deformation measurements were performed by a dynamometric method based on a stress relaxation control provided during irradiation under constant load applied. During the experiment 13 deformation curves were obtained for different stress levels ranged up to 165 MPa. At the same time the yield stress of the irradiated specimen was periodically determined. The irradiation creep rate has been found to be proportional to the stress up to 110-120 MPa with the module equal to 3.3 × 10 -12 dpa -1Pa -1. At higher streses, a creep process essentially accelerates. The results on VTiCr alloy are discussed in respect to data obtained for stainless steels in earlier BR-10 reactor experiments.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    PubMed

    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.

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

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Smith, D.L.

    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.

  11. Dependence of the nitriding rate of ferritic and austenitic substrates on the crystallographic orientation of surface grains; gaseous nitriding of Fe-Cr and Ni-Ti alloys

    NASA Astrophysics Data System (ADS)

    Akhlaghi, M.; Jung, M.; Meka, S. R.; Fonović, M.; Leineweber, A.; Mittemeijer, E. J.

    2015-12-01

    Gaseous nitriding of ferritic Fe-Cr and austenitic Ni-Ti solid solutions reveals that the extent of the uptake of dissolved nitrogen depends on the crystallographic orientation of the surface grains of the substrate. In both ferritic and austenitic substrates, the surface nitrogen concentration and the nitriding depth decrease upon increasing the smallest angle between the surface normal and the normal of a {1 0 0} plane of the surface grain considered. This phenomenon could be ascribed to the residual compressive macrostress developed during nitriding which varies as a function of crystallographic orientation of the (surface) grains due to the elastically anisotropic nature of ferrite and austenite solid solutions investigated in this study.

  12. Properties of vanadium-base alloys irradiated in the dynamic helium charging experiment

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1996-02-01

    One property of vanadium-base alloys that is not well understood in terms of their potential use as fusion reactor structural materials is the effect of simultaneous generation of helium and neutron damage. In the present Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in the specimen at linear rates of {approx} 0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18--31 dpa at 425--600 C in Li-filled capsules in a sodium-cooled fast reactor. This paper presents results of postirradiation examination and tests of microstructure and mechanical properties of V-5Ti, V-3Ti-1Si, V-8Cr-6Ti, and V-4Cr-4Ti (the latter alloy has been identified as the most promising candidate vanadium alloy). Effects of helium on tensile strength and ductility were insignificant after irradiation and testing at > 420 C. However, postirradiation ductilities at < 250 C were higher than those of the non-DHCE specimens (< 0.1 appm helium), whereas strengths were lower, indicating that different types of hardening centers are produced during DHCE and non-DHCE irradiation. Ductile-brittle transition behavior of the DHCE specimens was also determined from bend tests and fracture appearance of transmission electron microscopy (TEM) disks and broken tensile specimens. No brittle behavior was observed at temperatures > {minus}150 C in DHCE specimens. Predominantly brittle-cleavage fracture morphologies were observed only at {minus}196 C in some specimens that were irradiated to 31 dpa at 425 C during the DHCE. For the helium generation rates in this experiment ({approx} 0.4--4.2 appm He/dpa), grain-boundary coalescence of helium microcavities was negligible and intergranular fracture was not observed.

  13. Surface modifications of hydrogen storage alloy by heavy ion beams with keV to MeV irradiation energies

    NASA Astrophysics Data System (ADS)

    Abe, Hiroshi; Tokuhira, Shinnosuke; Uchida, Hirohisa; Ohshima, Takeshi

    2015-12-01

    This study deals with the effect of surface modifications induced from keV to MeV heavy ion beams on the initial reaction rate of a hydrogen storage alloy (AB5) in electrochemical process. The rare earth based alloys like this sample alloy are widely used as a negative electrode of Ni-MH (Nickel-Metal Hydride) battery. We aimed to improve the initial reaction rate of hydrogen absorption by effective induction of defects such as vacancies, dislocations, micro-cracks or by addition of atoms into the surface region of the metal alloys. Since defective layer near the surface can easily be oxidized, the conductive oxide layer is formed on the sample surface by O+ beams irradiation, and the conductive oxide layer might cause the improvement of initial reaction rate of hydriding. This paper demonstrates an effective surface treatment of heavy ion irradiation, which induces catalytic activities of rare earth oxides in the alloy surface.

  14. Characterization of Nanostructural Features in Irradiated Reactor Pressure Vessel Model Alloys

    SciTech Connect

    Wirth, B D; Odette, G R; Asoka-Kumar, P; Howell, R H; Sterne, P A

    2001-08-01

    Irradiation embrittlement in nuclear reactor pressure vessel steels results from the formation of a high number density of nanometer-sized copper rich precipitates and sub-nanometer defect-solute clusters. We present results of small angle neutron scattering (SANS) and positron annihilation spectroscopy (PAS) characterization of the nanostructural features formed in binary and ternary Fe-Cu-Mn alloys irradiated at {approx}290 C. These complementary techniques provide insight into the composition and character of both types of nanoscale features. The SANS measurements indicate populations of copper-manganese precipitates and smaller vacancy-copper-manganese clusters. The PAS characterization, including both Doppler broadening and positron lifetime measurements, indicates the presence of essentially defect-free Cu precipitates in the Fe-Cu-Mn alloy and vacancy-copper clusters in the Fe-Cu alloy. Thus the SANS and PAS provide a self-consistent picture of nanostructures composed of copper-rich precipitates and vacancy solute cluster complexes and tend to discount high Fe concentrations in the CRPs.

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

    SciTech Connect

    Tsai, H.; Nowicki, L.J.; Billone, M.C.; Chung, H.M.; Smith, D.L.

    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.

  16. Electrical resistivity measurement of Fe-0.6%Cu alloy irradiated by neutrons at 14-19 K

    NASA Astrophysics Data System (ADS)

    Xu, Q.; Yokotani, T.; Sato, K.; Hori, F.

    2016-12-01

    Electrical resistivity measurement is a useful experimental method for investigating the recovery of defects that are induced by irradiation in metals and alloys. In this study, an Fe-0.6%Cu alloy, used to model steel from old commercial reactor pressure vessels, was irradiated by neutrons at a low temperature range of 14-19 K with a dose of about 1.3 × 1020 neutrons/m2 (E > 0.1 MeV) in the Kyoto University Reactor (KUR); electrical resistivity measurement was performed during irradiation and after annealing of the irradiated sample from 20 K to 300 K to investigate the migration of point defects in the Fe-0.6%Cu alloy. The electrical resistivity was measured at 14-19 K. With the increase in the irradiation dose, the electrical resistivity increased linearly. Four peaks appeared at 70 K, 100 K, 150 K, and 260 K, in the change of electrical resistivity during annealing of the irradiated sample up to 300 K. The former two peaks were caused by the recombination of interstitials and vacancies, and the latter two peaks were caused by the formation of interstitial clusters and the migration of vacancies. Compared with previous electron irradiation results, the former two peaks represent new data, as does the ratio of recombination caused by close-pair and correlation to that caused by migrations of mixed-interstitials Fe-Cu and vacancies decreased in neutron irradiation.

  17. The irradiation-induced microstructural development and the role of γ' on void formation in Ni-based alloys

    NASA Astrophysics Data System (ADS)

    Kato, Takahiko; Nakata, Kiyotomo; Masaoka, Isao; Takahashi, Heishichiro; Takeyama, Taro; Ohnuki, Soumei; Osanai, Hisashi

    1984-05-01

    The microstructural development for Inconel X-750, N1-13 at%A1, and Ni-11.5 at%Si alloys during irradiation was investigated. These alloys were previously heat-treated at temperatures of 723-1073 K, and γ' precipitates were produced. Irradiation was performed in a high voltage electron microscope (1000 kV) in the temperature range 673-823 K. In the case of solution-treated Inconel, interstitial dislocation loops were formed initially, while voids were nucleated after longer times. When the Inconel specimen containing a high number density of small γ' was irradiated, dislocation loops were formed in both the matrix and precipitate-matrix interface. The loops formed on the interface scarcely grew during irradiation. On the other hand, for the Ni-Al alloy fine γ' nucleated during irradiation, the large γ' precipitated by pre-aging, dissolved. A similar resolution process was also observed in Ni-Si alloy. Furthermore, in the Ni-Si alloy precipitates of γ' formed preferentially at interstitial dislocation loops and both specimen surfaces.

  18. Cast alumina forming austenitic stainless steels

    DOEpatents

    Muralidharan, Govindarajan; Yamamoto, Yukinori; Brady, Michael P

    2013-04-30

    An austenitic stainless steel alloy consisting essentially of, in terms of weight percent ranges 0.15-0.5C; 8-37Ni; 10-25Cr; 2.5-5Al; greater than 0.6, up to 2.5 total of at least one element selected from the group consisting of Nb and Ta; up to 3Mo; up to 3Co; up to 1W; up to 3Cu; up to 15Mn; up to 2Si; up to 0.15B; up to 0.05P; up to 1 total of at least one element selected from the group consisting of Y, La, Ce, Hf, and Zr; <0.3Ti+V; <0.03N; and, balance Fe, where the weight percent Fe is greater than the weight percent Ni, and wherein the alloy forms an external continuous scale comprising alumina, and a stable essentially single phase FCC austenitic matrix microstructure, the austenitic matrix being essentially delta-ferrite free and essentially BCC-phase-free. A method of making austenitic stainless steel alloys is also disclosed.

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

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Smith, D.L.; Matsui, H.

    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.

  20. Irradiation effects in rapidly and conventionally solidified alloys. Phase stability in rapidly solidified N i-Nb under Ni ion irradiation

    NASA Technical Reports Server (NTRS)

    1982-01-01

    Two alloy compositions in the Ni-Nb system (Ni60Nb40 and Ni85Nb15) were produced by rapidly quenching from the melt with the piston anvil technique. The Ni60Nb40 was transformed to a metastable, partially crystalline state by heat treatment in a differential scanning calorimeter. The Ni85Nb15 was fully crystalline, with the majority of the grains composed of collections of primary dendrite arms. Both compositions were irradiated with 4 MeV Ni++ ions. The irradiation induced microstructures were examined by transmission electron microscopy and compared with thermally aged samples. The thermal evolution was arrested by ion irradiation in the temperature range studied, by inhibiting the nucleation of the NiNb phase. No irradiation induced voids were observed. It is found that the ion irradiation drives the microstructure along a different path than thermal evolution.

  1. Tensile and electrical properties of copper alloys irradiated in a fission reactor

    SciTech Connect

    Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.; Rowcliffe, A.F.

    1996-04-01

    Postirradiation electrical sensitivity and tensile measurements have been completed on pure copper and copper alloy sheet tensile specimens irradiated in the SM-2 reactor to doses of {approx}0.5 to 5 dpa and temperatures between {approx}80 and 400{degrees}C. Considerable radiation hardening and accompanying embrittlement was observed in all of the specimens at irradiation temperature below 200{degrees}C. The radiation-induced electrical conductivity degradation consisted of two main components: solid transmutation effects and radiation damage (defect cluster and particle dissolution) effects. The radiation damage component was nearly constant for the doses in this study, with a value of {approx}1.2n{Omega}m for pure copper and {approx}1.6n{Omega}m for dispersion strengthened copper irradiated at {approx}100{degrees}C. The solid transmutation component was proportional to the thermal neutron flux, and became larger than the radiation damage component for fluences larger than {approx}5 10{sup 24} n.m{sup 2}. The radiation hardening and electrical conductivity degradation decreased with increasing irradiation temperature, and became negligible for temperatures above {approx}300{degrees}C.

  2. Nanoindentation on V-4Ti alloy irradiated by H and He ions

    NASA Astrophysics Data System (ADS)

    Yang, Yitao; Zhang, Chonghong; Meng, Yancheng; Liu, Juan; Gou, Jie; Xian, Yongqiang; Song, Yin

    2015-04-01

    V-4Ti and V samples were irradiated by H/He ions with various energies to produce a damage plateau in the region from surface to the depth of 1.5 um at room temperature. Nanoindentation was performed to investigate the hardening behavior of the two materials under irradiation. It is found that the relation of maximum depth of plastic zone and indentation depth is not a fixed value. The maximum depth of plastic zone decreases with increase of damage level. Nix and Gao model was used to fit the measured hardness to obtain a hardness value H0 excluding indentation size effect (ISE), which can be used to characterize the hardening effect induced by irradiation. After fitting the data of H0, it is found that there is an exponential relation between the H0 and damage level for both the V-4Ti and V materials. When the damage level is higher than ∼0.2 dpa, the hardness increases slowly, this indicates a slow increase of pinning centers in samples at this damage level. Comparing the hardening fraction of V-4Ti and V samples, significant hardening can be seen for V sample, and it becomes severe especially at damage higher than ∼0.2 dpa. The irradiation hardening resistance property of V-4Ti alloy is better than that of pure V.

  3. Ion irradiation induced nanocrystal formation in amorphous Zr 55Cu 30Al 10Ni 5 alloy

    NASA Astrophysics Data System (ADS)

    Carter, Jesse; Fu, E. G.; Martin, Michael; Xie, Guoqiang; Zhang, X.; Wang, Y. Q.; Wijesundera, D.; Wang, X. M.; Chu, Wei-Kan; McDeavitt, Sean M.; Shao, Lin

    2009-09-01

    Ion irradiation can be used to induce partial crystallization in metallic glasses to improve their surface properties. We investigated the microstructural changes in ribbon Zr 55Cu 30Al 10Ni 5 metallic glass after 1 MeV Cu-ion irradiation at room temperature, to a fluence of 1.0 × 10 16 cm -2. In contrast to a recent report by others that there was no irradiation induced crystallization in the same alloy [S. Nagata, S. Higashi, B. Tsuchiya, K. Toh, T. Shikama, K. Takahiro, K. Ozaki, K. Kawatusra, S. Yamamoto, A. Inouye, Nucl. Instr. and Meth. B 257 (2007) 420], we have observed nanocrystals in the as-irradiated samples. Two groups of nanocrystals, one with diameters of 5-10 nm and another with diameters of 50-100 nm are observed by using high resolution transmission electron microscopy. Experimentally measured planar spacings ( d-values) agree with the expectations for Cu 10Zr 7, NiZr 2 and CuZr 2 phases. We further discussed the possibility to form a substitutional intermetallic (Ni xCu 1-x)Zr 2 phase.

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

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Smith, D.L.

    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. Effects of iron concentration on the microstructure of V-Fe alloys after low-dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Oono, Naoko; Nita, Nobuyasu; Abe, Yosuke; Satoh, Yuhki; Matsui, Hideki

    2011-11-01

    To examine the mechanism of huge swelling of V-Fe alloys, TEM observation was applied to V- x at.%Fe ( x = 0, 0.3, 1, 5, 7), after low-dose neutron irradiation of 0.1-1.17 dpa at 400-600 °C. Concentric Multi-Dislocation Loops (CMDLs) and rafts were observed in V-Fe alloys irradiated to 0.1 dpa. Voids were observed only within dislocation loops in both types of configurations. Ashby-Brown (AB) contrasts were always observed around voids, indicating that shell-like iron segregation with substantial elastic strain exists on the inner surface of the voids. For 1.17 dpa, swelling V-Fe alloys was always greater than those of un-alloyed vanadium, whatever the dislocation microstructures. Implication of a segregation shell in the swelling mechanism is discussed in addition to a model based on the dislocation bias.

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

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    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.

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

    SciTech Connect

    Chung, H.M.; Nowicki, L.; Gazda, J.

    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.

  8. Metal ceramic alloy structure and surface layer modification during electron-ion-plasma irradiation of its surface

    NASA Astrophysics Data System (ADS)

    Ovcharenko, V. E.; Ivanov, Yu. F.; Shilko, E. V.; Mokhovikov, A. A.; Baohai, Yu; Tianyng, Xiong; Hua, Xu Yun; Lisheng, Zhong

    2016-11-01

    The paper presents research findings on the problems of electron-beam irradiation in noble gases plasma with different indexes of ionizing energy and atomic weight, and a surface layer structure modification versus a surface layer microhardness, wear and bending resistances and corrosion stability of 50% TiC/50% (Ni + 20% Cr) metal ceramic alloy samples. Discussions on the issues of the ways impulse electron-beam irradiation in the conditions of various types of noble gas plasma influences the mechanism of a metal ceramic alloy surface layer structure-phase state modification has been also presented.

  9. Microstructural evolution of metastable austenitic steel during high-pressure torsion and subsequent heat treatment

    NASA Astrophysics Data System (ADS)

    Chen, S.; Shibata, A.; Zhao, L. J.; Gao, S.; Tian, Y. Z.; Tsuji, N.

