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Sample records for austenitic alloys irradiated

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

    DOEpatents

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

    1982-01-01

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

  2. 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.

  3. Comparison of irradiation creep and swelling of an austenitic alloy irradiated in FFTF and PFR

    SciTech Connect

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

    1999-10-01

    comparative irradiation of identically constructed creep tubes in the Fast Flux Test Facility (FFTF) and the Prototypic Fast Reactor (PFR) shows that differences in irradiation conditions arising from both reactor operation and the design of the irradiation vehicle can have a significant impact on the void swelling and irradiation creep of austenitic stainless steels. In spite of these differences, the derived creep coefficients fall within the range of previously observed values for 316 SS.

  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. Electron irradiation-induced defects in Mo-diluted FeCrNi austenitic alloy during void swelling incubation

    NASA Astrophysics Data System (ADS)

    Wang, B. Y.; Lu, E. Y.; Zhang, C. X.; Xu, Q.; Jin, S. X.; Zhang, P.; Cao, X. Z.

    2016-01-01

    The microstructural features and the effect of Mo addition in FeCrNi austenitic alloy during incubation period were investigated using positron annihilation technique and micro- Vickers Hardness. The electron irradiation, which could induce vacancy defects in material, was performed at room temperature up to the dose of 1.7×10-4 and 5×10-4 dpa, respectively. The defect concentration was estimated about 10-4-10-7 though the standard trapping model. The added Mo atoms could trap vacancies to form Mo-vacancy complexes, which may restrain the migration and growth of vacancy defects during electron irradiation. In addition, the microstructural evolution during electron radiation resulted in hardening, while the added Mo might improve the hardening property of the alloy.

  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. 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.

  13. 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.

  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. Corrosion of austenitic alloys in aerated brines

    SciTech Connect

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

    1999-11-01

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

  16. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    In this last of a series of papers describing the evolution of microstructure, void swelling and mechanical properties of model austenitic alloys in response to differences in helium/dpa rates, shear punch testing is used to assess the relative effect of helium generation ratios and various important material and environmental variables. Shear punch data confirm the general trends observed in earlier tensile data derived from the 59Ni isotopic doping experiment. There is a convergence to a common saturation level of yield strength that depends on alloy composition, temperature and displacement rate, but not on starting condition. The approach to saturation can be sensitive to helium/dpa ratio, however, and may depend on the starting state. For reasons not yet known, shear punch tests appear to be more sensitive to such transient differences than are tensile tests.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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)

  4. 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

  5. Effect of FFTF irradiation on tensile properties of P- and Ti-modified model austenitic alloys with small amounts of boron

    NASA Astrophysics Data System (ADS)

    Kurishita, H.; Muroga, T.; Watanabe, H.; Yoshida, N.; Kayano, H.; Hamilton, M. L.

    1994-09-01

    The interactive effects of P and Ti additions, {helium}/{dpa} ratio, irradiation and test temperature on postirradiation tensile properties of a model Fe-16Cr-17Ni alloy were investigated. Miniature tensile specimens containing 64 and 522 appm 10B (0.75 and 3.8 {He}/{dpa} ratio), with and without 0.1 wt% P and 0.25 wt% Ti additions, were irradiated to 33 dpa at 703, 793 and 874 K in the Fast Flux Test Facility (FFTF). They were deformed at 300, 473 and 673 K. For all the alloy conditions the irradiation at 703 K increases the yield stress and decreases the uniform elongation significantly. An effect of 10B addition occurs, which is not related to helium generation, but it is small compared to the effects of P and Ti additions. Additions of P and Ti, especially their simultaneous addition, cause a significant strengthening but no appreciable change in uniform elongation. The cause of the strengthening and the observed changes in uniform elongation are discussed.

  6. 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.

  7. 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.

  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. 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.

  11. Radiation-induced segregation of deuterium in austenitic steels and vanadium alloys

    NASA Astrophysics Data System (ADS)

    Arbuzov, V. L.; Raspopova, G. A.; Vykhodets, V. B.

    The accumulation and distribution of implanted deuterium were studied through simultaneous analysis using the nuclear reaction D(d,p)T for some austenitic, austenitic-martensitic steels, Fe-16% Cr, V-4% Ti-4% Cr, V-10% Ti-5% Cr alloys, and vanadium. The implantation was carried out by 700-keV deuteron irradiation at room temperature with a total implantation dose of about 2 × 10 18 cm -2. It is shown that the deuterium segregation induced by ion irradiation in vanadium and the Fe-16% Cr alloy remained unchanged during room temperature holding after implantation. On the other hand, in the two-phase steel and the V-Ti(-Cr) alloys the holding led to a partial elimination of the concentration inhomogeneity of the implant in the irradiated portion, while in the austenitic steel deuterium segregation increased probably due to the migration of deuterium from the unirradiated volume to the irradiation zone. Possible reasons for different behavior of the implanted deuterium in different materials will be briefly discussed.

  12. 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.

  13. 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.

  14. 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.

  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 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.

  17. 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

  18. Effects of phosphorus, silicon and sulphur on microstructural evolution in austenitic stainless steels during electron irradiation

    NASA Astrophysics Data System (ADS)

    Fukuya, K.; Nakahigashi, S.; Ozaki, S.; Shima, S.

    1991-03-01

    Fe-18Cr-9Ni-1.5Mn austenitic alloys containing phosphorus, silicon and sulphur were irradiated by 1 MeV electrons at 573-773 K. Phosphorus increased the intersitial loop nucleation and decreased the void swelling by increasing void number density and suppressing void growth. Silicon had a similar effect to phosphorus but its effect was weaker than phosphorus. Sulphur enhanced void swelling through increasing the void density. Nickel enrichment at grain boundaries was suppressed only in the alloy containing phosphorus. These phosphorus effects may be explained by a strong interaction with interstitials resulting in a high density of sinks for point defects.

  19. Features of structure-phase transformations and segregation processes under irradiation of austenitic and ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Neklyudov, I. M.; Voyevodin, V. N.

    1994-09-01

    The difference between crystal lattices of austenitic and ferritic steels leads to distinctive features in mechanisms of physical-mechanical change. This paper presents the results of investigations of dislocation structure and phase evolution, and segregation phenomena in austenitic and ferritic-martensitic steels and alloys during irradiation with heavy ions in the ESUVI and UTI accelerators and by neutrons in fast reactors BOR-60 and BN-600. The influence of different factors (including different alloying elements) on processes of structure-phase transformation was studied.

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

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

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

  1. Alumina-forming Austenitic Alloys for Advanced Recuperators

    SciTech Connect

    Pint, Bruce A; Shingledecker, John P; Brady, Michael P; Maziasz, Philip J

    2007-01-01

    Materials selection for thin-walled recuperators has been extensively investigated over the past decade. In the latest generation of recuperated turbine engines, type 347 stainless steel has been replaced by higher alloyed steels and Ni-base chromia-forming alloys. However, high (linear) rates of chromia evaporation in exhaust gas fundamentally limits the oxidation lifetime of these chromia-forming alloys. One solution is to use alumina-forming alloys that are more resistant to this environment. The lower scale growth kinetics and resistance to evaporation in the presence of water vapor suggests an order of magnitude increase in lifetime for alumina-forming alloys. A significant problem with this strategy was the large drop in creep strength with the addition of sufficient Al to form an external alumina scale. However, new Fe-base austenitic compositions have been developed with sufficient strength for this application above 700 C.

  2. 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.

  3. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2011-02-01

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

  5. 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.

  6. 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.

  7. 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.

  8. Microstructure of austenitic stainless steels irradiated at 400°C in the ORR and the HFIR spectral tailoring experiment

    NASA Astrophysics Data System (ADS)

    Hashimoto, N.; Wakai, E.; Robertson, J. P.; Sawai, T.; Hishinuma, A.

    2000-07-01

    Microstructural evolution in solution-annealed Japanese-PCA (JPCA-SA) and four other austenitic stainless steels, irradiated at 400°C to 17.3 dpa in the ORR and the high flux isotope reactor (HFIR) spectrally tailored experiment, were investigated by transmission electron microscopy (TEM). The mean He/dpa ratio throughout the irradiation fell between 12 and 16 appm He/dpa , which is close to the He/dpa values expected for fusion. In all the specimens, a bi-modal size distribution of cavities was observed and the number densities were about 1.0×10 22 m -3. There was no significant difference between the number densities in the different alloys, although the root mean cubes of the cavity radius are quite different for each alloy. Precipitates of the MC type were also observed in the matrix and on grain boundaries in all alloys except a high-purity (HP) ternary alloy. The JPCA-SA (including 0.06% carbon and 0.027% phosphorus) and standard type 316 steel (including 0.06% carbon and 0.028% phosphorus) showed quite low-swelling values of about 0.016 and 0.015%, respectively, while a HP ternary austenitic alloy showed the highest swelling value of 2.9%. This suggests that the existence of impurities affects the cavity growth in austenitic stainless steels even at 400°C.

  9. Effects of dose rate on microsturctural evolution and swelling in austenitic steels under irradiation

    NASA Astrophysics Data System (ADS)

    Okita, T.; Kamada, T.; Sekimura, N.

    2000-12-01

    Effects of dose rate on microstructural evolution in a simple model austenitic ternary alloy are examined. Annealed specimens are irradiated with fast neutrons at several positions in the core and above core in FFTF/MOTA between 390°C and 435°C in a wide range of doses and dose rates. In Fe-15Cr-16Ni, swelling seems to increase linearly with dose without incubation dose. Cavities are observed even in the specimens irradiated to 0.07 dpa at 1.9×10-9 dpa/s. Both cavity nucleation and growth are enhanced by low dose rates. These are mainly caused by accelerated formation of dislocation loops at lower dose rates. Low dose rates enhance swelling by shortening incubation dose for the onset of steady-state swelling. In the specimens irradiated at higher dose rates to higher doses, high density of dislocation increases average cavity diameter, however decreases cavity density.

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

    DOE PAGESBeta

    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

  11. 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.

  12. 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.

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

    SciTech Connect

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

    2010-01-15

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

  14. 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)

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

    NASA Astrophysics Data System (ADS)

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

    2015-05-01

    The dynamics of deformation localization and dislocation channel formation were investigated 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 (TEM). Channel formation was observed at ∼70% of the polycrystalline yield stress of the irradiated materials (σ0.2). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ0.2, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young's modulus) in 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. 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. In the AISI 304 steel, channels in grains oriented close to <0 0 1>||TA (tensile axis) and <1 0 1>||TA were twin free and grain with <1 1 1>||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

  16. 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.

