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Sample records for austenitic alloys irradiated

  1. Magnetic phase formation in irradiated austenitic alloys

    SciTech Connect

    Gussev, Maxim N; Busby, Jeremy T; Tan, Lizhen; Garner, Francis A.

    2014-01-01

    Austenitic alloys are often observed to develop magnetic properties during irradiation, possibly associated with radiation-induced acceleration of the ferrite phase. Some of the parametric sensitivities of this phenomenon have been addressed using a series of alloys irradiated in the BOR-60 reactor at 593K. The rate of development of magnetic phase appears to be sensitive to alloy composition. To the first order, the largest sensitivities to accelerate ferrite formation, as explored in this experiment, are associated with silicon, carbon and manganese and chromium. Si, C, and Mn are thought to influence diffusion rates of point defects while Cr plays a prominent role in defining the chromium equivalent and therefore the amount of ferrite at equilibrium. Pre-irradiation cold working was found to accelerate ferrite formation, but it can play many roles including an effect on diffusion, but on the basis of these results the dominant role or roles of cold-work cannot be identified. Based on the data available, ferrite formation is most probably associated with diffusion.

  2. 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.

  3. Hydrogen isotope transfer in austenitic steels and high-nickel alloy during in-core irradiation

    SciTech Connect

    Polosukhin, B.G.; Sulimov, E.M.; Zyrianov, A.P.; Kalinin, G.M.

    1995-10-01

    The transfer of protium and deuterium in austenitic chromium-nickel steels and in a high-nickel alloy was studied in a specially designed facility. The transfer parameters of protium and deuterium were found to change greatly during in-core irradiation, and the effects of irradiation increased as the temperature decreased. Thus, at temperature T<673K, the relative increase in the permeability of hydrogen isotopes under irradiation can be orders of magnitude higher in these steels. Other radiation effects were also observed, in addition to the changes from the initial values in the effects of protium and deuterium isotopic transfer. 4 refs., 3 figs., 2 tabs.

  4. Microstructure evolution in austenitic Fe-Cr-Ni alloys irradiated with rotons: comparison with neutron-irradiated microstructures

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.

    2001-08-01

    Irradiation-induced microstructures of high purity and commercial purity austenitic stainless steels were investigated using proton-irradiation. For high purity alloys, Fe-20Cr-9Ni (HP 304 SS), Fe-20Cr-24Ni and Ni-18Cr-9Fe were irradiated using 3.2 MeV protons between 300°C and 600°C at a dose rate of 7×10 -6 dpa/ s to doses up to 3.0 dpa. The commercial purity alloys, CP 304 SS and CP 316 SS were irradiated at 360°C to doses between 0.3 and 5.0 dpa. The dose, temperature and composition dependence of the number density and size of dislocation loops and voids were characterized. The changes in yield strength due to irradiation were estimated from Vickers hardness measurements and compared to calculations using a dispersed-barrier-hardening (DBH) model. The dose and temperature dependence of proton-irradiated microstructure (loops, voids) and the irradiation hardening are consistent with the neutron-data trend. Results indicate that proton-irradiation can accurately reproduce the microstructure of austenitic alloys irradiated in LWR cores.

  5. Microstructure evolution in proton-irradiated austenitic Fe-Cr-Ni alloys under LWR core conditions

    NASA Astrophysics Data System (ADS)

    Gan, Jian

    1999-11-01

    Irradiation-induced microstructure of austenitic stainless steel was investigated using proton irradiation. High-purity alloys of Fe-20Cr-9Ni (UHP 304 SS), Fe-20Cr-24Ni and Ni-18Cr-9Fe were irradiated using 3.2 MeV protons at a dose rate of 7 × 10-6 dpa/s between 300°C and 600°C. The irradiation produced a microstructure consisting of dislocation loops and voids. The dose and temperature dependence of the number density and size of dislocation loops and voids were investigated. The changes in yield strength due to irradiation were estimated from Vickers hardness measurements and compared to calculations using a dispersed-barrier hardening model. The dose and temperature dependence of microstructure and hardness change for proton irradiation follows the same trend as that for neutron irradiation at comparable irradiation conditions. Commercial purity alloys of CP 304 SS and CP 316 SS were irradiated at 360°C to doses between 0.3 and 3.0 dpa. The irradiated microstructure consists of dislocation loops. No voids were detected at doses up to 3.0 dpa. Loop size distributions are in close agreement with that in the same alloys neutron-irradiated in a LWR core. The loop density also agrees with neutron irradiation data. The yield strength as a function of dose in proton irradiated commercial purity alloys is consistent with the neutron- data trend. A fast-reactor microstructure model was adapted for light water reactor (LWR) irradiation conditions (275°C, 7 × 10 -8 dpa/s) and then applied to proton irradiation under conditions (360°C, 7 × 10-6 dpa/s) relevant to LWRs. The original model was modified by including in-cascade interstitial clustering and the loss of interstitial clusters to sinks by cluster diffusion. It was demonstrated that loop nucleation for both LWR irradiation condition and proton irradiation are driven by in-cascade interstitial clustering. One important result from this modeling work is that the difference in displacement cascade between

  6. Stress corrosion cracking behavior of irradiated model austenitic stainless steel alloys.

    SciTech Connect

    Chung, H. M.; Karlsen, T. M.; Ruther, W. E.; Shack, W. J.; Strain, R. V.

    1999-07-16

    Slow-strain-rate tensile tests (SSRTs) and posttest fractographic analyses by scanning electron microscopy were conducted on 16 austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} and {approx}0.9 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E >1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. Following irradiation to a fluence of {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2}, a high-purity laboratory heat of Type 316L SS (Si {approx} 0.024 wt%) exhibited the highest susceptibility to IGSCC. The other 15 alloys exhibited negligible susceptibility to IGSCC at this low fluence. The percentage of TGSCC on the fracture surfaces of SSRT specimens of the 16 alloys at {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E > 1 MeV) could be correlated well with N and Si concentrations; all alloys that contained <0.01 wt.% N and <1.0 wt. % Si were susceptible, whereas all alloys that contained >0.01 wt.% N or >1.0 wt.% Si were relatively resistant. High concentrations of Cr were beneficial. Alloys that contain <15.5 wt.% Cr exhibited greater percentages of TGSCC and IGSCC than those alloys with {approx}18 wt.% Cr, whereas an alloy that contains >21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions.

  7. 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.

  8. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2003-12-01

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

  10. 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.

  11. 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.

  12. Observations of void swelling in selected austenitic alloys during ion irradiation under a rising temperature ramp

    NASA Astrophysics Data System (ADS)

    Mazey, D. J.; Williams, T. M.; Bolster, D. E. J.

    1988-07-01

    Results are given of a TEM study of the void swelling behaviour of CW AISI 321 (En58B), CW FV548 stainless steel and STA Nimonic PE16 alloy during 46 MeV nickel ion irradiation under a near-linear rising temperature ramp. The dose range for the ramp was 0-75 dpa and four temperature ranges were investigated using 50 ° C intervals from 500-550 ° C, 525-575 ° C, 575-625 ° C and 625-675 ° C. Swelling in CW 321 was found to be no greater than that observed after isothermal irradiation to 75 dpa at temperatures corresponding to the end-of-ramp temperatures. This was not the case for CW FV548 where the swelling for the temperature ranges of 525-575 ° C and 575-625 ° C was slightly greater than that observed under isothermal irradiation to 75 dpa. Swelling in STA PE16 in the lower temperature ranges of 500-550 ° C and 525-575°C was much higher than corresponding isothermal values but in the 600-700 ° C temperature range the swelling was close to isothermal values. The results on PE16 indicate that a steady increase in irradiation temperature from just below or just above the low temperature cut-off into the void swelling region produces a significant increase in swelling with respect to that expected under isothermal conditions. Possible interpretations of the observed swelling behaviour in these alloys are discussed.

  13. 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.

  14. 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.

  15. 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.

  16. Effects of pulsed dual-ion irradiation on phase transformations and microstructure in Ti-modified austenitic alloy

    SciTech Connect

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

    1983-01-05

    The influence of pulsed 4 MeV Ni ion bombardment, with and without simultaneous helium injection, has been explored in a low swelling, Ti-modified austenitic stainless steel. Irradiations were carried out to 70 dpa at 950/sup 0/K; the pulsing frequencies were either 60 s on/off or 1 s on/off. Compared to continuous irradiation, pulsing caused a decrease in the interstitial loop diameter at 1 dpa, although at higher doses the overall dislocation density was not affected. Pulsing and helium both promoted the stability of MC precipitates and retarded the subsequent G phase formation; in some cases G-phase was suppressed and eta phase formed instead. Small bubble-like cavities were observed to grow into large voids after steady dual beam irradiation to 70 dpa. However, this conversion was suppressed by pulse irradiation to 70 dpa and furthermore the sizes of the small cavities were somewhat reduced. The results are explained in terms of current mechanistic understanding of mean point defect kinetics and the evolution of microstructure and microcomposition during irradiation with superimposed annealing periods.

  17. Development and evaluation of advanced austenitic alloys

    SciTech Connect

    Swindeman, R.W.; Maziasz, P.J.; King, J.F.; Bolling, E.

    1990-01-01

    Research was performed on advanced austenitic alloys for tubing in heat recovery systems. Evaluations addressed the need to optimize strength, fabricability, and surface protection for specific environments and temperatures. Alloys studied included advanced lean austenitic stainless steels and higher chromium alloys to 760{degree}C, nickel-chromium-iron aluminides at temperature to 760{degree}C, and Ni--Cr alloys with capability for service to 1000{degree}C. Coordinated research was performed at a number of universities and industrial research facilities. Evaluation of the lean stainless steels focused on MC-forming alloys and a family of modified 316 stainless steels. Work nearing completion revealed that many of the alloy design criteria for the lean stainless steels could be met. With the judicious selection of thermal-mechanical processing, data indicated that high strength and ductility could be achieved in both base metal and weldments. Fabrication requirements needed to produce optimum performance called for high solution treating temperatures and small levels of cold or warm work. Evaluations of high chromium stainless steels and modifications of alloy 800H were encouraging, and good properties were observed for temperatures to 760{degree}C. Work on the alloys and claddings for service to 1000{degree}C was begun on two commercial alloys of nearest in PBFC hot gas cleanup systems. 20 refs., 3 figs., 2 tabs.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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)

  5. Weldability of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Asano, Kyoichi; Nishimura, Seiji; Saito, Yoshiaki; Sakamoto, Hiroshi; Yamada, Yuji; Kato, Takahiko; Hashimoto, Tsuneyuki

    1999-01-01

    Degradation of weldability in neutron irradiated austenitic stainless steel is an important issue to be addressed in the planning of proactive maintenance of light water reactor core internals. In this work, samples selected from reactor internal components which had been irradiated to fluence from 8.5 × 10 22 to 1.4 × 10 26 n/m 2 ( E > 1 MeV) corresponding to helium content from 0.11 to 103 appm, respectively, were subjected to tungsten inert gas arc (TIG) welding with heat input ranged 0.6-16 kJ/cm. The weld defects were characterized by penetrant test and cross-sectional metallography. The integrity of the weld was better when there were less helium and at lower heat input. Tensile properties of weld joint containing 0.6 appm of helium fulfilled the requirement for unirradiated base metal. Repeated thermal cycles were found to be very hazardous. The results showed the combination of material helium content and weld heat input where materials can be welded with little concern to invite cracking. Also, the importance of using properly selected welding procedures to minimize thermal cycling was recognized.

  6. Modeling of microstructure evolution in austenitic stainless steels irradiated under light water reactor condition

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.; Stoller, R. E.

    2001-10-01

    A model for microstructure development in austenitic alloys under light water reactor irradiation conditions is described. The model is derived from the model developed by Stoller and Odette to describe microstructural evolution under fast neutron or fusion reactor irradiation conditions. The model is benchmarked against microstructure measurements in 304 and 316 SS irradiated in a boiling water reactor core using one material-dependent and three irradiation-based parameters. The model is also adapted for proton irradiation at higher dose rate and higher temperature and is calibrated against microstructure measurements for proton irradiation. The model calculations show that for both neutron and proton irradiations, in-cascade interstitial clustering is the driving mechanism for loop nucleation. The loss of interstitial clusters to sinks by interstitial cluster diffusion was found to be an important factor in determining the loop density. The model also explains how proton irradiation can produce an irradiated dislocation microstructure similar to that in neutron irradiation.

  7. 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

  8. 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.

  9. 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.

  10. Deformation and thermal fatigue in high temperature austenitic alloys

    SciTech Connect

    Ferro, P.D.; Yost, B.; Swindeman, R.W.; Li, Che-Yu . Dept. of Materials Science and Engineering)

    1991-03-01

    The flow properties of modified austenitic alloys are reviewed. The important strengthening mechanisms discussed include precipitation hardening produced by a combination of cold work and aging and by creep aging. Grain boundary sliding enhanced by reduced grain size is shown to reduce the flow strength of these alloys. 5 refs., 11 figs., 2 tabs.

  11. 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.

  12. Effect of neutron irradiation on vanadium alloys

    SciTech Connect

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. Creep Behavior of a New Cast Austenitic Alloy

    SciTech Connect

    Shingledecker, John P; Maziasz, Philip J; Evans, Neal D; Pollard, Michael J

    2007-01-01

    A new cast austenitic alloy, CF8C-Plus, has been developed by Oak Ridge National Laboratory (ORNL) and Caterpillar for a wide range of high temperature applications including diesel exhaust components and turbine casings. The creep strength of the CF8C-Plus steel is much greater than that of the standard cast CF8C stainless steel and is comparable to the highest strength wrought commercial austenitic stainless steels and alloys, such as NF709. The creep properties of CF8C-Plus are discussed in terms of the alloy design methodology and the evaluation of some long-term creep tested specimens (over 20,000 hours). Microcharacterization shows that the excellent creep strength is due mainly to the precipitation of very fine nano-scale and stable MC carbides, without the formation of deleterious intermetallic phases.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Irradiation-assisted stress corrosion cracking of model austenitic stainless steel.

    SciTech Connect

    Chung, H. M.; Ruther, W. E.; Strain, R. V.; Shack, W. J.; Karlsen, T. M.

    1999-10-26

    Slow-strain-rate tensile (SSRT) tests were conducted on model austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup 2} and {approx} 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. At {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2}, a high-purity heat of Type 316L SS that contains a very low concentration of Si exhibited the highest susceptibility to IGSCC. In unirradiated state, Types 304 and 304L SS did not exhibit a systematic effect of Si content on alloy strength. However, at {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2}, yield and maximum strengths decreased significantly as Si content was increased to >0.9 wt.%. Among alloys that contain low concentrations of C and N, ductility and resistance to TGSCC and IGSCC were significantly greater for alloys with >0.9 wt.% Si than for alloys with <0.47 wt.% Si. Initial data at {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} were also consistent with the beneficial effect of high Si content. This indicates that to delay onset of and reduce susceptibility to irradiation-assisted stress corrosion cracking (IASCC), at least at low fluence levels, it is helpful to ensure a certain minimum concentration of Si. High concentrations of Cr were also beneficial; alloys that contain <15.5 wt.% Cr exhibited greater susceptibility to IASCC than alloys with {approx}18 wt.% Cr, whereas an alloy that contains >21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions.

  3. Alloy development for irradiation performance in fusion reactors

    NASA Astrophysics Data System (ADS)

    Harling, O. K.; Grant, N. J.

    1980-12-01

    The development of improved structural alloys for the fusion reactor first wall application is addressed. Several new alloys were produced by rapid solidification. Emphasis in alloy design and production was placed on producing austenitic Type 316SS with fine dispersions of TiC and Al2O3 particles. Results of mechanical and microstructural tests are presented. A number of neutron irradiations were initiated on samples fabricated from alloys produced. 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 was continued with emphasis on understanding the effect of recoil resolution on relatively stable second phase particles. The microstructure of several ZrB2 doped stainless steels was characterized.

  4. 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.

  5. 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

  6. 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.

  7. Dislocation loop evolution under ion irradiation in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Etienne, A.; Hernández-Mayoral, M.; Genevois, C.; Radiguet, B.; Pareige, P.

    2010-05-01

    A solution annealed 304 and a cold worked 316 austenitic stainless steels were irradiated from 0.36 to 5 dpa at 350 °C using 160 keV Fe ions. Irradiated microstructures were characterized by transmission electron microscopy (TEM). Observations after irradiation revealed the presence of a high number density of Frank loops. Size and number density of Frank loops have been measured. Results are in good agreement with those observed in the literature and show that ion irradiation is able to simulate dislocation loop microstructure obtained after neutron irradiation. Experimental results and data from literature were compared with predictions from the cluster dynamic model, MFVIC (Mean Field Vacancy and Interstitial Clustering). It is able to reproduce dislocation loop population for neutron irradiation. Effects of dose rate and temperature on the loop number density are simulated by the model. Calculations for ion irradiations show that simulation results are consistent with experimental observations. However, results also show the model limitations due to the lack of accurate parameters.

  8. Strain hardening and plastic instability properties of austenitic stainless steels after proton and neutron irradiation

    NASA Astrophysics Data System (ADS)

    Byun, T. S.; Farrell, K.; Lee, E. H.; Hunn, J. D.; Mansur, L. K.

    2001-10-01

    Strain hardening and plastic instability properties were analyzed for EC316LN, HTUPS316, and AL6XN austenitic stainless steels after combined 800 MeV proton and spallation neutron irradiation to doses up to 10.7 dpa. The steels retained good strain-hardening rates after irradiation, which resulted in significant uniform strains. It was found that the instability stress, the stress at the onset of necking, had little dependence on the irradiation dose. Tensile fracture stress and strain were calculated from the stress-strain curve data and were used to estimate fracture toughness using an existing model. The doses to plastic instability and fracture, the accumulated doses at which the yield stress reaches instability stress or fracture stress, were predicted by extrapolation of the yield stress, instability stress, and fracture stress to higher dose. The EC316LN alloy required the highest doses for plastic instability and fracture. Plastic deformation mechanisms are discussed in relation to the strain-hardening properties of the austenitic stainless steels.

  9. 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)

  10. The effects of neutron irradiation on fracture toughness of austenitic stainless steels.

    SciTech Connect

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

    1999-05-21

    Austenitic stainless steels are used extensively as structural alloys in reactor pressure vessel internal components because of their superior fracture toughness properties. However, exposure to high levels of neutron irradiation for extended periods leads to significant reduction in the fracture resistance of these steels. This paper presents results of fracture toughness J-R curve tests on four heats of Type 304 stainless steel that were irradiated to fluence levels of {approx}0.3 and 0.9 x 10{sup 21} n cm{sup {minus}2} (E >1 MeV) at {approx}288 C in a helium environment in the Halden heavy water boiling reactor. The tests were performed on 1/4-T compact tension specimens in air at 288 C; crack extensions were determined by both DC potential and elastic unloading compliance techniques.

  11. 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.

  12. Precipitation and cavity formation in austenitic stainless steels during irradiation

    SciTech Connect

    Lee, E.H.; Rowcliffe, A.F.; Mansur, L.K.

    1981-01-01

    Microstructural evolution in austenitic stainless steels subjected to displacement damage at high temperature is strongly influenced by the interactions between helium atoms and second phase particles. Cavity nucleation occurs by the trapping of helium at partially coherent particle-matrix interfaces. The recent precipitate point defect collector theory describes the more rapid growth of precipitate-attached cavities compared to matrix cavities where the precipitate-matrix interface collects point defects to augment the normal point deflect flux to the cavitry. Data are presented which support these ideas. It is shown that during nickel ion irradiation of a titanium-modified stainless steel at 675/sup 0/C the rate of injection of helium has a strong effect on the total swelling and also on the nature and distribution of precipitate phases.

  13. Precipitation sensitivity to alloy composition in Fe-Cr-Mn austenitic steels developed for reduced activation for fusion application

    SciTech Connect

    Maziasz, P.J.; Klueh, R.L.

    1988-01-01

    Special austenitic steels are being designed in which alloying elements like Mo, Nb, and Ni are replaced with Mn, W, V, Ti, and/or Ta to reduce the long-term radioactivity induced by fusion reactor irradiation. However, the new steels still need to have properties otherwise similar to commercial steels like type 316. Precipitation strongly affects strength and radiation-resistance in austenitic steels during irradiation at 400--600/degree/C, and precipitation is also usually quite sensitive to alloy composition. The initial stage of development was to define a base Fe-Cr-Mn-C composition that formed stable austenite after annealing and cold-working, and resisted recovery or excessive formation of coarse carbide and intermetallic phases during elevated temperature annealing. These studies produced a Fe-12Cr-20Mn-0.25C base alloy. The next stage was to add the minor alloying elements W, Ti, V, P, and B for more strength and radiation-resistance. One of the goals was to produce fine MC precipitation behavior similar to the Ti-modified Fe-Cr-Ni prime candidate alloy (PCA). Additions of Ti+V+P+B produced fine MC precipitation along network dislocations and recovery/recrystallization resistance in 20% cold worked material aged at 800/degree/C for 166h, whereas W, Ti, W+Ti, or Ti+P+B additions did not. Addition of W+Ti+V+P+B also produced fine MC, but caused some sigma phase formation and more recrystallization as well. 29 refs., 14 figs., 9 tabs.

  14. 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.

  15. 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.

  16. 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.

  17. Corrosion of austenitic stainless steels and nickel-base alloys in supercritical water and novel control methods

    SciTech Connect

    Tan, Lizhen; Allen, Todd R.; Yang, Ying

    2012-01-01

    This chapter contains sections titled: (1) Introduction; (2) Thermodynamics of Alloy Oxidation; (3) Corrosion of Austenitic Stainless Steels and Ni-Base Alloys in SCW; (4) Novel Corrosion Control Methods; (5) Factors Influencing Corrosion; (6) Summary; and (7) References.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

    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. This difference is attributed 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.

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

    NASA Astrophysics Data System (ADS)

    Danilchenko, Vitalij

    2016-02-01

    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.% C 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.

  4. Irradiation-assisted stress corrosion cracking of austenitic stainless steels: Recent progress and new approaches

    SciTech Connect

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

    1996-09-01

    Irradiation-assisted stress corrosion cracking (IASCC) of several types of BWR field components fabricated from solution-annealed austenitic stainless steels (SSs), including a core internal weld, were investigated by means of slow-strain-rate test (SSRT), scanning electron microscopy (SEM), Auger electron spectroscopy (AES), and field-emission-gun advanced analytical electron microscopy (FEG-AAEM). Based on the results of the tests and analyses, separate effects of neutron fluence, tensile properties, alloying elements and major impurities identified in the American Society for Testing and Materials (ASTM) specifications, minor impurities, water chemistry, and fabrication-related variables were determined. The results indicate strongly that minor impurities not specified by the ASTM-specifications play important roles, probably through a complex synergism with grain-boundary Cr depletion. These impurities, typically associated with steelmaking and component fabrication processes, are very low or negligible in solubility in steels and are the same impurities that have been known to promote intergranular SCC significantly when they are present in water as ions or soluble compounds. It seems obvious that IASCC is a complex integral problem which involves many variables that are influenced strongly by not only irradiation conditions, water chemistry, and stress but also iron and steelmaking processes, fabrication of the component, and joining and welding. Therefore, for high-stress components in particular, it would be difficult to mitigate IASCC problems at high fluence based on the consideration of water chemistry alone, and other considerations based on material composition and fabrication procedure would be necessary as well.

  5. 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.

  6. 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

  7. 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

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

  9. TEM, XRD and nanoindentation characterization of Xenon ion irradiation damage in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Huang, H. F.; Li, J. J.; Li, D. H.; Liu, R. D.; Lei, G. H.; Huang, Q.; Yan, L.

    2014-11-01

    Cross-sectional and bulk specimens of a 20% cold-worked 316 austenitic stainless steel (CW 316 SS) has been characterized by TEM, XRD and nanoindentation to determine the microstructural evolution and mechanical property changes of 316 SS after irradiation with 7 MeV Xe26+ ions. TEM results reveal the presence of dislocation loops with a number density of approximately 3 × 1022 m-3 and sizes between 3 to 10 nm due to the collapse of vacancy rich cores inside displacement cascades. Peak broadening observed in XRD diffraction patters reveal systematic changes to lattice parameters due to irradiation. The calculated indentation values in irradiated 316 SS were found to be much higher in comparison to the unirradiated specimen, indicating the dose dependent effect of irradiation on hardness. The relationship between irradiation induced microstructural evolution and the changes to the mechanical properties of CW 316 SS are discussed in the context of fluence and irradiation temperature.

  10. 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.

  11. Alloy 33, a new corrosion resistant austenitic material for the refinery industry and related applications

    SciTech Connect

    Koehler, M.; Heubner, U.; Eichenhofer, K.W.; Renner, M.

    1995-09-01

    A new corrosion resistant austenitic material alloyed with nominally (wt. %) 33 Cr, 32 Fe, 31 Ni, 1.6 Mo, 0.6 Cu, 0.4 N exhibits excellent resistance to general and local corrosion in hot mineral acids and chloride bearing solutions. Furthermore, the new alloy stands out for its superior corrosion resistance in many other corrosive environments from acidic to alkaline including resistance to stress corrosion cracking. In mixed HNO{sub 3}/HF acids the corrosion resistance of Alloy 33 is superior to high chromium nickel-base alloys. In NAOH solutions the new alloy is applicable to conditions where the known stainless steels fail. Due to its high nitrogen content the new alloy exhibits a small grain size in its solution annealed condition and, consequently, a high yield strength and excellent toughness CP properties. Alloy 33 is easily welded without filler or using matching filler metal. Typical applications of Alloy 33 include heat exchangers, condenser tubes and other equipment for the Refinery Industry and the Chemical Process Industry as well as light weight structures in the Offshore Industry. Especially the multi-purpose character of Alloy 33 with respect to its corrosion resistance as well to acidic and alkaline media as to chloride bearing cooling waters opens a wide variety of applications.

  12. 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.

  13. 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.

  14. Plastic Localization Phenomena in a Mn-Alloyed Austenitic Steel

    NASA Astrophysics Data System (ADS)

    Scavino, G.; D'Aiuto, F.; Matteis, P.; Russo Spena, P.; Firrao, D.