    2014-08-01

    Metastable austenite in a Fe-24Ni-0.3C (wt.%) alloy was processed by high-pressure torsion and subsequently heat-treated. The HPT-processed material had lamellae structures composed of highly deformed austenite and deformation-induced martensite. The reverse transformation of the deformation-induced martensite and recovery/recrystallization of the retained austenite completed above 500 °C and resulted in fully annealed and single-phase austenite with different grain sizes. The ultrafine-grained and nanocrystalline austenite showed high yield strength and large ductility due to transformation-induced plasticity.

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

    NASA Astrophysics Data System (ADS)

    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<100> dislocation loops, a/2<111> 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<111> dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and a<100> 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.

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

    DOE PAGES

    Field, Kevin G.; Briggs, Samuel A.; Hu, Xunxiang; ...

    2016-11-01

    FeCrAl alloys are an attractive materials class for nuclear power applications due to their increased environmental compatibility over more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300-400 °C have shown post-irradiation microstructures containing dislocation loops and Cr-rich ' phase. Although these initial works established the post-irradiation microstructures, little to no focus was applied towards the influence of pre-irradiation microstructures on this response. Here, a well annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 dpa at 382 °C and then the role of random high angle grain boundariesmore » on the spatial distribution and size of dislocation loops, dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and an increased size of dislocation loops in the vicinity directly adjacent to the grain boundary. Lastly, these results suggest the importance of the pre-irradiation microstructure on the radiation tolerance of FeCrAl alloys.« less

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

    SciTech Connect

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

    2016-11-01

    FeCrAl alloys are an attractive materials class for nuclear power applications due to their increased environmental compatibility over more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300-400 °C have shown post-irradiation microstructures containing dislocation loops and Cr-rich ' phase. Although these initial works established the post-irradiation microstructures, little to no focus was applied towards the influence of pre-irradiation microstructures on this response. Here, a well annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 dpa at 382 °C and then the role of random high angle grain boundaries on the spatial distribution and size of dislocation loops, dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and an increased size of dislocation loops in the vicinity directly adjacent to the grain boundary. Lastly, these results suggest the importance of the pre-irradiation microstructure on the radiation tolerance of FeCrAl alloys.

  13. Irradiation performance of Fast Flux Test Facility drivers using D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1994-06-01

    Six test assemblies similar in design to the FFTF driver assembly but employing the advanced alloy D9 in place of Type 316 stainless steel for duct, cladding, and wire wrap material were irradiated to demonstrate the improved performance and lifetime capability of this design. A single pinhole-type breach was incurred in one of the high exposure tests after a peak fuel burnup of 155 MWd/kgM and peak fast neutron fluence of 25 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV). Postirradiation examinations were performed on four of the test assemblies and measured results were compared to analytical evaluations. A revised swelling correlation for D9 Alloy was developed to provide improved agreement between calculated and measured cladding deformation results. A fuel pin lifetime design criterion of 5% calculated hoop strain was derived. Alternatively, fuel pin lifetimes were developed for two irradiation parameters using statistical failure analyses. For a 99.99% reliability, the analyses indicated a peak fast fluence lifetime of 21.0 {times} 10{sup 22} n/cm{sup 2}, or a peak fuel burnup greater than 120 MWd/kgM. The extended lifetime capability of this design would reduce fuel supply requirements for the FFTF by a third relative to the reference driver design.

  14. High Mn austenitic stainless steel

    DOEpatents

    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.

  15. Nano-Scale Fission Product Phases in an Irradiated U-7Mo Alloy Nuclear Fuel

    SciTech Connect

    Dennis Keiser, Jr.; Brandon Miller; James Madden; Jan-Fong Jue; Jian Gan

    2014-09-01

    Irradiated nuclear fuel is a very difficult material to characterize. Due to the large radiation fields associated with these materials, they are hard to handle and typically have to be contained in large hot cells. Even the equipment used for performing characterization is housed in hot cells or shielded glove boxes. The result is not only a limitation in the techniques that can be employed for characterization, but also a limitation in the size of features that can be resolved The most standard characterization techniques include light optical metallography (WM), scanning electron microscopy (SEM), and electron probe microanalysis (EPMA). These techniques are applied to samples that are typically prepared using grinding and polishing approaches that will always generate some mechanical damage on the sample surface. As a result, when performing SEM analysis, for example, the analysis is limited by the quality of the sample surface that can be prepared. However, a new approach for characterizing irradiated nuclear fuel has recently been developed at the Idaho National Laboratory (INL) in Idaho Falls, Idaho. It allows for a dramatic improvement in the quality of characterization that can be performed when using an instrument like an SEM. This new approach uses a dual-beam scanning microscope, where one of the beams isa focused ion beam (FIB), which can be used to generate specimens of irradiated fuel (-10µm x 10µm) for microstructural characterization, and the other beam is the electron beam of an SEM. One significant benefit of this approach is that the specimen surface being characterized has received much less damage (and smearing) than is caused by the more traditional approaches, which enables the imaging of nanometer­ sized microstructural features in the SEM. The process details are for an irradiated low-enriched uranium (LEU) U-Mo alloy fuel Another type of irradiated fuel that has been characterized using this technique is a mixed oxide fuel.

  16. Void evolution and porosity under arsenic ion irradiation in GaAs1‑x Sb x alloys

    NASA Astrophysics Data System (ADS)

    Alkhaldi, H. S.; Kluth, P.; Kremer, F.; Lysevych, M.; Li, L.; Ridgway, M. C.; Williams, J. S.

    2017-03-01

    We have studied the formation of porosity in crystalline GaAs0.25Sb0.75 and GaAs0.5Sb0.5 alloys under irradiation with 140 keV As‑ ions over a wide range of temperature (‑180 to 400 °C) and ion fluences ranging from 1× {{10}13} to 2× {{10}17} ions cm‑2. The GaAs0.25Sb0.75 alloy showed only little swelling (in comparison with GaSb), with void formation and sputtering both playing an important role in the materials modification. The initiation of voids and their evolution in the alloy strongly depends on the ion fluence and irradiation temperature, as well as the As content in the alloy. Porosity is largely suppressed in the GaAs0.25Sb0.75 alloy, with the major change being void formation. For the GaAs0.5Sb0.5 alloy, it was rendered amorphous with no apparent pores or void structures and only sputtering effects were observed at high ion fluence. In addition, the transformations from crystalline to amorphous and to a void or a porous structure occurred simultaneously in the GaAs0.25Sb0.75 alloy. The mechanisms responsible for such changes are consistent with point defect movement and segregation.

  17. High strength, tough alloy steel

    DOEpatents

    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.

  18. Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2014-01-01

    The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

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

    SciTech Connect

    Chung, H.M.; Gazda, J.; Smith, D.L.

    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.

  20. Modeling the post-yield flow behavior after neutron and electron irradiation of steels and iron-base alloys.

    SciTech Connect

    Dimelfi, R. J.

    1999-01-13

    Irradiation hardening is an issue of practical importance as it relates to the remanent life and the nature of failure of reactor components exposed to displacement-producing radiation. For example, irradiation-induced yield strength increases in pressure vessel steels are directly related to increases in the ductile-to-brittle-transition-temperature of these materials. Other issues associated with hardening, such as reductions in ductility, toughness and fatigue life of structural steels are also of concern. Understanding these phenomena requires studies of fundamental microstructural mechanisms of hardening. Because of the limited supply of neutron-irradiated surveillance material, difficulties posed by the radioactivity of neutron-exposed samples and the uncertainty of irradiation conditions in this case, fundamental studies are often conducted using well-controlled experiments involving irradiation by electrons instead of neutrons. Also, in such studies, simple model alloys are used in place of steels to focus on the influence of specific alloy constituents. It is, therefore, important to understand the relationship between the results of this kind of experiment and the effects of in-reactor neutron exposure in order to use them to make predictions of significance to reactor component life. In this paper, we analyze the tensile behavior of pressure vessel steels (A212B and A350) irradiated by neutrons and electrons. The results show that the post-yield true stress/true strain behavior can provide fingerprints of the different hardening effects that result from irradiation by the two particles, which also reflect the influence of alloy content. Microstructurally-based models for irradiation-induced yield strength increases, combined with a model for strain hardening, are used to make predictions of the different effects of irradiation by the two particles on the entire flow curve that agree well with data.

  1. Stable atomic structure of NiTi austenite

    NASA Astrophysics Data System (ADS)

    Zarkevich, Nikolai A.; Johnson, Duane D.

    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.

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

    SciTech Connect

    Kohnert, Aaron A.; Dasgupta, Dwaipayan; Wirth, Brian; Linton, Kory D.

    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.

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

    PubMed

    Huang, Chen-Hsi; Marian, Jaime

    2016-10-26

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  6. In situ observation of defect annihilation in Kr ion-irradiated bulk Fe/amorphous-Fe 2 Zr nanocomposite alloy

    DOE PAGES

    Yu, K. Y.; Fan, Z.; Chen, Y.; ...

    2014-08-26

    Enhanced irradiation tolerance in crystalline multilayers has received significant attention lately. However, little is known on the irradiation response of crystal/amorphous nanolayers. We report on in situ Kr ion irradiation studies of a bulk Fe96Zr4 nanocomposite alloy. Irradiation resulted in amorphization of Fe2Zr and formed crystal/amorphous nanolayers. α-Fe layers exhibited drastically lower defect density and size than those in large α-Fe grains. In situ video revealed that mobile dislocation loops in α-Fe layers were confined by the crystal/amorphous interfaces and kept migrating to annihilate other defects. This study provides new insights on the design of irradiation-tolerant crystal/amorphous nanocomposites.

  7. A combined APT and SANS investigation of α' phase precipitation in neutron-irradiated model FeCrAl alloys

    DOE PAGES

    Briggs, Samuel A.; Edmondson, Philip D.; Littrell, Kenneth C.; ...

    2017-03-01

    Here, FeCrAl alloys are currently under consideration for accident-tolerant fuel cladding applications in light water reactors owing to their superior high-temperature oxidation and corrosion resistance compared to the Zr-based alloys currently employed. However, their performance could be limited by precipitation of a Cr-rich α' phase that tends to embrittle high-Cr ferritic Fe-based alloys. In this study, four FeCrAl model alloys with 10–18 at.% Cr and 5.8–9.3 at.% Al were neutron-irradiated to nominal damage doses up to 7.0 displacements per atom at a target temperature of 320 °C. Small angle neutron scattering techniques were coupled with atom probe tomography to assessmore » the composition and morphology of the resulting α' precipitates. It was demonstrated that Al additions partially destabilize the α' phase, generally resulting in precipitates with lower Cr contents when compared with binary Fe-Cr systems. The precipitate morphology evolution with dose exhibited a transient coarsening regime akin to previously observed behavior in aged Fe-Cr alloys. Similar behavior to predictions of the LSW/UOKV models suggests that α' precipitation in irradiated FeCrAl is a diffusion-limited process with coarsening mechanisms similar to those in thermally aged high-Cr ferritic alloys.« less

  8. Pitting corrosion resistant austenite stainless steel

    DOEpatents

    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.

  9. The effects of double austenitization on the mechanical properties of a 0. 34C containing low-alloy Ni-Cr-Mo-V steel

    SciTech Connect

    Chang, E.; Chang, C.Y. . Dept. of Materials Science and Engineering); Liu, C.D. )

    1994-03-01

    This article considers five different microstructures of a tempered martensitic 0.34C, 3Ni-1.3Cr-0.4Mo-0.1V steel through various heat treatments, including double austenitization (DA) treatments, and how the impact toughnesses are influenced by microstructure. Of the four mechanisms considered to explain the beneficial effect of DA treatment, the roles of retained austenite, grain-boundary embrittlement by impurity segregation, and matrix flow stress are discounted. The 50 pct fracture appearance transition temperature (FATT) of this steel is found to be dependent on both the grain size and the carbide dissolution. The conventionally treated steel contains mainly platelike M[sub 3]C carbides. The DA treatment helps to dissolve VC carbides and coarsen and spheroidize M[sub 3]C carbides in favor of the precipitation of short rodlike M7C3 carbides with a lower aspect ratio. The improvement of impact toughness (upper shelf energy, ductile-to-brittle transition temperature (DBTT), and lower shelf energy) by DA treatment, explained in detail, is attributed to a change of this material's tensile and work-hardening behavior affected by a variation of carbide morphology.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    Edwards, D.J.; 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.

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

    NASA Astrophysics Data System (ADS)

    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).

  13. Changes in electromagnetic properties of a low-alloy steel caused by neutron irradiation

    SciTech Connect

    Goto, Toru; Kamimura, Takeo; Kumano, Shintaro; Takeuchi, Iwao; Maeda, Noriyoshi; Yamaguchi, Atsunori

    1999-10-01

    In order to develop a method for the nondestructive evaluation of material deterioration in nuclear pressure vessels, changes in the electromagnetic properties of the low-alloy steel A533B, Class 1 and its weld metal caused by neutron irradiation up to {approximately}3 {times} 10{sup 23} n/m{sup 2} of neutron fluence at 561 K were measured. Electrical resistance, coercivity and Barkhausen noise were selected as the electromagnetic properties to measure. It was found that decreases of several percent in the readings of electrical resistance and coercivity, and an increase of several percent in the Barkhausen noise occurred due to neutron irradiation. Good correlations between the changes in the electromagnetic properties and those in the mechanical properties were confirmed. Furthermore, an equation using the results of the three tests was found to estimate well the transition temperature and yield strength. From this, the authors conclude that the electromagnetic tests have potential as methods for nondestructive evaluation of material deterioration in the reactor vessels of nuclear power plants.

  14. Research on long-time self-irradiation of alloy δ-Pu 242-Ga

    NASA Astrophysics Data System (ADS)

    Somenkov, V. A.; Blanter, M. S.; Glazkov, V. P.; Laushkin, A. V.; Orlov, V. K.

    2011-06-01

    Self-irradiation of Pu-Ga alloy was studied by neutron diffraction, which provided information about the crystalline structure and yielded mean-square atom displacements < u2> from the Debye-Waller factor. The measurements were performed at room temperature using the sample based on the isotope Pu 242 with low neutron absorption cross-section to which the short-living isotope Pu 238 (1.4%) was added to accelerate self-irradiation. This composition of the sample sped up the aging processes by the factor of four and allowed us to obtain the maximum equivalent time of 23.5 years. Fcc structure remains the same throughout this interval. The changes in < u2> due to static displacements occur in two stages, viz. a relatively rapid growth (roughly by 50%) over the first 5-6 equivalent years and slow decline during 6-23 equivalent years making a close approach to the initial values. The latter stage has not been described in the literature. It can be explained by the confluence of point defects into helium bubbles and dislocation loops accumulated over time.

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

    SciTech Connect

    Gazda, J.; Meshii, M.; Chung, H.M.

    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.

  16. A replica technique for extracting precipitates from neutron-irradiated or thermal-aged vanadium alloys for TEM analysis

    NASA Astrophysics Data System (ADS)

    Fukumoto, K.; Iwasaki, M.

    2014-06-01

    A carbon replica technique has been developed to extract precipitates from vanadium alloys. Using this technique, precipitation phases can be extracted from neutron-irradiated or thermal-aged V-4Cr-4Ti alloys. Precipitate identification using EDS X-ray analysis and electron diffraction was facilitated. Only NaCl type of Ti(OCN) precipitate was formed in the thermal-aged V-4Cr-4Ti alloys at 600 °C for 20 h and cation sub-lattice was only occupied by Ti atoms. However, the thin plate of precipitates with NaCl type of crystallographic structure could be seen in the V-4Cr-4Ti alloys irradiated at 593 °C in the JOYO fast reactor. The precipitate contained chromium and vanadium atoms on the cation sub-lattice as well as titanium atoms. It is considered that the phase of MX type (M = Ti, V, Cr and X = O, N, C) is a metastable phase under neutron irradiation.

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

    NASA Technical Reports Server (NTRS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    Wei, Tao; Zhu, Hanliang; Ionescu, Mihail; Dayal, Pranesh; Davis, Joel; Carr, David; Harrison, Robert; Edwards, Lyndon

    2015-04-01

    A 45XD TiAl alloy possessing a lamellar microstructure was irradiated using 5 MeV helium ions to a fluence of 5 × 1021 ion m-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 α2 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.