  17. 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.

  18. Swelling in several commercial alloys irradiated to very high neutron fluence

    SciTech Connect

    Gelles, D.S.; Pintler, J.S.

    1983-01-01

    Swelling values have been obtained from a set of commercial alloys irradiated in EBR-II to a peak fluence of 2.5 x 10/sup 23/ n/cm/sup 2/ (E > 0.1 MeV) or approx. 125 dpa covering the range 400 to 650/sup 0/C. The alloys can be ranked for swelling resistance from highest to lowest as follows: the martensitic and ferritic alloys, the niobium based alloys, the precipitation strengthened iron and nickel based alloys, the molybdenum alloys and the austenitic alloys.

  19. Microstructure inhomogeneity of Fe-31%Ni alloy and stabilization of austenite.

    PubMed

    Dzevin, Ievgenij M

    2015-01-01

    Сrystal structure and mechanism of crystallization of Fe-Ni alloys were studied by methods of X-ray diffraction and metallography. It has been found that macro- and microstructure of austenitic alloy was essentially heterogeneous at the contact and free surfaces and in the volume of a ribbon. The indentified peculiarities of the austenitic phase in different areas of the ribbon are attributed to different cooling rates and the melt crystallization conditions. PMID:25852411

  20. Direct decomposition of austenite in two Fe-V-C alloys

    NASA Astrophysics Data System (ADS)

    Mishima, Y.; Horn, R. M.; Zackay, V. F.; Parker, E. R.

    1980-03-01

    Investigations of austenite decomposition have been undertaken in (1) Fe-0.5Mn-1V-0.2C and (2) Fe-0.5Mn-3Ni-1V-0.2C alloys. Isothermal transformation characteristics were determined using dilatometric and thermo-electric potential techniques. Also, micro-structural features were observed using optical and transmission electron microscopy for treatments of interest following isothermal austenite decomposition in the 550 to 750° C range. Associated mechanical properties were measured with emphasis being placed on Charpy impact behavior. Both alloys exhibited two temperature regions in which “C-curve” austenite decomposition occurred. In the upper region a combination of fibrous and fine particle VC precipitation was observed in both alloys. In the lower transformation region, bainitic microstructures resulted from the isothermal treatments. Additionally, the alloy containing 3 pct Ni exhibited VC precipitation in the austenite prior to ferrite formation. In both alloys, complete isothermal transformation produced microstructures with poor impact properties. However, a good combination of strength and toughness was produced in the 3 pct Ni alloy using the heat treatment that promoted VC precipitation in austenite but avoided total isothermal austenite decomposition.

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

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

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

  2. 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.

  3. Evaluation of advanced austenitic alloys relative to alloy design criteria for steam service

    SciTech Connect

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

    1991-06-01

    The results are summarized for a task within a six-year activity to evaluate advanced austenitic alloys for heat recovery systems. Commercial, near-commercial, and development alloys were evaluated relative to criteria for metallurgical stability, fabricability, weldability, mechanical properties, and corrosion in fireside and steamside environments. Alloys that were given special attention in the study were 800HT{reg sign}, NF709{reg sign}, HR3C{reg sign}, and a group of 20/25% chromium-30% nickel-iron alloys identified as HT- UPS (high-temperature, ultrafine-precipitation strengthened) alloys. Excellent metallurgical stability and creep strength were observed in the NF709 and HR3C steels that contained niobium and nitrogen. One group of HT-UPS alloys was strengthened by solution treating to temperatures above 1150{degrees}C and subsequent cold or warm working. Test data to beyond 35,000 h were collected. The ability to clad some of the alloys for improved fireside corrosion resistance was demonstrated. Weldability of the alloys was a concern. Hot cracking and heat-affected-zone (HAZ) liquation cracking were potential problems in the HR3C stainless steel and HT-UPS alloys, and the use of dissimilar metal filler wire was required. By the reduction of phosphorous content and selection of either a nickel-base filler metal or alloy 556 filler metal, weldments were produced with minimum HAZ cracking. The major issues related to the development of the advanced alloys were identified and methods to resolve the issues suggested. 56 refs., 19 figs., 8 tabs.

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

    SciTech Connect

    Woollin, P.

    1996-12-31

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

  5. Welding techniques for high alloy austenitic stainless steel

    SciTech Connect

    Gooch, T.G.; Woollin, P.

    1996-11-01

    Factors controlling corrosion resistance of weldments in high alloy austenitic stainless steel are described, with emphasis on microsegregation, intermetallic phase precipitation and nitrogen loss from the molten pool. The application is considered of a range of welding processes, both fusion and solid state. Autogenous fusion weldments have corrosion resistance below that of the parent, but low arc energy, high travel speed and use of N{sub 2}-bearing shielding gas are recommended for best properties. Conventional fusion welding practice is to use an overalloyed nickel-base filler metal to avoid preferential weld metal corrosion, and attention is given to the effects of consumable composition and level of weldpool dilution by base steel. With non-matching consumables, overall joint corrosion resistance may be limited by the presence of a fusion boundary unmixed zone: better performance may be obtained using solid state friction welding, given appropriate component geometry. Overall, the effects of welding on superaustenitic steels are understood, and the materials have given excellent service in welded fabrications. The paper summarizes recommendations on preferred welding procedure.

  6. 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.

  7. Irradiation assisted stress corrosion cracking of austenitic stainless steels

    SciTech Connect

    Was, G.S.; Atzmon, M.

    1990-06-01

    Samples of ultra high purity stainless steel have been fabricated into 2mm {times} 2mm rectangular bars and irradiated to one dpa ({approximately}l {times} 10{sup 19} p{sup +}/cm{sup 2}) using 3.4 MeV protons (>20{mu}A) while controlling the sample temperature at 400{degree}C. Samples are pressed onto a water-cooled and electrically heated copper block with a thin layer of Sn in between to improve thermal conductivity. The irradiation produced a significant prompt radiation field but sample activation was limited to {beta}-decay and this decayed rapidly in less than 48 h. Samples were hydrogen charged and strained at slow rates at {minus}30{degree}C insitu in the Auger electron spectrometer to successfully fracture several samples intergranularly for grain boundary composition analysis. An ultra-high purity (UHP) alloy of Fe-19Cr-9Ni was irradiated to 1 dpa at 400C {plus minus} 5C and 7 {times} 10{sup {minus}9} torr in the tandem accelerator of the Michigan Ion Beam Laboratory, resulting in a dislocation network density of 1.8 {times} 10{sup 9} cm{sup 2} and a dislocation loop density of 7 {times} 10{sup 16} cm{sup {minus}3} along with the dissolution of small precipitates present in the unirradiated sample. EPR experiments on the UHP irradiated alloy showed no significant increase in charge passed upon reactivation, over an unirradiated sample experiencing the same thermal history. An SCC waterloop and autoclave system has been completed and a sample has been designed to measure the susceptibility of the irradiated microstructure as compared to the unirradiated microstructure.

  8. 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.

  9. 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.

  10. 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.

  11. Bainitic stabilization of austenite in low alloy sheet steels

    NASA Astrophysics Data System (ADS)

    Brandt, Mitchell L.

    The stabilization of retained austenite in 'triple phase' ferrite/bainite/austenite sheet steels by isothermal bainite transformation after intercritical annealing has been studied in 0.27C-1.5Si steels with 0.8 to 2.4Mn. Dilatometric studies show that cooling rates comparable to CAPL processing result in approximately 30% conversion of austenite to epitaxial ferrite, but the reaction can be suppressed by the faster cooling rate of salt bath quenching. Measured isothermal transformation kinetics at 350 to 450sp°C shows a maximum overall rate near 400sp°C. X-ray diffraction shows that the amount of austenite retained from 400sp°C treatment peaks at 3 minutes but the carbon content increases monotonically to a saturation level. The stability of austenite in this type of steel has been quantified for the first time by direct measurement of the characteristic Msbsps{sigma} temperature. With variations in processing conditions and test temperatures, the tensile uniform ductility has been correlated with the amount and stability of retained austenite, while maintaining a constant 3% flow of 83 ksi. Consistent with previous transformations plasticity studies an optimal austenite stability is found at approximately 10 K above the Msbsps{sigma} temperature, demonstrating a maximum uniform ductility of 44% for an austenite content of 16%. Correlations indicate that desired uniform ductility levels of 20 to 25% could be achieved with only approximately 5% austenite if stability is optimized by placing Msbsps{sigma} 10 K below ambient temperature. Measured uniform ductility in plane strain tension shows similar trends with processing conditions, but models predict that stress state effects will shift the Msbsps{sigma} temperature approximately 5 K higher than that for uniaxial tension. The measured dependence of Msbsps{sigma} on austenite composition and particle size has been modeled via heterogeneous nucleation theory. The composition dependence is consistent with

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

    NASA Astrophysics Data System (ADS)

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

    2016-03-01

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

  13. 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.

  14. 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.

  15. The temperature dependent role of phosphorus and titanium in microstructural evolution of FeCrNi alloys irradiated in FFTF

    NASA Astrophysics Data System (ADS)

    Watanabe, H.; Muroga, T.; Yoshida, N.

    1996-04-01

    The influence of combined addition of phosphorus and titanium on the microstructure of model FeCrNi austenitic alloys irradiated at 700-873 K to 60 dpa has been investigated in comparison with that of a complex austenitic alloy (JPCA-2). At all temperatures, void swelling of the model alloys was suppressed with increasing phosphorus content or co-addition of phosphorus and titanium. The microstructures observed and calculation of defect processes showed that the suppressed void swelling was due to an interaction of phosphorus in solution with defects at lower temperature and needle-like phosphide acting as defect sinks at higher temperature.