    2010-03-01

    A 0.5 wt pct C, 22 wt pct Mn austenitic steel, recently proposed for fabricating automotive body structures by cold sheet forming, exhibits plastic localizations (PLs) during uniaxial tensile tests, yet showing a favorable overall strength and ductility. No localization happens during biaxial Erichsen cupping tests. Full-thickness tensile and Erichsen specimens, cut from as-produced steel sheets, were polished and tested at different strain rates. During the tensile tests, the PL phenomena consist first of macroscopic deformation bands traveling along the tensile axis, and then of a series of successive stationary deformation bands, each adjacent to the preceding ones; both types of bands involve the full specimen width and yield a macroscopically observable surface relief. No comparable surface relief was observed during the standard Erichsen tests. Because the stress state is known to influence PL phenomena, reduced-width Erichsen tests were performed on polished sheet specimens, in order to explore the transition from biaxial to uniaxial loading; surface relief lines were observed on a 20-mm-wide specimen, but not on wider ones.

  15. Irradiation behavior of weldments of austenitic stainless steel made by various welding techniques

    SciTech Connect

    Shiba, Kiyoyuki; Sawai, Tomotsugu; Jitsukawa, Shiro; Hishinuma, Akimichi; Pawel, J.E.

    1996-12-31

    Austenitic stainless steel is one of the candidate materials for nuclear fusion reactor applications. Here, an austenitic stainless steel, 316F, irradiated in the High Flux Isotope Reactor to doses of about 8 to 33 dpa at 400 and 500 C was investigated. Electron beam (EB) welding and metal inert gas (MIG) welding techniques were used to make weldment specimens. Weldment specimens were made from their weld metal or weld joint (including heat affected zone) regions of the weldments. Base metal was also studied for comparison. Microstructures of these specimens were observed by TEM. Tensile tests were carried out at the nominal irradiation temperature in vacuum. Solution annealed 316F showed the large irradiation hardening at 400 C, while the change in yield stress observed at 500 C was not so large. Weldments specimens had the same temperature and dose dependence as the base metal. The differences between EB and MIG after irradiation were small, compared to the differences before irradiation, except for the slight less ductility of MIG weldments. The defect microstructures of weldments were the same as base metal.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Irradiation creep behavior of V-4Cr-4Ti alloys irradiated in a liquid sodium environment at the JOYO fast reactor

    NASA Astrophysics Data System (ADS)

    Fukumoto, Ken-ichi; Matsui, Hideki; Narui, Minoru; Yamazaki, Masanori

    2013-06-01

    Irradiation experiments on V-4Cr-4Ti alloys with sodium-enclosed irradiation capsules in the JOYO fast reactor were conducted using pressurized creep tubes (PCTs). The irradiation creep strain was significantly larger than the thermal creep strain below 686 °C, but there was no swelling of the neutron-irradiated V-4Cr-4Ti alloys. At temperatures below 500 °C, the irradiation creep was found to be proportional to the square root of the neutron dose and linear with the stress level. Above 500 °C, it was expected to be proportional to the stress level to a power greater than unity, because the irradiation creep mechanism could change from the stress-induced preferred absorption mechanism (SIPA) to the preferred absorption glide mechanism (PGA). By comparing annealed PCT specimens with cold-worked specimens, the cold-worked V-4Cr-4Ti alloys exhibited a larger irradiation creep strain compared with the annealed alloys. The irradiation creep compliance of the V-4Cr-4Ti alloys were ˜10 × 10-6 MPa-1 dpa-1 below 500 °C and 50-200 × 10-6 MPa-1 dpa-1 above 500 °C, a value greater than that of commercial V-4Cr-4Ti alloys, austenitic steels and ferritic steels.

  1. Analysis of tensile deformation and failure in austenitic stainless steels: Part II - Irradiation dose dependence

    NASA Astrophysics Data System (ADS)

    Kim, Jin Weon; Byun, Thak Sang

    2010-01-01

    Irradiation effects on the stable and unstable deformation and fracture behavior of austenitic stainless steels (SSs) have been studied in detail based on the equivalent true stress versus true strain curves. An iterative finite element simulation technique was used to obtain the equivalent true stress-true strain data from experimental tensile curves. The simulation result showed that the austenitic stainless steels retained high strain hardening rate during unstable deformation even after significant irradiation. The strain hardening rate was independent of irradiation dose up to the initiation of a localized necking. Similarly, the equivalent fracture stress was nearly independent of dose before the damage (embrittlement) mechanism changed. The fracture strain and tensile fracture energy decreased with dose mostly in the low dose range <˜2 dpa and reached nearly saturation values at higher doses. It was also found that the fracture properties for EC316LN SS were less sensitive to irradiation than those for 316 SS, although their uniform tensile properties showed almost the same dose dependencies. It was confirmed that the dose dependence of tensile fracture properties evaluated by the linear approximation model for nominal stress was accurate enough for practical use without elaborate calculations.

  2. Response of austenitic steels to radiation damage

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.

    1983-01-01

    Austenitic stainless steels are prominent contenders as first wall and blanket structural materials for early fusion power reactors. Properties affecting the performance of this class of alloys in the fusion irradiation environment, such as swelling, tensile elongation, irradiation creep, fatigue, and crack growth, have been identified. These properties and the effects of neutron irradiation on them are discussed in this paper. Emphasis is placed on the present status of understanding of irradiation effects.

  3. Structure and Composition of Nanometer-Sized Nitrides in a Creep-Resistant Cast Austenitic Alloy

    NASA Astrophysics Data System (ADS)

    Evans, Neal D.; Maziasz, Philip J.; Shingledecker, John P.; Pollard, Michael J.

    2010-12-01

    The microstructure of a new and improved high-temperature creep-resistant cast austenitic alloy, CF8C-Plus, was characterized after creep-rupture testing at 1023 K (750 °C) and 100 MPa. Microstructures were investigated by detailed scanning electron microscopy, transmission electron microscopy, and energy-dispersive X-ray spectroscopy (EDS). Principal component analysis of EDS spectrum images was used to examine the complex precipitate morphology. Thermodynamic modeling was performed to predict equilibrium phases in this alloy as well as the compositions of these phases at relevant temperatures. The improved high-temperature creep strength of CF8C-Plus over its predecessor CF8C is suggested to be due to the modified microstructure and phase stability in the alloy, including the absence of δ-ferrite in the as-cast condition and the development of a stable, slow-growing precipitation hardening nitride phase—the tetragonal Z-phase—which has not been observed before in cast austenitic stainless steels.

  4. Structure and composition of nanometer-sized nitrides in a creep resistant cast austenitic alloy

    SciTech Connect

    Evans, Neal D; Maziasz, Philip J; Shingledecker, John P.; Pollard, Michael J

    2010-01-01

    The microstructure of a new and improved high-temperature creep-resistant cast austenitic alloy, CF8C-Plus, was characterized after creep-rupture testing at 1023 K (750 C) and 100 MPa. Microstructures were investigated by detailed scanning electron microscopy, transmission electron microscopy, and energy-dispersive X-ray spectroscopy (EDS). Principal component analysis of EDS spectrum images was used to examine the complex precipitate morphology. Thermodynamic modeling was performed to predict equilibrium phases in this alloy as well as the compositions of these phases at relevant temperatures. The improved high-temperature creep strength of CF8C-Plus over its predecessor CF8C is suggested to be due to the modified microstructure and phase stability in the alloy, including the absence of {delta}-ferrite in the as-cast condition and the development of a stable, slow-growing precipitation hardening nitride phase - the tetragonal Z-phase - which has not been observed before in cast austenitic stainless steels.

  5. 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.

  6. 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.

  7. Investigating the martensite-austenite transformation on mechanically alloyed FeNi solid solutions

    NASA Astrophysics Data System (ADS)

    Martínez-Bianco, D.; Gorria, P.; Blanco, J. A.; Smith, R. I.

    2011-10-01

    The martensite-austenite transformation on Fe70Ni30 and Fe75Ni25 nanostructured solid solutions has been investigated by neutron thermo-diffraction experiments carried out between 300 and 1000 K. We observe that the difference between the temperatures at which the martensitic transformation starts (Ai) and finishes (Af) exceeds 250 K, being five times larger than that of the as-cast coarse-grained conventional alloys. The main reason for this striking phenomenon is the drastic microstructural changes produced during the severe mechanical milling process, giving rise to a large reduction of the crystalline mean size (below 20 nm) and the generation of a considerable microstain (reaching 1%).

  8. 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.

  9. Structure and composition of phases occurring in austenitic stainless steels in thermal and irradiation environments

    SciTech Connect

    Lee, E.H.; Maziasz, P.J.; Rowcliffe, A.F.

    1980-01-01

    Transmission electron diffraction techniques coupled with quantitative x-ray energy dispersive spectroscopy have been used to characterize the phases which develop in austenitic stainless steels during exposure to thermal and to irradiation environments. In AISI 316 and Ti-modified stainless steels some thirteen phases have been identified and characterized in terms of their crystal structure and chemical composition. Irradiation does not produce any completely new phases. However, as a result of radiation-induced segregation principally of Ni and Si, and of enhanced diffusion rates, several major changes in phase relationships occur during irradiation. Firstly, phases characteristic of remote regions of the phase diagram appear unexpectedly and dissolve during postirradiation annealing (radiation-induced phases). Secondly, some phases develop with their compositions significantly altered by the incorporation of Ni or Si (radiation-modified phases).

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

    NASA Astrophysics Data System (ADS)

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

    1998-04-01

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

  11. Progress with alloy 33 (UNS R20033), a new corrosion resistant chromium-based austenitic material

    SciTech Connect

    Koehler, M.; Heubner, U.; Eichenhofer, K.W.; Renner, M.

    1996-11-01

    Alloy 33 (UNS R20033), a new chromium-based corrosion resistant austenitic material with nominally (wt. %) 33 Cr, 32 Fe, 31 Ni, 1.6 Mo, 0.6 Cu, 0.4 N has been introduced to the market in 1995. This paper provides new data on this alloy with respect to mechanical properties, formability, weldability, sensitization characteristics and corrosion behavior. Mechanical properties of weldments including ductility have been established, and match well with those of wrought plate material, without any degradation of ISO V-notch impact toughness in the heat affected zone. When aged up to 8 hours between 600 C and 1,000 C the alloy is not sensitized when tested in boiling azeotropic nitric acid (Huey test). Under field test conditions alloy 33 shows excellent resistance to corrosion in flowing 96--98.5% H{sub 2}SO{sub 4} at 135 C--140 C and flowing 99.1% H{sub 2}SO{sub 4} at 150 C. Alloy 33 has also been tested with some success in 96% H{sub 2}SO{sub 4} with nitrosyl additions at 240 C. In nitric acid alloy 33 is corrosion resistant up to 85% HNO{sub 3} and 75 C or even more. Alloy 33 is also corrosion resistant in 1 mol. HCl at 40 C and in NaOH/NaOCl-solutions. In artificial seawater the pitting potential remains unchanged up to 75 C and is still well above the seawater`s redox potential at 95 C. Alloy 33 can be easily manufactured into all product forms required. The new data provided support the multipurpose character of alloy 33 to cope successfully with many requirements of the Chemical Process Industry, the Oil and Gas Industry and the Refinery Industry.

  12. Phase Field Modeling of Cyclic Austenite-Ferrite Transformations in Fe-C-Mn Alloys

    NASA Astrophysics Data System (ADS)

    Chen, Hao; Zhu, Benqiang; Militzer, Matthias

    2016-08-01

    Three different approaches for considering the effect of Mn on the austenite-ferrite interface migration in an Fe-0.1C-0.5Mn alloy have been coupled with a phase field model (PFM). In the first approach (PFM-I), only long-range C diffusion is considered while Mn is assumed to be immobile during the phase transformations. Both long-range C and Mn diffusions are considered in the second approach (PFM-II). In the third approach (PFM-III), long-range C diffusion is considered in combination with the Gibbs energy dissipation due to Mn diffusion inside the interface instead of solving for long-range diffusion of Mn. The three PFM approaches are first benchmarked with isothermal austenite-to-ferrite transformation at 1058.15 K (785 °C) before considering cyclic phase transformations. It is found that PFM-II can predict the stagnant stage and growth retardation experimentally observed during cycling transformations, whereas PFM-III can only replicate the stagnant stage but not the growth retardation and PFM-I predicts neither the stagnant stage nor the growth retardation. The results of this study suggest a significant role of Mn redistribution near the interface on reducing transformation rates, which should, therefore, be considered in future simulations of austenite-ferrite transformations in steels, particularly at temperatures in the intercritical range and above.

  13. Microstructural changes of austenitic steels caused by proton irradiation under various conditions

    NASA Astrophysics Data System (ADS)

    Fukuda, T.; Sagisaka, M.; Isobe, Y.; Hasegawa, A.; Sato, M.; Abe, K.; Nishida, Y.; Kamada, T.; Kaneshima, Y.

    2000-12-01

    In austenitic steels used for light water reactors (LWRs), neutron irradiation induces many kinds of degradation. For example, irradiation assisted stress corrosion cracking (IASCC) and swelling are two forms of this degradation. Although there are a great number of studies on radiation induced segregation (RIS) and void swelling at high temperatures (>400° C) corresponding to fast and fusion reactor conditions, up to now there have been few irradiation studies at low temperatures. This paper presents microchemical and microstructural changes in type 347 and 310+Nb stainless steels due to light ion irradiation. These samples were implanted with He+ and irradiated with 2 MeV H2+ at 300°C, 350°C and 400°C. This simulated generation of transmutant He in a fusion environment. In addition, at 300°C test pieces stressed close to the yield stress were also irradiated with the same ions. The irradiation tests were carried out using the Dynamitron accelerator at Tohoku University.

  14. 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.

  15. Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Stoenescu, R.; Schäublin, R.; Gavillet, D.; Baluc, N.

    2007-02-01

    The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.

  16. Procurement and screening test data for advanced austenitic alloys for 650/degree/C steam service: Part 2, final report

    SciTech Connect

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

    1988-08-01

    The results of screening tests on alloys from three compositional groups are summarized and compared to the alloy design and performance criteria identified as needed for austenitic alloys suitable as superheater/reheater tubing in advanced heat recovery systems. The three alloy groups included lean (nominally 14% Cr and 16% Ni) austenitic stainless steels that were modifications of type 316 stainless steel, 20Cr-30Ni-Fe alloys that were modifications of alloy 800H, and Ni-Cr aluminides, (Ni,Cr)/sub 3/Al. The screening tests covered fabricability, mechanical properties, weldability, and oxidation behavior. The lean stainless steels were found to possess excellent strength and ductility if cold-worked to an equivalent strain in the range 5 to 10% prior to testing. However, they possessed marginal weldability, poor oxidation resistance, and sensitivity to aging. The modified alloy 800H alloys also exhibited good strength and ductility in the cold-worked condition. The weldability was marginal, while the oxidation resistance was good. The aluminides were difficult to fabricate by methods typically used to produce superheater tubing alloys. The alloys that could be worked had marginal strength and ductility. An aluminide cast alloy, however, was found to be very strong and ductile. 23 refs., 19 figs., 13 tabs.

  17. 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.

  18. 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.

  19. Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Stoenescu, R.; Schäublin, R.; Gavillet, D.; Baluc, N.

    2007-03-01

    The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.

  20. Effects of Co and Al Contents on Cryogenic Mechanical Properties and Hydrogen Embrittlement for Austenitic Alloys

    NASA Astrophysics Data System (ADS)

    Li, X. Y.; Ma, L. M.; Li, Y. Y.

    2004-06-01

    The effects of Co and Al content on ambient and cryogenic mechanical properties, microstructure and hydrogen embrittlement of a high strength precipitate-strengthened austenitic alloy (Fe-Ni-Cr-Mo system) had been investigated with temperature range from 293K to 77 K. Hydrogen embrittlement tests were conducted using the method of high pressure thermal hydrogen charging. It was found that increasing Co content can cause increasing in ambient and cryogenic ductility, but has less effect on ultimate tensile strength. When Co content is 9.8%, obvious decrease was found in cryogenic yield strength. Increasing Al content can result in decreasing ambient and cryogenic ductility and severe hydrogen embrittlement, but slight increase in cryogenic yield strength. Increasing Co content, reducing Al content, and decreasing test temperature tend to decrease the hydrogen embrittlement tendency for the alloys. This work showed that the alloy with composition of Fe-31%Ni-15%Cr-5%Co-4.5%Mo-2.4%Ti-0.3%Al-0.3%Nb-0.2%V has the superior cryogenic mechanical properties and lower hydrogen embrittlement tendency, is a good high strength cryogenic hydrogen-resistant material.

  1. Void swelling and precipitation in a titanium-modified austenitic stainless steel under proton irradiation

    NASA Astrophysics Data System (ADS)

    Kimoto, T.; Shiraishi, H.

    1985-07-01

    The correlation between void swelling and precipitation behavior in a 10% cold worked Fe-16.2Ni-14.6Cr-2.37Mo-1.79Mn-0.53Si-0.24Ti-0.06C alloy was examined with 200 keV proton irradiation. Swelling peak temperature after the proton irradiation to 10 dpa was about 823 K, and void swelling decreased steeply with increase in irradiation temperature from 823 to 923 K. Void swelling increased rapidly from 1.9 to 12.1% with increase in irradiation dose from 20 to 45 dpa at 873 K. Fine intragranular TiC precipitates, which were formed during initial stage of irradiation, dissolved gradually with increase in irradiation dose from 10 to 45 dpa at 873 K, while the amount of precipitation of needle-shaped Fe 2P phase containing titanium increased with increasing dose. The reduction of sink strength of the TiC precipitates due to the dissolution during irradiation was thought to cause the increase of swelling rate with increase in irradiation dose from 20 to 45 dpa at 873 K.

  2. 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.

  3. 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.

  4. Investigation on the Behavior of Austenite and Ferrite Phases at Stagnation Region in the Turning of Duplex Stainless Steel Alloys

    NASA Astrophysics Data System (ADS)

    Nomani, J.; Pramanik, A.; Hilditch, T.; Littlefair, G.

    2016-06-01

    This paper investigates the deformation mechanisms and plastic behavior of austenite and ferrite phases in duplex stainless steel alloys 2205 and 2507 under chip formation from a machine turning operation. SEM images and EBSD phase mapping of frozen chip root samples detected a build-up of ferrite bands in the stagnation region, and between 65 and 85 pct, more ferrite was identified in the stagnation region compared to austenite. SEM images detected micro-cracks developing in the ferrite phase, indicating ferritic build-up in the stagnation region as a potential triggering mechanism to the formation of built-up edge, as transgranular micro-cracks found in the stagnation region are similar to micro-cracks initiating built-up edge formation. Higher plasticity of austenite due to softening under high strain is seen responsible for the ferrite build-up. Flow lines indicate that austenite is plastically deforming at a greater rate into the chip, while ferrite shows to partition most of the strain during deformation. The loss of annealing twins and activation of multiple slip planes triggered at high strain may explain the highly plastic behavior shown by austenite.

  5. Crevice corrosion testing of austenitic, superaustenitic, superferritic, and superduplex stainless type alloys in seawater

    SciTech Connect

    Zeuthen, A.W.; Kain, R.M.

    1997-12-31

    In industry, many problems from corrosion occurring in crevices have been experienced and reported. These include the refining industry, offshore drilling platforms, fossil and nuclear power plants, chemical plants and the public utilities. The services are highly variable. Corrosion mechanisms and the results experienced are influenced by severe environments which cannot always be avoided. Corrosion testing is considered useful not only in comparing materials, but also in selecting materials from the design standpoint. The ultimate goal is to use materials which are superior to those currently in use. This will result in fewer outages, reduce repairs and significantly lower costs. This paper provides the results from four seawater test programs addressing crevice corrosion resistance of a number of superferritic, superaustenitic, and superduplex alloys, along with conventional 300 Series stainless steel. These programs included exposure to natural fouling organisms which can produce crevices, and testing which comprised several different manmade crevice configurations. Alloys found to be resistant under some test conditions were prone to attack under others. All of the super stainless steels were found to be more resistant to crevice corrosion than conventional austenitic grades, but some were susceptible to some degree.

  6. 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.

  7. 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.

  8. Defect and solute properties in dilute Fe-Cr-Ni austenitic alloys from first principles

    NASA Astrophysics Data System (ADS)

    Klaver, T. P. C.; Hepburn, D. J.; Ackland, G. J.

    2012-05-01

    to the <100> dumbbell in the tensile site by 0.1 eV and was repelled from mixed and compressive sites. In contrast, Cr showed a preferential binding to interstitials. Calculation of tracer diffusion coefficients found that Ni diffuses significantly more slowly than both Cr and Fe, which is consistent with the standard mechanism used to explain radiation-induced segregation effects in Fe-Cr-Ni austenitic alloys by vacancy-mediated diffusion. Comparison of our results with those for bcc Fe showed strong similarity for pure Fe and no correlation with dilute Ni and Cr.

  9. High Temperature Irradiation Effects in Selected Generation IV Structural Alloys

    SciTech Connect

    Nanstad, Randy K; McClintock, David A; Hoelzer, David T; Tan, Lizhen; Allen, Todd R.

    2009-01-01

    In the Generation IV Materials Program cross-cutting task, irradiation and testing were carried out to address the issue of high temperature irradiation effects with selected current and potential candidate metallic alloys. The materials tested were (1) a high-nickel iron-base alloy (Alloy 800H); (2) a nickel-base alloy (Alloy 617); (3) two advanced nano-structured ferritic alloys (designated 14YWT and 14WT); and (4) a commercial ferritic-martensitic steel (annealed 9Cr-1MoV). Small tensile specimens were irradiated in rabbit capsules in the High-Flux Isotope Reactor at temperatures from about 550 to 700 C and to irradiation doses in the range 1.2 to 1.6 dpa. The Alloy 800H and Alloy 617 exhibited significant hardening after irradiation at 580 C; some hardening occurred at 660 C as well, but the 800H showed extremely low tensile elongations when tested at 700 C. Notably, the grain boundary engineered 800H exhibited even greater hardening at 580 C and retained a high amount of ductility. Irradiation effects on the two nano-structured ferritic alloys and the annealed 9Cr-1MoV were relatively slight at this low dose.

  10. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  12. Study of the dynamical features of the austenite-martensite phase transition in the Ni50(Mn, 1%Fe)34In16 alloy using scanning Hall probe imaging

    NASA Astrophysics Data System (ADS)

    Chattopadhyay, M. K.; Morrison, K.; Dupas, A.; Sharma, V. K.; Sharath Chandra, L. S.; Cohen, L. F.; Roy, S. B.

    2012-03-01

    We have performed scanning Hall probe imaging experiments to study the martensite to austenite phase transition in the Ni50(Mn, 1%Fe)34In16 alloy as a function of temperature and magnetic field. We observe that the martensite and austenite phase regions are separated by a distinct interface. The relative growth of phase across the phase transition is associated with the movement of this interface. The movement of the interface becomes arrested at low temperature, which leads to the formation of a "magnetic glass" state in the alloy. The dynamics of the martensite to austenite phase transition in the Ni50(Mn, 1%Fe)34In16 alloy is found to be qualitatively different when the transition is field induced than what it is when the same transition is induced by temperature. While both nucleation and growth of the martensite phase is observed during the austenite to martensite phase transition in the alloy during cooling down, the martensite to austenite phase transition during warming up appears to be growth oriented. In contrast, both nucleation and growth of the product phases are observed during the field induced martensite to austenite phase transition both during increasing and decreasing field experiments. The physical reasons behind these different observations are explored.

  13. Radiation-induced evolution of austenite matrix in silicon-modified AISI 316 alloys

    SciTech Connect

    Garner, F.A.; Brager, H.R.

    1980-01-01

    The microstructures of a series of silicon-modified AISI 316 alloys irradiated to fast neutron fluences of about 2-3 and 10 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV at temperatures ranging from 400/sup 0/C to 600/sup 0/C have been examined. The irradiation of AISI 316 leads to an extensive repartition of several elements, particularly nickel and silicon, between the matrix and various precipitate phases. The segregation of nickel at void and grain boundary surfaces at the expense of other faster-diffusing elements is a clear indication that one of the mechanisms driving the microchemical evolution is the Inverse Kirkendall effect. There is evidence that at one sink this mechanism is in competition with the solute drag process associated with interstitial gradients.

  14. 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.

  15. 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.

  16. Improved microstructure for creep strength in high-temperature austenitic alloys for energy conversion applications

    NASA Astrophysics Data System (ADS)

    Rayner, Garrett

    The current dominant role of fossil fuels for use in energy conversion applications is unlikely to change in the foreseeable future. In order to ensure the continued availability of these limited resources, it is critically important that remaining fossil fuel reserves are utilized as efficiently as possible. Increasing operating temperature in power plants is the most straightforward method of increasing plant efficiency, but over long life cycles in the harsh operating conditions of modern supercritical coal-fired power plants, current-generation materials are cannot be used above ˜620°C due to corrosion and/or creep-strength limitations. One possible class of materials for higher-temperature use are dispersion-strengthened alumina-forming austenitic stainless steels: in this work, Fe-20Cr-(20-30)Ni-2Nb-5Al at. % strengthened by a fine Fe2Nb C14 Laves phase dispersion. While the Laves phase has not been successfully used as a strengthener before, some prior research has indicated that the Laves phase could act as a stable high-temperature strengthener, if it could be more finely dispersed. This work attempted to refine the Laves phase by first solutionizing the alloy, then cold-working to introduce a dense dislocation structure, and finally aging in order to allow the Laves phase to nucleate on these dislocations. Transmission electron microscopy and scanning electron microscopy were used to analyze the material after thermomechanical processing. Final results showed that the size, scale, homogeneity of dispersion, and volume fraction of precipitated Laves phase particles were all altered by prestraining, and at high levels of prestrain (90% reduction in thickness), a significantly finer Laves phase dispersion was obtained when compared with the non-prestrained aged material.

  17. The use of potential drop techniques for the evaluation of environment assisted cracking of austenitic alloys

    SciTech Connect

    Lidar, P.; Hwang, I.S.; Ballinger, R.G.

    1992-12-31

    The DC potential drop technique has been adapted for strain measurements. An AC potential drop technique has been developed for crack detection. It has been demonstrated that a DC with switched polarity can have a strain accuracy of {+-}0.12% after yielding and a multi-frequency AC system with up to 200 kHz frequency can have a sensitivity of 50 {mu}m for crack detection. The techniques have been applied to environment assisted cracking tests of Ni-Cr-Fe alloys in 350{degrees}C water and austenitic stainless steel in 288{degrees}C water using both static and dynamic loading. Long term signal stability is maintained given: (1) rigid probe attachments; (2) local preamplification; and (3) adequate lead grounding and shielding. The AC potential drop technique is found to be more suitable for constant load than for dynamic loading and is also compatible with aggressive environments. During dynamic loading, sensitivity can be significantly reduced due to serrated yielding. An analytical model has been developed to predict the DC and AC potential drops in a round bar geometry. A sensitivity analysis has been made to determine test to test variation associated with variations in supplied current, specimen dimensions, and probe spacing. Multifrequency analysis shows that the measured data agrees with the prediction at frequencies up to about 100 kHz. Above 100 kHz induced signal in the probes results in increases in potential drop and phase angle. The induced signal is reproducible and therefore may be related to the crack mouth opening displacement.