  20. Effect of annealing on VmHn complexes in hydrogen ion irradiated Fe and Fe-0.3%Cu alloys

    NASA Astrophysics Data System (ADS)

    Zhang, Peng; Jin, Shuoxue; Lu, Eryang; Wang, Baoyi; Zheng, Yongnan; Yuan, Daqing; Cao, Xingzhong

    2015-04-01

    The effect of annealing on VmHn complexes and Cu precipitate behaviours in hydrogen ion irradiated Fe and Fe-0.3%Cu alloys was investigated by positron annihilation spectroscopy using a slow positron beam. The results of S parameters indicated that the room temperature irradiation was benefit for the formation of the VmHn complex compared to the elevated temperature irradiation. The S-W results confirmed the formation of Cu precipitates in Fe-0.3%Cu even at the irradiation dose of 0.1 dpa. The formation of the evident S value peaks in the damage region after annealing treatment suggested that the VmHn complexes were broken and a larger of hydrogen atoms were escaping. The residual vacancy defects would migrate towards both the surface region and the opposite direction with the increasing annealing temperature.

  1. Atomistic simulation of defects formation and structure transitions in U-Mo alloys at swift heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Kolotova, L. N.; Starikov, S. V.

    2017-01-01

    At irradiation of swift heavy ions, the track formation frequently takes place in nuclear materials. There is a large interest to understanding of the mechanisms of defects/track formation at this phenomenon. In this work, the atomistic simulation of defects formation and melting in U-Mo alloys at irradiation of swift heavy ions has been carried out. We use the two-temperature atomistic model with explicit account of electron pressure and electron thermal conductivity. This two-temperature model describes ionic subsystem by means of molecular dynamics while the electron subsystem is considered in the continuum approach. The various mechanisms of structure changes at irradiation are examined. In particular, the simulation results indicate that the defects formation may be produced without melting and subsequent crystallization. Threshold stopping power of swift ions for the defects formation at irradiation in the various conditions are calculated.

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

    PubMed Central

    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

  3. Mechanism of Austenite Formation from Spheroidized Microstructure in an Intermediate Fe-0.1C-3.5Mn Steel

    NASA Astrophysics Data System (ADS)

    Lai, Qingquan; Gouné, Mohamed; Perlade, Astrid; Pardoen, Thomas; Jacques, Pascal; Bouaziz, Olivier; Bréchet, Yves

    2016-07-01

    The austenitization from a spheroidized microstructure during intercritical annealing was studied in a Fe-0.1C-3.5Mn alloy. The austenite grains preferentially nucleate and grow from intergranular cementite. The nucleation at intragranular cementite is significantly retarded or even suppressed. The DICTRA software, assuming local equilibrium conditions, was used to simulate the austenite growth kinetics at various temperatures and for analyzing the austenite growth mechanism. The results indicate that both the mode and the kinetics of austenite growth strongly depend on cementite composition. With sufficiently high cementite Mn content, the austenite growth is essentially composed of two stages, involving the partitioning growth controlled by Mn diffusion inside ferrite, followed by a stage controlled by Mn diffusion within austenite for final equilibration. The partitioning growth results in a homogeneous distribution of carbon within austenite, which is supported by NanoSIMS carbon mapping.

  4. Fluctuations of chemical composition of austenite and their consequence on shape memory effect in Fe-Mn-(Si, Cr, Ni, C, N) alloys

    SciTech Connect

    Bliznuk, V.V.; Gavriljuk, V.G. . E-mail: gavr@imp.kiev.ua; Kopitsa, G.P.; Grigoriev, S.V.; Runov, V.V.

    2004-09-20

    Polycrystalline samples of shape memory iron-based alloys containing 17, and 30 mass% Mn and alloyed with Si, Cr, Ni, C, N were studied by means of small angle scattering of polarized neutrons (SAPNS). A direct correlation between chemical homogeneity of the Fe-Mn, Fe-Mn-Si, Fe-Mn-Si-Cr, Fe-Mn-Si-Cr-Ni solid solutions and the values of reversible strain caused by the {gamma} {yields} {epsilon} {yields} {gamma} martensitic transformation was found. The addition of silicon to the Fe-Mn alloys significantly improves chemical homogeneity of the fcc solid solution on the scale of larger than several nm, which correlates with the essential increase of reversible strain. A similar to silicon but weaker effect was observed in the case of nitrogen addition to the Fe-Mn-Si-Cr, Fe-Mn-Si-Cr-Ni alloys. Based on the obtained experimental data and in consistency with the previously expressed idea by Sade et al., the positive effect of silicon and nitrogen on chemical homogeneity and SME in Fe-Mn alloys is attributed to the short-range atomic ordering induced by these elements.

  5. Effects of alloying elements on radiation hardening based on loop formation of electron-irradiated light water reactor pressure vessel model steels

    NASA Astrophysics Data System (ADS)

    Nishi, Takakuni; Hashimoto, N.; Ohnuki, S.; Yamamoto, T.; Odette, G. R.

    2011-10-01

    Electron irradiations using a high voltage electron microscope were conducted on several reactor pressure vessel model alloys in order to investigate the effects of alloying elements on the formation and development of defect clusters. In addition, the effects of alloying elements on yield stress change after irradiation were considered, comparing the mean size and number density of dislocation loops with the irradiation-induced hardening. High Cu alloys formed Cu and Mn-Ni-Si rich clusters, and these are important in determining the yield stress increase. High Ni alloys formed a high density of small dislocation loops and probably Mn-Ni-Si rich cluster, which have the effect of increasing the yield stress. High P enhanced radiation-induced segregation on grain boundary, helping prevent dislocation movement.

  6. Mechanical properties of high-nickel alloys EP-753 and РЕ-16 after neutron irradiation to 54 dpa at 400-650 °С

    NASA Astrophysics Data System (ADS)

    Konobeev, Yu. V.; Porollo, S. I.; Ivanov, A. A.; Shulepin, S. V.; Budylkin, N. I.; Mironova, E. G.; Garner, F. A.

    2011-05-01

    Short-term mechanical properties and void swelling were investigated for high-nickel alloys РЕ-16 and three compositional variants of Russian alloy EP-753 and in various starting conditions after side-by-side irradiation in the BN-350 fast reactor at 400, 500, 600 and 650 °С to 54 dpa. For both alloys irradiation resulted in significant hardening and ductility reduction dependent on their chemical composition and initial heat treatment. At test temperatures equal to the irradiation values both alloys exhibited a high level of strength and satisfactory ductility. In the test temperature range of 550-650 °С the phenomenon of high-temperature irradiation embrittlement was observed.

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

    SciTech Connect

    Rowcliffe, A.F.; Tsai, H.C.; Smith, D.L.

    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.

  8. Nanopatterns induced by pulsed laser irradiation on the surface of an Fe-Al alloy and their magnetic properties

    SciTech Connect

    Yoshida, Yutaka; Oosawa, Kazuya; Watanabe, Seiichi; Kaiju, Hideo; Kondo, Kenji; Ishibashi, Akira; Yoshimi, Kyosuke

    2013-05-06

    We have studied nanopatterns induced by nanosecond pulsed laser irradiation on (111) plane surfaces of a polycrystalline iron-aluminum alloy and evaluated their magnetic properties. Multiple nanosecond pulsed laser irradiation induces a wavelength-dependent surface transformation of the lattice structure from a B2-type to a supersaturated body centered cubic lattice. The selective formation of surface nanopatterns consisting of holes, stripes, polygonal networks, and dot-like nanoprotrusions can be observed. Furthermore, focused magneto-optical Kerr effect measurements reveal that the magnetic properties of the resultant nanostructured region changes from a paramagnetic to a ferromagnetic phase in accordance with the number of laser pulses.

  9. Role of the Bogachev - Mints Concept of Metastability of Austenite in Choosing Wear-Resistant Materials

    NASA Astrophysics Data System (ADS)

    Schastlivtsev, V. M.; Filippov, M. A.

    2005-01-01

    The significance of the Bogachev - Mints concept of metastability of austenite for the choice of strain-hardenable steel, cast iron, and facing alloys resisting mechanical kinds of wear (cavitation-, erosion-, and abrasion-induced) is discussed.

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

    SciTech Connect

    Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J.

    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.

  11. Self-organization of Cu-based immiscible alloys under irradiation: An atom-probe tomography study

    NASA Astrophysics Data System (ADS)

    Stumphy, Brad D.

    The stability of materials subjected to prolonged irradiation has been a topic of renewed interest in recent years due to the projected growth of nuclear power as an alternative energy source. The irradiating particles impart energy into the material, thereby causing atomic displacements to occur. These displacements result in the creation of point defects and the random ballistic mixing of the atoms. Consequently, the material is driven away from its equilibrium structure. The supersaturation of defects can lead to the degradation of mechanical properties, but a high density of internal interfaces, which act as defect sinks, will suppress the supersaturation and long-range transport of defects. The microstructural evolution of the material is controlled by the ballistic mixing as well as the mobility of the point defects. In immiscible alloys, these two processes compete against one another, as the ballistic mixing acts to solutionize the alloy components, and the thermal diffusion of the large number of defects acts to phase separate the components. The work presented in this dissertation examines the effect of heavy-ion irradiation on immiscible, binary Cu-based alloys. Dilute alloys of Cu-Fe, Cu-V, and V-Cu have been subjected to irradiation, and atom-probe tomography has been utilized in order to better understand the complex nature of the response of these simple model systems to an irradiation environment. The results show that a steady-state, nano-scale patterning structure, with a high density of unsaturable defect sinks, can be maintained under prolonged irradiation. Additionally, precipitation from a supersaturated solid solution is shown to be a function of both the thermal diffusion and the ballistic mixing. Solvent-rich secondary precipitates, termed "cherry-pits," are observed inside of the solute-rich primary precipitates. Through a combination of simulation work and analyzing multiple alloys experimentally, it was determined that this cherry

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

    SciTech Connect

    Garner, F.A.; Oliver, B.M.; Greenwood, L.R.

    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.

  13. Positron Annihilation Spectroscopy and Small Angle Neutron Scattering Characterization of Nanostructural Features in Irradiated Fe-Cu-Mn Alloys

    SciTech Connect

    Wirth, B D; Asoka-Kumar, P; Howell, R H; Odette, G R; Sterne, P A

    2001-01-01

    Radiation embrittlement of nuclear reactor pressure vessel steels results from a high number density of nanometer sized Cu-Mn-Ni rich precipitates (CRPs) and sub-nanometer matrix features, thought to be vacancy-solute cluster complexes (VSC). However, questions exist regarding both the composition of the precipitates and the defect character and composition of the matrix features. We present results of positron annihilation spectroscopy (PAS) and small angle neutron scattering (SANS) characterization of irradiated and thermally aged Fe-Cu and Fe-Cu-Mn alloys. These complementary techniques provide insight into the composition and character of both types of nanoscale features. The SANS measurements indicate populations of CRPs and VSCs in both alloys. The CRPs are coarser in the Fe-Cu alloy and the number densities of CRP and VSC increase with the addition of Mn. The PAS involved measuring both the positron lifetimes and the Doppler broadened annihilation spectra in the high momentum region to provide elemental sensitivity at the annihilation site. The spectra in Fe-Cu-Mn specimens thermally aged to peak hardness at 450 C and irradiated at 288 C are nearly identical to elemental Cu. Positron lifetime and spectrum measurements in Fe-Cu specimens irradiated at 288 C clearly show the existence of long lifetime ({approx}500 ps) open volume defects, which also contain Cu. Thus the SANS and PAS provide a self-consistent picture of nanostructures composed of CRPs and VSCs and tend to discount high Fe concentrations in the CRPs.

  14. Influence of transmutation on microstructure, density change, and embrittlement of vanadium and vanadium alloys irradiated in HFIR

    SciTech Connect

    Ohnuki, S.; Takahashi, H.; Shiba, K.; Hishinuma, A.; Pawel, J.; Garner, F.A.

    1994-06-01

    Addition of 1 at.% nickel to vanadium and V-10Ti, followed by irradiation along with the nickel-free metals in HFIR to 2.3 {times} 10{sup 22}n cm{sup {minus}2}, E > 0.1MeV (corresponding to 17.7 dpa) at 400 C, has been used to study the influence of helium on microstructural evolution and embrittlement. Approximately 15.3% of the vanadium transmuted to chromium in these alloys. The {approximately}50 appm helium generated from the {sup 58}Ni(n,{gamma}){sup 59}Ni(n,{alpha}){sup 56}Fe sequence was found to exert much less influence than either the nickel directly or the chromium formed by transmutation. The V-10Ti and V-10Ti-1Ni alloys developed an extreme fragility and broke into smaller pieces in response to minor physical insults during density measurements. A similar behavior was not observed in pure V or V-1Ni. Helium`s role in determination of mechanical properties and embrittlement of vanadium alloys in HFIR is overshadowed by the influence of alloying elements such as titanium and chromium. Both elements have been shown to increase the ductile-to-brittle transition temperature (DBTT) rather rapidly in the region of 10% (Cr + Ti). Since Cr is produced by transmutation of V, this is a possible mechanism for the embrittlement. Large effects on the DBTT may have also resulted from uncontrolled accumulation of interstitial elements such as C, N, and O during irradiation.

  15. Heat treatment giving a stable high temperature micro-structure in cast austenitic stainless steel

    DOEpatents

    Anton, Donald L.; Lemkey, Franklin D.

    1988-01-01

    A novel micro-structure developed in a cast austenitic stainless steel alloy and a heat treatment thereof are disclosed. The alloy is based on a multicomponent Fe-Cr-Mn-Mo-Si-Nb-C system consisting of an austenitic iron solid solution (.gamma.) matrix reinforced by finely dispersed carbide phases and a heat treatment to produce the micro-structure. The heat treatment includes a prebraze heat treatment followed by a three stage braze cycle heat treatment.

  16. Microstructure of RERTR Du-Alloys Irradiated with Krypton Ions up to 100 dpa

    SciTech Connect

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

    2011-04-01

    The radiation stability of the interaction product formed at the fuel–matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al)3, (U, Mo)(Si, Al)3, UMo2Al20, U6Mo4Al43, and UAl4. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 °C to doses up to 100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al)3 and UAl4 phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al)3 and UMo2Al20 phases become amorphous at 1 and 2 dpa, respectively, and show no evidence of voids at 100 dpa. The U6Mo4Al43 phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

  17. Atomistic simulations of deuterium irradiation on iron-based alloys in future fusion reactors

    DOE PAGES

    Safi, E.; Polvi, J.; Lasa, A.; ...

    2016-10-14

    Iron-based alloys are now being considered as plasma-facing materials for the first wall of future fusion reactors. Therefore, the iron (Fe) and carbon (C) erosion will play a key role in predicting the life-time and viability of reactors with steel walls. In this work, the surface erosion and morphology changes due to deuterium (D) irradiation in pure Fe, Fe with 1% C impurity and the cementite, are studied using molecular dynamics (MD) simulations, varying surface temperature and impact energy. The sputtering yields for both Fe and C were found to increase with incoming energy. In iron carbide, C sputtering wasmore » preferential to Fe and the deuterium was mainly trapped as D2 in bubbles, while mostly atomic D was present in Fe and Fe–1%C. The sputtering yields obtained from MD were compared to SDTrimSP yields. At lower impact energies, the sputtering mechanism was of both physical and chemical origin, while at higher energies (>100 eV) the physical sputtering dominated.« less

  18. Atomistic simulations of deuterium irradiation on iron-based alloys in future fusion reactors

    SciTech Connect

    Safi, E.; Polvi, J.; Lasa, A.; Nordlund, K.

    2016-10-14

    Iron-based alloys are now being considered as plasma-facing materials for the first wall of future fusion reactors. Therefore, the iron (Fe) and carbon (C) erosion will play a key role in predicting the life-time and viability of reactors with steel walls. In this work, the surface erosion and morphology changes due to deuterium (D) irradiation in pure Fe, Fe with 1% C impurity and the cementite, are studied using molecular dynamics (MD) simulations, varying surface temperature and impact energy. The sputtering yields for both Fe and C were found to increase with incoming energy. In iron carbide, C sputtering was preferential to Fe and the deuterium was mainly trapped as D2 in bubbles, while mostly atomic D was present in Fe and Fe–1%C. The sputtering yields obtained from MD were compared to SDTrimSP yields. At lower impact energies, the sputtering mechanism was of both physical and chemical origin, while at higher energies (>100 eV) the physical sputtering dominated.

  19. Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

    DOE PAGES

    Jin, Ke; Guo, Wei; Lu, Chenyang; ...