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

    NASA Astrophysics Data System (ADS)

    Schibli, Raluca; Schäublin, Robin

    2013-11-01

    Irradiated austenitic stainless steels, because of their low stacking fault energy and high shear modulus, should exhibit a high ratio of stacking fault tetrahedra relative to the overall population of radiation induced nanometric defects. Experimental observations of stacking fault tetrahedra by transmission electron microscopy in commercial-purity stainless steels are however scarce, while they abundantly occur in high-purity or model austenitic alloys irradiated at both low and high temperatures, but not at around 673 K. In commercial alloys, the little evidence of stacking fault tetrahedra does not follow such a trend. These contradictions are reviewed and discussed. Reviewing the three possible formation mechanisms identified in the literature, namely the Silcox and Hirsch Frank loop dissociation, the void collapse and the stacking fault tetrahedra growth, it seems that the later dominates under irradiation. Black dots, are very small defect clusters, smaller than 1 nm in diameter, which cannot be resolved in TEM being below its spatial resolution in diffraction contrast. They can be created directly from the collapse of the cascade as undefined 3D clusters of point defects, namely vacancies, interstitials or impurities, or could be already well-defined nanometric voids, vacancy or interstitial dislocation loops [7]. Dislocation loops, either Frank or perfect dislocation loops, are generated by vacancies or interstitials coalescing as platelets between two adjacent {1 1 1} close-packed planes. Perfect loops are scarcer than Frank loops. For irradiation temperatures below 573 K some authors identified that Frank loops are of interstitial nature, while black dots are predominantly of vacancy nature [8-11]. More recent studies [12] contradict this statement and conclude that Frank loops with sizes in the range of 1-30 nm can be either vacancy or interstitial type. Stacking fault tetrahedra (SFT) are three-dimensional stacking fault configurations in the shape of

  17. Stress corrosion cracking and intergranular corrosion of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Fukuya, K.; Shima, S.; Kayano, H.; Narui, M.

    1992-09-01

    The effects of irradiation on stress corrosion cracking (SCC) and intergranular corrosion (IGC) susceptibility were investigated in solution-treated Fe19Cr9NiMn alloys and JPCA irradiated to 5.3×1024 n/m2 (E > 1 MeV) at 573 K. In Fe19Cr9NiMn alloys, the irradiation enhanced IGC i n boiling HNO3 + Cr6+ solution when the alloys contained phosphorus and silicon and induced SCC in all the alloys with strain rate tensile tests in 571 K water containing 32 ppm oxygen. With increasing phosphorus and silicon contents. IGC was promoted but IGSCC was suppressed after irradiation. The results indicated that these elements are not the main contributors to irradiation-assisted SCC, although they affect SCC behavior. The Japanese Prime Candidate Alloy (JPCA) had better SCC resistance than Fe19Cr9NiMn alloys under the present irradiation condition.

  18. Influence of kinetics of supercooled austenite decomposition on structure formation in sparingly-alloyed tool steel

    NASA Astrophysics Data System (ADS)

    Krylova, S. E.; Yakovleva, I. L.; Tereshchenko, N. A.; Priimak, E. Yu.; Kletsova, O. A.

    2013-10-01

    The decomposition of supercooled austenite in 70Kh3G2VTB steel under isothermal conditions and continuous cooling have been studied. The isothermal and continuous cooling tranformation curves of the decomposition of austenite in the experimental steel have been constructed. The effect of alloying elements on phase transformations in the steel under heating and cooling have been established. The features of the formation of a microstructure in the 70Kh3G2VTB steel after different regimes of heat treatment have been described. The optimal parameters of hardening heat treatment have been developed.

  19. 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.

  20. 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.

  1. 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

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

    DOEpatents

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

    1977-03-08

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

  3. 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.

  4. Microstructural evolution of austenitic stainless steels irradiated in spectrally tailored experiment in ORR at 400°C

    NASA Astrophysics Data System (ADS)

    Sawai, T.; Maziasz, P. J.; Kanazawa, H.; Hishinuma, A.

    1992-09-01

    Several different heats of austenitic stainless steel, including Japanese-PCA(JPCA), were irradiated in the spectrally tailored ORR experiment at 400°C to 7.4 dpa. The levels of helium generated were 155 appm for JPCA (16Ni, 30 wppm B) and 102 appm for standard type 316 steel (13Ni). The mean He: dpa ratio throughout the irradiation falls between 15 and 20 appm He/dpa, which is close to the He/dpa values expected for fusion. Swelling was measured by transmission electron microscopy and by precision immersion densitometry. All the CW alloys showed swelling that was at or below the detection limit of the densitometer (0.1%). No measurable swelling was detected in the SA JPCA alloy, while the highest value of 0.8% was observed in the SA high-purity alloy. One Ti-modified steel with low C also showed a relatively high swelling value of 0.5%, while standard type 316 steel showed only 0.15% swelling. TEM observation gave consistent but slightly larger values of swelling.

  5. 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

  6. Influence of radiation-induced voids and bubbles on physical properties of austenitic structural alloys

    NASA Astrophysics Data System (ADS)

    Balachov, Iouri I.; Shcherbakov, E. N.; Kozlov, A. V.; Portnykh, I. A.; Garner, F. A.

    2004-08-01

    Void swelling in austenitic stainless steels induces significant changes in their electrical resistivity and elastic moduli, as demonstrated in this study using a Russian stainless steel irradiated as fuel pin cladding in BN-600. Precipitation induced by irradiation also causes second-order changes in these properties, but can dominate the measurement for small swelling levels. When cavities are full of helium as expected under some fusion irradiation conditions, additional second-order changes are expected but they will be small enough to exclude from the analysis.

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

    NASA Astrophysics Data System (ADS)

    Kaynak, Yusuf

    2014-09-01

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

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

    SciTech Connect

    Guertsman, Valery Y.; Bruemmer, Stephen M.

    2001-05-25

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

  9. Characterization of high performance austenitic and ODS alloys for SCWR conditions

    SciTech Connect

    Penttila, S.; Toivonen, A.; Auerkari, P.; Novotny, R.

    2012-07-01

    High temperature oxidation resistance is of critical importance for the in-reactor components of the European supercritical water reactor (SCWR) concept. To consider candidate materials for this purpose, selected austenitic steels, iron based ODS (Oxide Dispersion Strengthened) alloys and one Ni alloy have been tested by exposure to supercritical water at 650 deg. C/25 MPa up to 2000 h. Results of observed mass change, oxide thickness and composition after exposure are shown and discussed regarding implications for long term oxidation performance. (authors)

  10. Physical and welding metallurgy of Gd-enriched austenitic alloys for spent nuclear fuel applications. Part II, nickel base alloys.

    SciTech Connect

    Mizia, Ronald E.; Michael, Joseph Richard; Williams, David Brian; Dupont, John Neuman; Robino, Charles Victor

    2004-06-01

    The physical and welding a metallurgy of gadolinium- (Gd-) enriched Ni-based alloys has been examined using a combination of differential thermal analysis, hot ductility testing. Varestraint testing, and various microstructural characterization techniques. Three different matrix compositions were chosen that were similar to commercial Ni-Cr-Mo base alloys (UNS N06455, N06022, and N06059). A ternary Ni-Cr-Gd alloy was also examined. The Gd level of each alloy was {approx}2 wt-%. All the alloys initiated solidification by formation of primary austenite and terminated solidification by a Liquid {gamma} + Ni{sub 5}Gd eutectic-type reaction at {approx}1270 C. The solidification temperature ranges of the alloys varied from {approx}100 to 130 C (depending on alloy composition). This is a substantial reduction compared to the solidification temperature range to Gd-enriched stainless steels (360 to 400 C) that terminate solidification by a peritectic reaction at {approx}1060 C. The higher-temperature eutectic reaction that occurs in the Ni-based alloys is accompanied by significant improvements in hot ductility and solidification cracking resistance. The results of this research demonstrate that Gd-enriched Ni-based alloys are excellent candidate materials for nuclear criticality control in spent nuclear fuel storage applications that require production and fabrication of large amounts of material through conventional ingot metallurgy and fusion welding techniques.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. The microstructural, mechanical, and fracture properties of austenitic stainless steel alloyed with gallium

    NASA Astrophysics Data System (ADS)

    Kolman, D. G.; Bingert, J. F.; Field, R. D.

    2004-11-01

    The mechanical and fracture properties of austenitic stainless steels (SSs) alloyed with gallium require assessment in order to determine the likelihood of premature storage-container failure following Ga uptake. AISI 304 L SS was cast with 1, 3, 6, 9, and 12 wt pct Ga. Increased Ga concentration promoted duplex microstructure formation with the ferritic phase having a nearly identical composition to the austenitic phase. Room-temperature tests indicated that small additions of Ga (less than 3 wt pct) were beneficial to the mechanical behavior of 304 L SS but that 12 wt pct Ga resulted in a 95 pct loss in ductility. Small additions of Ga are beneficial to the cracking resistance of stainless steel. Elastic-plastic fracture mechanics analysis indicated that 3 wt pct Ga alloys showed the greatest resistance to crack initiation and propagation as measured by fatigue crack growth rate, fracture toughness, and tearing modulus. The 12 wt pct Ga alloys were least resistant to crack initiation and propagation and these alloys primarily failed by transgranular cleavage. It is hypothesized that Ga metal embrittlement is partially responsible for increased embrittlement.

  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. PMID:26860715

  19. Tensile properties and damage microstructures in ORR/HFIR-irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Wakai, E.; Hashimoto, N.; Robertson, J. P.; Jistukawa, S.; Sawai, T.; Hishinuma, A.

    2000-12-01

    The synergistic effect of displacement damage and helium generation under neutron irradiation on tensile behavior and microstructures of austenitic stainless steels was investigated. The steels were irradiated at 400°C in the spectrally-tailored (ST) Oak Ridge research reactor/high flux isotope reactor (ORR/HFIR) capsule to 17 dpa with a helium production of about 200 appm and in the HFIR target capsule to 21 and 34 dpa with 1590 and 2500 appm He, respectively. The increase of yield strength in the target irradiation was larger than that in the ST irradiation because of the high-number density of Frank loops, bubbles, voids, and carbides. Based on the theory of dispersed barrier hardening, the strengths evaluated from these clusters coincide with the measured increase of yield strengths. This analysis suggests that the main factors of radiation hardening in the ST and the target irradiation at 400°C are Frank-type loops and cavities, respectively.