  18. Microstructures and Mechanical Properties of Irradiated Metals and Alloys

    SciTech Connect

    Zinkle, Steven J

    2008-01-01

    The effects of neutron irradiation on the microstructural evolution of metals and alloys are reviewed, with an emphasis on the roles of crystal structure, neutron dose and temperature. The corresponding effects of neutron irradiation on mechanical properties of metals and alloys are summarized, with particular attention on the phenomena of low temperature radiation hardening and embrittlement. The prospects of developing improved high-performance structural materials with high resistance to radiation-induced property degradation are briefly discussed.

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

    SciTech Connect

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

    1981-01-01

    Swelling values have been determined for a set of commercial alloys irradiated to a peak fluence of 17.8 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV) over the temperature range of 400 to 650/sup 0/C. The alloys studied fall into three classes: the ferritic alloys AISI 430F, AISI 416, EM-12, H-11 and 2 1/4 Cr-1 Mo; the superalloys Inconel 718 and Inconel X-750; and the refractory alloys TZM and Nb-1 Zr. After irradiation to a peak fluence approaching goal exposures envisioned for advanced fusion reactor first walls, all of the 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. Modeling precipitate evolution in zirconium alloys during irradiation

    NASA Astrophysics Data System (ADS)

    Robson, J. D.

    2016-08-01

    The second phase precipitates (SPPs) in zirconium alloys are critical in controlling their performance. During service, SPPs are subject to both thermal and irradiation effects that influence volume fraction, number, and size. In this paper, a model has been developed to capture the combined effect of thermal and irradiation exposure on the Zr(Fe,Cr)2 precipitates in Zircaloy. The model includes irradiation induced precipitate destabilization integrated into a classical size class model for nucleation, growth and coarsening. The model has been applied to predict the effect of temperature and irradiation on SPP evolution. Increasing irradiation displacement rate is predicted to strongly enhance the loss of particles that arises from coarsening alone. The effect of temperature is complex due to competition between coarsening and irradiation damage. As temperature increases, coarsening is predicted to become increasingly important compared to irradiation induced dissolution and may increase resistance to irradiation induced dissolution by increasing particle size.

  5. 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

  6. 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

  7. 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

  8. Local phase transformation in alloys during charged-particle irradiation

    SciTech Connect

    Lam, N.Q.; Okamoto, P.R.

    1984-10-01

    Among the various mechanisms and processes by which energetic irradiation can alter the phase stability of alloys, radiation-induced segregation is one of the most important phenomena. Radiation-induced segregation in alloys occurs as a consequence of preferential coupling between persistent fluxes of excess defects and solute atoms, leading to local enrichment or depletion of alloying elements. Thus, this phenomenon tends to drive alloy systems away from thermodynamic equilibrium, on a local scale. During charged-particle irradiations, the spatial nonuniformity in the defect production gives rise to a combination of persistent defect fluxes, near the irradiated surface and in the peak-damage region. This defect-flux combination can modify the alloy composition in a complex fashion, i.e., it can destabilize pre-existing phases, causing spatially- and temporally-dependent precipitation of new metastable phases. The effects of radiation-induced segregation on local phase transformations in Ni-based alloys during proton bombardment and high-voltage electron-microscope irradiation at elevated temperatures are discussed.

  9. 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.

  10. 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.

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

    DOE PAGES

    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 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

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

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

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

  13. 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.

  14. Self-irradiation of Pu, its alloys and compounds

    NASA Astrophysics Data System (ADS)

    Timofeeva, L. F.

    2000-07-01

    Self-irradiation of Pu, its alloys and compounds by products of known α-decomposition is a continuous complicated process, which includes numerous different phenomena. The accumulation of Pu decomposition products causes material structure and properties change. This problem is the subject of many works, most of them concerned with the behavior of Pu and its alloys at low (liquid He and N) temperatures. The survey is given of the results of our experiments connected with radiogenic helium behavior, crystal structure and properties of Pu metallic compounds and Pu oxide ceramics in a self-irradiation process at room temperature under isochronal heat treatments.

  15. 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.

  16. 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.

  17. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    NASA Astrophysics Data System (ADS)

    Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

    2011-09-01

    The C3M irradiation test, which was conducted in the experimental fast reactor, "Joyo", demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, "Monju". The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  18. Stressed capsules of austenitic and martensitic steels irradiated in SINQ Target-4 in contact with liquid lead bismuth eutectic

    NASA Astrophysics Data System (ADS)

    Dai, Y.; Gavillet, D.; Restani, R.

    2008-06-01

    In the MEGAPIE target, the steels used for the proton beam entrance window and other components in the spallation reaction zone suffer not only from the irradiation damage produced by protons and neutrons but also from the corrosion and embrittlement induced by liquid lead-bismuth eutectic (LBE). Although these effects have been separately studied by a number of authors, the synergistic effects of irradiation, LBE corrosion and embrittlement are little understood. This work presents detailed analyses of two stressed capsules made of the austenitic steel EC316LN and the martensitic steel 9Cr2WVTa, which were irradiated in SINQ Target-4 in contact with LBE at calculated temperatures of 315 and 225 °C, respectively. The Electron Probe Microanalysis (EPMA) on the cross-sections of the capsules showed that the stagnant LBE induced only slight corrosion on both capsules and no cracks existed in the wall of the EC316LN capsule. Some cracks were observed in the electron beam weld (EBW) and its vicinity of the 9Cr2WVTa capsule, which can be attributed to the high stress inside the wall, the hardening of the material induced by either welding (without re-tempering) or irradiation, and the effects of LBE embrittlement.

  19. 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.

  20. 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.

  1. 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

    2014-12-01

    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.

  2. Microstructural origins of radiation-induced changes in mechanical properties of 316 L and 304 L austenitic stainless steels irradiated with mixed spectra of high-energy protons and spallation neutrons

    NASA Astrophysics Data System (ADS)

    Sencer, B. H.; Bond, G. M.; Hamilton, M. L.; Garner, F. A.; Maloy, S. A.; Sommer, W. F.

    2001-07-01

    A number of candidate alloys were exposed to a particle flux and spectrum at Los Alamos Neutron Science Center (LANSCE) that closely match the mixed high-energy proton/neutron spectra expected in accelerator production of tritium (APT) window and blanket applications. Austenitic stainless steels 316 L and 304 L are two of these candidate alloys possessing attractive strength and corrosion resistance for APT applications. This paper describes the dose dependence of the irradiation-induced microstructural evolution of SS 316 L and 304 L in the temperature range 30-60°C and consequent changes in mechanical properties. It was observed that the microstructural evolution during irradiation was essentially identical in the two alloys, a behavior mirrored in their changes in mechanical properties. With one expection, it was possible to correlate all changes in mechanical properties with visible microstructural features. A late-term second abrupt decrease in uniform elongation was not associated with visible microstructure, but is postulated to be a consequence of large levels of retained hydrogen measured in the specimens. In spite of large amounts of both helium and hydrogen retained, approaching 1 at.% at the highest exposures, no visible cavities were formed, indicating that the gas atoms were either in solution or in subresolvable clusters.

  3. 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

    2016-09-01

    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.

  4. Corrosion processes of austenitic stainless steels and copper-based materials in gamma-irradiated aqueous environments

    SciTech Connect

    Glass, R.S.

    1985-09-01

    The US Department of Energy is evaluating a site located at Yucca Mountain in Nye County, Nevada, as a potential high-level nuclear waste repository. The rock at the proposed repository horizon (above the water table) is densely welded, devitrified tuff, and the fluid environment in the repository is expected to be primarily air-steam. A more severe environment would be present in the unlikely case of intrusion of vadose groundwater into the repository site. For this repository location, austenitic stainless steels and copper-based materials are under consideration for waste container fabrication. This study focuses on the effects of gamma irradiation on the electrochemical mechanisms of corrosion for the prospective waste container materials. The radiolytic production of such species as hydrogen peroxide and nitric acid are shown to exert an influence on corrosion mechanisms and kinetics.

  5. 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.

  6. 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.

  7. Influence of silicon on swelling and microstructure in Russian austenitic stainless steel EI-847 irradiated to high neutron doses

    NASA Astrophysics Data System (ADS)

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

    2008-08-01

    Void swelling and microstructural development of niobium-stabilized EI-847 austenitic stainless steel with a range of silicon levels were investigated by destructive examination of fuel pin cladding irradiated in three fast reactors located in either Russia or Kazakhstan. The tendency of void swelling to be progressively reduced by increasing silicon concentration appears to be a very general phenomenon in this steel, whether observed in simple, single-variable experiments on well-defined materials or when observed in multivariable, time-dependent irradiations conducted on commercially produced steels over a wide range of irradiation temperatures, neutron spectra and dpa rates. The role of silicon on microstructural development is expressed both in the solid solution via its influence on dislocation and void microstructure and via its influence on formation of radiation-induced phases that in turn alter the matrix composition. Surprisingly, increases in silicon level in this study do not accelerate the formation of silicon-rich G-phase, but act to increase the formation of Nb (C,N) precipitates. Such precipitates are known to be associated with delayed void swelling.

  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. The interaction of point defects with line dislocations in HVEM (high voltage electron microscope) irradiated Fe-Ni-Cr alloys

    SciTech Connect

    King, S.L.; Jenkins, M.L. . Dept. of Materials); Kirk, M.A. ); English, C.A. . Materials Development Div.)

    1990-05-01

    This paper presents results of a study of the interaction of point defects produced by high voltage electron microscope (HVEM) irradiation with pre-existing dislocations in austenitic Fe-15% 25%Ni-17%Cr alloys, aimed at the determination of the mechanisms of climb of dissociated dislocations. Dislocations were initially characterized at sub-threshold voltages (here 200kV) using the weak-beam technique. These dislocations were then irradiated with 1MeV electrons in the Argonne HVEM before being returned to a lower voltage microscope for post-irradiation characterization. Interstitial climb was seen only at particularly favorable sites, such as pre-existing jogs, whilst vacancies clustered near dislocations, forming stacking fault tetrahedra (SFT). Partial separations were also observed to have decreased after irradiation. The post-irradiation configuration was found to depend strongly on both dislocation character and pre-irradiation dislocation configuration. These results, and their relevance to the void swelling problem, are discussed. 52 refs., 8 figs.

  10. A study of the micro- and nanoscale deformation behavior of individual austenitic dendrites in a FeCrMoVC cast alloy using micro- and nanoindentation experiments

    NASA Astrophysics Data System (ADS)

    Zeisig, J.; Hufenbach, J.; Wendrock, H.; Gemming, T.; Eckert, J.; Kühn, U.

    2016-04-01

    Micro- and nanoindentation experiments were conducted to investigate the deformation mechanisms in a Fe79.4Cr13Mo5V1C1.6 (wt. %) cast alloy. This alloy consists of an as cast microstructure mainly composed of austenite, martensite, and a complex carbide network. During microhardness testing, metastable austenite transforms partially into martensite confirmed by electron backscatter diffraction. For nanoindentation tests, two different indenter geometries were applied (Berkovich and cube corner type). Load-displacement curves of nanoindentation in austenitic dendrites depicted pop-ins after transition into plastic deformation for both nanoindenters. Characterizations of the region beneath a nanoindent by transmission electron microscopy revealed a martensitic transformation as an activated deformation mechanism and suggest a correlation with the pop-in phenomena of the load-displacement curves. Furthermore, due to an inhomogeneous chemical composition within the austenitic dendrites, more stabilized regions deform by mechanical twinning. This additional deformation mechanism was only observed for the cube corner indenter with the sharper geometry since higher shear stresses are induced beneath the contact area.

  11. Complete and Incomplete Wetting of Ferrite Grain Boundaries by Austenite in the Low-Alloyed Ferritic Steel

    NASA Astrophysics Data System (ADS)

    Straumal, B. B.; Kucheev, Y. O.; Efron, L. I.; Petelin, A. L.; Majumdar, J. Dutta; Manna, I.

    2012-05-01

    Low-carbon low-alloyed ferritic steels are the main material for the production of high-strength pipes for the transportation of oil and gas. The formation of brittle carbide network during the lifetime of a pipeline could be a reason for a catastrophic failure. Among other reasons, it can be controlled by the morphology of grain boundary (GB) carbides. The microstructure of a low-alloyed ferritic steel containing 0.09 at.% C and small amounts of Si, Mn, Nb, Cu, Al, Ni, and Cr was studied between 300 and 900 °C. The samples were annealed very long time (700 to 4000 h) in order to produce the equilibrium morphology of phases. The (α-Fe)/(α-Fe) GBs can be either completely or incompletely wetted (covered) by the γ-Fe (austenite) above the temperature of eutectoid transition. The portion of (α-Fe)/(α-Fe) GBs completely wetted by γ-Fe is around 90% and does not change much between 750 and 900 °C. The (α-Fe)/(α-Fe) GBs can be either completely or incompletely wetted (covered) by the Fe3C (cementite) below the temperature of eutectoid transition. The portion of (α-Fe)/(α-Fe) GBs completely wetted by Fe3C changes below 680 °C between 67 and 77%. The formation of the network of brittle cementite layers between ductile ferrite grains can explain the catastrophic failure of gas- and oil-pipelines after a certain lifetime.

  12. Effect of Microstructure on Retained Austenite Stability and Tensile Behaviour in an Aluminum-Alloyed TRIP Steel

    NASA Astrophysics Data System (ADS)

    Chiang, Jasmine Sheree

    Transformation-induced plasticity (TRIP) steels have excellent strength, ductility and work hardening behaviour, which can be attributed to a phenomenon known as the TRIP effect. The TRIP effect involves a metastable phase, retained austenite (RA), transforming into martensite as a result of applied stress or strain. This transformation absorbs energy and improves the work hardening rate of the steel, delaying the onset of necking. This work describes two distinct TRIP steel microstructures and focuses on how microstructure affects the RA-to-martensite transformation and the uniaxial tensile behaviour. A two-step heat treatment was applied to an aluminum-alloyed TRIP steel to obtain a microstructure consisting of equiaxed grains of ferrite surrounded by bainite, martensite and RA -- the equiaxed microstructure. The second microstructure was produced by first austenitizing and quenching the steel to produce martensite, followed by the two-step heat treatment. The resulting microstructure (labelled the lamellar microstructure) consisted of elongated grains of ferrite with bainite, martensite and RA grains. Both microstructural variants had similar initial volume fractions of RA. A series of interrupted tensile tests and ex-situ magnetic measurements were conducted to examine the RA transformation during uniform elongation. Similar tests were also conducted on an equiaxed microstructure and a lamellar microstructure with similar ultimate tensile strengths. Results show that the work hardening rate is directly related to the RA transformation rate. The slower transformation rate, or higher RA stability, that was observed in the lamellar microstructure enables sustained work hardening at high strains. In contrast, the equiaxed microstructure has a lower RA stability and thus exhibits high values of work hardening at low strains, but the effect is quickly exhausted. Several microstructural factors that affect RA stability were examined, including RA grain size, aspect

  13. Post-irradiation annealing effects of austenitic stainless steels in IASCC

    SciTech Connect

    Katsura, Ryoei; Ishiyama, Yoshihide; Yokota, Norikatu; Kato, Takahiko; Nakata, Kiyotomo; Fukuya, Kouji; Sakamoto, Hiroshi; Asano, Kyoichi

    1998-12-31

    Post-irradiation annealing effects on the thermal sensitization and IASCC recovery for highly irradiated types 304 and 316L stainless steels were investigated using EPR and SSR tests. Irradiated type 316L stainless steel (neutron fluence: 8 x 10{sup 25} n/m{sup 2}, E > 1 MeV) was not sensitized and IGSCC susceptibility significantly was reduced to 7--0% at 400--700 C (x1h) from 23% at as-irradiated condition. Irradiated type 304 stainless steel (neutron fluence: 1.2 x 10{sup 26} n/m{sup 2}, E > 1MeV) was more easily sensitized than unirradiated material and IGSCC susceptibility was reduced to 62--45% at 400--500 C from 95% at the as-irradiated condition. These results on types 304 and 316L stainless steels indicated that the thermal healing technic enhanced IASCC recovery.

  14. 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.

  15. 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.

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

    NASA Astrophysics Data System (ADS)

    Khan, Mahmud; Brock, Jeffrey; Sugerman, Ian

    2016-02-01

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

  17. 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.

  18. 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.

  19. 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.

  20. Effects of silicon, carbon and molybdenum additions on IASCC of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Nakano, J.; Miwa, Y.; Kohya, T.; Tsukada, T.

    2004-08-01

    To study the effects of minor elements on irradiation assisted stress corrosion cracking (IASCC), high purity type 304 and 316 stainless steels (SSs) were fabricated and minor elements, Si or C were added. After neutron irradiation to 3.5 × 10 25 n/m 2 ( E>1 MeV), slow strain rate tests (SSRTs) of irradiated specimens were conducted in oxygenated high purity water at 561 K. Specimen fractured surfaces were examined using a scanning electron microscope (SEM) after the SSRTs. The fraction of intergranular stress corrosion cracking (IGSCC) on the fractured surface after the SSRTs increased with neutron fluence. In high purity SS with added C, the fraction of IGSCC was the smallest in the all SSs, although irradiation hardening level was the largest of all the SSs. Addition of C suppressed the susceptibility to IGSCC.

  1. 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.

  2. 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).

  3. 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).

  4. 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)

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

    SciTech Connect

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

    1995-08-01

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

  6. 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.

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

    SciTech Connect

    Farmer, J C

    2008-11-25

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

  8. 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.

  9. 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.

  10. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2000-01-01

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

  12. 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.

  13. 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.

  14. 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.

  15. The effect of alloying elements on the defect structural evolution in neutron irradiated Ni alloys

    NASA Astrophysics Data System (ADS)

    Yoshiie, T.; Xu, Q.; Satoh, Y.; Ohkubo, H.; Kiritani, M.

    2000-12-01

    The effect of alloying elements, Si (-5.8%: the volume size factor in Ni), Ge (+14.76%) and Sn (+74.08%), on void swelling in neutron irradiated Ni at 573 K was studied by transmission electron microscope (TEM) observation and positron annihilation lifetime measurement. Neutron irradiation dose was changed widely from 0.001 to 0.4 dpa using two reactors, the Kyoto University reactor (KUR) and the Japan materials testing reactor (JMTR). Voids were observed in pure Ni by TEM even after very small irradiation dose of 0.001 dpa. With increasing dose, the density of voids did not change much while their size increased. The same tendency was observed in Ni-2at.%Ge. In Ni-2at.%Sn and Ni-2at.%Si, however, no voids were observed by TEM at a damage dose of 0.4 dpa. But positron lifetime measurement revealed the existence of microvoids at a medium dose of irradiation. When irradiation dose increased to 0.4 dpa in Ni-2at.%Si and 0.13 dpa in Ni-2at.%Sn, their existence was not detected. Suppression of microvoids in these alloys is discussed from the standpoint of solute point defect interactions.

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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

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

    SciTech Connect

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

    2009-11-01

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

  7. 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.

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

    DOE PAGES

    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.; et al

    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

  9. 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.

  10. Relationship of microstructure and tensile properties for neutron-irradiated vanadium alloys

    SciTech Connect

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

    1990-01-01

    The microstructures in V-15Cr-5Ti, V-10Cr-5RTi, V-3Ti-1Si, V-15Ti-7.5Cr, and V-20Ti alloys were examined by transmission electron microscopy after neutron irradiation at 600{degree}C to 21--84 atom displacements per atom in the Materials Open Test Assembly of the Fast Flux Test Facility. The microstructures in these irradiated alloys were analyzed to determine the radiation-produced dislocation density, precipitate number density and size, and void number density and size. The results of these analyses were used to compute increases in yield stress and swelling of the irradiated alloys. The computed increase in yield stress was compared with the increase in yield stress determined from tensile tests on these irradiated alloys. This comparison made it possible to evaluate the influence of alloy composition on the evolution of radiation-damaged microstructures and the resulting tensile properties. 11 refs.

  11. 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.

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

    NASA Astrophysics Data System (ADS)

    Ohnuki, Soumei; Takahashi, Heishichiro; Takeyama, Taro

    1984-05-01

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

  13. Effect of irradiation on the tensile properties of niobium-base alloys

    SciTech Connect

    Grossbeck, M.L.; Heestand, R.L.; Atkin, S.D.

    1986-11-01

    The alloys Nb-1Zr and PWC-11 (Nb-1Zr-0.1C) were selected as prime candidate alloys for the SP-100 reactor. Since the mechanical properties of niobium alloys irradiated to end-of-life exposure levels of about 2 x 10SW neutrons/mS (E > 0.1 MeV) at temperatures above 1300 K were not available, an irradiation experiment (B-350) in EBR-II was conducted. Irradiation creep, impact properties, bending fatigue, and tensile properties were investigated; however, only tensile properties will be reported in this paper. The tensile properties were studied since they easily reveal the common irradiation phenomena of hardening and embrittlement. Most attention was directed to testing at the irradiation temperature. Further testing was conducted at lower temperatures in order to scope the behavior of the alloys in cooldown conditions.

  14. 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.

  15. 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.

  16. Effect of cold-work on self-welding susceptibility of austenitic stainless steel (alloy D9) in high temperature flowing sodium

    NASA Astrophysics Data System (ADS)

    Meikandamurthy, C.; Kumar, Hemant; Chakraborty, Gopa; Albert, S. K.; Ramakrishnan, V.; Rajan, K. K.; Bhaduri, A. K.

    2010-12-01

    Self-welding susceptibility of alloy D9 (15Cr-15Ni-2Mo titanium-modified austenitic stainless steel), used as wrapper in the fuel subassemblies of sodium cooled fast reactor, was studied in flowing sodium. Specimens were tested at 823 K in annealed and in 20% cold-worked condition up to a maximum contact stress of 24.5 MPa and maximum duration of 9 months. The results showed that the annealed alloy D9 showed good resistance to self-welding in all the tests. But 20% cold-worked alloy D9 got self-welded in all the tests except in the test carried out for 3 months duration indicating that tests conducted at high contact stresses and long duration reduce the resistance of the steel to self-weld. Microstructural changes observed in the cold-worked alloy D9 at the location of contact between the mating surfaces indicate dynamic recovery resulting from high contact stress and temperature facilitating self-weld.

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

    NASA Astrophysics Data System (ADS)

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

    2010-12-01

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

  18. Microstructure and mechanical properties of 2024-T3 and 7075-T6 aluminum alloys and austenitic stainless steel 304 after being exposed to hydrogen peroxide

    NASA Astrophysics Data System (ADS)

    Sofyan, Nofrijon Bin Imam

    The effect of hydrogen peroxide used as a decontaminant agent on selected aircraft metallic materials has been investigated. The work is divided into three sections; bacterial attachment behavior onto an austenitic stainless steel 304 surface; effect of decontamination process on the microstructure and mechanical properties of aircraft metallic structural materials of two aluminum alloys, i.e. 2024-T3 and 7075-T6, and an austenitic stainless steel 304 as used in galley and lavatory surfaces; and copper dissolution rate into hydrogen peroxide. With respect to bacterial attachment, the results show that surface roughness plays a role in the attachment of bacteria onto metallic surfaces at certain extent. However, when the contact angle of the liquid on a surface increased to a certain degree, detachment of bacteria on that surface became more difficult. In its relation to the decontamination process, the results show that a corrosion site, especially on the austenitic stainless steel 304 weld and its surrounding HAZ area, needs more attention because it could become a source or a harborage of bio-contaminant agent after either incidental or intentional bio-contaminant delivery. On the effect of the decontamination process on the microstructure and mechanical properties of aircraft metallic structural materials, the results show that microstructural effects are both relatively small in magnitude and confined to a region immediately adjacent to the exposed surface. No systematic effect is found on the tensile properties of the three alloys under the conditions examined. The results of this investigation are promising with respect to the application of vapor phase hydrogen peroxide as a decontaminant agent to civilian aircraft, in that even under the most severe circumstances that could occur; only very limited damage was observed. The results from the dissolution of copper by concentrated liquid hydrogen peroxide showed that the rate of copper dissolution increased for

  19. 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.

  20. 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.

  1. 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.

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

    DOE PAGES

    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.

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

    DOE PAGES

    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.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

  4. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

    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 α‧ 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. 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.

  6. 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.

  7. Development of creep resistant austenitic stainless steels for advanced steam cycle superheater application. [Uses of radiation effects to guide alloy development

    SciTech Connect

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

    1987-01-01

    The compositions of several 14Cr-16Ni austenitic stainless steels were modified with combinations of minor and residual alloying elements to produce excellent creep strength based on unique precipitate microstructures. These modifications produce fine MC and phosphide precipitates in the matrix for strength and various coarser carbide phases along the grain boundaries for ductility and rupture resistance. Creep-rupture resistance of these modified 14-16 steels is much better than that of type 316 or Inconel 800H and better than that of 17-14CuMo at 700C in the mill-annealed condition. Analysis of microstructure and correlation with creep properties suggests that precipitate effects are primarily responsible for the properties improvement. The ideas and insight for design of the novel precipitate microstructures stem from microcompositional information obtained using state-of-the-art analytical electron microscopy (AEM). 5 refs., 13 figs., 1 tab.

  8. The effect of MC and MN stabilizer additions on the creep rupture properties of helium implanted Fe-25% Ni-15% Cr austenitic alloy

    NASA Astrophysics Data System (ADS)

    Yamamoto, Norikazu; Nagakawa, Johsei; Shiraishi, Haruki

    1995-10-01

    Helium embrittlement resistance of Fe-25% Ni-15% Cr austenitic alloys with various MX (M = V, Ti, Nb, Zr; X = C, N) stabilizers was compared through post helium implantation creep testing at 923 K. While significant deterioration by helium in terms of creep rupture time and elongation occurred for all materials investigated, the suppression of the deterioration, especially in rupture time, was discerned for the materials in which semi-coherent MC (M = Ti, Ti + Nb, V + Ti) particles were distributed at high density. The material which contains the incoherent M 23C 6 as predominant precipitates seems to be less degraded by helium than those containing the MXs (M = Zr, V; X = C, N), if compared at the same number density of precipitates. Therefore, it is suggested that the high density dispersion of incoherent M 23C 6 as well as semi-coherent Ti containing MC particles would be beneficial in reducing the detrimental helium influences on mechanical properties.