    2016-12-01

    Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 1013 to 3 × 1016 cm-2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluence regime, which ismore » consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.« less

  20. Effects of Fe concentration on the ion-irradiation induced defect evolution and hardening in Ni-Fe solid solution alloys

    SciTech Connect

    Jin, Ke; Guo, Wei; Lu, Chenyang; Ullah, Mohammad W.; Zhang, Yanwen; Weber, William J.; Wang, Lumin; Poplawsky, Jonathan D.; Bei, Hongbin

    2016-12-01

    Understanding alloying effects on the irradiation response of structural materials is pivotal in nuclear engineering. In order to systematically explore the effects of Fe concentration on the irradiation-induced defect evolution and hardening in face-centered cubic Ni-Fe binary solid solution alloys, single crystalline Ni-xFe (x = 0–60 at%) alloys have been grown and irradiated with 1.5 MeV Ni ions. The irradiations have been performed over a wide range of fluences from 3 × 1013 to 3 × 1016 cm-2 at room temperature. Ion channeling technique has shown reduced damage accumulation with increasing Fe concentration in the low fluence regime, which is consistent to the results from molecular dynamic simulations. We did not observe any irradiation-induced compositional segregation in atom probe tomography within the detection limit, even in the samples irradiated with high fluence Ni ions. Transmission electron microscopy analyses have further demonstrated that the defect size significantly decreases with increasing Fe concentration, indicating a delay in defect evolution. Furthermore, irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but they all follow a similar trend for the increase of hardness as a function of irradiation fluence.

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

    NASA Astrophysics Data System (ADS)

    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.

  2. Peculiarities of structure and hardening of Ni-Ti alloy surface layers formed by 84Kr15+ ions irradiation at 147 MeV energy at high temperatures

    NASA Astrophysics Data System (ADS)

    Poltavtseva, V.; Larionov, A.; Zheltova, G.

    2017-01-01

    The consistent patterns of changes in nanostructure and nanohardness of Ni-Ti alloy after irradiation with 84Kr15+ ions with 147 MeV energy to the fluence of 1·1019 m-2 at 250 and 3000C temperatures depending on phase composition have been experimentally studied. It was shown that significant (44 – 94%) softening of surface layers for the single-phase and two-phase Ni-Ti alloys is connected with the formation of bubble nanostructured defects and complete sputtering of the process layers. The role of nanostructure in roughness of the irradiated Ni-Ti alloy surface of various phase composition has been established.

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    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

  5. Multicomponent interdiffusion in austenitic nickel-, iron-nickel-base alloys and L1(2)-nickel-aluminum intermetallic for high-temperature applications

    NASA Astrophysics Data System (ADS)

    Garimella, Narayana

    Interdiffusion in multicomponent-multiphase alloys is commonly encountered in many materials systems. The developments of multicomponent-multiphase alloys require control of microstructure through appropriate heat treatment, involving solid-state transformations, precipitation processes, and surface modification, where the interdiffusion processes play a major role. In addition, interdiffusion processes often control degradation and failure of these materials systems. Enhanced performance and reliable durability always requires a detailed understanding of interdiffusion. In this study, ternary and quaternary interdiffusion in Ni-Cr-X (X = Al, Si, Ge, Pd) at 900°C and 700°C, Fe-Ni-Cr-X (X = Si, Ge) at 900°C, and Ni3Al alloyed with Ir, Ta and Re at 1200°C were examined using solid-to-solid diffusion couples. Interdiffusion fluxes of individual components were calculated directly from experimental concentration profiles determined by electron probe microanalysis. Moments of interdiffusion fluxes were examined to calculate main and cross interdiffusion coefficients averaged over selected composition ranges from single diffusion couple experiments. Consistency in the magnitude and sign of ternary and quaternary interdiffusion coefficient were verified with interdiffusion coefficients determined by Boltzmann-Matano analysis that requires multiple diffusion couples with intersecting compositions. Effects of alloying additions, Al, Si, Ge and Pd, on the interdiffusion in Ni-Cr-X and Fe-Ni-Cr-X alloys were examined with respect to Cr2O 3-forming ability at high temperature. Effects of Ir, Ta and Re additions on interdiffusion in Ni3Al were examined with respect to phase stability and site-preference. In addition, a numerically refined approach to determine average ternary interdiffusion coefficients were developed. Concentrations and moments of interdiffusion fluxes are employed to generate multiple combinations of multicomponent interdiffusion coefficient as a function

  6. Manganese-stabilized austenitic stainless steels for fusion applications

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.

    1990-08-07

    An austenitic stainless steel that is comprised of Fe, Cr, Mn, C but no Ni or Nb and minimum N. To enhance strength and fabricability minor alloying additions of Ti, W, V, B and P are made. The resulting alloy is one that can be used in fusion reactor environments because the half-lives of the elements are sufficiently short to allow for handling and disposal.

  7. Manganese-stabilized austenitic stainless steels for fusion applications

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.

    1990-01-01

    An austenitic stainless steel that is comprised of Fe, Cr, Mn, C but no Ni or Nb and minimum N. To enhance strength and fabricability minor alloying additions of Ti, W, V, B and P are made. The resulting alloy is one that can be used in fusion reactor environments because the half-lives of the elements are sufficiently short to allow for handling and disposal.

  8. Method for residual stress relief and retained austenite destabilization

    DOEpatents

    Ludtka, Gerard M.

    2004-08-10

    A method using of a magnetic field to affect residual stress relief or phase transformations in a metallic material is disclosed. In a first aspect of the method, residual stress relief of a material is achieved at ambient temperatures by placing the material in a magnetic field. In a second aspect of the method, retained austenite stabilization is reversed in a ferrous alloy by applying a magnetic field to the alloy at ambient temperatures.

  9. Development of Cast Alumina-Forming Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    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.

  10. The influence of silicon and aluminum on austenite deformation behavior during fatigue and tensile loading

    NASA Astrophysics Data System (ADS)

    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

  11. Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe-Cr model alloys

    NASA Astrophysics Data System (ADS)

    Bergner, F.; Pareige, C.; Hernández-Mayoral, M.; Malerba, L.; Heintze, C.

    2014-05-01

    An attempt is made to quantify the contributions of different types of defect-solute clusters to the total irradiation-induced yield stress increase in neutron-irradiated (300 °C, 0.6 dpa), industrial-purity Fe-Cr model alloys (target Cr contents of 2.5, 5, 9 and 12 at.% Cr). Former work based on the application of transmission electron microscopy, atom probe tomography, and small-angle neutron scattering revealed the formation of dislocation loops, NiSiPCr-enriched clusters and α‧-phase particles, which act as obstacles to dislocation glide. The values of the dimensionless obstacle strength are estimated in the framework of a three-feature dispersed-barrier hardening model. Special attention is paid to the effect of measuring errors, experimental details and model details on the estimates. The three families of obstacles and the hardening model are well capable of reproducing the observed yield stress increase as a function of Cr content, suggesting that the nanostructural features identified experimentally are the main, if not the only, causes of irradiation hardening in these model alloys.

  12. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

    NASA Astrophysics Data System (ADS)

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; Wang, Lumin

    2016-02-01

    Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.

  13. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

    DOE PAGES

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; ...

    2016-02-01

    We report that energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters farmore » exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.« less

  14. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

    SciTech Connect

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; Wang, Lumin

    2016-02-01

    We report that energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.

  15. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys

    PubMed Central

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; Wang, Lumin

    2016-01-01

    Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance. PMID:26829570

  16. Mechanical property changes and microstructures of dispersion-strengthened copper alloys after neutron irradiation at 411, 414, and 529 degree C

    SciTech Connect

    Anderson, K.R.; Stubbins, J.F. ); Garner, F.A.; Hamilton, M.L. )

    1990-06-01

    Dispersion strengthened copper alloys have shown promise for certain high heat flux applications in both near term and long term fusion devices. This study examines mechanical properties changes and microstructural evolution in several oxide dispersion strengthened alloys which were subjected to high levels of irradiation-induced displacement damage. Irradiations were carried out in FFTF to 34 and 50 dpa at 411--414{degree}C and 32 dpa at 529{degree}C. The alloys include several oxide dispersion-strengthened alloys based on the Cu-Al system, as well as ones based on the Cu-Cr and Cu-Hf systems. Of this group, certain of the Cu-Al alloys, those produced by an internal oxidation technique to contain alumina weight fractions of 0.15 to 0.25% outperformed the other alloys in all respects. These alloys, designated CuAl15, CuAl20, and CuAl25, were found to be resistant to void swelling up to 50 dpa at 414{degree}C, and to retain their superior mechanical and physical properties after extended irradiation. The major factor which controls the stability during irradiation was found to be the dispersoid volume fraction and distribution. The other alloys examined were less resistant to radiation-induced properties changes for a variety of reasons. Some of these include dispersoid redistribution by ballistic resolution, effects of retained dissolved oxygen, and non-uniformity of dispersion distribution. The effect of laser welding was also examined. This joining technique was found to be unacceptable since it destroys the dispersoid distribution and thereby the resistance of the alloys to radiation-induced damage.

  17. Oxidation resistant high creep strength austenitic stainless steel

    DOEpatents

    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.

  18. Subcascade formation ratio in neutron-irradiated stainless steels

    NASA Astrophysics Data System (ADS)

    Yoshiie, T.; Satoh, Y.; Huang, S. S.; Horiki, M.; Sato, K.; Xu, Q.

    2016-01-01

    High-energy-particle irradiation in metals produces cascade damage. If the particle energy is high enough, a cascade is divided into subcascades. In each subcascade, a vacancy rich area is surrounded by an interstitial area. Vacancy clusters are expected to form directly in the vacancy rich area. In this study, the vacancy cluster formation ratio in subcascades was estimated by positron annihilation lifetime spectroscopy and transmission electron microscopy in commercial stainless steels and their model alloys. The vacancy cluster formation ratio was 1.7×10-3 and 9.1×10-5 in austenitic stainless steel and ferritic/martensitic stainless steel, respectively

  19. Detection of helium in irradiated Fe9Cr alloys by coincidence Doppler broadening of slow positron annihilation

    NASA Astrophysics Data System (ADS)

    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.

  20. Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking

    SciTech Connect

    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.

  1. Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

    NASA Astrophysics Data System (ADS)

    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.

  2. Revised ANL-reported tensile data for unirradiated and irradiated (FFTF, HFIR) V-Ti and V-Cr-Ti alloys

    SciTech Connect

    Billone, M.C.

    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.

  3. The development of alumina-forming austenitic stainless steels for high-temperature structural use

    SciTech Connect

    Brady, Michael P; Yamamoto, Yukinori; Santella, Michael L; Maziasz, Philip J; Pint, Bruce A; Lu, Zhao Ping; Liu, Chain T; Bei, Hongbin

    2008-01-01

    Efforts at Oak Ridge National Laboratory to developAl2O3-forming austenitic (AFA) stainless steels for high-temperature (600-900 aC) structural use under aggressive oxidizing conditions are overviewed. Data obtained to date indicate the potential to achieve superior oxidation resistance to conventional Cr2O3-forming Fe- and Ni-base heat-resistant alloys, with creep strength comparable to state-of-the-art advanced austenitic stainless steels. Preliminary assessment also indicates the developed alloys are amenable to welding. Details of the alloy design approach and composition-microstructure-property relationships are presented.

  4. Effects of heavy-ion irradiation on the grain boundary chemistry of an oxide-dispersion strengthened Fe-12 wt.% Cr alloy

    NASA Astrophysics Data System (ADS)

    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.

  5. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    SciTech Connect

    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.

  6. Texture evolution of cold rolled and reversion annealed metastable austenitic CrMnNi steels

    NASA Astrophysics Data System (ADS)

    Weidner, A.; Fischer, K.; Segel, C.; Schreiber, G.; Biermann, H.

    2015-04-01

    A thermo-mechanical process consisting of cold rolling and subsequent reversion annealing was applied to high-alloy metastable austenitic CrMnNi steels with different nickel contents. As a result of the reversion annealing ultrafine grained material with a grain size in the range between 500 nm up to 4 μm were obtained improving the strength behavior of the material. The evolution of the texture of both the cold rolled states and the reversion-annealed states was studied either by X-ray diffraction or by EBSD measurements. The nickel content has a significant influence on the austenite stability and consequently also on the amount of the martensitic phase transformation. However, the developed textures in both steel variants with different austenite stability revealed the same behavior. In both investigated steels the texture of the reverted austenite is a pronounced Bs-type texture as developed also for the deformed austenite

  7. Effect of double ion implantation and irradiation by Ar and He ions on nano-indentation hardness of metallic alloys

    NASA Astrophysics Data System (ADS)

    Dayal, P.; Bhattacharyya, D.; Mook, W. M.; Fu, E. G.; Wang, Y.-Q.; Carr, D. G.; Anderoglu, O.; Mara, N. A.; Misra, A.; Harrison, R. P.; Edwards, L.

    2013-07-01

    In this study, the authors have investigated the combined effect of a double layer of implantation on four different metallic alloys, ODS steel MA957, Zircaloy-4, Ti-6Al-4V titanium alloy and stainless steel 316, by ions of two different species - He and Ar - on the hardening of the surface as measured by nano-indentation. The data was collected for a large number of indentations using the Continuous Stiffness Method or "CSM" mode, applying the indents on the implanted surface. Careful analysis of the data in the present investigations show that the relative hardening due to individual implantation layers can be used to obtain an estimate of the relative hardening effect of a combination of two separate implanted layers of two different species. This combined hardness was found to lie between the square root of the sum of the squares of individual hardening effects, (ΔHA2 + ΔHB2)0.5 as the lower limit and the sum of the individual hardening effects, (ΔHA + ΔHB) as the upper limit, within errors, for all depths measured. The hardening due to irradiation by different species of ions was calculated by subtracting the average hardness vs. depth curve of the un-irradiated or "virgin" material from that of the irradiated material. The combined hardening of the irradiated samples due to Ar and He irradiation was found to be described well by an approximate upper bound given by the simple linear sum of the individual hardening (L) and a lower bound given by the square root of the sum of the squares (R) of the individual hardening effects due to Ar and He irradiation along the full depth of the indentation. The peak of the combined hardness of Ar and He irradiated material appears at the depth predicted by both the R and the L curves, in all samples. The combined hardness increase due to Ar and He irradiation lies near the upper limit (L curve) for the ODS steel MA957, somewhere in between L and R curves for Zircaloy-4, and near the R curve for the stainless steel 316

  8. Neutron Absorbing Alloys

    SciTech Connect

    Mizia, Ronald E.; Shaber, Eric L.; DuPont, John N.; Robino, Charles V.; Williams, David B.

    2004-05-04

    The present invention is drawn to new classes of advanced neutron absorbing structural materials for use in spent nuclear fuel applications requiring structural strength, weldability, and long term corrosion resistance. Particularly, an austenitic stainless steel alloy containing gadolinium and less than 5% of a ferrite content is disclosed. Additionally, a nickel-based alloy containing gadolinium and greater than 50% nickel is also disclosed.

  9. The Effect of H and He on Irradiation Performance of Fe and Ferritic Alloys

    SciTech Connect

    James F. Stubbins

    2010-01-22

    This research program was designed to look at basic radiation damage and effects and mechanical properties in Fe and ferritic alloys. The program scope included a number of materials ranging from pure single crystal Fe to more complex Fe-Cr-C alloys. The range of materials was designed to examine materials response and performance on ideal/model systems and gradually move to more complex systems. The experimental program was coordinated with a modeling effort. The use of pure and model alloys also facilitated the ability to develop and employ atomistic-scale modeling techniques to understand the inherent physics underlying materials performance

  10. Assessment of the performance potential of the martensitic alloy HT-9 for liquid-metal fast-breeder-reactor applications

    SciTech Connect

    Straalsund, J.L.; Gelles, D.S.

    1983-05-01

    Martensitic stainless steels appear to provide attractive alternatives to austenitic stainless steels for liquid metal fast breeder reactors (LMFBR). The United States National Cladding/Duct (NCD) Materials Development Program has selected Sandvik alloy HT-9 (12CrMoW) as one of six prime candidate alloys for advanced in-core structural materials having very high peak burnup capabilities. The NCD program, since 1974, has been accumulating engineering data for HT-9. Properties include swelling, irradiation creep and microstructure as a function of fluence, postirradiation mechanical properties, thermal creep, sodium compatibility and hardware fabrication technology. Tests results are presented.

  11. Phase-field Model for Interstitial Loop Growth Kinetics and Thermodynamic and Kinetic Models of Irradiated Fe-Cr Alloys

    SciTech Connect

    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.

  12. Effet d'un enrichissement en nickel sur la stabilite mecanique de l'austenite de reversion lorsque soumise a de la fatigue oligocyclique

    NASA Astrophysics Data System (ADS)

    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

  13. Austenitic stainless steel for high temperature applications

    DOEpatents

    Johnson, Gerald D.; Powell, Roger W.