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

    SciTech Connect

    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 1,000 C for 15 to 360 min and austempered at 360 C for 180 min. Results showed that the austenitizing temperature, T{sub {gamma}}, and time, t{sub {gamma}} have a significant effect on the impact properties of the alloy. This has been attributed to the influence of these variables on the carbon kinetics. Microstructures of samples austenitized at 950 and 1,000 C contain no pro-eutectoid ferrite. The impact properties of the former structures are independent of t{sub {gamma}}, while those solution treated at 1,000 C are generally low and show wide variation over the range of soaking time investigated. For fully ausferritic structures, impact properties fall with an increase in T{sub {gamma}}. This is particularly evident at 1,000 C. As the T{sub {gamma}} increases, the amount of carbon dissolved in the original austenite increases. This slows down the rate of austenite transformation and results in coarser structures with lower mechanical properties. Optimum impact properties are obtained following austenitizing between 900 and 950 C for 120 to 180 min.

  1. 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.

  2. 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.

  3. Interim fatigue design curves for carbon, low-alloy, and austenitic stainless steels in LWR environments

    SciTech Connect

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

    1993-01-01

    Both temperature and oxygen affect fatigue life; at the very low dissolved-oxygen levels in PWRs and BWRs with hydrogen water chemistry, environmental effects on fatigue life are modest at all temperatures (T) and strain rates. Between 0.1 and 0.2 ppM, the effect of dissolved-oxygen increases rapidly. In oxygenated environments, fatigue life depends strongly on strain rate and T. A fracture mechanics model is developed for predicting fatigue lives, and interim environmentally assisted cracking (EAC)-adjusted fatigue curves are proposed for carbon steels, low-alloy steels, and austenitic stainless steels.

  4. 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.

  5. Electroslag remelt processing of irradiated vanadium alloys

    NASA Astrophysics Data System (ADS)

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

    1996-10-01

    This paper describes experimental efforts to investigate the potential of a slag remelting process for reducing radioactivity of irradiated vanadium alloys used in a fusion power production facility. The experiment determined the removal characteristics of four surrogate transmutation isotopes significant to accident safety expected in a V5Ti5Cr alloy irradiated under fusion conditions (Ca, Y, to simulate Sc, Mn, and Ar). Removal of these isotopes could decrease the accident risk of reprocessing irradiated vanadium and reusing it in fusion reactors. An electroslag remelt furnace was used in the experiment to melt and react the constituents using a calcium fluoride slag. The process achieved 90% removal of calcium and over 99% removal of yttrium. Analyses indicate that 40% of the manganese has been removed. Argon analysis of the refined ingots indicates that 99% of the argon was removed from the vanadium alloy.

  6. 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.

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

    NASA Astrophysics Data System (ADS)

    Chen, Hao; Zhu, Benqiang; Militzer, Matthias

    2016-06-01

    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.

  8. 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.

  9. 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.

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

    SciTech Connect

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

    2009-03-10

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

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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).

  16. Alloy Design, Combinatorial Synthesis, and Microstructure-Property Relations for Low-Density Fe-Mn-Al-C Austenitic Steels

    NASA Astrophysics Data System (ADS)

    Raabe, D.; Springer, H.; Gutierrez-Urrutia, I.; Roters, F.; Bausch, M.; Seol, J.-B.; Koyama, M.; Choi, P.-P.; Tsuzaki, K.

    2014-09-01

    We present recent developments in the field of austenitic steels with up to 18% reduced mass density. The alloys are based on the Fe-Mn-Al-C system. Here, two steel types are addressed. The first one is a class of low-density twinning-induced plasticity or single phase austenitic TWIP (SIMPLEX) steels with 25-30 wt.% Mn and <4-5 wt.% Al or even <8 wt.% Al when naturally aged. The second one is a class of κ-carbide strengthened austenitic steels with even higher Al content. Here, κ-carbides form either at 500-600°C or even during quenching for >10 wt.% Al. Three topics are addressed in more detail, namely, the combinatorial bulk high-throughput design of a wide range of corresponding alloy variants, the development of microstructure-property relations for such steels, and their susceptibility to hydrogen embrittlement.

  17. 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.

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

    SciTech Connect

    Shiba, Kiyoyuki; Ioka, Ikuo; Jitsukawa, Shiro; Hamada, Shozo; Hishinuma, Atkinichi; Robertson, J.P.

    1999-10-01

    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 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 not only for base metal specimens but also for the weld joint and the weld metal specimens.

  19. Analysis of Tensile Deformation and Failure in Austenitic Stainless Steels: Part II- Irradiation Dose Dependence

    SciTech Connect

    Kim, Jin Weon; Byun, Thak Sang

    2010-01-01

    Irradiation effects on stable and unstable deformations and fracture behaviors in irradiated austenitic stainless steels (SSs) have been studied in detail based on the equivalent true stress versus true strain curves. An iterative technique in finite element simulation was used to obtain the equivalent true stress-true strain data from experimental tensile curves. It was shown that the strain hardening rate was retained at a high level on unstable deformation after significant irradiation and was independent of the irradiation dose up to the initiation of a localized necking. The equivalent fracture stress was nearly independent of irradiation dose before the damage (embrittlement) mechanism changed. In low dose range (< ~ 2dpa), the fracture strain and tensile fracture energy decreased rapidly with dose and at higher doses they decreased gradually to saturated levels, which were still high for irradiated materials. It was also found that the fracture properties for EC316LN SS were less sensitive to irradiation dose 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.

  20. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

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

  2. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

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

  4. Temperature dependence of the dislocation microstructure of PCA austenitic stainless steel irradiated in ORR spectrally-tailored experiments

    NASA Astrophysics Data System (ADS)

    Maziasz, P. J.

    1992-09-01

    Specimens of solution-annealed (SA) and 25% cold-worked (CW) prime-candidate-alloy (PCA) austenitic stainless steel were irradiated in ORR in spectrally-tailored experiments specially designed to produce fusion-relevant He/dpa ratios (12-18 appm He/dpa). SA and CW PCA were irradiated at 330 and 400°C to 13 dpa while only CW PCA was irradiated at 60, 200, 330 and 400°C to 7.4 dpa. Cavities and fine MC precipitates were only detectable at 330 and 400°C. Dislocations were a major component of the radiation-induced microstructure at 60-400°C. Mixtures of tiny “black-spot” loops, larger Frank loops, and network components of the total dislocation structure were very temperature dependent. Both SA and CW PCA contained Frank loops and network dislocations at 330 and 400°C, with SA PCA having more of both. Frank loop concentrations were maximum at 330°C and dislocations evolved most with dose at 400°C. At 60 and 200°C, the microstructure was dominated by very dense dispersions of tiny (1-3 nm diam) “black-spot” loops. No Frank loops were found at 60°C. Surprisingly, significant radiation-induced recovery of the as-cold-worked dislocation network occured in CW PCA at all temperatures. The nature of the radiation-induced microstructure makes a transition between 200 and 330°C.

  5. 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.

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

    SciTech Connect

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

    2004-06-28

    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.

  7. Influence of sulfate-reducing bacteria on alloy 625 and austenitic stainless steel weldments

    SciTech Connect

    Enos, D.G.; Taylor, S.R.

    1996-11-01

    A series of welded austenitic stainless steel and alloy 625 clad specimens were exposed to natural lake water inoculated with a mixed culture of anaerobic organisms high in sulfate-reducing bacteria. Total exposure was 300 days. The water and bacteria were taken from an actual service water system. Electrochemical testing included electrochemical impedance spectroscopy, monitoring of open-circuit potential (E{sub oc}), and zero resistance ammetry tests. Comparison of electrochemical and visual observations to sterile controls indicated electrochemical behavior of all materials in the test matrix was influenced by the bacteria. Polarization resistance and E{sub oc} values were reduced dramatically. Attack was along the fusion line of the weld. The magnitude of these effects followed a trend predicted by the pitting index for each material.

  8. Microstructural development in reduced activation ferritic alloys irradiated to 200 dpa at 420$deg;C

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    1994-09-01

    Density change and microstructural development are reported for nine reduced activation ferritic steels covering the range 2.3 to 12Cr with varying additions of V and/or W for hardening and up to 6.5 Mn for austenite stability. Specimens were examined following irradiation in FFTF/MOTA at 420°C to a dose exceeding 200 dpa. Void swelling was found, but the swelling remained at 5% or below, with the worst case in an alloy of 9Cr-2Mn-1WV. The carbide structure pinning martensite lath boundaries remained in place.

  9. Swelling in commercial Fe-Cr-Ni based alloys under electron irradiation

    NASA Astrophysics Data System (ADS)

    Thomas, L. E.; Gelles, D. S.

    1982-08-01

    Electron irradiation in a 1 MeV electron microscope has been used to study the void swelling response of several commercial austenitic stainless steels and iron-nickel based superalloys. Use of the 1 MeV microscope permits direct, continuous observation of the void development during elevated-temperature irradiations at displacement rates about 10 000 times greater then those in a fast breeder reactor. The alloys examined in this work included AISI 310, RA 330, A286, M813, Nimonic PE16, Inconel 706, Inconel 718 and Incoloy 901. Both helium preinjected specimens and uninjected specimens were studied. In all of the above alloys, swelling proceeds by formation of irradiation-induced dislocations and voids, followed by growth of the voids. The swelling rates and peak swelling temperatures vary considerably with alloy composition, heat treatment and helium preinjection. Comparisons of these results with recently reported swelling data from the same alloys after high fluence neutron irradiation in the EBR-II reactor shows good qualitative agreement in most cases. Helium preinjection of the electron irradiated specimens generally produced a poorer simulation than no helium preinjection. In one or two cases where the electron and neutron irradiation results strongly disagree, the differences appear to result from differences in irradiation-induced precipitation. Although the correlations between neutron and electron irradiation results are inadequate to obtain reliable engineering data by simulation, in-reactor swelling behavior is in general qualitatively well-represented by swelling response in the 1 MeV electron microscope. Nimonic is the registered trademark of Henry Wiggin and Company, UK. Inconel and Incoloy are registered trademarks of the International Nickel Company, Inc.

  10. Factors which control the swelling of FeCrNi ternary austenitic alloys

    NASA Astrophysics Data System (ADS)

    Garner, F. A.; Black, C. A.; Edwards, D. J.