  9. 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

  10. Mechanisms of Neutron Irradiation Hardening in Impurity-Doped Ferritic Alloys

    NASA Astrophysics Data System (ADS)

    Nishiyama, Y.; Liu, X. Y.; Kameda, J.

    2008-05-01

    Mechanisms of neutron irradiation hardening in phosphorus (P)-doped, sulfur (S)-doped, and copper (Cu)-doped ferritic alloys have been studied by applying a rate theory to the temperature dependence of the yield strength. Hardening behavior induced by neutron irradiation at various temperatures (473 to 711 K) is characterized in terms of the variations in athermal stress and activation energy for plasticity controlled by precipitation or solid solution, and kink-pair formation with the content and type of impurities. In P-doped alloys, neutron irradiation below 563 K brings about a remarkable increase in the athermal stress and activation energy, due to the dispersion of fine (˜1.7-nm) P-rich precipitates that is more extensive than that for the Cu-rich precipitates reported in irradiated steel. During neutron irradiation above 668 K, precipitation hardening occurs to some extent in Cu-doped and S-doped alloys, compared to small or negligible hardening in the P-doped alloys. In alloys with a low to moderate content of various dissolved impurities subjected to high-temperature irradiation, the formation of kink pairs becomes considerably difficult. Differing dynamic interactions of dissolved and precipitated impurities, i.e., P and Cu, with the nucleation and growth of dislocations are discussed, giving rise to irradiation hardening.

  11. 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.

  12. Deuterium ion irradiation induced precipitation in Fe-Cr alloy: Characterization and effects on irradiation behavior

    NASA Astrophysics Data System (ADS)

    Liu, P. P.; Yu, R.; Zhu, Y. M.; Zhao, M. Z.; Bai, J. W.; Wan, F. R.; Zhan, Q.

    2015-04-01

    A new phase was found to precipitate in a Fe-Cr model alloy after 58 keV deuterium ion irradiation at 773 K. The nanoscale radiation-induced precipitate was studied systematically using high resolution transmission electron microscopy (HRTEM), image simulation and in-situ ultrahigh voltage transmission electron microscopy (HVEM). B2 structure is proposed for the new Cr-rich phase, which adopts a cube-on-cube orientation relationship with regard to the Fe matrix. Geometric phase analysis (GPA) was employed to measure the strain fields around the precipitate and this was used to explain its characteristic 1-dimensional elongation along the <1 0 0> Fe direction. The precipitate was stable under subsequent electron irradiation at different temperatures. We suggest that the precipitate with a high interface-to-volume ratio enhances the radiation resistance of the material. The reason for this is the presence of a large number of interfaces between the precipitate and the matrix, which may greatly reduce the concentration of point defects around the dislocation loops. This leads to a significant decrease in the growth rate.

  13. 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.

  14. 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.

  15. Effect of neutron irradiation on fracture toughness behaviour of copper alloys

    NASA Astrophysics Data System (ADS)

    Tähtinen, S.; Pyykkönen, M.; Karjalainen-Roikonen, P.; Singh, B. N.; Toft, P.

    1998-10-01

    One of the most important factors in deciding about the applicability of materials in the structural components of ITER, is the effect of neutron irradiation on the fracture toughness behaviour of these materials. In the present work, the fracture toughness properties of two candidate materials for the first wall and divertor components of ITER, namely precipitation hardened CuCrZr and dispersion hardened CuAl25 alloys, have been studied in the unirradiated and irradiated conditions. In parallel, tensile properties of these alloys have been also investigated in the unirradiated and irradiated conditions.

  16. 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.

  17. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2013-07-01

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

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Irradiation creep of 11Cr-0.5Mo-2W,V,Nb ferritic-martensitic, modified 316, and 15Cr-20Ni austenitic S.S. irradiated in FFTF to 103-206 dpa

    NASA Astrophysics Data System (ADS)

    Uehira, A.; Mizuta, S.; Ukai, S.; Puigh, R. J.

    2000-12-01

    The irradiation creep of 11Cr-0.5Mo-2W-0.2V-0.05Nb ferritic-martensitic (PNC-FMS), modified 316 (PNC316) and 15Cr-20Ni base austenitic S.S. were determined by the gas pressurized capsule irradiation test using MOTA in FFTF. The pressurized capsules and open tubes were irradiated at 678-943 K to a peak dose of 206 dpa. The irradiation creep coefficients were derived from the diametral change differences between the capsules and open tubes, accounting for the stress-induced swelling. The creep compliance B0 and creep-swelling coupling coefficient D for PNC-FMS were found to be 0.43-0.76×10-6 MPa-1 dpa-1 and 0.85-2.5×10-2 MPa-1 for volumetric swelling, respectively. For both PNC316 and 15Cr-20Ni base S.S. the irradiation creep properties were very similar. B0 and D range from 0.55 to -1.5×10-6 MPa-1 dpa-1 and from 1.2 to -2.8×10-3 MPa-1, respectively.

  6. 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.

  7. 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.

  8. 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.

  9. Microstructural Development in Irradiated U-7Mo/6061 Al Alloy Matrix Dispersion Fuel

    SciTech Connect

    Dennis D. Keiser, Jr.; Adam B. Robinson; Jan-Fong Jue; Pavel G. Medvedev; Daniel M. Wachs; M. Ross Finlay

    2009-09-01

    A U-7Mo alloy/6061 Al alloy matrix dispersion fuel plate was irradiated in the Advanced Test Reactor and then destructively examined using optical metallography and scanning electron microscopy to characterize the developed microstructure. Results were compared to the microstructures of as-fabricated dispersion fuel to identify changes that occurred during irradiation. The interaction layers that formed on the surface of the fuel U-7Mo particles during fuel fabrication exhibited stable irradiation performance as a result of the ~0.88 wt% Si present in the fuel meat matrix. During irradiation, the interaction layers changed very little in thickness and composition. The overall irradiation performance of the fuel plate to moderate power and burnup was considered excellent.

  10. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2011-07-01

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

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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

    2016-01-01

    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

  17. 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.

  18. 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

  19. 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

  20. Mechanical properties of HIP bonded joints of austenitic stainless steel and Cu-alloy for fusion experimental reactor blanket

    NASA Astrophysics Data System (ADS)

    Sato, S.; Kuroda, T.; Kurasawa, T.; Furuya, K.; Togami, I.; Takatsu, H.

    1996-10-01

    Tensile, fatigue and impact properties have been measured for hot isostatic pressing (HIP) bonded joints of type 316 austenitic stainless steel (SS316)/SS316, and of SS316/Al 2O 3 dispersion strengthened copper (DSCu). The HIP bonded joints of SS316/SS316 had almost the same tensile and fatigue properties as those of the base metal. The HIP bonded joints of SS316/DSCu had also almost the same tensile properties as those of the base metal of the DSCu, though total elongation and fatigue strength were slightly lower than those of the DSCu base metal. Further data accumulation, even with further optimization of fabrication conditions, is required, especially for HIP bonded SS316/DSCu joints, to confirm above data and reflect to blanket/first wall design.

  1. Nanostructure evolution under irradiation in FeMnNi alloys: A "grey alloy" object kinetic Monte Carlo model

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

    This work extends the object kinetic Monte Carlo model for neutron irradiation-induced nanostructure evolution in Fe-C binary alloys developed in [1], introducing the effects of substitutional solutes like Mn and Ni. The objective is to develop a model able to describe the nanostructural evolution of both vacancy and self-interstitial atom (SIA) defect cluster populations in Fe(C)MnNi neutron-irradiated model alloys at the operational temperature of light water reactors (∼300 °C), by simulating specific reference irradiation experiments. To do this, the effects of the substitutional solutes of interest are introduced, under simplifying assumptions, using a "grey alloy" scheme. Mn and Ni solute atoms are not explicitly introduced in the model, which therefore cannot describe their redistribution under irradiation, but their effect is introduced by modifying the parameters that govern the mobility of both SIA and vacancy clusters. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proved to be key to explain the experimentally observed disappearance of detectable defect clusters with increasing solute content. Solute concentration is explicitly taken into account in the model as a variable determining the slowing down of self-interstitial clusters; small vacancy clusters, on the other hand, are assumed to be significantly slowed down by the presence of solutes, while for clusters bigger than 10 vacancies their complete immobility is postulated. The model, which is fully based on physical considerations and only uses a few parameters for calibration, is found to be capable of reproducing the experimental trends in terms of density and size distribution of the irradiation-induced defect populations with dose, as compared to the reference experiment, thereby providing insight into the physical mechanisms that influence the nanostructural evolution undergone by this material during irradiation.

  2. Effect of neutron irradiation on the tensile properties and microstructure of several vanadium alloys

    SciTech Connect

    Braski, D.N.

    1986-01-01

    Specimens of V-15Cr-5Ti, VANSTAR-7, and V-3Ti-1Si were encapsulated in TZM tubes containing /sup 7/Li to prevent interstitial pickup and irradiated in FFTF (MOTA experiment) to a damage level of 40 dpa. The irradiation temperatures were 420, 520, and 600/sup 0/C. For a better simulation of fusion reactor conditions, helium was preimplanted in some specimens using a modified version of the ''tritium trick.'' The V-15Cr-5Ti alloy was most susceptible to irradiation hardening and helium embrittlement, followed by VANSTAR-7 and V-3Ti-1Si. VANSTAR-7 exhibited a relatively high maximum void swelling of approx.6% at 520/sup 0/C while V-15Cr-5Ti and V-3Ti-1Si had values of less than 0.3% at all three temperatures. The V-3Ti-1Si clearly outperformed the other two vanadium alloys in resisting the effects of neutron irradiation.

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

    DOE PAGES

    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.

  4. 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.

  5. 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.

  6. Effect of phosphorus on Fe grain boundary self-diffusion in austenitic Fe-20Ni-10Cr-xP alloys

    SciTech Connect

    Cermak, J.; Ruzickova, J.; Pokorna, A. . Inst. of Physical Metallurgy)

    1994-08-15

    Austenitic Ni-Cr steels are extensively used as refractory and corrosion-resistant materials. Frequently, they are used at temperatures approaching T[sub m]/2 (T[sub m] is melting point). In this temperature range, all transport processes controlled by diffusion almost exclusively involve high-diffusivity paths, while the lattice diffusion is frozen. In a single-phase material without interphase interfaces, grain boundaries (GB's) and dislocation pipes act as high-diffusivity paths. It is well known that impurities, the solubility limit of which are low, concentrate at GB's. Because of the very small thickness of a GB, [delta], they form relatively highly concentrated thin zones along GB's even if their mean concentration is very low. Moreover, the changes of chemical composition of these zones may be even more complicated owing to e.g., cooperative and site-competitive interactions between atoms in the GB, and/or to heterogeneous precipitation on GB's induced by segregation. Therefore, diffusion along GB's is strongly affected by the chemistry of the GB, especially by segregated so-called surface active elements. At present, there is no universal theory describing all details of diffusion along segregated GB's, which is, partially, caused by lack of experimental data. In the present paper, the effect of small concentrations of phosphorus, that is one of most deleterious impurities in iron alloys, on GB self-diffusion of iron in Fe-Ni-Cr-xP system is studied.

  7. 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.

  8. 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.

  9. 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.

  10. Resistivity measurements of neutron-irradiated pure metals and Al-Zn alloys

    NASA Technical Reports Server (NTRS)

    Horak, J. A.

    1968-01-01

    Report presents resistivity measurements and their interpretation for neutron-irradiated pure metals and Al-Zn alloys. The influence of temperature, the role of point defects, and the aging behavior on resistivity are considered. The experimental procedures and results are discussed in detail.

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

    NASA Astrophysics Data System (ADS)

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

    2002-08-01

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

  12. 84 MeV C-ions irradiation effects on Zr-45Ti-5Al-3V alloy

    NASA Astrophysics Data System (ADS)

    Wang, Weipeng; Li, Zhengcao; Zhang, Zhengjun; Zhang, Chonghong

    2014-09-01

    Newly developed Zr-45Ti-5Al-3V alloy were irradiated by 84 MeV carbon ions with doses of 4 * 1015 ions/cm2 and 12 * 1015 ions/cm2, respectively. XRD, SEM, TEM, SAD and tensile tests were performed to study the microstructural evolution and mechanical properties modification upon high energy carbon ion irradiation. XRD patterns show no phase change while the diffraction peak position and intensity vary with irradiation doses. Tensile tests verify monotonic change of alloy strengths and elongations upon irradiation. Microstructure observations of the irradiated samples reveal the irradiation-induced precipitation of (Zr,Ti)3C2, which was believed contributing to the alloy hardening. Superlattice was discovered by the SAD patterns of original and irradiated samples and the high energy C-ions implantation was demonstrated to promote the disorder-order transition by introducing lattice defects.

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

    DOE PAGES

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

    2016-02-17

    Model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) have been neutron irradiated 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. This is significantly lower than the Cr-content of α' in binary FeCr alloys. As a result, significant partitioning of the Al from themore » α' precipitates was also observed.« less

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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}.

  19. 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.

  20. Experimental study and numerical modelling of the irradiation damage recovery in zirconium alloys

    NASA Astrophysics Data System (ADS)

    Ribis, J.; Onimus, F.; Béchade, J.-L.; Doriot, S.; Barbu, A.; Cappelaere, C.; Lemaignan, C.

    2010-08-01

    Neutron irradiation damage in zirconium alloys used as fuel cladding tubes for Pressurized Water Reactors in the nuclear industry consists mainly in a high density of small prismatic dislocation loops. During post-irradiation heat treatment thermal annealing of loops occurs. This phenomenon has been investigated by transmission electron microscopy and microhardness tests. It has been shown that the loop density decreases while their mean size increases. Furthermore it was demonstrated that only vacancy loops remain present in the material after a long term annealing at high temperature. A mechanism based on vacancies diffusion has been proposed to explain the loop evolution during annealing. A cluster dynamic model, originally developed to compute the evolution of the microstructure under irradiation, has been adapted to the modelling of the annealing for zirconium alloys. This physically based model reproduces the loop size and density evolution during a large variety of heat treatments and also provides a better understanding of the mechanisms involved in the loop recovery.

  1. 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.

  2. 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.

  3. 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.

  4. Surface microstructure and antibacterial property of an active-screen plasma alloyed austenitic stainless steel surface with Cu and N.

    PubMed

    Dong, Y; Li, X; Bell, T; Sammons, R; Dong, H

    2010-10-01

    Antibacterial modification of medical materials has already been developed as a potentially effective method for preventing device-associated infections. However, the thin layer generated, often less than 1 µm, cannot ensure durability for metal devices in constant use. A novel stainless steel surface with both a quick bacterial killing rate and durability has been developed by synthesizing Cu and a supersaturated phase (S-phase) using a new active screen plasma alloying technology. This paper investigated the microstructure of a multilayer (using EDS/WDS, SEM, TEM and XRD) and the viability of bacteria attached to biofunctional surfaces (using the spread plate method). The experimental results demonstrate that the plasma alloyed multilayered surface case consists of three sublayers: a nano-crystalline (Fe, Cr, Ni)3N deposition layer (∼200 nm), a unique Cu-containing face-centred cubic (f.c.c.) γ'-M4N (M=Fe, Cr, Ni, Cu) layer and a Cu/N S-phase layer. The thicknesses of the total treated case and the Cu-containing layers are 15 and 8 µm, respectively. Copper exists as substitutional atoms in the γ'-M4N (with a constant concentration of about 5 at%) and in the S-phase lattice (reduces from 5 to 0 at%). The crystal constant of the Cu/N S-phase layer ranged from 0.386 to 0.375 nm, which is expanded by γ from 4.4% to 7.5%. An effective reduction of 99% of Escherichia coli (E. coli) within 3 h was achieved by contact with the homogeneous Cu alloyed surface. No viable E. coli was found after 6 h (100% killed). PMID:20876967

  5. Investigation of Effects of Neutron Irradiation on Tantalum Alloys for Radioisotope Power System Applications

    SciTech Connect

    Barklay, Chadwick D.; Kramer, Daniel P.; Talnagi, Joseph

    2007-01-30

    Tantalum alloys have been used by the U.S. Department of Energy as structural alloys for space nuclear power systems such as Radioisotopic Thermoelectric Generators (RTG) since the 1960s. Tantalum alloys are attractive for high temperature structural applications due to their high melting point, excellent formability, good thermal conductivity, good ductility (even at low temperatures), corrosion resistance, and weldability. A number of tantalum alloys have been developed over the years to increase high-temperature strength (Ta-10%W) and to reduce creep strain (T-111). These tantalum alloys have demonstrated sufficient high-temperature toughness to survive the increasing high pressures of the RTG's operating environment resulting from the alpha decay of the 238-plutonium dioxide fuel. However, 238-plutonium is also a powerful neutron source. Therefore, the RTG operating environment produces large amounts of 3-helium and neutron displacement damage over the 30 year life of the RTG. The literature to date shows that there has been very little work focused on the mechanical properties of irradiated tantalum and tantalum alloys and none at the fluence levels associated with a RTG operating environment. The minimum, reactor related, work that has been reported shows that these alloys tend to follow trends seen in the behavior of other BCC alloys under irradiation. An understanding of these mechanisms is important for the confident extrapolation of mechanical-property trends to the higher doses and gas levels corresponding to actual service lifetimes. When comparing the radiation effects between samples of Ta-10%W and T-111 (Ta-8%W-2%Hf) subjected to identical neutron fluences and environmental conditions at temperatures <0.3Tm ({approx}700 deg. C), evidence suggests the possibility that T-111 will exhibit higher levels of internal damage accumulation and degradation of mechanical properties compared to Ta-10%W.

  6. Irradiation-induced phase transformations in zirconium alloys

    SciTech Connect

    Howe, L.M.; Phillips, D.; Motta, A.T.; Okamoto, P.R.

    1994-02-01

    {sup 40}Ar and {sup 209}Bi ion irradiations of Zr{sub 3}Fe were performed at 35--725 K using 15-1500 keV ions. Results are presented on role of deposited-energy density on nature of the damaged regions in individual cascades produced by ion bombardment of Zr{sub 3}Fe. Comparison is also made between irradiation-induced amorphization of Zr{sub 3}Fe during electron irradiation and under ion bombardments. Dependence of damage production on incident electron energy in Zr{sub 3}Fe was also determined. Preliminary results are also discussed for amorphization of ZrFe{sub 2}, Zr(Cr,Fe){sub 2} and ZrCr{sub 2} by electron irradiation. Results of a recent investigation on amorphization of Zr(Cr,Fe){sub 2} and Zr{sub 2}(Ni,Fe) precipitates in Zircaloy-4 are discussed in context of previous experimental results of neutron and electron irradiations and likely amorphization mechanisms are proposed.

  7. In situ HVEM studies of phase transformation in Zr alloys and compounds under irradiation

    SciTech Connect

    Motta, A.T.; Faldowski, J.A.; Howe, L.M.; Okamoto, P.R.

    1996-01-01

    The High Voltage Electron Microscope (HVEM)/Tandem facility at Argonne National Laboratory has been used to conduct detailed studies of the phase stability and microstructural evolution in zirconium alloys and compounds under ion and electron irradiation. Detailed kinetic studies of the crystalline-to-amorphous transformation of the intermetallic compounds Zr{sub 3}(Fe{sub 1-x}Ni{sub x}), Zr(Fe{sub 1-x},Cr{sub x}){sub 2}, Zr{sub 3}Fe, and Zr{sub 1.5} Nb{sub 1.5} Fe, both as second phase precipitates and in bulk form, have been performed using the in-situ capabilities of the Argonne facility, under a variety of irradiation conditions (temperature, dose rate). Results include a verification of a dose rate effect on amorphization and the influence of material variables (stoichiometry x, presence of stacking faults, crystal structure) on the critical temperature and on the critical dose for amorphization. Studies were also conducted of the microstructural evolution under irradiation of specially tailored binary and ternary model alloys. The stability of the {omega}-phase in Zr-20%Nb under electron and Ar ion irradiation was investigated as well as the {beta}-phase precipitation in Zr-2.5%Nb under Ar ion irradiation. The ensemble of these results is discussed in terms of theoretical models of amorphization and of irradiation-altered solubility.

  8. Tensile impact properties of vanadium-base alloys irradiated at <430{degree}C.

    SciTech Connect

    Chung, H. M.

    1998-05-18

    Tensile and impact properties were investigated at <430 C on V-Cr-Ti, V-Ti-Si, and V-Ti alloys after irradiation to {approx}2-46 dpa at 205-430 C in lithium or helium in the Fast Flux Test Facility (FFTF), High Flux Isotope Reactor (HFIR), Experimental Breeder Reactor II (EBR-II), and Advanced Test Reactor (ATR). A 500-kg heat of V-4Cr-4Ti exhibited high ductile-brittle transition temperature and minimal uniform elongation as a result of irradiation-induced loss of work-hardening capability. Work-hardening capabilities of 30- and 100-kg heats of V-4Cr-4Ti varied significantly with irradiation conditions, although the 30-kg heat exhibited excellent impact properties after irradiation at {approx}390-430 C. The origin of the significant variations in the work-hardening capability of V-4Cr-4Ti is not understood, although fabrication variables, annealing history, and contamination from the irradiation environment are believed to play important roles. A 15-kg heat of V-3Ti-1Si exhibited good work-hardening capability and excellent impact properties after irradiation at {approx}390-430 C. Helium atoms, either charged dynamically or produced via transmutation of boron in the alloys, promote work-hardening capability in V-4Cr-4Ti and V-3Ti-1Si.

  9. The effects of low dose rate irradiation and thermal aging on reactor structural alloys

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Trybus, C. L.; Cole, J. I.

    As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10 -8 dpa/s) irradiation at 380-410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.

  10. 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.

  11. 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).

  12. 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.

  13. 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.

  14. Voids in neutron-irradiated metals and alloys

    SciTech Connect

    Hendricks, R.W.

    1980-01-01

    Small-angle x-ray and neutron scattering are powerful analytical tools for investigating long-range fluctuations in electron (x-rays) or magnetic moment (neutrons) densities in materials. In recent years they have yielded valuable information about voids, void size distributions, and swelling in aluminum, aluminum alloys, copper, molybdenum, nickel, nickel-aluminum, niobium and niobium alloys, stainless steels, graphite and silicon carbide. In the case of aluminum, information concerning the shape of the voids and the ratio of specific surface energies was obtained. The technique of small-angle scattering and its application to the study of voids is reviewed in the paper. Emphasis is placed on the conditions which limit the applicability of the technique, on the interpretation of the data, and on a comparison of the results obtained with companion techniques such as transmission electron microscopy and bulk density. 8 figures, 41 references.

  15. Transmutations of elements under irradiation and its impact on alloys composition

    SciTech Connect

    Gomes, I.C.; Smith, D.L.

    1994-09-01

    This study presents a comparison of nuclear transmutation rates for candidate fusion first wall/blanket structural materials in available fission test reactors with those produced in a typical fusion spectrum. The materials analyzed in this study include a vanadium alloy (V-4Cr-4Ti), a reduced activation martensitic steel (Fe-9Cr-2WVTa), a high conductivity copper alloy (Cu-Cr-Zr), and the SiC compound. The fission irradiation facilities considered include the EBR-II (Experimental Breeder Reactor) fast reactor, and two high flux mixed spectrum reactors, HFIR (High Flux Irradiation Reactor) and SM-3 (Russian reactor). The transmutation and dpa rates that occur in these test reactors are compared with the calculated transmutation and dpa rates characteristic of a D-T fusion first wall spectrum. In general, past work has shown that the displacement damage produced in these fission reactors can be correlated to displacement damage in a fusion spectrum; however, the generation of helium and hydrogen through threshold reactions [(n,x{alpha}) and (n,xp)] are much higher in a fusion spectrum. As shown in this study, the compositional changes for several candidate structural materials exposed to a fast fission reactor spectrum are very low, similar to those for a characteristic fusion spectrum. However, the relatively high thermalized spectrum of a mixed spectrum reactor produces transmutation rates quite different from the ones predicted for a fusion reactor, resulting in substantial differences in the final composition of several candidate alloys after relatively short irradiation time. As examples, the transmutation rates of W, Ta, V, Cu, among others, differ considerably when the irradiation is performed under a mixed spectrum reactor`s and fusion first wall`s spectrum. Fast reactors (EBR-II) provide the only possibility for obtaining high damage rates without producing significant compositional effects in vanadium alloys, ferritic steels and copper alloys.

  16. Microstructure analysis of magnesium alloy melted by laser irradiation

    NASA Astrophysics Data System (ADS)

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

    2005-12-01

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

  17. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-08-01

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

  19. 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.

  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. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Effect of neutron-irradiation on the microstructure of a Fe-12at.%Cr alloy

    NASA Astrophysics Data System (ADS)

    Kuksenko, V.; Pareige, C.; Genevois, C.; Cuvilly, F.; Roussel, M.; Pareige, P.

    2011-08-01

    A nanoscale description of the microstructure in a Fe-12at%Cr model alloy of low purity which has been neutron irradiated at 300°C up to 0.6 dpa, has been performed owing to atom probe tomography (APT). APT investigations have shown that the impurities are also involved in the microstructural evolution under irradiation. Two different populations of clusters have been observed: Cr-enriched and NiSiPCr-enriched clusters. As expected with a process of enhanced precipitation, Cr-enriched clusters are homogeneously distributed inside grains and certainly not correlated to dislocation loops. The NiSiPCr-enriched clusters, which are probably radiation induced segregations, are independent of the Cr-enriched clusters and are also homogeneously distributed. A quantitative description of these objects is presented in this paper and results are compared to SANS data of the literature obtained for the same model alloy.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

  15. 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.

  16. Effects of neutron irradiation on mechanical properties and microstructures of dispersion and precipitation hardened copper alloys

    NASA Astrophysics Data System (ADS)

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

    1996-11-01

    Tensile specimens of Cusbnd Al2O3, CuCrZr and CuNiBe alloys were irradiated with fission neutrons to fluences of 5 × 1022, 5 × 1023and1 × 1024n/m2 (E > 1MeV) at 47°C. Tensile properties and Vickers hardness were determined at 22°C. Microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope and the fractured surfaces were investigated in a scanning electron microscope. The most significant effect of irradiation is a drastic decrease in the ductility of copper alloys at a dose level as low as 0.2 dpa. The loss of ductility appears to be related to the intrinsic hardness of the grain interior and not to the grain boundary embrittlement. It is suggested that the irradiation-induced hardening and the lack of uniform elongation may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) and/or impurity atoms.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. NANOPARTICLES: Formation of the alloy of Au and Ag nanoparticles upon laser irradiation of the mixture of their colloidal solutions

    NASA Astrophysics Data System (ADS)

    Izgaliev, Andrei T.; Simakin, Aleksandr V.; Shafeev, Georgii A.