    1985-01-01

    This invention describes a composition for an austenitic stainless steel which has been found to exhibit improved high temperature stress rupture properties. The composition of this alloy is about (in wt. %): 12.5 to 14.5 Cr; 14.5 to 16.5 Ni; 1.5 to 2.5 Mo; 1.5 to 2.5 Mn; 0.1 to 0.4 Ti; 0.02 to 0.08 C; 0.5 to 1.0 Si; 0.01 maximum, N; 0.02 to 0.08 P; 0.002 to 0.008 B; 0.004-0.010 S; 0.02-0.05 Nb; 0.01-0.05 V; 0.005-0.02 Ta; 0.02-0.05 Al; 0.01-0.04 Cu; 0.02-0.05 Co; 0.03 maximum, As; 0.01 maximum, O; 0.01 maximum, Zr; and with the balance of the alloy being essentially iron. The carbon content of the alloy is adjusted such that wt. % Ti/(wt. % C+wt. % N) is between 4 and 6, and most preferably about 5. In addition the sum of the wt. % P+wt. % B+wt. % S is at least 0.03 wt. %. This alloy is believed to be particularly well suited for use as fast breeder reactor fuel element cladding.

  14. Nano-scale chemical evolution in a proton-and neutron-irradiated Zr alloy

    NASA Astrophysics Data System (ADS)

    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.

  15. Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

    SciTech Connect

    Marquis, Emmanuelle; Wirth, Brian; Was, Gary

    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

  16. Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation

    SciTech Connect

    Long, Fei; Daymond, Mark R. Yao, Zhongwen; Kirk, Marquis A.

    2015-03-14

    The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic 〈a〉 dislocations has been dynamically observed before and after irradiation at room temperature and 300 °C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by 〈a〉 dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, (011{sup ¯}1)〈01{sup ¯}12〉 twinning was identified in the irradiated sample deformed at 300 °C.

  17. Microstructural evolution in NF616 (P92) and Fe-9Cr-0.1C-model alloy under heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Topbasi, Cem; Kaoumi, Djamel; Motta, Arthur T.; Kirk, Mark A.

    2015-11-01

    In this comparative study, in situ investigations of the microstructure evolution in a Fe-9Cr ferritic-martensitic steel, NF616, and a Fe-9Cr-0.1C-model alloy with a similar ferritic-martensitic microstructure have been performed. NF616 and Fe-9Cr-0.1C-model alloy were irradiated to high doses (up to ∼10 dpa) with 1 MeV Kr ions between 50 and 673 K. Defect cluster density increased with dose and saturated in both alloys. The average size of defect clusters in NF616 was constant between 50 and 573 K, on the other hand average defect size increased with dose in Fe-9Cr-0.1C-model alloy around ∼1 dpa. At low temperatures (50-298 K), alignment of small defect clusters resulted in the formation of extensive defects in Fe-9Cr-0.1C-model alloy around ∼2-3 dpa, while similar large defects in NF616 started to form at a high temperature of 673 K around ∼5 dpa. Interaction of defect clusters with the lath boundaries were found to be much more noticeable in Fe-9Cr-0.1C-model alloy. Differences in the microstructural evolution of NF616 and Fe-9Cr-0.1C-model alloy are explained by means of the defect cluster trapping by solute atoms which depends on the solute atom concentrations in the alloys.

  18. Microstructural evolution in NF616 (P92) and Fe–9Cr–0.1C-model alloy under heavy ion irradiation

    SciTech Connect

    Topbasi, Cem; Kaoumi, Djamel; Motta, Arthur T.; Kirk, Mark A.

    2015-11-01

    In this comparative study, in situ investigations of the microstructure evolution in a Fee9Cr ferritic emartensitic steel, NF616, and a Fee9Cre0.1C-model alloy with a similar ferriticemartensitic microstructure have been performed. NF616 and Fee9Cre0.1C-model alloy were irradiated to high doses (up to ~10 dpa) with 1 MeV Kr ions between 50 and 673 K. Defect cluster density increased with dose and saturated in both alloys. The average size of defect clusters in NF616 was constant between 50 and 573 K, on the other hand average defect size increased with dose in Fee9Cre0.1C-model alloy around ~1 dpa. At low temperatures (50e298 K), alignment of small defect clusters resulted in the formation of extensive defects in Fee9Cre0.1C-model alloy around ~2e3 dpa, while similar large defects in NF616 started to form at a high temperature of 673 K around ~5 dpa. Interaction of defect clusters with the lath boundaries were found to be much more noticeable in Fee9Cre0.1C-model alloy. Differences in the microstructural evolution of NF616 and Fee9Cre0.1C-model alloy are explained by means of the defect cluster trapping by solute atoms which depends on the solute atom concentrations in the alloys.

  19. Investigation of the radiation resistance of triple-junction a-Si:H alloy solar cells irradiated with 1.00 MeV protons

    NASA Technical Reports Server (NTRS)

    Lord, Kenneth R., II; Walters, Michael R.; Woodyard, James R.

    1993-01-01

    The effect of 1.00 MeV proton irradiation on hydrogenated amorphous silicon alloy triple-junction solar cells is reported for the first time. The cells were designed for radiation resistance studies and included 0.35 cm(sup 2) active areas on 1.0 by 2.0 cm(sup 2) glass superstrates. Three cells were irradiated through the bottom contact at each of six fluences between 5.10E12 and 1.46E15 cm(sup -2). The effect of the irradiations was determined with light current-voltage measurements. Proton irradiation degraded the cell power densities from 8.0 to 98 percent for the fluences investigated. Annealing irradiated cells at 200 C for two hours restored the power densities to better than 90 percent. The cells exhibited radiation resistances which are superior to cells reported in the literature for fluences less than 1E14 cm(sup -2).

  20. In situ observation of defect annihilation in Kr ion-irradiated bulk Fe/amorphous-Fe 2 Zr nanocomposite alloy

    SciTech Connect

    Yu, K. Y.; Fan, Z.; Chen, Y.; Song, M.; Liu, Y.; Wang, H.; Kirk, M. A.; Li, M.; Zhang, X.

    2014-08-26

    Enhanced irradiation tolerance in crystalline multilayers has received significant attention lately. However, little is known on the irradiation response of crystal/amorphous nanolayers. We report on in situ Kr ion irradiation studies of a bulk Fe96Zr4 nanocomposite alloy. Irradiation resulted in amorphization of Fe2Zr and formed crystal/amorphous nanolayers. α-Fe layers exhibited drastically lower defect density and size than those in large α-Fe grains. In situ video revealed that mobile dislocation loops in α-Fe layers were confined by the crystal/amorphous interfaces and kept migrating to annihilate other defects. This study provides new insights on the design of irradiation-tolerant crystal/amorphous nanocomposites.

  1. Combined nano-SIMS/AFM/EBSD analysis and atom probe tomography, of carbon distribution in austenite/ε-martensite high-Mn steels.

    PubMed

    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.

  2. Investigation of the thermo-mechanical behavior of neutron-irradiated Fe-Cr alloys by self-consistent plasticity theory

    NASA Astrophysics Data System (ADS)

    Xiao, Xiazi; Terentyev, Dmitry; Yu, Long; Bakaev, A.; Jin, Zhaohui; Duan, Huiling

    2016-08-01

    The thermo-mechanical behavior of non-irradiated (at 223 K, 302 K and 573 K) and neutron irradiated (at 573 K) Fe-2.5Cr, Fe-5Cr and Fe-9Cr alloys is studied by a self-consistent plasticity theory, which consists of constitutive equations describing the contribution of radiation defects at grain level, and the elastic-viscoplastic self-consistent method to obtain polycrystalline behaviors. Attention is paid to two types of radiation-induced defects: interstitial dislocation loops and solute rich clusters, which are believed to be the main sources of hardening in Fe-Cr alloys at medium irradiation doses. Both the hardening mechanism and microstructural evolution are investigated by using available experimental data on microstructures, and implementing hardening rules derived from atomistic data. Good agreement with experimental data is achieved for both the yield stress and strain hardening of non-irradiated and irradiated Fe-Cr alloys by treating dislocation loops as strong thermally activated obstacles and solute rich clusters as weak shearable ones.

  3. Structural, mechanical and magnetic properties studies on high-energy Kr-ion irradiated Fe3O4 material (main corrosion layer of Fe-based alloys)

    NASA Astrophysics Data System (ADS)

    Sun, Jianrong; Wang, Zhiguang; Zhang, Hongpeng; Song, Peng; Chang, Hailong; Cui, Minghuan; Pang, Lilong; Zhu, Yabin; Li, Fashen

    2014-12-01

    The Fe-based (T91 and RAFM) alloys are considered as the promising candidate structural materials for DEMO and the first fusion power plant, and these two kinds of steels suffered more serious corrosion attack at 450 °C in liquid PbBi metal. So in order to further clarify the applicability of Fe-based structural materials in nuclear facilities, we should study not only the alloys itself but also its corrosion layers; and in order to simplify the discussion and clarify the irradiation effects of the different corrosion layer, we abstract the Fe3O4 (main corrosion layer of Fe-based alloys) to study the structural, micro-mechanical and magnetic properties under 2.03 GeV Kr-ion irradiation. The initial crystallographic structure of the Fe3O4 remains unaffected after irradiation at low damage levels, but as the Kr-ion fluence increases and the defects accumulate, the macroscopic magnetic properties (Ms, Hc, etc.) and micro-mechanical properties (nano-hardness and Young's modulus) are sensitive to high-energy Kr-ion irradiation and exhibit excruciating uniform changing regularities with varying fluences (firstly increases, then decreases). And these magnetism, hardening and softening phenomena can be interpreted very well by the effects related to the stress and defects (the production, accumulation and free) induced by high-energy ions irradiation.

  4. Austenite decomposition in ternary manganese, molybdenum and tungsten steels

    NASA Astrophysics Data System (ADS)

    Hackenberg, Robert Errol

    A survey of austenite decomposition in Fe-(0.1, 0.2)C-(3, 4.2)Mn has revealed kinetic and morphological transitions which take place at substantial undercoolings below the paraequilibrium Ae3 temperature. An unusually long interval of transformation stasis was found in Fe-0.1C-3Mn, during which time the ferrite was free of carbides. A nodular product containing rod particles was observed in several of these alloys. The grain boundary bainite (GBB) and twin boundary bainite (TBB) morphologies at the bay in Fe-0.24C-4Mo were significantly more complex than previously assumed, with differing arrangements of bainite subunits; their thickening rates also differed. TEM revealed 10 nm steps at the bainite-austenite interfaces in GBB. Mo enrichment was found within GBB-austenite interfaces and extended ˜10 nm into the austenite. The M2C carbides are always enriched in Mo, possessing a non-equilibrium Mo content at earlier reaction times. The energies stored in the ferrite-carbide interfacial area and in carbides possessing non-equilibrium Fe/Mo ratios were considered to reduce the driving force for diffusion by up to 20%. GBB and TBB were found at and above the bay in Fe-0.3C-6.3W, while the bainite formed below the bay consisted of elongated subunits. M6C was found at all temperatures, while M2C was found only below the bay, both of which exhibited W partition. A dark-etching constituent of very high carbide density transformed the remaining pools of austenite at the late stages of reaction, a result consistent with the level of carbon in austenite rising with time. Transitions in carbide morphology were explored in Fe-0.2C-63W. At lower reaction temperatures, M6C precipitates with ferrite. At higher temperatures the cellular precipitation of quasilamellar M 6C in austenite occurs, and is considered to take place inside the ferrite + austenite + M6C three-phase field. The austenite inside the quasilamellar carbide nodules reverts to ferrite at long times, indicating a

  5. Synthesis of carbon-supported PtRh random alloy nanoparticles using electron beam irradiation reduction method

    NASA Astrophysics Data System (ADS)

    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.

  6. Development of Alumina-Forming Austenitic Stainless Steels

    SciTech Connect

    Brady, Michael P; Yamamoto, Yukinori; Bei, Hongbin; Santella, Michael L; Maziasz, Philip J

    2009-01-01

    This paper presents the results of the continued development of creep-resistant, alumina-forming austenitic (AFA) stainless steel alloys, which exhibit a unique combination of excellent oxidation resistance via protective alumina (Al2O3) scale formation and high-temperature creep strength through the formation of stable nano-scale MC carbides and intermetallic precipitates. Efforts in fiscal year 2009 focused on the characterization and understanding of long-term oxidation resistance and tensile properties as a function of alloy composition and microstructure. Computational thermodynamic calculations of the austenitic matrix phase composition and the volume fraction of MC, B2-NiAl, and Fe2(Mo,Nb) base Laves phase precipitates were used to interpret oxidation behavior. Of particular interest was the enrichment of Cr in the austenitic matrix phase by additions of Nb, which aided the establishment and maintenance of alumina. Higher levels of Nb additions also increased the volume fraction of B2-NiAl precipitates, which served as an Al reservoir during long-term oxidation. Ageing studies of AFA alloys were conducted at 750 C for times up to 2000 h. Ageing resulted in near doubling of yield strength at room temperature after only 50 h at 750 C, with little further increase in yield strength out to 2000 h of ageing. Elongation was reduced on ageing; however, levels of 15-25% were retained at room temperature after 2000 h of total ageing.

  7. The isothermal decomposition of austenite in hot-rolled microalloyed steels

    NASA Astrophysics Data System (ADS)

    Crooks, M. J.; Chilton, J. M.

    1984-06-01

    The isothermal decomposition of austenite has been examined in a set of 0.1 C, 1.4 Mn steels containing small amounts of Ti, V, or Nb. The volume fraction of ferrite was measured as a function of transformation temperature and holding time, after hot rolling. Precipitation of carbonitrides, in both the austenite and the ferrite, was examined by electron microscopy of extraction replicas. The decomposition is slowest in the Nb-alloyed steel, in which the start of transformation is delayed and ferrite growth rates are much lower than in the other steels. In the V-alloyed steels, ferrite growth rates are lower than in the plain carbon or Ti alloyed steels. These results are discussed in terms of the effects of carbonitride precipitation in the austenite during high temperature deformation and in the ferrite during transformation. The roles of V and Nb in solution are also considered.

  8. Long term corrosion resistance of alumina forming austenitic stainless steels in liquid lead

    NASA Astrophysics Data System (ADS)

    Ejenstam, Jesper; Szakálos, Peter

    2015-06-01

    Alumina forming austenitic steels (AFA) and commercial stainless steels have been exposed in liquid lead with 10-7 wt.% oxygen at 550 °C for up to one year. It is known that chromia forming austenitic stainless steels, such as 316L and 15-15 Ti, have difficulties forming protective oxides in liquid lead at temperatures above 500 °C, which is confirmed in this study. By adding Al to austenitic steels, it is in general terms possible to increase the corrosion resistance. However this study shows that the high Ni containing AFA alloys are attacked by the liquid lead, i.e. dissolution attack occurs. By lowering the Ni content in AFA alloys, it is possible to achieve excellent oxidation properties in liquid lead. Following further optimization of the microstructural properties, low Ni AFA alloys may represent a promising future structural steel for lead cooled reactors.

  9. Effects of proton irradiation on nanocluster precipitation in ferritic steel containing fcc alloying additions

    SciTech Connect

    Zhang, Zhongwu; Liu, C T; Wang, Xun-Li; Miller, Michael K; Ma, Dong; Chen, Guang; Williams, J R; Chin, Bryan

    2012-01-01

    Newly-developed precipitate-strengthened ferritic steels with and without pre-existing nanoscale precipitates were irradiated with 4 MeV protons to a dose of ~5 mdpa at 50 C and subsequently examined by nanoindentation and atom probe tomography (APT). Irradiation-enhanced precipitation and coarsening of pre-existing nanoscale precipitates were observed. Copper partitions to the precipitate core along with a segregation of Ni, Al and Mn to the precipitate/matrix interface after both thermal aging and proton irradiation. Proton irradiation induces the precipitation reaction and coarsening of pre-existing nanoscale precipitates, and these results are similar to a thermal aging process. The precipitation and coarsening of nanoscale precipitates are responsible for the changes in hardness. The observation of the radiation-induced softening is essentially due to the coarsening of the pre-existing Cu-rich nanoscale precipitates. The implication of the precipitation on the embrittlement of reactor-pressure-vessel steels after irradiation is discussed.