    1997-06-01

    In agreement with limited earlier studies, a comprehensive irradiation experiment conducted in both EBR-II and FFTF demonstrates that while cold-working usually decreases void swelling of ternary FeCrNi alloys at relatively low irradiation temperatures, it in general increases swelling at higher irradiation temperatures. Aging of cold-worked specimens to produce cellular dislocation networks tends to further increase swelling, especially at higher nickel levels. The swelling of ternary alloy at lower nickel levels also appears to be sensitive to details of the preirradiation annealing treatment. The differences in the details of reactor operating conditions also exert an influence on void nucleation and thereby on the duration of the transient regime of swelling. In the current irradiation series this leads to the swelling developed in EBR-II at ˜ 30 dpa being consistently larger than that in FFTF. All of these results confirm an earlier conclusion that the primary variability of void swelling of FeCrNi alloys lies in the incubation and transient regimes, rather than in the steady-state swelling rate regime. Under certain conditions, the transient regime can be made to approach 0 dpa.

  11. 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.

  12. 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.

    2016-04-01

    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.

  13. 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.

  14. 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.

  15. Post-irradiation annealing effect on helium diffusivity in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Katsura, R.; Morisawa, J.; Kawano, S.; Oliver, B. M.

    2004-08-01

    As an experimental basis for helium induced weld cracking of neutron irradiated austenitic stainless steels, helium diffusivity has been evaluated by measuring helium release at high temperature. Isochronal and isothermal experiments were performed at temperatures between 700 and 1300 °C for 304 and 316L stainless steels. In 1 h isochronal experiments, helium was released beginning at ˜900 °C and reaching almost 100% at 1300 °C. No apparent differences in helium release were observed between the two stainless steel types. At temperatures between 900 and 1300 °C, the diffusion rate was calculated from the time dependence of the helium release rate to be: D0=4.91 cm 2/s, E=289 kJ/mol. The observed activation energy suggests that the release of helium from the steels is associated with the removal of helium from helium bubbles and/or from vacancy diffusion.

  16. Post-Irradiation Annealing Effect on Helium Diffusivity in Austenitic Stainless Steels

    SciTech Connect

    Katsura, Ryoei; Morisawa, J; Kawano, S; Oliver, Brian M.

    2004-08-01

    As an experimental basis for helium induced weld cracking of neutron irradiated austenitic stainless steels, helium diffusivity has been evaluated by measuring helium release rates at high temperature. Isochronal and isothermal experiment were performed at temperatures between 700 and 1300 for Type 304 and 316L stainless steels. In 1 hour isochronal experiments, helium was released beginning at {approx}900 and reaching near 100% at 1300. No apparent differences in helium release rate were observed between Type 304 and 316L stainless steels. At temperatures between 1100 and 1300, the diffusion rate was calculated from the time dependence of the helium release rate to be:?D0=3.42?104 cm2/s, E=173.2 kJ/mol. The observed activation energy suggests that the release of helium from the steels is associated with the removal of helium from helium bubbles.

  17. 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.

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

    NASA Astrophysics Data System (ADS)

    Kajiwara, S.

    2003-10-01

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

  19. 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.

  20. Irradiation-induced sensitization of austenitic stainless steel in-core components

    SciTech Connect

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

    1990-10-01

    High- and commercial-purity specimens of Type 304 SS from BWR absorber rod tubes, irradiated during service to fluence levels of 6 {times} 10{sup 20} to 2 {times} 10{sup 21} n{center dot}cm{sup {minus}2} (E > 1 MeV) in two reactors, were examined by Auger electron spectroscopy to characterize irradiation-induced grain boundary segregation and depletion of alloying and impurity elements, which have been associated with irradiation-assisted stress corrosion cracking (IASCC) of the steel. Ductile and intergranular fracture surfaces were produced by bending of hydrogen-charged specimens in the ultra-high vacuum of Auger microscope. The intergranular fracture surfaces in high-fluence commercial-purity material were characterized by relatively high levels of Si, P, and In segregation. An Auger energy peak at 59 eV indicated either segregation of an unidentified element or formation of an unidentified compound on the grain boundary. In contrast to the commercial-purity material, segregation of the impurity elements and intergranular failure in the high-purity material were negligible for a similar fluence level. However, grain boundary depletion of Cr was more significant in high-purity material than in commercial-purity material, which indicates that irradiation-induced segregation of impurity elements and depletion of alloying elements are interdependent. 7 refs., 10 figs., 2 tabs.

  1. Evolution of magnetic properties of cladding austenitic steel under irradiation in a reactor

    NASA Astrophysics Data System (ADS)

    Chukalkin, Yu. G.; Kozlov, A. V.; Evseev, M. V.

    2014-03-01

    Magnetic properties of samples of austenitic steel ChS-68 cut from the cladding of a fuel element, which was irradiated in a BN-600 fast-neutron reactor to a maximal damage dose of ˜80 displacements per atom (dpa) at temperatures of 370-587°C, have been investigated. It has been established that irradiation with fast neutrons leads to the formation of ferromagnetic microregions, the effective sizes and concentration of which depend on the damage dose. It has been shown that, at damage doses higher than ˜55 dpa, small spontaneous magnetization and magnetization hysteresis, which are characteristic of the ferromagnetic state, appear in the samples. It is assumed that the ferromagnetic microregions are the nuclei of the α' phase and the radiation-induced segregation microregions, in which the spacing between the nearest iron atoms exceeds the critical distance that determines the change in the sign of exchange interaction. Arguments in favor of this assumption are presented.

  2. Irradiation-induced microstructural changes in alloy X-750

    SciTech Connect

    Kenik, E.A.

    1997-04-01

    Alloy X-750 is a nickel base alloy that is often used in nuclear power systems for it`s excellent corrosion resistance and mechanical properties. The present study examines the microstructure and composition profiles in a heat of Alloy X-750 before and after neutron irradiation.

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

    SciTech Connect

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

    2010-09-15

    Oak Ridge National Laboratory (ORNL) and Capstone Turbine Corporation (CTC) participated in an in-kind cost share cooperative research and development agreement (CRADA) effort under the auspices of the Energy Efficiency and Renewable Energy (EERE) Technology Maturation Program to explore the feasibility for use of developmental ORNL alumina-forming austenitic (AFA) stainless steels as a material of construction for microturbine recuperator components. ORNL delivered test coupons of three different AFA compositions to CTC. The coupons were exposed in steady-state elevated turbine exit temperature (TET) engine testing, with coupons removed for analysis after accumulating ~1,500, 3,000, 4,500, and 6,000 hours of operation. Companion test coupons were also exposed in oxidation testing at ORNL at 700-800°C in air with 10% H2O. Post test assessment of the coupons was performed at ORNL by light microscopy and electron probe microanalysis. The higher Al and Nb containing AFA alloys exhibited excellent resistance to oxidation/corrosion, and thus show good promise for recuperator applications.

  4. 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.

  5. 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.

  6. 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.

  7. Triple ion-beam studies of radiation damage effects in a 316LN austenitic alloy for a high power spallation neutron source

    SciTech Connect

    Lee, E.H.; Rao, G.R.; Hunn, J.D.; Rice, P.M.; Lewis, M.B.; Cook, S.W.; Farrell, K.; Mansur, L.K.

    1997-09-01

    Austenitic 316LN alloy was ion-irradiated using the unique Triple Ion Beam Facility (TIF) at ORNL to investigate radiation damage effects relevant to spallation neutron sources. The TIF was used to simulate significant features of GeV proton irradiation effects in spallation neutron source target materials by producing displacement damage while simultaneously injecting helium and hydrogen at appropriately high gas/dpa ratios. Irradiations were carried out at 80, 200, and 350 C using 3.5 MeV Fe{sup ++}, 360 keV He{sup +}, and 180 keV H{sup +} to accumulate 50 dpa by Fe, 10,000 appm of He, and 50,000 appm of H. Irradiations were also carried out at 200 C in single and dual ion beam modes. The specific ion energies were chosen to maximize the damage and the gas accumulation at a depth of {approximately} 1 {micro}m. Variations in microstructure and hardness of irradiated specimens were studied using transmission electron microscopy (TEM) and a nanoindentation technique, respectively. TEM investigation yielded varying damage defect microstructures, comprising black dots, faulted and unfaulted loops, and a high number density of fine bubbles (typically less than 1 nm in diameter). With increasing temperature, faulted loops had a tendency to unfault, and bubble microstructure changed from a bimodal size distribution to a unimodal distribution. Triple ion irradiations at the three temperatures resulted in similar increases in hardness of approximately a factor of two. Individually, Fe and He ions resulted in a similar magnitude of hardness increase, whereas H ions showed only a very small effect. The present study has yielded microstructural information relevant to spallation neutron source conditions and indicates that the most important concern may be radiation induced hardening and associated ductility loss.

  8. Effect of nitrogen and vanadium on austenite grain growth kinetics of a low alloy steel

    SciTech Connect

    Stasko, Renata . E-mail: rstasko@ap.Cracow.pl; Adrian, Henryk . E-mail: adrian@uci.agh.edu.pl; Adrian, Anna . E-mail: adrian@metal.agh.edu.pl

    2006-06-15

    Austenite grain growth kinetics in a steel containing 0.4% C, 1.8% Cr with different nitrogen contents (in the range 0.0038-0.0412%) and a micralloying addition of 0.078% V were investigated. The investigations were carried out in an austenitising temperature range of 840-1200 deg. C for 30 min. The results of investigations showed that N promotes the grain growth of austenite. The microalloying addition of vanadium protects the austenite grain growth because of carbonitride V(C,N) precipitation and the grain boundary pinning effect of undissolved particles of V(C,N). Using a thermodynamic model, the carbonitride V(C,N) content, undissolved at the austenitising temperature was calculated. At temperatures when a coarsening and dissolution of carbonitride occurs, the austenite grains start to growth. The effect of nitrogen on the type of chord length distribution of austenite grains was analysed.

  9. 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.

  10. The ferrite and austenite lattice parameters of Fe-Co and Fe-Cu binary alloys as a function of temperature

    SciTech Connect

    Velthuis, S.G.E. te; Sietsma, J.; Rekveldt, M.T.; Zwaag, S. van der; Root, J.H.

    1998-09-18

    The lattice parameters of Fe-15 Cu, Fe-2% Cu, Fe-1% Co, and Fe-2% Co binary alloys were determined by means of neutron diffraction at temperatures around the austenite-ferrite phase transformation (860--1350 K). While the thermal expansion coefficients prove to be similar to those of Fe for all alloys, Cu and Co have an opposite effect on the lattice parameter of Fe. Addition of Cu increases the lattice parameter in both ferrite ({alpha}) and austenite ({gamma}), while Co decreases the lattice parameter. For all alloys, the {alpha} {leftrightarrow} {gamma} phase transformation introduces a volume change of 1.0%. Evidence is found that both ferrite and austenite are slightly strained ({epsilon} < 8 {times} 10{sup {minus}4}) when both phases are present simultaneously.