    2004-01-01

    The formation dynamics of the alloy of gold and silver nanoparticles is studied upon laser irradiation of the mixture of these nanoparticles and factors affecting the alloy formation are determined. Individual nanoparticles are obtained by ablation of the corresponding metals in water or ethanol by copper vapour laser radiation at a wavelength of 510.6 nm close to the maximum of the plasmon resonance of gold particles at 518 nm. The intermediate phase of the alloy characterised by an anomalous red shift of the absorption spectrum is found for the first time. The dependences of the absorption spectrum of the alloy of colloidal particles of these metals and their morphology on the irradiation time are obtained. It is found that the rate of the alloy formation depends on the concentrations of nanoparticles and surfactants in the mixture.

  2. Effect of He+ irradiation on Fe-Cr alloys: Mössbauer-effect study

    NASA Astrophysics Data System (ADS)

    Dubiel, S. M.; Cieślak, J.; Reuther, H.

    2013-03-01

    Effect of He ions irradiation on three model Fe100-xCrx alloys (x = 5.8, 10.75 and 15.15) was investigated with the conversion electron Mössbauer spectroscopy. The study of the alloys irradiated with 25 keV ions revealed that the strongest effect occured in the Fe84.85Cr15.15 sample where an inversion of a short-range-order (SRO) parameter was found. Consequently, the investigation of the influence of the irradiation dose, D, was carried out on the chromium - most concentrated sample showing that the average hyperfine field, , the average angle between the normal to the sample's surface and the magnetization vector, <θ>, as well as the actual distribution of Fe/Cr atoms, as expressed by SRO parameters, strongly depend on D. In particular: (a) increases with D, and its maximum increase corresponds to a decrease of Cr content within the two-shell volume around the probe 57Fe nuclei by ˜2.3 at%, <⊝> decreases by ˜13° at maximum, (c) SRO-parameter averaged over the two-shell volume increases with D from weakly negative value (indicative of Cr atoms ordering) to weakly positive value (indicative of Cr atoms clustering). The inversion takes place at D ≈ 7 dpa.

  3. 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.

  4. Phase stability of decomposed Ni-Al alloys under ion irradiation

    SciTech Connect

    Schmitz, G.; Ewert, J. C.; Harbsmeier, F.; Uhrmacher, M.; Haider, F.

    2001-06-01

    The disordering of a decomposed Ni-12at.%-Al alloy under irradiation with 300-keV Ni{sup +} ions is studied in direct comparison to homogeneous Ni{sub 3}Al at various temperatures. For that, the degree of long-range order is quantitatively determined by electron diffractometry. Heterogeneous alloys disorder significantly faster than the homogeneous intermetallic. In addition, their disordering rate depends on the precipitate size of the microstructure. At 823 K a steady state with an intermediate degree of order is reached. The experimental results are compared with model calculation based on a Cahn-Hilliard-type continuum model. At higher temperatures the model yields a correct description of the experiments. However, discrepancies at room temperature suggest a direct coupling of local composition and degree of order during the collision cascade. This interpretation is further supported by Monte Carlo simulations of overlapping cascade events.

  5. Atomic kinetic Monte Carlo modeling of multi-component Fe dilute alloys under irradiation

    NASA Astrophysics Data System (ADS)

    Domain, C.; Becquart, C. S.

    2014-06-01

    The ageing of pressure vessel steels under radiation has been correlated with the formation of more or less dilute solute clusters which are investigated in this work using a multiscale approach based on ab initio and atomistic kinetic Monte Carlo (AKMC) simulations. The microstructure evolution of Fe alloys is modeled by AKMC on a lattice, using pair interactions adjusted on DFT calculations. Several substitutional elements (Cu, Ni, Mn, Si, P) and foreign interstitials (C, N) are taken into account to describe the alloy. The point defect created by the irradiation, i.e. the vacancies and self interstitials have a tendency to form clusters. The evolution of these clusters is governed by the migration energy of the individual point defects which is very heavy in terms of computing time due to the large number of AKMC steps required. The structure of all the possible objects that can form is complex and some optimized and accelerated methods will be presented.

  6. Microstructure and mechanical properties of neutron irradiated V-20Ti alloy

    NASA Astrophysics Data System (ADS)

    Higashiguchi, Y.; Kayano, H.; Morozumi, S.

    1985-08-01

    Neutron irradiation damage and its effects on mechanical properties were studied for V-20Ti alloy with three different fluences: 5 × 10 24, 2 × 10 24 and 1× 10 23 n m -2 ( E > l MeV) , named A-, B- and C-specimens, respectively. Radiation-induced precipitates (RIP) were observed in A and B specimens and no voids in all specimens. It was clear from the detailed investigation that the precipitates are circular and planar coherent particles lying on (100) planes and have the fee structures. The mechanisms for suppression of void formation and for softening the irradiation effects on the mechanical properties were suggested based on the relation with radiation-induced precipitates (RIP) which were considered as a non-stoichiometric suboxide and contained many vacancies at lattice sites.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Microstructural evolution of Fesbnd 22%Cr model alloy under thermal ageing and ion irradiation conditions studied by atom probe tomography

    NASA Astrophysics Data System (ADS)

    Korchuganova, Olesya A.; Thuvander, Mattias; Aleev, Andrey A.; Rogozhkin, Sergey V.; Boll, Torben; Kulevoy, Timur V.

    2016-08-01

    Nanostructure evolution during ion irradiation of two thermally aged binary Fee22Cr alloys has been investigated using atom probe tomography. Specimens aged at 500 °C for 50 and 200 h were irradiated by 5.6 MeV Fe ions at room temperature up to fluences of 0.3 × 1015 ions/cm2 and 1 × 1015 ions/cm2. The effect of irradiation on the material nanostructure was examined at a depth of 1 μm from the irradiated surface. The analysis of Cr radial concentration functions reveals that dense α‧-phase precipitates in the 200 h aged alloy become diffuse and thereby larger when subjected to irradiation. On the other hand, less Cr-enriched precipitates in the alloy aged for 50 h are less affected. The CreCr pair correlation function analysis shows that matrix inhomogeneity decreases under irradiation. Irradiation leads to a decrease in the number density of diffuse clusters, whereas in the case of well-developed precipitates it remains unchanged.

  12. Investigation of Phase Transformations in High-Alloy Austenitic TRIP Steel Under High Pressure (up to 18 GPa) by In Situ Synchrotron X-ray Diffraction and Scanning Electron Microscopy

    NASA Astrophysics Data System (ADS)

    Ackermann, Stephanie; Martin, Stefan; Schwarz, Marcus R.; Schimpf, Christian; Kulawinski, Dirk; Lathe, Christian; Henkel, Sebastian; Rafaja, David; Biermann, Horst; Weidner, Anja

    2016-01-01

    In order to clarify the difference between the deformation-induced ɛ-martensite ( ɛ 1) and the pressure-induced ɛ-iron ( ɛ 2), high-pressure quasi-hydrostatic experiments were performed on a low-carbon, high-alloy metastable austenitic steel. In situ synchrotron X-ray diffraction measurements as well as post-mortem investigations of the microstructure by electron backscatter diffraction were carried out to study the microstructural transformations. Three processes were observed during compression experiments: first, the formation of deformation-induced hexagonal ɛ 1-martensite, as well as small nuclei of deformation-induced bcc α'-martensite ( α 1') within the fcc γ-matrix due to non-hydrostaticity in the experiments; second, the onset of the phase transformation from the metastable fcc γ-austenite into the hexagonal pressure-induced ɛ 2-iron phase occurred at around 6 GPa; third, during decompression, the hexagonal pressure-induced ɛ 2-iron transformed partially into bcc α'-martensite ( α 2'). Completely different characteristics with regard to habitus as well as to orientation relationships were observed between the pressure-induced phases ( ɛ 2-iron phase and α 2'-martensite) and the deformation-induced martensites ( ɛ 1- and α 1'-martensite).

  13. Improved Austenitic Steels for Power Plant Applications

    SciTech Connect

    Alman, David E.; Dunning, John S.; Schrems, Karol K.; Rawers, James C.; Wilson, Rick D.; Hawk, Jeffrey A.; Petty, Arthur V., Jr.

    2002-08-06

    Using alloy design principles, an austenitic alloy, with base composition of Fe-16Cr-16Ni-2Mn-1Mo (in weight percent, wt%), was formulated to which up to 5 wt% Si and/or Al were added specifically to improve the oxidation resistance. Cyclic oxidation tests were carried out in air at 700 and 800 C for 1000 hours. For comparison, Fe-18Cr-8Ni type-304 stainless steel alloys was also tested. The results showed that at 700 C, all the alloys were twice as oxidation resistant as the type-304 alloy (i.e., the experimental alloys showed weight gains about half that of type-304). Surprisingly, at 800 C, alloys that contained both Al and Si additions were less oxidation resistant than the type-304 alloy. However, alloys containing only Si additions were significantly more oxidation resistant than the type 304 alloys (i.e., showed weight gains 4 times less than the type-304 alloy). Further, alloys with only Si additions pre-oxidized at 800 C, showed zero weight gain in subsequent testing for 1000 hours at 700 C. This implies the potential for producing in-situ protective coating for these alloys. Preliminary exposure tests (1%H2S at 700 C for 360 hrs) indicated that the Si-modified alloys are more sulfidation resistant than type-304 alloy. The mechanical properties of the alloys, modified with carbide forming elements, were also evaluated; and at 600, 700 and 800 C the yield stresses of the carbide modified alloys were twice that of type-304 stainless steel. In this temperature range, the tensile properties of these alloys were comparable to literature values for type-347 stainless steel. It should be emphasized that the microstructures of the carbide forming alloys were not optimized with respect to grain size, carbide size and/or carbide distribution. Also, presented are initial results of vari-strain weld tests used to determine parameters for joining these alloys.

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

    SciTech Connect

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

    2006-06-01

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

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

    NASA Astrophysics Data System (ADS)

    Pannu, Compesh; Bala, Manju; Singh, U. B.; Srivastava, S. K.; Kabiraj, D.; Avasthi, D. K.

    2016-07-01

    AuFe alloy nanoparticles embedded in silica matrix are synthesized using atom beam sputtering technique and subsequently irradiated with 100 MeV Au ions at various fluences ranging from 1 × 1013 to 6 × 1013 ions/cm2. The X-ray diffraction, absorption spectroscopy, X-ray photo electron spectroscopy and transmission electron microscopy results show that swift heavy ion irradiation leads to decomposition of AuFe alloy nanoparticles from surface region and subsequent reprecipitation of Au and Fe nanoparticles occur. The process of phase decomposition and reprecipitation of individual element nanoparticles is explained on the basis of inelastic thermal spike model.

  16. 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.

  17. Ion irradiation studies on the void swelling behavior of a titanium modified D9 alloy

    NASA Astrophysics Data System (ADS)

    Balaji, S.; Mohan, Sruthi; Amirthapandian, S.; Chinnathambi, S.; David, C.; Panigrahi, B. K.

    2015-12-01

    The sensitivity of Positron Annihilation Spectroscopy (PAS) for probing vacancy defects and their environment is well known. Its applicability in determination of swelling and the peak swelling temperature was put to test in our earlier work on ion irradiated D9 alloys [1]. Upon comparison with the peak swelling temperature determined by conventional step height measurements it was found that the peak swelling temperature determined using PAS was 50 K higher. It was conjectured that the positrons trapping in the irradiation induced TiC precipitation could have caused the shift. In the present work, D9 alloys have been implanted with 100 appm helium ions and subsequently implanted with 2.5 MeV Ni ions up to peak damage of 100 dpa. The nickel implantations have been carried out through a range of temperatures between 450 °C and 650 °C. The evolution of cavities and TiC precipitates at various temperatures has been followed by TEM and this report provides an experimental verification of the conjecture.

  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. 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.

  20. 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.

  1. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-09-01

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

  3. Effect of fast neutron irradiation on fatigue-crack growth behavior of three nickel-base alloys

    SciTech Connect

    James, L.A.

    1981-04-01

    Nickel-base alloys are often employed in reactor structural applications where high strength, resistance to corrosion, or swelling resistance are important considerations. In this study, fatigue-crack growth rate tests were conducted at 427/degree/C on Inconel 600, Inconel X-750, and Incoloy 800 irradiated in the Experimental Breeder Reactor II to total fluences ranging between 2.5 and 6.0*10/sup 22/ n/cm/sup 2/. Following irradiation, minor increases were noted in the crack growth rates of Inconel 600, minor decreases in the growth rates of Inconel X-750, and no irradiation effect in the cracking behavior of Incoloy 800. 14 refs.

  4. 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.

  5. 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. PMID:27541350

  6. 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.

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

    PubMed

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

    2016-12-01

    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. PMID:27440080

  8. 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.

  9. Morphological control and evolution of octahedral and truncated trisoctahedral Pt-Au alloy nanocrystals under microwave irradiation

    NASA Astrophysics Data System (ADS)

    Dai, Lei; Zhao, Yanxi; Chi, Quan; Liu, Hanfan; Li, Jinlin; Huang, Tao

    2014-08-01

    Uniform and well-defined truncated trisoctahedral and octahedral Pt-Au alloy nanocrystals were fabricated by co-reducing H2PtCl6-HAuCl4 with tetraethylene glycol (TEG) under microwave irradiation for only 140 s. Iodide ions were critical to the morphological control and evolution of Pt-Au alloy nanostructures. The as-prepared Pt-Au alloy nanocrystals exhibited efficient electrocatalytic activities.Uniform and well-defined truncated trisoctahedral and octahedral Pt-Au alloy nanocrystals were fabricated by co-reducing H2PtCl6-HAuCl4 with tetraethylene glycol (TEG) under microwave irradiation for only 140 s. Iodide ions were critical to the morphological control and evolution of Pt-Au alloy nanostructures. The as-prepared Pt-Au alloy nanocrystals exhibited efficient electrocatalytic activities. Electronic supplementary information (ESI) available: Experimental details; SEM, TEM and HAADF-STEM images, UV-vis absorbance spectra, XRD. See DOI: 10.1039/c4nr01864h

  10. 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.; 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.

  11. 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.

  12. 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.

  13. Stable atomic structure of NiTi austenite

    SciTech Connect

    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.

  14. 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.

  15. Nanocavity formation and hardness increase by dual ion beam irradiation of oxide dispersion strengthened FeCrAl alloy

    NASA Astrophysics Data System (ADS)

    Kögler, R.; Anwand, W.; Richter, A.; Butterling, M.; Ou, Xin; Wagner, A.; Chen, C.-L.

    2012-08-01

    Open volume defects generated by ion implantation into oxide dispersion strengthened (ODS) alloy and the related hardness were investigated by positron annihilation spectroscopy and nanoindentation measurements, respectively. Synchronized dual beam implantation of Fe and He ions was performed at room temperature and at moderately enhanced temperature of 300 °C. For room temperature implantation a significant hardness increase after irradiation is observed which is more distinctive in heat treated than in as-received ODS alloy. There is also a difference between the simultaneous and sequential implantation mode as the hardening effect for the simultaneously implanted ODS alloy is stronger than for sequential implantation. The comparison of hardness profiles and of the corresponding open volume profiles shows a qualitative agreement between the open volume defects generated on the nanoscopic scale and the macroscopic hardness characteristics. Open volume defects are drastically reduced for performing the simultaneous dual beam irradiation at 300 °C which is a more realistic temperature under application aspects. Few remaining defects are clusters of 3-4 vacancies in connection with Y oxide nanoparticles. These defects completely disappear in a shallow layer at the surface. The results are in agreement with hardness measurements showing little hardness increase after irradiation at 300 °C. Suitable characteristics of ODS alloy for nuclear applications and the close correlation between He-related open volume defects and the hardness characteristics are verified.

  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. Accumulation and annealing of radiation defects and the hydrogen effect thereon in an austenitic steel 16Cr15Ni3Mo1Ti upon low-temperature neutron and electron irradiation

    NASA Astrophysics Data System (ADS)

    Arbuzov, V. L.; Gothchitskii, B. N.; Danilov, S. E.; Zaluzhnyi, A. G.; Zuev, Yu. N.; Kar'kin, A. E.; Parkhomenko, V. D.; Sagaradze, V. V.

    2016-01-01

    The effect of hydrogen, accumulation and annealing of radiation defects on the physicomechanical properties of an austenitic Kh16N15M3T1 steel (16Cr15Ni3Mo1Ti) has been investigated upon low-temperature (77 K) neutron and electron irradiations. It has been shown that, when its concentration is about 300 at ppm, hydrogen reduces plasticity by 25%. The presence of helium (2.0-2.5 at ppm) introduced by the tritium-trick method exerts an effect on the yield strength and hardly affects embrittlement. Upon both electron and neutron irradiation, there is a linear relation between the increment of the yield strength and the square root of the increment of the residual electrical resistivity (the concentration of radiation defects). The annealing of vacancies occurs in the neighborhood of 300 K (energy for vacancy migration is 1.0-1.0 eV). Vacancy clusters dissociate near 480 K (energy for dissociation is 1.4-1.5 eV).

  18. Temperature and thermal stress fields during the pulse train of long-pulse laser irradiating aluminium alloy plate

    NASA Astrophysics Data System (ADS)

    Zhang, Wei; Jin, Guangyong; Gu, Xiu-ying

    2014-12-01

    Based on Von Mises yield criterion and elasto-plastic constitutive equations, an axisymmetric finite element model of a Gaussian laser beam irradiating a metal substrate was established. In the model of finite element, the finite difference hybrid algorithm is used to solve the problem of transient temperature field and stress field. Taking nonlinear thermal and mechanical properties into account, transient distributions of temperature field and stress fields generated by the pulse train of long-pulse laser in a piece of aluminium alloy plate were computed by the model. Moreover,distributions as well as histories of temperature and stress fields were obtained. Finite element analysis software COMSOL is used to simulate the Temperature and thermal stress fields during the pulse train of long-pulse laser irradiating 7A04 aluminium alloy plate. By the analysis of the results, it is found that Mises equivalent stress on the target surface distribute within the scope of the center of a certain radius. However, the stress is becoming smaller where far away from the center. Futhermore, the Mises equivalent stress almost does not effect on stress damage while the Mises equivalent stress is far less than the yield strength of aluminum alloy targets. Because of the good thermal conductivity of 7A04 aluminum alloy, thermal diffusion is extremely quick after laser irradiate. As a result, for the multi-pulsed laser, 7A04 aluminum alloy will not produce obvious temperature accumulation when the laser frequency is less than or equal to 10 Hz. The result of this paper provides theoretical foundation not only for research of theories of 7A04 aluminium alloy and its numerical simulation under laser radiation but also for long-pulse laser technology and widening its application scope.

  19. 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.

  20. 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.

  1. 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.

  2. 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 degrees 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, {01 (1) over bar1}< 0 (1) over bar 12 > twinning was identified in the irradiated sample deformed at 300 degrees C.

  3. 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.

  4. Computer simulations of the Ni2MnGa alloys

    NASA Astrophysics Data System (ADS)

    Breczko, Teodor M.; Nelayev, Vladislav; Dovzhik, Krishna; Najbuk, Miroslaw

    2008-07-01

    This article reports an computer simulations of physical properties of Heusler NiMnGa alloy. Computer simulation are devoted to austenite phase. The chemical composition of researched specimens causes generation martesite and austenite phases.

  5. 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.

  6. 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.

  7. 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

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

    NASA Astrophysics Data System (ADS)

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

    1988-07-01

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

  9. 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.

  10. Microchemical effects in irradiated Fe-Cr alloys as revealed by atomistic simulation

    NASA Astrophysics Data System (ADS)

    Malerba, L.; Bonny, G.; Terentyev, D.; Zhurkin, E. E.; Hou, M.; Vörtler, K.; Nordlund, K.

    2013-11-01

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

  11. Phase stability under irradiation in alloys with a positive heat of mixing: Effective thermodynamics description

    SciTech Connect

    Enrique, Raul A.; Bellon, Pascal

    1999-12-01

    The alteration of phase stability due to the continuous production of forced atomic displacements, as is the case under irradiation, is studied. A simple kinetic model of a binary alloy exhibiting phase separation is investigated, and two limiting cases are considered: nearest-neighbor ballistic exchanges and arbitrary-length ballistic exchanges. The model is simultaneously studied by direct kinetic Monte Carlo simulations, and by theoretical approaches based in the description of the steady state by effective thermodynamics. Two theoretical frameworks are considered: a microscopic description based in effective interactions, and a macroscopic description based in effective free energies constructed over a modified Cahn-Hilliard equation. Developments of these models are proposed to allow for quantitative predictions. For nearest-neighbor ballistic exchanges, the steady-state phase diagram is evaluated for each of the approaches and the results are directly compared. In the case of arbitrary-length ballistic exchanges, the appearance of labyrinthine mesoscopic structures at steady state is rationalized in terms of the competition between short-range attractive and long-range repulsive effective interactions. The favorable comparison between the theoretical results and the direct Kinetic Monte Carlo simulations shows that the long-term behavior of this inherently far-from-equilibrium system can be described in terms of effective thermodynamic potentials. (c) 1999 The American Physical Society.

  12. 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; 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 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

  13. Effect of Cr content on the nanostructural evolution of irradiated ferritic/martensitic alloys: An object kinetic Monte Carlo model

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

    Self-interstitial cluster diffusivity in Fe-Cr alloys, model materials for high-Cr ferritic/martensitic steels, is known to be reduced in a non-monotonic way as a function of Cr concentration: it first decreases, then increases. This non-monotonic behaviour is caused by a relatively long-ranged attractive interaction between Cr atoms and crowdions and correlates well with the experimentally observed swelling in these alloys under neutron irradiation, also seen to first decrease and then increase with increasing Cr content, under comparable irradiation conditions. Moreover, recent studies reveal that C atoms dispersed in the Fe matrix form under irradiation complexes with vacancies which, in turn, act as trap for one-dimensionally migrating self-interstitial clusters. The mobility of one-dimensional migrating clusters is considered key to determine swelling susceptibility. However, no model has ever been built that quantitatively describes the dependence of swelling on Cr content, allowing for the presence of C in the matrix. In this work we developed physically-based sets of parameters for object kinetic Monte Carlo (OKMC) simulations intended to study the nanostructure evolution under irradiation in Fe-Cr-C alloys. The nanostructural evolution in Fe-C and in four Fe-Cr-C alloys (containing 2.5, 5, 9 and 12 wt.% Cr) neutron irradiated up to ∼0.6 dpa at 563 K was simulated according to the model and reference experiments were reproduced. Our model shows that the SIA cluster reduced mobility has a major influence on the nanostructural evolution: it increases the number of vacancy-SIA recombinations and thus leads to the suppression of voids formation. This provides a clear framework to interpret the non-monotonic dependence of swelling in Fe-Cr alloys versus Cr content. Our model also suggests that the amount of C in the matrix has an equally important role: high amounts of it may counteract the beneficial effect that Cr has in reducing swelling.

  14. The effect of prolonged irradiation on defect production and ordering in Fe-Cr and Fe-Ni alloys.

    PubMed

    Vörtler, K; Juslin, N; Bonny, G; Malerba, L; Nordlund, K

    2011-09-01

    The understanding of the primary radiation damage in Fe-based alloys is of interest for the use of advanced steels in future fusion and fission reactors. In this work Fe-Cr alloys (with 5, 6.25, 10 and 15% Cr content) and Fe-Ni alloys (with 10, 40, 50 and 75% Ni content) were used as model materials for studying the features of steels from a radiation damage perspective. The effect of prolonged irradiation (neglecting diffusion), i.e. the overlapping of single 5 keV displacement cascade events, was studied by molecular dynamics simulation. Up to 200 single cascades were simulated, randomly induced in sequence in one simulation cell, to study the difference between fcc and bcc lattices, as well as initially ordered and random crystals. With increasing numbers of cascades we observed a saturation of Frenkel pairs in the bcc alloys. In fcc Fe-Ni, in contrast, we saw a continuous accumulation of defects: the growth of stacking-fault tetrahedra and a larger number of self-interstitial atom clusters were seen in contrast to bcc alloys. For all simulations the defect clusters and the short range order parameter were analysed in detail depending on the number of cascades in the crystal. We also report the modification of the repulsive part of the Fe-Ni interaction potential, which was needed to study the non-equilibrium processes. PMID:21846941

  15. Creating poly(ethylene glycol) film on the surface of NiTi alloy by gamma irradiation

    NASA Astrophysics Data System (ADS)

    Yu, Hongyan; Yan, Jin; Ma, Huiling; Zeng, Xinmiao; Liu, Yang; Zhao, Xinqing

    2015-07-01

    NiTi alloy has been extensively utilized as biomaterials owing to its unique shape memory effect, superelasticity and biocompatibility. However, concern with the toxic and allergic responses of nickel potentially releasing from implants stimulated lots of researches of modification on NiTi alloy surface. Creating chemical bond attachment of bioorganic film on NiTi alloy surface could effectively inhibit Ni releasing and obtain bioactive functions for further application. In this work, to get a bioorganic surface, NiTi alloy was modified with poly(ethylene glycol) (PEG) film by gamma ray induced grafting or crosslinking. X-ray diffraction (XRD) spectrum, water contact angle geometer and X-ray photoelectron spectroscopy (XPS) techniques were used to characterize the NiTi surface. The results indicated that PEG was covalent bonded on NiTi alloy surface. Fluorescence microscope (FM) images for morphology of 1 day osteoblast culture on the PEG coated NiTi surface showed that PEG could improve cell proliferation on NiTi surface. Our work offers a way to introduce a bioorganic metal surface by gamma irradiation.

  16. 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

  17. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part 1: Effects of minor alloying elements on precipitate phases in melt products and implication in alloy fabrication

    NASA Astrophysics Data System (ADS)

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

    2000-01-01

    In an effort to develop alloys for fission and fusion reactor applications, 28Fe-15Ni-13Cr base alloys were fabricated by adding various combinations of the minor alloying elements, Mo, Ti, C, Si, P, Nb, and B. The results showed that a significant fraction of undesirable residual oxygen was removed as oxides when Ti, C, and Si were added. Accordingly, the concentrations of the latter three essential alloying elements were reduced also. Among these elements, Ti was the strongest oxide former, but the largest oxygen removal (over 80%) was observed when carbon was added alone without Ti, since gaseous CO boiled off during melting. This paper recommends an alloy melting procedure to mitigate solute losses while reducing the undesirable residual oxygen. In this work, 14 different types of precipitate phases were identified. Compositions of precipitate phases and their crystallographic data are documented. Finally, stability of precipitate phases was examined in view of Gibbs free energy of formation.