  10. Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

    SciTech Connect

    Busby, Jeremy T; Gussev, Maxim N

    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

  11. Enhancement of Curie Temperature (T c) and Magnetization of Fe-Ni Invar alloy Through Cu Substitution and with He+2 Ion Irradiation

    NASA Astrophysics Data System (ADS)

    Khan, Sajjad Ahmad; Ziya, Amer Bashir; Ibrahim, Ather; Atiq, Shabbar; Usman, Muhammad; Ahmad, Naseeb; Shakeel, Muhammad

    2016-04-01

    The magnetic properties of ternary Fe-Ni-Cu invar alloys are affected by ion irradiation, which goes on increasing with increasing ion fluence (Φ), and by increasing Cu content. In the present study, the ions used are He+2 with 2 MeV energy and with 1 × 1013 cm-2, 1 × 1014 cm-2, 5 × 1014 cm-2, 1 × 1015 cm-2 and 5 × 1015 cm-2 fluence (dose) for irradiation purpose. The face centered cubic structure of the alloy was investigated after ion irradiation using x-ray diffraction (XRD) and found unchanged. However, the peaks become broader with increasing ion dose. Additionally, the lattice fluctuations were observed in XRD study. Curie temperature (T c) is also increased after irradiation. Many factors are considered here for the reason for increasing T c, such as the stopping of incident ions, atomic mixing effect at micro scale level owing to ion irradiation, which might change local concentration and ordering already reported in diffuse scattering, and as a result the Fe-Fe interatomic distance and the Fe-Fe coupling are changed. A comparative study shows that the effect of irradiation on T c and magnetization with increasing ion fluence is more distinctive than the addition of Cu.

  12. Positron Annihilation Spectroscopy and Small Angle Neutron Scattering Characterization of the Effect of Mn on the Nanostructural Features formed in Irradiated Fe-Cu-Mn Alloys

    SciTech Connect

    Glade, S C; Wirth, B D; Asoka-Kumar, P; Odette, G R; Sterne, P A; Howell, R H

    2003-02-27

    The size, number density and composition of the nanometer defects responsible for the hardening and embrittlement in irradiated Fe-0.9wt.% Cu and Fe-0.9wt.% Cu-1.0wt% Mn model reactor pressure vessel alloys were measured using small angle neutron scattering and positron annihilation spectroscopy. These alloys were irradiated at 290 C to relatively low neutron fluences (E > 1 MeV, 6.0 x 10{sup 20} to 4.0 x 10{sup 21} n/m{sup 2}) in order to study the effect of manganese on the nucleation and growth of copper rich precipitates and secondary defect features. Copper rich precipitates were present in both alloys following irradiation. The Fe-Cu-Mn alloy had smaller precipitates and a larger number density of precipitates, suggesting Mn segregation at the iron matrix-precipitate interface which reduces the interfacial energy and in turn the driving force for coarsening. Mn also retards the precipitation kinetics and inhibits large vacancy cluster formation, suggesting a strong Mn-vacancy interaction which reduces radiation enhanced diffusion.

  13. Microstructural studies of advanced austenitic steels

    SciTech Connect

    Todd, J. A.; Ren, Jyh-Ching

    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.

  14. Nucleation of Cr precipitates in Fe-Cr alloy under irradiation

    SciTech Connect

    Dai, Y. Y.; Ao, L.; Sun, Qing- Qiang; Yang, L.; Nie, JL; Peng, SM; Long, XG; Zhou, X. S.; Zu, Xiaotao; Liu, L.; Sun, Xin; Terentyev, Dimtry; Gao, Fei

    2015-04-01

    The nucleation of Cr precipitates induced by overlapping of displacement cascades in Fe-Cr alloys has been investigated using the combination of molecular dynamics (MD) and Metropolis Monte Carlo (MMC) simulations. The results reveal that the number of Frenkel pairs increases with the increasing of overlapped cascades. Overlapping cascades could promote the formation of Cr precipitates in Fe-Cr alloys, as analyzed using short range order (SRO) parameters to quantify the degree of ordering and clustering of Cr atoms. In addition, the simulations using MMC approach show that the presence of small Cr clusters and vacancy clusters formed within cascade overlapped region enhance the nucleation of Cr precipitates, leading to the formation of large Cr dilute precipitates.

  15. Development of Computational Tools for Predicting Thermal- and Radiation-Induced Solute Segregation at Grain Boundaries in Fe-based Alloys

    SciTech Connect

    Yang, Ying

    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.

  16. Vanadium alloy irradiation experiment X530 in EBR-II{sup *}

    SciTech Connect

    Tsai, H.; Strain, R.V.; Hins, A.G.

    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.

  17. Subtask 12H1: Vanadium alloy irradiation experiment X530 in EBR-II

    SciTech Connect

    Tsai, H.; Strain, R.V.; Hins, A.G.; Chung, H.M.; Nowicki, L.J.; Smith, D.L.

    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.

  18. Wear behavior of austenite containing plate steels

    NASA Astrophysics Data System (ADS)

    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

  19. Transmission electron microscopy investigation of the microstructure of Fe-Cr alloys induced by neutron and ion irradiation at 300 °C

    NASA Astrophysics Data System (ADS)

    Hernández-Mayoral, M.; Heintze, C.; Oñorbe, E.

    2016-06-01

    Four Fe-Cr binary alloys, with Cr content from 2.5 up to 12wt%, were neutron or ion irradiated up to a dose of 0.6 dpa at 300 °C. The microstructural response to irradiation has been characterised using Transmission Electron Microscopy (TEM). Both, neutrons and ions, gave rise to the formation of dislocation loops. The most striking difference between ion and neutron irradiation is the distribution of these loops in the sample. Except for the lowest Cr content, loops are distributed mainly along grain boundaries and dislocations in the neutron irradiated samples. The inhomogeneous distribution of dislocation loops could be related to the presence of α‧ precipitates in the matrix. In contrast, a homogeneous distribution is observed in all ion irradiated samples. This important difference is attributed to the orders of magnitude difference in dose rate between these two irradiation conditions. Moreover, the density of loops depends non-monotonically on Cr content in case of neutron irradiation, while it seems to increase with Cr content for ion implantation. Differences are also observed in terms of cluster size, with larger sizes for neutron irradiation than for ion implantation, again pointing towards an effect of the dose rate.

  20. Effect of the Content of Retained Austenite and Grain Size on the Fatigue Bending Strength of Steels Carburized in a Low-Pressure Atmosphere

    NASA Astrophysics Data System (ADS)

    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.

  1. Characterization of the sodium corrosion behavior of commercial austenitic steels

    SciTech Connect

    Shiels, S.A.; Bagnall, C.; Keeton, A.R.; Witkowski, R.E.; Anantatmula, R.P.

    1980-01-01

    During the course of an on-going evaluation of austenitic alloys for potential liquid metal fast breeder reactor (LMFBR) fuel pin cladding application, a series of commercial alloys was selected for study. The data obtained led to the recognition of an underlying pattern of behavior and enabled the prediction of surface chemistry changes. The changes in surface topographical development from alloy to alloy are shown and the important role played by the element molybdenum in this development is indicated. The presentation also illustrates how a total damage equation was evolved to encompass all aspects of weight loss and metal/sodium interactions: wall thinning ferrite layer formation and intergranular attack. The total damage equation represents a significant departure from the classical description of sodium corrosion in which weight loss is simply translated into wall thinning.

  2. Intermetallic strengthened alumina-forming austenitic steels for energy applications

    NASA Astrophysics Data System (ADS)

    Hu, Bin

    In order to achieve energy conversion efficiencies of >50 % for steam turbines/boilers in power generation systems, materials required are strong, corrosion-resistant at high temperatures (>700°C), and economically viable. Austenitic steels strengthened with Laves phase and Ni3Al 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-Fe 2Nb + L12-Ni3Al precipitates) without harmful effects on oxidation resistance. This research starts with microstructural and microchemical analyses of these intermetallic strengthened alumina-forming austenitic steels in a scanning electron microscope. The microchemistry of precipitates, as determined by energy-dispersive x-ray spectroscopy and transmission electron microscope, is also studied. Different thermo-mechanical treatments were carried out to these stainless steels in an attempt to further improve their mechanical properties. The microstructural and microchemical analyses were again performed after the thermo-mechanical processing. Synchrotron X-ray diffraction was used to measure the lattice parameters of these steels after different thermo-mechanical treatments. Tensile tests at both room and elevated temperatures were performed to study mechanical behaviors of this novel alloy system; the deformation mechanisms were studied by strain rate jump tests at elevated temperatures. Failure analysis and post-mortem TEM analysis were performed to study the creep failure mechanisms of these alumina-forming austenitic steels after creep tests. Experiments were carried out to study the effects of boron and carbon additions in the aged alumina-forming austenitic steels.

  3. Austenite Grain Growth and Precipitate Evolution in a Carburizing Steel with Combined Niobium and Molybdenum Additions

    NASA Astrophysics Data System (ADS)

    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.

  4. Solidification and solid state transformations of austenitic stainless steel welds

    SciTech Connect

    Brooks, J A; Williams, J C; Thompson, A W

    1982-05-01

    The microstructure of austenitic stainless steel welds can contain a large variety of ferrite morphologies. It was originally thought that many of these morphologies were direct products of solidification. Subsequently, detailed work on castings suggested the structures can solidify either as ferrite or austenite. However, when solidification occurs by ferrite, a large fraction of the ferrite transforms to austenite during cooling via a diffusion controlled transformation. It was also shown by Arata et al that welds in a 304L alloy solidified 70-80% as primary ferrite, a large fraction of which also transformed to austenite upon cooling. More recently it was suggested that the cooling rates in welds were sufficiently high that diffusionless transformations were responsible for several commonly observed ferrite morphologies. However, other workers have suggested that even in welds, delta ..-->.. ..gamma.. transformations are diffusion controlled. A variety of ferrite morphologies have more recently been characterized by Moisio and coworkers and by David. The purpose of this paper is to provide further understanding of the evaluation of the various weld microstructures which are related to both the solidification behavior and the subsequent solid state transformations. To accomplish this, both TEM and STEM (Scanning Transmission Electron Microscopy) techniques were employed.

  5. Phase stability in thermally-aged CASS CF8 under heavy ion irradiation

    SciTech Connect

    Li, Meimei; Miller, Michael K.; 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

  6. Austenite grain growth simulation considering the solute-drag effect and pinning effect

    PubMed Central

    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

  7. A framework for predicting the yield stress, Charpy toughness and one hundred-year activation level for irradiated fusion power plant alloys

    NASA Astrophysics Data System (ADS)

    Windsor, Colin; Cottrell, Geoff; Kemp, Richard

    2011-04-01

    Recent papers have demonstrated that the yield stress and the Charpy ductile to brittle transition temperature shift at the high irradiation levels of a fusion power plant may be predicted from measurements at lower irradiation levels using neural networks. It was demonstrated that the extrapolation inherent in such predictions could be validated provided that network complexity was appropriately low. Simultaneous predictions of these metallurgical properties at the 100 dpa irradiation level and 400 °C irradiation temperature of a possible fusion power plant have been made for a series of ferritic/martensitic steels, albeit based on mainly fission data. Together with the readily available one hundred-year activation level, benefit functions are defined which can be used to predict the most suitable alloys for a fusion power plant from within existing databases. Our model is sufficiently flexible to allow a variety of possible benefit functions to be defined. The F82H, Eurofer and LA12 alloy series all receive a favourable rating, although all results presented here must be tempered with caution until more data at relevant irradiation levels and with relevant energy spectra become available.

  8. Local electronic effects and irradiation resistance in high-entropy alloys

    DOE PAGES

    Egami, Takeshi; Stocks, George Malcolm; Nicholson, Don; ...

    2015-01-01

    High-entropy alloys are multicomponent solid solutions in which various elements with different chemistries and sizes occupy the same crystallographic lattice sites. Thus, none of the atoms perfectly fit the lattice site, giving rise to considerable local lattice distortions and atomic-level stresses. These characteristics can be beneficial for performance under both radiation and in a high-temperature environment, making them attractive candidates as nuclear materials. We discuss electronic origin of the atomic-level stresses based upon first-principles calculations using a density functional theory approach.

  9. Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

    SciTech Connect

    Jones, R.H.; Gelles, D.S.

    1993-08-01

    Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry.

  10. Roles of Vacancy/Interstitial Diffusion and Segregation in the Microchemistry at Grain Boundaries of Irradiated Fe-Cr-Ni alloys

    SciTech Connect

    Yang, Ying; Field, Kevin G.; Allen, Todd R.; Busby, Jeremy T.

    2016-02-23

    A detailed analysis of the diffusion fluxes near and at grain boundaries of irradiated Fe–Cr–Ni alloys, induced by preferential atom-vacancy and atom-interstitial coupling, is presented. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. The preferential atom-vacancy 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. The calculated fluxes up to 10 dpa 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 modified by the segregation induced by irradiation, leading to the oscillatory behavior of alloy compositions in this region.

  11. Roles of vacancy/interstitial diffusion and segregation in the microchemistry at grain boundaries of irradiated Fe-Cr-Ni alloys

    NASA Astrophysics Data System (ADS)

    Yang, Ying; Field, Kevin G.; Allen, Todd R.; Busby, Jeremy T.

    2016-05-01

    This work presents a detailed analysis of the diffusion fluxes near and at grain boundaries of irradiated Fe-Cr-Ni alloys, induced by preferential atom-vacancy and atom-interstitial coupling. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. The preferential atom-vacancy 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. The calculated fluxes up to 10 dpa 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 modified by the segregation induced by irradiation, leading to the oscillatory behavior of alloy compositions in this region.

  12. In situ high-energy X-ray diffraction study of tensile deformation of neutron-irradiated polycrystalline Fe-9%Cr alloy

    SciTech Connect

    Zhang, Xuan; Li, Meimei; Park, Jun -Sang; Kenesei, Peter; Almer, Jonathan; Xu, Chi; Stubbins, James F.

    2016-12-30

    The effect of neutron irradiation on tensile deformation of a Fe-9wt.%Cr alloy was investigated using in situ high-energy synchrotron X-ray diffraction during room-temperature uniaxial tensile tests. New insights into the deformation mechanisms were obtained through the measurements of lattice strain evolution and the analysis of diffraction peak broadening using the modified Williamson-Hall method. Two neutron-irradiated specimens, one irradiated at 300 °C to 0.01 dpa and the other at 450 °C to 0.01dpa, were tested along with an unirradiated specimen. The macroscopic stress–strain curves of the irradiated specimens showed increased strength, reduced ductility and work-hardening exponent compared to the unirradiated specimen. The evolutions of the lattice strain, the dislocation density and the coherent scattering domain size in the deformation process revealed different roles of the submicroscopic defects in the 300°C/0.01 dpa specimen and the TEM-visible nanometer-sized dislocation loops in the 450°C/0.01 dpa specimen: submicroscopic defects extended the linear work hardening stage (stage II) to a higher strain, while irradiation-induced dislocation loops were more effective in dislocation pinning. Lastly, while the work hardening rate of stage II was unaffected by irradiation, significant dynamic recovery in stage III in the irradiated specimens led to the early onset of necking without stage IV as observed in the unirradiated specimen.

  13. Shape memory alloy thaw sensors

    DOEpatents

    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.

  14. Changes in cluster magnetism and suppression of local superconductivity in amorphous FeCrB alloy irradiated by Ar+ ions

    NASA Astrophysics Data System (ADS)

    Okunev, V. D.; Samoilenko, Z. A.; Szymczak, H.; Szewczyk, A.; Szymczak, R.; Lewandowski, S. J.; Aleshkevych, P.; Malinowski, A.; Gierłowski, P.; Więckowski, J.; Wolny-Marszałek, M.; Jeżabek, M.; Varyukhin, V. N.; Antoshina, I. A.

    2016-02-01

    We show that сluster magnetism in ferromagnetic amorphous Fe67Cr18B15 alloy is related to the presence of large, D=150-250 Å, α-(Fe Cr) clusters responsible for basic changes in cluster magnetism, small, D=30-100 Å, α-(Fe, Cr) and Fe3B clusters and subcluster atomic α-(Fe, Cr, B) groupings, D=10-20 Å, in disordered intercluster medium. For initial sample and irradiated one (Φ=1.5×1018 ions/cm2) superconductivity exists in the cluster shells of metallic α-(Fe, Cr) phase where ferromagnetism of iron is counterbalanced by antiferromagnetism of chromium. At Φ=3×1018 ions/cm2, the internal stresses intensify and the process of iron and chromium phase separation, favorable for mesoscopic superconductivity, changes for inverse one promoting more homogeneous distribution of iron and chromium in the clusters as well as gigantic (twice as much) increase in density of the samples. As a result, in the cluster shells ferromagnetism is restored leading to the increase in magnetization of the sample and suppression of local superconductivity. For initial samples, the temperature dependence of resistivity ρ(T) T2 is determined by the electron scattering on quantum defects. In strongly inhomogeneous samples, after irradiation by fluence Φ=1.5×1018 ions/cm2, the transition to a dependence ρ(T) T1/2 is caused by the effects of weak localization. In more homogeneous samples, at Φ=3×1018 ions/cm2, a return to the dependence ρ(T) T2 is observed.

  15. The effects of ion irradiation on the micromechanical fracture strength and hardness of a self-passivating tungsten alloy

    NASA Astrophysics Data System (ADS)

    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.

  16. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

    DOEpatents

    Johnson, Jr., A. Burtron; Levy, Ira S.; Trimble, Dennis J.; Lanning, Donald D.; Gerber, Franna S.