  11. Phase transformation studies in unirradiated and proton beam irradiated Ni-Ti alloy between 25 and 100°C

    NASA Astrophysics Data System (ADS)

    Ayub, Rana; Afzal, Naveed; Ahmad, R.

    2012-06-01

    The stress-induced phase transformation characteristics of unirradiated and proton beam irradiated NiTi alloy were investigated at different tests temperatures. The wire-shaped NiTi specimens were irradiated by 2 MeV proton beam for 30 min at room temperature to a flux of 1019 protons/m2 s. Engineering stress-strain (S-S) curves of both unirradiated and irradiated specimens were obtained using a materials testing machine at 25, 50, 75 and 100°C. The results indicate a single-stage phase transformation from austenite to martensite (B2-B19‧) in unirraidated specimens at all the test temperatures. In contrast, in the case of the irradiated specimens, a two-stage austenite-rhombohedral-martensite (B2-R-B19‧) phase transformation is observed at 25 and 50°C. The B2-R-B19‧ phase transformation, however, is found to change into B2-B19‧ transformation at 75 and 100°C. The stress required to initiate the B19‧ phase transformation (σMS) and the plateau range are found to be lower in irradiated specimens compared with those of the unirradiated specimens. The results obtained are discussed on the basis of the formation of Ni4Ti3 precipitates in irradiated specimens and their consequences on the phase transformations.

  12. 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.

  13. 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.

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

    NASA Astrophysics Data System (ADS)

    Schmidt, Eric

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

  15. 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.

  16. 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

  17. 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.

  18. Microstructural development due to long-term aging and ion irradiation behavior in weld metals of austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Nakata, K.; Ikeda, S.; Hamada, S.; Hishinuma, A.

    1996-10-01

    In a candidate austenitic stainless steel (316F) for fusion reactor structural materials, irradiation behavior of the weld metal produced by electron-beam welding (containing 7.9 vol% δ-ferrite) was investigated in terms of microstructural development. The densities of interstitial clusters in the γ-phase of the weld metal irradiated with He-ions at 673 and 773 K were about four times larger than those in 316F. Voids were formed in the δ-ferrite of the weld irradiated at 773 K. The number of clusters decreased in the weld metal (γ-phase) aged at 773 to 973 K, compared with that in the as-welded metal. The change in cluster density could be attributed to a Ni concentration increase in the γ-phase of the weld metal during aging.

  19. Irradiation performance of FFTF drivers using the D9 alloy

    SciTech Connect

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

    1994-12-31

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

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

    SciTech Connect

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

    2009-03-23

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

  1. 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.

  2. 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

  3. 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

  4. 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.

  5. Temperature effect on characteristics of void population formed in the austenitic steel under neutron irradiation up to high damage dose

    NASA Astrophysics Data System (ADS)

    Kozlov, A. V.; Portnykh, I. A.; Skryabin, L. A.; Kinev, E. A.

    2002-12-01

    Radiation-induced porosity in fuel pin cladding of the BN-600 reactor fabricated of cold-worked austenitic steel 16Cr-15Ni-2Mo-2Mn irradiated to different damage dose 20-90 dpa at 410-600 °C has been examined by transmission electron microscopy. Formation and growth of various types of voids were shown to occur according to their both duration and mechanism of nucleation. Dependencies of average diameters and concentration of all void types on neutron irradiation damage dose were plotted for various temperature ranges. The change of void population with increasing dose at various temperature ranges was analyzed based on point defect kinetic. The contribution of different types of voids to swelling was examined.

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

    DOE PAGESBeta

    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

  7. Flow stress and microstructural evolution during hot working of alloy 22Cr-13Ni-5Mn-0.3N austenitic stainless steel

    SciTech Connect

    Mataya, M.C.; Perkins, C.A.; Thompson, S.W.; Matlock, D.K.

    1996-05-01

    The stress-strain behavior and the development of microstructure between 850 C and 1,150 C in an austenitic stainless steel, 22Cr-13Ni-5Mn-0.3N, were investigated by uniaxial compression of cylindrical specimens at strain rates between 0.01 and 1 s{sup {minus}1} up to a strain of one. The measured (anisothermal) and corrected (isothermal) flow curves were distinctly different. The flow stress at moderate hot working temperatures, compared to a number of other austenitic alloys, was second only to that of alloy 718. Both static and dynamic recrystallization were observed. Recrystallization was sluggish in comparison to alloy 304L, apparently due to the presence of a fine Cr- and Nb-rich second-phase dispersion, identified as Z phase, which tended to pin the high-angle grain boundaries even at a high temperature of 1,113 C. Recrystallization may also be retarded by preferential restoration through the competitive process of recovery, which is consistent with the relatively high stacking-fault energy for this alloy. It is concluded that this alloy must be hot worked at temperatures higher than usual for austenitic stainless steels in order to minimize flow stress and refine grain size.

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

    SciTech Connect

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

    2001-06-15

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

  9. 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. PMID:25250795

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

    PubMed

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

    2014-11-01

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

  11. Irradiation Stability of Uranium Alloys at High Exposures

    SciTech Connect

    McDonell, W.R.

    2001-03-26

    Postirradiation examinations were begun of a series of unrestrained dilute uranium alloy specimens irradiated to exposures up to 13,000 MWD/T in NaK-containing stainless steel capsules. This test, part of a program of development of uranium metal fuels for desalination and power reactors sponsored by the Division of Reactor Development and Technology, has the objective of defining the temperature and exposure limits of swelling resistance of the alloyed uranium. This paper discusses those test results.

  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 PAGESBeta

    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. 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.

  16. 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.

  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. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

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

    SciTech Connect

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

    1988-01-01

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

  4. Neutron irradiation creep in stainless steel alloys

    NASA Astrophysics Data System (ADS)

    Schüle, Wolfgang; Hausen, Hermann

    1994-09-01

    Irradiation creep elongations were measured in the HFR at Petten on AMCR steels, on 316 CE-reference steels, and on US-316 and US-PCA steels varying the irradiation temperature between 300°C and 500°C and the stress between 25 and 300 MPa. At the beginning of an irradiation a type of "primary" creep stage is observed for doses up to 3-5 dpa after which dose the "secondary" creep stage begins. The "primary" creep strain decreases in cold-worked steel materials with decreasing stress and decreasing irradiation temperature achieving also negative creep strains depending also on the pre-treatment of the materials. These "primary" creep strains are mainly attributed to volume changes due to the formation of radiation-induced phases, e.g. to the formation of α-ferrite below about 400°C and of carbides below about 700°C, and not to irradiation creep. The "secondary" creep stage is found for doses larger than 3 to 5 dpa and is attributed mainly to irradiation creep. The irradiation creep rate is almost independent of the irradiation temperature ( Qirr = 0.132 eV) and linearly dependent on the stress. The total creep elongations normalized to about 8 dpa are equal for almost every type of steel irradiated in the HFR at Petten or in ORR or in EBR II. The negative creep elongations are more pronounced in PCA- and in AMCR-steels and for this reason the total creep elongation is slightly smaller at 8 dpa for these two steels than for the other steels.

  5. 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.

  6. 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.

  7. 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.

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

    NASA Astrophysics Data System (ADS)

    Porollo, S. I.; Konobeev, Yu. V.; Garner, F. А.

    2009-08-01

    Austenitic stainless steel ChS-68 serving as fuel pin cladding was irradiated in the 20% cold-worked condition in the BN-600 fast reactor in the range 56-84 dpa. This steel was developed to replace EI-847 which was limited by its insufficient resistance to void swelling. Comparison of swelling between EI-847 and ChS-68 under similar irradiation conditions showed improvement of the latter steel by an extended transient regime of an additional ˜10 dpa. Concurrent with swelling was the development of a variety of phases. In the temperature range 430-460 °С where the temperature peak of swelling was located, the principal type of phase generated during irradiation was G-phase, with volume fraction increasing linearly with dose to ˜0.5% at 84 dpa. While the onset of swelling is concurrent with formation of G-phase, the action of G-phase cannot be confidently ascribed to significant removal from solution of swelling-suppressive elements such as silicon. A plausible mechanism for the higher resistance to void swelling of ChS-68 as compared with EI-847 may be related to an observed higher stability of faulted dislocation loops in ChS-68 that impedes the formation of a glissile dislocation network. The higher level of boron in ChS-68 is thought to be one contributor that might play this role.

  9. 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.

  10. 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.

  11. Hot Ductility Behaviors in the Weld Heat-Affected Zone of Nitrogen-Alloyed Fe-18Cr-10Mn Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Moon, Joonoh; Lee, Tae-Ho; Hong, Hyun-Uk

    2015-04-01

    Hot ductility behaviors in the weld heat-affected zone (HAZ) of nitrogen-alloyed Fe-18Cr-10Mn austenitic stainless steels with different nitrogen contents were evaluated through hot tension tests using Gleeble simulator. The results of Gleeble simulations indicated that hot ductility in the HAZs deteriorated due to the formation of δ-ferrite and intergranular Cr2N particles. In addition, the amount of hot ductility degradation was strongly affected by the fraction of δ-ferrite.

  12. 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.

  13. Defect structures before steady-state void growth in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Yoshiie, T.; Sato, K.; Cao, X.; Xu, Q.; Horiki, M.; Troev, T. D.

    2012-10-01

    In the radiation damage process of austenitic stainless steels, there exists an incubation period before steady-state void growth, and the defect formation behaviors during that period strongly depend on alloy composition. Using the technique of positron annihilation lifetime measurement, the evolution of defect clusters during the incubation period in neutron, electron, and H-ion irradiations was studied for a variety of austenitic stainless steels including commercial and model alloys. The lifetime measurements indicated that in fission neutron irradiation to 0.2 dpa at 363 K, single vacancies were predominantly formed in the commercial alloys, SUS316L and Ti added, modified SUS316, while large voids were formed in Ni and Fe-Cr-Ni. After neutron irradiation at 573 K, stacking fault tetrahedra and/or precipitates were detected in the commercial alloys, while large voids were detected in the model alloys. In the 30 MeV electron irradiation to a dose of 0.012 dpa, the effect of alloying elements on lifetime data was less significant at 353 K, but a significant difference was found between model alloys and commercial alloys at 573 K. The H-ion irradiation at 2 MeV was also performed at room temperature. Defect evolution during the incubation period is discussed on the basis of the neutron, electron and H-ion irradiation results.