  18. 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.

  19. Evaluation of ferritic alloy Fe-2-1/4Cr-1Mo after neutron irradiation - irradiation creep and swelling

    SciTech Connect

    Gelles, D.S.; Puigh, R.J.

    1983-10-01

    Irradiation creep and swelling measurements are reported for Fe-2-1/4Cr-1Mo after irradiation by fast neutrons over the temperature range 390 to 560/sup 0/C. Diameter change measurements on thin walled pressurized tubes in a bainitic condition and density change measurements on rods in a nonstandard condition were made following irradiation in the Experimental Breeder Reactor II. The irradiation creep specimens were irradiated to a fluence of 5.7 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV) or 30 dpa and the swelling specimens were irradiated to a peak fluence of 2.4 x 10/sup 23/ n/cm/sup 2/ or 115 dpa. These results have been used as a basis to establish in-reactor creep and swelling correlations for 2-1/4Cr-1Mo in a bainitic condition. The correlations predict moderate swelling and moderate irradiation enhanced creep at 390/sup 0/C.

  20. Understanding of copper precipitation under electron or ion irradiations in FeCu0.1 wt% ferritic alloy by combination of experiments and modelling

    NASA Astrophysics Data System (ADS)

    Radiguet, B.; Barbu, A.; Pareige, P.

    2007-02-01

    This work is dedicated to the understanding of the basic processes involved in the formation of copper enriched clusters in low alloyed FeCu binary system (FeCu0.1 wt%) under irradiation at temperature close to 300 °C. Such an alloy was irradiated with electrons or with ions (Fe+ or He+) in order to deconvolute the effect of displacement cascades and the associated generation of point defect clusters (ion irradiations), and the super-saturation of mono-vacancies and self-interstitial atoms (electron irradiation). The microstructure of this alloy was characterised by tomographic atom probe. Experimental results were compared with results obtained with cluster dynamic model giving an estimation of the evolution of point defects (free or agglomerated) under irradiation on the one hand and describing homogeneous enhanced precipitation of copper on the other hand. The comparison between the results obtained on the different irradiation conditions and the model suggests that the point defect clusters (dislocation loops and/or nano-voids) created in displacement cascades play a major role in copper clustering in low copper alloy irradiated at 573 K.

  1. 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.

  2. 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.

  3. 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.

  4. Mechanism of instability of carbides in Fe-TaC alloy under high energy electron irradiation at 673 K

    NASA Astrophysics Data System (ADS)

    Abe, Hiroaki; Ishizaki, Takahiro; Kano, Sho; Li, Feng; Satoh, Yuhki; Tanigawa, Hiroyasu; Hamaguchi, Dai; Nagase, Takeshi; Yasuda, Hidehiro

    2014-12-01

    Reduced activation ferritic/martensitic steels (RAFMs), such as F82H steel, are designed to enhance the high-temperature strength by formation of MX-type nanometer-scale precipitates, mainly TaC. However, their instability under irradiation was recently reported. The purpose of this work, therefore, is to clarify the mechanism employing simultaneous observations under electron irradiation at elevated temperature in a high voltage electron microscope. In this work, Fe-0.2 wt.% TaC was fabricated as a model alloy of F82H steel. The instability of the precipitates was observed under electron irradiation at 1 MeV or above. The remarkable shrinkage and disappearance were clearly observed under irradiation with 1.5 MeV and above. On the contrary, the precipitates were mostly stable below 0.75 MeV. Two kinds of mechanism of the irradiation-induced instability were deduced from the electron-energy dependence. One is the dissolution and diffusion of tantalum from precipitates in ferrite matrix. The other is the displacements of tantalum in precipitates that introduce dissolution of Ta into matrix.

  5. 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.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; et al

    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

  6. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys.

    PubMed

    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.

  7. Direct Observation of Defect Range and Evolution in Ion-Irradiated Single Crystalline Ni and Ni Binary Alloys.

    PubMed

    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

  8. 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.

  9. 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

  10. The effect of oxygen on void stability in ion-irradiated steel

    NASA Astrophysics Data System (ADS)

    Seitzman, Larry E.; Dodd, R. Arthur; Kulcinski, Gerald L.

    1990-07-01

    The effect of oxygen on void stability in an Fe-17Ni-13Cr austenitic ternary alloy has been investigated using 15 MeV nickel-ion irradiation at elevated temperatures and preimplantation of 6 MeV oxygen at room temperature. The nickel irradiation was performed over a temperature range of 550 °C to 650 °C. Utilizing transverse specimen preparation techniques, the irradiated steel was examined by transmission electron microscopy (TEM). As little as 10 appm preimplanted oxygen caused a significant increase in the void number density when the steel was irradiated at 550 °C. A near-surface void-denuded zone occurs in the irradiated steel, while a region depleted of visible voids also occurs in the steel injected with 300 appm oxygen or greater and irradiated at 550 °C.

  11. 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.

    2016-09-01

    Cast Fe-Ni-Cr chromia-forming austenitic stainless steels with Ni levels up to 45 wt.% are used at high temperatures in a wide range of industrial applications that demand microstructural stability, corrosion resistance, and creep strength. Although alumina scales offer better corrosion protection at these temperatures, designing cast austenitic alloys that form a stable alumina scale and achieve creep strength comparable to existing cast chromia-forming alloys is challenging. This work outlines the development of cast Fe-Ni-Cr-Al austenitic stainless steels containing about 25 wt.% Ni with good creep strength and the ability to form a protective alumina scale for use at temperatures up to 800-850°C in H2O-, S-, and C-containing environments. Creep properties of the best alloy were comparable to that of HK-type cast chromia-forming alloys along with improved oxidation resistance typical of alumina-forming alloys. Challenges in the design of cast alloys and a potential path to increasing the temperature capability are discussed.

  12. Formation and dissolution of oxide film on zirconium alloys in 288°C pure water under γ-ray irradiation

    NASA Astrophysics Data System (ADS)

    Nishino, Yoshitaka; Endo, Masao; Ibe, Eishi; Yasuda, Takayoshi

    1997-09-01

    Laboratory corrosion tests for zirconium alloys which were based on Zircaloy-2 were performed in 288°C oxygenated pure water for 100 days both with and without 60Co γ-ray irradiation. No nodular oxide was observed. Corrosion weight gains of the alloy which had the lowest nodular corrosion resistance were lower for the irradiated condition than the non-irradiated condition. On the other hand, the alloys which had higher nodular corrosion resistances showed almost the same weight gains for both conditions. Differences of weight gain with and without irradiation were attributed to dissolution of the oxide film in the high temperature water. Dissolution tests of single crystal yttria-doped ZrO 2 indicated that 30-40% larger amounts dissolved into water under the γ-ray irradiation. Low angle incident X-ray diffraction analysis showed that the tetragonal-ZrO 2 fraction in the oxide film was lower with irradiation than without it, especially for the near surface area. The water radiolysis species accelerated the dissolution of the oxide film, especially film on the alloy with lower nodular corrosion resistance. This dissolution led to the lower tetragonal-ZrO 2 fraction and was considered to be one of the factors causing a localized breakdown of the barrier oxide film to make the nodular oxide.

  13. Formation and evolution of intermetallic nanoparticles and vacancy defects under irradiation in Fesbnd Nisbnd Al ageing alloy characterized by resistivity measurements and positron annihilation

    NASA Astrophysics Data System (ADS)

    Druzhkov, A. P.; Danilov, S. E.; Perminov, D. A.; Arbuzov, V. L.

    2016-08-01

    In this paper, we study the effects of intermetallic nanoparticles like Ni3Al on the evolution of vacancy defects in the fcc Fesbnd Nisbnd Al alloy under electron irradiation using positron annihilation spectroscopy. Electrical resistivity measurements have been used as a testing method for characterizing the evolution in the underlying precipitate microstructure due to heat treatment and irradiation. It was shown that the nanosized (∼4.5 nm) intermetallic precipitates homogeneously distributed in the alloy matrix caused a several-fold decrease in the accumulation of vacancies as compared to their accumulation in the pre-quenched alloy. This effect was enhanced with the irradiation temperature. The irradiation-induced growth of intermetallic nanoparticles was also observed in the pre-quenched Fesbnd Nisbnd Al alloy under irradiation at 573 K. Thus, resistivity measurement and positron confinement in ultrafine intermetallic particles, which we revealed earlier, provided the control over the evolution of coherent precipitates, along with vacancy defects, during irradiation and annealing.

  14. The microstructure of rapidly solidified path a prime candidate alloys following irradiation with Fe and He ions

    NASA Astrophysics Data System (ADS)

    Arnberg, L.; Vander Sande, J. B.; Frost, H. J.; Harling, O. K.

    A rapidly solidified,titanium modified, cold worked Type 316 stainless steel has been irradiated simultaneously with Fe and He ions at temperatures between 450 and 650°C. The material shows no void swelling at 450°C even at 50 dpa. At 550°C the swelling is very low below 40 dpa (<0.01%) but increases with fluence to 0.4% at 60 dpa. At 650°C both the cavity number density and size increase to give a swelling of 0.8% at 40 dpa. A conventionally ingot cast alloy with the same composition shows comparable swelling but the void number density in this material is significantly higher. Small (<200 Å) TiC particles which were originally present in the rapidly solidified material are dissolved at fluences above 40 dpa at all temperatures. A silicon and nickel rich phase has precipitated during irradiations at 550 and 650°C.

  15. 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.

  16. 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.

  17. Neutron Absorbing Alloys

    DOEpatents

    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.

  18. 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.

  19. 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.

  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. Laser etching of austenitic stainless steels for micro-structural evaluation

    NASA Astrophysics Data System (ADS)

    Baghra, Chetan; Kumar, Aniruddha; Sathe, D. B.; Bhatt, R. B.; Behere, P. G.; Afzal, Mohd

    2015-06-01

    Etching is a key step in metallography to reveal microstructure of polished specimen under an optical microscope. A conventional technique for producing micro-structural contrast is chemical etching. As an alternate, laser etching is investigated since it does not involve use of corrosive reagents and it can be carried out without any physical contact with sample. Laser induced etching technique will be beneficial especially in nuclear industry where materials, being radioactive in nature, are handled inside a glove box. In this paper, experimental results of pulsed Nd-YAG laser based etching of few austenitic stainless steels such as SS 304, SS 316 LN and SS alloy D9 which are chosen as structural material for fabrication of various components of upcoming Prototype Fast Breeder Reactor (PFBR) at Kalpakkam India were reported. Laser etching was done by irradiating samples using nanosecond pulsed Nd-YAG laser beam which was transported into glass paneled glove box using optics. Experiments were carried out to understand effect of laser beam parameters such as wavelength, fluence, pulse repetition rate and number of exposures required for etching of austenitic stainless steel samples. Laser etching of PFBR fuel tube and plug welded joint was also carried to evaluate base metal grain size, depth of fusion at welded joint and heat affected zone in the base metal. Experimental results demonstrated that pulsed Nd-YAG laser etching is a fast and effortless technique which can be effectively employed for non-contact remote etching of austenitic stainless steels for micro-structural evaluation.

  2. Irradiation effect of nano-bubble dispersion strengthened (N-BDS) alloy

    NASA Astrophysics Data System (ADS)

    Oono, Naoko; Kawano, Ryohei; Shi, Shi; Ukai, Shigeharu; Hayashi, Shigenari; Kondo, Sosuke; Hashitomi, Okinobu; Kimura, Akihiko

    2013-11-01

    Nano-bubble dispersion strengthened (N-BDS) Fe was made from Fe and polymethylmethacrylate (PMMA) powder and irradiated by 6.4 MeV Fe3+ ions to investigate the cavity strengthening and the bubble to void evolution. The bubbles accelerated the irradiation-induced cavity growth. The hardness of the N-BDS Fe was 500 MPa higher than that of unalloyed Fe and the hardness increased by irradiation, while that of unalloyed Fe did not increase. Cavity is probably the origin of the irradiation hardening of N-BDS Fe.

  3. 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.

  4. Defect evolution in a Nisbnd Mosbnd Crsbnd Fe alloy subjected to high-dose Kr ion irradiation at elevated temperature

    NASA Astrophysics Data System (ADS)

    de los Reyes, Massey; Voskoboinikov, Roman; Kirk, Marquis A.; Huang, Hefei; Lumpkin, Greg; Bhattacharyya, Dhriti

    2016-06-01

    A candidate Nisbnd Mosbnd Crsbnd Fe alloy (GH3535) for application as a structural material in a molten salt nuclear reactor was irradiated with 1 MeV Kr2+ ions (723 K, max dose of 100 dpa) at the IVEM-Tandem facility. The evolution of defects like dislocation loops and vacancy- and self-interstitial clusters was examined in-situ. For obtaining a deeper insight into the true nature of these defects, the irradiated sample was further analysed under a TEM post-facto. The results show that there is a range of different types of defects formed under irradiation. Interaction of radiation defects with each other and with pre-existing defects, e.g., linear dislocations, leads to the formation of complex microstructures. Molecular dynamics simulations used to obtain a greater understanding of these defect transformations showed that the interaction between linear dislocations and radiation induced dislocation loops could form faulted structures that explain the fringed contrast of these defects observed in TEM.

  5. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    NASA Astrophysics Data System (ADS)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  6. 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.

  7. 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

    NASA Astrophysics Data System (ADS)

    Long, Fei; Daymond, Mark R.; Yao, Zhongwen; Kirk, Marquis A.

    2015-03-01

    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 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 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, {01 1 ¯ 1 }⟨0 1 ¯ 12 ⟩ twinning was identified in the irradiated sample deformed at 300 °C.

  8. 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.

  9. 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.

  10. 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

  11. Enhancing the absorption of aluminum alloys by irradiation with an excimer laser

    NASA Astrophysics Data System (ADS)

    Scott, Graeme; Williams, Stewart W.; Morgan, P. C.; Dempster, M.

    1994-09-01

    Aluminum alloys typically have as received reflectivities of 85 - 95% at 10.6 micrometers making many laser processes difficult or impossible. These values have been reduced to as low as 1 - 2% by optimizing the processing parameters of an excimer laser used to modify the surface structure of 8090 and 2024 Al alloys and pure Al prior to their exposure to a CO2 laser. The most significant excimer processing parameters were found to be the scan pattern of the excimer beam, the number of pulses per scan pattern step (dwell time) and the laser fluence. Optimizing these parameters allows the production of a rough oxide rich surface and reflectivities at 10.6 micrometers routinely below 10%. Preliminary results are presented from the practical implementation of the technique to a dual wavelength (CO2/excimer) cutting system. Increases in cutting speeds of between 2 - 4 times are demonstrated with 8090 Al-Li alloy using dual wavelength laser processing.

  12. 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.

  13. 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.

  14. 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).

  15. Correlation of radiation-induced changes in mechanical properties and microstructural development of Alloy 718 irradiated with mixed spectra of high-energy protons and spallation neutrons

    NASA Astrophysics Data System (ADS)

    Sencer, B. H.; Bond, G. M.; Garner, F. A.; Hamilton, M. L.; Maloy, S. A.; Sommer, W. F.

    2001-07-01

    Alloy 718 is a γ '(Ni 3(Al,Ti))-γ″(Ni 3Nb) hardenable superalloy with attractive strength, and corrosion resistance. This alloy is a candidate material for use in accelerator production of tritium (APT) target and blanket applications, where it would have to withstand low-temperature irradiation by high-energy protons and spallation neutrons. The existing data base, relevant to such irradiation conditions, is very limited. Alloy 718 has therefore been exposed to a particle flux and spectrum at the Los Alamos Neutron Science Center (LANSCE), closely matching those expected in the APT target and blanket applications. The yield stress of Alloy 718 increases with increasing dose up to ˜0.5 dpa, and then decreases with further increase in dose. The uniform elongation, however, drastically decreases with increasing dose at very low doses (<0.5 dpa), and does not recover when the alloy later softens somewhat. Transmission electron microscopy (TEM) investigation of Alloy 718 shows that superlattice spots corresponding to the age-hardening precipitate phases γ ' and γ″ are lost from the diffraction patterns for Alloy 718 by only 0.6 dpa, the lowest proton-induced dose level achieved in this experiment. Examination of samples that were neutron irradiated to doses of only ˜0.1 dpa showed that precipitates are faintly visible in diffraction patterns but are rapidly becoming invisible. It is proposed that the γ ' and γ″ first become disordered (by <0.6 dpa), but remain as solute-rich aggregates that still contribute to the hardness at relatively low dpa levels, and then are gradually dispersed at higher doses.

  16. 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

  17. Weak-beam imaging of dissociated dislocations in HVEM-irradiated Fe-Ni-Cr alloys

    SciTech Connect

    King, S.L.; Jenkins, M.L.; Kirk, M.A.; English, C.A.

    1992-06-01

    We report here on studies by weak-beam electron microscopy of the evolution of microstructures at and near preexisting line dislocations in a number of Fe-Ni-Cr alloys under electronirradiation in a high-voltage electron microscope (HVEM). The detailed observations are discussed in terms of dislocation climb mechanisms in these materials and a model based on interstitial pipe diffusion.

  18. 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.

  19. Alloy

    NASA Astrophysics Data System (ADS)

    Cabeza, Sandra; Garcés, Gerardo; Pérez, Pablo; Adeva, Paloma

    2014-07-01

    The Mg98.5Gd1Zn0.5 alloy produced by a powder metallurgy route was studied and compared with the same alloy produced by extrusion of ingots. Atomized powders were cold compacted and extruded at 623 K and 673 K (350 °C and 400 °C). The microstructure of extruded materials was characterized by α-Mg grains, and Mg3Gd and 14H-LPSO particles located at grain boundaries. Grain size decreased from 6.8 μm in the extruded ingot, down to 1.6 μm for powders extruded at 623 K (350 °C). Grain refinement resulted in an increase in mechanical properties at room and high temperatures. Moreover, at high temperatures the PM alloy showed superplasticity at high strain rates, with elongations to failure up to 700 pct.

  20. Microstructure evolution and hardness change in ordered Ni3V intermetallic alloy by energetic ion irradiation

    NASA Astrophysics Data System (ADS)

    Hashimoto, A.; Kaneno, Y.; Semboshi, S.; Yoshizaki, H.; Saitoh, Y.; Okamoto, Y.; Iwase, A.

    2014-11-01

    Ni3V bulk intermetallic compounds with ordered D022 structure were irradiated with 16 MeV Au ions at room temperature. The irradiation induced phase transformation was examined by means of the transmission electron microscope (TEM), the extended X-ray absorption fine structure measurement (EXAFS) and the X-ray diffraction (XRD). We also measured the Vickers hardness for unirradiated and irradiated specimens. The TEM observation shows that by the Au irradiation, the lamellar microstructures and the super lattice spot in diffraction pattern for the unirradiated specimen disappeared. This TEM result as well as the result of XRD and EXAFS measurements means that the intrinsic D022 structure of Ni3V changes into the A1 (fcc) structure which is the lattice structure just below the melting point in the thermal equilibrium phase diagram. The lattice structure change from D022 to A1 (fcc) accompanies a remarkable decrease in Vickers microhardness. The change in crystal structure was discussed in terms of the thermal spike and the sequential atomic displacements induced by the energetic heavy ion irradiation.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  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. 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.

  8. 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

  9. Local electronic effects and irradiation resistance in high-entropy alloys

    SciTech Connect

    Egami, Takeshi; Stocks, George Malcolm; Nicholson, Don; Khorgolkhuu, Od; Ojha, Madhusudan

    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.

  10. 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.

  11. 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.

  12. 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.

  13. Effects of composition and helium injection on dislocation loop development in pure FeNiCr alloys under Ni ion irradiation

    NASA Astrophysics Data System (ADS)

    Kimoto, Takayoshi

    1993-08-01

    Pure Fe35Ni7Cr, Fe45Ni7Cr, Fe40Ni13Cr and Fe45Ni15Cr alloys were irradiated by 4MeV Ni 2+ ions at 948 K to doses of about 0.05, 0.3 and 1.0 dpa without helium injection or with simultaneous helium injection. With increasing Ni content and decreasing Cr content, the diameter of radiation-induced dislocation loops increased, and the dose at which the dislocation loops disappeared to develop into dislocation networks decreased. The diameter of dislocation loops induced by Ni 2+ ions irradiation with simultaneous helium injection was larger than that without helium injection for the Fe35Ni7Cr and Fe45Ni7Cr alloys.

  14. Formation of highly corrosion resistant stainless steel surface alloys for marine environments by laser surface alloying

    SciTech Connect

    Sridhar, K.; Deshmukh, M.B.; Khanna, A.S.; Wissenbach, K.

    1998-12-31

    Austenitic stainless steels (SS) such as UNS S30403 are being used for numerous industrial applications due to their goad mechanical properties and weldability. However in aggressive marine environments such as seawater, they suffer from localized corrosion. Even though newly developed highly alloyed SS`s possess very high pitting resistance, they are susceptible to the formation of secondary phases. In the present study, a laser surface alloying technique was employed for the formation of highly alloyed austenitic stainless steel surfaces on conventional 304 SS substrate. Microstructural characterization by optical and SEM revealed finer cells of austenitic phase in the laser alloyed zones with molybdenum contents in the range of 3 to 15 wt%. The pitting corrosion resistance of the surface alloys were ascertained by immersion and potentiodynamic polarization tests and the repassivation behavior by cyclic polarization tests. Also the influence of microstructural features on pitting behavior of highly alloyed and laser surface alloyed steels is studied.

  15. 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.

  16. 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

  17. 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 cluster 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.

  18. Effect of Austenitizing Heat Treatment on the Microstructure and Hardness of Martensitic Stainless Steel AISI 420

    NASA Astrophysics Data System (ADS)

    Barlow, L. D.; Du Toit, M.

    2012-07-01

    The effect of austenitizing on the microstructure and hardness of two martensitic stainless steels was examined with the aim of supplying heat-treatment guidelines to the user that will ensure a martensitic structure with minimal retained austenite, evenly dispersed carbides and a hardness of between 610 and 740 HV (Vickers hardness) after quenching and tempering. The steels examined during the course of this examination conform in composition to medium-carbon AISI 420 martensitic stainless steel, except for the addition of 0.13% vanadium and 0.62% molybdenum to one of the alloys. Steel samples were austenitized at temperatures between 1000 and 1200 °C, followed by oil quenching. The as-quenched microstructures were found to range from almost fully martensitic structures to martensite with up to 35% retained austenite after quenching, with varying amounts of carbides. Optical and scanning electron microscopy was used to characterize the microstructures, and X-ray diffraction was employed to identify the carbide present in the as-quenched structures and to quantify the retained austenite contents. Hardness tests were performed to determine the effect of heat treatment on mechanical properties. As-quenched hardness values ranged from 700 to 270 HV, depending on the amount of retained austenite. Thermodynamic predictions (using the CALPHAD™ model) were employed to explain these microstructures based on the solubility of the carbide particles at various austenitizing temperatures.

  19. 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.

  20. Creep-Resistant, Al2O3- Forming Austenitic Stainless Steels

    SciTech Connect

    Yamamoto, Yukinori; Brady, Michael P; Lu, Zhao Ping; Maziasz, Philip J; Liu, Chain T; Pint, Bruce A; More, Karren Leslie; Meyer III, Harry M; Payzant, E Andrew

    2007-01-01

    A family of inexpensive, Al2O3-forming, high creep strength austenitic stainless steels have been developed. The alloys are based on Fe-20Ni-14Cr-2.5 Al wt.%, with strengthening achieved via nanodispersions of NbC. These alloys offer the potential to significantly increase the operating temperatures of structural components, and can be used under the aggressive oxidizing conditions encountered in energy conversion systems. Protective Al2O3 scale formation was achieved at lower levels of Al in austenitic alloys than previously used, provided that the Ti and V alloying additions frequently used for strengthening were eliminated. The lower levels of Al permitted stabilization of the austenitic matrix structure, and made it possible to obtain excellent creep resistance. Creep rupture lifetime in excess of 2000 h at 750 aC and 100 MPa in air, and resistance to oxidation in air + 10% water vapor environments at 650 and 800 aC are demonstrated

  1. Interface alloying due to Kr-irradiation on Ni/Si system

    NASA Astrophysics Data System (ADS)

    Datta, D.; Bhattacharyya, S. R.

    2007-03-01

    This study deals with an investigation of silicide formation at the interface of Ni thin film and Si substrate induced by 300 keV 84Kr2+ ions at a number of doses under ambient condition. The implanted samples were subsequently annealed using a rapid thermal annealing system (RTA) in vacuum environment. The modification of the interface and surface properties has been analyzed using various methods like scanning electron microscope with energy dispersive X-ray analysis (SEM/EDAX), X-ray diffraction (XRD), etc. The energy dispersive X-ray spectra (EDS) show a decrease of Ni content in the samples with increasing dose, which indicates the sputter erosion due to ion irradiation. The SEM micrographs show no considerable surface modification under ion bombardment but the XRD spectra clearly reveal the formation of different phases of Ni2Si and NiSi at lower doses which transforms to NiSi as dose increases.

  2. 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.

  3. 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.

  4. Irradiation effects in ferritic steels

    NASA Astrophysics Data System (ADS)

    Lechtenberg, Thomas

    1985-08-01

    Since 1979 the Alloy Development for Irradiation Performance (ADIP) task funded by the US Department of Energy has been studying the 2-12Cr class of ferritic steels to establish the feasibility of using them in fusion reactor first wall/breeding blanket (FW/B) applications. The advantages of ferritic steels include superior swelling resistance, low thermal stresses compared to austenitic stainless steels, attractive mechanical properties up to 600°C. and service histories exceeding 100 000 h. These steels are commonly used in a range of microstructural conditions which include ferritic, martensitic. tempered martensitic, bainitic etc. Throughout this paper where the term "ferritic" is used it should be taken to mean any of these microstructures. The ADIP task is studying several candidate alloy systems including 12Cr-1MoWV (HT-9), modified 9Cr-1MoVNb, and dual-phased steels such as EM-12 and 2 {1}/{4}Cr-Mo. These materials are ferromagnetic (FM), body centered cubic (bcc), and contain chromium additions between 2 and 12 wt% and molybdenum additions usually below 2%. The perceived issues associated with the application of this class of steel to fusion reactors are the increase in the ductile-brittle transition temperature (DBTT) with neutron damage, the compatibility of these steels with liquid metals and solid breeding materials, and their weldability. The ferromagnetic character of these steels can also be important in reactor design. It is the purpose of this paper to review the current understanding of these bcc steels and the effects of irradiation. The major points of discussion will be irradiation-induced or -enhanced dimensional changes such as swelling and creep, mechanical properties such as tensile strength and various measurements of toughness, and activation by neutron interactions with structural materials.