    1990-01-01

    An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.

  17. Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

    DOEpatents

    Johnson, A.B. Jr.; Levy, I.S.; Trimble, D.J.; Lanning, D.D.; Gerber, F.S.

    1990-04-10

    An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-based materials is disclosed. Samples of zirconium-based materials having different compositions and/or fabrication methods are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280 to 316 C). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hydriding for the same materials when subject to in-reactor (irradiated) corrosion. 1 figure.

  18. Austenite Stability and Tensile Properties of Warm-Extruded Trip Steels

    DTIC Science & Technology

    1976-05-01

    ductility in war-extruded TRIP steel. The austenite stability could be adjusted, however, by a tempering treatment to remove some carbon from solid ... solution , giving tensile properties equivalent or superior to those obtained by warm rolling. Difficulties in alloy composition control or temperature

  19. Delta ferrite-containing austenitic stainless steel resistant to the formation of undesirable phases upon aging

    DOEpatents

    Leitnaker, J.M.

    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 to 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.

  20. Delta ferrite-containing austenitic stainless steel resistant to the formation of undesirable phases upon aging

    DOEpatents

    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.

  1. Alumina-Forming Austenitics: A New Class of Heat-Resistant Stainless Steels

    SciTech Connect

    Brady, Michael P; Yamamoto, Yukinori; Lu, Zhao Ping; Maziasz, Philip J; Liu, Chain T; Pint, Bruce A; Santella, Michael L

    2008-01-01

    A family of alumina (Al2O3)-forming austenitic (AFA) stainless steels is under development. These alloys offer the potential for significantly higher operating temperature and environmental durability than conventional chromia (Cr2O3)-forming stainless steels, without sacrificing other critical characteristics such as cost, creep resistance, and weldability. An overview of the alloy development approach and details of the oxidation and creep resistance properties achieved to date are presented.

  2. Dynamic recrystallization in friction surfaced austenitic stainless steel coatings

    SciTech Connect

    Puli, Ramesh Janaki Ram, G.D.

    2012-12-15

    Friction surfacing involves complex thermo-mechanical phenomena. In this study, the nature of dynamic recrystallization in friction surfaced austenitic stainless steel AISI 316L coatings was investigated using electron backscattered diffraction and transmission electron microscopy. The results show that the alloy 316L undergoes discontinuous dynamic recrystallization under conditions of moderate Zener-Hollomon parameter during friction surfacing. - Highlights: Black-Right-Pointing-Pointer Dynamic recrystallization in alloy 316L friction surfaced coatings is examined. Black-Right-Pointing-Pointer Friction surfacing leads to discontinuous dynamic recrystallization in alloy 316L. Black-Right-Pointing-Pointer Strain rates in friction surfacing exceed 400 s{sup -1}. Black-Right-Pointing-Pointer Estimated grain size matches well with experimental observations in 316L coatings.

  3. All-proportional solid-solution Rh-Pd-Pt alloy nanoparticles by femtosecond laser irradiation of aqueous solution with surfactant

    NASA Astrophysics Data System (ADS)

    Sarker, Md. Samiul Islam; Nakamura, Takahiro; Sato, Shunichi

    2015-06-01

    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.

  4. Effects of helium and hydrogen on radiation-induced microstructural changes in austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Jin, Hyung-Ha; Ko, Eunsol; Lim, Sangyeop; Kwon, Junhyun

    2015-09-01

    Microstructural changes in austenitic stainless steel by helium, hydrogen, and iron ion irradiation were investigated with transmission electron microscopy. Typical radiation-induced changes, such as the formation of Frank loops in the matrix and radiation-induced segregation (RIS) or depletion at grain boundaries, were observed after ion irradiation. The helium ion irradiation led to the formation of cavities both at grain boundaries and in the matrix, as well as the development of smaller Frank loops. The hydrogen ion irradiation generated stronger RIS behavior at the grain boundaries compared to irradiation with helium and iron ions. The effects of helium and hydrogen on radiation-induced microstructural changes were discussed.

  5. Effect of austenite on mechanical properties in high manganese austenitic stainless steel with two phase of martensite and austenite

    NASA Astrophysics Data System (ADS)

    Kim, Y. H.; Kim, J. H.; Hwang, T. H.; Lee, J. Y.; Kang, C. Y.

    2015-05-01

    The effect of the austenite phase on mechanical properties of austenitic stainless steels was investigated using specimens with different volume fractions of retained and reversed austenite. Stainless steels with dual-phase coexisting martensite and austenite were successfully synthesized by deformation and reverse transformation treatment in the cold-rolled high manganese austenitic stainless steel and the ultrafine reverse austenite with less than 0.5 µm in size was formed by reverse transformation treatment in the temperature range of 500-750 °C for various times. With the increase of deformation degree, the volume fraction of retained austenite decreased, while that of the reversed austenite increased as the annealing time increased. From the results of the mechanical properties, it was obvious that as the volume fraction of retained and reversed austenite increased, hardness and strength rapidly decreased, while elongation increased. With regard to each austenite, reversed austenite indicated higher value of hardness and strength, while elongation suggested a lower value because of strengthening owing to grain refinement.

  6. On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

    PubMed

    Lindgren, Kristina; Stiller, Krystyna; Efsing, Pål; Thuvander, Mattias

    2017-03-21

    Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

  7. A review on nickel-free nitrogen containing austenitic stainless steels for biomedical applications.

    PubMed

    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.

  8. Study of biocompatibility of medical grade high nitrogen nickel-free austenitic stainless steel in vitro.

    PubMed

    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.

  9. Tailoring plasticity of austenitic stainless steels for nuclear applications: Review of mechanisms controlling plasticity of austenitic steels below 400 °C

    NASA Astrophysics Data System (ADS)

    Meric de Bellefon, G.; van Duysen, J. C.

    2016-07-01

    AISI 304 and 316 austenitic stainless steels were invented in the early 1900s and are still trusted by materials and mechanical engineers in numerous sectors because of their good combination of strength, ductility, and corrosion resistance, and thanks to decades of experience and data. This article is part of an effort focusing on tailoring the plasticity of both types of steels to nuclear applications. It provides a synthetic and comprehensive review of the plasticity mechanisms in austenitic steels during tensile tests below 400 °C. In particular, formation of twins, extended stacking faults, and martensite, as well as irradiation effects and grain rotation are discussed in details.

  10. In situ high-energy X-ray diffraction study of tensile deformation of neutron-irradiated polycrystalline Fe-9%Cr alloy

    DOE PAGES

    Zhang, Xuan; Li, Meimei; Park, Jun -Sang; ...

    2016-12-30

    The effect of neutron irradiation on tensile deformation of a Fe-9wt.%Cr alloy was investigated using in situ high-energy synchrotron X-ray diffraction during room-temperature uniaxial tensile tests. New insights into the deformation mechanisms were obtained through the measurements of lattice strain evolution and the analysis of diffraction peak broadening using the modified Williamson-Hall method. Two neutron-irradiated specimens, one irradiated at 300 °C to 0.01 dpa and the other at 450 °C to 0.01dpa, were tested along with an unirradiated specimen. The macroscopic stress–strain curves of the irradiated specimens showed increased strength, reduced ductility and work-hardening exponent compared to the unirradiated specimen.more » The evolutions of the lattice strain, the dislocation density and the coherent scattering domain size in the deformation process revealed different roles of the submicroscopic defects in the 300°C/0.01 dpa specimen and the TEM-visible nanometer-sized dislocation loops in the 450°C/0.01 dpa specimen: submicroscopic defects extended the linear work hardening stage (stage II) to a higher strain, while irradiation-induced dislocation loops were more effective in dislocation pinning. Lastly, while the work hardening rate of stage II was unaffected by irradiation, significant dynamic recovery in stage III in the irradiated specimens led to the early onset of necking without stage IV as observed in the unirradiated specimen.« less

  11. Mechanical properties of type 316 stainless steel materials after irradiation at 515/sup 0/C and 585/sup 0/C

    SciTech Connect

    Blackburn, L.D.; Greenslade, D.L.; Ward, A.L.

    1981-04-01

    Three different heats of type 316 SS base metal plate metals and three different weld metals produced by shielded metal arc, submerged arc, and gas tungsten arc processes with type 316 SS filler metal were used in the Gas-Cooled Fast Reactor (GCFR) Structural Materials Irradiation Experiment. Pre-irradiation strength and ductility properties over the range 24/sup 0/C to 650/sup 0/C were very similar for the three base metal heats and were within the expected range for this alloy. The three welds showed variations in strength and ductility before irradiation, but properties were generally within the range of previous experience for austenitic stainless steel welds. Weld metals showed higher yield strength but lower uniform and total elongations than those of base metal.

  12. In-situ determination of austenite and martensite formation in 13Cr6Ni2Mo supermartensitic stainless steel

    SciTech Connect

    Bojack, A.; Zhao, L.; Morris, P.F.; Sietsma, J.

    2012-09-15

    In-situ analysis of the phase transformations in a 13Cr6Ni2Mo supermartensitic stainless steel (X2CrNiMoV13-5-2) was carried out using a thermo-magnetic technique, dilatometry and high temperature X-ray diffractometry (HT-XRD). A combination of the results obtained by the three applied techniques gives a valuable insight in the phase transformations during the austenitization treatment, including subsequent cooling, of the 13Cr6Ni2Mo supermartensitic stainless steel, where the magnetic technique offers a high accuracy in monitoring the austenite fraction. It was found by dilatometry that the austenite formation during heating takes place in two stages, most likely caused by partitioning of Ni into austenite. The in-situ evolution of the austenite fraction is monitored by high-temperature XRD and dilatometry. The progress of martensite formation during cooling was described with a Koistinen-Marburger relation for the results obtained from the magnetic and dilatometer experiments. Enhanced martensite formation at the sample surface was detected by X-ray diffraction, which is assumed to be due to relaxation of transformation stresses at the sample surface. Due to the high alloy content and high thermodynamic stability of austenite at room temperature, 4 vol.% of austenite was found to be stable at room temperature after the austenitization treatment. - Highlights: Black-Right-Pointing-Pointer We in-situ analyzed phase transformations and fractions of a 13Cr6Ni2Mo SMSS. Black-Right-Pointing-Pointer Higher accuracy of the austenite fraction was obtained from magnetic technique. Black-Right-Pointing-Pointer Austenite formation during heating takes place in two stages. Black-Right-Pointing-Pointer Enhanced martensite formation at the sample surface detected by X-ray diffraction.

  13. Irradiation damage from low-dose high-energy protons on mechanical properties and positron annihilation lifetimes of Fe-9Cr alloy

    NASA Astrophysics Data System (ADS)

    Xu, Q.; Fukumoto, K.; Ishi, Y.; Kuriyama, Y.; Uesugi, T.; Sato, K.; Mori, Y.; Yoshiie, T.

    2016-01-01

    Nuclear reactions in accelerator-driven systems (ADS) result in the generation of helium within the ADS materials. The amount of helium produced in this way is approximately one order of magnitude higher than that generated by nuclear fusion. As helium is well-known to induce degradation in the mechanical properties of metals, its effect on ADS materials is an important factor to assess. The results obtained in this study show that low-dose proton irradiation (11 MeV at 573 K to 9.0 × 10-4 dpa and 150 MeV at room temperature to 2.6 × 10-6 dpa) leads to a decrease in yield stress and ultimate tensile strength in a Fe-9Cr alloy. Moreover, interstitial helium and hydrogen atoms, as well as the annihilation of dislocation jogs, were identified as key factors that determine the observed softening of the alloy.

  14. Specification of CuCrZr Alloy Properties after Various Thermo-Mechanical Treatments and Design Allowables including Neutron Irradiation Effects

    SciTech Connect

    Barabash, Vladimir; Kalinin, G. M.; Fabritsiev, Sergei A.; Zinkle, Steven J

    2012-01-01

    Precipitation hardened CuCrZr alloy is a promising heat sink and functional material for various applica- tions in ITER, for example the first wall, blanket electrical attachment, divertor, and heating systems. Three types of thermo-mechanical treatment were identified as most promising for the various applica- tions in ITER: solution annealing, cold working and ageing; solution annealing and ageing; solution annealing and ageing at non-optimal condition due to specific manufacturing processes for engineer- ing-scale components. The available data for these three types of treatments were assessed and mini- mum tensile properties were determined based on recommendation of Structural Design Criteria for the ITER In-vessel Components. The available data for these heat treatments were analyzed for assess- ment of neutron irradiation effect. Using the definitions of the ITER Structural Design Criteria the design allowable stress intensity values are proposed for CuCrZr alloy after various heat treatments.

  15. Capability assessment for application of clay mixture as barrier material for irradiated zirconium alloy structure elements long-term processing for storage during decommissioning of uranium-graphite nuclear reactors

    NASA Astrophysics Data System (ADS)

    Kotlyarevskiy, S. G.; Pavliuk, A. O.; Zakharova, E. V.; Volkova, A. G.

    2016-06-01

    The radionuclide composition and the activity level of the irradiated zirconium alloy E110, the radionuclide immobilization strength and the retention properties of the mixed clay barrier material with respect to the radionuclides identified in the alloy were investigated to perform the safety assessment of handling structural units of zirconium alloy used for the technological channels in uranium-graphite reactors. The irradiated zirconium alloy waste contained the following activation products: 93mNb and the long-lived 94Nb, 93Zr radionuclides. Radionuclides of 60Co, 137Cs, 90Sr, and actinides were also present in the alloy. In the course of the runs no leaching of niobium and zirconium isotopes from the E110 alloy was detected. Leach rates were observed merely for 60Co and 137Cs present in the deposits formed on the internal surface of technological channels. The radionuclides present were effectively adsorbed by the barrier material. To ensure the localization of radionuclides in case of the radionuclide migration from the irradiated zirconium alloy into the barrier material, the sorption properties were determined of the barrier material used for creating the long-term storage point for the graphite stack from uranium-graphite reactors.

  16. Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

    SciTech Connect

    Chen, Y.; Alexandreanu, B.; Natesan, K.

    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.

  17. Retained Austenite in SAE 52100 Steel Post Magnetic Processing and Heat Treatment

    SciTech Connect

    Pappas, Nathaniel R; Watkins, Thomas R; Cavin, Odis Burl; Jaramillo, Roger A; Ludtka, Gerard Michael

    2007-01-01

    Steel is an iron-carbon alloy that contains up to 2% carbon by weight. Understanding which phases of iron and carbon form as a function of temperature and percent carbon is important in order to process/manufacture steel with desired properties. Austenite is the face center cubic (fcc) phase of iron that exists between 912 and 1394 C. When hot steel is rapidly quenched in a medium (typically oil or water), austenite transforms into martensite. The goal of the study is to determine the effect of applying a magnetic field on the amount of retained austenite present at room temperature after quenching. Samples of SAE 52100 steel were heat treated then subjected to a magnetic field of varying strength and time, while samples of SAE 1045 steel were heat treated then subjected to a magnetic field of varying strength for a fixed time while being tempered. X-ray diffraction was used to collect quantitative data corresponding to the amount of each phase present post processing. The percentage of retained austenite was then calculated using the American Society of Testing and Materials standard for determining the amount of retained austenite for randomly oriented samples and was plotted as a function of magnetic field intensity, magnetic field apply time, and magnetic field wait time after quenching to determine what relationships exist with the amount of retained austenite present. In the SAE 52100 steel samples, stronger field strengths resulted in lower percentages of retained austenite for fixed apply times. The results were inconclusive when applying a fixed magnetic field strength for varying amounts of time. When applying a magnetic field after waiting a specific amount of time after quenching, the analyses indicate that shorter wait times result in less retained austenite. The SAE 1045 results were inconclusive. The samples showed no retained austenite regardless of magnetic field strength, indicating that tempering removed the retained austenite. It is apparent

  18. Unusual response of the binary V-2Si alloy to neutron irradiation in FFTF at 430-600{degrees}C

    SciTech Connect

    Ohnuki, S.; Konoshita, H.; Takahaski, H.; Garner, F.A.

    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.

  19. Processing and characterization of a hipped oxide dispersion strengthened austenitic steel

    NASA Astrophysics Data System (ADS)

    Zhou, Zhangjian; Yang, Shuo; Chen, Wanhua; Liao, Lu; Xu, Yingli

    2012-09-01

    An oxide dispersion strengthened (ODS) austenitic steel with a nominal chemical composition of Fe-18Cr-8Ni-1Mo-0.5Ti-0.35Y2O3 (in wt.%) was prepared by mechanical alloying (MA) combined with hot isostatic pressing (HIP). The morphology of MA powders was observed by SEM. The microstructure of the HIPed ODS austenitic steels and chemical composition of the oxide particles were examined by TEM combined with an energy dispersive spectrometry. The oxide dispersion particles with sizes less than 20 nm were determined to be complex Y-Ti-Si-O oxides. The tensile test showed that the fabricated ODS austenitic steel had very high strength and good ductility. The ultimate tensile strength was around 1000 MPa with a total elongation of 33.5% at room temperature, while at temperature of 700 °C, the ultimate tensile strength still reached around 500 MPa.