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

    SciTech Connect

    Maziasz, PJ

    2004-09-30

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

  15. Irradiation-Induced Nanoprecipitation in Ni-W Alloys

    NASA Astrophysics Data System (ADS)

    Lee, Jaeyel; Lear, Calvin R.; Zhang, Xuan; Bellon, Pascal; Averback, Robert S.

    2015-03-01

    The evolution of Ni-W alloy thin films subjected to Kr irradiation at room temperature and subsequent annealing at 1123 K (850 °C) was studied by X-ray diffraction and transmission electron microscopy. Irradiation resulted in significant increase in grain size, from ~20 nm in the as-grown state to over 300 nm after irradiation and annealing. The compositions selected for the study, 18 and 23 at. pct W, resulted in the formation of an ordered Ni4W matrix after annealing. Remarkably, in the Ni-23 at. pct W films, irradiation followed by annealing induced the precipitation of two families of Ni2W4C carbides, large blocky ones at grain boundaries, and intragranular nanocarbides, ~5 to 20 nm in size and with a high number density, 9.0 × 1022 m-3. In contrast, only blocky Ni6W6C carbides formed in control specimens directly subjected to annealing. The intragranular Ni2W4C nanocarbides displayed an orientation relationship with the Ni4W matrix, and they appear to be effective traps for implanted Kr ions, since nanobubbles formed on their periphery. The results suggest that non-equilibrium processing can be used to nucleate nanocarbides in the grain interiors of Ni-W alloys, and that this may improve alloy properties, including radiation resistance.

  16. 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).

  17. 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).

  18. 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.

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

    SciTech Connect

    Elthalabawy, Waled M.; Khan, Tahir I.

    2010-07-15

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

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

    SciTech Connect

    Swindeman, R.W.

    1993-07-01

    Commercial and developmental alloys were evaluated in support of advanced steam cycle and combined cycle technology. Working with industrial groups, Grade 91 steel, which is a candidate for main steam line piping and superheater tubing in advanced steam cycle plants, was re-evaluated to examine metallurgical factors that influence long-time performance to 600{degree}C. Deformation models and aging effect models were developed. Testing of corrosion-resistant filler metals for tubing was extended to times approaching 30,000 h. Good strengths were observed. Modified Type 310 stainless steels were examined to 927{degree}C. It was found that these steels had up to twice the strength of standard Type 310H stainless steel. The behavior of aluminum-bearing, alloys and high chromium alloys was examined for potential applications to 870{degree}C. Thermal cycling of clad tubing was undertaken, and good performance was found.

  1. Explosion bonding: aluminum-magnesium alloys bonded to austenitic stainless steel

    SciTech Connect

    Patterson, R.A.

    1982-01-01

    The explosion bonding of 5000 series aluminum alloys to 300 series stainless steel alloys is summarized. The process technique involves a parallel gap arrangement with copper or aluminum bonding aids. Successful bonds have been achieved using either a single shot process for joining the trilayer clad or a sequential shot technique for each metal component. Bond success is monitored through a combined metallographic and tensile strength evaluation. Tensile properties are shown to be strongly dependent upon process parameters and the amount of intermetallic formation at the aluminum bond interface. Empirical data has been compared with experimental and destructive test results to determine the optimum procedures.

  2. 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.

  3. 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.

  4. Effects of self-irradiation in plutonium alloys

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

    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.

  5. Effect of ferrite formation on abnormal austenite grain coarsening in low-alloy steels during hot rolling process

    SciTech Connect

    Asahi, Hitoshi; Ueno, Masakatsu; Yagi, Akira

    1998-05-01

    Abnormal coarsening of austenite ({gamma}) grains occurred in low-alloy steels during a seamless pipe hot-rolling process. Often, the grains became several hundred micrometer in diameter. This made it difficult to apply direct quenching to produce high-performance pipes. The phenomenon of grain coarsening was successfully reproduced using a thermomechanical simulator, and the factors which affected grain coarsening were clarified. The mechanism was found to be basically strain-induced grain growth which occurred during reheating at around 930 C. Furthermore, once a pipe temperature decreased to the dual-phase region after the minimal hot working and prior to the reheating process, the grain coarsening was more pronounced. It was understood that the formation of ferrite along grain boundaries had the role of reducing the migration of grain boundaries into neighboring grains, leaving a strain-free, recrystallized region behind. This abnormal grain coarsening was found to be effectively prevented by an addition of Nb, the content of which varied depending on the C content. The effect of the Nb addition was confirmed by an in-line test.

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

    SciTech Connect

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

    2011-09-01

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

  7. Effect of ferrite formation on abnormal austenite grain coarsening in low-alloy steels during the hot rolling process

    NASA Astrophysics Data System (ADS)

    Asahi, Hitoshi; Yagi, Akira; Ueno, Masakatsu

    1998-05-01

    Abnormal coarsening of austenite (γ) grains occurred in low-alloy steels during a seamless pipe hotrolling process. Often, the grains became several hundred micrometers in diameter. This made it difficult to apply direct quenching to produce high-performance pipes. The phenomenon of grain coarsening was successfully reproduced using a thermomechanical simulator, and the factors which affected grain coarsening were clarified. The mechanism was found to be basically strain-induced grain rowth which occurred during reheating at around 930 °C. Furthermore, once a pipe temperature decreased to the dual-phase region after the minimal hot working and prior to the reheating process, the grain coarsening was more pronounced. It was understood that the formation of ferrite along grain boundaries had the role of reducing the migration of grain boundaries into neighboring grains, leaving a strain-free, recrystallized region behind. This abnormal grain coarsening was found to be effectively prevented by an addition of Nb, the content of which varied depending on the C content. The effect of the Nb addition was confirmed by an in-line test.

  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. 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.

  10. Microstructural development of neutron irradiated W?Re alloys

    NASA Astrophysics Data System (ADS)

    Nemoto, Yoshiyuki; Hasegawa, Akira; Satou, Manabu; Abe, Katsunori

    2000-12-01

    Tungsten (W) alloys are candidate materials to be used as high-heat-flux materials in fusion reactors. In our previous work, W-26 wt% Re showed drastic hardening and embrittlement after the neutron irradiation. In this study, to clarify the irradiation hardening and embrittlement behavior of W-26 wt% Re, from the viewpoint of microstructural development, the microstructure observation of the neutron irradiated W-26 wt% Re was carried out using transmission electron microscope (TEM). The specimens were irradiated at the materials open test assembly of the fast flux test facility (FFTF/MOTA-2A cycle 11) up to ˜1×10 27 n/m2, ( En>0.1 MeV). The irradiation temperatures were 646, 679, 792, 873 and 1073 K. In all neutron irradiated W-26 wt% Re samples, sigma-phase precipitates and chi-phase precipitates were observed, while in the thermally aged specimen, only sigma-phase precipitates were observed. Irradiation effects on microstructural development are discussed.

  11. Physical and mechanical modelling of neutron irradiation effect on ductile fracture. Part 1. Prediction of fracture strain and fracture toughness of austenitic steels

    NASA Astrophysics Data System (ADS)

    Margolin, Boris; Sorokin, Alexander; Smirnov, Valeriy; Potapova, Vera

    2014-09-01

    A physical-and-mechanical model of ductile fracture has been developed to predict fracture toughness and fracture strain of irradiated austenitic steels taking into account stress-state triaxiality and radiation swelling. The model is based on criterion of plastic collapse of a material unit cell controlled by strain hardening of a material and criterion of voids coalescence due to channel shearing of voids. The model takes into account deformation voids nucleation and growth of deformation and vacancy voids. For justification of the model experimental data on fracture strain and fracture toughness of austenitic steel 18Cr-10Ni-Ti grade irradiated up to maximal dose 150 dpa with various swelling were used. Experimental data on fracture strain and fracture toughness were compared with the results predicted by the model. It has been shown that for prediction of the swelling effect on fracture toughness the dependence of process zone size on swelling should be taken into account.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. Irradiation assisted stress corrosion cracking of austenitic stainless steels. Progress report, September 30, 1989--June 30, 1990

    SciTech Connect

    Was, G.S.; Atzmon, M.

    1990-06-01

    Samples of ultra high purity stainless steel have been fabricated into 2mm {times} 2mm rectangular bars and irradiated to one dpa ({approximately}l {times} 10{sup 19} p{sup +}/cm{sup 2}) using 3.4 MeV protons (>20{mu}A) while controlling the sample temperature at 400{degree}C. Samples are pressed onto a water-cooled and electrically heated copper block with a thin layer of Sn in between to improve thermal conductivity. The irradiation produced a significant prompt radiation field but sample activation was limited to {beta}-decay and this decayed rapidly in less than 48 h. Samples were hydrogen charged and strained at slow rates at {minus}30{degree}C insitu in the Auger electron spectrometer to successfully fracture several samples intergranularly for grain boundary composition analysis. An ultra-high purity (UHP) alloy of Fe-19Cr-9Ni was irradiated to 1 dpa at 400C {plus_minus} 5C and 7 {times} 10{sup {minus}9} torr in the tandem accelerator of the Michigan Ion Beam Laboratory, resulting in a dislocation network density of 1.8 {times} 10{sup 9} cm{sup 2} and a dislocation loop density of 7 {times} 10{sup 16} cm{sup {minus}3} along with the dissolution of small precipitates present in the unirradiated sample. EPR experiments on the UHP irradiated alloy showed no significant increase in charge passed upon reactivation, over an unirradiated sample experiencing the same thermal history. An SCC waterloop and autoclave system has been completed and a sample has been designed to measure the susceptibility of the irradiated microstructure as compared to the unirradiated microstructure.