  5. Effect of manganese and nitrogen on the solidification mode in austenitic stainless steel welds

    NASA Astrophysics Data System (ADS)

    Suutala, N.

    1982-12-01

    The macrostructures and microstructures of thirty different austenitic stainless welds alloyed with manganese and Jor nitrogen are analyzed. Comparison of the results with those obtained from normal welds of the AISIJAWS 300 series indicates that the solidification mode and Ferrite Number can be predicted adequately using chromium and nickel equivalents. The solidification mode in the normal and nitrogen-alloyed welds can be best described by the equivalents developed by Hammar and Svensson and the Ferrite Number by the conventional Schaeffler-DeLong diagram. Both of these descriptions are invalid at high manganese content values (5 to 8 pct), however, in which case Hull’s equivalents give a better correlation between the composition and the solidification mode or Ferrite Number. The complicated role of manganese and the austenite-favoring effect of nitrogen in austenitic stainless steels are discussed.

  6. An On-Heating Dilation Conversional Model for Austenite Formation in Hypoeutectoid Steels

    NASA Astrophysics Data System (ADS)

    Lee, Seok-Jae; Clarke, Kester D.; van Tyne, Chester J.

    2010-09-01

    Dilatometry is often used to study solid-state phase transformations. While most steel transformation studies focus on the decomposition of austenite, this article presents an on-heating dilation conversional model to determine phase fraction based on measured volume changes during the formation of austenite in ferrite-pearlite hypoeutectoid steels. The effect of alloying elements on the transformation strain is incorporated into the model. Comparison of the conversional model predictions to measured transformation temperature ( A c3) shows excellent agreement. The pearlite decomposition finish temperature ( A pf ) predicted by the conversional model more closely matches experimental results when compared to standard lever rule calculations. Results show that including the effects of substitutional alloying elements (in addition to carbon) improves phase fraction predictions. The conversional model can be used to quantitatively predict intercritical austenite fraction with application to modeling, induction heating, intercritical annealing, and more complex heat treatments for hypoeutectoid steels.

  7. Enhancing catalytic performance of palladium in gold and palladium alloy nanoparticles for organic synthesis reactions through visible light irradiation at ambient temperatures.

    PubMed

    Sarina, Sarina; Zhu, Huaiyong; Jaatinen, Esa; Xiao, Qi; Liu, Hongwei; Jia, Jianfeng; Chen, Chao; Zhao, Jian

    2013-04-17

    The intrinsic catalytic activity of palladium (Pd) is significantly enhanced in gold (Au)-Pd alloy nanoparticles (NPs) under visible light irradiation at ambient temperatures. The alloy NPs strongly absorb light and efficiently enhance the conversion of several reactions, including Suzuki-Miyaura cross coupling, oxidative addition of benzylamine, selective oxidation of aromatic alcohols to corresponding aldehydes and ketones, and phenol oxidation. The Au/Pd molar ratio of the alloy NPs has an important impact on performance of the catalysts since it determines both the electronic heterogeneity and the distribution of Pd sites at the NP surface, with these two factors playing key roles in the catalytic activity. Irradiating with light produces an even more profound enhancement in the catalytic performance of the NPs. For example, the best conversion rate achieved thermally at 30 °C for Suzuki-Miyaura cross coupling was 37% at a Au/Pd ratio of 1:1.86, while under light illumination the yield increased to 96% under the same conditions. The catalytic activity of the alloy NPs depends on the intensity and wavelength of incident light. Light absorption due to the Localized Surface Plasmon Resonance of gold nanocrystals plays an important role in enhancing catalyst performance. We believe that the conduction electrons of the NPs gain the light absorbed energy producing energetic electrons at the surface Pd sites, which enhances the sites' intrinsic catalytic ability. These findings provide useful guidelines for designing efficient catalysts composed of alloys of a plasmonic metal and a catalytically active transition metal for various organic syntheses driven by sunlight.

  8. Enhancing catalytic performance of palladium in gold and palladium alloy nanoparticles for organic synthesis reactions through visible light irradiation at ambient temperatures.

    PubMed

    Sarina, Sarina; Zhu, Huaiyong; Jaatinen, Esa; Xiao, Qi; Liu, Hongwei; Jia, Jianfeng; Chen, Chao; Zhao, Jian

    2013-04-17

    The intrinsic catalytic activity of palladium (Pd) is significantly enhanced in gold (Au)-Pd alloy nanoparticles (NPs) under visible light irradiation at ambient temperatures. The alloy NPs strongly absorb light and efficiently enhance the conversion of several reactions, including Suzuki-Miyaura cross coupling, oxidative addition of benzylamine, selective oxidation of aromatic alcohols to corresponding aldehydes and ketones, and phenol oxidation. The Au/Pd molar ratio of the alloy NPs has an important impact on performance of the catalysts since it determines both the electronic heterogeneity and the distribution of Pd sites at the NP surface, with these two factors playing key roles in the catalytic activity. Irradiating with light produces an even more profound enhancement in the catalytic performance of the NPs. For example, the best conversion rate achieved thermally at 30 °C for Suzuki-Miyaura cross coupling was 37% at a Au/Pd ratio of 1:1.86, while under light illumination the yield increased to 96% under the same conditions. The catalytic activity of the alloy NPs depends on the intensity and wavelength of incident light. Light absorption due to the Localized Surface Plasmon Resonance of gold nanocrystals plays an important role in enhancing catalyst performance. We believe that the conduction electrons of the NPs gain the light absorbed energy producing energetic electrons at the surface Pd sites, which enhances the sites' intrinsic catalytic ability. These findings provide useful guidelines for designing efficient catalysts composed of alloys of a plasmonic metal and a catalytically active transition metal for various organic syntheses driven by sunlight. PMID:23566035

  9. Release of helium from irradiation damage in Fe?9Cr ferritic alloy

    NASA Astrophysics Data System (ADS)

    Ono, K.; Arakawa, K.; Shibasaki, H.; Kurata, H.; Nakamichi, I.; Yoshida, N.

    2004-08-01

    Thermal desorption of helium from Fe-9Cr and pure Fe which were irradiated with 5 keV He + ions at 85 or 473 K have been studied by thermal desorption spectrometry (TDS) at temperatures from room temperature to 1270 K and related microstructure changes by TEM. It is found that five TDS peaks appear in Fe-9Cr specimen and these peaks are well correlated with microstructure changes. From these results, the five peaks I Cr, II Cr, III Cr, IV Cr and V Cr may be attributed to release of trapped helium atoms by the break-up of vacancy-helium-self-interstitial atom complexes or very tiny loops, glide to the specimen surface or coalescence of interstitial type dislocation loops, shrinkage of dislocation loops, bubble migration to the specimen surface and α-γ phase transformation, respectively. Precipitation of Cr around the loops and bubbles in Fe-9Cr is revealed by STEM-EELS and this could shift the related TDS peaks to higher temperatures.

  10. 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.

  11. 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.

  12. 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.

  13. Formation and evolution of MnNi clusters in neutron irradiated dilute Fe alloys modelled by a first principle-based AKMC method

    NASA Astrophysics Data System (ADS)

    Ngayam-Happy, R.; Becquart, C. S.; Domain, C.; Malerba, L.

    2012-07-01

    An atomistic Monte Carlo model parameterised on electronic structure calculations data has been used to study the formation and evolution under irradiation of solute clusters in Fe-MnNi ternary and Fe-CuMnNi quaternary alloys. Two populations of solute rich clusters have been observed, which can be discriminated by whether or not the solute atoms are associated with self-interstitial clusters. Mn-Ni-rich clusters are observed at a very early stage of the irradiation in both modelled alloys, whereas the quaternary alloys contain also Cu-containing clusters. Mn-Ni-rich clusters nucleate very early via a self-interstitial-driven mechanism, earlier than Cu-rich clusters; the latter, however, which are likely to form via a vacancy-driven mechanism, grow in number much faster than the former, helped by the thermodynamic driving force to Cu precipitation in Fe, thereby becoming dominant in the low dose regime. The kinetics of the number density increase of the two populations is thus significantly different. Finally the main conclusion suggested by this work is that the so-called late blooming phases might as well be neither late, nor phases.

  14. 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.

  15. Characterisation of Cr, Si and P distribution at dislocations and grain-boundaries in neutron irradiated Fe-Cr model alloys of low purity

    NASA Astrophysics Data System (ADS)

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

    2013-03-01

    Segregations at some dislocations and grain boundaries in Fe-5%Cr, Fe-9%Cr and Fe-12%Cr model alloys of low purity after neutron irradiation at 300 °C up to 0.6 dpa have been analyzed with atom probe tomography. All dislocation lines and low- and high-angle grain boundaries (GBs) which have been observed were enriched with Cr, Si and P. The segregations reveal the different dislocation structures associated to different type of analysed GBs. Cr and Si atoms were found to be nonhomogenously distributed around the dislocation cores because of the non isotropic stress field induced by edge dislocation lines. Concerning GBs, precipitate free zones (PFZs) are exhibited around the planar defects which were analysed in Fe-9%Cr and Fe-12%Cr model alloys. These PFZ are size dependant with the nominal level of Cr.

  16. 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.

  17. Microstructural development in waste form alloys cast from irradiated cladding residual from the electrometallurgical treatment of EBR-II spent fuel.

    SciTech Connect

    Keiser, D. D., Jr.

    1999-06-10

    A metallic waste form alloy that consists primarily of stainless steel and zirconium is being developed by Argonne National Laboratory to contain metallic waste constituents that are residual from an electrometallurgical treatment process for spent nuclear fuel. Ingots have been cast in an induction furnace in a hot cell using actual, leftover, irradiated, EBR-II cladding hulls treated in an electrorefiner. The as-cast ingots have been sampled using a core-drilling and an injection-casting technique. In turn, generated samples have been characterized using chemical analysis techniques and a scanning electron microscope equipped with energy dispersive and wavelength-dispersive spectrometers. As-cast ingots contain the predicted concentration levels of the various constituents, and most of the phases that develop are analogous to those for alloys generated using non-radioactive surrogates for the various fission products.

  18. 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. PMID:25175259

  19. 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.

  20. 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.

  1. 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.

  2. Microstructures of laser deposited 304L austenitic stainless steel

    SciTech Connect

    BROOKS,JOHN A.; HEADLEY,THOMAS J.; ROBINO,CHARLES V.

    2000-05-22

    Laser deposits fabricated from two different compositions of 304L stainless steel powder were characterized to determine the nature of the solidification and solid state transformations. One of the goals of this work was to determine to what extent novel microstructure consisting of single-phase austenite could be achieved with the thermal conditions of the LENS [Laser Engineered Net Shape] process. Although ferrite-free deposits were not obtained, structures with very low ferrite content were achieved. It appeared that, with slight changes in alloy composition, this goal could be met via two different solidification and transformation mechanisms.

  3. 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.

  4. 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.

  5. The effects of silicon and titanium on void swelling and phase transformations in neutron irradiated 12Cr-15Ni steels

    NASA Astrophysics Data System (ADS)

    Boothby, R. M.; Williams, T. M.

    1988-05-01

    12Cr-15Ni-0.25Ti steels with Si additions of 0.5, 0.9 and 1.4 wt% have been irradiated to a maximum dose of 47 dpa at temperatures ranging from 399 to 649°C. Detailed microstructural examinations of void swelling, precipitation behaviour and austenite instability have been made. Assessments of swelling and matrix phase transformations have also been made using density and induced magnetization measurements respectively. Austenite instability was increased by Si additions; the transformation product was usually ferrite although some martensite was also observed, and compositional fluctuations in untransformed austenite were detected. Precipitation, particularly of G phase, became more extensive and swelling in solution-treated alloys was reduced at higher Si contents. Enhanced growth of voids attached to G phase precipitates was observed. Cold-working decreased both swelling and ferrite formation. A fine dispersion of TiC was effective in suppressing swelling at high irradiation temperature as long as the precipitates remained stable. The stability of TiC was increased by cold-working but reduced by Si additions.

  6. Ballistic effects on the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-Cu alloy.

    PubMed

    Xu, Donghua; Certain, Alicia; Lee Voigt, Hyon-Jee; Allen, Todd; Wirth, Brian D

    2016-09-14

    Studies of solute precipitation and precipitate phase stability in nuclear structural materials under concurrent irradiation and heat often lead to contradictory results due to the complex nature of the phenomena which is far from well understood. Here, we present a comprehensive atomistically based continuum model for the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-0.78 at. % Cu alloy. Our model incorporates thermal and irradiation enhanced diffusion of atomic Cu, clustering of Cu into sub-nanometer and nanometer sized precipitates, thermal dissociation of the precipitates and, in particular, a cascade re-dissolution parameter that has been made available by recent molecular dynamics simulations. Our model suggests that the Cu precipitates may form, re-dissolve, or coarsen under different irradiation and thermal conditions depending on the competition between the thermal and the ballistic effects. The quantitative predictions of our model are compared with available experiments including limited atom probe tomography data acquired in this study. The work highlights the importance of combining thermal and ballistic effects in the understanding of phase stability in extreme nuclear environments. PMID:27634272

  7. Ballistic effects on the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-Cu alloy

    NASA Astrophysics Data System (ADS)

    Xu, Donghua; Certain, Alicia; Lee Voigt, Hyon-Jee; Allen, Todd; Wirth, Brian D.

    2016-09-01

    Studies of solute precipitation and precipitate phase stability in nuclear structural materials under concurrent irradiation and heat often lead to contradictory results due to the complex nature of the phenomena which is far from well understood. Here, we present a comprehensive atomistically based continuum model for the copper precipitation and re-dissolution kinetics in an ion irradiated and thermally annealed Fe-0.78 at. % Cu alloy. Our model incorporates thermal and irradiation enhanced diffusion of atomic Cu, clustering of Cu into sub-nanometer and nanometer sized precipitates, thermal dissociation of the precipitates and, in particular, a cascade re-dissolution parameter that has been made available by recent molecular dynamics simulations. Our model suggests that the Cu precipitates may form, re-dissolve, or coarsen under different irradiation and thermal conditions depending on the competition between the thermal and the ballistic effects. The quantitative predictions of our model are compared with available experiments including limited atom probe tomography data acquired in this study. The work highlights the importance of combining thermal and ballistic effects in the understanding of phase stability in extreme nuclear environments.

  8. Corrosion of austenitic and ferritic-martensitic steels exposed to supercritical carbon dioxide

    SciTech Connect

    Tan, Lizhen; Anderson, Mark; Taylor, D; Allen, Todd R.

    2011-01-01

    Supercritical carbon dioxide (S-CO{sub 2}) is a potential coolant for advanced nuclear reactors. The corrosion behavior of austenitic steels (alloys 800H and AL-6XN) and ferritic-martensitic (FM) steels (F91 and HCM12A) exposed to S-CO{sub 2} at 650 C and 20.7 MPa is presented in this work. Oxidation was identified as the primary corrosion phenomenon. Alloy 800H had oxidation resistance superior to AL-6XN. The FM steels were less corrosion resistant than the austenitic steels, which developed thick oxide scales that tended to exfoliate. Detailed microstructure characterization suggests the effect of alloying elements such as Al, Mo, Cr, and Ni on the oxidation of the steels.

  9. Computational design of precipitation strengthened austenitic heat-resistant steels

    NASA Astrophysics Data System (ADS)

    Lu, Qi; Xu, Wei; van der Zwaag, Sybrand

    2013-09-01

    A new genetic alloy design approach based on thermodynamic and kinetic principles is presented to calculate the optimal composition of MX carbonitrides precipitation strengthened austenitic heat-resistant steels. Taking the coarsening of the MX carbonitrides as the process controlling the life time for steels in high temperature use, the high temperature strength is calculated as a function of steel chemistry, service temperature and time. New steel compositions for different service conditions are found yielding optimal combinations of strength and stability of the strengthening precipitation for specific applications such as fire-resistant steels (short-time property guarantee) and creep-resistant steels (long-time property guarantee). Using the same modelling approach, the high temperature strength and lifetime of existing commercial austenitic creep-resistant steels were also calculated and a good qualitative agreement with reported experimental results was obtained. According to the evaluation parameter employed, the newly defined steel compositions may have higher and more stable precipitation strengthening factors than existing high-temperature precipitate-strengthened austenite steels.

  10. Thermodynamic Calculation Study on Effect of Manganese on Stability of Austenite in High Nitrogen Stainless Steels

    NASA Astrophysics Data System (ADS)

    Wang, Qingchuan; Zhang, Bingchun; Yang, Ke

    2016-07-01

    A series of high nitrogen steels were studied by using thermodynamic calculations to investigate the effect of manganese on the stability of austenite. Surprisingly, it was found that the austenite stabilizing ability of manganese was strongly weakened by chromium, but it was strengthened by molybdenum. In addition, with an increase of manganese content, the ferrite stabilizing ability of chromium significantly increased, but that of molybdenum decreased. Therefore, strong interactions exist between manganese and the other alloying elements, which should be the main reason for the difference among different constituent diagrams.

  11. General and Localized corrosion of Austenitic and Borated Stainless Steels in Simulated Concentrated Ground Waters

    SciTech Connect

    D. Fix; J. Estill; L. Wong; R. Rebak

    2004-05-28

    Boron containing stainless steels are used in the nuclear industry for applications such as spent fuel storage, control rods and shielding. It was of interest to compare the corrosion resistance of three borated stainless steels with standard austenitic alloy materials such as type 304 and 316 stainless steels. Tests were conducted in three simulated concentrated ground waters at 90 C. Results show that the borated stainless were less resistant to corrosion than the witness austenitic materials. An acidic concentrated ground water was more aggressive than an alkaline concentrated ground water.

  12. General and Localized Corrosion of Austenitic And Borated Stainless Steels in Simulated Concentrated Ground Waters

    SciTech Connect

    Estill, J C; Rebak, R B; Fix, D V; Wong, L L

    2004-03-11

    Boron containing stainless steels are used in the nuclear industry for applications such as spent fuel storage, control rods and shielding. It was of interest to compare the corrosion resistance of three borated stainless steels with standard austenitic alloy materials such as type 304 and 316 stainless steels. Tests were conducted in three simulated concentrated ground waters at 90 C. Results show that the borated stainless were less resistant to corrosion than the witness austenitic materials. An acidic concentrated ground water was more aggressive than an alkaline concentrated ground water.

  13. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part II: Effects of minor elements on precipitate phase stability during thermal aging

    NASA Astrophysics Data System (ADS)

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

    2000-01-01

    The precipitate phase stability in Fe-15Ni-13Cr base austenitic alloys was investigated as a function of minor alloying additions after thermally aging at 600°C and 675°C for times ranging from 24 h to one year. Seven major precipitate phases were found in aged specimens, including M 23C 6, Laves, Eta (η), TiO, NbC, MC, and M 2P. The types and amounts of precipitate phases varied with alloying element additions, aging temperature, and aging time. By analyzing the composition of each individual particle, it was possible to determine the essential constituent elements for each phase. From this information, a strategy to promote or suppress certain precipitate phases was developed. Among the seven phases, the most desirable precipitate phases were considered to be MC and M 2P, because these particles form on a fine scale with a high number density and, therefore, can serve as effective gas atom trap sites under irradiation.

  14. 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.

  15. An approach to prior austenite reconstruction

    SciTech Connect

    Abbasi, Majid; Nelson, Tracy W.; Sorensen, Carl D.; Wei Lingyun

    2012-04-15

    One area of interest in Friction Stir Welding (FSW) of steels is to understand microstructural evolution during the process. Most of the deformation occurs in the austenite temperature range. Quantitative microstructural measurements of prior austenite microstructure are needed in order to understand evolution of the microstructure. Considering the fact that room temperature microstructure in ferritic steels contains very little to no retained austenite, prior austenite microstructure needs to be recovered from the room temperature ferrite. In this paper, an approach based on Electron Backscattered Diffraction (EBSD) is introduced to detect Bain zones. Bain zone detection is used to reconstruct prior austenite grain structure. Additionally, a separate approach based on phase transformation orientation relationships is introduced in order to recover prior austenite orientation. - Highlights: Black-Right-Pointing-Pointer This approach provides a tool to reconstruct large-scale austenite microstructures. Black-Right-Pointing-Pointer It recovers prior austenite orientation without relying on retained austenite. Black-Right-Pointing-Pointer It utilizes EBSD data from the room temperature microstructure. Black-Right-Pointing-Pointer Higher number of active variants leads to more accurate reconstructions. Black-Right-Pointing-Pointer At least two variants are needed in order to recover prior austenite orientation.

  16. COMPOSITIONAL AND TEMPERATURE DEPENDENCE OF VOID SWELLING IN MODEL Fe-Cr BASE ALLOYS IRRADIATED IN THE EBR-II FAST REACTOR

    SciTech Connect

    Sencer, Bulent H.; Garner, Francis A.

    2001-04-01

    A series of annealed and aged Fe-xCr, Fe-12Cr-yC and Fe-12Cr-0.1C-zMo model alloys were irradiated in EBR-II at eight temperatures between 400 and 650C and dose levels ranging from 35 to 131 dpa. Swelling-induced density changes observed in the binary alloys generally peaked at mid-chromium levels, with the chromium and temperature dependence expressed primarily in the duration of the transient regime. The steady state swelling rate at the lower irradiation temperatures was much higher than previous estimates, reaching ~0.2%/dpa and possibly still climbing at higher neutron exposures. The dependence of swelling on molybdenum and carbon was more complex, depending on whether the temperature was relatively low or high. At temperatures of 482oC and above the effect of carbon additions was very pronounced, with swelling of Fe-12Cr jumping dramatically from near zero at 0.002%C to 6-10% at 0.1%C. This indicates that the major determinant of the composition and temperature dependence probably lies in the duration of the nucleation-dominated transient regime of swelling and not primarily in the steady-state swelling rate as previously envisioned. This raises the possibility that significant swelling may occur earlier in fusion and spallation neutron spectra where high gas generation rates may assist void nucleation.

  17. 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.; 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,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

  18. 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

  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. Effects of Solute Nb Atoms and Nb Precipitates on Isothermal Transformation Kinetics from Austenite to Ferrite

    NASA Astrophysics Data System (ADS)

    Wang, Li; Parker, Sally; Rose, Andrew; West, Geoff; Thomson, Rachel

    2016-07-01

    Nb is a very important micro-alloying element in low-carbon steels, for grain size refinement and precipitation strengthening, and even a low content of Nb can result in a significant effect on phase transformation kinetics from austenite to ferrite. Solute Nb atoms and Nb precipitates may have different effects on transformation behaviors, and these effects have not yet been fully characterized. This paper examines in detail the effects of solute Nb atoms and Nb precipitates on isothermal transformation kinetics from austenite to ferrite. The mechanisms of the effects have been analyzed using various microscopy techniques. Many solute Nb atoms were found to be segregated at the austenite/ferrite interface and apply a solute drag effect. It has been found that solute Nb atoms have a retardation effect on ferrite nucleation rate and ferrite grain growth rate. The particle pinning effect caused by Nb precipitates is much weaker than the solute drag effect.

  1. Shape memory alloy thaw sensors

    DOEpatents

    Shahinpoor, M.; Martinez, D.R.

    1998-04-07

    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. 16 figs.

  2. 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.

  3. 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.

  4. Effect of electron irradiation on the 3C-4H transformation in a Co-Fe alloy

    SciTech Connect

    Allen, C.W.; Mori, H.

    1996-03-01

    Preliminary results are presented of a study of electron irradiation on the martensitic transformation in a relatively dilute terminal solid solution of Co+5.75 wt% Fe. The experiments demonstrate qualitatively the profound effect that the damage structure produced by electron irradiation can have on these simple martensitic transformations. Further study is being pursued to attempt to quantify the transformation retardation effect.

  5. 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.

  6. 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.

  7. Experimental investigation of stress effect on swelling and microstructure of Fe-16Cr-15Ni-3Mo-Nb austenitic stainless steel under low-temperature irradiation up to high damage dose in the BOR-60 reactor

    NASA Astrophysics Data System (ADS)

    Neustroev, V. S.; Ostrovsky, Z. E.; Shamardin, V. K.

    2004-08-01

    The present paper was devoted to investigation of the stress effect on swelling and microstructure evolution of the Fe-15.8Cr-15.3Ni-2.8Mo-0.6Nb steel irradiated in the BOR-60 reactor at temperatures from 395 to 410 °C and damage doses from 79 to 98 dpa. Was found out that the stress increase leads to an increase of swelling, that can be associated with a decrease in incubation period with a practically constant swelling rate. Voids concentration increases at the first stage of irradiation when the void sizes are practically constant, and then the concentration reaches some saturation and swelling increase is caused by void growth.

  8. Duct and cladding alloy

    DOEpatents

    Korenko, Michael K.

    1983-01-01

    An austenitic alloy having good thermal stability and resistance to sodium corrosion at 700.degree. C. consists essentially of 35-45% nickel 7.5-14% chromium 0.8-3.2% molybdenum 0.3-1.0% silicon 0.2-1.0% manganese 0-0.1% zirconium 2.0-3.5% titanium 1.0-2.0% aluminum 0.02-0.1% carbon 0-0.01% boron and the balance iron.

  9. 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.

    2016-09-01

    Microstructures of a ductile iron alloy at different solidification stages were captured in quenching experiments. Etched microstructures showed that spheroidal graphite particles and austenite dendrites nucleated independently to a significant extent. Growth of the austenite dendrite engulfed the spheroidal graphite particles after first contacting the nodule and then by forming an austenite shell around the spheroidal graphite particle. Statistical analysis of the graphite size distribution was used to determine the nodule diameter when the austenite shell was completed. In addition, multiple graphite nucleation events were discerned from the graphite particle distributions. Majority of graphite growth occurred when the graphite was in contact with the austenite. Circumferential growth of curved graphene layers appeared as faceted growth fronts sweeping around the entire surface of a spheroidal graphite particle which was at the early growth stage. Mismatches between competing graphene growth fronts created gaps, which divided the spheroidal graphite particle into radially oriented conical substructures. Graphene layers continued growing in each conical substructure to further extend the size of the spheroidal graphite particle.