  20. Strength of "Light" Ferritic and Austenitic Steels Based on the Fe - Mn - Al - C System

    NASA Astrophysics Data System (ADS)

    Kaputkina, L. M.; Svyazhin, A. G.; Smarygina, I. V.; Kindop, V. E.

    2017-01-01

    The phase composition, the hardness, the mechanical properties at room temperature, and the resistance to hot (950 - 1000°C) and warm (550°C) deformation are studied for cast deformable "light" ferritic and austenitic steels of the Fe - (12 - 25)% Mn - (0 - 15)% Al - (0 - 2)% C system alloyed additionally with about 5% Ni. The high-aluminum high-manganese low-carbon and carbonless ferritic steels at a temperature of about 0.5 T melt have a specific strength close to that of the austenitic steels and may be used as weldable scale-resistant and wear-resistant materials. The high-carbon Fe - (20 - 24)% Mn - (5 - 9)% Al - 5% Ni - 1.5% C austenitic steels may be applied as light high-strength materials operating at cryogenic temperatures after a solution treatment and as scale- and heat-resistant materials in an aged condition.

  1. Role of Localized Deformation in Irradiation-Assisted Stress Corrosion Cracking Initiation

    NASA Astrophysics Data System (ADS)

    West, Elaine A.; McMurtrey, Michael D.; Jiao, Zhijie; Was, Gary S.

    2012-01-01

    Intergranular cracking of irradiated austenitic alloys depended on localized grain boundary stress and deformation in both high-temperature aqueous and argon environments. Tensile specimens were irradiated with protons to doses of 1 to 7 dpa and then strained in high-temperature argon, simulated boiling water reactor normal water chemistry, and supercritical water environments. Quantitative measurements confirmed that the initiation of intergranular cracks was promoted by (1) the formation of coarse dislocation channels, (2) discontinuous slip across grain boundaries, (3) a high inclination of the grain boundary to the tensile axis, and (4) low-deformation propensity of grains as characterized by their Schmid and Taylor factors. The first two correlations, as well as the formation of intergranular cracks at the precise locations of dislocation channel-grain boundary intersections are evidence that localized deformation drives crack initiation. The latter two correlations are evidence that intergranular cracking is promoted at grain boundaries experiencing elevated levels of normal stress.

  2. Nanostructured nickel-free austenitic stainless steel/hydroxyapatite composites.

    PubMed

    Tulinski, Maciej; Jurczyk, Mieczyslaw

    2012-11-01

    In this work Ni-free austenitic stainless steels with nanostructure and their nanocomposites with hydroxyapatite are presented and characterized by means of X-ray diffraction and optical profiling. The samples were synthesized by mechanical alloying, heat treatment and nitriding of elemental microcrystalline powders with addition of hydroxyapatite (HA). In our work we wanted to introduce into stainless steel hydroxyapatite ceramics that have been intensively studied for bone repair and replacement applications. Such applications were chosen because of their high biocompatibility and ability to bond to bone. Since nickel-free austenitic stainless steels seem to have better mechanical properties, corrosion resistance and biocompatibility compared to 316L stainless steels, it is possible that composite made of this steel and HA could improve properties, as well. Mechanical alloying and nitriding are very effective technologies to improve the corrosion resistance of stainless steel. Similar process in case of nanocomposites of stainless steel with hydroxyapatite helps achieve even better mechanical properties and corrosion resistance. Hence nanocrystalline nickel-free stainless steels and nickel-free stainless steel/hydroxyapatite nanocomposites could be promising bionanomaterials for use as a hard tissue replacement implants, e.g., orthopedic implants. In such application, the surface roughness and more specifically the surface topography influences the proliferation of cells (e.g., osteoblasts).

  3. Evaluation of Alumina-Forming Austenitic Foil for Advanced Recuperators

    SciTech Connect

    Pint, Bruce A; Brady, Michael P; Yamamoto, Yukinori; Santella, Michael L; Maziasz, Philip J; Matthews, Wendy

    2011-01-01

    A corrosion- and creep-resistant austenitic stainless steel has been developed for advanced recuperator applications. By optimizing the Al and Cr contents, the alloy is fully austenitic for creep strength while allowing the formation of a chemically stable external alumina scale at temperatures up to 900 C. An alumina scale eliminates long-term problems with the formation of volatile Cr oxy-hydroxides in the presence of water vapor in exhaust gas. As a first step in producing foil for primary surface recuperators, three commercially cast heats have been rolled to 100 m thick foil in the laboratory to evaluate performance in creep and oxidation testing. Results from initial creep testing are presented at 675 C and 750 C, showing excellent creep strength compared with other candidate foil materials. Laboratory exposures in humid air at 650 800 C have shown acceptable oxidation resistance. A similar oxidation behavior was observed for sheet specimens of these alloys exposed in a modified 65 kW microturbine for 2871 h. One composition that showed superior creep and oxidation resistance has been selected for the preparation of a commercial batch of foil. DOI: 10.1115/1.4002827

  4. Electrochemical evaluation of sensitization in austenitic stainless steels using miniaturized specimens*1

    NASA Astrophysics Data System (ADS)

    Inazumi, T.; Bell, G. E. C.; Kiuchi, K.

    1991-03-01

    An electrochemical testing system was developed to evaluate the sensitization of neutron-irradiated austenitic stainless steels using miniaturized disk-type specimens, 3 mm in diameter and 0.25 mm thick. The system consists of a specimen holder in which a miniaturized specimen is mounted as the working electrode, a test cell designed to handle radioactive materials and waste, a computer-controlled potentiostat/galvanostat and a surface preparation equipment. Sensitization of a thermally-aged Ti-modified austenitic stainless steel was successfully detected by the single-loop electrochemical potentiokinetic reactivation (SL-EPR) method.

  5. Roles of Vacancy/Interstitial Diffusion and Segregation in the Microchemistry at Grain Boundaries of Irradiated Fe-Cr-Ni alloys

    DOE PAGES

    Yang, Ying; Field, Kevin G.; Allen, Todd R.; ...

    2016-02-23

    A detailed analysis of the diffusion fluxes near and at grain boundaries of irradiated Fe–Cr–Ni alloys, induced by preferential atom-vacancy and atom-interstitial coupling, is presented. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. The preferential atom-vacancy 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. The calculated fluxes up to 10 dpa suggested the dominant diffusion mechanism for chromium and iron is via vacancy,more » 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 modified by the segregation induced by irradiation, leading to the oscillatory behavior of alloy compositions in this region.« less

  6. Overview of strategies for high-temperature creep and oxidation resistance of alumina-forming austenitic stainless steels

    SciTech Connect

    Yamamoto, Yukinori; Brady, Michael P; Santella, Michael L; Bei, Hongbin; Maziasz, Philip J; Pint, Bruce A

    2011-01-01

    A family of creep-resistant, alumina-forming austenitic (AFA) stainless steel alloys is under development for structural use in fossil energy conversion and combustion system applications. The AFA alloys developed to date exhibit comparable creep-rupture lives to state-of-the-art advanced austenitic alloys, and superior oxidation resistance in the {approx}923 K to 1173 K (650 C to 900 C) temperature range due to the formation of a protective Al{sub 2}O{sub 3} scale rather than the Cr{sub 2}O{sub 3} scales that form on conventional stainless steel alloys. This article overviews the alloy design approaches used to obtain high-temperature creep strength in AFA alloys via considerations of phase equilibrium from thermodynamic calculations as well as microstructure characterization. Strengthening precipitates under evaluation include MC-type carbides or intermetallic phases such as NiAl-B2, Fe{sub 2}(Mo,Nb)-Laves, Ni{sub 3}Al-L1{sub 2}, etc. in the austenitic single-phase matrix. Creep, tensile, and oxidation properties of the AFA alloys are discussed relative to compositional and microstructural factors.

  7. Overview of Strategies for High-Temperature Creep and Oxidation Resistance of Alumina-Forming Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Yamamoto, Y.; Brady, M. P.; Santella, M. L.; Bei, H.; Maziasz, P. J.; Pint, B. A.

    2011-04-01

    A family of creep-resistant, alumina-forming austenitic (AFA) stainless steel alloys is under development for structural use in fossil energy conversion and combustion system applications. The AFA alloys developed to date exhibit comparable creep-rupture lives to state-of-the-art advanced austenitic alloys, and superior oxidation resistance in the ~923 K to 1173 K (650 °C to 900 °C) temperature range due to the formation of a protective Al2O3 scale rather than the Cr2O3 scales that form on conventional stainless steel alloys. This article overviews the alloy design approaches used to obtain high-temperature creep strength in AFA alloys via considerations of phase equilibrium from thermodynamic calculations as well as microstructure characterization. Strengthening precipitates under evaluation include MC-type carbides or intermetallic phases such as NiAl-B2, Fe2(Mo,Nb)-Laves, Ni3Al-L12, etc. in the austenitic single-phase matrix. Creep, tensile, and oxidation properties of the AFA alloys are discussed relative to compositional and microstructural factors.

  8. Tuning of the optical properties of In-rich In{sub x}Ga{sub 1−x}N (x=0.82−0.49) alloys by light-ion irradiation at low energy

    SciTech Connect

    De Luca, Marta; Polimeni, Antonio; Capizzi, Mario; Pettinari, Giorgio; Ciatto, Gianluca; Fonda, Emiliano; Amidani, Lucia; Boscherini, Federico; Knübel, Andreas; Cimalla, Volker; Ambacher, Oliver; Giubertoni, Damiano; Bersani, Massimo

    2013-12-04

    The effects of low-energy irradiation by light ions (H and He) on the properties of In-rich In{sub x}Ga{sub 1−x}N alloys are investigated by optical and structural techniques. H-irradiation gives rise to a remarkable blue-shift of light emission and absorption edge energies. X-ray absorption measurements and first-principle calculations address the microscopic origin of these effects.

  9. Investigation of austenitizing temperature on wear behavior of austempered gray iron (AGI)

    NASA Astrophysics Data System (ADS)

    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.

  10. Examination of Spheroidal Graphite Growth and Austenite Solidification in Ductile Iron

    NASA Astrophysics Data System (ADS)

    Qing, Jingjing; Richards, Von L.; Van Aken, David C.

    2016-12-01

    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.

  11. Deformation Mechanisms in Austenitic TRIP/TWIP Steel as a Function of Temperature

    NASA Astrophysics Data System (ADS)

    Martin, Stefan; Wolf, Steffen; Martin, Ulrich; Krüger, Lutz; Rafaja, David

    2016-01-01

    A high-alloy austenitic CrMnNi steel was deformed at temperatures between 213 K and 473 K (-60 °C and 200 °C) and the resulting microstructures were investigated. At low temperatures, the deformation was mainly accompanied by the direct martensitic transformation of γ-austenite to α'-martensite (fcc → bcc), whereas at ambient temperatures, the transformation via ɛ-martensite (fcc → hcp → bcc) was observed in deformation bands. Deformation twinning of the austenite became the dominant deformation mechanism at 373 K (100 °C), whereas the conventional dislocation glide represented the prevailing deformation mode at 473 K (200 °C). The change of the deformation mechanisms was attributed to the temperature dependence of both the driving force of the martensitic γ → α' transformation and the stacking fault energy of the austenite. The continuous transition between the ɛ-martensite formation and the twinning could be explained by different stacking fault arrangements on every second and on each successive {111} austenite lattice plane, respectively, when the stacking fault energy increased. A continuous transition between the transformation-induced plasticity effect and the twinning-induced plasticity effect was observed with increasing deformation temperature. Whereas the formation of α'-martensite was mainly responsible for increased work hardening, the stacking fault configurations forming ɛ-martensite and twins induced additional elongation during tensile testing.

  12. Nanoscale patterning of chemical order induced by displacement cascades in irradiated L10 alloys: Scaling analysis of the fluctuations of order

    NASA Astrophysics Data System (ADS)

    Ye, Jia; Bellon, Pascal

    2006-06-01

    Atomistic kinetic Monte Carlo simulations are employed to analyze the dynamical stabilization of nanoscale patterning of L10 chemical order in a model binary alloy subjected to sustained irradiation. The effect of irradiation-induced displacement cascades on the chemical order is modeled by the introduction at a controlled rate of nearly fully disordered spherical zones, which compete with the reordering promoted by the thermally activated migration of vacancies. When the size of the disordered zones is small, the alloy reaches a steady state that is either long-range ordered at low irradiation-induced ballistic jump frequency, Γb , or disordered at high Γb , with a first-order dynamical transition between these two steady states at Γb=Γbc . Furthermore, in the disordered steady state, the intensity of order fluctuations scales with the reduced variable Γb/Γbc , a scaling that is consistent with an effective temperature approach. For larger cascade sizes, however, an additional steady state is stabilized at intermediate ballistic jump frequency, with a microstructure comprised of well-ordered nanoscale domains. In this patterning-of-order steady state, the above rescaling breaks down but we show that, after deconvolution of the structure factor into Gaussian and Lorentzian components, scaling of the Gaussian component is recovered by introducing a new reduced variable, Γb/Γbp , where 1/Γbp is interpreted as the characteristic time for new domains to form in a disordered zone. This new scaling relationship provides a rigorous definition of the regime of patterning of order. This regime corresponds to the steady states stabilized by cascade sizes and ballistic jump frequencies satisfying Γbc≤Γb≤Γbp . A dynamical phase diagram based on this new criterion is constructed and it agrees well with direct visualization of atomic configurations. Extensions to nonstoichiometric compositions are investigated. Consequences for the direct synthesis of functional

  13. Nanostructurization of Fe-Ni Alloy

    NASA Astrophysics Data System (ADS)

    Danilhenko, Vitaliy E.

    2017-03-01

    Data about an effect of cyclic γ-α-γ martensitic transformations on the structure state of reverted austenite Fe-31.7 wt.% Ni-0.06 wt.% C alloy are presented. The effect of multiple direct γ-α and reverse α-γ martensitic transformations on fragmentation of austenitic grains has been investigated by electron microscopy and X-ray diffraction methods. An ultrafine structure has been formed by nanofragmentation inside the initial austenite grains due to the successive misorientation of their crystal lattice. Austenite was nanofragmented as a result of multiple γ-α-γ martensitic transformations. Slow heating of the nanofragmented alloy at a rate below 2 °C/s results in nanograin refinement of the structure by multiplication of the reverted γ-phase orientations. The conditions of structure refinement up to ultrafine and nanocrystalline levels as a result of both shear and diffusion mechanisms of reverse α-γ transformation are determined.

  14. Ion mass dependence of irradiation-induced local creation of ferromagnetism in Fe{sub 60}Al{sub 40} alloys

    SciTech Connect

    Fassbender, J.; Liedke, M. O.; Strache, T.; Moeller, W.; Menendez, E.; Sort, J.; Rao, K. V.; Deevi, S. C.; Nogues, J.

    2008-05-01

    Ion irradiation of Fe{sub 60}Al{sub 40} alloys results in the phase transformation from the paramagnetic, chemically ordered B2 phase to the ferromagnetic, chemically disordered A2 phase. The magnetic phase transformation is related to the number of displacements per atom (dpa) during the irradiation. For heavy ions (Ar{sup +}, Kr{sup +}, and Xe{sup +}), a universal curve is observed with a steep increase in the fraction of the ferromagnetic phase that reaches saturation, i.e., a complete phase transformation, at about 0.5 dpa. This proves the purely ballistic nature of the disordering process. If light ions are used (He{sup +} and Ne{sup +}), a pronounced deviation from the universal curve is observed. This is attributed to bulk vacancy diffusion from the dilute collision cascades, which leads to a partial recovery of the thermodynamically favored B2 phase. Comparing different noble gas ion irradiation experiments allows us to assess the corresponding counteracting contributions. In addition, the potential to create local ferromagnetic areas embedded in a paramagnetic matrix is demonstrated.

  15. Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.

    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.

  16. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    SciTech Connect

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  17. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    DOE PAGES

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; ...

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019more » n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.« less

  18. Method of making high strength, tough alloy steel

    DOEpatents

    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.

  19. Mechanical Properties of Austenitic Stainless Steel Made by Additive Manufacturing

    PubMed Central

    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. PMID:26601037

  20. Mechanical Properties of Austenitic Stainless Steel Made by Additive Manufacturing.

    PubMed

    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.