  17. Analytical description of true stress-true strain curves for neutron-irradiated stainless austenitic steels

    SciTech Connect

    Gussev, Maxim N; Byun, Thak Sang; Busby, Jeremy T

    2012-01-01

    This paper summarizes the results of an investigation for the deformation hardening behaviors of neutron-irradiated stainless steels in terms of true stress( ) true strain( ) curves. It is commonly accepted that the - curves are more informative for describing plastic flow, but there are few papers devoted to using the true curves for describing constitutive behaviors of materials. This study uses the true curves obtained from stainless steel samples irradiated to doses in the range of 0 55 dpa by various means: finite element calculation, optic extensomentry, and recalculation of engineering curves. It is shown that for the strain range 0 0.6 the true curves can be well described by the Swift equation: =k ( - 0)0.5. The influence of irradiation on the parameters of the Swift equation is investigated in detail. It is found that in most cases the k-parameter of this equation is not changed significantly by irradiation. Since large data scattering was observed for the 0-parameter, a modified Swift equation =k*( - 0 2/k2)0.5 was proposed and evaluated. This equation is based on the concept of zero stress, which is, in general, close to yield stress. The relationships among k, 0, and damage dose are discussed in detail, so as to more accurately describe the true curves for irradiated stainless steels.

  18. 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.

  19. A Study of the Properties of a Room Temperature Martensitic Binary Nitinol Alloy Above and Below its Martensite to Austenite Transformation Temperature

    NASA Astrophysics Data System (ADS)

    Norwich, Dennis W.

    2011-07-01

    Room temperature martensitic Nitinol alloys provide a challenge to end users of the material because they are martensitic and soft at room temperature. These are commonly referred to as Shape Memory alloys as they revert to their superelastic (pseudoelastic) form and austenitic structure at a temperature above ambient. For this study, a NiTi wire, Ti-55.3 wt.%Ni in composition (Alloy-B) and heat-treated to an Af ≈ 60 °C was used. Tensile testing was performed to fully characterize the performance of the material at a series of temperatures above and below its transformation temperature. This article will summarize the properties of the material along with the effects of multiple strains on key material performance characteristics.

  20. Radiation hardening of V C, V O, V N alloys neutron-irradiated to high fluences

    NASA Astrophysics Data System (ADS)

    Chuto, Toshinori; Satou, Manabu; Abe, Katsunori

    1998-10-01

    Vanadium has a large affinity for interstitial impurities such as C, N and O. Mechanical properties and irradiation performance of vanadium alloys are affected by the impurities. Radiation hardening and defect microstructures of vanadium alloys doped with relatively large amounts of these interstitial elements were studied. Neutron irradiation was conducted in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA-1F) to 47.9 dpa at temperatures of 679, 793 and 873 K. Irradiation hardening decreased with increasing irradiation temperature. Increase in hardness for the V-C alloy was relatively greater after irradiation at the low temperatures. Decorated dislocations and voids were observed depending on the alloying elements. The factors for irradiation hardening were different for each interstitial element in the alloys irradiated at 873 K to 47.9 dpa.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Long-Term Oxidation of Candidate Cast Iron and Advanced Austenitic Stainless Steel Exhaust System Alloys from 650-800 C in Air with Water Vapor

    DOE PAGESBeta

    Brady, Michael P; Muralidharan, Govindarajan; Leonard, Donovan N; Haynes, James A

    2014-01-01

    The oxidation behavior of SiMo cast iron, Ni-resist D5S cast iron, cast chromia-forming austenitic stainless steels of varying Cr/Ni content based on CF8C plus, HK, and HP, and a developmental cast alumina-forming austenitic (AFA) stainless steel of interest for diesel exhaust system components were studied for up to 5000 h at 650-800 C in air with 10% H2O. At 650 C, the Ni-resist D5S exhibited moderately better oxidation resistance than did the SiMo cast iron. However, the D5S suffered from oxide scale spallation issues at 700 C and higher, whereas the oxide scales formed on SiMo cast iron remained adherentmore » from 700-800 C despite oxide scales hundreds of microns thick. The oxidation of the SiMo cast iron exhibited unusual temperature dependence, with periods of slower oxidation kinetics at 750-800 C compared to 650-700 C due to continuous silica-rich scale formation at the higher temperatures. The oxidation of the cast chromia-forming austenitics trended with the level of Cr and Ni additions, with small mass losses consistent with Cr oxy-hydroxide volatilization processes for the higher 25Cr/25-35Ni HK and HP type alloys, and transition to rapid Fe-base oxide formation and scale spallation in the lower 19Cr/12Ni CF8C plus type alloy. In contrast, small positive mass changes consistent with protective alumina scale formation were observed for the cast AFA alloy under all conditions studied. Implications of these findings for diesel exhaust system components are discussed.« less

  6. The Effect of Oversize Solute Additions on the Irradiation-Assisted Stress Corrosion Cracking Resistance of Austenitic Stainless Steels

    SciTech Connect

    M Hackett; G Was

    2005-08-12

    Solute additions of zirconium are believed to decrease RIS and dislocation density through point defect trapping and recombination, which in turn reduces grain boundary sensitization and IGSCC. In this work, the effect of zirconium on the microstructure, microchemistry, hardening and IGSCC behavior of 316SS doped with zirconium to levels of 0.31 and 0.45 wt% was studied. These alloys were then irradiated with 3.2 MeV protons to doses up to 7 dpa at a temperature of 400 C. Zr additions had relatively little effect on radiation hardening. Dislocation densities were reduced and average sizes slightly increased for the +Zr alloys relative to the 316SS. Although a low amount of swelling was seen in 316SS at 3 dpa, no voids were observed in either of the +Zr alloys at 3 or 7 dpa. The difference in RIS of Cr and Ni between 316SS and 316+LoZr at 3 dpa was negligible, though RIS for 316+HiZr was considerably less than 316+LoZr at 7 dpa. The link between the oversize solute addition of Zr and its effect on IASCC shows that although the percent strain to failure increased substantially for 316+LoZr compared to the 316SS, cracking behavior was substantially worse as the number of cracks and total crack length was increased by more than an order of magnitude.

  7. 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)

  8. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

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

  10. 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.

  11. 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

  12. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2005-05-01

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

  14. 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.

  15. 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.

  16. Post-irradiation mechanical properties of an AlMgSi alloy

    NASA Astrophysics Data System (ADS)

    Ismail, Z. H.; Birt, B.

    1995-03-01

    The effect of fast-neutron irradiation on the tensile properties and hardness of the age-hardenable alloy AlMgSi is investigated. Post-irradiation tensile tests are carried out in the temperature range 298 to 628 K. The results show that the degree of irradiation-produced hardening is dependent upon the initial condition of the alloy. The alloy in its soft condition exhibits a higher degree of irradiation hardening compared with that in the hard condition. The implication of the results is discussed in terms of the variation in the microstructures involved and compared with previosly published data.

  17. 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.

  18. Neutron irradiation effects on the microstructure of low-activation ferritic alloys*1

    NASA Astrophysics Data System (ADS)

    Kimura, A.; Matsui, H.

    1994-09-01

    Microstructures of low-activation ferritic alloys, such as 2.25% Cr-2% W, 7% Cr-2% W, 9% Cr-2% W and 12% Cr-2% W alloys, were observed after FFTF irradiation at 698 K to a dose of 36 dpa. Martensite in 7% Cr-2% W, 9% Cr-2% W and 12% Cr-2% W alloys and bainite in 2.25% Cr-2% W alloy were fairly stable after the irradiation. Microvoids were observed in the martensite in each alloy but not in bainite and δ-ferrite in 12% Cr-2% W alloys. An addition of 0.02% Ti to 9% Cr-2% W alloy considerably reduced the void density. Spherical (Ta, W) and Ti-rich precipitates were observed in the Ti-added 9% Cr-2% W alloy. Precipitates observed in 9% Cr-2% W and 7% Cr-2% W alloys are mainly Cr-rich M 23C 6 (Ta, W) and Ta(W)-rich M 6C and Fe-rich Laves phase. In 2.25% Cr-2% W alloy, high density of fine (Ta, W)-rich M 2C type precipitates as well as M 6C were observed. Spherical small α' Cr-rich particles were observed in both martensite and α-ferrite in 12% Cr-2% W alloys. Correlation between postirradiation microstructure and irradiation hardening is shown and discussed for these alloys.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

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

    DOE PAGESBeta

    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

  4. 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.

  5. 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.

  6. 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.

  7. Microstructures of deformed VTiCrSi type alloys after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Satou, Manabu; Abe, Katsunori; Kayano, Hideo

    1996-10-01

    The alloy of V5Ti5Cr1SiAl,Y (nominal composition, weight percentage) was developed to improve oxidation properties and high temperature strength, and has been studied as one of the candidates for fusion applications. This alloy showed low swelling properties and enough tensile ductility after neutron irradiation to high fluence levels. The dislocation microstructures after tensile deformation and defect microstructures in the neutron-irradiated alloy to high fluences were studied. Irradiation was conducted in the Materials Open Test Assembly of the Fast Flux Test Facility (FFTF/MOTA-2A) at 406°C to 46 dpa and the deformation microstructures were examined by transmission electron microscopy. Slip dislocations were developed inhomogeneously in the specimen deformed at ambient temperature after neutron irradiation. Dislocation loops contributed mainly to hardening of the alloy after irradiation; however, cavities and radiation-induced precipitates did not so much.

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

    NASA Astrophysics Data System (ADS)

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

    2012-04-01

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

  9. 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.

  10. The Nature and Origin of "Double Expanded Austenite" in Ni-Based Ni-Ti Alloys Developing Upon Low Temperature Gaseous Nitriding

    NASA Astrophysics Data System (ADS)

    Fonović, Matej; Leineweber, Andreas; Robach, Odile; Jägle, Eric A.; Mittemeijer, Eric J.

    2015-09-01

    Gaseous nitriding of Ni-4 wt pct Ti alloy plates led to the development of double expanded austenite ( γ N1 and γ N2) at the surface of the nitride plates. Grazing-incidence X-ray diffraction analysis demonstrated that the component γ N1 is located close to the surface and the component γ N2 is located at a certain depth below the specimen surface, in correspondence with a layered character of the nitrided zone beneath the surface as revealed by optical microscopy. Electron probe microanalysis, atom probe tomography, and Laue microdiffraction analysis did not reveal a significant difference in nitrogen content of the γ N1 and γ N2 sublayers. By X-ray diffraction stress analysis it was shown that the only significant differences of the two expanded austenite layers is a pronounced difference in compressive stress parallel to the surface: the γ N1 layer is subjected to a huge compressive stress, as large as a few GPa, whereas a relatively modest stress prevails in the γ N2 layer.