  10. Effect of bainite transformation and retained austenite on mechanical properties of austempered spheroidal graphite cast steel

    NASA Astrophysics Data System (ADS)

    Takahashi, Toshio; Abe, Toshihiko; Tada, Shuji

    1996-06-01

    Austempered ductile iron (ADI) has excellent mechanical properties, but its Young's modulus is low. Austempered spheroidal graphite cast steel (AGS) has been developed in order to obtain a new material with superior mechanical properties to ADI. Its carbon content (approximately 1.0 pct) is almost one-third that of a standard ADI; thus, the volume of graphite is also less. Young's modulus of AGS is 195 to 200 GPa and is comparable to that of steel. Austempered spheroidal graphite cast steel has an approximately 200 MPa higher tensile strength than ADI and twice the Charpy absorbed energy of ADI. The impact properties and the elongation are enhanced with increasing volume fraction of carbon-enriched retained austenite. At the austempering temperature of 650 K, the volume fraction of austenite is approximately 40 pct for 120 minutes in the 2.4 pct Si alloy, although it decreases rapidly in the 1.4 pct Si alloy. The X-ray diffraction analysis shows that appropriate quantity of silicon retards the decomposition of the carbon-enriched retained austenite. For austempering at 570 K, the amount of the carbon-enriched austenite decreases and the ferrite is supersaturated with carbon, resulting in high tensile strength but low toughness.

  11. Effect of bainite transformation and retained austenite on mechanical properties of austempered spheroidal graphite cast steel

    SciTech Connect

    Takahashi, Toshio; Abe, Toshihiko; Tada, Shuji

    1996-06-01

    Austempered ductile iron (ADI) has excellent mechanical properties, but its Young`s modulus is low. Austempered spheroidal graphite cast steel (AGS) has been developed in order to obtain a new material with superior mechanical properties to ADI. Its carbon content (approximately 1.0 pct) is almost one-third that of a standard ADI; thus, the volume of graphite is also less. Young`s modulus of AGS is 195 to 200 GPa and is comparable to that of steel. Austempered spheroidal graphite cast steel has an approximately 200 MPa higher tensile strength than ADI and twice the Charpy absorbed energy of ADI. The impact properties and the elongation are enhanced with increasing volume fraction of carbon-enriched retained austenite. At the austempering temperature of 650 K, the volume fraction of austenite is approximately 40 pct for 120 minutes in the 2.4 pct Si alloy, although it decreases rapidly in the 1.4 pct Si alloy. The X-ray diffraction analysis shows that appropriate quantity of silicon retards the decomposition of the carbon-enriched retained austenite. For austempering at 570 K, the amount of the carbon-enriched austenite decreases and the ferrite is supersaturated with carbon, resulting in high tensile strength but low toughness.

  12. Impact Properties of Copper-Alloyed and Nickel-Copper Alloyed ADI

    NASA Astrophysics Data System (ADS)

    Batra, Uma; Ray, Subrata; Prabhakar, S. R.

    2007-08-01

    The influence of austenitization and austempering parameters on the impact properties of copper-alloyed and nickel-copper-alloyed austempered ductile irons (ADIs) has been studied. The austenitization temperature of 850 and 900 °C have been used in the present study for which austempering time periods of 120 and 60 min were optimized in an earlier work. The austempering process was carried out for 60 min for three austempering temperatures of 270, 330, and 380 °C to study the effect of austempering temperature. The influence of the austempering time on impact properties has been studied for austempering temperature of 330 °C for time periods of 30-150 min. The variation in impact strength with the austenitization and austempering parameters has been correlated to the morphology, size and amount of austenite and bainitic ferrite in the austempered structure. The fracture surface of ADI failed under impact has been studied using SEM.

  13. 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.

  14. 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

  15. 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. PMID:26601037

  16. 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.

  17. Tensile properties of vanadium-base alloys irradiated in the Fusion-1 low-temperature experiment in the BOR-60 reactor

    SciTech Connect

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

    1998-09-01

    The irradiation has been completed and the test specimens have been retrieved from the lithium-bonded capsule at the Research Institute of Atomic Reactors (RIAR) in Russia. During this reporting period, the Argonne National Laboratory (ANL) tensile specimens were received from RIAR and initial testing and examination of these specimens at ANL has been completed. The results, corroborating previous findings showed a significant loss of work hardening capability in the materials. There appears to be no significant difference in behavior among the various heats of vanadium-base alloys in the V-(4-5)Cr-(4-5)Ti composition range. The variations in the preirradiation annealing conditions also produced no notable differences.

  18. Irradiation effects on 17-7 PH stainless steel, A-201 carbon steel, and titanium-6-percent-aluminum-4-percent-vanadium alloy

    NASA Technical Reports Server (NTRS)

    Hasse, R. A.; Hartley, C. B.

    1972-01-01

    Irradiation effects on three materials from the NASA Plum Brook Reactor Surveillance Program were determined. An increase of 105 K in the nil-ductility temperature for A-201 steel was observed at a fluence of approximately 3.1 x 10 to the 18th power neutrons/sq cm (neutron energy E sub n greater than 1.0 MeV). Only minor changes in the mechanical properties of 17-7 PH stainless steel were observed up to a fluence of 2 x 10 to the 21st power neutrons/sq cm (E sub n greater than 1.0 MeV). The titanium-6-percent-aluminum-4-percent-vanadium alloy maintained its notch toughness up to a fluence of 1 x 10 to the 21st power neutrons/sq cm (E sub n greater than 1.0 MeV).

  19. 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.

  20. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    DOE PAGES

    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 × 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

  1. Scattering effects and high-spatial-frequency nanostructures on ultrafast laser irradiated surfaces of zirconium metallic alloys with nano-scaled topographies.

    PubMed

    Li, Chen; Cheng, Guanghua; Sedao, Xxx; Zhang, Wei; Zhang, Hao; Faure, Nicolas; Jamon, Damien; Colombier, Jean-Philippe; Stoian, Razvan

    2016-05-30

    The origin of high-spatial-frequency laser-induced periodic surface structures (HSFL) driven by incident ultrafast laser fields, with their ability to achieve structure resolutions below λ/2, is often obscured by the overlap with regular ripples patterns at quasi-wavelength periodicities. We experimentally demonstrate here employing defined surface topographies that these structures are intrinsically related to surface roughness in the nano-scale domain. Using Zr-based bulk metallic glass (Zr-BMG) and its crystalline alloy (Zr-CA) counterpart formed by thermal annealing from its glassy precursor, we prepared surfaces showing either smooth appearances on thermoplastic BMG or high-density nano-protuberances from randomly distributed embedded nano-crystallites with average sizes below 200 nm on the recrystallized alloy. Upon ultrashort pulse irradiation employing linearly polarized 50 fs, 800 nm laser pulses, the surfaces show a range of nanoscale organized features. The change of topology was then followed under multiple pulse irradiation at fluences around and below the single pulse threshold. While the former material (Zr-BMG) shows a specific high quality arrangement of standard ripples around the laser wavelength, the latter (Zr-CA) demonstrates strong predisposition to form high spatial frequency rippled structures (HSFL). We discuss electromagnetic scenarios assisting their formation based on near-field interaction between particles and field-enhancement leading to structure linear growth. Finite-difference-time-domain simulations outline individual and collective effects of nanoparticles on electromagnetic energy modulation and the feedback processes in the formation of HSFL structures with correlation to regular ripples (LSFL). PMID:27410083

  2. Corrosion properties of S-phase layers formed on medical grade austenitic stainless steel.

    PubMed

    Buhagiar, Joseph; Dong, Hanshan

    2012-02-01

    The corrosion properties of S-phase surface layers formed in AISI 316LVM (ASTM F138) and High-N (ASTM F1586) medical grade austenitic stainless steels by plasma surface alloying with nitrogen (at 430°C), carbon (at 500°C) and both carbon and nitrogen (at 430°C) has been investigated. The corrosion behaviour of the S-phase layers in Ringer's solutions was evaluated using potentiodynamic and immersion corrosion tests. The corrosion damage was evaluated using microscopy, hardness testing, inductive coupled plasma mass spectroscopy and X-ray diffraction. The experimental results have demonstrated that low-temperature nitriding, carburising and carbonitriding can improve the localised corrosion resistance of both industrial and medical grade austenitic stainless steels as long as the threshold sensitisation temperature is not reached. Carburising at 500°C has proved to be the best hardening treatment with the least effect on the corrosion resistance of the parent alloy.

  3. Phase transformation of Mg-Fe alloys

    SciTech Connect

    Yoneda, Yasuhiro; Abe, Hiroshi; Ohshima, Takeshi; Uchida, Hirohisa

    2010-05-15

    An Mg-Fe alloy system prepared through mechanical alloying (MA) was structurally analyzed. MA can produce single-phase bcc alloys using Mg concentrations up to about 15 mol %. Use of conventional average structure analysis and x-ray pair-distribution function method enabled the long-range and short-range order structures of the Mg-Fe alloys to be bridged. The substituted Mg atoms were randomly arranged in the low-Mg composition but started to have an order structure. The partially ordered Mg-Fe alloy undergoes an austenitic (cubic) to martensitic (orthorhombic) phase change, as increasing Mg composition.

  4. The impact of transmutant helium on weldability of austenitic steel

    NASA Astrophysics Data System (ADS)

    Fabritsiev, S. A.; Pokrovsky, A. S.; Brovko, V. A.

    1996-10-01

    The results of the investigation of 0.05-0.15 dpa neutron irradiation impact on the Cr16Ni11Mo3Ti austenitic steel weldability are presented. Samples were irradiated to doses of 10 20 n/cm 2 and 3 × 10 20 n/cm 2 ( E > 0.1 MeV) in the RBT-10 reactor, thus providing helium accumulation of 1 appm and 2.5 appm, respectively. Flat samples, 1 mm in thickness, were welded by an automatic device for argon arc welding in a hot chamber. Low-cycle fatigue (LCF) testing in bending was used to assess impact of helium on the degradation of welded joint properties. LCF tests showed that the transmutant helium accumulation resulted in a decrease in the number of cycles to failure at Ttest = 20°C and 350°C. It is concluded that repeated welding will present in the repair of ITER materials.

  5. Formation of microcraters and hierarchically-organized surface structures in TiNi shape memory alloy irradiated with a low-energy, high-current electron beam

    SciTech Connect

    Meisner, L. L. Meisner, S. N.; Markov, A. B. Ozur, G. E. Yakovlev, E. V.; Rotshtein, V. P.; Gudimova, E. Yu.

    2015-10-27

    The regularities of surface cratering in TiNi alloy irradiated with a low-energy, high-current electron beam (LEHCEB) in dependence on energy density and number of pulses are studied. LEHCEB processing of TiNi samples was carried out using RITM-SP facility. Energy density E{sub s} was varied from 1 to 5 J/cm{sup 2}, pulse duration was 2.5–3.0 μs, the number of pulses n = 1–128. The dominant role of non-metallic inclusions [mainly, TiC(O)] in the nucleation of microcraters was found. It was revealed that at small number of pulses (n = 2), an increase in energy density leads both to increasing average diameter and density of microcraters. An increase in the number of pulses leads to a monotonic decrease in density of microcraters, and, therefore, that of the proportion of the area occupied by microcraters, as well as a decrease in the surface roughness. The multiple LEHCEB melting of TiNi alloy in crater-free modes enables to form quasi-periodical, hierarchically-organized microsized surface structures.

  6. Shape-Memory-Alloy Actuator For Flight Controls

    NASA Technical Reports Server (NTRS)

    Barret, Chris

    1995-01-01

    Report proposes use of shape-memory-alloy actuators, instead of hydraulic actuators, for aerodynamic flight-control surfaces. Actuator made of shape-memory alloy converts thermal energy into mechanical work by changing shape as it makes transitions between martensitic and austenitic crystalline phase states of alloy. Because both hot exhaust gases and cryogenic propellant liquids available aboard launch rockets, shape-memory-alloy actuators exceptionally suited for use aboard such rockets.

  7. Evaluation of ferritic alloy Fe-2-1/4Cr-1Mo after neutron irradiation - microstructure development

    SciTech Connect

    Gelles, D.S.

    1984-05-01

    Microstructural examinations are reported for nine specimen conditions of 2-1/4Cr-1Mo steel which had been irradiated by fast neutrons over the temperature range 390 to 510/sup 0/C. Two heats of material were involved, each with a different preirradiation heat treatment, one irradiated to a peak fluence of 5.1 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV) or 24 dpa and the other to 2.4 x 10/sup 23/ n/cm/sup 2/ (E > 0.1 MeV) or 116 dpa. Void swelling is found following irradiation at 400/sup 0/C in both conditions and to 480/sup 0/C in the higher fluence conditions. Concurrently dislocation structure and precipitation formed. Peak void swelling, void density, dislocation density and precipitate number density developed at the lowest temperature, approx. 400/sup 0/C, whereas mean void size, and mean precipitate size increased with increasing irradiation temperature. The examination results are used to provide interpretation of in-reactor creep, density change and post irradiation tensile behavior.

  8. The effect of chemical composition and austenite conditioning on the transformation behavior of microalloyed steels

    SciTech Connect

    Mousavi Anijdan, S.H.; Rezaeian, Ahmad; Yue, Steve

    2012-01-15

    In this investigation, by using continuous cooling torsion (CCT) testing, the transformation behavior of four microalloyed steels under two circumstances of austenite conditioning and non-conditioning was studied. A full scale hot-rolling schedule containing a 13-pass deformation was employed for the conditioning of the austenite. The CCT tests were then employed till temperature of {approx} 540 Degree-Sign C and the flow curves obtained from this process were analyzed. The initial and final microstructures of the steels were studied by optical and electron microscopes. Results show that alloying elements would decrease the transformation temperature. This effect intensifies with the gradual increase of Mo, Nb and Cu as alloying elements added to the microalloyed steels. As well, austenite conditioning increased the transformation start temperature due mainly to the promotion of polygonal ferrite formation that resulted from a pancaked austenite. The final microstructures also show that CCT alone would decrease the amount of bainite by inducing ferrite transformation in the two phase region. In addition, after the transformation begins, the deformation might result in the occurrence of dynamic recrystallization in the ferrite region. This could lead to two different ferrite grain sizes at the end of the CCT. Moreover, the Nb bearing steels show no sign of decreasing the strength level after the transformation begins in the non-conditioned situation and their microstructure is a mix of polygonal ferrite and bainite indicating an absence of probable dynamic recrystallization in this condition. In the conditioned cases, however, these steels show a rapid decrease of the strength level and their final microstructures insinuate that ferrite could have undergone a dynamic recrystallization due to deformation. Consequently, no bainite was seen in the austenite conditioned Nb bearing steels. The pancaking of austenite in the latest cases produced fully polygonal ferrite

  9. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    NASA Astrophysics Data System (ADS)

    Brémier, S.; Inagaki, K.; Capriotti, L.; Poeml, P.; Ogata, T.; Ohta, H.; Rondinella, V. V.

    2016-11-01

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15-20 μm and migration of cladding elements to the fuel.

  10. Low-Temperature Nitriding of Deformed Austenitic Stainless Steels with Various Nitrogen Contents Obtained by Prior High-Temperature Solution Nitriding

    NASA Astrophysics Data System (ADS)

    Bottoli, Federico; Winther, Grethe; Christiansen, Thomas L.; Dahl, Kristian Vinter; Somers, Marcel A. J.

    2016-08-01

    In the past decades, high nitrogen steels (HNS) have been regarded as substitutes for conventional austenitic stainless steels because of their superior mechanical and corrosion properties. However, the main limitation to their wider application is their expensive production process. As an alternative, high-temperature solution nitriding has been applied to produce HNS from three commercially available stainless steel grades (AISI 304L, AISI 316, and EN 1.4369). The nitrogen content in each steel alloy is varied and its influence on the mechanical properties and the stability of the austenite investigated. Both hardness and yield stress increase and the alloys remain ductile. In addition, strain-induced transformation of austenite to martensite is suppressed, which is beneficial for subsequent low-temperature nitriding of the surface of deformed alloys. The combination of high- and low-temperature nitriding results in improved properties of both bulk and surface.

  11. Oxidation resistant alloys, method for producing oxidation resistant alloys

    DOEpatents

    Dunning, John S.; Alman, David E.

    2002-11-05

    A method for producing oxidation-resistant austenitic alloys for use at temperatures below 800 C. comprising of: providing an alloy comprising, by weight %: 14-18% chromium, 15-18% nickel, 1-3% manganese, 1-2% molybdenum, 2-4% silicon, 0% aluminum and the balance being iron; heating the alloy to 800 C. for between 175-250 hours prior to use in order to form a continuous silicon oxide film and another oxide film. The method provides a means of producing stainless steels with superior oxidation resistance at temperatures above 700 C. at a low cost

  12. Oxidation resistant alloys, method for producing oxidation resistant alloys

    DOEpatents

    Dunning, John S.; Alman, David E.

    2002-11-05

    A method for producing oxidation-resistant austenitic alloys for use at temperatures below 800.degree. C. comprising of: providing an alloy comprising, by weight %: 14-18% chromium, 15-18% nickel, 1-3% manganese, 1-2% molybdenum, 2-4% silicon, 0% aluminum and the balance being iron; heating the alloy to 800.degree. C. for between 175-250 hours prior to use in order to form a continuous silicon oxide film and another oxide film. The method provides a means of producing stainless steels with superior oxidation resistance at temperatures above 700.degree. C. at a low cost

  13. Control of cryogenic intergranular fracture in high-manganese austenitic steels

    SciTech Connect

    Strum, M.J.

    1986-12-01

    The sources of cryogenic intergranular embrittlement in high-Mn austenitic steels and the conditions necessary for its control are examined. It is shown that the high-Mn alloys are inherently susceptible to intergranular embrittlement due to both their low grain boundary cohesion and heterogeneous deformation characteristics. Extrinsic sources of embrittlement which could account for the transition behavior are not observed. An Auger electron spectroscopy (AES) study shows no indication of impurity-segregation-induced embrittlement. No grain boundary precipitation is observed, and austenite stabilization does not ensure ductile fracture. The influence of chemistry modifications on the ductile-to-brittle transition behavior were also examined through additions of N, Cr, and C to binary Fe-31 Mn. Nitrogen additions increase the 77K yield strength at a rate of 2200 MPa per weight percent N, and increase the austenite stability, but also increase the susceptibility of ternary alloys to intergranular fracture. Quaternary Cr additions are effective in increasing the N solubility, and lower the transition temperature. Carbon additions result in complete suppression of intergranular fracture at 77K. Qualitatively significant changes in the deformation heterogeneity with chemistry modifications are not observed. The temper-toughening of Fe-Mn-Cr-N alloys is associated with the grain boundary segregation of boron and the redistribution of N. Both boron and carbon are expected to inhibit intergranular fracture through increases in grain boundary cohesion.

  14. Effect of Internal Hydrogen on Delayed Cracking of Metastable Low-Nickel Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Papula, Suvi; Talonen, Juho; Todoshchenko, Olga; Hänninen, Hannu

    2014-10-01

    Metastable austenitic stainless steels, especially manganese-alloyed low-nickel grades, may be susceptible to delayed cracking after forming processes. Even a few wppm of hydrogen present in austenitic stainless steels as an inevitable impurity is sufficient to cause cracking if high enough fraction of strain-induced α'-martensite and high residual tensile stresses are present. The role of internal hydrogen content in delayed cracking of several metastable austenitic stainless steels having different alloying chemistries was investigated by means of Swift cup tests, both in as-supplied state and after annealing at 673 K (400 °C). Hydrogen content of the test materials in each state was analyzed with three different methods: inert gas fusion, thermal analysis, and thermal desorption spectroscopy. Internal hydrogen content in as-supplied state was higher in the studied manganese-alloyed low-nickel grades, which contributed to susceptibility of unstable grades to delayed cracking. Annealing of the stainless steels reduced their hydrogen content by 1 to 3 wppm and markedly lowered the risk of delayed cracking. Limiting drawing ratio was improved from 1.4 to 1.7 in grade 204Cu, from 1.7 to 2.0 in grade 201 and from 1.8 to 2.12 in grade 301. The threshold levels of α'-martensite and residual stress for delayed cracking at different hydrogen contents were defined for the test materials.

  15. Development and Exploratory Scale-Up of Alumina-Forming Austenitic (AFA) Stainless Steels

    SciTech Connect

    Brady, Michael P; Magee, John H; Yamamoto, Yukinori; Maziasz, Philip J; Santella, Michael L; Pint, Bruce A; Bei, Hongbin

    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 750C for times up to 2000 h. Ageing resulted in near doubling of yield strength at room temperature after only 50 h at 750C, 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.

  16. On the cryogenic magnetic transition and martensitic transformation of the austenite phase of 7MoPLUS duplex stainless steel

    NASA Astrophysics Data System (ADS)

    Lo, K. H.; Lai, J. K. L.

    2010-08-01

    The magnetic behaviour and martensitic transformation at cryogenic temperatures (down to 4 K) of the austenite phase of the duplex stainless steel (DSS), 7MoPLUS, were studied. As regards the prediction of Neel temperature, the empirical expressions for austenitic stainless steels are not applicable to the austenite phase of 7MoPLUS, although the composition of the austenite phase falls within the composition ranges within which the expressions were developed. Regarding the prediction of martensitic point Ms, the applicability of 'old' and recently developed expressions has been examined. The recently developed expressions, which take into account more alloying elements and their interactions, are not suitable for the austenite phase of 7MoPLUS. But for the 'old', simpler expressions, they seem to be valid in the sense that they all predict high stability of the austenite phase. Results obtained from 7MoPLUS were qualitatively the same as those obtained from another DSS, designated as 2205. Reasons for the applicability and inapplicability of these empirical expressions are suggested.

  17. Effect of Austenite Stability on Microstructural Evolution and Tensile Properties in Intercritically Annealed Medium-Mn Lightweight Steels

    NASA Astrophysics Data System (ADS)

    Song, Hyejin; Sohn, Seok Su; Kwak, Jai-Hyun; Lee, Byeong-Joo; Lee, Sunghak

    2016-06-01

    The microstructural evolution with varying intercritical-annealing temperatures of medium-Mn ( α + γ) duplex lightweight steels and its effects on tensile properties were investigated in relation to the stability of austenite. The size and volume fraction of austenite grains increased as the annealing temperature increased from 1123 K to 1173 K (850 °C to 900 °C), which corresponded with the thermodynamic calculation data. When the annealing temperature increased further to 1223 K (950 °C), the size and volume fraction were reduced by the formation of athermal α'-martensite during the cooling because the thermal stability of austenite deteriorated as a result of the decrease in C and Mn contents. In order to obtain the best combination of strength and ductility by a transformation-induced plasticity (TRIP) mechanism, an appropriate mechanical stability of austenite was needed and could be achieved when fine austenite grains (size: 1.4 μm, volume fraction: 0.26) were homogenously distributed in the ferrite matrix, as in the 1123 K (850 °C)—annealed steel. This best combination was attributed to the requirement of sufficient deformation for TRIP and the formation of many deformation bands at ferrite grains in both austenite and ferrite bands. Since this medium-Mn lightweight steel has excellent tensile properties as well as reduced alloying costs and weight savings, it holds promise for new automotive applications.

  18. Effects of the shape of the foil corners on the irradiation performance of U10Mo alloy based monolithic mini-plates

    SciTech Connect

    Ozaltun, Hakan; Medvedev, Pavel G

    2015-06-01

    Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These fuel elements are comprised of a high density, low enrichment, U-Mo alloy based fuel foil, sandwiched between Zirconium liners and encapsulated in Aluminum cladding. The use of a high density fuel in a foil form presents a number of fabrication and operational concerns, such as: foil centering, flatness of the foil, fuel thickness variation, geometrical tilting, foil corner shape etc. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance have been evaluated. As a part of these series of sensitivity studies, the shape of the foil corners were studied. To understand the effects of the corner shapes of the foil on thermo-mechanical performance of the plates, a behavioral model was developed for a selected plate from RERTR-12 experiments (Plate L1P785). Both fabrication and irradiation processes were simulated. Once the thermo-mechanical behavior the plate is understood for the nominal case, the simulations were repeated for two additional corner shapes to observe the changes in temperature, displacement and stress-strain fields. The results from the fabrication simulations indicated that the foil corners do not alter the post-fabrication stress-strain magnitudes. Furthermore, the irradiation simulations revealed that post-fabrication stresses of the foil would be relieved very quickly in operation. While, foils with chamfered and filleted corners yielded stresses with comparable magnitudes, they are slightly lower in magnitudes, and provided a more favorable mechanical response compared with the foil with sharp corners.

  19. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    DOE PAGES

    Maloy, Stuart A.; Saleh, Tarik A.; Anderoglu, Osman; Romero, Tobias J.; Odette, G. Robert; Yamamoto, Takuya; Li, S.; Cole, James I.; Fielding, Randall

    2015-08-06

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

  20. Low dose of continuous – wave microwave irradiation did not cause temperature increase in muscles tissue adjacent to titanium alloy implants – an animal study

    PubMed Central

    2013-01-01

    Background Research studies on the influence of radiofrequency electromagnetic radiation on implants in vitro have failed to investigate temperature changes in the tissues adjacent to the implants under microwave therapy. We therefore, used a rabbit model in an effort to determine the impact of microwave therapy on temperature changes in tissues adjacent to the titanium alloy implants and the safety profile thereof. Methods Titanium alloy internal fixation plates were implanted in New Zealand rabbits in the middle of femur. Microwave therapy was performed by a 2450 MHz microwave generator 3 days after the surgery. Temperature changes of muscles adjacent to the implants were recorded under exposure to dose-gradient microwave radiation from 20w to 60w. Results Significant difference between control and microwave treatment group at peak temperatures (Tpeak) and temperature gap (Tgap= Tpeak-Tvally) were observed in deep muscles (Tpeak, 41.63 ± 0.21°C vs. 44.40 ± 0.17°C, P < 0.01; Tgap, 5.33 ± 0.21°C vs. 8.10 ± 0.36°C, P < 0.01) and superficial muscles (Tpeak, 41.53 ± 0.15°C vs. 42.03 ± 0.23°C, P = 0.04; Tgap, 5.23 ± 0.21°C vs. 5.80 ± 0.17°C, P = 0.013) under 60 w, and deep muscles (Tpeak, 40.93 ± 0.25°C vs. 41.87 ± 0.23°C, P = 0.01; Tgap, 4.73 ± 0.20°C vs. 5.63 ± 0.35°C, P = 0.037) under 50w, but not under 20, 30 and 40w. Conclusion Our results suggest that low-dose (20w-40w) continuous-wave microwave irradiation delivered by a 2450 MHz microwave generator might be a promising treatment for patients with titanium alloy internal fixation, as it did not raise temperature in muscle tissues adjacent to the titanium alloy implant. PMID:24365389